Text

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.
	ML14356A426
Person / Time
Site:	Point Beach
Issue date:	12/22/2014
From:	Mccartney E; Point Beach
To:	; Document Control Desk, Division of Operating Reactor Licensing
References
	NRC 2014-0088
	Download: ML14356A426 (68)
	v • d • e

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December 22, 2014 NRC 2014-0088 10 CFR 50.54(f)

U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Point Beach Nuclear Plant, Units 1 and 2 Docket 50-266 and 50-301 Renewed License Nos. DPR-24 and DPR-27 NextEra Energy Point Beach. LLC'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)

References:

(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, dated March 12, 2012 (ML12073A348)

(2) NEI Letter, Proposed Path Forward for NTTF Recommendation 2.1: Seismic Reevaluations, dated April9, 2013, (ML13101A379)

(3) NRC Letter, Electric Power Research Institute Report 3002000704, "Seismic Evaluation Guidance: Augmented Approach for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1: Seismic," as an Acceptable Alternative to the March 12, 2012, Information Request for Seismic Reevaluations, dated May 7, 2013, (ML13106A331)

On March 12, 2012, the Nuclear Regulatory Commission (NRC) issued a 50.54(f) letter to all power reactor licensees and holders of construction permits in active or deferred status. of Reference (1) requested each addressee located in the Central and Eastern United States (CEUS) to submit a Seismic Hazard Evaluation and Screening Report within 1.5 years from the date of Reference (1).

In Reference (2), the Nuclear Energy Institute (NEI) requested NRC agreement to delay submittal of the final CEUS Seismic Hazard Evaluation and Screening Reports so that an update to the Electric Power Research Institute (EPRI) ground motion attenuation model could be completed and used to develop that information. NEI proposed that descriptions of subsurface materials and properties and base case velocity profiles be submitted to the NRC by September 12, 2013, with the remaining seismic hazard and screening information submitted by March 31,2014. NRC agreed with that proposed path forward in Reference (3).

NextEra Energy Point Beach, LLC, 6610 Nuclear Road, Two Rivers, WI 54241

Document Control Desk Page2 Reference (1) requested that licensees provide interim evaluations and actions taken or planned to address the higher seismic hazard relative to the design basis, as appropriate, prior to completion of the risk evaluation. In accordance with the NRC endorsed guidance in Reference (3), the attached Expedited Seismic Evaluation Process Report for NextEra Energy Point Beach, LLC provides the information described in Section 7 of Electrical Power Research Institute Report 3002000704 in accordance with the schedule identified in Reference (2).

This letter contains seven new Regulatory Commitments. These commitments are listed in Section 8.4, Summary of Regulatory Commitments. There are no changes to any existing Regulatory Commitments.

If you have any questions please contact Mr. Michael Millen, Licensing Manager, at 920/755-7845.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on December 22, 2014.

Very truly yours, NextEra Energy Point Beach, LLC f'rvz~

Eric McCartney Site Vice President Enclosure cc: Director, Office of Nuclear Reactor Regulation Administrator, Region Ill, USNRC Resident Inspector, Point Beach Nuclear Plant, USNRC Project Manager, Point Beach Nuclear Plant, USNRC

ENCLOSURE NEXTERA ENERGY POINT BEACH, LLC UNITS ONE AND TWO EXPEDITED SEISMIC EVALUATION PROCESS(ESEP)REPORT Page 1 of 66

EXPEDITED SEISMIC EVALUATION PROCESS 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 [Ref. 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 the Point Beach Nuclear Plant (PBNP). The intent of the ESEP is to perform an interim action in response to the NRC's 50.54(f) letter [Ref. 1] 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 EPRI 3002000704, Seismic Evaluation Guidance: Augmented Approach for the Resolution of Fukushima Near-Term Task Force Recommendation 2. 1: Seismic [Ref. 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 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 Point Beach FLEX strategies for Reactor Core Cooling and Heat Removal, Reactor Inventory Control/Long-term Subcriticality, Spent Fuel Pool (SFP) cooling and Containment Function are summarized below. This summary is derived from the Point Beach Overall Integrated Plan (OIP) in Response to the March 12, 2012, Commission Order EA-12-049 [Ref. 3].

Upon the reactor trip, reactor core cooling is accomplished by natural circulation of the Reactor Coolant System (RCS) through the Steam Generators (SGs). The SGs are supplied by the Auxiliary Feedwater (AFW) system and steam pressure is initially controlled by the Atmospheric Dump Valves (ADVs). If instrument air is unavailable steam pressure will be controlled by the operation of the Main Steam Safety Valves (MSSVs) until the ADVs are manually controlled. The main active component Page 2 of 66

associated with this strategy is the Turbine-Drive Auxiliary Feedwater Pump (TDAFW) pump, which is automatically actuated to provide feedwater from the CSTs to the SGs for the removal of reactor core decay heat. A modification will be performed on both CSTs to provide seismic qualification and protection from tornado generated missiles to a tank level of 6 feet which will provide a volume of 14,100 gallons of available water per tank. Operator action is initiated to swap the suction supply from the CST to Service Water (SW). SW will be supplied by the Diesel Driven Fire Pump (DDFP) via a cross connection between fire water and service water. The DDFP is being replaced and upgraded to make it seismically robust.

Several actions are required during Phase 2 following the event for reactor core cooling.

The main strategy is dependent upon the continual operation of the TDAFW pumps, which are only capable of feeding the Steam Generators as long as there is sufficient steam pressure to drive the TDAFW pump turbines.

If SGs are unavailable in MODES 5 and 6 and the refueling cavity is not flooded, the RCS will heat up and boil. Makeup flow to the RCS will be established from the accumulator(s) via the fill line. The accumulator fill line is connected to the Safety Injection (SI) cold leg injection line and when aligned will provide make up directly to the reactor vessel. In MODE 5 and 6 and SGs are unavailable, at least one accumulator will be procedurally controlled and maintained available with a hot leg vent path established whenever possible. For Phase 2 MODES 5 and 6 a Portable Diesel Driven Pump (PDDP), capable of at least 300 gpm to address boric acid precipitation concerns, will supply borated water from the Refueling Water Storage Tank (RWST) to the RCS using pre-established primary or secondary connection points on the Residual Heat Removal (RHR) system piping.

Reactor Inventory Control/Long-term Subcriticality strategy consists of reactor coolant system borated make-up via the primary make-up connections and a portable diesel driven pump. Cooldown of the RCS will commence approximately 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after the Beyond Design Basis External Event (BDBEE). The Reactor Coolant Pump (RCP) seals will be upgraded with low leakage Westinghouse Generation 3 SHIELD RCP seals.

Since the low leakage seals will allow negligible RCS inventory losses, RCS makeup is no longer required to achieve a stable steady state in Phase 1 with the reactor core being cooled.

Reactor coolant system (RCS) inventory reduction is a result of water volume reduction due to cooldown, reactor coolant pump seal leakage, and letdown via head-vents and/or pressurizer Power Operated Relief Valve (PORVs). To avoid adverse effects on the RCS natural circulation flow, the accumulator block isolation valves are electrically closed prior to commencing the cooldown to prevent nitrogen injection into the reactor coolant system.

There are no Phase 1 FLEX actions to maintain containment integrity. During Phase 2, containment pressure and temperature are monitored to ensure the containment safety function is not challenged. For the at-power event leakage from the RCS to containment is limited by the low leakage RCP seals. For the shutdown event the RCS is allowed to boil and steam is released to containment. If containment conditions warrant, a PDDP will supply water to the containment spray system via an adapter that will replace the cover of a spray pump discharge check valve. Manual venting is also an option.

Page 3 of 66

The Spent Fuel Pool (SFP) temperature is allowed to increase to the boiling point.

Water will be added (Phase 2) well before fuel becomes uncovered. The Primary Auxiliary Building (PAB) will be vented by opening the PAB truck access doors and the 66' Elevation personnel doors as necessary based on PAB conditions. Water is added to the SFP with a POOP and hoses using either direct addition or spray. The POOP will draw raw water from the Pump House Forebay, Pump bay, or directly from Lake Michigan. A connection point has been added that will allow the addition of raw water from the POOP to the SFP without accessing the refueling deck.

The safety-related 125V system consists of four main distribution buses: D-01, D-02, D-03, and D-04. The D-01 (train A) and D-02 (train B) main DC distribution buses supply power for control, emergency lighting, and the red and blue 120 VAC Vital Instrument bus (Y) inverters. The D-03 (train A) and D-04 (train B) main DC distribution buses supply power for control and the white and yellow 120 VAC Vital Instrument (Y) buses. A battery load management strategy has been developed to provide power to credited installed equipment (e.g., DC Motor Operated Valves (MOVs), Solenoid Operated Valves (SOVs), etc) and at least one channel of credited instrumentation during Phase 1. During Phase 2 onsite portable equipment is used to restore battery chargers, replenish fuel oil tanks, and augment plant lighting, ventilation, freeze protection, and communication systems as necessary. 480 VAC Portable Diesel Generator (PDG) will be used to power credited installed equipment via the safety related 480 VAC distribution system. The primary connection points will be at 1B-03 and 2B-03 which are A Train 480V vital buses located in the Cable Spreading Room (CSR).

The Phase 2 portable equipment and connection points will maintain the safety functions for an extended time. Point Beach did not identify any specific Phase 3 requirements.

Equipment provided by the Regional Response Centers can be used to replace phase 2 equipment and for recovery. A connection point(s) for a 4kV portable generator has been identified as a backup and to support recovery.

3.0 Equipment Selection Process and ESEL The selection of equipment for the Expedited Seismic Equipment List (ESEL) followed the guidelines of EPRI 3002000704 [Ref. 2]. The Point Beach design and FLEX strategy relies on some common equipment that supports both units. Backup strategies also rely on opposite unit equipment. Because of the reliance on common equipment and opposite unit equipment a combined ESEL was developed for the Point Beach units that support the Point Beach Overall Integrated Plan (OPI). The ESEL for Unit 1 and Unit 2 is presented in Attachment A.

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 Phases 1, 2 and 3 mitigation of a Beyond Design Basis External Event (BDBEE), as outlined in the OIP in Response to the March 12, 2012, Commission Order EA-12-049 [Ref. 3].

The OIP provides the Point Beach 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, spent fuel pool cooling and containment integrity consistent with the Point Beach OIP [Ref. 3].

FLEX recovery actions are excluded from the ESEP scope per Page 4 of 66

EPRI 3002000704 [Ref. 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, subcriticality, containment integrity and spent fuel pool cooling functions. Portable and pre-staged FLEX equipment (not permanently installed) are excluded from the ESEL per EPRI 3002000704

[Ref. 2].

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 Point Beach OIP [Ref. 3].

2) The scope of components is limited to installed plant equipment, and FLEX connections necessary to implement the Point Beach OIP [Ref. 3] 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

[Ref. 2] guidance are:

Structures (e.g. containment, reactor building, control building, auxiliary building, etc.)

Piping, cabling, conduit, HVAC, and their supports.

Manual 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. reactor pressure vessel 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 is included in the ESEL.

Page 5 of 66

3.1.1 ESEL Development The ESEL was developed by reviewing the Point Beach OIP [Ref. 3] to determine the major equipment involved in the FLEX strategies. Further reviews of plant drawings (e.g., Process and Instrumentation Diagrams (P&IDs) and Electrical One Line Diagrams) were performed to identify the boundaries of the flowpaths to be used in the FLEX strategies and to identify specific components in the flowpaths 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 I branch lines off the defined strategy electrical or fluid flowpath. 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, etc., as necessary.

3.1.2 Power Operated Valves Page 3-3 of EPRI 3002000704 [Ref. 2] notes that power operated valves not required to change state 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/AFW trips)." To address this concern, the following guidance is applied in the Point Beach ESEL for functional failure modes associated with power operated valves:

Power operated valves that remain energized during the Extended Loss of all 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 add to the ESELs as these components provide completely passive locations for pulling or installing cables. No breaks or connections in the cabling are included in pull boxes. Pull boxes were considered part of conduit and cabling, which are excluded in accordance with EPRI 3002000704 [Ref. 2].

Page 6 of 66

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.

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 Phase 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 Point Beach OIP [Ref. 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 [Ref. 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.

3.1. 7 Relays and Contactors The FLEX Phase 1 components were reviewed to identify relays and contactors. The relays and contactors were reviewed to identify relays and contactors that may lead to circuit seal-ins or lock-outs, and take the electrical circuit to a state different than is desired in the FLEX strategy.

Those relays and contactors leading to sealing-in or locking-out circuits were included on the ESEL.

3.1.8 Breakers Generally, the seismic qualification relied upon for the fragility evaluation is based on the panel/cabinet, not on individual components. Breakers were evaluated by including the panel/cabinet on the ESEL unless there was a seal-in circuit /lock out relay that could preclude manual operation.

3.2 No exceptions were taken for use of equipment that is not the primary means for FLEX Implementation.

The complete ESEL for Unit 1 and Unit 2 is presented in Attachment A Page 7 of 66

4.0 Ground Motion Response Spectrum (GMRS) 4.1 Plot of GMRS Submitted by the Licensee As discussed in Section 3.2 of the March submittal report [Ref. 4] the SSE Control Point elevation is +8.0 ft., which is the highest foundation of key safety-related structures.

The GMRS provided in the March submittal report [Ref. 4] is tabulated and graphed below:

TABLE 4-1 PBNP GMRS Freq (Hz) GMRS (g) Freq (Hz) GMRS (g) 0.1 8.60E-03 4 2.32E-01 0.125 1.08E-02 5 2.44E-01 0.15 1.29E-02 6 2.38E-01 0.2 1.72E-02 7 2.42E-01 0.25 2.15E-02 8 2.52E-01 0.3 2.58E-02 9 2.58E-01 0.35 3.01 E-02 10 2.67E-01 0.4 3.44E-02 12.5 2.75E-01 0.5 4.30E-02 15 2.67E-01 0.6 4.78E-02 20 2.47E-01 0.7 5.11 E-02 25 2.31 E-01 0.8 5.45E-02 30 2.14E-01 0.9 5.89E-02 35 2.00E-01 1 6.50E-02 40 1.86E-01 1.25 9.19E-02 50 1.63E-01 1.5 1.15E-01 60 1.49E-01 2 1.45E-01 70 1.44E-01 2.5 1.71 E-01 80 1.41 E-01 3 1.87E-01 90 1.40E-01 3.5 2.13E-01 100 1.40E-01 Page 8 of 66

FIGURE 4-1 PBNP GMRS PLOT 0.3 0.25 I

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4.2 Comparison to SSE As identified in the March submittal report [Ref. 4], the GMRS exceeds the SSE in the 1-10 hz range as shown in the table and graph below:

TABLE 4-2 PBNP GMRS vs. SSE Freq. GMRS Horizontal (Hz) (unsealed, g) SSE (g) 1 0.065 0.110 1.25 0.092 0.130 1.5 0.115 0.149 2 0.126 0.160 2.5 0.145 0.169 3 0.171 0.180 3.5 0.187 0.185 4 0.213 0.190 5 0.232 0.194 6 0.244 0.200 7 0.238 0.184 8 0.242 0.171 9 0.252 0.159 10 0.258 0.149 Page 10 of 66

FIGURE 4-2 PBNP GMRS vs. SSE PLOT 0.3

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5.0 Review Level Ground Motion (RLGM) 5.1 Description of RLGM Selected The RLGM for PBNP was determined in accordance with Section 4 of EPRI 30020000704 [Ref. 2] as being derived by linearly scaling the PBNP SSE by the maximum ratio of the GMRS/SSE between the 1 and 10 hertz range.

The ratio between the GMRS and SSE at 5% damping is tabulated below. Note that the acceleration values for the SSE spectrum or the GMRS which are not provided explicitly in the source documentation at intermediate points are developed by interpolating between the nearest available values.

TABLE 5-1 RATIO BETWEEN GMRS AND SSE Freq. GMRS Horizontal SSE SF=

(Hz) (unsealed, g) (Q) GMRS/SSE 1 0.065 0.110 0.59 1.25 0.092 0.130 0.71 1.5 0.115 0.149 0.77 1.67 0.126 0.160 0.79 2 0.145 0.169 0.86 2.5 0.171 0.180 0.95 3 0.187 0.185 1.01 3.5 0.213 0.190 1.12 4 0.232 0.194 1.20 5 0.244 0.200 1.22 6 0.238 0.184 1.29 7 0.242 0.171 1.42 8 0.252 0.159 1.58 9 0.258 0.149 1.73 10 0.267 0.140 1.91 The maximum ratio between the 5% damping GMRS and horizontal SSE occurs at 10Hz and equals 1.91.

The resulting RLGM based on increasing the horizontal SSE by the maximum ratio of 1.91 is plotted below. Per DG-C03 [Ref. 19], the vertical response spectrum is equal to 2/3 times the horizontal ground response spectrum.

Therefore, the vertical RLGM is equal to 2/3 times the horizontal RLGM.

Page 12 of 66

TABLE 5-2 PBNP RLGM RLGM Freq.

Horizontal Vertical (Hz)

(g) (g) 0.33 0.0860 0.0573 0.50 0.1222 0.0815 1.00 0.2101 0.1401 1.25 0.2483 0.1655 1.67 0.3056 0.2037 2.50 0.3438 0.2292 5.00 0.3820 0.2547 10.00 0.2674 0.1783 12.50 0.2292 0.1528 16.67 0.2292 0.1528 25.00 0.2292 0.1528 35.71 0.2292 0.1528 Page 13 of 66

FIGURE 5-2 PLOT OF HORTIZONTAL RLGM 1

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Page 14 of 66

FIGURE 5-3 PLOT OF VERTICAL RLGM

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Page 15 of 66

5.2 Method to Estimate ISRS The method used to derive the ESEP in-structure response spectra (ISRS) was to scale the existing SSE-based ISRS obtained from DG-C03, Revision 3, "Seismic Design Criteria Guideline" [Ref. 19] by the maximum ratio of 1.91. The scaled ISRS was determined for all buildings and elevations where ESEL items are located at PBNP. These scaled ISRS are sometimes referred to as the In-Structure Review Level Ground Motion (ISRLGM).

An exception has been made for the Recirculating Water Storage Tanks (RWST). These tanks are founded on an independent slab which is isolated from the surrounding buildings and located on grade. Because the effect of the slab on the seismic demand of the RWST is negligible and the slab responds independently from the nearby buildings, the seismic demand at the control point of El 8.0 ft. may be used as the seismic demand for the tank. Because the fluid-structure modal frequency of the tank is lower than the frequency of the applicable ground response spectrum at the peak acceleration, the Soil-Structure Interaction (SSI) effects on these tanks may be ignored, per Step 4 of Section 7.3.2 of Seismic Qualification Utilities Group (SQUG), "Generic Implementation Procedure (GIP) for Seismic Verification of Nuclear Plant Equipment", Rev. 2a

[Ref. 20]. As such, the GMRS was used directly as the RLGM (seismic demand) for these tanks.

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 peak ground acceleration (PGA) for which there is a high confidence of a low probability of failure (HCLPF). The PGA is associated with a specific spectral shape, in this case the 5°/o-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 [Ref. 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, A Methodology for Assessment of Nuclear Power Plant Seismic Margin (Revision 1) [Ref. 7].

2. Probabilistic approach using the fragility analysis methodology of EPRI TR-103959, Methodology for Developing Seismic Fragilities [Ref. 8].

For PBNP, the deterministic approach using the CDFM methodology of EPRI NP-6041

[Ref. 7] was used to determine HCLPFs.

6.1 Summary of methodologies used PBNP applied the methodology of EPRI NP-6041 [Ref. 7] to all items on the ESEL. The screening walkdowns used the screening tables from Chapter 2 of EPRI NP-6041 [Ref. 7]. The walkdowns were conducted by engineers who as a minimum attended the SQUG Walkdown Screening and Seismic Evaluation Page 16 of 66

Training Course. The walkdowns were documented on Screening Evaluation Work Sheets (SEWS) from EPRI NP-6041 [Ref. 7]. Anchorage capacity calculations were determined using the CDFM criteria from EPRI NP-6041

[Ref. 7] with PBNP specific allowables and material strengths used as applicable.

Seismic demand was the RLGM provided in Table 5-2 and Figures 5-2 and 5-3.

6.2 HCLPF screening process The peak spectral acceleration of the RLGM (amplified PGA) for PBNP equals 0.382 (Table 5-2). Table 2-4 of EPRI NP-6041 [Ref. 7] is based on ground peak spectral accelerations of 0.8g and 1.2g. These both exceed the PBNP RLGM peak spectral acceleration. The PBNP ESEL components were screened against the 0.8g column of Table 2-4 of NP-6041 [Ref. 7].

The combined Unit 1 and Unit 2 ESEL contains 241 items. The components in the ESEL were evaluated to the EPRI NP-6041 [Ref. 7] caveats and documented on the equipment SEWS.

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 [Ref. 2], which refers to EPRI NP-6041

[Ref. 7] for the Seismic Margin Assessment process. Pages 2-26 through 2-30 of EPRI NP-6041 [Ref. 7] 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 Page 17 of 66

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.

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 Seismic 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 Sf [Seismic lnteraction 1] 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 0 gives guidance for sampling selection of EPRI 3002000704 [Ref. 2], which refers to EPRI NP-6041 [Ref. 7] for the Seismic Margin Assessment process. Pages 2-26 through 2-30 of EPRI NP-6041 [7] describe the seismic wa/kdown criteria, including the following key criteria.

The PBNP walkdowns included as a minimum a 100% walk-by of all items on the ESEL except as noted in Section 7.0. Any previous walkdown information that was relied upon for SRT judgment is documented in Section 6.3.2.

6.3.2 Application of Previous Walkdown Information Documentation available via PBNP's Seismic Qualification Utility Group (SQUG) program was frequently used to enhance the screening process.

The walkdown information from the SQUG program was used as a basis for acceptability in the ESEP for the following components: 2TE-451A and 1

EPRI 3002000704 [Ref. 2] page 5-4 limits the ESEP seismic interaction reviews to "nearby block walls" and "piping attached to tanks" which are reviewed "to address the possibility of failures due to differential displacements." Other potential seismic interaction evaluations are "deferred to the full seismic risk evaluations performed in accordance with EPRI 1025287 [Ref. 15].

Page 18 of 66

2TE-451 C, as well as the internal mountings for all electrical panels/cabinets that were not opened and inspected by the SRT.

Previous NTTF 2.3 seismic walkdowns [Ref. 17] were not used to support the ESEP seismic evaluations.

6.3.3 Significant Walkdown Findings The following findings were noted during the walkdowns.

The lateral support for Valve 2SC-953 was found to have two missing bolts.

AR 01955412 was written to address these missing bolts. The two bolts have been installed and the AR has been closed.

The lateral support for Valve 1SC-953 was found to have two missing bolts.

AR 01998370 was written to address these missing bolts. Bolt hole misalignment prevented installing both bolts. The SRT has concluded that a single bolt provides adequate lateral support. One bolt has been installed and the AR has been closed.

Several block walls were identified in the proximity of ESEL equipment.

These block walls were assessed for their structural adequacy to withstand the seismic loads resulting from the RLGM. For these cases, the block wall is noted on the ESEL HCLPF tables in Attachment B. Where the HCLPF is below the RLGM plant modifications will be performed as identified in section 8.2 Identification of Planned Modifications.

No other significant outliers or anchorage concerns were identified during the PBNP seismic walkdowns.

6.4 HCLPF Screening Process ESEL items were evaluated using the criteria in EPRI NP-6041 [Ref. 7]. Those evaluations included the following steps:

Performing seismic capability walkdowns for equipment to evaluate the equipment installed plant conditions

Performing screening evaluations using the screening tables in EPRI NP-6041 [Ref. 7] as described in Section 6.2 and

Performing HCLPF calculations considering various failure modes that include both structural failure modes (e.g. anchorage, load path etc.) and functional failure modes [Note: Functional failure modes are for relays only].

All HCLPF calculations were performed using the CDFM methodology.

Anchorage for components was evaluated using SRT judgment, reviewing large margins in existing design basis calculations, or performing CDFM HCLPF calculations [Ref. 10]. These evaluations are summarized in Attachment B. For components located higher than 40 feet above grade, Table 2-4 of NP-6041

[Ref. 7] is not valid. Page 5-4 of EPRI 3002000704 [Ref. 2] references the EPRI document 1019200, "Seismic Fragility Applications Guide Update" [21] with Page 19 of 66

respect to screening criteria beyond 40 feet above grade. Section 4-2 of this document specifies 1.5 as an appropriate factor to evaluate the HCLPF capacity of structure-mounted items. As such, the Table 2-4 screening lanes' spectral accelerations are multiplied by a factor of 1.5 in order to account for spectral acceleration at the base of the component. This screening level at the base of the components is compared to the ISRLGM corresponding to the RLGM.

6.5 Functional Evaluation of Relays A HCLPF evaluation was performed for all relays and contactors included on the PBNP ESEL.

For relay evaluations, NP-6041-SL Appendix Q describes the following evaluation steps:

Calculate in-cabinet response spectra (ICRS)

Establish a clipping factor to be applied to the ICRS

Determine a relay's capacity based on GERS or component testing

Establish adjustment factors to convert the relay's capacity to a CDFM level

Compare demand to the capacity HCLPF capacities for the relays on the PBNP ESEL were calculated and are presented in Attachment B. Note that clipping factors were not used in the evaluations because they were not needed to show that relays' capacities are acceptable. Parent components are not assigned the HCLPF of the contained relays in Attachment B.

6.6 Tabulated ESEL HCLPF Values (Including Key Failure Modes)

Tabulated ESEL HCLPF values including the key failure modes are included in Attachment B for all items on the ESEL.

For items screened out using NP 6041 [Ref. 7] screening tables, the screening level can be provided as RLGM and the failure mode can be 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 relays evaluated "functional failure" is listed as the failure mode.

Page 20 of 66

7.0 Inaccessible Items 7.1 Identification of ESEL items inaccessible for walkdowns The following table lists the ESEL items that were not walked down, a discussion on why these items were not walked down, and states whether further action (i.e.

future walkdown) is required.

-= _:_

-- ~

Further_

-Eq'-lipmE'!nt -

,-Description_

_Building Discussion action ID _- --

req'd?-'-

This component is located in a high radiation area and is not easily accessible. For ALARA purposes, this U1 Nl Wide Unit 1 component was not walked down. The 1N-40 No Range Containment component itself is inherently rugged and is judged not to be a concern. Anchorage screened by large available margin in existing design basis calculation.

For ALARA purposes, this item was not walked down. Based upon the inherent seismic ruggedness of thermocouples and by comparison to a number of In core evaluations for similar thermocouples Unit 1 1TE-00019 Thermocouple at (see 1TE-450A, 1TE-450D, HE-451 B, No Containment B-5 1TE-451 C, 2TE-450B, 2TE-450C, 2TE-451A, 2TE-451 B, 2TE-451 C, and 2TE-451 D), the adequacy of this equipment is judged to be acceptable. No further action is reg_uired.

For ALARA purposes, this item was not walked down. Based upon the inherent seismic ruggedness of thermocouples and by comparison to a number of lncore evaluations for similar thermocouples Unit 1 1TE-00037 Thermocouple at (see 1TE-450A, 1TE-450D, HE-451 B, No Containment K-11 1TE-451 C, 2TE-450B, 2TE-450C, 2TE-451 A, 2TE-451 B, 2TE-451 C, and 2TE-451 D), the adequacy of this equipment is judged to be acceptable. No further action is required.

This component is located in a high radiation area and is not easily accessible. For ALARA purposes, this U2 Nl Wide Unit 2 component was not walked down. The 2N-40 No Range Containment component itself is inherently rugged and is judged not to be a concern. Anchorage screened by large available margin in existing design basis calculation.

Page 21 of 66

Equipltleht * * * * * * : Further

_: action

-..',,:*'.,** ID . '~,,, 3i r~q*~?

Equipment added to ESEL after Unit 2 RV Head Vent Unit 2 Containment walkdown. Equipment is 2RC-570A Yes Solenoid Containment scheduled to be walked down during next Unit 2 outage, fall 2015.

Equipment added to ESEL after Unit 2 RV Head Vent Unit 2 Containment walkdown. Equipment is 2RC-570B Yes Solenoid Containment scheduled to be walked down during next Unit 2 outage, fall 2015.

Equipment added to ESEL after Unit 2 RV!T-1 PZR Unit 2 Containment walkdown. Equipment is 2RC-575A Vent Header to Yes Containment scheduled to be walked down during next PRT Solenoid Unit 2 outage fall 2015.

RV!T-1 PZR Gas Equipment added to ESEL after Unit 2 Ventto Cont. Unit2 Containment walkdown. Equipment is 2RC-575B Yes Standpipe Containment scheduled to be walked down during next Solenoid Unit 2 outage, fall 2015.

For ALARA purposes, this item was not walked down. Based upon the inherent seismic ruggedness of thermocouples and by comparison to a number of In core evaluations for similar thermocouples Unit2 2TE-00023 Thermocouple at (see 1TE-450A, 1TE-450D, HE-451 B, No Containment D-7 HE-451 C, 2TE-450B, 2TE-450C, 2TE-451 A, 2TE-451 B, 2TE-451 C, and 2TE-451 D), the adequacy of this equipment is judged to be acceptable. No further action is r~uired.

For ALARA purposes, this item was not walked down. Based upon the inherent seismic ruggedness of thermocouples and by comparison to a number of In core evaluations for similar thermocouples Unit 2 2TE-00038 Thermocouple at (see 1TE-450A, 1TE-450D, HE-451 B, No Containment L-10 1TE-451 C, 2TE-450B, 2TE-450C, 2TE-451 A, 2TE-451 B, 2TE-451 C, and 2TE-451 D), the adequacy of this equipment is judged to be acceptable. No further action is re_quired.

Item is inaccessible because it is in a RC Loop B Cold LHRA. Equipment judged acceptable Leg Unit 2 upon review of documentation and photo 2TE-451A No Temperature Containment provided in A-46 SEWS SQ-001 052 and RTD by comparison to similar equipment (2TE-450C).

Page 22 of 66

- Further

Equipment Description -Building Discussion action ID req'd? .

Item is inaccessible because it is in a RC Loop B Cold LHRA. Equipment judged acceptable Leg Unit 2 upon review of documentation and photo 2TE-451C No Temperature Containment provided in A-46 SEWS SQ-001 056 and RTD by comparison to similar equipment (2TE-450C).

Diesel Driven Equipment not yet installed. Evaluation C-601 Fire Pump CWPH Yes required after installation.

Control Panel Diesel Driven Equipment not yet installed. Evaluation D-600 Fire Pump CWPH Yes required after installation.

Battery Rack Diesel Fire Equipment not yet installed. Evaluation FP-3715 Pump Relief CWPH Yes required after installation.

Valve Fire water Equipment not yet installed. Evaluation FP-448 CWPH Yes header isolation required after installation.

Fire water to SW Equipment not yet installed. Evaluation FP-536 CWPH Yes cross connection required after installation.

Diesel Driven Equipment not yet installed. Evaluation P-358 CWPH Yes Fire Pump required after installation.

Diesel Driven Equipment not yet installed. Evaluation P-358-E Fire Pump CWPH Yes required after installation.

Engine Diesel Fire Equipment not yet installed. Evaluation T-30 CWPH Yes Pump Fuel Tank required after installation.

Page 23 of 66

7.2 Planned Walkdown I Evaluation Schedule I Close Out The schedule for performing the walkdowns for the inaccessible and late addition components as listed in Section 7.1 is during the Unit 2 Refueling Outage U2R34 schedule for the fall 2015. The screening and evaluation of these components will be complete within 90 days following the conclusion of the U2R34 refueling outage. The Commitments associated with these tasks are included in Section 8.4.

Equip Planned . . Evaluation Close Description Building Discussion 10 WD Out Equipment added U2 Outage 41r QTR 1s QTR 2RC- RV Head Vent Unit2 to ESEL after Unit Fall2015 2015 2016 570A Solenoid Contain. 2 Containment walkdown.

Equipment added U2 Outage 4mQTR 1s1 QTR 2RC- RV Head Vent Unit 2 to ESEL after Unit Fall2015 2015 2016 5708 Solenoid Contain. 2 Containment walkdown.

Equipment added U2 Outage 4mQTR 1s QTR RVff-1 PZR 2RC- Unit 2 to ESEL after Unit Fall2015 2015 2016 Vent Header to 575A Contain. 2 Containment PRT Solenoid walkdown.

RVff-1 PZR U2 Outage 4mQTR 1sT QTR Equipment added Gas Vent to Fall2015 2015 2016 2RC- Unit 2 to ESEL after Unit Cont.

5758 Contain. 2 Containment Standpipe walkdown.

Solenoid C-601 Diesel Driven CWPH Equipment not yet Prior to U2 4mQTR 1s1 QTR Fire Pump installed. Outage Fall 2015 2016 Control Panel 2015 D-600 Diesel Driven CWPH Equipment not yet Prior to U2 4mQTR 1s QTR Fire Pump installed. Outage Fall 2015 2016 Battery_ Rack 2015 FP-3715 Diesel Fire CWPH Equipment not yet Prior to U2 4mQTR 1s1 QTR Pump Relief installed. Outage Fall 2015 2016 Valve 2015 FP-448 Fire water CWPH Equipment not yet Prior to U2 4" QTR 1s QTR header installed. Outage Fall 2015 2016 isolation 2015 FP-536 Fire water to CWPH Equipment not yet Prior to U2 4Tr QTR 1sT QTR SW cross installed. Outage Fall 2015 2016 connection 2015 P-358 Diesel Driven CWPH Equipment not yet Prior to U2 4mQTR 1s1 QTR Fire Pump installed. Outage Fall 2015 2016 2015 P-358-E Diesel Driven CWPH Equipment not yet Prior to U2 4m QTR 1s QTR Fire Pump installed. Outage Fall 2015 2016 EnQine 2015 T-30 Diesel Fire CWPH Equipment not yet Prior to U2 4mQTR 1sT QTR Pump Fuel installed. Outage Fall 2015 2016 Tank 2015 Page 24 of 66

8.0 ESEP Conclusions and Results 8.1 Supporting Information PBNP has performed the ESEP as an interim action in response to the NRC's 50.54(f) letter [Ref. 1]. It was performed using the methodologies in the NRC endorsed guidance in EPRI 3002000704 [Ref. 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 PBNP response to the NRC's 50.54(f) letter [1 ].

On March 12, 2014, NEI submitted to the NRC results of a study [Ref. 12] 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 [Ref. 14]

concluded that the "fleetwide 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 PBNP was included in the fleet risk evaluation submitted in the March 12, 2014 NEI letter [Ref. 12] therefore, the conclusions in the NRC's May 9 letter [Ref. 14] also apply to PBNP.

In addition, the March 12, 2014 NEIIetter [Ref. 12] provided an attached "Perspectives on the Seismic Capacity of Operating Plants," which (1) assessed a number of qualitative reasons why the design of 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 structures, systems and components (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 Page 25 of 66

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

Bounding requirements in codes and standards

Use of minimum strength requirements of structural components (concrete and steel)

Bounding testing requirements, and

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.

8.2 Identification of Planned Modifications Insights from the ESEP identified the following four items where the HCLPF is below the RLGM and plant modifications will be made in accordance with EPRI 3002000704 [Ref. 2] to enhance the seismic capacity of the plant:

1. Masonry Wall 111-2/23 has a HCLPF below the RLGM and requires modification. This wall is located along the West side of the control room.

The ESEL items affected by this wall are 2Y-01, 2Y-03, C-01, 1C-03, 1C-04, C-02, 2C-03, & 2C-04. The proposed modification includes the addition of a post at mid-span of the wall in order to reduce the span length of the wall, reducing in-plane wall stresses to acceptable levels.

2. Masonry Wall 111-4N/23 has a HCLPF below the RLGM and requires modification. This wall is located along the West side of the control room.

The ESEL items affected by this wall are C-01, 1C-03, 1C-04, C-02, 2C-03, and 2C-04. The proposed modification includes the addition of a post at mid-span of the wall in order to reduce the span length of the wall, reducing in-plane wall stresses to acceptable levels.

3. The Work Control Center (WCC) block walls on the Turbine Deck have a HCLPF below the RLGM and requires modification. The ESEL items affected by this wall are LT-4038, LT-4041, T-24A, & T-24B. A modification (i.e. reinforcement of the block walls or relocation of soft targets away from the path of falling debris) must be installed such that falling debris will not affect the level transmitters or sensitive tubing attached to the Condensate Storage Tanks located below the WCC.

Page 26 of 66

4. The evaluation of the anchorage for the Condensate Storage Tanks {T-24A and T -24B) is acceptable only after the installation of the approved anchorage modification (Engineering Change (EC) 279034, NRC Order Fukushima FLEX CSTs- Seismically Upgrade and Missile Protect Bottom 6 feet). The seismic upgrade of the CST was listed as a Pending Action in the Point Beach Overall Integrated Plan (OIP) in Response to the March 12, 2012, Commission Order EA-12-049 [Ref. 3]. Installation of anchorage modifications is scheduled to be completed prior to U2 Outage Fall2015.

8.3 Modification Implementation Schedule Plant modifications will be performed in accordance with the schedule identified in NEI letter dated April 9, 2013 [Ref 13], which states that plant modifications not requiring a planned refueling outage will be completed by December 2016 and modifications requiring a refueling outage will be completed within two planned refueling outages after December 31, 2014.

The modification of the three walls (Masonry Wall 111-2/23, Masonry Wall 111-4N/23 and Work Control Center (WCC) block walls) has not yet proceeded to a level of development to determine if a refueling outage is required to implement the modifications. As such, if a refueling outage is not required to implement these modifications, modification of the three walls will be complete no later than December, 31, 2016. If a refueling outage is required to implement, the modifications will be completed by the end of the second planned refueling outage after December 31, 2014. The second Unit 1 planned refueling outage after December 31, 2016 is U1 R37 currently scheduled to end in the 41h quarter 2017 and the second Unit 2 planned refueling outage after December 31, 2014 is U2R35 scheduled to end 2nd quarter 2017.

Page 27 of 66

8.4 Summary of Regulatory Commitments Item Commitment Date NextEra Energy Point Beach, LLC (NextEra) will Restart of Unit 2 at the complete walkdowns for the inaccessible and late completion of its fall 2015 1 addition components listed in Section 7.1 of this refueling outage.

enclosure.

Within 90 days following NextEra will complete screening and evaluation of the restart of Unit 2 at the inaccessible and late addition components listed in 2 completion of its fall 2015 Section 7.1 of this enclosure.

refueling outage.

NextEra will provide the screening and High Confidence Within 120 days following Low Probability of Failure (HCLPF) results for the restart of Unit 2 at the 3 inaccessible and late addition components listed in completion of its fall 2015 Section 7.1 of this enclosure to the NRC. refueling outage.

December 31, 2016 if the modification(s) do not require an outage on either unit, or

' the latter of the following:

Restart of Unit 1 at the completion of its fall 2017 NextEra will implement modification to Masonry Wall refueling outage if the 4

111-2/23 to raise the HCLPF above the RLGM. modification(s) require a Unit 1 outage, or Restart of Unit 2 at the completion of its spring 2017 refueling outage if the modification(s) require a Unit 2 outage.

December 31, 2016 if the modification(s) do not require an outage on either unit, or the latter of the following:

Restart of Unit 1 at the completion of its fall 2017 NextEra will implement modification to Masonry Wall refueling outage if the 5

111-4N/23 to raise the HCLPF above the RLGM. modification(s) require a Unit 1 outage, or Restart of Unit 2 at the completion of its spring 2017 refueling outage if the modification(s) require a Unit 2 outage.

Page 28 of 66

December 31, 2016 if the modification(s) do not require an outage on either unit, or the latter of the following:

Restart of Unit 1 at the NextEra will implement modification to Work Control completion of its fall2017 Center (WCC) block walls to raise the HCLPF above refueling outage if the 6 the RLGM or relocation of soft targets away from the modification(s) require a path of falling debris or protection of soft targets from Unit 1 outage, falling debris.

or Restart of Unit 2 at the completion of its spring 2017 refueling outage if the modification(s) require a Unit 2 outaQe.

Within 60 days following

~extEra will _submit a letter to NRC confirming 7 completion of all above noted 1mplementat1on of the above noted modification(s).

modifications.

Page 29 of 66

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-lchi Accident," March 12, 2012.

2) Seismic Evaluation Guidance: Augmented Approach for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1 -Seismic. EPRI, Palo Alto, CA:

May 2013. 3002000704.

3) NextEra Energy Point Beach, LLC's Overall Integrated Plan 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), dated February 22, 2013 (ML13053A401)

Updated by:

NextEra Energy Point Beach, LLC's 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) NRC 2014-0052 dated August 28, 2014

4) NRC 2014-0024, "NextEra Energy Point Beach, LLC, Seismic Hazard and Screening 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", March 31, 2014

5) Nuclear Regulatory Commission, NUREG-1407, Procedural and Submittal Guidance for the Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities, June 1991

6) Nuclear Regulatory Commission, Generic Letter No. 88-20 Supplement 4, Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities-10CFR 50.54(f), June 1991

7) A Methodology for Assessment of Nuclear Power Plant Seismic Margin, Rev. 1, August 1991, Electric Power Research Institute, Palo Alto, CA. EPRI NP 6041

8) Methodology for Developing Seismic Fragilities, August 1991, EPRI, Palo Alto, CA.1994, TR-1 03959

9) {Plant Seismic Margin Assessment} (Not Used)

10) Calculation 1400224-C-002, Rev. 0, "HCLPF Evaluations for ESEP".

11) Nuclear Regulatory Commission, NUREG/CR-0098, Development of Criteria for Seismic Review of Selected Nuclear Power Plants, published May 1978

12) 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 Page 30 of 66

13) Nuclear Energy Institute (NEI), A. Pietrangelo, Letter to D. Skeen of the USNRC, "Proposed Path Forward for NTTF Recommendation 2.1: Seismic Reevaluations",

April9, 2013

14) 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) Regarding Seismic Hazard Re-Evaluations for Recommendation 2.1 of the Near-Term Task Force Review of Insights From the Fukushima Dai-lchi Accident," May 9, 2014.

15) Seismic Evaluation Guidance: Screening, Prioritization and Implementation Details (SPID) for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1: Seismic. EPRI, Palo Alto, CA: February2013.1025287.

16) NRC (E Leeds) Letter to NEI (J Pollock), "Electric Power Research Institute Final Draft Report xxxxx, "Seismic Evaluation Guidance: Augmented Approach for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1: Seismic," as an Acceptable Report 1400224-RPT-002 Revision 0 Attachment A Sheet A26 Preliminary for Owner's Review (11/14/14) Page 27 of 28 Alternative to the March 12, 2012, Information Request for Seismic Reevaluations," May 7, 2013

17) NTTF 2.3 Seismic Walkdown Submittals:

Seismic Walkdown Report, rev. 1, In Response to the 50.54(f) Information Request Regarding Fukushima Near-Term Task Force Recommendation 2.3: Seismic for the Point Beach Nuclear Plant Unit 1, NRC Docket No. 50-266, dated May 2014.

Seismic Walkdown Report, rev. 1, In Response to the 50.54(f) Information Request Regarding Fukushima Near-Term Task Force Recommendation 2.3: Seismic for the Point Beach Nuclear Plant Unit 2, NRC Docket No. 50-301, dated May 2014.

18) Point Beach Nuclear Plant Updated Final Safety Analysis Report (UFSAR), 2013

19) DG-C03, Revision 3, "Seismic Design Criteria Guideline"

20) Seismic Qualification Utilities Group (SQUG), "Generic Implementation Procedure (GIP) for Seismic Verification of Nuclear Plant Equipment", Rev. 2a

21) EPRI document 1019200, "Seismic Fragility Applications Guide Update" Page 31 of 66

ATTACHMENT A NEXTERA ENERGY POINT BEACH, LLC UNITS ONE AND TWO EXPEDITED SEISMIC EVALUATION PROCESS(ESEP)REPORT PBNP UNITS 1 AND 2 EXPEDITED SEISMIC EQUIPMENT LIST (ESEL)

Page 32 of 66

TABLE A- UNITS 1 AND 2 EXPEDITED SEISMIC EQUIPMENT LIST TagiD Description Normal Desired State Comments State Turbine driven AFW pump & valves 1P-29 Turbine-driven AFW Standby Operating pump 2P-29 Turbine-driven AFW Standby Operating pump 1AF-4006 SW supply to TDAFW Closed Open pump 2AF-4006 SW supply to TDAFW Closed Open pump 1AF-4000 TDAFW supply to B SG Throttled Throttle/Close Only one SG will be used for decay heat removal and cooldown 1AF-4001 TDAFW supply to A SG Throttled Th rattle/Close Only one SG will be used for decay heat removal and cooldown 2AF-4000 TDAFW supply to B SG Throttled Throttle/Close Only one SG will be used for decay heat removal and cooldown 2AF-4001 TDAFW supply to A SG Throttled Throttle/Close Only one SG will be used for decay heat removal and cooldown 1AF-4002 TDAFW Recirc Closed Open/Close Close when forward flow is adequate 2AF-4002 TDAFW Recirc Closed Open/Close Close when forward flow is adequate 1MS-2018 A main steam isolation Open Close Close to prevent uncontrolled cooldown - Fail closed valve 1MS-2017 B main steam isolation Open Close Close to prevent uncontrolled cooldown - Fail closed valve Page 33 of 66

TagiD Description Normal Desired State Comments State 2MS-2018 A main steam isolation Open Close Close to prevent uncontrolled cooldown - Fail closed valve 2MS-2017 B main steam isolation Open Close Close to prevent uncontrolled cooldown - Fail closed valve 1MS-5958 A SG blowdown Open Close Close to prevent loss of SG isolation inventory - Fail closed valve 1MS-5959 B SG blowdown Open Close Close to prevent loss of SG isolation inventory 2MS-5958 A SG blowdown Open Close Close to prevent loss of SG isolation inventory 2MS-5959 B SG blowdown Open Close Close to prevent loss of SG isolation inventory 1MS-2083 A SG Sample Isolation Open Close Close to prevent loss of SG valve inventory 1MS-2084 B SG Sample Isolation Open Close Close to prevent loss of SG valve inventory 2MS-2083 A SG Sample Isolation Open Close Close to prevent loss of SG valve inventory 2MS-2084 B SG Sample Isolation Open Close Close to prevent loss of SG valve inventory RS-SA-09 U1 Radwaste Steam Open Closed Isolates non-seismic portion of Trip valve steam supply piping RS-SA-10 U2 Radwaste Steam Open Closed Isolates non-seismic portion of Trip valve steam supply piping 1MS-2020 A Steam admission Closed Open Only one SG will be used for valve decay heat removal and cooldown 1MS-2019 B Steam admission Closed Open Only one SG will be used for valve decay heat removal and cooldown 2MS-2020 A Steam admission Closed Open Only one SG will be used for valve decay heat removal and cooldown Page 34 of 66

TagiD Description Normal Desired State Comments State 2MS-2019 B Steam admission Closed Open Only one SG will be used for valve decay heat removal and cooldown 1MS-2082 TDAFW trip valve Open Open Need the ability to re-open following low suction pressure trip 2MS-2082 TDAFW trip valve Open Open Need the ability to re-open following low suction pressure trip SG Relief Valves 1MS-2016 A SG atmospheric Closed Throttled Local manual operation. Only steam dump one SG will be used for decay heat removal and cooldown 1MS-2015 B SG atmospheric Closed Throttled Local manual operation. Only steam dump one SG will be used for decay heat removal and cooldown 2MS-2016 A SG atmospheric Closed Throttled Local manual operation. Only steam dump one SG will be used for decay heat removal and cooldown 2MS-2015 B SG atmospheric Closed Throttled Local manual operation. Only steam dump one SG will be used for decay heat removal and cooldown 1MS-201 0 A SG safety valve Closed Open/Reseat Expect lowest set safety valve to lift until ADV is manually opened 1MS-2005 B SG safety valve Closed Open/Reseat Expect lowest set safety valve to lift until ADV is manually opened 2MS-2010 A SG safety valve Closed Open/Reseat Expect lowest set safety valve to lift until ADV is manually opened 2MS-2005 B SG safety valve Closed Open/Reseat Expect lowest set safety valve to lift until ADV is manually opened Page 35 of 66

TagiD Description Normal Desired State Comments State Storage Tanks T-24A CST Intact > 6 ft of water T-248 CST Intact > 6 ft of water normal level H-13 RWST Intact Lower portion normal level intact 2T-13 RWST Intact Lower portion normal level intact 1T-6A BAST Intact Intact normal Secondary connection for normal level level boric acid, primary is RWST OT-68 BAST Intact Intact normal Secondary connection for normal level level boric acid, primary is RWST 2T-6C BAST Intact Intact normal Secondary connection for normal level level boric acid, primary is RWST 1T-34A A Accumulator Intact Intact normal Required for Mode 5 and 6 normal level level response H-348 8 Accumulator Intact Intact normal Required for Mode 5 and 6 normal level level response 2T-34A A Accumulator Intact Intact normal Required for Mode 5 and 6 normal level level response 2T-348 8 Accumulator Intact Intact normal Required for Mode 5 and 6 normal level level response OT-30 Diesel Driven Fire Pump Intact Intact normal Fuel Oil Day Tank normal level level RCS injection valves 1SI-841A Accumulator isolation Open Close valve 1SI-8418 Accumulator isolation Open Close valve 2SI-841A Accumulator isolation Open Close valve Page 36 of 66

TagiD Description Normal Desired State Comments State 2SI-8418 Accumulator isolation Open Close valve 1CV-1298 Regenerative HX outlet Open Open Need the ability to open the MOV valve if closed 2CV-1298 Regenerative HX outlet Open Open Need the ability to open the MOV valve if closed 1CV-1296 Aux charging Closed Open AOV inside containment lA will not be available Valve lifts with a 248 psid 2CV-1296 Aux charging Closed Open AOV inside containment lA will not be available Valve lifts with a 248 psid 1SI-835A A Accumulator fill valve Closed Open Required for Mode 5 and 6 response 1SI-8358 8 Accumulator fill valve Closed Open Required for Mode 5 and 6 response 2SI-835A A Accumulator fill valve Closed Open Required for Mode 5 and 6 response 2SI-8358 8 Accumulator fill valve Closed Open Required for Mode 5 and 6 response RCS letdown path valves 1RC-570A RV Head Vent Solenoid Closed Open Letdown path for boration prior to asymmetric cooldown 1RC-5708 RV Head Vent Solenoid Closed Open Letdown path for boration prior to asymmetric cooldown 1RC-575A RV!f-1 PZR Vent Closed Open Letdown path for boration prior Header to PRT Solenoid to asymmetric cooldown 1RC-5758 RV!f-1 PZR Gas Vent Closed Open Letdown path for boration prior to Cont. Standpipe to asymmetric cooldown Solenoid 2RC-570A RV Head Vent Solenoid Closed Open Letdown path for boration prior to asymmetric cooldown Page 37 of 66

TagiD Description Normal Desired State Comments State 2RC-570B RV Head Vent Solenoid Closed Open Letdown path for boration prior to asymmetric cooldown 2RC-575A RVff-1 PZR Vent Closed Open Letdown path for boration prior Header to PRT Solenoid to asymmetric cooldown 2RC-575B RVff-1 PZR Gas Vent Closed Open Letdown path for boration prior to Cont. Standpipe to asymmetric cooldown Solenoid RCS boundary Cl valves 1SC-951 Pressurizer Steam Closed Closed Sample 1SC-953 Pressurizer Liquid Closed Closed Sample 1SC-955 RCS Hot Leg Sample Open Closed 2SC-951 Pressurizer Steam Closed Closed Sample 2SC-953 Pressurizer Liquid Closed Closed Sample 2SC-955 RCS Hot Leg Sample Open Closed 1CV-371 Letdown Isolation valve Open Closed 2CV-371 Letdown Isolation valve Open Closed 1CV-313A RCP Seal Return Open Closed Isolation 2CV-313A RCP Seal Return Open Closed Isolation Batteries D-05 SR Station Battery (A, In Service In Service Red)

D-06 SR Station Battery (B, In Service In Service Blue)

D-105 SR Station Battery (A, In Service In Service White)

Page 38 of 66

TagiD Description Normal Desired State Comments State D-106 SR Station Battery (B, In Service In Service Yellow)

D-305 SR Station Battery Standby In Service DC Load Management (Spare)

DC distribution panels D-01 D-05 Battery Bus (A, In Service In Service Red)

D-02 D-06 Battery Bus (B, In Service In Service Blue)

D-03 D-1 05 Battery Bus (A, In Service In Service 1DY-03 White)

D-04 D-106 Battery Bus (B, In Service In Service 2DY-04 Yellow)

D-11 DC Distribution (A, Red) In Service In Service 1DY-01, 1MS-2015, 1AF-4001, 1P-29 Control Pnl. 1C-328, D-16 D-12 DC Distribution (A, Red) In Service In Service 2DY-01, D-22 D-13 DC Distribution (B, In Service In Service 2MS-2020, 2AF-4000, 2P-29 Blue) Control Pnl. 2C-328, D-18 D-14 DC Distribution (B, In Service In Service 1C-20, 2C-20, (U1 and U2 Blue) Head Vent SOVs)

D-16 DC Distribution (A, Red) In Service In Service 1C-03 (1 AF-4001 indication),

1C-04 (U1 PORV), C-01 (1 Sf 835A&B)

D-18 DC Distribution (B, In Service In Service 2C-03 (2AF-4000 indication),

Blue) 2C-04 (U2 PORV), C-01 (2SI 835A&B)

D-21 DC Distribution (B, In Service In Service 1C-04 (U1 PORV)

Blue)

D-22 DC Distribution (A, Red) In Service In Service 2C-04 (U2 PORV)

D-26 DC Distribution (A, Red) In Service In Service 1C-20, 2C-20, (U1 and U2 Head Vent SOVs)

D-27 DC Distribution (B, In Service In Service D-21 Blue)

Page 39 of 66

TagiD Description Normal Desired State Comments State D-63 DC Distribution (A, In Service In Service 1MS-2020, 1AF-4000, 1AF-White) 4006, 1MS-2082, 1AF-4002-S, 1P-29 Control Pnl. 1C-328 D-64 DC Distribution (B, In Service In Service 2MS-2019, 2AF-4001, 2AF-Yellow) 4006, 2MS-2082, 2AF-4002-S, 2P-29 Control Pnl. 2C-328 D-301 Battery Switching Bus Standby In Service DC Load Management D-302 Battery Switching Bus Standby In Service DC Load Management DC MCCs I Switchgear Vital AC distribution panels 1B-03 A Train 480V vital bus Normal Power via 1B-32, 1B-39 power portable supply diesel generator 1B-04 B Train 480V vital bus Normal Power via 1B-42, 1 B-49 power portable supply diesel generator 2B-03 A Train 480V vital bus Normal Power via 2B-32, 2B-39 power portable supply diesel generator 2B-04 B Train 480V vital bus Normal Power via 2B-42, 2B-49 power portable supply diesel generator 1B-32 A Train 480V vital motor Normal Power via For 1B42-3212H Contactor, control center power portable 1SI-841A supply diesel generator 1B-42 B Train 480V vital motor Normal Power via For 1Sl-841 B control center power portable supply diesel generator Page 40 of 66

TagiD Description Normal Desired State Comments State 2B-32 A Train 480V vital motor Normal Power via For 281-841 A control center power portable supply diesel generator 2B-42 B Train 480V vital motor Normal Power via For 2B42-4212B Contactor, control center power portable 281-841 B supply diesel generator 1B-39 480V motor control Normal Power via For D-07 battery charger center power portable supply diesel generator 2B-49 480V motor control Normal Power via ForD-08 battery charger center power portable supply diesel generator 2B-39 480V motor control Normal Power via ForD-1 07 and D-09 battery center power portable charger supply diesel generator 1B-49 480V motor control Normal Power via For D-1 08 and D-09 battery center power portable charger supply diesel generator 1B42-3212H Battery Charger Open Closed ForD-1 09 battery charger Contact or 2B42-4212B Battery Charger Open Closed ForD-1 09 battery charger Contactor 2B4212B- Battery Charger D-1 09 To 2B42 To 2B42 ForD-1 09 battery charger B811M Transfer Switch 1 B42-391 Battery Charge Open Closed For D-07 battery charger Contactor 1B42-491 Battery Charge Open Closed For D-09 battery charger Contactor 1B42-494 Battery Charge Open Closed For D-1 08 battery charger Contactor Page 41 of 66

TagiD Description Normal Desired State Comments State 2B42-391 Battery Charge Open Closed For D-09 battery charger Contactor 2B42-394 Battery Charge Open Closed For D-1 07 battery charger Contact or 2B42-491 Battery Charge Open Closed For D-08 battery charger Contactor Battery chargers D-07 Battery charger Normal Power via Desire flexibility to use any of power portable the battery chargers supply diesel generator D-08 Battery charger Normal Power via Desire flexibility to use any of power portable the battery chargers supply diesel generator D-09 Battery charger Normal Power via Desire flexibility to use any of power portable the battery chargers supply diesel generator D-107 Battery charger Normal Power via Desire flexibility to use any of power portable the battery chargers supply diesel generator D-108 Battery charger Normal Power via Desire flexibility to use any of power portable the battery chargers supply diesel generator D-109 Battery charger Normal Power via Desire flexibility to use any of power portable the battery chargers supply diesel generator Inverters 1DY-01 U1 Instrument Inverter In Service In Service (Red) 2DY-01 U2 Instrument Inverter In Service In Service (Red)

Page 42 of 66

TagiD Description Normal Desired State Comments State 1 DY-03 U1 Instrument Inverter In Service In Service (White) 2DY-04 U2 Instrument Inverter In Service In Service (Yellow)

DY-OA Spare Inverter (Red) Stand By Backs up normal inverters DY-OC Spare Inverter (White) Stand By Backs up normal inverters DY-OD Spare Inverter (Yellow) Stand By Backs up normal inverters DY-13 Alternate Shutdown In Service Re-align Inverter Instrument Racks 1Y-01 Instrument Distribution In Service In Service Pnl (Red) 1Y-101 Instrument Distribution In Service In Service Pnl (Red) 2Y-01 Instrument Distribution In Service In Service Pnl (Red) 2Y-101 Instrument Distribution In Service In Service Pnl (Red) 1Y-03 Instrument Distribution In Service In Service Pnl (White) 1Y-103 Instrument Distribution In Service In Service Pnl (White) 2Y-04 Instrument Distribution In Service In Service Pnl (Yellow) 2Y-104 Instrument Distribution In Service In Service Pnl (Yellow)

Red Spec 200 In Service In Service 1C-170 Instrument Cabinet White Spec 200 In Service In Service 1C-171B Instrument Cabinet Red Spec 200 In Service In Service 2C-170 Instrument Cabinet Page 43 of 66

TagiD Description Normal Desired State Comments State Yellow Spec 200 In Service In Service 2C-173B Instrument Cabinet Control Channel I Panel In Service lri Service 1C-112 (Red)

Control Channel II In Service In Service 1C-114 Panel (White)

Control Channel I Panel In Service In Service 2C-112 (Red)

Control Channel IV In Service In Service 2C-117 Panel (Yellow)

White Spec 200 In Service In Service 1C-171A Instrument Cabinet Yellow Spec 200 In Service In Service 2C-173A Instrument Cabinet Control Channel I Panel In Service In Service 1C-111 (Red)

Control Channel II In Service In Service 1C-113 Panel (White)

Control Channel I Panel In Service In Service 2C-111 (Red)

Sl And Auxiliary Coolant In Service In Service 1C-1 09 Sys Panel 1C-129 RCS And SIS Panel In Service In Service Sl And Auxiliary Coolant In Service In Service 2C-109 Sys Panel De- In Service 1C-132 1N-32 Source Range energized De- In Service 1C-133 1N-31 Source Range energized De- In Service 2C-133 2N-31 Source Range energized 1C-205 Normal Re-align Page 44 of 66

TagiD Description Normal Desired State Comments State 2C-205 Normal Re-align C-207 Normal Re-align Transmitters 1HX-1A SG WR Level In Service In Service 1LT-460A Transmitter 1HX-1A SG WR Level In Service In Service 1LT-4608 Transmitter 1HX-18 SG WR Level In Service In Service 1LT-470A Transmitter 2HX-1A SG WR Level In Service In Service 2LT-460A Transmitter 2HX-1 8 SG WR Level In Service In Service 2LT-470A Transmitter 2HX-1 8 SG WR Level In Service In Service 2LT-4708 Transmitter HX-1A SG Steam In Service In Service 1 PT-468 Pressure Transmitter HX-1A SG Steam In Service In Service 1PT-469 Pressure Transmitter HX-1 8 SG Steam In Service In Service 1PT-483 Pressure Transmitter HX-1A SG Steam In Service In Service 2PT-468 Pressure Transmitter HX-18 SG Steam In Service In Service 2PT-479 Pressure Transmitter HX-18 SG Steam In Service In Service 2PT-483 Pressure Transmitter Aux Feedwater to 1HX- In Service In Service 1FT-4036 1A SG 1P-29 AFP Discharge In Service In Service Flow 1FT-4002 Page 45 of 66

TagiD Description Normal Desired State Comments State Aux Feedwater To 2HX- In Service In Service 2FT-4037 18 SG 2P-29 AFP Discharge In Service In Service 2FT-4002 Flow T-24A CST Level In Service In Service LT-4038 Transmitter T -248 CST Level In Service In Service LT-4041 Transmitter U1 RC Loop A In Service In Service Intermediate Leg WR 1PT-420A Press U1 RC Loop A Hot Leg In Service In Service 1PT-420C WR Pressure U2 RC Loop 8 In Service In Service Intermediate Leg WR 2PT-4208 Press U2 RC Loop A Hot Leg In Service In Service 2PT-420C WR Pressure H-1 Pzr NR Level In Service In Service 1LT-426 Transmitter H-1 Pzr NR Level In Service In Service 1LT-427 Transmitter 2T-1 Pzr NR Level In Service In Service 2LT-426 Transmitter R-1 RV Wide Range In Service In Service 1LT-494 Level R-1 RV Narrow Range In Service In Service 1LT-496 Level R-1 RV Wide Range In Service In Service 2LT-495 Level R-1 RV Narrow Range In Service In Service Level 2LT-497 Page 46 of 66

TagiD Description Normal Desired State Comments State lncore Thermocouple at In Service In Service 1TE-00037 K-11 lncore Thermocouple at In Service In Service 1TE-00019 B-5 lncore Thermocouple at In Service In Service 2TE-00023 D-7 lncore Thermocouple at In Service In Service 2TE-00038 L-10 RC Loop A Cold Leg In Service In Service 1TE-450A Temperature RTD RC Loop B Cold Leg In Service In Service 1TE-451C Temperature RTD RC Loop A Hot Leg In Service In Service 1TE-450D Temperature RTD RC Loop B Hot Leg In Service In Service 1TE-451B Temperature RTD RC Loop A Cold Leg In Service In Service 2TE-450C Temperature RTD RC Loop B Cold Leg In Service In Service 2TE-451A Temperature RTD RC Loop B Cold Leg In Service In Service 2TE-451C Temperature RTD RC Loop A Hot Leg In Service In Service 2TE-450B Temperature RTD RC Loop B Hot Leg In Service In Service 2TE-451B Temperature RTD RC Loop B Hot Leg In Service In Service 2TE-451D Temperature RTD T-34A Sl Accumulator In Service In Service 1PT-940 Pressure T-34A Sl Accumulator In Service In Service 1 PT-941 Pressure Page 47 of 66

TagiD Description Normal Desired State Comments State T-348 Sl Accumulator In Service In Service 1PT-936 Pressure T-348 Sl Accumulator In Service In Service 1PT-937 Pressure T-34A Sl Accumulator In Service In Service 2PT-940 Pressure T-348 Sl Accumulator In Service In Service 2PT-936 Pressure De- In Service 1N-31 U1 Nl Source Range energized De- In Service 1N-32 U1 Nl Source Range energized 1N-40 U1 Nl Wide Range In Service Re-align De- In Service 2N-31 U2 Ni Source Range energized 2N-40 U2 Nl Wide Range In Service Re-align U1 Containment WR In Service In Service 1PT-968 Pressure U2 Containment WR In Service In Service 2PT-968 Pressure U2 Containment WR In Service In Service 2PT-969 Pressure El 66' U 1 Containment In Service In Service HE-3292 Temperature El 66' U2 Containment In Service In Service 2TE-3293 Temperature Fire Protection P-358 Diesel Driven Fire Pump Standby Operating Replacement and seismic upgrade reference EC259770 P-358-E Diesel Driven Fire Pump Standby Operating Replacement and seismic Engine upgrade reference EC259770 Page 48 of 66

TagiD Description Normal Desired State Comments State FP-448 Fire water header Open Close New valve reference isolation EC259770 FP-536 Fire water to SW cross Close Open New valve reference connection EC259770 FP-3715 Diesel Fire Pump Relief Close Open as Replacement and seismic Valve required upgrade reference EC259770 T-30 Diesel Fire Pump Fuel Intact Intact normal Replacement and seismic Tank normal level level upgrade reference EC259770 D-600 Diesel Driven Fire Pump In Service In Service Replacement and seismic Battery Rack upgrade reference EC259770 C-601 Diesel Driven Fire Pump In Service In Service Replacement and seismic Control Panel upgrade reference EC259770 Service Water 1SW-2880 Unit 1 Turbine Hall Open Close MOV will be hand cranked Supply closed 2SW-2880 Unit 2 Turbine Hall Open Close MOV will be hand cranked Supply closed SW-4478 Water Treatment Open Close MOV will be hand cranked closed SW-2817 Water Treatment Open Close MOV will be hand cranked closed SW-4479 Service and Aux. Open Close MOV will be hand cranked Building closed SW-2816

Service and Aux. Open Close MOV will be hand cranked Building closed Relay/Contactor 1-62/04044 1MS-2082, 1P-29 Functional Functional Relay would seal in if contacts Trip/Throttle Valve bounce close. This would result in tripping P-29 Trip/Throttle valve 1MS-2082 closed. Valve could then be manually reset from the CR.

(MS -2082 would trip closed on any TDR momentary contact closure and the TDR Page 49 of 66

TagiD Description Normal Desired State Comments State would not have to seals in) 1SMS-2019 1MS-2019, 1P-29 Functional Functional If contactor contacts bounced Steam Supply MOV close when the valve is full open, the contactor would seal in and fully close the valve.

The valve would then automatically re-open.

1SMS-2020 1MS-2020, 1 P-29 Functional Functional If contactor contacts bounced Steam Supply MOV close when the valve is full open, the contactor would seal in and fully close the valve.

The valve would then automatically re-open.

1SAF-4006 1AF-4006, 1P29 SW Functional Functional If contactor contacts bounced Suction MOV close when the valve is full open, the contactor would seal in and fully close the valve.

The valve would stay open until manually closed using the control switch. This may result in CST inventory wasted via the SW overboard path.

1-62/4044C 1AF-4006, 1P29 SW Functional Functional This relay does not seal in, but Suction MOV a momentary contact closure would result in the open contactor sealing in causing the valve to fully open and remain open until manually closed using the control switch. This may result in CST inventory wasted via the SW overboard path.

1-4077LLL- 1AF-4006, 1P29 SW Functional Functional This relay does not seal in, but X Suction MOV a momentary contact closure would result in the open 2AF-4006, 2P29 SW contactor sealing in causing Suction MOV the valve to fully open and remain open until manually closed using the control switch. This may result in CST inventory wasted via the Page 50 of 66

TagiD Description Normal Desired State Comments State SW overboard path.

2-4078LLL- 1AF-4006, 1P29 SW Functional Functional This relay does not seal in, but X Suction MOV a momentary contact closure would result in the open 2AF-4006, 2P29 SW contactor sealing in causing Suction MOV the valve to fully open and remain open until manually closed using the control switch. This may result in CST inventory wasted via the SW overboard path.

1SAF-4067 1AF-4067, 2P53 SW Functional Functional If contactor contacts bounced Suction MOV close when the valve is full open, the contactor would seal in and tully close the valve.

The valve would stay open until manually closed using the control switch. This may result in CST inventory wasted via the SW overboard path.

2-62/04044 2MS-2082, 2P-29 Functional Functional Relay would seal in if contacts Trip/Throttle Valve bounce close. This would result in tripping P-29 Trip/Throttle valve 2MS-2082 closed. Valve could then be manually reset from the CR.

(MS -2082 would trip closed on any TOR momentary contact closure and the TOR would not have to seals in) 2SMS-2019 2MS-2019, 2P-29 Functional Functional If contactor contacts bounced Steam Supply MOV close when the valve is full open, the contactor would seal in and tully close the valve.

The valve would then automatically re-open.

2SMS-2020 2MS-2020, 2P-29 Functional Functional If contactor contacts bounced Steam Supply MOV close when the valve is full open, the contactor would seal in and tully close the valve.

The valve would then Page 51 of 66

TagiD Description Normal Desired State Comments State automatically re-open.

2SAF-4006 2AF-4006, 2P29 SW Functional Functional This relay does not seal in, but Suction MOV a momentary contact closure would result in the open contactor sealing in causing the valve to fully open and remain open until manually closed using the control switch. This may result in CST inventory wasted via the SW overboard path.

2-62/4044C 2AF-4006, 2P29 SW Functional Functional This relay does not seal in, but Suction MOV a momentary contact closure would result in the open contactor sealing in causing the valve to fully open and remain open until manually closed using the control switch. This may result in CST inventory wasted via the SW overboard path.

2SAF-4067 2AF-4067, 2P53 SW Functional Functional If contactor contacts bounced Suction MOV close when the valve is full open, the contactor would seal in and fully close the valve.

The valve would stay open until manually closed using the control switch. This may result in CST inventory wasted via the SW overboard path.

Rack/Panel 1C-03 Main Control Board In Service In Service Contains Various Circuit Breakers. Some controls and/or indicators used for FLEX.

1C-04 Main Control Board In Service In Service Contains Various Circuit Breakers. Some controls and/or indicators used for FLEX.

Page 52 of 66

TagiD Description Normal Desired State Comments State C-01 Main Control Board In Service In Service Contains Various Circuit Breakers. Some controls and/or indicators used for FLEX.

C-02 Main Control Board In Service In Service Contains Various Circuit Breakers. Some controls and/or indicators used for FLEX.

2C-03 Main Control Board In Service In Service Contains Various Circuit Breakers. Some controls and/or indicators used for FLEX.

2C-04 Main Control Board In Service In Service Contains Various Circuit Breakers. Some controls and/or indicators used for FLEX.

1C-20 ASIP Panel In Service In Service Contains Various Circuit Breakers. Some controls and/or indicators used for FLEX.

2C-20 ASIP Panel In Service In Service Contains Various Circuit Breakers. Some controls and/or indicators used for FLEX.

1C-197 P-29 AFP SUCTION In Service In Service PRESSURE CONTROL Contains seal in relay or PANEL contactor.

1SMS-2019 Motor Starter In Service In Service Contains seal in relay or contactor.

1SMS-2020 Motor Starter In Service In Service Contains seal in relay or contactor.

1SAF-4006 Motor Starter In Service In Service Contains seal in relay or contactor.

1SAF-4067 Motor Starter In Service In Service Contains seal in relay or contactor.

Page 53 of 66

TagiD Description Normal Desired State Comments State 2C-197 P-29 AFP SUCTION In Service In Service PRESSURE CONTROL Contains seal in relay or PANEL contactor.

2SMS-2019 Motor Starter In Service In Service Contains seal in relay or contactor.

2SMS-2020 Motor Starter In Service In Service Contains seal in relay or contactor.

2SAF-4006 Motor Starter In Service In Service Contains seal in relay or contactor.

2SAF-4067 Motor Starter In Service In Service Contains seal in relay or contactor.

Page 54 of 66

ATTACHMENT B NEXTERA ENERGY POINT BEACH, LLC UNITS ONE AND TWO EXPEDITED SEISMIC EVALUATION PROCESS(ESEP)REPORT ESEP HCLPF VALUES AND FAILURE MODES TABULATION Page 55 of 66

Note: Refer to section 6.6 for discussion regarding this table of ESEL HCLPF values.

Failure Equipment ID HCLPF Additional Discussion Mode Anchorage screened by large available margin in existing ODY-13 Screened ;:: 0.229 design basis calculation.

Equipment not yet installed. Evaluation required after OT-30 T8D T8D installation.

OT-68 Anchorage ;::o.229 Functional 1-4077LLL-X ;:: 0.229 Failure Functional 1-62/4044C ;::0.229 Failure Functional 1-62-4044 ;:: 0.229 Failure Operator offset acceptable based on a comparison to 1AF-4000 Screened ;:: 0.229 existinQ SQUG documentation & analysis.

Operator offset acceptable based on a comparison to 1AF-4001 Screened ;:: 0.229 existing SQUG documentation & analysis.

1AF-4002 Screened ;:: 0.229 1AF-4006 Screened ;:: 0.229 Anchorage screened by large available margin in existing 18-03 Screened ;:: 0.229 desiQn basis calculation.

Anchorage screened by large available margin in existing 18-04 Screened ;:: 0.229 design basis calculation.

Anchorage screened by large available margin in existing 18-32 Screened ;:: 0.229 desiQn basis calculation.

18-39 Screened ;:: 0.229 Anchorage screened by engineering judgment.

Systems Adjacent block wall controls the equipment HCLPF 18391 ;:: 0.229 Interaction capacity.

Anchorage screened by large available margin in existing 18-42 Screened ;:: 0.229 design basis calculation.

1842-3212H Screened ;::o.229 Anchorage screened by engineering judgment.

18-49 Screened ;:: 0.229 Anchorage screened by engineering judgment.

18491 Screened ;:: 0.229 Anchorage screened by engineering judgment.

18494 Screened ;:: 0.229 Anchorage screened by engineering judgment.

Adjacent block walls (Wall # 111-4N/23 & 111-2/23)

Systems 1C-03 0.139 control the equipment HCLPF capacity. Modification of Interaction block wall is required.

Adjacent block walls (Wall # 111-4N/23 & 111-2/23)

Systems 1C-04 0.139 control the equipment HCLPF capacity. Modification of Interaction block wall is required.

Systems Adjacent block wall controls the equipment HCLPF 1C-109 ;:: 0.229 Interaction capacity.

Systems Adjacent block wall controls the equipment HCLPF 1C-111 ;:: 0.229 Interaction capacity.

Systems Adjacent block wall controls the equipment HCLPF 1C-112 ;::0.229 Interaction capacity.

Systems Adjacent block wall controls the equipment HCLPF 1 C-113 ;:: 0.229 Interaction capacity.

Systems Adjacent block wall controls the equipment HCLPF 1C-114 ;:: 0.229 Interaction capacity.

Page 56 of 66

Failure Equipment ID HCLPF Additional Discussion Mode Anchorage screened by large available margin in existing 1C-129 Screened ~ 0.229 design basis calculation.

Anchorage screened by large available margin in existing 1C-132 Screened ~ 0.229 design basis calculation.

Anchorage screened by large available margin in existing 1C-133 Screened ~ 0.229 design basis calculation.

Anchorage screened by large available margin in existing 1C-170 Screened ~ 0.229 design basis calculation.

Anchorage screened by large available margin in existing 1C-171A Screened ~ 0.229 design basis calculation.

Anchorage screened by large available margin in existing 1C-171B Screened ~ 0.229 design basis calculation.

Anchorage screened by large available margin in existing 1C-197 Screened ~ 0.229 design basis calculation.

Anchorage screened by large available margin in existing 1C-20 Screened ~ 0.229 design basis calculation.

1C-205 Screened ~ 0.229 Anchorage screened by engineering judgment.

1CV-1296 Screened ~ 0.229 Operator weight & offset do not meet the criteria of NP-Equipment 1CV-1298 ~ 0.229 6041. Valve yoke stresses have been shown to be Capacity acceptable for the RLGM.

1CV-313A Screened ~ 0.229 1CV-371 Screened ~ 0.229 Anchorage screened by large available margin in existing 1DY-01 Screened ~ 0.229 design basis calculation.

1DY-03 Anchorage ~ 0.229 1 FT-4002 Screened ~ 0.229 Anchorage screened by engineering judgment.

1 FT-4036 Screened ~ 0.229 Anchorage screened by engineering judgment.

1LT-426 Screened ~ 0.229 Anchorage screened by engineering judgment.

1LT-427 Screened ~ 0.229 Anchorage screened by engineering judgment.

1LT-460A Screened ~ 0.229 Anchorage screened by engineering judgment.

1LT-460B Screened ~ 0.229 Anchorage screened by engineering judgment.

1LT-470A Screened ~ 0.229 Anchorage screened by engineering judgment.

1LT-494 Screened ~ 0.229 Anchorage screened by engineering judgment.

1LT-496 Screened ~ 0.229 Anchorage screened by engineering judgment.

1MS-2005 Screened ~ 0.229 1MS-2010 Screened ~ 0.229 1MS-2015 Screened ~ 0.229 1MS-2016 Screened ~ 0.229 1MS-2017 Screened ~ 0.229 1MS-2018 Screened ~ 0.229 Operator offset acceptable based on a comparison to 1MS-2019 Screened ~ 0.229 existing SQUG documentation & analysis.

Operator offset acceptable based on a comparison to 1MS-2020 Screened ~ 0.229 existing_ SQUG documentation & analy_sis.

Operator offset acceptable based on a comparison to 1MS-2082 Screened ~ 0.229 existing SQUG documentation & analysis.

Page 57 of 66

Failure Equipment ID HCLPF Additional Discussion Mode 1MS-2083 Screened ;:: 0.229 1MS-2084 Screened ;:: 0.229 1MS-5958 Screened ;:: 0.229 1MS-5959 Screened ;:: 0.229 1N-31 is a drawer in cabinet 1C-133. See 1C-133 for 1N-31 Screened ;:: 0.229 acceptability of anchoraae for cabinet 1C-133.

1N-32 is a drawer in cabinet 1C-132. See 1C-132 for 1N-32 Screened ;:: 0.229 acceptability of anchoraQe for cabinet 1C-132.

This component is located in a high radiation area and is not easily accessible. For ALARA purposes, this component was not walked down. The component itself is 1N-40 Screened ;:: 0.229 inherently rugged and is judged not to be a concern.

Anchorage screened by large available margin in existing desiQn basis calculation.

Anchorage screened by large available margin in existing 1P-29 Screened ;:: 0.229 design basis calculation.

1PT-420A Screened ;:: 0.229 Anchorage screened by engineering judgment.

1PT-420C Screened 2:0.229 Anchorage screened by engineering judgment.

Anchorage screened by large available margin in existing 1PT-468 Screened ;:: 0.229 desiQn basis calculation.

Anchorage screened by large available margin in existing 1PT-469 Screened ;:: 0.229 design basis calculation.

Anchorage screened by large available margin in existing 1 PT-483 Screened ;:: 0.229 desiQn basis calculation.

Anchorage screened by large available margin in existing 1PT-936 Screened ;:: 0.229 design basis calculation.

Anchorage screened by large available margin in existing 1PT-937 Screened 2:0.229 desiQn basis calculation.

Anchorage screened by large available margin in existing 1 PT-940 Screened ;:: 0.229 desiQn basis calculation.

Anchorage screened by large available margin in existing 1 PT-941 Screened 2:0.229 desiQn basis calculation.

1PT-968 Screened ;:: 0.229 Anchorage screened by engineering judgment.

1RC-570A Screened ;:: 0.229 1 RC-5708 Screened ;:: 0.229 1RC-575A Screened ;:: 0.229 1RC-575B Screened ;:: 0.229 Systems Adjacent block wall controls the equipment HCLPF 1SAF-4006 ;:: 0.229 Interaction capacity.

1SAF-4006 Functional

0.229

_(Relay) Failure Systems Adjacent block wall controls the equipment HCLPF 1SAF-4067 ;:: 0.229 Interaction capacity.

1SAF-4067 Functional 2:0.229 (Relay) Failure 1SC-951 Screened ;:: 0.229 Page 58 of 66

Failure Equipment ID HCLPF Additional Discussion Mode Walkdown identified missing bolts in the operator support anchorage. AR 01998370 was written to address the 1SC-953 Screened ;:: 0.229 issue. One bolt has been installed and the AR has been closed. One bolt is sufficient to prevent movement of the operator relative to the tube.

1SC-955 Screened 2':0.229 1SI-835A Screened ;:: 0.229 1SI-835B Screened ;:: 0.229 1SI-841A Screened ;:: 0.229 1SI-841B Screened ;:: 0.229 1SMS-2019 Screened ;:: 0.229 Anchorage screened by engineering judgment.

1SMS-2019 Functional

0.229 (Relay) Failure Systems Adjacent block wall controls the equipment HCLPF 1SMS-2020 ;

: 0.229 Interaction capacity.

1SMS-2020 Functional

0.229 (Relay). Failure 1SW-2880 Screened ;

: 0.229 The evaluation of the RWSTs uses alternate criteria in determining the RLGM for the tank as discussed in 1T-13 Anchorage 2':0.229 Section 5.2. The HCLPF listed is for comparison to the other equipment HCLPF values which used the typical RLGM (1.91 x SSE).

1T-34A Anchorage ;:: 0.229 1T-34B Anchorage ;:: 0.229 1T-6A Anchorage 2':0.229 For ALARA purposes, this item was not walked down.

Based upon the inherent seismic ruggedness of thermocouples and by comparison to a number of evaluations for similar thermocouples (see 1TE-450A, 1TE-00019 Screened ;:: 0.229 1TE-450D, 1TE-451 B, 1TE-451 C, 2TE-450B, 2TE-450C, 2TE-451 A, 2TE-451 B, 2TE-451 C, and 2TE-451 D), the adequacy of this equipment is judged to be acceptable.

No further action is required.

For ALARA purposes, this item was not walked down.

Based upon the inherent seismic ruggedness of thermocouples and by comparison to a number of evaluations for similar thermocouples (see 1TE-450A, 1TE-00037 Screened 2':0.229 1TE-450D, 1TE-451 B, 1TE-451 C, 2TE-450B, 2TE-450C, 2TE-451 A, 2TE-451 B, 2TE-451 C, and 2TE-451 D), the adequacy of this equipment is judged to be acceptable.

No further action is required.

1TE-3292 Screened ;:: 0.229 Anchorage screened by engineering judgment.

1TE-450A Screened ;:: 0.229 1TE-450D Screened ;:: 0.229 1TE-451 B Screened ;:: 0.229 1TE-451 C Screened ;:: 0.229 Systems Adjacent block wall controls the equipment HCLPF 1Y-01 2':0.229 Interaction capacity.

Page 59 of 66

Failure Equipment ID HCLPF Additional Discussion Mode Systems Adjacent block wall controls the equipment HCLPF 1Y-03 ;::: 0.229 Interaction capacity.

1Y-101 Screened ;::: 0.229 Anchorage screened by engineering judgment.

1Y-103 Screened ;::: 0.229 Anchorage screened by engineering judgment.

Functional 2-4078LLL-X <::0.229 Failure Functional 2-62/4044C <::0.229 Failure Functional 2-62-4044 ;::: 0.229 Failure Operator offset acceptable based on a comparison to 2AF-4000 Screened ;::: 0.229 existinQ SQUG documentation & analysis.

Operator offset acceptable based on a comparison to 2AF-4001 Screened ;::: 0.229 existing SQUG documentation & analysis.

2AF-4002 Screened ;::: 0.229 2AF-4006 Screened ;::: 0.229 Anchorage screened by large available margin in existing 28-03 Screened ;::: 0.229 design basis calculation.

Anchorage screened by large available margin in existing 28-04 Screened ;::: 0.229 design basis calculation.

Anchorage screened by large available margin in existing 28-32 Screened ;::: 0.229 design basis calculation.

28-39 Screened ;::: 0.229 Anchorage screened by engineering judgment.

Systems Adjacent block wall controls the equipment HCLPF 28391 ;::: 0.229 Interaction capacity.

28394 Screened ;::: 0.229 Anchorage screened by engineering judgment.

Anchorage screened by large available margin in existing 28-42 Screened ;::: 0.229 design basis calculation.

2842128-Anchorage ;::: 0.229 8811M 2842-42128 Screened ;::: 0.229 Anchorage screened by engineering judgment.

28-49 Screened ;::: 0.229 Anchorage screened by engineering judgment.

28491 Screened ;::: 0.229 Anchorage screened by engineering judgment.

Adjacent block walls (Wall # 111-4N/23 & 111-2/23)

Systems 2C-03 0.139 control the equipment HCLPF capacity. Modification of Interaction block wall is reguired.

Adjacent block walls (Wall# 111-4N/23 & 111-2/23)

Systems 2C-04 0.139 control the equipment HCLPF capacity. Modification of Interaction block wall is required.

The anchorage has a lower HCLPF value than the 2C-109 Anchorage 0.229 adjacent block wall. As such, adjacent block wall does not control the equipment HCLPF capacity.

The anchorage has a lower HCLPF value than the 2C-111 Anchorage 0.229 adjacent block wall. As such, adjacent block wall does not control the equipment HCLPF capacity.

The anchorage has a lower HCLPF value than the 2C-112 Anchorage 0.229 adjacent block wall. As such, adjacent block wall does not control the equipment HCLPF capacity.

2C-117 Anchorage ;::: 0.229 2C-133 Anchorage <':0.229 Page 60 of 66

Failure Equipment ID HCLPF Additional Discussion Mode Anchorage screened by large available margin in existing 2C-170 Screened ;::: 0.229 design basis calculation.

Anchorage screened by large available margin in existing 2C-173A Screened ;::: 0.229 design basis calculation.

Anchorage screened by large available margin in existing 2C-173B Screened ;::: 0.229 design basis calculation.

Anchorage screened by large available margin in existing 2C-197 Screened ;::: 0.229 design basis calculation.

Anchorage screened by large available margin in existing 2C-20 Screened ;::: 0.229 design basis calculation.

2C-205 Screened ;::: 0.229 Anchorage screened by engineering judgment.

2CV-1296 Screened ;::: 0.229 Operator weight & offset do not meet the criteria of NP-Equipment 2CV-1298 ;::: 0.229 6041. Valve yoke stresses have been shown to be Capacity acceptable for the RLGM.

2CV-313A Screened 2:0.229 2CV-371 Screened ;::: 0.229 Anchorage screened by large available margin in existing 2DY-01 Screened 2:0.229 design basis calculation.

2DY-04 Anchorage 2:0.229 2FT-4002 Screened ;::: 0.229 Anchorage screened by engineering judgment.

Anchorage screened by large available margin in existing 2FT-4037 Screened ;::: 0.229 design basis calculation.

2LT-426 Screened 2:0.229 Anchorage screened by engineering judgment.

2LT-460A Screened ;::: 0.229 Anchorage screened by engineering judgment.

2LT-470A Screened ;::: 0.229 Anchorage screened by engineering judgment.

2LT-470B Screened ;::: 0.229 Anchorage screened by engineering judgment.

2LT-495 Screened ;::: 0.229 Anchorage screened by engineering judgment.

2LT-497 Screened ;::: 0.229 Anchorage screened by engineering judgment.

2MS-2005 Screened ;::: 0.229 2MS-2010 Screened ;::: 0.229 2MS-2015 Screened ;::: 0.229 2MS-2016 Screened ;::: 0.229 2MS-2017 Screened ;::: 0.229 2MS-2018 Screened 2:0.229 Operator offset acceptable based on a comparison to 2MS-2019 Screened ;::: 0.229 existing SQUG documentation & analysis.

Operator offset acceptable based on a comparison to 2MS-2020 Screened ;::: 0.229 existing SQUG documentation & analysis.

Operator offset acceptable based on a comparison to 2MS-2082 Screened ;::: 0.229 existing SQUG documentation & analysis.

2MS-2083 Screened ;::: 0.229 2MS-2084 Screened 2:0.229 2MS-5958 Screened ;::: 0.229 2MS-5959 Screened 2:0.229 2N-31 Anchorage ;::: 0.229 Page 61 of 66

Failure Equipment ID HCLPF Additional Discussion Mode This component is located in a high radiation area and is not easily accessible. For ALARA purposes, this component was not walked down. The component itself is 2N-40 Screened ;::: 0.229 inherently rugged and is judged not to be a concern.

Anchorage screened by large available margin in existing design basis calculation.

Anchorage screened by large available margin in existing 2P-29 Screened ;::: 0.229 design basis calculation.

2PT-420B Screened ;::: 0.229 Anchorage screened by engineering judgment.

2PT-420C Screened ;::: 0.229 Anchorage screened by engineering judgment.

Anchorage screened by large available margin in existing 2PT-468 Screened ;::: 0.229 design basis calculation.

Anchorage screened by large available margin in existing 2PT-479 Screened ;::: 0.229 design basis calculation.

Anchorage screened by large available margin in existing 2PT-483 Screened ;::: 0.229 design basis calculation.

Anchorage screened by large available margin in existing 2PT-936 Screened ;::: 0.229 design basis calculation.

Anchorage screened by large available margin in existing 2PT-940 Screened ;::: 0.229 design basis calculation.

2PT-968 Screened ;::: 0.229 Anchorage screened by engineering judgment.

2PT-969 Screened ;::: 0.229 Anchorage screened by engineering judgment.

Equipment added to ESEL after Unit 2 Containment 2RC-570A TBD TBD walkdown. Equipment is scheduled to be walked down during next Unit 2 outage.

Equipment added to ESEL after Unit 2 Containment 2RC-5708 TBD TBD walkdown. Equipment is scheduled to be walked down during next Unit 2 outage.

Equipment added to ESEL after Unit 2 Containment 2RC-575A TBD TBD walkdown. Equipment is scheduled to be walked down duril}g next Unit 2 outage.

Equipment added to ESEL after Unit 2 Containment 2RC-575B TBD TBD walkdown. Equipment is scheduled to be walked down during next Unit 2 outage.

Systems Adjacent block wall controls the equipment HCLPF 2SAF-4006 ;::: 0.229 Interaction ca[>_acity.

2SAF-4006 Functional

0.229 (Relay) Failure Systems Adjacent block wall controls the equipment HCLPF 2SAF-4067 ;

:: 0.229 Interaction ca[>_acity_.

2SAF-4067 Functional

0.229 (Relay) Failure 2SC-951 Screened ;

:: 0.229 Walkdown identified missing bolts in the operator support anchorage. AR 01955412 was written to address the 2SC-953 Screened ;::: 0.229 issue. The bolts have been installed and the AR has been closed.

2SC-955 Screened ;::: 0.229 2SI-835A Screened ;::: 0.229 2SI-8358 Screened ;::: 0.229 Page 62 of 66

Failure Equipment ID HCLPF Additional Discussion Mode 2SI-841A Screened ;:: 0.229 2SI-841B Screened ;:: 0.229 Systems Adjacent block wall controls the equipment HCLPF 2SMS-2019 ;:: 0.229 Interaction capacity.

2SMS-2019 Functional

0.229 (Relay) Failure 2SMS-2020 Screened ;

: 0.229 Anchorage screened by engineering judgment.

2SMS-2020 Functional

0.229 (Relay) Failure 2SW-2880 Screened ;

:0.229 The evaluation of the RWSTs uses alternate criteria in determining the RLGM for the tank as discussed in 2T-13 Anchorage ;::0.229 Section 5.2. The HCLPF listed is for comparison to the other equipment HCLPF values which used the typical RLGM (1.91 x SSE).

2T-34A Anchorage ;:: 0.229 2T-34B Anchorage ;:: 0.229 2T-6C Anchorage ;::0.229 For ALARA purposes, this item was not walked down.

Based upon the inherent seismic ruggedness of thermocouples and by comparison to a number of evaluations for similar thermocouples (see 1TE-450A, 2TE-00023 Screened ;:: 0.229 HE-450D, HE-451 B, HE-451 C, 2TE-450B, 2TE-450C, 2TE-451 A, 2TE-451 B, 2TE-451 C, and 2TE-451 D), the adequacy of this equipment is judged to be acceptable.

No further action is required.

For ALARA purposes, this item was not walked down.

Based upon the inherent seismic ruggedness of thermocouples and by comparison to a number of evaluations for similar thermocouples (see HE-450A, 2TE-00038 Screened ;::0.229 HE-450D, HE-4518, HE-451 C, 2TE-450B, 2TE-450C, 2TE-451 A, 2TE-451 B, 2TE-451 C, and 2TE-451 D), the adequacy of this equipment is judged to be acceptable.

No further action is re_guired.

2TE-3293 Screened ;:: 0.229 Anchorage screened by engineering judgment.

2TE-450B Screened ;:: 0.229 2TE-450C Screened ;:: 0.229 Item is inaccessible because it is in a LHRA. Equipment judged acceptable upon review of documentation and 2TE-451A Screened ;:: 0.229 photo provided in A-46 SEWS SQ-001 052 and by comparison to similar equipment (2TE-450C).

2TE-451B Screened ;:: 0.229 Item is inaccessible because it is in a LHRA. Equipment judged acceptable upon review of documentation and 2TE-451C Screened ;:: 0.229 photo provided in A-46 SEWS SQ-001 056 and by comparison to similar equipment (2TE-450C).

2TE-451D Screened ;:: 0.229 Systems Adjacent block wall (Wall# 111-2/23) controls the 2Y-01 0.139 Interaction equipment HCLPF capacity. Modification is required.

Page 63 of 66

Failure Equipment ID HCLPF Additional Discussion Mode Systems Adjacent block wall (Wall # 111-2/23) controls the 2Y-04 0.139 Interaction equipment HCLPF capacity. Modification is required.

2Y-101 Screened 2: 0.229 Anchorage screened by engineering judgment.

2Y-104 Screened 2: 0.229 Anchorage screened by engineering judgment.

Adjacent block walls (Wall # 111-4N/23 & 111-2/23)

Systems C-01 0.139 control the equipment HCLPF capacity. Modification of Interaction block wall is required.

Adjacent block walls (Wall # 111-4N/23 & 111-2/23)

Systems C-02 0.139 control the equipment HCLPF capacity. Modification of Interaction block wall is required.

Anchorage screened by large available margin in existing C-207 Screened 2: 0.229 design basis calculation.

Equipment not yet installed. Evaluation required after C-601 TBD TBD installation.

Systems Adjacent block wall controls the equipment HCLPF D-01 2: 0.229 Interaction capacity.

Systems Adjacent block wall controls the equipment HCLPF D-02 2: 0.229 Interaction capacity.

Anchorage screened by large available margin in existing D-03 Screened 2: 0.229 design basis calculation.

Anchorage screened by large available margin in existing D-04 Screened 2: 0.229 design basis calculation.

Systems Adjacent block wall controls the equipment HCLPF D-05 2: 0.229 Interaction capacity.

Systems Adjacent block wall controls the equipment HCLPF D-06 2: 0.229 Interaction capacity.

Systems Adjacent block wall controls the equipment HCLPF D-07 2:0.229 Interaction capacity.

Systems Adjacent block wall controls the equipment HCLPF D-08 2: 0.229 Interaction capacity.

Systems Adjacent block wall controls the equipment HCLPF D-09 2: 0.229 Interaction capacity.

Anchorage screened by large available margin in existing D-105 Screened 2: 0.229 design basis calculation.

Anchorage screened by large available margin in existing D-106 Screened 2: 0.229 design basis calculation.

D-107 Anchorage 2: 0.229 D-108 Anchorage 2: 0.229 D-109 Anchorage 2: 0.229 Anchorage screened by large available margin in existing D-11 Screened 2: 0.229 design basis calculation.

Anchorage screened by large available margin in existing D-12 Screened 2: 0.229 design basis calculation.

Anchorage screened by large available margin in existing D-13 Screened 2:0.229 design basis calculation.

Anchorage screened by large available margin in existing D-14 Screened 2:0.229 design basis calculation.

Systems Adjacent block wall controls the equipment HCLPF D-16 2: 0.229 Interaction capacity.

Systems Adjacent block wall controls the equipment HCLPF D-18 2: 0.229 Interaction capacity.

Page 64 of 66

Failure Equipment ID HCLPF Additional Discussion Mode Systems Adjacent block wall controls the equipment HCLPF D-21 ;::: 0.229 Interaction capacity.

Systems Adjacent block wall controls the equipment HCLPF D-22 ;::o.229 Interaction capacity.

Anchorage screened by large available margin in existing D-26 Screened ;::: 0.229 design basis calculation.

Anchorage screened by large available margin in existing D-27 Screened ;::: 0.229 design basis calculation.

Anchorage screened by large available margin in existing D-301 Screened ;::: 0.229 design basis calculation.

Anchorage screened by large available margin in existing D-302 Screened ;::: 0.229 design basis calculation.

D-305 Anchorage 0.229 Equipment not yet installed. Evaluation required after D-600 TBD TBD installation.

The anchorage has a lower HCLPF value than the D-63 Anchorage ;::: 0.229 adjacent block wall. As such, adjacent block wall does not control the equipment HCLPF capacity.

The anchorage has a lower HCLPF value than the D-64 Anchorage ;::: 0.229 adjacent block wall. As such, adjacent block wall does not control the equipment HCLPF capacity.

Anchorage screened by large available margin in existing DY-OA Screened ;::: 0.229 design basis calculation.

DY-OC Anchorage ;::: 0.229 DY-00 Anchorage ;::: 0.229 Equipment not yet installed. Evaluation required after FP-3715 TBD TBD installation.

Equipment not yet installed. Evaluation required after FP-448 TBD TBD installation.

Equipment not yet installed. Evaluation required after FP-536 TBD TBD installation.

Systems Adjacent block wall (WCC block wall) controls the LT-4038 0.158 Interaction equipment HCLPF capacity. Modification is required.

Systems Adjacent block wall (WCC block wall) controls the LT-4041 0.158 Interaction equipment HCLPF capacity. Modification is required.

Equipment not yet installed. Evaluation required after P-358 TBD TBD installation.

Equipment not yet installed. Evaluation required after P358-E TBD TBD installation.

RS-SA-09 Screened ;::: 0.229 RS-SA-10 Screened ;::: 0.229 SW-2816 Screened ;::: 0.229 SW-2817 Screened ;:::0.229 SW-4478 Screened ;:::0.229 SW-4479 Screened ;::: 0.229 Adjacent block wall (WCC block wall) controls the Systems equipment HCLPF capacity. Modification of block wall is T-24A 0.158 Interaction required. Anchorage is acceptable after the approved anchorage modification is installed.

Page 65 of 66

Failure Equipment ID HCLPF Additional Discussion Mode Adjacent block wall (WCC block wall) controls the Systems equipment HCLPF capacity. Modification of block wall is T-248 0.158 Interaction required. Anchorage is acceptable after the approved anchor~ge modification is installed.

Equipment not yet installed. Evaluation required after T-30 TBD TBD installation.

Page 66 of 66