Overview of ABWR Safety Features
Transcript of Overview of ABWR Safety Features
Copyright 2013 GE Hitachi Nuclear Energy International, LLC - All rights reserved
Overview of ABWR Safety Features
INPRO Dialogue Forum
November 19-23, 2013
J. Alan Beard Principal Engineer
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Wilmington, NC USA
•Nuclear Power Plants: ABWR,
ESBWR and PRISM •Nuclear Services
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Enrichment
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Generation Technology
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GE Hitachi nuclear alliance and businesses
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Vallecitos – USA
Dresden 1 – USA
Laguna Verde - Mexico
Tarapur 1&2 – India
Dodewaard - Netherlands KKM - Switzerland
Garigliano - Italy
Santa María de Garoña - Spain Lungmen - Taiwan
K6/K7 - Japan
KRB - Germany
BWR legacy around the world
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Recent project experience
Images copyright TEPCO, Hokuriku Electric Power, Chugoku Electric Power, and J-Power; Provided by Hitachi GE Nuclear Energy
Continuously building for 58 years
Hamaoka-5 ABWR
Kashiwazaki-Kariwa 6/7 ABWR
Shika-2 ABWR
COD 1996/1997
COD 2006
COD 2005
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Current project status
Under Construction
Ohma ABWR
Shimane-3 ABWR
Lungmen-1/2 ABWR
Under Construction
Under Construction
Images copyright TEPCO, Hokuriku Electric Power, Chugoku Electric Power, and J-Power; Provided by Hitachi GE Nuclear Energy
Ohma 1 • 38% complete
Shimane 3 • 94% complete
• Approaching fuel load
Lungmen 1&2 • 94% complete
• Startup and Pre-Op testing
Continuously building for 58 years
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GE Hitachi’s new reactor portfolio
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ABWR
PRISM ESBWR
Operational Gen III
technology
• Lowest core damage frequency
of any Generation III reactor
• Extensive operational
experience since 1996
• Licensed in US, Taiwan & Japan
Evolutionary Gen III+
technology
• Lowest core damage frequency
of any Generation III+ reactor
• Passive cooling for >7 days w/o
AC power or operator action
• Lowest projected operations,
maintenance and staffing costs1
• 25% fewer pumps, valves and
motors than active safety plants
Revolutionary technology
with a rich, 40-year heritage
• Passive air-cooling; no operator
or mechanical actions needed
• The answer to the used fuel
dilemma - can reduce nuclear
waste to ~300-year radiotoxicity2
while generating electricity
1 Claims based on the U.S. DOE commissioned ‘Study of Construction Technologies and Schedules, O&M Staffing and Cost, and Decommissioning Costs and Funding
Requirements for Advanced Reactor Designs’ and an ESBWR staffing study
performed by a leading independent firm 2 To reach the same level of radiotoxicity as natural uranium
Copyright 2013 GE Hitachi Nuclear Energy International, LLC - All rights reserved
ABWR Basic Parameters
3,926 Megawatt core thermal power
~1,360 MWe gross
• Nominal summer rating
Reactor Recirculation with Internal Pumps
• External recirculation piping eliminated
Active safety systems
• 72 hours automated capability
Operating cycle length of 12 to 24 months
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Dresden 1 KRB
Oyster Creek Dresden 2
ABWR ESBWR
BWR Evolution
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Containment Evolution
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ABWR Overall Flowchart
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ECCS Systems Evolution
HPCI
LPCS
LPCI
LPCI
LPCS
LPCI
LPCI
LPCI/RHR
LPCI
High pressure capacity
Low Pressure capacity
No. of large pipes below core
Peak clad temperature
(App K), °F
Operator action
5000
42000
12
1600
Required
LPCI/RHR
LPCS
1900
29000
12
1100
Required
HPCS
Typical BWR/5 BWR/6
Typical BWR/4
HPCF
LPCF/RHR
HPCF
LPCF/RHR
RCIC LPCF/RHR
ABWR
2800
19000
0
No uncovery
Automated >72 hours
RHR RHR
ADS ADS
ADS
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Emergency Core Cooling System
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Emergency Core Cooling
Core is always under water
ADS
HPCF
LPCF/RHR
RCIC
LPCF/RHR
HPCF
LPCF/RHR
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ABWR Reactor Core Isolation Cooling (RCIC)
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ABWR High Pressure Core Flooder (HPCF)
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ABWR Residual Heat Removal (RHR)
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Automatic Depressurization System (ADS)
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ABWR Key Design Features – Onsite AC Power
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Severe Accident Features
ABWR passive features which
mitigate severe accidents:
• Inerted Containment
• Lower Drywell flood
capability
• Lower Drywell special
concrete & sump protection
• Suppression pool - fission products scrubbing &
retention
• Containment overpressure
protection
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AC Independent Water Addition (ACIWA)
Hard piped fire water connections
Seismic Category 1
On-site fire protection system connection
Truck connection
• Ties to RHR C
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References
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ABWR
Design Certification Document, Rev. 4: http://www.nrc.gov/reactors/new-reactors/design-cert/abwr.html#dcd
Safety Evaluation Report, NUREG-1503 and NUREG-1503 Supplement 1: http://pbadupws.nrc.gov/docs/ML0806/ML080670592.html
Design Certification Rule, 10CFR52 Appendix A: http://www.nrc.gov/reading-rm/doc-collections/cfr/part052/part052-appa.html
Revision 5 of the ABWR DCD can be found here: http://pbadupws.nrc.gov/docs/ML1100/ML110040323.html
Reference BWR (Grand Gulf NPP) NUREG-1150
http://www.nrc.gov/reading-rm/doc-collections/nuregs/staff/sr1150/