ML16049A609
| ML16049A609 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 02/29/2016 |
| From: | Frankie Vega Japan Lessons-Learned Division |
| To: | Braun R Public Service Enterprise Group |
| Vega, Frankie NRR/JLD 415-1617 | |
| References | |
| CAC MF3924 | |
| Download: ML16049A609 (22) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 February 29, 2016 Mr. Robert Braun President and Chief Nuclear Officer PSEG Nuclear LLC - N09 P. 0. Box 236, Hancocks Bridge, NJ 08038
SUBJECT:
HOPE CREEK GENERATING STATION - 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 AND STAFF CLOSURE OF ACTIVITES ASSOCIATED WITH RECOMMENDATION 2.1, "SEISMIC" (CAC NO. MF3924)
Dear Mr. Braun:
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 28, 2014, PSEG Nuclear, LLC (PSEG, the licensee), responded to this request for Hope Creek Generating Station (Hope Creek).
The NRC staff has reviewed the information provided related to the reevaluated seismic hazard for Hope Creek and, as documented in the enclosed staff assessment, determined that you provided sufficient information in response to Enclosure 1, Items (1) - (9) of the 50.54(f) letter.
In indicated in NRC letter dated October 27, 2015 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML15194A015), PSEG was requested to submit a High Frequency (HF) confirmation (Item 4). By letter dated December 23, 2015 (ADAMS Accession No. ML15358A138), PSEG submitted its high frequency confirmation. The NRC staff has reviewed this confirmation and has documented its review by NRC letter dated February 18, 2016 (ADAMS Accession No. ML15364A544).
The NRC staff concludes that the licensee responded appropriately and has completed its response to Enclosure 1 of the 50.54(f) letter.
R. Braun Furthermore, the NRC staff concludes that no further responses or regulatory actions associated with Phase 2 of Near-Term Task Force (NTTF) Recommendation 2.1 "Seismic" are needed for Hope Creek. This closes out the NRC's efforts associated with Phase 1 and 2 of NTTF Recommendation 2.1 "Seismic" (CAC No. MF3924) for Hope Creek.
If you have any questions, please contact me at (301) 415-1617 or at Frankie.Vega@nrc.gov.
Sincerely, Frat: Vega, ;,oject Manager Hazards Management Branch Japan Lessons-Learned Division Office of Nuclear Reactor Regulation Docket No. 50-354
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 HOPE CREEK GENERATING STATION DOCKET NO. 50-354
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, 2011b). 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).
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 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 10, 2013
(Davison, 2013), PSEG Nuclear LLC (PSEG, the licensee) submitted at least partial site response information for Hope Creek Generating Station (Hope Creek). By letter dated March 28, 2014 (Davison, 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 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 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) GMMs. 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
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 28, 2014 (Davison, 2014), the licensee provided the Hope Creek SHSR.
The licensee's SHSR indicates that the site GMRS over the frequency range of 1 to 1O Hertz (Hz) is bounded by the site SSE. Above 1O Hz, the Hope Creek GMRS is bounded by either the site SSE or the Individual Plant Examination of External Events (IPEEE) plant-level high confidence of low probability of failure (HCLPF) spectrum (IHS). Following the guidance in Section 3.3 of the SPID, the licensee provided an evaluation of its IPEEE program in order to use the IHS as the plant seismic capacity for the screening comparison with the GMRS. For frequencies above 10 Hz, the IHS also bounds the licensee's site GMRS. However, the full scope IPEEE detailed review of relay chatter required in SPID Section 3.3.1 has not been completed by the licensee. Therefore, the licensee stated that it will complete the relay chatter review consistent with NEI letter to NRC dated October 3, 2013 (Keithline, 2013) on the same schedule as the HF confirmation as proposed in the NEI letter dated April 9, 2013 (Pietrangelo, 2013), and accepted in NRC's letter dated May 7, 2013 (NRC, 2013a). The licensee stated that the HF confirmation per the SPID Section 3.4 will be performed only if the relay chatter review is not successful in demonstrating relay adequacy based on the GMRS.
On May 9, 2014 (NRC, 2014a), the NRC staff issued a letter providing the outcome of its 30- day, preliminary, screening and prioritization evaluation. In the letter, the NRC staff characterized the Hope Creek site as screened out. The NRC staff confirmed that the licensee's GMRS, as well as the staff's confirmatory GMRS is bounded by the SSE for Hope Creek over the frequency range of 1 to 30 Hz, and therefore, a plant seismic risk evaluation and SFP evaluation are not merited. Additionally, in the frequency range above 10 Hz, the GMRS exceeds the SSE and is bounded by the IHS. As stated in the October 27, 2015 (NRC, 2015),
letter, the NRC staff requested PSEG to submit a HF confirmation. By letter dated December 23, 2015 (Davison, 2015), PSEG submitted its HF confirmation. The NRC staff has reviewed this confirmation and has documented its review in letter dated February 19, 2016 (NRC, 2016).
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 re-evaluated 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 describes its seismic design bases for the Hope Creek site and states that the SSE is defined in terms of a PGA and a design response spectrum and was developed in accordance with 10 CFR Part 100, Appendix A through an evaluation of the maximum earthquake potential for the region surrounding the site. The licensee states that the SSE for Hope Creek is based on a postulated maximum potential earthquake either from the intensity VII (Modified Mercalli Scale) 1871 Wilmington, Delaware earthquake occurring near the site or the intensity VI I northern New Jersey earthquake of 1927 occurring near the site. Based on these earthquakes, the licensee anchored a Regulatory Guide 1.60 (RG 1.60) (NRC, 1973) response spectral shape was 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 specifies that the SSE control point is defined at the foundation level at a depth of 62 ft (19 m) below grade.
The NRC staff reviewed the licensee's description of its SSE in the SHSR for the Hope Creek site. With regard to the Hope Creek site SSE, based on its review of the SHSR and Updated Final Safety Analysis Report (UFSAR)(Hope Creek, 2012), the NRC staff confirmed that the licensee's SSE is defined in terms of a PGA and a RG 1.60 design response spectrum anchored at 0.2 g, as described by the licensee. Finally, based on its review of the SHSR and the UFSAR (Hope Creek, 2012), the NRC staff confirmed that the licensee's control point elevation for Hope Creek site SSE is defined at a depth of 62 ft (19 m) below grade and is consistent with the guidance provided in the SPID.
3.2 Probabilistic Seismic Hazard Analysis In Section 2.2 of its SHSR, the licensee states 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, 2004, 2006, 2013). The licensee used a minimum magnitude (M) of 5.0, as specified in the 50.54(f) letter. The licensee further states that it included the CEUS-SSC background sources out to a distance of 400 miles (640 km) around the site and included the Charleston, Charlevoix, and Wabash Valley repeated large magnitude earthquake (ALME) sources, which lie within 621 mi (1,000 km) of the site. The ALME 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 in SHSR Section 2.2.2 since a site response analysis is necessary to determine the control point seismic hazard curves. The licensee provides its control point
seismic hazard curves in Section 2.3.7 of its SHSR. The NRC 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 the Hope Creek site. As input, the NRC staff used the CEUS-SSC model in NUREG-2115 (NRC, 2012b) along with the Updated EPRI GMM model (EPRI 2013).
Consistent with the guidance provided in the SPID, the NRC staff included all CEUS-SSC background seismic sources within a 31 O mi {500 km) radius of the Hope Creek 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, 2004, 2006, 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 CEUS-SSC model and the updated EPRI GMM.
3.3 Site Response Evaluation After completing PSHA calculations for reference rock site conditions, Attachment 1 to of the 50.54(f) letter requests that licensees 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 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 provides detailed site profile descriptions in Sections 2.3.1 and 2.3.2 of its SHSR based on information provided in the PSEG Early Site Permit Application (ESPA) (PSEG, 2013)
and the guidance in Appendix B of the SPID. According to the licensee, the Hope Creek site is located in the eastern U.S. on the Delaware River within the Coastal Plain physiographic province. The Coastal Plain is underlain by a thick wedge of unconsolidated sediment ranging from Cretaceous to recent in age and the bedrock is estimated to be at a depth of about 1,800 ft (548 m).
In Table 2-1 of its SHSR, the licensee provides a brief description of the subsurface materials in terms of the geologic units and layer thicknesses. In Table 2-1 of its SHSR, the licensee provides the shear-wave velocities determined from the results detailed in the PSEG ESPA (PSEG, 2013). The PSEG ESPA site is the location of potential construction at the Hope Creek site. The PSEG ESPA site investigations included 4 compression (P) and shear (S) wave P-S suspension logging surveys ranging to a depth of approximately 300 to 630 ft (91 to 192 m), 2 cross-hole velocity testing boreholes extending to a depth of approximately 200 ft (61 m), one down-hole seismic velocity measurement to a depth of approximately 200 ft (61 m), and one deep production well extending to the top of basement (at approximately 1,800 ft (548 m))
located beneath the Hope Creek site.
In Section 2.3.2.1 of its SHSR, the licensee describes its three base case shear-wave velocity profiles for the Hope Creek site. The licensee used a natural log standard deviation of 0.20 to calculate the lower and upper base case shear-wave velocity profiles to a depth of approximately 516 ft (157 m) and a natural log standard deviation of 0.35 to calculate the lower and upper base case shear-wave velocity profiles from 516 ft (157 m) depth to bedrock at 1,800 ft (548 m). Table 2-2 and Figure 2-1 of its 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.
In Section 2.3.2.2 of its SHSR, the licensee states that no site-specific dynamic material properties were determined in the initial siting of Hope Creek site. Therefore, the licensee followed the SPID guidance for soil sites and selected two alternative characterizations of nonlinear dynamic material behavior. In one characterization, the licensee used the EPRI soil shear modulus reduction and hysteretic damping curves over the upper 516 ft (157 m) and modeled linear behavior with no damping below 516 ft (157 m). In the second characterization, the licensee used the Peninsular Range shear modulus reduction and hysteretic damping curves over the upper 516 ft (157 m) and linear behavior with no damping below 516 ft (157 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 Hope Creek site, the licensee provided estimates of kappa in Table 2-3 of the SHSR, calculated using the empirical relation of Campbell (2009), as described in the SPID Appendix B, for shallow soil sites having a depth to bedrock of less than 3,000 ft (1,000 m). The licensee also added an additional kappa of 0.006 sec to account for the damping in the underlying base rock material. The licensee's total profile kappa values for the best estimate, upper, and lower base case velocity profiles are all equal to 0.038 sec.
To account for randomness in material properties across the plant site, the licensee states in Section 2.3.3 of its SHSR that it randomized its base case shear-wave velocity profiles following Appendix B of the SPID. In addition, as stated in Section 2.3.2.1 of its SHSR, the licensee randomized the depth to bedrock by +/-51 Oft (+/-155 m), which corresponds to 30 percent of the total profile thickness. The licensee states 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 states 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 of its SHSR, the licensee describes 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 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 NRG staff performed its own site response calculations for the Hope Creek site. The NRG staff reviewed the licensee's site response analysis and performed confirmatory calculations to independently confirm the licensee's calculation following the guidance in Appendix B of the SPID.
Because the site is well characterized and the velocities are reported in the PSEG ESPA (PSEG, 2013), the NRC staff used the same three base case shear-wave velocity profiles that were used by the licensee for its site response analysis. However, to capture the uncertainty in the depth to base rock, the NRG staff used a value of +/-106 ft [+/-32 m], which is less than the
+/-51 Oft (+/-155 m) depth to base rock randomization used by the licensee. Figure 3.3.-1 of this assessment illustrates the velocity profiles the NRG staff used in its confirmatory analysis.
In the absence of any site-specific dynamic material property measurements, the NRG staff followed the SPID guidance for soil sites and selected two alternative characterizations of nonlinear dynamic material behavior. In one characterization, the NRG staff used the EPRI soil shear modulus reduction and hysteretic damping curves over the upper 516 ft (157 m) and modeled linear behavior with no damping below 516 ft (157 m). In the second characterization, the NRG staff used the Peninsular Range shear modulus reduction and hysteretic damping curves over the upper 516 ft (157 m) and linear behavior with no damping below 516 ft (157 m).
The NRG staff assigned equal weights to the two characterizations.
To determine kappa for its velocity profiles, the NRG staff used the empirical relation of Campbell (2009), as described in the SPID Appendix B, for shallow soil sites having a depth to bedrock of less than 3,000 ft (1,000 m) and also added an additional kappa of 0.006 sec to account for the damping in the underlying base rock material. The NRG staff's total profile kappa values for the best estimate, upper, and lower base case velocity profiles are all equal to 0.038 sec.
Figure 3.3-2 of this assessment shows a comparison of the NRG staff's and licensee's median site amplification factors 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.
The NRG staff conducted sensitivity tests to evaluate the licensee's characterization of the site and its use of (1) multiple base case velocity profiles, (2) kappa values, and (3) depth to bedrock randomization (BR) value. Of these three factors, the NRG staff used a smaller BR value (5 percent), rather than the larger value used by the licensee (30 percent). The NRG staff used a BR value of 5 percent, because the deep production well located at the Hope Creek site identified the basement rock at approximately 1,800 ft (548 m). This data indicates that the location of the basement rock is relatively well understood beneath the Hope Creek site and, thus, the NRG staff tested calculating the Hope Creek site hazard using a smaller BR value than that used by the licensee. However, as seen in Figure 3.3-2 of this assessment, both the NRG staff and licensee's site amplification factors are similar, such that the difference in BR value does not produce large difference in site amplification factors at the Hope Creek site.
Therefore, based on these sensitivity analyses, the NRG staff concludes that the licensee's evaluation for Hope Creek site adequately captures the site amplification occurring as a result of bedrock ground motions travelling upward through the soil/rock column to the control point elevation.
Overall, the licensee's approach to modeling the subsurface rock properties and uncertainties results in similar site amplification factors to the NRG staff's site amplification factors.
Additionally, as shown in Figure 3.3-3 of this assessment, the NRG staff notes that the slight differences in the site response analysis 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 NRG staff concludes that the licensee's site response was conducted using present-day guidance and methodology, including the NRG-endorsed SPID. The NRG staff performed independent calculations to confirm 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 Hope Creek site.
3.4 Ground Motion Response Spectra In Section 2.4 of its SHSR, the licensee states that it used the control point hazard curves, described in SHSR Section 2.3.7, to develop the 10-4 and 10-5 uniform hazard response spectra (UHAS) and then computed the GMRS using the criteria in RG 1.208.
The NRC staff independently calculated the 10-4 and 10-5 UHAS 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. 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 of this assessment, the licensee's GMRS shape is generally similar to that calculated by the NRC staff at frequencies less than 1O Hz. However, NRC staff's confirmatory GMRS is somewhat higher than the licensees at frequencies above 1O Hz. As described above in Section 3.3, the NRC staff concludes that these minor differences over the higher frequency range are primarily due to the differences in the site response analyses performed by the licensee and the NRC staff. The NRC staff concludes that these differences are acceptable for this application because the licensee followed the guidance provided in the SPID with respect to both the PSHA and site response analysis for the Hope Creek site.
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 Hope Creek 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 Hope Creek 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 of Item (4) identified in of the 50.54(f) letter.
In reaching this determination, the NRC staff confirmed the licensee's conclusion that the licensee's GMRS is bounded by the SSE in the frequency ranges of 1 to 30 and 40 to 100 Hz.
As such, a seismic risk evaluation (Item 8) and a SFP evaluation (Item 9) are not merited.
Additionally, the NRC staff confirms that the licensee's GMRS exceeds the SSE in the frequency range of 30 to 40 Hz. As stated in the October 27, 2015, letter, the NRC staff requested that a HF confirmation (Item 4) be performed.
By letter dated December 23, 2015 (Davison, 2015) PSEG submitted its high frequency confirmation. The NRC staff has reviewed this confirmation and has documented its review in NRC letter dated February 19, 2016 (NRC, 2016) . Based upon the preceding analysis, the NRC staff concludes that the licensee responded appropriately to Enclosure 1 of the 50.54(f) letter.
REFERENCES Note: ADAMS Accession Nos. refers to documents available through NRC's Agencywide Documents 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 Japan," 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 21st 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), 2011 c, "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), 2011d, "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 1O 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.
NRC (U.S. Nuclear Regulatory Commission) 2014b. Letter from D. H. Dorman (NRC) to Select Operating Power Reactor Licensees, 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. ML14258A043.
NRC (U.S. Nuclear Regulatory Commission), 2015. Letter from W. Dean (NRC), Director, Office of Nuclear Reactor Regulation to All Power Reactor Licensees and holders of Construction Permits in Active or Deferred Status, Final Determination of Licensee Seismic Probabilistic Risk Assessments Under Request for Information Pursuant to Title 1O of the Code of Federal Regulations 50.54 (f) Regarding Recommendation 2.1 "Seismic" of the Near Term Task Force Review of Insights from Fukushima Dai-lchi Accident, October 27, 2015, ADAMS Accession No. ML15194A015.
NRC (U.S. Nuclear Regulatory Commission), 2016. Letter from S. Wyman (NRC), Project Manager, Office of Nuclear Reactor Regulation, "Staff Review of High Frequency Confirmation Associated with Reevaluated Seismic Hazard in Response to March 12, 2012 50.54 (f) Request for Information", February 18, 2016, ADAMS Accession No. ML15364A544.
Other References Campbell, K. W. (2009). "Estimates of shear-wave Q and KO for unconsolidated and semiconsolidated sediments in Eastern North America." Bull. Seism. Soc. Am, 99(4),
2365-2392.
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 Ground Motion Model Review Final Report, June 3, 2013, ADAMS Accession No. ML13155A553.
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.
Keithline, 2013, Letter from K. Keithline, Senior Project Manager, NEI, to U.S. Nuclear Regulatory Commission, "Relay Chatter Reviews for Seismic Hazard Screening,"
October 3, 2013, ADAMS Accession No. ML13281A308.
Davison, 2013, Letter from P.J. Davison, Site Vice President, PSEG Nuclear to U.S. Nuclear Regulatory Commission, "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 - Base Case Velocity Profiles With Supporting Subsurface Materials and Properties," September 10, 2013, ADAMS Accession No. ML13254A166.
Davison, 2014, Letter from P.J. Davison, Site Vice President, PSEG Nuclear to U.S. Nuclear Regulatory Commission, "PSEG Nuclear LLC's Seismic Hazard and Screening Report (CEUS Sites) 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 - Hope Creek Generating Station," March 28, 2014, ADAMS Accession No. ML14087A436.
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.
PSEG Nuclear Early Site Permit Application (ESPA), Site Safety Analysis Report (SSAR),
Revision 2, April 8, 2013, ADAMS Accession No. ML13098A775.
Hope Creek, 2012, Updated Final Safety Analysis Report (UFSAR), Revision 19, November 5, 2012, ADAMS Accession No. ML12314A402, (Non-Public).
Davison, 2015, Letter from P.J. Davison, Site Vice President, PSEG Nuclear to U.S. Nuclear Regulatory Commission, "High Frequency Supplement to Seismic Hazard Screening
Report, 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". December 23, 2015, ADAMS Accession No. ML15358A138.
Figure 3.3-1 Plot of the Staff's and Licensee's Base Case Shear-Wave Velocity Profiles for the Hope Creek site Shear Wave Velocity (ft/s) 0 1000 2000 3000 4000 5000 6000 0
I I I 8 I I I N
I I 8'<t __J g
c:
- 5 8
Q.
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0 00
~
u 0
3:
..2 8 rl QI a:i
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- c. 0 0
QI N c
rl 0
0 rl
- HCGSBa se Case 0
0 - HCGSLo wer Case rl
- HCGSUp per Case 0
~
rl
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 Hope Creek site 3.5
- NRC PGA=0.05
- NRC PGA=0 .3
- HCGS PGA=0 .05 3 - HCGS PGA=0.292
- NRC PGA=0 .05 Sigma ln(AF)
- NRC PGA=0 .3 Sigma ln (AF) 2.5 - HCGS PGA=0 .05 Sigma In (AF)
.... - HCGS PGA=0.292 Sigma ln(AF) 0 u
LL"'c:
,g 2
~""'
0.
<(
E 1.5 0.5 0 .- -
0 .1 Frequency (Hz) 10 100
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 Hope Creek site lE-01
- HCGSPGA
- HCGSlOHz
- HCGS lHz lE-02
- NRCPGA
., - NRC 10 Hz u
c:
NRC 1 Hz u
lE-03 x
w
'15
.,~c:
- J lE-04 O'
u.
~
- J c:
c:
ct c:
lE-05
- ?
lE-06 lE-07 0.001 0.01 0.1 1 10 Spectral Acceleration (g)
Figure 3.4-1 Comparison of the Staff's GMRS, Licensee's GMRS, Hope Creek site SSE, and the Hope Creek site IHS for the Hope Creek site 1.4 1.2 0.8
~
<(
Ill 0.6 0.4 0 .2 0
0 .1 Frequency (Hz) lo 100
- HCGSSSE - HCGS GMRS 30% BR NRC GMRS 5% BR - HCGSIHS
ML16049A609 *via email OFFICE NRR/JLD/JHMB/PM NRR/JLD/LA NRO/DSEA/RGS1/BC*
NAME FVega SLent DJackson DATE 02/23/2016 02/22/2016 12/21/2015 OFFICE NRRJLD/JHMB/BC NRR/JLD/JHMB/PM NAME MShams FVega DATE 02/28/2016 02/29/2016