The seismic design criteria applied to siting commercial nuclear power plants operating in the United States received increased attention following the March 11, 2011, earthquake and tsunami that devastated Japan's Fukushima Daiichi nuclear power station. Since the event, a magnitude 5.8 earthquake near Mineral, VA, on August 23, 2011, precipitated the temporary shutdown of Dominion Power's North Anna nuclear power plant. Some Members of Congress have questioned whether U.S nuclear plants are more vulnerable to seismic threats than previously assessed, particularly given the Nuclear Regulatory Commission's (NRC's) ongoing reassessment of seismic risks at certain plant sites. Currently, 104 commercial nuclear power plants operating in the United States use variations in light water reactor designs and construction. Figure 1 shows the locations of all 104 nuclear power reactors operating in the United States. Light water reactors use ordinary water as a neutron moderator and coolant, and uranium fuel enriched in fissile uranium-235. 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 chain reaction. 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. 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. 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 in the United States 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. 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. In addition to the custom-design features of each plant, designers also had to contend with earthquake hazards unique to each plant site. Designs for structures, systems, and components important to a nuclear power plant operation must withstand earthquakes without losing their intended safety-related function. 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. General design criteria for nuclear power plants require that structures and components important to safety withstand the effects of earthquakes, tornados, hurricanes, floods, tsunamis, and seiche waves without losing the capability to perform their safety function. These "safety-related" structures, systems, and components are those necessary to assure: The capability to maintain the reactor coolant pressure, The capability to shut down the reactor and maintain it in a safe condition, or The capability to prevent or mitigate the consequences of accidents, which could result in potential offsite radiation exposures. All BWR plants operating in the United States use variations of a General Electric design. The more numerous PWR plants use Babcock & Wilcox, Combustion Engineering, and Westinghouse designs. Table 1 summarizes the various reactor types. The sections that follow discuss them further. A boiling water reactor generates steam directly inside the reactor vessel as water flows upward through the reactor's core (see Figure 2 ). 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 water and directly discharges it to 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 570" F and pressures of 1,000 pounds per square inch (psi) atmospheric. In the case of events that cause a nuclear power plant to exceed its operating parameters (for example, an earthquake or a critical component's failure) design safety features must provide a means to control reactivity and cool the reactor. During normal operation, reactor cooling relies on the water that enters the reactor vessel and the generated steam that exits. During safe shutdown, after the fission process is halted, the reactor core continues to generate heat by radioactive decay and generates steam. The heat from this radioactive decay initially equals about 6% of the heat produced by the reactor at full power and gradually declines. 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. The most severe operating condition affecting a BWR is a loss of coolant accident (LOCA). In the absence of coolant, the uncovered reactor core continues to generate heat through radioactive decay. 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 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. 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)). The NRC views the DEGB as an extremely unlikely event (likely to occur only once per 100,000 years of reactor operation). Only General Electric boiling water reactors 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, they have fewer components than pressurized water reactors. The original BWR design-types have been decommissioned, but Type 2 through Type 6 BWRs continue to operate. Some of the BWR evolutionary design features are summarized in Table 2 . Along with the evolution in BWR reactor design, containment structure designs have also evolved ( Figure 3 , Figure 4 , and Figure 5 ). 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 heat exchanger/steam generator. A secondary system then transfers steam to a combined high-pressure/low-pressure turbine generator ( Figure 6 ). 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 water and directly discharges it to a lake, river, or ocean. Unlike a BWR design, the cooling water that flows through the reactor core never contacts the turbine-generator. Under normal operating conditions, reactor cooling-water does not contact the environment. 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. 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. 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. As with BWRs, the most severe operating condition affecting a PWR is the loss of coolant accident (LOCA), the extreme case represented by the double-ended guillotine break (DEGB) of 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 consists 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 responds to 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). All U.S. reactors have primary containment structures 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 the reactor cooling system may release during an accident. The largest containment designs 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. 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). A dry containment system consists of a steel shell surrounded by a concrete biological shield that protects the reactor against outside elements, for example, debris driven by hurricane winds or an aircraft strike. The outer shield does not have a design function 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. A summary of containment building design features appears in Table 4 . 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. 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. Earthquakes occur when stresses in the earth exceed the strength of a rock mass, creating a fault or mobilizing an existing fault. 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: (1) amplitude—a measure of the peak wave height, and (2) period—the time interval between the arrival of successive peaks or valleys. The seismic wave's arrival causes ground motion. The intensity of ground motion depends primarily 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. In general, for a given magnitude earthquake, the shallower the focus, the stronger the wave will be when reaching the surface. In addition, the intensity of ground shaking diminishes with increasing distance from the earthquake focus. Sites with deep, soft soils or loosely compacted fill will experience stronger ground motion than sites with stiff soils or rock. In 1973, the concept of the "safe shutdown earthquake" (SSE) was introduced in Title 10 Part 100 of the Code of Federal Regulations (10 CFR 100), Appendix A—Seismic and Geologic Siting Criteria for Nuclear Power Plants . The NRC defines the Safe Shutdown Earthquake as the maximum earthquake in which certain structures, systems, and components, important to safety, must remain functional. Under an "operating basis earthquake," the reactor could continue operation without undue risk to the safety of the public. Ground motion at any specific location, such as a nuclear plant site, depends on the earthquake source, magnitude, distance to the source, and the attenuation (dampening) caused by rock and soil characteristics. A nuclear power plant responds to an earthquake depending on how its individual structures, systems, and components resonate, or vibrate, with the ground shaking. Heavier and more massive structures resonate at lower frequencies, while light components resonate at higher frequencies. During an earthquake, ground motion transmits vibrations to a nuclear power plant's foundation and structure. The vibrations cause back-and-forth acceleration of a structure, system or components that is measured relative to the earth's gravitational acceleration constant (g). Both vertical and horizontal components of ground acceleration place loads, or stresses, on a nuclear power plant's structure. Peak Ground Acceleration (PGA) is a measure that has been widely used in developing nuclear power plant "fragility estimates," which represent the sensitivity of nuclear plant structures, systems, and components (SSCs) to the inertial effects of acceleration during ground shaking. Structural damage to nuclear power plants occurs when the cumulative effects of ground acceleration (seismically induced vibrations) cross a certain threshold. The Electric Power Research Institute (EPRI) developed the concept of "cumulative absolute velocity" (CAV) in 1988 as an index for indicating the onset of structural damage from the cumulative effects of ground acceleration. The threshold between damaging and non-damaging earthquakes (for well-designed buildings) conservatively occurs at ground motions with cumulative absolute velocities (CAV) greater than 0.16 g-seconds. In simple terms, CAV is the sum of various ground acceleration frequencies (measured in terms of g) and the duration of their acceleration (measured in seconds). An example of this phenomenon is a wire coat-hanger that breaks from metal fatigue after being rapidly bent multiple times. Experimental and empirical seismic data have provided insights into the behavior of different structures under various acceleration and shaking conditions. For example, welded steel piping at nuclear power plants rarely failed when peak ground accelerations remained below 0.5g. Other types of structures exhibit different behaviors. Engineers design the various plant structures to withstand a certain severity of earthquake and estimates of ground shaking specific to each plant site. 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. Various plant structures, depending upon their elevation above the foundation, vibrate at different frequencies during an earthquake. Vibrations in the range of 1 to 10 Hz are particularly problematic, because a wide range of structures are susceptible to damaging resonance at those frequencies. These accelerations and the corresponding shaking frequencies are factors in the Probabilistic Seismic Hazard Analysis (PSHA, discussed below). The full seismic spectrum often can be characterized by two intervals: peak ground acceleration (PGA) and spectral acceleration (SA) averaged between 5 and 10 hertz (Hz). In the western United States (WUS), where earthquakes with frequencies below 15 Hz predominate, earthquake magnitude is a principal design consideration for nuclear power plants. Earthquakes below the 10 Hz frequency range pose the greatest hazard to nuclear power plants. In the central and eastern United States (CEUS), designs considered both earthquake magnitude and Modified Mercalli Intensity (MMI) due to sparse recordings of actual earthquake events. While plants designed to operate in the CEUS must also withstand low frequency earthquakes, the earthquakes that do occur are associated more often with higher frequencies than in the WUS. Higher frequency earthquakes are less damaging to large structures but may adversely affect small components. By the late 1940s, structural engineers had begun considering the shear forces caused by earthquakes that structures must resist. To supplement their design calculations, they referred to the Seismic Zone Map published by the Uniform Building Code (UBC). (Refer to Appendix Figure B -1 .) The UBC map divided the United State into six distinct seismic zones representing various degrees of seismic risk. The map expressed peak ground acceleration as the decimal ratio of the gravitational acceleration constant (g) that applied to a Maximum Credible Earthquake (MCE) and an Operating Basis Earthquake (OBE). UBC defined a maximum credible earthquake and its associated ground motion as the largest magnitude earthquake that could reasonably occur along the recognized faults or within a particular seismic source. An operating basis earthquake was defined as having the greatest level of ground motion likely to occur during the economic life of a structure. Designs for nuclear power plants granted construction permits during the 1960s and 1970s applied a deterministic approach to seismic design based on site-specific investigations of local and regional seismology, geology, and geotechnical soil conditions to determine the maximum credible earthquake from a single source (fault). Deterministic Seismic Hazard Analysis (DSHA) attempted to quantify the effects of a maximum credible earthquake based on known seismic sources sufficiently near the site and available historical seismic and geological data to estimate ground motion at the plant site. Appendix A to 10 CFR 100 requires an investigation of fault and earthquake occurrences to provide the basis for determining a safe shutdown earthquake. Appendix A to 10 CFR 100 notes the limitations for basing seismic design criteria on literature reviews of geophysical and geologic information, and requires supplementing the investigation with studies for vibratory ground motion, evidence of surface faulting, and evidence of seismically induced floods and water waves that have or could have affected the site. Under 10 CFR 100.23 (Geologic and Seismic Siting Criteria), designs for new nuclear power plants will base their Safe Shutdown Earthquake on Probabilistic Seismic Hazard Analysis (PSHA). The methodology has also found widespread use in U.S. engineering practice for non-nuclear structures. Where DSHA had based peak ground acceleration (PGA) on a single earthquake source, PSHA uses up-to-date interpretations of earthquake sources, earthquake recurrence, and strong ground motion estimates to estimate the probability of exceeding various levels of earthquake-caused ground motion at a given location in a given future time period. It quantifies a site's seismic hazard characteristics from seismic hazard curves or "response spectra" developed in part by identifying and characterizing each seismic source in terms of maximum magnitude, magnitude recurrence relationship, and source geometry. A response spectrum is a plot of the maximum response (acceleration, velocity, or displacement) of a family of oscillations (ground or structures). When derived from a earthquake record, the site-specific ground motion response spectrum appears as an irregular graph of peaks and valleys that combines a number of individual response spectra from past earthquakes ( Figure 7 ). Each earthquake produces a spectrum of ground motions that vary in frequency and acceleration. The seismic spectra important to nuclear power plant design are peak ground accelerations between 5 and 10 Hz. The NRC has developed Design Response Spectra statistically from response spectra of past strong motion earthquakes. The former Atomic Energy Commission (AEC) (the NRC's predecessor) published Regulatory Guide 1.60, Design Response Spectra of Nuclear Power Reactors in 1973 to provide spectral shapes for horizontal and vertical ground movements that designs must respond to (design response). A Safe Shutdown Earthquake is defined by 10 CFR 100 Appendix A as the response spectra corresponding to the maximum vibratory accelerations at the elevations of the nuclear power plant structural foundations. NRC may credit nuclear power plant foundations with a 5% dampening affect in reducing the transmission of ground accelerations. In the example of Figure 8 , the range of maximum accelerations (g) is plotted against the range of corresponding frequencies (Hz) for the Surry Nuclear Power Plant containment building. The NRC requires that nuclear plant designs account for site-specific ground motions and has specified a minimum ground motion level for nuclear plant designs. The NRC Regulatory Guide 1.208 endorses either the EPRI or Lawrence Livermore National Laboratory (LLNL) seismic hazard models as a starting point for conducting a PSHA in siting nuclear power plants. In 2007, the NRC contracted with Brookhaven National Laboratory to produce Evaluation of the Seismic Design Criteria in ASCE/SEI Standard 43-05 for Application to Nuclear Power Plants (NUREG/CR-6926). The report presents the results of a review of 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." As its title implies, this standard provides seismic design criteria for safety-related structures, systems, and components (SSCs) in a broad spectrum of nuclear facilities. In 2008, the U.S. Geological Survey (USGS) released an update of the National Seismic Hazard Maps (NSHM). 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 USGS revises the NHSM every six years to reflect newly published earthquake data. The NHSM are used 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. Although the seismic hazard is highest in the most tectonically active regions of the western United States, including Alaska, the Central and Eastern United States (CEUS) also contain regions of elevated seismic hazard. The USGS has mapped the New Madrid Seismic Zone (NMSZ)—which includes parts of Arkansas, Tennessee, Missouri, Illinois, and Kentucky—as comparable to the more seismically active portions of California. Portions of Indiana are also subject to elevated seismic hazard from the Wabash Valley Seismic Zone along the border of Illinois and Indiana. Earthquakes also occur in the central United States outside of these two documented zones, however, as demonstrated by the November 6, 2011, magnitude 5.6 earthquake that occurred 40 miles east of Oklahoma City. In the 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. The elevated seismic hazard in the New Madrid Seismic Zone is due primarily to three large earthquakes of magnitude 7.0 or greater that occurred during 1811-1812 and the chances of an earthquake of similar magnitude striking the region again. Because of the elevated hazard, portions of the NMSZ in the central United States require monitoring by a relatively dense array of seismometers compared to the northeastern and southeastern United States. As discussed earlier, most of the nation's nuclear power plants are located in the CEUS. The dominant earthquake hazard in the northeastern United States is related to seismic activity in the St. Lawrence rift zone, which extends from northeastern New York State into Canada. However, other parts of the Northeast have also experienced earthquakes, including western New York, Massachusetts, New Jersey, Pennsylvania, Ohio, and Maine. The seismic hazard in the southeastern United States is primarily near Charleston, SC, and along the Eastern Tennessee Seismic Zone. The largest historically documented earthquake that occurred in the CEUS other than the 1811-1812 earthquakes in the NMSZ took place near Charleston in 1886. The Eastern Tennessee Seismic Zone comprises a swath including portions of Alabama, Georgia, Tennessee, North Carolina, and Virginia. The magnitude 5.8 earthquake that struck Virginia on August 23, 2011, may have been in the northernmost portion of the Eastern Tennessee Seismic Zone. Figure 9 and Figure 10 show the locations of nuclear plant sites on the 2008 USGS National Seismic Hazard Map for the United States. These maps display quantitative information about seismic ground motion hazards expressed as horizontal ground acceleration ( g ) of a particle at ground level moving horizontally during an earthquake. CRS cautions against drawing any conclusion regarding a plant's seismic risk from the figures. Figure 11 shows the proximity of plant sites to Quaternary period faults based on the USGS Quaternary Fault and Fold Database of the United States. (Quaternary-active faults are those that have slipped in Quaternary time—the last 1.6 million years.) The map is not a prediction of an earthquake event. The USGS Database has information on faults and associated folds in the United States that are believed to be sources of greater than magnitude 6 earthquakes during the Quaternary period. Geologists think that these faults are the most likely source of future great earthquakes, so it is important to know what they are, where they are, and how they work. The NRC does not rank nuclear plants by seismic risk, and the NRC has not published a regulatory guide that recommends using the USGS national hazard maps for siting nuclear power plants. However, it recently considered the implications of updated USGS seismic hazard models on the seismic risk of nuclear power plants sites operating in the Central and Eastern United states. The NRC has required that each nuclear plant built meet certain structural specifications based on the earthquake susceptibility of each plant site. The NRC may re-evaluate some of those design specifications in light of the 2008 USGS seismic hazard maps. In 2010, the NRC published Review Implications of Updated Probabilistic Seismic H azard Estimates in the Central and Eastern United States on Existing Plants (GI-199 Safety/Risk Assessment), a two-stage assessment that determines the implications of the 2008 USGS updated probabilistic seismic hazard maps in the CEUS on existing nuclear power plant sites. 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 assessment first evaluated the change in seismic hazard with respect to previous estimates at individual nuclear power plants, and then estimated the change in Seismic Core Damage Frequency (SCDF) resulting from the 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 necessarily imply that a core meltdown or loss of containment (associated with a radiological release) would 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 nuclear power plants 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 March 2011, the NRC announced that it had identified 27 nuclear reactors operating in the CEUS (listed in Table 5 ) subject to priority earthquake safety reviews. 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. On April 6, 2011, the Subcommittee on Oversight and Investigations of the House Energy and Commerce Committee held a hearing on the U.S. Government Response to the Nuclear Power Plant Incident in Japan. 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. Legislation introduced in the Senate and House would reauthorize appropriations and make some changes to the ongoing National Earthquake Hazards Reduction Program (NEHRP). H.R. 1379 and S. 646 would amend the Earthquake Hazards Reduction Act of 1977 (42 U.S.C. 7704), and reauthorize appropriations through FY2015 (authorization for appropriations for NEHRP activities ended in FY2009). H.R. 3479 would also make some changes to NEHRP, and authorize appropriations through FY2014. Activities conducted under NEHRP are intended to improve the scientific understanding of earthquakes and their effects on people and infrastructure, develop effective measures to reduce earthquake hazards, and promote the adoption of earthquake hazard reduction activities. Results from these activities could improve the current understanding of how earthquake-caused shaking would affect nuclear power plants. 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 requirements for nuclear power plant licenses: 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. 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. The USGS developed a 2011 report in response to a request from the NRC evaluating seismic monitoring capabilities in the CEUS, with particular emphasis on meeting the current and future needs of the NRC. The USGS report recommends adding 100 new stations to existing arrays of seismometers in the CEUS: 38 deployed in seismically active zones in the CEUS to capture information from infrequent earthquakes for use in improving the understanding of how, where, and why earthquakes occur in the region; and 62 stations deployed near nuclear power plants to understand the site-specific response to earthquakes and improve assessments of seismic risk at the facilities. According to the USGS, the 38 proposed stations in seismically active zones would bolster earthquake monitoring where it is currently inadequate. The new stations would fill in where existing stations are farther than 50 kilometers apart (about 31 miles) or where existing stations could go off-scale during an earthquake of expected magnitude for the seismic zone. These stations would also help improve assessments of how well the seismic waves travel, or propagate, from the epicenter outwards, and inversely how the seismic energy is lost, or attenuated, as the earthquake waves move away from the epicenter. As people experienced first-hand, Virginia's August 2011 earthquake broadcast widespread ground shaking. For similar magnitude earthquakes, seismic waves travel more efficiently in general in the CEUS and shaking is felt over a broader area compared to the western United States. According to its report, the USGS proposes deploying 62 additional stations near nuclear power plants to help characterize earthquake effects for site-specific engineering and emergency-response applications. This would aid in the location and degree of ground shaking and estimation of damage to structures in near-real time. The combined 100 new stations proposed by the USGS would supplement the Advanced National Seismic System (ANSS) to compute earthquake properties more accurately in the region and improve the likelihood of capturing important earthquake data within 50 kilometers of the earthquake epicenter for moderate and large earthquakes. The USGS report indicates that this information would be useful to the NRC. The USGS report also includes cost information for the seismometers, which range from $15,000 to $20,000 to purchase and install "Class A" systems, and about $10,000 for "Class B" systems. Class A systems have higher resolution and other features superior to Class B systems. Operating costs range from $2,000 to $5,000 per year. For 100 new stations, costs to purchase and install Class A systems would range between $1.5 million and $2.0 million versus $1.0 million for 100 Class B systems. Operating costs would range between $200,000 to $500,000 per year, depending on location and options for transmitting data from the stations. 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 an earthquake's size 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 Moment Magnitude (M), which accounts for the strength of the rock that ruptured, the area of the fault that ruptured, and the average amount of slip. Moment is a physical quantity proportional to the slip on the fault times the area of the fault surface that slips. It relates 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 moment magnitude 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. In 1935, Charles F. Richter of the California Institute of Technology developed the Richter magnitude scale based on the behavior of a specific seismograph 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 term Richter Scale is so common in use 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 assigned intensity factors include 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 damage to 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. 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 increasing magnitude (M). Various methods developed relate the magnitude of an earthquake to the amplitude of acceleration it induces, and different methods may result in significant variations in results. Appendix B. Early Seismic Zone Map In the late 1940s, structural engineers began considering the seismic-based shear forces that structures must resist. To supplement their design calculations, they referred to the Seismic Zone Map published by the Uniform Building Code (UBC, in Figure B -1 ). The UBC map divided the United States into six distinct seismic zones representing various degrees of seismic risk. The map expressed peak ground acceleration as the decimal ratio of the acceleration due to gravity (g) that applied to a Maximum Credible Earthquake (MCE) and an Operating Basis Earthquake (OBE). UBC defined a maximum credible earthquake as producing the greatest level of ground motion at a site. An operating basis earthquake was be defined as the greatest level of ground motion likely to occur during the economic life of a structure. Appendix C. 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 damage 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 3%-5% fissile uranium-235. Includes BWR and PWR types. Loss of Coolant Accident (LOCA) is an accident involving a broken pipe, stuck-open valve, or other leak in the reactor coolant system that results in a loss of the water cooling the reactor core. Operating Basis Earthquake is the maximum vibratory ground motion for which 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.

Summary: The earthquake and subsequent tsunami that devastated Japan's Fukushima Daiichi nuclear power station and the earthquake that forced the North Anna, VA, nuclear power plant's temporary shutdown have focused attention on the seismic criteria applied to siting and designing commercial nuclear power plants. Some Members of Congress have questioned whether U.S nuclear plants are more vulnerable to seismic threats than previously assessed, particularly given the Nuclear Regulatory Commission's (NRC's) ongoing reassessment of seismic risks at certain plant sites. The design and operation of commercial nuclear power plants operating in the United States vary considerably because 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 are BWR designs. One of the most severe operating conditions a reactor may face 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 water containing boron (borated water slows the fission process) into the reactor coolant system following a pipe rupture or other water loss. 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)). The NRC considers the DEGB to be an extremely unlikely event; however, even unlikely events can occur, as the magnitude 9.0 earthquake and resulting tsunami that struck Fukushima Daiichi proves. U.S. nuclear power plants designed in the 1960s and 1970s used a deterministic statistical approach to addressing the risk of damage from shaking caused by a large earthquake (termed Deterministic Seismic Hazard Analysis, or DSHA). Since then, engineers have adopted a more comprehensive approach to design known as Probabilistic Seismic Hazard Analysis (PSHA). PSHA estimates the likelihood that various levels of ground motion will be exceeded at a given location in a given future time period. New nuclear plant designs will apply PSHA. In 2008, the U.S Geological Survey (USGS) updated the National Seismic Hazard Maps (NSHM) that were last revised in 2002. 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). Spectral acceleration refers to ground motion over a range, or spectra, of frequencies. Seismic hazards are greatest in the WUS, particularly in California, Oregon, and Washington, as well as Alaska and Hawaii. In 2010, the NRC examined the implications of the updated NSHM for nuclear power plants operating in the CEUS, and concluded that NSHM data suggest that the probability for earthquake ground motions may be above the seismic design basis for some nuclear plants in the CEUS. In late March 2011, NRC announced that it had identified 27 nuclear reactors operating in the CEUS that would receive priority earthquake safety reviews.