Nasr Ghoniem | University of California, Los Angeles (original) (raw)

Papers by Nasr Ghoniem

Fusion Engineering and Design, Dec 1, 2008

This article appeared in a journal published by Elsevier. The attached copy is furnished to the a... more This article appeared in a journal published by Elsevier. The attached copy is furnished to the author for internal non-commercial research and education use, including for instruction at the authors institution and sharing with colleagues. Other uses, including reproduction and distribution, or selling or licensing copies, or posting to personal, institutional or third party websites are prohibited. In most cases authors are permitted to post their version of the article (e.g. in Word or Tex form) to their personal website or institutional repository. Authors requiring further information regarding Elsevier's archiving and manuscript policies are encouraged to visit: http://www.elsevier.com/copyright

Fusion Technology, May 1, 1991

Journal of Nuclear Materials, Oct 1, 2011

Radiation damage parameters in SiC/SiC composite structures are determined in both magnetic (MFE)... more Radiation damage parameters in SiC/SiC composite structures are determined in both magnetic (MFE) and inertial (IFE) confinement fusion systems. Variations in the geometry, neutron energy spectrum, and pulsed nature of neutron production result in significant differences in damage parameters between the two systems. With the same neutron wall loading, the displacement damage rate in the first wall in an IFE system is $10% lower than in an MFE system, while gas production and burnup rates are a factor of 2 lower. Self-cooled LiPb and Flibe blankets were analyzed. While using LiPb results in higher displacement damage, Flibe yields higher gas production and burnup rates. The effects of displacement damage and helium production on defect accumulation in SiC/SiC composites are also discussed.

Nuclear Engineering and Design. Fusion, 1985

The effects of radiation on the structural performance of fusion reactor structures is recognized... more The effects of radiation on the structural performance of fusion reactor structures is recognized as a major issue for the development of fusion reactor technology. Neutron irradiation changes the mechanical properties of structural components resulting in a general degradation of these properties. In addition to the mechanical loads (pressure and weight) and the thermal strains, non-uniform inelastic strain fields are induced by radiation swelling and creep in fusion structures. In this paper, we describe a new computer code, STAIRE, for ST_Tress Analysis Including Radiation Effects. This code is based on standard beam theory for pipe-bends. The theory is modified in two areas: (1) consideration of the pipes' cross-section deformation as the radius of curvature changes; and (2) inclusion of inelastic radiation and thermal strains (swelling and creep). An efficient analytical/numerical approach is developed for the solution of indeterminate beam problems. As an application of the method, the stress distribution and deflections of toroidal blanket pipes in the Mirror Advanced Reactor Study (MARS) are evaluated. Swelling strains are identified as a major source of stress and deformation in the proposed blanket design, and possible soliJtions to the problem are outlined.

Journal of Applied Physics, Apr 15, 1987

An approximation of the scattering integral is derived by expanding the potential about the dista... more An approximation of the scattering integral is derived by expanding the potential about the distance of closest approach and truncating the expansion so that the integral can be performed analytically. The results are improved by using the approximate integrand only in the region near the closest approach, and assuming zero potential for large distances. The analytical solution, which requires little computation time relative to other solution methods, is shown to yield sufficient accuracy for a wide range of particle energies and impact parameters, using both the Moliere and ‘‘Universal’’ potentials.

Nuclear technology/fusion, Oct 1, 1982

Tiii report >u prepared as an account of work sponsored by in agency of the United States Governm... more Tiii report >u prepared as an account of work sponsored by in agency of the United States Government. Neither the United Sutet Government nor any agency thereof, nor any of their employees, makes any warranty, express or implied, or assumes any legal liability or responsi bility tor the accuracy, completeness, or usefulness of any information, apparatus, product, or process disclosed, or represents that its use would not infringe privately owned rights. Refer ence herein to any specific commercial product, process, or service by trade name, trademark, manufacturer, or otherwise does not necessarily constitute or imply its endorsement, recom mendation, or favoring by the United States Government or any agency thereof. The views and opinions of authors expressed herein do not necessarily state Oi reflect those of the United Slates Government or any agency thereof. Support by a Unimsit, of California Lo. An.ele. Cha-cello* V^^L iSSlSa acknowledsed. The support of US. Depart-ent of Enerty, Grant #DE-|p(l3-iOER52061, to UCLA is appreciated. " s as^mtm^^^^^5^^ DISCLAIMER This report was prepared as an account of work spon sored by an agency of the United States Government.

Nuclear Engineering and Design. Fusion, 1986

The sensitivity of lifetime predictions for fusion reactor blanket structures is investigated by ... more The sensitivity of lifetime predictions for fusion reactor blanket structures is investigated by applying the Monte Carlo numerical technique. A structural computer code, Stress Analysis Including Radiation Effects (STAIRE), developed for the analysis of mirror fusion blankets, is used as a deterministic model for the prediction of the lifetime of semicircular coolant tubes. Uncertainties in material variables are treated as probabilistic inputs to the STAIRE code and output distributions are obtained. Irradiation creep rates are shown to be sufficient for relaxation of swelling-induced stresses under most conditions. In absence of high stresses, the creep limit seems to be life-limiting, although this depends on the design-dependent swelling limit. In the case of the Mirror Advanced Reactor Study MARS) blanket design, a lifetime of several hundred dpa is shown to be highly probable.

Fusion Engineering and Design, 2017

FUSION-9565; No. of Pages 35 2 C.E. Kessel et al. / Fusion Engineering and Design xxx (2017) xxx-... more FUSION-9565; No. of Pages 35 2 C.E. Kessel et al. / Fusion Engineering and Design xxx (2017) xxx-xxx and thermal-hydraulics, liquid metal thermal hydraulics, transient thermo-mechanics, tritium analysis, maintenance assessment, magnet specification and analysis, materials assessments, core and scrape-off layer (SOL)/divertor plasma examinations.

Fusion Science and Technology, 2015

The proposed Fusion Nuclear Science Facility (FNSF) represents the first facility to enter the co... more The proposed Fusion Nuclear Science Facility (FNSF) represents the first facility to enter the complex fusion nuclear regime, and its technical mission and attributes are being developed. The FNSF represents one part of the fusion energy development pathway to the first commercial power plant with other major components being the pre-FNSF research and development, research in parallel with the FNSF, pre-DEMO research and development, and the demonstration power plant (DEMO). The Fusion Energy Systems Studies group is developing the technical basis for the FNSF in order to provide a better understanding of the demands on the fusion plasma and fusion nuclear science programs. *higher β N operation can allow higher neutron wall load **1 year has been allocated at the end of each phase in addition to the specified maintenance time

Fusion Engineering and Design, 1993

The operating conditions of a compact, high-neutron-wall-loading fusion reactor severely limit th... more The operating conditions of a compact, high-neutron-wall-loading fusion reactor severely limit the choices for structural, shield, insulator, and breeder materials. In particular the response of plasma-facing materials to radiation, thermal and pressure stresses, and their compatibility with coolants are of primary concern. Material selection issues are investigated for the compact, high mass-power-density TITAN-I reactor design study. In this paper the major findings regarding material performance are discussed. The retention of mechanical strength at relatively high temperatures, low thermal stresses, and compatibility with liquid lithium make vanadium-base alloys a promising material for structural components. Based on limited data, the thermal creep behaviour of V-3Ti-ISi and V-15Cr-5Ti alloys is approximated using the modified minimum committment method. In addition, the effects of irradiation and helium generation are superimposed on the creep behavior of V-3Ti-ISi. Coolant compatibility issues are investigated. The liquid lithium compatibility of the two vanadium alloys, V-15Cr-5Ti and V-3Ti-ISi, are compared, and the latter was chosen as the primary structural-material candidate for the liquid-lithium-cooled TITAN-I reactor. Electrically insulating materials, capable of operating at high temperatures are necessary throughout the fusion reactor device. Electrical insulator-material issues of concern include irradiation induced swelling and conductivity. Both issues are investigated and operating temperatures for minimum swelling and dielectric breakdown strength are identified for spinel (MgAIzO4).

Fusion Engineering and Design, 1993

The TITAN-II reactor is a compact, high-neutron-wall-loading (18 MW/m 2) design. The TITAN-II fus... more The TITAN-II reactor is a compact, high-neutron-wall-loading (18 MW/m 2) design. The TITAN-II fusion power core (FPC) is cooled by an aqueous lithium-salt solution that also acts as the breeder material. The use of an aqueous solution imposes special constraints on the selection of structural and breeder material because of corrosion concerns, hydrogen embrittlement, and radiolytic effects. In this paper, the materials engineering and design considerations for the TITAN-II FPC are presented. Material selection criteria, based on electrochemical corrosion mechanisms of aqueous solutions coupled with radiolysis of water by ionizing radiation, resulted in the choice of a low-activation ferritic steel as structural material for TITAN-II. Stress corrosion cracking, hydrogen embrittlement, and changes in the ductile-to-brittle transition temperature of ferritic alloys are discussed. Lithium-nitrate (LiNO 3) salt was chosen over lithium hydroxide (LiOH) because it is less corrosive and reduces the net radiolytic decomposition rate of the water. The dissolved salt in the coolant changes the thermophysical properties of the coolant results in trade-offs between the lithium concentration in the coolant, neutronics performance, thermal and structural design. The TITAN-II design requires a neutron multiplier to achieve an adequate tritium breeding ratio. Beryllium is the primary neutron multiplier, assuming a maximum swelling of about 10% based on continuous self-limiting microcracking/sintering cycles.

Fusion Engineering and Design, 1993

TITAN-II is a compact, high-power-density reversed-field pinch fusion power reactor design based ... more TITAN-II is a compact, high-power-density reversed-field pinch fusion power reactor design based on the aqueous lithium solution fusion power core concept. The selected breeding and structural materials are LiNO 3 and 9-C low activation ferritic steel, respectively. TITAN-II is a viable alternative to the TITAN-I lithium self-cooled design for the reversed-field pinch reactor to operate at a neutron wall loading of 18 MW/m 2. Submerging the complete fusion power core and the primary loop in a large pool of cool water will minimize the probability of radioactivity release. Since the protection of the large pool integrity is the only requirement for the protection of the public, TITAN-II is a level 2 of passive safety assurance design.

Journal of Nuclear Materials, 1978

Physical Review B, 1989

Atomic clustering into circular planar disks is an important process responsible for interstitial... more Atomic clustering into circular planar disks is an important process responsible for interstitialloop formation in the bulk of irradiated materials, and the evolution of atomic planes during thinfilm growth. We develop a stochastic theory for the formation of planar-atomic clusters by atomic dift'usion. The theory accounts for the transient coupling between master equations representing small-size atomic clusters and a Fokker-Planck {FP) equation for larger ones. The FP equation is solved self-consistently, together with the master equations by the moments method. Equations for the rates of change of atomic species and for the nucleation rate of atomic clusters are simultaneously solved with appropriate equations for the average size and various moments of the distribution function. An application of the theory is given by comparing the results of calculations with experimental data on interstitial-loop formation in ion-irradiated nickel.

Nuclear Technology, 1981

The Fuel Rod Analysis Program (FRAP-S3) is a fairly comprehensive computer code that is developed... more The Fuel Rod Analysis Program (FRAP-S3) is a fairly comprehensive computer code that is developed for the analysis of light water reactor fuel elements during steady-state operation. However, the code predicts an increase in the fuel radial temperature distribution with an increase in the fuel density, which is contrary to experiments. A simple modification of the code was used where the thermal conductivity is treated as porosity independent in the inner iteration loops of the program. The resulting temperature profile is corrected for the effects of porosity after it has converged. The modified code shows good agreement with the IFA-11 series of experiments using the Halden Boiling Water Reactor in Sweden.

計算力学講演会講演論文集, Nov 25, 2007

1993 Jannaf Propulsion Meeting Volume 2, Nov 1, 1993

Iridium-coated rhenium (Ir/Re) combustion chambers provide high temperature, oxidation-resistant ... more Iridium-coated rhenium (Ir/Re) combustion chambers provide high temperature, oxidation-resistant operation for radiation-cooled liquid-fueled rocket engines. A 22-N (5-lb(sub f)) chamber has been operated for 15 hours at 2200 C (4000 F) using nitrogen tetroxide/monomethyl hydrazine (NTO/MMH) propellant, with negligible internal erosion. The oxidation resistance of these chambers could be further increased by the addition of refractory oxide coatings, providing longer life and/or operation in more oxidizing and higher temperature environments. The oxide coatings would serve as a thermal and diffusion barrier for the iridium coating, lowering the temperature of the iridium layer while also preventing the ingress of oxygen and egress of iridium oxides. This would serve to slow the failure mechanisms of Ir/Re chambers, namely the diffusion of rhenium to the inner surface and the oxidation of iridium. Such protection could extend chamber lifetimes by tens or perhaps hundreds of hours, and allow chamber operation on stoichiometric or higher mixture ratio oxygen/hydrogen (O2/H2) propellant. Extensive thermomechanical, thermochemical, and mass transport modeling was performed as a key material/structure design tool. Based on the results of these analyses, several 22-N oxide-coated Ir/Re chambers were fabricated and delivered to NASA Lewis Research Center for hot-fire testing.

Handbook of Materials Modeling, 2005

Afundamentaldescriptionofplasticdeformationhasbeenrecentlypursued in many parts of the world as a... more Afundamentaldescriptionofplasticdeformationhasbeenrecentlypursued in many parts of the world as a result of dissatisfaction with the limitations of continuum plasticity theory. Although continuum models of plastic defor- mation are extensively used in engineering practice, their range of application is limited by the underlying database. The reliability of continuum plasticity descriptions is dependent on the accuracy and range of available experimental data. Under complex loading situations, however, the database is often hard to establish. Moreover, the lack of a characteristic length scale in continuum plas- ticity makes it difficult to predict the occurrence of critical localized deforma- tion zones. Although homogenization methods have played a significant role in determining the elastic properties of new materials from their constituents (e.g., composite materials), the same methods have failed to describe plastic- ity. It is widely appreciated that plastic strain is fundamentally heterogenous, displaying high strains concentrated in small material volumes, with virtually undeformed regions in-between. Experimental observations consistently show thatplasticdeformationisheterogeneousatalllength-scales. Dependingonthe deformation mode, heterogeneous dislocation structures appear with definitive wavelengths. A satisfactory description of realistic dislocation patterning and strain localization hasbeen ratherelusive.Attemptsaimedatthisquestionhave been based on statistical mechanics, reaction-diffusion dynamics, or the theory ofphasetransitions.Muchoftheeffortshaveaimedatclarifyingthefundamen- tal origins of inhomogeneousplastic deformation. On the other hand, engineer- ingdescriptionsofplasticity havereliedonexperimentallyverifiedconstitutive equations. At the macroscopic level, shear bands are known to localize plastic strain, leading to material failure. At smaller length scales, dislocation distributions are mostly heterogeneous in deformed materials, leading to the formation of

Fusion Engineering and Design, Dec 1, 2008

This article appeared in a journal published by Elsevier. The attached copy is furnished to the a... more This article appeared in a journal published by Elsevier. The attached copy is furnished to the author for internal non-commercial research and education use, including for instruction at the authors institution and sharing with colleagues. Other uses, including reproduction and distribution, or selling or licensing copies, or posting to personal, institutional or third party websites are prohibited. In most cases authors are permitted to post their version of the article (e.g. in Word or Tex form) to their personal website or institutional repository. Authors requiring further information regarding Elsevier's archiving and manuscript policies are encouraged to visit: http://www.elsevier.com/copyright

Fusion Technology, May 1, 1991

Journal of Nuclear Materials, Oct 1, 2011

Radiation damage parameters in SiC/SiC composite structures are determined in both magnetic (MFE)... more Radiation damage parameters in SiC/SiC composite structures are determined in both magnetic (MFE) and inertial (IFE) confinement fusion systems. Variations in the geometry, neutron energy spectrum, and pulsed nature of neutron production result in significant differences in damage parameters between the two systems. With the same neutron wall loading, the displacement damage rate in the first wall in an IFE system is $10% lower than in an MFE system, while gas production and burnup rates are a factor of 2 lower. Self-cooled LiPb and Flibe blankets were analyzed. While using LiPb results in higher displacement damage, Flibe yields higher gas production and burnup rates. The effects of displacement damage and helium production on defect accumulation in SiC/SiC composites are also discussed.

Nuclear Engineering and Design. Fusion, 1985

The effects of radiation on the structural performance of fusion reactor structures is recognized... more The effects of radiation on the structural performance of fusion reactor structures is recognized as a major issue for the development of fusion reactor technology. Neutron irradiation changes the mechanical properties of structural components resulting in a general degradation of these properties. In addition to the mechanical loads (pressure and weight) and the thermal strains, non-uniform inelastic strain fields are induced by radiation swelling and creep in fusion structures. In this paper, we describe a new computer code, STAIRE, for ST_Tress Analysis Including Radiation Effects. This code is based on standard beam theory for pipe-bends. The theory is modified in two areas: (1) consideration of the pipes' cross-section deformation as the radius of curvature changes; and (2) inclusion of inelastic radiation and thermal strains (swelling and creep). An efficient analytical/numerical approach is developed for the solution of indeterminate beam problems. As an application of the method, the stress distribution and deflections of toroidal blanket pipes in the Mirror Advanced Reactor Study (MARS) are evaluated. Swelling strains are identified as a major source of stress and deformation in the proposed blanket design, and possible soliJtions to the problem are outlined.

Journal of Applied Physics, Apr 15, 1987

An approximation of the scattering integral is derived by expanding the potential about the dista... more An approximation of the scattering integral is derived by expanding the potential about the distance of closest approach and truncating the expansion so that the integral can be performed analytically. The results are improved by using the approximate integrand only in the region near the closest approach, and assuming zero potential for large distances. The analytical solution, which requires little computation time relative to other solution methods, is shown to yield sufficient accuracy for a wide range of particle energies and impact parameters, using both the Moliere and ‘‘Universal’’ potentials.

Nuclear technology/fusion, Oct 1, 1982

Tiii report >u prepared as an account of work sponsored by in agency of the United States Governm... more Tiii report >u prepared as an account of work sponsored by in agency of the United States Government. Neither the United Sutet Government nor any agency thereof, nor any of their employees, makes any warranty, express or implied, or assumes any legal liability or responsi bility tor the accuracy, completeness, or usefulness of any information, apparatus, product, or process disclosed, or represents that its use would not infringe privately owned rights. Refer ence herein to any specific commercial product, process, or service by trade name, trademark, manufacturer, or otherwise does not necessarily constitute or imply its endorsement, recom mendation, or favoring by the United States Government or any agency thereof. The views and opinions of authors expressed herein do not necessarily state Oi reflect those of the United Slates Government or any agency thereof. Support by a Unimsit, of California Lo. An.ele. Cha-cello* V^^L iSSlSa acknowledsed. The support of US. Depart-ent of Enerty, Grant #DE-|p(l3-iOER52061, to UCLA is appreciated. " s as^mtm^^^^^5^^ DISCLAIMER This report was prepared as an account of work spon sored by an agency of the United States Government.

Nuclear Engineering and Design. Fusion, 1986

The sensitivity of lifetime predictions for fusion reactor blanket structures is investigated by ... more The sensitivity of lifetime predictions for fusion reactor blanket structures is investigated by applying the Monte Carlo numerical technique. A structural computer code, Stress Analysis Including Radiation Effects (STAIRE), developed for the analysis of mirror fusion blankets, is used as a deterministic model for the prediction of the lifetime of semicircular coolant tubes. Uncertainties in material variables are treated as probabilistic inputs to the STAIRE code and output distributions are obtained. Irradiation creep rates are shown to be sufficient for relaxation of swelling-induced stresses under most conditions. In absence of high stresses, the creep limit seems to be life-limiting, although this depends on the design-dependent swelling limit. In the case of the Mirror Advanced Reactor Study MARS) blanket design, a lifetime of several hundred dpa is shown to be highly probable.

Fusion Engineering and Design, 2017

FUSION-9565; No. of Pages 35 2 C.E. Kessel et al. / Fusion Engineering and Design xxx (2017) xxx-... more FUSION-9565; No. of Pages 35 2 C.E. Kessel et al. / Fusion Engineering and Design xxx (2017) xxx-xxx and thermal-hydraulics, liquid metal thermal hydraulics, transient thermo-mechanics, tritium analysis, maintenance assessment, magnet specification and analysis, materials assessments, core and scrape-off layer (SOL)/divertor plasma examinations.

Fusion Science and Technology, 2015

The proposed Fusion Nuclear Science Facility (FNSF) represents the first facility to enter the co... more The proposed Fusion Nuclear Science Facility (FNSF) represents the first facility to enter the complex fusion nuclear regime, and its technical mission and attributes are being developed. The FNSF represents one part of the fusion energy development pathway to the first commercial power plant with other major components being the pre-FNSF research and development, research in parallel with the FNSF, pre-DEMO research and development, and the demonstration power plant (DEMO). The Fusion Energy Systems Studies group is developing the technical basis for the FNSF in order to provide a better understanding of the demands on the fusion plasma and fusion nuclear science programs. *higher β N operation can allow higher neutron wall load **1 year has been allocated at the end of each phase in addition to the specified maintenance time

Fusion Engineering and Design, 1993

The operating conditions of a compact, high-neutron-wall-loading fusion reactor severely limit th... more The operating conditions of a compact, high-neutron-wall-loading fusion reactor severely limit the choices for structural, shield, insulator, and breeder materials. In particular the response of plasma-facing materials to radiation, thermal and pressure stresses, and their compatibility with coolants are of primary concern. Material selection issues are investigated for the compact, high mass-power-density TITAN-I reactor design study. In this paper the major findings regarding material performance are discussed. The retention of mechanical strength at relatively high temperatures, low thermal stresses, and compatibility with liquid lithium make vanadium-base alloys a promising material for structural components. Based on limited data, the thermal creep behaviour of V-3Ti-ISi and V-15Cr-5Ti alloys is approximated using the modified minimum committment method. In addition, the effects of irradiation and helium generation are superimposed on the creep behavior of V-3Ti-ISi. Coolant compatibility issues are investigated. The liquid lithium compatibility of the two vanadium alloys, V-15Cr-5Ti and V-3Ti-ISi, are compared, and the latter was chosen as the primary structural-material candidate for the liquid-lithium-cooled TITAN-I reactor. Electrically insulating materials, capable of operating at high temperatures are necessary throughout the fusion reactor device. Electrical insulator-material issues of concern include irradiation induced swelling and conductivity. Both issues are investigated and operating temperatures for minimum swelling and dielectric breakdown strength are identified for spinel (MgAIzO4).

Fusion Engineering and Design, 1993

The TITAN-II reactor is a compact, high-neutron-wall-loading (18 MW/m 2) design. The TITAN-II fus... more The TITAN-II reactor is a compact, high-neutron-wall-loading (18 MW/m 2) design. The TITAN-II fusion power core (FPC) is cooled by an aqueous lithium-salt solution that also acts as the breeder material. The use of an aqueous solution imposes special constraints on the selection of structural and breeder material because of corrosion concerns, hydrogen embrittlement, and radiolytic effects. In this paper, the materials engineering and design considerations for the TITAN-II FPC are presented. Material selection criteria, based on electrochemical corrosion mechanisms of aqueous solutions coupled with radiolysis of water by ionizing radiation, resulted in the choice of a low-activation ferritic steel as structural material for TITAN-II. Stress corrosion cracking, hydrogen embrittlement, and changes in the ductile-to-brittle transition temperature of ferritic alloys are discussed. Lithium-nitrate (LiNO 3) salt was chosen over lithium hydroxide (LiOH) because it is less corrosive and reduces the net radiolytic decomposition rate of the water. The dissolved salt in the coolant changes the thermophysical properties of the coolant results in trade-offs between the lithium concentration in the coolant, neutronics performance, thermal and structural design. The TITAN-II design requires a neutron multiplier to achieve an adequate tritium breeding ratio. Beryllium is the primary neutron multiplier, assuming a maximum swelling of about 10% based on continuous self-limiting microcracking/sintering cycles.

Fusion Engineering and Design, 1993

TITAN-II is a compact, high-power-density reversed-field pinch fusion power reactor design based ... more TITAN-II is a compact, high-power-density reversed-field pinch fusion power reactor design based on the aqueous lithium solution fusion power core concept. The selected breeding and structural materials are LiNO 3 and 9-C low activation ferritic steel, respectively. TITAN-II is a viable alternative to the TITAN-I lithium self-cooled design for the reversed-field pinch reactor to operate at a neutron wall loading of 18 MW/m 2. Submerging the complete fusion power core and the primary loop in a large pool of cool water will minimize the probability of radioactivity release. Since the protection of the large pool integrity is the only requirement for the protection of the public, TITAN-II is a level 2 of passive safety assurance design.

Journal of Nuclear Materials, 1978

Physical Review B, 1989

Atomic clustering into circular planar disks is an important process responsible for interstitial... more Atomic clustering into circular planar disks is an important process responsible for interstitialloop formation in the bulk of irradiated materials, and the evolution of atomic planes during thinfilm growth. We develop a stochastic theory for the formation of planar-atomic clusters by atomic dift'usion. The theory accounts for the transient coupling between master equations representing small-size atomic clusters and a Fokker-Planck {FP) equation for larger ones. The FP equation is solved self-consistently, together with the master equations by the moments method. Equations for the rates of change of atomic species and for the nucleation rate of atomic clusters are simultaneously solved with appropriate equations for the average size and various moments of the distribution function. An application of the theory is given by comparing the results of calculations with experimental data on interstitial-loop formation in ion-irradiated nickel.

Nuclear Technology, 1981

The Fuel Rod Analysis Program (FRAP-S3) is a fairly comprehensive computer code that is developed... more The Fuel Rod Analysis Program (FRAP-S3) is a fairly comprehensive computer code that is developed for the analysis of light water reactor fuel elements during steady-state operation. However, the code predicts an increase in the fuel radial temperature distribution with an increase in the fuel density, which is contrary to experiments. A simple modification of the code was used where the thermal conductivity is treated as porosity independent in the inner iteration loops of the program. The resulting temperature profile is corrected for the effects of porosity after it has converged. The modified code shows good agreement with the IFA-11 series of experiments using the Halden Boiling Water Reactor in Sweden.

計算力学講演会講演論文集, Nov 25, 2007

1993 Jannaf Propulsion Meeting Volume 2, Nov 1, 1993

Iridium-coated rhenium (Ir/Re) combustion chambers provide high temperature, oxidation-resistant ... more Iridium-coated rhenium (Ir/Re) combustion chambers provide high temperature, oxidation-resistant operation for radiation-cooled liquid-fueled rocket engines. A 22-N (5-lb(sub f)) chamber has been operated for 15 hours at 2200 C (4000 F) using nitrogen tetroxide/monomethyl hydrazine (NTO/MMH) propellant, with negligible internal erosion. The oxidation resistance of these chambers could be further increased by the addition of refractory oxide coatings, providing longer life and/or operation in more oxidizing and higher temperature environments. The oxide coatings would serve as a thermal and diffusion barrier for the iridium coating, lowering the temperature of the iridium layer while also preventing the ingress of oxygen and egress of iridium oxides. This would serve to slow the failure mechanisms of Ir/Re chambers, namely the diffusion of rhenium to the inner surface and the oxidation of iridium. Such protection could extend chamber lifetimes by tens or perhaps hundreds of hours, and allow chamber operation on stoichiometric or higher mixture ratio oxygen/hydrogen (O2/H2) propellant. Extensive thermomechanical, thermochemical, and mass transport modeling was performed as a key material/structure design tool. Based on the results of these analyses, several 22-N oxide-coated Ir/Re chambers were fabricated and delivered to NASA Lewis Research Center for hot-fire testing.

Handbook of Materials Modeling, 2005

Afundamentaldescriptionofplasticdeformationhasbeenrecentlypursued in many parts of the world as a... more Afundamentaldescriptionofplasticdeformationhasbeenrecentlypursued in many parts of the world as a result of dissatisfaction with the limitations of continuum plasticity theory. Although continuum models of plastic defor- mation are extensively used in engineering practice, their range of application is limited by the underlying database. The reliability of continuum plasticity descriptions is dependent on the accuracy and range of available experimental data. Under complex loading situations, however, the database is often hard to establish. Moreover, the lack of a characteristic length scale in continuum plas- ticity makes it difficult to predict the occurrence of critical localized deforma- tion zones. Although homogenization methods have played a significant role in determining the elastic properties of new materials from their constituents (e.g., composite materials), the same methods have failed to describe plastic- ity. It is widely appreciated that plastic strain is fundamentally heterogenous, displaying high strains concentrated in small material volumes, with virtually undeformed regions in-between. Experimental observations consistently show thatplasticdeformationisheterogeneousatalllength-scales. Dependingonthe deformation mode, heterogeneous dislocation structures appear with definitive wavelengths. A satisfactory description of realistic dislocation patterning and strain localization hasbeen ratherelusive.Attemptsaimedatthisquestionhave been based on statistical mechanics, reaction-diffusion dynamics, or the theory ofphasetransitions.Muchoftheeffortshaveaimedatclarifyingthefundamen- tal origins of inhomogeneousplastic deformation. On the other hand, engineer- ingdescriptionsofplasticity havereliedonexperimentallyverifiedconstitutive equations. At the macroscopic level, shear bands are known to localize plastic strain, leading to material failure. At smaller length scales, dislocation distributions are mostly heterogeneous in deformed materials, leading to the formation of