Trends in Reactor Pressure Vessel and Circuit DevelopmentR. W. Nichols |
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Page vii
... International Atomic Energy Agency Specialist Meeting on this very topic led me to believe it would be inappropriate to ... IAEA International Working Group on the Reliability of Reactor Pressurised Components , held in Innsbruck , 22-24 ...
... International Atomic Energy Agency Specialist Meeting on this very topic led me to believe it would be inappropriate to ... IAEA International Working Group on the Reliability of Reactor Pressurised Components , held in Innsbruck , 22-24 ...
Page 69
... IAEA SPECIALISTS ' MEETING IN VIENNA , AUSTRIA I. K. TERENTIEV IAEA , Vienna , Austria 1. INTRODUCTION The specialists ' meeting on ' Irradiation embrittlement , thermal annealing and surveillance of reactor pressure vessel steels ' was ...
... IAEA SPECIALISTS ' MEETING IN VIENNA , AUSTRIA I. K. TERENTIEV IAEA , Vienna , Austria 1. INTRODUCTION The specialists ' meeting on ' Irradiation embrittlement , thermal annealing and surveillance of reactor pressure vessel steels ' was ...
Page 248
... IAEA Panel on ISI , Pilsen ( Oct. 1966 ) . 2. S. ONODERA , M. NAGATA and H. TSUKADA , ' Larger Size , Integrated Type Steel Forgings as Intended for Easier ISI , IAEA Tech . Meeting , Kobé , Japan ( Apr. 1977 ) . 3. A. G. VINCKIER and ...
... IAEA Panel on ISI , Pilsen ( Oct. 1966 ) . 2. S. ONODERA , M. NAGATA and H. TSUKADA , ' Larger Size , Integrated Type Steel Forgings as Intended for Easier ISI , IAEA Tech . Meeting , Kobé , Japan ( Apr. 1977 ) . 3. A. G. VINCKIER and ...
Contents
Current Technical Technical Aspects Concerning PWR Vessel | 7 |
Some Remarks Related to the Requirements for Development | 15 |
Utilisation of Indian Point Unit 1 for Research to Support | 27 |
Copyright | |
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Common terms and phrases
alloy analysis ASME Code austenite bainite brittle capsules carbide casting CEGB cladding components cooling corrosion crack arrest temperature crack growth rates defect ductile effects examination fabrication factor failure fatigue crack growth ferritic flange flaws forging fracture mechanics fracture toughness grain boundary grain boundary segregation heat treatment heat-affected zone hydrogen ingot inservice inspection integral irradiation Japan light water reactor load manufacturing material mechanical properties metallurgical method microstructure mock-up neutron nondestructive nondestructive testing notch nozzle nuclear power operation parameters pipe plant plate pressure boundary pressure vessel steels Pressurised problems procedures programme quenching reactor pressure vessel reactor vessel recrystallised reduce reliability requirements safety Section shell shown in Fig Škoda specimens steam steelmaking stress corrosion cracking stress intensity stress relief cracks structure surface surveillance temper embrittlement tensile thickness transients transition temperature tube ultrasonic variation weld seams yield strength