Transcript The Fukushima earthquake and tdalevere natural

The Fukushima earthquake +
tsunami and other recent external
events that have challenged the
design basis for commercial
nuclear power plants
Peter Lobner
18 January 2012
Agenda



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
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

Conclusions
Definition 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
Safety 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:



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

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)
Safety 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-offailure” (HCLPF) level.
Source: 10CFR50 Appendix A, General Design Criterion 2
The design basis is not static

10CFR50.54,Conditions of License, paragraph (f):




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:



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.
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
Station 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



44 U.S. NPP implemented AC-independent solutions

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Batteries only
Maximum coping duration 4 hours
60 U.S. NPPs implemented alternate AC power sources. For
example:

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

“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.
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.
BWR mitigating systems and
system dependencies
Key mitigating systems available at the
Fukushima Daiichi units
BWR Mark-I containment steel shell
BWR Mark-I containment arrangement
within the reactor building
BWR Mark I containment performance
improvement (CPI) program

Resolution of Generic Safety Issue 157, Containment Performance,
resulted in significant modifications to BWR Mark 1 containments.




In Generic Letter 89-16 (1989), NRC requested each licensee to
provide cost estimates for implementation of a hardened vent.


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).
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.
BWR Mark-I containment refueling floor
arrangement
Isolation 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
Reactor Core Isolation Cooling (RCIC)
System
System Dependencies





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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 < 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
High-Pressure Coolant Injection (HPCI) System
System Dependencies

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Automatic actuation on
reactor vessel low water
level or drywell high
pressure, or remotemanual
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 < 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
Low-Pressure ECCS and Residual Heat
Removal (RHR)
System Dependencies
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

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
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Automatic pump actuation
on reactor vessel low
water level or drywell high
pressure, or remotemanual
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
Niigata Chuetsu-Oki Earthquake (NCOE),
Japan, magnitude 6.8, 16 July 2007
Niigata Chuetsu-Oki Earthquake (NCOE),
Japan, magnitude 6.8, 16 July 2007
Source: TEPCO
Source: EQECAT Inc
Kashiwazaki-Kariwa NPP

World’s 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.
Kashiwazaki-Kariwa NPP
Source: TEPCO
NCOE observed seismic data

The observed seismic accelerations largely exceeded
the design basis values.
Source: TEPCO
NPP 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.
NPP 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-safetyrelated 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
Improved 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.
Japan’s 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
Standard 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
Post-NCOE safety actions

Install seismic reinforcements to tolerate seismic motion
of 1000 Gal (~1.5 times NCOE max)



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

Perform facility integrity evaluation

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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
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.
Re-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.

Great East Japan Earthquake,
magnitude 9.0, 11 March 2011
Great East Japan Earthquake,
magnitude 9.0, 11 March 2011, 14:46 JST
Source: EQECAT Inc
Source: USGS
• 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.
Fukushima Daiichi NPP

One of Japan’s
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)
Fukushima
Daiichi Site
Arrangement
Source: INPO
Observed 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.
Tsunami following the
Great East Japan Earthquake,
11 March 2011
Tsunami timeline at Fukushima Daiichi
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15:27: First of seven tsunami waves arrived. Height
about 13’ (4 m) was less than the design basis tsunami
and was mitigated by the breakwater.
15:35: 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.
Tsunami wave arrives at Fukushima Daiichi
Tsunami wave arrives at Fukushima Daiichi
Site inundation
Site inundation
Site inundation
Tsunami effects on storage tank
Fukushima Daiichi site inundation
Source: IAEA
Fukushima 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.
Fukushima Daiichi plant responses
to the 11 March 2011
earthquake + tsunami
Decay heat – reactor units 1, 2, 3
Source: MIT
Decay heat – spent fuel pools
Fuel response to severe accident
progression
Unit 1 sequence of events
Adapted from INPO
Unit 1 sequence of events (continued)
Adapted from INPO
Unit 1 Hydrogen Explosion, 12 March 2011
Loss if lighting in the control room
Source: TEPCO
Unit 2 sequence of events
Adapted from INPO
Unit 2 sequence of events (continued)
Adapted from INPO
Unit 3 sequence of events
Adapted from INPO
Unit 3 sequence of events (continued)
Adapted from INPO
Unit 3 Hydrogen Explosion, 14 March 2011
Unit 4 sequence of events
Adapted from INPO
Unit 4 after hydrogen explosion
Possible hydrogen leak path to Unit 4
Source: INPO
Units 1-4 before the tsunami & explosion
Units 1-4 after the explosion
Unit 5 & 6 sequence of events
Source: SECY-11-0093
Severe 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:
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
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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
Severe accident response issues

Was there a re-criticality at Unit 2?
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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”
Protective actions
Source: SECY-11-0093
Cleanup and Decommissioning Plan

December 2011: TEPCO released its 40-year plan to
decommission the plan:

Phase 1: Post cold shutdown stabilization and planning
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
Phase 2: Removal of fuel from the spent fuel pools

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Maintain stable reactor & site conditions
Conduct R&D for later phases
Complete within 2 years (by end of 2013)
Remove fuel from spent fuel pools in all units
Process accumulated water
Conduct R&D for later phase
Complete within 10 years (by end of 2021)
Phase 3: Removal of fuel debris through final decommissioning
& cleanup


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Fuel debris removal in U1, 2 and 3
Decommissioning and site cleanup
Complete in 30-40 years (by 2041 – 2051)
International Response to the
Fukushima Daiichi Accident
USA

Aug 2011: NRC released the results of its 90-day review of
Fukushima lessons learned

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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

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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
USA


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Nov 2011: Proposed ballot initiative in California calls for immediate
shutdown of PG&E’s Diablo Canyon and SCE’s 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 supplies—such as fresh water and
diesel fuel that are available on-site—supplemented 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.
European 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




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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/
France



Current fleet of 58 NPPs has a generating capacity of 63,130 MWe
and produce >75% of France’s 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.
Germany


Current fleet of 17 NPPs has a generating capacity of
20,429 MWe and produce about 23% of Germany’s
electricity.
30 June 2011: the country's parliament voted to phase
out Germany's nuclear fleet



The 8 oldest reactors (> 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
Switzerland



Current fleet of 5 NPPs has a generating capacity of 3,220 MWe
and produce about 38% of Switzerland’s 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




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.


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.
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.
Belgium



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.
Elsewhere in Europe

Italy:



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 state’s energy dependence on Russia.
Finland:


Italy has no NPPs
In a 12-13 June 2011 referendum, voters rejected government
plans to build new nuclear plants.
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 Poland’s first NPP in 2020.
Japan



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 Japan’s 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.
Japan

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
Japan

Oct 2011:


Nuclear Safety Commission will mandate that Japan’s 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).


Dec 2011:



First stress test to be reported to NISA for consideration on
restarting a shutdown reactor.
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 Japan’s Nuclear Plant Operations Law
to limit NPP operating life to 40 years
China


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%.
Northern VA earthquake
magnitude 5.8
23 August 2011
U.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 area’s
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.
Northern VA earthquake
magnitude 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”.
Northern VA earthquake
magnitude 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.
North Anna NPP




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


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
Plant 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 (>5% change in 2.5 seconds).
 Root cause is believed to be a synergistic combination of seismicallyinduced 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
Reactor power during earthquake, before
scram
Scram 
North Anna earthquake timeline
Date
Time
Events at North Anna NPP
23 Aug 2011
1351
5.8 magnituide earthquake
Automatic reactor trip
Loss of offsite power and automatic main turbine trip
Automatic actuation of auxiliary feedwater system, charging system, emergency
diesel generators, and service water system
1403
Alert declared. Operators focus on stabilizing each unit and restoring offsite
power.
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
U.S. restart requirements and guidance

Appendix A to 10CFR100—Paragraph 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 > 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)
Dominion 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-lowprobability-of-failure (HCLPF) capacity of <0.3g and recommend
potential improvements
Source: Dominion 31 Oct 2011 letter to NRC and 1 Nov 11 presentation
Basis for post-earthquake integrity of North
Anna structures, systems & components
0.16 g-sec -------
Source: Dominion 1 Nov 11 presentation to NRC
Missouri River Flooding
Fort Calhoun NPP
June – August 2011
U.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 structuralfailure 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)
Fort Calhoun NPP site
Source: ORNL-NSIC-55, V1
Missouri 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 level:1009’ MSL
• Current design basis flood level: 1014’ MSL with NPP main buildings &
switchyard protected by temporary barrier (AquaDam®)
Missouri River floods Fort Calhoun NPP site
AquaDam® temporary barrier
OPPD refers to
the water-filled
AquaDam® as a
supplemental
flood protection
measure that
provides
protection up to
1014’ MSL.
Equipment at or below grade in the auxiliary
building that must be protected from flooding

1007’ level:

Both divisions of AC and DC power






Alternate shutdown panel
New fuel storage
989’ level:



Diesel generators
Batteries
4160 VAC, 480 VAC and 125 VDC electric panels
Emergency feedwater pumps
480 v Class 1E panels
971’ level:


High pressure safety injection (ECCS) pumps
Low pressure safety injection / shutdown cooling pumps
Fort 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.
Hurricanes
Hurricane 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.
Hurricane Andrew - 1992
Hurricane Irene - 2011

Category 3 Hurricane Irene – 2011
 Only
two NPPs in the hurricane’s track were shut
down:


 All
In Maryland, one reactor at the Calvert Cliffs plant
automatically went off-line when wind blew a piece of
aluminum siding into the unit’s 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.
others remained on-line throughout the storm.
Hurricane Irene - 2011
Conclusions




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.