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 & Associates Engmeermg So/1111onsfor N11clear Energy 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-RPT-005 Rev. 0 Page 2 of49 REVISION RECORD Initial Issue (Rev. 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 Page 3 of 49 Introduction ............................................................................................................................. 6 I . I Purpose ............................................................................................................................. 6 1.2 Background ...................................................................................................................... 6 I .3 Approach .......................................................................................................................... 7 1.4 Plant Screening ................................................................................................................. 7 2 Selection of Components for High-Frequency Screening ...................................................... 8 2.1 Reactor Trip/Scram .......................................................................................................... 8 2.2 Reactor Vessel Inventory Control .................................................................................... 8 2.3 Reactor Vessel Pressure Control .................................................................................... 10 2.4 Core Cooling .................................................................................................................. 11 2.5 AC/DC Power Support Systems .................................................................................... 13 2.6 Summary of Selected Components ................................................................................ 17 3 Seismic Evaluation ................................................................................................................ 18 3.1 Horizontal Seismic Demand ........................................................................................... 18 3.2 Vertical Seismic Demand ............................................................................................... 18 3 .3 Component Horizontal Seismic Demand ....................................................................... 21 3.4 Component Vertical Seismic Demand ........................................................................... 22 4 Contact Device Evaluations .................................................................................................. 23 5 Conclusions ........................................................................................................................... 24 5.1 General Conclusions ...................................................................................................... 24 5 .2 Identification of Follow-Up Actions .............................................................................. 24 6 References ............................................................................................................................. 25 A. Representative Sample Component Evaluations .................................................................. 32 A.1 High Frequency Seismic Demand .................................................................................. 32 A.2 High Frequency Capacity ............................................................................................... 36 B. Components Identified for High Frequency Confirmation ................................................... 38 TABLE OF TABLES: Table 3-1: Soil Mean Shear Wave Velocity vs. Depth Profile ..................................................... 19 Table 3-2: Horizontal and Vertical Ground Motions Response Spectra ...................................... 20 Table B-1: Components Identified for High Frequency Confirmation ........................................ 38 Table B-2: Reactor Coolant Leak Path Valves Identified for High Frequency Confirmation ..... 48 SA 50.54(f) N!TF 2.1 Seismic High Frequency Confirmation EXECUTIVE SUMMARY I 6C4J84-RPT-005 Rev. 0 Page 4 of49 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 Rev. 0 Page 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 16C4384-RPT-005 Rev. 0 Page 6 of49 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 that licensees and holders 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 6C43 84-RPT-005 Rev. 0 Page 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 reevaluated 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 Confirmation 16C4384-RPT-005 Rev. 0 Page 8 of 49 2 SELECTION OF COMPONENTS FOR HIGH-FREQUENCY SCREENING The fundamental objective of the high frequency confirmation review is to determine whether the occurrence of a seismic event could cause credited equipment to fail to perform as necessary. An optimized evaluation process is applied that focuses on achieving a safe and stable plant state following a seismic event. As described in Reference [4], this state is achieved by confirming that key plant safety functions critical to immediate plant safety are preserved (reactor trip, reactor vessel inventory and pressure control, and core cooling) and that the plant operators have the necessary power available to achieve and maintain this state immediately following the seismic event (AC/DC power support systems). Within the applicable functions, the components that would need a high frequency confirmation are contact control devices subject to intermittent states in seal-in or lockout (SILO) circuits. Accordingly, the objective of the review as stated in Section 4.2.1 of Reference [4] is to determine if seismic induced high frequency relay chatter would prevent the completion of the following key functions.* 2.1 Reactor Trip/Scram The reactor trip/SCRAM function is identified as a key function in Reference [4] to be considered in the High Frequency Confirmation. The same report also states that, "the design requirements preclude the application of seal-in or lockout circuits that prevent reactor trip/SCRAM/unctions" and that "No high-frequency review of the reactor trip/SCRAM is necessary. " 2.2 Reactor Vessel Inventory Control The reactor coolant system/reactor vessel inventory control systems were reviewed for contact control devices in seal-in and lockout (SILO) circuits that would create a Loss of Coolant Accident (LOCA). The focus of the review was contact control devices that could lead to a significant leak path. Check valves in series with active valves would prevent significant leaks due to misoperation of the active valve; therefore, SILO circuit reviews were not required for those active valves. Reactor coolant system/reactor vessel inventory control system reviews were performed for valves associated with the following functions:

• Main Steam,

• High Pressure Core Injection,

• Residual Heat Removal,

• Control Rod Drive,

• Reactor Water Clean-Up *The selection of components for high frequency screening is described in Stevenson & Associates report J 6C4384-RPT-OO J [72] and is summarized herein.

SA 50.54(f) NTTF 2.1 Seismic High Frequency Confirmation I 6C43 84-RPT-005 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-A080AIBICID, MS-AO-A086AIBICID 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-M074, MS-MOV-M077 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-71ARVIBRVICRVIERVIGRVIHRV 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 DRVIFRV 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 t6C4384-RPT-005 Rev. 0 Page to of 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 is no seal-in and the valves reclose without operator intervention, chatter is acceptable and no contact devices in this circuit meet the selection criteria. 2.2.4 Control Rod Drive Valves Control Rod Manual Positioning Valves CRD-SOV-S0/20, CRD-SOV-S0/21, CRD-SOVS0/22, CRD-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 Clean-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 "required post event pressure control is typically provided by passive devices" and that "no specific high frequency component chatter review is required 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 Rev. 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 milliseconds at any one time. When contact chatter applies power to the coil of a time delay relay with delay times significantly longer than this, the coil is not continually energized long enough to satisfy the timing function and thus the time delay relay will not change state.

SA 50.54(f) NTTF 2.1 Seismic High Frequency Confirmation 16C4384-RPT-005 Rev. 0 Page 12 of49 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 high 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 closing 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 analysis, there are no moving contact devices in the control circuit of this valve that meet the selection criteria. Pump Discharge Valve RCIC-M0-21 is normally closed 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 criteria. : Manual RC!C initiation is presumed to include operator alignment of valves via the RC!C system controls, including pump suction lo tht: dt:sired source.

SA 50.54(t) NTTF 2.1 Seismic High Frequency Confirmation 2.5 AC/DC Power Support Systems 16C4384-RPT-OOS Rev. 0 Page 13 of49 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 Ancillary 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 chatter 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 Chargers 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 power for their two (2) divisions of Class IE loads, with one 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 not 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 undervoltage relaying detecting the LOOP, the Class IE control systems must automatically shed loads, start the DGs, and sequentially 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 event, cause 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 divisions. The selected contact devices for all divisions are included in Table 8-1.

SA 50.54(t) NTTF 2.1 Seismic High Frequency Confirmation 2.5.1 Emergency Diesel Generators 16C4384-R.PT-005 Rev. 0 Page 14 of49 The analysis of the Emergency Diesel 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. Generator Protective 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 Relay. Chatter in the 50151-1 I FE, 50/51-2 I FE, and 50151-3 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 Systems In order to start and operate 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-RPT-005 Rev. 0 Page 15 of49 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 battery 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 Relays 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 respectively [ 42]. Lockout Relay 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, 120V AC, 250 VDC, and 125V DC Distribution and MCCs 16C4384-RPT-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 16C4384-RPT-005 Rev. 0 Page 18 of49 Per Reference [4], Section 4.3, the basis for calculating high-frequency seismic demand on the subject components in the horizontal direction 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. However, 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 50.54(f) NTTF 2.1 Seismic High Frequency Confirmation I 6C4384-RPT-005 Rev. 0 Page 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 10.0 1,020 0.0098 0.0098 2 14.5 4.5 1,020 0.0044 0.0142 3 24.5 10.0 1.030 0.0097 0.0239 4 34.5 10.0 1,040 0.0096 0.0335 5 40.5 6.0 1,040 0.0058 0.0393 6 49.5 9.0 l,120 0.0080 0.0473 7 59.5 10.0 1.620 0.0062 0.0535 1,369 8 69.5 10.0 1.760 0.0057 0.0592 9 79.5 10.0 1,760 0.0057 0.0649 10 84.5 5.0 1,760 0.0028 0.0677 II 94.5 10.0 2.750 0.0036 0.0714 12 97.0 2.5 7.292 0.0003 0.0717 13 98.4 1.4 7.294 0.0002 0.0719 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 (Vs30) 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 (Vs1).

• 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;Ns1]).

• The velocity of a shear wave traveling from a depth of 30m to the surface is therefore the total distance (30m) divided by the total time; i.e., Vs30 = (30m)/L[d;Ns1J. 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 site soil class is determined, the mean vertical vs. horizontal GMRS ratios (V/H) at each frequency are determined by using the site soil class and its associated V/H values in Reference [4], Table 3-2. The vertical GMRS is then calculated by multiplying the mean V/H ratio at each frequency by the horizontal GMRS acceleration at the corresponding frequency. It is 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-RPT-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 (!!) 100 0.241 0.78 0.188 90 0.242 0.82 0.198 80 0.245 0.86 0.211 70 0.249 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 0.197 2.5 0.209 0.67 0.140 2 0.162 0.67 0.109 1.5 0.116 0.67 0.078 1.25 0.096 0.67 0.064 I 0.082 0.67 0.055 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 0.017 0.67 0.011 0.125 0.014 0.67 0.009 0.1 0.011 0.67 0.007 SA 50.54(t) NTTF 2.1 Seismic High Frequency Confirmation 16C4384-RPT-005 Rev. 0 Page 21of49 0.60 -HGMRS -VGMRS 0.50 ---V/H Ratio (A-Intermediate) 0.40 ,, I ' I ' 1.00 ,' \ 0.90 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 uency 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 AFc 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, AFc 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 16C4384-RPT-005 Rev. 0 Page 22 of 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 Afsv to account for seismic amplification at floor elevations above the host building's foundation

• Vertical in-cabinet amplification factor Afc to account for seismic amplification within the host equipment (cabinet, switchgear, motor control center, etc.) The in-structure amplification factor AFsv 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) NTIF 2.1 Seismic High Frequency Confirmation 4 CONTACT DEVICE EVALUATIONS 16C4384-RPT-005 Rev. 0 Page 23 of 49 Per Reference [4], seismic capacities (the highest seismic test level reached by the contact device without chatter or other malfunction) for each subject contact device are determined by the following procedures: (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 General Conclusions 16C4384-RPT-005 Rev. 0 Page 24 of 49 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 'Seismic' 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) -Response 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 l6C4384-RPT-005 Rev. 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. 091 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 Diagram, Automatic Blowdown System. [19] Cooper Nuclear Station Document 791E253 Sheet 3 Rev. 12/AA, Elementary Diagram, Automatic Blowdown 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 16C4384-RPT-005 Rev. 0 Page 28 of49 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. [45] Cooper Nuclear Station Document 2077 Rev. 78/AA, Flow Diagram, Diesel 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, Control Elementary Diagrams. [50] Cooper Nuclear Station Document G5-262-743 Sheet l OA Rev. 06/ AD, Electrical Schematic, Emergency Diesel Generator #2.

SA 50.54(t) NTTF 2.1 Seismic High Frequency Confirmation 16C4384-RPT-005 Rev. 0 Pagt: 29 of 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, Emergency Diesel Generator #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 16C4384-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 1201240 VAC 3-Wire 60Hz. [66] Cooper Nuclear Station Document 20-100288 Sheet I Rev. 00/AA, Schematic, JOkVA Static Switch 2-Pole 1201240 VAC 1-Phase 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 Centers 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 One 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 Relays," 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 16C4384-RPT-OOS Rev. 0 Page 32 of49 A. REPRESENTATIVE SAMPLE COMPONENT EVALUATIONS A detailed example analysis of two components is provided within this section. This example is intended to illustrate each step of the high frequency analysis methodology given in Section 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 components. 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]): SAaMRS = 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: ELround = 903 ft Diesel Generator Building ELround == 860 ft Reactor Building Component Floor Elevation: ELcomp = 903 ft DG-LMS-DGJ 630SDL Distance Between Component Floor and Foundation Elevation: ELcomp = 932 ft EE-REL-I FA 86 hcomp == ELcomp -ELround == 0 ft DG-LMS-DGJ 630SDL hcomp = ELcomp -ELround = 72 ft EE-REL-I FA 86 SA 50.54(t) NTTF 2.1 Seismic High Frequency Con ti nnation 16C4384-RPT-005 Rev. 0 Pagt! 33 of49 Calculate the horizontal 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 < hcomp < 40ft: Intercept of Amplification Factor Line with Amplification Factor Axis: Horizontal In-Structure Amplification Factor (Ref. [4], p.4-1 l): ffih = 2.1-1.2 = 0.0225 2_ 40fc-Oft ft AFsH(hcomp) = (mh

• hcomp+ bh) ifhcomp <= 40ft 2.1 otherwise AF::rn(hcomp) = 1.2 DG-LMS-DG! 630SDL AFs1-1(hcomp) = 2.1 EE-REL-JFA 86 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 86" Horizontal In-Cabinet Amplification Factor (Ref [4], p. 4-13): AFc.h(cab) = 3.6 if cab= "MCC" 7.2 if cab= "Switchgear" 4.5 if cab= "Control Cabinet" 1.0 if cab= "Rigid" AFch(cabl) =4.5 AFch(cab2) =7.2 Multiply the peak horizontal GMRS acceleration by the horizontal in-structure and in-cabinet amplification factors to determine the in-cabinet response spectrum demand on the components: Horizontal In-Cabinet Response Spectrum: lCRSc h = Af s1-1

• AFc h

• SAoMRS ICRSc.h =1.2*4.5*0.463=2.Sg DG-LMS-DGI 630SDL lCRSc h =2.1 *7.2*0.463=7g EE-REL-I FA 86 SA 50.54(f) NTTF 2.1 Seismic High Frequency Confirmation 16C4384-RPT-005 Rev. 0 Page 34 of 49 A.1.2 Vertical Seismic Demand 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]): SAaMRS = 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): PGAaMRS = 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) s3o --l: (-,--di.) Vst where, di: Thickness of the layer (ft), Vs;: 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: VsJO = 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 16C4384-RPT-005 Rev. 0 Page 35 of 49 Detennine the peak acceleration of the vertical GMRS (SAVGMRs) 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: hcomp = ELcomp -ELround = 0 ft DG-LMS-DGJ 630SDL hcomp = ELcomp -ELround = 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 mv= = lOOft-Oft. 0.017 2:.. ft with Amplification Factor Axis: bv = 1.0 Vertical In-Structure Amplification Factor: AFsv(hcomp) = mv

• hcomp + bv AFsv(hcomp) = l.O DG-LMS-DGI 630SDL AFsv(hcomp) = 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 16C4384-RPT-005 Rev. 0 Page 36 of49 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): ICRSc.v = AFsv

• AFcv

• SAVGMRS ICRSc.v =1.000*4. 7*0.31=1.458g DG-LMS-DGI 630SDL ICRSc 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 16C4384-RPT-005 Rev. 0 Page 37 of49 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 correction 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 FMs =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 86 SA 50.54(1) NTTF 2.1 Seismic High Frequency Confirmation B. COMPONENTS IDENTIFIED FOR HIGH FREQUENCY CONFIRMA TIO'.'J Table B-1: Components Identified for High Frequency Confirmation Component l!:ndosurc No. *= System :;> Device ID Type Manuracturer Model ID *rn.* Function 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-AISOSS LRP-PNL-25* Control Switch Pressure 58 Cabinet J 1 RCIC*PS*87A Process Reactor l>rcssurc Stauc-0-Rins 5N6-BBJ-U8-LRP-PNL-25-Control Swirch ClA-TTNQ 58 Cabmel 4 I RCIC*PS-878 Process Reactor Pressure Staltc-0-Rmg 5N6*BB3-U8-LRP-PNL-25* Control Switch CIA-ITNQ 58 Cabinet 5 1 RCIC*l'S*87C Plocess Reactor PrCllisure Stahc-0-Rmg SN6-BBJ-U8-LRP-PNL-25-Control SW11Ch CIA-TI'NQ 58 Cabmet 6 1 RCIC-PS-8ID Process Reactor Pressure 5N6-BB3-U8-LRP-PNL*25-Control Switch CIA-ITNQ 58 Cabinet 7 I RCIC-REL*KIO 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 Excess 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 Bradley 700*RTC-LRP*P L-9-Control Relay D1fTeren11al Pressure llllOUI 30 Cabinet 10 I RCIC-REL-KlJ Coolrol Reactor Pressure Gt.-ncral Electric 1'.?llGAllAS2f I.RP-Pi 1...-q-Con1rol Relay 30 Cabinet II I RCIC*REL-Kl S Control RCJC Auto bolatmn General Electric 12HFASIA42F LRP*PNL-9-Control Relay Signal 30 Cabinet 12 I RCJC *REL-K6 Control Turbme Exhaust H1gt'I General Electric I 2HGA 11 l\52F LRP-PNL*9-Con1rol Relay Pressure 30 Cab met Control Pump Suction Low National LRJ*-l"NL-9-Cont rot 13 I RCIC-REl,K7 :-<TS-812 Relay Pressure JO Cabinet 14 I RCIC*REL-K8 Control furb1nc Tnp General Electnc IZICFASIA42f LRP*PNL-9-Cootrol Relay 30 Cabinet 16C4384-RPT-005 Rov 0 Page 38 of 49 Floor Componenl i:Y*lu.ation BuiJdlnc Elev. Basil for Evaluation (0) C*pecity Result Rll 881 GERS Operntor Action RB 881 GERS Operator Ac11on Rll 881 Vendor Cap>Dem Report RB 881 Vendor Cap> Dein Report Rll 881 Vendor Cap>Dem Report RH 881 Vendor Cap'> Dem Report CB 903 GERS Cap.,. Ucm CB 903 GERS Cap> Dem rn 901 SQURTS Cap> Dern Report CB 903 GERS \ap>Oem cu 901 GERS C1p>Oem CB 901 GERS Cap> Dem f'NS CB 903 Cap> Dem Repon Cl\ 90] GERS Cap'>Oem No. *;; ;;> Device ID Type lj I RCIC*REL*KJO Con1rot Relay 16 I RCIC-REL-KJ l Control Relay 17 I KCIC-1<.1'.L-KJ2 Contro! Relay 18 1 KCIC*KEL-KJ3 Control Relay 19 I RCIC-TS-nA Process Switch 20 l RClC-TS-79ll Proct!Ss Switch 21 I RCIC-TS-79C Process Swuch 22 I RCIC-TS-790 Process 5W1tch Z3 I RC!C-TS-80A Process SW1tch 24 I RC!C-TS-8011 Process Switch 25 I RCIC-TS-80C Process SWTtcli 26 I RCIC-TS-800 Process 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 50.54(1) NTIF 2.1 Seismic High Frequency Confirmation Table B-1: Components Identified for High Frequency Confirmation Component Enclosure S)'l'ltrm Man"racturer Model ID TYJI* Fu*ction I.me Space Excess Generul Ele4:mc 1.lHGA 11 LRJ>-PNL*9-Concr"OI Temper.uure 33 Cuburnt Steam Lme Space Excess General Eloctnc 12HGAl 1.'\l2F LRP*PNl.-9-Control Temperature 3l Cabinet Steam I.inc H1ch Allen Bradley 700-RTC-LRP-PNL,-9-Control DttTerential Pressure llllOUI JJ Cabinet RCIC Auto lsolatioo General Elccmc 12HFA1lli\2F LRP-PNL-9-Control Signal J3 Cnbmet Steam Line Space Patel I Fenwnl 01-170230-090 N/A(Local) R1g1d Skid Temperature Mounted Stc:sm Line Space E."'tcess Pa1el I Fenwal 01-170230-090 NIA (Local) Rigtd Skid Temperature Mounted Slcam Line Space Excess Patel I Fcnwal 01-170230-090 N/A(local) R1g1d Skid Temperature. Mounted Steam Lme Sp3ce Excess Patel I Fenwal 01-1 70230-0QO NIA (Local) K1g1d Skid Temperature Mounted Steam Line Space E'<ccss Patel I Fenwal 0 I -I 70230--090 NIA (Local) R1g1d Skid Temperature Mounkd Steam Line Sp.:u:e Excess Patel I Fenwal 01-170:?]0-()0() !Local) R191d kid Temperature Moun1ed Une Spare Excess / FemYal 01-170230...@0 NfA (Lccal) R1g1d Skid Temperature Mounted StCllm Line Space E'<ccss Patel I Fenwal U I -1 70230-090 NI A (L-Ocal) Skid Temperature Mounl<Xi Stcuin Line Space Excess Pntol / Fenwal 01-170210*090 NIA (Local) R1g1d Skid Temperature Mounted Steam Line Space Excess Pacel *' Fenwal 01-170230-090 N/A (L-Ocal) R1gtd Skid Temperature ).ioonled Steam Line Space Excess l'aiel I fenwal 01 *1702J0-090 'llA (Local) R1g1d Sl..1d Temperature Mounted 16C4384-RPT-005 Rev 0 Page 39 of 49 fi'lOGr Compont!nt Evalualion Bui.ldine, Elev. Basi1 for Evaluation (ft) Capacity Rt1u.lt CB 903 GERS Cap>Dem CB 90'.\ GERS Cap>Dcm CB 903 SQURTS Cap >Dem Report CB 903 EPRI Hf Cap> Dem Test Rll 860 CNS Crtp>Dem Repor1 RB 860 CNS Cap>Dcm Report RB 881 CNS Cap>Dem Rt.11ort RB 881 OIS Cap> Dern Report RB 860 CNS Cap>Dem Roport RB 860 CNS Cnp>Dcm Report RB 881 CNS Cap> Dem Report RB 881 CNS Cap'> Dem Report RD 860 CNS Cap> Dem Report 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 RCIC-TS-828 Process Swuch 33 I RCfC-TS-82C Process Swttch 34 I RCIC

• TS*82D Process Switch JS I DG-LMS-DGI l'roc<!"S 630SDL SW1tch 36 I DG-l.MS-DG1 Process 630SDR Switcn 37 I DG-REL-DGI 140S Control Relay 38 I DG-REL-DGI Control l4RX3 Relay 39 I DG-REL*DGI Control 14RYI Relay 40 I 00-REl.-DGI 27-59 Protective Kclay 41 I DG-REL*DGI 40 Protective Relay 42 I DG*REL-DGI Control 481SEX Rclny 43 I DG-REL-DGI Control 4E.'lllX Relay 44 I DG-REL*DGI Control 4EMX3 Relay SA 50.54(!) NTIF 2.1 Seismic lligh frequency Confirmation Table B-1: Components Identified for High Frequency Confirmation Component Enclo.ture Sys rem M*our*cturer Modd ID Type Fu onion Steam Une Space Excess Patel I Fcnwal 01-1 70230-090 NIA (Local) R1g1d Skid Tempera1urc Mounted Steam Une Sp3ce Excess I Fenwal 01*170230-090 N/A(Local) R1g1d Skid Temperature Muunlud Stctm Line Space Excess Patel I Fcm\111 01-1 ;o230-090 NIA (l.ocal) R1g1d Skid Temperawre Mounted Sleu.m Line Spece Excess Patel I Fenwal 01-170:?30*090 N/A(Local) Rigid Skrd Temperature Mounted Steiim Line Space Excess Pote! I Fenwal 01-170230-090 NIA (Local) Rigid Skid Temperature Mounted Overspaxl Namco Controls E,\ 180-32302 DG-PNL-Control DGI ECP Cabinet Engine Overspued Namco Controls EA 18t>-Jll02 DG-PNL-Control DGI ECP Cabinet Engine Overspeed Potter & KRPl4DG-125 DG-PNL-Control Brumfield DGI ECP Co.bmct Engine Rumung Potter&. KRPI 4DG-125 DG-PNL-Control Brumfield OGI ECP Cabuu.-t Enij:ine Running Allen 700.RTC-DG-PNL-Control 11020UI DGI ECP Cabmcl Generator Abnormal Electric 121AV7JAIA DG-PNL-Control Voltai,::o OGT EC'P Cnhrnct Generator Field Failure General electnc 12CEHSIAIA UG-PNL-Control OGl ECP Cobmet E11gmc (ncomplctc Sturt Potter& KRP I 4DG-1 25 DG*PNL-Control Sequence Brumfield DGI ECP Ciibuu .. -t Emergency Engine Start Potter& KRJ' I 4DG-I 25 Control Brumlidd DGI EC!' Cabinet Fmcrgency Eng111e S1J11t Potter& KRPl4DG-125 DG-PNL* Control Brumfield DGI ECI' Cab1m .. 't 16C4384-RPT-005 Rev 0 40 of49 Floor Component Ellalua1lon Building Elev. Basis ror E*aluation (0) Capa('ity ReJult RB 881 CNS C1tp>r>cm Report RB RM CNS Cap>Dem Report RB 860 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 DGI 903 SQURfS Cap> Dem Report DGI 903 CNS \ap >Dem Report DGI 903 CNS Cap>Oem Repon DGI 903 CNS Cap >Dem Report DGI 903 C:Dem R..:pon DGI 903 CNS Dem Report No. .. ;;> Device lD Type 45 I DG-REl.-DGI SI i\ Prote<..11ve Relay 46 I DG-RF.l.*DGI SIB Pmlecllve Kcl.ay 47 I DG-REL-DGI 51 C Protective Relay 48 ( DG-REL*DGI Control 62Cl.X Relay 49 ( OG-REL-OGI Control 630SDX Rela.y so I OG-REL-DGI 67 Protective Relay 51 I DG*REL*DGI 86 Control Relay l2 I DG*REL-DGI 87 A Protective Relay l3 I DG*R.EL-DGI 87 B Protecttvc Relay 54 ( OG-REL-OGI 87 C Protective Relay ll I DG-REL-DGI RI04 Control Relay 56 I DG-KT-Jl42 Process Switch 57 I DG-LMS-OG2 Process 630SOL Switch l8 l OG-LMS-DG2 Process 630SOR SWltch l9 I DG-REL-OG2 1405 Control Relay 50.54(t) NTH' 2.1 Seismic High Frequency Confirmation Table B-1: Components Identified for High Frequency Confirmation Component [nclosure System M1nufactun:r Model ID Type Funciion Phase Overcurrent General Electric IFCVSIAD DG*PNL-Comrol DGI ECP Cab1nel Phase Overcurrem General Electric !FCVlli\D OG-PNL-Con1rol DGI ErP Cabinet Phase Overcurrent General Efectnc IFCV51AO DG*PNL-Conlrol OGI ECP Cab1m .. "t Engine CranlC1ng L1m1t Agastat Relay E70 I 2PDOO<l Nam"-o Controls EA I 80-3 I 302 DG*PNL-Control DG2 ECP Cabinet Engine Potter& KRP141JG-125 DG-PNL-Control Orumfield DG2f'C'P Cabinet I 6C4384-RPT-005 Rev 0 Pa e 41 of 49 Ji'foor Component Evalu1'tion Buildine, f.lev. Basis for Evaluation (rt) Capacity Rtsull DGI 903 Vendor Cap> D<rn Report DGI 903 Vendor Cap> Dem Report OGI 903 Vendor Cap >Dem Repo1t OGI 903 EPlllHF Cap>Dem Test OGI 903 CNS Cap> Dem Repon DGI 'IOJ SQURTS M111g:iuon Report StralCJ?1es DGI 903 EPRI HF Cap>Dem Test DGI 903 CNS M1ug:uion Report S1ratcc1es DGI 901 CNS Mmg11uon Report Stnuegtcs DGI 903 CNS M1t1g11.1ioo Report Stratcgu:s DGI 903 GERS Cap> Dem DGI 903 CNS Cap> Dem Rcpon DG2 903 CNS Cnp> Dem Report DG2 903 CNS Cap>Ocm Kcpon DG2 903 C:'>S C1p>Oem Report No. *;; :> Device ID Type 60 I DG-REL-DG2 Control 14RX3 Relay 61 I DG-REL-DG2 Control 14RYI Relay 62 I DG-REL-DG2 27-59 Protective Relay 63 I DG-REL-DGZ 40 Pn:Mthve Relay 64 I DG-REL-DG2 Contra I 481SEX Relay 65 I DG-REL-DG2 Control 4EMX Kclay 66 I DG-REL-DG2 Control 4EMX3 K.clay 67 I DG-REL-DG2 51 A Protective Relay 68 I DG-REL-DG2 51 B Protcchvc Relay 69 I DG-REL-DG2 SI C Protective Relay 70 I DG-REL-DG2 Control 62CLX Relay 71 I DG-REL-IJG2 Conuol 6JOSDX Relay 72 I DG-REL-DG2 67 Protective Relay 73 I DG-REL-DGZ 86 Conical Relay 74 I DG-REL-DG2 87 A Prult.."Cf1ve Ralay SA 50.54(t) Tff 2.1 Seismic High Frequency Confirmation Table B-1: Components Identified for High Frequency Confirmation Compoaent Enclosure System l\tanurachlrer Model ID Type Function Engine Running Potter& KRPl4DG-J25 DG-PNL-Control Brumfield DG2 ECP Cabinet Engine Runnrng Allen Bradley 700-RTC-DC"PNL-Control I 1020UI DG2 ECP Cabinet Generator Abnom\a I General Electric 121AV7JAIA DG-PNL-Control Volta cc DGZ ECP Cab met Generator Field Failure Guieral Electnc 12Cf-H5JAIA DG-PNL-Control Cabinet Engine Jncompk:tc Start Pouer& KRPl4DG-125 [>G.PNL-Control Sequence Brumfield DG2cCP Cab met Emergency Fngmc Stan Potter& KRl'l4DG-125 JJG-rNL-Control Brumfield DG2 ECI' Cabinet F.me*cency Engine Suut Potter & KRP14DG-125 DG-PNL-Control Brumfield DGZ ECP Caliim:t Phase Overcurrent General Electnc IFCVSIAD Control DG2 ECP Cabmel Phase Ovcrcurrcn t General Electric IFCVSIAD DG-P L-Control DG2 ECP Cabinet Phase Owrcurrent GenL'Tal Electnc !FCVSJAD DG-PNL-Control DG2 ECP Cabmel Engmc Crankmg Lcmu Thomas& Betts E7012PD004 DG-PNL-Control Timer DGZ ECP Cabinet Rngtne Over.speed Potter & KRP J 41JG-l 25 DG-PNL-Control Shutdown Brumfield DG2 ECP Cob met D1rec1ionnl Ovcr<:urrcnt General Electnc !CW-SIA DG-PNL-Control DG2 ECP Cnbmel Diesel Gi:ocrator Lockout Gcntrnl Electnc 12JIEA61 DG-PNL-Control Relay DG2 ECP C11b1nct Generator D1ffercnt1JI General Electnc CFD-120 DG-P L-Control D<.;2 cl'P C11bmet I 6C4384-RPT-005 Rev 0 Page 42 of49 Floor Compooen1 Evaluation Building [lev. Buis ror [valuation (fl) C*pae:ity R"ult DG2 903 CNS Report DG2 903 SQURTS Cap>Dcm Report 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 903 Vt..-ndor Cap> Dem Report DG2 903 Vendor Cap> Dem Report DG2 903 EPRI HF Cap>Oem DG2 901 CNS Cap>Oem Report DG2 903 SQt;RTS Mitigation Report S1ra1eg1es DG2 903 EPR! HF Cap">Dem T .. 1 DG2 903 CNS M1t1galton Reporl Slrategies No. . ., ;;, Device ID Type 75 I DG-REL-DG2 87 B Pmtcct1ve 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 Prott.'Chve 79 I KJ Relay Protective 80 I KJ Protocttve 81 I KJ Relay Proloct1ve 82 I KJ Rehay 83 I EE-CB-4 I 60DGI MV Circuit EGI Rre:aker 84 I EE*CB-4160DG2 MY Circuit E02 85 I EE-CB-4 I 60f I FE MY Circuit Bre.lker 86 I EE-CB-4 I 60F SS IF MVCircu1t Breaker 87 EE-CB-416-0P MY Circuit I SWPIA Brcu.kBr 88 1 EE-CB-4160F MVCircutl SWl1IC Breoker 89 I EE-REL-IPA 5 I A Protective Rtlay 50.54(f) TIF 2.1 Seismic High frequency Confirmation Table B-1: Components Identified for High Frequency Confirmation Component Eodoimrt" Sy1fem Model ID 'fype Function Generamr Oifferent111.I Gcricral Eleanc CFD-126 DG-PNL-Control DG2 ECP Cahmct Genera1or D11Tt:tt.'1ll1al Gmeral Eleclnc CFD-128 OG-PNL-Control DG2 ECP Cabinet Engine Speed Agastat Relay EGPBOO<I DG-r L-Control Co D02ECP Cabinet Oynalco Corp SST-2400AN-140 DG*PNL-Control DG2 ECP Cabmet C&D EE-CHG-125 Control Overvoli.age Shutdo\.Yfl Technolog1es ARRIJOICOOF IA Cabmct Inc C&D EE-CH0-125 Control Ovcrvoltagc Shutdown rechnolog1es ARRIJOK200F IB Cabmet In<: C&D EE-CHG-250 Conlrol O"ervoltage Shutdown Technolocies ARR260K100F IA Cabmcl Inc C&D EE-CHG-250 Conlrol Overvoltage Shutdown Technologies ARR260K200F IB 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 76-250-EE-SW GR-Switchgear Transfonner Lockout ID 4160F Service Water Pump S1emt:ns 5GEHU-Jl0* EE-SW GR-SwitchKt:ar Lockout 1200-78 4160F Service Water Pump S1en11..-ns lGEHU-350-EE-SWGR-Switchgear Lockout 1200-78 4160F Phase Overcurrent General l!IC1..-1:nc 121AC5JA EE-SWGR* Switchgear 4160F 16C4384-RPT-005 Rev 0 Page 43 of49 Floor Componenl Enluation Building Basi.1for 1£valuation (ft) Result DG2 '1()3 css Mitigation Report Strategics DG2 903 CNS M1t1gation Report Strategies DG2 903 GERS Cap">Dem llG2 903 CNS C3p':>Dcm Report 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 M111gat1on Strategies No. .. ::> Device ID Type 90 I EE-Kl!L-1 t"A 51 H Pru1cct1ve Relay 91 I EE-REL-I FA 51 C P1otcct1ve Reloy 92 I EE-REL-IF A 86 Control Relay 93 I EE-REL-I FE 50-51 Prot< .. -ct1ve A Relay 94 I EE-REL-JFE 50-51 Proteciive B Relay 95 I EE-REL-I FE 50-51 Protective c Relay 96 I EE-REL-I FE 86 Control Relay 97 I EE-REL-IFS 51 A Protective l\elay 98 I EE-REL-IFS 51 B Protective Relay 99 I EE-REL-IFS 51 C Protcclrve Relay JOO I EE-REL-IFS 86 Conlrol Relay JOI I EE-REL-SSIF 50-51 ProtL"Ctive A Relay 102 I EE-REL-SS IF 50-51 1>rotcct1vc c Relay 103 I EE-REL-SWP I A 50-Protecltve 50-ll A Relay 104 I EE-REL-SWPIA 50-Protecuvc 50-51 c Relay 50.54(f) NTTF 2.1 Seismic High Frequency Confirmation Table B-1: Components Identified for High l<'requcncy Confirmation Endo1un System Mao11facturer Model ID Tyl"' Function Phase Overcunent Genera I Elcctnc t2JAC53A EE-SW GR-Switchgelr 4160F Phase Overcurrcnt Gcm.-rml Electric 121AC53A EE-SWGR-4160F Normal Feed Lockout Gencrsl EJectn..; 12HEA61 EE-SWGR-Sw11chg;ear 4160P Phase Overcurrenl Gencrul Elccmc 121AC5313812A EE-SWGR-Switctlgmr 4160F Phase Overcurrent Gcnernl Elc.:;111c l 21AC53BSI 2A EE-SWGR-Sw11chgi:ar 4160f Phase Overcurren1 Generul Electric 12IAC5Jll812A EE-SWGR-Swttchucar 4160F Bus Lockout General Eleclnc 12HEAbl EE*SWGR-SW1tchgear 4160f Phase Overcurrent General Eleclnc 12fAC5JA EE-SWGR-Swllchgt:ar 4160f Phase Overcurrcnl General Electm: 121AC53A EE-SWGR-Switchgear 4160F Phase Overcurrent G4...-neral Electric 121AC53A EE-SWGR-Switchgear 4160f Emergency Starrup General 12Hl.:A61 EE-SWGR-Switchgear Transfom1er Feed Lockout 4160f Phase Overcurrcnt General Electric 121AC53 EE-SWGR-Sw11chgcar 4160F Phase Overcmn:nt General Eloctric 1211\(53 EE-SW GR-Switchgeor 4160f Phase Ove1<.*1ment General Electnc 121AC66 EE-SWGR-Switchgear 4160F Phase Overcurrent General Electnc 121AC66 EE-SW GR-Swnchgear 4160F l6C4384-RPT-005 Rev 0 44 of 49 Floor Component Enlualitm BuiJdinc Elev. Batis for Evaluation (rr) Capacity Re1ult RD 932 GERS Mi11gat1un Strategies JUI 932 GERS Mitigation Str3.teg1es RB 932 EPRJ HF Cap>Dem TC$! RB 932 SQURTS Mitignt1on Report Strategics RB 932 SQURTS Mitigation Report RB 932 SQ UR TS Mill!llU1on Report Slrategu:::s RB 932 HP Cap>Dem Test RJ3 932 GERS M1tig.allon S1ru1eg1t:s RD 932 GERS M1t1gaoon StniH.-gies RB 932 GCKS Mihgallun Strategies RB 932 EPRJ HF Co.p>Dem Test RB 932 GERS Mit1g1111on Strategies RJ3 932 GERS Mi1iga11on Sm.1tcg1cs RB 9)2 EPRl HF M1tigat1on Test Strutcg1es RB 932 EPRJ HF M1t1gat1on Test Stmtcg1es SA 16C4384-RPT-005 Rev 0 50.54(f) NTTF 2.1 Seismic High Frequency Confinnation Page 45 of 49 Table B-1: Components Identified for High Frequency Confirmation Cumpunent EnclO!"ure lli'loor Component [valuation No. *= System Buildina: Elev. Basis for Evaluation "' Device ID Type Manufacl\lrer Model ID Type (fr) Function Capacity Retulf 105 EE.REL-SWPI/\ 86 Control Service Water L'ump General Elecmc lZllEA61 EE-SWGR-Switchgear Ril 932 EPRJ HF Cap>Dem Relay L<>ckoul 4160F Test 106 EE-REL-SWPIC 50-Protec1ive OvcrcUrTen t General Elecrnc 121AC66 EE-SWGR-Switchgear Ril 932 EPRJ HF Mitigation 50-51 A Rday 4160F Test S1ra1cg1cs 107 EE-REL-SWPIC 50-Protc:ctivc Pha.c;c Ovcrcurrent General Hcclnc 121AC6'i EErSWGR-Switchgear RB 932 EPRJ Hf "41t1gation 50-51 c Relay 4160f Test Strategies 108 EE-REl,SWPI C 86 Control Service W!Her Pump General Electric 12HEA61 EE-SW GR-SW'!tchgcar Ril 932 EPRI HF Cap>Dem Re1ay Lockout 4160F Test 109 F.F-CR-4160G IGE MVCircuit Switchgear Feeder General Electric AMH-4 76-250-EE-SW GR-Sw11chgcar RH 932 Not Mitiga1ion Breaker Lockout ID 4160G Available Strategies 110 EE-CB-4 I 60G SS 1 G MV Circuit talion Service General Eleclric l\)..tfl-4 76-250-EE-SWGR-Swttchgea.r RJl 932 N9t Breaker Tronsfonner Lockout 1D 4160G Available Strategies Ill EE-CB-4 I 60G P...fV C1rcu1t Station Service Water General Electric AMH*4 76-250* EE-SWGR-Sw1tctigear RH 932 Nol Mitigation SWPlB Oreaker Pump Lockout ID 41MG Av111lablc Strategics 112 EE-CA-4 I 60G MVCirc:uit Sl.lltion Service Water Siemens 5GEllU-J50-EE-SWGR-Swiichgear Rll 932 CNS Cap>Dcm SWPID Break.er Pump Lockoul 1200-78 4160G Report Ill EE-REL-I GB 51 /\ Protective Phase Overcurre11f General Electric 121AC5JA F.E-SWGR-SW1tchgear RB 932 GERS Mit1gat1on Relay 4160G Stralcgirs 114 EE-REL-I GB 51 B Protechve Phase Oven::urrent General Elecmc 12JAC5JA EE-SWGR-Swttchgear RB 932 GE!lS MillK'!llOO Relay 4160G Strategies tt5 EE-REL-I GB S 1 C' Proteclive Phase Overcurrnnl G1.!ncral Elccmc 12IACS3A EE-SW GR-Swttchgc.3.r "'" 932 liERS Miugauon Relay 4160G Strategies 116 EE-REL-I GB 86 Control 1:eed Lockout General 12HEA61 EE-SWGR-Swttctlgear RB 932 EPRJ lfF Cap> Dem Relay 1160G Test 117 EE-REL-I GE 50-51 P1oleet.1ve Phase Overcunent General Flectnc I 21AC53B812A EE-SW GR-Switchgear RB 932 SQURTS Mitigation A Relay 4160G Report Strategies 118 EE-REL-I GE 50-5 I Procective Phase Overcu1reo1 General Electn\; t21/\C53B812/\ EE-SWGR-Switdigear Ril 932 SQURTS Mitigation B Relay 4160G Report Slr:iteg1cs 119 EE-REL* I GE 50-51 Protective l'hasc Overcuncnt General Electric 121AC53B812A EE-SW GR-Sw.tchgear RB 932 SQURTS M1t1gatton c Kclay 4160G Report Strareg1cs No. *;; "' Devier.ID Type 120 I EE-RE!,.\ GE 86 Control Relay 121 I EE-REL-\ GS II A Protccuve Relay 122 I EE-REL-I GS II B Protective Relay 123 I EE-REL* I GS Protechve Relay 124 I EE-REL-\ GS 86 Control Relay 125 I F.F.-REL-SS I G S0-51 l'rotcct1vc A Relay 126 I EE-REi,.SSIG 50-51 Protective c Relay 127 I EE-RE!,.SWPIB 50-Protective 50-ST A Relay 128 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 Relay 113 I EE-CB-480F MCC-LV Circuit K Breaker \]4 I EE-CB-480F "1CC-LV Ci1cu1t LX Breaker SA 50.54(!) NTTF 2.1 Seismic High Frequency Confirmation Table 8-1: Components Identified for High Frequency Confirmation Cumpunenf Encfo.mre Systnn M*n11f11chtrer Model ID Type Function Bus Lockoul General Electric 12HEi\6\ EE-SWGR* SW'llchgcar 4160G Phase Ovcrcurrcnt Generitl Eleclric 12TAC5JA EE-SWGR* sw;1chgear 4160G Phase Ove<<:urrenl Gcncr::al Elcctnc 121AC5JA EE-SWGR-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 Lockout 4160G Phase Overcurrent General Electnc 12\AC5JB EE-SWGR-SWltchgear 4160G Phase Ovcrcum:nt General F.leclric 121AC53B EE-SWGR-Switchgear 4160G Overcurrent General Elt:"Ctnc 12\AC66K EE-SWGR* Switchgear 4\60G Overcurrcr11 Gtmeral Electric 12TAC66K EE-SWGR-Swi1chgeor 4160G Service Water Pump General Electnc 12HEA6\ EE-SW GR-Lockout 4\60G Phase Overcuuent General Electnc 121AC66K EE-SWGR* Swnchgear 4160G Phase Ovcrcurrent General Electric 121AC66K EE-SWGR-Swttchgcar 4\60G Service Water Pump Genera.I Electric 12HEA6\ EE*SWGR-Switchgear Lockout 4160G MCC Feeder Lockout Wes11nghouse DB-50 EE-SWGR-Swuchgear 480F MCC t'ccdcr Lockout Westinghouse DB-50 EE-SW GR-Switchsear 480F 16C4384-RPT-005 Rev 0 Page 46 of49 Floer Component Evaluation Buildin& i:lev. Basis for [valuation (R) Capaciry Rau It RB 932 EPRJ HF Cap> Dem Test RB 932 GERS Mrtigauon Stm1eg1cs RB 932 GERS Mi11gat1on Strategics RB 932 Gt RS M1t1galloo Stratqt1es RB 932 EPRJ HF Cap> Dem Test RB 932 SQURTS Mitigation Rep on S1rateg1es RB 932 SQ UR TS M1trgallon Rep on Suategics RB 932 EPRJ HF M1t1ga11on 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 CNS Cap>Dttm Letter Rll 932 CNS Cap>Oem Letter No. *;; :;, Device ID Typo 135 I EF.-C'B-4800 MCC* LY Circuit s Breaker 136 I EE-CB-4800 MCC

• LY Circuit TX ll<eakef 50.54(t) NTIF 2.1 Seismic High Frequency Confirmation Table B-1: Components Identified for High Frequency Confirmation Component End Mure S19tem Maauf*cturer Model ID Typo Fuadioo MCC' Feeder Lockout Wcslmghouse DB-SO EE-SWGR-Switchgeot 480G MCC Feeder Lockout Westinghouse DB-50 EE-SWOR-SW1tchgear 4800 I 6C4384-RPT-005 Rev 0 Page 47 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) is normally MS-AOV-738AV 2028 [79) closed at power and deactivated by 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 at power and deactivated by manual valve PC-V-561 (no need to evaluate). MS-AO-A080A 2041 [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-71CRV 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 2043 f35l Simple Check Valve (no need to evaluate). RF-CV-15 2043 [35] Simple Check Valve (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 evaluate). RWCU-MOV-MOl5 2042 Sh. I [821 R WCU Isolation RWCU-MOV-MOL8 2042 Sh. I [821 R WCU Isolation CRD-SOV-SO 120 2039 l83 J Control Rod Manual Positioning CRD-SOV-S012l 2039 [83] Control Rod Manual Positioning SA 50.54(t) NTTF 2.1 Seismic High Frequency Confirmation 16C4384-RPT-005 Rev. 0 Page 49 of49 Table B-2! Reactor Coolant Leak Path Valves Identified for High Frequency Confirmation Valve ID P&ID Comment CRD-SO V-SO 122 2039 [83) Normally Open; would only be a leak path ifCRD-SOV-SO 121 or CRD-SOV-SO 123 docs not close CRD-SOV-SOl23 2039 [83) Control Rod Manual Positioning CRD-AOV-CVl26 2039 [83) Control Rod Scram CS-CV-18 2045 Sh. I [84) Simple Check Valve (no need to evaluate). CS-MOV-MOJ2A 2045 Sh. I [84] Leak path blocked by upstream check valve CS-CV-18 (no need to evaluate). CS-CV-19 2045 Sh. I [84] Simple Check Valve (no need to evaluate). CS-MOV-MOl2B 2045 Sh. I [84] Leak path blocked by upstream check valve CS-CV-19 (no need to evaluate). SLC-CV-13 2045 Sh. 2 (85] Simple Check Valve (no need to evaluate). RF-CV-14 2044 (86) Simple Check Valve (no need to evaluate). RF-CV-13 2044 [86] Simple Check Valve (no need to evaluate). I-IPCl-CV-29 2044 [861 Simple Check Valve (no need to evaluate).