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#REDIRECT [[JAFP-17-0029, Revision to Entergy'S Expedited Seismic Evaluation Process Report (CEUS Sites) Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of Near-Term Task Force Review of Insights...]]
| number = ML17117A699
| issue date = 04/27/2017
| title = James A. Fitzpatrick, Revision to Entergy'S Expedited Seismic Evaluation Process Report (CEUS Sites) Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of Near-Term Task Force Review of Insights
| author name = Pacher J E
| author affiliation = Exelon Generation Co, LLC
| addressee name =
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000333
| license number = DPR-059
| contact person =
| case reference number = JAFP-17-0029
| document type = Letter, Response to Request for Additional Information (RAI)
| page count = 48
| project =
| stage = Response to RAI
}}
 
=Text=
{{#Wiki_filter:JAFP-17-0029 April 27, 2017
 
U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 
 
==Subject:==
Revision to Entergy's Expedited Seismic Evaluation Process Report (CEUS Sites) Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident James A. FitzPatrick Nuclear Power Plant Docket No. 50-333 License No. DPR-059 
 
==Reference:==
: 1. NRC letter, Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(f) Regarding Recommendations 2.1, 2.3, and 9.3 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident, ML12053A340, March 12, 2012 2. NEI letter, Proposed Path Forward for NTTF Recommendation 2.1: Seismic Reevaluations, ML13101A345, dated April 9, 2013 3. ENOI letter, Entergy's Expedited Seismic Evaluation Process Report (CEUS Sites), Response NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident, JAFP 0143, dated December 30, 2014 4. ENOI letter, Entergy's Response to Request for Additional Information for Expedited Seismic Evaluation Process Report, JAFP-15-0094, dated August 4, 2015
 
==Dear Sir or Madam:==
 
On March 12, 2012, the NRC issued a 50.54(f) letter, in Reference 1, requesting that licensees provide interim evaluations and actions taken or planned to address the higher seismic hazard relative to the design basis, as appropriate, per the Near-Term Task Force (NTTF) Recommendation 2.1: Seismic. In Reference 2, the Nuclear Energy Institute (NEI) proposed a path forward for Recommendation 2.1: Seismic. On December 30, 2014, Entergy Nuclear Operations Inc. (ENOI) submitted the Expedited Seismic Evaluation Process Report (ESEP) for JAF in Reference 3. Reference 3 included commitments to complete seismic walkdowns, perform High Confidence of Low Probability of Failure (HCLPF) evaluations, and implement modifications for inaccessible items listed in Section 7.1 of the ESEP. Pursuant to commitments made in Reference 4, this letter summarizes the JAF HCLPF results and confirms implementation of any plant modifications. Exelon Generation Company, LLCJames A. FitzPatrick NPP P.O. Box 110 Lycoming, NY 13093 Tel 315-342-3840 Joseph E. Pacher Site Vice President - JAF JAFP-17-0029 Page 2 of 2 Based on the completed walkdowns during Refueling Outage R22, there were no additional inaccessible items identified. No seismic concerns were identified and no detailed HCLPF evaluations were required as a result of outage walkdowns and HCLPF evaluations. In addition, there are no plant modifications required. The Enclosure contains Revision 1 to the ESEP report. This letter contains no new regulatory commitments. Should you have any questions regarding this submittal, please contact Mr. William C. Drews, Regulatory Assurance Manager at (315) 349-6562. I declare under penalty of perjury that the foregoing is true and correct. Executed on 27th day of April, 2017. Sincerely, JEP/WCD/mh
 
==Enclosure:==
Expedited Seismic Evaluation Process (ESEP) Report for James A. FitzPatrick Nuclear Power Plant (JAF), Revision 1 cc: NRC Regional Administrator NRC Resident Inspector NRC Project Manager NYSPSC NYSE RDA 
 
JAFP-17-0029  Enclosure  Expedited Seismic Evaluation Process (ESEP) Report for  James A. FitzPatrick Nuclear Power Plant (JAF), Revision 1  (45 Pages)
EXPEDITED SEISMIC EVALUATION  PROCESS (ESEP) REPORT FOR JAMES A. FITZPATRICK NUCLEAR POWER PLANT (JAF) Page 1 James A. FitzPatrick ESEP Report  Table of Contents      Page LIST OF TABLES ............................................................................................................................................ 4 LIST OF FIGURES .......................................................................................................................................... 5 1.0 PURPOSE AND OBJECTIVE ............................................................................................................... 6 2.0 BRIEF SUMMARY OF THE FLEX SEISMIC IMPLEMENTATION STRATEGIES ...................................... 6 3.0 EQUIPMENT SELECTION PROCESS AND ESEL.................................................................................. 7 3.1 Equipment Selection Process and ESEL .............................................................................. 7 3.1.1 ESEL Development ............................................................................................... 8 3.1.2 Power Operated Valves ....................................................................................... 9 3.1.3 Pull Boxes ............................................................................................................. 9 3.1.4 Termination Cabinets ........................................................................................... 9 3.1.5 Critical Instrumentation Indicators .................................................................... 10 3.1.6 Phase 2 and 3 Piping Connections ..................................................................... 10 3.2 Justification for Use of Equipment That is Not the Primary Means for FLEX Implementation ................................................................................................................ 10 4.0 GROUND MOTION RESPONSE SPECTRUM (GMRS) ...................................................................... 10 4.1 Plot of GMRS Submitted by the Licensee ......................................................................... 10 4.2 Comparison to SSE ............................................................................................................ 12 5.0 REVIEW LEVEL GROUND MOTION (RLGM) ................................................................................... 13 5.1 Description of RLGM Selected .......................................................................................... 13 5.2 Method to Estimate In-Structure Response Spectra (ISRS) .............................................. 14 6.0 SEISMIC MARGIN EVALUATION APPROACH ................................................................................. 15 6.1 Summary of Methodologies Used .................................................................................... 15 6.2 HCLPF Screening Process .................................................................................................. 15 6.3 Seismic Walkdown Approach ........................................................................................... 16 6.3.1 Walkdown Approach ......................................................................................... 16 6.3.2 Application of Previous Walkdown Information ............................................... 17 6.3.3 Significant Walkdown Findings .......................................................................... 17 6.4 HCLPF Calculation Process ................................................................................................ 17 6.5 Functional Evaluations of Relays ...................................................................................... 18 6.6 Tabulated ESEL HCLPF Values (Including Key Failure Modes) .......................................... 18 7.0 INACCESSIBLE ITEMS ..................................................................................................................... 18 7.1 Identification of ESEL Item Inaccessible for Walkdowns .................................................. 18 7.2 Planned Walkdown / Evaluation Schedule / Close Out .................................................... 18 8.0 ESEP CONCLUSIONS AND RESULTS ............................................................................................... 19 8.1 Supporting Information .................................................................................................... 19 Page 2 James A. FitzPatrick ESEP Report Table of Contents (continued)  Page 8.2 Identification of Planned Modifications ........................................................................... 20 8.3 Modification Implementation Schedule ........................................................................... 20 8.4 Summary of Regulatory Commitments ............................................................................ 20
 
==9.0 REFERENCES==
.................................................................................................................................. 21 ATTACHMENT A - JAMES A. FITZPATRICK ESEL .......................................................................................1 ATTACHMENT B - ESEP HCLPF VALUES AND FAILURE MODES TABULATION ........................................ 1 Page 3 James A. FitzPatrick ESEP Report  List of Tables    Page TABLE 4-1:  GMRS FOR JAMES A. FITZPATRICK ......................................................................................... 10 TABLE 4-2:  SSE FOR JAMES A. FITZPATRICK ............................................................................................. 12 TABLE 5-1:  RLGM FOR JAMES A. FITZPATRICK ......................................................................................... 14    Page 4 James A. FitzPatrick ESEP Report  List of Figures    Page FIGURE 4-1:  GMRS FOR JAMES A. FITZPATRICK ....................................................................................... 12 FIGURE 4-2:  GMRS TO SSE COMPARISON FOR JAMES A. FITZPATRICK ................................................... 13 FIGURE 5-1:  RLGM FOR JAMES A. FITZPATRICK ....................................................................................... 14  Page 5 James A. FitzPatrick ESEP Report  1.0 PURPOSE AND OBJECTIVE Following the accident at the Fukushima Dai-ichi nuclear power plant resulting from the March 11, 2011, Great Tohoku Earthquake and subsequent tsunami, the Nuclear Regulatory Commission (NRC) established a Near-Term Task Force (NTTF) to conduct a systematic review of NRC processes and regulations and to determine if the agency should make additional improvements to its regulatory system. The NTTF developed a set of recommendations intended to clarify and strengthen the regulatory framework for protection against natural phenomena. Subsequently, the NRC issued a 50.54(f) letter on March 12, 2012 [1], requesting information to assure that these recommendations are addressed by all U.S. nuclear power plants. The 50.54(f) letter requests that licensees and holders of construction permits under 10 CFR Part 50 reevaluate the seismic hazards at their sites against present-day NRC requirements and guidance. Depending on the comparison between the reevaluated seismic hazard and the current design basis, further risk assessment may be required. Assessment approaches acceptable to the staff include a seismic probabilistic risk assessment (SPRA), or a seismic margin assessment (SMA). Based upon the assessment results, the NRC staff will determine whether additional regulatory actions are necessary. This report describes the Expedited Seismic Evaluation Process (ESEP) undertaken for James A. FitzPatrick. The intent of the ESEP is to perform an interim action in response to the NRC's 50.54(f) letter to demonstrate seismic margin through a review of a subset of the plant equipment that can be relied upon to protect the reactor core following beyond design basis seismic events. The ESEP is implemented using the methodologies in the NRC endorsed guidance in Electric Power Research Institute (EPRI) 3002000704, Seismic Evaluation Guidance: Augmented Approach for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1: Seismic [2]. The objective of this report is to provide summary information describing the ESEP evaluations and results. The level of detail provided in the report is intended to enable the NRC to understand the inputs used, the evaluations performed, and the decisions made as a result of the interim evaluations. 2.0 BRIEF SUMMARY OF THE FLEX SEISMIC IMPLEMENTATION STRATEGIES The James A. FitzPatrick FLEX strategies for Reactor Core Cooling and Heat Removal, Reactor Inventory Control/Long Term Subcriticality, and Containment Function are summarized below. This summary is derived from the James A. FitzPatrick Overall Integrated Plan (OIP) in Response to the March 12, 2012, Commission Order EA-12-049 [3] and is consistent with the third six-month Status Report [4]. As the ESEP is an interim action to assess available seismic margin against beyond design basis events, equipment selection and the corresponding seismic evaluations are based on FLEX strategies at the time of initial ESEP report submittal to the NRC. Consequently, updates to FLEX strategy after initial ESEP submittal, and changes to supporting plant documentation, are not captured within this report. Core cooling and inventory control are achieved during the first five (5) hours using the Reactor Core Isolation Cooling (RCIC) system initially aligned to take suction from the Condensate Storage Tank (CST), with the suction swapped to the torus when the operators determine the event is a Beyond Design Basis External Event (BDBEE). Pressure control and heat removal are accomplished by Safety Relief Valves (SRVs) venting to the torus. At approximately five (5) hours, suction will be swapped back to the CST for torus temperature control and a controlled depressurization is commenced using RCIC and cycling the SRVs. Page 6 James A. FitzPatrick ESEP Report  At about ten (10) hours (beginning of Phase 2), the operators will need to connect and run a portable 200 kW FLEX diesel generator to the Class 1E 600 VAC electrical buses to re-power the battery chargers to maintain DC control power. At 23 hours, the torus will be vented via the hardened containment vent to maintain containment parameters within acceptable limits and within the limits that support continued use of the RCIC system. The torus and CST will enable the RCIC to provide make-up for at least 35 hours without replacement. Prior to depletion of the CST, James A. FitzPatrick will establish the flow path from the seismically-qualified diesel-driven fire pump 76P-1 to provide makeup directly to the reactor pressure vessel. For Phase 3, the reactor core cooling strategy is to place one loop of RHR into the shutdown cooling mode. This will be accomplished by powering an RHR pump from either Class 1E emergency bus 10500 or Class 1E emergency bus 10600 utilizing the 4160 VAC FLEX portable diesel generator. A modification will be implemented to provide a cross-connection between the fire protection system and one train of the RHR service water system. The seismically qualified diesel-driven fire pump (76P-1) will be used to provide lake water to RHR service water side of the appropriate RHR heat exchanger. Necessary electrical components are outlined in the James A. FitzPatrick FLEX OIP submittal, and primarily entail 125 VDC power buses, motor control centers, vital batteries, battery chargers, 600 VAC buses, and 4160 VAC buses. The FLEX strategy credits the monitoring of plant parameters, either from the control room, using available electric power supplied from the batteries or taken locally. If instrumentation is to be monitored from the control room, it will be powered from 125 VDC either directly or through (future) inverters for 120 VAC to some instruments. Figures 1 and 2 in the James A. FitzPatrick FLEX OIP submittal [3] provide the FLEX flow paths for James A. FitzPatrick Phases 1, 2 and 3. 3.0 EQUIPMENT SELECTION PROCESS AND ESEL The selection of equipment for the Expedited Seismic Equipment List (ESEL) followed the guidelines of EPRI 3002000704 [2]. The ESEL for James A. FitzPatrick is presented in Attachment A. Information presented in Attachment A is drawn from the following references [3], [4], [5], [6], [7], [8], [9], [10], [11], [12], [13], [14], [15], [16], [17], [18], [19], [20], [21], [22], [23], [24], [25], [26], [27], [28], [29], [30], [31], [32], [33], [34], [35], [36], [37], and [38]. 3.1 Equipment Selection Process and ESEL The selection of equipment to be included on the ESEL was based on installed plant equipment credited in the FLEX strategies during Phase 1, 2 and 3 mitigation of a BDBEE, as outlined in the James A. FitzPatrick OIP in Response to the March 12, 2012, Commission Order EA-12-049 [3] and is consistent with the third six-month Status Report [4]. The OIP provides the James A. FitzPatrick FLEX mitigation strategy and serves as the basis for equipment selected for the ESEP. The scope of "installed plant equipment" includes equipment relied upon for the FLEX strategies to sustain the critical functions of core cooling and containment integrity consistent with the James A. FitzPatrick OIP. FLEX recovery actions are excluded from the ESEP scope per EPRI 3002000704 [2]. The overall list of planned FLEX modifications and the scope for consideration herein is limited to those required to support core cooling, reactor coolant inventory and subcriticality, and containment Page 7 James A. FitzPatrick ESEP Report  integrity functions. Portable and pre-staged FLEX equipment (not permanently installed) are excluded from the ESEL per EPRI 3002000704. The ESEL component selection followed the EPRI guidance outlined in Section 3.2 of EPRI 3002000704. 1. The scope of components is limited to that required to accomplish the core cooling and containment safety functions identified in Table 3-2 of EPRI 3002000704. The instrumentation monitoring requirements for core cooling/containment safety functions are limited to those outlined in the EPRI 3002000704 guidance, and are a subset of those outlined in the James A. FitzPatrick OIP. 2. The scope of components is limited to installed plant equipment, and FLEX connections necessary to implement the James A. FitzPatrick OIP as described in Section 2. 3. The scope of components assumes the credited FLEX connection modifications are implemented, and are limited to those required to support a single FLEX success path (i.e.,
either "Primary" or "Back-up/Alternate"). 4. The "Primary" FLEX success path is to be specified. Selection of the "Back-up/Alternate" FLEX success path must be justified. 5. Phase 3 coping strategies are included in the ESEP scope, whereas recovery strategies are excluded. 6. Structures, systems, and components excluded per the EPRI 3002000704 [2] guidance are:  Structures (e.g. containment, reactor building, control building, auxiliary building, etc.). Piping, cabling, conduit, HVAC, and their supports. Manual valves, check valves and rupture disks. Power-operated valves not required to change state as part of the FLEX mitigation strategies. Nuclear steam supply system components (e.g. RPV and internals, reactor coolant pumps and seals, etc.). 7. For cases in which neither train was specified as a primary or back-up strategy, then only one train component (generally 'A' train) is included in the ESEL. 3.1.1 ESEL Development The ESEL was developed by reviewing the James A. FitzPatrick OIP [3] to determine the major equipment involved in the FLEX strategies. Further reviews of plant drawings (e.g., Piping and Instrumentation Diagrams (P&IDs) and Electrical One Line Diagrams) were performed to identify the boundaries of the flow paths to be used in the FLEX strategies and to identify specific components in the flow paths needed to support implementation of the FLEX strategies. Boundaries were established at an electrical or mechanical isolation device (e.g., isolation amplifier, valve, etc.) in branch circuits /
branch lines off the defined strategy electrical or fluid flow path. P&IDs were the primary reference documents used to identify mechanical components and instrumentation. The flow paths used for FLEX strategies were selected and specific components were identified using detailed equipment and instrument drawings, piping isometrics, electrical schematics and one-line drawings, system descriptions, design basis documents, as necessary. Page 8 James A. FitzPatrick ESEP Report  Cabinets containing electrical and instrumentation that could be affected by earthquake motions and that impact the operation of equipment in the ESEL are required to be on the ESEL. These cabinets and components were identified in the ESEL. For Phase 1, RCIC is the primary path for inventory control and core cooling. Therefore, the RCIC system was used as the basis for the Phase 1 and 2 ESEL. For Phase 2 and Phase 3, the RHR is used to provide the pathway for reactor pressure vessel injection utilizing the seismic grade fire pump or portable injection pumps. For each parameter monitored during the FLEX implementation, a single indication was selected for inclusion in the ESEL. For each parameter indication, the components along the flow path from measurement to indication were included, since any failure along the path would lead to failure of that indication. Components such as flow elements were considered as part of the piping and were not included in the ESEL. 3.1.2 Power Operated Valves Page 3-3 of EPRI 3002000704 [2] notes that power operated valves not required to change state as part of the FLEX mitigation strategies are excluded from the ESEL. Page 3-2 also notes that "functional failure modes of electrical and mechanical portions of the installed Phase 1 equipment should be considered (e.g., RCIC)."  To address this concern, the following guidance is applied in the James A. FitzPatrick ESEL for functional failure modes associated with power operated valves:  Power operated valves that remain energized during the Extended Loss of AC Power (ELAP) events (such as DC powered valves), were included on the ESEL. Power operated valves not required to change state as part of the FLEX mitigation strategies were not included on the ESEL. The seismic event also causes the ELAP event; therefore, the valves are incapable of spurious operation as they would be de-energized. Power operated valves not required to change state as part of the FLEX mitigation strategies during Phase 1, and are re-energized and operated during subsequent Phase 2 and 3 strategies, were not evaluated for spurious valve operation as the seismic event that caused the ELAP has passed before the valves are re-powered. 3.1.3 Pull Boxes Pull boxes were deemed unnecessary to be added to the ESEL as these components provide completely passive locations for pulling or installing cables. No breaks or connections in the cabling were included in pull boxes. Pull boxes were considered part of conduit and cabling, which were excluded in accordance with EPRI 3002000704 [2]. 3.1.4 Termination Cabinets Termination cabinets, including cabinets necessary for FLEX Phase 2 and Phase 3 connections, provide consolidated locations for permanently connecting multiple cables. The termination cabinets and the internal connections provide a completely passive function; however, the cabinets are included in the ESEL to ensure industry knowledge on panel/anchorage failure vulnerabilities is addressed. Page 9 James A. FitzPatrick ESEP Report  3.1.5 Critical Instrumentation Indicators Critical indicators and recorders are typically physically located on panels/cabinets and are included as separate components; however, seismic evaluation of the instrument indication may be included in the panel/cabinet seismic evaluation (rule-of-the-box). 3.1.6 Phase 2 and 3 Piping Connections Item 2 in Section 3.1 above notes that the scope of equipment in the ESEL includes "- FLEX connections necessary to implement the James A. FitzPatrick OIP [3] as described in Section 2."  Item 3 in Section 3.1 also notes that "The scope of components assumes the credited FLEX connection modifications are implemented, and are limited to those required to support a single FLEX success path (i.e., either "Primary" or "Back-up/Alternate")." Item 6 in Section 3.1 above goes on to explain that "Piping, cabling, conduit, HVAC, and their supports" are excluded from the ESEL scope in accordance with EPRI 3002000704 [2]. Therefore, piping and pipe supports associated with FLEX Phase 2 and Phase 3 connections are excluded from the scope of the ESEP evaluation. However, any active valves in FLEX Phase 2 and Phase 3 connection flow path are included in the ESEL. 3.2 Justification for Use of Equipment That is Not the Primary Means for FLEX Implementation RCIC is the primary system for Phases 1 and 2 and is presented as the single success path in the James A. FitzPatrick ESEL. RHR, Train A, is the primary system for Phase 3. Therefore, no additional justification is required. 4.0 GROUND MOTION RESPONSE SPECTRUM (GMRS) 4.1 Plot of GMRS Submitted by the Licensee The Safe Shutdown Earthquake (SSE) control point elevation is defined at depth 12 ft, which is the top of the Oswego sandstone where all plant structures are founded [39]. Table 4-1 shows the GMRS acceleration for a range of spectral frequencies [40]. The GMRS at the control point is shown in Figure 4-1. Table 4-1:  GMRS for James A. FitzPatrick Frequency (Hz) GMRS (g) 100 1.20E-01 90 1.20E-01 80 1.22E-01 70 1.25E-01 60 1.33E-01 50 1.55E-01 40 1.87E-01 35 1.99E-01 Page 10 James A. FitzPatrick ESEP Report  Table 4-1:  GMRS for James A. FitzPatrick (continued) Frequency (Hz) GMRS (g) 30 2.09E-01 25 2.16E-01 20 2.31E-01 15 2.41E-01 12.5 2.39E-01 10 2.33E-01 9 2.26E-01 8 2.15E-01 7 2.04E-01 6 1.88E-01 5 1.70E-01 4 1.44E-01 3.5 1.27E-01 3 1.14E-01 2.5 9.44E-02 2 8.47E-02 1.5 7.50E-02 1.25 7.13E-02 1 6.38E-02 0.9 6.05E-02 0.8 5.66E-02 0.7 5.11E-02 0.6 4.47E-02 0.5 3.76E-02 0.4 3.00E-02 0.35 2.63E-02 0.3 2.25E-02 0.25 1.88E-02 0.2 1.50E-02 0.15 1.13E-02 0.125 9.39E-03 0.1 7.51E-03 Page 11 James A. FitzPatrick ESEP Report      Figure 4-1:  GMRS for James A. FitzPatrick 4.2 Comparison to SSE The SSE corresponds to a horizontal acceleration of 0.15g. The SSE is defined in Figure 2.6-2 of the FSAR [39] in terms of a Peak Ground Acceleration (PGA) and a design response spectrum. These spectra have been digitized and tabulated [40] [41]. Table 4-2 shows the spectral acceleration values at selected frequencies for the 5% damped horizontal SSE. Table 4-2:  SSE for James A. FitzPatrick Frequency (Hz) Spectral Acceleration (g) 100 0.15 25 0.15 10 0.15 5 0.21 2.5 0.22 1 0.13 0.5 0.064  0.000.050.100.150.200.250.300.1110100SA (g)Frequency (Hz)GMRS at Control Point for James A. FitzPatrick Nuclear Power Plant, 5% DampingJAF GMRSPage 12 James A. FitzPatrick ESEP Report    Figure 4-2:  GMRS to SSE Comparison for James A. FitzPatrick The SSE envelops the GMRS for lower frequencies up to nearly 6 Hz. The GMRS exceeds the SSE beyond that point. As the GMRS exceeds the SSE in the 1 to 10 Hz range, the plant does not screen out of the ESEP according to Section 2.2 of EPRI 3002000704 [2]. The two special screening considerations as described in Section 2.2.1 of EPRI 3002000704, namely a) Low-frequency GMRS exceedances at Low Seismic Hazard Sites and b) Narrow Band Exceedances in the 1 to 10 Hz range, provide criteria for accepting specific GMRS exceedances. However, the GMRS exceedances are not limited to the low frequency range and there are no narrow-banded exceedances. Therefore, these special screening considerations do not apply for James A. FitzPatrick and hence High Confidence of a Low Probability of Failure (HCLPF) evaluations are to be performed. 5.0 REVIEW LEVEL GROUND MOTION (RLGM) 5.1 Description of RLGM Selected The RLGM is selected based on Approach 1 in Section 4 of EPRI 3002000704 [2]. The RLGM is developed based on the SSE. The maximum GMRS/SSE ratio between 1 and 10 Hz range occurs at 10 Hz where the ratio is 0.233/0.15 = 1.55. The GMRS/SSE ratio is set to the scaling factor value of 1.55 for James A. FitzPatrick in accordance with Section 4 of EPRI 3002000704. Table 5-1 lists the horizontal ground RLGM acceleration at 5% damping at selected frequencies and the plot is shown in Figure 5-1. The RLGM is generated by plotting the digitized data on a log/linear graph paper, and connecting the points with straight lines. Page 13 James A. FitzPatrick ESEP Report  Table 5-1:  RLGM for James A. FitzPatrick Frequency (Hz) RLGM at 5% Damping  (g) 0.50 0.099 1.00 0.202 2.50 0.342 5.00 0.326 10.00 0.233 25.00 0.233 100.00 0.233  Figure 5-1:  RLGM for James A. FitzPatrick 5.2 Method to Estimate In-Structure Response Spectra (ISRS) The RLGM ISRS for James A. FitzPatrick are generated by scaling the SSE ISRS [39]. The following steps are used to generate the RLGM ISRS. 1. Obtain the horizontal direction SSE ISRS for a particular damping value. 2. Calculate the horizontal RLGM ISRS by scaling the horizontal direction SSE ISRS by a factor of 1.55. 3. Repeat steps 1 and 2 to obtain RLGM ISRS for multiple damping values. The vertical direction RLGM ISRS is generated by repeating steps 1-3 above using vertical direction SSE ISRS as input for multiple damping values and elevations. Page 14 James A. FitzPatrick ESEP Report  6.0 SEISMIC MARGIN EVALUATION APPROACH It is necessary to demonstrate that ESEL items have sufficient seismic capacity to meet or exceed the demand characterized by the RLGM. The seismic capacity is characterized as the PGA for which there is a HCLPF. The PGA is associated with a specific spectral shape, in this case the 5%-damped RLGM spectral shape. The HCLPF capacity must be equal to or greater than the RLGM PGA. The criteria for seismic capacity determination are given in Section 5 of EPRI 3002000704 [2]. There are two basic approaches for developing HCLPF capacities: 1. Deterministic approach using the conservative deterministic failure margin (CDFM) methodology of EPRI NP-6041-SL, A Methodology for Assessment of Nuclear Power Plant Seismic Margin (Revision 1) [42]. 2. Probabilistic approach using the fragility analysis methodology of EPRI TR-103959, Methodology for Developing Seismic Fragilities [43]. 6.1 Summary of Methodologies Used James A. FitzPatrick performed a 0.3g focused-scope SMA in accordance with the methodology of NUREG-1407[44] in 1996 as part of Individual Plant Examination of External Events (IPEEE) program. The SMA is documented in [45] and consisted of screening evaluations, seismic walkdowns, and a review of the plant seismic design basis. The SMA was performed in accordance with EPRI NP-6041-SL [42]. The evaluation of mechanical and electrical equipment relied heavily on the walkdowns conducted for the USI A-46 seismic evaluation. Section 3.3 and Appendix B of [40] established that the results of the James A. FitzPatrick IPEEE are adequate to support screening of the updated seismic hazard for James A. FitzPatrick. Consequently, for ESEP, the results of HCLPF evaluations performed for IPEEE are used to screen out components with capacity that exceeds RLGM. For ESEP, the SMA consisted of screening walkdowns and HCLPF calculations. The screening walkdowns used the screening tables from Chapter 2 of EPRI NP-6041-SL. The walkdowns were conducted by engineers trained in EPRI NP-6041-SL and were documented on Screening Evaluation Work Sheets (SEWS) from EPRI NP-6041-SL. Anchorage capacity calculations used the CDFM criteria from EPRI NP-6041-SL. Seismic demand was based on EPRI 3002000704 [2] using an RLGM of 1.55xSSE with a PGA of 0.233g, Figure 5-1. 6.2 HCLPF Screening Process For ESEP, the components are screened considering the RLGM (1.55xSSE) with a 0.233g PGA. The screening tables in EPRI NP-6041-SL [42] are based on ground peak spectral accelerations of 0.8g and 1.2g. These both exceed the RLGM peak spectral acceleration. The ESEL components were prescreened based on Table 2-4 of EPRI NP-6041-SL [42]. Additional pre-screening, specifically for anchorage, considered walkdown results and documentation from NTTF 2.3 and SEWS from IPEEE and USI A-46. Equipment anchorage was screened out in cases where previous evaluations showed large available margin against SSE. The remaining components (i.e., components that do not screen out), were identified as requiring HCLPF calculations. ESEL components were walked down and based on the equipment and anchorage conditions, prescreening decisions were confirmed and a final list of required HCLPF calculations was generated. Equipment for which the screening caveats were met and for which the anchorage capacity exceeded the RLGM seismic demand Page 15 James A. FitzPatrick ESEP Report  are screened out from ESEP seismic capacity determination because the HCLPF capacity exceeds the RLGM. The James A. FitzPatrick ESEL contains 145 items. Of these, 45 are valves. In accordance with Table 2-4 of EPRI NP-6041-SL [42], active valves may be assigned a functional capacity of 0.8g peak spectral acceleration without any review other than looking for valves with large extended operators on small diameter piping, and anchorage is not a failure mode. Therefore, valves on the ESEL are screened out from ESEP seismic capacity determination, subject to the caveat regarding large extended operators on small diameter piping. The non-valve components in the ESEL are screened based on the SMA results. If the SMA showed that the component met the EPRI NP-6041-SL [42] screening caveats and the CDFM capacity exceeded the RLGM demand, the components are screened out from the ESEP capacity determination. Additionally, items with HCLPF capacities greater than RLGM that were calculated in [45] were also screened out. Block walls located in the proximity of ESEL equipment were assessed for potential seismic interaction impact resulting from the RLGM by reviewing the existing plant documents and found to be acceptable. 6.3 Seismic Walkdown Approach 6.3.1 Walkdown Approach Walkdowns were performed in accordance with the criteria provided in Section 5 of EPRI 3002000704 [2], which refers to EPRI NP-6041-SL [42] for the Seismic Margin Assessment process. Pages 2-26 through 2-30 of EPRI NP-6041-SL [42] describe the seismic walkdown criteria, including the following key criteria. "The SRT [Seismic Review Team] should "walk by" 100% of all components which are reasonably accessible and in non-radioactive or low radioactive environments. Seismic capability assessment of components which are inaccessible, in high-radioactive environments, or possibly within contaminated containment, will have to rely more on alternate means such as photographic inspection, more reliance on seismic reanalysis, and possibly, smaller inspection teams and more hurried inspections. A 100% "walk by" does not mean complete inspection of each component, nor does it mean requiring an electrician or other technician to de-energize and open cabinets or panels for detailed inspection of all components. This walkdown is not intended to be a QA or QC review or a review of the adequacy of the component at the SSE level. If the SRT has a reasonable basis for assuming that the group of components are similar and are similarly anchored, then it is only necessary to inspect one component out of this group. The "similarity-basis" should be developed before the walkdown during the seismic capability preparatory work (Step 3) by reference to drawings, calculations or specifications. The one component or each type which is selected should be thoroughly inspected which probably does mean de-energizing and opening cabinets or panels for this very limited sample. Generally, a spare representative component can be found so as to enable the inspection to be performed while the plant is in operation. At least for the one component of each type which is selected, anchorage should be thoroughly inspected. Page 16 James A. FitzPatrick ESEP Report  The walkdown procedure should be performed in an ad hoc manner. For each class of components the SRT should look closely at the first items and compare the field configurations with the construction drawings and/or specifications. If a one-to-one correspondence is found, then subsequent items do not have to be inspected in as great a detail. Ultimately the walkdown becomes a "walk by" of the component class as the SRT becomes confident that the construction pattern is typical. This procedure for inspection should be repeated for each component class; although, during the actual walkdown the SRT may be inspecting several classes of components in parallel. If serious exceptions to the drawings or questionable construction practices are found then the system or component class must be inspected in closer detail until the systematic deficiency is defined. The 100% "walk by" is to look for outliers, lack of similarity, anchorage which is different from that shown on drawings or prescribed in criteria for that component, potential SI [Seismic Interaction] problems, situations that are at odds with the team members' past experience, and any other areas of serious seismic concern. If any such concerns surface, then the limited sample size of one component of each type for thorough inspection will have to be increased.
The increase in sample size which should be inspected will depend upon the number of outliers and different anchorages, etc., which are observed. It is up to the SRT to ultimately select the sample size since they are the ones who are responsible for the seismic adequacy of all elements which they screen from the margin review. Appendix D gives guidance for sampling selection." 6.3.2 Application of Previous Walkdown Information Several ESEL items were previously walked down during the James A. FitzPatrick seismic IPEEE program, for the USI A-46 evaluation program, and NTTF Recommendation 2.3. Those walkdown results were reviewed and the following steps were taken to confirm that the previous walkdown conclusions remained valid. A walk by was performed to confirm that the equipment material condition and configuration is consistent with the walkdown conclusions and that no new significant interactions related to block walls or piping attached to tanks exist. If the ESEL item was screened out based on previous walkdowns, that screening evaluation was reviewed and reconfirmed for the ESEP. 6.3.3 Significant Walkdown Findings Consistent with the guidance from EPRI NP-6041-SL [42], no significant outliers or anchorage concerns were identified during the James A. FitzPatrick seismic walkdowns. Based on walkdown results, no HCLPF capacity evaluations were required. 6.4 HCLPF Calculation Process ESEL items identified for ESEP at James A. FitzPatrick were evaluated using the criteria in EPRI NP-6041-SL [42] and Section 5 of EPRI 3002000704 [2]. Those evaluations included the following steps:  Performing seismic capability walkdowns for equipment not included in previous seismic walkdowns (USI A-46, IPEEE, or NTTF 2.3) to evaluate the equipment installed plant conditions  Performing screening evaluations using the screening tables in EPRI NP-6041-SL as described in Section 6.2 Page 17 James A. FitzPatrick ESEP Report  6.5 Functional Evaluations of Relays No seal in /lockout type relays were identified on James A. FitzPatrick ESEL. Therefore, no relay evaluations were performed. 6.6 Tabulated ESEL HCLPF Values (Including Key Failure Modes) Tabulated ESEL HCLPF values are provided in Attachment B. The following notes apply to the information in the tables. For items screened out using EPRI NP-6041-SL [42] screening tables, the HCLPF capacity is provided as >RLGM and the failure mode is listed as "Screened", (unless the controlling HCLPF value is governed by anchorage). For items where anchorage controls the HCLPF value, the HCLPF value is listed in the table and the failure mode is noted as "anchorage."  For the items where the component function controls the HCLPF value, the HCLPF value is listed in the table and the failure mode is noted as "functional." ESEL components were determined to have adequate capacity for the design basis loads and HCLPF greater than RLGM. 7.0 INACCESSIBLE ITEMS 7.1 Identification of ESEL Item Inaccessible for Walkdowns Forty-four (44) components on the ESEL were previously identified as inaccessible and not walked down. All of these components were subsequently waked down, by plant personnel or the SRT, prior to or during the R22 refueling outage in January 2017. The results of the subsequent walkdowns and evaluations have been captured and summarized in Attachment B of this report. No seismic concerns were identified and no detailed HCLPF evaluations were required as a result of outage walkdowns and evaluations. 7.2 Planned Walkdown / Evaluation Schedule / Close Out No follow up walkdowns are required. Page 18 James A. FitzPatrick ESEP Report  8.0 ESEP CONCLUSIONS AND RESULTS 8.1 Supporting Information James A. FitzPatrick has performed the ESEP as an interim action in response to the NRC's 50.54(f) letter [1]. It was performed using the methodologies in the NRC endorsed guidance in EPRI 3002000704 [2]. The ESEP provides an important demonstration of seismic margin and expedites plant safety enhancements through evaluations and potential near-term modifications of plant equipment that can be relied upon to protect the reactor core following beyond design basis seismic events. The ESEP is part of the overall James A. FitzPatrick response to the NRC's 50.54(f) letter. On March 12, 2014, NEI submitted to the NRC results of a study [47] of seismic core damage risk estimates based on updated seismic hazard information as it applies to operating nuclear reactors in the Central and Eastern United States (CEUS). The study concluded that "site-specific seismic hazards show that there has not been an overall increase in seismic risk for the fleet of U.S. plants" based on the re-evaluated seismic hazards. As such, the "current seismic design of operating reactors continues to provide a safety margin to withstand potential earthquakes exceeding the seismic design basis."  The NRC's May 9, 2014 NTTF 2.1 Screening and Prioritization letter [46] concluded that the "fleet wide seismic risk estimates are consistent with the approach and results used in the Gl-199 safety/risk assessment."  The letter also stated that "As a result, the staff has confirmed that the conclusions reached in Gl-199 safety/risk assessment remain valid and that the plants can continue to operate while additional evaluations are conducted." An assessment of the change in seismic risk for James A. FitzPatrick was included in the fleet risk evaluation submitted in the March 12, 2014 NEI letter [47] therefore, the conclusions in the NRC's May 9 letter also apply to James A. FitzPatrick. In addition, the March 12, 2014 NEI letter provided an attached "Perspectives on the Seismic Capacity of Operating Plants," which (1) assessed a number of qualitative reasons why the design of Structures, Systems and Components (SSCs) inherently contain margin beyond their design level, (2) discussed industrial seismic experience databases of performance of industry facility components similar to nuclear SSCs, and (3) discussed earthquake experience at operating plants. The fleet of currently operating nuclear power plants was designed using conservative practices, such that the plants have significant margin to withstand large ground motions safely. This has been borne out for those plants that have actually experienced significant earthquakes. The seismic design process has inherent (and intentional) conservatisms, which result in significant seismic margins within SSCs. These conservatisms are reflected in several key aspects of the seismic design process, including:  Safety factors applied in design calculations  Damping values used in dynamic analysis of SSCs  Bounding synthetic time histories for in-structure response spectra calculations  Broadening criteria for in-structure response spectra  Response spectra enveloping criteria typically used in SSC analysis and testing applications  Response spectra based frequency domain analysis rather than explicit time history based time domain analysis Page 19 James A. FitzPatrick ESEP Report    Bounding requirements in codes and standards  Use of minimum strength requirements of structural components (concrete and steel)  Bounding testing requirements  Ductile behavior of the primary materials (that is, not crediting the additional capacity of materials such as steel and reinforced concrete beyond the essentially elastic range, etc.) These design practices combine to result in margins such that the SSCs will continue to fulfill their functions at ground motions well above the SSE. Based on the results of the screening evaluation performed in [40], James A. FitzPatrick screens-out of a risk evaluation. The NRC Screening and Prioritization Results letter concluded James A. FitzPatrick conditionally screens-in for the seismic risk evaluation [46] for the purpose of prioritizing and conducting additional evaluations. Consistent with [40] and detailed in this submittal, the IPEEE HCLPF Spectrum (IHS) bound the GMRS in the 1 Hz to 10 Hz range [48]. Upon further evaluation of information provided by Entergy, The NRC concluded that the IHS could be used for comparison with the GMRS for the screening decision. A seismic risk evaluation is not merited [49]. Contingent upon NRC staff review and acceptance of Entergy's full scope IPEEE relay chatter review and spent fuel pool evaluation, the seismic hazard evaluation identified in enclosure 1 of the 50.54(f) letter [1] will be completed. 8.2 Identification of Planned Modifications Insights from the ESEP identified that there is no plant modification required. 8.3 Modification Implementation Schedule There is no plant modification required. 8.4 Summary of Regulatory Commitments No follow up actions or regulatory commitments are required for ESEP.  [50]  Page 20 James A. FitzPatrick ESEP Report 
 
==9.0 REFERENCES==
: 1. NRC (E Leeds and M Johnson) Letter to All Power Reactor Licensees et al., "Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(f) Regarding Recommendations 2.1, 2.3 and 9.3 of the Near-Term Task Force Review of Insights from the Fukushima Dai-Ichi Accident," March 12, 2012, NRC ADAMS Accession No. ML12053A340. 2. EPRI 3002000704, "Seismic Evaluation Guidance, Augmented Approach for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1: Seismic," May 2013. 3. Entergy Letter to U.S. NRC, letter number JAFP-13-0025 "Overall Integrated Plan in Response to March 12, 2012, Commission Order to Modify Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events (Order Number EA-12-049)," February 28, 2013, NRC ADAMS Accession No. ML13063A287. 4. Entergy Letter to U.S. NRC, letter number JAFP-14-0105, "Third Six-Month Status Report in Response to March 12, 2012 Commission Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond Design Basis External Events (Order Number EA-12-049)," August 28, 2014, NRC ADAMS Accession No. ML14241A261. 5. Entergy Drawing FM-22A, Rev. 56, "Flow Diagram, Reactor Core Isolation Cooling, System 13." 6. Entergy Drawing FM-29A, Rev. 58, "Flow Diagram, Main Steam, System 29." 7. Entergy Drawing FM-47A, Rev. 52, Flow Diagram, Nuclear Boiler Vessel Instruments, System 02-3." 8. Entergy Drawing FM-18A, Rev. 57, "Flow Diagram, Drywell Inserting C.A.D. and Purge, System 27." 9. Entergy Drawing FM-20A, Rev. 72, "Flow Diagram, Residual Heat Removal, System 10." 10. Entergy Drawing FM-20B, Rev. 72, "Flow Diagram, Residual Heat Removal, System 10." 11. Entergy Drawing FE-1AH, Rev. 32, "125V DC One Line Diagram, Sheet 1." 12. Entergy Drawing FE-1AJ, Rev. 21, "125V DC One Line Diagram, Sheet 2." 13. Entergy Drawing FE-1AL, Rev. 28, "125V DC One Line Diagram, Sheet 4." 14. Entergy Drawing FE-1AX, Rev. 20, "125V DC One Line Diagram, Sheet 7." 15. Entergy Drawing FE-1H, Rev. 14, "4160V One Line Diagram, Sh. 4, Emergency Bus 10500." 16. Entergy Drawing FE-1BH, Rev. 11, "600V One Line Diagram, Sh.17, 71MCC-156 & 71MCC-166." 17. Entergy Drawing FE-1R, Rev. 29, "600V One Line Diagram, Sh.7, 71MCC-131, 141, 252, & 262." 18. Entergy Drawing FE-1Z, Rev. 26, "600V One Line Diagram, Sh.15, 71MCC-253, 263, 254, & 264." 19. Entergy Drawing FE-3DD, Rev. 16, "External Connections, Residual Heat Removal Panel 09-32, Sh. 2, System 10." 20. Entergy Drawing SE-9NM, Rev. 25, "Distribution Panel 71ACA2 Emergency Control & Instrument Bus A2." 21. Entergy Drawing SE-9PL, Rev. 10, "71UPP Uninterruptable Power Supply UPS Static Inverter." 22. Entergy Drawing SE-11A, Rev. 17, "Distribution Panel 71ACAUPS Uninterruptable Power." Page 21 James A. FitzPatrick ESEP Report  23. Entergy Drawing SE-11D, Rev. 11, "Distribution Panel 71ESSA1 Safeguard Control & Instrument Bus A1." 24. Entergy Drawing 1.61-154, Rev. 13, "Elem Diag RCIC Sys." 25. Entergy Drawing 1.61-156, Rev. 6, "Elem Diag RCIC Sys." 26. Entergy Drawing 1.49-164, Rev. 1, "DC/AC Inverter 71-INV-1A & 1B Schematic Diagram." 27. Entergy Drawing LP-02-3AD, Rev. 2, "LOOP Diagram, NBI Reactor Wide Range Level Transmitter." 28. Entergy Drawing LP-02-3AA, Rev. 1, "LOOP Diagram, Reactor Vessel Shroud Level, NBI/RHR Interlock Level." 29. Entergy Drawing LP-33-209, Rev. 3, "Condensate Storage Tanks 12A & 12B Level." 30. Entergy Drawing LP-06A, Rev. 2, "LOOP Diagram FWC ECCS Monitor, Reactor Pressure." 31. Entergy Drawing LP-27-115A1, Rev. 2, "I&C LOOP Diagram Drywell Pressure (NR) (Div. I - RED)." 32. Entergy Drawing LP-27-115A2, Rev. 2, "I&C LOOP Diagram Drywell Pressure (WR) (Div. I - RED)." 33. Entergy Drawing LP-27-118, Rev. 4, "LOOP Diagram Reactor Building Suppression Chamber Pressure." 34. Entergy Drawing LP-16-1-60, Rev. 3, "Reactor Building Drywell Temperature A." 35. Entergy Drawing LP-16-1-50, Rev. 4, "Reactor Building Suppression Pool Temperature A." 36. Entergy Drawing LP-23AJ, Rev. 2, "LOOP Diagram HPCI, Containment Wide Range Level." 37. Entergy Drawing LP-23AG, Rev. 4, "LOOP Diagram HPCI, Suppression Pool Water." 38. Entergy Drawing FB-48A, Rev. 34, "Flow Diagram, Fire Protection Water Piping, System 76." 39. "James A. FitzPatrick Nuclear Power Plant FSAR Update," Docket No. 50-333, 2013. 40. Entergy Letter to NRC, letter number JAFP-14-0039, "Entergy's Seismic Hazard and Screening Report (CEUS Sites), Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident," March 31, 2014, NRC ADAMS Accession No. ML14090A243. 41. EPRI Document, "Fitzpatrick Seismic Hazard and Screening Report," Revision 1, February 27, 2014. 42. EPRI-NP-6041-SL, "Methodology for Assessment of Nuclear Power Plant Seismic Margin," Revision 1, August 1991. 43. EPRI TR-103959, "Methodology for Developing Seismic Fragilities," July 1994. 44. NRC NUREG-1407, "Procedural and Submittal Guidance for the Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities," June 1991. 45. Entergy Document JAF-RPT-MISC-02211, "James A. FitzPatrick Nuclear Power Plant Individual Plant Examination of External Events," Revision 0, June 1996. 46. NRC (E. Leeds) Letter to All Power Reactor Licensees et al., "Screening and Prioritization Results Regarding Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(F) Page 22 James A. FitzPatrick ESEP Report  Regarding Seismic Hazard Re-Evaluations for Recommendation 2.1 of the Near-Term Task Force Review of Insights From the Fukushima Dai-Ichi Accident," May 9, 2014, NRC ADAMS Accession No. ML14111A147. 47. Nuclear Energy Institute (NEI), A. Pietrangelo, Letter to D. Skeen of the USNRC, "Seismic Core Damage Risk Estimates Using the Updated Seismic Hazards for the Operating Nuclear Plants in the Central and Eastern United States," March 12, 2014. 48. Memorandum to David Skeens, Director, Japan Lessons Learned Project Directorate, Office of Nuclear Reactor Regulation from Scott Flanders, Director, Division of Site Safety and Environmental Analysis, Office of New Reactors,
 
==Subject:==
"Support Document for Screening and Prioritization Results Regarding Seismic Hazard Re-Evaluations for Operating Reactors in the Central and Eastern United States," May 21, 2014, NRC ADAMS Accession No. ML14136A126. 49. NRC (F. Vega) Letter to Vice President, Operations, James A. FitzPatrick Nuclear Power Plant, "James A. FitzPatrick Nuclear Power Plant - Staff Assessment of Information Provided Pursuant to Title 10 of the Code of Federal Regulations Part 50, Section 50.54(f), Seismic Hazard Reevaluations for Recommendation 2.1 of the Near-Term Task Force Review of Insights From the Fukushima Dai-ichi Accident (CAC No. MF3725)" Feburary 18, 2016, NRC ADAMS Accession No. ML16043A411. 50. Entergy Document EC52427, "Fukushima - Acceptance of Expedited Seismic Evaluation Program (ESEP) Documentation," the following AREVA documents are captured in the plant document management system: a. AREVA Document 51-9219585-003, "ESEP Expedited Seismic Equipment List (ESEL) - James A. FitzPatrick Nuclear Power Plant." Page 23 James A. FitzPatrick ESEP Report  ATTACHMENT A - JAMES A. FITZPATRICK ESEL    Page A-1 James A. FitzPatrick ESEP Report  ESEL Item Number Equipment Operating State Notes/Comments References ID Description Normal State Desired State 1 13P-1 RCIC Turbine Driven Pump Off On Powered from 125VDC (71DC-A2) [5] 2 13E-1 RCIC Barometric Condenser Available Available - [5] 3 13E-2 RCIC Turbine Lube Oil Cooler Available Available - [5] 4 13TK-1 RCIC Vacuum Tank Available Available - [5] 5 13MOV-18 RCIC Pump Suct from Cond Stor Isol Valve Open Cycled Powered from 125VDC (BMCC-1).  [5] 6 13MOV-41 RCIC Pump Suct From Suppr Pool INBD Isol Valve Closed Cycled Powered from 125VDC (BMCC-3) [5] 7 13MOV-39 RCIC Pump Suct From Suppr Pool Outboard Isol Valve Closed Cycled Powered from 125VDC (BMCC-3) [5] 8 13MOV-15 RCIC Steam Supply INBD Isol Valve Open Open Powered from AC. May be excluded since normally open, required open. [5] 9 13MOV-16 RCIC Turbine Steam Supply Outbd Isol Valve Open Open Powered from 125VDC (BMCC-1) [5] 10 13MOV-131 RCIC Turbine Steam Inlet Isol Valve Closed Open Powered from 125VDC (BMCC-3) [5] 11 13HOV-1 RCIC Trip Valve Open Open Powered from 125VDC (71DC-A2) [5] 12 13HOV-2 RCIC Turbine Governor Valve Open Throttled Powered from 125VDC (71DC-A2) [5] 13 13MOV-132 RCIC Turb Lube Oil Cooler Water Supply Isol Valve Closed Open Powered from 125VDC (BMCC-3) [5] 14 13PCV-23 RCIC Turb Lube Oil Cooler Water Supply Press Control Valve Open Throttled Self-actuated [5]  Page A-2 James A. FitzPatrick ESEP Report  ESEL Item Number Equipment Operating State Notes/Comments References ID Description Normal State Desired State 15 13MOV-20 RCIC Pump Disch to Reactor Outbd Isol Valve Open Open Powered from 125VDC (BMCC-1). May be excluded since normally open, required open. [5] 16 13MOV-21 RCIC Pump Disch to Reactor Inbd Isol Valve Closed Open Powered from 125VDC (BMCC-1) [5] 17 13P-3 RCIC Barometric Cndsr Vacuum Pump Off On Powered from 125VDC (BMCC-1) [5] 18 13P-4 RCIC Condensate Pump Off On Powered from 125VDC (BMCC-1) [5] 19 33TK-12A Condensate Storage Tank A Available Available - [5] 20 33TK-12B Condensate Storage Tank B Available Available - [5] 21 Torus Suppression Pool Available Available - [5] 22 02RV-71A ADS Main Steam Line A Safety/Relief Valve Closed Cycle - [6] 23 39ACC-256A IAS 02RV-71A Air Accumulator Available Available - [6] 24 02SOV-71A1 ADS/MST A 02TV-71A Auto/CR Manual Pilot Solenoid Valve Closed Cycle Powered from 125VDC (71DC-A2) [6] 25 02RV-71B ADS Main Steam Line A Safety/Relief Valve Closed Cycle - [6] 26 39ACC-256B IAS 02RV-71B Air Accumulator Available Available The accumulator is shown on reference for 02RV-71A only - typical for all, except 71E and F share an accumulator. [6] 27 02SOV-71B1 ADS/MST A 02RV-71B Auto/CR Manual Pilot Solenoid Valve Closed Cycle Powered from 125VDC (71DC-A2) Shown on reference only for 02SOV71A1 - typical for all SOVs. [6]  Page A-3 James A. FitzPatrick ESEP Report  ESEL Item Number Equipment Operating State Notes/Comments References ID Description Normal State Desired State 28 02RV-71C ADS Main Steam Line B Safety/Relief Valve Closed Cycle  [6] 29 39ACC-256C IAS 02RV-71C Air Accumulator Available Available The accumulator is shown on reference for 02RV-71A only - typical for all, except 71E and F share an accumulator. [6] 30 02SOV-71C1 ADS/MST B 02RV-71C Auto/CR Manual Pilot Solenoid Valve Closed Cycle Powered from 125VDC (71DC-A2) Shown on reference only for 02SOV71A1 - typical for all SOVs. [6] 31 02RV-71D ADS Main Steam Line B Safety/Relief Valve Closed Cycle - [6] 32 39ACC-256D IAS 02RV-71D Air Accumulator Available Available The accumulator is shown on reference for 02RV-71A only - typical for all, except 71E and F share an accumulator. [6] 33 02SOV-71D1 ADS/MST B 02RV-71D Auto/CR Manual Pilot Solenoid Valve Closed Cycle Powered from 125VDC (71DC-A2) Shown on reference only for 02SOV71A1 - typical for all SOVs. [6] 34 02RV-71E ADS Main Steam Line C Safety/Relief Valve Closed Cycle - [6] 35 39ACC-256E IAS 02RV-71E/F Air Accumulator Available Available The accumulator is shown on reference for 02RV-71A only - typical for all, except 71E and F share an accumulator. [6] 36 02SOV-71E1 ADS/MST C 02RV-71E Auto/CR Manual Pilot Solenoid Valve Closed Cycle Powered from 125VDC (71DC-A2) Shown on reference only for 02SOV71A1 - typical for all SOVs. [6] 37 02RV-71F Main Steam Line C Manual Safety Relief Valve Closed Cycle - [6] 38 02SOV-71F1 MST C 02RV-71F Control Room Manual Pilot Solenoid Valve Available Available Powered from 125VDC (71DC-A2) Shown on reference only for 02SOV71A1 - typical for all SOVs. [6] 39 02RV-71G ADS Main Steam Line C Safety/Relief Valve Closed Cycle - [6] Page A-4 James A. FitzPatrick ESEP Report  ESEL Item Number Equipment Operating State Notes/Comments References ID Description Normal State Desired State 40 39ACC-256G IAS 02RV-71G Air Accumulator Closed Cycle The accumulator is shown on reference for 02RV-71A only - typical for all, except 71E and F share an accumulator. [6] 41 02SOV-71G1 ADS/MST C 02RV-71G Auto/CR Manual Pilot Solenoid Valve Available Available Powered from 125VDC (71DC-A2) Shown on reference only for 02SOV71A1 - typical for all SOVs. [6] 42 02RV-71H ADS Main Steam Line D Safety/Relief Valve Closed Cycle - [6] 43 39ACC-256H IAS 02RV-71H Air Accumulator Closed Cycle The accumulator is shown on reference for 02RV-71A only - typical for all, except 71E and F share an accumulator. [6] 44 02SOV-71H1 ADS/MST D 02RV-71H Auto/CR Manual Pilot  Solenoid Valve Available Available Powered from 125VDC (71DC-A2) Shown on reference only for 02SOV71A1 - typical for all SOVs. [6] 45 02RV-71J Main Steam Line D Manual Safety Relief Valve Closed Cycle - [6] 46 39ACC-256J IAS 02RV-71J Air Accumulator Closed Cycle The accumulator is shown on reference for 02RV-71A only - typical for all, except 71E and F share an accumulator. [6] 47 02SOV-71J1 MST D 02RV-71J Control Room Manual Pilot Solenoid Valve Available Available Powered from 125VDC (71DC-A2) Shown on reference only for 02SOV71A1 - typical for all SOVs. [6] 48 02RV-71K Main Steam Line A Manual Safety Relief Valve Closed Cycle - [6] 49 39ACC-256K IAS 02RV-71K Air Accumulator Closed Cycle The accumulator is shown on reference for 02RV-71A only - typical for all, except 71E and F share an accumulator. [6] 50 02SOV-71K1 MST A 02RV-71K Control Room Manual Pilot Solenoid Valve Available Available Powered from 125VDC (71DC-A2) Shown on reference only for 02SOV71A1 - typical for all SOVs. [6]  Page A-5 James A. FitzPatrick ESEP Report  ESEL Item Number Equipment Operating State Notes/Comments References ID Description Normal State Desired State 51 02RV-71L Main Steam Line D Manual Safety Relief Valve Closed Cycle - [6] 52 39ACC-256L IAS 02RV-71L Air Accumulator Closed Cycle The accumulator is shown on reference for 02RV-71A only - typical for all, except 71E and F share an accumulator. [6] 53 02SOV-71L1 MST D 02RV-71L Control Room Manual Pilot Solenoid Valve Available Available Powered from 125VDC (71DC-A2) Shown on reference only for 02SOV71A1 - typical for all SOVs. [6] 54 27TK-7A Safety-related Nitrogen Tank Closed Cycle Provide backup for instrument air for SRVs [8] 55 27TK-7B Safety-related Nitrogen Tank Available Available Provide backup for instrument air for SRVs [8] 56 76P-1 West Diesel Fire Pump Off On Seismic qualified [38] 57 TBD Reliable Hardened Vent  Closed Cycled Not yet installed [3] 58 02-3LI-85A RX Water Lvl Available Available Powered from 13P/S-107 [27] 59 02-3LT-85A Reactor Vessel Wide Range Level Xmitter Available Available Powered from 13P/S-107 [27] 60 13INV-152 Inverter 13-152 Available Available Powered from 71DC-A2 [24] 61 13P/S-107 Single Nest Power Supply Available Available Powered from 13INV-152 [24] 62 06-LI-094A Reactor Water A Level Indicator Available Available Power from DC A [7] 63 06-LI-094C Rx Water Lvl A Available Available Power from DC A [7]  Page A-6 James A. FitzPatrick ESEP Report  ESEL Item Number Equipment Operating State Notes/Comments References ID Description Normal State Desired State 64 02-3LI-91 RX Wtr Lvl - Fuel Zone Available Available Battery A [28] 65 02-3LT-73 Reactor Vessel RHR Interlock Level Xmitter Available Available Input to 02-LI-91 [28] 66 02-3MTU-273 Containment Spray Perm Master Trip Unit Available Available Input to 02-LI-91 [28] 67 33LI-101A CST Level Available Available Power from 33E/S-G [29] 68 33LT-101 Condensate Storage Tanks Level Xmitter Available Available Power from 33E/S-G [29] 69 33E/S-G BOP Inst Pwr Supp Available Available Power from 71ACUPS-1. In panel 09BOP-P/S-1 [29] 70 09BOP/PS-1 BOP Inst Pwr Supp Panel Available Available  [29] 71 06PI-61A Reactor Vessel Press Indic Available Available 120VAC (71ACA2) -Backup 71INV-1A (71DC-A5) [30] 72 06PT-61A ECCS Loop A Feedwater Control Reactor Press Xmitter Available Available 120VAC (71ACA2) -Backup 71INV-1A (71DC-A5) [30] 73 06SCM-61A Reactor Press "A" Signal Conditioner Available Available 120VAC (71ACA2) -Backup 71INV-1A (71DC-A5) [30] 74 06SDM-61A Reactor Press "A" Sig Dist Module Available Available 120VAC (71ACA2) -Backup 71INV-1A (71DC-A5) [30] 75 27PI-115A1 NR PC Press Indicator Available Available 120VAC (71ACA2) -Backup 71INV-1A (71DC-A5) [31] 76 27PT-115A1 Drywell Div 1 Narrow Range Press Xmitter Available Available 120VAC (71ACA2) -Backup 71INV-1A (71DC-A5) [31] 77 27SCM-115A CAD Drywell Press Div 1 Input Module Available Available 120VAC (71ACA2) -Backup 71INV-1A (71DC-A5) [31]  Page A-7 James A. FitzPatrick ESEP Report  ESEL Item Number Equipment Operating State Notes/Comments References ID Description Normal State Desired State 78 27SDM-115A CAD Drywell Press Div 1 Distribution Module Available Available 120VAC (71ACA2) -Backup 71INV-1A (71DC-A5) [31] 79 27PI-115A2 WR PC Press Indicator Available Available 120VAC (71ACA2) -Backup 71INV-1A (71DC-A5) [32] 80 27PT-115A2 Drywell Div I Wide Range Press Xmitter Available Available 120VAC (71ACA2) -Backup 71INV-1A (71DC-A5) [32] 81 27PR-101A Suppression Chamber Monitor Press Recorder Available Available Powered from 10P/S-100A [33] 82 27PT-101A Torus Wide Range Press Xmitter Available Available Powered from 10P/S-100A [32] 83 27SDM-101A Suppression Chamber Monitor Signal Dist Module Available Available Powered from 10P/S-100A [32] 84 16-1TR-108 LRT Drywell Temp Mon Temp Recorder Available Available Powered from 10P/S-100A [34] 85 16-1RTD-108 LRT Drywell Area 4 Resist Temp Detector Available Available Powered from 10P/S-100A [34] 86 16-1SDM-108 LRT Drywell Temp Mon Signal Dist Module Available Available Powered from 10P/S-100A [34] 87 10P/S-100A Power Supply Available Available Powered from 120VAC (71ESSA1) [34] 88 16-1TR-131A Torus Bulk Temp Mon Average Temp Recorder Available Available Powered from 23E/S-200A [35] 89 16-1RTD-131A Torus Bulk Temp Monitor 0 Azimuth Bay L X-232 Resist Temp Detector Available Available Powered from 23E/S-200A [35] 90 16-1SDM-131A Torus Temp Mon A Signal Dist Module Available Available Powered from 23E/S-200A [35] 91 23E/S-200A Power Supply PS-1A Available Available Powered from 120VAC (71ACA2), backup by 71INV-1A (71DC-A5) [35]  Page A-8 James A. FitzPatrick ESEP Report  ESEL Item Number Equipment Operating State Notes/Comments References ID Description Normal State Desired State 92 23LI-203A PC Lvl Indicator Available Available 120VAC (71ACA2), backup by 71INV-1A (71DC-A5) [36] 93 23LT-203A1 Wide Range Containment Level HPCI Logic Level Xmitter (HI Tap) Available Available 120VAC (71ACA2), backup by 71INV-1A (71DC-A5) [36] 94 23LT-203A2 Wide Range Containment Level HPCI Logic Level Xmitter (LO Tap) Available Available 120VAC (71ACA2), backup by 71INV-1A (71DC-A5) [36] 95 23SCM-203A HPCI Drywell Sump Level Div I Input Signal Module Available Available 120VAC (71ACA2), backup by 71INV-1A (71DC-A5) [36] 96 23SUM-203A HPCI Drywell/Torus Diff Press Subtraction Module Available Available 120VAC (71ACA2), backup by 71INV-1A (71DC-A5) [36] 97 23LI-202A Suppression Chamber Water Level Indic Available Available 120VAC (71ACA2), backup by 71INV-1A (71DC-A5) [37] 98 23LT-202A Suppression Pool HPCI Logic Level Xmitter Available Available 120VAC (71ACA2), backup by 71INV-1A (71DC-A5) [37] 99 23SCM-202A HPCI Suppression Chamber Level Div I Input Signal Module Available Available 120VAC (71ACA2), backup by 71INV-1A (71DC-A5) [37] 100 23SDM-203A HPCI Drywell Sump Level Div I Signal Distrib Module Available Available 120VAC (71ACA2), backup by 71INV-1A (71DC-A5) [37] 101 25-05 Reactor Protection and NSSS System Rack Available Available - [27][30] 102 25-51 Jet Pump Instrument Rack 25-51 Available Available - [28] 103 27MAP Monitoring Analysis Panel Available Available - [30] 104 09-3 Nuclear Station Main Control Board Available Available - [28][30]  Page A-9 James A. FitzPatrick ESEP Report  ESEL Item Number Equipment Operating State Notes/Comments References ID Description Normal State Desired State 105 09-4 RWCU & Recirc Control Panel Available Available - [28] 106 09-5 Reactor Control Main Control Board Available Available - [27] 107 09-6 BOP Main Control Board Panel(MECH) Available Available - [29] 108 09-24 Process Instrumentation Panel (Div I) Available Available - [27] 109 09-30 Relay Cabinet Channel 'A' RCIC Panel Available Available - [25] 110 09-32 Channel 'A' RHR/RCIC Relay Panel Available Available Relay panel (for various valve control circuits - reference is an example) [19] 111 09-95 Emergency Core Cooling System DIV 1 A/C Trip Cabinet Available Available - [28] 112 TBD RHV Instrumentation Available Available Not yet installed [3] 113 71SB-1 125 Volt Station Battery A Operating Operating  - [11] 114 71BC-1A 125 VDC Station Battery Charger Operating Operating Powered from 71MCC-252 [11] 115 71BCB-2A Battery Control Board A Operating Operating  - [11] 116 71DC-A2 Relay Room Distribution Cabinet Operating Operating - [13] 117 71DC-A5 Relay Room Distribution Cabinet Operating Operating - [14]  Page A-10 James A. FitzPatrick ESEP Report  ESEL Item Number Equipment Operating State Notes/Comments References ID Description Normal State Desired State 118 71BMCC-1 Reactor Building West Crescent Motor Control Center Operating Operating - [12] 119 71BMCC-3 Reactor Building West Crescent Motor Control Center Operating Operating - [12] 120 71ESSA1 Relay Room Safeguard Power Distribution Panel Operating Operating 120VAC Instrument power. Powered from 71MCC-252 [23] 121 TBD 120V Instrument Power Inverter 1 Operating Operating To power required instruments from battery. Not yet installed [3] 122 71ACA2 Relay Room Emergency Power Distribution Panel Operating Operating Instrument Power [20] 123 71ACUPS Dist. Panel - Uninterruptible Bus Operating Operating Powered from 71UPP [22] 124 71UPP UPS Static Inverter Operating Operating Powered from 71MCC-262 (not selected train), backup from 71MCC-252 or 71BCB-2A [21] 125 71INV-1A Instrument power inverter Operating Operating Instrument power. Powered from 71DC-A5 [26] 126 71H05 4160V Switchgear Distribution (Bus 10500) Operating Operating 4160V FLEX generator connection point - primarily for RHR [15] 127 71MCC-156 600V Motor Control Center  (Bus 115600) Operating Operating May be required to open 10MOV-18 [16] 128 71MCC-252 600V Motor Control Center Bus 125200 Operating Operating Power for battery charger 71BC-1A [17] 129 71MCC-254 600V Motor Control Center Bus 125400 Operating Operating Power for 71MCC-252 [18] 130 10P-3A Residual Heat Removal Pump A Off On Phase 3 installed equipment  [9]  Page A-11 James A. FitzPatrick ESEP Report  ESEL Item Number Equipment Operating State Notes/Comments References ID Description Normal State Desired State 131 10P-2A RHR Keep-Full Pump A Off On Phase 3 installed equipment [9] 132 10E-2A Residual Heat Removal System Heat Exchanger A Available Available Phase 3 installed equipment [10] 133 10MOV-25A RHR A LPCI Inbd Inj Valve Closed Open Phase 2/3 installed equipment [9] 134 10MOV-66A RHR Heat Exch A Bypass Valve Open Closed Phase 3 installed equipment [9] 135 10MOV-27A RHR A LPCI Outbd Inj Valve Open Open Phase 2/3 installed equipment. May be excluded since normally open, required open. [9] 136 10MOV-13A RHR Pump A Suction Torus Isol. Valve Open Closed 71MCC-153 [9] 137 10MOV-15A RHR Pump A SDC Suction Isol Valve Closed Open 71MCC-153 [9] 138 10MOV-17 RHR SDC Outbd Isol. Valve Closed Open 71BMCC-4 [9] 139 10MOV-18 RHR SDC Inbd Isol. Valve Closed Open  [9] 140 10MOV-148A RHRSW A to RHR Cross Tie Upstr Isol Valve Closed Open Phase 2 installed equipment [10] 141 10MOV-149A RHRSW A to RHR Cross Tie Dnstr Isol Valve Closed Open Phase 2 installed equipment [10] 142 10RHR-432 RHRSW - Fire Protection Cross-Tie Isol Valve Closed Open Manually opened [10] 143 10MOV-89A RHR Heat Exch A Service Water Outlet Isol Valve Closed Closed/ Open Closed (Ph 2) Open (Ph 3) [10]  Page A-12 James A. FitzPatrick ESEP Report  ESEL Item Number Equipment Operating State Notes/Comments References ID Description Normal State Desired State 144 TBD Fire Pump 76P-1 Temporary Connection Isolation Valve to RHRSW Piping Closed Open To be manually Opened Not yet installed [3] 145 TBD Jib Crane for FLEX Pump N/A Available To be permanently installed. Powered from FLEX DG. Not yet installed [3]  Page A-13 James A. FitzPatrick ESEP Report  ATTACHMENT B - ESEP HCLPF VALUES AND FAILURE MODES TABULATION    Page B-1 James A. FitzPatrick ESEP Report  Item No. Equipment ID Equipment Description HCLPF (g) / Screening Level Failure Mode Comments 1 13P-1 RCIC Turbine Driven Pump >RLGM Screened Note 1 2 13E-1 RCIC Barometric Condenser >RLGM Screened Note 2 3 13E-2 RCIC Turbine Lube Oil Cooler >RLGM Screened Note 2:  In-Line Component 4 13TK-1 RCIC Vacuum Tank >RLGM Screened Note 2 5 13MOV-18 RCIC Pump Suct from Cond Stor Isol Valve >RLGM Screened  6 13MOV-41 RCIC Pump Suct From Suppr Pool INBD Isol Valve >RLGM Screened  7 13MOV-39 RCIC Pump Suct From Suppr Pool Outboard Isol Valve >RLGM Screened  8 13MOV-15 RCIC Steam Supply INBD Isol Valve >RLGM Screened  9 13MOV-16 RCIC Turbine Steam Supply Outbd Isol Valve >RLGM Screened  10 13MOV-131 RCIC Turbine Steam Inlet Isol Valve >RLGM Screened  11 13HOV-1 RCIC Trip Valve >RLGM Screened  12 13HOV-2 RCIC Turbine Governor Valve >RLGM Screened  13 13MOV-132 RCIC Turb Lube Oil Cooler Water Supply Isol Valve >RLGM Screened  14 13PCV-23 RCIC Turb Lube Oil Cooler Water Supply Press Control Valve >RLGM Screened  15 13MOV-20 RCIC Pump Disch to Reactor Outbd Isol Valve >RLGM Screened  16 13MOV-21 RCIC Pump Disch to Reactor Inbd Isol Valve >RLGM Screened  17 13P-3 RCIC Barometric Cndsr Vacuum Pump >RLGM Screened Note 2 18 13P-4 RCIC Condensate Pump >RLGM Screened Note 2 19 33TK-12A Condensate Storage Tank A >RLGM Screened Note 1 20 33TK-12B Condensate Storage Tank B >RLGM Screened Note 1  Page B-2 James A. FitzPatrick ESEP Report  Item No. Equipment ID Equipment Description HCLPF (g) / Screening Level Failure Mode Comments 21 Torus Suppression Pool >RLGM Screened Note 3 22 02RV-71A ADS Main Steam Line A Safety/Relief Valve >RLGM Screened  23 39ACC-256A IAS 02RV-71A Air Accumulator >RLGM Screened Note 2 24 02SOV-71A1 ADS/MST A 02TV-71A Auto/CR Manual Pilot Solenoid Valve >RLGM Screened  25 02RV-71B ADS Main Steam Line A Safety/Relief Valve >RLGM Screened  26 39ACC-256B IAS 02RV-71B Air Accumulator >RLGM Screened Note 2 27 02SOV-71B1 ADS/MST A 02RV-71B Auto/CR Manual Pilot Solenoid Valve >RLGM Screened  28 02RV-71C ADS Main Steam Line B Safety/Relief Valve >RLGM Screened  29 39ACC-256C IAS 02RV-71C Air Accumulator >RLGM Screened Note 2 30 02SOV-71C1 ADS/MST B 02RV-71C Auto/CR Manual Pilot Solenoid Valve >RLGM Screened  31 02RV-71D ADS Main Steam Line B Safety/Relief Valve >RLGM Screened  32 39ACC-256D IAS 02RV-71D Air Accumulator >RLGM Screened Note 2 33 02SOV-71D1 ADS/MST B 02RV-71D Auto/CR Manual Pilot Solenoid Valve >RLGM Screened  34 02RV-71E ADS Main Steam Line C Safety/Relief Valve >RLGM Screened  35 39ACC-256E IAS 02RV-71E/F Air Accumulator >RLGM Screened Note 2 36 02SOV-71E1 ADS/MST C 02RV-71E Auto/CR Manual Pilot Solenoid Valve >RLGM Screened  37 02RV-71F Main Steam Line C Manual Safety Relief Valve >RLGM Screened  38 02SOV-71F1 MST C 02RV-71F Control Room Manual Pilot Solenoid Valve >RLGM Screened  39 02RV-71G ADS Main Steam Line C Safety/Relief Valve >RLGM Screened    Page B-3 James A. FitzPatrick ESEP Report  Item No. Equipment ID Equipment Description HCLPF (g) / Screening Level Failure Mode Comments 40 39ACC-256G IAS 02RV-71G Air Accumulator >RLGM Screened Note 2 41 02SOV-71G1 ADS/MST C 02RV-71G Auto/CR Manual Pilot Solenoid Valve >RLGM Screened  42 02RV-71H ADS Main Steam Line D Safety/Relief Valve >RLGM Screened  43 39ACC-256H IAS 02RV-71H Air Accumulator >RLGM Screened Note 2 44 02SOV-71H1 ADS/MST D 02RV-71H Auto/CR Manual Pilot  Solenoid Valve >RLGM Screened  45 02RV-71J Main Steam Line D Manual Safety Relief Valve >RLGM Screened  46 39ACC-256J IAS 02RV-71J Air Accumulator >RLGM Screened Note 2 47 02SOV-71J1 MST D 02RV-71J Control Room Manual Pilot Solenoid Valve >RLGM Screened  48 02RV-71K Main Steam Line A Manual Safety Relief Valve >RLGM Screened  49 39ACC-256K IAS 02RV-71K Air Accumulator >RLGM Screened Note 2 50 02SOV-71K1 MST A 02RV-71K Control Room Manual Pilot Solenoid Valve >RLGM Screened  51 02RV-71L Main Steam Line D Manual Safety Relief Valve >RLGM Screened  52 39ACC-256L IAS 02RV-71L Air Accumulator >RLGM Screened Note 2 53 02SOV-71L1 MST D 02RV-71L Control Room Manual Pilot Solenoid Valve >RLGM Screened  54 27TK-7A Safety-related Nitrogen Tank >RLGM Screened Note 1 55 27TK-7B Safety-related Nitrogen Tank >RLGM Screened Note 1 56 76P-1 West Diesel Fire Pump >RLGM Screened Note 2 57 TBD Reliable Hardened Vent Not Applicable Not Applicable New FLEX Component to be seismically designed. 58 02-3LI-85A RX Water Lvl >RLGM Screened Note 1  Page B-4 James A. FitzPatrick ESEP Report  Item No. Equipment ID Equipment Description HCLPF (g) / Screening Level Failure Mode Comments 59 02-3LT-85A Reactor Vessel Wide Range Level Xmitter >RLGM Screened  60 13INV-152 Inverter 13-152 >RLGM Screened Note 1 61 13P/S-107 Single Nest Power Supply >RLGM Screened Note 1 62 06-LI-094A Reactor Water A Level Indicator >RLGM Screened Note 1 63 06-LI-094C Rx Water Lvl A >RLGM Screened Note 1 64 02-3LI-91 RX Wtr Lvl - Fuel Zone >RLGM Screened Note 1 65 02-3LT-73 Reactor Vessel RHR Interlock Level Xmitter >RLGM Screened  66 02-3MTU-273 Containment Spray Perm Master Trip Unit >RLGM Screened Note 1 67 33LI-101A CST Level >RLGM Screened Note 1 68 33LT-101 Condensate Storage Tanks Level Xmitter >RLGM Screened  69 33E/S-G BOP Inst Pwr Supp >RLGM Screened  70 09BOP/PS-1 BOP Inst Pwr Supp Panel >RLGM Screened Note 1 71 06PI-61A Reactor Vessel Press Indic >RLGM Screened Note 1 72 06PT-61A ECCS Loop A Feedwater Control Reactor Press Xmitter >RLGM Screened  73 06SCM-61A Reactor Press "A" Signal Conditioner >RLGM Screened Note 1 74 06SDM-61A Reactor Press "A" Sig Dist Module >RLGM Screened Note 1 75 27PI-115A1 NR PC Press Indicator >RLGM Screened Note 1 76 27PT-115A1 Drywell Div 1 Narrow Range Press Xmitter >RLGM Screened  77 27SCM-115A CAD Drywell Press Div 1 Input Module >RLGM Screened Note 1 78 27SDM-115A CAD Drywell Press Div 1 Distribution Module >RLGM Screened Note 1  Page B-5 James A. FitzPatrick ESEP Report  Item No. Equipment ID Equipment Description HCLPF (g) / Screening Level Failure Mode Comments 79 27PI-115A2 WR PC Press Indicator >RLGM Screened Note 1 80 27PT-115A2 Drywell Div I Wide Range Press Xmitter >RLGM Screened  81 27PR-101A Suppression Chamber Monitor Press Recorder >RLGM Screened Note 1 82 27PT-101A Torus Wide Range Press Xmitter >RLGM Screened  83 27SDM-101A Suppression Chamber Monitor Signal Dist Module >RLGM Screened  84 16-1TR-108 LRT Drywell Temp Mon Temp Recorder >RLGM Screened Note 1 85 16-1RTD-108 LRT Drywell Area 4 Resist Temp Detector >RLGM Screened  86 16-1SDM-108 LRT Drywell Temp Mon Signal Dist Module >RLGM Screened  87 10P/S-100A Power Supply >RLGM Screened  88 16-1TR-131A Torus Bulk Temp Mon Average Temp Recorder >RLGM Screened Note 1 89 16-1RTD-131A Torus Bulk Temp Monitor 0 Azimuth Bay L X-232 Resist Temp Detector >RLGM Screened  90 16-1SDM-131A Torus Temp Mon A Signal Dist Module >RLGM Screened Note 1 91 23E/S-200A Power Supply PS-1A >RLGM Screened Note 1 92 23LI-203A PC Lvl Indicator >RLGM Screened Note 1 93 23LT-203A1 Wide Range Containment Level HPCI Logic Level Xmitter (HI Tap) >RLGM Screened  94 23LT-203A2 Wide Range Containment Level HPCI Logic Level Xmitter (LO Tap) >RLGM Screened  95 23SCM-203A HPCI Drywell Sump Level Div I Input Signal Module >RLGM Screened Note 1 96 23SUM-203A HPCI Drywell/Torus Diff Press Subtraction Module >RLGM Screened Note 1 97 23LI-202A Suppression Chamber Water Level Indic >RLGM Screened Note 1 98 23LT-202A Suppression Pool HPCI Logic Level Xmitter >RLGM Screened    Page B-6 James A. FitzPatrick ESEP Report  Item No. Equipment ID Equipment Description HCLPF (g) / Screening Level Failure Mode Comments 99 23SCM-202A HPCI Suppression Chamber Level Div I Input Signal Module >RLGM Screened Note 1 100 23SDM-203A HPCI Drywell Sump Level Div I Signal Distrib Module >RLGM Screened Note 1 101 25-05 Reactor Protection and NSSS System Rack >RLGM Screened Note 1 102 25-51 Jet Pump Instrument Rack 25-51 >RLGM Screened Note 1 103 27MAP Monitoring Analysis Panel >RLGM Screened Note 1 104 09-3 Nuclear Station Main Control Board >RLGM Screened Note 1 105 09-4 RWCU & Recirc Control Panel >RLGM Screened Note 1 106 09-5 Reactor Control Main Control Board >RLGM Screened Note 1 107 09-6 BOP Main Control Board Panel(MECH) >RLGM Screened Note 1 108 09-24 Process Instrumentation Panel (Div I) >RLGM Screened Note 1 109 09-30 Relay Cabinet Channel 'A' RCIC Panel >RLGM Screened Note 1 110 09-32 Channel 'A' RHR/RCIC Relay Panel >RLGM Screened Note 1 111 09-95 Emergency Core Cooling System DIV 1 A/C Trip Cabinet >RLGM Screened Note 1 112 TBD RHV Instrumentation Not Applicable Not Applicable New FLEX Component to be seismically designed. 113 71SB-1 125 Volt Station Battery A >RLGM Screened Note 1 114 71BC-1A 125 VDC Station Battery Charger >RLGM Screened Note 1 115 71BCB-2A Battery Control Board A >RLGM Screened Note 2 116 71DC-A2 Relay Room Distribution Cabinet >RLGM Screened Note 2 117 71DC-A5 Relay Room Distribution Cabinet >RLGM Screened Note 2  Page B-7 James A. FitzPatrick ESEP Report  Item No. Equipment ID Equipment Description HCLPF (g) / Screening Level Failure Mode Comments 118 71BMCC-1 Reactor Building West Crescent Motor Control Center >RLGM Screened Note 1 119 71BMCC-3 Reactor Building West Crescent Motor Control Center >RLGM Screened Note 1 120 71ESSA1 Relay Room Safeguard Power Distribution Panel >RLGM Screened Note 2 121 TBD 120V Instrument Power Inverter 1 Not Applicable Not Applicable New FLEX Component to be seismically designed. 122 71ACA2 Relay Room Emergency Power Distribution Panel >RLGM Screened Note 1 123 71ACUPS Dist. Panel - Uninterruptible Bus >RLGM Screened Note 1 124 71UPP UPS Static Inverter >RLGM Screened Note 1 125 71INV-1A Instrument power inverter >RLGM Screened Note 2 126 71H05 4160V Switchgear Distribution (Bus 10500) >RLGM Screened Note 1 127 71MCC-156 600V Motor Control Center  (Bus 115600) >RLGM Screened Note 1 128 71MCC-252 600V Motor Control Center Bus 125200 >RLGM Screened Note 1 129 71MCC-254 600V Motor Control Center Bus 125400 >RLGM Screened Note 1 130 10P-3A Residual Heat Removal Pump A >RLGM Screened Note 1 131 10P-2A RHR Keep-Full Pump A >RLGM Screened Note 2 132 10E-2A Residual Heat Removal System Heat Exchanger A >RLGM Screened Note 3 133 10MOV-25A RHR A LPCI Inbd Inj Valve >RLGM Screened  134 10MOV-66A RHR Heat Exch A Bypass Valve >RLGM Screened  135 10MOV-27A RHR A LPCI Outbd Inj Valve >RLGM Screened  136 10MOV-13A RHR Pump A Suction Torus Isol. Valve >RLGM Screened  137 10MOV-15A RHR Pump A SDC Suction Isol Valve >RLGM Screened    Page B-8 James A. FitzPatrick ESEP Report  Item No. Equipment ID Equipment Description HCLPF (g) / Screening Level Failure Mode Comments 138 10MOV-17 RHR SDC Outbd Isol. Valve >RLGM Screened  139 10MOV-18 RHR SDC Inbd Isol. Valve >RLGM Screened  140 10MOV-148A RHRSW A to RHR Cross Tie Upstr Isol Valve >RLGM Screened  141 10MOV-149A RHRSW A to RHR Cross Tie Dnstr Isol Valve >RLGM Screened  142 10RHR-432 RHRSW - Fire Protection Cross-Tie Isol Valve >RLGM Screened  143 10MOV-89A RHR Heat Exch A Service Water Outlet Isol Valve >RLGM Screened  144 TBD Fire Pump 76P-1 Temporary Connection Isolation Valve to RHRSW Piping Not Applicable Not Applicable New FLEX Component to be seismically designed. 145 TBD Jib Crane for FLEX Pump Not Applicable Not Applicable New FLEX Component to be seismically designed. Notes: 1. Anchorage screened out based on available margin during walkdown by SRT. 2. Anchorage screened out during walkdown validation by SRT. 3. Anchorage screened out after walkdown and evaluation of photographs and other existing information by SRT. Page B-9}}

Latest revision as of 11:20, 6 April 2019

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