An assessment of the failure rate for the beltline region of PWR pressure vessels during normal operation and certain transient conditions /

The octavia computer code : PWR reactor pressure vessel failure probabilities due to operationally caused pressure transients /

Vesely, W. E.
Goldberg, F. F.
Lynn, E. K.
U.S. Nuclear Regulatory Commission. Office of Nuclear Regulatory Research. Probabilistic Analysis Staff.
U.S. Nuclear Regulatory Commission. Office of Nuclear Regulatory Research. Metallurgy and Materials Research Branch.

Prepared in cooperation with Probabilities Analysis Branch.March 1978."NUREG-0258."Includes bibliogr...

Preliminary analysis of the effect of fatigue loading and crack propagation on crack acceptance criteria for nuclear power plant components /

Anderson, W. F.
Johns, E. C. (Earl C.)
Weidenhamer, G. H.
U.S. Nuclear Regulatory Commission. Division of Engineering Standards.

"NUREG-0726.""Date published: April 1981."Includes bibliographical references.Mode of access: Intern...

Summary of tube integrity operating experience with once-through steam generators /

Liaw, Boen-dar.
Strosnider, J.
U.S. Nuclear Regulatory Commission. Division of Operating Reactors.

"NUREG-0571.""Date published: March 1980."Includes bibliographical references.Mode of access: Intern...

Investigation and evaluation of cracking incidents in piping in pressurized water reactors /

U.S. Nuclear Regulatory Commission.
PWR Pipe Crack Study Group (U.S.)

January 1980

"NUREG-0691.""Date published : September 1980."Includes bibliographical references.Mode of access: I...

Reactor vessel pressure transient protection for pressurized water reactors : final report /

Zech, Gary.
U.S. Nuclear Regulatory Commission. Division of Operating Reactors.

"Date published: September 1978."NUREG-0223."Includes bibliographical references.Mode of access: Int...

Evaluation of the behavior of waterlogged fuel rod failure in LWRs /

Siegel, B.
U.S. Nuclear Regulatory Commission. Division of Systems Safety.

"Date published: March l978.""NUREG-0303."Includes bibliographical references.Mode of access: Intern...

Fatigue crack growth rates in pressure vessel and piping steels in LWR environments : final report /

Cullen, W. H.
U.S. Nuclear Regulatory Commission. Office of Nuclear Regulatory Research. Division of Engineering Technology.
Materials Engineering Associates.

January 1987

"Date published: March 1987.""NRC FIN B8900.""MEA-2175.""NUREG/CR-4724."Bibliography: p. 50-54.Mode ...

Guidelines for confirmatory inplant tests of safety-relief valve discharges for BWR plants /

Su, T. M.
U.S. Nuclear Regulatory Commission. Office of Nuclear Reactor Regulation. Division of Safety Technology.

"Date published: May 1981.""NUREG-0763."Includes bibliographical references (p. 15).Mode of access: ...

Computerized reactor pressure vessel materials information system /

Strosnider, J.
Tether, C. D.
Kenworthy, L. D.
Monserrate, C.
U.S. Nuclear Regulatory Commission. Office of Management and Program Analysis.
U.S. Nuclear Regulatory Commission. Office of Nuclear Reactor Regulation.

"NUREG-0688.""Date published: October 1980."Includes bibliographical references.Mode of access: Inte...

Transient response of Babcock & Wilcox-designed reactors /

U.S. Nuclear Regulatory Commission. B & W Reactor Transient Response Task Force.

January 1980

"NUREG-0667.""Date published: May 1980."Includes bibliographical references.Mode of access: Internet

Generic assessment of delayed reactor coolant pump trip during small break loss-of-coolant accidents in pressurized water reactors /

Sheron, B. (Brian)
U.S. Nuclear Regulatory Commission. Office of Nuclear Reactor Regulation. Bulletins and Orders Task Force.

January 1979

"NUREG-0623.""Date Published: November 1979."Includes bibliographical references.Mode of access: Int...

Technical bases for estimating fission product behavior during LWR accidents /

Sandia National Laboratories.
Oak Ridge National Laboratory.
Battelle Memorial Institute. Columbus Laboratories.
U.S. Nuclear Regulatory Commission. Office of Nuclear Reactor Regulation.
U.S. Nuclear Regulatory Commission. Office of Nuclear Regulatory Research.

January 1981

"NUREG-0772.""Date published: June 1981."Includes bibliographical references.Mode of access: Interne...

Generic evaluation of feedwater transients and small break loss-of-coolant accidents in Combustion Engineering designed operating plants /

U.S. Nuclear Regulatory Commission. Office of Nuclear Reactor Regulation. Bulletins and Orders Task Force.

January 1980

"NUREG-0635.""Date published: January 1980."Includes bibliographical references.Mode of access: Inte...

Report on the integrity of reactor vessels for light-water power reactors /

U.S. Atomic Energy Commission. Advisory Committee on Reactor Safeguards.
U.S. Atomic Energy Commission.

January 1974

"January 1974.""UC-78."Includes bibliographical references (page 76).Mode of access: Internet

Hydraulic studies of the lower plenum of a pressure water reactor vessel /

Weber, E.
Dittrich, W.
Germany. Bundesministerium für Forschung und Technologie.
Simens Aktiengesellschaft.

January 1977

"Translation series.""Date published by NRC: October 1977.""Original report number: BMFT-FB K 73-03....

The octavia computer code : PWR reactor pressure vessel failure probabilities due to operationally caused pressure transients /

Vesely, W. E.
Goldberg, F. F.
Lynn, E. K.
U.S. Nuclear Regulatory Commission. Office of Nuclear Regulatory Research. Probabilistic Analysis Staff.
U.S. Nuclear Regulatory Commission. Office of Nuclear Regulatory Research. Metallurgy and Materials Research Branch.

Prepared in cooperation with Probabilities Analysis Branch.March 1978."NUREG-0258."Includes bibliogr...

Preliminary analysis of the effect of fatigue loading and crack propagation on crack acceptance criteria for nuclear power plant components /

Anderson, W. F.
Johns, E. C. (Earl C.)
Weidenhamer, G. H.
U.S. Nuclear Regulatory Commission. Division of Engineering Standards.

"NUREG-0726.""Date published: April 1981."Includes bibliographical references.Mode of access: Intern...

Summary of tube integrity operating experience with once-through steam generators /

Liaw, Boen-dar.
Strosnider, J.
U.S. Nuclear Regulatory Commission. Division of Operating Reactors.

"NUREG-0571.""Date published: March 1980."Includes bibliographical references.Mode of access: Intern...

Investigation and evaluation of cracking incidents in piping in pressurized water reactors /

U.S. Nuclear Regulatory Commission.
PWR Pipe Crack Study Group (U.S.)

January 1980

"NUREG-0691.""Date published : September 1980."Includes bibliographical references.Mode of access: I...

Reactor vessel pressure transient protection for pressurized water reactors : final report /

Zech, Gary.
U.S. Nuclear Regulatory Commission. Division of Operating Reactors.

"Date published: September 1978."NUREG-0223."Includes bibliographical references.Mode of access: Int...

Evaluation of the behavior of waterlogged fuel rod failure in LWRs /

Siegel, B.
U.S. Nuclear Regulatory Commission. Division of Systems Safety.

"Date published: March l978.""NUREG-0303."Includes bibliographical references.Mode of access: Intern...

Fatigue crack growth rates in pressure vessel and piping steels in LWR environments : final report /

Cullen, W. H.
U.S. Nuclear Regulatory Commission. Office of Nuclear Regulatory Research. Division of Engineering Technology.
Materials Engineering Associates.

January 1987

"Date published: March 1987.""NRC FIN B8900.""MEA-2175.""NUREG/CR-4724."Bibliography: p. 50-54.Mode ...

Guidelines for confirmatory inplant tests of safety-relief valve discharges for BWR plants /

Su, T. M.
U.S. Nuclear Regulatory Commission. Office of Nuclear Reactor Regulation. Division of Safety Technology.

"Date published: May 1981.""NUREG-0763."Includes bibliographical references (p. 15).Mode of access: ...

Computerized reactor pressure vessel materials information system /

Strosnider, J.
Tether, C. D.
Kenworthy, L. D.
Monserrate, C.
U.S. Nuclear Regulatory Commission. Office of Management and Program Analysis.
U.S. Nuclear Regulatory Commission. Office of Nuclear Reactor Regulation.

"NUREG-0688.""Date published: October 1980."Includes bibliographical references.Mode of access: Inte...

Transient response of Babcock & Wilcox-designed reactors /

U.S. Nuclear Regulatory Commission. B & W Reactor Transient Response Task Force.

January 1980

"NUREG-0667.""Date published: May 1980."Includes bibliographical references.Mode of access: Internet

Generic assessment of delayed reactor coolant pump trip during small break loss-of-coolant accidents in pressurized water reactors /

Sheron, B. (Brian)
U.S. Nuclear Regulatory Commission. Office of Nuclear Reactor Regulation. Bulletins and Orders Task Force.

January 1979

"NUREG-0623.""Date Published: November 1979."Includes bibliographical references.Mode of access: Int...

Technical bases for estimating fission product behavior during LWR accidents /

Sandia National Laboratories.
Oak Ridge National Laboratory.
Battelle Memorial Institute. Columbus Laboratories.
U.S. Nuclear Regulatory Commission. Office of Nuclear Reactor Regulation.
U.S. Nuclear Regulatory Commission. Office of Nuclear Regulatory Research.

January 1981

"NUREG-0772.""Date published: June 1981."Includes bibliographical references.Mode of access: Interne...

Generic evaluation of feedwater transients and small break loss-of-coolant accidents in Combustion Engineering designed operating plants /

U.S. Nuclear Regulatory Commission. Office of Nuclear Reactor Regulation. Bulletins and Orders Task Force.

January 1980

"NUREG-0635.""Date published: January 1980."Includes bibliographical references.Mode of access: Inte...

Report on the integrity of reactor vessels for light-water power reactors /

U.S. Atomic Energy Commission. Advisory Committee on Reactor Safeguards.
U.S. Atomic Energy Commission.

January 1974

"January 1974.""UC-78."Includes bibliographical references (page 76).Mode of access: Internet

Hydraulic studies of the lower plenum of a pressure water reactor vessel /

Weber, E.
Dittrich, W.
Germany. Bundesministerium für Forschung und Technologie.
Simens Aktiengesellschaft.

January 1977

"Translation series.""Date published by NRC: October 1977.""Original report number: BMFT-FB K 73-03....

The octavia computer code : PWR reactor pressure vessel failure probabilities due to operationally caused pressure transients /

Vesely, W. E.
Goldberg, F. F.
Lynn, E. K.
U.S. Nuclear Regulatory Commission. Office of Nuclear Regulatory Research. Probabilistic Analysis Staff.
U.S. Nuclear Regulatory Commission. Office of Nuclear Regulatory Research. Metallurgy and Materials Research Branch.

Prepared in cooperation with Probabilities Analysis Branch.March 1978."NUREG-0258."Includes bibliogr...

Preliminary analysis of the effect of fatigue loading and crack propagation on crack acceptance criteria for nuclear power plant components /

Anderson, W. F.
Johns, E. C. (Earl C.)
Weidenhamer, G. H.
U.S. Nuclear Regulatory Commission. Division of Engineering Standards.

"NUREG-0726.""Date published: April 1981."Includes bibliographical references.Mode of access: Intern...

Summary of tube integrity operating experience with once-through steam generators /

Liaw, Boen-dar.
Strosnider, J.
U.S. Nuclear Regulatory Commission. Division of Operating Reactors.

"NUREG-0571.""Date published: March 1980."Includes bibliographical references.Mode of access: Intern...

An assessment of the failure rate for the beltline region of PWR pressure vessels during normal operation and certain transient conditions /

Abstract

Extracted data

An assessment of the failure rate for the beltline region of PWR pressure vessels during normal operation and certain transient conditions /

Abstract

Extracted data

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