Richard Mattas - Academia.edu (original) (raw)

Papers by Richard Mattas

Research paper thumbnail of Fusion Reactor Materials

Materials Science and Technology, 2006

Research paper thumbnail of Phase-IIB experiment of JAERI/USDOE collaborative program on fusion blanket neutronics

Research paper thumbnail of U.S. Assessment of advanced limiter-divertor plasma-facing systems (ALPS) design, analysis, and R and D needs

The purpose of the ALPS program is to identify and evaluate advanced limiter/divertor systems tha... more The purpose of the ALPS program is to identify and evaluate advanced limiter/divertor systems that will enhance the attractiveness of fusion power. The highest priority goals at present are achieving high power density, up to 50 MW/m2, and showing compatibility of plasma-facing surfaces with plasma operation. Personnel representing a wide range of disciplines from a number of institutions are engaged in the program, where an evaluation phase of the program is planned for three years. Successful identification of promising concepts in the evaluation phase should lead to an R&D phase that includes proof-of-principle experiments. B J J J J J J J J

Research paper thumbnail of Fabrication of a vanadium-stainless steel test section for MHD testing of insulator coatings in flowing lithium

Fusion Technology, 1996

To test the magnetohydrodynamic (MHD) pressure drop reduction performance of candidate insulator ... more To test the magnetohydrodynamic (MHD) pressure drop reduction performance of candidate insulator coatings for the ITER Vanadium/Lithium Breeding Blanket, a test section comprised of a V-4Cr-4Ti liner inside a stainless steel pipe was designed and fabricated. Theoretically, the MHD pressure drop reduction benefit resulting from an electrically insulating coating on a vanadium-lined pipe is identical to the benefit derived from

Research paper thumbnail of Results of R and D for lithium/vanadium breeding blanket design

The self-cooled lithium/vanadium blanket concept has several attractive features for fusion power... more The self-cooled lithium/vanadium blanket concept has several attractive features for fusion power systems, including reduced activation, resistance to radiation damage, accommodation of high heat loads and operating to temperatures of 650--700 C. The primary issue associated with the lithium/vanadium concept is the potentially high MHD pressure drop experienced by the lithium as it flows through the high magnetic field of

Research paper thumbnail of The Potential for Advanced Limiter/Divertor Systems

Journal of Fusion Energy, 1998

The Advanced Limiter-divertor Plasrna-facing Systems (ALPS) program was initiated in FY 1998 in o... more The Advanced Limiter-divertor Plasrna-facing Systems (ALPS) program was initiated in FY 1998 in order to evaluate the potential for improved performance and lifetime for plasma-facing systems. The main goal of the program is; to demonstrate the advantages of advanced limiter/divertor systems over conventional systems in terms of power density capability, component lifetime, and power conversion efficiency, while providing for safe operation and minimizing impurity concerns for the plasma.

Research paper thumbnail of Materials selection for the U.S. INTOR divertor collector plate

Journal of Nuclear Materials, 1981

The divertor collector plate in the INTOR reactor will be subjected to high heat, particle, and n... more The divertor collector plate in the INTOR reactor will be subjected to high heat, particle, and neutron fluxes, making it the most severely damaged torus component. The collector plate is composed of a protection plate, which is directly exposed to the particle flux, and a heat sink which provides support for the protection plate and carries the water coolant. The

Research paper thumbnail of Self-pumping impurity control

Research paper thumbnail of EVOLVE - an advanced first wall/blanket system

A new concept for an advanced fusion first wall and blanket has been identified. The key feature ... more A new concept for an advanced fusion first wall and blanket has been identified. The key feature of the concept is the use of the heat of vaporization of lithium (about 10 times higher than water) as the primary means for capturing and removing the fusion power. A reasonable range of boiling temperatures of this alkali metal is 1200 to 1400 C, corresponding with a saturation pressure of 0.035 to 0.2 MPa. Calculations indicate that a evaporative system with Li at â¼1200 C can remove a first wall surface heat flux of >2 MW/m2 with an accompanying neutron wall load of >10 MW/m2. Work to date shows that the system provides adequate tritium breeding and shielding, very high thermal conversion efficiency, and low system pressure. Tungsten is used as the structural material, and it is expected to operate at a surface wall load of 2 MW/m2 at temperatures above 1200 C.

Research paper thumbnail of Materials issues in the design of the ITER first wall, blanket, and divertor

Journal of Nuclear Materials, 1992

Research paper thumbnail of Important material considerations in INTOR

Journal of Nuclear Materials, 1981

Research paper thumbnail of Low-Temperature Internal Friction in the Niobium - Hydrogen and Niobium - Oxygen - Hydrogen Systems

Research paper thumbnail of Effect of thin YO and SiO films on in-situ formed CaO coatings on V-4%Cr4%Ti in liquid 2.8 at.% CaLi

Research paper thumbnail of A review of ITER blanket designs

Fusion Technology, 1994

Description/Abstract Changes in ITER requirements and conditions in the Engineering Design Activi... more Description/Abstract Changes in ITER requirements and conditions in the Engineering Design Activity (EDA), and the desire to obtain greater operating flexibility, led to a reconsideration of the ITER Conceptual Design Activity (CDA) blanket designs. The ...

Research paper thumbnail of Tritium production and distribution in a zircaloy-clad Li7Pb2 assembly irradiated in the oak ridge research reactor

Journal of Nuclear Materials, 1983

A specimen of Li 7 Pb 2 (lithium in natural abundance) contained in a Zircaloy capsule was irradi... more A specimen of Li 7 Pb 2 (lithium in natural abundance) contained in a Zircaloy capsule was irradiated in the Oak Ridge Research Reactor Poolside Facility for -700 h. The design of the irradiation assembly (designated TBC-07) permitted operational control of the breeder material at a temperature of (390 ±25)OC throughout the irradiation. The amount of tritium determined to be present in the capsule parts during post-irradiation chemical analysis was (52 ± 5)Ci, corresponding to the 1 burnup of -8% of the original 6Li in the capsule. The estimated tritium production based on simple neutronics calculations, using an assumed total flux of 1.2 x 10 14 neutrons/cm 2 s, was -56 Ci. The value derived from a limited dosimetry measurement on the TBC-07 thermocouple wire was -80 Ci. The axial tritium distribution profile along the Zircaloy capsule wall and the Li 7 Pb 2 cylinder indicated a nonuniform temperature during irradiation, with the capsule ends probably being at lower' temperatur~ than the center region. There was no evidence of tritium loss from the experiment during or after irradiation.

Research paper thumbnail of Sn-Li, a new coolant/breeding material for fusion applications

Research paper thumbnail of APEX ADVANCED FERRITIC STEEL, FLIBE SELF-COOLED FIRST WALL AND BLANKET DESIGN

As an element in the U.S. Advanced Power Extraction (APEX) program, we evaluated the design optio... more As an element in the U.S. Advanced Power Extraction (APEX) program, we evaluated the design option of using advanced nanocomposite ferritic steel (AFS) as the structural material and Flibe as the tritium breeder and coolant. We selected the recirculating flow configuration as our reference design. Based on the material properties of AFS, we found that the reference design can handle a maximum surface heat flux of 1 MW/m 2 , and a maximum neutron wall loading of 5.4 MW/m 2 , with a gross thermal efficiency of 47%, while meeting all the tritium breeding and structural design requirements. This paper covers the results of the following areas of evaluation: materials selection, first wall and blanket design configuration, materials compatibility, components fabrication, neutronics analysis, thermal hydraulics analysis including MHD effects, structural analysis, molten salt and helium closed cycle power conversion system, and safety and waste disposal of the recirculating coolant design.

Research paper thumbnail of The impact of swelling on fusion reactor first wall lifetime

Journal of Nuclear Materials, 1984

The impact of swelling in 20% cold-worked Type 316 stainless steel on first wall lifetime is asse... more The impact of swelling in 20% cold-worked Type 316 stainless steel on first wall lifetime is assessed for the INTOR, DEMO, and STARFIRE first wall designs.

Research paper thumbnail of Material problems and requirements related to the development of fusion blankets: The designer point of view

Journal of Nuclear Materials, 1994

The structural materials considered for solid and liquid metal breeder blankets are the austeniti... more The structural materials considered for solid and liquid metal breeder blankets are the austenitic and martensitic steels and vanadium alloys. The principal concerns with these materials are: (a) the high-temperature-induced swelling of the austenitic steels, (b) the low temperature irradiation embrittlement of martensitic steels, and (c) the exact specification of the preferred alloy composition(s), properties during and following irradiation, and technological aspects (fabrication and welding) for the vanadium alloys. Solid breeder blankets are based on the use of lithiated ceramics such as Li,O, LiA102, Li,SiO, and Li,ZrO, and beryllium as a neutron multiplier. The main uncertainty with these materials is their behaviour under irradiation, particularly at higher bumups and fluences than have been achieved hitherto. Liquid metal blankets, utilising pure Li or the LiPb eutectic as the tritium breeding material, can be either self-or separately-cooled; separate coolants include water (with LiPb) and helium. The important materials issues with the LiPb are the development of permeation barriers to contain the tritium and, for the self-cooled option, electrical insulators to reduce the MHD pressure drop to acceptable levels. 0022-3115/94/$07.00 0 1994 Elsevier Science B.V. All rights reserved SSDI 0022-3115(94)00066-W

Research paper thumbnail of Near term and long term materials issues and development needs for plasma interactive components

Journal of Nuclear Materials, 1986

Plasma interactive components (PICs), including the first wall, limiter blades, divertor collecto... more Plasma interactive components (PICs), including the first wall, limiter blades, divertor collector plates, halo scrapers, and RF launchers, are exposed to high particle fluxes that can result in high sputtering erosion rates and high heat fluxes. In addition, the materials in reactors are exposed to high neutron fluxes which will degrade the bulk properties. This severe environment will limit the materials and designs which can be used in fusion devices. In order to provide a reasonable degree of confidence that plasma interactive components will operate successfully, a comprehensive development program is needed.

Research paper thumbnail of Fusion Reactor Materials

Materials Science and Technology, 2006

Research paper thumbnail of Phase-IIB experiment of JAERI/USDOE collaborative program on fusion blanket neutronics

Research paper thumbnail of U.S. Assessment of advanced limiter-divertor plasma-facing systems (ALPS) design, analysis, and R and D needs

The purpose of the ALPS program is to identify and evaluate advanced limiter/divertor systems tha... more The purpose of the ALPS program is to identify and evaluate advanced limiter/divertor systems that will enhance the attractiveness of fusion power. The highest priority goals at present are achieving high power density, up to 50 MW/m2, and showing compatibility of plasma-facing surfaces with plasma operation. Personnel representing a wide range of disciplines from a number of institutions are engaged in the program, where an evaluation phase of the program is planned for three years. Successful identification of promising concepts in the evaluation phase should lead to an R&D phase that includes proof-of-principle experiments. B J J J J J J J J

Research paper thumbnail of Fabrication of a vanadium-stainless steel test section for MHD testing of insulator coatings in flowing lithium

Fusion Technology, 1996

To test the magnetohydrodynamic (MHD) pressure drop reduction performance of candidate insulator ... more To test the magnetohydrodynamic (MHD) pressure drop reduction performance of candidate insulator coatings for the ITER Vanadium/Lithium Breeding Blanket, a test section comprised of a V-4Cr-4Ti liner inside a stainless steel pipe was designed and fabricated. Theoretically, the MHD pressure drop reduction benefit resulting from an electrically insulating coating on a vanadium-lined pipe is identical to the benefit derived from

Research paper thumbnail of Results of R and D for lithium/vanadium breeding blanket design

The self-cooled lithium/vanadium blanket concept has several attractive features for fusion power... more The self-cooled lithium/vanadium blanket concept has several attractive features for fusion power systems, including reduced activation, resistance to radiation damage, accommodation of high heat loads and operating to temperatures of 650--700 C. The primary issue associated with the lithium/vanadium concept is the potentially high MHD pressure drop experienced by the lithium as it flows through the high magnetic field of

Research paper thumbnail of The Potential for Advanced Limiter/Divertor Systems

Journal of Fusion Energy, 1998

The Advanced Limiter-divertor Plasrna-facing Systems (ALPS) program was initiated in FY 1998 in o... more The Advanced Limiter-divertor Plasrna-facing Systems (ALPS) program was initiated in FY 1998 in order to evaluate the potential for improved performance and lifetime for plasma-facing systems. The main goal of the program is; to demonstrate the advantages of advanced limiter/divertor systems over conventional systems in terms of power density capability, component lifetime, and power conversion efficiency, while providing for safe operation and minimizing impurity concerns for the plasma.

Research paper thumbnail of Materials selection for the U.S. INTOR divertor collector plate

Journal of Nuclear Materials, 1981

The divertor collector plate in the INTOR reactor will be subjected to high heat, particle, and n... more The divertor collector plate in the INTOR reactor will be subjected to high heat, particle, and neutron fluxes, making it the most severely damaged torus component. The collector plate is composed of a protection plate, which is directly exposed to the particle flux, and a heat sink which provides support for the protection plate and carries the water coolant. The

Research paper thumbnail of Self-pumping impurity control

Research paper thumbnail of EVOLVE - an advanced first wall/blanket system

A new concept for an advanced fusion first wall and blanket has been identified. The key feature ... more A new concept for an advanced fusion first wall and blanket has been identified. The key feature of the concept is the use of the heat of vaporization of lithium (about 10 times higher than water) as the primary means for capturing and removing the fusion power. A reasonable range of boiling temperatures of this alkali metal is 1200 to 1400 C, corresponding with a saturation pressure of 0.035 to 0.2 MPa. Calculations indicate that a evaporative system with Li at â¼1200 C can remove a first wall surface heat flux of >2 MW/m2 with an accompanying neutron wall load of >10 MW/m2. Work to date shows that the system provides adequate tritium breeding and shielding, very high thermal conversion efficiency, and low system pressure. Tungsten is used as the structural material, and it is expected to operate at a surface wall load of 2 MW/m2 at temperatures above 1200 C.

Research paper thumbnail of Materials issues in the design of the ITER first wall, blanket, and divertor

Journal of Nuclear Materials, 1992

Research paper thumbnail of Important material considerations in INTOR

Journal of Nuclear Materials, 1981

Research paper thumbnail of Low-Temperature Internal Friction in the Niobium - Hydrogen and Niobium - Oxygen - Hydrogen Systems

Research paper thumbnail of Effect of thin YO and SiO films on in-situ formed CaO coatings on V-4%Cr4%Ti in liquid 2.8 at.% CaLi

Research paper thumbnail of A review of ITER blanket designs

Fusion Technology, 1994

Description/Abstract Changes in ITER requirements and conditions in the Engineering Design Activi... more Description/Abstract Changes in ITER requirements and conditions in the Engineering Design Activity (EDA), and the desire to obtain greater operating flexibility, led to a reconsideration of the ITER Conceptual Design Activity (CDA) blanket designs. The ...

Research paper thumbnail of Tritium production and distribution in a zircaloy-clad Li7Pb2 assembly irradiated in the oak ridge research reactor

Journal of Nuclear Materials, 1983

A specimen of Li 7 Pb 2 (lithium in natural abundance) contained in a Zircaloy capsule was irradi... more A specimen of Li 7 Pb 2 (lithium in natural abundance) contained in a Zircaloy capsule was irradiated in the Oak Ridge Research Reactor Poolside Facility for -700 h. The design of the irradiation assembly (designated TBC-07) permitted operational control of the breeder material at a temperature of (390 ±25)OC throughout the irradiation. The amount of tritium determined to be present in the capsule parts during post-irradiation chemical analysis was (52 ± 5)Ci, corresponding to the 1 burnup of -8% of the original 6Li in the capsule. The estimated tritium production based on simple neutronics calculations, using an assumed total flux of 1.2 x 10 14 neutrons/cm 2 s, was -56 Ci. The value derived from a limited dosimetry measurement on the TBC-07 thermocouple wire was -80 Ci. The axial tritium distribution profile along the Zircaloy capsule wall and the Li 7 Pb 2 cylinder indicated a nonuniform temperature during irradiation, with the capsule ends probably being at lower' temperatur~ than the center region. There was no evidence of tritium loss from the experiment during or after irradiation.

Research paper thumbnail of Sn-Li, a new coolant/breeding material for fusion applications

Research paper thumbnail of APEX ADVANCED FERRITIC STEEL, FLIBE SELF-COOLED FIRST WALL AND BLANKET DESIGN

As an element in the U.S. Advanced Power Extraction (APEX) program, we evaluated the design optio... more As an element in the U.S. Advanced Power Extraction (APEX) program, we evaluated the design option of using advanced nanocomposite ferritic steel (AFS) as the structural material and Flibe as the tritium breeder and coolant. We selected the recirculating flow configuration as our reference design. Based on the material properties of AFS, we found that the reference design can handle a maximum surface heat flux of 1 MW/m 2 , and a maximum neutron wall loading of 5.4 MW/m 2 , with a gross thermal efficiency of 47%, while meeting all the tritium breeding and structural design requirements. This paper covers the results of the following areas of evaluation: materials selection, first wall and blanket design configuration, materials compatibility, components fabrication, neutronics analysis, thermal hydraulics analysis including MHD effects, structural analysis, molten salt and helium closed cycle power conversion system, and safety and waste disposal of the recirculating coolant design.

Research paper thumbnail of The impact of swelling on fusion reactor first wall lifetime

Journal of Nuclear Materials, 1984

The impact of swelling in 20% cold-worked Type 316 stainless steel on first wall lifetime is asse... more The impact of swelling in 20% cold-worked Type 316 stainless steel on first wall lifetime is assessed for the INTOR, DEMO, and STARFIRE first wall designs.

Research paper thumbnail of Material problems and requirements related to the development of fusion blankets: The designer point of view

Journal of Nuclear Materials, 1994

The structural materials considered for solid and liquid metal breeder blankets are the austeniti... more The structural materials considered for solid and liquid metal breeder blankets are the austenitic and martensitic steels and vanadium alloys. The principal concerns with these materials are: (a) the high-temperature-induced swelling of the austenitic steels, (b) the low temperature irradiation embrittlement of martensitic steels, and (c) the exact specification of the preferred alloy composition(s), properties during and following irradiation, and technological aspects (fabrication and welding) for the vanadium alloys. Solid breeder blankets are based on the use of lithiated ceramics such as Li,O, LiA102, Li,SiO, and Li,ZrO, and beryllium as a neutron multiplier. The main uncertainty with these materials is their behaviour under irradiation, particularly at higher bumups and fluences than have been achieved hitherto. Liquid metal blankets, utilising pure Li or the LiPb eutectic as the tritium breeding material, can be either self-or separately-cooled; separate coolants include water (with LiPb) and helium. The important materials issues with the LiPb are the development of permeation barriers to contain the tritium and, for the self-cooled option, electrical insulators to reduce the MHD pressure drop to acceptable levels. 0022-3115/94/$07.00 0 1994 Elsevier Science B.V. All rights reserved SSDI 0022-3115(94)00066-W

Research paper thumbnail of Near term and long term materials issues and development needs for plasma interactive components

Journal of Nuclear Materials, 1986

Plasma interactive components (PICs), including the first wall, limiter blades, divertor collecto... more Plasma interactive components (PICs), including the first wall, limiter blades, divertor collector plates, halo scrapers, and RF launchers, are exposed to high particle fluxes that can result in high sputtering erosion rates and high heat fluxes. In addition, the materials in reactors are exposed to high neutron fluxes which will degrade the bulk properties. This severe environment will limit the materials and designs which can be used in fusion devices. In order to provide a reasonable degree of confidence that plasma interactive components will operate successfully, a comprehensive development program is needed.