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{{#Wiki_filter:~ENERGY,R. Michael GloverH. B. Robinson SteamElectric Plant Unit 2Site Vice PresidentDuke Energy Progress3581 West Entrance RoadHartsville, SC 295500:843 857 1704F: 843 857 1319Mike. Glover~a duke-energy.comRN P-RA/14-0129December 17, 201410 CFR 50.54(f)ATTN: Document Control DeskU. S. Nuclear Regulatory CommissionWashington, DC 20555-0001H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2DOCKET NO. 50-261 1 RENEWED LICENSE NO. DPR-23
{{#Wiki_filter:~ENERGY,R. Michael GloverH. B. Robinson SteamElectric Plant Unit 2Site Vice President Duke Energy Progress3581 West Entrance RoadHartsville, SC 295500:843 857 1704F: 843 857 1319Mike. Glover~a duke-energy.com RN P-RA/14-0129 December 17, 201410 CFR 50.54(f)ATTN: Document Control DeskU. S. Nuclear Regulatory Commission Washington, DC 20555-0001 H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2DOCKET NO. 50-261 1 RENEWED LICENSE NO. DPR-23


==Subject:==
==Subject:==
H. B. Robinson Steam Electric Plant, Unit No. 2 Expedited Seismic Evaluation ProcessReport (CEUS Sites), Response to NRC Request for Information Pursuant to 10 CFR50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review ofInsights from the Fukushima Dai-ichi Accident
 
H. B. Robinson Steam Electric Plant, Unit No. 2 Expedited Seismic Evaluation ProcessReport (CEUS Sites), Response to NRC Request for Information Pursuant to 10 CFR50.54(f)
Regarding Recommendation 2.1 of the Near-Term Task Force Review ofInsights from the Fukushima Dai-ichi Accident


==References:==
==References:==
: 1. NRC Letter, Request for Information Pursuant to Title 10 of the Code of Federal Regulations50.54(0 Regarding Recommendations 2.1, 2.3, and 9.3, of the Near-Term Task Force Reviewof Insights from the Fukushima Dai-ichi Accident, dated March 12, 20122. NEI Letter, Proposed Path Forward for NTTF Recommendation 2.1: Seismic Reevaluations,dated April 9, 2013, ADAMS Accession No. ML13101A3793. NRC Letter, Electric Power Research Institute Report 3002000704, "Seismic EvaluationGuidance: Augmented Approach for the Resolution of Fukushima Near-Term Task ForceRecommendation 2.1: Seismic," as an Acceptable Alternative to the March 12, 2012,Information Request for Seismic Reevaluations, dated May 7, 2013, ADAMS Accession No.ML13106A331Ladies and Gentlemen:On March 12, 2012, the Nuclear Regulatory Commission (NRC) issued a 50.54(f) letter to all powerreactor licensees and holders of construction permits in active or deferred status. Enclosure 1 ofReference 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 ofReference 1.In Reference 2, the Nuclear Energy Institute (NEI) requested NRC agreement to delay submittal ofthe final CEUS Seismic Hazard Evaluation and Screening Reports so that an update to the ElectricPower Research Institute (EPRI) ground motion attenuation model could be completed and used todevelop that information. NEI proposed that descriptions of subsurface materials and propertiesand base case velocity profiles be submitted to the NRC by September 12, 2013, with theremaining seismic hazard and screening information submitted by March 31, 2014. NRC agreedwith that proposed path forward in Reference 3. Ac {
: 1. NRC Letter, Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(0 Regarding Recommendations 2.1, 2.3, and 9.3, of the Near-Term Task Force Reviewof Insights from the Fukushima Dai-ichi  
Serial: RNP-RA/14-0129U. S. Nuclear Regulatory CommissionPage 2Reference 1 requested that licensees provide interim evaluations and actions taken or planned toaddress the higher seismic hazard relative to the design basis, as appropriate, prior to completionof the risk evaluation. In accordance with the NRC endorsed guidance in Reference 3, the attachedExpedited Seismic Evaluation Process Report for H. B. Robinson Steam Electric Plant, Unit No. 2provides the information described in Section 7 of Reference 3 in accordance with the scheduleidentified in Reference 2.This letter contains no new regulatory commitments.If you have any questions or require additional information, please contact Richard Hightower,Manager, Nuclear Regulatory Affairs at (843)-857-1329.I declare under the penalty of perjury that the foregoing is true and correct.Executedon '2_0 1Sincerely,R.ý Mihe GýILoveR. Michael GloverSite Vice PresidentRMG/shc
: Accident, dated March 12, 20122. NEI Letter, Proposed Path Forward for NTTF Recommendation 2.1: Seismic Reevaluations, dated April 9, 2013, ADAMS Accession No. ML13101A379
: 3. NRC Letter, Electric Power Research Institute Report 3002000704, "Seismic Evaluation Guidance:
Augmented Approach for the Resolution of Fukushima Near-Term Task ForceRecommendation 2.1: Seismic,"
as an Acceptable Alternative to the March 12, 2012,Information Request for Seismic Reevaluations, dated May 7, 2013, ADAMS Accession No.ML13106A331 Ladies and Gentlemen:
On March 12, 2012, the Nuclear Regulatory Commission (NRC) issued a 50.54(f) letter to all powerreactor licensees and holders of construction permits in active or deferred status. Enclosure 1 ofReference 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 ofReference 1.In Reference 2, the Nuclear Energy Institute (NEI) requested NRC agreement to delay submittal ofthe final CEUS Seismic Hazard Evaluation and Screening Reports so that an update to the ElectricPower Research Institute (EPRI) ground motion attenuation model could be completed and used todevelop that information.
NEI proposed that descriptions of subsurface materials and properties and base case velocity profiles be submitted to the NRC by September 12, 2013, with theremaining seismic hazard and screening information submitted by March 31, 2014. NRC agreedwith that proposed path forward in Reference  
: 3. Ac {
Serial: RNP-RA/14-0129 U. S. Nuclear Regulatory Commission Page 2Reference 1 requested that licensees provide interim evaluations and actions taken or planned toaddress the higher seismic hazard relative to the design basis, as appropriate, prior to completion of the risk evaluation.
In accordance with the NRC endorsed guidance in Reference 3, the attachedExpedited Seismic Evaluation Process Report for H. B. Robinson Steam Electric Plant, Unit No. 2provides the information described in Section 7 of Reference 3 in accordance with the scheduleidentified in Reference 2.This letter contains no new regulatory commitments.
If you have any questions or require additional information, please contact Richard Hightower,
: Manager, Nuclear Regulatory Affairs at (843)-857-1329.
I declare under the penalty of perjury that the foregoing is true and correct.Executedon  
'2_0 1Sincerely, R.ý Mihe GýILoveR. Michael GloverSite Vice President RMG/shc


==Enclosure:==
==Enclosure:==
Expedited Seismic Evaluation Process Report for H. B. Robinson Steam ElectricPlant, Unit No. 2cc: Ms. M. C. Barillas, NRC Project Manager, NRRMr. K. M. Ellis, NRC Senior Resident InspectorMr. V. M. McCree, NRC Region II Administrator Expedited Seismic Evaluation Process ReportExpedited Seismic Evaluation Process ReportForH. B. Robinson Steam Electric Plant, Unit No. 2Page 3 of 46 Expedited Seismic Evaluation Process ReportEXPEDITED SEISMIC EVALUATION PROCESS REPORTTABLE OF CONTENT1.0 Purpose and Objective ................................................................................. 072.0 Brief Summary of the FLEX Seismic Implementation Strategies .......................... 073.0 Equipment Selection Process and ESEL ......................................................... 133.1 Equipment Selection Process and ESEL ............................................... 133.1.1 ESEL Development ................................................................ 143.1.2 Power Operated Valves .......................................................... 143.1.3 Pull Boxes ........................................................................... 143.1.4 Termination Cabinets .............................................................. 153.1.5 Critical Instrumentation Indicators .............................................. 153.1.6 Phase 2 and Phase 3 Piping Connections .................................. 153.2 Justification for Use of Equipment That is Not the Primary Means forFLEX Implementation ..................................................................... 154.0 Ground Motion Response Spectrum .............................................................. 164.1 Plot of GMRS Submitted by H.B. Robinson Steam Electric Plant ............... 164.2 Comparison to SSE .......................................................................... 195.0 Review Level Ground Motion (RLGM) ............................................................ 235.1 Description of RLGM Selected ............................................................ 235.2 Method to Estimate ISRS .................................................................. 266.0 Seismic Margin Evaluation Approach ............................................................ 286.1 Summary of Methodologies Used ...................................................... 286.2 HCLPF Screening Process ................................................................ 296.3 Seismic Walkdown Approach ............................................................ 306.3.1 W alkdown Approach .............................................................. 306.3.2 Application of Previous W alkdown Information ............................ 326.3.3 Significant Walkdown Findings ................................................. 336.4 HCLPF Calculation Process .............................................................. 336.5 Functional Evaluation of Relays .......................................................... 336.6 Tabulated ESEL HCLPF Values (Including Key Failure Modes) ................. 347.0 Inaccessible Items ..................................................................................... 367.1 Identification of ESEL Items Inaccessible for Walkdown .......................... 367.2 Planned Walkdown/Evaluation Schedule/Close Out ................................ 368.0 ESEP Conclusions and Results .................................................................... 378.1 Supporting Information ...................................................................... 378.2 Identification of Planned Modifications ................................................. 388.3 Modification Implementation Schedule ................................................... 398.4 Summary of Planned Actions ............................................................ 399.0 References .............................................................................................. 40Page 4 of 46 Expedited Seismic Evaluation Process ReportList of FiguresFigure 2.1: Mark-Up Showing Addition of FLEX Tee Connection (AFW-166)to SDAFW Discharge at AFW -121 .......................................................... 9Figure 2.2: Mark-Up Showing Alternate Mechanical FLEX Connection (AFW-165) insidethe MDAFW Room on Line 4-AFW-23 and Upstream of AFW-54 .................. 10Figure 4.1: Plot of 1 E-4 and 1 E-5 UHRS and GMRS at Control Point forthe H.B. Robinson Steam Electric Plant ................................................... 18Figure 4.2: Comparison of the GMRS, SSE, and Ground LevelResponse Spectrum from Time History ................................................... 22Figure 5-1: Plot of 5% Damping 2 x SSE, 2 x Ground Level Response Spectrum,a nd G M R S ....................................................................................... 2 6Figure 6.1: Comparison of the H.B. Robinson Steam Electric Plant IPEEE RLE, ESEPRLGM, SSE, and Ground Level (El. 226ft) Spectrum from Time History,and 2 x Ground Level (El. 226ft) Spectrum from Time History ............... 29List of TablesTable 4-1: GMRS and UHRS for the H.B. Robinson Steam Electric Plant ...................... 17Table 4-2a: Original SSE Based on 0.2g Housner Spectrum for the H.B. RobinsonS team E lectric P lant ............................................................................ 20Table 4-2b: Ground Level Response Spectrum Based on Time History for theH.B. Robinson Steam Electric Plant ........................................................ 21Table 5-1: RLGM for H.B. Robinson Steam Electric Plant .......................................... 24Table 5-2: Ratio of G M RS to SSE ......................................................................... 25Table 6-1: Functional and Anchorage HCLPF Capacity Results .................................. 35AttachmentsAttachment A -H.B. Robinson Steam Electric Plant ESELAttachment B -H.B. Robinson Steam Electric Plant Unit 2 RCS Cooling StrategiesAttachment C -H.B. Robinson Steam Electric Plant Unit 2 RCS Boration and Makeup StrategiesAttachment D -FLEX Flow PathPage 5 of 46 Expedited Seismic Evaluation Process ReportEXECUTIVE SUMMARYAn Expedited Seismic Evaluation Process has been completed for the H.B. Robinson SteamElectric Plant site based on endorsed guidance outlined in Electric Power Research Institute (EPRI)3002000704 (Reference 2). The work includes screening, equipment selection, development of theRLGM and in-structure demands, evaluating seismic capacity of components and development ofHigh Confidence of Low Probability of Failure (HCLPF) calculations, and implementation ofnecessary plant modifications. HCLPF calculations revealed that Motor Control Center (MCC-A)required modification for the beyond design basis ground motion. Modifications have beendeveloped and implemented for MCC-A and a similar cabinet, MCC-B. Seismic margin above 2XSSE was also added to a group of instrument racks (Hagan Racks) by validating the boltingintegrity of the top braces. All items in the ESEL have seismic capacity that exceeds the demand ofthe RLGM. The ESEL has been updated to consider new equipment in FLEX strategy as outlinedin the updated Overall Integrated Plan. The FLEX strategy was subjected to critical path analysisand all the items required under the ESEP guidelines are included in the ESEL list.Page 6 of 46 Expedited Seismic Evaluation Process Report1.0 Purpose and ObjectiveFollowing 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 NRCprocesses and regulations and to determine if the agency should make additional improvements toits regulatory system. The NTTF developed a set of recommendations intended to clarify andstrengthen the regulatory framework for protection against natural phenomena. Subsequently, theNRC issued a 10 CFR 50.54(f) letter on March 12, 2012 (Reference 1), requesting information toassure 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 reevaluatethe seismic hazards at their sites against present-day NRC requirements and guidance.Depending on the comparison between the reevaluated seismic hazard and the current designbasis, further risk assessment may be required. Assessment approaches acceptable to the staffinclude a seismic probabilistic risk assessment (SPRA), or a seismic margin assessment (SMA).Based upon the assessment results, the NRC staff will determine whether additional regulatoryactions are necessary.This report describes the Expedited Seismic Evaluation Process (ESEP) undertaken for H.B.Robinson Steam Electric Plant (RNP). The intent of the ESEP is to perform an interim action inresponse to the NRC's 50.54(f) letter (Reference 1) to demonstrate seismic margin through areview of a subset of the plant equipment that can be relied upon to protect the reactor corefollowing beyond design basis seismic events.The ESEP is implemented using the methodologies in the NRC endorsed guidance in EPRI3002000704, Seismic Evaluation Guidance: Augmented Approach for the Resolution of FukushimaNear-Term Task Force Recommendation 2.1: Seismic (Reference 2).The objective of this report is to provide summary information describing the ESEP evaluations andresults. The level of detail provided in the report is intended to enable NRC to understand theinputs used, the evaluations performed, and the decisions made as a result of the interimevaluations.2.0 Brief Summary of the FLEX Seismic Implementation StrategiesThe H.B. Robinson Steam Electric Plant FLEX strategies for Reactor Core Cooling and HeatRemoval, Reactor Inventory Control/Long-term Subcriticality, and Containment Function aresummarized below. The FLEX flow path is shown in Attachment D. The summary is derived fromthe H.B. Robinson Steam Electric Plant Overall Integrated Plan (OIP) in Response to the March 12,2012, Commission Order EA-12-049 (Reference 3), as supplemented by six-month updates(References 30, 31, and 32). Note that the H.B. Robinson Overall Integrated Plan (as amended in 6month updates) is based on Engineering Change (EC) 88926 (Reference 33).Reactor Core Cooling and Heat RemovalNEI 12-06, Diverse and Flexible Coping Strategies (FLEX) Implementation Guideline, Revision 0(Reference 34), requires that Auxiliary Feedwater (AFW) cooling be available to provide secondarymakeup sufficient to maintain or restore Steam Generator (SG) level with installed equipment to thegreatest extent possible. Beyond the use of installed equipment, steam generators must be able tobe depressurized in order to support makeup via portable pumps. Multiple and diverse connectionpoints for the portable pumps must be provided and cooling water must be available indefinitely.Refer to Attachment B (Reactor Coolant System Cooling Strategies) for depiction of the followingdiscussion.Page 7 of 46 Expedited Seismic Evaluation Process ReportThe H.B Robinson Steam Electric Plant FLEX strategies require that the AFW be in operationwithin 61 minutes of event initiation. With the loss of AC power, a minimum of one steam supplyvalve (MS-V1-8A, MS-V1-8B, or MS-V1-8C) to Steam Driven Auxiliary Feedwater Pump (SDAFWP)and one AFW valve (AFW-V2-14A, AFW-V2-14B, AFW-V2-14C) to the steam generators must bemanually operated. These required valves are all located in seismic Class 1 bay of the TurbineBuilding.Additional portable backup for Steam Generator makeup is required per Section 3.2.2(13) of NEI12-06. The H.B. Robinson Steam Electric Plant has two strategies for portable backup. The firststrategy developed to satisfy this requirement is staging of two (2) intermediate pressure pumps(300 gpm at pressure of 1,000 psig) for all seismic events as described in detail below. The secondstrategy developed to satisfy the condition of Section 3.2.2(13) of NEI 12-06 is to store a Halepumper in a seismically robust Permanent FLEX Storage Building (PFSB). This strategy will involvethe use of the same primary and alternate connections described in the following paragraph, andwill require SG depressurization.The two (2) pre-staged portable pumps (300 gpm at 1,000 psig) eliminate the need to depressurizethe Steam Generators in the event the backup AFW feed capability is needed due to an AFWinterruption early in the ELAP transient as a result of seismic event. Either of the portable pumpscan take suction from a variety of plant sources (described below) and can be tied directly into theauxiliary feedwater system. Engineering Change 95266, Isolation Valves And Connection For AFW-FUKUSHIMA-Admin (Reference 48) was developed to add a FLEX tee connection (AFW-166) tothe SDAFWP discharge at AFW-121 (see Figure 2-1). Access to this primary connection is throughthe seismically qualified Turbine Building Class 1 bay. Engineering Change 90623, New Pipe TeeAnd Standard Connection For NTTF 4.2 (FLEX) (Reference 47) develops an alternate mechanicalFLEX connection (AFW-165) inside the MDAFWP room on line 4-AFW-23 and upstream of AFW-54 (See Figure 2-2). EC90623 will be implemented during Refueling Outage, R0229. TheMDAFW room is housed in the seismic Class 1 Reactor Auxiliary Building (RAB).Page 8 of 46 Expedited Seismic Evaluation Process ReportFigure 2-1: Mark-Up Showing Addition of FLEX Tee Connection (AFW-166) to SDAFWDischarge at AFW-121Page 9 of 46 Expedited Seismic Evaluation Process ReportFigure 2-2: Mark-Up Showing Alternate Mechanical FLEX Connection (AFW-165) inside theMDAFWP Room on Line 4-AFW-23 and Upstream of AFW-54There are several sources for sustained cooling water supply. The primary source of AFWinventory is the seismically qualified condensate storage tank (CST) and its level instrumentation.The CST is seismically robust and is the installed source of AFW to the SDAFWP. However, theCST inventory is not sufficient for indefinite coping (mission time is approximately 4 hours using theSDAFWP). A secondary source of AFW inventory is the "Tank Farm" (portable pump) inside theprotected area that supplies the two pre-staged portable pumps (each with capacity of 300 gpmand 1,000 psig pressure as noted in the seismic strategy above). This source has a capacity ofapproximately 120,000 gallons and 10 hours of mission time using a pre-staged portable pump.The only other assured source of water is the Ultimate Heat Sink (Lake Robinson) which perrestrictions outlined in NEI 12-06 can only be accessed using portable equipment (assumes normalPage 10 of 46 Expedited Seismic Evaluation Process Reportaccess to the ultimate heat sink is lost). Given these limitations, one Phase 2/3 seismic strategy isto provide an indefinite supply of water to the CST and the SDAFWP by staging a portable dieselpumper at Lake Robinson with hoses routed to the CST. EC 90622, Standard Piping ConnectionsFor NTTF 4.2 (FLEX) (Reference 43) adds a FLEX connection at valve C-66 to provide anindefinite water supply to the CST. This can be accomplished during the initial CST/Tank Farmmission time of 14 hours.The H.B. Robinson Steam Electric Plant has developed several options for the Steam Generatordepressurization capability. The Steam Generator Power Operated Relief Valves (PORVs) arenormally operated using the Instrument Air System or with backup Nitrogen System and alignedusing Attachment 2 of EOP-ECA-0.0 (Reference 35). However, neither the primary Instrument Airnor the backup Nitrogen System are seismically qualified. Therefore, the primary Instrument Air andthe backup Nitrogen System cannot be relied upon during or after seismic events. The Main SteamSafety Valves are an alternate option to depressurize the Steam Generators but this option is notrecommended per the PA-PSC-0965, PWROG Core Cooling Position Paper (Reference 37) andWCAP-17601-P, Revision 1, Reactor Coolant System response to Extended Loss of AC PowerEvent for Westinghouse, Combustion Engineering, and Babcock & Wilcox NSSS Designs forPhase Boration, August 2012 (Reference 36), which state that remaining on the Main Steam SafetyValves for an extended period may lead to failure of the valve(s) which subsequently will causeexcessive and uncontrolled RCS cooldown.Current strategy is to align portable nitrogen tanks to the Steam Generator PORV header usingAttachment 1 (Connecting Emergency Pressure Source to Operate SG PORVS) or Attachment 2(S/G Manual Depressurization) of RNP procedure EDMG-004, Steam Generators (Reference 38).In addition to the SG PORV capabilities recommended in Reference 37, the H.B. Robinson SteamElectric Plant has also developed a strategy to cooldown the RCS using the main steam lineisolation valve bypass lines. The strategy is detailed in Section 3.27 (Cooldown Using MSIV BypassLines) of calculation RNP-M/MECH-1712, Appendix R Mechanical Basis (Reference 39). Thiscapability results in a cooldown rate of 830/hr which bounds the recommended Westinghousecooldown rate of 750/hr described in Reference 37.After initiation of depressurization, it is desirable to isolate the Safety Injection (SI) Accumulators inorder to prevent injection of nitrogen into the RCS which will impede natural circulation cooldown.During an ELAP, power to the SI Accumulator isolation valves is lost. Although, the isolation valvescan be operated manually, they are located inside the Containment Building and it is undesirable toperform this operation at this time due to personnel safety. The valves are powered by MCC 5 andMCC 6 and will be re-powered via Emergency Buses El and E2 with portable diesel generatorsstaged in the seismic Class 1 Reactor Auxiliary Building (Drumming Room) for re-powering the Aand B Battery Chargers (see EC 90617 [Reference 40]). DB-50 Bus Feed Adapters can beinstalled in each of the Emergency Buses El and E2 and will be connected to the output of theDiesel Generators. As part of the Phase 2 strategy, Steam Generator pressure will be maintainedabove the pressure corresponding to the SI Accumulator injection (240 psig) until the SIAccumulator isolation valves are closed using FLEX Support Guideline (FSG) 10, Passive RCSInjection Isolation (Reference 41).Reactor Inventory Control/Long-Term SubcriticalityRefer to Attachment C (Reactor Coolant System Boration and Makeup Strategies) for a depiction ofthe following discussion. There is no installed means of providing borated makeup following anELAP. The primary method of boration and inventory control is the use of portable high pressureand low volume pump directly connected to the Charging Lines or Safety Injection Headers fromthe Refueling Water Storage Tank (RWST) or a portable tanker containing borated water (seeEC95216, NTTF 2.1 Interim Action RCS Injection [Reference 42]). The RWST is seismicallyPage 11 of 46 Expedited Seismic Evaluation Process Reportdesigned and will remain operational during and after a design basis seismic event. The makeupcapacity of the portable pump is 60 gpm at a pressure of 2,000 psig which is adequate for thebounding analysis in WCAP-1760-P (Reference 36). Phase 3 inventory control will beaccomplished using the same portable Phase 2 boration/makeup strategy. Portable high pressurepumping and portable tanker capability will be stored in the PFSB to support this strategy.EC 90622 (Reference 43) adds a FLEX connection to the exposed end downstream of the normallylocked closed drain valve (SI-837) located at the base of the RWST to access this borated water ifit available. This portable strategy will deliver borated water to the RCS through valves CVC-121A/B (primary) or SI-888P/S (alternate).Containment FunctionCalculation RNP-M/MECH-1877, RNP Extended Loss of AC Power (ELAP) Containment Response(Reference 45) was developed to determine the containment temperature and pressure responseassuming an ELAP and a trip from 100% reactor power at 100 days into the cycle. Results inReference 45 indicate that the Containment Building design limits for temperature and pressure willnot be challenged in the first 43 days following the event. This analysis assumes that: (1) no actionis taken to cool, spray, or vent the containment; and (2) low leakage RCP seals are installed.Therefore, Phase 1 and 2 strategies are not required. There is sufficient time and resources inPhase 3 to assemble a strategy using the National Safer Response Center (NSRC) pumpers andgenerators, prefabricated electrical connections, and prefabricated SW connections that will bestored in the PFSB. FSG-12, Alternate Containment Cooling (Reference 46) provides instructionsfor several existing strategies including external containment cooling which does not require use ofany plant system. These particular activities will be determined and directed by the EmergencyResponse Organization (Technical Support Center) based on the effects of the Beyond DesignBasis External Event (BDBEE) and the state of existing equipment.InstrumentationInstrumentation channels that are powered by station batteries will be lost upon depletion of thebatteries. FLEX strategies to improve battery coping occur by extending Phase 1. Phase 1 isextended by strategic load shedding followed by additional deep load shedding in the first hour ofthe event to extend battery coping times to 3.25 -3.75 hours. Phases 2 and 3 battery copingrequire portable diesel generators to power the vital battery chargers. Two FLEX diesel generatorswill be mounted in their deployed positions near the battery chargers and within the ReactorAuxiliary Building. Each generator will be sized to power two vital battery chargers, room air supplyand exhaust fans, and safety injection accumulator isolation valves. Electrical cables and pre-installed connectors will be routed from the FLEX diesel generators to the battery room for quickconnection of the cables to each of the battery chargers. The primary strategy is to power the A andB vital battery chargers from one or both of the pre-staged FLEX generators. The alternate is topower the A-1 and B-1 vital battery chargers from one or both of the pre-staged FLEX generators.See Reference 40 for complete details of this strategy.Page 12 of 46 Expedited Seismic Evaluation Process Report3.0 Equipment Selection Process and ESELThe selection of equipment for the Expedited Seismic Equipment List (ESEL) followed theguidelines of EPRI 3002000704. The complete ESEL for H. B. Robinson Unit 2 is presented inAttachment A.3.1 Equipment Selection Process and ESELThe selection of equipment to be included on the ESEL was based on installed plant equipmentcredited in the FLEX strategies during Phase 1, 2 and 3 mitigation of a BDBEE as described in theH.B. Robinson Steam Electric Plant OIP (Reference 3) in response to the March 12, 2012Commission Order EA-12-049 as revised in References 30 through 32.The scope of "installed plant equipment" includes equipment relied upon for the FLEX strategies tosustain the critical functions of core cooling and containment integrity consistent with Reference 3and References 30 through 32. FLEX recovery actions are excluded from the ESEP scope perEPRI 3002000704. The overall list of planned FLEX modifications and the scope for considerationherein is limited to those required to support core cooling, reactor coolant inventory andsubcriticality, and containment integrity functions. Portable and pre-staged FLEX equipment (notpermanently installed) are excluded from the ESEL per EPRI 3002000704.The ESEL component selection followed the EPRI guidance outlined in Section 3.2 of EPRI3002000704.1. The scope of components is limited to that required to accomplish the core cooling andcontainment safety functions identified in Table 3-2 of EPRI 3002000704. The instrumentationmonitoring requirements for core cooling/containment safety functions are limited to thoseoutlined in the EPRI 3002000704 guidance, and are a subset of those outlined in the H.B.Robinson Steam Electric Plant OIP and as revised in the first , second and third six-monthstatus reports.2. The scope of components is limited to installed plant equipment, and FLEX connectionsnecessary to implement the H.B. Robinson Steam Electric Plant OIP (Reference 3) inresponse to the March 12, 2012 Commission Order EA-12-049 and as revised in References30 through 32. and as described in Section 2.3. The scope of components assumes the credited FLEX connection modifications areimplemented, and are limited to those required to support a single FLEX success path (i.e.,either "Primary" or "Back-up/Alternate").4. The "Primary" FLEX success path is to be specified. Selection of the "Back-up/Alternate"FLEX success path must be justified.5. Phase 3 coping strategies are included in the ESEP scope, whereas recoverystrategies are excluded.6. Structures, systems, and components excluded per the EPRI 3002000704 (Reference 2)guidance are:" Structures (e.g. Reactor Containment Building, Reactor Auxiliary Building, etc.)" Piping, cabling, conduit, HVAC, and their supports." Manual valves and rupture disks." Power-operated valves not required to change state as part of the FLEX mitigationstrategies.Page 13 of 46 Expedited Seismic Evaluation Process Report* Nuclear steam supply system components (e.g. reactor pressure vessel and internals,reactor coolant pumps and seals, etc.)7. For cases in which neither train was specified as a primary or back-up strategy, then only onetrain component (generally 'A' train) is included in the ESEL.3.1.1 ESEL DevelopmentThe ESEL was developed by reviewing the H.B. Robinson Steam Electric Plant OIP (Reference 3)and revisions in three subsequent six-month status reports to determine the major equipmentinvolved in the FLEX strategies. Further reviews of plant drawings (e.g., Process andInstrumentation Diagrams (P&IDs)(EC92103R0, Attachment Z03RO Mechanical Documents[Reference 49]), and Electrical One Line Diagrams (EC92103R0, Attachment Z05R0 ElectricalDocuments [Reference 50]) were performed to identify the boundaries of the flowpaths to be usedin the FLEX strategies and to identify specific components in the flowpaths needed to supportimplementation of the FLEX strategies. Boundaries were established at an electrical or mechanicalisolation device (e.g., isolation amplifier, valve, etc.) in branch circuits / branch lines off the definedstrategy electrical or fluid flowpath. P&IDs were the primary reference documents used to identifymechanical components and instrumentation. The flow paths used for FLEX strategies wereselected and specific components were identified using detailed equipment and instrumentdrawings, piping isometrics, electrical schematics and one-line drawings, system descriptions,design basis documents, etc., as necessary.3.1.2 Power Operated ValvesPage 3-3 of EPRI 3002000704 notes that power operated valves not required to change state areexcluded from the ESEL. Page 3-2 also notes that "functional failure modes of electrical andmechanical portions of the installed Phase 1 equipment should be considered (e.g. AFW trips)."To address this concern, the following guidance is applied in the H.B. Robinson Steam ElectricPlant ESEL for functional failure modes associated with power operated valves:" Power operated valves not required to change state as part of the FLEX mitigation strategieswere not included on the ESEL. The seismic event also causes the ELAP event; therefore, thevalves 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 strategiesduring Phase 1, and are re-energized and operated during subsequent Phase 2 and 3strategies, were not evaluated for spurious valve operation as the seismic event that causedthe ELAP has passed before the valves are re-powered.3.1.3 Pull BoxesPull boxes were deemed unnecessary to add to the ESEL as these components provide completelypassive locations for pulling or installing cables. No breaks or connections in the cabling areincluded in pull boxes. Pull boxes were considered part of conduit and cabling, which are excludedin accordance with EPRI 3002000704.Page 14 of 46 Expedited Seismic Evaluation Process Report3.1.4 Termination CabinetsTermination cabinets, including cabinets necessary for FLEX Phase 2 and Phase 3 connections,provide consolidated locations for permanently connecting multiple cables. The terminationcabinets and the internal connections provide a completely passive function; however, the cabinetsare included in the ESEL to ensure industry knowledge on panel/anchorage failure vulnerabilities isaddressed.3.1.5 Critical Instrumentation IndicatorsCritical indicators and recorders are typically physically located on panels/cabinets and are includedas separate components.3.1.6 Phase 2 and Phase 3 Piping ConnectionsItem 2 in Section 3.1 above notes that the scope of equipment in the ESEL includes "... FLEXconnections necessary to implement the H.B. Robinson Steam Electric Plant OIP as described inSection 2." Item 3 in Section 3.1 also notes that "The scope of components assumes the creditedFLEX connection modifications are implemented, and are limited to those required to support asingle FLEX success path (i.e., either "Primary" or "Back-up/Alternate")."Item 6 in Section 3.1 above goes on to explain that "Piping, cabling, conduit, HVAC, and theirsupports" are excluded from the ESEL scope in accordance with EPRI 3002000704.Therefore, piping and pipe supports associated with FLEX Phase 2 and Phase 3 connections areexcluded from the scope of the ESEP evaluation. However, any active valves in FLEX Phase 2and Phase 3 connection flow path are included in the ESEL.3.2 Justification for Use of Equipment That Is Not The Primary Means for FLEXImplementationIn accordance with EPRI 3002000704, the H.B. Robinson Steam Electric Plant used equipmentthat is the primary means of implementing FLEX strategy. The complete ESEL for the H.B.Robinson Steam Electric Plant is presented in Attachment A.Page 15 of 46 Expedited Seismic Evaluation Process Report4.0 Ground Motion Response Spectrum (GMRS)4.1 Plot of GMRS Submitted by the H.B. Robinson Steam Electric PlantFollowing completion of the seismic hazard re-evaluation as requested in Reference 1, the NRC 10CFR 50.54(f) letter, a screening process is needed to determine if an interim seismic risk evaluationlike the EPRI ESEP is required. The screening GMRS was determined with control point seismichazard re-evaluation. In accordance with the 50.54(f) letter and following the guidance in EPRIScreening, Prioritization, and Implementation Details (SPID) (Reference 15), Probabilistic SeismicHazard Analysis (PSHA) was performed using the 2012 CEUS Seismic Source Characterization forNuclear Facilities (Reference 20), a Regional Seismic Catalog Correction (Reference 61), andupdated EPRI Ground Motion Model (GMM) for the CEUS (Reference 21). Development of theH.B. Robinson Steam Electric Plant Ground Motion Response Spectra (GMRS) is documented inReferences 4 and 62. The GMRS and Uniform Hazard Response Spectra (UHRS) are tabulated inTable 4-1 and then compared in Figure 4-1 with the 5% damped horizontal SSE. Note thatadditional seismic hazard analysis and GMRS development is underway for H.B. Robinson SteamElectric Plant to support completion of the seismic probabilistic risk analysis. In the analysis, newlyacquired geophysical testing results are being used to update the site response analysis. Theresults of the screening evaluation discussed will not change as a result of the newly acquiredgeophysical testing. These new geophysical testing data allow for a more accurate representationof seismic hazard and seismic probabilistic risk assessment by eliminating a significant source ofuncertainty.Page 16 of 46 Expedited Seismic Evaluation Process ReportTable 4-1: GMRS and UHRS for the H.B. Robinson Steam Electric PlantFreq. (Hz) 10-4 UHRS (g) 105 UHRS (g) GMRS100 4.20E-01 9.17E-01 4.71E-0190 4.23E-01 9.31 E-O1 4-77E-0180 4.27E-01 9-48E-01 4.85E-0170 4.35E-01 9.73E-01 4.97E-0160 4-54E-01 1.02E+00 5.19E-0150 4.98E-01 1.11E+00 5.66E-0140 5.74E-01 1.25E+00 6.43E-0135 6.21 E-01 1.35E+00 6.95E-0130 6.63E-01 1.46E+00 7.50E-0125 7.23E-01 1.61E+00 8.21E-0120 7.92E-01 1.75E+00 8.97E-0115 8.09E-01 1.82E+00 9.27E-O112.5 8.35E-01 1.82E+00 9.36E-0110 8.52E-01 1.86E+00 9.55E-019 8.40E-01 1.84E+00 9.42E-018 8.58E-01 1.84E+00 9.49E-017 8.98E-01 1-92E+00 9.88E-016 8.87E-01 1.95E+00 9.99E-015 8.57E-01 1.87E+00 9.61E-014 8.40E-01 1.83E&#xf7;00 9.39E-013.5 7.71 E-01 1.76E+00 8.94E-013 6.79E-01 1.59E+00 8.04E-012.5 6.08E-01 1.38E+00 7.04E-O12 5.37E-01 1.30E+00 6.52E-011.5 3.97E-01 1.05E+00 5.20E-011.25 3.23E-01 8.58E-01 4.23E-011 2.26E-01 6.44E-01 3.13E-010.9 1.87E-01 5.52E-01 2.67E-010.8 1.56E-01 4.69E-01 2.26E-010.7 1.31E-01 3.95E-01 1.90E-010.6 1.10E-01 3.25E-01 1.57E-010.5 8.86E-02 2.51E-01 1.22E-010.4 7.09E-02 2.01E-01 9.79E-020.35 6.20E-02 1.76E-01 8.57E-020.3 5.32E-02 1.51 E-01 7.34E-020.25 4-43E-02 1.26E-01 6.12E-020.2 3.55E-02 1.00E-01 4.90E-020.15 2-66E-02 7.54E-02 3.67E-020.125 2.22E-02 6.28E-02 3.06E-020.1 1.77E-02 5.02E-02 2.45E-02Page 17 of 46 Expedited Seismic Evaluation Process ReportMean Soil UHRS and GMRS at Robinson2-52.-1E-5 UHRSL1.50.0.1 1 10 100Spectral frequency, HzFigure 4-1: Plot of 1 E-4 and 1 E-5 UHRS and GMRS at Control Point for the H.B. RobinsonSteam Electric Plant (5% Damped Response Spectra)Control point hazard curves were used to develop the UHRS and the GMRS. The methodologydescribed in SPID (Reference 15) was used to compute site-specific control point hazard curves.The selection of control point elevation is based on recommendations in Section 2.4.2 of the SPID(Reference 15). The control point elevation for the H.B. Robinson Steam Electric Plant is at El. 226feet based on information in Sections 2.5 and 2.7 of the Updated Final Safety Analysis Report(Reference 51).Page 18 of 46 Expedited Seismic Evaluation Process Report4.2 Comparison to SSEOriginal design of the H.B. Robinson Steam Electric Plant was based on the 0.2g HousnerSpectrum. Table 4-2a shows the spectral acceleration values as a function of frequency for the 5%damped horizontal SSE. As will be discussed in more detail in Section 5.2, original design in-structure response spectra was developed based on conservative time history. The Ground LevelResponse Spectrum that results from this time history is reported in Table 4-2b.A comparison of the Ground Level Response Spectrum, SSE, and GMRS is shown in Figure 4-2.As shown in Figure 4-2, in the 1 to 10 Hz frequency range of the response spectrum, the GMRSexceeds the SSE and the Ground Level Response Spectrum. The GMRS also exceeds the SSEand the Ground Level Response Spectrum at frequency values higher than 10 Hz.Page 19 of 46 Expedited Seismic Evaluation Process ReportTable 4-2a: Original SSE Based on 0.2g Housner Spectrum for the H.B. Robinson Steam ElectricPlant (5% Damping)FrequencySSE(Hz) (g)1.0 0.171.5 0.2302.0 0.2602.5 0.2903.0 0.33.5 0.3104.0 0.325.0 0.3056.0 0.2907.0 0.2658.0 0.2559.0 0.24010.0 0.2312.50 0.21015.0 0.220.0 0.225.0 0.230.0 0.233.0 0.235.0 0.2Page 20 of 46 Expedited Seismic Evaluation Process ReportTable 4-2b: Ground Level Response Spectrum Based on Time History for H.B. Robinson SteamElectric Plant (5% Damping)Frequency Ground Level(Hz) ResponseSpectrum (g)from TimeHistory1.0 0.3001.5 0.4552.0 0.4412.5 0.4173.0 0.4453.5 0.4684.0 0.4895.0 0.4556.0 0.4157.0 0.3808.0 0.3519.0 0.31610.0 0.28112.50 0.22115.0 0.23220.0 0.24625.0 0.25830.0 0.26733.0 0.27335.0 0.275Page 21 of 46 Expedited Seismic Evaluation Process Report1.200 --1.00 from :1m1Hi tory00.800 -~ -- -----o0.600 _0.40C,0.2000.0000.1 1.0 10.0 100.0Frequency (Hz)Figure 4-2: Comparison of GMRS, SSE and Ground Level Response Spectrum from Time HistoryPage 22 of 46 Expedited Seismic Evaluation Process Report5.0 Review Level Ground Motion (RLGM)5.1 Description of RLGM SelectedPlants for which the GMRS exceeds the SSE in the 1.0 to 10.0 Hz frequency range do not screenout of the ESEP and require further seismic evaluation. The further seismic evaluation is performedto a Review Level Ground Motion which consists of a response spectrum above the SSE level. TheRLGM is defined as a response spectrum reflecting an earthquake level that is above the plant'sdesign basis SSE. The RLGM can be computed using one of the following criteria as described inReference 2:1. The RLGM can be derived by linearly scaling the SSE by the maximum ratio of the horizontalGMRS to the 5% damped SSE, between the 1 and 10 Hz frequency range, but not to exceed aratio greater than 2 times the SSE. The in-structure seismic motions corresponding to theRLGM would be derived using existing SSE-based In-Structure Response Spectra (ISRS)scaled with the same factor.2. Alternatively, licensees who have developed appropriate structural/soil-structure interaction(SSI) models capable of calculating ISRS based on site GMRS/Uniform Hazard ResponseSpectrum (UHRS) input may opt to use these ISRS in lieu of scaled SSE ISRS. In this case,the GMRS would represent the RLGM. EPRI 1025287 and the American Society ofMechanical Engineers/American Nuclear Society (ASME/ANS) PRA Standard give guidanceon acceptable methods to compute both the GMRS and the associated ISRS. Section 4 ofReference 2 contains full description of this task.The RLGM for the H.B. Robinson Steam Electric Plant was developed in Reference 52 and inaccordance with the methodology and objectives in EPRI ESEP guidance Reference 2. The RLGMis the SSE multiplied by a factor of 2.0. Table 5-1 is the RLGM as a function of frequency andacceleration at 5% damping. As discussed under Sections 4.2 and 5.2, original design in-structureresponse spectra were developed based on a conservative time history. The Ground LevelResponse Spectrum that resulted from this time history is reported in Table 4-2b and Figure 5-2.For consistency between component screening and component evaluations, the Ground LevelResponse Spectrum was scaled by 2 to represent an effective RLGM for component screening.Therefore, both screening and evaluation of ESEL items were conservatively based on 2 x GroundLevel Response Spectrum (see Figure 6-1 for plot of 2 x Ground Level Response Spectrum)instead of 2 x SSE.Page 23 of 46 Expedited Seismic Evaluation Process ReportTable 5-1: RLGM for H.B. Robinson Steam Electric PlantFrequency SSE RLGM(Hz) (g) (g)1.0 0.17 0.341.5 0.230 0.4602.0 0.260 0.5202.5 0.290 0.583.0 0.3 0.603.5 0.310 0.624.0 0.32 0.645.0 0.305 0.616.0 0.290 0.587.0 0.265 0.538.0 0.255 0.519.0 0.240 0.4810.0 0.23 0.4612.50 0.210 0.4215.0 0.2 0.420.0 0.2 0.425.0 0.2 0.430.0 0.2 0.433.0 0.2 0.435.0 0.2 0.4The ratio of the GMRS to the SSE is summarized in Table 5-2. The maximum ratio of the GMRS toSSE is 4.635 and this occurs at frequency of approximately 15Hz. In the frequency range of 1 to10Hz, the maximum ratio of the GMRS to SSE is 4.152. As limited in EPRI 3002000704, the RLGMis determined multiplying the SSE by a factor of 2.0.Page 24 of 46 Expedited Seismic Evaluation Process ReportTable 5-2: Ratio of GMRS to SSEFrequency GMRS SSE GMRSISSE(Hz) (g) (g)1.0 0.313 0.17 1.8411.5 0.520 0.230 2.2612.0 0.652 0.260 2.5082.5 0.704 0.290 2.4283.0 0.804 0.3 2.6803.5 0.894 0.310 2.8844.0 0.939 0.32 2.9345.0 0.961 0.305 3.1516.0 0.999 0.290 3.4457.0 0.988 0.265 3.7288.0 0.949 0.255 3.7229.0 0.942 0.240 3.92510.0 0.955 0.23 4.15212.50 0.936 0.210 4.45715.0 0.927 0.2 4.63520.0 0.897 0.2 4.48525.0 0.821 0.2 4.10530.0 0.750 0.2 3.75033.0 0.717 0.2 3.58535.0 0.695 0.2 3.475Page 25 of 46 Expedited Seismic Evaluation Process Report1.2001.000S0o.80000.6000.4000.200L20.0000.1 1.0 10.0 100.0Frequency (Hz)Figure 5-1: Plot of 5% Damping 2xSSE, 2 x Ground Level Response Spectrum, and GMRS5.2 Method to Estimate ISRSThe seismic demand of the ESEL items/element mounted rigidly to the structure can be specified interms of the In-Structure Response Spectra (ISRS). For use in the ESEP, the in-structure seismicdemand for an element listed in the ESEL is defined by the ISRS scaled by the same factor used toobtain the RLGM from the SSE. The guidance under Section 4 of Reference 7 recommendsbroadening the peaks of the ISRS to account for the uncertainty in the civil structure frequencycalculation. The extent of broadening is suggested to be at least 15 percent of the frequencyapproaching and proceeding spectral peaks but can be increased beyond the minimumrecommendation based on the level of uncertainty associated with the structural model.The original design basis ISRS for the H.B. Robinson Steam Electric Plant were generated in 1970by Westinghouse Electric Corporation using mathematical building models developed by EbascoServices, Inc. The original ISRS or floor spectra generated by Westinghouse was limited in scopeand only considered the 0.20g design basis earthquake at damping ratio of 0.005 (0.5 percent).These ISRS include conservatisms that result from conservative selection of the time history andexcessive bounding of design spectra. Figure 4-2 shows plot of: (1) Ground Level ResponseSpectrum; (2) GMRS; and (3) SSE.Additional ISRS for other damping values were generated. The task of generating the additionalfloor response spectra was complicated by lack of availability of time history data from the originalWestinghouse analysis. Consequently, synthetic ground motion time history that generates ISRSPage 26 of 46 Expedited Seismic Evaluation Process Reportcomparable to the original Westinghouse floor spectra was used. The ISRS were generated byinputting the synthetic ground motion through the original Ebasco structural models. Scale factorsas a function of frequency were developed by comparing the spectra at the desired damping ratioagainst the 0.50 percent damping spectra. The factors were then used to scale the originalWestinghouse 0.50 percent damped spectra to the desired damping ratio. The reconstituted ISRSat the various damping ratios have been incorporated into the H.B. Robinson Steam Electric Plant'sdesign basis ISRS documentation in Reference 18.The ISRS from Reference 18 were peak broadened in accordance with guidance in RegulatoryGuide 1.122 (Reference 19). Since the ISRS in Reference 18 are already broadened, these spectraare scaled by a factor of 2.0 for ESEP.In summary, in-structure response spectra developed with the conservative Ground LevelResponse Spectrum were scaled by a factor of 2 for use in ESEP. Figure 5-1 shows plot of the 2 xSSE (RLGM), 2 x Ground Level Response Spectrum, and GMRS.Page 27 of 46 Expedited Seismic Evaluation Process Report6.0 Seismic Margin Evaluation ApproachIt is necessary to demonstrate that ESEL items have sufficient seismic capacity to meet or exceedthe demand characterized by the RLGM and the corresponding scaled in-structure responsespectra. The seismic capacity is characterized as the peak ground acceleration (PGA) for whichthere is a high confidence of a low probability of failure (HCLPF). The PGA is associated with aspecific spectral shape, in this case the 5%-damped 2 x Ground Level Response Spectrum shape.The HCLPF capacity must be equal to or greater than the RLGM PGA. The criteria for seismiccapacity determination are given in Section 5 of EPRI 3002000704.There are two basic approaches for developing HCLPF capacities:1. Deterministic approach using the conservative deterministic failure margin (CDFM)methodology of EPRI NP-6041, A Methodology for Assessment of Nuclear Power PlantSeismic Margin (Revision 1) (Reference 7).2. Probabilistic approach using the fragility analysis methodology of EPRI TR-103959,Methodology for Developing Seismic Fragilities (Reference 8).6.1 Summary of Methodologies UsedThe H. B. Robinson Steam Electric Plant completed a seismic margin assessment (SMA) in 1993.The SMA is documented in Reference 9 and consisted of screening, walkdowns by SRT, andHCLPF anchorage calculations. The screening and walkdowns used the screening tables fromChapter 2 of EPRI NP-6041 (Reference 7) for peak spectral acceleration less than 0.8g. Thewalkdowns were conducted by engineers trained in EPRI NP 6041 (the engineers attended theEPRI SMA Add-On course in addition to the SQUG Walkdown Screening and Seismic EvaluationTraining Course), and were documented on Screening Evaluation Work Sheets from EPRI NP-6041. Anchorage capacity calculations used the CDFM criteria from EPRI NP-6041. Seismicdemand was the IPEEE Review Level Earthquake (RLE) for SMA (mean NUREG/CR-0098[Reference 11] ground response spectrum anchored to 0.3g PGA).Figure 6-1 shows the mean NUREG/CR-0098 ground response spectrum used as the IPEEE RLEcompared to the 2 x Ground Level Response Spectrum. The figure shows that the ESEP inputmotion enveloped the IPEEE RLE at all frequencies except between 10 Hz and 15 Hz where theIPEEE RLE slightly exceed the ESEP input motion. The frequency of interest for ESEL items isbetween 1 Hz and 10Hz.The ESEP methodology included screening and extensive walkdown by the Seismic Review Team(SRT), and HCLPF calculations to evaluate structural capacity of the ESEL items against theRLGM. Function evaluation of relays was also performed. The walkdowns were documented onScreening Evaluation Worksheets (SEWS) from EPRI NP-6041. Based on outcome of the seismicwalkdown and documentation in SEWS, six (6) HCLPF calculations were performed to envelopethe thirteen (13) ESEL items identified during the walkdowns.Page 28 of 46 Expedited Seismic Evaluation Process Report1.2001.000-2 X Ground LevelTim History-.SS-IPEEE RILE-Ground Level SpaHistory,-&#xfd;2 XSSE1(RLGMSfromn04..CO)0.8000.6000.4000.2000.0000.11.010.0100.0Frequency (Hz)Figure 6-1.Comparison of the H.B. Robinson Steam Electric Plant IPEEE RLE, ESEPRLGM, SSE, Ground Level (El. 226ft) Spectrum from Time History, and 2 xGround Level (El. 226ft) Spectrum from Time History6.2 HCLPF Screening ProcessThe HCLPF screening and calculations were based on 2 x Ground Level Response Spectrum peakground acceleration. Screening tables in EPRI NP-6041 (Reference 7) are based on peak spectralacceleration of < 0.8g, 0.8g to 1.2g, and > 1.2g. Since 2 x Ground Level Response Spectrum peakground acceleration is 0.978g, screening of ESEL items was based on the 0.8g to 1.2g rangecriteria. The screening guidelines were supplemented by Appendix A of EPRI NP-6041 SL whichprovides the basis for the seismic capacity screening guidelines.Anchorage capacity calculations were based on 2 x Ground Level Response Spectrum. Equipmentfor which the screening caveats were met and for which the anchorage capacity exceeded 2 xGround Level Response Spectrum seismic demand were screened out from ESEP seismiccapacity determination.Page 29 of 46 Expedited Seismic Evaluation Process Report6.3 Seismic Walkdown Approach6.3.1 Walkdown ApproachWalkdowns were performed in accordance with the criteria provided in Section 5 of EPRI3002000704 (Reference 2), which refers to EPRI NP-6041 (Reference 7) for the Seismic MarginAssessment process. Pages 2-26 through 2-30 of EPRI NP-6041 describe the seismic walkdowncriteria, including the following key criteria."The SRT [Seismic Review Team] should "walk by" 100% of all components which are reasonablyaccessible and in non-radioactive or low radioactive environments. Seismic capability assessmentof components which are inaccessible, in high-radioactive environments, or possibly withincontaminated containment, will have to rely more on alternate means such as photographicinspection, more reliance on seismic reanalysis, and possibly, smaller inspection teams and morehurried 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 orpanels for detailed inspection of all components. This walkdown is not intended to be a QA or QCreview or a review of the adequacy of the component at the SSE levelIf the SRT has a reasonable basis for assuming that the group of components are similar and aresimilarly 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 capabilitypreparatory work (Step 3) by reference to drawings, calculations or specifications. The onecomponent or each type which is selected should be thoroughly inspected which probably doesmean de-energizing and opening cabinets or panels for this very limited sample. Generally, aspare representative component can be found so as to enable the inspection to be performed whilethe plant is in operation. At least for the one component of each type which is selected, anchorageshould be thoroughly inspected.The walkdown procedure should be performed in an ad hoc manner. For each class ofcomponents the SRT should look closely at the first items and compare the field configurations withthe construction drawings and/or specifications. If a one-to-one correspondence is found, thensubsequent items do not have to be inspected in as great a detail. Ultimately the walkdownbecomes a "walk by" of the component class as the SRT becomes confident that the constructionpattern 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 inparallel. If serious exceptions to the drawings or questionable construction practices are foundthen the system or component class must be inspected in closer detail until the systematicdeficiency is defined.The 100% "walk by" is to look for outliers, lack of similarity, anchorage which is different from thatshown on drawings or prescribed in criteria for that component, potential SI [Seismic Interaction1]problems, situations that are at odds with the team members' past experience, and any other areasof serious seismic concern. If any such concerns surface, then the limited sample size of onecomponent of each type for thorough inspection will have to be increased. The increase in samplesize 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'EPRI 3002000704 page 5-4 limits the ESEP seismic interaction reviews to "nearby block walls" and "pipingattached to tanks" which are reviewed "to address the possibility of failures due to differential displacements." Otherpotential seismic interaction evaluations are "deferred to the full seismic risk evaluations performed in accordancewith EPRI 1025287'Page 30 of 46 Expedited Seismic Evaluation Process Reportones who are responsible for the seismic adequacy of all elements which they screen from themargin review. Appendix D gives guidance for sampling selection."As part of the ESEP, demonstration that the components listed in the ESEL have a HCLPFcapacity that exceeds the effective RLGM (2 x Ground Level Response Spectrum) verifiesadequate seismic ruggedness. Section 5 of EPRI ESEP guidance specifies that the methodology inEPRI NP-6041 SL may be used for the development of the HCLPF capacity. The major steps inReference 7 include pre-screening, walkdowns, and the CDFM HCLPF calculations.In order to ensure efficiency while performing the walkdowns and during seismic capacityevaluations, each of the items listed in the ESEL were subjected to pre-screening. The initial pre-screening effort consisted of data collection in the form of drawings, calculations, specifications,and vendor documents for each item in the ESEL. After identification of documentation for aspecific item, the pre-screening process followed the general seismic capacity screening guidelinespresented in Reference 7 for civil structures, equipment, and subsystems to be consideredscreened out from further review. The caveats and footnoted exceptions and restrictions listed arefollowed.For the purpose of completing the ESEP for the H. B. Robinson Steam Electric Plant, only Table 2-4 of Reference 7 is relevant for applying seismic screening criteria for plant equipment listed in theESEL. In addition to using the screening criteria in Reference 7 during plant walkdown, the SRTalso exercised their collective experience and judgment while using the criteria for specificcomponent. The screening criteria can be used for equipment that is approximately 40ft abovegrade or lower. EPRI Report No 1019200 (Reference 23) provides guidance on screening criteriafor equipment that is greater than 40ft above grade. Screening criteria in Reference 7 do notinclude considerations for anchorage. Therefore, structural integrity of anchorage was evaluatedseparately. Some simple cases were documented on the SEWS form.Plant walkdowns were performed for items in the ESEL using guidance in Reference 7. Informationextracted from existing documentation such as equipment location, seismic input elevation,relevant drawing details, and previous seismic capacity calculations were recorded on the ESEPSEWS and used during the walkdowns. In accordance with the ESEP guidance, the SEWS thatwere used in the ESEP walkdowns were consistent with content and format of the SEWSpresented in Appendix F of EPRI NP-6041 SL.A major part of the ESEP walkdowns was the investigation of equipment anchorages. Therefore,cabinets with anchorages located internally were opened. Furthermore, the ESEP guidance statesthat components that are anchored to sub-structural elements that may not have the same capacityas the main structural system (e.g. block walls, frames, stanchions etc.) should also be reviewed.Nearby block walls were identified and evaluated as necessary. Piping attached to tanks were alsoreviewed. Other potential seismic interaction evaluations were deferred to a full Seismic RiskEvaluation (SRE) as discussed in the SPID References 14 and 15, and were not addressed in theESEP walkdowns.Walkdown assessment for the H.B. Robinson Steam Electric Plant ESEL items were completed bythe SRT between August 2013 and February 2014. Some of the components were previouslywalked down during the IPEEE, USI A-46, or NTTF 2.3: Seismic and relevant information such asthe equipment location, seismic input elevation, drawing details and previous seismic calculationswere recorded on the ESEP SEWS. Previous walkdowns were credited since they were performedby qualified Seismic Review Team. A walk-by of these components was performed anddocumented. The objective of the walk-by is to confirm and verify that the components and theiranchorage have not degraded since the previous walkdown.Items included in the ESEL that have not been previously walked down and evaluated, wereautomatically included for a detailed walkdown.Page 31 of 46 Expedited Seismic Evaluation Process ReportThe SRT was comprised of at least two SQUG trained engineers and often included two additionalstructural engineers (Reference 57). The results of the walkdowns were documented on the SEWSfor each item. The completed SEWS and pictures taken during the walkdowns for the ESEL aredocumented in Reference 55. Follow-up inspections and walkdowns were completed whereadditional information was necessary.6.3.2 Application of Previous Walkdown InformationPrevious seismic walkdowns from IPEEE and USI A-46 were used to support the ESEP seismicevaluations. Some of the components and items on the ESEL were included in the NTTF 2.3seismic walkdowns (Reference 17). Those walkdowns were well documented and recent enoughthat they did not need to be repeated for the ESEP.Several ESEL items were previously walked down during the H.B. Robinson Steam Electric PlantSeismic IPEEE program. Those walkdown results were reviewed and the following steps weretaken to confirm that the previous walkdown conclusions remained valid.* A walk by was performed to confirm that the equipment material condition and configurationis consistent with the walkdown conclusions and that no new significant interactions relatedto block walls or piping attached to tanks exist." If the ESEL item was screened out based on the previous walkdown, that screeningevaluation was reviewed and reconfirmed for the ESEP.Page 32 of 46 Expedited Seismic Evaluation Process Report6.3.3 Significant Walkdown FindingsConsistent with guidance from NP-6041, no significant outliers or anchorage concerns (exceptMCC-A) were identified during the H.B. Robinson Steam Electric Plant seismic walkdowns. Thefollowing findings were noted during the walkdowns.* Nearby block walls were identified in the proximity of ESEL item. These block walls wereassessed for their structural adequacy to withstand the seismic loads resulting from theRLGM. There is no case where the block wall represented the HCLPF failure mode for anESEL item.* Piping attached to tanks were reviewed and evaluated for their structural integrity towithstand seismic-induced loads from RLGM.* Cabinets with anchorage located internally were opened and evaluated against RLGM." Thirteen (13) components were identified by the SRT during the plant walkdowns and six(6) HCLPF calculations were performed to envelope the thirteen components identified.6.4 HCLPF Calculation ProcessESEL items not included in the previous IPEEE evaluations at H.B. Robinson Steam Electric Plantwere evaluated using the criteria in EPRI NP-6041. Those evaluations included the followingsteps:* Performing seismic capability walkdowns for equipment not included in previous seismicwalkdowns (SQUG, IPEEE, or NTTF 2.3) to evaluate the equipment installed plantconditions. Results of the walkdowns which are documented in the ESEP SEWS identifiedthirteen (13) components that require HCLPF calculation.* Performing screening evaluations using the screening tables in EPRI NP-6041 as describedin Section 6.2 and* Performing HCLPF calculations considering various failure modes that include structuralfailure modes (e.g. anchorage, load path etc.) and functional failure modes.Items based on similarity of model, function and anchorage were grouped together. Based on EPRINP-6041-SL rule of similarity, a bounding anchorage evaluation was performed for equipmentgrouped together. The calculations evaluate the demand and capacity of the equipment anchorageand derived a HCLPF capacity from the results of the anchorage evaluation. The functional failuremode(s) are also evaluated.Equipment that were identified as requiring a HCLPF capacity calculation in Reference 55 wereevaluated using the CDFM methodology as outlined in EPRI NP-6041-SL. The HCLPF calculationsare documented in Reference 10 and References 25 through 29. Thirteen components wereidentified by the SRT during walkdown and six HCLPF calculations were completed to envelope allthe components which include I&C and Hagan rack; Pressure Vessel; MCC; Battery Charger; andAuxiliary DC Panel.6.5 Functional Evaluations of RelaysBased on review of ESEL and associated single line diagrams, two relays (Under-Voltage AlarmRelay 27/MCC-A and Under-Voltage Alarm Relay 27/MCC-B) were identified. However, thesePage 33 of 46 Expedited Seismic Evaluation Process Reportrelays do not have lockout or seal-in mechanism (Reference 59) and are not required during FLEXimplementation. 27/MCC-A and 27/MCC-B are not designed to operate during and following DBEand BDBEE. Therefore, these relays were not included on the ESEL list. Extensive review of thesingle line diagrams did not identify any other relay or contactor that will be of concern.6.6 Tabulated ESEL HCLPF Values (Including Key Failure Modes)Tabulated ESEL HCLPF values are provided in Table 6-1. The following notes apply to theinformation in the table:* For items screened out using NP 6041 screening tables, the screening level can beprovided as >RLGM and the failure mode can be listed as "Screened", (unless thecontrolling HCLPF value is governed by anchorage).* For items where anchorage controls the HCLPF value, the HCLPF value is listed in thetable and the failure mode is noted as "anchorage."Six HCLPF calculations were performed for items listed in the ESEL. Items that are based onsimilarity of equipment model, function, and anchorage are grouped together. Based on EPRI NP-6041 SL rule of similarity, some items were grouped together and a bounding anchorage evaluationwas performed. The six HCLPF capacity evaluations are documented in Reference 10 andReferences 25 through 29. Each capacity calculation evaluates the demand and capacity of theequipment anchorage and derives a HCLPF capacity from the results of the anchorage evaluation.The functional failure modes for each ESEL item were identified and documented in the calculation.The functional and anchorage HCLPF capacity of items identified by the SRT for a seismic capacityevaluation is presented in Table 6-1.Page 34 of 46 Expedited Seismic Evaluation Process ReportTable 6-1: Functional and Anchorage HCLPF Capacity ResultsFunctional Anchorage/StructuralEquipment Group Equipment HCLPF Achora ctyCapacityHCLPF CapacityCapacityInstrumentation and ControlPanels and Ck Main Control Board > 0.40g 0.414gPanels and RackRack -4Rack -11Hagan Racks > 0.40g 0.445gRack -12Rack -130.541gPressure Vessels Boron Injection Tank > 0.40gBattery Charger -ABattery Charger -AlBattery Chargers > 0.40g 0.755gBattery Charger -BBattery Charger -B1> 0.40gMotor Control Centers MCC-A 0.250g> 0.40gMCC-B 0.406g> 0.40gAuxiliary DC Panel GD AUX-PNL-GD 0.596gPage 35 of 46 Expedited Seismic Evaluation Process Report7.0 Inaccessible Items7.1 Identification of ESEL items inaccessible for walkdownsAll ESEL items were accessible with the exception of TE-423. This temperature element is ruggedand due to installation internal to the pipe, it is also protected from seismic interaction. Anevaluation was performed based on available information and this item was determined to beacceptable by the SRT with no visual examination.7.2 Planned Walkdown / Evaluation Schedule / Close OutNo ESEL item requires future walkdown.Page 36 of 46 Expedited Seismic Evaluation Process Report8.0 ESEP Conclusions and Results8.1 Supporting InformationThe H.B. Robinson Steam Electric Plant has performed the ESEP as an interim action in responseto Reference 1, the NRC's 10 CFR 50.54(f) letter. It was performed using the methodologies inReference 2, the NRC endorsed guidance in EPRI 3002000704.The ESEP provides an important demonstration of seismic margin and expedites plant safetyenhancements through evaluations and potential near-term modifications of plant equipment thatcan be relied upon to protect the reactor core following beyond design basis seismic events.The ESEP is part of the overall H.B. Robinson Steam Electric Plant response to the NRC's 50.54(f)letter. On March 12, 2014, NEI submitted to the NRC results of Reference 12, a study of seismiccore damage risk estimates based on updated seismic hazard information as it applies to operatingnuclear 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 thefleet of U.S. plants" based on the re-evaluated seismic hazards. As such, the "current seismicdesign of operating reactors continues to provide a safety margin to withstand potentialearthquakes exceeding the seismic design basis."The NRC's May 9, 2014 NTTF 2.1 Screening and Prioritization letter (Reference 14) concluded thatthe "fleetwide seismic risk estimates are consistent with the approach and results used in the GI-199 safety/risk assessment." The letter also stated that "As a result, the staff has confirmed thatthe conclusions reached in GI-199 safety/risk assessment remain valid and that the plants cancontinue to operate while additional evaluations are conducted."An assessment of the change in seismic risk for H.B. Robinson Steam Electric Plant was includedin the fleet risk evaluation submitted in the March 12, 2014 NEI letter therefore, the conclusions inthe NRC's May 9 letter also apply to H.B. Robinson Steam Electric Plant.In addition, Reference 12, the March 12, 2014 NEI letter, provided an attached "Perspectives onthe Seismic Capacity of Operating Plants," which (1) assessed a number of qualitative reasons whythe design of SSCs inherently contain margin beyond their design level, (2) discussed industrialseismic experience databases of performance of industry facility components similar to nuclearSSCs, 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 hasbeen borne out for those plants that have actually experienced significant earthquakes. Theseismic design process has inherent (and intentional) conservatisms which result in significantseismic margins within structures, systems and components (SSCs). These conservatisms arereflected in several key aspects of the seismic design process, including:* Safety factors applied in design calculations* Damping values used in dynamic analysis of SSCs* Bounding synthetic time histories for in-structure response spectra calculations* Broadening criteria for in-structure response spectra" Response spectra enveloping criteria typically used in SSC analysis and testing applications" Response spectra based frequency domain analysis rather than explicit time history basedtime domain analysis* Bounding requirements in codes and standards* Use of minimum strength requirements of structural components (concrete and steel)* Bounding testing requirements, andPage 37 of 46 Expedited Seismic Evaluation Process Report0 Ductile behavior of the primary materials (that is, not crediting the additional capacity ofmaterials 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 theirfunctions 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 todemonstrate seismic margin through a review of a subset of the plant equipment that can be reliedupon to protect the reactor core following beyond design basis seismic events. In order tocomplete the ESEP in an expedited amount of time, the RLGM used for the ESEP evaluation is ascaled version of the plant's SSE rather than the actual GMRS. To more fully characterize the riskimpacts of the seismic ground motion represented by the GMRS on a plant specific basis, a moredetailed seismic risk assessment (SPRA or risk-based SMA) is to be performed in accordance withEPRI 1025287 (Reference 15). As identified in Reference 4, the H. B. Robinson Steam ElectricPlant Seismic Hazard and GMRS submittal, the H.B. Robinson Steam Electric Plant screens in fora seismic risk evaluation. The complete seismic risk evaluation will more completely characterizethe probabilistic seismic ground motion input into the plant, the plant response to that probabilisticseismic ground motion input, and the resulting plant risk characterization. H.B. Robinson SteamElectric Plant will complete that evaluation in accordance with the schedule identified in Reference13, NEI's letter dated April 9, 2013 and endorsed by the NRC in Reference 16, their May 7, 2013letter.8.2 Identification of Planned ModificationsThere are no planned future modifications for ESEP. The ESEP identified MCC-A as having aHCLPF capacity below the RLGM and not meeting the requirements of EPRI ESEP and NTTFRecommendation 2.1: Seismic. MCC-A has since been modified in accordance with EPRI3002000704 to increase its seismic capacity to the RLGM. This was achieved by bracing thecabinet at the top. This modification eliminated flexible modes and resulted in reduced tensile loadapplied to the concrete expansion anchors. The HCLPF capacity of MCC-A is now greater than0.4g.The ESEP determined that the HCLPF capacity of MCC-B was slightly above the RLGM and meetsthe requirements of the EPRI ESEP such that no modification was required. However, amodification similar to that discussed above for MCC-A was implemented in order to increase thecapacity of MCC-B anchorage and eliminate potential inertial forces at the top entry cable tray andconduit.Seismic margin above 2 x SSE was also added to a group of instrument racks (Hagan Racks) byvalidating the bolting integrity of the top braces (a relatively minor scope of work). The HCLPFcapacity of the Main Control Board is higher than the RLGM and meets the requirements of theEPRI ESEP. However, greater seismic capacity can be demonstrated by additional inspection ofplug welds that form part of the anchorage. The additional inspection should confirm plug weldthickness and quality. Table 6-1 shows the capacities of the thirteen ESEL items that requiredHCLPF calculation. No additional modifications are planned for the H.B. Robinson Steam ElectricPlant related to ESEP.Page 38 of 46 Expedited Seismic Evaluation Process Report8.3 Modification Implementation ScheduleThe only ESEL item that required modification based on the seismic walkdown and HCLPFcapacity calculation was MCC-A. The modification has been developed and implemented asdiscussed in Section 8.2. The anchorage system for MCC-B is slightly different from that of MCC-Aand has higher structural capacity. The HCLPF capacity of MCC-B slightly exceeds RLGMdemand. However, similar modification developed for MCC-A was also implemented on MCC-B.Although, not considered a modification, the Hagan Rack cabinets bolts were tightened to improvestructural capacity.8.4 Summary of Planned ActionsThe H.B. Robinson Steam Electric Plant has no follow-up action or planned modification to supportthe ESEP. All of the items identified in the ESEL currently have a HCLPF capacity at or above theRLGM and do not require further evaluation. The ESEL has been updated to consider newequipment that account for the changes in the FLEX strategy. The new FLEX strategy wassubjected to critical path analysis and those items that fall under the ESEP guidelines have beenadded to the ESEL.Page 39 of 46 Expedited Seismic Evaluation Process Report9.0 References1) 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 Regulations50.54(f) Regarding Recommendations 2.1, 2.3 and 9.3 of the Near-Term Task ForceReview of Insights from the Fukushima Dai-lchi Accident," March 12, 2012.2) Seismic Evaluation Guidance: Augmented Approach for the Resolution ofFukushima Near-Term Task Force Recommendation 2.1 -Seismic. EPRI, PaloAlto, CA: May 2013. 3002000704.3) Updated H.B. Robinson Steam Electric Plant Overall Integrated Plan (OIP) inResponse to the March 12, 2012, Commission Order EA-12-049, August 2014.4) H.B. Robinson Steam Electric Plant Seismic Hazard and GMRS submittal, datedMarch 31, 2014.5) Nuclear Regulatory Commission, NUREG-1407, Procedural and Submittal Guidancefor the Individual Plant Examination of External Events (IPEEE) for Severe AccidentVulnerabilities, June 19916) Nuclear Regulatory Commission, Generic Letter No. 88-20 Supplement 4, IndividualPlant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities -10CFR 50.54(f), June 19917) A Methodology for Assessment of Nuclear Power Plant Seismic Margin, Rev. 1,August 1991, Electric Power Research Institute, Palo Alto, CA. EPRI NP 60418) Methodology for Developing Seismic Fragilities, August 1991, EPRI, Palo Alto, CA.1994, TR-1039599) Appendix A to The H.B. Robinson Steam Electric Plant Unit No. 2 Individual PlantExamination for External Event Submittal: Seismic IPEEE10) Calculation RNP-13-05-600-005,"High Confidence of Low Probability of Failure(HCLPF) for the H.B. Robinson Steam Electric Plant Motor Control Center A and B(MCC-A and MCC-B)}11) Nuclear Regulatory Commission, NUREG/CR-0098, Development of Criteria forSeismic Review of Selected Nuclear Power Plants, published May 197812) Nuclear Energy Institute (NEI), A. Pietrangelo, Letter to D. Skeen of the USNRC,"Seismic Core Damage Risk Estimates Using the Updated Seismic Hazards for theOperating Nuclear Plants in the Central and Eastern United States", March 12, 201413) Nuclear Energy Institute (NEI), A. Pietrangelo, Letter to D. Skeen of the USNRC,"Proposed Path Forward for NTTF Recommendation 2.1: Seismic Reevaluations",April 9, 201314) NRC (E Leeds) Letter to All Power Reactor Licensees et al., "Screening andPrioritization Results Regarding Information Pursuant to Title 10 of the Code ofFederal Regulations 50.54(F) Regarding Seismic Hazard Re-Evaluations forRecommendation 2.1 of the Near-Term Task Force Review of Insights From theFukushima Dai-lchi Accident," May 9, 2014.15) Seismic Evaluation Guidance: Screening, Prioritization and Implementation Details(SPID) for the Resolution of Fukushima Near-Term Task Force Recommendation2.1: Seismic. EPRI, Palo Alto, CA: February 2013. 1025287.16) NRC (E Leeds) Letter to NEI (J Pollock), "Electric Power Research Institute FinalDraft Report xxxxx, "Seismic Evaluation Guidance: Augmented Approach for theResolution of Fukushima Near-Term Task Force Recommendation 2.1: Seismic," asan Acceptable Alternative to the March 12, 2012, Information Request for SeismicReevaluations," May 7, 201317) H.B. Robinson Steam Electric Plant NTTF 2.3 Seismic Walkdown Submittal datedFebruary 27, 2014.Page 40 of 46 Expedited Seismic Evaluation Process Report18) Carolina Power and Light Company (CP&L), Specification No CPL-HBR2-C-008,"Specification for Floor Response Spectra", Revision 1, 1991.19) United States Nuclear Regulatory Commission, Regulatory Guide1.122,"Development of Floor Design Response Spectra for Seismic Design of Floor-Supported Equipment or Components", Revision 1, February 1978.20) United States Nuclear Regulatory Commission NUREG-2115, Department of EnergyOffice of Nuclear Energy (DOE/NE)-0140, EPRI 1021097,"Central and EasternUnited States Seismic Source Characterization for Nuclear Facilities", 6 Volumes,2012.21) Electric Power Research Institute (EPRI), Final Report No 3002000717,"EPRI(2004, 2006) Ground Motion Model (GMM) Review Project", June 2013.22) URS Energy and Construction Calculation RNP-13-05-600-001,"Review LevelGround Motion (RLGM) and In-Structure Response Spectra (ISRS) for H.B.Robinson Steam Electric Plant Unit 2, Revision 0, July 2013.23) Electric Power Research Institute (EPRI) Final Report No 1019200,"Seismic FragilityApplications Guide Update", 2009.24) EC 92103,"Fukushima NTTF Recommendation 2.1: Seismic Reevaluation"25) Calculation RNP-13-05-600-006,"High Confidence of Low Probability of Failure(HCLPF) for the H.B. Robinson Steam Electric Plant Battery Chargers"26) Calculation RNP-13-05-600-004,"High Confidence of Low Probability of Failure(HCLPF) for the H.B. Robinson Steam Electric Plant Boron Injection Tank"27) Calculation RNP-13-05-600-003,"High Confidence of Low Probability of Failure(HCLPF) for the H.B. Robinson Steam Electric Plant East Hagan Rack"28) Calculation RNP-13-05-600-007,"High Confidence of Low Probability of Failure(HCLPF) for the H.B. Robinson Steam Electric Plant Auxiliary DC Panel GD".29) Calculation RNP-13-05-600-002,"High Confidence of Low Probability of Failure(HCLPF) for the H.B. Robinson Steam Electric Plant Main Control Board".30) RNP/RA-13-0087, First Six Month Status Report (Order EA-12-049) H. B. RobinsonSteam Electric Plant (RNP), Unit 2.31) RNP/RA-14-0008, Second Six Month Status Report (Order EA-12-049) H. B.Robinson Steam Electric Plant (RNP), Unit 2.32) RNP/RA-14-0083, Third Six Month Status Report (Order EA-12-049) H. B. RobinsonSteam Electric Plant (RNP), Unit 2.33) Engineering Change (EC) 88926, FLEX Strategies and Implementation Plan, Rev. 3.34) NEI 12-06, Diverse and Flexible Coping Strategies (FLEX) ImplementationGuideline, Revision 0.35) EOP-ECA-0.0, Loss Of All AC Power, Revision 0.36) WCAP-1 760-P, Revision 1, Reactor Coolant System Response to Extended Loss ofAC Power Event for Westinghouse, Combustion Engineering, and Babcock &Wilcox NSSS Designs for Phase Boration, August 2012.37) PA-PSC-0965, PWROG Core Cooling Position Paper.38) EDMG-004, Steam Generators.39) Calculation RNP-M/MECH-1712, Appendix R Mechanical Basis, Section 3.27,Cooldown Using MSIV Bypass Lines.40) EC 90617, Pre-Staged Diesel Generator Design To Power 125VDC -A Train and BTrain Battery Chargers For Fukushima Support (NTFF 4.2 -FLEX).41) FSG-10, Passive RCS Injection Isolation.42) EC 95216, NTTF 2.1 Interim Action RCS Injection.43) EC 90622, Standard Piping Connections For NTTF 4.2 (FLEX).44) EC 94745, Boric Acid and RCS Make Up Connections To The Safety InjectionSystem NTTF 4.2 -Flexible Coping StrategiesPage 41 of 46 Expedited Seismic Evaluation Process Report45) Calculation RNP-M/MECH-1877, RNP Extended Loss of AC (ELAP) PowerContainment Response46) FSG-12, Alternate Containment Cooling47) EC 90623, New Pipe Tee And Standard Connection For NTTF 4.2 (FLEX)48) EC 95266, Isolation Valves And Connection For AFW -FUKUSHIMA-Admin Rev49) EC 92103R0, Attachment Z03RO Mechanical Documents50) EC 92103R0, Attachment Z05RO Electrical Documents51) UFSAR, Section 02, Site Characteristics"52) EC 92103R0, Attachment Z06RO53) EC 92103, Attachment Z18R054) EC 92103, Attachment Z09R055) EC 92103, Attachment Z1ORO56) EC 92103, Attachment Z01 RO57) EC 92103, Attachment Z16RO58) NRC Letter from NRC to Duke Energy and South Carolina Electric and GasCompany, Request for Additional Information Associated with Near-Term ForceRecommendation 2.1, Seismic Re-Evaluations Related to Southeastern CatalogChanges (TAC NOS MF3724, MF3736, MF 3738, and MF 3831), dated October 23,2014, ADAM Accession No ML14268A516.59) H.B. Robinson Steam Electric Plant Control Wiring Diagram (CWD) B-190628, SH00955 and 00956.60) EC 92501, Attachment Z09Rl,"Additional Seismic Interim Actions Studies for theH.B. Robinson Steam Electric Plant".61) Letter Dated August 28, 2014 from EPRI to NRC Review of EPRI 1021097Earthquake Catalog for RIS Earthquakes in the Southeastern U.S. and Earthquakesin South Carolina Near Time of the 1886 Charleston Earthquake Sequence.62) Response to Request for Additional Information Associated with Near-Term TaskForce Recommendation 2.1, Seismic Re-Evaluations Related to SoutheasternCatalog Changes (TAC NOS. MF3724, MF3736, MF3737, MF3738, and MF3831),November 12, 2014Page 42 of 46 Expedited Seismic Evaluation Process ReportAttachment A -H.B. Robinson Steam Electric Plant ESELPage 43 of 46


Expedited Seismic Evaluation Process ReportAttachment B -H.B. Robinson Steam Electric Plant Unit 2 RCS Cooling StrategiesPage 44 of 46 MNP72" MS HEADERMain Feed PumpsBLAplanREDStraGREStraBLUCapa-w'*-Lake Robinson______ IAJFW-TNK4/ -'IDischarge/ ICanal na/ I -3 E AFW-TNK 9LEGEND I l I Portable ...-.. 20k gli,CK -Installed \ o\ I.sel P pmnt equipment \ ." -/-FLEXLP LP\ " -tegles NN-EN -FLEX HP -.tegies rlrpLI IE -Additional I Iabilities I I E08Q1I I AFW-TNK-1 AF17I I I ,201 , g//iAFW-TN K-8EIWIS -017p I-F-TNK-20k g.1AW-TNK-620W-171AFW-TNK-320k galI-F.1F.Attachment BPage I of I/F-AFW-PMP-2Portable1000 psi/300 gpmDiesel Pumpers/RCS COOLINGSEISMIC STRATEGIESvMDAFWPs Expedited Seismic Evaluation Process ReportAttachment C -H.B. Robinson Steam Electric Plant Unit 2 RCS Boration and MakeupStrategiesPage 45 of 46 LEGENDBLACK -Installedplant equipmentn~~i 7 LakeRobinsonLEGENDMagenta -Mode 5/6Makeup (ReconfigureCheck Valve Covers)LCV-11SA toC/CS HUT,RED -PortableBoration andMakeup StrategyUsing the ChargingSystemE--3 SI-iGREEN -PortableBoration and MakeupStrategy Using theSafety Injection System//IF ,RCS BORATIONand MAKEUPSI ACC nt Expedited Seismic Evaluation Process ReportAttachment D -H.B. Robinson Steam Electric Plant FLEX Flow PathPage 46 of 46}}
Expedited Seismic Evaluation Process Report for H. B. Robinson Steam ElectricPlant, Unit No. 2cc: Ms. M. C. Barillas, NRC Project Manager, NRRMr. K. M. Ellis, NRC Senior Resident Inspector Mr. V. M. McCree, NRC Region II Administrator Expedited Seismic Evaluation Process ReportExpedited Seismic Evaluation Process ReportForH. B. Robinson Steam Electric Plant, Unit No. 2Page 3 of 46 Expedited Seismic Evaluation Process ReportEXPEDITED SEISMIC EVALUATION PROCESS REPORTTABLE OF CONTENT1.0 Purpose and Objective
.................................................................................
072.0 Brief Summary of the FLEX Seismic Implementation Strategies
..........................
073.0 Equipment Selection Process and ESEL .........................................................
133.1 Equipment Selection Process and ESEL ...............................................
133.1.1 ESEL Development
................................................................
143.1.2 Power Operated Valves ..........................................................
143.1.3 Pull Boxes ...........................................................................
143.1.4 Termination Cabinets
..............................................................
153.1.5 Critical Instrumentation Indicators
..............................................
153.1.6 Phase 2 and Phase 3 Piping Connections
..................................
153.2 Justification for Use of Equipment That is Not the Primary Means forFLEX Implementation
.....................................................................
154.0 Ground Motion Response Spectrum
..............................................................
164.1 Plot of GMRS Submitted by H.B. Robinson Steam Electric Plant ...............
164.2 Comparison to SSE ..........................................................................
195.0 Review Level Ground Motion (RLGM) ............................................................
235.1 Description of RLGM Selected
............................................................
235.2 Method to Estimate ISRS ..................................................................
266.0 Seismic Margin Evaluation Approach
............................................................
286.1 Summary of Methodologies Used ......................................................
286.2 HCLPF Screening Process ................................................................
296.3 Seismic Walkdown Approach
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306.3.1 W alkdown Approach
..............................................................
306.3.2 Application of Previous W alkdown Information
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326.3.3 Significant Walkdown Findings
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336.4 HCLPF Calculation Process ..............................................................
336.5 Functional Evaluation of Relays ..........................................................
336.6 Tabulated ESEL HCLPF Values (Including Key Failure Modes) .................
347.0 Inaccessible Items .....................................................................................
367.1 Identification of ESEL Items Inaccessible for Walkdown
..........................
367.2 Planned Walkdown/Evaluation Schedule/Close Out ................................
368.0 ESEP Conclusions and Results ....................................................................
378.1 Supporting Information
......................................................................
378.2 Identification of Planned Modifications
.................................................
388.3 Modification Implementation Schedule
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398.4 Summary of Planned Actions ............................................................
399.0 References
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40Page 4 of 46 Expedited Seismic Evaluation Process ReportList of FiguresFigure 2.1: Mark-Up Showing Addition of FLEX Tee Connection (AFW-166) to SDAFW Discharge at AFW -121 ..........................................................
9Figure 2.2: Mark-Up Showing Alternate Mechanical FLEX Connection (AFW-165) insidethe MDAFW Room on Line 4-AFW-23 and Upstream of AFW-54 ..................
10Figure 4.1: Plot of 1 E-4 and 1 E-5 UHRS and GMRS at Control Point forthe H.B. Robinson Steam Electric Plant ...................................................
18Figure 4.2: Comparison of the GMRS, SSE, and Ground LevelResponse Spectrum from Time History ...................................................
22Figure 5-1: Plot of 5% Damping 2 x SSE, 2 x Ground Level Response
: Spectrum, a nd G M R S .......................................................................................
2 6Figure 6.1: Comparison of the H.B. Robinson Steam Electric Plant IPEEE RLE, ESEPRLGM, SSE, and Ground Level (El. 226ft) Spectrum from Time History,and 2 x Ground Level (El. 226ft) Spectrum from Time History ...............
29List of TablesTable 4-1: GMRS and UHRS for the H.B. Robinson Steam Electric Plant ......................
17Table 4-2a: Original SSE Based on 0.2g Housner Spectrum for the H.B. RobinsonS team E lectric P lant ............................................................................
20Table 4-2b: Ground Level Response Spectrum Based on Time History for theH.B. Robinson Steam Electric Plant ........................................................
21Table 5-1: RLGM for H.B. Robinson Steam Electric Plant ..........................................
24Table 5-2: Ratio of G M RS to SSE .........................................................................
25Table 6-1: Functional and Anchorage HCLPF Capacity Results ..................................
35Attachments Attachment A -H.B. Robinson Steam Electric Plant ESELAttachment B -H.B. Robinson Steam Electric Plant Unit 2 RCS Cooling Strategies Attachment C -H.B. Robinson Steam Electric Plant Unit 2 RCS Boration and Makeup Strategies Attachment D -FLEX Flow PathPage 5 of 46 Expedited Seismic Evaluation Process ReportEXECUTIVE SUMMARYAn Expedited Seismic Evaluation Process has been completed for the H.B. Robinson SteamElectric Plant site based on endorsed guidance outlined in Electric Power Research Institute (EPRI)3002000704 (Reference 2). The work includes screening, equipment selection, development of theRLGM and in-structure
: demands, evaluating seismic capacity of components and development ofHigh Confidence of Low Probability of Failure (HCLPF) calculations, and implementation ofnecessary plant modifications.
HCLPF calculations revealed that Motor Control Center (MCC-A)required modification for the beyond design basis ground motion. Modifications have beendeveloped and implemented for MCC-A and a similar cabinet, MCC-B. Seismic margin above 2XSSE was also added to a group of instrument racks (Hagan Racks) by validating the boltingintegrity of the top braces. All items in the ESEL have seismic capacity that exceeds the demand ofthe RLGM. The ESEL has been updated to consider new equipment in FLEX strategy as outlinedin the updated Overall Integrated Plan. The FLEX strategy was subjected to critical path analysisand all the items required under the ESEP guidelines are included in the ESEL list.Page 6 of 46 Expedited Seismic Evaluation Process Report1.0 Purpose and Objective Following the accident at the Fukushima Dai-ichi nuclear power plant resulting from the March 11,2011, Great Tohoku Earthquake and subsequent
: tsunami, the Nuclear Regulatory Commission (NRC) established a Near Term Task Force (NTTF) to conduct a systematic review of NRCprocesses and regulations and to determine if the agency should make additional improvements toits regulatory system. The NTTF developed a set of recommendations intended to clarify andstrengthen the regulatory framework for protection against natural phenomena.
Subsequently, theNRC issued a 10 CFR 50.54(f) letter on March 12, 2012 (Reference 1), requesting information toassure 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 designbasis, further risk assessment may be required.
Assessment approaches acceptable to the staffinclude 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 H.B.Robinson Steam Electric Plant (RNP). The intent of the ESEP is to perform an interim action inresponse to the NRC's 50.54(f) letter (Reference
: 1) to demonstrate seismic margin through areview of a subset of the plant equipment that can be relied upon to protect the reactor corefollowing beyond design basis seismic events.The ESEP is implemented using the methodologies in the NRC endorsed guidance in EPRI3002000704, Seismic Evaluation Guidance:
Augmented Approach for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1: Seismic (Reference 2).The objective of this report is to provide summary information describing the ESEP evaluations andresults.
The level of detail provided in the report is intended to enable NRC to understand theinputs used, the evaluations performed, and the decisions made as a result of the interimevaluations.
2.0 Brief Summary of the FLEX Seismic Implementation Strategies The H.B. Robinson Steam Electric Plant FLEX strategies for Reactor Core Cooling and HeatRemoval, Reactor Inventory Control/Long-term Subcriticality, and Containment Function aresummarized below. The FLEX flow path is shown in Attachment D. The summary is derived fromthe H.B. Robinson Steam Electric Plant Overall Integrated Plan (OIP) in Response to the March 12,2012, Commission Order EA-12-049 (Reference 3), as supplemented by six-month updates(References 30, 31, and 32). Note that the H.B. Robinson Overall Integrated Plan (as amended in 6month updates) is based on Engineering Change (EC) 88926 (Reference 33).Reactor Core Cooling and Heat RemovalNEI 12-06, Diverse and Flexible Coping Strategies (FLEX) Implementation Guideline, Revision 0(Reference 34), requires that Auxiliary Feedwater (AFW) cooling be available to provide secondary makeup sufficient to maintain or restore Steam Generator (SG) level with installed equipment to thegreatest extent possible.
Beyond the use of installed equipment, steam generators must be able tobe depressurized in order to support makeup via portable pumps. Multiple and diverse connection points for the portable pumps must be provided and cooling water must be available indefinitely.
Refer to Attachment B (Reactor Coolant System Cooling Strategies) for depiction of the following discussion.
Page 7 of 46 Expedited Seismic Evaluation Process ReportThe H.B Robinson Steam Electric Plant FLEX strategies require that the AFW be in operation within 61 minutes of event initiation.
With the loss of AC power, a minimum of one steam supplyvalve (MS-V1-8A, MS-V1-8B, or MS-V1-8C) to Steam Driven Auxiliary Feedwater Pump (SDAFWP)and one AFW valve (AFW-V2-14A, AFW-V2-14B, AFW-V2-14C) to the steam generators must bemanually operated.
These required valves are all located in seismic Class 1 bay of the TurbineBuilding.
Additional portable backup for Steam Generator makeup is required per Section 3.2.2(13) of NEI12-06. The H.B. Robinson Steam Electric Plant has two strategies for portable backup. The firststrategy developed to satisfy this requirement is staging of two (2) intermediate pressure pumps(300 gpm at pressure of 1,000 psig) for all seismic events as described in detail below. The secondstrategy developed to satisfy the condition of Section 3.2.2(13) of NEI 12-06 is to store a Halepumper in a seismically robust Permanent FLEX Storage Building (PFSB). This strategy will involvethe use of the same primary and alternate connections described in the following paragraph, andwill require SG depressurization.
The two (2) pre-staged portable pumps (300 gpm at 1,000 psig) eliminate the need to depressurize the Steam Generators in the event the backup AFW feed capability is needed due to an AFWinterruption early in the ELAP transient as a result of seismic event. Either of the portable pumpscan take suction from a variety of plant sources (described below) and can be tied directly into theauxiliary feedwater system. Engineering Change 95266, Isolation Valves And Connection For AFW-FUKUSHIMA-Admin (Reference
: 48) was developed to add a FLEX tee connection (AFW-166) tothe SDAFWP discharge at AFW-121 (see Figure 2-1). Access to this primary connection is throughthe seismically qualified Turbine Building Class 1 bay. Engineering Change 90623, New Pipe TeeAnd Standard Connection For NTTF 4.2 (FLEX) (Reference
: 47) develops an alternate mechanical FLEX connection (AFW-165) inside the MDAFWP room on line 4-AFW-23 and upstream of AFW-54 (See Figure 2-2). EC90623 will be implemented during Refueling Outage, R0229. TheMDAFW room is housed in the seismic Class 1 Reactor Auxiliary Building (RAB).Page 8 of 46 Expedited Seismic Evaluation Process ReportFigure 2-1: Mark-Up Showing Addition of FLEX Tee Connection (AFW-166) to SDAFWDischarge at AFW-121Page 9 of 46 Expedited Seismic Evaluation Process ReportFigure 2-2: Mark-Up Showing Alternate Mechanical FLEX Connection (AFW-165) inside theMDAFWP Room on Line 4-AFW-23 and Upstream of AFW-54There are several sources for sustained cooling water supply. The primary source of AFWinventory is the seismically qualified condensate storage tank (CST) and its level instrumentation.
The CST is seismically robust and is the installed source of AFW to the SDAFWP. However, theCST inventory is not sufficient for indefinite coping (mission time is approximately 4 hours using theSDAFWP).
A secondary source of AFW inventory is the "Tank Farm" (portable pump) inside theprotected area that supplies the two pre-staged portable pumps (each with capacity of 300 gpmand 1,000 psig pressure as noted in the seismic strategy above). This source has a capacity ofapproximately 120,000 gallons and 10 hours of mission time using a pre-staged portable pump.The only other assured source of water is the Ultimate Heat Sink (Lake Robinson) which perrestrictions outlined in NEI 12-06 can only be accessed using portable equipment (assumes normalPage 10 of 46 Expedited Seismic Evaluation Process Reportaccess to the ultimate heat sink is lost). Given these limitations, one Phase 2/3 seismic strategy isto provide an indefinite supply of water to the CST and the SDAFWP by staging a portable dieselpumper at Lake Robinson with hoses routed to the CST. EC 90622, Standard Piping Connections For NTTF 4.2 (FLEX) (Reference
: 43) adds a FLEX connection at valve C-66 to provide anindefinite water supply to the CST. This can be accomplished during the initial CST/Tank Farmmission time of 14 hours.The H.B. Robinson Steam Electric Plant has developed several options for the Steam Generator depressurization capability.
The Steam Generator Power Operated Relief Valves (PORVs) arenormally operated using the Instrument Air System or with backup Nitrogen System and alignedusing Attachment 2 of EOP-ECA-0.0 (Reference 35). However, neither the primary Instrument Airnor the backup Nitrogen System are seismically qualified.
Therefore, the primary Instrument Air andthe backup Nitrogen System cannot be relied upon during or after seismic events. The Main SteamSafety Valves are an alternate option to depressurize the Steam Generators but this option is notrecommended per the PA-PSC-0965, PWROG Core Cooling Position Paper (Reference
: 37) andWCAP-17601-P, Revision 1, Reactor Coolant System response to Extended Loss of AC PowerEvent for Westinghouse, Combustion Engineering, and Babcock & Wilcox NSSS Designs forPhase Boration, August 2012 (Reference 36), which state that remaining on the Main Steam SafetyValves for an extended period may lead to failure of the valve(s) which subsequently will causeexcessive and uncontrolled RCS cooldown.
Current strategy is to align portable nitrogen tanks to the Steam Generator PORV header usingAttachment 1 (Connecting Emergency Pressure Source to Operate SG PORVS) or Attachment 2(S/G Manual Depressurization) of RNP procedure EDMG-004, Steam Generators (Reference 38).In addition to the SG PORV capabilities recommended in Reference 37, the H.B. Robinson SteamElectric Plant has also developed a strategy to cooldown the RCS using the main steam lineisolation valve bypass lines. The strategy is detailed in Section 3.27 (Cooldown Using MSIV BypassLines) of calculation RNP-M/MECH-1712, Appendix R Mechanical Basis (Reference 39). Thiscapability results in a cooldown rate of 830/hr which bounds the recommended Westinghouse cooldown rate of 750/hr described in Reference 37.After initiation of depressurization, it is desirable to isolate the Safety Injection (SI) Accumulators inorder to prevent injection of nitrogen into the RCS which will impede natural circulation cooldown.
During an ELAP, power to the SI Accumulator isolation valves is lost. Although, the isolation valvescan be operated
: manually, they are located inside the Containment Building and it is undesirable toperform this operation at this time due to personnel safety. The valves are powered by MCC 5 andMCC 6 and will be re-powered via Emergency Buses El and E2 with portable diesel generators staged in the seismic Class 1 Reactor Auxiliary Building (Drumming Room) for re-powering the Aand B Battery Chargers (see EC 90617 [Reference 40]). DB-50 Bus Feed Adapters can beinstalled in each of the Emergency Buses El and E2 and will be connected to the output of theDiesel Generators.
As part of the Phase 2 strategy, Steam Generator pressure will be maintained above the pressure corresponding to the SI Accumulator injection (240 psig) until the SIAccumulator isolation valves are closed using FLEX Support Guideline (FSG) 10, Passive RCSInjection Isolation (Reference 41).Reactor Inventory Control/Long-Term Subcriticality Refer to Attachment C (Reactor Coolant System Boration and Makeup Strategies) for a depiction ofthe following discussion.
There is no installed means of providing borated makeup following anELAP. The primary method of boration and inventory control is the use of portable high pressureand low volume pump directly connected to the Charging Lines or Safety Injection Headers fromthe Refueling Water Storage Tank (RWST) or a portable tanker containing borated water (seeEC95216, NTTF 2.1 Interim Action RCS Injection
[Reference 42]). The RWST is seismically Page 11 of 46 Expedited Seismic Evaluation Process Reportdesigned and will remain operational during and after a design basis seismic event. The makeupcapacity of the portable pump is 60 gpm at a pressure of 2,000 psig which is adequate for thebounding analysis in WCAP-1760-P (Reference 36). Phase 3 inventory control will beaccomplished using the same portable Phase 2 boration/makeup strategy.
Portable high pressurepumping and portable tanker capability will be stored in the PFSB to support this strategy.
EC 90622 (Reference
: 43) adds a FLEX connection to the exposed end downstream of the normallylocked closed drain valve (SI-837) located at the base of the RWST to access this borated water ifit available.
This portable strategy will deliver borated water to the RCS through valves CVC-121A/B (primary) or SI-888P/S (alternate).
Containment FunctionCalculation RNP-M/MECH-1877, RNP Extended Loss of AC Power (ELAP) Containment Response(Reference
: 45) was developed to determine the containment temperature and pressure responseassuming an ELAP and a trip from 100% reactor power at 100 days into the cycle. Results inReference 45 indicate that the Containment Building design limits for temperature and pressure willnot be challenged in the first 43 days following the event. This analysis assumes that: (1) no actionis taken to cool, spray, or vent the containment; and (2) low leakage RCP seals are installed.
Therefore, Phase 1 and 2 strategies are not required.
There is sufficient time and resources inPhase 3 to assemble a strategy using the National Safer Response Center (NSRC) pumpers andgenerators, prefabricated electrical connections, and prefabricated SW connections that will bestored in the PFSB. FSG-12, Alternate Containment Cooling (Reference
: 46) provides instructions for several existing strategies including external containment cooling which does not require use ofany plant system. These particular activities will be determined and directed by the Emergency Response Organization (Technical Support Center) based on the effects of the Beyond DesignBasis External Event (BDBEE) and the state of existing equipment.
Instrumentation Instrumentation channels that are powered by station batteries will be lost upon depletion of thebatteries.
FLEX strategies to improve battery coping occur by extending Phase 1. Phase 1 isextended by strategic load shedding followed by additional deep load shedding in the first hour ofthe event to extend battery coping times to 3.25 -3.75 hours. Phases 2 and 3 battery copingrequire portable diesel generators to power the vital battery chargers.
Two FLEX diesel generators will be mounted in their deployed positions near the battery chargers and within the ReactorAuxiliary Building.
Each generator will be sized to power two vital battery chargers, room air supplyand exhaust fans, and safety injection accumulator isolation valves. Electrical cables and pre-installed connectors will be routed from the FLEX diesel generators to the battery room for quickconnection of the cables to each of the battery chargers.
The primary strategy is to power the A andB vital battery chargers from one or both of the pre-staged FLEX generators.
The alternate is topower the A-1 and B-1 vital battery chargers from one or both of the pre-staged FLEX generators.
See Reference 40 for complete details of this strategy.
Page 12 of 46 Expedited Seismic Evaluation Process Report3.0 Equipment Selection Process and ESELThe selection of equipment for the Expedited Seismic Equipment List (ESEL) followed theguidelines of EPRI 3002000704.
The complete ESEL for H. B. Robinson Unit 2 is presented inAttachment A.3.1 Equipment Selection Process and ESELThe selection of equipment to be included on the ESEL was based on installed plant equipment credited in the FLEX strategies during Phase 1, 2 and 3 mitigation of a BDBEE as described in theH.B. Robinson Steam Electric Plant OIP (Reference
: 3) in response to the March 12, 2012Commission Order EA-12-049 as revised in References 30 through 32.The scope of "installed plant equipment" includes equipment relied upon for the FLEX strategies tosustain the critical functions of core cooling and containment integrity consistent with Reference 3and References 30 through 32. FLEX recovery actions are excluded from the ESEP scope perEPRI 3002000704.
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 andsubcriticality, and containment integrity functions.
Portable and pre-staged FLEX equipment (notpermanently installed) are excluded from the ESEL per EPRI 3002000704.
The ESEL component selection followed the EPRI guidance outlined in Section 3.2 of EPRI3002000704.
: 1. The scope of components is limited to that required to accomplish the core cooling andcontainment safety functions identified in Table 3-2 of EPRI 3002000704.
The instrumentation monitoring requirements for core cooling/containment safety functions are limited to thoseoutlined in the EPRI 3002000704
: guidance, and are a subset of those outlined in the H.B.Robinson Steam Electric Plant OIP and as revised in the first , second and third six-month status reports.2. The scope of components is limited to installed plant equipment, and FLEX connections necessary to implement the H.B. Robinson Steam Electric Plant OIP (Reference
: 3) inresponse to the March 12, 2012 Commission Order EA-12-049 and as revised in References 30 through 32. and as described in Section 2.3. The scope of components assumes the credited FLEX connection modifications areimplemented, and are limited to those required to support a single FLEX success path (i.e.,either "Primary" or "Back-up/Alternate").
: 4. The "Primary" FLEX success path is to be specified.
Selection of the "Back-up/Alternate" FLEX success path must be justified.
: 5. Phase 3 coping strategies are included in the ESEP scope, whereas recoverystrategies are excluded.
: 6. Structures,
: systems, and components excluded per the EPRI 3002000704 (Reference 2)guidance are:" Structures (e.g. Reactor Containment
: Building, Reactor Auxiliary
: Building, etc.)" Piping, cabling,
: conduit, HVAC, and their supports.
" Manual valves and rupture disks." Power-operated valves not required to change state as part of the FLEX mitigation strategies.
Page 13 of 46 Expedited Seismic Evaluation Process Report* Nuclear steam supply system components (e.g. reactor pressure vessel and internals, reactor coolant pumps and seals, etc.)7. For cases in which neither train was specified as a primary or back-up strategy, then only onetrain component (generally
'A' train) is included in the ESEL.3.1.1 ESEL Development The ESEL was developed by reviewing the H.B. Robinson Steam Electric Plant OIP (Reference 3)and revisions in three subsequent six-month status reports to determine the major equipment involved in the FLEX strategies.
Further reviews of plant drawings (e.g., Process andInstrumentation Diagrams (P&IDs)(EC92103R0, Attachment Z03RO Mechanical Documents
[Reference 49]), and Electrical One Line Diagrams (EC92103R0, Attachment Z05R0 Electrical Documents
[Reference 50]) were performed to identify the boundaries of the flowpaths to be usedin the FLEX strategies and to identify specific components in the flowpaths needed to supportimplementation of the FLEX strategies.
Boundaries were established at an electrical or mechanical isolation device (e.g., isolation amplifier, valve, etc.) in branch circuits
/ branch lines off the definedstrategy electrical or fluid flowpath.
P&IDs were the primary reference documents used to identifymechanical components and instrumentation.
The flow paths used for FLEX strategies wereselected and specific components were identified using detailed equipment and instrument
: drawings, piping isometrics, electrical schematics and one-line
: drawings, system descriptions, design basis documents, etc., as necessary.
3.1.2 Power Operated ValvesPage 3-3 of EPRI 3002000704 notes that power operated valves not required to change state areexcluded from the ESEL. Page 3-2 also notes that "functional failure modes of electrical andmechanical portions of the installed Phase 1 equipment should be considered (e.g. AFW trips)."To address this concern, the following guidance is applied in the H.B. Robinson Steam ElectricPlant ESEL for functional failure modes associated with power operated valves:" 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, thevalves 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 3strategies, were not evaluated for spurious valve operation as the seismic event that causedthe ELAP has passed before the valves are re-powered.
3.1.3 Pull BoxesPull boxes were deemed unnecessary to add to the ESEL as these components provide completely passive locations for pulling or installing cables. No breaks or connections in the cabling areincluded in pull boxes. Pull boxes were considered part of conduit and cabling, which are excludedin accordance with EPRI 3002000704.
Page 14 of 46 Expedited Seismic Evaluation Process Report3.1.4 Termination CabinetsTermination
: cabinets, including cabinets necessary for FLEX Phase 2 and Phase 3 connections, provide consolidated locations for permanently connecting multiple cables. The termination cabinets and the internal connections provide a completely passive function;
: however, the cabinetsare included in the ESEL to ensure industry knowledge on panel/anchorage failure vulnerabilities isaddressed.
3.1.5 Critical Instrumentation Indicators Critical indicators and recorders are typically physically located on panels/cabinets and are includedas separate components.
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
"... FLEXconnections necessary to implement the H.B. Robinson Steam Electric Plant OIP as described inSection 2." Item 3 in Section 3.1 also notes that "The scope of components assumes the creditedFLEX connection modifications are implemented, and are limited to those required to support asingle FLEX success path (i.e., either "Primary" or "Back-up/Alternate")."
Item 6 in Section 3.1 above goes on to explain that "Piping,
: cabling, conduit, HVAC, and theirsupports" are excluded from the ESEL scope in accordance with EPRI 3002000704.
Therefore, piping and pipe supports associated with FLEX Phase 2 and Phase 3 connections areexcluded from the scope of the ESEP evaluation.
: However, any active valves in FLEX Phase 2and Phase 3 connection flow path are included in the ESEL.3.2 Justification for Use of Equipment That Is Not The Primary Means for FLEXImplementation In accordance with EPRI 3002000704, the H.B. Robinson Steam Electric Plant used equipment that is the primary means of implementing FLEX strategy.
The complete ESEL for the H.B.Robinson Steam Electric Plant is presented in Attachment A.Page 15 of 46 Expedited Seismic Evaluation Process Report4.0 Ground Motion Response Spectrum (GMRS)4.1 Plot of GMRS Submitted by the H.B. Robinson Steam Electric PlantFollowing completion of the seismic hazard re-evaluation as requested in Reference 1, the NRC 10CFR 50.54(f) letter, a screening process is needed to determine if an interim seismic risk evaluation like the EPRI ESEP is required.
The screening GMRS was determined with control point seismichazard re-evaluation.
In accordance with the 50.54(f) letter and following the guidance in EPRIScreening, Prioritization, and Implementation Details (SPID) (Reference 15), Probabilistic SeismicHazard Analysis (PSHA) was performed using the 2012 CEUS Seismic Source Characterization forNuclear Facilities (Reference 20), a Regional Seismic Catalog Correction (Reference 61), andupdated EPRI Ground Motion Model (GMM) for the CEUS (Reference 21). Development of theH.B. Robinson Steam Electric Plant Ground Motion Response Spectra (GMRS) is documented inReferences 4 and 62. The GMRS and Uniform Hazard Response Spectra (UHRS) are tabulated inTable 4-1 and then compared in Figure 4-1 with the 5% damped horizontal SSE. Note thatadditional seismic hazard analysis and GMRS development is underway for H.B. Robinson SteamElectric Plant to support completion of the seismic probabilistic risk analysis.
In the analysis, newlyacquired geophysical testing results are being used to update the site response analysis.
Theresults of the screening evaluation discussed will not change as a result of the newly acquiredgeophysical testing.
These new geophysical testing data allow for a more accurate representation of seismic hazard and seismic probabilistic risk assessment by eliminating a significant source ofuncertainty.
Page 16 of 46 Expedited Seismic Evaluation Process ReportTable 4-1: GMRS and UHRS for the H.B. Robinson Steam Electric PlantFreq. (Hz) 10-4 UHRS (g) 105 UHRS (g) GMRS100 4.20E-01 9.17E-01 4.71E-0190 4.23E-01 9.31 E-O1 4-77E-0180 4.27E-01 9-48E-01 4.85E-0170 4.35E-01 9.73E-01 4.97E-0160 4-54E-01 1.02E+00 5.19E-0150 4.98E-01 1.11E+00 5.66E-0140 5.74E-01 1.25E+00 6.43E-0135 6.21 E-01 1.35E+00 6.95E-0130 6.63E-01 1.46E+00 7.50E-0125 7.23E-01 1.61E+00 8.21E-0120 7.92E-01 1.75E+00 8.97E-0115 8.09E-01 1.82E+00 9.27E-O112.5 8.35E-01 1.82E+00 9.36E-0110 8.52E-01 1.86E+00 9.55E-019 8.40E-01 1.84E+00 9.42E-018 8.58E-01 1.84E+00 9.49E-017 8.98E-01 1-92E+00 9.88E-016 8.87E-01 1.95E+00 9.99E-015 8.57E-01 1.87E+00 9.61E-014 8.40E-01 1.83E&#xf7;00 9.39E-013.5 7.71 E-01 1.76E+00 8.94E-013 6.79E-01 1.59E+00 8.04E-012.5 6.08E-01 1.38E+00 7.04E-O12 5.37E-01 1.30E+00 6.52E-011.5 3.97E-01 1.05E+00 5.20E-011.25 3.23E-01 8.58E-01 4.23E-011 2.26E-01 6.44E-01 3.13E-010.9 1.87E-01 5.52E-01 2.67E-010.8 1.56E-01 4.69E-01 2.26E-010.7 1.31E-01 3.95E-01 1.90E-010.6 1.10E-01 3.25E-01 1.57E-010.5 8.86E-02 2.51E-01 1.22E-010.4 7.09E-02 2.01E-01 9.79E-020.35 6.20E-02 1.76E-01 8.57E-020.3 5.32E-02 1.51 E-01 7.34E-020.25 4-43E-02 1.26E-01 6.12E-020.2 3.55E-02 1.00E-01 4.90E-020.15 2-66E-02 7.54E-02 3.67E-020.125 2.22E-02 6.28E-02 3.06E-020.1 1.77E-02 5.02E-02 2.45E-02Page 17 of 46 Expedited Seismic Evaluation Process ReportMean Soil UHRS and GMRS at Robinson2-52.-1E-5 UHRSL1.50.0.1 1 10 100Spectral frequency, HzFigure 4-1: Plot of 1 E-4 and 1 E-5 UHRS and GMRS at Control Point for the H.B. RobinsonSteam Electric Plant (5% Damped Response Spectra)Control point hazard curves were used to develop the UHRS and the GMRS. The methodology described in SPID (Reference
: 15) was used to compute site-specific control point hazard curves.The selection of control point elevation is based on recommendations in Section 2.4.2 of the SPID(Reference 15). The control point elevation for the H.B. Robinson Steam Electric Plant is at El. 226feet based on information in Sections 2.5 and 2.7 of the Updated Final Safety Analysis Report(Reference 51).Page 18 of 46 Expedited Seismic Evaluation Process Report4.2 Comparison to SSEOriginal design of the H.B. Robinson Steam Electric Plant was based on the 0.2g HousnerSpectrum.
Table 4-2a shows the spectral acceleration values as a function of frequency for the 5%damped horizontal SSE. As will be discussed in more detail in Section 5.2, original design in-structure response spectra was developed based on conservative time history.
The Ground LevelResponse Spectrum that results from this time history is reported in Table 4-2b.A comparison of the Ground Level Response
: Spectrum, SSE, and GMRS is shown in Figure 4-2.As shown in Figure 4-2, in the 1 to 10 Hz frequency range of the response
: spectrum, the GMRSexceeds the SSE and the Ground Level Response Spectrum.
The GMRS also exceeds the SSEand the Ground Level Response Spectrum at frequency values higher than 10 Hz.Page 19 of 46 Expedited Seismic Evaluation Process ReportTable 4-2a: Original SSE Based on 0.2g Housner Spectrum for the H.B. Robinson Steam ElectricPlant (5% Damping)Frequency SSE(Hz) (g)1.0 0.171.5 0.2302.0 0.2602.5 0.2903.0 0.33.5 0.3104.0 0.325.0 0.3056.0 0.2907.0 0.2658.0 0.2559.0 0.24010.0 0.2312.50 0.21015.0 0.220.0 0.225.0 0.230.0 0.233.0 0.235.0 0.2Page 20 of 46 Expedited Seismic Evaluation Process ReportTable 4-2b: Ground Level Response Spectrum Based on Time History for H.B. Robinson SteamElectric Plant (5% Damping)Frequency Ground Level(Hz) ResponseSpectrum (g)from TimeHistory1.0 0.3001.5 0.4552.0 0.4412.5 0.4173.0 0.4453.5 0.4684.0 0.4895.0 0.4556.0 0.4157.0 0.3808.0 0.3519.0 0.31610.0 0.28112.50 0.22115.0 0.23220.0 0.24625.0 0.25830.0 0.26733.0 0.27335.0 0.275Page 21 of 46 Expedited Seismic Evaluation Process Report1.200 --1.00 from :1m1Hi tory00.800 -~ -- -----o0.600 _0.40C,0.2000.0000.1 1.0 10.0 100.0Frequency (Hz)Figure 4-2: Comparison of GMRS, SSE and Ground Level Response Spectrum from Time HistoryPage 22 of 46 Expedited Seismic Evaluation Process Report5.0 Review Level Ground Motion (RLGM)5.1 Description of RLGM SelectedPlants for which the GMRS exceeds the SSE in the 1.0 to 10.0 Hz frequency range do not screenout of the ESEP and require further seismic evaluation.
The further seismic evaluation is performed to a Review Level Ground Motion which consists of a response spectrum above the SSE level. TheRLGM is defined as a response spectrum reflecting an earthquake level that is above the plant'sdesign basis SSE. The RLGM can be computed using one of the following criteria as described inReference 2:1. The RLGM can be derived by linearly scaling the SSE by the maximum ratio of the horizontal GMRS to the 5% damped SSE, between the 1 and 10 Hz frequency range, but not to exceed aratio greater than 2 times the SSE. The in-structure seismic motions corresponding to theRLGM would be derived using existing SSE-based In-Structure Response Spectra (ISRS)scaled with the same factor.2. Alternatively, licensees who have developed appropriate structural/soil-structure interaction (SSI) models capable of calculating ISRS based on site GMRS/Uniform Hazard ResponseSpectrum (UHRS) input may opt to use these ISRS in lieu of scaled SSE ISRS. In this case,the GMRS would represent the RLGM. EPRI 1025287 and the American Society ofMechanical Engineers/American Nuclear Society (ASME/ANS)
PRA Standard give guidanceon acceptable methods to compute both the GMRS and the associated ISRS. Section 4 ofReference 2 contains full description of this task.The RLGM for the H.B. Robinson Steam Electric Plant was developed in Reference 52 and inaccordance with the methodology and objectives in EPRI ESEP guidance Reference
: 2. The RLGMis the SSE multiplied by a factor of 2.0. Table 5-1 is the RLGM as a function of frequency andacceleration at 5% damping.
As discussed under Sections 4.2 and 5.2, original design in-structure response spectra were developed based on a conservative time history.
The Ground LevelResponse Spectrum that resulted from this time history is reported in Table 4-2b and Figure 5-2.For consistency between component screening and component evaluations, the Ground LevelResponse Spectrum was scaled by 2 to represent an effective RLGM for component screening.
Therefore, both screening and evaluation of ESEL items were conservatively based on 2 x GroundLevel Response Spectrum (see Figure 6-1 for plot of 2 x Ground Level Response Spectrum) instead of 2 x SSE.Page 23 of 46 Expedited Seismic Evaluation Process ReportTable 5-1: RLGM for H.B. Robinson Steam Electric PlantFrequency SSE RLGM(Hz) (g) (g)1.0 0.17 0.341.5 0.230 0.4602.0 0.260 0.5202.5 0.290 0.583.0 0.3 0.603.5 0.310 0.624.0 0.32 0.645.0 0.305 0.616.0 0.290 0.587.0 0.265 0.538.0 0.255 0.519.0 0.240 0.4810.0 0.23 0.4612.50 0.210 0.4215.0 0.2 0.420.0 0.2 0.425.0 0.2 0.430.0 0.2 0.433.0 0.2 0.435.0 0.2 0.4The ratio of the GMRS to the SSE is summarized in Table 5-2. The maximum ratio of the GMRS toSSE is 4.635 and this occurs at frequency of approximately 15Hz. In the frequency range of 1 to10Hz, the maximum ratio of the GMRS to SSE is 4.152. As limited in EPRI 3002000704, the RLGMis determined multiplying the SSE by a factor of 2.0.Page 24 of 46 Expedited Seismic Evaluation Process ReportTable 5-2: Ratio of GMRS to SSEFrequency GMRS SSE GMRSISSE(Hz) (g) (g)1.0 0.313 0.17 1.8411.5 0.520 0.230 2.2612.0 0.652 0.260 2.5082.5 0.704 0.290 2.4283.0 0.804 0.3 2.6803.5 0.894 0.310 2.8844.0 0.939 0.32 2.9345.0 0.961 0.305 3.1516.0 0.999 0.290 3.4457.0 0.988 0.265 3.7288.0 0.949 0.255 3.7229.0 0.942 0.240 3.92510.0 0.955 0.23 4.15212.50 0.936 0.210 4.45715.0 0.927 0.2 4.63520.0 0.897 0.2 4.48525.0 0.821 0.2 4.10530.0 0.750 0.2 3.75033.0 0.717 0.2 3.58535.0 0.695 0.2 3.475Page 25 of 46 Expedited Seismic Evaluation Process Report1.2001.000S0o.80000.6000.4000.200L20.0000.1 1.0 10.0 100.0Frequency (Hz)Figure 5-1: Plot of 5% Damping 2xSSE, 2 x Ground Level Response
: Spectrum, and GMRS5.2 Method to Estimate ISRSThe seismic demand of the ESEL items/element mounted rigidly to the structure can be specified interms of the In-Structure Response Spectra (ISRS). For use in the ESEP, the in-structure seismicdemand for an element listed in the ESEL is defined by the ISRS scaled by the same factor used toobtain the RLGM from the SSE. The guidance under Section 4 of Reference 7 recommends broadening the peaks of the ISRS to account for the uncertainty in the civil structure frequency calculation.
The extent of broadening is suggested to be at least 15 percent of the frequency approaching and proceeding spectral peaks but can be increased beyond the minimumrecommendation based on the level of uncertainty associated with the structural model.The original design basis ISRS for the H.B. Robinson Steam Electric Plant were generated in 1970by Westinghouse Electric Corporation using mathematical building models developed by EbascoServices, Inc. The original ISRS or floor spectra generated by Westinghouse was limited in scopeand only considered the 0.20g design basis earthquake at damping ratio of 0.005 (0.5 percent).
These ISRS include conservatisms that result from conservative selection of the time history andexcessive bounding of design spectra.
Figure 4-2 shows plot of: (1) Ground Level ResponseSpectrum; (2) GMRS; and (3) SSE.Additional ISRS for other damping values were generated.
The task of generating the additional floor response spectra was complicated by lack of availability of time history data from the originalWestinghouse analysis.
Consequently, synthetic ground motion time history that generates ISRSPage 26 of 46 Expedited Seismic Evaluation Process Reportcomparable to the original Westinghouse floor spectra was used. The ISRS were generated byinputting the synthetic ground motion through the original Ebasco structural models. Scale factorsas a function of frequency were developed by comparing the spectra at the desired damping ratioagainst the 0.50 percent damping spectra.
The factors were then used to scale the originalWestinghouse 0.50 percent damped spectra to the desired damping ratio. The reconstituted ISRSat the various damping ratios have been incorporated into the H.B. Robinson Steam Electric Plant'sdesign basis ISRS documentation in Reference 18.The ISRS from Reference 18 were peak broadened in accordance with guidance in Regulatory Guide 1.122 (Reference 19). Since the ISRS in Reference 18 are already broadened, these spectraare scaled by a factor of 2.0 for ESEP.In summary, in-structure response spectra developed with the conservative Ground LevelResponse Spectrum were scaled by a factor of 2 for use in ESEP. Figure 5-1 shows plot of the 2 xSSE (RLGM), 2 x Ground Level Response
: Spectrum, and GMRS.Page 27 of 46 Expedited Seismic Evaluation Process Report6.0 Seismic Margin Evaluation ApproachIt is necessary to demonstrate that ESEL items have sufficient seismic capacity to meet or exceedthe demand characterized by the RLGM and the corresponding scaled in-structure responsespectra.
The seismic capacity is characterized as the peak ground acceleration (PGA) for whichthere is a high confidence of a low probability of failure (HCLPF).
The PGA is associated with aspecific spectral shape, in this case the 5%-damped 2 x Ground Level Response Spectrum shape.The HCLPF capacity must be equal to or greater than the RLGM PGA. The criteria for seismiccapacity determination are given in Section 5 of EPRI 3002000704.
There are two basic approaches for developing HCLPF capacities:
: 1. Deterministic approach using the conservative deterministic failure margin (CDFM)methodology of EPRI NP-6041, A Methodology for Assessment of Nuclear Power PlantSeismic Margin (Revision
: 1) (Reference 7).2. Probabilistic approach using the fragility analysis methodology of EPRI TR-103959, Methodology for Developing Seismic Fragilities (Reference 8).6.1 Summary of Methodologies UsedThe H. B. Robinson Steam Electric Plant completed a seismic margin assessment (SMA) in 1993.The SMA is documented in Reference 9 and consisted of screening, walkdowns by SRT, andHCLPF anchorage calculations.
The screening and walkdowns used the screening tables fromChapter 2 of EPRI NP-6041 (Reference
: 7) for peak spectral acceleration less than 0.8g. Thewalkdowns were conducted by engineers trained in EPRI NP 6041 (the engineers attended theEPRI SMA Add-On course in addition to the SQUG Walkdown Screening and Seismic Evaluation Training Course),
and were documented on Screening Evaluation Work Sheets from EPRI NP-6041. Anchorage capacity calculations used the CDFM criteria from EPRI NP-6041.
Seismicdemand was the IPEEE Review Level Earthquake (RLE) for SMA (mean NUREG/CR-0098
[Reference 11] ground response spectrum anchored to 0.3g PGA).Figure 6-1 shows the mean NUREG/CR-0098 ground response spectrum used as the IPEEE RLEcompared to the 2 x Ground Level Response Spectrum.
The figure shows that the ESEP inputmotion enveloped the IPEEE RLE at all frequencies except between 10 Hz and 15 Hz where theIPEEE RLE slightly exceed the ESEP input motion. The frequency of interest for ESEL items isbetween 1 Hz and 10Hz.The ESEP methodology included screening and extensive walkdown by the Seismic Review Team(SRT), and HCLPF calculations to evaluate structural capacity of the ESEL items against theRLGM. Function evaluation of relays was also performed.
The walkdowns were documented onScreening Evaluation Worksheets (SEWS) from EPRI NP-6041.
Based on outcome of the seismicwalkdown and documentation in SEWS, six (6) HCLPF calculations were performed to envelopethe thirteen (13) ESEL items identified during the walkdowns.
Page 28 of 46 Expedited Seismic Evaluation Process Report1.2001.000-2 X Ground LevelTim History-.SS-IPEEE RILE-Ground Level SpaHistory,-&#xfd;2 XSSE1(RLGM Sfromn04..CO)0.8000.6000.4000.2000.0000.11.010.0100.0Frequency (Hz)Figure 6-1.Comparison of the H.B. Robinson Steam Electric Plant IPEEE RLE, ESEPRLGM, SSE, Ground Level (El. 226ft) Spectrum from Time History, and 2 xGround Level (El. 226ft) Spectrum from Time History6.2 HCLPF Screening ProcessThe HCLPF screening and calculations were based on 2 x Ground Level Response Spectrum peakground acceleration.
Screening tables in EPRI NP-6041 (Reference
: 7) are based on peak spectralacceleration of < 0.8g, 0.8g to 1.2g, and > 1.2g. Since 2 x Ground Level Response Spectrum peakground acceleration is 0.978g, screening of ESEL items was based on the 0.8g to 1.2g rangecriteria.
The screening guidelines were supplemented by Appendix A of EPRI NP-6041 SL whichprovides the basis for the seismic capacity screening guidelines.
Anchorage capacity calculations were based on 2 x Ground Level Response Spectrum.
Equipment for which the screening caveats were met and for which the anchorage capacity exceeded 2 xGround Level Response Spectrum seismic demand were screened out from ESEP seismiccapacity determination.
Page 29 of 46 Expedited Seismic Evaluation Process Report6.3 Seismic Walkdown Approach6.3.1 Walkdown ApproachWalkdowns were performed in accordance with the criteria provided in Section 5 of EPRI3002000704 (Reference 2), which refers to EPRI NP-6041 (Reference
: 7) for the Seismic MarginAssessment process.
Pages 2-26 through 2-30 of EPRI NP-6041 describe the seismic walkdowncriteria, including the following key criteria.
"The SRT [Seismic Review Team] should "walk by" 100% of all components which are reasonably accessible and in non-radioactive or low radioactive environments.
Seismic capability assessment of components which are inaccessible, in high-radioactive environments, or possibly withincontaminated containment, will have to rely more on alternate means such as photographic inspection, more reliance on seismic reanalysis, and possibly, smaller inspection teams and morehurried 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 orpanels for detailed inspection of all components.
This walkdown is not intended to be a QA or QCreview or a review of the adequacy of the component at the SSE levelIf the SRT has a reasonable basis for assuming that the group of components are similar and aresimilarly
: 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 onecomponent or each type which is selected should be thoroughly inspected which probably doesmean de-energizing and opening cabinets or panels for this very limited sample. Generally, aspare representative component can be found so as to enable the inspection to be performed whilethe 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 ofcomponents the SRT should look closely at the first items and compare the field configurations withthe construction drawings and/or specifications.
If a one-to-one correspondence is found, thensubsequent items do not have to be inspected in as great a detail. Ultimately the walkdownbecomes 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 inparallel.
If serious exceptions to the drawings or questionable construction practices are foundthen 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 thatshown on drawings or prescribed in criteria for that component, potential SI [Seismic Interaction 1]problems, situations that are at odds with the team members' past experience, and any other areasof serious seismic concern.
If any such concerns
: surface, then the limited sample size of onecomponent of each type for thorough inspection will have to be increased.
The increase in samplesize 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'EPRI 3002000704 page 5-4 limits the ESEP seismic interaction reviews to "nearby block walls" and "pipingattached to tanks" which are reviewed "to address the possibility of failures due to differential displacements."
Otherpotential seismic interaction evaluations are "deferred to the full seismic risk evaluations performed in accordance with EPRI 1025287'Page 30 of 46 Expedited Seismic Evaluation Process Reportones who are responsible for the seismic adequacy of all elements which they screen from themargin review. Appendix D gives guidance for sampling selection."
As part of the ESEP, demonstration that the components listed in the ESEL have a HCLPFcapacity that exceeds the effective RLGM (2 x Ground Level Response Spectrum) verifiesadequate seismic ruggedness.
Section 5 of EPRI ESEP guidance specifies that the methodology inEPRI NP-6041 SL may be used for the development of the HCLPF capacity.
The major steps inReference 7 include pre-screening, walkdowns, and the CDFM HCLPF calculations.
In order to ensure efficiency while performing the walkdowns and during seismic capacityevaluations, each of the items listed in the ESEL were subjected to pre-screening.
The initial pre-screening effort consisted of data collection in the form of drawings, calculations, specifications, and vendor documents for each item in the ESEL. After identification of documentation for aspecific item, the pre-screening process followed the general seismic capacity screening guidelines presented in Reference 7 for civil structures, equipment, and subsystems to be considered screened out from further review. The caveats and footnoted exceptions and restrictions listed arefollowed.
For the purpose of completing the ESEP for the H. B. Robinson Steam Electric Plant, only Table 2-4 of Reference 7 is relevant for applying seismic screening criteria for plant equipment listed in theESEL. In addition to using the screening criteria in Reference 7 during plant walkdown, the SRTalso exercised their collective experience and judgment while using the criteria for specificcomponent.
The screening criteria can be used for equipment that is approximately 40ft abovegrade or lower. EPRI Report No 1019200 (Reference
: 23) provides guidance on screening criteriafor equipment that is greater than 40ft above grade. Screening criteria in Reference 7 do notinclude considerations for anchorage.
Therefore, structural integrity of anchorage was evaluated separately.
Some simple cases were documented on the SEWS form.Plant walkdowns were performed for items in the ESEL using guidance in Reference
: 7. Information extracted from existing documentation such as equipment
: location, seismic input elevation, relevant drawing details, and previous seismic capacity calculations were recorded on the ESEPSEWS and used during the walkdowns.
In accordance with the ESEP guidance, the SEWS thatwere used in the ESEP walkdowns were consistent with content and format of the SEWSpresented in Appendix F of EPRI NP-6041 SL.A major part of the ESEP walkdowns was the investigation of equipment anchorages.
Therefore, cabinets with anchorages located internally were opened. Furthermore, the ESEP guidance statesthat components that are anchored to sub-structural elements that may not have the same capacityas the main structural system (e.g. block walls, frames, stanchions etc.) should also be reviewed.
Nearby block walls were identified and evaluated as necessary.
Piping attached to tanks were alsoreviewed.
Other potential seismic interaction evaluations were deferred to a full Seismic RiskEvaluation (SRE) as discussed in the SPID References 14 and 15, and were not addressed in theESEP walkdowns.
Walkdown assessment for the H.B. Robinson Steam Electric Plant ESEL items were completed bythe SRT between August 2013 and February 2014. Some of the components were previously walked down during the IPEEE, USI A-46, or NTTF 2.3: Seismic and relevant information such asthe equipment
: location, seismic input elevation, drawing details and previous seismic calculations were recorded on the ESEP SEWS. Previous walkdowns were credited since they were performed by qualified Seismic Review Team. A walk-by of these components was performed anddocumented.
The objective of the walk-by is to confirm and verify that the components and theiranchorage have not degraded since the previous walkdown.
Items included in the ESEL that have not been previously walked down and evaluated, wereautomatically included for a detailed walkdown.
Page 31 of 46 Expedited Seismic Evaluation Process ReportThe SRT was comprised of at least two SQUG trained engineers and often included two additional structural engineers (Reference 57). The results of the walkdowns were documented on the SEWSfor each item. The completed SEWS and pictures taken during the walkdowns for the ESEL aredocumented in Reference
: 55. Follow-up inspections and walkdowns were completed whereadditional information was necessary.
6.3.2 Application of Previous Walkdown Information Previous seismic walkdowns from IPEEE and USI A-46 were used to support the ESEP seismicevaluations.
Some of the components and items on the ESEL were included in the NTTF 2.3seismic walkdowns (Reference 17). Those walkdowns were well documented and recent enoughthat they did not need to be repeated for the ESEP.Several ESEL items were previously walked down during the H.B. Robinson Steam Electric PlantSeismic IPEEE program.
Those walkdown results were reviewed and the following steps weretaken 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 relatedto 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.Page 32 of 46 Expedited Seismic Evaluation Process Report6.3.3 Significant Walkdown FindingsConsistent with guidance from NP-6041, no significant outliers or anchorage concerns (exceptMCC-A) were identified during the H.B. Robinson Steam Electric Plant seismic walkdowns.
Thefollowing findings were noted during the walkdowns.
* Nearby block walls were identified in the proximity of ESEL item. These block walls wereassessed for their structural adequacy to withstand the seismic loads resulting from theRLGM. There is no case where the block wall represented the HCLPF failure mode for anESEL item.* Piping attached to tanks were reviewed and evaluated for their structural integrity towithstand seismic-induced loads from RLGM.* Cabinets with anchorage located internally were opened and evaluated against RLGM." Thirteen (13) components were identified by the SRT during the plant walkdowns and six(6) HCLPF calculations were performed to envelope the thirteen components identified.
6.4 HCLPF Calculation ProcessESEL items not included in the previous IPEEE evaluations at H.B. Robinson Steam Electric Plantwere evaluated using the criteria in EPRI NP-6041.
Those evaluations included the following steps:* Performing seismic capability walkdowns for equipment not included in previous seismicwalkdowns (SQUG, IPEEE, or NTTF 2.3) to evaluate the equipment installed plantconditions.
Results of the walkdowns which are documented in the ESEP SEWS identified thirteen (13) components that require HCLPF calculation.
* Performing screening evaluations using the screening tables in EPRI NP-6041 as described in Section 6.2 and* Performing HCLPF calculations considering various failure modes that include structural failure modes (e.g. anchorage, load path etc.) and functional failure modes.Items based on similarity of model, function and anchorage were grouped together.
Based on EPRINP-6041-SL rule of similarity, a bounding anchorage evaluation was performed for equipment grouped together.
The calculations evaluate the demand and capacity of the equipment anchorage and derived a HCLPF capacity from the results of the anchorage evaluation.
The functional failuremode(s) are also evaluated.
Equipment that were identified as requiring a HCLPF capacity calculation in Reference 55 wereevaluated using the CDFM methodology as outlined in EPRI NP-6041-SL.
The HCLPF calculations are documented in Reference 10 and References 25 through 29. Thirteen components wereidentified by the SRT during walkdown and six HCLPF calculations were completed to envelope allthe components which include I&C and Hagan rack; Pressure Vessel; MCC; Battery Charger; andAuxiliary DC Panel.6.5 Functional Evaluations of RelaysBased on review of ESEL and associated single line diagrams, two relays (Under-Voltage AlarmRelay 27/MCC-A and Under-Voltage Alarm Relay 27/MCC-B) were identified.
: However, thesePage 33 of 46 Expedited Seismic Evaluation Process Reportrelays do not have lockout or seal-in mechanism (Reference
: 59) and are not required during FLEXimplementation.
27/MCC-A and 27/MCC-B are not designed to operate during and following DBEand BDBEE. Therefore, these relays were not included on the ESEL list. Extensive review of thesingle line diagrams did not identify any other relay or contactor that will be of concern.6.6 Tabulated ESEL HCLPF Values (Including Key Failure Modes)Tabulated ESEL HCLPF values are provided in Table 6-1. The following notes apply to theinformation in the table:* For items screened out using NP 6041 screening tables, the screening level can beprovided as >RLGM and the failure mode can be listed as "Screened",
(unless thecontrolling HCLPF value is governed by anchorage).
* For items where anchorage controls the HCLPF value, the HCLPF value is listed in thetable and the failure mode is noted as "anchorage."
Six HCLPF calculations were performed for items listed in the ESEL. Items that are based onsimilarity of equipment model, function, and anchorage are grouped together.
Based on EPRI NP-6041 SL rule of similarity, some items were grouped together and a bounding anchorage evaluation was performed.
The six HCLPF capacity evaluations are documented in Reference 10 andReferences 25 through 29. Each capacity calculation evaluates the demand and capacity of theequipment anchorage and derives a HCLPF capacity from the results of the anchorage evaluation.
The functional failure modes for each ESEL item were identified and documented in the calculation.
The functional and anchorage HCLPF capacity of items identified by the SRT for a seismic capacityevaluation is presented in Table 6-1.Page 34 of 46 Expedited Seismic Evaluation Process ReportTable 6-1: Functional and Anchorage HCLPF Capacity ResultsFunctional Anchorage/Structural Equipment Group Equipment HCLPF Achora ctyCapacityHCLPF CapacityCapacityInstrumentation and ControlPanels and Ck Main Control Board > 0.40g 0.414gPanels and RackRack -4Rack -11Hagan Racks > 0.40g 0.445gRack -12Rack -130.541gPressure Vessels Boron Injection Tank > 0.40gBattery Charger -ABattery Charger -AlBattery Chargers
> 0.40g 0.755gBattery Charger -BBattery Charger -B1> 0.40gMotor Control Centers MCC-A 0.250g> 0.40gMCC-B 0.406g> 0.40gAuxiliary DC Panel GD AUX-PNL-GD 0.596gPage 35 of 46 Expedited Seismic Evaluation Process Report7.0 Inaccessible Items7.1 Identification of ESEL items inaccessible for walkdowns All ESEL items were accessible with the exception of TE-423. This temperature element is ruggedand due to installation internal to the pipe, it is also protected from seismic interaction.
Anevaluation was performed based on available information and this item was determined to beacceptable by the SRT with no visual examination.
7.2 Planned Walkdown
/ Evaluation Schedule
/ Close OutNo ESEL item requires future walkdown.
Page 36 of 46 Expedited Seismic Evaluation Process Report8.0 ESEP Conclusions and Results8.1 Supporting Information The H.B. Robinson Steam Electric Plant has performed the ESEP as an interim action in responseto Reference 1, the NRC's 10 CFR 50.54(f) letter. It was performed using the methodologies inReference 2, the NRC endorsed guidance in EPRI 3002000704.
The ESEP provides an important demonstration of seismic margin and expedites plant safetyenhancements through evaluations and potential near-term modifications of plant equipment thatcan be relied upon to protect the reactor core following beyond design basis seismic events.The ESEP is part of the overall H.B. Robinson Steam Electric Plant response to the NRC's 50.54(f)letter. On March 12, 2014, NEI submitted to the NRC results of Reference 12, a study of seismiccore 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 thefleet of U.S. plants" based on the re-evaluated seismic hazards.
As such, the "current seismicdesign 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 (Reference
: 14) concluded thatthe "fleetwide seismic risk estimates are consistent with the approach and results used in the GI-199 safety/risk assessment."
The letter also stated that "As a result, the staff has confirmed thatthe conclusions reached in GI-199 safety/risk assessment remain valid and that the plants cancontinue to operate while additional evaluations are conducted."
An assessment of the change in seismic risk for H.B. Robinson Steam Electric Plant was includedin the fleet risk evaluation submitted in the March 12, 2014 NEI letter therefore, the conclusions inthe NRC's May 9 letter also apply to H.B. Robinson Steam Electric Plant.In addition, Reference 12, the March 12, 2014 NEI letter, provided an attached "Perspectives onthe Seismic Capacity of Operating Plants,"
which (1) assessed a number of qualitative reasons whythe design of SSCs inherently contain margin beyond their design level, (2) discussed industrial seismic experience databases of performance of industry facility components similar to nuclearSSCs, 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 hasbeen borne out for those plants that have actually experienced significant earthquakes.
Theseismic design process has inherent (and intentional) conservatisms which result in significant seismic margins within structures, systems and components (SSCs). These conservatisms arereflected in several key aspects of the seismic design process, including:
* Safety factors applied in design calculations
* Damping values used in dynamic analysis of SSCs* Bounding synthetic time histories for in-structure response spectra calculations
* Broadening criteria for in-structure response spectra" Response spectra enveloping criteria typically used in SSC analysis and testing applications
" Response spectra based frequency domain analysis rather than explicit time history basedtime domain analysis* Bounding requirements in codes and standards
* Use of minimum strength requirements of structural components (concrete and steel)* Bounding testing requirements, andPage 37 of 46 Expedited Seismic Evaluation Process Report0 Ductile behavior of the primary materials (that is, not crediting the additional capacity ofmaterials 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 theirfunctions 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 todemonstrate seismic margin through a review of a subset of the plant equipment that can be reliedupon to protect the reactor core following beyond design basis seismic events. In order tocomplete the ESEP in an expedited amount of time, the RLGM used for the ESEP evaluation is ascaled version of the plant's SSE rather than the actual GMRS. To more fully characterize the riskimpacts of the seismic ground motion represented by the GMRS on a plant specific basis, a moredetailed seismic risk assessment (SPRA or risk-based SMA) is to be performed in accordance withEPRI 1025287 (Reference 15). As identified in Reference 4, the H. B. Robinson Steam ElectricPlant Seismic Hazard and GMRS submittal, the H.B. Robinson Steam Electric Plant screens in fora seismic risk evaluation.
The complete seismic 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.
H.B. Robinson SteamElectric Plant will complete that evaluation in accordance with the schedule identified in Reference 13, NEI's letter dated April 9, 2013 and endorsed by the NRC in Reference 16, their May 7, 2013letter.8.2 Identification of Planned Modifications There are no planned future modifications for ESEP. The ESEP identified MCC-A as having aHCLPF capacity below the RLGM and not meeting the requirements of EPRI ESEP and NTTFRecommendation 2.1: Seismic.
MCC-A has since been modified in accordance with EPRI3002000704 to increase its seismic capacity to the RLGM. This was achieved by bracing thecabinet at the top. This modification eliminated flexible modes and resulted in reduced tensile loadapplied to the concrete expansion anchors.
The HCLPF capacity of MCC-A is now greater than0.4g.The ESEP determined that the HCLPF capacity of MCC-B was slightly above the RLGM and meetsthe requirements of the EPRI ESEP such that no modification was required.
: However, amodification similar to that discussed above for MCC-A was implemented in order to increase thecapacity of MCC-B anchorage and eliminate potential inertial forces at the top entry cable tray andconduit.Seismic margin above 2 x SSE was also added to a group of instrument racks (Hagan Racks) byvalidating the bolting integrity of the top braces (a relatively minor scope of work). The HCLPFcapacity of the Main Control Board is higher than the RLGM and meets the requirements of theEPRI ESEP. However, greater seismic capacity can be demonstrated by additional inspection ofplug welds that form part of the anchorage.
The additional inspection should confirm plug weldthickness and quality.
Table 6-1 shows the capacities of the thirteen ESEL items that requiredHCLPF calculation.
No additional modifications are planned for the H.B. Robinson Steam ElectricPlant related to ESEP.Page 38 of 46 Expedited Seismic Evaluation Process Report8.3 Modification Implementation ScheduleThe only ESEL item that required modification based on the seismic walkdown and HCLPFcapacity calculation was MCC-A. The modification has been developed and implemented asdiscussed in Section 8.2. The anchorage system for MCC-B is slightly different from that of MCC-Aand has higher structural capacity.
The HCLPF capacity of MCC-B slightly exceeds RLGMdemand. However, similar modification developed for MCC-A was also implemented on MCC-B.Although, not considered a modification, the Hagan Rack cabinets bolts were tightened to improvestructural capacity.
8.4 Summary of Planned ActionsThe H.B. Robinson Steam Electric Plant has no follow-up action or planned modification to supportthe ESEP. All of the items identified in the ESEL currently have a HCLPF capacity at or above theRLGM and do not require further evaluation.
The ESEL has been updated to consider newequipment that account for the changes in the FLEX strategy.
The new FLEX strategy wassubjected to critical path analysis and those items that fall under the ESEP guidelines have beenadded to the ESEL.Page 39 of 46 Expedited Seismic Evaluation Process Report9.0 References
: 1) NRC (E Leeds and M Johnson)
Letter to All Power Reactor Licensees et al.,"Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(f)
Regarding Recommendations 2.1, 2.3 and 9.3 of the Near-Term Task ForceReview of Insights from the Fukushima Dai-lchi Accident,"
March 12, 2012.2) Seismic Evaluation Guidance:
Augmented Approach for the Resolution ofFukushima Near-Term Task Force Recommendation 2.1 -Seismic.
EPRI, PaloAlto, CA: May 2013. 3002000704.
: 3) Updated H.B. Robinson Steam Electric Plant Overall Integrated Plan (OIP) inResponse to the March 12, 2012, Commission Order EA-12-049, August 2014.4) H.B. Robinson Steam Electric Plant Seismic Hazard and GMRS submittal, datedMarch 31, 2014.5) Nuclear Regulatory Commission, NUREG-1407, Procedural and Submittal Guidancefor the Individual Plant Examination of External Events (IPEEE) for Severe AccidentVulnerabilities, June 19916) Nuclear Regulatory Commission, Generic Letter No. 88-20 Supplement 4, Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities
-10CFR 50.54(f),
June 19917) A Methodology for Assessment of Nuclear Power Plant Seismic Margin, Rev. 1,August 1991, Electric Power Research Institute, Palo Alto, CA. EPRI NP 60418) Methodology for Developing Seismic Fragilities, August 1991, EPRI, Palo Alto, CA.1994, TR-103959
: 9) Appendix A to The H.B. Robinson Steam Electric Plant Unit No. 2 Individual PlantExamination for External Event Submittal:
Seismic IPEEE10) Calculation RNP-13-05-600-005,"High Confidence of Low Probability of Failure(HCLPF) for the H.B. Robinson Steam Electric Plant Motor Control Center A and B(MCC-A and MCC-B)}11) Nuclear Regulatory Commission, NUREG/CR-0098, Development of Criteria forSeismic Review of Selected Nuclear Power Plants, published May 197812) Nuclear Energy Institute (NEI), A. Pietrangelo, Letter to D. Skeen of the USNRC,"Seismic Core Damage Risk Estimates Using the Updated Seismic Hazards for theOperating Nuclear Plants in the Central and Eastern United States",
March 12, 201413) Nuclear Energy Institute (NEI), A. Pietrangelo, Letter to D. Skeen of the USNRC,"Proposed Path Forward for NTTF Recommendation 2.1: Seismic Reevaluations",
April 9, 201314) NRC (E Leeds) Letter to All Power Reactor Licensees et al., "Screening andPrioritization Results Regarding Information Pursuant to Title 10 of the Code ofFederal Regulations 50.54(F)
Regarding Seismic Hazard Re-Evaluations forRecommendation 2.1 of the Near-Term Task Force Review of Insights From theFukushima Dai-lchi Accident,"
May 9, 2014.15) Seismic Evaluation Guidance:
Screening, Prioritization and Implementation Details(SPID) for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1: Seismic.
EPRI, Palo Alto, CA: February 2013. 1025287.16) NRC (E Leeds) Letter to NEI (J Pollock),
"Electric Power Research Institute FinalDraft Report xxxxx, "Seismic Evaluation Guidance:
Augmented Approach for theResolution of Fukushima Near-Term Task Force Recommendation 2.1: Seismic,"
asan Acceptable Alternative to the March 12, 2012, Information Request for SeismicReevaluations,"
May 7, 201317) H.B. Robinson Steam Electric Plant NTTF 2.3 Seismic Walkdown Submittal datedFebruary 27, 2014.Page 40 of 46 Expedited Seismic Evaluation Process Report18) Carolina Power and Light Company (CP&L), Specification No CPL-HBR2-C-008,"Specification for Floor Response Spectra",
Revision 1, 1991.19) United States Nuclear Regulatory Commission, Regulatory Guide1.122,"Development of Floor Design Response Spectra for Seismic Design of Floor-Supported Equipment or Components",
Revision 1, February 1978.20) United States Nuclear Regulatory Commission NUREG-2115, Department of EnergyOffice of Nuclear Energy (DOE/NE)-0140, EPRI 1021097,"Central and EasternUnited States Seismic Source Characterization for Nuclear Facilities",
6 Volumes,2012.21) Electric Power Research Institute (EPRI), Final Report No 3002000717,"EPRI (2004, 2006) Ground Motion Model (GMM) Review Project",
June 2013.22) URS Energy and Construction Calculation RNP-13-05-600-001,"Review LevelGround Motion (RLGM) and In-Structure Response Spectra (ISRS) for H.B.Robinson Steam Electric Plant Unit 2, Revision 0, July 2013.23) Electric Power Research Institute (EPRI) Final Report No 1019200,"Seismic Fragility Applications Guide Update",
2009.24) EC 92103,"Fukushima NTTF Recommendation 2.1: Seismic Reevaluation"
: 25) Calculation RNP-13-05-600-006,"High Confidence of Low Probability of Failure(HCLPF) for the H.B. Robinson Steam Electric Plant Battery Chargers"
: 26) Calculation RNP-13-05-600-004,"High Confidence of Low Probability of Failure(HCLPF) for the H.B. Robinson Steam Electric Plant Boron Injection Tank"27) Calculation RNP-13-05-600-003,"High Confidence of Low Probability of Failure(HCLPF) for the H.B. Robinson Steam Electric Plant East Hagan Rack"28) Calculation RNP-13-05-600-007,"High Confidence of Low Probability of Failure(HCLPF) for the H.B. Robinson Steam Electric Plant Auxiliary DC Panel GD".29) Calculation RNP-13-05-600-002,"High Confidence of Low Probability of Failure(HCLPF) for the H.B. Robinson Steam Electric Plant Main Control Board".30) RNP/RA-13-0087, First Six Month Status Report (Order EA-12-049)
H. B. RobinsonSteam Electric Plant (RNP), Unit 2.31) RNP/RA-14-0008, Second Six Month Status Report (Order EA-12-049)
H. B.Robinson Steam Electric Plant (RNP), Unit 2.32) RNP/RA-14-0083, Third Six Month Status Report (Order EA-12-049)
H. B. RobinsonSteam Electric Plant (RNP), Unit 2.33) Engineering Change (EC) 88926, FLEX Strategies and Implementation Plan, Rev. 3.34) NEI 12-06, Diverse and Flexible Coping Strategies (FLEX) Implementation Guideline, Revision 0.35) EOP-ECA-0.0, Loss Of All AC Power, Revision 0.36) WCAP-1 760-P, Revision 1, Reactor Coolant System Response to Extended Loss ofAC Power Event for Westinghouse, Combustion Engineering, and Babcock &Wilcox NSSS Designs for Phase Boration, August 2012.37) PA-PSC-0965, PWROG Core Cooling Position Paper.38) EDMG-004, Steam Generators.
: 39) Calculation RNP-M/MECH-1712, Appendix R Mechanical Basis, Section 3.27,Cooldown Using MSIV Bypass Lines.40) EC 90617, Pre-Staged Diesel Generator Design To Power 125VDC -A Train and BTrain Battery Chargers For Fukushima Support (NTFF 4.2 -FLEX).41) FSG-10, Passive RCS Injection Isolation.
: 42) EC 95216, NTTF 2.1 Interim Action RCS Injection.
: 43) EC 90622, Standard Piping Connections For NTTF 4.2 (FLEX).44) EC 94745, Boric Acid and RCS Make Up Connections To The Safety Injection System NTTF 4.2 -Flexible Coping Strategies Page 41 of 46 Expedited Seismic Evaluation Process Report45) Calculation RNP-M/MECH-1877, RNP Extended Loss of AC (ELAP) PowerContainment Response46) FSG-12, Alternate Containment Cooling47) EC 90623, New Pipe Tee And Standard Connection For NTTF 4.2 (FLEX)48) EC 95266, Isolation Valves And Connection For AFW -FUKUSHIMA-Admin Rev49) EC 92103R0, Attachment Z03RO Mechanical Documents
: 50) EC 92103R0, Attachment Z05RO Electrical Documents
: 51) UFSAR, Section 02, Site Characteristics"
: 52) EC 92103R0, Attachment Z06RO53) EC 92103, Attachment Z18R054) EC 92103, Attachment Z09R055) EC 92103, Attachment Z1ORO56) EC 92103, Attachment Z01 RO57) EC 92103, Attachment Z16RO58) NRC Letter from NRC to Duke Energy and South Carolina Electric and GasCompany, Request for Additional Information Associated with Near-Term ForceRecommendation 2.1, Seismic Re-Evaluations Related to Southeastern CatalogChanges (TAC NOS MF3724, MF3736, MF 3738, and MF 3831), dated October 23,2014, ADAM Accession No ML14268A516.
: 59) H.B. Robinson Steam Electric Plant Control Wiring Diagram (CWD) B-190628, SH00955 and 00956.60) EC 92501, Attachment Z09Rl,"Additional Seismic Interim Actions Studies for theH.B. Robinson Steam Electric Plant".61) Letter Dated August 28, 2014 from EPRI to NRC Review of EPRI 1021097Earthquake Catalog for RIS Earthquakes in the Southeastern U.S. and Earthquakes in South Carolina Near Time of the 1886 Charleston Earthquake Sequence.
: 62) Response to Request for Additional Information Associated with Near-Term TaskForce Recommendation 2.1, Seismic Re-Evaluations Related to Southeastern Catalog Changes (TAC NOS. MF3724, MF3736, MF3737, MF3738, and MF3831),November 12, 2014Page 42 of 46 Expedited Seismic Evaluation Process ReportAttachment A -H.B. Robinson Steam Electric Plant ESELPage 43 of 46
 
Expedited Seismic Evaluation Process ReportAttachment B -H.B. Robinson Steam Electric Plant Unit 2 RCS Cooling Strategies Page 44 of 46 MNP72" MS HEADERMain Feed PumpsBLAplanREDStraGREStraBLUCapa-w'*-Lake Robinson______ IAJFW-TNK4
/ -'IDischarge/
ICanal na/ I -3 E AFW-TNK 9LEGEND I l I Portable  
...-.. 20k gli,CK -Installed  
\ o\ I.sel P pmnt equipment  
\ ." -/-FLEXLP LP\ " -tegles NN-EN -FLEX HP -.tegies rlrpLI IE -Additional I Iabilities I I E08Q1I I AFW-TNK-1 AF17I I I ,201 , g//iAFW-TN K-8EIWIS -017p I-F-TNK-20k g.1AW-TNK-620W-171AFW-TNK-3 20k galI-F.1F.Attachment BPage I of I/F-AFW-PMP-2 Portable1000 psi/300 gpmDiesel Pumpers/RCS COOLINGSEISMIC STRATEGIES vMDAFWPs Expedited Seismic Evaluation Process ReportAttachment C -H.B. Robinson Steam Electric Plant Unit 2 RCS Boration and MakeupStrategies Page 45 of 46 LEGENDBLACK -Installed plant equipment n~~i 7 LakeRobinson LEGENDMagenta -Mode 5/6Makeup (Reconfigure Check Valve Covers)LCV-11SA toC/CS HUT,RED -PortableBoration andMakeup StrategyUsing the ChargingSystemE--3 SI-iGREEN -PortableBoration and MakeupStrategy Using theSafety Injection System//IF ,RCS BORATIONand MAKEUPSI ACC nt Expedited Seismic Evaluation Process ReportAttachment D -H.B. Robinson Steam Electric Plant FLEX Flow PathPage 46 of 46}}

Revision as of 06:47, 1 July 2018

H. B. Robinson, Unit 2 - Expedited Seismic Evaluation Process Report (CEUS Sites), Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fuku
ML14365A105
Person / Time
Site: Robinson Duke Energy icon.png
Issue date: 12/17/2014
From: Glover R M
Duke Energy Corp, Duke Energy Progress
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
RNP-RA/14-0129
Download: ML14365A105 (56)
v • d • e


Text

~ENERGY,R. Michael GloverH. B. Robinson SteamElectric Plant Unit 2Site Vice President Duke Energy Progress3581 West Entrance RoadHartsville, SC 295500:843 857 1704F: 843 857 1319Mike. Glover~a duke-energy.com RN P-RA/14-0129 December 17, 201410 CFR 50.54(f)ATTN: Document Control DeskU. S. Nuclear Regulatory Commission Washington, DC 20555-0001 H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2DOCKET NO. 50-261 1 RENEWED LICENSE NO. DPR-23

Subject:

H. B. Robinson Steam Electric Plant, Unit No. 2 Expedited Seismic Evaluation ProcessReport (CEUS Sites), Response to NRC Request for Information Pursuant to 10 CFR50.54(f)

Regarding Recommendation 2.1 of the Near-Term Task Force Review ofInsights from the Fukushima Dai-ichi Accident

References:

1. NRC Letter, Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(0 Regarding Recommendations 2.1, 2.3, and 9.3, of the Near-Term Task Force Reviewof Insights from the Fukushima Dai-ichi
Accident, dated March 12, 20122. NEI Letter, Proposed Path Forward for NTTF Recommendation 2.1: Seismic Reevaluations, dated April 9, 2013, ADAMS Accession No. ML13101A379
3. NRC Letter, Electric Power Research Institute Report 3002000704, "Seismic Evaluation Guidance:

Augmented Approach for the Resolution of Fukushima Near-Term Task ForceRecommendation 2.1: Seismic,"

as an Acceptable Alternative to the March 12, 2012,Information Request for Seismic Reevaluations, dated May 7, 2013, ADAMS Accession No.ML13106A331 Ladies and Gentlemen:

On March 12, 2012, the Nuclear Regulatory Commission (NRC) issued a 50.54(f) letter to all powerreactor licensees and holders of construction permits in active or deferred status. Enclosure 1 ofReference 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 ofReference 1.In Reference 2, the Nuclear Energy Institute (NEI) requested NRC agreement to delay submittal ofthe final CEUS Seismic Hazard Evaluation and Screening Reports so that an update to the ElectricPower Research Institute (EPRI) ground motion attenuation model could be completed and used todevelop that information.

NEI proposed that descriptions of subsurface materials and properties and base case velocity profiles be submitted to the NRC by September 12, 2013, with theremaining seismic hazard and screening information submitted by March 31, 2014. NRC agreedwith that proposed path forward in Reference

3. Ac {

Serial: RNP-RA/14-0129 U. S. Nuclear Regulatory Commission Page 2Reference 1 requested that licensees provide interim evaluations and actions taken or planned toaddress the higher seismic hazard relative to the design basis, as appropriate, prior to completion of the risk evaluation.

In accordance with the NRC endorsed guidance in Reference 3, the attachedExpedited Seismic Evaluation Process Report for H. B. Robinson Steam Electric Plant, Unit No. 2provides the information described in Section 7 of Reference 3 in accordance with the scheduleidentified in Reference 2.This letter contains no new regulatory commitments.

If you have any questions or require additional information, please contact Richard Hightower,

Manager, Nuclear Regulatory Affairs at (843)-857-1329.

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

'2_0 1Sincerely, R.ý Mihe GýILoveR. Michael GloverSite Vice President RMG/shc

Enclosure:

Expedited Seismic Evaluation Process Report for H. B. Robinson Steam ElectricPlant, Unit No. 2cc: Ms. M. C. Barillas, NRC Project Manager, NRRMr. K. M. Ellis, NRC Senior Resident Inspector Mr. V. M. McCree, NRC Region II Administrator Expedited Seismic Evaluation Process ReportExpedited Seismic Evaluation Process ReportForH. B. Robinson Steam Electric Plant, Unit No. 2Page 3 of 46 Expedited Seismic Evaluation Process ReportEXPEDITED SEISMIC EVALUATION PROCESS REPORTTABLE OF CONTENT1.0 Purpose and Objective

.................................................................................

072.0 Brief Summary of the FLEX Seismic Implementation Strategies

..........................

073.0 Equipment Selection Process and ESEL .........................................................

133.1 Equipment Selection Process and ESEL ...............................................

133.1.1 ESEL Development

................................................................

143.1.2 Power Operated Valves ..........................................................

143.1.3 Pull Boxes ...........................................................................

143.1.4 Termination Cabinets

..............................................................

153.1.5 Critical Instrumentation Indicators

..............................................

153.1.6 Phase 2 and Phase 3 Piping Connections

..................................

153.2 Justification for Use of Equipment That is Not the Primary Means forFLEX Implementation

.....................................................................

154.0 Ground Motion Response Spectrum

..............................................................

164.1 Plot of GMRS Submitted by H.B. Robinson Steam Electric Plant ...............

164.2 Comparison to SSE ..........................................................................

195.0 Review Level Ground Motion (RLGM) ............................................................

235.1 Description of RLGM Selected

............................................................

235.2 Method to Estimate ISRS ..................................................................

266.0 Seismic Margin Evaluation Approach

............................................................

286.1 Summary of Methodologies Used ......................................................

286.2 HCLPF Screening Process ................................................................

296.3 Seismic Walkdown Approach

............................................................

306.3.1 W alkdown Approach

..............................................................

306.3.2 Application of Previous W alkdown Information

............................

326.3.3 Significant Walkdown Findings

.................................................

336.4 HCLPF Calculation Process ..............................................................

336.5 Functional Evaluation of Relays ..........................................................

336.6 Tabulated ESEL HCLPF Values (Including Key Failure Modes) .................

347.0 Inaccessible Items .....................................................................................

367.1 Identification of ESEL Items Inaccessible for Walkdown

..........................

367.2 Planned Walkdown/Evaluation Schedule/Close Out ................................

368.0 ESEP Conclusions and Results ....................................................................

378.1 Supporting Information

......................................................................

378.2 Identification of Planned Modifications

.................................................

388.3 Modification Implementation Schedule

...................................................

398.4 Summary of Planned Actions ............................................................

399.0 References

..............................................................................................

40Page 4 of 46 Expedited Seismic Evaluation Process ReportList of FiguresFigure 2.1: Mark-Up Showing Addition of FLEX Tee Connection (AFW-166) to SDAFW Discharge at AFW -121 ..........................................................

9Figure 2.2: Mark-Up Showing Alternate Mechanical FLEX Connection (AFW-165) insidethe MDAFW Room on Line 4-AFW-23 and Upstream of AFW-54 ..................

10Figure 4.1: Plot of 1 E-4 and 1 E-5 UHRS and GMRS at Control Point forthe H.B. Robinson Steam Electric Plant ...................................................

18Figure 4.2: Comparison of the GMRS, SSE, and Ground LevelResponse Spectrum from Time History ...................................................

22Figure 5-1: Plot of 5% Damping 2 x SSE, 2 x Ground Level Response

Spectrum, a nd G M R S .......................................................................................

2 6Figure 6.1: Comparison of the H.B. Robinson Steam Electric Plant IPEEE RLE, ESEPRLGM, SSE, and Ground Level (El. 226ft) Spectrum from Time History,and 2 x Ground Level (El. 226ft) Spectrum from Time History ...............

29List of TablesTable 4-1: GMRS and UHRS for the H.B. Robinson Steam Electric Plant ......................

17Table 4-2a: Original SSE Based on 0.2g Housner Spectrum for the H.B. RobinsonS team E lectric P lant ............................................................................

20Table 4-2b: Ground Level Response Spectrum Based on Time History for theH.B. Robinson Steam Electric Plant ........................................................

21Table 5-1: RLGM for H.B. Robinson Steam Electric Plant ..........................................

24Table 5-2: Ratio of G M RS to SSE .........................................................................

25Table 6-1: Functional and Anchorage HCLPF Capacity Results ..................................

35Attachments Attachment A -H.B. Robinson Steam Electric Plant ESELAttachment B -H.B. Robinson Steam Electric Plant Unit 2 RCS Cooling Strategies Attachment C -H.B. Robinson Steam Electric Plant Unit 2 RCS Boration and Makeup Strategies Attachment D -FLEX Flow PathPage 5 of 46 Expedited Seismic Evaluation Process ReportEXECUTIVE SUMMARYAn Expedited Seismic Evaluation Process has been completed for the H.B. Robinson SteamElectric Plant site based on endorsed guidance outlined in Electric Power Research Institute (EPRI)3002000704 (Reference 2). The work includes screening, equipment selection, development of theRLGM and in-structure

demands, evaluating seismic capacity of components and development ofHigh Confidence of Low Probability of Failure (HCLPF) calculations, and implementation ofnecessary plant modifications.

HCLPF calculations revealed that Motor Control Center (MCC-A)required modification for the beyond design basis ground motion. Modifications have beendeveloped and implemented for MCC-A and a similar cabinet, MCC-B. Seismic margin above 2XSSE was also added to a group of instrument racks (Hagan Racks) by validating the boltingintegrity of the top braces. All items in the ESEL have seismic capacity that exceeds the demand ofthe RLGM. The ESEL has been updated to consider new equipment in FLEX strategy as outlinedin the updated Overall Integrated Plan. The FLEX strategy was subjected to critical path analysisand all the items required under the ESEP guidelines are included in the ESEL list.Page 6 of 46 Expedited Seismic Evaluation Process Report1.0 Purpose and Objective Following the accident at the Fukushima Dai-ichi nuclear power plant resulting from the March 11,2011, Great Tohoku Earthquake and subsequent

tsunami, the Nuclear Regulatory Commission (NRC) established a Near Term Task Force (NTTF) to conduct a systematic review of NRCprocesses and regulations and to determine if the agency should make additional improvements toits regulatory system. The NTTF developed a set of recommendations intended to clarify andstrengthen the regulatory framework for protection against natural phenomena.

Subsequently, theNRC issued a 10 CFR 50.54(f) letter on March 12, 2012 (Reference 1), requesting information toassure 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 designbasis, further risk assessment may be required.

Assessment approaches acceptable to the staffinclude 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 H.B.Robinson Steam Electric Plant (RNP). The intent of the ESEP is to perform an interim action inresponse to the NRC's 50.54(f) letter (Reference

1) to demonstrate seismic margin through areview of a subset of the plant equipment that can be relied upon to protect the reactor corefollowing beyond design basis seismic events.The ESEP is implemented using the methodologies in the NRC endorsed guidance in EPRI3002000704, Seismic Evaluation Guidance:

Augmented Approach for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1: Seismic (Reference 2).The objective of this report is to provide summary information describing the ESEP evaluations andresults.

The level of detail provided in the report is intended to enable NRC to understand theinputs used, the evaluations performed, and the decisions made as a result of the interimevaluations.

2.0 Brief Summary of the FLEX Seismic Implementation Strategies The H.B. Robinson Steam Electric Plant FLEX strategies for Reactor Core Cooling and HeatRemoval, Reactor Inventory Control/Long-term Subcriticality, and Containment Function aresummarized below. The FLEX flow path is shown in Attachment D. The summary is derived fromthe H.B. Robinson Steam Electric Plant Overall Integrated Plan (OIP) in Response to the March 12,2012, Commission Order EA-12-049 (Reference 3), as supplemented by six-month updates(References 30, 31, and 32). Note that the H.B. Robinson Overall Integrated Plan (as amended in 6month updates) is based on Engineering Change (EC) 88926 (Reference 33).Reactor Core Cooling and Heat RemovalNEI 12-06, Diverse and Flexible Coping Strategies (FLEX) Implementation Guideline, Revision 0(Reference 34), requires that Auxiliary Feedwater (AFW) cooling be available to provide secondary makeup sufficient to maintain or restore Steam Generator (SG) level with installed equipment to thegreatest extent possible.

Beyond the use of installed equipment, steam generators must be able tobe depressurized in order to support makeup via portable pumps. Multiple and diverse connection points for the portable pumps must be provided and cooling water must be available indefinitely.

Refer to Attachment B (Reactor Coolant System Cooling Strategies) for depiction of the following discussion.

Page 7 of 46 Expedited Seismic Evaluation Process ReportThe H.B Robinson Steam Electric Plant FLEX strategies require that the AFW be in operation within 61 minutes of event initiation.

With the loss of AC power, a minimum of one steam supplyvalve (MS-V1-8A, MS-V1-8B, or MS-V1-8C) to Steam Driven Auxiliary Feedwater Pump (SDAFWP)and one AFW valve (AFW-V2-14A, AFW-V2-14B, AFW-V2-14C) to the steam generators must bemanually operated.

These required valves are all located in seismic Class 1 bay of the TurbineBuilding.

Additional portable backup for Steam Generator makeup is required per Section 3.2.2(13) of NEI12-06. The H.B. Robinson Steam Electric Plant has two strategies for portable backup. The firststrategy developed to satisfy this requirement is staging of two (2) intermediate pressure pumps(300 gpm at pressure of 1,000 psig) for all seismic events as described in detail below. The secondstrategy developed to satisfy the condition of Section 3.2.2(13) of NEI 12-06 is to store a Halepumper in a seismically robust Permanent FLEX Storage Building (PFSB). This strategy will involvethe use of the same primary and alternate connections described in the following paragraph, andwill require SG depressurization.

The two (2) pre-staged portable pumps (300 gpm at 1,000 psig) eliminate the need to depressurize the Steam Generators in the event the backup AFW feed capability is needed due to an AFWinterruption early in the ELAP transient as a result of seismic event. Either of the portable pumpscan take suction from a variety of plant sources (described below) and can be tied directly into theauxiliary feedwater system. Engineering Change 95266, Isolation Valves And Connection For AFW-FUKUSHIMA-Admin (Reference

48) was developed to add a FLEX tee connection (AFW-166) tothe SDAFWP discharge at AFW-121 (see Figure 2-1). Access to this primary connection is throughthe seismically qualified Turbine Building Class 1 bay. Engineering Change 90623, New Pipe TeeAnd Standard Connection For NTTF 4.2 (FLEX) (Reference
47) develops an alternate mechanical FLEX connection (AFW-165) inside the MDAFWP room on line 4-AFW-23 and upstream of AFW-54 (See Figure 2-2). EC90623 will be implemented during Refueling Outage, R0229. TheMDAFW room is housed in the seismic Class 1 Reactor Auxiliary Building (RAB).Page 8 of 46 Expedited Seismic Evaluation Process ReportFigure 2-1: Mark-Up Showing Addition of FLEX Tee Connection (AFW-166) to SDAFWDischarge at AFW-121Page 9 of 46 Expedited Seismic Evaluation Process ReportFigure 2-2: Mark-Up Showing Alternate Mechanical FLEX Connection (AFW-165) inside theMDAFWP Room on Line 4-AFW-23 and Upstream of AFW-54There are several sources for sustained cooling water supply. The primary source of AFWinventory is the seismically qualified condensate storage tank (CST) and its level instrumentation.

The CST is seismically robust and is the installed source of AFW to the SDAFWP. However, theCST inventory is not sufficient for indefinite coping (mission time is approximately 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> using theSDAFWP).

A secondary source of AFW inventory is the "Tank Farm" (portable pump) inside theprotected area that supplies the two pre-staged portable pumps (each with capacity of 300 gpmand 1,000 psig pressure as noted in the seismic strategy above). This source has a capacity ofapproximately 120,000 gallons and 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> of mission time using a pre-staged portable pump.The only other assured source of water is the Ultimate Heat Sink (Lake Robinson) which perrestrictions outlined in NEI 12-06 can only be accessed using portable equipment (assumes normalPage 10 of 46 Expedited Seismic Evaluation Process Reportaccess to the ultimate heat sink is lost). Given these limitations, one Phase 2/3 seismic strategy isto provide an indefinite supply of water to the CST and the SDAFWP by staging a portable dieselpumper at Lake Robinson with hoses routed to the CST. EC 90622, Standard Piping Connections For NTTF 4.2 (FLEX) (Reference

43) adds a FLEX connection at valve C-66 to provide anindefinite water supply to the CST. This can be accomplished during the initial CST/Tank Farmmission time of 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br />.The H.B. Robinson Steam Electric Plant has developed several options for the Steam Generator depressurization capability.

The Steam Generator Power Operated Relief Valves (PORVs) arenormally operated using the Instrument Air System or with backup Nitrogen System and alignedusing Attachment 2 of EOP-ECA-0.0 (Reference 35). However, neither the primary Instrument Airnor the backup Nitrogen System are seismically qualified.

Therefore, the primary Instrument Air andthe backup Nitrogen System cannot be relied upon during or after seismic events. The Main SteamSafety Valves are an alternate option to depressurize the Steam Generators but this option is notrecommended per the PA-PSC-0965, PWROG Core Cooling Position Paper (Reference

37) andWCAP-17601-P, Revision 1, Reactor Coolant System response to Extended Loss of AC PowerEvent for Westinghouse, Combustion Engineering, and Babcock & Wilcox NSSS Designs forPhase Boration, August 2012 (Reference 36), which state that remaining on the Main Steam SafetyValves for an extended period may lead to failure of the valve(s) which subsequently will causeexcessive and uncontrolled RCS cooldown.

Current strategy is to align portable nitrogen tanks to the Steam Generator PORV header usingAttachment 1 (Connecting Emergency Pressure Source to Operate SG PORVS) or Attachment 2(S/G Manual Depressurization) of RNP procedure EDMG-004, Steam Generators (Reference 38).In addition to the SG PORV capabilities recommended in Reference 37, the H.B. Robinson SteamElectric Plant has also developed a strategy to cooldown the RCS using the main steam lineisolation valve bypass lines. The strategy is detailed in Section 3.27 (Cooldown Using MSIV BypassLines) of calculation RNP-M/MECH-1712, Appendix R Mechanical Basis (Reference 39). Thiscapability results in a cooldown rate of 830/hr which bounds the recommended Westinghouse cooldown rate of 750/hr described in Reference 37.After initiation of depressurization, it is desirable to isolate the Safety Injection (SI) Accumulators inorder to prevent injection of nitrogen into the RCS which will impede natural circulation cooldown.

During an ELAP, power to the SI Accumulator isolation valves is lost. Although, the isolation valvescan be operated

manually, they are located inside the Containment Building and it is undesirable toperform this operation at this time due to personnel safety. The valves are powered by MCC 5 andMCC 6 and will be re-powered via Emergency Buses El and E2 with portable diesel generators staged in the seismic Class 1 Reactor Auxiliary Building (Drumming Room) for re-powering the Aand B Battery Chargers (see EC 90617 [Reference 40]). DB-50 Bus Feed Adapters can beinstalled in each of the Emergency Buses El and E2 and will be connected to the output of theDiesel Generators.

As part of the Phase 2 strategy, Steam Generator pressure will be maintained above the pressure corresponding to the SI Accumulator injection (240 psig) until the SIAccumulator isolation valves are closed using FLEX Support Guideline (FSG) 10, Passive RCSInjection Isolation (Reference 41).Reactor Inventory Control/Long-Term Subcriticality Refer to Attachment C (Reactor Coolant System Boration and Makeup Strategies) for a depiction ofthe following discussion.

There is no installed means of providing borated makeup following anELAP. The primary method of boration and inventory control is the use of portable high pressureand low volume pump directly connected to the Charging Lines or Safety Injection Headers fromthe Refueling Water Storage Tank (RWST) or a portable tanker containing borated water (seeEC95216, NTTF 2.1 Interim Action RCS Injection

[Reference 42]). The RWST is seismically Page 11 of 46 Expedited Seismic Evaluation Process Reportdesigned and will remain operational during and after a design basis seismic event. The makeupcapacity of the portable pump is 60 gpm at a pressure of 2,000 psig which is adequate for thebounding analysis in WCAP-1760-P (Reference 36). Phase 3 inventory control will beaccomplished using the same portable Phase 2 boration/makeup strategy.

Portable high pressurepumping and portable tanker capability will be stored in the PFSB to support this strategy.

EC 90622 (Reference

43) adds a FLEX connection to the exposed end downstream of the normallylocked closed drain valve (SI-837) located at the base of the RWST to access this borated water ifit available.

This portable strategy will deliver borated water to the RCS through valves CVC-121A/B (primary) or SI-888P/S (alternate).

Containment FunctionCalculation RNP-M/MECH-1877, RNP Extended Loss of AC Power (ELAP) Containment Response(Reference

45) was developed to determine the containment temperature and pressure responseassuming an ELAP and a trip from 100% reactor power at 100 days into the cycle. Results inReference 45 indicate that the Containment Building design limits for temperature and pressure willnot be challenged in the first 43 days following the event. This analysis assumes that: (1) no actionis taken to cool, spray, or vent the containment; and (2) low leakage RCP seals are installed.

Therefore, Phase 1 and 2 strategies are not required.

There is sufficient time and resources inPhase 3 to assemble a strategy using the National Safer Response Center (NSRC) pumpers andgenerators, prefabricated electrical connections, and prefabricated SW connections that will bestored in the PFSB. FSG-12, Alternate Containment Cooling (Reference

46) provides instructions for several existing strategies including external containment cooling which does not require use ofany plant system. These particular activities will be determined and directed by the Emergency Response Organization (Technical Support Center) based on the effects of the Beyond DesignBasis External Event (BDBEE) and the state of existing equipment.

Instrumentation Instrumentation channels that are powered by station batteries will be lost upon depletion of thebatteries.

FLEX strategies to improve battery coping occur by extending Phase 1. Phase 1 isextended by strategic load shedding followed by additional deep load shedding in the first hour ofthe event to extend battery coping times to 3.25 -3.75 hours8.680556e-4 days <br />0.0208 hours <br />1.240079e-4 weeks <br />2.85375e-5 months <br />. Phases 2 and 3 battery copingrequire portable diesel generators to power the vital battery chargers.

Two FLEX diesel generators will be mounted in their deployed positions near the battery chargers and within the ReactorAuxiliary Building.

Each generator will be sized to power two vital battery chargers, room air supplyand exhaust fans, and safety injection accumulator isolation valves. Electrical cables and pre-installed connectors will be routed from the FLEX diesel generators to the battery room for quickconnection of the cables to each of the battery chargers.

The primary strategy is to power the A andB vital battery chargers from one or both of the pre-staged FLEX generators.

The alternate is topower the A-1 and B-1 vital battery chargers from one or both of the pre-staged FLEX generators.

See Reference 40 for complete details of this strategy.

Page 12 of 46 Expedited Seismic Evaluation Process Report3.0 Equipment Selection Process and ESELThe selection of equipment for the Expedited Seismic Equipment List (ESEL) followed theguidelines of EPRI 3002000704.

The complete ESEL for H. B. Robinson Unit 2 is presented inAttachment A.3.1 Equipment Selection Process and ESELThe selection of equipment to be included on the ESEL was based on installed plant equipment credited in the FLEX strategies during Phase 1, 2 and 3 mitigation of a BDBEE as described in theH.B. Robinson Steam Electric Plant OIP (Reference

3) in response to the March 12, 2012Commission Order EA-12-049 as revised in References 30 through 32.The scope of "installed plant equipment" includes equipment relied upon for the FLEX strategies tosustain the critical functions of core cooling and containment integrity consistent with Reference 3and References 30 through 32. FLEX recovery actions are excluded from the ESEP scope perEPRI 3002000704.

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 andsubcriticality, and containment integrity functions.

Portable and pre-staged FLEX equipment (notpermanently installed) are excluded from the ESEL per EPRI 3002000704.

The ESEL component selection followed the EPRI guidance outlined in Section 3.2 of EPRI3002000704.

1. The scope of components is limited to that required to accomplish the core cooling andcontainment safety functions identified in Table 3-2 of EPRI 3002000704.

The instrumentation monitoring requirements for core cooling/containment safety functions are limited to thoseoutlined in the EPRI 3002000704

guidance, and are a subset of those outlined in the H.B.Robinson Steam Electric Plant OIP and as revised in the first , second and third six-month status reports.2. The scope of components is limited to installed plant equipment, and FLEX connections necessary to implement the H.B. Robinson Steam Electric Plant OIP (Reference
3) inresponse to the March 12, 2012 Commission Order EA-12-049 and as revised in References 30 through 32. and as described in Section 2.3. The scope of components assumes the credited FLEX connection modifications areimplemented, and are limited to those required to support a single FLEX success path (i.e.,either "Primary" or "Back-up/Alternate").
4. The "Primary" FLEX success path is to be specified.

Selection of the "Back-up/Alternate" FLEX success path must be justified.

5. Phase 3 coping strategies are included in the ESEP scope, whereas recoverystrategies are excluded.
6. Structures,
systems, and components excluded per the EPRI 3002000704 (Reference 2)guidance are:" Structures (e.g. Reactor Containment
Building, Reactor Auxiliary
Building, etc.)" Piping, cabling,
conduit, HVAC, and their supports.

" Manual valves and rupture disks." Power-operated valves not required to change state as part of the FLEX mitigation strategies.

Page 13 of 46 Expedited Seismic Evaluation Process Report* Nuclear steam supply system components (e.g. reactor pressure vessel and internals, reactor coolant pumps and seals, etc.)7. For cases in which neither train was specified as a primary or back-up strategy, then only onetrain component (generally

'A' train) is included in the ESEL.3.1.1 ESEL Development The ESEL was developed by reviewing the H.B. Robinson Steam Electric Plant OIP (Reference 3)and revisions in three subsequent six-month status reports to determine the major equipment involved in the FLEX strategies.

Further reviews of plant drawings (e.g., Process andInstrumentation Diagrams (P&IDs)(EC92103R0, Attachment Z03RO Mechanical Documents

[Reference 49]), and Electrical One Line Diagrams (EC92103R0, Attachment Z05R0 Electrical Documents

[Reference 50]) were performed to identify the boundaries of the flowpaths to be usedin the FLEX strategies and to identify specific components in the flowpaths needed to supportimplementation of the FLEX strategies.

Boundaries were established at an electrical or mechanical isolation device (e.g., isolation amplifier, valve, etc.) in branch circuits

/ branch lines off the definedstrategy electrical or fluid flowpath.

P&IDs were the primary reference documents used to identifymechanical components and instrumentation.

The flow paths used for FLEX strategies wereselected and specific components were identified using detailed equipment and instrument

drawings, piping isometrics, electrical schematics and one-line
drawings, system descriptions, design basis documents, etc., as necessary.

3.1.2 Power Operated ValvesPage 3-3 of EPRI 3002000704 notes that power operated valves not required to change state areexcluded from the ESEL. Page 3-2 also notes that "functional failure modes of electrical andmechanical portions of the installed Phase 1 equipment should be considered (e.g. AFW trips)."To address this concern, the following guidance is applied in the H.B. Robinson Steam ElectricPlant ESEL for functional failure modes associated with power operated valves:" 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, thevalves 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 3strategies, were not evaluated for spurious valve operation as the seismic event that causedthe ELAP has passed before the valves are re-powered.

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

Page 14 of 46 Expedited Seismic Evaluation Process Report3.1.4 Termination CabinetsTermination

cabinets, including cabinets necessary for FLEX Phase 2 and Phase 3 connections, provide consolidated locations for permanently connecting multiple cables. The termination cabinets and the internal connections provide a completely passive function;
however, the cabinetsare included in the ESEL to ensure industry knowledge on panel/anchorage failure vulnerabilities isaddressed.

3.1.5 Critical Instrumentation Indicators Critical indicators and recorders are typically physically located on panels/cabinets and are includedas separate components.

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

"... FLEXconnections necessary to implement the H.B. Robinson Steam Electric Plant OIP as described inSection 2." Item 3 in Section 3.1 also notes that "The scope of components assumes the creditedFLEX connection modifications are implemented, and are limited to those required to support asingle FLEX success path (i.e., either "Primary" or "Back-up/Alternate")."

Item 6 in Section 3.1 above goes on to explain that "Piping,

cabling, conduit, HVAC, and theirsupports" are excluded from the ESEL scope in accordance with EPRI 3002000704.

Therefore, piping and pipe supports associated with FLEX Phase 2 and Phase 3 connections areexcluded from the scope of the ESEP evaluation.

However, any active valves in FLEX Phase 2and Phase 3 connection flow path are included in the ESEL.3.2 Justification for Use of Equipment That Is Not The Primary Means for FLEXImplementation In accordance with EPRI 3002000704, the H.B. Robinson Steam Electric Plant used equipment that is the primary means of implementing FLEX strategy.

The complete ESEL for the H.B.Robinson Steam Electric Plant is presented in Attachment A.Page 15 of 46 Expedited Seismic Evaluation Process Report4.0 Ground Motion Response Spectrum (GMRS)4.1 Plot of GMRS Submitted by the H.B. Robinson Steam Electric PlantFollowing completion of the seismic hazard re-evaluation as requested in Reference 1, the NRC 10CFR 50.54(f) letter, a screening process is needed to determine if an interim seismic risk evaluation like the EPRI ESEP is required.

The screening GMRS was determined with control point seismichazard re-evaluation.

In accordance with the 50.54(f) letter and following the guidance in EPRIScreening, Prioritization, and Implementation Details (SPID) (Reference 15), Probabilistic SeismicHazard Analysis (PSHA) was performed using the 2012 CEUS Seismic Source Characterization forNuclear Facilities (Reference 20), a Regional Seismic Catalog Correction (Reference 61), andupdated EPRI Ground Motion Model (GMM) for the CEUS (Reference 21). Development of theH.B. Robinson Steam Electric Plant Ground Motion Response Spectra (GMRS) is documented inReferences 4 and 62. The GMRS and Uniform Hazard Response Spectra (UHRS) are tabulated inTable 4-1 and then compared in Figure 4-1 with the 5% damped horizontal SSE. Note thatadditional seismic hazard analysis and GMRS development is underway for H.B. Robinson SteamElectric Plant to support completion of the seismic probabilistic risk analysis.

In the analysis, newlyacquired geophysical testing results are being used to update the site response analysis.

Theresults of the screening evaluation discussed will not change as a result of the newly acquiredgeophysical testing.

These new geophysical testing data allow for a more accurate representation of seismic hazard and seismic probabilistic risk assessment by eliminating a significant source ofuncertainty.

Page 16 of 46 Expedited Seismic Evaluation Process ReportTable 4-1: GMRS and UHRS for the H.B. Robinson Steam Electric PlantFreq. (Hz) 10-4 UHRS (g) 105 UHRS (g) GMRS100 4.20E-01 9.17E-01 4.71E-0190 4.23E-01 9.31 E-O1 4-77E-0180 4.27E-01 9-48E-01 4.85E-0170 4.35E-01 9.73E-01 4.97E-0160 4-54E-01 1.02E+00 5.19E-0150 4.98E-01 1.11E+00 5.66E-0140 5.74E-01 1.25E+00 6.43E-0135 6.21 E-01 1.35E+00 6.95E-0130 6.63E-01 1.46E+00 7.50E-0125 7.23E-01 1.61E+00 8.21E-0120 7.92E-01 1.75E+00 8.97E-0115 8.09E-01 1.82E+00 9.27E-O112.5 8.35E-01 1.82E+00 9.36E-0110 8.52E-01 1.86E+00 9.55E-019 8.40E-01 1.84E+00 9.42E-018 8.58E-01 1.84E+00 9.49E-017 8.98E-01 1-92E+00 9.88E-016 8.87E-01 1.95E+00 9.99E-015 8.57E-01 1.87E+00 9.61E-014 8.40E-01 1.83E÷00 9.39E-013.5 7.71 E-01 1.76E+00 8.94E-013 6.79E-01 1.59E+00 8.04E-012.5 6.08E-01 1.38E+00 7.04E-O12 5.37E-01 1.30E+00 6.52E-011.5 3.97E-01 1.05E+00 5.20E-011.25 3.23E-01 8.58E-01 4.23E-011 2.26E-01 6.44E-01 3.13E-010.9 1.87E-01 5.52E-01 2.67E-010.8 1.56E-01 4.69E-01 2.26E-010.7 1.31E-01 3.95E-01 1.90E-010.6 1.10E-01 3.25E-01 1.57E-010.5 8.86E-02 2.51E-01 1.22E-010.4 7.09E-02 2.01E-01 9.79E-020.35 6.20E-02 1.76E-01 8.57E-020.3 5.32E-02 1.51 E-01 7.34E-020.25 4-43E-02 1.26E-01 6.12E-020.2 3.55E-02 1.00E-01 4.90E-020.15 2-66E-02 7.54E-02 3.67E-020.125 2.22E-02 6.28E-02 3.06E-020.1 1.77E-02 5.02E-02 2.45E-02Page 17 of 46 Expedited Seismic Evaluation Process ReportMean Soil UHRS and GMRS at Robinson2-52.-1E-5 UHRSL1.50.0.1 1 10 100Spectral frequency, HzFigure 4-1: Plot of 1 E-4 and 1 E-5 UHRS and GMRS at Control Point for the H.B. RobinsonSteam Electric Plant (5% Damped Response Spectra)Control point hazard curves were used to develop the UHRS and the GMRS. The methodology described in SPID (Reference

15) was used to compute site-specific control point hazard curves.The selection of control point elevation is based on recommendations in Section 2.4.2 of the SPID(Reference 15). The control point elevation for the H.B. Robinson Steam Electric Plant is at El. 226feet based on information in Sections 2.5 and 2.7 of the Updated Final Safety Analysis Report(Reference 51).Page 18 of 46 Expedited Seismic Evaluation Process Report4.2 Comparison to SSEOriginal design of the H.B. Robinson Steam Electric Plant was based on the 0.2g HousnerSpectrum.

Table 4-2a shows the spectral acceleration values as a function of frequency for the 5%damped horizontal SSE. As will be discussed in more detail in Section 5.2, original design in-structure response spectra was developed based on conservative time history.

The Ground LevelResponse Spectrum that results from this time history is reported in Table 4-2b.A comparison of the Ground Level Response

Spectrum, SSE, and GMRS is shown in Figure 4-2.As shown in Figure 4-2, in the 1 to 10 Hz frequency range of the response
spectrum, the GMRSexceeds the SSE and the Ground Level Response Spectrum.

The GMRS also exceeds the SSEand the Ground Level Response Spectrum at frequency values higher than 10 Hz.Page 19 of 46 Expedited Seismic Evaluation Process ReportTable 4-2a: Original SSE Based on 0.2g Housner Spectrum for the H.B. Robinson Steam ElectricPlant (5% Damping)Frequency SSE(Hz) (g)1.0 0.171.5 0.2302.0 0.2602.5 0.2903.0 0.33.5 0.3104.0 0.325.0 0.3056.0 0.2907.0 0.2658.0 0.2559.0 0.24010.0 0.2312.50 0.21015.0 0.220.0 0.225.0 0.230.0 0.233.0 0.235.0 0.2Page 20 of 46 Expedited Seismic Evaluation Process ReportTable 4-2b: Ground Level Response Spectrum Based on Time History for H.B. Robinson SteamElectric Plant (5% Damping)Frequency Ground Level(Hz) ResponseSpectrum (g)from TimeHistory1.0 0.3001.5 0.4552.0 0.4412.5 0.4173.0 0.4453.5 0.4684.0 0.4895.0 0.4556.0 0.4157.0 0.3808.0 0.3519.0 0.31610.0 0.28112.50 0.22115.0 0.23220.0 0.24625.0 0.25830.0 0.26733.0 0.27335.0 0.275Page 21 of 46 Expedited Seismic Evaluation Process Report1.200 --1.00 from :1m1Hi tory00.800 -~ -- -----o0.600 _0.40C,0.2000.0000.1 1.0 10.0 100.0Frequency (Hz)Figure 4-2: Comparison of GMRS, SSE and Ground Level Response Spectrum from Time HistoryPage 22 of 46 Expedited Seismic Evaluation Process Report5.0 Review Level Ground Motion (RLGM)5.1 Description of RLGM SelectedPlants for which the GMRS exceeds the SSE in the 1.0 to 10.0 Hz frequency range do not screenout of the ESEP and require further seismic evaluation.

The further seismic evaluation is performed to a Review Level Ground Motion which consists of a response spectrum above the SSE level. TheRLGM is defined as a response spectrum reflecting an earthquake level that is above the plant'sdesign basis SSE. The RLGM can be computed using one of the following criteria as described inReference 2:1. The RLGM can be derived by linearly scaling the SSE by the maximum ratio of the horizontal GMRS to the 5% damped SSE, between the 1 and 10 Hz frequency range, but not to exceed aratio greater than 2 times the SSE. The in-structure seismic motions corresponding to theRLGM would be derived using existing SSE-based In-Structure Response Spectra (ISRS)scaled with the same factor.2. Alternatively, licensees who have developed appropriate structural/soil-structure interaction (SSI) models capable of calculating ISRS based on site GMRS/Uniform Hazard ResponseSpectrum (UHRS) input may opt to use these ISRS in lieu of scaled SSE ISRS. In this case,the GMRS would represent the RLGM. EPRI 1025287 and the American Society ofMechanical Engineers/American Nuclear Society (ASME/ANS)

PRA Standard give guidanceon acceptable methods to compute both the GMRS and the associated ISRS. Section 4 ofReference 2 contains full description of this task.The RLGM for the H.B. Robinson Steam Electric Plant was developed in Reference 52 and inaccordance with the methodology and objectives in EPRI ESEP guidance Reference

2. The RLGMis the SSE multiplied by a factor of 2.0. Table 5-1 is the RLGM as a function of frequency andacceleration at 5% damping.

As discussed under Sections 4.2 and 5.2, original design in-structure response spectra were developed based on a conservative time history.

The Ground LevelResponse Spectrum that resulted from this time history is reported in Table 4-2b and Figure 5-2.For consistency between component screening and component evaluations, the Ground LevelResponse Spectrum was scaled by 2 to represent an effective RLGM for component screening.

Therefore, both screening and evaluation of ESEL items were conservatively based on 2 x GroundLevel Response Spectrum (see Figure 6-1 for plot of 2 x Ground Level Response Spectrum) instead of 2 x SSE.Page 23 of 46 Expedited Seismic Evaluation Process ReportTable 5-1: RLGM for H.B. Robinson Steam Electric PlantFrequency SSE RLGM(Hz) (g) (g)1.0 0.17 0.341.5 0.230 0.4602.0 0.260 0.5202.5 0.290 0.583.0 0.3 0.603.5 0.310 0.624.0 0.32 0.645.0 0.305 0.616.0 0.290 0.587.0 0.265 0.538.0 0.255 0.519.0 0.240 0.4810.0 0.23 0.4612.50 0.210 0.4215.0 0.2 0.420.0 0.2 0.425.0 0.2 0.430.0 0.2 0.433.0 0.2 0.435.0 0.2 0.4The ratio of the GMRS to the SSE is summarized in Table 5-2. The maximum ratio of the GMRS toSSE is 4.635 and this occurs at frequency of approximately 15Hz. In the frequency range of 1 to10Hz, the maximum ratio of the GMRS to SSE is 4.152. As limited in EPRI 3002000704, the RLGMis determined multiplying the SSE by a factor of 2.0.Page 24 of 46 Expedited Seismic Evaluation Process ReportTable 5-2: Ratio of GMRS to SSEFrequency GMRS SSE GMRSISSE(Hz) (g) (g)1.0 0.313 0.17 1.8411.5 0.520 0.230 2.2612.0 0.652 0.260 2.5082.5 0.704 0.290 2.4283.0 0.804 0.3 2.6803.5 0.894 0.310 2.8844.0 0.939 0.32 2.9345.0 0.961 0.305 3.1516.0 0.999 0.290 3.4457.0 0.988 0.265 3.7288.0 0.949 0.255 3.7229.0 0.942 0.240 3.92510.0 0.955 0.23 4.15212.50 0.936 0.210 4.45715.0 0.927 0.2 4.63520.0 0.897 0.2 4.48525.0 0.821 0.2 4.10530.0 0.750 0.2 3.75033.0 0.717 0.2 3.58535.0 0.695 0.2 3.475Page 25 of 46 Expedited Seismic Evaluation Process Report1.2001.000S0o.80000.6000.4000.200L20.0000.1 1.0 10.0 100.0Frequency (Hz)Figure 5-1: Plot of 5% Damping 2xSSE, 2 x Ground Level Response

Spectrum, and GMRS5.2 Method to Estimate ISRSThe seismic demand of the ESEL items/element mounted rigidly to the structure can be specified interms of the In-Structure Response Spectra (ISRS). For use in the ESEP, the in-structure seismicdemand for an element listed in the ESEL is defined by the ISRS scaled by the same factor used toobtain the RLGM from the SSE. The guidance under Section 4 of Reference 7 recommends broadening the peaks of the ISRS to account for the uncertainty in the civil structure frequency calculation.

The extent of broadening is suggested to be at least 15 percent of the frequency approaching and proceeding spectral peaks but can be increased beyond the minimumrecommendation based on the level of uncertainty associated with the structural model.The original design basis ISRS for the H.B. Robinson Steam Electric Plant were generated in 1970by Westinghouse Electric Corporation using mathematical building models developed by EbascoServices, Inc. The original ISRS or floor spectra generated by Westinghouse was limited in scopeand only considered the 0.20g design basis earthquake at damping ratio of 0.005 (0.5 percent).

These ISRS include conservatisms that result from conservative selection of the time history andexcessive bounding of design spectra.

Figure 4-2 shows plot of: (1) Ground Level ResponseSpectrum; (2) GMRS; and (3) SSE.Additional ISRS for other damping values were generated.

The task of generating the additional floor response spectra was complicated by lack of availability of time history data from the originalWestinghouse analysis.

Consequently, synthetic ground motion time history that generates ISRSPage 26 of 46 Expedited Seismic Evaluation Process Reportcomparable to the original Westinghouse floor spectra was used. The ISRS were generated byinputting the synthetic ground motion through the original Ebasco structural models. Scale factorsas a function of frequency were developed by comparing the spectra at the desired damping ratioagainst the 0.50 percent damping spectra.

The factors were then used to scale the originalWestinghouse 0.50 percent damped spectra to the desired damping ratio. The reconstituted ISRSat the various damping ratios have been incorporated into the H.B. Robinson Steam Electric Plant'sdesign basis ISRS documentation in Reference 18.The ISRS from Reference 18 were peak broadened in accordance with guidance in Regulatory Guide 1.122 (Reference 19). Since the ISRS in Reference 18 are already broadened, these spectraare scaled by a factor of 2.0 for ESEP.In summary, in-structure response spectra developed with the conservative Ground LevelResponse Spectrum were scaled by a factor of 2 for use in ESEP. Figure 5-1 shows plot of the 2 xSSE (RLGM), 2 x Ground Level Response

Spectrum, and GMRS.Page 27 of 46 Expedited Seismic Evaluation Process Report6.0 Seismic Margin Evaluation ApproachIt is necessary to demonstrate that ESEL items have sufficient seismic capacity to meet or exceedthe demand characterized by the RLGM and the corresponding scaled in-structure responsespectra.

The seismic capacity is characterized as the peak ground acceleration (PGA) for whichthere is a high confidence of a low probability of failure (HCLPF).

The PGA is associated with aspecific spectral shape, in this case the 5%-damped 2 x Ground Level Response Spectrum shape.The HCLPF capacity must be equal to or greater than the RLGM PGA. The criteria for seismiccapacity determination are given in Section 5 of EPRI 3002000704.

There are two basic approaches for developing HCLPF capacities:

1. Deterministic approach using the conservative deterministic failure margin (CDFM)methodology of EPRI NP-6041, A Methodology for Assessment of Nuclear Power PlantSeismic Margin (Revision
1) (Reference 7).2. Probabilistic approach using the fragility analysis methodology of EPRI TR-103959, Methodology for Developing Seismic Fragilities (Reference 8).6.1 Summary of Methodologies UsedThe H. B. Robinson Steam Electric Plant completed a seismic margin assessment (SMA) in 1993.The SMA is documented in Reference 9 and consisted of screening, walkdowns by SRT, andHCLPF anchorage calculations.

The screening and walkdowns used the screening tables fromChapter 2 of EPRI NP-6041 (Reference

7) for peak spectral acceleration less than 0.8g. Thewalkdowns were conducted by engineers trained in EPRI NP 6041 (the engineers attended theEPRI SMA Add-On course in addition to the SQUG Walkdown Screening and Seismic Evaluation Training Course),

and were documented on Screening Evaluation Work Sheets from EPRI NP-6041. Anchorage capacity calculations used the CDFM criteria from EPRI NP-6041.

Seismicdemand was the IPEEE Review Level Earthquake (RLE) for SMA (mean NUREG/CR-0098

[Reference 11] ground response spectrum anchored to 0.3g PGA).Figure 6-1 shows the mean NUREG/CR-0098 ground response spectrum used as the IPEEE RLEcompared to the 2 x Ground Level Response Spectrum.

The figure shows that the ESEP inputmotion enveloped the IPEEE RLE at all frequencies except between 10 Hz and 15 Hz where theIPEEE RLE slightly exceed the ESEP input motion. The frequency of interest for ESEL items isbetween 1 Hz and 10Hz.The ESEP methodology included screening and extensive walkdown by the Seismic Review Team(SRT), and HCLPF calculations to evaluate structural capacity of the ESEL items against theRLGM. Function evaluation of relays was also performed.

The walkdowns were documented onScreening Evaluation Worksheets (SEWS) from EPRI NP-6041.

Based on outcome of the seismicwalkdown and documentation in SEWS, six (6) HCLPF calculations were performed to envelopethe thirteen (13) ESEL items identified during the walkdowns.

Page 28 of 46 Expedited Seismic Evaluation Process Report1.2001.000-2 X Ground LevelTim History-.SS-IPEEE RILE-Ground Level SpaHistory,-ý2 XSSE1(RLGM Sfromn04..CO)0.8000.6000.4000.2000.0000.11.010.0100.0Frequency (Hz)Figure 6-1.Comparison of the H.B. Robinson Steam Electric Plant IPEEE RLE, ESEPRLGM, SSE, Ground Level (El. 226ft) Spectrum from Time History, and 2 xGround Level (El. 226ft) Spectrum from Time History6.2 HCLPF Screening ProcessThe HCLPF screening and calculations were based on 2 x Ground Level Response Spectrum peakground acceleration.

Screening tables in EPRI NP-6041 (Reference

7) are based on peak spectralacceleration of < 0.8g, 0.8g to 1.2g, and > 1.2g. Since 2 x Ground Level Response Spectrum peakground acceleration is 0.978g, screening of ESEL items was based on the 0.8g to 1.2g rangecriteria.

The screening guidelines were supplemented by Appendix A of EPRI NP-6041 SL whichprovides the basis for the seismic capacity screening guidelines.

Anchorage capacity calculations were based on 2 x Ground Level Response Spectrum.

Equipment for which the screening caveats were met and for which the anchorage capacity exceeded 2 xGround Level Response Spectrum seismic demand were screened out from ESEP seismiccapacity determination.

Page 29 of 46 Expedited Seismic Evaluation Process Report6.3 Seismic Walkdown Approach6.3.1 Walkdown ApproachWalkdowns were performed in accordance with the criteria provided in Section 5 of EPRI3002000704 (Reference 2), which refers to EPRI NP-6041 (Reference

7) for the Seismic MarginAssessment process.

Pages 2-26 through 2-30 of EPRI NP-6041 describe the seismic walkdowncriteria, including the following key criteria.

"The SRT [Seismic Review Team] should "walk by" 100% of all components which are reasonably accessible and in non-radioactive or low radioactive environments.

Seismic capability assessment of components which are inaccessible, in high-radioactive environments, or possibly withincontaminated containment, will have to rely more on alternate means such as photographic inspection, more reliance on seismic reanalysis, and possibly, smaller inspection teams and morehurried 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 orpanels for detailed inspection of all components.

This walkdown is not intended to be a QA or QCreview or a review of the adequacy of the component at the SSE levelIf the SRT has a reasonable basis for assuming that the group of components are similar and aresimilarly

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 onecomponent or each type which is selected should be thoroughly inspected which probably doesmean de-energizing and opening cabinets or panels for this very limited sample. Generally, aspare representative component can be found so as to enable the inspection to be performed whilethe 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 ofcomponents the SRT should look closely at the first items and compare the field configurations withthe construction drawings and/or specifications.

If a one-to-one correspondence is found, thensubsequent items do not have to be inspected in as great a detail. Ultimately the walkdownbecomes 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 inparallel.

If serious exceptions to the drawings or questionable construction practices are foundthen 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 thatshown on drawings or prescribed in criteria for that component, potential SI [Seismic Interaction 1]problems, situations that are at odds with the team members' past experience, and any other areasof serious seismic concern.

If any such concerns

surface, then the limited sample size of onecomponent of each type for thorough inspection will have to be increased.

The increase in samplesize 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'EPRI 3002000704 page 5-4 limits the ESEP seismic interaction reviews to "nearby block walls" and "pipingattached to tanks" which are reviewed "to address the possibility of failures due to differential displacements."

Otherpotential seismic interaction evaluations are "deferred to the full seismic risk evaluations performed in accordance with EPRI 1025287'Page 30 of 46 Expedited Seismic Evaluation Process Reportones who are responsible for the seismic adequacy of all elements which they screen from themargin review. Appendix D gives guidance for sampling selection."

As part of the ESEP, demonstration that the components listed in the ESEL have a HCLPFcapacity that exceeds the effective RLGM (2 x Ground Level Response Spectrum) verifiesadequate seismic ruggedness.

Section 5 of EPRI ESEP guidance specifies that the methodology inEPRI NP-6041 SL may be used for the development of the HCLPF capacity.

The major steps inReference 7 include pre-screening, walkdowns, and the CDFM HCLPF calculations.

In order to ensure efficiency while performing the walkdowns and during seismic capacityevaluations, each of the items listed in the ESEL were subjected to pre-screening.

The initial pre-screening effort consisted of data collection in the form of drawings, calculations, specifications, and vendor documents for each item in the ESEL. After identification of documentation for aspecific item, the pre-screening process followed the general seismic capacity screening guidelines presented in Reference 7 for civil structures, equipment, and subsystems to be considered screened out from further review. The caveats and footnoted exceptions and restrictions listed arefollowed.

For the purpose of completing the ESEP for the H. B. Robinson Steam Electric Plant, only Table 2-4 of Reference 7 is relevant for applying seismic screening criteria for plant equipment listed in theESEL. In addition to using the screening criteria in Reference 7 during plant walkdown, the SRTalso exercised their collective experience and judgment while using the criteria for specificcomponent.

The screening criteria can be used for equipment that is approximately 40ft abovegrade or lower. EPRI Report No 1019200 (Reference

23) provides guidance on screening criteriafor equipment that is greater than 40ft above grade. Screening criteria in Reference 7 do notinclude considerations for anchorage.

Therefore, structural integrity of anchorage was evaluated separately.

Some simple cases were documented on the SEWS form.Plant walkdowns were performed for items in the ESEL using guidance in Reference

7. Information extracted from existing documentation such as equipment
location, seismic input elevation, relevant drawing details, and previous seismic capacity calculations were recorded on the ESEPSEWS and used during the walkdowns.

In accordance with the ESEP guidance, the SEWS thatwere used in the ESEP walkdowns were consistent with content and format of the SEWSpresented in Appendix F of EPRI NP-6041 SL.A major part of the ESEP walkdowns was the investigation of equipment anchorages.

Therefore, cabinets with anchorages located internally were opened. Furthermore, the ESEP guidance statesthat components that are anchored to sub-structural elements that may not have the same capacityas the main structural system (e.g. block walls, frames, stanchions etc.) should also be reviewed.

Nearby block walls were identified and evaluated as necessary.

Piping attached to tanks were alsoreviewed.

Other potential seismic interaction evaluations were deferred to a full Seismic RiskEvaluation (SRE) as discussed in the SPID References 14 and 15, and were not addressed in theESEP walkdowns.

Walkdown assessment for the H.B. Robinson Steam Electric Plant ESEL items were completed bythe SRT between August 2013 and February 2014. Some of the components were previously walked down during the IPEEE, USI A-46, or NTTF 2.3: Seismic and relevant information such asthe equipment

location, seismic input elevation, drawing details and previous seismic calculations were recorded on the ESEP SEWS. Previous walkdowns were credited since they were performed by qualified Seismic Review Team. A walk-by of these components was performed anddocumented.

The objective of the walk-by is to confirm and verify that the components and theiranchorage have not degraded since the previous walkdown.

Items included in the ESEL that have not been previously walked down and evaluated, wereautomatically included for a detailed walkdown.

Page 31 of 46 Expedited Seismic Evaluation Process ReportThe SRT was comprised of at least two SQUG trained engineers and often included two additional structural engineers (Reference 57). The results of the walkdowns were documented on the SEWSfor each item. The completed SEWS and pictures taken during the walkdowns for the ESEL aredocumented in Reference

55. Follow-up inspections and walkdowns were completed whereadditional information was necessary.

6.3.2 Application of Previous Walkdown Information Previous seismic walkdowns from IPEEE and USI A-46 were used to support the ESEP seismicevaluations.

Some of the components and items on the ESEL were included in the NTTF 2.3seismic walkdowns (Reference 17). Those walkdowns were well documented and recent enoughthat they did not need to be repeated for the ESEP.Several ESEL items were previously walked down during the H.B. Robinson Steam Electric PlantSeismic IPEEE program.

Those walkdown results were reviewed and the following steps weretaken 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 relatedto 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.Page 32 of 46 Expedited Seismic Evaluation Process Report6.3.3 Significant Walkdown FindingsConsistent with guidance from NP-6041, no significant outliers or anchorage concerns (exceptMCC-A) were identified during the H.B. Robinson Steam Electric Plant seismic walkdowns.

Thefollowing findings were noted during the walkdowns.

  • Nearby block walls were identified in the proximity of ESEL item. These block walls wereassessed for their structural adequacy to withstand the seismic loads resulting from theRLGM. There is no case where the block wall represented the HCLPF failure mode for anESEL item.* Piping attached to tanks were reviewed and evaluated for their structural integrity towithstand seismic-induced loads from RLGM.* Cabinets with anchorage located internally were opened and evaluated against RLGM." Thirteen (13) components were identified by the SRT during the plant walkdowns and six(6) HCLPF calculations were performed to envelope the thirteen components identified.

6.4 HCLPF Calculation ProcessESEL items not included in the previous IPEEE evaluations at H.B. Robinson Steam Electric Plantwere evaluated using the criteria in EPRI NP-6041.

Those evaluations included the following steps:* Performing seismic capability walkdowns for equipment not included in previous seismicwalkdowns (SQUG, IPEEE, or NTTF 2.3) to evaluate the equipment installed plantconditions.

Results of the walkdowns which are documented in the ESEP SEWS identified thirteen (13) components that require HCLPF calculation.

  • Performing screening evaluations using the screening tables in EPRI NP-6041 as described in Section 6.2 and* Performing HCLPF calculations considering various failure modes that include structural failure modes (e.g. anchorage, load path etc.) and functional failure modes.Items based on similarity of model, function and anchorage were grouped together.

Based on EPRINP-6041-SL rule of similarity, a bounding anchorage evaluation was performed for equipment grouped together.

The calculations evaluate the demand and capacity of the equipment anchorage and derived a HCLPF capacity from the results of the anchorage evaluation.

The functional failuremode(s) are also evaluated.

Equipment that were identified as requiring a HCLPF capacity calculation in Reference 55 wereevaluated using the CDFM methodology as outlined in EPRI NP-6041-SL.

The HCLPF calculations are documented in Reference 10 and References 25 through 29. Thirteen components wereidentified by the SRT during walkdown and six HCLPF calculations were completed to envelope allthe components which include I&C and Hagan rack; Pressure Vessel; MCC; Battery Charger; andAuxiliary DC Panel.6.5 Functional Evaluations of RelaysBased on review of ESEL and associated single line diagrams, two relays (Under-Voltage AlarmRelay 27/MCC-A and Under-Voltage Alarm Relay 27/MCC-B) were identified.

However, thesePage 33 of 46 Expedited Seismic Evaluation Process Reportrelays do not have lockout or seal-in mechanism (Reference
59) and are not required during FLEXimplementation.

27/MCC-A and 27/MCC-B are not designed to operate during and following DBEand BDBEE. Therefore, these relays were not included on the ESEL list. Extensive review of thesingle line diagrams did not identify any other relay or contactor that will be of concern.6.6 Tabulated ESEL HCLPF Values (Including Key Failure Modes)Tabulated ESEL HCLPF values are provided in Table 6-1. The following notes apply to theinformation in the table:* For items screened out using NP 6041 screening tables, the screening level can beprovided as >RLGM and the failure mode can be listed as "Screened",

(unless thecontrolling HCLPF value is governed by anchorage).

  • For items where anchorage controls the HCLPF value, the HCLPF value is listed in thetable and the failure mode is noted as "anchorage."

Six HCLPF calculations were performed for items listed in the ESEL. Items that are based onsimilarity of equipment model, function, and anchorage are grouped together.

Based on EPRI NP-6041 SL rule of similarity, some items were grouped together and a bounding anchorage evaluation was performed.

The six HCLPF capacity evaluations are documented in Reference 10 andReferences 25 through 29. Each capacity calculation evaluates the demand and capacity of theequipment anchorage and derives a HCLPF capacity from the results of the anchorage evaluation.

The functional failure modes for each ESEL item were identified and documented in the calculation.

The functional and anchorage HCLPF capacity of items identified by the SRT for a seismic capacityevaluation is presented in Table 6-1.Page 34 of 46 Expedited Seismic Evaluation Process ReportTable 6-1: Functional and Anchorage HCLPF Capacity ResultsFunctional Anchorage/Structural Equipment Group Equipment HCLPF Achora ctyCapacityHCLPF CapacityCapacityInstrumentation and ControlPanels and Ck Main Control Board > 0.40g 0.414gPanels and RackRack -4Rack -11Hagan Racks > 0.40g 0.445gRack -12Rack -130.541gPressure Vessels Boron Injection Tank > 0.40gBattery Charger -ABattery Charger -AlBattery Chargers

> 0.40g 0.755gBattery Charger -BBattery Charger -B1> 0.40gMotor Control Centers MCC-A 0.250g> 0.40gMCC-B 0.406g> 0.40gAuxiliary DC Panel GD AUX-PNL-GD 0.596gPage 35 of 46 Expedited Seismic Evaluation Process Report7.0 Inaccessible Items7.1 Identification of ESEL items inaccessible for walkdowns All ESEL items were accessible with the exception of TE-423. This temperature element is ruggedand due to installation internal to the pipe, it is also protected from seismic interaction.

Anevaluation was performed based on available information and this item was determined to beacceptable by the SRT with no visual examination.

7.2 Planned Walkdown

/ Evaluation Schedule

/ Close OutNo ESEL item requires future walkdown.

Page 36 of 46 Expedited Seismic Evaluation Process Report8.0 ESEP Conclusions and Results8.1 Supporting Information The H.B. Robinson Steam Electric Plant has performed the ESEP as an interim action in responseto Reference 1, the NRC's 10 CFR 50.54(f) letter. It was performed using the methodologies inReference 2, the NRC endorsed guidance in EPRI 3002000704.

The ESEP provides an important demonstration of seismic margin and expedites plant safetyenhancements through evaluations and potential near-term modifications of plant equipment thatcan be relied upon to protect the reactor core following beyond design basis seismic events.The ESEP is part of the overall H.B. Robinson Steam Electric Plant response to the NRC's 50.54(f)letter. On March 12, 2014, NEI submitted to the NRC results of Reference 12, a study of seismiccore 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 thefleet of U.S. plants" based on the re-evaluated seismic hazards.

As such, the "current seismicdesign 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 (Reference

14) concluded thatthe "fleetwide seismic risk estimates are consistent with the approach and results used in the GI-199 safety/risk assessment."

The letter also stated that "As a result, the staff has confirmed thatthe conclusions reached in GI-199 safety/risk assessment remain valid and that the plants cancontinue to operate while additional evaluations are conducted."

An assessment of the change in seismic risk for H.B. Robinson Steam Electric Plant was includedin the fleet risk evaluation submitted in the March 12, 2014 NEI letter therefore, the conclusions inthe NRC's May 9 letter also apply to H.B. Robinson Steam Electric Plant.In addition, Reference 12, the March 12, 2014 NEI letter, provided an attached "Perspectives onthe Seismic Capacity of Operating Plants,"

which (1) assessed a number of qualitative reasons whythe design of SSCs inherently contain margin beyond their design level, (2) discussed industrial seismic experience databases of performance of industry facility components similar to nuclearSSCs, 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 hasbeen borne out for those plants that have actually experienced significant earthquakes.

Theseismic design process has inherent (and intentional) conservatisms which result in significant seismic margins within structures, systems and components (SSCs). These conservatisms arereflected in several key aspects of the seismic design process, including:

  • Safety factors applied in design calculations
  • Damping values used in dynamic analysis of SSCs* Bounding synthetic time histories for in-structure response spectra calculations
  • Broadening criteria for in-structure response spectra" Response spectra enveloping criteria typically used in SSC analysis and testing applications

" Response spectra based frequency domain analysis rather than explicit time history basedtime domain analysis* Bounding requirements in codes and standards

  • Use of minimum strength requirements of structural components (concrete and steel)* Bounding testing requirements, andPage 37 of 46 Expedited Seismic Evaluation Process Report0 Ductile behavior of the primary materials (that is, not crediting the additional capacity ofmaterials 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 theirfunctions 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 todemonstrate seismic margin through a review of a subset of the plant equipment that can be reliedupon to protect the reactor core following beyond design basis seismic events. In order tocomplete the ESEP in an expedited amount of time, the RLGM used for the ESEP evaluation is ascaled version of the plant's SSE rather than the actual GMRS. To more fully characterize the riskimpacts of the seismic ground motion represented by the GMRS on a plant specific basis, a moredetailed seismic risk assessment (SPRA or risk-based SMA) is to be performed in accordance withEPRI 1025287 (Reference 15). As identified in Reference 4, the H. B. Robinson Steam ElectricPlant Seismic Hazard and GMRS submittal, the H.B. Robinson Steam Electric Plant screens in fora seismic risk evaluation.

The complete seismic 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.

H.B. Robinson SteamElectric Plant will complete that evaluation in accordance with the schedule identified in Reference 13, NEI's letter dated April 9, 2013 and endorsed by the NRC in Reference 16, their May 7, 2013letter.8.2 Identification of Planned Modifications There are no planned future modifications for ESEP. The ESEP identified MCC-A as having aHCLPF capacity below the RLGM and not meeting the requirements of EPRI ESEP and NTTFRecommendation 2.1: Seismic.

MCC-A has since been modified in accordance with EPRI3002000704 to increase its seismic capacity to the RLGM. This was achieved by bracing thecabinet at the top. This modification eliminated flexible modes and resulted in reduced tensile loadapplied to the concrete expansion anchors.

The HCLPF capacity of MCC-A is now greater than0.4g.The ESEP determined that the HCLPF capacity of MCC-B was slightly above the RLGM and meetsthe requirements of the EPRI ESEP such that no modification was required.

However, amodification similar to that discussed above for MCC-A was implemented in order to increase thecapacity of MCC-B anchorage and eliminate potential inertial forces at the top entry cable tray andconduit.Seismic margin above 2 x SSE was also added to a group of instrument racks (Hagan Racks) byvalidating the bolting integrity of the top braces (a relatively minor scope of work). The HCLPFcapacity of the Main Control Board is higher than the RLGM and meets the requirements of theEPRI ESEP. However, greater seismic capacity can be demonstrated by additional inspection ofplug welds that form part of the anchorage.

The additional inspection should confirm plug weldthickness and quality.

Table 6-1 shows the capacities of the thirteen ESEL items that requiredHCLPF calculation.

No additional modifications are planned for the H.B. Robinson Steam ElectricPlant related to ESEP.Page 38 of 46 Expedited Seismic Evaluation Process Report8.3 Modification Implementation ScheduleThe only ESEL item that required modification based on the seismic walkdown and HCLPFcapacity calculation was MCC-A. The modification has been developed and implemented asdiscussed in Section 8.2. The anchorage system for MCC-B is slightly different from that of MCC-Aand has higher structural capacity.

The HCLPF capacity of MCC-B slightly exceeds RLGMdemand. However, similar modification developed for MCC-A was also implemented on MCC-B.Although, not considered a modification, the Hagan Rack cabinets bolts were tightened to improvestructural capacity.

8.4 Summary of Planned ActionsThe H.B. Robinson Steam Electric Plant has no follow-up action or planned modification to supportthe ESEP. All of the items identified in the ESEL currently have a HCLPF capacity at or above theRLGM and do not require further evaluation.

The ESEL has been updated to consider newequipment that account for the changes in the FLEX strategy.

The new FLEX strategy wassubjected to critical path analysis and those items that fall under the ESEP guidelines have beenadded to the ESEL.Page 39 of 46 Expedited Seismic Evaluation Process Report9.0 References

1) NRC (E Leeds and M Johnson)

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

Regarding Recommendations 2.1, 2.3 and 9.3 of the Near-Term Task ForceReview of Insights from the Fukushima Dai-lchi Accident,"

March 12, 2012.2) Seismic Evaluation Guidance:

Augmented Approach for the Resolution ofFukushima Near-Term Task Force Recommendation 2.1 -Seismic.

EPRI, PaloAlto, CA: May 2013. 3002000704.

3) Updated H.B. Robinson Steam Electric Plant Overall Integrated Plan (OIP) inResponse to the March 12, 2012, Commission Order EA-12-049, August 2014.4) H.B. Robinson Steam Electric Plant Seismic Hazard and GMRS submittal, datedMarch 31, 2014.5) Nuclear Regulatory Commission, NUREG-1407, Procedural and Submittal Guidancefor the Individual Plant Examination of External Events (IPEEE) for Severe AccidentVulnerabilities, June 19916) Nuclear Regulatory Commission, Generic Letter No. 88-20 Supplement 4, Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities

-10CFR 50.54(f),

June 19917) A Methodology for Assessment of Nuclear Power Plant Seismic Margin, Rev. 1,August 1991, Electric Power Research Institute, Palo Alto, CA. EPRI NP 60418) Methodology for Developing Seismic Fragilities, August 1991, EPRI, Palo Alto, CA.1994, TR-103959

9) Appendix A to The H.B. Robinson Steam Electric Plant Unit No. 2 Individual PlantExamination for External Event Submittal:

Seismic IPEEE10) Calculation RNP-13-05-600-005,"High Confidence of Low Probability of Failure(HCLPF) for the H.B. Robinson Steam Electric Plant Motor Control Center A and B(MCC-A and MCC-B)}11) Nuclear Regulatory Commission, NUREG/CR-0098, Development of Criteria forSeismic Review of Selected Nuclear Power Plants, published May 197812) Nuclear Energy Institute (NEI), A. Pietrangelo, Letter to D. Skeen of the USNRC,"Seismic Core Damage Risk Estimates Using the Updated Seismic Hazards for theOperating Nuclear Plants in the Central and Eastern United States",

March 12, 201413) Nuclear Energy Institute (NEI), A. Pietrangelo, Letter to D. Skeen of the USNRC,"Proposed Path Forward for NTTF Recommendation 2.1: Seismic Reevaluations",

April 9, 201314) NRC (E Leeds) Letter to All Power Reactor Licensees et al., "Screening andPrioritization Results Regarding Information Pursuant to Title 10 of the Code ofFederal Regulations 50.54(F)

Regarding Seismic Hazard Re-Evaluations forRecommendation 2.1 of the Near-Term Task Force Review of Insights From theFukushima Dai-lchi Accident,"

May 9, 2014.15) Seismic Evaluation Guidance:

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

EPRI, Palo Alto, CA: February 2013. 1025287.16) NRC (E Leeds) Letter to NEI (J Pollock),

"Electric Power Research Institute FinalDraft Report xxxxx, "Seismic Evaluation Guidance:

Augmented Approach for theResolution of Fukushima Near-Term Task Force Recommendation 2.1: Seismic,"

asan Acceptable Alternative to the March 12, 2012, Information Request for SeismicReevaluations,"

May 7, 201317) H.B. Robinson Steam Electric Plant NTTF 2.3 Seismic Walkdown Submittal datedFebruary 27, 2014.Page 40 of 46 Expedited Seismic Evaluation Process Report18) Carolina Power and Light Company (CP&L), Specification No CPL-HBR2-C-008,"Specification for Floor Response Spectra",

Revision 1, 1991.19) United States Nuclear Regulatory Commission, Regulatory Guide1.122,"Development of Floor Design Response Spectra for Seismic Design of Floor-Supported Equipment or Components",

Revision 1, February 1978.20) United States Nuclear Regulatory Commission NUREG-2115, Department of EnergyOffice of Nuclear Energy (DOE/NE)-0140, EPRI 1021097,"Central and EasternUnited States Seismic Source Characterization for Nuclear Facilities",

6 Volumes,2012.21) Electric Power Research Institute (EPRI), Final Report No 3002000717,"EPRI (2004, 2006) Ground Motion Model (GMM) Review Project",

June 2013.22) URS Energy and Construction Calculation RNP-13-05-600-001,"Review LevelGround Motion (RLGM) and In-Structure Response Spectra (ISRS) for H.B.Robinson Steam Electric Plant Unit 2, Revision 0, July 2013.23) Electric Power Research Institute (EPRI) Final Report No 1019200,"Seismic Fragility Applications Guide Update",

2009.24) EC 92103,"Fukushima NTTF Recommendation 2.1: Seismic Reevaluation"

25) Calculation RNP-13-05-600-006,"High Confidence of Low Probability of Failure(HCLPF) for the H.B. Robinson Steam Electric Plant Battery Chargers"
26) Calculation RNP-13-05-600-004,"High Confidence of Low Probability of Failure(HCLPF) for the H.B. Robinson Steam Electric Plant Boron Injection Tank"27) Calculation RNP-13-05-600-003,"High Confidence of Low Probability of Failure(HCLPF) for the H.B. Robinson Steam Electric Plant East Hagan Rack"28) Calculation RNP-13-05-600-007,"High Confidence of Low Probability of Failure(HCLPF) for the H.B. Robinson Steam Electric Plant Auxiliary DC Panel GD".29) Calculation RNP-13-05-600-002,"High Confidence of Low Probability of Failure(HCLPF) for the H.B. Robinson Steam Electric Plant Main Control Board".30) RNP/RA-13-0087, First Six Month Status Report (Order EA-12-049)

H. B. RobinsonSteam Electric Plant (RNP), Unit 2.31) RNP/RA-14-0008, Second Six Month Status Report (Order EA-12-049)

H. B.Robinson Steam Electric Plant (RNP), Unit 2.32) RNP/RA-14-0083, Third Six Month Status Report (Order EA-12-049)

H. B. RobinsonSteam Electric Plant (RNP), Unit 2.33) Engineering Change (EC) 88926, FLEX Strategies and Implementation Plan, Rev. 3.34) NEI 12-06, Diverse and Flexible Coping Strategies (FLEX) Implementation Guideline, Revision 0.35) EOP-ECA-0.0, Loss Of All AC Power, Revision 0.36) WCAP-1 760-P, Revision 1, Reactor Coolant System Response to Extended Loss ofAC Power Event for Westinghouse, Combustion Engineering, and Babcock &Wilcox NSSS Designs for Phase Boration, August 2012.37) PA-PSC-0965, PWROG Core Cooling Position Paper.38) EDMG-004, Steam Generators.

39) Calculation RNP-M/MECH-1712, Appendix R Mechanical Basis, Section 3.27,Cooldown Using MSIV Bypass Lines.40) EC 90617, Pre-Staged Diesel Generator Design To Power 125VDC -A Train and BTrain Battery Chargers For Fukushima Support (NTFF 4.2 -FLEX).41) FSG-10, Passive RCS Injection Isolation.
42) EC 95216, NTTF 2.1 Interim Action RCS Injection.
43) EC 90622, Standard Piping Connections For NTTF 4.2 (FLEX).44) EC 94745, Boric Acid and RCS Make Up Connections To The Safety Injection System NTTF 4.2 -Flexible Coping Strategies Page 41 of 46 Expedited Seismic Evaluation Process Report45) Calculation RNP-M/MECH-1877, RNP Extended Loss of AC (ELAP) PowerContainment Response46) FSG-12, Alternate Containment Cooling47) EC 90623, New Pipe Tee And Standard Connection For NTTF 4.2 (FLEX)48) EC 95266, Isolation Valves And Connection For AFW -FUKUSHIMA-Admin Rev49) EC 92103R0, Attachment Z03RO Mechanical Documents
50) EC 92103R0, Attachment Z05RO Electrical Documents
51) UFSAR, Section 02, Site Characteristics"
52) EC 92103R0, Attachment Z06RO53) EC 92103, Attachment Z18R054) EC 92103, Attachment Z09R055) EC 92103, Attachment Z1ORO56) EC 92103, Attachment Z01 RO57) EC 92103, Attachment Z16RO58) NRC Letter from NRC to Duke Energy and South Carolina Electric and GasCompany, Request for Additional Information Associated with Near-Term ForceRecommendation 2.1, Seismic Re-Evaluations Related to Southeastern CatalogChanges (TAC NOS MF3724, MF3736, MF 3738, and MF 3831), dated October 23,2014, ADAM Accession No ML14268A516.
59) H.B. Robinson Steam Electric Plant Control Wiring Diagram (CWD) B-190628, SH00955 and 00956.60) EC 92501, Attachment Z09Rl,"Additional Seismic Interim Actions Studies for theH.B. Robinson Steam Electric Plant".61) Letter Dated August 28, 2014 from EPRI to NRC Review of EPRI 1021097Earthquake Catalog for RIS Earthquakes in the Southeastern U.S. and Earthquakes in South Carolina Near Time of the 1886 Charleston Earthquake Sequence.
62) Response to Request for Additional Information Associated with Near-Term TaskForce Recommendation 2.1, Seismic Re-Evaluations Related to Southeastern Catalog Changes (TAC NOS. MF3724, MF3736, MF3737, MF3738, and MF3831),November 12, 2014Page 42 of 46 Expedited Seismic Evaluation Process ReportAttachment A -H.B. Robinson Steam Electric Plant ESELPage 43 of 46

Expedited Seismic Evaluation Process ReportAttachment B -H.B. Robinson Steam Electric Plant Unit 2 RCS Cooling Strategies Page 44 of 46 MNP72" MS HEADERMain Feed PumpsBLAplanREDStraGREStraBLUCapa-w'*-Lake Robinson______ IAJFW-TNK4

/ -'IDischarge/

ICanal na/ I -3 E AFW-TNK 9LEGEND I l I Portable

...-.. 20k gli,CK -Installed

\ o\ I.sel P pmnt equipment

\ ." -/-FLEXLP LP\ " -tegles NN-EN -FLEX HP -.tegies rlrpLI IE -Additional I Iabilities I I E08Q1I I AFW-TNK-1 AF17I I I ,201 , g//iAFW-TN K-8EIWIS -017p I-F-TNK-20k g.1AW-TNK-620W-171AFW-TNK-3 20k galI-F.1F.Attachment BPage I of I/F-AFW-PMP-2 Portable1000 psi/300 gpmDiesel Pumpers/RCS COOLINGSEISMIC STRATEGIES vMDAFWPs Expedited Seismic Evaluation Process ReportAttachment C -H.B. Robinson Steam Electric Plant Unit 2 RCS Boration and MakeupStrategies Page 45 of 46 LEGENDBLACK -Installed plant equipment n~~i 7 LakeRobinson LEGENDMagenta -Mode 5/6Makeup (Reconfigure Check Valve Covers)LCV-11SA toC/CS HUT,RED -PortableBoration andMakeup StrategyUsing the ChargingSystemE--3 SI-iGREEN -PortableBoration and MakeupStrategy Using theSafety Injection System//IF ,RCS BORATIONand MAKEUPSI ACC nt Expedited Seismic Evaluation Process ReportAttachment D -H.B. Robinson Steam Electric Plant FLEX Flow PathPage 46 of 46

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