ANTHONY R. PIETRANGELO Senior Vice President and Chief Nuclear Officer 1201 F Street, NW, Suite 1100 Washington, DC 20004 P: 202.739.8081 arp@nei.org nei.org

March 12, 2014 Mr. Eric J. Leeds Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Subject: Seismic Risk Evaluations for Plants in the Central and Eastern United States Project Number: 689 Dear Mr. Leeds: On March 12, 2012, NRC issued a request for information to power reactor licensees and holders of construction permits in accordance with 10 CFR Part 50, Section 50.54(f) [ML12053A340]. In Enclosure (1) of that letter, NRC requested specific information related to updated seismic hazard estimates and associated risk evaluations. In a letter dated February 15, 2013 [ML12319A074], NRC endorsed the industry guidance for performing those evaluations, Seismic Evaluation Guidance: Screening, Prioritization, and Implementation Details (SPID) for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1: Seismic, Electric Power Research Institute (EPRI) Report 1025287 (February 2013). On May 7, 2013 [ML13106A331], NRC endorsed industry guidance for an additional deterministic evaluation, the Expedited Seismic Evaluation described in EPRI Report 3002000704 (May 2013), and agreed with a modified schedule for completing all of the seismic tasks associated with the 50.54(f) letter. In accordance with that schedule, licensees and construction permit holders in the Central and Eastern United States will submit seismic hazard/screening reports to NRC by March 31, 2014. In a letter dated February 20, 2014 [ML14030A046], NRC provided supplemental information related to the 50.54(f) request for information. NRC noted that since the seismic hazard reevaluations being performed pursuant to the 50.54(f) letter are considered to be distinct from the current design or licensing basis of operating plants, the results of those analyses are generally not expected to call into question the operability or functionality of systems, structures, or components. Therefore, the results are not expected to be reportable pursuant to 10 CFR 50.72, "Immediate notification requirements for operating nuclear power reactors," and 10 CFR 50.73, "Licensee event report system." NRC acknowledged the benefit of the Expedited Seismic Evaluation as a timely method for demonstrating additional seismic margin through near-term evaluations and enhancing safety through potential plant modifications.

Mr. Eric J. Leeds March 12, 2014 Page 2 In the February 20, 2014 letter, NRC also requested that licensees’ seismic hazard/screening reports include an interim evaluation or actions to demonstrate that the plants can cope with the reevaluated hazard while the expedited evaluations and more comprehensive risk evaluations are conducted. In response to that request, Attachment 1 contains a report transmitted from EPRI to NEI on March 11, 2014, Fleet Seismic Core Damage Frequency Estimates for Central and Eastern U.S. Nuclear Power Plants Using New Site-Specific Seismic Hazard Estimates. In this study, initial estimates of seismic core damage frequency (SCDF) were calculated using the latest seismic hazard information and compared to earlier estimates of SCDF that had been developed for the Generic Issue 199 Safety/Risk Assessment in 2010. As shown in Attachment 1, the overall distribution of SCDFs for the fleet indicates that the impact of the updated seismic hazard has been to reduce risk for most plants relative to estimates obtained using either the 2008 USGS or the 1994 LLNL hazard assessments, with all plants still falling in the range of 1E-7/year to 1E-4/year. This observation is further supported by the information provided in Attachment 2, Perspective on the Seismic Capacity of Operating Plants, which describes how seismic ruggedness is achieved through the design process and has been demonstrated by earthquake experience. Thus, the conclusions reached in 2010 remain valid (i.e., existing operating reactors have margin to withstand potential earthquakes exceeding their original design bases and no concern exists regarding adequate protection), and the course of action proposed in NEI letter dated April 9, 2013 [ML13101A379] remains appropriate. Please feel free to contact me or Kimberly Keithline (202-739-8121, kak@nei.org), if you have any questions. Sincerely,

Anthony R. Pietrangelo Attachments c: Mr. Nilesh C. Chokshi, NRO/DSEA, NRC

March 11, 2014 Anthony R. Pietrangelo Senior Vice President and Chief Nuclear Officer, Nuclear Generation Nuclear Energy Institute 1201 F Street, NW, Suite 1100 Washington, DC 20004 Subject: Fleet Seismic Core Damage Frequency Estimates for Central and Eastern U.S. Nuclear

Power Plants Using New Site-Specific Seismic Hazard Estimates Dear Mr. Pietrangelo: The Electric Power Research Institute (EPRI) has recently completed site-specific seismic hazard evaluations for nuclear plants in the central and eastern United States (CEUS) using the guidance in Electric Power Research Institute (EPRI) 1025287 (EPRI 2013a). To provide perspective regarding the safety implications of these new seismic hazard estimates, EPRI has performed an initial assessment of the changes in the seismic core-damage frequency relative to earlier fleet-wide estimates. A description of the fleet evaluation is attached. If you have questions or would like to discuss this evaluation, please contact John Richards at 704-595-2707 or jrichards@epri.com. Sincerely, Stuart Lewis Program Manager Risk and Safety Management RSM-031114-077 Attachment

ATTACHMENT 1

1

Fleet Seismic Core Damage Frequency Estimates for Central and Eastern U.S. Nuclear Power Plants Using New Site-Specific Seismic Hazard Estimates

Electric Power Research Institute Project Manager J. Richards

This evaluation was prepared by Simpson Gumpertz & Heger Inc., under contract to the Electric Power Research Institute.

The principal authors are G. Hardy, T. Graf, F. Grant, and Y. Tang.

1 BACKGROUND

Following the accident at the Fukushima Daiichi Nuclear Power Plant resulting from the March 11, 2011, Great Tohoku Earthquake and subsequent tsunami, the U.S. Nuclear Regulatory Commission (USNRC) established a Near Term Task Force (NTTF) to conduct a systematic review of USNRC processes and regulations, and to determine if the agency should make additional improvements to its regulatory system. The NTTF developed a set of recommendations intended to clarify and strengthen the regulatory framework for protection against natural phenomena such as earthquakes. Subsequently, the USNRC issued a 50.54(f) letter that requests information to ensure that all U.S. nuclear power plants address these recommendations. This letter requests that licensees and holders of construction permits under 10 CFR Part 50 reevaluate the seismic hazards at their sites against present-day USNRC requirements and guidance. In response to the 50.54(f) letter, site-specific seismic hazard estimates have been developed for nuclear plants in the central and eastern United States (CEUS) using the guidance in Electric Power Research Institute (EPRI) 1025287 (EPRI, 2013a) and in a 2013 letter from the Nuclear Energy Institute (NEI, 2013). These hazards form the basis for determining whether further seismic evaluation may be needed on a plant-by-plant basis. The USNRC has requested that interim actions (that is, actions that can be implemented before more extensive seismic evaluations could be completed) be taken for plants whose ground motion response spectrum (GMRS) exceeds the design basis (USNRC, 2012; USNRC, 2014). In response to this request, the U.S. nuclear industry proposed an Expedited Seismic Evaluation Process (ESEP) as an effective interim action. Guidance for conducting such an evaluation was developed by EPRI (EPRI, 2013b), and the process and guidance were endorsed by the USNRC (USNRC, 2013). The expedited evaluation is being carried out for any site with a GMRS that exceeds the safe shutdown earthquake (SSE) in the spectral frequency range from 1 to 10 Hz. As an input to the consideration of whether additional interim actions may be warranted, EPRI has estimated, for the fleet of nuclear power plants operating in the CEUS, the seismic core-damage frequencies (SCDFs) based on the newly completed site-specific seismic hazards.

2

2 OBJECTIVES

Because it does not explicitly account for the capability of a nuclear power plant to maintain a safe condition during an earthquake, the GMRS calculated from the new seismic hazard characterization provides an incomplete perspective regarding overall seismic safety. The objective of this study is to provide an initial assessment of the safety implications of the new seismic hazard estimates across the CEUS fleet of operating plants. This assessment involves comparing SCDF estimates reflecting the new seismic hazard estimates for the fleet of operating plants to SCDF estimates previously developed by the USNRC in its 2010 Safety / Risk Assessment for GI-199 (USNRC, 2010). To perform this assessment, point estimates of the SCDF have been developed using (1) the methods defined by the USNRC in the 2010 Safety / Risk Assessment for GI-199, (2) the plant-level fragilities determined by the USNRC in the GI-199 Assessment, and (3) new site-specific seismic hazard estimates. The resulting SCDF estimates are compared with the baseline SCDFs developed by the USNRC in 2010 using the 2008 U.S. Geological Survey (USGS) and 1994 Lawrence Livermore National Laboratory (LLNL) seismic hazard curves. These are, respectively, the most recent seismic hazard assessment available at the time of the 2010 study, and the hazard assessment used by the USNRC in its review of seismic evaluations submitted as part of the Individual Plant Examination of External Events (IPEEE) in the 1990s.

3 ESTIMATING SEISMIC CORE DAMAGE FREQUENCY

As described in Section 1, new probabilistic seismic hazard analyses (PSHA) have been completed for all U.S. nuclear power plant sites located in the CEUS. The potential safety and risk implications of these new seismic hazard estimates can most comprehensively be assessed with a modern Seismic Probabilistic Risk Assessment (SPRA) in accordance with the PRA Standard (ASME, 2013), but these modern SPRAs are not yet available for most plants. In 2010, the USNRC used a simplified approach to estimate the SCDF for all of the CEUS plants as part of the GI-199 program. The GI-199 program was associated with the changing understanding of seismic hazards in much of the United States and the implications of that understanding for nuclear plant safety. The USNRC simplified seismic risk estimation approach involved estimating the plant seismic fragility (i.e., conditional probability of plant damage at a given seismic hazard input level) from the results of the earlier IPEEE submittals, and convolving that plant fragility estimate with the new seismic hazard to obtain an SCDF estimate. EPRI is conducting a similar assessment of SCDFs for the fleet of CEUS plants using the same IPEEE-derived plant-level fragilities combined with the new site-specific seismic hazard curves.

4 SITE-SPECIFIC SEISMIC HAZARDS

The first major step in responding to Enclosure 1 of the 50.54(f) letter (USNRC, 2012) is to calculate seismic hazards at existing plant sites following the USNRC endorsed guidance in the Screening, Prioritization and Implementation Details (SPID) (EPRI, 2013a). These seismic hazards incorporate PSHA methods using the recently developed CEUS Seismic Source Characterization (CEUS-SSC) for Nuclear Facilities (CEUS-SSC, 2012), together with an updated ground-motion model (GMM) for the CEUS (EPRI, 2013c), and site-specific site amplification calculations. CEUS plants will submit these site-specific seismic hazards by March 31, 2014, in accordance with NEI’s letter dated 9 April 2013 (NEI, 2013). These newly developed seismic hazard characterizations were used for the subject fleet SCDF calculations.

3

5 PLANT-LEVEL FRAGILITIES

Plant-level fragility curves for each GI-199 plant were developed by the USNRC as part of the 2010 Safety / Risk Assessment based on information provided in the IPEEE submittals. Appendix C of the USNRC GI-199 report (USNRC, 2010) defines three methods for estimating a plant-level fragility from information reported in the IPEEE submittals. The methods are briefly summarized in Table C.1 of the USNRC report (2010), which is reproduced below as Table 1. About one-third of the plants in the CEUS performed an SPRA as part of their IPEEE program. Many of the plants that performed SPRAs provided plant-level fragility information in their IPEEE submittals (Method 1a below), and the remaining plants that performed SPRAs provided SCDF estimates based on a variety of seismic hazard curves (EPRI 1989, LLNL1994, or site-specific curves developed specifically for the IPEEE program). For these remaining plants, plant-level fragility values were approximated in the USNRC GI-199 assessment by estimating and matching the reported SCDFs and using engineering judgment (Methods 1b and 1c below). In cases where reasonable engineering judgments could not be readily made, the USNRC performed sensitivity studies to estimate the potential plant level fragilities. This resulted in more than one potential plant level fragility for a number of specific plant sites / units. Two-thirds of the plants conducted a seismic margins analysis (SMA) as part of their IPEEE program. For these plants, the USNRC estimated the plant-level fragility based on the reported plant-level high confidence of a low probability of failure (HCLPF) value and an estimate of the composite variability, βc (Methods 2, 3a, and 3b below). The USNRC used a βc of 0.4 to develop

the plant-level fragilities for the SMA plants.

Table 1 –Summary of USNRC GI-199 Methods for Estimating Plant Damage State Fragilities

Bases for Establishing Plant-Level Fragility Curves Parameters From IPEEE Information

Basis Source Parameters*

1a SPRA C50 and βC determined by probability plot of the reported plant-level

fragility curve

1b SPRA

C50 found by matching the computed SCDF to the SCDF stated in

the IPEEE for the specified hazard curve (EPRI, LLNL, or plant-specific). Assumed βC = 0.4.

1c SPRA C50 and βC determined by matching computed SCDFs to IPEEE

SCDFs for a pair of hazard curves.

2 SMA

(HCLPF < RLE)

C50 found by using the stated HCLPF

Assumed βC = 0.4.

3a SMA

(HCLPF = RLE)

C50 found by using the stated HCLPF/RLE

Assumed βC = 0.4

Note: The RLE is a lower bound on the actual HCLPF.

3b SMA

(HCLPF = RLE = SSE)

C50 found by using the stated HCLPF/RLE/SSE

Assumed βC = 0.4

Note: The SSE is a lower bound on the actual HCLPF; applies to reduced scope SMA plants.

* C50 is the median (50th percentile) plant-level acceleration capacity and βc is the composite variability

4

These plant-level fragility values developed by the USNRC (USNRC, 2010) were used directly for the SCDF calculations in this current EPRI fleet risk assessment, which allows for a direct comparison of the SCDF estimates using the newly developed seismic hazards and the USNRC’s SCDF estimates in 2010 using the 2008 USGS and 1994 LLNL seismic hazards. For convenience, the plant-level fragilities from the GI-199 Safety / Risk Assessment are reproduced in Table 2 below. As noted above, some SPRA plants have more than one plant-level fragility estimate (sensitivity studies were conducted in the NRC GI-199 Safety / Risk Assessment for those plants where adequate information was not submitted as part of the IPEEE process). The columns to the right side of Table 2 summarize the sixty-one CEUS sites for which new site-specific seismic hazards have been calculated. For purposes of the SCDF calculations in this study, the following decisions are made relative to calculating a single SCDF for each of these sixty-one sites:

For sites with multiple units, the plant-level fragility that results in the highest SCDF estimate is conservatively selected (most sites with multiple units have the same plant-level fragilities defined due to similarity, but several sites had submitted different plant-level fragilities as part of their IPEEE efforts).

For sites where the USNRC defined multiple plant-level fragilities (due to uncertainty in the correct spectral ratios from the IPEEE submittals), the plant-level fragility that results in the highest resulting SCDF value is conservatively selected.

These sixty-one plant level “bounding” fragilities are documented on the right half of Table 2.

5

Table 2 – Plant-Level Fragilities from USNRC Safety / Risk Assessment for GI-199 (USNRC, 2010)

GI-199 Safety Report - Appendix C “Plant” Data

Point

Plant-Level Fragility from Appendix C of 2010 USNRC Safety/Risk Assessment CEUS Site with New

Hazard Estimates

Bounding Case *

Plant Level Fragility from Appendix C of 2010 USNRC Safety/Risk Assessment

PGA Fragility ** Spectral Ratios PGA Fragility ** Spectral Ratios

C50 (g) βC 10 Hz 5 Hz 1 Hz C50 (g) βC 10 Hz 5 Hz 1 Hz

Arkansas Nuclear One 1 0.76 0.4 1.87 2.12 0.96 Arkansas Nuclear 0.76 0.4 1.87 2.12 0.96

Arkansas Nuclear One 2 0.76 0.4 1.87 2.12 0.96

Beaver Valley 1 0.36 0.26 1.71 1.54 0.68 Beaver Valley 0.36 0.26 1.71 1.54 0.68

Beaver Valley 2 0.53 0.34 1.71 1.54 0.68

Braidwood 1 0.76 0.4 1.87 2.12 0.96 Braidwood 0.76 0.4 1.87 2.12 0.96

Braidwood 2 0.76 0.4 1.87 2.12 0.96

Browns Ferry 1 0.76 0.4 1.87 2.12 0.96

Browns Ferry 0.66 0.4 1.87 2.12 0.96 Browns Ferry 2 0.66 0.4 1.87 2.12 0.96

Browns Ferry 3 0.66 0.4 1.87 2.12 0.96

Brunswick 1 0.76 0.4 1.85 2.12 1.32 Brunswick 0.76 0.4 1.85 2.12 1.32

Brunswick 2 0.76 0.4 1.85 2.12 1.32

Byron 1 0.76 0.4 1.87 2.12 0.96 Byron 0.76 0.4 1.87 2.12 0.96

Byron 2 0.76 0.4 1.87 2.12 0.96

Callaway 0.76 0.4 1.85 2.12 1.32 Callaway 0.76 0.4 1.85 2.12 1.32

Calvert Cliffs 1 0.62 0.4 1.38 1.72 0.6 Calvert Cliffs 0.58 0.4 1.38 1.72 0.6

Calvert Cliffs 2 0.58 0.4 1.38 1.72 0.6

Catawba 1 0.44 0.63 1.87 2.12 0.96 Catawba 0.44 0.63 1.87 2.12 0.96

Catawba 2 0.44 0.63 1.87 2.12 0.96

Clinton (0098) 0.76 0.4 1.85 2.12 1.32 Clinton 0.76 0.4 1.67 1.81 0.59

Clinton(UHS) 0.76 0.4 1.67 1.81 0.59

Comanche Peak 1 0.30 0.4 2.26 2.56 1.28 Comanche Peak 0.3 0.4 2.26 2.56 1.28

Comanche Peak 2 0.30 0.4 2.26 2.56 1.28

Cooper 0.76 0.4 1.85 2.12 1.32 Cooper 0.76 0.4 1.85 2.12 1.32

Crystal River 3 0.25 0.4 1.22 1.51 1.58 Crystal River 0.25 0.4 1.22 1.51 1.58

D.C. Cook 1 0.48 0.27 2.27 2.13 0.65 D.C. Cook 0.48 0.27 2.27 2.13 0.65

D.C. Cook 2 0.48 0.27 2.27 2.13 0.65

Davis-Besse 0.66 0.4 1.87 2.12 0.96 Davis-Besse 0.66 0.4 1.87 2.12 0.96

Dresden 2 0.51 0.4 1.87 2.12 0.96 Dresden 0.51 0.4 1.87 2.12 0.96

Dresden 3 0.51 0.4 1.87 2.12 0.96

Duane Arnold 0.30 0.4 1.85 2.68 1.07 Duane Arnold 0.3 0.4 1.85 2.68 1.07

6

GI-199 Safety Report - Appendix C “Plant” Data

Point

Plant-Level Fragility from Appendix C of 2010 USNRC Safety/Risk Assessment CEUS Site with New

Hazard Estimates

Bounding Case *

Plant Level Fragility from Appendix C of 2010 USNRC Safety/Risk Assessment

PGA Fragility ** Spectral Ratios PGA Fragility ** Spectral Ratios

C50 (g) βC 10 Hz 5 Hz 1 Hz C50 (g) βC 10 Hz 5 Hz 1 Hz

Farley 1 (1st spectral ratios) 0.25 0.4 1.87 2.12 0.96

Farley 0.25 0.4 1.87 2.12 0.96 Farley 1 (2nd spectral ratios) 0.25 0.4 1.85 2.12 1.32

Farley 2 (1st spectral ratios) 0.25 0.4 1.87 2.12 0.96

Farley 2 (2nd spectral ratios) 0.25 0.4 1.85 2.12 1.32

Fermi 2 0.76 0.4 1.87 2.12 0.96 Fermi 0.76 0.4 1.87 2.12 0.96

FitzPatrick 0.56 0.4 1.87 2.12 0.96 FitzPatrick 0.56 0.4 1.87 2.12 0.96

Fort Calhoun 0.63 0.4 1.85 2.12 1.32 Fort Calhoun 0.63 0.4 1.85 2.12 1.32

Ginna 0.51 0.4 2.14 2.42 1.36 Ginna 0.51 0.4 2.14 2.42 1.36

Grand Gulf 1 0.38 0.4 1.92 2.65 1.33 Grand Gulf 0.38 0.4 1.92 2.65 1.33

Harris 1 0.74 0.4 1.87 2.12 0.96 Harris 0.74 0.4 1.87 2.12 0.96

Hatch 1 0.76 0.4 1.85 2.12 1.32 Hatch 0.76 0.4 1.85 2.12 1.32

Hatch 2 0.76 0.4 1.85 2.12 1.32

Hope Creek 1 1.66 0.7 1.97 2.27 0.98 Hope Creek 1.66 0.7 1.97 2.27 0.98

Indian Point 2 0.68 0.4 1.62 1.23 0.41 Indian Point 0.34 0.34 1.56 1.61 0.81

Indian Point 3 0.34 0.34 1.56 1.61 0.81

Kewaunee 0.41 0.22 1.8 1.79 0.4 Kewaunee 0.41 0.22 1.8 1.79 0.4

La Salle 1 (0098) 1.32 0.4 1.85 2.12 1.32

La Salle 1.32 0.4 1.67 1.83 0.923

La Salle 1 (SSE) 1.32 0.4 1.85 2.62 1.31

La Salle 1 (UHS) 1.32 0.4 1.67 1.83 0.923

La Salle 2 (0098) 1.32 0.4 1.85 2.12 1.32

La Salle 2 (SSE) 1.32 0.4 1.85 2.62 1.31

La Salle 2 (UHS) 1.32 0.4 1.67 1.83 0.923

Limerick 1 0.38 0.4 2.59 2.47 1.18 Limerick 0.38 0.4 2.59 2.47 1.18

Limerick 2 0.38 0.4 2.59 2.47 1.18

McGuire 1 0.45 0.74 1.88 2.35 1.19 McGuire 0.45 0.74 1.88 2.35 1.19

McGuire 2 0.45 0.74 1.88 2.35 1.19

Millstone 2 0.63 0.4 1.87 2.12 0.96 Millstone 0.54 0.4 2.27 2.27 1.26

Millstone 3 0.54 0.4 2.27 2.27 1.26

Monticello 0.30 0.4 2.29 2.69 1.12 Monticello 0.3 0.4 2.29 2.69 1.12

Nine Mile Point 1 0.68 0.4 1.87 2.12 0.96 Nine Mile Point 0.58 0.4 1.87 2.12 0.96

Nine Mile Point 2 0.58 0.4 1.87 2.12 0.96

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GI-199 Safety Report - Appendix C “Plant” Data

Point

Plant-Level Fragility from Appendix C of 2010 USNRC Safety/Risk Assessment CEUS Site with New

Hazard Estimates

Bounding Case *

Plant Level Fragility from Appendix C of 2010 USNRC Safety/Risk Assessment

PGA Fragility ** Spectral Ratios PGA Fragility ** Spectral Ratios

C50 (g) βC 10 Hz 5 Hz 1 Hz C50 (g) βC 10 Hz 5 Hz 1 Hz

North Anna 1 (1st spectral ratios)

0.41 0.4 1.87 2.12 0.96

North Anna 0.41 0.4 1.85 2.12 1.32

North Anna 1 (2nd spectral ratios)

0.41 0.4 1.85 2.12 1.32

North Anna 2 (1st spectral ratios)

0.41 0.4 1.87 2.12 0.96

North Anna 2 (2nd spectral ratios)

0.41 0.4 1.85 2.12 1.32

Oconee 1 0.62 0.32 1.66 1.32 0.35

Oconee 0.62 0.32 1.66 1.32 0.35 Oconee 2 0.62 0.32 1.66 1.32 0.35

Oconee 3 0.62 0.32 1.66 1.32 0.35

Oyster Creek 0.57 0.36 2 1.78 0.796 Oyster Creek 0.57 0.36 2 1.78 0.796

Palisades 0.49 0.35 2.13 2.44 0.74 Palisades 0.49 0.35 2.13 2.44 0.74

Peach Bottom 2 0.51 0.4 1.87 2.12 0.96 Peach Bottom 0.51 0.4 1.87 2.12 0.96

Peach Bottom 3 0.51 0.4 1.87 2.12 0.96

Perry 1 0.76 0.4 1.87 2.12 0.96 Perry 0.76 0.4 1.87 2.12 0.96

Pilgrim 1 0.49 0.27 1.55 1.66 0.5 Pilgrim 0.49 0.27 1.55 1.66 0.5

Point Beach 1 0.45 0.45 1.78 1.75 0.675 Point Beach 0.45 0.45 1.78 1.75 0.675

Point Beach 2 0.45 0.45 1.78 1.75 0.675

Prairie Island 1 0.71 0.4 1.85 2.12 1.32 Prairie Island 0.71 0.4 1.85 2.12 1.32

Prairie Island 2 0.71 0.4 1.85 2.12 1.32

Quad Cities 1 0.23 0.4 1.87 2.12 0.96 Quad Cities 0.23 0.4 1.87 2.12 0.96

Quad Cities 2 0.23 0.4 1.87 2.12 0.96

River Bend 1 0.25 0.4 2.35 2.75 1.41 River Bend 0.25 0.4 2.35 2.75 1.41

Robinson 2 0.71 0.4 1.85 2.12 1.32 Robinson 0.71 0.4 1.85 2.12 1.32

Saint Lucie 1 (s4) 0.25 0.4 1.18 1.5 0.8

Saint Lucie 0.25 0.4 1.18 1.5 0.8 Saint Lucie 1 (s5) 0.25 0.4 1.18 1.5 0.8

Saint Lucie 2 (s4) 0.25 0.4 1.18 1.5 0.8

Saint Lucie 2 (s5) 0.25 0.4 1.18 1.5 0.8

Salem 1 1.31 0.84 1.97 2.27 0.68 Salem 1.31 0.84 1.97 2.27 0.68

Salem 2 1.31 0.84 1.97 2.27 0.68

Seabrook 1 0.90 0.52 2.223 2.42 1.36 Seabrook 0.9 0.52 2.223 2.42 1.36

Sequoyah 1 0.68 0.4 1.87 2.12 0.96 Sequoyah 0.68 0.4 1.87 2.12 0.96

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GI-199 Safety Report - Appendix C “Plant” Data

Point

Plant-Level Fragility from Appendix C of 2010 USNRC Safety/Risk Assessment CEUS Site with New

Hazard Estimates

Bounding Case *

Plant Level Fragility from Appendix C of 2010 USNRC Safety/Risk Assessment

PGA Fragility ** Spectral Ratios PGA Fragility ** Spectral Ratios

C50 (g) βC 10 Hz 5 Hz 1 Hz C50 (g) βC 10 Hz 5 Hz 1 Hz

Sequoyah 2 0.68 0.4 1.87 2.12 0.96

South Texas 1 0.38 0.59 2.47 2.97 1.53 South Texas 0.38 0.59 2.47 2.97 1.53

South Texas 2 0.38 0.59 2.47 2.97 1.53

Summer 0.56 0.4 1.87 2.12 0.96 Summer 0.56 0.4 1.87 2.12 0.96

Surry 1 0.74 0.66 2.08 1.95 0.97 Surry 0.74 0.66 2.08 1.95 0.97

Surry 2 0.74 0.66 2.08 1.95 0.97

Susquehanna 1 0.53 0.4 1.87 2.12 0.96 Susquehanna 0.53 0.4 1.87 2.12 0.96

Susquehanna 2 0.53 0.4 1.87 2.12 0.96

Three Mile Island 1 0.29 0.28 2.73 2.6 1.127 Three Mile Island 0.29 0.28 2.73 2.6 1.127

Turkey Point 3 0.38 0.4 1.26 1.58 0.85 Turkey Point 0.38 0.4 1.26 1.58 0.85

Turkey Point 4 0.38 0.4 1.26 1.58 0.85

Vermont Yankee 0.63 0.4 1.87 2.12 0.96 Vermont Yankee 0.63 0.4 1.87 2.12 0.96

Vogtle 1 0.76 0.4 1.85 2.12 1.32 Vogtle 0.76 0.4 1.85 2.12 1.32

Vogtle 2 0.76 0.4 1.85 2.12 1.32

Waterford 3 0.25 0.4 1.72 2.4 1.19 Waterford 0.25 0.4 1.72 2.4 1.19

Watts Bar 1 (rock) 0.76 0.4 1.87 2.12 0.96 Watts Bar 0.76 0.4 1.87 2.12 0.96

Watts Bar 1 (soil) 0.76 0.4 1.85 2.12 1.32

Wolf Creek 1 0.51 0.4 1.83 2.25 0.32 Wolf Creek 0.51 0.4 1.83 2.25 0.32

* Plant level fragility that results in the maximum SCDF for the site when combined with the newly developed site-specific seismic hazard (2013/2014) ** C50 is the median (50

th percentile) plant-level acceleration capacity and βc is the composite variability

9

6 QUANTIFICATION APPROACH

The USNRC used approximate methods to estimate the SCDF for each operating nuclear plant as part of their 2010 study to assess the safety implications of changing seismic hazards as part of GI-199. These approximate SCDF estimates were developed using a method that involved integrating the mean seismic hazard curve and an approximation of the mean plant-level fragility curve for each plant. This approximate method was first developed by Kennedy (Kennedy, 1999) and is discussed in Section 10-B.9 of the ASME/ANS RA-Sa-2009 Standard (ASME, 2009), as well as Appendix D of the SPID (EPRI, 2013a). This same approach is judged to be the most appropriate method to assess this latest set of new site-specific seismic hazard estimates developed in accordance with the USNRC’s 50.54(f) letter. In the NRC Safety/Risk Assessment of GI-199, SCDF estimates were computed at four spectral frequencies: 10 Hz, 5 Hz, 1 Hz and the peak ground acceleration (PGA). The terminology defined within the GI-199 Safety/Risk Assessment included the concept of a “derived SCDF estimate” which consisted of an estimate of the seismic core-damage frequency that was developed from these four spectral SCDF estimates:

SCDFpga = SCDF estimate obtained by using the PGA-based seismic hazard and plant-level

fragility curves

SCDF10 = SCDF estimate obtained by using the 10 Hz seismic hazard and plant-level fragility curves

SCDF5 = SCDF estimate obtained by using the 5 Hz seismic hazard and plant-level fragility curves

SCDF1 = SCDF estimate obtained by using the 1 Hz seismic hazard and plant-level fragility curves

The seismic core damage frequency for a plant can most accurately be generated by incorporating each individual seismic fragility function into the complete plant logic model and convolving with the hazard to develop the SCDF. However, since the plant logic model was not typically included as part of the IPEEE submittal, this approximate approach is the best alternative to estimating these SCDFs. Past SPRAs have demonstrated that the actual plant risk is a function of the seismic response at a variety of spectral frequencies. The plant risk is very site specific and is a function of:

Failure modes governing the lower capacity structures, systems and components

Soil frequencies for those structures founded on soil columns

Structure fundamental frequencies

Equipment fundamental frequencies

The frequency ranges that drive the plant seismic risk are typically very broad, including contributions from 1 Hz to PGA. One of the methods to account for the spectral frequency contribution to the SCDF used in the GI-199 Safety / Risk Assessment considered each of the four frequencies (1, 5, 10 Hz and PGA) to contribute equally to the overall SCDF. The resulting “derived SCDF estimate” associated with this spectral weighting is shown mathematically in the equation below:

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This averaging of the four frequencies approach is judged to be appropriate for this study as past SPRAs have demonstrated that typically there are risk contributions from all these frequencies due to the variety of equipment, systems and structures that end up contributing to the risk. In addition, EPRI has conducted some limited additional sensitivity studies related to this frequency weighting (expanding the number of frequencies from 4 to 6 and also considering an alternate approach in the GI-199 Safety / Risk Assessment referred to as the “IPEEE weighted average SCDF” approach) and the overall results and conclusions are relatively insensitive to the approach taken. EPRI does not recommend using any very conservative approaches to estimate the SCDF such as use of the maximum SCDFs calculated at any one frequency. This type of bounding approach is overly conservative and judged to not provide realistic risk estimates consistent with SCDFs calculated in actual SPRAs.

7 SCDF RESULTS

To provide an initial assessment of the safety implications of the new seismic hazard estimates across the fleet of CEUS operating plants, point estimates of the mean SCDF are developed using the new site-specific seismic hazard curves. These are compared with the baseline SCDFs developed by the USNRC in 2010 using the 2008 USGS and 1994 LLNL seismic hazard curves. Figure 1 provides a comparison of the cumulative fleet SCDF distribution calculated using the new site-specific seismic hazards, the 1994 LLNL seismic hazards, and the 2008 USGS seismic hazards. The SCDF values computed using the new hazard range from approximately 4E-7/year to 6E-5/year. The comparison shows that the overall distribution of SCDFs for the fleet has not changed significantly due to the new site-specific seismic hazards.

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Figure 1 – Comparison of CEUS NPP Site Cumulative Distribution of Seismic CDFs

8 CONCLUSIONS

In 2010, the USNRC conducted a Safety / Risk Assessment for the GI-199 program and developed simplified methods to calculate a point estimate of the SCDF. The USNRC developed an estimate of the seismic hazard at that time using the 2008 USGS seismic source to develop a new rock hazard, and EPRI site amplification factors. This 2008 hazard, along with the previously developed 1994 LLNL hazard, was then used to estimate the SCDFs for the fleet of U.S. plants using the plant-level fragilities estimated from each plant’s IPEEE submittals. The USNRC concluded in 2010 that the overall SCDF estimates are indicative of performance consistent with the Commission’s Safety Goal Policy Statement because they are within the subsidiary objective of 1E-4/year. The specific USNRC statement from the GI-199 Safety / Risk Assessment (USNRC, 2010) was:

“Overall seismic core damage risk estimates are consistent with the Commission’s Safety Goal Policy Statement because they are within the subsidiary objective of 10-4/year for core damage frequency. The GI-199 Safety / Risk Assessment, based in part on information from the U.S. Nuclear Regulatory Commission’s (NRC’s) Individual Plant Examination of External Events (IPEEE) program, indicates that no concern exists regarding adequate protection and that the current seismic design of operating reactors

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provides a safety margin to withstand potential earthquakes exceeding the original design basis.”

New seismic hazard analyses have been completed for all sixty-one CEUS nuclear power plant sites. EPRI calculated the approximate SCDFs for each of these sites using methods that the USNRC used to assess changing seismic hazard in the past. As can be seen from Figure 1 above, the overall distribution of SCDFs for the fleet indicates that the impact of the updated seismic hazard has been to reduce risk for most plants relative to estimates obtained using either the 2008 USGS or the 1994 LLNL hazard assessments.

The range of SCDFs still falls between 1E-7/year and 1E-4/year.

For individual plants, some plant SCDF estimates have increased, but the vast majority have decreased somewhat.

In the case of the sites for which increases were seen, none of the SCDF values approaches 1E-4/year.

Comparisons of the SCDF estimates developed in 2010 by the USNRC to the SCDF estimates developed by EPRI for the new site-specific seismic hazards show that there clearly has not been an overall increase in seismic risk for the fleet of U.S. plants. In addition, all sixty-one of the CEUS sites have SCDF estimates below the 1E-4/year threshold considered in the USNRC 2010 Safety / Risk Assessment. Thus it can be concluded that the current seismic design of operating reactors continues to provide a safety margin to withstand potential earthquakes exceeding the seismic design basis, as was concluded in the USNRC 2010 Safety / Risk Assessment.

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9 REFERENCES

ASME (2009). Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications. American Society of Mechanical Engineers and American Nuclear Society Standard ASME/ANS RA-Sb-2009 (Addenda to ASME/ANS RA-S-2008).

ASME (2013). Standard for Level 1/Large Early Release Frequency Probabilistic Risk

Assessment for Nuclear Power Plant Applications. American Society of Mechanical Engineers and American Nuclear Society Standard ASME/ANS RA-Sb-2013 (Addenda to ASME/ANS RA-S-2008), September 2013.

CEUS-SSC (2012). Central and Eastern United States Seismic Source Characterization for

Nuclear Facilities, U.S. Nuclear Regulatory Commission Report, NUREG-2115; Electric Power Research Institute Report 1021097, 6 Volumes; DOE Report# DOE/NE-0140.

EPRI (2013a). Seismic Evaluation Guidance: Screening, Prioritization and Implementation

Details (SPID) for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1: Seismic, Electric Power Research Institute Report 1025287, February 2013.

EPRI (2013b). Seismic Evaluation Guidance: Augmented Approach for the Resolution of

Fukushima Near-Term Task Force Recommendation 2.1 – Seismic, Electric Power Research Institute Report 3002000704, May 2013.

EPRI (2013c). EPRI (2004, 2006) Ground-Motion Model (GMM) Review Project, 2 volumes,

Electric Power Research Institute Report 3002000717, June, 2013. NEI (2013). “Proposed Path Forward for NTTF Recommendation 2.1: Seismic Reevaluations,”

A. Pietrangelo Letter to D. Skeen, Nuclear Energy Institute, 9 April 2013. Kennedy, R.P. (1999). “Overview of Methods for Seismic PRA and Margins Including Recent

Innovations,” Proceedings of the Organization for the Economic Cooperation and Development/Nuclear Energy Agency Workshop on Seismic Risk, Tokyo, Japan, 10 - 12 August 1999.

USNRC (2007). A Performance-Based Approach to Define the Site-Specific Earthquake

Ground Motion, U.S. Nuclear Regulatory Commission Reg. Guide 1.208, U.S. Nuclear Regulatory Commission, Washington, DC.

USNRC (2010). Implications of Updated Probabilistic Seismic Hazard Estimates In Central And

Eastern United States On Existing Plants Generic Issue 199 (GI-199), Safety Risk Assessment, U.S. Nuclear Regulatory Commission, Washington, DC, Aug.

USNRC (2012). “Request for Information Pursuant to Title 10 of the Code of Federal

Regulations 50.54(f) Regarding Recommendations 2.1, 2.3 and 9.3 of the Near-Term Task Force Review of Insights from the Fukushima Dai-Ichi Accident,” E. Leeds and M. Johnson Letter to All Power Reactor Licensees et al., U.S. Nuclear Regulatory Commission, Washington, DC, 12 March.

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USNRC (2013). Electric Power Research Institute Final Draft Report XXXXXX, “Seismic Evaluation Guidance: Augmented Approach for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1: Seismic,” as an Acceptable Alternative to the March 12, 2012, Information Request for Seismic Reevaluations, Eric Leeds Letter to Joseph Pollock (NEI), 7 May 2013.

USNRC (2014). “Supplemental Information Related to Request for Information Pursuant to Title

10 of the Code of Federal Regulations 50.54(f) Regarding Seismic Hazard Reevaluations for Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident,” E. Leeds Letter to All Power Reactor Licensees et al., U.S. Nuclear Regulatory Commission, Washington, DC 20 February 2014.

ATTACHMENT 2

Perspective on the Seismic Capacity of Operating Plants

The fleet of operating nuclear power plants was designed using conservative practices, such that the plants have significant margin to safely withstand large earthquake ground motions. This has been borne out for those plants that have actually experienced significant earthquakes in Japan and the United States in recent years.

Ruggedness Resulting from Design Practices

Operating nuclear power plants were designed based on a “deterministic” or “scenario-earthquake” basis that accounted for the largest earthquakes expected in the regional and local areas around the plant. Further margins were added to the predicted ground motions to provide enhanced robustness. The resulting ground motions comprise the design-basis earthquake, most commonly referred to as the safe shutdown earthquake (SSE).

The SSE ground motions were used to design the plants with conservative methods, resulting in significant seismic margins within structures, systems and components (SSCs). These conservatisms are reflected in several key aspects of the seismic design process, including:

• Safety factors applied in design calculations

• Damping values used in dynamic analysis of SSCs

• Bounding synthetic time histories for in-structure response spectra calculations

• Broadening criteria for in-structure response spectra

• Response spectra enveloping criteria typically used in SSC analysis and testing applications

• Response spectra based frequency domain analysis rather than explicit time history based time domain analysis

• Bounding requirements in codes and standards

• Use of minimum strength requirements of structural components (concrete and steel)

• Bounding testing requirements, and

• Additional capacity in the primary materials such as steel and reinforced concrete beyond the elastic capacity credited in designs

These design practices combine to result in margins such that the SSCs will continue to fulfill their functions at ground motions well above the SSE. In general, ground motions at levels 1½ to 2 times the SSE are typically expected to produce only a small probability of failure (e.g., about 1%) for safety-related SSCs. A common parameter used to characterize this margin is the acceleration level at which there remains a high confidence of a low probability of failure (HCLPF). These margins are accounted for in performing more realistic characterizations of seismic performance, such as those that comprise seismic probabilistic risk assessments (PRAs) or seismic margins analyses (SMAs).

Ruggedness Demonstrated by Experience

Earthquake experience at industrial facilities and nuclear power plants demonstrates that well designed (engineered) facilities perform well, even in very large earthquakes. The EPRI Seismic Qualification Utilities Group (SQUG) program has investigated the performance of equipment in large earthquakes at industrial facilities around the world for over 20 years. Insights from that experience have been applied in the design and evaluation of nuclear plant equipment.

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A SQUG “reference” ground motion response spectrum was developed to represent the ground motion levels from earthquake experience. The SQUG reference ground motion response spectrum illustrates that equipment in the facilities that experienced these earthquakes survived and continued to function at earthquake levels for which there is confidence in the performance of SSCs at nuclear power plants. Figure 1 shows the SQUG reference ground response spectrum1 compared with representative SSEs from a number of central and eastern United States (CEUS) nuclear plants. This figure shows that the ground motions based on experience bound the SSE spectra for typical nuclear power plants. Therefore, the insights applied from operating experience support the conclusion that well designed facilities perform well at and above the earthquake levels required for nuclear power plants.

Figure 1. Comparison of Experience-Based (SQUG) Response Spectrum to Representative SSEs

In recent years, several nuclear plants have experienced significant earthquakes. The experience at these plants has further confirmed the seismic ruggedness of the plants at earthquake levels even beyond the initial design basis. This experience is described below for some of these events.

• Chūetsu Offshore Earthquake (Japan, 2007). In 2007, the Kashiwazaki-Kariwa plant experienced ground motions beyond its seismic design basis from a large earthquake just offshore. While there was some damage to the non-safety, non-seismically designed portions of the plant, none of the safety-related, seismically designed portions of the plant were affected. All safety systems performed as designed to maintain all seven units in a safe condition.

1 SQUG used 2/3 of the reference ground response spectrum for NPP equipment qualification; however, the full reference spectrum represents the ground motions at the industrial sites.

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• Great Japan Earthquake (2011). In 2011, several nuclear plants on the eastern coast of Japan were affected by one of the largest earthquakes in recorded history. The nuclear plant closest to the source of the earthquake was the Onagawa plant, which safely shut down following the earthquake. Similar to the 2007 experience at Kashiwazaki-Kariwa, there was some damage to the non-safety, non-seismically designed portions of the plant; however, all of the safety-related, seismically designed portions of the plant performed as designed to maintain all three units in a safe condition.

The Fukushima Dai-ichi and Daini plants are located south of the Onagawa plant and somewhat farther away from the earthquake epicenter. All of the units at both of these plants performed well following the earthquake. It was only when the tsunami generated by the earthquake struck the sites that significant damage occurred. Prior to the time the tsunami reached the respective sites, the units that were operating had tripped, and the safety-related, seismically designed portions of the plants were successfully providing decay heat removal for all of the units. The extensive damage at the Fukushima Dai-ichi plant caused by the tsunami does not diminish the positive evidence of seismic robustness demonstrated by the thirteen affected units.

• Mineral, VA Earthquake (2011). Later in 2011, the North Anna plant experienced an earthquake originating in Mineral, Virginia. The ground motions at the plant site exceeded a portion of the seismic design basis; however, the safety-related, seismically designed portions of the plant were unaffected and performed as designed to maintain both units in a safe condition. Indeed, there was virtually no observable damage at North Anna, even to non-safety SSCs.

Collectively, the industrial earthquake experience documented by SQUG and the recent earthquake experience at nuclear power plants demonstrate that seismically designed structures, systems, and components are robust, even for ground motions that exceed the seismic design basis.

Significance of High-Frequency Ground Motions

In the CEUS, new ground motion estimates at many sites are expected to include a significant amount of energy content in frequencies above 10 Hz, which for nuclear plants are generally considered to be “high-frequency”.

Limited Potential for High-Frequency Motions to Cause Damage

Nuclear plant SSCs typically have dominant frequency response modes less than 10 Hz. Previous studies have concluded that high-frequency ground motions are not damaging to the majority of nuclear plant SSCs. Examples of these studies include the following EPRI reports:

• Procedure for Evaluating Nuclear Power Plant Relay Seismic Functionality (EPRI NP-7148) [1],

• Industry Approach to Severe Accident Policy Implementation (NP-7498) [2],

• The Effects of High-Frequency Ground Motion on Structures, Components, and Equipment in Nuclear Power Plants (EPRI 1015108) [3], and

• Seismic Screening of Components Sensitive to High-Frequency Vibratory Motions (EPRI 1015109) [4].

EPRI 1015108 summarizes a significant amount of empirical and theoretical evidence, as well as regulatory precedents, that support the conclusion that high-frequency vibratory motions above about 10 Hz are not damaging to the large majority of SSCs in a nuclear power plant. A potential exception is the functional performance of vibration-sensitive components, such as relays and other electrical and instrumentation

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devices whose output signals could be affected by high-frequency excitation. EPRI 1015109 provides guidance for identifying and evaluating potentially sensitive components in plant applications that may be subject to high-frequency motions.

Spectral Acceleration versus Spectral Displacements

When subjected to motions at high frequencies at a given spectral acceleration level, an object’s displacement is significantly less than the displacement at lower frequencies for the same spectral acceleration level. The very small displacement has little potential to cause damage. Design response spectra are typically illustrated as a plot of acceleration versus frequency, as in Figure 2 (although acceleration is only one parameter used for design of nuclear power plants).

A corresponding plot of displacement versus frequency over the same spectral acceleration range can be constructed, as shown in Figure 3. As this figure shows, there is only very small displacement at frequencies greater than 10 Hz. These small displacements do not cause damage to typical structures, systems, and components in a nuclear power plant. Some electrical components (such as relays, switches, and connectors), although, may be sensitive to high-frequency vibrations. As described below, an extensive program is underway to test components that may be sensitive to high-frequency motions.

Figure 2. Example Plot of GMRS and SSE, Indicating High-Frequency Exceedances

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Figure 3. Displacement Corresponding to Ground Accelerations in Figure 2

The NRC has acknowledged the relative inconsequence of high-frequency ground motions. For example, in NUREG-1793 [5], the NRC stated the following:

“At high frequencies of vibratory excitation, the relative displacement is small and produces insignificant increase in stress. As an example, at 25 Hz and a spectral acceleration of 1.0g, the relative displacement is 0.016 inch. This is too small to cause damage.”

Testing Program for Components Potentially Susceptible to High Frequency

EPRI established a test program to develop a more comprehensive understanding of the types of components that could be susceptible to high-frequency motions. As described in EPRI 1025287 [6], Phase 1 of the test program completed in late 2012, developed the appropriate high-frequency testing protocols. Phase 2 testing began in May 2013 using the protocols established in Phase 1, and will continue through March 2014.

As of November 2013, testing had been completed for 100 components. Approximately 75% of those components functioned properly at test levels high enough to bound any practical seismic ground motions. In fact, most of the components performed successfully up to the limits of the shake table (i.e., about 7.5g) as shown in Figure 4. The remaining component high-frequency capacities are above their typical design basis frequency range capacities. The results of the test program will be useful in screening out from further evaluations (i.e., in a SMA or seismic PRA) those components that have especially high capacities, and will help to determine the capacities of those components that cannot be screened out.

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Figure 4. High Frequency Component Capacities: Results of Testing through November 2013

Conclusions Regarding Seismic Ruggedness of Nuclear Power P lants

The SSCs used in nuclear power plants are intentionally designed using conservative methods and criteria to ensure that they have margins well above the required design bases. The ruggedness resulting from these margins has been demonstrated in actual earthquake experiences at nuclear power plants and other industrial facilities. This experience has shown engineered facilities perform well at earthquake ground motions even beyond their design levels.

Several previous studies have concluded that high-frequency ground motions are not damaging to most nuclear plant SSCs. Seismic testing of potentially sensitive components is underway to evaluate potential impacts of high-frequency ground motions. Results to date demonstrate that most components have substantial capacity, and that only a small portion of these potentially sensitive components may require further evaluation.

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1 Procedure for Evaluating Nuclear Power Plant Relay Seismic Functionality. Electric Power Research

Institute Report NP-7148, December 1990.

2 Industry Approach to Severe Accident Policy Implementation. Electric Power Research Institute Report NP-7498, November 1991.

3 Program on Technology Innovation: The Effects of High-Frequency Ground Motion on Structures, Components, and Equipment in Nuclear Power Plants. Electric Power Research Institute Report 1015108, June 2007.

4 Program on Technology Innovation: Seismic Screening of Components Sensitive to High-Frequency Vibratory Motions. Electric Power Research Institute Report 1015109, October 2007.

5. Final Safety Evaluation Report Related to Certification of the AP1000 Standard Design. U.S. Nuclear Regulatory Commission Report NUREG-1793, September 2004.

6 Seismic Evaluation Guidance: Screening, Prioritization and Implementation Details (SPID) for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1: Seismic. Electric Power Research Institute Technical Report 1025287, February 2013.