Evolution and Assessment of Core Disruptive Accident Scenarios in SFR: From PFBR to CFBR
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1 Evolution and Assessment of Core Disruptive Accident Scenarios in SFR: From PFBR to CFBR K. Natesan, T.Sathiyasheela, K. Velusamy & P. Chellapandi Reactor Design Group Indira Gandhi Centre for Atomic Research Kalpakkam, INDIA Int. Workshop on Prevention & Mitigation of Severe Accidents in SFRs, June 11-13, 2012, Tsuruga, Japan
2 Possible severe accident initiators in PFBR Loss of flow followed by failure of both the shutdown systems (ULOF) - Battery power to Primary sodium pumps - 2 independent & diverse shutdown systems Uncontrolled withdrawal of one Control rod followed by failure of both the shutdown systems (UTOP) ULOF and UTOP events are analyzed for PFBR
3 Various Phases of Accident Progression in PFBR Pre-disassembly Phase In this phase, the phenomena occurring are sodium boiling, fuelclad melting and fuel slumping. This phase ends when the peak fuel temperature in the highest rated subassembly reaches boiling point. This phase lasts from a few tens of seconds to a few minutes. Transition Phase At the end of the pre disassembly phase, the reactor can become sub critical if there are sufficient negative reactivity feedbacks. If the negative feedbacks are insufficient, melting of fuel-clad will continue to form a molten pool. This is known as transition phase, since the fuel attains gradual transition from solid to liquid phase. Disassembly Phase In this phase, the core has lost its integrity & whole core is taken as fluid and fuel starts dispersing. Fuel displacement feedback dominates & time scales are short (~milliseconds) and all other reactivity feedbacks except the Doppler are insignificant. This phase lasts till the reactor attains sub criticality due to fuel dispersal.
4 Tools used for analysis of pre-disassembly phase The initial analysis of CDA was done using Bethe & Tait model, which assumes gravity driven collapse of the core and hydrodynamic core disassembly. PREDIS In house developed 2D axi-symmetric point kinetics code for reactor core Models for coolant boiling and fuel melting Feedback reactivity models for Doppler, core expansion, coolant expansion, control rod drive expansion, coolant voiding, fuel melting and slumping Validation against International benchmark data. DYANA-P In house developed 1D system dynamics code for whole plant Hydraulic models of primary and secondary sodium systems and thermal models for core, IHX, pipe lines, pools and steam generator Validated against FBTR experiments and PHENIX end of life test Iterations performed between PREDIS and DYANA-P to obtain consistent reactor inlet temperature and thermal balance.
5 Tools used for analysis of disassembly phase At the end of the pre disassembly phase, the reactor can become sub critical if there are sufficient negative reactivity feedbacks. If the negative feedbacks are insufficient, melting of fuel-clad will continue to form a molten pool. This transition phase is highly complex and is accounted by conservative assumption. VENUS II ANL developed 2D coupled neutronics and hydrodynamics code Point kinetics used for power calculations Calculates core dynamics during prompt critical disassembly excursion Feedback reactivity models for material displacement and Doppler effects Validated against Kiwi-TNT, SNAPTRAN-2 and SNAPTRAN-3 reactor disassembly experiments Calculations performed for the pre-disassembly phase provides input for the disassembly phase calculation Mechanical energy release of the event is predicted by MERC subroutine attached to the Venus code Mechanical consequences of the event are predicted subsequently.
6 Parametric Studies on Various Fuels Fuels considered: Mixed Oxide, Mixed Carbide & Metallic Fuel Event Scenario considered: ULOFA with fast flow coating (2 s flow halving time) Results of the pre-disassembly phase Parameter Oxide Carbide Metal Thermal power (W) 6.9x x x10 12 Reactivity ($) Coolant void fraction Molten fuel fraction Coolant void react($/s) <1 <1 <1 Fuel slumping reactivity rate ($/s) Reactivity addition rate due to FCI ($/s) Phase duration (s) Ref: Om Pal Singh, R. Harish, Energetics of core disruptive accident for different fuels for a medium sized fast reactor, Annals of Nuclear Energy 29 (2002)
7 Results of the disassembly phase calculations Parameter Oxide Carbide Metal Reactivity ramp rate ($/s) Thermal energy release (MJ) Mechan. work-potential (MJ) Max. fuel temperature (K) Max. fuel vapor pres. (atm) Fuel vapor fraction Phase duration (ms) Ref: Om Pal Singh, R. Harish, Energetics of core disruptive accident for different fuels for a medium sized fast reactor, Annals of Nuclear Energy 29 (2002)
8 Major Findings of the Parametric Study The energy release is seen to stabilize at 1000 MJ beyond a reactivity insertion rate of 100 $/s. Revised estimate of reactivity insertion rate based on realistic scenarios based on TREAT & CABRI tests indicated a value of < 50 $/s and the associated energy release is 23 MJ, using the CV2M Crosss-section data A few fundamental experiments on molten fuel coolant interaction, carried out at IGCAR (to be presented in a companion paper) indicated that MFCI effects are insignificant. However, an energy potential of 100 MJ is pessimistically assumed for PFBR. Through backward calculations, it is estimated that the reactivity addition rate to achieve 100 MJ energy is 66 $/s. Ref.: SC.Chetal and P.Chellapandi, Severe Accident Assessment for PFBR: Designer s Perspective, Economic times.
9 Reference Analysis
10 Analysis of UTOPA Following cases of uncontrolled withdrawal of one control rod (CSR) have been considered BOL fresh core - conservative BOL fresh core - nominal BOEC core conservative Plant under normal operation Initiating event : Withdrawal of one CSR Several SCRAM parameters (, Power & Temp.) evolve Shutdown system fails Power in fuel pin increases due to reactivity ddition by CSR Fuel pin temperature increases Fuel pin melting starts from its centre Clad remains intact due to availability of cooling Due to the release of cover gas from the molten fuel, pressure builds up and pushes the molten fuel upwards and downwards (fuel squirting) Large negative reactivity inserted due to movement of fuel from most active region (core centre) to less active region (top and bottom) Reactor power reduces and new stable steady power condition achieved Does not result in CDA
11 Analysis of UTOPA - continued Modeling of fuel-squirting In-pin fuel motion of melted fuel is considered in the analysis after CSR withdrawal is complete (250 s) Only 7 % of melted fuel is shifted to axial blanket regions But % movement of melted fuel was observed in experiments Conclusion of HUT 52 A Experiments Full length heated pins and annular pellets enhance in-pin fuel motion Conclusion of TREAT Experiments Full length pins with 22 % PuO 2 tested in TREAT reactor TOPA simulated with an external reactivity addition rate of 5 c/s Pre failure movement of 25 % of melted fuel was observed even with solid fuel pellets CABRI experiments also confirm central hole formation during a simulated TOPA
12 UTOPA Fresh core Nominal analysis Length of CSR withdrawal = 40 cm Time for CSR withdrawal = 200 s Net reactivity added = 317 pcm Non-uniform power profile & low pellet-clad gap conductance The power reaches 145%. No fuel melting Feedback reactivity (Doppler, Core radial & Fuel axial expansions, Control rod drive expansion) arrests the transient Reactor power stabilises at ~ 115 % Reactor p ower, % nominal Time, s Predicted power evolution
13 UTOPA Fresh core Nominal analysis Tem perature, deg C Cold pool Hot pool Main vessel Tem perature, K Max. Hot pool temperature : C Max. cold pool temperature : C Max. main vessel temperature : C Time, s Predicted pool temperature evolution
14 UTOPA Fresh core Conservative analysis 200 Length of CSR withdrawal = 50 cm Time for CSR withdrawal = 250 s Net reactivity added = 479 pcm Fuel melting starts at 128 s & at 240s the melt fraction reached is 28 % Fuel squirting considered Power reduces and stabilises at ~ 110 % Reactor p ower, % Time, s Predicted power evolution
15 UTOPA Fresh core Conservative analysis Tem perature, deg C Cold pool Hot pool Main vessel Tem perature, K Max. Hot pool temperature : C Max. cold pool temperature : C Max. main vessel temperature : C Time, s Predicted pool temperature evolution
16 UTOPA BOEC core Conservative analysis Length of CSR withdrawal = 50 cm Time for CSR withdrawal = 250 s More uniform power profile & high pelletclad gap conductance Fuel melting starts at 133 s & at 250 s, the melt fraction reached is 28 % Fuel squirting occurs Power reduces and stabilises at ~ 45 % Power, % nominal Time, s Predicted power evolution
17 UTOP BOEC core Conservative analysis Temp erature, deg C Cold pool Hot pool Main vessel Temp erature, K Max. Hot pool temperature : C Max. cold pool temperature : C Max. main vessel temperature : C Time, s Predicted Pool temperature evolution
18 Summary of UTOPA Studies Case description Maximum temperatures of various parts of the primary circuit reached, 0 C Hot pool Cold pool Main vessel Minimum cover gas volume reached, % nominal Fresh core Nominal Fresh core Conservative BOEC core - Conservative
19 Analysis of ULOFA Initiating event This transient is initiated due to the loss of power supply in the plant at full power operating conditions Alternate power sources to drive pony motors are also not available Primary and secondary sodium pumps coast down naturally governed by the inertia of their respective drive system Steam water system is also considered to be lost due to power failure Shutdown systems fail to trip the plant in-spite of the availability of several SCRAM parameters
20 Scenario of ULOFA Pre-disassembly phase Reactor under normal operating conditions Initiating event Failure of all power sources (i) (ii) (iii) Primary sodium pumps coast down by inertia (pony motors do not take over) Secondary sodium pumps coast down by inertia Steam water system becomes unavailable Several SCRAM parameters (speed, P/Q & Temp.) evolve Shutdown systems fail to act Coolant dryout and clad & fuel melting Molten fuel slumps down Large positive reactivity insertion Fuel temperature reaches boiling point Core integrity lost Disassembly phase Core flow reduces Coolant boiling starts from top (- ve reactivity feedback) Coolant boiling propagates to middle of the core (+ ve reactivity feedback)
21 Scenario of ULOFA Disassembly phase Coolant flow completely stopped Power generated is deposited within core Reactivity insertion due to fuel slumping Super prompt power excursion initiated with associated Doppler feedback Core bubble gets pressurised due to heat generation Governed by equation of state of fuel Core bubble expands due to pressure gradient -ve reactivity feedback evolves due to Doppler and fuel displacement Keff becomes less than 1 End of disassembly phase Core bubble expands isentropically and exerts mechanical loading
22 ULOFA : Fuel Slumping Model Conservative slumping model ABBN-93 Cross Section Data are used instead of CV2M Cross Section Data
23 Analysis of ULOFA 200 Fuel melts at 77s and slumps after 78 s leading to core disruptive accident (CDA) Reactor power, % nominal Dominated by +ve reactivity due to core voiding Dominated by ve reactivity due to core expansion Time, s Predicted power evolution
24 Analysis of ULOFA Flow, % nominal Primary sodium Secondary sodium Feedwater Time, s Predicted flow evolution in the plant
25 Analysis of ULOFA Temp erature, deg C Cold pool Hot pool Main vessel Temp erature, K Max. Hot pool temperature : C Max. main vessel temperature : C Time, s Predicted pool temperature evolution
26 End of pre-disassembly and disassembly phases End of pre-disassembly phase End of disassembly phase Thermal energy released : MJ Mechanical energy released : 0.02 MJ Peak temperature at the end of disassembly : 3734 K Peak pressure : 0.14 MPa Phase duration : 42.2 ms
27 Parametric study on energy release Vs reactivity addition 100 MJ mechanical energy release has been considered for the evaluation of mechanical consequences Reactivity addition rate for this is 65.7 $/s Reactor containment is designed for 100 MJ of mechanical energy release
28 Future Directions Future R&D in the domain of Severe Accident will be directed towards development of (i) Transition phase modeling and (ii)an integrated Computer code for severe accident analysis incorporating models for reactor physics, thermal hydraulics and structural dynamics, on the lines of SIMMER. The R & D studies being performed at IGCAR towards these goals will be presented by Dr.P.Chellapandi, Director, Reactor Design Group, in a companion paper.
29 Thank You for your kind attention!
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