Nuclear Power Plant Design and Seismic
Safety Considerations
Anthony Andrews
Specialist in Energy and Defense Policy
May 2, 2011
Congressional Research Service
7-5700
www.crs.gov
R41805
CRS Report for Congress
P
repared for Members and Committees of Congress

---

Nuclear Power Plant Design and Seismic Safety Considerations

Summary
Since the March 11, 2011, earthquake and tsunami that devastated Japan’s Fukushima Daiichi
nuclear power station, the seismic criteria applied to siting commercial nuclear power plants
operating in the United States have received increased attention; particularly the Nuclear
Regulatory Commission’s (NRC’s) 2010 reassessment of seismic risks at certain plant sites.
Commercial nuclear power plants operating in the United States vary considerably, as most were
custom-designed and custom-built. Boiling water reactors (BWRs) directly generate steam inside
the reactor vessel. Pressurized water reactors (PWRs) use heat exchangers to convert the heat
generated by the reactor core into steam outside of the reactor vessel. U.S. utilities currently
operate 104 nuclear power reactors at 65 sites in 31 states; 69 are PWR designs and the 35
remaining are BWR designs.
One of the most severe operating conditions for a reactor is a loss of coolant accident (LOCA),
which can lead to a reactor core meltdown. The emergency core cooling system (ECCS) provides
core cooling to minimize fuel damage by injecting large amounts of cool, borated water into the
reactor coolant system following a pipe rupture or other water loss, and (secondarily) to provide
extra neutron poisons to ensure the reactor remains shut down. The ECCS must be sized to
provide adequate make-up water to compensate for a break of the largest diameter pipe in the
primary system (i.e., the so-called “double-ended guillotine break” (DEGB)). However, the NRC
considers the DEGB to be an extremely unlikely event. Nevertheless, even unlikely events can
occur, as the combined tsunami and magnitude 9.0 earthquake that struck Fukushima Daiichi
proves.
U.S. nuclear power plants have designs based on Deterministic Seismic Hazard Analysis
(DSHA). Since then, Probabilistic Seismic Hazard Analysis (PSHA) has been adopted as a more
comprehensive approach in engineering practice. Consequently, the NRC is reassessing the
probability of seismic core damage at existing plants.
In 2008, the U.S Geological Survey (USGS) released an update of the National Seismic Hazard
Maps (NSHM). USGS notes that the 2008 hazard maps differ significantly from the 2002 maps in
many parts of the United States, and generally show 10%-15% reductions in spectral and peak
ground acceleration across much of the Central and Eastern United States (CEUS), and about
10% reductions for spectral and peak horizontal ground acceleration in the Western United States
(WUS). Seismic hazards are greatest in the WUS, particularly in California, Oregon, and
Washington, as well as Alaska and Hawaii.
In 2010, NRC published its GI-199 Safety/Risk Assessment; a two-stage assessment of the
implications of USGS updated probabilistic seismic hazards analysis in the CEUS on existing
nuclear power plants sites. NRC does not rank nuclear plants by seismic risk. NRC’s objective in
GI-199 was to evaluate the need for further investigations of seismic safety for operating reactors
in the CEUS. The data evaluated in the assessment suggest that the probability for earthquake
ground motion above the seismic design basis for some nuclear plants in the CEUS, although still
low, is larger than previous estimates. In late March 2011, NRC announced that it had identified
27 nuclear reactors operating in the CEUS that would receive priority earthquake safety reviews.

Congressional Research Service

---

Nuclear Power Plant Design and Seismic Safety Considerations

Contents
Background on Seismic Standards............................................................................................... 1
Nuclear Power Plant Designs ...................................................................................................... 2
Boiling Water Reactor (BWR) Systems ................................................................................. 3
Safe-Shutdown Condition ............................................................................................... 4
Loss of Coolant Accident ................................................................................................ 4
BWR Design Evolution................................................................................................... 5
Pressurized Water Reactor Systems ....................................................................................... 8
PWR Design Configurations ........................................................................................... 9
Safe Shutdown Condition.............................................................................................. 10
Loss of Coolant Accident .............................................................................................. 10
Containment Structure Designs ........................................................................................... 11
Nuclear Power Plants Operating in the United States ........................................................... 13
Plant Seismic Siting Criteria...................................................................................................... 16
General Design Criteria....................................................................................................... 16
Site Investigations............................................................................................................... 17
Safe Shutdown Earthquake Condition ................................................................................. 18
National Seismic Hazard Maps............................................................................................ 20
NRC Priority Earthquake Safety Review ............................................................................. 24
Recent Legislative Activities ..................................................................................................... 25

Figures
Figure 1. Boiling Water Reactor (BWR) Plant ............................................................................. 4
Figure 2. GE BWR / Mark I Containment Structure..................................................................... 6
Figure 3. General Electric Mark II Containment Structure ........................................................... 7
Figure 4. General Electric Mark III Containment Structure.......................................................... 8
Figure 5. Pressurized Water Reactor (PWR) Plant ....................................................................... 9
Figure 6. Commercial Nuclear Power Plants Operating in the United States .............................. 14
Figure 7. Spectral Acceleration 5 Hz ......................................................................................... 19
Figure 8. Operating Nuclear Power Plant Sites and Seismic Hazard........................................... 22
Figure 9. Operating Nuclear Power Plant Sites and Mapped Quaternary Faults.......................... 23
Figure A-1. Spectral Acceleration (g) vs. Frequency (Hz) .......................................................... 28

Tables
Table 1. Reactor Type, Vendor, and Containment ......................................................................... 3
Table 2. BWR Design Evolution ................................................................................................. 5
Table 3. Containment Building Design Parameters .................................................................... 12
Table 4. Operating Nuclear Power Plants Subject to Earthquake Safety Reviews ....................... 24
Congressional Research Service

---

Nuclear Power Plant Design and Seismic Safety Considerations


Appendixes
Appendix A. Magnitude, Intensity, and Seismic Spectrum ......................................................... 27
Appendix B. Terms ................................................................................................................... 30

Contacts
Author Contact Information ...................................................................................................... 30
Acknowledgments .................................................................................................................... 30

Congressional Research Service

---

Nuclear Power Plant Design and Seismic Safety Considerations

Background on Seismic Standards
The seismic design criteria applied to siting commercial nuclear power plants operating in the
United States received increased attention following the March 11 earthquake and tsunami that
devastated Japan’s Fukushima Daiichi nuclear power station. Since the events, some in Congress
have begun to question whether U.S plants are vulnerable to a similar threat, particularly in light
of the Nuclear Regulatory Commission’s (NRC’s) ongoing reassessment of seismic risks at
certain plant sites.1
Commercial nuclear power plants operating in the United States use light water reactor designs,
but vary widely in design and construction. Light water reactors use ordinary water as a neutron
moderator and coolant, and uranium fuel enriched in fissile uranium-235.2 Designs fall into either
pressurized water reactor (PWR) or boiling water reactor (BWR) categories. Both have reactor
cores (the source of heat) consisting of arrays of uranium fuel bundles capable of sustaining a
controlled nuclear reaction.3 U.S. commercial nuclear power plants incorporate safety features
intended to ensure that, in the event of an earthquake, the reactor core would remain cooled, the
reactor containment would remain intact, and radioactive releases would not occur from spent
fuel storage pools. The Nuclear Regulatory Commission (NRC) defines this as the “safe-
shutdown condition.”
When utilities began building nuclear power plants in the 1960s-1970s era, they typically hired an
architect/engineering firm, then contracted with a reactor manufacturer (“nuclear vendors”) to
build the nuclear steam supply system (NSSS), consisting of the nuclear core, reactor vessel,
steam generators and pressurizer (in PWRs), and control mechanisms—representing about 10%
of the plant investment.4 The balance of the plant (BOP) consisted of secondary cooling systems,
feed-water systems, steam systems, control room, and generator systems. At the time, the four
vendors who offered designs for nuclear reactor systems were Babcock & Wilcox, Combustion
Engineering, General Electric, and Westinghouse. About 12 architect/engineering firms were
available to design the balance of the plant. Each architect/engineer had its own preferred
approach to designing the balance of plant systems. In addition, plant site-conditions varied due
to the different meteorological, seismic, and hydrological conditions. The custom design-and-
build industry approach resulted in problems verifying the safety of individual plants and in
transferring the safety lessons learned from one reactor to another.
The previous design approach to withstanding earthquakes had relied on Deterministic Seismic
Hazard Analysis (DSHA). Any new plant design is to consider Probabilistic Seismic Hazard
Analysis (PSHA), which has been widely adopted in engineering practice. Deterministic analysis
attempts to quantify the worst-case scenario based on the combination of earthquake sources at a
site’s location that results in the strongest ground-motion potentially generated.5 In other words,

1 This report does not discuss the risk from earthquake-caused tsunamis, as associated with the catastrophic damage to
the Fukushima plants.
2 Heavy water reactors, such as Canada’s CANDU reactor, use water enriched with a heavier hydrogen isotope and
natural uranium for fuel, which contains less than 3.5% uranium-235.
3 For further background uranium fuel, see CRS Report RL34234, Managing the Nuclear Fuel Cycle: Policy
Implications of Expanding Global Access to Nuclear Power, coordinated by Mary Beth Nikitin.
4 Office of Technology Assessment, Nuclear Power Plant Standardization: Light Water Reactors, NTIS order #PB81-
213589, April 1981, p. 11.
5 Julian J. Bommer, Norman A. Abrahamson, and Fleur O. Strasser, et al., “The Challenge of Defining Upper Bounds
(continued...)
Congressional Research Service
1

---

Nuclear Power Plant Design and Seismic Safety Considerations

the deterministic assessment focuses on a single earthquake event to determine the finite
probability of occurrence. PSHA is a methodology that estimates the likelihood that various levels
of earthquake-caused ground motion will be exceeded at a given location in a given future time
period.6 Due to possible uncertainties in geoscience data and in the models used to estimate
ground motion from earthquakes, multiple model interpretations are often possible. This has led
to disagreement among experts, which in turn has led to disagreement on the selection of ground
motion magnitudes for the design at a given site. PSHA traditionally quantified ground motion
based on peak ground acceleration (PGA).7 Today, the preferred parameter is Response Spectral
Acceleration (SA), which gives the maximum acceleration of an oscillating structure such as a
building or power plant.
In its 2010 study (GI-199), the NRC concludes that deterministic assessments (DSHA) do not
necessarily mean that the seismic design basis for the Safe Shutdown Earthquake (SSE) condition
was, or is, deficient in some fashion.8 The design approach to developing loadings on power plant
piping and equipment systems relies on the SSE condition. Existing nuclear plants designs
include considerable safety margins that enable them to withstand “deterministic” or “scenario
earthquake” ground motions that accounted for the largest earthquakes expected in the area
around the plant.9 The NRC study found that some plant sites might have an increased
probability, albeit relatively small, of exceeding their design basis ground motion. NRC considers
that the probabilities of seismic core damage occurring are lower than its guidelines for taking
immediate action, but has determined that some plants’ performance should be reassessed based
on updated seismic hazards.
This report presents some of the general design concepts of operating nuclear power plants in
order to discuss design considerations for seismic events. This report does not attempt to
conclude whether one design is inherently safer or less safe than another plant. Nor does it
attempt to conclude whether operating nuclear power plants are at any greater or lesser risk from
earthquakes given recent updates to seismic data and seismic hazard maps.
Nuclear Power Plant Designs
Currently, 104 nuclear power plants currently operate at 65 sites in 31 states; 69 are PWR designs
and the 35 remaining are BWR designs.

(...continued)
on Earthquake Ground Motions,” Seismological Research Letters, vol. 75, no. 1 (February 2004).
6 R. J. Budnitz, G. Apostolakis, and D. M. Boore, Recommendations for Probabilistic Seismic Hazard Analysis:
Guidance on Uncertainty and Use of Experts: Main Report, U.S. Nuclear Regulatory Commission, Nureg/CR-6372,
Lawrence Berkeley National Laboratory, CA, April 1997, http://www.nrc.gov/reading-rm/doc-collections/nuregs/
contract/cr6372/vol1/index.html#pub-info.
7 Edward (Ned) H. Field, Probabilistic Seismic Hazard Analysis (PSHA) - A Primer, http://www.relm.org/
tutorial_materials.
8 U.S. Nuclear Regulatory Commission, Implications of Updated Probabilistic Seismic Hazard Estimates in Central
and Eastern United States Existing Plants - Safety/Risk Assessment, Generic Issue 199 (GI-199), August 2010.
9 U.S. NRC, NRC frequently asked questions related to the March 11, 2011 Japanese Earthquake and Tsunami, March
2011, http://www.nrc.gov.
Congressional Research Service
2

---

Nuclear Power Plant Design and Seismic Safety Considerations

The more numerous PWR plants include Babcock & Wilcox, Combustion Engineering, and
Westinghouse designs. The BWR plants all use a General Electric design. Table 1 summarizes
the various reactor types. The sections that follow discuss them further.
Table 1. Reactor Type, Vendor, and Containment
Reactor Type Vendor
Containment Type
No. of Plants
PWR
Babcock & Wilcox 2-Loop Lower
Dry, Ambient Pressure
7

Combustion Engineering
Dry, Ambient Pressure
11

Combustion Engineering System 80 Large Dry, Ambient Pressure
3

Westinghouse 2-Loop
Dry, Ambient Pressure
6

Westinghouse 3-Loop
Dry, Ambient Pressure
7

Westinghouse 3-Loop
Dry, Sub-atmospheric
6

Westinghouse 4-Loop
Dry, Ambient Pressure
18

Westinghouse 4-Loop
Dry, Sub-atmospheric
1

Westinghouse 4-Loop
Wet, Ice Condenser
9

Westinghouse 4-Loop
Dry, Ambient Pressure
1


69




BWR
General Electric Type 2
Wet, Mark I
2

General Electric Type 3
Wet, Mark I
6

General Electric Type 4
Wet, Mark 1
15

General Electric Type 4
Wet, Mark II
4

General Electric Type 5
Wet, Mark II
4

General Electric Type 6
Wet, Mark III
4


35
Source: U.S. NRC.
Boiling Water Reactor (BWR) Systems
A boiling water reactor generates steam directly inside the reactor vessel as water flows upward
through the reactor’s core (see Figure 1).10 The water also cools the reactor core, and the reactor
operator is able to vary the reactor’s power by controlling the rate of water flow through the core
with recirculation pumps and jet pumps. The generated steam flows out the top of the reactor
vessel through pipelines to a combined high-pressure/low-pressure turbine-generator. After the
exhausted steam leaves the low-pressure turbine, it runs through a condenser/heat exchanger that
cools the steam and condenses it back to water. A series of pumps return the condensed water
back to the reactor vessel. The heat exchanger cycles cooling water through a cooling tower, or
takes in and discharges water with a lake, river, or ocean. The water that flows through the
reactor, steam turbines, and condenser is a closed loop that never contacts the outside
environment under normal operating conditions. Reactors of this design operate at temperatures
of approximately 570º F and pressures of 1,000 pounds per square inch (psi) atmospheric.

10 U.S. Nuclear Regulatory Commission, Reactor Concepts Manual, Boiling Water Reactor Systems,
http://www.nrc.gov/reading-rm/basic-ref/teachers/03.pdf - 2005-10-17.
Congressional Research Service
3

---

Nuclear Power Plant Design and Seismic Safety Considerations

Figure 1. Boiling Water Reactor (BWR) Plant
Generic Design Features

Source: U.S. Nuclear Regulatory Commission, Reactor Concepts Manual, Boiling Water Reactor Systems, 2005.
Safe-Shutdown Condition
During normal operation, reactor cooling relies on the water that enters the reactor vessel and the
generated steam that leaves. During safe shutdown, the core continues to generate heat by
radioactive decay and generates steam.11 Under this condition, the steam bypasses the turbine and
diverts directly to the condenser to cool the reactor. When the reactor vessel pressure decreases to
approximately 50 psi, the shutdown-cooling mode removes residual heat by pumping water from
the reactor recirculation loop through a heat exchanger and back to the reactor via the
recirculation loop. The recirculation loop design limits the number of pipes that penetrate the
reactor vessel.
Loss of Coolant Accident
The most severe operating condition that a reactor design must contend with is a loss of coolant
accident (LOCA). In the absence of coolant, the uncovered reactor core continues to generate heat
through fission. The resulting heat buildup can damage the fuel or fuel cladding and lead to a fuel
“meltdown.” Under such a condition, an emergency core cooling system (ECCS) provides water

11 During the sustained chain reaction in an operating reactor, the U-235 splits into highly radioactive fission products,
while the U-238 is partially converted to plutonium-239 by neutron capture, some of which also fissions. Further
neutron capture creates other radioactive elements. The process of radioactive decay transforms an atom to a more
stable element through the release of radiation—alpha particles (two protons and two neutrons), charged beta particles
(positive or negative electrons), or gamma rays (electromagnetic radiation).
Congressional Research Service
4

---

Nuclear Power Plant Design and Seismic Safety Considerations

to cool the reactor core. The ECCS is an independent high-pressure coolant injection system that
requires no auxiliary electrical power, plant air systems, or external cooling water systems to
provide makeup water under small and intermediate loss of coolant accidents. A low-pressure
ECCS sprays water from the suppression pool into the reactor vessel and on top of the fuel
assemblies.12 The ECCS must also be sized to provide adequate makeup water to compensate for
a break of the largest diameter pipe in the primary system (i.e., the so-called “double-ended
guillotine break” (DEGB)). However, the NRC views the DEGB as an extremely unlikely event
(likely to occur only once per 100,000 years of reactor operation).13
BWR Design Evolution
Currently, General Electric Type 2 through Type 6 BWRs operate in the United States (Table 1).
BWRs are inherently simpler designs than other light water reactor types. Since they heat water
and generate steam directly inside the reactor vessel, there are fewer components.
Table 2. BWR Design Evolution
Year
Model
Introduced
Design Feature
Typical Plants
BWR/1 1955
Natural
circulation
Dresden 1
First internal steam separation
Big Rock Point
Isolation condenser
Humboldt Bay
Pressure Suppression Containment

BWR/2
1963
Large direct cycle
Oyster Creek
BWR/3/4 1965/1966
First jet pump application
Dresden 2
Improved Emergency Core Cooling System (ECCS); spray and
Browns Ferry
flood
Reactor Core Isolation Cooling, (RCIC) system
BWR/5 1969
Improved
ECCS
systems
LaSalle
Valve recirculation flow control
9 Mile Point 2
BWR/6
1972
Improved jet pumps and steam separators
Clinton
Reduced fuel duty: 13.4 kW/ft, 44 kW/m
Grand Gulf
Improved ECCS performance
Perry
Gravity containment flooder
Solid-state nuclear system protection system (Option, Clinton
only)
Compact control room option
Source: M. Ragheb, Chapter 3, Boiling Water Reactors, https://netfiles.uiuc.edu/mragheb/www/
NPRE%20402%20ME%20405%20Nuclear%20Power%20Engineering/Boiling%20Water%20Reactors.pdf.
Note: All BWR/1 plants that operated in the United States have been decommissioned.

12 The NRC regulates the design, construction, and operation requirements of the ECCS under 10 C.F.R. 50.46,
“Acceptance criteria for emergency core cooling systems for light-water nuclear reactors”; Appendix K to 10 C.F.R.
Part 50, “ECCS Evaluation Models”; and Appendix A to 10 C.F.R. Part 50, “General Design Criteria [GDC] for
Nuclear Power Plants” (e.g., GDC 35, “Emergency Core Cooling”).
13 N.C. Chokshi, S.K. Shaukat, and A.L. Hiser, et al., Seismic Considerations for the Transition Break Size, U.S.
Nuclear Regulatory Commission, NUREG 1903, Brookhaven National Laboratory, February 2008.
Congressional Research Service
5

---

Nuclear Power Plant Design and Seismic Safety Considerations

Figure 2. GE BWR / Mark I Containment Structure
Showing Torus Suppression Pool

Source: General Electric, in NRC Boiling Water Reactor (BWR) Systems, http://www.nrc.gov/reading-rm/basic-ref/
teachers/03.pdf.
Note: Japan’s Fukushima Daiichi plants use designs similar to this.
Congressional Research Service
6

---

Nuclear Power Plant Design and Seismic Safety Considerations

Figure 3. General Electric Mark II Containment Structure

Source: General Electric, in NRC Boiling Water Reactor (BWR) Systems, http://www.nrc.gov/reading-rm/basic-ref/
teachers/03.pdf.
Congressional Research Service
7

---

Nuclear Power Plant Design and Seismic Safety Considerations

Figure 4. General Electric Mark III Containment Structure

Source: General Electric, in NRC Boiling Water Reactor (BWR) Systems, http://www.nrc.gov/reading-rm/basic-ref/
teachers/03.pdf.
Notes:
Reactor Building
Auxiliary Building
Fuel Building
1. Shield Building
16. Steam Line Channel
19. Spent Fuel Shipping cask
2. Free Standing Steel Containment
17. RHR System
20. Fuel Storage Pool
3. Upper Pool
18. Electrical Equipment Room
21. Fuel Transfer Pool
4. Refueling Platform

22. Cask Loading Pool
5. Reactor Water Cleanup

23. Cask Handling Crane
6. Reactor Vessel

24. Fuel Transfer Bridge
7. Steam Line

25. Fuel Cask Skid on Railroad Car
8. Feed-water Line


9. Recirculation Loop


10. Suppression Pool


11. Weir Wall


12. Horizontal Vent


13. Dry Wel


14. Shield Wall


15. Polar Crane


Pressurized Water Reactor Systems
A pressurized water reactor (PWR) generates steam outside the reactor vessel, unlike a BWR
design. A primary system (reactor cooling system) cycles superheated water from the core to a
Congressional Research Service
8

---

Nuclear Power Plant Design and Seismic Safety Considerations

heat exchanger/steam generator. A secondary system then transfers steam to a combined high-
pressure/ low-pressure turbine generator (Figure 5).14 Steam exhausted from the low-pressure
turbine runs through a condenser that cools and condenses it back to water. Pumps return the
cooled water back to the steam generator for reuse. The condenser cools the steam leaving the
turbine-generator through a third system by flowing past a heat-exchanger that recycles cooling
water through a cooling tower, or takes in and discharges water with a lake, river, or ocean.
Unlike a BWR design, the cooling water that flows through the reactor core never contacts the
turbine-generator. Nor does reactor cooling water contact the environment under normal
operating conditions.
Figure 5. Pressurized Water Reactor (PWR) Plant
Generic Design Features

Source: U.S. Nuclear Regulatory Commission, Reactor Concepts Manual, Boiling Water Reactor Systems, 2005.
Notes: PIZ – Pressurizer; S/G – Steam generator
To keep the reactor operating under ideal conditions, a pressurizer keeps water and steam
pressure under equilibrium conditions. The pressurizer is part of the reactor coolant system, and
consists of electrical heaters, pressure sprays, power-operated relief valves, and safety valves. For
example, if pressure rises too high, water spray cools the steam in the pressurizer; or if pressure is
too low, the heaters increase steam pressure. The cause of the pressure deviation is normally
associated with a change in the temperature of the reactor coolant system.
PWR Design Configurations
All PWR systems consist of the same major components, but arranged and designed differently.
For example, Westinghouse has built plants with two, three, or four primary coolant loops,
depending upon the power output of the plant.

14 U.S. NRC, Reactor Concepts Manual, Pressurized Water Reactor Systems, http://www.nrc.gov/reading-rm/basic-
ref/teachers/04.pdf - 2005-10-17.
Congressional Research Service
9

---

Nuclear Power Plant Design and Seismic Safety Considerations

• Two-loop Westinghouse reactors have two steam generators, two reactor coolant
pumps, a pressurizer, and 121 fuel assemblies; electrical output is approximately
500 megawatts. Six currently operate.15
• Three-loop Westinghouse reactors have three steam generators, three reactor
coolant pumps, a pressurizer, and 157 fuel assemblies; output ranges from 700 to
more than 900 megawatts. Thirteen currently operate.16
• Four-loop Westinghouse reactors have four steam generators, four reactor coolant
pumps, a pressurizer, and 193 fuel assemblies; output ranges from 950 to 1,250
megawatts.17 Twenty-nine currently operate.
The seven operating Babcock & Wilcox reactors have two once-through steam generators, four
reactor coolant pumps, and a pressurizer.18 These reactors have 177 fuel assemblies and produce
approximately 850 megawatts of electricity.
The 14 operating Combustion Engineering reactors have two steam generators, four reactor
coolant pumps, and a pressurizer.19 They produce from less than 500 to more than 1,200
megawatts.
Safe Shutdown Condition
During normal operation, a PWR does not generate steam directly. For cooling, it transfers heat
via the reactor primary coolant to a secondary coolant in the steam generators. There, the
secondary coolant water is boiled into steam and sent to the main turbine to generate electricity.
Even after shutdown (when the moderated uranium fission is halted), the reactor continues to
produce a significant amount of heat from decay of uranium fission products (decay heat). The
decay heat is sufficient to cause fuel damage if the core cooling is inadequate. Auxiliary feed-
water systems and the steam dump systems work together to remove the decay heat from the
reactor. If a system for dumping built-up steam is not available or inoperative, atmospheric relief
valves can dump the steam directly to the atmosphere. Under normal operating conditions, water
flowing through the secondary system does not contact the reactor core; dumped-steam does not
present a radiological release.
Loss of Coolant Accident
The most severe operating condition that reactor designs must contend with is the loss of coolant
accident (LOCA); the extreme case represented by the double-ended guillotine break (DEGB) of

15 The two-loop units in the United States are Ginna, Kewaunee, Point Beach 1 and 2, and Prairie Island 1 and 2.
16 The three-loop units in the United States are Beaver Valley 1 and 2, Farley 1 and 2, H. B. Robinson 2, North Anna 1
and 2, Shearon Harris 1, V. C. Summer, Surry 1 and 2, and Turkey Point 3 and 4.
17 The four-loop units in the United States are Braidwood 1 and 2, Byron 1 and 2, Callaway, Catawba 1 and 2,
Comanche Peak 1 and 2, D. C. Cook 1 and 2, Diablo Canyon 1 and 2, Indian Point 2 and 3, McGuire 1 and 2, Millstone
3, Salem 1 and 2, Seabrook, Sequoyah 1 and 2, South Texas Project 1 and 2, Vogtle 1 and 2, Watts Bar 1, and Wolf
Creek.
18 The Babcock & Wilcox units in the United States are Arkansas 1, Crystal River 3, Davis Besse, Oconee 1, 2, and 3,
and Three Mile Island 1.
19 The Combustion Engineering units in the United States are Arkansas 2, Calvert Cliffs 1 and 2, Fort Calhoun,
Millstone 2, Palisades, Palo Verde 1, 2, and 3, San Onofre 2 and 3, Saint Lucie 1 and 2, and Waterford 3.
Congressional Research Service
10

---

Nuclear Power Plant Design and Seismic Safety Considerations

large diameter pipe systems. In the event of a LOCA, the reactor’s emergency core cooling
system (ECCS) provides core cooling to minimize fuel damage by injecting large amounts of
cool, borated water into the reactor coolant system from a storage tank. The borated water stops
the fission process by absorbing neutrons, and thus aids in shutting down the reactor.
The ECCS on the PWR consist of four separate systems: the high-pressure injection (or charging)
system, the intermediate pressure injection system, the cold leg accumulators, and the low-
pressure injection system (residual heat removal). The high pressure injection system provides
water to the core during emergencies in which reactor coolant system pressure remains relatively
high (such as small breaks in the reactor coolant system, steam break accidents, and leaks of
reactor coolant through a steam generator tube to the secondary side). The intermediate pressure
injection system is designed to accommodate emergency conditions under which the primary
pressure stays relatively high; for example, small to intermediate size primary breaks. The cold
leg accumulators operate without electrical power by using a pressurized nitrogen gas bubble on
the top of tanks that contain large amounts of borated water. The low-pressure injection system
removes residual heat by injecting water from the refueling water storage tank into the reactor
coolant system during large breaks (which would cause very low reactor coolant-system
pressure).
Containment Structure Designs
All U.S. reactors are surrounded by a primary containment structure that is designed to minimize
releases of radioactive material into the environment. The PWR primary containment structure
must surround all the components of the primary cooling system, including the reactor vessel,
steam generators, and pressurizer. BWR primary containments typically are smaller, because
there are no steam generators or pressurizers.
Containments must be strong enough to withstand the pressure created by large amounts of steam
that may be released from the reactor cooling system during an accident. The largest
containments are designed to provide sufficient space for steam released by an accident to expand
and cool to keep pressure within the design parameters of the structure. Smaller containments,
such as those for BWRs, require pressure suppression systems to condense much of the released
steam into water. Smaller PWR containments also may include pressure suppression systems,
such as ice condensers.20
To further limit the leakage from the containment structure following an accident, a steel liner
that covers the inside surface of the containment building acts as a vapor-proof membrane to
prevent any gas from escaping through any cracks that may develop in the concrete of the
containment structure. Two systems act to reduce temperature and pressure within the
containment structure: a fan cooler system that circulates air through heat exchangers, and a
containment spray system.
All U.S. PWR designs include a containment system with Multiple Engineered Safety Features
(ESFs).21 A dry containment system consists of a steel shell surrounded by a concrete biological

20 Kazys Almenas and R. Lee, Nuclear Engineering: An Introduction (Berlin: Springer-Verlag, 1992), pp. 507-514.
21 M. Ragheb, Containment Structures (2011). University of Illinois Champaign-Urbana, https://netfiles.uiuc.edu/
mragheb/www/NPRE%20457%20CSE%20462%20Safety%20Analysis%20of%20Nuclear%20Reactor%20Systems/
Containment%20Structures.pdf.
Congressional Research Service
11

---

Nuclear Power Plant Design and Seismic Safety Considerations

shield that protects the reactor against outside elements, for example, debris driven by hurricane
winds or an aircraft strike.22 The outer shield is not designed as a barrier against the release of
radiation. Although the concrete structures in existing plants act as insulators against uncontrolled
releases of radioactivity to the environment, they will fail if the ESFs fail in their function. Some
containment building design features are summarized in Table 3.
Table 3. Containment Building Design Parameters
Containment Type, plant
Parameter Technical
Specification
Containment capability pressure
149 psiaa
Upper bound spike pressure
107 psia
Early failure physical y unreasonable
10 psi/hour
SP-1, Zion
best estimate pressure rise, including
heat sinks
Time to failure, best estimate with
16 hours
unlimited water in cavity



Containment capability pressure
134 psia
Upper bound spike pressure
107 psia
SP-2, Surry
Time to failure, early failure
Several days
physical y unreasonable best estimate
with dry cavity



Containment capability pressure
65 psia, 330 ºF
Upper bound loading pressure
70-100 psia
SP-3, Sequoyah
Lower bound loading pressure
50-70 psia
Thermal loads
500-700 ºF
Early failure
Quite likely



Containment capability pressure
132 psia, 330 ºF
Upper bound loading pressure
132 psia in 40 minutes
SP-4, Browns Ferry
Lower bound loading pressure
132 psia in 2 hours
Thermal loads
500-700 ºF
Early failure
Quite likely



Containment capability pressure
75 psia
Upper bound loading pressure
30 psia
Wal heat flux
1,000 to 10,000 Btu/hr-square foot
SP-6, Grand Gulf
Penetration seal temperature
345 ºF
Pressurization failure from diffusion
Unreasonable
flames
Seal failure
Unlikely



Containment capability pressure
155 psia, 330 ºF
Upper bound loading pressure
145 psia in 2-3 hours
SP-15, Limerick
Lower bound loading pressure
100 psia in 3 hours
Thermal loads
500-700 ºF
Early failure
Rather unlikely



Source: U.S. NRC, General Studies of Nuclear Reactors; BWR Type Reactors; Containment; Reactor Accidents; Leaks;
PWR Type Reactors; Accidents; Reactors; Water Cooled Reactors; Water Moderated Reactors, NUREG-1037, 1985, as
cited by M. Ragheb UICU.

22 NRC regulations require that new reactors be designed to withstand the impact of large commercial aircraft and that
existing plants develop strategies to mitigate the effects of large aircraft crashes. See CRS Report RL34331, Nuclear
Power Plant Security and Vulnerabilities, by Mark Holt and Anthony Andrews.
Congressional Research Service
12

---

Nuclear Power Plant Design and Seismic Safety Considerations

Notes: NUREG-1037 was never released, but draft versions were apparently circulated.
a. psia = pounds per square inch atmospheric.
The NRC Containment Performance Working Group studied containment buildings in 1985 to
estimate their potential leak rates as a function of increasing internal pressure and temperature
associated with severe accident sequences involving significant core damage.23 It indentified
potential leak paths through containment penetration assemblies (such as equipment hatches,
airlocks, purge and vent valves, and electrical penetrations) and their contributions to leakage
from for the containment. Because the group lacked reliable experimental data on the leakage
behavior of containment penetrations and isolation barriers at pressures beyond their design
conditions, it relied on an analytical approach to estimate the leakage behavior of components
found in specific reference plants that approximately characterize the various containment types.
Nuclear Power Plants Operating in the United States
The locations of all 104 nuclear power plants operating in the United States are shown on the map
in Figure 6.


23 U.S. Nuclear Regulatory Commission, General Studies of Nuclear Reactors; BWR Type Reactors; Containment;
Reactor Accidents; Leaks; PWR Type Reactors; Accidents; Reactors; Water Cooled Reactors; Water Moderated
Reactors, NUREG-1037, May 1, 1985.
Congressional Research Service
13

---

Nuclear Power Plant Design and Seismic Safety Considerations

Figure 6. Commercial Nuclear Power Plants Operating in the United States
One hundred and four (104) Operating Reactors

Source: Prepared by the Library of Congress Geography and Maps Division for CRS using U.S. NRC Find Operating Nuclear Reactors by Location or Name,
http://www.nrc.gov/info-finder/reactor/index.html#AlphabeticalList.
CRS-14

---

Nuclear Power Plant Design and Seismic Safety Considerations

Notes:
Unit Type
MW
Vendor
St.
Lic.
Unit Type MW Vendor St. Lic. Unit
Type MW Vendor St. Lic.
Arkansas Nuclear 1 PWR
843 B&W
AK 1974 Grand Gulf 1
BWR 1,297 GET6
MS 1984 Point Beach 1
PWR
512 W2L
WI 1970
Arkansas Nuclear 2 PWR
995 CE
AK 1974 Hatch 1
BWR
876 GET4
GA 1974 Point Beach 2
PWR
514 W2L
WI 1973
Beaver Val ey 1
PWR
892 W3L
PA 1976 Hatch 2
BWR
883 GET4
GA 1978 Prairie Island 1
PWR
551 W2L
MN 1874
Beaver Val ey 2
PWR
846 W3L
PA 1987 Robinson 2
PWR
710 W3L
SC 1970 Prairie Island 2
PWR
545 W2L
MN 1974
Braidwood 1
PWR 1,178 W4L
IL
1987 Hope Creek 1
BWR 1,061 GET4
NJ 1986 Quad Cities 1
BWR
867 GET3
IL
1972
Braidwood 2
PWR 1,152 W4L
IL
1988 Indian Point 2
PWR 1,023 W4L
NY 1973 Quad Cities 2
BWR
869 GET3
IL
1972
Browns Ferry 1
BWR 1,065 GET4
AL 1973 Indian Point 3
PWR 1,025 W4L
NY 1975 R. E. Ginna
PWR
498 W2L
NY 1969
Browns Ferry 2
BWR 1,104 GET4
AL 1974 Joseph M. Farley 1 PWR
851 W3L
AL 1977 River Bend 1
BWR
989 GET6
LA 1985
Browns Ferry 3
BWR 1,115 GET4
AL 1976 Joseph M. Farley 2 PWR
860 W3L
AL 1981 Salem 1
PWR 1,174 W4L
NJ 1976
Brunswick 1
BWR
938 GET4
NC 1976 Kewaunee
PWR
556 W2L
WI 1973 Salem 2
PWR 1,130 W4l
NJ 1981
Brunswick 2
BWR
937 GET4
NC 1974 LaSal e County 1
BWR 1,118 GET5
IL
1982 San Onofre 2
PWR 1,070 CE
CA 1982
Byron 1
PWR 1,164 W4L
IL
1985 LaSal e County 2
BWR 1,120 GET5
IL
1983 San Onofre 3
PWR 1,080 CE
CA 1992
Byron 2
PWR 1,136 W4L
IL
1987 Limerick 1
BWR 1,134 GET4
PA 1985 Seabrook 1
PWR 1,295 W4L
NH 1990
Cal away 1
PWR 1,236 WFL
MO 1984 Limerick 2
BWR 1,134 GET4
PA 1989 Sequoyah 1
PWR 1,148 W4L
TN 1980
Calvert Cliffs 1
PWR
873 CE
MD 1974 McGuire 1
PWR 1,100 W4L
NC 1981 Sequoyah 2
PWR 1,126 W4L
TN 1981
Calvert Cliffs 2
PWR
862 CE
MD 1976 McGuire 2
PWR 1,100 W4L
NC 1983 Shearon Harris 1 PWR
900 W3L
NC 1986
Catawba 1
PWR 1,129 W4L
SC 1985 Millstone 2
PWR
884 CE
CT 1975 South Texas 1
PWR 1,410 W4L
TX 1988
Catawba 2
PWR 1,129 W4L
SC 1986 Millstone 3
PWR 1,227 W4L
CT 1986 South Texas 2
PWR 1,410 W4L
TX 1989
Clinton 1
BWR 1,065 GET6
IL
1987 Monticel o
BWR
579 GET3
MN 1970 St. Lucie 1
PWR
839 CE
FL 1976
Columbia Gen. St. BWR 1,190 GET5
WA 1984 Nine Mile Pt .1
BWR
621 GET2
NY 1974 St. Lucie 2
PWR
839 CE
FL 1983
Comanche Peak 1 PWR 1,200 W4L
TX 1990 Nine Mile Pt. 2
BWR 1,140 GET5
NY 1987 Surry 1
PWR
799 W3L
VA 1972
Comanche Peak 2 PWR 1,150 W4L
TX 1993 North Anna 1
PWR
981 W3L
VA 1978 Surry 2
PWR
799 W3l
VA 1973
Cooper Station
BWR
830 GET4
NE 1974 North Anna 2
PWR
973 W3L
VA 1980 Susquehanna 1 BWR 1,149 GET4
PA 1982
Crystal River 3
PWR
838 B&WLL FL
1976 Oconee 1
PWR
846 B&WLL SC 1973 Susquehanna 2
BWR 1,140 GET4
PA 1984
Davis-Besse
PWR
893 B&WLL OH 1977 Oconee 2
PWR
846 B&WLL SC 1973 Three Mile Isl. 1 PWR
786 B&WLL PA 1974
Diablo Canyon 1
PWR 1,151 W4L
CA 1984 Oconee 3
PWR
846 B&WLL SC 1974 Turkey Point 3
PWR
720 W3L
FL 1972
Diablo Canyon 2
PWR 1149 W4L
CA 1985 Oyster Creek
BWR
619 GET2
NJ 1991 Turkey Point 4
PWR
720 W3l
FL 1973
Donald C. Cook 1 PWR 1,009 W4L
MI 1974 Palisades
PWR
778 CE
MI 1971 VC Summer
PWR
966 W3l
SC 1982
Donald C. Cook 2 PWR 1,060 W4L
MI 1977 Palo Verde 1
PWR 1,335 CES80
AZ 1985 Vermont Yankee BWR
510 GET4
VT 1972
Dresden 2
BWR
867 GET3
IL
1991 Palo Verde 2
PWR 1,335 CES80
AZ 1986 Vogtle 1
PWR 1,109 W4L
GA 1987
Dresden 3
BWR
867 GET3
IL
1971 Palo Verde 3
PWR 1,335 CES80
AZ 1987 Vogtle 2
PWR 1,127 W4L
GA 1989
Duane Arnold
BWR
640 GET4
IA
1974 Peach Bottom 2
BWR 1,112 GET4
PA 1973 Waterford 3
PWR 1,250 CE
LA 1985
Fermi 2
BWR 1,122 GET4
MI 1985 Peach Bottom 3
BWR 1,112 GET4
PA 1974 Watts Bar 1
PWR 1,123 W4l
TN 1996
Fitzpatrick
BWR
852 GET4
NY 1974 Perry 1
BWR 1,261 GET6
OH 1986 Wolf Creek 1
PWR 1,166 W4L
KS 1985
Fort Calhoun
PWR
500 CE
NE 1973 Pilgrim 1
BWR
685 GET3
MA 1972



Notes: No commercial nuclear power plants operate in Alaska or Hawaii. B&W: Babcock & Wilcox 2-Loop Lower; CE: Combustion Engineering; CE80: Combustion
Engineering System 80; W2L Westinghouse 2-Loop; W3L Westinghouse 3-Loop; W4L Westinghouse 4-Loop; GET2: General Electric Type 2; GET3: General Electric
Type 3; GET4: General Electric Type 4; GET5: General Electric Type 5; GET6: General Electric Type 6.

CRS-15

---

Nuclear Power Plant Design and Seismic Safety Considerations

Plant Seismic Siting Criteria
Earthquakes occur when stresses in the earth exceed the strength of a rock mass, creating a fault
or mobilizing an existing fault.24 The fault can slip laterally (a strike/slip fault, such as the San
Andreas Fault), move vertically (a thrust or reverse fault, such as the fault that caused the March
11 Japanese earthquake), or move in some combination of the two. The fault’s sudden release
sends seismic shock waves through the earth that have two primary characteristics: amplitude—a
measure of the peak wave height, and period—the time interval between the arrival of successive
peaks or valleys.25 The seismic wave’s arrival causes ground motion. The ground motion intensity
depends on three factors: the distance from the source (also known as focus or epicenter), the
amount of energy released (magnitude of the earthquake), and the type of soil or rock at the site.
The shallower the earthquake’s focus, the stronger the waves will be when they reach the surface.
Generally, the intensity of ground shaking diminishes with increasing distance from the
earthquake focus. The earthquake’s magnitude (M) is measured on a logarithmic scale
(sometimes referred to as the Richter scale), thus an M 7.0 earthquake has amplitude that is ten
times larger than an M 6.0, but releases 31.5 times more energy than an M 6.0 earthquake. Sites
with deep, soft soils or loosely compacted fill will experience stronger ground motion than sites
with stiff soils, soft rock, or hard rock.
Refer to Appendix A of this report for additional discussion on magnitude, Richter scale, and
intensity. For more detailed information about earthquake hazards, refer to CRS Report
RL33861, Earthquakes: Risk, Detection, Warning, and Research, by Peter Folger.
General Design Criteria
For nuclear power plants granted construction permits during the 1960s and 1970s, a design
approach emerged for considering seismic loads based on site-specific investigations of local and
regional seismology, geology and geotechnical engineering.26 The 1973 publication of 10 C.F.R.
100, Appendix A—Seismic and Geologic Siting Criteria for Nuclear Power Plants, included the
concept of a “safe shutdown earthquake” (SSE), which is discussed in a later section of this
report.
General design criteria for nuclear power plants require that structures and components important
to safety be designed to withstand the effects of earthquakes, tornados, hurricanes, floods,
tsunamis, and seiche27 waves without losing the capability to perform their safety function. These
“safety-related” structures, systems, and components are those necessary to assure:
1. the integrity of the reactor coolant pressure boundary,

24 The Applied Technology Council (ATC) and the Structural Engineers Association of California (SEAOC), Briefing
Paper 1 Building Safety and Earthquakes Part A: Earthquake Shaking and Building Response, Redwood City, CA,
http://www.atcouncil.org/.
25 The wave’s frequency is the inverse of the period (1/s), and is expressed as the number of wave cycles per second
(termed Hertz or Hz).
26 U.S. Nuclear Regulatory Commission, Evaluation of the Seismic Design Criteria in ASCE/SEI Standard 43-05 for
Application to Nuclear Power Plants, NUREG/CR-6926, Brookhaven National Laboratory, NY, March 2007.
27 Standing waves, or waves that move vertically but not horizontally. Seiche waves can be triggered by earthquakes,
strong winds, tides, and other causes.
Congressional Research Service
16

---

Nuclear Power Plant Design and Seismic Safety Considerations

2. the capability to shut down the reactor and maintain it in a safe condition, or
3. the capability to prevent or mitigate the consequences of accidents, which could
result in potential offsite exposures.
Refer to this report’s section on “Nuclear Power Plant Designs” for some discussion of safety-
related components.
The language in 10 C.F.R. 100, Appendix A, notes that the seismic criteria are based on limited
geophysical and geologic information, available at the time, on faults and earthquake
occurrences, and that the information would be revised when more information became available.
The information is based on a review of historical records and a site investigation. Ultimately, the
investigation provides the basis for determining a “safe shutdown earthquake,” alternately
referred to as the “design basis earthquake,” defined as the maximum vibratory ground motion for
which certain structures, systems, and components are designed to remain functional. Under an
“operating basis earthquake,” the reactor could continue operation without undue risk to the
safety of the public.
The NRC subsequently published a series of Regulatory Guides in support of Appendix A of 10
C.F.R. 100. These guides provide technical information, procedures, and design criteria that are
beyond the scope of this report.
• Regulatory Guide 1.60, Design Response Spectra of Nuclear Power Reactors
(1973), provides ground design response spectral shapes for horizontal and
vertical ground movements developed from a statistical analysis of response
spectra of past Western United States (WUS) strong-motion earthquakes
collected from a variety of different site conditions, primarily at deep soil sites.
• Regulatory Guide 1.165, Identification and Characterization of Seismic Sources
and Determination of Safe Shutdown Earthquake Ground Motion (1997)
,provided procedures for (1) conducting geological, geophysical, seismological,
and geotechnical investigations, (2) identifying and characterizing seismic
sources, (3) conducting probabilistic seismic hazard analysis (PSHA), and (4)
determining the safe shutdown earthquake for satisfying the requirements of 10
C.F.R. 100.23. The guide evolved out of investigations into seismic hazard
estimates for nuclear power plant sites operating in the Central and Eastern
United States (CEUS).
• NUREG/CR-6926, Evaluation of the Seismic Design Criteria in ASCE/SEI
Standard 43-05 for Application to Nuclear Power Plants (2007), provided
seismic design criteria for safety-related structures, systems, and components in a
broad spectrum of nuclear facilities.28
Site Investigations
The site investigations required under 10 C.F.R. 100, Appendix A, starts with a review of pertinent
literature and progresses to field investigations. The required investigations include:

28 Based on a review by the American Society of Civil Engineers/Structural Engineering Institute (ASCE/SEI)
Standard 43-05 - Seismic Design Criteria for Structures, Systems, and Components in Nuclear Facilities.
Congressional Research Service
17

---

Nuclear Power Plant Design and Seismic Safety Considerations

• Vibratory Ground Motion—examines lithology, stratigraphy, structural geology,
underlying tectonic structures, physical earthquake evidence, engineering
properties of underlying soil and rock, historically reported earthquakes,
earthquake epicenters within 200 miles of site, faults within 200 miles.
• Surface Faulting—evaluates lithology, stratigraphy, structural geology,
underlying tectonic structures, evidence of fault offsets, nearby faults greater than
1,000 feet in length, records of earthquakes associated with faults greater than
1,000 feet in length, epicenters of earthquakes with faults greater than 1,000 feet
in length.
• Seismically Induced Floods and Water Waves ─ looks at reports or evidence of
distantly and locally generated waves or tsunamis which have or could have
affected the site, and evidence for seismically induced floods and water waves
that have or could have affected the site.
Safe Shutdown Earthquake Condition
The NRC defines the Safe Shutdown Earthquake as the maximum earthquake potential for which
certain structures, systems, and components, important to safety, are designed to sustain and
remain functional.29 During an earthquake, ground motion sets up vibrations in a nuclear power
plant’s foundation and structure. In simple terms, the vibrations represent the back-and-forth
acceleration of an object (the distance moved is the amplitude). Vibration, or horizontal ground
acceleration, is measured in terms of the earth’s gravitational acceleration constant (g) for
structural design purposes.30 These vibrations place additional loads and displacements on the
nuclear power plant’s structure, equipment and piping systems. The additional loading must be
accounted for in the structural design of the piping systems supports.
Various plant structures, depending upon their elevation above the foundation, vibrate at different
frequencies during an earthquake. Low frequency vibrations in the range of 1 to 10 Hz (cycles per
second) are particularly problematic for a wide range of structures because such structures are
often susceptible to damaging resonance at those frequencies. These accelerations and the
corresponding shaking frequencies are used in the probabilistic seismic hazard analysis (PSHA)
discussed in this report’s “Background on Seismic Standards” section. The full seismic spectrum
can be characterized by two intervals: peak ground acceleration (PGA) and spectral acceleration
(SA) averaged between 5 and 10 Hz. PGA has been widely used to develop nuclear power plant
“fragility estimates” and represents the performance of nuclear plant structures, systems, and
components (SSCs) that are sensitive to inertial effects.
The maximum vibratory accelerations of the Safe Shutdown Earthquake must take into account
the characteristics of the underlying soil material in transmitting the earthquake-induced motions
at the various locations of the plant’s foundation. A multiple degree-of-freedom analysis is used to
simulate the effect of the earthquake on the piping systems.
Experimental and empirical seismic data have provided insights into the behavior of different
structures under various acceleration and shaking conditions. One conclusion reached regarded

29 http://www.nrc.gov/reading-rm/basic-ref/glossary/safe-shutdown-earthquake.html
30 Gravitation acceleration g = 32 feet/second/second (ft/second2).
Congressional Research Service
18

---

Nuclear Power Plant Design and Seismic Safety Considerations

the performance of welded steel piping at power plants during strong motion earthquakes.
Relatively small numbers of failures occurred when peak ground accelerations remained below
0.5g.31 Other types of structures would exhibit different behaviors, and engineers design the
various plant structures to withstand a certain severity of earthquake specific to each plant site.
The example of Figure 7 shows areas susceptible to shaking of a frequency of 5 Hz having a 5%
probability of occurring at least once within 50 years.32 The map shows the strength of the
expected acceleration (in g) for areas experiencing such an earthquake. The darker colors on the
map indicate areas of strongest shaking.
Figure 7. Spectral Acceleration 5 Hz
Return Period of 5% in 50 Years

Source: USGS National Seismic Hazard maps, USGS Open-File Report 2008-1128, 2008,
http://earthquake.usgs.gov/hazards/.
Notes: Areas that are susceptible to shaking at a frequency of 5 Hz with a 5% probability of occurring at least
once within 50 years. The strength of the expected acceleration is expressed in terms of earth’s gravitational

31 N.C. Chokshi, S.K. Shaukat, and A.L. Hiser, et al., Seismic Considerations for the Transition Break Size, U.S. NRC,
NUREG-1903, February 2008, pp. 29-30.
32 This collection of USGS seismic hazard maps includes probabilistic ground motion maps for Peak Ground
Acceleration (PGA), 1Hz (1.0 second SA), and 5Hz (0.2 second SA). (Refer to the report section on “Safe Shutdown
Earthquake” for a discussion of spectral acceleration.) Some additional spectral accelerations (SA) are also included for
central and southern California. Most figures correspond to the 2% in 50-year probability of exceedance, but there are a
few figures for the 10% in 50 year and the 5% in 50-year probability of exceedance as well a range of accelerations and
associated probabilities.
Congressional Research Service
19

---

Nuclear Power Plant Design and Seismic Safety Considerations

acceleration constant (g) for areas experiencing such an earthquake. The darker colors on the map indicate
areas of strongest shaking.
National Seismic Hazard Maps
In 2008, the U.S. Geological Survey (USGS) released an update of the National Seismic Hazard
Maps (NSHM).33 The purpose of the maps is to show the likelihood of a particular severity of
shaking within a specified time-period. The Seismic Hazard maps are the basis for seismic design
provisions of building codes to allow buildings, highways, and critical infrastructure to withstand
earthquake shaking without collapse. The NRC requires that every nuclear plant be designed for
site-specific ground motions that are appropriate for their site locations. In addition, the NRC has
specified a minimum ground motion level to which nuclear plants must be designed. (See
discussion above on design criteria.)
The USGS revises the NHSM every six years to reflect newly published earthquake data to
update building code seismic design provisions. USGS notes that the 2008 hazard maps differ
significantly from the 2002 maps in many parts of the United States:
The new maps generally show 10- to 15-percent reductions in acceleration across much of
the Central and Eastern United States [CEUS] for 0.2-s [second] and 1.0-s spectral
acceleration and peak horizontal ground acceleration for 2-percent probability of exceedance
in 50 years. The new maps for the Western United States [WUS] indicate about 10-percent
reductions for 0.2-s spectral acceleration and peak horizontal ground acceleration and up to
30-percent reductions in 1.0-s spectral acceleration at similar hazard levels.34
In the Central and Eastern United States (CEUS), the New Madrid Seismic Zone and the
Charleston area in southeast South Carolina comprise the dominant seismic hazard (at 2%
probability of exceedance in 50 years). Seismically active portions of eastern Tennessee and some
portions of the northeast also contribute to the seismic hazard. The hazard at the 2% probability
of exceedance in 50 years level is typically a factor of two to four times higher than the 10%
probability of exceedance in 50 years values in the seismically active portions of the CEUS.
Seismic hazards are greatest in the Western United States (WUS), particularly in California,
Oregon, and Washington, as well as Alaska and Hawaii. The hazard at the 2% probability of
exceedance in 50 years level is typically a factor of 1.5 to 2 times higher than the 10% in 50 years
values in coastal California and from 2 to 3.5 higher across the rest of the WUS.
CRS has mapped the proximity of plant sites to seismic hazards based on the USGS National
Seismic Hazard Map for the United States in Figure 8. This map displays quantitative
information about seismic ground motion hazards as horizontal ground acceleration (g) of a
particle at ground level moving horizontally during an earthquake.
CRS has also mapped the proximity of plant sites to Quaternary period faults based on the USGS
Quaternary Fault and Fold Database of the United States in Figure 9. The USGS Database has
information on faults and associated folds in the United States that are believed to be sources of

33 Mark D. Petersen, Arthur D. Frankel, and Stephen C. Harmsen, et al., Documentation for the 2008 Update of the
United States National Seismic Hazard Maps, U.S Geological Survey, Open-File Report 2008-1128, 2008,
http://earthquake.usgs.gov/hazards/.
34 Ibid.
Congressional Research Service
20

---

Nuclear Power Plant Design and Seismic Safety Considerations

greater than magnitude 6 earthquakes during the past 1,600,000 years ─ the Quaternary period of
the geologic time scale. The map is not a prediction of an earthquake event.

Congressional Research Service
21

---

Nuclear Power Plant Design and Seismic Safety Considerations

Figure 8. Operating Nuclear Power Plant Sites and Seismic Hazard
Seismic hazard expressed as horizontal ground acceleration (shown as a percent of gravity)

Source: Background map USGS Seismic Hazard Map for the United States, prepared for CRS by the Library of Congress Geography and Maps Division.
Notes: This map displays quantitative information about seismic ground motion hazards as horizontal ground acceleration (in terms of gravitational acceleration) of a
particle at ground level moving horizontal y during an earthquake. This map is not a prediction of an earthquake event. The NRC does not rank nuclear plants by seismic
risk. No commercial nuclear power plants operate in either Alaska or Hawaii.
CRS-22

---

Nuclear Power Plant Design and Seismic Safety Considerations

Figure 9. Operating Nuclear Power Plant Sites and Mapped Quaternary Faults

Source: CRS and the USGS Quaternary Fault and Fold Database of the United States.
Notes: To map the proximity of plant sites to faults, CRS referred to the USGS Quaternary Fault and Fold Database of the United States. This is information on faults and
associated folds in the United States that are believed to be sources of greater than moment magnitude 6 (M>6) earthquakes during the Quaternary (the past 1,600,000
years). This map is not a prediction of an earthquake event. No commercial nuclear power plants operate in either Alaska or Hawaii.
CRS-23

---

Nuclear Power Plant Design and Seismic Safety Considerations

NRC Priority Earthquake Safety Review
The NRC has required that each nuclear plant be built to certain structural specifications based on
the earthquake susceptibility of each plant site, but some of those design specifications may be re-
evaluated in light of new seismic analysis in the United States. In 2010 the NRC published GI-
199 Safety/Risk Assessment, a two-stage assessment that determines the implications of USGS
updated probabilistic seismic hazards in the Central and Eastern U.S. (CEUS) on existing nuclear
power plant sites.35 The assessment first evaluated the change in seismic hazard with respect to
previous estimates at individual NPPs, and then estimated the change in Seismic Core Damage
Frequency (SCDF) resulting from change in the seismic hazard. Seismic core damage frequency
is the probability of damage to the reactor core (fuel rods) resulting from a seismic initiating
event. It does not imply either a core meltdown or the loss of containment, which would be
required for radiological release to occur. The seismic hazard at each plant site depends on the
unique seismology and geology surrounding the site. Consequently, the report separately
determined the implications of updated probabilistic seismic hazard for each of the 96 operating
NPPs in the CEUS.36
The NRC does not rank nuclear plants by seismic risk. NRC’s objective in the GI-199 Safety/Risk
Assessment was to evaluate the need for further investigations of seismic safety for operating
reactors in the CEUS. The data evaluated in the assessment suggest that the probability for
earthquake ground motion above the seismic design basis for some nuclear plants in the CEUS,
although still low, is larger than previous estimates. In late March 2011, the NRC announced that
it had identified 27 nuclear reactors operating in the CEUS that would receive priority earthquake
safety reviews.37 Those 27 reactors are listed in Table 4.
Table 4. Operating Nuclear Power Plants Subject to Earthquake Safety Reviews
Plant
St. Type Plant
St. Type Plant
St. Type
Crystal River 3
FL
PWR North Anna 1 & 2
VA
PWR Sequoyah 1 & 2 TN PWR
Dresden 2 & 3
IL
BWR Oconee 1, 2 & 3
SC
PWR
Seabrook
NH PWR
Duane
Arnold
IA BWR Perry
1
OH BWR V.C.
Summer SC PWR
Joseph M. Farley 1 & 2 AL
PWR Peach Bottom 2 & 3 PA
BWR Watts Bar 1
TN PWR
Indian Point 2 & 3
NY PWR River Bend 1
LA
BWR Wolf Creek
KS
PWR
Limerick 1 & 2
PA
BWR Saint Lucie 1 & 2
FL
PWR


Source: The Energy Daily.
Note: The NRC has not announced a schedule for completing the seismic reviews at the time of this report.

35 U.S. Nuclear Regulatory Commission, Implications of Updated Probabilistic Seismic Hazard Estimates in Central
and Eastern United States Existing Plants—Safety/Risk Assessment, Generic Issue 199 (GI-199), August 2010.
36 Ibid.
37 George Lobsenz, “NRC Task Force To Review Safety: 27 Reactors Are Seismic Priorities,” The Energy Daily,
March 24, 2011.
Congressional Research Service
24

---

Nuclear Power Plant Design and Seismic Safety Considerations

Recent Legislative Activities
Within a few days following Japan’s nuclear crisis, Democrats on the House Energy and
Commerce Committee requested a hearing on U.S. Nuclear Power Plant Safety and
Preparedness.38
On March 17, 2011, the Senate Committee on Homeland Security and Governmental Affairs held
a hearing on Catastrophic Preparedness that looked at technologies and emergency procedures
used in the event of a large-scale earthquake or other natural disaster.39 On April 6, 2011, the
Subcommittee on Oversight and Investigations of the House Energy and Commerce Committee
held a hearing the U.S. Government Response to the Nuclear Power Plant Incident in Japan.40 On
April 7, 2011, the Subcommittee on Technology and Innovation of the House Science, Space, and
Technology Committee held a hearing on Earthquake Risk Reduction.41
Several bills have been introduced in the 112th Congress that are relevant to either nuclear power
plant safety of earthquake hazard assessment.
S. 646, the Natural Hazards Risk Reduction Act of 2011, would amend the Earthquake Hazards
Reduction Act of 1977 (42 U.S.C. 7704) to add program activities to research and develop
effective methods, tools, and technologies to reduce the risk posed by earthquakes, and authorize
the United States Geological Survey to conduct research and other activities necessary to
characterize and identify earthquake hazards, assess earthquake risks, monitor seismic activity,
and provide real-time earthquake information.
H.R. 1379, the Natural Hazards Risk Reduction Act of 2011, would also amend the Earthquake
Hazards Reduction Act of 1977 (42 U.S.C. 7704) to research and develop effective methods,
tools, and technologies to reduce the risk posed by earthquakes to the built environment,
especially to lessen the risk to existing structures and lifelines.
H.R. 1268, the Nuclear Power Licensing Reform Act of 2011, would amend Section 103 of the
Atomic Energy Act of 1954 (42 U.S.C. 2133), subsection c, by adding at the end the following:
‘Any such renewal shall be subject to the same criteria and requirements that would be applicable
for an original application for initial construction, and the Commission shall ensure that any
changes in the size or distribution of the surrounding population, or seismic or other scientific
data not available at time of original licensing, have not resulted in the facility being located at a
site at which a new facility would not be allowed to be built.

38 House Committee on Energy & Commerce Democrats, Committee Democrats Request Hearing on U.S. Nuclear
Power Plant Safety and Preparedness, http://democrats.energycommerce.house.gov/index.php?q=news/committee-
democrats-request-hearing-on-us-nuclear-power-plant-safety-and-preparedness.
39 Senate Committee on Homeland Security & Governmental Affairs, Catastrophic Preparedness: How Ready is FEMA
for the Next Big Disaster? http://hsgac.senate.gov/public/index.cfm?FuseAction=Hearings.Hearing&Hearing_ID=
a42880b1-22fc-4890-b82c-dd2a369e2aa2
40 House Energy & Commerce Committee, The U.S. Government Response to the Nuclear Power Plant Incident in
Japan, http://energycommerce.house.gov/hearings/hearingdetail.aspx?NewsID=8420.
41 House Committee on Science, Space, and Technology, Subcommittee Reviews Status of U.S. Earthquake
Preparedness, http://science.house.gov/press-release/subcommittee-reviews-status-us-earthquake-preparedness.

Congressional Research Service
25

---

Nuclear Power Plant Design and Seismic Safety Considerations

H.R. 1242, the Nuclear Power Safety Act of 2011, would amend the Atomic Energy Act to revise
regulations to ensure that nuclear facilities licensed under the act can withstand and adequately
respond to an earthquake, tsunami (for a facility located in a coastal area), strong storm, or other
event that threatens a major impact to the facility; a loss of the primary operating power source
for at least 14 days; and a loss of the primary backup operating power source for at least 72 hours.
Congressional Research Service
26

---

Nuclear Power Plant Design and Seismic Safety Considerations

Appendix A. Magnitude, Intensity, and Seismic
Spectrum
Earthquake magnitude is a measure of the strength of the earthquake as determined from
seismographic observations. Magnitude is essentially an objective, quantitative measure of the
size of an earthquake.42 The magnitude can be expressed in various ways based on seismographic
records (e.g., Richter Local Magnitude, Surface Wave Magnitude, Body Wave Magnitude, and
Moment Magnitude). Currently, the most commonly used magnitude measurement is the Moment
Magnitude (M) which is based on the strength of the rock that ruptured, the area of the fault that
ruptured, and the average amount of slip.43 Moment is a physical quantity proportional to the slip
on the fault times the area of the fault surface that slips; it is related to the total energy released in
the earthquake. The moment can be estimated from seismograms (and from geodetic
measurements). The Moment Magnitude provides an estimate of earthquake size that is valid over
the complete range of magnitudes, a characteristic that was lacking in other magnitude scales,
such as the Richter scale.
Because of the logarithmic basis of the scale, each whole number increase in magnitude
represents a tenfold increase in measured amplitude; as an estimate of energy, each whole number
step in the magnitude scale corresponds to the release of about 31 times more energy than the
amount associated with the preceding whole number value.
The Richter magnitude scale was developed in 1935 by Charles F. Richter of the California
Institute of Technology and was based on the behavior of a specific seismograph that was
manufactured at that time. The instruments are no longer in use and therefore the Richter
magnitude scale is no longer used in the technical community. However, the Richter Scale is a
term that is so commonly used by the public that scientists generally just answer questions about
“Richter” magnitude by substituting moment magnitude without correcting the misunderstanding.
The intensity of an earthquake is a qualitative assessment of effects of the earthquake at a
particular location. The intensity assigned is based on observed effects on humans, on human-
built structures, and on the earth’s surface at a particular location. The most commonly used scale
in the United States is the Modified Mercalli Intensity (MMI) scale, which has values ranging
from I to XII in the order of severity. MMI of I indicates an earthquake that was not felt except by
a very few, whereas MMI of XII indicates total damage of all works of construction, either
partially or completely. While an earthquake has only one magnitude, intensity depends on the
effects at each particular location.
Greater magnitude earthquakes are generally associated with greater lengths of fault ruptures.44 A
fault break of 100 miles might be associated with an M8 earthquake, while a break of several
miles might generate an M6 earthquake. The length of the fault break, however, is not directly
proportional to the energy released. The induced amplitude of acceleration (g) does increase with

42 US NRC, NRC frequently asked questions related to the March 11, 2011 Japanese Earthquake and Tsunami.
43 USGS, Measuring Earthquakes, http://earthquake.usgs.gov/learn/faq/?categoryID=2&faqID=23.
44 H. Bolton Seed, I. M. Idriss, and Fred. W. Kiefer, “Characteristics of Rock Motions During Earthquakes,” Journal of
Soil Mechanics and Foundation Division, Proceedings of the American Society of Civil Engineers, September 1969,
pp. 1199-1217.
Congressional Research Service
27

---

Nuclear Power Plant Design and Seismic Safety Considerations

increasing magnitude (M). Various methods have been developed to relate the magnitude of an
earthquake to the amplitude of acceleration it induces, and different methods may result in
significant variations in results.
The seismic spectrum can be characterized by two intervals—peak ground acceleration (PGA)
and spectral acceleration averaged between 5 and 10 Hz (SAAvg5-10). PGA has been widely
used to develop fragility estimates and represents the performance of nuclear plant structures,
systems, and components (SSCs) that are sensitive to inertial effects.
Figure A-1 shows a example of response spectra for several power plants.45 The frequency range,
of 1 to 10 Hz, is the subject of USGS earthquake hazard studies, as discussed above.
Figure A-1. Spectral Acceleration (g) vs. Frequency (Hz)
Curves are response spectral values (5% damping) at an annual exceedance frequency of10-5

Source: NRC Generic Issue -99, Implications of Updated Probabilistic Seismic Hazard Estimates in Central and
Eastern United States on Existing Plants, Figure 1, August 2010.
Notes: For illustrative purposes only. Originally prepared to compare seismic hazard results for four early site
permit submittals. Solid line represents submittals to 1989. Dashed Lines represent Electric Power Research
Institute Seismicity Owners Group Study.
NUREG/CR-6926 references the American Society of Civil Engineers (ASCE) Standard 7-05
Minimum Design Loads for Buildings and Other Structures seismic hazard maps. The maximum
considered earthquake (MCE) is based on spectral accelerations with 2%/50 yr probability (2%
probability of being equaled or exceeded in any single year in 50 years or otherwise stated as a
2% annual exceedance probability). (To obtain the design earthquake spectral response
accelerations (DS) used in structural design, the spectral accelerations are multiplied by 2/3.) At
sites in seismically active regions in the Western United States (WUS), the corresponding DS
hazard is approximately 10%/50 yr (return period of 475 yr). In the Central and Eastern United

45 Frequency Hz (Hertz) refers to the number of cycles per second (which is inverse of the ground motion wave period
─ the time between two wave peaks). Thus, 0.2-s is the equivalent of 5 Hz (1/0.2-s), and 1-s is the equivalent of 1 Hz
(1/1-s).
Congressional Research Service
28

---

Nuclear Power Plant Design and Seismic Safety Considerations

States (CEUS) this hazard is approximately 4%/50 yr (return period of approximately 1,200 yr),
These are due to differences in the typical slopes of seismic hazard curves in the WUS and
CEUS.
Congressional Research Service
29

---

Nuclear Power Plant Design and Seismic Safety Considerations

Appendix B. Terms
Boiling water reactor (BWR) directly generates steam inside the reactor vessel.
Deterministic Seismic Hazard Assessment (DSHA) focuses on a single earthquake event to
determine the finite probability of occurring.
Double-ended guillotine break (DEGB) represents a break of the largest diameter pipe in the
primary system that the emergency core cooling system (ECCS) must be sized to provide
adequate makeup water to compensate for.
Light water reactor systems use ordinary water as a fuel moderator and coolant, and uranium
fuel artificially enriched to 4.5%-5% fissile uranium-235. Includes BWR and PWR types.
Loss of Coolant Accident (LOCA) is the most severe operating condition for a reactor that can
contribute to a reactor core meltdown.
Operating Basis Earthquake is the maximum vibratory ground motion that a reactor could
continue operation without undue risk and safety of the public.
Pressurized water reactor (PWR) uses two major loops to convert the heat generated by the
reactor core into steam outside of the reactor vessel.
Probabilistic Seismic Hazard Assessments (PSHA) attempt to quantify the probability of
exceeding various ground-motion levels at a site given all possible earthquakes.
Safe Shutdown Earthquake (also design basis earthquake) is the maximum vibratory ground
motion at which certain structures, systems, and components are designed to remain functional.
Seismic Core Damage Frequency is the probability of damage to the core resulting from a
seismic initiating event.

Author Contact Information

Anthony Andrews

Specialist in Energy and Defense Policy
aandrews@crs.loc.gov, 7-6843

Acknowledgments
Jacqueline C. Nolan, Library of Congress, Geography and Maps Division
Richard J. Campbell, Specialist in Energy Policy, Congressional Research Service
Peter Folger, Specialist in Energy and Natural Resources Policy, Congressional Research Service

Congressional Research Service
30

---