Nuclear Systems Multiphase Flow Phenomena: Post-Fukushima

Transcription

1 Nuclear Systems Multiphase Flow Phenomena: Post-Fukushima Michael Corradini American Nuclear Society VP-Pres. Elect: UW Energy Institute Chair: 1

2 Fukushima-1 Accident Summary March 11 th earthquake/tsunami struck Fukushima Daiichi, a six-unit BWR nuclear power plant on the NE coast of Japan Earthquake/tsunami caused station blackout, disabling safety systems Extensive fuel damage occurred in the cores at Units 1-3 Containment failure at Units 1-3 resulted in large release of radioactivity in atmosphere and ocean Large environmental impact and minimal health impacts Nuclear Industry and Regulators are learning lessons * Info: TEPCO, NISA, MEXT 2

3 Overview of Boiling Water Reactor Typical BWR/3 and BWR/4 Reactor Design Similarities to BWR/4 Plants in Midwestern US 3

4 Mark 1 Containment and Reactor Building BWR/3 (460 MWe, 1F1) Mark 1 containment (drywell + torus suppression pool) SFP on top floor of the R/B Isolation condenser for core cooling (hi-press) HPCI (high pressure core injection, hi-press) Core spray system (CS at low pressure) after depressurization by SRVs BWR/4 (784 MWe, 1F2, 3, and 4) Mark I containment (drywell + torus suppression pool) SFP on top floor of the R/B RCIC (reactor core isolation cooling) HPCI (high pressure core injection) CS and RHR/LPCI (at lo-pressure) after depressurization by SRVs

5 After Loss of Water Injection in RPV After makeup water injection into the RPV is lost, all three units follow the same fate: " Water in RPV progressively boils off, thus water level decreases rapidly " The resulting steam is directed to the Suppression Pool (SP), which heats up " Water level in RPV reaches Top of Active Fuel (TAF); fuel is exposed " Fuel overheating leads to hydrogen generation (Zr/H 2 O reaction) and eventually fuel starts to melt " SP reaches saturation; containment pressure begins to rise " Molten fuel relocates to bottom of the vessel with water present and eventually RPV failure into cavity with water present Reactor building SRVs Drywell SP Boil-off curve for station blackout at Grand Gulf BWR (without ADS Actuation) RPV Spent fuel pool Containment vessel

6 Earthquake at 14:46: LOSP Tsunami at 15:41: SBO IC operating level loss Accident Sequence Summary Timeline of Major Fukushima Damage Sequences Unit 1 SC Saturated core damage Sea water injection Containment vent H2 Explosion RCS Repressurizes RCS Depressurized Unit 3 RCIC operating HPCI operating Level loss SC Saturated RPV Depressurization Sea water injection? Core damage? Sea water injection Containment vents H2 Explosion Unit 2 RCIC - CST RCIC from suppression pool SC Saturated Level loss RPV Depressurization Sea water injection Fuel damage Containment vent Noise from torus room Unit 4 (SFP) Explosion in Unit 4 Information Friday 6 11within this illustration Saturday was developed 12 from the INPO Sunday Special Report and the Monday Report of 14 Japanese Government Tuesday to IAEA 15 Ministerial Conference Wednesday on Nuclear 16 Safety Accident at TEPCO s Fukushima Nuclear Power Stations Transmitted by Permanent Mission of Japan to IAEA, June 7, 2011

7 Predicted BWR Severe Accident Response Is Different from That Expected of a PWR in Several Aspects More zirconium metal Isolated reactor vessel and bundle geometry Reduction in power factor in the outer core region Potential effects of safety relief valve actuations Progressive relocation of core structures Importance of core plate and lower internals Large amount of water in the vessel lower plenum External structures and water in pedestal region

8 Fukushima Lessons-Learned Issues That Require More Physical Insight: Hydrogen transport and mixing in reactor containment compartments as well as H2 mixing/recombination Effect of raw water addition on in-vessel cooling, accident progression, source term release In-vessel retention in BWR with water present Ex-vessel coolability in containment reactor cavity Instrumentation to better understand TH conditions Innovations for long-term decay heat removal 8

9 Severe Accident Physical Processes Selected two-phase issues that require insight: In-vessel or ex-vessel steam explosions considered Dynamic pressures that can fail RPV or Cavity structures In-vessel or ex-vessel debris coolability considered Water is a benefit given low SE energetics (SERENA) In-vessel coolability is complex for BWR geometry Ex-vessel coolability has been the focus of international experimental efforts (OECD MACE, CCI, SWICS) 9

10 Steam Explosions (In-vessel or Ex-vessel)

11 Severe Accident Phenomena: Steam Explosions Current Status on Steam Explosions in terms of Energetics Experimental Database Corium CR (%) CCM (ANL:Corium) 0 SUW (Winfrith: UO 2 +Mo) >1 WUMT (Winfrith: UO 2 +Mo) 0 MIXA Winfrith: UO 2 +Mo) 0 KROTOS (JRC: E/UO 2 -ZrO 2 ) FARO (JRC: E/UO 2 -ZrO 2 ) <0.2 COTELS (NUPEC: BWR) 0 TROI (KAERI: E/UO 2 -ZrO 2 )* <0.08 TROI (KAERI: NE/UO 2 -ZrO 2 )* 0 Very low * Before SERENA-2 Reality Simulant CR (%) EXO-FITS (SNL:Fe+Al 2 O 3 ) >1 FITS (SNL:Fe+Al 2 O 3 ) 1.1~7.4 ALPHA (JAEA:Fe+Al 2 O 3 ) KROTOS (JRC: Alumina) Metal (Various: Tin) <0.85 TROI (KAERI: ZrO 2 )* <0.18 ZREX (ANL: Zr) <3.85 Safety Margin High Upper Bounds

12 TEXAS for Melt Mixing and Quench Eulerian formulation for the vapor and water phases - 1-D models (multi-fluid) LaGrangian formulation for fuel particles (coherent jet and discrete pcls) Coherent Jet Discrete Relative velocities between fuel and coolant induce jet/pcl breakup (RT- Rayliegh-Taylor, KH-Kelvin-Helmholtz) Dynamic breakup of fuel both as a coherent jet and discrete particles create additional area to drive fuel mixing (NHTC 1994 and CSNI 1997 and NSE-2012)

13 TEXAS fragmentation/heat release Fuel jet fragmentation during mixing: The correlation of our theoretical model (Chu et al, 1986) is: D n+1 = D n (1 - C o ΔT + We 0.25 ) where We - Weber number for particles; ΔT + - dimensionless timestep = C o = ε where ε = (ρ c /ρ f ) 0.5 Explosive fuel fragmentation: thermal mechanisms (film collapse, jet) m fr = 6 C fr m p [(P P th )/(ρ f D f2 )] 0.5 F (F determined by axial const.) The fuel fragmented during the explosion is assumed to be quenched to the bulk coolant temperature and all the heat goes into vaporization. This simplified model is based on detailed modeling of single drop tests (C.Chu, B.Kim, M.Oh, J.Tang, S.Nisulwankosit, S.Wang, R.Chen) 13 U rel (t n+1 t n ) n D f We = ρ cu rel ρ c ρ f σ f 1/ 2 2 D f n

14 KROTOS Alumina Tests (K-38, K-42) w TEXAS

15 KROTOS Corium Tests (K-52 and K-53) w TEXAS Fuel fragments only if vapor film collapse can break through the solid fuel crust

16 Ex-vessel Vapor Explosion (PWR Example) Pressure outlet Fuel jet Water TEXASV domain ANSYS domain

17 Ex-Vessel Vapor Explosion: PWR Example Table 3. Results from TEXAS-V analysis Case U o [m/s] Trigger[s] M tot [kg] P max [MPa] KE max [MJ] I-A I-B I-C II-A II-B Table 4. Pressure and Impulse Load at the cavity wall Case P max [Mpa] P cw [MPa] Δtime [ms] Impulse [kpa-s] I-A I-B I-C II-A II-B Melt jet of 10 cm radius (corium jet with supht.) Dynamic frag. mixing Triggered at base (~1sec) Compare at cavity wall Dynamic pressures ~100MPa Kinetic energies <10 MJ Wall impulse is key factor Imp ~ Kinetic Energy Impulse < Struct. Strength (10kPa-s < 20kPa-s)

18 Ex-Vessel Core Melt - MCCI Accident Sequence: Beyond Design Basis accident occurs Reactor core is not cooled and melts Release of radionuclides in containment Corium melts through pressure vessel Corium relocates to reactor concrete containment floor Concrete decomposes due to high temp. (releasing H 2 and non-condensables) Water addition? Continued erosion while melt cools Concerns: Melt coolability? Containment integrity? Pressurization failure Basemat failure 18

19 Ex-vessel Debris Quench/Coolability Attainment of stable coolable state for debris Melt quench by water forming coolable debris and arrest attack MACE Tests did not show coolability => oscillatory behavior Crust formation hampered water ingress from above

20 Ex-vessel Debris Quench/Coolability Bulk Cooling Water Ingression Melt Dispersal Crust Breach (Farmer 2000)

21 MCCI Experimental Studies Substantial domestic and international research: Scale effects (most tests <0.3 m 2, reactor cavity >30 m 2 ) Decay heat simulation, high temperatures (2300+ C) Melt constituents, prototypic with large UO 2 content Enhances Coolability: Melt Eruptions & Water Ingression Degrades Coolability: Crust Anchoring scale issue vs. vs. Lomperski, S., Farmer, M.T., Measurements of the Mechanical Strength of Corium Crusts, ICAPP 2008, Anahein, CA, USA, paper Farmer, M.T., Basu, S., A Summary of Findings from the Melt Coolability and Concrete Interaction (MCCI) Program, ICAPP 2007, Nice, France, pape

22 MCCI Model Simulations Most Important Parameters: Collapsed melt height (cavity diameter & melt mass) Crust Anchoring Melt eruptions Water ingression δ t cr VENT:HOLE D h CONDUCTION:A:LIMITED BOUNDARY:LAYER:OF THICKNESS,: h CRUST:WITH:PERMEABLE CRACK:STRUCTURE CONDUCTION:A:LIMITED BOUNDARY:LAYER Modeling (CORQUENCH): ( ) g m bed + ρ t,c A b δ t,min + m wat applied load on crust C 2 geom σ t, f δ t,min crust mechanical strength CORIUM:MELT:POOL WITH:G AS :S PARG ING! δ t,min = ρ t,c A b g 2C geom σ t, f " $ ρ t,c A b g $ C # geom σ t, f % ' ' & 2 + 4g ( m bed + m l ) C geom σ t, f δ t δ t,min Crust Anchors Robb, K. R., Corradini, M. L., "Ex-Vessel Corium Coolability Sensitivity Study with CORQUENCH Code," NURETH-13, Sept. 2009

23 MCCI Model Simulations Most Important Parameters: Collapsed melt height (cavity diameter & melt mass) Crust Anchoring Melt eruptions Water ingression Modeling (CORQUENCH): " q c,dry = κ ρ v Δe lv (ρ l ρ v )g 2µ v " q c,dry # Δe = C lv (ρ l ρ v )g & dry % ( $ ' ν v 5/13 # % $ 2 Nk t,c ( Δe sat ) 2 c t,c Δe crack & ( ' 4/13 # ) # α c,exp T t, frz T sat + σ t, f + % % $ * + $ α t,ex E t,y &,& (. '-. ( ' 15/13

24 MCCI Model Simulations Most Important Parameters: Collapsed melt height (cavity diameter & melt mass) Crust Anchoring Melt eruptions Water ingression Modeling uncertainties: V = Ke Model Used Ablation [cm] liq V gas Quench Time [min] Ke = >>7200 Ke = >>7200 Ke = Ke = Robb, K. R., Corradini, M. L., "Ex-Vessel Corium Coolability Sensitivity Study with CORQUENCH Code," NURETH-13, Sept. 2009

25 Top Surface Cooling for Debris Coolability Melt Eruption Process Formation: Channel forms through crust Entrainment: Concrete decomposition gases travel through channels in crust Gas entrains melt thru channels and onto crust Result: Ejected melt quenches forming a part bed 10 1 Modified Model Comparison Water Particle Bed Crust Corium Melt NURETH Model Entrainment Khalil Parker Yoshinaga Todoroki Stenning Kouremenos White Jeelani Castro Drake Fujimoto Kamata 1 to 1 +30% -30% Experimental Entrainment Decomposition Gases Concrete

26 MCCI Simulations w Water (Water Ingression, Melt Eruptions, No Crust Anchoring) (CORQUENCH) Elevation (cm) 50 0 Crust Concrete hrs Time after melt relocation (Minutes) 5 days