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Safety

  • Yoshiaki Oka
  • Seiichi Koshizuka
  • Yuki Ishiwatari
  • Akifumi Yamaji
Chapter

Abstract

This chapter covers safety topics of the Super LWR. The safety principle is introduced first. The safety system design based on the safety principle is presented. Safety analysis codes for the deterministic approach to the Super LWR safety are introduced. Safety criteria for the integrities of fuel rod and pressure boundary are proposed. Selection, classification, and analyses of the possible abnormal transients and accidents in the Super LWR are presented. There are several key safety characteristics of the Super LWR that are inherent in the design features and their benefits have been identified through systematic safety analyses. A transient subchannel analysis code for the Super LWR is prepared and applied to the flow decreasing events to investigate the influence of cross flow in the fuel assemblies on the safety margin. For the probabilistic approach to the Super LWR safety, simplified level-1 PSA is performed.

Keywords

Fuel Assembly Coolant Flow Supercritical Pressure Coolant Flow Rate Core Damage 
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.

Notes

Glossary

ABWR

advanced boiling water reactor

ADS

automatic depressurization system

AFS

auxiliary feedwater system

ATWS

anticipated transient without scram

BDBE

beyond design basis event

BWR

boiling water reactor

CDF

core damage frequency/cumulative damage fraction

ECCS

emergency core cooling system

E/G

emergency diesel generator

FPP

fossil-fired power plant

HEM

homogeneous equilibrium model

LOCA

loss of coolant accident

LOSP

loss of offsite power

LPCI

low pressure core injection

LWR

light water reactor

MSIV

main steam isolation valve

PCMI

pellet cladding mechanical interaction

PCS

power conversion system

PCT

peak cladding temperature

PSA

probabilistic safety assessment

PWR

pressurized water reactor

RCIC

reactor core isolation cooling

RCP

reactor coolant pump

RHR

residual heat removal

ROSP

recovery of offsite power

RPS

reactor protection system

RPV

reactor pressure vessel

SCWR

supercritical pressure water cooled reactor

SG

steam generator

SLCS

standby liquid control system

SRV

safety relief valve

Super LWR

high temperature thermal reactor version of SCWR

References

  1. 1.
    Y. Ishiwatari, Y. Oka and S. Koshizuka, “Safety of the Super LWR,” Nuclear Engineering and Technology, Vol. 39(4), 257–272 (2007)CrossRefGoogle Scholar
  2. 2.
    Y. Ishiwatari, Y. Oka, S. Koshizuka, A. Yamaji and J. Liu, “Safety of Super LWR, (I) Safety System Design,” Journal of Nuclear Science and Technology, Vol. 42(11), 927–934 (2005)CrossRefGoogle Scholar
  3. 3.
    Y. Ishiwatari, “Safety of Super LWR,” Doctoral thesis, the University of Tokyo (2006) (in Japanese)Google Scholar
  4. 4.
    Y. Ishiwatari, Y. Oka, S. Koshizuka and J. Liu, “ATWS Characteristics of Super LWR with/without Alternative Action,” Journal of Nuclear Science and Technology, Vol. 44(4), 572–580 (2007)CrossRefGoogle Scholar
  5. 5.
    Y. Ishiwatari, Y. Oka, S. Koshizuka, A. Yamaji and J. Liu, “Safety of Super LWR, (II) Safety Analysis at Supercritical Pressure,” Journal of Nuclear Science and Technology, Vol. 42(11), 935–948 (2005)CrossRefGoogle Scholar
  6. 6.
    Y. Ishiwatari, Y. Oka, S. Koshizuka and J. Liu, “LOCA Analysis of Super LWR,” Journal of Nuclear Science and Technology, Vol. 43(3), 231–241 (2006)CrossRefGoogle Scholar
  7. 7.
    K. Kitoh, S. Koshizuka and Y. Oka, “Refinement of Transient Criteria and Safety Analysis for a High-temperature Reactor Cooled by Supercritical Water,” Nuclear Technology, Vol. 135, 252–264 (2001)Google Scholar
  8. 8.
    F. D. Coffman, Jr., “LOCA Temperature Criterion for Stainless Steel Clad Fuel,” NUREG-0065, (1976)Google Scholar
  9. 9.
    Y. F. Shen, Z. D. Cao and Q. G. Lu, “An Investigation of Cross Flow Mixing Effect Caused by Grid Spacer with Mixing Blades in Rod Bundle,” Nuclear Engineering and Design, Vol. 125(2), 111–119 (1991)CrossRefGoogle Scholar
  10. 10.
    F. W. Dittus and L. M. K. Boelter, “Heat Transfer in Automobile Radiators of the Tubular Type,” University of California Publications in English, Berkeley, Vol. 2, 443–461 (1930)Google Scholar
  11. 11.
    Proposed standard ANS-5.1 – 1971, American Nuclear Society (1971)Google Scholar
  12. 12.
    K. Kamei, “Core Design of Super LWR and Its Safety Analysis at Subcritical-pressure,” Master’s thesis, the University of Tokyo (2006) (in Japanese)Google Scholar
  13. 13.
    K. V. Moore and W. H. Rettig, “RELAP-4: A Computer Program for Transient Thermal-hydraulic Analysis,” ANCR-1127, Aerojet Nuclear Company (1973)Google Scholar
  14. 14.
    J. R. S. Thom, W. M. Walker, T. A. Fallon and G. F. S. Reising, “Boiling in Subcooled Water During Flow Up Heated Tubes or Annuli,” Proc. Inst. Mech. Eng. 180 (Part 3C) (1966)Google Scholar
  15. 15.
    V. E. Schrock and I. N. Grossman, “Forced Convection Boiling Studies, Final Report on Forced Convection Vaporization Project,” TID-14632 (1959)Google Scholar
  16. 16.
    J. B. McDonough, W. Milich and E. C. King, “Partial Film Boiling with Water at 2000 psig in a Round Vertical Tube,” MSA Research Corp., Technical Report 62 (1958) (NP-6976)Google Scholar
  17. 17.
    D. C. Groeneveld, L. K. H. Leung, A. Z. Vasic, Y. J. Guo and S. C. Cheng, “A Look up Table for Fully Developed Film Boiling Heat Transfer,” Nuclear Engineering and Design, Vol. 225, 83–97 (2003)CrossRefGoogle Scholar
  18. 18.
    D. C. Groeneveld, L. K. H. Leung, P. L. Kirillov, V. P. Bobkov, I. P. Smogalev, V. N. Vinogradov, X. C. Huang and E. Royer, “The 1995 Look-up Table for Critical Heat Flux in Tubes,” Nuclear Engineering and Design, Vol. 163, 1–23 (1996)CrossRefGoogle Scholar
  19. 19.
    R. C. Martinelli and D. B. Nelson, “Prediction of Pressure Drop During Forced-circulation Boiling of Water,” Transactions of ASME, Vol. 71, 695–702 (1948)Google Scholar
  20. 20.
    J. H. Lee, S. Koshizuka and Y. Oka, “Development of a LOCA Analysis Code for the Supercritical-pressure Light Water Cooled Reactors,” Annals of Nuclear Energy, Vol. 25 (16), 1341–1361 (1998)CrossRefGoogle Scholar
  21. 21.
    J. H. Lee, “LOCA Analysis and Safety System Consideration for the Supercritical-Water Cooled Reactor,” Doctoral thesis, the University of Tokyo (1996)Google Scholar
  22. 22.
    N. E. Todreas and M. S. Kazimi, “Nuclear Systems I – Thermal Hydraulic Fundamentals,” Hemisphere Publishing Corporation, ISBN 0-89116-935-0 (1990)Google Scholar
  23. 23.
    F. M. Bordelon, et al., “SATAN IV Program: Comprehensive Space-time Dependent Analysis of Loss of Coolant,” WCAP-8302 (1974)Google Scholar
  24. 24.
    A. Yamanouchi, “Effect of Core Spray Cooling in Transient Stat After Loss of Coolant Accident,” Journal of Nuclear Science and Technology, Vol. 5(11), 547–558 (1968)CrossRefGoogle Scholar
  25. 25.
    Y. Murao and T. Hojo, “Numerical Simulation of Reflooding Behavior in Tight-Lattice Rod Bundles,” Nuclear Technology, Vol. 80, 83 (1998)Google Scholar
  26. 26.
    Anticipated Transients Without Scram for Light Water Reactors, NUREG-0460, US-NRC (1978)Google Scholar
  27. 27.
    Preliminary Safety Analysis Report Lungmen Nuclear Power Station Units 1 & 2, GE Nuclear Energy (1997)Google Scholar
  28. 28.
    ABWR Standard Safety Analysis Report, 23A6100 Rev. 1, GE Nuclear Energy (1993)Google Scholar
  29. 29.
    D. Y. Oh, S. H. Ahn and I. G. Kim, “Sensitivity Study on the Safety Parameters During ATWS with/without AMSAC,” Proc. ICAPP’03, Cordoba, Spain, May 4–7, 2003, Paper 3149 (2003)Google Scholar
  30. 30.
    Y. Okano, S. Koshizuka, K. Kitoh and Y. Oka, “Flow-induced Accident and Transient Analyses of a Direct-cycle, Light-water Cooled, Fast Breeder Reactor Operating at Supercritical Pressure,” Journal of Nuclear Science and Technology, Vol. 33(4), 307–315 (1996)CrossRefGoogle Scholar
  31. 31.
    K. Kamei, A. Yamaji, Y. Ishiwatari, Y. Oka and J. Liu, “Fuel and Core Design of Super Light Water Reactor with Low Leakage Fuel Loading Pattern,” Journal of Nuclear Science and Technology, Vol. 43(2), 129–139 (2006)CrossRefGoogle Scholar
  32. 32.
    A. A. Amsden and F. H. Harlow, “The SMAC Method: A Numerical Technique for Calculating Incompressible Fluid Flows,” Los Alamos Scientific Laboratory, Report LA-4370 (1970)Google Scholar
  33. 33.
    J. H. Lee, Y. Oka and S. Koshizuka, “Safety System Consideration of a Supercritical-water Cooled Fast Breeder Reactor with Simplified PSA,” Reliability Engineering & System Safety, Vol. 64, 327–338 (1999)CrossRefGoogle Scholar
  34. 34.
    Reactor Safety Study, An Assessment of Accident Risks in US Commercial Nuclear Power Plants, WASH-1400, Appendix-I, US Nuclear Regulatory Commission (1974)Google Scholar
  35. 35.
    Reactor Safety Study, An Assessment of Accident Risks in US Commercial Nuclear Power Plants, WASH-1400, Appendix-II, US Nuclear Regulatory Commission (1974)Google Scholar
  36. 36.
    Analysis of Core Damage Frequency from Internal Events: Peach Bottom Unit 2. Science Applications International Corp, NUREG/CR-4550, Vol. 4, US Nuclear Regulatory Commission (1986)Google Scholar
  37. 37.
    Severe Accident Risks: An Assessment for Five US Nuclear Power Plants, NUREG-1150, Vol. 1, US Nuclear Regulatory Commission (1990)Google Scholar
  38. 38.
    Level-1 PSA of 1,100 MWe-class PWRs: Annual Report of FY1997, Nuclear Power Engineering Corporation (NUPEC), INS/M97-04 (1998) (in Japanese)Google Scholar
  39. 39.
    Development of Level-1 PSA Methods for BWR Plants: Annual Report of FY1999, Nuclear Power Engineering Corporation (NUPEC), INS/M99-15 (2001) (in Japanese)Google Scholar
  40. 40.
    Development of Level-1 PSA Methods for PWR Plants at Power: Annual Report of FY2000, Nuclear Power Engineering Corporation (NUPEC), INS/M00-04 (2001) (in Japanese)Google Scholar
  41. 41.
    Development of Level-1 PSA Methods for BWR Plants: Annual Report of FY2000, Nuclear Power Engineering Corporation (NUPEC), INS/M00-09 (2001) (in Japanese)Google Scholar

Copyright information

© Springer Science+Business Media, LLC 2010

Authors and Affiliations

  • Yoshiaki Oka
    • 1
  • Seiichi Koshizuka
    • 2
  • Yuki Ishiwatari
    • 3
  • Akifumi Yamaji
    • 4
  1. 1.Department of Nuclear Energy Graduate School of Advanced Science and EngineeringWaseda UniversityShinjuku-kuJapan
  2. 2.Department of Systems Innovation Graduate School of EngineeringUniversity of TokyoBunkyo-kuJapan
  3. 3.Department of Nuclear Engineering and Management Graduate School of EngineeringUniversity of TokyoBunkyo-kuJapan
  4. 4.Department of Nuclear Engineering and ManagementUniversity of TokyoBunkyo-kuJapan

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