Safety Analysis Results of Representative DEC Accidental Transients for the ALFRED Reactor

Transcription

1 FR13 - TECHNICAL SESSION 3.5: Fast reactor safety: post-fukushima lessons and goals for next-generation reactors Paper n. IAEA-CN-199/260 Safety Analysis Results of Representative DEC Accidental Transients for the ALFRED Reactor G. Bandini(ENEA), E. Bubelis, M. Schikorr (KIT), M.H. Stempnievicz (NRG), K. Tucek, A. Lázaro, (JRC-IET), P. Kudinov, K. Kööp, M. Jeltsov (KTH), L. Mansani(Ansaldo Nucleare) International Conference on Fast Reactors and Related Fuel Cycles: Safe Technologies and Sustainable Scenarios (FR13) Paris, 4-7 March 2013

2 Outline Introduction The ALFRED reactor DEC transients for ALFRED DEC transient results Conclusions 2

3 Introduction The conceptual design of the lead-cooled demonstrator reactor ALFRED was developed in the LEADER EU FP7 project to meet the safety objectives of the GEN-IV nuclear energy systems One of the main objectives of the project was the evaluation of the safety aspects and to perform a preliminary safety analysis of ALFRED Both Design Basis Conditions (DBC) and Design Extension Conditions (DEC) have been considered in the safety analysis of ALFRED The DEC accident scenarios are very low probability events which include the failure of prevention or mitigating systems ALFRED The main objective of DEC transient analysis is to evaluate the impact of the core and plant design features on the intrinsic safety behaviour of the plant More representative DEC events for ALFRED have been analysed by several research organizations using different system codes 3

4 ALFRED: Reactor block Horizontal section Vertical section Lead-cooled pool-type reactor of 300 MWth power 171 fuel assemblies in the core 8 pump -bayonet tube SG (MHX) connected to the 8 secondary circuits 4

5 ALFRED: Secondary circuits To DHR system Steam lines In-water pool isolation condenser (IC) Valve Feedwater Steam MHX Feedwater lines From DHR system DHR-1 system (4 IC loops) DHR-2 system (other 4 IC loops) 5

6 Steady-state at nominal power (EOC) Parameter Unit ALFRED RELAP5 SIM-LFR Reactor thermal power MW Total primary flow rate kg/s Total P in the primary circuit bar P through the core bar < Core inlet temperature C Upper plenum temperature C Max core outlet temperature (*) C Peak clad temperature C ~ Peak fuel temperature C ~ Feedwater temperature C Feedwater flow rate kg/s Steam temperature C Steam pressure bar (*) Hottest FA flow rate is ~120% of average FA flow rate 6

7 Analysis of DEC transients Organizations and Codes: ENEA (RELAP5, CATHARE), KIT (SIM-LFR), JRC-IET (TRACE), KTH (RELAP5), NRG (SPECTRA) TRANSIENT Initiating Event Reactor scram Primary pump trip MHX FW trip MSIV closure DHR startup UNPROTECTED PROTECTED UTOP ULOF Insertion of 250 pcm in 10 s All primary pumps coastdown No No No No No No 0 s No No No ULOHS All MHX feedwater trip No No 0 s 1 s ULOHS + ULOF Partial flow area blockage in the hottest FA Reduction of FW temperature + all primary pumps stop Increase of FW flow rate + all primary pumps stop SCS failure All primary pumps and MHXs feedwater trip 10% to 97.5% blockage at the hottest FA inlet T-fw: C in 1s + all p. pumps stop FW-flow +20% in 25 s + all p. pumps stop Depressurization of all secondary circuits No 0 s 0 s 1 s DHR-1 at 2 s (3 IC loops) DHR-1 at 2 s (3 IC loops) No No No No No 2 s, low pump speed 2 s, low pump speed 2 s, low sec. pressure 0 s 2 s 2 s 0 s 2 s 2 s DHR-1 at 3 s (4 IC loops) DHR-1 at 3 s (4 IC loops) No 2 s No No 7

8 DEC: Unprotected transients Objective:Verify the intrinsic safety behavior of the ALFRED plant and its response to more unlikely accidental events Analyzed transients without reactor scram: UTOP:Reactivity insertion of 250 pcm in 10 s (core compaction, core voiding following SGTR, etc.) ULOF:Loss of all primary pumps ULOHS: Loss of feedwater to all MHXs ULOHS + ULOF:Loss of feedwater to all MHXs + loss of all primary pumps Partial FA flow area blockageverify the maximum acceptable flow are blockage without fuel rod damage 8

9 Reactivity feedbacks at EOC REACTIVITY COEFFICIENT Unit Ref. Temperature Value Control rod differential expansion (*) pcm/k T upper plenum Coolant expansion (**) pcm/k Average T-core Axial clad expansion pcm/k Average T-clad Axial wrapper tube expansion pcm/k Average T-wrapper Radial clad expansion pcm/k Average T-clad Radial wrapper tube expansion pcm/k Average T-wrapper Diagrid radial core expansion pcm/k T-core inlet Pad radial core expansion pcm/k T-core outlet Axial fuel expansion: free pcm/k Average T-fuel Axial fuel expansion: linked pcm/k Average T-clad Doppler constant pcm Average T-fuel -566 (*) Prompt response (the delayed response has been neglected) (**) Calculated on the whole height of the fuel assembly (the other feedbacks are calculated only in the fissile zone) 9

10 UTOP transient (1/2) Insertion of 250 pcm in 10 s without reactor scram SCS in service No feedwater control on the secondary side Total reactivity and feedbacks Core and MHX powers Inserted Total Core power Pad + Diag. Fuel exp. (free) MHX power Doppler TRACE Results Maximum net reactivity insertion of 80 pcm Initial core power peak of 640 MW 10

11 UTOP transient (2/2) Primary coolant temperatures Max fuel and clad temperatures Core outlet Max fuel MHX inlet MHX outlet Core inlet Max clad TRACE Results Maximum clad temperature remains below 650 C Maximum fuel temperature of ~2820 C at t = 20 s (hottest FA, middle core plane, fuel pellet center)exceeds the MOX melting point (~2700 C) only local fuel melting 11

12 ULOF transient (1/4) All primary pumps coastdown without reactor scram SCS in service No feedwater control on the secondary side Active core flowrate Core and MHX powers MHX power Core power Core flow rate RELAP5 Results Natural circulation in the primary circuit stabilizes at 23% of nominal value Core power reduces down to about 200 MW due to negative reactivity feedbacks 12

13 ULOF transient (2/4) Core temperatures Total reactivity and feedbacks Core inlet Max clad Max lead Doppler Fuel exp. Cool. exp. C. Rods Pad + Diag. Core temperatures Max fuel Max clad Core inlet RELAP5 Results: Max coolant temperature of 710 C is well below the lead boiling point of 1740 C Initial clad peak temperature of 764 C, then max clad temp. stabilizes below 650 C Doppler and fuel exp. effects are counterbalanced by radial core exp. (Pad + Diag.), control rods and coolant exp. effects 13

14 ULOHS transient (1/2) Loss of feedwater to all MHXs without reactor scram SCS isolation and startup of DHR-1 system (3 out of 4 IC loops are in service) Core power and flow rate Core and vessel temperatures rel. units [fr] Core flow rate Core power Power_th Flow_Cool Temperature [ C] Fuelc_peak Cool_out T_wall Max fuel Max clad Clad_peak Cool_in Max vessel Time [sec] 200 SIM-LFR Results Time [sec] Core power progressively reduces down towards decay level After about one hour the core power can be removed by the DHR-1 system Maximum vessel and clad temperatures rise up to 663 C and 700 C after one hour 14

15 ULOHS transient (2/2) Max clad and vessel temperatures Max clad Max vessel No clad failure is calculated by SIM-LFR code in the short and long term Vessel integrity is guaranteed in the medium term, but not in the long term Clad Failure Time [sec] 1E+14 1E+12 1E+10 1E+08 1E+06 1E+04 1E+02 1E min Clad Failure Time [sec] Fission Gas Pressure [bar] Time [sec] Fission Gas Pressure [bar] SIM-LFR: Minimum clad failure time > 1.0E+6 s 15

16 ULOHS + ULOF transient (1/2) rel. units [fr] Loss of feedwater to all MHXs and all primary pumps without reactor scram SCS isolation and startup of DHR-1 system (3 out of 4 IC loops are in service) Core flow rate and power Power Flow rate Power_th Flow_Cool Time [sec] Temperature [ C] SIM-LFR Results Core and vessel temperatures Max fuel Fuelc_peak Cool_out T_wall Max lead, clad Clad_peak Cool_in Core inlet, vessel Time [sec] Sharp decrease of core power and flow rate in the initial transient phase and then their progressive decrease Core flow rate/power ratio is ~1/3 of nominal value Large ΔT through the core Maximum clad temperature rises up to ~800 C in 30 minutes, but the maximum vessel wall temperature remains rather low (Tw < 550 C) 16

17 ULOHS + ULOF transient (2/2) Max clad and vessel temperatures Max clad Max vessel The minimum clad failure time predicted by SIM-LFR is of about 3 hours Vessel integrity is guaranteed in the medium and long term 1E Clad Failure Time [sec] 1E+12 1E+10 1E+08 1E+06 1E+04 1E+02 1E min Clad Failure Time [sec] Fission Gas Pressure [bar] Time [sec] Fission Gas Pressure [bar] SIM-LFR: Minimum clad failure time > 1.0E+4 s 17

18 Partial FA blockage (RELAP5 results) Assumptions: Total ΔP over the FA = 1.0 bar ΔP at FA inlet = 0.22 bar Flow area blockage at hot FA inlet No heat exchange with surrounding FAs FA flowrate versus flow area blockage Max lead, clad and fuel temperatures versus flow area blockage Max fuel Max lead Max clad Main results: 75% FA flow area blockage 50% FA flowrate reduction 85% blockage T-max clad = 700 C No clad melting if area blockage < 95% Fuel melting if area blockage > 97.5% 50% inlet flow area blockage can be detected by TCs at FA outlet 18

19 Conclusions (1/2) In all simulated transients there is a very large margin to coolant boiling: the coolant is always at least 900 C below the lead boiling point (1740 C) Clad failure is not predicted in all simulated transients except for: Unprotected FA flow area blockage greater than ~85% which might be excluded by design (many orifices at the FA inlet) The very unlikely ULOHS+ULOF event, when the time-to-failure reduces down to few hours, but still leaving enough grace time for corrective operator actions Fuel melting is excluded in all simulated transients except for local fuel melting in the hottest pins in case of UTOP transient The vessel integrity seems guaranteed in the long term in all simulated transients except for the ULOHS transient, but also in this case there is enough grace time for corrective operator actions No relevant safety issues have been identified for ALFRED in case of representative DEC events In particular the ULOF transient can be accommodated without the need of corrective operator actions 19

20 Conclusions (2/2) The analysis of DEC transients with various codes has highlighted the very good intrinsic safety features of ALFRED design thanks to: Good natural convection in the primary circuit to limit core temperature increases Large thermal inertia to slow down the transients Prevalent negative reactivity feedbacks to reduce nuclear power excursions In all analyzed unprotected transients there is no risk for significant core damage and then for transient evolution towards severe accidents enough grace time is left to the operator to take opportune corrective actions for a safe reactor shutdown 20