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Increase In Heat Removal By Secondary System

Increase In Heat Removal By Secondary System. Case 1 : 99409-004 Sung Joo, Kim Case 3 : 2000-12046 Kyuhwan , Lee Case 2 : 2000-12043 Sujong, Yoon Case 4 : 2000-12052 Soon-Wook, Jeong Case 5 : 2000-12058 Byeong heon, Hwa n g. Introduction. Cool down events

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Increase In Heat Removal By Secondary System

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  1. Increase In Heat Removal By Secondary System Case 1 : 99409-004 Sung Joo, Kim Case 3 : 2000-12046 Kyuhwan, Lee Case 2 : 2000-12043 Sujong, Yoon Case 4 : 2000-12052 Soon-Wook, Jeong Case 5 : 2000-12058 Byeong heon, Hwang

  2. Introduction • Cool down events • 1 Additional feed water (condition 2) • 2 Increase feed water flow (condition 2) • 3 Increase secondary steam flow (condition 2) • 4 Inadvertent opening (condition 2) • 5 Piping failure (condition 3 can be 4)

  3. Common Situations • Some mistakes or faults • Increase in heat removal • Increase in Core power • Some Trip or transient situation • New status • End of situation

  4. Equations • Heat Transfer Formula • MTC (Moderator Temperature Coefficient)

  5. An Accidental Bypass Open Case 99409-004 Kim Sung Joo

  6. Causes and Mal function • A full opening of a bypass valve • A control system malfunction • An operator error. (human error) • Mal function • Open bypass valves(66,1)

  7. Results ; Core Coolant Temperature

  8. Results ; Net Reactivity

  9. Results ; Average Fuel Temperature

  10. Results ; DNBM

  11. Results ; Core Coolant Temperature, Net Reactivity, Fuel Temperature, DNBM

  12. An Excessive Increase in Secondary Steam Flow 2000- 12046 / Lee, Kyu-hwan

  13. Mal-function of the Accident • Applicable mal-function doesn’t exist in the mal-function list. (Trip doesn’t occur. / Moderate Frequency) → manually changed variables. • 10% step load increase within 15~100% of Full-Power doesn’t bring about the trip signals. (Reactor Control System is design to accommodate.)

  14. Accident Description 1. Excessive Load Increase → Increase in Secondary Steam Flow (also by operator or equipment-malfunction) 2. Power mismatch between the Reactor Core and the Turbine Load Demand 3. Increase in Steam Flow → Increase in Feed Water Flow 4. Increase in Heat Removal by Secondary System 5. Decrease in Coolant Temp. 6. Decrease in Moderator Temp. (Negative Coefficient) → Slight Power Increase 7. Increase in Moderator Temp. → Power Decrease New Equilibrium Condition ※ DNBR > 1.3 → Stable Operation!!!

  15. Turbine Load Increase Simulation • Compact Nuclear Simulator • Power : 70% → 100% • Set Increasing rate : 9.1518% per minute • Running Time : 132 counts (Real Time Simulation : 132 seconds) ▶Equilibrium Condition Time in FSAR : 100 seconds (in Automatic Control Mode)

  16. Accident Result Analysis - 1 27s 35s 35s 25s

  17. Accident Result Analysis - 2

  18. Comparison with FSAR - 1 Core Avg. Temperature 590 °F → 310 °C

  19. Comparison with FSAR - 2 Nuclear Power

  20. Comparison with FSAR - 3 DNBR

  21. An increase in FW flow 2000- 12043 Yoon, Sujong

  22. Causes and Mal function • A full opening of a feedwater control valve • A feedwater control system malfunction • An operator error. • Mal function • Open all FW valves(68,1)

  23. S/G high-high water level signal (15.3 m) (1) Rx trip (3) start S/G start Core FW PUMP Turbine Trip (2) Min. DNBM (1.29773) start FW Isolation valves closed(4) start Results of the analysis 20s 110s

  24. Conclusion • DNBR does not drop below the limit value • No fuel or clad damage is predicted. Comparison with FSAR

  25. Inadvertent Opening of a S/G Safety Valve 2000-12052 Jeong, Soon-Wook

  26. Event Description • Unintentional steam leakage through Safety Valve Depressurization in PRZ • Potential insecurity to decrease DNBR value and break down nuclear system. • Assumption • In case of 100% power output condition • One safety valve open by mal-operation • RCCA still operational during the simulation

  27. Reactor Trip Signal SIS Actuation Accident Analysis - 1 (25s): Depressurization in S/G & PRZ begins. (75s): Pressure decrease in PRZ causes reactor trip. (130s) Continuous pressure decrease in PRZ causes SIS. (140s) Boron acid begins to be injected in the core. 200s 400s

  28. Reactor Trip Signal SIS Actuation Accident Analysis - 2 (25s): Depressurization in S/G & PRZ begins. (75s): Reactor trip signal occurs and control rods withdrawn simultaneously. (140s) Boric acid in core has effect on decrease in net reactivity. 200s 400s

  29. Conclusion • Reactor trip prevent the reactor from further reaction in core. • SIS is to interrupt pressure increase in PRZ, ultimately to prevent DNBR lower than 1.3. • No further hazard after reactor trip and SIS actuated. • SAFE!!!

  30. Main Steam Line Break 2000-12058 Hwang, Byeong-Hyun

  31. Simulation Condition • Assumption : Control rod Inserted, but with the most reactive RCCA stuck out Single failure in ESF • Malfunction Condition : Inside containment, loop 1 Leak size 1000cm2

  32. 14 14 14 5 5 5 Results – Secondary System

  33. 10 7 7 Results – Primary System 1

  34. 14 14 10 10 Results – Primary System 2

  35. 14 14 16 9 Results – Secondary System 3

  36. 14 17 14 Results – Secondary System 4

  37. 15 9 14 Results – Primary System 1.27 Although the main steam line break happens, DNBdoes not occurs, thusthe reactorremains S A F E

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