LEVEL 2 PROBABILISTIC SAFETY ASSESSMENT MODEL FOR PAKS NUCLEAR POWER PLANT

Institute for Electric Power Research [email protected]

International Workshop On Level 2 PSA

and Severe Accident Management

Cologne, Germany 29th to the 31st of March 2004 LEVEL 2 - 1

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





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





VVER-440/213

Air traps

A256

A201_L A201_R

Trays

A257

A20

2

A20

3

Cor

ridor

Cor

ridor

A201_L A201_R





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





Interface with Level 1, grouping of sequences

• reactor core statusreactor pressure at the onset of core damagetype and amount of emergency cooling before and during core damage

• status of the containment systemscontainment 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)





Interface with Level 1, grouping of sequences (Cont’d)

Isolated containment Unisolated containment

Containment by-pass

(PRISE, interfaceLOCA)

Parameters at the moment ofcore melt

Spray availability Spray availabilitySump

leakageSpray availability

Pressure ECCS R H N R H N R H N

PDS A B C D E F G H I J- 0 3.53E-9 4.56E-6 1.35E-8 5.05E-10 2.09E-6 1.90E-9 6.02E-12

E-HA 1E 2 1.65E-6 3.22E-7 1.93E-9 3.14E-9 3.37E-9 5.88E-11 2.22E-10U 3 5.04E-7 2.93E-

121.94E-9

E/U 4

Very low(<7bar)

N 5 3.02E-9 1.49E-7 1.58E-5 1.34E-10 9.35E-8 6.70E-7 5.87E-6E 6 7.79E-11 5.03E-12U 7 9.57E-10 8.13E-8 1.12E-10

E/U 8 1.52E-9 2.50E-7 2.28E-10Low

(7bar< p<20bar) N 9 4.95E-11 1.86E-8 2.91E-10 6.93E-9

E 10 6.31E-11 7.69E-10 5.03E-12U 11 3.68E-10 1.47E-7 2.20E-

102.23E-9

E/U 12 1.25E-8 9.18E-7 6.79E-10

2.45E-10

Medium(20bar< p<60 bar)

N 13 6.59E-10 1.55E-7 8.29E-10 3.17E-9 5.08E-8E 14 1.83E-11U 15 1.97E-10 1.23E-8 1.85E-

118.12E-9 1.41E-11

E/U 16 1.12E-10 8.07E-10

High(>60bar)

N 17 4.53E-10 3.35E-8 8.69E-9 4.06E-9 8.84E-12





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





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 buildingVVER-440/213 specific containment thermal-hydraulics (pressurisation)hydrogen combustionengineered safety features (spray system)

transport of fission products (bubble condenser, spray, leak)





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





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





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 Sherman-

Berman conditions





PDS_05C no recovery,

0.00E+00

1.00E+05

2.00E+05

3.00E+05

4.00E+05

5.00E+05

6.00E+05

7.00E+05

8.00E+05

9.00E+05

2.60

E+

04

2.96

E+

04

3.32

E+

04

3.68

E+

04

4.04

E+

04

4.40

E+

04

4.76

E+

04

5.12

E+

04

5.48

E+

04

5.84

E+

04

6.20

E+

04

6.56

E+

04

6.92

E+

04

7.28

E+

04

7.64

E+

04

8.00

E+

04

8.36

E+

04

8.72

E+

04

9.08

E+

04

9.44

E+

04

9.80

E+

04

1.02

E+

05

1.05

E+

05

1.09

E+

05

1.12

E+

05

1.16

E+

05

1.20

E+

05

1.23

E+

05

Idő (s)

Pre

ssu

re L

oad

(P

a)

0.00E+00

5.00E-02

1.00E-01

1.50E-01

2.00E-01

2.50E-01

3.00E-01

Co

nce

ntr

atio

n (

-)

P-AICC

Pload

P

DDT(50%)

H2ave

O2ave

Calculated Pressure Load Due to Hydrogen Burn

in-vessel ex-vessel





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-vesselPAR ignition (from 10vol%, ignition: 1800 s, 20 vol%)

0

0.1

0.2

0.3

0.4

0.5

0.6

0.7

0.8

0.9

1

1.50E+05 2.00E+05 2.50E+05 3.00E+05 3.50E+05 4.00E+05 4.50E+05 5.00E+05

Pressure (bar)

Pro

ba

bil

ity

0.00E+00

5.00E-03

1.00E-02

1.50E-02

2.00E-02

2.50E-02

3.00E-02

3.50E-02

Pro

ba

bil

ity

de

ns

ityG(p). g(p)

G(p)g(p)





Containment Failurenodal 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`)

PDS_05C Load and Fragility

0

0,1

0,2

0,3

0,4

0,5

0,6

0,7

0,8

0,9

1

1,5 1,7 1,9 2,1 2,3 2,5 2,7 2,9 3,1 3,3 3,5

Pressure (bar, overpressure)

Lo

ad (

G(p

) an

d F

rag

ility

(f(p

)

0

0,05

0,1

0,15

0,2

0,25

0,3

Load Distribution

Containment FragilityDensityFunction

Conv. Int Value (Numerical integral)

Containment Failure Probability CFP= 0,23





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 onVESSEL code and hand calculations

• Containment overpressurization - calculated, comparison of the calculated 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)





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 termcategory

Description

1 High Pressure Core Melt or Steam Explosion, the reactor cavity isdamaged, the molten core is evacuated from the containment

2 By-pass cases, including arrested core melt3 Containment isolation failure or containment rupture, spray is inactive4 Early hydrogen burn, no containment rupture, spray is inactive5 Late hydrogen burn with containment rupture, spray is inactive6 Late hydrogen burn, no containment rupture, spray is inactive7 Containment isolation failure or containment rupture, spray operates8 Early hydrogen burn, no containment rupture, spray operates9 Late hydrogen burn with containment rupture, spray operates

10 Late hydrogen burn, no containment rupture, spray operates11 Intact containment, spray is inactive

11A Intact, filtered venting, spray is inactive11B Intact, basemat meltthroug, spray is inactive12 Intact containment, spray operates

12A Filtered venting, spray operates12B Intact, basemat meltthroug, spray operates13 Partial core damage





Accident Management

Objective Present

situation Strategy 1 Strategy 2

Prevention of RPV Failure

ECCS recovery ECCS recovery ECCS recovery + Cavity flooding

Hydrogen Management

- Igniters and recombiners

Igniters and recombiners

Limitation of Radioactive Releases

Spray recovery Spray recovery Spray recovery

Prevention of Cont. Slow

Overpressurization

Spray recovery

Spray recovery + Filtered venting

Spray recovery + Filtered venting

Preserve of Cavity Integrity

- Room A004 hermetization

(solved by cavity flooding)

Cooling of Molten Corium

- - (partly solved by cavity flooding)





A004 hermetization (Mitigating of the effect of the Cavity Door Failure)

Door ofA00041.

cavitydoorajtó

A004





Cavity Flooding External cooling of the reactor pressure vessel

Ventillation line

Isolation and radiation shield (in lower position)

water injection line





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) 0,119 0,015 0,015

Late Containment Failure (LCF. LCFS) 0,000 0,017 0,023

Late Containment Leak (LCL. LCLS) 0,395 0,005 0,185

Isolation Isolation Failure 0,030 0,030 0,030

Failure Filtered Vent Remains Open (FVO) 0,000 0,020 0,020

Total of failure states 0,537 0,090 0,276

Filtered Vent (FV) 0,000 0,416 0,251

Intact (I, IS) 0,254 0,265 0,264

Partly Damaged Core 0,199 0,199 0,199

Controlled release states 0,453 0,900 0,723

Remaining 1% of the PDS's 0,01 0,01 0,01

Total 1,00 1,00 1,00





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

Documents

LEVEL 2 PROBABILISTIC SAFETY ASSESSMENT MODEL FOR PAKS NUCLEAR POWER PLANT