ML14353A333: Difference between revisions

From kanterella
Jump to navigation Jump to search
(Created page by program invented by StriderTol)
(Created page by program invented by StriderTol)
 
(5 intermediate revisions by the same user not shown)
Line 1: Line 1:
{{Adams
#REDIRECT [[RS-14-300, Expedited Seismic Evaluation Process Report (CEUS Sites), Response to Nrg Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1..]]
| number = ML14353A333
| issue date = 12/19/2014
| title = Peach Bottom Atomic Power Station, Units 2 and 3 - Exelon Generation Company, LLC Expedited Seismic Evaluation Process Report (CEUS Sites), Response to Nrg Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1..
| author name = Barstow J
| author affiliation = Exelon Generation Co, LLC
| addressee name =
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000277, 05000278
| license number = DPR-044, DPR-056
| contact person =
| case reference number = RS-14-300
| document report number = 14Q4233-RPT-004, Rev 3
| document type = Letter, Report, Technical
| page count = 68
| project =
| stage = RAI
}}
 
=Text=
{{#Wiki_filter:Exelon Generation RS-14-300 December 19, 2014 U.S. Nuclear Regulatory Commission Attn: Document Control Desk 11555 Rockville Pike, Rockville.
MD 20852 Peach Bottom Atomic Power Station, Units 2 and 3 10 CFR 50.54(f) Renewed Facility Operating License Nos. DPR-44 and DPR-56 NRG Docket Nos. 50-277 and 50-278
 
==Subject:==
Exelon Generation Company, LLC Expedited Seismic Evaluation Process Report (CEUS Sites), Response to NRG Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident
 
==References:==
: 1. NRG Letter, Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(f) Regarding Recommendations 2.1, 2.3, and 9.3, of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident, dated March 12, 2012 (ML 12053A340)
: 2. NEI Letter, Proposed Path Forward for NTTF Recommendation 2.1: Seismic evaluations, dated April 9, 2013 (ML 13101 A379) 3. Seismic Evaluation Guidance: "Augmented Approach for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1 -Seismic'', EPRI, Palo Alto, CA: May 2013.3002000704(ML13102A142)
: 4. NRG Letter, Electric Power Research Institute Report 3002000704, "Seismic Evaluation Guidance:
Augmented Approach for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1: Seismic," as an Acceptable Alternative to the March 12, 2012, Information Request for Seismic Re-evaluations, dated May 7, 2013 (ML 13106A331)
: 5. Exelon Generation Company, LLC, Seismic Hazard and Screening Report (Central and Eastern United States (CEUS) Sites), Response to NRG Request for Information Pursuant to 1 O CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident (RS-14-071), dated March 31, 2014 (ML14090A247) . 6. Exelon Generation Company, LLC Response to NRG Request for Information Pursuant to 1 O CFR 50.54(f) Regarding the Seismic Aspects of Recommendation 2.1 of the Term Task Force Review of Insights from the Fukushima Dai-ichi Accident -1.5 Year Response for CEUS Sites (RS-13-205), dated September 12, 2013 (ML 13256A070)
U.S. Nuclear Regulatory Commission NTTF 2.1 Seismic Response for CEUS Sites December 19, 2014 Page2 On March 12, 2012, the Nuclear Regulatory Commission (NRG) issued a 50.54(f) letter to all power reactor licensees and holders of construction permits in active or deferred status. Enclosure 1 of Reference 1 requested each addressee located in the Central and Eastern United States (CEUS) to submit a Seismic Hazard Evaluation and Screening Report within 1.5 years from the date of Reference 1 . In Reference 2, the Nuclear Energy Institute (NEI) requested NRG agreement to delay submittal of the final CEUS Seismic Hazard Evaluation and Screening Reports so that an update to the Electric Power Research Institute (EPRI) ground motion attenuation model could be completed and used to develop that information.
NEI proposed that descriptions of subsurface materials and properties and base case velocity profiles be submitted to the NRG by September 12, 2013, (Reference 6), with the remaining seismic hazard and screening information submitted by March 31, 2014 (Reference 5). NRG agreed with that proposed path forward in Reference
: 4. Reference 1 requested that licensees provide interim evaluations and actions taken or planned to address the higher seismic hazard relative to the design basis, as appropriate, prior to completion of the risk evaluation.
In accordance with the NRG endorsed guidance in Reference 3, Enclosure 1 provides the Expedited Seismic Evaluation Process (ESEP) Report for Peach Bottom Atomic Power Station, Units 2 and 3, and the information described in the "ESEP Report" Section 7, of Reference 3 in accordance with the schedule identified in Reference
: 2. With the exception of the two (2) plant relays identified in Enclosure 1 , all other equipment evaluated for the ESEP for Peach Bottom Atomic Power Station Units 2 and 3 was found to have adequate capacity for the required seismic demand as defined by the Augmented Approach (ESEP) guidance (Reference 3). Further evaluation, and implementation of modifications, if required, to increase seismic margin for the affected plant relays, will be completed as identified in Enclosure
: 2. This ESEP report transmittal completes regulatory Commitment No. 3 of Reference
: 5. A list of new regulatory commitments contained in this letter is provided in Enclosure
: 2. If you have any questions regarding this report, please contact Ron Gaston at (630) 657-3359.
I declare under penalty of perjury that the foregoing is true and correct. Executed on the 19 1 h day of December 2014. Respectfully submitted, James Barstow Director -Licensing
& Regulatory Affairs Exelon Generation Company, LLC U.S. Nuclear Fle9ulatory Commission NTT F 2. 1 Seisrnic F{<::sponse for CEUS Sites DecembCJr 19, 2014 3
 
==Enclosures:==
: 1. Peach Bottorn Atornic Power Station, Units 2 and 3 Expedited Seismic Evaluation Process (ESEP) Report 2. Surnmary of Commitments cc: Director, Office of Nuclear Reactor Regulation Regional Adn1inistrator
-NRC Region I NRC Senior Resident Inspector
-Peach Bottom Atomic Power Station NRC Project Manager, NRR -Peach Bottom Atomic Power Station Mr. Nicholas J. DiFrancesco, NRR/JLD/JHMB, NRC Director, Bureau of Radiation Protection
-Pennsylvania Department of Environmental Resources S. T. Gray, State of Maryland R. R. Jana ti, Chief, Division of Nuclear Safety, Pennsylvania Department of Environmental Protection, Bureau of Radiation Protection Enclosure 1 Peach Bottom Atomic Power Station, Units 2 and 3 Expedited Seismic Evaluation Process (ESEP) Report (63 pages)
EXPEDITED SEISMIC EV A LUATION PROCESS (ESEP) RE PORT IN R ESPONSE TO THE 50.54(f) INFO RMA TION REQUEST REGARDING FUKUSHI MA NEA R-TERM TAS K FO CE RECOM M ENDATION 2.1: SBS MI C for the PHch Bottom Atomk Pow* Stlltlon I/nits 2 a I 1848 uy ROM/ /hll tll, Pminsylnnla l1J1.4-9012 F*clllly Optlt'atlng Llc*n* No. DPR-.U, DPR*S6 NRC Docket No. 50*277, 50*278 --**---*****----***---****------*----**-***-__________ Co rr.npondence
..No.i.RS.-, U..300-----**-----***--*--
---------*-------**--*-*** -----***-**--** ***---***** *-*------* --*** *-***** *-*fxetorr-G-eneratlon Company,*tte{EXetc>nr-
**-**--***-****-****-*-****----**----****-*--***-* PO Box 805398 Chicago, IL 60680-5398 Prepared by: Stevenson
& Associates 275 Mishawum Road, Suite 200 Woburn, MA 01801 -------******* ---****-*-**--------*-----*---*-* *-------ReportNum6er.i4Q4233--RPr=004.
-Rev.f* *-----------* ------------*****--------** Printed Name Siglllture J2ile. Preparer:
Douglas Seymour f!)/l'&sect;r 12/17 /2014 Reviewer:
A. Karavoussianis -
12/17 /2014 Approver:
A. Karavousslanis U/17/2014 Lead Respon sib le Tracey Gallagher Branch Manager: Frank Glaco Senior Manager Design Engineering:
__ o __ a __
Corporate Acceptance:
Jeffrey Clari<
Document Title:  EXPEDITED SEISMIC EVALUATION PROCESS (ESEP) REPORT IN RESPONSE TO THE 50.54(f) INFORMATION REQUEST REGARDING FUKUSHIMA NEAR-TERM TASK FORCE RECOMMENDATION 2.1:  SEISMIC for the Peach Bottom Atomic Power Station Units 2 and 3  Document Type: Report Report Number: 14Q4233-RPT-004 Project Name:
Exelon ESEP for Peach Bottom Job No.: 14Q4233 Client:
This document has been prepared in accordance with the S&A Quality Assurance Program Manual, Revision _17_ and project requirements:
Rev. 0 Prepared by:
Douglas Seymour Date: 12/5/2014 Reviewed by:
A. Karavoussianis Date: 12/5/2014 Approved by:  A. Karavoussianis Date: 12/5/2014 Revision Record: Revision No. Prepared by/
Date Reviewed by/
Date Approved by/
Date Description of Revision 1    D. Seymour 12/10/2014
 
A. Karavoussianis 12/10/2014 A. Karavoussianis 12/10/2014 Incorporation of client editorial
 
comment and update of Reference 4.
2 D. Seymour 12/15/2014
 
A. Karavoussianis 12/16/2014 A. Karavoussianis 12/16/2014 Incorporation of minor editorial comments.
3    D. Seymour 12/17/2014
 
A. Karavoussianis 12/17/2014 A. Karavoussianis 12/17/2014 Incorporation of minor editorial comments.
Stevenson & Associates DOCUMENT APPROVAL SHEET CONTRACT NO. 14Q4233 14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 3 of 63 Table of Contents 1 Purpose and Objective ................................................................................................... 6 2 Brief Summary of the FLEX Seismic Implementation Strategies
............................... 7 3 Equipment Selection Process and ESEL ...................................................................... 10  3.1 Equipment Selection Process and ESEL and Alternate Path Justifications ........................
10  3.1.1 ESEL Development
..............................................................................................................................
11  3.1.2 Power Operated Valves ..................................................................................................
..................
13  3.1.3 Pull Boxes .............................................................................................................
..................................
13  3.1.4 Termination Cabinets ...................................................................................................
.....................
13  3.1.5 Critical Instrumentation Indicators ....................................................................................
...........
13  3.1.6 Phase 2 and Phase 3 Piping Connections
..................................................................................
14  3.2 Justification for use of Equipment that is not the Primary Means for FLEX  Implementation ..............................................................................................................
.........................
14 4 Ground Motion Response Spectrum (GMRS) ............................................................ 15  4.1 Plot of GMRS Submitted by the Licensee ...................................................................................
.. 15  4.2 Comparison to SSE ........................................................................................................
.........................
16 5 Review Level Ground Motion (RLGM) ....................................................................... 17  5.1 Description of RLGM Selected .............................................................................................
..............
17  5.2 Method to Estimate ISRS ..................................................................................................
...................
19 6 Seismic Margin Evaluation Approach ........................................................................ 20  6.1 Summary of Methodologies used ............................................................................................
........ 20  6.2 HCLPF Screening Process ..................................................................................................
..................
20  6.3 Seismic Walkdown Approach ................................................................................................
............
21  6.3.1 Walkdown Approach ......................................................................................................
...................
21  6.3.2 Application of Previo us Walkdown Information .....................................................................
23  6.3.3 Significant Walkdown Observations ......................................................................................
...... 23  6.4 HCLPF Calculation Process ................................................................................................
..................
23  6.5 Functional Evaluation of Relays ..........................................................................................
...............
24  6.6 Tabulated ESEL HCLPF Values (including key failure modes) ................................................
24 7 Inaccessible Items
......................................................................................................... 26  7.1 Identification of ESEL Items Inaccessible for Walkdowns ........................................................
26  7.2 Planned Walkdown / Evaluation Schedule / Close Out ...........................................................
27 8 ESEP Conclusions and Results ..................................................................................... 28 14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 4 of 63 8.1 Supporting Information ...................................................................................................
....................
28  8.2 Identification of Planned Modi fications ..................................................................................
....... 30  8.3 Modification Implementation Schedule .....................................................................................
... 30  8.4 Summary of Regulatory Commitments
..........................................................................................
30 9 Refere nces ..................................................................................................................
... 31 Attachment A Peach Bottom Unit 2 and Common ESEL Attachment B Peach Bottom Unit 3 ESEL Attachment C ESEP HCLPF Values and Failure Modes Tabulation, Unit 2 and Common Attachment D ESEP HCLPF Values and Failure Modes Tabulation, Unit 3
 
14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 5 of 63 List of Tables Table 3.1-1 Flow Paths Credited for ESEP Table 4.1-1 Peach Bottom GMRS Table 4.2-1 Peach Bottom GMRS vs. SSE Table 5.1-1 Ratio between GMRS and SSE Table 5.1-2 Peach Bottom RLGM Table 7.1-1 Items Inaccessible for Walkdowns List of Figures Figure 4.1-1 Peach Bottom GMRS Figure 4.2-1 Peach Bottom GMRS vs. SSE Figure 5.1-1 Peach Bottom RLGM
 
14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 6 of 63 1 PURPOSE AND OBJECTIVE Following the accident at the Fukushima Dai-ichi nuclear power plant resulting from the March 11, 2011, Great Tohoku Earthquake and subsequent tsunami, the Nuclear Regulatory Commission (NRC) established a Near Term Task Force (NTTF) to conduct a systematic review of NRC processes and regulations and to determine if the agency should make additional improvements to its regulatory system. Th e NTTF developed a set of recommendations intended to clarify and strengthen the regulatory framework for protection against natural phenomena. Subsequently, the NRC issued a 50.54(f) letter on March 12, 2012 (Ref. 1) requesting information to assure that these recommendations are addressed by all U.S. nuclear power plants. The 50.54(f) letter requests that licensees and holders of construction permits under 10 CFR Part 50 reevaluate the seismic hazards at their sites against present-day NRC requirements and guidance. Depending on the comparison between the reevaluated seismic hazard and the current design basis, further risk assessment may be required. Assessment approaches acceptable to the staff include a seismic probabilistic risk assessment (SPRA), or a seismic margin assessment (SMA). Based upon the assessment results, the NRC staff will determine whether additional regulatory actions are necessary. This report describes the Expedited Seismic Evaluation Process (ESEP) undertaken for Peach Bottom Atomic Power Station (PBAPS). The intent of the ESEP is to perform an interim action in response to the NRC's 50.54(f) letter (Ref. 1) to demonstrate seismic margin through a review of a subset of the plant equipment that can be reli ed upon to protect the reactor core following beyond design basis seismic events. The ESEP is implemented using the methodologies in the NRC endorsed guidance in EPRI 3002000704, Seismic Evaluation Guidance: Augmented Approach for the Resolution of Fukushima Near-Term Task Force Recomme ndation 2.1: Seismic (Ref. 2). The objective of this report is to provide summary information describing the ESEP evaluations and results. The level of detail provided in the report is intended to enable NRC to understand the inputs used, the evaluations performed, and the decisions made as a result of the interim evaluations.
 
14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 7 of 63 2 BRIEF
 
==SUMMARY==
OF THE FLEX SEIS MIC IMPLEMENTATION STRATEGIES
* The Peach Bottom FLEX response strategies for Reactor Core Cooling and Heat Removal, Reactor Inventory Control, Core Cooling and Heat Removal (Modes 3 and 4), Containment Function and Spent Fuel Pool Control are summarized below. This summary is derived from the Overall Integrated Plan (OIP) in Response to the March 12, 2012, Commission Order EA-12-049 (Ref. 3) including the required 6-Month Updates that have been prepared since the OIP was submitted. Flex Phase 1, strategy relies on installed plant equipment. Reactor core cooling and heat removal is achieved via steam release from the Safety Relief Valves (SRV's) to the Torus and Reactor Core Isolation Cooling (RCIC) drive steam. Coolant ma keup is established and maintained from RCIC. Preferred suction for RCIC will be from the condensate storage tank (CST) if it is still viable. If not, suction will be from the Torus. The cool down rate of the reactor coolant system is controlled through manual operation of the SRV's and RCIC at a targeted 80 degrees per hour. The cool down will facilitate decay heat removal while keeping the Torus water temperature within limits and reactor pressure high enough to maintain RCIC operation. RCIC operation is DC and independent of emergency AC power. It can also be operated manually without power if necessary. SRV's operate mechanically at specific high set pressures and manually with DC power to solenoids with nitrogen accumulators to enable operation to establish a cool down below set pressures. Key reactor parameters are obtained via DC powered instrumentation. A DC load reduction strategy is employed to extend battery life. The Spent Fuel Pool makeup may be required if a full core off load was recently completed as time to boil in the pool is at its most limiting. Spent Fuel level is obtained from the new spent fuel pool wide range instrumentation installed under order EA-12-051. Flex Phase 2, strategy relies on installed plant equipment and portable equipment. If RCIC is running without challenging limits, it will cont inue in operation to control reactor level and pressure. It is recognized that area temperatures will increase. Opening plant doors and fans powered by portable diesel generators will be used to address the RCIC room environmental issues. Work to line-up and start operation of phase 2 equipment will commence no later than 1 hour after the event occurs. A portable 500 KW 480VAC diesel generator is used to provide power to 480 VAC Load Centers and Motor Control Centers. This will allow energizing the Division 1 battery chargers, battery room ventilation, Control Room Emergency Ventilation (CREV), diesel fuel oil transfer pump, and valves re quired to inject water from external sources to the reactor via the Residual Heat Removal (RHR) system. Portable diesel driven pumps will provide injection from two suction sources, either the Emergency Cooling Tower (ECT) or the Conowingo Pond (Susquehanna River) at the screen house. The discharge will be either into the RHR system directly or into the High Pressure Service Water system (HPSW) emergency cross-tie to the RHR system. RHR will then supply water to the Reactor, Torus and the Fuel Pool Cooling system. The Refuel Floor roof hatch and the Reac tor Building truck bay doors will need to be opened as a method to maintain temperature on the refuel floor.
* This section is based upon input received from Peach Bottom Atomic Power Station in (Refs. 4, 25, 26, and 27).
14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 8 of 63 Flex Phase 1 and 2 strategy will provide sufficient capability such that no additional Phase 3 strategies are required. Phase 3 relies on installed plant equipment and portable equipment as described in Phase 1 and Phase 2.
Peach Bottom Phase 1 Flex Strategy Safety Function Primary Method Alternate Method Core Cooling Reactor Core Cooling & Heat Removal
* RCIC pump with suction from Torus
* RCS cool down at 80
&deg;F/hour with SRV's
* RCIC pump with suction from CST
* Backup nitrogen to SRV
 
solenoids RCS Inventory Control & Heat Removal,
* RCIC pump with suction from Torus
* RCS cool down at 80
&deg;F/hour with SRV's
* RCIC pump with suction from CST
* Backup nitrogen to SRV
 
solenoids Key Reactor Parameters
* Temperature, pressure, level
* Use existing ba ttery powered indication
* Extend coping with deep DC load stripping Containment Containment Pressure Control
 
& Remove Heat
* None required - pressure and temperature below limits
* Containment venting if Torus temperature reaches 230
&deg;F Key Containment Parameters
* Temperature and pressure indication powered via vital buses
* Temperature and pressure indication powered via vital buses SFP Cooling Spent Fuel Cooling
* None required - temperature below limits
* None required - temperature below limits SFP Parameters
* SFP Wide Range Level Indicator
* SFP Wide Range Level Indicator
 
14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 9 of 63 Peach Bottom Phase 2 Flex Strategy Safety Function Primary Method Alternate Method Core Cooling Reactor Core Cooling & Heat Removal (Mode 3 and 4)
* Diesel driven portable pumps
* 500KW 480VAC FLEX generator to operate valves
* Diesel driven portable pumps at alternate suction and discharge locations
* 500KW 480VAC FLEX generator to operate valves at alternate location RCS Inventory Control & Heat
 
Removal (Mode 3 and 4)
* Diesel driven portable pumps
* 500KW 480VAC FLEX generator to operate valve
* Diesel driven pump at alternate suction and discharge locations
* 500KW 480VAC FLEX generator to operate valves at alternate location Key Reactor Parameters
* 500KW 480VAC FLEX generator repower one Vital Load center to repower battery charger
* Local indication determination at rack or penetration Containment Containment Pressure Control
& Remove Heat
* None required - pressure and temperature below limits
* None required - pressure and temperature below limits Key Containment Parameters
* Temperature and pressure indication powered via vital buses
* Local indication determination at rack or penetration SFP Cooling Spent Fuel Cooling
* RHR to Fuel Pool cooling crosstie
* Fire hose spray on the spent fuel SFP Parameters
* SFP Wide Range Level Indicator
* SFP Wide Range Level Indicator
 
14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 10 of 63 3 EQUIPMENT SELECTION PROCESS AND ESEL AND ALTERNATE PATH JUSTIFICATIONS The selection of equipment for the ESEL followed the guidelines of EPRI 3002000704 (Ref. 2). The ESELs per Ref. 22 for Unit 2 and Unit 3 are presented in Attachments A and B, respectively.
3.1 Equipment Selection Process and ESEL The selection of equipment to be included on the Expedited Seismic Equipment List (ESEL) was based on installed plant equipment credited in the FLEX strategies during mitigation of an Extended Loss of AC Power (ELAP), as outlined in the Peach Bottom Overall Integrated Plan (OIP) in Response to the March 12, 2012, Commission Order EA-12-049 (Ref. 3) including subsequent 6 month updates through August 2014
: f. The OIP provides the Peach Bottom FLEX 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 following a beyond design basis seismic event, consistent with the Peach Bottom OIP (Ref. 3) including subsequent 6 month updates through August 2014. FLEX recove ry actions are excluded from the ESEP scope per EPRI 3002000704 (Ref. 2). The overall list of planned FLEX modifications and the scope for consideration herein is limited to those required to support core cooling, reactor coolant inventory and sub-criticality, and containment integrity functions. Portable and pre-staged FLEX equipment (not permanently installed) are excluded from the ESEL per EPRI 3002000704 (Ref. 2). The ESEL component selection followed the EPRI guidance outlined in Section 3.2 of EPRI 3002000704 (Ref. 2). 
: 1. The scope of components is limited to that required to accomplish the core cooling and containment safety functions identified in Table 3-1 of EPRI 3002000704. The instrumentation monitoring requirements for core cooling/containment safety functions are limited to those outlined in the EPRI 3002000704 guidance, and are a subset of those outlined in the Peach Bottom OIP (Ref.3) including subsequent 6 month updates through August 2014. 
: 2. The scope of components is limited to installed plant equipment, and FLEX connections necessary to implement the Peach Bottom OIP (Ref. 3) including subsequent 6 month updates through August 2014 as described in Section 2.
: 3. The scope of components assumes the credited FLEX connection modifications are implemented, and are limited to those required to support a single FLEX success path (i.e., either "Primary" or "Back-up/Alternate"). 
: 4. The "Primary" FLEX success path is to be specified. Selection of the "Back-up/Alternate" FLEX success path must be justified.
f References 25 and 26 confirm that there are no changes between the February 2013 and August 2014 Flex Strategies submittals. This footnote applies to all references of the August 2014 Flex Strategies throughout this document.
14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 11 of 63
: 5. Structures, systems, and components excluded per the EPRI 3002000704 (Ref. 2) guidance are:
* Structures (e.g. containment, reactor building, control building, auxiliary building, etc.)
* Piping, cabling, conduit, HVAC, and their supports.
* Manual valves, check valves and rupture disks.
* Power-operated valves not required to change state as part of the FLEX mitigation strategies.
* Nuclear steam supply system components (e.g. reactor pressure vessel and internals, reactor coolant pumps and seals, etc.)
: 6. For cases in which neither train was specified as a primary or back-up strategy, then only one train component (generally 'A' train) is included in the ESEL.
3.1.1 ESEL Development The ESEL was developed by reviewing the Peach Bottom OIP (Ref. 3) including subsequent 6 month updates through August 2014 to determine equipment involved in the FLEX strategies. Further reviews of plant drawings (e.g., Process and Instrumentation Diagrams (P&IDs) and Electrical One Line Diagrams) were performed to identify the boundaries of the flow paths to be used in the FLEX strategies and to identify sp ecific components in the flow paths needed to support implementation of the FLEX strategies.
Boundaries were established at an electrical power distribution or mechanical isolation device in branch circuits / branch lines off the defined strategy electrical or fluid flow path. P&IDs were the primary reference documents used to identify mechanical components and instrumentation. The flow paths used for FLEX strategies were selected and specific components were identified using detailed equipment and instrument drawings, piping isometrics, electrical schematics and one-line drawings, system descriptions, design basis documents, etc., as necessary. The flow paths credited for ESEP for Peach Bottom per Ref. 22 are shown in Table 3.1-1.
 
14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 12 of 63 Table 3.1-1 Flow Paths Credited for ESEP Flow Path P&IDs Unit 2 Unit 3 Core Heat  Removal using RCIC system: Coolant from the Suppression Pool to the RCS via the RCIC pump and Reactor Recirculation Pump Discharge Piping. Main Steam providing motive force to the RCIC pump turbine and exhausted to the Suppression Pool. M-359 sh. 1 (Ref. 21.h)M-360 sh. 1 (Ref. 21.j) M-360 sh. 3 (Ref. 21.l) M-359 sh. 1 (Ref. 21.h)M-359 sh. 2 (Ref. 21.i)
M-360 sh. 1 (Ref. 21.j) M-360 sh. 2 (Ref. 21.k) M-360 sh. 4 (Ref. 21.m) RPV Pressure Control using ADS system: Main Steam relieved through the ADS Safety/Relief Valves to the Suppression Pool. M-351 sh. 1 (Ref. 21.d)M-351 sh. 2 (Ref. 21.e) M-372 sh. 1 (Ref. 21.t) M-333 sh. 1 (Ref. 21.b) M-351 sh. 3 (Ref. 21.f)M-351 sh. 4 (Ref. 21.g)M-372 sh. 1 (Ref. 21.t) M-333 sh. 2 (Ref. 21.c) RPV Make Up: Coolant from the Ultimate Heat Sink to the Suppression Pool via the FLEX pump and the RHR system. M-361 sh. 1 (Ref. 21.n)M-361 sh. 2 (Ref. 21.o) M-361 sh. 3 (Ref. 21.p)M-361 sh. 4 (Ref. 21.q) Hardened Containment Vent: Torus vented to atmosphere. M-367 sh. 1 (Ref. 21.r) M-367 sh. 2 (Ref. 21.s) Essential Service Water: Coolant from the Ultimate Heat Sink to the RHR system via a FLEX pump connection. M-315 sh. 1 (Ref. 21.a) M-315 sh. 1 (Ref. 21.a) Main Control Room Ventilation: Outside air through the Control Room Emergency Ventilation filters to a Control Room Emergency Ventilation Fan. Fan discharge into the Control Room. M-384 sh. 1 (Ref. 21.u)M-384 sh. 2 (Ref. 21.v) M-384 sh. 3 (Ref. 21.w) M-384 sh. 1 (Ref. 21.u)M-384 sh. 2 (Ref. 21.v) M-384 sh. 3 (Ref. 21.w) Battery Room Ventilation: Outside Air to the Emergency Switch Gear & Battery Room Vent. Fan supplied to both Units' Battery Rooms. Air from Battery Rooms exhausted through a Battery Room Exhaust Fan to atmosphere. M-399 sh. 1 (Ref. 21.x) M-399 sh. 2 (Ref. 21.y) M-399 sh. 3 (Ref. 21.z) M-399 sh. 4 (Ref. 21.aa) M-399 sh. 1 (Ref. 21.x) M-399 sh. 2 (Ref. 21.y) M-399 sh. 3 (Ref. 21.z) M-399 sh. 4 (Ref. 21.aa)
 
14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 13 of 63 3.1.2 Power Operated Valves Page 3-3 of EPRI 3002000704 (Ref. 2) notes that power operated valves not required to change state are excluded from the ESEL. Page 3-2 also notes that "functional fail ure modes of electrical and mechanical portions of the installed Phase 1 equipment should be considered (e.g. RCIC trips)."  To address this concern, the following guidance is applied in the Peach Bottom ESEL for functional failure modes associated with power operated valves:
* Power operated valves that must remain energized during the Extended Loss of all AC Power (ELAP) events in order to maintain a credited FLEX flow path or pressure boundary (such as DC powered solenoid-operated va lves), were included on the ESEL.
* Power operated valves not required to change state as part of the FLEX mitigation strategies were not included on the ESEL. The seismic event also causes the ELAP event; therefore, the valves are incapable of spurious operation as they would be de-energized.
* Power operated valves not required to change state as part of the FLEX mitigation strategies during Phase 1, and are re-energized and operated during subsequent Phase 2 and 3 strategies, were not evaluated for spurious valve operation as the seismic event that caused the ELAP has passed before the valves are re-powered.
3.1.3 Pull Boxes Pull boxes were deemed unnecessary to add to the ESELs as these components provide completely passive locations for pulling or installing cables. No breaks or connections in the cabling are included in pull boxes. Pull boxes were considered part of conduit and cabling, which are excluded in accord ance with EPRI 3002000704 (Ref. 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, and the connections are excluded from the ESEL.
3.1.5 Critical Instrumentation Indicators Critical indicators and recorders are typically physically located on panels/cabinets and are included as separate components; however, seismic evaluation of the instrument indication may be included in the panel/cabinet seismic evaluation (rule-of-the-box).
14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 14 of 63 3.1.6 Phase 2 and Phase 3 Piping Connections Item 2 in Section 3.1 above notes that the scope of equipment in the ESEL includes "- FLEX connections necessary to implement the Peach Bottom OIP (Ref. 3) including subsequent 6 month updates through August 2014 as described in Section 2."  Item 3 in Section 3.1 also notes that "The scope of components assumes the credited FLEX connection modifications are implemented, and are limited to those required to support a single FLEX success path (i.e., either "Primary" or "Back-up/Alternate")." Item 5 in Section 3.1 above goes on to explain that "Piping, cabling, conduit, HVAC, and their supports" are excluded from the ESEL scope in accordance with EPRI 3002000704 (Ref. 2). Therefore, piping and pipe supports associated with FLEX Phase 2 and Phase 3 connections are excluded from the scope of the ESEP evaluation. However, any active valves in FLEX Phase 2 and Phase 3 connection flow path are included in the ESEL.
3.2 Justification for use of Equipment that is not the Primary Means for FLEX Implementation All equipment used for FLEX implementation on the PBAPS ESEL are the primary path.
 
14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 15 of 63 4 GROUND MOTION RESPONSE SPECTRUM (GMRS) 4.1 Plot of GMRS Submitted by the Licensee In accordance with Section 2.4.2 of the SPID (Ref. 14), the licensing design basis definition of the SSE control point for Peach Bottom is used for comparison to the GMRS. Ref. 6 lists the Peach Bottom SSE as being located at 136 feet MSL with a PGA of 0.12g. The GMRS per the March 2014 submittal report (Ref. 6) is tabulated and graphed below: Table 4.1-1 Peach Bottom GMRS (5% Damping) Freq. (Hz) GMRS (unscaled, g)Freq. (Hz)GMRS (unscaled, g) 0.1 0.007 40.256 0.125 0.008 50.332 0.15 0.010 60.387 0.2 0.014 70.441 0.25 0.017 80.490 0.3 0.020 90.536 0.35 0.024 100.581 0.4 0.027 12.50.659 0.5 0.034 150.727 0.6 0.041 200.844 0.7 0.047 250.924 0.8 0.052 300.967 0.9 0.057 350.961 1 0.062 400.914 1.25 0.076 500.730 1.5 0.089 600.549 2 0.124 700.461 2.5 0.143 800.426 3 0.179 900.410 3.5 0.219 1000.402  Figure 4.1-1 Peach Bottom GMRS (5% Damping) 0.00 0.20 0.40 0.60 0.80 1.00 1.20 0.1 1 10 100 Acceleration (g)
Frequency (Hz) 14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 16 of 63 4.2 Comparison to SSE As identified in the March 2014 submittal report (Ref. 6), the GMRS exceeds the SSE in the 1-10 Hz range as shown in the table and graph below: Table 4.2-1 Peach Bottom GMRS vs. SSE (5% Damping) Freq. (Hz)GMRS (g)SSE (g)1 0.062 0.11 1.25 0.076 0.13 1.5 0.089 0.16 2 0.124 0.19 2.5 0.143 0.21 3 0.179 0.22 3.5 0.219 0.22 4 0.256 0.22 5 0.332 0.21 6 0.387 0.21 7 0.441 0.20 8 0.490 0.20 9 0.536 0.19 10 0.581 0.19 Figure 4.2-1 Peach Bottom GMRS vs. SSE (5% Damping) 0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 11 0Spectral Acceleration , 5% damping (g)
Frequency (Hz)
GMRSSSE 14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 17 of 63 5 REVIEW LEVEL GROUND MOTION (RLGM) 5.1 Description of RLGM Selected The RLGM for Peach Bottom was determined in accordance with Section 4 of EPRI 30020000704 (Ref. 2) as being derived by linearly scaling the Peach Bottom SSE by the maximum ratio of the GMRS/SSE between the 1 and 10 Hertz range, with an upper bound of 2.0. The ratio between the GMRS and SSE at 5% damping is tabulated below.
 
Table 5.1-1 Ratio between GMRS and SSE (5% Damping) Freq. (Hz) GMRS (g)SSE (g)Ratio GMRS/SSE1 0.0619 0.11 0.56 1.25 0.0759 0.13 0.58 1.5 0.0893 0.16 0.56 2 0.124 0.19 0.65 2.5 0.143 0.21 0.68 3 0.179 0.22 0.81 3.5 0.219 0.22 1.00 4 0.256 0.22 1.16 5 0.332 0.21 1.58 6 0.387 0.21 1.84 7 0.441 0.2 2.21 8 0.49 0.2 2.45 9 0.536 0.19 2.82 10 0.581 0.19 3.06 The maximum ratio between the 5% damping GMRS and horizontal SSE occurs at 10 Hz and equals 3.06. Based on Section 4 of EPRI 30020000704 (Ref. 2), the RLGM derived by linearly scaling the SSE need not exceed 2 x SSE; therefore the upper bound ratio of 2.0 applies. 
 
The resulting RLGM based on increasing the horizontal SSE by the maximum ratio of 2.0 is plotted below. Note that the RLGM PGA is 0.24g.
 
14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 18 of 63 Table 5.1-2 Peach Bottom RLGM (5% Damping) Freq. (Hz) RLGM (g)  Freq. (Hz) RLGM (g) 1 0.22 12.5 0.36 1.25 0.26 15 0.36 1.5 0.32 20 0.34 2 0.38 25 0.32 2.5 0.42 30 0.30 3 0.44 35 0.30 3.5 0.44 40 0.30 4 0.44 50 0.28 5 0.42 60 0.28 6 0.42 70 0.26 7 0.40 80 0.24 8 0.40 90 0.24 9 0.38 100 0.24 10 0.38    Figure 5.1-1 Peach Bottom RLGM, GMRS & SSE (5% Damping) 0 0.2 0.4 0.6 0.8 1 1.2 1 10 100Spectral Acceleration (g)
Frequency (Hz)Original SSE RLGM GMRS 14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 19 of 63 5.2 Method to Estimate ISRS The method used to derive the ESEP in-structure response spectra (ISRS) was to scale the existing SSE-based ISRS obtained from Peach Bottom Specification 11187-G-14, "General Project Requirements for Seismic Design and Analysis of Equipment and Equipment Supports for the Peach Bottom Atomic Power Station Units 2 & 3" (Ref. 18) and PBAPS calculation PS-0907, "SQUG - Radwaste/Turbine Building A46 Spectra" (Ref. 19), by the maximum ratio of 2.0. The scaled ISRS was determined for all buildings and elevations where ESEL items are located at Peach Bottom.
 
14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 20 of 63 6  SEISMIC MARGIN EVALUATION APPROACH It is necessary to demonstrate that ESEL items have sufficient seismic capacity to meet or exceed the demand characterized by the RLGM. The seismic capacity is characterized as the peak ground acceleration (PGA) for which there is a high confidence of a low probability of failure (HCLPF). The PGA is associated with a specific spectral shape, in this case the 5%-damped RLGM spectral shape. The HCLPF capacity must be equal to or greater than the RLGM PGA. The criteria for seismic capacity determination are given in Section 5 of EPRI 3002000704 (Ref. 2).
There are two basic approaches fo r developing HCLPF capacities: 1. Deterministic approach using the conserva tive deterministic failure margin (CDFM) methodology of EPRI NP-6041, A Methodology for Assessment of Nuclear Power Plant Seismic Margin (Revision 1) (Ref. 7). 2. Probabilistic approach using the fragility analysis methodology of EPRI TR-103959, Methodology for Developing Seismic Fragilities (Ref. 8). For Peach Bottom, the deterministic approach using the CDFM methodology of EPRI NP-6041 (Ref. 7) was used to determine HCLPF capacities.
6.1 Summary of Methodologies Used Peach Bottom applied the Deterministic Approach (i.e. Method 1 from the previous section) to all items on the ESEL. The screening walkdowns used the screening tables from Chapter 2 of EPRI NP-6041 (Ref. 7). The walkdowns were conducted by engineers who as a minimum attended the SQUG Walkdown Screening and Seismic Evaluation Training Course. The walkdowns were documented on Screening Evaluation Work Sheets from EPRI NP-6041 (Ref. 7). Anchorage capacity calculations used the CDFM criteria from EPRI NP-6041 (Ref. 7) with Peach Bottom specific allowables and material strength s used as applicable. Seismic demand was the RLGM provided in Table 5.1-2 and Figure 5.1-1.
6.2 HCLPF Screening Process The peak RLGM (amplified PGA) For Peach Bottom equals 0.44 g (Table 5.1-2). The screening tables in EPRI NP-6041 (Ref. 7) are based on ground peak spectral accelerations of 0.8g and 1.2g. All Peach Bottom ESEL components were screened against either the caveats of the <0.8g column (lane 1) or the 0.8g-1.2g column (lane 2) of Table 2-4 of NP-6041 (Ref. 7). Screening based on lane 1 with the RLGM spectral shape yields an equivalent HCLPF of 0.44g PGA (witness 0.8g/0.44g*0.24g PGA = 0.44g PGA). Screening based on lane 2 with the RLGM spectral shape yields an equivalent HCLPF of 0.65g PGA (witness 1.2g/0.44g*0.24g PGA = 0.65g PGA).
14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 21 of 63 A number of components were located above 40 feet from grade. For components located 40 feet above grade, screening based on ground peak spectral acceleration is not applicable and additional consideration is required. In accordance with Appendix B of EPRI 1019200 (Ref. 20), components that are above 40 feet from grade and have corresponding ISRS at the base of component not in exceedance of 1.2g in the component frequency range of interest may be screened using the caveats of the 1st screening column, and components that are above 40 feet from grade and have corresponding ISRS at the base of component not in exceedance of 1.8g in the component frequency range of interest may be screened using the caveats of the 2nd
 
screening column. The screening of anchorage for non-valve components was evaluated either by SRT judgment or simple analysis. For components whose anchorage could not readily be screened by SRT judgment or simple analysis, CDFM HCLPF calculations (Ref. 9) were performed. This is documented in Attachments C and D. Per Ref 9.a, the seismic spectra for the Reactor Building (RB) are scaled from the original design spectra, which were based on an OBE seismic input and thus were based on a structural damping which is conservative for CDFM analysis (See Ref. 17). Under scaled SSE loading, a level of critical damping of 5% is appropriate for this structure. Based on NP-6041 Appendix Q and consideration of a structural damping level of 5%, it is shown in Ref. 9.b that Reactor Building elevations 135' and 165' may be addressed for CDFM purposes with peak spectral accelerations of 1.0g and 1.4g respectively. These spectral peak values are thus used for the purposes of equipment qualification as per NP-6041 Table 2-4.
 
6.3 Seismic Walkdown Approach 6.3.1 Walkdown Approach Walkdowns for Peach Bottom were performed in accordance with the criteria provided in Section 5 of EPRI 3002000704 (Ref. 2), which refers to EPRI NP-6041 (Ref. 7) for the Seismic Margin Assessment process. Pages 2-26 through 2-30 of EPRI NP-6041 (Ref. 7) describe the seismic walkdown criteria, including the following key criteria. "The SRT [Seismic Review Team] should "w alk by" 100% of all components which are reasonably accessible and in non-radioactive or low radioactive environments. Seismic capability assessment of components which are inaccessible, in high-radioactive environments, or possibly within contaminated containment, will have to rely more on alternate means such as photographic inspection, more reliance on seismic reanalysis, and possibly, smaller inspection teams and more hurried inspections. A 100% "walk by" does not mean complete inspection of each component, nor does it mean requiring an electrician or other technician to de-energize and open cabinets or panels for detailed inspection of all components. This walkdown is not intended to be a QA or QC review or a review of the adequacy of the component at the SSE level.
14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 22 of 63 If the SRT has a reasonable basis for assuming that the group of components are similar and are similarly anchored, then it is only necessary to inspect one component out of this group. The "similarity-basis" should be developed before the walkdown during the seismic capability preparatory work (Step 3) by reference to drawings, calculations or specifications. The one component or each type which is selected should be thoroughly inspected which probably does mean de-energizing and opening cabinets or panels for this very limited sample. Generally, a spare representative component can be found so as to enable the inspection to be performed while the plant is in operation. At least for the one component of each type which is selected, anchorage should be thoroughly inspected. The walkdown procedure should be performed in an ad hoc manner. For each class of components the SRT should look closely at the first items and compare the field configurations with the construction drawings and/or specifications. If a one-to-one correspondence is found, then subsequent items do not have to be inspected in as great a detail. Ultimately the walkdown becomes a "walk by" of the component class as the SRT becomes confident that the construction pattern is typical. This procedure for inspection should be repeated for each component class; although, during the actual walkdown the SRT may be inspecting several classes of components in parallel. If serious exceptions to the drawings or questionable construction practices are found then the system or component class must be inspected in closer detail until the systematic deficiency is defined. The 100% "walk by" is to look for outliers, lack of similarity, anchorage which is different from that shown on drawings or prescribed in criteria for that component, potential SI
[Seismic Interaction
] problems, situations that are at odds with the team members' past experience, and any other areas of serious seismic concern. If any such concerns surface, then the limited sample size of one component of each type for thorough inspection will have to be increased. The increase in sample size which should be inspected will depend upon the number of outliers and different anchorages, etc., which are observed. It is up to the SRT to ultimately select the sample size since they are the ones who are responsible for the seismic adequacy of all elements which they screen from the margin review. Appendix D gives guidance for sampling selection." The Peach Bottom walkdowns included as a minimum a 100% walk-by of all items on the ESEL except as noted in Section 7. Any previous walkdown information that was relied upon for SRT judgment is documented in Section 6.3.2.
 
EPRI 3002000704 (Ref. 2) page 5-4 limits the ESEP seismic interaction reviews to "nearby block walls" and "piping attached to tanks" which are reviewed "to address the possibility of failures due to differential displacements."  Other potential seismic interaction evaluations are "deferred to the full seismic risk evaluations performed in accordance with EPRI 1025287 (Ref. 14)."
 
14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 23 of 63 6.3.2 Application of Previous Walkdown Information The seismic walkdowns for Peach Bottom included as a minimum a walk-by of all the components on the ESEL with the exception of the items which are discussed in Section 7. Previous seismic walkdowns were used to support the ESEP seismic evaluations. Some of the components on the ESEL were included in the NTTF Recommendation 2.3 seismic walkdowns (Ref. 16). Photos taken during the NTTF R2.3 seismic walkdowns (Ref. 16), although available to the SRT during the ESEP walkdowns, were not necessary to the SRT at Peach Bottom. A-46 and IPEEE notes were available to the SRT and were used where appropriate to reduce the number of equipment items that needed to be opened and evaluate equipment that were not completely accessible to the SRT. Several ESEL items were previously walked down during the Peach Bottom 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.
6.3.3 Significant Walkdown Observations Consistent with that guidance from NP-6041 (Ref. 7), no significant outliers or anchorage concerns were identified during the Peach Bottom Atomic Power Station seismic walkdowns.
* Several block walls were identified in the proximity of ESEL equipment. These block walls were assessed for their structural adequacy to withstand the seismic loads resulting from the RLGM. For any cases where the block wall represented the HCLPF failure mode for an ESEL item, it is noted in the tabulated HCLPF values described in Section 6.6.
 
6.4 HCLPF Calculation Process ESEL items were evaluated using the criteria in EPRI NP-6041 (Ref. 7). Those evaluations included the following steps:
EPRI 3002000704 (Ref. 2) page 5-4 limits the ESEP seismic interaction reviews to "nearby block walls" and "piping attached to tanks" which are reviewed "to address the possibility of failures due to differential displacements."  Other potential seismic interaction evaluations are "deferred to the full seismic risk evaluations performed in accordance with EPRI 1025287 (Ref. 14)."
 
14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 24 of 63
* Performing seismic capability walkdowns for equipment to evaluate the equipment installed plant conditions
* Performing screening evaluations using the screening tables in EPRI NP-6041 (Ref. 7) as described in Section 6.2 and
* Performing HCLPF calculations considering various failure modes that include both structural failure modes (e.g. anchorage, load path etc.) and functional failure modes. All HCLPF calculations were performed using the CDFM methodology and are documented in Peach Bottom calculations (Ref. 9).
6.5 Functional Evaluation of Relays A HCLPF evaluation is performed for all relays and switches which may negatively "seal in" or "lock out" on the PBAPS ESEL. For relay evaluations, NP-6041 Appendix Q describes the following steps:
* Calculate in-cabinet response spectra (ICRS)
* Establish a clipping factor to be applied to the ICRS
* Determine a relay's Generic Equipment Ruggedness Spectra (GERS) Capacity
* Establish adjustment factors to convert the relay's GERS capacity to a CDFM level
* Compare clipped-peak and Zero Period Acceleration (ZPA) demands to the GERS capacity/test capacity HCLPF capacities for the relays are calculated using the procedure described above. The switch HCLPF value was determined by using existing test data in lieu of GERS. HCLPFs are calculated in 14Q4233-CAL-004 (Ref. 9) and are presented in Attachment C and D. Attachments C and D identify four relays for which operator action will be undertaken to reset the relay if necessary. Section 8.2 identifies two relays which are found to have a HCLPF capacity below the RLGM, and for which additional modifications, tests or analysis will be performed by the site.
6.6 Tabulated ESEL HCLPF Values (including key failure modes)
Tabulated ESEL HCLPF values including the key failure modes are included in Attachment C for Unit 2 and in Attachment D for Unit 3 items.
* For items screened out using NP-6041 (Ref. 7) screening tables, the screening level is provided as ">RLGM" and the failure mode is listed as "Screened."
 
14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 25 of 63
* For items where anchorage controls the HCLPF value, the HCLPF value is listed in the table and the failure mode is noted as "Anchorage."
* For items where block wall interaction controls the HCLPF value, the HCLPF value is listed in the table and the failure mode is noted as "Block Wall Interaction."
* For items where a relay or switch HCLPF controls, the relay or switch HCLPF value is listed in the table and the failure mode is listed as "Functional Failure".
* For items where equipment capacity based upon the screening lane values of Table 2-4 of EPRI NP-6041 (Ref. 7) controls the HCLPF value (e.g. anchorage, block wall, or relay HCLPF capacity exceeds the equipment capacity derived from screening), the screening lane HCLPF value is listed in the table and the failure mode is noted as "equipment capacity." Based on NP-6041 Table 2-4 Lane 1, this limit is equal to 0.44g for items below 40 feet above grade.
14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 26 of 63 7 INACCESSIBLE ITEMS 7.1 Identification of ESEL Items Inaccessible for Walkdowns Thirty three ESEL items were not accessible to the SRT during the ESEP walkdowns at Peach Bottom Atomic Power Station. A description of circumstances and disposition for these items is provided below. Table 7.1-1 Items Inaccessible for Walkdowns ID Description Resolution3AT540 3AT545 3BT540 3BT545 3CT540 3CT545 3GT540 3GT545 3KT540 3KT545 Nitrogen accumulatorsTh ese items are located in the Unit 3 Drywell and Isolation Valve Compartment. Since Unit 3 did not have a sch eduled outage in 2014, these items were inaccessible.
 
The Unit 2 Drywell and Isolation Valve Compartment were walked down during the Unit 2 scheduled outage in October 2014. Equivalent items in Unit 2 were seen to be rugged and well secured. Based on these observations, drawings and overall comparison of similar items between the two units, these items are expected to be similar to those in the Unit 2, and judged by SRT to be acceptable on that basis, including consideration of seismic interaction with block walls and piping attached to tanks. MO-3-13-021 Motor-operated valveRV-3-02-071A RV-3-02-071B RV-3-02-071C RV-3-02-071G RV-3-02-071K Relief valves2AT545 2CT545 2GT545 AccumulatorsSRT did not see theseaccumulators due to physical accessibility restriction in the drywell, but did walkdown the 2BT545, 2KT545 accumulators (among others not on the ESEL) and reviewed drawings. All accumulators seen were similar, and were
 
very ruggedly supported as shown in the SEWS photos, and are judged by SRT to be acceptable on that basis, including consideration of seismic interaction with block walls and piping attached to tanks.
00E072 Reheat coilThe reheat coilwas not directly visible, but was located by drawings and the site escort to be in line with large overhead ductwork. The Reheat Coil was judged by SRT to be adequately secured to the ductwork and screen out, including consideration of seismic interaction with block walls. 2AC270(K3A) 2BC270(K3B)
 
2AC270(K3C) 2BC270(K3D) 3AC270(K3A) 3BC270(K3B) Component relaysThese relays are inside control room cabinets2AC270, 2BC270, 3AC270 and 3BC270. These cabinets could not be opened during operation; therefore the relays were not walked down. However, the site provided a test report for the cabinet for the purpose of 14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 27 of 63 3AC270(K3C) 3BC270(K3D) component relay evaluation. Relays were analyzed to be adequate per 14Q4233-CAL-004. Cabinet was walked down and there are no block wall interactions in the vicinity. PO-0-40W-00016 PO-0-40W-00019-02 PO-0-40W-00808 Damper actuatorsThese damper actuatorswere not seen by SRT. They are understood to be inside an AHU unit that is connected to OAV034. All damper actuators that were observed during the walkdowns were either model D-251 or D-9504. The SRT judged these light weight actuators to be seismically rugged and adequately supported and they were screened out, including consideration of seismic interaction with block walls. AO-2-07B-2511 AO-2-07B-80290 Air-operated valvesThese valves are located on the top of the Torus. The area in the vicinity of these valves was walked by and valves were located by site escort with SRT. However, the valves were not directly accessible to SRT due to piping interferences. Due to a review of drawings and the similarity of other valves between units, these valves screened out, including consideration of seismic interaction with block walls.
7.2 Planned Walkdown / Evaluation Schedule / Close Out Since all items that were inaccessible during the ESEP were resolved by alternative means to the satisfaction of the SRT as discussed in Table 7.1-1 above, no additional walkdowns are required.
14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 28 of 63 8 ESEP CONCLUSIONS AND RESULTS 8.1 Supporting Information Peach Bottom Atomic Power Station has performed the ESEP as an interim action in response to the NRC's 50.54(f) letter (Ref. 1). It was performed using the methodologies in the NRC endorsed guidance in EPRI 3002000704 (Ref. 2). The ESEP provides an important demonstration of seismic margin and expedites plant safety enhancements through evaluations and potential near-term modifications of plant equipment that can be relied upon to protect the reactor core following beyond design basis seismic events. The ESEP is part of the overall Peach Bottom response to the NRC's 50.54(f) letter (Ref. 1). On March 12, 2014, NEI submitted to the NRC results of a study (Ref. 11) of seismic core damage risk estimates based on updated seismic hazard information as it applies to operating nuclear reactors in the Central and Eastern United States (CEUS). The study concluded that "site-specific seismic hazards show that there has not been an overall increase in seismic risk for the fleet of U.S. plants" based on the re-evaluated seismic hazards. As such, the "current seismic design of
 
operating reactors continues to provide a safety margin to withstand potential earthquakes exceeding the seismic design basis."  The NRC's May 9, 2014 NTTF 2.1 Screening and Prioritization letter (Ref. 13) concluded that the "fleetwide seismic risk estimates are consistent with the approach and results used in the Gl-199 safety/risk assessment."  The letter also stated that "As a result, the staff has confirmed that the conclusions reached in Gl-199 safety/risk assessment remain valid and that the plants can continue to operate while additional evaluations are conducted." An assessment of the change in seismic risk for Peach Bottom was included in the fleet risk evaluation submitted in the March 12, 2014 NEI letter (Ref. 11) therefore, the conclusions in the NRC's May 9 letter (Ref. 13) also apply to Peach Bottom. In addition, the March 12, 2014 NEI letter (Ref. 11) provided an attached "Perspectives on the Seismic Capacity of Operating Plants," which (1) assessed a number of qualitative reasons why the design of SSCs inherently contain margin beyond their design level, (2) discussed industrial seismic experience databases of performance of industry facility components similar to nuclear SSCs, and (3) discussed earthquake experience at operating plants. The fleet of currently operating nuclear power plants was designed using conservative practices, such that the plants have significant margin to withstand large ground motions safely. This has been borne out for those plants that have actually experienced significant earthquakes. The seismic design process has inherent (and intentional) conservatisms which result in significant seismic margins within structures, systems and components (SSCs). These conservatisms are reflected in several key aspects of the seismic design process, including:
14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 29 of 63
* Safety factors applied in design calculations
* Damping values used in dynamic analysis of SSCs
* Bounding synthetic time histories for in-structure response spectra calculations
* Broadening criteria for in-structure response spectra
* Response spectra enveloping criteria typically used in SSC analysis and testing applications
* Response spectra based frequency domain analysis rather than explicit time history based time domain analysis
* Bounding requirements in codes and standards
* Use of minimum strength requirements of structural components (concrete and steel)
* Bounding testing requirements, and
* Ductile behavior of the primary materials (that is, not crediting the additional capacity of materials such as steel and reinforced concrete beyond the essentially elastic range, etc.). These design practices combine to result in margins such that the SSCs will continue to fulfill their functions at ground motions well above the SSE. The intent of the ESEP is to perform an interim action in response to the NRC's 50.54(f) letter (Ref. 1) to demonstrate seismic margin through a review of a subset of the plant equipment that can be relied upon to protect the reactor core fo llowing 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 plant's SSE rather than the actual GMRS. To more fully characterize the risk impacts of the seismic ground motion represented by the GMRS on a plant specific basis, a more detailed seismic risk assessment (SPRA or risk-based SMA) is to be performed in accordance with EPRI 1025287 (Ref. 14). As identified in the Peach Bottom Seismic Hazard and GMRS submittal (Ref. 6), Peach Bottom 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. Peach Bottom will complete that evaluation in accordance with the schedule identified in NEI's letter dated April 9, 2013 (Ref. 12) and endorsed by the NRC in their May 7, 2013 letter (Ref. 13).
 
14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 30 of 63 8.2 Identification of Planned Modifications The following two relays were identified with HCLPF capacities below the RLGM. These relays will be further evaluated and may require modification.
Component ID Resolution 2-13A-K033 See below for a list of potential solutions. 3-13A-K033 See below for a list of potential solutions.
Solutions which may be considered in qu alifying 2-13A-K033 and 3-13A-K033 include:
* Stiffen or replace host cabinets 20C033 and 30C033. A modification which significantly stiffens the host cabinet for each of the subject relays would lower the seismic demand.
* Relocate the subject relays to a mo re seismically favorable location.
* Replace the subject relays with a compatible relay model.
* Determine a higher capacity by shake table testing of this relay model.
* Reduce seismic demand through analysis or reduction of existing conservatisms
* Risk analysis 8.3 Modification Implementation Schedule The modification implementation schedule will be included in the transmittal letter from Exelon to the NRC for this report.
8.4 Summary of Regulatory Commitments Regulatory commitments for Peach Bottom modification implementation will be included in the transmittal letter from Exelon to the NRC for this report.
14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 31 of 63 9 REFERENCES
: 1. NRC (E Leeds and M Johnson) Letter to All Power Reactor Licensees et al., "Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(f) Regarding Recommendations 2.1, 2.3 and 9.3 of the Near-Term Task Force Review of Insights from the Fukushima Dai-Ichi Accident," March 12, 2012
: 2. Seismic Evaluation Guidance: Augmented Approach for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1 - Seismic. EPRI, Palo Alto, CA: May 2013. 3002000704
: 3. Peach Bottom Letters
: a. NRC Letter RS-13-024 from Peach Bottom (ML13059A305), "Overall Integrated Plan in Response to March 12, 2012 Commission Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events (Order Number EA-12-049)", February 28, 2013
: b. NRC Letter RS-13-127 from Peach Bottom (ML13246A412), "First Six-Month Status Report in Response to March 12, 2012 Commission Order Modifying Licenses with Regard to Requirements for Mitigation St rategies for Beyond-Design-Basis External Events (Order Number EA-12-049)", August 28, 2013
: c. NRC Letter RS-14-014 from Peach Bottom (ML14059A222), "Second Six-Month Status Report in Response to March 12, 2012 Commission Order Modifying Licenses with Regard to Requirements for Mitigation St rategies for Beyond-Design-Basis External Events (Order Number EA-12-049)", February 28, 2014
: d. NRC Letter RS-14-212 from Peach Bottom (ML14241A252), "Third Six-Month Status Report in Response to March 12, 2012 Commission Order Modifying Licenses with Regard to Requirements for Mitigation St rategies for Beyond-Design-Basis External Events (Order Number EA-12-049)", August 28, 2014
: 4. Peach Bottom Station Transmittal of Design Information to Stevenson & Associates, Tracking No: TODI AR#2397100-02, "PB Flex Strategy Rev. 1", December 9, 2014
: 5. Peach Bottom Station Transmittal of Design Information to Stevenson & Associates, Tracking No: PB 1570792-76, Input relating to Relays and Switches, December 4, 2014
: 6. Seismic Hazard and Screening Report in Response to the 50.54(f) Information Request Regarding Fukushima Near-Term Task Force Recommendation 2.1: Seismic for Peach Bottom Atomic Power Station dated 3/31/14, Correspondence No. RS-14-071 (Exelon Report EXLNPB056-PR-001, Revision 1)
: 7. A Methodology for Assessment of Nuclear Power Plant Seismic Margin, Rev. 1, August 1991, Electric Power Research Institute, Palo Alto, CA. EPRI NP-6041
: 8. Methodology for Developing Seismic Fragilities, August 1991, EPRI, Palo Alto, CA. 1994, TR-103959
: 9. Peach Bottom HCLPF Calculations for the ESEP project 14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 32 of 63
: a. 14Q4233-CAL-001, Rev. 0, Generation of In-Structure Response Spectra for use in ESEP Evaluations
: b. 14Q4233-CAL-002, Rev. 0, HCLPF Analysis for ESEP Evaluations for PBAPS
: c. 14Q4233-CAL-003, Rev. 0, ESEP Block Wall HCLPFs
: d. 14Q4233-CAL-004, Rev. 0, ESEP HCLPFs for Relays
: 10. Nuclear Regulatory Commission, NUREG/CR-0098, Development of Criteria for Seismic Review of Selected Nuclear Power Plants, published May 1978
: 11. Nuclear Energy Institute (NEI), A. Pietrangelo, Letter to D. Skeen of the USNRC, "Seismic Core Damage Risk Estimates Us ing the Updated Seismic Hazards for the Operating Nuclear Plants in the Central and Eastern United States", March 12, 2014
: 12. Nuclear Energy Institute (NEI), A. Pietrang elo, Letter to D. Skeen of the USNRC, "Proposed Path Forward for NTTF Recommendation 2.1: Seismic Reevaluations", April 9, 2013  13. NRC (E. Leeds) Letter to All Power Reacto r Licensees et al. (ML14111A147), "Screening and Prioritization Results Regarding Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(F) Regarding Seismic Hazard Re-Evaluations for Recommendation 2.1 of the Near-Term Task Force Review of Insights From the Fukushima Dai-Ichi Accident," May 9, 2014
: 14. Seismic Evaluation Guidance: Screening, Prioritization and Implementation Details (SPID) for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1: Seismic. EPRI, Palo Alto, CA: February 2013. 1025287
: 15. NRC (E. Leeds) Letter to NEI (J Pollock) (ML13106A331), "Electric Power Research Institute Final Draft Report xxxxxx, "Seismic Evaluation Guidance: Augmented Approach for the Resolution of Fukushima Near-Term Task Fo rce Recommendation 2.1: Seismic," as an Acceptable Alternative to the March 12, 2012, Information Request for Seismic Reevaluations," May 7, 2013
: 16. PBAPS NTTF 2.3 Seismic Walkdown Submittals for Unit 2 and Unit 3, dated 11/20/2014
: 17. Peach Bottom Atomic Power Station Updated Final Safety Analysis Report (UFSAR)
Appendix C, Rev.24
: 18. Peach Bottom Specification 11187-G-14 Rev.0, General Project Requirements for Seismic Design and Analysis of Equipment and Equipment Supports for the Peach Bottom Atomic Power Station Units 2 & 3
: 19. Peach Bottom Calculation PS-0907 Rev.0, SQUG - Radwaste/Turbine Building A46 Spectra 14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 33 of 63
: 20. EPRI Technical Report (TR) 1019200, "Seismic Fragility Applications Guide Update," December 2009
: 21. Peach Bottom Drawings
: a. M-315 Sheet 1, Rev. 65, P&I Diagram Emergency Service Water and High Pressure Service Water Systems
: b. M-333 Sheet 1, Rev. 57, P&ID Diagram Instrument Nitrogen (Unit 2)
: c. M-333 Sheet 2, Rev. 58, P&ID Diagram Instrument Nitrogen (Unit 3)
: d. M-351 Sheet 1, Rev. 78, P&ID Nuclear Boiler
: e. M-351 Sheet 2, Rev. 70, P&ID Nuclear Boiler
: f. M-351 Sheet 3, Rev. 74, P&ID Nuclear Boiler
: g. M-351 Sheet 4, Rev. 69, P&ID Nuclear Boiler
: h. M-359 Sheet 1, Rev. 50, P&ID Diagram Reactor Core Isolation Cooling
: i. M-359 Sheet 2, Rev. 48, P&ID Diagram Reactor Core Isolation Cooling
: j. M-360 Sheet 1, Rev. 56, R.C.I.C. Pump Turbine Details
: k. M-360 Sheet 2, Rev. 54, R.C.I.C. Pump Turbine Details
: l. M-360 Sheet 3, Rev. 47, R.C.I.C. Pump Turbine Details Lube Oil and Control System Unit 2 m. M-360 Sheet 4, Rev. 39, R.C.I.C. Pump Turbine Details Lube Oil and Control System Unit 3 n. M-361 Sheet 1, Rev. 82, P&I Diagram Residual Heat Removal Sys (Unit 2)
: o. M-361 Sheet 2, Rev. 68, P&I Diagram Residual Heat Removal Sys (Unit 2)
: p. M-361 Sheet 3, Rev. 70, P&I Diagram Residual Heat Removal Sys (Unit 3)
: q. M-361 Sheet 4, Rev. 072, Residual Heat Removal System (Unit 3)
: r. M-367 Sheet 1, Rev. 85, P&ID Diagram Co ntainment Atmospheric Control System (Unit 2) s. M-367 Sheet 2, Rev. 76, P&ID Diagram Co ntainment Atmospheric Control System (Unit 3) t. M-372 Sheet 1, Rev. 62, P&ID Diagram Co ntainment Atmosphere Dilution System
: u. M-384 Sheet 1, Rev. 39, P&I Diagram Control Room HVAC
: v. M-384 Sheet 2, Rev. 6, P&I Diagram Control Room HVAC 
: w. M-384 Sheet 3, Rev. 6, P&I Diagram Control Room HVAC
: x. M-399 Sheet 1, Rev. 32, P&I Diagram Emergency Switchgear, Battery Room, Laboratory Supply & Exhaust
: y. M-399 Sheet 2, Rev. 4, P&I Diagram Emergency Switchgear, Battery Room, Laboratory Supply & Exhaust
: z. M-399 Sheet 3, Rev. 2, P&I Diagram Emergency Switchgear, Battery Room, Laboratory Supply & Exhaust aa. M-399 Sheet 4, Rev. 5, P&I Diagram Emergency Switchgear, Battery Room, Laboratory Supply & Exhaust
: 22. 14Q4233-RPT-003, Rev.1, Validation of Expedited Seismic Equipment List 
: 23. Seismic Qualification Utility Group (SQUG), Generic Implementation Procedure (GIP) for Verification of Nuclear Plant Equipment, Revision 3A
: 24. NRC Order Number EA-12-049 (ML12054A735), "Issuance of Order to Modify Licenses with Regard to Requirements for Mitigation Strategies for Beyond-D esign-Basis External Events," dated March 12, 2012.
 
14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 34 of 63
: 25. Email to S&A from Ms. Tracey L. Gallagher (Exelon), LRC-009, Confirmation of "no changes" between February 2013 and August 2014 Flex Strategies, December 4, 2014
: 26. Email to S&A from Ms. Tracey L. Gallagher (Exelon), LRC-010, Confirmation from "Ops" of "no significant changes" between February 2013 and August 2014 Flex Strategies, December 4, 2014
: 27. Email to S&A from Ms. Tracey L. Gallagher (Exelon), LRC-025, Inputs to RPT-004 including that "Flex Phase 1 and 2 strategy will provide sufficient capability such that no additional Phase 3 strategies are required", December 17, 2014
 
14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 35 of 63
 
Attachment A PBAPS Unit 2 and Common ESEL
 
14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 36 of 63 Equipment ID Description Equipment Normal State Equipment Desired State Notes 00C133 Panel Energized Energized  00E068 CR Fresh Air Supply Preheat Coil Standby Standby  00E072 CR Vent Reheat Coil Standby Standby  00F039 CR Fresh Air Supply Roll Filter Standby Standby  00F043 OA Damper Emergency Switchgear & Battery Room Vent Supply Roll Filter Standby Standby 00T116 CAD Liquid Nitrogen Storage Tank Standby In Service Passive component 0AE073 OA Damper Emergency Switchgear &
Battery Room Vent Supply Heat Coil Standby Standby 0AF041 A Train HEPA Filter Standby Standby Passive Component 0AF042 A Train Charcoal Filter Standby Standby Passive Component 0AF050 A Train HEPA Filter Standby Standby Passive Component 0AV030 CR Room Emergency Vent Fan Standby Energized  0AV034 Emergency Switch Gear and Battery Room Supply Fan Standby Operating 0AV036 Battery Room Exhaust Fan Standby Operating  20C003 Reactor and Containment Cooling and Isolation Panel Energized Energized 20C004C RCIC Control Panel Energized Energized  20C005A Reactor Manual Control Panel Energized Energized  20C012 Plant Services Console Energized Energized  20C018 Panel Energized Energized  20C019 Panel Energized Energized Contains power supply for PT-2-13-068 20C032 Panel Energized Energized  20C033 Panel Energized Energized  20C034 RCIC Relay Panel Energized Energized  20C035 Panel Energized Energized  20C041 Panel Energized Energized  20C095 RCIC Instrument Rack Energized Energized  20C144 Panel Energized Energized  20C722A Accident Monitoring Instrumentation Panel Energized Energized 20C722B Panel Energized Energized  20C818 Reactor Water Level/Pressure Component Cabinet Energized Energized 20D021 (2PPA) 125V DC Station Distribution Energized Energized 20D023 (2PPC) 125V DC Station Distribution Energized Energized  20D024 Distribution Panel Energized Energized  20D037 Uninterruptable Power Supply Static Inverter Energized Energized Powers vital instrument bus during Phase 1 20D039 RCIC Barometric CDSR Vacuum Pump
 
Starter Standby Energized 20D040 RCIC Barometric CDSR Cond Pump
 
Starter Standby Energized 20P036 RCIC Pump Standby Operating  20P046 RCIC Barometric CDSR Vacuum Pump Standby Operating  20P048 RCIC Barometric CDSR Condensate
 
Pump Standby Operating 20P340 RCIC Turbine Driven Lube Oil Pump Standby Operating  20S038 RCIC Turbine Standby Operating Controlled via Included Governor Valve and Trip & Throttle Valve 20S315 Static Inverter Man Bypass/Isolation Switch Energized Bypassed and
 
Isolated  Phase 1 power is from inverter, Phase 2 is from 480/120V AC transformer 20S354 Load Center E-124/E324 Transfer Switch Standby Energized 20S700 Battery Charger Panel 2AD003 Transfer Switch 20S700 Energized Energized 20S701 Transfer Switch Energized Energized  20S703 Transfer Switch Energized Energized  20X133 Transformer Energized Energized  20X135 20Y035 Transformer Energized Energized  20X150 Transformer Energized Energized  20Y033 Distribution Panel Energized Energized 14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 37 of 63 Equipment ID Description Equipment Normal State Equipment Desired State Notes 20Y035 120 VAC 'Y' Power Panel Energized Energized  20Y050 Uninterruptable Power Supply Distribution Panel Energized Energized 2-13A-K004 RCIC Hi Temp Steam Leak Relay Standby Standby  2-13A-K006 RCIC Hi Temp Steam Leak Relay Standby Standby  2-13A-K007 RCIC Steam Line Hi DP Line Break Relay Standby Standby 2-13A-K010 RCIC Turbine Trip Aux Relay Standby Standby  2-13A-K011 RCIC Turbine Trip Aux Relay Standby Standby Closes Trip & Throttle Valve when energized 2-13A-K012 RCIC Auto Isolation Signal Relay Standby Standby  2-13A-K014 RCIC Pump Lo Suction Pressure Trip Relay Standby Standby 2-13A-K017 RCIC Turbine Exh Hi Pressure Trip Relay Standby Standby 2-13A-K-022 RCIC Auto Isolation Signal Relay Standby Standby  2-13A-K030 Reactor Hi Vessel Water Level Trip Relay Standby Standby 2-13A-K031 RCIC Steam Line Space Hi Temp Isolation Relay Standby Standby 2-13A-K032 RCIC Steam Line Space Hi Temp Isolation Relay Standby Standby 2-13A-K033 RCIC Steam Line Hi DP Line Break Relay Standby Standby 2-13A-K034 RCIC Auto Isolation Signal Relay Standby Standby  2-13A-K044 RCIC Reactor Hi Vessel Water Level Trip Relay Standby Standby 2-13A-K049 RCIC Low Steam Pressure Auto Isolation Relay Standby Standby 2-13A-K050 RCIC Low Steam Pressure Auto Isolation Relay Standby Standby 2-13A-K053 RCIC Reactor Hi Vessel Water Level
 
Trip Relay Auxiliary Standby Standby 2-13A-K054 RCIC Auto Isolation Signal Relay Standby Standby  2AC043 Emergency Shutdown Panel Standby Standby  2AC065 Rx Vessel Lvl and Pressure Inst Rack A Energized Energized  2AC091 Jet Pump Inst Rack A Energized Energized  2AC270 Panel Energized Standby  2AD001 2A 125V DC Battery Energized Energized  2AD003 Station Battery Charger 2A Energized Energized  2AD017 (2FPA) Battery Main Fuse Box Energized Energized  2AD018 (2DPA) 250V DC Distribution Panel Energized Energized  2AD019 (2FA) 250 Volt Fuse Box Energized Energized  2AD025 Distribution Panel Energized Energized  2AE024 Residual Heat Exchangers Standby Standby Passive component 2AS377 Back-Up N2 Supply to Ads RV's Standby Open Passive component 2AT545 2A Srv Inst N2 Accumulator Standby Standby Passive component 2BC043 Panel Standby Standby Contains control switch for LT-2-02-3-085A 2BC065 Rx Vessel Lvl and Pressure Inst Rack B Energized Energized  2BC172 Panel Standby Standby  2BC270 Panel Energized Standby  2BD001 2B 125V DC Battery Standby Energized Provides power for 120V AC Vital Instrument Power 2BD017 Battery Main Fuse Box Energized Energized  2BD018 250V DC Distribution Panel Div. II Energized Energized  2BE024 Residual Heat Exchangers Standby Standby Passive component 2BS377 Back-Up N2 Supply to Ads RV's Standby Open Passive component 2BS545 Automatic Transfer Switch Panel Energized Energized  2BT545 2B Srv Inst N2 Accumulator Standby Standby Passive component 2CC133 Panel Energized Energized  2CD001 2C 125V DC Battery Energized Energized  2CD003 Station Battery Charger 2C Energized Energized  2CD017 (2FPC) Battery Main Fuse Box Energized Energized 14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 38 of 63 Equipment ID Description Equipment Normal State Equipment Desired State Notes 2CD019 (2FC) 250 Volt Fuse Box Energized Energized  2CE024 Residual Heat Exchangers Standby Standby Passive component 2CS377 Back-Up N2 Supply to Ads RV's Standby Open Passive component 2CS545 Automatic Transfer Switch Panel Energized Energized  2CT545 2C Srv Inst N2 Accumulator Standby Standby Passive component 2DA-W-A (1201) RCIC MO-2-13-021 Breaker Energized Energized  2DA-W-A (1203) RCIC MO-2-13-030 Breaker Energized Energized  2DA-W-A (1204) RCIC MO-2-13-027 Breaker Energized Energized  2DA-W-A (1205) RCIC MO-2-13-041 Breaker Energized Energized  2DA-W-A (1206) RCIC MO-2-13-039 Breaker Energized Energized  2DA-W-A (1207) RCIC MO-2-13-132 Breaker Energized Energized  2DA-W-A (1209) RCIC MO-2-13-131 Breaker Energized Energized  2DA-W-A (1210) RCIC MO-2-13-018 Breaker Energized Energized  2DA-W-A (1214) RCIC Cond Vac PP 20P046 Breaker Energized Energized  2DA-W-A (1215) RCIC Vac Tank Cond PP 20P048 Breaker Energized Energized 2DA-W-A (20D012) RCIC 250VDC MCC Energized Energized  2DD001 2D 125V DC Battery Standby Energized Provides power for 120V AC Vital Instrument Power and RCIC B Logic 2DD017 Battery Main Fuse Box Energized Energized  2DD019 Fuse Box Energized Energized  2DE024 Residual Heat Exchangers Standby Standby Passive component 2GT545 2G Srv Inst N2 Accumulator Standby Standby Passive component 2KT545 2K Srv Inst N2 Accumulator Standby Standby Passive component 2OE032 RCIC Barometric Condenser (13-2) Standby Operating Passive Component 2OE104 RCIC Turb. Lube Oil Cooler (13-2) Standby Operating Passive Component 2AC270 (K3A) Relay De-Energized De-Energized  2BC270 (K3B) Relay De-Energized De-Energized  2AC270 (K3C) Relay De-Energized De-Energized  2BC270 (K3D) Relay De-Energized De-Energized  AO-2-07B-2511 Torus 18 Inch Vent Inboard Isol Valve
 
to Sbgt/Atmos Closed Open AO-2-07B-80290 Ctmt Emerg Vent Outboard Isolation Vlv to Atmos Closed Open E124 (1013) E124-R-C 20B036 Breake r  Energized Energized  E124 (1014) E124-T-B 20B059 Breaker Energized Energized E124 (20B010) E124 Load Center  Energized Energized E124-R-C (20B036) E124-R-C Motor Control Center Energized Energized E124-R-C (3606) MO-2-10-25A Norm Breaker Energized Energized E124-R-C (3691) Alt Feed Breaker 20D037 Y50 Energized Energized  E124-T-B (20B059) E124-T-B Motor Control Center Energized Energized  E124-T-B (5931) Norm Fdr for 125 V Battery Charger 'A' 2AD03 Energized Energized E324 (1213) E324-R-B 20B038 Breaker Energized Energized  E324 (1222) E324-T-B 00B049 Breaker Energized Energized E324 (20B012) E324 Load Center Energized Energized E324-R-B (20B038) MCC 20B038 Energized Energized E324-R-B (3821) MO-2-10-038A Breaker Energized Energized 14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 39 of 63 Equipment ID Description Equipment Normal State Equipment Desired State Notes E324-R-B (3822) MO-2-10-020 Breaker Locked De-Energized Energized E324-R-B (3824) MO-2-10-026A Breaker Energized Energized E324-R-B (3831) MO-2-10-039A Breaker Energized Energized E324-R-B (3832) MO-2-10-034A Breaker Energized Energized E324-R-B (3844) MO-2-10-031A Breaker Energized Energized E324-R-B (3862) MO-2-10-174 Breaker Energized Energized E324-R-B (3863) MO-2-10-176 Breaker De-Energized Energized E324-R-B (3882) Norm Fdr for 120V Instr Pnl 20Y35 Trans 20X135 Energized Energized E324-R-B (3893) 125 VDC Batt Charger 2CD03 Energized Energized  E324-T-B (00B049) MCC 00B049 for 0AV034, 0AV036, and 0AV030 Energized Energized INV-2-13-90 RCICs-125 VDC Bus 'A' Power Distribution Energized Energized  J-1648 Junction Box Standby Standby Contains resistor for PT-2-13-068 J2915 J-Box at E124 Load Center Standby Energized  J2916 J-Box at E324 Load Center Standby Energized LI-2-02 085A Reactor Vessel High Water Energized Energized  LI-2-02-3-113 Reactor Water Level Energized Energized LR/TR-8123A Torus Water Level/Temperature Recorder Energized Energized LT-2-02 072A Reactor Vessel Water Level Energized Energized  LT-2-02-3-113 Reactor Press Vessel Fuel Zone Wtr Level Energized Energized LT-8123A Torus Water Level Energized Energized  MO-2-10-020 RHR Loops A/B X-Tie Closed Open  MO-2 026A RHR Loop A D/W Spray O/B Closed Open/Closed  MO-2 031A RHR Loop A D/W Spray I/B Closed Open/Closed  MO-2 034A RHR Loop A FFT Valve Closed Open  MO-2 038A RHR Loop A Torus Spray Closed Open/Closed  MO-2 039A RHR Loop A Torus Valve Closed Open  MO-2-10-174 HPSW to RHR Inner X-Tie Closed Open  MO-2-10-176 HPSW to RHR Outer X-Tie Closed Open  MO-2-10-25A RHR Loop A I/B Disc Valve Closed Open Valve closure via control switch requires core spray relay logic permissive MO-2-13-018 RCIC Pump Suction from Condensate Storage Tank Open Open/Closed MO-2-13-021 RCIC Discharge to Feedwater Line B Closed Open  MO-2-13-027 RCIC Minimum Flow Valve Closed Open/Closed  MO-2-13-030 RCIC Full Flow Test Valve Closed Open/Closed  MO-2-13-039 RCIC Pump Torus Suction Outer Closed Open/Closed  MO-2-13-041 RCIC Pump Torus Suction Inner Closed Open/Closed  MO-2-13-131 RCIC Turbine Steam Supply Valve Closed Open/Closed  MO-2-13-132 RCIC Cooling Water Supply to Lo Clr +
Barometric Cdsr Closed Open N210025A Cabinet Provides Power to Valve MO-2-10-25A Energized Energized OAS384 N2 Tank Standby Standby Passive Component OAS385 N2 Tank Standby Standby Passive Component OBS384 N2 Tank Standby Standby Passive Component OBS385 N2 Tank Standby Standby Passive Component PCV-0-40W-70088A N2 Regulator - OAS384 Standby Open 14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 40 of 63 Equipment ID Description Equipment Normal State Equipment Desired State Notes PCV-0-40W-70088B N2 Regulator - OBS384 Standby Open  PCV-0-40W-70089A N2 Regulator - OAS385 Closed Open  PCV-0-40W-70089B N2 Regulator - OBS385 Closed Open  PI-2-06-090A Reactor Wide Range Press Ind Energized Energized  PI-2-06-090B Reactor Wide Range Press Ind Energized Energized  PI-2-06-090C Reactor Wide Range Press Ind Energized Energized  PI-2-13-094 RCIC Pump Turb Stm Press Energized Energized  PO-0-40D-00153-01 CR Emergency Vent Filters Inlet Damper Closed Open Fails open on loss of instrument air PO-0-40D-00153-02 CR Emergency Vent Filters Inlet Damper Closed Open Fails open on loss of instrument air PO-0-40D-00157-01 CR Emergency Vent Supply Fan Inlet Damper Closed Open Fails open on loss of instrument air PO-0-40D-00157-02 CR Emergency Vent Supply Fan Outlet Damper Closed Open Fails open on loss of instrument air PO-0-40W-00016 OA Damper Emergency Switchgear And Battery Room Vent Supply Fans Open Throttled Fails closed to minimum on loss of instrument air PO-0-40W-00019-01 OA Damper Emergency Switchgear And Battery Room Vent Supply Damper Standby Open PO-0-40W-00019-02 OA Damper Emergency Switchgear And Battery Room Vent Outlet Damper Standby Open PO-0-40W-00021-01 OA Damper Emergency Switchgear And Battery Room Vent Supply Damper Standby Open PO-0-40W-00031-01 Battery Room Exhaust Fan Inlet Damper Standby Open PO-0-40W-00031-02 Battery Room Exhaust Fan Outlet Damper Standby Open PO-0-40W-00782-01 OA Damper Emergency Switchgear and Battery Room Vent Supply Damper Standby Open PO-0-40W-00782-02 OA Damper Emergency Switchgear and Battery Room Vent Supply Damper Standby Open PO-0-40W-00808 OA Damper Emergency Switchgear and Battery Room Vent Outlet Damper Standby Open PO-0-40W-00822-01 OA Damper Emergency Switchgear and Battery Room Vent Supply Damper Standby Open PO-0-40W-00822-02 OA Damper Emergency Switchgear and Battery Room Vent Supply Damper Standby Open PO-0-40W-00822-03 OA Damper Emergency Switchgear and Battery Room Vent Supply Damper Standby Open PR/LR-2 096 Reactor Level/Steam Flow Ratio Energized Energized  PR/TR-4805 Containment Pressure/Temp Energized Energized  PS-2-13-67-1 Pressure Switch Open Open  PS-2-13-72A Pressure Switch Open Open  PS-2-13-72B Pressure Switch Open Open  PS-2-13-87A Pressure Switch Open Open  PS-2-13-87B Pressure Switch Open Open  PT-2-02 404A Reactor Pressure Energized Energized PT-2-02 404C Reactor Pressure Transmitter Energized Energized  PT-2-06-053A Reactor Wide Range -Pressure Energized Energized  PT-2-06-053B Reactor Wide Range -Pressure Energized Energized  PT-2-06-053C Reactor Wide Range -Pressure Energized Energized  PT-2-13-068 RCIC Turbine Steam Supply Press Energized Energized  PT-4805 Drywell Pressure Energized Energized  RV-2-02-071A 2A Safety Relief Valve Closed Open/Closed  RV-2-02-071B 2B Safety Relief Valve Closed Open/Closed  RV-2-02-071C 2C Safety Relief Valve  Closed Open/Closed  RV-2-02-071G 2G Safety Relief Valve Closed Open/Closed  RV-2-02-071K 2K Safety Relief Valve Closed Open/Closed SV-0-36B-00019 OA Damper Emergency Switchgear and
 
Battery Room Damper IA Standby Energized 
 
14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 41 of 63 Equipment ID Description Equipment Normal State Equipment Desired State Notes SV-0-36B-00031 Air Supply Shutoff for PO-0-40W-00031-01, PO-0-40W-00031-02 Standby Energized TI-2501 Ventilation Air Temperatures Energized Energized  TI-80146 Drywell Bulk Average Temp Indicator Energized Energized  TT-2501 Vent Air/Wtr Temp Energized Energized  XAM-2-02 117A Reactor Water Level Wide Range Energized Energized 14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 42 of 63
 
Attachment B PBAPS Unit 3 ESEL
 
14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 43 of 63 Equipment ID Description Equipment Normal State Equipment Desired State Notes 30C003 Reactor and Containment Cooling and Isolation Energized Energized  30C004C RCIC Control Panel Energized Energized  30C005A Reactor Manual Control Panel Energized Energized  30C012 Plant Services Console Energized Energized This panel is not credited for the Unit 3
 
ESEP-pertinent FLEX Response, however it remains on the list for conservatism 30C018 Panel Energized Energized  30C019 Panel Energized Energized  30C032 Panel  Energized Energized  30C033 Panel Energized Energized  30C034 RCIC Relay Panel Energized Energized  30C035 Panel Energized Energized  30C041 Panel Energized Energized  30C095 RCIC Instrument Rack Energized Energized  30C144 Panel Energized Energized  30C722A Accident Monitoring Instrumentation Panel Energized Energized  30C722B Panel Energized Energized  30C818 Reactor Water Level/Pressure Component Cabinet Energized Energized 30D021 (3PPA) 125V DC Station Distribution Energized Energized 30D023 (3PPC) 125V DC Station Distribution Energized Energized  30D024 Distribution Panel Energized Energized  30D037 Uninterruptable Power Supply Static Inverter Energized Bypassed and
 
Isolated  Powers vital instrument bus during Phase 1 30D039 RCIC Barometric Cdsr Vacuum Pump Starter Standby Energized 30D040 RCIC Barometric Cdsr Cond Pump Starter Standby Energized  30P036 RCIC Pump Standby Operating  30P046 RCIC Barometric Cdsr Vacuum Pump Standby Operating  30P048 RCIC Barometric Cdsr Condensate Pump Standby Operating  30P340 RCIC Turbine Driven Lube Oil Pump Standby Operating  30S038 RCIC Turbine Standby Operating  30S315 Static Inverter Man Bypass/Isolation Switch Energized Bypassed and Isolated  Phase 1 power is from inverter, Phase 2 is from 480/120V AC transformer 30S356 Load Center E134/E334 Transfer Switch Standby Energized  30S546 Transfer Switch Energized Energized  30S701 Transfer Switch Energized Energized  30S703 Transfer Switch Energized Energized  30S704 125V DC Battery Charger 3CD003 Transfer Sw 30S704 Energized Energized  30X133 Transformer Energized Energized  30X135 30Y035 Transformer Energized Energized  30X150 30Y050 Transformer Energized Energized  30Y033 Distribution Panel Energized Energized  30Y035 120 VAC 'Y' Power Panel Energized Energized  30Y050 120 VAC 'Y' Power Panel Energized Energized  3-13A-K004 RCIC Hi Temp Steam Leak Relay Standby Standby  3-13A-K006 RCIC Hi Temp Steam Leak Relay Standby Standby  3-13A-K007 RCIC Steam Line Hi DP Line Break Relay Standby Standby  3-13A-K010 RCIC Turbine Trip Aux Relay Standby Standby  3-13A-K011 RCIC Turbine Trip Aux Relay Standby Standby Closes Trip & Throttle Valve when energized 3-13A-K012 RCIC Auto Isolation Signal Relay Standby Standby  3-13A-K014 RCIC Pump Lo Suction Pressure Trip Relay Standby Standby  3-13A-K017 RCIC Turbine Exh Hi Pressure Trip Relay Standby Standby  3-13A-K-022 RCIC Auto Isolation Signal Relay Standby Standby  3-13A-K030 Reactor Hi Vessel Water Level Trip Relay Standby Standby  3-13A-K031 RCIC Steam Line Space Hi Temp Isolation Relay Standby Standby 3-13A-K032 RCIC Steam Line Space Hi Temp Isolation Relay Standby Standby 3-13A-K033 RCIC Steam Line Hi DP Line Break Relay Standby Standby 14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 44 of 63 Equipment ID Description Equipment Normal State Equipment Desired State Notes 3-13A-K034 RCIC Auto Isolation Signal Relay Standby Standby  3-13A-K044 RCIC Reactor Hi Vessel Water Level Trip Relay Standby Standby 3-13A-K049 RCIC Low Steam Pressure Auto Isolation Relay Standby Standby 3-13A-K050 RCIC Low Steam Pressure Auto Isolation Relay Standby Standby 3-13A-K053 RCIC Reactor Hi Vessel Water Level Trip
 
Relay Auxiliary Standby Standby 3-13A-K054 RCIC Auto Isolation Signal Relay Standby Standby  3AC043 Emergency Shutdown Panel Standby Standby  3AC065 Rx Vessel Lvl and Pressure Inst Rack A Energized Energized  3AC091 Jet Pump Inst Rack A Energized Energized  3AC270 Panel Energized Standby  3AD001 3A 125V DC Battery Energized Energized  3AD003 Station Battery Charger 3A Energized Energized  3AD017 (3FBA) Battery Main Fuse Box Energized Energized  3AD018 (3DPA) 250v DC Distribution Panel Energized Energized  3AD019 (3FA) 250 Volt Fuse Box Energized Energized  3AD025 (3PPAD) 3ppad 125V DC Distribution Panel Energized Energized  3AE024 Residual Heat Exchanger Standby Standby  3AS377 Back-Up N2 Supply to Ads RV's Standby Open  3AS456 Transfer Switch Energized Energized  3AT540 Instrument N2 Accumulator Standby Standby  3AT545 3A Srv Inst N2 Accumulator Standby Standby  3BC043 Panel Standby Standby Contains control switch for LT-3-02-3-085A 3BC065 Instrument Rack Standby Standby  3BC091 Instrument Rack Standby Standby  3BC270 Panel Energized Standby  3BD001 2B 125V DC Battery Standby Energized Provides power for 120V AC Vital Instrument Power 3BD017 Battery Main Fuse Box Energized Energized  3BD018 250V DC Distribution Panel Div. II Energized Energized  3BS377 Back-Up N2 Supply to Ads RV's Standby Open  3BS545 Automatic Transfer Switch Panel Energized Energized  3BT540 Instrument N2 Accumulator Standby Standby  3BT545 3B Srv Inst N2 Accumulator Standby Standby  3CD001 3C 125V DC Battery Energized Energized  3CD003 Station Battery Charger 3C Energized Energized  3CD017 (3FBC) Battery Main Fuse Box Energized Energized  3CD019 (3FC) 250 Volt Fuse Box Energized Energized  3CD025 (3PPCD) 3PPCD 125V DC Distribution Panel Energized Energized This panel is not credited as a power source for any item on this ESEL, it remains on the list for conservatism 3CE024 Residual Heat Exchanger Standby Standby  3CS377 Back-Up N2 Supply to Ads RV's Standby Open  3CS456 Control to Battery Charger 3CD003 Transfer Switch Energized Energized  3CT540 Instrument N2 Accumulator Standby Standby  3CT545 3C Srv Inst N2 Accumulator Standby Standby  3DA-W-A (1201) RCIC MO-3-13-021 Breaker Energized Energized  3DA-W-A (1203) RCIC MO-3-13-030 Breaker Energized Energized  3DA-W-A (1204) RCIC MO-3-13-027 Breaker Energized Energized  3DA-W-A (1205) RCIC MO-3-13-041 Breaker Energized Energized  3DA-W-A (1206) RCIC MO-3-13-039 Breaker Energized Energized  3DA-W-A (1207) RCIC MO-3-13-132 Breaker Energized Energized 14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 45 of 63 Equipment ID Description Equipment Normal State Equipment Desired State Notes 3DA-W-A (1209) RCIC MO-3-13-131 Breaker Energized Energized  3DA-W-A (1210) RCIC MO-3-13-018 Breaker Energized Energized  3DA-W-A (1214) RCIC Cond Vac PP 30P046 Breaker Energized Energized  3DA-W-A (1215) RCIC Vac Tank Cond PP 30P048 Breaker Energized Energized  3DA-W-A (30D012) RCIC 250VDC MCC Energized Energized  3DD001 2D 125V DC Battery Standby Energized Provides power for 120V AC Vital Instrument Power and RCIC B Logic 3DD017 Battery Main Fuse Box Energized Energized  3DD019 Fuse Box Energized Energized  3GT540 Instrument N2 Accumulator Standby Standby  3GT545 3G Srv Inst N2 Accumulator Standby Standby  3KT540 Instrument N2 Accumulator Standby Standby  3KT545 3K Srv Inst N2 Accumulator Standby Standby  3OE032 RCIC Barometric Condenser (13-2) Standby Operating  3OE104 RCIC Turb. Lube Oil Cooler (13-2)
Standby Operating  3AC270 (K3A) Relay De-Energized De-Energized  3BC270 (K3B) Relay De-Energized De-Energized  3AC270 (K3C) Relay De-Energized De-Energized  3BC270 (K3D) Relay De-Energized De-Energized  AO-3-07B-3511 Torus 18 Inch Vent Inboard Isol Valve to
 
Sbgt/Atmos Closed Open AO-3-07B-90290 Ctmt Emerg Vent Outboard Isolation Vlv to
 
Atmos Closed Open E134 (1014) E134-T-B 30B059 Breaker Energized Energized E134 (30B010) E134 Load Center Energized Energized  E134-T-B (30B059) E134-T-B Motor Control Center Energized Energized  E134-T-B (5924) Alt Feed For Uninterrupt AC Power Supp Inverter 30D37 Energized Energized  E134-T-B (5931) 125V. D.C. Battery Charger 3AD03 Energized Energized  E334 (1213) E334-R-B 30B038 Breaker Energized Energized E334 (30B012) E334 Load Center Energized Energized E334-R-B (30B038) MCC 30B038 Energized Energized E334-R-B (3821) MO-3-10-038A Breaker Energized Energized E334-R-B (3824) MO-3-10-026A Breaker Energized Energized E334-R-B (3831) MO-3-10-039A Breaker Energized Energized E334-R-B (3832) MO-3-10-034A Breaker Energized Energized E334-R-B (3844) MO-3-10-031A Breaker Energized Energized E334-R-B (3851) MO-3-10-25A Alt Breaker Energized Energized E334-R-B (3862) MO-3-10-174 Breaker Energized Energized E334-R-B (3863) MO-3-10-176 Breaker De-Energized Energized E334-R-B (3882) Norm Fdr for 120V Instr Pnl 30Y35 Trans 30X135 Energized Energized E334-R-B (3893) 125V Battery Charger C 3C Transfer Switch 30S704 Energized Energized  INV-3-13-90 RCICs-125 VDC Bus 'A' Power Distribution Energized Energized  J2919 J-Box at E134 LC Standby Energized  J2920 J-Box at E334 LC Standby Energized LI-3-02 085A Reactor Vessel High Water Energized Energized  LI-3-02-3-113 Reactor Water Level Energized Energized
 
14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 46 of 63 Equipment ID Description Equipment Normal State Equipment Desired State Notes LR/TR-9123A Torus Water Level/Temperature Recorder Energized Energized  LT-3-02-3-072A Reactor Vessel Water Level Energized Energized  LT-3-02-3-113 Level Transmitter Energized Energized  LT-9123A Torus Water Level Energized Energized  MO-3 025A RHR Inner Injection Valve to Recirc Loop A Closed Open Valve closure via control switch requires core spray relay logic permissive MO-3 026A RHR Loop A D/W Spray O/B Closed Open/Closed  MO-3 031A RHR Loop A D/W Spray I/B Closed Open/Closed  MO-3 034A RHR Loop A FFT Valve Closed Open/Closed  MO-3-10-038A RHR Loop A Torus Spray Closed Open/Closed  MO-3-10-039A RHR Loop A Torus Valve Closed Open  MO-3-10-174 HPSW to RHR Inner X-Tie Closed Open/Closed  MO-3-10-176 HPSW to RHR Outer X-Tie Closed Open/Closed  MO-3-13-018 RCIC Pump Suction from Condensate Storage Tank Open Open/Closed MO-3-13-021 RCIC Discharge to Feedwater Line B Closed Open  MO-3-13-027 RCIC Minimum Flow Valve Closed Open/Closed  MO-3-13-030 RCIC Full Flow Test Valve Closed Open/Closed  MO-3-13-039 RCIC Pump Torus Suction Outer Closed Open/Closed  MO-3-13-041 RCIC Pump Torus Suction Inner Closed Open/Closed  MO-3-13-131 RCIC Turbine Steam Supply Valve Closed Open  MO-3-13-132 RCIC Cooling Water Supply to Lo Clr +
Barometric Cdsr Closed Open N310025A Cabinet Energized Energized  PI-3-06-090A Reactor Wide Range Press Ind Energized Energized  PI-3-06-090B Reactor Wide Range Press Ind Energized Energized  PI-3-06-090C Reactor Wide Range Press Ind Energized Energized  PI-3-13-094 RCIC Pump Turb Stm Press Energized Energized  PR/LR-3 096 Reactor Level/Steam Flow Ratio Energized Energized  PR/TR-5805 Containment Pressure/Temp Energized Energized  PS-3-13-67-1 Pressure Switch Open Open  PS-3-13-72A Pressure Switch Open Open  PS-3-13-72B Pressure Switch Open Open  PS-3-13-87A Pressure Switch Open Open  PS-3-13-87B Pressure Switch Open Open  PT-3-02 404A Reactor Pressure Energized Energized PT-3-02 404C Reactor Pressure Transmitter Energized Energized  PT-3-06-053A Reactor Wide Range -Pressure Energized Energized PT-3-06-053B Reactor Wide Range -Pressure Energized Energized PT-3-06-053C Reactor Wide Range -Pressure Energized Energized PT-3-13-068 RCIC Turbine Steam Supply Pressure Energized Energized  PT-5805 Drywell Pressure Energized Energized  RV-3-02-071A 3A Safety Relief Valve Closed Open/Closed  RV-3-02-071B 3B Safety Relief Valve Closed Open/Closed  RV-3-02-071C 3C Safety Relief Valve Closed Open/Closed  RV-3-02-071G 3G Safety Relief Valve Closed Open/Closed  RV-3-02-071K 3K Safety Relief Valve Closed Open/Closed  XAM-3-02 117A Reactor Water Level Wide Range Energized Energized 14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 47 of 63
 
Attachment C ESEP HCLPF Values and Failure Modes Tabulation, Unit 2 and Common
 
14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 48 of 63
 
Equipment ID Description Failure Mode HCLPF (g) Basis 00C133 Panel Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 00E068 CR Fresh Air Supply Preheat Coil Equipment Capacity >0.24 Component is Rule Of Box to 0AF041. Parent component is evaluated in 14Q4233-CAL-002. 00E072 CR Vent Reheat Coil Screened >RLGM Component screened by SRT judgment. 00F039 CR Fresh Air Supply Roll Filter Equipment Capacity >0.24 Component is Rule Of Box to 0AF041. Parent component is evaluated in 14Q4233-CAL-002. 00F043 OA Damper Emergency Switchgear & Battery Room Vent Supply Roll Filter Screened >RLGM Component screened by SRT judgment. 00T116 CAD Liquid Nitrogen Storage Tank Anchorage 0.276 Component evaluated by HCLPF calculation 14Q4233-CAL-002.
0AE073 OA Damper Emergency Switchgear & Battery Room Vent Supply Heat Coil Equipment Capacity >0.24 Component is Rule Of Box to 0AV034. Parent component is evaluated in 14Q4233-CAL-002. 0AF041 A Train HEPA Filter Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 0AF042 A Train Charcoal Filter Equipment Capacity >0.24 Component is Rule Of Box to 0AF041. Parent component is evaluated in 14Q4233-CAL-002. 0AF050 A Train HEPA Filter Equipment Capacity >0.24 Component is Rule Of Box to 0AF041. Parent component is evaluated in 14Q4233-CAL-002.
0AV030 CR Room Emergency Vent Fan Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002.
0AV034 Emergency Switch Gear and Battery Room Supply Fan Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 0AV036 Battery Room Exhaust Fan Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 20C003 Reactor and Containment
 
Cooling and Isolation Panel Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 20C004C RCIC Control Panel Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 20C005A Reactor Manual Control
 
Panel Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 20C012 Plant Services Console Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 20C018 Panel Equipment Capacity 0.44 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 20C019 Panel Equipment Capacity 0.44 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 20C032 Panel Equipment Capacity 0.44 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 20C033 Panel Equipment Capacity 0.44 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 20C034 RCIC Relay Panel Equipment Capacity 0.44 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 20C035 Panel Equipment Capacity 0.44 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 20C041 Panel Equipment Capacity 0.44 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 20C095 RCIC Instrument Rack Anchorage 0.41 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 20C144 Panel Equipment Capacity 0.44 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 20C722A Accident Monitoring
 
Instrumentation Panel Equipment Capacity 0.44 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 20C722B Panel Equipment Capacity 0.44 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 20C818 Reactor Water Level/Pressure Component Cabinet Equipment Capacity 0.44 Component evaluated by HCLPF calculation 14Q4233-CAL-002.
14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 49 of 63
 
Equipment ID Description Failure Mode HCLPF (g) Basis 20D021 (2PPA) 125V DC Station Distribution Anchorage >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 20D023 (2PPC) 125V DC Station
 
Distribution Anchorage >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 20D024 Distribution Panel Anchorage 0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 20D037 Uninterruptable Power Supply Static Inverter Equipment Capacity 0.44 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 20D039 RCIC Barometric CDSR
 
Vacuum Pump Starter Screened >RLGM Component screened by SRT analysis. 20D040 RCIC Barometric CDSR Cond Pump Starter Screened >RLGM Component screened by SRT analysis. 20P036 RCIC Pump Screened >RLGM Component screened by SRT analysis. 20P046 RCIC Barometric CDSR
 
Vacuum Pump Anchorage 0.36 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 20P048 RCIC Barometric CDSR Condensate Pump Anchorage 0.36 Component is Rule Of Box to 20P046. Parent component is evaluated in 14Q4233-CAL-002. 20P340 RCIC Turbine Driven Lube Oil Pump Screened >RLGM Component is Rule Of Box to 20P036. Parent component screens. 20S038 RCIC Turbine Screened >RLGM Component is Rule Of Box to 20P036. Parent component screens. 20S315 Static Inverter Man Bypass/Isolation Switch Screened >RLGM Component screened by SRT analysis. 20S354 Load Center E-124/E324 Transfer Switch Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 20S700 Battery Charger Panel 2AD003 Transfer Switch 20S700 Anchorage >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 20S701 Transfer Switch Anchorage >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 20S703 Transfer Switch Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 20X133 Transformer Block wall interaction 0.27 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 20X135 20Y035 Transformer Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 20X150 Transformer Block wall interaction 0.27 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 20Y033 Distribution Panel Block wall interaction 0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 20Y035 120 VAC 'Y' Power Panel Screened >RLGM Component screened by SRT analysis. 20Y050 Uninterruptable Power Supply Distribution Panel Block wall interaction 0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 2-13A-K004 RCIC Hi Temp Steam Leak Relay Functional Failure 0.34 Component evaluated by HCLPF calculation 14Q4233-CAL-004. 2-13A-K006 RCIC Hi Temp Steam Leak Relay Functional Failure 0.34 Component evaluated by HCLPF calculation 14Q4233-CAL-004. 2-13A-K007 RCIC Steam Line Hi DP Line Break Relay Functional Failure 0.29 Component evaluated by HCLPF calculation 14Q4233-CAL-004. 2-13A-K010 RCIC Turbine Trip Aux Relay Functional Failure 0.14 Component evaluated by HCLPF calculation 14Q4233-CAL-004. (Resolved by Operator Action per TODI PB 1570792-76) 2-13A-K011 RCIC Turbine Trip Aux Relay Functional Failure 0.14 Component evaluated by HCLPF calculation 14Q4233-CAL-004. (Resolved by Operator Action per TODI PB 1570792-76) 2-13A-K012 RCIC Auto Isolation Signal Relay Functional Failure 0.36 Component evaluated by HCLPF calculation 14Q4233-CAL-004. 2-13A-K014 RCIC Pump Lo Suction Pressure Trip Relay Functional Failure 0.34 Component evaluated by HCLPF calculation 14Q4233-CAL-004. 2-13A-K017 RCIC Turbine Exh Hi Pressure Trip Relay Functional Failure 0.34 Component evaluated by HCLPF calculation 14Q4233-CAL-004.
14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 50 of 63
 
Equipment ID Description Failure Mode HCLPF (g) Basis 2-13A-K-022 RCIC Auto Isolation Signal Relay Functional Failure 0.36 Component evaluated by HCLPF calculation 14Q4233-CAL-004. 2-13A-K030 Reactor Hi Vessel Water Level Trip Relay Functional Failure 0.34 Component evaluated by HCLPF calculation 14Q4233-CAL-004. 2-13A-K031 RCIC Steam Line Space Hi Temp Isolation Relay Functional Failure 0.34 Component evaluated by HCLPF calculation 14Q4233-CAL-004. 2-13A-K032 RCIC Steam Line Space Hi Temp Isolation Relay Functional Failure 0.34 Component evaluated by HCLPF calculation 14Q4233-CAL-004. 2-13A-K033 RCIC Steam Line Hi DP Line Break Relay Functional Failure 0.20 Component evaluated by HCLPF calculation 14Q4233-CAL-004. (Modification/Resolution Required) 2-13A-K034 RCIC Auto Isolation Signal Relay Functional Failure 0.36 Component evaluated by HCLPF calculation 14Q4233-CAL-004. 2-13A-K044 RCIC Reactor Hi Vessel Water Level Trip Relay Functional Failure 0.29 Component evaluated by HCLPF calculation 14Q4233-CAL-004. 2-13A-K049 RCIC Low Steam Pressure
 
Auto Isolation Relay Functional Failure 0.36 Component evaluated by HCLPF calculation 14Q4233-CAL-004. 2-13A-K050 RCIC Low Steam Pressure
 
Auto Isolation Relay Functional Failure 0.36 Component evaluated by HCLPF calculation 14Q4233-CAL-004. 2-13A-K053 RCIC Reactor Hi Vessel Water Level Trip Relay Auxiliary Functional Failure 0.34 Component evaluated by HCLPF calculation 14Q4233-CAL-004. 2-13A-K054 RCIC Auto Isolation Signal Relay Functional Failure 0.36 Component evaluated by HCLPF calculation 14Q4233-CAL-004. 2AC043 Emergency Shutdown Panel Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 2AC065 Rx Vessel Lvl and Pressure
 
Inst Rack A Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 2AC091 Jet Pump Inst Rack A Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 2AC270 Panel Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 2AD001 2A 125V DC Battery Anchorage 0.26 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 2AD003 Station Battery Charger 2A Block wall interaction 0.27 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 2AD017 (2FPA) Battery Main Fuse Box Anchorage >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 2AD018 (2DPA) 250V DC Distribution Panel Anchorage >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 2AD019 (2FA) 250 Volt Fuse Box Anchorage >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 2AD025 Distribution Panel Anchorage >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 2AE024 Residual Heat Exchangers Anchorage 0.26 Component evaluated by HCLPF calculation 14Q4233-CAL-002.
2AS377 Back-Up N2 Supply to Ads
 
RV's Screened >RLGM Component screened by SRT analysis. 2AT545 2A Srv Inst N2 Accumulator Screened >RLGM Component screened by SRT judgment. 2BC043 Panel Equipment Capacity >0.24 Component is Rule Of Box to 2AC043. Parent component is evaluated in 14Q4233-CAL-002. 2BC065 Rx Vessel Lvl and Pressure Inst Rack B Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 2BC172 Panel Equipment Capacity 0.44 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 2BC270 Panel Equipment Capacity >0.24 Component is Rule Of Box to 2AC270. Parent component is evaluated in 14Q4233-CAL-002. 2BD001 2B 125V DC Battery Anchorage 0.26 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 2BD017 Battery Main Fuse Box Anchorage >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002.
14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 51 of 63
 
Equipment ID Description Failure Mode HCLPF (g) Basis 2BD018 250V DC Distribution Panel Div. II Anchorage >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 2BE024 Residual Heat Exchangers Anchorage 0.26 Component evaluated by HCLPF calculation 14Q4233-CAL-002.
2BS377 Back-Up N2 Supply to Ads
 
RV's Screened >RLGM Component is Rule Of Box to 2AS377. Parent component screens.
2BS545 Automatic Transfer Switch
 
Panel Anchorage >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 2BT545 2B Srv Inst N2 Accumulator Screened >RLGM Component screened by SRT judgment. 2CC133 Panel Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 2CD001 2C 125V DC Battery Anchorage 0.26 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 2CD003 Station Battery Charger 2C Block wall interaction 0.27 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 2CD017 (2FPC) Battery Main Fuse Box Block wall interaction 0.27 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 2CD019 (2FC) 250 Volt Fuse Box Anchorage >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 2CE024 Residual Heat Exchangers Anchorage 0.26 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 2CS377 Back-Up N2 Supply to Ads
 
RV's Screened >RLGM Component is Rule Of Box to 2AS377. Parent component screens. 2CS545 Automatic Transfer Switch
 
Panel Anchorage >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 2CT545 2C Srv Inst N2 Accumulator Screened >RLGM Component screened by SRT judgment. 2DA-W-A (1201) RCIC MO-2-13-021 Breaker Anchorage 0.43 Component is Rule Of Box to 2DA-W-A (20D012).
Parent component is evaluated in 14Q4233-CAL-002. 2DA-W-A (1203) RCIC MO-2-13-030 Breaker Anchorage 0.43 Component is Rule Of Box to 2DA-W-A (20D012).
Parent component is evaluated in 14Q4233-CAL-002. 2DA-W-A (1204) RCIC MO-2-13-027 Breaker Anchorage 0.43 Component is Rule Of Box to 2DA-W-A (20D012).
Parent component is evaluated in 14Q4233-CAL-002. 2DA-W-A (1205) RCIC MO-2-13-041 Breaker Anchorage 0.43 Component is Rule Of Box to 2DA-W-A (20D012).
Parent component is evaluated in 14Q4233-CAL-002. 2DA-W-A (1206) RCIC MO-2-13-039 Breaker Anchorage 0.43 Component is Rule Of Box to 2DA-W-A (20D012).
Parent component is evaluated in 14Q4233-CAL-002. 2DA-W-A (1207) RCIC MO-2-13-132 Breaker Anchorage 0.43 Component is Rule Of Box to 2DA-W-A (20D012).
Parent component is evaluated in 14Q4233-CAL-002. 2DA-W-A (1209) RCIC MO-2-13-131 Breaker Anchorage 0.43 Component is Rule Of Box to 2DA-W-A (20D012).
Parent component is evaluated in 14Q4233-CAL-002. 2DA-W-A (1210) RCIC MO-2-13-018 Breaker Anchorage 0.43 Component is Rule Of Box to 2DA-W-A (20D012). Parent component is evaluated in 14Q4233-CAL-002. 2DA-W-A (1214) RCIC Cond Vac PP 20P046 Breaker Anchorage 0.43 Component is Rule Of Box to 2DA-W-A (20D012).
Parent component is evaluated in 14Q4233-CAL-002. 2DA-W-A (1215) RCIC Vac Tank Cond PP 20P048 Breaker Anchorage 0.43 Component is Rule Of Box to 2DA-W-A (20D012).
Parent component is evaluated in 14Q4233-CAL-002.
2DA-W-A (20D012) RCIC 250VDC MCC Anchorage 0.43 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 2DD001 2D 125V DC Battery Anchorage 0.26 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 2DD017 Battery Main Fuse Box Anchorage >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 2DD019 Fuse Box Anchorage >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 2DE024 Residual Heat Exchangers Anchorage 0.26 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 2GT545 2G Srv Inst N2 Accumulator Screened >RLGM Component screened by SRT judgment. 2KT545 2K Srv Inst N2 Accumulator Screened >RLGM Component screened by SRT judgment.
14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 52 of 63
 
Equipment ID Description Failure Mode HCLPF (g) Basis 2OE032 RCIC Barometric Condenser (13-2)
Equipment Capacity 0.44 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 2OE104 RCIC Turb. Lube Oil Cooler (13-2) Screened >RLGM Component is Rule Of Box to 20P036. Parent component screens. 2AC270 (K3A) Relay Equipment Capacity >0.24 Component is Rule Of Box to 2AC270. Parent component is evaluated in 14Q4233-CAL-002. Component evaluated by HCLPF calculation 14Q4233-CAL-004. 2BC270 (K3B) Relay Equipment Capacity >0.24 Component is Rule Of Box to 2BC270. Parent component is evaluated in 14Q4233-CAL-002. Component evaluated by HCLPF calculation 14Q4233-CAL-004. 2AC270 (K3C) Relay Equipment Capacity >0.24 Component is Rule Of Box to 2AC270. Parent component is evaluated in 14Q4233-CAL-002. Component evaluated by HCLPF calculation 14Q4233-CAL-004. 2BC270 (K3D) Relay Equipment Capacity >0.24 Component is Rule Of Box to 2BC270. Parent component is evaluated in 14Q4233-CAL-002. Component evaluated by HCLPF calculation 14Q4233-CAL-004. AO-2-07B-2511 Torus 18 Inch Vent Inboard
 
Isol Valve to Sbgt/Atmos Screened >RLGM Component screened by SRT judgment. AO-2-07B-80290 Ctmt Emerg Vent Outboard Isolation Vlv to Atmos Screened >RLGM Component screened by SRT judgment. E124 (1013) E124-R-C 20B036 Breaker Equipment Capacity >0.24 Component is Rule Of Box to E124 (20B010). Parent component is evaluated in 14Q4233-CAL-002. E124 (1014) E124-T-B 20B059 Breaker Equipment Capacity >0.24 Component is Rule Of Box to E124 (20B010). Parent component is evaluated in 14Q4233-CAL-002. E124 (20B010) E124 Load Center Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. E124-R-C (20B036) E124-R-C Motor Control Center Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. E124-R-C (3606) MO-2-10-25A Norm Breaker Equipment Capacity >0.24 Component is Rule Of Box to E124-R-C (20B036).
Parent component is evaluated in 14Q4233-CAL-002. E124-R-C (3691) Alt Feed Breaker 20D037 Y50 Equipment Capacity >0.24 Component is Rule Of Box to E124-R-C (20B036).
Parent component is evaluated in 14Q4233-CAL-002. E124-T-B (20B059) E124-T-B Motor Control Center Block wall interaction 0.27 Component evaluated by HCLPF calculation 14Q4233-CAL-002. E124-T-B (5931) Norm Fdr for 125 V Battery Charger 'A' 2AD03 Block wall interaction 0.27 Component is Rule Of Box to E124-T-B (20B059).
Parent component is evaluated in 14Q4233-CAL-002. E324 (1213) E324-R-B 20B038 Breaker Equipment Capacity >0.24 Component is Rule Of Box to E324 (20B012). Parent component is evaluated in 14Q4233-CAL-002. E324 (1222) E324-T-B 00B049 Breaker Equipment Capacity >0.24 Component is Rule Of Box to E324 (20B012). Parent component is evaluated in 14Q4233-CAL-002. E324 (20B012) E324 Load Center Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002.
E324-R-B (20B038) MCC 20B038 Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. E324-R-B (3821) MO-2-10-038A Breaker Equipment Capacity >0.24 Component is Rule Of Box to E324-R-B (20B038).
Parent component is evaluated in 14Q4233-CAL-002. E324-R-B (3822) MO-2-10-020 Breaker Equipment Capacity >0.24 Component is Rule Of Box to E324-R-B (20B038).
Parent component is evaluated in 14Q4233-CAL-002. E324-R-B (3824) MO-2-10-026A Breaker Equipment Capacity >0.24 Component is Rule Of Box to E324-R-B (20B038).
Parent component is evaluated in 14Q4233-CAL-002. E324-R-B (3831) MO-2-10-039A Breaker Equipment Capacity >0.24 Component is Rule Of Box to E324-R-B (20B038).
Parent component is evaluated in 14Q4233-CAL-002. E324-R-B (3832) MO-2-10-034A Breaker Equipment Capacity >0.24 Component is Rule Of Box to E324-R-B (20B038).
Parent component is evaluated in 14Q4233-CAL-002. E324-R-B (3844) MO-2-10-031A Breaker Equipment Capacity >0.24 Component is Rule Of Box to E324-R-B (20B038).
Parent component is evaluated in 14Q4233-CAL-002. E324-R-B (3862) MO-2-10-174 Breaker Equipment Capacity >0.24 Component is Rule Of Box to E324-R-B (20B038).
Parent component is evaluated in 14Q4233-CAL-002.
14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 53 of 63
 
Equipment ID Description Failure Mode HCLPF (g) Basis E324-R-B (3863) MO-2-10-176 Breaker Equipment Capacity >0.24 Component is Rule Of Box to E324-R-B (20B038). Parent component is evaluated in 14Q4233-CAL-002. E324-R-B (3882) Norm Fdr for 120V Instr Pnl 20Y35 Trans 20X135 Equipment Capacity >0.24 Component is Rule Of Box to E324-R-B (20B038).
Parent component is evaluated in 14Q4233-CAL-002. E324-R-B (3893) 125 VDC Batt Charger 2CD03 Equipment Capacity >0.24 Component is Rule Of Box to E324-R-B (20B038).
Parent component is evaluated in 14Q4233-CAL-002. E324-T-B (00B049) MCC 00B049 for 0AV034, 0AV036, and 0AV030 Block wall interaction 0.27 Component evaluated by HCLPF calculation 14Q4233-CAL-002. INV-2-13-90 RCICs-125 VDC Bus 'A' Power Distribution Equipment Capacity 0.44 Component is Rule Of Box to 20C019. Parent component is evaluated in 14Q4233-CAL-002. J-1648 Junction Box Screened >RLGM Component screened by SRT analysis. J2915 J-Box at E124 Load Center Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. J2916 J-Box at E324 Load Center Screened >RLGM Component screened by SRT analysis. LI-2-02-3-085A Reactor Vessel High Water Equipment Capacity >0.24 Component is Rule Of Box to 20C005A. Parent component is evaluated in 14Q4233-CAL-002. LI-2-02-3-113 Reactor Water Level Equipment Capacity >0.24 Component is Rule Of Box to 20C003. Parent component is evaluated in 14Q4233-CAL-002. LR/TR-8123A Torus Water Level/Temperature Recorder Equipment Capacity >0.24 Component is Rule Of Box to 20C004C. Parent component is evaluated in 14Q4233-CAL-002. LT-2-02-3-072A Reactor Vessel Water Level Equipment Capacity >0.24 Component is Rule Of Box to 2AC065. Parent component is evaluated in 14Q4233-CAL-002. LT-2-02-3-113 Reactor Press Vessel Fuel Zone Wtr Level Equipment Capacity >0.24 Component is Rule Of Box to 2AC091. Parent component is evaluated in 14Q4233-CAL-002. LT-8123A Torus Water Level Screened >RLGM Component screened by SRT analysis. MO-2-10-020 RHR Loops A/B X-Tie Screened >RLGM Component screened by SRT analysis. MO-2-10-026A RHR Loop A D/W Spray O/B Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. MO-2-10-031A RHR Loop A D/W Spray I/B Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. MO-2-10-034A RHR Loop A FFT Valve Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. MO-2-10-038A RHR Loop A Torus Spray Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. MO-2-10-039A RHR Loop A Torus Valve Screened >RLGM Component screened by SRT analysis. MO-2-10-174 HPSW to RHR Inner X-Tie Screened >RLGM Component screened by SRT analysis. MO-2-10-176 HPSW to RHR Outer X-Tie Screened >RLGM Component screened by SRT analysis. MO-2-10-25A RHR Loop A I/B Disc Valve Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. MO-2-13-018 RCIC Pump Suction from Condensate Storage Tank Screened >RLGM Component screened by SRT analysis. MO-2-13-021 RCIC Discharge to Feedwater Line B Screened >RLGM Component screened by SRT analysis. MO-2-13-027 RCIC Minimum Flow Valve Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. MO-2-13-030 RCIC Full Flow Test Valve Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. MO-2-13-039 RCIC Pump Torus Suction Outer Screened >RLGM Component screened by SRT analysis. MO-2-13-041 RCIC Pump Torus Suction Inner Screened >RLGM Component screened by SRT analysis. MO-2-13-131 RCIC Turbine Steam Supply
 
Valve Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. MO-2-13-132 RCIC Cooling Water Supply to Lo Clr + Barometric Cdsr Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002.
14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 54 of 63
 
Equipment ID Description Failure Mode HCLPF (g) Basis N210025A Cabinet Provides Power to Valve MO-2-10-25A Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. OAS384 N2 Tank Screened >RLGM Component screened by SRT analysis. OAS385 N2 Tank Screened >RLGM Component is Rule Of Box to OBS385. Parent component screens. OBS384 N2 Tank Screened >RLGM Component is Rule Of Box to OAS384. Parent component screens. OBS385 N2 Tank Screened >RLGM Component screened by SRT analysis. PCV-0-40W-70088A N2 Regulator - OAS384 Screened >RLGM Component is Rule Of Box to OAS384. Parent component screens. PCV-0-40W-70088B N2 Regulator - OBS384 Screened >RLGM Component is Rule Of Box to OBS384. Parent component screens. PCV-0-40W-70089A N2 Regulator - OAS385 Screened >RLGM Component is Rule Of Box to OAS385. Parent component screens. PCV-0-40W-70089B N2 Regulator - OBS385 Screened >RLGM Component is Rule Of Box to OBS385. Parent component screens. PI-2-06-090A Reactor Wide Range Press
 
Ind Equipment Capacity >0.24 Component is Rule Of Box to 20C005A. Parent component is evaluated in 14Q4233-CAL-002. PI-2-06-090B Reactor Wide Range Press
 
Ind Equipment Capacity >0.24 Component is Rule Of Box to 20C005A. Parent component is evaluated in 14Q4233-CAL-002. PI-2-06-090C Reactor Wide Range Press
 
Ind Equipment Capacity >0.24 Component is Rule Of Box to 20C005A. Parent component is evaluated in 14Q4233-CAL-002. PI-2-13-094 RCIC Pump Turb Stm Press Equipment Capacity >0.24 Component is Rule Of Box to 20C004C. Parent component is evaluated in 14Q4233-CAL-002. PO-0-40D-00153-01 CR Emergency Vent Filters
 
Inlet Damper Equipment Capacity >0.24 Component is Rule Of Box to 0AF041. Parent component is evaluated in 14Q4233-CAL-002. PO-0-40D-00153-02 CR Emergency Vent Filters
 
Inlet Damper Equipment Capacity >0.24 Component is Rule Of Box to 0AF041. Parent component is evaluated in 14Q4233-CAL-002. PO-0-40D-00157-01 CR Emergency Vent Supply Fan Inlet Damper Equipment Capacity >0.24 Component is Rule Of Box to 0AF041. Parent component is evaluated in 14Q4233-CAL-002. PO-0-40D-00157-02 CR Emergency Vent Supply Fan Outlet Damper Screened >RLGM Component screened by SRT judgment. PO-0-40W-00016 OA Damper Emergency Switchgear And Battery Room Vent Supply Fans Screened >RLGM Component is Rule Of Box to 00F043. Parent component is screened. PO-0-40W-00019-01 OA Damper Emergency Switchgear And Battery Room Vent Supply Damper Equipment Capacity >0.24 Component is Rule Of Box to 0AV034. Parent component is evaluated in 14Q4233-CAL-002. PO-0-40W-00019-02 OA Damper Emergency Switchgear And Battery Room Vent Outlet Damper Screened >RLGM Component is Rule Of Box to 00F043. Parent component is screened. PO-0-40W-00021-01 OA Damper Emergency Switchgear And Battery Room Vent Supply Damper Equipment Capacity >0.24 Component is Rule Of Box to 0AV034. Parent component is evaluated in 14Q4233-CAL-002. PO-0-40W-00031-01 Battery Room Exhaust Fan
 
Inlet Damper Screened >RLGM Component screened by SRT judgment. PO-0-40W-00031-02 Battery Room Exhaust Fan Outlet Damper Screened >RLGM Component screened by SRT judgment. PO-0-40W-00782-01 OA Damper Emergency Switchgear and Battery Room Vent Supply Damper Equipment Capacity >0.24 Component is Rule Of Box to 0AV034. Parent component is evaluated in 14Q4233-CAL-002. PO-0-40W-00782-02 OA Damper Emergency Switchgear and Battery Room Vent Supply Damper Equipment Capacity >0.24 Component is Rule Of Box to 0AV034. Parent component is evaluated in 14Q4233-CAL-002. PO-0-40W-00808 OA Damper Emergency Switchgear and Battery Room Vent Outlet Damper Screened >RLGM Component screened by SRT judgment. PO-0-40W-00822-01 OA Damper Emergency Switchgear and Battery Room Vent Supply Damper Equipment Capacity >0.24 Component is Rule Of Box to 0AV034. Parent component is evaluated in 14Q4233-CAL-002.
14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 55 of 63
 
Equipment ID Description Failure Mode HCLPF (g) Basis PO-0-40W-00822-02 OA Damper Emergency Switchgear and Battery Room Vent Supply Damper Equipment Capacity >0.24 Component is Rule Of Box to 0AV034. Parent component is evaluated in 14Q4233-CAL-002. PO-0-40W-00822-03 OA Damper Emergency Switchgear and Battery Room Vent Supply Damper Equipment Capacity >0.24 Component is Rule Of Box to 0AV034. Parent component is evaluated in 14Q4233-CAL-002. PR/LR-2-06-096 Reactor Level/Steam Flow Ratio Equipment Capacity >0.24 Component is Rule Of Box to 20C005A. Parent component is evaluated in 14Q4233-CAL-002. PR/TR-4805 Containment Pressure/Temp Equipment Capacity >0.24 Component is Rule Of Box to 20C003. Parent component is evaluated in 14Q4233-CAL-002. PS-2-13-67-1 Pressure Switch Anchorage 0.41 Component is Rule Of Box to 20C095. Parent component is evaluated in 14Q4233-CAL-002.
Component evaluated by HCLPF calculation 14Q4233-CAL-004. PS-2-13-72A Pressure Switch Functional Failure 0.27 Component evaluated by HCLPF calculation 14Q4233-CAL-004. PS-2-13-72B Pressure Switch Functional Failure 0.27 Component evaluated by HCLPF calculation 14Q4233-CAL-004. PS-2-13-87A Pressure Switch Functional Failure 0.27 Component evaluated by HCLPF calculation 14Q4233-CAL-004. PS-2-13-87B Pressure Switch Functional Failure 0.27 Component evaluated by HCLPF calculation 14Q4233-CAL-004. PT-2-02-3-404A Reactor Pressure Equipment Capacity >0.24 Component is Rule Of Box to 2AC065. Parent component is evaluated in 14Q4233-CAL-002. PT-2-02-3-404C Reactor Pressure Transmitter Equipment Capacity >0.24 Component is Rule Of Box to 2AC091. Parent component is evaluated in 14Q4233-CAL-002. PT-2-06-053A Reactor Wide Range -
Pressure Equipment Capacity >0.24 Component is Rule Of Box to 2AC065. Parent component is evaluated in 14Q4233-CAL-002. PT-2-06-053B Reactor Wide Range -
Pressure Equipment Capacity >0.24 Component is Rule Of Box to 2BC065. Parent component is evaluated in 14Q4233-CAL-002. PT-2-06-053C Reactor Wide Range -
Pressure Equipment Capacity >0.24 Component is Rule Of Box to 2AC065. Parent component is evaluated in 14Q4233-CAL-002. PT-2-13-068 RCIC Turbine Steam Supply Press Anchorage 0.41 Component is Rule Of Box to 20C095. Parent component is evaluated in 14Q4233-CAL-002. PT-4805 Drywell Pressure Screened >RLGM Component screened by SRT analysis. RV-2-02-071A 2A Safety Relief Valve Screened >RLGM Component screened by SRT judgment. RV-2-02-071B 2B Safety Relief Valve Screened >RLGM Component screened by SRT judgment. RV-2-02-071C 2C Safety Relief Valve  Screened >RLGM Component screened by SRT judgment. RV-2-02-071G 2G Safety Relief Valve Screened >RLGM Component screened by SRT judgment. RV-2-02-071K 2K Safety Relief Valve Screened >RLGM Component screened by SRT judgment. SV-0-36B-00019 OA Damper Emergency Switchgear and Battery Room Damper IA Screened >RLGM Component is Rule Of Box to 00F043. Parent component is Screened. SV-0-36B-00031 Air Supply Shutoff for PO-0-40W-00031-01, PO-0-40W-00031-02 Screened >RLGM Component screened by SRT analysis. TI-2501 Ventilation Air Temperatures Equipment Capacity >0.24 Component is Rule Of Box to 20C012. Parent component is evaluated in 14Q4233-CAL-002. TI-80146 Drywell Bulk Average Temp
 
Indicator Equipment Capacity >0.24 Component is Rule Of Box to 20C012. Parent component is evaluated in 14Q4233-CAL-002. TT-2501 Vent Air/Wtr Temp Equipment Capacity 0.44 Component is Rule Of Box to 2BC172. Parent component is evaluated in 14Q4233-CAL-002. XAM-2-02-3-117A Reactor Water Level Wide Range Equipment Capacity 0.44 Component is Rule Of Box to 20C818. Parent component is evaluated in 14Q4233-CAL-002.
 
14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 56 of 63
 
Attachment D ESEP HCLPF Values and Failure Modes Tabulation, Unit 3
 
14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 57 of 63 Equipment ID Description Failure Mode HCLPF (g) Basis 30C003 Reactor and Containment Cooling and Isolation Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 30C004C RCIC Control Panel Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 30C005A Reactor Manual Control
 
Panel Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 30C012 Plant Services Console Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 30C018 Panel Equipment Capacity 0.44 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 30C019 Panel Equipment Capacity 0.44 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 30C032 Panel Equipment Capacity 0.44 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 30C033 Panel Equipment Capacity 0.44 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 30C034 RCIC Relay Panel Equipment Capacity 0.44 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 30C035 Panel Equipment Capacity 0.44 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 30C041 Panel Equipment Capacity 0.44 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 30C095 RCIC Instrument Rack Anchorage 0.41 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 30C144 Panel Equipment Capacity 0.44 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 30C722A Accident Monitoring
 
Instrumentation Panel Equipment Capacity 0.44 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 30C722B Panel Equipment Capacity 0.44 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 30C818 Reactor Water Level/Pressure Component Cabinet Equipment Capacity 0.44 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 30D021 (3PPA) 125V DC Station Distribution Anchorage 0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 30D023 (3PPC) 125V DC Station Distribution Anchorage >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 30D024 Distribution Panel Anchorage >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 30D037 Uninterruptable Power Supply Static Inverter Equipment Capacity 0.44 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 30D039 RCIC Barometric Cdsr Vacuum Pump Starter Screened >RLGM Component screened by SRT analysis. 30D040 RCIC Barometric Cdsr Cond
 
Pump Starter Screened >RLGM Component screened by SRT analysis. 30P036 RCIC Pump Anchorage >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 30P046 RCIC Barometric Cdsr
 
Vacuum Pump Anchorage 0.36 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 30P048 RCIC Barometric Cdsr Condensate Pump Anchorage 0.36 Component is Rule Of Box to 30P046. Parent component is evaluated in 14Q4233-CAL-002. 30P340 RCIC Turbine Driven Lube Oil Pump Anchorage >0.24 Component is Rule Of Box to 30P036. Parent component is evaluated in 14Q4233-CAL-002. 30S038 RCIC Turbine Anchorage >0.24 Component is Rule Of Box to 30P036. Parent component is evaluated in 14Q4233-CAL-002. 30S315 Static Inverter Man Bypass/Isolation Switch Screened >RLGM Component screened by SRT analysis. 30S356 Load Center E134/E334 Transfer Switch Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002.
14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 58 of 63 Equipment ID Description Failure Mode HCLPF (g) Basis 30S546 Transfer Switch Block wall interaction 0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 30S701 Transfer Switch Anchorage >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 30S703 Transfer Switch Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 30S704 125V DC Battery Charger 3CD003 Transfer Sw 30S704 Anchorage >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 30X133 Transformer Block wall interaction 0.27 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 30X135 30Y035 Transformer Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 30X150 30Y050 Transformer Block wall interaction 0.27 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 30Y033 Distribution Panel Block wall interaction 0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 30Y035 120 VAC 'Y' Power Panel Screened >RLGM Component screened by SRT analysis. 30Y050 120 VAC 'Y' Power Panel Block wall interaction 0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 3-13A-K004 RCIC Hi Temp Steam Leak Relay Functional Failure 0.34 Component evaluated by HCLPF calculation 14Q4233-CAL-004. 3-13A-K006 RCIC Hi Temp Steam Leak Relay Functional Failure 0.34 Component evaluated by HCLPF calculation 14Q4233-CAL-004. 3-13A-K007 RCIC Steam Line Hi DP Line Break Relay Functional Failure 0.29 Component evaluated by HCLPF calculation 14Q4233-CAL-004. 3-13A-K010 RCIC Turbine Trip Aux Relay Functional Failure 0.14 Component evaluated by HCLPF calculation 14Q4233-CAL-004. (Resolved by Operator Action per TODI PB 1570792-76) 3-13A-K011 RCIC Turbine Trip Aux Relay Functional Failure 0.14 Component evaluated by HCLPF calculation 14Q4233-CAL-004.
(Resolved by Operator Action per TODI PB 1570792-76) 3-13A-K012 RCIC Auto Isolation Signal Relay Functional Failure 0.29 Component evaluated by HCLPF calculation 14Q4233-CAL-004. 3-13A-K014 RCIC Pump Lo Suction Pressure Trip Relay Functional Failure 0.34 Component evaluated by HCLPF calculation 14Q4233-CAL-004. 3-13A-K017 RCIC Turbine Exh Hi Pressure Trip Relay Functional Failure 0.34 Component evaluated by HCLPF calculation 14Q4233-CAL-004. 3-13A-K-022 RCIC Auto Isolation Signal Relay Functional Failure 0.36 Component evaluated by HCLPF calculation 14Q4233-CAL-004. 3-13A-K030 Reactor Hi Vessel Water Level Trip Relay Functional Failure 0.34 Component evaluated by HCLPF calculation 14Q4233-CAL-004. 3-13A-K031 RCIC Steam Line Space Hi Temp Isolation Relay Functional Failure 0.34 Component evaluated by HCLPF calculation 14Q4233-CAL-004. 3-13A-K032 RCIC Steam Line Space Hi Temp Isolation Relay Functional Failure 0.34 Component evaluated by HCLPF calculation 14Q4233-CAL-004. 3-13A-K033 RCIC Steam Line Hi DP Line Break Relay Functional Failure 0.20 Component evaluated by HCLPF calculation 14Q4233-CAL-004. (Modification/Resolution Required) 3-13A-K034 RCIC Auto Isolation Signal Relay Functional Failure 0.36 Component evaluated by HCLPF calculation 14Q4233-CAL-004. 3-13A-K044 RCIC Reactor Hi Vessel Water Level Trip Relay Functional Failure 0.36 Component evaluated by HCLPF calculation 14Q4233-CAL-004. 3-13A-K049 RCIC Low Steam Pressure
 
Auto Isolation Relay Functional Failure 0.36 Component evaluated by HCLPF calculation 14Q4233-CAL-004. 3-13A-K050 RCIC Low Steam Pressure
 
Auto Isolation Relay Functional Failure 0.29 Component evaluated by HCLPF calculation 14Q4233-CAL-004. 3-13A-K053 RCIC Reactor Hi Vessel Water Level Trip Relay Auxiliary Functional Failure 0.34 Component evaluated by HCLPF calculation 14Q4233-CAL-004.
14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 59 of 63 Equipment ID Description Failure Mode HCLPF (g) Basis 3-13A-K054 RCIC Auto Isolation Signal Relay Functional Failure 0.29 Component evaluated by HCLPF calculation 14Q4233-CAL-004. 3AC043 Emergency Shutdown Panel Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 3AC065 Rx Vessel Lvl and Pressure
 
Inst Rack A Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 3AC091 Jet Pump Inst Rack A Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 3AC270 Panel Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 3AD001 3A 125V DC Battery Anchorage 0.26 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 3AD003 Station Battery Charger 3A Block wall interaction 0.27 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 3AD017 (3FBA) Battery Main Fuse Box Anchorage >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 3AD018 (3DPA) 250v DC Distribution Panel Anchorage >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 3AD019 (3FA) 250 Volt Fuse Box Anchorage >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 3AD025 (3PPAD) 3ppad 125V DC Distribution
 
Panel Anchorage >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 3AE024 Residual Heat Exchanger Anchorage 0.26 Component evaluated by HCLPF calculation 14Q4233-CAL-002.
3AS377 Back-Up N2 Supply to Ads
 
RV's Screened >RLGM Component screened by SRT analysis. 3AS456 Transfer Switch Anchorage >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 3AT540 Instrument N2 Accumulator Screened >RLGM Component screened by SRT judgment. 3AT545 3A Srv Inst N2 Accumulator Screened >RLGM Component screened by SRT judgment. 3BC043 Panel Equipment Capacity >0.24 Component is Rule Of Box to 3AC043. Parent component is evaluated in 14Q4233-CAL-002. 3BC065 Instrument Rack Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 3BC091 Instrument Rack Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 3BC270 Panel Equipment Capacity >0.24 Component is Rule Of Box to 3AC270. Parent component is evaluated in 14Q4233-CAL-002. 3BD001 2B 125V DC Battery Anchorage 0.26 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 3BD017 Battery Main Fuse Box Anchorage >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 3BD018 250V DC Distribution Panel
 
Div. II Anchorage >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002.
3BS377 Back-Up N2 Supply to Ads
 
RV's Screened >RLGM Component is Rule Of Box to 3AS377. Parent component screens.
3BS545 Automatic Transfer Switch
 
Panel Anchorage >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 3BT540 Instrument N2 Accumulator Screened >RLGM Component screened by SRT judgment. 3BT545 3B Srv Inst N2 Accumulator Screened >RLGM Component screened by SRT judgment. 3CD001 3C 125V DC Battery Anchorage 0.26 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 3CD003 Station Battery Charger 3C Block wall interaction 0.27 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 3CD017 (3FBC) Battery Main Fuse Box Anchorage >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 3CD019 (3FC) 250 Volt Fuse Box Anchorage >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002.
14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 60 of 63 Equipment ID Description Failure Mode HCLPF (g) Basis 3CD025 (3PPCD) 3PPCD 125V DC Distribution Panel Anchorage >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 3CE024 Residual Heat Exchanger Anchorage 0.26 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 3CS377 Back-Up N2 Supply to Ads
 
RV's Screened >RLGM Component is Rule Of Box to 3AS377. Parent component screens. 3CS456 Control to Battery Charger 3CD003 Transfer Switch Block wall interaction 0.27 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 3CT540 Instrument N2 Accumulator Screened >RLGM Component screened by SRT judgment. 3CT545 3C Srv Inst N2 Accumulator Screened >RLGM Component screened by SRT judgment. 3DA-W-A (1201) RCIC MO-3-13-021 Breaker Anchorage 0.43 Component is Rule Of Box to 3DA-W-A (30D012). Parent component is evaluated in 14Q4233-CAL-002. 3DA-W-A (1203) RCIC MO-3-13-030 Breaker Anchorage 0.43 Component is Rule Of Box to 3DA-W-A (30D012).
Parent component is evaluated in 14Q4233-CAL-002. 3DA-W-A (1204) RCIC MO-3-13-027 Breaker Anchorage 0.43 Component is Rule Of Box to 3DA-W-A (30D012).
Parent component is evaluated in 14Q4233-CAL-002. 3DA-W-A (1205) RCIC MO-3-13-041 Breaker Anchorage 0.43 Component is Rule Of Box to 3DA-W-A (30D012).
Parent component is evaluated in 14Q4233-CAL-002. 3DA-W-A (1206) RCIC MO-3-13-039 Breaker Anchorage 0.43 Component is Rule Of Box to 3DA-W-A (30D012).
Parent component is evaluated in 14Q4233-CAL-002. 3DA-W-A (1207) RCIC MO-3-13-132 Breaker Anchorage 0.43 Component is Rule Of Box to 3DA-W-A (30D012).
Parent component is evaluated in 14Q4233-CAL-002. 3DA-W-A (1209) RCIC MO-3-13-131 Breaker Anchorage 0.43 Component is Rule Of Box to 3DA-W-A (30D012).
Parent component is evaluated in 14Q4233-CAL-002. 3DA-W-A (1210) RCIC MO-3-13-018 Breaker Anchorage 0.43 Component is Rule Of Box to 3DA-W-A (30D012).
Parent component is evaluated in 14Q4233-CAL-002. 3DA-W-A (1214) RCIC Cond Vac PP 30P046 Breaker Anchorage 0.43 Component is Rule Of Box to 3DA-W-A (30D012).
Parent component is evaluated in 14Q4233-CAL-002. 3DA-W-A (1215) RCIC Vac Tank Cond PP 30P048 Breaker Anchorage 0.43 Component is Rule Of Box to 3DA-W-A (30D012).
Parent component is evaluated in 14Q4233-CAL-002.
3DA-W-A (30D012) RCIC 250VDC MCC Anchorage 0.43 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 3DD001 2D 125V DC Battery Anchorage 0.26 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 3DD017 Battery Main Fuse Box Anchorage >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 3DD019 Fuse Box Anchorage >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 3GT540 Instrument N2 Accumulator Screened >RLGM Component screened by SRT judgment. 3GT545 3G Srv Inst N2 Accumulator Screened >RLGM Component screened by SRT judgment. 3KT540 Instrument N2 Accumulator Screened >RLGM Component screened by SRT judgment. 3KT545 3K Srv Inst N2 Accumulator Screened >RLGM Component screened by SRT judgment. 3OE032 RCIC Barometric Condenser (13-2) Block wall interaction 0.44 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 3OE104 RCIC Turb. Lube Oil Cooler (13-2) Anchorage >0.24 Component is Rule Of Box to 30P036. Parent component is evaluated in 14Q4233-CAL-002. 3AC270 (K3A) Relay Equipment Capacity >0.24 Component is Rule Of Box to 3AC270. Parent component is evaluated in 14Q4233-CAL-002.
Component evaluated by HCLPF calculation 14Q4233-CAL-004. 3BC270 (K3B) Relay Equipment Capacity >0.24 Component is Rule Of Box to 3BC270. Parent component is evaluated in 14Q4233-CAL-002. Component evaluated by HCLPF calculation 14Q4233-CAL-004. 3AC270 (K3C) Relay Equipment Capacity >0.24 Component is Rule Of Box to 3AC270. Parent component is evaluated in 14Q4233-CAL-002. Component evaluated by HCLPF calculation 14Q4233-CAL-004.
14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 61 of 63 Equipment ID Description Failure Mode HCLPF (g) Basis 3BC270 (K3D) Relay Equipment Capacity >0.24 Component is Rule Of Box to 3BC270. Parent component is evaluated in 14Q4233-CAL-002. Component evaluated by HCLPF calculation 14Q4233-CAL-004. AO-3-07B-3511 Torus 18 Inch Vent Inboard
 
Isol Valve to Sbgt/Atmos Screened >RLGM Component screened by SRT analysis. AO-3-07B-90290 Ctmt Emerg Vent Outboard Isolation Vlv to Atmos Screened >RLGM Component screened by SRT analysis. E134 (1014) E134-T-B 30B059 Breaker Equipment Capacity >0.24 Component is Rule Of Box to E134 (30B010). Parent component is evaluated in 14Q4233-CAL-002. E134 (30B010) E134 Load Center Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. E134-T-B (30B059) E134-T-B Motor Control Center Block wall interaction 0.27 Component evaluated by HCLPF calculation 14Q4233-CAL-002. E134-T-B (5924) Alt Feed For Uninterrupt AC Power Supp Inverter 30D37 Block wall interaction 0.27 Component is Rule Of Box to E134-T-B (30B059).
Parent component is evaluated in 14Q4233-CAL-002. E134-T-B (5931) 125V. D.C. Battery Charger 3AD03 Block wall interaction 0.27 Component is Rule Of Box to E134-T-B (30B059).
Parent component is evaluated in 14Q4233-CAL-002. E334 (1213) E334-R-B 30B038 Breaker Equipment Capacity >0.24 Component is Rule Of Box to E334 (30B012). Parent component is evaluated in 14Q4233-CAL-002. E334 (30B012) E334 Load Center Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002.
E334-R-B (30B038) MCC 30B038 Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. E334-R-B (3821) MO-3-10-038A Breaker Equipment Capacity >0.24 Component is Rule Of Box to E334-R-B (30B038).
Parent component is evaluated in 14Q4233-CAL-002. E334-R-B (3824) MO-3-10-026A Breaker Equipment Capacity >0.24 Component is Rule Of Box to E334-R-B (30B038).
Parent component is evaluated in 14Q4233-CAL-002. E334-R-B (3831) MO-3-10-039A Breaker Equipment Capacity >0.24 Component is Rule Of Box to E334-R-B (30B038).
Parent component is evaluated in 14Q4233-CAL-002. E334-R-B (3832) MO-3-10-034A Breaker Equipment Capacity >0.24 Component is Rule Of Box to E334-R-B (30B038).
Parent component is evaluated in 14Q4233-CAL-002. E334-R-B (3844) MO-3-10-031A Breaker Equipment Capacity >0.24 Component is Rule Of Box to E334-R-B (30B038).
Parent component is evaluated in 14Q4233-CAL-002. E334-R-B (3851) MO-3-10-25A Alt Breaker Equipment Capacity >0.24 Component is Rule Of Box to E334-R-B (30B038).
Parent component is evaluated in 14Q4233-CAL-002. E334-R-B (3862) MO-3-10-174 Breaker Equipment Capacity >0.24 Component is Rule Of Box to E334-R-B (30B038).
Parent component is evaluated in 14Q4233-CAL-002. E334-R-B (3863) MO-3-10-176 Breaker Equipment Capacity >0.24 Component is Rule Of Box to E334-R-B (30B038). Parent component is evaluated in 14Q4233-CAL-002. E334-R-B (3882) Norm Fdr for 120V Instr Pnl 30Y35 Trans 30X135 Equipment Capacity >0.24 Component is Rule Of Box to E334-R-B (30B038).
Parent component is evaluated in 14Q4233-CAL-002. E334-R-B (3893) 125V Battery Charger C 3C Transfer Switch 30S704 Equipment Capacity >0.24 Component is Rule Of Box to E334-R-B (30B038).
Parent component is evaluated in 14Q4233-CAL-002. INV-3-13-90 RCICs-125 VDC Bus 'A' Power Distribution Equipment Capacity 0.44 Component is Rule Of Box to 30C019. Parent component is evaluated in 14Q4233-CAL-002. J2919 J-Box at E134 LC Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. J2920 J-Box at E334 LC Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. LI-3-02-3-085A Reactor Vessel High Water Equipment Capacity >0.24 Component is Rule Of Box to 30C005A. Parent component is evaluated in 14Q4233-CAL-002. LI-3-02-3-113 Reactor Water Level Equipment Capacity >0.24 Component is Rule Of Box to 30C003. Parent component is evaluated in 14Q4233-CAL-002. LR/TR-9123A Torus Water Level/Temperature Recorder Equipment Capacity >0.24 Component is Rule Of Box to 30C004C. Parent component is evaluated in 14Q4233-CAL-002. LT-3-02-3-072A Reactor Vessel Water Level Equipment Capacity >0.24 Component is Rule Of Box to 3AC065. Parent component is evaluated in 14Q4233-CAL-002. LT-3-02-3-113 Level Transmitter Equipment Capacity >0.24 Component is Rule Of Box to 3BC091. Parent component is evaluated in 14Q4233-CAL-002. LT-9123A Torus Water Level Screened >RLGM Component screened by SRT analysis.
14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 62 of 63 Equipment ID Description Failure Mode HCLPF (g) Basis MO-3-10-025A RHR Inner Injection Valve to Recirc Loop A Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. MO-3-10-026A RHR Loop A D/W Spray O/B Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. MO-3-10-031A RHR Loop A D/W Spray I/B Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. MO-3-10-034A RHR Loop A FFT Valve Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. MO-3-10-038A RHR Loop A Torus Spray Screened >RLGM Component screened by SRT analysis. MO-3-10-039A RHR Loop A Torus Valve Screened >RLGM Component screened by SRT analysis. MO-3-10-174 HPSW to RHR Inner X-Tie Screened >RLGM Component screened by SRT analysis. MO-3-10-176 HPSW to RHR Outer X-Tie Screened >RLGM Component screened by SRT analysis. MO-3-13-018 RCIC Pump Suction from Condensate Storage Tank Screened >RLGM Component screened by SRT analysis. MO-3-13-021 RCIC Discharge to Feedwater Line B Screened >RLGM Component screened by SRT judgment. MO-3-13-027 RCIC Minimum Flow Valve Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. MO-3-13-030 RCIC Full Flow Test Valve Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. MO-3-13-039 RCIC Pump Torus Suction Outer Screened >RLGM Component screened by SRT analysis. MO-3-13-041 RCIC Pump Torus Suction Inner Screened >RLGM Component screened by SRT analysis. MO-3-13-131 RCIC Turbine Steam Supply
 
Valve Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. MO-3-13-132 RCIC Cooling Water Supply to Lo Clr + Barometric Cdsr Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. N310025A Cabinet Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. PI-3-06-090A Reactor Wide Range Press
 
Ind Equipment Capacity >0.24 Component is Rule Of Box to 30C005A. Parent component is evaluated in 14Q4233-CAL-002. PI-3-06-090B Reactor Wide Range Press
 
Ind Equipment Capacity >0.24 Component is Rule Of Box to 30C005A. Parent component is evaluated in 14Q4233-CAL-002. PI-3-06-090C Reactor Wide Range Press
 
Ind Equipment Capacity >0.24 Component is Rule Of Box to 30C005A. Parent component is evaluated in 14Q4233-CAL-002. PI-3-13-094 RCIC Pump Turb Stm Press Equipment Capacity >0.24 Component is Rule Of Box to 30C004C. Parent component is evaluated in 14Q4233-CAL-002. PR/LR-3-06-096 Reactor Level/Steam Flow
 
Ratio Equipment Capacity >0.24 Component is Rule Of Box to 30C005A. Parent component is evaluated in 14Q4233-CAL-002. PR/TR-5805 Containment Pressure/Temp Equipment Capacity >0.24 Component is Rule Of Box to 30C003. Parent component is evaluated in 14Q4233-CAL-002. PS-3-13-67-1 Pressure Switch Anchorage 0.41 Component is Rule Of Box to 30C095. Parent component is evaluated in 14Q4233-CAL-002. Component evaluated by HCLPF calculation 14Q4233-CAL-004. PS-3-13-72A Pressure Switch Functional Failure 0.27 Component evaluated by HCLPF calculation 14Q4233-CAL-004. PS-3-13-72B Pressure Switch Functional Failure 0.27 Component evaluated by HCLPF calculation 14Q4233-CAL-004. PS-3-13-87A Pressure Switch Functional Failure 0.27 Component evaluated by HCLPF calculation 14Q4233-CAL-004. PS-3-13-87B Pressure Switch Functional Failure 0.27 Component evaluated by HCLPF calculation 14Q4233-CAL-004. PT-3-02-3-404A Reactor Pressure Equipment Capacity >0.24 Component is Rule Of Box to 3AC065. Parent component is evaluated in 14Q4233-CAL-002. PT-3-02-3-404C Reactor Pressure Transmitter Equipment Capacity >0.24 Component is Rule Of Box to 3AC091. Parent component is evaluated in 14Q4233-CAL-002.
14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 63 of 63 Equipment ID Description Failure Mode HCLPF (g) Basis PT-3-06-053A Reactor Wide Range -Pressure Equipment Capacity >0.24 Component is Rule Of Box to 3AC065. Parent component is evaluated in 14Q4233-CAL-002. PT-3-06-053B Reactor Wide Range -
Pressure Equipment Capacity >0.24 Component is Rule Of Box to 3BC065. Parent component is evaluated in 14Q4233-CAL-002. PT-3-06-053C Reactor Wide Range -
Pressure Equipment Capacity >0.24 Component is Rule Of Box to 3AC065. Parent component is evaluated in 14Q4233-CAL-002. PT-3-13-068 RCIC Turbine Steam Supply Pressure Anchorage 0.41 Component is Rule Of Box to 30C095. Parent component is evaluated in 14Q4233-CAL-002. PT-5805 Drywell Pressure Screened >RLGM Component screened by SRT analysis. RV-3-02-071A 3A Safety Relief Valve Screened >RLGM Component screened by SRT judgment. RV-3-02-071B 3B Safety Relief Valve Screened >RLGM Component screened by SRT judgment. RV-3-02-071C 3C Safety Relief Valve Screened >RLGM Component screened by SRT judgment. RV-3-02-071G 3G Safety Relief Valve Screened >RLGM Component screened by SRT judgment. RV-3-02-071K 3K Safety Relief Valve Screened >RLGM Component screened by SRT judgment. XAM-3-02 117A Reactor Water Level Wide Range Equipment Capacity 0.44 Component is Rule Of Box to 30C818. Parent component is evaluated in 14Q4233-CAL-002.
 
1 . 2. Enclosure 2 Peach Bottom Atomic Power Station, Units 2 and 3
 
==SUMMARY==
OF REGULATORY COMMITMENTS The following tat)le identifies commitments made in this document. (Any other actions discussed in the submittal represent intended or planned actions. They are described to the NRC for the NRC's information and are not regulatory commitments.)
COMMITMENT TYPE COMMITTED COMMITMENT DATE OR. ONE-TIME ACTION PROGRAMMATIC "OUTAGE" (Yes/No) (Yes/No) Complete further evaluation and implement Unit 3 -P3R21 Yes No modifications, if required, to increase seismic (Fall 2017) rna1*gin for the following plant relay: Unit 3 Relay 3-13A-K033 Corn further evaluation and Unit 2 -P2R22 Yes No modifications, if required, to increase seismic (Fall 2018) margin for the following plant relay: Unit 2 Relay 2-13A-K033 Exelon Generation RS-14-300 December 19, 2014 U.S. Nuclear Regulatory Commission Attn: Document Control Desk 11555 Rockville Pike, Rockville.
MD 20852 Peach Bottom Atomic Power Station, Units 2 and 3 10 CFR 50.54(f) Renewed Facility Operating License Nos. DPR-44 and DPR-56 NRG Docket Nos. 50-277 and 50-278
 
==Subject:==
Exelon Generation Company, LLC Expedited Seismic Evaluation Process Report (CEUS Sites), Response to NRG Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident
 
==References:==
: 1. NRG Letter, Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(f) Regarding Recommendations 2.1, 2.3, and 9.3, of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident, dated March 12, 2012 (ML 12053A340)
: 2. NEI Letter, Proposed Path Forward for NTTF Recommendation 2.1: Seismic evaluations, dated April 9, 2013 (ML 13101 A379) 3. Seismic Evaluation Guidance: "Augmented Approach for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1 -Seismic'', EPRI, Palo Alto, CA: May 2013.3002000704(ML13102A142)
: 4. NRG Letter, Electric Power Research Institute Report 3002000704, "Seismic Evaluation Guidance:
Augmented Approach for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1: Seismic," as an Acceptable Alternative to the March 12, 2012, Information Request for Seismic Re-evaluations, dated May 7, 2013 (ML 13106A331)
: 5. Exelon Generation Company, LLC, Seismic Hazard and Screening Report (Central and Eastern United States (CEUS) Sites), Response to NRG Request for Information Pursuant to 1 O CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident (RS-14-071), dated March 31, 2014 (ML14090A247) . 6. Exelon Generation Company, LLC Response to NRG Request for Information Pursuant to 1 O CFR 50.54(f) Regarding the Seismic Aspects of Recommendation 2.1 of the Term Task Force Review of Insights from the Fukushima Dai-ichi Accident -1.5 Year Response for CEUS Sites (RS-13-205), dated September 12, 2013 (ML 13256A070)
U.S. Nuclear Regulatory Commission NTTF 2.1 Seismic Response for CEUS Sites December 19, 2014 Page2 On March 12, 2012, the Nuclear Regulatory Commission (NRG) issued a 50.54(f) letter to all power reactor licensees and holders of construction permits in active or deferred status. Enclosure 1 of Reference 1 requested each addressee located in the Central and Eastern United States (CEUS) to submit a Seismic Hazard Evaluation and Screening Report within 1.5 years from the date of Reference 1 . In Reference 2, the Nuclear Energy Institute (NEI) requested NRG agreement to delay submittal of the final CEUS Seismic Hazard Evaluation and Screening Reports so that an update to the Electric Power Research Institute (EPRI) ground motion attenuation model could be completed and used to develop that information.
NEI proposed that descriptions of subsurface materials and properties and base case velocity profiles be submitted to the NRG by September 12, 2013, (Reference 6), with the remaining seismic hazard and screening information submitted by March 31, 2014 (Reference 5). NRG agreed with that proposed path forward in Reference
: 4. Reference 1 requested that licensees provide interim evaluations and actions taken or planned to address the higher seismic hazard relative to the design basis, as appropriate, prior to completion of the risk evaluation.
In accordance with the NRG endorsed guidance in Reference 3, Enclosure 1 provides the Expedited Seismic Evaluation Process (ESEP) Report for Peach Bottom Atomic Power Station, Units 2 and 3, and the information described in the "ESEP Report" Section 7, of Reference 3 in accordance with the schedule identified in Reference
: 2. With the exception of the two (2) plant relays identified in Enclosure 1 , all other equipment evaluated for the ESEP for Peach Bottom Atomic Power Station Units 2 and 3 was found to have adequate capacity for the required seismic demand as defined by the Augmented Approach (ESEP) guidance (Reference 3). Further evaluation, and implementation of modifications, if required, to increase seismic margin for the affected plant relays, will be completed as identified in Enclosure
: 2. This ESEP report transmittal completes regulatory Commitment No. 3 of Reference
: 5. A list of new regulatory commitments contained in this letter is provided in Enclosure
: 2. If you have any questions regarding this report, please contact Ron Gaston at (630) 657-3359.
I declare under penalty of perjury that the foregoing is true and correct. Executed on the 19 1 h day of December 2014. Respectfully submitted, James Barstow Director -Licensing
& Regulatory Affairs Exelon Generation Company, LLC U.S. Nuclear Fle9ulatory Commission NTT F 2. 1 Seisrnic F{<::sponse for CEUS Sites DecembCJr 19, 2014 3
 
==Enclosures:==
: 1. Peach Bottorn Atornic Power Station, Units 2 and 3 Expedited Seismic Evaluation Process (ESEP) Report 2. Surnmary of Commitments cc: Director, Office of Nuclear Reactor Regulation Regional Adn1inistrator
-NRC Region I NRC Senior Resident Inspector
-Peach Bottom Atomic Power Station NRC Project Manager, NRR -Peach Bottom Atomic Power Station Mr. Nicholas J. DiFrancesco, NRR/JLD/JHMB, NRC Director, Bureau of Radiation Protection
-Pennsylvania Department of Environmental Resources S. T. Gray, State of Maryland R. R. Jana ti, Chief, Division of Nuclear Safety, Pennsylvania Department of Environmental Protection, Bureau of Radiation Protection Enclosure 1 Peach Bottom Atomic Power Station, Units 2 and 3 Expedited Seismic Evaluation Process (ESEP) Report (63 pages)
EXPEDITED SEISMIC EV A LUATION PROCESS (ESEP) RE PORT IN R ESPONSE TO THE 50.54(f) INFO RMA TION REQUEST REGARDING FUKUSHI MA NEA R-TERM TAS K FO CE RECOM M ENDATION 2.1: SBS MI C for the PHch Bottom Atomk Pow* Stlltlon I/nits 2 a I 1848 uy ROM/ /hll tll, Pminsylnnla l1J1.4-9012 F*clllly Optlt'atlng Llc*n* No. DPR-.U, DPR*S6 NRC Docket No. 50*277, 50*278 --**---*****----***---****------*----**-***-__________ Co rr.npondence
..No.i.RS.-, U..300-----**-----***--*--
---------*-------**--*-*** -----***-**--** ***---***** *-*------* --*** *-***** *-*fxetorr-G-eneratlon Company,*tte{EXetc>nr-
**-**--***-****-****-*-****----**----****-*--***-* PO Box 805398 Chicago, IL 60680-5398 Prepared by: Stevenson
& Associates 275 Mishawum Road, Suite 200 Woburn, MA 01801 -------******* ---****-*-**--------*-----*---*-* *-------ReportNum6er.i4Q4233--RPr=004.
-Rev.f* *-----------* ------------*****--------** Printed Name Siglllture J2ile. Preparer:
Douglas Seymour f!)/l'&sect;r 12/17 /2014 Reviewer:
A. Karavoussianis -
12/17 /2014 Approver:
A. Karavousslanis U/17/2014 Lead Respon sib le Tracey Gallagher Branch Manager: Frank Glaco Senior Manager Design Engineering:
__ o __ a __
Corporate Acceptance:
Jeffrey Clari<
Document Title:  EXPEDITED SEISMIC EVALUATION PROCESS (ESEP) REPORT IN RESPONSE TO THE 50.54(f) INFORMATION REQUEST REGARDING FUKUSHIMA NEAR-TERM TASK FORCE RECOMMENDATION 2.1:  SEISMIC for the Peach Bottom Atomic Power Station Units 2 and 3  Document Type: Report Report Number: 14Q4233-RPT-004 Project Name:
Exelon ESEP for Peach Bottom Job No.: 14Q4233 Client:
This document has been prepared in accordance with the S&A Quality Assurance Program Manual, Revision _17_ and project requirements:
Rev. 0 Prepared by:
Douglas Seymour Date: 12/5/2014 Reviewed by:
A. Karavoussianis Date: 12/5/2014 Approved by:  A. Karavoussianis Date: 12/5/2014 Revision Record: Revision No. Prepared by/
Date Reviewed by/
Date Approved by/
Date Description of Revision 1    D. Seymour 12/10/2014
 
A. Karavoussianis 12/10/2014 A. Karavoussianis 12/10/2014 Incorporation of client editorial
 
comment and update of Reference 4.
2 D. Seymour 12/15/2014
 
A. Karavoussianis 12/16/2014 A. Karavoussianis 12/16/2014 Incorporation of minor editorial comments.
3    D. Seymour 12/17/2014
 
A. Karavoussianis 12/17/2014 A. Karavoussianis 12/17/2014 Incorporation of minor editorial comments.
Stevenson & Associates DOCUMENT APPROVAL SHEET CONTRACT NO. 14Q4233 14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 3 of 63 Table of Contents 1 Purpose and Objective ................................................................................................... 6 2 Brief Summary of the FLEX Seismic Implementation Strategies
............................... 7 3 Equipment Selection Process and ESEL ...................................................................... 10  3.1 Equipment Selection Process and ESEL and Alternate Path Justifications ........................
10  3.1.1 ESEL Development
..............................................................................................................................
11  3.1.2 Power Operated Valves ..................................................................................................
..................
13  3.1.3 Pull Boxes .............................................................................................................
..................................
13  3.1.4 Termination Cabinets ...................................................................................................
.....................
13  3.1.5 Critical Instrumentation Indicators ....................................................................................
...........
13  3.1.6 Phase 2 and Phase 3 Piping Connections
..................................................................................
14  3.2 Justification for use of Equipment that is not the Primary Means for FLEX  Implementation ..............................................................................................................
.........................
14 4 Ground Motion Response Spectrum (GMRS) ............................................................ 15  4.1 Plot of GMRS Submitted by the Licensee ...................................................................................
.. 15  4.2 Comparison to SSE ........................................................................................................
.........................
16 5 Review Level Ground Motion (RLGM) ....................................................................... 17  5.1 Description of RLGM Selected .............................................................................................
..............
17  5.2 Method to Estimate ISRS ..................................................................................................
...................
19 6 Seismic Margin Evaluation Approach ........................................................................ 20  6.1 Summary of Methodologies used ............................................................................................
........ 20  6.2 HCLPF Screening Process ..................................................................................................
..................
20  6.3 Seismic Walkdown Approach ................................................................................................
............
21  6.3.1 Walkdown Approach ......................................................................................................
...................
21  6.3.2 Application of Previo us Walkdown Information .....................................................................
23  6.3.3 Significant Walkdown Observations ......................................................................................
...... 23  6.4 HCLPF Calculation Process ................................................................................................
..................
23  6.5 Functional Evaluation of Relays ..........................................................................................
...............
24  6.6 Tabulated ESEL HCLPF Values (including key failure modes) ................................................
24 7 Inaccessible Items
......................................................................................................... 26  7.1 Identification of ESEL Items Inaccessible for Walkdowns ........................................................
26  7.2 Planned Walkdown / Evaluation Schedule / Close Out ...........................................................
27 8 ESEP Conclusions and Results ..................................................................................... 28 14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 4 of 63 8.1 Supporting Information ...................................................................................................
....................
28  8.2 Identification of Planned Modi fications ..................................................................................
....... 30  8.3 Modification Implementation Schedule .....................................................................................
... 30  8.4 Summary of Regulatory Commitments
..........................................................................................
30 9 Refere nces ..................................................................................................................
... 31 Attachment A Peach Bottom Unit 2 and Common ESEL Attachment B Peach Bottom Unit 3 ESEL Attachment C ESEP HCLPF Values and Failure Modes Tabulation, Unit 2 and Common Attachment D ESEP HCLPF Values and Failure Modes Tabulation, Unit 3
 
14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 5 of 63 List of Tables Table 3.1-1 Flow Paths Credited for ESEP Table 4.1-1 Peach Bottom GMRS Table 4.2-1 Peach Bottom GMRS vs. SSE Table 5.1-1 Ratio between GMRS and SSE Table 5.1-2 Peach Bottom RLGM Table 7.1-1 Items Inaccessible for Walkdowns List of Figures Figure 4.1-1 Peach Bottom GMRS Figure 4.2-1 Peach Bottom GMRS vs. SSE Figure 5.1-1 Peach Bottom RLGM
 
14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 6 of 63 1 PURPOSE AND OBJECTIVE Following the accident at the Fukushima Dai-ichi nuclear power plant resulting from the March 11, 2011, Great Tohoku Earthquake and subsequent tsunami, the Nuclear Regulatory Commission (NRC) established a Near Term Task Force (NTTF) to conduct a systematic review of NRC processes and regulations and to determine if the agency should make additional improvements to its regulatory system. Th e NTTF developed a set of recommendations intended to clarify and strengthen the regulatory framework for protection against natural phenomena. Subsequently, the NRC issued a 50.54(f) letter on March 12, 2012 (Ref. 1) requesting information to assure that these recommendations are addressed by all U.S. nuclear power plants. The 50.54(f) letter requests that licensees and holders of construction permits under 10 CFR Part 50 reevaluate the seismic hazards at their sites against present-day NRC requirements and guidance. Depending on the comparison between the reevaluated seismic hazard and the current design basis, further risk assessment may be required. Assessment approaches acceptable to the staff include a seismic probabilistic risk assessment (SPRA), or a seismic margin assessment (SMA). Based upon the assessment results, the NRC staff will determine whether additional regulatory actions are necessary. This report describes the Expedited Seismic Evaluation Process (ESEP) undertaken for Peach Bottom Atomic Power Station (PBAPS). The intent of the ESEP is to perform an interim action in response to the NRC's 50.54(f) letter (Ref. 1) to demonstrate seismic margin through a review of a subset of the plant equipment that can be reli ed upon to protect the reactor core following beyond design basis seismic events. The ESEP is implemented using the methodologies in the NRC endorsed guidance in EPRI 3002000704, Seismic Evaluation Guidance: Augmented Approach for the Resolution of Fukushima Near-Term Task Force Recomme ndation 2.1: Seismic (Ref. 2). The objective of this report is to provide summary information describing the ESEP evaluations and results. The level of detail provided in the report is intended to enable NRC to understand the inputs used, the evaluations performed, and the decisions made as a result of the interim evaluations.
 
14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 7 of 63 2 BRIEF
 
==SUMMARY==
OF THE FLEX SEIS MIC IMPLEMENTATION STRATEGIES
* The Peach Bottom FLEX response strategies for Reactor Core Cooling and Heat Removal, Reactor Inventory Control, Core Cooling and Heat Removal (Modes 3 and 4), Containment Function and Spent Fuel Pool Control are summarized below. This summary is derived from the Overall Integrated Plan (OIP) in Response to the March 12, 2012, Commission Order EA-12-049 (Ref. 3) including the required 6-Month Updates that have been prepared since the OIP was submitted. Flex Phase 1, strategy relies on installed plant equipment. Reactor core cooling and heat removal is achieved via steam release from the Safety Relief Valves (SRV's) to the Torus and Reactor Core Isolation Cooling (RCIC) drive steam. Coolant ma keup is established and maintained from RCIC. Preferred suction for RCIC will be from the condensate storage tank (CST) if it is still viable. If not, suction will be from the Torus. The cool down rate of the reactor coolant system is controlled through manual operation of the SRV's and RCIC at a targeted 80 degrees per hour. The cool down will facilitate decay heat removal while keeping the Torus water temperature within limits and reactor pressure high enough to maintain RCIC operation. RCIC operation is DC and independent of emergency AC power. It can also be operated manually without power if necessary. SRV's operate mechanically at specific high set pressures and manually with DC power to solenoids with nitrogen accumulators to enable operation to establish a cool down below set pressures. Key reactor parameters are obtained via DC powered instrumentation. A DC load reduction strategy is employed to extend battery life. The Spent Fuel Pool makeup may be required if a full core off load was recently completed as time to boil in the pool is at its most limiting. Spent Fuel level is obtained from the new spent fuel pool wide range instrumentation installed under order EA-12-051. Flex Phase 2, strategy relies on installed plant equipment and portable equipment. If RCIC is running without challenging limits, it will cont inue in operation to control reactor level and pressure. It is recognized that area temperatures will increase. Opening plant doors and fans powered by portable diesel generators will be used to address the RCIC room environmental issues. Work to line-up and start operation of phase 2 equipment will commence no later than 1 hour after the event occurs. A portable 500 KW 480VAC diesel generator is used to provide power to 480 VAC Load Centers and Motor Control Centers. This will allow energizing the Division 1 battery chargers, battery room ventilation, Control Room Emergency Ventilation (CREV), diesel fuel oil transfer pump, and valves re quired to inject water from external sources to the reactor via the Residual Heat Removal (RHR) system. Portable diesel driven pumps will provide injection from two suction sources, either the Emergency Cooling Tower (ECT) or the Conowingo Pond (Susquehanna River) at the screen house. The discharge will be either into the RHR system directly or into the High Pressure Service Water system (HPSW) emergency cross-tie to the RHR system. RHR will then supply water to the Reactor, Torus and the Fuel Pool Cooling system. The Refuel Floor roof hatch and the Reac tor Building truck bay doors will need to be opened as a method to maintain temperature on the refuel floor.
* This section is based upon input received from Peach Bottom Atomic Power Station in (Refs. 4, 25, 26, and 27).
14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 8 of 63 Flex Phase 1 and 2 strategy will provide sufficient capability such that no additional Phase 3 strategies are required. Phase 3 relies on installed plant equipment and portable equipment as described in Phase 1 and Phase 2.
Peach Bottom Phase 1 Flex Strategy Safety Function Primary Method Alternate Method Core Cooling Reactor Core Cooling & Heat Removal
* RCIC pump with suction from Torus
* RCS cool down at 80
&deg;F/hour with SRV's
* RCIC pump with suction from CST
* Backup nitrogen to SRV
 
solenoids RCS Inventory Control & Heat Removal,
* RCIC pump with suction from Torus
* RCS cool down at 80
&deg;F/hour with SRV's
* RCIC pump with suction from CST
* Backup nitrogen to SRV
 
solenoids Key Reactor Parameters
* Temperature, pressure, level
* Use existing ba ttery powered indication
* Extend coping with deep DC load stripping Containment Containment Pressure Control
 
& Remove Heat
* None required - pressure and temperature below limits
* Containment venting if Torus temperature reaches 230
&deg;F Key Containment Parameters
* Temperature and pressure indication powered via vital buses
* Temperature and pressure indication powered via vital buses SFP Cooling Spent Fuel Cooling
* None required - temperature below limits
* None required - temperature below limits SFP Parameters
* SFP Wide Range Level Indicator
* SFP Wide Range Level Indicator
 
14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 9 of 63 Peach Bottom Phase 2 Flex Strategy Safety Function Primary Method Alternate Method Core Cooling Reactor Core Cooling & Heat Removal (Mode 3 and 4)
* Diesel driven portable pumps
* 500KW 480VAC FLEX generator to operate valves
* Diesel driven portable pumps at alternate suction and discharge locations
* 500KW 480VAC FLEX generator to operate valves at alternate location RCS Inventory Control & Heat
 
Removal (Mode 3 and 4)
* Diesel driven portable pumps
* 500KW 480VAC FLEX generator to operate valve
* Diesel driven pump at alternate suction and discharge locations
* 500KW 480VAC FLEX generator to operate valves at alternate location Key Reactor Parameters
* 500KW 480VAC FLEX generator repower one Vital Load center to repower battery charger
* Local indication determination at rack or penetration Containment Containment Pressure Control
& Remove Heat
* None required - pressure and temperature below limits
* None required - pressure and temperature below limits Key Containment Parameters
* Temperature and pressure indication powered via vital buses
* Local indication determination at rack or penetration SFP Cooling Spent Fuel Cooling
* RHR to Fuel Pool cooling crosstie
* Fire hose spray on the spent fuel SFP Parameters
* SFP Wide Range Level Indicator
* SFP Wide Range Level Indicator
 
14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 10 of 63 3 EQUIPMENT SELECTION PROCESS AND ESEL AND ALTERNATE PATH JUSTIFICATIONS The selection of equipment for the ESEL followed the guidelines of EPRI 3002000704 (Ref. 2). The ESELs per Ref. 22 for Unit 2 and Unit 3 are presented in Attachments A and B, respectively.
3.1 Equipment Selection Process and ESEL The selection of equipment to be included on the Expedited Seismic Equipment List (ESEL) was based on installed plant equipment credited in the FLEX strategies during mitigation of an Extended Loss of AC Power (ELAP), as outlined in the Peach Bottom Overall Integrated Plan (OIP) in Response to the March 12, 2012, Commission Order EA-12-049 (Ref. 3) including subsequent 6 month updates through August 2014
: f. The OIP provides the Peach Bottom FLEX 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 following a beyond design basis seismic event, consistent with the Peach Bottom OIP (Ref. 3) including subsequent 6 month updates through August 2014. FLEX recove ry actions are excluded from the ESEP scope per EPRI 3002000704 (Ref. 2). The overall list of planned FLEX modifications and the scope for consideration herein is limited to those required to support core cooling, reactor coolant inventory and sub-criticality, and containment integrity functions. Portable and pre-staged FLEX equipment (not permanently installed) are excluded from the ESEL per EPRI 3002000704 (Ref. 2). The ESEL component selection followed the EPRI guidance outlined in Section 3.2 of EPRI 3002000704 (Ref. 2). 
: 1. The scope of components is limited to that required to accomplish the core cooling and containment safety functions identified in Table 3-1 of EPRI 3002000704. The instrumentation monitoring requirements for core cooling/containment safety functions are limited to those outlined in the EPRI 3002000704 guidance, and are a subset of those outlined in the Peach Bottom OIP (Ref.3) including subsequent 6 month updates through August 2014. 
: 2. The scope of components is limited to installed plant equipment, and FLEX connections necessary to implement the Peach Bottom OIP (Ref. 3) including subsequent 6 month updates through August 2014 as described in Section 2.
: 3. The scope of components assumes the credited FLEX connection modifications are implemented, and are limited to those required to support a single FLEX success path (i.e., either "Primary" or "Back-up/Alternate"). 
: 4. The "Primary" FLEX success path is to be specified. Selection of the "Back-up/Alternate" FLEX success path must be justified.
f References 25 and 26 confirm that there are no changes between the February 2013 and August 2014 Flex Strategies submittals. This footnote applies to all references of the August 2014 Flex Strategies throughout this document.
14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 11 of 63
: 5. Structures, systems, and components excluded per the EPRI 3002000704 (Ref. 2) guidance are:
* Structures (e.g. containment, reactor building, control building, auxiliary building, etc.)
* Piping, cabling, conduit, HVAC, and their supports.
* Manual valves, check valves and rupture disks.
* Power-operated valves not required to change state as part of the FLEX mitigation strategies.
* Nuclear steam supply system components (e.g. reactor pressure vessel and internals, reactor coolant pumps and seals, etc.)
: 6. For cases in which neither train was specified as a primary or back-up strategy, then only one train component (generally 'A' train) is included in the ESEL.
3.1.1 ESEL Development The ESEL was developed by reviewing the Peach Bottom OIP (Ref. 3) including subsequent 6 month updates through August 2014 to determine equipment involved in the FLEX strategies. Further reviews of plant drawings (e.g., Process and Instrumentation Diagrams (P&IDs) and Electrical One Line Diagrams) were performed to identify the boundaries of the flow paths to be used in the FLEX strategies and to identify sp ecific components in the flow paths needed to support implementation of the FLEX strategies.
Boundaries were established at an electrical power distribution or mechanical isolation device in branch circuits / branch lines off the defined strategy electrical or fluid flow path. P&IDs were the primary reference documents used to identify mechanical components and instrumentation. The flow paths used for FLEX strategies were selected and specific components were identified using detailed equipment and instrument drawings, piping isometrics, electrical schematics and one-line drawings, system descriptions, design basis documents, etc., as necessary. The flow paths credited for ESEP for Peach Bottom per Ref. 22 are shown in Table 3.1-1.
 
14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 12 of 63 Table 3.1-1 Flow Paths Credited for ESEP Flow Path P&IDs Unit 2 Unit 3 Core Heat  Removal using RCIC system: Coolant from the Suppression Pool to the RCS via the RCIC pump and Reactor Recirculation Pump Discharge Piping. Main Steam providing motive force to the RCIC pump turbine and exhausted to the Suppression Pool. M-359 sh. 1 (Ref. 21.h)M-360 sh. 1 (Ref. 21.j) M-360 sh. 3 (Ref. 21.l) M-359 sh. 1 (Ref. 21.h)M-359 sh. 2 (Ref. 21.i)
M-360 sh. 1 (Ref. 21.j) M-360 sh. 2 (Ref. 21.k) M-360 sh. 4 (Ref. 21.m) RPV Pressure Control using ADS system: Main Steam relieved through the ADS Safety/Relief Valves to the Suppression Pool. M-351 sh. 1 (Ref. 21.d)M-351 sh. 2 (Ref. 21.e) M-372 sh. 1 (Ref. 21.t) M-333 sh. 1 (Ref. 21.b) M-351 sh. 3 (Ref. 21.f)M-351 sh. 4 (Ref. 21.g)M-372 sh. 1 (Ref. 21.t) M-333 sh. 2 (Ref. 21.c) RPV Make Up: Coolant from the Ultimate Heat Sink to the Suppression Pool via the FLEX pump and the RHR system. M-361 sh. 1 (Ref. 21.n)M-361 sh. 2 (Ref. 21.o) M-361 sh. 3 (Ref. 21.p)M-361 sh. 4 (Ref. 21.q) Hardened Containment Vent: Torus vented to atmosphere. M-367 sh. 1 (Ref. 21.r) M-367 sh. 2 (Ref. 21.s) Essential Service Water: Coolant from the Ultimate Heat Sink to the RHR system via a FLEX pump connection. M-315 sh. 1 (Ref. 21.a) M-315 sh. 1 (Ref. 21.a) Main Control Room Ventilation: Outside air through the Control Room Emergency Ventilation filters to a Control Room Emergency Ventilation Fan. Fan discharge into the Control Room. M-384 sh. 1 (Ref. 21.u)M-384 sh. 2 (Ref. 21.v) M-384 sh. 3 (Ref. 21.w) M-384 sh. 1 (Ref. 21.u)M-384 sh. 2 (Ref. 21.v) M-384 sh. 3 (Ref. 21.w) Battery Room Ventilation: Outside Air to the Emergency Switch Gear & Battery Room Vent. Fan supplied to both Units' Battery Rooms. Air from Battery Rooms exhausted through a Battery Room Exhaust Fan to atmosphere. M-399 sh. 1 (Ref. 21.x) M-399 sh. 2 (Ref. 21.y) M-399 sh. 3 (Ref. 21.z) M-399 sh. 4 (Ref. 21.aa) M-399 sh. 1 (Ref. 21.x) M-399 sh. 2 (Ref. 21.y) M-399 sh. 3 (Ref. 21.z) M-399 sh. 4 (Ref. 21.aa)
 
14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 13 of 63 3.1.2 Power Operated Valves Page 3-3 of EPRI 3002000704 (Ref. 2) notes that power operated valves not required to change state are excluded from the ESEL. Page 3-2 also notes that "functional fail ure modes of electrical and mechanical portions of the installed Phase 1 equipment should be considered (e.g. RCIC trips)."  To address this concern, the following guidance is applied in the Peach Bottom ESEL for functional failure modes associated with power operated valves:
* Power operated valves that must remain energized during the Extended Loss of all AC Power (ELAP) events in order to maintain a credited FLEX flow path or pressure boundary (such as DC powered solenoid-operated va lves), were included on the ESEL.
* Power operated valves not required to change state as part of the FLEX mitigation strategies were not included on the ESEL. The seismic event also causes the ELAP event; therefore, the valves are incapable of spurious operation as they would be de-energized.
* Power operated valves not required to change state as part of the FLEX mitigation strategies during Phase 1, and are re-energized and operated during subsequent Phase 2 and 3 strategies, were not evaluated for spurious valve operation as the seismic event that caused the ELAP has passed before the valves are re-powered.
3.1.3 Pull Boxes Pull boxes were deemed unnecessary to add to the ESELs as these components provide completely passive locations for pulling or installing cables. No breaks or connections in the cabling are included in pull boxes. Pull boxes were considered part of conduit and cabling, which are excluded in accord ance with EPRI 3002000704 (Ref. 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, and the connections are excluded from the ESEL.
3.1.5 Critical Instrumentation Indicators Critical indicators and recorders are typically physically located on panels/cabinets and are included as separate components; however, seismic evaluation of the instrument indication may be included in the panel/cabinet seismic evaluation (rule-of-the-box).
14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 14 of 63 3.1.6 Phase 2 and Phase 3 Piping Connections Item 2 in Section 3.1 above notes that the scope of equipment in the ESEL includes "- FLEX connections necessary to implement the Peach Bottom OIP (Ref. 3) including subsequent 6 month updates through August 2014 as described in Section 2."  Item 3 in Section 3.1 also notes that "The scope of components assumes the credited FLEX connection modifications are implemented, and are limited to those required to support a single FLEX success path (i.e., either "Primary" or "Back-up/Alternate")." Item 5 in Section 3.1 above goes on to explain that "Piping, cabling, conduit, HVAC, and their supports" are excluded from the ESEL scope in accordance with EPRI 3002000704 (Ref. 2). Therefore, piping and pipe supports associated with FLEX Phase 2 and Phase 3 connections are excluded from the scope of the ESEP evaluation. However, any active valves in FLEX Phase 2 and Phase 3 connection flow path are included in the ESEL.
3.2 Justification for use of Equipment that is not the Primary Means for FLEX Implementation All equipment used for FLEX implementation on the PBAPS ESEL are the primary path.
 
14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 15 of 63 4 GROUND MOTION RESPONSE SPECTRUM (GMRS) 4.1 Plot of GMRS Submitted by the Licensee In accordance with Section 2.4.2 of the SPID (Ref. 14), the licensing design basis definition of the SSE control point for Peach Bottom is used for comparison to the GMRS. Ref. 6 lists the Peach Bottom SSE as being located at 136 feet MSL with a PGA of 0.12g. The GMRS per the March 2014 submittal report (Ref. 6) is tabulated and graphed below: Table 4.1-1 Peach Bottom GMRS (5% Damping) Freq. (Hz) GMRS (unscaled, g)Freq. (Hz)GMRS (unscaled, g) 0.1 0.007 40.256 0.125 0.008 50.332 0.15 0.010 60.387 0.2 0.014 70.441 0.25 0.017 80.490 0.3 0.020 90.536 0.35 0.024 100.581 0.4 0.027 12.50.659 0.5 0.034 150.727 0.6 0.041 200.844 0.7 0.047 250.924 0.8 0.052 300.967 0.9 0.057 350.961 1 0.062 400.914 1.25 0.076 500.730 1.5 0.089 600.549 2 0.124 700.461 2.5 0.143 800.426 3 0.179 900.410 3.5 0.219 1000.402  Figure 4.1-1 Peach Bottom GMRS (5% Damping) 0.00 0.20 0.40 0.60 0.80 1.00 1.20 0.1 1 10 100 Acceleration (g)
Frequency (Hz) 14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 16 of 63 4.2 Comparison to SSE As identified in the March 2014 submittal report (Ref. 6), the GMRS exceeds the SSE in the 1-10 Hz range as shown in the table and graph below: Table 4.2-1 Peach Bottom GMRS vs. SSE (5% Damping) Freq. (Hz)GMRS (g)SSE (g)1 0.062 0.11 1.25 0.076 0.13 1.5 0.089 0.16 2 0.124 0.19 2.5 0.143 0.21 3 0.179 0.22 3.5 0.219 0.22 4 0.256 0.22 5 0.332 0.21 6 0.387 0.21 7 0.441 0.20 8 0.490 0.20 9 0.536 0.19 10 0.581 0.19 Figure 4.2-1 Peach Bottom GMRS vs. SSE (5% Damping) 0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 11 0Spectral Acceleration , 5% damping (g)
Frequency (Hz)
GMRSSSE 14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 17 of 63 5 REVIEW LEVEL GROUND MOTION (RLGM) 5.1 Description of RLGM Selected The RLGM for Peach Bottom was determined in accordance with Section 4 of EPRI 30020000704 (Ref. 2) as being derived by linearly scaling the Peach Bottom SSE by the maximum ratio of the GMRS/SSE between the 1 and 10 Hertz range, with an upper bound of 2.0. The ratio between the GMRS and SSE at 5% damping is tabulated below.
 
Table 5.1-1 Ratio between GMRS and SSE (5% Damping) Freq. (Hz) GMRS (g)SSE (g)Ratio GMRS/SSE1 0.0619 0.11 0.56 1.25 0.0759 0.13 0.58 1.5 0.0893 0.16 0.56 2 0.124 0.19 0.65 2.5 0.143 0.21 0.68 3 0.179 0.22 0.81 3.5 0.219 0.22 1.00 4 0.256 0.22 1.16 5 0.332 0.21 1.58 6 0.387 0.21 1.84 7 0.441 0.2 2.21 8 0.49 0.2 2.45 9 0.536 0.19 2.82 10 0.581 0.19 3.06 The maximum ratio between the 5% damping GMRS and horizontal SSE occurs at 10 Hz and equals 3.06. Based on Section 4 of EPRI 30020000704 (Ref. 2), the RLGM derived by linearly scaling the SSE need not exceed 2 x SSE; therefore the upper bound ratio of 2.0 applies. 
 
The resulting RLGM based on increasing the horizontal SSE by the maximum ratio of 2.0 is plotted below. Note that the RLGM PGA is 0.24g.
 
14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 18 of 63 Table 5.1-2 Peach Bottom RLGM (5% Damping) Freq. (Hz) RLGM (g)  Freq. (Hz) RLGM (g) 1 0.22 12.5 0.36 1.25 0.26 15 0.36 1.5 0.32 20 0.34 2 0.38 25 0.32 2.5 0.42 30 0.30 3 0.44 35 0.30 3.5 0.44 40 0.30 4 0.44 50 0.28 5 0.42 60 0.28 6 0.42 70 0.26 7 0.40 80 0.24 8 0.40 90 0.24 9 0.38 100 0.24 10 0.38    Figure 5.1-1 Peach Bottom RLGM, GMRS & SSE (5% Damping) 0 0.2 0.4 0.6 0.8 1 1.2 1 10 100Spectral Acceleration (g)
Frequency (Hz)Original SSE RLGM GMRS 14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 19 of 63 5.2 Method to Estimate ISRS The method used to derive the ESEP in-structure response spectra (ISRS) was to scale the existing SSE-based ISRS obtained from Peach Bottom Specification 11187-G-14, "General Project Requirements for Seismic Design and Analysis of Equipment and Equipment Supports for the Peach Bottom Atomic Power Station Units 2 & 3" (Ref. 18) and PBAPS calculation PS-0907, "SQUG - Radwaste/Turbine Building A46 Spectra" (Ref. 19), by the maximum ratio of 2.0. The scaled ISRS was determined for all buildings and elevations where ESEL items are located at Peach Bottom.
 
14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 20 of 63 6  SEISMIC MARGIN EVALUATION APPROACH It is necessary to demonstrate that ESEL items have sufficient seismic capacity to meet or exceed the demand characterized by the RLGM. The seismic capacity is characterized as the peak ground acceleration (PGA) for which there is a high confidence of a low probability of failure (HCLPF). The PGA is associated with a specific spectral shape, in this case the 5%-damped RLGM spectral shape. The HCLPF capacity must be equal to or greater than the RLGM PGA. The criteria for seismic capacity determination are given in Section 5 of EPRI 3002000704 (Ref. 2).
There are two basic approaches fo r developing HCLPF capacities: 1. Deterministic approach using the conserva tive deterministic failure margin (CDFM) methodology of EPRI NP-6041, A Methodology for Assessment of Nuclear Power Plant Seismic Margin (Revision 1) (Ref. 7). 2. Probabilistic approach using the fragility analysis methodology of EPRI TR-103959, Methodology for Developing Seismic Fragilities (Ref. 8). For Peach Bottom, the deterministic approach using the CDFM methodology of EPRI NP-6041 (Ref. 7) was used to determine HCLPF capacities.
6.1 Summary of Methodologies Used Peach Bottom applied the Deterministic Approach (i.e. Method 1 from the previous section) to all items on the ESEL. The screening walkdowns used the screening tables from Chapter 2 of EPRI NP-6041 (Ref. 7). The walkdowns were conducted by engineers who as a minimum attended the SQUG Walkdown Screening and Seismic Evaluation Training Course. The walkdowns were documented on Screening Evaluation Work Sheets from EPRI NP-6041 (Ref. 7). Anchorage capacity calculations used the CDFM criteria from EPRI NP-6041 (Ref. 7) with Peach Bottom specific allowables and material strength s used as applicable. Seismic demand was the RLGM provided in Table 5.1-2 and Figure 5.1-1.
6.2 HCLPF Screening Process The peak RLGM (amplified PGA) For Peach Bottom equals 0.44 g (Table 5.1-2). The screening tables in EPRI NP-6041 (Ref. 7) are based on ground peak spectral accelerations of 0.8g and 1.2g. All Peach Bottom ESEL components were screened against either the caveats of the <0.8g column (lane 1) or the 0.8g-1.2g column (lane 2) of Table 2-4 of NP-6041 (Ref. 7). Screening based on lane 1 with the RLGM spectral shape yields an equivalent HCLPF of 0.44g PGA (witness 0.8g/0.44g*0.24g PGA = 0.44g PGA). Screening based on lane 2 with the RLGM spectral shape yields an equivalent HCLPF of 0.65g PGA (witness 1.2g/0.44g*0.24g PGA = 0.65g PGA).
14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 21 of 63 A number of components were located above 40 feet from grade. For components located 40 feet above grade, screening based on ground peak spectral acceleration is not applicable and additional consideration is required. In accordance with Appendix B of EPRI 1019200 (Ref. 20), components that are above 40 feet from grade and have corresponding ISRS at the base of component not in exceedance of 1.2g in the component frequency range of interest may be screened using the caveats of the 1st screening column, and components that are above 40 feet from grade and have corresponding ISRS at the base of component not in exceedance of 1.8g in the component frequency range of interest may be screened using the caveats of the 2nd
 
screening column. The screening of anchorage for non-valve components was evaluated either by SRT judgment or simple analysis. For components whose anchorage could not readily be screened by SRT judgment or simple analysis, CDFM HCLPF calculations (Ref. 9) were performed. This is documented in Attachments C and D. Per Ref 9.a, the seismic spectra for the Reactor Building (RB) are scaled from the original design spectra, which were based on an OBE seismic input and thus were based on a structural damping which is conservative for CDFM analysis (See Ref. 17). Under scaled SSE loading, a level of critical damping of 5% is appropriate for this structure. Based on NP-6041 Appendix Q and consideration of a structural damping level of 5%, it is shown in Ref. 9.b that Reactor Building elevations 135' and 165' may be addressed for CDFM purposes with peak spectral accelerations of 1.0g and 1.4g respectively. These spectral peak values are thus used for the purposes of equipment qualification as per NP-6041 Table 2-4.
 
6.3 Seismic Walkdown Approach 6.3.1 Walkdown Approach Walkdowns for Peach Bottom were performed in accordance with the criteria provided in Section 5 of EPRI 3002000704 (Ref. 2), which refers to EPRI NP-6041 (Ref. 7) for the Seismic Margin Assessment process. Pages 2-26 through 2-30 of EPRI NP-6041 (Ref. 7) describe the seismic walkdown criteria, including the following key criteria. "The SRT [Seismic Review Team] should "w alk by" 100% of all components which are reasonably accessible and in non-radioactive or low radioactive environments. Seismic capability assessment of components which are inaccessible, in high-radioactive environments, or possibly within contaminated containment, will have to rely more on alternate means such as photographic inspection, more reliance on seismic reanalysis, and possibly, smaller inspection teams and more hurried inspections. A 100% "walk by" does not mean complete inspection of each component, nor does it mean requiring an electrician or other technician to de-energize and open cabinets or panels for detailed inspection of all components. This walkdown is not intended to be a QA or QC review or a review of the adequacy of the component at the SSE level.
14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 22 of 63 If the SRT has a reasonable basis for assuming that the group of components are similar and are similarly anchored, then it is only necessary to inspect one component out of this group. The "similarity-basis" should be developed before the walkdown during the seismic capability preparatory work (Step 3) by reference to drawings, calculations or specifications. The one component or each type which is selected should be thoroughly inspected which probably does mean de-energizing and opening cabinets or panels for this very limited sample. Generally, a spare representative component can be found so as to enable the inspection to be performed while the plant is in operation. At least for the one component of each type which is selected, anchorage should be thoroughly inspected. The walkdown procedure should be performed in an ad hoc manner. For each class of components the SRT should look closely at the first items and compare the field configurations with the construction drawings and/or specifications. If a one-to-one correspondence is found, then subsequent items do not have to be inspected in as great a detail. Ultimately the walkdown becomes a "walk by" of the component class as the SRT becomes confident that the construction pattern is typical. This procedure for inspection should be repeated for each component class; although, during the actual walkdown the SRT may be inspecting several classes of components in parallel. If serious exceptions to the drawings or questionable construction practices are found then the system or component class must be inspected in closer detail until the systematic deficiency is defined. The 100% "walk by" is to look for outliers, lack of similarity, anchorage which is different from that shown on drawings or prescribed in criteria for that component, potential SI
[Seismic Interaction
] problems, situations that are at odds with the team members' past experience, and any other areas of serious seismic concern. If any such concerns surface, then the limited sample size of one component of each type for thorough inspection will have to be increased. The increase in sample size which should be inspected will depend upon the number of outliers and different anchorages, etc., which are observed. It is up to the SRT to ultimately select the sample size since they are the ones who are responsible for the seismic adequacy of all elements which they screen from the margin review. Appendix D gives guidance for sampling selection." The Peach Bottom walkdowns included as a minimum a 100% walk-by of all items on the ESEL except as noted in Section 7. Any previous walkdown information that was relied upon for SRT judgment is documented in Section 6.3.2.
 
EPRI 3002000704 (Ref. 2) page 5-4 limits the ESEP seismic interaction reviews to "nearby block walls" and "piping attached to tanks" which are reviewed "to address the possibility of failures due to differential displacements."  Other potential seismic interaction evaluations are "deferred to the full seismic risk evaluations performed in accordance with EPRI 1025287 (Ref. 14)."
 
14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 23 of 63 6.3.2 Application of Previous Walkdown Information The seismic walkdowns for Peach Bottom included as a minimum a walk-by of all the components on the ESEL with the exception of the items which are discussed in Section 7. Previous seismic walkdowns were used to support the ESEP seismic evaluations. Some of the components on the ESEL were included in the NTTF Recommendation 2.3 seismic walkdowns (Ref. 16). Photos taken during the NTTF R2.3 seismic walkdowns (Ref. 16), although available to the SRT during the ESEP walkdowns, were not necessary to the SRT at Peach Bottom. A-46 and IPEEE notes were available to the SRT and were used where appropriate to reduce the number of equipment items that needed to be opened and evaluate equipment that were not completely accessible to the SRT. Several ESEL items were previously walked down during the Peach Bottom 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.
6.3.3 Significant Walkdown Observations Consistent with that guidance from NP-6041 (Ref. 7), no significant outliers or anchorage concerns were identified during the Peach Bottom Atomic Power Station seismic walkdowns.
* Several block walls were identified in the proximity of ESEL equipment. These block walls were assessed for their structural adequacy to withstand the seismic loads resulting from the RLGM. For any cases where the block wall represented the HCLPF failure mode for an ESEL item, it is noted in the tabulated HCLPF values described in Section 6.6.
 
6.4 HCLPF Calculation Process ESEL items were evaluated using the criteria in EPRI NP-6041 (Ref. 7). Those evaluations included the following steps:
EPRI 3002000704 (Ref. 2) page 5-4 limits the ESEP seismic interaction reviews to "nearby block walls" and "piping attached to tanks" which are reviewed "to address the possibility of failures due to differential displacements."  Other potential seismic interaction evaluations are "deferred to the full seismic risk evaluations performed in accordance with EPRI 1025287 (Ref. 14)."
 
14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 24 of 63
* Performing seismic capability walkdowns for equipment to evaluate the equipment installed plant conditions
* Performing screening evaluations using the screening tables in EPRI NP-6041 (Ref. 7) as described in Section 6.2 and
* Performing HCLPF calculations considering various failure modes that include both structural failure modes (e.g. anchorage, load path etc.) and functional failure modes. All HCLPF calculations were performed using the CDFM methodology and are documented in Peach Bottom calculations (Ref. 9).
6.5 Functional Evaluation of Relays A HCLPF evaluation is performed for all relays and switches which may negatively "seal in" or "lock out" on the PBAPS ESEL. For relay evaluations, NP-6041 Appendix Q describes the following steps:
* Calculate in-cabinet response spectra (ICRS)
* Establish a clipping factor to be applied to the ICRS
* Determine a relay's Generic Equipment Ruggedness Spectra (GERS) Capacity
* Establish adjustment factors to convert the relay's GERS capacity to a CDFM level
* Compare clipped-peak and Zero Period Acceleration (ZPA) demands to the GERS capacity/test capacity HCLPF capacities for the relays are calculated using the procedure described above. The switch HCLPF value was determined by using existing test data in lieu of GERS. HCLPFs are calculated in 14Q4233-CAL-004 (Ref. 9) and are presented in Attachment C and D. Attachments C and D identify four relays for which operator action will be undertaken to reset the relay if necessary. Section 8.2 identifies two relays which are found to have a HCLPF capacity below the RLGM, and for which additional modifications, tests or analysis will be performed by the site.
6.6 Tabulated ESEL HCLPF Values (including key failure modes)
Tabulated ESEL HCLPF values including the key failure modes are included in Attachment C for Unit 2 and in Attachment D for Unit 3 items.
* For items screened out using NP-6041 (Ref. 7) screening tables, the screening level is provided as ">RLGM" and the failure mode is listed as "Screened."
 
14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 25 of 63
* For items where anchorage controls the HCLPF value, the HCLPF value is listed in the table and the failure mode is noted as "Anchorage."
* For items where block wall interaction controls the HCLPF value, the HCLPF value is listed in the table and the failure mode is noted as "Block Wall Interaction."
* For items where a relay or switch HCLPF controls, the relay or switch HCLPF value is listed in the table and the failure mode is listed as "Functional Failure".
* For items where equipment capacity based upon the screening lane values of Table 2-4 of EPRI NP-6041 (Ref. 7) controls the HCLPF value (e.g. anchorage, block wall, or relay HCLPF capacity exceeds the equipment capacity derived from screening), the screening lane HCLPF value is listed in the table and the failure mode is noted as "equipment capacity." Based on NP-6041 Table 2-4 Lane 1, this limit is equal to 0.44g for items below 40 feet above grade.
14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 26 of 63 7 INACCESSIBLE ITEMS 7.1 Identification of ESEL Items Inaccessible for Walkdowns Thirty three ESEL items were not accessible to the SRT during the ESEP walkdowns at Peach Bottom Atomic Power Station. A description of circumstances and disposition for these items is provided below. Table 7.1-1 Items Inaccessible for Walkdowns ID Description Resolution3AT540 3AT545 3BT540 3BT545 3CT540 3CT545 3GT540 3GT545 3KT540 3KT545 Nitrogen accumulatorsTh ese items are located in the Unit 3 Drywell and Isolation Valve Compartment. Since Unit 3 did not have a sch eduled outage in 2014, these items were inaccessible.
 
The Unit 2 Drywell and Isolation Valve Compartment were walked down during the Unit 2 scheduled outage in October 2014. Equivalent items in Unit 2 were seen to be rugged and well secured. Based on these observations, drawings and overall comparison of similar items between the two units, these items are expected to be similar to those in the Unit 2, and judged by SRT to be acceptable on that basis, including consideration of seismic interaction with block walls and piping attached to tanks. MO-3-13-021 Motor-operated valveRV-3-02-071A RV-3-02-071B RV-3-02-071C RV-3-02-071G RV-3-02-071K Relief valves2AT545 2CT545 2GT545 AccumulatorsSRT did not see theseaccumulators due to physical accessibility restriction in the drywell, but did walkdown the 2BT545, 2KT545 accumulators (among others not on the ESEL) and reviewed drawings. All accumulators seen were similar, and were
 
very ruggedly supported as shown in the SEWS photos, and are judged by SRT to be acceptable on that basis, including consideration of seismic interaction with block walls and piping attached to tanks.
00E072 Reheat coilThe reheat coilwas not directly visible, but was located by drawings and the site escort to be in line with large overhead ductwork. The Reheat Coil was judged by SRT to be adequately secured to the ductwork and screen out, including consideration of seismic interaction with block walls. 2AC270(K3A) 2BC270(K3B)
 
2AC270(K3C) 2BC270(K3D) 3AC270(K3A) 3BC270(K3B) Component relaysThese relays are inside control room cabinets2AC270, 2BC270, 3AC270 and 3BC270. These cabinets could not be opened during operation; therefore the relays were not walked down. However, the site provided a test report for the cabinet for the purpose of 14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 27 of 63 3AC270(K3C) 3BC270(K3D) component relay evaluation. Relays were analyzed to be adequate per 14Q4233-CAL-004. Cabinet was walked down and there are no block wall interactions in the vicinity. PO-0-40W-00016 PO-0-40W-00019-02 PO-0-40W-00808 Damper actuatorsThese damper actuatorswere not seen by SRT. They are understood to be inside an AHU unit that is connected to OAV034. All damper actuators that were observed during the walkdowns were either model D-251 or D-9504. The SRT judged these light weight actuators to be seismically rugged and adequately supported and they were screened out, including consideration of seismic interaction with block walls. AO-2-07B-2511 AO-2-07B-80290 Air-operated valvesThese valves are located on the top of the Torus. The area in the vicinity of these valves was walked by and valves were located by site escort with SRT. However, the valves were not directly accessible to SRT due to piping interferences. Due to a review of drawings and the similarity of other valves between units, these valves screened out, including consideration of seismic interaction with block walls.
7.2 Planned Walkdown / Evaluation Schedule / Close Out Since all items that were inaccessible during the ESEP were resolved by alternative means to the satisfaction of the SRT as discussed in Table 7.1-1 above, no additional walkdowns are required.
14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 28 of 63 8 ESEP CONCLUSIONS AND RESULTS 8.1 Supporting Information Peach Bottom Atomic Power Station has performed the ESEP as an interim action in response to the NRC's 50.54(f) letter (Ref. 1). It was performed using the methodologies in the NRC endorsed guidance in EPRI 3002000704 (Ref. 2). The ESEP provides an important demonstration of seismic margin and expedites plant safety enhancements through evaluations and potential near-term modifications of plant equipment that can be relied upon to protect the reactor core following beyond design basis seismic events. The ESEP is part of the overall Peach Bottom response to the NRC's 50.54(f) letter (Ref. 1). On March 12, 2014, NEI submitted to the NRC results of a study (Ref. 11) of seismic core damage risk estimates based on updated seismic hazard information as it applies to operating nuclear reactors in the Central and Eastern United States (CEUS). The study concluded that "site-specific seismic hazards show that there has not been an overall increase in seismic risk for the fleet of U.S. plants" based on the re-evaluated seismic hazards. As such, the "current seismic design of
 
operating reactors continues to provide a safety margin to withstand potential earthquakes exceeding the seismic design basis."  The NRC's May 9, 2014 NTTF 2.1 Screening and Prioritization letter (Ref. 13) concluded that the "fleetwide seismic risk estimates are consistent with the approach and results used in the Gl-199 safety/risk assessment."  The letter also stated that "As a result, the staff has confirmed that the conclusions reached in Gl-199 safety/risk assessment remain valid and that the plants can continue to operate while additional evaluations are conducted." An assessment of the change in seismic risk for Peach Bottom was included in the fleet risk evaluation submitted in the March 12, 2014 NEI letter (Ref. 11) therefore, the conclusions in the NRC's May 9 letter (Ref. 13) also apply to Peach Bottom. In addition, the March 12, 2014 NEI letter (Ref. 11) provided an attached "Perspectives on the Seismic Capacity of Operating Plants," which (1) assessed a number of qualitative reasons why the design of SSCs inherently contain margin beyond their design level, (2) discussed industrial seismic experience databases of performance of industry facility components similar to nuclear SSCs, and (3) discussed earthquake experience at operating plants. The fleet of currently operating nuclear power plants was designed using conservative practices, such that the plants have significant margin to withstand large ground motions safely. This has been borne out for those plants that have actually experienced significant earthquakes. The seismic design process has inherent (and intentional) conservatisms which result in significant seismic margins within structures, systems and components (SSCs). These conservatisms are reflected in several key aspects of the seismic design process, including:
14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 29 of 63
* Safety factors applied in design calculations
* Damping values used in dynamic analysis of SSCs
* Bounding synthetic time histories for in-structure response spectra calculations
* Broadening criteria for in-structure response spectra
* Response spectra enveloping criteria typically used in SSC analysis and testing applications
* Response spectra based frequency domain analysis rather than explicit time history based time domain analysis
* Bounding requirements in codes and standards
* Use of minimum strength requirements of structural components (concrete and steel)
* Bounding testing requirements, and
* Ductile behavior of the primary materials (that is, not crediting the additional capacity of materials such as steel and reinforced concrete beyond the essentially elastic range, etc.). These design practices combine to result in margins such that the SSCs will continue to fulfill their functions at ground motions well above the SSE. The intent of the ESEP is to perform an interim action in response to the NRC's 50.54(f) letter (Ref. 1) to demonstrate seismic margin through a review of a subset of the plant equipment that can be relied upon to protect the reactor core fo llowing 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 plant's SSE rather than the actual GMRS. To more fully characterize the risk impacts of the seismic ground motion represented by the GMRS on a plant specific basis, a more detailed seismic risk assessment (SPRA or risk-based SMA) is to be performed in accordance with EPRI 1025287 (Ref. 14). As identified in the Peach Bottom Seismic Hazard and GMRS submittal (Ref. 6), Peach Bottom 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. Peach Bottom will complete that evaluation in accordance with the schedule identified in NEI's letter dated April 9, 2013 (Ref. 12) and endorsed by the NRC in their May 7, 2013 letter (Ref. 13).
 
14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 30 of 63 8.2 Identification of Planned Modifications The following two relays were identified with HCLPF capacities below the RLGM. These relays will be further evaluated and may require modification.
Component ID Resolution 2-13A-K033 See below for a list of potential solutions. 3-13A-K033 See below for a list of potential solutions.
Solutions which may be considered in qu alifying 2-13A-K033 and 3-13A-K033 include:
* Stiffen or replace host cabinets 20C033 and 30C033. A modification which significantly stiffens the host cabinet for each of the subject relays would lower the seismic demand.
* Relocate the subject relays to a mo re seismically favorable location.
* Replace the subject relays with a compatible relay model.
* Determine a higher capacity by shake table testing of this relay model.
* Reduce seismic demand through analysis or reduction of existing conservatisms
* Risk analysis 8.3 Modification Implementation Schedule The modification implementation schedule will be included in the transmittal letter from Exelon to the NRC for this report.
8.4 Summary of Regulatory Commitments Regulatory commitments for Peach Bottom modification implementation will be included in the transmittal letter from Exelon to the NRC for this report.
14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 31 of 63 9 REFERENCES
: 1. NRC (E Leeds and M Johnson) Letter to All Power Reactor Licensees et al., "Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(f) Regarding Recommendations 2.1, 2.3 and 9.3 of the Near-Term Task Force Review of Insights from the Fukushima Dai-Ichi Accident," March 12, 2012
: 2. Seismic Evaluation Guidance: Augmented Approach for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1 - Seismic. EPRI, Palo Alto, CA: May 2013. 3002000704
: 3. Peach Bottom Letters
: a. NRC Letter RS-13-024 from Peach Bottom (ML13059A305), "Overall Integrated Plan in Response to March 12, 2012 Commission Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events (Order Number EA-12-049)", February 28, 2013
: b. NRC Letter RS-13-127 from Peach Bottom (ML13246A412), "First Six-Month Status Report in Response to March 12, 2012 Commission Order Modifying Licenses with Regard to Requirements for Mitigation St rategies for Beyond-Design-Basis External Events (Order Number EA-12-049)", August 28, 2013
: c. NRC Letter RS-14-014 from Peach Bottom (ML14059A222), "Second Six-Month Status Report in Response to March 12, 2012 Commission Order Modifying Licenses with Regard to Requirements for Mitigation St rategies for Beyond-Design-Basis External Events (Order Number EA-12-049)", February 28, 2014
: d. NRC Letter RS-14-212 from Peach Bottom (ML14241A252), "Third Six-Month Status Report in Response to March 12, 2012 Commission Order Modifying Licenses with Regard to Requirements for Mitigation St rategies for Beyond-Design-Basis External Events (Order Number EA-12-049)", August 28, 2014
: 4. Peach Bottom Station Transmittal of Design Information to Stevenson & Associates, Tracking No: TODI AR#2397100-02, "PB Flex Strategy Rev. 1", December 9, 2014
: 5. Peach Bottom Station Transmittal of Design Information to Stevenson & Associates, Tracking No: PB 1570792-76, Input relating to Relays and Switches, December 4, 2014
: 6. Seismic Hazard and Screening Report in Response to the 50.54(f) Information Request Regarding Fukushima Near-Term Task Force Recommendation 2.1: Seismic for Peach Bottom Atomic Power Station dated 3/31/14, Correspondence No. RS-14-071 (Exelon Report EXLNPB056-PR-001, Revision 1)
: 7. A Methodology for Assessment of Nuclear Power Plant Seismic Margin, Rev. 1, August 1991, Electric Power Research Institute, Palo Alto, CA. EPRI NP-6041
: 8. Methodology for Developing Seismic Fragilities, August 1991, EPRI, Palo Alto, CA. 1994, TR-103959
: 9. Peach Bottom HCLPF Calculations for the ESEP project 14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 32 of 63
: a. 14Q4233-CAL-001, Rev. 0, Generation of In-Structure Response Spectra for use in ESEP Evaluations
: b. 14Q4233-CAL-002, Rev. 0, HCLPF Analysis for ESEP Evaluations for PBAPS
: c. 14Q4233-CAL-003, Rev. 0, ESEP Block Wall HCLPFs
: d. 14Q4233-CAL-004, Rev. 0, ESEP HCLPFs for Relays
: 10. Nuclear Regulatory Commission, NUREG/CR-0098, Development of Criteria for Seismic Review of Selected Nuclear Power Plants, published May 1978
: 11. Nuclear Energy Institute (NEI), A. Pietrangelo, Letter to D. Skeen of the USNRC, "Seismic Core Damage Risk Estimates Us ing the Updated Seismic Hazards for the Operating Nuclear Plants in the Central and Eastern United States", March 12, 2014
: 12. Nuclear Energy Institute (NEI), A. Pietrang elo, Letter to D. Skeen of the USNRC, "Proposed Path Forward for NTTF Recommendation 2.1: Seismic Reevaluations", April 9, 2013  13. NRC (E. Leeds) Letter to All Power Reacto r Licensees et al. (ML14111A147), "Screening and Prioritization Results Regarding Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(F) Regarding Seismic Hazard Re-Evaluations for Recommendation 2.1 of the Near-Term Task Force Review of Insights From the Fukushima Dai-Ichi Accident," May 9, 2014
: 14. Seismic Evaluation Guidance: Screening, Prioritization and Implementation Details (SPID) for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1: Seismic. EPRI, Palo Alto, CA: February 2013. 1025287
: 15. NRC (E. Leeds) Letter to NEI (J Pollock) (ML13106A331), "Electric Power Research Institute Final Draft Report xxxxxx, "Seismic Evaluation Guidance: Augmented Approach for the Resolution of Fukushima Near-Term Task Fo rce Recommendation 2.1: Seismic," as an Acceptable Alternative to the March 12, 2012, Information Request for Seismic Reevaluations," May 7, 2013
: 16. PBAPS NTTF 2.3 Seismic Walkdown Submittals for Unit 2 and Unit 3, dated 11/20/2014
: 17. Peach Bottom Atomic Power Station Updated Final Safety Analysis Report (UFSAR)
Appendix C, Rev.24
: 18. Peach Bottom Specification 11187-G-14 Rev.0, General Project Requirements for Seismic Design and Analysis of Equipment and Equipment Supports for the Peach Bottom Atomic Power Station Units 2 & 3
: 19. Peach Bottom Calculation PS-0907 Rev.0, SQUG - Radwaste/Turbine Building A46 Spectra 14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 33 of 63
: 20. EPRI Technical Report (TR) 1019200, "Seismic Fragility Applications Guide Update," December 2009
: 21. Peach Bottom Drawings
: a. M-315 Sheet 1, Rev. 65, P&I Diagram Emergency Service Water and High Pressure Service Water Systems
: b. M-333 Sheet 1, Rev. 57, P&ID Diagram Instrument Nitrogen (Unit 2)
: c. M-333 Sheet 2, Rev. 58, P&ID Diagram Instrument Nitrogen (Unit 3)
: d. M-351 Sheet 1, Rev. 78, P&ID Nuclear Boiler
: e. M-351 Sheet 2, Rev. 70, P&ID Nuclear Boiler
: f. M-351 Sheet 3, Rev. 74, P&ID Nuclear Boiler
: g. M-351 Sheet 4, Rev. 69, P&ID Nuclear Boiler
: h. M-359 Sheet 1, Rev. 50, P&ID Diagram Reactor Core Isolation Cooling
: i. M-359 Sheet 2, Rev. 48, P&ID Diagram Reactor Core Isolation Cooling
: j. M-360 Sheet 1, Rev. 56, R.C.I.C. Pump Turbine Details
: k. M-360 Sheet 2, Rev. 54, R.C.I.C. Pump Turbine Details
: l. M-360 Sheet 3, Rev. 47, R.C.I.C. Pump Turbine Details Lube Oil and Control System Unit 2 m. M-360 Sheet 4, Rev. 39, R.C.I.C. Pump Turbine Details Lube Oil and Control System Unit 3 n. M-361 Sheet 1, Rev. 82, P&I Diagram Residual Heat Removal Sys (Unit 2)
: o. M-361 Sheet 2, Rev. 68, P&I Diagram Residual Heat Removal Sys (Unit 2)
: p. M-361 Sheet 3, Rev. 70, P&I Diagram Residual Heat Removal Sys (Unit 3)
: q. M-361 Sheet 4, Rev. 072, Residual Heat Removal System (Unit 3)
: r. M-367 Sheet 1, Rev. 85, P&ID Diagram Co ntainment Atmospheric Control System (Unit 2) s. M-367 Sheet 2, Rev. 76, P&ID Diagram Co ntainment Atmospheric Control System (Unit 3) t. M-372 Sheet 1, Rev. 62, P&ID Diagram Co ntainment Atmosphere Dilution System
: u. M-384 Sheet 1, Rev. 39, P&I Diagram Control Room HVAC
: v. M-384 Sheet 2, Rev. 6, P&I Diagram Control Room HVAC 
: w. M-384 Sheet 3, Rev. 6, P&I Diagram Control Room HVAC
: x. M-399 Sheet 1, Rev. 32, P&I Diagram Emergency Switchgear, Battery Room, Laboratory Supply & Exhaust
: y. M-399 Sheet 2, Rev. 4, P&I Diagram Emergency Switchgear, Battery Room, Laboratory Supply & Exhaust
: z. M-399 Sheet 3, Rev. 2, P&I Diagram Emergency Switchgear, Battery Room, Laboratory Supply & Exhaust aa. M-399 Sheet 4, Rev. 5, P&I Diagram Emergency Switchgear, Battery Room, Laboratory Supply & Exhaust
: 22. 14Q4233-RPT-003, Rev.1, Validation of Expedited Seismic Equipment List 
: 23. Seismic Qualification Utility Group (SQUG), Generic Implementation Procedure (GIP) for Verification of Nuclear Plant Equipment, Revision 3A
: 24. NRC Order Number EA-12-049 (ML12054A735), "Issuance of Order to Modify Licenses with Regard to Requirements for Mitigation Strategies for Beyond-D esign-Basis External Events," dated March 12, 2012.
 
14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 34 of 63
: 25. Email to S&A from Ms. Tracey L. Gallagher (Exelon), LRC-009, Confirmation of "no changes" between February 2013 and August 2014 Flex Strategies, December 4, 2014
: 26. Email to S&A from Ms. Tracey L. Gallagher (Exelon), LRC-010, Confirmation from "Ops" of "no significant changes" between February 2013 and August 2014 Flex Strategies, December 4, 2014
: 27. Email to S&A from Ms. Tracey L. Gallagher (Exelon), LRC-025, Inputs to RPT-004 including that "Flex Phase 1 and 2 strategy will provide sufficient capability such that no additional Phase 3 strategies are required", December 17, 2014
 
14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 35 of 63
 
Attachment A PBAPS Unit 2 and Common ESEL
 
14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 36 of 63 Equipment ID Description Equipment Normal State Equipment Desired State Notes 00C133 Panel Energized Energized  00E068 CR Fresh Air Supply Preheat Coil Standby Standby  00E072 CR Vent Reheat Coil Standby Standby  00F039 CR Fresh Air Supply Roll Filter Standby Standby  00F043 OA Damper Emergency Switchgear & Battery Room Vent Supply Roll Filter Standby Standby 00T116 CAD Liquid Nitrogen Storage Tank Standby In Service Passive component 0AE073 OA Damper Emergency Switchgear &
Battery Room Vent Supply Heat Coil Standby Standby 0AF041 A Train HEPA Filter Standby Standby Passive Component 0AF042 A Train Charcoal Filter Standby Standby Passive Component 0AF050 A Train HEPA Filter Standby Standby Passive Component 0AV030 CR Room Emergency Vent Fan Standby Energized  0AV034 Emergency Switch Gear and Battery Room Supply Fan Standby Operating 0AV036 Battery Room Exhaust Fan Standby Operating  20C003 Reactor and Containment Cooling and Isolation Panel Energized Energized 20C004C RCIC Control Panel Energized Energized  20C005A Reactor Manual Control Panel Energized Energized  20C012 Plant Services Console Energized Energized  20C018 Panel Energized Energized  20C019 Panel Energized Energized Contains power supply for PT-2-13-068 20C032 Panel Energized Energized  20C033 Panel Energized Energized  20C034 RCIC Relay Panel Energized Energized  20C035 Panel Energized Energized  20C041 Panel Energized Energized  20C095 RCIC Instrument Rack Energized Energized  20C144 Panel Energized Energized  20C722A Accident Monitoring Instrumentation Panel Energized Energized 20C722B Panel Energized Energized  20C818 Reactor Water Level/Pressure Component Cabinet Energized Energized 20D021 (2PPA) 125V DC Station Distribution Energized Energized 20D023 (2PPC) 125V DC Station Distribution Energized Energized  20D024 Distribution Panel Energized Energized  20D037 Uninterruptable Power Supply Static Inverter Energized Energized Powers vital instrument bus during Phase 1 20D039 RCIC Barometric CDSR Vacuum Pump
 
Starter Standby Energized 20D040 RCIC Barometric CDSR Cond Pump
 
Starter Standby Energized 20P036 RCIC Pump Standby Operating  20P046 RCIC Barometric CDSR Vacuum Pump Standby Operating  20P048 RCIC Barometric CDSR Condensate
 
Pump Standby Operating 20P340 RCIC Turbine Driven Lube Oil Pump Standby Operating  20S038 RCIC Turbine Standby Operating Controlled via Included Governor Valve and Trip & Throttle Valve 20S315 Static Inverter Man Bypass/Isolation Switch Energized Bypassed and
 
Isolated  Phase 1 power is from inverter, Phase 2 is from 480/120V AC transformer 20S354 Load Center E-124/E324 Transfer Switch Standby Energized 20S700 Battery Charger Panel 2AD003 Transfer Switch 20S700 Energized Energized 20S701 Transfer Switch Energized Energized  20S703 Transfer Switch Energized Energized  20X133 Transformer Energized Energized  20X135 20Y035 Transformer Energized Energized  20X150 Transformer Energized Energized  20Y033 Distribution Panel Energized Energized 14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 37 of 63 Equipment ID Description Equipment Normal State Equipment Desired State Notes 20Y035 120 VAC 'Y' Power Panel Energized Energized  20Y050 Uninterruptable Power Supply Distribution Panel Energized Energized 2-13A-K004 RCIC Hi Temp Steam Leak Relay Standby Standby  2-13A-K006 RCIC Hi Temp Steam Leak Relay Standby Standby  2-13A-K007 RCIC Steam Line Hi DP Line Break Relay Standby Standby 2-13A-K010 RCIC Turbine Trip Aux Relay Standby Standby  2-13A-K011 RCIC Turbine Trip Aux Relay Standby Standby Closes Trip & Throttle Valve when energized 2-13A-K012 RCIC Auto Isolation Signal Relay Standby Standby  2-13A-K014 RCIC Pump Lo Suction Pressure Trip Relay Standby Standby 2-13A-K017 RCIC Turbine Exh Hi Pressure Trip Relay Standby Standby 2-13A-K-022 RCIC Auto Isolation Signal Relay Standby Standby  2-13A-K030 Reactor Hi Vessel Water Level Trip Relay Standby Standby 2-13A-K031 RCIC Steam Line Space Hi Temp Isolation Relay Standby Standby 2-13A-K032 RCIC Steam Line Space Hi Temp Isolation Relay Standby Standby 2-13A-K033 RCIC Steam Line Hi DP Line Break Relay Standby Standby 2-13A-K034 RCIC Auto Isolation Signal Relay Standby Standby  2-13A-K044 RCIC Reactor Hi Vessel Water Level Trip Relay Standby Standby 2-13A-K049 RCIC Low Steam Pressure Auto Isolation Relay Standby Standby 2-13A-K050 RCIC Low Steam Pressure Auto Isolation Relay Standby Standby 2-13A-K053 RCIC Reactor Hi Vessel Water Level
 
Trip Relay Auxiliary Standby Standby 2-13A-K054 RCIC Auto Isolation Signal Relay Standby Standby  2AC043 Emergency Shutdown Panel Standby Standby  2AC065 Rx Vessel Lvl and Pressure Inst Rack A Energized Energized  2AC091 Jet Pump Inst Rack A Energized Energized  2AC270 Panel Energized Standby  2AD001 2A 125V DC Battery Energized Energized  2AD003 Station Battery Charger 2A Energized Energized  2AD017 (2FPA) Battery Main Fuse Box Energized Energized  2AD018 (2DPA) 250V DC Distribution Panel Energized Energized  2AD019 (2FA) 250 Volt Fuse Box Energized Energized  2AD025 Distribution Panel Energized Energized  2AE024 Residual Heat Exchangers Standby Standby Passive component 2AS377 Back-Up N2 Supply to Ads RV's Standby Open Passive component 2AT545 2A Srv Inst N2 Accumulator Standby Standby Passive component 2BC043 Panel Standby Standby Contains control switch for LT-2-02-3-085A 2BC065 Rx Vessel Lvl and Pressure Inst Rack B Energized Energized  2BC172 Panel Standby Standby  2BC270 Panel Energized Standby  2BD001 2B 125V DC Battery Standby Energized Provides power for 120V AC Vital Instrument Power 2BD017 Battery Main Fuse Box Energized Energized  2BD018 250V DC Distribution Panel Div. II Energized Energized  2BE024 Residual Heat Exchangers Standby Standby Passive component 2BS377 Back-Up N2 Supply to Ads RV's Standby Open Passive component 2BS545 Automatic Transfer Switch Panel Energized Energized  2BT545 2B Srv Inst N2 Accumulator Standby Standby Passive component 2CC133 Panel Energized Energized  2CD001 2C 125V DC Battery Energized Energized  2CD003 Station Battery Charger 2C Energized Energized  2CD017 (2FPC) Battery Main Fuse Box Energized Energized 14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 38 of 63 Equipment ID Description Equipment Normal State Equipment Desired State Notes 2CD019 (2FC) 250 Volt Fuse Box Energized Energized  2CE024 Residual Heat Exchangers Standby Standby Passive component 2CS377 Back-Up N2 Supply to Ads RV's Standby Open Passive component 2CS545 Automatic Transfer Switch Panel Energized Energized  2CT545 2C Srv Inst N2 Accumulator Standby Standby Passive component 2DA-W-A (1201) RCIC MO-2-13-021 Breaker Energized Energized  2DA-W-A (1203) RCIC MO-2-13-030 Breaker Energized Energized  2DA-W-A (1204) RCIC MO-2-13-027 Breaker Energized Energized  2DA-W-A (1205) RCIC MO-2-13-041 Breaker Energized Energized  2DA-W-A (1206) RCIC MO-2-13-039 Breaker Energized Energized  2DA-W-A (1207) RCIC MO-2-13-132 Breaker Energized Energized  2DA-W-A (1209) RCIC MO-2-13-131 Breaker Energized Energized  2DA-W-A (1210) RCIC MO-2-13-018 Breaker Energized Energized  2DA-W-A (1214) RCIC Cond Vac PP 20P046 Breaker Energized Energized  2DA-W-A (1215) RCIC Vac Tank Cond PP 20P048 Breaker Energized Energized 2DA-W-A (20D012) RCIC 250VDC MCC Energized Energized  2DD001 2D 125V DC Battery Standby Energized Provides power for 120V AC Vital Instrument Power and RCIC B Logic 2DD017 Battery Main Fuse Box Energized Energized  2DD019 Fuse Box Energized Energized  2DE024 Residual Heat Exchangers Standby Standby Passive component 2GT545 2G Srv Inst N2 Accumulator Standby Standby Passive component 2KT545 2K Srv Inst N2 Accumulator Standby Standby Passive component 2OE032 RCIC Barometric Condenser (13-2) Standby Operating Passive Component 2OE104 RCIC Turb. Lube Oil Cooler (13-2) Standby Operating Passive Component 2AC270 (K3A) Relay De-Energized De-Energized  2BC270 (K3B) Relay De-Energized De-Energized  2AC270 (K3C) Relay De-Energized De-Energized  2BC270 (K3D) Relay De-Energized De-Energized  AO-2-07B-2511 Torus 18 Inch Vent Inboard Isol Valve
 
to Sbgt/Atmos Closed Open AO-2-07B-80290 Ctmt Emerg Vent Outboard Isolation Vlv to Atmos Closed Open E124 (1013) E124-R-C 20B036 Breake r  Energized Energized  E124 (1014) E124-T-B 20B059 Breaker Energized Energized E124 (20B010) E124 Load Center  Energized Energized E124-R-C (20B036) E124-R-C Motor Control Center Energized Energized E124-R-C (3606) MO-2-10-25A Norm Breaker Energized Energized E124-R-C (3691) Alt Feed Breaker 20D037 Y50 Energized Energized  E124-T-B (20B059) E124-T-B Motor Control Center Energized Energized  E124-T-B (5931) Norm Fdr for 125 V Battery Charger 'A' 2AD03 Energized Energized E324 (1213) E324-R-B 20B038 Breaker Energized Energized  E324 (1222) E324-T-B 00B049 Breaker Energized Energized E324 (20B012) E324 Load Center Energized Energized E324-R-B (20B038) MCC 20B038 Energized Energized E324-R-B (3821) MO-2-10-038A Breaker Energized Energized 14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 39 of 63 Equipment ID Description Equipment Normal State Equipment Desired State Notes E324-R-B (3822) MO-2-10-020 Breaker Locked De-Energized Energized E324-R-B (3824) MO-2-10-026A Breaker Energized Energized E324-R-B (3831) MO-2-10-039A Breaker Energized Energized E324-R-B (3832) MO-2-10-034A Breaker Energized Energized E324-R-B (3844) MO-2-10-031A Breaker Energized Energized E324-R-B (3862) MO-2-10-174 Breaker Energized Energized E324-R-B (3863) MO-2-10-176 Breaker De-Energized Energized E324-R-B (3882) Norm Fdr for 120V Instr Pnl 20Y35 Trans 20X135 Energized Energized E324-R-B (3893) 125 VDC Batt Charger 2CD03 Energized Energized  E324-T-B (00B049) MCC 00B049 for 0AV034, 0AV036, and 0AV030 Energized Energized INV-2-13-90 RCICs-125 VDC Bus 'A' Power Distribution Energized Energized  J-1648 Junction Box Standby Standby Contains resistor for PT-2-13-068 J2915 J-Box at E124 Load Center Standby Energized  J2916 J-Box at E324 Load Center Standby Energized LI-2-02 085A Reactor Vessel High Water Energized Energized  LI-2-02-3-113 Reactor Water Level Energized Energized LR/TR-8123A Torus Water Level/Temperature Recorder Energized Energized LT-2-02 072A Reactor Vessel Water Level Energized Energized  LT-2-02-3-113 Reactor Press Vessel Fuel Zone Wtr Level Energized Energized LT-8123A Torus Water Level Energized Energized  MO-2-10-020 RHR Loops A/B X-Tie Closed Open  MO-2 026A RHR Loop A D/W Spray O/B Closed Open/Closed  MO-2 031A RHR Loop A D/W Spray I/B Closed Open/Closed  MO-2 034A RHR Loop A FFT Valve Closed Open  MO-2 038A RHR Loop A Torus Spray Closed Open/Closed  MO-2 039A RHR Loop A Torus Valve Closed Open  MO-2-10-174 HPSW to RHR Inner X-Tie Closed Open  MO-2-10-176 HPSW to RHR Outer X-Tie Closed Open  MO-2-10-25A RHR Loop A I/B Disc Valve Closed Open Valve closure via control switch requires core spray relay logic permissive MO-2-13-018 RCIC Pump Suction from Condensate Storage Tank Open Open/Closed MO-2-13-021 RCIC Discharge to Feedwater Line B Closed Open  MO-2-13-027 RCIC Minimum Flow Valve Closed Open/Closed  MO-2-13-030 RCIC Full Flow Test Valve Closed Open/Closed  MO-2-13-039 RCIC Pump Torus Suction Outer Closed Open/Closed  MO-2-13-041 RCIC Pump Torus Suction Inner Closed Open/Closed  MO-2-13-131 RCIC Turbine Steam Supply Valve Closed Open/Closed  MO-2-13-132 RCIC Cooling Water Supply to Lo Clr +
Barometric Cdsr Closed Open N210025A Cabinet Provides Power to Valve MO-2-10-25A Energized Energized OAS384 N2 Tank Standby Standby Passive Component OAS385 N2 Tank Standby Standby Passive Component OBS384 N2 Tank Standby Standby Passive Component OBS385 N2 Tank Standby Standby Passive Component PCV-0-40W-70088A N2 Regulator - OAS384 Standby Open 14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 40 of 63 Equipment ID Description Equipment Normal State Equipment Desired State Notes PCV-0-40W-70088B N2 Regulator - OBS384 Standby Open  PCV-0-40W-70089A N2 Regulator - OAS385 Closed Open  PCV-0-40W-70089B N2 Regulator - OBS385 Closed Open  PI-2-06-090A Reactor Wide Range Press Ind Energized Energized  PI-2-06-090B Reactor Wide Range Press Ind Energized Energized  PI-2-06-090C Reactor Wide Range Press Ind Energized Energized  PI-2-13-094 RCIC Pump Turb Stm Press Energized Energized  PO-0-40D-00153-01 CR Emergency Vent Filters Inlet Damper Closed Open Fails open on loss of instrument air PO-0-40D-00153-02 CR Emergency Vent Filters Inlet Damper Closed Open Fails open on loss of instrument air PO-0-40D-00157-01 CR Emergency Vent Supply Fan Inlet Damper Closed Open Fails open on loss of instrument air PO-0-40D-00157-02 CR Emergency Vent Supply Fan Outlet Damper Closed Open Fails open on loss of instrument air PO-0-40W-00016 OA Damper Emergency Switchgear And Battery Room Vent Supply Fans Open Throttled Fails closed to minimum on loss of instrument air PO-0-40W-00019-01 OA Damper Emergency Switchgear And Battery Room Vent Supply Damper Standby Open PO-0-40W-00019-02 OA Damper Emergency Switchgear And Battery Room Vent Outlet Damper Standby Open PO-0-40W-00021-01 OA Damper Emergency Switchgear And Battery Room Vent Supply Damper Standby Open PO-0-40W-00031-01 Battery Room Exhaust Fan Inlet Damper Standby Open PO-0-40W-00031-02 Battery Room Exhaust Fan Outlet Damper Standby Open PO-0-40W-00782-01 OA Damper Emergency Switchgear and Battery Room Vent Supply Damper Standby Open PO-0-40W-00782-02 OA Damper Emergency Switchgear and Battery Room Vent Supply Damper Standby Open PO-0-40W-00808 OA Damper Emergency Switchgear and Battery Room Vent Outlet Damper Standby Open PO-0-40W-00822-01 OA Damper Emergency Switchgear and Battery Room Vent Supply Damper Standby Open PO-0-40W-00822-02 OA Damper Emergency Switchgear and Battery Room Vent Supply Damper Standby Open PO-0-40W-00822-03 OA Damper Emergency Switchgear and Battery Room Vent Supply Damper Standby Open PR/LR-2 096 Reactor Level/Steam Flow Ratio Energized Energized  PR/TR-4805 Containment Pressure/Temp Energized Energized  PS-2-13-67-1 Pressure Switch Open Open  PS-2-13-72A Pressure Switch Open Open  PS-2-13-72B Pressure Switch Open Open  PS-2-13-87A Pressure Switch Open Open  PS-2-13-87B Pressure Switch Open Open  PT-2-02 404A Reactor Pressure Energized Energized PT-2-02 404C Reactor Pressure Transmitter Energized Energized  PT-2-06-053A Reactor Wide Range -Pressure Energized Energized  PT-2-06-053B Reactor Wide Range -Pressure Energized Energized  PT-2-06-053C Reactor Wide Range -Pressure Energized Energized  PT-2-13-068 RCIC Turbine Steam Supply Press Energized Energized  PT-4805 Drywell Pressure Energized Energized  RV-2-02-071A 2A Safety Relief Valve Closed Open/Closed  RV-2-02-071B 2B Safety Relief Valve Closed Open/Closed  RV-2-02-071C 2C Safety Relief Valve  Closed Open/Closed  RV-2-02-071G 2G Safety Relief Valve Closed Open/Closed  RV-2-02-071K 2K Safety Relief Valve Closed Open/Closed SV-0-36B-00019 OA Damper Emergency Switchgear and
 
Battery Room Damper IA Standby Energized 
 
14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 41 of 63 Equipment ID Description Equipment Normal State Equipment Desired State Notes SV-0-36B-00031 Air Supply Shutoff for PO-0-40W-00031-01, PO-0-40W-00031-02 Standby Energized TI-2501 Ventilation Air Temperatures Energized Energized  TI-80146 Drywell Bulk Average Temp Indicator Energized Energized  TT-2501 Vent Air/Wtr Temp Energized Energized  XAM-2-02 117A Reactor Water Level Wide Range Energized Energized 14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 42 of 63
 
Attachment B PBAPS Unit 3 ESEL
 
14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 43 of 63 Equipment ID Description Equipment Normal State Equipment Desired State Notes 30C003 Reactor and Containment Cooling and Isolation Energized Energized  30C004C RCIC Control Panel Energized Energized  30C005A Reactor Manual Control Panel Energized Energized  30C012 Plant Services Console Energized Energized This panel is not credited for the Unit 3
 
ESEP-pertinent FLEX Response, however it remains on the list for conservatism 30C018 Panel Energized Energized  30C019 Panel Energized Energized  30C032 Panel  Energized Energized  30C033 Panel Energized Energized  30C034 RCIC Relay Panel Energized Energized  30C035 Panel Energized Energized  30C041 Panel Energized Energized  30C095 RCIC Instrument Rack Energized Energized  30C144 Panel Energized Energized  30C722A Accident Monitoring Instrumentation Panel Energized Energized  30C722B Panel Energized Energized  30C818 Reactor Water Level/Pressure Component Cabinet Energized Energized 30D021 (3PPA) 125V DC Station Distribution Energized Energized 30D023 (3PPC) 125V DC Station Distribution Energized Energized  30D024 Distribution Panel Energized Energized  30D037 Uninterruptable Power Supply Static Inverter Energized Bypassed and
 
Isolated  Powers vital instrument bus during Phase 1 30D039 RCIC Barometric Cdsr Vacuum Pump Starter Standby Energized 30D040 RCIC Barometric Cdsr Cond Pump Starter Standby Energized  30P036 RCIC Pump Standby Operating  30P046 RCIC Barometric Cdsr Vacuum Pump Standby Operating  30P048 RCIC Barometric Cdsr Condensate Pump Standby Operating  30P340 RCIC Turbine Driven Lube Oil Pump Standby Operating  30S038 RCIC Turbine Standby Operating  30S315 Static Inverter Man Bypass/Isolation Switch Energized Bypassed and Isolated  Phase 1 power is from inverter, Phase 2 is from 480/120V AC transformer 30S356 Load Center E134/E334 Transfer Switch Standby Energized  30S546 Transfer Switch Energized Energized  30S701 Transfer Switch Energized Energized  30S703 Transfer Switch Energized Energized  30S704 125V DC Battery Charger 3CD003 Transfer Sw 30S704 Energized Energized  30X133 Transformer Energized Energized  30X135 30Y035 Transformer Energized Energized  30X150 30Y050 Transformer Energized Energized  30Y033 Distribution Panel Energized Energized  30Y035 120 VAC 'Y' Power Panel Energized Energized  30Y050 120 VAC 'Y' Power Panel Energized Energized  3-13A-K004 RCIC Hi Temp Steam Leak Relay Standby Standby  3-13A-K006 RCIC Hi Temp Steam Leak Relay Standby Standby  3-13A-K007 RCIC Steam Line Hi DP Line Break Relay Standby Standby  3-13A-K010 RCIC Turbine Trip Aux Relay Standby Standby  3-13A-K011 RCIC Turbine Trip Aux Relay Standby Standby Closes Trip & Throttle Valve when energized 3-13A-K012 RCIC Auto Isolation Signal Relay Standby Standby  3-13A-K014 RCIC Pump Lo Suction Pressure Trip Relay Standby Standby  3-13A-K017 RCIC Turbine Exh Hi Pressure Trip Relay Standby Standby  3-13A-K-022 RCIC Auto Isolation Signal Relay Standby Standby  3-13A-K030 Reactor Hi Vessel Water Level Trip Relay Standby Standby  3-13A-K031 RCIC Steam Line Space Hi Temp Isolation Relay Standby Standby 3-13A-K032 RCIC Steam Line Space Hi Temp Isolation Relay Standby Standby 3-13A-K033 RCIC Steam Line Hi DP Line Break Relay Standby Standby 14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 44 of 63 Equipment ID Description Equipment Normal State Equipment Desired State Notes 3-13A-K034 RCIC Auto Isolation Signal Relay Standby Standby  3-13A-K044 RCIC Reactor Hi Vessel Water Level Trip Relay Standby Standby 3-13A-K049 RCIC Low Steam Pressure Auto Isolation Relay Standby Standby 3-13A-K050 RCIC Low Steam Pressure Auto Isolation Relay Standby Standby 3-13A-K053 RCIC Reactor Hi Vessel Water Level Trip
 
Relay Auxiliary Standby Standby 3-13A-K054 RCIC Auto Isolation Signal Relay Standby Standby  3AC043 Emergency Shutdown Panel Standby Standby  3AC065 Rx Vessel Lvl and Pressure Inst Rack A Energized Energized  3AC091 Jet Pump Inst Rack A Energized Energized  3AC270 Panel Energized Standby  3AD001 3A 125V DC Battery Energized Energized  3AD003 Station Battery Charger 3A Energized Energized  3AD017 (3FBA) Battery Main Fuse Box Energized Energized  3AD018 (3DPA) 250v DC Distribution Panel Energized Energized  3AD019 (3FA) 250 Volt Fuse Box Energized Energized  3AD025 (3PPAD) 3ppad 125V DC Distribution Panel Energized Energized  3AE024 Residual Heat Exchanger Standby Standby  3AS377 Back-Up N2 Supply to Ads RV's Standby Open  3AS456 Transfer Switch Energized Energized  3AT540 Instrument N2 Accumulator Standby Standby  3AT545 3A Srv Inst N2 Accumulator Standby Standby  3BC043 Panel Standby Standby Contains control switch for LT-3-02-3-085A 3BC065 Instrument Rack Standby Standby  3BC091 Instrument Rack Standby Standby  3BC270 Panel Energized Standby  3BD001 2B 125V DC Battery Standby Energized Provides power for 120V AC Vital Instrument Power 3BD017 Battery Main Fuse Box Energized Energized  3BD018 250V DC Distribution Panel Div. II Energized Energized  3BS377 Back-Up N2 Supply to Ads RV's Standby Open  3BS545 Automatic Transfer Switch Panel Energized Energized  3BT540 Instrument N2 Accumulator Standby Standby  3BT545 3B Srv Inst N2 Accumulator Standby Standby  3CD001 3C 125V DC Battery Energized Energized  3CD003 Station Battery Charger 3C Energized Energized  3CD017 (3FBC) Battery Main Fuse Box Energized Energized  3CD019 (3FC) 250 Volt Fuse Box Energized Energized  3CD025 (3PPCD) 3PPCD 125V DC Distribution Panel Energized Energized This panel is not credited as a power source for any item on this ESEL, it remains on the list for conservatism 3CE024 Residual Heat Exchanger Standby Standby  3CS377 Back-Up N2 Supply to Ads RV's Standby Open  3CS456 Control to Battery Charger 3CD003 Transfer Switch Energized Energized  3CT540 Instrument N2 Accumulator Standby Standby  3CT545 3C Srv Inst N2 Accumulator Standby Standby  3DA-W-A (1201) RCIC MO-3-13-021 Breaker Energized Energized  3DA-W-A (1203) RCIC MO-3-13-030 Breaker Energized Energized  3DA-W-A (1204) RCIC MO-3-13-027 Breaker Energized Energized  3DA-W-A (1205) RCIC MO-3-13-041 Breaker Energized Energized  3DA-W-A (1206) RCIC MO-3-13-039 Breaker Energized Energized  3DA-W-A (1207) RCIC MO-3-13-132 Breaker Energized Energized 14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 45 of 63 Equipment ID Description Equipment Normal State Equipment Desired State Notes 3DA-W-A (1209) RCIC MO-3-13-131 Breaker Energized Energized  3DA-W-A (1210) RCIC MO-3-13-018 Breaker Energized Energized  3DA-W-A (1214) RCIC Cond Vac PP 30P046 Breaker Energized Energized  3DA-W-A (1215) RCIC Vac Tank Cond PP 30P048 Breaker Energized Energized  3DA-W-A (30D012) RCIC 250VDC MCC Energized Energized  3DD001 2D 125V DC Battery Standby Energized Provides power for 120V AC Vital Instrument Power and RCIC B Logic 3DD017 Battery Main Fuse Box Energized Energized  3DD019 Fuse Box Energized Energized  3GT540 Instrument N2 Accumulator Standby Standby  3GT545 3G Srv Inst N2 Accumulator Standby Standby  3KT540 Instrument N2 Accumulator Standby Standby  3KT545 3K Srv Inst N2 Accumulator Standby Standby  3OE032 RCIC Barometric Condenser (13-2) Standby Operating  3OE104 RCIC Turb. Lube Oil Cooler (13-2)
Standby Operating  3AC270 (K3A) Relay De-Energized De-Energized  3BC270 (K3B) Relay De-Energized De-Energized  3AC270 (K3C) Relay De-Energized De-Energized  3BC270 (K3D) Relay De-Energized De-Energized  AO-3-07B-3511 Torus 18 Inch Vent Inboard Isol Valve to
 
Sbgt/Atmos Closed Open AO-3-07B-90290 Ctmt Emerg Vent Outboard Isolation Vlv to
 
Atmos Closed Open E134 (1014) E134-T-B 30B059 Breaker Energized Energized E134 (30B010) E134 Load Center Energized Energized  E134-T-B (30B059) E134-T-B Motor Control Center Energized Energized  E134-T-B (5924) Alt Feed For Uninterrupt AC Power Supp Inverter 30D37 Energized Energized  E134-T-B (5931) 125V. D.C. Battery Charger 3AD03 Energized Energized  E334 (1213) E334-R-B 30B038 Breaker Energized Energized E334 (30B012) E334 Load Center Energized Energized E334-R-B (30B038) MCC 30B038 Energized Energized E334-R-B (3821) MO-3-10-038A Breaker Energized Energized E334-R-B (3824) MO-3-10-026A Breaker Energized Energized E334-R-B (3831) MO-3-10-039A Breaker Energized Energized E334-R-B (3832) MO-3-10-034A Breaker Energized Energized E334-R-B (3844) MO-3-10-031A Breaker Energized Energized E334-R-B (3851) MO-3-10-25A Alt Breaker Energized Energized E334-R-B (3862) MO-3-10-174 Breaker Energized Energized E334-R-B (3863) MO-3-10-176 Breaker De-Energized Energized E334-R-B (3882) Norm Fdr for 120V Instr Pnl 30Y35 Trans 30X135 Energized Energized E334-R-B (3893) 125V Battery Charger C 3C Transfer Switch 30S704 Energized Energized  INV-3-13-90 RCICs-125 VDC Bus 'A' Power Distribution Energized Energized  J2919 J-Box at E134 LC Standby Energized  J2920 J-Box at E334 LC Standby Energized LI-3-02 085A Reactor Vessel High Water Energized Energized  LI-3-02-3-113 Reactor Water Level Energized Energized
 
14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 46 of 63 Equipment ID Description Equipment Normal State Equipment Desired State Notes LR/TR-9123A Torus Water Level/Temperature Recorder Energized Energized  LT-3-02-3-072A Reactor Vessel Water Level Energized Energized  LT-3-02-3-113 Level Transmitter Energized Energized  LT-9123A Torus Water Level Energized Energized  MO-3 025A RHR Inner Injection Valve to Recirc Loop A Closed Open Valve closure via control switch requires core spray relay logic permissive MO-3 026A RHR Loop A D/W Spray O/B Closed Open/Closed  MO-3 031A RHR Loop A D/W Spray I/B Closed Open/Closed  MO-3 034A RHR Loop A FFT Valve Closed Open/Closed  MO-3-10-038A RHR Loop A Torus Spray Closed Open/Closed  MO-3-10-039A RHR Loop A Torus Valve Closed Open  MO-3-10-174 HPSW to RHR Inner X-Tie Closed Open/Closed  MO-3-10-176 HPSW to RHR Outer X-Tie Closed Open/Closed  MO-3-13-018 RCIC Pump Suction from Condensate Storage Tank Open Open/Closed MO-3-13-021 RCIC Discharge to Feedwater Line B Closed Open  MO-3-13-027 RCIC Minimum Flow Valve Closed Open/Closed  MO-3-13-030 RCIC Full Flow Test Valve Closed Open/Closed  MO-3-13-039 RCIC Pump Torus Suction Outer Closed Open/Closed  MO-3-13-041 RCIC Pump Torus Suction Inner Closed Open/Closed  MO-3-13-131 RCIC Turbine Steam Supply Valve Closed Open  MO-3-13-132 RCIC Cooling Water Supply to Lo Clr +
Barometric Cdsr Closed Open N310025A Cabinet Energized Energized  PI-3-06-090A Reactor Wide Range Press Ind Energized Energized  PI-3-06-090B Reactor Wide Range Press Ind Energized Energized  PI-3-06-090C Reactor Wide Range Press Ind Energized Energized  PI-3-13-094 RCIC Pump Turb Stm Press Energized Energized  PR/LR-3 096 Reactor Level/Steam Flow Ratio Energized Energized  PR/TR-5805 Containment Pressure/Temp Energized Energized  PS-3-13-67-1 Pressure Switch Open Open  PS-3-13-72A Pressure Switch Open Open  PS-3-13-72B Pressure Switch Open Open  PS-3-13-87A Pressure Switch Open Open  PS-3-13-87B Pressure Switch Open Open  PT-3-02 404A Reactor Pressure Energized Energized PT-3-02 404C Reactor Pressure Transmitter Energized Energized  PT-3-06-053A Reactor Wide Range -Pressure Energized Energized PT-3-06-053B Reactor Wide Range -Pressure Energized Energized PT-3-06-053C Reactor Wide Range -Pressure Energized Energized PT-3-13-068 RCIC Turbine Steam Supply Pressure Energized Energized  PT-5805 Drywell Pressure Energized Energized  RV-3-02-071A 3A Safety Relief Valve Closed Open/Closed  RV-3-02-071B 3B Safety Relief Valve Closed Open/Closed  RV-3-02-071C 3C Safety Relief Valve Closed Open/Closed  RV-3-02-071G 3G Safety Relief Valve Closed Open/Closed  RV-3-02-071K 3K Safety Relief Valve Closed Open/Closed  XAM-3-02 117A Reactor Water Level Wide Range Energized Energized 14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 47 of 63
 
Attachment C ESEP HCLPF Values and Failure Modes Tabulation, Unit 2 and Common
 
14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 48 of 63
 
Equipment ID Description Failure Mode HCLPF (g) Basis 00C133 Panel Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 00E068 CR Fresh Air Supply Preheat Coil Equipment Capacity >0.24 Component is Rule Of Box to 0AF041. Parent component is evaluated in 14Q4233-CAL-002. 00E072 CR Vent Reheat Coil Screened >RLGM Component screened by SRT judgment. 00F039 CR Fresh Air Supply Roll Filter Equipment Capacity >0.24 Component is Rule Of Box to 0AF041. Parent component is evaluated in 14Q4233-CAL-002. 00F043 OA Damper Emergency Switchgear & Battery Room Vent Supply Roll Filter Screened >RLGM Component screened by SRT judgment. 00T116 CAD Liquid Nitrogen Storage Tank Anchorage 0.276 Component evaluated by HCLPF calculation 14Q4233-CAL-002.
0AE073 OA Damper Emergency Switchgear & Battery Room Vent Supply Heat Coil Equipment Capacity >0.24 Component is Rule Of Box to 0AV034. Parent component is evaluated in 14Q4233-CAL-002. 0AF041 A Train HEPA Filter Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 0AF042 A Train Charcoal Filter Equipment Capacity >0.24 Component is Rule Of Box to 0AF041. Parent component is evaluated in 14Q4233-CAL-002. 0AF050 A Train HEPA Filter Equipment Capacity >0.24 Component is Rule Of Box to 0AF041. Parent component is evaluated in 14Q4233-CAL-002.
0AV030 CR Room Emergency Vent Fan Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002.
0AV034 Emergency Switch Gear and Battery Room Supply Fan Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 0AV036 Battery Room Exhaust Fan Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 20C003 Reactor and Containment
 
Cooling and Isolation Panel Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 20C004C RCIC Control Panel Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 20C005A Reactor Manual Control
 
Panel Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 20C012 Plant Services Console Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 20C018 Panel Equipment Capacity 0.44 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 20C019 Panel Equipment Capacity 0.44 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 20C032 Panel Equipment Capacity 0.44 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 20C033 Panel Equipment Capacity 0.44 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 20C034 RCIC Relay Panel Equipment Capacity 0.44 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 20C035 Panel Equipment Capacity 0.44 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 20C041 Panel Equipment Capacity 0.44 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 20C095 RCIC Instrument Rack Anchorage 0.41 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 20C144 Panel Equipment Capacity 0.44 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 20C722A Accident Monitoring
 
Instrumentation Panel Equipment Capacity 0.44 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 20C722B Panel Equipment Capacity 0.44 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 20C818 Reactor Water Level/Pressure Component Cabinet Equipment Capacity 0.44 Component evaluated by HCLPF calculation 14Q4233-CAL-002.
14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 49 of 63
 
Equipment ID Description Failure Mode HCLPF (g) Basis 20D021 (2PPA) 125V DC Station Distribution Anchorage >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 20D023 (2PPC) 125V DC Station
 
Distribution Anchorage >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 20D024 Distribution Panel Anchorage 0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 20D037 Uninterruptable Power Supply Static Inverter Equipment Capacity 0.44 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 20D039 RCIC Barometric CDSR
 
Vacuum Pump Starter Screened >RLGM Component screened by SRT analysis. 20D040 RCIC Barometric CDSR Cond Pump Starter Screened >RLGM Component screened by SRT analysis. 20P036 RCIC Pump Screened >RLGM Component screened by SRT analysis. 20P046 RCIC Barometric CDSR
 
Vacuum Pump Anchorage 0.36 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 20P048 RCIC Barometric CDSR Condensate Pump Anchorage 0.36 Component is Rule Of Box to 20P046. Parent component is evaluated in 14Q4233-CAL-002. 20P340 RCIC Turbine Driven Lube Oil Pump Screened >RLGM Component is Rule Of Box to 20P036. Parent component screens. 20S038 RCIC Turbine Screened >RLGM Component is Rule Of Box to 20P036. Parent component screens. 20S315 Static Inverter Man Bypass/Isolation Switch Screened >RLGM Component screened by SRT analysis. 20S354 Load Center E-124/E324 Transfer Switch Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 20S700 Battery Charger Panel 2AD003 Transfer Switch 20S700 Anchorage >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 20S701 Transfer Switch Anchorage >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 20S703 Transfer Switch Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 20X133 Transformer Block wall interaction 0.27 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 20X135 20Y035 Transformer Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 20X150 Transformer Block wall interaction 0.27 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 20Y033 Distribution Panel Block wall interaction 0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 20Y035 120 VAC 'Y' Power Panel Screened >RLGM Component screened by SRT analysis. 20Y050 Uninterruptable Power Supply Distribution Panel Block wall interaction 0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 2-13A-K004 RCIC Hi Temp Steam Leak Relay Functional Failure 0.34 Component evaluated by HCLPF calculation 14Q4233-CAL-004. 2-13A-K006 RCIC Hi Temp Steam Leak Relay Functional Failure 0.34 Component evaluated by HCLPF calculation 14Q4233-CAL-004. 2-13A-K007 RCIC Steam Line Hi DP Line Break Relay Functional Failure 0.29 Component evaluated by HCLPF calculation 14Q4233-CAL-004. 2-13A-K010 RCIC Turbine Trip Aux Relay Functional Failure 0.14 Component evaluated by HCLPF calculation 14Q4233-CAL-004. (Resolved by Operator Action per TODI PB 1570792-76) 2-13A-K011 RCIC Turbine Trip Aux Relay Functional Failure 0.14 Component evaluated by HCLPF calculation 14Q4233-CAL-004. (Resolved by Operator Action per TODI PB 1570792-76) 2-13A-K012 RCIC Auto Isolation Signal Relay Functional Failure 0.36 Component evaluated by HCLPF calculation 14Q4233-CAL-004. 2-13A-K014 RCIC Pump Lo Suction Pressure Trip Relay Functional Failure 0.34 Component evaluated by HCLPF calculation 14Q4233-CAL-004. 2-13A-K017 RCIC Turbine Exh Hi Pressure Trip Relay Functional Failure 0.34 Component evaluated by HCLPF calculation 14Q4233-CAL-004.
14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 50 of 63
 
Equipment ID Description Failure Mode HCLPF (g) Basis 2-13A-K-022 RCIC Auto Isolation Signal Relay Functional Failure 0.36 Component evaluated by HCLPF calculation 14Q4233-CAL-004. 2-13A-K030 Reactor Hi Vessel Water Level Trip Relay Functional Failure 0.34 Component evaluated by HCLPF calculation 14Q4233-CAL-004. 2-13A-K031 RCIC Steam Line Space Hi Temp Isolation Relay Functional Failure 0.34 Component evaluated by HCLPF calculation 14Q4233-CAL-004. 2-13A-K032 RCIC Steam Line Space Hi Temp Isolation Relay Functional Failure 0.34 Component evaluated by HCLPF calculation 14Q4233-CAL-004. 2-13A-K033 RCIC Steam Line Hi DP Line Break Relay Functional Failure 0.20 Component evaluated by HCLPF calculation 14Q4233-CAL-004. (Modification/Resolution Required) 2-13A-K034 RCIC Auto Isolation Signal Relay Functional Failure 0.36 Component evaluated by HCLPF calculation 14Q4233-CAL-004. 2-13A-K044 RCIC Reactor Hi Vessel Water Level Trip Relay Functional Failure 0.29 Component evaluated by HCLPF calculation 14Q4233-CAL-004. 2-13A-K049 RCIC Low Steam Pressure
 
Auto Isolation Relay Functional Failure 0.36 Component evaluated by HCLPF calculation 14Q4233-CAL-004. 2-13A-K050 RCIC Low Steam Pressure
 
Auto Isolation Relay Functional Failure 0.36 Component evaluated by HCLPF calculation 14Q4233-CAL-004. 2-13A-K053 RCIC Reactor Hi Vessel Water Level Trip Relay Auxiliary Functional Failure 0.34 Component evaluated by HCLPF calculation 14Q4233-CAL-004. 2-13A-K054 RCIC Auto Isolation Signal Relay Functional Failure 0.36 Component evaluated by HCLPF calculation 14Q4233-CAL-004. 2AC043 Emergency Shutdown Panel Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 2AC065 Rx Vessel Lvl and Pressure
 
Inst Rack A Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 2AC091 Jet Pump Inst Rack A Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 2AC270 Panel Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 2AD001 2A 125V DC Battery Anchorage 0.26 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 2AD003 Station Battery Charger 2A Block wall interaction 0.27 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 2AD017 (2FPA) Battery Main Fuse Box Anchorage >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 2AD018 (2DPA) 250V DC Distribution Panel Anchorage >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 2AD019 (2FA) 250 Volt Fuse Box Anchorage >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 2AD025 Distribution Panel Anchorage >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 2AE024 Residual Heat Exchangers Anchorage 0.26 Component evaluated by HCLPF calculation 14Q4233-CAL-002.
2AS377 Back-Up N2 Supply to Ads
 
RV's Screened >RLGM Component screened by SRT analysis. 2AT545 2A Srv Inst N2 Accumulator Screened >RLGM Component screened by SRT judgment. 2BC043 Panel Equipment Capacity >0.24 Component is Rule Of Box to 2AC043. Parent component is evaluated in 14Q4233-CAL-002. 2BC065 Rx Vessel Lvl and Pressure Inst Rack B Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 2BC172 Panel Equipment Capacity 0.44 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 2BC270 Panel Equipment Capacity >0.24 Component is Rule Of Box to 2AC270. Parent component is evaluated in 14Q4233-CAL-002. 2BD001 2B 125V DC Battery Anchorage 0.26 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 2BD017 Battery Main Fuse Box Anchorage >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002.
14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 51 of 63
 
Equipment ID Description Failure Mode HCLPF (g) Basis 2BD018 250V DC Distribution Panel Div. II Anchorage >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 2BE024 Residual Heat Exchangers Anchorage 0.26 Component evaluated by HCLPF calculation 14Q4233-CAL-002.
2BS377 Back-Up N2 Supply to Ads
 
RV's Screened >RLGM Component is Rule Of Box to 2AS377. Parent component screens.
2BS545 Automatic Transfer Switch
 
Panel Anchorage >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 2BT545 2B Srv Inst N2 Accumulator Screened >RLGM Component screened by SRT judgment. 2CC133 Panel Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 2CD001 2C 125V DC Battery Anchorage 0.26 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 2CD003 Station Battery Charger 2C Block wall interaction 0.27 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 2CD017 (2FPC) Battery Main Fuse Box Block wall interaction 0.27 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 2CD019 (2FC) 250 Volt Fuse Box Anchorage >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 2CE024 Residual Heat Exchangers Anchorage 0.26 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 2CS377 Back-Up N2 Supply to Ads
 
RV's Screened >RLGM Component is Rule Of Box to 2AS377. Parent component screens. 2CS545 Automatic Transfer Switch
 
Panel Anchorage >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 2CT545 2C Srv Inst N2 Accumulator Screened >RLGM Component screened by SRT judgment. 2DA-W-A (1201) RCIC MO-2-13-021 Breaker Anchorage 0.43 Component is Rule Of Box to 2DA-W-A (20D012).
Parent component is evaluated in 14Q4233-CAL-002. 2DA-W-A (1203) RCIC MO-2-13-030 Breaker Anchorage 0.43 Component is Rule Of Box to 2DA-W-A (20D012).
Parent component is evaluated in 14Q4233-CAL-002. 2DA-W-A (1204) RCIC MO-2-13-027 Breaker Anchorage 0.43 Component is Rule Of Box to 2DA-W-A (20D012).
Parent component is evaluated in 14Q4233-CAL-002. 2DA-W-A (1205) RCIC MO-2-13-041 Breaker Anchorage 0.43 Component is Rule Of Box to 2DA-W-A (20D012).
Parent component is evaluated in 14Q4233-CAL-002. 2DA-W-A (1206) RCIC MO-2-13-039 Breaker Anchorage 0.43 Component is Rule Of Box to 2DA-W-A (20D012).
Parent component is evaluated in 14Q4233-CAL-002. 2DA-W-A (1207) RCIC MO-2-13-132 Breaker Anchorage 0.43 Component is Rule Of Box to 2DA-W-A (20D012).
Parent component is evaluated in 14Q4233-CAL-002. 2DA-W-A (1209) RCIC MO-2-13-131 Breaker Anchorage 0.43 Component is Rule Of Box to 2DA-W-A (20D012).
Parent component is evaluated in 14Q4233-CAL-002. 2DA-W-A (1210) RCIC MO-2-13-018 Breaker Anchorage 0.43 Component is Rule Of Box to 2DA-W-A (20D012). Parent component is evaluated in 14Q4233-CAL-002. 2DA-W-A (1214) RCIC Cond Vac PP 20P046 Breaker Anchorage 0.43 Component is Rule Of Box to 2DA-W-A (20D012).
Parent component is evaluated in 14Q4233-CAL-002. 2DA-W-A (1215) RCIC Vac Tank Cond PP 20P048 Breaker Anchorage 0.43 Component is Rule Of Box to 2DA-W-A (20D012).
Parent component is evaluated in 14Q4233-CAL-002.
2DA-W-A (20D012) RCIC 250VDC MCC Anchorage 0.43 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 2DD001 2D 125V DC Battery Anchorage 0.26 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 2DD017 Battery Main Fuse Box Anchorage >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 2DD019 Fuse Box Anchorage >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 2DE024 Residual Heat Exchangers Anchorage 0.26 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 2GT545 2G Srv Inst N2 Accumulator Screened >RLGM Component screened by SRT judgment. 2KT545 2K Srv Inst N2 Accumulator Screened >RLGM Component screened by SRT judgment.
14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 52 of 63
 
Equipment ID Description Failure Mode HCLPF (g) Basis 2OE032 RCIC Barometric Condenser (13-2)
Equipment Capacity 0.44 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 2OE104 RCIC Turb. Lube Oil Cooler (13-2) Screened >RLGM Component is Rule Of Box to 20P036. Parent component screens. 2AC270 (K3A) Relay Equipment Capacity >0.24 Component is Rule Of Box to 2AC270. Parent component is evaluated in 14Q4233-CAL-002. Component evaluated by HCLPF calculation 14Q4233-CAL-004. 2BC270 (K3B) Relay Equipment Capacity >0.24 Component is Rule Of Box to 2BC270. Parent component is evaluated in 14Q4233-CAL-002. Component evaluated by HCLPF calculation 14Q4233-CAL-004. 2AC270 (K3C) Relay Equipment Capacity >0.24 Component is Rule Of Box to 2AC270. Parent component is evaluated in 14Q4233-CAL-002. Component evaluated by HCLPF calculation 14Q4233-CAL-004. 2BC270 (K3D) Relay Equipment Capacity >0.24 Component is Rule Of Box to 2BC270. Parent component is evaluated in 14Q4233-CAL-002. Component evaluated by HCLPF calculation 14Q4233-CAL-004. AO-2-07B-2511 Torus 18 Inch Vent Inboard
 
Isol Valve to Sbgt/Atmos Screened >RLGM Component screened by SRT judgment. AO-2-07B-80290 Ctmt Emerg Vent Outboard Isolation Vlv to Atmos Screened >RLGM Component screened by SRT judgment. E124 (1013) E124-R-C 20B036 Breaker Equipment Capacity >0.24 Component is Rule Of Box to E124 (20B010). Parent component is evaluated in 14Q4233-CAL-002. E124 (1014) E124-T-B 20B059 Breaker Equipment Capacity >0.24 Component is Rule Of Box to E124 (20B010). Parent component is evaluated in 14Q4233-CAL-002. E124 (20B010) E124 Load Center Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. E124-R-C (20B036) E124-R-C Motor Control Center Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. E124-R-C (3606) MO-2-10-25A Norm Breaker Equipment Capacity >0.24 Component is Rule Of Box to E124-R-C (20B036).
Parent component is evaluated in 14Q4233-CAL-002. E124-R-C (3691) Alt Feed Breaker 20D037 Y50 Equipment Capacity >0.24 Component is Rule Of Box to E124-R-C (20B036).
Parent component is evaluated in 14Q4233-CAL-002. E124-T-B (20B059) E124-T-B Motor Control Center Block wall interaction 0.27 Component evaluated by HCLPF calculation 14Q4233-CAL-002. E124-T-B (5931) Norm Fdr for 125 V Battery Charger 'A' 2AD03 Block wall interaction 0.27 Component is Rule Of Box to E124-T-B (20B059).
Parent component is evaluated in 14Q4233-CAL-002. E324 (1213) E324-R-B 20B038 Breaker Equipment Capacity >0.24 Component is Rule Of Box to E324 (20B012). Parent component is evaluated in 14Q4233-CAL-002. E324 (1222) E324-T-B 00B049 Breaker Equipment Capacity >0.24 Component is Rule Of Box to E324 (20B012). Parent component is evaluated in 14Q4233-CAL-002. E324 (20B012) E324 Load Center Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002.
E324-R-B (20B038) MCC 20B038 Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. E324-R-B (3821) MO-2-10-038A Breaker Equipment Capacity >0.24 Component is Rule Of Box to E324-R-B (20B038).
Parent component is evaluated in 14Q4233-CAL-002. E324-R-B (3822) MO-2-10-020 Breaker Equipment Capacity >0.24 Component is Rule Of Box to E324-R-B (20B038).
Parent component is evaluated in 14Q4233-CAL-002. E324-R-B (3824) MO-2-10-026A Breaker Equipment Capacity >0.24 Component is Rule Of Box to E324-R-B (20B038).
Parent component is evaluated in 14Q4233-CAL-002. E324-R-B (3831) MO-2-10-039A Breaker Equipment Capacity >0.24 Component is Rule Of Box to E324-R-B (20B038).
Parent component is evaluated in 14Q4233-CAL-002. E324-R-B (3832) MO-2-10-034A Breaker Equipment Capacity >0.24 Component is Rule Of Box to E324-R-B (20B038).
Parent component is evaluated in 14Q4233-CAL-002. E324-R-B (3844) MO-2-10-031A Breaker Equipment Capacity >0.24 Component is Rule Of Box to E324-R-B (20B038).
Parent component is evaluated in 14Q4233-CAL-002. E324-R-B (3862) MO-2-10-174 Breaker Equipment Capacity >0.24 Component is Rule Of Box to E324-R-B (20B038).
Parent component is evaluated in 14Q4233-CAL-002.
14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 53 of 63
 
Equipment ID Description Failure Mode HCLPF (g) Basis E324-R-B (3863) MO-2-10-176 Breaker Equipment Capacity >0.24 Component is Rule Of Box to E324-R-B (20B038). Parent component is evaluated in 14Q4233-CAL-002. E324-R-B (3882) Norm Fdr for 120V Instr Pnl 20Y35 Trans 20X135 Equipment Capacity >0.24 Component is Rule Of Box to E324-R-B (20B038).
Parent component is evaluated in 14Q4233-CAL-002. E324-R-B (3893) 125 VDC Batt Charger 2CD03 Equipment Capacity >0.24 Component is Rule Of Box to E324-R-B (20B038).
Parent component is evaluated in 14Q4233-CAL-002. E324-T-B (00B049) MCC 00B049 for 0AV034, 0AV036, and 0AV030 Block wall interaction 0.27 Component evaluated by HCLPF calculation 14Q4233-CAL-002. INV-2-13-90 RCICs-125 VDC Bus 'A' Power Distribution Equipment Capacity 0.44 Component is Rule Of Box to 20C019. Parent component is evaluated in 14Q4233-CAL-002. J-1648 Junction Box Screened >RLGM Component screened by SRT analysis. J2915 J-Box at E124 Load Center Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. J2916 J-Box at E324 Load Center Screened >RLGM Component screened by SRT analysis. LI-2-02-3-085A Reactor Vessel High Water Equipment Capacity >0.24 Component is Rule Of Box to 20C005A. Parent component is evaluated in 14Q4233-CAL-002. LI-2-02-3-113 Reactor Water Level Equipment Capacity >0.24 Component is Rule Of Box to 20C003. Parent component is evaluated in 14Q4233-CAL-002. LR/TR-8123A Torus Water Level/Temperature Recorder Equipment Capacity >0.24 Component is Rule Of Box to 20C004C. Parent component is evaluated in 14Q4233-CAL-002. LT-2-02-3-072A Reactor Vessel Water Level Equipment Capacity >0.24 Component is Rule Of Box to 2AC065. Parent component is evaluated in 14Q4233-CAL-002. LT-2-02-3-113 Reactor Press Vessel Fuel Zone Wtr Level Equipment Capacity >0.24 Component is Rule Of Box to 2AC091. Parent component is evaluated in 14Q4233-CAL-002. LT-8123A Torus Water Level Screened >RLGM Component screened by SRT analysis. MO-2-10-020 RHR Loops A/B X-Tie Screened >RLGM Component screened by SRT analysis. MO-2-10-026A RHR Loop A D/W Spray O/B Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. MO-2-10-031A RHR Loop A D/W Spray I/B Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. MO-2-10-034A RHR Loop A FFT Valve Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. MO-2-10-038A RHR Loop A Torus Spray Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. MO-2-10-039A RHR Loop A Torus Valve Screened >RLGM Component screened by SRT analysis. MO-2-10-174 HPSW to RHR Inner X-Tie Screened >RLGM Component screened by SRT analysis. MO-2-10-176 HPSW to RHR Outer X-Tie Screened >RLGM Component screened by SRT analysis. MO-2-10-25A RHR Loop A I/B Disc Valve Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. MO-2-13-018 RCIC Pump Suction from Condensate Storage Tank Screened >RLGM Component screened by SRT analysis. MO-2-13-021 RCIC Discharge to Feedwater Line B Screened >RLGM Component screened by SRT analysis. MO-2-13-027 RCIC Minimum Flow Valve Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. MO-2-13-030 RCIC Full Flow Test Valve Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. MO-2-13-039 RCIC Pump Torus Suction Outer Screened >RLGM Component screened by SRT analysis. MO-2-13-041 RCIC Pump Torus Suction Inner Screened >RLGM Component screened by SRT analysis. MO-2-13-131 RCIC Turbine Steam Supply
 
Valve Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. MO-2-13-132 RCIC Cooling Water Supply to Lo Clr + Barometric Cdsr Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002.
14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 54 of 63
 
Equipment ID Description Failure Mode HCLPF (g) Basis N210025A Cabinet Provides Power to Valve MO-2-10-25A Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. OAS384 N2 Tank Screened >RLGM Component screened by SRT analysis. OAS385 N2 Tank Screened >RLGM Component is Rule Of Box to OBS385. Parent component screens. OBS384 N2 Tank Screened >RLGM Component is Rule Of Box to OAS384. Parent component screens. OBS385 N2 Tank Screened >RLGM Component screened by SRT analysis. PCV-0-40W-70088A N2 Regulator - OAS384 Screened >RLGM Component is Rule Of Box to OAS384. Parent component screens. PCV-0-40W-70088B N2 Regulator - OBS384 Screened >RLGM Component is Rule Of Box to OBS384. Parent component screens. PCV-0-40W-70089A N2 Regulator - OAS385 Screened >RLGM Component is Rule Of Box to OAS385. Parent component screens. PCV-0-40W-70089B N2 Regulator - OBS385 Screened >RLGM Component is Rule Of Box to OBS385. Parent component screens. PI-2-06-090A Reactor Wide Range Press
 
Ind Equipment Capacity >0.24 Component is Rule Of Box to 20C005A. Parent component is evaluated in 14Q4233-CAL-002. PI-2-06-090B Reactor Wide Range Press
 
Ind Equipment Capacity >0.24 Component is Rule Of Box to 20C005A. Parent component is evaluated in 14Q4233-CAL-002. PI-2-06-090C Reactor Wide Range Press
 
Ind Equipment Capacity >0.24 Component is Rule Of Box to 20C005A. Parent component is evaluated in 14Q4233-CAL-002. PI-2-13-094 RCIC Pump Turb Stm Press Equipment Capacity >0.24 Component is Rule Of Box to 20C004C. Parent component is evaluated in 14Q4233-CAL-002. PO-0-40D-00153-01 CR Emergency Vent Filters
 
Inlet Damper Equipment Capacity >0.24 Component is Rule Of Box to 0AF041. Parent component is evaluated in 14Q4233-CAL-002. PO-0-40D-00153-02 CR Emergency Vent Filters
 
Inlet Damper Equipment Capacity >0.24 Component is Rule Of Box to 0AF041. Parent component is evaluated in 14Q4233-CAL-002. PO-0-40D-00157-01 CR Emergency Vent Supply Fan Inlet Damper Equipment Capacity >0.24 Component is Rule Of Box to 0AF041. Parent component is evaluated in 14Q4233-CAL-002. PO-0-40D-00157-02 CR Emergency Vent Supply Fan Outlet Damper Screened >RLGM Component screened by SRT judgment. PO-0-40W-00016 OA Damper Emergency Switchgear And Battery Room Vent Supply Fans Screened >RLGM Component is Rule Of Box to 00F043. Parent component is screened. PO-0-40W-00019-01 OA Damper Emergency Switchgear And Battery Room Vent Supply Damper Equipment Capacity >0.24 Component is Rule Of Box to 0AV034. Parent component is evaluated in 14Q4233-CAL-002. PO-0-40W-00019-02 OA Damper Emergency Switchgear And Battery Room Vent Outlet Damper Screened >RLGM Component is Rule Of Box to 00F043. Parent component is screened. PO-0-40W-00021-01 OA Damper Emergency Switchgear And Battery Room Vent Supply Damper Equipment Capacity >0.24 Component is Rule Of Box to 0AV034. Parent component is evaluated in 14Q4233-CAL-002. PO-0-40W-00031-01 Battery Room Exhaust Fan
 
Inlet Damper Screened >RLGM Component screened by SRT judgment. PO-0-40W-00031-02 Battery Room Exhaust Fan Outlet Damper Screened >RLGM Component screened by SRT judgment. PO-0-40W-00782-01 OA Damper Emergency Switchgear and Battery Room Vent Supply Damper Equipment Capacity >0.24 Component is Rule Of Box to 0AV034. Parent component is evaluated in 14Q4233-CAL-002. PO-0-40W-00782-02 OA Damper Emergency Switchgear and Battery Room Vent Supply Damper Equipment Capacity >0.24 Component is Rule Of Box to 0AV034. Parent component is evaluated in 14Q4233-CAL-002. PO-0-40W-00808 OA Damper Emergency Switchgear and Battery Room Vent Outlet Damper Screened >RLGM Component screened by SRT judgment. PO-0-40W-00822-01 OA Damper Emergency Switchgear and Battery Room Vent Supply Damper Equipment Capacity >0.24 Component is Rule Of Box to 0AV034. Parent component is evaluated in 14Q4233-CAL-002.
14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 55 of 63
 
Equipment ID Description Failure Mode HCLPF (g) Basis PO-0-40W-00822-02 OA Damper Emergency Switchgear and Battery Room Vent Supply Damper Equipment Capacity >0.24 Component is Rule Of Box to 0AV034. Parent component is evaluated in 14Q4233-CAL-002. PO-0-40W-00822-03 OA Damper Emergency Switchgear and Battery Room Vent Supply Damper Equipment Capacity >0.24 Component is Rule Of Box to 0AV034. Parent component is evaluated in 14Q4233-CAL-002. PR/LR-2-06-096 Reactor Level/Steam Flow Ratio Equipment Capacity >0.24 Component is Rule Of Box to 20C005A. Parent component is evaluated in 14Q4233-CAL-002. PR/TR-4805 Containment Pressure/Temp Equipment Capacity >0.24 Component is Rule Of Box to 20C003. Parent component is evaluated in 14Q4233-CAL-002. PS-2-13-67-1 Pressure Switch Anchorage 0.41 Component is Rule Of Box to 20C095. Parent component is evaluated in 14Q4233-CAL-002.
Component evaluated by HCLPF calculation 14Q4233-CAL-004. PS-2-13-72A Pressure Switch Functional Failure 0.27 Component evaluated by HCLPF calculation 14Q4233-CAL-004. PS-2-13-72B Pressure Switch Functional Failure 0.27 Component evaluated by HCLPF calculation 14Q4233-CAL-004. PS-2-13-87A Pressure Switch Functional Failure 0.27 Component evaluated by HCLPF calculation 14Q4233-CAL-004. PS-2-13-87B Pressure Switch Functional Failure 0.27 Component evaluated by HCLPF calculation 14Q4233-CAL-004. PT-2-02-3-404A Reactor Pressure Equipment Capacity >0.24 Component is Rule Of Box to 2AC065. Parent component is evaluated in 14Q4233-CAL-002. PT-2-02-3-404C Reactor Pressure Transmitter Equipment Capacity >0.24 Component is Rule Of Box to 2AC091. Parent component is evaluated in 14Q4233-CAL-002. PT-2-06-053A Reactor Wide Range -
Pressure Equipment Capacity >0.24 Component is Rule Of Box to 2AC065. Parent component is evaluated in 14Q4233-CAL-002. PT-2-06-053B Reactor Wide Range -
Pressure Equipment Capacity >0.24 Component is Rule Of Box to 2BC065. Parent component is evaluated in 14Q4233-CAL-002. PT-2-06-053C Reactor Wide Range -
Pressure Equipment Capacity >0.24 Component is Rule Of Box to 2AC065. Parent component is evaluated in 14Q4233-CAL-002. PT-2-13-068 RCIC Turbine Steam Supply Press Anchorage 0.41 Component is Rule Of Box to 20C095. Parent component is evaluated in 14Q4233-CAL-002. PT-4805 Drywell Pressure Screened >RLGM Component screened by SRT analysis. RV-2-02-071A 2A Safety Relief Valve Screened >RLGM Component screened by SRT judgment. RV-2-02-071B 2B Safety Relief Valve Screened >RLGM Component screened by SRT judgment. RV-2-02-071C 2C Safety Relief Valve  Screened >RLGM Component screened by SRT judgment. RV-2-02-071G 2G Safety Relief Valve Screened >RLGM Component screened by SRT judgment. RV-2-02-071K 2K Safety Relief Valve Screened >RLGM Component screened by SRT judgment. SV-0-36B-00019 OA Damper Emergency Switchgear and Battery Room Damper IA Screened >RLGM Component is Rule Of Box to 00F043. Parent component is Screened. SV-0-36B-00031 Air Supply Shutoff for PO-0-40W-00031-01, PO-0-40W-00031-02 Screened >RLGM Component screened by SRT analysis. TI-2501 Ventilation Air Temperatures Equipment Capacity >0.24 Component is Rule Of Box to 20C012. Parent component is evaluated in 14Q4233-CAL-002. TI-80146 Drywell Bulk Average Temp
 
Indicator Equipment Capacity >0.24 Component is Rule Of Box to 20C012. Parent component is evaluated in 14Q4233-CAL-002. TT-2501 Vent Air/Wtr Temp Equipment Capacity 0.44 Component is Rule Of Box to 2BC172. Parent component is evaluated in 14Q4233-CAL-002. XAM-2-02-3-117A Reactor Water Level Wide Range Equipment Capacity 0.44 Component is Rule Of Box to 20C818. Parent component is evaluated in 14Q4233-CAL-002.
 
14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 56 of 63
 
Attachment D ESEP HCLPF Values and Failure Modes Tabulation, Unit 3
 
14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 57 of 63 Equipment ID Description Failure Mode HCLPF (g) Basis 30C003 Reactor and Containment Cooling and Isolation Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 30C004C RCIC Control Panel Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 30C005A Reactor Manual Control
 
Panel Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 30C012 Plant Services Console Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 30C018 Panel Equipment Capacity 0.44 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 30C019 Panel Equipment Capacity 0.44 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 30C032 Panel Equipment Capacity 0.44 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 30C033 Panel Equipment Capacity 0.44 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 30C034 RCIC Relay Panel Equipment Capacity 0.44 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 30C035 Panel Equipment Capacity 0.44 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 30C041 Panel Equipment Capacity 0.44 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 30C095 RCIC Instrument Rack Anchorage 0.41 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 30C144 Panel Equipment Capacity 0.44 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 30C722A Accident Monitoring
 
Instrumentation Panel Equipment Capacity 0.44 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 30C722B Panel Equipment Capacity 0.44 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 30C818 Reactor Water Level/Pressure Component Cabinet Equipment Capacity 0.44 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 30D021 (3PPA) 125V DC Station Distribution Anchorage 0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 30D023 (3PPC) 125V DC Station Distribution Anchorage >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 30D024 Distribution Panel Anchorage >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 30D037 Uninterruptable Power Supply Static Inverter Equipment Capacity 0.44 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 30D039 RCIC Barometric Cdsr Vacuum Pump Starter Screened >RLGM Component screened by SRT analysis. 30D040 RCIC Barometric Cdsr Cond
 
Pump Starter Screened >RLGM Component screened by SRT analysis. 30P036 RCIC Pump Anchorage >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 30P046 RCIC Barometric Cdsr
 
Vacuum Pump Anchorage 0.36 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 30P048 RCIC Barometric Cdsr Condensate Pump Anchorage 0.36 Component is Rule Of Box to 30P046. Parent component is evaluated in 14Q4233-CAL-002. 30P340 RCIC Turbine Driven Lube Oil Pump Anchorage >0.24 Component is Rule Of Box to 30P036. Parent component is evaluated in 14Q4233-CAL-002. 30S038 RCIC Turbine Anchorage >0.24 Component is Rule Of Box to 30P036. Parent component is evaluated in 14Q4233-CAL-002. 30S315 Static Inverter Man Bypass/Isolation Switch Screened >RLGM Component screened by SRT analysis. 30S356 Load Center E134/E334 Transfer Switch Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002.
14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 58 of 63 Equipment ID Description Failure Mode HCLPF (g) Basis 30S546 Transfer Switch Block wall interaction 0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 30S701 Transfer Switch Anchorage >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 30S703 Transfer Switch Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 30S704 125V DC Battery Charger 3CD003 Transfer Sw 30S704 Anchorage >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 30X133 Transformer Block wall interaction 0.27 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 30X135 30Y035 Transformer Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 30X150 30Y050 Transformer Block wall interaction 0.27 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 30Y033 Distribution Panel Block wall interaction 0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 30Y035 120 VAC 'Y' Power Panel Screened >RLGM Component screened by SRT analysis. 30Y050 120 VAC 'Y' Power Panel Block wall interaction 0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 3-13A-K004 RCIC Hi Temp Steam Leak Relay Functional Failure 0.34 Component evaluated by HCLPF calculation 14Q4233-CAL-004. 3-13A-K006 RCIC Hi Temp Steam Leak Relay Functional Failure 0.34 Component evaluated by HCLPF calculation 14Q4233-CAL-004. 3-13A-K007 RCIC Steam Line Hi DP Line Break Relay Functional Failure 0.29 Component evaluated by HCLPF calculation 14Q4233-CAL-004. 3-13A-K010 RCIC Turbine Trip Aux Relay Functional Failure 0.14 Component evaluated by HCLPF calculation 14Q4233-CAL-004. (Resolved by Operator Action per TODI PB 1570792-76) 3-13A-K011 RCIC Turbine Trip Aux Relay Functional Failure 0.14 Component evaluated by HCLPF calculation 14Q4233-CAL-004.
(Resolved by Operator Action per TODI PB 1570792-76) 3-13A-K012 RCIC Auto Isolation Signal Relay Functional Failure 0.29 Component evaluated by HCLPF calculation 14Q4233-CAL-004. 3-13A-K014 RCIC Pump Lo Suction Pressure Trip Relay Functional Failure 0.34 Component evaluated by HCLPF calculation 14Q4233-CAL-004. 3-13A-K017 RCIC Turbine Exh Hi Pressure Trip Relay Functional Failure 0.34 Component evaluated by HCLPF calculation 14Q4233-CAL-004. 3-13A-K-022 RCIC Auto Isolation Signal Relay Functional Failure 0.36 Component evaluated by HCLPF calculation 14Q4233-CAL-004. 3-13A-K030 Reactor Hi Vessel Water Level Trip Relay Functional Failure 0.34 Component evaluated by HCLPF calculation 14Q4233-CAL-004. 3-13A-K031 RCIC Steam Line Space Hi Temp Isolation Relay Functional Failure 0.34 Component evaluated by HCLPF calculation 14Q4233-CAL-004. 3-13A-K032 RCIC Steam Line Space Hi Temp Isolation Relay Functional Failure 0.34 Component evaluated by HCLPF calculation 14Q4233-CAL-004. 3-13A-K033 RCIC Steam Line Hi DP Line Break Relay Functional Failure 0.20 Component evaluated by HCLPF calculation 14Q4233-CAL-004. (Modification/Resolution Required) 3-13A-K034 RCIC Auto Isolation Signal Relay Functional Failure 0.36 Component evaluated by HCLPF calculation 14Q4233-CAL-004. 3-13A-K044 RCIC Reactor Hi Vessel Water Level Trip Relay Functional Failure 0.36 Component evaluated by HCLPF calculation 14Q4233-CAL-004. 3-13A-K049 RCIC Low Steam Pressure
 
Auto Isolation Relay Functional Failure 0.36 Component evaluated by HCLPF calculation 14Q4233-CAL-004. 3-13A-K050 RCIC Low Steam Pressure
 
Auto Isolation Relay Functional Failure 0.29 Component evaluated by HCLPF calculation 14Q4233-CAL-004. 3-13A-K053 RCIC Reactor Hi Vessel Water Level Trip Relay Auxiliary Functional Failure 0.34 Component evaluated by HCLPF calculation 14Q4233-CAL-004.
14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 59 of 63 Equipment ID Description Failure Mode HCLPF (g) Basis 3-13A-K054 RCIC Auto Isolation Signal Relay Functional Failure 0.29 Component evaluated by HCLPF calculation 14Q4233-CAL-004. 3AC043 Emergency Shutdown Panel Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 3AC065 Rx Vessel Lvl and Pressure
 
Inst Rack A Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 3AC091 Jet Pump Inst Rack A Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 3AC270 Panel Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 3AD001 3A 125V DC Battery Anchorage 0.26 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 3AD003 Station Battery Charger 3A Block wall interaction 0.27 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 3AD017 (3FBA) Battery Main Fuse Box Anchorage >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 3AD018 (3DPA) 250v DC Distribution Panel Anchorage >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 3AD019 (3FA) 250 Volt Fuse Box Anchorage >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 3AD025 (3PPAD) 3ppad 125V DC Distribution
 
Panel Anchorage >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 3AE024 Residual Heat Exchanger Anchorage 0.26 Component evaluated by HCLPF calculation 14Q4233-CAL-002.
3AS377 Back-Up N2 Supply to Ads
 
RV's Screened >RLGM Component screened by SRT analysis. 3AS456 Transfer Switch Anchorage >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 3AT540 Instrument N2 Accumulator Screened >RLGM Component screened by SRT judgment. 3AT545 3A Srv Inst N2 Accumulator Screened >RLGM Component screened by SRT judgment. 3BC043 Panel Equipment Capacity >0.24 Component is Rule Of Box to 3AC043. Parent component is evaluated in 14Q4233-CAL-002. 3BC065 Instrument Rack Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 3BC091 Instrument Rack Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 3BC270 Panel Equipment Capacity >0.24 Component is Rule Of Box to 3AC270. Parent component is evaluated in 14Q4233-CAL-002. 3BD001 2B 125V DC Battery Anchorage 0.26 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 3BD017 Battery Main Fuse Box Anchorage >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 3BD018 250V DC Distribution Panel
 
Div. II Anchorage >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002.
3BS377 Back-Up N2 Supply to Ads
 
RV's Screened >RLGM Component is Rule Of Box to 3AS377. Parent component screens.
3BS545 Automatic Transfer Switch
 
Panel Anchorage >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 3BT540 Instrument N2 Accumulator Screened >RLGM Component screened by SRT judgment. 3BT545 3B Srv Inst N2 Accumulator Screened >RLGM Component screened by SRT judgment. 3CD001 3C 125V DC Battery Anchorage 0.26 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 3CD003 Station Battery Charger 3C Block wall interaction 0.27 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 3CD017 (3FBC) Battery Main Fuse Box Anchorage >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 3CD019 (3FC) 250 Volt Fuse Box Anchorage >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002.
14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 60 of 63 Equipment ID Description Failure Mode HCLPF (g) Basis 3CD025 (3PPCD) 3PPCD 125V DC Distribution Panel Anchorage >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 3CE024 Residual Heat Exchanger Anchorage 0.26 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 3CS377 Back-Up N2 Supply to Ads
 
RV's Screened >RLGM Component is Rule Of Box to 3AS377. Parent component screens. 3CS456 Control to Battery Charger 3CD003 Transfer Switch Block wall interaction 0.27 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 3CT540 Instrument N2 Accumulator Screened >RLGM Component screened by SRT judgment. 3CT545 3C Srv Inst N2 Accumulator Screened >RLGM Component screened by SRT judgment. 3DA-W-A (1201) RCIC MO-3-13-021 Breaker Anchorage 0.43 Component is Rule Of Box to 3DA-W-A (30D012). Parent component is evaluated in 14Q4233-CAL-002. 3DA-W-A (1203) RCIC MO-3-13-030 Breaker Anchorage 0.43 Component is Rule Of Box to 3DA-W-A (30D012).
Parent component is evaluated in 14Q4233-CAL-002. 3DA-W-A (1204) RCIC MO-3-13-027 Breaker Anchorage 0.43 Component is Rule Of Box to 3DA-W-A (30D012).
Parent component is evaluated in 14Q4233-CAL-002. 3DA-W-A (1205) RCIC MO-3-13-041 Breaker Anchorage 0.43 Component is Rule Of Box to 3DA-W-A (30D012).
Parent component is evaluated in 14Q4233-CAL-002. 3DA-W-A (1206) RCIC MO-3-13-039 Breaker Anchorage 0.43 Component is Rule Of Box to 3DA-W-A (30D012).
Parent component is evaluated in 14Q4233-CAL-002. 3DA-W-A (1207) RCIC MO-3-13-132 Breaker Anchorage 0.43 Component is Rule Of Box to 3DA-W-A (30D012).
Parent component is evaluated in 14Q4233-CAL-002. 3DA-W-A (1209) RCIC MO-3-13-131 Breaker Anchorage 0.43 Component is Rule Of Box to 3DA-W-A (30D012).
Parent component is evaluated in 14Q4233-CAL-002. 3DA-W-A (1210) RCIC MO-3-13-018 Breaker Anchorage 0.43 Component is Rule Of Box to 3DA-W-A (30D012).
Parent component is evaluated in 14Q4233-CAL-002. 3DA-W-A (1214) RCIC Cond Vac PP 30P046 Breaker Anchorage 0.43 Component is Rule Of Box to 3DA-W-A (30D012).
Parent component is evaluated in 14Q4233-CAL-002. 3DA-W-A (1215) RCIC Vac Tank Cond PP 30P048 Breaker Anchorage 0.43 Component is Rule Of Box to 3DA-W-A (30D012).
Parent component is evaluated in 14Q4233-CAL-002.
3DA-W-A (30D012) RCIC 250VDC MCC Anchorage 0.43 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 3DD001 2D 125V DC Battery Anchorage 0.26 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 3DD017 Battery Main Fuse Box Anchorage >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 3DD019 Fuse Box Anchorage >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 3GT540 Instrument N2 Accumulator Screened >RLGM Component screened by SRT judgment. 3GT545 3G Srv Inst N2 Accumulator Screened >RLGM Component screened by SRT judgment. 3KT540 Instrument N2 Accumulator Screened >RLGM Component screened by SRT judgment. 3KT545 3K Srv Inst N2 Accumulator Screened >RLGM Component screened by SRT judgment. 3OE032 RCIC Barometric Condenser (13-2) Block wall interaction 0.44 Component evaluated by HCLPF calculation 14Q4233-CAL-002. 3OE104 RCIC Turb. Lube Oil Cooler (13-2) Anchorage >0.24 Component is Rule Of Box to 30P036. Parent component is evaluated in 14Q4233-CAL-002. 3AC270 (K3A) Relay Equipment Capacity >0.24 Component is Rule Of Box to 3AC270. Parent component is evaluated in 14Q4233-CAL-002.
Component evaluated by HCLPF calculation 14Q4233-CAL-004. 3BC270 (K3B) Relay Equipment Capacity >0.24 Component is Rule Of Box to 3BC270. Parent component is evaluated in 14Q4233-CAL-002. Component evaluated by HCLPF calculation 14Q4233-CAL-004. 3AC270 (K3C) Relay Equipment Capacity >0.24 Component is Rule Of Box to 3AC270. Parent component is evaluated in 14Q4233-CAL-002. Component evaluated by HCLPF calculation 14Q4233-CAL-004.
14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 61 of 63 Equipment ID Description Failure Mode HCLPF (g) Basis 3BC270 (K3D) Relay Equipment Capacity >0.24 Component is Rule Of Box to 3BC270. Parent component is evaluated in 14Q4233-CAL-002. Component evaluated by HCLPF calculation 14Q4233-CAL-004. AO-3-07B-3511 Torus 18 Inch Vent Inboard
 
Isol Valve to Sbgt/Atmos Screened >RLGM Component screened by SRT analysis. AO-3-07B-90290 Ctmt Emerg Vent Outboard Isolation Vlv to Atmos Screened >RLGM Component screened by SRT analysis. E134 (1014) E134-T-B 30B059 Breaker Equipment Capacity >0.24 Component is Rule Of Box to E134 (30B010). Parent component is evaluated in 14Q4233-CAL-002. E134 (30B010) E134 Load Center Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. E134-T-B (30B059) E134-T-B Motor Control Center Block wall interaction 0.27 Component evaluated by HCLPF calculation 14Q4233-CAL-002. E134-T-B (5924) Alt Feed For Uninterrupt AC Power Supp Inverter 30D37 Block wall interaction 0.27 Component is Rule Of Box to E134-T-B (30B059).
Parent component is evaluated in 14Q4233-CAL-002. E134-T-B (5931) 125V. D.C. Battery Charger 3AD03 Block wall interaction 0.27 Component is Rule Of Box to E134-T-B (30B059).
Parent component is evaluated in 14Q4233-CAL-002. E334 (1213) E334-R-B 30B038 Breaker Equipment Capacity >0.24 Component is Rule Of Box to E334 (30B012). Parent component is evaluated in 14Q4233-CAL-002. E334 (30B012) E334 Load Center Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002.
E334-R-B (30B038) MCC 30B038 Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. E334-R-B (3821) MO-3-10-038A Breaker Equipment Capacity >0.24 Component is Rule Of Box to E334-R-B (30B038).
Parent component is evaluated in 14Q4233-CAL-002. E334-R-B (3824) MO-3-10-026A Breaker Equipment Capacity >0.24 Component is Rule Of Box to E334-R-B (30B038).
Parent component is evaluated in 14Q4233-CAL-002. E334-R-B (3831) MO-3-10-039A Breaker Equipment Capacity >0.24 Component is Rule Of Box to E334-R-B (30B038).
Parent component is evaluated in 14Q4233-CAL-002. E334-R-B (3832) MO-3-10-034A Breaker Equipment Capacity >0.24 Component is Rule Of Box to E334-R-B (30B038).
Parent component is evaluated in 14Q4233-CAL-002. E334-R-B (3844) MO-3-10-031A Breaker Equipment Capacity >0.24 Component is Rule Of Box to E334-R-B (30B038).
Parent component is evaluated in 14Q4233-CAL-002. E334-R-B (3851) MO-3-10-25A Alt Breaker Equipment Capacity >0.24 Component is Rule Of Box to E334-R-B (30B038).
Parent component is evaluated in 14Q4233-CAL-002. E334-R-B (3862) MO-3-10-174 Breaker Equipment Capacity >0.24 Component is Rule Of Box to E334-R-B (30B038).
Parent component is evaluated in 14Q4233-CAL-002. E334-R-B (3863) MO-3-10-176 Breaker Equipment Capacity >0.24 Component is Rule Of Box to E334-R-B (30B038). Parent component is evaluated in 14Q4233-CAL-002. E334-R-B (3882) Norm Fdr for 120V Instr Pnl 30Y35 Trans 30X135 Equipment Capacity >0.24 Component is Rule Of Box to E334-R-B (30B038).
Parent component is evaluated in 14Q4233-CAL-002. E334-R-B (3893) 125V Battery Charger C 3C Transfer Switch 30S704 Equipment Capacity >0.24 Component is Rule Of Box to E334-R-B (30B038).
Parent component is evaluated in 14Q4233-CAL-002. INV-3-13-90 RCICs-125 VDC Bus 'A' Power Distribution Equipment Capacity 0.44 Component is Rule Of Box to 30C019. Parent component is evaluated in 14Q4233-CAL-002. J2919 J-Box at E134 LC Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. J2920 J-Box at E334 LC Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. LI-3-02-3-085A Reactor Vessel High Water Equipment Capacity >0.24 Component is Rule Of Box to 30C005A. Parent component is evaluated in 14Q4233-CAL-002. LI-3-02-3-113 Reactor Water Level Equipment Capacity >0.24 Component is Rule Of Box to 30C003. Parent component is evaluated in 14Q4233-CAL-002. LR/TR-9123A Torus Water Level/Temperature Recorder Equipment Capacity >0.24 Component is Rule Of Box to 30C004C. Parent component is evaluated in 14Q4233-CAL-002. LT-3-02-3-072A Reactor Vessel Water Level Equipment Capacity >0.24 Component is Rule Of Box to 3AC065. Parent component is evaluated in 14Q4233-CAL-002. LT-3-02-3-113 Level Transmitter Equipment Capacity >0.24 Component is Rule Of Box to 3BC091. Parent component is evaluated in 14Q4233-CAL-002. LT-9123A Torus Water Level Screened >RLGM Component screened by SRT analysis.
14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 62 of 63 Equipment ID Description Failure Mode HCLPF (g) Basis MO-3-10-025A RHR Inner Injection Valve to Recirc Loop A Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. MO-3-10-026A RHR Loop A D/W Spray O/B Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. MO-3-10-031A RHR Loop A D/W Spray I/B Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. MO-3-10-034A RHR Loop A FFT Valve Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. MO-3-10-038A RHR Loop A Torus Spray Screened >RLGM Component screened by SRT analysis. MO-3-10-039A RHR Loop A Torus Valve Screened >RLGM Component screened by SRT analysis. MO-3-10-174 HPSW to RHR Inner X-Tie Screened >RLGM Component screened by SRT analysis. MO-3-10-176 HPSW to RHR Outer X-Tie Screened >RLGM Component screened by SRT analysis. MO-3-13-018 RCIC Pump Suction from Condensate Storage Tank Screened >RLGM Component screened by SRT analysis. MO-3-13-021 RCIC Discharge to Feedwater Line B Screened >RLGM Component screened by SRT judgment. MO-3-13-027 RCIC Minimum Flow Valve Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. MO-3-13-030 RCIC Full Flow Test Valve Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. MO-3-13-039 RCIC Pump Torus Suction Outer Screened >RLGM Component screened by SRT analysis. MO-3-13-041 RCIC Pump Torus Suction Inner Screened >RLGM Component screened by SRT analysis. MO-3-13-131 RCIC Turbine Steam Supply
 
Valve Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. MO-3-13-132 RCIC Cooling Water Supply to Lo Clr + Barometric Cdsr Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. N310025A Cabinet Equipment Capacity >0.24 Component evaluated by HCLPF calculation 14Q4233-CAL-002. PI-3-06-090A Reactor Wide Range Press
 
Ind Equipment Capacity >0.24 Component is Rule Of Box to 30C005A. Parent component is evaluated in 14Q4233-CAL-002. PI-3-06-090B Reactor Wide Range Press
 
Ind Equipment Capacity >0.24 Component is Rule Of Box to 30C005A. Parent component is evaluated in 14Q4233-CAL-002. PI-3-06-090C Reactor Wide Range Press
 
Ind Equipment Capacity >0.24 Component is Rule Of Box to 30C005A. Parent component is evaluated in 14Q4233-CAL-002. PI-3-13-094 RCIC Pump Turb Stm Press Equipment Capacity >0.24 Component is Rule Of Box to 30C004C. Parent component is evaluated in 14Q4233-CAL-002. PR/LR-3-06-096 Reactor Level/Steam Flow
 
Ratio Equipment Capacity >0.24 Component is Rule Of Box to 30C005A. Parent component is evaluated in 14Q4233-CAL-002. PR/TR-5805 Containment Pressure/Temp Equipment Capacity >0.24 Component is Rule Of Box to 30C003. Parent component is evaluated in 14Q4233-CAL-002. PS-3-13-67-1 Pressure Switch Anchorage 0.41 Component is Rule Of Box to 30C095. Parent component is evaluated in 14Q4233-CAL-002. Component evaluated by HCLPF calculation 14Q4233-CAL-004. PS-3-13-72A Pressure Switch Functional Failure 0.27 Component evaluated by HCLPF calculation 14Q4233-CAL-004. PS-3-13-72B Pressure Switch Functional Failure 0.27 Component evaluated by HCLPF calculation 14Q4233-CAL-004. PS-3-13-87A Pressure Switch Functional Failure 0.27 Component evaluated by HCLPF calculation 14Q4233-CAL-004. PS-3-13-87B Pressure Switch Functional Failure 0.27 Component evaluated by HCLPF calculation 14Q4233-CAL-004. PT-3-02-3-404A Reactor Pressure Equipment Capacity >0.24 Component is Rule Of Box to 3AC065. Parent component is evaluated in 14Q4233-CAL-002. PT-3-02-3-404C Reactor Pressure Transmitter Equipment Capacity >0.24 Component is Rule Of Box to 3AC091. Parent component is evaluated in 14Q4233-CAL-002.
14Q4233-RPT-004Rev. 3 Correspondence No.: RS-14-300 Page 63 of 63 Equipment ID Description Failure Mode HCLPF (g) Basis PT-3-06-053A Reactor Wide Range -Pressure Equipment Capacity >0.24 Component is Rule Of Box to 3AC065. Parent component is evaluated in 14Q4233-CAL-002. PT-3-06-053B Reactor Wide Range -
Pressure Equipment Capacity >0.24 Component is Rule Of Box to 3BC065. Parent component is evaluated in 14Q4233-CAL-002. PT-3-06-053C Reactor Wide Range -
Pressure Equipment Capacity >0.24 Component is Rule Of Box to 3AC065. Parent component is evaluated in 14Q4233-CAL-002. PT-3-13-068 RCIC Turbine Steam Supply Pressure Anchorage 0.41 Component is Rule Of Box to 30C095. Parent component is evaluated in 14Q4233-CAL-002. PT-5805 Drywell Pressure Screened >RLGM Component screened by SRT analysis. RV-3-02-071A 3A Safety Relief Valve Screened >RLGM Component screened by SRT judgment. RV-3-02-071B 3B Safety Relief Valve Screened >RLGM Component screened by SRT judgment. RV-3-02-071C 3C Safety Relief Valve Screened >RLGM Component screened by SRT judgment. RV-3-02-071G 3G Safety Relief Valve Screened >RLGM Component screened by SRT judgment. RV-3-02-071K 3K Safety Relief Valve Screened >RLGM Component screened by SRT judgment. XAM-3-02 117A Reactor Water Level Wide Range Equipment Capacity 0.44 Component is Rule Of Box to 30C818. Parent component is evaluated in 14Q4233-CAL-002.
 
1 . 2. Enclosure 2 Peach Bottom Atomic Power Station, Units 2 and 3
 
==SUMMARY==
OF REGULATORY COMMITMENTS The following tat)le identifies commitments made in this document. (Any other actions discussed in the submittal represent intended or planned actions. They are described to the NRC for the NRC's information and are not regulatory commitments.)
COMMITMENT TYPE COMMITTED COMMITMENT DATE OR. ONE-TIME ACTION PROGRAMMATIC "OUTAGE" (Yes/No) (Yes/No) Complete further evaluation and implement Unit 3 -P3R21 Yes No modifications, if required, to increase seismic (Fall 2017) rna1*gin for the following plant relay: Unit 3 Relay 3-13A-K033 Corn further evaluation and Unit 2 -P2R22 Yes No modifications, if required, to increase seismic (Fall 2018) margin for the following plant relay: Unit 2 Relay 2-13A-K033}}

Latest revision as of 02:21, 28 April 2019

Redirect to:

Retrieved from "https://kanterella.com/w/index.php?title=ML14353A333&oldid=1531990"