Text

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.
	ML14353A332
Person / Time
Site:	Oyster Creek
Issue date:	12/19/2014
From:	Jim Barstow; Exelon Generation Co
To:	; Document Control Desk, Division of Operating Reactor Licensing
References
	RA-14-083, RS-14-299
	Download: ML14353A332 (45)
	v • d • e

Exelon Generation RS-14-299 RA-14-083 December 19, 2014 U.S. Nuclear Regulatory Commission Attn: Document Control Desk 11555 Rockville Pike, Rockville. MD 20852 Oyster Creek Nuclear Generating Station Renewed Facility Operating License No. DPR-16 NRG Docket No. 50-219 10 CFR 50.54(f)

Subject:

Exelon Generation Company, LLC Expedited Seismic Evaluation Process Report (CEUS Sites), Response to NRG Request for Information Pursuant to 1 O CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident

References:

1. NRG 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, Proposed Path Forward for NTTF Recommendation 2.1: Seismic Re-evaluations, dated April 9, 2013 (ML13101A379)

3. Seismic Evaluation Guidance: "Augmented Approach for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1 -Seismic", EPRI, Palo Alto, CA: May 2013.3002000704(ML13102A142)

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

5. Exelon Generation Company, LLC, Seismic Hazard and Screening Report (Central and Eastern United States (CEUS) Sites), Response to NRG 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 (RS-14-070), dated March 31, 2014 (ML14090A241)

6. Exelon Generation Company, LLC Response to NRG Request for Information Pursuant to 1 O 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 - 1.5 Year Response for CEUS Sites (RS-13-205), dated September 12, 2013 (ML13256A070)

U.S. Nuclear Regulatory Commission NTTF 2.1 Seismic Response for CEUS Sites December 19, 2014 Page2 On March 12, 2012, the Nuclear Regulatory Commission (NRC) issued a 50.54(f) letter to all power reactor licensees and holders of construction permits in active or deferred status. of Reference 1 requested each addressee located in the Central and Eastern United States (CEUS) to submit a Seismic Hazard Evaluation and Screening Report within 1.5 years from the date of Reference 1.

In Reference 2, the Nuclear Energy Institute (NEI) requested NRC agreement to delay submittal of the final CEUS Seismic Hazard Evaluation and Screening Reports so that an update to the Electric Power Research Institute (EPRI) ground motion attenuation model could be completed and used to develop that information. NEI proposed that descriptions of subsurface materials and properties and base case velocity profiles be submitted to the NRC by September 12, 2013, (Reference 6) with the remaining seismic hazard and screening information submitted by March 31, 2014 (Reference 5). NRC agreed with that proposed path forward in Reference 4.

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 enclosed Expedited Seismic Evaluation Process (ESEP) Report for Oyster Creek Nuclear Generating Station provides the information described in the "ESEP Report" Section 7, of Reference 3 in accordance with the schedule identified in Reference 2.

All equipment evaluated for the ESEP for Oyster Creek Nuclear Generating Station was found to have adequate capacity for the required seismic demand as defined by the Augmented Approach (ESEP) guidance (Reference 3). Therefore, no equipment modifications are required.

This ESEP report transmittal completes regulatory Commitment No. 1 of Reference 5.

No new regulatory commitments result from this transmittal.

If you have any questions regarding this report, please contact Ron Gaston at (630) 657-3359.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 191h day of December 2014.

Respectfully submitted, James Barstow Director - Licensing & Regulatory Affairs Exelon Generation Company, LLC

Enclosure:

Oyster Creek Nuclear Generating Station Expedited Seismic Evaluation Process (ESEP)

Report

U.S. Nuclear Regulatory Commission NTTF 2.1 Seismic Response for CEUS Sites December 19, 2014 Page 3 cc:

Director, Office of Nuclear Reactor Regulation Regional Administrator - NRC Region I NRC Senior Resident Inspector - Oyster Creek Nuclear Generating Station NRC Project Manager, NRR - Oyster Creek Nuclear Generating Station Mr. Nicholas J. DiFrancesco, NRR/JLD/JHMB, NRC Manager, Bureau of Nuclear Engineering - New Jersey Department of Environmental Protection Mayor of Lacey Township, Forked River, NJ

Enclosure Oyster Creek Nuclear Generating Station Expedited Seismic Evaluation Process (ESEP) Report (41 pages)

EXPEDITED SEISMIC EVALUATION PROCESS (ESEP) REPORT IN RESPONSE TO THE S0.54(f) INFORMATION REQUEST REGARDING FUKUSHIMA NEAR-TERM TASK FORCE RECOMMENDATION 2.1: SEISMIC for the Oyster Creek Generating Station Route 9 South P.O. Box 388 Forked River, New Jersey 08731 Faclllty Operating License No. DPR-16 NRC Docket No. STN 50-219 Correspondence No.: RS-14*299

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Stevenson & Associates 275 Mishawum Road, Suite 200 Woburn, MA 01801 Report Number: 1404241-RPT-003, Rev. 5 Printed Name Preparer:

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S&A Report 1404241-RPT-003 Rev. 5 Correspondence No. RS-14-299 EXPEDITED SEISMIC EVALUATION PROCESS (ESEP) IN RESPONSE TO THE 50.54(f)

INFORMATION REQUEST REGARDING FUKUSHIMA NEAR-TERM TASK FORCE RECOMMENDATION 2.1: SEISMIC for the OYSTER CREEK GENERATING STATION Document Type:

Criteria 0 Interface 0 Report [;8J Specification 0 Other 0 Drawing 0 Project Name: Exelon ESEP for Oyster Creek Job No. : 1404241 Client: ~$1 Exelon This document has been prepared in accordance with the S&A Quality Assurance Program Manual, Revision _1l_ and project requirements:

Initial Issue (Rev. 0)

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Date: 11/26/2014 Approved by: Walter Diordievic trJ !J1fl Date: 11 /26/2014 Revision Record:

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' DOCUMENT APPROVAL SHEET CONTRACT NO.

1404241 Stevenson & Associates Page 2 of 41

Contents S&A Report 1404241-RPT00'.3 Rev. 5 Correspondence No. HS-14-29D

1.0 Purpose and Objective................. **--........ ***--***........................... "............ ***-*-- 6 2.0 Brief SurnnH:lry of tl1e FLEX Seisrnic lrnplernentation Strategies.................................... 7 3.0 Equipment Selection Process and ESEL, Alternate Path Justifications, and Determination of the Reduced ESEL........................

.......................................................... 9 3.1 Equipment Selection Process and ESEL.................................................................... 9 3.1.*1 ESEL Development.............................................................................................. 11 3.1.2 Power Operated Valves...................................................................................... 12 3.1.3 Pull Boxes............................................................................................................ 12 3.-1.4 Termination Cabinets.......................................................................................... *12 3.1.5 Critical Instrumentation Indicators_....................................................................... i 2 3.1.6 Phase 2 and Phase 3 Piping Connections........................................................... 13 3.2 Justification for use of Equipment that is not the Primary Means for FLEX Implementation........................................................................................................... 13 3.3 Determination of the Reduced ESEL.......................................................................... 13 4.0 Ground Motion Response Spectrum (GMRS)................................................................. 15

4. I Plot of GMRS Submitted by the Licensee................................................................... 15 4.2 Comparison to SSE.................................................................................................... 16 5.0 Review Level Ground Motion (RLGM)........................................................................... 18 5.1 Description of HLGM selected..................................................................................... 18 5.2 Method to Estimate ISRS.................................................................................................... 20 6.0 Seismic Margin Evaluation Approach.............................................................................. 21

6. '!

Surnmary of Methodologies Used................................................................................ 21 6.2 HCLPF Screening Process.............................................................................................. 21 6.3 Seismic Wa!kdown Approach.......................................................................................... 22 6.3.1 Walkdown Approach.................................................................................................. 22 6.3.2 Application of Previous Walkdown Information.................................................... 23 6.3.3 Significant Walk down Findings.............................................................................. 23 6A HCLPF Calculation Process....................................................................................... 24 6.5 Functional evaluation of re!ays..................................................................................... 26 6.6 Tabulated ESEL HCLPF Values (Including Key Failure Modes).................................. 26 7.0 Inaccessible ltems.......................................................................................................... 26 7.1 Identification of ESEL Items Inaccessible for Walkdowns............................................26 7.2 Planned Walkdown I Evaluation Schedule I Close Out............................................... 26 8.0 ESEP Conclusions and Results.................................................................................... 27 8.1 Supporting Information.............................................................................................. 27 Page 3 of 4*1

S&A Report 1404241-RPT-003 Rev. 5 Correspondence No. RS-14-299 8.2 Summary of ESEP Identified and Planned Modifications............................................ 28 8.3 Modification Implementation Schedule........................................................................ 28 8.4 Summary of Regulatory Commitments........................................................................ 28 9.0 References..................................................................................................................... 29 Attachment A: Oyster Creek ESEL............................................................................................ 31 Attachment B: Oyster Creek Reduced ESEL............................................................................ 38 Attachment C: ESEL HCLPF Values and Failure Modes Tabulation.........................................40 Page 4 of 41

List of Tables S&A Report 1404241-RPT-003 Rev. 5 Correspondence No. RS-14-299 Table 3-1: Flow Paths Credited for ESEP................................................................................. 11 Table 4-1: Oyster Creek GMRS (5% Damping)......................................................................... 15 Table 4-2: Oyster Creek GMRS vs. SSE (5% Damping)........................................................... 16 Table 5-1: Ratio Between GMRS AND SSE (5% Damping)...................................................... 18 Table 5-2: RLGM (5% Damping)............................................................................................... 19 Table 6-1: HCLPF Calculation Summary................................................................................... 25 List of Figures Figure 4-1: Oyster Creek GMRS Plot (5% Damping)................................................................ 16 Figure 4-2: Oyster Creek GMRS vs. SSE Plot (5% Damping)................................................... 17 Figure 5-1: Plot of RLGM (5% Damping)................................................................................... 19 Page 5 of 41

1.0 Purpose and Objective S&A Report 1404241-RPT-003 Rev. 5 Correspondence No. RS-14-299 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 Oyster Creek Generating Station. The intent of the ESEP is to perform an interim action in response to the NRC's 50.54(f) letter [1] to demonstrate seismic margin through a review of a subset of the plant equipment that can be relied upon to protect the reactor core following beyond design basis seismic events.

The ESEP is implemented using the methodologies in the NRC endorsed guidance in EPRI 3002000704, Seismic Evaluation Guidance: Augmented Approach for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1: Seismic [2]. EPRI 3002000704 also contains a scope reduction allowance for low seismic hazard sites with a ground motion response spectrum (GMRS) that only exceeds the safe shutdown earthquake (SSE) at low frequencies. Section 4 of the Oyster Creek Seismic Hazard and Screening Report [4] presents justification for classifying the plant as a low seismic hazard site as well as a discussion of the GMRS to SSE exceedance at low frequencies. The allowed reduction in scope will limit the ESEP to equipment items with potential susceptibility to damage from spectral accelerations at low frequencies.

The objective of this report is to provide summary information describing the ESEP evaluations and results. The level of detail provided in the report is intended to enable NRC to understand the inputs used, the evaluations performed, and the decisions made as a result of the interim evaluations.

Page 6 of 41

S&A Report 1404241-RPT-003 Rev. 5 Correspondence No. RS-14-299 2.0 Brief Summary of the FLEX Seismic Implementation Strategies The Oyster Creek FLEX response strategies to maintain Core Cooling, Containment, Spent Fuel Pool Cooling, and Safety Function Support are summarized below. This summary is derived from the Oyster Creek Overall Integrated Plan (OIP), including all 6 month FLEX updates through August 2014, in Response to the March 12, 2012, NRC Order EA-12 049 [3].

Flex Phase 1, 1.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months , strategy relies on installed plant equipment. The Reactor will automatically isolate maintaining RPV inventory. Reactor Core Cooling, and Decay Heat Removal is achieved through the Isolation Condenser System (ICS). The ICS is comprised of two heat exchangers. The ICS is placed into service by opening a single DC powered condensate return valve in each system. The condensate return valves open automatically and are then manually cycled to limit RPV cool down rate. The ICS removes decay heat and deposits it into the environment and not into Containment. The ICS is a closed loop system; RPV inventory is not lost due to ICS operation. Oyster Creek is a hot shutdown design and as long as water is supplied to the ICS shells coping can extend indefinitely. The ICS can provide decay heat removal for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months 40 minutes without makeup water being provided to the condenser shells. Although not credited in the FLEX time line off site redundant fire diesels can provide water to the ICS shells if the fire system was not damaged in the initiating event. The fire protection system is considered a defense in-depth system, use if available, but does not affect the FLEX primary strategy.

Key Reactor Parameters are obtained via DC powered and locally installed instrumentation. A DC load shedding strategy is employed to extend battery life.

No specific Containment Control is required in Phase 1 as both temperature and pressure stay within design limits for the first 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months of the event. Key containment parameters are obtained from DC powered instrumentation or from locally installed gauges.

No specific Spent Fuel Pool control is required in Phase 1 as both temperature and level stay within design limits for the first 14.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months of the event. Spent Fuel Pool level is obtained from the new Spent Fuel Pool wide range instrumentation installed under order EA-12-051 [20].

No specific Safety Function Support actions are required during phase 1.

Flex Phase 2, 1.5 to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , strategy relies on installed plant equipment and portable equipment.

Core Cooling is ensured by providing water to the ICS shells within 1.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months . Makeup water is supplied using the FLEX pump taking suction from the intake or discharge canal. Reactor Inventory control is managed using a connection to Core Spray System I and uses the same FLEX pump that provides makeup water to ICS shells. The ICS system reduces reactor pressure to the point that the low pressure FLEX pump can inject into the RPV. Reactor Inventory loss and containment energy addition are from reactor recirculation pump seal leakage and unidentified leakage, with the major contributor being recirculation pump seal leakage.

During phase 2, electrical power is restored at 2.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months . A portable 500KW 480VAC diesel generator is installed at 480 VAC Unit Substations USS 1 A2 or USS 182. This re-powered USS will provide power to Battery chargers, ICS MOVs, and Control Rod Drive (CRD) pump if the Condensate Storage Tank (CST) is available. The use of the CRD pump is a FLEX defense in-Page 7 of 41

S&A Report 1404241-RPT-003 Rev. 5 Correspondence No. RS-14-299 depth strategy. The CST is not a protected or seismically qualified water source but if available, would provide a clean high pressure injection source to the RPV.

Electrical power is used to isolate reactor recirculation pumps, limiting RPV losses and energy addition to the containment from recirculation pump seal leakage. This restored power is also used to re-power station battery chargers ensuring the continued availability of DC power to provide critical instrumentation and DC valve operation.

Key Reactor Parameters are obtained via DC powered instrumentation or via the 500KW 480 VAC generator to re-power Motor Control Centers (MCCs) required to provide additional instrumentation.

No specific Containment Control is required in Phase 2 as both temperature and pressure stay within design limits for the first 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months of the event. Key Containment Parameters are obtained from DC powered instrumentation, local gauges, or instrumentation re-powered from the FLEX generator.

Spent Fuel Pool control is required in Phase 2. At 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months a connection from the FLEX pump will be made to the Spent Fuel Pool Cooling (SFPC) systems existing 8.5.b connection to provide makeup water to the fuel pool. An alternate SFPC strategy is to provide a water spray directly to the fuel pool on the refueling floor.

Spent Fuel Pool level is obtained from the new Spent Fuel Pool wide range instrumentation installed under order EA-12-051 [20).

Safety Functions Support strategies in phase 2 include the control room, battery room, and refuel floor habitability. The strategies include opening of doors and roof hatches, and the use of portable fans and blowers, to provide ventilation to affected areas.

Flex Phase 3, hour 24 to 72, strategy relies on installed plant equipment and portable equipment.

Phase 1 and 2 strategy will provide sufficient capability that no additional Phase 3 strategies are required. Phase 3 equipment for Oyster Creek includes backup portable pumps and generators.

The portable pumps will be capable of providing the necessary flow and pressure as outlined in Phase 2 response for Core Cooling & Decay Heat Removal, RCS Inventory Control and Spent Fuel Pool Cooling. The portable generators will be capable of providing the necessary 480 VAC power requirements as outlined in Phase 2 response for Safety Functions Support.

Page 8 of 41

S&A Report 1404241-RPT-003 Rev. 5 Correspondence No. RS-14-299 3.0 Equipment Selection Process and ESEL, Alternate Path Justifications, and Determination of the Reduced ESEL The selection of equipment for the Expedited Seismic Equipment List (ESEL) followed the guidelines of EPRI 3002000704 [2]. Per the EPRI guidance, a full ESEL [17] was first developed without considering the allowed reduction for low seismic hazard sites having only low frequency exceedance of GMRS to SSE (<2.5 Hz). The full ESEL for Oyster Creek is presented in Attachment A and the reduced ESEL is presented in Attachment B. Section 3.1 and 3.2 of this report detail the selection process for the full ESEL while the selection process for the reduced ESEL is discussed in Section 3.3.

3.1 Equipment Selection Process and ESEL The selection of equipment 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 Oyster Creek Overall Integrated Plan (OIP) in Response to the March 12, 2012, Commission Order EA-12-049 [3]. The OIP, including 6 month updates through August 2014, provides the Oyster Creek FLEX mitigation strategy and serves as the basis for the 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 Oyster Creek OIP (3]. 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, subcriticality, and containment integrity functions. Portable and pre-staged FLEX equipment (not permanently installed/anchored) are excluded from the ESEL per EPRI 3002000704 (2].

The ESEL component selection followed the EPRI guidance outlined in Section 3.2 of EPRI 3002000704 [2].

1. The scope of components is limited to those required to accomplish the core cooling and containment safety functions identified in Table 3-1 of EPRI 3002000704 [2]. The instrumentation monitoring requirements for core cooling/containment safety functions are limited to those outlined in the EPRI 3002000704 [2] guidance, and are a subset of those outlined in the Oyster Creek OIP [3].

2. The scope of components is limited to installed plant equipment, and FLEX connections necessary to implement the Oyster Creek OIP [3] as described in Section 2.

3. The scope of components assumes the credited FLEX modifications, including connections, 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 if used.

5. Phase 3 coping strategies are included in the ESEP scope, whereas recovery strategies are excluded.

Page 9 of 41

S&A Report 1404241-RPT-003 Rev. 5 Correspondence No. RS-14-299

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 and rupture disks.

Power-operated valves not required to change state as part of the FLEX mitigation strategies.

Nuclear steam supply system components (e.g. reactor pressure vessel and internals, reactor coolant pumps and seals, etc.)

7. For cases in which neither train was specified as a primary or back-up strategy, then only one train component (generally 'A' train) is included in the ESEL.

Page 10 of 41

3.1.1 ESEL Development S&A Report 1404241-RPT-003 Rev. 5 Correspondence No. RS-14-299 The ESEL was developed by reviewing the Oyster Creek OIP, including all 6 month FLEX updates through August 2014, [3] to determine the major equipment involved in the FLEX strategies. Further reviews of plant drawings (e.g., Process and Instrumentation Diagrams (P&IDs) and Electrical One Line Diagrams) were performed to identify the boundaries of the flow paths to be used in the FLEX strategies and to identify specific components in the flow paths needed to support implementation of the FLEX strategies. Boundaries were established at an electrical or mechanical isolation device (e.g., isolation amplifier, valve, etc.) in branch circuits I branch lines off the defined electrical or fluid flow path. P&IDs were the primary reference documents used to identify mechanical components and instrumentation. The flow paths used for FLEX strategies were selected and specific components were identified using detailed equipment and instrument drawings, piping isometrics, electrical schematics and one-line diagrams, system descriptions, and design basis documents.

The flow paths credited for the Oyster Creek ESEP are shown in Table 3-1 below.

Table 3-1: Flow Paths Credited for ESEP Flow Path i

FLEX Drawing P&IDs Steam from the Reactor Pressure Vessel to the FLEX Second Six-Emergency Condensers and condensate from the Month Status Report GE 148F262 Sh. 1 (19.1]

Emergency Condensers to the Reactor Recirculation (3.3]

GE 237E798 [19.2]

Piping Make up coolant from the Ultimate Heat Sink to the FLEX Second Six-Emergency Condenser Secondary Side via a FLEX Month Status Report GE 148F262 Sh. 1 [19. 1]

pump and resulting steam vented to Atmosphere [3.3]

RPV/RCS make up coolant from the Ultimate Heat FLEX Second Six-Month Status Report GE 8850781 Sh. 1 [19.3]

Sink to Core Spray System via Flex pump connection [3.3]

Drywell and Torus Hardened Containment Ventilation GU 3E 243-21-1000 Sh. 1 [19.4]

None BR 2011 Sh. 2 (19.5]

System, vents structures to atmosphere SN 13432.19-1 Sh.1 (19.6]

Coolant from the Ultimate Heat Sink to Containment FLEX Second Six-Spray system via FLEX pump connection to control Month Status Report GE 148F740 Sh. 1 [19.7]

ContainmenUDrywell pressure [3.3]

Isolation of the Reactor Recirculation Pump seals to None GE 237E798 [19.2]

minimize RPV/RCS leakage Fuel Oil from the Diesel Generator Fuel Oil Tank to FLEX Second Six-the FLEX Connection Point Month Status Report GU 3E-862-21-1000 Sh. 1 [19.8] [3.3]

Page 11 of 41

3.1.2 Power Operated Valves S&A Report 1404241-RPT-003 Rev. 5 Correspondence No. RS-14-299 Page 3-3 of EPRI 3002000704 [2] notes that power operated valves not required to change state are excluded from the ESEL. Page 3-2 also notes that "functional failure modes of electrical and mechanical portions of the installed Phase 1 equipment should be considered (e.g. RCIC/AFW trips)". To address this concern, the following guidance is applied in the Oyster Creek ESEL for functional failure modes associated with power operated valves:

Power operated valves that must remain energized during the Extended Loss of all AC Power (ELAP) events in order to maintain a credited FLEX flow path or pressure boundary (such as DC powered solenoid-operated valves), were included on the ESEL.

Power operated valves not required to change state as part of the FLEX mitigation strategies were not included on the ESEL. The seismic event also causes the ELAP event; therefore, the valves are incapable of spurious operation as they would be de-energized.

Power operated valves not required to change state as part of the FLEX mitigation strategies during Phase 1, and are re-energized and operated during subsequent Phase 2 and 3 strategies, were not evaluated for spurious valve operation as the seismic event that caused the ELAP has passed before the valves are re-powered.

3.1.3 Pull Boxes Pull boxes were deemed unnecessary to add to the ESEL as these components provide completely passive locations for pulling or installing cables. No breaks or connections in the cabling are included in pull boxes. Pull boxes were considered part of conduit and cabling, which are excluded in accordance with EPRI 3002000704 [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 and the connections are excluded from the ESEL.

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).

Page 12 of 41

3.1.6 Phase 2 and Phase 3 Piping Connections S&A Hc1porl. 1404241 **HPT003 Ftev 5 Corre~;pon<Jence No RS* '14--299 Item 2 in Section 3. *1 above notes that the scope of equipment in the includes ... FLEX connections necessary to irnplement the Oyster Creek OIP [3] as described in Section 2."

Item 3 in Section 3. *1 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 goes on to explain that "Piping, cabling, conduit, HVAC, and their supports" are excluded from the ESEL scope in accordance with EPRI 3002000704 [2J.

There-fore, piping and pipe supports associated with FLEX Pl1ase 2 and Phase 3 connections are excluded from the scope of the ESEP evaluation. However, any active valves in the 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 All equipment used for FLEX implementation on the Oyster Creek ESEL are primary path.

3.3 Determination of the Reduced ESEL EPRI 3002000704 [2] contains an ESEL reduction allowance for plants qualifying as low seismic llazard sites under Section 3.2.1. I of EPRI 1025287 (14]. This provision allows qualifying plants to limit the ESEL to equipment that are potentially susceptible to damage from spectral accelerations at low frequencies. Section 4 of the Oyster Creek Seismic Hazard and Screening Report [4] presents justification for classifying the plant as a low seismic hazard site. An excerpt from the Oyster Creek Seismic Hazard and Screening Report [4] is shown below. Refer to Section 4.0 for plots of the Oyster Creek SSE and GMRS..

In the frequency range of 1 to 1 OHz, the Oyster Creek Nuclear Generating Station (OCNGS) SSE spectral acceleration exceeds tfJat of the GMRS except for frequencies below approximately 1.9 Nz [4]. According to the Screening, Prioritization and Implementation Details (SPID), Section J.1 the OCNGS SSE exceedances of the GMRS in the frequency range of'/ to 101-lz are classified as low-frequency exceedances. Fwthe1: the GMRS spectra! acceleration does not exceed the low lwzard threshold of 0. 4g peak spectral acceleration. For most Structures, Systems, and Components (SSCs), exceec.1ances below 2.5 Hz are non-consequential as the fundamental frequency of these SS Cs exceeds 2. 5 Hz. Because of this and the low likelihood of any seismically designec1 SSC being damagc-;d by ground motion with a peak spectral acceleration less than the tow hazard threshold, the expected seismic risk at OCNGS is low [14]. As a result~ the SPID, Section 3.2.1.1 [14] limits the seismic risk assessment to evaluation of safety-significant SSCs that are potentially suscepti/1/e to ground motions at frequencies less than 1.9 Hz for OCNGS.

Examples of SSCs and failure mocfes potentially susceptible to damage from spectral accelerations at low frequencies are provided in the SP!O, Section 3.2.1.1 [14] and reproduced below. Based upon furtf1er review of equipment natural frequencies, an additional component type was identified as potentially susceptible to low frequency Page 13 of 41

S&A Report 1404241-RPT-003 Rev. 5 Correspondence No. RS-14-299 acceleration: equipment mounted on vibration isolators. The SSC and failure mode types, along with examples of specific potentially safety-significant OCNGS SSCs, are listed below.

Liquid sloshing in atmospheric pressure storage tanks o

Diesel generator fuel oil storage tank, T-39-2 o

Condensate storage tank, T-11-1 Very flexible distribution systems with frequencies less than 1.9 Hz o

Cable tray raceways o

Conduit raceways o

Flexible piping systems Sliding and rocking of unanchored components o

Emergency diesel generators, M-39-001 and M-39-002 o

Fire water pump house (controlling failure mode is sliding)

Fuel assemblies inside the reactor vessel Soi/ liquefaction o

Emergency diesel generator building o

Turbine building o

Fire water buried piping Equipment mounted on vibration isolators o

Batt & M-G room exhaust and supply fans, EF-1-20 and SF-1-20 o

Switchgear room "A" main exhaust and supply fans, FN-56-4 and FN-56-7 The above Structures, Systems, and Components (SSCs) were compared against the full ESEL presented in Attachment A. The only overlapping item is the diesel generator fuel oil storage tank, tag number T-39-2. Per Section 2.2.1.1 of EPRI 3002000704 [2] the ESEL is therefore reduced to only include tank T-39-2, as presented in Attachment 8.

Page 14 of 41

4.0 Ground Motion Response Spectrum (GMRS) 4.1 Plot of GMRS Submitted by the Licensee S&A Report 1404241-RPT-003 Rev. 5 Correspondence No. RS-14-299 In accordance with Section 2.4.2 of the Screening, Prioritization and Implementation Details (SPID) (14], the licensing design basis definition of the SSE control point for Oyster Creek is used for comparison to the GMRS. The Oyster Creek March 31, 2014 Submittal [4] states that the site SSE, anchored to a peak ground acceleration (PGA) of 0.184g, is defined at elevation 23 feet.

The GMRS, taken from the Oyster Creek March 31, 2014 Submittal report [4] is shown in Table 4-1 and Figure 4-1.

Table 4-1: Oyster Creek GMRS (5% Damping)

Freq. (Hz)

GMRS (unscaled, g) 1 0.168 1.25 0.196 1.5 0.220 2

0.256 2.5 0.270 3

0.296 3.5 0.312 4

0.320 5

0.328 6

0.311 7

0.297 8

0.286 9

0.275 10 0.266 Page 15 of 41

S&A Report 1404241-RPT-003 Rev. 5 Correspondence No. RS-14-299 Figure 4-1: Oyster Creek GMRS Plot (5% Damping)

0. 70 1--------*--------0----------->----c-----+--*----;----~-

§ 0.60 c:

0

~ 0.50 Cl)

Qi 8 0.40

<(

0.30 0.20 0.10 0.00 4.2 Comparison to SSE 10 Frequency (Hz)

GMRS As identified in the March submittal report, the GMRS only exceeds the SSE below 1.9 Hz within the 1-10 Hz range. A comparison of the GMRS to the SSE between 1-10Hz is shown in Table 4-2 and Figure 4-2. Per EPRI 3002000704 (2], low-frequency GMRS exceedances (below 2.5 Hz) at low seismic hazard sites do not require a plant to perform a full ESEP.

Table 4-2: Oyster Creek GMRS vs. SSE (5% Damping)

Freq. (Hz)

GMRS (unscaled, g)

Horizontal SSE (g) 1 0.168 0.110 1.25 0.196 0.150 1.5 0.220 0.190 2

0.256 0.270 2.5 0.270 0.290 3

0.296 0.360 3.5 0.312 0.390 4

0.320 0.410 5

0.328 0.440 6

0.311 0.430 7

0.297 0.420 8

0.286 0.390 9

0.275 0.370 10 0.266 0.360 Page 16 of 41

S&A Report 1404241-RPT-003 Rev. 5 Correspondence No. RS-14-299 Figure 4-2: Oyster Creek GMRS vs. SSE Plot (5% Damping) 1.00 --------*---------------------

0.90 1---------------~.*----"-~-~--f--+~~

0.80 1-------~-------_,__*~--~-.,---'--i----c

0. 70 <--*-------~--*---*-----~----~-~---+--~-

Oi

'C' 0.60 0

~ 0.50 Ql

~ 0.40

<(

0.30 0.20 0.10 0.00 10 Frequency (Hz)

GMRS

--- SSE Page 17 of 41

5.0 Review Level Ground Motion (RLGM) 5.1 Description of RLGM selected S&A Report 1404241-RPT-003 Rev. 5 Correspondence No. RS-14-299 The RLGM for Oyster Creek was determined in accordance with Section 4 of EPRI 30020000704 [2] by linearly scaling the SSE by the maximum GMRS/SSE ratio between the 1 and 10 Hz range. This calculation is shown in Table 5-1.

Table 5-1: Ratio Between GMRS AND SSE (5% Damping)

Freq. (Hz)

GMRS (unscaled, g)

Horizontal SSE lol GM RS/SSE 1

0.168 0.110 1.53 1.25 0.196 0.150 1.31 1.5 0.220 0.190 1.16 2

0.256 0.270 0.95 2.5 0.270 0.290 0.93 3

0.296 0.360 0.82 3.5 0.312 0.390 0.80 4

0.320 0.410 0.78 5

0.328 0.440 0.75 6

0.311 0.430 0.72 7

0.297 0.420 0.71 8

0.286 0.390 0.73 9

0.275 0.370 0.74 10 0.266 0.360 0.74 As shown above, the maximum GMRS/SSE ratio for Oyster Creek occurs at 1.0 Hz and equals 1.53, which is conservatively rounded up to 1.60.

The resulting 5% damped RLGM, based on scaling the horizontal SSE by the scale factor of 1.60, is shown below in Table 5-2 and Figure 5-1 below. Note that the RLGM peak ground acceleration (PGA) is 0.29g. Seismic capacities for equipment will be compared against the PGA of the RLGM.

Page 18 of 41

1.00 0.90 0.80 0.70

§ 0.60 r::::

0

~ 0.50 Q) a;

~ 0.40 0.30 0.20 0.10 0.00 Table 5-2: RLGM (5% Damping)

Frea. (Hz\\

RLGM fo\\

1 0.18 1.25 0.24 1.5 0.30 2

0.43 2.5 0.46 3

0.58 3.5 0.62 4

0.66 5

0.70 6

0.69 7

0.67 8

0.62 9

0.59 10 0.58 12.5 0.50 15 0.42 20 0.35 25 0.32 50 0.29 100 0.29 S&A Report 1404241-RPT-003 Rev. 5 Correspondence No. RS-14-299 Figure 5-1: Plot of RLGM (5% Damping) i i i i /i i

j l

,/

[

I

/

I i

I i i I!

i

..... \\.

I ro\\

\\.

l I :

10 Frequency (Hz) 1 I, I

i i i

l i

l

-RLGM 1'-.~

I i '

i I l 100 Page 19 of 41

5.2 Method to Estimate ISRS S&A Report 1404241-RPT-003 Rev. 5 Correspondence No. RS-14-299 The method used to derive the ESEP in-structure response spectra (ISRS) was to uniformly scale existing SSE-based ISRS from 50124-R-001 [16) by the maximum scale factor of 1.60 from Table 5-1. Scaled ISRS are calculated for all locations where ESEL items are located at Oyster Creek. These scaled ISRS are documented within calculation 1404241-CAL-001 (10].

Page 20 of 41

6.0 Seismic Margin Evaluation Approach S&A Report 1404241-RPT-003 Rev. 5 Correspondence No. RS-14-299 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 highest PGA for which there is a high confidence of a low probability of failure (HCLPF). The PGA is associated with a particular spectral shape, in this case the 5%-damped RLGM spectral shape. The calculated HCLPF capacity must be equal to or greater than the RLGM PGA (0.290g from Table 5-2). 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 [7].

2.

Probabilistic approach using the fragility analysis methodology of EPRI TR-103959 [8].

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

[7] was used to determine HCLPF capacities.

6.1 Summary of Methodologies Used Oyster Creek performed a probabilistic risk assessment (PRA) that was concluded in 2001. The PRA is documented in the Oyster Creek IPEEE report [9] and consisted of walkdowns and HCLPF calculations. The walkdowns were conducted by engineers trained in EPRI NP-6041 and PRA. Walkdown results were documented on Screening Evaluation Work Sheets (SEWS) from EPRI NP-6041 [7] in concert with the Unresolved Safety Issue (USI) A-46 evaluation of Oyster Creek.

The screening walkdowns used Table 2-4 of EPRI NP-6041 [7]. The walkdowns were conducted by engineers who as a minimum attended the Seismic Qualification Utility Group (SQUG) Walkdown Screening and Seismic Evaluation Training Course. The walkdowns were documented on Screening Evaluation Work Sheets (contained within report 1404241-RPT-005

[1 O]) from EPRI NP-6041 [7]. Anchorage capacity calculations used the CDFM criteria from EPRI NP-6041 [7]. The input seismic demand was RLGM shown in Table 5-2 and Figure 5-1.

6.2 HCLPF Screening Process The spectral peak RLGM for Oyster Creek reaches approximately 0.70g at 5 Hz (Table 5-2:).

The screening tables in EPRI NP-6041 [7] are based on ground peak spectral accelerations of O.Bg and 1.2g. These both exceed the RLGM peak spectral acceleration. The Oyster Creek reduced ESEL components were screened against the 0.8g column of Table 2-4 of NP-6041.

The Oyster Creek reduced ESEL (Attachment B) contains one item: the diesel generator fuel oil storage tank, tag number T-39-2. In accordance with Table 2-4 of EPRI NP-6041 [7], all atmospheric storage tanks require HLCPF evaluation. The HCLPF evaluation for tank T-39-2 is performed within calculation 1404241-CAL-002 [1 O], and results are summarized in Attachment C of this report. HCLPF capacities are compared against the RGLM peak ground acceleration identified in Section 5.1.

Page 21 of 41

6.3 Seismic Walkdown Approach 6.3.1 Walkdown Approach S&A Report 1404241-RPT-003 Rev. 5 Correspondence No. RS-14-299 Walkdowns for Oyster Creek were performed in accordance with the criteria provided in Section 5 of EPRI 3002000704 [2], which refers to EPRI NP-6041 [7] for the Seismic Margin Assessment process. Pages 2-26 through 2-30 of EPRI NP-6041 [7] describe the seismic walkdown criteria, including the following key criteria.

"The SRT {Seismic Review Team] should "walk by" 100% of all components which are reasonably accessible and in non-radioactive or low radioactive environments. Seismic capability assessment of components which are inaccessible, in high-radioactive environments, or possibly within contaminated containment, will have to rely more on alternate means such as photographic inspection, more reliance on seismic reanalysis, and possibly, smaller inspection teams and more hurried inspections. A 100% "walk by" does not mean complete inspection of each component, nor does it mean requiring an electrician or other technician to de-energize and open cabinets or panels for detailed inspection of all components. This walkdown is not intended to be a QA or QC review or a review of the adequacy of the component at the SSE level.

If the SRT has a reasonable basis for assuming that the group of components are similar and are similarly anchored, then it is only necessary to inspect one component out of this group. The "similarity-basis" should be developed before the walkdown during the seismic capability preparatory work (Step 3) by reference to drawings, calculations or specifications. The one component of each type which is selected should be thoroughly inspected which probably does mean de-energizing and opening cabinets or panels for this very limited sample. Generally, a spare representative component can be found so as to enable the inspection to be performed while the plant is in operation. At least for the one component of each type which is selected, anchorage should be thoroughly inspected.

The walkdown procedure should be performed in an ad hoc manner. For each class of components the SRT should look closely at the first items and compare the field configurations with the construction drawings and/or specifications. If a one-to-one correspondence is found, then subsequent items do not have to be inspected in as great a detail. Ultimately the walkdown becomes a "walk by" of the component class as the SRT becomes confident that the construction pattern is typical. This procedure for inspection should be repeated for each component class; although, during the actual walkdown the SRT may be inspecting several classes of components in parallel. Is serious exceptions to the drawings or questionable construction practices are found then the system or component class must be inspected in closer detail until the systematic deficiency is defined.

Page 22 of 41

S&A Report 1404241-RPT-003 Rev. 5 Correspondence No. RS-14-299 The 100% "walk by" is to look for outliers, lack of similarity, anchorage which is different from that shown on drawings or prescribed in criteria for that component, potential SI

[Seismic lnteraction;1 problems, situations that are at odds with the team members' past experience, and any other areas of serious seismic concern. If any such concerns surface, then the limited sample size of one component of each type for thorough inspection will have to be increased. The increase in sample size which should be inspected will depend upon the number of outliers and different anchorages, etc., which are observed. It is up to the SRT to ultimately select the sample size since they are the ones who are responsible for the seismic adequacy of all elements which they screen from the margin review. Appendix D gives guidance for sampling selection.

As shown in Attachment 8, the only item on the reduced ESEL is the diesel generator fuel oil storage tank, tag number T-39-2. However, the SRT deemed it prudent to walk down an expanded set of equipment that could potentially be susceptible to damage from low frequency spectral accelerations, namely the motor control centers, battery racks, and isolation condensers listed in Attachment A. Upon visual inspection, the SRT judged these items to have a natural frequency well above 1.9 Hz (the GMRS-SSE intersection point) and confirmed that they could be excluded from the reduced ESEL.

The diesel generator fuel oil storage tank is located within a confined space and the SRT was not permitted to access the area during the time of the walkdown. The tank was previously walked down during NTTF 2.3 [15] and USI A-46 [18] and it was determined that enough preexisting information was available to preclude the need to enter the confined space around the tank. Furthermore, EPRI 3002000704 [2] limits the ESEP seismic interaction reviews to nearby block walls and piping attached to tanks 1. Given that no block walls exist within the tank enclosure and previous walkdown information shows that piping exhibits adequate flexibility, a future walkdown to check tank seismic interactions is not necessary. Previous walkdown information that was relied upon is documented in Section 6.3.2.

6.3.2 Application of Previous Walkdown Information As discussed in Section 6.3.1, the confined space around the diesel generator fuel oil storage tank (T-39-2) prevented access during the time of the walkdowns. Previous walkdown information from NTTF 2.3 [15], along with existing calculations and SEWS from the USI A-46 evaluation [18], were determined to provide a sufficient amount of information for the purposes of ESEP.

6.3.3 Significant Walkdown Findings Consistent with the guidance from NP-6041 [7], no significant outliers or anchorage concerns were identified during the Oyster Creek ESEP walkdowns.

1 EPRI 3002000704 [2] page 5-4 limits the ESEP seismic interaction reviews to "nearby block walls" and "piping attached to tanks" which are reviewed "to address the possibility of failures due to differential displacements." Other potential seismic interaction evaluations are "deferred to the full seismic risk evaluations performed in accordance with EPRI 1025287 [14]."

Page 23 of 41

6.4 HCLPF Calculation Process S&A Report 1404241-RPT-003 Rev. 5 Correspondence No. RS-14-299 ESEL items were evaluated using the criteria in EPRI NP-6041 [7]. Those evaluations included the following steps:

Performing seismic capability walkdowns for equipment to evaluate the equipment installed plant conditions Performing screening evaluations using the screening tables in EPRI NP-6041 as described in Section 6.2 Performing HCLPF calculations considering various failure modes that include both structural (e.g. anchorage, load path etc.) and functional failure modes.

HCLPF calculations were performed using the CDFM methodology and are documented in calculation 1404241-CAL-002 [1 O], with results summarized in Attachment C of this report.

HCLPF capacities are compared against the RGLM peak ground acceleration identified in Section 5.1.

The CDFM analysis criteria established in Section 6 of EPRI NP-6041 [7] are used when detailed analysis is required. The relevant CDFM criteria from EPRI NP-6041 [7] are summarized in Table 6-1.

Page 24 of 41

S&A Report 1404241-RPT-003 Rev. 5 Correspondence No. RS-14-299 Table 6-1:

Calculation Summary Load combination:

Normal+ Ee

---~*.~~.. -*---*---~---------

Ground response spectrum:

Conservatively specified (84°/t) non-exceedance probability)

--*-~~*--~------------------ -~--*--~--~--*-*--~*-e*--**----------**--*----*-*-----**-*--*---------*

Damping:

Conservative estimate of median damping.

Structural model:

Best estimate (median) + uncertainty variation in frequency.

Soil-structure interaction Best estimate (median) + parameter variation Material strength:

Code specified minimum strength 01* 95%, exceedance of actual strength if test data is available.

Code ultimate strength (ACI), maximum strength (AISC), Service Static capacity equations:

Level D (ASME) or functional limits. If test data is available to demonstrate excessive conservatism of code equations then use 84% exceedance of test data for capacity equations.

For non-brittle failure modes and linear analysis, use 80% of Inelastic energy absorption:

computed seismic stress in capacity evaluation to account for ductility benefits or perform nonlinear analysis and use 95%

exceedance ductility levels.

In-structure (floor) spectra Use frequency shifting rather than peak broadening to account for generation:

unce1*tainty and use median damping.

The HCLPF capacity is equal to the PGA at which the strength limit is reached. The HCLPF earthquake load is calculated as follows:

U = Normal + Ee Where:

111 U = Ultimate strength per Section 6 of EPRI NP-6041 [7]

Ii Ee = HCLPF earthquake load

~

Normal= Norma! operating loads (dead live load expected to be present, etc.. )

For this calculation, the HCLPF earthquake !oad is related to a fixed reference earthquake:

Ee = SFcEref Where:

e Eref = reference earthquake from the relevant in-structure response spectrum (ISRS)

SFc = component-specific scale factor that satisfies U = Normal +Ee The HCLPF will be defined as the PGA produced by Ee. The Oyster Creek RLGM PGA is 0.290g, therefore:

HCLPF = 0.290g*SFc Page of 4*1

6.5 Functional evaluation of relays S&A Report 1404241-RPT-003 Rev. 5 Correspondence No. RS-14-299 Relays are not considered vulnerable to low frequency spectral accelerations and therefore do not need to be included in the reduced ESEL per section 2.2.1.1 of EPRI 3002000704 [2].

6.6 Tabulated ESEL HCLPF Values (Including Key Failure Modes)

Tabulated ESEL HCLPF values including key failure modes for low frequency ESEL items are included in Attachment C. Anchorage failure controls the diesel generator fuel oil storage tank HCLPF; therefore, the anchorage HCLPF value is listed in the table and the failure mode is set to "Anchorage".

7.0 Inaccessible Items 7.1 Identification of ESEL Items Inaccessible for Walkdowns As discussed in Section 6.3.2, the confined space around the diesel generator fuel oil storage tank (T-39-2) prevented access during the time of the walkdowns. Previous walkdown information from NTTF 2.3 [15] and USI A-46 [18] was determined by the Seismic Review Team (SRT) to provide sufficient information for the purposes of ESEP. Detailed analysis performed in 1404241-CAL-002 [1 O] found the tank (T-39-2) to be acceptable. A future walkdown of tank T-39-2 is not required.

7.2 Planned Walkdown I Evaluation Schedule I Close Out No additional walkdowns are required.

Page 26 of 41

8.0 ESEP Conclusions and Results 8.1 Supporting Information S&A Report 1404241-RPT-003 Rev. 5 Correspondence No. RS-14-299 Oyster Creek Generating Station 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 Oyster Creek response to the NRC's 50.54(f) letter [1 ]. On March 12, 2014, NEI submitted to the NRC results of a study [12) of seismic core damage risk estimates based on updated seismic hazard information as it applies to operating nuclear reactors in the Central and Eastern United States (CEUS). The study concluded that "site-specific seismic hazards show that there has not been an overall increase in seismic risk for the fleet of U.S. plants" based on the re-evaluated seismic hazards. As such, the "current seismic design of operating reactors continues to provide a safety margin to withstand potential earthquakes exceeding the seismic design basis."

The NRC's May 9, 2014 NTTF 2.1 Screening and Prioritization letter [13] concluded that the "fleetwide seismic risk estimates are consistent with the approach and results used in the Gl-199 safety/risk assessment." The letter also stated that "As a result, the staff has confirmed that the conclusions reached in Gl-199 safety/risk assessment remain valid and that the plants can continue to operate while additional evaluations are conducted."

An assessment of the change in seismic risk for Oyster Creek was included in the fleet risk evaluation submitted in the March 12, 2014 NEI letter [12] therefore, the conclusions in the NRC's May 9 letter [13] also apply to Oyster Creek.

In addition, the March 12, 2014 NEI letter [12] 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 Page 27 of 41

S&A Report 1404241-RPT-003 Rev. 5 Correspondence No. RS-14-299

Response spectra based frequency domain analysis rather than explicit time history based time domain analysis

Bounding requirements in codes and standards

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

Bounding testing requirements, and

Ductile behavior of the primary materials (that is, not crediting the additional capacity of materials such as steel and reinforced concrete beyond the essentially elastic range, etc.).

These design practices combine to result in margins such that the SSCs will continue to fulfill their functions at ground motions well above the SSE.

8.2 Summary of ESEP Identified and Planned Modifications The results of the Oyster Creek ESEP performed as ah interim action in response to the NRC's 50.54(f) letter [1] using the methodologies in the NRC endorsed guidance in EPRI 3002000704

[2] show that evaluated equipment are adequate in resisting the seismic loads expected to result from the site RLGM. Therefore, no plant modifications are required as a result of the Oyster Creek ESEP.

8.3 Modification Implementation Schedule No modification implementation schedule is required because no modifications are required.

8.4 Summary of Regulatory Commitments No regulatory commitments are required.

Page 28 of 41

9.0 References S&A Report 1404241-RPT-003 Rev. 5 Correspondence No. RS-14-299 1

NRC (E Leeds and M Johnson) Letter to All Power Reactor Licensees et al., "Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(f)

Regarding Recommendations 2.1, 2.3 and 9.3 of the Near-Term Task Force Review of Insights from the Fukushima Dai-lchi Accident," March 12, 2012.

2 Seismic Evaluation Guidance: Augmented Approach for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1 - Seismic. EPRI, Palo Alto, CA: May 2013. 3002000704.

3 Order Number EA-12-049 responses:

3.1 NRC Letter RS-13-023 from Oyster Creek (ML13060A126), "Overall Integrated Plan in Response to March 12, 2012 Commission Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events (Order Number EA-12-049)", February 28, 2013 3.2 NRC Letter RS-13-125 from Oyster Creek (ML13240A263), "First Six-Month Status Report in Response to March 12, 2012 Commission Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events (Order Number EA-12-049)", August 28, 2013 3.3 NRC Letter RS-14-013 from Oyster Creek (ML14059A220), "Second Six-Month Status Report in Response to March 12, 2012 Commission Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events (Order Number EA-12-049)", February 28, 2014 3.4 NRC Letter RS-14-211 from Oyster Creek (ML14241A253), "Third Six-Month Status Report in Response to March 12, 2012 Commission Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events (Order Number EA-12-049)", August 28, 2014 4

Oyster Creek Seismic Hazard and GMRS Submittal, Correspondence No. RS-14-070, dated March 31, 2014.

5 Nuclear Regulatory Commission, NUREG-1407, Procedural and Submittal Guidance for the Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities, June 1991 6

Nuclear Regulatory Commission, Generic Letter No. 88-20 Supplement 4, Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities -

1 OCFR 50.54(f), June 1991 7

A Methodology for Assessment of Nuclear Power Plant Seismic Margin, Rev. 1, August 1991, Electric Power Research Institute, Palo Alto, CA. EPRI NP 6041 8

Methodology for Developing Seismic Fragilities, August 1991, EPRI, Palo Alto, CA.

1994, TR-103959 9

Staff Evaluation Report of Individual Plant Examination of External Events (IPEEE) submittal for the Oyster Creek Nuclear Generating Station, dated February 8, 2001 1 O Oyster Creek ESEP Calculations:

10.1 S&A Calculation 1404241-CAL-001 Rev. 1, Generation of In-Structure Response Spectra for use in ESEP Evaluations Page 29 of 41

S&A Report 1404241-RPT-003 Rev. 5 Correspondence No. RS-14-299 10.2 S&A Calculation 1404241-CAL-002 Rev. 2, HCLPF Seismic Capacity of Diesel Oil Storage Tank 10.3 S&A Report 1404241-RPT-005, Rev. 5, Oyster Creek ESEP SEWS 11 Nuclear Regulatory Commission, NUREG/CR-0098, Development of Criteria for Seismic Review of Selected Nuclear Power Plants, published May 1978 12 Nuclear Energy Institute (NEI), A. Pietrangelo, Letter to D. Skeen of the USN RC, "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 13 NRC (E Leeds) Letter to All Power Reactor Licensees et al., "Screening and Prioritization Results Regarding Information Pursuant to Title 1 O of the Code of Federal Regulations 50.54(F) Regarding Seismic Hazard Re-Evaluations for Recommendation 2.1 of the Near-Term Task Force Review of Insights From the Fukushima Dai-lchi Accident," May 9, 2014.

14 Seismic Evaluation Guidance: Screening, Prioritization and Implementation Details (SPID) for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1:

Seismic. EPRI, Palo Alto, CA: February 2013. 1025287.

15 Oyster Creek NTTF 2.3 SeismicWalkdown Submittal, Correspondence No. RS-12-177, dated November 19, 2012 16 Report No. 50124-R-001, Rev. 0, In-Structure Response Spectra for the Oyster Creek Nuclear Generating Station Compilation of Response Spectra for Use in USI A-46 Program 17 S&A Report 1404241-RPT-004 Rev. 1, Validation of Expedited Seismic Equipment List 18 A-46 Seismic Qualification SQ-OC-T-39-002 Rev. 1, Diesel Oil Storage Tank T-39-002 19 Oyster Creek P&IDs:

19.1 GE 148F262 Sheet 1, Rev. 55, Emergency Condenser Flow Diagram 19.2 GE 237E798, Rev. 36, Recirculation System Flow Diagram 19.3 GE 885D781 Sheet 1, Rev. 73, Core Spray System Flow Diagram 19.4 GU 3E-243-21-1000 Sheet 1, Rev. 29, Drywell and Torus Vacuum Relief System Flow Diagram 19.5 BR 2011 Sheet 2, Rev. 62, Reactor Building VentHation Flow Diagram 19.6 SN 13432.19-1 Sheet 1, Rev. 33, Nitrogen Supply System Flow Diagram 19.7 GE 148F740 Sheet 1, Rev. 44, Containment Spray System Flow Diagram 19.8 GU 3E-862-21-1000 Sheet 1, Rev. 24, Emergency Diesel Generator Diesel Fuel Oil Storage & Transfer System Flow Diagram 20 NRC Order EA-12-051, "Order Modifying Licenses with Regard to Reliable Spent Fuel Pool Instrumentation" Page 30 of 41

S&A Report 1404241-RPT-003 Rev. 5 Correspondence No. RS-14-299 Attachment A: Oyster Creek ESEL Page 31of41

ESEL Item Number 2

3 4

5 6

7 8

9 10 11 12 USS 1.1\\;>

MCC 1A21 MCC 1A21A VMCC 1A2 BTCHC3 C1 Battery Bank C DC-C 125V DC-F DC-2 125VDC CD-14-18 LT-IG0006B Ll-211-'1215 V*14-35 USS '182 VMCC iB2 Table A-'! Oyster Creek.

Equipment Description 4HOVAC Vtta! f(1,-,actor Bldg Bus A Power to RE~c1rculation Loop Isolation Valves Power to F~ocirculation Loop Isolation Valves Vital Motor Control Center 1A2 C Station Battery Solid State Static Charger C'I Vital Bank C Station Batte1y

'125VDC Distribution Center C

'I 25VDC Power Panel DC-F 125VDC Motor Control CTR for Reactor Building B Isolation Condenser (NE01B)

B Isolation Condenser Shell Level XMITR B Isolation Condenser Local Shell level Indication B Isolation Condenser Condensate Return Valve 480Vfl.C Vital Reactor Bldg Bus B Vital Motor Control Center 1 B2 S&J\\ Hoporl i 4042.4 i-HPT*003 F~ev. 5 Conicc!.,;pondence No. f~S-14**299 Operating State Normal Desired State St:i;te In Service In Service In Service In Service In Service In Service In Service In Service Standby In Service In Service Closed In Service In Service In Service In Service In Servic<3 In Service In Service In Secv1ce In Service In Service In Service In Service as required In Service In Service Open/Closed Service Service Notes Isolation Condl'"nser, Core Spray and EMRV control/logic power Passive component The indicator for thi.s transmitter is localed in panel 1 F/2F Mechanical instrument f---------+------------+----------------*-----------**---------**------+---*---------e---*-------j---------;

VIV1CC 1/\\82 Vital Motor Control Cen!er 1AB2 (Recirculation Pump Isolation Valve Power)

In Service Service ATS 1AB2 is contained in VMCC 1AB2 f----------+-------------*--l--*--~------------**-~--*--+-----------1--------+--------i MCC 182'iA Power to i'<ecirculation Loop Isolation Valves In Service In Service

!--------!-*--*---------+--*---*---------*------------- --------f---------1 MCC 1821 Power to Static Charger and Recirculation Pump Isolation Valves In Service In Service

>--------+-------*--- ----------------f----------+---------+--

19 20 21 22 A/B Station Batteries Solid State Static Charger STATIC CHGR In Service In Service

--+-------+--------------------+--------- ---------+-----------<

Vital Bank B Station Battery (Lead Acid)

Battery Bank B In Service In Service DC-8 i25V 125VDC Distribution Panel B In Service In Service

1-------------------+-----~---------r---------

DC-0 125VDC Power Panel D In Service In Service Isolation Condenser, Core Spray and EMRV control/logic power Page 32of41

ESEL Item Number ID 23 DC-1 125VDC 24 CD-14-1A 25 LT-IG0006A 26 Ll-211-1214 27 V-14-34 28 V-20-15 29 RK-3 30 PT-IP0007 31 1F/2F 32 5F/6F 33 RSP 34 16R 35 18R 36 11F 37 V-23-13 S&A Report 1404241-RPT-003 Rev. 5 Correspondence No. RS-14-299 Table A-1 Oyster Creek ESEL Equipment Operating State Normal Desired Notes Description State State 125VDC Isolation Valves Motor ATS DC-1 is In Service In Service contained in Control Center MCC DC-1 A Isolation Condenser (NE01A)

Standby In Service as Passive required component The indicator A Isolation Condenser Shell Level for this In Service In Service transmitter is XMITR located in panel 1 F/2F A Isolation Condenser Local Shell In Service In Service Mechanical Level Indication Indicator A Isolation Condenser Condensate Closed Open/Closed Return Valve AC powered Core Spray to Reactor Parallel Valve valve which will Closed Open be manually System 1 operated during ELAP Contains separately Instrument Rack 03 In Service In Service powered PT-IP0007 instrument transmitter Containment Pressure Transmitter In Service In Service Phase 2 MCR Control Reactor & Drywell In Service In Service Cooling Panel Contains separately Main Control Room Panel 5F/6F In Service In Service powered instrument indicators Contains power supplies for, Remote Shutdown Panel In Service In Service and elements of, credited instruments Monitors Containment H2/02 Panel In Service In Service containment parameters Main Control Room Panel 18R Contains Reactor Protection In Service In Service instruments from the IOP Routes power MCR Panel 11 F In Service In Service to panel 12XR via internal fuse 6F7 Drywell N2 Purge Valve/Containment In Service In Service Isolation Valve for Hardened Vent Page 33 of 41

ESEL Item Number ID 38 V-23-14 39 V-23-15 40 V-23-16 41 DPT-622-1009 42 PT-622-1018 43 T-39-2 44 V-37-09 45 V-37-10 46 V-37-11 47 V-37-20 48 V-37-21 49 V-37-22 50 V-37-31 51 V-37-32 52 V-37-33 53 V-37-42 54 V-37-43 55 V-37-44 56 V-37-53 57 V-37-54 58 V-37-55 S&A Report 1404241-RPT-003 Rev. 5 Correspondence No. RS-14-299 Table A-1 Oyster Creek ESEL Equipment Operating State Description Normal Desired Notes State State Drywell N2 Purge Valve/Containment Isolation Valve for Hardened Vent In Service In Service Torus N2 Purge Valve/Containment In Service Isolation Valve for Hardened Vent In Service Torus N2 Purge Valve/Containment In Service Isolation Valve for Hardened Vent In Service The indicator Reactor Fuel Zone Level Wide Range for this I Transmitter (Channel C)

Standby In Service transmitter is located in panel 5F/6F The indicator Reactor Wide Range Pressure for this Transmitter (Channel C)

Standby In Service transmitter is located in panel 5F/6F Diesel Generator Fuel Oil Storage Passive Tank Standby Standby Component Reactor Recirculation Pump NG01-A-Suction Isolation Valve Open Closed Reactor Recirculation Pump NG01-A Discharge Isolation Valve Open Closed Reactor Recirculation Loop 'A' Bypass Valve NG08-A Open Closed Reactor Recirculation Pump NG01-B Suction Isolation Valve Open Closed Reactor Recirculation Pump NG01-B Discharge Isolation Valve Open Closed Reactor Recirculation Loop 'B' Bypass Valve NG08-B Open Closed Reactor Recirculation Pump NG01-C Suction Isolation Valve Open Closed Reactor Recirculation Pump NG01-C Discharge Isolation Valve Open Closed Reactor Recirculation Loop 'C' Bypass Valve NG08-C Open Closed Reactor Recirculation Pump NG01-D Suction Isolation Valve Open Closed Reactor Recirculation Pump NG01-D Discharge Isolation Valve Open Closed Reactor Recirculation Loop 'D' Bypass Valve NG08-D Open Closed Reactor Recirculation Pump NG01-E Suction Isolation Valve Open Closed Reactor Recirculation Pump NG01-E Discharge Isolation Valve Open Closed Reactor Recirculation Loop 'E' Bypass Valve NG08-E Open Closed Page 34 of 41

ESEL Item Number ID 59 LSP-1AB2 60 3F 61 IP-4 62 IT-4 63 IT-48 64 10R 65 ER-622-080 66 ATS DC-D 67 6K3A 68 6K3B 69 6K5A 70 6K5B 71 6K4A 72 6K4B S&A Report 1404241-RPT-003 Rev. 5 Correspondence No. RS-14-299 Table A-1 Oyster Creek ESEL Equipment Operating State Normal Desired Notes Description State State Contains Local Shutdown Panel Standby Standby elements of control for valve V-37-54 Contains control switches for recirculation pump valves.

Panel In Service In Service Valve control power provided from MCC via internal control transformer 120V AC Vital Power Distribution Provides power Panel In service In service for credited instruments Provides power for 120V AC Automatic Transfer Switch In Service In Service vital power distribution panel IP-4 Provides power Transformer In Service In Service for automatic transfer switch IT-4 Contains power supplies for, Panel In Service In Service and elements of, credited instrumentation Contains power supplies for, Panel In Service In Service and elements of, credited instrumentation Provides power Automatic Transfer Switch In Service In Service for 125V DC distribution panel DC-D Isolation Condenser Valve Hi Flow Energized Energized CR120AD0424 Isolation Logic 1AA relay Isolation Condenser Valve Hi Flow Energized Energized CR120AD0424 Isolation Logic 1AA relay Isolation Condenser Valve Hi Flow Energized Energized CR120AD0424 Isolation Logic 1AA relay Isolation Condenser Valve Hi Flow Energized Energized CR120AD0424 Isolation Logic 1AA relay Isolation Condenser Valve Hi Flow Energized Energized CR120AD0424 Isolation Logic 1AA relay Isolation Condenser Valve Hi Flow Energized Energized CR120AD0424 Isolation Logic 1AA relay Page 35 of 41

ESEL Item Number ID 73 6K6A 74 6K6B 75 6K7A 76 6K7B 77 6K8A 78 6K8B 79 Y-6-42 80 Y-6-43 81 Y-6-44 82 V-6-953 83 V-6-954 84 V-6-902 85 V-6-903 86 V-6-950 87 V-6-899 88 V-6-898 89 CIP-3 S&A Report 1404241-RPT-003 Rev. 5 Correspondence No. RS-14-299 Table A-1 Oyster Creek ESEL Equipment Operating State Normal Desired Notes Description State State Isolation Condenser valve Hi Flow Energized Energized CR 120AD0424 Isolation Logic 1AA relay Isolation Condenser Valve Hi Flow Energized Energized CR120AD0424 Isolation Logic 1AA relay 27s Time Delay Isolation Condenser Valve Hi Flow Drop-Out Relay Isolation Logic Energized Energized Model 700RTC 11200 U1 27s Time Delay Isolation Condenser Valve Hi Flow Drop-Out Relay Isolation Logic Energized Energized Model 700RTC 11200 U1 27s Time Delay Isolation Condenser Valve Hi Flow Drop-Out Relay Isolation Logic Energized Energized Model 700RTC 11200 U1 27s Time Delay Isolation Condenser Valve Hi Flow Drop-Out Relay isolation logic Energized Energized Model 700RTC11200 U1 Back-Up Air Supply Accumulator for Functional Functional Passive Valve V-23-0013 Component Back-Up Air Supply Accumulator for Functional Functional Passive Valve V-23-0014 Component Back-Up Air Supply Accumulator for Functional Functional Passive Valve V-23-0015&16 Component Pilot Solenoid Air Supply Valve for V-De-Energized Energized 23-0015 Pilot Solenoid Air Supply Valve for V-De-Energized Energized 23-0016 Pilot Solenoid Air Supply Valve for V-De-Energized Energized 23-0013 Pilot Solenoid Air Supply Valve for V-De-Energized Energized 23-0014 McMaster-Carr Instrument Air Regulating Valve Functional Functional Supply Co, 382M, Model:

4959K1 McMaster-Carr Instrument Air Regulating Valve Functional Functional Supply Co, 382M, Model:

4959K1 Fisher Controls Instrument Air Regulating Valve Functional Functional International LLC Model 67CFR-239 Continuous Instrument Panel No. 3 Energized Energized Page 36 of 41

ESEL Item Number ID 90 ROTARY INVERTER 91 12XR 92 IT-3 S&A Report 1404241-RPT-003 Rev. 5 Correspondence No. RS-14-299 Table A-1 Oyster Creek ESEL Equipment Operating State Normal Desired Notes Description State State 120V AC Supply for CIP-3 208/120V, Energized Energized 3PH, 4W Contains PNL-822-12XRCS1 Key lock bypass switch Panel In Service In Service for purge

valves, bypasses Isolation relays for hardened vent valves Automatic Transfer Switch In Service In Service Page 37of41

S&A Heport l4CJ4241*-HPf"-003 R<w. '.i Correspondence No. f~S *14-2D\\l Attachment B: Oyster Creek Reduced ESEL (low frequency items)

Tl1e reduced ESEL listed on the following table contains those items from Attachment A which are susceptible to damage from low frequency spectral accelerations, as defined in Section 2.2:1.1 of EPRI 3002000104 [2].

Page 38 of 41

ESEL Item Number ID 43 T-39-2 S&A Report 1404241-RPT-003 Rev. 5 Correspondence No. RS-14-299 Table B-1 Oyster Creek Reduced ESEL Equipment Operating State Notes Description Normal State Desired State Diesel Generator Fuel Oil Storage Tank Standby Standby Passive Component Page 39 of 41

S&A Report 1404241-RPT-003 Rev. 5 Correspondence No. RS-14-299 Attachment C: ESEL HCLPF Values and Failure Modes Tabulation Page 40 of 41

ESEL Item Number 43 S&A Report 14Q4241-RPT-003 Rev. 5 Correspondence No. RS-14-299 Table C-1 Oyster Creek ESEP HCLPF Values and Failure Mode Tabulation Equipment ID Failure Mode HCLPF (g)

Additional Discussion T-39-2 Anchorage 0.53 HCLPF calculated in 1404241-CAL-002 [10)

Page 41 of 41