Text

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...
	ML17117A699
Person / Time
Site:	FitzPatrick
Issue date:	04/27/2017
From:	Joseph Pacher; Exelon Generation Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	JAFP-17-0029
	Download: ML17117A699 (48)
	v • d • e

Exelon Generation Company, LLC James A. FitzPatrick NPP P.O. Box 110 Lycoming, NY 13093 Tel 315-342-3840 Joseph E. Pacher Site Vice President - JAF JAFP-17-0029 April 27, 2017 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001

Subject:

Revision to Entergys 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, Entergys 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, Entergys 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.

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)

For Information Only EXPEDITED SEISMIC EVALUATION PROCESS (ESEP) REPORT FOR JAMES A.

FITZPATRICK NUCLEAR POWER PLANT (JAF)

Page 1

For Information Only 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

For Information Only 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 ....................................................................................... A-1 ATTACHMENT B - ESEP HCLPF VALUES AND FAILURE MODES TABULATION ......................................... B-1 Page 3

For Information Only 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

For Information Only 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

For Information Only 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 NRCs 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.

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For Information Only 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 hours2.662037e-4 days 0.00639 hours 3.80291e-5 weeks 8.7515e-6 months , 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 hours4.050926e-4 days 0.00972 hours 5.787037e-5 weeks 1.33175e-5 months 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

For Information Only 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.

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For Information Only 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.

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For Information Only 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 GMRS (Hz) (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

For Information Only James A. FitzPatrick ESEP Report Table 4-1: GMRS for James A. FitzPatrick (continued)

Frequency GMRS (Hz) (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

For Information Only James A. FitzPatrick ESEP Report GMRS at Control Point for James A. FitzPatrick Nuclear Power Plant, 5% Damping 0.30 JAF GMRS 0.25 0.20 SA (g) 0.15 0.10 0.05 0.00 0.1 1 10 100 Frequency (Hz)

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 Spectral Acceleration (Hz) (g) 100 0.15 25 0.15 10 0.15 5 0.21 2.5 0.22 1 0.13 0.5 0.064 Page 12

For Information Only 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.

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For Information Only James A. FitzPatrick ESEP Report Table 5-1: RLGM for James A. FitzPatrick Frequency RLGM at 5% Damping (Hz) (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.

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For Information Only 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

For Information Only 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

For Information Only 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

For Information Only 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 Fortyfour (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

For Information Only 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 NRCs 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 NRCs 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 NRCs 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 NRCs 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

For Information Only 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 Entergys 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]

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For Information Only 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

For Information Only 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

For Information Only 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

For Information Only James A. FitzPatrick ESEP Report ATTACHMENT A - JAMES A. FITZPATRICK ESEL Page A-1

For Information Only James A. FitzPatrick ESEP Report ESEL Equipment Operating State Item Notes/Comments References Number 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]

RCIC Pump Suct from Cond 5 13MOV-18 Open Cycled Powered from 125VDC (BMCC-1). [5]

Stor Isol Valve RCIC Pump Suct From Suppr 6 13MOV-41 Closed Cycled Powered from 125VDC (BMCC-3) [5]

Pool INBD Isol Valve RCIC Pump Suct From Suppr 7 13MOV-39 Closed Cycled Powered from 125VDC (BMCC-3) [5]

Pool Outboard Isol Valve RCIC Steam Supply INBD Isol Powered from AC. May be excluded since 8 13MOV-15 Open Open [5]

Valve normally open, required open.

RCIC Turbine Steam Supply 9 13MOV-16 Open Open Powered from 125VDC (BMCC-1) [5]

Outbd Isol Valve RCIC Turbine Steam Inlet Isol 10 13MOV-131 Closed Open Powered from 125VDC (BMCC-3) [5]

Valve 11 13HOV-1 RCIC Trip Valve Open Open Powered from 125VDC (71DC-A2) [5]

RCIC Turbine Governor 12 13HOV-2 Open Throttled Powered from 125VDC (71DC-A2) [5]

Valve RCIC Turb Lube Oil Cooler 13 13MOV-132 Closed Open Powered from 125VDC (BMCC-3) [5]

Water Supply Isol Valve RCIC Turb Lube Oil Cooler 14 13PCV-23 Water Supply Press Control Open Throttled Self-actuated [5]

Valve Page A-2

For Information Only James A. FitzPatrick ESEP Report ESEL Equipment Operating State Item Notes/Comments References Number ID Description Normal State Desired State Powered from 125VDC (BMCC-1). May be RCIC Pump Disch to Reactor 15 13MOV-20 Open Open excluded since normally open, required [5]

Outbd Isol Valve open.

RCIC Pump Disch to Reactor 16 13MOV-21 Closed Open Powered from 125VDC (BMCC-1) [5]

Inbd Isol Valve RCIC Barometric Cndsr 17 13P-3 Off On Powered from 125VDC (BMCC-1) [5]

Vacuum Pump 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]

ADS Main Steam Line A 22 02RV-71A Closed Cycle - [6]

Safety/Relief Valve IAS 02RV-71A Air 23 39ACC-256A Available Available - [6]

Accumulator ADS/MST A 02TV-71A 24 02SOV-71A1 Auto/CR Manual Pilot Closed Cycle Powered from 125VDC (71DC-A2) [6]

Solenoid Valve ADS Main Steam Line A 25 02RV-71B Closed Cycle - [6]

Safety/Relief Valve The accumulator is shown on reference IAS 02RV-71B Air 26 39ACC-256B Available Available for 02RV-71A only - typical for all, except [6]

Accumulator 71E and F share an accumulator.

ADS/MST A 02RV-71B Powered from 125VDC (71DC-A2) 27 02SOV-71B1 Auto/CR Manual Pilot Closed Cycle Shown on reference only for 02SOV71A1 [6]

Solenoid Valve - typical for all SOVs.

Page A-3

For Information Only James A. FitzPatrick ESEP Report ESEL Equipment Operating State Item Notes/Comments References Number ID Description Normal State Desired State ADS Main Steam Line B 28 02RV-71C Closed Cycle [6]

Safety/Relief Valve The accumulator is shown on reference IAS 02RV-71C Air 29 39ACC-256C Available Available for 02RV-71A only - typical for all, except [6]

Accumulator 71E and F share an accumulator.

ADS/MST B 02RV-71C Powered from 125VDC (71DC-A2) 30 02SOV-71C1 Auto/CR Manual Pilot Closed Cycle Shown on reference only for 02SOV71A1 [6]

Solenoid Valve - typical for all SOVs.

ADS Main Steam Line B 31 02RV-71D Closed Cycle - [6]

Safety/Relief Valve The accumulator is shown on reference IAS 02RV-71D Air 32 39ACC-256D Available Available for 02RV-71A only - typical for all, except [6]

Accumulator 71E and F share an accumulator.

ADS/MST B 02RV-71D Powered from 125VDC (71DC-A2) 33 02SOV-71D1 Auto/CR Manual Pilot Closed Cycle Shown on reference only for 02SOV71A1 [6]

Solenoid Valve - typical for all SOVs.

ADS Main Steam Line C 34 02RV-71E Closed Cycle - [6]

Safety/Relief Valve The accumulator is shown on reference IAS 02RV-71E/F Air 35 39ACC-256E Available Available for 02RV-71A only - typical for all, except [6]

Accumulator 71E and F share an accumulator.

ADS/MST C 02RV-71E Powered from 125VDC (71DC-A2) 36 02SOV-71E1 Auto/CR Manual Pilot Closed Cycle Shown on reference only for 02SOV71A1 [6]

Solenoid Valve - typical for all SOVs.

Main Steam Line C Manual 37 02RV-71F Closed Cycle - [6]

Safety Relief Valve MST C 02RV-71F Control Powered from 125VDC (71DC-A2) 38 02SOV-71F1 Room Manual Pilot Solenoid Available Available Shown on reference only for 02SOV71A1 [6]

Valve - typical for all SOVs.

ADS Main Steam Line C 39 02RV-71G Closed Cycle - [6]

Safety/Relief Valve Page A-4

For Information Only James A. FitzPatrick ESEP Report ESEL Equipment Operating State Item Notes/Comments References Number ID Description Normal State Desired State The accumulator is shown on reference IAS 02RV-71G Air 40 39ACC-256G Closed Cycle for 02RV-71A only - typical for all, except [6]

Accumulator 71E and F share an accumulator.

ADS/MST C 02RV-71G Powered from 125VDC (71DC-A2) 41 02SOV-71G1 Auto/CR Manual Pilot Available Available Shown on reference only for 02SOV71A1 [6]

Solenoid Valve - typical for all SOVs.

ADS Main Steam Line D 42 02RV-71H Closed Cycle - [6]

Safety/Relief Valve The accumulator is shown on reference IAS 02RV-71H Air 43 39ACC-256H Closed Cycle for 02RV-71A only - typical for all, except [6]

Accumulator 71E and F share an accumulator.

ADS/MST D 02RV-71H Powered from 125VDC (71DC-A2) 44 02SOV-71H1 Auto/CR Manual Pilot Available Available Shown on reference only for 02SOV71A1 [6]

Solenoid Valve - typical for all SOVs.

Main Steam Line D Manual 45 02RV-71J Closed Cycle - [6]

Safety Relief Valve The accumulator is shown on reference IAS 02RV-71J Air 46 39ACC-256J Closed Cycle for 02RV-71A only - typical for all, except [6]

Accumulator 71E and F share an accumulator.

MST D 02RV-71J Control Powered from 125VDC (71DC-A2) 47 02SOV-71J1 Room Manual Pilot Solenoid Available Available Shown on reference only for 02SOV71A1 [6]

Valve - typical for all SOVs.

Main Steam Line A Manual 48 02RV-71K Closed Cycle - [6]

Safety Relief Valve The accumulator is shown on reference IAS 02RV-71K Air 49 39ACC-256K Closed Cycle for 02RV-71A only - typical for all, except [6]

Accumulator 71E and F share an accumulator.

MST A 02RV-71K Control Powered from 125VDC (71DC-A2) 50 02SOV-71K1 Room Manual Pilot Solenoid Available Available Shown on reference only for 02SOV71A1 [6]

Valve - typical for all SOVs.

Page A-5

For Information Only James A. FitzPatrick ESEP Report ESEL Equipment Operating State Item Notes/Comments References Number ID Description Normal State Desired State Main Steam Line D Manual 51 02RV-71L Closed Cycle - [6]

Safety Relief Valve The accumulator is shown on reference IAS 02RV-71L Air 52 39ACC-256L Closed Cycle for 02RV-71A only - typical for all, except [6]

Accumulator 71E and F share an accumulator.

MST D 02RV-71L Control Powered from 125VDC (71DC-A2) 53 02SOV-71L1 Room Manual Pilot Solenoid Available Available Shown on reference only for 02SOV71A1 [6]

Valve - typical for all SOVs.

Provide backup for instrument air for 54 27TK-7A Safety-related Nitrogen Tank Closed Cycle [8]

SRVs Provide backup for instrument air for 55 27TK-7B Safety-related Nitrogen Tank Available Available [8]

SRVs 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]

Reactor Vessel Wide Range 59 02-3LT-85A Available Available Powered from 13P/S-107 [27]

Level Xmitter 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]

Reactor Water A Level 62 06-LI-094A Available Available Power from DC A [7]

Indicator 63 06-LI-094C Rx Water Lvl A Available Available Power from DC A [7]

Page A-6

For Information Only James A. FitzPatrick ESEP Report ESEL Equipment Operating State Item Notes/Comments References Number ID Description Normal State Desired State 64 02-3LI-91 RX Wtr Lvl - Fuel Zone Available Available Battery A [28]

Reactor Vessel RHR Interlock 65 02-3LT-73 Available Available Input to 02-LI-91 [28]

Level Xmitter 02-3MTU- Containment Spray Perm 66 Available Available Input to 02-LI-91 [28]

273 Master Trip Unit 67 33LI-101A CST Level Available Available Power from 33E/S-G [29]

Condensate Storage Tanks 68 33LT-101 Available Available Power from 33E/S-G [29]

Level Xmitter Power from 71ACUPS-1. In panel 09BOP-69 33E/S-G BOP Inst Pwr Supp Available Available [29]

P/S-1 70 09BOP/PS-1 BOP Inst Pwr Supp Panel Available Available [29]

120VAC (71ACA2) -Backup 71INV-1A 71 06PI-61A Reactor Vessel Press Indic Available Available [30]

(71DC-A5)

ECCS Loop A Feedwater 120VAC (71ACA2) -Backup 71INV-1A 72 06PT-61A Control Reactor Press Available Available [30]

(71DC-A5)

Xmitter Reactor Press "A" Signal 120VAC (71ACA2) -Backup 71INV-1A 73 06SCM-61A Available Available [30]

Conditioner (71DC-A5)

Reactor Press "A" Sig Dist 120VAC (71ACA2) -Backup 71INV-1A 74 06SDM-61A Available Available [30]

Module (71DC-A5) 120VAC (71ACA2) -Backup 71INV-1A 75 27PI-115A1 NR PC Press Indicator Available Available [31]

(71DC-A5)

Drywell Div 1 Narrow Range 120VAC (71ACA2) -Backup 71INV-1A 76 27PT-115A1 Available Available [31]

Press Xmitter (71DC-A5)

CAD Drywell Press Div 1 120VAC (71ACA2) -Backup 71INV-1A 77 27SCM-115A Available Available [31]

Input Module (71DC-A5)

Page A-7

For Information Only James A. FitzPatrick ESEP Report ESEL Equipment Operating State Item Notes/Comments References Number ID Description Normal State Desired State CAD Drywell Press Div 1 120VAC (71ACA2) -Backup 71INV-1A 78 27SDM-115A Available Available [31]

Distribution Module (71DC-A5) 120VAC (71ACA2) -Backup 71INV-1A 79 27PI-115A2 WR PC Press Indicator Available Available [32]

(71DC-A5)

Drywell Div I Wide Range 120VAC (71ACA2) -Backup 71INV-1A 80 27PT-115A2 Available Available [32]

Press Xmitter (71DC-A5)

Suppression Chamber 81 27PR-101A Available Available Powered from 10P/S-100A [33]

Monitor Press Recorder Torus Wide Range Press 82 27PT-101A Available Available Powered from 10P/S-100A [32]

Xmitter Suppression Chamber 83 27SDM-101A Available Available Powered from 10P/S-100A [32]

Monitor Signal Dist Module LRT Drywell Temp Mon 84 16-1TR-108 Available Available Powered from 10P/S-100A [34]

Temp Recorder LRT Drywell Area 4 Resist 85 16-1RTD-108 Available Available Powered from 10P/S-100A [34]

Temp Detector 16-1SDM- LRT Drywell Temp Mon 86 Available Available Powered from 10P/S-100A [34]

108 Signal Dist Module 87 10P/S-100A Power Supply Available Available Powered from 120VAC (71ESSA1) [34]

Torus Bulk Temp Mon 88 16-1TR-131A Available Available Powered from 23E/S-200A [35]

Average Temp Recorder Torus Bulk Temp Monitor 0 16-1RTD-89 Azimuth Bay L X-232 Resist Available Available Powered from 23E/S-200A [35]

131A Temp Detector 16-1SDM- Torus Temp Mon A Signal 90 Available Available Powered from 23E/S-200A [35]

131A Dist Module Powered from 120VAC (71ACA2), backup 91 23E/S-200A Power Supply PS-1A Available Available [35]

by 71INV-1A (71DC-A5)

Page A-8

For Information Only James A. FitzPatrick ESEP Report ESEL Equipment Operating State Item Notes/Comments References Number ID Description Normal State Desired State 120VAC (71ACA2), backup by 71INV-1A 92 23LI-203A PC Lvl Indicator Available Available [36]

(71DC-A5)

Wide Range Containment 120VAC (71ACA2), backup by 71INV-1A 93 23LT-203A1 Level HPCI Logic Level Available Available [36]

(71DC-A5)

Xmitter (HI Tap)

Wide Range Containment 120VAC (71ACA2), backup by 71INV-1A 94 23LT-203A2 Level HPCI Logic Level Available Available [36]

(71DC-A5)

Xmitter (LO Tap)

HPCI Drywell Sump Level Div 120VAC (71ACA2), backup by 71INV-1A 95 23SCM-203A Available Available [36]

I Input Signal Module (71DC-A5)

HPCI Drywell/Torus Diff 120VAC (71ACA2), backup by 71INV-1A 96 23SUM-203A Available Available [36]

Press Subtraction Module (71DC-A5)

Suppression Chamber Water 120VAC (71ACA2), backup by 71INV-1A 97 23LI-202A Available Available [37]

Level Indic (71DC-A5)

Suppression Pool HPCI Logic 120VAC (71ACA2), backup by 71INV-1A 98 23LT-202A Available Available [37]

Level Xmitter (71DC-A5)

HPCI Suppression Chamber 120VAC (71ACA2), backup by 71INV-1A 99 23SCM-202A Level Div I Input Signal Available Available [37]

(71DC-A5)

Module HPCI Drywell Sump Level Div 120VAC (71ACA2), backup by 71INV-1A 100 23SDM-203A Available Available [37]

I Signal Distrib Module (71DC-A5)

Reactor Protection and NSSS 101 25-05 Available Available - [27][30]

System Rack Jet Pump Instrument Rack 102 25-51 Available Available - [28]

25-51 103 27MAP Monitoring Analysis Panel Available Available - [30]

Nuclear Station Main 104 09-3 Available Available - [28][30]

Control Board Page A-9

For Information Only James A. FitzPatrick ESEP Report ESEL Equipment Operating State Item Notes/Comments References Number ID Description Normal State Desired State RWCU & Recirc Control 105 09-4 Available Available - [28]

Panel Reactor Control Main 106 09-5 Available Available - [27]

Control Board BOP Main Control Board 107 09-6 Available Available - [29]

Panel(MECH)

Process Instrumentation 108 09-24 Available Available - [27]

Panel (Div I)

Relay Cabinet Channel A 109 09-30 Available Available - [25]

RCIC Panel Channel A RHR/RCIC Relay Relay panel (for various valve control 110 09-32 Available Available [19]

Panel circuits - reference is an example)

Emergency Core Cooling 111 09-95 System DIV 1 A/C Trip Available Available - [28]

Cabinet 112 TBD RHV Instrumentation Available Available Not yet installed [3]

113 71SB-1 125 Volt Station Battery A Operating Operating - [11]

125 VDC Station Battery 114 71BC-1A Operating Operating Powered from 71MCC-252 [11]

Charger 115 71BCB-2A Battery Control Board A Operating Operating - [11]

Relay Room Distribution 116 71DC-A2 Operating Operating - [13]

Cabinet Relay Room Distribution 117 71DC-A5 Operating Operating - [14]

Cabinet Page A-10

For Information Only James A. FitzPatrick ESEP Report ESEL Equipment Operating State Item Notes/Comments References Number ID Description Normal State Desired State Reactor Building West 118 71BMCC-1 Crescent Motor Control Operating Operating - [12]

Center Reactor Building West 119 71BMCC-3 Crescent Motor Control Operating Operating - [12]

Center Relay Room Safeguard 120VAC Instrument power. Powered 120 71ESSA1 Operating Operating [23]

Power Distribution Panel from 71MCC-252 120V Instrument Power To power required instruments from 121 TBD Operating Operating [3]

Inverter 1 battery. Not yet installed Relay Room Emergency 122 71ACA2 Operating Operating Instrument Power [20]

Power Distribution Panel Dist. Panel - Uninterruptible 123 71ACUPS Operating Operating Powered from 71UPP [22]

Bus Powered from 71MCC-262 (not selected 124 71UPP UPS Static Inverter Operating Operating train), backup from 71MCC-252 or [21]

71BCB-2A Instrument power. Powered from 71DC-125 71INV-1A Instrument power inverter Operating Operating [26]

A5 4160V Switchgear 4160V FLEX generator connection point -

126 71H05 Operating Operating [15]

Distribution (Bus 10500) primarily for RHR 600V Motor Control Center 127 71MCC-156 Operating Operating May be required to open 10MOV-18 [16]

(Bus 115600) 600V Motor Control Center 128 71MCC-252 Operating Operating Power for battery charger 71BC-1A [17]

Bus 125200 600V Motor Control Center 129 71MCC-254 Operating Operating Power for 71MCC-252 [18]

Bus 125400 Residual Heat Removal 130 10P-3A Off On Phase 3 installed equipment [9]

Pump A Page A-11

For Information Only James A. FitzPatrick ESEP Report ESEL Equipment Operating State Item Notes/Comments References Number ID Description Normal State Desired State 131 10P-2A RHR Keep-Full Pump A Off On Phase 3 installed equipment [9]

Residual Heat Removal 132 10E-2A Available Available Phase 3 installed equipment [10]

System Heat Exchanger A 133 10MOV-25A RHR A LPCI Inbd Inj Valve Closed Open Phase 2/3 installed equipment [9]

RHR Heat Exch A Bypass 134 10MOV-66A Open Closed Phase 3 installed equipment [9]

Valve Phase 2/3 installed equipment. May be 135 10MOV-27A RHR A LPCI Outbd Inj Valve Open Open excluded since normally open, required [9]

open.

RHR Pump A Suction Torus 136 10MOV-13A Open Closed 71MCC-153 [9]

Isol. Valve RHR Pump A SDC Suction 137 10MOV-15A Closed Open 71MCC-153 [9]

Isol Valve 138 10MOV-17 RHR SDC Outbd Isol. Valve Closed Open 71BMCC-4 [9]

139 10MOV-18 RHR SDC Inbd Isol. Valve Closed Open [9]

RHRSW A to RHR Cross Tie 140 10MOV-148A Closed Open Phase 2 installed equipment [10]

Upstr Isol Valve RHRSW A to RHR Cross Tie 141 10MOV-149A Closed Open Phase 2 installed equipment [10]

Dnstr Isol Valve RHRSW - Fire Protection 142 10RHR-432 Closed Open Manually opened [10]

Cross-Tie Isol Valve RHR Heat Exch A Service 143 10MOV-89A Closed Closed/ Open Closed (Ph 2) Open (Ph 3) [10]

Water Outlet Isol Valve Page A-12

For Information Only James A. FitzPatrick ESEP Report ESEL Equipment Operating State Item Notes/Comments References Number ID Description Normal State Desired State Fire Pump 76P-1 Temporary To be manually Opened 144 TBD Connection Isolation Valve Closed Open [3]

Not yet installed to RHRSW Piping To be permanently installed. Powered 145 TBD Jib Crane for FLEX Pump N/A Available [3]

from FLEX DG. Not yet installed Page A-13

For Information Only James A. FitzPatrick ESEP Report ATTACHMENT B - ESEP HCLPF VALUES AND FAILURE MODES TABULATION Page B-1

For Information Only James A. FitzPatrick ESEP Report HCLPF (g) /

Item Failure Equipment ID Equipment Description Screening Comments No. Mode Level 1 13P-1 RCIC Turbine Driven Pump >RLGM Screened Note 1 2 13E-1 RCIC Barometric Condenser >RLGM Screened Note 2 RCIC Turbine Lube Oil Note 2: In-Line 3 13E-2 >RLGM Screened Cooler Component 4 13TK-1 RCIC Vacuum Tank >RLGM Screened Note 2 RCIC Pump Suct from Cond 5 13MOV-18 >RLGM Screened Stor Isol Valve RCIC Pump Suct From 6 13MOV-41 >RLGM Screened Suppr Pool INBD Isol Valve RCIC Pump Suct From 7 13MOV-39 Suppr Pool Outboard Isol >RLGM Screened Valve RCIC Steam Supply INBD 8 13MOV-15 >RLGM Screened Isol Valve RCIC Turbine Steam Supply 9 13MOV-16 >RLGM Screened Outbd Isol Valve RCIC Turbine Steam Inlet 10 13MOV-131 >RLGM Screened Isol Valve 11 13HOV-1 RCIC Trip Valve >RLGM Screened RCIC Turbine Governor 12 13HOV-2 >RLGM Screened Valve RCIC Turb Lube Oil Cooler 13 13MOV-132 >RLGM Screened Water Supply Isol Valve RCIC Turb Lube Oil Cooler 14 13PCV-23 Water Supply Press Control >RLGM Screened Valve RCIC Pump Disch to 15 13MOV-20 >RLGM Screened Reactor Outbd Isol Valve RCIC Pump Disch to 16 13MOV-21 >RLGM Screened Reactor Inbd Isol Valve RCIC Barometric Cndsr 17 13P-3 >RLGM Screened Note 2 Vacuum Pump 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

For Information Only James A. FitzPatrick ESEP Report HCLPF (g) /

Item Failure Equipment ID Equipment Description Screening Comments No. Mode Level 21 Torus Suppression Pool >RLGM Screened Note 3 ADS Main Steam Line A 22 02RV-71A >RLGM Screened Safety/Relief Valve IAS 02RV-71A Air 23 39ACC-256A >RLGM Screened Note 2 Accumulator ADS/MST A 02TV-71A 24 02SOV-71A1 Auto/CR Manual Pilot >RLGM Screened Solenoid Valve ADS Main Steam Line A 25 02RV-71B >RLGM Screened Safety/Relief Valve IAS 02RV-71B Air 26 39ACC-256B >RLGM Screened Note 2 Accumulator ADS/MST A 02RV-71B 27 02SOV-71B1 Auto/CR Manual Pilot >RLGM Screened Solenoid Valve ADS Main Steam Line B 28 02RV-71C >RLGM Screened Safety/Relief Valve IAS 02RV-71C Air 29 39ACC-256C >RLGM Screened Note 2 Accumulator ADS/MST B 02RV-71C 30 02SOV-71C1 Auto/CR Manual Pilot >RLGM Screened Solenoid Valve ADS Main Steam Line B 31 02RV-71D >RLGM Screened Safety/Relief Valve IAS 02RV-71D Air 32 39ACC-256D >RLGM Screened Note 2 Accumulator ADS/MST B 02RV-71D 33 02SOV-71D1 Auto/CR Manual Pilot >RLGM Screened Solenoid Valve ADS Main Steam Line C 34 02RV-71E >RLGM Screened Safety/Relief Valve IAS 02RV-71E/F Air 35 39ACC-256E >RLGM Screened Note 2 Accumulator ADS/MST C 02RV-71E 36 02SOV-71E1 Auto/CR Manual Pilot >RLGM Screened Solenoid Valve Main Steam Line C Manual 37 02RV-71F >RLGM Screened Safety Relief Valve MST C 02RV-71F Control 38 02SOV-71F1 Room Manual Pilot >RLGM Screened Solenoid Valve ADS Main Steam Line C 39 02RV-71G >RLGM Screened Safety/Relief Valve Page B-3

For Information Only James A. FitzPatrick ESEP Report HCLPF (g) /

Item Failure Equipment ID Equipment Description Screening Comments No. Mode Level IAS 02RV-71G Air 40 39ACC-256G >RLGM Screened Note 2 Accumulator ADS/MST C 02RV-71G 41 02SOV-71G1 Auto/CR Manual Pilot >RLGM Screened Solenoid Valve ADS Main Steam Line D 42 02RV-71H >RLGM Screened Safety/Relief Valve IAS 02RV-71H Air 43 39ACC-256H >RLGM Screened Note 2 Accumulator ADS/MST D 02RV-71H 44 02SOV-71H1 Auto/CR Manual Pilot >RLGM Screened Solenoid Valve Main Steam Line D Manual 45 02RV-71J >RLGM Screened Safety Relief Valve IAS 02RV-71J Air 46 39ACC-256J >RLGM Screened Note 2 Accumulator MST D 02RV-71J Control 47 02SOV-71J1 Room Manual Pilot >RLGM Screened Solenoid Valve Main Steam Line A Manual 48 02RV-71K >RLGM Screened Safety Relief Valve IAS 02RV-71K Air 49 39ACC-256K >RLGM Screened Note 2 Accumulator MST A 02RV-71K Control 50 02SOV-71K1 Room Manual Pilot >RLGM Screened Solenoid Valve Main Steam Line D Manual 51 02RV-71L >RLGM Screened Safety Relief Valve IAS 02RV-71L Air 52 39ACC-256L >RLGM Screened Note 2 Accumulator MST D 02RV-71L Control 53 02SOV-71L1 Room Manual Pilot >RLGM Screened Solenoid Valve Safety-related Nitrogen 54 27TK-7A >RLGM Screened Note 1 Tank Safety-related Nitrogen 55 27TK-7B >RLGM Screened Note 1 Tank 56 76P-1 West Diesel Fire Pump >RLGM Screened Note 2 New FLEX Component Not Not 57 TBD Reliable Hardened Vent to be seismically Applicable Applicable designed.

58 02-3LI-85A RX Water Lvl >RLGM Screened Note 1 Page B-4

For Information Only James A. FitzPatrick ESEP Report HCLPF (g) /

Item Failure Equipment ID Equipment Description Screening Comments No. Mode Level Reactor Vessel Wide Range 59 02-3LT-85A >RLGM Screened Level Xmitter 60 13INV-152 Inverter 13-152 >RLGM Screened Note 1 61 13P/S-107 Single Nest Power Supply >RLGM Screened Note 1 Reactor Water A Level 62 06-LI-094A >RLGM Screened Note 1 Indicator 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 Reactor Vessel RHR 65 02-3LT-73 >RLGM Screened Interlock Level Xmitter Containment Spray Perm 66 02-3MTU-273 >RLGM Screened Note 1 Master Trip Unit 67 33LI-101A CST Level >RLGM Screened Note 1 Condensate Storage Tanks 68 33LT-101 >RLGM Screened Level Xmitter 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 ECCS Loop A Feedwater 72 06PT-61A Control Reactor Press >RLGM Screened Xmitter Reactor Press "A" Signal 73 06SCM-61A >RLGM Screened Note 1 Conditioner Reactor Press "A" Sig Dist 74 06SDM-61A >RLGM Screened Note 1 Module 75 27PI-115A1 NR PC Press Indicator >RLGM Screened Note 1 Drywell Div 1 Narrow 76 27PT-115A1 >RLGM Screened Range Press Xmitter CAD Drywell Press Div 1 77 27SCM-115A >RLGM Screened Note 1 Input Module CAD Drywell Press Div 1 78 27SDM-115A >RLGM Screened Note 1 Distribution Module Page B-5

For Information Only James A. FitzPatrick ESEP Report HCLPF (g) /

Item Failure Equipment ID Equipment Description Screening Comments No. Mode Level 79 27PI-115A2 WR PC Press Indicator >RLGM Screened Note 1 Drywell Div I Wide Range 80 27PT-115A2 >RLGM Screened Press Xmitter Suppression Chamber 81 27PR-101A >RLGM Screened Note 1 Monitor Press Recorder Torus Wide Range Press 82 27PT-101A >RLGM Screened Xmitter Suppression Chamber 83 27SDM-101A >RLGM Screened Monitor Signal Dist Module LRT Drywell Temp Mon 84 16-1TR-108 >RLGM Screened Note 1 Temp Recorder LRT Drywell Area 4 Resist 85 16-1RTD-108 >RLGM Screened Temp Detector LRT Drywell Temp Mon 86 16-1SDM-108 >RLGM Screened Signal Dist Module 87 10P/S-100A Power Supply >RLGM Screened Torus Bulk Temp Mon 88 16-1TR-131A >RLGM Screened Note 1 Average Temp Recorder Torus Bulk Temp Monitor 0 89 16-1RTD-131A Azimuth Bay L X-232 Resist >RLGM Screened Temp Detector Torus Temp Mon A Signal 90 16-1SDM-131A >RLGM Screened Note 1 Dist Module 91 23E/S-200A Power Supply PS-1A >RLGM Screened Note 1 92 23LI-203A PC Lvl Indicator >RLGM Screened Note 1 Wide Range Containment 93 23LT-203A1 Level HPCI Logic Level >RLGM Screened Xmitter (HI Tap)

Wide Range Containment 94 23LT-203A2 Level HPCI Logic Level >RLGM Screened Xmitter (LO Tap)

HPCI Drywell Sump Level 95 23SCM-203A >RLGM Screened Note 1 Div I Input Signal Module HPCI Drywell/Torus Diff 96 23SUM-203A >RLGM Screened Note 1 Press Subtraction Module Suppression Chamber 97 23LI-202A >RLGM Screened Note 1 Water Level Indic Suppression Pool HPCI 98 23LT-202A >RLGM Screened Logic Level Xmitter Page B-6

For Information Only James A. FitzPatrick ESEP Report HCLPF (g) /

Item Failure Equipment ID Equipment Description Screening Comments No. Mode Level HPCI Suppression Chamber 99 23SCM-202A Level Div I Input Signal >RLGM Screened Note 1 Module HPCI Drywell Sump Level 100 23SDM-203A >RLGM Screened Note 1 Div I Signal Distrib Module Reactor Protection and 101 25-05 >RLGM Screened Note 1 NSSS System Rack Jet Pump Instrument Rack 102 25-51 >RLGM Screened Note 1 25-51 103 27MAP Monitoring Analysis Panel >RLGM Screened Note 1 Nuclear Station Main 104 09-3 >RLGM Screened Note 1 Control Board RWCU & Recirc Control 105 09-4 >RLGM Screened Note 1 Panel Reactor Control Main 106 09-5 >RLGM Screened Note 1 Control Board BOP Main Control Board 107 09-6 >RLGM Screened Note 1 Panel(MECH)

Process Instrumentation 108 09-24 >RLGM Screened Note 1 Panel (Div I)

Relay Cabinet Channel A 109 09-30 >RLGM Screened Note 1 RCIC Panel Channel A RHR/RCIC Relay 110 09-32 >RLGM Screened Note 1 Panel Emergency Core Cooling 111 09-95 System DIV 1 A/C Trip >RLGM Screened Note 1 Cabinet New FLEX Component Not Not 112 TBD RHV Instrumentation to be seismically Applicable Applicable designed.

113 71SB-1 125 Volt Station Battery A >RLGM Screened Note 1 125 VDC Station Battery 114 71BC-1A >RLGM Screened Note 1 Charger 115 71BCB-2A Battery Control Board A >RLGM Screened Note 2 Relay Room Distribution 116 71DC-A2 >RLGM Screened Note 2 Cabinet Relay Room Distribution 117 71DC-A5 >RLGM Screened Note 2 Cabinet Page B-7

For Information Only James A. FitzPatrick ESEP Report HCLPF (g) /

Item Failure Equipment ID Equipment Description Screening Comments No. Mode Level Reactor Building West 118 71BMCC-1 Crescent Motor Control >RLGM Screened Note 1 Center Reactor Building West 119 71BMCC-3 Crescent Motor Control >RLGM Screened Note 1 Center Relay Room Safeguard 120 71ESSA1 >RLGM Screened Note 2 Power Distribution Panel New FLEX Component 120V Instrument Power Not Not 121 TBD to be seismically Inverter 1 Applicable Applicable designed.

Relay Room Emergency 122 71ACA2 >RLGM Screened Note 1 Power Distribution Panel Dist. Panel -

123 71ACUPS >RLGM Screened Note 1 Uninterruptible Bus 124 71UPP UPS Static Inverter >RLGM Screened Note 1 125 71INV-1A Instrument power inverter >RLGM Screened Note 2 4160V Switchgear 126 71H05 >RLGM Screened Note 1 Distribution (Bus 10500) 600V Motor Control Center 127 71MCC-156 >RLGM Screened Note 1 (Bus 115600) 600V Motor Control Center 128 71MCC-252 >RLGM Screened Note 1 Bus 125200 600V Motor Control Center 129 71MCC-254 >RLGM Screened Note 1 Bus 125400 Residual Heat Removal 130 10P-3A >RLGM Screened Note 1 Pump A 131 10P-2A RHR Keep-Full Pump A >RLGM Screened Note 2 Residual Heat Removal 132 10E-2A >RLGM Screened Note 3 System Heat Exchanger A 133 10MOV-25A RHR A LPCI Inbd Inj Valve >RLGM Screened RHR Heat Exch A Bypass 134 10MOV-66A >RLGM Screened Valve 135 10MOV-27A RHR A LPCI Outbd Inj Valve >RLGM Screened RHR Pump A Suction Torus 136 10MOV-13A >RLGM Screened Isol. Valve RHR Pump A SDC Suction 137 10MOV-15A >RLGM Screened Isol Valve Page B-8

For Information Only James A. FitzPatrick ESEP Report HCLPF (g) /

Item Failure Equipment ID Equipment Description Screening Comments No. Mode Level 138 10MOV-17 RHR SDC Outbd Isol. Valve >RLGM Screened 139 10MOV-18 RHR SDC Inbd Isol. Valve >RLGM Screened RHRSW A to RHR Cross Tie 140 10MOV-148A >RLGM Screened Upstr Isol Valve RHRSW A to RHR Cross Tie 141 10MOV-149A >RLGM Screened Dnstr Isol Valve RHRSW - Fire Protection 142 10RHR-432 >RLGM Screened Cross-Tie Isol Valve RHR Heat Exch A Service 143 10MOV-89A >RLGM Screened Water Outlet Isol Valve Fire Pump 76P-1 New FLEX Component Temporary Connection Not Not 144 TBD to be seismically Isolation Valve to RHRSW Applicable Applicable designed.

Piping New FLEX Component Not Not 145 TBD Jib Crane for FLEX Pump to be seismically Applicable Applicable 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

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...

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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...

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