ML18093A445

Ad 2, Response to 03/12/2012 Information Request Seismic Probabilistic Risk Assessment for Recommendation 2.1

ML18093A445
Person / Time
Site:	North Anna
Issue date:	03/28/2018
From:	Stoddard D Virginia Electric & Power Co (VEPCO)
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
18-086
Download: ML18093A445 (185)

Text

VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261 March 28, 2018 U.S. Nuclear Regulatory Commission Serial No.18-086 Attention: Document Control Desk NRA/DEA RO Washington, DC 20555 Docket Nos.: 50-338/339 License Nos.: NPF-4/7 VIRGINIA ELECTRIC AND POWER COMPANY NORTH ANNA POWER STATION UNITS 1 AND 2 RESPONSE TO MARCH 12, 2012 INFORMATION REQUEST SEISMIC PROBABILISTIC RISK ASSESSMENT FOR RECOMMENDATION 2.1 References.:-

1. NRG Letter, "Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(f) Regarding Recommendations 2.1, 2.3, and 9.3, of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident," dated March 12, 2012 [ADAMS Accession Nos. ML12056A046 and ML12053A340].

2. 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." [ADAMS Accession No. ML12333A170].

3. Virginia Electric and Power Company Letter to NRG, "North Anna Power Station Units 1 and 2 Response to March 12, 2012 Information Request - Seismic Hazard and Screening Report (CEUS Sites) for Recommendation 2.1," dated March 31, 2014 [ADAMS Accession No. ML14092A416].

4. NRG Letter, "North Anna Power 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 Relating to Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dal-lchi Accident (TAC Nos. MF3797 and MF3798)," dated April 20, 2015 [ADAMS Accession No. ML15057A249].

5. NRG Letter, ""Final Determination of Licensee Seismic Probabilistic Risk Assessments Under the Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(f) Regarding Recommendation 2.1 "Seismic" of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident,"

dated October 27, 2015 [ADAMS Accession No. ML15194A015].

On March 12, 2012, the Nuclear Regulatory Commission (NRG) issued a request for information pursuant to 10 CFR 50.54(f) associated with the recommendations of the Fukushima Near-Term Task Force (NTTF) (Reference 1). Enclosure 1 of Reference 1 requested each licensee to reevaluate the seismic hazards at their sites using present-day NRG requirements and guidance, and to identify actions taken or planned to address plant-specific vulnerabilities associated with the updated seismic hazards.

Serial No.18-086 Docket Nos. 50-338/339 Page 2 of 3 Reference 2 contains industry guidance developed by EPRI that provides the screening, prioritization and implementation details for the resolution of Fukushima Near-Term Task Force Recommendation 2.1: Seismic. The SPID (Reference 2) was used to compare the reevaluated seismic hazard to the design basis hazard. The North Anna Power Station (NAPS), Units 1 and 2 Seismic Hazard and Screening Report (Reference

3) concluded that the ground motion response spectrum (GMRS) exceeded the design basis seismic response spectrum in the 1 to 1O Hz range, and therefore a seismic probabilistic risk assessment was required.

Reference 4 contains the NRC Staff Assessment of the NAPS Units 1 and 2 seismic hazard submittal and concluded that the reevaluated seismic hazard prepared for NAPS is suitable for other activities associated with the NRC Near-Term Task Force Recommendation 2.1: Seismic.

Reference 5 contains the NRC letter "Final Determination of Licensee Seismic Probabilistic Risk Assessments." In that letter (Table 1 a - Recommendation 2.1 Seismic

- Information Requests), the NRC instructed that a Seismic Probabilistic Risk Assessment (SPRA) be submitted for NAPS Units 1 and 2 by March 31, 2018.

The Attachment to this letter contains the NAPS Units 1 and 2 SPRA Summary Report, which provides the information requested in Enclosure 1, Item (8)8 of Reference 1.

If you have any questions regarding this information, please contact Diane E. Aitken at (804) 273-2694.

Sincerely, Daniel G. Stoddard DIANE E. AITKEN Senior Vice President and Chief Nuclear Officer NOTARY PUBLIC REG. #7763114 Virginia Electric and Power Company COMMONWEALTH OF VIRGINIA MY COMMISSION EXPIRES MARCH 31, 2022 COMMONWEALTH OF VIRGINIA COUNTY OF HENRICO The foregoing document was acknowledged before me, in and for the County and Commonwealth aforesaid, today by Daniel G. Stoddard, who is Senior Vice President and Chief Nuclear Officer of Virginia Electric and Power Company. He has affirmed before me that he is duly authorized to execute and file the foregoing document in behalf of that company, and that the statements in the document are true to the best of his knowledge and belief.

Acknowledged before me this 2. 'i day of }1) ofGh , 2018.

My Commission Expires:

Notary Public

Serial No.18-086 Docket Nos. 50-338/339 Page 3 of 3 Commitments made in this letter: No new regulatory commitments.

Attachment:

North Anna Power Station Units 1 and 2 Seismic Probabilistic Risk Assessment in Response to 10 CFR 50.54(f) Letter with Regard to NTTF 2.1 Seismic -

Summary Report cc: U.S. Nuclear Regulatory Commission Region II Marquis One Tower 245 Peachtree Center Ave., NE Suite 1200 Atlanta, Georgia 30303-1257 Mr. J. E. Reasor, Jr.

Old Dominion Electric Cooperative Innsbrook Corporate Center 4201 Dominion Blvd.

Suite 300 Glen Allen, Virginia 23060 Mr. James R. Hall NRC Senior Project Manager-North Anna U. S. Nuclear Regulatory Commission One White Flint North Mail Stop 0-8 G9A 11555 Rockville Pike Rockville, Maryland 20852-2738 Ms. K. R. Cotton-Gross Project Manager - Surry U.S. Nuclear Regulatory Commission One White Flint North Mail Stop 08 G-9A 11555 Rockville Pike Rockville, MD 20852-2738 NRC Senior Resident Inspector North Anna Power Station 50.54f_ Seismic.Resource@nrc.gov

Serial No.18-086 Docket No. 50-338/339 ATTACHMENT North Anna Power Station Units 1 and 2 Seismic Probabilistic Risk Assessment in Response to 10 CFR 50.54(f) Letter with Regard to NTTF 2.1 Seismic Summary Report VIRGINIA ELECTRIC AND POWER COMPANY NORTH ANNA POWER STATION UNITS 1 AND 2

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NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Summary Report March 2018 NORTH ANNA POWER STATION UNITS 1 AND 2 SEISMIC PROBABILISTIC RISK ASSESSMENT IN RESPONSE TO 10 CFR 50.54(1) LETTER WITH REGARD TO NTTF 2.1 SEISMIC

SUMMARY

REPORT MARCH2018 Page 1 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Summary Report March 2018 North Anna Power Station Units 1 and 2 Seismic Probabilistic Risk Assessment Summary Report Contents Executive Summary 1.0 Purpose and Objective 2.0 Information Provided in This Report 3.0 NAPS Seismic Hazard and Plant Response 4.0 Determination of Seismic Fragilities for the SPRA 5.0 Plant Seismic Logic Model 6.0 Conclusions 7.0 References 8.0 Acronyms Appendix A SPRA Technical Adequacy Assessment and Peer Review Page 2 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Summary Report March 2018 EXECUTIVE

SUMMARY

In response to the NRC 10 CFR 50.54(f) letter of March 12, 2012, a seismic probabilistic risk assessment (SPRA) was performed for North Anna Power Station (NAPS) Units 1 and 2. The SPRA effort included performing a probabilistic seismic hazard analysis (PSHA) to develop seismic hazard and response spectra at the plant using the state-of-the-art seismic source model and attenuation equations; site response analyses; dynamic analyses of structures; fragility analyses of structures, systems and components (SSCs); developing a logic model; and performing risk quantification. Each element of the SPRA effort underwent an in-process independent expert review and a final peer review by a team of experts. The comments and 1 suggestions of the reviewers were addressed and incorporated into the SPRA as applicable.

The SPRA identified risk-significant sequences and SSCs with their risk rankings, and showed that for both North Anna units, the seismic Core Damage Frequency (SCDF) is 6.0E-5 per year and the seismic Large Early Release Frequency (SLERF) is l.6E-5 per year.

Sensitivity studies were performed to identify critical assumptions, test the sensitivity to quantification parameters and the seismic hazard, and identify potential areas to consider for the reduction of seismic risk. These sensitivity studies demonstrated that the model results are robust with respect to the modeling and assumptions used.

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NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Summary Report March 2018 1.0 Purpose and Objective Following the accident at the Fukushima Dai-ichi nuclear power plant resulting from the March 11, 2011, Great Tohoku Earthquake and subsequent tsunami, the Nuclear Regulatory Commission (NRC) established a Near Term Task Force (NTTF) to conduct a systematic review of NRC processes and regulations and to determine if the agency should make additional improvements to its regulatory system. The NTTF developed a set of recommendations intended to clarify and strengthen the regulatory framework for protection against natural phenomena. Subsequently, the NRC issued a 10 CFR 50.54(f) letter on March 12, 2012 [1], requesting information to assure that these recommendations are addressed by all U.S. nuclear power plants. The 50.54(f) letter requests that licensees and holders of construction permits under 10 CFR Part 50 reevaluate the seismic hazards at their sites against present-day NRC requirements and guidance.

A comparison between the reevaluated seismic hazard and the design basis for NAPS has been performed, in accordance with the guidance in EPRI 1025287, "Screening, Prioritization and Implementation Details (SPID) for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1: Seismic" [2], and was previously submitted to NRC [3]. That comparison concluded that the ground motion response spectrum (GMRS), which was developed based on the reevaluated seismic hazard, exceeds the design basis seismic response spectrum in the 1 to 10 Hz range, and a seismic risk assessment is required. A seismic PRA (SPRA) has been developed to perform the seismic risk assessment for North Anna Power Station Units 1 and 2 (NAPS) in response to the 50.54(f) letter, specifically item (8) in Enclosure 1 of the 50.54(f) letter.

This report describes the seismic PRA developed for NAPS and provides the information requested in item (8)8 of Enclosure 1 of the 50.54(f) letter and in Section 6.8 of the SPID

[2]. The SPRA model has been peer reviewed (as described in Appendix A) and found to be of appropriate scope and technical capability for use in assessing the seismic risk for NAPS, identifying which structures, systems, and components (SSCs) are important to seismic risk, and describing plant-specific seismic issues and associated actions planned or taken in response to the 50.54(f) letter.

This report provides summary information regarding the SPRA as outlined in Section 2.

The level of detail provided in the report is intended to enable NRC to understand the inputs and methods used, the evaluations performed, and the decisions made as a result of the insights gained from the NAPS seismic PRA.

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NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Su_mmary Report March 2018 2.0 Information Provided in This Report The following information is requested in the 50.54(f) letter [1], Enclosure 1, "Requested Information" Section, paragraph (8)8, for plants performing a SPRA.

(1) The list of the significant contributors to seismic core damage frequency (SCDF) for each seismic acceleration bin, including importance measures (e.g., Risk Achievement Worth and Fussel-Vesely)

(2) A summary of the methodologies used to estimate the SCDF and seismic large early release frequency (SLERF), including the following:

i. Methodologies used to quantify the seismic fragilities of SSCs, together with key assumptions ii. SSC fragility values with reference to the method of seismic qualification, the dominant failure mode(s), and the source of information iii. Seismic fragility parameters iv. Important findings from plant walkdowns and any corrective actions taken

v. Process used in the seismic plant response analysis and quantification, including the specific adaptations made in the internal events PRA model to produce the seismic PRA model and their motivation vi. Assumptions about containment performance (3) Description of the process used to ensure that the SPRA is technically adequate, including the dates and findings of any peer reviews (4) Identified plant-specific vulnerabilities and actions that are planned or taken Note that 50.54(f) letter Enclosure 1 paragraphs 1 through 6, regarding the seismic hazard evaluation reporting, also apply, but have been satisfied through the previously submitted NAPS Seismic Hazard and Screening Report submittal [3]. Further, 50.54(f) letter Enclosure 1 paragraph 9 requests information on the Spent Fuel Pool, which was submitted separately [15].

Table 2-1 provides a cross-reference between the 50.54(f) reporting items noted above and the location in this report where the corresponding information is discussed.

The SPID [2] defines the principal parts of an SPRA, and the NAPS SPRA has been developed and documented in accordance with the SPID. The main elements of the SPRA performed for NAPS in response to the 50.54(f) letter correspond to those described in Section 6.1.1 of the SPID, i.e.:

Seismic hazard analysis Seismic structure response and SSC fragility analysis Systems/accident sequence (seismic plant response) analysis Risk quantification Page 5 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Summary Report March 2018 Table 2-2 provides a cross-reference between t~e reporting items noted in Section 6.8 of the SPID, other than those already listed in Table 2-1, and provides the location in this report where the corresponding information is discussed.

The NAPS SPRA and associated documentation has been peer reviewed against the PRA Standard [4] in accordance with the process defined in NEI 12-13 [S], as documented in the NAPS SPRA Peer Review Report. The NAPS SPRA, complete SPRA documentation, and details of the peer review are available for NRC review.

This submittal provides a summary of the SPRA development, results and insights, and the peer review process and results, sufficient to meet the 50.54(f) information request in a manner intended to enable NRC to understand and determine the validity of key input data and calculation models used, and to assess the sensitivity of the results to key aspects of the analysis.

The content of this report is organized as follows:

Section 3 provides information related to the NAPS seismic hazard analysis.

Section 4 provides information related to the determination of seismic fragilities for NAPS SSCs included in the seismic plant response.

Section 5 provides information regarding the plant seismic response model (seismic accident sequence model) and the quantification of results.

Section 6 summarizes the results and conclusions of the SPRA, including identified plant seismic issues and actions taken or planned.

Section 7 provides references.

Section 8 provides a list of acronyms used.

Appendix A provides an assessment of SPRA Technical Adequacy for Response to NTTF 2.1 Seismic 50.54(f) Letter, including a summary of NAPS SPRA peer review.

\

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NAPS Units 1 and 2 10 CFR 50.54(f) NTIF 2.1 Seismic PRA Summary Report March 2018 Table 2-1 Cross-Reference for S0.54(f) Enclosure 1 SPRA Reporting 50.54(f) Letter Reporting Item Description Location in this Report (1) List of the significant contributors Section 5 to SCDF for each seismic acceleration bin, including importance measures (2) Summary of the methodologies Sections 3, 4, 5 used to estimate the SCDF and LERF (2)i Methodologies used to quantify Section 4 the seismic fragilities of SSCs, together with key assumptions (2)ii SSC fragility values with Tables 5.4-2 and 5.5-2 provide fragilities reference to the method of (median acceleration capacity [Am] and seismic qualification, the aleatory [~r] and epistemic variability dominant failure mode(s), and [~u]), failure mode information, and the source of information method of determining fragilities for the top risk significant SSCs based on standard importance measures such as Fussell-Vesely (FV).

(2)iii Seismic fragility parameters Tables 5.4-2 and 5.5-2 provide fragilities (Am, ~r, ~u) information for the top risk significant SSCs based on standard importance measures such as FV.

(2)iv Important findings from plant Section 4.2 addresses walkdowns and walkdowns and any corrective walkdown insights.

actions taken (2)v Process used in the seismic plant Sections 5.1 and 5.2 provide this response analysis and information.

quantification, including specific adaptations made in the internal events PRA model to produce the seismic PRA model and their motivation Page 7 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTIF 2.1 Seismic PRA Summary Report March 2018 Table 2-1 Cross-Reference for 50.54(f) Enclosure 1 SPRA Reporting S0.54(f) Letter Reporting Item Description Location in this Report (2)vi Assumptions about containment Sections 4.3 and 5.5 address performance containment and related SSC performance (3) Description of the process used App. A describes the assessment of SPRA to ensure that the SPRA is technical adequacy for the 50.54(f) technically adequate, including submittal and results of the SPRA peer the dates and findings of any review peer reviews (4) Identified plant-specific Section 6 addresses this topic.

vulnerabilities and actions that are planned or taken Page 8 of 181 L_

NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Summary Report March 2018 Table 2-2 Cross-Reference for Additional SPID Section 6.8 SPRA Reporting SPID Section 6.8 Item 111 Description Location in this Report A report should be submitted to the NRC Entirety of the submittal addresses summarizing the SPRA inputs, methods, and this.

results.

The level of detail needed in the submittal Entirety of the submittal addresses should be sufficient to enable NRC to this. The summary report identifies understand and determine the validity of all key methods of analysis and input data and calculation models used referenced codes and standards.

The level of detail needed in the submittal Entirety of the submittal addresses should be sufficient to assess the sensitivity of this. Results sensitivities are the results to all key aspects of the analysis discussed in the following sections:

5.7 (SPRA model sensitivities) 4.4 Fragility screening (sensitivity)

The level of detail needed in the submittal Entirety of the submittal report should be sufficient to make necessary addresses this.

regulatory decisions as a part of NTTF Phase 2 activities.

It is not necessary to submit all of the SPRA Entire report addresses this. This documentation for such an NRC review. report summarizes important Relevant documentation should be cited in the information from the SPRA, with submittal, and be available for NRC review in detailed information in lower tier easily retrievable form. documentation.

Documentation criteria for a SPRA are This is an expectation relative to identified throughout the ASME/ANS Standard documentation of the SPRA that the

[4]. Utilities are expected to retain that utility retains to support application documentation consistent with the Standard. of the SPRA to risk-informed plant decision-making.

Note (1): The items listed here do not include those designated in SPID Section 6.8 as "guidan_ce".

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NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Summary Report March 2018 3.0 NAPS Seismic Hazard and Plant Response This section provides summary site information and pertinent features including location and site characterization. The subsections provide brief summaries of the site hazard and plant response characterization.

North Anna Power Station is a dual. unit Westinghouse 3-loop pressurized water reactor plant located on a peninsula on the southern shore of Lake Anna, approximately 45 miles northwest of Richmond, Virginia. The reactor buildings are founded on competent bedrock; other principal structures are founded on weathered bedrock or on structural fill overlying bedrock. The bedrock has been weathered unevenly into saprolitic soils of varying thickness, ranging from a few feet to as much as 100 ft below original grade.

Detailed studies carried out during the siting investigation for North Anna Units 1 and 2, and more recently for the proposed North Anna Unit 3, show that there are no capable faults within the site vicinity. Additional site description and composite profile development are described in the NAPS NTTF 2.1 Seismic Hazard and Screening Report submittal [3].

3.1 Seismic Hazard Analysis This section discusses the seismic hazard methodology, presents the final seismic hazard results used in the SPRA, and discusses important assumptions and important sources of uncertainty.

The seismic hazard analysis determines the annual frequency of exceedance for selected ground motion parameters. The analysis involves use of earthquake source models, ground motion attenuation models, characterization of the site response (e.g. soil column}, and accounts for the uncertainties and randomness of these parameters to arrive at the site seismic hazard. Detailed information regarding the NAPS site hazard was provided to NRC in the seismic hazard information submitted in response to the NTTF 2.1 Seismic information request [3]. That information was used in development of the NAPS SPRA.

3.1.1 Seismic Hazard Analysis Methodology The seismic hazard was developed for the NAPS SPRA as described in the NAPS NTTF 2.1 Seismic Hazard and Screening Report submittal [3]. A GMRS was developed from the uniform hazard response spectra (UHRS), which are based on hard-rock ground motions determined a5, ,part of the probabilistic seismic hazard analysis (PSHA) and the site response analysis, at the control point defined in accordance with the SPID [2]. The control point for NAPS is defined as the foundation bearing elevation of the highest rock-supported, safety-related structure, which corresponds to the Casing Cooling Tank and Pumphouse structure.

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NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Summary Report March 2018 The reference earthquake used in developing building response, fragility evaluations, and risk quantification corresponds to the GMRS at the control point. The GMRS has a peak ground acceleration (PGA) of 0.572g.

Horizontal foundation input response spectra (FIRS) were developed as input to the dynamic analyses of structures that are not founded on grade (shown in Figure 3-1).

The calculation of the horizontal FIRS is consistent with the methodology used to develop the GMRS, including development of soil column profiles, site amplification functions, and UHRS. As applicable, soil properties that are strain compatible with the FIRS are developed consistent with the approach suggested by SPID. The FIRS are directly used in the probabilistic SSI analysis of the Containment Building, Service Building, Auxiliary Building, and Main Steam Valve House Unit 2. The FIRS are further adjusted to generate SSI input response spectra which are suitable for deterministic analysis (per requirements of ISG-17) and used in the SSI analysis of the Service Water Pump House and Service Water Valve House and in fixed-base analyses of the Main Steam Valve House Unit 1 and Safeguards Buildings. The development of vertical FIRS is described in Section 3.1.4.

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NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Summary Report March 2018 Figure 3-1 : Horizontal GMRS and FIRS/ SSI Input 1.40

- service 1.20 Building

- Auxiliary

~ Building C - containment 0

~ 1.00

~

n,

- safeguards (II I.I

l 0.80 - Main Steam Valve House 1 n,

- Main Steam I.I Valve House 2

~ 0.60 II)

- Service Water ++---+--+-l+-+-7!ll'F*:f-t--~ k'--+-+-+~ ~

"C Pump House (II Q. - Service Water C

~ 0.40 Valve House ++-----,~ ~ ~

- - GMRS in 0.20 0.00 0.1 1 10 100 Frequency [Hz]

3.1.2 Seismic Hazard Analysis Technical Adequacy The NAPS SPRA hazard methodology and analysis associated with the horizontal GMRS were submitted to the NRC as part of the NAPS Seismic Hazard Submittal [3], and found to be technically acceptable by NRC for application to the NAPS SPRA [16].

The NAPS hazard analysis was also subjected to an independent peer review against the pertinent requirements in the PRA Standard [4]. The SPRA was peer reviewed relative to Capability Category II for the full set of requirements in the Standard and determined to be acceptable for use in SPRA applications [6].

The peer review assessment, and subsequent disposition of peer review findings, is described in Appendix A.

3.1.3 Seismic Hazard Analysis Results and Insights Table 3-1 provides the final seismic hazard results used as input to the NAPS SPRA, in terms of exceedance frequencies as a function of PGA level for the mean and several fractiles. Information on the vertical hazard is discussed in Section 3.1.4.

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NAPS Units 1 and 2 10 CFR 50.54(f) NTIF 2.1 Seismic PRA Summary Report March 2018 Table 3-1 NAPS Mean and Fractile Exceedance Frequencies Exceedance Frequency (/yr)

PGA (g) Mean 16th Fractile 50th Fractile 84th Fractile 0.0530 1.07E-03 6.26E-04 1.03E-03 1.07E-03 0.0648 1.05E-03 4.73E-04 8.83E-04 1.07E-03 0.0717 9.96E-04 4.11E-04 7.82E-04 1.07E-03 0.0793 9.06E-04 3.58E-04 6.85E-04 1.07E-03 0.1019 6.51E-04 2.49E-04 4.80E-04 1.03E-03 0.1524 3.71E-04 1.39E-04 2.74E-04 6.34E-04 0.2061 2.42E-04 8.66E-05 1.77E-04 3.97E-04 0.3082 1.37E-04 4.24E-05 9.80E-05 2.13E-04 0.5097 5.70E-05 1.72E-05 3.80E-05 9.48E-05 0.7248 2.81E-05 8.64E-06 1.92E-05 4.59E-05 1.0306 1.39E-05 3.57E-06 9.37E-06 2.16E-05 1.5411 5.34E-06 1.25E-06 3.31E-06 8.88E-06 2.0840 2.39E-06 4.77E-07 1.48E-06 3.89E-06 2.5483 1.40E-06 2.31E-07 8.26E-07 2.25E-06 3.1162 7.83E-07 1.08E-07 4.25E-07 L28E-06 3.6237 4.84E-07 5.52E-08 2.54E-07 8.11E-07 4.0071 3.48E-07 3.50E-08 1.78E-07 5.84E-07 5.1000 1.53E-07 1.00E-08 6.92E-08 2.51E-07 In the SPRA plant model, described in Section 5, the hazard data in Table 3-1 was discretized into 10 intervals, with parameters as listed in Table 3-2.

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NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Summary Report March 2018 Table 3-2 Acceleration Intervals and Interval Frequencies as Used in SPRA Model Interval Interval Lower Interval Upper Representative Interval Mean Designator Bound (g} Bound (g} Magnitude PGA Frequency (/yr}

(g}

%G01 0.06 0.3 0.13 9.21E-04

%G02 0.3 0.4 0.35 5.34E-05

%G03 0.4 0.5 0.45 3.0lE-05

%G04 0.5 0.6 0.55 1.79E-05

%GOS 0.6 0.7 0.65 1.llE-05

%GOG 0.7 0.8 0.75 7.0SE-06

%G07 0.8 1.0 0.89 8.26E-06

%GOS 1 1.5 1.22 9.09E-06

%G09 1.5 2.5 1.94 4.25E-06

%G10 2.5 5.1 2.75 1.48E-06 Uncertainties in the PSHA result from uncertainties in input models and parameters.

These have been investigated for the NAPS SPRA. The composited seismic hazard includes Background seismic sources and individual repeated large magnitude earthquake (RLME) sources: Charleston, New Madrid Seismic Zone (NMSZ), Wabash Valley, and the northern segment of the Eastern Rift Margin fault. For 1 Hz spectral frequency at a mean annual frequency of exceedance (MAFE) of lE-04, the Background seismic sources are dominant with Charleston and New Madrid together contributing about 15% of the total hazard. At lower MAFE levels, the Background sources become even more dominant. For 10 Hz spectral frequency a larger and almost complete contribution to the total hazard is from the Background seismic sources. The observation that high frequency is tending to be controlled by the background seismic sources and the low frequency having a larger contribution from the RLME seismic sources is commonly observed for sites in the CEUS. Sites located closer to a RLME would be expected to have a larger contribution from the RLME seismic sources, especially for the low frequency cases.

The ECC-AM seismic source, which is the background source zone in which the site is located, contributes the most individual hazard to the Background total with the combination of MESE-N and STUDY-R together contributing about the same hazard as ECC-AM. This observation is similar for both the 1 Hz and 10 Hz cases.

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NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Summary Report March 2018 Finally, the last sensitivity is for the individual ground motion models used in the PSHA, which demonstrates the epistemic variation in seismic hazard among the ground motion models for the Background and RLME seismic sources for 1 Hz and 10 Hz spectral frequencies. At hazard levels of about 10-4 to 10-6, the epistemic range is about a factor of 20 to 30 for the 1 Hz spectral frequency. For the 10 Hz spectral frequency, the epistemic range is somewhat narrower, about a factor of 10.

Based on these sensitivities, the largest variation is based on the individual ground motion models implemented in the PSHA. The host background seismic source zone, ECC-AM, is the controlling seismic source for the MAFE range of interest at both the low and high frequency cases with a more significant contribution for the high frequency case relative to the low frequency case. These observations and the other sensitivity results presented in the FSAR for North Anna Unit 3 [24] are in agreement with the general observation for sites located in the CEUS that are not relatively close to a given RLME seismic source.

3.1.4 Horizontal and Vertical Response Spectra The vertical response spectra (GMRS and FIRS) used as input to SPRA analyses were derived from the horizontal spectra by scaling using an appropriate frequency-dependent vertical-to-horizontal (V/H) ratio. The V/H ratio was developed in accordance with the guidance in Appendix J of NUREG/CR-6728 [17].

To illustrate the results of the vertical spectra development, Table 3-3 provides the frequency-specific data for the horizontal and vertical GMRS at the control point along with the corresponding V/H ratio and Figure 3-2 provides a plot of the horizontal and vertical GMRS.

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NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Summary Report March 2018 Table 3-3: NAPS Control Point GMRS and V/H Ratios Frequency (Hz) Horizontal GMRS (g) V/H Ratio Vertical GMRS (g) 100.000 0.5721 1.0000 0.5721 90.000 0.6149 1.0376 0.6380 80.000 0.6965 1.0901 0.7593 70.000 0.8132 1.1275 0.9169 60.000 0.9601 1.1371 1.0918 50.000 1.1145 1.1245 1.2532 45.000 1.1652 1.1024 1.2845 40.000 1.2155 1.0423 1.2669 35.000 1.2617 0.9808 1.2374 30.000 1.2226 0.9368 1.1453 25.000 1.1889 0.8800 1.0462 20.000 1.1670 0.8256 0.9635 15.000 1.1707 0.7882 0.9227 12.500 1.1525 0.7708 0.8883 10.000 1.0508 0.7500 0.7881 9.000 0.9622 0.7500 0.7217 8.000 0.8562 0.7500 0.6421 7.000 0.7346 0.7500 0.5510 6.000 0.6068 0.7500 0.4551 5.000 0.4847 0.7500 0.3635 4.000 0.3702 0.7500 0.2777 3.000 0.2667 0.7500 0.2QOO 2.500 0.2159 0.7500 0.1619 2.000 0.1770 0.7500 0.1327 1.500 0.1317 0.7500 0.0988 1.250 0.1065 0.7500 0.0799 Page 16 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Summary Report March 2018 Table 3-3: NAPS Control Point GMRS and V /H Ratios Frequency {Hz) Horizontal GMRS (g) V/H Ratio Vertical GMRS (g) 1.000 0.0806 0.7500 0.0605 0.900 0.0745 0.7500 0.0559 0.800 0.0677 0.7500 0.0508 0.700 0.0602 0.7500 0.0452 0.600 0.0522 0.7500 0.0391 0.500 0.0435 0.7500 0.0326 0.400 0.0347 0.7500 0.0260 0.300 0.0260 0.7500 0.0195 0.200 0.0174 0.7500 0.0130 0.167 0.0145 0.7500 0.0109 0:125 0.0109 0.7500 0.0082 0.100 0.0087 0.7500 0.0065 Page 17 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Summary Report March 2018 1.4

- - Horizontal GMRS 1.2

- - *Vertical GMRS

§ 1 C:

0

~

fC1J 0.8 aiu u

c:t 0.6

.......u

!ti C1J

~ 0.4 0.2 0

0.1 1 10 100 Frequency (Hz)

Figure 3-2: NAPS Horizontal and Vertical GMRS 4.0 Determination of Seismic Fragilities for the SPRA This section provides a summary of the process for identifying and developing fragilities for SSCs that participate in the plant response to a seismic event for the NAPS SPRA. The subsections provide brief summaries of these elements.

4.1 Seismic Equipment List For the NAPS SPRA, a seismic equipment list (SEL) was developed that includes those SSCs that are important to achieving safe shutdown following a seismic event, and for mitigating radioactivity release if core damage occurs, and that are included in the SPRA model. The methodology used to develop the SEL is consistent with the guidance provided in EPRI 3002000709, SPRA Implementation Guide [10].

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NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Summary Report March 2018 4.1.1 SEL Development The SEL includes the equipment and systems required to provide protection for all seismically induced initiating events, including those needed to address seismic induced fires and floods and to prevent early containment failure in an earthquake. The SEL forms the basis for the seismic fragility and systems analysis task~. The initial SEL was developed from the SSCs modeled in the internal events PRA model. The internal events PRA model is a detailed and comprehensive logic model that includes the failure of SSCs needed for mitigating the various initiating events that could occur at the site.

Additional SSCs were added to this initial list of SSCs that may have been screened out of the internal events PRA such as passive failures of buildings, structures, cable trays, HVAC ducts, block walls, and tanks. SSCs important for containment performance such as containment isolation and bypass events were added to the list.

SSCs Modeled in the Level 1 and 2 Internal Events PRA The SSCs modeled in the level 1 and 2 internal events PRA are modeled using basic events that model various failure modes of the SSCs. The internal events PRA model also includes other basic events that model operator actions, component alignment events, and other non-component basic events. Over 5100 basic events are contained in the model, which models both unit 1 and 2 Core Damage Frequency (CDF) and Large Early Release Frequency (LERF). Basic events from the PRA database that do not represent structures or equipment (except for post-initiator operator actions and recovery actions) were removed from the SEL list. Some examples of such basic events to remove include the following:

• Configuration events (such as percentage of time a specific train is running)

• Environmental events (such as percentage of time that a given temperature range exists and HVAC is required)

• Pre-initiator operator actions and operator actions that cause initiating events

• Maintenance events

• Common-cause failure basic events (unless the associated random failure basic events do not exist separately in the models)

Screening notes were documented to denote why SSCs or basic events were screened in or out of the SEL. After the screening, over 3800 basic events were screened out of the SEL using the screening criteria for screening out basic events.

SSCs can also be screened out of the SEL based on a number of reasons. For example, some SSCs are known to have significantly high seismic capacity such that they are considered to be inherently rugged. These were screened out of the SEL because their contribution to seismic risk would likely be very small. All SSCs reviewed and screened from the SEL, and the associated basis for screening, are documented.

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NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Summary Report March 2018 Passive SSCs While the SSCs added to the SEL from the internal events PRA include SSCs needed for mitigating initiating events, the internal events PRA may not model passive SSCs because the probability of their random failure is relatively low. However, during seismic events, the probability of failure of some passive SSCs could be high and have a significant contribution to risk, such as:

• Tanks

• Buildings

• Cable Trays and Conduit

• Ventilation Ducts

• Piping

• Soil Failures

• Pressure Boundaries

• Block Walls

• Cranes

• Passive Valves The general approach used in identifying passive SSCs to be included in the SEL was to obtain a list of all of the SSCs for the particular type from the plant equipment database and evaluate whether their failure impacts a mitigating function, causes flooding or fire, or impacts an operator action. Passive SSCs that were evaluated as screened out of the SEL, and the associated basis for screening, are documented. Those not screened out are modeled in the SPRA.

In addition, the plant areas housing SEL SSCs or in which operators would need to perform seismic response actions were reviewed for accessibility and evaluated for potential impact. The following structures were included in the SEL:

• Auxiliary Feedwater Pump Houses

• Service Water Pump House

• Service Water Valve House

• Service Building (including Control Room, Emergency Switchgear Rooms, and Emergency Diesel Generator Rooms)

• Auxiliary Building

• Containment Buildings

• Main Steam Valve Houses (and Quench Spray Pump Houses)

• Safeguards Buildings

• Fuel Oil Pump House

• Casing Cooling Pump Houses

• Beyond Design Basis Storage Building Page 20 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Summary Report March 2018 Cabinets and Panels The cabinets and panels included in the SEL are those that contain the following:

1. Indications and controls that Operators use to mitigate initiating events

2. Protection and control circuits that are used in reactor protection (RPS) and engineered safety feature (ESF) systems

3. Beyond Design Basis panels used to connect the cables from the FLEX 120VAC generators to the vital AC buses.

To mitigate transients, the operators follow the guidance in the emergency operating procedures (EOPs) to ensure the unit is safely shut down and the core remains covered and cooled. They rely on various instrumentation to verify successful operation of the mitigating safety functions. As part of the development of the SEL, the instrumentation required to safely shut the unit down was reviewed to determine what panels and cabinets should be evaluated for seismic capacity. The sensors and associated cabinets and control room panels are added to the composite SEL. Seismic failure of these cabinets and panels could impact operator actions.

Reactor protection circuits and sensors are not included for the following reasons.

Seismic events generally involve a loss of offsite power, which would fail power to the motor-generator sets and thus result in trip of the control rods. For seismic events where there is no loss of offsite power, the ground acceleration level is much lower than the seismic capacity of the reactor protection system sensors and cabinets that it is very unlikely that an automatic trip signal would fail due to the seismic event. In addition, the operators would manually trip the reactor if the automatic trip system failed. Note that failure of the control rods to insert is included in the SEL.

There* are a number of actuation systems that automatically actuate safety systems upon detection of adverse trends in key safety parameters. Instrumentation associated with the following was reviewed and added to the SEL:

• Safety Injection

• Containment Depressurization Actuation

• Phase A and B Containment Isolation

• Main Steam Isolation

• Undervoltage/Degraded Voltage

• Recirculation Mode Transfer

• Anticipated Transient Without Scram (ATWS) Mitigation System Actuation Circuitry The primary focus of the review of the actuation circuits was to identify the sensors that monitor the various plant parameters and the cabinets that contain the components necessary to process the signals (e.g. power supplies, comparator card, etc.). Thus, the Page 21 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTIF 2.1 Seismic PRA Summary Report March 2018 components added to the composite SEL from this review are mainly the sensors and cabinets.

The review of cabinets and panels described above resulted in including over 150 cabinets and panels in the SEL.

Containment Performance The main objective of th~ Containment Performance evaluation is to identify seismic vulnerabilities that involve early failure of containment functions. This includes consideration of Containment integrity, Containment isolation, and other Containment functions.

Section 6.5.1 of the SPID [2] provides guidance for the SSCs that should be included in the SEL that support the containment functions. Section 5.8 of the SPRA Implementation Guide [10] also includes guidance for developing a level 2 (LERF) model in seismic PRAs. Both documents provide similar guidance with respect to the SSCs that should be included in the SEL for containment analysis. SSCs associated with the following functions were added to the SEL based ori this guidance:

• Containment structure including pressure boundary

• Containment pressure suppression

• Containment isolation

• Interfacing system LOCA

• Hydrogen mitigation

• Containment vacuum

• Heat exchanger (inside Containment) pressure boundary Approximately 160 SSCs were included in the SEL for the Containment Performance functions.

Seismic-induced Fire and Flood Additional SSCs were added to the Initial SEL based on the seismic-fire and seismic-flood evaluations, as applicable.

A review was performed to identify potential plant vulnerabilities, given the combined effects of a seismic event and consequential internal fire hazard (i.e. a fire that occurs as a direct result of the seismic event), with a focus on seismically induced internal fires that may have the potential to significantly affect the plant seismic risk. Ignition sources, fire impact to SEL SSCs, spurious actuation of fire suppression systems (CO 2 and Halon), and impact on fire mitigation actions were reviewed. The seismic-induced fire review included plant walkdowns. The walkdowns and evaluations concluded that seismic-induced fire scenarios would not have a significant impact on seismic risk.

A review was performed to identify potential plant vulnerabilities, given the combined effects of a seismic event and consequential internal flood hazard (i.e. a flood that Page 22 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTIF 2.1 Seismic PRA Summary Report March 2018 occurs as a direct result of the seismic event), with a focus on seismically induced internal floods that may have the potential to significantly affect the plant seismic risk.

This evaluation sought to identify the potentially risk-significant seismically induced flood scenarios and to screen out those that were not expected to contribute significantly to plant risk or that would be subsumed by other damage states, such as building failures. Using the internal flooding PRA, a qualitative assessment was performed to identify potential seismic-induced flood scenarios that could be significant contributors to seismic risk. In addition to reviewing the North Anna Internal Flooding PRA, the following other sources were reviewed:

• Non-seismically qualified tanks

• Fire Protection piping

• Failure of Heat Exchangers

• Expansion Joints

• SpentFuelPool

• Sources within the non-seismic Turbine Building The seismic-induced flooding evaluation concluded that flooding of the Auxiliary Building due to failure of the Component Cooling Heat Exchanger service water nozzles could result in a significant contribution to risk and this flood source was added to the SPRA model. The other flood sources and scenarios screened out from unique consideration in the SPRA.

Miscellaneous Additions Relays and contactors that are prone to chatter, as identified from the relay chatter analysis (Section 4.1.2), were added to the SEL.

Additionally, in some cases, SSCs were added if potential seismic spatial interactions were identified between non-seismic SSCs near seismic SSCs (Seismic II over I} or other spatial issues were identified during walkdowns.

Other Inputs to SEL Development A number of other inputs were reviewed and SSCs added to the SEL. These include:

• Assumptions in the internal events systems model

• Review of plant process and instrumentation drawings

• Comparison with the Individual Plant Examination of External Events (IPEEE) Safe Shutdown Equipment List [8]

This final SEL includes approximately 800 SSCs (not including relays) for each unit and is documented in the SPRA documentation.

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NAPS Units 1 and 2 10 CFR 50.54(f) NTIF 2.1 Seismic PRA Summary Report March 2018 4.1.2 Relay Evaluation During a seismic event, vibratory ground motion can cause electrical contacts of seismically sensitive equipment (e.g., relays) to open or close (or 'chatter')

inadvertently. The chattering of device contacts can potentially result in spurious signals to equipment. Most electrical contact device (herein referred to as relays) chatter is either acceptable (i.e., does not impact the associated equipment), is self-correcting, or can be recovered by operator action.

An extensive relay chatter evaluation was performed for the NAPS SPRA, in accordance with SPID, Section 6.4.2 and ASME/ANS PRA Standard, Section 5-2.2. The evaluation resulted in most relay chatter scenarios screened from further evaluation based on no impact to component function. A summary of the relay evaluation is provided in Table 4-1.

Relays that could not be screened out were modeled in the SPRA. Relay-specific fragilities were determined for relays that were modeled using the separation of variables (SOV) approach.

Table 4 Relay Chatter Evaluation Summary Unit 1 Unit2 Total SSCs Evaluated 341 322 703 Devices Evaluated 2674 2332 5006 Relays/Contactors Screened In MCCs 15 15 30 4KV Breaker 24 24 48 EDG 17 17 34 Aux Relays 6 6 12 Total 62 62 124 4.2 Walkdown Approach This section provides a summary of the methodology and scope of the seismic walkdowns performed for the SPRA. Walkdowns were performed by personnel with appropriate qualifications and documented in accordance with the PRA Standard. The seismic review teams (SRT) included seismic engineering experts with extensive experience in fragility assessment.

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NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Summary Report March 2018 Walkdowns were performed to assess the as-installed condition of those SSCs included on the seismic equipment list for use in determining their seismic capacity and performing initial screening, to identify potential spatial interactions, and look for potential seismic-induced fire/flood interactions. The walkdowns included samples of distribution systems such as piping, cable trays, electrical conduits, and HVAC ducting.

The SSC walkdowns were performed in accordance with the criteria provided in EPRI NP 6041-SL [7] and/or Seismic Qualification Utility Group (SQUG) guidance in the Generic Implementation Procedure (GIP) [21]. Most SH components were reasonably accessible and in areas where inspection was possible. For the limited inaccessible components or those located in areas where significant ALARA concerns existed, alternate methods were used, such as photographs and reliance on design information.

Walkdown information obtained was used to refine the SEL, and provide input to the fragilities analysis (as-installed conditions, dimensions, interactions etc.) and SPRA modeling (e.g., regarding corre.lation and rule-of-the-box considerations). In some cases, information from previously performed walkdowns, such as the IPEEE / USI A-46 Program [8] walkdown results, was used. In these cases, a walk-by of the applicable SSCs was performed to confirm the installed condition of the SSC was consistent with the previously performed walkdown and that the results remained applicable. The walk-by included verifying that the current material conditions and configurations were consistent with the conclusions, and to identify potential spatial interaction concerns. If applicable, recent walkdowns performed for the NTTF Recommendation 2.3: Seismic effort [22], post-Mineral earthquake plant inspections performed to support NAPS restart, and ESEP [20] were used provided these walkdowns furnished the appropriate level of detail needed for the SPRA.

Seismic-induced fire and flood and operator pathways walkdowns were also performed.

The walkdown team included PRA Systems Analysts and plant Operations personnel as well as SRT members. The results of these walkdowns were used to refine the SEL as discussed in Section 4.1.

Walkdown procedures and results of walkdowns and walkbys (observations and conclusions) were documented as required per the PRA standard.

4.2.1 Significant Walkdown Results and Insights Components on the SEL were evaluated for seismic anchorage and interaction effects, effects of component degradation, such as corrosion and concrete cracking, for consideration in the development of SEL fragilities. In addition, walkdowns were performed on operator pathways, and the potential for seismic-induced fire and flooding scenarios was assessed. The information gathered during walkdown inspections was adequate for use in developing the SSC fragilities for the SPRA.

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NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Summary Report March 2018 No significant findings were noted during the NAPS seismic walkdowns. In a few instances, potential seismic spatial interaction concerns related to the higher seismic demand to be evaluated for the SPRA were identified. For example:

• Space heaters in the Emergency Diesel Generator (EDG) Rooms were identified as potential seismic spatial interaction concerns for nearby electrical cabinets and modifications to the heater supports have been developed to resolve the concern.

• Fire extinguishers and mobile firefighting carts were identified as potential seismic spatial interaction concerns for sensitive equipment in nearby cabinets and the firefighting equipment is being evaluated for restraint or relocation to resolve the concern.

No conditions that could challenge the NAPS seismic design basis were identified.

4.2.2 Seismic Equipment List and Seismic Walkdowns Technical Adequacy Initial SEL Independent Technical Review The Initial SEL is the result of screening SSCs from, or adding SSCs to, the final SEL using the general approach discussed above.

An independent in-process technical review of the initial draft SEL was performed by industry experts. The reviewers' overall assessment was that the SEL development was comprehensive and thorough. Comments from the review were resolved and documented in the SPRA documentation and the SEL was updated accordingly.

Walkdown Methodology Independent Technical Review The methodology used to perform SSC walkdowns was reviewed by industry experts.

Comments from the review were resolved and documented in the SPRA documentation.

The NAPS SPRA SEL development and walkdowns were subjected to an independent peer review against the pertinent requirements (i.e., the relevant SFR and SPR requirements) in the PRA Standard [4]. The peer review assessment, and subsequent disposition of peer review findings, is described in Appendix A, and establishes that the NAPS SPRA SEL and seismic walkdowns are suitable for this SPRA application.

4.3 Dynamic Analysis of Structures New dynamic analyses of structures that contain systems and components important to achieve safe shutdown were performed to develop structural responses and in-structure response spectra (ISRS). Scaling of responses from previous analyses (design basis, IPEEE etc.) was not performed for any structure because the shapes of the GMRS-based spectra at the foundations of structures in the SPRA are completely different Page 26 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Summary Report March 2018 when compared to the spectral shapes used in the past design-basis or other seismic analyses performed for NAPS.

NAPS is designated as a rock site since key safety-related structures (e.g., the reactor containments) are rock-founded. However, some auxiliary structures on the site are founded on soil, or partially on soil and partially on rock. Based on the founding condition, importance of the structure/components within it to the SPRA, and fidelity of the previous design-basis lumped-mass stick models (LMSM), various fixed-base and SSI analyses using either the previous LMSMs (with modifications where necessary to meet SPID requirements) or new finite element method (FEM) models for the key structures, were performed using deterministic and probabilistic methods, as appropriate. Table 4-2 shows the foundation condition, the type of model used, whether deterministic or probabilistic analysis was performed, and other relevant information for each structure that was analyzed for the SPRA.

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NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Summary Report March 2018 Table 4-2: Description of Structures and Dynamic Analysis Methods for North Anna SPRA Foundation Type of Analysis Structure Condition Model Method Comments/Other Information Shear Wave velocity> 5000 Reactor Containment Probabilistic ft/sec; SSI analysis performed Buildings (Units 1 and Rock LMSM*

SSI with incoherence, 30 SSI input 2) profiles used Service Water Pump Deterministic Soil LMSM* LB, BE, UB cases, 5 sets of T-H House SSI Structure is partially on soil, Probabilistic Service Building Rock/ Soil FEM partially on rock. 551 Analysis SSI with 30 SSI input profiles used Service Water Valve Deterrn in istic LB, BE, UB cases, 5 sets of T-H Soil LMSM*

House SSI used Structure is partially on soil, Probabilistic Auxiliary Building Rock/ Soil FEM partially on rock. 551 analysis 551 with 30 SSI input profiles used LB, BE, UB cases, 5 sets of T-H Safeguards Building Rock FEM Fixed Base used Auxiliary Feedwater One set of T-H (simple Pump Houses Rock LMSM* Fixed Base structure)

Unit 1 Main Steam LB, BE, UB cases, 5 sets of T-H Rock FEM Fixed base Valve House used Unit 2 Main Steam Probabilistic SSI Analysis, 30 SSI input Soil FEM Valve House SSI profiles used

• LMSM models were reviewed based on the criteria in EPRI SPID and found to be acceptable for use in the SPRA.

4.3.1 Input Motions for Structural Analyses The foundation input response spectra {FIRS) and 551 input response spectra, as applicable, were developed for each structure. These spectra were derived from site Page 28 of 181 L

NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Summary Report March 2018 response analyses and correspond to the GMRS. Time-histories (T-H) corresponding to these spectra were developed and used as input motions in the probabilistic and deterministic analyses of structures.

4.3.2 Damping Values In the structural dynamic analyses, 5% median damping value for concrete and 3% for steel were used per Table 3-4 of EPRI TR-103959. The 5% concrete damping is based on demands at approximately Yi the yield strength for reinforced concrete with cracking.

This value is also consistent with Table 4-1 of EPRI NP-6041-SL, Rev. 1, which recommends 5% damping for reinforced concrete with moderate cracking. An exception was the Auxiliary Feedwater Pump House, which is a simple structure, and therefore, 4% damping value for concrete was used to develop 84% responses using one set of time-history input. Median and 84 percentile ISRS were developed at various locations and elevations of structures at various damping ratios (e.g., at 1%, 2%, 3%, 4%,

5%, 7%, and 10%).

4.3.3 Fixed-base Dynamic Analyses As indicated in Table 4-2, three rock founded structures were analyzed as fixed-base because they are considered relatively low-mass buildings and the shear wave velocities at the foundation of each of these structures exceed 5000 ft/sec. Given the small footprint of these structures, their dynamic analyses were performed using coherent input motions.

Detailed dynamic analyses were performed for two structures - the Safeguards buildings (both units are similar, Unit 2 was used to develop responses for both buildings) and the Unit 1 Main Steam Valve House. New finite element models were developed for both these structures. Lower bound (LB), best estimate (BE) and upper bound (UB) cases were established by varying the structural stiffness by one-standard d~viation (through concrete Young's modulus Ee using logarithmic standard deviation of 0.3) from the BE values; this corresponds to approximately +/-15% variation of natural frequencies. A lower bound damping (LB-D) case was also analyzed with a lower damping of 3.7% (log standard deviation of 0.3) for the BE stiffness case. The input ground motions are applied using five sets of input time-histories which are spectrally matched to the SSI input response spectra. The ISRS were calculated at 301 frequencies at equal intervals in the logarithmic space between 0.1 Hz and 100 Hz (100 frequencies per decade). For each of the 20 seismic analysis cases, each node, and each damping ratio, the ISRS in the X direction are obtained by combining the acceleration response spectra (ARS) designated as XX (X response due to input in the X direction), XV (X response due to input in the V direction), and XZ (X response due to input in the Z direction) using the square root of sum of squares (SRSS) method. The V and Z direction ISRS are calculated similarly. For each node, each damping ratio, and each direction (X, V, and Z), the logarithmic mean of the ISRS due to the five sets of input time-histories for the BE case Page 29 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Summary Report March 2018 is calculated and used as the BE ISRS. Similarly, the LB, UB and LB-D ISRS are calculated from their respective five time-history cases. The variation obtained in the ISRS results from the five BE seismic analysis cases reflects the variation due to phase differences between the five sets of input time-histories. The aleatory variation of the response due to the time-history phase variation is estimated as the logarithmic standard deviation of the ISRS obtained from the five analysis cases. The variation of the response due to damping is estimated on a frequency-by-frequency basis as the natural logarithm of the ratio of the BE and LB-D ISRS. The median ISRS (SA 50 ) is estimated frequency-by-frequency as the logarithmic mean of the BE, LB, and UB ISRS results. The variation of the response due to stiffness effects is estimated on a frequency-by-frequency basis as the natural logarithm of the ratio of the envelope of the BE, LB, and UB ISRS to the median ISRS results. The broadening of the envelope ISRS is done to fill in potential gaps between the LB, BE, and UB results by connecting the ISRS peaks using straight lines.

Other sources of uncertainty, such as modelling, ground motion directivity effects and V/H ratio uncertainties are estimated separately. All uncertainties (aleatory and epistemic) are combined on a frequency-by-frequency basis to obtain the total composite uncertainty (/3c) for the ISRS. The 84th percentile ISRS are calculated as SA 50 x ePc. For the calculation of functional fragilities of equipment, peak clipped median and 84th percentile of ISRS are developed based on the methodology in Reference 9 (EPRI TR-103959).

The third structure analyzed as fixed base is the Auxiliary Feedwater Pump House. This is a simple structure and was analyzed using a lumped mass stick model (LMSM) with one set of 3-directional time-history input and 4% concrete damping to estimate 84%

non-exceedance probability (NEP) responses. ISRS were calculated at several elevations of the structure at various damping values.

4.3.4 Soil Structure Interaction (SSI) Dynamic Analyses As listed in Table 4-2, detailed probabilistic SSI analyses were performed for four key structures - Reactor Containment buildings (RCB - identical for both units), Service building (SB), Auxiliary building (AB), and the Unit 2 Main Steam Valve House (MSVH-2).

Deterministic 551 analyses were performed for the Service Water Pump House (SWPH) and Service Water Valve House (SWVH) structures.

For the RCB, the best estimate of the shear wave velocity of the supporting media below it is approximately 5200 fps. This is higher than the threshold provided by-SPID for fixed base analysis. However, an 551 analysis was performed for this building because the RC structure is tall and heavy and its response is expected to be affected by potential foundation rocking. Given the large building footprint and high-frequency-rich nature of the input motions, the ground motion incoherency was included in the development of ISRS from the 551 analysis. The existing LMSM of the RCB was considered adequate to capture the structural response and satisfied the SPID requirements for model adequacy. The horizontal and vertical FIRS for the RCB were calculated at the bottom of Page 30 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Summary Report March 2018 the mat foundation of RCB at elevation of 203 ft. These FIRS are calculated as geologic outcrop motion and are appropriate for use in the SSI analysis of the RCB as a surface structure. From the site response analysis, 30 sets of strain compatible soil properties consistent with FIRS and reflecting the rock property variations for the SSI analysis of the RCB were calculated. The SSI analysis used 30 sets of spectrally-matched time-histories which are tightly matched to the building FIRS Best estimate concrete strength of 5400 psi was used. The RCB structure was considered uncracked. Five engineering variables are identified for uncertainty modeling in the probabilistic SSI analyses: (1)

Young's modulus for concrete, (2) Structural damping ratio, (3) Dynamic soil profile properties (4) Ground motion directional variability, and (5) ground motion V/H variability. The best estimate and logarithmic standard deviation (log-SD) of all random variables were explicitly included in the analysis. Using Latin Hypercube Sampling (LHS),

30 sets of SSI input parameters were developed by combining the above variables in an unbiased fashion. Other sources of uncertainty, namely, modelling uncertainty and coherency uncertainty, are explicitly estimated and included in the calculation of the total composite uncertainty. The median and 84th percentile of the probabilistic ISRS with and without ground motion incoherency were calculated at various damping values and median values of displacements relative to the top of the RCB foundation, with and without rigid body rotations were calculated. Peak clipped ISRS were generated using the methodology in Reference 9 (EPRI TR-103959). From the probabilistic analysis, frequency dependent aleatory and epistemic variabilities due to SSI and structural response were calculated in each direction in addition to the median and 84% ISRS responses. Structure-soil-structure interaction (SSSI) effects from the RCB were evaluated on nearby structures. These effects were found only to be significant for the vertical ISRS of SG and MSVH structures by causing slightly more than 10% increase in certain frequency bands; the ISRS within these structures were adjusted to include the SSSI effects.

The probabilistic SSI analyses of SB, AB and MSVH-2 were performed in a similar manner as described above for the RCB. However, for these three structures, new finite element models were developed instead of using the previous LMSMs. Similar to the RCB, ground motion incoherency was included for the AB and SB evaluation. Note that ground motion incoherency effects were not included in evaluation of the MSVH-2, due to its small footprint. The SSSI effect of the AB on nearby structures were also evaluated and found to be negligible.

For the SWPH and SWVH structures, deterministic SSI analyses were performed using updated LMSMs based on those used in the design basis calculations. SSI analyses were performed for the LB, BE and UB soil cases, each with 5 sets of time-histories, which yields 15 SSI analysis cases. Note that the variation of the structural properties (e.g.,

stiffness and damping) were not considered significant for these buildings because their SSI response were found to be entirely dominated by the soil impedance. The same approach for combining spatial components and uncertainties as discussed for the fixed-Page 31 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Summary Report March 2018 base analysis of Safeguards building and MSVH-1 was used and the median and 84th percentile ISRS were developed at various elevations and damping values.

4.3.5 Structure Response Analysis Technical Adequacy The structural dynamic analyses were subjected to an in-process independent technical review by industry experts. Comments from the review were resolved and documented in the SPRA documentation.

The NAPS structural dynamic analyses were subjected to an independent peer review against the pertinent requirements in the PRA Standard [4]. The peer review assessment, and subsequent disposition of peer review findings, is described in Appendix A, and establishes that the NAPS structural dynamic analyses are suitable for this SPRA application.

4.4 Fragility Analyses of SSCs Seismic fragilities representing the conditional probabilities that a component would fail for a specified seismic ground motion or response as a function of that value were developed for SSCs in the SPRA seismic equipment list (SEL). The high-confidence-of-low-probability-of-failure (HCLPF) and median capacities were expressed as a fraction of the peak ground acceleration (PGA) of the control point GMRS. This PGA is 0.572g. With the exception of loss-of-offsite-power and LOCA events, which were based on the guidance in the SPRA Implementation Guide [10], seismic fragilities were plant-specific and were calculated in a realistic manner based on the actual conditions of the SSCs in the plant, as confirmed through detailed walkdowns.

This section summarizes the fragility analysis methodology, presents a tabulation of the fragilities with median capacity Am and randomness and uncertainty variabilities ~r and

~u, and the calculation method and failure modes for those SSCs determined to be sufficiently risk important, based on the final SPRA quantification. Important assumptions and important sources of uncertainty, and any particular fragility-related insights identified, are also discussed.

4.4.1 SSC Screening Approach Screening of SSCs primarily followed the guidance in Section 5.2 of SPRA Implementation Guide [10], and the guidance for screening in the Screening Prioritization and Implementation Details (SPID) - EPRl-1025287 [2]. The following methods were established for the screening of SSCs:

1. Screen inherently rugged SSCs. Inherently rugged SSCs were typically not retained in the logic model.

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NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Summary Report March 2018

2. Develop a screening HCLPF using the SPID capacity-based screening criterion.

This criterion can be applied during the walkdowns, and also via inspection of margins in the previous design basis, USI A-46 and/or IPEEE calculations.

3. Starting from the initial SPRA quantification, use a graded approach to screen SSCs and prioritize them for calculation of fragilities based on their risk significance. Review the validity of screening out SSCs that are not risk-significant via a surrogate screening event in the final logic model.

Using the North Anna control point hazard curve, the capacity-based screening HCLPF was calculated to be 1.8g. This was judged to be conservative; therefore, a l.Og HCLPF screening threshold was used. Even though SSCs had capacities greater than this 1.0g screening HCLPF, they were retained in the SPRA logic model. A surrogate event, with a HCLPF of 0.6g, was included in the logic model which provided confirmation that the contribution to SCDF and SLERF from SSCs that could have been screened out was very low.

4.4.2 SSC Fragility Analysis Methodology Detailed fragility analyses were performed for those SSCs that were not screened. The conservative deterministic failure margin (CDFM) approach per the guidance of EPRI NP-6041-SL, Revision 1 [7], supplemented by EPRl-1019200 [19] was initially used for most SSCs in the SEL, with the exception of relays. Using the CDFM approach, the HCLPF capacities were calculated using the 84% in-structure response spectra {ISRS). Detailed and more refined fragility analyses using the separation of variables (SOV) approach were performed for the top risk-important SSCs where the 50% confidence level ISRS were directly used to calculate their median capacities. The epistemic and aleatory variabilities for the fragilities calculated using the CDFM method were developed using one of the following two approaches: (a) Use the variabilities from the SPID, as appropriate, or (b) Use the detailed North Anna specific structural response variabilities (calculated frequency-by-frequency for each orthogonal direction), develop the equipment response variabilities per the guidance in EPRI TR-103959, and combine using SRSS. For the SOV method, variabilities were always calculated using approach (b) above.

In calculating the fragilities of SSCs, both structural and functional failure modes were considered. The seismic demand consisted of spectral accelerations up to a frequency of 20 Hz for structural failures such as bolted cabinets and also for functional failure modes with the exception of potentially high frequency sensitive SSCs, where a cut-off frequency limit of 40 Hz was used. For functional evaluation, peak clipped ISRS were used per the guidance in EPRI TR-103959. In many instances, functional capacities were based on Table 2-4 of EPRI NP-6041 with a modification that ISRS peaks, rather than ground peak spectral values, were used as recommended in EPRl-1019200. For SSCs covered by EPRI NP-6041 Table 2-4, 5% damped spectral peaks of only the horizontal ISRS (both directions) were compared to the modified peak spectral acceleration Page 33 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Summary Report March 2018 capacities of EPRI NP-6041 Table 2-4. An exception was the functional assessments of batteries and racks where vertical spectral peaks were also considered. EPRI NP-6041 Table 2-4 capacities were increased by a factor of 1.5 to obtain the HCLPF capacity (1%

non-exceedance probability) and/or by a factor of 4.0 to obtain the median capacity, as recommended in EPRl-1019200. For some SSCs such as relays, functional capacities were based on the available shake table test data. When a static analysis was used to determine the capacity of a beam or frame type equipment item (e.g., anchorage evaluation of a cabinet), and if the natural frequency of the item was not known, peak spectral accelerations (slightly reduced as discussed below) were used with no multi-mode factor (i.e., a multi-mode factor of unity). Where it was judged that SSCs are not significantly sensitive to seismic accelerations in one horizontal direction more than the other, calculations of HCLPF capacities based on Table 2-4 of EPRI NP-6041 were refined by using the geometric average of the spectral accelerations (i.e., clipped ISRS spectral peaks up to the 20 Hz cut-off) in the two horizontal directions rather than using the maximum of two horizontal directions. The use of the geometric averaging is consistent with EPRI NP-6041, which notes that the screening guidance provided in Tables 2-3 and 2-4 are "in terms of five percent-damped peak spectral ground acceleration (average of two orthogonal horizontal components)." Where applicable, similar SSCs in close proximity were grouped together to perform a single fragility calculation. For fragility -

analyses of SSCs in structures analyzed using SSI, frequency (peak) shifting or broadening was limited to +/-10% to address uncertainties in equipment natural frequencies because uncertainties in the soil and structural stiffnesses were already accounted for in the SSI analyses. However, for SSCs in structures analyzed using fixed-based dynamic analyses, the EPRl-recommended +/-20% peak shifting or peak broadening was used. When the natural frequency of an equipment item was not available or unknown, peak of the ISRS was used but with a slight modification. It was reasonably assumed that the component frequency has equal probability of lying within +/-15% of the frequency at which the peak spectral acceleration occurs and the spectral acceleration values within this +/-15% window centered on the peak were averaged to obtain the seismic demand. In limited cases, small reductions in the ISRS were obtained based on the coupled analyses of structures and equipment.

The fragilities of structures were initially based on Table 2-3 of EPRI NP-6041; however, detailed fragility analyses were subsequently performed for several structures because either (a) the caveats of EPRI NP-6041 Table 2-3 could not be satisfied, or (b) the use of EPRI NP-6041 Table 2-3 was conservative and more realistic fragilities were needed because the structure was high in the risk significance list of SSCs for CDF or LERF.

Fragilities of block walls in areas near the SEL items were developed by grouping the walls and analyzing the bounding cases.

T_he fragilities of reactor internals and other NSSS components were calculated using a scaling approach; these components have typically been demonstrated to have high capacities based on past SPRAs. Evaluations of representative distributed systems Page 34 of 181

NAPS Units 1 and 2 10 CFR 50.54(f} NTTF 2.1 Seismic PRA Summary Report March 2018 (piping, HVAC ducts, cable trays, and conduits} and associated components were performed; these components also have been shown to be generally rugged or have high capacities.

Correlation of components (or common cause failure} was considered in accordance with the ASME/ANS PRA Standard (4]. For the NAPS SPRA, if the equipment items were similar in design and physical orientation, with similar anchorage, and located in the same building on the same elevation, then these equipment items were assumed to be fully-correlated. In some cases, separate ISRS were used to develop location-specific fragilities for similar components located on the same floor. From the detailed finite element models of the structures, the seismic demand at different locations of the buildings was available. Since the seismic fragility of a component is a function of its seismic capacity and the seismic demand at the component location, similar components at different locations could have different demand, thus different fragilities. If the difference between the capacities of such components was small, then the components were considered correlated using the lower capacity value. However, if there was a significant difference in the fragilities of two similar components, then both detailed individual fragilities were entered in the logic model.

The impact of two (or multiple} failure modes, e.g., the functional and structural failure modes of a component, may cause the combined probability of failure to be slightly higher than the probability of either of the two failure modes, thus impacting the component's fragility. This occurs if the two failure modes are independent but not mutually exclusive (i.e., both could happen}. The probability that at least one failure will occur is expressed by the union of two events (failures} A and B or P(A U B), where P(A U B} = P(A} + P(B} - P(A} x P(B}. This consideration is more pronounced when the HCLPF capacities of the two failure modes of an item are within about 20% of each other. Thus the fragilities for two failure modes, if within 20% of each other, were combined for the top risk significant SSCs to obtain a more accurate estimate of the fragility.

4.4.3 SSC Fragility Analysis Results and Insights The final set of fragilities for the risk important contributors to SCDF and SLERF are summarized in Section 5, Table 5.4-2 (for SCDF} and Table 5.5-2 (for SLERF}. Refined fragility calculations were performed for the highest risk significant SSCs, as well as for selected other components.

4.4.4 SSC Fragility Analysis Technical Adequacy A sampling of NAPS fragility analyses were subjected to an in-process independent technical review by industry experts. Comments from the review were resolved and documented in the SPRA documentation.

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i 1

NAPS Units 1 and 2 10 CFR 50.54(f) NTIF 2.1 Seismic PRA Summary Report March 2018 The NAPS fragility analyses were subjected to an independent peer review against the pertinent requirements in the PRA Standard [4]. The peer review assessment, and subsequent disposition of peer review findings, is described in Appendix A, and establishes that the NAPS fragility analyses are suitable for this SPRA application.

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NAPS Units 1 and 2 10 CFR 50.54(f) NTIF 2.1 Seismic PRA Summary Report March 2018 5.0 Plant Seismic Logic Model The seismic plant response analysis models the various combinations of structural, equipment, and human failures given the occurrence of a seismic event that could initiate and propagate a seismic core damage or large early release sequence. This model is quantified to determine the overall SCDF and SLERF and to identify the important contributors, e.g., important accident sequences, SSC failures, and human actions. The quantification process also includes an evaluation of sources of uncertainty and provides a perspective on how such sources of uncertainty affect SPRA insights.

5.1 Development of the SPRA Plant Seismic Logic Model The NAPS seismic response model was developed by starting with the NAPS internal events at power PRA model of record as of March 30, 2017, and adapting the model in accordance with guidance in the SPID [2] and PRA Standard [4], including adding seismic fragility-related basic events to the appropriate portions of the internal events PRA, eliminating some parts of the internal events model that do not apply or that were screened-out, and adjusting the internal events PRA model human reliability analysis to account for response during and following a seismic event. The model is developed using the EPRI CAFTA software suite. This model credits FLEX equipment in the SBO sequences as well as low leakage reactor coolant pump (RCP) seals. Both random and seismic-induced failures of modeled SSCs are included. The modifications to develop the SCDF fault tree are summarized in Table 5.1-1. The following sections provide additional description in the development of the SPRA.

Seismic Equipment List A seismic equipment list (SEL) was developed to define the scope of SSCs to include in the SPRA. Guidance in the SPRA Implementation Guide [10) was used in the development of the SEL. The SSCs modeled in the internal events PRA was used as a start for the SEL. Plant drawings, procedures and other design and configuration resources were reviewed and SSCs are added to the SEL to capture specific failures that can occur during seismic events and are not modeled in the FPIE PRA. The SSCs on the SEL were also reviewed to identify relays that could impact the SSC function if the relay contacts chattered during a seismic event. The circuits for the SSCs were reviewed to determine which relay contacts could impact the SSC function. Over 120 relays screened in and are modeled in the SPRA. Section 4.1 contains additional details of the SEL.

Initiating Events and Accident Sequences The seismic hazard was modeled using 10 discrete hazard intervals (or bins) based on increasing peak ground acceleration. The seismic hazard bins are as listed in Table 3-2.

Each bin is treated as a seismic initiator and the SCDF (and SLERF) results are summed Page 37 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Summary Report March 2018 over all the bins to obtain the total SCDF (and SLERF). Bin-specific SSC fragilities are used in the accident sequences for each bin.

The SPRA models each seismic event (i.e., each bin) as possibly leading to transients and LOCAs (small, medium, large, and excess LOCA (e.g., reactor pressure vessel failure)),

without onsite AC power, and with response reflecting impact of the seismic event on mitigating systems. The event trees that model the seismic accident sequences are essentially the same as the event trees for the internal events core damage event trees.

The following seismic-induced initiating events are modeled:

• LOOP

• ATWS

• Small-small LOCA

• Small LOCA

• Medium LOCA

• Large LOCA

• SBO

• Damage - includes excessive LOCA, building failures, distributed systems, etc Modeling of Correlated Components Fully correlated components were assigned to correlated component groups so that all components in the group fail with the same probability based on the seismic magnitude for each hazard bin. The model assumes fully correlated response of same or very similar equipment in the same structure, elevation, and orientation. Correlated component groups were developed for all redundant components in the model that met these correlation criteria. The seismic capacity for the group was assigned the capacity of the weakest component in the group. If the components are located in different areas or there are significant differences in the capacities of the components in the group due to differences in in-structure response spectra, the components were modeled as uri-correlated. Section 4.4.2 contains additional information on correlation.

Modeling of Human Actions Human error probabilities (HEP) for operator actions in the SPRA model are developed using the same methodology as in the internal events PRA. The EPRI Human Reliability Analysis (HRA) Calculator software was used to develop and document the HEPs for the internal events actions and for new HEPs for mitigating seismic failures of mitigating functions. HEPs were then adjusted as a function of seismic magnitude using a performance shaping factor approach consistent with the EPRI seismic HRA methodology [18]. Each Operator action is modeled by four HEP basic events that model the probability of failure for four different seismic hazard intervals. The ten Page 38 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Summary Report March 2018 hazard intervals are binned into the four HRA bins, which allow adjusting the HEP probabilities to account for increased stress and other shaping factors due to higher ground motion. The importance of the four HEPs is combined to obtain the overall importance ofthe Operator action.

The HEPs in the SPRA contains logic for failing the HEPs if SSCs needed by the Operators to complete the actions are failed. For example, failure of the main control panels or the process cabinets fails the HEPs.

A complete dependency analysis was performed on all human actions (including both seismic-specific actions and actions included in the internal events model on which the SPRA is based) required for a response to a seismic event. The dependency module in the HRA Calculator was used to determine the level of dependencies and the probability of the dependent HEPs. The dependent HEPs are added to the cutsets using a recovery file.

SLERF Model The additional seismic initiating events, and their associated accident sequences, added to the core damage model were also added to the seismic LERF model. The seismic core damage accident sequences were mapped to the appropriate SLERF damage states based on the mapping in the internal events level 2 PRA. Most core damage sequences went to several SLERF damage states depending on failures in the Level 2 event trees from the internal events PRA. Some of the new core damage sequences, such as failure of the buildings and containment isolation, were directly mapped to SLERF. Others, such as a SBO sequences, were mapped based on similar core damage sequence mapping, using the level 2 event trees in the internal events PRA.

Additional SSC Failures Modeled in the SPRA Certain failures are modeled as leading directly to core damage given the potential for multiple system impacts or distributed system failures. These include seismic failure of:

• Distributed Systems - Cable Trays/Conduit

• Distributed Systems - Piping

• Building Failures - Reactor Containment Building, Auxiliary Building, Service Building i

• Excessive LOCA caused by failure of Reactor Vessel, Steam Generators, Reactor Coolant Pumps As part of the seismically-induced internal floods evaluation, seismic failure of the Component Cooling heat exchangers resulting in failure of the Service Water supply piping to the heat exchangers was included in the SPRA logic for failing SSCs in the Auxiliary building.

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NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Summary Report March 2018 Table 5.1-1 Summary of Modifications to Internal Events CDF Fault Tree to Create Seismic CDF Fault Tree Recovery of offsite power is not credited in the SBO sequences.

SBO event tree modified to not credit recovery of offsite power and added credit for selected FLEX actions:

• load shed batteries to extend vital 125VDC battery life

• repower 120VAC vital buses using FLEX generators

• installing FLEX RCS Injection Pump to makeup to RCS (SSLOCA assumed)

RCP Seal LOCA model revised to use the Flowserve N9000 low leakage seal failure probabilities. The seals for all RCPs at North Anna have been replaced with Flowserve seals.

Added spurious opening of the pressurizer PORVs due to seismic failure of

.reactor pressure signals.

• Revised HEPs in the seismic accident sequences to model four HEPs. The four seismic HEPs model the probability of failure at four different seismic ground motion bins.

Added seismic failures that impact Operator actions to fail HEPs. For example, seismic failure of the MCR panels, process cabinets, or instrumentation are modeled as failing HEPs.

Added over 160 fragility groups to the PRA fault trees that model seismic failure of the various SSCs that are used for mitigating seismic-induced accidents.

Various miscellaneous changes were made to the fault trees to accommodate new logic for the seismic model.

5.2 SPRA Plant Seismic Logic Model Technical Adequacy The initial NAPS SPRA seismic plant response logic model was reviewed by industry experts. Comments from the review were resolved and documented in the SPRA documentation.

The NAPS SPRA seismic plant response methodology and analysis were subjected to an independent peer review against the pertinent requirements in the PRA Standard [4].

The peer review assessment, and subsequent disposition of peer review findings, is described in Appendix A, and establishes that the NAPS SPRA seismic plant response analysis is suitable for this SPRA application.

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NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Summary Report March 2018 5.3 Seismic Risk Quantification In the SPRA risk quantification, the seismic hazard is integrated with the seismic response analysis model to calculate the frequencies of core damage and large early release of radioactivity to the environment. This section describes the SPRA quantification methodology and important modeling assumptions.

5.3.1 SPRA Quantification Methodology For the NAPS SPRA, the following approach was used to quantify the seismic plant response model and determine seismic CDF and LERF:

The EPRI FRANX software code was used to discretize the seismic hazard into the 10 seismic initiators. FRANX was also used to generate the fault tree gates that model seismic failure of the SSC fragility groups modeled in the systems fault trees. The Unit 1 and 2 seismic CDF and seismic LERF top gates were quantified using the EPRI PRAQuant code to obtain cutset files that were then processed using the EPRI Code ACUBE. ACUBE was used to obtain a more accurate CDF/LERF by calculating the exact probability on the set of SCDF/SLERF cutsets. ACUBE does not use the rare events approximation as is utilized in CAFTA's min cut upper bound estimation calculation and so ACUBE provides a more accurate solution. Additional details can be found in the following sections, along with descriptions of sensitivity studies, uncertainty estimations and a more complete description on the insights from _top contributors to SCDF/SLERF.

5.3.2 SPRA Model and Quantification Assumptions The following assumptions were made as part of the seismic PRA quantification:

1. Due to the relatively low fragility of the insulators on the switchyard transformers, a loss of offsite power (LOOP) is likely to occur during most seismic events. The model includes SEIS-LOOP in all sequences in the Seismic Event Tree.

2. The seismic capacity for small-small LOCA is assumed to be 0.12g, which is the Safe Shutdown Earthquake (SSE) for North Anna. Guidance in SPRA Implementation Guide [10] includes several options, but generally recommends using the SSE as the capacity if detailed fragility calculations and walkdowns of the RCS piping are not performed.

3. Chatter of multiple relays in series where the contacts of the relays have to chatter in unison is considered to have a very low likelihood and therefore is not considered in the relay chatter evaluation.

4. Some SSCs that are part of alternate or backup mitigating functions were not credited in the SPRA either due to their low seismic capacities or to reduce the scope of the fragility analyses. For example, the alternate AC diesel generator is not credited because the seismic capacity of the building and support SSCs is likely to be low. Likewise, the Condensate Storage Tanks (CSTs), which are used Page 41 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Summary Report March 2018 to supply the AFW pumps when the Emergency Condensate Storage Tank (normal AFW supply) is depleted, are not credited because the CSTs are unanchored, flat bottom tanks that typically have low capacity.

5. Seismic failure of the Component Cooling heat exchangers was assumed to result in flooding of the Auxiliary building SSCs from the failure of the Service Water supply piping to the heat exchangers. Isolation of the flood was not credited given the uncertainty in the size of the pipe breaks and the resulting flood flow rate.

6. Seismic failure of the Steam Generator (SG) tubes is not considered to be controlling and is subsumed by failure of the SG supports, which is assumed to result in an excessive LOCA.

7. Mission time is assumed to be 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. A sensitivity using a mission time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> showed little impact on the SPRA results.

5.4 SCDF Results This section presents the base SCDF results, a list of the SSCs that are significant contributors, including risk importance measures, a discussion of significant sequences and/or cutsets and their relative SCDF contributions. A discussion of sensitivity studies is provided in Section 5.7.

The seismic PRA performed for NAPS shows that the point estimate seismic CDF is 5

6.0xl0- for both Unit 1 and Unit 2. A discussion of the mean SCDF with uncertainty distribution reflecting the uncertainties in the hazard, fragilities, and model data is presented in Section 5.6. Important contributors are discussed in the following paragraphs.

The top SCDF accident sequences based on Fussell-Vesely (FV) importance of the sequence flags are documented in Table 5.4-1. These sequences contribute over 90% of the SCDF. Note that these sequences have been combined across all the hazard bin intervals. Three of the top seven sequences are seismic events with a loss of offsite power and failure of the EDGs due to relay chatter resulting in a Station Blackout (SBO}.

SSCs with the most significant seismic failure contributions to SCDF are listed in Table 5.4-2, sorted by FV importance. The seismic fragilities for each of the significant contributors are also provided in Table 5.4-2, along with the corresponding limiting seismic failure mode and method of fragility calculation. Importance analyses were performed for both SCDF and SLERF, using the ACUBE code. From the ACUBE output, FV values for the seismic failures (i.e. fragility groups) is the sum of the FV values for each hazard interval.

The FV listing shows the top individual contributor to SCDF as seismically induced Loss of Offsite Power (LOOP}, due to the low median seismic capacity assumed for offsite power failure following a seismic event. The fragility for LOOP is a value from the SPRA Implementation Guide [10] and considered reasonably representative for NAPS.

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NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Summary Report March 2018 The next highest contributor is seismically induced small-small LOCA (SSLOCA), which similar to LOOP, has a low median capacity. The capacity is based on the SPRA Implementation Guide, which provides guidance for modeling SSLOCA and recommends the capacity (i.e. HCLPF) be set to the Safe Shutdown Earthquake (SSE), which for NAPS is 0.12g.

Most of the top seismic failures involve chatter of relays that result in failure of emergency power, or key safety system pumps due to chatter of the 4kv breaker lockout relays. The capacity of these relays is relatively low and the seismic failures for each are assumed to be correlated (e.g., both trains of LHSI pumps fail due to lockout).

The model does not currently credit Operator action to reset the relays and restore the mitigating functions.

Other top seismic failures involve failure of the 120vac vital buses and the vital bus inverters, which not only fail the actuation systems and power to some SSCs, but also fails critical instrumentation relied on for Operator actions (i.e. fails Human Error Probability basic events in the model). Failure of the vital 125v DC buses and batteries also have significant FV importances, which have similar impacts as the vital buses.

Table 5.4-1 Summary of Top SCDF Accident Sequences FV Importance Accident Sequence Description Ul =8.3E-02 Station Blackout (SBO) with successful Auxiliary Feedwater (i.e.

(29%) Turbine-driven AFW pump) but either. Long Term Cooling fails, Cooldown and Depressurization fails, or the SI Accumulators fail.

U2 =7.7E-02 The dominant failures are:

(27%)

• SBO caused mainly by relay chatter of EOG output breaker or 4kv breaker supply to the 480V buses and MCCs; no credit for Operators recovery of the relay chatter.

• Seismic failure of the Steam Generator (SG) Power Operated Relief Valves (PO RVs)

• Seismic failure of the 120VAC vital buses, DC buses and inverters that power critical instrument transmitters required for Operator actions

• Seismic failure of Main Control Room panels Sequence Ux-SBO-SEIS-02 Page 43 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Summary Report March 2018 Table 5.4-1 Summary of Top SCDF Accident Sequences FV Importance Accident Sequence Description Ul = 6.SE-02 Loss of Offsite Power with a Small-small LOCA and successful AFW (24%) and long term cooling and failure of RCS makeup using the Charging pumps. The dominant failures are:

U2 = 7.2E-02

• Seismic failure of the RWST (25%)

• Seismic failure of SW pumps due to chatter of the lockout relays which fails cooling to the Charging pumps

• Seismic failure of the Charging pumps due to chatter of the lockout relays

• Seismic failure ofthe Low Head Safety Injection (LHSI) pumps due to relay chatter or due to failure of the Safeguards area ventilation where the pumps are located

• Seismic failure ofthe SG PORVs Sequence Ux-LOOP-SEIS-01 Ul = 4.2E-02 SBO with successful AFW and Long Term Cooling, but FLEX (15%) mitigation fails due to the following:

• Seismic failure of the RWST which fails RCS makeup from the U2 = 4.0E-02 FLEX RCS Injection Pump (14%)

• Seismic failure of the vital 125vdc batteries resulting in loss of critical instrumentation

• Seismic failure of the FLEX electrical distribution panel Sequence Ux-SBO-SEIS-01 Ul = 3.lE-02 SBO with failures that go directly to core damage due to (11%) insufficient time to mitigate (large, medium, small LOCAs, ATWS).

Dominant failures that result in this SBO direct core damage U2 = 3.0E-02 sequence are:

(11%)

• Small LOCA

• Control Rods

• Medium LOCA Sequence Ux-LOOP-SEIS-04 Page 44 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Summary Report March 2018 Table 5.4-1 Summary of Top SCDF Accident Sequences FV Importance Accident Sequence Description Ul = 2.lE-02 Loss of Offsite Power with successful AFW but long term cooling (7%) fails (i.e. align Service Water or Fire Protection to AFW after Emergency Condensate Storage Tank depletes) and Bleed & Feed U2 = 2.lE-02 fails. The dominant failures are:

(7%)

• Seismic failure of the 120VAC vital buses that power critical instrumentation (which .fails HEPs for long term cooling and Bleed & Feed)

• Failure of the MCR panels or process cabinets, which also fails HEPs

• Seismic failure of MCCs that power the MOVs for High Head SI and pressurizer PORVs

• Seismic failure of SW pumphouse or SW reservoir, which fails SW

• Seismic failure of process cabinets, which fails actuation signals and critical instrumentation Sequence Ux-LOOP-SEIS-03 Ul = 1.3E-02 Small LOCA (2" break) with successful AFW but with failure of the (4%) High Head SI injection. The dominant failures are:

U2 = 1.3E-02

• Chatter ofthe HHSI pump lockout relays results in failure of high head safety injection (5%)

• Chatter of the Service Water lockout relays results in failure of cooling to the HHSI pumps

• Seismic failure of the RWST

• Seismic failure of the Component Cooling heat exchangers results in a flood that fails the HHSI pumps Sequence Ux-SLOCA-SEIS-04 Page 45 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Summary Report March 2018 Table 5.4-1 Summary of Top SCDF Accident Sequences FV Importance Accident Sequence Description Ul =1.lE-02 Small LqCA (2" break) with successful AFW and High Head SI but (4%) with failure of the High Head SI recirculation when the RWST is depleted. The dominant failures are:

U2 =1.2E-02 (4%)

• Chatter of the Service Water lockout relays results in failure of containment sump cooling

• Chatter ofthe Low Head SI pump lockout relays results in failure of the LHSI pumps

• Failure of the Safeguards area ventilation due to seismic failure of the upper levels of the Auxiliary building, which fails the Safeguards area fans; Failure of the Safeguards are ventilation fails the LHSI pumps

• Chatter of relays in the Recirculation Spray (RS) pumps causing them to pre-maturely start before the containment sump contains water. Failure of the RS results in failure of containment sump recirculation since the pumps are required for sump cooling.

Sequence Ux-SLOCA-SEIS-01 Page 46 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTIF 2.1 Seismic PRA Summary Report March 2018 Table 5.4-2 SCDF Importance Measures Ranked by FV Failure Fragility Fragility ~roups Fragility Group Description Ul CDF FV U2 CDF FV Am Br Bu Mode Method EPRI Report SE IS-LOOP SEISMIC-INDUCED LOSS OF OFFSITE POWER 6.91E-01 6.90E-01 0.30 0.27 0.40 Generic [10]

HCLPF is set SEIS-SSLOCA SEISMIC-INDUCED SMALL-SMALL LOCA 9.51E-02 1.02E-01 0.30 0.28 0.28 Generic to SSE [10]

SEIS-EE-BKR-HJ8-RLY 4KV to 480V BUS BREAKERS - RELAY CHATTER 6.76E-02 6.90E-02 0.52 0.24 0.52 Functional sov SEIS-SW-P-lAB-RLY SERVICE WATER PUMPS- RELAY CHATTER 3.84E-02 3.96E-02 0.77 0.24 0.49 Functional sov SEIS-CH-P-lABC-RLY CHARGING PUMPS - RELAY CHATTER 3.63E-02 3.75E-02 0.77 0.24 0.49 Functional sov EPRI Report SEIS-SLOCA SEISMIC-INDUCED SMALL LOCA 3.33E-02 3.37E-02 1.00 0.30 0.40 Generic [10]

SEIS-VB-INV-1234 120 VAC VITAL BUS INVERTERS 3.26E-02 3.23E-02 1.10 0.19 0.58 Functional sov SEIS-SI-P-lAB-RLY LOW HEAD SI PUMP - RELAY CHATTER 2.83E-02 2.83E-02 0.77 0.24 0.49 Functional sov SEIS-FW-P-3AB-RLY MOTOR-DRIVEN AFW PUMPS - RELAY CHATTER 2.65E-02 2.65E-02 0.77 0.24 0.49 Functional sov SEIS-EE-BKR-HJ2-RLY EDG OUTPUT BREAKERS - RELAY l.90E-02 l.94E-02 0.77 0.24 0.49 Functional sov CDFM SEIS-EP-CB-12ABCD 125 VDC DISTRIBUTION PANELS l.46E-02 l.48E-02 1.15 0.24 0.38 Functional Hybrid CDFM SE IS-E P-CB-4ABCD 120 VAC VITAL BUS DISTRIBUTION PANELS l.40E-02 l.41E-02 1.16 0.24 0.38 Anchorage Hybrid EMERGENCY DIESEL GENERATORS - RELAY SEIS-EDG-HJ-RLY CHATTER l.09E-02 l.09E-02 0.70 0.24 0.83 Functional sov Structural failure of CDFM SEIS-BY-B-1-24 STATION BATTERIES 1-11 AND 1-IV 8.53E-03 8.38E-03 1.14 0.24 0.38 rack Hybrid CDFM SEIS-EI-CB-MCR-PNL SEISMIC FAILURE OF MCR BOARDS AND PANELS 7.55E-03 7.GlE-03 1.30 0.24 0.38 Functional Hybrid Page 47 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Summary Report March 2018 The most significant non-seismic SSC failures (e.g., random failures of modeled components during the SPRA mission time) are listed in Table 5.4-3. The unavailability of the diesel-driven fire pump and FLEX equipment (pumps and generators) constitutes the highest FV importance for SCDF. These SSCs support mitigation of a SBO.

Table 5.4-3 SCDF Importance Measures Ranked by FV for Non-Seismic Failures Unit 1 Model Basic Events Prob SCDFFV Description DIESEL-DRIVEN FIRE PUMP 1-FP-P-2 OUT OF SERVICE lFP-DDP--TM-2 3.16E-02 l.33E-02 FOR TEST OR MAINTENANCE OBDBEDG--FR-lA-FLEX 2.04E-02 7.0SE-03 FLEX DIESEL GENERATOR FAILS TO RUN FLEX RCS INJECTION PUMP (OO-BDB-P-3A) FAILS TO OBDBDDP--FS-3A-FLEX S.46E-03 1.84E-03 START OBDBEDG--FS-lA-FLEX 4.53E-03 l.53E-03 FLEX DIESEL GENERATOR FAILS TO START Ul TURBINE-DRIVEN AFW PUMP OUT OF SERVICE lFW-TRB--TM-2 2.81E-03 l.43E-03 FOR TEST OR MAINTENANCE lFW-TRB--FS-2 l.92E-03 9.63E-04 Ul TURBINE-DRIVEN AFW PUMP FAILS TO START OBDBEDG--FL-lA-FLEX 2.90E-03 9.59E-04 FLEX DIESEL GENERATOR FAILS TO LOAD lFW-TRB--FR-2 l.71E-03 8.SSE-04 Ul TURBINE-DRIVEN AFW PUMP FAILS TO RUN lFP-DDP--FR-2 2.13E-03 8.33E-04 DIESEL-DRIVEN FIRE PUMP 1-FP-P-2 FAILS TO RUN FLEX RCS INJECTION PUMP (OO-BDB-P-3A) FAILS TO OBDBDDP--FR-3A-FLEX 2.28E-03 7.37E-04 RUN Unit 2 Model Basic Events and FV Importance DIESEL-DRIVEN FIRE PUMP 1-FP-P-2 OUT OF SERVICE lFP-DDP--TM-2 3.16E-02 l.32E-02 FOR TEST OR MAINTENANCE OBDBEDG--FR-lA-FLEX 2.04E-02 6.88E-03 FLEX DIESEL GENERATOR FAILS TO RUN FLEX RCS INJECTION PUMP (OO-BDB-P-3A) FAILS TO OBDBDDP--FS-3A-FLEX S.46E-03 l.79E-03 START OBDBEDG--FS-lA-FLEX 4.53E-03 l.49E-03 FLEX DIESEL GENERATOR FAILS TO START U2 TURBINE-DRIVEN AFW PUMP OUT OF SERVICE 2FW-TRB--TM-2 2.81E-03 1.43E-03 FOR TEST OR MAINTENANCE 2FW-TRB--FS-2 l.92E-03 9.61E-04 U2 TURBINE-DRIVEN AFW PUMP FAILS TO START OBDBEDG--FL-lA-FLEX 2.90E-03 9.41E-04 FLEX DIESEL GENERATOR FAILS TO LOAD 2FW-TRB--FR-2 1.71E-03 8.53E-04 U2 TURBINE-DRIVEN AFW PUMP FAILS TO RUN lFP-DDP--FR-2 2.13E-03 8.24E-04 DIESEL-DRIVEN FIRE PUMP 1-FP-P-2 FAILS TO RUN FLEX RCS INJECTION PUMP (OO-BDB-P-3A) FAILS TO OBDBDDP--FR-3A-FLEX 2.28E-03 7.24E-04 RUN A summary of the SCDF results for each seismic hazard interval is presented in Table 5.4-

4. Figure 5.4-1 shows a bar chart of the Unit 1 SCDF as a function of PGA (Unit 2 results Page 48 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Summary Report March 2018 are the same as Unit 1). The seismic ground motions that contribute the most to SCDF are in the O.Sg to 1.0g range (%G04 - %G07). The small increase in SCDF contribution for the %G07 and %G08 intervals is due to the width of the intervals being larger than the lower intervals.

5.4-4 Contribution to SCDF by Acceleration Interval Initiator  % Total Ul  % Total U2 PGA Frequency UlCDF Ul CDF CCDP U2 CDF U2CDF CCDP

%G01 0.06g to <0.3g 9.21E-04 6.87E-08 0.1% 0.00 6.87E-08 0.1% 0.00

%G02 0.3g to <0.4g 5.34E-05 3.16E-06 5.3% 0.06 3.12E-06 5.2% 0.06

%G03 0.4g to <0.Sg 3.0lE-05 7.lOE-06 11.8% 0.24 6.99E-06 11.7% 0.23

%G04 O.Sg to <0.6g 1.79E-OS 1.06E-05 17.7% 0.59 1.06E-05 17.7% 0.59

%GOS 0.6g to <0.7g 1.llE-05 9.30E-06 15.5% 0.84 9.30E-06 15.5% 0.84

%G06 0.7g to <0.8g 7.08E-06 6.74E-06 11.2% 0.95 6.74E-06 11.3% 0.95

%G07 0.8g to <lg 8.26E-06 8.19E-06 13.7% 0.99 8.19E-06 13.7% 0.99

%GOS lg to <1.Sg 9.09E-06 9.09E-06 15.2% 1.00 9.09E-06 15.2% 1.00

%G09 1.Sg to <2.Sg 4.25E-06 4.25E-06 7.1% 1.00 4.25E-06 7.1% 1.00

%G10 >2.Sg 1.48E-06 1.48E-06 2.5% 1.00 1.48E-06 2.5% 1.00

• * * *** * *** * ** **** Unit 1 seismic tot= con1:r1b1.1tions

<<=**pt;"A_"==*-~~*~~<<<=*

0.06gto 03gto 0.4gto O.Sgto 0.6gto 0.7gto O.Bgto lgto 1..Sgto >2.5g

<0.3g <OAg -~0.5g <0.6g <0.7g <O.Sg <lg -~1.Sg <2.5g Ground Acceleration (PGA)

Figure 5.4-1 Unit 1 SCDF Contributions by PGA Page 49 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Summary Report March 2018 The most significant Operator actions modeled as Human Error Probability (HEPs) in the model are listed in Table 5.4-5. As discussed in Section 5.1, the seismic PRA models each Operator action using four HEP basic events per action, which model different failure probabilities for the higher ground motions. The FV importance of the Operator action is the sum of the FV importance for each of the four HEP basic events. The important actions involve restoring alternate cooling to the Charging pumps upon loss of normal SW cooling, aligning the turbine-driven AFW pump to all three SGs during a SBO, and installing and starting the FLEX RCS injection pump during a SBO.

5.4 SCDF Importance Measures Ranked by FV for Operator Actions HEP Basic Event SCDFFV Description Restore Cooling to the Charging Pumps from Fire HEP-C-OSW-CHP-ALT 6.61E-02 Protection or Primary Grade Water systems HEP-C-ALIG N-TDAFW 2.66E-02 Align turbine-driven AFW Pump to the Band C SGs HEP-C-FLEX-RIP l.55E-02 Install and Start FLEX RCS Injection Pump HEP-C-FLEX-LOADSHED 7.87E-03 Load shed the vital 125vdc batteries during SBO Open 1-SI-MOV-1836 to Align Alternate Flow Path for HEP-C-1SI-OPN1836 6.71E-03 HHSI Align SW OR Fire Protection Water to AFW Pumps When HEP-C-lFW-AFWSPLY 5.50E-03 ECST Depletes HEP-C-FLEX-VAC 5.36E-03 Install FLEX Generator to Power Vital Buses Isolate SW Flood in Auxiliary Building Caused by Seismic REC-SEIS-FLD-CCHX 5.llE-03 Failure of the CCW Heat Exchangers HEP SCDF FV Importance in Unit 2 Model Restore Cooling to the Charging Pumps from Fire HEP-C-OSW-CH P-ALT 6.90E-02 Protection or Primary Grade Water systems HEP-C-ALIG N-TDAFW 2.59E-02 Align TDAFW Pump to the B and C SGs HEP-C-FLEX-RIP l.51E-02 Install and Start FLEX RCS Injection Pump HEP-C-FLEX-LOADSHED 7.67E-03 Load shed the vital 125vdc batteries during SBO Open 2-SI-MOV-2836 to Align Alternate Flow Path for HEP-C-2SI-OPN2836 6.82E-03 HHSI Align SW or Fire Protection Water to AFW Pumps When HEP-C-2FW-AFWSPLY 5.57E-03 ECST Depletes HEP-C-FLEX-VAC 5.19E-03 Install FLEX Generator to Power Vital Buses Isolate SW Flood in Auxiliary Building Caused by Seismic REC-SEIS-FLD-CCHX 5.19E-03 Failure of the Component Cooling Heat Exchangers Page 50 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Summary Report March 2018 5.5 SLERF Results This section presents the seismic large early release frequency (SLERF) results, a list of the SSCs that are significant contributors, including risk importance measures, and a discussion of significant sequences and their relativ~ SLERF contributions.

The seismic PRA performed for NAPS shows that the point estimate seismic LERF is 1.6x10-5 for both Unit 1 and Unit 2. A discussion of the mean SLERF with uncertainty distribution reflecting the uncert.ainties in the hazard, fragilities, and model data is presented in Section 5.6. Important contributors are discussed in the following paragraphs.

The top SLERF accident sequences based on FV importance of the sequence flags are documented in Table 5.5-1. These sequences contribute over 80% of the SLERF. Note that these sequences have been combined across all the hazard bin intervals. Three of the top seven sequences are seismic events with a loss of offsite power and failure of the EDGs due to relay chatter resulting in a Station Blackout (SBO).

These core damage sequences progress to a release generally due to temperature-induced steam generator tube rupture caused by a loss of AFW which results in dry out of the SGs. Some ofthe sequences where there is a loss of containment sump cooling, a release occurs due to containment failure caused by containment overpressurization upon loss of heat removal from the sump.

Table 5.5-1 Summary of Top SLERF Accident Sequences FV Importance Accident Sequence Description Ul =7.35E-02 580 with successful AFW (i.e. Turbine-driven AFW pump) but (26.8%) either Long Term Cooling fails, Cooldown and Depressurization fails, or the SI Accumulators fail. The dominant failures are:

U2 =7.24E-02 (25.9%)

• SBO caused mainly by relay chatter of EDG output breaker or 4kv breaker supply to the 480V buses and MCCs; no credit for Operators recovery of the relay chatter.

• Seismic failure of the SG PO RVs

• Seismic failure of the 120VAC vital buses, DC buses and inverters that power critical instrument transmitters required for Operator actions

• Seismic failure of Main Control Room panels Sequence Ux-SBO-SEIS-02 Page 51 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTIF 2.1 Seismic PRA Summary Report March 2018 Table 5.5-1 Summary of Top SLERF Accident Sequences FV Importance Accident Sequence Description Ul = 4.37E-02 SBO with successful AFW and Long Term Cooling, but FLEX (15.9%) mitigation fails due to the following:

• Seismic failure of the RWST which fails RCS makeup from the U2 = 4.73E-02 (16.9%) FLEX RCS Injection Pump

• Seismic failure of the vital 125vdc batteries resulting in loss of critical instrumentation

• Seismic failure of the FLEX electrical distribution panel Sequence Ux-SBO-SEIS-01 Ul = 3.90E-02 SBO with failures that go directly to core damage due to (14.2%) insufficient time to mitigate (large, medium, small LOCAs, ATWS). Dominant failures that result in direct core damage are:

U2 = 4.16E-02 (14.9%)

• Small LOCA

• Control Rods Sequence Ux-SBO-SEIS-04 Ul = 2.09E-02 Small LOCA (2" break) with successful AFW and High Head SI but (7.6%) with failure of the High Head SI recirculation when the RWST is depleted. The dominant failures are:

U2 = 2.27E-02

• Failure of the Service Water MOVs in the Quench Spray (8.1%)

Pumphouse that need to open to provide cooling to the Recirculation Spray Heat Exchangers for containment sump cooling

• Chatter of relays in the Recirculation Spray (RS) pumps causing them to pre-maturely start before the containment sump contains water. Failure of the RS pumps result in failure of containment sump recirculation since the pumps are required for sump cooling.

• Sump recirculation fails due to failure of containment heat removal resulting in containment failure prior to core damage. Contributes to LERF with conditional probability of 1.0 since containment is open at the time of core damage.

Sequence Ux-SLOCA-SEIS-01 Page 52 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Summary Report March 2018 Table 5.5-1 Summary of Top SLERF Accident Sequences FV Importance Accident Sequence Description Ul = 1.86E-02 Loss of Offsite Power with successful AFW but long term cooling

{6.8%) fails (i.e. align Service Water or Fire Protection to AFW after Emergency Condensate Storage Tank depletes) and Bleed &

U2 = 2.05E-02 Feed fails. The dominant failures are:

(7.3%)

• Seismic failure of the 120VAC vital buses that power critical instrumentation (which fails HEPs for long term cooling and Bleed & Feed)

• Failure of the MCR panels or process cabinets, which also fails HEPs

• Operator actions to align an alternate source of water to AFW

• Seismic failure of the RWST

• Failure of the Safeguards area ventilation due to seismic failure of the upper levels of the Auxiliary building, which fails the Safeguards area fans; Failure of the Safeguards are ventilation fails the LHSI pumps Sequence Ux-LOOP-SEIS-03 Ul = 1.GlE-02 Loss of Offsite Power with failure of AFW and failure of Bleed &

(5.9%) Feed. The dominant failures are:

• Seismic failure of the turbine-driven AFW pump and relay U2 = 1.31E-02 chatter of the motor-driven AFW pumps (4.7%)

• Failure of the Safeguards area ventilation due to seismic failure of the upper levels of the Auxiliary building, which fails the Safeguards area fans; Failure of the Safeguards are ventilation fails the LHSI pumps

• Chatter of the HHSI pump lockout relays

• Failure ofthe Operator action to establish Bleed and Feed.

Sequence Ux-LOOP-SEIS-05 Ul = 1.42E-02 Seismic event causes damage to the reactor containment (5.2%) building resulting in failure of the RCS, core damage and large early release.

U2 = 1.49E-02 (5.3%) Sequence Ux-DMG-SEIS-05 SSCs with the most significant seismic failure contributions to SLERF are listed in Table 5.5-2, sorted by FV importance. The seismic fragilities for each of the Page 53 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NITF 2.1 Seismic PRA Summary Report March 2018 significant contributors are also provided in Table 5.5-2, along with the corresponding limiting seismic failure mode and method of fragility calculation.

Importance analyses were performed for SLERF using the ACUBE code. From the ACUBE output, FV values for the seismic failures (i.e. fragility groups) is the sum ofthe FV values for each hazard interval.

The FV listing shows the top individual contributor to SLERF as seismically induced Loss of Offsite Power (LOOP), due to the low median seismic capacity assumed for offsite power failure following a seismic event. The fragility for LOOP is a value from SPRA Implementation Guide (10] and considered reasonably representative for NAPS.

The next highest contributor is seismically induced small LOCA (SLOCA), which has a relatively low median capacity, and is based on the SPRA Implementation Guide (10]. The relay chatter failures of the Recirculation Spray pumps (needed for sump cooling) and the AFW pumps show up in these SLOCA cutsets, where these relays have relatively low capacities.

The next highest contributor to SLERF is seismic failure of the containment building, which is assumed to result in direct core damage as well as direct LERF'.

Other top contributors to SLERF are failures that fail containment sump cooling such as relay chatter of the RS pumps, seismic failure of the four RS heat exchangers as well as seismic failure of the Service Water MOVs in the Quench Spray Pumphouse basement that have to open to provide cooling to the RS heat exchangers. There are also a number of other seismic failures that have SLERF FV values greater than 0.005 that are in SBO, LOOP and SLOCA sequences.

Page 54 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTIF 2.1 Seismic PRA Summary Report March 2018 Table 5.5-2 SLERF Importance Measures Ranked by FV Fragility Fragility Groups Fragility Group Description Ul LERF FV U2 LERF FV Am Br Bu Failure Mode Method EPRI Report SE IS-LOOP SEISMIC-INDUCED LOSS OF OFFSITE POWER 5.0lE-01 5.05E-01 0.30 0.27 0.40 Generic [10]

EPRI Report SEIS-SLOCA SEISMIC-INDUCED SMALL LOCA 9.07E-02 9.19E-02 1.00 0.30 0.40 Generic [10]

SEIS-RS-P-lAB-RLY INSIDE RS PUMP - RELAY CHATIER 5.46E-02 5.25E-02 1.37 0.23 0.48 Functional sov SE IS-BLDG-RC REACTOR CONTAINMENT BUILDING 4.40E-02 4.31E-02 1.71 0.24 0.26 Structural CDFM Outside RS Pumps Spuriously Start due to Relay SEIS-RS-P-2AB-RLYSS Chatter 2.94E-02 2.81E-02 1.37 0.23 0.48 Functional sov SEIS-FW-P-3AB-RLY MOTOR-DRIVEN AFW PUMPS- RELAY CHATIER 2.43E-02 1.80E-02 0.77 0.24 0.49 Functional sov SEIS-RS-P-2AB OUTSIDE RECIRC SPRAY PUMPS 2.29E-02 2.18E-02 1.38 0.24 0.32 Anchorage CDFM SEIS-FW-P-2 TURBINE-DRIVEN AUXILIARY FEEDWATER PUMP 2.26E-02 2.36E-02 1.60 0.24 0.32 Functional CDFM SEIS-EE-BKR-HJ8-RLY 4KV TO 480V BUS BREAKERS - RELAY CHATIER 2.18E-02 2.20E-02 0.52 0.24 0.52 Functional sov SEIS-RS-E-lABCD RECIRC SPRAY HEAT EXCHANGERS 1.58E-02 1.50E-02 2.01 0.24 0.32 Structural CDFM SEIS-EI-CB-MCR-PNL SEISMIC FAILURE OF MCR BOARDS AND PANELS 1.42E~02 1.62E-02 1.30 0.24 0.38 Functional CDFM SE IS-BLDG-AB- Shear Wall LOWER AUX BLDG LOWER FLOORS FAIL 1.42E-02 1.39E-02 2.05 0.24 0.26 Failure CDFM SEIS-MS-TV-111AB MAIN STEAM TRIP VALVE TO TURBINE DRIVEN 1.81 SEIS-MS-TV-211AB AFW PUMP 1.39E-02 3.18E-03 2.51 0.24 0.32 Functional CDFM EPRI Report SEIS-SSLOCA SEISMIC-INDUCED SMALL-SMALL LOCA 1.37E-02 1.46E-02 0.30 0.28 0.28 Generic [10]

Page 55 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Summary Report March 2018 Table 5.5-2 SLERf Importance Measures Ranked by FV Fragility Fragility Groups Fragility Group Description Ul LERF FV U2 LERF FV Am Br Bu Failure Mode Method EPRI Report SEIS-LLOCA LARGE LOCA 1.34E-02 1.27E-02 2.50 0.30 0.40 Generic (10]

SEIS-EG-B-3 EDG lJ Battery 1.32E-02 4.07E-03 1.15 0.24 0.38 Functional CDFM SEIS-EG-P-lJ EDG lJ Fuel Oil Transfer Pumps l.29E-02 3.99E-03 1.16 0.24 0.38 Functional CDFM SEIS-MOV-QSPH- MOVs in QUENCH SPRAY PUMP HOUSE -SW RSHX Cooling to RS HXs l.19E-02 l.13E-02 2.13 0.24 0.32 Functional CDFM SEIS-VB-INV-1234 120 VAC VITAL BUS INVERTERS l.19E-02 l.35E-02 1.10 0.19 0.58 Functional sov SEIS-EP-CB-4ABCD 120 VAC VITAL BUS DISTRIBUTION PANELS l.14E-02 l.28E-02 1.16 0.24 0.38 Anchorage CDFM EPRI Report SEIS-MLOCA MEDIUM LOCA l.13E-02 l.22E-02 2.00 0.35 0.45 Generic (10]

SEIS-EE-BKR-HJ2-RLY EDG OUTPUT BREAKERS - RELAY l.05E-02 1.02E-02 0.77 0.24 0.49 Functional sov Tank SEIS-QS-TK-1 REFUELING WATER STORAGE TANK (RWST) 9.77E-03 l.lOE-02 1.07 0.15 0.29 Overturning sov Failure of Fuel SEIS-RC-CNTRL- Hold Down RODS 'REACTOR CONTROL RODS 9.68E-03 l.09E-02 1.26 0.24 0.32 Spring CDFM EMERGENCY DIESEL GENERATOR CONTROL SEIS-EI-CB-202 PANELS IN ESGR - Fails EDGs 8.83E-03 l.13E-02 1.40 0.24 0.38 Functional CDFM SEIS-EP-SS-1Hl-1Jl 480V LOAD CONTROL CENTERS lHl AND lJl 8.16E-03 l.20E-02 1.22 0.24 0.38 Functional CDFM EMERGENCY DIESEL GENERATOR CONTROL SEIS-EI-CB-201 PANELS IN EDG ROOM - Fails EDGs 7.98E-03 9.96E-03 1.45 0.24 0.38 Anchorage CDFM SEIS-CH-P-lABC-RLY CHARGING PUMPS- RELAY CHATIER 7.47E-03 7.30E-03 0.77 0.24 0.49 Functional sov SE IS-CV-TV-150ABCD Containment Vacuum Isolation Trip Valves 7.llE-03 6.96E-03 2.51 0.24 0.32 Functional CDFM Page 56 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Summary Report March 2018 Table 5.5-2 SLERF Importance Measures Ranked by FV Fragility Fragility Groups Fragility Group Description Ul LERF FV UZ LERF FV Am Br Bu Failure Mode Method EMERGENCY DIESEL GENERATORS - RELAY SEIS-EDG-HJ-RLY CHATIER 7.08E-03 7.13E-03 0.70 0.24 0.83 Functional sov SEIS-EP-SS-1H-1J 480V LOAD CONTROL CENTERS 1H AND lJ 6.87E-03 1.75E-02 0.99 0.24 0.38 Functional CDFM Structural SEIS-BY-B-1-24 STATION BATIERIES 1-11 AND 1-IV 6.70E-03 8.21E-03 1.14 0.24 0.38 Failure of Rack CDFM SEIS-SW-P-lAB-RLY SERVICE WATER PUMPS - RELAY CHATIER 6.21E-03 5.97E-03 0.77 0.24 0.49 Functional sov Combined Structural/

SEIS-EP-SW-1H-1J 4160V EMERGENCY BUSES 6.19E-03 5.85E-03 1.13 0.24 0.33 Function CDFM SEIS-EG-B-4 EDG 2J Battery 6.09E-03 1.71E-02 0.97 0.24 0.38 Functional CDFM SEIS-EG-P-2J EDG 2J Fuel Oil Transfer Pumps 5.78E-03 1.64E-02 1.00 0.24 0.38 Functional CDFM Combined Structural/

SEIS-EG-B-1 EDG 1H Battery 5.69E-03 2.24E-03 1.49 0.24 0.33 Function CDFM SEIS-FW-P-3AB MOTOR-DRIVEN AUXILIARY FEEDWATER PUMPS 5.67E-03 4.06E-03 1.62 0.24 0.32 Functional CDFM SE IS-BLDG-AB- Failure of Steel UPPER AUX BLDG UPPER FLOORS FAIL 5.40E-03 5.04E-03 1.02 0.24 0.26 Superstructure CDFM SEIS-RS-P-2AB- Outside RS Pumps Fail to Start due to Lockout RLYLO Relay 5.09E-03 4.64E-03 0.77 0.24 0.49 Functional sov SEIS-EI-CB-PROCESS PLANT PROCESS CABINETS 4.78E-03 5.19E-03 1.91 0.19 0.55 Functional sov BEYOND DESIGN BASIS (FLEX) DISTRIBUTION Seismic SEIS-BDB-DB-123 PANELS 4.lOE-03 5.71E-03 1.10 0.24 0.26 Interaction CDFM Combined Structural/

SEIS-EG-B-2 EDG 2H Battery 3.73E-03 1.21E~02 1.22 0.24 0.33 Function CDFM SEIS-EG-P-2H EDG 2H Fuel Oil Transfer Pumps 3.03E-03 1.0SE-02 1.40 0.24 0.38 Functional CDFM Page 57 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTIF 2.1 Seismic PRA Summary Report March 2018 The most significant non-seismic SSC failures (e.g., random failures of modeled components during the SPRA mission time) are listed in Table 5.5-3. The unavailability of the diesel-driven fire pump has the highest FV for both Units 1 and 2. As noted in the important non-seismic failures for SCDF, the diesel-driven fire pump is important for long term supply to the turbine-driven AFW when the ECST depletes during a SBO. The other non-seismic failures that are important for SLERF are SW supply headers and SW pumps that fail cooling to the RS heat exchangers and thus fails containment heat removal. The EDGs and FLEX equipment are also important for mitigating SBO sequences.

Table 5.5-3 SLERF Importance Measures Ranked by FV for Non-Seismic Failures Unit 1 Model Basic Events Prob SLERF FV Description DIESEL-DRIVEN FIRE PUMP 1-FP-P-2 OUT OF SERVICE lFP-DDP--TM-2 3.16E-02 5.14E-03 FOR TEST OR MAINTENANCE B SW HEADER IN OUT OF SERVICE FOR TEST OR OSW-HDR--TM-B 1.52E-02 2.95E-03 MAINTENANCE OBDBEDG--FR-lA-FLEK 2.04E-02 2.73E-03 FLEX DIESEL GENERATOR FAILS TO RUN A SW HEADER IN OUT OF SERVICE FOR TEST OR OSW-HDR--TM-A 1.52E-02 1.99E-03 MAINTENANCE Ul 18 SW PUMP OUT OF SERVICE FOR TEST OR 1SW-PAT~-TM-1B 8.SSE-03 1.43E-03 MAINTENANCE 1EE-EDG--FR-1H 2.79E-02 1.29E-03 Ul H DIESEL GENERATOR FAILS TO RUN 1EE-EDG--FR-1J 2.79E-02 1.14E-03 Ul J DIESEL GENERATOR FAILS TO RUN Ul lA SW PUMP OUT OF SERVICE FOR TEST OR 1SW-PAT--TM-1A 8.SSE-03 1.0lE-03 MAINTENANCE Ul H DIESEL GENERATOR OUT OF SERVICE FOR TEST 1EE-EDG--TM-1H 2.25E-02 9.13E-04 OR MAINTENANCE Ul J DIESEL GENERATOR OUT OF SERVICE FOR TEST 1EE-EDG--TM-1J 2.25E-02 8.65E-04 OR MAINTENANCE Unit 2 Model Basic Events and FV Importance DIESEL-DRIVEN FIRE PUMP 1-FP-P-2 OUT OF SERVICE lFP-DDP--TM-2 3.16E-02 5.25E-03 FOR TEST OR MAINTENANCE OBDBEDG--FR-lA-FLEX 2.04E-02 2.67E-03 FLEX DIESEL GENERATOR FAILS TO RUN A SW HEADER IN OUT OF SERVICE FOR TEST OR OSW-HDR--TM-A 1.52E-02 2.62E-03 MAINTENANCE 2EE-EDG--FR-2H 2.79E-02 2.31E-03 U2 H DIESEL GENERATOR FAILS TO RUN U2 H DIESEL GENERATOR OUT Of SERVICE FOR TEST 2EE-EDG--TM-2H 2.25E-02 1.79E-03 OR MAINTENANCE OSW-H DR--TM-B 1.52E-02 1.SSE-03 B SW HEADER IN OUT OF SERVICE FOR TEST OR Page 58 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Summary Report March 2018 Table 5.5-3 SLERF Importance Measures Ranked by FV for Non-Seismic Failures Unit 1 Model Basic Events Prob SLERF FV Description MAINTENANCE 2EE-EDG--FR-2J 2.79E-02 1.27E-03 U2 J DIESEL GENERATOR FAILS TO RUN U2 1B SW PUMP OUT OF SERVICE FOR TEST OR 2SW-PAT--TM-1B 8.SSE-03 1.12E-03 MAINTENANCE U2 J DIESEL GENERATOR OUT OF SERVICE FOR TEST 2EE-EDG--TM-2J 2.25E-02 9.71E-04 OR MAINTENANCE 2QS-PSB--FS-1A 5.59E-03 8.63E-04 U2 lA QS PUMP FAILS TO START A summary of the SLERF results for each seismic hazard interval is presented in Table 5.5-4. Figure 5.5-1 shows a bar chart of the unit 1 SLERF as a function of PGA (Unit 2 results are the same as Unit 1). The seismic ground motions that contribute the most to SLERF are in the 1.0 to 2.5g range (%GOS and %G09) which is generally the case in SPRA LERF results.

5.5-4 Contribution to SLERF by Acceleration Interval Initiator  % Total Ul  % Total U2 Ul LERF U2 LERF PGA Frequency Ul LERF CLERP U2 LERF CLERP

%G01 0.06g to <0.3g 9.21E-04 7.58E-10 0.00% 0.00 7.58E-10 0.00% 0.00

%G02 0.3g to <0.4g 5.34E-05 5.77E-08 0.37% 0.00 5.69E-08 0.36% 0.00

%G03 0.4g to <0.Sg 3.0lE-05 1.35E-07 0.87% 0.00 1.31E-07 0.84% 0.00

%G04 O.Sg to <0.6g 1.79E-05 2.75E-07 1.77% 0.02 2.68E-07 1.72% 0.01

%GOS 0.6g to <0.7g 1.llE-05 3.71E-07 2.39% 0.03 3.65E-07 2.34% 0.03

%GOG 0. 7g to <0.8g 7.08E-06 4.96E-07 3.19% 0.07 5.03E-07 3.23% 0.07

%G07 0.8g to <lg 8.26E-06 1.60E-06 10.29% 0.19 1.63E-06 10.45% 0.20

%GOS lg to <1.Sg 9.09E-06 6.89E-06 44.29% 0.76 6.91E-06 44.31% 0.76

%G09 1.Sg to <2.Sg 4.25E-06 4.25E-06 27.32% 1.00 4.25E-06 27.25% 1.00

%G10 >2.Sg 1.48E-06 1.48E-06 9.51% 1.00 1.48E-06 9.49% 1.00 Page 59 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Summary Report March 2018 45.0%

,, . . .,.Un.it 1 Seismic . lERFContributions....... ,......." . ...

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<0.3g <0.4g <O.Sg <O.&g <U7g <0.8g <1.g <LSg <2.Sg Ground Acceleration {PGA)

Figure 5.5-1 Unit 1 SLERF Contributions by PGA The most significant Operator actions for SLERF are listed in Table 5.5-5. As discussed in Section 5.1, the seismic PRA models each Operator action using four Human Error Probability (HEP) basic events per action, which model different failure probabilities for the four damage states. The FV importance of the Operator action is the sum of the FV importance for each of the four HEP basic events. The important actions involve depressurizing the RCS after core damage per the SAMGs. Other important actions are mainly important for mitigating core damage, such as aligning the turbine-driven AFW pump to the other SGs, initiating Bleed and Feed, and performing FLEX mitigating actions (battery load shed and installing RCS injection pump).

5.5 SLERF Importance Measures Ranked by FV for Operator Actions HEP Basic Event SLERF FV Description HEP-C-RCSDEP 2.71E-02 Depressurize the RCS Per SAM Gs Align turbine-driven AFW Pump to Band C HEP-C-ALIG N-TDAFW 2.26E-02 SGs HEP-C-1BAFE 1.17E-02 Initiate Bleed and Feed After AFW Fails HEP-C-FLEX-RIP 8.28E-03 Install and Start FLEX RCS Injection Pump Page 60 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Summary Report March 2018 5.5 SLERF Importance Measures Ranked by FV for Operator Actions HEP Basic Event SLERF FV Description HEP-C-lHV-SFGD-VENT 6.77E-03 Restore Safeguards Area Ventilation Isolate SW Flood in Auxiliary Building Caused by Failure of the Component REC-SEIS-FLD-CCHX 5.92E-03 Cooling Heat Exchangers Load shed the vital 12Svdc batteries during HEP-C-FLEX-LOADSH ED 5.23E-03 SBO HEP SLERF FV Importance in Unit 2 Model HEP-C-RCSDEP 2.56E-02 Depressurize the RCS Per SAM Gs Align turbine-driven AFW Pump to Band C HEP-C-ALIGN-TDAFW 2.34E-02 SGs HEP-C-2BAFE 9.27E-03 Initiate Bleed and Feed After AFW Fails HEP-C-FLEX-RIP 8.31E-03 Install and Start FLEX RCS Injection Pump HEP-C-2HV-SFGD-VENT 6.73E-03 Restore Safeguards Area Ventilation Isolate SW Flood in Auxiliary Building Caused by Failure of the Component REC-SEIS-FLD-CCHX 5.72E-03 Cooling Heat Exchangers Load shed the vital 12Svdc batteries during HEP-C-FLEX-LOADSHED 5.52E-03 SBO 5.6 SPRA quantification Uncertainty Analysis This section documents the parametric uncertainty analysis and the approach used to identify sources of model uncertainty.

Parametric Uncertainty Parameter uncertainty in seismic PRA results comes from seismic hazard curve uncertainty, the SSC fragility uncertainties, and uncertainties in the human interaction and random failure calculations. SPRA model parameter uncertainty was quantified using the EPRI UNCERT code. The results are provided in Table 5.6-1, and Figures 5.6-1 through 5.6-4 show the curves of cumulative probability and probability density function.

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NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Summary Report March 2018 Table 5.6 Seismic CDF and LERF Uncertainty Distributions Unit 1 CDF Unit 2 CDF Unit 1 LERF Unit 2 LERF Mean 6.32E-05 6.34E-05 1.93E-05 1.94E-05 5th Percentile 1.04E-05 1.03E-05 3.00E-06 2.91E-06 Median 4.32E-05 4.30E-05 1.30E-05 1.28E-05 95th Percentile 1.81E-04 1.84E-04 5.58E-05 5.71E-05 StdDev 6.77E-05 6.83E-05 2.0SE-05 2.24E-05 Skewness 4.4 4.7 3.9 5.5 The UNCERT runs were performed using the Monte Carlo method of sampling and a total of 20,000 samples. Both SCDF and SLERF runs solved 1,000 cutsets using ACUBE.

The distribution for both SCDF and SLERF appears generally uniform. The distribution (i.e. spread between 5th and 95th) for SCDF and SLERF is larger than that of the internal events uncertainty distribution, which is expected and reasonable given relatively large uncertainties in the seismic hazard curves and SSC fragility curves.

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NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Summary Report March 2018 V

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1E-Oi Figure 5.6 Unit 1 Seismic CDF Cumulative and Density Distribution Functions Page 63 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Summary Report March 2018 I

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NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Summary Report March 2018

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11;-ce 11::-os lE-01 Figure 5.6 Unit 1 Seismic LERF Cumula1tive and Density Distribution Functions Page 65 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Summary Report March 2018 0.8-+-----+----+----,f------1------+-----+-------+------!

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1EA02 lE-Ol Figure 5.6 Unit 2 Seismic LERF Cumulative and Density Distribution Functions Model Uncertainty Model uncertainty relates to the uncertainty associated with some aspect of a PRA model that can be represented by any one of several different modeling approaches.

Consequently, uncertainty is introduced into the PRA results since there may not be consensus about which model approach most appropriately represents the particular aspect of the plant being modeled. The uncertainty associated with a model and its constituent parts is typically addressed by making assumptions.

The guidance provided in EPRI reports in references [10] and [19] were used in the identification and characterization *of sources of model uncertainty and related assumptions. A generic list of uncertainty sources for seismic PRAs is contained in Appendix C of reference [19]. In addition to the generic sources of uncertainty, the assumptions made in the NAPS SPRA were also evaluated for sources of uncertainty.

Since many of the assumptions are considered reasonable and consistent with standard industry practices, only the assumptions that may involve a significant source of uncertainty were identified for potential sensitivities studies. The sources of uncertainty that were identified for further evaluation are discussed in the next section (5.7) for sensitivity studies.

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NAPS Units 1 and 2 10 CFR 50.54(f) NTIF 2.1 Seismic PRA Summary Report March 2018 5.7 SPRA Quantification Sensitivity Analysis As discussed in Section 5.6, various sources of model uncertainties were reviewed and examined to identify sources that may have a significant impact on the SCDF and SLERF.

The following sensitivity studies were performed to evaluate how the SCDF and SLERF are impacted by model assumptions, simplifications and uncertainties:

• Model Truncation and Convergence

• Relay Chatter

• Small-Small LOCA

• FLEX Credited in SBO

• Mission Time

• Building HCLPFs

• HEP to Isolate Flood from CC HX Failure

• HEPs at 5th and 95th Percentile

• LERF 5.7.1 Model Truncation and Convergence The baseline SPRA was quantified at lE-09 for SCDF and lE-10 for SLERF. Model convergence per the criteria in the PRA Standard was achieved at these levels.

5.7.2 Relay Chatter As part of the development of the North Anna SPRA, detailed circuit analyses were performed to identify relays that could impact SSC function if chatter occurs. The North Anna SPRA includes over 20 fragility groups that model relay chatter. The HCLPF capacity of many of the relays is relatively low, which results in loss of some key mitigating functions. These relay fragility groups show up as significant contributors to SCDF and SLERF (based on FV importance) partly due to not crediting Operator action to reset the relays to restore the mitigating functions. This sensitivity was performed to determine the reduction in SCDF and SLERF if all relays were assumed to not be vulnerable to chatter. A reduction of approximately 28% was realized in SCDF and 15% reduction for SLERF for both Unit 1 and 2.

The sensitivity shows the SCDF and SLERF results are impacted by the modeling of relay chatter. The HCLPF capacities of the relays are considered reasonable since they were developed using the current industry methods (separation of variables fragility method) for relay fragilities. The SPRA does not credit Operator action to reset the relays mainly Page 67 of 181 .

NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Summary Report March 2018 due to the time required to investigate the lockout condition before restoring the breakers. Therefore, no further refinement to the SPRA was made.

5.7.3 Small-Small LOCA As discussed in the PRA Standard [4] and the EPRI SPRA Implementation Guide [10], the SPRA must consider the potential occurrence of a small-small LOCA (SSLOCA). For NAPS, detailed walkdowns of the RCS were not performed to develop a fragility for the small-bore piping and instrument tubing that connects to the RCS. Therefore, SSLOCA is included in the SPRA with a HCLPF equal to the Safe Shutdown Earthquake of 0.12g per the guidance in the EPRI SPRA Implementation Guide [10). In this sensitivity, the HCLPF was increased to that of the small LOCA (SLOCA) HCLPF, which is 0.32g. The results showed the SCDF decreased by approximately 8% and slightly less than 1% for SLERF for both Units 1 and 2. Even though the SCDF may be reduced by 8%, this is not considered significant enough reduction to warrant changing the model.

5.7.4 FLEX Credited in SBO Sequences The SPRA does not credit recovery of offsite power given the possible damage to the offsite power sources and the likely long repair times to restore power. The SPRA does I

credit FLEX mitigating strategies in the SBO sequences to restore and maintain safety functions to prevent core damage as long as there is sufficient time available to install and start the FLEX equipment. FLEX is not credited for sequences where there is insufficient time to implement the FLEX. equipment before core damage. For example, FLEX is not credited if there is a large, medium, or small LOCA, or a large RCP seal LOCA coincident with the SBO since core damage would occur before the FLEX strategies could be implemented. Also, if the TDAFW pump fails to start and run, FLEX is not credited as there would be insufficient time before SG dryout. The probabilities for the FLEX HEPs and the FLEX equipment failures were identified as sources of uncertainty.

Two sensitivities were performed to assess the impact of these uncertainties.

FLEX HEP Probabilities The SPRA credits FLEX for mitigating seismic-induced SBO. Given the unique nature of the FLEX mitigating strategies as compared with standard actions in the EOPs, the uncertainties associated with the FLEX actions were evaluated. The FLEX mitigating actions modeled are:

• Load shedding the vital 125vdc station batteries to extend battery life

• Installing generators to power vital buses before batteries deplete

• Installing RCS Injection pump to makeup to RCS

• Refuel FLEX engine-driven SSCs Page 68 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Summary Report March 2018 The Human Reliability Analysis (HRA) for these actions followed the guidance in EPRI report 3002008093 [18]. The HRA for these actions did include some judgements with respect to using surrogates for estimating the failure probabilities of actions unique to FLEX strategies, such as transporting the FLEX equipment from the FLEX storage building. This sensitivity evaluates the impact on the SCDF and SLERF if these HEPs were increased by a factor of 5 to account for the selection of different 'commission errors.

The results show the SCDF increased by approximately 7% and SLERF increased by approximately 1% for both units. The use of surrogates for the execution error actually only increases the HEP probability by less than 15% if different (higher) probabilities are used in the HRA. So using a factor of 5 for this sensitivity is considered conservative for assessing the impact of using surrogates. This sensitivity also provides insight that the SPRA risk is not significantly impacted by changes in the FLEX HEP probabilities. No further refinement to the SPRA is considered necessary.

FLEX Equipment Reliability This sensitivity evaluates the impact if the reliability of the FLEX equipment is less than assumed in the SPRA. The following FLEX equipment is credited for maintaining power to the critical instrumentation and for RCS makeup:

• FLEX 120VAC Portable Generator; Used to repower the vital buses to maintain critical instrumentation (0-BDB-GEN-lA)

• Portable RCS Injection Pump; Used to makeup to the RCS {0-BDB-P-3A)

The failure probabilities used for the FLEX equipment are based on similar installed equipment (e.g. EDGs) for now until sufficient reliability data is available for the FLEX equipment. These failure probabilities for the FLEX equipment are not expected to be significantly different than the failure probabilities of the actual portable equipment.

However, since there is uncertainty in these failure probabilities, this sensitivity evaluates the impact if the failure probabilities are increased by a factor of 5. The results show the SCDF increased by less than 5% and SLERF increased by approximately 1% for both units. The FLEX equipment failure probabilities are not expected to increase by a factor of 5. The failure probabilities used in the SPRA are considered reasonable and the SCDF and SLERF are not significantly impacted by changes in the equipment probabilities. Therefore, no further refinement to the SPRA is considered necessary.

5.7.5 Mission Time The mission time assumed in the SPRA is 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. In this sensitivity, the mission time is changed to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. The results show the SCDF and SLERF increased by approximately 2%. Extending the mission time to longer than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> does not have much of an impact on the SCDF and SLERF. The majority of the SSC failures are due to seismic Page 69 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Summary Report March 2018 damage and not random failures of the SSCs during the mission time. No further refinement to the SPRA is considered necessary.

5.7.6 Building HCLPFs The SPRA assumes the HCLPF capacity of buildings represents gross failure of the building such that all SSCs in the building are failed. This is a very conservative assumption, as the reported HCLPF value often corresponds to a local failure, for which the majority of SSCs in the building will survive and not result in a complete or "gross" failure condition. Given the uncertainty in how the buildings fail, this sensitivity evaluates the impact on SCDF and SLERF if the HCLPF capacity of the buildings is increased, which would represent a higher HCLPF capacity that would result in gross failure of the building. The buildings evaluated in this sensitivity are all reinforced concrete, missile protected structures that will be assumed to have a HCLPF capacity of 3.0g in this sensitivity. A HCLPF of 3.0g is selected since it provides a reasonable estimate for the gross failure of the buildings without being overly optimistic. The results show less than 1% decrease in SCDF and approximately 10% decrease in SLERF.

The SCDF is not particularly sensitive to these building failures. SLERF decreased mainly due to the reactor containment, whose failure is direct LERF, and due to failure of the Service Water Pump House and Service Water Valve House, which results in loss of containment heat removal. No further refinement to the SPRA is considered necessary.

5.7.7 Isolating Service Water Flood Seismic failure of the Component Cooling (CC) heat exchangers was determined to result in a Service Water flood that could impact the Charging pumps if the flood was not mitigated in time. This flood scenario is modeled in the SPRA and the HEP for mitigating the flood is set to 1 because of the uncertainty of the size of the flood given that the four heat exchangers are assumed to be 100% correlated.

The SPRA models failure of the four CC heat exchangers in the Auxiliary building as a major flood due to failure of the SW piping that connects to the heat exchangers. There is uncertainty on the size of the flood and the flow rate from the pipe breaks. The model assumes the flood flow rate is large enough such that there is little time available to diagnose and isolate the flood before it damages the Charging pumps given failure of all four heat exchangers (assumed correlated). Therefore, the HEP for isolating the flood is set to 1. In this sensitivity, it was assumed that the flood rate from the SW lines of the four CC heat exchangers is low enough such that there is time available for Operators to isolate the breaks (i.e. the breaks are not complete guillotine breaks, but are splits in the pipe nozzles). The HEP probabilities for the four HRA bins were assumed to vary from 5E-03 to lE-01, which are considered reasonable estimates for isolating lower flow rate floods. The results show very little reduction in SCDF and SLERF, less than 1%, if the seismic failure of the CC heat exchangers is assumed to result in a low Page 70 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Summary Report March 2018 enough SW flood flow rate to allow crediting isolation of the flood. No further refinement to the SPRA is considered necessary.

5.7.8 HEP Probabilities This sensitivity evaluates the impact of the HEPs credited in the SPRA. There is uncertainty in the development of the adjustments to the HEPs in the model to account for the various impacts on the Operators taking mitigating actions after a seismic event.

This sensitivity quantifies the SPRA with the HEPs set to their 95th and to their 5th percentile probabilities. The results show that increasing the HEPs to their 95th percentile results in less than a 10% increase in SCDF and approximately 2.5% increase in SLERF. If the HEPs were reduced to their 5th percentile, the SCDF decreases approximately 5% and the SLERF decreases less than 2%. The results indicate that the model is not overly sensitive to the HEP probabilities. Therefore, no further refinement to the SPRA is considered necessary.

5.7.9 Delay Evacuation Impact on LERF This sensitivity evaluates the impact of delayed evacuations caused by damage to surrounding infrastructure (e.g. bridges, communication towers). The delayed evacuations results in LERF sequences that were previously screened out because they are not early releases that should be included in the LERF as releases before evacuations take place. A simplified approach was used in this sensitivity where all seismic events with magnitude >0.Sg result in sufficient delay in the evacuation time such that they are modeled as leading directly to the LERF end state. The results show that SLERF increases by a factor of 3.2. Two other cases were evaluated where all seismic events

>0.6g and >lg were assumed to result in SLERF. The results showed increases in SLERF by a factor of 2.5 and 1.1 for >0.6g and >l.Og, respectively.

This sensitivity used a very simplified approach for estimating the impact on SLERF due to delays in evacuations since not all SCDF sequences at the elevated ground motions would result in direct LERF. There is uncertainty in what size seismic events could significantly impact the surrounding infrastructure and thus cause delays in evacuations.

The sensitivity shows there may be some impact on SLERF if the infrastruct_ure is impacted at lower ground motions. However, due to the conservative approach used in the sensitivity, no further refinements to the SPRA are considered necessary.

5.7.10 SPRA Logic Model and Quantification Technical Adequacy The NAPS SPRA risk quantification and results interpretation methodology were subjected to an independent peer review against the pertinent requirements in the PRA Standard [4]. The peer review assessment, and subsequent disposition of peer review findings, is described in Appendix A, and establishes that the NAPS SPRA seismic plant response analysis is suitable for this SPRA application.

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NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Summary Report March 2018 6.0 Conclusions A seismic PRA has been performed for North Anna Power Station Units 1 and 2 in accordance with the guidance in the PRA Standard [4] and the SPID [2]. The SPRA shows 5

that the point estimate seismic CDF is 6.0xlff5/yr and the seismic LERF is l.6xlff /yr for both units. The PRA model provides insights and identifies the most important equipment relied upon for responding to a seismic event. No seismic hazard vulnerabilities were identified.

The SPRA as described in this submittal reflects the as-built/as-operated North Anna Power Station Units 1 and 2 as of the SPRA freeze date - January, 2015. An assessment is included in Appendix A of the impact on the results of plant changes not included in the model. No seismic hazard vulnerabilities were identified, and no plant actions have been taken or are planned given the insights from this study.

7.0 References

[1] NRC (E Leeds and M Johnson) Letter to All Power Reactor Licensees et al.,

"Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(f) Regarding Recommendations 2.1, 2.3 and 9.3 of the Near-Term Task Force Review of Insights from the Fukushima Dai-lchi Accident," March 12, 2012.

[2] EPRI 1025287, Seismic Evaluation Guidance: Screening, Prioritization and Implementation Details {SPJD) for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1: Seismic. Electric Power Research Institute, Palo Alto, CA: February 2013.

[3] Virginia Electric and Power Company Letter to NRC, "North Anna Power Station Units 1 and 2 Response to March 12, 2012 Information Request - Seismic Hazard and Screening Report (CEUS Sites) for Recommendation 2.1," dated March 31, 2014.

(4] ASME/ANS RA-S-2008, Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, including Addendum B, 2013.

[5] NEl-12-13, External Hazards PRA Peer Review Process Guidelines, Revision 0, Nuclear Energy Institute, Washington, DC, August 2012.

[6] Pressurized Water Reactor Owners Group Report, PWROG-17028-P, Peer Review of the North Anna Units 1 & 2 Seismic Probabilistic Risk Assessment, Revision 0, PWR Owners Group Risk Management Committee, PA-RMSC-0403, December 2017.

(7] . EPRI NP 6041-SL, A Methodology for Assessment of Nuclear Power Plant Seismic Margin, Rev. 1. , *Electric Power Research Institute, Palo Alto, CA, August 1991.

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NAPS Units 1 and 2 10 CFR 50.54(f} NTTF 2.1 Seismic PRA Summary Report March 2018

[8] Virginia Electric and Power Company Letter to U.S. NRC Document Control Desk, "North Anna Power Station, Units 1 and 2, Summary Report for Individual Plant Examination of External Events (IPEEE} - Seismic", Serial No.97-303 dated May 27, 1997.

[9] EPRI TR-103959, Methodology for Developing Seismic Fragilities, Electric Power Research Institute, Palo Alto, CA, June 1994.

[10] EPRI 3002000709, Seismic PRA Implementation Guide, Electric Power Research Institute, Palo Alto, CA, December 2013

[11] Regulatory Guide 1.200, Revision 2, "An Approach For Determining The Technical Adequacy Of Probabilistic Risk Assessment Results For Risk-Informed Activities,"

U.S. Nuclear Regulatory Commission, March 2009

[12] NAPS SPRA Summary or Quantification Report

[13] NUREG-1855, "Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decision Making", Rev. 0, March 2009

[14] EPRI 1016737, Treatment of Parameter and Modeling Uncertainty for

, Probabilistic Risk Assessments, Electric Power Research Institute, Palo Alto, CA, December 2008

[15] Virginia Electric and Power Company Letter, "North Anna Power Station Units 1 and 2 Response to March 12, 2012 Information Request - Spent Fuel Pool Seismic Evaluation for Recommendation 2.1," dated December 14, 2017.

[16] NRC Letter, "North Anna Power 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 Relating to Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dal-lchi Accident (TAC Nos. MF3797 and MF3798}," dated April 20, 2015

[17] McGuire, R.K., W. J. Silva, and C. J. Costantino (2001}. "Technical Basis for Revision of Regulatory Guidance on Design Ground Motions: Hazard- and Risk-Consistent Ground Motion Spectra Guidelines", NUREG/CR- 6728, (Non proprietary}, U.S. Nuclear Regulatory Commission, Washington, D.C., 2001.

[18] EPRI 3002008093, An Approach to Human Reliability Analysis for External Events with a Focus on Seismic, Electric Power Research Institute, Palo Alto, CA, December 2016

[19] EPRI 1019200, Seismic Fragility Applications Guide Update, Electric Power Research Institute, Palo Alto, CA, December 2009 Page 73 of 181

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[20] Virginia Electric and Power Company Letter, "North Anna Power Station Units 1 and 2 Response to March 12, 2012 Information Request - Expedited Seismic Evaluation Process Report for Recommendation 2.1," dated December 17, 2014.

[21] Generic Implementation Procedure (GIP) for Seismic Verification of Nuclear Plant Equipment," Revision 3A updated December 2001, prepared by the Seismic Qualification Utility Group (SQUG).

[22] Virginia Electric and Power Company Letter to U.S. NRC Document Control Desk, "North Anna Power Station Units 1 and 2, Report in Response to March 12, 2012 Information Request Regarding Seismic Aspects of Recommendation 2.3," Serial No.14-017 dated January 30, 2014.

[23] ASME/ANS RA-Sa-2009, "Addenda to ASME/ANS RA-S-2008 Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications," American Society of Mechanical Engineers, New York, NY, February 2009.

[24] North Anna Unit 3 Final Safety Analysis Report

[25] PWROG-14058-P, Revision 0, Assessing the Need for a PRA Peer Review Following a PRA Model Change 8.0 Acronyms AB Auxiliary Building AC Alternating Current AFW Auxiliary Feedwater ANS American Nuclear Society AOD Air Operated Damper AOV Air Operated Valve ASCE American Society of Civil Engineers ASME American Society of Mechanical Engineers ATWS Anticipated Transient without Scram BE Best Estimate CCDP Conditional Core Damage Probability CDF Core Damage Frequency CDFM Conservative Deterministic Failure Margin Page 74 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTIF 2.1 Seismic PRA Summary Report March 2018 CEUS Central and Eastern United States CLERP Conditional Large Early Release Probability CO 2 Carbon Dioxide CST Condensate Storage Tank DC Direct Current ECC-AM Extended Continental Crust-Atlantic Margin ECST Emergency Condensate Storage Tank EDG Emergency Diesel Generator EOP Emergency Operating Procedure EPRI Electric Power Research Institute ESEP Expedited Seismic Evaluation Program ESF Engineered Safeguards Features ESGR Emergency Switchgear Room FEM Finite Element Model FIRS Foundation Input Response Spectra FLEX Diverse and Flexible Mitigation Strategies I

FPIE Full Power Internal Events FV Fussell-Vesely GMPE Ground Motion Prediction Equation GMRS Ground Motion Response Spectra IPEEE Individual Plant Examination for External Events HCLPF High Confidence of a Low Probability of Failure HEP Human Error Probability HF High Frequency HHSI High Head Safety Injection HRA Human Reliability Analysis HVAC Heating, Ventilation, and Air Conditioning ISRS In-Structure Response Spectrum LB Lower Bound LCC Load Control Center Page 75 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Summary Report March 2018 LERF Large Early Release Frequency LHSI Low-Head Safety Injection LMSM Lumped Mass Stick Model LOCA Loss of Coolant Accident LOOP Loss of Offsite Power MAFE Mean Annual Frequency of Exceedance MCC Motor Control Center MCR Main Control Room MESE-N Mesozoic and younger extended prior - narrow MOD Motor Operated Damper MOV Motor Operated Valve MSVH Main Steam Valve House NAPS North Anna Power Station NEI Nuclear Energy Institute NMSZ New Madrid Seismic Zone NRC Nuclear Regulatory Commission NSSS Nuclear Steam Supply System NTTF Near Term Task Force PGA Peak Ground Acceleration PORV Power Operated Relief Valve PRA Probabilistic Risk Assessment PSHA Probabilistic Seismic Hazard Analysis QS Quench Spray RCB Reactor Containment Building RCP Reactor Coolant Pump RCS Reactor Coolant System RLME Repeated Large Magnitude Earthquake RPS Reactor Protection System RS Recirculation Spray RWST Refueling Water Storage Tank Page 76 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Summary Report March 2018 SAMG Severe Accident Management Guideline.s SB Service Building SBO Station Blackout SCDF Seismic Core Damage Frequency SEL Seismic Equipment List SFP Spent Fuel Pool SFR Seismic Fragility Element within ASME/ANS PRA Standard SG Safeguards Building, Steam Generator SHA Seismic Hazard Analysis Element Within ASME/ANS PRA Standard SHS Seismic Hazard Submittal SHSR Seismic Hazard and Screening Report SI Safety Injection SLERF Seismic Large Early Release Frequency SMA Seismic Margin Assessment SOV Solenoid Operated Valve, Separation of Variables SPID

• Screening, Prioritization and Implementation Details SPR Seismic PRA Modeling Element Within ASME/ANS PRA Standard SPRA Seismic Probabilistic Risk Assessment SQUG Seismic Qualification Utility Group SRSS Square Root Sum of the Squares SRT Seismic Review Team SSC Structure, System or Component SSE Safe Shutdown Earthquake SSEL Safe Shutdown Equipment List SSI Soil Structure Interaction SSLOCA Small-small LOCA SSSI Structure-Soil-Structure Interaction SW Service Water SWPH Service Water Pump House SWVH Service Water Valve House Page 77 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTIF 2.1 Seismic PRA Summary Report March 2018 T-H Time History UB Upper Bound UHS Ultimate Heat Sink USI Unresolved Safety Issue V/H Vertical to Horizontal Page 78 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Summary Report March 2018 Appendix A Summary of SPRA Peer Review and Assessment of PRA Technical Adequacy for Response to NTTF 2.1 Seismic S0.54(f) Letter This Appendix has two purposes:

1. Provide a summary of the SPRA peer review

2. Provide the bases for why the SPRA is technically adequate for the 50.54(f) response.

The NAPS SPRA was subjected to an independent peer review against the pertinent requirements in Part 5 of the ASME/ANS PRA Standard [4].

I The information presented here establishes that the SPRA has been peer reviewed by a team with adequate credentials to perform the assessment, establishes that the peer review process followed meets the intent of the peer review characteristics and attributes in Table 16 of RGl.200 R2 [11] and the requirements in Section 1-6 of the ASME/ANS PRA Standard [4], and presents the significant results of the peer review.

A.1. Overview of Peer Review The peer review assessment, and subsequent disposition of peer review findings, is summarized here. The scope of the review encompassed the set of technical elements and supporting requirements (SR) for the SHA (seismic hazard), SFR (seismic fragilities),

and SPR (seismic plant response) elements for seismic CDF and LERF. The peer review therefore addressed the set of SRs identified in Tables 6-4 through 6-6 of the SPID [2].

The NAPS SPRA peer review was conducted during the week of July 17, 2017 at the Dominion Energy Innsbrook Technical Center offices in Glen Allen, Virginia. As part of the peer review, a walk-down of portions of NAPS Units 1 & 2 was performed on July 18, 2017 by selected members of the peer review team.

A.2. Summary of the Peer Review Process The peer review was performed against the requirements in Part 5 (Seismic) of Addenda B of the PRA Standard [4], using the peer review process defined in NEI 12-13 [S]. The review was conducted over a four-day period, with a summary and exit meeting on the fifth day.

The SPRA peer review process defined in [5] involves an examination by each reviewer of their assigned PRA technical elements against the requirements in the Standard to ensure the robustness of the model relative to all of the requirements.

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NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Summary Report March 2018 Implementing the review involves a combination of a broad scope examination of the PRA elements within the scope of the review and a deeper examination of portions of the PRA elements based on what is found during the initial review. The supporting requirements (SRs) provide a structure which, in combination with the peer reviewers' PRA experience, provides the basis for examining the various PRA technical elements. If a reviewer identifies a question or discrepancy, that leads to additional investigation until the issue is resolved or a Fact and Observation (F&O) is written describing the issue and its potential impacts, and suggesting possible resolution.

For each technical element, i.e., SHA, SFR, SPR, a team of peer reviewers were assigned, one having lead responsibility for that area. For each SR reviewed, the responsible reviewers reached consensus regarding which of the capability categories defined in the Standard that the PRA meets for that SR, and the assignment of the capability category for each SR was ultimately based on the consensus of the full review team. The Standard also specifies high level requirements (HLR). Consistent with the guidance in the Standard, capability categories were not assigned to the HLRs, but a qualitative assessment of the applicable HLRs in the context of the PRA technical element summary was made based on the associated SR capability categories.

As part of the review team's assessment of capability categories, F&Os are prepared.

There are three types of F&Os defined in [S]: Findings, which identify issues that must be addressed in order for an SR (or multiple SRs) to meet Capability Category II; Suggestions, which identify issues that the reviewers have noted as potentially important but not requiring resolution to meet the SRs; and Best Practices, which reflect the reviewers' opinion that a particular aspect of the review exceeds normal industry practice. The focus in this Appendix is on Findings and their disposition relative to this submittal.

A.3. Peer Review Team Qualifications The review was conducted by Dr. Andrea Maioli of Westinghouse, Dr. Martin Mccann of Jack Benjamin & Associates, Dr. Glenn Rix of Geosyntec Consultants, Dr. James J.

Johnson of James J. Johnson and Associates, Mr. Frederic Grant of Simpson Gumpertz &

Heger, Mr. Benny Ratnagaran of Southern Nuclear Operating Company, Dr. Jonathan Lucero of Arizona Public Services and Mr. Edmond Wiegert of Duke Energy. Appendix D contains the resumes for the reviewers. The team was assembled by the peer review team lead. The lead and reviewer qualifications have been reviewed by Dominion and have been confirmed to be consistent with requirements in the ANS/ASME PRA Standard and the guidelines of NEl-12-13.

Consistent with the requirement in Section 1-6.2.2 of the ASME/ANS PRA Standard [4],

the members of the peer review team were independent of the North Anna Units 1 & 2 PRA. They were not involved in performing or directly supervising work on any element evaluated in the overall North Anna Units 1 & 2 seismic PRA.

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NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Summary Report March 2018 Dr. Andrea Maioli, the team lead, has over 10 years of experience at Westinghouse in the nuclear safety area generally and PRA specifically for both existing and new nuclear power plants. He is the technical lead for all seismic PRA activities with Westinghouse.

He has supported and led peer reviews for internal events, internal flooding, fire PRAs, high winds and other external hazards as well as seismic PRAs and is a member of the ASME/ANS JCNRM and of the JCNRM Subcommittee on Standard Maintenance, which is maintaining the ASME/ANS PRA Standard.

Dr. Martin Mccann was the lead for the review of the Seismic Hazard Analysis (SHA) technical element. He has over 35 years of experience in engineering seismology including site response analysis and specification of ground motion. Dr. Mccann has served as SHA lead reviewer for a number of recent SPRAs. He was assisted in the hazard review by Dr. Glenn Rix, who has more than 25 years of experience in the areas of geotechnical earthquake engineering and engineering seismology (particularly for the eastern and central U.S.}, seismic hazard assessment and risk mitigation for civil infrastructure including dams and power plants, and advanced near-surface geophysics investigations and interpretations across a range of applications. Dr. Rix also served as reviewer for multiple recent SPRAs peer reviews.

Dr. James Johnson, the lead reviewer for the SFR technical element, is an independent contractor with more than 40 years of experience mainly in the area of structural and engineering mechanics. He has been involved SPRAs for 35 nuclear power plants as well as in numerous peer reviews. He was assisted in fragility review by Mr. Frederic Grant and Mr. Benny Ratnagaran. Mr. Grant has 11 years of structural mechanics engineering experience, the majority of which has been in the commercial and government nuclear industries. His work in the nuclear industries involves seismic probabilistic risk assessments, seismic fragility analysis, seismic margin assessments, experience-based seismic qualification methods, walkdown of existing facilities, probabilistic seismic response analysis of structures, and analysis of damage indicating ground motion parameters. Most recently he served as reviewer for the Watts Bar SPRA peer review and he has. defended the Indian Point SPRA peer review. He is a member of the ASME/ANS JCNRM Working Group maintaining Part 5 of the ASME/ANS PRA Standard.

Mr. Ratnagaran has 5 years of experience and supported the Vogtle and Hatch Seismic PRA. He has defended the Vogtle Units 1 & 2 and 3 & 4 as well as Hatch SPRA peer reviews.

Mr. Edmond Wiegert was the lead reviewer for the SPR technical element. Mr. Wiegert has 25 years of experience in the nuclear industry and 17 years' experience in the areas of probabilistic risk assessment (PRA), now at Duke Energy as lead engineer in the PSA applications and models group. Mr. Wiegert has been supporting the SPRA modeling task for the Duke plants and supported numerous peer reviews. He was assisted in the SPR technical element review by Dr. Jonathan Lucero. Dr. Lucero has eight years of PRA and nuclear power experience in various aspects of PRA such as model maintenance, Page 81 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Summary Report March 2018 online and shutdown risk assessment, and regulatory oversight process. Dr. Lucero is the lead for the SPRA at Palo Verde and has supported numerous peer reviews for fire PRA and external hazards PRAs. Dr. Lucero was also the lead reviewer for the PRA configuration control element of the review.

Three working observers (Robert Keiser, Rusty Childs and Winston Stewart from Duke) supported the review of the SFR technical element, while David Gerlits, from Westinghouse supported as working observer the review of the SPR technical element.

Any observations and findings these working observers generated were given to the peer review team for their review and "ownership." As such, Mr. Keiser, Mr. Childs, Mr.

Stewart and Mr. Gerlits assisted with the review but were not formal members of the peer review team.

Finally, Mr. Gerald Dowdy (AEP) supported the review as a process observer. In this role Mr. Dowdy was not a formal reviewer.

A.4. Summary of the Peer Review Conclusions The review team's assessment of the SPRA elements is excerpted from the peer review report as follows. Where the review team identified issues, these are captured in peer review findings, for which the dispositions are summarized in the next section of this appendix.

Seismic Hazard (SHA)

The Standard requires the seismic hazard input to the SPRA be determined on the basis of a site-specific probabilistic seismic hazard analysis (PSHA). The PSHA performed for the North Anna site is a site-specific analysis that was performed by:

1. Using existing regional seismic source characterization (SSC) and ground motion characterization (GMC) models;

2. Assessing whether conditions local to the plant site and/or the availability of new data since the SSC and GMC models were developed require a revision of the regional-scale models to define a site-specific PSHA for the North Anna site;

3. Evaluating the effects of local site conditions on the ground motions; and

4. Considering potential ground failures caused by soil liquefaction, landslides, fault displacement, and other secondary hazards.

The regional-scale SSC model is the recently completed Central and Eastern U.S. (CEUS) seismic source model (NRC, EPRI, and DOE, 2012). The existing GMC model is based on the recent EPRI (2013) CEUS ground motion update project. Both models were the result of SSHAC Level 3 studies and, in the case of the GMC model, a SSHAC Level 2 update of a prior Level 3 study. The SSHAC process provides a structured approach to the use of experts and the evaluation and integration of available information, and provides minimum technical requirements to complete a PSHA. Using an appropriate Page 82 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Summary Report March 2018 "SSHAC level" when conducting a seismic hazard study ensures that data, methods, and models supporting the PSHA are appropriately assessed and incorporated and that uncertainties are fully considered in the process at a sufficient depth and level of detail necessary to satisfy scientific and regulatory requirements. Although the Standard does not define a minimum required SSHAC level of analysis, the available Level 3 studies satisfy SHA High Level Requirement A (SHA-A).

As part of the SSC SSHAC study, comprehensive datasets were compiled to support the evaluations of the Technical Integration Teams, including regional geological, seismological, and geophysical data for the CEUS. As part of the CEUS SSC project, an earthquake catalog of relevant historical, instrumental, and paleoseismic information was gathered and processed. These aspects of the SSC study satisfy SHA-B.

The CEUS SSC model defines seismic sources for the entire central and eastern U.S. For purposes of the North Anna PSHA, background seismic sources in the CEUS SSC model within 320 km of the site were included in the PSHA. To expedite the calculations, cells in the gridded seismicity model that are more than 1,000 km from the North Anna site were not included in the calculation. This simplification is reasonable and does not impact the estimate of the seismic hazard at the site. In addition, repeated large-magnitude earthquake (RLME) seismic sources were included in the North Anna PSHA.

The inclusion of RLME sources was based on the criterion that sources that contributed to 99 percent of the mean hazard for spectral acceleration of 1.0 Hz were included in the PSHA (sources contributing less than 1 percent were excluded). For the North Anna analysis, the Charleston, Reelfoot, New Madrid, and Wabash Valley RLME sources were included in the PSHA. Based on this selection process, "near-field" and "far-field" earthquake sources that are contributors to ground motions at North Anna were considered in the analysis. These aspects of the SSC study satisfy SHA-C.

As part of the GMC SSHAC studies in 2004 and 2013, available ground motion datasets and models were compiled and evaluated. The resulting ground motion prediction equations (GMPEs) account for epistemic and aleatory uncertainties. Accordingly, SHA-D is satisfied.

The effects of local site conditions are included (SHA-E) via amplification factors derived from site response analyses, which incorporate site-specific information on site topography, surficial geologic deposits, and site geotechnical properties. Epistemic uncertainty and aleatory variability in shear wave velocity, layer thickness, and nonlinear properties are considered in the site response analyses. However, inadequate justification is provided for the assumption that epistemic uncertainty is negligible (and thus not considered) compared to the aleatory variability.

Requirement SHA-F addresses the quantification of the seismic hazard; propagation of uncertainties and the results that are generated. Both the aleatory and epistemic uncertainties were addressed in the SSC and GMC parts of the analysis and were Page 83 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Summary Report March 2018 propagated through the hazard quantification. The PSHA results that were generated include fractile and mean hazard curves, uniform hazard response spectra, and magnitude-distance deaggregation plots. Sensitivity analysis results are presented in the report that document the contribution of seismic sources and alternative GMPEs to the site hazard. However, no sensitivity analyses were included to evaluate uncertainties in site response parameters.

The spectral shape used in the seismic PRA is based on the results of the site-specific PSHA (SHA-G). A GMRS was generated from the site-specific lE-4 and lE-5 UHRS and the associated design factors from Regulatory Guide 1.208. The UHRS developed in the PSHA was extended to a spectral frequency of 0.1 Hz. The process used in the extrapolation was based on the EPRI (2013) GMPEs for large-magnitude earthquakes (M 7 to 7.5) which predict constant spectral velocity in the 0.5 and 0.2 Hz and transition to constant spectral displacement at lower frequencies.

Vertical response spectra were developed for input to the seismic response analysis.

The vertical spectra were derived from the horizontal response spectra using vertical-to-horizontal (V/H) spectral ratios. The McGuire et al. (2001) hard rock V/H ratios were used to derive V/H spectral ratios appropriate for CEUS soil sites.

The CEUS SSC and the GMC models are existing, regional-scale models, and in principle are not site-specific. The requirements of SHA-H state that if an existing PSHA is used, 'it shall be confirmed that the basic data and interpretations are still valid in light of current information. In the context of an existing, regional-scale study, SHA-H requires that steps be taken to develop a North Anna site-specific PSHA model. Somewhat unique to the North Anna site was the occurrence of the 2011 Mineral, VA earthquake near the plant. To satisfy SHA-H, the PSHA analysts conducted a systematic data coll~ction and evaluation of geological, seismological, and geophysical data. This included a SSHAC Level 2 evaluation focused on the implications of the Mineral, VA earthquake, an update to the earthquake catalog that is the basis for the estimate of earthquake recurrence rates, and the evaluation of new information available in the literature to determine if there was a basis for making revisions to the SSC model or the addition of new, local seismic sources that would contribute to the ground motion hazard at the North Anna site. The evaluation of the Mineral, VA earthquake which included discussions/input from experts in the field, a literature review concluded there was no basis to revise or amend the SSC model for the North Anna PSHA.

In the case of the GMC model, a systematic assessment was not performed to assess whether an update was required. The reason for not conducting such an evaluation is the pending completion of the NGA East modeling effort that will develop new GMPEs for the CEUS.

The potential for induced earthquakes associated with hydraulic fracturing or waste fluid injection was not evaluated as part ohhe PSHA.

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NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Summary Report March 2018 The Standard requires that a screening analysis be performed to assess whether in addition to vibratory ground motion, other seismic hazards, such as fault displacement, landslide, soil liquefaction, or soil settlement, need to be included in the seismic PRA. As part of the PSHA a systematic evaluation should be carried out to identify if there are other seismic hazards that may impact the site. Analyses were performed that considered the (i) seismic stability of the dike for the service water reservoir, and (ii) potential for soil liquefaction. A screening assessment for other potential seismic hazards, such as fault displacement, ground settlement, seiche in the reservoir, flooding due to dike breach and uncontrolled release of the reservoir, etc. was not performed.

Furthermore, the evaluation of the potential in the power block area and instability of the service water reservoir dike lacked rigor and therefore a basis to confidently screen them out from the seismic PRA.

SHA-J defines the requirements for documentation of the PSHA. The documentation of the North Anna PSHA is a collection of documents and analyses for Unitsl/2 and 3. SHA-J sets a high bar with regard to the documentation that should be prepared for the PSHA and the needs (applications) it must satisfy (e.g., PRA applications, peer review, future updates). For the North Anna PSHA, a PSHA Summary Report that fully describes the methodology that was implemented, the rock PSHA results, the site response analysis, sensitivity studies, the control point motions, etc. was not prepared. The lack of a PSHA Summary report is further complicated by the fact that part of the Unit 1/2 seismic hazard story is based on the analyses and results that were performed for the Unit 3 combined license (COL) which is documented in the Unit 3 Final Safety Analysis Report (FSAR) and various supporting calculations.

Seismic Fragility (SFR)

Seismic fragility analyses were performed for the North Anna Power Station Units 1 and 2 structures, systems, and components (SSCs) that were included in the seismic equipment list (SEL). The SEL was developed from the internal event PRA basic events with additions and subtractions resulting from recognition that the seismic hazard applies loading conditions to passive as well as active equipment and components and recognition that some systems, components, and equipment are robust when subjected to seismic loading conditions. The focus was on SEL items that were significant contributors to seismic core damage frequency (SCDF) and seismic large early release frequency (SLERF). This was possible through efficient computational tools that permitted risk quantification analyses to be performed quickly for sensitivity study purposes.

The seismic fragility analyses were significantly enhanced through the NAPS systematic approach to solve complex technical problems associated with the SPRA. Position Papers were developed on fourteen (14) topics that represented either complex technical issues or issues that required coordination between various disciplines to ensure that rigorous treatment and consistency were maintained. The Position Papers Page 85 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTIF 2.1 Seismic PRA Summary Report March 2018 identified industry standards, methodologies, and best practices to be implemented. An external expert technical panel reviewed and commented on draft versions of the Position Papers; these comments were addressed and final versions of the Position Papers were issued. The Position Papers were consistently used as guidance for the work performed throughout the SPRA effort. The PWROG Peer Review considers this approach a "best practice."

The SSCs that are judged to be of high-seismic capacity were not screened out and are retained in the Seismic PRA model. Only the SSCs that are considered inherently rugged were screened out from the Seismic PRA model. The high-seismic capacity SSCs are assigned a screening level HCLPF of 1.0g. A sensitivity study was conducted to show that the risk contribution (SCDF) of SSC failure (modeled as direct core damage) with HCLPF capacity of 0.6g is very low.

Seismic input motions were based on the ground motion response spectra (GMRS) characterized by high frequency motion (greater than 10 Hz) and anchored to a horizontal peak ground acceleration (PGA) of 0.572g. Foundation input response spectra (FIRS) were developed from the probabilistic site response analyses for eight different locations in the site profile, each corresponding to structure foundation locations of one or more (grouped) structures.

Median-centered seismic response analyses were performed for all the structures that were included in the SPRA. Depending on the foundation (soil/rock properties) and embedment conditions either fixed-base, deterministic soil structure interaction (D-SSI),

or probabilistic SSI (P-SSI) analyses were performed. New finite element models were developed for structures if their existing lumped mass stick models (LMSM) were judged to be inadequate for use in the seismic response analyses. The preferred method of generating seismic responses for fragility development is the P-SSI approach. The NAPS SPRA Team implemented P-SSI analyses for soil/rock supported structures based on evaluations of their importance to risk, their physical attributes (supporting media, foundations, structure configuration) and the Team's intent to provide the best estimate of structure specific seismic responses (median and variabilities) for fragility development. The peer review team considered the response analysis results to be reasonable, and upon close inspection of the calculations, found that the methods and approaches were generally technically rigorous. For these reasons, the PWROG Peer Review considers the response analysis to represent a "best practice."

The seismic fragility analyses followed industry guidelines as described in Position Papers 3, 4, and 6. Generally, plant-specific data was used, including seismic qualification data, 2011 Mineral earthquake performance data and post-earthquake evaluation results, IPEEE data and analyses, and other available information. Plant specific data was supplemented by seismic experience data, based on EPRI NP-6041, Rev. 1 and generic test data such as GERS and relay tests. Various levels of screening were progressively implemented based on robust behavior of SEL items as recognized Page 86 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTIF 2.1 Seismic PRA Summary Report March 2018 by EP~I NP-6041, Rev. 1, and experience of the NAPS SPRA Team and consultants; bounding calculations of NAPS site specific hazard data and assumed fragility curve values to identify and confirm fragility functions (HCLPF and variability) that had minimal effects on risk metrics; and individual SEL items (or groups) that have minimal effects on the risk metrics as calculated by the sensitivity studies of risk quantification. Generally, plant-specific fragility functions for the remaining unscreened items were generated, including consideration of anchorage capacity and seismic systems interaction (11/1).

Seismically induced fire and flood initiators were walked down and evaluated. These approaches of screening, calculation of preliminary fragilities (generally conse~vative),

and performing sensitivity studies with risk quantification is acceptable. Some exceptions are noted below.

For future work (including resolution of findings of the PWROG Peer Review), the NAPS SPRA Team should focus on the following:

• Based on the ASME ANS "Addendum B," for Capability Categories II and 111, realistic fragilities based on site/plant specific data are required. Accurate risk insights require realism to avoid misinterpretation of risk important SSCs and phenomena (SR SFR A2). Examples are:

o Turbine Building (TB) and its contents were not credited in the quantification of the SCDF and SLERF. However, consequences of the assumed TB failure and its contents were not extensively evaluated.

Assumed structure failure likely causes failure of its contents, which includes large diameter piping systems (Circulating Water System) (4-96" lines) that could be a flooding source to the adjacent Emergency Switchgear Rooms. Other consequences of TB failure, such as potential interaction or impact with, or load redistribution onto neighboring structures, should also be evaluated. The evaluation should be documented.

o Structure fragilities based on EPRI NP-6041, Rev. 1 Table 2-3 provides I , HCLPF values for overall behavior when caveats are met. In addition, local sources of failure should be evaluated in a structure's focused walkdown, e.g., penetrations, relative displacement effects on systems running from structure-to-structure and supported therein. The review and evaluations of local sources of failure should be documented.

• Revisit all risk significant contributors to SCDF and SLERF, and verify their fragilities are site/plant specific and realistic, including MOVs, PORVs, reactor containment building, auxiliary feed water pump house, service water valve house, and emergency condensate storage tank.

• Fragilities calculated for other purposes, such as IPEEE, should be revisited to verify that all failure modes have been considered, e.g., the Emergency Condensate Storage Tank (ECST) fragility was previously based on tank failure -

currently, and its fragility is based on EPRI NP-6041 Rev. 1 Table 2-3 failure of the Page 87 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Summary Report March 2018 concrete missile shield (MS). This should be revisited, including penetrations in the MS.

• Facts and observations associated with the seismic walkdown including the following:

o Revise and improve the summary walkdown report to facilitate peer review and future applications of the SPRA.

o Review walkdown documentation for consistency between teams, equipment types, and locations. For example, focus on documentation of issues such as seismic systems interaction (11/1), systems extending from structure-to-structure and supporting structure-to-supporting structure.

o Document the presence of all walkdown personnel and their credentials including technical support personnel.

• Respond to facts and observations associated with the structure response analysis, such as structure damping values, concrete compressive strength for stiffness calculations, SASSI Modified Subtraction Method (MSM) verification, etc.

Seismic Plant Response (SPR)

The NAPS SPRA was developed starting from the internal events PRA and captures seismically induced failures along with random failures, unavailabilities and operator errors. The SPRA was determined to adequately model seismically induced initiating events: the process was systematic to identify, screen, and model the events.

The review team identified some scenarios were not identified or not addressed. No seismic fire scenarios were modeled. This is possibly the result of an aggressive screening approach. It is noted that North Anna does not have a fire PRA to support fully defending the screening of all seismically induced fire scenarios. Furthermore, some of the scenarios evaluated in the IPEEE fire evaluation, with a realistic potential for a seismic-induced equivalent, were not included or addressed in the model. The rationale used for the screening of the seismically induced fire scenario relies upon the SPRA Implementation Guide (SPRAIG) [10]. The SPRAIG has been recently recognized to underestimate the possibility of seismically induced fire scenarios for example, by not

• addressing the possibility of seismically induced fires generated by high energy cabinets.

Similarly, an important flood scenario discussed in the internal flooding PRA was not addressed completely for the seismic-related equivalent scenario.

The SPRA appropriately models the seismic failures in the system model. The effect of relay chatter is also addressed adequately. Operator actions are adequately addressed for the seismic related performance shaping factor. A progressive approach is used where important operator actions are addressed more in details. One exception was observed where a potentially significant operator action was not credited in the SPRA, possibly resulting in a slightly conservative estimation of LERF. As discussed above, the Flowserve RCP seal package is credited in the SPRA but was not peer reviewed as part of Page 88 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Summary Report March 2018 either this or previous peer reviews. The Flowserve RCP seal modeling in considered a model upgrade based on Reference 25, and as such will need to be peer reviewed.

The Seismic Equipment List appears to be comprehensively assembled, with the notable exception of the travelling screens that have been screened functionally before including them in the SEL for fragility considerations. This resulted in overlooking the structural failure mode of these components, which could prevent water flow. This appears to be an isolated issue in the process, that otherwise seems to be capturing all relevant components for fragility considerations.

The quantification of the SPRA was determined to be performed in accordance with standard practice. Meaningful insights can be retrieved from the quantification results and are effectively summarized and discussed. It was observed that no investigation was made on the sensitivity of the SPRA model to the number and size of the seismic acceleration bins used in the quantification of the SPRA. The eight seismic acceleration bins generated by default in FRANX have been retained. In other SPRAs across the industry, it was observed that some SPRA models are significantly sensitive to the number and especially size of the bins: SCDF can be overestimated up to 40% by a non-optimal binning size. It is recommended to perform a more extensive study of the stability of the model that could result in an appreciable modification of CDF and LERF.

It was also observed that the uncertainty assessment was limited to a .few standard sensitivities. Most notably, no sensitivities were made to test the model sensitivity to grouping and correlation of fragilities. The grouping and correlation of fragilities is known to have the potential to mask or bias results and is recognized to have an appreciable degree of epistemic uncertainties. There is no evidence that alternative grouping of components in fragility groups have been considered or even discussed with the fragility team. Furthermore, there was no investigation of the epistemic/model uncertainties associated with modeling of FLEX equipment. Modeling FLEX equipment can be impacted due to the limited knowledge and experience associated with current human reliability analysis (HRA) methods and component reliability data. Because of the very high importance of FLEX equipment in the NAPS SPRA, some investigation of these aspects appears to be necessary. There are other assumptions documented in the development of the SPRA, for which a disposition of the potential associated epistemic uncertainties was not performed. These quantification and uncertainty issues are more important on LERF than on CDF, given the higher absolute LERF value.

PRA Configuration Control The primary NAPS PRA Configuration Control procedure reviewed was NF-AA-PRA-410, Revision 8. This provided a good overview of the NAPS configuration control process.

This document also listed specific guidance documents to assist the PRA engineer on how to evaluate and assess PRA model impacts. Overall the NAPS PRA Configuration Page 89 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Summary Report March 2018 Control process meets the requirements of SMU. However, the following observations are made:

• Industry guidance documents are not listed.

• There's no specific step to evaluate a change impact as an "update" or "upgrade."

• There's no specific guidance on when to have a peer review performed.

It was also observed that all of the tools that are used to track design changes, errors or similar impacts to the internal events PRA model seem well organized and easy to use.

However, the tools are not set up for managing model impacts specifically against the seismic PRA model per se. This has been flagged as a possible limitation of the program, especially given that the SPRA is the first additional hazard (i.e., non-internal events) that has been added to the NAPS PRA.

The review team concluded that the North Anna SPRA realistically reflects the seismic risk profile of the plant, with no evident bias or conservatism. As a result of the above the North Anna SPRA is judged by the review team to be technically adequate for supporting risk-informed applications and risk-informed decision making.

A.5. Summary of the Assessment of Supporting Requirements and Findings Table A-1 presents a summary of the SRs graded as Not Met or Not Capability Category II, and lists the associated Finding F&Os and disposition for each. Table A-2 presents summary of the Finding F&Os and the disposition for each (included at the end of this Appendix due to size).

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NAPS Units 1 and 2 10 CFR 50.54(f) NTIF 2.1 Seismic PRA Summary Report March 2018 Table A-1: Summary of SRs Graded as Not Met or Capability Category I for Supporting Requ1remen t s Covere d b1yt h e NAPS SPRA Peer Rev1ew Assessed Associated Disposition to Achieve Met or SR Capability Finding F&Os Capability Category II Category SHA Associated F&Os have been resolved. SR is judged to ,

SHA-E2 CCI 20-3 be Met for Capability Category II.

Associated F&Os have been resolved. SR is judged to SHA-12 Not Met 20-5, 20-8 be Met.

Associated F&Os have been resolved. SR is judged to SHA-Jl Not Met 20-1 be Met.

SFR 23-8, 23-10, Associated F&Os have been resolved. SR is judged to SFR-A2 CCI 23-11, 23-12, be Met for Capability Category II.

24-2 Associated F&Os have been resolved. SR is judged to SFR-F2 Not Met 23-8, 24-2 be Met.

SPR None N/A - -

(S)MU Associated F&Os have been resolved. SR is judged to (S)MU-B4 Not Met 25-4 be Met.

A.6. Summary of Technical Adequacy of the SPRA for the 50.54(f) Response The set of supporting requirements from the ASME/ANS PRA Standard [4] that are identified in Tables 6-4 through 6-6 of the SPID [2] define the technical attributes of a PRA model required for a SPRA used to respond NTTF Recommendation 2.1: Seismic of the 10 CFR 50.54(f) letter [1]. The conclusions of the peer review discussed above and summarized in this submittal demonstrates that the NAPS SPRA model meets the expectations for PRA scope and technical adequacy as presented in RG 1.200, Revision 2

[11] as clarified in the SPID [2].

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NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Summary Report March 2018 The main body of this report provides a description of the SPRA methodology, including:

o Summary of the seismic hazard analysis (Section 3}

o Summary ofthe structures and fragilities analysis (Section 4) o Summary of the seismic walkdowns performed (Section 4) o Summary of the internal events at power PRA model on which the SPRA is based, for CDF and LERF (Section 5) o Summary of adaptations made in the internal events PRA model to produce the seismic PRA model and bases for the adaptations (Section 5)

Detailed archival information for the SPRA consistent with the listing in Section 4.1 of RG 1.200 Rev. 2 is available if required to facilitate the NRC staff's review of this submittal.

The NAPS SPRA reflects the as-built and as-operated plant as of the cutoff date for the SPRA, January 2015. There are no permanent plant changes that have not been reflected in the SPRA model except for those discussed further in section A.9.

The peer review observations and conclusions noted in Section A.4, the F&O finding dispositions noted in the discussion in Section A.5, and the discussion in Section A.7 demonstrate that the NAPS SPRA is technically adequate in all aspects for this submittal.

Subsequent to the SPRA peer review, the peer review findings have been appropriately dispositioned, and the SPRA model has been updated to reflect these dispositions and further refine several fragility values. The results presented in this submittal reflect the updated model as of January 2018. No changes were made in updating the model that would require a subsequent focused peer review except for the Flowserve RCP seal upgrade as discussed in F&O 25-9.

A.7. Summary of SPRA Capability Relative to SPID Tables 6-4 through 6-6 The PWR Owners Group performed a full scope peer review of the NAPS internal events PRA and internal flooding PRA that forms the basis for the SPRA to determine compliance with ASME PRA Standard, RA-S-2009 [23] and RG 1.200 Rev. 2 [11] in in November 2013. The ASME/ANS PRA standard contains a total of 316 numbered supporting requirements for internal events and internal flooding in nine technical elements and 10 configuration control supporting requirements. Eleven of the SRs were determined to be not applicable to the North Anna PRA. Of the 315 remaining SRs, 292 SRs, or 92%, were rated as SR Met, Capability Category 1/11, or greater. Three SRs were rated as Category I and 20 SRs were Not Met. A total of 72 F&Os were issued by the peer review team with 35 being findings, 35 suggestions, and 2 were best practices.

Since the peer review, the internal events PRA model has been revised to address 13 finding F&Os that were found to impact the PRA model logic and results. The remaining finding F&Os were considered to be documentation improvements and other changes that were not considered to impact the PRA model results. As part of the SPRA Page 92 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTIF 2.1 Seismic PRA Summary Report March 2018 development, these. F&Os were reviewed again and verified to not impact the SPRA model.

The PWR Owners Group performed a peer review of the NAPS SPRA in July, 2017 The results of this peer review are discussed above, including resolution of SRs assessed by the peer review as not meeting Capability Category 11, and resolution of peer review findings pertinent to this submittal. The peer review team expressed the opinion that the NAPS seismic PRA model is of good quality and integrates the seismic hazard, the seismic fragilities, and the systems-analysis aspects appropriately to quantify core damage frequency and large early release frequency. The general conclusion of the peer review was that the NAPS SPRA is judged to be suitable for use for risk-informed applications.

• Table A-1 provides a summary of the disposition of SRs judged by the peer review to be not met, or not meeting Capability Category II.

• Table A-2 (located at the end of this Appendix due to size) provides a summary of the disposition of the open SPRA peer review findings.

• Table A-3 provides an assessment of the expected impact on the results of the NAPS SPRA of the peer review Findings that have not been fully addressed.

Of the peer review finding-level Facts and Observations (F&Os) listed in Table A-2, most were associated with PRA Standard supporting requirements (SRs) that were deemed by the peer reviewers to be either "Met" or met at "Capability Category II." This indicates, as can be seen from the finding details, that these findings deal with relatively focused issues that have been adequately dispositioned within the reviewed methodologies, for the SPRA and for future risk-informed application. Many of these were documentation related.

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NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Summary Report March 2018 Table A-3 Summary of Impact of Not Met SRs and Unresolved Peer Review Findings Summary of Issue Not Fully SR Finding Impact on SPRA Results Resolved Flowserve seal model in NAPS SPRA is the same as the New Flowserve RCP seal Flowserve seal model in the model is considered an Surry PRA model, which has SPR-Bl 25-9 been peer reviewed. F&Os upgrade that has not been from the Surry Flowserve peer reviewed. peer review were reviewed and verified not to impact the SPRA results.

As this list indicates, there is only one finding F&O that has not been resolved, and it is not expected to impact the SPRA results as noted in the table. All of the other finding F&Os have been resolved and therefore, the SPRA is considered to be technically adequate to provide risk insights for the NTTF 2.1 submittal.

The SPID [2] defines the principal parts of an SPRA, and the NAPS SPRA has been developed and documented in accordance with the SPID. The information in the tables identified above demonstrates that the NAPS SPRA is of sufficient quality and level of detail for the response to the NTTF 2.1 Seismic SPRA submittal.

A.8. Identification of Key Assumptions and Uncertainties Relevant to the SPRA Results.

The PRA Standard [4] includes a number of requirements related to identification and evaluation of the impact of assumptions and sources of uncertainty on the PRA results.

NUREG-1855 [13] and EPRI 1016737 [14] provide guidance on assessment of uncertainty for applications of a PRA. As described in NUREG-1855, sources of uncertainty include "parametric" uncertainties, "modeling" uncertainties, and "completeness" (or scope and level of detail) uncertainties.

• Parametric uncertainty was addressed as part of the NAPS SPRA model quantification (see Section 5 of this submittal).

• Modeling uncertainties are considered in both the base internal events PRA and the SPRA. Assumptions are made during the PRA development as a way to address a particular modeling uncertainty because there is not a single definitive approach. Plant-specific assumptions made for each of the NAPS SPRA technical elements are noted in the SPRA documentation that was subject to peer review, and a summary of important modeling assumptions is included in Section 5.

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NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Summary Report March 2018

• Completeness uncertainty addresses scope and level of detail. Uncertainties associated with scope and level of detail are documented in the PRA but are only considered for their impact on a specific application. No specific issues of PRA completeness were identified in the SPRA peer review.

A summary of potentially important sources of uncertainty in the NAPS SPRA is listed in Table A-4.

Table A-4 Summary of Potentially Important Sources of Uncertainty PRA Summary of Treatment of Sources of Potential Impact on SPRA Element Uncertainty per Peer Review Results Seismic The NAPS SPRA peer review team noted The seismic hazard Hazard that both the aleatory and epistemic reasonably reflects sources of uncertainties have been addressed in uncertainty.

The conclusion in the site characterizing the seismic sources.

response analysis that Uncertainties in each step of the hazard epistemic uncertainty is analysis were propagated and displayed in negligible due to the the final quantification of hazard estimates extensive site investigations for the NAPS site. The peer review team has been further justified in noted that inadequate justification was response to peer review team provided in the site response analysis for F&O 20-3 and is not considered a significant the assumption that epistemic uncertainty source of uncertainty.

is negligible (and thus not considered) compared to the aleatory variability.

Seismic The fragility of some SSCs were identified The fragilities of the SSCs Fragilities by the peer review team as being overly noted by the peer review conservative resulting in low HCLPF team were revised to more capacities. appropriately model their seismic capacity. Also, a sensitivity was performed to assess the impact of building capacity on the SCDF and SLERF, which showed no significant impact on the results that warrant changes to the fragilities.

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NAPS Units 1 and 2 10 CFR 50.54{f) NTTF 2.1 Seismic PRA Summary Report March 2018 Table A-4 Summary of Potentially Important Sources of Uncertainty PRA Summary of Treatment of Sources of Potential Impact on SPRA Element Uncertainty per Peer Review Results Seismic Assumptions and sources of uncertainties Additional sensitivities were PRA in the SPR development were reviewed to performed to address the Model identify sources that may have an peer review team F&Os. The important impact on the SPRA results. results showed the SPR Sensitivities were performed as modeling as appropriate with documented in Section 5.7 to verify the no significant impact on the sources of uncertainty do not have a results.

significant impact on the SPRA results.

The peer review team assessed supporting requirement SPR-F3 as met for documenting sources of uncertainty. It did issue some F&Os recommending that additional sensitivities be performed.

Additional sensitivities were performed to confirm the adequacy of the SPRA model.

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NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Summary Report March 2018 A.9. Identification of Plant Changes Not Reflected in the SPRA The NAPS SPRA reflects the plant as of the cutoff date for the SPRA, which was January 2015. Table A-5 lists significant plant changes subsequent to this date and provides a qualitative assessment of the likely impact of those changes on the SPRA results and insights.

Table A-5 Summary of Significant Plant Changes Since SPRA Cutoff Date Description of Plant Change Impact on SPRA Results Motor-Control Center (MCC) The new MCC buckets have contactors that have Bucket Replacement - MCC higher HCLPF capacity for relay chatter of the Motor-breaker assemblies being operated valves (MOVs) they power. The bucket replacement is an ongoing effort that will continue replaced in various MCCs.

for the next couple of years. The SPRA reflects the latest configuration as of January 2018. As buckets are replaced, the HCLPF capacity of the corresponding MOVs will be improved and therefore the SCDF and SLERF will be reduced. The reduction in SCDF and SLERF is not expected to be significant given the relatively low FV importance of the fragility groups that model these MOV chatter failures.

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NAPS Units 1 and 2 10 CFR 50.54(f) NTIF 2.1 Seismic PRA Summary Report March 2018 Table A-5 Summary of Significant Plant Changes Since SPRA Cutoff Date Description of Plant Change Impact on SPRA Results EDG Room Heaters During the walkdowns of the Emergency Diesel Generators (EDGs), the supports for the unit heaters in the EDG rooms were identified as potential seismic spatial interaction concerns. The heaters are located such that if the supports failed and the I heater unit displaced' significantly, the attached steam supply and drain piping could breach and leak into the room, or the heater unit could fall and impact sensitive EDG equipment. Modifications have been developed to upgrade the heater supports to resolve the seismic spatial interaction concern.

The SPRA model assumes these modifications have been completed for all of the heaters. The modification for all but two of the heaters have been completed. The modifications are scheduled to be completed for the final two heaters during the Spring 2018 unit 1 refueling outage.

Flowserve Seals North Anna has replaced the RCP seals with Flowserve low leakage seals for all RCPs expect for the Unit 1 'C' RCP, which still has the Westinghouse seal. This seal will be replaced during the Spring 2018 unit 1 refueling outage. The SPRA assumes all RCP seals have been replaced with the Flowserve seals.

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NAPS Units 1 and 2 10 CFR 50.54(f) NTIF 2.1 Seismic PRA Summary Report March 2018 Table A-5 Summary of Significant Plant Changes Since SPRA Cutoff Date Description of Plant Change Impact on SPRA Results Fire Extinguishers During the seismic walkdowns, portable fire extinguishers were identified that were stored on a bracket configuration that could allow the fire extinguisher to fall during a severe seismic event.

For the fire extinguishers located in areas where there are mitigating SSCs, the fire extinguisher could potentially impact sensitive equipment upon falling.

Additionally, mobile CO2 firefighting carts were identified that could displace in the event of a severe seismic event and potentially impact sensitive equipment cabinets containing mitigating instruments.

The SPRA does not include these damage scenarios because of the uncertainty in whether the fire extinguishers could fall from the support bracket or the firefighting carts could displace sufficiently to impact sensitive cabinets.

The potential for seismic spatial interactions from this firefighting equipment is being addressed through engineering review to resolve the concern.

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NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Submittal March 2018 SHA-Jl, 20-1 The documentation of the As part of the PSHA Prepare a Summary Documentation of the North J3 North Anna PSHA is a documentation, a PSHA Report that Anna PSHA has been enhanced collection of documents summary report of the provides a complete to provide a complete and analyses for Units 1/2 seismic hazard description of the description of the PSHA, and 3. Documentation methodology and results PSHA, an overview including an overview and should be prepared for was not prepared. The and summary of the summary of the overall PSHA the PSHA that meets the lack of a summary report overall PSHA process, process, model uncertainties needs of PRA makes peer review model uncertainties and assumptions, and applications, peer review, difficult, fails to document and assumptions, and reference to intermediate and and future updates. For certain basic PSHA results, fully documents final seismic hazard documents the North Anna PSHA a and lacks reference (i.e., a intermediate and final supporting the North Anna single volume, a PSHA roadmap) to supporting seismic hazard SPRA.

Summary Report, that documents for more products supporting This finding is considered fully describes the detailed explanation of the North Anna SPRA.

resolved and there is no affect methodology that was each element of the on the SPRA results or implemented, the rock analysis.

conclusions.

PSHA results, the site While some elements of response analysis, the overall PSHA process sensitivity studies, etc.

are extremely well-was not prepared. The documented (e.g., CEUS lack of a PSHA Summary SSC model), there is no report makes peer review summary document that:

difficult and could compromise future 1. Summarizes the Page 100 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTIF 2.1 Seismic PRA Submittal March 2018 efforts to understand methodology that was what was done and used in the PSHA, implement future model including the updates. implementation of the PSHA methodology, the (This F&O originated from estimate of GMRS and SR SHA-Jl)

FIRS, the propagation of uncertainties in the analysis, etc.

2. Reports PSHA results for reference rock site conditions and control point motions,

3. Examines the potential for seismic hazards other than earthquake ground motion, and

4. Reports sensitivity analyses.

A summary document, as typically prepared, describes the methodology that is Page 101 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Submittal March 2018 implemented in the PSHA, provides an overview of the elements of the PSHA/GMRS/FIRS process, explains how the elements of the analysis fit together, provides a summary of work performed for each element, provides reference to supporting documents for more detailed explanation, describes the outputs from each element (and, if appropriate, where more complete outputs can be found), and describes, as appropriate, any independent external peer review to which each element was subjected.

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NAPS Units 1 and 2 10 CFR 50.54(f) NTIF 2.1 Seismic PRA Submittal March 2018 SHA-E2, 20-3 Epistemic uncertainty has The analysis of epistemic The site response The geologic conditions at the F3 not been included in the uncertainties in the site analysis should include North Anna site consist of site response analysis. response analysis involved a epistemic uncertainty in saprolitic soils near the surface at limited assessment that shear wave velocity that the site that transition into more concluded these reflects the potential for intact rock at depth. The (This F&O originated from uncertainties were different interpretations weathering across the site is SR SHA-E2) negligible. There are two of the available data. uneven, and the thicknesses of issues that contribute to Alternatively, sensitivity the various material layers within this requirement not being analyses may be the subsurface soil profile vary met: performed that widely and randomly throughout

1. Information on the site demonstrate the the site. The rock layers (zones) insensitivity of the are defined by both rock quality shear-wave velocity profile calculated surface designation (RQD) and shear does not necessarily

. hazard curves to the wave velocity. Considering the support th e cone Ius1on

. * "bl . t . assumed combination original site investigations at th ere 1s neg 11g1 e ep1s em1c

. t yin

. th e s1*t e of epistemic and Units 1 and 2 site and the more unce rt ain aleatory uncertainties. recent site investigations at the response analysis.

proposed site for Unit 3 (which

2. Available documentation shares the same geologic does not present an characteristics), the North Anna analysis of possible site is well characterized and epistemic uncertainties. As extensively investigated with a consequence there is no abundant high-quality data (>200 evidence to support the borings, including five deep conclusion that these borings with P-S suspension uncertainties are neglible. logging velocity measurements),

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NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Submittal March 2018 which reduces epistemic uncertainty in the site properties.

In addition to the foregoing, These data also provide supporting requirement information to characterize the SHA-E2 requires that aleatory variation in layer aleatory and epistemic thickness and shear wave uncertainties be included in velocity across the site. These the site response analysis to variations were included in satisfy Capability Category considerations of aleatory 11/111. uncertainties for the base-case Evaluation of Shear-Wave profile. No alternate profiles Velocity Profiles - The PSHA were considered because of the analysts reviewed the significant amount of recent site measured shear wave specific data and the relative velocity profiles and insignificance of epistemic concluded that: "two shear uncertainty with respect to the wave velocity profiles are aleatory variability for this site.

defined in the power block In addition, the seismic hazard area using 8-901, 8907 and results from the North Anna 8-909. One is for mostly Units 1 and 2 PSHA are unfractured rock consistent with (1) the results of throughout the profile, and the PSHA independently the other is for partially performed for North Anna Unit 3 fractured rock down to [24] and (2) the reuslts of the around El. 184 ft, underlain NRC confirmatory analysis PSHA by the same mostly [16] performed for the review of Page 104 of 181 L

NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Submittal March 2018 unfractured profile" the NAPS SHSR [3].

(Calculation 25161-G-Oll).

This finding is considered Based on the nature of the resolved and there is no effect on surficial geology at the site, the SPRA results or conclusions.

the PSHA analysts subsequently interpreted these two profiles (Profiles 1 and 2) to represent the lower and upper bounds of the aleatory variability in shear wave velocity and calculated the corresponding depth-dependent mean and standard deviation for a single base-case shear wave velocity profile. In arriving at this interpretation, the PSHA analysts did not include epistemic uncertainty in the shear wave velocity profile "because of the significant amount of recent site specific data and the relative insignificance of Page 105 of 181

---

~-----

NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Submittal March 2018 epistemic uncertainty with respect to the aleatory variability for this site."

(25784-000-KOC-0000-00006). The review team believes that it is possible, that other experts may have interpreted the available data differently and arrived at the conclusion that Profiles 1 and 2 represent the epistemic uncertainty in shear wave velocity.

Documentation of the Analysis of Epistemic Uncertainties - The current documentation of the site response analysis does not present an analysis and quantitative estimate of potential epistemic uncertainties that supports a determination and conclusion that they are negligible (e.g., effectively Page 106 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Submittal March 2018 zero, or small enough to be of no engineering significance).

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NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Submittal March 2018 SHA-F2 20-4 An important part of the Per this requirement, Sensitivity calculations Sensitivity calculations have been North Anna PSHA is the analyses should be carried should be performed to performed for the North Anna analysis of site response out to show the influence of illustrate the effect of Unit 3 (NA3) PSHA and site and its impact on the factors that are important the site response on the response analysis as documented ground motion hazard. to the site hazard. With plant ground motions, in the NA3 FSAR [24), Section Sensitivity analyses have respect to the site response effect of alternative site 2.5.2. A study was performed not been performed that analysis, sensitivities have Vs profiles, etc. on the that demonstrated applicability illustrate influence of site not been performed. site response and the of the North Anna Unit 3 PSHA to response on site motions, ground motion hazard. the North Anna Units 1 and 2 including variations in site PSHA based on similarity of velocity profiles, subsurface conditions/ soil interpretations of site profiles, common site location of velocity data, etc. the units, and similarity of the hard rock seismic hazard. As a result, sensitivity calculations (This F&O originated from performed for NA3 are SR SHA-F2) considered applicable for the North Anna Units 1 and 2 site response and ground motion hazard.

This finding is considered resolved and there is no effect on the SPRA results or conclusions.

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NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Submittal March 2018 SHA-11 20-5 As part of the seismic Analyses have been A systematic evaluation A review for other potential hazard analysis, an performed that considers that identifies potential seismic hazards has been evaluation must be the 1) seismic stability of other seismic hazards, performed. Potential seismic performed to assess the dike for the service and performs a hazards of seiche and fault whether the other hazards water reservoir and 2) screening evaluation for displacement are evaluated in that may be initiated by a potential for soil each. the UFSAR and determined not seismic event can be liquefaction. A screening to be credible based on screened out from the assessment for other geographical parameters, which seismic PRA, not inlcuded in potential seismic hazards, are not changed by consideration the assessment of seismic such as fault displacement, of the GMRS. Ground settlement risk, or whether they should ground settlement, seiche near important SSCs with respect be quantitatively evaluated in the reservoir, flooding to the GMRS seismic hazard has and included in the due to dike breach and been evaluated and determined analysis.There are a number uncontrolled release of the to be insignificant. Therefore, of 'other' seismic hazards reservoir, etc. have not these other seismic hazards were that should be considered been performed. screened out.

in the screening analysis.

An analysis of the slope stability of the Service Water Reservoir dike was performed (as indicated (This F&O originated from in the finding basis) and the SR SHA-11) results are included in the SPRA model. Therefore, Service Water Reservoir dike breach as a result of seismic activity has been evaluated.

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NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Submittal March 2018 This finding is considered resolved and there is no effect on the SPRA results or conclusions.

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NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Submittal March 2018 SHA-11 20-8 The analysis used to screen Regarding the potential for As an alternative to an The majority of Category I out liquefaction triggering liquefaction triggering ad-hoc approach, one structures within the power for structures within the based on SPT data for soils possible approach to block area are rock-founded for power block and for buried within the power block and address this issue is to which liquefaction is not a piping lacks rigor. for buried piping, begin with a review of consideration.

Calculation 25784-000-KOC- liquefaction phenomena Th C t t t

• th ree a egory I s rue ures in e 0000-00017 indicates that in residual soils (or lack bl k db . d power oc area an une (This F&O originated from an SPT-based approach thereof) and attempt to . . t t* II

. . piping are comp 1e eIy or par 1a y SR SHA-11) gives minimum FS values build a compelling founded on soil_ either residual less than 1.0 and that a argument that the soils or compacted engineered (lower-bound) Vs-based nature of residual soil backfill underlain by residual approach yields minimum deposits (strong fabric, soils. The soil material supporting FS values below 1.1. An ad- high variability, etc.)

these strucutures is saprolitic hoc argument is effectively precludes material. The engineered backfill constructed that these liquefaction in these consists of excavated saprolitic values are likely not soils.

soil.

accurate because of the nature of the soils at the The saprolite is classified into site. This approach to does two zones based on the not result in a compelling extensive site investigations argument that liquefaction carried out for North Anna Units may be screened out. 1 and 2 construction and for North Anna Unit 3 licensing. The material in these zones are described as:

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NAPS Units 1 and 2 10 CFR 50.54(f) NTIF 2.1 Seismic PRA Submittal March 2018 Zone IIA: Saprolite - medium dense silty sand, with some fine-grained layers Zone 118: Saprolite - very dense silty sand All Zone 1 material (residual clays and clayey silts) was removed during construction. Material underlying the Zones IIA and 118 material consists of zones of weathered to moderately weathered to fresh rock.

As would be expected with these residual Zone IIA and 118 soils, the fabric is that of the parent rock, mainly a biotitic quartz gneiss.

There is strong foliation in the saprolite, dipping at angles of about 50 degrees to the horizontal. The fabric is strongly anisotropic. The texture shows angular geometrically interlocking grains with a lack of void network. The mineralogy Page 112 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Submittal March 2018 also reflects the parent rock, with 30-40 percent quartz, 20 to 30 percent microline, 25 to 40 percent clay minerals, and 5 to 20 percent biotite (mica). The major clay mineral is halloysite (a hydrated form of kaolinite) with lesser amounts of illite and montmorillonite Much of the halloysite is in the form of aggregates that are larger than 2 micrometers and, therefore, would be classified as silt, allowing the sand to be classified as non-plastic. The fabric of the saprolite contrasts strongly with that of an alluvial or marine deposited sand. Such sand shows no foliation and no interlocking of grains, even though the grains can be quite angular. The fabric of saprolite is, therefore, not one of a transported soil but one of the parent rock material. Its age, fabric and interlocking angular grain structure, along with the Page 113 of 181 L

NAPS Units 1 and 2 10 CFR 50.54{f) NTIF 2.1 Seismic PRA Submittal March 2018 significant portion of low plasticity clay minerals present in the material, have been demonstrated to give the grain structure a low susceptibility to pore pressure build-up or liquefaction. This material would not lose a significant proportion of its shear strength during shaking. Although much of the fabric of the saprolite is lost during excavation and subsequent backfilling, some of its interlocking grain structure will remain, providing a low susceptibility of liquefaction of the saprolite fill.

On the basis of the types of soil materials supporting the soil-founded structures and buried piping, liquefaction can be screened out from further consideration.

However, the liquefaction analysis further evaluated the Page 114 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Submittal March 2018 potential for liquefaction based on correlations using blowcounts and shear wave velocity. For these evaluations, several conservatisms were included in determining the application of various correction factors, age factor, water table location (at surface), and material properties.

In addition, the correlations are intended for liquefiable soils and the benefits of fabric and texture of the Zones IIA and IIB soils are not reflected in the calculations.

Consequently, the results (factors of safety against liquefaction

[FS]) were considered conservatively low and in fact some FS were below the lower limit of 1.1. The documentation of the liquefaction evaluation provided a qualitative basis for the conclusion that liquefaction was screened out based on the conservatisms in the application of the correlations.

Page 115 of 181 L

NAPS Units 1 and 2 10 CFR 50.54(f) NTIF 2.1 Seismic PRA Submittal March 2018 Since the documentation provides a basis for screening liquefaction out from further consideration based on the susceptibility of the material alone, this finding is considered resolved. There is no effect on the results or conclusions of the SPRA.

Page 116 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Submittal March 2018 SHA-12 20-9 The analysis of the potential The objective of Section 6 It may be possible to As part of the design, liquefaction-induced of Calculation 2S784-000- address this issue via construction, and the licensing damage to the SWR lacks KOC-0000-00017 is to one or more of the process for NAPS Units 1 and 2, rigor. estimate an appropriate FS following approaches: the Service Water Reservoir against liquefaction for (i) construct a (SWR) and its component subsequent use in compelling argument materials have been subjected to (This F&O originated from calculating a HCLPF capacity that the nature of the extensive subsurface exploration, SR SHA-12) for liquefaction-induced soils at the site and/or laboratory testing, analyses, and stability ofthe SWR dike. the lack of continuity of instrumentation monitoring.

The analyses used to potentially liquefiable Twenty two borings were estimate the appropriate FS zones precludes performed with depths ranged are based on qualitative liquefaction-induced from 27 to 105 ft, and averaged arguments such as instability of the dike 70 ft. Borings used standard "Although the FS is low, it is and/or (ii) if liquefaction penetration test (SPT) sampling very improbable that is assumed to occur, and thin-walled tube samplers.

liquefaction of 10 ft of soil perform a post- The borings encountered fill, at 40 to 50 ft depth could earthquake stability residual soil, and saprolite cause significant settlement analysis using grading to sound rock. The soils or collapse" and "It is appropriately selected underlying the SWR area are possible that liquefaction of residual undrained primarily residual soils and 15 ft of soil at 40 to 55 ft shear strengths to saprolites similar to those depth could cause demonstrate that the encountered at the main plant significant settlement or post-earthquake factor site. As would be expected, these collapse and resulting water of safety is adequate. soils are erratic in terms of loss if the lateral extent of spatial and property distribution.

the liquefiable zone is Typically, silts of low-to-Page 117 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTIF 2.1 Seismic PRA Submittal March 2018 sufficiently large." No basis moderate plasticity (ML and MH) is provided for these are found in the upper portions conclusions, and thus the of the soil profile. These finer selection of FS = 1.34 as the materials grade to coarser-most appropriate value to grained saprolite soils (SP, SM, calculate a HCLPF capacity and SP-SM) which are is not compelling. encountered in the lower portions of the profile. Sound bedrock is found at depths of about 65 ft to 100 ft below original ground surface.

The saprolite is classified into two zones:

Zone IIA - Saprolite - medium dense silty sand, with some fine-grained layers Zone IIB - Saprolite - very dense silty sand All Zone 1 material (residual clays and clayey silts) was removed during construction. Material underlying the Zones IIA and IIB material consists of zones of weathered to moderately Page 118 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTIF 2.1 Seismic PRA Submittal March 2018 weathered to fresh rock.

Materials with N values over about 30 blows/ft will typically not liquefy. Since the N-value for Zone IIB material is generally more than 50 blows/ft, it is not expected to liquefy. The Zone IIA material was further evaluated.

As would be expected with these residual Zone IIA and IIB soils, the fabric is that of the parent rock, mainly a biotitic quartz gneiss.

There is strong foliation in the saprolite, dipping at angles of about 50 degrees to the horizontal. The fabric is strongly anisotropic. The texture shows angular geometrically interlocking grains with a lack of void network. The mineralogy also reflects the parent rock, with 30-40 percent quartz, 20 to 30 percent microline, 25 to 40 percent clay minerals, and 5 to 20 percent biotite (mica). The Page 119 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTIF 2.1 Seismic PRA Submittal March 2018 major clay mineral is halloysite (a hydrated form of kaolinite) with lesser amounts of ii lite and montmorillonite Much of the halloysite is in the form of aggregates that are larger than 2 micrometers and, therefore, would be classified as silt, allowing the sand to be classified as non-plastic. The fabric of the saprolite contrasts strongly with that of an alluvial or marine deposited sand. Such sand shows no foliation and no interlocking of grains, even though the grains can be quite angular. The fabric of saprolite is, therefore, not one of a transported soil but one of the parent rock material. Its age, fabric and interlocking angular grain structure, along with the significant portion of low plasticity clay minerals present in the material, have been demonstrated to give the grain structure a low susceptibility to Page 120 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Submittal March 2018 pore pressure build-up or liquefaction. This material would not lose a significant proportion of its shear strength during shaking. Although much of the fabric of the saprolite is lost during excavation and subsequent backfilling, some of its interlocking grain structure will remain, providing a low susceptibility of liquefaction of the saprolite fill.

On the basis of the types of soil materials supporting the SWR and comprising the construction of the dike, liquefaction can be screened out from further consideration.

However, the liquefaction analysis further evaluated the potential for liquefaction based on correlations using blowcounts and shear wave velocity. For these evaluations, several conservatisms were included in Page 121 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Submittal March 2018 determining the application of various correction factors, age factor, water table location (assumed at surface), and material properties. In addition, the correlations are intended for evaluation of liquefiable soils and the benefits of fabric and texture of the Zones IIA and IIB soils are not reflected in the calculations.

Consequently, the results (factors of safety against liquefaction

[FS]) were considered conservatively low and in fact some FS were below the lower limit of 1.1. The documentation of the liquefaction evaluation provided a qualitative basis for the conclusion that liquefaction potential remained limited based on the conservatisms in the application of the correlations.

Since the documentation provides a basis for screening liquefaction out from further consideration based on the Page 122 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Submittal March 2018 susceptibility of the material alone, the SPRA has been updated to screen out liquefaction from consideration for the SWR.

This finding is considered resolved. There is no effect on the results or conclusions of the SPRA.

Page 123 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Submittal March 2018 SHA-12 20-10 Epistemic uncertainty is not The analyses performed to Ensure that epistemic The SWR slope failure analysis adequately reflected in the estimate mean fragility uncertainty is included has been updated to include mean fragility curves for curves for slope failure of in estimates of the consideration of epistemic slope failure of the SWR the SWR dike and composite beta for the uncertainty and provide dike and liquefaction liquefaction triggering for mean fragility curves for appropriate input to the risk triggering for foundation foundation soils in the SWR slope failure of the SWR quantification.

soils in the SWR area. area are based on values of dike and liquefaction This finding is considered composite beta that do not triggering for resolved and there is no effect on include consideration of foundation soils in the the SPRA results or conclusions.

(This F&O originated from epistemic uncertainty. SWR area that are used SR SHA-12) While appropriate for for risk quantification.

calculating the HCLPF capacity, the resulting mean fragility curves are not appropriate for use in risk quantification.

Page 124 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Submittal March 2018 SHA-J2 20-11 The documentation of the The documentation of the Review the PSHA to These documentation methods and processes PSHA was should be ensure that each of the clarification items will be should be improved in a reviewed and improved at elements of the analysis considered for inclusion in any number of areas. least in the following areas: is fully described, required future revisions of the

1. The documentation including the overall applicable documents.

implementation.

in several calculation* The disposition of this (This F&O originated from packages(25784-000-KOC- documentation-related finding SR SHA-J2) 0000-00006, 25784-000- has no significant impact to the KOC-0000-00013, and SPRA results or conclusions.

25 784-000-KOC-0000-00019) should be revised to make it clear that site-specific modulus reduction and damping curves were used for Zone II and Ill materials.

2. The assumed variation of shear modulus with shear strain for Zones Ill-IV and IV should be more clearly stated.

3. In several calculation packages

{25784-000-KOC-OOOO-Page 125 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Submittal March 2018 00006, 25784-000-KOC-0000-00013, and 25784-000-KOC-0000-00019), plots are presented to verify that the median of the 60 simulated shear wave velocity profiles approximately matches the best-estimate profile.

Similar plots should be developed to demonstrate that the random variability in the simulated profiles is a reasonable approximation to the (assumed) randomness observed in measured shear wave velocity profiles.

4. In Calculation Package 25784-000-KOC-0000-00006, Table 1 provides the mean thickness and standard deviation of thickness for each stratum. Later (p. 19 of 54) it is stated that the Page 126 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTIF 2.1 Seismic PRA Submittal March 2018 variability in stratum thickness is+/- 20%, which is inconsistent with the values provided in Table 1.

5. Figure 11 from Calculation 25161-G-017, Rev. 006, North Anna COL Unit 3 should be included in the documentation of the site response analysis to more completely illustrate the interpretation of available shear wave velocity measurements to derive the best-estimate profile and associated variability.

6. In Calculation Package 25784-000-KOC-0000-00016, the choice of a minimum acceptable factor of safety for pseudo-static slope stability analysis implies some tolerable displacement as indicated Page 127 of 181

NAPS Units 1 and 2 10 CFR S0.54(f) NTIF 2.1 Seismic PRA Submittal March 2018 in Table 10.1 from Reference 16 (provided in Attachment 3). This relationship between factor of safety and tolerable displacement should be acknowledged and discussed in the calculation package.

7. In Calculation Package 25784-000-KOC-0000-00059, it would be helpful to include plots of the critical failure surfaces associated with each case included in Table 11.

8. Given the nature of the soil profile consisting of weathered material of varying thickness, the assumption that a one-dimensional site response is appropriate should be discussed and justified.

9. The implementation Page 128 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Submittal March 2018 of Approach 3 to combine the hard-rock seismic hazard curves with the site amplification functions should be documented in greater detail, particularly with respect to the use of fractile hard-rock hazard curves rather than the suite of individual hard-rock hazard curves.

Page 129 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Submittal March 2018 SHA-J3 20-12 A foundational element of Within in the scope of the Conduct a systematic Documentation of the North PSHA - as it has evolved PSHA there are sources of review and evaluation Anna PSHA has been enhanced over the past 30 years - is uncertainty that are not to identify (e.g., a to provide a complete the development and directly modeled and tabular summary) and description of the PSHA, implementation of methods assumptions that are made document (e.g., including an overview and to identify, evaluate, and for pragmatic or other discussion) sources of summary of the overall PSHA model sources of epistemic reasons. There are also model uncertainty and process, model uncertainties and (model and parametric) sources of model analysis assumptions in assumptions, and reference to uncertainty in the estimate uncertainty that are the PSHA. The intermediate and final seismic of ground motion hazards. embedded in the context of documentation of hazard documents supporting These methods look at the current practice that are uncertainties and the North Anna SPRA.

epistemic uncertainties 'accepted' and typically not assumptions should Site conditions for North Anna associated with data, subject to critical review. provide the SPRA Units 1 and 2 are consistent with models and methods that For instance, in the PSHA it analysts with insight the use of standard practice in could contribute to the is standard practice to and guidance as to modeling and analysis for the uncertainty in elements of assume that the temporal elements of the SHA.

PSHA. In addition, the seismic the PSHA. occurrence of earthquakes hazard results from the North is defined by a Poisson This supporting Anna Units 1 and 2 PSHA are process. This assumption is requirement states sources consistent with (1) the results of of model uncertainty and well accepted despite the the PSHA independently

. t b fact that it violates certain assump t ions mus e performed for North Anna Unit 3

• t d fundamental understanding d ocumen t ed . Wh at I oes [24] and (2) the reuslts of the

• th t . t of tectonic processes (strain not say, 1s ese op1cs mus NRC confirmatory analysis PSHA

. accumulation). A second b e d ocumen t ed in a [16] performed for the review of practice is the fact that manner th at suppo rt s th e the NAPS SHSR [3].

earthquake aftershocks are Page 130 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTIF 2.1 Seismic PRA Submittal March 2018 SPRA analysts ability to not modeled in the PSHAI The uncertainties in the PSHA are assess whether identified even though they may be ultimately captured in the hazard sources of uncertainty or significant events curve distribution (mean 16th assumptions may have (depending on the size of soth, 84th) that is used in ~he '

important implications to the main event).

parametric uncertainty analysis estimates of plant risk.

to estimate the distribution of the SPRA results due to In the spirit of this variability in the SSC seismic (This F&O originated from requirement it seems failure probabilities and seismic SR SHA-J3) appropriate that sources of hazard initiating event model uncertainty that are frequencies. The parametric modeled as well as sources uncertainty analysis uses the of uncertainty and EPRI UNCERT code that employs associated assumptions as the Monte Carlo technique to they relate to the site-generate random samples for specific analysis should be each probabilistically-varying identified/discussed and event and to quantify the their influence on the uncertainty distribution. The results discussed. The parametric uncertainty analysis model uncertainties and would be expected to encompass assumptions in a PSHA fall the effects of model uncertainty into the following and analysis assumptions that categories:

are not explicitly modeled since they are part of the standard-of-

1. Uncertainties that practice in the PSHA.

Page 131 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Submittal March 2018 are explicitly identified and This finding is considered modeled in the PSHA logic resolved and there is no effect on trees the SPRA results or conclusions.

2. Methods, sources of uncertainty or modeling assumptions that are not explicitly modeled since they are part of the standard-of-practice in PSHA (i.e., earthquake occurrence modeling), site response analysis, etc.

3. Detailed modeling assumptions that are made as part of specific calculations (e.g.,

liquefaction assessment, slope stability failure criterion).

The PSHA documentation addresses, at least in part, Page 132 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTIF 2.1 Seismic PRA Submittal March 2018 Items 1 and 3. What the available documentation does not do is provide a comprehensive summary of the model uncertainties and assumptions in the PSHA and insight to their possible implication to estimates of plant risk.

Page 133 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Submittal March 2018 SFR-ES 23-4 During the peer review It is unclear from the Review the walkdown With the exception of Item (b),

team walkdown performed available SPRA documentation for each of the listed items identified on 7/18/17, several documentation whether consistency across during the PRT walkdown were potential interaction the following potential various teams, dispositioned at the time of the sources were identified that interaction sources, which equipment types, peer review. When appropriate, were not in the SPRA were identified during the locations, etc., to assess document updates have been walkdown documentation. PRT walkdown, are whether the kinds of completed to address omissions significant to the SPRA: issues identified here and/ or document dispositions are contained to a provided. Item (b) identified one However, other SEWS forms limited extent within of a few SEL items that were not did identify potential a) Round duct in quench the documentation. walked down prior to the peer interaction sources, often in spray pump house that has Supplement the review. These items were detail. Therefore, it is clear fixed supports on either walkdowns and identified in the walkdown that interaction was side of a building joint. documentation as summary report. Since the peer considered, but for a Potential for seismic anchor necessary to provide review, walkdown inspections number SSCs, some motion across the building confidence that the have been completed for those potential interactions were separation joint was not review for potential SEL items missing walkdown not documented, and their documented in the SEWS or interactions was documentation including the disposition is likewise not evaluated subsequent to comprehensive. transmitters on Rack 1-802. In documented. walkdown. addition, subsequent walkdowns b) Rack 1-802 supports were conducted in various plant some instruments that have areas containing SEL equipment (This F&O originated from attached lines anchored to (with a particular emphasis on SR SFR-ES) the safeguards building* the Emergency Switchgear/

wall, while the rack is Instrument Rack arid Relay Rooms) to confirm the adequacy Page 134 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Submittal March 2018 anchored to the floor of the of walkdowns performed. There quench spray pump house. were no obvious deficiencies in The potential for seismic terms of identifying seismic anchor motion was not interactions.

identified on walkdown This finding is considered SEWS or evaluated resolved. There is no effect on subsequent to walkdown.

the results or conclusions of the c) The lH 4kV SPRA.

switchgear is in close proximity to a neighboring computer rack, 1-EI-CB-301A. The proximity was not noted on the SEWS or evaluated subsequent to the walkdown.

d) On both lineups of the lH 4kV switchgear, there is a copper bus bar on each end of the lineup that is flexible and free to slap against the side of the cabinet during an earthquake. This potential interaction was not noted on the SEWS or evaluated Page 135 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTIF 2.1 Seismic PRA Submittal March 2018 subsequent to the walkdown.

e) Inverter 1-VB-INV-02:

There is about a 3/8" to 1/2" gap between the inverter and a unistrut that is attached to a neighboring cabinet. The proximity issue was not noted on the inverter SEWS or evaluated subsequent to the walkdown.

f) There is a mobile CO2 firefighting cart located near the 1 EP CB 28A relay cabinet. It appears if the cart overturns, it could hit the cabinet and potentially affect the function of the relays. The potential interaction was not noted in the walkdown SEWS provided to the peer review team. The SPRA team indicated during the peer Page 136 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Submittal March 2018 review week that this issue was identified during the walkdown and subsequently evaluated and dispositioned but inadvertently omitted from the walkdown documentation.

g) A clamping mechanism on top of Relay Cabinet 1-EP-CB-28Ais 1/16 in. away from the top of cabinet 1HC-H2A-101 at the end of the lineup. The potential interaction was not noted in the SEWS or evaluated subsequently.

Page 137 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Submittal March 2018 SFR-F3, 23-5 Some numerical errors Page 3 of 9, Cabinet 1-EE- Correct the numerical The identified numerical error Gl were found in the relay BKR-15H8, Relays errors and verify the has been corrected and computations in Appendix B 12HFA151A2F, change does not affect documented. Calculated relay of Position Paper 9 for relay 12IJCV51B23A and the SPRA results. fragilities were only minimally fragility group SEIS-EE-BKR- 12PJC11AV1A, the demand changed as a result. The HJ8-RLY. acceleration should be identified error had no impact on 1.8225g instead of 1.821g SPRA results. An extent of and the associated Beta U condition assessment identified (This F&O originated from should be 0.2961 instead of no other similar errors.

SR SFR-Gl) 0.253.

This finding is considered resolved. There is no effect on the results or conclusions of the It appears that making this SPRA.

correction should not affect the SPRA results because these relays are non-governing. Therefore, this appears to be a documentation issue only.

Page 138 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Submittal March 2018 SFR-C4 23-6 A slightly conservative EPRI TR-103959 states that JUSTIFY how the The median concrete elastic estimate of median minimum specified 28-day conservative approach modulus is calculated using ACI concrete compressive strength should be used to used by NAPS to formulation after aging strength is used to calculate estimate structure evaluate the concrete considerations. The specified structure stiffnesses. stiffnesses for concrete compressive strength design compressive strength of According to EPRI TR- shear wall structures. It yield realistic structural concrete for the NAPS Units 1 103959, this approach is states that stiffnesses can loads and floor and 2 structures is 3000 psi. Test expected to overestimate alternatively be estimated response spectra for use results from 2032 cylinder test structure stiffnesses. using 0.7 times the median in the seismic PRA. specimens taken across different concrete strength. If either An appropriate structures at the site early in of these accepted industry- sensitivity could be plant life showed that 67.5% of (This F&O originated from standard approaches are performed to show that the specimens had 28-day SR SFR-C4) used, it is expected that the th II . t f strength of more than 4500 psi, e overa 1mpac o structure stiffnesses would t . th and, thus, the median strength is no using e be significantly lower than mentioned industry higher than 4500 psi. Therefore, those used in the NAPS accepted approach for the use of 3000 psi compressive SPRA. The stiffness change t'ff t* . strength would lead to un-s I ness ca 1cu 1a 10n 1s could affect structure forces t . 'f' ti conservatively low estimate of no s1gni 1can y and ISRS, both in amplitude impacting the final uncracked structural stiffness for and peak frequencies. fragility calculation. North Anna structures. On the other hand, recent research has shown that typical shear walls are flexible compared to the stiffness representations given in ASCE 43-05. Considering both of the above points, for the North Page 139 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Submittal March 2018 Anna SPRA effort and for use with the ASCE 43-05 formulation for elastic modulus, E, the median 28 day strength of concrete is judged to be 4500 psi and a 70% value of E and modulus of rigidity, G, need not be considered. Per recommendations of EPRI TR-103959, an aging factor of 1.2 is applied to obtain the median strength of concrete as f'c = 5400 psi. Thus, the concrete elastic modulus is calculated as Ee=

57000v( = 4189 ksi or 603,200 ksf.

This finding is considered resolved. There is no effect on the results or conclusions of the SPRA.

Page 140 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Submittal March 2018 SFR-A2, 23-8 Fragilities for some The PRT reviewed a sample Realistic fragilities Detailed fragility calculations for F2 significant SLERF of the significant should be developed for the Auxiliary Feedwater Pump contributors are not contributors to SCDF and the SSCs that are House and the Emergency realistic. If the fragilities SLERF. As defined in the significant contributors Condensate Storage Tank that 1 1 are refined, the SPRA PRA notebooks, significant to seismic risk. provide more realistic inputs to results and insights could SSCs have Fussel-Vesselly the SPRA have been performed be affected substantially. importance of 0.005 or and the results have been greater. Some significant incorporatd into the SPRA.

SLERF contributors are not Sensitivity studies have been (This F&O originated from realistic. performed for the Containment SR SFR-F2) and Service Water Valve House fragility values have been For example, the following performed to determine the significant structure and effect on the SPRA results.

tank fragilities are Although higher fragility values computed based on the provide some SPRA results screening level capacities in improvements, the changes are EPRI NP-6041 Tables 2.3: not significant.

- Reactor Containment Motor-operated valve (MOV) and Building the MS PORV fragility evaluations

- Auxiliary Feed Water have been refined where Pump House possible and the more realistic results have been incorporated

- Emergency Condensate into the SPRA.ln addition.

Storage Tank This finding is considered Page 141 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Submittal March 2018 resolved. The results of

- Service Water Valve House improved fragility evaluations have been incorporated into the Additionally, all the MOVs SPRA and minor improvements in that are credited in the SCDF and/or SLERF were realized.

SPRA (some of which are This finding is considered significant to SLERF) are resolved.

assigned a HCLPF of 0.6g based on the limiting fragility value of all those MOVs. This is conservative for most MOVs.

The PRT also reviewed the MS PORVs fragility as part of the sample review. The PORVs are the #5 top SLERF contributor according to Table 3-12 in Notebook SA.1. The HCLPF is 0.32g and appears to be based on a 1.8g generic spectral capacity. This is probably conservative for this valve, and if a component-specific Page 142 of 181

NAPS Units 1 and *2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Submittal March 2018 evaluation were performed based on plant-specific qualification levels or component-specific stress analysis, this fragility could likely be significantly refined.

Page 143 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Submittal March 2018 SFR-A2 23-10 When fragilities are The HDPRV factor in one For SOV calculations When functional fragilities are developed based on direction is the inverse of where the geomean of derived using EPRI NP-6041 SL-capacities from EPRI NP- the other direction such two horizontal Rl, Table 2-3 or 2-4, the use of 6041 Table 2-3 or 2-4, those that the geomean of the directions is used to geometric mean of the two capacities are interpreted factors in the two directions characterize demands, horizontal spectral peaks is as a geomean of two is always 1.0, and there is the HDPRV should be 'udged to be reasonable and will horizontal directions. no variability on this omitted from the be about the same as using the Accordingly, demands are geomean factor. variability calculations. arithmetic average of the two likewise characterized as For CDFM calculations, horizontal spectral values which geomean of two horizontal the 84 % NEP demands is recommended on p. 2-44 of directions for comparison should be adjusted to EPRI NP-6041 SL. An exception is to these capacities. when one direction clearly remove HDPRV.

governs; the spectral In these cases, horizontal Alternatively, assess accelerations in that direction direction peak response whether these changes were used.

variability (HDPRV) should might significantly affect not be included in the SPRA, and adjust the The majority of fragility separation of variables fragilities only as calculations for functional failure (SOV) or CDFM HCLPF mode using NP-6041-SL, Tables necessary to ensure calculations. Including 2-3 and 2-4 were performed with meaningful results and HDPRV in these cases the CDFM approach. In these insights.

conservatively calculations, 84% ISRS from the overestimates aleatory As another alternative, response analyses were used and variability in SOV compare the current the variabilities were used from calculations, and fragilities to the the SPID; thus HDPRV was not conservatively alternate, more considered explicitly for any Page 144 of 181 L

NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Submittal March 2018 overestimates 84% NEP conventional approach direction. Therefore, for CDFM demands for CDFM

• in which EPRI 6041 calculations, the 84% NEP calculations. capacities are compared demands need not be adjusted to the maximum since explicitly calculated (This F&O originated from direction response variabilities from structural SR SFR-A2) rather than the response were not used.

geomean. Assess When an SOV analysis is whether the current performed, the use of geometric fragilities as calculated mean could slightly overestimate can be justified based the aleatory variability since the on this comparison.

variabilities due to structural response were explicitly calculated. However, since the composite variability remains the same, a small redistribution of the aleatory and epistemic variabilities is judged not to affect the fragility curve significantly. In addition, the top risk contributor SSCs, where the SOV approach was used, include several relays; however, in these analyses the governing horizontal direction was used rather than the geometric mean. Other risk-significant components using the Page 145 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Submittal March 2018 SOV approach are the vital inverters where geometric mean was used. A review of this calculation shows that the anchorage controls the fragility and not the function (which was based on NP-6041 Table 2-4) therefore, there is no impact.

Other SSCs where SOV calculations were performed are not among the top contributors to risk. Therefore, if HDPRV was removed, the effect on the SPRA results and risk insights would be negligible.

This finding is considered resolved. There is no effect on the results or conclusions ofthe SPRA.

Page 146 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTIF 2.1 Seismic PRA Submittal March 2018 SFR-A2 23-11 The 0.6g HCLPF for the The ECST IPEEE HCLPF was The fragility should be A realistic fragility analysis has Emergency Condensate relatively low and identified refined to realistically been performed for the Storage Tank (ECST) is not as important in that characterize the Emergency Condensate Storage adequately justified as evaluation. Similarly, a dynamic response of Tank (ECST) missile shield representative of the fragility evaluation was the tank and the structure and the HCLPF and realistic failure behavior. performed for the SPRA, progression of failure. median fragility of the structure and it was likewise Alternatively, a is no longer based on EPRI NP-(This F&O originated from relatively low and sensitivity could be 6041 Table 2-3 for reinforced SR SFR-A2) important. The fragilities performed to assess concrete shear wall structures.

were based on failure of the how sensitive the SPRA The HCLPF capacity for the steel tank, and did not results are to structure is greater than lg.

address additional strength assumptions regarding The seismic fragility evaluation or dynamic influence of the progression of for the steel tank concluded that connection to concrete failure and the overturning and sliding were the missile shield. concrete's ability to governing failure modes. This retain the fluid.

The final fragility that is fragility analysis is not a realistic used to represent the ECST representation of the failure of in the SPRA is based on the function of the ECST since failure of the concrete the steel tank is completely missile shield. The fragility surrounded by a 2-foot thick is based on EPRI NP-6041 reinforced concrete missile shield Table 2-3 for reinforced that would restrict sliding or concrete shear wall overturning. Additionally, in the structures. Table 2-3 event of a breach of the pressure indicates penetrations must boundary of the tank within the Page 147 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Submittal March 2018 be evaluated, but there is missile shield, significant no documented evaluation inventory loss would not be of the penetrations in the expected since the reinforced ECST fragility calculation. It concrete shield is essentially a is not clear from the monolithic structure and available documentation penetrations through the shield whether the concrete shield are sealed by caulking or grout.

wall is capable of retaining For this case, the missile shield the ECST contents in case would function as the tank the steel tank fails. The pressure boundary and the penetrations, for example, limited displacement of the steel are sealed with elastomeric tank within the shield would not sealant, and there is no prevent the flow of tank contents evaluation whether this through the connected piping.

sealant would remain intact The mission time for the use of if the ECST were to fail. the tank contents is relatively short such that a small amount of The lower fragilities leakage through the shield representing failure of steel penetrations would not tank probably significantly affect available tank underestimate the actual inventory or the function of the fragility since they do not tank to provide an adequate credit the support provided water source to the Auxiliary by the concrete shield wall.

Feedwater System pumps.

The fragility representing shield wall failure, however, Therefore, the seismic fragility of may be unconservative the reinforced concrete missile Page 148 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Submittal March 2018 because it does not shield structure provides a adequately address the realistic representation of the capability of the shield wall ECST seismic fragility and was to retain the ECST contents used as the input to the SPRA in a useful state (e.g., no model.

documentation of the This finding is considered capability of the resolved. The results of the SPRA penetrations to retain the are improved slightly by the fluid).

refined fragility analysis of the ECST missile shield structure.

Page 149 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTIF 2.1 Seismic PRA Submittal March 2018 SFR-A2, 24-2 The Position Paper 12_RO Based on the review of Estimate realistic failure Additional failure modes have Dl, F2 (Service Bldg) evaluates plant design documents and modes for the Turbine been evaluated for the Turbine only one failure mode for observations made during Building (including Building (TB) and the the Turbine Building. the walkdown, there are potential seismic consequences have been additional failure modes for induced flood sources) characterized. The seismically-the Turbine Building which and characterize its induced structural damage within (This F&O originated from are not identified in the consequence. the TB has been evaluated to SR SFR-Dl) Seismic PRA model. determine the potential for significant flooding, fires, and toxic chemical releases that could adversely affect the function of core damage mitigating equipment or main control room (MCR) habitability.

The TB has been modeled using the finite element method and a linear dynamic analysis of the TB response to seismic ground motions has been performed.

The seismic ground motions were based on the re-evaluated seismic hazard, or Ground Motion Response Spectrum (GMRS), used for the seismic PRA.

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NAPS Units 1 and 2 10 CFR 50.54(f) NTIF 2.1 Seismic PRA Submittal March 2018 Results of the TB linear dynamic analysis were used, along with building walkdowns and design documentation reviews, to identify locations for first failure in the TB steel superstructure under GM RS-level loading conditions. Based on this information, a qualitative evaluation was made to determine bounding modes of failure for the TB. Two significant bounding modes of failure were evaluated for flooding, fire and impact on toxic chemical release; (1) complete collapse of the TB roof truss supporting structure and (2) derailment of the TB Unit 1 and 2 overhead bridge cranes resulting in crane free-fall to the TB operating deck.

Flooding: These bounding structural modes of failure were evaluated for their potential to damage systems that would Page 151 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Submittal March 2018 constitute significant flood volume sources. (1) For the postulated failure of the TB roof truss supporting structure and its subsequent collapse onto the TB operating deck, it was concluded that the relatively lightweight roof truss members would not penetrate the concrete TB operating deck or cause collapse of the TB operating deck supporting structure. Since there are no significant flood sources on or above the operating deck, there were no flooding consequences identified from this bounding failure mode. (2)

For the derailment of the overhead bridge cranes, each crane was assumed to free-fall to the operating deck as a result of the seismic motions. Significant flooding sources were identified in the Unit 1 TB, but are located in the basement and are west of the projected Unit 1 overhead Page 152 of 181 L_ - - - - -

NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Submittal March 2018 crane fall path. No significant flooding sources are located near the fall path of the Unit 2 overhead crane. The bounding failure mode evaluation assumed that the Unit 1 overhead crane would derail and fall to the operating deck below with the north end of the crane passing through a large opening in the deck and coming to rest in the truck bay below. It was concluded that the TB operating deck would withstand the impact of the Unit 1 overhead crane with only local member damage and further progressive collapse of the TB operating deck would not occur. This conclusion was based on the substantial steel framing and thick reinforced concrete slab construction of the TB operating deck, which is designed to support heavy turbine dismantling/ laydown equipment loads. Additionally, Page 153 of 181 L

NAPS Units 1 and 2 10 CFR 50.54(f) NTIF 2.1 Seismic PRA Submittal March 2018 Operating Experience from the AN0-1 stator drop (03/31/2013) was reviewed and provided support for the conclusion that operating deck damage would be limited to local member failure and not create a progressive collapse scenario. A review concluded that systems that constituted significant flood sources were located to the west of the fall zone of the Unit 1 overhead crane and that there would not be significant collateral damage to the TB operating deck from the postulated overhead crane drop that could adversely affect the water systems. Therefore, there were no flooding consequences identified from this bounding failure mode.

Fire: There are systems in the TB that contain flammable materials, such as hydrogen for main generator cooling and Page 154 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTIF 2.1 Seismic PRA Submittal March 2018 turbine lubricating oil. Structural damage in the TB could result in a breach of these systems and resulting fires. The evaluation concluded that firefighting would prevent the spread of fires to safety-related areas and that these areas are protected by fire-rated walls and doors.

Therefore, there were no fire-related consequences from TB structural damage.

Toxic Chemical Release/ MCR Habitability: There are systems in the TB that contain toxic chemicals. Structural damage in the TB could result in a breach of these systems and result in a release of toxic chemicals to the environment, potentially affecting MCR habitability. The evaluation concluded that based on the limited amount of chemicals in the TB, and the manual initiation of MCR isolation by the operators in the Page 155 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Submittal March 2018 event that a toxic atmosphere is detected, MCR habitability would not be affected. Therefore, there are no consequences of toxic chemical release due to TB damage.

Ba*sed on the evaluation of the effects of seismically-induced TB damage, this finding is considered resolved. There is no effect on the results or conclusions of the SPRA.

Page 156 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Submittal March 2018 SFR-C4 24-3 Median concrete damping The stress levels in most of Provide justification for A median damping ratio of 5%

of 5% is assigned for the the category I concrete the use of 5% median was used for concrete materials, seismic response analysis. structures are very low and damping for the per Table 3-4 of EPRI TR the assigned damping is not concrete structures 103959. This is based on

. consistent with the stress which does not undergo demands at approximately Yi the (This F&O originated from levels experienced by these cracking. yield strength for reinforced SR SFR-C4) structures. concrete with cracking. This value is also consistent with Table 4-1 of EPRI NP-6041-SL, Rev. 1, which recommends 5% damping for reinforced concrete with moderate cracking.

If higher demands are observed based on a best estimate evaluation, a higher damping ratio of 10% for reinforced concrete could be justified along with the use of cracked properties for concrete.

It is noted that more recent design codes such as ASCE 4-98 and ASCE 43-05 recommend the use of 4% damping ratio for uncracked concrete. While this value is appropriate for design, it is considered to be a Page 157 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Submittal March 2018 conservatively biased estimate of the median damping. Because the goal in the analysis supporting SPRA is to obtain an unbiased estimate of the median response, the use of a slightly higher damping ratio of 5% is considered appropriate.

Furthermore, the shear demands in major concrete shear walls of Service Building and Auxiliary Building were evaluated and found to be generally between 1.5 to 3 square roots of f'c. These levels of stress in concrete shear walls are considered consistent with the adopted median damping ratio o 5%.

It is also noted that some SPRA practitioners have used 7% or possibly higher concrete damping values in their structural dynamic analyses. For instance, a concrete damping of 7% was used for the Page 158 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Submittal March 2018 Watts Bar SPRA (Ref. NRC ADAMS Accession number ML17181A485) which has a lower GMRS than North Anna. Watts Bar assumed Damage Level 2 of ASCE/SEI Standard 43-05, 2005 and considered even the 7% damping somewhat conservatively biased relative to the likely damage state associated with the median seismic capacities of the SSCs.

Thus the 5% structural damping used in dynamic analyses of structures is appropriate.

This finding is considered resolved. There is no effect on the results or conclusions of the SPRA.

Page 159 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Submittal March 2018 SFR-C6 24-5 The 551 analysis of the The report does not provide Perform a sensitivity A sensitivity study has been embedded structures are *ustification on the study to validate the performed to compare the soil performed using Modified adequacy of the MSM used results from MSM by structure interaction (551)

Subtraction Method (MSM). for the embedded 551 comparing with those of analysis results using the analysis. Direct method or modified subtraction method Surrogate for the Direct (MSM), also referred to as the (This F&O originated from method. extended subtraction method SR SFR-C6) (ESM), and the direct method.

The study results showed no significant differences for the two representative structures studied.

This finding is considered resolved. There is no effect on the results or conclusions of the SPRA.

Page 160 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Submittal March 2018 SFR-G2 24-8 Additional documentation The statement 'This Incorporate the changes The administrative error in and corrections to the calculation utilizes an to CCW report and the calculation DMNNA023-CALC-existing reports should be unverified assumption fragility summary table 006 has been corrected.

included. about the accuracy of the spreadsheet.

provided in-structure response spectra, see Attachment 3 - SSC Fragility Attachments 1 and 2.' on Responses to the Block Summary Table of NAPS SA.5 has pages 4 and 41 of Walls screening been updated.

DMNNA023-CALC-006 Rev. approach, Collapse of 0 was inadvertently left in fuel building and spent the calculation from an fuel pool, Incoherency The documentation associated Modes, Sliding and with block wall evaluations has earlier draft and should have been removed. Overturning failure been updated to include a modes of the structures discussion of the approach for were provided during identifying block walls that could In Attachment 3 - SSC the on-site review and impact distributions systems Fragility Summary Table of should be documented. (including their support if they the calculation NAPS SA.5 are mounted on a block wall).

RO, the revision/version numbers of the reference The seismic capacity of the spent calculations should be fuel pool has been evaluated and updated for all the SSCs.

documented in accordance with the guidance in EPRI 3002009564 Section 2.2.8 of the report [15]. The SPRA documentation (NAPS SA.4 Rl) states that has been updated to reflect the Page 161 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Submittal March 2018

'Of the 481 block walls in conclusions ofthat evaluation.

the plant, 34 walls have been identified for further evaluation if their failure Ten (10) incoherency modes could impact mitigating were used with SRSS SSCs or operator pathways'. combination for the computation When pre-screening block of the 551 response due to walls, the documentation incoherent input ground motion.

should also discuss the While the number of incoherent approach adopted for modes selected was discussed identifying block walls that during the in-process peer could impact distributions review, no sensitivity studies on systems (including their the number of incoherent modes support if they are mounted were suggested or performed.

on a block wall). Based on the structural analyst's past experience with similar models and foundation The report (NAPS SA.4 Rl) dimensions, only the first few discusses the sloshing of incoherency modes have the water in the spent fuel significant contribution to the pool that could result in solution and the use of 10 water "spilling" out of the incoherency modes is considered pool and propagating into adequate.

the Auxiliary building basement via the pipe tunnel between the two Because of the high frequency Page 162 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Submittal March 2018 buildings. However, it is nature of the input motion at the unlikely there would be site, and small building enough water from the SFP displacements calculated at the sloshing that would result in GMRS level of input, for the submergence damage. The structures where specific fragility report does not consider calculations were performed, the the failure of the spent fuel sliding and overturning modes pool itself. The failure for the structures were judged modes (collapse of the fuel not to be governing. For building or the spent fuel structures where the fragility was pool) and its consequence determined from EPRI NP-6041 should be evaluated and Table 2-3, the capacity is based documented. on the information in the table and no specific failure modes were evaluated.

For all the SSI analysis that included ground motion incoherency, the reports This finding is considered does not provide resolved. There is no effect on

• ustification for the use of the results or conclusions of the 10 incoherency modes. SPRA.

The sliding and overturning failure modes for the structures have not been Page 163 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Submittal March 2018 evaluated. The basis for not including them as credible failure modes is not documented.

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NAPS Units 1 and 2 10 CFR 50.54(f) NTIF 2.1 Seismic PRA Submittal March 2018 SMU-B3 25-3 There's no explicit Supporting guidance Revise NF-AA-PRA-410 As a matter of practice, PRA instruction to make model document NF-AA-PRA-4040 Revision 8, Step 3.6.1 to model changes are developed changes in accordance with Revision 2 addresses the include ASME/ANS RA- and documented to meet the RG 1.200, Rev. 2 and PRA Model of Record Sa-2009, ASME/ANS RA- requirements of the ASME/ANS ASME/ANS RA-Sa-2009. Revision process. This Sb-2013, and NRC Reg. PRA Standard. But the PRA document primarily Guide 1.200, Revision 2. procedure, as noted by the peer describes the work scope review, lacked specific guidance (This F&O originated from management process in for ensuring this. The PRA SR SMU-B3) making model changes. The procedure was revised to include PRA model elements of guidance for revising the PRA in ASME/ANS RA-Sa-2009 are accordance with the ASME/ANS listed in definition item PRA standard.

5.3.10. However, there's no This finding is considered clearly stated direction for

• resolved. There is no effect on the PRA engineer to the results or conclusions of the evaluate the model change SPRA.

in light of these PRA model elements and the governing document ASME/ANS RA-Sa-2009. The intent is clear but a set of instruction steps are missing.

SMU-B4 25-4 There's no explicit There's a gap in the Revise NF-AA-PRA-410 As a matter of practice, PRA instruction to review model configuration control Revision 8, to include model changes are reviewed to changes to distinguish process where a peer steps to review the identify changes that are Page 165 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Submittal March 2018 between a PRA upgrade review will be required. model impact considered upgrades. A list of and an update. Furthermore, there's an classification and upgrades is maintained to track Furthermore, there's no increased emphasis on determine if it will resulttheir status with respect to guidance on when to follow-on peer reviews in in an upgrade or an undergoing a peer review. But require a peer review in order to fully implement update. Also, include a the PRA procedure, as noted by accordance with this SR. the guidance in NEI 05- step that stipulates a the peer review, lacked specific 04/07-12/12-06 Appendix PRA model upgrade guidance for ensuring this. The X: Close Out of F&Os. requires a follow-on PRA procedure has been revised (This F&O originated from peer review. to include guidance for reviewing SR SMU-B4) model changes to distinguish between upgrades and updates.

This finding is considered resolved. There is no effect on the results or conclusions of the SPRA.

SMU-El 25-6 There is no software quality According to the subject Develop an SQA PRA codes used in the assurance (SQA) report that matter expert PRA Software Code File in development of PRA models at documents the impact engineer, the version of accordance with Dominion are maintained under assessment, classification, FRANX used for the NAPS Administrative the Software Quality Assurance and verification/validation SPRA is 4.3. Testing was Procedure IT-AA-SQA- (SQA) program. As noted by the testing of FRANX 4.2 performed on FRANX 4.3 101 for the version of peer review team, the SQA code applied specifically for the but an SQA report has not FRANX that quantifies file for the FRANX version used in seismic PRA quantification. been issued for this version. the NAPS SPRA of the development of the SPRA Moreover, he stated that record. In addition, was not up to date. Subsequent FRANX Version 4.4 is ensure all relevant to the peer review, the SQA code Page 166 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Submittal March 2018 expected to be released documents the state the file was updated to match the (This F&O originated from soon and they will be version of FRANX in use. the version used in the SPRA SR SMU-El) updating the FRANX SQA development file for that version.

This finding is considered resolved. There is no effect on the results or conclusions of the SPRA.

SPR-E2, 25-8 Several sources of Many sources of modeling Review all the The sources of uncertainties E6 uncertainty are identified in uncertainties are found uncertainties listed in were updated and sensitivities the NAPS SPRA that are not unaddressed when Tables 7-1 and 7-2 of added to the Seismic addressed in the section, comparing the sources of SA.3. For the Quantification notebook as Sensitivity Studies Section modeling uncertainties uncertainties that are needed.

4.2 of NOTEBK-PRA-NAPS- identified in the NAPS SPRA not addressed with a This finding is considered SA.1 Revision 0. with those addressed in sensitivity case in resolved. There is no effect on section 4.2 of NOTEBK-PRA- section 4.2 of SA.1, the results or conclusions of the NAPS-SA.1 Revision 0. provide a sensitivity SPRA.

(This F&O originated from Specifically, the sources of case or document the SR SPR-E2) uncertainty were gleaned reason why a sensitivity from NOTEBK-PRA-NAPS- case is not necessary to SA.3 Revision 0, Table 7-1 satisfy SR QU-E4.

(Generic Sources of Uncertainty for Seismic PRA) and Table 7-2 (Plant-Specific Assumptions and Uncertainties). These Page 167 of 181 L

NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Submittal March 2018 uncertainties are compared to the Sensitivity Studies section (Section 4.2) of SA.1 Reva.

Per SR QU-E4, the above sources of uncertainty need to be addressed and their potential impact assessed.

SPR-Bl 25-9 The SPRA model builds on NOTEBK-PRA-NAPS-SA.3 Perform a focused peer The Flowserve RCP seal model an interim Internal Events Revision 0, Section 3.1, review and document in upgrade has not yet been peer PRA model that includes the Table 3-1, lists RCP low accordance with reviewed. However, th~

modeling of low leakage leakage Flowserve seal ASME/ANS RA-Sb-2013 Flowserve seal model in the RCP Flowserve seals. This is modeling as included in the 1-6.2.4 and 1-6.6, North Anna PRA (and SPRA) is a PRA model upgrade that SPRA in the Loss of RCP Seal respectively. nearly identical to the Flowserve according to ASME/ANS RA- Cooling Internal Events seal model in the Surry PRA, Sb-2013, Non mandatory Event Tree. However, the which had undergone a peer Appendix 1-A, requires a Internal Events PRA review in 2013. The F&Os from focused peer review prior notebook NOTEBK-PRA- the Surry peer review of the seal to crediting. NAPS-AS.1 Revision 5, model were reviewed for Page 168 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Submittal March 2018 Accident Sequence Analysis, applicability to the North Anna states that the logic for the PRA seal model. The conclusion (This F&O originated from Flowserve seals is disabled is that the F&Os either are not SR SPR-Bl) until the seals are replaced applicable to the North Anna seal in all RCPs. model or they have no impact on the results.

This F&O will remain as unresolved until a peer review is performed. However, the SPRA results are not impacted.

SPR-E2, 25-11 The uncertainties of It has been shown that Perform a sensitivity The number of hazard intervals E6 accelaration bin range and accelaration bin range and analysis on acceleration have been changed from 8 ACUBE parameters have ACUBE parameter selection bin ranges and intervals to 10 intervals, which not been identified and can have significant impacts demonstrate CDF and provides a better understanding evaluated. on CDF and LERF values. LERF stability. of which ground motions contribute the most to seismic risk. Several variations on the (This F&O originated from Per SR QU-E4, the above Perform a sensitivity number and size of the intervals SR SPR-E2) sources of uncertainty need analysis on ACUBE were performed to establish the to be addressed and their parameters and 10 intervals used in the final potential impact assessed. demonstrate CDF and SPRA.

LERF stability.

ACUBE was used to process the CDF and LERF cutsets using the Page 169 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Submittal March 2018 Binary Decision Diagram (BDD) to obtain a more accurate result that reduces the over-counting that can occur with the minimum cutset upper bound (MCUB) when high probabilities are present in the cutsets. All SCDF cutsets were processed through ACUBE to obtain the SCDF.

However, due to limitations in computer memory, not all SLERF cutsets were processed through ACUBE. The processing of the cutsets through ACUBE was refined to maximize the number of cutsets processed. For example, to process more SLERF cutsets, the SLERF for each hazard interval was processed through ACUBE separately, which allows processing nearly all of the cutsets for each initiator.

Additional improvements in the processing of the cutsets for importance of the SSCs as well as for the HEPs and accident Page 170 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Submittal March 2018 sequence flags as documented in the SPRA quantification results.

This finding is considered resolved. There is no effect on the results or conclusions of the SPRA.

Page 171 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Submittal March 2018 SPR-E2, 25-12 lnternalevents The internal events Review the internal The internal events PRA model E6 uncertainties are not uncertainties could have a events uncertainties uncertainties were reviewed for reviewed and evaluated significant impact on the and assumptions and applicability to the SPRA. The with respect to seismic seismic PRA if not evaluate them with results are documented in the impacts. sufficiently addressed. respect to the seismic SPRA Model Development PRA. notebook. The review concluded that the uncertainties are either (This F&O originated from Per SR QU-E4, the above not applicable in the SPRA or SR SPR-E2) sources of uncertainty need they are already included as a to be addressed and their source of uncertainty in the potential impact assessed. SPRA.

This finding is considered resolved. There is no effect on the results or conclusions of the SPRA.

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NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Submittal March 2018 SPR-E2, 25-13 The uncertainty of using of The FLEX related HFEs were For the model A clarification was added to the E6 HRA calculator surrogates reviewed: HEP-C-FLEX- uncertainty associated seismic HRA notebook that for HFEs was not evaluated. LOADSHED-5(1-4), HEP-C- with the use of HEP discusses the use of surrogates in FLEX-REFUEL-S(l-4), HEP-C- calculator surrogate, the HRA for the FLEX execution FLEX-RIPS( 1-4), and HEP-C- IDENTIFY how the PRA errors. Also, this was listed as a (This F&O originated from FLEX-VAC-5(1-4) and it was model is affected (e.g., source of uncertainty, which was SR SPR-E2) noticed that surrogate perform a sensitivity on evaluated by a sensitivity. The values are used to capture the actions addressed results show a relatively minor the contribution for the with this technique). impact on SCDF and SLERF.

unique nature of the This finding is considered actions taken in FLEX that resolved.

are outside the scope of the HRA calculator.

This approach has inherent uncertainty that should be evaluated. Per SR QU-E4, this source of uncertainty need to be addressed and their potential impact assessed.

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NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Submittal March 2018 SPR-E4 25-14 There was no evaluation of The basis for correlation is Perform appropriate The basis for correlating SSCs is correlation impact documented in section 4.3 sensitivity analyses of consistent with standard industry performed. of the SA.3 notebook. Some correlation for risk- methods _with respect to of the fragility groups that significant fragilities. correlating redundant SSCs that appear to be significant The fragility team may are located in the same area and (This F&O originated from may not be 100% correlated suggest additional have similar design and SR SPR-E4) given that they have possible correlation (for installation. Orientation of the different orientations (e.g. example based on SSCs may be considered for vital buses) or have orientation, design or uncorrelating SSCs if the SSCs are different designs (e.g. some other factors). oriented differently.

vital bus inverters are 20kva In the NAPS SPRA, redundant and others are 15kva).

SSCs that have different Correlation doesn't have to orientation were modeled as be 100% correlated. There uncorrelated only if the fragilities was no determination on of the SSCs were significantly whether or not the model is different. In the case of the vital sensitive to correlation.

bus panels, the HCLPF capacities of the panels are essentially the same regardless of orientation.

Therefore, modeling these panels as correlated is considered appropriate.

Likewise for the vital bus inverters, where one of the inverters has a higher power Page 174 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTIF 2.1 Seismic PRA Submittal March 2018 rating (resulting in slightly larger mass) than the other three. The HCLPF capacities calculated for the inverters are not different enough to considered them uncorrelated due to the weight difference. Therefore, modeling of the inverters as correlated is considered appropriate. The other SSCs were reviewed and verified to be modeled appropriately with respect to correlation.

This finding is considered resolved. There is no effect on the results or conclusions of the SPRA.

Page 175 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Submittal March 2018 SFR-Dl, 26-1 Service water travelling The travelling screens were Evaluate the travelling The design and configuration of SPR-Dl screens were screened at identified as components screens for potential SW traveling screens were the system analysis level as from the internal events seismic failure and reviewed to determine if their filters. A.s active moving PRA but were screened as a interactions that would failure could impact the SW components, they should filter. The failure modes of impact functionality of pumps. The review concluded have been*passed to the a passive component like a service water. that seismic failure of the screens fragility analysis as failure filter do not have the same would not impact the SW pumps.

mode did not match the potential seismic failure The SPRA documentation was plugging that could be modes and interaction as a updated to document the screened. travelling screen. conclusions of this review.

This finding is considered resolved. There is no effect on (This F&O originated from the results or conclusions of the SR SPR-Dl)

SPRA.

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NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Submittal March 2018 SPR-B9 26-2 A potential for a flood in The postulated collapse Confirm the Additional failure modes have excess of the plant flood failure of the turbine effectiveness of the been evaluated for the Turbine design in turbine building building could result in a berm to control the Building (TB). The seismically-(which is assumed to fail in flood with the inventories impacts of the flood OR induced structural damage within a seismic event) was of circulating water, include the flood the TB has been evaluated to identified during the plant condensate, feedwater, scenario in the SPRA. determine the potential for walkdown. condensate makeup, significant flooding that could condensate polishing, main adversely affect the function of steam, turbine lube oil and core damage mitigating (This F&O originated from any secondary side cooling equipment as described in the SR SPR-B9) water systems. The flood disposition of finding F&O 24-2.

volume retained behind the The disposition of F&O 24-2 wall would be reduced due concluded that there were no to debris filling the flooding consequences from the retention volume. The flood bounding failure modes for the sources alone would TB.

normally be in excess of Therefore, the existing flood what design basis flood barriers are adequate to protect protection in the form of the safety related AC and DC walls/berms would be power distribution systems designed to contain. The within the Emergency Switchgear propagation of this flood Room (ESGR).

beyond the flood wall would impact all safety This finding is considered related AC and DC power resolved. There is no effect on distribution resulting a high the results or conclusions of the Page 177 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTIF 2.1 Seismic PRA Submittal March 2018 Conditional Core Damage SPRA.

Probability (CCDP). The collapse of the turbine building may also preclude the use of an operator action to mitigate the flood by isolating the flood sources.

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NAPS Units 1 and 2 10 CFR 50.54(f) NTIF 2.1 Seismic PRA Submittal March 2018 SPR-Cl 26-3 Human Action to secure This action could alter the JUSTIFY how this The Operator action to isolate CCW HX flood sources has a SPRA CDF / LERF and there potential conservatism the SW flood is not credited in relatively high FV but is not is no discussion of this is imapcting the model. the SPRA due to the uncertainty credited without any potential conservatism. This can be done by in the size of the flood. This has evaluation of the potential performing a sensitivity been listed as a source of impact on the model. analysis that show the uncertainty. A sensitivity was effect of crediting this performed to evaluate the action. If this action is impact of crediting this action if (This F&O originated from included in the the flood size is lower and time is SR SPR-Cl) evaluation, an available to isolate it. The results appropriate feasibility show only a very little decrease assessment should be in SCDF and SLERF if this action is included. credited for smaller breaks in the SW piping.

This finding is considered resolved. There is no effect on the results or conclusions of the SPRA.

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NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Submittal March 2018 SPR-Al 26-5 Process for determining Recent industry experience Evaluate the impact of The seismic-induced fire earthquake caused and focus has heighted not screening the high evaluation has been revised to initiating events is outlined awareness and concern in energy electrical include the evaluation of high in SA.3 section 3.1 using the arena of high-energy equipment from futher energy electrical SSCs. The SPRAIG guidance - but all cabinet fires. The following consideration in the evaluation concluded that seismic fire interaction was examples show some areas/scenarios that seismic risk due to high energy screened. Industry potential significance for were significant in the electrical SSCs is low and that no experience has this issue: fire analysis. changes to the SPRA model were demonstrated that several required to model seismic failure of the SPRAIG guidance of high energy electrical SSCs.

component type listed as

• At Onagawa {2011) fire This finding is considered "neglible" should still be occurred in a non-resolved. There is no effect on considered. seismically qualified power the results or conclusions of the supply, but no count of SPRA.

total number of functional (This F&O originated from failures is provided.

SR SPR-Al)

• At Kashiwazaki-Kariwa (2007) there were fires in non-seismically qualified equipment (it did not say

• how many)

• At Kashiwazaki-Kariwa (2007) "Only minor damage to non-Class A or As SSCs was found, for example, a Page 180 of 181

NAPS Units 1 and 2 10 CFR 50.54(f) NTTF 2.1 Seismic PRA Submittal March 2018 house transformer fire of Unit 3

• There are 2 transformer fires and one low voltage switchgear fire in the SQUG database

• There was a medium voltage switchgear fire at a Kansai power sub-station (1995)

• Recent (post Fukushima) shake table testing in Japan has shown HEAF in switchgear can occur

• A recent study by FENOC and ABS concluded that HEAF due to seismic failure could not be excluded a priori (Screening of Seismic-Induced Fires by Lin, Wakefield and Reddington, PSAM 12)

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