L-17-234, High Frequency Supplement to Seismic Hazard Screening Report, Response to NRC Request for Information Pursuant to 10 CFR 50.54(0 Regarding Recommendation 2.1 of the Near-Term Task Force (NTTF) Review of Insights from the .

High Frequency Supplement to Seismic Hazard Screening Report, Response to NRC Request for Information Pursuant to 10 CFR 50.54(0 Regarding Recommendation 2.1 of the Near-Term Task Force (NTTF) Review of Insights from the .

ML17223A362
Person / Time
Site:	Perry
Issue date:	08/11/2017
From:	Hamilton D FirstEnergy Nuclear Operating Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
CAC MF3729, L-17-234
Download: ML17223A362 (81)

Text

{{#Wiki_filter:FENOC' ~ FirstEnergy Nuclear Operating Company David B. Hamilton Vice President August 11, 2017 L-17-234 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission 11555 Rockville Pike Rockville, MD 20852

SUBJECT:

Perry Nuclear Power Plant Docket No. 50-440, License No. NPF-58 Perry Nuclear Power Plant PO. Box 97 10 Center Road Perry, Ohio 44081 440-280-5382 10 CFR 50.54(f) High Frequency Supplement to Seismic Hazard Screening Report, Response to NRC Request for Information Pursuant to 10 CFR 50.54(0 Regarding Recommendation 2.1 of the Near-Term Task Force (NTTF) Review of Insights from the Fukushima Dai-ichi Accident {CAC Nos. MF3729) On March 12, 2012, the Nuclear Regulatory Commission (NRC) issued a Request for Information pursuant to 10 CFR 50.54(f) (Reference 1) to all power reactor licensees. The required response section of Enclosure 1 of Reference 1 indicated that licensees should provide a seismic hazard evaluation and screening report within 1.5 years from the date of the letter for central and eastern United States (CEUS) nuclear power plants. By letter dated May 7, 2013 (Reference 2), the NRC extended the date to submit the report to March 31, 2014. By letter dated May 9, 2014 (Reference 3), the NRC transmitted the results of the screening and prioritization review of the seismic hazards reevaluation report for Perry Nuclear Power Plant (PNPP) submitted by letter dated March 31, 2014 (Reference 4). In accordance with the screening, prioritization, and implementation details report (SPID) (References 5, 6, and 7), and Augmented Approach guidance (Reference 2), the reevaluated seismic hazard is used to determine if additional seismic risk evaluations are warranted for a plant. Specifically, the reevaluated horizontal ground motion response spectrum (GMRS) at the control point elevation is compared to the existing safe shutdown earthquake (SSE) or Individual Plant Examination for External Events (IPEEE) High Confidence of Low Probability of Failure (HCLPF) Spectrum (HIS) to determine if a plant is required to perform a high frequency confirmation evaluation. As noted in Enclosure 2 of Reference 3, PNPP is to conduct a limited scope high frequency evaluation ( confirmation).

Perry Nuclear Power Plant L-17-234 Page 2 Within Reference 3, the NRC acknowledged that these limited scope evaluations will require additional development of the assessment process. The Nuclear Energy Institute (NEI) submitted an Electric Power Research Institute (EPRI) report titled, High Frequency Program: Application Guidance for Functional Confirmation and Fragility Evaluation (EPRI 3002004396) for NRC review and endorsement (References 8 and 9). NRC endorsement was provided by Reference 10. Reference 11 provided the NRC final seismic hazard evaluation screening determination results and the associated schedules for submittal of the remaining seismic hazard evaluation activities. The enclosure to this letter provides the High Frequency Evaluation Confirmation Report for PNPP that confirms that all high frequency susceptible equipment evaluated with the scoping requirements and criteria for seismic demand have adequate seismic capacity. Therefore, no additional modifications or evaluations are necessary. The enclosure provides the requested information in response to Reference 1 associated with NTTF Recommendation 2.1 Seismic evaluation criteria. There are no new regulatory commitments contained in this letter. If there are any questions or if additional information is required, please contact Mr. Thomas A. Lentz, Manager - Fleet Licensing, at 330-315-6810. I declare under penalty of perjury that the foregoing is true and correct. Executed on August _iL, 2017. Respectfully, __ ( David B. Hamilton Enclosure Near-Term Task Force (NTTF) 2.1 High-Frequency Confirmation Submittal Perry Nuclear Power Plant

References:

1.

NRC 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, Agencywide Documents Access and Management System (ADAMS) Accession Number ML12053A340.

2.

NRC Letter, Electric Power Research Institute Report Final Draft Report XXXXXX, Seismic Evaluation Guidance: Augmented Approach for the Resolution of Fukushima Near-Term Task Force Recommendation 2. 1: Seismic, As An

Perry Nuclear Power Plant L-17-234 Page 3 Acceptable Alternative to the March 12, 2012, Information Request for Seismic Reevaluations, dated May 7, 2013, ADAMS Accession Number ML13106A331.

3.

NRC Letter, Screening and Prioritization Results Regarding Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(f) Regarding Seismic Hazard Re-evaluations for Recommendation 2.1 of the Near Term Task Force Review of Insights from the Fukushima Dai-ichi Accident, dated May 9, 2014, ADAMS Accession Number ML14111A147.

4.

FENOC Letter, FirstEnergy Nuclear Operating Company (FENOC) Seismic Hazard and Screening Report (CEUS Sites), Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force (NTTF) Review of Insights from the Fukushima Dai-ichi Accident, dated March 31, 2014, ADAMS Accession Number ML14092A203.

5.

NEI Letter, Final Draft of Industry Seismic Evaluation Guidance (EPRI 1025287), dated November 27, 2012, ADAMS Accession Numbers ML12333A168.

6.

EPRI Report 1025287, Seismic Evaluation Guidance, Screening, Prioritization and Implementation Details [SP/DJ for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1: Seismic, November 2012, ADAMS Accession Number ML12333A170.

7.

NRC Letter, Endorsement of Electric Power Research Institute Final Draft Report 1025287, Seismic Evaluation Guidance, dated February 15, 2013, ADAMS Accession Number ML12319A074.

8.

NEI Letter, Request for NRC Endorsement of High Frequency Program: Application Guidance for Functional Confirmation and Fragility Evaluation (EPRI 3002004396), dated July 30, 2015, ADAMS Accession Numbers ML15223A100.

9.

EPRI Report 3002004396, High Frequency Program: Application Guidance for Functional Confirmation and Fragility Evaluation, July 2015, ADAMS Accession Number ML15223A102.

10. NRC Letter, Endorsement of Electric Power Research Institute Final Draft Report 3002004396, High Frequency Program: Application Guidance for Functional Confirmation and Fragility, dated September 17, 2015, ADAMS Accession Number ML15218A569.

11. NRC 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 Number ML15194A015.

cc: Director, Office of Nuclear Reactor Regulation (NRR) NRC Region Ill Administrator NRC Resident Inspector NRR Project Manager

Enclosure L-17-234 Near-Term Task Force (NTTF) 2.1 High-Frequency Confirmation Submittal Perry Nuclear Power Plant (77 pages follow)

FIRST EF{ERGY I\\UCLEAR OPERATII\\G COMPAI\\Y I\\ear-Term Task Force (l{TTF) 2.1 High-Frequency Confirmation Submittal Perry Nuclear Power Plant

Rfport Nrne Ilrto: Hsvirion }lo.: APPROVALS Ncar-Term Task Forps (NTTF) Z.l High Frcquency Comfirrrdon Srrylemeutal Repoil Pary Nuclar Pourp'r Plfirt June 30,2017 Rr;visim 0 bl*=fua1 Nuclear Euginee#i IldE I R Hayes, Plant Sf'rt Eng* Dds 6 P Jr,Irluclear DdE tltfpn byr ft n- --r1 Levr,rencc A, I Specialirt* Nuplear EngiEEr* R NuslcuEryinec# Huclear Engincr* Specialirtr Engrcff* DdE 7 I slrr ffis nI=l*or.p 7 Dtte -t Dabrr = *llstly_ DEtc u A t ,+rJ*If $ystem Eilg* Paril Suporrrisor $upplySyste,m*ng* Dato I\\rfcthods* F. AIvi, ft,ojest [4auagert ffi

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AESGonsulting il3nl##,9 Near-Term Task Force (NTTF) 2.1 H i g h-F req uency Confi rmation Submittal Perry Nuclear Power Plant June 28,2017 2734298-R-015 Revision 0 Prepared for: FirstEnergy Nuclear Operating Company ABSG Consulting lnc.. 300 Commerce Drive, Suite 200. lruine, California 92602

2734298-R-015 Raision 0 lune 28,201.7 Page 2 of 49 NEAR-TERM TASKFORCE (NTTF) 2.1 HIGH.FREQUENCY CONFIRMATION SUBMITTAL PERRY NUCLEAR POWER PLANT ABSG Cor{sul,rrNc INC. Rnronr No. 2734298-R-015 RrvrsroN 0 RLZ.Z,O Rrponr N0. R12 12-4734 JuNn 28,2017 ABSG CONSULTING INC. RJIZ,,4O ASSOCIATES lSGonsultirtg (iFrzzo

2734298-R-01.s Rwision 0 lune 28,201.7 Page 3 of 49 Report Name: Date: Revision No.: Originator: Independent Verifier: Principal: Project Manager; ""fr"r.r@W June 28. 2017 Date June 28.2017 Date June 28, 2017 Date June 28,2017 Date June 28,2017 APPROVALS Near-Term Task Force (NTTF) 2.1 High-Frequency Confirmation Submiual Perry Nuclear Power Plant June 28,2017 0 Bradley Yagla Engineering Associate RIZZO Associates Eddie Guerra, P.E. Director of Structural Engineering F*IZZO Associates l-li*u"qo Nishikant R. Vaidya, Ph.D., P.E. Vice President P*IZZO Associates 7na Farzin R. Beigi, P.E. Consultant RIZZO Associates R. Roche, P.E. Vice President ABSG Consulting Inc. lEtGonsuttlttg ()Rtzzo Approver: Date

2734298-R-015 Raision 0 lune 28,201.7 Pase 4 of 49 CHANGE MANAGEMENT RECORI) Rrvrsrou N0. D,+.rr DrcscmrloNs oF CHANGES/AFFECTED P.t cBs 0 June 28,2017 Initial Submittal lEEGonsulting []Rrzzo

2734298-R-015 Ratision 0 lune 28,201.7 Pase 5 of 49 TABLE OF CONTENTS PAGE LIST OF TABLES 7 I 9 LIST OF FIGURES EXECUTIVE

SUMMARY

LIST OF ACRONYMS t1 t4 t4 t4 l5 l6 19 t9 19 25 25 3l 36 43 44 45 .37 ,37 .38 I.O INTRODUCTION 1.3 AppnoacH 1.4 Pr-,q.hrr ScnEeNn-rG........... 2.0 SELECTTON OF COMPONENTS FOR HrGH-FREQUENCY SCREENING 2.1 RBacron/Tnrp/SCRAM.. 2.2 Rsacron VssspI-INVENToRy Colrrnol 2.3 Rsecron Vsssnr-PnsssuRE Cournol I.1 Punposp 1.2 BncncRouND.... 2.4 Cone Coor-rhrc 2.5 AC/DC Powpn Supponr Svsreus 2.6 Suuuany oF Spr-ecrpn CoNapohrENTS 3.0 SEISMIC EVALUATION 3.1 HozuzoNrnr. Sersuuc DSUAND 3.2 Venucal Sersnnc Deunuo 3.3 CovrpoueNr HoruzoNTAL Sersrrc Dprrrenm 3.4 Conapoupxr VERUcAL Sersurc DEnaeun 4.0 CONTACT DEVICE EVALUATIONS 5.1 GeNpneL Colrcr,usroNs ..............46 5.2 InEhrrrrrcATroN or Follow-Up Acuous .46 .47 lESGonrulting (]Rtzzo

6.0 REFERENCES

2734298-R-01,5 Reuision 0 June 28,201-7 Page 6 of 49 APPENDICES: APPENDIX A APPENDIX B TABLE OF CONTENTS (coNTTNUED) REPRESENTATIVE SAMPLE COMPONENT EVALUATIONS COMPONENTS IDENTIFIED FOR HIGH.FREQUENCY CONFIRMATION AESConsulting tiRtzuo

2734298-R-015 Ranision 0 lune 28, 2017 Pase 7 of 49 LIST OF TABLES TABLE NO. TABLE 1.1 TABLE 1.2 TABLE 3.1 TABLE 3.2 TABLE 3.3 TITLE GMRS AT THE PNPP, EL 561 FT SSE AT THE PNPP PAGE SOIL MEAN SHEAR-WAVE VELOCITY AND DEPTH PROFILE FOR THE FIRST IOO FT; REACTOR BUTLDTNG FOUNDATTON (EL 561 FT)....... ...............38 HORIZONTAL AND VERTICAL GMRS FOR RB FOUNDATTON (EL s6r FT)....... .......40 HORIZONTAL AND VERTICAL FOTINDATION INPUT RESPONSE SPECTRA (FIRS) FOR DGB FOLTNDATTON (EL 6rs FT) t7 t8 4t [E$Gonsultlng tiRtzao

2734298-R-01.5 Rwision 0 lune 28,201,7 Page I of 49 FIGURE NO. FIGURE I.1 FIGURE 3-1 FIGURE 3.2 LIST OF FIGURES TITLE PAGE COMPARISON OF GMRS AND SSE AT THE PNPP CONTROL POINT (EL s61 FT). ........18 PLOT OF THE HORIZONTAL AND VERTICAL GROUND MOTION RESPONSE SPECTRA A^hID V/H RATTOS FOR EL 561 FT (RB FOUNDATTON)............42 PLOT OF THE HORIZONTAL AND VERTICAL GROUND MOTION RESPONSE SPECTRA AND V/H RATTOS FOR EL 61s FT (DGB FOUNDATION) 42 lE$Gonsulting tlFtzzo

2734298-R-015 Rwision 0 lune 28,2017 Pase I of 49 ABS CONSULTING AC ADS AUX BWR CC CEUS CST DC DGB ECCS EDG EL EPRI ESEP ESW FENOC FHB FIRS ft fl/s ob GMRS HFIRS HGMRS HPCS HX Hz LIST OF ACRONYMS ABSG CONSULTTNG INC. ALTERNATTNG CURRENT AUTOMATIC DEPRESSURIZATION SYSTEM AUXILIARY BUILDING BOILING WATER REACTOR CONTROL COMPLEX CENTRAL AND EASTERN UNITE,D STATES CONDENSATE STORAGE TANK DIRECT CURRENT DIESEL GENERATOR BUILDING EMERGENCY CORE COOLING SYSTEM EMERGENCY DIESEL GENERATORS ELEVATION ELECTRIC POWER RESEARCH INSTITUTE EXPEDITED SEISMIC EVALUATION PROCESS EMERGENCY SERVICE WATER FIRSTENERGY NUCLEAR OPERATING COMPANY FUEL HANDLING BUILDING FOLINDATION INPUT RESPONSE SPECTRA FEET FEET PER SECOND ACCELERATION OF GRAVITY GROI.]ND MOTION RESPONSE SPECTRA HORIZONTAL FIRS HORIZONTAL GMRS HIGH PRESSURE CORE SPRAY HEAT EXCHANGER HERTZ lESGonsulting []Ftzzo

2734298-R-015 Reaision 0 lune 28, 2017 Pase 10 of 49 ISRS LOCA LOOP LPCS m m/s MCC MOV MSIV NRC NSSS NTTF PGA PNPP RB RCIC RCS RHR RTZZO RPS RPV RWCU SA SILO SPRA SPID SRV SSC LIST OT ACRONYMS (coNTTNUED) IN.STRUCTURE RESPON SE SPECTRA LOSS OF COOLANT ACCIDENT LOSS OF OFFSITE POWER LOW PRESSURE CORE SPRAY METER METER PER SECOND MOTOR CONTROL CENTER MOTOR.OPERATED VALVE MATN STEAM ISOLATION VALVES UNITED STATES NUCLEAR REGULATORY COMMISSION NUCLEAR STEAM SUPPLY SHUTOFF NEAR.TERM TASK FORCE PEAK GROUND ACCELERATION PERRY NUCLEAR POWER PLANT REACTOR BUILDING REACTOR CORE ISOLATION COOLING REACTOR COOLANT SYSTEM RESIDUAL HEAT REMOVAL F-IZZO ASSOCIATES REACTOR PROTECTION SYSTEM REACTOR PRESSURE VESSEL REACTOR WATER CLEAN-UP SPECTRAL ACCELERATION SEAL.IN OR LOCK-OUT SEISMIC PROBABILISTIC RISK AS SESSMENT S CREENING, PRIORITIZATION, AND IMPLEMENTATI ON DETAILS SAFETY RELIEF VALVE STRUCTURES, SYSTEMS, AND COMPONENTS nESGonsulting []Rtzzo

2734298-R-01,5 Rmtision 0 lune 28,20L7 Page 11 of49 SSE UFSAR WUS V/H Vs VAC VFIRS VGMRS LIST OF ACRONYMS (coNTTNUED) SAFE SHUTDOWN EARTHQUAKE UPDATED SAFETY ANALYSIS REPORT WESTERN UNITED STATES VERTI CAL - TO. HOzuZ ONTAL SHEAR.WAVE VELOCITY VOLTS AC VERTICAL FIRS VERTICAL GMRS ABSCo:rsulting tlR.zzo

2734298-R-015 Ratision 0 June 28,2017 Page 1.2 of 49 NEAR-TERM TASK FORCE (NTTF) 2.1 HIGH-FREQUENCY CONFIRMATION SUBMITTAL PERRY NUCLEAR POWER PLANT EXECUTIVB

SUMMARY

The purpose of this report is to provide information as requested by the Nuclear Regulatory Commission (NRC) in its March 12,2012, letter issued to all power reactor licensees and holders of construction permits in active or deferred status (Reference 1). In particular, this report provides information requested to address the High-Frequency Confirmation requirements of Item (4), Enclosure l, Recommendation 2.1: Seismic, of the March 12,2012, letter (Reference l). Following the accident at the Fukushima Dai-ichi nuclear power plant resulting from the March 11, 2011, Great Tohoku Earthquake and subsequent tsunami, the 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 clariff and strengthen the regulatory framework for protection against natural phenomena. Subsequently, the NRC issued a 50.54(f) letter on March 12,2012 (Reference l), requesting information to assure that these recofirmendations 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. Included in the 50.54(0 letter was a request that licensees perform a "confirmation, if necessary, that SSCs, whichmay be affected by high-frequency ground motion, will maintain their functions important to safety". EPRI 1025287, "Seismic Evaluation Guidance: Screening, Prioritization and Implementation Details (SPID) for the resolution of Fukushima Near-Term Task Force Recommendation 2.1: Seismic" (Reference 2) provided screening, prioritization, and implementation details to the U.S. nuclear utility industry for responding to the NRC 50.54(0 letter. This report was developed with NRC participation and was subsequently endorsed by lESGonsulting ()Rtzzo

2734298-R-015 Rsuision 0 lune 28, 201.7 Page L3 of49 the NRC. The SPID included guidance for determining which plants should perform a High-Frequency Confirmation and identified the types of components that should be evaluated in the evaluation. Subsequent guidance for performing a High-Frequency Confirmation was provided in EPRI 3002004396, "High Frequency Program, Application Guidance for Functional Confirmation and Fragility Evaluation," (Reference 3) and was endorsed by the NRC in a letter dated September 17,2015 (Reference 4). Final screening identiffing plants needingto perform a High-Frequency Confirmation was provided by NRC in a letter dated October 27,2075 (Reference 5). This report describes the High-Frequency Confirmation evaluation undertaken for Perry Nuclear Power Plant (PNPP). The objective of this report is to provide srunmary information describing the High-Frequency Confirmation evaluations and results. The level of detail provided in the report is intended to enable NRC to understand the inputs used, the evaluations performed, and the decisions made as a result of the evaluations. EPRI 3002004396 (Reference 3) is used for the PNPP engineering evaluations described in this report. In accordance with Reference 3, the following topics are addressed in the subsequent sections of this report: r Process of Selecting Components and a List of Specific Components for Hi gh-Frequency Confi rmation . Estimation of a Vertical Ground Motion Response Spectrum (GMRS) r Estimation of In-Cabinet Seismic Demand for Subject Components . Estimation of In-Cabinet Seismic Capacity for Subject Components o Summary of Subject Components' High-Frequency Evaluations fiESGqrsulting

2734298-R-015 Raision 0 lune 28,201.7 Page 14 of49

1.0 INTRODUCTION

1.1 Punrosr The purpose of this report is to provide information as requested by the NRC in its March 12, 2012,50.54(0 letter issued to all power reactor licensees and holders of construction permits in active or deferred status (Reference 1). In particular, this report provides requested information to address the High-Frequency Confirmation requirements of Item (4), Enclosure 1, Recommendation 2.1: Seismic, of the March 1,2,2012, letter (Reference l). 1.2 BncrcnouND Following the accident at the Fukushima Dai-ichi nuclear power plant resulting from the March 11,2071, Great Tohoku Earthquake and subsequent tsunami, the NRC established a NTTF to conduct a systematic review ofNRC processes and regulations and to determine if the agency should make additional improvements to its regulatory system. The NTTF developed a set of recofirmendations intended to clariff and strengthen the regulatory framework for protection against natural phenomena. Subsequently, the NRC issued a 50.54(f) letter on March 12,2012 (Reference l), 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. Included in the 50.54(f) letter was a request that licensees perform a ooconfirmation, if necessary, that SSCs, which may be affected by high-frequency ground motion, will maintain their functions important to safety." EPRI 1025287, "Seismic Evaluation Guidance: Screening, Prioritization and Implementation Details (SPID) for the resolution of Fukushima Near-Term Task Force Recommendation 2.1: Seismic" (Reference 2) provided screening, prioritization, and implementation details to the U.S. nuclear utility industry for responding to the NRC 50.54(0 letter. This report was developed with NRC participation and is endorsed by the NRC. The SPID included guidance for determining which plants should perform a High-Frequency Confirmation and identified the types of components that should be evaluated in the evaluation. lESGonsulting []Rtzzo

2734298-R-015 Ratision 0 lune 28,2017 Pase L5 of49 Subsequent guidance for performing a High-Frequency Confirmation was provided in EPRI 3002004396, "High Frequency Program, Application Guidance for Functional Confirmation and Fragility Evaluation," (Reference 3) and was endorsed by the NRC in a letter dated September 17,2015 (Reference 4). Final screening identiffingplants needing to perform a High-Frequency Confrmation was provided by NRC in a letter dated October 27,2015 (Reference 5). On March 31,20t4, PNPP submitted areevaluated seismic hazardto theNRC as apart of the Seismic Hazard and Screening Report (Reference 6). By letter dated August 3,2015, the NRC staff concluded that the GMRS that was submified adequately characterizes the reevaluated seismic hazard for the PNPP site (Reference 8). The seismic hazard was later reevaluated under the Expedited Seismic Evaluation Process (ESEP) and submitted to the NRC on December 19, 2014 (Reference 7). The ESEP was accepted by the NRC by letter dated Septemb er 23,2015 (Reference l9). By letter dated October 27,2015 (Reference 5), the NRC transmitted the results of the screening and prioritization review of the seismic hazards reevaluation. This report describes the High-Frequency Confirmation evaluation undertaken for PNPP using the methodologies in EPRI 3002004396, "High Frequency Prograrn, Application Guidance for Functional Confirmation and Fragility Evaluation," as endorsed by the NRC in a letter dated September 17,2015 (Reference 4). The objective of this report is to provide summary information describing the High-Frequency Confirmation evaluations and results. The level of detail provided in the report is intended to enable NRC to understand the inputs used, the evaluations performed, and the decisions made as a result of the evaluations. 1.3 AprRoncn EPRI 3002004396 (Reference 3) is used for the PNPP engineering evaluations described in this report. Section 4.1 of Reference 3 provided general steps to follow for the High-Frequency Confirmation component evaluation. Accordingly, the following topics are addressed in the subsequent sections of this report: AESGonsulting []Rtz",o^

2734298-R-01.5 Raision A lune 28, 2017 Pase 16 of49 PNPP's SSE and GMRS Information Selection of Components and a List of Specific Components for High-Frequency Confirmation . Estimation of Seismic Demand for Subject Components . Estimation of Seismic Capacity for Subject Components Summary of Subject Components' High-Frequency Evaluations Summary of Results t.4 PLaF{r Scnnnxruc PNPP submitted the seismic hazafi and screening report in response to the NRC request for information Pursuant to 10 CFR 50.54(0 on March 31,2014 (Reference 6). By letter dated August 3,2015, the NRC staff concluded that the GMRS that was submitted adequately characterizes the reevaluated seismic hazard for the PNPP site (Reference 8). The NRC final screening determination letter concluded (Reference 5) that the GMRS to SSE comparison at the PNPP resulted in a need to perform a High-Frequency Confirmation in accordance with the screening criteria in the SPID (Reference 2). Subsequent to the March 31,2014 submiual, the seismic hazard was updated considering site specific damping in rock. The updated seismic hazard is the basis for the ESEP Reports submitted by FirstEnergy Nuclear Operating Company (FENOC) on December 19,2014 (ReferenceT), and also used inthe SPRA. The ESEP was acceptedby theNRC by letter dated September 23,2015 (Reference 19). Table 1-1, Table I-2, and Figure 1-I present the spectral accelerations characterizing the updated GMRSs and SSE at the PNPP. Figure 1-1 presents the comparison of SSE, ESEP GMRS (Reference 7) and the GMRS reported in the PNPP March 2014 submittal (Reference 6). The difference in the GMRS results is attributed to the material damping used for the rock material over the upper 500 feet (ft). While the GMRS reported in the March 2014 submittal is based on the low strain damping of approximately 3.2 percent over a depth of 500 ft below the Reactor Building (RB) foundation, the GMRS used in the ESEP limits this damping value to the upper 100 ft where the rock is considered as weathered or fractured. Below this depth, a low lESGonsulting tlFtzzo

2734298-R-015 Raision 0 lune 28,201-7 Pase L7 of49 strain damping of 1.0 percent is used based on the unweathered shale dynamic properties from Stokoe et al. (Reference 9). TABLE 1.1 GMRS AT THE PNPP, EL 561 FT FnreunNCY (Hz) GMRS (g) (ESEP, DncBnnnER 2014 SugnilrrrA.L) GMRS (g) (MARCH 2014 Sunnnmrnl) 0.r0 0.0030 0.003 0.13 0.0045 0.0044 0.16 0.0065 0.0065 0.24 0.009s 0.0095 0.26 0.0139 0.0139 0.33 0.0208 0.0209 0.42 0.0322 0.0323 0.50 0.0458 0.0458 0.s3 0.0489 0.0488 0.67 0.0626 0.062 0.85 0.0784 0.0778 1.00 0.089s 0.0886 l.08 0.0991 0.0978 1.37 0.1228 0.1206 1.74 0.1277 0.12,62 2.21 0.1489 0.1453 2.s0 0.r769 0.1 6s6 2.81 0.2r03 0.1944 3.s6 0.272r 0.2484 4.52 0.3287 0.3011 s.00 0.3618 0.3247 5.74 0.4020 0.3s54 7.2& 0.4664 0.4036 9.24 0.5424 0.4514 10.00 0.5663 0.4681 tt.tz 0.5773 0.4726 14.87 0.5851 0.4648 18.87 0.s976 0.4s93 23.95 0.s788 0.4282 25.00 0.5722 0.4207 30.39 0.s389 0.3854 38.57 0.4868 0.3476 48.94 0.4233 0.3183 62.1 0 0.3443 0.2661 78.80 0.2672 0.2048 100.00 0.2426 0.1883 frEBComsultlng ()Rtzzo

2734298-R-015 Rsuision 0 lune 28, 201.7 Page 18 of 49 FnneuENCY (llzl SSE lsl 0.10 0.013 4.25 0.07 2.50 0.47 9.00 0.39 33.00

0. 15 100.00 0.15 TABLEI.2 SSB AT THE PNPP 1.m 0.90 0.80 0.70 0.60 0.50 0.40 0.30 o.20 0.10 0.00 o.10 1.m Freguency {Hz}

10.00 1m.o0 FIGURE 1.1 coMpARISON OF'GMRS AI\\D SSE AT THE PNPP CONTROL POINT (EL 561 FT) fBtConsulffng ()Ffza:Q frID b, Eo t-+. Eg sfl( Eft lt -FGMRS [ESEP, Dec. z0l4Submitall ffi GMRS [March 2OI4 Submitall -ss PGA \\ \\ \\

2734298-R-41.5 Ratision A lune 28,2017 19 49 2.0 SELECTTON OF COMPONENTS FOR HIGH-FREQUENCY SCREENING The fundamental objective of the High-Frequency Confirmation review is to determine whether the occurrence of a seismic event could cause credited equipment to fail to perform as necessary. An optimized evaluation process is applied that focuses on achieving a safe and stable plant state following a seismic event. As described in Reference 3, this state is achieved by confirming that key plant safety functions critical to immediate plant safety are preserved (reactor ttip, reactor vessel inventory and pressure control, and core cooling) and that the plant operators have the necessary power available to achieve and maintain this state immediately following the seismic event (AD/DC power support systems). Within the applicable functions, the components that would need a High-Frequency Confirmation are contact control devices subject to intermittent states in seal-in or lockout circuits. Accordingly, the objective of the review as stated in Section4.2.l of Reference 3 is to determine if seismic induced high-frequency relay chatter would prevent the completion of the following key functions. 2.1 Rrncron/Tnrr/SCRAM The reactor trip/SCRAM function is identified as a key function in Reference 3 to be considered in the High-Frequency Confirmation. The same report also states that "the design requirements preclude the application of seal-in or lockout circuits that prevent reactor trip/SCRAM functions" and that "No high-frequency review of the reactor trip/SCRAM systems is necessary". 2.2 Rn.q.croRVESsEr.IrwnxroRyCournor, The reactor coolant system/reactor vessel inventory control systems were reviewed for contact control devices in seal-in and lock-out (S[O) circuits that would create a Loss of Coolant Accident (LOCA). The focus of the review was contact control devices that could lead to a significant leak path. Check valves in series with active valves would prevent significant leaks due to misoperation of the active valve; therefore, SILO circuit reviews were not required for those active valves. AESGonsulting tlRtzzo^

2734298-R-015 Ranision 0 lune 28, 201.7 Page 20 of49 Reactor coolant systefl/reactor vessel inventory control system reviews were performed for valves associated with the following functions: . Nuclear Steam Supply Shutoff . Reactor Water Clean-Up . Reactor Core Isolation Cooling . Residual Heat Removal . High Pressure Core Spray . Low Pressure Core Spray Nuclear Steam Suonlv Shutoff /I/^S^S,S) Valves Reactor Head Vent Valves The two reactor head vent valves (1821F0001 and lB2lF0002) are normally closed and in series with one another. Electrical control for these motor-operated valves is via a rugged hand switch. The motor contactors for these valves do not contain a seal-in and there are no other chatter sensitive contact devices involved in the control logic of these valves. Automatic Depre ssurization System (ADS) Valves The ADS valves include 1821F0041A, lB2lF0041B, 1821F0041E, 1821F0041F, 1821F0047D, 1821F0047H, 1821F005IC, and lB2lF005lG. These Safety Relief Valves (SRV) are operated via the solenoid valves SOVs 1821F0410A, 1821F04108, 82 1 F04 1 I A, tB21F04 1 1 B, I 82 I F04 I 4A, I 82 1 F04 1 48, 1 82 1F04 I 54, 1 B2 1 F041 58, I B2 1 F04224. 1 82 1 F04228, I B2 I F04254, 1 B2 1 F04258, 1 82 I F0442A, 1 B2 1 F 04428, 1821F0444A, ffid 1821F04448. Electrical control for the solenoid-operated pilot valves is via relays, which are controlled by the Reactor Pressure Vessel (RPV) Low Level Logic and the low pressure Emergency Core Cooling System (ECCS) pump pressure relays. This relay logic contains seal-ins and it is possible for the ADS valves to open following a seismic event. These relays are listed in Table B-1, below. AEEGursulting tlRtzzo

2734298-R-015 Rffiisian 0 June 28,201,7 Page 21 of 49 Safety Relief Valves In addition to the eight ADS SRVs listed above, PNPP has an additional I I SRVs: I 82 1 F00 4lC, 1 82 I F0041 D, 1 B2 I F004 I G, I 82 I F004 I K, I B2 I F00478, 1B2l F0047C, lB2lF0047F, 1821F0047G, lB2lF0051A, 1B2lF005lB, and 1821F005ID, operated via the solenoid valves tB21F04l2A,lB2lF0412B, 1821F0413A, 1821F04138, 1B2lF04l6,{, 1 B2 1 F04 I 68, I B2 I F04 1 7 A, tBz I F04 I 78, I 82 I F0420 A, 1 B2 I F04208, 1 B2 I F042t A, 1 B2 1 F042lB, 1 B2 1 F0423A, tBzl F04238, 1 82 I F0424A, I B2 I F04248, 7B2l F0440A, 1821F04408, 1B2lF044lA, 1821F0441B, 1821F0443A, ffid 1821F04438. The control logic which governs the Safety mode ofthe 19 SRVs contains seal-ins and it is possible for SRVs to open due to a seismic event. These relays are listed in Table B-7, below. Main Steam Isolation Valves (MSIV) The MSIVs include 1821F0022A, B, C, D, and 1821F002SA, Bo C, D. The MSIVs are controlled via solenoid valves. The solenoid-operated pilot valves are electrically controlled via relays, which are slaves to isolation logic relays. The later relays are energized for at-power operation and de-energizedto close the valves. In the energized state the isolation logic relays are sealed in and any chatter in the control logic would break the seal-in and close the valves. This action is a desired response to the seismic event and for this reason chatter is acceptable and no contact devices in this circuit meet the selection criteria. Main Steam Stop Valves The Main Steam Stop Valves (lNl1F0020A, B, C, D) are not required to be shut automatically upon isolation of the system, but provide a means of back-up isolation if necessary. The control logic for these normally open motor-operated valves contains no seal-in logic beyond the limit switch contactors. While it is possible for chatter of the contactors to close the Main Steam Shutoff Valves, this is the desired response to the seismic event and for this reason chatter is acceptable and no contact devices in this circuit meet the selection criteria. lESGonsultlng [iRtzzo

2734298-R-01.5 Reztsion 0 lune 28, 2017 Page 22 of 49 Main Steam Line Drain Valves The control logic for the normally-open Motor-Operated Valves 1821F0016 and 1821F0019 contains motor contactors which could chatter and seal-in, causing the valves to close. However, the closed position is the desired response to the seismic event and for this reason chatter is acceptable and no contact devices in this circuit meet the selection criteria. Reactor Water Clean-Up fiWCU Valves Reactor Water Clean-{Jp Flow Control Valve and Bottom Head Drain Flow Control Valves The RWCU Flow Control Valve 1G33F0102 is a nonnally-open motor-operated valve controlled by a hand switch, The relays, including the motor contactors for this valve, do not contain a seal-in and there are no other chatter sensitive contact devices involved in the control logic for this valve. The Bottom Head Drain Bypass Valve 1G33F0103 is a normally-open manual valve and is not susceptible to chatter. The Bottom Head Drain Valve 1G33F0101 is a nonnally closed motor-operated valve (MOV). This valve contains a motor contactor with a seal-in through which chatter could result in the valve opening. However, these valves are upstream of the RWCU Containment Isolation Valves 1G33F0001 and 1G33F0004 (see below) and are not relied upon for isolation of the system. No contact devices in this circuit meet the selection criteria. Reactor Water Clean-Up Isolation Valves The RWCU Containment Isolation Valves 1G33F0001 and 1G33F0004 are normally-open MOVs which close upon an isolation signal. Open limit switches in the opening circuit prevent seal-in of the opening contactor auxiliary contact and no contacts prevent valve closure via the control switch or isolation relay. These relays are energized for at-power operation and de-energized to close the valves. In the energized state the relays are sealed in and any chatter in the control logic would break the seal-in and close the valves. This action is a desired response to the seismic event and for this reason chatter is acceptable and no contact devices in this circuit meet the selection criteria. lEtGonsulting []Rtzzo

2734298-R-015 Reuision 0 lune 28,201,7 Page 23 of 49 Reactor Core Isolation Coolins &CIO Valves Reactor Core Isolation Coaling Steam Supply Line Isolation Valves The RCIC Steam Supply Line IsolationValves 1E51F0063 and 1E51F0064 are normally-open MOVs and are required to remain open to supply steam to the RCIC turbine. The control logic contains seal-ins through the motor contactors and it is possible for the valves to close due to chatter following a seismic event. There is no seal-in that would prevent the automatic closure of these valves on a valid isolation signal. Residual Heat Removal fiIIR) Valves Testable Check Valves The RHR Testable Check Valves 1El2F0041A, B, C are operated by the solenoid-operated valves 1E12F0597A,8, C which are controlled by rugged control switches. The control logic for these valves contains no SILO devices that would prevent the normal operation of these check valves. RHR Injection Valves The RHR Injection MOV (1El2F0042A, B, C) control logic contains relays and motor contactors which may chatter and result in the valves opening following a seismic event. However, the RHR testable check valves are between the injection MOVs and the RPV; an undesired opening of the RHR Injection MOVs would not result in a loss of reactor inventory and piping would not be exposed to reactor pressure. RHR Shutdown Cooling Injection Valves The RHR Shutdown Cooling Injection MOV (1E12F0053A, B) control logic relays, including the motor contactors for these valves, do not contain a seal-in and there are no other chatter sensitive contact devices involved in the control logic for this valve. Additionally, there is a check valve in series with these valves. No contact devices in this circuit meet the selection criteria. frESGonsulting (}Frzzo

2734298-R-0L5 Ratision 0 lune 28,201.7 Pase 24 of49 RHR Shutdown Cooling Isolation Valves The RHR Shutdown Cooling Isolation Valves 1E12F0008 and 1E12F0009 are normally-closed MOVs are opened via a control switch and relay permissive. While the plant is at power, the 1E12F0008 valve is de-energized by opening its disconnect; thereby preventing this valve from opening. If open, the valves will close automatically via an isolation signal. During a seismic event, chatter on the controlling relays or motor contactors could cause the 1E12F0009 valve to open, however the low reactor pressure permissive in control logic would prevent seal-in of the relays. After the period of strong shaking the normally-closed contact of the relays (isolation signal) would coflrmand the valve to reclose. Because there is no seal-in and the valves reclose without operator intervention, chatter is acceptable and no contact devices in this circuit meet the selection criteria. ftH/t Shutdown Cooltng Suction Vqlves The RHR Shutdown Cooling Suction MOV (1E12F0006A, B) control logic relays, including the motor contactors for these valves, do not contain a seal-in and there are no other chatter sensitive contact devices involved in the control logic for this valve. High Pressure Core Sprav Valve.s Testable Check Valve The HighPressure Core Spray (HPCS) Testable CheckValve, 1E22F0005, is operatedby a solenoid-operated valve, 1E22F0526, which is controlled by a rugged control switch. There are no SILO devices that would prevent the normal operation of this check valve. HPCS Injection Valve The HPCS Injection MOV (1E22F0004) control logic contains relays and motor contactors which may chatter and result in the valves opening following a seismic event. However, the lESGonsulting [iRtzz.o

2734298-R-015 Reuision 0 lune 28, 201,7 Pase 25 of 49 HPCS testable check valve is between the injection MOVs and the RPV; an undesired opening of the HPCS Injection MOV would not result in a loss of reactor inventory and piping would not be exposed to Reactor prsssure. Low Pressure Core Sprav Valves Testable Check Valve The LowPressure Core Spray Valves (LPCS) Testable CheckValve, 1E21F0006, is operatedby a solenoid-operated valve, lEZlF0524, which is controlled by a rugged control switch. There are no SILO devices that would prevent the normal operation of this check valve. LPCS Injection Valve The LPCS Injection MOV (1E21F0005) control logic contains relays and motor contactors which may chatter and result in the valves opening following a seismic event. However, the LPCS testable check valve is between the injection MOVs and the RPV; an undesired opening of the LPCS Injection MOV would not result in a loss of reactor inventory and piping would not be exposed to Reactor pressure. 2.3 RnncroR VESSEL PRESSURE Coxrnor-The reactor vessel pressure control function is identified as a key function in Reference 3 to be considered in the High-Frequency Confirmation. The same report also states that "required post event pressure control is typically provided by passive devices" and that o'no specific high frequency component chatter review is required for this function." 2.4 Cons Coolrrc The core cooling systems were reviewed for contact control devices in SILO circuits that would prevent at least a single train of non-AC power driven decay heat removal from functioning. lESGonsulting (IRtzzo

2734298-R-015 Rwision 0 June 28,201,7 Page 26 ofa9 The initial need for decay heat removal and the related scope of consideration varies based on the plant's NSSS system. The relay chatter impacts that could affect this function would be those that would cause the flow control valves to close and remain closed. For BWR plants, the decay heat removal mechanism involves the transfer of mass and energy from the reactor vessel to the suppression pool. This requires the replacement of that mass to the reactor vessel via some core cooling system; e.8., RCIC. Therefore, for this evaluation the following functions need to be checked. (l) Steam from the RPV to the RCIC turbine and exhausted to the suppression pool, (2) coolant from the suppression pool to the reactor via the RCIC pump, and (3) steam from the RPV vented to the suppression pool via the SRVs. The selection of contact devices for the SRVs overlaps with the Reactor Coolant System (RCS)/Reactor Vessel Inventory Control Category. In addition to RCIC, the HPCS system was also assessed, as this system is powered by an independent AC source. The cooling of the suppression pool, while ultimately required, is not an immediate need, so assessment of component chatter effects on systems supporting suppression pool cooling or other core cooling systems is not required. Reactor Core Isolation Cooling The selection of contact devices for RCIC was based onthe premise that RCIC operation is desired, thus any SILO which would lead to RCIC operation is beneficial and thus does not meet the criteria for selection. Only contact devices which could render the RCIC system unavailable were considered. RCIC Pump and Control Logic A vulnerability to RCIC operation following a seismic event is contact chatter leading to a false RCIC Isolation Signal or false turbine trip. A false steam line break trip has the potential to delay RCIC operation while confirmatory inspections are being made. Chatter in the contacts of RCIC Isolation Signal Relay or Steam Line High Differential Pressure Time Delay Relay may lead to a RCIC Isolation Signal and seal-in of the signal relay resulting in an Isolation of the RCIC system. Similar chatter in the contact devices that drive those relays could also lead to seal-in. lESGonsulting tiFrzzo

2734298-R-AIs Rruision 0 lune 28,201,7 Page 27 of 49 An additional vulnerability was identified involving contact chatter in the RCIC turbine trip logic and close the valve linkage arrangement forthe RCIC Trip and Throttle Valve, 1E51F0510. Closure of this linkage will require operator action to reopen the valve. These relays resulting in an undesired RCIC Isolation are listed in Table B-1, below. RCIC Injection MOV The RCIC injection MOV (1E51F0013) is normally closed, and is desired to open to permit RCIC injection. The control logic contains relays and motor contactors which include seal-ins, so chatter due to a seismic event may result in the valve opening. Opening of the injection valve without the pump running will not result in a potential RPV drain path due to the presence of the testable check valve, 1E51F0066, between the injection valve and the RPV. There are no seal-ins which would prevent the valve from opening when required. No contact devices were identified that met the criteria for selection. RCIC Steam Supply MOVs The normally closed RCIC Steam Supply MOV (1E51F0045) control logic was reviewed. Relay chaffer may result in opening of this MOV; however, this is the desired state. No contact devices were identified that would prevent the proper operation of this valve on a valid RCIC initiation signal. The normally-open RCIC Steam Supply Isolation Valves (1E51F0063, 1E51F0064) were initially reviewed in Section 2.1, above, fromtheNSSS perspective. The control logic forthese MOVs contains motor contactors that could seal-in and close these valves. There is no signal to automatically reopen these two AC-powered valves on a valid RCIC initiation signal. Therefore, chatter of these motor contactors could prevent the RCIC system from supply injection to the RPV. These motor contactors are listed inTable B-1, below. In addition, relays identified above that are associated with the RCIC Isolation Signal can also close 1E51F0064. These relays are listed in Table B-1, below. In addition, the control logic for the RCIC turbine exhaust to suppression pool valve (1E51F0068) was reviewed. This normally-open valve contains a motor contactor with a ABSGonsulting []Rtzz.o

2734298-R-015 Ratision 0 lunc 28,201,7 Page 28 of 49 seal-in, and it is possible for chaffer during a seismic event to result in closure of this valve. However, valid RCIC initiation conditions will automatically restore this valve to its desired open position. There are no seal-ins which would prevent this automatic restoration. No contact devices were identified that met the criteria for selection. Finally, the control logic forthe turbine trip and throttling valve (1E51F0510) was reviewed. The control logic for this normally-open valve does not contain any seal-in devices. No contact devices were identified that met the criteria for selection. RCIC Suction Supply MOVs The RCIC pump suction supply from Suppression Pool MOV (1E51F0031) and the suction supply from the Condensate Storage Tank MOV (1E51F0010) were reviewed for chatter impacts. Typically, the suction from the Condensate Storage Tank (CST) valve (1E51F0010) is open while the suction from the suppression pool valve (1E51F0031) is closed, as RCIC is always aligned to one suction supply or the other. The control logic for the normally-open 1E51F0010 includes motor contactors with seal-ins, and it is possible for chatter of the motor contactor or additional relays to result in closure of the normally-open valve. However, valid RCIC initiation conditions will automatically restore this valve to its desired open position. There are no seal-ins which would prevent this automatic restoration. The control logic for the normally closed 1E5 1F003 1 includes motor contactors with seal-ins, and it is possible for chatter of the motor contactor or additional relays to result in opening of this normally closed valve. However, if bothRCIC suction supply valves are open, 1E51F0010 will receive an automatic closure signal. Relays identified above that are associated with the RCIC Isolation Signal can close 1E51F0031 and inhibit the automatic signals to restore it. These relays are already listed in Table B-7, below. There are no other seal-ins which would prevent this automatic action. It is possible for the RCIC system to be in an alternate alignment with the suction supply from the suppression pool valve open and the suction supply from the CST closed. Again, it is possible for relay chatter to alter the states of these valves. As before, if both suction supply valves are closed, the 1E51F0010 valve will automatically open onaRCIC initiation signal. If hoth valves are open, the lE51F00l0 valve automatically closes. There are no other seal-ins AESGonsulting []Rtzza

2734298-R-015 Rmision 0 lurrc 28,201.7 Page 29 of49 which would prevent this automatic actiono beyond the already identified relays associated with the RCIC Isolation Signal. RCIC Minimum Flow Valve During operation of the RCIC system, the injection valve will cycle open and shut as the RPV level cycles between Level 2 and Level 8. During the times that the injection valve is shut, the minimum flowvalve (1E51F0019) is requiredto be opento protectthe RCIC pump from a deadhead condition. This valve is normally closed. The motor contactors contain seal-ins, however, no other control logic contains seal-ins. It is possible for chatter to cause the motor contactor to seal-in and open the valve, however, the control logic will then automatically restore the valve to its desired state, based on RCIC pump and RPV conditions. No contact devices were identified that met the criteria for selection. RCIC Test Return MOVs Potential diversion pathways through the RCIC test return to CST MOVs (1E51F0059, 1E51F0022) were reviewed. These valves arenornally closed. The control logic for 1E5tF0022 does not contain any devices that seal-in. The control logic for 1E51F0059 does contain a motor contactor with a seal-in as well as relays that may impact this motor contactor, however this logic is only tied to the valve closure, which is the expected and desired state. It is possible for chatter of the motor contactor itself in the open portion of the circuitry to result in the valve opening. However, if either valve is open it will automatically close on a RCIC initiation signal. This automatic action is not inhibited by any seal-in. No contact devices were identified that met the criteria for selection. High Pressure Core Spray The HPCS system is powered by an independent diesel generator. The selection of contact devices was based on the premise that HPCS operation is desired, thus any SILO which would lead to HPCS operation is beneficial and thus does not meet the criteria for selection. Only contact devices which could render the HPCS system unavailable were considered. Furthermore, component mispositions that would be automatically restored to their desired state on a Loss of Offsite Power (LOOP) or LOCA signal were screened from inclusion, unless the SILO inhibited the LOOP/LOCA signal from restoring the component to its desired state. lESGonsulting ()Rtzzo

2734298-R-015 Reuision 0 lune 28,20L7 Pase 30 of49 HPCS Pump and Control Logic The HPCS motor driven pump (1E22C0001) and control logic was reviewed to identiff any contact control devices in SILO circuits that would prevent the system from functioning. Circuits that contain seal-ins were identified, however, these seal-ins all pertain to the LOOP/LOCA initiation signal and would cause the system to initiate and the pump to start. As this is the desired state, these contact devices do not meet the criteria for selection. However, chatter on the IFC 50/51 relays located on the pump circuit breaker would result in the breaker tripping open and require an operator action to reset the lockout. These relays are listed in Table B-1,below. HPCS Injection MOV The HPCS Injection MOV (1E22F0004) is normally closed, and is desired to open to permit HPCS injection. The control logic contains relays and motor contactors which include seal-ins, so chatter due to a seismic event may result in the valve opening. Opening of the injection valve without the pump running will not result in a potential RPV drain path due to the presence of the testable check valve, 1E22F0005, between the injection valve and the RPV. There is an additional relay which may seal-in and hold the valve closed, thereby preventing it from opening. However, this seal-in is broken by a low RPV level signal as part of the LOCA initiation logic, and therefore will not prevent the valve from opening when HPCS injection is needed. Thus, no contact devices were identified that met the criteria for selection. HPCS Suction Supply Valves The HPCS suction supply from Suppression Pool MOV (1E22F0015) and the suction supply from the CST MOV (1E22F0001) were reviewed for chatter impacts. Typically, one of these valves is open while the other is closed, as the HPCS is always aligned to one suction supply or the other. The control logic includes motor contactors with seal-ins. It is possible for chatter to result in either or both of these valves to change state. However, the 1E22F0001 valve control logic also includes input fromthe 1E22F0015 limit switch. If the 1E22F0015 valve is fuIl open, 1E22F0001 will automatically close. Similarly, if 1822F0015 is closed, then lE22F000l will automatically open. Therefore, there will always be a single suction supply to the HPCS pump. AESGonsutting (IRrzzo

2734298-R-015 Rwision 0 lune 28,201.7 Page 31. of 49 There are no other relays that will seal-in or inhibit the automatic reposition due to the 1E22F001 5 limit switch. Thus, no contact devices were identified that met the criteria for selection. HPCS Minimum Flow Valve During operation of the HPCS system, the injection valve will cycle open and shut as the RPV level cycles between Level 2 and Level 8. During the times that the injection valve is shut, the minimum flow valve (1E22F0012) is required to be open to prevent the HPCS pump from failing. This valve is normally closed. The motor contactors contain seal-ins; however, no other control logic contains seal-ins. It is possible for chatter to cause the motor contactor to seal-in and open the valve; however, the control logic will then automatically restore the valve to its desired state, based on HPCS pump and RPV conditions. No contact devices were identified that met the criteria for selection. HPCS Test Return MOVs Potential diversion pathways through the HPCS test return to Suppression Pool MOV (1E22F0023), and the test return to CST MOVs (1E22F0010, 1E22F0011) were reviewed. These valves are all normally closed. The control logic for these valves contains a relay that may seal-in; however, the only consequence is a closure signal to these three valves. This is the desired state. No contact devices were identified that met the criteria for selection. 2.5 AC/DC Powrn Supponr SYSTEMS The AC and DC power support systems were reviewed for contact control devices in SILO circuits that prevent the availabilrty of DC and AC power sources. The following AC and DC power support systems were reviewed: e Emergency Diesel Generators, t Battery Chargers and Inverters, Emergency Diesel Generators (EDG) Ancillary Systems, and Switchgear, Load Centers, and MCCs. AESGonsulting (:Rtzzo

2734298-R-015 Ranision 0 lune 28,20L7 Pase 32 of49 Electrical power, especially DC, is necessary to support achieving and maintaining a stable plant condition following a seismic event. DC power relies on the availability of AC power to recharge the batteries. The availability of AC power is dependent upon the EDGs and their ancillary support systems. EPzu 3002004396 requires confirmation that the supply of emergency power is not challenged by a SILO device. The tripping of lockout devices or circuit breakers is expected to require some level of diagnosis to determine if the trip diagnose the fault condition is real or an artifact of seismically induced vibration, which could substantially delay the restoration of emergency power. In order to ensure contact chatter cannot compromise the emergency power system, control circuits were analyzed for the EDG, Battery Chargers, Vital AC Inverters, and Switchgear/Load Centers/I{CCs as necessary to distribute power from the EDGs to the battery chargers and EDG Ancillary Systems. General information on the ilrangement of safety-related AC and DC systems, as well as operation of the EDGs, was obtained from the PNPP Updated Final Safety Analysis Report (UFSAR). PNPP EDGs provide emergency power to the safety-related buses. PNPP has three divisions of Class lE loads with one EDG for each division. The analysis considers the reactor is operating at power with no equipment failures or LOCA prior to the seismic event. The EDGs are not operating hut are available. The seismic event is presumed to cause a LOOP and a noffnal reactor SCRAM. In response to bus under-voltage relaying detecting the LOOP, the Class I E control systems must automatically shed loads, start the EDGs, and sequentially load the Diesel Generators as designed. Ancillary systems required for EDG operation as well as Class lE battery chargers and inverters must function as necessary. The goal of this analysis is to identiff any vulnerable contact devices that could chatter during the seismic event, seal-in or lock-out, and prevent these systems from performing their intended safety-related function of supplying electrical power during the LOOP. The following sections contain a description of the analysis for each element of the AC/DL Support Systems. Contact devices are identified by description in this narrative and apply to all divisions. lESGonsulting tlRtzzo

2734298-R-01s Rruision 0 lune 28,20L7 33 49 Emersencv Diesel Generators The analysis of the EDGs is broken down into the generator protective relaying and diesel engine control. General descriptions of these systems and controls appear in the UFSAR. Diesel Engine Control and Protective Relaying Chatter analysis was performed for the diesel engine control logic and the diesel generator output circuit breaker, as well as the bus under-voltage and LOOP signal logic. This review also included the safety-related 4160 V switchgotr, due to interlocks and dependencies in this control logic. The control circuits for the EDG circuit breakers include bus overcurrent lockout (868) and protective relaying generator lockout (86G). Chatter of the generator lockout relay will prevent the diesel from starting. Chatter of the bus overcurrent lockout relay will cause the bus preferred supply breaker, alternate preferred supply breaker, and diesel generator supply breaker to trip open and prevent them from re-closing until the relay has been reset, The Division I and Division 2 Diesel Generator Up to Voltage auxiliary lockout relay (59DX) however will not result in a trip of the diesel output breaker and only provides a permissive for the diesel output breaker to close. An additional diesel generator lockout relay (86G1) associated with the definite time overcurrent protection, reverse power relays, are bypassed with a LOOP signal and will not preventthe diesel from starting or loading the bus if needed. The 59NX and 59EX diesel generator lockout relays are bypassed with permanently installed jumpers and will not prevent the diesel from starting or the diesel generator output breaker from closing. Chatter of the phase-overcurrent protection relays (5lA/B/CN) will result in an actuation of the included bus overcurrent lockout (868). Division 3 reverse power, definite time overcurrent protection or timing relay chatter will result in tripping the Diesel generator lockout relay (86G1) will prevent the diesel from starting or loading the bus if needed. Those relays whose chatter results in a lockout of the diesel generator and/or safety buses are listed in Table B-1, below. EDG Ancillarv Systems In order to start and operate the EDGs require a number of components and systems. For the purpose of identifying electrical contact devices, only systems and components which are electrically controlled are analyzed. Information in the UFSAR was used as appropriate for this analysis. IESGonsulting [iRtzza

2734298-R-01.5 Rwision 0 lune 28,201.7 Page 34 of 49 Starting Air Based on diesel generator availability as an initial condition the passive air reservolrs are presumed pressurized and the only active components in this system required to operate are the air start solenoids, which are covered under the EDG engine control analysis above. Combustion Air Intake and Exhaust The combustion air intake and exhaust for the Diesel Generators are passive systems which do not rely on electrical control. Lube Oil The Diesel Generators utilize engine-driven mechanical lubrication oil pumps which do not rely on electrical control. Fuel Oil The Diesel Generator Fuel Oil System is described in the UFSAR. The Diesel Generators utilize engine-driven mechanical pumps and DC-powered auxiliary pumps to supply fuel oil to the engines from the day tanks. The day tanks are re-supplied using AC-powered Diesel Oil Transfer Pumps. Chatter analysis of the control circuits for the electrically-powered auxiliary and transfer pumps concluded they do not include SILO devices. The mechanical pumps do not rely on electrical control. Cooling Water The Standby Diesel Generator Jacket Water Cooling System is described in the UFSAR. Engine-driven pumps are credited when the engine is operating. These mechanical pumps do not rely on electrical control. The electric jacket water pump is only used during shutdown periods and is thus not included in this analysis. lBSGonsulting []Ht7.zo

2734298-R-015 Rwision 0 June 28,20L7 Pase 35 of49 The Standby Diesel Generator Jacket Water Cooling System is cooled by the Emergency Service Water System (ESUD. The ESW pump (1P45C0001A8 and 1P45C0002) control logic was reviewed. Additionally, the control logic for the pump discharge MOVs (1P45F0130A/B and 1P45F0140) was reviewed. Note that the RHR Heat Exchanger (HX) inlet and outlet isolation valves (1P45F0014A/B and 1P45F0068A8) have been de-energized in the ooopen" position" There are no other MOVs along the key flowpaths to support required systems or to maintain minimum flow. Relays were identified through which chatter during a seismic event could start the ESW pumps; however, these relays do not seal-in. The control logic for the RHR HX inlet isolation valves contains contacts through which chatter during a seismic event could cause these normally-open valves to close. However, these valves will automatically open upon receipt of an ESW start signal or on a LOCA signal. This control logic does not contain any relays through which a seal-in would inhibit the automatic action. However, chatter on the IFC 50/51 and HFC 50A/C relays located on the ESW Pump A and B circuit breakers would result in those circuit breakers tripping open and require an operator action to reset. These relays are listed inTable B-1, below. No other contact devices were identifiedthat metthe criteria for selection. Ventilation The Diesel Generator Enclosure Ventilation System is described in the UFSAR. Ventilation for each Diesel Generator Enclosure is provided via two supply fans and one exhaust fan. In automatic mode the supply fans are startedviathe EDG start signal. Chatter analysis of the EDG start signal is included above. Other than SILO devices identified for the EDG start signal, chatter analysis of the control circuits for these fans concluded they do not include SILO devices. Batterv Charsers Chatter analysis on the battery chargers was performed using information from the UFSAR, as well as vendor schematic diagrams. The solid-state battery chargers each have a filtered DC output for float and equalizing modes. Each battery charger is equipped with a DC voltmeter, DC ammeter, charger failure relay, high battery voltage relay, and low battery voltage relay. The Division 3 Unit I and Unit 2 battery chargers have a high voltage shutdown circuit, which is intended to protect the batteries and DC loads from output overvoltage due to charger failure. The high voltage shutdown circuit has a magnetic latching output relay which disconnects the AESGon*ulting []Rtzu,c-

2734298-R-015 Rwision 0 lune 28,2017 Page 36 of 49 auxiliary voltage transformer, shutting the charger down. Chafier in the contacts of this output relay will cause the charger to trip and remain in a tripped state until manually reset. No other adverse impacts from chatter that would affect the availability of the battery chargers. fnverters At PNPP inverters are only used as a power supply to the Reactor Protection System (RPS). Any failure of the inverters would not prevent the RPS from performing its function to scram the reactor. No chatter analysis is necessary. Switchgear, Load Centers, and MCCs Power distribution from the EDGs to the necessary electrical loads (Battery Chargers, Fuel Oil Pumps, and EDG Ventilation Fans) was traced to identift any SILO devices, which could lead to a circuit breaker trip and internrption in power. This effort excluded the EDG circuit breakers and the ESW pump breakers which are covered in above, as well as component-specific contactors and their control devices, which are covered inthe analysis of each component above. The medium-and low-voltage power circuit breakers in switchgear and load centers supplying power to loads identified in this section are included in this evaluation. The Molded-Case Circuit Breakers used in the motor control centers are seismically rugged; and DC power distribution is via non-vulnerable disconnect switches. The only circuit breakers affected by contact devices (not already covered) were those that distribute power from the safety-related buses to the load centers. A chatter analysis of the control circuits for these circuit breakers indicates that chatter of the IFC 50/51 relays onthe 41601480 VAC transformer input circuit breakers could result in these breakers tripping open. There is no automatic closrxe signal; these breakers wouldhave to be manually reclosed. These relays are listed inTable B-1, below. 2.6 Sunamanv oF SELECTED ConnpouENTS In total 95 contact devices were identified that require a High-Frequency Confirmation. These 95 contact devices include 18 different model types encompassing 15 evaluations. A list of these contact devices requiring a High-Frequency Confirmation is provided in Appendix B. lESGonsulting tlRtzzo

2734298-R-015 Rasision 0 lune 28,201,7 Page 37 of 49 3.0 SEISMIC EVALUATION 3.1 HomzoNTAL Snrs*rrc Dnu.tun Per Reference 3, Section 4.3, the basis for calculating high-frequency seismic demand on the subject components in the horizontal direction is the PNPP horizontal GMRS, which was generated as part of the PNPP ESEP report (Reference 7) submitted to the NRC on December 19, 2014, and accepted by the NRC on September 23,2015 (Reference l9). It is noted in Reference 3 that a Foundation Input Response Spectrum (FIRS) may be necessary to evaluate buildings whose foundations are supported at elevations different than the Control Point elevation. However, for sites founded on rock, per Reference 3, "The Control Point GMRS developed for these rock sites are typically appropriate for all rock-founded structures and additional FIRS estimates are not deemed necessary for the High-Frequency Confirmation effort." The PNPP nominal plant grade elevation is 625 ft. Most major structures are founded in the Chagrin Shale bedrock at foundation elevations varying between 561 ft for the RB and the Auxiliary Building (AUX) to 564 ft for the Fuel Handling Building (FHB) and the Control Complex (CC) Building. The foundation of the Diesel Generator Building (DGB) is at elevation (EL) 615 ft founded on 30 ft of Class A backfill and 20 ft of glacial till, which extends to the in-situ rock at EL 565 ft. The design basis analysis applies the SSE ground motion at the respective building foundations. Therefore, the SSE, and the GMRS, Control Point elevation is taken to be the deepest foundation level, which is the base of the RB foundation, EL 561 ft. The bedrock immediately underlying the RB foundation (EL 561 ft) is characterized by shear-wave velocities (Vs) of about 5,200 feet per second (fl/s). The RB, AUX, FHB, and CC buildings at PNPP are founded on rock; therefore, the Control Point GMRS at EL 561 ft is representative of the input at the building foundation. For the DGB a separate FIRS is developed at EL 615 ft through a separate site response analysis to the base of the DGB at EL 615 ft. [E$Gonsulting []Rtzzo

2734298-R-01,5 Ratision 0 lune 28,20L7 38 49 The horizontal GMRS values for RB foundation (EL 561 ft) and horizontal FIRS for DGB foundation are provided in Table 3-2 and Table 3-3, respectively. 3.2 VnnTTcAL SEISMIC DEMAND As described in Section 3.2 of Reference 3, the horizontal GMRS and site soil conditions are used to calculate the vertical GMRS (VGMRS), which is the basis for calculating high-frequency seismic demand on the subject components in the vertical direction. The site's soil mean Vs vs. depth profile is provided in Reference 10, Table 5-3, and reproduced below in Table 3-l for RB foundation. TABLE 3-1 SOIL MEAN SHEAR-WAVE VELOCITY AFID I}EPTH PROFILE FOR THE FIRST 100 FT; REACTOR BUILDING FOIINDATION (EL s61 rT) Lavrn LnvnR Enn DBrru lfq Lavrcn Exu Ilnprn lml L,Lynn THrcrclBss di lftl Ysi Ift/sl di / Vsi EIdr/ Ysi I Vs30 Iftls] I 55 16.8 55 4772 0.01 1s3 0.011s3 4985 2 100 30.5 45 s273 0.008s3 0.02006 Using the Vs vs. depthprofile of RB foundation(Table 3-1),the velocrty of a shear wave traveling from a depth of 30m (100 ft) to the surface of the site (Vs30) is calculatedperthe methodology of Reference 3, Section 3.5. The time for a shear wave to travel through each soil layer is calculated by dividing the layer depth (d,) bV the shear-wave velocity (Vs) of the layer (Vs,). The total time for a wave to travel from a depth of 30mto the surface is calculated by adding the travel time through each layer from depths of 0m to 30m (E[diA/si]). The velocity of a shear wave traveling from a depth of 30m to the surface is therefore the total distance (30m) divided by the total time; i.e., Vs3g - (30 m)/E[d/Vs,]. lESGonsulting []Frzzo a a a

2734298-R-015 Raision 0 lune 28,2017 Page 39 of49 The vertical FIRS is derived using the Vertical-to-Horizontal (V/H) spectral ratio for rock sites in Western United States (WIJS) and Central and Eastern United States (CEUS) from NUREGICR-6728 (McGuire et a1., 2001) (Reference 1l). The average Vs in the upper 30 meters (100 ft) is used to weight the WUS and CEUS V/H values. The average Vs in the upper 30 meters (m) (Vs30) for EL 561 ft is 4,985 fl/sec (1,519 meters per second [rr/s]). The Vs30 for WUS and CEUS rock sites are 520 m/s and 2800 m/s, respectively (Reference ll). The V/H ratios at EL 561 ftuse aweightof (2800-1519y(2800-520):0.56 for WUS V/H ratios and ( 1 5 1 9 -520)l (2800-520):0.44 for CEUS V/H ratios. The V/H ratios from Reference 11 are also dependent on peak ground acceleration (PGA). The spectral ordinate of horizontal FIRS at 100 Hertz (Hz) is used as the PGA to determine the V/H ratios. Since the spectral acceleration (SA) of the horizontal FIRS at 100 Hz is 0.2439, the V/H ratios for the PGA range of 0.2g - 0.5g from Reference 11 are used. The vertical GMRS is then calculated by multiplying the mean V/H ratio at each frequency by the horizontal GMRS acceleration at the corresponding frequency. The resulting V/H ratios and VGMRS values for RB foundation (EL 561 ft) are provided in Table 3-2. Figare 3-1 below provides a plot of the horizontal GMRS, V/I{ ratios, and vertical GMRS for EL 561 ft at the PNPP. A similar process is used to determine the VIH spectral ratio for EL 615 ft (DGB foundation elevation). The VS30 for DGB foundation is 1,842 fl/s (562 n/s). This leads to weights of 0.98 for the WUS V/H ratios from McGuire et al., (2001) (Reference 11) and 0.02 for the CEUS V/H ratios. The 100-Hz SA value for the DGB FIRS is 0.4189, which corresponds to the V/H ratios for the 0.2g - 0.5g range. The final horizontal and vertical FIRS for EL 615 ft are shown on Table J-3 and Figure 3-2. lESGonsulting (}Rtzzo

2734298-R-A1,5 Ratision 0 lune 28,201-7 Page 40 ofa9 TABLE 3-2 HORTZONTAL AND VERTTCAL GMRS FOR RB FOUNTIATTON (EL s6r FT) FnreunNCY (Hz) HGMRS (s) v/H Rr.rto YGMRS (s) 0.10 0.0030 0.6425 0.0020 0.13 0.004s 0,642s 0.0029

0. 16 0.0065 0.6425 0.4042 0.20 0.0095 0.642s 0.006r 0.26 0.0139 0.6425 0.0089 0.33 0.0208 0.642s 0.0 t 34 0.42 0.0322 0.6266 0.0202 0.s0 0.0458 0.6146 0.0281 0.53 0.0489 0.61 3 0 0.0300 0.67 0"0626 0.6068 0.0380 0.8s 0.0784 0.595s 0.0467 1.00 0.089s 0.5882 0.0s27 1.08 0,0991 0.5863 0.0580 r.37 0.1228 0.s822 0.0713 t.74 0,1277 0.s803 0.0742 2.21

0. r 489 0.5842 0.0874 2.50 0.1769 0.5916 0.1 047 2.81 0.2103 0.6008 0.1262 3.56 0.2721 0.6258

0. I 701 4.52 0.3287 0.6636 0.2183 5.00 0.3618 0.6817 0.2465 5.74 0.4020 0.7113 0.28s9 7.28 0.4664 0.7795 0.3636 9.24 0.5424 0.8648 0.4688 10.00 0.5663 0.8956 0.s072 11.72 0.5773 0.9483 0.5474 14.87 0.5851 0.9803 0.s736 18.87 0.5976 0.9881 0.s897 23.9s 0.s788 0.9642 0.ss79 25.00 0.5722 0.9s86 0.5484 30.39 0.5389 0.9433 0.s081 38.57 0.4868 0.9455 0.4s99 48.94 0.4233 0.9708 0.4109 62.t0 0.3443 0.97 60 0.3357 78.80 0.2672 0.9s76 0.2s67 r 00.00 0.2426 0.9149 0.2219 lBEGonsutting

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2734298-R-015 Rwision 0 lune 28,20L7 Page 41. of 49 TABLE 3.3 HORIZONTAL AND VERTICAL FOUNDATION INPUT RESPONSE SPECTRA (FIRS) FOR DGB FOUNDATTON (EL 61s FT) FnnqunNCY ffiz) HFIRS (g) Y/H Rarro VF'IRS (e) 0.r0 0.0035 0.5618 0.0020 0.13 0.0049 0.5618 0.0028 0.r6 0.0071 0.s618 0.0040 0.20 0.0101 0.s618 0.0057 0.26 0.0147 0.5618 0.0082 0.33 0.02 r 8 0.s618 0.0122 0.42 0.0336 0.s340 0.0179 0.s0 0.0478 0.5 128 0.024s 0.53 0.0512 0.s 103 0.0261 0.67 0.0665 0.4994 0.0332 0.85 0.0834 0.4796 0.0400 1.00 0.0939 0.4668 0.043I 1.08 0.1 030 0.4635 0.0477 1.37 0.1297 0.4564 0.0s92 1.74

0. 1 430 0.4531 0.0648 2.21 0.1 883 0.4599 0.0866 2.50 0.2621 0.4727 0.1239 2.81 0.3677 0.4889

0. 1 798 3.56 0.s0s4 0.5327 0.2692 4.52 0.8000 0.5988 0.4790 s.00 0.9256 0.6304 0.s83s 5.74 1.04s2 0.6824 0.7132 7.28 1.1561 0.8016 0.9268 9.24 1.0515 0.9s 10 1.0000 10.00 1.0090 1.0048 1.0139 11.72 0.9249 1.0861 1.0045 t4.87 0.8937 1.12s 1 1.00s4 18.87 0.9118 1.1201 t.0214 23.95 0.890 1 1.0351 0.9214 25.00 0.879 1 1.0172 0.8942 30.39 0.7884 0.9450 0.745I 38.s7 0.6894 0.8909 0.6142 48.94 0.6096 0.8s67 0.5222 62.10 0.5203 0.8538 0.4443 78.80 0.4386 0.8s30 0.3742 100.00 0.4 r 78 0.85 10 0.3ss6 lESGsrsulting

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2734298-R-015 Rwision 0 lune 28, 2017 Page 42 of 49 HGMRS VGMRS Ratio ---r-O l/,, tair,{, \\ t-I li-tr I I rtt -r I { 'i rl i d4 / 0.70 0.60 0.50 0.40 0.30 0.20 0.10 o.m L.2 t o 0.8 ?it\\. o.6 o.4 1m.00 g tro .IPgg o C,t( th E, = (9 o.10 1.m 10.oo X'IGURE 3.1 PLOT OF'THE HORIZONTAL AND YERTICAL GROT]ND MOTION RESPONSE SPECTRA AND V/H RATIOS X'OR EL 561 rT GB X'OUNDATION) X'IGT]RE 3-2 PLOT OF'THE HORIZONTAL AND VERTICAL GROUND MOTION RESPONSE SPECTRA AND V/H RATIOS FOR EL 615 r'T (DGB X'OUNDATION) lISConsuffing ()Frzzo L.40 L.20 1.00 0.80 0.60 0.40 0.20 0.00 L.2 L o 0.9 H E,

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0.6 0.4 100.00 HFIRS VFIRS -,D - V lH Ratio ,-\\ I / 7 I ttt I I / \\, \\ \\\\ t a I / \\ \\ \\ l I It ] \\ \\\\ \\\\ t I L I ) t I aa aa 1, .-tJ ,/ L.00 ,A(I, Y tr .9 P tE l-o-o (JI Itt G,-l! 0.L0 10.00 Frequency(Hz)

2734298-R-015 Reaision 0 lune 28,201,7 Page 43 of49 3.3 CompouENT HomzoNTAL Srrsmtc Dnpr.q,nu The horizontal seismic demands to be used in this evaluation are the in-structure response spectra (ISRS) at the base of the equipment, amplified by amplification factors suggested in Reference 3 forthe specific type of equipment. Alternatively, ifthe seismic capacitiesto which the seismic demands are compared are based on assembly (e.g., cabinet) tests and the test spectra are defined at the base of the assembly, the horizontal amplification factor is taken as 1.0. The required 5% damped ISRS are obtained from Reference 12 which is developed as part of the Seismic Probabilistic Risk Assessment (SPRA) program at PNPP. If there are sharp peak(s) in the ISRS in the frequency range of interest, these peaks are clipped in accordance with the guidelines in EPRI NP-6041-SL (Reference 13). Per Reference 3, the peak horizontal acceleration is amplified using the horizontal in-cabinet amplification factor AFc to account for seismic amplification within the host equipment (cabinet, switchgear, or motor control center). The in-cabinet amplification factor, AF' is associated with a given type of cabinet construction. The three general cabinet types are identified in Reference 3 and Appendix I of EPRI NP-7148 (Reference 14) assuming 5% in-cabinet response spectrum damping. EPRI NP-7148 (Reference 14) classified the cabinet types as high amplification structures, such as switchgear panels and other similar large flexible panels; medium amplification structures, such as control panels and control room benchboard panels; and low amplification strucfures, such as motor control centers. All of the electrical cabinets containing the components subject to High-Frequency Confirmation (see Ta6le B-1 inAppendix B) can be categorized into one of the in-cabinet amplification categories in Reference 3 as follows: Switchgear cabinets 1R22S0006, 1R22S0007, artd 1R22S0009 are large cabinets consisting of a lineup of several interconnected sections typical of the high amplification cabinet category. Each section is a wide box-type structure with height-to-depth ratios of about 1.2 and may include wide stiffened panels. This results in lower stresses and hence less damping which increases the enclosure response. Components can be mounted on the wide panels, which results in the higher in-cabinet amplification factors. a lESGonsulting (:Rrzzo

2734298-R-015 Rwision 0 lune 28,2017 Page 44 of a9 t Control cabinets 1E22P0002, lHl3P06l 8, 1Hl3P0621, lHl3P0628, 1H13P0631, IHl3P0632, and lHl3P0642 arc in a lineup of several interconnected sections with moderate width. Each section consists of structures with height-to-depth ratios of in the range of 1.7 to 2.5, which results in moderate frame stresses and damping. The response levels are mid-range between motor control centers and switchgear and; therefore, these cabinets can be considered in the medium amplification category. Motor control centers 1R24S0018 and 1R24S0026 and battery chargers 1E22S0006 and 282250006 contain devices within the scope of the High-Frequency Confirmation. The seismic capacities of the devices utilize assembly based tests of the MCCs and battery chargers where the test spectra are defined at the bases of the assemblies. Therefore, amplification factors are taken as 1.0 for high frequency evaluation of the devices within these motor control centers and battery chargers. 3.4 ConnpoNENT VnnrrcAr, SErsMrc DEMANT) The component vertical demand is determined using the peak acceleration of the 5% damped vertical ISRS from Reference 12 between l5 Hz and 40 Hz and amplifuing it using the vertical in-cabinet amplification factor AFc to account for seismic amplification within the host equipment (cabinet, switchgear, or motor control center). The in-cabinet amplification factor, AF. is derived in Reference 3 and is 4.7 for all cabinet types. Altematively, if the seismic capacities to which the seismic demands are compared are based on assembly (e.g., cabinet) tests and the test spectra are defined at the base of the assembly, the vertical amplification factor is taken as 1.0. If there are sharp peak(s) in the ISRS in the frequency range of interest, these peaks are clipped in accordance with the guidelines in EPRI NP-6041-SL (Reference 13). lESGonsulting []Hru zo a

2734298-R-0L5 Raision 0 lune 28,2017 Page 45 of49 1 4.0 CONTACT DEVICE EVALUATIONS Per Reference 3, seismic capacities (the highest seismic test level reached by the contact device without chatter or other malfunction) for each subject contact device are determined by the following procedures: If a contact device was tested as part of the EPRI High-Frequency Testing program (Reference l5), then the component seismic capacrty from this program is used. 2 If a contact device was not tested as part of Reference 15, then one or more of the following means to determine the component capacity were used: Device-specific seismic test reports (either from the station or from the SQURTS testing program). Generic Equipment Ruggedness Spectra (GERS) capacities per Reference 16 and Referenc e 17. Assembly (e.g., electrical cabinet) tests where the component functional performance was monitored. The high-frequency capacrty of each device was evaluated with the component mounting point demand from Section 3.0 using the criteria in Section 4.5 of Reference 3. A total of 95 components are identified that required High-Frequency Confirmation evaluation. The 95 components are grouped into 15 main groups based on device type and capacity and enclosure dynamic characteristics and location. A summary of the high-frequency evaluation conclusions is provided in Table B-1 in Appendix B. lESGonsulting []Rtzzo a. b. c.

2734298-R-01.5 Reuision 0 lune 28,201-7 Pase 46 of49

5.0 CONCLUSION

S 5.1 Gnxnnar. CoNCLUSToNS The PNPP has performed a High-Frequency Confirmation evaluation in response to the NRC's 50.54(0 letter (Reference 1) using the methods in EPRI Report 3002004396 (Reference 3). The evaluation identified a total of 95 components that required High-Frequency Confirmation evaluation. The 95 components identified are grouped into 15 main groups based on device type and capacity and enclosure dynamic characteristics and location. The high-frequency evaluation is performed for the 15 main groups and the results are summarized in Table B-l in Appendix B, The evaluation shows that all 15 main groups (95 total components) have adequate seismic capacrty and none of the components required resolution following the criteria in Section 4.6 of Reference 3. 5.2 InBrurmrcATroN oF FoLLow-Up Acrrons For PNPP, all the identified 95 components have adequate seismic capacity and no follow-up actions were identified. lESGsnsulting fiFrzzo

2734298-R-015 Rwision 0 lune 28,241-7 Page 47 of49 I

6.0 REFERENCES

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 50.5a(f) Regarding Recommendations2.l,2.3, and 9.3 of the Near-Term Task Force Review of Insights from the Fukushima Dai-Ichi Accident," March 12, 2012, ADAMS Accession Number ML120534340. EPRI rc25287, 'oSeismic Evaluation Guidance: Screening, Prioritization and Implementation Details (SPID) for the Resolution of Fukushima Near-Term Task Force Recommendation 2. I : Seismic," February 2013. EPRI 3002004396, "High Frequency Program: Application Guidance for Functional Confirmation and Fragility Evaluation," July 2015. NRC (J. Davis) Letter to Nuclear Energy Institute (A. Mauer), "Endorsement of Electric Power Research Institute Final Draft Report 3002004396, High Frequency Program: Application Guidance for Functional Confirmation and Fragility," September 17, }}LS,ADAMS Accession Number ML152184569. NRC (W. Dean) Letter to the Power Reactor Licensees on the Enclosed List, "Final Determination of Licensee Seismic Probabilistic Risk Assessments Under the Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.5a(fl Regarding Recommendation 2.1 "Seismic" of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident," Octob er 27, 2015, ADAMS Accession Number ML I 5 1 94A01 5. FirstEnergy Nuclear Operating Company (FENOC) Seismic Hazard and Screening Report (CEUS Sites), Response to NRC Request for Information Pursuant to 10 CFR 50.54 (f) Regarding Recommendation 2.1 of the Near-Term Task Force (I{TTF) Review of Insights from the Fukushima Dai-ichi Accident: Enclosure D - NTTF 2.1 Seismic Hazard and Screening Report for Perry Nuclear Power Plant dated March 31,2014, ADAMS Accession Number ML14092A203. AESConsulting []Ftzzo 2 J 4 5 6

273U98-R-01.5 Reuision 0 lune 28,2017 Pase 48 of49 7 FirstEnergy Nuclear Operating Company (FENOC) Expedited Seismic Evaluation Process (ESEP) Reports, Response to NRC Request for Information Pursuant to t0 CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force (NTTF) Review of Insights from the Fukushima Dai-ichi Accident: Enclosure D - Expedited Seismic Evaluation Process (ESEP) Report for Perry Nuclear Power Plant dated December 19, 2014, ADAMS Accession Number ML143534.059. NRC Letter, Perry Nuclear Power Plant, Unit I - Staff Assessment of Information provided Pursuant to Title 10 of the Code of Federal Regulations Part 50, Section 50.54(f), Seismic Hazard Reevaluations for Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident, dated August 3,2015, ADAMS Accession Number MLl52084034. Stokoe, K. H., W. K. Choi, and F.-Y. Menq, 2003, "Summary Report: Dynamic Laboratory Tests: Unweathered and Weathered Shale Proposed Site of Building 9720-82 Y-12 National Security Complex, Oak Ridge, Tennessee," Department of Civil Engineering, The University of Texas at Austin, Austin, Texas, 2003. ABS ConsultinglRlZZo Associates, "Prohabilistic Seismic Hazard Analysis and Ground Motion Response Spectra, Perry Nuclear Power Plant, Seismic PRA Project," 27 34298-R-003, Revision l, 2014. McGuire, R.K, W. J. Silva, and C. J. Constantino,200l, "Technical Basis for Revision of Regulatory Guidance on Design Ground Motions: Hazard-and Risk-Consistent Ground Motion Spectra Guidelines," NUREG/CR-672 8, U.S. Nuclear Regulatory Commission, October 2001. ABS Consulting/RIZZO Associates Report 2734298-R-005, Part C and Part D, "Building Seismic Analysis, Perry Nuclear Power Plant, Seismic Probabilistic Risk Assessment Project," Reyision l, 2014. lESGonsulting tlRtzT.o I 9 10. 11. t2.

2734298-R-015 Reuision A lune 28,201.7 49 49 EPRI NP-6041-SL, "A Methodology for Assessment of Nuclear Power Plant Seismic Margin," Revision 1, Electric Power Research Institute, Jrxre 1994. Procedure for Evaluating Nuclear Power Plant Relay Seismic Functionality EPR[, Palo Alto, CA: 1990, NP-7148. EPRI 3002002997, "High Frequency Program: High Frequency Testing Summ&ry," September 2014.

16.

EPRI NP-7147-SL, "Seismic Ruggedness of Relays," August 1991. EPzu NP-7147 SQUG Advisory 2004-02, "Relay GERS Correctiorls," September 10, 2004. 13. 14. 15. t7. 18. 19. EPRI 1015109, "Program on Technology Innovation: Seismic Screening of Components Sensitive to High-Frequency Vibratory Motions," October 2007. NRC Letter, Perry Nuclear Power Plant Unit I - Staff Review of Interim Evaluation Associated with Reevaluated Seismic Hazard Implementating Near-Term Task Force Recommendation 2.1, dated September 23,2015, ADAMS Accession Number MLt 5240A032. lEEGonsutting {}Frzzo

2734298-R-015 Rasision 0 lune 28,201-7 Page A1 of A18 APPENDIXA REPRESENTATIVE SAMPLE COMPONENT EVALUATIONS AESGonsutting tiRtzz:o

A Representative Sample Component Evaluations A1.0 Purpose The purpose of hb calcuhtion b tc show tro e><amples of High Frequency Confirmation evaluaton fur sensitive conponenb hat required evabaton at Perry Nudear Power Phnt Thb calcubtiOn b in support of phnt response t NRC Near-TermTask Force recomnendaficn 2.1 for performing hph frequency confirmati6n. A2.0 Scope The compleb lst of the omponents selec{ed br Hph Frequency confinnatibn evaluafpn are hted in Table &1. The trvo componenb selecbd for e:<anple cabuhtidn are presenbd foi Table 41. These example cahulations show he dehihd procedure for he HPh Frequency confirnafpn evaluatbn. TaHe A-1: Components selected for sample High Frequerrcy confirmation ewluation LZHFA151A2H GE LE22P002 Diesel Generator 620'. 2 20.r&cP Energized (No/Nc) 1R22S0007-EL4 1R2250007-E15 1R2250007-E04 1R2250007-E13 1R2250009-001 1R2250009-E03 1R22S0009-00s 12!FC53A1A, 12rFC53B1A GE 1R2250005-E12 1R2250006-E13 1R2250006-E04 1R2250006-E09 Control Complex 620 3

3. MVSG De-Energized Report 2734298-R-015, Appendix A, Revision 0 Page A2 of A18

Report 2734298-R-015, Appendix A, Revision 0 50.54(0 NTTF 2.1 Seismic Hioh Frequencv Confirmation Example A3.0 Methodology The methodohgy in ReferenceAl willbe used to cahulate Capacity to Demand ratids for he subject rehys in he High Frequency range of 15-40 Hz. The capacity b obtained from tre EPRI HF Test program (Ref. A2), orftom GERS (Rets.47, AB &411), orothershake table tesb (Reft.44,46,412 &413) lf he EPRI Programdirl not include the specifc relay model. The EPRI HF Test Program reports a represenhfue average spectral acceleration (SA) in the high frequency range. While the capacities in ReferenoesA4, 46-48, &411-A13 are br he Low Frequenry region (i.e.,4.5-16 Hz), according to the condusions in ReferencesAl and A2, the Low Frequency capacities are always lower than he Hph Frequency capacities and therefore could be used conservatively in the HF confirmation program. The sebmic Demand to be used in his evaluatbn are tre in-structure re$ponse spectra atthe base of he equipment, amplified by amplifcation factors suggested in ReferenceAl forthe specific type of equipment. ReferenceA3 provkles the required 5% damped ISRS, which were devebped as part of he Sebmic PRA program at Perry Nuclear Power Phnt lf there are sharp peak(s) in the ISRS in the frequency range of interest (15 Hz to 40 Hz), these peaks are clipped in accordance with the guidelines in ReferenceAl0. \\iryhile not required for HF mnfirmation task, the C10% capacities are calculated and aEo reported here foreach relay using guldance in ReferenceA5. Page A3 of A18

Report 2734298-R-015, Appendix A, Revision 0 50.54(fl NTTF 2.1 Seismic Hioh Frequency Confirmation Example 440 41. M. 43. 44. 45. 46, 47. 48. A9.

410, Al l.

412, 413.

References EPRI Technical Report No. 3002004396, "High Frequency Program - Application Guidance for Functional Confirmation and Fragility Evaluation," Final Report, July 2015. EPRI Technical Report No. 3002002997, "High Frequency Program - High Frequency Testing Summary," Final Report, September 2014. ABS Consulting/Rizzo Associates Report 2734298-R-005/R7-124734, Part C & Part D, "Building Seismic Analysis of Perry Nuclear Power PIant Seismic PRA Project," Rev.1,2014. Trentec lnc. Test Report 58004.0, Rev.0, "Seismic Test Report for General Electric Relays," September 2008. NEI 12-06, Appendix H, December 2015. Electroswitch Technical Publication LOR-1, "High Speed Multi-Contact Lock-Out Relays For Power lndustry Applications," September 2A12. EPRI TR-105988,'GERS Formulated Using Data from the SQURTS Program,"April 1996. EPRI NP-7147-SL," Seismic Ruggedness of Relays," August 1991. FENOC, "Perry Relay ChatterAnalysis,'SPRA-011 Rev.0, May 18,2017. EPRI NP-6041-SL, 'A Methodology forAssessment of Nuclear Power Plant Seismic Margin," Rev.1, Electric Power Research Institute, June 1994. EPRI NP-5223-SLR1, "Generic Seismic Ruggedness of Power Plant Equipment", Rev.1, August 1991. Brown Boveri Electric, Inc. Report Number 37-51958-S, Rev.0, "Seismic Certification Report for ClasslE Electrical Equipment", March 1983. ABB PowerT&D Co., "Seismic Qualification Report RC-5503-A, Type 50D/H Overcurrent Relays," January 1997. Page A4 of A18

A5.0 High Frequency Confirmation Evaluations The HFconfinna06n of he rehysshown in SedionA2.0 aborre b perbnned in heficbnhg sedions. These evaluaficns use he rnefiodology cibd in RebrenceAl, as descrbed h Sec{ion A3.0 above. A5.1 HF Evaluation for GE Relay 12HFA151MH A5.1.1 Capacity SA := 21.309 HF sebmb capacrty of HFA151 relay in Energzed state fom Referen@ M,Tabh 5-12 Efiective spectral test capacity per Ref.A1 SA1 := SA + 0.6259 = 21.93.9 A5.1.2 Demand The 5% damped in-stucfure response specfa atthe location of he host panelin bofr horizontaland verti:al dhec{ions are shourn bebw (fiom RefierenoeA3). The HF dernand in he 1SHz b 40 lLare: SAH_O G6ZO-2:= 0.759 SAV DG62O 2:= 1.209 Maximum horizontal accebraton (X or Y direction) in the 1 SHzto 40Hz range (Note: No clipping required in the frequency range of interest) Maximum vertical accebratbn (z directbn) in he 1S{zto 40Hz range (Note: No clipping required in he frequency range of interest) Relay Model Relay Manufacturer Host Panel Additional Panels Building Elev. RRS Point EPRI Equip. Class 12HFA151A2H (excludes Code 06) GE LE22P0002 DG 620 2 20. lnstrument & Control Panel Report 2734298-R-015, Appendix A, Revision 0 50.54(fl NTTF 2.1 Seismic High Frequency Confirmation Example Page A5 of A18

Appendix A, Revision 0 50.34(fl NfTF 2.1 Seismic Hiqh Freouency Confirmation Example Page A6 of A18 Diesel Generator Bulldlng, E!. 620' Global X at Polnt 2 2.5 2 1.5 sA (e) 1 0.5 0 0.1 1 10 100 Frequency (Hzl li!tirit iitittttIi II I IItI tiaill tiIItfIr I{ itil ltttt! ltti t! -5% I!ti II {r \\ ^/ I I I / tttttl rlttt! IIttlt 2.5 Diesel Generator Buildin& El. 620' Global Y Response Spectra at Point 2 2 1.5 sA (el 1 0.5 0 0.1 1 10 100 Frequency (Hzl I \\ .A Jlt'J /

Report 2734298-R-015, Appendix A, Revision 0 Page A7 of A18 Diese! Generator Building El. 620' 3 Vertica! at Polnt 2 2.5 2 SA (g) 1.s 1 0.5 0 0.1 1 10 100 Frequency (Hz) .j i: fi t e576 il j i \\ \\nl t I l:t:{; lill\\: I i t I l I \\ \\ / / AFC H:= 4.5 Maxlnum horizontal in-cabinet amplificatibn factor ficr insfument and confol paneb per Ref.Al AFC_V:= 4.7 Maxtrrrm vertical lncbinet amplificaton hctorfor insfument and confol paneb per Ref.A1 ICRSH _pt2i= SAH_O G6ZO_Z'AFC_H [,laximum Horizontal ln-cablnet response specffa (Note: no chplng was ne@ssElry in the frequency range of lnterest) ICRSH _pt2= $.38.9 I C RSV_ pt2 i= SAV_O G62O _7 AFC_V tvlaximum vertitul in-cabinet response specffa (Note: no clpping was necessary in he frequency range of lnterest) ICRSV pe= 5.64'9

5.1.3 Capacity-Demand Ratio Fp := 1.56 FtrlS := 1.20 CDFM Knockdown factor for fiagility threshold tom hi;h frequency test program (Table 4-Z of Ref.A1) Multi-axb to single-axb correction hctor fiom section 4.5.2 of Ref. A1 I snr TRS:=l-Itn (tr*) = 16.87.9 effectivewiJe-band componentcapacity accehraton cDRH-Po,= o*ffio Capacity-Demand-Ratio in horEontal directbn CDRH pp = 5.00 > 1.0 TRS Capacity-Demand-Ratio in vertinl direction cDRv_pt2'=mffi; CDRV ptz = 2.99 > 1.0 rt Appendix A, Revision 0 50.54(fl NTTF 2.1 Seismic Hioh Frequency Confirmation Examole Page A8 of A18

Report 2734298-R-015, Appendix A, Revision 0 50.54(fl NTTF 2.1 Seismic Hiqh Frequencv Confirmation Example 5.1A HCLPF Capacities lCrr,and CrorJ The PGAused in devebping he Perry in-structure response spectra b 0.249 from RefurenceAS PGA:= 0.249 F" is the composite unceftainty for relays taken from Table H.1 of Reference A5-This composite uncertainty is considered to be Realistic Lower Bound Case according to Table H.1 of Reference A5, and it b suggested for use in cahulating the median capacity. 0":= 0.30 HCLPFHFA, s1_C1 o7o := mih (con*_tp, CDRy_pt2).PGA HCLPFHFA1 Sl C1% = 0.72.9 Am-HFA1E1-C1o7o := HCLPFHpnt s1-c1% 'e(z.rs o.) Am HFA1E1 C1o/o = 1.44'{ Ratiogl0% C1o7o:= 1.36 Ratio of Cro"rp,r, ftom Table H.1 of Ref. A5 HCLPFHFA1Sl C10% := Ratio6lg% CloT..HCLPFHFA1Sl C1% HCLPFHFA1Sl C10% = 0.98-( Page A9 of A18

A5.2 HF Evaluation for GE Relays l2lFCs3AlA and 12lFC53BiA N.2.1 Capacity According trc the guUance in RefercnceAl, he HF capacity can be establbhed based on rehy's low frequency capactty. The hph-ftequency capaclty of trb relay is establbhed based on its tesbd capacrty in he 4.5-16 Hz ftequency range in he hodzonhland verti:aldirectbns. The overall HF capacfi willbe cahuhbd by geornefric averaging of the capaciths in three orhogonaldireclions (consistentwttr Ref.Al recomrnendafon) TRS* := 9.889 Average De-EnergEed/No Contacts TRS accel. in Xdh 5o/o damping (fom test report in Ref. M, and shown in Table Below) - Tabb Limit TRSr:= 8.639 Average De-Energized/Nlo Contacts TRS accel. in Ydir, So/o damping (fom test report in Ref. M, and shown in Table Below) - Table Limit TRS=:= 8.639 Average De-Energized/No Contacts TRS accel. in Zdir, 5o/o damping (from test report in Ref.A4, and shown in Table Bebw) - Table Limit 1 sA:= (rns*.TRSy.TRSz) 3 - g.o3.g HF seismic capacity of lFC53 relays in De-Energized state 1R22S0AO7-EL4 1R2250007-E15 1R2250007-E04 1R2250007-E13 1R2250009-001 1R2250009-E03 1R2250009-005 L 12tFC53A1A 12r FC53 B1A GE 1R22S0006-E12 1R2250006-E13 1R2250006-E04 1R2250006-E09 cc 620 3

3. MVSG SA1 := SA - 9.03.g EfiectMe spectral test capacity per Ref.A1 Report 2734298-R-015, Appendix A, Revision 0 50.54(f) NTTF 2.1 Seismic Hiqh Frequency Confirmation Example Page A1 0 of A1 I

5u Danryllng TRSfroq Riltry.ate Al SSE Test#7 (De-Encr3izcdl Freq.(Hz) X-Dir. Acc. k) Y-Dl r. Acc. (g) Z-Dir. Acc. (gl 1.00 0.77 0.6 0.56 t.t2 4.76 0.8r 0.81 1.25 1.37 t.32 1.04 t.4l t.6 t.47 1.6 1.58 2.10 1.95 1.49 t.78 2.4 t.75 2.38 2.00 3.t2 2.82 3.73 2.24 3.s3 4.10 3.96 2.51 5.52 4.98 4,n 2.82 6.31 6.t7 7.00 3.t6 7.s8 7.02 6.92 3.55 9.37 7.35 7.38 3.98 9.s6 8.14 8.Tl 4.47 8.74 8.76 I 1.68 5.01 l1.80

10. l5 9.48 5.62 8.75 10.50 10.21 6.31 10.29 8.56 7.63 7.08 10.31 8.26 8.4 7.94 12.05 7.84 6.63 8.91 t0.27 6.37 7.20 10.00 10.93 8.73 8.21 11.22 8.98 10.06 I1.00 12.59 10.27 8.99 8.26 t4.t3 8.M 7.59 7.97 15.85 7.73 7.76 6.84 17.18 7.99 10.66 7.01 19.95 7.n 11.32 8.34 22.39 6.20 9.71 7.00 2s.t2 7.t8 7.81 5.12 28.18 6.36 7.00 5.93 3t.62 6.29 7.57 4,n 35.48 5.95 7.03 5.y 39.81 5.47 6.67 4.89 4.67 5.15 6.13 4.71 50.12 5.09 5.91 4.41 56.23 4.90 5.&

4.50 63.10 4.s4 s.84 4.71 70.79 4.58 5.7t 4.ffi 79.43 4.54 5.M 4.50 89.13 4.57 5.52 4.s0 100.00 4.58 5.46 4.96 Ave. Spec. Accel. (g) 9.88 8.63 9.63 Report 2734.298-R-015, Appendix A, Revision 0 50.5,4(fl,NfiF 2.1 Seismic High FrEuency Confirmation Example Page A1 1 of A18

A6'22 llemand The 5% damped h+frncture response specfra athe bcatibns of he swHrgeats (bcaEd at Pohb I and 3 h ContolCorplex E. 620) h boh horizonhland vertizldiedions arc drorvn behr (fiom ReftrcneA3). The HF demand h he 15}lz tc 40 Flz arc: SAFI_CC62O_1 != max(O.529,0.699) = 0.69.9 Maxinum horizontala@hraton (X or Y directbn) in he 15jLtc 40Hz range conespondlng to Pt1 (no *plng requfed) lvlaxirum vertical aeleratbn (z dlredbn) h he 15Flz tc 40Hz range @respondlng to Pt1 (no Splng requied) SAV_CC62O_1 := 0.799 S41_CC62O_3 := max(0.709,0.O49) = 0.70'9 [tlaxinum hor2ontala@eleraton (X or Y d[rection) ln he 151-lz t 40Hz range conespondlng t'c Pt3 (no

• phg requied)

Chped vertilal acehrafpn (z diedion) h he 15Hz to 40Hzrange @nespondhg b Pt3 (see dpplng bebw) S&V CC62O 3:= 0.719 o ,SRS at Hevatiq 620', Potnt I d UE Wol @tnplex: Contro! Complex Buildihg, EI. 620' Global X at Point 1 sA (el 1.8 1.6 L.4 L.2 1 0.8 0.6 0.4 0.2 0 \\ I I I r5% I \\ \\ I v / \\ sa + -.'/ 0.1 1 Frequency (Hzl 10 100 Report 27U298-R-015, Appendix A, Revision 0 Page A12 of A18

Report 273/.298-R{15, Appendix A, Revision 0 50,54(fl NTTF 2.1 $eismic Hiqh Frequency Confirmation Example Contrcl Complex BulldltrB, El. 620' Global Y Response Spectra at Point I 2 1.8 1.5 1.4 t.2 sA (gl 1 0.8 0.6 0.4 0.2 0 0.1 1 10 100 Frequency (Hzl -5% ,\\ ^l \\ / \\ It II \\ / ,,r-? Control Complex Buildin& El. 620' L.4 L.2 Vertical Response at Point I 1 0.8 sA (gl 0.6 0.4 0.2 0 0.1 1 10 100 Frequency (Hzl I I,. \\ e5r( ) \\_/ \\^r \\ \\ l\\ / r t -/J / Page A13 of A18

Reprt 273/,298-R-015, Appendix A, Revision 0 5Q.54(fi NTTF 2.1 Spismic Hiqh Frquency Confirmation Example O ,SRS at Hevahn 020', fuhil 3 d frE Wol @cx: Control Complex Bulldiht, El. 620' Global X Respoffie Spectra at Polnt 3 1.6 1.4 1.2 1 SA (g! 0.8 0.6 0.4 0.2 0 0.1 1 10 100 Frequency (Hzl -5% \\I\\/ \\ t I v \\ \\ / \\ t /.'/ Contrcl Complelr Buildin& El.620' Global Y Response Spectra at Point 3 t.4 t.2 1 0.8 sA (el 0.6 0.4 0.2 0 0.1 1 10 100 Frequency (Hzl e596 11 I \\ J t I \\ I Y \\ \\, a / Page A14 of A18

Report 27U298-R-015, Appendix A, Revision 0 Contrcl Complex Building, EI. 620' Vertical at Polnt 3 2.5 2 1.5 sA (el 1 0.5 0 0.1 1 10 100 Frequency (Hzl t tt $ tt iii t tl , TI

tt r itIiI

-5% til .) I "A \\ ./L / a t Cb RRS z atft=17l-lz f, := 17Hz S"J"akt= 0.989 Sa_p"ak.80% = 0.78.g t1:= 13.6H2 t2:= 20j12 AfO.8 := lZ-t1= 6.40'j1z Afo.g B:=-=0.38 fc cg:= 0.55 if BsA.2 0.4+ 0.75.8 if 0.2< B < 0.8 1.0 if B>0.8 cc = o'68 Sa_clip:= Cg.S"reak = 0.67.9 Sa_valley := 0.719 specfalaccelenafnn onesponding b he vaby at 11.SFlz sa-ctip-z := max(sa-clip, sa-vatby) = o'71'g Page A15 of A18

AFC_tt := 7.2 Maximum horkontal in+abinet amplification factor for medium voltage switchgeafti per Ref.A1 AFC y:= 4.7 Maximum verlical in+abinet amplification hctor for paneb per Ref A1 ICRSH pt1 := SAn CCOZO t.AFC H = 4.97'9 Maximum Horlzontal in-cabinet response spectra at Point 1 (Note: no clipping was necesffiry in the frequency range of interest) ICRSV pt1 := SAV CC620 t'AFC V = 3.71 'g Maximum verticalin-cabinet response spectra at Point 1 (Note: no cllpping wa$ necessary in he frequency range of interest) ICRSH_p13 := SA1_CC6Z0_3'AFC_H = 5.04'9 Maximum HorEontalin-cabinet response spectra at Point 3 (Note: no clipping was neoessary in the frequency range of interest) ICRSV_ptg := SfoV_CC620-3.AFC_V = 3.34.9 Maximum clipped vertimlin-cabinet response spectra at Point 3 Report 2734298-R-015, Appendix A, Revision 0 50.54(fl NTTF 2.1 Seismic High Frequencv Confirmation Example Page 416 of 418

A5.2.3 Capacity-Demand Ratio Fk := 1.20 FfUS := 1.20 CDFM Knockdown factor for IEEE qualificatbn test ftom Table 4-2 of Ref.A1 Mufti-axb to singh-axb conection Ectorfom section 4.5.2 of Ref.Al TRS:= t+l (rr=) = e.os-s efiective wkle-band component capacity acceleration TRS Capacity-Demand-Ratio in horizontaldirec{ion at Point 1 CDRH Pt1 IcRSH_Pt1 CDRH p11 = 1.82 > 1.0 CDRrr o+{ t= TRS v-' r ICRSV pt1 Capacity-Demand-Ratio in verti:aldirection at Point 1 CDRV pt1 = 2.43 > 1.0 cDRH-Prr'=,"*ft Capacity-Demand-Ratio in horEontaldirectbn at Point 3 CDRH p13 = 1.79 > 1.0 TRS Capacity-Demand-Ratio in vertitmldirection at Point 3 CDRrr D+2 t= Y-r.-' ICRSy p13 CDRV p13 = 2.71 > 1.0 Report 2734298-R-015, Appendix A, Revision 0 50.54(fl NTTF 2.1 Seismic Hiqh Frequency Confirmation Examole Page A17 of A18

A5.2.4 HCLPF Capacities (Crv" and CrorJ The PGAused in developing he Perry in-structure response spectra b 0.249 ftom ReferenoeA.?. PGA:= 0.249 F" is the composite uncertainty for relays taken from Table H.1 of Reference 45. This composite uncertainty is considered to be Realistic Lower Bound Gase according to Table H.1 of Reference A5, and it b suggested for use in cabulating the median capacity. 0.:= 0.30 HCLPFI FCS3_C 1 o/o_1 := m in (COnH-pt1, C DRy_ptr ) . eCn HCLPFIFCS3 C1o/o 1 = 0.44'( Am-tFCE3-c1 o/o-1 i=HcLpFtFCs3-c1 o/o-1."(2'3 ot) Ratio610% C1o7o:= 1.36 Ratio of C,,o"rpr* from Table H.1 of Ref. AS HCLPFTFCS3 C10% 1 := Ratioglg% C1I/o'HCLPFIFCSS C1% 1 HCLPFTFCSS C10% 1 =0.59.S H C L P F I FCSB_C 1 olo _Z i= m in (CO nH_pt3, C DRy_ptS) - eCn HCLPFIFCS3 C1 olo l= 0.43.9 Am-l FC53-C1 olo -3 i= HCLPF! 5c53-c 1 Yo-3'e (z.ss o") Am IFCS3 C1% 3 = 0.86.9 HCLPFIFCE3 C10% A = 0.58.8 1 = 0.88' IFC53 c1 HCLPFTFCS3 C10% 3 := RatioCl1o/o C1%'HCLPFIFCS3 C1% 3 Report 2734298-R-015, Appendix A, Revision 0 50.54(fl NTTF 2.1 Seismic Hioh Frequencv Confirmation Example Page A18 of A18

2734298-R-0L5 Reuision 0 lune 28, 201,7 PaseBl of 87 APPENDIX B COMPONENTS IDENTIFIED FOR HIGH-FREQUENCY CONFIRMATION lESGonsulting []Rtzzlo

2734298-R-015 Rwision 0 lune 28,20L7 PaseBZ of 87 TABLE B-1 COMPONENTS IDENTIFIED FOR HIGH-FREQUENCY CONF'IRMATION HF Rrlav Gnoup Uurr CowoNnnr Encrosunn Bullnnrc Floon Elnv. (ft) Conmoxunr EvaluATIoN C19/o* (s) C107o* G) ID Tvpr Svsrnna Funcrrox MamrracruRER Monnr No. ID Tvrs Brsrs ron Clpacrrv MtrrI. CtD Rarto Evalunrron Rrsulr I I 1E228-K0015 Control Relay Diesel Engine Lockout General Electric 12HEA6182341235 1E22P0002 Control Cabinet Diesel Generator 620 EPRI HF Test 3.06 Capacity > Demand 0.73 1.00 ) I 1R22Q06374 Protective Relay Actuates Bus Lockout General Electric I2IFC53AIA 1R22S0007-Et4 Switchgear Control Complex 620 IEEE/ANS r c37-98 Test t.79 Capacity > Demand 0.43 0.58 1R22Q06378 Protective Relav Actuates Bus Lockout 1R22S0007-Et4 rR22Q0637C Protective Relay Actuates Bus Lockout 1R2250007-El4 1R22Q0642A Protective Relay Actuates Bus Lockout 1R22S0007-El5 1R22Q06428 Protective Relay Actuates Bus Lockout 1R22S0007-Ets 1R22Q0642C Protective Relay Actuates Bus Lockout 1R22S0007-Els rR22Q0643 hotective Relay Actuates Bus Lockout 1R22S0007-El5 1R22Q07284 Protective Relay Actuates Bus Lockout 1R22S0006-El2 1R22Q07288 hotective Relay Actuates Bus Lockout 1R22S0006-Et2 1R22Q0728C Frotective Relay Actuates Bus Lockout 1R22S0006-Er2 1R22Q07324 Protective Relay Actuates Bus Lockout 1R22S0006-El3 1R22Q07328 Protective Relay Actuates Bus Lockout rR22S0006-El3 1R22Q0732C Protective Relay Actuates Bus Lockout 1R22S0006-El3 1R22Q08064 Protective Relay Actuates Bus Lockout 1R22S0009-001 1R22Q08068 Protective Relay Actuates Bus Lockout 1R2250009-001 1R22Q0806C Protective Relay Actuates Bus Lockout 1R22S0009-001 1R22Q0810A Protective Relav Actuates Bus Lockout 1R22S0009-E03 1R22Q08108 Protective Relay Actuates Bus Lockout 1R22S0009-E03 1R22Q0810C Protective Relay Actuates Bus Lockout 1R22S0009-E03 ABSConsulting tlRtzzo

2734298-R-01.5 Raision 0 lune 28,2017 Pase B3 of 87 TABLE B-1 COMPONBNTS IDENTIFIED FOR HIGH.F'REQUENCY C ONFIRMATION (coNTTNUED) HF Rnmv Gnoup Uurr Conpournrr Eucr,osuRn Burluuvc Floon Eluv. (ft) ComoFmxr EvnruATIoN Clo/"r' (s) Cl0Yo* (s) ID Tvpn Svsrrpr FuxcrroN MawurnCTURER Moosl, No. ID Tvrn Basrs ron Cnracrrv Mtrr. C/D Rarro Evlru^lrrou Rrsulr 2 I rR22Q0710.A Protective Relay Overcurrent Protection General Electric I2IFC53BIA 1R22S0006-E04 Switchgear Control Complex 620 IEEE/ANS r c37-98 Test 1.79 Capacity > Demand 0.43 0.58 1R22Q07108 Protective Relay Overcurrent Protection 1R22S0006-E04 lR22Q07r0C Protective Relay Overcurrent Protection 1R22S0006-E04 1R22Q07224 Protective Relay Overcurrent Protection rR22S0006-E09 1R22Q07228 Protective Relav Overcurent Protection 1R22S0006-E09 1R22Q0722C Protective Relay Overcurrent Protection rR22S0006-E09 1R22Q06r24. Protective Relay Overcurrent Protection 1R22S0007-E;04 1R22Q06128 Protective Relay Overcurrent Protection 1R2250007-E;04 lR22Q06l2C Protective Relay Overcurrent Protection lFJ2S0007-E04 rR22Q063sA hotective Relay Overcurrent Protection 1R22S0007-Et3 1R22Q063s8 Protective Relay Overcurrent Protection rR2250007-El3 1R22Q063sC Protective Relay Overcurent Protection 1R22S0007-Et3 1R22Q08214 Frotective Relay Overcurrent Protection 1R2230009-005 1R22Q08218 Protective Relay Overcurrent Protection tFJzS0009-00s lR22Q082rC Protective Relay Overcurrent Protection 1R22S0009-005 J I 868/EH12 Control Relay Bus Lockout Electro Switch 7805LR 1R22S0006-E02 Switchgear Confiol Complex 620 IEEE/ANS I c37-98 Test 2.48 Capacity > Demand 0.60 0.81 86G/EH12 Control Relay Diesel Generator Lockout 1R22S0006-E0l 868/EHl I Control Relay Bus Lockout 1R22S0007-E03 86G/EHI l Control Relay Diesel Generator Lockout 1R22S0007-E;02 lESConsulting riRtzzo

2734298-R-41,5 Rwision 0 lune 28,201.7 Pase B4 of B7 TABLE 8.1 COMPONBNTS IDENTIFIED FOR HIGH.FREQUENCY CONFIRMATION (coNrrNUED) HF RsLnv GRour Uxrr Coprpoxrxr Euclosunr Bun nmc FrooR Er,nv. (ft) Coprpourxr EvaluATIoN Cw"r' (e) C10o/o* (s) ID Tvpn Svsrru Fuxcrrou M,txuracTURER Mounl No. ID Tvpn Basrs roR C^q,pncrrv Mtr't. C/D Rarro Evalu*rrou Rnsur,r J I 868/EHr3 Conffol Relay Bus Lockout Elecfo Switch 7805LR 1R22S0009-E0l Switchgear Control Complex 620 IEEE/ANS r c37-98 Test 2.48 Capacity > Demand 0.60 0.81 86GIEH13 Control Relay Diesel Generator Lockout 1R2250009-001 4 1 42R (rEslF0063) Motor Contactor CIV Closure - RCIC Steam Supply Cutler Hammer C50C-1 Size I 1R24S0026 Motor Control Center Control Complex 620 GERS 1.24 Capacity > Demand 0.30 0.40 42R (1EsrF0064) Motor Contactor CIV Closure - RCIC Steam Supplv 1R24S0018 5 I lB2lC-K007A. Control Relay ADS Logic Amerace (Tyco) EGPD and EGPB 1H13P0628 Control Cabinet Conftol Complex 6s4 EPRI HF Test 3.26 Capacity > Demand 0.78 1.06 l82lc-K008E Control Relay ADS Logic 1H13P0628 rB2lc-K0078 Control Relay ADS Logic lHl3P063l l82lC-K008F Control Relay ADS Logic lH13P063l lB2lC-K0514 Confrol Relay ADS Logic lHl3P0628 lB2lC-K051E Control Relay ADS Logic 1H13P0628 lB2rC-K0518 Control Relay ADS Logic 1H13P0628 r82lc-Kos lF Confrol Relay ADS Logic lHl3P063t lE5lA-K008 Control Relay RCIC Steam Supply lHl3P0621 lE5lA-K0r5 Control Relav RCIC Steam Supplv lHl3P0621 lE5lA-K024 Control Relay RCIC Steam Supply 1Hr3P062l lEs lA-K033 Control Relay RCIC Steam Supplv lHt3P06l8 lE5lA-K066 Control Relay Rcic Isolation Sinnral lHl3P062l lE5lA-K067 Control Relay RCIC Steam Supply rHl3P062l lE5lA-K086 Control Relay Rcic Isolation Sisral lHr3P06l8 IESGonsulting rlRtzTo

2734298-R-015 Reuision 0 lune 28,2017 PaseBS of B7 TABLE 8.1 COMPONENTS IDENTIFIED FOR HIGH-FREQUENCY CONFIRMATION (coNTINUED) HF Rrlav Gnoup UxIr Courourur Enclosunr Burlurxc FI,ooR Elrv. (ft) Cowor*tnrur Evar,uATI oN ClYo* (s) Cl0Yo* (g) il) Tvpr Svsrrpr Furqcrrou MaUUTACTTIRER Monul No. ID Tvpr Basrs non Cnplcrrv Mrnq. C/D Rarro Evlluarlor,t Rrsulr { I lEs lA-K100 Control Relay RCIC Leak Detection Amerace (Tyco) EGPD and EGPB 1H13P0621 Control Cabinet Control Complex 654 EPRI }# Test 3.26 Capacity > Demand 0.78 1.06 lEs lA-Kl0l Control Relay RCIC Steam Supply lH13P06l8 6 I lE5rQ7064 Control Relay RCIC Isolation Sisral Agastat ETRI4B3BOO4 and ETRI4B3COO4 1Hl3P062l Control Cabinet Control Complex 6s4 EPRI HF Test 3.60 Capacity > Demand 0.86 1.17 lEslQ7065 Contol Relay RCIC Isolation Sienal lHr3P062l rEslQ7072 Control Relav RCIC Isolation Sienal lHt3P062l 1EslQ7084 Control Relay RCIC Isolation Sienal lH13P06l8 lE5lQ708s Control Relav RCIC Isolation Sipnal rHl3P06l8 7 I 1E22Q0008 Control Relav Impacts Diesel Lockout General Electric I2HFAI5IA2H rE22P0002 Control Cabinet Diesel Generator 620 EPRI HF Test 2.99 Capacity > Demand 0.72 0.98 1E22Q0009 Control Relay Impacts Diesel Lockout 1E22P0002 1E22Q0010 Control Relay Impacts Diesel Lockout 1E22P0002 1E22Q001 l Control Relay Impacts Diesel Lockout 1822P0002 lE22Q00r3 Control Relay Impacts Diesel Lockout 1E22P0002 I 1 1R22Q7021 Protective Relay Impacts Diesel Lockout Agastat ETO I2PB 1R22S0009-001 Switchgear Control Complex 620 EPRI HF Test 3.36 Capacity > Demand 0.81 1.10 9 I 1R22Q0638 Protective Relay Impacts Diesel Lockout General Electric T2IFC5IA2A 1R22S0007-Et4 Switchgear Control Complex 6?0 IEEE/ANS r c37-98 Test 1.49 Capacity > Demand 0.36 0.49 1R22Q0729 Protective Relay lmpacts Diesel Lockout 1R22S0006-Er2 1R22Q0733 Protective Relay Impacts Diesel Lockout 1R22S0006-E13 10 I rR22Q080lA Protective Relay Impacts Diesel Lockout General Electric l2rcw52B 1R22S0009-001 Switchgear Control Complex 620 IEEE/ANS r c37-98 Test 1.51 Capacity > Demand 0.36 0.49 rR22Q080lB Protective Relay Impacts Diesel Lockout 1R22S0009-001 l R22Q080 l C Protective Relay Impacts Diesel Lockout 1R2250009-001 1l I rR22Qr010 Protective Relay Impacts Diesel Lockout Brown Boveri Electric Inc. ITE.5OD lFJ2S0009-001 Switchgear Control Complex 620 IEEE/ANS r c37-98 Test 3.12 Capacity > Demand 0.75 1.02 lESGonsuEing riRlzzo

2734298-R-015 Rwision 0 June 28,2017 PaRe B6 of 87 TABLB B.I COMPONBNTS IDENTIFIED FOR HIGH-FRBQUENCY CONF'IRMATION (coNTTNUED) HF Rrrlv Gnour Unm Conrrpourxr Eucr,osuRn Burluruc Floon Elnv. (ft) CorwoFrnm Ev.r,luATIoN C1o/o* (s) C107o* (e) ID Twn Svsrnpr Fur.tcrlou M.LuuracruRER Monnl No. ID Tvrn B,lsls ron Caplclrv Mrx. CID R.q.rro Evar,uartom Rrsulr t2 I lE3lA-K005 (1E31N0702A) Control Relay RCIC Isolation Sipnal Tyco/Potter Brumfield KHS-17D12-5 lHl3P0632 Control Cabinet Control Complex 6s4 GERS 2.44 Capacity > Demand 0.59 0.80 lE3lA-K005 (lE3lN0702B) Control Relay RCIC Isolation Simal lHl3P0642 lE3lA-Ko13 (1E31N07024) Control Relay RCIC Isolation Sisral tHI3P0632 1E31A'-K013 (1E31No7o2B) Conhol Relay RCIC Isolation SiEral lHl3P0642 l3 I HVSD (1E22S0006) High Voltage Shutdown Relay Isolate Battery and Charger Potter Brumfield HVSD 1E22S0006 Battery Charger Control Complex 620 GERS 2.03 Capacity > Demand 0.49 0.66 HVSD (2E22S0006) High Voltage Shutdovrrn Relay Isolate Battery And Charger 2E22S0006 t4 1 lR22Q06r7A Protective Relay Lockout Breaker To ESW Pump General Electric I2IFC66KDIA 1R22S0007-E06 Switchgear Control Complex 620 IEEE/ANS r c37-98 Test 1,49 Capacity > Demand 0.36 0.49 lR22Q06l7B Protective Relay Lockout Breaker To ESW Pump 1R22S0007-E06 lR22Q06r7C Protective Relay Lockout Breaker To ESW Pump 1R22S0007-E06 1R22Q0712A' Protective Relay Lockout Breaker To ESW Pump 1R2250006-E05 1R22Q07128 Protective Relay Lockout Breaker To ESW Pump rR22S0006-E05 lR22Q07l2C Protective Relay Lockout Breaker To ESW Pump 1R22S0006-E05 1R22Q08144 Protective Relay Lockout Breaker To FIPCS Pump 1R22S0009-004 lR22Q08l4B Protective Relay Lockout Breaker To HPCS Pump 1R22S0009-004 lESGonsutting {}Rtzzo

2734298-R-0L5 Rruision 0 lune 28, 2017 Pase 87 of B7 TABLE 8.1 COMPONENTS IDENTIFIED FOR HIGH.FREQUENCY CONFIRMATION (coNTINUED) HF Rrlav Gnour Unm Conmonnur ErtclosuRu Buu,onvc Floon Emv. (ft) Coupouuur Evnr,uATIoN Clo/o* (s) C10g/o* (g) ID Tvpe Svsrrur Funcrtox M^l,xuracruRER Monnr, No. il) Tvru Basrs ron Ceplcrrv Mru. C/D Rarro Evnr,uarrou Rnsulr l4 I lR22Q08l4C Protective Relay Lockout Breaker To HPCS Pump General Elechic 12IFC66KD1A 1R22S0009-004 Switchgear Conffol Complex 620 IEEE/AI{SI C37-98 Test 1.49 Capacity > Demand 0.36 0.49 l5 1 lR22Q06r 8 Protective Relay Lockout Breaker To ESW Fump General Electric IZHFC22B2A 1R22S0007-E06 Switchgear Contol Complex 620 IEEE/ANSI C37-98 Test 1.74 Capacity > Demand 0.42 0.57 1R22Q0713 Protective Relay Lockout Breaker To ESW Pump 1R22S0006-E05 Ets @E!.dE ofG. lJ IIr b,O Brtq!-d rqr lESGonsulting []Rtzzo

ABSG CONSULTING INC. 300 Commerce, Suite 200 lrvine, CA 92602 Telephone 714-7344242 Fax 714-7344252 NORTH AMERICA ABSGonsulting ABS GROUP OF COMPAI.IIES,INC. 16855 Northchase Drive Houston, TX 77060 Telephone 281-673-2800 Fax 281473-2801 EUROPE Sofia, Bulgaria Telephone 359-2-9632049 Piraeus, Greece Telephone 30-21 04294046 Genoa, ltaly Telephone 39-01 0-2512090 Hamburg, Germany Telephone 4940-300-92-22-21 Las Arenas, Spain Telephone 34-944644444 Rotterdam, The Netherlands Telephone 31-10-206-0778 Amsterdam, The Netherlands Telephone 31-205-207-947 Goteborg, Sweden Telephone 46-70-283{234 Beqen, Norway Telephone 47-55-55-10-90 Oslo, Nonray Telephone 4747-57-2740 Stavanger, Nonray Telephone 47-51-93-92-20 Trondheim, Noruay Telephone 47-73-900-500 ASIA.PACIFIC Ahrnedabad, lndia Telephone 079 4000 9595 NaviMumbai, lndia Telephone 91-22-757{780 New Delhi, lndia Telephone 91-1 145634738 Yokohama, Japan Telephone 8145450-1250 Kuala Lumpur, Malaysia Telephone 603-79822455 Kuala Lumpur, Malaysia Telephone 603-2161-5755 Beijing, PR China Telephone 86-10-581 12921 Shanghai, PR China Telephone 86-21S876-9266 Busan, Korea Telephone 82-514524661 Seoul, Korea Telephone 82-2-5524661 Alexandra Point, Singapore Telephone 65S270S663 l(aohsiung, Taiwan, Republic of China Telephone 886-7-271-3463 Bangkok, Thailand Telephone 662-399-2420 West Perh, WA 6005 Telephone 61-8-9486-9909 INTERHET Additional office information can be found at: www,abslroup.com SOUTH AMERICA 2100 Space Park Drive, Suite 100 Houston, TX 77058 Telephone 713-929-6800 Energy Crossing ll, E. Building 1 501 1 Katy Freeway, Suite 100 Houston, TX 77094 1525 Wilson Boulevard, Suite 625 Arlington, VA 22209 Telephone 703-682-7373 Fax 703S82-7374 10301 Technology Drive Knoxville, TN 37932 Telephone 865-966-5232 Fax 865-966-5287 1745 Shea Center Drive, Suite 400 Highland Ranch, CO 80129 Telephone 303-674-2990 1390 Piccard Drive, Suite 350 Rockville, MD 20850 Telephone 301-907100 Fax 301-990-7185 31 15 East Lion Lane, Suite 160 Salt Lake City, UT 84121 Telephone 801-333-7676 Fax 801-333-7677 140 Heimer Road, Suite 300 San Antonio, TX 78232 Telephone 210495-5195 Fax 210495-5134 823 Congress Avenue, Suite 1510 Austin, TX 78701 Telep hone 512-7 32-2223 Fu 512-233-2210 55 Westport Plaza, Suite 700 St. Louis, M0 63146 Telephone 314-819-1550 Fax 314-819-1551 One Chelsea Street New London, CT 06320 Telephone 860-701{608 100 Danbury Road, Suite 105 Ridgefield, CT 06877 Telephone 203431{281 Fax 203431-3643 1360 Truxtun Avenue, Suite 103 North Charleston, SC 29405 Telephone 843-297-0690 152 Blades Lane, Suite N Glen Bumie, MD 21060 Telephone 410-5144450 MEXICO Maca6, Brazil Telephone 55-22-2763-7018 Rio de Janeiro, Brazil Telephone 55-21-3179-3182 Sao Paulo, Brazil Telephone 55-1 1-3707-1055 Vina del lt4ar, Chile Telephone 56-32-2381780 Bogota, Colombia Telephone 571-2960718 Chuao, Venezuela Telephone 58-21 2-959-7442 Lima, Peru Telephone 51-1 437-7430 Manaus, Brazil Telephone 55-92-3213-951 1 Montevideo, Uruguay Telephone 5982-2-901 33 UHITED KINGDOM EQE House, The Beacons Wanington Road Birchwood, Wanington Cheshire WA3 6WJ Telephone 44-1925-287300 3 Pdde Place Pride Park Derby DE24 8QR Telephone 444-1332-254-010 Unit 3b Damery Works Woodford, Berkley Gloucestershire GL13 9JR Telephone 44{-1454-269-300 ABS House 1 Frying Pan Alley London E1 7HR Telepho ne 44 -207 -377 4422 Aberdeen AB25 1XQ Telephone 44{-1224-392100 London W1T 4TQ Telephone 444-203-301 -5900 MIDDLE EAST Ciudad del Carmen, Mexico Telephone 52-938-3824530 Mexico City, Mexico Telephone 52-55-551 14240 Monteney, Mexico Telephone 52+1 +3194290 Reynosa, Mexico Telephone 52-899-920-2642 Veracruz, Mexico Telephone 52-229-980-8133 Dhahran, Kingdom of Saudi Arabia Telephone 966-3-868-9999 Ahmadi, Kuwait Telephone 965-3263886 Doha, State of Qatar Telephone 97444-13106 Muscal, Sultanate of Oman Telephone 968-597950 lstanbul, Turkey Telephone 90-2124614127 Abu Dhabi, United Arab Emirates Telephone 971-2912000 Dubai, United Arab Emirates Telephone 9714-33061 16}}