Transcript LEVEL 2 PROBABILISTIC SAFETY ASSESSMENT MODEL FOR PAKS NUCLEAR POWER PLANT Attila BAREITH, Gabor LAJTHA, Zsolt TÉCHY VEIKI INSTITUTE FOR ELECTRIC POWER RESEARCH József ELTER PAKS.

LEVEL 2 PROBABILISTIC SAFETY
ASSESSMENT MODEL FOR
PAKS NUCLEAR POWER PLANT
Attila BAREITH, Gabor LAJTHA, Zsolt TÉCHY
VEIKI INSTITUTE FOR ELECTRIC POWER RESEARCH
József ELTER
PAKS NUCEAR POWER PLANT LTD
International Workshop On Level 2 PSA
and Severe Accident Management
Institute for Electric Power Research Co.
[email protected]
LEVEL 2 - 1
Cologne, Germany 29th to the 31st of March 2004
Content
•
•
•
•
Introduction
Scope of Level 2 PSA
Interface with Level 1, grouping of sequences
Accident progression and containment analysis
• Containment Event Trees
• Conditional Probability of Nodes
• Release Categories
• Accident management
• Results
International Workshop On Level 2 PSA
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LEVEL 2 - 2
Cologne, Germany 29th to the 31st of March 2004
VVER-440/213
Air traps
A257
Trays
A201_R
A202
A203
A201_L
Corridor
Corridor
A256
A201_L
A201_R
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LEVEL 2 - 3
Cologne, Germany 29th to the 31st of March 2004
Level 2 PSA
• 3 years work (2000-2003)
–
–
–
•
•
•
•
Preparation of models, connection between Level 1 and 2 PSA
Level 2 PSA for the present status of the plant
Level 2 PSA with assumption of accident management strategies
KFKI Atomic Energy Research Institute (AEKI)
VEIKI Institute for Electric Power Research Co
Paks NPP Co.
ABS Consulting Co. was responsible for the fragility curve calculations.
The starting point of the Level 2 PSA is the Level 1 PSA study. The existing
Level 1 PSA covers accident
- internal initiating events emerging at shutdown states
- spent fuel storage pool (SFSP) events
– internal initiating events and internal hazards as fire and flooding at
nominal power
This is the scope of the Level 2 PSA study
International Workshop On Level 2 PSA
and Severe Accident Management
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LEVEL 2 - 4
Cologne, Germany 29th to the 31st of March 2004
Interface with Level 1,
grouping of sequences
• reactor core status
reactor pressure at the onset of core damage
type and amount of emergency cooling before and during core
damage
• status of the containment systems
containment initial leakage rate, isolation failure, structural
damage, primary to secondary leakage (PRISE), by-pass
availability of containment systems (spray, bubbler condenser
trays, recirculation and ventilation systems)
International Workshop On Level 2 PSA
and Severe Accident Management
Institute for Electric Power Research Co.
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LEVEL 2 - 5
Cologne, Germany 29th to the 31st of March 2004
Interface with Level 1,
grouping of sequences (Cont’d)
Parameters at the moment of
core melt
Isolated containment
Unisolated containment
Spray availability
Pressure
ECCS
Very low
(<7bar)
E-HA
E
U
PDS
0
1
2
3
E/U
N
E
U
E/U
N
E
U
4
5
6
7
8
9
10
11
9.57E-10
1.52E-9
E/U
12
N
E
U
13
14
15
E/U
N
16
17
Low
(7bar< p
<20bar)
Medium
(20bar< p
<60 bar)
High
(>60bar)
R
H
Containment bypass
(PRISE, interface
LOCA)
Sump
leakage
Spray availability
N
R
H
N
Spray availability
R
H
N
H
I
J
A
B
C
D
E
F
G
3.53E-9
4.56E-6
1.35E-8
5.05E-10
2.09E-6
1.90E-9
6.02E-12
1.65E-6
3.22E-7
5.04E-7
1.93E-9
2.93E12
3.14E-9
3.37E-9
1.94E-9
5.88E-11
2.22E-10
3.02E-9
1.58E-5
1.34E-10
9.35E-8
5.03E-12
1.12E-10
2.28E-10
6.70E-7
5.87E-6
2.91E-10
6.93E-9
6.31E-11
3.68E-10
1.49E-7
7.79E-11
8.13E-8
2.50E-7
4.95E-11
7.69E-10
1.47E-7
1.25E-8
9.18E-7
8.29E-10
3.17E-9
5.08E-8
1.97E-10
6.59E-10
1.83E-11
1.23E-8
1.85E11
8.12E-9
1.41E-11
3.35E-8
8.69E-9
4.06E-9
1.12E-10
8.07E-10
4.53E-10
1.86E-8
2.20E10
6.79E10
1.55E-7
5.03E-12
2.23E-9
2.45E-10
8.84E-12
International Workshop On Level 2 PSA
and Severe Accident Management
Institute for Electric Power Research Co.
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LEVEL 2 - 6
Cologne, Germany 29th to the 31st of March 2004
ACCIDENT PROGRESSION AND
CONTAINMENT ANALYSIS
• Type of code: MAAP4/VVER code developed from the
original MAAP code by Westinghouse (WESE)
• MAAP provides an integrated framework for evaluating
the timing of key accident events, thermodynamic
histories of the reactor coolant system, core and
containment, and corresponding estimates of fission
product release and transport.
• Supplemented with calculations performed with
CONTAIN, H2AICC, VESSEL, MVITA, ICARE
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LEVEL 2 - 7
Cologne, Germany 29th to the 31st of March 2004
ACCIDENT PROGRESSION AND
CONTAINMENT ANALYSIS (cont’d)
• Phenomena within the RPV
-core-heat-up and degradation
-zirconium oxidation
-fission product release from fuel and transport in primary circuit
-core degradation and loss of geometry
-vessel melt-through
• Phenomena within the reactor cavity
-debris ejection from vessel, direct containment (cavity) heating
-debris structure heat transfer (cavity door)
-high pressure melt ejection (fission product release)
-ex-vessel core-coolant interaction
-steam explosion
-core-concrete interaction
• Phenomena within the containment building
VVER-440/213 specific containment thermal-hydraulics (pressurisation)
hydrogen combustion
engineered safety features (spray system)
transport of fission products (bubble condenser, spray, leak)
International Workshop On Level 2 PSA
and Severe Accident Management
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LEVEL 2 - 8
Cologne, Germany 29th to the 31st of March 2004
Containment Event Trees
CET structure and nodal questions:
• Represented by 3 different time regimes
•Questions - Early phase
–
–
–
–
–
–
–
Temperature induced failure of the primary coolant system
Reactor cavity flooded
Melt progression arrested
Spray system recovery
Hydrogen management
Hydrogen burn
Containment failure mode
International Workshop On Level 2 PSA
and Severe Accident Management
Institute for Electric Power Research Co.
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LEVEL 2 - 9
Cologne, Germany 29th to the 31st of March 2004
Containment Event Trees
(Cont’d)
•Questions - Intermediate phase
–
–
–
–
High Pressure Melt Ejection (HPME)
RPV failure: pour
Steam explosion
Containment failure mode
•Questions - Late phase
–
–
–
–
–
–
Molten Core Concrete Interaction
Cavity door failure
Spray system recovery
Hydrogen burn
Filtered vent (open, close)
Containment failure mode
International Workshop On Level 2 PSA
and Severe Accident Management
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LEVEL 2 - 10
Cologne, Germany 29th to the 31st of March 2004
Hydrogen Burn
•Hydrogen production (MAAP calculation for each sequence)
•Hydrogen distribution in each volume (H2. CO, O2, CO2, H2O mole
fraction, pressure, temperature versus time) –
from MAAP calculation
• Combustion mechanism – three combustion mechanisms are
distinguished (burn and Deflagration Detonation Transition), for
the determination of containment pressure load the H2AICC
code is used with Modified Adiabatic Isochoric Complete
Combustion (AICC) model
•Pressure load due to hydrogen burn
• hydrogen deflagration - HBURN code calculated pressure versus time
• DDT
- time frame and probability based on ShermanBerman conditions
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and Severe Accident Management
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LEVEL 2 - 11
Cologne, Germany 29th to the 31st of March 2004
Calculated Pressure Load Due to Hydrogen Burn
in-vessel
PDS_05C
no recovery,
ex-vessel
9.00E+05
3.00E-01
8.00E+05
P-AICC
2.50E-01
Pload
7.00E+05
DDT(50%)
6.00E+05
2.00E-01
H2ave
O2ave
5.00E+05
1.50E-01
4.00E+05
3.00E+05
1.00E-01
2.00E+05
5.00E-02
1.00E+05
Idő (s)
1.23E+05
1.20E+05
1.16E+05
1.12E+05
1.09E+05
1.05E+05
1.02E+05
9.80E+04
9.44E+04
8.72E+04
9.08E+04
International Workshop On Level 2 PSA
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8.36E+04
8.00E+04
7.64E+04
7.28E+04
6.92E+04
6.56E+04
6.20E+04
5.84E+04
5.48E+04
5.12E+04
4.76E+04
4.40E+04
4.04E+04
3.68E+04
3.32E+04
0.00E+00
2.96E+04
0.00E+00
2.60E+04
Pressure Load (Pa)
P
LEVEL 2 - 12
Cologne, Germany 29th to the 31st of March 2004
Containment Failure due to
Hydrogen Burn
• Determination of the probability of Ignition –
• probability of ignition depends on the existence of igniting sources
and also on the hydrogen concentration, duration of different
hydrogen concentrations (recombiner)
• Determination of the probability of pressure load
PDS_05C In-vessel
PAR ignition (from 10vol%, ignition: 1800 s, 20 vol%)
G(p)
g(p)
1
3.50E-02
0.9
3.00E-02
0.8
g(p)
2.50E-02
Probability
0.6
2.00E-02
0.5
1.50E-02
0.4
0.3
Probability density
G(p).
0.7
1.00E-02
0.2
5.00E-03
0.1
0
1.50E+05
2.00E+05
2.50E+05
3.00E+05
3.50E+05
4.00E+05
4.50E+05
0.00E+00
5.00E+05
Pressure (bar)
International Workshop On Level 2 PSA
and Severe Accident Management
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LEVEL 2 - 13
Cologne, Germany 29th to the 31st of March 2004
Containment Failure
nodal question for containment failure due to hydrogen burn
Joint treatment of containment loads and fragility curves
Density function of the pressure load probability: f(p), distribution function: F(p).
The probability of the containment damage is described by the fragility curve: Frag(p) = P(pfail < p)
The Containment Failure Probability for the entire load pressure range is
CFP = integral dp f(p) Frag(p) = integral dp f(p) • integral dp` frag(p`)
1
0,3
PDS_05C
Load and Fragility
0,9
Load Distribution
Load (G(p) and Fragility(f(p)
0,8
0,25
Containment Failure Probability
CFP= 0,23
Containment Fragility
DensityFunction
0,7
0,6
0,2
Conv. Int Value (Numerical integral)
0,15
0,5
0,4
0,1
0,3
0,2
0,05
0,1
0
1,5
1,7
1,9
2,1
2,3
2,5
2,7
Pressure (bar, overpressure)
3,1
3,3
0
3,5
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2,9
LEVEL 2 - 14
Cologne, Germany 29th to the 31st of March 2004
Conditional Probability of CET Nodes
• Temperature induced hot leg failure for high pressure sequences MAAP calculation – failure was considered but it was not taken into account,
conservative assumption
•
•
•
•
Core melt arrested - recovery time was assumed with an exponential distribution
Containment failure due to hydrogen burn - calculated
Cavity failure due to DCH - cavity pressure calculated by CONTAIN code,
Cavity door seal failure - expert judgement based on
VESSEL code and hand calculations
• Containment overpressurization - calculated, comparison of the calcul
ated pressure and fragility curve
• Steam explosion - based on expert judgement
taking into account
available water mass and corium, corium temperature (superheated, saturated),
water temperature (saturated, subcooled)
International Workshop On Level 2 PSA
and Severe Accident Management
Institute for Electric Power Research Co.
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LEVEL 2 - 15
Cologne, Germany 29th to the 31st of March 2004
Release Categories
•
•
•
MAAP calculates the fission product release and transport (from fuel to environment)
Grouping of fission products (release time, height)
Binning of event tree and states into release categories
Source term
category
1
2
3
4
5
6
7
8
9
10
11
11A
11B
12
12A
12B
13
Description
High Pressure Core Melt or Steam Explosion, the reactor cavity is
damaged, the molten core is evacuated from the containment
By-pass cases, including arrested core melt
Containment isolation failure or containment rupture, spray is inactive
Early hydrogen burn, no containment rupture, spray is inactive
Late hydrogen burn with containment rupture, spray is inactive
Late hydrogen burn, no containment rupture, spray is inactive
Containment isolation failure or containment rupture, spray operates
Early hydrogen burn, no containment rupture, spray operates
Late hydrogen burn with containment rupture, spray operates
Late hydrogen burn, no containment rupture, spray operates
Intact containment, spray is inactive
Intact, filtered venting, spray is inactive
Intact, basemat meltthroug, spray is inactive
Intact containment, spray operates
Filtered venting, spray operates
Intact, basemat meltthroug, spray operates
Partial core damage
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and Severe Accident Management
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LEVEL 2 - 16
Cologne, Germany 29th to the 31st of March 2004
Accident Management
Objective
Present
situation
Strategy 1
Strategy 2
Prevention of RPV
Failure
Hydrogen
Management
Limitation of
Radioactive Releases
Prevention of Cont.
Slow
Overpressurization
Preserve of Cavity
Integrity
Cooling of Molten
Corium
ECCS recovery
ECCS recovery
-
Igniters and
recombiners
Spray recovery
ECCS recovery +
Cavity flooding
Igniters and
recombiners
Spray recovery
Spray recovery
Spray recovery +
Filtered venting
Spray recovery +
Filtered venting
-
Room A004
hermetization
-
(solved by cavity
flooding)
(partly solved by
cavity flooding)
Spray recovery
-
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LEVEL 2 - 17
Cologne, Germany 29th to the 31st of March 2004
A004 hermetization
(Mitigating of the effect of the Cavity Door Failure)
A004
Door of
A00041.
cavity
door
ajtó
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LEVEL 2 - 18
Cologne, Germany 29th to the 31st of March 2004
Cavity Flooding
External cooling of the reactor pressure vessel
water injection line
Ventillation
line
Isolation and radiation shield
(in lower position)
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and Severe Accident Management
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LEVEL 2 - 19
Cologne, Germany 29th to the 31st of March 2004
Results of Containment Performance Analysis
State of the Containment Structure
(Atmospheric Release)
Base
%
Acc. Man I. Acc. Man II.
Structural
High Pressure Vessel Failure (HPVF)
0,002
0,002
0,002
Failure
Early Containment Failure (ECF, ECFS)
Late Containment Failure (LCF. LCFS)
Late Containment Leak (LCL. LCLS)
0,119
0,000
0,395
0,015
0,017
0,005
0,015
0,023
0,185
Isolation
Isolation Failure
0,030
0,030
0,030
Failure
Filtered Vent Remains Open (FVO)
Total of failure states
0,000
0,537
0,020
0,090
0,020
0,276
Filtered Vent (FV)
Intact (I, IS)
Partly Damaged Core
Controlled release states
Remaining 1% of the PDS's
Total
0,000
0,254
0,199
0,453
0,01
1,00
0,416
0,265
0,199
0,900
0,01
1,00
0,251
0,264
0,199
0,723
0,01
1,00
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LEVEL 2 - 20
Cologne, Germany 29th to the 31st of March 2004
Conclusion
• Effective Reduction of Early Containment Failure Probability
– due to hydrogen management
• Effective Reduction of Late Containment Leak Probability
– due to A 004 compartment hermetization or cavity flooding)
• Effective reduction of basemat melt through
– due to cavity flooding
International Workshop On Level 2 PSA
and Severe Accident Management
Institute for Electric Power Research Co.
[email protected]
LEVEL 2 - 21
Cologne, Germany 29th to the 31st of March 2004