Simulations of large break LOCA without safety … 2015, Marseille March 24 – 26, 2015 Simulations...
Transcript of Simulations of large break LOCA without safety … 2015, Marseille March 24 – 26, 2015 Simulations...
ERMSAR 2015, Marseille March 24 – 26, 2015
Simulations of large break LOCA without safety injection
for EPR reactor using MELCOR computer code
Piotr Darnowski, Eleonora Skrzypek, Piotr Mazgaj,
Michał Gatkowski
Warsaw University of Technology,
Institute of Heat Engineering, Poland
Session IV: SA Scenarios
ERMSAR 2015, Marseille March 24 – 26, 2015
Outline
1. Introduction
2. MELCOR model
3. Scenario and assumptions
4. Results
5. Conclusions
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ERMSAR 2015, Marseille March 24 – 26, 2015
Introduction
Activities in the framework of the two projects: SARWUT (2013-2014)
and INSPE (2014-2015).
Initiation of the Severe Accident research – no activities in the field
before.
MELCOR 2.1 code. Models development: EPR, Zion PWRs and
BWR.
LBLOCA, SBO, Total Loss of AC Power and variations.
Cooperation with National Atomic Energy Agency (PAA), Areva NP,
GE-Hitachi NE and National Centre for Nuclear Research (NCBJ)
EPR selected because it is potential PWR design for the first polish
NPP (~ 3000 MWe, >2028).
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MELCOR model
Model based on publicly available data, Ref. [1],[2].
Relatively simple nodalization (59 CVH, 74 FL, 117 HS).
Containment with single CV.
Model parameters, coefficient and setup based on SOARCA
recomendations Ref. [5].
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LP
DC
UP
BYPASS
UH
Ring: 1 2 3 4 5 6
19181716151413121110987654321
Level:
CORE
Loop x3 Loop x1
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MELCOR model
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LP
DC
UP
BYPASS
UH
CORE
RPV and core model:
10 control volumes
19 axial levels and 6 rings
30 heat strucutres
EOC core
ERMSAR 2015, Marseille March 24 – 26, 2015
MELCOR model
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CV
H 1
20
CV
H 1
40
CVH 130
FL 1
20
FL 1
10
FL 1
30
FL 507FL 5
02
FL 1
40
CVH 150
CVH 110
CV
21
0/2
40
CV
20
0/2
30
CV 220/250
FL 2
21
/24
2
FL 197/182
FL 2
21
/25
1
FL 198/181
FL 2
11
/24
1
FL 201/230
SG Secondary Side SG Primary Side
CV, FP & HS
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MELCOR model
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501502
510
50
8
511512
509
511
507 503
512 510
509 508
500501
502507
558
71
0
560
71
3
PDS
PSV
PRT Tank
1 Loop
HS4
06
00
559
71
2
Rupture disk
HS4
05
8
HS4055971
1
561
550
551
71
47
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Containment
PRZ with single RCS loop and 1CV containment
ERMSAR 2015, Marseille March 24 – 26, 2015
Scenario
LBLOCA with complete loss of active safety injection - SA
for PWR type reactor.
Fast progressing sequence with core meltdown.
Comparison of two LB-LOCAs: 2A-LOCA (cold leg double
ended break) and SL-LOCA (surge line double ended
break).
Due to the Break Preclusion Principle 2A-LOCA is not a
design basis accident for the EPR reactor (Ref. [4],[6]).
Main motivation: to investigate differences between
2A-LOCA and SL-LOCA.
Only In-Vessel phase considered.
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Assumptions
2A-LOCA (Cold-Leg) and SL-LOCA (Surge-Line).
Active safety injections (MHSI & LHSI) are not available.
Accumulators are available.
SCRAM at time zero, pumps cost-down starts at time zero.
MFW isolated at accident initiation.
MSIV closure at accident initiation.
EFWS available.
Secondary Partial Cooldown activated at SI signal (MSRV setpoint 95.5 bars to 60 bars in 20 min).
Pressurizer Discharge System (PDS) activated at core exit gas T=650 °C.
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Results
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SL-LOCA reference MAAP4.0.4 results available in
Ref. [3].
Containment peak pressure: 3.5 and 4.3 bar for SL-
LOCA and 2A-LOCA respectively. Reference [3] – 3.5
bar.
For 2A-LOCA Cathare V2.5 predicts 4 bars in
Ref. [8] and PAREO9 4.3 bars Ref. [7].
2A-LOCA SL-LOCA SL-LOCA (Ref. 3)
0 0 0
Start of core uncovery (S) 0.2 57 60
SI signal 7.2 148.5 N/A
ACC injection 11 152 120
ACC depletion 37.5 228.5 180
Containment pressure peak 103 183 100
PDS valves open 497 1702 1500
Start of oxidation 591 1587 1560
Core fully uncovered (C) 919 1926 N/A
Start of core melting 1024 1968 1740
Massive core relocation 4764 5834 7380
Vessel failure 6070 6732 10680
TIME [s]KEY EVENT
Pipe rupture
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Results
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2A-LOCA SL-LOCA SL-LOCA (Ref. 3)
0 0 0
Start of core uncovery (S) 0.2 57 60
SI signal 7.2 148.5 N/A
ACC injection 11 152 120
ACC depletion 37.5 228.5 180
Containment pressure peak 103 183 100
PDS valves open 497 1702 1500
Start of oxidation 591 1587 1560
Core fully uncovered (C) 919 1926 N/A
Start of core melting 1024 1968 1740
Massive core relocation 4764 5834 7380
Vessel failure 6070 6732 10680
TIME [s]KEY EVENT
Pipe rupture
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2A-LOCA
Cladding Temperatures
for Inner Core Ring #1
SL-LOCA
Ref. [3]: 490 kg
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Results
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Mass in lower plenum before core plate failure (10-15 tons) – after ~160
tons.
16 minutes difference in massive relocation time.
11 minutes difference in RPV time-to-rupture.
RPV failure: 22 min and 15 min after relocation for 2A-LOCA and SL-LOCA
respectively.
SL-LOCA RPV breach – 1h 05min before MAAP4 (Ref. [3]).
2A-LOCA SL-LOCA SL-LOCA (Ref. 3)
0 0 0
Containment pressure peak 103 183 100
Start of core uncovery (S) 0.2 57 60
7.2 148.5 N/A
11 152 120
37.5 228.5 180
PDS valves open 497 1702 1500
Start of oxidation 591 1587 1560
Core fully uncovered (C) 919 1926 N/A
Start of core melting 1024 1968 1740
Massive core relocation 4764 5834 7380
6070 6732 10680
TIME [s]
SI signal
ACC injection
ACC depletion
Vessel failure
KEY EVENT
Pipe rupture
2A-LOCA SL-LOCA SL-LOCA (Ref. 3)
0 0 0
Containment pressure peak 103 183 100
Start of core uncovery (S) 0.2 57 60
7.2 148.5 N/A
11 152 120
37.5 228.5 180
PDS valves open 497 1702 1500
Start of oxidation 591 1587 1560
Core fully uncovered (C) 919 1926 N/A
Start of core melting 1024 1968 1740
Massive core relocation 4764 5834 7380
6070 6732 10680
TIME [s]
SI signal
ACC injection
ACC depletion
Vessel failure
KEY EVENT
Pipe rupture
ERMSAR 2015, Marseille March 24 – 26, 2015
Issues
Heavy Reflector is modeled as HS in MELCOR.
In EPR sideward reflector melt-through is expected and relocation through downcomer.
MELCOR allows relocation through core bypass (TMI like) but not through the downcomer.
New model with HR being a part of COR package is considered.
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TMI
LP
DC
UP
BYPASS
UH
CORE
EPR
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Conclusions
Containment pressure peak 0.8 bar difference for 2A-LOCA and SL-LOCA.
Similar H2 production.
10-15 minutes difference in core degradation phenomena beetwen 2A-LOCA and
SL-LOCA - core relocation and RPV rupture.
Vessel failure earlier in comparison to Ref. [3] results by one hour for SL-LOCA.
Model needs further development.
Uncertainty and sensitivity is recommended.
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Future plans
Sensitivity and Uncertainty studies.
Presented model is under development.
Detailed containment modifications in progress.
Detailed RPV model in progress.
Simulations of Ex-Vessel phase.
Other activities (ASTEC code, RELAP/SCADAP code,
experiments, ISPs, Fukushima, TMI, Re-criticality
during SA).
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References
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[1] Areva & EDF, 2013. The Pre-Construction Safety Report (PSCR) for UK EPR Reactor, www.epr-reactor.co.uk.
[2] U.S. EPR Application Documents, Final Safety Analysis Report, 2014
[3] UK EPR GDA Submission, The Pre-Construction Safety Report (PCSR), Chapter 16.2 RCC-B – Severe Accident Analysis
[4] Chapuliot, S., Migné, C., Break Preclusion Concept and its Application to the EPRTM Reactor, Nuclear Engineering and Design,
Volume 269, 97-102, 2014
[5] Sandia National Laboratories, MELCOR Best Practices as Applied in the State-of-the-Art Reactor Consequence Analyses
(SOARCA) Project, Nuclear Regulatory Commission, NUREG/CR-7008, 2014.
[6] UK EPR GDA Submission, The Pre-Construction Safety Report (PCSR), Chapter 5.2 – Integrity of the Reactor Coolant Pressure
Boundary
[7] UK EPR GDA Submission, The Pre-Construction Safety Report (PCSR), Chapter 16.4 – Specific Studies
[8] UK EPR Fundamental Safety Overview, Volume 2: Design and Safety, Chapter F: Containment Systems and Safeguard
Systems, Subchapter: F.2 Section F.2.1., 2014
ERMSAR 2015, Marseille March 24 – 26, 2015
Acknowledgments
The publication was financed and partially created in the framework of a
project: Innovative Nuclear and Sustainable Power Engineering (INSPE) on
The Faculty of Power and Aeronautical Engineering at Warsaw University
of Technology. The INSPE is co-financed with EU funds by European
Social Fund.
Presented model was partially created in the framework of a strategic
project NCBiR: “Technologies for the development of safe nuclear energy”,
Research Task No. 9 entitled “Development and implementation of safety
analysis methods in nuclear reactors during disturbances in heat removal
and severe accident conditions”
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Thank you for your attention
Contact:
ERMSAR 2015, Marseille March 24 – 26, 2015
Additional Slides
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Break Preclusion Principle
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Ref. [2]
Ref. [2]
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Results
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Additional slides
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Additional Slides
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LP
DC
UP
BYPASS
UH
CORE
LP
DC
UP
BYPASS
UH
Ring: 1 2 3 4 5 6
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Level:
CORE
ERMSAR 2015, Marseille March 24 – 26, 2015
Containment
IRWST – 2086 m3 of water
Containment RH=0.5, p=1 bar, V=78500 m3
47 PARS – default MELCOR
ACCx1 m=29667kg ACCx3 m=89334 kg water mass
ACC setpoint 45 bars
SI signal + Partial Cooldown = 115 bar
RT, TT – 135 bar
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Steady State
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Time [s]: -0.97
Pressurizer Pressure [Pa]: 15501700.00
HLx1 Flow [kg/s]: 5571.76
CLx1 Flow [kg/s]: 5574.31
HLx3 Flow [kg/s]: 16793.30
CLx3 Flow [kg/s]: 16789.60
CL Total Flow [kg/s]: 22363.90
HL Total Flow [kg/s]: 22365.00
RCS Water Inv. [kg]: 271151.00
PRZ Level [m]: 6.73
PRZ Water Mass [kg]: 22537.00
Feedwater Flow x1 [kg/s]: 657.51
Feedwater Flow x3 [kg/s]: 1972.52
SGx1 Outflow [kg/s]: 657.50
SGx3 Outflow [kg/s]: 1972.40
SGx1 Pressure [Pa]: 7706760.00
SGx3 Pressure [Pa]: 7708300.00
SGx1 Level [m]: 14.93
SGx3 Level [m]: 14.93
SGx1 Water Mass [kg]: 77970.60
SGx3 Water Mass [kg]: 233847.00
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MAAP4 – Ref. [3] Results
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MAAP4 – Ref. [3] Results
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Additional Slides
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Additional Slides
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From: EMUG 2010, New ‚Best Practice’ Default Values for MELCOR 2.1
Several tons of debris material was spotted in LP, before core plate
failure.
It is not unusual in MELCOR simulations.
EPR
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Additional Slides
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SL-LOCA Fuel Temperatures
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Additional Slides
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Additional Slides
Simulation setup
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No. Field(s) Type/Package Unit(s) SOARCA DFT186 DFT21
1 HFRZUO UO2 7500 1000 7500
2 HFRZZR Zr 7500 1000 7500
3 HFRZSS Steel 2500 1000 2500
4 HFRZZX ZrOX 7500 1000 7500
5 HFRZSX SSOX 2500 1000 2500
6 HFRZCP Poison 2500 1000 2500
7 FCELR Radial 0.1 0.25 0.1
8 FCELA Axial 0.1 0.25 0.1
9 4413(5) S.C. 1.00E-06 0.05
10 4414(1) S.C. 0.01 0.001 0.01
11 4415(1) S.C. 1 0.5 1
12 4055(2) S.C. 0.5 5.00E-04 0.5
13 1001(1,2) S.C. Zr - O2 reaction - low temp. range const coeff kg2/m4/s 26.7 50.4 26.7
14 1001(2,2) S.C. Zr - O2 reaction - low temp. range exp. constant K 17490 14630 17490
15 1001(3,2) S.C. Zr - O2 reaction - high temp. constant coeff kg2/m4/s 26.7 0 26.7
16 1001(4,2) S.C. Zr - O2 reaction - high temp. exp. constant 17490 0 17490
17 1001(5,2) S.C. Zr - O2 reaction - upper temp. Bound. For low range 9998 10000 9998
18 1001(6,2) S.C. Zr - O2 reaction - upper temp. Bound. For high range 9999 10000 9999
19 1004(1) S.C. Oxidation cutoff temperatures - minimum 1100 1100 1100
20 1250(1) S.C. COR package temp. For enhanced debris-LH conduction 2800 3200 2800
21 1505(1) S.C. - 0.05 0.001 0.05
22 1505(2) S.C. - 0.05 0.001 0.05
23 1600(1) S.C. - 1 0 1
24 1603(2) S.C. COR package min yield stress temperature K 1700 1800 1700
25 IACTV BUR -
26 KFLSH FL -
27 DRGAP COR m
28 HDBPN COR HTC from debris to penetration structures W/m2/K 100 1000 1000
29 HDBLH COR W/m2/K 100 1000 1000
30 MDHMPO COR - MODEL 1000 1000
31 MDHMPM COR - MODEL 1000 1000
32 TPFAIL COR Failure temp.of the penetrations or the LH K 9999 1273.15 1273.15
33 CDISPN COR Discharge coef. for debris through failed penetration - 1 1 1
34 HDBH2O COR W/m2/K 2000 100 100
35 VFALL COR m/s 0.01 1 1
36 IAICON COR - Not active Not active Not active
37 PORDP COR Porosity of particulate debris defaults N/A - 0.4 0.4 0.4
38 DHYPD COR Debris equivalent diamter Core/LP defaults N/A m 0.01/0.002 0.01/0.002 0.01/0.002
39 1132(1) S.C. K 2800 3100 3100
40 1141(2) S.C. kg/m/s 0.2 1 1
41 4401(3) S.C. Max numb. Of iter. Permitted before solution is repeated - 15 default default
42 CPFPL/CPFAL HS -
43 EMISWL HS - 0.27 Not active Not active
44 RMODL HS - EQ-BAND Not active Not active
45 PATHL HS m 0.1 Not active Not active
46 DEGAS HS DEGAS model - not active due to numerical problems -
47 MLT MP Material interactions parameters - nuemrical problems -
48 IRODDOMAGE COR - SOARCA Not active Not active
49 RCLADTHICKNESS COR Min. Unox. clad thick. under which Fuel Failure Table works m 5.70E-05 Not active Not active
50 DRZRMN COR m 1.00E-04 1.00E-04 1.00E-04
51 DRSSMN COR m 1.00E-04 1.00E-04 1.00E-04
COR package min. porosity for flow and HT
COR package 1-dim stress/strain distribution
HS temperature convergence criterion
CVH/FL direct versus iterative solution algorithm
COR package min. CVH volume fraction
COR_CHT W/m2/KCandling Heat transfer coefficient
Radiation heat transfer coefficientCOR_RF
Description
Not active
Not used
0.0/1.0 core; 0.5/0.5 elsewhere
BUR package was not active
Not active
Higher than 0
K
-
Burn package activation
Superheated pool flashing model
Thickness of gas gap
Flow blockage friction parameters
HTC from debris to LH
HTC from oxidic pool to lower head
HTC in-vessel failing debris to pool
Velocity of failing debris
Control rods release model
HTC from metalic pool to lower head
Nominal optical distatnce
Fuel Failure Table
Critical min. thickness of unox steel
Fuel rod collapse temeprature
Max. Molten Zr breakout flow rate
Pool fraction HS settings
Steel emissivity
Eq. Band radiation
Critical min. thickness of unox zircaloy
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Additional Slides
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Additional Slides
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Additional Slides
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