CNL-14-211, 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.
| ML14365A055 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 12/22/2014 |
| From: | James Shea Tennessee Valley Authority |
| To: | Document Control Desk, Division of Operating Reactor Licensing |
| References | |
| CNL-14-211 | |
| Download: ML14365A055 (64) | |
Text
Tennessee Valley Authority, 1101 Market Street, Chattanooga, Tennessee 37402 CNL-14-211 December 22, 2014 10 CFR 50.4 10 CFR 50.54(f)
Attn: Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 Sequoyah Nuclear Plant, Units 1 and 2 Facility Operating License Nos. DPR-77 and DPR-79 NRC Docket Nos. 50-327 and 50-328
Subject:
Tennessee Valley Authoritys Sequoyah Nuclear Plant 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
References:
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 (ML12056A046)
On March 12, 2012, the U.S. Nuclear Regulatory Commission (NRC) issued the referenced letter to all power reactor licensees and holders of construction permits in active or deferred status. Enclosure 1 of the referenced letter requested each addressee located in the Central and Eastern United States (CEUS) to submit a Seismic Hazard Evaluation that includes an interim evaluation 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 referenced letter above, TVA is enclosing the Expedited Seismic Evaluation Process (ESEP) Report for Sequoyah Nuclear Plant. provides a list of new regulatory commitments as described in Section 8.0 of the enclosed ESEP Report.
U.S. Nuclear Regulatory Commission CNL-14-211 Page 2 December 22, 2014 Should you have any questions concerning the content of this letter, please contact Kevin Casey at (423) 751-8523.
I declare under penalty of perjury that the foregoing is true and correct. Executed on the 22nd day of December 2014 .
. Shea President, Nuclear Licensing
Enclosures:
- 1. Expedited Seismic Evaluation Process (ESEP) Report for Sequoyah Nuclear Plant
- 2. List of Commitments cc (Enclosures):
NRR Director- NRC Headquarters NRO Director - NRC Headquarters NRR JLD Director- NRC Headquarters NRC Regional Administrator - Region II NRR Project Manager - Sequoyah Nuclear Plant NRR JLD Project Manager- Sequoyah Nuclear Plant NRC Senior Resident Inspector- Sequoyah Nuclear Plant
ENCLOSURE 1 EXPEDITED SEISMIC EVALUATION PROCESS (ESEP) REPORT FOR SEQUOYAH NUCLEAR PLANT
EXPEDITED SEISMIC EVALUATION PROCESS (ESEP) REPORT FOR SEQUOYAH NUCLEAR PLANT Page 1
Sequoyah Nuclear Plant ESEP Report Table of Contents Page LIST OF TABLES ............................................................................................................................................ 4 LIST OF FIGURES .......................................................................................................................................... 5 1.0 PURPOSE AND OBJECTIVE ............................................................................................................... 6 2.0 BRIEF
SUMMARY
OF THE FLEX SEISMIC IMPLEMENTATION STRATEGIES ...................................... 6 3.0 EQUIPMENT SELECTION PROCESS AND ESEL.................................................................................. 8 3.1 Equipment Selection Process and ESEL .............................................................................. 8 3.1.1 ESEL Development ............................................................................................... 9 3.1.2 Power Operated Valves ..................................................................................... 10 3.1.3 Pull Boxes ........................................................................................................... 10 3.1.4 Termination Cabinets......................................................................................... 10 3.1.5 Critical Instrumentation Indicators .................................................................... 10 3.1.6 Phase 2 and 3 Piping Connections ..................................................................... 10 3.2 Justification for Use of Equipment That is Not the Primary Means for FLEX Implementation ................................................................................................................ 11 4.0 GROUND MOTION RESPONSE SPECTRUM (GMRS) ...................................................................... 11 4.1 Plot of GMRS Submitted by the Licensee ......................................................................... 11 4.2 Comparison to SSE ............................................................................................................ 13 5.0 REVIEW LEVEL GROUND MOTION (RLGM) ................................................................................... 14 5.1 Description of RLGM Selected .......................................................................................... 14 5.2 Method to Estimate InStructure Response Spectra (ISRS) .............................................. 15 6.0 SEISMIC MARGIN EVALUATION APPROACH ................................................................................. 17 6.1 Summary of Methodologies Used .................................................................................... 17 6.2 HCLPF Screening Process .................................................................................................. 18 6.3 Seismic Walkdown Approach ........................................................................................... 19 6.3.1 Walkdown Approach.......................................................................................... 19 6.3.2 Application of Previous Walkdown Information ............................................... 20 6.3.3 Significant Walkdown Findings .......................................................................... 20 6.4 HCLPF Calculation Process ................................................................................................ 21 6.5 Functional Evaluations of Relays ...................................................................................... 21 6.6 Tabulated ESEL HCLPF Values (Including Key Failure Modes) .......................................... 22 7.0 INACCESSIBLE ITEMS ..................................................................................................................... 22 7.1 Identification of ESEL Item Inaccessible for Walkdowns .................................................. 22 7.2 Planned Walkdown / Evaluation Schedule / Close Out .................................................... 23 8.0 ESEP CONCLUSIONS AND RESULTS ............................................................................................... 24 8.1 Supporting Information .................................................................................................... 24 Page 2
Sequoyah Nuclear Plant ESEP Report Table of Contents (continued)
Page 8.2 Identification of Planned Modifications ........................................................................... 25 8.3 Modification Implementation Schedule ........................................................................... 25 8.4 Summary of Regulatory Commitments ............................................................................ 26
9.0 REFERENCES
.................................................................................................................................. 27 ATTACHMENT A - SEQUOYAH NUCLEAR PLANT ESEL ............................................................................. A1 ATTACHMENT B - ESEP HCLPF VALUES AND FAILURE MODES TABULATION FOR SEQUOYAH NUCLEAR PLANT......................................................................................................... B1 Page 3
Sequoyah Nuclear Plant ESEP Report List of Tables Page TABLE 41: GMRS FOR SEQUOYAH NUCLEAR PLANT ............................................................................... 11 TABLE 42: SSE FOR SEQUOYAH NUCLEAR PLANT ................................................................................... 13 TABLE 51: 2X SSE FOR SEQUOYAH NUCLEAR PLANT .............................................................................. 15 TABLE 81:
SUMMARY
OF REGULATORY COMMITMENTS ...................................................................... 26 TABLE A1: EXPEDITED SEISMIC EQUIPMENT LIST (ESEL) FOR SEQUOYAH NUCLEAR PLANT ................ A2 TABLE B1: ESEP HCLPF VALUES AND FAILURE MODES FOR SEQUOYAH NUCLEAR PLANT ................... B2 Page 4
Sequoyah Nuclear Plant ESEP Report List of Figures Page FIGURE 41: GMRS FOR SEQUOYAH NUCLEAR PLANT ............................................................................. 12 FIGURE 42: GMRS TO SSE COMPARISON FOR SEQUOYAH NUCLEAR PLANT ......................................... 14 FIGURE 51: 2X SSE FOR SEQUOYAH NUCLEAR PLANT ............................................................................ 15 FIGURE 52: NUREG/CR0098 (0.3G) VERSUS SEQUOYAH NUCLEAR PLANT SSE .................................... 16 FIGURE 61: 84TH PERCENTILE OF THE ENSEMBLE OF THE 30 RESPONSE SPECTRA .............................. 18 FIGURE 62: SEQUOYAH NUCLEAR PLANT IPEEE ADJUSTED HCLPF VS ESEP TARGET HCLPF .................. 18 Page 5
Sequoyah Nuclear Plant ESEP Report 1.0 PURPOSE AND OBJECTIVE Following the accident at the Fukushima Daiichi nuclear power plant resulting from the March 11, 2011, Great Tohoku Earthquake and subsequent tsunami, the Nuclear Regulatory Commission (NRC) established a NearTerm 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 presentday 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 Sequoyah Nuclear Plant Units 1 and 2. The intent of the ESEP is to perform an interim action in response to the NRCs 50.54(f) letter to demonstrate seismic margin through a review of a subset of the plant equipment that can be relied upon to protect the reactor core following beyond design basis seismic events.
The ESEP is implemented using the methodologies in the NRC endorsed guidance in Electric Power Research Institute (EPRI) 3002000704, Seismic Evaluation Guidance: Augmented Approach for the Resolution of Fukushima NearTerm 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.
This ESEP report is for both Sequoyah Unit 1 and Unit 2 which are identical. Unless noted otherwise, all descriptions in this report apply to both Unit 1 and Unit 2 structures, systems, and components. For this reason, unit designations are not included on equipment unit identifications in the FLEX strategy or Expedited Seismic Equipment List (ESEL) descriptions.
2.0 BRIEF
SUMMARY
OF THE FLEX SEISMIC IMPLEMENTATION STRATEGIES The Sequoyah FLEX strategies for Reactor Core Cooling and Heat Removal, Reactor Inventory Control, and Containment Function are summarized below. This summary is derived from the Sequoyah Overall Integrated Plan (OIP) in Response to the March 12, 2012, Commission Order EA12049 submitted in February 2013 [3] [4] and is consistent with the third six month status report issued to the NRC in August 2014 [5].
For At Power Conditions Core Cooling and Heat Removal Reactor core cooling and heat removal is achieved via steam release from the Steam Generators (SGs) with SG makeup from the Turbine Driven Auxiliary Feedwater Pump (TDAFWP) during FLEX Phase 1 with suction from the Condensate Storage Tank (CST) [5]. Local control and operation of the SG Page 6
Sequoyah Nuclear Plant ESEP Report Atmospheric Relief Valves (ARVs) and the TDAFWP system is available and proceduralized so that operation from the main control room is not required.
To provide an unlimited supply of water for core cooling during Phase 2, low pressure FLEX Pumps will be staged at the Intake Pump Station (IPS) and take suction from the intake channel and discharge to four, Emergency Raw Cooling Water (ERCW) FLEX connections inside the IPS. They will be used to pressurize the ERCW headers, which can then be used for direct supply to the TDAFWP suction.
When the TDAFWP becomes unavailable due to reduction in available steam pressure, a portable intermediate pressure FLEX pump will be used to continue to supply feedwater to the SGs. Suction would be from an ERCW FLEX connection. The discharge is routed by hose to the TDAFWP discharge FLEX connections downstream of Flow Element 3142.
Reactor Inventory Control For Phase 1, Reactor Coolant System (RCS) makeup will be provided by the cold leg accumulators. RCS depressurization and cool down will be initiated as soon as possible to reduce the Reactor Coolant Pump (RCP) seal leakage rate.
In Phase 2, RCS makeup will be provided by repowering existing Safety Injection (SI) pumps and using the pumps to inject borated water as needed into the RCS. The SI pumps will be repowered with a 6.9 kVA FLEX Diesel Generator. The SI pumps can be manually controlled with hand switches on Panel M
Later in Phase 2, when the RCS is depressurized sufficiently, a highpressure FLEX pump will be used to inject borated water into the RCS through SI piping. These pumps would be aligned with a suction hose from RWST FLEX connections and a discharge hose routed to a SI pump discharge FLEX header connection. The high pressure FLEX pumps are fed from and operated from the 480v Control and Auxiliary Building (C&A) Vent Boards 1A1A and 2A1A.
Containment Function There are no Phase 1 FLEX actions to maintain containment integrity. The primary Phase 2 FLEX strategy for containment integrity entails repowering one train of hydrogen igniters. Phase 2 may entail repowering the Containment Air Return Fans inside of containment.
Support Systems Key reactor parameters to be monitored during FLEX implementation are measured and indicated by instrumentation that is powered by the 125V DC vital battery. During Phase 1, the vital batteries provide power to needed instrumentation through the vital battery boards, vital inverters and vital instrument power boards.
During Phase 2, power to vital instrumentation will be maintained by supplying 480V AC power to the vital battery chargers through new, fused, FLEX distribution panels, which will be connected directly to the battery chargers. 480V AC power will be supplied to the distribution panels by prestaged, 480V AC FLEX diesel generators that will be located on the roof of the auxiliary building.
During the early portion of Phase 2, the 6.9kV switchgear and 6.9kV Shutdown Boards will be energized with a prestaged 6.9kV FLEX diesel generator that will be located in the additional diesel generator building. This will allow reenergizing the SI pumps for inventory control.
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Sequoyah Nuclear Plant ESEP Report For Shutdown Conditions During shutdown, both safety functions (maintaining core cooling and heat removal and maintaining RCS inventory control) are accomplished by the same FLEX strategy and rely on the same FLEX equipment needed for the at power condition. Core cooling and heat removal is achieved by coolant boil off. Injection of borated water to the RCS is needed to replenish the coolant lost to boiling. For shutdown configurations where the RCS is depressurized and open but the cavity is not flooded, gravity feed from the RWST may be used to maintain RCS inventory in Phase 1. A flow path from the RWST to the RCS would be established. If gravity feed is not sufficient to makeup coolant to the RCS, a pre staged, intermediate pressure FLEX pump will be used to maintain RCS inventory (in Phase 2).
Sufficient flushing flow will be needed to prevent boron precipitation. Connection of the FLEX pump discharge hoses will be to the safety injection piping using the same FLEX connections planned for RCS inventory control under at power conditions. The FLEX connections are shown in [6]
For shutdown configurations where the RCS head is off and the cavity is filled, there will be sufficient time to mobilize portable FLEX pumps to provide RCS makeup from the BATs or an alternate borated water source. The same FLEX connections to the safety injection system piping will be used in this mode.
In accordance with [7] (footnote to Table D1), some shutdown configurations where the RCS is closed or pressurized so that injection of borated water cannot be accomplished are considered outside of ESEP because these configurations have short durations.
3.0 EQUIPMENT SELECTION PROCESS AND ESEL The selection of equipment for the ESEL followed the guidelines of EPRI 3002000704 [2]. The ESEL for Sequoyah Units 1 and 2 is presented in Attachment A. Information presented in Attachment A is drawn from [8].
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 Sequoyah OIP in Response to the March 12, 2012, Commission Order EA12049 [3] and is consistent with the second and third six month status reports issued to the NRC
[4] [5]. The OIP provides the Sequoyah 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 Sequoyah 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 prestaged FLEX equipment (not permanently installed) are excluded from the ESEL per EPRI 3002000704.
The ESEL component selection followed the EPRI guidance outlined in Section 3.2 of EPRI 3002000704.
- 1. The scope of components is limited to that required to accomplish the core cooling and containment safety functions identified in Table 32 of EPRI 3002000704. The instrumentation monitoring requirements for core cooling/containment safety functions Page 8
Sequoyah Nuclear Plant ESEP Report are limited to those outlined in the EPRI 3002000704 guidance, and are a subset of those outlined in the Sequoyah OIP.
- 2. The scope of components is limited to installed plant equipment and FLEX connections necessary to implement the Sequoyah 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 Backup/Alternate).
- 4. The Primary FLEX success path is to be specified. Selection of the Backup/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 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.
Poweroperated 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 backup 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 Sequoyah Nuclear Plant OIP [3] [4] [5] 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 Single Line Diagrams) were performed to identify the boundaries of the flow paths to be used in the FLEX strategies and to identify specific components in the flow paths needed to support implementation of the FLEX strategies. Boundaries were established at an electrical or mechanical isolation device (e.g., isolation amplifier, valve, etc.) in branch circuits /
branch lines off the defined strategy electrical or fluid flow path. P&IDs were the primary reference documents used to identify mechanical components and instrumentation. The flow paths used for FLEX strategies were selected and specific components were identified using detailed equipment and instrument drawings, piping isometrics, electrical schematics and oneline drawings, system descriptions, design basis documents, etc., as necessary. Host components were identified for sub assemblies.
Cabinets and equipment controls containing relays, contactors, switches, potentiometers, circuit breakers and other electrical and instrumentation that could be affected by highfrequency 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 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 Page 9
Sequoyah Nuclear Plant ESEP Report 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 33 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 32 also notes that functional failure modes of electrical and mechanical portions of the installed Phase 1 equipment should be considered (e.g. Auxiliary Feedwater (AFW) trips). To address this concern, the following guidance is applied in the Sequoyah ESEL for functional failure modes associated with power operated valves:
Power operated valves that remain energized during the 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 deenergized.
Power operated valves not required to change state as part of the FLEX mitigation strategies during Phase 1, and are reenergized 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 repowered.
3.1.3 Pull Boxes Pull boxes were deemed unnecessary to be added to the ESELs 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.
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 (ruleofthebox).
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 Sequoyah OIP [3] [4] [5] 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 Backup/Alternate).
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Sequoyah Nuclear Plant ESEP Report 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 The Sequoyah Nuclear Plant ESEL is based on the primary means of implementing the FLEX strategy.
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 base of the Containment Structures, which corresponds to a depth of 64 ft. (Elevation 641 ft.) and is the deepest structure foundation elevation control point. Table 41 shows the GMRS accelerations for a range of frequencies. The GMRS at the control point elevation is shown in Figure 41 [9].
Table 41: GMRS for Sequoyah Nuclear Plant Frequency (Hz) GMRS (g) 100 3.79E01 90 3.83E01 80 3.89E01 70 3.98E01 60 4.18E01 50 4.65E01 40 5.54E01 35 6.14E01 30 6.72E01 25 7.41E01 20 7.59E01 15 7.57E01 12.5 7.49E01 10 7.06E01 9 6.82E01 8 6.53E01 7 6.11E01 6 5.58E01 5 5.00E01 Page 11
Sequoyah Nuclear Plant ESEP Report Table 41: GMRS for Sequoyah Nuclear Plant (Continued)
Frequency (Hz) GMRS (g) 4 4.05E01 3.5 3.78E01 3 3.13E01 2.5 2.50E01 2 2.30E01 1.5 1.92E01 1.25 1.68E01 1 1.42E01 0.9 1.36E01 0.8 1.25E01 0.7 1.14E01 0.6 9.98E02 0.5 8.34E02 0.4 6.67E02 0.35 5.84E02 0.3 5.00E02 0.25 4.17E02 0.2 3.34E02 0.15 2.50E02 0.125 2.08E02 0.1 1.67E02 Figure 41: GMRS for Sequoyah Nuclear Plant Page 12
Sequoyah Nuclear Plant ESEP Report 4.2 Comparison to SSE The SSE was developed in accordance with 10 CFR Part 100 Appendix A through an evaluation of the maximum earthquake potential for the region surrounding the site. Considering the historic seismicity of the site region, the maximum potential earthquake was determined to be an intensity VIII on the Modified Mercalli Intensity Scale of 1931. The SSE is defined in terms of a Peak Ground Acceleration (PGA) and a design response spectrum. Considering a site intensity of VIII, a PGA of 0.18g was estimated. To be consistent with EPRI Report Nos. EPRI 3002000704 [2] and EPRI 1025287 [22] the site licensing basis earthquake is used for the SSE to GMRS comparison in this report. The design basis earthquake was used in the prior TVA submittal [9] which concluded that a risk analysis would be performed. However, the application of either the design basis or the licensing basis SSE curve to the prior evaluation will not alter the conclusion. The Sequoyah licensing basis SSE is based on a peak ground acceleration of 0.18g with a Housner spectral shape. Table 42 shows the spectral acceleration values as a function of frequency for the 5% damped horizontal Sequoyah licensing basis SSE.
Table 42: SSE for Sequoyah Nuclear Plant Frequency Spectral Acceleration (Hz) (g) 100 0.18 25 0.18 10 0.19 5 0.27 2.5 0.26 1.0 0.14 0.5 0.08 Page 13
Sequoyah Nuclear Plant ESEP Report Figure 42: GMRS to SSE Comparison for Sequoyah Nuclear Plant The SSE and the GMRS in the low frequency range up to about 2.5Hz are essentially the same amplitude. The GMRS exceeds the Sequoyah Nuclear Plant SSE beyond about 2.5Hz. As the GMRS exceeds the SSE in the 1 to 10Hz 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 Seismic Hazard Site and b) Narrow Band Exceedances in the 1 to 10Hz range do not apply for Sequoyah Nuclear Plant and hence High Confidence of a Low Probability of Failure (HCLPF) evaluations are required.
5.0 REVIEW LEVEL GROUND MOTION (RLGM) 5.1 Description of RLGM Selected Section 4 of EPRI 3002000704 [2] presents two approaches for developing the RLGM to be used in the ESEP:
- 1. The RLGM may be derived by linearly scaling the SSE by the maximum ratio of the GMRS/SSE between the 1 and 10 Hz range (not to exceed 2x SSE). Instructure RLGM seismic motions would be derived using existing SSE based instructure response spectra (ISRS) with the same scale factor.
- 2. Alternately, licensees who have developed appropriate structural/soilstructure interaction (SSI) models capable of calculating ISRS based on site GMRS/uniform hazard response spectrum (UHRS) input may opt to use these ISRS in lieu of scaled SSE ISRS.
Based on a review of tabulated data in Table 41 and the SSE values in Table 42, in the range between 1 and 10 Hz the maximum ratio of GMRS to the SSE is calculated to be:
SFmax = SAGMRS(10 HZ)/SA SSE 10Hz) = 0.71g/0.19g = 3.7 Page 14
Sequoyah Nuclear Plant ESEP Report Since the computed scale factor is greater than 2.0, the RLGM would be set a level of 2x SSE. This is shown in Table 51 and Figure 51.
Table 51: 2x SSE for Sequoyah Nuclear Plant Frequency Spectral Acceleration (Hz) (g) 100 0.36 25 0.36 10 0.38 5 0.54 2.5 0.52 1.0 0.28 0.5 0.16 SQN 2x SSE Design Spectra , 5% Damping 0.6 0.5 Acceleration (g) 0.4 0.3 0.2 2x SSE 0.1 0
0.1 1 10 100 Frequency (Hz)
Figure 51: 2x SSE for Sequoyah Nuclear Plant 5.2 Method to Estimate InStructure Response Spectra (ISRS)
A full scope SMA was performed to support the IPEEE for Sequoyah Nuclear Plant Units 1 and 2 [11].
The Review Level Earthquake (RLE) is defined as the NUREG/CR0098 [10] median spectral shape for rock, anchored to 0.3g PGA. The RLE ISRS were defined at the 84% NonExceedance Probability (NEP).
To determine the 84% NEP response, a probabilistic method of generating ISRS was used which accounts for the uncertainty in both the ground motion description and in the structural and soil parameters.
Uncertainties in the structural properties are accounted for by representing structural natural frequencies and damping ratios as a lognormally distributed random variable with specified median and Coefficient of Variation (COV) values. A total of thirty (30) earthquake time histories (each with Page 15
Sequoyah Nuclear Plant ESEP Report three components) were generated such that the spectral ordinates were lognormally distributed with a COV equal to 0.25, and the 84% NEP value matches the NUREG/CR0098 median rock shape.
The results of the IPEEE for Sequoyah Nuclear Plant Units 1 and 2 were submitted to the NRC [11]. It should be noted that the NRC [12] took exception to the approach used for Sequoyah Nuclear Plant in that Tennessee Valley Authority (TVA) defined the RLE as being in the freefield at the top of the soil surface, whereas the NRC concluded the RLE should have been defined on a rock. TVA reviewed the NRC Request for Additional Information (RAI) and made adjustments to the originally defied HCLPF capacity of 0.3g. The results of the adjustments of the full scope seismic margin assessment were submitted to the NRC [13], concluding that Sequoyah Nuclear Plant Units 1 and 2 had a plant level HCLPF capacity of 0.23g. Subsequent to TVA's docketed response [13] the NRC issued their Staff Evaluation Report (SER) [14]. In the SER, the NRC recognized the TVA HCLPF capacity value of 0.23g for Sequoyah Nuclear Plant, but also acknowledged a lower HCLPF value of 0.2g that was developed by the staff consultant. During the IPEEE adequacy review, TVA reviewed the NRC staff consultant's opinion regarding a lower HCLPF capacity of 0.2g for Sequoyah Nuclear Plant and concluded that the technical basis described by the NRC staff consultant in the SER is technically correct. Consequently, TVA decided that the assignment of 0.2g HCLPF capacity was appropriate.
Because of the significant effort expended by TVA to develop an updated dynamic analysis of the safety related structures for Sequoyah Nuclear Plant described above and in the IPEEE submittal [11],
TVA felt this model provided improved dynamic behavior of Sequoyah Nuclear Plant structures.
Consequently, for the purpose of evaluating seismic capacity of ESEP components, TVA chose to scale the 84th percentile values by an increase scale factor of 1.5 (0.3g/0.2g) to achieve a response equivalent to a 0.3g NUREG/CR0098 shaped response. Figure 52 demonstrates that the use of a 0.3g NUREG/CR0098 shape response bounds 2x SSE for Sequoyah Nuclear Plant from 1 to 10 Hz.
0.7 0.6 0.5 Acceleration (g) 0.4 2x SSE 0.3 CR0098 0.3g SSE (.18g) 0.2 0.1 0
0.01 0.1 1 10 100 Frequency (Hz)
Figure 52: NUREG/CR0098 (0.3g) versus Sequoyah Nuclear Plant SSE Page 16
Sequoyah Nuclear Plant ESEP Report 6.0 SEISMIC MARGIN EVALUATION APPROACH It is necessary to demonstrate that ESEL items have sufficient seismic capacity to meet or exceed the demand characterized by the RLGM. The seismic capacity is characterized as the PGA for which there is a HCLPF. The PGA is associated with a specific spectral shape, in this case the 5%damped RLGM spectral shape. The HCLPF capacity must be equal to or greater than the RLGM PGA. The criteria for seismic capacity determination are given in Section 5 of EPRI 3002000704 [2].
There are two basic approaches for developing HCLPF capacities:
- 1. Deterministic approach using the conservative deterministic failure margin (CDFM) methodology of EPRI NP6041SL, A Methodology for Assessment of Nuclear Power Plant Seismic Margin (Revision 1) [15].
- 2. Probabilistic approach using the fragility analysis methodology of EPRI TR103959, Methodology for Developing Seismic Fragilities [16].
6.1 Summary of Methodologies Used Sequoyah Nuclear Plant completed SMA for Units 1 and 2 in 1995. The SMA is documented in [11] that consisted of development of a Safe Shutdown Equipment List (SSEL), probabilistic approach for determining seismic demand based on 84% NEP, new building models, associated generation of ISRS, screening walkdowns, and HCLPF capacity calculations.
The screening walkdowns used the screening tables from Chapter 2 of EPRI NP6041SL [15]. The walkdowns were conducted by engineers trained in EPRI NP6041SL (the engineers attended the EPRI SMA AddOn course in addition to the SQUG Walkdown Screening and Seismic Evaluation Training Course), and were documented on Screening Evaluation Work Sheets from EPRI NP6041SL.
Anchorage capacity calculations used the CDFM criteria from EPRI NP6041SL. The seismic demand is based on a probabilistic approach that involves the generation of an ensemble of artificial earthquake (ground motion) time histories as well as structural and soil parameters values. The probabilistic approach of determining seismic demand is based on guidance from EPRI NP6041SL:
"For the Specified SME, the elastic computed response (SME demand) of structures and components mounted thereon should be defined at the 84% nonexceedance probability (NEP)."
The results of the probabilistic approach for development of seismic demand for Sequoyah Nuclear plant is documented in [11].
Figure 61 shows the fit of the 84th percentile of the ensemble of the 30 response spectra (of the 30 generated time histories) to the target spectral shape (NUREG/CR0098 median rock spectrum). Note this figure represents the input motion assuming the target spectrum is at the top of free field on the soil surface. Figure 62 shows the adjusted Sequoyah Nuclear Plant IPEEE HCLPF RLE response spectrum adjusted to 0.2g, compared to the ESEP RLGM response spectrum used for the Sequoyah Nuclear Plant ESEP. Note both spectra are rock input motions at the base of the containment structure. This demonstrates that the ESEP RLGM envelopes the RLGM used for SMA at all frequencies by an amplitude factor of 1.5.
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Sequoyah Nuclear Plant ESEP Report Figure 61: 84th Percentile of the Ensemble of the 30 Response Spectra Figure 62: Sequoyah Nuclear Plant IPEEE adjusted HCLPF vs ESEP Target HCLPF 6.2 HCLPF Screening Process For ESEP, the components are screened at RLGM (NUREG/CR0098 curve) anchored at 0.3g PGA. The screening tables in EPRI NP6041SL [15] are based on ground peak spectral accelerations of 0.8g and 1.2g. These both exceed the RLGM peak spectral acceleration. The anchorage capacity calculations Page 18
Sequoyah Nuclear Plant ESEP Report were on based floor response spectra developed for the Sequoyah Nuclear Plant IPEEE and scaled to the adjusted RLGM. Equipment for which the screening caveats were met and for which the anchorage capacity exceeded the RLGM seismic demand, can be screened out from ESEP seismic capacity determination because the HCLPF capacity exceeds the RLGM.
The Unit 1 ESEL contains 182 items. Of these, 27 are valves, both poweroperated and relief. In accordance with Table 24 of EPRI NP6041SL [15], 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 may be screened out from ESEP seismic capacity determination, subject to the caveat regarding large extended operators on small diameter lines.
The nonvalve components in the ESEL are generally screened based on the SMA methodology. If the SMA showed that the component met the EPRI NP6041SL screening caveats and the CDFM capacity exceeded the RLE demand, then the component can be screened out from the ESEP capacity determination.
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 NP6041SL [15] for the Seismic Margin Assessment process. Pages 226 through 230 of EPRI NP6041SL 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 nonradioactive or low radioactive environments. Seismic capability assessment of components which are inaccessible, in highradioactive 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 deenergize 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 similaritybasis should be developed before the walkdown during the seismic capability preparatory work (Step 3) by reference to drawings, calculations or specifications. The one component or each type which is selected should be thoroughly inspected which probably does mean deenergizing 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 onetoone correspondence is found, then subsequent items do not have to be inspected in as great a detail. Ultimately the Page 19
Sequoyah Nuclear Plant ESEP Report walkdown becomes a walk by of the component class as the SRT becomes confident that the construction pattern is typical. This procedure for inspection should be repeated for each component class; although, during the actual walkdown the SRT may be inspecting several classes of components in parallel. If serious exceptions to the drawings or questionable construction practices are found then the system or component class must be inspected in closer detail until the systematic deficiency is defined.
The 100% walk by is to look for outliers, lack of similarity, anchorage which is different from that shown on drawings or prescribed in criteria for that component, potential SI [Seismic Interaction] problems, situations that are at odds with the team members past experience, and any other areas of serious seismic concern. If any such concerns surface, then the limited sample size of one component of each type for thorough inspection will have to be increased.
The increase in sample size which should be inspected will depend upon the number of outliers and different anchorages, etc., which are observed. It is up to the SRT to ultimately select the sample size since they are the ones who are responsible for the seismic adequacy of all elements which they screen from the margin review. Appendix D gives guidance for sampling selection.
6.3.2 Application of Previous Walkdown Information Several ESEL items were previously walked down during the Sequoyah Nuclear Plant Units 1 and 2 seismic IPEEE program. 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.
Except for inaccessible items as described below in Section 7, in all cases it was determined that the HCLPF capacities established for these items under the seismic IPEEE program remained valid. Thus, all ESEL components that were part of the IPEEE program have a HCLPF capacity of 0.3g or greater and are thus adequate for ESEP [9].
6.3.3 Significant Walkdown Findings Consistent with the guidance from EPRI NP6041SL [15], no significant outliers and only one (1) anchorage concern was identified during the Sequoyah Nuclear Plant seismic walkdowns. The following findings were noted during the walkdowns.
One anchor for Sequoyah Unit 2 TDAFW Pump Control Panel 2L381 was observed to be significantly corroded. An evaluation was performed of the configuration assuming that the anchor was inactive.
The evaluation determined that the configuration (using 3 of 4 anchors) satisfied design requirements.
The corroded anchored is scheduled to be replaced in upcoming U2R20 outage.
Based on walkdown results, HCLPF capacity evaluations were recommended for the following twelve (12) components, on a bounding basis:
Turbine Driven Auxiliary Feedwater Pump Instrument Rack Page 20
Sequoyah Nuclear Plant ESEP Report R Panels Benchboard M Panels Vertical M and L Panels Main Control Room Ceiling Wall Mounted Panel Boric Acid Storage Tank TDAFWP Control Panel PHMS Transformers and Distribution Panel Valves Block Walls 6.4 HCLPF Calculation Process ESEL items not included in the previous IPEEE evaluations at Sequoyah were evaluated using the criteria in EPRI NP6041 [7]. 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 NP6041 as described in Section 6.2 and Performing HCLPF calculations considering various failure modes that include both structural failure modes (e.g. anchorage, load path etc.) and functional failure modes.
All HCLPF calculations were performed using the CDFM methodology and are documented in a TVA Calculation: CDQ999 2014 000140 SQN Expedited Seismic Evaluation Process (ESEP) HCLPF Capacity Calculation [17].
6.5 Functional Evaluations of Relays ESEP considers cabinets and equipment controls containing relays, contactors, switches, circuit breakers and other electrical and instrumentation components that could be affected by high frequency earthquake motions and that impact the operation of equipment in the ESEL.
A full scope SMA was performed to support the IPEEE for Sequoyah Nuclear Plant Units 1 and 2 as summarized in the IPEEE submittal [11]. A slightly modified version of the EPRI NP6041SL [15]
recommended approach was implemented for the Sequoyah Nuclear Plant IPEEE in order to increase efficiency. All relays were screened out by: (1) ground rules and assumptions; (2) comparison of design qualification test spectrum (or generic equipment ruggedness spectra ) with Seismic Margin Earthquake (SME) incabinet response spectrum; or (3) analysis showing that relay chatter does not disable safe shutdown equipment without the possibility of recovery. In all cases, equipment actuation was determined to not affect the safe shutdown capability of the equipment in the SSEL. Low ruggedness relays were screened out only if the effects of chatter could be reset by operator action.
All but two lowruggedness relays fell into this category. The remaining two relays were found to not be used in safe shutdown equipment at Sequoyah Nuclear Plant. The principal conclusion from the Page 21
Sequoyah Nuclear Plant ESEP Report IPEEE relay evaluation was that safe shutdown systems will not be adversely affected by relay malfunction during or after an SME.
For the Sequoyah Nuclear Plant ESEP analysis, an evaluation was performed to identify components that are (1) needed for FLEX implementation, (2) not on the Sequoyah IPEEE SSEL, and (3) that have the potential for relay chatter issues. The only cases identified are the FCV117 and FCV118 steam isolation valves that can isolate the steam supply to the TDAFW pump. In the event of a steam line break, both of these valves can receive a close signal if high temperature is detected in the TDAFW pump room. However, because these valves are motor operated valves (MOVs), with a Loss of Offsite Power (LOOP) the valves will not isolate even with a spurious close signal. Therefore, these valves do not present a problem for successful FLEX implementation. No other relay chatter cases were identified. No seismic capacity to demand relay evaluations were necessary for ESEP.
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.
Items previously included in the seismic IPEEE programs are not listed. Walkby verifications re confirmed the HCLPF capacities from the IPEEE, and the IPEEE 0.3g RLE HCLPF capacity exceeds the RLGM [18].
HCLPF capacity evaluations were performed for the nonIPEEE items on the ESEL, addressing both structural/anchorage and functional failure modes. The HCLPF capacity of each item is listed in the tables, with associated governing failure mode.
Rugged items not specifically evaluated are conservatively assigned a 0.50g HCLPF capacity based on the EPRI screening tables or by engineering judgment.
New prestaged and permanently installed FLEX items are not listed. TVA design criteria SQN DCV48.0 [19] requires that new FLEX items have HCLPF capacity exceeding the RLGM.
All ESEP components have a HCLPF capacity greater than the RLGM for the frequency range of 1 to 10Hz.
7.0 INACCESSIBLE ITEMS 7.1 Identification of ESEL Item Inaccessible for Walkdowns There are four (4) valves and seven (7) instrument racks that could not be walked down since they are in the Unit 1 Reactor Building (inaccessible area). These components Unit 2 counter parts were walked down during the recent Unit 2 outage. The Unit 2 components were determined to be acceptable. It is expected that the same conclusion can be made for the Unit 1 components. The following is a list of the Unit 1 components located in the Reactor Building that were not walked down:
1FCV 63118 Cold leg Accumulator Isolation Valve #1 1FVC 6367 Cold leg Accumulator Isolation Valve #4 1FCV 6380 Cold leg Accumulator Isolation Valve #3 1FCV 6398 Cold leg Accumulator Isolation Valve #2 Instrument Rack 1L182 located in Fan Room 2 Page 22
Sequoyah Nuclear Plant ESEP Report Instrument Rack 1L183 located in Fan Room 1 Instrument Rack 1L179 Instrument Rack 1L185 Instrument Rack 1L704 Instrument Rack 1L706 Instrument Rack 1L194 Also there are two (2) panels that could not be walked down in the Unit 1 Auxiliary Building because the components are in a Contaminated and Radiation Area. They are:
Instrument Rack 1L196 Instrument Rack 1L216 In addition, a walk by inside Unit 1 containment was not possible.
7.2 Planned Walkdown / Evaluation Schedule / Close Out The following Unit 1 components will be walked down in upcoming Unit 1 outage:
FCV 63118 Cold Leg Accumulator Isolation Valve #1 FVC 6367 Cold Leg Accumulator Isolation Valve #4 FCV 6380 Cold Leg Accumulator Isolation Valve #3 FCV 6398 Cold Leg Accumulator Isolation Valve #2 Instrument Rack 1L 182 located in Fan Room 2 Instrument Rack 1L 183 located in Fan Room 1 Instrument Rack 1L179 Instrument Rack 1L185 Instrument Rack 1L704 Instrument Rack 1L706 Instrument Rack 1L194 Instrument Rack 1L196 Instrument Rack 1L216 In addition, as performed inside the Unit 2 containment, a walk by will be conducted to verify that HCLPF capacity of at least 0.3g is maintained for IPEEE items on the ESEL. It is expected that the same conclusions will be made for the Unit 1 components that were completed for the counterpart components in Unit 2.
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Sequoyah Nuclear Plant ESEP Report 8.0 ESEP CONCLUSIONS AND RESULTS 8.1 Supporting Information Sequoyah Nuclear Plant Units 1 and 2 have performed the ESEP as an interim action in response to the NRCs 50.54(f) letter [1]. It was performed using the methodologies in the NRC endorsed guidance in EPRI 3002000704 [2].
The ESEP provides an important demonstration of seismic margin and expedites plant safety enhancements through evaluations and potential nearterm 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 Sequoyah Nuclear Plant Units 1 and 2 in response to the NRCs 50.54(f) letter. On March 12, 2014, NEI submitted to the NRC results of a study [21] 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 sitespecific seismic hazards show that there has not been an overall increase in seismic risk for the fleet of U.S. plants based on the reevaluated seismic hazards. As such, the current seismic design of operating reactors continues to provide a safety margin to withstand potential earthquakes exceeding the seismic design basis.
The NRCs May 9, 2014 NTTF 2.1 Screening and Prioritization letter [20] concluded that the fleet wide seismic risk estimates are consistent with the approach and results used in the Gl199 safety/risk assessment. The letter also stated that As a result, the staff has confirmed that the conclusions reached in Gl199 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 Sequoyah Nuclear Plant Units 1 and 2 was included in the fleet risk evaluation submitted in the March 12, 2014 NEI letter [21]; therefore, the conclusions in the NRCs May 9 letter also apply to Sequoyah Nuclear Plant Units 1 and 2.
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 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 ISRS calculations Broadening criteria for ISRS Response spectra enveloping criteria typically used in SSC analysis and testing applications Page 24
Sequoyah Nuclear Plant ESEP Report 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 NRCs 50.54(f) letter to demonstrate seismic margin through a review of a subset of the plant equipment that can be relied upon to protect the reactor core following beyond design basis seismic events. In order to complete the ESEP in an expedited amount of time, the RLGM used for the ESEP evaluation is a scaled version of the plants 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 riskbased SMA) is to be performed in accordance with EPRI 1025287 [22] . As identified in the Sequoyah Nuclear Plant Units 1 and 2 Seismic Hazard and GMRS submittal [9],
Sequoyah Nuclear Plant Units 1 and 2 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. Sequoyah Nuclear Plant Units 1 and 2 will complete that evaluation in accordance with the schedule identified in NEIs letter dated April 9, 2013 [23] and endorsed by the NRC in their May 7, 2013 letter [24].
8.2 Identification of Planned Modifications One modification was identified in unit 2, to repair a corroded anchor observed for TDAFW pump Control Panel 2L381.
8.3 Modification Implementation Schedule Plant modifications will be performed in accordance with the schedule identified in NEI letter dated April 9, 2013 [23], which states that plant modifications not requiring a planned refueling outage will be completed by December 2016 and modifications requiring a refueling outage will be completed within two planned refueling outages after December 31, 2014.
The plant modification identified in Section 8.2 requires a refueling outage that will be performed in the upcoming unit 2 refueling outage (U2R20) and will be completed by the end of December 2015.
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Sequoyah Nuclear Plant ESEP Report 8.4 Summary of Regulatory Commitments The following actions will be performed as a result of the ESEP.
Table 81: Summary of Regulatory Commitments Action Equipment Equipment
- ID Description Action Description Completion Date 1 NA N/A Perform seismic walkdowns, generate No later than the end HCLPF calculations, and design and of the second implement any necessary planned Unit 1 modifications for Unit 1 inaccessible refueling outage items listed in Section 7.1 after December 31, 2014 2 2L381 TDAFP Modify anchorage to replace No later than the end Control corroded anchor such that HCLPF of U2R20 Refueling Panel >RLGM Outage, December 31, 2015 3 N/A N/A Submit a letter to NRC summarizing Within 60 days the HCLPF results of Item 1 and following completion confirming implementation of the of ESEP activities, plant modifications associated with including Items 1 Item 2 through 2 Page 26
Sequoyah Nuclear Plant ESEP Report
9.0 REFERENCES
- 1. NRC (E Leeds and M Johnson) Letter to All Power Reactor Licensees et al., Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(f) Regarding Recommendations 2.1, 2.3 and 9.3 of the NearTerm Task Force Review of Insights from the Fukushima DaiIchi Accident, March 12, 2012.
- 2. EPRI 3002000704, Seismic Evaluation Guidance, Augmented Approach for the Resolution of Fukushima NearTerm Task Force Recommendation 2.1: Seismic, May 2013.
- 3. TVA Letter to U.S. NRC, Tennessee Valley Authority - 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 EA12049),
February 28, 2013.
- 4. TVA Letter to U.S. NRC, Second SixMonth Status Report and Revised Overall Integrated Plan in Response to the March 12, 2012, Commission Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for BeyondDesignBasis External Events (Order Number EA12049) for Sequoyah Nuclear Plant, February 28, 2014.
- 5. TVA Letter to U.S. NRC, Third SixMonth Status Report in Response to the March 12, 2012, Commission Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond Design Basis External Events (Order Number EA12049) for Browns Ferry Nuclear Plant (TAC Nos. MF0864 and MF0865), August 28, 2014.
- 7. NEI 1206, Diverse and Flexible Coping Strategies (FLEX) Implementation Guide, Revision 0, August 2012.
Sequoyah Nuclear Plant.
- 9. TVA Letter to U.S. NRC, letter number CNL14038, Tennessee Valley Authoritys Seismic Hazard and Screening Report (CEUS Sites), response to NRC Request for Information Pursuant to 10 CFR 50.54(f) regarding Recommendation 2.1 of the NearTerm Task Force Review of Insights from the Fukushima Daiichi Accident, March 31, 2014
- 10. U.S. NRC NUREG/CR0098, Development of Criteria for Seismic Review of Selected Nuclear Power Plants, May 1978.
- 11. TVA Letter from R. H. Shell to U.S. NRC, Sequoyah Nuclear Plant (SQN) - Generic Letter GL 88 20, Supplement No. 4, Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities - 10CFR 50.54(f), June 29, 1995.
- 12. NRC Letter to TVA, Sequoyah Nuclear Plant, Units 1 and 2 - Request for Additional Information on Individual Plant Examination of External Events (TAC Nos. M86374 and M86375), August 2, 2000.
- 13. Letter from Pedro Salas to NRC, Sequoyah Nuclear Plant Units 1 and 2 - Response to Request for Additional Information of the Individual Plant Examination of External Events (IPEEE) (TAC Nos. M83674 and M83675), December 5, 2000.
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Sequoyah Nuclear Plant ESEP Report
- 14. Letter from NRC to J. A. Scalice, Sequoyah Nuclear Plant, Units 1 and 2 - Review of Sequoyah Individual Plant Examination of External Events Submittal (TAC Nos. M83764 and M83675),
February 21, 2001.
- 15. EPRINP6041SL, Methodology for Assessment of Nuclear Power Plant Seismic Margin, Revision 1, August 1991.
- 16. EPRI TR103959, Methodology for Developing Seismic Fragilities, July 1994.
HCLPF Capacity Calculation.
- 18. TVA letter to U.S. NRC, Letter number CNL14013, Highlights of Improvements to the Sequoyah Nuclear Plant IPEEE Seismic Analysis Results and Supplemental Response to NRC Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(f) regarding the Sequoyah Nuclear Plant Unit 1 Seismic Walkdown Results of Recommendations 2.3 of the NearTerm Task Force Review of Insights from the Fukushima Daiichi Accident, January 31, 2014.
- 20. 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 ReEvaluations for Recommendation 2.1 of the NearTerm Task Force Review of Insights From the Fukushima DaiIchi Accident, May 9, 2014.
- 21. 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.
- 22. EPRI 1025287, Seismic Evaluation Guidance: Screening, Prioritization and Implementation Details (SPID) for the Resolution of Fukushima NearTerm Task Force Recommendation 2.1:
Seismic. Electric Power Research Institute, February 2013.
- 23. 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.
- 24. 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 NearTerm Task Force Recommendation 2.1: Seismic, as an Acceptable Alternative to the March 12, 2012, Information Request for Seismic Reevaluations, May 7, 2013.
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Sequoyah Nuclear Plant ESEP Report ATTACHMENT A - SEQUOYAH NUCLEAR PLANT ESEL Page A1
Sequoyah Nuclear Plant ESEP Report TABLE A1: Expedited Seismic Equipment List (ESEL) for Sequoyah Nuclear Plant ESEL Equipment Operating State Item Number Normal Desired ID Description State State Notes/Comments Steam Generator #3 Main Steam Safety 1 VLV1512 Operational Operational Valve Steam Generator #2 Main Steam Safety 2 VLV1517 Operational Operational Valve Steam Generator #1 Main Steam Safety 3 VLV1522 Operational Operational Valve Steam Generator #4 Main Steam Safety 4 VLV1527 Operational Operational Valve 5 PCV15 Steam Generator #1 ARV (SG PORV) Operational Operational Fail closed on loss of Train A essential air 6 PCV112 Steam Generator #2 ARV (SG PORV) Operational Operational Fail closed on loss of Train B essential air 7 PCV123 Steam Generator #3 ARV (SG PORV) Operational Operational Fail closed on loss of Train A essential air 8 PCV130 Steam Generator #4 ARV (SG PORV) Operational Operational Fail closed on loss of Train B essential air Steam Generator #1 ARV (SG PORV) Hand Local control of steam generator PORV 9 PCV15 Operational Operational wheel during ELAP Steam Generator #2 ARV (SG PORV) Local Emergency control station per EA12 10 Air Bottles Operational Operational Control Station FLEX compressed air cylinders Steam Generator #3 ARV (SG PORV) Local Emergency control station per EA12 11 Air Bottles Operational Operational Control Station FLEX compressed air cylinders Steam Generator #4 ARV (SG PORV) Hand Local control of steam generator PORV 12 PCV130 Operational Operational wheel during ELAP Page A2
Sequoyah Nuclear Plant ESEP Report TABLE A1: Expedited Seismic Equipment List (ESEL) for Sequoyah Nuclear Plant (Continued)
ESEL Equipment Operating State Item Number Normal Desired ID Description State State Notes/Comments 13 L501 PCV112 Local Control Station Operational Operational Emergency control station per EA12 14 L502 PCV123 Local Control Station Operational Operational Emergency control station per EA12 15 PMP3142 Turbine Driven AFW Pump Standby Operating Automatic start on LOOP Normal power supply is from 125V Vital 16 FCV151 TDAFW Pump Trip and Throttle Valve Closed Open Battery Board III 17 FCV152 TDAFW Pump Governor Valve Closed Open Fails open on loss of DC control power These components are on the ESEL if manual 18 XS4657 AFWT AS Backup Control Transfer Switch Operational Operational operation of TDAFW is implemented TDAFW Pump Trip and Throttle Valve 19 HS151B Operational Operational Handswitch 20 SI4656B TDAFW Pump Speed Indicator Operational Operational 21 FIC4657 TDAFW Pump Master Speed Controller Operational Operational 22 L381 TDAFW Pump Control Panel Operational Operational TDAFW Pump Discharge Pressure 23 PI3138 Operational Operational Indicator Portable delta press gauge can be used for 24 L215 AFW Flow Monitoring Panel Operational Operational local monitoring of AFW flow Page A3
Sequoyah Nuclear Plant ESEP Report TABLE A1: Expedited Seismic Equipment List (ESEL) for Sequoyah Nuclear Plant (Continued)
ESEL Equipment Operating State Item Number Normal Desired ID Description State State Notes/Comments AFW Flow to Steam Generator #3 Flow 25 FT3147 Operational Operational Transmitter AFW Flow to Steam Generator #2 Flow 26 FT3155 Operational Operational Transmitter AFW Flow to Steam Generator #1 Flow 27 FT3163 Operational Operational Transmitter AFW Flow to Steam Generator #4 Flow 28 FT3170 Operational Operational Transmitter AFW Flow to Steam Generator #3 Flow 29 FI3147B Operational Operational Indication AFW Flow to Steam Generator #2 Flow 30 FI3155B Operational Operational Indication AFW Flow to Steam Generator #1 Flow 31 FI3163B Operational Operational Indication AFW Flow to Steam Generator #4 Flow 32 FI3170B Operational Operational Indication AFW Flow to Steam Generator #3 Flow 33 L341 Operational Operational Transmitter Rack AFW Flow to Steam Generator #2 Flow 34 L217 Operational Operational Transmitter Rack AFW Flow to Steam Generator #1 Flow 35 L216 Operational Operational Transmitter Rack AFW Flow to Steam Generator #4 Flow 36 L703 Operational Operational Indication Rack Page A4
Sequoyah Nuclear Plant ESEP Report TABLE A1: Expedited Seismic Equipment List (ESEL) for Sequoyah Nuclear Plant (Continued)
ESEL Equipment Operating State Item Number Normal Desired ID Description State State Notes/Comments Fails open on loss of AX power or control air.
37 LCV3172 Steam Generator #3 Level Control Valve Closed Open Backup air supply bottles available. Manual operation with handwheel is available.
Fails open on loss of AC power or control air.
38 LCV3173 Steam Generator #2 Level Control Valve Closed Open Backup air supply bottle available. Manual operation with handwheel is available.
Fails open on loss of AC power or control air.
39 LCV3174 Steam Generator #1 Level Control Valve Closed Open Backup air supply bottles available. Manual operation with handwheel is available.
Fails open on loss of AC power or control air.
40 LCV3175 Steam Generator #4 Level Control Valve Closed Open Backup air supply bottles available. Manual operation with handwheel is available.
Steam Generator #3 Level Control Valve 41 XS3172 Operational Operational Transfer Switch Steam Generator #2 Level Control Valve 42 XS3173 Operational Operational Transfer Switch Steam Generator #1 Level Control Valve 43 XS3174 Operational Operational Transfer Switch Steam Generator #4 Level Control Valve 44 XS3175 Operational Operational Transfer Switch 45 L11A Steam Generator Level Control Panel Operational Operational 46 L11B Steam Generator Level Control Panel Operational Operational Steam Generator #3 Level Control Valve 47 HS3172B Operational Operational Hand Switch Page A5
Sequoyah Nuclear Plant ESEP Report TABLE A1: Expedited Seismic Equipment List (ESEL) for Sequoyah Nuclear Plant (Continued)
ESEL Equipment Operating State Item Number Normal Desired ID Description State State Notes/Comments Steam Generator #2 Level Control Valve 48 HS3173B Operational Operational Hand Switch Steam Generator #1 Level Control Valve 49 HS3174B Operational Operational Hand Switch Steam Generator #4 Level Control Valve 50 HS3175B Operational Operational Hand Switch Steam Generator #2 Level Control Valve 51 L661 Operational Operational FLEX Backup Air Station Steam Generator #1 Level Control Valve 52 L662 Operational Operational FLEX Backup Air Station Steam Generator #3 Level Control Valve 53 L663 Operational Operational FLEX Backup Air Station Steam Generator #4 Level Control Valve 54 L664 Operational Operational FLEX Backup Air Station Steam Generator #1 Wide Range Level 55 LT343 Operational Operational Transmitter Steam Generator #2 Wide Range Level 56 1LT356 Operational Operational Transmitter Steam Generator #2 Wide Range Level Unit 2 SG3 Level Transmitter on Rack L182 57 2LT356 Operational Operational Transmitter Unit 1 SG3 Level Transmitter on Rack L706 Steam Generator #3 Wide Range Level 58 LT398 Operational Operational Transmitter Steam Generator #4 Wide Range Level 59 LT3111 Operational Operational Transmitter 60 L183 Steam Generator Level Transmitter Rack Operational Operational Page A6
Sequoyah Nuclear Plant ESEP Report TABLE A1: Expedited Seismic Equipment List (ESEL) for Sequoyah Nuclear Plant (Continued)
ESEL Equipment Operating State Item Number Normal Desired ID Description State State Notes/Comments Steam Generator Level Transmitter Rack 61 L706 Operational Operational (Unit 1) 62 L704 Steam Generator Level Transmitter Rack Operational Operational 63 L185 Steam Generator Level Transmitter Rack Operational Operational Steam Generator Level Transmitter Rack 64 L182 Operational Operational (Unit 2)
Steam Generator #1 Wide Range Level 65 LI343 Operational Operational Indicator Steam Generator #2 Wide Range Level 66 LI356 Operational Operational Indicator Steam Generator #3 Wide Range Level 67 LI398 Operational Operational Indicator Steam Generator #4 Wide Range Level 68 LI3111 Operational Operational Indicator Steam Generator #1 Discharge Pressure 69 PT12A Operational Operational 120V VIPB I Rack 3 Transmitter Steam Generator #2 Discharge Pressure 70 PT19A Operational Operational 120V VIPB I Rack 3 Transmitter Steam Generator #3 Discharge Pressure 71 PT120A Operational Operational 120V VIPB I Rack 4 Transmitter Steam Generator #4 Discharge Pressure 72 PT127A Operational Operational 120V VIPB I Rack 4 Transmitter Steam Generator Discharge Pressure 73 L194 Operational Operational Transmitter Rack Page A7
Sequoyah Nuclear Plant ESEP Report TABLE A1: Expedited Seismic Equipment List (ESEL) for Sequoyah Nuclear Plant (Continued)
ESEL Equipment Operating State Item Number Normal Desired ID Description State State Notes/Comments Steam Generator Discharge Pressure 74 L196 Operational Operational Transmitter Rack Steam Generator #1 Discharge Pressure 75 PI12D Operational Operational Channel A input Indicator Steam Generator #2 Discharge Pressure 76 PI19D Operational Operational Channel A input Indicator Steam Generator #3 Discharge Pressure 77 PI120D Operational Operational Channel A input Indication Steam Generator #4 Discharge Pressure 78 PI127D Operational Operational Channel A input Indication DCN 23191 will seismically qualify CST to 2x 79 1TNK0020229 Unit 1 Condensate Storage Tank Operational Operational SSE HCLPF DCN 23191 will seismically qualify CST to 2x 80 2TNK0020232 Unit 2 Condensate Storage Tank Operational Operational SSE HCLPF Switchover to ERCW header 480V MOV 81 FCV3179A ERCW Header B AFW Supply Valve Closed Open Board 1(2)B2B/11E or hand wheel Switchover to ERCW header 480V MOV 82 FCV3179B ERCW Header B AFW Supply Valve Closed Open Board 1(2)B2B/11B or hand wheel Switchover to ERCW header 480V MOV 83 FCV3136A ERCW Header A AFW Supply Valve Closed Open Board 1(2)A2A/2E or hand wheel Switchover to ERCW header 480V MOV 84 FCV3136B ERCW Header A AFW Supply Valve Closed Open Board 1(2)A2A/2B or hand wheel 85 FCV63118 Cold Leg Accumulator #1 Isolation Valve Open Closed 86 FCV6398 Cold Leg Accumulator #2 Isolation Valve Open Closed Page A8
Sequoyah Nuclear Plant ESEP Report TABLE A1: Expedited Seismic Equipment List (ESEL) for Sequoyah Nuclear Plant (Continued)
ESEL Equipment Operating State Item Number Normal Desired ID Description State State Notes/Comments 87 FCV6380 Cold Leg Accumulator #3 Isolation Valve Open Closed 88 FCV6367 Cold Leg Accumulator #4 Isolation Valve Open Closed 89 1PMP6310A Safety Injection Pump Standby Operational 90 2PMP6310A Safety Injection Pump Standby Operational 91 HS6310A Safety Injection Pump Hand Switch Operational Operational 92 M6 Panel M6 Operational Operational 93 1TNK0620239 Boric Acid Tank (BAT) A Available Available 94 2TNK0620239 Boric Acid Tank (BAT) B Available Available 95 0TNK0620243 Boric Acid Tank (BAT) C Available Available 96 HEX0740015 RHR Heat Exchanger 1A Intact Intact RWST gravity feed path 97 TNK063044 RWST Operational Operational 98 PCV0680340A RCS Pressurizer Power Relief Valve Operational Operational 125V DC Vital Battery Board I 99 HS68340 AA Pressurizer PORV Hand Switch Operational Operational Page A9
Sequoyah Nuclear Plant ESEP Report TABLE A1: Expedited Seismic Equipment List (ESEL) for Sequoyah Nuclear Plant (Continued)
ESEL Equipment Operating State Item Number Normal Desired ID Description State State Notes/Comments Containment Pressure Differential 100 PDT3044 Operational Operational Transmitter Containment Pressure Differential 101 PDI3044 Operational Operational Indicator 102 1FAN0300038 Unit 1 Containment Air Return Fan A Standby Operational Powered by 480V SDB 1A1 103 2FAN0300038 Unit 2 Containment Air Return Fan A Standby Operational Powered by 480V SDB 2A1 Containment Air Return Fan A Hand 104 HS3038A Operational Operational Switch 105 M9 Panel M9 Operational Operational 106 1XFA2681AA PHMS Xfrm 1A Operational Operational Power supply to hydrogen igniters 107 2XFA2682AA PHMS Xfrm 2A Operational Operational Power supply to hydrogen igniters 108 1PNL268YA 120V AC PHMS Distribution Panel 1A Operational Operational Power supply to hydrogen igniters 109 2PNL268YC 120V AC PHMS Distribution Panel 2A Operational Operational Power supply to hydrogen igniters Use to supply dieselpowered FLEX 110 TNK01838 1AA 7Day Oil Supply Tank Available Available equipment Use to supply dieselpowered FLEX 111 TNK01840 1BB 7Day Oil Supply Tank Available Available equipment Use to supply dieselpowered FLEX 112 TNK01839 2AA 7Day Oil Supply Tank Available Available equipment Page A10
Sequoyah Nuclear Plant ESEP Report TABLE A1: Expedited Seismic Equipment List (ESEL) for Sequoyah Nuclear Plant (Continued)
ESEL Equipment Operating State Item Number Normal Desired ID Description State State Notes/Comments Use to supply dieselpowered FLEX 113 TNK01841 2BB 7Day Oil Supply Tank Available Available equipment 114 1BDB201DJ 480V Shutdown Board A1A Operational Operational 115 1BDB201DK 480V Shutdown Board A2A Operational Operational 116 1BDB201DL 480V Shutdown Board B1B Operational Operational 117 1BDB201DM 480V Shutdown Board B2B Operational Operational 118 2BDB201DN 480V Shutdown Board A1A Operational Operational 119 2BDB201DO 480V Shutdown Board A2A Operational Operational 120 2BDB201DP 480V Shutdown Board B1B Operational Operational 121 2BDB201DQ 480V Shutdown Board B2B Operational Operational Operational Operational Power to safety injection accumulator 122 1BDC201GG 480V Reactor MOV Board 1A1A isolation MOVs Operational Operational Power to safety injection accumulator 123 1BDC201GJ 480V Reactor MOV Board 1B1B isolation MOVs Operational Operational Power to safety injection accumulator 124 2BDC201GL 480V Reactor MOV Board 2A1A isolation MOVs Operational Operational Power to safety injection accumulator 125 2BDC201GN 480V Reactor MOV Board 2B1B isolation MOVs Page A11
Sequoyah Nuclear Plant ESEP Report TABLE A1: Expedited Seismic Equipment List (ESEL) for Sequoyah Nuclear Plant (Continued)
ESEL Equipment Operating State Item Number Normal Desired ID Description State State Notes/Comments Operational Operational Power to high pressure FLEX pump 126 1BDC201JE 480V C&A Vent Board 1A1A Repowered by 6.9kV FLEX diesel generator Operational Operational Power to intermediate pressure FLEX pump 127 1BDC201JF 480V C&A Vent Board 1A2A Repowered by 6.9kV FLEX diesel generator Power to high pressure FLEX pump 128 2BDC201JJ 480V C&A Vent Board 2A1A Operational Operational Repowered by 6.9kV FLEX diesel generator Power to intermediate pressure FLEX Pump 129 2BDC201JK 480V C&A Vent Board 2A2A Operational Operational Repowered by 6.9kV FLEX diesel generator 130 BDG250KE 125V DC Vital Battery Board I Operational Operational 131 BDG250KF 125V DC Vital Battery Board II Operational Operational 132 BDG250KG 125V DC Vital Battery Board III Operational Operational 133 BDG250KH 125V DC Vital Battery Board IV Operational Operational 134 1BDE250NCD 120V AC Vital Instrument Power Board 1I Operational Operational 135 1BDE250NEE 120V AC Vital Instrument Power Board 1II Operational Operational 120V AC Vital Instrument Power Board 1 136 1BDE250NGF Operational Operational III 120V AC Vital Instrument Power Board 1 137 1BDE250NJG Operational Operational IV 138 2BDE250NDD 120V AC Vital Instrument Power Board 2I Operational Operational Page A12
Sequoyah Nuclear Plant ESEP Report TABLE A1: Expedited Seismic Equipment List (ESEL) for Sequoyah Nuclear Plant (Continued)
ESEL Equipment Operating State Item Number Normal Desired ID Description State State Notes/Comments 139 2BDE250NFE 120V AC Vital Instrument Power Board 2II Operational Operational 120V AC Vital Instrument Power Board 2 140 2BDE250NHF Operational Operational III 120V AC Vital Instrument Power Board 2 141 2BDE250NKG Operational Operational IV 142 0BATB250QV 125V DC Vital Battery I Operational Operational 143 0BATB250QW 125V DC Vital Battery II Operational Operational 144 0BATB250QX 125V DC Vital Battery III Operational Operational 145 0BATB250QY 125V DC Vital Battery IV Operational Operational 146 0CHGB250QE 125V DC Vital Battery Charger I Operational Operational 147 0CHGB250QG 125V DC Vital Battery Charger II Operational Operational 148 0CHGB250QH 125V DC Vital Battery Charger III Operational Operational 149 0CHGB250QJ 125V DC Vital Battery Charger IV Operational Operational 150 1INVB250QL 120V AC Vital Inverter 1I Operational Operational 151 1INVB250QN 120V AC Vital Inverter 1II Operational Operational Page A13
Sequoyah Nuclear Plant ESEP Report TABLE A1: Expedited Seismic Equipment List (ESEL) for Sequoyah Nuclear Plant (Continued)
ESEL Equipment Operating State Item Number Normal Desired ID Description State State Notes/Comments 152 1INVB250QR 120V AC Vital Inverter 1III Operational Operational 153 1INVB250QT 120V AC Vital Inverter 1IV Operational Operational 154 2INVB250QM 120V AC Vital Inverter 2I Operational Operational 155 2INVB250QP 120V AC Vital Inverter 2II Operational Operational 156 2INVB250QS 120V AC Vital Inverter 2III Operational Operational 157 2INVB250QU 120V AC Vital Inverter 2IV Operational Operational 158 1XE925001 N31 Neutron Detector Operational Operational Unit 1 NIS Channel 1 159 1XM925001A N31 Neutron Source Range Amplifier Operational Operational N31 Neutron Source Range Optical 160 1XM925001B Operational Operational Isolator 161 1XI925 N31 Signal Processor App R Operational Operational 162 1XX925001 N31 Source Range Indicator Operational Operational 163 1XI925001A N31B Source Range Indicator Operational Operational 164 2XE925002 N32 Neutron Detector Operational Operational Unit 2 NIS Channel 2 Page A14
Sequoyah Nuclear Plant ESEP Report TABLE A1: Expedited Seismic Equipment List (ESEL) for Sequoyah Nuclear Plant (Continued)
ESEL Equipment Operating State Item Number Normal Desired ID Description State State Notes/Comments 165 2XM925002A N32 Neutron Source Range Amplifier Operational Operational N32 Neutron Source Range Optical 166 2XE925002B Operational Operational Isolator 167 2XI925 N32 Signal Processor App R Operational Operational 168 2XX925002 N32 Source Range Indicator Operational Operational 169 2XI925002B N32B Source Range Indicator Operational Operational 170 L10 Instrument Rack Operational Operational 171 M4 Instrument Panel Operational Operational 172 M13 Instrument Panel Operational Operational 173 PT6869 RCS Loop WR Pressure Transmitter Loop 1 Operational Operational 174 PT6866 RCS Loop WR Pressure Transmitter Loop 3 Operational Operational 175 PI6869 RCS Loop WR Pressure Indication Loop 1 Operational Operational 176 PI6866A RCS Loop WR Pressure Indication Loop 3 Operational Operational 177 L388 Instrument panel Operational Operational Page A15
Sequoyah Nuclear Plant ESEP Report TABLE A1: Expedited Seismic Equipment List (ESEL) for Sequoyah Nuclear Plant (Continued)
ESEL Equipment Operating State Item Number Normal Desired ID Description State State Notes/Comments 178 L340 Instrument Panel Operational Operational 179 L649 Instrument Panel Operational Operational 180 R4 Instrument Rack Operational Operational 181 R5 Instrument Rack Operational Operational 182 TE6818 Cold Leg WR Temperature Element Loop 1 Operational Operational 183 TE6841 Cold Leg WR Temperature Element Loop 2 Operational Operational 184 TE6860 Cold Leg WR Temperature Element Loop 3 Operational Operational 185 TE6883 Cold Leg WR Temperature Element Loop 4 Operational Operational Cold Leg WR Temperature Indication Loop 186 TI6818 Operational Operational 1
Cold Leg WR Temperature Indication Loop 187 TI6841 Operational Operational 2
Cold Leg WR Temperature Indication Loop 188 TI6860 Operational Operational 3
Cold Leg WR Temperature Indication Loop 189 TI6883 Operational Operational 4
190 TE681 Hot Leg WR Temperature Element Loop 1 Operational Operational Page A16
Sequoyah Nuclear Plant ESEP Report TABLE A1: Expedited Seismic Equipment List (ESEL) for Sequoyah Nuclear Plant (Continued)
ESEL Equipment Operating State Item Number Normal Desired ID Description State State Notes/Comments 191 TE6824 Hot Leg WR Temperature Element Loop 2 Operational Operational 192 TE6843 Hot Leg WR Temperature Element Loop 3 Operational Operational 193 TE6865 Hot Leg WR Temperature Element Loop 4 Operational Operational Hot Leg WR Temperature Indication Loop 194 TI681 Operational Operational 1
Hot Leg WR Temperature Indication Loop 195 TI6824 Operational Operational 2
Hot Leg WR Temperature Indication Loop 196 TI6843 Operational Operational 3
Hot Leg WR Temperature Indication Loop 197 TI6865 Operational Operational 4
198 M5 Instrument Panel Operational Operational 199 R2 Instrument Rack Operational Operational 200 R6 Instrument Rack Operational Operational 201 LT68325C RCS Pressurizer Level Transmitter Operational Operational 202 LI68325C RCS Pressurizer Level Indication Operational Operational 203 L179 Instrumentation Panel Operational Operational Page A17
Sequoyah Nuclear Plant ESEP Report TABLE A1: Expedited Seismic Equipment List (ESEL) for Sequoyah Nuclear Plant (Continued)
ESEL Equipment Operating State Item Number Normal Desired ID Description State State Notes/Comments 204 0DG360000A 480V FLEX Diesel Generator Standby Operational 205 0DG360003A 6.9kV FLEX Diesel Generator Standby Operational 206 1PNLA082TU Diesel Generator 1BB Control Panel Standby Operational 207 1PNLA082TT Diesel Generator G 1AA Control Panel Standby Operational 208 2PNLA082TV Diesel Generator 2AA Control Panel Standby Operational 209 2PNLA082TW Diesel Generator 2BB Control Panel Standby Operational 210 0BD3600003A FLEX Diesel Generator 3A Switchgear Standby Operational 0BKR360 211 FLEX DG 3A Switchgear Breaker A2 Standby Operational DCN 23197 0003A/1/A2 0TANK360113 6900V 3MW FLEX DG Fuel Oil Storage 212 Standby Operational DCN 23197 Tank 3A 0SW360 6900V 3MW FLEX Diesel GEN 3A Fused 213 Standby Operational DCN 23197 0003A/1 Disconnect Switch 0XFMR360 6900V480V 3MW FLEX Diesel GEN 3A 20 214 Standby Operational DCN 23197 3A/1 KVA Dry Type Transformer 0XFMR360 480V120/240V 3MW FLEX Diesel GEN 3A 215 Standby Operational DCN 23197 3A/2 5 KVA Dry Type Transformer 0DPL360 480Volt Distribution Panel with 100A 216 Standby Operational DCN 23197 0003A/1 Main Circuit Breaker Page A18
Sequoyah Nuclear Plant ESEP Report TABLE A1: Expedited Seismic Equipment List (ESEL) for Sequoyah Nuclear Plant (Continued)
ESEL Equipment Operating State Item Number Normal Desired ID Description State State Notes/Comments 0DPL360 120/240 VAC Panelboard 217 Standby Operational DCN 23197 0003A/2 0FU1360 Primary Cntrl Fuse for Fuel Oil Pump A 218 Standby Operational DCN 23197 0103A Starter 0FU1360 Secondary Cntrl Fuse for Fuel Oil Pump A 219 Standby Operational DCN 23197 0103B Starter 0FU1360 Primary Cntrl Fuse for Fuel Oil Pump A 220 Standby Operational DCN 23197 0103C Starter 0HS360103C Fuel Oil Transfer Pmp A Emer Stop SW 221 Standby Operational DCN 23197 0PMP360103 Fuel Oil System Transfer Pump 3A 222 Standby Operational DCN 23197 0RES360003A 3MW FLEX Diesel Generator A Neutral 223 Standby Operational DCN 23197 Grounding Resistor 0STR3600103 3MW FLEX Diesel Generator Fuel Oil 224 Standby Operational DCN 23197 Transfer Pump Starter 0LS3600103 3MW FLEX Diesel Generator Fuel Oil Float 225 Standby Operational DCN 23197 Switch (Fill Pump Control) 0XSW082 226 Transfer Switch 1A Standby Operational DCN 23197 0001A 0XSW082 227 Transfer Switch 2A Standby Operational DCN 23197 0002A 0XSW082 228 Transfer Switch 2B Standby Operational DCN 23197 0003B 0XSW082 229 Transfer Switch 1B Standby Operational DCN 23197 0004B Page A19
Sequoyah Nuclear Plant ESEP Report TABLE A1: Expedited Seismic Equipment List (ESEL) for Sequoyah Nuclear Plant (Continued)
ESEL Equipment Operating State Item Number Normal Desired ID Description State State Notes/Comments 1BDA202CM 230 6.9kV Shutdown Board 1AA Operational Operational A
231 2BDA202COA 6.9kV Shutdown Board 2AA Operational Operational 232 1BDA202CN 6.9kV Shutdown Board 1BB Operational Operational 233 2BDA202CP 6.9kV Shutdown Board 2BB Operational Operational 234 1OXF202DL 480V Shutdown Transformer 1B1 Operational Operational Listed as 1XFA2020317 in IPEEE Listed as 1XFA2020319 in IPEEE 235 1OXF202DM 480V Shutdown Transformer 1B2 Operational Operational Listed as 1XFA2020313 in IPEEE 236 1OXF202DJ 480V Shutdown Transformer 1A1 Operational Operational Listed as 1XFA2020315 in IPEEE 237 1OXF202DK 480V Shutdown Transformer 1A2 Operational Operational 238 2OXF202DN 480V Shutdown Transformer 2A1 Operational Operational Listed as 2XFA2020315 in IPEEE Listed as 2XFA2020313 in IPEEE 239 2OXF202DO 480V Shutdown Transformer 2A2 Operational Operational Listed as 2XFA2020319 in IPEEE 240 2OXF202DP 480V Shutdown Transformer 2B1 Operational Operational Listed as 2XFA2020317 in IPEEE 241 2OXF202DQ 480V Shutdown Transformer 2B2 Operational Operational Powered by 480V C&A Vent BD 1A2, DCN 242 1PMP360IP01 U1 Intermediate Pressure FLEX Pump Standby Operational 23193 Page A20
Sequoyah Nuclear Plant ESEP Report TABLE A1: Expedited Seismic Equipment List (ESEL) for Sequoyah Nuclear Plant (Continued)
ESEL Equipment Operating State Item Number Normal Desired ID Description State State Notes/Comments Powered by 480V C&A Vent BD 2A2, DCN 243 2PMP360IP01 U2 Intermediate Pressure FLEX Pump Standby Operational 23193 Powered by 480V C&A Vent BD 1A1, DCN 244 1PMP360HPCS U1 High Pressure FLEX Pump Standby Operational 23193 Powered by 480V C&A Vent BD 2A1, DCN 245 2PMP360HPCS U2 High Pressure FLEX Pump Standby Operational 23193 SQN1IGN268 246 Unit 1 Train A Hydrogen Igniter Standby Operational 0142A SQN1IGN268 247 Unit 1 Train A Hydrogen Igniter Standby Operational 0130A SQN1IGN268 248 Unit 1 Train A Hydrogen Igniter Standby Operational 0125A SQN1IGN268 249 Unit 1 Train A Hydrogen Igniter Standby Operational 0123A SQN1IGN268 250 Unit 1 Train A Hydrogen Igniter Standby Operational 0116A SQN1IGN268 251 Unit 1 Train A Hydrogen Igniter Standby Operational 0128A SQN1IGN268 252 Unit 1 Train A Hydrogen Igniter Standby Operational 0129A SQN1IGN268 253 Unit 1 Train A Hydrogen Igniter Standby Operational 0114A SQN1IGN268 254 Unit 1 Train A Hydrogen Igniter Standby Operational 0133A SQN1IGN268 255 Unit 1 Train A Hydrogen Igniter Standby Operational 0102A Page A21
Sequoyah Nuclear Plant ESEP Report TABLE A1: Expedited Seismic Equipment List (ESEL) for Sequoyah Nuclear Plant (Continued)
ESEL Equipment Operating State Item Number Normal Desired ID Description State State Notes/Comments SQN1IGN268 256 Unit 1 Train A Hydrogen Igniter Standby Operational 0115A SQN1IGN268 257 Unit 1 Train A Hydrogen Igniter Standby Operational 0132A SQN1IGN268 258 Unit 1 Train A Hydrogen Igniter Standby Operational 0108A SQN1IGN268 259 Unit 1 Train A Hydrogen Igniter Standby Operational 0127A SQN1IGN268 269 Unit 1 Train A Hydrogen Igniter Standby Operational 0155A SQN1IGN268 270 Unit 1 Train A Hydrogen Igniter Standby Operational 0136A SQN1IGN268 271 Unit 1 Train A Hydrogen Igniter Standby Operational 0131A SQN1IGN268 272 Unit 1 Train A Hydrogen Igniter Standby Operational 0121A SQN1IGN268 273 Unit 1 Train A Hydrogen Igniter Standby Operational 0122A SQN1IGN268 274 Unit 1 Train A Hydrogen Igniter Standby Operational 0135A SQN1IGN268 275 Unit 1 Train A Hydrogen Igniter Standby Operational 0159A SQN1IGN268 276 Unit 1 Train A Hydrogen Igniter Standby Operational 0126A SQN1IGN268 277 Unit 1 Train A Hydrogen Igniter Standby Operational 0107A Page A22
Sequoyah Nuclear Plant ESEP Report TABLE A1: Expedited Seismic Equipment List (ESEL) for Sequoyah Nuclear Plant (Continued)
ESEL Equipment Operating State Item Number Normal Desired ID Description State State Notes/Comments SQN2IGN268 278 Unit 2 Train A Hydrogen Igniter Standby Operational 0226A SQN2IGN268 279 Unit 2 Train A Hydrogen Igniter Standby Operational 0235A SQN2IGN268 280 Unit 2 Train A Hydrogen Igniter Standby Operational 0202A SQN2IGN268 281 Unit 2 Train A Hydrogen Igniter Standby Operational 0223A SQN2IGN268 282 Unit 2 Train A Hydrogen Igniter Standby Operational 0222A SQN2IGN268 283 Unit 2 Train A Hydrogen Igniter Standby Operational 0231A SQN2IGN268 284 Unit 2 Train A Hydrogen Igniter Standby Operational 0206A SQN2IGN268 285 Unit 2 Train A Hydrogen Igniter Standby Operational 0205A SQN2IGN268 286 Unit 2 Train A Hydrogen Igniter Standby Operational 0234A SQN2IGN268 287 Unit 2 Train A Hydrogen Igniter Standby Operational 0214A SQN2IGN268 288 Unit 2 Train A Hydrogen Igniter Standby Operational 0208A SQN2IGN268 289 Unit 2 Train A Hydrogen Igniter Standby Operational 0259A SQN2IGN268 290 Unit 2 Train A Hydrogen Igniter Standby Operational 0250A Page A23
Sequoyah Nuclear Plant ESEP Report TABLE A1: Expedited Seismic Equipment List (ESEL) for Sequoyah Nuclear Plant (Continued)
ESEL Equipment Operating State Item Number Normal Desired ID Description State State Notes/Comments SQN2IGN268 291 Unit 2 Train A Hydrogen Igniter Standby Operational 0232A SQN2IGN268 292 Unit 2 Train A Hydrogen Igniter Standby Operational 0201A SQN2IGN268 293 Unit 2 Train A Hydrogen Igniter Standby Operational 0230A SQN2IGN268 294 Unit 2 Train A Hydrogen Igniter Standby Operational 0249A SQN2IGN268 295 Unit 2 Train A Hydrogen Igniter Standby Operational 0254A SQN2IGN268 296 Unit 2 Train A Hydrogen Igniter Standby Operational 0224A SQN2IGN268 297 Unit 2 Train A Hydrogen Igniter Standby Operational 0233A SQN2IGN268 298 Unit 2 Train A Hydrogen Igniter Standby Operational 0255A SQN2IGN268 299 Unit 2 Train A Hydrogen Igniter Standby Operational 0207A SQN2IGN268 300 Unit 2 Train A Hydrogen Igniter Standby Operational 0236A SQN2IGN268 301 Unit 2 Train A Hydrogen Igniter Standby Operational 0221A SQN2IGN268 302 Unit 2 Train A Hydrogen Igniter Standby Operational 0213A SQN2IGN268 303 Unit 2 Train A Hydrogen Igniter Standby Operational 0228A Page A24
Sequoyah Nuclear Plant ESEP Report TABLE A1: Expedited Seismic Equipment List (ESEL) for Sequoyah Nuclear Plant (Continued)
ESEL Equipment Operating State Item Number Normal Desired ID Description State State Notes/Comments SQN2IGN268 304 Unit 2 Train A Hydrogen Igniter Standby Operational 0216A SQN2IGN268 305 Unit 2 Train A Hydrogen Igniter Standby Operational 0229A SQN2IGN268 306 Unit 2 Train A Hydrogen Igniter Standby Operational 0242A SQN2IGN268 307 Unit 2 Train A Hydrogen Igniter Standby Operational 0227A SQN2IGN268 308 Unit 2 Train A Hydrogen Igniter Standby Operational 0215A SQN2IGN268 309 Unit 2 Train A Hydrogen Igniter Standby Operational 0225A SQN1IGN268 310 Unit 1 Train A Hydrogen Igniter Standby Operational 0105A SQN1IGN268 311 Unit 1 Train A Hydrogen Igniter Standby Operational 0154A SQN1IGN268 312 Unit 1 Train A Hydrogen Igniter Standby Operational 0106A SQN1IGN268 313 Unit 1 Train A Hydrogen Igniter Standby Operational 0113A SQN1IGN268 314 Unit 1 Train A Hydrogen Igniter Standby Operational 0149A SQN1IGN268 315 Unit 1 Train A Hydrogen Igniter Standby Operational 0101A SQN1IGN268 316 Unit 1 Train A Hydrogen Igniter Standby Operational 0150A Page A25
Sequoyah Nuclear Plant ESEP Report TABLE A1: Expedited Seismic Equipment List (ESEL) for Sequoyah Nuclear Plant (Continued)
ESEL Equipment Operating State Item Number Normal Desired ID Description State State Notes/Comments SQN1IGN268 317 Unit 1 Train A Hydrogen Igniter Standby Operational 0124A SQN1IGN268 318 Unit 1 Train A Hydrogen Igniter Standby Operational 0134A Page A26
Sequoyah Nuclear Plant ESEP Report ATTACHMENT B - ESEP HCLPF VALUES AND FAILURE MODES TABULATION FOR SEQUOYAH NUCLEAR PLANT Page B1
Sequoyah Nuclear Plant ESEP Report TABLE B1: ESEP HCLPF Values and Failure Modes for Sequoyah Nuclear Plant HCLPF Equipment Equipment Floor Equipment Description Building Failure Mode Capacity ID Class Elevation (g) 1PCV15 Steam Generator #1 ARV 7 ACB 734 Screen 0.50 2PCV15 Handwheel 1PCV130 Steam Generator #4 ARV 7 ACB 734 Screen 0.50 2PCV130 Handwheel 1L501 PCV112 Local Control 18 ACB 714 Functional 0.62 2L501 Station 1L502 PCV123 Local Control 18 ACB 714 Functional 0.62 2L502 Station 1PMP3142 TDAFW Pump 5 ACB 669 Functional 1.01 2PMP3142 1FCV151 TDAFW Pump Trip and 8A ACB 669 Screen 0.50 2FCV151 Throttle Valve 1FCV152 TDAFW Pump Governor 7 ACB 669 Functional 1.01 2FCV152 Valve 1XS4657 AFWT AS Backup Control 14 ACB 669 Functional 0.88 2XS4657 Transfer Service Water 1HS151B TDAFW Pump Trip and 20 ACB 669 Functional 0.88 2HS151B Throttle Valve HS 1L381 TDAFW Pump Control 20 ACB 669 Functional 0.85 2L381 Panel 1L215 AFW Flow Monitoring 18 ACB 669 Functional 0.62 2L215 Panel 1L341 AFW Flow to Steam 18 ACB 714 Functional 0.62 2L341 Generator #3 FT Rack 1L217 AFW Flow to Steam 18 ACB 714 Functional 0.62 2L217 Generator #2 FT Rack 1L216 AFW Flow to Steam 18 ACB 690 Functional 0.62 2L216 Generator #1 FT Rack AFW Flow to Steam 1L703 Generator #4 Flow 18 ACB 690 Functional 0.62 2L703 Indication Rack 1L11A Steam Generator Level 20 ACB 734 Functional 0.75 2L11A Control Panel 1L11B Steam Generator Level 20 ACB 734 Functional 0.75 2L11B Control Panel Page B2
Sequoyah Nuclear Plant ESEP Report TABLE B1: ESEP HCLPF Values and Failure Modes for Sequoyah Nuclear Plant (Continued)
HCLPF Equipment Equipment Floor Equipment Description Building Failure Mode Capacity ID Class Elevation (g) 1L183 Steam Generator Level 18 RB 697 Functional 0.62 2L183 Transmitter Rack 1L706 Steam Generator Level 18 RB 697 Functional 0.62 2L182 Transmitter Rack 1L704 Steam Generator Level 18 RB 697 Functional 0.62 2L704 Transmitter Rack 1L185 Steam Generator Level 18 RB 697 Functional 0.62 2L185 Transmitter Rack Steam Generator 1L194 Discharge Pressure 18 ACB 690 Functional 0.62 2L194 Transmitter Rack Steam Generator 1L196 Discharge Pressure 18 ACB 690 Functional 0.62 2L196 Transmitter Rack 1TNK002 Unit 1 Condensate 21 YARD 705 DCN 23191 >2x SSE 0229 Storage Tank 2TNK002 Unit 2 Condensate 21 YARD 705 DCN 23191 >2x SSE 0232 Storage Tank 1FCV63118 Cold Leg Accumulator #1 8a RB 693 Screen 0.50 2FCV63118 Isolation Valve 1FCV6398 Cold Leg Accumulator #2 8a RB 693 Screen 0.50 2FCV6398 Isolation Valve 1FCV6380 Cold Leg Accumulator #3 8a RB 693 Screen 0.50 2FCV6380 Isolation Valve 1FCV6367 Cold Leg Accumulator #4 8a RB 693 Screen 0.50 2FCV6367 Isolation Valve 1M6 Main Control MCR Benchboard M6 20 ACB 732 0.425 2M6 Room Ceiling 1TNK062 Overturning Boric Acid Tank (BAT) A 21 ACB 690 0.78 0239 Moment 2TNK062 Overturning Boric Acid Tank (BAT) B 21 ACB 690 0.78 0239 Moment 0TNK062 Overturning Boric Acid Tank (BAT) C 21 ACB 690 0.78 0243 Moment 1M9 Main Control MCR Vertical Panel M9 20 ACB 732 0.425 2M9 Room Ceiling Page B3
Sequoyah Nuclear Plant ESEP Report TABLE B1: ESEP HCLPF Values and Failure Modes for Sequoyah Nuclear Plant (Continued)
HCLPF Equipment Equipment Floor Equipment Description Building Failure Mode Capacity ID Class Elevation (g)
SQN1IGN 268 Hydrogen Igniters 0 RB SCV Screen 0.50 (MANY) 1XFA268 PHMS Xfrm 1A 4 ACB 759 Functional 0.36 1AA 2XFA268 PHMS Xfrm 2A 4 ACB 759 Functional 0.36 2AA 1PNL268 120V AC PHMS 14 ACB 759 Functional 0.36 YA Distribution Panel 1A 120V AC PHMS 2PNL268YC 14 ACB 759 Functional 0.36 Distribution Panel 2A 1AA 7Day Oil Supply TNK01838 21 DGB 719 Screen 0.50 Tank 1BB 7Day Oil Supply TNK01840 21 DGB 719 Screen 0.50 Tank 2AA 7Day Oil Supply TNK01839 21 DGB 719 Screen 0.50 Tank 2BB 7Day Oil Supply TNK01841 21 DGB 719 Screen 0.50 Tank 1XE925001 N31 Neutron Detector 18 RB 697 Screen 0.50 1XM92 N31 Neutron Source 20 ACB 734 Screen 0.50 5001A Range Amplifier 1XM92 N31 Neutron Source 20 ACB 734 Screen 0.50 5001B Range Optical Isolation 2XE925002 N32 Neutron Detector 18 RB 697 Screen 0.50 2XM92 N32 Neutron Source 20 ACB 714 Screen 0.50 5002A Range Amplifier Page B4
Sequoyah Nuclear Plant ESEP Report TABLE B1: ESEP HCLPF Values and Failure Modes for Sequoyah Nuclear Plant (Continued)
HCLPF Equipment Equipment Floor Equipment Description Building Failure Mode Capacity ID Class Elevation (g) 2XM92 N32 Neutron Source 20 ACB 714 Screen 0.50 5002B Range Optical Isolation 1L10 Remote Control Panel 20 ACB 734 Functional 0.425 2L10 L10 1M4 Main Control MCR Benchboard M4 20 ACB 732 0.425 2M4 Room Ceiling 1M13 Main Control MCR Vertical Panel M13 20 ACB 732 0.425 2M13 Room Ceiling 1L388 RCS Loop WR PT Loop 1 28 ACB 690 Functional 0.62 2L388 Instrument Rack 1L340 RCS Loop WR PT Loop 3 28 ACB 690 Functional 0.62 2L340 Instrument Rack 1R4 AIR Panel R4 20 ACB 685 Functional 0.64 2R4 1R5 AIR Panel R5 20 ACB 685 Functional 0.64 2R5 Cold Leg WR 1TE6818 Temperature Element 19 RB 693 Screen 0.50 2TE6818 Loop 1 Cold Leg WR 1TE6841 Temperature Element 19 RB 693 Screen 0.50 2TE6841 Loop 2 Cold Leg WR 1TE6860 Temperature Element 19 RB 693 Screen 0.50 2TE6860 Loop 3 Cold Leg WR 1TE6883 Temperature Element 19 RB 693 Screen 0.50 2TE6883 Loop 4 1TE681 Hot Leg WR Temperature 19 RB 679 Screen 0.50 2TE681 Element Loop 1 1TE6824 Hot Leg WR Temperature 19 RB 679 Screen 0.50 2TE6824 Element Loop 2 1TE6843 Hot Leg WR Temperature 19 RB 679 Screen 0.50 2TE6843 Element Loop3 1TE6865 Hot Leg WR Temperature 19 RB 679 Screen 0.50 2TE6865 Element Loop 4 1M5 Main Control MCR Benchboard M5 20 ACB 732 0.425 2M5 Room Ceiling Page B5
Sequoyah Nuclear Plant ESEP Report TABLE B1: ESEP HCLPF Values and Failure Modes for Sequoyah Nuclear Plant (Continued)
HCLPF Equipment Equipment Floor Equipment Description Building Failure Mode Capacity ID Class Elevation (g) 1R2 AIR Panel R2 20 ACB 685 Functional 0.64 2R2 1R6 AIR Panel R6 20 ACB 685 Functional 0.64 2R6 RCS Pressurizer Level 1L179 Transmitter Instrument 18 RB Functional 0.62 2L179 Rack 2XM92 N32 Neutron Source 20 ACB 714 Screen 0.50 5002B Range Optical Iso.
Main Control 1M2 Cabinet M2 20 ACB 732 0.425 Room Ceiling 1FCV3 136A ERCW Header A AFW 8a ACB 669 Screen 0.50 2FCV3 Supply Valve 136A 1FCV3136B ERCW Header A AFW 8a ACB 669 Screen 0.50 2FCV3136B Supply Valve 1FCV3 179A ERCW Header B AFW 8a ACB 669 Screen 0.50 2FCV3 Supply Valve 179A 1FCV3179B ERCW Header B AFW 8a ACB 669 Screen 0.50 2FCV3179B Supply Valve Page B6
ENCLOSURE 2 LIST OF COMMITMENTS
- 1. Perform seismic walkdowns, generate HCLPF calculations, and design and implement any necessary modification for the following Unit 1 inaccessible items no later than the end of the second planned Unit 1 refueling outage after December 31, 2014:
- a. FCV 63-118 - Cold Leg Accumulator Isolation Valve #1
- b. FCV 63-067 - Cold Leg Accumulator Isolation Valve #4
- c. FCV 63-080 - Cold Leg Accumulator Isolation Valve #3
- d. FCV 63-098 - Cold Leg Accumulator Isolation Valve #2
- e. Instrument Rack 1 - L - 182 located in Fan Room 2
- f. Instrument Rack 1 - L - 183 located in Fan Room 1
- g. Instrument Rack 1 - L - 179
- h. Instrument Rack 1 - L - 185
- i. Instrument Rack 1 - L - 704
- j. Instrument Rack 1 - L - 706
- k. Instrument Rack 1 - L - 194
- l. Instrument Rack 1 - L - 196
- m. Instrument Rack 1 - L - 216
- 2. Modify TDAFP Control Panel 2-L-381 anchorage to replace the corroded anchor such that HCLPH > RLGM no later than the end of the U2 Cycle 20 refueling outage.
- 3. Submit a letter to NRC summarizing the HCLPF results of Commitment 1 and confirming implementation of the plant modifications associated with Commitment 2 within 60 days following completion of ESEP activities.
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