ML13136A144

Beaver Valley, Units 1 and 2, Generic Safety Issue 191 Resolution Plan (TAC Nos. MC4665 and MC4666)

ML13136A144
Person / Time
Site:	Beaver Valley
Issue date:	05/16/2013
From:	Larson E A FirstEnergy Nuclear Operating Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
GSI- 191, L-13-176, TAC MC4665, TAC MC4666
Download: ML13136A144 (23)

Text

FENOC_EFirstEnergy Nuclear Operating CompanyBeave r V alley Powe r SfafionP.O. Box 4Shippingport, PA 15077Eric A. LarsonS[e Vice President724-682-5234Fax: 724-643-8069May 16, 2013L-1 3-1 76ATTN: Document Control DeskU. S. Nuclear Regulatory CommissionWashington, DC 20555-000110 cFR 50.54(f)SUBJECT:Beaver Valley Power Station, Unit Nos. 1 and 2BVPS-1 Docket No. 50-334, License No. DPR-66BVPS-2 Docket No. 50-4'12, License No. NPF-73Generic Safety lssue 191 Resolution Plan (TAC Nos. MC4665 and MC4666)This letter fonryards information regarding resolution of Generic Safety lssue 191,"Assessment of Debris Accumulation on Pressurized-Water Reactor SumpPerformance," for BeaverValley Power Station, Unit Nos. 1 (BVPS-1) and 2 (BVPS-2).Nuclear Regulatory Commission (NRC) staff has interacted with the industry andstakeholders to develop options forthe resolution of Generic Safety lssue 191" TheNRC staff paper SECY-12-0093 presents closure options for Generic Safety lssue 191.Attachments 1 and 2 provide information regarding the current status of efforts toaddress Generic Letter 20A4-02, "Potential lmpact of Debris Blockage on EmergencyRecirculation During Design Basis Accidents at Pressurized Water Reactors," anddescribe the Generic Safety lssue 191 closure option, resolution plan, and associatedimplementation schedule for BVPS-1 and BVPS-2, respectively. These attachmentsalso describe mitigation measures appropriate to support the implementation schedule.Attachment 3 provides references for information cited in Attachments 1 and 2, andrelated correspondence. Attachment 4 provides a list of regulatory commitmentsincluded in this submittal. lf there are any questions or if additional information isrequired, please contact Mr. Thomas A. Lentz, Manager - Fleet Licensing, at(330) 315-6810.

Beaver Valley Power Station, Unit Nos. 1 and 2L-13-176Page 2I declare under penalty of perjury that the foregoing is true and correct. Executed onMay I b ,2a8.Sincerely,ilf,/^-Eric A. LarsonAttachments:1. Beaver Valley Power Station, Unit No. 1,Generic Safety lssue 191, In-Vessel Effects Resolution Plan2. Beaver Valley Power Station, Unit No. 2,Generic Safety lssue 191, In-Vessel Effects Resolution Plan3. References4. Regulatory Commitment Listcc: NRC Region lAdministratorNRC Resident lnspectorNRC Project ManagerDirector BRP/DEPSite BRP/DEP Representative ATTACHMENT 1L-1 3-1 76Beaver Valley Power Station, Unit No. 1,Generic Safety lssue 191, In-Vessel Effects Resolution PlanPage 1 of 9Introduction:FirstEnergy Nuclear Operating Company (FENOC) has selected Option 2, deterministicpath, of Nuclear Regulatory Commission (NRC) staff paper SECY-12-0093, "ClosureOptions for Generic Safety lssue 191, Assessment of Debris Accumulation onPressurized-Water Reactor Sump Performance," for Beaver Valley Power Station, UnitNo. 1 (BVPS-1) and intends to pursue refinements to evaluation methods andacceptance criteria. ln addition, the resolution schedule defined in SECY-12-0093 willalso be adhered to as described herein. The Nuclear Energy Institute (NEl) closureoption template dated November 9, 2012 was used to develop this response. Thissubmittal provides a resolution plan that follows the deterministic path of Option 2(referred to as NEI template option 2a).To support use of this path, for the period required to complete the necessary analysisand testing, FENOC has evaluated the design and procedural capabilities that exist toidentify defense in depth measures to detect and mitigate in-vessel blockage. Adescription of these measures to detect and mitigate in-vessel blockage is providedlater in this document. A summary of the existing margins and conservatisms that existfor BVPS-1 are also included in this document. References cited in this attachment arelisted in Attachment 3.Current Gontainment Fiber Status:A bypass test was performed with an objective to collect and record the fibrous debrisbypass fraction for the BVPS-1 prototypical sump strainer. Sump strainer bypasstesting was completed for BVPS-1 in the spring of 2008, at the Alion HydraulicsLaboratory. This testing established the quantity of fiber that passed through thestrainer over a range of approach velocities and head losses. Scaled debris quantitiesused for these tests were derived from the primary line break within containment thatyields the most fibrous debris and the highest approach velocity.The strainer bypass testing was credited for determining the amount of fiber thatreaches the vessel. In addition, the bypass testing was reviewed and generally alignswith the concepts outlined by the NRC at the NEI Workshop held on October 18, 2012.For example, the debris was added to the tank in increments of approximately onesixteenth inch of bed thickness, and a bag capture method was used.Sump strainer head loss testing was conducted in the spring of 2008 in accordance withthe March 2008 protocol (Reference 1) prepared by the NRC. In the summer of 2010an additional strainer head loss test was performed for a 6 inch pressurizer safety reliefvalve line break. The total quantity of fibrous insulation was 1 1.8 pounds (lbs) for 2008

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1L-13-176Page 2 of Istrainer head loss Test 6 (Reference 2). This quantity (1 1.8 lbs) bounds breaks otherthan the pressurizer safety relief valve line break, including the debris quantities for theBVPS-1 loop break, pressurizer surge line break, safety injection line break, residualheat removal line break, and reactor vessel nozzle break. The total quantity of fibrousinsulation was 1A7 .4 lbs for 2010 strainer head loss Test 7 (Reference 3). This quantity(107.4 lbs) bounds the debris quantities for the BVPS-1 pressurizer safety relief valveline break. Thirty pounds of latent fiber must be added to these quantities (the quantityof 30 lbs is based on testing and bounds the calculated quantity). This makes the totaldebris quantity that could be transported to the strainer, 41.8 lbs (1 1 .8 lbs plus 30 lbs)for Test 6 and 137.4lbs (1 07 .4 lbs plus 30 lbs) for Test 7.The fiber bypass percentage was 8.0 percent, based on the results of plant specificstrainer bypass testing. The bypass fraction can be applied over the range of breakssince the testing showed that there was open screen area with the bounding fiber load.The amount of fiber bypass that may reach each fuel assembly was calculated on a perassembly basis. The amount of fiber per assembly in grams (g) was calculated bymultiplying the total fiber (41 .8 lbs or 137.4 lbs) by the fiber bypass percentage(8.0 percent) and converting the result from pounds to grams and then dividing that bythe number of fuel assemblies (157). Thus, the total fiber bypass load was'9.7 gramsper fuel assembly (g/FA) for Test 6 and 31 .8 grams per fuel assembly for Test 7, asshown in the following equations.41.8 lbs x0.080 x 453.6 glb=Test 6:Test 7:157 FAs.7 3-FA= 31.8 gFA137 .4lbs x0.080 x 453.6 gtb157 FATherefore, for BVPS-1 , the only break that exceeds the acceptance criteria of 15 g/FA(Reference 4) was a break associated with a pressurizer safety relief valve line. Otherbreaks fall within the acceptance criteria. Actions to address the pressurizer safetyrelief valve line are described under the heading Resolution Schedule below.Strainer Head Loss Status:FirstEnergy Nuclear Operating Company has previously provided the results of strainerhead loss testing, including chemical effects, in references 2 and 3. These testsdemonstrated acceptable results with regard to allowable strainer head loss. Concernsof the NRC staff associated with Generic Safety lssue 191 , "Assessment of DebrisAccumulation on Pressurized Water Reactor Sump Performance," Generic Letter

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1L-1 3-1 76Page 3 of 92004-0l "Potential lmpact of Debris Blockage on Emergency Recirculation DuringDesign Basis Accidents at Pressurized Water Reactors," and BVPS-1 strainer head losshave been addressed as described in references 2, 3, and 5. There are no outstandingissues with respect to head loss testing.Characterization of I n-Vessel Effects :FirstEnergy Nuclear Operating Company intends to follow the resolution strategyproposed by the Pressurized Water Reactor Owners Group (PWROG) to establishin-vessel debris fimits for the BVPS-1 type plant design. This approach is expected toestablish in-vessel debris limits in excess of that currently established by WestinghousereportWCAP-16793, Revision 2 (Reference 6).Licensing Basis Commitments:There is currently only one open FENOC commitment for BVPS-1 regarding GenericLetter 2004-02. The commitment to the NRC is contained in a June 30, 2009 FENOCletter (Reference 2) that states:It is recognized thatthe NRC review of WCAP-16793-NP, Revision 1, has not beencompleted. Any additional actions required to address NRC questions will beaddressed. The due date is, within 90 days after issuance of the final NRC safetyevaluation on WCAP-16793-NP, Revision 1.The NRC has not issued a safety evaluation forWCAP-16793-NP, Revision 1. In theinterim WCAP-16793-NP, Revision 2, October zAfi (Reference 6), was issued byWestinghouse for review and approval by the NRC. A safety evaluation for Revision 2of WCAP-16793-NP was made available by the NRC on April 16, 2013 (Reference 7).The above commitment, listed as Commitment 1 in Attachment 2 of Reference 2, ishereby replaced with the commitments described under the Resolution Scheduleheading below and listed in Attachment 4.Resolution Schedule:FirstEnergy Nuclear Operating Company will achieve closure of Generic Safetylssue 191 and address Generic Letter 2004-02 per the schedule provided below. Theresults of the PWROG in-vessel effects testing effort will be utilized to close theremaining issue for BVPS-1 if it provides a conclusion that supports the current fibrousfuel assembly loading forthe BVPS-1 limiting break (pressurizer safety relief valve line).The PWROG program results are expected to be available and provided to the NRCstaff for review in the fall of 2014 (Reference 8).Measurements will be taken in preparation for insulation modifications associated withthe BVPS-1 pressurizer safety relief valve inlet lines, if it is determined that the PWROGin-vessel effects testing effort does not support closure of Generic Safety lssue 191 and

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1L-13-176Page 4 of 9Generic Letter 2AA4-02 for BVPS-1. This action will be accomplished during theBVPS-1 refueling outage in the fall of 2013.Insulation for the BVPS-1 pressurizer safety relief valve inlet lines will be replaced ormodified as appropriate if it is determined that the PWROG in-vessel effects testingeffort does not support closure of Generic Safety lssue 191 and Generic Letter 2004-02for BVPS-1. This action will be accomplished by the end of the refueling outage in thefafl of 2A16, if needed.The final Generic Letter 2004-02 supplemental response for BVPS-1 will be provided tothe NRC within 6 months after the NRC approves, by issuance of a safety evaluation,the new PWROG topical report addressing additional in-vessel effects testing effortsthat are currently being pursued.Updated Final Safety Analysis Report changes will be completed to update the currentlicensing basis for BVPS-1 as appropriate, following NRC acceptance of the finaldocketed Generic Safety lssue 191 response for BVPS-1 and completion of anyidentified insulation modifications for BVPS-1 that may be required.Summary of Actions Gompleted:A strainer replacement was installed at BVPS-1 during the fall 2007 refueling outage(1R18). The new replacement strainer is of Control Components Incorporated (CCl)design, and increased the available surface area from approximately 130 square feet to3400 square feet.The BVPS-1 start signal for the recirculation spray system pumps was changed from afixed time delay to an engineered safety features actuation system signal based on arefueling water storage tank level low coincident with a containment pressure high-highsignal. This change was completed during the fall 2007 refueling outage (1R18), andwill allow sufficient pool depth to cover the sump strainer before initiating recirculationflow.Beaver Valley Power Station, Unit No. 1, is considered a low to medium fiber plantbased on approved industry and NRC standards associated with Generic Safetyf ssue 191 and Generic Letter 2004-02 This has been accomplished through extensivereplacement of fibrous and particulate insulation with reflective metal insulation.The sodium hydroxide containment sump buffer was replaced in the spring 2012refueling outage (1R21). The replacement buffer is sodium tetraborate. This lowers thechemical loading for BVPS-1 as discussed in references 9 and 10.A containment coatings inspection and assessment program and containment cleaningprogram became effective for BVPS in April 2008 and applies to BVPS-1 refueling

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1L-1 3-1 76Page 5 of 9outages beginning with the spring 2009 refueling outage (1 R19). This reduces thequantity of unqualified coatings and latent debris that may be transported to the sump.lodine filters, containing a significant amount of thin aluminum that would have beensubmerged, were removed from the BVPS-1 containment. This supported a significantreduction in the generation of chemical precipitates.Details of additional actions and modifications completed to address Generic Safetylssue 191 for BVPS-1 are provided in FENOC letters dated June 30, 2009 andSeptember 28, 2010 (references 2 and 3).Summary of Margins and Gonseruatisms:Margins and conservatisms with respect to debris generation, debris transpott, strainerhead loss and chemical effects have been summarized in previous Generic Letter2004-02 submittals (references 2 and 3). Additional information is ofiered here.Reactor In-vessel Fiber LoadinqThe limiting break is a pressurizer safety relief valve line break, which generates 31.8g/FA of fiber. Significant fiber reduction has been conducted in containment such thatno other evaluated breaks result in excess of 15 g/FA of fiber.The fibrous debris limits were generated from testing conducted at limiting reactorcoofant system flows (that is, 44.7 and 15.5 gallons-per-minute per fuel assembly[gpm/FA]). Pressurized Water Reactor Owners Group fuel assembly testing hasdemonstrated that maximum head loss occurs at high flow conditions (44.7 gpm/FA).A small break loss of coolant accident at a 6 inch pipe near the top of the pressurizer(such as the pressurizer safety relief valve line break) will experience a greatly reducedflow due to the significantly smaller break size and large elevation difference betweenthe reactor core and the break. An analysis of the pressurizer safety relief valve linebreak using the containment transient analysis code MAAP-DBA, determined theemergency core cooling system flow rate to be approximately 23.7 gpm/FA. Fuelassembly testing performed per Westinghouse fuel assembly test report, WCAP 17057-P, Revision 1 (Reference 11), used flow rates of 44.7 gpm and 15.5 gpm. These testsconsistently demonstrated that maximum head loss occurs at high flow conditions;therefore, the smaller flow rate associated with the pressurizer safety relief valve linebreak would translate into a higher acceptable in-vessel fiber load.Hot Les Drivinq HeadAccording to WCAP-16793-NP, Rev. 2, the PWROG testing demonstrated that sincethe hot leg break is limiting with respect to allowable fiber loading, the calculation of thehot leg available driving head is the relevant value when determining if a reduction incore flow occurs. For the pressurizer safety relief valve line break with 31.8 g/FA offiber, the RCS'loops will remain pressurized due to the size and elevation of the break;

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1L-1 3-1 76Page 6 of 9therefore, the hot leg driving head methodology provided by WCAP-16793 does notapply to this break.The available driving head value for a hot leg break can be compared to the differentialpressure value recorded from the test conducted with 15 grams of fiber to demonstratethat significant margin exists between the expected pressure loss due to a debris bedand the expected driving head available to support core flow. The available driving headfor a hot leg break at BVPS-1 is 16.37 psi; the limiting AREVA Enterprises Inc. (AREVA)fuel assembly test that uses 15 grams of fiber (12-FG-FPC) resulted in a total fuelassembly pressure drop of only 2.7 psid. This confirms that significant margin existsbetween the hot leg break available driving head and the expected pressure lossthrough the debris bed, and there will be no significant reduction in core flow.Several tests were performed at Westinghouse, which show debris head loss resultsthat are comparable to the available driving head at BVPS-1. Tests ClB49, ClB50, andCf B51 each use 50 g/FA of fiber and produced debris head losses of 14.94,17.80, and15.35 psid, respectively (following addition of chemical precipitates). Since a BVPS-1hot leg break will produce less than 15 g/FA of fiber, significant margin exists betweenthe actual fiber quantity at BVPS-1 and that which was used in these Westinghouse FAtests.Reactor Vessel DesiqnReactor vessel designs were considered (Reference 6) and a limiting design waschosen based upon the vessel design that would be the most limiting with respect tocore inlet flow blockage. Three Westinghouse vessel designs were considered;designed barrel/baffle (B/B) upflow, converted B/B upflow and B/B downflow. ForWestinghouse designed plants, the most limiting vessel design is the B/B downflow,since the only means for the flow to enter the core is through the lower core plate. AWestinghouse B/B downflow plant was evaluated using the WCOBRA/TRAC thermal-hydraulic computer code. For this evaluation, it was concluded that sufficient liquid canenter the core to remove core decay heat once the plant has switched to sumprecirculation with up to 99.4 percent core blockage. Beaver Valley Power Station, UnitNo. 1, is a converted barrel/baffle upflow plant, which is less limiting for fuel assemblyblockage as stated above and provides an alternate path to provide cooling to the core.Boron PrecipitationBoron precipitation is an issue that becomes problematic with cold leg breaks. Asstated in WCAP-16793-NP, Revision 2 (Reference 6):The limiting scenario for boric acid precipitation is a cold leg break where the coreflow is stagnant with only enough core inlet flow to replace core boil-off.Therefore, boron precipitation is not an issue for any other breaks including the limitingBVPS-1 pressurizer safety relief valve line break.

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1L-1 3-1 76Page 7 of 9Summary of Defense-in-Depth MeasuresBVPS-1 has a low concentration of precipitates formed during a loss of coolantaccident, due to the reduction of available submerged aluminum in containment and thechange of the containment sump buffering agent from sodium hydroxide to sodiumtetraborate. The existing emergency operating procedure guidance for transfer to hotleg recirculation is 6.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> after initiation of a loss of coolant accident. For this timeframe, considering injection time and transfer to recirculation time, it is expected that areduced concentration of precipitates would be formed. However, to enhance thecapability to address postulated core inlet blockage, the BVPS-1 emergency operatingprocedures will be revised using recent guidance from the PWROG to implement earlyswitchover to hot leg recirculation should plant parameters indicate that core blockage isoccurring. This action will be taken prior to transfer to the existing "Response toDegraded Core Cooling" procedure. Appropriate operator training will be completed toaddress this emergency operating procedure revision prior to implementation. Theseactions will be completed within six months of the submittal date of this Generic Safetylssue 191 resolution plan.Existing defense-in-depth mitigative measures are described below.Containment Sump ScreenActions taken in response to NRC Bulletin 2003-01, "Potential lmpact of DebrisBlockage on Emergency Sump Recirculation at Pressurized-Water Reactors," aredescribed in Referen ce 12. These actions continue to remain in effect at BVPS-1 .ln-VesselThe following describes the plant specific design features and procedural capabilitiesthat exist for detecting and mitigating fuel blockage.Detection of Inadequate Reactor Core Flow- Increasing core exit thermocouple temperature indicationCore exit thermocouples are monitored as part of emergency operating proceduremonitoring of status trees and the safety parameter display system. As part of operatortraining, the operating crew must demonstrate the ability to detect increases in core exitthermocouple temperature indication and transition to the appropriate emergencyoperating procedure for dealing with this condition. This guidance is provided in theBVPS emergency operating procedure "User's Guide" section titled "Control RoomUsage of Status Trees," and the BVPS-1 "Core Cooling Status Trees" procedure.- Decreasing reactor water level indicationThe reactor vessel level indication system is monitored throughout the emergencyoperating procedures. Through continuing training, operators demonstrate the ability tomonitor and understand the implications of a decreasing reactor vessel water level and

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1L-13-176Page 8 of 9appropriately transition within the emergency operating procedure framework to mitigatethis condition.- Increasing containment or auxiliary building radiation levelsIncreasing radiation levels will be indicated by alarms in the control room with specificprocedural steps in both alarm response procedures and the emergency operatingprocedures for addressing the condition.Mitigation of Inadequate Reactor Core Flow- Start a reactor coolant pumpBeaver Valley Power Station, Unit No. 1, procedure titled, "Response to InadequateCore Cooling," provides direction to start a reactor coolant pump if core exitthermocouple temperature indication is greater than 12A} degrees Fahrenheit. Thisaction would aid in the removal of the established blockage to the core to once againallow normal recirculation injection flow paths to become effective at maintainingadequate core cooling.- lmplementation of Severe Accident Management GuidelinesSevere Accident Management Guidelines (SAMG) provide additional guidance andactions for addressing inadequate core flow conditions. At BVPS-1 the SAMGs areentered from the following function restoration procedure for inadequate core cooling.Procedure titled, "Response to Inadequate Core Cooling," when core exitthermocouple temperatures are greater than 1200 degrees Fahrenheit and actionsto cool the core are not successful.The SAMGs provide for alternate injection paths into the reactor coolant system (RCS).Specific SAMGs that provide guidance in this area are listed below."lnject lnto the RCS" guideline"RCS Injection to Recover Core" guidelineThe SAMGs provide for flooding containment to provide for convective circulationcooling of the reactor. Specific SAMGs that provide guidance in this area are listedbelow."lnject Into Containment" guideline"Flood Containment" guideline"Containment Water Level And Volume" guideline

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1L-13-176Page 9 of 9Although these measures are not expected to be required based on the very lowprobability of an event that would result in significant quantities of debris beingtransported to the reactor vessel that would inhibit the necessary cooling of the fuel,they do provide additional assurance that the health and safety of the public would beprotected. These measures provide support for the extension of time required tocompletely address Generic Letter 2004-02fior BVPS-1.ConclusionThe Generic Safety lssue 191 resolution path for BVPS-1 is acceptable, based on theinformation provided in this document. The execution of the actions identified in thisdocument will result in successful resolution of Generic Safety lssue 191 and closure ofGeneric Letter 20A4-02.

ATTACHMENT 2L-13-176Beaver Valley Power Station, Unit No. 2,Generic Safety lssue 191, In-Vessel Effects Resolution PlanPage 1 of 8Introduction:FirstEnergy Nuclear Operating Company (FENOC) has selected Option 2, deterministicpath, of Nuclear Regulatory Commission (NRC) staff paper SECY-12-0093, "ClosureOptions for Generic Safety lssue 191, Assessment of Debris Accumulation onPressurized-Water Reactor Sump Performance," for Beaver Valley Power Station,Unit No. 2 (BVPS-2) and intends to pursue refinements to evaluation methods andacceptance criteria. In addition, the resolution schedule defined in SECY-12-0093 willalso be adhered to as described herein. The Nuclear Energy Institute (NEI) closureoption template dated November 9, 2012 was used to develop this response. Thissubmittal provides a resolution plan that follows the deterministic path of Option 2(referred to as NEI template option 2a).To support use of this path, for the period required to complete the necessary analysisand testing, FENOC has evaluated the design and procedural capabilities that exist toidentify defense in depth measures to detect and mitigate in-vessel blockage. Adescription of these measures to detect and mitigate in-vessel blockage is providedlater in this document. A summary of the existing margins and conservatisms that existfor BVPS-2 are also included in this document. References cited in this attachment arelisted in Attachment 3.Current Gontainment Fiber Status:A bypass test was performed with an objective to collect and record the fibrous debrisbypass fraction for the BVPS-2 prototypical sump strainer. Sump strainer bypasstesting was completed for BVPS-2 in the fall of 2008, at the Alion Hydraulics Laboratory.This testing established the quantity of fiber that passed through the strainer over arange of approach velocities and head losses. Scaled debris quantities used for thesetests were derived from the primary line break within containment that yields the mostfibrous debris and the highest approach velocityThe strainer bypass testing was credited for determining the amount of fiber thatreaches the vessel. In addition, the bypass testing was reviewed and generally alignswith the concepts outlined by the NRC at the NEI Workshop held on October 18,2012.For example, the debris was added to the tank in increments of approximately onesixteenth inch of bed thickness, and a bag capture method was used.Sump strainer head loss testing was conducted in the fall of 2008 in accordance withthe March 2008 protocol (Reference 1) prepared bythe NRC. The total quantity offibrous insulation was 36.4 pounds (lbs) for BVPS-2 strainer head loss Test 1A(references 2 and 3). This quantity bounds the debris quantities for BVPS-2 breaks.

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2L-13-176Page 2 of 8Thirty pounds of latent fiber must be added to this quantity (the quantity of 30 lbs isbased on testing and bounds the calculated quantity). This makes the total debrisquantity that could be transported to the strainer, 66.4 lbs (36.4 lbs plus 30 lbs). Thistotal value was conservative based on the conservatisms utilized for the analyses.The fiber bypass percentage was 4.2 percent, based on the results of plant specificstrainer bypass testing. The amount of fiber bypass that may reach each fuel assemblywas calculated on a per assembly basis. The amount of fiber per fuel assembly ingrams (g) was calculated by multiplying the total fiber (66.4 lbs) by the fiber bypasspercentage (4.2 percent) and converting the result from pounds to grams and thendividing that by the number of fuel assemblies (1 57). Thus, the total fiber bypass loadfor BVPS-2 was 8.1 grams per fuel assembly (g/FA), as shown in the followingequation.66.41bsx0.042x453.6 gBased on this fiber bypass load (8.1 g/FA), bounding breaks for BVPS-2 met the fibrouslimit per fuel assembly of 15 grams per fuel assembly (Reference 4 andSECY-12-0093). However, FENOC plans to continue to support the Pressurized WaterReactor Owners Group (PWROG) effort to establish improved margin for in-vesseleffects that apply to BVPS-2.Strainer Head Loss Status:FirstEnergy Nuclear Operating Company has previously provided the results of strainerhead loss testing, including chemical effects in references 2, and 3. These testsdemonstrated acceptable results with regard to allowable strainer head loss. Concernsof the NRC staff associated with Generic Safety lssue 191, "Assessment of DebrisAccumulation on Pressurized Water Reactor Sump Performance," Generic Letter2004 -02, "Potential lmpact of Debris Blockage on Emergency Recirculation DuringDesign Basis Accidents at Pressurized Water Reactors," and BVPS-2 strainer head losshave been addressed as described in references 2, 3, and 5. There are no outstandingissues with respect to head loss testing.Characterization of I n-Vessel Effects :FirstEnergy Nuclear Operating Company intends to follow the resolution strategyproposed by the PWROG to establish in-vessel debris limits for the BVPS-2 type plantdesign. This approach is expected to establish in-vessel debris limits in excess of thatcurrently established by Westinghouse report WCAP-16793, Revision 2 (Reference 6).= 8.1 8-FA

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2L-1 3-1 76Page 3 of 8Licensing Basis Commitments:There is currently only one open FENOC commitment for BVPS-2 regarding GenericLetter 2004-02. The commitment to the NRC is contained in a June 30, 2009 FENOCletter (Reference 2) that states:It is recognized that the NRC review of WCAP-16793-NP, Revision 1, has not beencompleted. Any additional actions required to address NRC questions will beaddressed. The due date is, within 90 days after issuance of the final NRC safetyevaluation on WCAP-16793-NP, Revision 1.The NRC has not issued a safetyevaluation forWCAP-16793-NP, Revision 1. In theinterim WCAP-16793-NP, Revision 2, October 2011 (Reference 6), was issued byWestinghouse for review and approval by the NRC. A safety evaluation for Revision 2of WCAP-16793-NP was made available by the NRC on April 16,2013 (Reference 7).The above commitment, listed as Commitment 1 in Attachment 2 of Referenc,e 2,hereby replaced with the commitments described under the Resolution Scheduleheading below and listed in Attachment 4.Resolution Schedule:FirstEnergy Nuclear Operating Company will achieve closure of Generic Safetylssue 191 and address Generic Letter 2004-02 perthe schedule provided below. Thefuel assembly fibrous debris loading for BVPS-2 is within the limits established byWestinghouse evaluation report WCAP-16793, Revision 2. The results of the ongoingPWROG in-vessel effects testing effort will be utilized to close this remaining issue forBVPS-2, if it provides a conclusion that supports additional margin for the currentBVPS-2 fuel assembly fibrous debris loading. The PWROG program results areexpected to be available and provided to the NRC staff for review in the fall of 2Q14(Reference 8).The final Generic Letter 2004-02 supplemental response for BVPS-2 will be provided tothe NRC within 6 months after the NRC approves, by issuance of a safety evaluation,the new PWROG topical report addressing additional in-vessel effects testing effortsthat are currently being pursued.Updated Final Safety Analysis Report changes will be completed to update the currentlicensing basis for BVPS-2 as appropriate, following NRC acceptance of the finaldocketed Generic Safety lssue 191 response for BVPS-2.Summary of Actions Gompleted:A strainer replacement was installed at BVPS-2 during the fall 2006 refueling outage(2R12). The new replacement strainer is of Enercon design with bypass eliminators,

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2L-1 3-1 76Page 4 of 8which increased the available surface area from approximately 150 square feet to3300 square feet.The BVPS-2 start signal for the recirculation spray system pumps was changed from afixed time delay to an engineered safety features actuation system signal based on arefueling water storage tank level low coincident with a containment pressure high-highsignal. This change was completed during the spring 2008 refueling outage (2R13),and will allow sufficient pool depth to cover the sump strainer before initiatingrecirculation flow.Beaver Valley Power Station, Unit No. 2, is considered a low fiber plant based onapproved industry and NRC standards associated with Generic Safety lssue 191 andGeneric Letter 2004-02. This has been accomplished through extensive replacement offibrous and particulate insulation with reflective metal insulation.The sodium hydroxide containment sump buffer was replaced in the fall 2009 refuelingoutage (2R14). The replacement buffer is sodium tetraborate. This lowers the chemicalfoading for BVPS-2 as discussed in references 13,14 and 15.A containment coatings inspection and assessment program and containment cleaningprogram became effective for BVPS in April 2008 and applies to BVPS-2 refuelingoutages beginning with the spring 2008 refueling outage (2R13). This reduces thequantity of unqualified coatings and latent debris that may be transported to the sump.lodine filters, containing a significant amount of thin aluminum that would have beensubmerged, were removed from the BVPS-2 containment. This supported a significantreduction in the generation of chemical precipitates.Details of additional actions and modifications completed to address Generic Safetylssue 191 for BVPS-2 are provided in FENOC letters dated June 30,2009 andSeptember 28, 2010 (references 2 and 3).Summary of Margins and Conservatisms:Margins and conservatisms with respect to debris generation, debris transport, strainerhead loss and chemical effects have been summarized in previous Generic Letter2004-A2 submittals (references 2 and 3). Additional information is offered here.Reactor In-vessel Fiber LoadinqThe BVPS-2 strainer design includes Enercon cylindrical top-hat style strainerassemblies with debris eliminators. This limits the amount of debris that can bypass thestrainer and make it into the vessel. In addition, a significant debris reduction effortwasundertaken. This, in conjunction with the debris eliminators and the associated bypassfraction, yields an in-vessel loading of less than 15 g/FAfor all evaluated breaks.

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2L-1 3-1 76Page 5 of IHot Leq Drivino HeadAccording to WCAP-16793-NP, Rev. 2, the PWROG testing demonstrated that sincethe hot leg break is limiting with respect to allowable fiber loading, the calculation of thehot leg available driving head is the relevant value when determining if a reduction incore flow occurs. The available driving head value for a hot leg break can be comparedto the differential pressure value recorded from the test conducted with 15 grams offiber to demonstrate that significant margin exists between the expected pressure lossdue to a debris bed and the expected driving head available to support core flow. Theavailable driving head for a hot leg break at BVPS-2 is 16.51 psi; the limiting AREVAfuel assembly testthat uses 15 grams of fiber (12-FG-FPC) resulted in a total fuelassembly pressure drop of only 2.7 psid. This confirms that significant margin existsbetween the hot leg break available driving head and the expected pressure lossthrough the debris bed, and there will be no significant reduction in core flow.Several tests were performed at Westinghouse, which show debris head loss resultsthat are comparable to the available driving head at BVPS-2. Tests ClB49, ClB50, andCfB51 each use 50 g/FA of fiber and produced debris head losses of 14.94, 17.80, and15.35 psid, respectively (following addition of chemical precipitates). Since a BVPS-2hot leg breakwill produce less than 15 g/FA of fiber, significant margin exists betweenthe actual fiber quantity at BVPS-2 and that which was used in these Westinghouse FAtests.ReactolVessel DesiqnReactor vessel designs were considered (Reference 6) and a limiting design waschosen based upon the vessel design that would be the most limiting with respect tocore inlet flow bfockage. Three Westinghouse vessel designs were considered;designed barrel/baffle (B/B) upflow, converted B/B upflow and B/B downflow. ForWestinghouse designed plants, the most limiting vessel design is the B/B downflow,since the only means for the flow to enter the core is through the lower core plate. AWestinghouse B/B downflow plant was evaluated using the WCOBRA/TRAC thermal-hydraulic computer code. For this evaluation, it was concluded that sufficient liquid canenter the core to remove core decay heat once the plant has switched to sumprecirculation with up to 99.4 percent core blockage. Beaver Valley Power Station, UnitNo. 2, is a designed barrel/baffle upflow plant, which is less limiting for fuel assemblyblockage as stated above and provides an alternate path to provide cooling to the core.BVPS-2 Chemical PrecipitatesThe BVPS-2 containment sump buffer was changed from sodium hydroxide to sodiumtetraborate during the fall 2009 refueling outage (2R14). Additionally, iodine filters thatcontain a significant amount of thin aluminum that would have been submerged wereremoved from the BVPS-2 containment. These modifications have resulted in adecrease in the quantity of post-LOCA chemical precipitates aluminum oxyhydroxideand sodium aluminum silicate. For the loop break, no aluminum oxyhydroxide is

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2L-13-176Page 6 of 8predicted to be formed, and the other chemical precipitates of concern are predicted tobe in the form of sodium aluminum silicate.Tests have been performed by the Argonne National Laboratory that compare the headloss properties of atuminum oxyhydroxide, sodium aluminum silicate, and other potentialchemical precipitates. These tests have shown that much more sodium aluminumsilicate than aluminum oxyhydroxide is needed to cause significant head loss. In theWestinghouse fuel assembly test program, chemical precipitates were represented byal um inum o4yhyd roxide.Therefore, post-LOCA chemical precipitates of concern have been reduced, and FAtesting is based on a more limiting chemical precipitate (aluminum oxyhydroxide) thanthe form predicted to be present.Summary of Defense-in-Depth MeasuresBVPS-2 has a low level concentration of precipitates formed during a loss of coolantaccident due to the reduction of available submerged aluminum in containment and thechange of the containment sump buffering agent from sodium hydroxide to sodiumtetraborate. The existing emergency operating procedure guidance for transfer to hotleg recirculation is 6.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> after initiation of a loss of coolant accident. For this timeframe, considering injection time and transfer to recirculation time, it is expected that areduced concentration of precipitates would be formed. However, to enhance thecapability to address possible core inlet blockage, the BVPS-2 emergency operatingprocedures will be revised using recent guidance from the PWROG to implement earlyswitchover to hot leg recirculation should plant parameters indicate that core blockage isoccurring. This action will be taken prior to transfer to the existing "Response toDegraded Core Cooling" procedure. Appropriate operator training will be completed toaddress this emergency operating procedure revision prior to implementation. Theseactions will be completed within six months of the submittal date of this Generic Safetylssue 191 resolution plan.Existing defense-in-depth mitigative measures are described below.Containment Sump ScreenActions taken in response to NRC Bulletin 2003-01, "Potential lmpact of DebrisBlockage on Emergency Sump Recirculation at Pressurized-Water Reactors,n aredescribed in Referen ce 12. These actions continue to remain in effect at BVPS-2.In-VesselThe following describes the plant specific design features and procedural capabilitiesthat exist for detecting and mitigating fuel blockage.

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2L-1 3-1 76Page 7 of 8Detection of Inadequate Reactor Core Flow- Increasing core exit thermocouple temperature indicationCore exit thermocouples are monitored as part of emergency operating proceduremonitoring of status trees and the safety parameter display system. As part of operatortraining, the operating crew must demonstrate the ability to detect increases in core exitthermocouple temperature indication and transition to the appropriate emergencyoperating procedure for dealing with this condition. This guidance is provided in theBVPS emergency operating procedure .User's Guide" section titled "Control RoomUsage of Status Trees," and the BVPS-2 "Core Cooling Status Trees" procedure.- Decreasing reactor water level indicationThe reactor vessel level indication system is monitored throughout the emergencyoperating procedures. Through continuing training, operators demonstrate the ability tomonitor and understand the implications of a decreasing reactor vessel water level andappropriately transition within the emergency operating procedure framework to mitigatethis condition.- Increasing containment or auxiliary building radiation levelsIncreasing radiation levels will be indicated by alarms in the control room with specificprocedural steps in both alarm response procedures and the emergency operatingprocedures for addressing the condition.Mitigation of Inadequate Reactor Core Flow- Start a reactor coolant pumpBeaver Valley Power Station, Unit No. 2, procedure titled, "Response to InadequateCore Cooling," provides direction to start a reactor coolant pump if core exitthermocouple temperature indication is greater than 12A0 degrees Fahrenheit. Thisaction would aid in the removal of the established blockage to the core to once againallow normal recirculation injection flow paths to become effective at maintainingadequate core cooling.- lmplementation of Severe Accident Management GuidelinesSevere Accident Management Guidelines (SAMG) provide additional guidance andactions for addressing inadequate core flow conditions. At BVPS-2 the SAMGs areentered from the following function restoration procedure for inadequate core cooling.Procedure titled, "Response to Inadequate Core Cooling," when core exitthermocouple temperatures are greater than 120A degrees Fahrenheit and actions tocool the core are not successful.

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2L-13-176Page 8 of IThe SAMGs provide for alternate injection paths into the reactor coolant system (RCS).Specific SAMGs that provide guidance in this area are listed below."lnject Into the RCS' guideline"RCS Injection to Recover Core" guidelineThe SAMGs provide for flooding containment to provide for convective circulationcooling of the reactor. Specific SAMGs that provide guidance in this area are listedbelow."lnject Into Containment" guideline"Flood Containment" guideline"Containment Water Level And Volume" guidelineAlthough these measures are not expected to be required based on the very lowprobability of an event that would result in significant quantities of debris beingtransported to the reactor vessel that would inhibit the necessary cooling of the fuel,they do provide additional assurance that the health and safety of the public would beprotected. These measures provide support for the extension of time required tocompfetely address Generic Letter 2004-02 for BVPS-2.ConclusionThe Generic Safety lssue 191 resolution path for BVPS-2 is acceptable, based on theinformation provided in this document. The execution of the actions identified in thisdocument will result in successful resolution of Generic Safety lssue 191 and closure ofGeneric Letter 2044-02.

1.2.ATTACHMENT 3L-13-176ReferencesPage 1 of 2NRC Staff Review Guidance Regarding Generic Letter 2A04-A2 Closure in the Areaof Strainer Head Loss and Vortexing, dated March 2008, Accession No.M1080230038.FENOC Letter L-09-152, Subject: Supplemental Response to Generic Letter2AA4-02 (TAC Nos. MC4665 and MC4666), dated June 30, 2009,Accession No. ML091830390.FENOC Letter L-10-1 15, Subject: Response to Request for Additional InformationRelated to Generic Letter 2004-02 (TAC Nos. MC4665 and MC4666), datedSeptember 28, 2010, Accession No. ML1 02770023.NRC Letter, Subject: NRC Review of Nuclear Energy lnstitute Clean PlantAcceptance Criteria for Emergency Core Cooling Systems, dated May 2,2A12,Accession No. ML1 20730181 .NRC Letter, Subject: Summary of April 21,2010 Category 1 Teleconference withFirstEnergy Nuclear Operating Company on Generic Letter 2004-02 (TAC Nos.MC4665 and MC4666), dated May 18,2010, Accession No. ML101320665.Westinghouse Report WCAP-16793-NP, Revision 2, "Evaluation of Long-TermCooling Considering Particulate, Fibrous and Chemical Debris in the RecirculatingFluid," dated October 2011, Accession No. ML1 12924421.NRC Letter, Subject: Final Safety Evaluation for Pressurized Water Reactor OwnersGroup Topical Report WCAP-16793-NP, Revision 2, Evaluation of Long-TermCooling Considering Particulate Fibrous and Ghemical Debris in the RecirculatingFluid" (TAC No. ME1234), dated April 8,2013, Accession Nos. ML13084A152 andM113084A154.Pressurized Water Reactor Owners Group Letter OG-12-395, Subject: PWR OwnersGroup GSI-191 In-Vessel Debris Program (PA-SEE-0312 Rev. 4 and PA-SEE-4872,Revision 0), dated September 20, 2012, Accession No. ML122900033.FENOC Letter L-1 1 -141, Subject: License Amendment Request 10-021,Replacement of Beaver Valley Power Station Unit No. 1 Spray Additive System byContainment Sump pH Control System, dated May 27,2A11 , Accession No.ML1 1 1510646.10. NRC Letter, Subject: Beaver Valley Power Station, Unit Nos. 1 and 2 - lssuance ofAmendments Regarding the Spray Additive System by Containment Sump pHControl System (TAC Nos. ME6352 and ME6353), dated March 14,2012,Accession No. ML120530591 .3.4.5.6.7.8.9.

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3L-1 3-1 76Page 2 of 211. Pressurized Water Reactor Owners Group Letter OG-1 1-291, Subject PWR OwnersGroup For Information Only - WCAP-17057-P/NP, Revision 1, "GSl-191 FuelAssembly Test Report for PWROG," (PA-SEE-0312, Revision 2) datedOctober 12, 2011 , Accession No. ML1 1293A09812.NRC Letter, Subject: BeaverValley PowerStation Unit Nos. 1 and 2 (BVPS-1 and 2)Response to NRC Bulletin 2003-01, Potential lmpact of Debris Blockage onEmergency Sump Recirculation at Pressurized-Water Reactors (TAC Nos. MB9554and M89555), dated September 6, 2005, Accession No. ML052410375.13. FENOC Letter L-08-236, Subject: License Amendment Request No.08-006,Replacement of Beaver Valley Power Station Unit No. 2 Spray Additive System byContainment Sump pH Control System, dated September24,2008, AccessionNo. M1082730716.14. FENOC Letter L-08-350, Subject. Response to Request for SupplementalInformation Regarding Containment Spray Additive System License AmendmentRequest (TAC Nos. MD9734 and MD9735), dated November 10, 2008, AccessionNo. M1083180133.15.NRC Letter, Subject: BeaverValley Power Station, Unit Nos. 1 and 2 - lssuance ofAmendments Re: Spray Additive System by Containment Sump pH Control(TAC Nos. MD9734 and MD9735), dated April 16, 2009, AccessionNo. M1090780352.

ATTACHMENT 4L-13-176Regulatory Commitment ListPage 1 of 2The following list identifies those actions committed to by FirstEnergy Nuclear OperatingCompany (FENOC) for Beaver Valley Power Station (BVPS) Unit Nos. 1 and 2 in thisdocument. Any other actions discussed in the submittal represent intended or plannedactions by FENOC. They are described only as information and are not RegulatoryCommitments. Please notify Mr. Thomas A. Lentz, Manager - Fleet Licensing, at330-315-6810 of any questions regarding this document or associated RegulatoryCommitments.Requlatory Commitment Due Date1. Measurements will be taken in This action will be accomplishedpreparation for insulation modifications during the BVPS-1 refueling outageassociated with the BVPS-1 pressurizer in the fall of 2013.safety relief valve inlet lines, if it isdetermined that the PWROG in-vesseleffects testing effort does not supportclosure of Generic Safety lssue 191 andGeneric Letter 2004-02 for BVPS-1.2. lnsulation for the BVPS-1 pressurizer This action will be accomplished bysafety relief valve inlet lines will be the end of the refueling outage inreplaced or modified as appropriate if it is the fall af 2A16, if needed.determined that the PWROG in-vesseleffects testing effort does not supportclosure of Generic Safety lssue 191 andGeneric Letter 2004-02 for BVPS-1.3. The final Generic Letter 2OO4-02 Within 6 months after the NRCsupplemental response for BVPS-1 will approves, by issuance of a safetybe provided to the NRC. evaluation, the new PWROG topicalreport addressing additional in-vessel effects testing efforts thatare currently being pursued.4. Updated Final Safety Analysis Report Following NRC acceptance of thechanges will be completed to update the final docketed Generic Safetycurrent licensing basis for BVPS-1 as lssue 191 response for BVPS-1 andappropriate. completion of any identified_insulation modifications for BVPS-1that may be required.

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4L-13-176Page2 ot 2Requlatorv Commitment Due Date5. The BVPS-I emergency operating These actions will be completedprocedures will be revised using recent within six months of the submittalguidance from the PWROG to implement date of this Generic Safetyearly switchover to hot leg recirculation lssue 191 resolution plan.should plant parameters indicate thatcore blockage is occuning. This actionwill be taken prior to transfer to theexisting "Response to Degraded CoreCooling' procedure. Appropriate operatortraining will be completed to address thisemergency operating procedure revisionprior to implementation6. The final Generic Letter 2004-02 Within 6 months after the NRCsupplemental response for BVPS-2 will approves, by issuance of a safetybe provided to the NRC. evaluation, the new PWROG topicalreport addressing additional in-::::1,:J'",iJ":T1"3,"J,:S'n"'7. Updated Final Safety Analysis Report Following NRC acceptance of thechanges will be completed to update the final docketed Generic Safetycurrent licensing basis for BVPS-2 as lssue 191 response for BVPS-2.appropriate.8. The BVPS-2 emergency operating These actions will be completedprocedures will be revised using recent within six months of the submittalguidance from the PWROG to implement date of this Generic Safetyearly switchover to hot leg recirculation lssue 191 resolution plan.should plant parameters indicate thatcore blockage is occurring. This actionwill be taken prior to transfer to theexisting 'Response to Degraded CoreCooling" procedure. Appropriate operatortraining will be completed to address thisemergency operating procedure revisionprior to implementation.