Nuclear Engineering and Technology

Korean Nuclear Society (한국원자력학회)
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UNCERTAINTY AND SENSITIVITY ANALYSIS OF TMI-2 ACCIDENT SCENARIO USING SIMULATION BASED TECHNIQUES

  • Rao, R. Srinivasa (Nuclear Safety Analysis Division, Atomic Energy Regulatory Board, Anushaktinagar) ;
  • Kumar, Abhay (Light Water Reactor Division, Bhabha Atomic Research Centre) ;
  • Gupta, S.K. (Nuclear Safety Analysis Division, Atomic Energy Regulatory Board, Anushaktinagar) ;
  • Lele, H.G. (Reactor Safety Division, Bhabha Atomic Research Centre)
  • Received : 2011.07.12
  • Accepted : 2011.12.20
  • Published : 2012.10.25
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Abstract

The Three Mile Island Unit 2 (TMI-2) accident has been studied extensively, as part of both post-accident technical assessment and follow-up computer code calculations. The models used in computer codes for severe accidents have improved significantly over the years due to better understanding. It was decided to reanalyze the severe accident scenario using current state of the art codes and methodologies. This reanalysis was adopted as a part of the joint standard problem exercise for the Atomic Energy Regulatory Board (AERB) - United States Regulatory Commission (USNRC) bilateral safety meet. The accident scenario was divided into four phases for analysis viz., Phase 1 covers from the accident initiation to the shutdown of the last Reactor Coolant Pumps (RCPs) (0 to 100 min), Phase 2 covers initial fuel heat up and core degradation (100 to 174 min), Phase 3 is the period of recovery of the core water level by operating the reactor coolant pump, and the core reheat that followed (174 to 200 min) and Phase 4 covers refilling of the core by high pressure injection (200 to 300 min). The base case analysis was carried out for all four phases. The majority of the predicted parameters are in good agreement with the observed data. However, some parameters have significant deviations compared to the observed data. These discrepancies have arisen from uncertainties in boundary conditions, such as makeup flow, flow during the RCP 2B transient (Phase 3), models used in the code, the adopted nodalisation schemes, etc. In view of this, uncertainty and sensitivity analyses are carried out using simulation based techniques. The paper deals with uncertainty and sensitivity analyses carried out for the first three phases of the accident scenario.

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References

  1. J. C. Helton et al., "Survey of Sampling-based Methods for Uncertainty and Sensitivity Analysis", Reliability Engineering and System Safety, Volume 91, 1175-1209 (2006). https://doi.org/10.1016/j.ress.2005.11.017
  2. IAEA safety report series, 2008, "Best Estimate Safety Analysis for Nuclear Power Plants: Uncertainty Evaluation", IAEA, Vienna.
  3. M. Khatib-Rahbar and M. J. Zavisca, "An Overview of the Three Mile Island Unit 2 Accident and Plant Design", Energy Research Inc., Maryland (November 2006).
  4. M. L. Corradini et al., "SCDAP/RELAP5 Independent Peer Review", Los Alamos National Laboratory, Las Alamos, LA-12481, UC-000 (January 1993).
  5. M. Khatib-Rahbar, E. Cazzoli, M. Lee, H. Nourbakhsh, R. Davis and E. Schmidt, "A Probabilistic Approach to Quantifying Uncertainties in the Progression of Severe Accidents", Nuclear Science & Engineering, Volume 102, 219-259 (1989). https://doi.org/10.13182/NSE89-A27476
  6. SCDAP/RELAP5/MOD3.2 CODE MANUAL, VOLUME V, The SCDAP/RELAP5 Development Team, October 1997, Idaho National Engineering and Environmental Laboratory Lockheed Martin Idaho Technologies Idaho Falls, Idaho 83415
  7. M. Khatib-Rahbar et. al, "Analysis of Uncertainties Associated with the MELCOR Prediction of the TMI-2 Accident", Presentation made in Meeting of NRC/AERB, Mumbai, India (February, 2008).
  8. SCDAP/RELAP5/MOD3.2 CODE MANUAL, VOLUME II, The SCDAP/RELAP5 Development Team, October 1997, Idaho National Engineering and Environmental Laboratory Lockheed Martin Idaho Technologies Idaho Falls, Idaho 83415.
  9. Sang-Ryeol Park et al., "Development of an Uncertainty Quantification Method of the Best Estimate Large LOCA Analysis", Nuclear Engineering and Design, Volume 135, 367-378 (1992) https://doi.org/10.1016/0029-5493(92)90203-8
  10. D. M. Hamby, "A Review of Sensitivity Analysis Techniques (U)", WSRC-MS-93-576, Westinghouse Savannah River Company, (March, 1994).