UB Paderborn / Katalog / Suche / in CDI

Sortiert nach

Einfacher Volltextzugriff von (fast) überall

Installieren Sie die LibKey Nomad-Browsererweiterung und erhalten Sie von vielen Webseiten einfachen Zugriff auf lizenzierte Volltexte oder Open Access Publikationen. Mehr erfahren...

Crack Initiation Mechanism in Non-ductile Cracking of Irradiated 304L Stainless Steels under BWR Water Environment

Journal of nuclear science and technology, 2006-08, Vol.43 (8), p.851-865

2006

Volltextzugriff (PDF)

Crack Initiation Mechanism in Non-ductile Cracking of Irradiated 304L Stainless Steels under BWR Water Environment

Journal of nuclear science and technology, 2006-08, Vol.43 (8), p.851

2006

Volltextzugriff (PDF)

Creep deformation and rupture properties of unirradiated Zircaloy-4 nuclear fuel cladding tube at temperatures of 727 to 857 K

Journal of nuclear materials, 1990-12, Vol.175 (1), p.135-142

1990

Volltextzugriff (PDF)

Aging Degradation of Component Materials for Light Water Reactor

Materia Japan, 1999/09/20, Vol.38(9), pp.689-692

Onchi, Takeo

1999

Volltextzugriff (PDF)

Creep deformation of an unirradiated zircaloy nuclear fuel cladding tube under dry storage conditions

Journal of nuclear materials, 1990-05, Vol.171 (2), p.381-388

1990

Volltextzugriff (PDF)

Direction to the Real Solution of SCC Initiation Mechanism

Journal of the Atomic Energy Society of Japan, 2007, Vol.49(11-12), pp.783-787

ONCHI, Takeo

2007

Volltextzugriff (PDF)

Benchmarking Studies of Probabilistic Fracture Mechanics Analysis Code, PROFMAC-II, for Assessing Pressurized Thermal Shock Events of Reactor Pressure Vessel Integrity Issues

Journal of nuclear science and technology, 1996-01, Vol.33 (1), p.87-98

1996

Volltextzugriff (PDF)

Radiation Damage of Light Water Reactor Materials

Materia Japan, 1995/01/20, Vol.34(1), pp.25-35

Onchi, Takeo

1995

Volltextzugriff (PDF)

Failure Propensities of Pressurized Zircaloy Tube Containing Iodine and Other Chemical Species

Journal of nuclear science and technology, 1993-12, Vol.30 (12), p.1288-1298

1993

Volltextzugriff (PDF)

応力腐食割れ（SCC）発生機構の本質的な理解を目指して

日本原子力学会誌ATOMOΣ, 2007, Vol.49(11-12), pp.783-787

恩地, 健雄

2007

Volltextzugriff (PDF)

Inhomogeneous plastic deformation and its relevance to iodine stress corrosion cracking susceptibility in irradiated Zircaloy-2 Tubing

Journal of nuclear materials, 1983-06, Vol.116 (2), p.211-218

1983

Volltextzugriff (PDF)

Effect of Neutron Irradiation on Iodine Stress Corrosion Cracking Susceptibility of Zircaloy-2 Tubing

Journal of nuclear science and technology, 1982-09, Vol.19 (9), p.740-748

1982

Volltextzugriff (PDF)

Benchmarking Studies of Probabilistic Fracture Mechanics Analysis Code, PROFMAG-II, for Assessing Pressurized Thermal Shock Events of Reactor Pressure Vessel Integrity Issues

Journal of nuclear science and technology, 1996-01, Vol.33 (1), p.87

1996

Volltextzugriff (PDF)

Iodine Stress Corrosion Cracking of Unirradiated Zircaloy-2 Tubing by Means of Ring Tension Test

Journal of nuclear science and technology, 1981-12, Vol.18 (12), p.930-939

1981

Volltextzugriff (PDF)

Effect of Neutron Irradiation on Fracture Behavior of Zirconium

Journal of nuclear science and technology, 1980-05, Vol.17 (5), p.359-369

1980

Volltextzugriff (PDF)

Effect of Neutron Irradiation on Deformation Behavior of Zirconium

Journal of nuclear science and technology, 1977-05, Vol.14 (5), p.359-369

1977

Volltextzugriff (PDF)

Irradiation effects of boron carbide used as control rod elements in fast breeder reactors

Journal of nuclear materials, 1978-06, Vol.74 (1), p.114-122

1978

Volltextzugriff (PDF)

Effects of Temperature, Strain Rate, and Specimen Orientation on Localized Plastic Deformation of Irradiated Zircaloy-2

Journal of nuclear science and technology, 1980-11, Vol.17 (11), p.848-856

1980

Volltextzugriff (PDF)

Yield Point Phenomenon in Neutron Irradiated Zirconium

Journal of nuclear science and technology, 1978-11, Vol.15 (11), p.868-871

1978

Volltextzugriff (PDF)

Mechanical properties of austenitic stainless steels irradiated at 323 K in the Japan Materials Testing reactor

Journal of nuclear materials, 1996-10, Vol.233, p.188-191

1996

Volltextzugriff (PDF)

The applicability of the strain-hardening rule to creep deformation of Zircaloy fuel cladding tube under dry storage condition

Journal of nuclear materials, 1991, Vol.178 (1), p.73-79

1991

Volltextzugriff (PDF)

A Method to Evaluate the Maximum Allowable Temperature of Spent Fuel in Dry Storage During a Postulated Accident

Nuclear technology, 1991-03, Vol.93 (3), p.382-388

1991

Volltextzugriff (PDF)

The inhomogeneous deformation behaviour of neutron irradiated Zircaloy-2

Journal of nuclear materials, 1980-01, Vol.88 (2), p.226-235

1980

Volltextzugriff (PDF)

Iodine Stress Corrosion Cracking of Zircaloy

CORROSION ENGINEERING, 1983/04/15, Vol.32(4), pp.225-231

1983

Volltextzugriff (PDF)

Latest Research Activities on Stress Corrosion Cracking of BWR Core Shrouds and Major Piping

Journal of the Atomic Energy Society of Japan / Atomic Energy Society of Japan, 2005-06, Vol.47 (6), p.385-397

2005

Volltextzugriff (PDF)

1 – 25 von 32

Suche als RSS-Feed

Suchergebnisse filtern

Filter anzeigen

Nur zeigen

Geprüft (Peer-Review)
(24)
Mit Volltextzugriff
(29)

Publikationsform

Artikel
(32)

Sprache

Englisch
(28)
Japanisch
(6)

Erscheinungsjahr

n.n

Zeitschrift / Serie

5 weitere anzeigen...

Weniger anzeigen...

Menü

Weitere Dienste

Einstellungen

Crack Initiation Mechanism in Non-ductile Cracking of Irradiated 304L Stainless Steels under BWR Water Environment

Crack Initiation Mechanism in Non-ductile Cracking of Irradiated 304L Stainless Steels under BWR Water Environment

Creep deformation and rupture properties of unirradiated Zircaloy-4 nuclear fuel cladding tube at temperatures of 727 to 857 K

Aging Degradation of Component Materials for Light Water Reactor

Creep deformation of an unirradiated zircaloy nuclear fuel cladding tube under dry storage conditions

Direction to the Real Solution of SCC Initiation Mechanism

Benchmarking Studies of Probabilistic Fracture Mechanics Analysis Code, PROFMAC-II, for Assessing Pressurized Thermal Shock Events of Reactor Pressure Vessel Integrity Issues

Radiation Damage of Light Water Reactor Materials

Failure Propensities of Pressurized Zircaloy Tube Containing Iodine and Other Chemical Species

応力腐食割れ（SCC）発生機構の本質的な理解を目指して

Inhomogeneous plastic deformation and its relevance to iodine stress corrosion cracking susceptibility in irradiated Zircaloy-2 Tubing

Effect of Neutron Irradiation on Iodine Stress Corrosion Cracking Susceptibility of Zircaloy-2 Tubing

Benchmarking Studies of Probabilistic Fracture Mechanics Analysis Code, PROFMAG-II, for Assessing Pressurized Thermal Shock Events of Reactor Pressure Vessel Integrity Issues

Iodine Stress Corrosion Cracking of Unirradiated Zircaloy-2 Tubing by Means of Ring Tension Test

Effect of Neutron Irradiation on Fracture Behavior of Zirconium

Effect of Neutron Irradiation on Deformation Behavior of Zirconium

Irradiation effects of boron carbide used as control rod elements in fast breeder reactors

Effects of Temperature, Strain Rate, and Specimen Orientation on Localized Plastic Deformation of Irradiated Zircaloy-2

Yield Point Phenomenon in Neutron Irradiated Zirconium

Mechanical properties of austenitic stainless steels irradiated at 323 K in the Japan Materials Testing reactor

The applicability of the strain-hardening rule to creep deformation of Zircaloy fuel cladding tube under dry storage condition

A Method to Evaluate the Maximum Allowable Temperature of Spent Fuel in Dry Storage During a Postulated Accident

The inhomogeneous deformation behaviour of neutron irradiated Zircaloy-2

Iodine Stress Corrosion Cracking of Zircaloy

Latest Research Activities on Stress Corrosion Cracking of BWR Core Shrouds and Major Piping

Suchergebnisse filtern