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High Frequency Supplement to Seismic Hazard Screening Report, Response to NRC Request for Information, Per 10CFR50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force (NTTF) Review of Insights from the Fukishima Dai-ichi Accident
	ML17214A639
Person / Time
Site:	Davis Besse
Issue date:	08/02/2017
From:	Bezilla M; FirstEnergy Nuclear Operating Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	CAC MF3728, L-17-233
	Download: ML17214A639 (81)
	v • d • e

FENOC Davrs-Bess e Nuclear Power Sfaft'on 5501 North Sfafe Route 2 Fhst E rr.E/ rurchar +E,'Ary Cof'Wny Oak Haftor, OH 43449 Mark B, Bezilla 419-436-1340 Sffe Vr'ce Presrdenf August 2,2017 L-17-233 10 cFR 50.54(0 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission 1 1555 Rockville Pike Rockville, MD 20852

SUBJECT:

Davis-Besse Nuclear Power Station Docket No. 50-346, License No. NPF-3 Hiqh Frequen cv Supplement to Seismic Hazard Screeninq Report. se to NRC Reouest for lnformation Pu rsuant to 10 CFR 50.54(fl Reoardino Recom mendation 2.1 of the Near-Term Task F orce (NTTF) of lnsiohts from the Fu ima Dai-ichi Accident (CAC Nos. MF3728)

On March 12,2012, the Nuclear Regulatory Commission (NRC) issued a Request for lnformation 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 Davis-Besse Nuclear Power Station (DBNPS) submitted by letter dated March 31,2414 (Reference 4). In accordance with the screening, prioritization, and implementation details repoft (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 lndividual Plant Examination for External Events (IPEEE) High Confidence of Low Probability of Failure (HCLPF)

Spectrum (HlS) to determine if a plant is required to perform a high frequency confirmation evaluation. As noted in Enclosure 2 of Reference 3, DBNPS is to conduct a limited scope high frequency evaluation (confirmation).

Davis-Besse Nuclear Power Station L-17-233 Page 2 Wilhin Reference 3, the NRC acknowledged that these limited scope evaluations will require additional development of the assessment process. The Nuclear Energy lnstitute (NEl) submitted an Electric Power Research lnstitute (EPRI) report titled, High Frcquency Program: Application Guidance for Functional Confirmation and Fngility Evaluation (EPRI 30020A4396,,} for NRC review and endorsement (References I 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 DBNPS 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 commifnents contained in this letter. lf there are any questions or if additional information is required, please contact Mr. Thomas A. Lentz, Manager- Fleet Licensing, at 330-3154810.

I declare under penalty of perjury that the foregoing is true and correct. Executed on August A , 2017.

Respectfully, Mark B. Bezilla Enclosure Near-Term Task Force (NTTF) 2.1 High-Frequency Confirmation Submittal Davis-Besse Nuclear Power Station

References:

1, NRC Letter, Request for lnformation Pursuant to Title 10 of the Code of Fedeml Regulations 50.54(0 Regarding Recommendations 2.1,2.3, and 9.3, of the Near-Term Task Force Review of lnsights from the Fukushima Dai-ichi Accident, dated March 12,2012, Agencywide Documents Access and Management System (ADAMS) Accession Number MLI 20534340.

2. NRC Letter, Electric Power Research lnstitute Repoil Final Draft Report XXXXXX, Ser'smic Evaluation Guidance Augmented Approach for the Reso/ufton of Fukushima Near-Term Task Force Recommendation 2.1: Seismh, As An

Davis-Besse Nuclear Power Station L-17-233 Page 3 Acceptable Alternative to the March 12,2012, lnformation Request for Seismic Reevaluations, dated May 7 , 2013, ADAMS Accession Number ML13106A331 .

3. NRC Letter, Screening and Prioritization Results Regarding lnformation 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 lnsights from the Fukushima Dai-ichi Accident, dated May 9, 2014, ADAMS Accession Number ML141114147.

4. FENOC Letter, FirstEnergy Nuclear Operating Company (FENOC) Seismic Hazard and Screening Report (CEUS Sites), Response to NRC Request for lnformation Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force (NTTF) Review of lnsights from the Fukushima DaLichi Accident, dated March 31, 2014, ADAMS Accession Number ML140924203.

5. NEI Letter, Fina! Draft of lndustry Seismic Evaluation Guidance (EPRI 1025287),

dated November 27,2A12, ADAMS Accession Numbers ML12333A168.

6. EPRI Report 1025287, Ser'smic Evaluation Guidance, Screening, Pioritization and lmplementation Details ISPID] for the Resolufion of Fukushima Near-Term Task Force Recommendation 2.1: Serbmrc, November 2012, ADAMS Accession Number M1123334170.

7. NRC Letter, Endorsement of Electric Power Research lnstitute Final Draft Report 1025287, Seismic Evaluation Guidance, dated February 15, 2013, ADAMS Accession Number ML1 2319AOT 4.

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 ML152234102.

10. NRC Letter, Endorsement of Electric Power Research lnstitute Final Draft Report 3002004396, High Frequency Program: Application Guidance for Functional Confirmation and Fragility, dated September 17,2015, ADAMS Accession Number M1152184569.

11. NRC Letter, Final Determination of Licensee Seismic Probabilistic Risk Assessments Under the Request for lnformation Pursuant to Title 10 of the Code of Federal Regulations 50.54(0 Regarding Recommendation 2.1 "Seismic" of the Near-Term Task Force Review of lnsights from the Fukushima Dai-ichi Accident, dated October 27 , 2015, ADAMS Accession Number ML151944015.

CC Director, Office of Nuclear Reactor Regulation (NRR)

NRC Region lll Administrator NRC Resident lnspector NRR Project Manager Utility Radiological Safety Board

Enclosure L-17-233 Near-Term Task Force (NTTF) 2.1 High-Frequency Confirmation Submittal Davis-Besse Nuclear Power Station (77 pages follow)

FIRST ENE RGYNUCLEAR o PERATING COMPANY Near-Term Task Force (NTTF) 2.1 High-Frequency Confirmation Submittal Ilavis-Besse lTuclear Power Station

APPROYALS Report Name: Near-Term Task Force OrrTF) 2.1 High-Frequency Confirmation Submittal Davis-Besse Nuclear Power Station Ilate: June 12,2017 Revision No.: Revision 0 (FEN0C):

2d T., r'r6 'l*lt Z M. Andrews, Engineer* Date 6'L1 - l l N. Aldrich, Nuclear Engineer* Date

.H b-??-t t W. Marley, Date See Forrn 10 rrfi t ?'t,st-t Eric W. J Nuclear Engineer* Date

-?r-1" Craig A. Date 6'Z'I-?jl T Nuclear Engineer+ Date Nuclear Eugineer*

Approved by (['ENOC]:

tlr Nuclear Analytical Methods* Date 6fr R. ) Nuc Elec System Engineering* Date J Nuc Supply System Engineering*

(,-?g-t7 Supv, Nuc Engineering+ Date G*L8..-\ 7 F. AIvi, Project Manager+ Date

+See comment review sheets for applicable sections reviewed

IESGonsulting 2734296-R-015 Revision 0 Near-Term Task Force (NTTF) 2.1 H ig h -Freq uency Confi rmation Submittal Davis-Besse Nuclear Power Station June 12,2017 Preparedfor:

FirstEnergy Nuclear Operating Company ABSG Consulting lnc. . 300 Commerce Drive, Suite 200 . lrvine, Califomia 92602

2734296-R-075 Rwision 0 lune 72,2477 Paxe 2 of46 NEAR-TERM TASK FORCE (NTTF) 2.1 HIGH.FREQUENCY CONFIRMATION SUBMITTAL DAVIS-BESSE NUCLEAR POWER STATION ABSG Cor'{suLTrNc INC. Rnronr No. 2734296-R-015 RrvrsroN 0 HIZ,Z,O Rrponr N0. Rl2 12-4737 Jur,{u l1r20l7 ABSG CONSULTING INC.

RIZZO ASSOCIATES lESGonsulfing

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2734296-R-01.5 Reuision 0 lune L2,20L7 Page 3 of46 APPROVALS Report Name: Near-Term Task Force (NTTF) 2.1 High-Frequency Confi rmation Submittal Davis-Besse Nuclear Power Station Date: June 12,2017 Revision No.: 0 Originator: 'fir-"Or{,rW Bradley Yagla June 12,2017 Date Engineering Associate RIZZO Associates Independent Verifier: June 12.2017 Eddie Guerra, P.E. Date Director of Structural Engineering F.IZZO Associates J-}L' uo""r,+

Principal:

( June 12.2017 Nishikant R. Vaidya, Ph.D., P.E. Date Vice President RIZZO Associates Project Manager:

7tuT June 12.2017 Farzin R. Beigi, P.E. Date Consultant F*IZZO Associates Approver: Iune 12.2017 R. Roche, P.E. Date Vice President ABSG Consulting Inc.

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2734296-R-01s Rwision 0 lune L2,2077 Pa*e 4 of46 CHANGE MANAGEMENT RECORI}

Rnvrsronr No.

Illrp I}nscnrpTloNs or CHANGES/AFFECTED PIcgs 0 June 12,2017 Initial Submiual lESGonsutfing tlRtzzo

2734296-R-01.s Ranision A lune L2,2017 Page 5 of45 TABLE OF CONTENTS PAGE LTST OF TABLES 7 LIST OF FIGURES .8 LIST OF ACRONYMS 9 EXECUTIVE

SUMMARY

ll I.O INTRODUCTION r3 l.l Punrosp l3 1.2 Bncrcnol.rND..... t3 1.3 AppnoncH........... l5 1.4 PlnNrScREEunqG........... l5 2.0 SELECTION OF COMPONENTS FOR HrGH-FREQUENCY SCREENING l9 2,I REacroR Trup/SCRAM l9 2.2 Reecron Vnssel lr-rvnNroRy CoNTRoL .........20 2.3 Reecron Vpssel Pnnssunr Cohrrnor .,.........,24 2.4 ConE Coolrmc .25 2.5 AC/DC Pownn SuppoRr Svsrnus .27 2.6 Sunnuanv op SsI-scrp,n ConmoNENTS .34 3.0 SEISMTC EVALUATION.... .35 3.1 Horuzournr- SErsurc Dnuaun .3s 3.2 Venucel Sr,lstvttc Del{et*tn .36 3.3 Courpour,rur HoruzoNTAL SErsnnc Dnueun .39 3.4 Cotvtpot*tet*trVpnrICALSgtstvflcDnti,tat*tn .41 4-O CONTACT DEVICE EVALUATIONS......... .42

5.0 CONCLUSION

S .43 5-I GENSRAI. CoucIUSIoNS.. .43 5.2 Insr*rrrrrcATroN op Follow-Up Acrrohrs ........43

6.0 REFERENCES

.44 lESGonsulting

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2734?96-R-475 Rwisim 0 lune 12,201.7 Page 6 of 46 TABLE OF CONTENTS (coNTTNUED)

APPENDICES:

APPENDIX A REPRESENTATIVE SAMPLE COMPONENT EVALUATTONS APPENDIX B COMPONENTS IDENTIFIED FOR HIGH.FREQUENCY CONFIRMATION lB$Goneulting

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2734296-R-01,s Rwision 0 lune 12,2077 Page 7 of 46 LIST OF TABLES TABLE NO. TITLE PAGE TABLE I.I GMRS AT THE DBNPS, EL 540 FT t7 TABLE I.2 SSE AT THE DBNPS. 18 TABLE 3-I SOIL MEAN SHEAR-WAVE VELOCITY VS. DEPTH PROFILE FOR THE FIRST lOO FT ...36 TABLE 3.2 HORIZONTAL AND VERTICAL GROUND MOTIONS RESPONSE SPECTRA .........38 lff,Gonsulffng tiRtzz$

2734296-R-075 Rwision 0 lune 12,2017 Page I of 46 LIST OF FIGURES FIGURE NO. TITLE PAGE FIGURE I-I COMPARISON OF GMRS AND SSE AT THE DBNPS CONTROL POrNT ELEVATION (EL s40 FT)........ .....18 FIGURE 3-I PLOT OF THE HORIZONTAL AND VERTICAL GROUND MOTIONS RESPONSE SPECTRA AND V/H RATIOS .39 lESGoneulting

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2734296-R-01,5 Reuision 0 lune 1.2,201,7 Prye I of 46 LIST OF'ACROI\WMS ABS CONSULTING ABSG CONSULTING INC.

AC ALTERNATING CURRENT AFW AUXILIARY FEEDWATER APRM AVERAGE POWER RANGE MONITOR AVV ATMOSPHERIC VENT VALVE CCW COMPONENT COOLTNG WATER CEUS CENTRAL AND EASTERN UNITED STATES DBNPS DAVIS.BESSE NUCLEAR POWER STATION DC DIRECT CURRENT DDFW DIESEL-DRIVEN FEEDWATER SYSTEM DGB DIESEL GENERATOR BUILDTNG EDG EMERGENCY DTESEL GENERATOR EFW EMERGENCY FEEDWATER EL ELEVATTON EPRI ELECTRIC POWER RESEARCH INSTITUTE ESEP EXPEDITED SEISMIC EVALUATION PROCESS FENOC FIRSTENERGY NUCLEAR OPERATING COMPANY FIRS FOUNDATION INPUT RESPONSE SPECTRA ft FEET ftls FEET PER SECOND (r

D ACCELERATION OF GRAVITY GERS GENERIC EQUIPMENT RUGGEDNESS SPECTRA GMRS GROUND MOTION RESPONSE SPECTRA Hz HERTZ ISRS IN.STRUCTURE RESPONSE SPECTRA LOCA LOSS OF COOLANT ACCIDENT LOOP LOSS OF OFFSITE POWER m METER lESGonsulting

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2734296-R-075 Rutision 0 lune 1.2,2017 Pase L0 of46 LIST OF ACROI\ryMS (coNTTNUED)

MCC MOTOR CONTROL CENTER MOV MOTOR-OPERATED VALVES m/s METER PER SECOND NRC UNITED STATES NUCLEAR REGULATORY COMMISSION NSSS NUCLEAR STEAM SUPPLY SYSTEM NTTF NEAR-TERM TASK FORCE PGA PEAK GROUND ACCELERATION PORV POWER-OPERATED RELIEF VALVE PVTR PRESSURI ZED WATER REAC TOR RB REACTOR BUILDING RCS REACTOR COOLANT SYSTEM RTZZO R-IZZO ASSOCTATES SDFW STEAM DRIVEN FEEDWATER SYSTEM SFAS SAFETY FEATURES ACTUATION SYSTEM SFRCS STEAM FEED RUPTURE CONTROL SYSTEM SILO SEAL.TN AND LOCK-OUT SCREENING, PRIORITIZATION, AND IMPLEMENTATI ON SPID DETAILS SPRA SEISMTC PROBABILISTIC RISK ASSESSMENT SEISMTC QUALIFICATION REPORTING AND TESTTNG SQURTS STANDARDIZATION SSCs STRUCTURES, SYSTEMS, AND COMPONENTS SSE SAFE SHUTDOWN EARTHQUAKE SW SERVICE WATER UFSAR UPDATED SAFETY ANALYSIS REPORT UPS UNINTERRUPTABLE POWER SUPPLY WUS WESTERN UNITED STATES V/I{ VERTI CAL.TO-HORIZONTAL Vs SHEAR-WAVE VELOCITY lESGonculting

2734296-R-075 Revision 0 lune 12,2017 Paxe L1. of 46 NEAR-TERM TASK FORCE (I{TTF'} 2.1 HIGH.FREQUENCY CONF'IRMATION SUBMITTAL DAVIS-BESSE NUCLEAR POWER STATION EXECUTIVE

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 l). In particular, this report provides information requested to address the High-Frequency Confirmation requirements of Item (4), Enclosure l, Recommendation2.l: Seismic, of the March 12,2012, letter (Reference l).

Following the accident at the Fukushima Dai-ichi nuclear power plant resulting from the March I l, 201l, 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 detcrmine 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 50.5a(fl 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 perfonn a "confirmation, 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 lmplementation Details (SPID) for the resolution of Fukushima Near-Term Task Force Recommendation 2.1:

Seismic" (Reference 6) provided screening, prioritization, and implementation details to the U.S. nuclear utility industry for responding to the NRC 50.54(0 letter. This repoft was developed with NRC participation and was subsequently endorsed by the NRC. The SPID AESGonculffng

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2734296-R-075 Rwisian 0 lune 1.2,2017 12 45 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 8) and was endorsed by the NRC in a letter dated September 17,2015 (Reference 3). Final screening identiffing plants needing to perform a High-Frequency Confirmation was provided by NRC in a leffer dated October 27,2015 (Reference 2).

This report describes the High-Frequency Confirmation evaluation undertaken for the Davis-Besse Nuclear Power Station (DBNPS). The objective of this report is to provide summary information describing the High-Frequency Confimration 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 8) is used for the DBNPS engineering evaluations described in this report. In accordance with Reference 8, the following topics are addressed in the subsequent sections of this report:

. Process of selecting components and a list of specific components for High-Frequency Confi rmation

. Estimation of a vertical ground motion response spectrum (GMRS) o 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 lESGonsulting iliRtzzo

2734296-R-01.5 Rruision 0 lune L2,2077 Pa*e 1"3 of 46

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(f) letter issued to all power reactor licensees and holders of construction permits in active or deferred status (Reference l). In particularn this report provides requested information to address the High-Frequency Confirmation requirements of Item (4), Enclosure l, Recommendation 2.1: Seismic, of the March 12,2012 letter (Reference l).

1.2 BncxcRouNr)

Following the accident at the Fukushima Dai-ichi nuclear power plant resulting from the March I l, 201 I , Great Tohoku Earthquake and subsequent tsunami, the NRC established a 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 strenglhen the regulatory framework for protection against natural phenomena. Subsequently, the NRC issued a 50.5a(fl letter on March 12,2fr12 (Reference l), requesting information to assure that these recommendations aro addressed hy 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 perfoffn a "confirmation, 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: SPID for the resolution of Fukushima Near-Term Task Force Recommendation 2.1: Seismic" (Reference 6) provided screening, prioritization, and implementation details to the U.S. nuclear utility industry for responding to the NRC 50.54(f) 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.

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2734296-R-015 Reuision 0 June 72,2077 Page 74 of46 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 8) and was endorsed hy the NRC in a letter dated September 17,2015 (Reference 3). Final screening identiffingplants needing to perform a High-Frequency Confirmation was provided by NRC in a letter dated October 27,2015 (Reference 2).

On March 31, 2014, DBNPS submitted a reevaluated seismic hazard to the NRC as a part of the Seismic Hazard and Screening Report (Reference 4). By letter dated August 25,2015, the NRC staff concluded that the GMRS that was submitted adequately characterizes the reevaluated seismic hazard for the DBNPS site (Reference 20). The seismic hazard was later reevaluated under the Expedited Seismic Evaluation Process @SEP) and submiued to the NRC on December 19, 2014 (Reference 13). The ESEP was accepted by the NRC by letter dated October 19,2015 (Reference l4). By letterdated October 27,2015 (Reference 2), theNRC transmitted the results of the screening and prioritization review of the seismic hazards reevaluation.

This report describes the High-Frequoncy Confirmation evaluation undertaken for DBNPS using the methodologies in EPRI 3002004396, "High Frequency Program, Application Guidance for Functional Confirmation and Fragility Evaluation," as endorsed by the NRC in a letter dated September 17, 2015 (Reference 3).

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.

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2734296-R-01.5 Reaision 0 June 1.2,2AL7 Page 1.5 of 46 1.3 AppnoncH EPRI 3002004396 (Reference 8) is used for the DBNPS engineering evaluations described in this report. Section 4.1 of Reference 8 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:

. DBNPS safe shutdown earthquake (SSE) and GMRS Information o Selection of components and a list of specific compontrnts for High-Frequency Confirmation o Estimation of seismic demand for subject components r Estimation of seismic capacity for subject components t Summary of subject components' high-frequency evaluations

. Summary of Results 1.4 PLnr'rr ScRrnt{Iuc The DBNPS submitted the Seismic Hazard and Screening Report in Response to the NRC Request forlnformation Pursuantto 10 CFR 50.54 (f) on March, 312014 (Reference 4). By letter dated August 25,2A15, the NRC staff concluded that the GMRS that was submitted adequately characterizes the reevaluated seismic hazard for the DBNPS site (Reference 20).

The NRC final screening determination letter concluded (Reference 2) that the GMRS to SSE comparison at the DBNPS resulted in a need to perform a High-Frequency Confirmation in accordance with the screening criteria in the SPID (Reference 6).

Subsequent to the March 31, 2014 submittal, 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 (Reference l3), and also used in the SPRA. The ESEP was accepted by the NRC by letter dated October 19,2015 (Reference l4).

Table 1-1, Table 1-2 and Figure I-l present the spectral accelerations characterizing the updated GMRSs and SSE at the DBNPS. Figure 1-1 presents the comparison of SSE, ESEP lESGsrrulting

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2734296-R-01_5 Raision 0 June 72,20L7 PaXe 1.6 of 46 GMRS (Reference 13) and the GMRS reported in the DBNPS March 2014 submittal (Reference 4). 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 strain damping of 1.0 percent is used based on the unweathered shale dynamic properties from Stokoe et al. (Reference l9).

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2734296-R-01.s Ratision 0 lune L2,201.7 Page 17 af46 TABLE 1-I GMRS AT THE I}BNPS, EL 540 FT GMRS (g)

FnreurNCY GMRS (g)

(Hz) [ESEP, DscrunrR 2014 Sunurrr,ul Sunmrrrall [Mlncn 2014 0.10 0.0032 0.0032 0.13 0.0047 0.0046 0.16 0.0067 0.0067 0.20 0.0097 0.0097 0.26 0.0140 0.0141 0.33 0.0205 0.0206 0.42 0.0307 0.0308 0.50 0.0422 0.042s 0.53 0.0443 0.0446 0.67 0.0545 0.0549 0.85 0.066s 0.0666 1.00 0.0719 0.071 I 1.08 0.0766 0.0764 1.37 0.0840 0.0834 1.74 0.0826 0.0816 2.21 0.0880 0.0869 2.50 0.0953 0.094 2.8 t 0.l l l4 0.1 094 3.56 0. l 666 0. I 609 4.52 0.2470 0.2317 5.00 0.2951 0.27 5 5.74 43744 0.3385 7.28 0.466s 0.4216 9.24 0.s30s 0.4751 10.00 0.s444 0.4889 11.72 0.s362 0.47s 1 14.87 0.4896 0.4308 18.87 0.4708 0.4013 23.95 0.4239 0.3455 2s.00 0.41 l0 0.335 30.39 0.39r 3 0.31ts 38.57 0.3877 03022 48.94 0.3572 0.2752 62.1 0 0.2878 0.2237 78.80 0.2200 0.1787 100.00 0.1993 0.1676 lESGoilsutting riRrzzo

2734296-R-015 Reuisian 0 Iutu 1.2,2077 Pase 1,8 of 46 TABLEIA SSE AT THE DBNPS ssE FREaUENcY(Hz) tgl 0.10 0.004 a37 0.06 2.31 0.374 8.00 0.374 33.00 0.15 100.00 0.15 0.60

[ESEP, Dec. 2OL4 submittalf ffiGMRS

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J J I 0.00 -/ -fllJ 0.10 1.00 10.00 100.00 Frequency (HZl FIGURE 1-1 COMPARISON OF GMRS AI{D SSE AT THE DBNPS CONTROL POINT ELEVATION (EL s40 FT) lICffing

2734296-R-01"5 Raision 0 lune 72,2077 Page 19 of46 2.0 SELECTIONOF'COMPONENTS FORHIGH-F-REQUENCY SCREENING The fundamental objective of the High-Frequoncy 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 8, this state is achieved by confirming that key plant safety functions critical to immediate plant safety are preserved (reactor trip, 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 (AC/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. Circuits that require two simultaneous relay chaffers in order to reposition or lock-out a component are deemed improbable. However, if one set of contacts may seal-in, the second set of contacts were still analyzed. Conhol switches and local pushbuttons are not subject to high-frequency chatter and thus are not considered. This analysis is based on the plant operating at 100% power and all normal equipment is available and in the normal configuration; i.e., no equipment is out of service, o.9., for maintenance or testing.

Accordingly, the objective of the review as stated in Section 4.2.1 of Reference I is to determine if seismic induced high-frequency relay chatter would prevent the completion of the following key functions.

2.1 Rnncron Trur/SCRAM The reactor trip/SCRAM function is identified as a key function in Reference I to be considered in the High-Frequency Confirmation. Thb 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.

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2734296-R-01,5 Ratision 0 June 1.2,2017 Paxe 20 of46 2.2 RmcroR VESSEL Ir{vEI,[ToRy Cournol The reactor coolant system/reactor vessel inventory control systems were reviewed for contact control devices in seal-in and lockout (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 (valves opening or the inability to close the valves). 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.

Reactor coolant system/reactor vessel inventory control system reviews were performed for valves associated with the following functions:

I Pressurizer Safety Valves (PORV, Sample Lines, High Point Vents)

. Letdown Line Valves

. Low Pressure Injection / Core Flood Tank Line Valves

. High Pressure Injection Line Valves

. Reactor Coolant Pump Seal Return Isolation Valves Pressurizer Safetv Valves PORV, Sample Lines, Hish Point Vents)

Power-Operated Relief Valve (RC2A [PORV])

The PORV is a solenoid-operated pilot valve controlled via relays, which are controlled by reactor coolant system (RCS) pressure relays. The RCS pressure relays are solid state and do not seal-in, thus there is no seal-in or lockout relays in this logic which would cause a loss of RCS inventory out of the PORV.

IRC,S Sample Line Valves (RC21?A, RC239B, RC240A, RC240B, RC200, and RC4632)

These are noffnally-closed motor-operated valves (MOVs) and solenoid valve (RC4632). In order to lose RCS inventory through the sample line, two to three valves would need to open.

The MOVs' open circuitry are controlled by hand switches which are not susceptible to chatter, while the closed circuit can be controlled by the hand switch or (for RC240A and RC240B) a Safety Features Actuation System (SFAS) signal. Relay chatter of the MOV contactor (42lO) is possible and could lead to a potential of a very small leak path of the RCS to the Quench Tank or lreGonsulting

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2734296-R-01.5 Reaision 0 lune L2,2017 PaXe 2L of46 the Sampling System through the opening of these valves; therefore, these relays were included for evaluation. The relays evaluated include BEI l8ll42 (RC240A), BFl285l42 (RC200),

BF I I 26142 (RC239A), BF rl27l42 (RC239B), and BF I 128/42 (RC240B). Solenoid Valve RC4632 has a relay (SV4632/4) whose contacts could chatter, seal-in, and open the valve; therefore, that relay was included for evaluation.

High Point Vent Valves (RC4608A, RC46088, RC46I0A, and RC4610B)

RC46084, RC46088, RC46I0A, and RC46I0B are all normally-closed solenoid valves. They are controlled by control switches in the control room, which are not susceptible to chatter.

There are no seal-in contacts to cause the valves to open. Therefore, there are no SILO devices susceptible to chatter.

Letdown Line Valves Letdown Line Valves (MU?B, MUIA, MUIB, MUZA, and MUj)

MUZB, MU2A, MUIA, and MUIB are normally-open MOVs. MU3 is a normally-open air-operated valve. To isolate letdown either MU2B, MU2A, or MU3, or the combination of MUIA and MUIB need to close.

Letdown Isolation Valve MU2B can be closed by the hand switch. Additionally, the valve will automatically close if there is high pressure at the Reactor Coolant Letdown Cooler Component Cooling Outlet Valve or high temperature on the Letdown Line Delay Coil. The closing coil does have seal-in contacts, but the limit switch contacts and torque switch contacts open when the valve is closed, releasing the seal-in. Additionally, for this purpose closing the valve is beneficial and these contacts can be screened. The open portion of the MU2B circuit is controlled by a hand switch only. Since the valve is normally open, chatter of the 42lO contacts could only re-open the valve after the valve had started to close. The open coil does have seal-in contacts, but the limit switch contacts and torque switch contacts open when the valve is open (returned to its normal position), releasing the seal-in. Control of the valve is returned to its hand switch and the automatic interlocks; therefore, this relay was not evaluated.

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2734296-R-01.5 Reuision 0 lune 12,2077 Page 22 of46 Letdown Coolers Outlet Valve MU2A can be closed by the hand switch. Additionally, if there is a SFAS Level 2 signal (RCS pressure below 1600 PSI or Containment Pressure greaterthan 18.7 PSI), the SFAS contacts will close to energize the 42lC coil. The closing coil does have seal-in contacts, but the limit switch contacts and torque switch contacts open when the valve is closed, releasing the seal-in. Additionally, for this pu{pose, closing the valve is beneficial so these contacts can be screened. The open portion of the MU2A circuit is controlled by the hand switch only. Since the valve is normally open, chatter of the 42lO contacts could only re-open the valve after the valve had started to close or is closed. The open coil does have seal-in contacts, but the limit switch contacts and torque switch contacts open when the valve is open (returned to its normal position), releasing the seal-in. Control of the valve is returned to its hand switch; therefore, this relay was not evaluated.

Letdown Cooler Inlet Valves MUIA and MUIB can be closed by the hand switch. Additionally, if there is high pressure at the associated Reactor Coolant Letdown Cooler Component Cooling Outlet Valve or high temperature on the Letdown Line Delay Coil, the valve will automatically close. The closing coil does have seal-in contacts, but the limit switch contacts and torque switch contacts open when the valve is closed, releasing the seal-in. Additionally, for this purpose, closing the valve is beneficial so these contacts can be screened. The open portion of the MUIA and MUIB circuit is controlled by position of the Component Cooling Water to the Decay Heat Exchanger Valve (CC1409, CC14l0). If CCl409 (CCl4l0) is fully opon, then MUIA (MUIB) will receive an open signal sincethe 33/ao contactswould close. The 42/O open coil for MUIA (MUIB) does have seal-in contacts, butthe limit switch contacts and torque switch contacts open with the valve is open (returned to its normal position), releasing the seal-in. Because the position switch contacts in this circuit are opcn when the valve is in the contacts could only open MUIA (MUIB) afterthe valve had startedto close. Therefore, the MUIA/I{UIB 42lO relay was not evaluated.

Due to the interlock between CCl409/CCl4l0 and MUIAA{UIB, the circuits for CC1409/CCl4l0 were also reviewed. These valves are noffnally-open MOVs that can he closed using the hand switch for MUIA/MUIB through time delay relay contacts in the MUIA/IVIUIB closing circuit. The time delay relay contacts do not seal-in so they can be screened. The CCl409lCCl4l0 closing coil does have seal-in contacts, but the limit switch contacts and torque switch contacts open when the valve is closed, releasing the seal-in. tf CCl409lccl4l0 is lESGontulting

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2734296-R-01s RutisionA lune 1,2,2017 Pase 2i of46 closed, letdown temperature will increase and TSH3745A contacts will close resulting in a closure of MUIAA{UIB. Closure of CCl409/CCl410 does not prevent MUIA/IvIUIB from closing and therefore has no impact on letdown isolation. Therefore, the CCl409/CCl410 42lC relay contacts can be screened. The open portion of the CCl409/CC14l0 circuit is controlled bythe MUIA/I{UlB hand switch only. Since the valve is normally open, chatterof the CC 1409/CC 14 I 0 42lO contacts could only re-open the valve after the valve had started to close. The open coil does have seal-in contacts, but the limit switch contacts and torque switch contacts open when the valve is open (returned to its normal position), releasing the seal*in.

Control of the valve is returned to the hand switch; therefore, this relay was not evaluated.

Pneumatically operated MU3 is normally open; when the solenoid is de-energized the valve will close. Power will be removed via SFAS Level 2 signal should a signal be generated, resulting in opening of contacts causing the valve to close. The valve can also be de-energized if the accumulator has low pressurs causing the pressure switch contacts to open. Chatter of either the SFAS contact or the pressure switch can result in closure of the valve, which is a beneficial result for this analysis. Thereforeo there are no seal-in or lock-out relays present that would prevent MU3 from closing.

Low Pressure IniectionlCore Flood Line Valves Check Valves There are two core flood tank to reactor coolant system check valves (CF30/CF3l) inside containment preventing flow from the RCS into the Low Pressure Injection/ Core Flood line.

The check valves are not chatter sensitive and can be credited to remain closed; thus, these valves do not need to be analyzed.

DHI I and DHIZ The two RCS to decay heat system MOVs DHI I and DHl2 arc normally-closed MOVs in series. If both were to unexpectedly open, it could result in an Inter-System Loss of Coolant Accident (ISLOCA). Therefore, these valves were reviewed for SILO devices. DHI I and DHl2 have control power removed during plant heat up and power operations by opening a set of contacts (control room switch) preventing power from passing to the rest of the circuit; thus, lESGonsulting

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2734296-R-0L5 Reaision 0 lune 1.2,20L7 Paxe 24 of46 chatter of downstream contacts cannot cause the valves to reposition. Additionally, a high pressure switch in the open circuitry would have open contacts while at power. DHI I and DH12 were found to not have any relays that could seal-in and cause the valve to reposition open without simultaneous chatter from the control power contacts and the pressure switch contacts.

With DHll and DHl2 remaining closed, an ISLOCA would be avoided and no relayswere evaluated for these valves.

Hish Pressure Iniection Valves Check Valves In each of the four High Pressure Injection lines into the reactor coolant system, there are two check valves in series inside containment (HP48/HP50, HP49/HP5I, HP56/HP58, and HP57/HP59) preventing flow from the RCS into the High Pressure Injection line. The check valves are not chatter sensitive and can be credited to remain closed; thus, these valves do not need to be analyzed.

Reactor Coolant Pump SeaI Return Valves Valves MU59A, MU59B, MU59C, and MU59D To isolate Seal Return, all MU59A-D valves need to close, MUsgA-D are all normally-open valves that need to reposition closed. MUS9A-D uses a similar control diagram such that a momentary chatter of the 42lC contacts would result in a closure of the valve. For this purpose closing the valve is beneficial and the 42lC contacts can be screened. There are limit switches that do not close unless the valve is some percentage closed or is closed, thus preventing the valve from opening due to a chatter of the 42lO contacts. Therefore, open position switches in the opening circuit prevent seal-in of the opening contactor auxiliary contact and no other contacts prevent valve closure via the control switch. Thus, these valves are not affected by seal-in or lock-out.

2.3 RB.+,croR VESSEL PRESSURE COr{TROL The reactor vessel pressure control function is identified as a key function in Reference 8 to be considered in the High-Frequency Confirmation. The same report also states that'orequired post lESGonsulting (iRrzzo

2734296-R-01.5 Reoision 0 lune L2,20L7 Pnse 25 of46 event pressure control is typically provided by passive devices" and that "no specific high-frequency component chatter review is required for this function." As confirmation, the pressure control function at Davis-Besse was reviewed.

The DBNPS RCS pressure control is provided by essential Pressurizer Heaters, the Power-Operated Relief Valve (PORV), and the Pressurizer Safety Valves (PSVs).

A review of the essential Pressurizer Heater control circuits found no relays that could seal-in or lock-out and prevent control room use of the heaters.

The PORV (RC2A) is a solenoid-operated pilot valve controlled via relays, which are controlled by RCS pressure relays. The RCS pressure relays are solid state and do not seal-in, thus there is no seal-in or lockout relays in this logic which would cause a loss of RCS inventory out of the PORV (and a reduction in pressure).

The PSVs are spring-operated valves that open when the RCS pressure exceeds the lift set point.

There are no relays that would seal-in or lock-out and cause the PSVs to open. Therefore, there are no relays that could impact RCS pressure control.

2.4 Conn Coor,Ir.tc 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.

The initial need for decay heat removal and the related scope of consideration varies based on the plant's nuclear steam supply system (NSSS). 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 Pressurized Water Reactor (PWR) plants, it is expected that the availability of an AC-independent (steam or diesel-driven) alternate/emergency feedwater pump to supply water to at least one steam generator will be sufficient to satisfy the immediate decay heat removal needs.

Therefore, for this evaluation, the following function needs to be checked: circuitry to detect and isolate flow to a faulted steam generator and MOVs relied upon for feeding the steam generator.

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2734296-R-01,5 Ratision 0 lune 12,201.7 Pase 26 of46 The selection of contact devices for the Steam Driven Feedwater System ([SDFW] - Auxiliary Feedwater [AFW] at Davis-Besse] and Diesel-Driven Feedwater System ([DDFW] - Emergency Feedwater BFW] at Davis-Besse) were based on the premise that AFWIEFW operation is desired; thus, any SILO which would lead to AFWEFW operation is beneficial and thus does not meet the criteria for selection. Only contact devices which could render the AFWIEFW system inoperable were considered. A review of the AFW system was broken down into the steam side and the flow side.

The steam side included a review of the valves required to allow steam to pass from the steam generator to the Auxiliary Feed Pump Turbine (AFPT). The valves reviewed include MOVs MSl06, MSI06A, MSl07, and MSIO7A, and air-operated valves (AOVs) MS5889A, and MS58898. For the MOVs, there are contactors with potential to seal-in and prevent the Main Steam valves from going to their desired position (open); however, a torque switch would open once the valve has gone fully closed and clear the seal-in. After that point, the Steam and Feedwater Rupture Control System (SFRCS) would reposition the valve to the desired position.

The AOVs (MS5889A, MS5889B) fail open on loss of power to the associated solenoid valves.

SFRCS deenergizes the solenoids. There are no relays in the circuit which could chatter, seal-in, and prevent SFRCS from de-energizing the solenoid valves to allow the valves to open.

Therefore, there are no contact devices in the steam line valves to the AFPTs identified for evaluation. Additionally, if an air-operated Main Steam lsolation Valve (MSl00, MSl0l) fails to close during a seismic evento the open pathway could depressurize the associated steam generator and divert steam from the AFPT. Therefore, the circuits for MSl00 and MSlOl were reviewed. The spring-loaded AOVs fail closed on loss of power to the associated solenoid valves or on low air pressure. SFRCS deenergizes the solenoids to close the valves. There are no relays in the circuits which could chatter, seal-in, and prevent SFRCS from de-energizing the solenoid valves to cause the valves to close. Therefore, those valves were screened.

The flow side included a review of the valves required to move water from the safety-related feedwater source (service water) to the SGs. The Condensate Storage Tank is not seismically qualified and therefore is not credited during the seismic event. The valves reviewed include MOVs AF599, AF608, AF3869, AF3870, AF387l, AF3872, SWl382, and SWI383, and solenoid valves FV645l and FV6452. Additionally, the Service Water (SW) Pump circuits were reviewed. From review of the MOVs required to allow flow from the Auxiliary Feed Pumps to the Steam Generators there are some contactors with potential to seal-in and prevent the AFW lESGonsulting tiRtzzo

2734296-R-075 Reaision A lune 1.2,2017 Paxe 27 of46 flow to the SGs; however, a torque or limit switch would open once the valve has gone fully closed and clear the seal-in. At that point, either SFRCS would reposition the valve to the desired position, or the SW valve would open when low suction pressure is sensed. Solenoid valves FV645l and FV6452 automatically throttle flow based on steam generator water level and the valves fail open on loss of power. A review of their circuits confirmed there are no relays whose contacts could chatter, seal-in, and prevent the valves from allowing the flow of water to the SGs.

A chatter analysis of the SW pump circuit breaker control circuits indicates the bus lockout, phase overcurrento and ground fault relays all could prevent automatic breaker closure following the seismic event and were identified for evaluation.

Since the diesel-driven EFW system is manually started and manually terminated by hand switch operation only, there are no contact devices in the pump circuit that meet the selection criteria.

The flow path was also reviewed and found to contain no contact devices that would seal-in or lock-out thus preventing flow to the Steam Generators. Steam relief from the SGs to the atmosphere will be done through local operator action of the Atmosphere Vent Valves (AVVs) using installed remote controllerso therefore relay chatter from the AVVs is irrelevant. In addition, there are no high-frequency sensitive devices within the control circuitry of the AVVs that would seal-in or lock-out and prevent manual operation of the AWs.

2.5 AC/DC Powrn Supronr Svsrrprs The AC and DC power support systems were reviewed for contact control devices in SILO circuits that prevent the availability of DC and AC power sources. The following AC and DC power support systems were reviewed:

. Emergency Diesel Generators (EDGs),

t Battery Chargers and Inverters, o EDG Ancillary Systems, and o Switchgear, Load Centers, and MCCs.

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 lBSGonsulting

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2734295-R-075 Reaision A lune 72,2477 Paxe 28 of46 recharge the batteries. The availability of AC power is dependent upon the EDGs and their ancillary support systems. EPRI 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 of the faulted condition could delay the restoration of emergency power.

In order to ensure contact chatter cannot compromise the emergency power system, control circuits were analyzed for the EDGs, Battery Chargers, Vital AC Inverters, and Switchgear/Load Centers/JvlCCs as necessary to distribute power from the EDGs to the Battery Chargers and EDG Ancillary Systems. General information on the arrangement of safety-related AC and DC systems, as well as operation of the EDGs, was obtained from the DBNPS UFSAR. The DBNPS EDGs provide emergency power to the safety-related busses. The DBNPS contains two trains of Class lE loads with one EDG for each train.

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 but are available. The seismic event is presumed to cause a Loss of Offsite Power (LOOP) and a normal reactor SCRAM.

In response to bus under-voltage relaying detecting the LOOP, the Class lE 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/DC Support Systems. Contact devices are identified by description in this narrative and apply to both trains.

E mergencv D ie+ el 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.

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2734296-R-015 Rutision 0 lune L2,2017 PaRe 29 of45 Generator Protective Relaying The control circuits for the EDG circuit breakers include bus lockout, EDG lockout, and phase over-current protective relays. Chatter in any of the bus lockout or EDG lockout relays may prevent closure of the EDG circuit breaker.

The bus lockout relay could actuate due to relay chatter, resulting in closure of contacts that would cause the EDG output breaker from closing and the EDG would not re-power the bus.

When the bus lockout relay is tripped, it prevents closure of the EDG, CCW pump, SW pump, and load center transformer circuit breakers.

The relays that could lead to a bus lockout include:

o Relays ACl03/5 l-1, AC103/5 IGS-1, ACl03/5 l-2, ACl03/5l-lx, ACl03/51-2, ACl03/51-3, ACl03l5lGS, and ACI l0/5lX can cause the bus lockout relay to actuate and prevent EDG 1 from loading the 4160 bus Cl_41.

I Relays ADI 03/5 l-1, AD103/5 ICS-l, ADl03/51-2, ADl03/5 l-lX, ADl03/51-2, AD103/51-3, ADl03/5lGS, and ADI l0/5lX can cause the bus lockout relay to actuate and prevent EDG 2 from loading the 4160 bus DI EA.

The EDG output breaker can be prevented from closing due to actuation of the over-current relay. The over-current relays AClDllSlV2DG or ADI01/51V2DG contacts could chatter and seal-in resulting in an energization of the EDG output breaker trip coil.

The EDG lockout relays are not actuated until after the EDG has started and reached at least 200 RPM. The current is stopped by the open contacts SS2X and KIX; thus, if there were chatter of any of the EDG protection circuits (ground over current, phase over current, differential, or engine auto shutdown) the lockout relays would not energize and seal-in. If chatter of the protection relays were to occur after the EDG was started due to under voltage and has reached 200 RPM, the R3 contacts would open and prevent the 86- l coil from being energized. Only chatter of the shutdown relay contacts (SDRX or SDRXI) could result in energizing the 86-2 coil and seal-in causing the EDG to shut down under those conditions.

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2734296-R-075 Reaision 0 lune 1.2, 201.7 Paxe 30 of45 The SFAS interlock can also cause the output breaker trip coil to energize. Chatter of the 94-l relay or chatter of R3Xl can cause the 94-l coil to energize thus causing the EDG trip coil to energize.

Diesel Engine Control Chatter analysis for the diesel engine control was performed on the start and shutdown circuits of each EDG. The start circuit is blocked by seal-in of the engine trouble SDRXI shutdown or start failure relays. Chatter of the seal-in contacts of these relays or other relay contacts that actuate the shutdown relay (SDR) may prevent EDG start.

The start failure relay (TDl) is actuated if the cooling water output pressure is less than 20 PSI, or Diesel Speed does not reach 200 RPM after 7 seconds following a start signal. If either of those conditions is not reached within 7 seconds, the fail to start coil (TDl) actuates causing the TDI contacts to close and energize the R2 contact resulting in passing current to the shutdown coils that seal-in and cause the generator to reduce speed. Chatter in the contacts of the R2 relay may also energize the shutdown relay (SDRX) and seal-in.

The EDG is emergency started by an under-voltage signal that closes2TZ-3 contacts l-2. Once closed the R3X coil is energized, closing the R3X contacts starting the redundant fuel pump, and energizing the RIX coil. The RIX coil then closes the RIX contacts, energizing the AVIA and AV2A coils which opens the air start solenoid valves and the engine starts to pick up in speed.

Once the engine reaches 40 RPM the SSIA contacts close, and the SSIX coils energize. As the EDG increases in speed, the SS2A contacts close energizing the SS2X coil at 200 RPM, the SS3A contacts close energizing the SS3 coil at 400 RPM, and the SS4A contacts close energizing the SS4 coil at 800 RPM.

As the engine comes up in speed, at 400 RPM, the field flash occurs from closure of the SS3 contacts, and the generator starts developing an output voltage. Once the field flash occurs, and voltage raises the CR3X coil is energized by a solid-state "V" relay (elecho-mechanical relay output contacts), causing the EDG output breaker to close. If the SS3 contacts would chatter and the EDG did not receive its start signal, the field flash could potentially burn itself out such that when called upon during emergency actuation, the field would not flash and voltage would not lESGonsulting

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2734296-R-075 Rutision 0 lune 1.2,2017 Page 3L of46 be developed and the CR3X coil would not energize thus the EDG output breaker would not close.

Chatter of the SDR R7 contacts could energize the EDG output breaker trip coil and is therefore included for evaluation.

Chatter of the 86-2 DG relay contacts could lead to seal-in of the SDR. Chatter of the overspeed trip relay (OTR) contacts themselves can also lead to seal-in of the SDR. The emergency stop control switch is non-vulnerable. Chatter from the relays for the low lube oil pressure and high jacket water temperature switches could lead to seal-in of the SDR.

Chaffer of the switches controlling low lube oil pressure or high jacket water temperature are not susceptible to chatter, thus they would not lead to seal-in of the SDR.

Note that during Diesel Generator emergency operation, the Safety Features Actuation starto Under-voltage start, and control room manual start, contacts block the low lube oil pressure and high jacket water temperature chatter; however, this feature is not operating before the start signal is given and thus will not prevent the seal-in of the engine trouble SDR should coincident chatter occur in these circuits prior to DG start.

EDG Svslems A number of components and systems are required to start and operate the EDGs. For identiffing electrical contact devices, only systems and components which are electrically controlled are analyzed. Information in the UFSAR was used as appropriate for this analysis.

Starting Air Based on Diesel Generator availability as an initial condition, the passive air reservoirs are presumed pressurized and the only electrically active components in this system required to operate are the air start solenoids, which are covered under the EDG engine control analysis above.

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2734296-R-075 Reaision 0 lune L2,201.7 Page 32 of46 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 use engine-driven mechanical lubrication oil pumps which do not rely on electrical control. The lube oil system also contains AC Turbo Oil pumps which circulate the lube oil. These pumps do not contain seal-in or lockout relays. These pumps would stop if there is a thermal overload condition. In addition, there are DC Turbo Oil pumps which circulate the lube oil and start on low lube oil pressure and stop after lube oil pressure has been raised. The DC oil pumps do not contain seal-in or lockout relays and will function as required in a seismic event.

Fuel Oil The Diesel Generators use engine-driven mechanical pumps and DC-powered auxiliary pumps to supply fuel oil to the engines from the day tanks. The day tanks are rc-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 and therefore there is no impact due to relay chatter.

Cooling Water This system consists ofjacket water and an aftercooler. The aftercooler is cooled by jacket water and the jacket water is cooled by Component Cooling Water (CCW). The CCW is then cooled by SW. Engine-driven pumps operating in the cooling loops 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.

Three CCW and SW pumps provide cooling water to the heat exchangers associated with the two EDGs. [n automatic mode these pumps are started via closure of the EDG Output breaker.

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2734296-R-0L5 Reaision 0 lune L2,201-7 Paxe 33 of46 If there were chatter of the SAlUl3 such that it opens and recloses itself (i.e., it does not seal-in) could result in an increase in delay to start the CCW and SW pumps. This momentary chatter does not result in an extensive delay and is therefore not evaluated in this analysis. A chatter analysis of the CCW and SW pump circuit breaker control circuits indicates the bus lockout, phase overcurrent, and ground fault relays all could prevent automatic (sequential) breaker closure following the seismic event. These breakers are ACll3 (CC\ry Pump l), ADl13 (CCW Pump 2), ACl08/ADl08 (Swing CCW Pump 3), ACl07 (SW Pump l), AD107 (SW Pump 2), and ACl09/ADl09 (Swing SW Pump 3).

Ventilation Ventilation for each Diesel Generator Enclosure is provided via two supply fans. In automatic mode these fans are started when the EDG reaches 40 RPM. This permissive does not prevent the ventilation fans from running and does not contain a SILO device. Successful ventilation also requires proper damper alignment. No SILO devices were found to prevent the dampers from opening as required.

Batterv Chargers Chatter analysis on the battery chargers was performed using information from the UFSAR as well as vendor schematic diagrams. Each battery charger has an under-voltage relay on the input side set to alarm when battery charge input voltage drops below the design capability of the batteries, and one under-voltage alarm relay on the output side. Contacts l-3 of Relay K301 can operate the trip coil of the battery charger 480VAC input breaker only if the Low AC Voltage Disconnect Switch is enabled, which is done by taking the switch to Test. On the in-service battery charger, this switch is never in TEST, so the K30l contacts cannot energize the trip coil.

The operate coil of this relay is controlled by a non-vulnerable solid-state circuit. The Master Control Board does not have any relays installed; therefore, no further review is required.

Contacts shown on the Battery Charger Schematic Drawings were determined to have no adverse effect on the capability of a battery charger to provide an output voltage. No other vulnerable contact device affects the availability of the battery chargers.

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2734296-R-0L5 Reaision 0 lune 72,2077 Paxe 34 of46 Inverters Analysis of schematics for the Static Inverters, and the Average Power Range Monitor (APRM)

Unintemrptable Power Supply (UPS) Inverters revealed no vulnerable contact devices and thus chatter analysis is unnecessary.

Switcheear, Load Centery-, and MCCs Power distribution from the EDGs to the necessary electrical loads (Battery Chargerso Inverters, Fuel Oil Pumps, and EDG Ventilation Fans) was traced to identiff any SILO devices which could lead to a circuit breaker trip and intemrption in power. This effort excluded the EDG circuit breakers and the CCW/SW pump breakers which are covered in above, as well as component-specific contactors and their control devices, which are covered in the 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 essential busses to the load centers. A chaffer analysis of the control circuits for these circuit breakers indicates the bus lockout, overcurrent, and ground fault relays all could prevent automatic (sequential) breaker closure following the seismic event. The relays listed above are included for evaluation and reported in Table B-1 inAppendix B.

2.6 Sutupr.+ny oF SELECTED ConaroFrENTs A list of the contact devices requiring a High-Frequency Confirmation is provided in Appendix B.

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2734296-R-0Ls Ratision 0 lune 1.2,2077 Pase 35 of46 3.0 SEISMIC EVALUATION 3.1 IIonrzoNTAL SusuIC I}rMAFIu Per Reference 8, Sect. 4.3, the basis for calculating high-frequency seismic demand on the subject components in the horizontal direction is the DBNPS horizontal GMRS, which was generated as part of the DBNPS ESEP report (Reference 13) submitted to the NRC on December 19, 2014 and accepted by the NRC on October 19,2015 (Reference 14).

It is noted in Refersnce I that a Foundation Input Response Spectrum (FIRS) *ay 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 8, "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.'o For sites founded on soil, the soil layers will shift the frequency range of seismic input towards the lower frequency range of the response spectrum by engineering judgment.

Therefore, for purposes of high-frequency evaluations in this report, the GMRS is an adequate substitute for the FIRS for sites founded on soil.

The DBNPS site bedrock occurs at elevation (EL) 555 ft and consists of massive and bedded dolomite layers. The deepest foundation elevation of these structures is EL 540 ft and is associated with the RB. Therefore, the GMRS, Control Point elevation is taken to be the base of the RB foundation, EL 540 ft. The bedrock immediately underlying the bottom of the RB foundation (EL 540 ft) is characterized by shear-wave velocities (Vs) of about 5,200 feet per second (ff/s).

The applicable buildings at DBNPS are founded on rock; therefore, the Control Point GMRS is representative of the input at the building foundation.

The horizontal GMRS values are provided in Table 3-2.

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2734296-R-01_5 Reaision 0 lune 72,2017 Page 36 of46 3.2 VentIc.dr, SEISMIC DEMAND As described in Section 3.2 of Reference 8, 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 shear-wave velocity vs. depth profile is provided in Reference 15, Table 5-3 and reproduced below in Toble 3-1.

TABLE 3-1 SOIL MEAI\I SHEAR.WAVE \IELOCITY VS. DEPTH PROF'ILE FOR THE FIRST lOO FT LAYER LAYER PROFILE LAYER ELEVATION LAYER END END THICKNESS Vsi [fl/sl di / Vsi EIdr/ VsS0 DEPTH I}EPTH Vsr I Ift/sl IftI dr [ft1 IftI lml 540 I 12 3.7 t2 4948 0.00243 0.00243

,, .),)

528 LL 6.t l0 3970 0.00252 0.00494 518 3 32 9.8 l0 5790 0.00173 0.00667 4548 508 4 80 24.4 48 4071 0.01179 0.01846 460 5 100 30.5 20 5672 0.00353 0.02199 Using the shear-wave velocity vs. depth profile, the velocity of a shear wave traveling from a depth of 30m (98.4 ft) to the surface of the site (Vs30) is calculated perthe methodology of Reference 8, Section 3.5.

o The time for a shear wave to travel through each soil layer is calculated by dividing the layer depth (*) by the shear-wave velocity of the layer (Vsi).

t The total time for a wave to travel from a depth of 30m to the surface is calculated by adding the travel time through each layer from depths of 0m to 30m (E[diA/sr]).

o 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., Vs30 (30m)/E[diA/s,].

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2734296-R-01s Rwision 0 lune L2,20L7 Page 37 of46 The vertical FIRS is derived using the Vertical-to-Horizontal (V/H) spectral ratio for rock sites in Western United States (WUS) and Central and Eastern United States (CEUS) from NUREG/CR-6728 (McGuire et a1.,2001) (Reference l6). The average Vs in the upper 30 meters (m) (100 ft) is used to weight the WUS and CEUS VIH values. The average Vs in the upper 30 meters (Vs30) for EL 540 ft is 4548 fl/sec (1386 meters per second [r/s]). The Vs30 for WUS and CEUS rock sites are 520 m/s and 2800 m/s, respectively (Reference 16). The V/H ratios at EL 540 ft use a weight of (2800-1386y(2800-520):0.62 for WUS V/H ratios and

( I 3 86-520y(28 00-5 20;:9.3 I for CEUS V/H ratios.

The V/H ratios from Reference 16 are also dependent on peak ground acceleration (PGA). The spectral ordinate of horizontal FIRS at 100 Hertz (Hr) is used as the PGA to determine the V/H ratios. For EL 540 ft, the 100-Hz SA for the horizontal FIRS is 0.19939. Because this value is at the boundary between two PGA-level bins (i.e., < 0.2g and 0.2 - 0.5g) corresponding to different V/t{ spectral ratios in Reference 16, the arithmetic average of the V/H spectral ratio for the two bins is 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 V/FI ratios and VGMRS values are provided in Table 3-2 of this report.

Figure J-I below provides a plot of the horizontal GMRS, V/t{ ratios, and vertical GMRS for DBNPS.

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2734296-R-015 Reaision 0 lune L2,201.7 Pnxe 38 of46 TABLE 3-2 HORIZONTAL AND YERTICAL GROUND MOTIONS RESPONSE SPECTRA Frequency HGMRS VGMRS (Hz) (s) V/H Ratio (g) 0.10 0.0032 0.5938 0.0019 0.13 0.0047 0.5957 0.0028 0.16 0.0067 0.5970 0.0040 0.20 0.0097 0.5979 0.0058 0.26 0.0140 0.6000 0.0084 0.33 0.0205 0.6000 0.0123 0.42 0.0307 0.5863 0.0180 0.50 0.0422 0.571I 0.0241 0.53 0.0443 0.5688 fr.42s2 0.67 0.0s45 0.5651 0.0308 0.85 0.0665 0.5564 0.0370 1.00 0.0719 0.5494 0.039s 1.08 0.0766 0.5470 0.0419 1.37 0.0840 0.s429 0.0456 1.74 0.0826 0.s387 0.0445 2.21 0.0880 0.s398 0.0475 2.50 0.09s3 0.s467 0.0521 2.81 0.1l r4 0.5557 0.0619 3.56 0. l 666 0.5774 0.0962 4.52 0.2470 0.6134 0.1515 5.00 0.2951 0.6306 0.1861 s.7 4 0.3?44 0.6s84 0.2465 7.28 0.466s 0.7226 0.3371 9.24 0.5305 0.8026 0.4258 10.00 0.5444 0.8297 0.4517 tt.72 0.5362 0.8790 0.4713 t4.87 0.4896 0.9079 0.4445 18.87 0.4708 0.9150 0.4308 23.95 0.4239 0.8846 0.37s0 25.00 0.4110 0.8779 0.3608 30.39 0.3913 0.851s 0.3332 38.s7 0.3877 0.8463 0.3281 48.94 0.3572 0.8583 0.3066 62.10 0.2878 0.8687 0.2500 78.80 0.2200 0.8591 0. l 890 100.00 0.1993 0.8194 0. I 633 lE$Gon*ulting riRtzzo

2734296-R-015 Reuision 0 lune 12,201.7 Pase 39 of 46 0.6 1 HGMRS x-rV6MRS

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,a 0.1 0 0.4 0.1 I Frequenry(Hz) 10 1m FIGURE 3.1 PLOT OF TIIE HORIZONTAL AI\ID VERTICAL GROUND MOTIONS RESPONSE SPECTRA AI\ID V/H RATIOS 3.3 CourpoNENT HoRTzoNTAL SBrsruc DEMAr\D The horizontal seismic demand to be used in this evaluation are the in-suructure response spectra at the base of the equipment, amplified by amplification factors suggested in Reference 8 for the specific type of equipment. The required 5% in-structure response spectra ISRS are obtained from Reference 17 which is developed as part of the Seismic Probabilistic Risk Assessment (SPRA) program at DBNPS Unit l. 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 l8).

Per Reference 8, 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, motor control center, etc.)

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2734296-R-015 Ranision 0 lune L2,2017 Page 40 of45 The in-cabinet amplification factor, AF* is associated with a given type of cabinet construction.

The three general cahinet types are identified in Reference I and Appendix I of EPRI NP-7148 (Reference I l) assuming SYo in-cabinet response spectrum damping. EPRI NP-7148 (Reference 11) 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 conffol room benchboard panels and low amplification structures such as motor control centers.

All of the electrical cabinets containing the components subject to High-Frequency Confirmation (see Table B-1 inAppendix B) can be categorized into one of the in-cabinet amplification categories in Reference I as follows:

t MCCs Fl2A, Fl lA, and El lB are typical motor control center cabinets consisting of a lineup of several interconnected sections. Each section is a relatively naffow cabinet structure with height-to-depth ratios of about 4.5 that allow the cabinet framing to be efficiently used in flexure for the dynamic response loading, primarily in the front-to-back direction. This results in higher frame stresses and hence more damping which lowers the cabinet response. In addition, the subject components are not located on large unstiffened panels that could exhibit high local amplifications. These cabinets qualiff as low amplification cabinets.

a Switchgear cabinets Cl and Dl 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 0.87 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 Relay panels C3615, C3616, C3617 and C3618 are in lineups of three interconnected sections with moderate width. Each section consists of structures with height-to-depth ratios of about 1.8 which results in moderate frame stresses and damping. Relay panels C3621 and C3622 are slender panels mounted on the EDGs and braced slightly above their mid-heights.

Relay panel RC4607 is a small and lightrneight panel attached to a stiffened floor-mounted rack. The response levels are mid-range between motor control centers and switchgear and; therefore, these panels can be considered in the medium amplification category.

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2734296-R-015 Rerision 0 lune 1.2,2017 Paxe 41 of46 3.4 Conrponnhrr VERTICAL Sotsprrc DBpr,+.un The component vertical demand is determined using the peak acceleration of the 5% damped vertical ISRS from Reference 17 between 15 Hz and 40 Hz and amplifying it using the vertical in-cabinet amplification factor AF" to account for seismic amplification within the host equipment; e.g., switchgear, motor control center, or relay panel. The in-cabinet amplification factor, AF", is derived in Reference I and is 4.7 for all cabinet types. 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 l8).

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2734296-R-075 Reaision 0 lune 1.2,2017 Pa*e 42 of iI6 4.0 CONTACT DEVICE EV^A.LUATIONS Per Reference 8, seismic eapacities (the highest seismic test level reached by thc contact device without chatter or other malfunction) for each subject contact device are determined by the following procedures:

I If a contact device was tested as part of the EPRI High-Frequency Testing program (Reference 7), then the component seismic capacity from this program is used.

2. If a contact device was not tested as part of Reference 7, then one or more of the following means to determine the component capacity were used:

a) Device-specific seismic test reports (either from the station or from the SQURTS testing program.

b) Generic Equipment Ruggedness Spectra (GERS) capacities per Reference 9 and Reference 10.

c) Assembly (e.9., electrical cabinet) tests where the component functional performance was monitored.

The high-frequency capacity of each device was evaluated with the component mounting point demand from Section J.0 using the criteria in Section 4.5 of Reference 8. A total of 80 components are identified that required High-Frequency Confirmation evaluation. The 80 components are grouped into 16 main groups based on device type and capacity and enc losure dynamic characteristics and location.

A summary of the high-frequency evaluation conclusions is provided in Table.B-I in Appendix B.

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2734296-R-015 Ranision 0 lune 1.2,2017 Pa*e 43 of46

5.0 CONCLUSION

S 5.I GrNrnaL CoNCLUSIoNS DBNPS has performed a High-Frequency Confirmation evaluation in responsc to the NRC's 50.54(0 letter (Reference l) using the methods in EPRI Report 3002004396 (Reference 8).

The evaluation identified a total of 80 components that required High-Frequency Confirmation evaluation. The 80 components identified are grouped into 16 main groups based on device type and capacity and enclosure dynamic characteristics and location. The high-frequency evaluation is performed for the 16 main groups and the results are summarized in Table B-1 inAppendix B.

As showninAppendix B, Table B-1,11 components out of a total of 80 components have capacity less than demand when evaluated in accordance with Reference I guidance. These components were then reevaluated through mitigation strategies included in Appendix H of Reference 5 and shownto be adequate (i.e., HCLPFcrooz) PGAcrrans) and do not impactthe credited path for mitigation strategies.

5.2 Inu,xrrucATroN oF FoLI,tlw-Up Acrrous For DBNPS, all the identified 80 components have adequate seismic capacity and no follow-up actions were identified.

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2734296-R-01.5 Ranision 0 lune 1.2,201,7 Paxe 44 of46 6.0 REF'ERENCES I NRC (E. Leeds and M. Johnson) Leffer to All Power Reactor Licensees et al., "Request for lnformation Pursuant to Title l0 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," March 12,2fr12, ADAMS Accession Number MLI20534340.

.l L NRC (W. Dean) Letter to the Power Reactor Licensees on the Enclosed List. "Final Determination of Licensee Seismic Prohabilistic Risk Assessments Under the Request for Information Pursuant to Title l0 of the Code of Federal Regulations 50.5a(f) Regarding Recommendation 2.1 'Seismic' of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident," October2l,2015, ADAMS Accession Number MLI51944015.

J NRC (J. Davis) Letter to Nuclear Energy Institute (A. Mauer), "Endorsement of Eleckic Power Research Institute Final Draft Report 3002004396,'High Frequency Program:

Application Guidance for Functional Confirmation and Fragility," September 17,2015, ADAMS Accession Number ML I 521 8A569.

4 FirstEnergy Nuclear Operating Company (FENOC) seismic Haeard and Screening Report (CEUS Sites), 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: Enclosure C-NTTF 2.1 Seismic Hazard and Screening Report for Davis-Besse Nuclear Power Station dated March 31, 2014, ADAMS Accession Number ML14092A203.

5 NEI 12-06, Revision 2, Diverse and Flexible Coping Strategies (FLEX) Implementation Guide, December 2015, ADAMS Accession Number ML16005A625.

6 EPRI 1025287, "Seismic Evaluation Guidance: Screening, Prioritization and Implementation Details (SPID) for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1: Seismic," February 2013.

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2734296-R-01.5 Reaision 0 lune L2,AAfi Pase 45 of46 7 EPRI 3002002997, "High Frequency Program: High Frequency Testing Summary.'n September 2014.

I EPRI 3002004396, "High Frequency Program: Application Guidance for Functional Confirmation and Fragility Evaluation," July 2015.

9. EPRI NP-7147-SL, "Seismic Ruggedness of Relays," August 1991.

10. EPRI NP-7147 SQUG Advisory 2004-02, "Relay GERS Corrections," September 10, 2004.

ll. EPRI NP-7148-SL, "Procedure for Evaluating Nuclear Power Plant Relay Seismic Functionality," December I 990.

12. Not used.

13. FirstEnergy Nuclear Operating Company (FENOC) Expedited Seismic Evaluation Process (ESEP) Reports, Response to NRC Request for lnformation Pursuant to 10 CFR 50.54(0 Regarding Recommendation 2.1 of the Near-Term Task Force (NTTF)

Review of lnsights from the Fukushima Dai-ichi Accident: Enclosure C-Expedited Seismic Evaluation Process (ESEP) Report for Davis-Besse Nuclear Power Station dated December 19, 2014, ADAMS Accession Number ML14353A059,

14. NRC Letter, Davis-Besse Nuclear Power Station, Unit I - Staff Review of Interim Evaluation Associated with Reevaluated Seismic Hazard Implementing Near-Term Task Force Recommendation 2.1, dated October 19, 2015, ADAMS Accession Number MLI5273A237.

15. ABS ConsultinglRlZZ0 Associates, "Probabilistic Seismic Hazard Analysis and Ground Motion Response Spectra, Davis-Besse Nuclear Power Station, Seismic PRA Project,"

2734296-R-003, Revision l, 2014.

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2734296-R-01.5 Ratision 0 June L2,20L7 Paxe 45 of46

16. McGuireo R. K, Silva, W. J., and Costantino, C.J.,200l, "Technical Basis forRevision of Regulatory Guidance on Design Ground Motions: Hazard- and Risk-Consistent Ground Motion Spectra Guidelines," NUREG/CR-6728, U.S. Nuclear Regulatory Commission, October 2001.

t7. ABS ConsultingtLlZZ0 Associates, "Building Seismic Analysis of Davis-Besse Nuclear Power Station: Seismic PRA Project," Revision l, ABS Consulting Report 27 34296-R-005, 201 4.

18. EPRI NP-6041-SL, "A Methodology for Assessment of Nuclear Power Plant Seismic Margin," Revision l, Electric Power Research Institute, June 1994.

19. 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.

20 NRC Letter, Davis-Besse Nuclear Power Station, Unit I - Staff Assessment of lnformation Provided Pursuantto Title l0 of the Code of Federal Regulations Part 50, Section 50.54(0, Seismic Hazard Reevaluations for Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident, dated August 25,2015, ADAMS Accession Number MLI5230A289.

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2734296-R-41,5 Ranision 0 lune 1.2,zilfi Paxe Al of A22 APPENI}IXA REPRESENTATIVE SAMPLE COMPONENT EVALUATIONS lESConsulting

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Report 2734296-R-015, Appendix A, 50.54(f) NTTF 2.1 Seismic Revision 0 High Frequency Confirmation Example A Representative Sample Component Evaluations A{. Purpose The purpose of this calculation is to show two examples of High Frequency Confirmation evaluation for sensitive components that rcquired evaluation at Davis Besse Nuclear Power Station. This calculation is in support of plant response to NRG Near-Term Task Force recommendation 2.1 for performing high frequency confirmation.

A2.0 Scope The complete list of the components selected for High Frequency confirmation evaluation are listed in Table &1 The two components selected for sample calculation are presented in Table A-l. These example calculations show the detailed prccedurc for the High Frequency confirmation evaluation.

Table A-1: Components selected for sample High Frcquency Gonfirmation Evaluation F11A Aux. 7 603 2 De-Energized ElectroMechanical Cutler Contactor E11B Aux. 8 585 1 D+Energized pc01K1C Hammer F12A Aux 6 603 2 De-Energized HFA51 A42H G.E. c3615, C3616 Aux. 6 585 3 D+Energized (NO)

A3.0 Methodology The methodology in Reference A1 will be used to calculate Capacity to Pemand ratios for the subject relays, and contactors in the High Frequency range of 1540 Hz. The capacity is obtained from either EPRI HF Test program (Ref. A2), or from GERS (Refs. A8, A9 & A10), or other shake table tests (Refs. A5, A6, A7) if the EPRI Program did not include the specific relay model. The EPRI HF Test Program reports a representative average spectral acceleration (SA) in the high frequency range. \Mrile the capacities in References A5 thru A10 are for the Low Frequency region (i.e., 4.5-16 Hz), according to the conclusions in References A1 and A2, the Low Frequency capacities are always lower than the High Frequency capacities and therefore could be used conservatively in the HF confirmation program.

The seismic Demand to be used in this evaluation are the in-structure response spectra at the base of the equipment, amplified by amplification factors suggested in Reference A1 for the specific type of equipment.

Reference 43 provides the required 5% ISRS, wtrich were developed as part of the Seismic PRA program at Davis Besse Nuclear Povver Station Unit 1. !f there are sharp peak(s) in the ISRS in the frequency range of interest (15 jlz lo 40 Hz), these peaks are clipped in accordance with the guidelines in Reference A4.

Wrile not required for HF confirmation task, the C10% capacities are calculated and also reported here for each relay or contactor using guidance in Reference A11.

Page A2 of A22

Report 2734296-R-015, Appendix A, 50.54(fl NTTF 2.1 Seismic Revision 0 High Frequency Confirmation Example A4.0 References

41. EPRI Technical Report No. 3002004396, "High Frequency Program - Application Guidance for Functional Confirmation and Fragility Evaluation," Final Report, July 2015.

42. EPRI Technical Repoft No. 3002002997, "High Frequency Program - High Frequency Testing Summary," Final Report, September 2Q14, A3. ABS Consulting/Rizzo Associates, "Building Seismic Analysis of Davis Besse Nuclear Power Station: Seismic PRA Project," Rev.1. ABS Report 2734296-R-005 (Rizzo Report R7-124737).

44. EPRI NP-6041-SL, "A Methodology forAssessment of Nuclear Power Plant Seismic Margin," Rev.1, Electric Power Research lnstitute, June 1994.

A5. Anco Test Repoft 1633.03, "Relay and Switch Tests," Rev.1.0, September 1996.

A6. FirstEnergy Calculation No. C-CSS-100.00-155, "Seismic Evaluation for Revised ABB Relay Seismic Testing ZPA Values," Rev.0, 6/13/2003.

A7. FirstEnergy Calculation No. C-CSS-O04.01-014, "Seismic Evaluation for 4.16 kV Switchgear 50/51 Phase A & Phase C Relay Replacement with ABB COM-S Model," Rev.2, Dated 512612011.

48. EPRI TR-105988-V1, 'GERS Formulated Using Data from the SQURTS Program," April 1996, and SQUG Advisory 2004-02, "Relay GERS Corrections," September 7 ,2004.

49. EPRI NP-7147-SL, "Seismic Ruggedness of Relays," August 1991.

A10. EPRI NP-7147-SL, Vol.2, Addendum 2, "Seismic Ruggedness of Relays,"April 1995.

All. NEI 12-06, Appendix H, December 1, 2015.

Page A3 of 422

Report 2734296-R-015, Appendix A, 50.54(0 NTTF 2.1 Seismic Revision 0 Hish Frequency Confirmation Example A5.0 High Frequency Gonftmation Evaluations The HF confirmation of the relays and contactors identified in Secfon M.0 above is performed in the following Sections. These evaluations use the methodology cited in Reference Al, as described in Sec0'on A3.0 above.

A5.1 HF Evaluation for Gutler Hammer Gontactor A201Kl C F11A Aux. 7 603 2 1. MCC Cutler E11B Aux. 8 585 3 1. MCC F12A Aux. 6 603 2 1. MCC A5.{.1 Gapacity SA := 19.09 HF seismic capacity of Cutler Hammer Electro Mechanical ContactorA20lKlC in De-Energized state from Reference 42, Table 56 SAf := SA + 0.6259 = 19.63.9 Effective spectral test capacity per Ref. A1 Page A4 of 422

Report 2734296-R-015, Appendix A, 50,54(fl NTTF 2.1 Seismic Revision 0 High Frequencv Confi rmation Examole 45.1.2 Demand The 5% damped in-structure response spectra at the Iocations of the MCCs that house the contactors (located at Point 2 inAux. 6, El. 603', Point 2 in Aux. 7, EI. 603'& Point 1 inAux. 8, El. 585') in both horizontal and vertical directions are shown below and the HF demand in the 1SHz to 40 Hz.

SAH_Au*6_60g_2 := max (0.24g, 0.329) Maximum horizontal acceleration (X or Y direction) in the 15Hz to 40Hz range SAH_Aux6_603_2 : 0.32. g corresponding to Pt. 2, AUX6, E1.603' (no clipping required) sAv-Ar*G-GoB-2 :: 1'039 Maximum vertical acceleration in Z-direction in the 15Hz to 40Hz range coresponding to Pt.z, AUX6, E1.603'(no clipping required)

SAn-lr*7_60g 2 := max(0.6ag, 1.379) Maximum horizontal acceleration (X or Y direction) in the 15Hz to 40Hz range corresponding to Pt.2, AUX7, E1.603' SAH Ar*7 60g Z= 1.37.g (see clippings on the following sections)

SAV Ar*7 60g 2:= 0.929 Maximum veftica! acceleration in Z-direction in the 1SHz to 40Hz range conesponding to Pt.z, AUX7, E1.603' (see clipping on the following sections)

$AH Ar*g SgS 3 r: lrlox(0.789,0.739) Maximum horizontal acceleration (X or Y direction) in the 15Hz to 40Hz range colresponding to Pt.3, AUX8, E1.585' SAH Ar*g SBE g = 0.78.9 (see clippings on the following sections)

SAV Auxg SgS *:= 0.619 Maximum veftical acceleration in Z-direction in the 1SHz to 40Hz range conesponding to Pt.3, AUX8, El.585' (see clipping on the following sections)

Page A5 of A22

t Appendix A, Revision 0 a ,SRS at Auxiliary Bldg. 6, Elevation 603', Point 2:

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Report 2734296-R-015, Appendix A, 50.34(fl NTTF 2.1 Seismic Revision 0 Hioh Frequencv Confirmation Examole Aux. 6 Building, El. 603' Vertical Point 2 1.8 1.6 I

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0.1 1 10 100 Frequency (Hzl Page A7 of M2

Reprt 2734;Zs-R-015, Appendix A, 50.3dffI NTTF 2.1 Sclernic Revision 0 Hlqh Fle,ouencv Confirmatbn Examgh

,SRS at Auxiliary Bldg. 7, Elevation 403', Point 2:

Aux hrlldlryT,E).68' Gbbrl!( nrpomc rpGctrl, Polnt2 1.4 12 1

-516 I

I o"8 I

I I \

t A

UI o.6

\

J t o.4

\

02 I \r E

.Jdal J

o o.1 1 10 lm Frcquanclllhl Qlio BRS x at fc=11.5F12 f, := 1 1.51'lz Sa3Ceak :* 1.169 S"J"ak.80o/o = 0.93.g fi_l := 10.25j12 t2_1 ,* 12.5j1z:

Af0.8:=tZ 1-f1 I =2.25'l4z Afo.a B:=-=0.20 f,

cg:= 0.55 if B<0.2 cg = o'55 0.4 + 0.75.8 lf 0.2 sB s 0.8 1.0 if B>0.8 Sa-clip-x-Aux7-60 g-2-5% := C6' SaJ*ak = 0'64' g Clipped horizontal X acceleration at ArD{. 7, EL.603', Point 2, 5o/o damping in the frquency range of I5j1tr to 40Flz PageAB of M2

Report Z7ffi2ffit-015, ApprdixA, 50.84{fl NTTF 2.1, Sebmb Revision 0 Hklh Fraomnor Conlinrdion Examole Aux BuildirU 7rEl.5O3' Global Y rceponsc specta, Point 2 2

rt I

I 1A I

1,6 1,4 -5r 7,.7 A

Y I

1 I \

fi I

\

oa

\

o.6

\

n 0.4 r A I o.2 rFaJ/

o o.1 1 10 1m Frcrycnca{Hz}

Clio BBS v at fc=11.5F12 f, := 1 1.5H2 Sa3reek l= 1.899 S"-F"gg'80?6 = 1.51'g t1_1r= 10.25t12 tZ-1 := 15.25112 Afg.g i= tZ_1 - t1_1= 5.00.H2 Afo.B B:=-=0.43 t

c6:= 0.55 if B<0.2 0.4 + 0.75.8 if 0.2 < B s 0.8 cc = o'73 1.0 if B>0.8 Sa-clip3-Aux7-60 g-2-5o/o := Cc' Sa-p"ak = 1'37' g Clipped horizontal Y acceleratlon at Aux. 7, EL.603', Point 2, 5.r{D damping in tte frequency range of 151-lz to 40FE Page A9 of PC2

Report 2734.2S-R{15, Appendix A, 50.t{(0 NTTF 2.1 Seisrnic Revisbn 0 Hhh Frcqrcncy $onfifmation ExamoF Aux Building 7, El.6O3' Global Z rcsponsc spectra, Point 2 L.2 I

1 r5r

\"

o.t I, t f o'u

\

o.4

\ \r I

\

o:

/

o o.l 1 10 100 Fruquanca0El Clip RBS z at fc=10.5Ft fa := 10.51'lz SalCeak := 1.039 S"-1*"g.80% = 0.82.g t1_1:= 8.5H2 t2-1 := 15.4H2 AfO.g := t2_1- f1_1 = 6.90.H2 Afo.B B:= =0.66 fc c6:= 0.55 if B<0.2 0.4 + 0.75.8 if 0.2s8 s 0.8 cc = o'89 1.0 if B>0.8 Sa-clip-z-Aux7-603-2-5 % := Cg' Sa-paak = 0'92' g Clippe.d vertical Z acceleration at Aux 7, EL.603', Point 2, ioh damping in the frequency range of 15Flz to 40Ftr PageAlO of M2

Report 27U296-R{15, Appendix A, t0.54(fl NTTF ?.1 Sqigmh Revision 0 Hlqh Frequeqcy Confirmf,thn Examolg a ,SRS at Auxiliary Bldg. 8, Elevation 585', Point 3:

Arrx. I tulldlry, El.S8!i' Glob.l X raponrc +Gctrr, Pdnt 3 1.4 I I

I L.2 I

1

-5*)

\

^0.8 UI t \

\I o.u 0.4

\

/

0.2 /

o 0.1 1 to 1m FruqucncylHz)

Clio RRS x at fc=13.5F12 fa :* 13.5H2 Sa3""k := 1.2059 SaJ"sj.80% = 0.96.g fl_l := 12.6jfr. t2_11= 17Hz AfO.8 ,= t1_l- f1_1 = 4.40.J12 Afo.s B:=-=0.33 fc c6:= 0.55 if B<0.2 0.4 + 0.75.8 if 0.2sB s 0.8 cc = 0'64 1.0 if B>0.8 Sa-clip-x-Aux8-585-3-5 o/o := C6'S"-p"ak = 0'78'9 Clipped horizontal X accelenation at Aux. 8, EL.585', Point 3, 5% damping in the frequency lltnge of lsFlz to 401'lz PageAl 1 otAP;2

Rcport 27Y296+t-015, Appendix A, Rcvisbn 0 fuir. t Bulldh6 El.585' Gbb.lY Pdnt3 oa o.7 I

\

It I

0.6

-5I 0.5

\

.^ \

-t 0.4 I

0.3 I t I \

o2 J

/

0.1 I a, o

o.1 1 10 1m FruquancryHzl Clip RFSJ at fc=13.5Ftr fs := 13.5llz Sa;eeek:= 0.7349 Sa_p"gj.80% = 0.59.9 f1 := 1 1.31'tr t2:= 22jfr.

AfO.8 := lZ- f1 = 10.70'HZ Afo.8 B:=-=0.79 fc cg:= 0.55 if Bs0.2 0.4 + 0.75.8 if 0.2 sB < 0.8 cc = o'99 1.0 if B>0.8 Sa-clip3-Aux8-585-3-5 oh := C6' S"--p"ak = 0'73' g Clipped horizontd Y accderation at Aux. 8, EL.585', Point 3, 5% dampirp in the frequency range of 151'tr to 40FE Page A12ot M2

Report 27U29&R-015, Appendix A, 5_0.54(0. NTTF 2.1 $eismlc Revision 0 Hhh FT?guencv Confirmation Erclmb An B Bulldlry, E!.585' VGrdGrl upolre rpGctn, Polnt3 o.7 LJ 0.6 I t I

o.5

-5t3 v o.{

^rUtl

o.,

0.2

/

o.1 /

o o.1 1 10 lm Frs$GncU[Hr]

Qlip RRS 4at fc=12.51'tr

f. := 12.5l'12 Salceakt= 0.6079 Sajc"ek.8O% = 0.49.9 f1 := 9.3H2 t2:o 19.51'lz Af0.8 := tZ- fl = 10.20 'jlz Afo.B B:=-=0.82 fc c6:= 0.55 if B<0.2 0.4 + 0.75.8 if 0.2 sB < 0.8 cc = 1'oo 1.0 if B>0.8 Sa-clip-z AuxS-585-3-5 o/o := cc' S"--p"ak = 0'6 1' g Clipped vertical Z *celeration at Aux 8, EL.585', Pdnt 3, 50/6 damping in the frequency range of 15Flz to 40Flz Page A13 of FAz

Report 2734296-R-015, Appendix A, 50.54(fl NTTF 2.1 Seismic Revision 0 Hish Frequencv Confi rmation Example AFC tt:= 3.6 Maximum horizontal in+abinet amplification factor for motor control centers per Ref. A1 AFC V := 4.7 Maximum vertical in+abinet amplification factor for motor control centers per Ref. 41 ICRSH Aux6 603:= SAH Maximum Horizontal in-cabinet response 4r*6 603 2'AFC H = 1.15'g spectra at Point 2, AUX6, E1.603' I c Rsv_Aux6_608 := sAy_4ux6_608_2' nFc-y : 4' 84' g Maximum vertical in-cabinet response spectra at Point 2, AUX6, E1.603' ICRSH AuxT_603 := SAg AuxT_603 2'AFC_H: 4.93'9 Maximum Horizontal in+abinet response spectra at Point 2, AUX7, EI.603' ICRSV_Aux7_603 := SAy_4ux7_603 2'AFC_V = 4,32'9 Maximum vertical in+abinet response spectra at Point 2, AUX7, E1.603' ICRSH AuxB_SBS := SAg AuxB_SBS_g'AFC_U = 2.81'g Maximum Horizontal in+abinet response spectra at Point 3, AUX8, E1.585' ICRSV_Auxg_5BS := SAy-4uxB_SB5_g'AFC_V = 2.87' g Maximum vertical in-cabinet response spectra at Point 3, AUX8, EI.585' Page A14 of A22

Report 2734296-R-015, Appendix A, 50.54(fl NTTF 2.1 Seismic Revision 0 High Frequencv Confi rmation Example A5.1.3 Gapacity-D,emand Ratio Fp := 1.56 CDFM Knockdown factor for fragility threshold from high frequency test program (Table 4-2 of Ref. 41)

Fplg := 1.20 Multi-axis to single-axis conection factor from Section 4.5.2 of Ref. A1

( sR-\

TRS := ("*) :151s effective wide-band component capacity acceleration

[n]

TRS CDRH Aux6_608 '= Capacity-Demand-Ratio in horizontal direction at Aux6,

' lcRsH_Aux6;603 E1.603', Point 2 CDRH Aux6 60g = 13.10 > 1.0 TRS cDRY-4rxE-603' r= Capacity-Demand-Ratio in vertical direction at Aux6,

-lcRsv Aux6 603 E1.603', Point 2 CDRV Aux6 693 = 3.12 > 1.0 CDRH AuxT-GO3'=

- TRS Capacity-Demand-Ratio in horizontal direction at Aux7, lfu E1.603', Points 2 CDRH AuxT 563 = 3.06 > 1.0 TRS CDRV_A,x7_608 i= Capacity-Demand-Ratio in veftical direction at Aux7,

' ICRSV AuxT 603 El.603', Point 2 CDRV AuxT 693 : 3.49 > 1.0 TRS CDRH AuxB-EBS'= Capacity-Demand-Ratio in horizontal direction at Aux8,

-I E1.585', Point 1 CDRH Auxg 565 = 5.38 > 1.0 TRS cDRV-Auxg-SBf := Capacity-Demand-Ratio in vertical direction at Aux8,

' ICRSV Auxg EgS EI.585', Point 1 CDRV Auxg 5g5 = 5.27 > 1,0

- Page 415 of A22

Report 2734296-R-015, Appendix A, 50.54(fl NTTF 2.1 Seismic Revision 0 High Frequencv Confirmation Example A5.{.4 HCLPF Gapacities (Gr"re and G1s"/"}

The PGA used in developing the Davis Besse in-structure response spectra is 0.209.

PGA := 0.209 F" is the composite uncertainty for relays taken from Table H.1 of Reference A11. This composite uncertainty is considered to be Realistic Lower Bound Case according to Table H.1 of Reference A11, and it is suggested for use in calculating the median capacity.

0r:: 0.30 I Contactors in MCC FlzA HCLPFAzo 1 K-Aux6-60g-c 1 o7o := min (CDRH Aux6-603, c DRv-Aux6-60 g)' PGA HCLPFAZg1K Aux6 609 C1% = 0.62.(

t)

Am-A201 K-Aux6-603-C1 o7o := HCLPFnZO, K-Aux6-603-C1 * *("'

1 3 C1 olo = 1.25' Ratiogl0% Clyo:: 1.36 Ratio of C,, oml0t* from Table H.1 of Ref. A11 HCLPFA2g1K Aux6 603 C10o/o r= Ratio610% C1%.HCLPFA2g1K Aux6 603 C1%

HCLPFA261K Aux6 60g C10% = 0.85.9 a Contactors in MCC FlIA HCLPFAzo1 K-Aux7-603-c 1 o7o := min (ODRH AuxT-603 , GDRv-Aux7-603) ' PGA HCLPFA2g1K AuxT 608 C1% = 0.61.

(e ss or)

Am A201K AuxT 603 c17o := HGLPFRzotK AuxT 603 c1%'e 1 60g C1 olo = 1.23' Ratio610%_C1o7o:= 1.36 Ratio of Cro.a/Cr* from Table H.1 of Ref. A11 Hc LP FA20 1 KJuxT-60 3-c 1 0o/o r= Rati og 1 0 %-c 1 %' H c LP F Azo 1K-Aux7-6 og-c 1 %

HCLPFA2g1K AuxT 603 C1 Ao/o= 0.83'g Page 416 of A22

Report 2734296-R-015, Appendix A, 50.54(fl NTTF 2.1 Seismic Revision 0 Hiqh Freouency Confirmation Examole a Contactors rn MCC FllB H c LPFA2o 1 K-AUXB-sBs-c 1 o7o := m in (c D RH AuxB-E BS, c DRv-AuxB-sBs)' PGA HCLPFAZ61K AUX6 SgS C1o/o = 1.05.(

(z.aa.o.)

Am-A20 1 K-AUXB-sgs-c 1 o7o := HCLPFRzo t K.,,RUXB-5B5-C 1 %' e A2O1K AUXS C1o/o

2.12 Ratiogl0% Clyo :: 1.36 Ratio of C,, oml0r* from Table H.1 of Ref. A11 HCLPFA2g1K AUXB SBS C1O%:= Ratiog rc% C1%.HCLPFA2g1K AUXB SBE G1%

HCLPFA2g1K AUXS 585 C10% : 1.43.!

Page A17 of 422

Report 2734296-R-015, Appendix A, 50.54(0 NTTF 2.1 Seismic Revision 0 High Frequency Confirmation Example A5.2 HF Evaluation for General Electric Relay HFA51 A42H c3615, 20. Dist HFAsLA42H GE Aux. 5 58s 3 c3516 Panel A5.2.1 Capacity TRS, :- 10.09 TRS, := 10.09 GERS Capacity from Ref. A10 for Non-Operate, Normally Open state TRS= :- 10.09 1

3 sA :- (rns*.TRSy.TRSz) = 1o.o.g HF seismic capacity of GE HFA51 A42H relays in De-Energized/NO state SAf :: SA = 10.0.9 Effective spectral test capacity per Ref. A1 45.2.2 Demand The 5% damped in-structure response spectra at the locations of the relay panels (located at Points 3 in Aux.

6, El. 585) in both horizontal and vertical dircctions are shown belonr. The HF demand in the 15Hz to 40 Hz are:

SAH_Aux6_5gS_3 r= lrl?X(0.419, 0.33g) - 0.41'g Maximum horizontal acceleration (X or Y direction) in the 1 5Hz to 40Hz range coresponding to Pt.3, AUX6, E1.585' (no clipping required) sAv-Rux6-58s-3 := 1'149 Maximum vertical acceleration (Z direction) in the 15Hz to 40Hz range coresponding to Pt.3, AUX6, E1.585' (see below for clipping of vertica! spectra)

Page A18 of A22

Appendix A, 50.34(fl NTTF 2.1 Seismic Revision 0 High Freouengv,Confirmatbn Example a ,SRS at Auxiliary BIdg. 6, Elevation 585', Point 3:

Aur. 6 BulHlry, El. 585' Polnt 3 o.9 o.8 o.7 I

J o.5 -5t6r il I

\

E o.5 I

f; o.+ \

o3 I

\

o.2 o.1

/

7 J

o 0.1 I 10 tm Frcqucnc?lHzl Arfr. 6 BulHlrg, El. 585' Gl$el Y rerponse spectr+ Folnt 3 oa o.7 o.6

-5r 1 o.5 a

t0 Y O.4

,^

fi I o3 \^t

\

o2 )

J - \ \

o.1 A7 o

0.1 1 10 100 Fraqucncy lHzl Page A19 of M2

Report 27U296+t-015, Appendix A, Revision 0 ArDr. 6 BdHkr& B. 585' thr$el respursG rpectrr, Folnt 3 2

1a I 1

1.6 II 1.4 -5X) I T2 A

I frIoa Y.

I

\

o.6 o.4

\ \

o.2

/

o o.t 1 10 1m Fraqucncy[Hz]

Qio.BFS z at fc=1,3.51-lz f, := 13.51',lz Sa-p"ak t= 1.879 S"-1c"Eg.80% = 1.50.g f1 := 11.26112 t2:= 15.01'lZ Afg.B := tZ- fl = 3.74'Hz Afo.8 B := = o'28 tt c6:= 0.55 lf B<0.2 =0.61 0.4 + 0.75.8 lt 0.2 sB s 0.8 1.0 if B>0.9 Sa_clip:= CC.S"Jr"ak = 1.14.9 Clipped vertical Z accelenation SaJ"ak-z-AUx6J85-3-5 i= Sa-clip = 1' 14'g at Aux 6, EL.585', Point 3, 5016 damping in the frequency range of lsFlz to 40Fts Page M0 of A22

Report 2734296-R-015, Appendix A, 50.54(fl NTTF 2.1 Seismic Revision 0 High Frequencv Confi rmation Examole AFC ;1 := 4.5 Maximum horizontal in+abinet amplification factor for Control Cabinets per Ref. 41 AFC V := 4.7 Maximum vertical in+abinet amplification factor for Control Cabinets per Ref. A1 ICRSH Aux6_SBE_3:= SAg Aux6_EgE_e.AFC_H = 1,85.9 Maximum Horizontal in+abinet response spectra at Point 3, AUX6, E1.585' I CRSV_Aux6_SBE_3 := SAy_4ux6_SBS_S' AFC_V = 5.36' g Maximum vertical in-cabinet response spectra at Point 3, AUX6, EI.585' A5.2.3 Gapacity-Demand Ratio F;o := 1.50 CDFM Knockdown factor for GERS qualification test from Table 4-2 ot Ref. Al Fplg := 1.20 Multi-axis to single-axis conection factor from Section 4.5.2 of Ref. A1

/ sR-\

rRsD:[*,J(rrr) :Boos effective wide-band component capacity acceleration TRSD CDRH AuxB-sBE:=- IUlaDH Capacity-Demand-Ratio in horizontal direction at Aux6, AuxG 585 3 El.585', Point 3 CDRH Aux6 56g = 4.34 > 1.0 TRSp cDRv_Aux6_Egs l= CapacitpDemand-Ratio in vertical direction at Aux6,

' ICRSV Aux6 SgS g EI.585', Point 3 CDRV Aux6 SgS = 1,49 > 1.0 Page 421 of A22

5 Appendix A, 50.54(0 NTTF 2.1 Seismic Revision 0 Hish Frequency Confirmation Fxample A5.2.4 HCLPF Gapacities (Grg6 and C1sy")

The PGA used in developing the Davis Besse in-structure response spectra is 0,209.

PGA := 0.209 F, is the composite uncertainty for relays taken from Table H.1 of ReferenceAll. This composite uncertainty is considered to be Realistic Lower Bound Case according to Table H.1 of Reference 411, and it is suggested for use in calculating the median capacity, 0c 0 .30 HCLPFHFA5IA-c1o7o := min(GDRH Aux6-s85, cDRy-4ux6-s's)'PGA HCLPFHFAS1A C1% = 0.30'!

(z.sa.or)

Am-HFAg l A-c 1o/o i= HCLPFHFAS1 A-c1 o/o' e 1 C1o/o = 0.60 Ratiogl 0o/o_C1o7o

= 1.36 Ratio of C1 owlOtuo from Table H.1 of Ref. A11 HCLPFHFAE1A_C1 0% := Ratiogl 0%_C1 % " HCLPFHFASlA_Cl %

HCLPFHFAS1A C10% = 0.41.I Page 422 of 422

2734296-R-01.5 Ranision 0 lune 12 201.7 Page Bl of B5 APPENDIX B COMPONENTS IT)ENTIFIED FOR HIGH.FREQUENCY CONFIRMATION lffiGonrulting

[iRrzzo

273l{ZFR475 Rf,iqtr 0

[w72,2417 Prce 82 ulB6 TABLE Fl COMPONENTS II'EI{TIFIED FOR HIGH-FREQUENCY CONI'IRMATION CopIPoNENT ENCL0SURE CoITfoNENT EVALUATToN No.

Rruv/ ELEv. C I o/'.{'} cl0%G)

UNTT BLDG.

ID SYSIEM CoITTAC'roF/ (ft) BAsls FoR c/D EvALUATIoN G) G)

TYPE MANUFACIURER ID TVPE tr'rTNsuoN BREAEER Capacrrv Rarrool RESULT MoDELNo.

ACl03A6-r/C1 AC/DC Power ACt03A6-2iC I Suppon Syste4 ADt03A6-liDl Actuates BN ADt03/86-21D2 Lockout AC103/5r-lX ltot$live IEEEiANSI Caprcity >

I AC 103/5r -r Relay G,E, }IFA53K9IF CI,DI Switchgetr Au.6 585 C37-9E r6t t.6t Demd 0_32 0,44 AC I03/5t -2 ACiDC Pows AC l0li5l -3 Suppon System, ADl03/51-lX OverClJlIut ADt03/51-l hol,ection ADl03/51-2 AD103/51-3 AC/DC Power ACl 10/514 hot*tive Support Systsrn, IEEE/ANSI C8prcity >

) I ADI l0/514 Relay Over Cmnt Westinghow co-2 cl,Dt $witchgeu Am,6 585 C37-9E tcst 1.93 Detmd 0_39 0,53 Protection ACiDC Power AC I l0/51'5 Pmtctive Support Syste4 IEEE/ANSI Captrity >

3 I Clmnt Westiqlow co.5 CI,DI Switchgetu Au.6 585 tst 2.90 0.J8 o.'t9 ADr r0/5r-5 Relay Over C37-98 Delmd hotection ACr l0/5lGS-3 AC ICE I t.50/51 AC/DC Power AC lCE 12.50/5 1 Protetive Support Syste4 IEEE/ANSI Capacity >

4 I ADr r0/5lGs-3 Relay OwrCulmt Westinghow co-l1 CI,DI Switchgffi Aq.6 5E5 C37-98 tert 4.51 Demd 0.90 1.23 ADTDFT r-50/51 Protection AD tDFl2-50/51 AC I 0?-50i5 I AC t0t-50i5 I AC 109-50/5 I AC/DC Power AC lt S-5oifl Protetive Support System, IEEEiAI{SI Caprcity >

I ABB COM.5 CI, DI Swiichgear AUX.6 585 2.90 0.58 0.79 ADt07-50/51 Relay OYer ClJrent C37-98 test Demd AD t0t-50/5 t Protection 4D109-50/51 ADI r t-50/5 t

[lCmnfthe QRTZZO

27il29(rR415 Rsisin0 lffi72,2t17 P$e83dB6 TABLEEI COMPONENTS IDENTIFIEI' TOR HIGH-F'REQUENCY CONI'IRMATION (CONTINUED)

CoMPoNEl{T ENcl-osuRE CotroNENT EVTLUATToN No, RELAYi ELEV clo/$) Cl0'/r)

Un-IT BLDG.

SYSTEM CoNTAc.IoR/ (ft) Blsts ron c/r) EVALUATIoN (B) (8)

ID TYPE MANUFACTURER ID TYPE tr'uNCrIoN BREAxER CAPACrIY f,4116{t) RESULT M0DELNo.

AC 107/50G$

AC/DCPowq AC I I3/5OGS SupFort Systerf 6 ACICEI I/5OGS Groud Sasug ADl07/50GS Relay ADIDFI2/5OGS Protmtive IEEE/ANSI Cspacity >

ABB ITE Type 50D CI,DI Switchgear Aux.6 585 5.00 1.00 t_36 Relay C3?-98 r6t Dcmd AC103/5 lCS-1 AC/DC Powg AC103/5 rGS-2 Support SystE, 7 I ADt03/5tGS.t O1,trCmt ADl03i5lCS-2 hotection AC/DC Power ACl l0/5lx hotqtivc Suppofl Systeq Capacity >

I 1 ADt t0/5 tx Relay OvEr Clrctrt G.E, HGAtTC6l C1,Dl Switchgeu Au.6 585 GERS r.55 DeImd 0.31 0.42 Protection AC/DC Power ACI0ti5tv2DG Prctelive Sup,pon Systffi; IEEE/ANSI Capacity >

I I ADt0t/51v2DG Relay Back Up OC Trip ABB cov-9 ct, Dl Switcbgear Au.6 5t5 C37-98 t6t 4,51 Deffid 0.90 1.23 ofACl0l RCS Invatory Conuol; Conrol of Contol Capacity >

I BFl285/42 Relay Valve RC20O; FI24 Aq.6 603 3. l2 Derud 0.62 0.85 Drain wlve to RCS Elmlro-Qrnch Tank Mehanical EPRI l0 Cutler Hamer Contaotor MCC IIF TeEt RCS lnvfltory A2OIKIC ConEol; Contsol of Control Capacity >

I BFl26t42 Valve RC2394; FI IA Aux.7 lub 0.61 0_83 Relay RCS Sample Demd Isolotion Vslve lllCardtq

[iRrzzo

27t291t,R475 Xru.cfr 0

[me12,2077 Pt*t &4 ilBG TABLE BT COMPONENTS IDENTIF'IED T'OR HIGH.FREQUENCY CONFIRMATION (CONTINUED)

CoIlPoNENT ENCLoSURE Cotr0oNENT EVALUATIoN RET.AY/ ELEV. Cl./.(r) ClO./.o)

No. UNIT BLDc. (d SYSTEM CoNTAcIoR/ (ft) BAsIs FDn C/D EVALUATIoN (c)

ID TYPE MaNtrFn(rltRER ID TYPE FUNcTIoN BREAI(ER CAPACIIY R.lrtofl, RESULT MoDELNo.

RCS Invenlory Contol; Conlrcl Control I BFLt27l42 of Vslve RC239Bt.

Rehy RCS Sample Isolation Valve Capaoity >

FI lA Au,7 603 3.06 Demd 0.61 0.t3 RCS lnventory Elrctro-Cotrlrol ; Conlrol Control Mebanical EPRI l0 I BF l 128/42 of Valve RC240B; Cutler Hsmr Conlsctor MCC Relay IIF Te$t RCS Sample A2OIKIC Isolalion Valve RC$ hventory Contol; Conhol Cmtrol of Valve Capacity >

I BEI t8t/42 Relay RC240A; RCS EI IB Au.8 585 5.27 DeImd 1,05 1.43 Sample Isolation Valve EDcl Shutdoffi R7_1 c3621 Relry EDG2 Shutdom R72 Relay c3672 EMERDSL GEN OTR_I I-I OVERSPEED c3621 TRIP RELAY Contol IEEE/ANSI Capacity >

II I Relay EMERDSL GEN SqureD 850tKPDl3V63 Relay Panel Au6 585 C37-98 r6t 3.06 Demd 0,61 0.83 OTR_2 I-2OVERSPEED c1622 TRIP RELAY EMERDSL GEN R?I I.I FAIL TO C362I START REI,AY EMLRI'SL OEN R_2_2 1.2 FAIL TO c3622 START RET,AY fff Candt*te

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2734?s.R{1s I&ar'.q;m O ftnc12,2077 PtccBS ofB6 TABLE El COMPONENTS IDENTIFIED FOR HIGH-FREQUENCY CONFIRMATION (coNTrmrED)

CoMPoNENT ENcLosuRE CowoNEIr'r EVALUATToN RELAY/ ELEv BAsrs Cl'loo) Cl0./to)

No. UNrI CoNTAcron/ BLDc.

(ft) f,0R C/D EvALUATto (8) (B)

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300 Commerce, Suite 200 lrvine, CA 92602 AESGonsulting ABS GROUP OF COMPAHIES,INC.

16855 Northchase Drive Houston, TX 77060 Telephone 714-7344242 Telephone 281- 673-2800 Fax 714-7344252 Fax 281-673-2801 HORTH AMERICA SOUTH AMERICA EUROPE 2100 Space Park Drive, Suite 100 Maca6, Brazil Sofia, Bulgaria Houston, TX 77058 Telephone 55-22-2763-701 I Telephone 359-2-9632049 Telephone 713-929-6800 Rio de Janeiro, Brazil Piraeus, Greece Energy Crossing ll, E. Building Telephone 55-21 -3 1 79-31 82 Telephone 30-2104294046 1501 1 Katy Freeway, Suite 100 Sao Paulo, Brazil Genoa, ltaly Houston, TX 77094 Telephone 55-1 1 -3707-1 055 Telephone 39-010-251 2090 1525 Wilson Boulevard, Suite 625 Vina del Mar, Chile Hamburg, Germany Arlington, VA 22209 Telephone 56-32-2381 780 Telephone 4940"300-92-22-21 Telephone 703482-7373 Bogota, Colombia Las Arenas, Spain Fax 703482-7374 Telephone 571-2960718 Telephone 34-94464-0444 10301 Technology Drive Chuao, Venezuela Rotterdam, The Netrerlands Knoxville, TN 37932 Teleph one 58-212-959 -7 442 Telephone 31 -1 0-206-0778 Telephone 865-966-5232 Fax 865-966-5287 Lima, Peru Amsterdam, The Netherlands Telephone 51-1437-7430 Telephone 31-205-207-947 1745 Shea Center Drive, Suite 400 Highland Ranch, CO 80129 Manaus, Brazil Griteborg, Sweden Telephone 303S74-2990 Telephone 55-92-3213-951 1 Telephone 46-70-283-0234 1390 Piccard Drive, Suite 350 Montevideo, Uruguay Bergen, Nonrrray Rockville, MD 20850 Telephone 5982-2-901 33 Telephone 47-55-55-1 0-90 Telephone 301-907-9 1 00 UNITED KINGDOM Oslo, Nonray Fax 301490-7185 Telephone 47 51 -27 40 31 15 East Lion Lane, Suite 160 EQE House, The Beacons Stavanger, Nonrray Salt Lake City, UT 84121 Wanington Road Telephone 47-51-93-92-20 Telephone 801 -333-7676 Birchwood, Wanington Trondheim, Norway Fax 801-333-7677 Cheshire WA3 6WJ Telephone 44-1925-287300 Telephone 47-73-900-500 140 Heimer Road, Suite 300 San Antonio, TX 78232 3 Pdde Place ASIA.PACIFIC Telephone 210495-5195 Pride Park Derby DE24 8QR Ahmedabad, lndia Fax 210495-5134 Telephone 44{-1332-254410 Telephone 079 4000 9595 823 Congress Avenue, Suite 1510 Unit 3b Damery Works Navi Mumbai, lndia Austin, TX 78701 Woodford, Berkley Telephone 91-22-757$780 Telephone 512-7 32-2223 Fax 512-233-2210 Gloucestershire GL13 9JR New Delhi, lndia Telephone 444-145+269-300 Telephone 91-1 145634738 55 Wes$ort Plaza, Suite 700 St. Louis, M0 63146 ABS House Yokohama, Japan Telephone 314+19-1550 1 Frying Pan Alley Telephone 81 45450-1 250 Fax 314419-1551 London E1 7HR Kuala Lumpur, Malaysia Telephone 44-207 -377 4422 Telephone 603-79822455 One Chelsea Street New London, CT 06320 Aberdeen AB25 1XQ Kuala Lumpur, Malaysia Telephone 860-7014608 Telephone 44-0-122+392100 Telephone 603-21 61-5755 100 Danbury Road, Suite 105 London WlT 4TQ Beijing, PR China Ridgefield, CT 06877 Telephone 44-0-203-30 1 -5900 Telephone 86-10-581 1 2921 Telephone 2034314281 MIDDLE EAST Shanghai, PR China Fax 203431-3643 Telephone 86-214876-9266 1360 Truxtun Avenue, Suite 103 Dhahran, Kingdom of SaudiArabia Busan, Korea Norh Charleston, SC 29405 Telephone 966-3-868-9999 Telephone 82-51452466 1 Telephone 843-2974690 Ahmadi, Kuwait Seoul, Korea 152 Blades Lane, Suite N Telephone 965-3263886 Telephone 82-2-5524661 Glen Burnie, MD 21060 Doha, State of Qatar Telephone 410-5144450 Alexandra Point, Singapore Telephone 974{4-13106 Telephone 65-6270{663 MEXIC0 Muscat, Sultanate of Oman Kaohsiung, Taiwan, Republic of China Telephone 968-597950 Ciudad del Carmen, Mexico Telephone 886-7-271-3463 lstanbul, Turkey Telephone 52-938-3824530 Bangkok, Thailand Telephone 90-21246141 27 Mexico City, Mexico Telephone 662-399-2420 Abu Dhabi, United Arab Emirates Telephone 52-55-551 14240 West Perth, WA 6005 Telephone 971 -2491 2000 Monteney, Mexico Telephone 61-8-9486-9909 Dubai, United Arab Emirates Telephone 52-81 -831 94290 INTERHET Telephone 9714-33061 16 Reynosa, Mexico Telephone 52-899-920-2642 Additional office information can be found at:

mnv.abslroup.com Veracruz, Mexico Telephone 52-229-980-81 33

v t e Letter Sequence Supplement
CAC:MF3728, (Approved, Closed)
Initiation Acceptance... Supplement, Supplement Administration Questions Answers Results Approval Other:
MONTHYEAR2017-01-12 [Table View]

L-17-233, High Frequency Supplement to Seismic Hazard Screening Report, Response to NRC Request for Information, Per 10CFR50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force (NTTF) Review of Insights from the Fukishima Dai-ichi Accident

Text

SUBJECT:

References:

SUMMARY

5.0 CONCLUSION

6.0 REFERENCES

SUMMARY

1.0 INTRODUCTION

5.0 CONCLUSION

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L-17-233, High Frequency Supplement to Seismic Hazard Screening Report, Response to NRC Request for Information, Per 10CFR50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force (NTTF) Review of Insights from the Fukishima Dai-ichi Accident

Text

SUBJECT:

References:

SUMMARY

5.0 CONCLUSION

6.0 REFERENCES

SUMMARY

1.0 INTRODUCTION

5.0 CONCLUSION

Navigation menu

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