Comprehensive Safety Assessment for Comprehensive Safety Assessment for Kashiwazaki Kashiwazaki Kariwa Kariwa NPS NPS International Experts International Experts ’ ’ Meeting on Reactor and Spent Meeting on Reactor and Spent Fuel Safety in the Light of the Accident at the Fuel Safety in the Light of the Accident at the Fukushima Daiichi Nuclear Power Plant Fukushima Daiichi Nuclear Power Plant IAEA, Vienna IAEA, Vienna Mar. 19 Mar. 19 – – 22, 2012 22, 2012 Hideki Masui Hideki Masui Seismic Research Manager Seismic Research Manager Tokyo Electric Power Company Tokyo Electric Power Company
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Comprehensive Safety Assessment for Comprehensive Safety Assessment for KashiwazakiKashiwazaki KariwaKariwa NPSNPS
International ExpertsInternational Experts’’ Meeting on Reactor and Spent Meeting on Reactor and Spent Fuel Safety in the Light of the Accident at the Fuel Safety in the Light of the Accident at the
Fukushima Daiichi Nuclear Power PlantFukushima Daiichi Nuclear Power PlantIAEA, ViennaIAEA, Vienna
Mar. 19 Mar. 19 –– 22, 201222, 2012Hideki MasuiHideki Masui
Seismic Research ManagerSeismic Research ManagerTokyo Electric Power CompanyTokyo Electric Power Company
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Contents
1. Background and Concept of Comprehensive Safety Assessment
2. Methodology and Result of Assessment
3. Further Additional Safety Measures in Light of Fukushima Daiichi Accident
4. Continuous Improvement
3
1. Background and Concept of Comprehensive Safety Assessment
4
Background
Minister of Economy Trade and Industry and other two Ministers laid out a plan for assessmentJul.11, 2011
Reports of KK 1/7 were re-submitted to NISA after correction of editorial errorsMar.12, 2012
IAEA mission visited NISA to review NISA’s approach to assessment
Jan.23-31, 2012
Reports of Primary Assessment of Kashiwazaki Kariwa 1/7 were submitted to NISAJan.16 2012
NISA issued direction to utilities for implementing comprehensive safety assessmentJul.22, 2011
Chair of NSC issued request to Minister of Economy Trade and Industry for comprehensive safety assessment (stress test)
Jul.6, 2011
Fukushima Daiichi AccidentMar. 2011
EventDate
5
Objectives of Stress Test
To identify and improve potential vulnerabilities of plant by clarifying
quantitative safety margins
To secure assurance and trust of general public and local residents
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Two-Step Approach for Stress Test
Same as primaryCombination of SBO
and LUHS(Other events may be
considered)
Earthquake, TsunamiCombination of quake
and tsunamiSBO, LUHS, SAM
Event
Realistic approachConservative approachMethod
Reactor, SFPReactor, SFPFacility
All operating plantPlants ready to start-up after outageTarget
SecondaryAssessment
PrimaryAssessment
Scope of this presentation
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Primary and Secondary Assessment
Material strength confirmed by testing etc.
Prim
ary S
econ
dary
Allowable stress limits in code and standard (σd)
Calculated stress for design base earthquake (σc)
Tota
l Saf
ety
Mar
gin
Seismic margin for primary assessment = σd/ σc
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TEPCO Nuclear Power StationsFukushima Daiichi NPS 4696MWeFukushima Daiichi NPS 4696MWe
・LOOP (<1.0)・LOPA・Loss of RCW・Loss of DC power (2.27)・Loss of I/C(1.70)
Before safety measures: 1.37After safety measures : 1.58
Cliff edge after safety measure(RPV anchor volt damage)
Detailed Assessment by ET
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Result of Stress Test (KK7, Earthquake)
SFPReactor
1.37(Loss of EDG)
1.47(RPV anchor volt damage)
After implemented
measure
1.37(Loss of EDG)
1.37(Loss of EDG)
Before implemented
measure
DBEGM=1209 gal
DBEGM: Design Basis Earthquake Ground Motion
Reactor: Due to implemented safety measure (power supply car etc), cliff edge will shift from LOOP to RPV damage
SFP: EDG failure can be covered by power supply car. However, margin of EDG is larger than that of MUW (alternative SFP injection) . Thereby, cliff edge stay unchanged.
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Result of Stress Test (KK1, Earthquake)
SFPReactor
1.45(R/B overhead crane
damage)
1.29(PCV stabilizer damage)
After implemented
measure
1.32(Loss of RCW)
1.29(PCV stabilizer damage)
Before implemented
measure
DBEGM=2300 gal
DBEGM: Design Basis Earthquake Ground Motion
Reactor: For initiating event “PCV damage,” no mitigation function is expected for primary assessment. Therefore, cliff edge stay unchanged
SFP: Due to implemented safety measure (power supply car etc), cliff edge will shift from “Loss of RCW” to “R/B overhead crane damage”
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Result of Stress Test (KK1/7, Tsunami)
Before
3.3m
After
5m
15mKK1 (Reactor, SFP)
Before
3.3m
After
12m
15m
Design Tsunami Height
11.7m 11.7m
KK7 (Reactor, SFP)
Before implementation of safety measure: Allowable tsunami height is conservatively estimated to be equal to site height.
After implementation of safety measure: Allowable tsunami height is the one under the assumption of which water seal measures were conducted.
Margin Margin
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Combination of Earthquake and Tsunami (KK7)
Allowable Tsunami height
1.47
15.0m
1.37
12.0m
Seismic Margin Expanded Safety Margin
0 3.3mDesign Tsunami Height
Site Height
Safety margin was expanded due to implementation of safety measures
Secondary Assessment
Secondary Assessment
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• Tolerance time is estimated after SBO or LUHS
• Tolerance time is determined by whichever shorter of the followings- Water supply time- Power supply time
• Conservative conditions are assumed:- All 7 reactors and SFPs become SBO or LUHS- No support is expected from outside
Assessment for SBO and LUHS
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Result of Stress Test (SBO, KK7, Reactor in Operation)
Water
Power
Time
Water
Power
Time
CSP 0.7d(Minimum water level was changed)
Battery 16h
Fresh water 4.9d Sea water 7d
Power supply car 94d
Before
After
Battery 16h(Actual Life, 8h on design basis)
CSP 10h
About 10 hours
About 12 days
Due to concern of salt damage, 7 days was assumed.
Determined by capacity of fuel
Reactor + SFP
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Result of Stress Test (SBO, KK7, Reactor Shutdown)
Water
Power
Time
Water
Power
Time
CSP 0.9d
Fresh water 5d Sea water 7d
Power supply car 99d
Before
After
About 5 hours (Time to water temperature reaches 100℃)
About 12 days
Due to concern of salt damage, 7 days was assumed.
Before safety measures, there was no specific procedures to inject water after SBO
SFP
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Before After
196 days
Result of Stress Test (LUHS, KK7 Reactor/ SFP)
1 day
Determined by capacity of CSP and purified water tank
Determined by capacity of on-site fuel storage for power supply car
Mobile Heat Exchanger
Power Supply Car
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Summary of Stress Test Result
• KK 1/7 have sufficient robustness against events beyond design basis through additional safety measures implemented after Fukushima Daiichi Accident
• In near future, secondary assessment will be conducted to identify and address potential vulnerabilities of plant.
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3.Further Additional Safety Measures in light of Fukushima Daiichi Accident
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④Support for emergencyresponse
Course of Event and Countermeasures<Event>Tsunami
SBO LUHS
Core damage
Hydrogen explosionRadiation Release
<Countermeasure>
①Protection from tsunami
②Prevention of core damage upon
Loss of AC/DC,LUHS
③Mitigation upon core damage
•Tsunami barrier•Tide barrier•Water proof sealing
Inundation of building
•Backup power •Heat removal•Diversified injection
•Roof venting•Reliable PCV venting
•Communication •Radiation Protection•Training
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(1)Protection from Tsunami
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Tsunami Barrier for Entire Site
Tsunami Barrier to prevent or deter the force of tsunami
Cable Laying Power Supply Car Connection of FP-line Balloon light
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Development of New Procedures
・RCIC manual startup・SRV operation by battery・Injection by FP system・Manual operation of venting valve
・Securing power source・Extension of RCIC life・Injection by FP system・Seawater injection by fire engine・Operation of roof vent
Contents
Extensive Damage Mitigation GuidelineResponses against unforeseeable cause-free events
Severe Accident Management Guideline for TsunamiOutline
EDMGTsunami SAMG
To make improvised action more organized one, 2 new procedures have been developed
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4.Continuous Improvement
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Continuous Improvement
• From Fukushima Daiichi Accident, we have learned a lot of lessons.
• Based upon these lessons, we have been implementing additional safety measures.
• We are committed to continuously improve safety of our plants through collecting new findings domestically and internationally.
• Followings are some examples of initiatives to improve safety.
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Enhanced Reliability for HP Injection
RP
V
PCV
FP MUW
MO
B系 A系
Isolation Condenser
1B
10A10B
3A
3B
PLR
MO
MO
MO
MO
MO
MO
MOMO
MO
1A
2B2A
4A
4B
MSIV
Valves of IC are of failure-close design (Power loss of piping rupture detection circuit closes corresponding isolation valve) Approaches to enhanced reliability for HP injection are considered
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Enhanced Reliability for PCV Venting
AO
MO Main Stack
AO
AO
AOIA
IA
Solenoid Valve
Solenoid Valve
Failure close or open?
Necessity of those valves?
Relocation for less exposure?
Necessity of rupture disk?
Bypass line to control release timing?
PCV
D/W
S/C
RPV
RPV
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Other Issues for Improving Safety
• Development of instrumentation resistant to severe accident condition
• Diversification of cooling function without AC power
• Reorganization of emergency response and operation staffing to deal with prolonged accident
• Establishment of transportation base for emergency material and staffing near plant
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We are committed to improve the safety of the nuclear power plant through comprehensive safety assessment (Stress Test) and lessons learned from Fukushima Daiichi Accident.
For TEPCO, stress test is an opportunity to identify and address weakness of plants.
Concluding Remarks
We are going to keep you informed of any update in our English website. http://www.tepco.co.jp/en/index-e.html