ML16014A188

Staff Assessment of Information Provided Pursuant to Title 10 of the Code of Federal Regulations Part 50 Section 50.54(f) Seismic Hazard Reevaluations for Recommendation 2.1of the Near-Term Task Force Review of Insights

ML16014A188
Person / Time
Site:	Braidwood
Issue date:	01/22/2016
From:	Frankie Vega Japan Lessons-Learned Division
To:	Bryan Hanson Exelon Generation Co
Vega, Frankie NRR/JLD 415-1617
References
CAC MF3886, CAC MF3887
Download: ML16014A188 (20)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 January 22, 2016 Mr. Bryan C. Hanson Senior Vice President Exelon Generation Company, LLC President and Chief Nuclear Officer Exelon Nuclear 4300 Winfield Road Warrenville, IL 60555

SUBJECT:

BRAIDWOOD STATION, UNITS 1 AND 2 - STAFF ASSESSMENT OF INFORMATION PROVIDED PURSUANT TO TITLE 10 OF THE CODE OF FEDERAL REGULA T/ONS PART 50, SECTION 50.54(f), SEISMIC HAZARD REEVALUATIONS FOR RECOMMENDATION 2.1 OF THE NEAR-TERM TASK FORCE REVIEW OF INSIGHTS FROM THE FUKUSHIMA DAl-ICHI ACCIDENT (CAC NOS. MF3886 AND MF3887)

Dear Mr. Hanson:

On March 12, 2012, the U.S. Nuclear Regulatory Commission (NRC) issued a request for information pursuant to Title 10 of the Code of Federal Regulations, Part 50, Section 50.54(f}

(hereafter referred to as the 50.54(f) letter). The purpose of that request was to gather information concerning, in part, seismic hazards at each operating reactor site and to enable the NRC staff, using present-day NRC requirements and guidance, to determine whether licenses should be modified, suspended, or revoked.

By letter dated March 31; 2014, Exelon Generation Company, LLC (Exelon, the licensee).

responded to this request for Braidwood Station, Units 1 and 2 (Braidwood).

The NRC staff has reviewed the information provided related to the reevaluated seismic hazard for Braidwood and, as documented in the enclosed staff assessment, determined that you provided sufficient information in response to Requested Information Items (1) - (3), (5) - (9) and the comparison portion to Item (4), identified in Enclosure 1 of the 50.54(f) letter. Further, the NRC staff concludes that the licensee's reevaluated seismic hazard is suitable for other actions associated with Near-Term Task Force Recommendation 2.1, "Seismic".

Contigent upon the NRC's review and acceptance of Exelon's high frequency confirmation (Item

4) for Braidwood, the Seismic Hazard Evaluation identified in Enclosure 1 of the 50.54(f) letter will be completed.

B. Hanson If you have any questions, please contact me at (301) 415-1617 or at Frankie.Vega@nrc.gov.

Sincerely, Fra1::f:cl Manager Hazards Management Branch Japan Lessons-Learned Division Office of Nuclear Reactor Regulation Docket Nos. 50-456 and 50-457

Enclosure:

Staff Assessment of Seismic Hazard Evaluation and Screening Report cc w/encl: Distribution via Listserv

STAFF ASSESSMENT BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO SEISMIC HAZARD AND SCREENING REPORT BRAIDWOOD STATION. UNITS 1 AND 2 DOCKET NOS. 50-456 AND 50-457

1.0 INTRODUCTION

By letter dated March 12, 2012 (NRC, 2012a), the U.S. Nuclear Regulatory Commission (NRC or Commission) issued a request for information to all power reactor licensees and holders of construction permits in active or deferred status, pursuant to Title 10 of the Code of Federal Regulations (10 CFR), Section 50.54(f) "Conditions of license" (hereafter referred to as the "50.54(f) letter"). The request and other regulatory actions were issued in connection with implementing lessons-learned from the 2011 accident at the Fukushima Dai-ichi nuclear power plant as documented in the "Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident" (NRC, 2011 b). 1 In particular, the NRC Near-Term Task Force (NTTF)

Recommendation 2.1, and subsequent Staff Requirements Memoranda (SRM) associated with Commission Papers SECY-11-0124 (NRC, 2011c) and SECY-11-0137 (NRC, 2011d),

instructed the NRC staff to issue requests for information to licensees pursuant to 10 CFR 50.54(f). to the 50.54(f) letter requests that addressees perform a reevaluation of the seismic hazards at their sites using present-day NRC requirements and guidance to develop a ground motion response spectrum (GMRS).

The required response section of Enclosure 1 requests that each addressee provide the following information:

(1) Site-specific hazard curves (common fractiles and mean) over a range of spectral frequencies and annual exceedance frequencies, (2) Site-specific, performance-based GMRS developed from the new site-specific seismic hazard curves at the control point elevation, (3) Safe Shutdown Earthquake (SSE) ground motion values including specification of the control point elevation, (4) Comparison of the GMRS and SSE. A high frequency (HF) evaluation (if necessary),

1 Issued as an enclosure to Commission Paper SECY-11-0093 (NRC, 2011a).

Enclosure

(5) Additional information such as insights from NTTF Recommendation 2.3 walkdown and estimates of plant seismic capacity developed from previous risk assessments to inform NRC screening and prioritization, (6) Interim evaluation and actions taken or planned to address the higher seismic hazard relative to the design basis, as appropriate, prior to completion of the risk evaluation (if necessary),

(7) Selected risk evaluation approach (if necessary),

(8) Seismic risk evaluation (if necessary), and (9) Spent fuel pool (SFP) evaluation (if necessary).

Present-day NRC requirements and guidance with respect to characterizing seismic hazards use a probabilistic approach in order to develop a risk-informed performance-based GMRS for the site. Regulatory Guide (RG) 1.208, A Performance-based Approach to Define the Site-Specific Earthquake Ground Motion (NRC, 2007), describes this approach. As described in the 50.54(f) letter, if the reevaluated seismic hazard, as characterized by the GMRS, is not bounded by the current plant design basis SSE, further seismic risk evaluation of the plant is merited.

By letter dated November 27, 2012 (Keithline, 2012), NEI submitted EPRI report "Seismic Evaluation Guidance: Screening, Prioritization, and Implementation Details (SPID) for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1 Seismic"(EPRI, 2012),

hereafter called the SPID. The SPID supplements the 50.54(f) letter with guidance necessary to perform seismic reevaluations and report the results to NRC in a manner that will address the Requested Information Items in Enclosure 1 of the 50.54(f) letter. By letter dated February 15, 2013 (NRC, 2013b), the staff endorsed the SPID.

The required response section of Enclosure 1 to the 50.54(f) letter specifies that Central and Eastern United States (CEUS) licensees provide their Seismic Hazard and Screening Report (SHSR) by 1.5 years after issuance of the 50.54(f) letter. However, in order to complete its update of the EPRI seismic ground motion models (GMM) for the CEUS (EPRI, 2013), industry proposed a six-month extension to March 31, 2014, for submitting the SHSR. Industry also proposed that licensees perform an expedited assessment, referred to as the Augmented Approach, for addressing the requested interim evaluation (Item 6 above), which would use a simplified assessment to demonstrate that certain key pieces of plant equipment for core cooling and containment functions, given a loss of all alternating current power, \/\'.OUld be able to withstand a seismic hazard up to two times the design basis. Attachment 2 to the April 9, 2013, letter (Pietrangelo, 2013) provides a revised schedule for plants needing to perform (1) the Augmented Approach by implementing the Expedited Seismic Evaluation Process and (2) a seismic risk evaluation. By letter dated May 7, 2013 (NRC, 2013a), the NRC determined that the modified schedule was acceptable and by letter dated August 28, 2013 (NRC, 2013c), the NRC determined that the updated GMM (EPRI, 2013) is an acceptable GMM for use by CEUS plants in developing a plant-specific GMRS.

By letter dated April 9, 2013 (Pietrangelo, 2013), industry committed to following the SPID to develop the SHSR for existing nuclear power plants. By letter dated September 12, 2013 (Kaegi, 2013), Exelon Generation Company, LLC (Exelon, the licensee) submitted partial site response information for the Braidwood Station, Units 1 and 2 (Braidwood) site. By letter dated March 31, 2014 (Kaegi, 2014), Exelon submitted its SHSR.

2.0 REGULATORY BACKGROUND The structures, systems, and components (SSCs) important to safety in operating nuclear power plants are designed either in accordance with, or meet the intent of Appendix A to 10 CFR Part 50, General Design Criteria (GDC) 2: "Design Bases for Protection Against Natural Phenomena;" and Appendix A to 10 CFR Part 100, "Reactor Site Criteria." The GDC 2 states that SSCs important to safety at nuclear power plants shall be designed to withstand the effects of natural phenomena such as earthquakes, tornadoes, hurricanes, floods, tsunami, and seiches without loss of capability to perform their safety functions.

For initial licensing, each licensee was required to develop and maintain design bases that, as defined by 10 CFR 50.2, identify the specific functions that an SSC of a facility must perform, and the specific values or ranges of values chosen for controlling parameters as reference bounds for the design. The design bases for the SSCs reflect appropriate consideration of the most severe natural phenomena that had been historically reported for the site and surrounding area. The design bases also considered limited accuracy, quantity, and period of time in which the historical data have been accumulated.

The seismic design bases for currently operating nuclear power plants were either developed in accordance with, or meet the intent of GDC 2 and 10 CFR Part 100, Appendix A. Although the regulatory requirements in Appendix A to 10 CFR Part 100 are fundamentally deterministic, the NRC process for determining the seismic design basis ground motions for new reactor applications after January 10, 1997, as described in 10 CFR 100.23, requires that uncertainties be addressed through an appropriate analysis such as a probabilistic seismic hazard analysis (PSHA).

Section 50.54(f) of 10 CFR states that a licensee shall at any time before expiration of its license, upon request of the Commission, submit written statements, signed under oath or affirmation, to enable the Commission to determine whether or not the license should be modified, suspended, or revoked. On March 12, 2012, the NRC staff issued requests for licensees to reevaluate the seismic hazards at their sites using present-day NRC requirements and guidance, and identify actions planned to address plant-specific vulnerabilities associated with the updated seismic hazards. to Enclosure 1 of the 50.54(f) letter describes an acceptable approach for performing the seismic hazard reevaluation for plants located in the CEUS. Licensees are expected to use the CEUS Seismic Source Characterization (CEUS-SSC) model in NUREG- 2115 (NRC, 2012b) along with the appropriate EPRI (2004, 2006) GMMs. The SPID

provides further guidance regarding the appropriate use of GM Ms for the CEUS. Specifically, Section 2.3 of the SPID recommends the use of the updated GMM (EPRI 2013) and, as such, licensees used the NRG-endorsed updated EPRI GMM instead of the older EPRI (2004, 2006)

GMM to develop PSHA base rock hazard curves. Finally, Attachment 1 requests that licensees conduct an evaluation of the local site response in order to develop site-specific hazard curves and GMRS for comparison with the plant SSE.

2.1 Screening Evaluation Results By letter dated March 31, 2014 (Kaegi, 2014), the licensee provided its SHSR for the Braidwood site. The licensee's SHSR indicates that the site SSE bounds the GMRS for the Braidwood site over the frequency range of 1 to 10 Hertz (Hz). As such, Braidwood screens out of performing a seismic risk evaluation, as well as a SFP evaluation. However, since the GMRS exceeds the SSE above 10 Hz, the licensee stated that it will perform a HF confirmation.

On May 9, 2014 (NRC, 2014), the NRC staff issued a letter providing the outcome of its 30-day screening and prioritization evaluation. As indicated in the letter, the NRC staff confirmed the licensee's screening results. The licensee's GMRS, as well as the staff's confirmatory GMRS, are bounded by the SSE for Braidwood over the frequency range of 1 to approximately 11 Hz.

Therefore, a seismic risk evaluation and SFP evaluation are not merited. The NRC staff also confirmed the licensee's conclusion that a HF confirmation for Braidwood is merited because the GMRS exceeds the SSE at frequencies above 10 Hz.

3.0 TECHNICAL EVALUATION

The NRC staff evaluated the licensee's submittal to determine if the provided information responded appropriately to Enclosure 1 of the 50.54(f) letter with respect to characterizing the reevaluated seismic hazard.

3.1 Plant Seismic Design Basis of the 50.54(f) letter requests the licensee provide the SSE ground motion values, as well as the specification of the control point elevation(s) for comparison to the GMRS. For operating reactors licensed before 1997, the SSE is the plant licensing basis earthquake and is characterized by (1) a peak ground acceleration (PGA) value which anchors the response spectra at high frequencies (typically at 20 to 30 Hz for the existing fleet of nuclear power plants; (2) a response spectrum shape which depicts the amplified response at all frequencies below the PGA; and (3) a control point where the SSE is defined.

In Section 3.1 of its SHSR, the licensee described its seismic design bases for the Braidwood site by stating that the SSE for the Braidwood site has a RG 1.60 spectral shape with a PGA that is based on a postulated Intensity VIII earthquake. Based on this earthquake, the response spectral shape was anchored at a PGA of 0.20 g (20 percent of the acceleration due to earth's gravity). In Section 3.2 of its SHSR, the licensee specified that the SSE control point is defined

at an elevation of 562 ft. (171 m) at the top of the Pennsylvanian limestone, which is the bedrock-till interface at the Braidwood site.

The NRC staff reviewed the licensee's description of its SSE in the SHSR for the Braidwood site. With regard to the Braidwood site SSE, based on its review of the SHSR and Updated Final Safety Analysis Report (UFSAR) (Exelon, 2014), the NRC staff confirmed that the licensee's SSE is defined in terms of a RG 1.60 design response spectrum anchored at a PGA of 0.2 g, as described by the licensee. Based on its review of the SHSR and the UFSAR (Exelon, 2014 and Kaegi, 2014), the NRC staff confirmed that the licensee's control point elevation for the Braidwood site SSE is consistent with the guidance provided in the SPID.

3.2 Probabilistic Seismic Hazard Analysis In Section 2.2 of its SHSR, the licensee stated that, in accordance with the 50.54(f) letter and the SPID, it performed a PSHA using the CEUS-SSC model and the updated EPRI GMM for the CEUS (EPRI, 2013). The licensee used a minimum magnitude (M) of 5.0, as specified in the 50.54(f) letter. The licensee further stated that it included the CEUS-SSC background sources out to a distance of 400 miles (640 km) around the site and included the Commerce, Eastern Rift Margin Fault northern segment (ERM-N), Eastern Rift Margin Fault southern segment (ERM-S), Marianna, New Madrid Fault System (NMFS), and Wabash Valley repeated large magnitude earthquake (RLME) sources, which lie within 621 mi (1,000 km) of the site. The RLME sources are those source areas or faults for which more than one large magnitude (M ~

6.5) earthquake has occurred in the historical or paleo-earthquake (geologic evidence for prehistoric seismicity) record. The licensee used the mid-continent version of the updated EPRI GMM (EPRI, 2013) for each of the CEUS-SSC sources. Consistent with the SPID, the licensee did not provide its base rock seismic hazard curves since a site response analysis is necessary to determine the control point seismic hazard curves. The licensee provided its control point seismic hazard curves in Section 2. 3. 7 of its SHSR. The staff's review of the licensee's control point seismic hazard curves is provided in Section 3.3 of this staff assessment.

As part of its confirmatory analysis of the licensee's GMRS, the NRC staff performed its own PSHA calculations for base rock site conditions at the Braidwood site. As input, the NRC staff used the CEUS-SSC model as documented in NUREG-2115 (NRC, 2012b) along with the updated EPRI GMM (EPRI, 2013). Consistent with the guidance provided in the SPID, the NRC staff included all CEUS-SSC background seismic sources within a 310 mi (500 km) radius of the Braidwood site. In addition, the NRC staff included RLME sources which lie within 621 mi (1,000 km) of the site. For each of the CEUS-SSC sources used in the PSHA, the NRC staff used the mid-continent version of the updated EPRI GMM (EPRI, 2013). The NRC staff used the resulting base rock seismic hazard curves together with a confirmatory site response analysis, described in the next section, to develop control point seismic hazard curves and a GMRS for comparison with the licensee's results.

Based on its review of the SHSR, the NRC staff concludes that the licensee followed guidance provided in the SPID for selecting PSHA input models and parameters for the site. This

includes the licensee's use and implementation of the CE US-SSC model and the updated EPRI GMM.

3.3 Site Response Evaluation After completing PSHA calculations for reference rock site conditions, Attachment 1 to Enclosure 1 of the 50.54(f) letter requests that the licensee provide a GMRS developed from the site-specific seismic hazard curves at the control point elevation. In addition, the 50.54(f) letter specifies that the subsurface site response model, for both soil and rock sites, should extend to sufficient depth to reach the generic or base rock conditions as defined in the GMMs used in the PSHA. To develop site-specific hazard curves at the control point elevation, Attachment 1 requests that the licensee perform a site response analysis.

Detailed site response analyses were not typically performed for many of the older operating plants; therefore, Appendix B of the SPID provides detailed guidance on the development of site-specific amplification factors (including the treatment of uncertainty) for sites that do not have detailed, measured soil and rock parameters to extensive depths.

The purpose of the site response analysis is to determine the site amplification that will occur as a result of bedrock ground motions propagating upwards through the soil/rock column to the surface. The critical parameters that determine what frequencies of ground motion are affected by the upward propagation of bedrock motions are the layering of soil and/or soft rock, the thicknesses of these layers, the shear-wave velocities and low-strain damping of the layers, and the degree to which the shear modulus and damping change with increasing input bedrock amplitude.

3.3.1 Site Base Case Profiles The licensee provided detailed site profile descriptions in Sections 2.3.1 and 2.3.2 of its SHSR based on information cited in the UFSAR (Exelon, 2014). The Braidwood site is located near Joliet, Illinois within the Central Lowland Physiographic Province and is underlain by approximately 42 ft. (13 m) of soils overlying 5,000 ft. (1,524 m) of firm sedimentary rock. The licensee indicates that the site consists of approximately 42 ft. (13 m) of eolian and lacustrine deposits on top of approximately 31 ft. (9 km) of Pleistocene age soil deposits. These soils overlay Pennsylvanian to Cambrian bedrock deposits consisting of sandstone, siltstone, shale, limestone, and dolomite formations. The licensee described the Precambrian basement rock as primarily medium to coarse-grained granite.

In Table 2.3.2-1 of its SHSR, the licensee provides the shear-wave velocities determined from seismic refraction and shear-wave velocity surveys performed for siting the independent spent fuel storage installation (ISFSI) at the Braidwood site. As described by the licensee, these geophysical surveys extend 700 ft. (213 m) below the depth of the control point elevation which is located at 562 ft. at the top of the Pennsylvanian limestone. The shear-wave velocity in the upper 224 ft. (68 m) range from 3200 to 6800 ft/sec (975 to 2072 m/sec).

To capture the uncertainty in the subsurface geology, the licensee developed three site base case profiles for its site response analysis. The licensee developed the middle, or best estimate, profile using the measured shear-wave velocities from nearby downhole geophysical investigations to a depth of 224 ft. (68 m). Below a depth of 224 ft. (68 m), the licensee's extrapolated the shear wave velocities based on measurements made in the sound rock and assumed that base reference rock conditions are reached at an assumed depth of 5,062 ft.

(1,543 m). To develop the lower and upper base case profiles, the licensee used a natural log standard deviation of 0.2 and 0.35 respectively. Table 2.3.2-1 and Figure 2.3.2-1 of the SHSR provide the licensee's shear-wave velocity profile for each of the three base cases. Figure 3.3-1 of this assessment shows the licensee's three shear-wave velocity base case profiles.

As described in Section 2.3.2.1 of the SHSR in the absence of any site-specific dynamic material property measurements, the licensee followed the SPID guidance for rock sites and selected two alternative characterizations of site-specific dynamic material behavior. In the first characterization, the licensee modeled non-linear behavior using the generic EPRI rock shear modulus and damping curves over the upper 500 ft. (152 m). In the second characterization, the licensee modeled linear behavior with a low strain damping value from the EPRI rock damping curves in the upper 500 ft. (152 m). The licensee assigned equal weights to the two characterizations.

The licensee also considered the impact of kappa, or small strain damping, on site response.

Kappa is measured in units of seconds (sec), and is the damping contributed by both intrinsic hysteretic damping, as well as scattering due to wave propagation in heterogeneous material.

For the Braidwood site, the licensee used the low strain damping values, shear-wave velocities, layer thicknesses, and bedrock damping to arrive at kappa values for the best estimate, upper, and lower base case velocity profiles of 0.024, 0.031, and 0.019 sec, respectively. The licensee also applied a scale factor of 1.68 to the base-case profile kappa values to calculate lower range kappa values of 0.014, 0.018, and 0.011 sec and upper range kappa values of 0.040, 0.040, and 0.032 sec.

To account for randomness in material properties across the Braidwood site, the licensee stated in Section 2.3.3 of its SHSR, that it randomized its base case shear-wave velocity profiles in accordance with the SPID. In addition, as stated in Section 2.3.2 of its SHSR, the licensee randomized the depth to bedrock by +/-1,519 ft. (+/-462 m), which corresponds to 30 percent of the total profile thickness. The licensee stated that this randomization did not represent the actual uncertainty in the depth to bedrock, but was used to broaden the spectral peaks.

3.3.2 Site Response Method and Results In Section 2.3.4 of its SHSR, the licensee stated that it followed the guidance in Appendix B of the SPID to develop input ground motions for the site response analysis and in Section 2.3.5, the licensee described its implementation of the random vibration theory (RVT) approach to perform its site response calculations. Finally, Section 2.3.6 of the SHSR shows the resulting amplification functions and associated uncertainties for two of the eleven input loading levels for the base case profile and EPRI soil and rock shear modulus and damping curves.

In order to develop probabilistic site-specific control point hazard curves, as requested in Requested Information Item 1 of the 50.54(f) letter, the licensee used Method 3, described in Appendix B of the SPID. The licensee's use of Method 3 involved computing the site-specific control point elevation hazard curves for a broad range of spectral accelerations by combining the site-specific bedrock hazard curves, determined from the initial PSHA (Section 3.2 of this assessment), and the amplification functions and their associated uncertainties, determined from the site response analysis.

3.3.3 Staff Confirmatory Analysis To confirm the licensee's site response analysis, the NRC staff performed site response calculations for the Braidwood site. The NRC staff independently developed a shear-wave velocity profile, damping values, and modeled the potential nonlinear behavior of the subsurface using measurements and geologic information provided in the Braidwood UFSAR and ISFSI information, and Appendix B of the SPID.

Due to the abundant amount of subsurface data for the Braidwood site, which the licensee incorporated into its best-estimate base case velocity profile, the NRC staff used the same profile with only minor modifications for its best-estimate profile. To capture the uncertainty in the depth to base rock, the NRC staff used a value of +/-500 ft. (+/-152 m), which is less than the

+/-1,519 ft. (+/-463 m), depth to base rock randomization used by the licensee. Similar to the licensee's approach, the NRC staff also developed lower and upper profiles to capture the uncertainty in the subsurface shear wave velocities across the site. Figure 3.3-1 of this assessment illustrates the licensee's and staff's base case shear wave velocity profiles.

Consistent with the guidance in the SPID and the approach used by the licensee, the NRC staff used two alternative characterizations of site-specific dynamic material behavior. In one characterization, the NRC staff modeled non-linear behavior using the generic EPRI rock shear modulus and damping curves over the upper 500 ft. [152 m] of the site profile. In the second characterization, the NRC staff modeled linear behavior with a low strain damping value from the EPRI rock damping curves. Below a depth of 500 ft [152 m], the NRC staff assumed linear behavior for the rock with no damping.

The NRC staff considered the impact of kappa on site response. To determine kappa for its case profiles, the NRC staff used the low strain damping values, shear-wave velocities, layer thicknesses, and bedrock damping to arrive at kappa values for the best estimate, upper, and lower base case velocity profiles of 0.006, 0.008, and 0.005 sec, respectively, which are similar to the licensee's kappa values for its three profiles.

Figure 3.3-2 of this assessment shows a comparison of the staff's and licensee's median site amplification functions and uncertainties (+/-1 standard deviation) for 2 of the 11 input loading levels. The staff's median site amplification factors are similar to the licensee's but peak between 2 to 5 Hz whereas the licensee's amplification factors peak at between 3 to 7 Hz frequency range.

Based on the staff's confirmatory calculations, the NRC staff concludes that the licensee's evaluation for the Braidwood site adequately captures the site amplification occurring as a result of bedrock ground motions traveling upward through the rock column to the control point elevation. Overall, the licensee's approach to modeling the subsurface rock properties and to modeling their uncertainty results in similar amplification factors as those developed by the NRC staff. As shown in Figure 3.3-3 of this assessment, the NRC staff notes these differences in the site response do not have a large impact on the control point seismic hazard curves or the resulting GMRS, as discussed below. Appendix B of the SPID provides guidance for performing site response analyses, including capturing the uncertainty for sites with less subsurface data; however, the guidance is neither entirely prescriptive nor comprehensive. As such, various approaches in performing site response analyses, including the modeling of uncertainty, are acceptable for the 50. 54(f) response.

In summary, the NRC staff concludes that the licensee's site response was conducted using present-day guidance and methodology, including the NRG-endorsed SPID. The NRC staff performed independent calculations which confirmed that the licensee's amplification factors and control point hazard curves adequately characterize the site response, including the uncertainty associated with the subsurface material properties, for the Braidwood site.

3.4 Ground Motion Response Spectra In Section 2.4 of its SHSR, the licensee stated that it used the control point hazard curves, described in SHSR Section 2.3.7, to develop the 10-4 and 10-5 (mean annual frequency of exceedance) uniform hazard response spectra (UHRS) and then computed the GMRS using the criteria in RG 1.208.

The NRC staff independently calculated the 10-4 and 10-5 UHRS using the results of its confirmatory PSHA and site response analyses, as described in Sections 3.2 and 3.3 of this staff assessment, respectively. As shown in Figure 3.4-1 of this assessment, the licensee's GMRS shape is generally similar to that calculated by the NRC staff. However, the licensee's GMRS is higher than the staff's in the range of approximately 5 to 30 Hz. As described above in Section 3.3, the NRC staff concludes that these differences are primarily due to the differences in the site response analyses performed by the licensee and NRC staff.

The NRC staff confirms that the licensee used the present-day guidance and methodology outlined in RG 1.208 and the SPID to calculate the horizontal GMRS, as requested in the 50.54(f) letter. The NRC staff performed both a PSHA and site response confirmatory analysis and achieved results consistent with the licensee's horizontal GMRS. As such, the NRC staff concludes that the GMRS determined by the licensee adequately characterizes the reevaluated hazard for the Braidwood site. Therefore, this GMRS is suitable for use in subsequent evaluations and confirmations, as needed, for the response to the 50.54(f) letter.

4.0 CONCLUSION

The NRC staff reviewed the information provided by the licensee for the reevaluated seismic hazard for the Braidwood site. Based on its review, the NRC staff concludes that the licensee conducted the hazard reevaluation using present-day methodologies and regulatory guidance, it appropriately characterized the site given the information available, and met the intent of the guidance for determining the reevaluated seismic hazard. Based upon the preceding analysis, the NRC staff concludes that the licensee provided an acceptable response to Requested Information Items (1) - (3), (5) - (7), and the comparison portion to Item (4) identified in of the 50.54(f) letter. Further, the licensee's reevaluated seismic hazard is acceptable to address other actions associated with NTTF Recommendation 2.1: "Seismic".

In reaching this determination, the NRC staff confirms the licensee's conclusion that the licensee's GMRS for the Braidwood site is bounded by the SSE in the 1 to 11 Hz range, and exceeds the SSE above approximately 11 Hz. As such, a seismic risk evaluation (Item 8) and SFP evaluation (Item 9) are not merited, however a HF confirmation (Item (4)) is merited. NRC review and acceptance of Exelon's HF confirmation for Braidwood will complete the Seismic Hazard Evaluation identified in Enclosure 1 of the 50.54(f) letter.

REFERENCES Note: ADAMS Accession Nos. refer to documents available through NRC's Agency wide Document Access and Management System (ADAMS). Publicly-available ADAMS documents may be accessed through http://www.nrc.gov/reading-rm/adams.html.

U.S. Nuclear Regulatory Commission Documents and Publications NRC (U.S. Nuclear Regulatory Commission), 2007, A Performance-based Approach to Define the Site-Specific Earthquake Ground Motion, Regulatory Guide (RG) 1.208, March 2007.

NRC (U.S. Nuclear Regulatory Commission), 1973, Design Response Spectra for Seismic Design of Nuclear Power Plants, Regulatory Guide (RG) 1.60, December 1973.

NRC (U.S. Nuclear Regulatory Commission), 2011a, "Near-Term Report and Recommendations for Agency Actions Following the Events in Japqn," Commission Paper SECY-11-0093, July 12, 2011, ADAMS Accession No. ML11186A950.

NRC (U.S. Nuclear Regulatory Commission), 2011 b, "Recommendations for Enhancing Reactor Safety in the 21 51 Century: The Near-Term Task Force Review of Insights from the Fukushima Dai-lchi Accident," Enclosure to SECY-11-0093, July 12, 2011, ADAMS Accession No. ML11186A950.

NRC (U.S. Nuclear Regulatory Commission), 2011c, "Recommended Actions to be Taken Without Delay from the Near-Term Task Force Report," Commission Paper SECY 0124, September 9, 2011, ADAMS Accession No. ML11245A158.

NRC (U.S. Nuclear Regulatory Commission), 2011 d, "Prioritization of Recommended Actions to be Taken in Response to Fukushima Lessons Learned," Commission Paper SECY 0137, October 3, 2011, ADAMS Accession No. ML11272A111.

NRC (U.S. Nuclear Regulatory Commission), 2012a, letter from Eric J. Leeds, Director, Office of Nuclear Reactor Regulation and Michael R. Johnson, Director, Office of New Reactors, to All Power Reactor Licensees and Holders of Construction Permits in Active or Deferred Status, March 12, 2012, ADAMS Accession No. ML12053A340.

NRC (U.S. Nuclear Regulatory Commission), 2012b, "Central and Eastern United States Seismic Source Characterization for Nuclear Facilities", NUREG-2115, ADAMS stores the NUREG as multiple ADAMS documents, which are accessed through the web page http://www.nrc.gov/reading-rm/doc-collections/nuregs/staff/sr2115/.

NRC (U.S. Nuclear Regulatory Commission), 2013a. Letter From Eric J. Leeds, to Joseph Pollock, Executive Director NEI, Acceptance Letter for NEI Submittal of Augmented Approach, Ground Motion Model Update Project, and 10 CFR 50.54(f) Schedule

Modifications Related to the NTTF Recommendation 2.1, Seismic Reevaluations, May 7, 2013, ADAMS Accession No. ML13106A331.

NRC (U.S. Nuclear Regulatory Commission), 2013b, letter from David L. Skeen, Director, Japan Lessons-Learned Directorate, to Joseph E. Pollock, Executive Director, Nuclear Energy Institute, Endorsement of Electric Power Research Institute Draft Report 1025287, "Seismic Evaluation Guidance," February 15, 2013, ADAMS Accession No. ML12319A074.

NRC (U.S. Nuclear Regulatory Commission), 2013c. Letter from D. L. Skeen (NRC) to K. A Keithline (NEI), Approval of Electric Power Research Institute Ground Motion Model Review Project Final Report for Use by Central and Eastern United States Nuclear Power Plants, August 28, 2013 ADAMS Accession No. ML13233A102.

NRC (U.S. Nuclear Regulatory Commission) 2014a. Letter from Eric J. Leeds, Director, Office of Nuclear Reactor Regulation to All Power Reactor Licensees and holders of Construction Permits in Active or Deferred Status, Seismic Screening and Prioritization Results Regarding Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(f)

Regarding Seismic Hazard Reevaluations for Recommendations 2.1 of the Near-Term Task Force Review of Insights, May 9, 2014, ADAMS Accession No. ML14111A147.

Other References AMEC, 2012, Data for Site Amplifications- McGuire Phase 2 EPRI Seismic Attenuation and GMRS Project, McGuire Nuclear Station, Project No. 6234-12-0031, July 26, 2012.

Electric Power Research Institute (EPRI), 2004. EPRI Report 1009684, "CEUS Ground Motion Project Final Report." Palo Alto, CA, 2004.

Electric Power Research Institute (EPRI), 2006. EPRI Report 1014381, "Truncation of the Lognormal Distribution and Value of the Standard Deviation for Ground Motion Models in the Central and Eastern United States." Palo Alto, CA, 2006.

Electric Power Research Institute (EPRI), 2012. EPRI Report 1025287 "Seismic Evaluation Guidance, Screening, Prioritization and Implementation Details [SPID] for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1: Seismic" November 27, 2012, ADAMS Accession No. ML12333A170.

Electric Power Research Institute (EPRI), 2013. EPRI Report 1025770, "EPRI (2004, 2006)

Ground Motion Model (GMM) Review Final Report." Palo Alto, CA, 2013.

Exelon Generation Company (Exelon), 2014, Braidwood Station, Units 1 and 2, Updated Final Safety Analysis Report (UFSAR), Revision 15, December 14, 2014, ADAMS Accession No. ML14363A393.

Kaegi, G.T., 2013, Letter from G.T. Kaegi (Exelon Generation Company, LCC) to NRC, Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding the Seismic Aspects of Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident - 1.5 Year Response for CEUS Sites, September 12, 2013, ADAMS Accession No. ML13256A070.

Kaegi, G.T., 2014, Letter from G.T. Kaegi (Exelon Generation Company, LCC) to NRC, Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.50(f) regarding Recommendations 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident, March 31, 2014, ADAMS Accession Nos. ML14091A005 and ML14091A006.

Keithline, 2012, Letter from Kimberly Keithline, Senior Project Manager, NEI, to David L. Skeen, Director, Japan Lessons Learned Project Directorate, NRC, Final Draft of Industry Seismic Evaluation Guidance (EPRI 1025287), November 27, 2012, ADAMS Accession No. ML12333A168.

Pietrangelo, 2013. Letter from A. R. Pietrangelo (NEI) to D. L. Skeen (NRC), Proposed Path Forward for NTTF Recommendation 2.1: Seismic Reevaluations, April 9, 2013, ADAMS Accession No. ML13101A379.

Figure 3.3-1 Plot of Staff's and Licensee's Base Case Shear-Wave Velocity Profiles for the Braidwood site Shear-Wave Veloc ity (ft/sec) 2000 3000 4000 5000 6000 7000 8000 9000 10000 0 , _____ I I 1__ _

c::::_....____.._____.____~---.----~-~-.----~-------.-~_...,

1000 I +

I

.I I

I t

I I

I I

I I

I

~ 2000 I

c

• c; Q.

f-I I

I I

I I

~

c0

(.)

~ .,. i 3000 l Qi I I  :

ID

-s<>.

I I

I

';, I I \ I c I I

4000 I

I I

I

I I

\I

- - Licensee Base Case I I

I

\1

- - License Lower Base Case

- - Licensee Upper Base Case I

I \;

---

* NRG Low er Base Case

':~

i 5000 - * - NRG Upper Ba se Case *-*------------------------------- !- -----~

......... NRG Base Case

Figure 3.3-2 Plot Comparing the Staff's and the License's Median Amplification Functions and Uncertainties for two input loading levels for the Braidwood site 2.00 i-;:::======================:::;-~~~~~~~~i-~~~~~~~~~-i

- NRG PGA= 0.05g

- NRG PGA= 0.3g

- -

• NRG PGA= 0.05g Sigma n(AF)

- -

• NRG PGA = 0.3g Sig ma ln(AF)

- Licensee PGA = 0.0459g

- Licensee PGA =0.292g 1.50 - -

• Licensee PGA = 0.0459g Sig ma ln(AF) ! - - - - -- - - - - - - - - + - - - - - - - - - - - - -

- -

• Licensee PGA = 0.292g Sigma ln(AF) 0.1 1 10 100 Frequency (Hz)

Figure 3.3-3 Plot Comparing the Staff's and the Licensee's Mean Control Point Hazard Curves at a Variety of Frequencies for the Braidwood site 1.E-02

- u censeePGA

- Licensee 10 Hz

- - Licensee 1 Hz

- - NRCPGA l .E-03 NRC 10 Hz NRC 1 Hz

.8 Cl

"'Cl Cl l .E-04 8

JS

....0

.ti ell

=

er l .E-05

...=

<(

1 .E-06 l .E-07 -+---------,---------.------------f'-~~---~-___,

0 .001 0 .01 0 .1 1 10 Spectral Aa:elerat ion (g)

Figure 3.4-1 Comparison of the Staff's GMRS with Licensee's GMRS and the SSE for the Braidwood site 1.4 1.2 c

.,g

., 0.8 8

~ 0.6 0.4 0.2 10 100 Frequency (Hz)

- Broldwod SSE - LicenseeGM RS - NRCGIA RS

ML16014A188 *via email OFFICE NRR/JLD/JHMB/PM NRR/JLD/LA NRO/DSEA/RGS/BC* NRR/JLD/JHMB/BC NRR/JLD/JHMB/PM NAME FVega Slent DJackson MShams FVega DATE 01/20/2016 01/15/2016 01/08/2016 01/21/2016 01/22/2016