Title: The Fukushima earthquake tsunami and other recent external events that have challenged the design basis for commercial nuclear power plants
1The Fukushima earthquake tsunami and other
recent external events that have challenged the
design basis for commercial nuclear power plants
Peter Lobner 18 January 2012
2Agenda
- Definition of design basis
- BWR mitigating systems and system dependencies
- Earthquake
- Niigata Chuetsu-oki, Japan, magnitude 6.8,16 July
2007 - Great East Japan, magnitude 9.0, 11 March 2011
- Tsunami
- Fukushima Daiichi, 11 March 2011
- Fukushima Daiichi plant responses to the 11 March
2011 earthquake tsunami - International Response to the events at Fukushima
Daiichi - Recent external event challenges to U.S. NPPs
- Missouri River flooding, July August 2011
- Northern VA earthquake, magnitude 5.8, 23 August
2011 - Hurricanes
- Conclusions
3Definition of design basis
- Design bases that information which identifies
the specific functions to be performed by a
structure, system, or component (SSC) of a
facility, and the specific values or ranges of
values chosen for controlling parameters as
reference bounds for design. - These values may be
- (1) restraints derived from generally accepted
"state of the art" practices for achieving
functional goals, or - (2) requirements derived from analysis (based on
calculation and/or experiments) of the effects of
a postulated accident for which an SSC must meet
its functional goals.
Source 10CFR50.2
4Safety design basis and safety functions
- Safety design basis focuses on assuring that
nuclear power plant (NPP) safety functions
defined in 10CFR50 Appendix A, General Design
Criteria, can be accomplished when required to
protect the integrity of multiple fission product
barriers - Accomplish reactor shutdown (GDC 20, 29)
- Maintain primary system integrity (GDC 14, 15,
31) - Maintain reactor core cooling (GDC 33 37)
- Maintain containment integrity (GDC 16, 38-43,
50, 51, 52-57) - Maintain the cooling water heat transport path to
the ultimate heat sink (GDC 44-46) - Prevent an uncontrolled release of radioactive
material to the environment from fuel and waste
systems (GDC 60-64)
5Safety design basis for protection against severe
natural phenomena
- GDC 2, Design bases for protection against
natural phenomena. - SSCs important to safety shall be designed to
withstand the effects of natural phenomena such
as earthquakes, tornadoes, hurricanes, floods,
tsunami, and seiches without loss of capability
to perform their safety functions. - The design bases for these SSCs shall reflect
- Most severe historically reported for the site
and surrounding area, with sufficient margin - Combinations of the effects of normal and
accident conditions with the effects of the
natural phenomena, and - The importance of the safety functions to be
performed. - Generic Letter 88-20, Supplement 4, Individual
Plant Evaluation of External Events (IPEEE) for
Severe Accident Vulnerabilities - Licensees requested to perform analyses to
determine vulnerabilities to beyond-design-basis
external events and determine if any improvements
are needed. - SSCs were examined to estimate their
high-confidence-of-low-probability-of-failure
(HCLPF) level.
Source 10CFR50 Appendix A, General Design
Criterion 2
6The design basis is not static
- 10CFR50.54,Conditions of License, paragraph (f)
- The licensee shall at any time before expiration
of the license, upon request of the Commission,
submit, as specified in 50.4, written
statements, signed under oath or affirmation, to
enable the Commission to determine whether or not
the license should be modified, suspended, or
revoked. - Except for information sought to verify licensee
compliance with the current licensing basis for
that facility, the NRC must prepare the reason or
reasons for each information request prior to
issuance to ensure that the burden to be imposed
on respondents is justified in view of the
potential safety significance of the issue to be
addressed in the requested information. - Value / impact ratio used for prioritizing safety
issue resolution is determined using the
conversion factor of 2,000/person-rem, which was
approved by the Commission in September 1995. - Resolving generic safety issues may require
utilities to implement changes. For example - Station blackout rule (SBO) (10CFR50.63, 1988)
- Mark I containment hard vents (Generic Ltr 89-16,
1989) - Utilities may choose to improve NPP operating
capability. - Longer operating cycles between refueling
- Power increase
7Station blackout rule (SBO)
- 10CFR50.63 Loss of all alternating current
power, requires that each NPP be able to cope
with and recover from an SBO event of specified
duration - Cope means that the core is cooled and
appropriate containment integrity is maintained
in the event of a station blackout for the
specified duration. - Implemented by Regulatory Guide 1.155 and
industry document NUMARC 87-00. - 44 U.S. NPP implemented AC-independent solutions
- Batteries only
- Maximum coping duration 4 hours
- 60 U.S. NPPs implemented alternate AC power
sources. For example - Emergency diesel generator from an adjacent unit
- Gas turbine or other diesel generators, hydro
generator. - Coping duration 4 16 hours
- Non-electric driven pumps (steam, diesel) provide
important capabilities for operating cooling and
makeup systems during SBO.
8BWR mitigating systems and system dependencies
9Key mitigating systems available at the
Fukushima Daiichi units
10BWR Mark-I containment steel shell
11BWR Mark-I containment arrangement within the
reactor building
12BWR Mark I containment performance improvement
(CPI) program
- Resolution of Generic Safety Issue 157,
Containment Performance, resulted in significant
modifications to BWR Mark 1 containments. - All affected BWRs had in place emergency
procedures directing the operator to vent via the
non-pressure bearing Standby Gas Treatment System
(SGTS) ducting under certain circumstances
(primarily to avoid exceeding the primary
containment pressure limit). - A hard pipe vent path bypassing the SGTS and
capable of withstanding the anticipated severe
accident pressure loadings would eliminate the
problems with venting the containment wetwell
during a severe accident. - The vent isolation valves should be remotely
operable from the control room and should be
provided with a power supply independent of
normal or emergency AC power (i.e., operable
during SBO). - In Generic Letter 89-16 (1989), NRC requested
each licensee to provide cost estimates for
implementation of a hardened vent. - GE reports that US operators installed hardened
vents in their Mark I BWRs. - In 1992, Japan's Nuclear Safety Commission
rejected establishing a regulatory requirement
for a hardened wetwell vent for Mark 1 BWR
containments, leaving it to the NPP operators to
decide to install a hardened vent. - GE reports that Japanese operators, including
TEPCO, installed hardened vents in their Mark I
BWRs.
13BWR Mark-I containment refueling floor arrangement
14Isolation Condenser (IC) System
- System Dependencies
- Automatic start on reactor vessel high pressure
or low water level, or remote manual, - DC power to open the normally closed valve in the
condensate return line - AC power to operate normally open valves in the
steam supply and condensate return lines - Steam supply from main steam line to isolation
condenser - Periodic water supply to the secondary-side of
the isolation condenser to make up for
evaporation to the environment - Periodic makeup to the primary system to make up
for coolant shrinkage during cooldown
Source NUREG/CR-5640
15Reactor Core Isolation Cooling (RCIC) System
- System Dependencies
- Automatic start on reactor vessel low water
level, or remote manual, - DC power to open RCIC turbine steam supply
valves, injection valve, wetwell suction valves
(when needed) - Steam from main steam line
- Turbine exhaust path to wetwell and wetwell
pressure lt turbine backpressure trip setpoint. - Water supply from condensate storage tank or
wetwell. - Automatic pump suction realignment on CST low
level - Pump room cooling by service water
- No cooling for the pump itself.
Source NRC BWR Concepts Manual
16High-Pressure Coolant Injection (HPCI) System
- System Dependencies
- Automatic actuation on reactor vessel low water
level or drywell high pressure, or remote-manual - DC power to open HPCI turbine steam supply
valves, injection valve, wetwell suction valves
(when needed) and operate the aux lube oil pump
during startup - Steam from main steam line
- Turbine exhaust path to wetwell and wetwell
pressure lt turbine backpressure trip setpoint. - Water supply from condensate storage tank or
wetwell. - Automatic pump suction realignment on CST low
level - Pump room cooling by service water
- No cooling for the pump itself.
Source NRC BWR Concepts Manual
17Low-Pressure ECCS and Residual Heat Removal (RHR)
- System Dependencies
- Automatic pump actuation on reactor vessel low
water level or drywell high pressure, or
remote-manual - Automatic Depressurization System (ADS) actuation
on low vessel level high drywell level LP
ECCS pump running - AC power for LPCS and LPCI (RHR) pumps valves
- DC power to open ADS valves
- Water supply from wetwell.
- Pump room cooling by service water
- RHR pump cooling by service water
Source NRC BWR Concepts Manual
18Niigata Chuetsu-Oki Earthquake (NCOE), Japan,
magnitude 6.8, 16 July 2007
19Niigata Chuetsu-Oki Earthquake (NCOE), Japan,
magnitude 6.8, 16 July 2007
Source TEPCO
Source EQECAT Inc
20Kashiwazaki-Kariwa NPP
- Worlds largest nuclear power facility 7,965
MWe net from 7 BWR units. - U1 5 BWR, 1067 MWe
- U6 7 ABWR, 1315 MWe
- During NCOE
- 3 operating at rated power (U3, U4 U7)
- 1 starting up (U2)
- 3 shutdown for periodic inspection (U1, 5 U6)
- 16 km from NCOE epicenter.
- No tsunami.
21Kashiwazaki-Kariwa NPP
Source TEPCO
22NCOE observed seismic data
- The observed seismic accelerations largely
exceeded the design basis values.
Source TEPCO
23NPP response to NCOE (1/2)
- Units operating (Units 3, 4 7) and being
started up (Unit 2) automatically scrammed on
detection of large seismic acceleration. - Off-site power remained available during and
after NCOE. - Reactor vessel water level maintained in all
units. - Reactor cooldown and depressurization
accomplished. - Reactor coolant at all units cooled to below
100ºC. - Reactor pressure in each unit reduced to
atmospheric pressure - Stable cold shutdown condition achieved by 17
July. - In spite of significantly exceeding the original
seismic design basis, the safety-related
structures, systems and components at all seven
units demonstrated good performance and
accomplished their intended safety functions.
24NPP response to NCOE (2/2)
- No change in fission product concentration in
reactor coolant and spent fuel water, indicating
that fuel in all units was sound. - Minor releases of radioactive material
- Some water sloshed out of the Unit 6 spent fuel
pool. - Many containers of LLW overturned, some lids came
off. - Minor release via main stack detected on 17 July
at Unit 7. - Relatively minor physical damage, mainly to
non-safety-related items. - Mechanical anchorages, ducting to main stacks,
various water, oil air leaks - Structural wall embankment cracking
- Ground deformations, with potential to damage
underground tunnels pipeways and surface roads
drainage paths. - Transformer fire
25Improved understanding of site seismicity
- The NCOE seismic intensity exceeded the original
seismic design basis for all NPP units. - The NCOE seismic intensity also exceeded the
seismic intensity estimated from an empirical
evaluation of a magnitude 6.8 earthquake. - Japans newer (2006) seismic design guidelines
redefine active faults and the process for
defining a Standard Seismic Ground Motion (SSGM)
to be used in design. - Post-NCOE seismic study findings
- New and extended fault lines.
- Geologic structure amplifies seismic motion from
sea-side. - Differences between the Unit 1-4 and Unit 5-7
sites, which are 1 km apart.
Source TEPCO
26Standard seismic ground motion (SSGM) defined for
Kashiwazaki-Kariwa NPPs.
- Post-NCOE seismic hazard studies yielded the
largest values for ground motion ever considered
for a nuclear power plant site.
Source TEPCO
27Post-NCOE safety actions
- Install seismic reinforcements to tolerate
seismic motion of 1000 Gal (1.5 times NCOE max) - Add more pipe snubbers pipe supports
- Reinforce reactor building roof truss structure
- Reinforce reactor building overhead crane,
including derailment prevention - Reinforce refueling machinery, including
derailment prevention - Add vibration control device for stacks
- Perform facility integrity evaluation
- Confirm NCOE loads on each equipment was within
applicable elastic limits. - Perform equipment, system plant-level
functional inspections tests - EPRI supporting evaluation of hidden damage
- Improve the spent fuel storage pool structure to
prevent radioactive water overflow (from
seismic-induced sloshing) by Sep 2012.
28Re-start status
- May 2009 Unit 7 re-started (22 mos)
- August 2009 Unit 6 re-started (25 mos)
- May 2010 Unit 1 re-started (34 mos)
- November 2010 Unit 5 restarted (40 mos)
- Units 2 4 investigations, modifications
tests in-progress. Unit 3 likely to be next unit
restarted.
29Great East Japan Earthquake, magnitude 9.0, 11
March 2011
30Great East Japan Earthquake, magnitude 9.0, 11
March 2011, 1446 JST
Source USGS
Source EQECAT Inc
- An earthquake of this magnitude is
unprecedented in this region. - Megathrust rupture on the Japan Trench
subduction zone - Earthquake lasted about 2 -2.3 minutes
- 11 aftershocks on 11 March, ranging from 6.0 to
7.4.
31Fukushima Daiichi NPP
- One of Japans larger nuclear power facilities
4,696 MWe net from 6 BWR units. - U1 BWR 3
- U2 5 BWR 4
- U6 BWR 5
- During earthquake
- 3 operating at rated power (U1, 2 3)
- 3 shutdown for periodic inspection (U4, 5 6)
- 112 miles from epicenter.
- Design basis tsunami 18.8 (5.7m)
32FukushimaDaiichi SiteArrangement
Source INPO
33Observed seismic data at Fukushima Daiichi
- Design Basis Earthquake maximum acceleration
exceeded at Units 2, 3 and 5. -
- The power lines connecting the site to the
off-site transmission grid were damaged - during the earthquake, resulting in a loss of
all off-site power. - All reactor safety functions were successfully
performed after the - earthquake and all units were in a safe state
prior to the arrival of the tsunami.
34Tsunami following theGreat East Japan
Earthquake, 11 March 2011
35Tsunami timeline at Fukushima Daiichi
- 1527 First of seven tsunami waves arrived.
Height about 13 (4 m) was less than the design
basis tsunami and was mitigated by the
breakwater. - 1535 Second tsunami wave arrived. Height
unknown. Tidal gauge failed. - Five more tsunami waves. At least one of the
waves measured 46 49 (14 15 m) high based
on water level indications on the buildings. - Unit 1 4 site area inundated to a depth of 13
16 (4 5 m) above grade. - Grade level at the Unit 5 6 site area is 3 m
higher, so inundation there was less.
36Tsunami wave arrives at Fukushima Daiichi
37Tsunami wave arrives at Fukushima Daiichi
38Site inundation
39Site inundation
40Site inundation
41Tsunami effects on storage tank
42Fukushima Daiichi site inundation
Source IAEA
43Fukushima Daiichi Units 1 4 inundation
Source INPO
- Flooding resulted in common cause failure and
loss of the ability to perform key safety
functions - Intake structure, pumps, and flow paths to the
ultimate heat sink (the ocean) at Units 1 - 6. - Most main and safety-related AC and DC electric
power sources and distribution rooms / areas
needed to support active safety systems at Units
1 - 5. DC in Units 3, 5 6 survived.
44Fukushima Daiichi plant responses to the 11
March 2011 earthquake tsunami
45Decay heat reactor units 1, 2, 3
Source MIT
46Decay heat spent fuel pools
47Fuel response to severe accident progression
48Unit 1 sequence of events
Adapted from INPO
49Unit 1 sequence of events (continued)
Adapted from INPO
50Unit 1 Hydrogen Explosion, 12 March 2011
51Loss if lighting in the control room
Source TEPCO
52Unit 2 sequence of events
Adapted from INPO
53Unit 2 sequence of events (continued)
Adapted from INPO
54Unit 3 sequence of events
Adapted from INPO
55Unit 3 sequence of events (continued)
Adapted from INPO
56Unit 3 Hydrogen Explosion, 14 March 2011
57Unit 4 sequence of events
Adapted from INPO
58Unit 4 after hydrogen explosion
59Possible hydrogen leak path to Unit 4
Source INPO
60Units 1-4 before the tsunami explosion
61Units 1-4 after the explosion
62Unit 5 6 sequence of events
Source SECY-11-0093
63Severe accident response issues
- TEPCO confirmed that adverse conditions in the
drywell may have resulted in boiling of the
reference legs of the reactor vessel water level
instruments, causing indicated water level to be
higher than actual level. - TEPCO severe accident procedures provided
guidance for venting containment - If core damage has not occurred, vent at
containment maximum operating pressure 62.4
psig for U1, 55.1 psig for U2 U5 - If core damage has occurred, delay venting until
pressure approaches twice the maximum operating
pressure. - In Units 1, 2, and 3, the extended duration of
high temperature and pressure conditions inside
containment may have damaged the drywell head
seals, contributing to - Hydrogen leaks into the upper level of the
reactor building and the subsequent explosions,
and - Ground-level radiation releases
64Severe accident response issues
- Was there a re-criticality at Unit 2?
- While examining gases taken from the reactor,
short-lived fission product Xe-133 was detected
on 2 November 2011 - Boric acid water injected
- TEPCO general manager "Given the signs, it's
certain that fission is occurring." - The next day, TEPCO spokesman "Analysis suggests
that it was not a criticality
65Protective actions
Source SECY-11-0093
66Cleanup and Decommissioning Plan
- December 2011 TEPCO released its 40-year plan
to decommission the plan - Phase 1 Post cold shutdown stabilization and
planning - Maintain stable reactor site conditions
- Conduct RD for later phases
- Complete within 2 years (by end of 2013)
- Phase 2 Removal of fuel from the spent fuel
pools - Remove fuel from spent fuel pools in all units
- Process accumulated water
- Conduct RD for later phase
- Complete within 10 years (by end of 2021)
- Phase 3 Removal of fuel debris through final
decommissioning cleanup - Fuel debris removal in U1, 2 and 3
- Decommissioning and site cleanup
- Complete in 30-40 years (by 2041 2051)
67International Response to theFukushima Daiichi
Accident
68USA
- Aug 2011 NRC released the results of its 90-day
review of Fukushima lessons learned - No "imminent threat, but some issues require
immediate action - Ability to withstand prolonged loss of AC power
- Ability to respond to earthquakes and flooding,
and - Ability to monitor the condition of spent fuel
pools. - Sep 2011 NRC issues, Recommendations for
Enhancing Reactor Safety in the 21st Century,
with 12 recommendations, including - Balance defense in depth and risk
considerations - As needed, upgrade design basis seismic and flood
protection - Strengthen prolonged station blackout mitigation
- Study adequacy of hydrogen control
- Enhance spent fuel makeup capability and
instrumentation - Strengthen on-site emergency response accident
management
69USA
- Nov 2011 Proposed ballot initiative in
California calls for immediate shutdown of PGEs
Diablo Canyon and SCEs San Onofre NPPs, which
generate 16 of California's power. - Dec 2011 NRC approved the Westinghouse AP1000
standard plant design - 13 Jan 2012 Industry NRC meeting to recommend
an approach for post-Fukushima improvements - Diverse and flexible coping strategy (FLEX) for
preventing fuel damage. - FLEX differs from Severe Accident Management
Guidelines (SAMGs), which come into play after
core damage. - FLEX is designed to expand the margin of safety
at nuclear energy facilities and ensure they can
cope with extended loss of power using pre-staged
backup equipment and suppliessuch as fresh water
and diesel fuel that are available
on-sitesupplemented by off-site resources
established for this purpose. - Approach builds on concepts used to provide
additional contingency at U.S. nuclear facilities
after the 9/11 attacks.
70European Union (EU) stress test
- The European Council of 24-25 March 2011
requested that the safety of all EU NPPs be
reviewed on the basis of a stress test. - A reassessment of NPP safety margins in the light
of the events that occurred at Fukushima - Extreme natural events challenging the plant
safety functions and leading to a severe
accident. - A deterministic sequential loss of lines of
defense is assumed, irrespective of the
probability of the loss. - The final country-specific reports were due to be
submitted to the European Nuclear Safety
Regulators Group (ENSREG) by December 31, 2011. - The next stage is a peer review of the
country-specific reports, to be completed by
April 30, 2012 - A consolidated EU report will be issued in June
2012. - These reports are publically available on the
ENSREG web site - http//www.ensreg.eu/
71France
- Current fleet of 58 NPPs has a generating
capacity of 63,130 MWe and produce gt75 of
Frances electricity. - One new 1600 MWe EPR unit is under construction
and one more committed in Nov 2011. - Nov 2011 Green and Socialist parties call for
shutting down 24 NPPs across France by 2024. - President Sarkozy said the proposal would cost
French consumers 5 B (6.63 B) a year. - Dec 2011 First phase of EU stress test
completed. - Jan 2012 French Nuclear Safety Authority (ASN)
stated that current NPPs have a sufficient
safety level, but called for significant safety
investment from EDF on the order of 10 B (about
13.5 B) over 10 years. Identified safety
improvements include - Flood-proof diesel generators, and
- Bunkered remote back-up control rooms
- Nuclear Fast Response Force available to support
an NPP site within 24 hours - EDF is planning to operate its fleet of PWRs for
60 years.
72Germany
- Current fleet of 17 NPPs has a generating
capacity of 20,429 MWe and produce about 23 of
Germanys electricity. - 30 June 2011 the country's parliament voted to
phase out Germany's nuclear fleet - The 8 oldest reactors (gt 8,000 MWe) already have
been disconnected from the grid - The remaining 9 reactors will be retired by 2022
- Sep 2011 International Energy Agency warns
German government of risky phase-out strategy
73Switzerland
- Current fleet of 5 NPPs has a generating capacity
of 3,220 MWe and produce about 38 of
Switzerlands electricity - Parliament approved nuclear phase-out in 2011.
- Preliminary phase-out plan
- Beznau I in 2019 (365 MWe)
- Beznau II and Muehleberg in 2022 (720 MWe
combined), - Goesgen in 2029 (970 MWe)
- Leibstadt in 2034 (1165 MWe)
- Sources of replacement power
- Development of hydro-electric plants and other
renewable energy - Possibly importing electricity.
- If necessary the country could also fall back on
electricity produced by fossil fuels. - It has been estimated that the cost of reshaping
the country's energy resources, offset by
measures to cut consumption, would cost the
country between 0.4 - 0.7 of gross domestic
product per year. - 2010 GDP was 524 B, so phase-out costs 2.1
3.7 B / year - Swiss nuclear safety authority ENSI requires EU
stress tests applied to Swiss NPPs.
74Belgium
- Current fleet of 7 NPPs has a generating capacity
of 5,885 MWe, which represents 92 of domestic
energy generation and 22 of domestic energy
consumption. Belgium imports most of its energy. - In October 2011, the Belgian government committed
to implementing the nuclear exit law of 2003. - The plan calls for the following shutdown
schedule - The three oldest NPPs by 2015 (1787 MWe)
- The remaining four NPPs by 2025 (4098 MWe)
- This plan is conditional on finding enough energy
from alternative sources to prevent electric
supply shortages and significant change in the
price of electricity.
75Elsewhere in Europe
- Italy
- Italy has no NPPs
- In a 12-13 June 2011 referendum, voters rejected
government plans to build new nuclear plants. - Lithuania
- July 2011 GE-Hitachi was selected to build a
new BWR NPP to replace the Ignalina NPP, which is
being decommissioned - Will reduce Baltic states energy dependence on
Russia. - Finland
- October 2011 First in Europe to approve a new
green-field NPP site since the Fukushima Daiichi
accident. - Poland
- Still moving ahead to select NPP supplier in
2013, with initial operation of Polands first
NPP in 2020.
76Japan
- In 2010, the Japanese government approved a plan
to build 14 new NPPs and increase reliance on
nuclear energy. - Current fleet of 48 NPPs (excluding 6 units at
Fukushima Daiichi) has a generating capacity of
42,300 MWe and produce about 25 of Japans
electricity. - Since the Fukushima Daiichi accident, all
reactors that have been shut for regular
maintenance have been kept offline as part of
efforts to assuage public concerns about nuclear
safety. - Only 6 NPPs operating in Japan at the end of
2011. - July 2011 Japanese Prime Minister states the
country must eliminate dependence on nuclear
power.
77Japan
- Tepco proposed to install a system of tide
barriers with watertight doors at Kashiwazaki
Kariwa units 1 to 4. - In addition, TEPCO has installed facilities on
the upland part of the site to provide backup
power and water injection to the reactors and
spent fuel pools, and taken measures to ensure
cooling functions in the event of tsunamis
flooding the reactor buildings
78Japan
- Oct 2011
- Nuclear Safety Commission will mandate that
Japans utilities install reinforced sources of
electric power at all NPPs - Kansai Electric submit the results of the stress
test for Ohi Unit 3 to the Nuclear and
Industrial Safety Agency (NISA). - First stress test to be reported to NISA for
consideration on restarting a shutdown reactor. - Dec 2011
- New nuclear safety agency is being formed under
the Environment Ministry from the merger of the
Nuclear and Industrial Safety Agency of the
Ministry of Economy Trade and Industry and the
Nuclear Safety Commission of Japan - Parliament appoints an independent panel formed
to investigate the Fukushima Daiichi incident - Jan 2012
- Japanese Prime vowed to revive the region
surrounding the Fukushima Daiichi nuclear plant - Amendment proposed to Japans Nuclear Plant
Operations Law to limit NPP operating life to 40
years
79China
- Japan's Fukushima nuclear disaster in March led
China to delay all nuclear project approvals. - Dec 2011 China has approved a five-year nuclear
safety plan, which is a prelude to their nuclear
development plan that is expected to reduce the
2020 nuclear capacity target by about 10.
80Northern VA earthquakemagnitude 5.823 August
2011
81U.S seismic design basis
- Licensing bases for existing NPPs considers
historical data at each site. - Data are used to determine design basis loads
from the areas maximum credible earthquake, with
an additional margin included. - In Generic Letter 88-20, the NRC required
existing NPPs to assess their potential
vulnerability to earthquake events, including
those that might exceed the design basis. - Following the events of September 11, 2001, NRC
required all nuclear plant licensees to take
additional steps to protect public health and
safety in the event of a large fire or explosion.
If needed, these additional steps could also be
used to mitigate severe natural phenomena. - The NRC examined new Central Eastern US (CEUS)
earthquake hazard information under Generic
Issues GI-199 and completed a limited scope
screening analysis for the seismic issue in
December 2007. - New CEUS seismic data were compared with earlier
seismic evaluations. - This analysis confirmed that operating nuclear
power plants remain safe with no need for
immediate action.
82Northern VA earthquakemagnitude 5.8, 23 August
2011
- Very short duration peak acceleration (1 3
sec). - No fault associated with the earthquake
- epicenter and aftershocks.
- No surface ruptures during the earthquake.
- NRC classifies as blind reverse fault.
83Northern VA earthquakemagnitude 5.8, 23 August
2011
- Although the U.S. east of the Rockies has fewer
and generally smaller earthquakes than the West,
due to geologic differences, eastern earthquakes
affect areas 10 time than western ones of the
same magnitude. (ref NJ Geologic Survey) - Hard ground and fewer faults
- Effective in conducting seismic waves over long
distances. - USGS estimated the earthquake produced a peak
ground acceleration of 0.26g at the North Anna
NPP - First time that an earthquake has exceeded the
design basis for a U.S. NPP.
84North Anna NPP
- 2 unit Westinghouse PWRs
- Net 1,806 MWe
- Both operating at 100 power when earthquake
occurred - Site includes an independent spent fuel storage
installation - 11 miles from epicenter
- Seismic design basis
- DBE, structures on rock 0.12g horiz, 0.08 g
vert - DBE, structures on soil 0.18g horiz, 0.12g
vert - OBE ½ DBE
85Plant response to the earthquake
- Reactor tripped automatically
- Reactor trip system does not include an automatic
seismic scram. - Direct cause for both Units 1 2 reactor trip
was detection of high rate of change of neutron
flux (decreasing) in the power range nuclear
instruments (gt5 change in 2.5 seconds). - Root cause is believed to be a synergistic
combination of seismically-induced conditions - Core barrel, core detector movement.
- Momentary change in thermal boundary layer
conditions along the fuel rods. - Momentarily under-moderated core with oscillatory
but overall decreasing flux. - Turbine tripped automatically and offsite power
lost - Main turbines tripped because of main transformer
lockout, which interrupted the connection to
the off-site grid. - The earthquake caused multiple transformers to
lockout due to activation of sudden pressure
relays, which operated as designed due to
earthquake-induced pressure pulses within the
transformer, not due to an electrical fault. - NPP connection to offsite power restored about 7
hours later. - Mitigating systems started automatically
86Reactor power during earthquake, before scram
Scram ?
87North Anna earthquake timeline
Date Time Events at North Anna NPP
23 Aug 2011 1351 5.8 magnituide earthquake
23 Aug 2011 1351 Automatic reactor trip
23 Aug 2011 1351 Loss of offsite power and automatic main turbine trip
23 Aug 2011 1351 Automatic actuation of auxiliary feedwater system, charging system, emergency diesel generators, and service water system
23 Aug 2011 1403 Alert declared. Operators focus on stabilizing each unit and restoring offsite power.
23 Aug 2011 2055 NPP connection to offsite power restored
24 Aug Unit 1 cooldown to cold shutdown started. Unit 2 cooldown started after Unit 1 cooldown completed
26 Aug Review of seismic data determined that seismic acceleration potentially exceeded the Design Basis Earthquake at frequencies above 5 Hz.
Aug - Oct Plant walkdowns, inspections, tests and analysis do not reveal any significant physical or functional damage to safety-related structures, systems or components, and only limited damage to non-safety, non-seismic SSC.
1 Nov Public meeting with NRC to address readiness to re-start
11 Nov NRC approves re-start
18 Nov Unit 1 back at 100 power
21 Nov Unit 2 back at 100 power
88U.S. restart requirements and guidance
- Appendix A to 10CFR100Paragraph V(a)(2)
- If vibratory ground motion exceeding that of the
Operating Basis Earthquake occurs, shutdown of
the nuclear power plant will be required. - Prior to resuming operations, the licensee will
be required to demonstrate to the Commission that
no functional damage occurred to those features
necessary for continued operation without undue
risk to the health and safety of the public. - Regulatory Guide 1.166, Pre-earthquake planning
and immediate NPP Operator Post-earthquake
Actions (1997) - Cumulative Absolute Velocity (CAV) is a measure
of the damage potential of earthquake ground
motion - NRC, EPRI and industry agree on a CAV threshold
- If CAV calculation gt 0.16 g-sec, then OBE
exceeded - Regulatory Guide 1.167, Restart of Nuclear Power
Plant Shut Down by a Seismic Event (1997) - EPRI NP-6695, Guidelines for Nuclear Power Plant
Response to an Earthquake (1990)
89Dominion report of readiness to re-start
- Acceleration criteria were briefly exceeded in
certain directions and frequencies by a strong,
but very short duration earthquake - Previous IPEEE evaluations establish that safe
shutdown systems, structures and components can
handle peak accelerations above design basis - No safety-related systems, structures or
components required repair due to the earthquake - No significant damage was found or should have
been expected and results of expanded tests and
inspections have confirmed expectations - Commitments
- By February 2012 With Westinghouse, develop a
plan for additional evaluations or inspections to
assure long-term reliability of reactor
internals. - By December 2012 Improve seismic monitoring
equipment. - By March 2013 Reevaluate equipment identified
in the Individual Plant Evaluation of External
Events (IPEEE) with a high-confidence-of-low-proba
bility-of-failure (HCLPF) capacity of lt0.3g and
recommend potential improvements
Source Dominion 31 Oct 2011 letter to NRC and 1
Nov 11 presentation
90Basis for post-earthquake integrity of North Anna
structures, systems components
0.16 g-sec -------
Source Dominion 1 Nov 11 presentation to NRC
91Missouri River FloodingFort Calhoun NPPJune
August 2011
92U.S. design basis flood and flood protection
- A design-basis flood is a flood caused by one or
an appropriate combination of several
hydrometeorological, geoseimic, or
structural-failure phenomena, which results in
the most severe hazards to structures, systems,
and components (SSCs) important to the safety of
a nuclear power plant (NUREG/CR-7046). - Sources of requirements guidance
- USNRC Regulatory Guide 1.59, Design Basis Floods
for NPPs (1977) - USNRC Regulatory Guide 1.102 (R1), Flood
Protection for NPPs (1976) - Standard Review Plan 3.4.1, R2, Flood
Protection (1981) - NUREG/CR-7046, Design-Basis Flood Estimation for
Site Characterization at Nuclear Power Plants in
the United States of America (Nov 2011) - Temporary flood barriers, such as sandbags,
plastic sheeting, portable panels, etc., which
must be installed prior to the advent of the
DBFL, are not acceptable for issuance of a
construction permit. - However, unusual circumstances could arise after
construction that would warrant consideration of
such barriers. - One example of unusual circumstances that might
justify use of temporary barriers is a
post-construction change in the flood-producing
characteristics of the drainage area.. In such
circumstances, and with strong justification, the
staff may accept temporary barriers (RG 1.102)
93Fort Calhoun NPP site
Source ORNL-NSIC-55, V1
94Missouri River floods Fort Calhoun NPP site
- Site grade elevation 1004 MSL, includes an
independent spent fuel storage - installation
- Alert level 1006 MSL
- Auxiliary building ground floor level 1007 MSL
- Tech Spec reactor shutdown level1009 MSL
- Current design basis flood level 1014 MSL with
NPP main buildings - switchyard protected by temporary barrier
(AquaDam)
95Missouri River floods Fort Calhoun NPP site
96AquaDam temporary barrier
OPPD refers to the water-filled AquaDam as a
supplemental flood protection measure that
provides protection up to 1014 MSL.
97Equipment at or below grade in the auxiliary
building that must be protected from flooding
- 1007 level
- Both divisions of AC and DC power
- Diesel generators
- Batteries
- 4160 VAC, 480 VAC and 125 VDC electric panels
- Alternate shutdown panel
- New fuel storage
- 989 level
- Emergency feedwater pumps
- 480 v Class 1E panels
- 971 level
- High pressure safety injection (ECCS) pumps
- Low pressure safety injection / shutdown cooling
pumps
98Fort Calhoun flood timeline
Date Events at Fort Calhoun NPP
9 Apr 2011 NPP in cold shutdown for routing refueling
6 Jun Notice of Unusual Event (NOUE) due to high river level
8 Jun Fire in switchgear temporarily disables spent fuel pool cooling
17 Jun NRC implements 24 hr/day augmented coverage
26 Jun Worker punctures AquaDam with bobcat. Plant temporarily disconnected from offsite power to protect switchyard equipment that might become flooded. NPP loads supplied from the diesel generators.
29 Jun Missouri river crests at Blair
11 Jul Re-installation of AquaDam complete, water within the confines of the dam perimeter removed.
23 Jul Second Missouri river crest at Blair
29-30 Jul NOUE rescinded, NRC suspends 24 hr/day augmented coverage.
10 Aug OPPD issues post-flood recovery plan. Updated 30 Aug.
6 Oct NRC issues finding of inadequate flood protection strategies.
14 Dec NRC delays plans to restart Fort Calhoun at least to 2012 Q2.
17 Dec OPPD issues LER for inadequate flood protection for intake structure and auxiliary building due to unsealed wall and ceiling/floor penetrations and other reasons.
99Hurricanes
100Hurricane Andrew - 1992
- Category 4 Hurricane Andrew 1993
- First time a hurricane significantly affected a
U.S. NPP - Hurricane passed over 2-unit Turkey Point NPP,
which was shut down 4 hours prior to the onset of
hurricane strength winds - 145 mph winds, gusts to 175 mph
- The onsite damage included loss of all offsite
power for more than 5 days, complete loss of
communication systems, closing of the access
road, and damage to the fire protection and
security systems and warehouse facilities. - No damage to the safety-related systems except
for minor water intrusion. - There was no radioactive release to the
environment.
101Hurricane Andrew - 1992
102Hurricane Irene - 2011
- Category 3 Hurricane Irene 2011
- Only two NPPs in the hurricanes track were shut
down - In Maryland, one reactor at the Calvert Cliffs
plant automatically went off-line when wind blew
a piece of aluminum siding into the units main
transformer in the switchyard. The second unit
remained online - In New Jersey, the Oyster Creek NPP was taken
offline as a precaution ahead of expected high
winds and storm surge. - All others remained on-line throughout the storm.
103Hurricane Irene - 2011
104Conclusions
- NPPs have demonstrated their robustness and
ability to withstand some beyond design basis
severe natural events and then be able to return
to operation. - The magnitude of some beyond design basis severe
natural events were much greater than expected
based on pre-event knowledge of historical events
and site characteristics. - The common cause failure potential for some
beyond design basis severe natural events has
been grossly underestimated. - It is time to redefine the nuclear regulatory
process and develop a more effective approach for
assuring that nuclear safety functions can be
accomplished when required so nuclear power
plants can cope with events and combinations of
events that exceed the traditional design basis.