Книги з теми "Nuclear reactor vessel"

Server, William L., and Milan Brumovský, eds. International Review of Nuclear Reactor Pressure Vessel Surveillance Programs. 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959: ASTM International, 2018. http://dx.doi.org/10.1520/stp1603-eb.

Johnson, R. E. Radiation effects on reactor pressure vessel supports. Washington, DC: U.S. Nuclear Regulatory Commission, 1996.

Johnson, R. Radiation effects on reactor pressure vessel supports. Washington, DC: Division of Engineering Technology, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 1996.

Hawthorne, J. R. Accelerated irradiation test of Gundremmingen reactor vessel trepan material. Washington, DC: Division of Engineering, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 1992.

Croneberg, August W. In-vessel zircaloy oxidation/hydrogen generation behavior during severe accidents. Washington, D.C: Division of Systems Research, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 1990.

Croneberg, August W. In-vessel zircaloy oxidation/hydrogen generation behavior during severe accidents. Washington, D.C: Division of Systems Research, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 1990.

McCabe, Donald E. Fracture evaluation of surface cracks embedded in reactor vessel cladding. Washington, DC: Division of Engineering, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 1989.

U.S. Nuclear Regulatory Commission. Office of Nuclear Regulatory Research. Division of Engineering., University of Tennessee Knoxville, and Oak Ridge National Laboratory, eds. Extrapolation of the J-R curve for predicting reactor vessel integrity. Washington, DC: Division of Engineering, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 1992.

Remec, I. Neutron spectra at different high flux isotope reactor (HFIR) pressure vessel surveillance locations. Washington, DC: Division of Safety Issue Resolution, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 1993.

Remec, I. Neutron spectra at different high flux isotope reactor (HFIR) pressure vessel surveillance locations. Washington, DC: Division of Safety Issue Resolution, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 1993.

Hawthorne, J. R. Irradiation-anneal-reirradiation (IAR) studies of prototypic reactor vessel weldments. Washington, DC: Division of Engineering, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 1989.

Hawthorne, J. R. Irradiation-anneal-reirradiation (IAR) studies of prototypic reactor vessel weldments. Washington, DC: Division of Engineering, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 1989.

Hawthorne, J. R. Irradiation-anneal-reirradiation (IAR) studies of prototypic reactor vessel weldments. Washington, DC: Division of Engineering, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 1989.

Steele, LE, ed. Radiation Embrittlement of Nuclear Reactor Pressure Vessel Steels: An International Review (Fourth Volume). 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959: ASTM International, 1993. http://dx.doi.org/10.1520/stp1170-eb.

Steele, LE, ed. Radiation Embrittlement of Nuclear Reactor Pressure Vessel Steels: An International Review (Third Volume). 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959: ASTM International, 1989. http://dx.doi.org/10.1520/stp1011-eb.

Steele, LE, ed. Radiation Embrittlement of Nuclear Reactor Pressure Vessel Steels: An International Review (Second Volume). 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959: ASTM International, 1986. http://dx.doi.org/10.1520/stp909-eb.

Chatterjee, S. Integrity assessment of taps reactor pressure vessel at extended EOL using surveillance test results. Mumbai: Bhabha Atomic Research Centre, 2008.

Chambers, R. S. A finite element analysis of a reactor pressure vessel during a severe accident. Washington, DC: Division of Systems Research, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 1989.

Corwin, WR, FM Haggag, and WL Server, eds. Small Specimen Test Techniques Applied to Nuclear Reactor Vessel Thermal Annealing and Plant Life Extension. 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959: ASTM International, 1993. http://dx.doi.org/10.1520/stp1204-eb.

Damiano, B. Current applications of vibration monitoring and neutron noise analysis: Detection and analysis strutural degradation of reactor vessel internals from operational aging. Washington, DC: Division of Engineering, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 1990.

Asgari, M. Determination of the neutron and gamma flux distribution in the pressure vessel and cavity of a boiling water reactor. Washington, DC: Division of Engineering, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 1990.

Hawthorne, J. R. Influence of fluence rate on radiation-induced mechanical property changes in reactor pressure vessel steels: Final report on exploratory experiments. Washington, DC: Division of Engineering, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 1990.

Kikō, Genshiryoku Anzen Kiban. Fukuzatsu keijōbu kiki haikan kenzensei jisshō (IAF) jigyō: Genshiro atsuryoku yōki no izai yōsetsubu ni kansuru kōon zairyō tokusei dēta-shū = Project of integrity assessment of flawed components with structural discontinuity (IAF) : material properties data book at high temperature for dissimilar metal welding in reactor pressure vessel. [Tōkyō-to Minato-ku]: Genshiryoku Anzen Kiban Kikō, 2013.

Kikō, Genshiryoku Anzen Kiban. Fukuzatsu keijōbu kiki haikan kenzensei jisshō (IAF) jigyō: Yōsetsu zanryu ōryoku kaiseki hyōka dēta-shū : yōki kantsūbu yōsetsubu (sashikomi tsugite) = Project of integrity assessment of flawed components with structural discontinuity (IAF) : data book for residual stress analysis in weld joint : weld joint around penetrations in reactor vessel (insert joint). Tōkyō-to Minato-ku: Genshiryoku Anzen Kiban Kikō, 2012.

Kikō, Genshiryoku Anzen Kiban. Fukuzatsu keijōbu kiki haikan kenzensei jisshō (IAF) jigyō: Ōryoku kakudai keisu hyōka dēta-shū : yōki kantsūbu ICM haujingu no hyōmen kiretsu = Project of integrity assessment of flawed components with structural discontinuity (IAF) : data book for estimation stress intensity factor : surface crack on ICM housing for penetration in reactor vessel. Tōkyō-to Minato-ku: Genshiryoku Anzen Kiban Kikō, 2012.

Ludwigsen, John S. Posttest analysis of the steel containment vessel model. Washington, DC: Division of Engineering Technology, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 2000.

Ludwigsen, J. S. Posttest analysis of the steel containment vessel model. Washington, D.C: U.S. Nuclear Regulatory Commission, 2000.

Schuster, G. J. Characterization of flaws in U.S. reactor pressure vessels. Washington, DC: Division of Engineering Technology, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 1998.

Rosdahl, Ö. Assessment of RELAP5/MOD 2 against Marviken jet impingement test 11 level swell. Washington, D.C: U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, 1986.

Rosdahl, Ö. Assessment of RELAP5/MOD 2 against critical flow data from Marviken tests JIT 11 and CFT 21. Washington, D.C: U.S. Nuclear Regulatory Commission, 1986.

Rosdahl, Ö. Assessment of RELAP5/MOD 2 against critical flow data from Marviken tests JIT 11 and CFT 21. Washington, D.C: U.S. Nuclear Regulatory Commission, 1986.

Rosdahl, Ö. Assessment of RELAP5/MOD 2 against critical flow data from Marviken tests JIT 11 and CFT 21. Washington, D.C: U.S. Nuclear Regulatory Commission, 1986.

Rosdahl, Ö. Assessment of RELAP5/MOD 2 against Marviken jet impingement test 11 level swell. Washington, D.C: U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, 1986.

Rolfe, S. T. The behavior of shallow flaws in reactor pressure vessels: Status report. Washington, DC: Division of Engineering, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 1991.

Pennell, W. E. Mission survey for the Pressure Vessel Research User's Facility (PVRUF). Washington, D.C: Division of Engineering, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 1989.

International Conference on Structural Mechanics in Reactor Technology. (9th 1987 August 17-21 Lausanne, Switzerland). Transactions of the 9th International Conference on Structural Mechanics in Reactor Technology, Lausanne, 17-21 August 1987: Structural Mechanics in Reactor Technology. Edited by Wittmann F. H. Rotterdam: A.A. Balkema, 1987.

Libmann, Jacques. Approche et analyse de la sûreté des réacteurs à eau sous pression. Gif-sur-Yvette, France: Commissariat à l'énergie atomique, Institut national des sciences et techniques nucléaires, 1987.

J, Elliot B., and U.S. Nuclear Regulatory Commission. Office of Nuclear Reactor Regulation. Division of Engineering, eds. Reactor pressure vessel status report. Washington, DC: Division of Engineering, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, 1996.

E, Wright J., Odette G. R, U.S. Nuclear Regulatory Commission. Office of Nuclear Regulatory Research. Division of Engineering Technology., Modeling and Computing Services (Firm), and University of California, Santa Barbara., eds. Improved embrittlement correlations for reactor pressure vessel steels. Washington, DC: Division of Engineering Technology, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 1998.

E, Wright J., Odette G. R, U.S. Nuclear Regulatory Commission. Office of Nuclear Regulatory Research. Division of Engineering Technology., Modeling and Computing Services (Firm), and University of California, Santa Barbara., eds. Improved embrittlement correlations for reactor pressure vessel steels. Washington, DC: Division of Engineering Technology, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 1998.

E, Lipinski R., and U.S. Nuclear Regulatory Commission. Office of Nuclear Regulatory Research. Division of Engineering Technology., eds. Radiation effects on reactor pressure vessel supports. Washington, DC: Division of Engineering Technology, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 1996.

Application of Surveillance Programme Results to Reactor Pressure Vessel Integrity Assessment. International Atomic Energy Agency, 2005.

IAEA. Integrity of Reactor Pressure Vessels in Nuclear Power Plants: Assessment of Irradiation Embrittlement Effects in Reactor Pressure Vessel Steels. International Atomic Energy Agency, 2009.

J, McAfee W., U.S. Nuclear Regulatory Commission. Office of Nuclear Regulatory Research. Division of Engineering Technology., and Oak Ridge National Laboratory, eds. Biaxial loading effects on fracture toughness of reactor pressure vessel steel. Washington, DC: U.S. Nuclear Regulatory Commission, 1995.

U.S. Nuclear Regulatory Commission. Office of Nuclear Regulatory Research. Division of Engineering., University of Tennessee Knoxville, and Oak Ridge National Laboratory, eds. Extrapolation of the J-R curve for predicting reactor vessel integrity. Washington, DC: Division of Engineering, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 1992.

U.S. Nuclear Regulatory Commission. Office of Nuclear Regulatory Research. Division of Engineering., University of Tennessee Knoxville, and Oak Ridge National Laboratory, eds. Extrapolation of the J-R curve for predicting reactor vessel integrity. Washington, DC: Division of Engineering, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 1992.

Agency, International Atomic Energy, ed. Application surveillance programme results to reactor pressure vessel integrity assessment: Results of a coordinated research project, 2000-2004. Vienna: International Atomic Energy Agency, 2005.

1928-, Steele L. E., ASTM Committee E-10 on Nuclear Technology and Applications., International Atomic Energy Agency, and Symposium on Irradiation Embrittlement and Aging of Reactor Pressure Vessels (1987 : Philadelphia, Pa.), eds. Radiation embrittlement of nuclear reactor pressure vessel steels: An international review (third volume). Philadelphia, PA: American Society for Testing and Materials, 1989.

1928-, Steele L. E., American Society for Testing Materials., International Atomic Energy Agency. International Working Group on Plant Life Management., and IAEA Specialists' Meeting on Radiation Embrittlement of Nuclear Reactor Pressure Vessel Steels (1990 : Balatonfüred, Hungary), eds. Radiation embrittlement of nuclear reactor pressure vessel steels: An international review (fourth volume). Philadelphia, PA: ASTM, 1993.

U.S. Nuclear Regulatory Commission, ed. Status Report: Reactor Vessel Integrity Database... NUREG-1612... U.S. Nuclear Regulatory Commission... 1997. [S.l: s.n., 1998.