Advertisement

PWR and BWR Anticipated and Abnormal Plant Transient Research Sponsored by the U.S. Nuclear Regulatory Commission

  • William D. Beckner
  • Fuat Odar
  • L. Harold Sullivan

Abstract

Increased attention is being given to the analysis of anticipated plant transients and abnormal plant operating conditions. This is in contrast to past regulatory activities devoted primarily to the analysis of design basis accidents (DBA), such as the large-break loss-of-coolant accident (LOCA), which were designed to represent worst or bounding cases. This change has come about due to the realization that anticipated transients with multiple failures (including operator actions) are the major contributors to risk from reactor operations and that the low probability DBAs are relatively small contributors to risk. The accident at Three Mile Island Unit 2 (TMI-2) has shown that anticipated transients, combined with multiple equipment and/or operator failures, can result in significant core damage.

Keywords

Steam Generator Electric Power Research Institute Pressurize Water Reactor Nuclear Regulatory Commission Boiling Water Reactor 
These keywords were added by machine and not by the authors. This process is experimental and the keywords may be updated as the learning algorithm improves.

Preview

Unable to display preview. Download preview PDF.

Unable to display preview. Download preview PDF.

References

  1. 1.
    “Clarification of TMI Action Plan Requirements,” NUREG-0737, U.S. Nuclear Regulatory Commission, November 1980.Google Scholar
  2. 2.
    “Generic Evaulation of Small Break Loss-of-Coolant Accident Behavior in Babcock & Wilcox Designed 177-FA Operating Plants,” NUREG-0565, U.S. Nuclear Regulatory Commission, January 1980.Google Scholar
  3. 3.
    W.S. Hwang, “BWR Small Break Simulation Tests With and Without Degraded ECC Systems-BWR Blowdown/Emergency Core Cooling Program,” NUREG CR-2230, General Electric Co., January 1982.Google Scholar
  4. 4.
    W.S. Hwang, “Analysis of TLTA Small Break Test Results,” NEDO-24823, General Electric Co., August 1980.Google Scholar
  5. 5.
    C.B. Davis, “RELAP5 Calculations of the Effect of Primary Coolant Pump Operation During Semiscale Small Fireak Experiments,” EGG-CAAD-5531, Idaho National Engineering Laboratory, August 1981.Google Scholar
  6. 6.
    K.G. Condie, et al., “Four-Inch Equivalent Break Loss-of Coolant Experiments: Posttest Analysis of LOFT Experiments L3-1, L3-5, (Pumps Off), and L3-6 (Pumps On),” EGG-LOFT-5480, Idaho National Engineering Laboratory, October 1981.Google Scholar
  7. 7.
    D.J. Shimeck, et al., “Analysis of Primary Feed and Bleed Cooling in PWR Systems,” EGG-SEMI-6022, Idaho National Engineering Laboratory, September 1982.Google Scholar
  8. 8.
    F.X. Dolan, et al., “Facility and Test Design Report-1/2 Scale Thermal Mixing Project,” NUREG/CR-3426, Creare, September 1983.Google Scholar
  9. 9.
    U.S. Rohatgi, et al., “Assessment of Selected TRAC and RELAP5 Calculations for Oconee-1 Pressurized Thermal Shock Study,” Brookhaven National Laboratory, To Be Published.Google Scholar
  10. 10.
    C.D. Fletcher, et al., “Thermal Shock Sequences for the Oconee-1 Pressurized Water Reactor,” EGG-NSMD-6343, Idaho National Engineering Laboratory, July 1983.Google Scholar
  11. 11.
    B. Bassitt, et al., “TRAC Analysis of Severe Overcooling Transients for the Oconee-1 PWR,” Los Alomos National Laboratory, to be published.Google Scholar
  12. 12.
    C.L. Nalezny, “Summary of Nuclear Regulatory Commission LOFT Program Experiments,” NUREG/CR-3214, Idaho National Engineering Laboratory, July 1983.Google Scholar
  13. 13.
    G.G. Loomis and K. Soda, “Results of the Semiscale M0D-2A Natural Circulation Experiments,” NUREG/CR-2335, Idaho National engineering Laboratory, October 1982.Google Scholar
  14. 14.
    W.W. Tingle, “Test Data Report on Westinghouse Reactor Vessel Level Indicating System Performance During Semiscale Test S-UT-8,” EGG-SEMI-5827, Idaho National Engineering Laboratory, March 1982.Google Scholar
  15. 15.
    M.T. Leonard, “Vessel Mass Depletion During A Small Break LOCA,” EGG-SEMI-6010, Idaho National Engineering Laboratory, Sept. 1982.Google Scholar
  16. 16.
    T.J. Boucher and R.A. Dimenna, “Semiscale MOD-2A Intermediate Break Test Series Results Comparison,” NUREG/CR-3126, Idaho National Engineering Laboratory, Feburary 1983.Google Scholar
  17. 17.
    J.E. Steit, “Posttest RELAP5 Simulations of the Semiscale MOD-2A Feedwater line Break Tests S-SF-1, 2, and 3C,” EGG-SEMI-6062, Idaho National Engineering Laboratory, October 1982.Google Scholar
  18. 18.
    R.A. Shaw “Posttest RELAP5 Simulations of the Semiscale S-SF-4 & 5 Steam Line Break Experiments,” EGG-SEMI-6106, Idaho National Engineering Laboratory, November 1982.Google Scholar
  19. 19.
    D.J. Shimeck, “Analysis of Semiscale M0D-2A System UHI/SBLOCA Experiments,” NUREG/CR-3195, Idaho National Engineering Laboratory, April 1983.Google Scholar
  20. 20.
    C. Johnsen, “Semiscale Loss-of-Offsite Power Test Results,” to be presented at the NRC 11th Water Reactor Safety Research Information Meeting, Gaithersburg, Maryland, October 24-28, 1983.Google Scholar
  21. 21.
    R. Dimenna, “Semiscale Steam Generator Tube Rupture Test Result,” to to presented at the NRC 11th Water Reactor Safety Research Information Meeting, Gaithersburg, Maryland, October 24-28, 1983.Google Scholar
  22. 22.
    J.E. Thompson, “BWR Full Integral Simulation Test (FIST) Program Test Plan,” NUREG/CR-2575, General Electric Co., April 1982.Google Scholar
  23. 23.
    “Test Advisory Group Final Report-Integral System Testing Program For B&W-Designed NSS,” BAW-1787, Babcock & Wilcox, June 1983.Google Scholar
  24. 24.
    “Multiloop Integral Systems Test (MIST) Facility Specifications,” RDP:84:4091-01-01:01, Babcock & Wilcox, to be published.Google Scholar
  25. 25.
    “Thermal and Hydraulic Code Verification: ATHOS2 and Model Boiler No. 2 Data,” NP-2887, Electric Power Research Institute, February 1983.Google Scholar
  26. 26.
    B.E. Boyack, “TRAC-PF1 Developmental Assessment,” NUREG/CR-3280, Los Alamos National Laboratory, July 1983.Google Scholar
  27. 27.
    J.W. Spore, et al., “TRAC-BD1: An Advanced Best Estimate Computer Program for Boiling Water Reactor Loss-of-Coolant Accident Analysis,” NUREG/CR-2178, Idaho National Engineering Laboratory, October 1981.Google Scholar
  28. 28.
    W. Wulff, et al., “A Description and Assessment of RAMONA-3B A Computer Code With Three Dimensional Neutron Kinetics for BWR System Transients,” Brookhaven National Laboratory, To Be Published.Google Scholar
  29. 29.
    V.H. Ransom, et al., “RELAP5/M0D1 Code Manual,” NUREG/CR-1826, Idaho National Engineering Laboratory, March 1982.Google Scholar
  30. 30.
    “Nuclear-Power-Plant Malfunction Analysis,” IEEE SPECTRUM, June 1983, pp. 53-58.Google Scholar
  31. 31.
    W. Wulff, et al., “On-Line Prediction of BWR Transients in Support of Plant Operation and Safety Analyses,” ANS Topical Meeting On Anticipated and Abnormal Plant Transients in Light Water Reactors, Jackson, Wyoming, September 1983.Google Scholar

Copyright information

© Springer Science+Business Media New York 1984

Authors and Affiliations

  • William D. Beckner
    • 1
  • Fuat Odar
    • 1
  • L. Harold Sullivan
    • 2
  1. 1.U.S. Nuclear Regulatory CommissionUSA
  2. 2.Los Alamos National LaboratoryLos AlamosUSA

Personalised recommendations