Transcript スライド タイトルなし - Nuclear Energy Agency

Level 2 PSA Results Considered Mitigation Accident
Managements for a Japanese PWR Plant
O. KAWABATA
Safety Evaluation and Analysis Group
Safety Analysis and Evaluation Division,
Japan Nuclear Energy Safety Organization (JNES),
Kamiya-cho MT BLDG,
4-3-20, Toranomon, Minato-ku, Tokyo,
105-0001 Japan
Presented in Level2 PSA Workshop of OECD/CSNI Wgrisk , March 30 in 2004.
Contents
1. Major safety features of the PWR plant
2. Accident Management Counter-measures
3. Point-Estimate of Containment Failure
Frequency
4. Effectiveness of AMs to Prevent CV failure
5. Source Terms Analysis
6. Conclusion
1. Major safety features of the PWR plant
Safety Functions
Safety Functions
(1) High pressure injection system
 High pressure injection pump
(2) Low pressure injection system
(Number of RHR pumps and heat exchanger)
Reactor cooling
2,2
4
(4) Auxiliary feed water system
 Motor driven pump
 Turbine driven pump
2
1
(6) Number of PORVs
(1) Containment spray pumps and heat exchanger
Supporting
system
2
(3) Accumulator
(5) ECCS re-circulation mode change
CV heat removal
4 Loop Dry
(2) Containment Type
Automatic
2
2, 2
Pre-stressed Concrete
Containment Vessel
(3) Containment Free Volume
74,000m3
(4) Containment Design Pressure
0.49MPa
Component cooling water system
Train isolation with motoroperated valves
1
2. Accident Management Counter-measures
Water Injection into CV
Containment Vessel
Containment cooling
by Natural Convection
CV Spray Ring
Raw
Water
Tank
Fire P
Cooling Coil
CCWS
Forced Depressurization
of the RCS
Cooling down
MSRV
Turbine
SG
Pressurizer
Relief Valve
Alternative
Recirculation
Refueling
Water
Storage
Pit
CV Spray P
Recirculation
RV
RHRP
HPIP
2
Mitigation Accident Management Measures
Containment
Vessel
Opening Pressurizer
PORVs
Pressurizer
Reactor vessel
Pressurizer
Forced depression of primary
system
To decrease the primary system
pressure by opening pressurizer relief
valve in order to prevent scattering of
melted debris from reactor cavity
region even when the bottom of the
reactor vessel is melted through.
Relief Tank
Re-circulation
Sump
3
Fire water injection into the
containment vessel
To inject fire water into the
containment vessel to prevent the
penetration failure of containment
base-mat concrete even when the
bottom of the reactor vessel is melted
and the debris goes down the reactor
cavity region.
4
Containment
Vessel
Spray Ring
Steam
Generator
Non-safety
CV Fan
Cooler
C/C
CCW
CCW
Pressurizer
Reactor vessel
Natural convection containment
vessel cooling
To supply the pressurized component
cooling water to the air-conditioning
cooler, and to cool the containment
vessel atmosphere by the natural
convection at the loss of CV spray
system.
Re-circulation
Sump
5
3. Point-Estimate of Containment Failure Frequency
Core Damage
Sequence
Typical
Sequence
PDS
1
- Similarity of
Accident
Progression
Containment
Failure
Sequence
Failure Modes
Quantification
CET
2
4
- Heading
- Failure Mode
- Branch
Probability
MELCOR Calculation
3
6
Core Damage Frequencies without AM (4 Loop Plant)
P2.2%
TE’0.9% SE0.4%
AE0.2%
AL1.9%
SE’ 2.3%
AEC3.3%
TE0.1%
V3.8%
SL5%
SLC(37%)
G7%
SE”(7%)
ALC(15%)
TEC(17%
)
AE :Large&Medium LOCA/Early Core Damage
/Without CV Spray
AEC:Large&Medium LOCA/Early Core Damage
/With CV Spray
AL :Large&Medium LOCA/Late Core Damage
/Without CV Spray
ALC:Large&Medium LOCA/Late Core damage
/With CV Spray
SE :Small LOCA/ Early Core Damage
/Without CV Spray
SE’ :SBO/RC Pump Seal LOCA
SE” :CCWS Failure/RC Pump Seal LOCA
SEC:Small LOCA/Early Core Damage
/With CV Spray
SL :Small LOCA/Late Core Damage
/Without CV Spray
SLC:Small LOCA/Late Core Damage
/With CV Spray
TE :Transient/Early Core Damage
/Without CV Spray
TE’ :SBO
TEC:Transient/Early Core Damage/With CV Spray
G :SGTR
P :Containment Failure before Core Damage
V :Interface-System LOCA
Core Damage Frequency 2.3E-07/RY
7
CET Heading Probability
Branching probability of headings regarding to physico-chemical
phenomena of severe accident sequence
Headings
Probability
Determined by PDS definition
1.0
Uncertainty phenomena
0.5
Unlikely
0.1
Remotely possible with uncertainty
1E-02
Remotely possible with little uncertainty
1E-04
8
Branching probabilities of AM Headings (4loop Plant)
Equipment failure and human factor regarding to each AM were considered.
AM
measures
Forced
depress
Fire water
injection
Natural
Convection
PDS
Available
time
Branching
probabilities
SE’,TE’
-
1.0
SEC,SE,SE”,SLC,SL
10 min.
0.21 to 0.37*
TEC,TE
10 min.
0.40 to 0.50*
AEC,AE,ALC,AL
-
1.0
SE’,TE’
30 min.
0.55*
Other PDSs excluding P,G,V
30 min.
0.49*
SE’,TE’
60 min.<
4.0E-02
SE”
-
1.0
Other PDSs excluding P,G,V
60 min.<
1.3E-02
*: These values are included failure probabilities of Level 1 PSA.
9
Containment Event Tree for a long term
period after the reactor vessel failure
10
4. Effectiveness of AMs to Prevent CV failure
Reduction=1/5
without AM
Effect of Natural
Convection
1.0E-06
6.9E-08
Effect of Forced DepressUrization & CV Injection
1.0E-07
1.5E-08
1.0E-08
Effect of CV
Injection
1.0E-09
Frequency (/RY)
with AM
1.0E-10
Effect of Cool down
& re-circulation
1.0E-11
1.0E-12
1.0E-13
1.0E-14
1.0E-15
1.
2.
3.
4.
To
O
C
IS
L
ta
l
A
R
C
H
G
T
S
es
s
E
x-
V
on
Containment Failure Mode
C
or
ec
D
X
lS
te
su
re
s
rp
O
ve
ro
yd
H
cr
e
re
ur
n
ge
n
B
Fa
ilu
on
Is
ol
at
i
In
-V
es
s
el
S
X
re
1.0E-16
The CFF becomes 1.5x10-8 / ry after AM measures implementation, and it decreases
to one-fifth of that before AM measures implementation.
The frequency of the overpressure sequence is decreased to one-twelfth by the
natural convection containment vessel cooling.
The frequency of the sequence contained concrete corrosion was decreased to
one-twentieth by fire water injection into the containment vessel.
While the water injection into the containment vessel increases the possibility of
11
steam explosion and hydrogen detonation with adverse effects.
Containment Failure Fraction with AM (4Loop Plant)
Isolation failure 3%
Melt through 1%
SGTR 15%
Interface LOCA 58%
Over-pressure 14%
Total Failure Frequency 1.5E-08/RY
The largest contribution to CFF after AM measures implementation is loss of
isolation function, and the second contribution is overpressure by steam.
12
5. Source Terms Analysis
1. A hydraulic impact analysis for an IS-LOCA sequence
(Short Term MELCOR Analysis)
CV
RV
Outlet
Pipe
RHR
Relief
Valve
A
Atmosphere
B
Isolation
Valve
Branch Pipe
RHR
Relief
Valve
E
D
Check
Atmosphere
Valve RHR
Cooler
C
RHR
Relief
RHR
Valve
Pump
RHR
Cooler
Check
Valve
Inside CV
Check
Valve
LP Safety
Injection Pipe
Tie-line
Check
Valve
LP Safety
Injection Pipe
Inside CV
The containment isolation valve of the RHR pump suction side was assumed
to be opened with 0.1 seconds.
13
Pressure Transient in the RHR piping during an IS-LOCA sequence
35
Node A
Node B
Node C
Node D
Node E
Pressure (MPa)
30
25
20
15
10
5
0
0
0.2
0.4
0.6
0.8
1
Time(s)
This figure shows the pressure transient in RHR piping when three RHR
relief valves do not open.
The maximum pressure in the RHR system piping obtained the pressure
of about 32MPa.
14
2. A structural dynamic response analysis of RHR piping
A structural dynamic response phenomenon of the RHR piping with
AUTODYN-2D code was analyzed by using the pressure transients
calculated with MELCOR code.
0.20
0.20
0.15
Strain (-)
Strain (-)
0.15
S1
S2
S3
S4
S5
S6
0.10
0.10
0.05
0.05
0.00
0.0
0.00
0.2
0.4
0.6
0.8
1.0
Time(s)
Transients of equivalent plastic strain of
some points in the RHR pipe
-0.05
0.0
20.0
40.0
60.0
80.0 100.0 120.0
Distance from RCS (m)
140.0
160.0
Distance distribution of final equivalent
plastic strain for the RHR pipe
(1) The plastic strain of any position is increasing to the period from 100
msec to 400 msec.
(2) The maximum plastic strain of about 0.16 was obtained at the exit
piping of the RHR cooler.
(3) Since the fracture strain of 304 stainless steel piping is 0.19, the pipes
in the residual heat removal system are unlikely to rupture.
15
3. Source term analysis
Assuming the fail close of three RHR relief valves as an initiating event, a
source term analysis was performed with the MELCOR code for a case of
the pipe break at the RHR heat exchanger outlet side with about 2 inches
Refueling Storage Pit
diameter piping.
High Pressure Injection
Atmosphere
RHR pipe
in inside CV
Release
opening
Low Pressure
Injection
CV
isolation
valve
RHR Coolers
Low Pressure
Injection
RHR pipe
in outside CV
Auxiliary
Building
LOCA opening
MELCOR Control Volumes of RHR system
for a IS-LOCA sequence
(Long Term Analysis)
16
FP Release Fraction to Environment during the IS-LOCA sequence
Event chronology for the IS-LOCA sequence
Events
Time
RHR pipe failure
0.0 sec
FP Release Fraction for 2 inches IS-LOCA
1
Xe
CsI
Te
Sr
Initiation of Safety
Injection Signal
40 sec
Re-circulation Failure of
ECCCS
7.2 hours
Core Uncoverd
8.9 hours
Failure of Fuel Cladding
9.2 hours
Failure of Core Support
Plate
10.5 hours
Failure of Reactor Vessel
14.4 hours
Release Fraction (-)
0.8
0.6
0.4
Failure of
Fuel
Cladding
0.2
0
0
10
20
Time(h)
30
40
The CsI released to the environment during 40 hours from an accident
initiation was obtained to be 0.48 of the inventory.
17
6. Conclusion
Containment Failure Frequency Evaluation
(1) Containment failure frequency in the case without AMs was reduced
about 1/5 by implementation of AMs.
(2) Dominant sequence in the AM case led to large FP release :
Containment Bypass sequence caused by a pipe break of RHR
system by multiple failure of isolation valves (IS-LOCA), has the
largest frequency for the reference PWR.
Source Terms Analysis
(1) The calculation results to confirm integrity of piping for the IS-LOCA
sequence indicated that the pipes in the residual heat removal system
are unlikely to rupture.
(2) But a source term analysis has been performed with MELCOR
code by a case of the pipe break on the RHR heat exchanger outlet
side. The CsI released to the environment during 40 hours from an
accident initiation was obtained to be 0.48 of the inventory.
18