Text

Supplemental Response to NRC Generic Letter 2004-02
	ML20335A564
Person / Time
Site:	Beaver Valley
Issue date:	11/30/2020
From:	Grabnar J; Energy Harbor Nuclear Corp
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	GL-2004-02, L-20-162
	Download: ML20335A564 (62)
	v • d • e

energy Energy Harbor Nuclear Corp.

Beaver Valley Power Station harbor P.O. Box4 Shippingport, PA 15077 John J. Grabnar 724-682-5234 Site Vice President, Beaver Valley Nuclear November 30, 2020 L-20-162 10 CFR 50.54(f)

ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

Subject:

Beaver Valley Power Station, Unit Nos. 1 and 2 Docket No. 50-334, License No. DPR-66 Docket No. 50-412, License No. NPF-73 Supplemental Response to NRC Generic Letter 2004-02 This submittal provides a final supplemental response for Beaver Valley Power Station Unit Nos. 1 and 2 (BVPS-1 and BVPS-2), to Generic Letter (GL) 2004-02, dated September 13, 2004, "Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors."

By letter dated February 27, 2020 (Accession No. ML20030A440), the Nuclear Regulatory Commission (NRC) staff issued amendment numbers 304 and 194 to the facility operating licenses for BVPS-1 and BVPS-2, respectively. The amendments transferred operating authority for the facilities from FirstEnergy Nuclear Operating Company to Energy Harbor Nuclear Corp. The following paragraphs refer to letters regarding BVPS-1 and BVPS-2 submitted by FirstEnergy Nuclear Operating Company (FENOC).

By letter dated June 30, 2009 (Accession No. ML091830390), FENOC submitted a supplemental response to GL 2004-02 for BVPS-1 and BVPS-2. By letter dated September 28, 2010 (Accession No. ML102770023), FENOC submitted a response to the request for additional information (RAI) issued by the NRC in response to the supplemental response. The attachment to this letter documents the changes to the information provided in the supplemental response and associated RAI response that resulted from plant modifications and analysis updates implemented after September 28, 2010. The sections of the supplemental response updated in this attachment include the following:

1.0 - Overall Compliance 2.0 - General Description of and Schedule for Corrective Actions

Beaver Valley Power Station, Unit Nos. 1 and 2 L-20-162 Page 2 3.0 - Specific Information for Review Areas 3.a - Break Selection 3.b - Debris Generation / Zone of Influence (excluding coatings), including RAIs 2 through 6 3.c - Debris Characteristics 3.d - Latent Debris 3.e - Debris Transport 3.f - Head Loss and Vortexing, including RAIs 7 through 19 3.i - Debris Source Term Refinements 3.n - Downstream Effects - Fuel and Vessel 3.o - Chemical Effects, including RAI 26 3.p - Licensing Basis The remaining evaluation for BVPS-1 and BVPS-2 with respect to GL 2004-02 was the in-vessel downstream effects evaluation to demonstrate reasonable assurance that long-term core cooling could be adequately maintained for postulated accident scenarios that require sump recirculation. The in-vessel downstream effects evaluation has been completed for BVPS-1 and BVPS-2, and is documented in the attachment to this letter. This evaluation supersedes Section 3.n, Downstream Effects - Fuel and Vessel, of the June 30, 2009 supplemental response in its entirety.

By letter dated May 16, 2013 (Accession No. ML13136A144), FENOC submitted resolution plans to address Generic Safety Issue 191 (GSI-191), Assessment of Debris Accumulation on Pressurized-Water Reactor Sump Performance, for BVPS-1 and BVPS-2. FENOC selected Option 2, deterministic path, of NRC staff paper SECY-12-0093, Closure Options for Generic Safety Issue - 191, Assessment of Debris Accumulation on Pressurized-Water Reactor Sump Performance, for BVPS-1 and BVPS-2 with refinements to evaluation methods and acceptance criteria. The status of commitments identified in the May 16, 2013 GSI-191 resolution plan letter is described below.

Commitments 1 and 2 identified in the May 16, 2013 GSI-191 resolution plan letter were to address BVPS-1 pressurizer safety relief valve inlet line insulation modifications.

Commitment 1 to take measurements in preparation for the insulation modifications was completed in the fall of 2013. Commitment 2 to replace or modify insulation as appropriate will be closed without making insulation changes as the in-vessel downstream effects evaluation concluded that BVPS-1 is within fibrous debris limits with the current insulation configuration.

Commitments 3 and 6 identified in the May 16, 2013 GSI-191 resolution plan letter were to provide the NRC with the final GL 2004-02 supplemental response for BVPS-1 and BVPS-2, respectively. The attached final supplemental response to GL 2004-02 for

Beaver Valley Power Station, Unit Nos. 1 and 2 L-20-162 Page 3 BVPS-1 and BVPS-2, completes commitments 3 and 6, respectively, identified in the May 16, 2013 letter.

The BVPS-1 and BVPS-2 Updated Final Safety Analysis Report changes to reflect the current licensing basis for GSl-191 will be completed following NRC acceptance of the final docketed GSl-191 response [that is, the final GL 2004-02 supplemental response referred to above] as indicated in commitments 4 and 7, respectively, of the May 16, 2013 GSl-191 resolution plan letter.

In response to commitments 5 and 8 identified in the May 16, 2013 GSl-191 resolution plan letter, and based on plant specific accident analysis, emergency operating procedures were revised in 2013 to implement early switchover to hot leg recirculation for BVPS-2, and simultaneous hot leg and cold leg recirculation for BVPS-1, should certain plant parameters indicate that core blockage is occurring. In addition, appropriate operator training to address the emergency operating procedure revisions was completed in 2013. Therefore, commitments 5 and 8 identified in the May 16, 2013 GSl-191 resolution plan letter have been completed.

Completion of the NRC staff review of the attached response to GL 2004-02 for BVPS-1 and BVPS-2 is requested by May 31, 2021 to support concurrent implementation of containment debris limits in the Updated Final Safety Analysis Reports and related containment sump technical specifications as described in Section 3.p, "Licensing Basis," of the attachment. '

There are no regulatory commitments contained in this submittal. If there are any questions or if additional information is required, please contact Mr. Phil H. Lashley, Manager- Fleet Licensing, at (330) 696-7208.

I declare under penalty of perjury that the foregoing is true and correct. Executed on November 3 o, 2020.

Sincerely,

~

John J. Grabnar

Attachment:

Beaver Valley Power Station, Unit Nos. 1 and 2, Final Supplemental Response to Generic Letter 2004-02 cc: NRG Region I Administrator NRC Resident Inspector NRR Project Manager Director BRP/DEP Site BRP/DEP Representative

Beaver Valley Power Station, Unit Nos. 1 and 2, Final Supplemental Response to Generic Letter 2004-02 Page 1 of 59 Table of Contents 1.0 Overall Compliance 1.1 Overview of BVPS-1 and BVPS-2 Response to Generic Letter 2004-02 1.2 Correspondence Background 1.3 General Plant System Description 1.4 General Description of Containment Sump Strainers 2.0 General Description of and Schedule for Corrective Actions 3.0 Specific Information for Review Areas 3.a Break Selection 3.b Debris Generation / Zone of Influence (Excluding Coatings) 3.c Debris Characteristics 3.d Latent Debris 3.e Debris Transport 3.f Head Loss and Vortexing 3.i Debris Source Term Refinements 3.n Downstream Effects - Fuel and Vessel 3.o Chemical Effects 3.p Licensing Basis 4.0 References

Beaver Valley Power Station, Unit Nos. 1 and 2 Final Supplemental Response to Generic Letter 2004-02 Page 2 of 59 1.0 Overall Compliance NRC Issue Provide information requested in GL 2004-02, "Requested Information." Item 2(a) regarding compliance with regulations. That is, provide confirmation that the emergency core cooling system (ECCS) and containment spray system (CSS) recirculation functions under debris loading conditions are or will be in compliance with the regulatory requirements listed in the Applicable Regulatory Requirements section of this generic letter. This submittal should address the configuration of the plant that will exist once all modifications required for regulatory compliance have been made and this licensing basis has been updated to reflect the results of the analysis described above.

Energy Harbor Nuclear Corp. Response By letter dated February 27, 2020 (Reference 1), the Nuclear Regulatory Commission (NRC) staff issued amendment numbers 304 and 194 to the facility operating licenses for Beaver Valley Power Station (BVPS), Unit No. 1 (BVPS-1) and Unit No. 2 (BVPS-2),

respectively. The amendments transferred the operating authority for the facilities from FirstEnergy Nuclear Operating Company to Energy Harbor Nuclear Corp. The final supplemental response to Generic Letter (GL) 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation during Design Basis Accidents at Pressurized-Water Reactors (Reference 2), provided below refers to letters regarding BVPS-1 and BVPS-2 submitted by FirstEnergy Nuclear Operating Company (FENOC).

In accordance with Commission paper SECY-12-0093 (Reference 3) and as identified in a FENOC letter to the NRC dated May 16, 2013 (Reference 4), FENOC selected Generic Safety Issue 191 (GSI-191), Assessment of Debris Accumulation on Pressurized-Water Reactor Sump Performance, Closure Option 2 - Deterministic for BVPS-1 and BVPS-2, and identified in-vessel downstream effects as the last outstanding issue. Topical Report (TR) WCAP-17788-P, Revision 1,Comprehensive Analysis and Test Program for GSI-191 Closure (PA-SEE-1090), (References 5, 6, 7) provides evaluation methods and results to address these effects.

As discussed in NRC Technical Evaluation Report of In-Vessel Debris Effects, (Reference 8), the NRC staff has performed an independent analysis of WCAP-17788-P. Although the NRC staff did not issue a Safety Evaluation for WCAP-17788, as discussed in U.S. Nuclear Regulatory Commission Staff Review Guidance for In-Vessel Downstream Effects Supporting Review of Generic Letter 2004-02 Responses (Reference 9), the staff expects that many of the methods developed in WCAP-17788 can be used by pressurized water reactor (PWR) licensees to demonstrate adequate long-term core cooling (LTCC). Analyses completed by Energy Harbor Nuclear Corp. demonstrate compliance with 10 CFR 50.46, Acceptance Criteria for Emergency Core Cooling Systems for Light-Water Nuclear Power Plants, (b)(5), Long-Term Cooling, as it relates to in-vessel downstream debris effects for BVPS-1 and BVPS-2.

Beaver Valley Power Station, Unit Nos. 1 and 2 Final Supplemental Response to Generic Letter 2004-02 Page 3 of 59 1.1 Overview of BVPS-1 and BVPS-2 Response to Generic Letter 2004-02 By letter dated June 30, 2009 (Reference 10), FENOC submitted a supplemental response to GL 2004-02 for BVPS-1 and BVPS-2. By letter dated September 28, 2010 (Reference 11), FENOC submitted a response to the NRC Request for Additional Information (RAI) related to the supplemental response. Changes to the information provided in the supplemental response and associated RAI responses that resulted from plant modifications and analysis updates implemented after September 28, 2010 are presented in this attachment. The sections of the supplemental response updated in this attachment include the following:

1.0 - Overall Compliance 2.0 - General Description of and Schedule for Corrective Actions 3.0 Specific Information for Review Areas The June 30, 2009 supplemental response also identified some of the conservatisms in the approach to addressing GSI-191 concerns. The following conservatisms identified in the June 30, 2009 supplemental response are being revised (or supplemented).

Latent Debris Conservatism: No distinction was made regarding the type of tags and labels installed (qualified vs. unqualified).

Revision: The postulated miscellaneous debris load was reduced at BVPS-2 by removing qualified tags from the miscellaneous debris inventory. Metal tags were also removed as they will not transport to the containment sump strainers.

Net Positive Suction Head Conservatism: For BVPS-1 an additional volume of water (4,700 to 8,500 gallons) is injected from the chemical addition system. This volume is conservatively not credited for the purpose of calculating sump inventory and available net positive suction head (NPSH).

Revision: Retirement of the BVPS-1 chemical addition system removes the additional volume of water from the chemical addition tank. This conservatism no longer applies to BVPS-1.

Conservatism: Conservatively, the effects of chemical precipitates is applied to the containment sump strainer head loss at the onset of the accident.

Revision: Testing performed in WCAP-17788-P, Volume 5 (Reference 7), concludes that chemical precipitates do not form in the sump until a minimum of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after initiation of a loss-of-coolant accident (LOCA). However, head loss from chemical precipitates is conservatively applied to the containment sump strainer head loss when the containment sump temperature is reduced below 150 degrees Fahrenheit (°F) for BVPS-1 and 140°F for BVPS-2. These sump temperatures are reached at a maximum of 8.6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months (BVPS-1) and 16.7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months (BVPS-2) after initiation of a LOCA.

Beaver Valley Power Station, Unit Nos. 1 and 2 Final Supplemental Response to Generic Letter 2004-02 Page 4 of 59 Chemical Effects Conservatism: An additional conservatism that exists for chemical effects is that the chemical precipitates will not readily form until containment pool temperature has decreased below the precipitate associated value. This will not occur until later in the event at which time the containment water level will be considerably higher, providing a greater available NPSH. Chemical precipitate loading was considered at initiation of recirculation.

Revision: Head loss from chemical precipitates is added when the containment sump temperature reaches 150°F for BVPS-1 (approximately 8.6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months post-LOCA) or 140°F for BVPS-2 (approximately 16.7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months post-LOCA) rather than at the onset of the accident. This remains conservative as testing performed in WCAP-17788-P, Volume 5 (Reference 7), demonstrates that chemical precipitates do not form until a minimum of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months post-LOCA for both BVPS-1 and BVPS-2.

In addition, some of the conservatisms in the in-vessel downstream effects analysis are as follows:

The in-vessel fibrous debris acceptance criterion is based on the core inlet debris limit from WCAP-17788-P, Volume 1, Revision 1 (Reference 5). This assumes that all fibrous debris that enters the reactor vessel will accumulate at the core inlet. Some fraction of fibrous debris will pass through the core or bypass the core inlet via alternate flow paths.

The break scenario that produces the most fibrous debris for BVPS-1 is a 6-inch break in the pressurizer safety valve inlet piping. This break is evaluated for in-vessel downstream effects using the hot leg break methodology, which assumes all ECCS flow passes through the core. Due to the elevation of the break at the top of the pressurizer, a portion of the ECCS flow and the debris it contains will be diverted through the steam generator tubes, out the break, and return to the sump.

The design inputs for the BVPS-2 hot leg break methodology and BVPS-1 pressurizer safety valve piping break methodology are biased toward maximizing the injection rate of fibrous debris into the reactor vessel.

Margins exist between several BVPS-1 and BVPS-2 plant parameters and those key parameters used in the thermal-hydraulic model which established the fibrous debris limits in WCAP-17788. Therefore, BVPS-1 and BVPS-2 can tolerate fibrous debris in excess of the WCAP-17788 limits without compromising core cooling.

1.2 Correspondence Background The following table provides a listing of significant correspondence related to GL 2004-02 that was issued by the NRC or submitted by FENOC for BVPS.

Beaver Valley Power Station, Unit Nos. 1 and 2 Final Supplemental Response to Generic Letter 2004-02 Page 5 of 59 Table 1.2-1: Generic Letter 2004-02 Correspondence ADAMS Document Date Accession Document Number Nuclear Regulatory Commission (NRC) Generic Letter (GL) 2004-02, "Potential Impact of Debris September 13, 2004 ML042360586 Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors" March 4, 2005 ML050680211 FENOC Response to GL 2004-02 June 3, 2005 ML051530267 First NRC Request for Additional information (RAI)

July 22, 2005 ML052080167 First FENOC RAI Response September 6, 2005 ML052510411 Second FENOC Response to GL 2004-02 February 9, 2006 ML060380342 Second NRC RAI NRC Letter to Nuclear Energy Institute (NEI),

March 3, 2006 ML060650335 GL 2004-02 RAI Responses NRC letter to Licensees, Alternative Approach for March 28, 2006 ML060870274 Responding to NRC RAI Letter Regarding GL 2004-02 SECY-06-0078, "Status of Resolution of GSI -191, "Assessment of Effect of Debris Accumulation on March 31, 2006 ML053620174 PWR Pressurized Water Reactor Sump Performance,""

April 3, 2006 ML060960442 FENOC Supplemental Response to GL 2004-02 NRC letter to FENOC, Approves BVPS-2 Schedule May 18, 2006 ML061380273 Extension FENOC Letter to NRC, Revised Commitment Dates September 29, 2006 ML062830044 Relevant to FENOC letter dated April 3, 2006 FENOC Letter to NRC, Revised Commitment Date December 21, 2006 ML063610096 Relevant to FENOC Letter dated September 29, 2006 NRC Letter to Licensees, Alternative Approach for January 4, 2007 ML063460258 Responding to NRC RAI Letter Regarding GL 2004-02 NRC E-Mail to NEI, NRC Staff Review March 29, 2007 ML071280350 Considerations for Buffer Changes

Beaver Valley Power Station, Unit Nos. 1 and 2 Final Supplemental Response to Generic Letter 2004-02 Page 6 of 59 Table 1.2-1: Generic Letter 2004-02 Correspondence ADAMS Document Date Accession Document Number NRC Letter to NEI, Content Guide for GL 2004-02 August 15, 2007 ML071060091 Supplemental Responses NRC letter to NEI, Plant-Specific Requests for November 8, 2007 ML073060581 Extension of Time to Complete One or More Corrective Actions for GL 2004-02.

NRC Letter to NEI, Revised Content Guide for November 21, 2007 ML073110389 GL 2004-02 Supplemental Responses FENOC Letter to NRC, GL 2004-02, - Request for December 20, 2007 ML073620201 Extension of Completion Date for Corrective Actions NRC Letter to FENOC, GL 2004 Extension December 27, 2007 ML073600373 Request Approval for BVPS Units 1 and 2 FENOC Letter to NRC, GL 2004 Request for February 14, 2008 ML080510246 Extension of Completion Date for Corrective Actions NRC Letter to FENOC, Extension Request Approval February 29, 2008 ML081230116 Letter Regarding GL 2004-02 FENOC Letter to NRC, Supplemental Response to February 29, 2008 ML080660597 GL 2004-02 NRC Letter to NEI, Revised Guidance for Review of March 28, 2008 ML080230234 Final Licensee Responses to GL 2004-02 Closure FENOC Letter to NRC, GL 2004 Request for August 28, 2008 ML082480045 Extension of Completion Date for Corrective Actions NRC Letter to FENOC, Extension Request Approval September 30, 2008 ML082740241 Letter Regarding GL 2004-02 FENOC Letter to NRC, Supplemental Response to October 29, 2008 ML083080094 GL 2004-02 FENOC Letter to NRC, Supplemental BVPS-1 March 11, 2009 ML090750619 Information Regarding Response to GL 2004-02 FENOC Letter to NRC, Revised BVPS-2 March 12, 2009 ML090750618 Commitment Date Relevant to FENOC Correspondence to NRC, Dated August 28, 2008 NRC Letter to FENOC, BVPS-2 Extension Request March 31, 2009 ML090900193 Approval Letter Regarding GL 2004-02

Beaver Valley Power Station, Unit Nos. 1 and 2 Final Supplemental Response to Generic Letter 2004-02 Page 7 of 59 Table 1.2-1: Generic Letter 2004-02 Correspondence ADAMS Document Date Accession Document Number NRC Letter to FENOC, BVPS-2 Extension Request April 21, 2009 ML091050100 Approval Letter Regarding GL 2004-02 FENOC Letter to NRC, GL 2004 Request for April 30, 2009 ML091250180 Extension of Corrective Action Completion Dates NRC Letter to FENOC, BVPS-1 Extension Request May 5, 2009 ML091240030 Approval Letter Regarding GL 2004-02 FENOC Letter to NRC, Supplemental Response to June 30, 2009 ML091830390 GL 2004-02 NRC Letter to FENOC, RAI Regarding BVPS-1 and February 18, 2010 ML100290318 BVPS-2 Supplemental Response to GL 2004-02 NRC Internal Memo, Forthcoming April 21, 2010 March 31, 2010 ML100840756 Teleconference with FENOC Regarding GL 2004-02 Response for BVPS-1 and BVPS-2 NRC Summary of April 21, 2010, Category 1 May 18, 2010 ML101320665 Teleconference with FENOC on GL 2004-02 FENOC Letter to NRC, Response to RAI Related to September 28, 2010 ML102770023 GL 2004-02 FENOC Letter to NRC, Generic Safety Issue 191 May 16, 2013 ML13136A144 Resolution Plan 1.3 General Plant System Description Description of BVPS-1 Plant Systems BVPS Unit 1 is a Westinghouse Electric Company LLC (Westinghouse) three loop PWR design. The nuclear steam supply system consists of a reactor vessel, three steam generators, three reactor coolant pumps, one pressurizer, and reactor coolant system (RCS) piping.

A reinforced concrete primary shield wall forms the reactor cavity at the center of the containment structure. Located concentrically to the primary shield wall is the reinforced concrete crane wall. Extending approximately radially between the primary shield wall and the crane wall are reinforced concrete walls which separate the internals into cubicles. Each reactor coolant system loop (consisting of a steam generator, reactor coolant pump, and connecting piping) and pressurizer are located in this area within separate cubicles.

Beaver Valley Power Station, Unit Nos. 1 and 2 Final Supplemental Response to Generic Letter 2004-02 Page 8 of 59 Following a LOCA, the low head safety injection (LHSI), high head safety injection (HHSI), and quench spray (QS) pumps are automatically started. Initially, two LHSI and two HHSI pumps take suction from the refueling water storage tank (RWST) and discharge into the RCS cold legs. Two QS pumps each take suction from the RWST and discharge to a separate 360 degree spray ring header located approximately 96 feet above the containment building operating floor. When the RWST level reaches the low level set point, four recirculation spray (RS) pumps automatically start. Each pump takes suction from the containment sump and discharges into a separate 180 degree spray ring header located approximately 80 feet above the containment building operating floor.

When the RWST level reaches the extreme low level setpoint, the transfer to safety injection recirculation is initiated. The LHSI pump suction flow paths are realigned to take suction from the containment sump, while the HHSI pump flow paths take suction from the LHSI pump discharge. Prior to the switchover, two recirculation spray pumps are secured to limit head loss across the sump strainers.

Approximately 6.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months after the loss of coolant accident, the cold leg only recirculation mode will be terminated, and the simultaneous cold leg and hot leg recirculation mode initiated. The change-over is accomplished by realigning the LHSI pumps to deliver to the hot legs. During the change-over, the HHSI pumps will continue injection to the cold legs.

Schematics for the BVPS-1 systems that perform the ECCS and CSS functions are provided in Section 3.f.1 of the June 30, 2009 supplemental response. An updated schematic of the quench spray system is provided in Section 3.f.1 of this response; this schematic accounts for retirement of the sodium hydroxide chemical addition system.

Description of BVPS-2 Plant Systems BVPS Unit 2 is a Westinghouse three loop PWR design. The nuclear steam supply system consists of a reactor vessel, three steam generators, three reactor coolant pumps, one pressurizer, and RCS piping.

A reinforced concrete primary shield wall forms the reactor cavity at the center of the containment structure. Located concentrically to the primary shield wall is the reinforced concrete crane wall. Extending approximately radially between the primary shield wall and the crane wall are reinforced concrete walls which separate the internals into cubicles. Each reactor coolant system loop (consisting of a steam generator, reactor coolant pump, and connecting piping) and pressurizer are located in this area within separate cubicles.

Following a LOCA, the injection mode of ECCS operation is initiated automatically and requires no operator action. During injection, the LHSI and HHSI/charging pumps take suction from the RWST and deliver borated water to the cold legs.

Each of the two QS pumps take suction from the RWST and discharge into separate 360-degree spray rings near the top of containment. When the RWST level reaches

Beaver Valley Power Station, Unit Nos. 1 and 2 Final Supplemental Response to Generic Letter 2004-02 Page 9 of 59 the low level setpoint, RS pumps start automatically and provide recirculation of the containment sump water through the RS coolers to the RS headers. Each 360-degree spray ring header is fed by two risers, where each riser originates from one of the recirculation coolers.

Switchover from injection to cold leg recirculation occurs automatically as the level in the RWST drops to the extreme low-level set point. The LHSI pumps are stopped, while the discharge flow paths for two RS pumps are automatically realigned to the RCS cold legs and the HHSI pump suctions.

After the transfer to recirculation takes place and approximately six hours after the LOCA occurs, the operators initiate hot leg recirculation, at which time all ECCS flow paths are re-aligned from the cold legs to the hot legs. The alignment is switched between hot leg and cold leg recirculation 9.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months after the initial transfer to hot leg recirculation.

Schematics for the BVPS-2 systems that perform the ECCS and CSS functions are provided in Section 3.f.1 of the June 30, 2009 supplemental response (Reference 10).

An updated schematic of the quench spray system is provided in Section 3.f.1 of this letter; this schematic accounts for retirement of the sodium hydroxide chemical addition system.

1.4 General Description of Containment Sump Strainers The BVPS-1 containment sump strainer assembly engineered and manufactured by Control Components Incorporated (CCI), provides 3,493 square feet (ft²) of strainer area through the use of 1/16-inch perforations. The strainer assembly has 13 cassette-type strainer modules, located near the outside wall of containment. The two strings of modules provide clean containment water flow to a channel box. The channel boxes are connected to each of the module strings and are connected to a common channel box which enters the suction box surrounding the containment sump. Refer to Section 3.j.1 of the June 30, 2009 supplemental response for detailed descriptions of the BVPS-1 strainer assembly and components.

The BVPS-2 containment sump strainer assembly, engineered by Enercon and fabricated by Transco, provides 3,396 ft² of strainer area through the use of 3/32-inch perforations. The strainer assembly consists of 113 top-hat strainers, placed in a rectangular grid arrangement, located near the outside wall of containment. The top-hats are constructed of a series of perforated plate tubes. Debris eliminator mesh is installed between the perforated plates to prevent excessive fibrous debris from passing through the strainers. Refer to Section 3.j.1 of the June 30, 2009 supplemental response for detailed descriptions of the BVPS-2 strainer assembly and components.

The surface areas for the containment sump strainers are summarized below.

Beaver Valley Power Station, Unit Nos. 1 and 2 Final Supplemental Response to Generic Letter 2004-02 Page 10 of 59 Table 1.4-1: Containment Sump Strainer Parameters Strainer Manufacturer Strainer Type Surface Area (ft2)

BVPS-1 Strainer CCI Cassette 3,493 BVPS-2 Strainer Enercon and Top Hat 3,396 Transco 2.0 General Description of and Schedule for Corrective Actions NRC Issue Provide a general description of actions taken or planned, and dates for each. For actions planned beyond December 31, 2007, reference approved extension requests or explain how regulatory requirements will be met as per "Requested Information" Item 2(b). That is provide a general description of and implementation schedule for all corrective actions, including any plant modifications, that you identified while responding to this generic letter. Efforts to implement the identified actions should be initiated no later than the first refueling outage starting after April 1, 2006. All actions should be completed by December 31, 2007. Provide justification for not implementing the identified actions during the first refueling outage starting after April 1, 2006. If all corrective actions will not be completed by December 31, 2007, describe how the regulatory requirements discussed in the Applicable Regulatory Requirements section will be met until the corrective actions are completed.

Energy Harbor Nuclear Corp. Response Energy Harbor Nuclear Corp. has performed analyses to determine the susceptibility of the ECCS and CSS recirculation functions for BVPS-1 and BVPS-2 to the adverse effects of post-accident debris blockage and operation with debris-laden fluids. These analyses conform to the guidance report NEI 04-07 methodology (Reference 12) as approved by the NRC Safety Evaluation dated December 6, 2004 (Reference 13), with certain exceptions as described in the June 30, 2009 supplemental response. FENOC has completed the following GL 2004-02 actions, analyses and modifications.

The following are plant modifications, program changes, analyses, and licensing basis changes completed prior to the June 30, 2009 supplemental response:

Strainer replacements have been installed at both units. At BVPS-2, the new replacement strainer, which increased the available surface area from approximately 150 ft² to 3,396 ft², was installed during the fall 2006 refueling outage (2R12). At BVPS-1, the new replacement strainer, which increased the available surface area from approximately 130 ft² to 3,493 ft², was installed during the fall 2007 refueling outage (1R18). As part of this modification, bellmouth transition pieces were installed on ends of the pump suction piping (for two outside containment RSS pumps and two LHSI pumps at BVPS-1, and four RSS pumps at BVPS-2) in the containment sumps to increase available NPSH.

Beaver Valley Power Station, Unit Nos. 1 and 2 Final Supplemental Response to Generic Letter 2004-02 Page 11 of 59

Replacement of BVPS-1 and modification of BVPS-2 high pressure safety injection cold leg throttle valves have been completed to increase the throttle valve gap and thereby reduce flow restrictions that may become blocked by debris. BVPS-1 modifications were performed in the fall 2007 refueling outage (1R18). BVPS-2 modifications were performed in the spring 2008 refueling outage (2R13).

The BVPS-1 and BVPS-2 start signal for the RS pumps has been changed from a fixed time delay to an engineered safety features actuation system signal based on a RWST level low coincident with a containment pressure high-high signal to allow sufficient pool depth to cover the sump strainer before initiating recirculation flow.

These start signal changes required certain technical specification changes that were found acceptable by the NRC as described in BVPS-1 License Amendment No. 280 (issued by the NRC on October 5, 2007, Reference 14), and BVPS-2 License Amendment No. 164 (issued by the NRC on March 11, 2008, Reference 15).

Borated Temp-MatTM insulation encapsulated in reflective metal insulation (RMI) on the BVPS-1 reactor vessel closure head has been replaced with RMI during the spring 2006 refueling outage (1R17) to reduce debris loading on the sump strainer.

New RMI was installed on the BVPS-1 replacement steam generators and associated piping in the vicinity of the replacement steam generators during the spring 2006 refueling outage (1R17). This resulted in a reduced quantity of insulation that could contribute to debris loading on the sump strainer.

Borated Temp-MatTM insulation encapsulated in RMI on the BVPS-2 reactor vessel closure head flange has been replaced with RMI during the spring 2008 refueling outage (2R13) to reduce debris loading on the sump strainer. Min-KTM insulation encapsulated in RMI on portions of the BVPS-2 reactor coolant system piping has been replaced with Transco Thermal WrapTM insulation encapsulated in RMI during the spring 2008 refueling outage (2R13) to reduce debris loading on the sump strainer.

A containment coatings inspection and assessment program and a containment cleaning program became effective for BVPS in April of 2008 and apply to refueling outages beginning with the BVPS-2 spring 2008 (2R13) and BVPS-1 spring 2009 (1R19) refueling outages.

BVPS-1 and BVPS-2 reactor cavity drain cross bars that have the potential to collect debris and block water flow to the containment sump were removed during the fall 2007 (1R18) and spring 2008 (2R13) refueling outages.

Various portions of fibrous and particulate insulation in the BVPS-1 RCS loop cubicles were replaced with RMI during the spring 2009 refueling outage (1R19).

This included insulation on the reactor coolant loop piping.

Iodine filters, containing a significant amount of thin aluminum that would have been submerged, were removed from the BVPS-1 containment during the spring 2009

Beaver Valley Power Station, Unit Nos. 1 and 2 Final Supplemental Response to Generic Letter 2004-02 Page 12 of 59 refueling outage (1R19) and from BVPS-2 containment during the spring 2008 refueling outage (2R13).

LOCADM analyses were conducted for both BVPS-1 and BVPS-2 in accordance with WCAP-16793-NP, Revision 2 (Reference 16). Both BVPS-1 and BVPS-2 satisfy the maximum clad temperature and total deposition thickness limits of WCAP-16793-NP.

License Amendment No. 167 for BVPS-2, issued March 26, 2009 (Reference 17),

authorized changes to the licensing basis as described in the BVPS-2 Updated Final Safety Analysis Report regarding the method of calculating the NPSH available to the RS pumps by crediting containment overpressure. The licensing basis was revised April 7, 2009.

Tests and Evaluations

Walkdowns were performed to identify potential debris sources in containment.

Inventories of insulation, unqualified coatings, aluminum, and latent debris were created from the information gathered.

Debris generation and transport analyses were performed to determine the quantities of debris that may reach the sump strainers following a variety of LOCA scenarios.

Strainer prototype debris and chemical effects testing of the new strainer designs were performed (during the time periods April to June of 2008 and June 2010 for BVPS-1, and October to November 2008 and August 2009 for BVPS-2) to determine the head loss across the strainer modules. Test results were used to verify that adequate NPSH is available to the ECCS and RS pumps following a LOCA.

Prototype strainer bypass testing was performed for the new BVPS-1 strainer design in May 2008 and for the new BVPS-2 strainer design in November and December of 2008. This testing determined the percentage of fibrous debris that penetrates the strainers.

Downstream effects analyses for BVPS-1 and BVPS-2 were performed to verify debris-induced blockage of tight clearances, abrasive wear of rotating pump components, and erosive wear of valve components does not occur for ECCS and RSS flow path components.

Air ingestion due to deaeration, flashing, and vortex formation associated with flow through the BVPS-1 and BVPS-2 containment sump strainers was evaluated and quantified to determine the impact on the required NPSH available to the downstream pumps.

The following are GL 2004-02 actions, analyses and modifications made following the June 30, 2009 supplemental response:

Beaver Valley Power Station, Unit Nos. 1 and 2 Final Supplemental Response to Generic Letter 2004-02 Page 13 of 59

Temp-MatTM fibrous insulation installed on the BVPS-1 reactor vessel inlet and outlet nozzles was replaced with RMI during the fall 2010 refueling outage (1R20).

Various portions of fibrous and particulate insulation in the BVPS-2 RCS loop cubicles were replaced with RMI during the fall 2009 refueling outage (2R14).

Additional portions of calcium silicate insulation were removed from the BVPS-1 RCS loop cubicles during the fall 2010 refueling outage (1R20).

Min-KTM insulation on the BVPS-2 steam generator level instrumentation tubing was replaced with jacketed Thermal WrapTM insulation during the spring 2011 (2R15) refueling outage. The Min-KTM insulation was inside an RCS loop break zone of influence while the jacketed Thermal WrapTM insulation on the instrumentation tubing for the steam generators is outside the zone of influence.

NUKON fibrous insulation blankets installed on the BVPS-2 pressurizer power operated relief valves, and associated inlet piping and pipe supports (hereafter referred to as pressurizer power operated relief valve inlet piping) was replaced by jacketed Thermal WrapTM insulation secured with Sure-Hold banding where practical during the spring 2011 refueling outage (2R15).

Additional portions of calcium silicate insulation were removed from the BVPS-2 RCS loop cubicles during the spring 2011 refueling outage (2R15).

The BVPS-1 and BVPS-2 liquid sodium hydroxide chemical addition systems used for containment sump pH control were replaced with baskets of powdered sodium tetraborate located at the lowest floor in each containment. This results in a reduced quantity of chemical precipitates that may accumulate on the sump strainers following a loss-of-coolant accident. NRC letters dated April 16, 2009 (Reference 18) and March 14, 2012 (Reference 19) issued license amendments 168 and 289 for BVPS-2 and BVPS-1, respectively. The amendments revised technical specifications to address requirements for the new containment sump pH control systems that use sodium tetraborate.

BVPS-2 has six manway hatches between the refueling cavity and reactor cavity, three of which are normally sealed shut during operation with the remaining three covered with course grating. Procedures were modified to install grating over all six hatches after refueling (beginning in fall 2009 refueling outage - 2R14) to provide additional pressure relief pathways for a reactor vessel nozzle break.

A re-evaluation of the quantity of tape in BVPS-1 containment was performed, reducing the debris load from unqualified tags, labels, and tape by 202 ft², or 37 percent (%).

A significant portion of the unqualified tags, labels, and tape in BVPS-2 containment were removed during the fall 2009 refueling outage (2R14) or shown to be qualified for the post-LOCA containment environment. The debris load for unqualified tags, labels, and tape was reduced by 633 ft², or 85%.

Beaver Valley Power Station, Unit Nos. 1 and 2 Final Supplemental Response to Generic Letter 2004-02 Page 14 of 59

Emergency operating procedures for BVPS-1 were revised to secure two RS pumps prior to the transfer to ECCS recirculation. This prevents excessive head loss across the strainers due to the sump flow rate exceeding the flow rate analyzed in strainer head loss testing.

Emergency operating procedures for BVPS-2 were revised to shut down one of the RS pumps supplying the spray header when the containment pressure is reduced below a predetermined value. This lowers the pressure differential on the strainers during the time of greatest structural loading.

Emergency operating procedures for BVPS-1 and BVPS-2 were revised to initiate early transfer to hot leg recirculation (BVPS-2) or simultaneous hot and cold leg recirculation (BVPS-1) should plant parameters indicate that core blockage is occurring.

An evaluation of the effects of fibrous debris transported to the reactor vessel was performed in accordance with WCAP-17788-P, Volume 1, Revision 1 (Reference 5).

Quantities of fibrous debris are within WCAP limits, and a core cooling flow path is verified to be available at all times following a design basis LOCA.

Energy Harbor Nuclear Corp. has no outstanding corrective actions associated with GL 2004-02 for BVPS-1 or BVPS-2.

3.0 Specific Information for Review Areas As stated in the June 30, 2009 supplemental response as well as subsequent RAI responses submitted on September 28, 2010 (Reference 11), Energy Harbor Nuclear Corp. has addressed review areas 3.a through 3.p. This submittal provides new, revised, or supplemental information for the following listed review areas. Section 3.n of this submittal replaces, in its entirety, Section 3.n of the BVPS supplemental response dated June 30, 2009.

3.a - Break Selection 3.b - Debris Generation / Zone of Influence (Excluding Coatings), including RAIs 2 through 6 3.c - Debris Characteristics 3.d - Latent Debris 3.e - Debris Transport 3.f - Head Loss and Vortexing, including RAIs 7 through 19.

3.i - Debris Source Term Refinements 3.n - Downstream Effects - Fuel and Vessel 3.o - Chemical Effects, including RAI 26 3.p - Licensing Basis

Beaver Valley Power Station, Unit Nos. 1 and 2 Final Supplemental Response to Generic Letter 2004-02 Page 15 of 59 3.a Break Selection The BVPS-2 plant modification discussed in the September 28, 2010 RAI 1 response has been completed. This modification replaced NUKON insulation on the BVPS-2 pressurizer power operated relief valve inlet piping with stainless steel jacketed Thermal WrapTM insulation secured with Sure-Hold bands where practical.

3.b Debris Generation / Zone of Influence (Excluding Coatings)

Assumptions utilized for Transco Thermal WrapTM insulation in the debris generation analysis are updated as follows:

Transco Thermal WrapTM has been used in BVPS-2 as a replacement for Min-KTM in the loop compartments. In the upper pressurizer cubicle, Transco Thermal WrapTM is installed on the pressurizer power operated relief valve inlet piping; the majority of which is jacketed. Transco Thermal WrapTM is a low density fiberglass insulation with a density of 2.4 lb/ft3 that is equivalent to NUKON. Therefore, the material characteristics for NUKON are assumed. A 17.0D zone of influence is used; however, a reduced zone of influence of 2.4D is credited for jacketed Thermal WrapTM secured with Sure-Hold banding, consistent with the NEI 04-07 Safety Evaluation Report (Reference 13).

BVPS-1 Table 3.b-2 is updated to show the BVPS-1 insulation debris quantities following the revised debris generation analysis, including evaluation of the pressurizer safety valve inlet piping break and implementation of the fall 2010 insulation removal modifications.

The pressurizer safety valve inlet piping break was previously described in Appendix V, BVPS-1 June 2010 Test 7 Summary, of the RAI response letter dated September 28, 2010 (Reference 11).

Table 3.b-2: BVPS-1 Insulation Debris Quantities Pressurizer 6-inch SIS (3) 6-inch Material Loop RPV (2)

Surge Line Injection PSV (4) Inlet Types(1) LBLOCA(1) Nozzle Break Break Point Piping RMI 19,523 ft2 5,224 ft2 4,921 ft² 18,241 ft² 6,264 ft² Temp-MatTM 4.4 ft3 0 ft3 0 ft³ 3.9 ft³ 44.0 ft³ Cal-Sil 42.0 lb. 0 lb. 96.0 lb. 3.0 lb. 0 lb.

Min-KTM 0 lb. 4.8 lb. 16.0 lb. 0 lb. 0 lb.

Foamglas 46.5 lb. 0 lb 0 lb. 0 lb. 0 lb.

Notes:

1. Break locations were evaluated for hot leg, cold leg, and the cross-over leg; with the limiting values presented as loop large break loss of coolant accident (LBLOCA).

2. Reactor pressure vessel (RPV)

Beaver Valley Power Station, Unit Nos. 1 and 2 Final Supplemental Response to Generic Letter 2004-02 Page 16 of 59

3. Safety injection system (SIS)

4. Pressurizer safety valve (PSV)

5. RMI is reflective metal insulation, and Cal-Sil is calcium silicate BVPS-2 As previously discussed, the insulation replacement on the pressurizer power operated relief valve inlet piping has been implemented. NUKON insulation has been replaced with jacketed Transco Thermal WrapTM. The replacement insulation is secured with Sure-Hold banding, where practical. A 2.4D zone of influence is applied to the Thermal WrapTM secured with Sure-Hold banding, while a 17.0D zone of influence is applied to the other low-density fiberglass insulation in the area (including Fiberglas Thermal Insulating Wool [TIW] on the pressurizer spray line, which is outside the zone of influence).

Table 3.b-3 is updated to show the BVPS-2 insulation debris quantities following the revised debris generation analysis, including evaluation of the pressurizer power operated relief valve piping break and implementation of the fall 2009 and spring 2011 insulation replacement/removal modifications.

Table 3.b-3: BVPS-2 Insulation Debris Quantities Material Types Loop RPV Pressurizer 6-inch SIS 6-inch LBLOCA (1) Nozzle Surge Line Injection PORV (2)

Break Break Point Inlet RMI 32,137 ft² 11,106 ft² 4,431 ft² 4,993 ft² 3,670 ft² Thermal WrapTM 1.5 ft³ 0 ft³ 0 ft³ 0.1 ft³ 16.0 ft³ Damming 0.1 ft³ 0 ft³ 0 ft³ 0.1 ft³ 0 ft³ Material Temp-MatTM 7.4 ft³ 0 ft³ 0 ft³ 1.0 ft³ 0 ft³ Cal-Sil 57.0 lb. 0 lb. 3.0 lb. 0 lb. 0 lb.

Min-KTM 2.4 lb. 0 lb. 10.4 lb. 0 lb. 0 lb.

Microtherm 0 lb. 303.0 lb. 0 lb. 0 lb. 0 lb.

Notes:

1. Break locations were evaluated for Hot Leg, Cold Leg, and the Cross-over Leg; with the limiting values presented as Loop Large Break Loss of Coolant Accident (LBLOCA).

2. Pressurizer Power Operated Relief Valve (PORV)

Two additional debris sources were identified at BVPS-2.

Rubber insulation was installed on the reactor coolant pump motor leads in each reactor coolant pump cubicle. This insulation was installed to protect the conduits

Beaver Valley Power Station, Unit Nos. 1 and 2 Final Supplemental Response to Generic Letter 2004-02 Page 17 of 59 from damage due to contact with nearby platforms. An RCS loop piping break in the loop cubicles may dislodge or destroy this insulation.

Each hot leg and cold leg pipe is wrapped in borated silicone rubber sheets in order to reduce neutron radiation passing through the openings in the reactor cavity shield wall. An RCS piping break inside the reactor cavity may dislodge or destroy this material.

The debris category of thick elastic materials is created to include both potential debris sources, as the debris generation and transport characteristics of these debris sources are unique when compared to existing categories.

A LBLOCA inside the loop cubicle is assumed to destroy the reactor coolant pump conduit insulation installed in the corresponding cubicle. An RPV nozzle break inside the reactor cavity is assumed to destroy the silicone neutron shielding on the broken nozzle and two adjacent nozzles, crediting the reactor vessel as a robust barrier.

Surface areas of destroyed materials are summarized below.

Table 3.b-4: Debris Generation - Thick Elastic Materials Debris Surface Break Scenario Debris Source Area Reactor Coolant Pump Loop LBLOCA 7.4 ft2 Conduit Insulation RPV Nozzle Silicone Neutron 104.4 ft2 Break Shielding 3.c Debris Characteristics The following paragraph provides a description of new debris sources evaluated since 2009.

Thermal WrapTM installed on BVPS-2 pressurizer power operated relief valve inlet piping The size distribution applied to Thermal WrapTM secured with Sure-Hold banding is 100% fines, given the forces associated with the 2.4D zone of influence. Thermal WrapTM not secured with Sure-Hold banding (17.0D zone of influence) is assumed to fail as 60% fines, 40% large pieces, consistent with the size distribution recommended by guidance report NEI 04-07 (Reference 12) for Transco fiber blankets.

Thick Elastic Materials Based on the July 10, 2008 supplemental GL 2004-02 response for the Diablo Canyon Power Plant (Reference 20), silicone rubber insulation materials have a macroscopic density of 58 pounds mass (lbm) per cubic foot (ft³) and are assumed to fail as small

Beaver Valley Power Station, Unit Nos. 1 and 2 Final Supplemental Response to Generic Letter 2004-02 Page 18 of 59 pieces ranging in size from -inch by -inch to 1/2-inch by 1/2-inch; small enough to be transportable but large enough to block holes in the sump screens. This size range is appropriate for the BVPS-2 silicone rubber neutron shielding, as it is composed of concentric 1/2-inch thick sheets wrapped around the pipe.

3.d Latent Debris The following paragraphs describe activities implemented since 2009 that reduced the latent debris load.

Tape and Equipment Labels BVPS-1 The BVPS-1 latent debris walkdown identified approximately 207 ft² of tape in containment. This walkdown was performed during a refueling outage, when large amounts of tape are used during outage work activities. The walkdown report did not distinguish tape used during power operations from tape used for outage-related activities and thus likely counted tape that would be removed prior to startup.

It is highly unlikely that the quantity of tape observed during the latent debris walkdown would be left in containment during power operation. There are a number of controls in place and reasons to conclude that tape counted during the walkdown is not representative of operating plant conditions.

1. The containment closeout inspection ensures that loose material is removed prior to operation.

2. The containment cleaning program undertaken in response to the containment sump issue has cleaned all accessible areas of containment since the original walkdown.

3. Since the original walkdown, only red tape is used in containment to increase visibility and decrease the likelihood it will remain in containment during operations.

4. As discussed below, laser scans of the containment interior were reviewed and identified very little tape in containment.

The containment closeout inspection procedures verify loose material is removed prior to operation, and states in part:

The purpose of this procedure is to verify that all debris and unauthorized, non-permanently mounted equipment or material has been removed from Containment. . . .

A visual inspection verifies that no loose debris (rags, trash, clothing, etc) is present in Containment which could be transported to the containment sump and cause restriction of the pump suctions during LOCA conditions. The inspection is required:

Beaver Valley Power Station, Unit Nos. 1 and 2 Final Supplemental Response to Generic Letter 2004-02 Page 19 of 59

1. For all accessible area[s] of the Containment prior to establishing Containment Operability. . . .

The following items should be checked during the containment tour, based on past surveillance performances: . . .

General area cleanliness, all tape removed . . .

Laser scans of the BVPS-1 containment interior were utilized to identify tape left inside containment following the closeout inspection. These scans provide 360-degree pictures of the containment interior and are of sufficient detail to provide dimensional information to support construction of the replacement sump strainers. The lower containment (Elevation 692) was reviewed, with approximately 66% of this area visible from the scans.

Review of the laser scans identified that approximately 5% of the tape identified in the walkdown remained in containment following the closeout inspection. This accounted for the areas not visible from the scans and assumed 75% of the tape in the visible areas was identified. This review concluded that a value of 25% of the original tape allowance, or 51.9 ft² of tape, is appropriate.

Accounting for the 30% increase applied to all miscellaneous debris, the BVPS-1 tape reduction evaluation reduced the miscellaneous debris load from 543 ft² to 341 ft², a reduction of 202 ft² or 37%.

BVPS-2 When the original walkdown was conducted, no information was located to justify the qualification status of the tags and labels listed in the walkdown report. Electromark conduit labels, ID triangle markers, and cable tray labels were later determined to be qualified for the post-LOCA containment environment. These labels were removed from the miscellaneous debris inventory following the 2R14 refueling outage, resulting in a surface area reduction in the tag and label inventory of 395.4 ft².

A considerable portion of the miscellaneous debris inventory consisted of nuclear instrumentation system (NIS) labels. These labels were applied to the nuclear instrumentation conduit and junction boxes during construction to provide a visual aid for cable installation but are no longer used. During the fall 2009 refueling outage (2R14), the accessible NIS labels were removed from containment, reducing the NIS label surface area from 241.9 ft² to 34.1 ft². The 30% margin applied to other tags and labels was not applied to the NIS labels since an exact count of the labels remaining in containment was made.

Additional tag and label removal efforts in the fall 2009 refueling outage include the following.

Adhesive markers used during pre-operational testing

Beaver Valley Power Station, Unit Nos. 1 and 2 Final Supplemental Response to Generic Letter 2004-02 Page 20 of 59

Electrical tape used to outline penetrations

Hand-marked tape labels

Unqualified plastic tags (larger tags replaced with qualified equipment labels)

The combined total of miscellaneous debris removed from containment is 29.9 ft²,

including the 30% increase applied to all miscellaneous debris.

Metal tags were included in the original miscellaneous debris inventory. Strainer head loss testing did not include RMI debris because even small pieces of RMI will not remain suspended in the recirculation pool; debris must remain suspended in order to reach the sump strainers as they are raised above the containment floor. Since metal tags are thicker than small RMI debris, they have a higher settling velocity and will not remain suspended or be transported to the strainer modules. Metal tags were removed from the debris inventory and reduced the miscellaneous debris load by 67.9 ft².

A summary of the miscellaneous debris reduction activities at BVPS-2 is provided in the table below:

Table 3.d-5: Summary of BVPS-2 Latent Debris Reduction Debris Source Reason for Debris Reduction Surface Area Reduction Conduit Labels Qualified for Containment Environment 178.0 ft² ID Triangle Markers Qualified for Containment Environment 108.1 ft² Cable Tray Labels Qualified for Containment Environment 109.3 ft² NIS Labels Removed from Containment 207.8 ft² Metal Tags Does not transport to strainers 67.9 ft² Other Tags and Labels Removed from Containment 26.3 ft² Total 697.4 ft2 The BVPS-2 miscellaneous debris inventory is currently 59 ft², which includes the changes described above and some minor additions to the inventory since 2009.

3.e Debris Transport Due to the revision of the debris generation tables as described above in Section 3.b, the BVPS-1 debris transport tables 3.e-6 through 3.e-9 and BVPS-2 debris transport tables 3.e-10 through 3.e-13 have been revised to reflect the revised quantities of relevant debris types.

Beaver Valley Power Station, Unit Nos. 1 and 2 Final Supplemental Response to Generic Letter 2004-02 Page 21 of 59 Table 3.e-9a is added to show the overall debris transport for the 6-inch pressurizer safety valve inlet piping break at BVPS-1, and Table 3.e-14 was added to show the overall debris transport for the 6-inch power operated relief valve inlet piping break at BVPS-2.

BVPS-1 Table 3.e-6 Overall Debris Transport (Bounding RCS Loop Break) - BVPS-1 Debris Debris Debris Debris Type Debris Size Quantity Transport Quantity at Generated Fraction Sump Small Pieces 13,861.6 ft2 47% 6,515.0 ft2

(<4 inch)

RMI Large Pieces 5,661.8 ft2 1% 56.6 ft2 (4 inch)

Total 19,523.4 ft2 34% 6,571.6 ft2 Fines 0.89 ft3 100% 0.89 ft3 Small Pieces 3.54 ft3 100% 3.54 ft3

(<6 inch)

Temp-Mat' Large Pieces 0 ft3 10% 0 ft3 (6 inch)

Intact Pieces 0 ft3 0% 0 ft3 (6 inch)

Total 4.4 ft3 100% 4.4 ft3 Foamglas Total (Fines) 46.5 lbm 100% 46.5 lbm Cal-Sil Total (Fines) 42.0 lbm 100% 42.0 lbm Coatings Inside ZOI Total (Fines) 194.9 lbm 100% 194.9 lbm Exposed Unqualified Total (Fines) 66.2 lbm 100% 66.2 lbm Coatings Outside ZOI Dirt/Dust Total (Fines) 134.7 lbm 100% 134.7 lbm Latent Fiber Total (Fines) 9.9 ft3 100% 9.9 ft3 Miscellaneous Debris Total 341 ft2 100% 341 ft2

Beaver Valley Power Station, Unit Nos. 1 and 2 Final Supplemental Response to Generic Letter 2004-02 Page 22 of 59 Table 3.e-7 Overall Debris Transport (Pressurizer Surge Line Break) - BVPS-1 Debris Debris Debris Debris Type Debris Size Quantity Transport Quantity at Generated Fraction Sump Small Pieces 3,494.0 ft2 100% 3,494.0 ft2

(<4 inch)

RMI Large Pieces 1,427.2 ft2 100% 1,427.2 ft2

(>4 inch)

Total 4,921.2 ft2 100% 4,921.2 ft2 Cal-Sil Total (Fines) 96 lbm 100% 96 lbm Min-KTM Total (Fines) 16.0 lbm 100% 16.0 lbm Coatings Inside ZOI Total (Fines) 8.2 lbm 100% 8.2 lbm Exposed Unqualified Total (Fines) 66.2 lbm 100% 66.2 lbm Coatings Outside ZOI Dirt/Dust Total (Fines) 134.7 lbm 100% 134.7 lbm Latent Fiber Total (Fines) 9.9 ft3 100% 9.9 ft3 Miscellaneous Debris Total 341 ft2 100% 341 ft2

Beaver Valley Power Station, Unit Nos. 1 and 2 Final Supplemental Response to Generic Letter 2004-02 Page 23 of 59 Table 3.e-8 Overall Debris Transport (Reactor Vessel Nozzle Break) - BVPS-1 Debris Debris Debris Debris Type Debris Size Quantity Transport Quantity at Generated Fraction Sump Small Pieces 3,708.7 ft2 100% 3,708.7 ft2

(<4 inch)

RMI Large Pieces 1,514.9 ft2 100% 1,514.9 ft2

(>4 inch)

Total 5,223.6 ft2 100% 5,223.6 ft2 Fines 0 ft3 100% 0 ft3 Small Pieces 0 ft3 100% 0 ft3

(<6 inch)

Temp-Mat' Large Pieces 0 ft3 100% 0 ft3

(>6 inch)

Intact Pieces 0 ft3 100% 0 ft3

(>6 inch)

Total 0 ft3 100% 0 ft3 Cal-Sil Total (Fines) 0 lbm 100% 0 lbm Min-K' Total (Fines) 4.8 lbm 100% 4.8 lbm Coatings Inside ZOI Total (Fines) 112.9 lbm 100% 112.9 lbm Exposed Unqualified Total (Fines) 66.2 lbm 100% 66.2 lbm Coatings Outside ZOI Dirt/Dust Total (Fines) 134.7 lbm 100% 134.7 lbm Latent Fiber Total (Fines) 9.9 ft3 100% 9.9 ft3 Miscellaneous Debris Total 341 ft2 100% 341 ft2

Beaver Valley Power Station, Unit Nos. 1 and 2 Final Supplemental Response to Generic Letter 2004-02 Page 24 of 59 Table 3.e-9 Overall Debris Transport (6-inch SIS Line Break) - BVPS-1 Debris Debris Debris Debris Type Debris Size Quantity Transport Quantity at Generated Fraction Sump Small Pieces 12,951.3 ft2 100% 12,951.3 ft2

(<4 inch)

RMI Large Pieces 5,289.9 ft2 100% 5,289.9 ft2

(>4 inch)

Total 18,241.2 ft2 100% 18,241.2 ft2 Fines 0.8 ft3 100% 0.8 ft3 Small Pieces 3.1 ft3 100% 3.1 ft3

(<6 inch)

Temp-Mat' Large Pieces 0 ft3 100% 0 ft3

(>6 inch)

Intact Pieces 0 ft3 100% 0 ft3

(>6 inch)

Total 3.9 ft3 100% 3.9 ft3 Cal-Sil Total (Fines) 3 lbm 100% 3 lbm Min-K' Total (Fines) 0 lbm 100% 0 lbm Coatings Inside ZOI Total (Fines) 10.6 lbm 100% 10.6 lbm Exposed Unqualified Total (Fines) 66.2 lbm 100% 66.2 lbm Coatings Outside ZOI Dirt/Dust Total (Fines) 134.7 lbm 100% 134.7 lbm Latent Fiber Total (Fines) 9.9 ft3 100% 9.9 ft3 Miscellaneous Debris Total 341 ft2 100% 341 ft2

Beaver Valley Power Station, Unit Nos. 1 and 2 Final Supplemental Response to Generic Letter 2004-02 Page 25 of 59 Table 3.e-9a Overall Debris Transport (6-inch Pressurizer Safety Valve Inlet Line Break) -

BVPS-1 Debris Debris Debris Debris Type Debris Size Quantity Transport Quantity at Generated Fraction Sump Small Pieces 4,447.5 ft2 100% 4,447.5 ft2

(<4 inch)

RMI Large Pieces 1,816.5 ft2 100% 1,816.5 ft2

(>4 inch)

Total 6,264 ft2 100% 6,264 ft2 Fines 44 ft3 100% 44 ft3 Small Pieces 0 ft3 100% 0 ft3

(<6 inch)

Temp-Mat' Large Pieces 0 ft3 100% 0 ft3

(>6 inch)

Intact Pieces 0 ft3 100% 0 ft3

(>6 inch)

Total 44 ft3 100% 44 ft3 Cal-Sil Total (Fines) 0 lbm 100% 0 lbm Min-K' Total (Fines) 0 lbm 100% 0 lbm Coatings Inside ZOI Total (Fines) 6.5 lbm 100% 6.5 lbm Exposed Unqualified Total (Fines) 66.2 lbm 100% 66.2 lbm Coatings Outside ZOI Dirt/Dust Total (Fines) 134.7 lbm 100% 134.7 lbm Latent Fiber Total (Fines) 9.9 ft3 100% 9.9 ft3 Miscellaneous Debris Total 341 ft2 100% 341 ft2

Beaver Valley Power Station, Unit Nos. 1 and 2 Final Supplemental Response to Generic Letter 2004-02 Page 26 of 59 BVPS-2 Table 3.e-10 Overall Debris Transport (Bounding RCS Loop Break) - BVPS-2 Debris Debris Debris Debris Type Debris Size Quantity Transport Quantity at Generated Fraction Sump Small Pieces 22,817.3 ft2 94.6% 21,585.2 ft2

(<4 inch)

RMI Large Pieces 9,319.7 ft2 98.4% 9,170.6 ft2

(>4 inch)

Total 32,137.0 ft2 95.7% 30,755.8 ft2 Fines 1.5 ft3 100% 1.5 ft3 Small Pieces 5.9 ft3 81.7% 4.8 ft3

(<6 inch)

Temp-Mat' Large Pieces 0 ft3 N/A 0 ft3

(>6 inch)

Intact Pieces 0 ft3 N/A 0 ft3

(>6 inch)

Total 7.4 ft3 85.5% 6.3 ft3 Fines 0.3 ft3 100% 0.3 ft3 Small Pieces 1.2 ft3 81.7% 1.0 ft3

(<6 inch)

Large Pieces Thermal WrapTM 0 ft3 N/A 0 ft3

(>6 inch)

Intact Pieces 0 ft3 N/A 0 ft3

(>6 inch)

Total 1.5 ft3 86.7% 1.3 ft3 Damming Material Total (Fines) 0.1 ft3 100% 0.1 ft3 Cal-Sil Total (Fines) 57.0 lbm 100% 57.0 lbm Min-KTM Total (Fines) 2.4 lbm 100% 2.4 lbm Coatings Inside ZOI Total (Fines) 321.7 lbm 100% 321.7 lbm Exposed Unqualified Total (Fines) 177.8 lbm 100% 177.8 lbm Coatings Outside ZOI Dirt/Dust Total (Fines) 156.4 lbm 100% 156.4 lbm Latent Fiber Total (Fines) 11.5 ft3 100% 11.5 ft3 Miscellaneous Debris Total 59.0 ft2 100% 59.0 ft2

Beaver Valley Power Station, Unit Nos. 1 and 2 Final Supplemental Response to Generic Letter 2004-02 Page 27 of 59 Table 3.e-11 Overall Debris Transport (Pressurizer Surge Line Break) - BVPS-2 Debris Debris Debris Debris Type Debris Size Quantity Transport Quantity at Generated Fraction Sump Small Pieces 3,146.0 ft2 100% 3,146.0 ft2

(<4 inch)

RMI Large Pieces 1,285.0 ft2 100% 1,285.0 ft2

(>4 inch)

Total 4,431.0 ft2 100% 4,431.0 ft2 Fines 0 ft3 100% 0 ft3 Small Pieces 0 ft3 100% 0 ft3

(<6 inch)

Temp-Mat' Large Pieces 0 ft3 100% 0 ft3

(>6 inch)

Intact Pieces 0 ft3 100% 0 ft3

(>6 inch)

Total 0 ft3 100% 0 ft3 Cal-Sil Total (Fines) 3.0 lbm 100% 3.0 lbm Min-K' Total (Fines) 10.4 lbm 100% 10.4 lbm Coatings Inside ZOI Total (Fines) 11.0 lbm 100% 11.0 lbm Exposed Unqualified Total (Fines) 177.8 lbm 100% 177.8 lbm Coatings Outside ZOI Dirt/Dust Total (Fines) 156.4 lbm 100% 156.4 lbm Latent Fiber Total (Fines) 11.5 ft3 100% 11.5 ft3 Miscellaneous Debris Total 59.0 ft2 100% 59.0 ft2

Beaver Valley Power Station, Unit Nos. 1 and 2 Final Supplemental Response to Generic Letter 2004-02 Page 28 of 59 Table 3.e-12 Overall Debris Transport (Reactor Vessel Nozzle Break) - BVPS-2 Debris Debris Debris Debris Type Debris Size Quantity Transport Quantity at Generated Fraction Sump Small Pieces 7,885.3 ft2 100% 7,885.3 ft2

(<4 inch)

RMI Large Pieces 3,220.7 ft2 100% 3,220.7 ft2

(>4 inch)

Total 11,106.0 ft2 100% 11,106.0 ft2 Microtherm Total (Fines) 303.0 lbm 100% 303.0 lbm Coatings Inside ZOI Total (Fines) 84.4 lbm 100% 84.4 lbm Exposed Unqualified Total (Fines) 177.8 lbm 100% 177.8 lbm Coatings Outside ZOI Dirt/Dust Total (Fines) 156.4 lbm 100% 156.4 lbm Latent Fiber Total (Fines) 11.5 ft3 100% 11.5 ft3 Miscellaneous Debris Total 59.0 ft2 100% 59.0 ft2

Beaver Valley Power Station, Unit Nos. 1 and 2 Final Supplemental Response to Generic Letter 2004-02 Page 29 of 59 Table 3.e-13 Overall Debris Transport (6-inch SIS Line Break) - BVPS-2 Debris Debris Debris Debris Type Debris Size Quantity Transport Quantity at Generated Fraction Sump Small Pieces 3,545.2 ft2 100% 3,545.2 ft2

(<4 inch)

RMI Large Pieces 1,448.0 ft2 100% 1,448.0 ft2

(>4 inch)

Total 4,993.2 ft2 100% 4,993.2 ft2 Fines 1.0 ft3 100% 1.0 ft3 Small Pieces 0 ft3 100% 0 ft3

(<6 inch)

Temp-Mat' Large Pieces 0 ft3 100% 0 ft3

(>6 inch)

Intact Pieces 0 ft3 100% 0 ft3

(>6 inch)

Total 1.0 ft3 100% 1.0 ft3 Damming Material Total (Fines) 0.1 ft3 100% 0.1 ft3 Thermal Wrap' Total (Fines) 0.1 ft3 100% 0.1 ft3 Coatings Inside ZOI Total (Fines) 18.8 lbm 100 18.8 lbm Exposed Unqualified Total (Fines) 177.8 lbm 100 177.8 lbm Coatings Outside ZOI Dirt/Dust Total (Fines) 156.4 lbm 100% 156.4 lbm Latent Fiber Total (Fines) 11.5 ft3 100% 11.5 ft3 Miscellaneous Debris Total 59.0 ft2 100% 59.0 ft2 BVPS-2 Power Operated Relief Valve Inlet Piping Break As described in Section 3.c, the power operated relief valve inlet piping break analysis applies a size distribution of 60% fines and 40% large pieces to the unbanded Thermal WrapTM insulation in the upper pressurizer cubicle. Thermal WrapTM fines are assumed to transport 100% to the strainer. No large debris is assumed to transport to the sump strainers. The upper pressurizer compartment is fully enclosed with the exception of two doors and a few equipment and piping penetrations in the floor. Since large piece debris would likely be blown against walls or other structures and would not transport as readily with the blowdown flow as it moves around corners, the majority of the large piece debris would remain in the pressurizer compartment at the end of the blowdown phase. The debris transport quantities for the BVPS-2 power operated relief valve inlet piping break are summarized in Table 3.e-14 below.

Beaver Valley Power Station, Unit Nos. 1 and 2 Final Supplemental Response to Generic Letter 2004-02 Page 30 of 59 Table 3.e-14, Overall Debris Transport (6-Inch Power Operated Relief Valve Inlet Piping Break) - BVPS-2 Debris Debris Debris Debris Type Debris Size Quantity Transport Quantity at Generated Fraction Sump Small Pieces 2,605.7 ft2 100% 2,605.7 ft2

(<4 inch)

RMI Large Pieces 1,064.3 ft2 0% 0 ft2

(>4 inch)

Total 3,670.0 ft2 2,605.7 ft2 Fines 9.9 ft3 100% 9.9 ft3 Thermal Wrap' Large Pieces 6.1 ft3 0% 0 ft3 Total 16.0 ft3 9.9 ft3 Total Coatings Inside ZOI 19.4 lbm 100% 19.4 lbm (Fines)

Exposed Unqualified Total 177.8 lbm 100% 177.8 lbm Coatings Outside ZOI (Fines)

Dirt/Dust Total (Fines) 156.4 lbm 100% 156.4 lbm Latent Fiber Total (Fines) 11.5 ft3 100% 11.5 ft3 Miscellaneous Debris Total 59.0 ft2 100% 59.0 ft2 Thick Elastic Materials As indicated above in Section 3.c, thick elastic materials in the vicinity of a break are assumed to fail as -inch by -inch to 1/2-inch by 1/2-inch pieces, based on the methodology outlined in the Diablo Canyon Power Plant 2008 supplemental response to GL 2004-02 (Reference 20). For debris to transport to the containment sump screens and impact strainer head loss, it must remain suspended in the recirculation pool as the BVPS-2 strainer screens are raised above the floor.

The Diablo Canyon Power Plant supplemental response (Table 1 of Section 3e) describes the debris characteristics of silicone rubber sleeving as follows:

Terminal Settling Velocity 0.256 ft/sec Calculated Minimum Turbulent Kinetic Energy (TKE) 0.098 ft²/sec² Required to Suspend Flow Velocity Associated with Incipient Tumbling 0.372 ft/sec The table below compares these characteristics to those of RMI, latent particulate, and individual fibers.

Beaver Valley Power Station, Unit Nos. 1 and 2 Final Supplemental Response to Generic Letter 2004-02 Page 31 of 59 Table 3.e-15: Transport Metrics for Various Debris Types - BVPS-2 Material Settling Velocity TKE Reqired to Flow Required for (ft/sec) Suspend (ft2/sec2) Tumbling (ft/sec)

Silicone Rubber 0.256 0.098 0.372 Small RMI 0.37 0.21 0.28 Latent Particulate 0.0016 4.00E-06 N/A Individual Fibers 0.0074 8.20E-05 N/A Individual fibers and latent particulate remain suspended and fully transport to the sump strainers. The impact of RMI on strainer head loss is negligible because the maximum approach velocity at the strainer is an order of magnitude lower than the settling velocity of RMI. Although the settling velocity and TKE required to suspend silicone rubber is less than that for small pieces of RMI, these settling velocities are comparable in magnitude and far greater than that of fine debris assumed to fully transport to the sump. The settling velocity of silicone rubber is over 23 times greater than the strainer approach velocity of 0.0108 ft/sec; therefore, the amount of thick elastic debris that will remain suspended in the recirculation pool and reach the sump strainers will be minimal.

3.f Head Loss and Vortexing The following updates to Sections 3.f.1 and 3.f.14 reflect plant modifications and revisions to the containment analyses implemented since 2010.

3.f.1 Provide a schematic diagram of the emergency core cooling system (ECCS) and containment spray systems (CSS).

As previously discussed, the sodium hydroxide (NaOH) chemical addition systems have been retired at both BVPS-1 and BVPS-2. Schematic diagrams for the BVPS-1 and BVPS-2 quench spray systems are updated to depict the current plant configurations.

Beaver Valley Power Station, Unit Nos. 1 and 2 Final Supplemental Response to Generic Letter 2004-02 Page 34 of 59 3.f.14 State whether containment accident pressure was credited in evaluating whether flashing would occur across the strainer surface, and if so, summarize the methodology used to determine the available containment pressure This section of the June 30, 2009 supplemental response states that a potential exists for minor flashing to occur across the strainer at BVPS-1 for a brief period of time shortly after the RS pumps start. The containment analysis was revised to apply the head loss from chemical precipitates when the sump water temperature reaches 150°F, as described in Appendix 2 of the September 28, 2010 RAI response (Reference 11).

The results of the flashing evaluation confirm that no flashing occurs at BVPS-1 when accounting for delayed chemical effects.

The BVPS-1 and BVPS-2 containment sump void fraction analyses were also updated to account for the revised containment analyses. The maximum void fraction calculated for a BVPS-1 break is 0.23%. The maximum void fraction for a BVPS-2 break is 0.26%.

Although both values have increased from those provided in the June 30, 2009 supplemental response, void fractions remain less than the 0.3% assumed in the NPSH computations.

3.i Debris Source Term Refinements 3.i.6 Recent or planned insulation change-outs in the containment that will reduce the debris burden at the sump strainers The following insulation modifications have been implemented since the submittal of the September 28, 2010 RAI response letter (Reference 11).

BVPS-1

Temp-MatTM fibrous insulation installed on the BVPS-1 reactor vessel inlet and outlet nozzles was replaced with RMI during the fall 2010 refueling outage (1R20). Fibrous debris loads for a reactor vessel nozzle break are now within the tested parameters.

Additional portions of calcium silicate insulation were removed from the RCS loop and pressurizer cubicles. With the exception of a short run of 3/4-inch piping, the insulation was not replaced. The scope of the modification was limited to sample lines; the insulation is not required to successfully deliver the samples to the appropriate locations.

No insulation modifications are planned for the Temp-MatTM insulation boxes installed on the pressurizer safety valve inlet piping. A break of the 6-inch piping has been shown to be adequate with respect to both strainer head loss and in-vessel downstream effects; therefore, no modifications are necessary. Regulatory Commitment No. 2 noted in the GSI-191 Resolution Plan letter, dated May 16, 2013 (Reference 4), will be closed without making insulation changes.

Beaver Valley Power Station, Unit Nos. 1 and 2 Final Supplemental Response to Generic Letter 2004-02 Page 35 of 59 BVPS-2

NUKON fibrous insulation blankets installed on the pressurizer power operated relief valve inlet piping were replaced by jacketed Thermal WrapTM insulation during the spring 2011 refueling outage (2R15). The Thermal WrapTM insulation was secured with Sure-Hold banding to the extent practical to reduce the zone of influence from 17.0D to 2.4D.

Min-KTM insulation on the BVPS-2 steam generator level instrumentation tubing was replaced with jacketed Thermal WrapTM insulation during the spring 2011 refueling outage. With respect to an RCS loop piping break at a steam generator nozzle, the steam generator level instrumentation tubing is inside the 28.6D zone of influence of Min-KTM insulation, but outside the 17.0D zone of influence of Thermal WrapTM.

Additional portions of calcium silicate insulation were removed from the RCS loop and pressurizer cubicles. The insulation was not replaced. The scope of the modification was limited to sample lines; the insulation is not required to successfully deliver the samples to the appropriate locations.

3.i.8 Modifications to equipment or systems conducted to reduce the debris burden at the sump strainers The sodium hydroxide (NaOH) chemical addition systems at both BVPS-1 and BVPS-2 have been replaced with baskets of powdered sodium tetraborate (NaTB) mounted to the lower containment floor. Use of sodium tetraborate reduces the maximum pH of the recirculation pool and containment sprays when compared to sodium hydroxide, thereby reducing the corrosion rate of susceptible materials, such as metallic aluminum. The reductions in both pH and chemical precipitate loads are demonstrated in the tables below:

Table 3.i.8-1 BVPS-1 Bounding Break (RCS Loop Break)

BVPS-1 Plant Parameter NaOH NaTB Maximum pH 10.1 8.28 Ultimate pH 8.79 7.79 Sodium Aluminum Silicate 110.25 lb 103.05 lb Aluminum Oxyhydroxide 136.84 lb 34.24 lb

Beaver Valley Power Station, Unit Nos. 1 and 2 Final Supplemental Response to Generic Letter 2004-02 Page 36 of 59 Table 3.i.8-2 BVPS-2 Bounding Break Excluding Reactor Vessel Nozzle Break (RCS Loop Break)

BVPS-2 Plant Parameter NaOH NaTB Maximum pH 9.93 8.27 Ultimate pH 9.03 7.80 Sodium Aluminum Silicate 118.51 lb 106.64 lb Aluminum Oxyhydroxide 42.04 lb 0.00 lb Table 3.i.8-3: BVPS-2 Reactor Vessel Nozzle Break BVPS-2 Plant Parameter NaOH NaTB Maximum pH 9.93 8.27 Ultimate pH 9.03 7.80 Sodium Aluminum Silicate 191.47 lb 106.42 lb Aluminum Oxyhydroxide 25.32 lb 0.00 lb BVPS-2 has six manway hatches between the refueling cavity and reactor cavity, three of which are normally sealed shut during operation with the remaining three covered with coarse grating. Procedures were modified to install grating over all six hatches after refueling (beginning in fall 2009 refueling outage - 2R14) to provide additional pressure relief pathways for a reactor vessel nozzle break. This modification was necessary to credit the reduced zone of influence of the limited offset reactor vessel nozzle break.

3.n Downstream Effects - Fuel and Vessel This section replaces Section 3.n of the June 30, 2009 supplemental response, in its entirety.

NRC Issue:

The objective of the downstream effects, fuel and vessel section is to evaluate the effects that debris carried downstream of the containment sump screen and into the reactor vessel has on core cooling.

Show that the in-vessel effects evaluation is consistent with, or bounded by, the industry generic guidance (WCAP-16793), as modified by NRC staff comments on that document. Briefly summarize the application of the methods. Indicate where the WCAP methods were not used or exceptions were taken and summarize the evaluation of those areas.

Beaver Valley Power Station, Unit Nos. 1 and 2 Final Supplemental Response to Generic Letter 2004-02 Page 37 of 59 Energy Harbor Nuclear Corp. Response:

Topical Report WCAP-17788-P, Revision 1 (References 5, 6, 7) provides evaluation methods to address in-vessel downstream effects. As discussed in NRC Technical Evaluation Report of In-Vessel Debris Effects, (Reference 8), NRC staff has performed a detailed review of WCAP-17788-P. Although the NRC staff did not approve WCAP-17788-P for use, as discussed further in U.S. Nuclear Regulatory Commission Staff Review Guidance for In-Vessel Downstream Effects Supporting Review of GL 2004-02 Responses (Reference 9), the staff expects that many of the methods developed in the TR may be used by PWR licensees in demonstrating adequate LTCC. Energy Harbor Nuclear Corp. has elected to use methods and analytical results developed in WCAP-17788-P, Revision 1 to address in-vessel downstream debris effects for BVPS-1 and BVPS-2 and has evaluated the applicability of the methods and analytical results for both units.

3.n.1 Sump Strainer Fiber Penetration Strainer bypass testing was completed in May 2008 for BVPS-1 and December 2008 for BVPS-2. At the time, no NRC approved protocol for strainer bypass testing existed.

The following aspects of sump strainer fiber bypass testing are discussed in detail for both BVPS-1 and BVPS-2:

Test Flume Design

Test Flow Rate

Debris Preparation

Debris Introduction

Transport Efficiency

Capture/Quantification of Bypassed Debris

Characterization of Captured Downstream Debris

Bypass Test Results 3.n.1.1 BVPS-1 Sump Strainer Fiber Penetration Test Flume Design The prototype strainer module cassettes and cartridges used in the fiber penetration test are identical to those used in the BVPS-1 containment sump strainer modules. The screen opening size, 1/16 inch, represents the full-scale size. The module used for testing contains 4 cartridges of 16 pockets each, for a total of 64 pockets with an effective area of 76.60 ft2. The tested debris quantity was scaled down by a ratio of the prototype strainer area to the effective full-size strainer area.

An isometric view of the strainer module used during testing is shown in Figure 3.n.1.1-1:

Beaver Valley Power Station, Unit Nos. 1 and 2 Final Supplemental Response to Generic Letter 2004-02 Page 39 of 59 Figure 3.n.1.1-2 also displays the sparger system used to keep the debris-laden water turbulent enough to carry the debris to the strainers without settling on the tank floor.

The sparger splits the return flow into two pathways that are injected at the bottom side of the tank and discharges the water towards the opposite side of the tank across the width of the tank floor. This method of stirring eliminates the need to constantly hand stir the tank to keep debris suspended.

The water level is maintained 1-inch above the strainer module to ensure it remains fully submerged at all times. Head loss, water temperature, flow rate, turbidity, and filter differential pressure were recorded during testing.

Test Flow Rate The flow rate through the prototype strainer was also reduced by the ratio of prototype to full strainer effective area, such that the approach velocity during the test is equal to the approach velocity at the sump strainers during maximum flow conditions. The planned maximum approach velocity of 0.0105 ft/sec is equivalent to 14,500 gallons per minute (gpm) of flow through the full-size sump strainers and corresponded to a scaled flow rate of 360 gpm through the prototype strainer.

Due to high differential pressure across the filter bags, the planned maximum flow rate was not achieved during strainer bypass testing. The maximum tested flow rate was 237 gpm, corresponding to an unscaled flow rate of (14,500 gpm)*(237/360) = 9,546 gpm. This is less than the maximum sump flow rate of 12,302 gpm at the time of cold leg switchover for the spectrum of large break LOCAs. The effects of this limitation may be considered minimal for the following three reasons.

1. A higher flow rate forces large fibers onto the strainer more effectively in the early stages of the event. This increases the filtering efficiency of the debris bed, reducing the penetration of the fine fibrous debris. A lower flow rate transports a higher percentage of fine fiber to the strainer early in the test, thereby increasing the bypass fraction.

2. The flow rate was increased by thirty percent to account for the higher approach velocity effect on the bed. Although this velocity did not reach that corresponding to maximum flow, it is high enough to demonstrate and account for the possible effect of pulling debris through the bed.

3. During strainer head loss testing, downstream samples of water were taken before each subtest of each test and no fibers were observed visually in the water samples, which were filtered and observed under a magnifying glass. The series of head loss tests were run at a variety of flow rates including a representative maximum flow rate.

Beaver Valley Power Station, Unit Nos. 1 and 2 Final Supplemental Response to Generic Letter 2004-02 Page 41 of 59 Debris Introduction All debris was added to the front right side of the test tank at the top. A total of three separate batches were added to the test tank, forming a fiber bed with a theoretical thickness of 0.06 inches. Each subsequent batch of fiber was added once it was determined visually that the previous fiber load has been loaded onto the strainers to the extent possible, with a maximum waiting period of five pool turnovers between batches.

Transport Efficiency No observations are noted in the test logs of fiber settling during the strainer bypass testing. With the aid of a mixing motor and hand stirring, the debris accumulated on the strainer screen at the maximum prototypical flow rate. The sparger system installed on the return line is designed to create turbulence near the tank floor to prevent settling of debris, eliminating the need to constantly hand stir the tank to keep debris suspended.

During strainer head loss testing with the maximum fibrous debris load, it was explicitly noted that all fibrous debris reached the strainers. The test tank setup for strainer head loss testing uses a sparger system installed on the return line and mechanical mixers in the tank, similar to those used in strainer bypass testing.

Capture/Quantification of Bypassed Debris The filter bags used were polyester microfiber filter bags with absolute 5-micron (m) rating. According to the manufacturer, the filter bags will retain at least 90% of particles of the specified micron size. The use of these filter bags is considered appropriate because fiber diameter of Temp-MatTM is 9 microns.

Before testing, the filter bags used to sieve the debris from the recirculating water were dried in a drying closet and weighed to obtain a baseline, clean filter bag mass. After testing, the filter bags were again dried and weighed in the same fashion. The mass difference between these two masses of each filter bag was used to determine the amount of fiber that bypassed the strainer during testing.

Characterization of Captured Downstream Debris Microscopic examinations of the captured debris were performed to determine the length distribution of the fibers that passed through the prototype strainer module during bypass testing. Length measurements were performed using both Optical Microscopy and Scanning Electron Microscope (SEM) imaging. The SEM imagery uses a larger field of view and is therefore able to more accurately measure longer fibers compared to the optical method. The fiber size distribution obtained from the SEM imagery is as follows:

Fibers greater than 250 m = 74%

Fibers greater than 500 m = 47%

Beaver Valley Power Station, Unit Nos. 1 and 2 Final Supplemental Response to Generic Letter 2004-02 Page 42 of 59 Fibers greater than 750 m = 28%

Fibers greater than 1000 m = 17%

The evaluation results indicate a size distribution that has larger fibers than that used during the Pressurized Water Reactors Owners Group (PWROG) fuel assembly testing.

Westinghouse evaluated the BVPS-1 fibrous debris bypass size distribution with respect to PWROG fuel assembly testing program. It was found that the BVPS-1 SEM distribution follows the same trend as the PWROG distribution but does not fall within the expected range. However, the distribution is close enough or equivalent to the bypass size that was used for the PWROG fuel assembly debris capture testing, as it is very similar to three of the plant specific fiber size distributions used to develop the PWROG distribution.

The larger fiber size distribution is likely due to testing with Temp-MatTM insulation rather than the NUKON insulation used in Westinghouse fuel assembly head loss testing. A vendor analysis of bypass samples obtained from a series of fiber bypass length characterization tests performed for the PWROG, determined typical size distributions for various types of fibrous debris including Temp-MatTM and NUKON. The analysis showed that Temp-MatTM produces larger fibers than NUKON.

The larger fiber distribution at BVPS-1 is representative of the bounding break conditions as the insulation boxes installed on the pressurizer safety valve inlet piping are composed of Temp-MatTM insulation.

Bypass Test Results The results of strainer bypass testing show that of the 725 grams of fiber added to the test tank, 58 grams of fiber was captured in the downstream filter bags. The bypass fraction is therefore (58 grams / 725 grams) = 0.080, or 8.0%. The bypass test used 100% fiber fines; therefore, the bypass fraction of 0.080 may be used in the in-vessel downstream effects analysis.

3.n.1.2 BVPS-2 Sump Strainer Fiber Penetration Test Flume Design The prototype top-hat strainer module used in the fiber penetration test is similar to those used in the BVPS-2 containment sump strainer assembly. The screen opening size, 3/32 inch, represents the full-scale size. The test used two full size Enercon Top-Hat strainer modules, each with a screen area of 26.25 ft2, for a total screen area of 52.5 ft2. The tested debris quantity was scaled down by a ratio of the prototype strainer area to the effective full-size strainer area.

A diagram of the test tank is illustrated below:

Beaver Valley Power Station, Unit Nos. 1 and 2 Final Supplemental Response to Generic Letter 2004-02 Page 45 of 59 Debris Preparation Though there was no specific guidance for debris preparation for strainer bypass testing, debris preparation of fiber fines was performed in accordance with the existing standards at the time for containment sump strainer head loss testing. The debris preparation procedure used for BVPS-2 strainer bypass testing is identical to that used during a strainer head loss test witnessed and accepted by NRC staff in January 2009 at the Alion Hydraulics Laboratory (Reference 22).

The test was conducted using Temp-MatTM as a surrogate for high-density fiberglass and NUKON as a surrogate for low density fiberglass and latent fiber. The test was conducted using both fibrous fines and smalls debris. The smalls were prepared by first cutting fiber blankets into smaller pieces and then feeding the pieces through a commercially available leaf shredder. Then to obtain fiber fines, the fiber smalls were again fed through the leaf shredder. Size requirements were referenced to the NUREG/CR-6808 size classification of fibrous debris, shown in Figure 3.n.1.1-3, provided in the previous section.

Fiber fines were inspected to ensure the size requirements of classes 1 through 3 were met, while smalls were inspected to ensure the size requirements of classes 1 through 4 were met. Both fiber sizes were then boiled in water for at least 10 minutes to remove the binder that exists in the fibrous debris. For the "fines," the fiber was then placed in buckets of approximately 4 gallons of water in quantities of approximately 0.25 lbs. and then beaten for 4 minutes with a paint mixer attached to an electric drill.

For the "smalls," the fiber was mixed thoroughly with a paint mixer attached to an electric drill until a homogeneous slurry was formed.

Debris Introduction All debris was added to the front and rear right corners of the test tank. Additional batches of fiber were added once it was determined visually that the previous fiber load had been loaded onto the strainers to the extent possible, with a maximum waiting period of five pool turnovers.

The first debris batch consisted of a fiber amount equal to a 0.05-inch theoretical bed thickness. Another batch consisted of another 0.01-inch of theoretical fiber bed (equating to a 0.06-inch total bed thickness). The final batch was the remaining amount of scaled fibrous debris created by the break (0.01 inch more resulting in 0.07-inch total bed thickness).

Transport Efficiency No observations are noted in the test logs of fiber settling during the strainer bypass testing. With the aid of a mixing motor and hand stirring, the debris accumulated on the strainer screen at the maximum prototypical flow rate. The sparger system installed on the return line is designed to create turbulence near the tank floor to prevent settling of debris.

Beaver Valley Power Station, Unit Nos. 1 and 2 Final Supplemental Response to Generic Letter 2004-02 Page 46 of 59 During strainer head loss testing with the maximum fibrous debris load, it was explicitly noted that all fibrous debris reached the strainers without the use of hand-stirring. The test tank setup for strainer head loss testing uses a sparger system installed on the return line and mechanical mixers in the tank, similar to those used in strainer bypass testing.

Capture/Quantification of Bypassed Debris The BVPS-2 bypass testing used polyester microfiber filter bags with absolute 5-micron rating. According to the manufacturer, the filter bags will retain at least 90% of the particles of the specified size. The use of these bags is appropriate since the diameter of Temp-MatTM fibers is 9 microns and the diameter of NUKON fibers is 7 microns.

Before testing, the filter bags used to sieve the debris from the recirculating water were dried in a drying closet and weighed to obtain a baseline, clean filter bag mass. After testing, the filter bags were again dried and weighed in the same fashion. The mass difference between these two masses of each filter bag was used to determine the amount of fiber that bypassed the strainer.

Characterization of Captured Downstream Debris Microscopic examinations of the captured debris were performed to determine the length distribution of the fibers that passed through the prototype strainer module during bypass testing. Length measurements were performed using Scanning Electron Microscope (SEM) imaging. The fiber size distribution obtained from the SEM imagery is as follows:

Fibers shorter than 500 m = 60%

Fibers between 500 m and 1000 m = 15%

Fibers longer than 1000 m = 25%

Westinghouse evaluated the BVPS-2 fibrous debris bypass size distribution with respect to the PWROG fuel assembly testing program. It was concluded that when considering the program being performed by the PWROG to increase in-vessel fiber quantities (Reference 5), the BVPS-2 fiber length distribution is both bounded by and closely approximated by the fiber length distributions used in testing. Thus, the data of the PWROG program to increase the fiber debris limit per fuel assembly is applicable to BVPS-2.

Bypass Test Results The results of strainer bypass testing show that of the 568 grams of fiber added to the test tank, 24 grams of fiber was captured in the filter bags. The bypass fraction is therefore (24 grams / 568 grams) = 0.042, or 4.2%. This includes all sizes of fibrous debris; both fines and smalls.

The in-vessel downstream effects analysis assumes all fibrous debris is in the form of fines. A bypass fraction for fiber fines may be calculated from the existing test data by

Beaver Valley Power Station, Unit Nos. 1 and 2 Final Supplemental Response to Generic Letter 2004-02 Page 47 of 59 subtracting the smalls debris from the initial fiber load while maintaining the same bypassed fiber quantity. This conservatively assumes that 100% of the smalls debris was captured by the strainer and all captured downstream debris originated from the fiber fines. The total amount of fiber fines added to the test tank was 0.911 lbm (413 g).

The resulting fiber fines bypass fraction is (24 grams / 413 grams) = 0.058, or 5.8%.

3.n.2 Applicability to WCAP-17788 Methods and Analysis Results BVPS-1 and BVPS-2 are Westinghouse 3-loop upflow barrel/baffle plants. Per Section 3.0 of the September 4, 2019 NRC Staff Review Guidance (Reference 9), it is necessary to confirm that BVPS-1 and BVPS-2 are within the key parameters of the WCAP-17788-P, Revision 1 methods and analysis. Each of the key parameters is discussed below.

The postulated break that generates the most fibrous debris at BVPS-1 is a break in the 6-inch pressurizer safety valve inlet piping. Although this type of break is not specifically addressed in WCAP-17788, use of the hot leg break methodology is appropriate. This is a conservative approach since, due to the elevation of the break at the top of the pressurizer (located on a hot leg), a portion of the ECCS flow and entrained fiber will bypass the core via backflow through the steam generators.

Application of the hot leg break methodology directs all ECCS flow and entrained fiber to the reactor vessel, thereby maximizing the quantity of accumulated reactor vessel fiber for this break scenario. A separate evaluation is performed for the hot leg break and pressurizer safety valve inlet piping breaks at BVPS-1.

A single in-vessel downstream effects evaluation was performed that bounds the hot leg break and small break cases that produce comparable quantities of fibrous debris at BVPS-2, such as a 6-inch pressurizer power operated relief valve inlet piping break.

3.n.3 Fuel Design BVPS-1 and BVPS-2 use the Westinghouse 17x17 Robust Fuel Assembly (RFA-2) fuel design.

3.n.4 WCAP-17788 Debris Limit The proprietary total in-vessel (core inlet and heated core) fibrous debris limit contained in Section 6.5 of WCAP-17788-P, Volume 1, Revision 1 (Reference 5) applies to both BVPS-1 and BVPS-2.

3.n.5 Methodology used to calculate the fibrous debris amounts The amount of fibrous debris calculated to arrive at the reactor vessel is determined for BVPS-1 and BVPS-2 following the method described in WCAP-17788-P, Volume 1, Revision 1, Section 6.5 (Reference 5). This method credits the strainer bypass testing results and debris bypassing the core via the recirculation spray system. The iterative methodology provided in Section 6.5.3 was implemented using an Excel spreadsheet

Beaver Valley Power Station, Unit Nos. 1 and 2 Final Supplemental Response to Generic Letter 2004-02 Page 48 of 59 and run until the time of hot leg switchover (beyond the termination criteria provided in Section 6.5.5). The quantity of fiber reaching the reactor vessel was then confirmed using WCAP-17788-P, Volume 1, Equation 6-27, which calculates the total quantity of fiber ingested into the reactor vessel as a function of time.

3.n.6 Confirm maximum combined amount of fiber that may arrive at the core inlet and heated core for hot leg break is below the WCAP-17788 fiber limit The BVPS-1 maximum amount of fiber calculated to potentially reach the reactor vessel is 4.8 grams (g) per fuel assembly (FA) for a large hot leg break and 13.4 g/FA for a 6-inch pressurizer safety valve inlet piping break.

The BVPS-2 maximum amount of fiber calculated to potentially reach the reactor vessel is 7.7 g/FA for a large hot leg break.

These calculated fiber amounts are less than the proprietary in-vessel fibrous debris limit provided in Section 6.5 of WCAP-17788-P, Volume 1, Revision 1 (Reference 5).

3.n.7 Confirmation that the core inlet fiber amount is less than the WCAP-17788-P, Revision 1 threshold The applicable WCAP-17788-P, Revision 1 core inlet fiber threshold is provided in Table 6-3 of WCAP-17788-P, Volume 1, Revision 1 (Reference 5).

The core inlet fiber amount for BVPS-1 is calculated to be 4.8 g/FA for a large hot leg break and 12.9 g/FA for a pressurizer safety valve inlet piping break.

The core inlet fiber amount for BVPS-2 is calculated to be 7.2 g/FA for a large hot leg break.

These calculated core inlet fiber amounts are less than the applicable WCAP-17788-P, Volume 1, Revision 1 (Reference 5) core inlet fiber threshold.

3.n.8 Confirmation that the earliest sump switchover (SSO) time is 20 minutes or greater The earliest possible SSO time for BVPS-1 is 1768 seconds (29.46 minutes).

The earliest possible SSO time for BVPS-2 is 2476 seconds (41.27 minutes).

3.n.9 Predicted chemical precipitation timing from WCAP-17788-P, Revision 1, Volume 5 testing and the specific test group considered to be representative of the plant BVPS-1 and Test Group 27 Chemical precipitation timing is dependent on the plant buffer, sump pool pH, volume and temperature, and debris types and quantities. Table 3.n.9-1, Key Parameter Values for BVPS-1 Chemical Testing, summarizes the key chemical precipitation

Beaver Valley Power Station, Unit Nos. 1 and 2 Final Supplemental Response to Generic Letter 2004-02 Page 49 of 59 parameters and values for BVPS-1 and compares them to Test Group 27 from WCAP-17788-P, Volume 5, Revision 1 (Reference 7).

Table 3.n.9-1, Key Parameter Values for BVPS-1 Chemical Testing Parameter BVPS-1 Value Test Group 27 Value Buffer NaTB NaTB pH 8.28 8.28 Minimum Sump Volume (ft3) 34,049 23,041 Maximum Sump Pool Temperature (°F) 232 250 Calcium-Silicate (lbm) 42 101.5 E-glass (lbm) 41.8 86.4 Silica (lbm) 0 16 Mineral Wool (ft3) 0 0 Aluminum Silicate (ft3) 0 0 Concrete (ft2) 611 811 Interam (ft3) 0 0 Aluminum (ft2) 4,384 10,373 Galvanized Steel (ft2) N/A N/A The BVPS-1 values provided in this table are specific to a hot leg break in the RCS loop cubicle, as the aluminum surface area provided is specific to an RCS loop piping break.

2,850 ft2 of the 4,384 ft2 of aluminum surface area is attributed to aluminum fins in the reactor coolant pump motor coolers; the coolers are recessed within the motor and are not exposed to containment spray. A break in the RCS loop cubicle is assumed to destroy the coolers of one motor such that the cooling fins become exposed to containment spray.

Although the pressurizer safety valve inlet piping break generates more fibrous (E-glass) debris than the tested quantities above (137.4 lbm), the aluminum from the reactor coolant pump motor coolers will remain intact; the total aluminum surface area for this break is approximately 15% of that which was tested and 35% of that calculated for a hot leg break. WCAP-17788-P, Volume 5, Revision 1, Table 7-6 (Reference 7),

shows that a decrease in aluminum relative to the test aluminum decreases the risk of precipitation and increases tchem, while the insulation material mass has no effect on tchem.

Beaver Valley Power Station, Unit Nos. 1 and 2 Final Supplemental Response to Generic Letter 2004-02 Page 50 of 59 Based on the discussion in Section 5.5.8 of WCAP-17788-P, Volume 5, Revision 1 (Reference 7) the long drain times observed for Test Group 27 samples taken at eight hours or earlier were attributed to particulate material rather than chemical products.

Therefore, it can be concluded that chemical precipitation occurs at 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for the conditions represented by Test Group 27. This is confirmed in PWROG-16073-P (Reference 23), Table 4.4-1.

Based on the comparison of the hot leg break plant parameters in Table 3.n.9-1, Test Group 27 is representative of BVPS-1; thus, the predicted chemical precipitation timing (tchem) is 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

BVPS-2 and Test Group 17 Test Group 17 was intended to be representative of BVPS-2. Table 3.n.9-2, Key Parameter Values for BVPS-2 Chemical Testing, summarizes the key chemical precipitation parameters for BVPS-2 and compares them to Test Group 17 from WCAP-17788-P, Volume 5, Revision 1 (Reference 7).

Table 3.n.9-2, Key Parameter Values for BVPS-2 Chemical Testing Parameter BVPS-2 Value Test Group 17 Value Buffer NaTB NaTB pH 8.27 8.27 Minimum Sump Volume (ft3) 46,229 50,460 Maximum Sump Pool Temerature (°F) 237 253 Calcium-Silicate (lbm) 60.0 92.8 E-glass (lbm) 66.0 66.8 Silica (ft3) 0.65 1 Mineral Wool (ft3) 0 0 Aluminum Silicate (ft3) 0 0 Concrete (ft2) 611 811 Interam (ft3) 0 0 Aluminum (ft2) 1,291 1,539 Galvanized Steel (ft2) N/A N/A The Test Group 17 parameters bound BVPS-2 values, except for minimum sump volume. According to WCAP-17788-P, Volume 5, Revision 1, Section 3, (Reference 7) the sump volume is used as an input to calculate the scaled debris quantities used in the autoclave test.

Beaver Valley Power Station, Unit Nos. 1 and 2 Final Supplemental Response to Generic Letter 2004-02 Page 51 of 59 Aluminum is the main element in the materials causing long drain times in the autoclave testing. If the scaled surface area of the aluminum coupon used in the autoclave test is greater than the scaled surface area calculated using the actual BVPS-2 aluminum surface area and sump volume, Test Group 17 bounds the BVPS-2 plant conditions.

The scaled aluminum surface area for Test Group 17 and the equivalent BVPS-2 scaled aluminum surface area are calculated below using the method described in Reference 7, Section 3.

1,539 2 1 3 17 = 50 = 0.0538 2 50,460 3 28.32 1,290 2 1 3 2 = 50 = 0.0497 2 46,229 3 28.32 The equivalent BVPS-2 scaled aluminum surface area is less than the surface area of the aluminum coupon used in the autoclave test; therefore, Test Group 17 remains bounding with respect to aluminum surface area.

3.n.10 Confirmation that chemical effects will not occur earlier than latest time to implement boric acid precipitation mitigation measures As described in BVPS-1 Updated Final Safety Analysis Report, Section 6.3.3.9, the simultaneous cold leg and hot leg recirculation mode is initiated at BVPS-1 approximately 6.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months after the loss of coolant accident.

As described in BVPS-2 Updated Final Safety Analysis Report, Table 6.3-7, the hot leg recirculation mode is initiated at BVPS-2 approximately 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after the loss of coolant accident.

Both values are less than the minimum time to chemical effects, which is 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months as stated in the previous subsection.

3.n.11 WCAP-17788 tblock value for the RCS design category BVPS-1 and BVPS-2 are 3-loop Westinghouse upflow barrel/baffle plants. Based on WCAP-17788, Volume 1, Revision 1, Table 6-1, (Reference 5) tblock for BVPS-1 and BVPS-2 is 143 minutes.

3.n.12 Confirmation that chemical effects do not occur prior to tblock As stated in the previous subsections, the time to chemical precipitation is 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for both BVPS-1 and BVPS-2, which is greater than the applicable tblock value of 143 minutes.

Beaver Valley Power Station, Unit Nos. 1 and 2 Final Supplemental Response to Generic Letter 2004-02 Page 52 of 59 3.n.13 Plant rated thermal power compared to the analyzed power level for the RCS design category Both BVPS-1 and BVPS-2 have a rated thermal power of 2900 mega watts thermal (MWt). BVPS-1 and BVPS-2 are Westinghouse 3-loop plants and the applicable analyzed thermal power is 3658 MWt as provided in WCAP-17788-P, Volume 4, Revision 1, Table 6-1 (Reference 6). The rated thermal power for BVPS-1 and BVPS-2 are less than the analyzed power; therefore, this parameter is considered bounded by the WCAP-17788-P, Revision 1 alternate flow path analysis.

3.n.14 Plant alternate flow path (AFP) resistance compared to the analyzed AFP resistance for the plant RCS design category BVPS-1 is a Westinghouse upflow barrel/baffle plant. The proprietary analyzed AFP resistance is provided in Table 6-1 of WCAP-17788-P, Volume 4, Revision 1 (Reference 6). The proprietary BVPS-1 specific AFP resistance is provided in Table RAI-4.2-24. The BVPS-1 specific AFP resistance is less than the analyzed value; therefore, the BVPS-1 AFP resistance is bounded by the resistance applied to the AFP analysis.

BVPS-2 is a Westinghouse upflow barrel/baffle plant. The proprietary analyzed AFP resistance is provided in Table 6-1 of WCAP-17788-P, Volume 4, Revision 1 (Reference 6). The proprietary BVPS-2 specific AFP resistance is provided in Table RAI-4.2-24. The BVPS-2 specific AFP resistance is greater than the resistance used in the AFP analysis.

RAI 4.2 of WCAP-17788-P, Volume 4, Revision 1 (Reference 6) provides an equivalent maximum AFP resistance for those plants whose rated thermal power is less than that used in the thermal-hydraulic model, as less flow is required to match core boil-off in a core with lower rated thermal power.

WCAP-17788-P, Volume 4, Revision 1, Table RAI-4.2-24 (Reference 6) shows that the BVPS-2 AFP resistance, when adjusted for rated thermal power, is less than the analyzed AFP resistance. Therefore, the BVPS-2 AFP resistance is bounded by the resistance applied to the AFP analysis.

3.n.15 Consistency between the minimum ECCS flow per FA assumed in the AFP analyses and that at the plant BVPS-1 is a Westinghouse upflow barrel/baffle plant. The AFP analysis for Westinghouse upflow plants analyzed a range of ECCS recirculation flow rates from 8 to 40 gpm/FA, as shown in Table 6-1 of WCAP-17788-P, Volume 4, Revision 1 (Reference 6). The BVPS-1 ECCS recirculation flow rates used in the bounding in-vessel debris analyses are 33.8 gpm/FA for a large hot leg break and 24.8 gpm/FA for a pressurizer safety valve Inlet Piping break. Both are within the range of ECCS recirculation flow rates considered in the AFP analysis.

Beaver Valley Power Station, Unit Nos. 1 and 2 Final Supplemental Response to Generic Letter 2004-02 Page 53 of 59 BVPS-2 is a Westinghouse upflow barrel/baffle plant. The BVPS-2 ECCS recirculation flow rate used in the bounding in-vessel debris analysis is 40 gpm/FA for a large hot leg break. This is equal to the maximum flow rate considered in the AFP analysis.

3.n.16 Summary A comparison of key parameters used in the WCAP-17788 AFP analysis to the BVPS-1 specific values is summarized in Table 3.n.16-1. Based on these comparisons, BVPS-1 is bounded by the key parameters and the WCAP-17788 methods and results are applicable.

Table 3.n.16-1, Key Parameter Values for BVPS-1 In-Vessel Debris Effects WCAP-17788 BVPS-1 Parameter Evaluation Value Value Maximum Total Volume 1 Maximum in-vessel fiber load is In-Vessel Fiber Load 13.4 Section 6.5 13% of WCAP-17788 limit.

(g/FA)

Maximum Core Inlet Volume 1 Maximum in-vessel fiber load is 12.9 Fiber Load (g/FA) Table 6-3 25% of WCAP-17788 limit.

Minimum Sump Later switchover time results in Switchover Time a lower decay heat at the time of (minutes) 20 29.5 debris arrival, reducing the potential for debris induced core uncovery and heatup.

Minimum Chemical Potential for complete core inlet Precipitate Time 2.38 24 blockage due to chemical (hours) (tblock) (tchem) product generation would occur much later than assumed.

Maximum Hot Leg 24 Latest hot leg switchover occurs Switchover Time 7 well before the earliest potential (hours) (tchem) chemical product generation.

Rated Thermal Lower thermal power results in 3658 2900 Power (MWt) lower decay heat.

Maximum Alternate AFP resistance is less than the Volume 4 Volume 4 Flow Path (AFP) analyzed value, which increases Table 6-1 Table RAI-4.2-24 Resistance the effectiveness of the AFP.

Minimum ECCS Maximum debris bed resistance Recirculation Flow 8 24.8 at the core inlet occurs at lower (gpm/FA) flow rates.

A comparison of key parameters used in the WCAP-17788 AFP analysis to the BVPS-2 specific values is summarized in Table 3.n.16-2. Based on these comparisons, BVPS-2 is bounded by the key parameters and the WCAP-17788 methods and results are applicable.

Beaver Valley Power Station, Unit Nos. 1 and 2 Final Supplemental Response to Generic Letter 2004-02 Page 54 of 59 Table 3.n.16-2, Key Parameter Values for BVPS-2 In-Vessel Debris Effects WCAP-17788 BVPS-2 Parameter Evaluation Value Value Maximum Total Volume 1 Maximum in-vessel fiber load is In-Vessel Fiber Load 7.7 Section 6.5 15% of WCAP-17788 limit.

(g/FA)

Maximum Core Inlet Volume 1 Maximum in-vessel fiber load is 7.2 Fiber Load (g/FA) Table 6-3 8% of WCAP-17788 limit.

Minimum Sump Later switchover time results in Switchover Time a lower decay heat at the time of (minutes) 20 41.3 debris arrival, reducing the potential for debris induced core uncovery and heatup.

Minimum Chemical Potential for complete core inlet Precipitate Time blockage due to chemical 2.38 24 (hours) product generation would occur much later than assumed.

Maximum Hot Leg Latest hot leg switchover occurs Switchover Time 24 6 well before earliest potential (hours) chemical product generation.

Rated Thermal Lower thermal power results in 3658 2900 Power (MWt) lower decay heat.

Maximum Alternate AFP resistance is less than the Flow Path (AFP) Volume 4 analyzed value when accounting Volume 4 Resistance Table RAI-4.2-24 for differences in rated thermal Table 6-1 (adjusted) power, which increases the effectiveness of the AFP.

Minimum ECCS Maximum debris bed resistance Recirculation Flow 8 40 at the core inlet occurs at lower (gpm/FA) flow rates.

3.o Chemical Effects NRC Issue The objective of the chemical effects section is to evaluate the effect that chemical precipitates have on head loss and core cooling.

1) Provide a summary of evaluation results that show that chemical precipitates formed in the post-LOCA containment environment, either by themselves or combined with debris, do not deposit at the sump screen to the extent that an unacceptable head loss results, or deposit downstream of the sump screen to the extent that long-term core cooling is unacceptably impeded.

Beaver Valley Power Station, Unit Nos. 1 and 2 Final Supplemental Response to Generic Letter 2004-02 Page 55 of 59 Energy Harbor Nuclear Corp. Response The BVPS-1 and BVPS-2 chemical effects analysis of the sump strainers was submitted in the June 30, 2009 supplemental response as well as subsequent RAI responses submitted on September 28, 2010 (Reference 11).

Since these submittals, the sodium hydroxide chemical addition systems at both BVPS-1 and BVPS-2 have been retired. Baskets of powdered sodium tetraborate mounted to the floor of lower containment were installed to perform the function of containment sump pH buffer. Use of sodium tetraborate reduces the pH of the containment sump and spray fluid, thereby reducing the corrosion rate of susceptible materials, such as metallic aluminum. As previously shown in Tables 3.i.8-1 through 3.i.8-3, the buffer switchovers resulted in a significant reduction in the chemical precipitate loading for both BVPS-1 and BVPS-2.

Strainer head loss testing was performed for BVPS-1 and BVPS-2 with maximum chemical debris loads plus a 10% margin. Therefore, all chemical debris loads listed in Section 3.o of the June 30, 2009 supplemental response, remains bounding with respect to the design basis strainer head loss testing.

Chemistry parameters for BVPS-1 containment sump with a sodium tetraborate buffer were not included in the June 30, 2009 supplemental response, as no buffer change was planned at that time. Those parameters provided for the BVPS-2 NaTB buffer in the June 2009 submittal are provided below in this section for the BVPS-1 NaTB buffer.

The response to RAI #5 in the June 30, 2009 supplemental response provides the range of recirculation pool pH values at both the beginning of life and end of life periods of the fuel cycle. The range of BVPS-2 pH values are provided, as the buffer switchover was planned at the time of response submittal. The buffer switchover for BVPS-1 was not planned at that time; the range of BVPS-1 pH values with the sodium tetraborate buffer are provided below:

Minimum sump pH (beginning of life, or BOL): 7.16 Maximum sump pH (end of life, or EOL): 7.59 The response to RAI 6 in the June 30, 2009 supplemental response compares the expected containment pool conditions to the Integrated Chemical Effects Test (ICET) conditions. Table RAI 6-1 provides the expected conditions at BVPS-1 and BVPS-2 with NaOH buffer and Table RAI 6-2 provides the expected conditions at BVPS-2 with NaTB buffer.

Table RAI 6-2 is revised to include expected containment pool conditions at BVPS-1 with NaTB buffer.

Beaver Valley Power Station, Unit Nos. 1 and 2 Final Supplemental Response to Generic Letter 2004-02 Page 56 of 59 Table RAI 6-2 CHEMICAL ICET #5 BVPS-1 BVPS-2 VALUE MID- MID-POINT POINT pH 8.0 - 8.5 7.1 7.5 7.3 7.2 7.4 7.3 (MIN) (MAX) (MIN) (MAX)

Corresponding 2800 2580 2150 2365 2588.2 2269 2428.6 BORIC ACID (as ppm Boron)

Corresponding 5574.5 1391 1417 1404 1205.4 1219.3 1212.4 SODIUM TETRABORATE (ppm)

The BVPS-1 and BVPS-2 in-vessel chemical effects analysis is described in Sections 3.n.9 through 3.n.12. These sections conclude that formation of chemical precipitates does not occur until a minimum of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months post-LOCA at both BVPS-1 and BVPS-2 as determined through testing described in WCAP-17788-P, Volume 5, Revision 1 (Reference 7).

3.p Licensing Basis NRC Issue:

The objective of the licensing basis section is to provide information regarding any changes to the plant licensing basis due to the sump evaluation or plant modifications.

1) Provide the information requested in GL 04-02 Requested Information Item 2(e) regarding changes to the plant licensing basis. The effective date for changes to the licensing basis should be specified. This date should correspond to that specified in the 10 CFR 50.59 evaluation for the change to the licensing basis.

Energy Harbor Nuclear Corp. Response:

Since the June 30, 2009 supplemental response was issued, the sodium hydroxide containment sump pH buffer was replaced with sodium tetraborate at BVPS-1. The NRC issued License Amendment No. 289 by letter dated March 14, 2012 (Reference 19) to change the pH buffer from sodium hydroxide to sodium tetraborate prior to achieving Mode 4 during startup from the BVPS-1 refueling outage in the spring of 2012.

Beaver Valley Power Station, Unit Nos. 1 and 2 Final Supplemental Response to Generic Letter 2004-02 Page 57 of 59 A July 10, 2020 license amendment request letter (Reference 24) was submitted to the NRC to propose the addition of a new technical specification for the containment sump, and make related changes to ECCS technical specifications in accordance with Technical Specifications Task Force, Improved Standard Technical Specifications Change Traveler No. 567 (TSTF-567), Revision 1, Add Containment Sump TS to Address GSI-191 Issues, (Reference 25). NRC approval of this amendment was requested by the end of January 2021.

Energy Harbor Nuclear Corp. plans to update the BVPS-1 and BVPS-2 licensing basis in accordance with the requirements of 10 CFR 50.71(e) to incorporate the GL 2004-02 response as appropriate, including the values of analyzed debris limits associated with the new containment sump technical specification, at the same time the containment sump technical specification changes are implemented and within 60 days of receiving the NRC closure letter for actions to address GL 2004-02 at BVPS. This will ensure that the analyzed debris limits referenced in the technical specifications are available in a licensing basis document (that is, the Updated Final Safety Analysis Report).

4.0 References

1. NRC Letter to Energy Harbor Nuclear Corp., Beaver Valley Power Station, Unit Nos.

1 and 2; Davis-Besse Nuclear Power Station, Unit No. 1; and Perry Nuclear Power Plant, Unit No. 1 - Issuance of Amendment Nos. 304, 194, 299, and 187, Respectively, Re: Order Approving Transfer of Licenses and Conforming License Amendments (EPID L-2020-LLM-0000), dated February 27, 2020 (ADAMS Accession No. ML20030A440).

2. NRC Generic Letter 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors, dated September 13, 2004. (ADAMS Accession No. ML042360586).

3. SECY-12-0093, Closure Options for Generic Safety Issue - 191, Assessment of Debris Accumulation on Pressurized-Water Reactor Sump Performance, dated July 9, 2012 (ADAMS Accession No. ML121310648).

4. FENOC Letter to NRC L-13-176, Generic Safety Issue 191 Resolution Plan (TAC Nos. MC4665 and MC4666), dated May 16, 2013 (ADAMS Accession No. ML13136A144).

5. Topical Report WCAP-17788-P, Volume 1, Revision 1, Comprehensive Analysis and Test Program for GSI-191 Closure (PA-SEE-1090), dated December 12, 2019.

6. Westinghouse Topical Report WCAP-17788-P, Volume 4, Revision 1, Comprehensive Analysis and Test Program for GSI-191 Closure (PA-SEE-1090) -

Thermal-Hydraulic Analysis of Large Hot Leg Break with Simulation of Core Inlet Blockage, dated February 28, 2020.

Beaver Valley Power Station, Unit Nos. 1 and 2 Final Supplemental Response to Generic Letter 2004-02 Page 58 of 59

7. Topical Report WCAP-17788-P, Volume 5, Revision 1, Comprehensive Analysis and Test Program for GSI-191 Closure (PA-SEE-1090) - Autoclave Chemical Effects Testing for GSI-191 Long-Term Cooling, dated December 13, 2019.

8. NRC Document, Technical Evaluation Report of In-Vessel Debris Effects, dated June 13, 2019 (ADAMS Accession No. ML19178A252).

9. NRC Document, U.S. Nuclear Regulatory Commission Staff Review Guidance for In-Vessel Downstream Effects Supporting Review of Generic Letter 2004-02 Responses, dated September 4, 2019 (ADAMS Accession No. ML19228A011).

10. FENOC Letter to NRC L-09-152, Supplemental Response to Generic Letter 2004-02 (TAC Nos. MC4665 and MC4666), dated June 30, 2009 (ADAMS Accession No. ML091830390).

11. FENOC Letter to NRC L-10-115, Response to Request for Additional Information Related to Generic Letter 2004-02 (TAC Nos. MC4665 and MC4666), dated September 28, 2010 (ADAMS Accession No. ML102770023).

12. Nuclear Energy Institute Guidance Report NEI 04-07, Revision 0, Pressurized Water Reactor Sump Performance Evaluation Methodology, dated December 2004 (ADAMS Accession No. ML050550138).

13. Safety Evaluation by the Office of the Nuclear Reactor Regulation Related to NRC Generic Letter 2004-02, Nuclear Energy Institute Guidance Report NEI 04-07, Pressurized Water Reactor Sump Performance Evaluation Methodology, dated December 6, 2004 (ADAMS Accession No. ML043280007).

14. NRC Letter to FENOC, Beaver Valley Power Station, Unit No. 1 - Issuance of Amendment No. 280 RE: Changes to the Recirculation Spray System Pump Start Signal Due to the Containment Sump Screen Modification (TAC No. MD4290), dated October 5, 2007 (ADAMS Accession No. ML072680397).

15. NRC Letter to FENOC, Beaver Valley Power Station Unit No. 2 - Issuance of Amendment No. 164 RE: Changes to the Recirculation Spray System Pump Start Signal Due to the Containment Sump Screen Modification (TAC No. MD4291), dated March 11, 2008 (ADAMS Accession No. ML080730018).

16. Westinghouse Report WCAP-16793-NP, Revision 2, Evaluation of Long-Term Cooling Considering Particulate, Fibrous and Chemical Debris in the Recirculating Fluid, dated October 2011 (ADAMS Accession No. ML11292A021).

17. NRC Letter to FENOC, Beaver Valley Power Station, Unit No. 2 - Issuance of Amendment RE: The Use of Containment Accident Pressure in Determining Net Positive Suction Head of Recirculation Spray Pumps (TAC No. ME0098), dated March 26, 2009 (ADAMS Accession No. ML090270068).

Beaver Valley Power Station, Unit Nos. 1 and 2 Final Supplemental Response to Generic Letter 2004-02 Page 59 of 59

18. NRC Letter to FENOC, Beaver Valley Power Station, Unit Nos. 1 and 2 - Issuance of Amendments RE: Spray Additive System by Containment pH Control (TAC Nos.

MD9734 and MD9735), dated April 16, 2009, (ADAMS Accession No. ML090780352).

19. NRC Letter to FENOC, Beaver Valley Power Station, Unit Nos. 1 and 2 - Issuance of Amendments Regarding the Spray Additive System by Containment Sump pH Control System (TAC Nos. ME6352 and ME6353), dated March 14, 2012, (ADAMS Accession No. ML120530591).

20. Pacific Gas and Electric Company Letter to NRC DCL-08-059, Supplemental Response to Generic Letter 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized Water Reactors (Revision 1), Dated July 10, 2008 (ADAMS Accession No. ML081980104).

21. NUREG/CR-6808, Knowledge Base for the Effect of Debris on Pressurized Water Reactor Emergency Core Cooling Sump Performance, dated February 2003, (ADAMS Accession No. ML030780733).

22. NRC Document, Staff Observations of Testing for Generic Safety Issue 191 during January 27 to 30, 2009, Trip to the Alion Hydraulics Laboratory, dated February 23, 2009, (ADAMS Accession No. ML090500230).

23. Pressurized Water Reactor Owners Group Report PWROG-16073-P, Revision 0, TSTF-567 Implementation Guidance, Evaluation of In-Vessel Debris Effects, Submittal Template for Final Response to Generic Letter 2004-02 and FSAR Changes, dated February 28, 2020.

24. FENOC Letter to NRC L-20-144, License Amendment Request - Application to Revise Technical Specifications to Adopt TSTF-567, Add Containment Sump TS to Address GSI-191 Issues, dated July 10, 2020 (ADAMS Accession No. ML20192A210).

25. Technical Specifications Task Force Letter TSTF-17-04, Transmittal of TSTF-567, Revision 1, Add Containment Sump TS to Address GSI-191 Issues, dated August 2, 2017 (ADAMS Accession No. ML17089A628).

v t e Letter Sequence Request
EPID:L-2017-LRC-0000, TSTF-567, TSTF-567, TSTF-567, TSTF-567, TSTF-567 (Approved, Closed)
Initiation Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request Acceptance... Supplement, Supplement Administration Meeting, Meeting, Meeting, Meeting, Meeting, Meeting, Meeting Questions 23 August 2022 2 March 2021 16 July 2021 22 April 2022 9 September 2022 21 December 2022 11 January 2024 Answers 27 May 2021 Results Approval Other: ML21161A299, ML21203A314, ML21211A082, ML21214A178, ML21260A161, ML22013A339, ML22019A279, ML22061A056, ML22109A175, ML22164A861, ML22167A170, ML23188A020, ML23193A938, ML23200A183, ML23243A910
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