Skip to main content
SpringerLink
Log in

Wichtige Untersysteme des Druckwasserreaktors

  • Chapter
Reaktor-Sicherheitstechnik

  • 56 Accesses

Zusammenfassung

Der Druckwasserreaktor (DWR, engl. PWR, pressurized water reactor) hat heute eine mehr als 20jährige Entwicklung hinter sich. Er wurde zunächst für den Antrieb von Kriegsschiffen entwickelt und gebaut; die erste Anlage zur Stromerzeugung (136 MW) wurde 1957 in Shippingport, USA, fertiggestellt. Der Druckwasserreaktor ist der heute am häufigsten für die kommerzielle Stromerzeugung eingesetzte Typ. Seine Ausführungsform ist weitgehend standardisiert. Bei gleichem grundsätzlichen Aufbau gibt es jedoch bei den verschiedenen Herstellern Unterschiede in der Detailausführung der einzelnen Untersysteme, die interessante Vergleiche ermöglichen, das Verständnis für ihre Wirkungsweise vertiefen und dadurch den Stand der heute erreichten Sicherheitstechnik verdeutlichen. In diesem Kapitel sollen insbesondere die Systeme der Firmen Westinghouse (W), USA, und Kraftwerk-Union (KWU), Bundesrepublik Deutschland, einander gegenübergestellt werden.

This is a preview of subscription content, log in via an institution to check access.

Access this chapter

Log in via an institution
Chapter
USD 29.95
Price excludes VAT (USA)
  • Available as PDF
  • Read on any device
  • Instant download
  • Own it forever
eBook
USD 49.99
Price excludes VAT (USA)
  • Available as PDF
  • Read on any device
  • Instant download
  • Own it forever
Softcover Book
USD 64.99
Price excludes VAT (USA)
  • Compact, lightweight edition
  • Dispatched in 3 to 5 business days
  • Free shipping worldwide - see info

Tax calculation will be finalised at checkout

Purchases are for personal use only

Institutional subscriptions

Preview

Unable to display preview. Download preview PDF.

Unable to display preview. Download preview PDF.

Literatur zu Kapitel 3

  1. Trojan Nuclear Plant, Portland General Electric Company. Final Safety Analysis Report, Docket No. 50–344

    Google Scholar 

  2. Kraftwerk Union, A.G., Kernkraftwerk Philippsburg G.m.b.H., Sicherheitsbericht Kernkraftwerk Phüippsburg (KKPII), Sept. 1975

    Google Scholar 

  3. Reactor Safety Study. An Assessment of Accident Risks in the U.S. Commercial Nuclear Power Plants. Main Report (WASH-1400), U.S. Department of Commerce, National Technical Information Service, PB-248 201, Oct. 1975

    Google Scholar 

  4. Reactor Safety Study, Appendix II, Fault Trees (WASH-1400). U.S. Department of Commerce, Technical Information Service, PB-248 203, Oct. 1975

    Google Scholar 

  5. ASME Boiler and Pressure Vessel Code, Section III-Div. 1, Rules for Construction of Nuclear Power Plant Components. The American Society of Mechanical Engineers, New York, July 1974

    Google Scholar 

  6. Effect of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials. Regulatory Guide 1.99, U.S. Regulatory Commission, Washington, D.C., July 1975

    Google Scholar 

  7. Hawthorne, J. R.: Radiation Effects on Vessel Steels and Welds with Varied Residual Element Content and Embrittlement Relief by Post-Irradiation Annealing; 4th Water Reactor Safety Research Information Meeting. Nucl. Regulatory Commission, Sept. 1976

    Google Scholar 

  8. Licensing of Production and Utilisation Facilities, Title 10 — Chapter 1, Part 50. Fracture Toughness and Surveillance Program Requirements, Federal Register, Vol. 38, No. 130, July 1973

    Google Scholar 

  9. United Kingdom Atomic Energy Authority. An Assessment of the Integrity of PWR Pressure Vessels. Report by a Study Group under the Chairmanship of Dr. W. Marshall, October 1, 1976

    Google Scholar 

  10. Integrity of Reactor Vessels for Light Water Power Reactors, ACRS Report WASH-1285 (Jan. 1974)

    Google Scholar 

  11. Whitman, G. D.; Robinson, G. C; Savolainen, A. W. (eds.): Technology of Steel Pressure Vessels for Water Cooled Nuclear Reactors, a Review of Current Practice in Design, Analysis, Fabrication, Inspection and Test. ORNL NSIC-21 (Dec. 1967), Zusammenfassung von G. D. Whitman in Nucl. Eng. Des. 8 (1968)

    Google Scholar 

  12. Cowan, A.; Kirby, N.; Nichols, R. W.: The Integrity of Pressure Vessels for LWR Systems. TRG Report 2183 (C) (1972)

    Google Scholar 

  13. Report to the American Physical Society by the Study Group on Light Water Reactor Safety. Rev. Mod. Phys. 47, Suppl. No. 1 (1975)

    Google Scholar 

  14. Vinckier, A. G.; Pense, A. W.: A Review of Underclad Cracking in Pressure Vessel Components, Whitaker Laboratory 4 Task Group on Under-clad Cracking, Thermal and Mechanical Treatments Sub-committee, 3rd Draft (January 1st, 1974)

    Google Scholar 

  15. Derby, R. W., et al.: Test of 6 Inch Thick Pressure Vessels, Series 1 Intermediate Test Vessels V1 and V2, HSST Program ORNL-4895, Feb. 1974

    Book  Google Scholar 

  16. Mager, T. R.; Landes, J. D.; Moon, D. M.; McLoughlin, V. J.: The Effect of Low Frequencies on the Fatigue Crack Growth Characteristics of A533 Grade B, Class 1 Plate in an Environment of High Temperature Primary Grade Nuclear Reactor Water. HSST Program Technical Report No. 35 (1973)

    Google Scholar 

  17. Paris, P. C; Bucci, R. J.; Wessel, E. T.; Clar, W. G.; Mager, T. R.: Stress Analysis and Growth of Cracks, ASTM-STP-513 (American Society for Testing and Materials, Philadelphia 1972) p. 141

    Google Scholar 

  18. Mager, T. R.; McLoughlin, V. J.: The Effect of an Environment of High Temperature Primary Grade Nuclear Reactor Water on the Fatigue Growth Characteristics of A533 Grade B Class 1 Plate and Weldment Material. HSST Program Technical Report No. 16 (1971)

    Google Scholar 

  19. Kondo, T.; Kikuyama, T.; Nakajima, H.; Shindo, M.: Mechanical Behaviour of Materials, Vol. III (The Society of Materials Science, Japan), 1972, p. 319

    Google Scholar 

  20. Kondo, T.; Kikuyama, T.; Nakajima, H.; Shindo, M.; Nagasaki, R.: Corrosion Fatigue: Chemistry, Mechanics and Microstructure (International Corrosion Conference Series, NACE-2). Devereaux, O.; McEvily, A. J.; Staehle, R. W. (eds.), (National Association of Corrosion Engineers, Houston, 1972) p. 539

    Google Scholar 

  21. Kondo, T.; Kikuyama, T.; Makajima, H.; Shindo, M.: Fatigue Crack Propagation Behaviour of ASTM A533B and A302B Steels in High Temperature Aqueous Environment, Paper No. 6, HSST Program 6th Annual Information Meeting, ORNL, April 25–26, 1972

    Google Scholar 

  22. Results on Fatigue at 1 and 60 Cycles/min in PWR water at R = 0.63 and 0.7 Tabled by Westinghouse at a Discussion at AERE Harwell, January, 20–22, 1975 (zitiert in [9])

    Google Scholar 

  23. U.S.A.E.C. Technical Report, Analysis of Pressure Vessel Statistics from Fossil Fuelled Power Plant Service and Assessment of Reactor Vessel Reliability in Nuclear Power Plant Service, WASH 1318 (July 1974)

    Google Scholar 

  24. Kellermann, O.; Seipel, H. G.: Containment and Siting of Nuclear Power Plants, I.A.E.A. Conf. Vienna, April 3–7, 1967, p. 403

    Google Scholar 

  25. Kellermann, O.; Kraegeloh, E.; Kußmaul, K.; Sturm, D.: 2nd Conf. Pressure Vessel Technology, Pt. I, San Antonio, Texas, Oct. 1 – 4, 1973, p. 25

    Google Scholar 

  26. Phillips, C. A. G.; Warwick, R. G.: A Survey of Defects in Pressure Vessels Built to High Standards Construction and its Relevance to Nuclear Primary Circuit Envelopes. UKAEA Report, AHSB (S) R 162 (1967)

    Google Scholar 

  27. Smith, T. A.; Warwick, R. G.: Int. J. Pressure Vessels and Piping 2 (1974) 283

    Article  Google Scholar 

  28. Kußmaul, K.; Ewald, J.; Maier, G.; Schellhammer, W.: Enhancement of the Quality of the Reactor Pressure Vessel Used in Light Water Power Plants by Advanced Material, Fabrication and Testing Technologies. 4th International Conf. on Structural Mechanics in Reactor Technology, August 15–19, 1977, San Francisco, USA. Deutsche Übersetzung: 2. VGB Kraft-werkstechnik, 58, Heft 6, S. 439–448 (1978)

    Google Scholar 

  29. Cheverton, R. D.: Studies Associated with the LWR LOCA-ECC Thermal Shock. 4th Water Reactor Safety Research Information Meeting, Nuclear Regulatory Commission, Sept. 27–30, 1976

    Google Scholar 

  30. Bryan, R. H.: Evaluation of Sustained Load Effects in Reactor Pressure Vessels by Means of Intermediate-Scale Flawed Vessel Tests. 4th Water Reactor Safety Research Information Meeting, Nuclear Regulatory Commission, Sept. 27–30, 1976

    Google Scholar 

  31. ASME Boiler and Pressure Vessel Code Section XI, Article A 3000, Method for KI-Determination, Article 4000, Definition of Material Properties. The American Society of Mechanical Engineers, New York

    Google Scholar 

  32. Smidt, D.: Reaktortechnik, 2 Bde., 2. Aufl., Karlsruhe: Braun 1976

    Google Scholar 

  33. Merkle, J. G.; Whitman, G. D.; Bryan, R. H.: An Evaluation of the HSST Program Intermediate Pressure Vessel Tests in Terms of Light Water Reactor Pressure Vessel Safety. ORNLTM-5090, November 1975

    Google Scholar 

  34. Derby, R. W., et al.: Test of Six-Inch-Thick Pressure Vessels, Series I: Intermediate Test Vessels V-1 and V-2. ORNL-4895, February 1974

    Book  Google Scholar 

  35. Bryan, R. H., et al.: Test of Six-Inch-Thick Pressure Vessels, Series II: Intermediate Test Vessels V-3, V-4, and V-6, ORNL-5059, November 1975

    Google Scholar 

  36. Merkle, J. G., et al.: Test of Six-Inch-Thick-Pressure Vessels, Series III: Intermediate Test Vessels V-7, ORNL-5059, August 1976

    Book  Google Scholar 

  37. Whitman, G. D.: Heavy Section Steel Technology Program Quarterly Progress Report tor January thru March 1976, ORNL-TM-28, July 1976

    Book  Google Scholar 

  38. Kußmaul, K.; Ewald, J.: Assessment of Toughness and Cracking in the Heat-Affected Zone ot Light Water Reactor Components. Third International Conference on Pressure Vessel Technology, Tokyo 1977, to be published

    Google Scholar 

  39. Kellermann, O., et al.: Considerations about the Reliability of Nuclear Pressure Vessels, Status and Research Planning, in ASME Second International Conference on Pressure Vessel Technology, San Antonio, Texas, October 1–4, 1973, Part I: Design and Analysis, ASME, New York (1973)25–28

    Google Scholar 

  40. Kußmaul, K.; Kraegeloh, E.: Formation, Significance and Evaluation of Welding Detects in Pressure Vessels, IIW, Annual Assembly 1972 — Public Session, Subdivision 3, Toronto

    Google Scholar 

  41. Ewald J.; Kußmaul, K.: Investigation of the Properties of the Heat-Affected Zone (Methods for Assessing the Properties of the Haze) Review Document March 1976, CEC-NEA Expert s Group on Mechanic and Material Problems Relating to the Safety of Steel Components in Nuclear Plants

    Google Scholar 

  42. Kußmaul, K.; Ewald, J.; Maier, G.: Verfahren zur Simulation der Warmeeinflußzonen von Schmelzschweißverbindungen. Schweißen + Schneiden 28 (1976) 250–255

    Google Scholar 

  43. Kußmaul K.; Blind, D.; Ewald, J.: Methoden zum Nachweis und zur Untersuchung von Korngrenzenschwächungen und Rissen in der Wärmeeinflußzone von Schweißverbindungen an Druckbehältern. Schweißen + Schneiden 28 (1975) 219–223

    Google Scholar 

  44. Forschungsprogramm Zentrale Auswertung von HersteUungsfehlern und Schäden im Hinblick auf druckführende Anlagenteile von Kernkraftwerken SR 10. Durchgeführt im Auftrag des Bundesministeriums des Innern, l.Techn. Bericht: Schellhammer, W.; Maile, K.; Maier, S., Staatliche Materialprüfungsanstalt (MPA), Universität Stuttgart, Januar 1976

    Google Scholar 

  45. Kußmaul, K.; Blind, D.; Ewald, J.: Investigation for the Detection and Study of Stress-Reliet Cracking. International Journal of Pressure Vessel and Piping, to be published

    Google Scholar 

  46. Kußmaul, K.: Persönliche Mitteilung

    Google Scholar 

  47. Application of the Single Failure Criterion to Nuclear Power Plant Protection Systems. US-NRC Reg. Guide 1.53

    Google Scholar 

  48. Leitlinien für Druckwasserreaktoren der Reaktorsicherheitskommission.

    Google Scholar 

  49. Zienkiewicz, O.: The Finite Element Method in Engineering Science. London: McGraw-Hill: 1971

    MATH  Google Scholar 

  50. Griffith, A. A.: The Phenomena of Rupture and Flow in Solids. Philos. Trans. R. Soc. London, A 221 (1921) 163–198

    Article  Google Scholar 

  51. Irwin, G. R.: Fracture. In: Handbuch der Physik, Bd. VI, Elastizität und Plastizität. Berlin, Göttingen, Heidelberg: Springer 1958

    Google Scholar 

  52. Hahn, H. G.: Bruchmechanik. Teubner 1976

    MATH  Google Scholar 

  53. Bryan, R. H., et al.: Heavy Section Steel Technology Program Intermediate-Scale Pressure Vessel Test. 4th International Conference on Structural Mechanics in Reactor Technology, San Francisco, Aug. 15–19, 1977

    Google Scholar 

  54. Gesellschaft für Reaktorsicherheit, Köln. Persönliche Mitteilung

    Google Scholar 

  55. US-NRCReg. Guide 1.53, 10 CFR 50

    Google Scholar 

  56. USAEC, Acceptance Criteria for Emergency Core Cooling Systems, Light-Water-Cooled Nuclear Power Reactors, Docket No. RM-50–1

    Google Scholar 

  57. Kußmaul, K.: Persönliche Mitteilung

    Google Scholar 

  58. Eickelpasch, N.: Untersuchungen zur Reduzierung der Strahlenbelastung des in Kernkraftwerken eingesetzten Personals. Dissertation Univ. Karlsruhe, 1978

    Google Scholar 

  59. Bush, S. H.: Probability of Damage to Nuclear Components Due to Turbine Failure. Nucl. Saf., 14 (1973) 187–201

    Google Scholar 

  60. Yeh, G. C. K.: Probability and Containment of Turbine Missiles, Paper O 2/6. International Seminar on Extreme Load Conditions and Limit Analysis Procedures for Structural Reactor Safeguards and Containment Structures, Berlin, September 8–11, 1975

    Google Scholar 

  61. O’Connell, W. J.; Baschiere, R. J.: Design Applications of Turbine Missile Impact Analysis. Second ASCE Specialty Conference on Structural Design of Nuclear Plant Facilities, New Orleans, Louisiana, December 8–10, 1975, Preprint Vol. I-A, pp. 541–561

    Google Scholar 

  62. Steele, L. E.: Neutron Irradiation Embrittlement of Reactor Pressure Vessel Steels. At. Energy Rev. 7 (1969) No. 2

    Google Scholar 

  63. Leitz, C.: Berücksichtigung der Strahlenversprödung bei der Auslegung von Reaktordruckbehältern. Atomwirtschaft 17 (1972) 614–619

    Google Scholar 

  64. Varsik, I. D., et al.: An Empirical Evaluation of the Irradiation Sensitivity of Reactor Pressure Vessel Materials. 9th ASTM Int. Symp. on Effects of Radiation on Structural Materials. Richland, Wash., July 1978

    Google Scholar 

  65. Steele, L. E.: Neutron Irradiation Embrittlement of Reactor Pressure Vessel Steels. IAEA, Technical Report Series No. 163, Vienna (1975)

    Google Scholar 

  66. Kußmaul, K.: Maßnahmen und Prüfkonzept zur weiteren Verbesserung der Qualität von Reaktordruckbehältern für Leichtwasser-Kernkraftwerke, 2. VGB — Kraftwerkstechnik 58, Heft 6, S. 439–448(1978)

    Google Scholar 

  67. Kußmaul, K.; Stoppler, W.: Temperaturführung bei und nach dem Schweißen. Atomwirtschaft 23, Nr. 7/8, S. 354–361 (1978)

    Google Scholar 

Download references

Author information

Authors and Affiliations

  1. Institut für Reaktortechnik, Universität Karlsruhe, Deutschland

    Professor Dr. Dieter Smidt

  2. Institut für Reaktortechniklung, Kernforschungszentrums Karlsruhe, Deutschland

    Professor Dr. Dieter Smidt

Authors
  1. Professor Dr. Dieter Smidt
    View author publications

    You can also search for this author in PubMed Google Scholar

Rights and permissions

Reprints and permissions

Copyright information

© 1979 Springer-Verlag, Berlin, Heidelberg

About this chapter

Cite this chapter

Smidt, D. (1979). Wichtige Untersysteme des Druckwasserreaktors. In: Reaktor-Sicherheitstechnik. Springer, Berlin, Heidelberg. https://doi.org/10.1007/978-3-642-50225-5_3

Download citation

  • DOI: https://doi.org/10.1007/978-3-642-50225-5_3

  • Publisher Name: Springer, Berlin, Heidelberg

  • Print ISBN: 978-3-642-50226-2

  • Online ISBN: 978-3-642-50225-5

  • eBook Packages: Springer Book Archive

Publish with us

Policies and ethics

Search

Navigation