ML14357A061
| ML14357A061 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 12/16/2014 |
| From: | Dent J Entergy Nuclear Operations |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| Download: ML14357A061 (58) | |
Text
'. Entergy Entergy Nuclear Operations, Inc.
Pilgrim Nuclear Power Station 600 Rocky Hill Road Plymouth, MA 02360 John A. Dent, Jr.
Site Vice President December 16, 2014 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk 11555 Rockville Pike Rockville, MD 20852
SUBJECT:
Pilgrim's Expedited Seismic Evaluation Process Report (CEUS Sites), Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident Pilgrim Nuclear Power Station Docket No. 50-293 License No. DPR-35 LETTER NUMBER 2.14.082
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 (ML12053A340)
- 2. NEI Letter to NRC, "Proposed Path Forward for NTTF Recommendation 2.1: Seismic Reevaluations", dated April 9, 2013 (ML13101A345)
- 3. NRC Letter, "Electric Power Research Institute Final Draft Report XXXXXX, Seismic Evaluation Guidance: Augmented Approach for the Resolution of 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 (MLI13106A331)
Dear Sir or Madam:
On March 12, 2012, the Nuclear Regulatory Commission (NRC) issued Reference 1 to all power reactor licensees and holders of construction permits in active or deferred status. Enclosure 1 of Reference 1 requested each addressee located in the Central and Eastern United.States (CEUS) to submit a Seismic Hazard Evaluation within 1.5 years from the date of Reference 1.
PNPS Letter 2.14.082 Page 2 of 3 In Reference 2, the Nuclear Energy Institute (NEI) requested NRC agreement to delay submittal of the final CEUS Seismic Hazard 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.
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 Pilgrim Nuclear Power Station (Attachment 1) provides the information described in Section 7 of Reference 3 in accordance with the schedule identified in Reference 2.
This letter contains new regulatory commitments as shown in Attachment 2.
Should you have any questions concerning the content of this letter, please contact Mr. Everett (Chip) Perkins Jr. at (508) 830-8323.
I declare under penalty of perjury that the foregoing is true and correct. Executed on December 16, 2014.
Sincerely, Jon. Dent Jr.
Site /ice President JAD/rmb
Attachment:
1] Expedited Seismic Evaluation Process Report for Pilgrim Nuclear Power Station 2] List of Regulatory Commitments for Pilgrim Nuclear Power Station
PNPS Letter 2.14.082 Page 3 of 3 cc: Mr. Daniel H. Dorman Regional Administrator, Region 1 U.S. Nuclear Regulatory Commission 2100 Renaissance Boulevard, Suite 100 King of Prussia, PA 19406-1415 U. S. Nuclear Regulatory Commission Director, Office of Nuclear Reactor Regulation One White Flint North 11555 Rockville Pike Rockville, MD 20852 Ms. Nadiyah Morgan, Project Manager Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Mail Stop O-8C2A Washington, DC 20555 Mr. John Giarrusso Jr.
Planning, Preparedness & Nuclear Section Chief Mass. Emergency Management Agency 400 Worcester Road Framingham, MA 01702 U. S. Nuclear Regulatory Commission ATTN: Robert J. Fretz Jr.
Mail Stop OWFN/4A15A 11555 Rockville Pike Rockville, MD 20852-2378 U. S. Nuclear Regulatory Commission ATTN: Robert L. Dennig Mail Stop OWFN/10E1 11555 Rockville Pike Rockville, MD 20852-2378 NRC Senior Resident Inspector Pilgrim Nuclear Power Station
ATTACHMENT 1 to PNPS Letter 2.14.082 EXPEDITED SEISMIC EVALUATION PROCESS REPORT FOR PILGRIM NUCLEAR POWER STATION
EXPEDITED SEISMIC EVALUATION PROCESS (ESEP) REPORT FOR PILGRIM NUCLEAR POWER STATION (PNPS)
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Pilgrim Nuclear Power Station ESEP Report Table of Contents Page LIST OF TABLES ....................................................................................................................................... 4 LIST OF FIGURES ..................................................................................................................................... 5 1.0 PURPOSE AND OBJECTIVE ............................................................................................... ..... 6 2.0 BRIEF
SUMMARY
OF THE FLEX SEISMIC IMPLEMENTATION STRATEGIES ................................. 6 3.0 EQUIPM ENT SELECTION PROCESS AND ESEL ........................................................................... 7 3.1 Equipm ent Selection Process and ESEL ....................................................................... 7 3.1.1 ESEL Development ......................................................................................... 8 3.1.2 Power Operated Valves ................................................................................ 9 3.1.3 Pull Boxes .................................................................................................. 9 3.1.4 Term ination Cabinets .................................................................................... 9 3.1.5 Critical Instrum entation Indicators ............................................................. 10 3.1.6 Phase 2 and 3 Piping Connections .............................................................. 10 3.2 Justification for Use of Equipment That is Not the Primary Means for FLEX Im plem entation ............................................................................................................ 10 4.0 GROUND MOTION RESPONSE SPECTRUM (GM RS) .............................................................. 10 4.1 Plot of GM RS Subm itted by the Licensee .................................................................. 10 4.2 Com parison to SSE ........................................................................................................ 12 5.0 REVIEW LEVEL GROUND M OTION (RLGM ) ........................................................................... 13 5.1 Description of RLGM Selected .................................................................................... 13 5.2 Method to Estimate In-Structure Response Spectra (ISRS) ....................................... 15 6.0 SEISM IC MARGIN EVALUATION APPROACH ......................................................................... 15 6.1 Sum m ary of Methodologies Used ............................................................................. 16 6.2 HCLPF Screening Process .......................................................................................... 16 6.3 Seism ic W alkdown Approach .................................................................................... 17 6.3.1 W alkdown Approach .................................................................................. 17 6.3.2 Application of Previous W alkdown Information .......................................... 18 6.3.3 Significant W alkdown Findings ................................................................... 18 6.4 HCLPF Calculation Process ........................................................................................ 18 6.5 Functional Evaluations of Relays ................................................................................ 19 6.6 Tabulated ESEL HCLPF Values (Including Key Failure Modes) ..................................... 19 7.0 INACCESSIBLE ITEM S ................................................................................................................ 19 7.1 Identification of ESEL Item Inaccessible for W alkdowns ............................................ 19 7.2 Planned W alkdown / Evaluation Schedule / Close Out .............................................. 20 8.0 ESEP CONCLUSIONS AND RESULTS ...................................................................................... 20 8.1 Supporting Inform ation ................................................................................................ 20 Page 2
Pilgrim Nuclear Power Station ESEP Report Table of Contents (continued)
Page 8.2 Identification of Planned Modifications ..................................................................... 22 8.3 Modification Implementation Schedule ..................................................................... 22 8.4 Summary of Regulatory Commitments ..................................................................... 22 9 .0 REFER ENCES ............................................................................................................................. 22 ATTA CH M EN T A - PN PS ESEL .............................................................................................................. A-1 ATTACHMENT B - ESEP HCLPF VALUES AND FAILURE MODES TABULATION ................................... B-1 Page 3
Pilgrim Nuclear Power Station ESEP Report List of Tables Page TABLE 4-1: GM RS FOR PNPS ................................................................................................................ 10 TABLE 4-2: SSE FOR PNPS .................................................................................................................... 12 TABLE 5-1: RLGM FOR PNPS ................................................................................................................ 14 Page 4
Pilgrim Nuclear Power Station ESEP Report List of Figures Page FIGURE 4-1: GM RS FOR PNPS .............................................................................................................. 12 FIGURE 4-2: GM RS TO SSE COM PARISON FOR PNPS ........................................................................ 13 FIGURE 5-1: RLGM FOR PNPS .............................................................................................................. 15 Page 5
Pilgrim Nuclear Power Station ESEP Report 1.0 PURPOSE AND OBJECTIVE Following the accident at the Fukushima Dai-ichi nuclear power plant resulting from the March 11, 2011, Great Tohoku Earthquake and subsequent tsunami, the Nuclear Regulatory Commission (NRC) established a Near-Term Task Force (NTTF) to conduct a systematic review of NRC processes and regulations and to determine if the agency should make additional improvements to its regulatory system. The NTTF developed a set of recommendations intended to clarify and strengthen the regulatory framework for protection against natural phenomena. Subsequently, the NRC issued a 50.54(f) letter on March 12, 2012 [1], requesting information to assure that these recommendations are addressed by all U.S. nuclear power plants. The 50.54(f) letter requests that licensees and holders of construction permits under 10 CFR Part 50 reevaluate the seismic hazards at their sites against present-day NRC requirements and guidance. Depending on the comparison between the reevaluated seismic hazard and the current design basis, further risk assessment may be required. Assessment approaches acceptable to the staff include a seismic probabilistic risk assessment (SPRA), or a seismic margin assessment (SMA). Based upon the assessment results, the NRC staff will determine whether additional regulatory actions are necessary.
This report describes the Expedited Seismic Evaluation Process (ESEP) undertaken for Pilgrim Nuclear Power Station (PNPS). The intent of the ESEP is to perform an interim action in response to the NRC's 50.54(f) letter to demonstrate seismic margin through a review of a subset of the plant equipment that can be relied upon to protect the reactor core following beyond design basis seismic events.
The ESEP is implemented using the methodologies in the NRC endorsed guidance in Electric Power Research Institute (EPRI) 3002000704, Seismic Evaluation Guidance: Augmented Approach for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1: Seismic [2].
The objective of this report is to provide summary information describing the ESEP evaluations and results. The level of detail provided in the report is intended to enable the NRC to understand the inputs used, the evaluations performed, and the decisions made as a result of the interim evaluations.
2.0 BRIEF
SUMMARY
OF THE FLEX SEISMIC IMPLEMENTATION STRATEGIES The PNPS FLEX strategies for Reactor Core Cooling and Heat Removal, Reactor Inventory Control, and Containment Function are summarized below. This summary is derived from the PNPS Overall Integrated Plan (OIP) in Response to the March 12, 2012, Commission Order EA-12-049 submitted in February 2013 [3] and is consistent with the third six month status report issued to the NRC in August 2014 [4].
For Phase 1 Core cooling and inventory control are achieved during the first six (6) hours using the Reactor Core Isolation Cooling (RCIC) system aligned to take suction from the torus. Pressure control and heat removal are accomplished by Safety Relief Valves (SRV) venting to the torus. At six (6) hours, a controlled depressurization is commenced based on the Emergency Operating Procedure (EOP) heat capacity temperature limit curve. The depressurization is carried out over three (3) hours using RCIC and cycling the SRVs.
At nine (9) hours (beginning of Phase 2), the operators will transition from the installed RCIC system to diesel powered FLEX low pressure injection pumps, taking suction from the Ultimate Heat Sink (UHS) and connecting through the Condensate Storage Tank (CST) suction line for injection via either the High Pressure Coolant Injection (HPCI) or RCIC idle pump and normal pump discharge path to the Reactor Pressure Vessel (RPV) feedwater lines. An alternate FLEX injection point is to the Residual Heat Page 6
Pilgrim Nuclear Power Station ESEP Report Removal (RHR) system via the readily accessible Firewater to Service Water Cross-tie to RHR, which provides a path into the RPV, Drywell Spray, or Torus via the RHR system.
At 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />, the torus will be vented via the hardened containment vent to provide containment heat removal, and to begin a long term strategy of reactor feedwater makeup and boiling to protect the core and containment.
At 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> (beginning of Phase 3), the water source for the FLEX injection pumps will be transitioned to the mobile water tank fed from the on-site ground water wells. The ground water wells will be powered by a portable 100 kVA generator. The flow from the mobile storage tank will be passed through a FLEX Demineralizer vessel and injected into the RPV via the CST storage tank suction line (same flow path as Phase 2).
Initially containment integrity is maintained by normal design features of the containment (e.g.,
containment isolation valves). At 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />, when torus water temperature reaches 280'F, torus venting will commence through the Hardened Containment Venting System to provide containment heat removal and protect containment integrity. Containment venting will not begin until after reactor depressurization to ensure sufficient containment pressure and net positive suction head for RCIC or HPCI operation.
Necessary electrical components are outlined in the PNPS FLEX OIP submittal, and primarily entail a 125 volt motor control center, vital batteries, battery chargers, and 250 volt DC batteries and battery chargers. Other supporting components include monitoring instrumentation for core cooling, reactor coolant inventory, and containment integrity.
Figure 1 through Figure 5 of [3] provide the FLEX flow paths for PNPS Phases 1 through 3.
3.0 EQUIPMENT SELECTION PROCESS AND ESEL The selection of equipment for the Expedited Seismic Equipment List (ESEL) followed the guidelines of EPRI 3002000704 [2]. The ESEL for PNPS is presented in Attachment A. Information presented in Attachment A is drawn from the following references [3], [4], [5], [6], [7], [8], [9], [10], [11], [12], [13],
[14], [15], [16], [17], [18], [19], [20], [21], [22], [23], [24], [25], [26], [27], [28], [29], [30], [31], [32], [33],
[34], [35], and [36].
3.1 Equipment Selection Process and ESEL The selection of equipment to be included on the ESEL was based on installed plant equipment credited in the FLEX strategies during Phase 1, 2 and 3 mitigation of a Beyond Design Basis External Event (BDBEE), as outlined in the PNPS OIP in Response to the March 12, 2012, Commission Order EA-12-049 [3]. The OIP provides the PNPS FLEX mitigation strategy and serves as the basis for equipment selected for the ESEP.
The scope of "installed plant equipment" includes equipment relied upon for the FLEX strategies to sustain the critical functions of core cooling and containment integrity consistent with the PNPS OIP.
FLEX recovery actions are excluded from the ESEP scope per EPRI 3002000704 [2]. The overall list of planned FLEX modifications and the scope for consideration herein is limited to those required to support core cooling, reactor coolant inventory and subcriticality, and containment integrity functions.
Portable and pre-staged FLEX equipment (not permanently installed) are excluded from the ESEL per EPRI 3002000704.
The ESEL component selection followed the EPRI guidance outlined in Section 3.2 of EPRI 3002000704.
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Pilgrim Nuclear Power Station ESEP Report
- 1. The scope of components is limited to that required to accomplish the core cooling and containment safety functions identified in Table 3-1 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 PNPS OIP.
- 2. The scope of components is limited to installed plant equipment, and FLEX connections necessary to implement the PNPS OIP as described in Section 2.
- 3. The scope of components assumes the credited FLEX connection modifications are implemented, and are limited to those required to support a single FLEX success path (i.e.,
either "Primary" or "Back-up/Alternate").
- 4. The "Primary" FLEX success path is to be specified. Selection of the "Back-up/Alternate" FLEX success path must be justified.
- 5. Phase 3 coping strategies are included in the ESEP scope, whereas recovery strategies are excluded.
- 6. Structures, systems, and components excluded per the EPRI 3002000704 [2] guidance are:
- Structures (e.g. containment, reactor building, control building, auxiliary building, etc.).
- Piping, cabling, conduit, HVAC, and their supports.
- Manual valves, check valves and rupture disks.
" Power-operated valves not required to change state as part of the FLEX mitigation strategies.
- Nuclear steam supply system components (e.g. RPV and internals, reactor coolant pumps and seals, etc.).
- 7. For cases in which neither train was specified as a primary or back-up strategy, then only one train component (generally 'A' train) is included in the ESEL.
3.1.1 ESEL Development The ESEL was developed by reviewing the PNPS OIP [3] to determine the major equipment involved in the FLEX strategies. Further reviews of plant drawings (e.g., Piping and Instrumentation Diagrams (P&IDs) and Electrical One Line Diagrams) were performed to identify the boundaries of the 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 / 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.
Cabinets and equipment controls containing relays, contactors, switches, potentiometers, circuit breakers and other electrical and instrumentation that could be affected by high-frequency earthquake motions and that impact the operation of equipment in the ESEL are required to be on the ESEL. These cabinets and components were identified in the ESEL. For the ESEL, the relays identified were in the Page 8
Pilgrim Nuclear Power Station ESEP Report RCIC and Automatic Depressurization System (ADS), and malfunction of these relays during a seismic event could lead to the failure of the reactor core cooling safety function.
For Phase 1, RCIC is the primary path for inventory control and core cooling. Therefore, the RCIC system was used as the basis for the Phase 1 ESEL. For Phase 2 and Phase 3, the RCIC system was also used to provide the pathway for RPV injection utilizing portable injection pumps. Relays that could malfunction during a seismic event and prevent successful RCIC or ADS operation were included in the ESEL.
For each parameter monitored during the FLEX implementation, a single indication was selected for inclusion in the ESEL. For each parameter indication, the components along the flow path from measurement to indication were included, since any failure along the path would lead to failure of that indication. Components such as flow elements were considered as part of the piping and were not included in the ESEL.
3.1.2 Power Operated Valves Page 3-3 of EPRI 3002000704 [2] notes that power operated valves not required to change state as part of the FLEX mitigation strategies are excluded from the ESEL. Page 3-2 also notes that "functional failure modes of electrical and mechanical portions of the installed Phase 1 equipment should be considered (e.g. RCIC/AFW trips)." To address this concern, the following guidance is applied in the PNPS 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 be added to the ESEL as these components provide completely passive locations for pulling or installing cables. No breaks or connections in the cabling were included in pull boxes. Pull boxes were considered part of conduit and cabling, which were excluded in accordance with EPRI 3002000704 [2].
3.1.4 Termination Cabinets Termination cabinets, including cabinets necessary for FLEX Phase 2 and Phase 3 connections, provide consolidated locations for permanently connecting multiple cables. The termination cabinets and the internal connections provide a completely passive function; however, the cabinets are included in the ESEL to ensure industry knowledge on panel/anchorage failure vulnerabilities is addressed.
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Pilgrim Nuclear Power Station ESEP Report 3.1.5 Critical Instrumentation Indicators Critical indicators and recorders are typically physically located on panels/cabinets and are included as separate components; however, seismic evaluation of the instrument indication may be included in the panel/cabinet seismic evaluation (rule-of-the-box).
3.1.6 Phase 2 and 3 Piping Connections Item 2 in Section 3.1 above notes that the scope of equipment in the ESEL includes "... FLEX connections necessary to implement the PNPS OIP [3] as described in Section 2." Item 3 in Section 3.1 also notes that "The scope of components assumes the credited FLEX connection modifications are implemented, and are limited to those required to support a single FLEX success path (i.e., either "Primary" or "Back-up/Alternate")."
Item 6 in Section 3.1 above goes on to explain that "Piping, cabling, conduit, HVAC, and their supports" are excluded from the ESEL scope in accordance with EPRI 3002000704 [2].
Therefore, piping and pipe supports associated with FLEX Phase 2 and Phase 3 connections are excluded from the scope of the ESEP evaluation. However, any active valves in FLEX Phase 2 and Phase 3 connection flow path are included in the ESEL.
3.2 Justification for Use of Equipment That is Not the Primary Means for FLEX Implementation RCIC is the primary system for Phase 1 and was presented as the single success path in the PNPS ESEL.
Therefore, no additional justification is required.
4.0 GROUND MOTION RESPONSE SPECTRUM (GMRS) 4.1 Plot of GMRS Submitted by the Licensee The Safe Shutdown Earthquake (SSE) control point elevation is defined at the bottom of the Reactor Building foundation at elevation -26 ft MSL which is 48 ft below grade based on Section 2.5.3.3.2, Section 2.5.2.4.3, and Figure 12.2-6 of the Final Safety Analysis Report (FSAR) [37]. Table 4-1 shows the GMRS acceleration for a range of frequencies [38]. The GMRS at the control point is shown in Figure 4-1.
Table 4-1: GMRS for PNPS Frequency GMRS (Hz) (g) 100 5.05E-01 90 5.09E-01 80 5.18E-01 70 5.40E-01 60 5.93E-01 50 7.31E-01 40 8.57E-01 35 9.21E-01 Page 10
Pilgrim Nuclear Power Station ESEP Report Table 4-1: GMRS for PNPS (continued)
Frequency GMRS (Hz) (g) 30 9.51E-01 25 9.22E-01 20 9.09E-01 15 9.61E-01 12.5 1.06E+00 10 1.18E+00 9 1.16E+00 8 1.10E+00 7 9.75E-01 6 8.07E-01 5 6.09E-01 4 3.83E-01 3.5 2.93E-01 3 2.19E-01 2.5 1.61E-01 2 1.29E-01 1.5 1.OOE-01 1.25 7.82E-02 1 6.22E-02 0.9 5.87E-02 0.8 5.55E-02 0.7 5.18E-02 0.6 4.67E-02 0.5 3.92E-02 0.4 3.14E-02 0.35 2.74E-02 0.3 2.35E-02 0.25 1.96E-02 0.2 1.57E-02 0.15 1.18E-02 0.125 9.80E-03 0.1 7.84E-03 Page 11
Pilgrim Nuclear Power Station ESEP Report GMRS at Control Point for Pilgrim Nuclear Power Station, 5% Damping 1.40 1.20 1.00 0.80 0.60 0.40 0.20 0.00 0.1 10 100 Frequency (Hz)
Figure 4-1: GMRS for PNPS 4.2 Comparison to SSE The SSE is defined in the FSAR in terms of a Peak Ground Acceleration (PGA) and a design response spectrum. These spectra have been digitized and tabulated [39]. Table 4-2 shows the spectral acceleration values at selected frequencies for the 5% damped horizontal SSE.
Table 4-2: SSE for PNPS Frequency Spectral Acceleration (Hz) (g) 100 0.15 33 0.15 25 0.15 10 0.184 9 0.194 5 0.238 2.5 0.225 1 0.126 0.5 0.071 Page 12
Pilgrim Nuclear Power Station ESEP Report GMRS to SSE Comparison for Pilgrim Nuclear Power Station, 5% Damping 1.40 1.20 1.00 080 0.60 0.40 0.20 0.00 0.1 1 10 100 Frequency (Hz)
Figure 4-2: GMRS to SSE Comparison for PNPS The SSE envelops the GMRS in the low frequency range up to approximately 3 Hz. The GMRS exceeds the SSE beyond that point. As the GMRS exceeds the SSE in the 1 to 10 Hz range, the plant does not screen out of the ESEP according to Section 2.2 of EPRI 3002000704 [2]. The two special screening considerations as described in Section 2.2.1 of EPRI 3002000704, namely a) Low-frequency GMRS exceedances at Low Seismic Hazard Sites and b) Narrow Band Exceedances in the 1 to 10 Hz range, provide criteria for accepting specific GMRS exceedances. However, the GMRS exceedances are not limited to the low frequency range and there are no narrow-banded exceedances. Therefore, these special screening considerations do not apply for PNPS and hence High Confidence of a Low Probability of Failure (HCLPF) evaluations were performed.
5.0 REVIEW LEVEL GROUND MOTION (RLGM) 5.1 Description of RLGM Selected The RLGM is selected based on Approach 1 in Section 4 of EPRI 3002000704 [2]. The RLGM is developed based on the SSE [39].
The maximum GMRS/SSE ratio between 1 and 10 Hz range occurs at 10 Hz where the ratio is 1.18/0.184 = 6.41. As the maximum ratio of the GMRS to the SSE over the 1 to 10 Hz range exceeds a value of 2, the GMRS/SSE ratio is set to the maximum scaling factor value of 2.0 for PNPS in accordance with Section 4 of EPRI 3002000704. Table 5-1 lists the horizontal ground RLGM acceleration at 5%
damping at selected frequencies and the plot is shown in Figure 5-1. The RLGM are generated by plotting the digitized data on a linear/linear graph paper, and connecting the points with straight lines.
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Pilgrim Nuclear Power Station ESEP Report Table 5-1: RLGM for PNPS Frequency RLGM at 5% Damping (Hz) (g) 0.10 0.016 0.30 0.094 0.60 0.166 1.00 0.252 1.40 0.318 2.00 0.412 2.50 0.450 3.00 0.476 3.50 0.488 4.00 0.488 4.50 0.484 5.00 0.476 5.50 0.466 6.00 0.450 7.00 0.428 8.00 0.406 10.00 0.368 12.00 0.338 14.00 0.308 18.00 0.300 24.00 0.300 33.00 0.300 Page 14
Pilgrim Nuclear Power Station ESEP Report Review Level Ground Motion (2xSSE) Response Spectra - Horizontal Direction 0.60 - - ' -' * *-
156%0Damping _
0.40
- I I _I 0.20 44 :'
-f------ -
... .... . . t - .. .. . 1-!.. .. .. .
010 0.00, ,
4 8 12e16 20 24 28 32 Frequency (Hz)
Figure 5-1: RLGM for PNPS 5.2 Method to Estimate In-Structure Response Spectra (ISRS)
The RLGM ISRS for PNPS are generated by scaling the SSE ISRS [39]. The following steps are used to generate the RLGM ISRS.
- 2. Calculate the horizontal RLGM ISRS by scaling the horizontal direction SSE ISRS by a factor of 2.0.
The vertical direction RLGM ISRS is obtained by scaling the vertical amplified ground response spectrum.
6.0 SEISMIC MARGIN EVALUATION APPROACH It is necessary to demonstrate that ESEL items have sufficient seismic capacity to meet or exceed the demand characterized by the RLGM. The seismic capacity is characterized as the PGA for which there is a HCLPF. The PGA is associated with a specific spectral shape, in this case the 5%-damped RLGM spectral shape. The HCLPF capacity must be equal to or greater than the RLGM PGA. The criteria for seismic capacity determination are given in Section 5 of EPRI 3002000704 [2].
There are two basic approaches for developing HCLPF capacities:
- 1. Deterministic approach using the conservative deterministic failure margin (CDFM) methodology of EPRI NP-6041-SL, A Methodology for Assessment of Nuclear Power Plant Seismic Margin (Revision 1) [40].
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Pilgrim Nuclear Power Station ESEP Report
- 2. Probabilistic approach using the fragility analysis methodology of EPRI TR-103959, Methodology for Developing Seismic Fragilities [41].
6.1 Summary of Methodologies Used PNPS performed a SPRA in 1994 as part of Individual Plant Examination for External Events (IPEEE) program. The SPRA is documented in [23] and consisted of screening walkdowns, fragility analysis and three-dimensional soil structure interaction (SSI) analysis. The SPRA was a hybrid of the conventional PRA and seismic margin assessment approaches. The seismic walkdowns for IPEEE were performed simultaneously with USI A-46 evaluations. Section 3.3 of [38] established that the results of the PNPS SPRA performed as part of IPEEE are not sufficient to serve as the basis for PNPS to screen-out of further risk assessment.
For ESEP, the SMA consisted of screening walkdowns and HCLPF calculations. The screening walkdowns used the screening tables from Chapter 2 of EPRI NP-6041-SL [40]. The walkdowns were conducted by engineers trained in EPRI NP-6041-SL and were documented on Screening Evaluation Work Sheets (SEWS) from EPRI NP-6041-SL. Anchorage capacity calculations used the CDFM criteria from EPRI NP-6041-SL. Seismic demand was based on EPRI 3002000704 [2] using an RLGM of 2xSSE with a PGA of 0.3g, Figure 5-1.
6.2 HCLPF Screening Process For ESEP, the components are screened considering the RLGM (2xSSE) with a 0.3g PGA. The screening tables in EPRI NP-6041-SL [40] are based on ground peak spectral accelerations of 0.8g and 1.2g. These both exceed the RLGM peak spectral acceleration.
The ESEL components were prescreened based on Table 2-4 of EPRI NP-6041-SL. Additional pre-screening, specifically for anchorage, considered walkdown results and documentation from NTTF 2.3 and SEWS from IPEEE and USI A-46. Equipment anchorage was screened out in cases where previous evaluations showed large available margin against SSE. The remaining components (i.e., components that do not screen out), were identified as requiring HCLPF calculations. ESEL components were walked down and based on the equipment and anchorage conditions, prescreening decisions were confirmed and a final list of required HCLPF calculations was generated. Equipment for which the screening caveats were met and for which the anchorage capacity exceeded the RLGM seismic demand are screened out from ESEP seismic capacity determination because the HCLPF capacity exceeds the RLGM.
The PNPS ESEL contains 187 items. Of these, 16 are valves. In accordance with Table 2-4 of EPRI NP-6041-SL, active valves may be assigned a functional capacity of 0.8g peak spectral acceleration without any review other than looking for valves with large extended operators on small diameter piping, and anchorage is not a failure mode. Therefore, valves on the ESEL are screened out from ESEP seismic capacity determination, subject to the caveat regarding large extended operators on small diameter piping.
The non-valve components in the ESEL are screened based on the SMA results. If the SMA showed that the component met the EPRI NP-6041-SL screening caveats and the CDFM capacity exceeded the RLGM demand, the components are screened out from the ESEP capacity determination.
Page 16
Pilgrim Nuclear Power Station ESEP Report Six (6) block walls were identified in the proximity of ESEL equipment. These block walls were assessed for potential seismic interaction impact resulting from the RLGM by reviewing the existing plant documents and or by generating new analysis and found to be acceptable.
6.3 Seismic Walkdown Approach 6.3.1 Walkdown Approach Walkdowns were performed in accordance with the criteria provided in Section 5 of EPRI 3002000704
[2], which refers to EPRI NP-6041-SL [40] for the Seismic Margin Assessment process. Pages 2-26 through 2-30 of EPRI NP-6041-SL 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-radioactiveor low radioactiveenvironments. Seismic capability assessment of components which are inaccessible, in high-radioactiveenvironments, or possibly within contaminatedcontainment, will have to rely more on alternatemeans such as photographicinspection, 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 reasonablebasisfor 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 preparatorywork (Step 3) by reference to drawings, calculationsor specifications. The one component or each type which is selected should be thoroughly inspected which probably does mean de-energizing and opening cabinets or panelsfor this very limited sample. Generally, a spare representativecomponent can be found so as to enable the inspection to be performed while the plant is in operation. At leastfor the one component of each type which is selected, anchorageshould 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 correspondenceis 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 procedurefor inspection should be repeatedfor 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 arefound, then the system or component class must be inspected in closer detail until the systematic deficiency is defined.
The 100% "walk by" is to look for outliers, lack of similarity, anchoragewhich is differentfrom that shown on drawings or prescribedin criteriafor that component, potentialSI [Seismic Interaction]problems, situationsthat are at odds with the team members' past experience, and any other areas of seriousseismic 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.
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Pilgrim Nuclear Power Station ESEP Report 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 responsiblefor the seismic adequacy of all elements which they screen from the margin review. Appendix D gives guidancefor sampling selection."
6.3.2 Application of Previous Walkdown Information Several ESEL items were previously walked down during the PNPS seismic IPEEE program, for seismic IPEEE outlier resolutions in accordance with USI A-46 evaluation program and NTTF Recommendation 2.3. Those walkdown results were reviewed and the following steps were taken to confirm that the previous walkdown conclusions remained valid.
- A walk by was performed to confirm that the equipment material condition and configuration is consistent with the walkdown conclusions and that no new significant interactions related to block walls or piping attached to tanks exist.
- If the ESEL item was screened out based on the previous walkdown, that screening evaluation was reviewed and reconfirmed for the ESEP.
6.3.3 Significant Walkdown Findings Consistent with the guidance from EPRI NP-6041-SL [40], no significant outliers or anchorage concerns were identified during the PNPS seismic walkdowns. Based on walkdown results, HCLPF capacity evaluations were recommended for the following eight (8) components:
" C7, Containment Isolation and Ventilation Vertical Board
- C2205A, Reactor Protection and NSS Inst. Rack
- C2251A, Jet Pump Instrument Rack A
- D16, 125 VDC Bus A
- D1, 125 VDC Battery Rack A
- D2, 125 VDC Battery Rack B
- D3, 125 VDC Battery Rack B
" EG-23, Vital MG Set 6.4 HCLPF Calculation Process ESEL items identified for ESEP at PNPS were evaluated using the criteria in EPRI NP-6041-SL [40] and Section 5 of EPRI 3002000704 [2]. Those evaluations included the following steps:
- Performing seismic capability walkdowns for equipment not included in previous seismic walkdowns (SQUG, IPEEE, or NTTF 2.3) to evaluate the equipment installed plant conditions
- Performing screening evaluations using the screening tables in EPRI NP-6041-SL as described in Section 6.2
- Performing HCLPF calculations considering various failure modes that include both structural failure modes (e.g. anchorage, load path etc.) and functional failure modes All HCLPF calculations were performed using the CDFM methodology. A total of seven (7) HCLPF calculations were performed to address the eight (8) components.
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Pilgrim Nuclear Power Station ESEP Report
- C7, "Containment Isolation and Ventilation Vertical Board"
- C2205A, "Reactor Protection and NSS Inst. Rack"
- C2251A, "Jet Pump Instrument Rack A"
- D16, "125 VDC Bus A"
" D2 and D3, "125 VDC Battery Rack B"
- D1, "125 VDC Battery Rack A"
- EG-23, "Vital MG Set" 6.5 Functional Evaluations of Relays Five (5) relays 13A-K3, -K5, -K7, K10, and -K22 associated with "RCIC Relay Vertical Board" cabinet C930, and three (3) relays 13A-K31, -K32, and -K33 associated with "Channel B Vertical Board" cabinet C933 were identified as seal in/lockout type needing HCLPF calculation. The relays were of three types General Electric 12HFA151A2F, General Electric 12HGA11A52F and Agastat 7014PB.
The relays were evaluated using the guidance provided in EPRI NP-6041-SL [40] for equipment qualified by testing. Subject relays were determined to have higher HCLPF values than the plant RLGM.
6.6 Tabulated ESEL HCLPF Values (Including Key Failure Modes)
Tabulated ESEL HCLPF values are provided in Attachment B. The following notes apply to the information in the tables.
- For items screened out using EPRI NP-6041-SL [40] screening tables, the HCLPF capacity is provided as >RLGM and the failure mode is listed as "Screened", (unless the controlling HCLPF value is governed by anchorage).
- For items where anchorage controls the HCLPF value, the HCLPF value is listed in the table and the failure mode is noted as "anchorage." For the items where the component function controls the HCLPF value, the HCLPF value is listed in the table and the failure mode is noted as "functional."
After performing the HCLPF calculations, the anchorage was determined to have adequate capacity for the design basis loads and HCLPF greater than RLGM for all components except EG-23. A modification is planned for EG-23 and the HCLPF capacity presented in Attachment B includes the proposed modifications.
7.0 INACCESSIBLE ITEMS 7.1 Identification of ESEL Item Inaccessible for Walkdowns There are total of 13 Torus Water Temperature Elements (TEs) on the ESEL. Six (6) of the TEs were not walked down since they are located in a high dose area. The evaluation of subject TEs were done by comparison and similarity to the other seven (7) TEs that were walked down. The following is the list of the TEs that were not walked down:
- TE5021-01A
" TE5021-06A Page 19
Pilgrim Nuclear Power Station ESEP Report
- TE5021-07A
- TE5021-08A
- TE5021-10A
- TE5021-12A Also, the two (2) valves and two (2) accumulator tanks listed below were not walked down, since they are located in the Dry well (inaccessible area). Subject components were evaluated based on the available photos, drawings, existing SEWS, and vendor information.
- RV-203-3B, Safety Relief Valve
- RV-203-3C, Safety Relief Valve
- T-221B, Accumulator Tank for SRV B
- T-221C, Accumulator Tank for SRV C In addition two (2) junction boxes J599 and J600 were not walked down since they were not accessible for visual inspection due to their location. These items were assessed and found to be acceptable by comparison and similarity to J601 and J602 respectively, and by reviewing their A-46 SEWS, and existing analysis information.
7.2 Planned Walkdown / Evaluation Schedule / Close Out There are no components that require follow up seismic walkdowns.
8.0 ESEP CONCLUSIONS AND RESULTS 8.1 Supporting Information PNPS has performed the ESEP as an interim action in response to the NRC's 50.54(f) letter [1]. It was performed using the methodologies in the NRC endorsed guidance in EPRI 3002000704 [2].
The ESEP provides an important demonstration of seismic margin and expedites plant safety enhancements through evaluations and potential near-term modifications of plant equipment that can be relied upon to protect the reactor core following beyond design basis seismic events.
The ESEP is part of the overall PNPS response to the NRC's 50.54(f) letter. On March 12, 2014, NEI submitted to the NRC results of a study [43] 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
[38]. 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 [42] concluded that the "fleet wide seismic risk estimates are consistent with the approach and results used in the GI-199 safety/risk assessment." The letter also stated that "As a result, the staff has confirmed that the conclusions reached in GI-199 safety/risk assessment remain valid and that the plants can continue to operate while additional evaluations are conducted."
Page 20
Pilgrim Nuclear Power Station ESEP Report An assessment of the change in seismic risk for PNPS was included in the fleet risk evaluation submitted in the March 12, 2014 NEI letter [43]; therefore, the conclusions in the NRC's May 9 letter also apply to PNPS.
In addition, the March 12, 2014 NEI letter provided an attached "Perspectives on the Seismic Capacity of Operating Plants," which (1) assessed a number of qualitative reasons why the design of Structures, Systems and Components (SSCs) inherently contain margin beyond their design level, (2) discussed industrial seismic experience databases of performance of industry facility components similar to nuclear SSCs, and (3) discussed earthquake experience at operating plants.
The fleet of currently operating nuclear power plants was designed using conservative practices, such that the plants have significant margin to withstand large ground motions safely. This has been borne out for those plants that have actually experienced significant earthquakes. The seismic design process has inherent (and intentional) conservatisms which result in significant seismic margins within SSCs.
These conservatisms are reflected in several key aspects of the seismic design process, including:
- Safety factors applied in design calculations
- Damping values used in dynamic analysis of SSCs
- Bounding synthetic time histories for in-structure response spectra calculations
" Broadening criteria for in-structure response spectra
" Response spectra enveloping criteria typically used in SSC analysis and testing applications
- Response spectra based frequency domain analysis rather than explicit time history based time domain analysis
- Bounding requirements in codes and standards
- Use of minimum strength requirements of structural components (concrete and steel)
- Bounding testing requirements
" Ductile behavior of the primary materials (that is, not crediting the additional capacity of materials such as steel and reinforced concrete beyond the essentially elastic range, etc.)
These design practices combine to result in margins such that the SSCs will continue to fulfill their functions at ground motions well above the SSE.
The intent of the ESEP is to perform an interim action in response to the NRC's 50.54(f) letter to demonstrate seismic margin through a review of a subset of the plant equipment that can be relied upon to protect the reactor core following beyond design basis seismic events. The RLGM used for the ESEP evaluation is a scaled version of the plant's SSE rather than the actual GMRS. To more fully characterize the risk impacts of the seismic ground motion represented by the GMRS on a plant specific basis, a more detailed seismic risk assessment (SPRA or risk-based SMA) is to be performed in accordance with EPRI 1025287 [44] . As identified in the PNPS Seismic Hazard and GMRS submittal
[38], PNPS screens in for a risk evaluation. The complete risk evaluation will more completely characterize the probabilistic seismic ground motion input into the plant, the plant response to that probabilistic seismic ground motion input, and the resulting plant risk characterization. PNPS will complete that evaluation in accordance with the schedule identified in NEI's letter dated April 9, 2013
[45] and endorsed by the NRC in their May 7, 2013 letter [46].
Page 21
Pilgrim Nuclear Power Station ESEP Report 8.2 Identification of Planned Modifications Insights from the ESEP identified the following item where the HCLPF is below the RLGM and plant modifications will be made in accordance with EPRI 3002000704 [2] to enhance the seismic capacity of the plant.
- Vital MG Set EG-23 anchorage had a HCLPF capacity below RLGM. A modification is planned to provide additional seismic margin such that the HCLPF will exceed the RLGM.
8.3 Modification Implementation Schedule Plant modifications described in Section 8.2 will be performed in accordance with the schedule identified in NEI letter dated April 9, 2013 [45], 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.
8.4 Summary of Regulatory Commitments The following actions will be performed as a result of the ESEP.
Equipment Action # Equipment ID Description Action Description Completion Date 1 EG-23 Vital MG Set Modify anchorage such that As described in HCLPF > RLGM Section 8.3 2 N/A N/A Submit a letter to NRC Within 60 days summarizing the HCLPF results of following Item 1 confirming completion of ESEP implementation of the plant activities, including modification associated with item 1 item 1
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, NRC ADAMS Accession No. ML12053A340.
- 2. EPRI 3002000704, "Seismic Evaluation Guidance, Augmented Approach for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1: Seismic," May 2013.
- 3. Entergy Letter to U.S. NRC, letter number 2.13.012 "Overall Integrated Plan in Response to March 12, 2012, Commission Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond Design Basis Events (Order Number EA-12-049)," February 28, 2013, NRC ADAM Accession No. ML13063A063.
- 4. Entergy Letter to U.S. NRC, letter number 2.14.061 "Pilgrim Nuclear Power Station's Third Six Month Status Report in Response to March 12, 2012, Commission Order Modifying Licenses Page 22
Pilgrim Nuclear Power Station ESEP Report with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events (Order Number EA-12-049)," August 28, 2014, NRC ADAMS Accession No. ML14253A189.
- 6. Entergy Drawing E13, Revision E83, "Single Line Relay & Meter Diagram 125V and 250V DC Systems."
8.. Entergy Drawing M1G13-11, Revision E17, "Elementary Diagram RCIC System, Sheet 3 of 9."
- 10. Entergy Drawing MlG12-12, Revision E14, "Elementary Diagram RCIC System, Sheet 2 of 9."
- 11. Entergy Drawing M252SH1, Revision 69, "P & ID Nuclear Boiler."
- 12. Entergy Drawing M1R4-10, Revision 25, "Elementary Diagram Automatic Blowdown System, Sheet 1 of 2."
- 13. Entergy Drawing M253SH1, Revision 45, "Nuclear Boiler Vessel Instrumentation."
- 14. Entergy Drawing M253SH2, Revision 29, "Nuclear Boiler Vessel Instrumentation."
- 15. Entergy Drawing E91, Revision E8, "Wiring Block Diagram RCIC System, Sheet 1 of 2."
- 16. Entergy Drawing M241SH1, Revision 87, "P&ID Residual Heat Removal System."
- 17. Entergy Drawing M227SH1, Revision 60, "P&ID Containment Atmospheric Control System."
- 18. Entergy Drawing E14SH1, Revision 39, "Single Line Diagram 120 V Instrument AC Vital and Reactor Protection AC System & +/-24 VDC Power System."
- 19. Entergy Drawing M1R8-2, Revision 10, "Elementary Diagram Automatic Blowdown System."
- 20. Entergy Document ELNRC1.2.96.085, BECo Letter 96-085, "Summary Report, Generic Letter 87-02, Verification of Seismic Adequacy of Mechanical and Electrical Equipment in Operating Reactors, Unresolved Safety Issue (USI) A-46."
- 21. Entergy Drawing E727, Revision E7, "Elementary Diagram Emergency Core Cooling System Analog Trip Cabinet C2233A Section A."
- 22. Entergy Drawing M1P361-2, Revision E3, "Arrangement Diagram Reactor Core Isolation Cooling System Instrument Rack C2258, Sheet 3 of 4."
- 23. Entergy Correspondence IPEEE, "Pilgrim Nuclear Power Station Individual Plant Examination for External Events (GL 88-20)," dated July 1994.
- 24. Entergy Drawing M1P302-15, Revision Ell, "Arrangement Drawing Control Room Panel C903, Sheet 2 of 2."
- 25. Entergy Drawing E692, Revision E6, "Elementary Diagram Torus Water Temperature Monitoring System Channel A."
- 26. Entergy Drawing M206B, Revision EO, "Control Room and Local Panel Instruments."
- 27. Entergy Drawing E401SH2, Revision E0, "Schematic Diagram Containment Atmospheric Control System."
Page 23
Pilgrim Nuclear Power Station ESEP Report
- 28. Entergy Drawing M227A32-10, Revision EO, "C170/C171 Post Accident Monitoring System."
- 29. Entergy Document EQDFDPT1001-604A, Revision 7, "Equipment Qualification Data Sheet DFPT1001-604A."
- 30. Entergy Drawing M227-154, Revision E8, "Front View Layout for Containment Ventilation Isolation & Gas Treatment - Vertical Board C-7."
- 31. Entergy Drawing E401SH3, Revision El, "Schematic Diagram Containment Atmospheric Control System."
- 32. Entergy Drawing M227C1, Revision EO, "Arrangement Drawing Torus Water Temperature Monitoring Sys. Panel C179."
- 33. Entergy Drawing MlG17-7 Revision E6, "Elementary Diagram RCIC System, Sheet 7 of 9."
- 34. Entergy Drawing M1P355-5 Revision 4, "Arrangement Drawing Leak Detection System Instrument Rack C2257."
- 35. Entergy Document EC45555 Revision 1, "FLEX Alternate Power to 125VDC and 250VDC Battery Chargers (Base EC)."
- 36. Entergy Document EC42259 Revision 0, "PNPS FLEX Strategy Master EC for Beyond-Design-Basis External Events (BDBEEs) Diverse & Flexible Coping Strategy (FLEX) Implementation."
- 37. "Pilgrim Nuclear Power Station - Final Safety Analysis Report," Revision 29, Docket No. 50-293, October 2013.
- 38. Entergy Letter Number 2.14.026, John A. Dent Jr. to NRC, "Entergy's Seismic Hazard and Screening Report (CEUS Sites), Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding the Seismic Aspects of Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident dated March 31, 2014." NRC ADAMS Accession No. ML14092A023.
- 39. Entergy Document C114ERQE1, Revision El, "Seismic Response Spectra," October 2005. (Stored in Merlin as C114ERQEO)
- 40. EPRI-NP-6041-SL, "Methodology for Assessment of Nuclear Power Plant Seismic Margin,"
Revision 1, August 1991.
- 41. EPRI TR-103959, "Methodology for Developing Seismic Fragilities," July 1994.
- 42. 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-Ichi Accident," May 9, 2014, NRC ADAMS Accession No. ML14111A147.
- 43. 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.
- 44. EPRI 1025287, "Seismic Evaluation Guidance: Screening, Prioritization and Implementation Details (SPID) for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1:
Seismic. Electric Power Research Institute," February 2013.
Page 24
Pilgrim Nuclear Power Station ESEP Report
- 45. Nuclear Energy Institute (NEI), A. Pietrangelo, Letter to D. Skeen of the USNRC, "Proposed Path Forward for NTTF Recommendation 2.1: Seismic Reevaluations," April 9, 2013, NRC ADAMS Accession No. ML13101A379.
- 46. 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 Alternative to the March 12, 2012, Information Request for Seismic Reevaluations," May 7, 2013, NRC ADAMS Accession No. ML13106A331.
- 47. Entergy Document EC53987, "Expedited Seismic Evaluation Process (ESEP) Report (Submittal to NRC - Fukushima 50.54F Response", the following AREVA documents are captured in the plant document management system:
Pilgrim Nuclear Power Station."
- b. AREVA Calculation 32-9224839-001, "Pilgrim ESEP HCLPF Calculation - Reactor Protection and NSS Instrument Rack, C2205A."
- c. AREVA Calculation 32-9225449-001, "Pilgrim ESEP HCLPF Calculation - Containment Isolation and Ventilation Vertical Board, C7."
- e. AREVA Calculation 32-9225626-001, "Pilgrim ESEP HCLPF Calculation - Jet Pump A Instrument Rack C2251A."
- g. AREVA Calculation 32-9226906-001, "Pilgrim ESEP HCLPF Calculation - Motor-Generator Set, EG-23."
- 48. Entergy Calculation C15.0.3623, "FLEX Transfer Switch Mounting Evaluation for Battery Chargers Dl1, D12, D13, D14, D15," Revision OA.
- 49. Entergy Calculation C15.0.3624, "FLEX Transfer Switch Mounting Evaluation for 120 VAC Panels Y3, Y31, Y4, Y41, Y13, and Y14," Revision 0.
- 50. Entergy Calculation C15.0.3631, "Nitrogen Cylinder and Pressure Regulator Supports - Backup Nitrogen Supply to Main Steam SRVs RV-203-3B and RV-203-3C," Revision OB Page 25
Pilgrim Nuclear Power Station ESEP Report ATTACHMENT A - PNPS ESEL Page A-1
Pilgrim Nuclear Power Station ESEP Report ESEL Equipment>'. Operating State ,
Item . Nom. ** De" r.. d* . otes/Comments t References
- .*,;* *.* .. ;.::° ,*'Norm ali .
- Desired'. *: : * , '" ." .....; ,.:...". . .
Number ID Description.. ., . te . State. * ,..
I MO-1301-22 RCIC Pump CST Suction Must close to swap RCIC suction to the [5]
VaCvC Psuppression pool 2 D754 DC Bus D7 Supply Energized Energized Supply power to MO-1301-22 [6]
Breaker to MO-1301-22 The FLEX Primary External Water Source Injection Point is to the HPCI & RCIC FLEX Primary External System 18-inch CST common suction line 3 TBD Water Source Injection Closed Open and shall be established at the CST T- [3]
Point 105A/B piping vault on the plant-side of the 26-HO-78 & 79 CST 18-inch manual isolation butterfly valves.
4 Normally closed, must open to supply [5]
M0-1301-25 4 RCIC Suction from Torus Closed Open RCIC suction from torus D761 DC Bus D7 supply Energized Energized Supply power to MO-1301-25 [6]
breaker to MO-1301-25 6 MO-1301-26 RCIC Suction from Torus Closed Open Normally closed, must open to supply [5]
RCIC suction from torus 7 DC Bus D7 supply Energized Energized Supply power to MO-1301-26 [6]
7D764 breaker to MO-1301-26 8 P-206 RCIC Pump Idle Operating Provides RPV makeup in Phase 1 [7]
9 ft~f12A1Af ValveNormally closed, must open to supply ]
MO-1301-49 RCIC Discharge Valve Isolation Closed Open RCICrCaC cton flow injection flow[5]
10 D774 DC Bus D7 Supply Energized Energized Supply power to MO-1301-49 [6]
Breaker to MO-1301-49 RCIC Lube Oil Cooling Open/Close Controls cooling water flow to RCIC lube 11 PCV 1301-43 Water Pressure Control Open as needed oil cooler[7]
_ Valve as needed oIl Iooler Page A-2
Pilgrim Nuclear Power Station ESEP Report ESEL , Equipment. Operating State Item'
- State State
- Notes/Comments , . References DesripionNormal - Desired ~
Number ID Decito.
RCICTurine ubeOilNormally closed, opens to allow cooling 1MO1062inlet Valve Closed Open water flow to RCIC turbine lube oil [7]
cooler 13 D794 DC Bus D7 Supply Energized Energized Supply power to MO-1301-62 [6]
Breaker to MO-1301-62 RCIC Barometric Collects and condenses RCIC gland seal 14 E-201 Condenser and Vacuum Idle Operating leakage [7]
_________ ~~Tank ______
15 E-204 RCIC Lube Oil Cooler Idle Operating Cools RCIC lube oil [7]
16 P-221 RCIC Vacuum Tank Idle Operating Pumps condensate from vacuum tank to [7]
Condensate Pump pump suction 17 D712 DC Bus D7 Supply Energized Energized Supply power to P-221 [6]
Breaker to P-221 18 P-222 . RCIC Vacuum Pump Idle Operating Maintains vacuum on barometric [7]
condenser 19 D714 DC Bus D7 Supply Energized Energized Supply power to P-222 [6]
Breaker to P-222 20 MO-1301-61 RCIC Turbine Steam CoeOpn Opens on RCIC start to admit steam to [7]
Inlet Valve CoeOpn RCIC turbine 21 D751 DC Bus D7 Supply Energized Energized Supply power to MO-1301-61 [6]
Breaker to MO-1301-61 22 SV-1301-1 RCIC Turbine Trip Open Open Closes on protective signals that will be [7]
Throttle Valve bypassed by procedure 23 HO-1301-159 RCIC Turbine Governor Open Open/Close Modulates steam flow to RCIC turbine [7]
Valve as needed 24 X-202 RCIC Turbine Idle Operating Supplies motive force to RCIC pump [7]
Page A-3
Pilgrim Nuclear Power Station ESEP Report ESEL -Equipment Operating State . -w Item . N. Notes/Comments .,References Number ID Description Normal Desired, State State RCIC Pump Flow Transmits RCIC flow to flow control 25 FT-1360-4 Transmitter for Turbine On On Transit [5][22]
Control 26 SQRT1340-10 RCIC Flow Square Root Energized Energized RCIC flow integrator for RCIC flow [5]
Converter control 27 FIC-1340-1 RCIC Flow Indicating Energized Energized Controls RCIC flow [5]
Controller 28 DC/AC 1340- DC/AC Inverter for RCIC Energized Supplies AC Power to RCIC flow control [5][8]
16 Flow Controller circuitry 29 C1303 RCIC Local Controls Energized Energized Local control of RCIC turbine [15]
30 D4-3 Supply Breaker for RCIC Closed Closed Power to the logic and indications in [6]
in Panels C904 and C939 C904 and C939 31 13A-K1 RCIC Auto Initiation Deenergized Energized Start RCIC on low-low water level (C930) [10]
Logic 32 13A-K1O RCIC Steam Supply Low Deenergized Deenergized Could prevent RCIC operation [10]
Pressure 33 13A-K11 Turbine Trip Auxiliary Deenergized Deenergized Could prevent RCIC operation [10]
Relay 34 13A-K13 Pump Discharge Low Deenergized Deenergized Could prevent RCIC operation [10]
Flow 35 Pump Suction Low Deenergized Deenergized Could prevent RCIC operation [10]
13A-K14 P Pressure 36 13A-K17 Turbine Exhaust High Deenergized Deenergized Could prevent RCIC operation [10]
Pressure 37 13A-K8 MO-1301-25 Position Deenergized Energized Monitors valve position [10]
RCACKAutMonitor n 38 13A-K2 RCIC Auto Initiation Deenergized Energized Start RCIC on low-low water level (C930) [10]
Logic Page A-4
Pilgrim Nuclear Power Station ESEP Report ESEL Equipmient Operating State Item *ii*.
- *:*:**. _ ;. : .. .:':. Normal* Desired - *****i!:j-:.Notes/Comments
,,,;,-***ii\. ;:. :, References..
Number ID DeRscript ionS 39 13A-K22 RCIC Auto Isolation Deenergized Deenergized Could prevent RCIC operation [10]
Relay 40 13A-K3 Pump/Turbine Room Deenergized Deenergized Could prevent RCIC operation [10]
High Temperature 41 13A-K31 Pump/Turbine Room Deenergized Deenergized Could prevent RCIC operation [9]
High Temperature 42 13A-K32 RCIC Valve Station High Deenergized Deenergized Could prevent RCIC operation [9]
Temperature 43 13A-K33 RCIC Steam Line High Deenergized Deenergized Could prevent RCIC operation [9]
Differential Pressure 44 13A-K34 RCIC Auto Isolation Deenergized Deenergized Could prevent RCIC operation [9]
Relay 45 13A-K5 RCIC Valve Station High Deenergized Deenergized Could prevent RCIC operation [10]
Temperature 46 13A-K7 RCIC Steam Line High Deenergized Deenergized Could prevent RCIC operation [10]
Differential Pressure 47 RV-203-3B Safety Relief Valve B Shut Open Used for pressure control and cooldown [11]
48 RV-203-3C Safety Relief Valve C Shut Open Used for pressure control and cooldown [11]
49 SV-203-3B Solenoid Pilot Valve for Deenergized Energized Pilot valve for associated SRV [11][19]
SRV B 50 SV-203-3C Solenoid Pilot Valve for Deenergized Energized Pilot valve for associated SRV [11][19]
SRV C 51 T-221B Accumulator Tank for Operable Operable Holds operating nitrogen for SRV [11]
SRV B 52 T-221C Accumulator Tank for Operable Operable Holds operating nitrogen for SRV [11]
SRVCPae A Page A-5
Pilgrim Nuclear Power Station ESEP Report ESEL . Equipment.. : Operating State..*;, 'No o Item .. *..... . ... :-Notes/Comments . References Number . Normal : .- Desired-"
ID Description.
. .... ", State . . State* - ' _" ' o Pressure Control Valve Controls the pressure supplied to the [11]
53 PCV-203-11 for Backup Nitrogen Idle In service SRV accumulators Supply Used to inhibit automatic initiation of 54 2E-S1OA ADS Inhibit Switch Normal Inhibit Automatic Depressurization System [12]
(ADS) 55 2E-S1B SRV B Valve Actuation Auto Open Open SRV B for cooldown [12]
56 2E-S1C SRV C Valve Actuation Auto Open Open SRV C for cooldown [12][19]
57 2E-K12AInelk CSCS Pump Running Deenergized Deenergized Included for relay chatter impacts [12][20]
Interlock 58 2E-K12B CSCS Pump Running Deenergized Deenergized Included for relay chatter impacts [12][20]
Interlock 59 SRV Power Supply Energized Energized Included for relay chatter impacts [12][20]
2E-K13BControl Logic SRV Power Supply Energized Energized Included for relay chatter impacts [12][16][19]
60 2E-K13C Control Logic [20]
61 2E-K7A Reactor Water Low Low Deenergized Energized Included for relay chatter impacts [12][20]
Level Interlock Low 62 2E-K7B Reactor Water Low Low Deenergized Energized Included for relay chatter impacts [12][20]
Level Interlock 63 LI-263-100A Reactor Water Level Operable Operable Reactor level indication [13][20]
Narrow Range Voltage Current CE/-263-72A Operable Operable Reactor level indication [13][20]
ReaConverter Level 65 LI-263-106A Reactor Water Level Operable Operable Reactor level indication [13][20]
Indicator Page A-6
Pilgrim Nuclear Power Station ESEP Report ESEL Equipment Operating State .
Item Numb.. . Norm ,al .- .. . Des, red .,. .,.. : ., . Notes/Comments,
.. . . , .. . . .... .. .. References Number ID Description State Desired
.tate,. State Voltage Current CE/-263-73A Operable Operable Reactor level indication [13][20]
Converter 67 LIS-263-73A Level Indicating Switch Operable Operable Reactor level indication [131 68 LT-263-73A Level Transmitter Operable Operable Reactor level indication [13][20]
69 LIS-263-72A Reactor Water Level Operable Operable Reactor level indication [13]
Narrow Range 70 LT-263-72A Level Transmitter Operable Operable Reactor level indication [13][20]
71 PI-640-25A Reactor Pressure Operable Operable Reactor pressure indication [13]
72 PT-647A Reactor Pressure Operable Operable Reactor pressure indication [13][20]
Transmitter 73 PI-263-49A Reactor Pressure Operable Operable Reactor pressure indication [13][20]
74 Voltage Current Operable Operable Reactor pressure indication [13][20]
E/l-23-49AConverter 75 PIS-263-49A Reactor Pressure Operable Operable Reactor pressure indication [13]
76 PT-263-49A Reactor Pressure Operable Operable Reactor pressure indication [13][20]
Transmitter 77 C2205A Reactor Protection and Operable Operable Pressure and level transmitter support [13]
C2205A NSS Instrument Rack Emergency Core Cooling 78 C2233A System (ECCS) Analog Operable Operable Pressure and level transmitter support [13][15]
Trip Cabinet 79 C2251A Jet Pump Instrument Operable Operable Pressure and level transmitter support [13][14]
Rack A Page A-7
Pilgrim Nuclear Power Station ESEP Report ESEL Equipment .. Operating State References Item.Notes/Comments Number ID, "dDescription Normal Desired
____S. Sate,;C
- ~State C904 80 RWCU and Recirc. Bench Operable Operable Powered from D4, contains RCIC [5][6][15][27]
Board controls Reactor and [5][6][12][13]
81 C903 Containment Cooling Operable Operable Indications provided here [15][16][24]
Bench Board [25]
82 C905 Reactor Control Bench Operable Operable Indications provided here, powered [6][13]
Board from D6 Containment Pressure 83 C129B Switch Instrument Rack Operable Operable Indications provided here [16][21]
Switchen Insruen forRCcklo 84 C2258 RCIC Instrument Rack Operable Operable Instrument rack for RCIC flow [15][22]
transmitter 85 C930 RCIC Relay Vertical Operable Operable Contains RCIC logic. IPEEE correlated [15][23]
Board failure of C930, C932, C933.
86 C932 Channel A Vertical Operable Operable Contains RCIC logic. IPEEE correlated [12][15][23]
Board failure of C930, C932, C933.
87 C933 Channel B Vertical Operable Operable Contains RCIC logic. IPEEE correlated [15][23]
Board failure of C930, C932, C933.
88 C941 Primary Containment OperableContains isolation logic that could [6]
Isolation Relay Cabinet disable RCIC or HPCI operation 89 PIS-1001-89A Drywell Pressure Indicating Switch Operable Operable Containment pressure indication [16][21]
90 PT-1001-89A Drywell Pressure Operable Operable Containment pressure indication [16][21]
Transmitter Page A-8
Pilgrim Nuclear Power Station ESEP Report ESEL Equipment . Operating-State ., References Item... ........................
. . . Norma~ll. Desired Desired.......,.. Notes/Comments.
Numer ID ;-,Description State - ae a
91 LI-1001-604A Torus Water level Energized Energized Wide range torus water level in Phase 2 [16][28]
Indicator 92 DPT1001 - Torus Water level Energized Energized Wide range torus water level in Phase 2 [16][29]
604A Transmitter 93 Torus Water Bulk Torus bulk water temperature for Phase [16][24][25]
TI-5021-02A Temperature Energized Energized 2, powered from Y31 TRU-5021- Torus Water Processes bulk and local torus water 94 01A Torus Water Energized Energized temperatures for Phase 2, powered [16][25][26]
01A Temperature Recorder from Y31 95 TE5021-01A Torus Water Torus water temperature element for [16][251 Temperature Element Phase 2, powered by Y31 96 TE5021-02A TrsWtrEnergized Energized Trswtrem rauelmntfr [16][25]
97 TE5021-0A Torus Water Torus water temperature element for Temperature Element Energized Energized Phase 2, powered by Y31 [16][25]
97 TE5021-04A Temperture1Eement Torus Water Energized Energized Phae6,]oweedby]3 Torus water temperature element for [16][25]
Temperature Element Phase 2, powered by Y31 99 TE5021-05A Torus Water Torus water temperature element for [16][25]
Temperature Element Phase 2, powered by Y31 100 TE5021-06A Torus Water Torus water temperature element for [16][25]
Temperature Element Energized Energized Phase 2, powered by Y31 Page A-9
Pilgrim Nuclear Power Station ESEP Report ESEL Equipment Oper~ting State'.
Item Normal"" Dired.". * .Notes/Comments,. References R
Number ID Description -4 State ~State TE5021-07A 1 Torus Water Torus water temperature element for 101 ~~Temperature Element Eegzd nried Phase 2, powered by Y31 [6[5 102 TE5021-8A Torus Water EnergizedTorus water temperature element for [16][25]
Temperature Element Phase 2, powered by Y31 103 TE5021-09A Torus Water Energized ETorus water temperature element for [16][25]
Temperature Element Phase 2, powered by Y31 104 TE5021-10A Torus Water Energized Energized Torus water temperature element for [16][25]
Temperature Element Phase 2, powered by Y31 105 TE5021-11A Torus Water Energized Energized Torus water temperature element for [16][25]
Temperature Element Phase 2, powered by Y31 106 TE5021-12A Torus Water Energized ETorus water temperature element for [16][25]
Temperature Element Phase 2, powered by Y31 107 TE5021-13A Torus Water Torus water temperature element for [16][25]
Temperature Element Energized Energized Phase 2, powered by Y31 Containment Isolation Included because the controls for the 108 C7 and Ventilation Vertical Operable Operable direct torus vent path are located on [27][30]
Board this panel 109 C170 PAM panel Operable Operable Indications provided here [14][16][26]
Page A-10
Pilgrim Nuclear Power Station ESEP Report ESELqui,:..entOperating State Item.. *,, Notes/Comments References.
Number ,ID Description PASS StatioeVlv 110 C174 PASS Isolation Valve Operable Operable Indications provided here [6]
Control Panel Torus Water Torus temperature signals processed 111 C179 Temperature Signal Operable Operable here [25][32]
Processing Cabinet Inboard valve for hardened containment 112 AO-5042B Isolation Valve (inboard) Closed Open venting system. Powered from 125 VDC [17]
bus D5.
Torus Purge Exhaust 113 SV-5042B Isolation Solenoid Valve Deenergized Energized Solenoid valve for AO-5042B for DTV [17][31]
(inboard) 114 D5-15 DC Breaker Supply to C7 Energized Energized Power supply for solenoid valve for AO- [6][31]
5042B for DTV Locked Open to provide flow path for 115 AO-5025 Direct Torus Vent Closed Open containment heat removal during Phase [17][27]
- 2. Powered from 125VDC Bus D4.
116 SV-5025 Direct Torus Vent Deenergized Energized Solenoid valve for AO-5025 for DTV [17][27]
Solenoid Valve 117 D4-15 DC Breaker Supply to C7 Energized Energized Power supply for solenoid valve for AO- [6][27]
5025 for DTV N2-401A Backup Nitrogen Bottles N2-402A for operation of AO- TBD TBD Local backup nitrogen for direct torus 118 N2-401BA vent valves [3]
N2-402B 5042B and AO-5025 119 D7 125VDC Motor Control Energized Energized Power for RCIC valves [6][15]
Center for RCIC 120 72-166 Supply for D7 Closed Closed Power for RCIC valves [6]
121 D36 Extension of 125 VDC Energized Energized Power for analog trip system [6]
PanelA Page A-li
Pilgrim Nuclear Power Station ESEP Report ESEL E~quipment Operating State Item Noml~ DsrdNotes/Comments References Number ID.. : Description . .tate.
.' .State * . .*-.. State :! ,=." .,.**. .. . .. * *,,, ,State ECCS Analog Trip 122 D36-8 Cabinet C2233A Power Closed Closed Power for analog trip system [6]
Breaker 123 D4 125 VDC Distribution Energized Energized 125 DC Bus A for CSCS logic [6]
PanelA 124 72-165 Supply for D4 Closed Closed Power for D4 [6]
125 72-16A D16 Internal Breaker Closed Closed Bus continuity [6]
126 D16 125 VDC Bus A Energized Energized 125 DC Power Bus A [6]
127 72-161 Battery A Output Closed Closed Supply power to D16 [6]
Breaker 128 D29 Battery A Current Intact Intact Protective device [6]
Limiter 129 D1 125 VDC Battery Rack A Float Charge Discharge 125 VDC A power source [6]
130 72-162 Battery Charger Dl1 Closed Closed Connection for portable generator [3][6]
Supply to D16 131 D37 Extension of 125 VDC Energized Energized Power for analog trip system [6]
Panel B 132 D5 125 VDC Distribution Energized Energized 125 DC Bus B for CSCS logic [6]
PanelB 133 72-175 Supply Breaker for Bus Closed Closed Power for D5 [6]
D5 134 D17 125 VDC Bus B Energized Energized 125 VDC power Bus B [6]
135 72-17A D17 Internal Breaker Closed Closed Bus continuity [6]
Page A-12
Pilgrim Nuclear Power Station ESEP Report ES EL . Eq u pme nt . , , O pe r a tin g St a t e .. ' .. References Item " Notes/Comments References Number ID .Description . Noml Dsirted' State,< 4 Stt 136 D2 125 VDC Battery B Float Charge Discharge 125 VDC B power source [6]
137 72-171 Battery B Output Closed Closed Supply power to D17 [6]
Breaker 138 D30 Battery B Current Intact Intact Protective device [6]
Limiter 139 72-172 Battery Charger Supply to D17D12 Closed Closed Connection for portable generator [3][6]
140 Y2 Vital Services Power Energized Energized Vital bus for Indication and control [18]
Supply 141 Y12 Auto Transfer Switch for MG Set MG Set Power from vital motor-generator set to [18]
Y2 Supply Supply Y2 Vital Motor-Generator 142 EG-23 Set Operating Operating Supply power to Y2 and requires D6 [18]
143 72-1022 D10 Breaker to Vital Closed Closed Supply power to DC motor - Vital motor- [6][18]
Motor-Generator generator set 144 D10 250 VDC Power Bus Energized Energized 250 VDC power bus [6][18]
145 72-1013 250 VDC Battery D3 Closed Closed Supply power to D10 [6]
Output Breaker 146 D31 250 VDC Battery D3 Intact Intact Protective Device [6]
Current Limiter 147 D3 250 VDC Battery Float Charge Discharge 250 VDC power Source [6]
148 72-1014 Battery Charger D13 Closed Closed Connection for portable generator [6]
Supply to D10 Energized 149250 VDC Battery Energized from mobile Connection for portable generator [6]
generator Page A-13
Pilgrim Nuclear Power Station ESEP Report ESEL Equipment " -, Operating State ".
Item .. Notes/Comments , References,.:
Number ID Description State state 150 D6 125VDC Distribution Energized Energized Vital instruments and controls [6]
PanelC 151 Y1O 125VDC Control Power Operable Operable D6 normally supplied from D16 [6]
Transfer 152 D32 D16 Control Logic Y10 Switching Closed Closed D6 normally supplied from D16 [6]
D17 Control Logic Y10 153 D33 Switching Open Open D6 normally supplied from D16 [6]
Safeguard 120VAC "A" 154 Y3 Control Power Supply Energized Energized Power supply for I&C [18]
Panel Safeguard 120VAC "A" 155 Y31 H2 0 2 Control Power Energized Energized Power supply for I&C [18]
Supply Panel Energized 125 VDC Battery 156 Dl 2Chargery Energized from mobile Connection for portable generator [6]
generator Energized 157 D12 125 VDC Battery Energized from mobile Connection for portable generator [6]
Crrgenerator 158 C2257B Instrument Rack 2257B Operable Operable [5][34]
159 dPIS-1360-1A dP Switch for RCIC Operable Operable [5][34]
Steam Isolation 160 dPIS-1360-1B dP Switch for RCIC Operable Operable [5][34]
Steam Isolation 161 PS-1360-9A Pressure Switch for RCIC Operable Operable [5][341 Steam Isolation 7162 PS-1360-9B3 Pressure Switch for RCIC Operable Operable [5][34]
Steam Isolation Page A-14
Pilgrim Nuclear Power Station ESEP Report ESEL .Equipment . Operating State -.
Item. Notes/Comments References Number ID Description Normal .. *,.Desired
.State *.: .. ', State 163 PS-1360-9C Pressure Switch for RCIC Operable Operable [5][34]
Steam Isolation Oeal prbe[]3 164 PS-1360-9D) Pressure Switch for RCIC Operable Operable [5][34]
Steam Isolation OperableOperable_[5][34]
165 J315 Junction Box Operable Operable [10]
166 J317 Junction Box Operable Operable [9]
167 J602 Junction Box Operable Operable [9]
168 J599 Junction Box Operable Operable [10]
169 J600 Junction Box Operable Operable [10]
170 J601 Junction Box Operable Operable [91 171 TS-1360-14C Temperature Switch for Operable Operable [10]
RCIC Isolation 172 TS-1360-15A Temperature Switch for Operable Operable [10]
RCIC Isolation 173 TS-1360-15C Temperature Switch for Operable Operable [10]
RCIC Isolation 174 TS-1360-16C Temperature Switch for Operable Operable [10]
RCIC Isolation 175 TS-1360-16D Temperature Switch for Operable Operable [9]
RCIC Isolation Oprbeprbl__
176 TS-1360-17A Temperature Switch for Operable Operable [10]
RCIC Isolation OpeabeOerbl [_0]
Page A-15
Pilgrim Nuclear Power Station ESEP Report ESEL Equipment Operating'State . . ,
Item Number .
.Normal . "".:Notes/Comments
- , ,/:Desired -' , References' uIDr Description, :.< ,ta,,.
State State 177 TS-1360-17B Temperature Switch for Operable Operable 19]
RCIC Isolation 178 TS-1360-17C Temperature Switch for Operable Operable [10]
RCIC Isolation 179 TS-1360-17D Temperature Switch for Operable Operable 19]
RCIC Isolation 180 TS-1360-15B Temperature Switch for Operable Operable 19]
RCIC Isolation 181 TS-1360-14D Temperature Switch for Operable Operable 19]
RCIC Isolation 182 TS-1360-15D Temperature Switch for Operable Operable 19]
RCIC Isolation FLEX AC Power Transfer 183 D1513 Swtc Switch to PowerTaf D11 to Repower Operable Operable [3][6][35][36]
184 D1413 FLEX AC Power Transfer Operable Operable [3][35][36]
Switch to Repower D12 185 D1414B FLEX AC Power Transfer Operable Operable [3][6][35][36]
Switch to Repower D13 186 N17115* FLEX AC Power Transfer Operable Operable [3][36]
Switch to Repower Y3 FLEX AC Power Transfer Switch to Repower Y31 187 N17115* *Note that a single Operable Operable [3][36]
switch, N17115, repowers Y3 and Y31 Page A-16
Pilgrim Nuclear Power Station ESEP Report ATTACHMENT B - ESEP HCLPF VALUES AND FAILURE MODES TABULATION Page B-1
Pilgrim Nuclear Power Station ESEP Report Item HCP g " Failure.
No.. Equipment ID Equipment Description Screening . Mode Comments
. . .... Level 1 MO-1301-22 RCIC Pump CST Suction Valve >RLGM Screened 2 D754 DC Bus D7 Supply Breaker to >RLGM Screened MO-1301-22 3 TBD FLEX Primary External Water >RLGM Screened Note 5 Source Injection Point 4 MO-1301-25 RCIC Suction from Torus >RLGM Screened 5 D761 DC Bus D7 supply breaker to >RLGM Screened MO-1301-25 6 MO-1301-26 RCIC Suction from Torus >RLGM Screened 7 D764 DC Bus D7 supply breaker to >RLGM Screened MO-1301-26 8 P-206 RCIC Pump >RLGM Screened Note 1 9 MO-1301-49 RCIC Discharge Isolation >RLGM Screened Valve 10 D774 DC Bus D7 Supply Breaker to >RLGM Screened MO-1301-49 11 PCV 1301-43 RCIC Lube Oil Cooling Water >RLGM Screened Pressure Control Valve 12 MO-1301-62 RCIC Turbine Lube Oil Inlet >RLGM Screened Valve 13 D794 DC Bus D7 Supply Breaker to >RLGM Screened MO-1301-62 14 E-201 RCIC Barometric Condenser >RLGM Screened Note 1 and Vacuum Tank 15 E-204 RCIC Lube Oil Cooler >RLGM Screened Note 1 RCIC Vacuum Tank 16 P-221 >RLGM Screened Note 1 Condensate Pump 17 D712 DC Bus D7 Supply Breaker to >RLGM Screened P-221 18 P-222 RCIC Vacuum Pump >RLGM Screened Note 1 19 D714 DC Bus D7 Supply Breaker to >RLGM Screened P-222 20 MO-1301-61 RCIC Turbine Steam Inlet >RLGM Screened Valve Page B-2
Pilgrim Nuclear Power Station ESEP Report HCLPF (g)/ Falr, Cmet i.ia lu re ':,.::.: - .. *. ... .:. . - . : .
Fa.
Ie m Ite .
Equipment ID: ... ....
. i.Equipment Description Screening ." .:
S Comments No. Level 21 D751 DC Bus D7 Supply Breaker to >RLGM Screened MO-1301-61 22 SV-1301-1 RCIC Turbine Trip Throttle >RLGM Screened Valve 23 HO-1301-159 RCIC Turbine Governor Valve >RLGM Screened 24 X-202 RCIC Turbine >RLGM Screened Note 1 25 FT-1360-4 RCIC Pump Flow Transmitter >RLGM Screened for Turbine Control 26 SQRT1340-10 RCIC Flow Square Root >RLGM Screened Converter 27 FIC-1340-1 RCIC Flow Indicating >RLGM Screened Controller 28 DC/AC 1340-16 DC/AC Inverter for RCIC Flow >RLGM Screened Controller 29 C1303 RCIC Local Controls >RLGM Screened Note 2 30 D4-3 Supply Breaker for RCIC in >RLGM Screened Panels C904 and C939 31 13A-K1 RCIC Auto Initiation Logic >RLGM Screened RCIC Steam Supply Low Relay Pressure Function 33 13A-K11 Turbine Trip Auxiliary Relay >RLGM Screened 34 13A-K13 Pump Discharge Low Flow >RLGM Screened 35 13A-K14 Pump Suction Low Pressure >RLGM Screened 36 13A-K17 Turbine Exhaust High >RLGM Screened Pressure 37 13A-K18 MO-1301-25 Position >RLGM Screened Monitor 38 13A-K2 RCIC Auto Initiation Logic >RLGM Screened Relay 13A-K22 RCIC Auto Isolation Relay 0.33 Function 39 40 13A-K3 Pump/Turbine Room High 0.39 Relay Temperature Function Page B-3
Pilgrim Nuclear Power Station ESEP Report HCLPF (g alreCmet Nte. Equipment ID Equipmen.t Description Screening Modent Pump/Turbine Room High 0.39 Relay 41 13A-K31 Temperature Function 42 13A-K32 RCIC Valve Station High 0.39 Relay Temperature Function RCIC Steam Line High Relay Differential Pressure Function 44 13A-K34 RCIC Auto Isolation Relay >RLGM Screened 45 13A-K5 RCIC Valve Station High 0.39 Relay Temperature Function RCIC Steam Line High Relay Differential Pressure Function 47 RV-203-3B Safety Relief Valve B >RLGM Screened 48 RV-203-3C Safety Relief Valve C >RLGM Screened 49 SV-203-3B Solenoid Pilot Valve for SRV B >RLGM Screened 50 SV-203-3C Solenoid Pilot Valve for SRV C >RLGM Screened 51 T-221B Accumulator Tank for SRV B >RLGM Screened Note 2 52 T-221C Accumulator Tank for SRV C >RLGM Screened Note 2 53 PCV-203-11 Pressure Control Valve for >RLGM Screened Backup Nitrogen Supply 54 2E-SlOA ADS Inhibit Switch >RLGM Screened 55 2E-S1B SRV B Valve Actuation >RLGM Screened 56 2E-S1C SRV C Valve Actuation >RLGM Screened 57 2E-K12A CSCS Pump Running Interlock >RLGM Screened 58 2E-K12B CSCS Pump Running Interlock >RLGM Screened 59 2E-K13B SRV Power Supply Control >RLGM Screened Logic 60 2E-K13C SRV Power Supply Control >RLGM Screened Logic Page B-4
Pilgrim Nuclear Power Station ESEP Report Item Eqimn .HLF()/ Failure Cmet S Equipment D..: Equipment Description Screening. Momnodts Level 61 2E-K7A Reactor Water Low Low Level >RLGM Screened Interlock 62 2E-K7B Reactor Water Low Low Level >RLGM Screened Interlock Reactor Water Level Narrow 63 LI-263-100A Range Range >RLGM Screened 64 E/I-263-72A Voltage Current Converter >RLGM Screened 65 LI-263-106A Reactor Water Level >RLGM Screened Indicator 66 E/1-263-73A Voltage Current Converter >RLGM Screened 67 LIS-263-73A Level Indicating Switch >RLGM Screened 68 LT-263-73A Level Transmitter >RLGM Screened 69 LIS-263-72A Reactor Water Level Narrow >RLGM Screened Range 70 LT-263-72A Level Transmitter >RLGM Screened 71 PI-640-25A Reactor Pressure >RLGM Screened 72 PT-647A Reactor Pressure Transmitter >RLGM Screened 73 PI-263-49A Reactor Pressure >RLGM Screened 74 E/I-263-49A Voltage Current Converter >RLGM Screened 75 PIS-263-49A Reactor Pressure >RLGM Screened 76 PT-263-49A Reactor Pressure Transmitter >RLGM Screened 77 C2205A Reactor Protection and NSS 0.38 Functional Instrument Rack Emergency Core Cooling 78 C2233A System (ECCS) Analog Trip >RLGM Screened Note 1 Cabinet 79 C2251A Jet Pump Instrument Rack A 0.44 Anchorage 80 C904 RWCU and Recirc. Bench >RLGM Screened Note 1 Board Page B-5
Pilgrim Nuclear Power Station ESEP Report Item HCLPF (g)/ Failure Comments Nte No. Equipment ID.. .. Equipment Description Screening
.node Modeomme M*.
- Level 81 C903 Reactor and Containment >RLGM Screened Note 1 Cooling Bench Board 82 C905 Reactor Control Bench Board >RLGM Screened Note 1 Containment Pressure Switch 83 C129B Intrument rack Instrument Rack >RLGM Screened Note 1 84 C2258 RCIC Instrument Rack >RLGM Screened Note 1 Relay Nt 85 C930 RCIC Relay Vertical Board 0.33 Function ote 1 86 C932 Channel A Vertical Board >RLGM Screened Note 1 87 C933 Channel B Vertical Board 0.39 Relay Function Note 1 88 C941 Primary Containment >RLGM Screened Note I Isolation Relay Cabinet 89 PIS-1001-89A Drywell Pressure Indicating >RLGM Screened Switch 90 PT-1001-89A Drywell Pressure Transmitter >RLGM Screened 91 LI-1001-604A Torus Water level Indicator >RLGM Screened 92 DPT1001 -604A Torus Water level >RLGM Screened Note 2 Transmitter 93 TI-5021-02A Torus Water Bulk >RLGM Screened Temperature 94 TRU-5021-01A Torus Water Temperature >RLGM Screened Recorder 95 TE5021-01A Torus Water Temperature >RLGM Screened Element 96 TE5021-02A Torus Water Temperature >RLGM Screened Element 97 TES021-03A Torus Water Temperature >RLGM Screened Element 98 TE5021-04A Torus Water Temperature >RLGM Screened Element 99 TE5021-05A Torus Water Temperature >RLGM Screened Element 100 TE5021-06A Torus Water Temperature >RLGM Screened Element Page B-6
Pilgrim Nuclear Power Station ESEP Report Item HLF()/ Failure N. .Equipment ID... Equipment Description Screening M Comments Level 101 TE5021-07A Torus Water Temperature >RLGM Screened Element 102 TE5021-08A Torus Water Temperature >RLGM Screened Element 103 TE5021-09A Torus Water Temperature >RLGM Screened Element 104 TE5021-10A Torus Water Temperature >RLGM Screened Element 105 TE5021-11A Torus Water Temperature >RLGM Screened Element 106 TE5021-12A Torus Water Temperature >RLGM Screened Element 107 TE5021-13A Torus Water Temperature >RLGM Screened Element 108 C7 Containment Isolation and 0.31 Anchorage Ventilation Vertical Board 109 C170 PAM panel >RLGM Screened Note 1 110 C174 PASS Isolation Valve Control >RLGM Screened Note 1 Panel 111 C179 Torus Water Temperature >RLGM Screened Note 1 Signal Processing Cabinet 112 AO-5042B Torus Purge Exhaust Isolation >RLGM Screened Valve (inboard) 113 SV-5042B Torus Purge Exhaust Isolation >RLGM Screened Solenoid Valve (inboard) 114 D5-15 DC Breaker Supply to C7 >RLGM Screened 115 AO-5025 Direct Torus Vent >RLGM Screened 116 SV-5025 Direct Torus Vent Solenoid Vle>RLGM Screened Valve 117 D4-15 DC Breaker Supply to C7 >RLGM Screened N2-401A Backup Nitrogen Bottles for 118 N2-402A operation of AO-5042B and >RLGM Screened Note 6 N2-402B AO-5025 119 D7 125VDC Motor Control >RLGM Screened Note 1 Center for RCIC 120 72-166 Supply for D7 >RLGM Screened Page B-7
Pilgrim Nuclear Power Station ESEP Report Item. Equipment ID . . .P. .g . Failure Comments
. EquipmentD Equipment
Description:
... Screening e ComentsM.od Level.
121 D36 Extension of 125 VDC Panel A >RLGM Screened Note 2 122 D36-8 ECCS Analog Trip Cabinet >RLGM Screened C2233A Power Breaker 123 D4 125 VDC Distribution Panel A >RLGM Screened Note 2 124 72-165 Supply for D4 >RLGM Screened 125 72-16A D16 Internal Breaker >RLGM Screened 126 D16 125 VDC Bus A 0.37 Anchorage 127 72-161 Battery A Output Breaker >RLGM Screened 128 D29 Battery A Current Limiter >RLGM Screened Note 2 129 D1 125 VDC Battery Rack A 0.31 Anchorage 130 72-162 Battery Charger Dl1 Supply >RLGM Screened to D16 131 D37 Extension of 125 VDC Panel B >RLGM Screened Note 2 132 D5 125 VDC Distribution Panel B >RLGM Screened Note 2 133 72-175 Supply Breaker for Bus D5 >RLGM Screened 134 D17 125 VDC Bus B >RLGM Screened Note 1 135 72-17A D17 Internal Breaker >RLGM Screened 136 D2 125 VDC Battery B 0.30 Anchorage 137 72-171 Battery B Output Breaker >RLGM Screened 138 D30 Battery B Current Limiter >RLGM Screened Note 2 139 72-172 Battery Charger D12 Supply >RLGM Screened to D17 140 Y2 Vital Services Power Supply >RLGM Screened Note 2 141 Y12 Auto Transfer Switch for Y2 >RLGM Screened Note 1 Page B-8
Pilgrim Nuclear Power Station ESEP Report Item Eqimn DHLF()/ Failure Cmet
. Equipment ID Equipment Description Screening Comments Level HCLPF calculated 142 EG-23 Vital Motor-Generator Set 0.32 Anchorage with modifications to anchorage.
143 72-1022 D10 Breaker to Vital Motor- >RLGM Screened Generator 144 DiO 250 VDC Power Bus >RLGM Screened Note 1 145 72-1013 250 VDC Battery D3 Output >RLGM Screened Breaker 146 D31 250 VDC Battery D3 Current >RLGM Screened Note 2 Limiter 147 D3 250 VDC Battery 0.40 Anchorage 148 72-1014 Battery Charger D13 Supply >RLGM Screened to DiO 149 D13 250 VDC Battery Charger >RLGM Screened Note 1 150 D6 125VDC Distribution Panel C >RLGM Screened Note 2 151 Y Transfer Power >RLGM Screened Note 1 152 D32 D16 Control Logic Y10 >RLGM Screened Note 2 Switching 153 D33 D17 Control Logic Y10 >RLGM Screened Note 2 Switching 154 Y3 Safeguard 120VAC "A" >RLGM Screened Note 1 Control Power Supply Panel 155 Y31 Safeguard 120VAC "A" H202 >RLGM Screened Note 2 Control Power Supply Panel 156 D1 125 VDC Battery Charger A >RLGM Screened Note 1 157 D12 125 VDC Battery Charger B >RLGM Screened Note 1 158 C2257B Instrument Rack 2257B >RLGM Screened Note 1 159 dPIS-1360-1A dP Switch for RCIC Steam >RLGM Screened Isolation 160 dPIS-1360-1B dP Switch for RCIC Steam >RLGM Screened Isolation 161 PS-1360-9A Pressure Switch for RCIC >RLGM Screened Steam Isolation Page B-9
Pilgrim Nuclear Power Station ESEP Report IHCLPF (g)/ -
No. Equipment ID Equipment Description screenings Failure Comments Item.. Level' Md 162 PS-1360-9B3 Pressure Switch for RCIC >RLGM Screened Steam Isolation 163 PS-1360-9C Pressure Switch for RCIC >RLGM Screened Steam Isolation 164 PS-1360-9B Pressure Switch for RCIC >RLGM Screened Steam Isolation 165 J315 Junction Box >RLGMV Screened Note 2 166 J317 Junction Box >RLGMV Screened Note 2 167 J602 Junction Box >RLGM Screened Note 1 168 J599 Junction Box >RLGM Screened Note 2 169 J600 Junction Box >RLGM Screened Note 1 170 J601 Junction Box >RLGM Screened Note 2 171 TSJ136004C Temperature Switch for RCIC >RLGM Screened Isolation 172 TS-1360-15A Temperature Switch for RCIC >RLGM Screened Isolation 173 TS-1360-15C Temperature Switch for RCIC >RLGM Screened Isolation 174 TS-1360-16C Temperature Switch for RCIC >RLGM Screened Isolation 175 TS-1360-16D Temperature Switch for RCIC >RLGM Screened Isolation 176 TS-1360-17A Temperature Switch for RCIC >RLGM Screened Isolation 177 TS-1360-17B Temperature Switch for RCIC >RLGM Screened Isolation 178 TS-1360-17C Temperature Switch for RCIC >RLGM Screened Isolation 179 TS-1360-17D Temperature Switch for RCIC >RLGM Screened Isolation 180 TS-1360-15B Temperature Switch for RCIC >RLGM Screened Isolation 181 TS-1360-14D Temperature Switch for RCIC >RLGM Screened Isolation Page B-10
Pilgrim Nuclear Power Station ESEP Report Item HCLPF (g)/ Failure Comments N. Equipment ID Equipment Description Screening Comments
. ......... Level 182 TS-1360-15D Temperature Switch for RCIC >RLGM Screened Isolation 183 D1513 FLEX AC Power Transfer >RLGM Screened Note 3 Switch to Repower D11 FLEX AC Power Transfer 184 D1413 FLEX C IPower Taf >RLGM Screened Note 3 Switch to Repower D12 185 D1414B FLEX AC Power Transfer >RLGM Screened Note 3 Switch to Repower D13 FLEX AC Power Transfer 186 N17115* FLEX ACto Repower Switch PowerT Y3 s >RLGM Screened Note 4 FLEX AC Power Transfer Switch to Repower Y31 187 N17115* *Note that a single switch, >RLGM Screened Note 4 N17115, repowers Y3 and Y31 Notes:
- 1. Anchorage screened out based on available margin during walkdown by SRT.
- 2. Anchorage screened out during walkdown validation by SRT.
- 3. This component is evaluated in Entergy Calculation C15.0.3623 [48] for 2xSSE.
- 4. This component is evaluated in Entergy Calculation C15.0.3624 [49] for 2xSSE.
- 5. Entergy document no. EC-0000042259 [36] is reviewed. Design input 59 within this document states that seismic demand is based on 2xSSE. Thus, this component, which is yet to be installed, is screened for ESEP based on seismic design criteria.
- 6. This component is evaluated in Entergy Calculation C15.0.3631 [50] for 2xSSE.
Page B-11
ATTACHMENT 2 to PNPS Letter 2.14.082 LIST OF REGULATORY COMMITMENTS FOR PILGRIM NUCLEAR POWER STATION to PNPS Letter 2.14.082 Page 1 of 1 List of Regulatory Commitments The following table identifies those actions committed to by Entergy in this document. Any other statements in this submittal are provided for information purposes and are not considered to be regulatory commitments.
TYPE (Check One) SCHEDULED COMMITMENT ONE- COMPLETION DATE ONE CONTINUING (If Required)
TIME ACTION COMPLIANCE Modify Vital MG Set EG-23 anchorage.
Vital MG Set EG-23 anchorage had a *On a schedule High Confidence of a Low Probability of specified in Section Failure (HCLPF) capacity below the 8.4 of the Expedited Review Level Ground Motion (RLGM). A [] Seismic Evaluation modification is planned to provide Process Report additional seismic margin such that the HCLPF will exceed RLGM. May 31, 2017 NRC Commitment No. A16866 Submit a letter to NRC summarizing the Within 60 days HCLPF results of item 1 above confirming following completion implementation of the plant modification of ESEP activities, associated with item 1. including item 1 above.
NRC Commitment No. A16867 July 30, 2017
- 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 to the Vital MG-Set EG-23 will be done during a planned refueling outage.
'. Entergy Entergy Nuclear Operations, Inc.
Pilgrim Nuclear Power Station 600 Rocky Hill Road Plymouth, MA 02360 John A. Dent, Jr.
Site Vice President December 16, 2014 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk 11555 Rockville Pike Rockville, MD 20852
SUBJECT:
Pilgrim's Expedited Seismic Evaluation Process Report (CEUS Sites), Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident Pilgrim Nuclear Power Station Docket No. 50-293 License No. DPR-35 LETTER NUMBER 2.14.082
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 (ML12053A340)
- 2. NEI Letter to NRC, "Proposed Path Forward for NTTF Recommendation 2.1: Seismic Reevaluations", dated April 9, 2013 (ML13101A345)
- 3. NRC Letter, "Electric Power Research Institute Final Draft Report XXXXXX, Seismic Evaluation Guidance: Augmented Approach for the Resolution of 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 (MLI13106A331)
Dear Sir or Madam:
On March 12, 2012, the Nuclear Regulatory Commission (NRC) issued Reference 1 to all power reactor licensees and holders of construction permits in active or deferred status. Enclosure 1 of Reference 1 requested each addressee located in the Central and Eastern United.States (CEUS) to submit a Seismic Hazard Evaluation within 1.5 years from the date of Reference 1.
PNPS Letter 2.14.082 Page 2 of 3 In Reference 2, the Nuclear Energy Institute (NEI) requested NRC agreement to delay submittal of the final CEUS Seismic Hazard 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.
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 Pilgrim Nuclear Power Station (Attachment 1) provides the information described in Section 7 of Reference 3 in accordance with the schedule identified in Reference 2.
This letter contains new regulatory commitments as shown in Attachment 2.
Should you have any questions concerning the content of this letter, please contact Mr. Everett (Chip) Perkins Jr. at (508) 830-8323.
I declare under penalty of perjury that the foregoing is true and correct. Executed on December 16, 2014.
Sincerely, Jon. Dent Jr.
Site /ice President JAD/rmb
Attachment:
1] Expedited Seismic Evaluation Process Report for Pilgrim Nuclear Power Station 2] List of Regulatory Commitments for Pilgrim Nuclear Power Station
PNPS Letter 2.14.082 Page 3 of 3 cc: Mr. Daniel H. Dorman Regional Administrator, Region 1 U.S. Nuclear Regulatory Commission 2100 Renaissance Boulevard, Suite 100 King of Prussia, PA 19406-1415 U. S. Nuclear Regulatory Commission Director, Office of Nuclear Reactor Regulation One White Flint North 11555 Rockville Pike Rockville, MD 20852 Ms. Nadiyah Morgan, Project Manager Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Mail Stop O-8C2A Washington, DC 20555 Mr. John Giarrusso Jr.
Planning, Preparedness & Nuclear Section Chief Mass. Emergency Management Agency 400 Worcester Road Framingham, MA 01702 U. S. Nuclear Regulatory Commission ATTN: Robert J. Fretz Jr.
Mail Stop OWFN/4A15A 11555 Rockville Pike Rockville, MD 20852-2378 U. S. Nuclear Regulatory Commission ATTN: Robert L. Dennig Mail Stop OWFN/10E1 11555 Rockville Pike Rockville, MD 20852-2378 NRC Senior Resident Inspector Pilgrim Nuclear Power Station
ATTACHMENT 1 to PNPS Letter 2.14.082 EXPEDITED SEISMIC EVALUATION PROCESS REPORT FOR PILGRIM NUCLEAR POWER STATION
EXPEDITED SEISMIC EVALUATION PROCESS (ESEP) REPORT FOR PILGRIM NUCLEAR POWER STATION (PNPS)
Page 1
Pilgrim Nuclear Power Station ESEP Report Table of Contents Page LIST OF TABLES ....................................................................................................................................... 4 LIST OF FIGURES ..................................................................................................................................... 5 1.0 PURPOSE AND OBJECTIVE ............................................................................................... ..... 6 2.0 BRIEF
SUMMARY
OF THE FLEX SEISMIC IMPLEMENTATION STRATEGIES ................................. 6 3.0 EQUIPM ENT SELECTION PROCESS AND ESEL ........................................................................... 7 3.1 Equipm ent Selection Process and ESEL ....................................................................... 7 3.1.1 ESEL Development ......................................................................................... 8 3.1.2 Power Operated Valves ................................................................................ 9 3.1.3 Pull Boxes .................................................................................................. 9 3.1.4 Term ination Cabinets .................................................................................... 9 3.1.5 Critical Instrum entation Indicators ............................................................. 10 3.1.6 Phase 2 and 3 Piping Connections .............................................................. 10 3.2 Justification for Use of Equipment That is Not the Primary Means for FLEX Im plem entation ............................................................................................................ 10 4.0 GROUND MOTION RESPONSE SPECTRUM (GM RS) .............................................................. 10 4.1 Plot of GM RS Subm itted by the Licensee .................................................................. 10 4.2 Com parison to SSE ........................................................................................................ 12 5.0 REVIEW LEVEL GROUND M OTION (RLGM ) ........................................................................... 13 5.1 Description of RLGM Selected .................................................................................... 13 5.2 Method to Estimate In-Structure Response Spectra (ISRS) ....................................... 15 6.0 SEISM IC MARGIN EVALUATION APPROACH ......................................................................... 15 6.1 Sum m ary of Methodologies Used ............................................................................. 16 6.2 HCLPF Screening Process .......................................................................................... 16 6.3 Seism ic W alkdown Approach .................................................................................... 17 6.3.1 W alkdown Approach .................................................................................. 17 6.3.2 Application of Previous W alkdown Information .......................................... 18 6.3.3 Significant W alkdown Findings ................................................................... 18 6.4 HCLPF Calculation Process ........................................................................................ 18 6.5 Functional Evaluations of Relays ................................................................................ 19 6.6 Tabulated ESEL HCLPF Values (Including Key Failure Modes) ..................................... 19 7.0 INACCESSIBLE ITEM S ................................................................................................................ 19 7.1 Identification of ESEL Item Inaccessible for W alkdowns ............................................ 19 7.2 Planned W alkdown / Evaluation Schedule / Close Out .............................................. 20 8.0 ESEP CONCLUSIONS AND RESULTS ...................................................................................... 20 8.1 Supporting Inform ation ................................................................................................ 20 Page 2
Pilgrim Nuclear Power Station ESEP Report Table of Contents (continued)
Page 8.2 Identification of Planned Modifications ..................................................................... 22 8.3 Modification Implementation Schedule ..................................................................... 22 8.4 Summary of Regulatory Commitments ..................................................................... 22 9 .0 REFER ENCES ............................................................................................................................. 22 ATTA CH M EN T A - PN PS ESEL .............................................................................................................. A-1 ATTACHMENT B - ESEP HCLPF VALUES AND FAILURE MODES TABULATION ................................... B-1 Page 3
Pilgrim Nuclear Power Station ESEP Report List of Tables Page TABLE 4-1: GM RS FOR PNPS ................................................................................................................ 10 TABLE 4-2: SSE FOR PNPS .................................................................................................................... 12 TABLE 5-1: RLGM FOR PNPS ................................................................................................................ 14 Page 4
Pilgrim Nuclear Power Station ESEP Report List of Figures Page FIGURE 4-1: GM RS FOR PNPS .............................................................................................................. 12 FIGURE 4-2: GM RS TO SSE COM PARISON FOR PNPS ........................................................................ 13 FIGURE 5-1: RLGM FOR PNPS .............................................................................................................. 15 Page 5
Pilgrim Nuclear Power Station ESEP Report 1.0 PURPOSE AND OBJECTIVE Following the accident at the Fukushima Dai-ichi nuclear power plant resulting from the March 11, 2011, Great Tohoku Earthquake and subsequent tsunami, the Nuclear Regulatory Commission (NRC) established a Near-Term Task Force (NTTF) to conduct a systematic review of NRC processes and regulations and to determine if the agency should make additional improvements to its regulatory system. The NTTF developed a set of recommendations intended to clarify and strengthen the regulatory framework for protection against natural phenomena. Subsequently, the NRC issued a 50.54(f) letter on March 12, 2012 [1], requesting information to assure that these recommendations are addressed by all U.S. nuclear power plants. The 50.54(f) letter requests that licensees and holders of construction permits under 10 CFR Part 50 reevaluate the seismic hazards at their sites against present-day NRC requirements and guidance. Depending on the comparison between the reevaluated seismic hazard and the current design basis, further risk assessment may be required. Assessment approaches acceptable to the staff include a seismic probabilistic risk assessment (SPRA), or a seismic margin assessment (SMA). Based upon the assessment results, the NRC staff will determine whether additional regulatory actions are necessary.
This report describes the Expedited Seismic Evaluation Process (ESEP) undertaken for Pilgrim Nuclear Power Station (PNPS). The intent of the ESEP is to perform an interim action in response to the NRC's 50.54(f) letter to demonstrate seismic margin through a review of a subset of the plant equipment that can be relied upon to protect the reactor core following beyond design basis seismic events.
The ESEP is implemented using the methodologies in the NRC endorsed guidance in Electric Power Research Institute (EPRI) 3002000704, Seismic Evaluation Guidance: Augmented Approach for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1: Seismic [2].
The objective of this report is to provide summary information describing the ESEP evaluations and results. The level of detail provided in the report is intended to enable the NRC to understand the inputs used, the evaluations performed, and the decisions made as a result of the interim evaluations.
2.0 BRIEF
SUMMARY
OF THE FLEX SEISMIC IMPLEMENTATION STRATEGIES The PNPS FLEX strategies for Reactor Core Cooling and Heat Removal, Reactor Inventory Control, and Containment Function are summarized below. This summary is derived from the PNPS Overall Integrated Plan (OIP) in Response to the March 12, 2012, Commission Order EA-12-049 submitted in February 2013 [3] and is consistent with the third six month status report issued to the NRC in August 2014 [4].
For Phase 1 Core cooling and inventory control are achieved during the first six (6) hours using the Reactor Core Isolation Cooling (RCIC) system aligned to take suction from the torus. Pressure control and heat removal are accomplished by Safety Relief Valves (SRV) venting to the torus. At six (6) hours, a controlled depressurization is commenced based on the Emergency Operating Procedure (EOP) heat capacity temperature limit curve. The depressurization is carried out over three (3) hours using RCIC and cycling the SRVs.
At nine (9) hours (beginning of Phase 2), the operators will transition from the installed RCIC system to diesel powered FLEX low pressure injection pumps, taking suction from the Ultimate Heat Sink (UHS) and connecting through the Condensate Storage Tank (CST) suction line for injection via either the High Pressure Coolant Injection (HPCI) or RCIC idle pump and normal pump discharge path to the Reactor Pressure Vessel (RPV) feedwater lines. An alternate FLEX injection point is to the Residual Heat Page 6
Pilgrim Nuclear Power Station ESEP Report Removal (RHR) system via the readily accessible Firewater to Service Water Cross-tie to RHR, which provides a path into the RPV, Drywell Spray, or Torus via the RHR system.
At 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />, the torus will be vented via the hardened containment vent to provide containment heat removal, and to begin a long term strategy of reactor feedwater makeup and boiling to protect the core and containment.
At 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> (beginning of Phase 3), the water source for the FLEX injection pumps will be transitioned to the mobile water tank fed from the on-site ground water wells. The ground water wells will be powered by a portable 100 kVA generator. The flow from the mobile storage tank will be passed through a FLEX Demineralizer vessel and injected into the RPV via the CST storage tank suction line (same flow path as Phase 2).
Initially containment integrity is maintained by normal design features of the containment (e.g.,
containment isolation valves). At 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />, when torus water temperature reaches 280'F, torus venting will commence through the Hardened Containment Venting System to provide containment heat removal and protect containment integrity. Containment venting will not begin until after reactor depressurization to ensure sufficient containment pressure and net positive suction head for RCIC or HPCI operation.
Necessary electrical components are outlined in the PNPS FLEX OIP submittal, and primarily entail a 125 volt motor control center, vital batteries, battery chargers, and 250 volt DC batteries and battery chargers. Other supporting components include monitoring instrumentation for core cooling, reactor coolant inventory, and containment integrity.
Figure 1 through Figure 5 of [3] provide the FLEX flow paths for PNPS Phases 1 through 3.
3.0 EQUIPMENT SELECTION PROCESS AND ESEL The selection of equipment for the Expedited Seismic Equipment List (ESEL) followed the guidelines of EPRI 3002000704 [2]. The ESEL for PNPS is presented in Attachment A. Information presented in Attachment A is drawn from the following references [3], [4], [5], [6], [7], [8], [9], [10], [11], [12], [13],
[14], [15], [16], [17], [18], [19], [20], [21], [22], [23], [24], [25], [26], [27], [28], [29], [30], [31], [32], [33],
[34], [35], and [36].
3.1 Equipment Selection Process and ESEL The selection of equipment to be included on the ESEL was based on installed plant equipment credited in the FLEX strategies during Phase 1, 2 and 3 mitigation of a Beyond Design Basis External Event (BDBEE), as outlined in the PNPS OIP in Response to the March 12, 2012, Commission Order EA-12-049 [3]. The OIP provides the PNPS FLEX mitigation strategy and serves as the basis for equipment selected for the ESEP.
The scope of "installed plant equipment" includes equipment relied upon for the FLEX strategies to sustain the critical functions of core cooling and containment integrity consistent with the PNPS OIP.
FLEX recovery actions are excluded from the ESEP scope per EPRI 3002000704 [2]. The overall list of planned FLEX modifications and the scope for consideration herein is limited to those required to support core cooling, reactor coolant inventory and subcriticality, and containment integrity functions.
Portable and pre-staged FLEX equipment (not permanently installed) are excluded from the ESEL per EPRI 3002000704.
The ESEL component selection followed the EPRI guidance outlined in Section 3.2 of EPRI 3002000704.
Page 7
Pilgrim Nuclear Power Station ESEP Report
- 1. The scope of components is limited to that required to accomplish the core cooling and containment safety functions identified in Table 3-1 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 PNPS OIP.
- 2. The scope of components is limited to installed plant equipment, and FLEX connections necessary to implement the PNPS OIP as described in Section 2.
- 3. The scope of components assumes the credited FLEX connection modifications are implemented, and are limited to those required to support a single FLEX success path (i.e.,
either "Primary" or "Back-up/Alternate").
- 4. The "Primary" FLEX success path is to be specified. Selection of the "Back-up/Alternate" FLEX success path must be justified.
- 5. Phase 3 coping strategies are included in the ESEP scope, whereas recovery strategies are excluded.
- 6. Structures, systems, and components excluded per the EPRI 3002000704 [2] guidance are:
- Structures (e.g. containment, reactor building, control building, auxiliary building, etc.).
- Piping, cabling, conduit, HVAC, and their supports.
- Manual valves, check valves and rupture disks.
" Power-operated valves not required to change state as part of the FLEX mitigation strategies.
- Nuclear steam supply system components (e.g. RPV and internals, reactor coolant pumps and seals, etc.).
- 7. For cases in which neither train was specified as a primary or back-up strategy, then only one train component (generally 'A' train) is included in the ESEL.
3.1.1 ESEL Development The ESEL was developed by reviewing the PNPS OIP [3] to determine the major equipment involved in the FLEX strategies. Further reviews of plant drawings (e.g., Piping and Instrumentation Diagrams (P&IDs) and Electrical One Line Diagrams) were performed to identify the boundaries of the 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 / 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.
Cabinets and equipment controls containing relays, contactors, switches, potentiometers, circuit breakers and other electrical and instrumentation that could be affected by high-frequency earthquake motions and that impact the operation of equipment in the ESEL are required to be on the ESEL. These cabinets and components were identified in the ESEL. For the ESEL, the relays identified were in the Page 8
Pilgrim Nuclear Power Station ESEP Report RCIC and Automatic Depressurization System (ADS), and malfunction of these relays during a seismic event could lead to the failure of the reactor core cooling safety function.
For Phase 1, RCIC is the primary path for inventory control and core cooling. Therefore, the RCIC system was used as the basis for the Phase 1 ESEL. For Phase 2 and Phase 3, the RCIC system was also used to provide the pathway for RPV injection utilizing portable injection pumps. Relays that could malfunction during a seismic event and prevent successful RCIC or ADS operation were included in the ESEL.
For each parameter monitored during the FLEX implementation, a single indication was selected for inclusion in the ESEL. For each parameter indication, the components along the flow path from measurement to indication were included, since any failure along the path would lead to failure of that indication. Components such as flow elements were considered as part of the piping and were not included in the ESEL.
3.1.2 Power Operated Valves Page 3-3 of EPRI 3002000704 [2] notes that power operated valves not required to change state as part of the FLEX mitigation strategies are excluded from the ESEL. Page 3-2 also notes that "functional failure modes of electrical and mechanical portions of the installed Phase 1 equipment should be considered (e.g. RCIC/AFW trips)." To address this concern, the following guidance is applied in the PNPS 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 be added to the ESEL as these components provide completely passive locations for pulling or installing cables. No breaks or connections in the cabling were included in pull boxes. Pull boxes were considered part of conduit and cabling, which were excluded in accordance with EPRI 3002000704 [2].
3.1.4 Termination Cabinets Termination cabinets, including cabinets necessary for FLEX Phase 2 and Phase 3 connections, provide consolidated locations for permanently connecting multiple cables. The termination cabinets and the internal connections provide a completely passive function; however, the cabinets are included in the ESEL to ensure industry knowledge on panel/anchorage failure vulnerabilities is addressed.
Page 9
Pilgrim Nuclear Power Station ESEP Report 3.1.5 Critical Instrumentation Indicators Critical indicators and recorders are typically physically located on panels/cabinets and are included as separate components; however, seismic evaluation of the instrument indication may be included in the panel/cabinet seismic evaluation (rule-of-the-box).
3.1.6 Phase 2 and 3 Piping Connections Item 2 in Section 3.1 above notes that the scope of equipment in the ESEL includes "... FLEX connections necessary to implement the PNPS OIP [3] as described in Section 2." Item 3 in Section 3.1 also notes that "The scope of components assumes the credited FLEX connection modifications are implemented, and are limited to those required to support a single FLEX success path (i.e., either "Primary" or "Back-up/Alternate")."
Item 6 in Section 3.1 above goes on to explain that "Piping, cabling, conduit, HVAC, and their supports" are excluded from the ESEL scope in accordance with EPRI 3002000704 [2].
Therefore, piping and pipe supports associated with FLEX Phase 2 and Phase 3 connections are excluded from the scope of the ESEP evaluation. However, any active valves in FLEX Phase 2 and Phase 3 connection flow path are included in the ESEL.
3.2 Justification for Use of Equipment That is Not the Primary Means for FLEX Implementation RCIC is the primary system for Phase 1 and was presented as the single success path in the PNPS ESEL.
Therefore, no additional justification is required.
4.0 GROUND MOTION RESPONSE SPECTRUM (GMRS) 4.1 Plot of GMRS Submitted by the Licensee The Safe Shutdown Earthquake (SSE) control point elevation is defined at the bottom of the Reactor Building foundation at elevation -26 ft MSL which is 48 ft below grade based on Section 2.5.3.3.2, Section 2.5.2.4.3, and Figure 12.2-6 of the Final Safety Analysis Report (FSAR) [37]. Table 4-1 shows the GMRS acceleration for a range of frequencies [38]. The GMRS at the control point is shown in Figure 4-1.
Table 4-1: GMRS for PNPS Frequency GMRS (Hz) (g) 100 5.05E-01 90 5.09E-01 80 5.18E-01 70 5.40E-01 60 5.93E-01 50 7.31E-01 40 8.57E-01 35 9.21E-01 Page 10
Pilgrim Nuclear Power Station ESEP Report Table 4-1: GMRS for PNPS (continued)
Frequency GMRS (Hz) (g) 30 9.51E-01 25 9.22E-01 20 9.09E-01 15 9.61E-01 12.5 1.06E+00 10 1.18E+00 9 1.16E+00 8 1.10E+00 7 9.75E-01 6 8.07E-01 5 6.09E-01 4 3.83E-01 3.5 2.93E-01 3 2.19E-01 2.5 1.61E-01 2 1.29E-01 1.5 1.OOE-01 1.25 7.82E-02 1 6.22E-02 0.9 5.87E-02 0.8 5.55E-02 0.7 5.18E-02 0.6 4.67E-02 0.5 3.92E-02 0.4 3.14E-02 0.35 2.74E-02 0.3 2.35E-02 0.25 1.96E-02 0.2 1.57E-02 0.15 1.18E-02 0.125 9.80E-03 0.1 7.84E-03 Page 11
Pilgrim Nuclear Power Station ESEP Report GMRS at Control Point for Pilgrim Nuclear Power Station, 5% Damping 1.40 1.20 1.00 0.80 0.60 0.40 0.20 0.00 0.1 10 100 Frequency (Hz)
Figure 4-1: GMRS for PNPS 4.2 Comparison to SSE The SSE is defined in the FSAR in terms of a Peak Ground Acceleration (PGA) and a design response spectrum. These spectra have been digitized and tabulated [39]. Table 4-2 shows the spectral acceleration values at selected frequencies for the 5% damped horizontal SSE.
Table 4-2: SSE for PNPS Frequency Spectral Acceleration (Hz) (g) 100 0.15 33 0.15 25 0.15 10 0.184 9 0.194 5 0.238 2.5 0.225 1 0.126 0.5 0.071 Page 12
Pilgrim Nuclear Power Station ESEP Report GMRS to SSE Comparison for Pilgrim Nuclear Power Station, 5% Damping 1.40 1.20 1.00 080 0.60 0.40 0.20 0.00 0.1 1 10 100 Frequency (Hz)
Figure 4-2: GMRS to SSE Comparison for PNPS The SSE envelops the GMRS in the low frequency range up to approximately 3 Hz. The GMRS exceeds the SSE beyond that point. As the GMRS exceeds the SSE in the 1 to 10 Hz range, the plant does not screen out of the ESEP according to Section 2.2 of EPRI 3002000704 [2]. The two special screening considerations as described in Section 2.2.1 of EPRI 3002000704, namely a) Low-frequency GMRS exceedances at Low Seismic Hazard Sites and b) Narrow Band Exceedances in the 1 to 10 Hz range, provide criteria for accepting specific GMRS exceedances. However, the GMRS exceedances are not limited to the low frequency range and there are no narrow-banded exceedances. Therefore, these special screening considerations do not apply for PNPS and hence High Confidence of a Low Probability of Failure (HCLPF) evaluations were performed.
5.0 REVIEW LEVEL GROUND MOTION (RLGM) 5.1 Description of RLGM Selected The RLGM is selected based on Approach 1 in Section 4 of EPRI 3002000704 [2]. The RLGM is developed based on the SSE [39].
The maximum GMRS/SSE ratio between 1 and 10 Hz range occurs at 10 Hz where the ratio is 1.18/0.184 = 6.41. As the maximum ratio of the GMRS to the SSE over the 1 to 10 Hz range exceeds a value of 2, the GMRS/SSE ratio is set to the maximum scaling factor value of 2.0 for PNPS in accordance with Section 4 of EPRI 3002000704. Table 5-1 lists the horizontal ground RLGM acceleration at 5%
damping at selected frequencies and the plot is shown in Figure 5-1. The RLGM are generated by plotting the digitized data on a linear/linear graph paper, and connecting the points with straight lines.
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Pilgrim Nuclear Power Station ESEP Report Table 5-1: RLGM for PNPS Frequency RLGM at 5% Damping (Hz) (g) 0.10 0.016 0.30 0.094 0.60 0.166 1.00 0.252 1.40 0.318 2.00 0.412 2.50 0.450 3.00 0.476 3.50 0.488 4.00 0.488 4.50 0.484 5.00 0.476 5.50 0.466 6.00 0.450 7.00 0.428 8.00 0.406 10.00 0.368 12.00 0.338 14.00 0.308 18.00 0.300 24.00 0.300 33.00 0.300 Page 14
Pilgrim Nuclear Power Station ESEP Report Review Level Ground Motion (2xSSE) Response Spectra - Horizontal Direction 0.60 - - ' -' * *-
156%0Damping _
0.40
- I I _I 0.20 44 :'
-f------ -
... .... . . t - .. .. . 1-!.. .. .. .
010 0.00, ,
4 8 12e16 20 24 28 32 Frequency (Hz)
Figure 5-1: RLGM for PNPS 5.2 Method to Estimate In-Structure Response Spectra (ISRS)
The RLGM ISRS for PNPS are generated by scaling the SSE ISRS [39]. The following steps are used to generate the RLGM ISRS.
- 2. Calculate the horizontal RLGM ISRS by scaling the horizontal direction SSE ISRS by a factor of 2.0.
The vertical direction RLGM ISRS is obtained by scaling the vertical amplified ground response spectrum.
6.0 SEISMIC MARGIN EVALUATION APPROACH It is necessary to demonstrate that ESEL items have sufficient seismic capacity to meet or exceed the demand characterized by the RLGM. The seismic capacity is characterized as the PGA for which there is a HCLPF. The PGA is associated with a specific spectral shape, in this case the 5%-damped RLGM spectral shape. The HCLPF capacity must be equal to or greater than the RLGM PGA. The criteria for seismic capacity determination are given in Section 5 of EPRI 3002000704 [2].
There are two basic approaches for developing HCLPF capacities:
- 1. Deterministic approach using the conservative deterministic failure margin (CDFM) methodology of EPRI NP-6041-SL, A Methodology for Assessment of Nuclear Power Plant Seismic Margin (Revision 1) [40].
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Pilgrim Nuclear Power Station ESEP Report
- 2. Probabilistic approach using the fragility analysis methodology of EPRI TR-103959, Methodology for Developing Seismic Fragilities [41].
6.1 Summary of Methodologies Used PNPS performed a SPRA in 1994 as part of Individual Plant Examination for External Events (IPEEE) program. The SPRA is documented in [23] and consisted of screening walkdowns, fragility analysis and three-dimensional soil structure interaction (SSI) analysis. The SPRA was a hybrid of the conventional PRA and seismic margin assessment approaches. The seismic walkdowns for IPEEE were performed simultaneously with USI A-46 evaluations. Section 3.3 of [38] established that the results of the PNPS SPRA performed as part of IPEEE are not sufficient to serve as the basis for PNPS to screen-out of further risk assessment.
For ESEP, the SMA consisted of screening walkdowns and HCLPF calculations. The screening walkdowns used the screening tables from Chapter 2 of EPRI NP-6041-SL [40]. The walkdowns were conducted by engineers trained in EPRI NP-6041-SL and were documented on Screening Evaluation Work Sheets (SEWS) from EPRI NP-6041-SL. Anchorage capacity calculations used the CDFM criteria from EPRI NP-6041-SL. Seismic demand was based on EPRI 3002000704 [2] using an RLGM of 2xSSE with a PGA of 0.3g, Figure 5-1.
6.2 HCLPF Screening Process For ESEP, the components are screened considering the RLGM (2xSSE) with a 0.3g PGA. The screening tables in EPRI NP-6041-SL [40] are based on ground peak spectral accelerations of 0.8g and 1.2g. These both exceed the RLGM peak spectral acceleration.
The ESEL components were prescreened based on Table 2-4 of EPRI NP-6041-SL. Additional pre-screening, specifically for anchorage, considered walkdown results and documentation from NTTF 2.3 and SEWS from IPEEE and USI A-46. Equipment anchorage was screened out in cases where previous evaluations showed large available margin against SSE. The remaining components (i.e., components that do not screen out), were identified as requiring HCLPF calculations. ESEL components were walked down and based on the equipment and anchorage conditions, prescreening decisions were confirmed and a final list of required HCLPF calculations was generated. Equipment for which the screening caveats were met and for which the anchorage capacity exceeded the RLGM seismic demand are screened out from ESEP seismic capacity determination because the HCLPF capacity exceeds the RLGM.
The PNPS ESEL contains 187 items. Of these, 16 are valves. In accordance with Table 2-4 of EPRI NP-6041-SL, active valves may be assigned a functional capacity of 0.8g peak spectral acceleration without any review other than looking for valves with large extended operators on small diameter piping, and anchorage is not a failure mode. Therefore, valves on the ESEL are screened out from ESEP seismic capacity determination, subject to the caveat regarding large extended operators on small diameter piping.
The non-valve components in the ESEL are screened based on the SMA results. If the SMA showed that the component met the EPRI NP-6041-SL screening caveats and the CDFM capacity exceeded the RLGM demand, the components are screened out from the ESEP capacity determination.
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Pilgrim Nuclear Power Station ESEP Report Six (6) block walls were identified in the proximity of ESEL equipment. These block walls were assessed for potential seismic interaction impact resulting from the RLGM by reviewing the existing plant documents and or by generating new analysis and found to be acceptable.
6.3 Seismic Walkdown Approach 6.3.1 Walkdown Approach Walkdowns were performed in accordance with the criteria provided in Section 5 of EPRI 3002000704
[2], which refers to EPRI NP-6041-SL [40] for the Seismic Margin Assessment process. Pages 2-26 through 2-30 of EPRI NP-6041-SL 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-radioactiveor low radioactiveenvironments. Seismic capability assessment of components which are inaccessible, in high-radioactiveenvironments, or possibly within contaminatedcontainment, will have to rely more on alternatemeans such as photographicinspection, 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 reasonablebasisfor 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 preparatorywork (Step 3) by reference to drawings, calculationsor specifications. The one component or each type which is selected should be thoroughly inspected which probably does mean de-energizing and opening cabinets or panelsfor this very limited sample. Generally, a spare representativecomponent can be found so as to enable the inspection to be performed while the plant is in operation. At leastfor the one component of each type which is selected, anchorageshould 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 correspondenceis 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 procedurefor inspection should be repeatedfor 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 arefound, then the system or component class must be inspected in closer detail until the systematic deficiency is defined.
The 100% "walk by" is to look for outliers, lack of similarity, anchoragewhich is differentfrom that shown on drawings or prescribedin criteriafor that component, potentialSI [Seismic Interaction]problems, situationsthat are at odds with the team members' past experience, and any other areas of seriousseismic 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.
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Pilgrim Nuclear Power Station ESEP Report 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 responsiblefor the seismic adequacy of all elements which they screen from the margin review. Appendix D gives guidancefor sampling selection."
6.3.2 Application of Previous Walkdown Information Several ESEL items were previously walked down during the PNPS seismic IPEEE program, for seismic IPEEE outlier resolutions in accordance with USI A-46 evaluation program and NTTF Recommendation 2.3. Those walkdown results were reviewed and the following steps were taken to confirm that the previous walkdown conclusions remained valid.
- A walk by was performed to confirm that the equipment material condition and configuration is consistent with the walkdown conclusions and that no new significant interactions related to block walls or piping attached to tanks exist.
- If the ESEL item was screened out based on the previous walkdown, that screening evaluation was reviewed and reconfirmed for the ESEP.
6.3.3 Significant Walkdown Findings Consistent with the guidance from EPRI NP-6041-SL [40], no significant outliers or anchorage concerns were identified during the PNPS seismic walkdowns. Based on walkdown results, HCLPF capacity evaluations were recommended for the following eight (8) components:
" C7, Containment Isolation and Ventilation Vertical Board
- C2205A, Reactor Protection and NSS Inst. Rack
- C2251A, Jet Pump Instrument Rack A
- D16, 125 VDC Bus A
- D1, 125 VDC Battery Rack A
- D2, 125 VDC Battery Rack B
- D3, 125 VDC Battery Rack B
" EG-23, Vital MG Set 6.4 HCLPF Calculation Process ESEL items identified for ESEP at PNPS were evaluated using the criteria in EPRI NP-6041-SL [40] and Section 5 of EPRI 3002000704 [2]. Those evaluations included the following steps:
- Performing seismic capability walkdowns for equipment not included in previous seismic walkdowns (SQUG, IPEEE, or NTTF 2.3) to evaluate the equipment installed plant conditions
- Performing screening evaluations using the screening tables in EPRI NP-6041-SL as described in Section 6.2
- Performing HCLPF calculations considering various failure modes that include both structural failure modes (e.g. anchorage, load path etc.) and functional failure modes All HCLPF calculations were performed using the CDFM methodology. A total of seven (7) HCLPF calculations were performed to address the eight (8) components.
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Pilgrim Nuclear Power Station ESEP Report
- C7, "Containment Isolation and Ventilation Vertical Board"
- C2205A, "Reactor Protection and NSS Inst. Rack"
- C2251A, "Jet Pump Instrument Rack A"
- D16, "125 VDC Bus A"
" D2 and D3, "125 VDC Battery Rack B"
- D1, "125 VDC Battery Rack A"
- EG-23, "Vital MG Set" 6.5 Functional Evaluations of Relays Five (5) relays 13A-K3, -K5, -K7, K10, and -K22 associated with "RCIC Relay Vertical Board" cabinet C930, and three (3) relays 13A-K31, -K32, and -K33 associated with "Channel B Vertical Board" cabinet C933 were identified as seal in/lockout type needing HCLPF calculation. The relays were of three types General Electric 12HFA151A2F, General Electric 12HGA11A52F and Agastat 7014PB.
The relays were evaluated using the guidance provided in EPRI NP-6041-SL [40] for equipment qualified by testing. Subject relays were determined to have higher HCLPF values than the plant RLGM.
6.6 Tabulated ESEL HCLPF Values (Including Key Failure Modes)
Tabulated ESEL HCLPF values are provided in Attachment B. The following notes apply to the information in the tables.
- For items screened out using EPRI NP-6041-SL [40] screening tables, the HCLPF capacity is provided as >RLGM and the failure mode is listed as "Screened", (unless the controlling HCLPF value is governed by anchorage).
- For items where anchorage controls the HCLPF value, the HCLPF value is listed in the table and the failure mode is noted as "anchorage." For the items where the component function controls the HCLPF value, the HCLPF value is listed in the table and the failure mode is noted as "functional."
After performing the HCLPF calculations, the anchorage was determined to have adequate capacity for the design basis loads and HCLPF greater than RLGM for all components except EG-23. A modification is planned for EG-23 and the HCLPF capacity presented in Attachment B includes the proposed modifications.
7.0 INACCESSIBLE ITEMS 7.1 Identification of ESEL Item Inaccessible for Walkdowns There are total of 13 Torus Water Temperature Elements (TEs) on the ESEL. Six (6) of the TEs were not walked down since they are located in a high dose area. The evaluation of subject TEs were done by comparison and similarity to the other seven (7) TEs that were walked down. The following is the list of the TEs that were not walked down:
- TE5021-01A
" TE5021-06A Page 19
Pilgrim Nuclear Power Station ESEP Report
- TE5021-07A
- TE5021-08A
- TE5021-10A
- TE5021-12A Also, the two (2) valves and two (2) accumulator tanks listed below were not walked down, since they are located in the Dry well (inaccessible area). Subject components were evaluated based on the available photos, drawings, existing SEWS, and vendor information.
- RV-203-3B, Safety Relief Valve
- RV-203-3C, Safety Relief Valve
- T-221B, Accumulator Tank for SRV B
- T-221C, Accumulator Tank for SRV C In addition two (2) junction boxes J599 and J600 were not walked down since they were not accessible for visual inspection due to their location. These items were assessed and found to be acceptable by comparison and similarity to J601 and J602 respectively, and by reviewing their A-46 SEWS, and existing analysis information.
7.2 Planned Walkdown / Evaluation Schedule / Close Out There are no components that require follow up seismic walkdowns.
8.0 ESEP CONCLUSIONS AND RESULTS 8.1 Supporting Information PNPS has performed the ESEP as an interim action in response to the NRC's 50.54(f) letter [1]. It was performed using the methodologies in the NRC endorsed guidance in EPRI 3002000704 [2].
The ESEP provides an important demonstration of seismic margin and expedites plant safety enhancements through evaluations and potential near-term modifications of plant equipment that can be relied upon to protect the reactor core following beyond design basis seismic events.
The ESEP is part of the overall PNPS response to the NRC's 50.54(f) letter. On March 12, 2014, NEI submitted to the NRC results of a study [43] 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
[38]. 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 [42] concluded that the "fleet wide seismic risk estimates are consistent with the approach and results used in the GI-199 safety/risk assessment." The letter also stated that "As a result, the staff has confirmed that the conclusions reached in GI-199 safety/risk assessment remain valid and that the plants can continue to operate while additional evaluations are conducted."
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Pilgrim Nuclear Power Station ESEP Report An assessment of the change in seismic risk for PNPS was included in the fleet risk evaluation submitted in the March 12, 2014 NEI letter [43]; therefore, the conclusions in the NRC's May 9 letter also apply to PNPS.
In addition, the March 12, 2014 NEI letter provided an attached "Perspectives on the Seismic Capacity of Operating Plants," which (1) assessed a number of qualitative reasons why the design of Structures, Systems and Components (SSCs) inherently contain margin beyond their design level, (2) discussed industrial seismic experience databases of performance of industry facility components similar to nuclear SSCs, and (3) discussed earthquake experience at operating plants.
The fleet of currently operating nuclear power plants was designed using conservative practices, such that the plants have significant margin to withstand large ground motions safely. This has been borne out for those plants that have actually experienced significant earthquakes. The seismic design process has inherent (and intentional) conservatisms which result in significant seismic margins within SSCs.
These conservatisms are reflected in several key aspects of the seismic design process, including:
- Safety factors applied in design calculations
- Damping values used in dynamic analysis of SSCs
- Bounding synthetic time histories for in-structure response spectra calculations
" Broadening criteria for in-structure response spectra
" Response spectra enveloping criteria typically used in SSC analysis and testing applications
- Response spectra based frequency domain analysis rather than explicit time history based time domain analysis
- Bounding requirements in codes and standards
- Use of minimum strength requirements of structural components (concrete and steel)
- Bounding testing requirements
" Ductile behavior of the primary materials (that is, not crediting the additional capacity of materials such as steel and reinforced concrete beyond the essentially elastic range, etc.)
These design practices combine to result in margins such that the SSCs will continue to fulfill their functions at ground motions well above the SSE.
The intent of the ESEP is to perform an interim action in response to the NRC's 50.54(f) letter to demonstrate seismic margin through a review of a subset of the plant equipment that can be relied upon to protect the reactor core following beyond design basis seismic events. The RLGM used for the ESEP evaluation is a scaled version of the plant's SSE rather than the actual GMRS. To more fully characterize the risk impacts of the seismic ground motion represented by the GMRS on a plant specific basis, a more detailed seismic risk assessment (SPRA or risk-based SMA) is to be performed in accordance with EPRI 1025287 [44] . As identified in the PNPS Seismic Hazard and GMRS submittal
[38], PNPS screens in for a risk evaluation. The complete risk evaluation will more completely characterize the probabilistic seismic ground motion input into the plant, the plant response to that probabilistic seismic ground motion input, and the resulting plant risk characterization. PNPS will complete that evaluation in accordance with the schedule identified in NEI's letter dated April 9, 2013
[45] and endorsed by the NRC in their May 7, 2013 letter [46].
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Pilgrim Nuclear Power Station ESEP Report 8.2 Identification of Planned Modifications Insights from the ESEP identified the following item where the HCLPF is below the RLGM and plant modifications will be made in accordance with EPRI 3002000704 [2] to enhance the seismic capacity of the plant.
- Vital MG Set EG-23 anchorage had a HCLPF capacity below RLGM. A modification is planned to provide additional seismic margin such that the HCLPF will exceed the RLGM.
8.3 Modification Implementation Schedule Plant modifications described in Section 8.2 will be performed in accordance with the schedule identified in NEI letter dated April 9, 2013 [45], 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.
8.4 Summary of Regulatory Commitments The following actions will be performed as a result of the ESEP.
Equipment Action # Equipment ID Description Action Description Completion Date 1 EG-23 Vital MG Set Modify anchorage such that As described in HCLPF > RLGM Section 8.3 2 N/A N/A Submit a letter to NRC Within 60 days summarizing the HCLPF results of following Item 1 confirming completion of ESEP implementation of the plant activities, including modification associated with item 1 item 1
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, NRC ADAMS Accession No. ML12053A340.
- 2. EPRI 3002000704, "Seismic Evaluation Guidance, Augmented Approach for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1: Seismic," May 2013.
- 3. Entergy Letter to U.S. NRC, letter number 2.13.012 "Overall Integrated Plan in Response to March 12, 2012, Commission Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond Design Basis Events (Order Number EA-12-049)," February 28, 2013, NRC ADAM Accession No. ML13063A063.
- 4. Entergy Letter to U.S. NRC, letter number 2.14.061 "Pilgrim Nuclear Power Station's Third Six Month Status Report in Response to March 12, 2012, Commission Order Modifying Licenses Page 22
Pilgrim Nuclear Power Station ESEP Report with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events (Order Number EA-12-049)," August 28, 2014, NRC ADAMS Accession No. ML14253A189.
- 6. Entergy Drawing E13, Revision E83, "Single Line Relay & Meter Diagram 125V and 250V DC Systems."
8.. Entergy Drawing M1G13-11, Revision E17, "Elementary Diagram RCIC System, Sheet 3 of 9."
- 10. Entergy Drawing MlG12-12, Revision E14, "Elementary Diagram RCIC System, Sheet 2 of 9."
- 11. Entergy Drawing M252SH1, Revision 69, "P & ID Nuclear Boiler."
- 12. Entergy Drawing M1R4-10, Revision 25, "Elementary Diagram Automatic Blowdown System, Sheet 1 of 2."
- 13. Entergy Drawing M253SH1, Revision 45, "Nuclear Boiler Vessel Instrumentation."
- 14. Entergy Drawing M253SH2, Revision 29, "Nuclear Boiler Vessel Instrumentation."
- 15. Entergy Drawing E91, Revision E8, "Wiring Block Diagram RCIC System, Sheet 1 of 2."
- 16. Entergy Drawing M241SH1, Revision 87, "P&ID Residual Heat Removal System."
- 17. Entergy Drawing M227SH1, Revision 60, "P&ID Containment Atmospheric Control System."
- 18. Entergy Drawing E14SH1, Revision 39, "Single Line Diagram 120 V Instrument AC Vital and Reactor Protection AC System & +/-24 VDC Power System."
- 19. Entergy Drawing M1R8-2, Revision 10, "Elementary Diagram Automatic Blowdown System."
- 20. Entergy Document ELNRC1.2.96.085, BECo Letter 96-085, "Summary Report, Generic Letter 87-02, Verification of Seismic Adequacy of Mechanical and Electrical Equipment in Operating Reactors, Unresolved Safety Issue (USI) A-46."
- 21. Entergy Drawing E727, Revision E7, "Elementary Diagram Emergency Core Cooling System Analog Trip Cabinet C2233A Section A."
- 22. Entergy Drawing M1P361-2, Revision E3, "Arrangement Diagram Reactor Core Isolation Cooling System Instrument Rack C2258, Sheet 3 of 4."
- 23. Entergy Correspondence IPEEE, "Pilgrim Nuclear Power Station Individual Plant Examination for External Events (GL 88-20)," dated July 1994.
- 24. Entergy Drawing M1P302-15, Revision Ell, "Arrangement Drawing Control Room Panel C903, Sheet 2 of 2."
- 25. Entergy Drawing E692, Revision E6, "Elementary Diagram Torus Water Temperature Monitoring System Channel A."
- 26. Entergy Drawing M206B, Revision EO, "Control Room and Local Panel Instruments."
- 27. Entergy Drawing E401SH2, Revision E0, "Schematic Diagram Containment Atmospheric Control System."
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Pilgrim Nuclear Power Station ESEP Report
- 28. Entergy Drawing M227A32-10, Revision EO, "C170/C171 Post Accident Monitoring System."
- 29. Entergy Document EQDFDPT1001-604A, Revision 7, "Equipment Qualification Data Sheet DFPT1001-604A."
- 30. Entergy Drawing M227-154, Revision E8, "Front View Layout for Containment Ventilation Isolation & Gas Treatment - Vertical Board C-7."
- 31. Entergy Drawing E401SH3, Revision El, "Schematic Diagram Containment Atmospheric Control System."
- 32. Entergy Drawing M227C1, Revision EO, "Arrangement Drawing Torus Water Temperature Monitoring Sys. Panel C179."
- 33. Entergy Drawing MlG17-7 Revision E6, "Elementary Diagram RCIC System, Sheet 7 of 9."
- 34. Entergy Drawing M1P355-5 Revision 4, "Arrangement Drawing Leak Detection System Instrument Rack C2257."
- 35. Entergy Document EC45555 Revision 1, "FLEX Alternate Power to 125VDC and 250VDC Battery Chargers (Base EC)."
- 36. Entergy Document EC42259 Revision 0, "PNPS FLEX Strategy Master EC for Beyond-Design-Basis External Events (BDBEEs) Diverse & Flexible Coping Strategy (FLEX) Implementation."
- 37. "Pilgrim Nuclear Power Station - Final Safety Analysis Report," Revision 29, Docket No. 50-293, October 2013.
- 38. Entergy Letter Number 2.14.026, John A. Dent Jr. to NRC, "Entergy's Seismic Hazard and Screening Report (CEUS Sites), Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding the Seismic Aspects of Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident dated March 31, 2014." NRC ADAMS Accession No. ML14092A023.
- 39. Entergy Document C114ERQE1, Revision El, "Seismic Response Spectra," October 2005. (Stored in Merlin as C114ERQEO)
- 40. EPRI-NP-6041-SL, "Methodology for Assessment of Nuclear Power Plant Seismic Margin,"
Revision 1, August 1991.
- 41. EPRI TR-103959, "Methodology for Developing Seismic Fragilities," July 1994.
- 42. 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-Ichi Accident," May 9, 2014, NRC ADAMS Accession No. ML14111A147.
- 43. 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.
- 44. EPRI 1025287, "Seismic Evaluation Guidance: Screening, Prioritization and Implementation Details (SPID) for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1:
Seismic. Electric Power Research Institute," February 2013.
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Pilgrim Nuclear Power Station ESEP Report
- 45. Nuclear Energy Institute (NEI), A. Pietrangelo, Letter to D. Skeen of the USNRC, "Proposed Path Forward for NTTF Recommendation 2.1: Seismic Reevaluations," April 9, 2013, NRC ADAMS Accession No. ML13101A379.
- 46. 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 Alternative to the March 12, 2012, Information Request for Seismic Reevaluations," May 7, 2013, NRC ADAMS Accession No. ML13106A331.
- 47. Entergy Document EC53987, "Expedited Seismic Evaluation Process (ESEP) Report (Submittal to NRC - Fukushima 50.54F Response", the following AREVA documents are captured in the plant document management system:
Pilgrim Nuclear Power Station."
- b. AREVA Calculation 32-9224839-001, "Pilgrim ESEP HCLPF Calculation - Reactor Protection and NSS Instrument Rack, C2205A."
- c. AREVA Calculation 32-9225449-001, "Pilgrim ESEP HCLPF Calculation - Containment Isolation and Ventilation Vertical Board, C7."
- e. AREVA Calculation 32-9225626-001, "Pilgrim ESEP HCLPF Calculation - Jet Pump A Instrument Rack C2251A."
- g. AREVA Calculation 32-9226906-001, "Pilgrim ESEP HCLPF Calculation - Motor-Generator Set, EG-23."
- 48. Entergy Calculation C15.0.3623, "FLEX Transfer Switch Mounting Evaluation for Battery Chargers Dl1, D12, D13, D14, D15," Revision OA.
- 49. Entergy Calculation C15.0.3624, "FLEX Transfer Switch Mounting Evaluation for 120 VAC Panels Y3, Y31, Y4, Y41, Y13, and Y14," Revision 0.
- 50. Entergy Calculation C15.0.3631, "Nitrogen Cylinder and Pressure Regulator Supports - Backup Nitrogen Supply to Main Steam SRVs RV-203-3B and RV-203-3C," Revision OB Page 25
Pilgrim Nuclear Power Station ESEP Report ATTACHMENT A - PNPS ESEL Page A-1
Pilgrim Nuclear Power Station ESEP Report ESEL Equipment>'. Operating State ,
Item . Nom. ** De" r.. d* . otes/Comments t References
- .*,;* *.* .. ;.::° ,*'Norm ali .
- Desired'. *: : * , '" ." .....; ,.:...". . .
Number ID Description.. ., . te . State. * ,..
I MO-1301-22 RCIC Pump CST Suction Must close to swap RCIC suction to the [5]
VaCvC Psuppression pool 2 D754 DC Bus D7 Supply Energized Energized Supply power to MO-1301-22 [6]
Breaker to MO-1301-22 The FLEX Primary External Water Source Injection Point is to the HPCI & RCIC FLEX Primary External System 18-inch CST common suction line 3 TBD Water Source Injection Closed Open and shall be established at the CST T- [3]
Point 105A/B piping vault on the plant-side of the 26-HO-78 & 79 CST 18-inch manual isolation butterfly valves.
4 Normally closed, must open to supply [5]
M0-1301-25 4 RCIC Suction from Torus Closed Open RCIC suction from torus D761 DC Bus D7 supply Energized Energized Supply power to MO-1301-25 [6]
breaker to MO-1301-25 6 MO-1301-26 RCIC Suction from Torus Closed Open Normally closed, must open to supply [5]
RCIC suction from torus 7 DC Bus D7 supply Energized Energized Supply power to MO-1301-26 [6]
7D764 breaker to MO-1301-26 8 P-206 RCIC Pump Idle Operating Provides RPV makeup in Phase 1 [7]
9 ft~f12A1Af ValveNormally closed, must open to supply ]
MO-1301-49 RCIC Discharge Valve Isolation Closed Open RCICrCaC cton flow injection flow[5]
10 D774 DC Bus D7 Supply Energized Energized Supply power to MO-1301-49 [6]
Breaker to MO-1301-49 RCIC Lube Oil Cooling Open/Close Controls cooling water flow to RCIC lube 11 PCV 1301-43 Water Pressure Control Open as needed oil cooler[7]
_ Valve as needed oIl Iooler Page A-2
Pilgrim Nuclear Power Station ESEP Report ESEL , Equipment. Operating State Item'
- State State
- Notes/Comments , . References DesripionNormal - Desired ~
Number ID Decito.
RCICTurine ubeOilNormally closed, opens to allow cooling 1MO1062inlet Valve Closed Open water flow to RCIC turbine lube oil [7]
cooler 13 D794 DC Bus D7 Supply Energized Energized Supply power to MO-1301-62 [6]
Breaker to MO-1301-62 RCIC Barometric Collects and condenses RCIC gland seal 14 E-201 Condenser and Vacuum Idle Operating leakage [7]
_________ ~~Tank ______
15 E-204 RCIC Lube Oil Cooler Idle Operating Cools RCIC lube oil [7]
16 P-221 RCIC Vacuum Tank Idle Operating Pumps condensate from vacuum tank to [7]
Condensate Pump pump suction 17 D712 DC Bus D7 Supply Energized Energized Supply power to P-221 [6]
Breaker to P-221 18 P-222 . RCIC Vacuum Pump Idle Operating Maintains vacuum on barometric [7]
condenser 19 D714 DC Bus D7 Supply Energized Energized Supply power to P-222 [6]
Breaker to P-222 20 MO-1301-61 RCIC Turbine Steam CoeOpn Opens on RCIC start to admit steam to [7]
Inlet Valve CoeOpn RCIC turbine 21 D751 DC Bus D7 Supply Energized Energized Supply power to MO-1301-61 [6]
Breaker to MO-1301-61 22 SV-1301-1 RCIC Turbine Trip Open Open Closes on protective signals that will be [7]
Throttle Valve bypassed by procedure 23 HO-1301-159 RCIC Turbine Governor Open Open/Close Modulates steam flow to RCIC turbine [7]
Valve as needed 24 X-202 RCIC Turbine Idle Operating Supplies motive force to RCIC pump [7]
Page A-3
Pilgrim Nuclear Power Station ESEP Report ESEL -Equipment Operating State . -w Item . N. Notes/Comments .,References Number ID Description Normal Desired, State State RCIC Pump Flow Transmits RCIC flow to flow control 25 FT-1360-4 Transmitter for Turbine On On Transit [5][22]
Control 26 SQRT1340-10 RCIC Flow Square Root Energized Energized RCIC flow integrator for RCIC flow [5]
Converter control 27 FIC-1340-1 RCIC Flow Indicating Energized Energized Controls RCIC flow [5]
Controller 28 DC/AC 1340- DC/AC Inverter for RCIC Energized Supplies AC Power to RCIC flow control [5][8]
16 Flow Controller circuitry 29 C1303 RCIC Local Controls Energized Energized Local control of RCIC turbine [15]
30 D4-3 Supply Breaker for RCIC Closed Closed Power to the logic and indications in [6]
in Panels C904 and C939 C904 and C939 31 13A-K1 RCIC Auto Initiation Deenergized Energized Start RCIC on low-low water level (C930) [10]
Logic 32 13A-K1O RCIC Steam Supply Low Deenergized Deenergized Could prevent RCIC operation [10]
Pressure 33 13A-K11 Turbine Trip Auxiliary Deenergized Deenergized Could prevent RCIC operation [10]
Relay 34 13A-K13 Pump Discharge Low Deenergized Deenergized Could prevent RCIC operation [10]
Flow 35 Pump Suction Low Deenergized Deenergized Could prevent RCIC operation [10]
13A-K14 P Pressure 36 13A-K17 Turbine Exhaust High Deenergized Deenergized Could prevent RCIC operation [10]
Pressure 37 13A-K8 MO-1301-25 Position Deenergized Energized Monitors valve position [10]
RCACKAutMonitor n 38 13A-K2 RCIC Auto Initiation Deenergized Energized Start RCIC on low-low water level (C930) [10]
Logic Page A-4
Pilgrim Nuclear Power Station ESEP Report ESEL Equipmient Operating State Item *ii*.
- *:*:**. _ ;. : .. .:':. Normal* Desired - *****i!:j-:.Notes/Comments
,,,;,-***ii\. ;:. :, References..
Number ID DeRscript ionS 39 13A-K22 RCIC Auto Isolation Deenergized Deenergized Could prevent RCIC operation [10]
Relay 40 13A-K3 Pump/Turbine Room Deenergized Deenergized Could prevent RCIC operation [10]
High Temperature 41 13A-K31 Pump/Turbine Room Deenergized Deenergized Could prevent RCIC operation [9]
High Temperature 42 13A-K32 RCIC Valve Station High Deenergized Deenergized Could prevent RCIC operation [9]
Temperature 43 13A-K33 RCIC Steam Line High Deenergized Deenergized Could prevent RCIC operation [9]
Differential Pressure 44 13A-K34 RCIC Auto Isolation Deenergized Deenergized Could prevent RCIC operation [9]
Relay 45 13A-K5 RCIC Valve Station High Deenergized Deenergized Could prevent RCIC operation [10]
Temperature 46 13A-K7 RCIC Steam Line High Deenergized Deenergized Could prevent RCIC operation [10]
Differential Pressure 47 RV-203-3B Safety Relief Valve B Shut Open Used for pressure control and cooldown [11]
48 RV-203-3C Safety Relief Valve C Shut Open Used for pressure control and cooldown [11]
49 SV-203-3B Solenoid Pilot Valve for Deenergized Energized Pilot valve for associated SRV [11][19]
SRV B 50 SV-203-3C Solenoid Pilot Valve for Deenergized Energized Pilot valve for associated SRV [11][19]
SRV C 51 T-221B Accumulator Tank for Operable Operable Holds operating nitrogen for SRV [11]
SRV B 52 T-221C Accumulator Tank for Operable Operable Holds operating nitrogen for SRV [11]
SRVCPae A Page A-5
Pilgrim Nuclear Power Station ESEP Report ESEL . Equipment.. : Operating State..*;, 'No o Item .. *..... . ... :-Notes/Comments . References Number . Normal : .- Desired-"
ID Description.
. .... ", State . . State* - ' _" ' o Pressure Control Valve Controls the pressure supplied to the [11]
53 PCV-203-11 for Backup Nitrogen Idle In service SRV accumulators Supply Used to inhibit automatic initiation of 54 2E-S1OA ADS Inhibit Switch Normal Inhibit Automatic Depressurization System [12]
(ADS) 55 2E-S1B SRV B Valve Actuation Auto Open Open SRV B for cooldown [12]
56 2E-S1C SRV C Valve Actuation Auto Open Open SRV C for cooldown [12][19]
57 2E-K12AInelk CSCS Pump Running Deenergized Deenergized Included for relay chatter impacts [12][20]
Interlock 58 2E-K12B CSCS Pump Running Deenergized Deenergized Included for relay chatter impacts [12][20]
Interlock 59 SRV Power Supply Energized Energized Included for relay chatter impacts [12][20]
2E-K13BControl Logic SRV Power Supply Energized Energized Included for relay chatter impacts [12][16][19]
60 2E-K13C Control Logic [20]
61 2E-K7A Reactor Water Low Low Deenergized Energized Included for relay chatter impacts [12][20]
Level Interlock Low 62 2E-K7B Reactor Water Low Low Deenergized Energized Included for relay chatter impacts [12][20]
Level Interlock 63 LI-263-100A Reactor Water Level Operable Operable Reactor level indication [13][20]
Narrow Range Voltage Current CE/-263-72A Operable Operable Reactor level indication [13][20]
ReaConverter Level 65 LI-263-106A Reactor Water Level Operable Operable Reactor level indication [13][20]
Indicator Page A-6
Pilgrim Nuclear Power Station ESEP Report ESEL Equipment Operating State .
Item Numb.. . Norm ,al .- .. . Des, red .,. .,.. : ., . Notes/Comments,
.. . . , .. . . .... .. .. References Number ID Description State Desired
.tate,. State Voltage Current CE/-263-73A Operable Operable Reactor level indication [13][20]
Converter 67 LIS-263-73A Level Indicating Switch Operable Operable Reactor level indication [131 68 LT-263-73A Level Transmitter Operable Operable Reactor level indication [13][20]
69 LIS-263-72A Reactor Water Level Operable Operable Reactor level indication [13]
Narrow Range 70 LT-263-72A Level Transmitter Operable Operable Reactor level indication [13][20]
71 PI-640-25A Reactor Pressure Operable Operable Reactor pressure indication [13]
72 PT-647A Reactor Pressure Operable Operable Reactor pressure indication [13][20]
Transmitter 73 PI-263-49A Reactor Pressure Operable Operable Reactor pressure indication [13][20]
74 Voltage Current Operable Operable Reactor pressure indication [13][20]
E/l-23-49AConverter 75 PIS-263-49A Reactor Pressure Operable Operable Reactor pressure indication [13]
76 PT-263-49A Reactor Pressure Operable Operable Reactor pressure indication [13][20]
Transmitter 77 C2205A Reactor Protection and Operable Operable Pressure and level transmitter support [13]
C2205A NSS Instrument Rack Emergency Core Cooling 78 C2233A System (ECCS) Analog Operable Operable Pressure and level transmitter support [13][15]
Trip Cabinet 79 C2251A Jet Pump Instrument Operable Operable Pressure and level transmitter support [13][14]
Rack A Page A-7
Pilgrim Nuclear Power Station ESEP Report ESEL Equipment .. Operating State References Item.Notes/Comments Number ID, "dDescription Normal Desired
____S. Sate,;C
- ~State C904 80 RWCU and Recirc. Bench Operable Operable Powered from D4, contains RCIC [5][6][15][27]
Board controls Reactor and [5][6][12][13]
81 C903 Containment Cooling Operable Operable Indications provided here [15][16][24]
Bench Board [25]
82 C905 Reactor Control Bench Operable Operable Indications provided here, powered [6][13]
Board from D6 Containment Pressure 83 C129B Switch Instrument Rack Operable Operable Indications provided here [16][21]
Switchen Insruen forRCcklo 84 C2258 RCIC Instrument Rack Operable Operable Instrument rack for RCIC flow [15][22]
transmitter 85 C930 RCIC Relay Vertical Operable Operable Contains RCIC logic. IPEEE correlated [15][23]
Board failure of C930, C932, C933.
86 C932 Channel A Vertical Operable Operable Contains RCIC logic. IPEEE correlated [12][15][23]
Board failure of C930, C932, C933.
87 C933 Channel B Vertical Operable Operable Contains RCIC logic. IPEEE correlated [15][23]
Board failure of C930, C932, C933.
88 C941 Primary Containment OperableContains isolation logic that could [6]
Isolation Relay Cabinet disable RCIC or HPCI operation 89 PIS-1001-89A Drywell Pressure Indicating Switch Operable Operable Containment pressure indication [16][21]
90 PT-1001-89A Drywell Pressure Operable Operable Containment pressure indication [16][21]
Transmitter Page A-8
Pilgrim Nuclear Power Station ESEP Report ESEL Equipment . Operating-State ., References Item... ........................
. . . Norma~ll. Desired Desired.......,.. Notes/Comments.
Numer ID ;-,Description State - ae a
91 LI-1001-604A Torus Water level Energized Energized Wide range torus water level in Phase 2 [16][28]
Indicator 92 DPT1001 - Torus Water level Energized Energized Wide range torus water level in Phase 2 [16][29]
604A Transmitter 93 Torus Water Bulk Torus bulk water temperature for Phase [16][24][25]
TI-5021-02A Temperature Energized Energized 2, powered from Y31 TRU-5021- Torus Water Processes bulk and local torus water 94 01A Torus Water Energized Energized temperatures for Phase 2, powered [16][25][26]
01A Temperature Recorder from Y31 95 TE5021-01A Torus Water Torus water temperature element for [16][251 Temperature Element Phase 2, powered by Y31 96 TE5021-02A TrsWtrEnergized Energized Trswtrem rauelmntfr [16][25]
97 TE5021-0A Torus Water Torus water temperature element for Temperature Element Energized Energized Phase 2, powered by Y31 [16][25]
97 TE5021-04A Temperture1Eement Torus Water Energized Energized Phae6,]oweedby]3 Torus water temperature element for [16][25]
Temperature Element Phase 2, powered by Y31 99 TE5021-05A Torus Water Torus water temperature element for [16][25]
Temperature Element Phase 2, powered by Y31 100 TE5021-06A Torus Water Torus water temperature element for [16][25]
Temperature Element Energized Energized Phase 2, powered by Y31 Page A-9
Pilgrim Nuclear Power Station ESEP Report ESEL Equipment Oper~ting State'.
Item Normal"" Dired.". * .Notes/Comments,. References R
Number ID Description -4 State ~State TE5021-07A 1 Torus Water Torus water temperature element for 101 ~~Temperature Element Eegzd nried Phase 2, powered by Y31 [6[5 102 TE5021-8A Torus Water EnergizedTorus water temperature element for [16][25]
Temperature Element Phase 2, powered by Y31 103 TE5021-09A Torus Water Energized ETorus water temperature element for [16][25]
Temperature Element Phase 2, powered by Y31 104 TE5021-10A Torus Water Energized Energized Torus water temperature element for [16][25]
Temperature Element Phase 2, powered by Y31 105 TE5021-11A Torus Water Energized Energized Torus water temperature element for [16][25]
Temperature Element Phase 2, powered by Y31 106 TE5021-12A Torus Water Energized ETorus water temperature element for [16][25]
Temperature Element Phase 2, powered by Y31 107 TE5021-13A Torus Water Torus water temperature element for [16][25]
Temperature Element Energized Energized Phase 2, powered by Y31 Containment Isolation Included because the controls for the 108 C7 and Ventilation Vertical Operable Operable direct torus vent path are located on [27][30]
Board this panel 109 C170 PAM panel Operable Operable Indications provided here [14][16][26]
Page A-10
Pilgrim Nuclear Power Station ESEP Report ESELqui,:..entOperating State Item.. *,, Notes/Comments References.
Number ,ID Description PASS StatioeVlv 110 C174 PASS Isolation Valve Operable Operable Indications provided here [6]
Control Panel Torus Water Torus temperature signals processed 111 C179 Temperature Signal Operable Operable here [25][32]
Processing Cabinet Inboard valve for hardened containment 112 AO-5042B Isolation Valve (inboard) Closed Open venting system. Powered from 125 VDC [17]
bus D5.
Torus Purge Exhaust 113 SV-5042B Isolation Solenoid Valve Deenergized Energized Solenoid valve for AO-5042B for DTV [17][31]
(inboard) 114 D5-15 DC Breaker Supply to C7 Energized Energized Power supply for solenoid valve for AO- [6][31]
5042B for DTV Locked Open to provide flow path for 115 AO-5025 Direct Torus Vent Closed Open containment heat removal during Phase [17][27]
- 2. Powered from 125VDC Bus D4.
116 SV-5025 Direct Torus Vent Deenergized Energized Solenoid valve for AO-5025 for DTV [17][27]
Solenoid Valve 117 D4-15 DC Breaker Supply to C7 Energized Energized Power supply for solenoid valve for AO- [6][27]
5025 for DTV N2-401A Backup Nitrogen Bottles N2-402A for operation of AO- TBD TBD Local backup nitrogen for direct torus 118 N2-401BA vent valves [3]
N2-402B 5042B and AO-5025 119 D7 125VDC Motor Control Energized Energized Power for RCIC valves [6][15]
Center for RCIC 120 72-166 Supply for D7 Closed Closed Power for RCIC valves [6]
121 D36 Extension of 125 VDC Energized Energized Power for analog trip system [6]
PanelA Page A-li
Pilgrim Nuclear Power Station ESEP Report ESEL E~quipment Operating State Item Noml~ DsrdNotes/Comments References Number ID.. : Description . .tate.
.' .State * . .*-.. State :! ,=." .,.**. .. . .. * *,,, ,State ECCS Analog Trip 122 D36-8 Cabinet C2233A Power Closed Closed Power for analog trip system [6]
Breaker 123 D4 125 VDC Distribution Energized Energized 125 DC Bus A for CSCS logic [6]
PanelA 124 72-165 Supply for D4 Closed Closed Power for D4 [6]
125 72-16A D16 Internal Breaker Closed Closed Bus continuity [6]
126 D16 125 VDC Bus A Energized Energized 125 DC Power Bus A [6]
127 72-161 Battery A Output Closed Closed Supply power to D16 [6]
Breaker 128 D29 Battery A Current Intact Intact Protective device [6]
Limiter 129 D1 125 VDC Battery Rack A Float Charge Discharge 125 VDC A power source [6]
130 72-162 Battery Charger Dl1 Closed Closed Connection for portable generator [3][6]
Supply to D16 131 D37 Extension of 125 VDC Energized Energized Power for analog trip system [6]
Panel B 132 D5 125 VDC Distribution Energized Energized 125 DC Bus B for CSCS logic [6]
PanelB 133 72-175 Supply Breaker for Bus Closed Closed Power for D5 [6]
D5 134 D17 125 VDC Bus B Energized Energized 125 VDC power Bus B [6]
135 72-17A D17 Internal Breaker Closed Closed Bus continuity [6]
Page A-12
Pilgrim Nuclear Power Station ESEP Report ES EL . Eq u pme nt . , , O pe r a tin g St a t e .. ' .. References Item " Notes/Comments References Number ID .Description . Noml Dsirted' State,< 4 Stt 136 D2 125 VDC Battery B Float Charge Discharge 125 VDC B power source [6]
137 72-171 Battery B Output Closed Closed Supply power to D17 [6]
Breaker 138 D30 Battery B Current Intact Intact Protective device [6]
Limiter 139 72-172 Battery Charger Supply to D17D12 Closed Closed Connection for portable generator [3][6]
140 Y2 Vital Services Power Energized Energized Vital bus for Indication and control [18]
Supply 141 Y12 Auto Transfer Switch for MG Set MG Set Power from vital motor-generator set to [18]
Y2 Supply Supply Y2 Vital Motor-Generator 142 EG-23 Set Operating Operating Supply power to Y2 and requires D6 [18]
143 72-1022 D10 Breaker to Vital Closed Closed Supply power to DC motor - Vital motor- [6][18]
Motor-Generator generator set 144 D10 250 VDC Power Bus Energized Energized 250 VDC power bus [6][18]
145 72-1013 250 VDC Battery D3 Closed Closed Supply power to D10 [6]
Output Breaker 146 D31 250 VDC Battery D3 Intact Intact Protective Device [6]
Current Limiter 147 D3 250 VDC Battery Float Charge Discharge 250 VDC power Source [6]
148 72-1014 Battery Charger D13 Closed Closed Connection for portable generator [6]
Supply to D10 Energized 149250 VDC Battery Energized from mobile Connection for portable generator [6]
generator Page A-13
Pilgrim Nuclear Power Station ESEP Report ESEL Equipment " -, Operating State ".
Item .. Notes/Comments , References,.:
Number ID Description State state 150 D6 125VDC Distribution Energized Energized Vital instruments and controls [6]
PanelC 151 Y1O 125VDC Control Power Operable Operable D6 normally supplied from D16 [6]
Transfer 152 D32 D16 Control Logic Y10 Switching Closed Closed D6 normally supplied from D16 [6]
D17 Control Logic Y10 153 D33 Switching Open Open D6 normally supplied from D16 [6]
Safeguard 120VAC "A" 154 Y3 Control Power Supply Energized Energized Power supply for I&C [18]
Panel Safeguard 120VAC "A" 155 Y31 H2 0 2 Control Power Energized Energized Power supply for I&C [18]
Supply Panel Energized 125 VDC Battery 156 Dl 2Chargery Energized from mobile Connection for portable generator [6]
generator Energized 157 D12 125 VDC Battery Energized from mobile Connection for portable generator [6]
Crrgenerator 158 C2257B Instrument Rack 2257B Operable Operable [5][34]
159 dPIS-1360-1A dP Switch for RCIC Operable Operable [5][34]
Steam Isolation 160 dPIS-1360-1B dP Switch for RCIC Operable Operable [5][34]
Steam Isolation 161 PS-1360-9A Pressure Switch for RCIC Operable Operable [5][341 Steam Isolation 7162 PS-1360-9B3 Pressure Switch for RCIC Operable Operable [5][34]
Steam Isolation Page A-14
Pilgrim Nuclear Power Station ESEP Report ESEL .Equipment . Operating State -.
Item. Notes/Comments References Number ID Description Normal .. *,.Desired
.State *.: .. ', State 163 PS-1360-9C Pressure Switch for RCIC Operable Operable [5][34]
Steam Isolation Oeal prbe[]3 164 PS-1360-9D) Pressure Switch for RCIC Operable Operable [5][34]
Steam Isolation OperableOperable_[5][34]
165 J315 Junction Box Operable Operable [10]
166 J317 Junction Box Operable Operable [9]
167 J602 Junction Box Operable Operable [9]
168 J599 Junction Box Operable Operable [10]
169 J600 Junction Box Operable Operable [10]
170 J601 Junction Box Operable Operable [91 171 TS-1360-14C Temperature Switch for Operable Operable [10]
RCIC Isolation 172 TS-1360-15A Temperature Switch for Operable Operable [10]
RCIC Isolation 173 TS-1360-15C Temperature Switch for Operable Operable [10]
RCIC Isolation 174 TS-1360-16C Temperature Switch for Operable Operable [10]
RCIC Isolation 175 TS-1360-16D Temperature Switch for Operable Operable [9]
RCIC Isolation Oprbeprbl__
176 TS-1360-17A Temperature Switch for Operable Operable [10]
RCIC Isolation OpeabeOerbl [_0]
Page A-15
Pilgrim Nuclear Power Station ESEP Report ESEL Equipment Operating'State . . ,
Item Number .
.Normal . "".:Notes/Comments
- , ,/:Desired -' , References' uIDr Description, :.< ,ta,,.
State State 177 TS-1360-17B Temperature Switch for Operable Operable 19]
RCIC Isolation 178 TS-1360-17C Temperature Switch for Operable Operable [10]
RCIC Isolation 179 TS-1360-17D Temperature Switch for Operable Operable 19]
RCIC Isolation 180 TS-1360-15B Temperature Switch for Operable Operable 19]
RCIC Isolation 181 TS-1360-14D Temperature Switch for Operable Operable 19]
RCIC Isolation 182 TS-1360-15D Temperature Switch for Operable Operable 19]
RCIC Isolation FLEX AC Power Transfer 183 D1513 Swtc Switch to PowerTaf D11 to Repower Operable Operable [3][6][35][36]
184 D1413 FLEX AC Power Transfer Operable Operable [3][35][36]
Switch to Repower D12 185 D1414B FLEX AC Power Transfer Operable Operable [3][6][35][36]
Switch to Repower D13 186 N17115* FLEX AC Power Transfer Operable Operable [3][36]
Switch to Repower Y3 FLEX AC Power Transfer Switch to Repower Y31 187 N17115* *Note that a single Operable Operable [3][36]
switch, N17115, repowers Y3 and Y31 Page A-16
Pilgrim Nuclear Power Station ESEP Report ATTACHMENT B - ESEP HCLPF VALUES AND FAILURE MODES TABULATION Page B-1
Pilgrim Nuclear Power Station ESEP Report Item HCP g " Failure.
No.. Equipment ID Equipment Description Screening . Mode Comments
. . .... Level 1 MO-1301-22 RCIC Pump CST Suction Valve >RLGM Screened 2 D754 DC Bus D7 Supply Breaker to >RLGM Screened MO-1301-22 3 TBD FLEX Primary External Water >RLGM Screened Note 5 Source Injection Point 4 MO-1301-25 RCIC Suction from Torus >RLGM Screened 5 D761 DC Bus D7 supply breaker to >RLGM Screened MO-1301-25 6 MO-1301-26 RCIC Suction from Torus >RLGM Screened 7 D764 DC Bus D7 supply breaker to >RLGM Screened MO-1301-26 8 P-206 RCIC Pump >RLGM Screened Note 1 9 MO-1301-49 RCIC Discharge Isolation >RLGM Screened Valve 10 D774 DC Bus D7 Supply Breaker to >RLGM Screened MO-1301-49 11 PCV 1301-43 RCIC Lube Oil Cooling Water >RLGM Screened Pressure Control Valve 12 MO-1301-62 RCIC Turbine Lube Oil Inlet >RLGM Screened Valve 13 D794 DC Bus D7 Supply Breaker to >RLGM Screened MO-1301-62 14 E-201 RCIC Barometric Condenser >RLGM Screened Note 1 and Vacuum Tank 15 E-204 RCIC Lube Oil Cooler >RLGM Screened Note 1 RCIC Vacuum Tank 16 P-221 >RLGM Screened Note 1 Condensate Pump 17 D712 DC Bus D7 Supply Breaker to >RLGM Screened P-221 18 P-222 RCIC Vacuum Pump >RLGM Screened Note 1 19 D714 DC Bus D7 Supply Breaker to >RLGM Screened P-222 20 MO-1301-61 RCIC Turbine Steam Inlet >RLGM Screened Valve Page B-2
Pilgrim Nuclear Power Station ESEP Report HCLPF (g)/ Falr, Cmet i.ia lu re ':,.::.: - .. *. ... .:. . - . : .
Fa.
Ie m Ite .
Equipment ID: ... ....
. i.Equipment Description Screening ." .:
S Comments No. Level 21 D751 DC Bus D7 Supply Breaker to >RLGM Screened MO-1301-61 22 SV-1301-1 RCIC Turbine Trip Throttle >RLGM Screened Valve 23 HO-1301-159 RCIC Turbine Governor Valve >RLGM Screened 24 X-202 RCIC Turbine >RLGM Screened Note 1 25 FT-1360-4 RCIC Pump Flow Transmitter >RLGM Screened for Turbine Control 26 SQRT1340-10 RCIC Flow Square Root >RLGM Screened Converter 27 FIC-1340-1 RCIC Flow Indicating >RLGM Screened Controller 28 DC/AC 1340-16 DC/AC Inverter for RCIC Flow >RLGM Screened Controller 29 C1303 RCIC Local Controls >RLGM Screened Note 2 30 D4-3 Supply Breaker for RCIC in >RLGM Screened Panels C904 and C939 31 13A-K1 RCIC Auto Initiation Logic >RLGM Screened RCIC Steam Supply Low Relay Pressure Function 33 13A-K11 Turbine Trip Auxiliary Relay >RLGM Screened 34 13A-K13 Pump Discharge Low Flow >RLGM Screened 35 13A-K14 Pump Suction Low Pressure >RLGM Screened 36 13A-K17 Turbine Exhaust High >RLGM Screened Pressure 37 13A-K18 MO-1301-25 Position >RLGM Screened Monitor 38 13A-K2 RCIC Auto Initiation Logic >RLGM Screened Relay 13A-K22 RCIC Auto Isolation Relay 0.33 Function 39 40 13A-K3 Pump/Turbine Room High 0.39 Relay Temperature Function Page B-3
Pilgrim Nuclear Power Station ESEP Report HCLPF (g alreCmet Nte. Equipment ID Equipmen.t Description Screening Modent Pump/Turbine Room High 0.39 Relay 41 13A-K31 Temperature Function 42 13A-K32 RCIC Valve Station High 0.39 Relay Temperature Function RCIC Steam Line High Relay Differential Pressure Function 44 13A-K34 RCIC Auto Isolation Relay >RLGM Screened 45 13A-K5 RCIC Valve Station High 0.39 Relay Temperature Function RCIC Steam Line High Relay Differential Pressure Function 47 RV-203-3B Safety Relief Valve B >RLGM Screened 48 RV-203-3C Safety Relief Valve C >RLGM Screened 49 SV-203-3B Solenoid Pilot Valve for SRV B >RLGM Screened 50 SV-203-3C Solenoid Pilot Valve for SRV C >RLGM Screened 51 T-221B Accumulator Tank for SRV B >RLGM Screened Note 2 52 T-221C Accumulator Tank for SRV C >RLGM Screened Note 2 53 PCV-203-11 Pressure Control Valve for >RLGM Screened Backup Nitrogen Supply 54 2E-SlOA ADS Inhibit Switch >RLGM Screened 55 2E-S1B SRV B Valve Actuation >RLGM Screened 56 2E-S1C SRV C Valve Actuation >RLGM Screened 57 2E-K12A CSCS Pump Running Interlock >RLGM Screened 58 2E-K12B CSCS Pump Running Interlock >RLGM Screened 59 2E-K13B SRV Power Supply Control >RLGM Screened Logic 60 2E-K13C SRV Power Supply Control >RLGM Screened Logic Page B-4
Pilgrim Nuclear Power Station ESEP Report Item Eqimn .HLF()/ Failure Cmet S Equipment D..: Equipment Description Screening. Momnodts Level 61 2E-K7A Reactor Water Low Low Level >RLGM Screened Interlock 62 2E-K7B Reactor Water Low Low Level >RLGM Screened Interlock Reactor Water Level Narrow 63 LI-263-100A Range Range >RLGM Screened 64 E/I-263-72A Voltage Current Converter >RLGM Screened 65 LI-263-106A Reactor Water Level >RLGM Screened Indicator 66 E/1-263-73A Voltage Current Converter >RLGM Screened 67 LIS-263-73A Level Indicating Switch >RLGM Screened 68 LT-263-73A Level Transmitter >RLGM Screened 69 LIS-263-72A Reactor Water Level Narrow >RLGM Screened Range 70 LT-263-72A Level Transmitter >RLGM Screened 71 PI-640-25A Reactor Pressure >RLGM Screened 72 PT-647A Reactor Pressure Transmitter >RLGM Screened 73 PI-263-49A Reactor Pressure >RLGM Screened 74 E/I-263-49A Voltage Current Converter >RLGM Screened 75 PIS-263-49A Reactor Pressure >RLGM Screened 76 PT-263-49A Reactor Pressure Transmitter >RLGM Screened 77 C2205A Reactor Protection and NSS 0.38 Functional Instrument Rack Emergency Core Cooling 78 C2233A System (ECCS) Analog Trip >RLGM Screened Note 1 Cabinet 79 C2251A Jet Pump Instrument Rack A 0.44 Anchorage 80 C904 RWCU and Recirc. Bench >RLGM Screened Note 1 Board Page B-5
Pilgrim Nuclear Power Station ESEP Report Item HCLPF (g)/ Failure Comments Nte No. Equipment ID.. .. Equipment Description Screening
.node Modeomme M*.
- Level 81 C903 Reactor and Containment >RLGM Screened Note 1 Cooling Bench Board 82 C905 Reactor Control Bench Board >RLGM Screened Note 1 Containment Pressure Switch 83 C129B Intrument rack Instrument Rack >RLGM Screened Note 1 84 C2258 RCIC Instrument Rack >RLGM Screened Note 1 Relay Nt 85 C930 RCIC Relay Vertical Board 0.33 Function ote 1 86 C932 Channel A Vertical Board >RLGM Screened Note 1 87 C933 Channel B Vertical Board 0.39 Relay Function Note 1 88 C941 Primary Containment >RLGM Screened Note I Isolation Relay Cabinet 89 PIS-1001-89A Drywell Pressure Indicating >RLGM Screened Switch 90 PT-1001-89A Drywell Pressure Transmitter >RLGM Screened 91 LI-1001-604A Torus Water level Indicator >RLGM Screened 92 DPT1001 -604A Torus Water level >RLGM Screened Note 2 Transmitter 93 TI-5021-02A Torus Water Bulk >RLGM Screened Temperature 94 TRU-5021-01A Torus Water Temperature >RLGM Screened Recorder 95 TE5021-01A Torus Water Temperature >RLGM Screened Element 96 TE5021-02A Torus Water Temperature >RLGM Screened Element 97 TES021-03A Torus Water Temperature >RLGM Screened Element 98 TE5021-04A Torus Water Temperature >RLGM Screened Element 99 TE5021-05A Torus Water Temperature >RLGM Screened Element 100 TE5021-06A Torus Water Temperature >RLGM Screened Element Page B-6
Pilgrim Nuclear Power Station ESEP Report Item HLF()/ Failure N. .Equipment ID... Equipment Description Screening M Comments Level 101 TE5021-07A Torus Water Temperature >RLGM Screened Element 102 TE5021-08A Torus Water Temperature >RLGM Screened Element 103 TE5021-09A Torus Water Temperature >RLGM Screened Element 104 TE5021-10A Torus Water Temperature >RLGM Screened Element 105 TE5021-11A Torus Water Temperature >RLGM Screened Element 106 TE5021-12A Torus Water Temperature >RLGM Screened Element 107 TE5021-13A Torus Water Temperature >RLGM Screened Element 108 C7 Containment Isolation and 0.31 Anchorage Ventilation Vertical Board 109 C170 PAM panel >RLGM Screened Note 1 110 C174 PASS Isolation Valve Control >RLGM Screened Note 1 Panel 111 C179 Torus Water Temperature >RLGM Screened Note 1 Signal Processing Cabinet 112 AO-5042B Torus Purge Exhaust Isolation >RLGM Screened Valve (inboard) 113 SV-5042B Torus Purge Exhaust Isolation >RLGM Screened Solenoid Valve (inboard) 114 D5-15 DC Breaker Supply to C7 >RLGM Screened 115 AO-5025 Direct Torus Vent >RLGM Screened 116 SV-5025 Direct Torus Vent Solenoid Vle>RLGM Screened Valve 117 D4-15 DC Breaker Supply to C7 >RLGM Screened N2-401A Backup Nitrogen Bottles for 118 N2-402A operation of AO-5042B and >RLGM Screened Note 6 N2-402B AO-5025 119 D7 125VDC Motor Control >RLGM Screened Note 1 Center for RCIC 120 72-166 Supply for D7 >RLGM Screened Page B-7
Pilgrim Nuclear Power Station ESEP Report Item. Equipment ID . . .P. .g . Failure Comments
. EquipmentD Equipment
Description:
... Screening e ComentsM.od Level.
121 D36 Extension of 125 VDC Panel A >RLGM Screened Note 2 122 D36-8 ECCS Analog Trip Cabinet >RLGM Screened C2233A Power Breaker 123 D4 125 VDC Distribution Panel A >RLGM Screened Note 2 124 72-165 Supply for D4 >RLGM Screened 125 72-16A D16 Internal Breaker >RLGM Screened 126 D16 125 VDC Bus A 0.37 Anchorage 127 72-161 Battery A Output Breaker >RLGM Screened 128 D29 Battery A Current Limiter >RLGM Screened Note 2 129 D1 125 VDC Battery Rack A 0.31 Anchorage 130 72-162 Battery Charger Dl1 Supply >RLGM Screened to D16 131 D37 Extension of 125 VDC Panel B >RLGM Screened Note 2 132 D5 125 VDC Distribution Panel B >RLGM Screened Note 2 133 72-175 Supply Breaker for Bus D5 >RLGM Screened 134 D17 125 VDC Bus B >RLGM Screened Note 1 135 72-17A D17 Internal Breaker >RLGM Screened 136 D2 125 VDC Battery B 0.30 Anchorage 137 72-171 Battery B Output Breaker >RLGM Screened 138 D30 Battery B Current Limiter >RLGM Screened Note 2 139 72-172 Battery Charger D12 Supply >RLGM Screened to D17 140 Y2 Vital Services Power Supply >RLGM Screened Note 2 141 Y12 Auto Transfer Switch for Y2 >RLGM Screened Note 1 Page B-8
Pilgrim Nuclear Power Station ESEP Report Item Eqimn DHLF()/ Failure Cmet
. Equipment ID Equipment Description Screening Comments Level HCLPF calculated 142 EG-23 Vital Motor-Generator Set 0.32 Anchorage with modifications to anchorage.
143 72-1022 D10 Breaker to Vital Motor- >RLGM Screened Generator 144 DiO 250 VDC Power Bus >RLGM Screened Note 1 145 72-1013 250 VDC Battery D3 Output >RLGM Screened Breaker 146 D31 250 VDC Battery D3 Current >RLGM Screened Note 2 Limiter 147 D3 250 VDC Battery 0.40 Anchorage 148 72-1014 Battery Charger D13 Supply >RLGM Screened to DiO 149 D13 250 VDC Battery Charger >RLGM Screened Note 1 150 D6 125VDC Distribution Panel C >RLGM Screened Note 2 151 Y Transfer Power >RLGM Screened Note 1 152 D32 D16 Control Logic Y10 >RLGM Screened Note 2 Switching 153 D33 D17 Control Logic Y10 >RLGM Screened Note 2 Switching 154 Y3 Safeguard 120VAC "A" >RLGM Screened Note 1 Control Power Supply Panel 155 Y31 Safeguard 120VAC "A" H202 >RLGM Screened Note 2 Control Power Supply Panel 156 D1 125 VDC Battery Charger A >RLGM Screened Note 1 157 D12 125 VDC Battery Charger B >RLGM Screened Note 1 158 C2257B Instrument Rack 2257B >RLGM Screened Note 1 159 dPIS-1360-1A dP Switch for RCIC Steam >RLGM Screened Isolation 160 dPIS-1360-1B dP Switch for RCIC Steam >RLGM Screened Isolation 161 PS-1360-9A Pressure Switch for RCIC >RLGM Screened Steam Isolation Page B-9
Pilgrim Nuclear Power Station ESEP Report IHCLPF (g)/ -
No. Equipment ID Equipment Description screenings Failure Comments Item.. Level' Md 162 PS-1360-9B3 Pressure Switch for RCIC >RLGM Screened Steam Isolation 163 PS-1360-9C Pressure Switch for RCIC >RLGM Screened Steam Isolation 164 PS-1360-9B Pressure Switch for RCIC >RLGM Screened Steam Isolation 165 J315 Junction Box >RLGMV Screened Note 2 166 J317 Junction Box >RLGMV Screened Note 2 167 J602 Junction Box >RLGM Screened Note 1 168 J599 Junction Box >RLGM Screened Note 2 169 J600 Junction Box >RLGM Screened Note 1 170 J601 Junction Box >RLGM Screened Note 2 171 TSJ136004C Temperature Switch for RCIC >RLGM Screened Isolation 172 TS-1360-15A Temperature Switch for RCIC >RLGM Screened Isolation 173 TS-1360-15C Temperature Switch for RCIC >RLGM Screened Isolation 174 TS-1360-16C Temperature Switch for RCIC >RLGM Screened Isolation 175 TS-1360-16D Temperature Switch for RCIC >RLGM Screened Isolation 176 TS-1360-17A Temperature Switch for RCIC >RLGM Screened Isolation 177 TS-1360-17B Temperature Switch for RCIC >RLGM Screened Isolation 178 TS-1360-17C Temperature Switch for RCIC >RLGM Screened Isolation 179 TS-1360-17D Temperature Switch for RCIC >RLGM Screened Isolation 180 TS-1360-15B Temperature Switch for RCIC >RLGM Screened Isolation 181 TS-1360-14D Temperature Switch for RCIC >RLGM Screened Isolation Page B-10
Pilgrim Nuclear Power Station ESEP Report Item HCLPF (g)/ Failure Comments N. Equipment ID Equipment Description Screening Comments
. ......... Level 182 TS-1360-15D Temperature Switch for RCIC >RLGM Screened Isolation 183 D1513 FLEX AC Power Transfer >RLGM Screened Note 3 Switch to Repower D11 FLEX AC Power Transfer 184 D1413 FLEX C IPower Taf >RLGM Screened Note 3 Switch to Repower D12 185 D1414B FLEX AC Power Transfer >RLGM Screened Note 3 Switch to Repower D13 FLEX AC Power Transfer 186 N17115* FLEX ACto Repower Switch PowerT Y3 s >RLGM Screened Note 4 FLEX AC Power Transfer Switch to Repower Y31 187 N17115* *Note that a single switch, >RLGM Screened Note 4 N17115, repowers Y3 and Y31 Notes:
- 1. Anchorage screened out based on available margin during walkdown by SRT.
- 2. Anchorage screened out during walkdown validation by SRT.
- 3. This component is evaluated in Entergy Calculation C15.0.3623 [48] for 2xSSE.
- 4. This component is evaluated in Entergy Calculation C15.0.3624 [49] for 2xSSE.
- 5. Entergy document no. EC-0000042259 [36] is reviewed. Design input 59 within this document states that seismic demand is based on 2xSSE. Thus, this component, which is yet to be installed, is screened for ESEP based on seismic design criteria.
- 6. This component is evaluated in Entergy Calculation C15.0.3631 [50] for 2xSSE.
Page B-11
ATTACHMENT 2 to PNPS Letter 2.14.082 LIST OF REGULATORY COMMITMENTS FOR PILGRIM NUCLEAR POWER STATION to PNPS Letter 2.14.082 Page 1 of 1 List of Regulatory Commitments The following table identifies those actions committed to by Entergy in this document. Any other statements in this submittal are provided for information purposes and are not considered to be regulatory commitments.
TYPE (Check One) SCHEDULED COMMITMENT ONE- COMPLETION DATE ONE CONTINUING (If Required)
TIME ACTION COMPLIANCE Modify Vital MG Set EG-23 anchorage.
Vital MG Set EG-23 anchorage had a *On a schedule High Confidence of a Low Probability of specified in Section Failure (HCLPF) capacity below the 8.4 of the Expedited Review Level Ground Motion (RLGM). A [] Seismic Evaluation modification is planned to provide Process Report additional seismic margin such that the HCLPF will exceed RLGM. May 31, 2017 NRC Commitment No. A16866 Submit a letter to NRC summarizing the Within 60 days HCLPF results of item 1 above confirming following completion implementation of the plant modification of ESEP activities, associated with item 1. including item 1 above.
NRC Commitment No. A16867 July 30, 2017
- 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 to the Vital MG-Set EG-23 will be done during a planned refueling outage.