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Development of the Methodologies for Evaluating Severe Accident
Management
Lim Joong Taek
Department of human resource development team
December 13, 2016
IAEA Technical Meeting on the Verification and Validation of Severe Accident Management Guidelines
December 12-14, 2016 IAEA Headquarters, Vienna, Austria
Contents
Introduction
Theoretical Background
Methodology
Preliminary Results
Concluding Remarks & Future Work
2
Nuclear Power Generation in Korea
25 Power Reactors Operation • 21 PWRs and 4 PHWRs
• Installed capacity of 23.1 GWe
• 30 ~ 35% share of electricity supply
3 PWRs(APR1400) under Construction
4 PWRs(APR1400) under Planning
The Most Economical, Reliable & Semi-domestic Electricity Source in Korea
Operation Construction Planning
Hanbit 6 units
Hanul 6 units
Shin-Hanul 4 units
Wolsong 4 units
Shin-Wolsung 2 units
Kori 4 units
Shin-Kori 6 units
4
Domestic severe accident legislation
Amendment of Nuclear safety act • Purpose
Severe accident management performed by administrative order with weak legal basis
Clarification of responsibility for accident management and regulatory requirements in Nuclear Safety Act
Even in case of a severe accident, to minimize the release of radioactive materials and to restore the NPP to a safe condition through accident management program
To introduce accident management program as a licensing document
• Application New NPP : AMP is required to obtain operating license
NPP in operation : AMP is required to submit within 3 years(June 23, 2019)
5
Purpose
The new nuclear safety act
Purpose • Evaluate the effectiveness of
the severe accident management strategies using probabilistic manner
• Combine PSA, AMP, and severe accident research
6
Accident Management
Program
Severe Accident Research
Integrated Methodology
Probabilistic Safety
Analysis
Integrated ROAAM • ROAAM + level 2 PSA
• PSA is a tool for determining relevant sequences that must be mitigated in the SAM approach
• ROAAM increases the credibility and transparency of the level 2 PSA study providing a framework for modeling and quantifying complex physical phenomena.
• Application Westinghouse AP600 : passive LWR design
SAMEM • Performed the development and
demonstration of integrated models for the evaluation of severe accident management strategies using PSA and ROAAM
8
Background
Probabilistic framework for containment failure due to hydrogen combustion
- Application of the risk oriented accident analysis methodology (ROAAM)to severe accident management in the AP600 advanced light water reactor : J.H.Scobel, et al(1998) - The development and demonstration of integrated models for the evaluation of sever accident management strategies-SAMEM : M.L. Ang(2001)
10
Level 1 PSA Level 2 PSA
Fission Product
Noble Gas
I
Cs
Te
Sb
Sr
Ru
Ba
La
Ce
Source Term Category
STC1 (SGTR)
STC2 (ISLOCA)
STC3 (NOTISOCS)
STC4 (NOTISONOCS)
STC5 (CFBRB)
STC6 (MELTSTOP)
STC7 (NOCF)
STC8 (BMT)
STC9 (ECF1)
STC10 (ECF2)
STC11 (ECF3)
STC12 (ECF4)
STC13 (LCF1)
STC14 (LCF2)
STC15 (LCF3)
STC16 (LCF4)
STC17 (LCF5)
Containment Failure Mode
BYPASS
NOTISO
NOCF
BMT
ECF
LCF
Severe Accident
Phenomena
Molten Core- Concrete
Interaction
In-Vessel Steam Explosion
Ex-Vessel Steam Explosion
Hydrogen Generation &
Reaction
High Pressure Melt Ejection
Direct Containment
Heating
Containment Failure Mitigation
CIS Containment Isolation before Core Damage
SDR Sudden Depressurization
for mitigation
RACV Power Recovery before Reactor Vessel Failure
RACC Power Recovery before
Containment Failure
INJ Coolant Injection to Reactor Vessel
CFS Reactor Cavity Flooding
HMSI Hydrogen Control System
Operation
CHR Containment Heat Removal
Initiating Event
LOCA
Large LOCA
Medium LOCA
Small LOCA
SGTR
ISLOCA
RVR
T r an s i en t
LSSB-IN/OUT
LOFW
LOCV
LOIA
PLOCCW
TLOCCW
LOKV(4.16kV)
LODC(125V)
LOOP
SBO
GTRN
ATWS
Core Damage Prevention
EVENT Tree
Core Damage Frequency
Containment Failure Probability
Plant Damage State Frequency
Fission Product Release Frequency
Korean Level 2 PSA Process
Database the level 2 PSA information
Transfer to database • Accident progress
Core Damage Prevention function
Severe accident mitigation function
• PDS group of Accident Sequences
• Screening PDS ET Containment failure sequence
Frequency : Below 10-14 /RY
• Success/failure of the functions Success : (AAC)
Failure : AAC
11
Plant Damage State Event Tree
12
Database the level 2 PSA information
Source term release fraction • The mass of the initial fission
product in a reactor core and the mass fraction of the fission product released into the environment
• Each PDS has its own 17 STC quantification values and the probabilities by containment failure modes
• 10 fission products Noble gas, I, Cs, Te, Sb, Sr, Ru,
Ba, La, Ce
Source Term Category Release fraction
13
Initiating Event Core Damage
Prevention Functions Severe Accident
Mitigation Function Source Term
Database the level 2 PSA information
Preliminary application
APR1400 • Thermal power : 4,000 MWth
• Hot-leg temperature
: 615 ℉(323.9 ℃)
• Severe accident respond facilities Cavity flooding system
Passive auto-catalytic recombiner
ECSBS
Selection of initiating event • Based on CDF contribution
14
Initial Event
Co
re D
amag
e Fr
equ
en
cy SBO
SLOCA
Risk information to display
Severe Accident Prevention / Mitigation • Core Damage Frequency
• Conditional Containment Failure Probability
Containment failure / Source term release • Containment Failure Frequency
• Source Term Release Fraction
15
𝐶𝐹𝐹 = 𝐹𝑅𝐸𝑄 × 𝐶𝐹𝑃
𝐶𝐶𝐹𝑃 =𝐶𝐹𝐹
𝐶𝐷𝐹=
𝐹𝑅𝐸𝑄
𝐶𝐷𝐹 × 𝐶𝐹𝑃
* FREQ : Frequency of PDS ET end state
Database of the risk information
17
STATION
BLACKOU
T
AAC
UNAVAIL
ABLE
DELIVER
AUX.
FEEDWATER
(TD PUMPS)
RCP
SEAL
FAILURE
RECOVER
AC
POWER
(IN
1HRS)
DEL.
AFW
AND
REM.
STEAM
(MSSV)
DEL.
AFW
AND
REM.
STEAM
(ADV)
RECOVER
AC
POWER
LATE
REFUEL
AAC
STORAGE
TANK
SHUT
DOWN
COOLING
MAINTAIN
SECONDARY
HEAT
REMOVAL
RECOVER
AC
POWER
AND
REFUELING
AAC
SAFETY
DEPRESSU
RIZATION
SAFETY
INJECTION
FOR
FEED
(2/4)
CONTAIN
MENT
HEAT
REMOVAL
(COOL
IRWST)
CONTAIN
MENT
HEAT
REMOVAL
(CS
SPRAY)
CONTAIN
MENT
ISOLATIO
N
SYSTEM
RAPID
DEPRESSU
RIZATION
RECOVER
AC
POWER
PRIOR
TO
VESSEL
RECOVER
AC
POWER
PRIOR
TO
CONTAIN
MENT
IN-VESSEL
INJECTIO
N
CAVITY
FLOODIN
G
SYSTEM
HYDROGEN
CONTROL
SYSTEM
IGNITORS
CONTAIN
MENT
HEAT
REMOVA
L
FREQ
(Accident
Sequence
Frequency
)
CLASS
(PDS No.)I Cs
83 SBO (AAC) - - - (SHR1) - RACL-1 REFUEL - (MSHR) RACL-2 - - - - 3.39E-10 (CIS) - (RACV) - (INJ) (CFS) (HMSI) (CHR1) 1.31E-10 113 3.57E-02 1.21E-11 2.01E-03 9.68E-04
85 SBO (AAC) - - - (SHR1) - RACL-1 REFUEL - (MSHR) RACL-2 - - - - 3.39E-10 (CIS) - (RACV) - (INJ) (CFS) HMSI (CHR1) 3.54E-11 113 9.61E-03 3.26E-12 2.01E-03 9.68E-04
134 SBO (AAC) - - - (SHR1) - RACL-1 REFUEL - MSHR - - - - - 6.38E-10 (CIS) SDR (RACV) - (INJ) (CFS) (HMSI) (CHR1) 6.27E-11 113 9.04E-03 5.77E-12 2.01E-03 9.68E-04
136 SBO (AAC) - - - (SHR1) - RACL-1 REFUEL - MSHR - - - - - 6.38E-10 (CIS) SDR (RACV) - (INJ) (CFS) HMSI (CHR1) 1.40E-11 113 2.01E-03 1.28E-12 2.01E-03 9.68E-04
261 SBO AAC (AFW-T) (RSF) - - - RACL-1 - - - - - - - - 6.23E-07 (CIS) - (RACV) - (INJ) - (HMSI) (CHR1) 3.16E-07 113 4.67E-02 2.91E-08 2.01E-03 9.68E-04
263 SBO AAC (AFW-T) (RSF) - - - RACL-1 - - - - - - - - 6.23E-07 (CIS) - (RACV) - (INJ) - HMSI (CHR1) 1.15E-07 113 1.70E-02 1.06E-08 2.01E-03 9.68E-04
418 SBO AAC AFW-T (RSF) RACE - - - - - - - - - - - 6.88E-09 (CIS) SDR (RACV) - (INJ) (CFS) (HMSI) (CHR1) 5.66E-10 113 7.57E-03 5.21E-11 2.01E-03 9.68E-04
420 SBO AAC AFW-T (RSF) RACE - - - - - - - - - - - 6.88E-09 (CIS) SDR (RACV) - (INJ) (CFS) HMSI (CHR1) 1.46E-10 113 1.95E-03 1.34E-11 2.01E-03 9.68E-04
434 SBO AAC AFW-T (RSF) RACE - - - - - - - - - - - 6.88E-09 (CIS) SDR RACV (RACC) (INJ) (CFS) (HMSI) (CHR1) 2.77E-10 113 3.70E-03 2.54E-11 2.01E-03 9.68E-04
436 SBO AAC AFW-T (RSF) RACE - - - - - - - - - - - 6.88E-09 (CIS) SDR RACV (RACC) (INJ) (CFS) HMSI (CHR1) 6.21E-11 113 8.31E-04 5.72E-12 2.01E-03 9.68E-04
CCFP
(=CFF/CDF)
CFF
(=FREQ*CFP)
Source TermContainment Failure MitigationCore Damage Prevention
No.
CDF
(Level1
Accident
Sequence
CDF)
CDF
Frequency of PDS ET end state
PDS Group
CCFP
CFF
Distribution chart
18
1.00E-12
1.00E-11
1.00E-10
1.00E-09
1.00E-08
1.00E-07
1.00E-06
1.00E-05
1.00E-07 1.00E-06 1.00E-05 1.00E-04 1.00E-03 1.00E-02 1.00E-01 1.00E+00
Co
re D
amag
e Fr
equ
ency
Conditional Containment Failrue Probability
SLOCA Accident Sequence
CDF Vs CCFP • Core Damage Frequency
• Conditional Containment Failure Probability
PDS 18
PDS 19
PDS 40PDS 41
PDS 76
PDS 77
PDS 92
PDS 93
PDS 4
PDS 18
PDS 19
PDS 40
PDS 41PDS 76
PDS 77
PDS 92
PDS 93
PDS 4
1.00E-12
1.00E-11
1.00E-10
1.00E-09
1.00E-08
1.00E-07
1.00E-06
1.00E-05
1.00E-07 1.00E-06 1.00E-05 1.00E-04 1.00E-03 1.00E-02 1.00E-01 1.00E+00
Co
re D
am
ag
e F
req
ue
ncy
Conditional Containment Failrue Probability
CHR1 Failure in SLOCA
19
1.00E-07
1.00E-06
1.00E-05
1.00E-04
1.00E-03
1.00E-02
1.00E-01
1.00E+00
1.00E-16 1.00E-15 1.00E-14 1.00E-13 1.00E-12 1.00E-11 1.00E-10 1.00E-09 1.00E-08 1.00E-07
방
출
분
율
(
S
T)
원자로건물손상빈도 (CFF)Containment Failure Frequency
Sou
rce
Term
Rel
ease
Fra
ctio
n
1.00E-12
1.00E-11
1.00E-10
1.00E-09
1.00E-08
1.00E-07
1.00E-06
1.00E-050
0.001
0.002
0.003
0.004
0.005
0.006
0.007
Co
re D
am
ag
e F
req
ue
ncy
So
urc
e T
erm
Re
lea
se F
ract
ion
Containment Failure Probability
CCFP-CDF-ST
Source term release fraction • CFF Vs RF (2D)
• CCFP Vs CDF Vs RF (3D)
Distribution chart
Insight from this works
Effective evaluation for SAMP • Legislation
• Verify the impact of mitigation success and failure
Prioritization of severe accident management strategies • Selection of the most important mitigation functions
• Application Improvement of the equipment
Modification of Technical specification
Personnel training
21
Further Work
Concept for evaluating severe accident
Methodology for DET/CET branch probability calculation
Uncertainty analysis of CET using computational code
Level 2 PSA re-quantification • The new index for measuring the effectiveness of AMP
22
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