ML15296A492
| ML15296A492 | |
| Person / Time | |
|---|---|
| Site: | Limerick  |
| Issue date: | 11/06/2015 |
| From: | Vega F G Japan Lessons-Learned Division |
| To: | Bryan Hanson Exelon Nuclear |
| Vega, Frankie, NRR/JLD 415-1617 | |
| References | |
| TAC MF3864, TAC MF3865 | |
| Download: ML15296A492 (19) | |
Text
Mr. Bryan C. Hanson President and Chief Nuclear Officer Exelon Nuclear 4300 Winfield Rd. Warrenville, IL 60555 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 November 6, 2015
SUBJECT:
LIMERICK GENERATING STATION, UNITS 1 AND 2 -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 2.1 OF THE NEAR-TERM TASK FORCE REVIEW OF INSIGHTS FROM THE FUKUSHIMA DAl-ICHI ACCIDENT (TAC NOS. MF3864 AND MF3865)
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 Limerick Generating Station, Units 1 and 2 (Limerick).
The NRC staff has reviewed the information provided related to the reevaluated seismic hazard for Limerick 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 the Exe!on's high frequency confirmation (Item 4) for Limerick, 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.
Docket Nos. 50-352 and 50-353
Enclosure:
Staff Assessment of Seismic Hazard Evaluation and Screening Report cc w/encl: Distribution via Listserv SinfV? Frankie Vega, Project Manager Hazards Management Branch Japan Lessons-Learned Division Office of Nuclear Reactor Regulation STAFF ASSESSMENT BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO SEISMIC HAZARD AND SCREENING REPORT LIMERICK GENERATING STATION, UNITS 1 AND 2 DOCKET NOS. 50-352 AND 50-353
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 1 O 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 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 1 O CFR 50.54(f).
Enclosure 1 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) Statement if a seismic risk evaluation is 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), the Nuclear Energy Institute (NEI) submitted Electric Power Research Institute (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, would 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 (ESEP) 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 NRG determined that the updated GMM (EPRI, 2013) is an acceptable GMM for use by GEUS plants in developing a plant-specific GMRS. By letter dated April 9, 2013 (Pietrangelo, 2013), industry agreed to follow 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 at least partial site response information for Limerick Generating Station, Units 1 and 2 (LGS). By letter dated March 31, 2014 (Barstow, 2014), the licensee 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 1 O 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 NRG process for determining the seismic design-basis ground motions for new reactor applications after January 10, 1997, as described in 1 O 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. Attachment 1 to Enclosure 1 of the 50.54(f) letter described 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) ground motion models. The SPID provided further guidance regarding the appropriate use of GMMs for the CEUS. Specifically, Section 2.3 of the SPID recommended 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 requested 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 (Barstow, 2014), the licensee provided its SHSR for the LGS site. The licensee's SHSR indicates that the plant GMRS is bounded by the SSE. The licensee stated that a seismic risk evaluation will not be performed.
Also, a SFP evaluation will not be performed.
Additionally, due to exceedances at frequencies above 10 Hertz (Hz), the licensee indicated that a HF confirmation will be performed.
On May 9, 2014 (NRC, 2014a), the NRC staff issued a letter providing the outcome of its 30-day screening and prioritization evaluation.
In the letter, the NRC staff characterized the LGS site as conditionally screened-in, because additional information was needed to support a screening and prioritization decision.
On October 3, 2014 (NRC, 2014b), the NRC staff issued a letter providing the outcome of its final seismic screening and prioritization results. The licensee's GMRS, as well as the staff's confirmatory GMRS, is bounded by the SSE for LGS over the frequency range of 1 to 10 Hz, but exceeds the SSE for LGS above approximately10 Hz. Therefore, a seismic risk evaluation and SFP evaluation are not merited but a HF confirmation is merited for LGS. 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 Enclosure 1 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 ground motion and is characterized by (1) a peak ground acceleration (PGA) value which anchors the response spectra at high frequencies (typically at 33 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-basis for LGS. The licensee stated that the design-basis of LGS was determined from an evaluation of the regional and local maximum earthquake potential.
The licensee determined that the maximum potential earthquake might be either an intensity VII event at the closest approach of the Fall Zone or an intensity VI event very near to the site. Accounting for uncertainties, the licensee specified a maximum potential earthquake as the equivalent of the intensity VI I 1871 Wilmington, Delaware event near the site. The design earthquake for LGS is a Newmark spectral shape anchored at 0.15g (15 percent of the acceleration due to earth's gravity).
The licensee specified that because the Updated Final Safety Analysis Report (UFSAR), (Exelon, 2012) does not explicitly define the SSE control point, it followed the guidance for a rock site in Section 2.4.2 of the SPID to define the control point at the top of the rock surface at elevation 204 ft. (62 m). Based on its review of the licensee's submittal and the UFSAR, the NRC staff confirms that the licensee's SSE spectrum and SSE control point elevation determination is consistent with information provided in the LGS UFSAR, as well as guidance 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 cutoff of M5.0, 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 mi (640 km) around the site and included the Charleston, Charlevoix, and Wabash Valley Repeated Large Magnitude Earthquake (RLME) sources, which lie within 620 mi (1,000 km) of LGS. 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 for each of the CE US-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 PSHA calculations for base rock site conditions at the LGS 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, and licensee's approach, the NRC staff included all CEUS-SSC background seismic sources within a 310 mi (500 km) radius of the LGS site. In addition, the NRC staff included the Charleston, Charlevoix, and Wabash Valley RLME sources, which lie within 620 km (1,000 mi) of the LGS site. For each of the CE US-SSC sources used in the PSHA, the NRC staff used the mid-continent version of the updated EPRI GMM. Based on its review of the SHSR, the NRC staff concludes that the licensee appropriately followed the guidance provided in the SPID for selecting the 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 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 reference rock conditions, as defined in the ground motion models used in the PSHA. To develop site-specific hazard curves at the control point elevation, Attachment 1 requests that licensees 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 would 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 these layers, and the degree to which the shear modulus and damping change with increasing input bedrock amplitude.
To develop site-specific hazard curves at the control point, the licensee performed a site response analysis.
3.3.1 Site Base Case Profiles In its SHSR, the licensee indicated that it performed a site response analysis for LGS. According to the licensee, the site rests on 10 ft. (3 m) of Cretaceous residual soils such as clay, silt, sand, and gravel underlain by almost 8,000 ft. (2,438 m) of sound Triassic sedimentary rocks including siltstone, sandstone, and shale. Precambian crystalline basement rocks are encountered at a depth of more than 8,000 ft. (2,438 m). The licensee provided site profile descriptions in Sections 2.3.1 and 2.3.2 of its SHSR based on information in the LGS UFSAR (Exelon, 2012) and the geotechnical information gathered for the Independent Spent Fuel Storage Installation (ISFSI) (SGH, 2012). The licensee noted that shear wave velocity measurements from the UFSAR were likely based on compressional-wave velocities obtained from refraction surveys and assumed Poisson ratios. More recent downhole testing for the ISFSI investigations provide a different and wider range in shear wave velocities.
Because it is more recent testing, the licensee used the ISFSI measurements, along with guidance provided in the SPID, to develop the best-estimate base case firm rock profile. The licensee assumed a shear wave velocity of 3,452 feet per second (fps, 1,052 meters per second (m/s)) in the shallow portion of the profile and assumed a gradient for sedimentary rock of 0.5 m/s/m. This resulted in a shear wave velocity of 7,400 fps (2,255 mis) at a depth of 8,000 ft. (2,438 m). For the best-estimate profile, the licensee developed the lower base case profiles using a scale factor of 1.57, which reflects a natural log standard deviation of 0.35 following guidance in the SPID. For the upper range base profile, the licensee assumed a shear wave velocity at the SSE control point of between 5,800 and 6, 100 fps (1,768 and 1,859 mis) with a gradient of 0.5 m/s/m resulting in a hard rock shear wave velocity of 9,285 fps (2,830 mis) at a depth of 6,734 ft. (2,052 m). The licensee stated that no site-specific dynamic material properties were determined during the initial investigations of the LGS site. Therefore, the licensee assumed that the rock material in the upper 500 ft. (150 m) could be modeled as either linear or non-linear using two sets of shear modulus reduction and hysteretic damping curves. Consistent with the SPID, the licensee determined that the EPRI rock curves (model M1) were appropriate to represent the upper range nonlinearity and linear analyses (model M2) were an equally plausible rock response across loading levels assuming low strain damping from the EPRI rock curves as the constant damping values in the upper 500 ft. (150 m). 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 LGS site, with 8,000 ft. (2,438 m) of firm sedimentary rock below the SSE, the licensee based the kappa estimates on the average shear wave velocity in the upper 100 ft. (30 m) for each base case profile. This resulted in kappa values for the best, lower and upper profiles of 0.023 sec, 0.036 sec, and 0.012 sec, respectively.
To account for aleatory variability in material properties across the plant site in its site response calculations, the licensee stated that it randomized its base case profiles in accordance with Appendix B of the SPID. The licensee stated that it also randomized the depth to correspond to 30 percent of the depth to bedrock. The licensee stated that this randomization did not represent actual uncertainty in the depth to reference rock, but was used to broaden the spectral peak. 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 the eleven input loading levels for the each base case profile. Consistent with the SPID, the licensee used a minimum median amplification value of 0.5 in the analysis.
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-6.0 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 reference rock hazard curves, determined from the initial PSHA (Section 3.2 of this assessment), and the amplification function and their associated uncertainties, determined from the site response analysis. 3.3.3 Staff Confirmatory Analysis The NRC staff used the ISFSI shear wave velocity data to characterize the dynamic properties in the upper 50 ft. (15 m). The NRC staff used the SPID template velocity profiles and a constant velocity gradient (as shown in Figure 3.3-1 of this assessment), to estimate the shear wave velocity at depths greater than 50 ft. (15 m). The NRC staff used a scale factor of 1.28, corresponding to a natural log standard deviation of 0.25, to estimate lower and upper base case velocity profiles.
The NRC staff velocity profiles differ moderately from those proposed by the licensee.
The NRC staff assumed the upper 500 ft. of rock material could demonstrate either linear or non-linear dynamic properties.
The NRC staff used EPRI Rock shear modulus degradation and damping curves to model non-linear behavior and a constant damping of 1 percent for the linear case. Kappa was also implemented to account for damping of seismic waves at the site. The NRC staff used the empirical relationship from the SPID for CEUS rock sites with at least 3000 ft. (1000 m) of firm sedimentary rock overlying hard rock to estimate the kappa for each base case velocity profile. Upper and lower kappa values were estimated using a scale factor of 1.49 corresponding to a log normal standard deviation of 0.4. The base case kappa values implemented in the NRC staff evaluations are 0.029, 0.037, 0.021, 0.027, 0.037, and 0.02 sec for NRC base cases P1, P2, P3, P4, P5, and P6, respectively.
Total site kappa was not allowed to exceed 0.04 sec. Figure 3.3-2 of this assessment shows a comparison of the NRC staff's and licensee's median site amplification functions and uncertainties for 2 of the 11 input loading levels. At low acceleration amplitudes, the amplification functions are slightly greater than the licensee amplification functions (Figure 3.3-2a). The difference between NRC and licensee amplification functions above 1 Hz reduces with larger input acceleration amplitudes.
The amplification function standard deviation computed by the NRC staff and licensee are similar. The difference between the licensee and NRC staff shear wave velocity profiles is the most significant factor leading to differences the amplification functions.
The differences in amplification function lead to minor differences in the hazard curves shown in Figure 3.3-3 of this assessment.
The shape of the control point hazard curves are similar. At 1 Hz, the licensee hazard curve is slightly steeper. 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, alternative approaches in performing site response analyses, including the modeling of uncertainty, are acceptable for this application.
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 LGS 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 analysis, as described in Sections 3.2 and 3.3 of this staff assessment, respectively.
Figure 3.4-1 of this assessment shows a comparison of the GMRS determined by the licensee to that determined by the NRC staff. As shown in Figure 3.4-1 below, the licensee's GMRS shape is generally similar to that calculated by the NRC staff. At frequencies where the amplitude of the GMRS determined by the licensee is different from that determined by the NRC staff, the staff's GMRS generally exceeds that of the licensee.
These differences in GMRS are the result of differences in the site response analyses performed by the licensee and NRC staff as discussed in Section 3.3 above. 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 LGS site. Therefore, this GMRS is suitable for use in subsequent evaluations and confirmations, as needed, for the licensee's 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 LGS site. Based on its review, the NRC staff concludes that the licensee conducted the seismic hazard reevaluation using present-day methodologies and regulatory guidance, appropriately characterized the site given the information available, and met the intent of the guidance for determining the reevaluated seismic hazard. Based on 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 Enclosure 1 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 confirmed the licensee's conclusion that the licensee's GMRS for the LGS is bounded by the SSE for the LGS site in the frequency range of 1 to 10 Hz and exceeds the SSE in the frequency range above approximately 10 Hz. As such, a seismic risk evaluation (Item 8) and SFP evaluation (Item 9) are not merited. However, a HF confirmation (i.e., Item (4)) is merited for the LGS site. The NRC review and acceptance of Exelon's HF confirmation for LGS will complete 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 Agencywide 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), 2011a, "Near-Term Report and Recommendations for Agency Actions Following the Events in Japan," Commission Paper SECY-11-0093, July 12, 2011, ADAMS Accession No. ML 11186A950.
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. ML 11186A950.
NRC (U.S. Nuclear Regulatory Commission), 2011c, "Recommended Actions to be Taken Without Delay from the Near-Term Task Force Report," Commission Paper SECY-11-0124, September 9, 2011, ADAMS Accession No. ML 11245A158.
NRC (U.S. Nuclear Regulatory Commission), 2011d, "Prioritization of Recommended Actions to be Taken in Response to Fukushima Lessons Learned," Commission Paper SECY-11-0137, October 3, 2011, ADAMS Accession No. ML 11272A111.
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. ML 12053A340.
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 1 O CFR 50.54(f) Schedule Modifications Related to the NTTF Recommendation 2.1, Seismic Reevaluations, May 7, 2013, ADAMS Accession No. ML 13106A331.
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. ML 12319A074.
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. ML 13233A 102. 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. ML 14111A147.
NRC (U.S. Nuclear Regulatory Commission), 2014b. Letter from D. Dorman, Acting Director Office of Nuclear Reactor Regulation to Operating Power Reactor Licensees on the Enclosed List, Screening and Prioritization Results Regarding Seismic Hazard Reevaluations for Recommendation 2.1 of the Near-Term Task. Force Review of Insights from the Fukushima Dai-ichi Accident, October 3, 2014, ADAMS Accession No. ML 14258A043.
Other References Barstow S. 2014, Letter from S. Barstow (Exelon Generation Company) to NRC, Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident, March 31, 2014, ADAMS Accession No. ML 14090A236.
Exelon Generation Company, 2012, Limerick Generating Station, Units 1 and 2, Updated Final Safety Analysis Report (UFSAR), Revision 16, September 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 Log normal 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. ML 12333A 170. Electric Power Research Institute (EPRI), 2013. EPRI Report 3002000717 "EPRI (2004, 2006) Ground-Motion Model Review Project Final Report," Palo Alto, CA, 2013. Kaegi, G., 2013, Letter from G. Kaegi (Exelon Generation Company) 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. ML 13256A070.
Keith line, 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. ML 12333A168.
SGH Report No. 128018-R-01, Revision 1, Review of Existing Site Response Data for the Exelon Nuclear Fleet, July 17, 2012. Keithline, 2013, Submittal of EPRI (2004, 2006) Ground Motion Model Review Final Report, June 3, 2013, ADAMS Accession No. ML 13170A378.
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. ML 13101A379. Figure 3.3-1 Plot of NRC's and Licensee's Base Case Shear-Wave Velocity Profiles for the LGS Site Shea r Wave Velocity (ft/s) Shear Wave Velocity (ft/s) 0 2000 4000 6000 8000 10000 12000 0 2000 4000 6000 8000 10000 12000 0 0 1000 1000 -;::: ;::: :::-2000 :::-2000 c: c: *a *a 3000 3000 ... ... --c: c: 8 4000 8 4000 5000 5000 Q. 6000 Q. 6000 Q) Q) 0 0 7000 7000 8000 -Licensee P1 ---N R C P1 -Licensee P1 ---N R C P4 -Licensee P2 ----N R C P2 -Licensee P2 ----N R C P5 -Licensee P3 N R CP3 -Licensee P3 N R CP6 Figure 3.3-1 Plot Comparing the NRC's and the Licensee's Median Amplification Functions and Uncertainties for the LGS site. 3 a) -Licensee P1 Amp ----NRC P1 Amp 2.5 ----NRC P4Amp -Licensee P1 Sigma Ln(Amp) c: 2 0 ----NRC P1 Sigma Ln(Amp) NRC P4 S i gma Ln (Amp) +: ro (..) 1.5 c. E <( 1 0.5 0 -----0.1 Frequency (Hz) 1 0 100 3 b) 2.5 ,, 2 -----......
,,-----v ', c: ------..... : ,.,,,.---," .. , ', 0 .. , \ ._, +: ... ' \ ' ro ,, ', (..) 1.5 ii= ' \ ...... ' ' c. ' \ E ' \ <( 1 0.5 --0 0.1 1 Frequency (Hz) 10 100 3 1 C) 0 ...I ro =s 0.5 rt; z 0 3 c: 2.5 0 +: ro > (],) 2 0 "'C '-ro "'C 1.5 c: ro -(/) C) 1 0 ...I ro '-=s 0.5 -ro z 0 Figure 3.3-2 Plot Comparing the NRC's and the Licensee's Mean Control Point Hazard Curves at a Variety of Frequencies for the LGS site (],) (J c C'O "O (],) (],) (J >< w .... 0 >. (J c (],) :::J C"' (],) "-u.. C'O :::J c c <( c C'O (],) 2 1.E-02 1.E-03 1.E-04 1.E-05 1.E-06 1.E-07 0.01 -Licnesee 1 Hz -Licensee 10 Hz -L i censee PGA --*NRC 1 Hz -NRC 10 Hz --*NRC PGA t 0.1 1 10 Spectral Acceleration (g) Figure 3.4-1 Comparison of the NRC's GMRS with Licensee's GMRS and the SSE for the LGS site -C> -s:: 1 0 ;:; <'Cl I.. Cl) Cl) 0 0 <( <'Cl I.. -0.5 0.. Cl) 0.1 -SSE -Licensee GMRS -NRCGMRS t 1 10 100 Frequency (Hz)
B. Hanson If you have any questions, please contact me at (301) 415-1617 or at Frankie.Vega@nrc.gov.
Docket Nos. 50-352 and 50-353
Enclosure:
Staff Assessment of Seismic Sincerely, IRA/ Frankie Vega, Project Manager Hazards Management Branch Japan Lessons-Learned Division Office of Nuclear Reactor Regulation Hazard Evaluation and Screening Report cc w/encl: Distribution via Listserv DISTRIBUTION:
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