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EUROTRANS – DM1 Analysis of Protected Accidental Transients in EFIT with RELAP5 Code

EUROTRANS – DM1 Analysis of Protected Accidental Transients in EFIT with RELAP5 Code. G. Bandini, P. Meloni, M. Polidori. Update of RELAP5 Model. Primary system layout by D1.26 of ANSALDO (November 2007) Primary circuit pressure drops according to new ANSALDO data

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EUROTRANS – DM1 Analysis of Protected Accidental Transients in EFIT with RELAP5 Code

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  1. EUROTRANS – DM1 Analysis of Protected Accidental Transients in EFIT with RELAP5 Code G. Bandini, P. Meloni, M. Polidori

  2. Update of RELAP5 Model • Primary system layout by D1.26 of ANSALDO (November 2007) • Primary circuit pressure drops according to new ANSALDO data • Gagging at core inlet according to SIM-ADS • DHR modeling data according to detailed ANSALDO analysis RELAP5 Nodalization Scheme

  3. Main EFIT Parameters • PRIMARY SYSTEM: • Total power = 395.2 MW • Lead mass flowrate = 33243 kg/s • Lead temperature = 400 / 480 C • Total primary circuit pressure drop = 1.37 bar (core = 0.7 bar, SG = 0.4 bar, Pump = 0.27 bar ) • Total mass of lead = 5880 tons (ANSALDO data = 5954 tons) • Lead free levels = 1.085 / 1.495 / 0.448 (ANSALDO data = 1.085 / 1.473 / 0.406) • SECONDARY SYSTEM: • Feedwater flow rate (4 SGs) = 244.4 kg/s, Temperature = 335 C • Steam pressure = 140 bar • Steam temperature = 452 C (Superheating of 115 C)

  4. RELAP5 Steady-State Calculation

  5. Transients to be Analyzed by ENEA (1)

  6. Analysis of Protected Transients • PLOF: Total loss of forced circulation (4 pumps) • PLOF-1: Stop of 1 pump (pump rotor seizure) • PLOH: Total Loss of Heat Sink • PLOH-1: Loss of feedwater to 1 steam Generator • PLOF + PLOH (Station blackout): Total loss of forced circulation and secondary loops with beam trip • PTOP: Overpower Transient (+10%) REACTOR TRIP (Proton beam switch-off): • Core outlet temperature > 515 - 525 C (> DTass-max x 1.2 + 400 C) • Pump speed close to 0 (in case of total loss of forced circulation) • Delay in reactor trip = 2 s

  7. Safety Limits • According to PDS-XADS safety analysis: • Lead temperature always below 1500 C • Clad temperature below 550 C during normal operation • Clad temperature in the range: • 550 – 600 C for less than 600 s, • 600 – 650 C for less than 180 s, in transient conditions • Vessel wall temperature below 450 C

  8. Sensitivity Study on Reactor Trip Set-Points (1)

  9. Sensitivity Study on Reactor Trip Set-Points (2)

  10. PLOF + PLOH – Station Blackout (1) Pump Vessel Core • Lead free level fluctuations and stabilization after primary pump stop • Core mass flow rate reduces down to 0 in the initial transient

  11. PLOF + PLOH – Station Blackout (2) • Very fast pump coastdown due to low initial pump velocity and low moment of inertia • Pump velocity and mass flow rate reverse in the initial transient following lead free level stabilization

  12. Inner core Inner core Middle core Middle core Outer core Outer core PLOF + PLOH – Station Blackout (3) • Maximum lead and clad peak temperature in hot channel of outer core zone • Maximum clad temperature is well below safety limits for transient conditions

  13. PLOF + PLOH – Station Blackout (4) DHR Core Core DHR • Natural circulation in the primary circuit and through the DHR stabilizes in about 500 s • Core decay power is totally removed by the DHR after about 1000 s

  14. PLOF + PLOH – Station Blackout (5) Core out (average) DHR in Core in DHR out • Primary hot lead temperature at core outlet stabilizes at about 453 C after 10000 s • Vessel wall temperature reaches a maximum of 435 C after 5000 s

  15. PLOF – Total Loss of Forced Circulation (1) • Pump speed = 0 after 0.8 s  Reactor trip with 2 s delay at 2.8 s • Initial core mass flow rate undershoot and fluctuations after primary pump stop

  16. Inner core Inner core Middle core Middle core Outer core Outer core PLOF – Total Loss of Forced Circulation (2) • Enhanced lead and clad temperature peaks after reactor trip at 2.8 s • Maximum clad temperature (hot channel of outer core) is below safety limits for transient conditions (less then 180 s in the range 600 – 650 C)

  17. PLOF-1 – Stop of 1 Primary Pump (1) Running pumps Running pumps Vessel Stopped pump Core Stopped pump • No reverse pump velocity is allowed after pump stop • Lead free level above stopped pump lies between vessel and core free levels after pump stop (reverse flow from vessel to core zone)

  18. PLOF-1 – Stop of 1 Primary Pump (2) Running pumps Stopped pump • Slight increase in running pump mass flow rate and reverse flow through stopped pump • Core mass flow rate reduces by about 30%

  19. PLOF-1 – Stop of 1 Primary Pump (3) Trip set-point Core SGs Reactor trip on Hot–average T-channel • Reactor trip at 10 s (on hot channels T > trip set-point) or 86 s (on average channels T > trip set-point) • SG removal power decreases by about 12% after pump stop due to primary circuit mass flow rate reduction

  20. Inner core Inner core Middle core Middle core Outer core Outer core PLOF-1 – Stop of 1 Primary Pump (4) • Clad temperature is kept below safety limits for transient conditions (less than 600 s in the range 550 – 600 C) • Maximum fuel temperature is below acceptable value

  21. Inner core Middle core Outer core PLOH – Total Loss of Heat Sink (1) Trip set-point Reactor trip on Hot–average T-channel • Reactor trip at 53 s (on hot channels T > trip set-point) or 76 s (on average channels T > trip set-point) • Clad temperature is kept well below safety limits for transient conditions

  22. PLOH – Total Loss of Heat Sink (2) DHR Core • DHR mass flow rate stabilizes at approximately 3500 kg/s after about 3 hours • DHR system reaches maximum performances (about 20 MW for 3 units) after about 400 s

  23. PLOH – Total Loss of Heat Sink (3) Core in Core out DHR in DHR out • Primary lead temperature stabilizes at about 435 C after 10000 s • Maximum vessel wall temperature around 3000 s is below safety limit (450 C)

  24. PLOH-1 – Loss of Feedwater to 1 SG (1) Trip set-point Core SGs Reactor trip on Hot–average T-channel • Reactor trip at 146 s (on hot channels T > trip set-point) or 340 s (on average channels T > trip set-point) • SGs removal power reduces down to 75% at transient initiation

  25. Inner core Inner core Middle core Middle core Outer core Outer core PLOH-1 – Loss of Feedwater to 1 SG (2) • Clad temperature is kept below safety limits for transient conditions (less than 600 s in the range 550 – 600 C) • Maximum fuel temperature is below acceptable value

  26. PTOP – Overpower Transient (+10%) Core and SGs Power Max Clad Temperature Max Lead Temperature Max Fuel Temperature

  27. CONCLUSIONS • Core decay heat removal by natural circulation through the DHR system is efficient in all investigated scenarios with total loss of heat removal by the secondary loops • Core and vessel structures temperatures are kept well below safety limits in all investigated protected scenarios even under the most conservative assumptions on reactor trip occurrence based on core outlet temperature measurements and trip set-point • Reactor trip on low pump rotational speed signal is needed in case of total loss of forced circulation in the primary circuit to keep clad temperature peak within acceptable values

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