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Ductile-to-brittle transition temperature for high-burnup cladding alloys exposed to simulated drying-storage conditions

Journal of nuclear materials, 2013-02, Vol.433 (1-3), p.431-448

2013

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High-temperature steam-oxidation behavior of Zr-1Nb cladding alloy E110

Journal of nuclear materials, 2009-09, Vol.393 (3), p.433-448

2009

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Design and material issues for high performance SiCf/SiC-based fusion power cores

Fusion engineering and design, 2001-05, Vol.55 (1), p.55-95

2001

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Fusion power core engineering for the ARIES-ST power plant

Fusion engineering and design, 2003-02, Vol.65 (2), p.215-261

2003

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Vanadium-base alloys for fusion first-wall/blanket applications

International journal of refractory metals & hard materials, 2000, Vol.18 (4), p.213-224

2000

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Influence of temperature and hydrogen content on stress-induced radial hydride precipitation in Zircaloy-4 cladding

Journal of nuclear materials, 2014-10, Vol.453 (1-3), p.131-150

2014

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Determination of actinide and fission-product isotopes in very-high-burnup spent nuclear fuel

Journal of radioanalytical and nuclear chemistry, 2008-07, Vol.277 (1), p.59-64

2008

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Thermal Creep of Irradiated Zircaloy Cladding

Journal of ASTM International, 2006-01, Vol.3 (1), p.1-16

2006

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Radial-hydride Embrittlement of High-burnup Zircaloy-4 Fuel Cladding

Journal of nuclear science and technology, 2006-09, Vol.43 (9), p.1054

2006

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Materials integration issues for high performance fusion power systems

Journal of nuclear materials, 1998-10, Vol.258, p.65-73

1998

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Irradiation creep of vanadium-base alloys

Journal of nuclear materials, 1998-10, Vol.258, p.1471-1475

1998

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Performance limits for fusion first-wall structural materials

Journal of nuclear materials, 2000-12, Vol.283-287, p.716-720

2000

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Status of beryllium development for fusion applications

Fusion engineering and design, 1995-03, Vol.27 (1-2), p.179-190

1995

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Revision of the tensile database for V–Ti and V–Cr–Ti alloys tested at ANL

Journal of nuclear materials, 1998-10, Vol.258, p.1523-1527

1998

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Implications of radiation-induced reductions in ductility to the design of austenitic stainless steel structures

Journal of nuclear materials, 1996-10, Vol.233 (1-3), p.207-212

1996

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High temperature oxidation behavior of Zr-1Nb cladding alloy E110

Journal of nuclear materials, 2009-09, Vol.393 (3 ; Sep. 15, 2009)

2009

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Effects of low-temperature neutron irradiation on mechanical properties of vanadium-base alloys

Journal of nuclear materials, 2000-12, Vol.283-287, p.362-366

2000

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Modeling of tritium release from ceramic breeders: Status and some implications for next-step devices

Journal of Nuclear Materials, 1992-03, Vol.187 (1), p.1-31

1992

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TIARA analysis of tritium inventory in Li2O

Fusion engineering and design, 1995-03, Vol.28 (1-2), p.313-318

BILLONE, M. C

1995

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Thermal and tritium transport in Li2O and Li2ZrO3

Journal of nuclear materials, 1996-10, Vol.233-237, p.1462-1466

Billone, M.C.

1996

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Tensile and impact properties of V–4Cr–4Ti alloy heats 832665 and 832864

Journal of nuclear materials, 2000-12, Vol.283-287, p.633-636

2000

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Diffusion/desorption of tritium from irradiated beryllium

Journal of nuclear materials, 1994-09, Vol.212, p.948-953

1994

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Modeling the behavior of metallic fast reactor fuels during extended transients

Journal of nuclear materials, 1993-09, Vol.204, p.203-211

1993

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Examination of Spent Pressurized Water Reactor Fuel Rods After 15 Years in Dry Storage

Nuclear technology, 2003-11, Vol.144 (2), p.186-200

2003

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Materials for breeding blankets

Journal of nuclear materials, 1996-10, Vol.233, p.72-81

1996

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Ductile-to-brittle transition temperature for high-burnup cladding alloys exposed to simulated drying-storage conditions

High-temperature steam-oxidation behavior of Zr-1Nb cladding alloy E110

Design and material issues for high performance SiCf/SiC-based fusion power cores

Fusion power core engineering for the ARIES-ST power plant

Vanadium-base alloys for fusion first-wall/blanket applications

Influence of temperature and hydrogen content on stress-induced radial hydride precipitation in Zircaloy-4 cladding

Determination of actinide and fission-product isotopes in very-high-burnup spent nuclear fuel

Thermal Creep of Irradiated Zircaloy Cladding

Radial-hydride Embrittlement of High-burnup Zircaloy-4 Fuel Cladding

Materials integration issues for high performance fusion power systems

Irradiation creep of vanadium-base alloys

Performance limits for fusion first-wall structural materials

Status of beryllium development for fusion applications

Revision of the tensile database for V–Ti and V–Cr–Ti alloys tested at ANL

Implications of radiation-induced reductions in ductility to the design of austenitic stainless steel structures

High temperature oxidation behavior of Zr-1Nb cladding alloy E110

Effects of low-temperature neutron irradiation on mechanical properties of vanadium-base alloys

Modeling of tritium release from ceramic breeders: Status and some implications for next-step devices

TIARA analysis of tritium inventory in Li2O

Thermal and tritium transport in Li2O and Li2ZrO3

Tensile and impact properties of V–4Cr–4Ti alloy heats 832665 and 832864

Diffusion/desorption of tritium from irradiated beryllium

Modeling the behavior of metallic fast reactor fuels during extended transients

Examination of Spent Pressurized Water Reactor Fuel Rods After 15 Years in Dry Storage

Materials for breeding blankets

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