ML17244A281

16C4384-RPT-005, Rev 005, 50.54(f) NTTF 2.1 Seismic High Frequency Confirmation.

ML17244A281
Person / Time
Site:	Cooper
Issue date:	05/08/2017
From:	Masiunas A Stevenson & Associates
To:	Office of Nuclear Reactor Regulation, Nebraska Public Power District (NPPD)
References
16C4384-RPT-005, Rev 005
Download: ML17244A281 (49)

Text

SA Stevenson

& Assoc i ate s En g m ee rm g So/1111 o n sfo r N 11 c l ear En e r gy Document No: 16C4384-RPT-005 Revision 0 May 8, 2017 50.54(f) NTTF 2.1 Seismic High Frequency Confirmation Prepared for: Nebraska Public Power District Cooper Nuclear Station Brownville, Nebraska Stevenson

& Associates 1626 North Litchfield Road, Suite 170 Goodyear, AZ 85395 SA 50.54(t) NTTF 2.1 Seismic High Frequency Confirmation 16C4384-R PT-005 Rev. 0 P a ge 2 of49 REVISION RECORD Initial I ssue (Re v. 0) Prepared by: 5/8/2017 Reviewed by: 5/8/2017 ons antmos konomou Approved by: 51812017 Revision Historv Rev. Prepared by/ Reviewed by/ Approved by/ Description of Revision No. Date Date Date SA 50.54 (t) NTTF 2.1 Seismic Hi¥h Frequency Confirmation TABLE OF CONTENTS:

l 6C4384-RPT-005 Rev. 0 P age 3 of 49 Intr o ducti o n .........................................

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.. 6 I . I Purpose .........................

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..... 6 1.2 Background

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....... 6 I .3 Approach ..............................................

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7 1.4 Plant Screenin g .............................................

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......... 7 2 Selection of Components for High-Frequency Screening

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8 2.1 Reactor Trip/Scram ..............

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...... 8 2.2 Reactor Vessel Inventor y Control .....................................................................

............... 8 2.3 Reactor Vessel Pre ss ure Control ......................

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.. 10 2.4 Core Cooling .................

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......... 11 2.5 AC/DC Power Support Systems .........................

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.......................... 13 2.6 Summary of Selected Components

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............................. 17 3 Seismic Evaluation

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18 3.1 Horizontal Seismic Demand ............................................

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18 3.2 Vertical Seismic Demand ...........................

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.................................. 18 3 .3 Component Horizontal Seismic Demand ................................

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.............. 21 3.4 Component Vertical Seismic Demand ...............................................

............................ 22 4 Contact Device E valuations

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...... 23 5 Conclusions

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24 5.1 General Conclusions

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.......... 24 5 .2 Identification of Follow-Up Actions .............

................................................................. 24 6 References

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25 A. Representative Sample Component Ev aluation s ....................................................

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32 A.1 High Frequency Seismic Demand ............................

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....... 32 A.2 H ig h Frequenc y Capacity ..................

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36 B. Components Identified for High Frequency Confirmation

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........................ 38 TABLE OF TABLES: Table 3-1: Soil Mean Shear Wave Velocity vs. Depth Profile ...........................

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19 Table 3-2: Horizontal and Vertical Ground Motions Respon se Spectra .................

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.. 20 Table B-1: Components Identified for High Frequency Co nfirmation

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.. 38 Table B-2: Reactor Coolant Leak Path Valves Identified for High F requenc y Confirmation

..... 48 SA 50.54(f) N!TF 2.1 Seismic High Frequency Confirmation EXECUTIVE

SUMMARY

I 6 C4J84-RP T-00 5 R ev. 0 Pa g e 4 o f4 9 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 deterred status [1]. In particular, this report provides information requested to address the High Frequency Confirmation requirements of ltem (4), Enclosure l, Recommendation 2.1: Seismic , of the March 12 , 2012 letter [ l ]. Following the accident at the Fukushima Dai-ichi nuclear power plant resulting from the March 11, 2011 , Great Tohoku Earthquake and subsequent tsunami, the Nuclear Regulatory Commission (NRC) established a Near Tenn Task Force (NTTF) to conduct a systematic review ofNRC processes and regulations and to determine if the agency should make additional improvements to its regulatory system. The NTTF developed a set of recommendations intended to clarify and strengthen the regulatory framework for protection against natural phenomena

[2]. Subsequently, the NRC issued a 50.54(f) letter on March 12 , 2012 [I], requesting information to assure that these recommendations are addressed by all U.S. nuclear power plants. The 50.54(t) 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(t) letter was a request that licensees perform a " confirmation, if necessary , that SSCs , which may be affected by high-frequency ground motion , will maintain their functions important to safety." EPRI I 025287 , "Seismic Evaluation Guidance:

Screening, Prioritization and Implementation Details (SPID) for the resolution of Fukushima Near-Term Task Force Recommendation 2.1: Seismic" [3] provided screening, prioritization, and implementation details to the U.S. nuclear utility industry for responding to the NRC 50.54(t) letter. This report was developed with NRC participation and was subsequently 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.

Subsequent guidance for performing a High Frequency Confirmation was provided in EPRI 3002004396, "High Frequency Program , Application Guidance for Functional Confirmation and Fragility Evaluation

," [4] and was endorsed by the NRC in a letter dated September 17, 2015 [5]. Final screening identifying plants needing to perform a High Frequency Confirmation was provided by NRC in a letter dated October 27 , 2015 [6). This report describes the High Frequency Confirmation evaluation undertaken for Cooper Nuclear Station. The objective of this report is to provide summary information describing the High Frequency Confinnation 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.

SA 50.54(t) N!TF 2.1 Seismic High Frequency Confirmation

!6C4384-RPT-005 Re v. 0 Pa ge 5 of 49 EPRI 3002004396

[4] is used for the Cooper Nuclear Station engineering evaluations described in this report. In accordance with Reference

[4], 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 confirmation

• Estimation of a vertical ground motion response spectrum (GMRS)

• Estimation of in-cabinet seismic demand for subject components

• Estimation of in-cabinet seismic capacity for subject components

• Summary of subject components' high-frequency evaluations SA 50.54(t) N!fF 2. l Seismic High Frequency Confirmation INTRODUCTION

1.1 Purpose

1 6C 4384-RP T-00 5 R ev. 0 Page 6 o f49 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 [ 1 ). In particular , this report provides requested information to address the High Frequency Confirmation requirements of Item ( 4), Enclosure l, Recommendation

2. l: Seismic, of the March 12 , 2012 letter [l]. 1.2 Background Following the accident at the Fukushima Dai-ichi nuclear power plant resulting from the March l I , 2011, Great Tohoku Earthquake and subsequent tsunami , the Nuclear Regulatory Commission (NRC) established a Near Term Task Force (NTTF) to conduct a systematic review ofNRC processes and regulations and to determine ifthe agency should make additional improvements to its regulatory system. The NTTF developed a set of recommendations intended to clarify and strengthen the regulatory framework for protection against natural phenomena

[2]. Subsequently , the NRC issued a 50.54(t) letter on March 12 , 2012 [ 1 ], requesting in formation to assure that these recommendations are addressed by all U.S. nuclear power plants. The 50.54(t) letter requests t h at licensees and holde r s of construction permits under I 0 CFR Part 50 reevaluate the seismic hazards at their sites against present-day NRC requirements and guidance. Included in the 50.54(f) letter was a request that licensees perform a " 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 Implementation Details (SPID) for the resolution of Fukushima Near-Term Task Force Recommendation

2. l: Seismic" [3) 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.

Subsequent guidance for performing a High Frequency Confirmation was provided in EPRl 3002004396 , "High Frequency Program, Application Guidance for Functional Confirmation and Fragility Evaluation," [ 4) and was endorsed by the NRC in a letter dated September 17 , 2015 [5]. Final screening identifying plants needing to perform a High Frequency Confirmation was provided by NRC in a letter dated October 27 , 2015 [6]. On March 31, 2014, Cooper Nuclear Station submitted a reevaluated seismic hazard to the NRC as a part of the Seismic Hazard and Screening Report [7]. By letter dated October 27, 2015 [6], the NRC transmitted the results of the screening and prioritization review of the seismic hazards reevaluation.

SA 50.54(f) N!TF 2.1 Seismic High Frequency Confirmation l 6 C4 3 84-RPT-00 5 Re v. 0 P age 7 of 49 This report describes the High Frequency Confirmation evaluation undertaken for Cooper Nuclear Station 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 [5]. The objective of this report is to provide summary information describing the High Frequency Confirmation evaluations and results. The level of detail provided in the report is intended to enable NRC to understand the inputs used , the evaluations performed , and the decisions made as a result of the evaluations.

1.3 Approach

EPRI 3002004396 [ 4 J is used for the Cooper Nuclear Station engineering evaluations described in this report. Section 4.1 of Reference (4] 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:

• Cooper Nuclear Station's SSE and GMRS Information

• Selection of components and a list of specific components for high-frequency confirmation

• Estimation of seismic demand for subject components

• Estimation of seismic capacity for subject components

• Summary of subject components' high-frequency evaluations

• Summary of Results 1.4 Plant Screening Cooper Nuclear Station submitted reeva l uated seismic hazard information including GMRS and seismic hazard information to the NRC on March 3 l , 2014 (7] and amended this information on February I 1 , 2015 [8]. In a letter dated September 8, 2015 , the NRC staff concluded that the submitted GMRS adequately characterizes the reevaluated seismic hazard for the Cooper Nuclear Station site [9]. The NRC final screening determination letter [6] concluded that the Cooper Nuclear Station GMRS to SSE comparison resulted in a need to perform a High Frequency Confirmation in accordance with the screening criteria in the SPID [3].

SA 50.54(t) N!TF 2.1 Seismic High Frequency Confirmat i on 1 6 C 4 38 4-RPT-00 5 R ev. 0 Page 8 of 4 9 2 SELEC T ION OF COMPONEN T S FO R HIGH-FREQUENCY SCREENING The fundamental objective of the high frequency confirmation review is to determine whether the occurrence of a seismic event could cause credited equipment to fail to perform as necessar y. An optimized evaluation process is applied that focuses on achieving a safe and stable plant state following a seismic event. As described in Reference

[4], this state is achieved b y confirming that key plant safety functions critical to immediate plant safety are pre s erved (reactor trip , reactor vessel in v entory and pressure control , and core cooling) and that the plant operators have the necessary p ower available t o achieve and maintain this state immediatel y following the seismic event (AC/DC power support system s). Within the ap p licable func t ions, the com p onents that would need a high frequency confirmation a r e contact control devices subject to intermittent states in seal-in or lockout (SILO) circuit s. Accordingly , the objective of the review as stated in Section 4.2.1 of Reference

[4] i s t o determine if seismic induced high frequency relay chatter would prevent the comp l etion of the following key functions.* 2.1 R e actor Trip/Sc ram The reactor trip/SCRAM function is ide n tified as a key function in Reference

[4] to b e conside r e d in th e High Freq u e n cy Confi r mation. The same report also states that, "the design requirements preclude the ap p lication of s e al-in or lockout circuits that pr eve nt r e actor trip/SCRAM/uncti o n s" and that " No high-fr e qu e nc y re v ie w of th e reactor trip/S CRAM is necessary. " 2.2 R e actor Vessel I n ve nt ory Co ntrol The reactor coolant s ystem/reactor vessel invent o ry control s ystems were re v iewed for contact co n trol dev i ces in seal-in a n d lockout (S I LO) circuits that would create a L o ss of Coolant Accident (LOCA). The focus of the review wa s contact control devices that could lead t o a significant leak path. Check valves in series with active valves would prevent significant leaks due to miso p e r ation of t h e active valve; t h erefore , SILO circuit reviews were not required for those active valves. Reactor coolant s y stem/reactor vessel inventory co ntrol s y s tem revi e ws were performed for valves associated with the following function s:

• Main Steam ,

• High Pres s ure Core Injection ,

• Residual Heat Removal,

• Cont r o l Rod Drive ,

• Reactor Water Clean-Up *T h e sel e ct i o n of c om p o n e nt s for hi gh fre qu e ncy sc r eening is descr i be d in St e ve n s o n & A sso cia t e s r e p ort J 6C438 4-RPT-OO J [72] a nd is s um ma r ize d herein.

SA 50.54(f) NTTF 2.1 Seismic High Frequency Confirmation I 6 C43 84-RPT-0 05 Rev. 0 Page 9 of49 A table listing the valves selected for analysis and their associated P&ID is included as Table B-2 of this report. 2.2. I Main Steam Valves Main Steam Isolation Valves MS-AO-A0 8 0A I B I C I D , MS-AO-A086 AI B IC I D Electrical control for the solenoid-operated pilot valves is via relays l 6A-K 14, l 6A-K 16, l K5 l and l 6A-K52. These relays are slaves to l 6A-K7 A/B/C/D isolation logic relays [ 10, 11 ]. These relays are energized for at-power operation and de-energized to close the valves [12, 13]. In the energized state l 6A-K7 A/B/C/D are sealed in and any chatter in the control logic would break the seal-in and close the valves. This action is a desired response to the seismic event and for this reason chatter is acceptable and no contact devices in this circuit meet the selection criteria.

Main Steam Line Drain Valves MS-MOV-M0 7 4 , MS-MOV-M0 77 These normally-open motor-operated valves close on an isolate signal from 16A-K7A/B/C/D via slave relays 16A-K56 and 16A-K57 [14, 15, 16]. Limit switches in the opening circuits prevent seal-in of the opening contactors and there are no permissive contacts in the close circuit which could block valve closure manually or automatically via an isolation signal. Auto Slowdown Valves MS-RV-71ARV I BRV I CRV I ERV I GRV I HRV Electrical control for the solenoid-operated pilot valves is via relays 2E-K6A/B and 2E-K 7 A/B. These relays are controlled by the Reactor Pressure Vessel (RPV) Low Level Logic, the Residual Heat Removal (RHR) Pump Discharge Pressure relays 1OA-K101 A/B and 1OA-K102A/B, and the Core Spray Pump Discharge Pressure relays 14A-K23A/B and 14A-K25A/B

[17, 18, 19]. The RHR and Core Spray Pump Pressure relays do not seal-in [20, 21, 221 and, based on initial conditions at the time of the event, would block any inadvertent seal-in of the RPV Low Level Logic. Thus, there are no SILO relays in this logic which could cause the Auto Slowdown Valves to remain open following a seismic event. Main Blowdown Valves MS-RV-71 DRV I FRV Electrical control for the solenoid-operated pilot valves is via relays 821 M-2E-K20A/B and 821 M-2E-K21A/B. Seal-in of these relays is blocked by pressure switches 2-3-51 Band 2-3-51 D [23]. 2.2.2 High Pre sure Core Injection Valves High Pressure Core Injection Steam Supply Line Isolation Valves HPCI-MOV-15 , HPC/-MOV-16 These normally-open motor-operated valves supply steam to the HPCI turbine. The opening circuit is controlled by a rugged hand switch and permissive from 23A-K5 I, 23A-K44, 23A-K 15, and 23A-K34 [24]. There is no seal-in in the opening circuit. The closing circuit is controlled manually by a rugged hand switch or automatically via the auto isolation relays K34 and 23A-K34, or the low steam pressure relays 23A-Kl5 and 23A-KS I [25, 26]. Any chatter in the isolation or low steam pressure logic would close the valves. Since RCCC, not HPCI , is credited for core cooling this seal-in causing valve closure is not a selection criterion.

SA 50.54(f) NTTF 2.1 Seismic High Frequency Confirmation t 6 C4384-RPT-00 5 Re v. 0 Pa ge to o f 49 There is no SILO which would prevent closure of these valves and thus no contact devices in this circuit meet the selection criteria.

2.2.3 Residual

Heat Removal Valves RHR Suction Cooling lsola1ion Valves RIIR-MOV-M0/7 , RHR-MOV-M0/8 These normally-closed motor-operated valves are opened via a normally-open control switch and relay permissive.

The valves can be closed manually via the control switch and automatically via an isolation signal. Sympathetic chatter on 16A-K29 and 42/0 auxiliary contact could cause valve RHR-MOV-MO 18 to open; and sympathetic chatter on l 6A-K30 and 72/10 auxiliary contact could cause valve RHR-MOV-MO 17 to open [27]. However , the low reactor pressure permissive in the control logic would prevent a seal-in of I 6A-K29 or I 6A-K30 [28]. After the period of strong shaking the normally-closed contacts of I 6A-K29 and I 6A-K30 would command these valves to reclose. Because there i s no seal-in and the valves reclose without operator intervention, chatter is acceptable and no contact devices in this circuit meet the selection criteria.

2.2.4 Control

Rod Drive Valves Control Rod Manual Positioning Valves CRD-SOV-S0/20 , CRD-SOV-S0/21, C RD-SO VS0/22 , C RD-SOV-S0/23; Control Rod Scram Valve CRD-AOV-CV/26 These valves are part of the Control Rod Drive Hydraulic Positioning System [29] and as such they are covered under the Reactor Trip/Scram category.

For more information, see Section 2.1 above. 2.2.5 Reactor Water Clean-Up Valves Reactor Water Cl e an-Up Isolation Valves RWCU-MOV-M015 , RWCU-MOV-M0/8 These are normally-open motor-operated valves which close upon an isolation signal. Open limit switches in the opening circuit prevent seal-in of the opening contactor auxiliary contact and no contacts prevent valve closure via the control switch or isolation relays I 6A-K26 and I 6A-K27 [27). These relays are energized for at-power operation and de-energized to close the valves [28]. In the energized state 16A-K26 and 16A-K27 are sealed in and any chatter in the control logic would break the seal-in and close the valves. This action is a desired response to the seismic event and for this reason chatter is acceptable and no contact devices in this circuit meet the selection criteria. 2.3 Reactor Vessel Pressure Control The reactor vessel pressure control function is identified as a key function in Reference

[4] to be considered in the High Frequency Confirmation.

The same report also states that " requir e d post event pressure control is typically provided by pa s siv e devi c es" and that " no sp e cific high frequenc y compon e nt chatter review i s requir e d for this fanction." L 4 , pp. 4-6)

SA 50.54(f) NTTF 2.1 Seismic High Frequency Confirmation 2.4 Core Cooling 16C4384-RPT-005 Re v. 0 Page 11 of49 Core cooling is also a key function in Reference

[ 4]. The core cooling systems were reviewed for contact control devices in seal-in and lockout circuits that would prevent at least a single train of non-AC power driven decay heat removal from functioning.

For BWR plants, the decay heat removal mechanism involves the transfer of mass and energy from the reactor vessel to the suppression pool. This requires the replacement of that mass to the reactor vessel via some core cooling system, e.g., reactor core isolation cooling (RCIC). Therefore, for this evaluation the following functions need to be checked: ( 1) Steam from the reactor pressure vessel to the RCIC turbine and exhausted to the suppression pool; (2) coolant from the suppression pool to the reactor via the RCIC pump; and (3) steam from the reactor pressure vessel vented to the suppression pool via the Safety Relief Valves (SR Vs). The selection of contact devices for the SR Vs overlaps with the RCS/Reactor Vessel Inventory Control Category.

The selection of contact devices for RCIC was based on the premise that RCIC operation is desired, thus any SILO which would lead to RCIC operation is beneficial and , for that reason, does not meet the criteria for selection. Only contact devices which could render the RCIC system inoperable were considered.

Seismically-induced contact chatter could lead to a false RCIC isolation Signal or false Turbine Trip, which would prevent RCIC operation. A false steam line break trip has the potential to delay RCIC operation while confirmatory inspections are being made. Chatter in the contacts of RCIC Isolation Signal Relay 13A-K 15 , the Steam Line High Differential Pressure Time Delay Relay RCIC-TDR-Kl2, the Steam Line Space Excess Temperature Relays 13A-KIO and KI I, or the Reactor Pressure Relay 13A-Kl3 may lead to a RCiC Isolation Signal and seal-in of 13A-Kl5 [30]. This would cause the RCIC isolation Valves to close and the RCIC Trip and Throttle Valve to trip. Simultaneous chatter in identical contact devices controlling these relays could also lead to seal-in: TS-I 3-79A/C, TS-13-SOA/C, TS-13-8 IA/C , TS-13-82A/C, and PS-13-87 A/C. (The 3.5 second time delay t associated with RCIC-TDR-K 12 [31] will mask any chatter on dPIS-13-83, so it is excluded.)

The same selection rationale applies to the identical Division 2 devices: 13A-K33 , RCIC-TDR-K32, 13A-K30 , 13A-K3 l , TS-13-798/D , TS-13-808/D , TS-13-818/D, TS-13-828/D , and PS-13-878/D [32]. Any chatter that may lead to the energization of the Trip and Throttle Valve Remote Trip Circuit is considered as SILO , as it will close the valve and require a manual reset prior to restoration of the RCiC system. Chatter in Turbine Trip Auxiliary Relay 13A-K8, or in the devices which control this relay; the Turbine Exhaust High Pressure Relay l 3A-K6, the Pump Suction Low Pressure Relay l3A-K7, and the isolation Signal Relay I 3A-K 15 [30]. Similar chatter in the contact devices that drive those relays (and not already covered in the RCJC Isolation Signal t High frequem;y t:hatter is not expected to cause continuous contact closure for more than 100 millisecond s at any one time. When contact chatter applies power to the coil of a time delay relay wit h delay times significantly longer than this, the coil is n ot co ntinuall y energized long enough to satisfy the timing function and thus the time delay relay will not c hange state.

SA 50.54 (f) NTTF 2.1 Seismic High Frequency Confirmation 1 6C 4 3 8 4-RPT-00 5 Rev. 0 Pa g e 12 of 4 9 analysis) could also lead to a turbine trip: PS-13-72A/B. (The time delay associated with K7 will mask any chatter on PS-13-67-1, so it is excluded.) In addition to control of the RCIC Isolation Valves, several other valves need to be properly aligned for RCIC operation.

Steam-to-Turbine Valve RCIC-M0-131 is normally closed and opens on a reactor low level signal or control hand switch [33 , 34 , 30]. Once open, it is reclosed on a reactor hi gh water level or control hand switch. Chatter in the closing circuit of this normally-closed valve is blocked by open rugged limit and torque switches. Chatter in the opening circuit could lead to valve opening , which would be beneficial to RCIC operation. Based on this analysis, there are no moving contact devices in the control circuit of this valve that meet the selection criteria.

Pump Suction from Suppression Pool Valve RCIC-M0-41 is normally closed and opens on an Emergency Condensate Storage Tank (ECST) low level signal or control hand switch [35 , 34 , 32, 36]. Once open , it is reclosed by a control hand switch only. Chatter in the closing circuit of this normally-closed valve is blocked by open rugged limit and torque switches. Chatter in the opening circuit could lead to valve opening, which would in turn close Pump Suction from ECST Valve RCIC-M0-18 and align pump suction from the suppression pool. This would not impact RClC's ability to provide core cooling and based on this , there are no moving contact devices in the control circuit of this valve that meet the selection criteria. Pump Suction from Emergency Condensate Storage Tank Valve RCIC-M0-18 is normally open and closes automatically when Suppression Pool Valve RCIC-M0-41 is fully open, or manually via a control hand switch [35 , 36 , 30 , 32]. Contact chatter in the valve clo s ing circuit could close the valve , however the valve would reopen automatically in response to RCIC initiation on a low reactor level signal , or would open upon operator command via a control hand switch!. Based on this analysis, there are no moving contact devices in the control circuit of this valve that meet the selection criteria.

Pump Discharge Valve RCIC-M0-20 is normally open and closes via a control hand switch only [35 , 36 , 30]. Chatter in the closing contactor auxiliary contacts could cause valve closure , however the valve would reopen automatically in response to RCIC initiation on a low reactor level signal, or would open upon operator command via a control hand switch. Based on this anal y sis , there are n o moving contact device s in the control circuit of this valve that meet the selection criteria.

Pump Discharge Valve RCIC-M0-21 is normall y clo s ed and opens on a reactor low level signal or control hand switch [35 , 36, 30]. Once open, it is reclosed by a control hand switch only. Chatter in the closing circuit of this normally-closed valve is blocked by open rugged limit and torque switches.

Chatter in the opening circuit could lead to valve opening , which would be beneficial to RCIC operation.

Based on this analysis , there are no moving contact devices in the control circuit of this valve that meet the selection crite r ia. : M a nu al R C!C in i tia tion i s pr es um ed t o inc lude o perat o r al i g nm e nt of valv e s v ia t he R C!C sys t e m c o nt ro l s, inc l uding pump s uctio n l o tht: dt: s ired so urce.

SA 50.54(t) NTTF 2.1 Seismic High Frequency Confirmation 2.5 AC/DC Power Support Systems 1 6 C4 3 84-RP T-OO S Rev. 0 Pag e 13 o f49 The AC and DC power support systems were reviewed for contact control devices in seal-in and lockout circuits that prevent the availability of DC and AC power sources. The following AC and DC power support systems were reviewed:

• Emergency Diesel Generators ,

• Battery Chargers and Inverters ,

• EDG Ancillar y Systems , and

• 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 recharge the batteries. The availability of AC power is dependent upon the Emergency Diesel Generators and their ancillary support systems. EPRI 3002004396 (4) 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 ifthe trip was spurious due to contact chatter or in response to an actual system fault. The actions taken to diagnose the fault condition could substantially delay the restoration of emergency power. In order to ensure contact c h atter cannot compromise the emergency power system, control circuits were analyzed for the Diesel Generators (DG), Battery Chargers , Vital AC Inverters , and Switchgear

/Load Centers/MCCs as necessary to distribute power from the DGs to the Battery Charger s and DG Ancillary Systems. General information on the arrangement of safety-related AC and DC systems , as well as operation of the DGs, was obtained from Cooper's UFSAR [37). Cooper has two (2) DGs which provide emergency powe r for their two (2) divisions of Class IE loads, with o n e DG for each division (38). Four (4) battery chargers provide DC power and battery recharging functions

[39). (The output disconnect switches of the 250V IC and 125V IC chargers are locked open and for this reason were n o t considered in this analysis.)

The analysis considers the reactor is operating at power with no equipment failures or LOCA prior to the seismic event. The Diesel Generators 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 und e rvoltage relaying detecting the LOOP, the Class IE control systems must automatically shed loads , start the DGs , and sequentiall y load the diesel generators as designed.

Ancillary systems required for DG operation as well as Class 1 E battery chargers and inverters must function as necessary. The goal of this analysis is to identify any vulnerable contact devices which could chatter during the seismic e v ent, c au se a circuit 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 all division s. The selected contact device s for all divi s ions are included in Table 8-1.

SA 50.54(t) NTTF 2.1 Seismic High Frequency Confirmation 2.5.1 E mergency Diesel Generators 1 6 C4 3 8 4-R.PT-005 R ev. 0 Pa g e 14 of 4 9 The anal y sis of the E mergency D i esel Generators is broken down into the generator protective relaying and diesel engine control. General descriptions of these systems and controls appear in the UFSAR [37]. The control circuitry associated with each train is identical and for this reason only one train is described herein , however Table B-1 includes both trains. G e nerator Protecti v e Relaying The closure of the 52 EG 1 DG Circuit Breaker is prevented when either the 86 DG l Generator Lockout Relay or 86 JFE Bus Lockout Relay is tripped [40]. The control circuits for the DG Lockout Relay [41] include the 40 DGl Field Failure; 87-1DGl , 87-2 DGJ , and 87-3 DGl Differential Trip; 51 V-1 DG 1, 51 V-2 DG 1 , and 51 V-3 DG l Phase Overcurrent; 67 DG I Directional Overcurrent; and 27/59DG1 Abnormal Voltage protective relays. Chatter in any of these relays may trip the DG Lockout Rela y. Chatter in the 50 1 51-1 I FE , 50/51-2 I FE , and 50 1 51-3 I FE Phase Overcurrent protective relays associated with the normal power feed could lead to the tripping of the Bus Lockout Relay [42]. Diesel Engine Control Starting of the DG is blocked when the 86 DG 1 Generator Lockout Relay is tripped; and chatter in the 481SEX Incomplete Start Sequence , 630SDX Overspeed Shutdown , 4EMX or 4EMX3 Emergency Master, 14RX3 Running Master, or 14RY1 Running Slave relays could break the start seal-in and shut down the engine [43]. Chatter in the 62CLX Cranking Limit Timer may seal in the Incomplete Start Sequence Relay 48ISEX which would prevent engine start [43]. The coil of 62CLX is energized at the beginning of the start sequence.

Any chatter in the contacts comprising the coil circuit would be beneficial as it would reset the timer and prevent tripping the Incomplete Start Sequence circuit. The Overspeed Shutdown Relay may seal-in if chatter occurs in the 630SDL or 630SDR Overspeed Switches; or in the 140S Overspeed Auxiliary Relay, RI04 Auxiliary Speed Relay, or RT Relay Tachometer

[43 , 44J. The Running Master Relay 14RX3 is energized by either the RT Relay Tachometer RT , via Auxiliary Speed Relay RI 02 , or by the Magnetic Pickup Bypass Relay 14MPFB [43]. It is unlikely that chatter would occur in these diverse input contacts simultaneously in a way that would drop out I 4RX3 , and thus they are not considered in this analysis. Running Relay I 4RY 1 is energized by 14RX3 and is therefore covered by its analysis.

2.5.2 EDG Ancillary System s In order to start and o perate the Diesel Generators require a number of components and systems. For the purpose of identifying electrical contact devices, only systems and components which are electrically controlled are analyzed.

Information in the UFSAR [37] was used as appropriate for this analysis.

SA Starting Air 50.54(t) NTTF 2.1 Seismic High Frequency Confinnation 16C4384-RP T-00 5 Re v. 0 Page 15 o f49 Based on Diesel Generator availability as an initial condition the passive air reservoirs are presumed pressurized and the only active components in this system required to operate are the air start solenoids

[45], which are covered under the DG engine control analysis above. Combustion Air Intake and Exhaust The combustion air intake and exhaust for the Diesel Generators are passive systems [ 46] which do not rely on electrical control. Lube Oil The Diesel Generators utilize engine-driven mechanical lubrication oil pumps [47] which do not rely on electrical control. Fuel Oil The Diesel Generators utilize engine-driven mechanical pumps and DC-powered booster pumps to supply fuel oil to the engines from the day tanks [45]. The day tanks are re-supplied using AC-powered Diesel Oil Transfer Pumps. Chatter analysis of the control circuits for the electrically-powered transfer [48 , 49] and booster pumps [44, 50], as well as the Fuel Oil Shutoff Solenoid Operated Valves [51, 52] concluded they do not include SILO devices. The mechanical pumps do not rely on electrical control. Cooling Water The Diesel Generator Cooling Water System is described in the UFSAR [37]. This system consists of two cooling loops, jacket water and Service Water (SW). Engine driven pumps are credited for jacket water when the engine is operating (53]. These mechanical pumps do not rely on electrical control. Four SW pumps, 1 A, I B, IC , and ID, provide cooling water to the heat exchangers associated with the two DGs (54 , 55, 45 , 56]. There are no electrically operated valves in this flow path. In automatic mode, these pumps are started via a low discharge pressure signal and sequencing signal following DG start [57]. In standby mode, these pumps are sequenced to start automatically following a DG start. There is no SILO associated with the low discharge pressure signal. Chatter analysis of the DG start signal is included in the DG engine control analysis above. An analysis of the 52 SWPIA (52 SWPIB, 52 SWPIC , 52 SWPID) SW pump circuit breaker trip control circuits indicates chatter in the Pump Lockout Relay 86 S WP! A (86 SWPIB , 86 SWPIC , 86 SWPI D) or the Phase Overcurrent Relays 50/51 0A SWPlA and 50/51 0C SWPIA (50/51 0A SWPIB, 50/51 0C SWPIB , 50/5 l 0A SWPIC , 50/5 l 0C SWPlC , 50/51 0A S WP l D , 50/5 l 0C S WP 1 0) could trip the circuit breaker and prevent pump operation following the seismic event. Ventilation The Diesel Generator Building Ventilation System is described in the UFSAR [37]. During Diesel Generator Operation, ventilation is provided by Heating and Ventilation Units lC and HV-DG-10 and Exhaust Fans EF-DG-lA and EF-DG-IB [58]. In automatic mode, SA 50.54(t) NTTF 2.1 Seismic High Frequency Confirmation 16C4384-RPT-005 Rev. 0 Page 16 of49 these fans are started via the DG Start Signal. Chatter analysis of the DG start signal is included in the DG engine control analysis above. Other than SILO devices identified for the DG start signal, chatter analysis of the control circuits for these ventilation components

[59, 60] concluded they do not include SILO devices. 2.5.3 Battery Chargers Chatter analysis on the battery chargers was performed using information from the UFSAR as well as vendor schematic diagrams [61 , 62, 63]. Each batter y charger has a high voltage shutdown circuit which is intended to protect the batteries and DC loads from output overvoltage due to charger failure. The K3 High Voltage Shutdown (HVSD) circuits [64] in the 125V and 250V chargers have an output relay which shunt-trips the AC input circuit breaker, shutting the charger down. Chatter in the contacts of these output relays may disable the battery chargers , and for this reason meet the selection criteria.

2.5 .4 Inverters Analysis of schematics for the I A Static Inverter [65, 66] revealed no vulnerable contact devices and thus chatter analysis is unnecessary. 2.5.5 Switchgear, Load Centers, and MCCs Power distribution from the DGs to the necessary electrical loads (Battery Chargers , Inverters, Fuel Oil Pumps, and DG Ventilation Fans) was traced to identify any SILO devices which could lead to a circuit breaker trip and interruption in power. This effort excluded the DG circuit breakers , and the SW Pump breakers which are covered 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 circuit breakers in 4 I 60V and 480V AC Switchgear

[38] supplying power to loads identified in this section (battery chargers , EDG ancillary systems , etc.) have been identified for evaluation

52 I FE , 52SS1 F, 52 MCC K, 52 MCC LX; 52 lGE, 52 SSlG, 52 MCC S, 52 MCC TX. Bus Feeder Breaker Power from the Diesel Generator is fed to the 4 l 60 Switchgear Critical Bus IF (I G) via the I FE (I GE) circuit breakers.

This circuit breaker is tripped and locked-out by Lockout Relay s 86 I FE and 86 DG I, which are covered above, as well as Lockout Relays 86 IF A and 86 IFS, associated with the Normal Feeder Breaker and the Emergency Startup Transformer Breaker respectivel y [ 42]. Lockout Rela y 86 1 FA is tripped by Phase Overcurrent Relays 5 I 0A IF A , 51 0B 1 FA, 51 0C 1 FA. Lockout Relay 86 l FS is tripped by Phase Overcurrent Relays 51 0A IFS, 51 0B IFS , 51 0C IFS [67]. Chatter in any of these relays could trip the Bus Feeder Breaker. Station Service Step-Down Transformer The close control for the Station Service Step-Down Transformer IF circuit breaker is via a normally-open manual control switch. For this reason, any chatter that leads to a circuit breaker trip would not be automatically reset, leaving the breaker in the tripped position.

There are two potentially vulnerable contact devices which could trip this breaker if they chatter , the Phase Overcurrent Relays 50/51 0A SS IF and 50/51 0C SS IF [67].

SA 50.54(f) N!fF 2.1 Seismic High Frequency Confirmation 480V AC , 120 V AC, 250 VDC , and 125V DC Distribution and MCCs 16C4384-RP T-005 Rev. 0 Page 17 of 49 The 480V AC Load Centers and MCCs, and the 120V AC, 250 VDC, and 125V DC Distribution

[38, 68 , 69, 70, 71, 39) all use either Molded-Case Circuit Breakers or fused disconnect switches, both of which are seismically rugged [ 4, pp. 2-11]. 2.6 Summary of Selected Components The investigation of high-frequency contact devices as described above was performed in Ref. [72]. A list of the contact devices requiring a high frequency confirmation is provided in Appendix B, Table B-1.

SA 50.54(t) NTTF 2.1 Seismic High Frequency Confirmation 3 SEISMIC EVALUATION

3.1 Horizontal

Seismic Demand 1 6 C 4 384-RPT-0 05 Re v. 0 P age 18 of 49 Per Reference

[4], Section 4.3 , the basis for calculating high-frequency seismic demand on the subje c t components in the horizontal directi o n is the Cooper Nuclear Station horizontal ground motion response spectrum (GMRS), which was generated as part of the Cooper Nuclear Station Seismic Hazard and Screening Report [7] submitted to the NRC on March 31 , 2014 , amended on February I l, 2015 [8], and accepted by the NRC on September 8 , 2015 [9]. It is noted in Reference

[4] that a Foundation Input Response Spectrum (FIRS) may be necessary to evaluate buildings whose foundations are supported at elevations different than the Control Point elevation. H o wever , for sites founded on rock , per Reference

[4], " 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." 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 applicable buildings at Cooper Nuclear Station are founded on soil and have only the Control Point GMRS defined; therefore, the Control Point GMRS is conservatively used as the input at the building foundation.

The horizontal GMRS values are provided in Table 3-2. 3.2 Vertical Seismic Demand As described in Section 3.2 of Reference

[4], 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

[7] Table 2.3.2-2 , and below in Table 3-1.

SA 50.5 4(f) NTTF 2.1 Seismic H i gh Frequency Confirmation I 6C4384-RP T-005 R ev. 0 P age 19 of 49 Table 3-1: Soil Mean Shear Wave Velocity vs. Depth Profile Layer Depth Thickness, di VSi di I Vsi I [di I Vsi) Vs30 (ft) (ft) (ft/s) (ft/s) I 10.0 1 0.0 1,020 0.0 098 0.0 098 2 14.5 4.5 1,020 0.00 44 0.01 4 2 3 24.5 10.0 1.03 0 0.00 97 0.02 3 9 4 34.5 1 0.0 1 ,0 4 0 0.0096 0.033 5 5 40.5 6.0 1,0 4 0 0.00 58 0.0393 6 49.5 9.0 l , 1 20 0.0 080 0.047 3 7 59.5 10.0 1.620 0.0062 0.053 5 1,369 8 69.5 10.0 1.760 0.0057 0.0 59 2 9 79.5 10.0 1,760 0.0057 0.06 4 9 1 0 84.5 5.0 1 ,760 0.0028 0.06 77 II 94.5 10.0 2.7 50 0.0 03 6 0.0 7 1 4 12 97.0 2.5 7.292 0.0 003 0.0 7 1 7 13 98.4 1.4 7.2 94 0.0002 0.0 71 9 Using the shear wave velocity vs. depth profile , the velocity of a shear wave traveling from a depth of30m (98.4ft) to the surface of the site (V s30) is calculated per the methodology of Reference

[4], Section 3.2.

• The time for a shear wave to travel through each soil layer is calculated by dividing the layer depth (d;) by the shear wave velocity of the layer (V s 1).

• 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 ofOm to 30m (E[d;N s 1]).

• 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., V s30 = (30m)/L[d;N s 1J. The site's soil class is determined by using the site's shear wave velocity (V s30) and the peak ground acceleration (PGA) of the GMRS and comparing them to the values within Reference

[4], Table 3-1. Based on the PGA of 0.241 g and the shear wave velocity of I 369ft/s , the site soil class is A-Intermediate. Once a s ite soil class is determined , the mean vertical v s. horizontal GMRS ratios (V/H) at each frequenc y are determined by using the site soil class and its associated V/H values in Reference

[4], Table 3-2. The vertical GMRS i s then calculated by multiplying the mean V/H ratio at each frequency by the horizontal GMRS acceleration at the corresponding frequency.

It i s noted that Reference

[4], Table 3-2 values are constant between 0.1 Hz and l 5Hz. The V/H ratios and VGMRS values are provided in Table 3-2 of this report. Figure 3-1 below provides a plot of the horizontal GMRS , V/H ratios , and vertical GMRS for Cooper Nuclear Station.

50.54(f) NTTF 2.1 Seismic High Frequency Confirmation 16C4384-R PT-005 Rev. 0 Page 20 of49 Table 3-2: Horizontal and Vertical Ground Motions Response Spectra Frequency HG MRS V/H VG MRS (Hz) (!!) Ratio (!!) 1 00 0.241 0.78 0.188 90 0.242 0.82 0.198 80 0.245 0.86 0.211 70 0.2 49 0.91 0.227 60 0.258 0.93 0.240 50 0.282 0.95 0.268 40 0.321 0.91 0.292 35 0.342 0.86 0.294 30 0.359 0.79 0.284 25 0.386 0.72 0.278 20 0.417 0.67 0.279 15 0.463 0.67 0.310 12.5 0.486 0.67 0.326 10 0.465 0.67 0.312 9 0.449 0.67 0.301 8 0.430 0.67 0.288 7 0.417 0.67 0.279 6 0.422 0.67 0.283 5 0.454 0.67 0.304 4 0.415 0.67 0.278 3.5 0.364 0.67 0.244 3 0.294 0.67 0.19 7 2.5 0.209 0.67 0.14 0 2 0.162 0.67 0.10 9 1.5 0.116 0.6 7 0.078 1.25 0.096 0.67 0.064 I 0.082 0.67 0.0 55 0.9 0.076 0.67 0.051 0.8 0.069 0.67 0.046 0.7 0.063 0.67 0.042 0.6 0.060 0.67 0.040 0.5 0.055 0.67 0.037 0.4 0.044 0.67 0.030 0.35 0.039 0.67 0.026 0.3 0.033 0.67 0.022 0.25 0.028 0.67 0.019 0.2 0.022 0.67 0.015 0.15 0.0 17 0.67 0.011 0.1 25 0.0 14 0.67 0.009 0.1 0.011 0.67 0.007 SA 50.54(t) NTTF 2.1 Seismic High Frequency Confirmation 1 6 C4384-RPT-00 5 Rev. 0 Page 21o f 49 0.60 -HGMRS -VGMRS 0.50 ---V/H Ratio (A-Intermediate) 0.40 ,, I ' I ' 1.0 0 ,' \ 0.9 0 I \ I \ , \ I \ I \ \ 0.80

• 0 +; c 0 :Q cu 0.30 cu a::: .... Q) o; 0 0.20 *------------------

0.10 0.00 0.1 _ _.._ ... ___ .--... ., 10 Fre uen cy Hz I I I 0.70 > 0.60 0.50 100 Figure 3-1: Plot of the Horizontal and Vertical Ground Motions Response Spectra and V/H Ratios 3.3 Component Horizontal Seismic Demand Per Reference

[4] the peak horizontal acceleration is amplified using the following two factors to determine the horizontal in-cabinet response spectrum:

• Horizontal in-structure amplification factor AFsH to account for seismic amplification at floor elevations above the host building's foundation

• Horizontal in-cabinet amplification factor AF c to account for seismic amplification within the host equipment (cabinet , switchgear , motor control center , etc.) The in-structure amplification factor AfsH is derived from Figure 4-3 in Reference

[4]. The cabinet amplification factor, AF c is associated with a given type of cabinet construction.

The three general cabinet types are identified in Reference

[4] and Appendix I of EPRI NP-7148 [73] assuming 5% in-cabinet response spectrum damping. EPRI NP-7148 [73] classified the cabinet types as high amplification structures such as switchgear panels and other similar large flexible panels, medium amplification structures such as control panels and control room benchboard panels and low amplification structures such as motor control centers.

SA 50.54(t) NTIF 2.1 Seismic High Frequency Confirmation 1 6 C4384-RPT-0 05 Rev. 0 Pag e 22 o f 49 All of the electrical cabinets containing the components subject to high frequency confirmation (see Table B-l in Appendix B) can be categorized into one of the in-cabinet amplification categories in Reference

[4] as follows:

• Typical motor control center cabinets consisting of a lineup of several interconnected sections.

Each section is a relatively narrow 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 qualify as low amplification cabinets.

• Switchgear cabinets EE-SWGR-4160F , EE-SWGR-4160G , EE-SWGR-480F, SWGR-480G, EE-SWGR-4160DGI and EE-SWGR-4160DG2 are large cabinets consisting of a lineup of several interconnected sections typical of the high amplification cabinet category.

Each section is a wide box-type structure with height-to-depth ratios of about 1.5 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.

• Control cabinets DG-PNL-DG I ECP, DG-PNL-DG2 ECP , EE-CHG-125 l A , EE-CHG-125 lB, EE-CHG-250 IA, EE-CHG-250 IB, LRP-PNL-25-58, LRP-PNL-9-30 and PNL-9-3 l are in a lineup of several interconnected sections with moderate width. Each section consists of structures with height-to-depth ratios of about 3 which results in moderate frame stresses and damping. The response levels are mid-range between MCCs and switchgear and therefore these cabinets can be considered in the medium amplification category. 3.4 Component Vertical Seismic Demand The component vertical demand is determined using the peak acceleration of the VGMRS between 15 Hz and 40 Hz and amplifying it using the following two factors:

• Vertical in-structure amplification factor Af s v to account for seismic amplification at floor elevations above the host building's foundation

• Vertical in-cabinet amplification factor Af c to account for seismic amplification within the host equipment (cabinet, switchgear , motor control center , etc.) The in-structure amplification factor AF s v is derived from Figure 4-4 in Reference

[4]. The cabinet amplification factor, AFc is derived in Reference

[4] and is 4.7 for all cabinet types.

SA 50.54 (f) N TIF 2.1 Seismic High Frequenc y Confirm a ti o n 4 CONTACT DEVICE EVALUATIONS 1 6C 4 3 84-RP T-00 5 Re v. 0 Pag e 2 3 of 49 Per Reference

[4], seismic capacit i e s (the highest seismic test level reached by the contact device without chatter or other malfunction) for each subject contact device are determined by the following procedures: (I) If a contact device was tested as part of the EPRI High Frequency Testing program [74], then the component seismic capacity from this program is used. (2) If a contact device was not tested as part of [74J , 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, manufacturer

/vendor , or from the SQURTS testing program). (b) Generic Equipment Ruggedness Spectra (GERS) capacities per [75] and [76]. (c) Assembly (e.g. 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 3 using the criteria in Section 4.5 of Reference

[4]. The high-frequency evaluations as described above were performed in Ref. [77]. Where applicable , operator actions that are included in existing station procedures

[78] are used to resolve functional failures of contact devices that impact the operation of essential plant components.

A summary of the high-frequency evaluation conclusions is provided in Table B-1 in Appendix B.

SA 50.54(t) NTTF 2.1 Seismic High Frequency Confirmation 5 CONCLUSIONS 5.1 Gene r al Conclusions 1 6C 4 3 84-RPT-00 5 Re v. 0 P ag e 24 o f 4 9 Cooper Nuclear Station has performed a High Frequency Confirmation evaluation in response to the NRC's 50.54(f) letter [I] using the methods in EPRI report 3002004396

[4]. The evaluation identified a total of 136 components that required evaluation.

As summarized in Table 8-1 in Appendix 8, 89 of the devices have adequate seismic capacity, two (2) have existing plant procedures to cope with the effect of contact chatter, and 45 components required resolution following the criteria in Section 4.6 of Reference

[4]. To improve plant safety , Cooper Nuclear Station intends to address equipment sensitive to high frequency ground motion for the reevaluated seismic hazard information through mitigation strategies in lieu of a separate resolution of the 45 components identified under the letter [I] which do not impact the credited path for mitigation strategies.

5.2 Identification

of Follow-Up Actions Based on the general conclusions above , no follow-up actions are necessary.

SA 50.54(f) NTTF 2. I Seismic High Frequency Confirmation 6 REFERENCES 16C4384-RPT-005 Rev. 0 Page 25 of 49 [I] NRC (E. Leeds and M. Johnson) Letter to All Power Reactor Licensees et al., "Request for information Pursuant to Title 10 of the Code of Federal Regulations 50.54(t) Regarding Recommendations 2.1, 2.3 and 9.3 ofthe Near-Term Task Force Review of Insights from the Fukushima Dai-Ichi Accident," ADAMS Accession Number MLl2053A340, March 12 , 2012. [2] NRC Report, "Recommendations for Enhancing Reactor Safety in the 21st Century," ADAMS Accession Number MLl I I 861807 , July 12 , 2011. [3] EPRI Report 1025287, "Seismic Evaluation Guidence:

Screening, Prioritization, and lmplimentation Details (SPID) for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1: Seismic ," Final Report , February 2013. [4] EPRI Report 3002004396, "High Frequency Program: Application Guidance for Functional Confirmation and Fragility Evaluation," Final Report , July 2015. [5] NRC (J. Davis) Letter to Nuclear Energy lnstitute (A. Mauer), "Endorsement of Electric Power Research Institute Final Draft Report 3002004396, 'High Frequency Program: Application Guidance for Functional Confirmation and Fragility.'," ADAMS Accession Number ML I 52 l 8A569 , September 17 , 2015. [6] NRC (W. Dean) Letter to the Power Reactor Licensees on the Enclosed List, "Final Determination of Licensee Seismic Probabilistic Risk Assessments Under the Request for Information Pursuant to Title I 0 of the Code of Federal Regulations 50.54(t) Regarding Recommendation

2. I 'Seis mic' of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident," ADAMS Accession Number ML 15I94AO15, October 27, 2015. [7] NPPD Letter (NLS2014027) to NRC, "Nebraska Public Power District's Seismic Hazard and Screening Report (CEUS Sites) -Re sponse to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation

2. I of the Near-Term Task Force Review of insights from the Fukushima Dai-ichi Accident," ADAMS Accession Number MLl4094A040, March 31, 2014. [8] NPPD Letter ( LS2015017) to NRC, "Revision to Nebraska Public Power District's Response to Nuclear Regulatory Commission Request for Information Pursuant to I OCFR 50.54(t) Regarding the Seismic Aspects of Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident ," ADAMS Accession Number ML!5050A 165 , February l I , 2015.

SA 50.54 (f) NTTF 2.1 Seismic High Frequency Confirmation l 6C 4384-RPT-00 5 Re v. 0 Page 26 of 49 [9] NRC (F. Vega) Letter to NPPD (0. Limpias), "Cooper Nuclear Station -Staff Assessment of Information Provided Pursuant to Title I 0 of the Code of Federal Regulations Part 50, Section 50.54(f), Seismic Hazard Reevaluations for Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident ," ADAMS Accession Number ML I 5240A030 , September 8, 2015. [10] Cooper Nuclear Station Document 791 E266 Sheet 10 Rev. 14/AA , Elementary Diagram, Primary Containment Isolation System. [ 11] Cooper Nuclear Station Document 791 E266 Sheet 11 Rev. 13/ AB , Elementary Diagram , Primary Containment Isolation System. [12] Cooper Nuclear Station Document 791 E266 Sheet 5 Rev. 15/AA, Elementary Diagram, Primary Containment Isolation System. [ 13] Cooper Nuclear Station Document 791 E266 Sheet 6 Rev. 16/ AA, Elementary Diagram , Primmy Containment Isolation System. [14] Cooper Nuclear Station Document 791 E266 Sheet 7 Rev. 30/AA, Elementary Diagram , Primary Containment Isolation System. [ 15] Cooper Nuclear Station Document 791 E266 Sheet 8 Rev. 09 1 AA, Elementary Diagram , Primary Containment Isolation System. [ 16] Cooper Nuclear Station Document 791 E266 Sheet 14 Rev. 04/ AA, Elementary Diagram , Primary Containment Isolation System. [ 17] Cooper Nuclear Station Document 791 E253 Sheet I Rev. 30/ AA, Elementary Diagram , Automatic Blowdown System. [ 18] Cooper Nuclear Station Document 791 E253 Sheet 2 Rev. 28/ AA, Elementary Dia gram, Automatic Blowdown System. [19] Cooper Nuclear Station Document 791E253 Sheet 3 Rev. 12/AA , Elementary Diagram , Automatic Bl owdown System. (20] Cooper Nuclear Station Document 791 E26 l Sheet 5 Rev. 23/ AA , Elementary Diagram , Residual Heat Removal System. (21] Cooper Nuclear Station Document 791 E26 l Sheet 8 Rev. 23/ AA, Elementary Diagram , Residual Heat Removal System.

SA 50.54(t) N}TF 2.1 Seismic High Frequency Confirmation 16C4384-RPT-005 Rev. 0 Page 27 of 49 (22] Cooper Nuclear Station Document 791E265 Sheet 2 Rev. 23/AA, Elementary Diagram, Core Spray System. (23] Cooper Nuclear Station Document 944E689 Sheet I Rev. 13/AA, Elementary Diagram, Low-Low Set. (24] Cooper Nuclear Station Document 791E271Sheet7 Rev. 25/AA, Elementary Diagram , High Pressure Core Injection System. [25] Cooper Nuclear Station Document 791 E271 Sheet 3 Rev. 23/ AA, Elementary Diagram, High Pressure Core Injection System. [26] Cooper Nuclear Station Document 791 E27 I Sheet 4 Rev. 24/ !\A, Elementary Diagram , High Pressure Core Injection System. [27] Cooper Nuclear Station Document 791 E266 Sheet 12 Rev. 19/ AC , Elementary Diagram, Primary Containment Isolation System. [28] Cooper Nuclear Station Document 791 E266 Sheet 13 Rev. 25/ AC, Elementary Diagram, Primary Containment Isolation System. [29] Cooper Nuclear Station Document l04R907BB Rev. 06/AA , "P&ID , Control Rod Drive Hydraulic System". l 30 J Cooper Nuclear Station Document 791 E264 Sheet 2 Rev. 28/ AA, Elementary Diagram, Reactor Core Isolation Cooling System. [31] Cooper Nuclear Station Surveillance Procedure

6. l RCIC.30 I Rev. I 0 , "RCIC Steam Line High Flow Channel Caibration (Division l )". [32] Cooper Nuclear Station Document 791 E264 Sheet 3 Rev. 21/AA, Elementary Diagram. Reactor Core Isolation Cooling System. [33] Cooper Nuclear Station Document 2041 Rev. 87/AA, Flow Diagram , Reactor Building Main Steam System. [34] Cooper Nuclear Station Document 791 £264 Sheet 7 Rev. 15/AA, Elementary Diagram , Reactor Core Isolation Cooling System. (35] Cooper Nuclear Station Document 2043 Rev. 56/AC , Flow Diagram , Reactor Core Isolation Coolant and Reactor Feed Systems.

SA 50.54(!) NTTF 2.1 Seismic High Frequency Confirmation 1 6C43 84-RPT-0 05 Re v. 0 Page 28 o f49 l36J Cooper Nuclear Station Document 791 E264 Sheet 6 Rev. 13/AA , Elementary Diagram , Reactor Core Isolation Cooling System. [37] Cooper Nuclear Station, "Updated Safety Analysis Report ," List of Effective Pages XXVII 5. (38] Cooper Nuclear Station Document 3002 Sheet I Rev. 52/AE , Auxiliary One line Diagram , Motor Control Center Z , Switchgear Bus I A , I B, IE. and Critical Switchgear Bus IF , 1 G. [39] Cooper Nuclear Station Document 3058 Rev. 66/AI, DC One Line Diagram. [40] Cooper Nuclear Station Document 3024 Sheet 8 Rev. 35/AE, Elementary Diagrams, 4160 V Switchgear. [41] Cooper Nuclear Station Document 14EK-0144 Rev. 23/AA , Schematic Diagram , Diesel Engine Generator.

[ 42] Cooper Nuclear Station Document 3020 Sheet 4 Rev. 20/ AA , Elementary Diagrams , 4/60V Switchgear. [43] Cooper Nuclear Station Document G5-262-743 Sheet I Rev. 26/AA , Electrical Schematic, Emergency Diesel Generator#!. [44] Cooper Nuclear Station Document G5-262-743 Sheet IA Rev. 12/AD , Electrical Schematic, Emergency Diesel Generator

1. 1. [45] Cooper Nuclear Station Document 2077 Rev. 78/AA , Flow Diagram , Dies e l Generator Building Service Water, Starting Air , Fuel Oil. Sump System, and Roof Drains. [46] Cooper Nuclear Station Document KSV96-3 Rev. 06/AA , Piping Schematic , Air Intake and Exhaust. [ 47] Cooper Nuclear Station Document KSV 46-5 Rev. 26/ AB , Piping Schematic , Lube Oil. [48] Cooper Nuclear Station Document 3040 Sheet 9 Rev. 38/AK, Control Elementary Diagrams. (49] Cooper Nuclear Station Document 3045 Sheet 14 Rev. 50/AB, C ontrol Elementary Diagrams.

[50] Cooper Nuclear Station Document G5-262-743 Sheet l OA Rev. 06/ AD, Electrical Schematic , Emergency Diesel Generator

1. 2.

SA 50.54(t) NTTF 2.1 Seismic High Frequency Confirmation 1 6 C4384-RPT-0 05 Rev. 0 Pagt: 29 o f 49 [51] Cooper Nuclear Station Document G5-262-743 Sheet 2 Rev. 20/ AC , Electrical Schematic, Emergency Diesel Generator#!. [52] Cooper Nuclear Station Document GS-262-743 Sheet 11 Rev. 14/AC , Electrical Schematic, Em e rgency Diesel Generator

1. 2. [53] Cooper Nuclear Station Document KSV47-9NP Rev. 08/AJ , Piping Schematic, Jacket Water. [54] Cooper Nuclear Station Document 2006 Sheet I Rev. 90/AN , Flow Diagram, Circulating , Screen Wash , and Service Water Systems. [55] Cooper Nuclear Station Document 2006 Sheet 3 Rev. 55/AC , Flow Diagram , Circulating , Screen Wash , and Service Water Systems. [56] Cooper Nuclear Station Document KSV47-8 Rev. 27/AA, Piping Schematic , Diesel Generator I and 2 Cooling Water. [57] Cooper Nuclear Station Document 3022 Sheet 6 Rev. 49/AH, Elementary Diagrams , 4160 V Switchgear. [58] Cooper Nuclear Station Document 2024 Sheet 2 Rev. 38/AA, Flow Diagram , HVAC Miscellaneous Service Buildings. [59] Cooper Nuclear Station Document 3065 Sheet 17 Rev. 47/AB, Control Elementary Diagrams. (60] Cooper Nuclear Station Document 3065 Sheet 17A Rev. 12/AB, Control Elementary Diagrams. [61] Cooper Nuclear Station Document INV-3C-70048 Sheet 2 Rev. 02/AA , Schematic Diagram , ARRI 30K200F. [62] Cooper Nuclear Station Document INV-4C-01410 Sheet 2 Rev. 02/AA, Schematic Diagram , ARR260K200F.

[63] Cooper Nuclear Station Document VM-0228 Rev. 19, Vendor Manual , Batteries and Chargers.

[64] Cooper Nuclear Station Document MBC-2920 Sheet Bl Rev. 00/AA, Schematic , High Voltage Shutdown.

SA 50.54(f) N:11F 2.1 Seismic High Frequency Confinnat1on 1 6C 4384-RPT-005 Rev. 0 Page 30 of 49 [65] Cooper Nuclear Station Document 20-100287 Sheet I Rev. 0 I/AA, Schematic , JOkVA Inverter 210-280 VDC 120 1 240 VAC 3-Wire 60Hz. [66] Cooper Nuclear Station Document 20-100288 Sheet I Rev. 00/AA, Schematic, JOkVA Static Switch 2-Pole 120 1 240 VAC 1-Pha se 60Hz. [67] Cooper Nuclear Station Document 3025 Sheet 9 Rev. 29/AH , Elementary Diagrams , 4160V Switchgear.

[68] Cooper Nuclear Station Document 3004 Sheet 3 Rev. 22/AA, Auxiliary One line Diagram , Motor Control Ce nters C , D , H, .!, DG!, DG2. [69] Cooper Nuclear Station Document 3006 Sheet 5 Rev. 84/ AG , Auxiliary One Line Diagram , Starter Racks lZ and TZ, Motor Control Centers K , l, LX, RA , RX, S , T, TX, X [70] Cooper Nuclear Station Document 3010 Sheet I Rev. 82/AH , Vital On e line Diagram. [71] Cooper Nuclear Station Document 30 I 0 Sheet 2 Rev. l 0/ AE, load and Fuse Schedule , Critical Distribution Panel CDP/A. [72] Stevenson

& Associates Report 16C4384-RPT-001, Rev. 2, "Selection of Relays and Switches for High Frequency Seismic Evaluation".

[73] EPRI Report NP-7148-SL, "Procedure for Evaluating Nuclear Power Plant Relay Seismic Functionality," Final Report December 1990. [74] EPRI Report 3002002997, "High Frequency Program: High Frequency Testing Summary," Final Report , September 2014. [75] EPRf Report NP-7147-SL, "Seismic Ruggedness of Rela ys," Final Report August 1991. [76] SQUG Advisory 2004-02, "Relay GERS Corrections," September 7, 2004. [77J Stevenson

& Associates Calculation l6C4384-CAL-OOI , Rev. 0 , "High Frequency Functional Confinnation and Fragility Evaluation of Relays". [78] Cooper Nuclear Station Emergency Procedure 5.8. l Rev. 27 , "RPV Pressure Control Systems".

[79] Cooper Nuclear Station Document 2028 Rev. 52/AA, Flow Diagram , Reactor Building and Drywell Equipment Drain System.

SA 50.54(t) NTTF 2.1 Seismic High Frequency Confirmation 16C4384-RPT-005 Rev. 0 Page 31of49 (80] Cooper Nuclear Station Document 2040 Sheet 1 Rev. 82/AA, Flow Diagram. Residual Heat Removal System. [81] Cooper Nuclear Station Document 2040 Sheet 2 Rev. 19/ AB, Flow Diagram. Residual Heat Removal System Loop B. [82] Cooper Nuclear Station Document 2042 Sheet I Rev. 35/AA, Flow Diagram, Reactor Water Clean-Up System. [83] Cooper Nuclear Station Document 2039 Rev. 61/AD, Flow Diagram , Control Rod Drive Hydraulic System. (84] Cooper Nuclear Station Document 2045 Sheet I Rev. 58/AA, Flow Diagram, Core Spray System. (85] Cooper Nuclear Station Document 2045 Sheet 2 Rev. 21/AA, Flow Diagram, Standby Liquid Control System. [86] Cooper Nuclear Station Document 2044 Rev. 74/AB, Flow Diagram , High Pressure Coolant Injection and Reactor Feed Systems.

SA 50.54(f) NTTF 2.1 Seismic High Frequency Confirmation 1 6C 4 3 84-RP T-OOS R ev. 0 Pa ge 32 o f4 9 A. REPRESENTATIVE SAMPLE COMPONENT EVALUATIONS A detailed example analysis of two components i s provided within thi s section. Thi s example is intended to illustrate each step of the high frequency analysi s methodology given in Secti on 4 of Ref. [ 4]. A.1 High Frequency Seismic Demand Calculate the high-frequency seismic demand on the components per the methodology from Reference

[ 4]. Sample calculations for the high-frequency seismic demand of components DG-LMS-DG 1 630SDL contained in control cabinet DG-PNL-DG I ECP and located in the Diesel Generator Building at elevation 903' and EE-REL-I FA 86 contained in switchgear EE-SWGR-4 l 60F and located in the Reactor Building at elevation 932'. Ref. (77] calculates the high-frequency seismic demand for all the subject component s. A.1.1 Horizontal Seismic Demand The horizontal site-specific CNS GMRS data can be found in Section 6 of Ref. (77]. Determine the peak acceleration of the horizontal GMRS between 15 Hz and 40 Hz: Peak Acceleration of Horizontal GMRS between 15 Hz and 40 Hz (see Table 6.2 of Ref. [77]): SA a MRS = 0.463g (at 15 Hz) Work the distance between the component floor and foundation with Ref. [4], Fig. 4-3 to calculate the horizontal in-structure amplification factor: Bottom of Deepest Foundation Elevation:

EL ro und = 903 ft Diesel Generator Building EL rou nd == 860 ft Reactor Building Component Floor Elevation:

E L co mp = 903 ft DG-LMS-DGJ 630SDL Distance Between Component Floor and Foundation Elevation: E L c om p = 932 ft EE-REL-I FA 8 6 h co m p == EL comp -E L ro u n d == 0 ft DG-LMS-DGJ 630SDL h co mp = EL comp -ELroun d = 72 ft EE-REL-I FA 86 SA 50.54(t) NTTF 2.1 Seismic High Frequenc y Con ti nnation 16C 4 3 8 4-RP T-0 0 5 R ev. 0 Pag t! 33 o f49 Calculate the horiz o ntal in-structure amplification factor based on the distance between the bottom of the foundation elevation and the subject floor elevation: Slope of Amplification Factor Line , Oft < h c o mp < 40ft: Intercept of Amplification Factor Line with Amplification Factor Axis: Horizontal In-Structure Amplification Factor (Ref. [4], p.4-1 l): ffi h = 2.1-1.2 = 0.0225 2_ 40fc-Oft f t AF s H(h comp) = (m h

• hc om p+ b h) ifhcom p <= 40ft 2.1 otherwise AF::rn(h com p) = 1.2 DG-L MS-D G! 630SDL AF s1-1 (h co m p) = 2.1 EE-REL-JFA 8 6 Calculate the horizontal in-cabinet amplification factor based on the type of cabinet that contains the subject component:

Type of Cabinet: cab I ="Control Cabinet for DG-LMS-DGJ 630SDL" (enter "MCC", " Switchgear" , "Control Cabinet" , or "Rigid") cab2 ="Switchgear for EE-REL-I FA 8 6" Horizontal In-Cabinet Amplification Factor (Ref [4], p. 4-13): AF c.h (cab) = 3.6 if cab= " MCC" 7.2 if cab= "Swit c hgear" 4.5 if cab= "Control Cabinet" 1.0 i f cab= " Rigid" AF c h (cabl) =4.5 AF c h (cab2) =7.2 Multiply the peak h o rizontal GMRS acceleration by t he horizontal in-structure and in-cabinet amplification factors t o determine the in-cabinet response spectrum demand o n the components

Horizontal In-Cabinet Response Spectrum: lCRS c h = Af s 1-1

• AF c h

• SA o MR S ICRS c.h =1.2*4.5*0.463=2.Sg DG-L MS-DGI 630 S DL lCRS c h =2.1 *7.2*0.463=7g EE-RE L-I FA 8 6 SA 50.54 (f) NTTF 2.1 Seismic High Frequency Confirmation 1 6C 4 3 84-RPT-005 R ev. 0 Pa g e 34 o f 4 9 A.1.2 Vertical Seismic Demand Determine the peak acceleration of the hor i zontal GMRS between 15 Hz and 40 Hz: Peak Acceleration of Horizontal GMRS between 15 Hz and 40 Hz (see Table 6.2 of Ref. [77]): SA a M RS = 0.463g (at 15 Hz) Obtain the peak ground acceleration (PGA) of the horizontal GMRS (See Table 6.2 of Ref. (77]): Peak Ground Acceleration (GMRS): PGAaMR S = 0.241 g Calculate the shear wave velocity traveling from a depth of30m (98.4 ft) to the surface of the site (Vs30) from Ref. [4]: Shear Wave Velocity: V _ (30m) s3 o --l: (-,--d i.) V st where, d i: Thickness of the layer (ft), V s;: Shear wave velocity of the layer (ft/s) Per Table 6.1 of Ref. [77], the sum of thickness of each layer over shear wave velocity of each layer is 0.0719 sec. The shear wave velocity is calculated as: Shear Wave Velocity:

V sJO = 98.4ft I 0.07 l 9sec = 1369 ft/sec Work the PGA and shear wave velocity with Ref. [4], Table 3-1 to determine the soil class of the site. Based on the PGA of 0.241 g and shear wave velocity of 1369 ft/sec at C S , the site soil class is A-Intermediate.

Work the site soil class with Ref. [4], Table 3-2 to determine the mean vertical vs. horizontal GMRS ratios (V/H) at each spectral frequency.

Multiply the V/H ratio at each frequency between 15Hz and 40Hz by the corresponding horizontal GMRS acceleration at each frequency to calculate the vertical GMRS. Table 6.2 of Ref. [77] calculates the vertical GMRS (equal to (V/H) x horizontal GMRS).

SA 50.54(f) NTTF 2.1 Seismic High Frequency Confinnation 1 6C 4 3 8 4-RP T-00 5 Re v. 0 Page 3 5 o f 49 Detennine the peak acceleration of the vertical GMRS (SAV G MRs) between frequencies of JSHz and 40Hz: V/H Ratio at ISI-Iz (See Table 6.2 of Ref. [77]): Horizontal GMRS at Frequency of Peak Vertical GMRS (at I SHz) (See Table 6.2 of Ref. [77]): Peak Acceleration of Vertical GMRS between 15 Hz and 40 Hz: VH=0.67 HGMRS = 0.463g SA v GMRS = VH

• HGMRS = 0.67*0.463=0.3 I Og (at 15 Hz) Work the distance between the component floor and foundation with Ref. [4], Fig. 4-4 to calculate the vertical in-structure amplification factor: Distance Between Component Floor and Foundation Elevation:

h c omp = ELco m p -EL ro u n d = 0 ft DG-LMS-DGJ 630SDL hcomp = EL co mp -EL ro und = 72 ft EE-REL-JFA 86 Calculate the vertical in-structure amplification factor based on the distance between the plant foundation elevation and the subject floor elevation:

Slope of Amplification Factor Line: Intercept of Amplification Factor Line 2.7-1.0 m v= = lOOft-O f t. 0.017 2:.. ft with Amplification Factor Axis: b v = 1.0 Vertical In-Structure Amplification Factor: AFsv(hcomp)

= mv

• h c omp + bv AF s v(hc o mp) = l.O DG-LMS-DGI 630SDL AF s v(h co m p) = 2.224 EE-REL-I FA 86 Per Ref. [4] the vertical in-cabinet amplification factor is 4.7 regardless of cabinet type: Vertical In-Cabinet Amplification Factor: A Fe v =4.7 SA 50.54(t) NTTF 2.1 Seismic High Frequency Confirmation 1 6C 4384-RPT-0 0 5 Re v. 0 Pag e 36 o f49 Multiply the peak vertical GMRS acceleration by the vertical in-structure and in-cabinet amplification factors to determine the in-cabinet response spectrum demand on the component:

Vertical In-Cabinet Response Spectrum (Ref. [4], p. 4-12 , Eq. 4-lb): ICRS c.v = AFsv

• AF cv

• SAVGMR S ICRS c.v =1.000*4.

7*0.31=1.458g DG-LMS-DGI 630SDL ICRS c v =2.224*4.7*0.3 l=3.243g EE-REL-1 FA 86 A.2 High Frequency Capacity A sample calculation for the high-frequency seismic capacity of components DG-LMS-DG 1 630SDL (contained in DG-P L-DGl ECP) and EE-REL-lFA 86 (contained in EE-SWGR-4160F) is presented here. A.2.1 Seismic Test Capacity The high frequency seismic capacity of a component can be determined from the EPRI High Frequency Testing Program or other broad banded low frequency capacity data such as the Generic Equipment Ruggedness Spectra (GERS) or other qualification reports. The model for component DG-LMS-DG I 630SDL is a Namco Controls EA 180-32302 relay per Table I. I of Ref. [77) and was not tested as part of the high-frequency testing program. The seismic capacity was calculated in Table 9-1of16C4384-RPT-OOI (72] to be 9.52g per a low frequency qualification test. The model for component EE-REL-IF A 86 is a General Electric I 2HEA6 I relay per Table 1.1 of Ref. [77] and was tested as part of the high-frequency testing program. High Frequency capacity was determined to be 2 l.8g per l6C4384-RPT-001

[72J. A.2.2 Seismic Capacitv Knockdown Factor Determine the seismic capacity knockdown factor for the subject relay based on the type of testing used to determine the seismic capacity of the relay. Using Table 4-2 of Ref. [4], the knockdown factors are chosen as: Seismic Capacity Knockdown Factor: Fk = 1.2 Lowest Level Without Chatter DG-LMS-DG 1 630SDL Fk = 1.11 Test Table Capacity EE-REL-1 FA 86 A.2.3 Seismic Te ting Single-Axis Correction Factor Determine the seismic testing single-axis correction factor of the subject relay, which is based on whether the equipment housing to which the relay is mounted has well-separated horizontal and SA 50.54(t) NTIF 2. l Seismic High Frequency Confirmation 1 6 C 4384-RPT-00 5 Rev. 0 P a ge 3 7 o f4 9 vertical motion or not. Per Ref. [4], pp. 4-17 to 4-18 , relays mounted within cabinets that are braced, bolted together in a row, mounted to both floor and wall, etc. will have a correction factor of 1.00. Relays mounted within cabinets that are bolted only to the floor or otherwise not well-braced will have a correcti o n factor of 1.2. per Ref. [ 4 ], pp. 4-18. Single-Axis Correction Factor (Ref. [ 4], pp. 4-17 to 4-18 and Table 6.4 of Ref. [77]): F MS = 1.2 DG-LMS-DGJ 630SDL EE-REL-JFA 86 FM s =LO A.2.4 Effective Wide-Band Component Capacity Acceleration Calculate the effective wide-band component capacity acceleration per Ref. [4], Eq. 4-5: Effective Wide-Band Component Capacity Acceleration (Ref. [4], Eq. 4-5): A.2.5 Component Margin TRS = 9.52g DG-LMS-DGJ 630SDL TRS = l9.64g EE-REL-JFA 86 Calculate the high-frequency seismic margin for relays per Ref. [4J , Eq. 4-6: (A sample calculation for the high-frequency seismic demand ofrelay components DG I 630SDL and EE-REL-I FA 86 is presented here. A table that calculates the high-frequency seismic margin for all of the subject relays listed in Table 6.4 of Ref. [77].) Horizontal Seismic Margin TRS 3.81>1.0 , OK DG-LMS-DGJ 630SDL (Ref. [4], Eq. 4-6): = ICRSc. h 2.8 l> 1.0 , OK EE-REL-JFA 86 Vertical Seismic Margin TRS 6.53> 1.0, OK DG-LMS-DG 1 630SDL (Ref. [4], Eq. 4-6): = ICRSc. v 6.06> 1.0 , OK EE-REL-JFA 8 6 SA 50.54(1) NTTF 2.1 Seis mic Hi gh Frequency Co n firmation B. COMPONENTS IDENTIFI ED FOR HIGH FREQUENCY CONFIRMA TIO'.'J Table B-1: Components Identified for High Frequency Confirmation Co mponent l!: ndosurc No. *= Syste m :;> Devic e ID Type Manuracturer Model ID *rn.* Fu nction I I RCIC-PS*72A rrocess Turhine Exhaust High Barksdale D2H-Al50SS LRP*PNl.r25*

Control Swttch Pressure 58 Cab met 2 1 RCIC-PS-728 Process Turhme Exhaust High Barksdale D2H-A I SOSS LRP-PNL-25* Control Switch Pressure 58 Cabinet J 1 RCIC*PS*87A Process Reactor l>rcssurc Stauc-0-Rins 5N6-BB J-U8-LRP-PNL-25-Control Swirch C l A-TTNQ 58 Cabmel 4 I RCIC*PS-878 Process Reactor Pressure Staltc-0-Rmg 5N6*BB 3-U8-LRP-PNL-25*

Control Switch CI A-ITNQ 58 Cabinet 5 1 RCIC*l'S*87C Plocess Reactor PrCllisure Stahc-0-Rmg SN6-B B J-U8-LRP-PNL-25-Con tr ol SW11Ch CIA-TI'NQ 58 Cabmet 6 1 RCIC-PS-8ID Process Rea ctor Pressure 5N 6-BB3-U8-LRP-PNL*25-Control Switch C IA-ITNQ 58 Cabinet 7 I RCIC-REL*K IO Cocurol S1cam Lme Space Excess Ekctnc 12HGAllA52F LRP-PNL-9-Control Relay Temperature JO Cabinet 8 I RrIC'-REl.-Kl I Control Stmun Linc Space E xcess General Electric I 2J-IGA 11 AS2F LRP*rNL*9*

Control Relay Temperature 10 Cabinet 9 I RCIC-REL-K12 Control Steam Lrnc High Allen Brad l ey 700*RTC-LRP*P L-9-Control Relay D1fTeren11al Pressure ll l l O U I 30 Cabinet 10 I RCIC-REL-KlJ Coolrol Reactor Pressure Gt.-ncral Electric 1'.?llGA llAS2f I.RP-P i 1...-q-C o n1rol Relay 30 Cabi net II I RCIC*R E L-Kl S Control RCJC Auto bolatmn General Electric 12HFASIA4 2 F LRP*PNL-9-Co n trol Re l ay Signal 30 Cabinet 12 I RCJC *REL-K6 Control Turbme Exhaust H1gt'I General Electric I 2 HGA 1 1 l\52F LRP-PNL*9-Con1rol Relay Pressure 3 0 Cab met Control Pump Suction Low National LRJ*-l" NL-9-Cont rot 13 I RCIC-REl , K7

-<TS-812 Relay Pressure J O Cabinet 1 4 I RCIC*REL-K8 Control furb1nc Tnp General Electnc IZICFASIA42f LRP*PNL-9-Coot r o l Rela y 30 Cabinet 16C4384-RPT-005 Rov 0 Page 38 of 49 Floor Co mponen l i:Y*lu.ation BuiJdlnc Elev. Bas il for Evaluation (0) C*p ecity Result Rll 881 GERS Operntor Action RB 88 1 GERS Operator Ac11on Rll 88 1 Vendor Cap>Dem Report RB 881 Vendor Cap> Dein Report Rll 881 Vendor Cap>Dem Report RH 881 Ve ndo r Cap'> Dem Report CB 903 GERS C ap.,. U c m CB 903 GERS Cap> Dem rn 901 SQURTS Cap> Dern Report CB 903 GERS \ap>Oem c u 9 01 GERS C 1p>Oem CB 901 GERS Cap> Dem f'NS CB 9 03 C ap> Dem Repon Cl\ 90] GERS Cap'>Oem No. *;; ;;> Devic e ID Ty p e lj I RCIC*REL*KJO Con1rot Relay 16 I RCIC-REL-KJ l Con t rol Re lay 17 I KCIC-1<.1'.L-KJ2 Contro! Relay 18 1 KCIC*KEL-KJ3 Con trol Re lay 1 9 I RCIC-TS-nA Process Swi t ch 20 l R C l C-TS-79ll Proct!Ss Switch 21 I R CIC-TS-79C Process Swuc h 22 I R CIC-TS-790 Process 5W1tch Z3 I RC!C-TS-80 A Process SW1 t ch 24 I R C!C-TS-80 11 Process Switch 25 I RCIC-TS-80C Process SWTtcli 26 I RCIC-T S-800 P rocess SW1lCh 27 I RCIC-TS-81 A Process Switch 28 I RCIC-TS-810 Process Switch 29 I RCIC-TS-8IC Proc-ess Switch SA 5 0.54 (1) NTIF 2.1 S ei s m ic High Fre qu e n cy C onfirmati o n Tab l e B-1: Components Identified for High Frequenc y Confirmation Comp o nent Enclosur e S)'l'l trm Man"r ac tur e r Model ID TYJI* Fu*ction I.me Space Excess Generul Ele4:mc 1.l H GA 1 1 LRJ>-PNL*9-Concr"O I Temper.uure 33 Cuburnt Steam Lme Space Excess General Eloct n c 12HGAl 1.'\l2F LRP*PNl.-9-Control Temperature 3l Cabinet S team I.i nc H1ch Allen Bradley 700-RTC-LRP-PNL,-9

-Control DttTerential Pressure llllOUI JJ Cabinet RCIC Auto lsolatioo General Elccmc 12HFA1lli\2F LRP-PNL Contro l Signal J 3 Cn b met Steam Line Space P a t el I Fenwnl 01-170230-090 N/A (Loca l) R1g1d Skid Temperature Mounted Stc:sm Li ne Space E."'tcess Pa1el I Fenwal 0 1-170230-090 NIA (Local) R igt d S k id Temperature Mounted Slcam Line Space Excess Patel I Fc n wal 01-1 70230-090 N/A(loca l) R 1g 1d Skid Temperature.

Mounted Stea m Lm e Sp3ce Excess Pat el I Fe n wal 01-1 70230-0QO N I A (Local) K1g1d Skid Temperature Mounted Steam Line Space E'<ccss Patel I Fenwal 0 I -I 70230--090 NIA (Loca l) R1g 1 d Skid Temperature M o unkd Steam Line Sp.:u:e Excess Pate l I Fenwal 01-170:?]0-()0() !L ocal) R 191d kid Tem p erature M o un1ed Une S pare Excess / FemYal 01-170230...@0 N f A (Lcca l) R1g1d Skid Temperature Mounted StCllm Line Space E'<ccss Pate l I Fenwal U I -1 70230-090 N I A (L-Ocal)

Skid Temperature Mounl<Xi S tcu in Line Space Excess Pntol / Fe n wal 01-170210*090 NIA (Loca l) R1g1d Skid Temperature Mounted Stea m Line Space Excess Pa c el *' Fenwa l 0 1-170230-090 N/A (L-Ocal) R1gtd Skid Tempe r atu r e ).ioonled Steam Line Space E xcess l'aiel I fenwal 01 *1 702J0-090

'l l A (Loca l) R 1 g1d Sl..1d Te m perature Mounted 16C4384-RPT-005 Rev 0 Page 39 of 49 fi'lOGr Compont!nt Evalualion Bui.ldine, Elev. Basi1 for E valuati o n (ft) C apacity R t1 u.lt CB 903 GE RS Cap>Dem C B 90'.\ G ERS Cap>Dcm CB 903 SQURTS Cap >Dem Report CB 903 EP RI Hf Cap> Dem Test Rll 860 CNS Crtp>Dem Repo r1 RB 860 CNS Cap>Dcm Report RB 881 CN S Cap>Dem R t.11o rt RB 881 OIS Cap> Dern Rep ort RB 860 CNS Cap>Dem Roport RB 8 60 CNS Cnp>Dcm Report RB 881 CNS Cap> Dem Report RB 881 CNS Cap'> Dem Report RD 860 CNS Cap> Dem Repo rt RB 860 CNS Cap> Dem Report RB 881 CNS Cap'>Oem Rl.-port No. *= "'

ID Type JO I RCIC-TS-810 ProcesJ Swnch JI I RCIC-TS*82A Process SW\!Ch 32 I R C IC-TS-82 8 Process Swuch 33 I R Cf C-TS-8 2 C Process Swttch 34 I RCIC

• TS*82D Process Switch JS I DG-LMS-DGI l'roc<!"S 630SDL SW1 t ch 36 I DG-l.MS-DG1 Process 630SDR Switcn 37 I DG-R EL-DG I 140S Control Relay 38 I DG-REL-DGI Con trol l4 RX 3 Relay 39 I DG-REL*DGI Control 14 RYI Re lay 40 I 00-REl.-DG I 2 7-59 Protective K clay 41 I DG-REL*DGI 40 Protective Relay 42 I DG*REL-DGI Contro l 481SEX Rclny 43 I DG-REL-DG I Control 4E.'lllX Relay 4 4 I DG-REL*DGI Control 4EMX3 Relay SA 50.54 (!) NTIF 2.1 Sei s m ic lligh f requ e nc y Confirmation Table B-1: Compo nent s Identified for High Frequency Confirmation Co mp onent Enclo.ture Sys r em M*our*ctu r er Modd ID Type Fu oni o n Steam Une Space E xcess Patel I Fcnwal 0 1-1 70230-090 NIA (Local) R1g1d Skid Tempera1urc Mounted Steam U ne Sp 3 ce Excess I Fenwal 01*170230-090 N/A(Loca l) R1g1d Skid Temperat ur e Muunlud Stctm Line Space Excess Patel I Fcm\111 01-1 ;o230-090 N I A (l.ocal) R1g1d Skid Temperawre Mounted Sleu.m Line Spece Excess Patel I Fenwal 01-170:?3 0*090 N/A(Local)

Rigid Skrd Temperature Mounted Steiim Line Space Excess Pote! I Fenwal 01-170230-090 NIA (Loca l) R ig id S k id Temperature Mounted Over s paxl N a mco Con tr ols E,\ 18 0-32302 DG-PNL-Co n trol DGI ECP Cab i net Engine Overspued Namco Con t rols EA 1 8t>-Jll02 DG-PNL-Control DGI ECP Ca b inet Eng in e Overspeed P otter & KRPl4DG-125 DG-PNL-Contro l Brumfie ld DGI ECP Co.bmct Engine Rumung Pott er&. KRPI 4DG-125 DG-PNL-Control Brumfield OGI ECP Cabuu.-t Enij:ine Running Allen 700.R TC-DG-PNL-Control 11020UI DGI ECP Cabmcl Generator Abnormal Electric 1 2 1A V7JAIA DG-PN L-Co nt ro l Vo ltai,:: o OGT EC'P Cnhrnct Genera t or F i eld Failure General electnc 1 2CEHSIAIA UG-PNL-Control OGl ECP Cobmet E11gmc (ncomp l ctc St urt Potter& KRP I 4 DG-1 25 DG*P NL-Con tr o l Sequence Brumfield DGI ECP Ciibuu .. -t Emergency Engine Start Potter& KRJ' I 4 D G-I 25 Cont r ol Brumlidd DG I E C!' Cabinet Fmcrgency Eng111e S1J11t Potter& KRPl4DG-125 DG-PNL* Contro l Br umfi eld DGI ECI' Cab1m .. 't 16C4384-RPT-005 Rev 0 40 of49 Floo r Co mponent Ell alua1lon Buildin g Elev. B as is ror E*aluati o n (0) Ca pa('ity ReJult RB 881 CNS C1tp>r>cm Report RB RM CNS C a p>Dem Report RB 8 6 0 CNS Cap>Dem Report Rli 881 CNS Cap>Dem Report RU S81 CNS Cap> Dem Report OGI 903 CNS Cap> Dem Report Q(jJ 90) CNS Cap>Dem Report DGI 903 CNS C11p>Dcm Report DGI 903 CNS Cap>Dem Repon DG I 903 SQURfS Cap> Dem Report DGI 903 CNS \ap >Dem R epo rt DGI 9 03 CNS Cap>Oem Repon DGI 903 CNS Cap >Dem Report DGI 903 C:Dem R..:pon DG I 903 CNS Dem Repo rt No. .. ;;> Device lD Type 45 I DG-R El.-DGI S I i\ Prote<..11ve Relay 46 I DG-R F.l.*D G I SI B Pm l ec ll ve K cl.ay 47 I DG-RE L-DGI 51 C Protective Rel a y 48 ( DG-RE L*DG I Control 62Cl.X R elay 49 ( O G-R EL-O GI Con t ro l 63 0 S DX R ela.y so I O G-RE L-DG I 67 Protectiv e Relay 51 I DG*RE L*D GI 86 Con t ro l R elay l2 I DG*REL-DGI 87 A P r otective R elay l3 I DG*R.EL-D GI 87 B Pro t ecttvc R elay 54 ( O G-RE L-O GI 87 C Protective Re lay ll I DG-REL-DGI RI 04 Con t ro l R elay 56 I DG-K T-Jl42 Process Switch 57 I DG-LMS-OG2 P rocess 630 SOL Swi t c h l8 l O G-LM S-D G2 P rocess 630S OR SWltch l9 I DG-R E L-O G2 1 405 Control Relay 50.54 (t) N T H' 2.1 Sei s m ic H ig h Fre quenc y Confirmat io n Table B-1: Compo nent s Identified for High Frequency Confirmatio n Component

[nclosure Sy s tem M 1 nufactun:r Mod e l ID T y pe Funci i on Phase Overcurrent General Electric IFCVSIAD DG*PNL-Comrol DGI ECP Cab1nel Phase Overcu rr em General E lectr ic !FCVlli\D OG-PNL-Con1ro l DGI ErP Cabinet Phase Ove r c urr ent General Efectnc IFCV51AO DG*PN L-Co n lrol OG I EC P Cab1m .. "t Engine CranlC1 n g L1m1 t Agasta t Relay E70 I 2PDOO<l Na m"-o Controls EA I 80-3 I 302 DG*PNL-Con t rol DG2 ECP C abinet Eng i n e Potter& KRP141JG-125 DG-PNL-Control O r u m fie l d DG2f'C'P Cabine t I 6C4384-RPT-005 Rev 0 P a e 41 of 4 9 Ji'foor Component Evalu1'tion Bui l dine, f.lev. Basis for Evaluation (rt) C apacity R ts ull D G I 903 Ve n dor Cap> D<rn R e p o rt DGI 903 Vendo r Cap> Dem R eport OG I 903 Ven d o r Cap >Dem R e po1t OG I 903 EPlllHF Cap>Dem Test OG I 903 CNS Cap> Dem Repon DG I 'IOJ SQU RT S M 1 1 1 g:iuon R eport Stra l CJ?1es D GI 903 EPRI H F Cap>Dem Tes t DGI 903 CNS M 1 ug:uion Report S 1 ra t cc1es DGI 901 CNS Mmg 11 uon Report Stnu e gtcs DGI 903 CNS M1t1g 11.1ioo Report Stra tc gu:s DGI 903 GERS Cap> De m DGI 903 CNS Cap> Dem Rcpo n DG2 903 C NS Cnp> Dem Re p ort DG2 903 CNS Cap>Ocm Kcpo n DG2 903 C:'>S C1p>Oem R eport No. *;; :> Device ID Type 60 I DG-REL-DG2 Control 14RX3 Relay 61 I DG-REL-DG2 Control 14R YI Relay 62 I DG-REL-DG2 27-59 Protec t i ve Relay 63 I DG-REL-DGZ 40 Pn:Mth v e Relay 64 I DG-REL-DG2 Contra I 481SEX R e la y 65 I DG-REL-DG2 Co n t rol 4EMX K clay 66 I DG-R EL-DG2 Control 4 EMX3 K.cla y 67 I DG-REL-DG2 51 A P rotective Relay 68 I DG-REL-DG2 5 1 B Protcchvc Rela y 69 I DG-REL-DG2 SI C Protective R elay 70 I DG-REL-DG2 Cont ro l 62CLX R elay 7 1 I DG-REL-IJG2 Co nu ol 6JOSDX Rela y 72 I DG-RE L-DG2 67 Protective Relay 73 I DG-REL-D GZ 86 Conica l Relay 74 I DG-REL-DG2 87 A Prult.."C f1ve Ralay SA 50.54 (t) T ff 2.1 Sei s m ic H ig h Fre qu e nc y C onfirmation T able B-1: Co mponents Identified for High Frequency Co nfirmation Compoaent E nclosur e System l\tanurachlrer Model ID Type Function Engine Runn ing Potter& KRPl4DG-J25 DG-PNL-Control Brumfield DG2 ECP Ca binet Engine Runn r n g Allen Bradley 700-RTC-DC" PN L-Control I 1020UI DG2 EC P Cabine t Generator Abno m\a I Genera l Elect ri c 12 1A V7JAI A DG-PNL-Control Volt a cc D GZ ECP Cab met Generator Field Failure Guiera l Electnc 12Cf-H5JAIA DG-PNL-Co ntr o l Cabin et Engine Jncompk:tc Start Pouer& KRPl4DG-1 25 [>G.PNL-Control Sequence Brumfield D G2cC P Cab met Emergency Fngmc Stan P otte r& KRl'l4 D G-125 JJG-rNL-Co ntr ol B rumfi e ld DG2 EC I' Cabinet F.me*ce n cy Engine Suut Potter & KRP1 4DG-125 DG-PNL-Control Brum field DG Z EC P Caliim: t Phase Over current Gene ra l Electn c IFCVSIAD Control DG2 EC P Cabm el Pha s e Ovcrcurrcn t Genera l Electric IFCVSIAD D G-P L-Contro l DG2 ECP C abine t Phase Owrcurrent GenL'Tal E l ectnc !F CVSJ AD DG-PNL-Con tro l DG2 ECP Cabmel Engmc Crankmg Lcmu Thomas& Betts E7012PD004 DG-PNL-Control Timer DGZ EC P Cabin et R n gtne Ove r.speed Potter & KRP J 41JG-l 25 DG-P NL-Control Shutdown Brumfield DG2 ECP Cob met D1rec1ionnl Ovcr<: urrcnt General Electnc !CW-SIA D G-PNL-Control DG2 EC P Cnbmel Diesel Gi:ocrator Lockou t Gcntrnl Electnc 12JIEA61 DG-PNL-Control Re l ay DG2 ECP C11 b 1nct Generator D 1 ffercnt1JI General Electnc CF D-12 0 DG-P L-Con t rol D<.;2 cl'P C11bmet I 6C4384-RPT-005 Rev 0 Page 42 of49 Floor Co mp o o e n1 Evaluation Building [le v. B uis ror [v a lu ation (fl) C*pa e:ity R"ult DG2 903 CNS Rep o rt DG2 903 SQU R TS Cap>Dcm R eport DGZ 903 CNS Cap> Dem Report DG2 903 CNS Cap>Dem Report DG2 90J CNS Cap >Dem Report DG2 903 CNS C3p>Dem Rcpon DGZ 903 CNS Ctip>Dem Repon DG2 903 Vendor Cap >Dem Report IJ(j2 9 03 Vt..-ndor Cap> Dem Repo r t DG2 903 Vendor Cap> Dem Report DG2 903 EPRI HF Cap>Oem DG2 901 CNS Cap>Oem Report DG2 903 SQt;RTS Mitigation R e p ort S1ra1eg1es DG2 903 EPR! HF Cap">Dem T .. 1 DG2 903 CNS M1t1galton Reporl Slrategies N o. . ., ;;, Device ID Type 75 I DG-REL-DG2 87 B Pmtcct 1 ve Relay 76 I DG-R.EL-DG2 87 C Protective Kclay 77 I DG-REL-OG2 RIO<I Conlrol Relay 78 I DG-RT-3143 Process Swuch P rott.'Chve 79 I KJ Relay P rotect i ve 80 I KJ Protocttve 81 I KJ Relay Proloct1ve 82 I KJ Rehay 83 I EE-CB-4 I 60DGI MV Circ u it EGI Rre:aker 84 I EE*CB-4160DG2 MY Ci rcuit E02 85 I EE-CB-4 I 60 f I FE MY Circuit Bre.lker 86 I EE-CB-4 I 60F SS IF MVCircu1t Breaker 87 EE-C B-41 6-0P MY Circuit I SWPIA Brcu.kBr 88 1 EE-CB-4160F MVCircutl SWl 1 I C Breoker 89 I EE-REL-IPA 5 I A Protective Rtlay 5 0.54 (f) TIF 2.1 Se i s m ic Hi g h fr equ e nc y C o nfirmation Table B-1: C omponents Identified for High Frequency Confirmation Component Eodoimrt" Sy1fem Model ID 'fype Functi o n Generamr Oifferent111.I Gcricral E l eanc CFD-1 26 DG-PNL-Contro l DG2 ECP Cahmct Genera1or D11Tt:tt.'1ll 1 al Gmeral Eleclnc CFD-128 OG-PNL-Control DG2 ECP Cabinet Engine Speed Agastat Relay EGPBOO<I DG-r L-Contro l Co D02ECP Cabinet Oynalco Corp SST-2400AN-140 DG*PNL-Control DG2 ECP Cabmet C&D EE-CHG-125 Contro l Overvo li.age Shutdo\.Yfl Technolog1es ARRIJOICOOF IA Cabmct I nc C&D EE-CH0-125 Control Ovcrvoltagc Shutdown rechnolog1es ARRIJOK200F IB Cabmet In<: C&D EE-CHG-250 Conlrol O"ervoltage Shutdown Technoloc i es ARR260K100F IA Cabmcl Inc C&D EE-CHG-250 Conlrol Overvoltage Shutdown Technologies ARR260K200F I B Cabinet r .. DG Output Lockout General Elcctnc AMH-4 76°250-EE-SWGR-Sw1tchr,ear ID 4160DGI DG Output Lockout General Electric AMH-4 76-250-EE-SWOR-Switchgear ID 4160DG2 SW1td1gcar Feeder General Electnc Ai\ifH-4 76-250-EE-SWOR-SW1tchgear Lockout ID 4160F Station Service General Electric AMH-4 76-25 0-EE-SW GR-Switchgear Transfonner Lockout ID 4160F Service Wate r Pump S1emt:ns 5GEHU-Jl0*

EE-SW GR-SwitchKt:ar Lockout 1 200-78 41 6 0F Service Wa t er Pump S1en11..-ns lGEHU-350-EE-SWGR-Switchgear Lockout 1200-78 4160F Phase Overcurrent General l!IC1..-1:nc 12 1AC5J A EE-SWGR* Switchgear 4160F 16C4384-RPT-005 Rev 0 Page 43 of49 Floor Compo n enl Enluation Building Basi.1for 1£valuati o n (ft) Result DG2 '1()3 css M i tigation Report Strategics DG2 9 0 3 CNS M1t1gation Report Strategies DG2 903 GE RS Cap">Dem llG2 903 CNS C3p':>Dcm Rep ort CNS CB ')()] Kcport Cap>Dcm CNS CR 'I03 Rep on Cap> Dem CNS CB Q03 Report Cap >Dem CNS CB 903 Report Cap>Dcm DGI 903 Not MH1ga11on Av:11lablc S1rateg1es D02 Q03 Not Mit1gat10t1 Available Strategu:s RB 93::? Not M1t1gat1on Available Strategies RB 932 Not M1t1gallon Available Strc1tt:g1cs RB QJ2 CNS Cap> Dem Report RB 932 CNS Cap> Dem Repon KB 932 <11-: RS M111 gat1on Strategies No. .. ::> D evice ID Type 90 I EE-Kl!L-1 t" A 5 1 H Pru1cc t1v e Relay 91 I EE-R EL-I FA 51 C P 1otcct 1 ve Reloy 92 I EE-REL-I F A 86 Con trol Re l a y 93 I EE-REL-I FE 50-51 Pr ot< .. -ct1ve A Relay 94 I EE-REL-JFE 50-51 P roteci i ve B Relay 95 I EE-REL-I FE 5 0-5 1 Protective c Re lay 96 I EE-REL-I F E 86 Control Relay 97 I EE-REL-IF S 51 A P ro tective l\ela y 98 I EE-R EL-IFS 51 B Protective Rela y 99 I EE-REL-IFS 51 C Pr otcclrv e Re lay J OO I EE-REL-I FS 86 Co nl rol Relay JOI I EE-REL-SSIF 50-51 Pro tL"Ct i ve A Re l a y 10 2 I EE-REL-SS I F 50-51 1>rotcc t 1vc c Rela y 10 3 I EE-REL-SWP I A 50-Proteclt v e 50-l l A R elay 1 04 I EE-RE L-SWPI A 50-Protecuvc 50-51 c Rela y 50.5 4(f) NTTF 2.1 Sei s m ic High Fre quency Confirmat io n Table B-1: Components Identified for High l<'requcncy Confirmation Endo1un System Mao11facturer Model ID T y l"' Function Phas e Overcunent Genera I El cctnc t2J AC 53A EE-S W GR-Switchgelr 4160F P hase Overcurrcnt Gcm.-rml E lect ric 121AC53A EE-SWGR-4160F Normal Feed Lockout Gencrsl EJectn..;

12 H E A 61 EE-SWG R-Sw11chg;ear 4160P Phas e Ove r cu rr enl Gencrul Elccmc 121AC5313812A EE-SWGR-Switctlgmr 4160 F Pha se Overcurrent Gcnernl Elc.:;111c l 2 1A C53 BSI 2A E E-SWGR-Sw 1 1chgi:ar 41 60f Pha se Overcurren1 Generul Electric 12IAC5Jll812A EE-S WGR-Swttchucar 4160F Bus Lockout General Elecln c 12 HEA b l EE*SWGR-SW1tchgear 4 160 f Phase Overcurrent Genera l Eleclnc 1 2 fA C5JA EE-S WGR-Swllchgt:ar 41 60 f Phase Overcurrcnl General Electm: 121AC53A EE-S WGR-Switchgear 41 60 F Phase Overcurrent G4...-ner a l Electric 121AC53A EE-SWGR-Switchgear 41 60f Emergency Starrup Gene ral 1 2 Hl.:A 61 EE-S W G R-Switc h gear Transfom1er Feed Locko ut 4160f Phase Over c urrcnt Genera l Electric 121AC53 EE-SWGR-Sw11chgcar 4160F Phase Overcmn:nt General Eloctric 12 11\(53 EE-SW GR-Switchgeor 4160f Phase Ove1<.*1ment Genera l Electnc 1 21AC66 EE-SWG R-Swi t chgear 4160F Phase Overcurrent Ge n eral E l ectnc 12 1 AC66 EE-SW GR-Swn ch gear 4160F l6C4384-RPT-005 Rev 0 44 of 49 Floor Component Enluali t m BuiJdinc Elev. Batis fo r Evaluation (rr) Capacity R e1 ult RD 932 GE RS M i 11 ga t1 un Stra teg ies JUI 932 GE R S Mitigation S tr3.teg1es RB 932 EPRJ HF Cap>D em TC$! RB 932 SQURTS Mitignt1on Report S trat egics RB 9 3 2 SQURTS Mitigati o n Report RB 932 SQ UR TS Mill!l l U1 o n Report Slrategu:::s RB 932 H P Cap>Dem Test RJ3 932 GE R S M1tig.all on S 1ru1 eg1t:s RD 932 GERS M1t1gaoon StniH.-gies RB 9 32 GCKS M ihgallun Strategies RB 932 EPRJ HF Co.p>Dem Test RB 932 GERS Mit1g1111on Strateg ies RJ3 932 GERS Mi1iga11on Sm.1tcg1cs RB 9)2 EPRl HF M1 tig at 1on Test Strutcg 1 es RB 932 EPRJ HF M1t1ga t1on Test Stmtcg1es SA 16C4 384-RPT-005 Re v 0 50.54(f) N TT F 2.1 Se i sm ic H i gh Fr e que n c y Co n fi nn ation Page 45 of 49 Table B-1: Co mp o nent s Iden t ified fo r Hi g h F r e qu e nc y C onfirmati o n C umpun en t E ncl O!" ure lli'l oo r Co m ponent [v a lu a tion N o. *= S ys t e m Bu i ld in a: E l ev. B asis f o r Eva lu a t io n "' D evic e ID Type Man u fa c l\l r er M odel ID T y pe (fr) F un cti on Ca p acity R etulf 105 E E.REL-SWPI/\

86 Co ntrol Service Water L'ump General Elecmc lZll EA 61 EE-SWG R-Switchgear Ril 932 EPRJ HF Cap>Dem Rela y L<>c k o ul 4160 F Test 106 EE-REL-SWP I C 50-P ro tec 1ive OvcrcUrTen t G en eral Elecrnc 121AC66 EE-SWGR-Switchgear Ril 932 EPRJ HF Mitigation 50-51 A Rda y 4160F Test S1ra1cg 1 cs 107 EE-REL-SWP IC 50-Pr otc:ct i v c P ha.c;c Ovcrcurrent Ge n eral Hcclnc 121AC 6'i EErSWG R-Switchgear RB 932 E PRJ Hf "4 1t1ga t io n 50-51 c Relay 4160 f Tes t S tra tegies 1 08 EE-REl,SWPI C 86 Co ntrol Service W!Her Pump General Elec tric 1 2HEA61 EE-SW GR-SW'!tchgcar Ril 932 E PR I HF Cap>Dem Re1ay L ockout 4160F Test 109 F.F-C R-4 160G IGE MVCircuit Switchgear Feeder Genera l Elec tric AMH-4 7 6-250-EE-S W GR-Sw 1 1ch gcar RH 932 Not Mitiga1i on B r eaker Lock o u t ID 41 60G A vail able Strat egies 1 10 EE-CB-4 I 60G SS 1 G MV Ci rcuit talion Service General E l eclric l\)..tfl-4 76-250-EE-SWG R-Swt tc hgea.r RJl 932 N9t Brea ker T ro n sfo nne r Lockout 1D 4160G A va ilab l e Strateg i es Ill EE-CB-4 I 60G P...fV C1rcu1t Sta t io n Service Water General Elect r ic AMH*4 76-250* EE-S WGR-Sw1tctigear RH 932 Nol Mitigation SWPlB Oreaker Pump Lockout I D 4 1MG Av111 l ablc Strategics 112 EE-C A-4 I 60 G MVCirc:uit Sl.llt io n Service Water S i emens 5G Ell U-J50-EE-SWGR-Swiichgear Rll 932 CNS Cap>Dcm SWPI D Break.er Pump Lockoul 1 200-78 4160G Report Il l EE-REL-I GB 5 1 /\ Protective Phase Overcurre11f General Electric 121AC5JA F.E-SWG R-SW1tchgear RB 932 GERS Mit 1 gat1on Rela y 4160G Stralcgirs 11 4 EE-REL-I GB 5 1 B Pr otechve P ha s e Oven::urrent General Elecmc 12J AC 5JA EE-SWG R-Swttchgear RB 932 GE!lS M i l lK'!llOO Relay 4160G Stra tegies tt5 EE-RE L-I GB S 1 C' Proteclive Phase Overcurrnnl G1.!ncral E lccm c 12IACS3A EE-S W G R-Swtt c hgc.3.r "'" 932 liERS Miugauon Rela y 4 160G St ra tegies 116 EE-REL-I GB 86 Cont r o l 1:eed Lockout General 12HEA 61 EE-SWGR-Swttctlgear RB 932 E PRJ lfF Cap> Dem Rela y 1160G Test 1 17 EE-REL-I G E 50-51 P1oleet.1ve Phase Overcunent Genera l Flectnc I 2 1 AC53 B8 1 2A EE-SW GR-Swit ch gear RB 932 SQURTS Mitigation A Relay 4160G Repo rt Strategies 118 E E-RE L-I GE 50-5 I Procec t ive Phase Ove rcu 1reo1 Genera l Electn\; t21/\C53 B81 2/\ EE-SWG R-Switdigear Ril 932 SQURTS Mitiga ti on B Relay 4 16 0G Report S lr:ite g1 cs 119 EE-RE L* I G E 50-51 Protect ive l'h asc Over c uncnt General Electr i c 1 21AC53B8 1 2A EE-S W GR-Sw.tchgear RB 932 SQURTS M 1t 1 gatton c K cl a y 4160G Report Stra r eg1cs No. *;; "' Dev ier.ID Type 120 I EE-RE!,.\

GE 86 Control Re lay 121 I EE-REL-\ GS II A Protccuve Relay 122 I EE-REL-I GS II B Protecti ve Relay 123 I EE-REL* I GS Protechve Relay 124 I EE-REL-\ GS 86 Control Relay 1 25 I F.F.-REL-SS I G S0-51 l'r otcc t1vc A Rela y 126 I E E-REi,.SSIG 50-51 P rotectiv e c Relay 127 I EE-RE!,.SWPI B 50-Protective 50-ST A Rela y 1 28 I EE-REJ..SWPIB 50-Protective 50-51 c Relay 129 I EE-REl-SWPIR R6 Control Re by 130 I EE*REL-SWI'\

0 SO* Pro1ect1vc SO-SI A Relay \JI I EE-REi,.SWPID 50-Protective S0-51 C Relay 132 I EE-REL-SWPI 0 86 Control R e lay 113 I EE-CB-480F MCC-LV C i rcuit K Breaker \]4 I EE-CB-480F "1CC-LV Ci1cu1t LX Breaker SA 50.54(!) NTTF 2.1 Seismic High F r e quenc y Co nfirm ation Table 8-1: Co mponent s Identified for Hi g h Frequency Co nfirmation Cumpunenf Encfo.mre Systnn M*n11f11chtrer Model ID Type F unction Bus Lockoul General Electric 12HEi\6\ EE-SWGR* SW'llchgcar 416 0 G Phase Ovcrcu rrcn t Generitl Eleclric 12TAC5JA EE-SWGR* sw;1chgear 4160G Phase Ove<<:urrenl Gcncr::al Elcctnc 12 1A C5JA EE-SWG R-Switchgear 4\60G Pho.sc Overcurn:nl Gcncrnl Electnc 121AC5JA EE-SW GR-Sw1tctlgc.1r 4160G Emergency Startup General Electric 12HEA61 EE-SWGR-Switchgear Transfom1er Feed Locko ut 4160G Phase Overcurrent Genera l Electnc 12\A C5JB EE-SWGR-SWltchgear 4160G Phase Ovcrcum:nt General F.leclric 121AC53B EE-SWGR-Switchgear 4160G Overcurrent General Elt:"Ctnc 1 2\AC66 K EE-SWGR* Switchgear 4\60G Overcurrcr11 Gt mera l Elect r ic 12TAC66K EE-SWGR-Swi1chgeor 4160G Service Water Pump Genera l Electnc 12HEA6\ EE-SW GR-Lockout 4\60G Phase Overcuuent General Electnc 12 1AC 66 K EE-SWGR* Swnchgear 4160G Phase Ovcrcurrent Gene r al Electric 12 1A C66K EE-SWGR-Swttchgcar 4\60G Service Water Pump Genera.I E l ectric 12HEA6\ EE*SWGR-Switchgear Lockout 4160G MCC Feeder Lockout Wes11nghouse DB-50 EE-SWGR-Swuchgear 480F MCC t'ccdcr Locko ut Westinghouse DB-50 EE-SW GR-Switchsear 480F 16C4384-RPT-005 Rev 0 Page 46 of49 Floer C omponent Evaluation Buildin& i:lev. B as is for [valuatio n (R) Capaciry R a u It RB 932 EPRJ HF Cap> Dem Test RB 932 GE RS Mrtigauon Stm1eg1cs RB 932 GERS Mi11gat1on Strategics RB 932 Gt RS M1t1galloo Stratqt 1 es RB 932 EPRJ HF Cap> Dem Test RB 932 SQU RTS Mit i gation Rep on S1rateg1es RB 932 SQ UR TS M1trgallon Rep on Suategics RB 932 EPRJ HF M1t1ga1 1on Tcs1 Stralcgtcs RB 932 EPRJ TIF M1uga1100.

Tt:St Strategies RB 932 EPRJ HF Cap>Dcm Tes! RB 932 EPRT HF Mitigation Test Strategies RB 932 EPRI HF Test Strategics RB 932 EPRI HP Cap>Oem Test Rli 932 C NS Cap>Dttm Letter Rll 932 CNS Cap>Oem Letter No. *;; :;, Device ID Typo 135 I EF.-C'B-4800 MCC* L Y Circui t s B reaker 136 I EE-C B-4800 MCC

• L Y Circuit TX ll<eakef 50.54 (t) NTIF 2.1 Seismic High Freque nc y Confirmation Table B-1: Components Identified for High Frequency Confirmation Component End Mure S19tem Maauf*cturer Model ID Typo Fuadioo MC C' F eed e r Lockout Wcslmghouse DB-SO EE-SWG R-Swi t chgeo t 480G MCC Feeder Lockout Wes t ing h ouse DB-50 EE-SWO R-SW1tchgear 4800 I 6C4384-RPT-005 R ev 0 P age 4 7 of49 llloor Compueo* Evalaatioo Build Inc Elev. Buiafor J:nluaOO.

(n) Cap*citr Result RB 932 CNS Cup>Dem Letter RB 932 CNS Cap>Dem Letter SA 50.54(f) N!TF 2. I Seismic High Frequency Contirmat1on 16C4384-RPT

-005 Rev. 0 Page 48 of 49 Table B-2: Reactor Coolant Leak Path Valves Identified for High Frequency Confirmation Valve ID P&ID Comment I lead Vent (called 738A V976P on the P&ID) i s normally MS-AOV-738AV 2028 [79) clo s ed at power and deactivated b y manual valve PC-V-563 (no need to evaluate). Head Vent (called 739AV976P on the P&ID) is normally MS-AOV-739A V 2028 [79) closed a t power and deactivated by manual valve PC-V-561 (no need to evaluate).

MS-AO-A080A 20 41 [33) MS-/\0-A086J\

2041 [33) MS-AO-A0808 2041 (33) MS-J\O-J\0868 2041 [33] MS-AO-A080C 2041 (33) MS-/\O-J\086C 2041 (331 MS-AO-A080D 2041 (33) MS-A0-/\0860 2041 [331 llPCl-MOV-15 204 l (33] HPCl-MOV-16 2041 [33J MS-MOV-M074 2041 [33] MS-MOV-M077 2041 [33] MS Drain; Normally open drain line would only be a leak path if MS-MOV-M074 does not close MS-RV-71ARV 2028 [79] MS-RV-71BRV 2028 [79) MS-RV-71C RV 2028 [79) MS-RV-71DRV 2028 (79) MS-RV-7LERV 2028 [79] MS-RV-71FRV 2028 [791 MS-RV-71GRV 2028 [79) MS-RV-71HRV 2028 [79) RCIC-CV-26 2043 (35) Simple Check Valve (no need to evaluate). RF-CV-16 20 4 3 f 35l Simple Check Va l ve (no need to evaluate).

RF-CV-15 2043 [35] Simple Check Val v e (no need to evaluate).

RHR-MOV-M017 2040 Sh. I [80] RHR Isolation RHR-MOV-M018 2040 Sh. l (80) RHR Isolation RHR-CV-27 2040 Sh. 2 [81] Simple Check Valve (no need to evaluate).

RHR-MOV

• M025B 2040 Sh. 2 [81] Leak path blocked by upstream check valve RHR-CV-27 (no need to evaluate). RHR-CV-26 2040 Sh. l (80] Simple Check Valve (no need to evaluate). RHR-MOV-M025A 2040 Sh. I [80) Leak path blocked by upstream check valve RHR-CV-26 (no need to evalua te). RWC U-M OV-MOl5 2042 Sh. I [821 R WCU Isolation RWC U-MOV-MOL8 2042 Sh. I [821 R WCU Isolation CRD-SOV-SO 1 20 2 0 39 l83 J Control Rod Manual Positioning CRD-SOV-S012l 2039 [83] Control Rod Manual Positioning SA 5 0.54(t) NTTF 2.1 Seis m ic High Fre qu e n cy Confirmat i o n 16C4384-RPT-005 Rev. 0 Page 49 of49 Table B-2! Reactor Cool a nt Leak Path Valves Identified for High Frequenc y Confirmation Valve ID P&ID Comment CRD-SO V-SO 1 22 2039 [83) Normally O p en; would only be a leak path ifC R D-SOV-SO 121 o r CRD-SOV-SO 1 23 docs not close CRD-SOV-SO l 23 2039 [83) Co n trol R o d Ma nu al Positioning CRD-AOV-CVl26 2039 [83) Control Rod Sc r am CS-CV-18 2045 Sh. I [84) Simple Check Va l ve (n o n eed to evaluate). CS-MOV-MO J 2A 2045 Sh. I [84] Leak pat h b l ocked by upst r ea m check valve CS-CV-18 (no n eed to eva l uate). CS-CV-19 2045 Sh. I [8 4] Simple Check Valve (no need to evaluate).

CS-MOV-MOl2B 2045 Sh. I [84] Leak path blocked by ups t ream check valve CS-CV-19 (no need to evaluate).

SLC-CV-1 3 2045 Sh. 2 (85] Simp l e C h eck Valve (no n eed t o ev al uate). RF-CV-1 4 20 4 4 (86) S im ple C h eck V alve (n o n ee d to eval u ate). RF-CV-13 2044 [86] Simple Check Valve (no need to evaluate). I-IPCl-CV-29 20 4 4 [861 Simple C h eck Valve (no need to evaluate).