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LEADER Lead-cooled European Advanced DEmonstration Reactor. Safety Analysis Results of the DEC Transients of ALFRED. G. Bandini ( ENEA), E. Bubelis, M. Schikorr (KIT), A. Lazaro , K. Tucek (JRC-IET) - PowerPoint PPT Presentation
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Safety Analysis Results of the DEC Transients of ALFRED
LEADERLead-cooled European Advanced DEmonstration Reactor
G. Bandini (ENEA), E. Bubelis, M. Schikorr (KIT), A. Lazaro, K. Tucek (JRC-IET)P. Kudinov, K. Kööp, M. Jeltsov (KTH), M. H. Stempnievicz (NRG), Z. Youpeng, K. Mikityuk (PSI)
Technical Workshop to Review Safety and Design Aspects of ALFRED, ELFR and ELECTRAJRC-IET, Petten, 27-28 February 2013
2
Outline
Introduction The ALFRED reactor DEC transients for ALFRED DEC transient results Conclusions
3
Introduction
One of the main objectives of the LEADER EU project was the evaluation of the safety aspects of the lead-cooled demonstrator reactor ALFRED
Both Design Basis Conditions (DBC) and Design Extensions 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
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
4
ALFRED: Reactor block
Vertical section
Horizontal section
Pool-type reactor of 300 MWth power 171 fuel assemblies in the core 8 pump-bayonet tube SG connected to
the 8 secondary circuits
5
ALFRED: Secondary circuits
DHR System (4 x 2 IC loops)
In-water pool isolation condenser (IC)
Valve
Water
Hot Lead
Cold Lead
Steam
Water
Hot Lead
Cold Lead
Steam
SG
Feedwater
Steam
From DHR system
To DHR system
Steam lines
Feedwater lines
6
Steady-state at nominal power (EOC)
Parameter Unit ALFRED RELAP5 CATHARE SIM-LFR
Reactor thermal power MW 300 300 300 300Total primary flow rate kg/s 25980 25250 25460 25682Total ΔP in the primary circuit bar 1.5 1.5 1.5 1.5ΔP through the core bar < 1.0 1.0 1.0 1.0Core inlet temperature °C 400 400 400 400Upper plenum temperature °C 400 480 480 480Max core outlet temperature (*) °C - 483 483 487Peak clad temperature °C ~550 508 518 514Peak fuel temperature °C ~2000 1991 1985 2064Feedwater temperature °C 335 335 335 335Feedwater flow rate kg/s 192.8 192.8 196.6 193.6Steam temperature °C 450 450 451 450Steam pressure bar 180 180 180 180
(*) Hottest FA flow rate is ~120% of average FA flow rate
7
Analysis of DEC transients
Organizations and codes: ENEA (RELAP5, CATHARE), KIT (SIM-LFR), JRC-IET (TRACE, SIMMER), KTH (RELAP5),
NRG (SPECTRA), PSI (TRACE/FRED)
TRANSIENT Initiating Event Reactor scram
Primary pump trip
MHX FW trip
MSIV closure DHR startup
TR-4: UTOP Insertion of 250 pcm in 10 s No No No No No
TDEC-1: ULOF All primary pumps coastdown No 0 s No No No
TDEC-3: ULOHS All MHX feedwater trip No No 0 s 1 s DHR-1 at 2 s (3 IC loops)
T-DEC4: ULOHS+ULOF All primary pumps and MHXs feedwater trip No 0 s 0 s 1 s DHR-1 at 2 s
(3 IC loops)
T-DEC5: Partial block. in the hottest FA
10% to 97.5% blockage at the hottest FA inlet No No No No No
TO-3: All prim. pumps stop + reduction of FW temperature
T-fw: 335330°C in 1s + all p. pumps stop
2 s, low pump speed 0 s 2 s 2 s DHR-1 at 3 s
(4 IC loops)
TO-6: All prim. pumps stop + increase of FW flow rate
FW-flow +20% in 25 s + all p. pumps stop
2 s, low pump speed 0 s 2 s 2 s DHR-1 at 3 s
(4 IC loops)
T-DEC6: SCS failure Depressurization of all secondary circuits
2 s, low sec. pressure No 2 s No No
UN
PRO
TECT
EDPR
OTE
CTED
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DEC: Unprotected transients
Objective: Verify the intrinsic safety behaviour of the ALFRED plant and its response to more unlikely accidental events
Analysed transients without reactor scram: UTOP: Reactivity insertion of 25 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 blockage verify the maximum acceptable
flow are blockage without fuel rod damage
9
Reactivity feedbacks at EOC
REACTIVITY COEFFICIENT Unit Ref. Temperature Value
Control rod differential expansion (*) pcm/K T upper plenum -0.218
Coolant expansion (**) pcm/K Average T-core -0.268
Axial clad expansion pcm/K Average T-clad 0.039
Axial wrapper tube expansion pcm/K Average T-wrapper 0.023
Radial clad expansion pcm/K Average T-clad 0.011
Radial wrapper tube expansion pcm/K Average T-wrapper 0.003
Diagrid radial core expansion pcm/K T-core inlet -0.152
Pad radial core expansion pcm/K T-core outlet -0.430
Axial fuel expansion: free pcm/K Average T-fuel -0.155
Axial fuel expansion: linked pcm/K Average T-clad -0.242
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)
UTOP transient (1/4)
Insertion of 250 pcm in 10 s without reactor scram No feedwater control on secondary side Codes used: TRACE, SIM-LFR, RELAP5, CATHARE, TRACE/FRED, SPECTRA
10
Total reactivity and feedbacks Core and MHX powers
RELAP5 Results
Total
Inserted
Doppler
Fuel exp.
Core power
MHX power
Maximum net reactivity insertion of 85 pcm Initial core power peak of 680 MW
UTOP transient (2/4)
11
Core temperatures Max clad and fuel temperatures
RELAP5 Results
Core outlet
MHX inlet
MHX outlet
Core inlet
Max fuel
Max clad
Maximum clad temperature remains below 650 °C Maximum fuel temperature of ~2930 °C at t = 50 s (hottest pin, middle core plane,
fuel pellet centre) exceeds the MOX melting point (~2700 °C) only local fuel melting
Core outlet
Max clad
Core inlet
UTOP transient (3/4)
12
Differences in fuel expansion reactivity feedback (free/linked effects) and fuel rod gap dynamic modelling
Only local fuel melting in the hottest pin is confirmed by all codes
Peak power and max fuel temperatures: RELAP5: 679 MW and 2930 °C SIM-LFR: 656 MW and 2996 °C CATHARE: 735 MW and 2866 °C TRACE/FRED: 642 MW and 2779 °C
UTOP transient (4/4)
13
SIM-LFR:Minimum clad failure time >> 1.0E+7 s
Different heat transfer correlations used by RELAP5 and CATHARE for fuel rod bundle
Maximum clad temperature is below 650 °C
1E+00
1E+02
1E+04
1E+06
1E+08
1E+10
1E+12
1E+14
0 20 40 60 80 100 120 140 160 180 200Time [sec]
Cla
d Fa
ilure
Tim
e [s
ec]
19.5
20.0
20.5
21.0
21.5
22.0
Fiss
ion
Gas
Pre
ssur
e [b
ar]
Clad Failure Time [sec]
Fission Gas Pressure [bar]30 min
14
ULOF transient (1/4)
All primary pumps coastdown without reactor scram No feedwater control on secondary side Codes used: RELAP5, SIM-LFR, CATHARE, TRACE, TRACE/FRED, SPECTRA
Active core flowrate Core and MHX powers
RELAP5 Results
Core power
MHX power
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
15
ULOF transient (2/4)
Core temperatures
Core temperatures
RELAP5 Results Initial clad peak temperature of 764 °C Max clad temp. stabilizes below 650 °C Positive Doppler and fuel exp. effects are
mainly counterbalanced by negative radial core exp. (Pad + Diag.), control rods and coolant exp. effects
Total reactivity and feedbacks
Max clad
Max lead
Core inlet
Max clad
Core inlet
Max fuel
DopplerFuel exp.
C. Rods
Pad + Diag.
Cool. exp.
16
ULOF transient (3/4)
Slight deviations in the initial core flow rate transient, but good agreement in stabilized natural circulation flow rate in the primary circuit
Core power at t = 200 s is slightly under predicted by SIM-LFR (P = 177 MW) and TRACE/FRED (P = 180 MW) with respect to RELAP5 (P = 195 MW) and CATHARE (P = 198 MW)
17
ULOF transient (4/4)
SIM-LFR:Minimum clad failure time >1.0E+5 s
The initial clad peak temperature is calculated in the range 730° C–764°C
Maximum clad temperature predicted by the codes at t = 200 s is around 650 °C
No clad failure is expected under ULOF in the short and long term
No vessel wall temperature increase (Tw < 400 ° C during ULOF transient)
1E+00
1E+02
1E+04
1E+06
1E+08
1E+10
1E+12
1E+14
0 50 100 150 200 250 300Time [sec]
Cla
d Fa
ilure
Tim
e [s
ec]
19.2
19.4
19.6
19.8
20.0
20.2
20.4
Fiss
ion
Gas
Pre
ssur
e [b
ar]
Clad Failure Time [sec]
Fission Gas Pressure [bar]
30 min
Min. Pin Clad Failure Time = 1.6E+5 sec at transient time t = 17.8 sec
18
ULOHS transient (1/4)
Loss of feedwater to all MHXs without reactor scram Startup of DHR-1 (3 out of 4 IC loops are in service) Codes used: RELAP5, SIM-LFR, CATHARE, TRACE, TRACE/FRED, SPECTRA
Core power progressively reduces down towards decay level removed by DHR-1 Maximum clad and vessel temperatures rise up to ~700 °C after about one hour
Core and MHX powers Core and vessel temperatures
Core power
MHX power
Max vessel
Max clad
CATHARE Results
19
ULOHS transient (2/4)
Total reactivity and feedbacksCore temperatures
Fuel temperature reduces down close to clad temperature Positive Doppler and fuel and clad expansion effects are mainly counterbalanced by
negative radial core expansion (Pad + Diag.), coolant expansion and control rods effects
Max clad
Core inlet
Max fuel
Doppler
Fuel exp.
Clad exp.
C. Rods
Cool. exp.
Pad + Diag.
CATHARE Results
20
ULOHS transient (3/4)
Vessel wall temperature is over predicted by RELAP5 and CATHARE (no heat losses from the external wall surface) with respect to SIM-LFR
Maximum vessel temperature rises over about 650 °C in 30 minutes no vessel failure is expected in the medium term vessel integrity is not guaranteed in the long term
21
ULOHS transient (4/4)
SIM-LFR:Minimum clad failure time > 1.0E+6 s
Maximum clad temperature stabilizes around 700 °C after one hour transient
No clad failure is calculated by SIM-LFR code in the short and long term
1E+00
1E+02
1E+04
1E+06
1E+08
1E+10
1E+12
1E+14
0 500 1000 1500 2000 2500 3000 3500Time [sec]
Cla
d Fa
ilure
Tim
e [s
ec]
0
5
10
15
20
25
30
Fiss
ion
Gas
Pre
ssur
e [b
ar]
Clad Failure Time [sec]
Fission Gas Pressure [bar]30 min
22
ULOHS+ULOF transient (1/4)
Loss of feedwater to all MHXs and all primary pumps without reactor scram Startup of DHR-1 (3 out of 4 IC loops are in service) Codes used: SIM-LFR, RELAP5, CATHARE, SPECTRA
0.0
0.2
0.4
0.6
0.8
1.0
0 500 1000 1500 2000 2500 3000 3500Time [sec]
rel.
units
[fr]
Power_thFlow_Cool
200
700
1200
1700
2200
0 500 1000 1500 2000 2500 3000 3500Time [sec]
Tem
pera
ture
[°C
] Fuelc_peak Clad_peakCool_out Cool_inT_wall
Core flow rate and power Core and vessel temperatures
Max fuel
Max lead, cladCore inlet, vessel
Power
Flow rate
SIM-LFR Results 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
23
ULOHS+ULOF transient (2/4)
24
ULOHS+ULOF transient (3/4)
Similar evolution of core flow rate and core power is calculated by the codes
Calculated vessel wall temperature is in the range 440 °C - 520 °C after 30 min.
Vessel integrity is guaranteed in the medium term and likely also in the long term according to RELAP5 results
25
ULOHS+ULOF transient (4/4)
SIM-LFR:Minimum clad failure time > 1.0E+4 s
The maximum clad temperature is over predicted of 25° C – 30 °C by RELAP5 and CATHARE with respect to SIM-LFR
The minimum clad failure time predicted by SIM-LFR is of about 3 hours
1E+00
1E+02
1E+04
1E+06
1E+08
1E+10
1E+12
1E+14
0 500 1000 1500 2000 2500 3000 3500Time [sec]
Cla
d Fa
ilure
Tim
e [s
ec]
0
5
10
15
20
25
30
Fiss
ion
Gas
Pre
ssur
e [b
ar]Clad Failure Time [sec]
Fission Gas Pressure [bar]
30 min
26
Partial FA blockage (RELAP5 results)
Code used: RELAP5, SIM-LFR, SIMMER
RELAP5 assumptions: Total ΔP over the FA = 1.0 bar ΔP at FA inlet = 0.22 bar Flow area blockage at FA inlet No heat exchange with
surrounding FAs
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
27
SCS Failure (1/2) (RELAP5 results) Secondary pressure
Core and MHX powers
Primary lead temperatures
Depressurization of all secondary circuits at t = 0 s (no availability of the DHR)
Reactor scram at t = 2 s on low secondary pressure
Initial MHX power increase up to 850 MW no risk for lead freezing
MHXs
Core
MHX outlet
28
SCS Failure (2/2) (RELAP5 results)
Core decay and MHX powers Core and vessel temperatures
No risk for lead freezing in the initial transient phase Slow primary temperature increase due to large thermal inertia of the primary
system large grace time for the operator to take opportune corrective actions
Core
MHX
29
Conclusions (1/2)
In all simulated transients there is a very large margin to coolant boiling since 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: Undetected FA 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 even 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
30
Conclusions (2/2)
The analysis of DEC transients with various codes has highlighted the very good intrinsic safety features of ALFRED design thanks to:
Benign characteristics of the coolant Good natural convection in the primary circuit Large thermal inertia to slow down the transients Prevalent negative reactivity feedbacks to limit 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 the opportune corrective actions and bring the plant in safe conditions in the medium and long term