ML22019A265

Us NRC Staff Review of Documentation Provided by Firstenergy Nuclear Operating Co. for Beaver Valley, Units 1&2 Concerning Resolution of Generic Letter 2004-02 - Potential Impact of Debris Blockage on Emergency Recirculation During Design B

ML22019A265
Person / Time
Site:	Beaver Valley
Issue date:	01/27/2022
From:	Victor Cusumano NRC/NRR/DSS/STSB
To:	James Danna Office of Nuclear Reactor Regulation
Russell A, NRR/DSS, 301-415-8553
Shared Package
ML22013A372	List: ML22019A265
References
EPID L-2017-LRC-0000, GL 2004-02
Download: ML22019A265 (42)

Text

January 27, 2022 MEMORANDUM TO: James G. Danna, Chief Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation FROM: Victor G. Cusumano, Chief /RA/

Technical Specification Branch Division of New and Renewed Licenses Office of Nuclear Reactor Regulation

SUBJECT:

BEAVER VALLEY POWER STATION, UNIT NOS. 1 AND 2 -

CLOSEOUT OF GENERIC LETTER 2004-02, POTENTIAL IMPACT OF DEBRIS BLOCKAGE ON EMERGENCY RECIRCULATION DURING DESIGN BASIS ACCIDENTS AT PRESSURIZED-WATER REACTORS (EPID L-2017-LRC-0000)

DATED By letter dated May 16, 2013 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML13136A144), FirstEnergy Nuclear Operating Company (FeNoc),

which was later renamed Energy Harbor Nuclear Corporation (the licensee) stated that they will pursue Option 2 (deterministic) for the closure of Generic Safety Issue (GSI)-191 and GL 2004-02 for Beaver Valley Power Station Units 1 and 2 (Beaver Valley).

On July 23, 2019 (ADAMS Package Accession No. ML19203A303), GSI-191 was closed. It was determined that the technical issues identified in GSI-191 were now well understood and therefore GSI-191 could be closed. Prior to and in support of closing the GSI, the Office of Nuclear Reactor Regulation (NRR) staff issued a technical evaluation report on in-vessel downstream effects (IVDEs) (ADAMS Accession Nos. ML19178A252 and ML19073A044 (non-public version)). Following the closure of the GSI, NRR staff also issued review guidance for IVDEs to support review of the GL 2004-02 responses NRC Staff Review Guidance for In-Vessel Downstream Effects Supporting Review of Generic Letter 2004-02 Responses (ADAMS Accession No. ML19228A011).

The NRC staff has performed a thorough review of the licensees responses to requests for additional information and supplements to GL 2004-02. Based on the evaluations, the Nuclear Regulatory Commission staff finds the licensee has provided adequate information as requested by GL 2004-02.

CONTACTS: Steve Smith, NRR/DSS/STSB 301-415-3190 Andrea Russell, NRR/DSS/STSB 301-415-8553

J. Danna The stated purpose of GL 2004-02 was focused on demonstrating compliance with Title 10 of the Code of Federal Regulations, Part 50.46. Specifically, the GL requested addressees to perform an evaluation of the emergency core cooling system (ECCS) and core spray system (CSS) recirculation and, if necessary, take additional action to ensure system function in light of the potential for debris to adversely affect long-term core cooling (LTCC). The NRC staff finds the information provided by the licensee demonstrates that debris will not inhibit the ECCS or CSS performance following a postulated loss-of-coolant accident. Therefore, the ability of the systems to perform their safety functions, to assure adequate LTCC following a design-basis accident, as required by 10 CFR 50.46, has been demonstrated.

Therefore, the NRC staff finds the licensees responses to GL 2004-02 are adequate and considers GL 2004-02 closed for Beaver Valley.

Docket Nos: 50-334 and 50-412

Enclosure:

Safety Evaluation

ML22019A265 *via e-mail OFFICE NRR/DSS/STSB* NRR/DSS/STSB* NRR/DSS/STSB/BC*

NAME SSmith ARussell VCusumano DATE 9/13/2021 9/13/2021 1/27/2022 U.S. NUCLEAR REGULATORY COMMISSION STAFF REVIEW OF THE DOCUMENTATION PROVIDED BY FIRSTENERGY NUCLEAR OPERATING COMPANY FOR BEAVER VALLEY POWER STATION UNIT 1 AND UNIT 2 CONCERNING RESOLUTION OF GENERIC LETTER 2004-02 POTENTIAL IMPACT OF DEBRIS BLOCKAGE ON EMERGENCY RECIRCULATION DURING DESIGN-BASIS ACCIDENTS AT PRESSURIZED-WATER REACTORS

1.0 INTRODUCTION

A fundamental function of the emergency core cooling system (ECCS) is to recirculate water that has collected at the bottom of the containment through the reactor core following a break in the reactor coolant system (RCS) piping to ensure long-term removal of decay heat from the reactor fuel. Leaks from the RCS, hypothetical scenarios known as loss-of-coolant accidents (LOCAs), are part of a plants design-basis. Hence, nuclear plants are designed and licensed with the expectation that they are able to remove reactor decay heat following a LOCA to prevent core damage. Long-term cooling following a LOCA is a basic safety function for nuclear reactors. The recirculation sump provides a water source to the ECCS in pressurized-water reactors (PWRs) once the primary water source has been depleted.

If a LOCA occurs, piping thermal insulation and other materials may be dislodged by the two-phase coolant jet emanating from the broken RCS pipe. This debris may transport, via flows coming from the RCS break or from the containment spray system (CSS), to the pool of water that collects at the bottom of containment following a LOCA. Once transported to the sump pool, the debris could be drawn towards the ECCS sump strainers, which are designed to prevent debris from entering the ECCS and the reactor core. If this debris were to clog the strainers and prevent coolant from entering the reactor core, containment cooling could be lost and result in core damage and containment failure.

It is also possible that some debris would pass through (termed bypass) the sump strainer and lodge in the reactor core. This could result in reduced core cooling and potential core damage.

If the ECCS strainer were to remain functional, even with core cooling reduced, containment cooling would be maintained, and the containment function would not be adversely affected.

Findings from research and industry operating experience raised questions concerning the adequacy of PWR sump designs. Research findings demonstrated that, compared to other LOCAs, the amount of debris generated by a high-energy line break (HELB) could be greater.

The debris from a HELB could also be finer (and thus more easily transportable) and could be comprised of certain combinations of debris (i.e., fibrous material plus particulate material) that could result in a substantially greater flow restriction than an equivalent amount of either type of debris alone. These research findings prompted the U.S. Nuclear Regulatory Commission (NRC) to open Generic Safety Issue (GSI) - 191, Assessment of Debris Accumulation on PWR Sump Performance, in 1996. This resulted in new research for PWRs in the late 1990s.

Enclosure

GSI-191 focuses on reasonable assurance that the provisions of Title 10 of the Code of Federal Regulations (10 CFR) Section 50.46(b)(5) are met. This rule, which is deterministic, requires maintaining long-term core cooling (LTCC) after initiation of the ECCS. The objective of GSI-191 is to ensure that post-accident debris blockage will not impede or prevent the operation of the ECCS and CSS in recirculation mode at PWRs during LOCAs or other HELB accidents for which sump recirculation is required. The NRC completed its review of GSI-191 in 2002 and documented the results in a parametric study that concluded that sump clogging at PWRs was a credible concern.

GSI-191 concluded that debris clogging of sump strainers could lead to recirculation system ineffectiveness as a result of a loss of net positive suction head (NPSH) for the ECCS and CSS recirculation pumps. Resolution of GSI-191 involves two distinct but related safety concerns:

(1) potential clogging of the sump strainers that results in ECCS and/or CSS pump failure; and (2) potential clogging of flow channels within the reactor vessel because of debris bypass of the sump strainer (in-vessel effects). Clogging at either the strainer or in-vessel channels can result in loss of the long-term cooling safety function.

After completing the technical assessment of GSI-191, the NRC issued Bulletin 03-01, Potential Impact of Debris Blockage on Emergency Sump Recirculation at Pressurized-Water Reactors (Agencywide Documents Access and Management System (ADAMS) Accession No. ML031600259), on June 9, 2003. The Office of Nuclear Reactor Regulation (NRR) requested and obtained the review and endorsement of the bulletin from the Committee to Review Generic Requirements (CRGR) (ADAMS Accession No. ML031210035). As a result of the emergent issues discussed in Bulletin 03-01, the NRC staff requested an expedited response from PWR licensees on the status of their compliance of regulatory requirements concerning the ECCS and CSS recirculation functions based on a mechanistic analysis. The NRC staff asked licensees, who chose not to confirm regulatory compliance, to describe any interim compensatory measures that they had implemented or will implement to reduce risk until the analysis could be completed. All PWR licensees responded to Bulletin 03-01. The NRC staff reviewed all licensees Bulletin 03-01 responses and found them acceptable.

In developing Bulletin 03-01, the NRC staff recognized that it might be necessary for licensees to undertake complex evaluations to determine whether regulatory compliance exists in light of the concerns identified in the bulletin and that the methodology needed to perform these evaluations was not currently available. As a result, that information was not requested in Bulletin 03-01, but licensees were informed that the NRC staff was preparing a generic letter (GL) that would request this information. GL 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), was the follow-on information request referenced in Bulletin 03-01. This document set the expectations for resolution of PWR sump performance issues identified in GSI-191, to ensure the reliability of the ECCS and CSS at PWRs. NRR requested and obtained the review and endorsement of the GL from the CRGR (ADAMS Accession No. ML040840034).

The GL 2004-02 requested that addressees perform an evaluation of the ECCS and CSS recirculation functions in light of the information provided in the letter and, if appropriate, take additional actions to ensure system function. Additionally, addressees were requested to submit the information specified in GL 2004-02 to the NRC. This request is based on the

identified potential susceptibility of PWR recirculation sump screens to debris blockage during design-basis accidents requiring recirculation operation of ECCS or CSS and on the potential for additional adverse effects due to debris blockage of flow paths necessary for ECCS and CSS recirculation and containment drainage. The GL 2004-02 required addressees to provide the NRC a written response in accordance with 10 CFR 50.54(f).

By letter dated May 28, 2004 (ADAMS Accession No. ML041550279), the Nuclear Energy Institute (NEI) submitted a report describing a methodology for use by PWRs in the evaluation of containment sump performance. NEI requested that the NRC review the methodology. The methodology was intended to allow licensees to address and resolve GSI-191 issues in an expeditious manner through a process that starts with a conservative baseline evaluation. The baseline evaluation serves to guide the analyst and provide a method for quick identification and evaluation of design features and processes that significantly affect the potential for adverse containment sump blockage for a given plant design. The baseline evaluation also facilitates the evaluation of potential modifications that can enhance the capability of the design to address sump debris blockage concerns and uncertainties and supports resolution of GSI-191. The report offers additional guidance that can be used to modify the conservative baseline evaluation results through revision to analytical methods or through modification to the plant design or operation.

By letter dated December 6, 2004 (ADAMS Accession No. ML043280641), the NRC issued an evaluation of the NEI methodology. The NRC staff concluded that the methodology, as approved in accordance with the NRC staff safety evaluation (SE), provides an acceptable overall guidance methodology for the plant-specific evaluation of the ECCS or CSS sump performance following postulated design-basis accidents(DBAs).

In response to the NRC staff SE conclusions on NEI 04-07 Pressurized Water Reactor Sump Performance Evaluation Methodology (ADAMS Accession Nos. ML050550138 and ML050550156), the Pressurized Water Reactor Owners Group (PWROG) sponsored the development of the following Westinghouse Commercial Atomic Power (WCAP) Topical Reports (TRs)

• TR-WCAP-16406-P-A, Evaluation of Downstream Sump Debris Effects in Support of GSI-191, Revision 1 (not publicly available), to address the effects of debris on piping systems and components (SE at ADAMS Accession No. ML073520295).

• TR-WCAP-16530-NP-A, Evaluation of Post-accident Chemical Effects in Containment Sump Fluids to Support GSI-191, issued March 2008 (ADAMS Accession No. ML081150379), to provide a consistent approach for plants to evaluate the chemical effects that may occur post-accident in containment sump fluids (SE at ADAMS Accession No. ML073521072).

• TR-WCAP-16793-NP-A, Evaluation of Long-Term Cooling Considering Particulate, Fibrous and Chemical Debris in the Recirculating Fluid, Revision 2 issued July 2013 (ADAMS Accession No. ML13239A114), to address the effects of debris on the reactor core (SE at ADAMS Accession No. ML13084A154).

The NRC staff reviewed the TRs and found them acceptable to use (as qualified by the limitations and conditions stated in the respective SEs). A more detailed evaluation of how the TRs were used by the licensee is contained in the evaluations below.

After the NRC staff evaluated licensee responses to GL 2004-02, the NRC staff found that there was a misunderstanding between the industry and the NRC on the level of detail necessary to respond to GL 2004-02. The NRC staff in concert with stakeholders developed a content guide for responding to requests for additional information (RAIs) concerning GL 2004-02. By letter dated August 15, 2007 (ADAMS Accession No. ML071060091), the NRC issued the content guide describing the necessary information to be submitted to allow the NRC staff to verify that each licensees analyses, testing, and corrective actions associated with GL 2004-02 are adequate to demonstrate that the ECCS and CSS will perform their intended function following any DBA. By letter dated November 21, 2007 (ADAMS Accession No. ML073110389), the NRC issued a revised content guide.

The content guide described the following information needed to be submitted to the NRC:

• Corrective Actions for GL 2004-02

• Break Selection

• Debris Generation/Zone of Influence (ZOI) (Excluding Coatings)

• Debris Characteristics

• Latent Debris

• Debris Transport

• Head Loss and Vortexing

• Net Positive Suction Head

• Coatings Evaluation

• Debris Source Term

• Screen Modification Package

• Sump Structural Analysis

• Upstream Effects

• Downstream Effects - Components and Systems

• Downstream Effects - Fuel and Vessel

• Chemical Effects

• Licensing Basis Based on the interactions with stakeholders and the results of the industry testing, the NRC staff, in 2012, developed three options to resolve GSI-191. These options were documented and proposed to the Commission in 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. ML121320270). The options are summarized as follows:

• Option 1 would require licensees to demonstrate compliance with 10 CFR 50.46, Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors, through approved models and test methods. These will be low fiber plants with less than 15 grams of fiber per fuel assembly

• Option 2 requires implementation of additional mitigating measures and allows additional time for licensees to resolve issues through further industry testing or use of a risk informed approach.

o Option 2 Deterministic: Industry to perform more testing and analysis and submit the results for NRC review and approval (in-vessel only).

o Option 2 Risk Informed: Use the South Texas Project pilot approach currently under review with NRR staff.

• Option 3 involves separating the regulatory treatment of the sump strainer and in-vessel effects.

The options allowed industry alternative approaches for resolving GSI-191. The Commission issued a Staff Requirement Memorandum on December 14, 2012 (ADAMS Accession No. ML12349A378), approving all three options for closure of GSI-191.

By letter dated May 16, 2013 (ADAMS Accession No. ML13136A144), FirstEnergy Nuclear Operating Company (the licensee) stated that they will pursue Option 2 (deterministic) for the closure of GSI-191 and GL 2004-02 for Beaver Valley Power Station Units 1 and 2 (Beaver Valley).

On July 23, 2019 (ADAMS Package Accession No. ML19203A303), GSI-191 was closed. It was determined that the technical issues identified in GSI-191 were now well understood and therefore GSI-191 could be closed. Prior to and in support of closing the GSI, NRR staff issued a technical evaluation report on in-vessel downstream effects (IVDEs) (ADAMS Accession Nos.

ML19178A252 and ML19073A044 (non-public version)). Following the closure of the GSI, NRR staff also issued review guidance for IVDEs to support review of the GL 2004-02 responses NRC Staff Review Guidance for In-Vessel Downstream Effects Supporting Review of Generic Letter 2004-02 Responses (ADAMS Accession No. ML19228A011).

The following is a list of documentation provided by the licensee in response to GL 2004-02:

RESPONSES TO GL 2004-02 DOCUMENT DATE ACCESSION NUMBER DOCUMENT TYPE March 4, 2005 ML050680211 90-day response June 3, 2005 ML051530267 1st NRC RAI July 22, 2005 ML052080167 Licensee Response September 6, 2005 ML052510411 Supplemental response February 9, 2006 ML060380342 2nd NRC RAI April 3, 2006 ML060960442 Supplemental response February 29, 2008 ML080660597 Licensee Response October 29, 2008 ML083080094 Supplemental Response March 11, 2009 ML090750619 Supplemental Response June 30, 2009 ML091830390 Supplemental Response February 18, 2010 ML100290318 3rd NRC RAI

September 28, 2010 ML102770023 Licensee Response May 16, 2013 ML13136A144 Closure letter November 30, 2020 ML20335A564 Final Supplemental

Response

March 23, 2021 ML21082A494 RAI May 27, 2021 ML21147A146 Licensee Response The NRC staff reviewed the information provided by the licensee in response to GL 2004-02 and all RAIs. The following is a summary of the NRC staff review.

2.0 GENERAL DESCRIPTION OF CORRECTIVE ACTIONS FOR THE RESOLUTION OF GL-2004-02 GL 2004-02 Requested Information Item 2(b) requested a general description of and implementation schedule for all corrective actions. The following is a list of corrective actions completed by the licensee at Beaver Valley in support of the resolution of GL 2004-02:

• Evaluation of long term cooling debris effects using the guidance of NEI 04-07.

• Ex-vessel downstream effects evaluation using the TR-WCAP-16406-P-A, Revision 1 methodology.

• Replaced or modified high pressure safety injection (HPSI) cold leg throttle valves to prevent blockage with debris.

• IVDEs evaluation using the TR-WCAP-17788-P methodology, Comprehensive Analysis and Test Program for GSI-191 Closure, Revision 1 (ADAMS Package Accession No. ML20010F181).

• Containment walkdowns using the guidance of NEI 02-01, Condition Assessment Guidelines: Debris Sources Inside PWR Containments, April 19, 2002 (ADAMS Accession No. ML021490212).

• The modification process and maintenance process have been enhanced relative to GL 2004-02 controls to insure operability of the containment sumps.

• Installation of new strainers (Unit 1 approximately 3,493 square feet (ft2), Unit 2 3,396 ft2).

• Removal of insulation from ZOI and any area in containment that could result in debris generation not bounded by strainer testing.

• ECCS sump strainer performance was confirmed by performing a prototype head loss test including chemical precipitates.

• Changed the logic for recirculation spray system (RSS) pumps to ensure that the strainers are fully submerged prior to being placed in service.

• Removed structures that could result in debris collection and holdup of water intended to drain to the ECCS sump.

• Removed aluminum from containment to reduce chemical source term.

• Replaced borated Temp-Mat insulation encapsulated in reflective metal insulation (RMI) with RMI on Unit 1 reactor vessel closure head.

• Replaced borated Temp-Mat insulation encapsulated in RMI with RMI on Unit 2 reactor vessel closure head flange. Replaced Min-K insulation encapsulated in RMI on portions of the Unit 2 RCS piping with Transco Thermal Wrap insulation encapsulated in RMI.

• Installed new RMI on Unit 1 replacement steam generators and associated piping.

• Implemented a containment coatings inspection and assessment program and a containment cleaning program.

• Replaced fibrous and particulate insulation in Unit 1 RCS loop cubicles with RMI, including insulation on reactor coolant loop piping.

• Conducted loss-of-coolant accident deposition model analyses in accordance with WCAP-16793 that confirmed the maximum clad temperature and total deposition thickness limits of WCAP-16793 are satisfied.

• Updated method for calculating NPSH available to the recirculation spray (RS) pumps by crediting containment overpressure.

• Replaced Min-K insulation on Unit 2 steam generator level instrumentation tubing with jacketed Thermal Wrap.

• Replaced NUKON fibrous insulation blankets on Unit 2 pressurizer power operated relief valves (PORVs) and associated inlet piping and pipe supports with jacketed Thermal Wrap insulation secured with Sure-Hold banding, where practical.

• Removed additional portions of calcium silicate insulation from Unit 2 RCS loop cubicles in 2011 and Unit 1 in 2010.

• Replaced fibrous and particulate insulation in Unit 2 RCS loop cubicles with RMI.

• Replaced liquid sodium hydroxide (NaOH) chemical addition systems with baskets of powdered sodium tetraborate (NaTB) to lower pH and reduce the quantity of post-LOCA chemical precipitates.

• Modified procedures to install grating over all six hatches between the reactor and refueling cavities to provide additional pressure relief pathways for a reactor vessel nozzle break.

• Performed a re-evaluation of the quantity of tape in Unit 1.

• Removed or qualified a significant portion of unqualified tags, labels, and tape in Unit 2 containment.

• Revised emergency operating procedures (EOPs) for Unit 1 to secure two RS pumps prior to transfer to ECCS recirculation.

• Revised EOPs for Unit 2 to shut down one of the RS pumps supplying the spray header when containment pressure is reduced below a predetermined value.

• Revised EOPs to initiate early transfer to hot leg recirculation (Unit 2) or simultaneous hot and cold leg recirculation (Unit 1) should plant parameters indicate a core blockage.

Based on the actions taken and information provided by the licensee, the NRC staff considers GL 2004-02 closed.

3.0 BREAK SELECTION The objective of the break selection process is to identify the break sizes and locations that present the greatest challenge to post-accident sump performance. The term ZOI used in this section refers to the spherical zone representing the volume of space affected by a hypothetical piping rupture.

NRC STAFF REVIEW:

The NRC staff review is based on documentation provided by the licensee through November 30, 2020. The guidance documents used for the review include the revised content guide dated November 21, 2007, Regulatory Guide (RG) 1.82, Water Sources for Long-Term Recirculation Cooling Following a Loss-of-Coolant Accident, and the NEI 04-07 guidance report (GR) and associated NRC staff SE (GR/SE).

The licensee used the approved methodology as stated in the GR/SE for break selection. The limiting characteristics of break selection analyzed by the licensee include:

• breaks with the largest potential for debris generation,

• large breaks with two or more different types of debris,

• breaks with the most direct path to the sump,

• medium and large breaks with the highest particulate to fibrous debris ratio, and

• breaks that generate a thin bed For both units, the licensee evaluated large breaks in the RCS loops and at the reactor pressure vessel nozzles. The licensee also evaluated smaller RCS piping including the pressurizer surge

line and other branch piping. The RCS loop piping breaks were initially found to have a greater amount of fibrous debris and particulate debris. The analysis of the debris sources for both units from an RCS loop piping break concluded that the fibrous bed, when applied uniformly across the screen, would result in a fiber bed being less than 1/8 inch. However, the total amount and resulting fiber bed thickness is greater than the thickness resulting from the limiting breaks which are the pressurizer surge line break (for Unit 1) and the reactor cavity nozzle break (for Unit 2) and has a higher impact on head loss. Later discovery of fibrous insulation near the safety relief valve lines on Unit 1 resulted in a higher fibrous debris source, but testing showed that the head loss from the overall debris load for this break was lower than the limiting break. On Unit 2, a similar discovery found that a break on the PORV line could produce more fibrous debris than the limiting RCS loop break. For this condition, the licensee committed to remediating the insulation in the area of the PORV line to reduce the potential amount of debris generation so that it is no longer the limiting break. The limiting breaks for head loss were identified by the break selection process.

During the resolution of GL 2004-02, the licensee discovered additional debris sources that were not originally considered in the break selection evaluation. The licensee appropriately added the new debris sources to their evaluations and removed debris sources as necessary to ensure that the debris loads are maintained within the design bounds. During the evaluation, the licensee stated that additional break locations were evaluated to ensure that the potential debris loads were bounded by the analysis.

The spectrum of breaks evaluated by the licensee is consistent with that recommended in staff-approved methodology in the GR/SE. The licensee performed debris generation calculations for breaks in the RCS loop piping and smaller piping attached to the main loops.

The licensee also evaluated breaks at the reactor nozzles. The majority of debris generated by the various breaks is RMI which is not a major contributor to strainer head loss. Testing for Beaver Valley Units 1 and 2 bounded the worst-case debris loads.

In addition to the above, secondary breaks were considered in the break selection process.

Secondary breaks do not require the plants to enter the recirculation mode for safe shutdown.

Therefore, these breaks are not required to be evaluated for debris generation.

The licensee stated that the plant modifications discussed in the September 28, 2010 RAI 1 response were completed where NUKON insulation on the Unit 2 pressurizer PORV inlet piping was replaced with stainless steel jacketed Thermal Wrap insulation secured with Sure-Hold bands where practical.

NRC STAFF CONCLUSION:

The break location analysis completed by the licensee is in accordance with the revised content guide dated November 21, 2007 and the GR/SE and used an acceptable method to determine break locations to analyze for maximum debris generation results. The NRC staff finds the licensees method of choosing the limiting break locations that present the greatest challenge to post accident sump performance acceptable. Therefore, the NRC staff concludes that the break selection evaluation for Beaver Valley Units 1 and 2 is acceptable. Based on the information provided by the licensee, the NRC staff considers this area closed for GL 2004-02 for Beaver Valley Units 1 and 2.

4.0 DEBRIS GENERATION/ZONE OF INFLUENCE (EXCLUDING COATINGS)

The objective of the debris generation/ZOI evaluation is to determine the limiting amounts and combinations of debris that can occur from the postulated breaks in the RCS.

NRC STAFF REVIEW:

The NRC staff review is based on documentation provided by the licensee through May 27, 2021. The NRC staff initial review found that the methodology being used by the licensee was per approved guidance. Because the licensee discovered potentially problematic debris sources and planned to revise the debris generation methodology this area remained open until all potential sources were adequately evaluated.

The licensees evaluation of debris generation generally followed NEI 04-07 guidance, except as discussed below. The licensee used the GR Section 4.2.2.1.1 ZOI refinement of debris-specific spherical ZOI to specify different ZOIs for different material types depending upon the robust nature of the materials.

The licensee state that the following ZOIs were assumed per the approved methodology (GR/SE):

• 2.0D for Transco RMI,

• 28.6D for Diamond Mirror RMI,

• 17.0D for NUKON and Thermal Wrap (not secured with Sure-Hold Banding)

• 11.7D for Temp-Mat with stainless steel wire retainer

• 5.45D for Cal-Sil (aluminum clad, stainless steel bands)

• 28.6D for Encapsulated Min-K

• 2.4D for Encapsulated Thermal Wrap with Sure-Hold Banding installed The licensee provided insulation debris quantities for 5 evaluated breaks.

The licensee initially determined by walk down that signs, placards, tags, tape, and similar miscellaneous materials amounted to 543.9 ft2 in Unit 1 and 750.8 ft2 in Unit 2. These values were reduced, most significantly on Unit 2, by removing or qualifying miscellaneous materials within the containments. The final values for miscellaneous materials are 341 ft2 for Unit 1 and 59 ft2 for Unit 2.

After the initial review of debris generation, the licensee performed strainer head loss testing for both units to identify acceptable debris loading for the strainers. The debris loads were determined using approved guidance, so the staff determined that the debris generation area had been acceptably addressed.

After the head loss testing had been completed, the licensee identified additional debris sources within containment. One additional material that was identified is Benelex 401. Benelex 401 was stated to be a high density wood based shielding material. The licensee stated that the material had been evaluated to withstand a pipe rupture of 120 pounds per square inch gauge (psig) and therefore assumed a 2D ZOI. Details of the evaluation were not provided.

Therefore, additional information was required to allow the staff to evaluate the acceptability of the ZOI assigned to this material. The licensee responded that the Benelex material is installed

only in Unit 1 and that the material had been evaluated and found to be a hard stable material that can withstand RCS blowdown and the post-LOCA environment over a 30 day period. The detailed material characteristics were provided in an appendix to the RAI response submittal.

The potential for Benelex degradation due to water spray and immersion, erosion, and chemical effects were evaluated and shown to be non-problematic. In addition, the licensee stated that due to its location the Benelex would not be subjected to jet impingement from a pipe break.

Based on these conditions, the licensee determined that jet impingement and post-LOCA environment are not credible damage mechanisms for the Benelex.

One other issue the staff identified in Beaver Valleys response was that the predicted Cal-Sil debris generation was reduced from 100 percent as recommended in the NEI 04-07 guidance to 50 percent. The licensee revised their methodology to include 100 percent of the Cal-Sil within the ZOI as debris. In addition, the licensee stated that any Cal-Sil greater than the amount included in strainer head loss testing would be removed thus ensuring that the testing bounds the plant condition.

The licensee discovered additional fibrous insulation within the containment that could become debris following a LOCA. As described in the Break Selection area (Section 3.0) above, this source material was appropriately evaluated by performing head loss testing (Unit 1) or remediated (Unit 2).

The licensee used an alternate methodology to calculate a ZOI for Microtherm installed on the RCS near the reactor vessel nozzles. The methodology was used by boiling-water reactors to evaluate breaks that could not fully separate because of restraints that would limit the displacement of the piping following a break. The use of the Utility Resolution Guide for ECCS Suction Strainer Blockage methodology was not approved by the staff for PWRs. Therefore, the staff requested that the licensee provide additional information to justify the use of the alternate method. The licensee stated that a 40 psi destruction pressure, resulting in a 4D ZOI was used in the evaluation. The destruction pressure basis was provided in the June 30, 2009 supplemental response and was based on construction similar to an RMI type that has a destruction pressure of 114 psi, and additional protection from relatively robust neutron shielding. The licensee evaluated the displacements of the piping at the potential break locations using a computer model and used displacement values larger than those calculated by the computer model for ZOI sizing. The break displacements were used to determine jet characteristics as defined in the ANSI/ANS 58.2 standard for jet modeling. The licensee used multiple methods to reshape the ZOI for each break and selected the break and ZOI shape that resulted in the largest debris generation volume. The methodology included reshaping the ZOI to account for solid volumes within the ZOI to conserve jet forces within the confined space of the reactor annulus without accounting for pressure losses due to deflections. The methodology assumed that any cassette with any portion inside the ZOI was completely destroyed. The largest potential Microtherm debris generation from a reactor nozzle break was found to be 20.2 cubic feet (ft3) and represents about 1/3 of the Microtherm installed on the reactor and reactor nozzles. The licensee also stated that additional strainer head loss testing had been performed for Microtherm debris loading and that the testing included Microtherm amounts greater than those calculated by the alternate ZOI calculation discussed in this RAI.

In its March 23, 2021 RAI, the NRC staff requested the licensee to state which strainer head loss test was used to evaluate the Unit 2 pressurizer-relief valve (PRV) line break, In its

response, the licensee used an alternate methodology to calculate the amount of Thermal Wrap insulation that could be generated. The method combined two previously approved methods and evaluated debris generated within a 2.4D ZOI for the encapsulated Thermal Wrap protected by Sure Hold Bands. One of the methods provided a four-zone debris size distribution. The second method was the baseline two-zone method from the GR/SE. The licensee did not address the size distribution of the material within the 2.4D ZOI for Sure Hold Bands separately as they had in the November 30, 2020 submittal of where they assumed 100 percent fine debris was generated within a 2.4D ZOI. Using the two models, the licensee calculated that the fibrous Thermal Wrap debris generation for the PRV line break would be 2.9 ft3 of fine debris and 8.7 ft3 of small debris and that all of this would transport to the strainer. The NRC staff did not agree with the assumptions used by the licensee for its debris generation calculation for this break.

The NRC staff concluded that all of the debris generated within a 2.4D ZOI should be fine as originally assumed by the licensee (refer to ADAMS Accession No. ML110140145, NRC letter dated October 3, 2014, transmitting revision to staff SE on NEI 04-07). Also, the licensee neglected to account for erosion of small and large pieces to fines in its updated analysis.

Finally, the combination of two different debris generation models is questionable and appears to have been performed to minimize the amount of fine debris reaching the strainer. These factors all reduced the amount of fine debris that can reach the strainer from the Unit 2 PRV line break.

In its November 30, 2020, submittal, the licensee identified Foamglas as a debris source for the Unit 1 Loop large-break LOCA (LBLOCA) case. This type of debris had not been previously identified. In its March 23, 2021 RAI, the NRC staff requested the licensee to describe how Foamglas was accounted for in testing for this break case. In its response dated May 27, 2021, the licensee stated that the Foamglas is considered to fail as particulate and stated that the debris amounts in Test 6 bound those for the Unit 1 Loop LBLOCA scenario. The staff reviewed the referenced information and found the response acceptable. The staff also noted that more recent information on Foamglas indicates that it is unlikely to significantly affect head loss.

The majority of potential debris materials and their amounts were evaluated using staff approved guidance. There were three cases where guidance did not exist or was not followed by the licensee. One case was for the Benelex material which had not previously been evaluated by the NRC. Another case was the use of a ZOI for Microtherm based on a break size reduced due to limited piping displacement. These two cases are discussed in more detail below. The third case involved a discrepancy in calculating the fibrous debris amounts for the Unit 2 PRV line break.

For the evaluation of the Microtherm in the reactor cavity, the 40 psi destruction pressure used by the licensee was evaluated by the staff and determined to be adequate based on similarity of design to an RMI type with a significantly higher destruction pressure, and the presence of neutron shielding which is relatively robust and shields a majority of the Microtherm except that on the broken piping. The staff also found that the pipe separation values used to determine the ZOIs were conservative since they were larger than separations predicted by the licensee. The calculation of the ZOI also contained adequate conservatism such that the staff concludes that the volume of Microtherm predicted to become debris is acceptable. Even though the potential for destruction within a confined volume has some uncertainty, the licensees evaluation contained adequate conservatism to ensure that the predicted amount of Microtherm generated is bounding of that which could actually be generated by the most limiting break that could affect

the debris source. Since the licensees strainer test program inputs included a larger amount of Microtherm than that calculated to become debris there is additional margin. Therefore, the staff found the evaluation of the Microtherm in the reactor cavity to be acceptable.

The staff concludes that Benelex 401 would not contribute to the debris loading or to chemical effects at Beaver Valley Unit 1 based on the chemical and physical properties of the material, the distance of the material from a potential break, and the robust barriers between the material and the postulated breaks.

To address the discrepancy in the fibrous debris calculation for the Unit 2 PRV line break, the NRC staff performed confirmatory calculations to estimate the amount of fine and small debris predicted to reach the strainer by applying the 4 zone model that was used by the licensee to estimate generation of fibrous debris for other breaks. The staff assumed all debris generated within a 2.4D ZOI is 100 percent fine and accounted for erosion of the large and small fibrous debris generated by the break at further distances. The NRC calculation found that 3.57 ft3 (8.57 pounds (lb.)) of fine debris and 8.23 ft3 (19.8 lb.) of small debris would be generated. The fine debris load for this break also includes 11.5 ft3 (27.6 lb.) of latent fiber for a total of 36.2 lb.

of fine fibrous debris. Test 1A included 31.7 lb. of Nukon fine and 16.5 lb. of Temp-Mat fine fiber on the plant scale for a total of 48.2 lb. The staff concluded that the substitution of some fine Temp-Mat, a high-density fiberglass (HDFG), for Thermal-Wrap, a low-density fiberglass (LDFG), was acceptable for this test because of conservatisms noted below and because LDFG and HDFG fine fiber are expected to behave relatively similarly in head loss testing. The test included 1.6 lb. of Nukon small piece debris and 16.5 lb. of Temp-Mat small piece debris. The staff did not agree that small piece HDFG can be used in a test to substitute for LDFG, especially on a mass basis. The NRC staff found that Test 1A was deficient in Thermal Wrap (LDFG) small piece debris by 18.2 lb. The test included 16.5 lb. of Temp-Mat (HDFG) small pieces. For the following reasons, the NRC staff found that there was adequate overall conservatism in the test to reasonably assure that the head loss from the test bounds that which could occur in the plant. The following discussion uses NRC calculated values for fiber because they are more conservative than the licensee values and align with NRC guidance. First, the test included about 12 lb. of excess fine fiber which would have a greater influence on head loss than the 18.2 lb. deficiency in small debris arising from crediting HDFG small pieces as equivalent to LDFG small pieces. The 16.5 lb. of Temp-Mat small pieces provides some debris of the correct size category. The test included Cal-Sil and Min-K which have significant effects on head loss and none of this type of debris is generated from the PRV line break. There are other conservatisms (e.g. debris generation and transport) in the overall head loss analysis that can be considered when looking at a single scenario without significantly reducing the overall conservatism in the analysis. Therefore, even though the staff disagrees with the licensees methodology for calculating the debris generation and transport values for this break, the overall analysis remains valid. This issue applies to other sections evaluated in this staff summary (head loss, debris characteristics, and transport sections).

NRC STAFF CONCLUSION:

The licensee used staff approved guidance to address debris generation/ZOI, except as described herein. After reviewing the responses to questions regarding methodologies and debris sources that were not conducted per approved guidance, the NRC staff concludes that those area of debris generation had been addressed in a manner that provides adequate

assurance that the strainer will perform its function. All other evaluations in the debris generation area followed approved guidance. Therefore, the NRC staff concludes that the debris generation/ZOI evaluation for Beaver Valley Units 1 and 2 is acceptable. The NRC staff considers this item closed for GL 2004-02.

5.0 DEBRIS CHARACTERISTICS The objective of the debris characteristics determination process is to establish a conservative debris characteristics profile for use in determining the transportability of debris and its contribution to strainer head loss.

NRC STAFF REVIEW:

The NRC staff review is based on documentation provided by the licensee through May 27, 2021. The NRC found that the licensee followed approved guidance to perform the debris characteristics evaluation in almost all areas. One material identified by the licensee had not been previously evaluated as a potential debris source, so the NRC performed a more detailed evaluation of that material, Benelex. The NRC also determined that the licensees assumptions regarding the debris characteristics of calcium silicate were potentially non-conservative.

The licensee referred to testing performed by Ontario Power Generation (OPG) which showed that less than 50 percent of calcium silicate became fine debris when exposed to a steam jet with pressure equivalent to that for the approved ZOI for the material. The staff reviewed the testing and determined that a significant portion of the target material during the test was not exposed to the intended jet pressure. This would likely result in less material becoming fine debris than if the entire target had been exposed to the intended pressure. The staff also questioned whether the installation of the calcium silicate at Beaver Valley was at least as robust as the system tested at OPG. To address this issue, the licensee changed its treatment of calcium silicate such that all of that type of material within the approved ZOI was assumed to be rendered into fine debris. This is consistent with NRC guidance and this question was resolved.

With respect to the Benelex, the licensee provided significant information regarding its material and chemical properties, and its location in containment with respect to potential LOCA jets.

The licensee stated that the material is present only in the Unit 1 containment and stated that it is a hard stable material that will withstand the initial impulse of a LOCA jet as well as mildly alkaline water associated with the Beaver Valley post-LOCA environment. The licensee provided information that justified that the Benelex is relatively distant from any potential LOCA break location and that there is intervening equipment that would prevent a jet from directly impinging the material. Based on the material properties and the location of the Benelex, the NRC determined that it would not become debris due to a LOCA jet because any impingement pressure would be well below that required to damage the material. The licensee also provided information to show that the Benelex is not water soluble, will not erode, and will not form particulate or fibrous debris under post-LOCA conditions. Therefore, the NRC staff found that Benelex did not need to be considered as a post-LOCA debris source for Beaver Valley.

In its November 30, 2020 submittal, the licensee stated that it also assumed that all Thermal-Wrap installed with Sure-Hold Bands would fail as 100 percent fines due to the

relatively small ZOI for this insulation system. This is conservative since fine fibrous material results in greater transport and strainer head loss. The licensees May 27, 2021 RAI response changed this assumption. See the Debris Generation Section for additional information regarding this issue.

NRC STAFF CONCLUSIONS:

The majority of potential debris materials were evaluated using staff approved guidance. There were three cases where guidance did not exist or was not followed by the licensee. These were assumptions Benelex material which had not previously been evaluated by the NRC, the use of a debris size distribution for calcium silicate that was not in accordance with NRC guidance, and the size distribution assumption for Thermal-Wrap secured with Sure-Hold Bands within a 2.4D ZOI.

The licensee used staff approved guidance to address debris characteristics, except as described above. The licensee decided to revert to the NRC approved characteristics for calcium silicate and provided information that justified that Benelex would not become a post-LOCA debris source. The NRC staff performed confirmatory calculations to address the Thermal-Wrap size distribution issue. After reviewing the licensees responses with respect to debris characteristics, the NRC staff concludes that the area had been addressed in a conservative manner. With the exception of the Benelex material and the Thermal-Wrap size distribution, the licensees final debris characteristics evaluation ultimately followed approved guidance. The licensees disposition of the Benelex material was found to be acceptable to the staff and the staff evaluated the Thermal-Wrap sizing issue. Therefore, the NRC staff concludes that the debris characteristics evaluation for Beaver Valley Units 1 and 2 is acceptable. The NRC staff considers this item closed for GL 2004-02.

6.0 LATENT DEBRIS The objective of the latent debris evaluation process is to provide a reasonable approximation of the amount and types of latent debris (e.g., miscellaneous fiber, dust, dirt) existing within the containment and its potential impact on sump screen head loss. The guidance documents used for the review include the Revised Content Guide dated November 2007, the GR/SE, and NEI 02-01.

NRC STAFF REVIEW:

The NRC staff review is based on documentation provided by the licensee through November 30, 2020. The NRC found that the licensee followed approved guidance to perform the latent debris evaluation.

The licensee performed containment walkdowns and sampling in accordance with NRC accepted guidance and also performed calculations to extrapolate the sample results to the entire containment. The licensee assumed that 15 percent of the latent debris was fibrous in accordance with approved guidance. The licensee also performed walkdowns to determine the amount of miscellaneous debris that could block the strainer area. The surface area of miscellaneous material identified during walkdowns was increased by 30 percent to ensure conservatism. Beaver Valley assumed that 75 percent of the signs, labels, etc. would be

available to block the strainer surface. This assumption is in accordance with approved guidance.

In its November 30, 2020 submittal, the licensee stated that it reduced the postulated miscellaneous debris load at Unit 2 by removing qualified tags from the miscellaneous debris inventory. It also removed metal tags as they will not transport to the containment sump strainers. The reduction of labels in Unit 2 were significant and result in a miscellaneous debris load of 59 ft2. The previous miscellaneous debris estimate was 750.8 ft2. The licensee also reduced the amount of tape assumed as miscellaneous debris in Unit 1. The original tape source term was determined during an outage when there is a significant amount of tape being used in containment for outage activities. The licensee conducted additional inspections at the end of a refueling outage, following a containment closeout inspection. Based on the inspection the miscellaneous debris load was reduced to from 543 ft2 to 341 ft2 for Unit 1.

NRC STAFF CONCLUSION:

The licensee used staff approved guidance to address the area of latent debris. After reviewing the licensees supplemental response and RAI responses, the NRC staff concludes that the area had been addressed in a conservative manner. The reduction in miscellaneous debris was based on inspection and verification that some tags would not fail or transport to the strainer.

These methods are consistent with NRC guidance. Therefore, the NRC staff concludes that the latent debris evaluation for Beaver Valley Units 1 and 2 is acceptable. The NRC staff considers this item closed for GL 2004-02.

7.0 DEBRIS TRANSPORT The objective of the debris transport evaluation process is to estimate the fraction of debris that would be transported from debris sources within containment to the sump suction strainers.

NRC STAFF REVIEW:

The NRC staff review is based on documentation provided by the licensee through May 27, 2021. The NRC found that the licensee followed approved guidance and made generally conservative assumptions in the performance of the debris transport evaluation.

The NRC staff determined that the debris transport results for Unit 1 are clearly conservative, with 100 percent transport assumed for almost all types of generated debris with RMI as the lone exception. Since RMI is not considered to be detrimental to strainer head loss under most conditions, including those at Beaver Valley, the staff found that transport was treated conservatively for Unit 1.

The results for Unit 2 also appear to be conservative, with all types of generated debris having an assumed transport percentage of 82 percent or higher, with most percentages set to 100 percent. For Unit 2, several analytical assumptions were made that appeared questionable, but these assumptions were compensated for by conservative assumptions that led to the final transport fractions that are conservative rather than best-estimate. The potential non-conservative assumptions identified by the staff include small debris capture on surfaces during blowdown and the hold up of small and large debris pieces on gratings.

The licensee credited small debris capture of 5 percent on surfaces based on the results of the Drywell Debris Transport Study which found that 17 percent of small debris would be inertially captured on surfaces when encountering a 90 degree bend in the transport path. The licensee stated that the transport path to upper containment was significantly tortuous so that some debris would be captured and assumed that 5 percent was a reasonable value.

The licensees assumption of holdup of small debris on gratings was not implemented when the debris loads for the head loss testing were performed. That is, during head loss testing, the debris loading was calculated assuming no hold up of small debris on gratings. The final treatment of small piece fiber debris for the head loss testing assumed that 18 percent of small pieces were held up in upper containment and eroded by spray flow at 1 percent. The remaining small fiber was all assumed to reach the strainer. The testing included 50 percent of the small fibrous debris as fines and 50 percent as small pieces. This ratio is conservative with respect to the expected size distribution at the strainer since finer debris is associated with higher head losses during testing.

The licensee did not assume and holdup of debris in inactive cavities and assumed 100 percent transport of all fine debris to the strainer. These assumptions significantly increase the overall conservatism in the strainer evaluation.

In its November 30, 2020 submittal, the licensee identified an additional break on the Unit 2 PRV line and provided debris transport metrics for that break. The NRC staff identified issues with the evaluation of this break because head loss testing did not appear to bound the amounts of debris predicted to be generated and transported for this break. The staff requested that the licensee provide additional information for the evaluation of this break. In its RAI 2 response dated May 27, 2021, the licensee provided information, but did not account for erosion of large and small debris for the Unit 2 PRV line break. Other aspects of the transport analysis for this break were conducted in accordance with NRC approved guidance. This issue is discussed in the debris generation section above.

NRC STAFF CONCLUSION:

The licensee used staff approved guidance to address the area of debris transport. After reviewing the licensees supplemental response and RAI responses, the NRC staff concludes that the area had been addressed in a conservative manner. The Unit 1 evaluation was clearly conservative. The Unit 2 evaluation contained some assumptions not fully justified by the licensee, but also contained conservatism that more than offset the potential non-conservatism of the aforementioned assumptions. Therefore, the NRC staff concludes that the debris transport evaluation for Beaver Valley Units 1 and 2 is acceptable. The NRC staff considers this item closed for GL 2004-02.

8.0 HEAD LOSS AND VORTEXING The objectives of the head loss and vortexing evaluations are to calculate head loss across the sump strainer and to evaluate the susceptibility of the strainer to vortex formation.

NRC STAFF REVIEW:

The NRC staff review is based on documentation provided by the licensee through May 27, 2021. The NRC found that the licensee followed approved guidance and made generally conservative assumptions in the performance of the head loss and vortexing evaluation with the exception of the Unit 2 PRV line break as discussed below. The evaluation for Beaver Valley had to consider differences between units because each unit used a different strainer as strainer supplier.

Prior to its June 30, 2009 supplemental response, the licensee did not provide significant detail regarding the head loss and vortexing evaluation. Until that time, the licensee had been working to reduce potential debris sources within containment and had not finalized the design basis debris load so that the final tests could not be performed.

One of the issues identified early in the review was that air could be ingested into the strainer due to water cascading in the area of the strainer and entraining air bubbles that could be sucked into the strainer and eventually transport to the ECCS pump suctions. The licensee addressed this issue by ensuring that the strainers were covered to prevent water from falling in the area. The licensee also determined that water falling in the area of the strainer would be in droplets which would be small enough so that significant air bubbles would not be created. The staff also noted that the strainer submergence increases relatively quickly reducing the time during which this phenomenon is able to result in air ingestion.

An additional consideration in the staff evaluation of air ingestion was the flow rate used during testing for vortex formation. The Beaver Valley strainers consist of long trains of modules connected in series. This configuration results in the flow at the modules closest to the pump suctions being much higher than the average flow, until debris arrives at the strainer. Vortex testing was conducted at twice the average flow velocity for the strainer. For Unit 2, the NRC found that the strainer submergence was very large so that vortex formation would not occur.

For Unit 1, the submergence is not as great. During testing, the submergence was at the top of the strainer which is at least 2 inches lower than the minimum water level for small breaks and significantly lower than for large breaks. The low level during testing is conservative. The licensee provided additional information regarding the vortex testing of the Unit 1 strainer and stated that the velocities present during the vendors generic vortex testing bounded the plant velocities. Results of the vortex analysis found that a very small amount of air entrainment may occur under the most conservative conditions, but that the amount would be less than 0.0012 percent which would have a negligible impact on pump performance. The staff found that the licensee had addressed the issue appropriately and that it was unlikely that any entrainment from vortexing would occur in the plant because the evaluation contains conservatism.

The NRC staff questioned the basis for the assumption that any gasses liberated from the fluid as it passed through the debris bed would be reabsorbed prior to the fluid reaching the pump suction. The licensee provided the results of additional calculations that do not credit the reabsorption of gasses into the fluid. The calculations show that the amount of gasses reaching the pump suctions is much lower than NRC guidance recommends. The licensee also determined that the longest time that air evolution may occur following a LOCA is less than 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />. The licensee also evaluated the potential for gasses to accumulate in the strainer.

Based on calculations and computational fluid dynamics analysis the licensee determined that

air bubbles will continue with flow through the strainer and be transported to the pumps. The licensee also stated that if the bubbles collect in the top of the strainer, the amount of gas collected would be small due to the short time that degasification may occur. The Unit 1 evaluation was most limiting because the velocities within the strainer are lower than those of Unit 2. The staff found that the licensee had appropriately evaluated the potential for degasification of the fluid and its effects.

The NRC staff requested additional information to validate that all debris sources predicted to reach the strainer were appropriately included in the strainer head loss testing. Information in the supplemental response caused the staff to conclude that some of the debris types had been under-represented. The licensee stated that testing included the appropriate amounts of debris and that the test amounts were based on mass and that the conversion of mass to density in equating different types of fiber resulted in the potential appearance of under-representing some fiber types in the testing. The NRC accepts the substitution of some types of fine fiber with other types on a limited bases as long as the mass of fiber is consistent. The NRC staff found that the correct amount of fibrous debris was included in the strainer testing.

The NRC staff questioned the assumption that only 0.25 ft3 of latent fiber was included in the debris source term. The licensee stated that this volume is based on a density of individual fibers and not on the as-manufactured density of fiber. The staff found that the appropriate latent fibrous debris amount, based on mass, was included during testing in the correct form.

The NRC staff also questioned the methodology used to calculate the clean strainer head loss (CSHL) for Unit 1 because it was not clear that the licensee had assumed equal flow through all strainer modules as is considered realistic by the staff. The Unit 2 CSHL appropriately assumed equal flow through each module. The licensee provided a physical description of the strainer layout and also provided details of how the CSHL was calculated. The licensee acknowledged that the presence of a debris bed would increase the CSHL, however, the licensee calls the debris laden CSHL clean side head loss. So, when there is no debris the clean side head loss is equal to the CSHL. As debris head loss increases the clean side head loss increases. The licensee calculated clean side head losses for a large range of debris loads and assumed the maximum clean side head loss in their NPSH calculations. The methods employed by the licensee are acceptable to the staff.

The NRC staff also questioned which water sources were credited for sump inventory for the small break LOCA (SBLOCA) to ensure that the strainer submergence was calculated correctly.

The staff was concerned that the accumulators may not discharge for all breaks, thus reducing sump inventory. The licensee stated that the sump inventory is calculated using a computer code that tracks the amount of water released from the RCS to the sump. The sources include the refueling water storage tank, reactor vessel, piping, pressurizer, and accumulators. The licensee stated that the code correctly determines the amount of inventory from each source and provided an example for small breaks for each unit. The NRC staff found that the licensee had properly accounted for the potential for the accumulators to not inject following a small break.

The NRC staff requested additional information on the debris characteristics of the fiber used during testing and the debris introduction procedures. The licensee stated that the it followed March 2008 staff guidance, NRC Staff Review Guidance Regarding Generic Letter 2004-02

Closure in the Area of Strainer Head Loss and Vortexing (ADAMS Accession No. ML080230038), for debris preparation and provided additional details regarding the debris preparation procedure. The licensee provided photographs of the prepared fiber which show that its size distribution met staff guidance for fine fiber in head loss testing. The licensee also stated that the March 2008 guidance was followed for debris introduction and provided additional details of the debris introduction procedures and the test set up. The NRC staff noted that they had witnessed tests at the test facility using a very similar test set up and had found the debris introduction to be acceptable for those tests. Based on the information provided, the NRC staff found the debris preparation and introduction methods used by the licensee to be acceptable.

The NRC requested a justification for adding paint chips prior to other debris during head loss testing when guidance states that the most easily transportable debris should be added first.

The licensee stated that the unqualified coatings were added as chips and also as particulate, doubling the coating load from this source, because it was not clear whether a fibrous debris bed would fully cover the strainer. Guidance states that if a fiber bed fully covers the strainer unqualified coatings should be added as particulate. If a fiber bed does not fully cover the strainer, it may be more conservative to add the unqualified coatings as chips. Guidance also allows the licensee to determine the characteristics of the coatings and the ability of the coating chips to transport to the strainer if they have plant specific information that justifies the chosen treatment. The particulate debris was added prior to the fine fibrous debris. The chips were also added prior to the fine fiber. The licensee stated that several steps were undertaken to ensure that the debris addition sequence did not affect the results of the test non-conservatively.

The licensee also noted that some paint chips settled during the test under conditions much more turbulent than would be present in the plant. The NRC staff evaluated the licensee methodology and determined that the paint chips had an opportunity to transport to the strainer and block strainer area and that even if the chips blocked the strainer area prior to fiber arriving that the head loss effect would be similar. The licensee also noted that the settled paint chips did not prevent less transportable debris from reaching the strainer. The NRC found that the addition of the paint chips prior to the addition of fine fiber did not have a significant effect on the test results.

The staff requested additional information on the method used to extrapolate the test results to higher fluid temperatures. The licensee stated that the testing included flow sweeps to change the velocity through the debris bed to assist in determining the appropriate correction to temperatures expected post-LOCA. The temperature correction was calculated based on a head loss correlation that accounts for both viscous (laminar) and kinetic (turbulent or density driven) changes in head loss caused by temperature changes. The correlation was developed for NUREG/CR-6224, Parametric Study of the Potential for BWR ECCS Strainer Blockage Due to LOCA Generated Debris (ADAMS Accession No. ML083290498) and has been accepted by the NRC staff for interpolation and extrapolation of test results, especially for temperature correction. The flow sweeps conducted during testing provided the behavior of the flow through the bed so that the correlation could be employed to extrapolate the test results to plant design temperatures. The methodology used by the licensee has been accepted by the staff.

Therefore, this issue was addressed appropriately.

The NRC staff requested information on piping that penetrates the Unit 1 strainer. It was stated that the loop seal is maintained full of water. The staff concern is that if the loop seal dries out

the strainer would be vented, and its design would have to consider alternate conditions. The licensee provided information that justified that the loop seal is normally filled because it branches off of the quench spray system which is maintained full of water to an elevation near the top of the loop seal. Once quench spray is initiated it provides flow to the loop seal. The quench spray system ensures that the loop seal is completely filled with adequate fluid to prevent breaching of the seal. This response was determined to be acceptable.

The NRC staff requested additional information regarding a statement by the licensee that the Unit 2 strainer was divided into two trains. The staff did not understand the purpose for this division. The licensee stated that the division of the strainer into two trains was carried over from the design of the previous strainer. The new strainer design is not subject to passive failure as the original strainer was so the division is unnecessary for the new design. The divider consists of perforated plates within the two sections of the strainer furthest from the pump suctions and a solid divider within the section of the strainer closest to the pump suction.

The licensee removed divider plate covers in the section nearest the pumps opening 8 holes, greater than 9 inches in diameter. These holes are too large to become blocked. This modification essentially makes the strainer a single train by allowing free flow between the previously potentially divided sections. The staff reviewed the licensee response and found that the strainer was modified to be essentially a single train.

The NRC staff questioned why the head loss testing included a flow sweep prior to the addition of all debris to one of the tests and also prior to the head loss becoming stable. The licensee stated that the flow sweep was completed after the full design basis debris load had been added to the test and the test stability criterion (1 percent change in head loss in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />) had been achieved. The flow sweep was conducted at this point in the test. After the flow sweep, an additional 10 percent of the chemical debris load was added to the test. The added chemical load was intended to demonstrate increased margin. The licensee also stated that the test under question no longer represents the design basis debris load for the strainer since a later test resulted in higher head loss. The NRC staff finds this response acceptable because the flow sweep was conducted after the design basis debris load was added and the test was superseded by a later test.

In its November 30, 2020 submittal the licensee discussed that the NaOH chemical addition systems were retired at both units and provided updated schematic diagrams for the quench spray systems.

The licensee also stated in its November 30, 2020 submittal that its head loss analysis for Unit 1 was revised to apply head loss from chemical precipitates only after the sump temperature decreased to below 150 °F to prevent flashing across the strainer and debris bed. The submittal further addressed the void fraction analyses for both units and stated that these were revised to account for the revised containment analyses. The maximum void fractions at the pump suctions were calculated to be 0.23 percent for Unit 1 and 0.26 percent for Unit 2. These values were confirmed to be less than the value assumed in the NPSH analyses (0.30 percent.)

In its November 30, 2020, submittal, the licensee identified an additional break for consideration in its Unit 2 analysis. The break is on the Unit 2 PRV line. The NRC staff identified that it appeared that the amount of debris that reached the strainer for this break was not bounded by any of the licensees head loss tests. The type of debris that was not bounded was Thermal

Wrap. The licensee responded to the question in its May 27, 2021 RAI response. The issue affects the head loss, debris generation, transport, and debris characteristics sections of this summary. The issue is discussed in detail in the debris generation section because the NRC staff identified several issues with the updated debris generation and transport evaluations associated with the break location. See the debris generation section for the NRC staff evaluation of this issue.

NRC STAFF CONCLUSION:

The licensee generally used staff approved guidance to address the area of head loss and vortexing. After reviewing the licensees supplemental response and RAI responses, the NRC staff concludes that the area had been addressed in a conservative manner. The NRC had several questions regarding the licensees treatment of the area. The licensee provided responses in the form of updating the evaluation in response to the questions or they provided information that justified the methodology that was used. Although some of the evaluation practices did not strictly follow NRC guidance, the staff was able to determine that the deviations from following the guidance would not significantly affect the evaluation results.

Therefore, the NRC staff concludes that the head loss and vortexing evaluation for Beaver Valley Units 1 and 2 is acceptable. The NRC staff considers this item closed for GL 2004-02.

9.0 NET POSITIVE SUCTION HEAD The objective of the NPSH section is to calculate the NPSH margin for the ECCS and CSS pumps that would exist during a LOCA considering a spectrum of break sizes.

NRC STAFF REVIEW:

The NRC staff review is based on documentation provided by the licensee through May 27, 2021. The NRC found that the licensee followed approved guidance and made generally conservative assumptions in the performance of the NPSH evaluation. Based on the initial licensee submittal, the NRC staff had several questions regarding the NPSH evaluation.

Several of these were answered in the updated supplemental response of June 30, 2009. The NPSH evaluations for each unit were different, so some questions apply only to a single unit.

For Unit 1, the staff requested clarification that the NPSH results bounded both the cold leg recirculation lineup and simultaneous cold and hot leg recirculation lineup lineups or that the licensee provide additional results for the most limiting condition. The staff also requested clarification on whether the NPSH margin results account for the strainer and debris bed head loss. Finally, the staff asked for the accident scenarios and times at which the minimum NPSH margins occur. In response the licensee stated that the flows through the low head safety injection (LHSI) pumps are limited by cavitating venturis which makes both cold leg and simultaneous hot and cold leg flow rates equal. The licensee provided tables with NPSH results for various cases, and plots that displayed the NPSH trend with time for the limiting NPSH cases. The information provided adequately answered these questions.

For Unit 2, the staff requested the final results for the NPSH required, NPSH available, and NPSH margin for the pumps taking suction from the containment recirculation sump following a LOCA. The staff also asked for confirmation that the results presented bound both the cold leg

recirculation lineup and simultaneous cold and hot leg recirculation lineup. The staff requested clarification as to whether the NPSH margin results account for the strainer and debris bed head loss. The staff also requested the accident scenarios and times at which the minimum NPSH margins occur. The licensee provided the final NPSH results for Unit 2 and stated that the system design results in lower sump flows for the hot leg recirculation mode than those for cold leg recirculation. This leads to greater NPSH margins for the hot leg recirculation case. The licensee provided a table with the NPSH results for various cases and a plot that displayed the trend with time for the limiting NPSH case. The staff found that the licensee adequately answered the questions except for one issue regarding the hot leg recirculation flow rates. This issue is discussed below.

The NRC asked for additional detail on the water sources that contribute to the post-LOCA sump inventory. The licensee provided more detailed information in its June 30, 2009 supplemental response concerning the water sources assumed in the minimum water level calculations, including the volumes of water assumed to contribute to sump inventory. The response did not include an adequate accounting of potential water hold up in the RCS. This issue is discussed below.

After reviewing the information provided in the June 30, 2009 supplemental response, additional information was requested from the licensee to ensure that the NPSH area had been adequately addressed. First, the staff requested that the licensee provide information regarding the volumes that could hold up water and prevent it from reaching the sump pool or displace water and result in increased sump level. The staff also questioned what overall effect these inputs could have on sump level. The licensee response stated that the Modular Accident.

Analysis Program-Design Basis Accident computer code is used to determine the amount of water hold up in the containment, and the amount of water released from the RCS. The model integrates information from the primary system and the containment. The response provided information regarding the model inputs and assumptions. The response also included, graphically, the water distribution in containment for a SBLOCA for each unit. In addition, the licensee provided tabular information showing minimum NPSH margins and associated strainer head losses for various plant operating conditions. The NRC staff found that the licensee provided adequate information to justify that holdups and equipment credited to displace water in the sump were adequately accounted for, and that the assumptions used in the calculation were appropriate.

The NRC staff requested that the licensee provide additional information regarding the potential for blockage of the refueling cavity drain. The licensee provided figures to depict the two units reactor cavities and other potential holdup mechanisms. The licensee provided a detailed description of the flow paths and predicted inventory accumulation within holdup volumes for each unit. The debris expected to reach the refueling and reactor cavities is described in the RAI response, and tables of the debris types and sizes were provided. The licensee stated that all debris predicted to transport to the refueling cavity was conservatively assumed to transport over the curb around the reactor seal ring. For Unit 1, the licensee stated that it was possible for debris to cause blockage at the openings around the neutron shield ring and evaluated the potential for debris to cause blockage. An evaluation showed that the potential head loss due to blockage was minimal and would not cause adverse inventory hold ups. The licensee also stated that an alternate flow path exists such that water could flow out of the reactor annulus penetrations and directly to the sump pool. This was stated to be advantageous because less

water would be held up in the lower reactor cavity. The NRC staff found the information provided by the licensee demonstrated that blockages in the refueling cavity would not result in the hold up of coolant flow to the Unit 1sump. The licensee also described the flowpath for water from the refueling cavity to the sump for Unit 2. For Unit 2, the licensee stated that the loop break does not result in adequate debris amounts to create significant blockage. The NRC staff found that the licensee demonstrated that debris would not inhibit the flow of coolant from the Unit 2 refueling cavity to the ECCS sump.

The NRC staff requested that the licensee provide additional information that justifies that the cold leg recirculation NPSH calculations are conservative with respect to all potential hot leg recirculation flow lineups for Unit 2, specifically for the case where the RSS pumps recirculate fluid through the LHSI system piping to the hot legs. The licensee stated that the case for hot leg recirculation from the RSS pumps via the LHSI piping was bounded by the cold leg recirculation alignment. The licensee provided information regarding the potential flow paths and stated that the LHSI pumps will be providing less flow in the hot leg recirc alignment. This results in lower NPSH requirements for the pumps. The licensee additionally noted that NPSH margin increases in the 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> period prior to hot leg switchover due to cooling of the fluid and increases in the sump pool level. The NRC found the information provided by the licensee acceptable to demonstrate that they had identified the bounding conditions for LHSI pump NPSH margin.

The NRC staff requested that the licensee provide additional information regarding the divider plate installed between two channels of the Unit 2 strainer. Additional information was also requested for Unit 1 if a divider plate was installed in that strainer. The licensee responded that the divider plate in Unit 2 was removed and Unit 1 does not contain a divider. The NRC staff determined the response was acceptable based on the removal of the divider plate on Unit 2 and the lack of divider plate on Unit 1.

The NRC staff requested that the licensee clarify the elevations of the breaks assumed for the SBLOCA cases, especially whether a break location that would allow complete refill of the RCS was bounded by the evaluation. The licensee stated that if the break occurred at the top of the pressurizer that inventory holdup that was not previously considered could occur. The licensee stated that the containment and NPSH analyses were revised to include a break at the top of the pressurizer and that it was found that this location was more limiting. The evaluation found that strainer submergence was decreased by 0.8 inches for both units and this new value was used as the limiting case for strainer submergence. The licensee stated that since the SBLOCAs are much less limiting for NPSH results that the small reduction in pump suction head would not impact the limiting NPSH cases. The staff found that the licensee actions were acceptable to address this issue because the licensee reevaluated the potential decrease in sump levels due to the break at a higher elevation and included the effects of such a break in the strainer submergence and the NPSH margin evaluations.

The NRC staff requested that the licensee provide additional information regarding parameters with respect to securing an RSS pump on Unit 2 upon the reduction of containment pressure to a predetermined value. This action was proposed to prevent the strainer structural limit from being exceeded at low sump water temperatures. The licensee provided the total strainer flow rates before and after the RSS pump is secured. The licensee stated that the operator is directed to shut down one of the RSS pumps when containment pressure reaches a value of

13.5 psi absolute. The shortest time to reach the structural pressure limit was estimated at about 18 days following the initiation of the event. This timing was based on updated strainer head loss testing that provided additional margin. The licensee stated that based on the timing and low temperature required to reach the limit that the operator action to secure the RSS pump was no longer considered to be a critical action. Based on the information provided by the licensee, especially the updated information based on new strainer testing, the staff agrees that the securing of the RSS pump is not a critical operator action and that adequate time is available for the action to be taken, if required at all.

In its November 30, 2020 submittal, the licensee stated it retired the Unit 1 chemical addition system, which removes the additional volume of water from the chemical addition tank and therefore this conservatism no longer applies. The licensee also stated that head loss from chemical precipitates is conservatively applied to the containment sump strainer head loss when the containment sump temperature is reduced below 150 °F for Unit 1 and 140 °F for Unit 2.

These sump temperatures are reached at a maximum of 8.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for Unit 1 and 16.7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> for Unit 2 following initiation of a LOCA.

Because of the new Unit 2 PRV break identified by the licensee in its November 30, 2020, submittal, the NRC staff requested that the licensee confirm that the limiting head loss values for the Unit 2 strainer provided in Table AI-4 of its September 28, 2010, submittal remain the limiting values. In its May 27, 2021 RAI response, the licensee stated that those values remain limiting.

NRC STAFF CONCLUSION:

The licensee generally used staff approved guidance to address the NPSH area. After reviewing the licensees supplemental response and RAI responses, the NRC staff concludes that the area had been addressed in a conservative manner. The NRC had several questions regarding the licensees treatment of the area. The licensee provided responses in the form of updating the evaluation in response to the questions or providing information that justified the methodology that was used. Therefore, the NRC staff concludes that the NPSH evaluation for Beaver Valley Units 1 and 2 is acceptable. The NRC staff considers this item closed for GL 2004-02.

10.0 COATINGS EVALUATION The objective of the coatings evaluation section is to determine the plant-specific ZOI and debris characteristics for coatings for use in determining the eventual contribution of coatings to overall head loss at the sump screen.

NRC STAFF REVIEW:

Although the review guidance calls for the dry film thicknesses, this information was not listed in the Beaver Valley response. Although missing, the staff does not consider this a significant gap in information since conservative quantities of coating debris were used in testing and analysis.

The break zone ZOI used for calculating the amount of qualified epoxy coating debris was 5D, which is acceptable per the staffs review of WCAP-16568-P. All qualified and unqualified

coatings in the ZOI fail as fine particulate and all debris generated by unqualified coatings in containment fail as fine particulate to maximize transport, which is acceptable by the NRC SE to NEI 04-07.

The licensee did not observe a thin bed during testing and treated all of the generated coatings debris as fine particulate. The licensee also added an additional 100 percent of the unqualified coatings debris as paint chips of a size equivalent or slightly larger than the area of the sump screen openings to maximize head loss for strainer testing. This is acceptable as justified by the NRC SE to NEI 04-07. Also, the amount of unqualified coating debris used in testing is doubled and the staff finds this to be conservative.

The surrogate material, ground silica, used for testing is acceptable to the staff. The type of paint chips used in head loss testing was not specified, but the staff does not find this to be a significant gap in information since a conservative amount of unqualified coatings debris was used in head loss testing.

The licensees coating assessment program met expectations.

NRC STAFF CONCLUSION:

For this review area, the licensee has provided information such that the NRC staff has reasonable assurance that the subject review area has been addressed conservatively or prototypically. Therefore, the NRC staff concludes that the coatings evaluation for Beaver Valley Units 1 & 2 is acceptable. The NRC staff considers this item closed for GL 2004-02.

11.0 DEBRIS SOURCE TERM The objective of the debris source term section is to identify any significant design and operational measures taken to control or reduce the plant debris source term to prevent potential adverse effects on the ECCS and CSS recirculation functions.

NRC STAFF REVIEW:

The NRC staff review is based on documentation provided by the licensee through November 30, 2020. The NRC found that the licensee followed approved guidance to perform the debris source term evaluation.

The licensee provided information regarding the controls that the units use to reduce the potential for materials in containment to become debris and affect the recirculation function.

The licensee stated that the containment buildings are maintained in a clean condition as required by technical specifications. This ensures that inspections are conducted to ensure that loose debris is not present prior to starting up from outages. The licensee also has programs in place to ensure that materials taken into containment are removed as appropriate following maintenance activities. The licensee has also implemented a containment cleaning program to further reduce latent debris in the containments. A plant label and tag program is in place to ensure that labels and tags meet post-LOCA requirements. The licensee has also implemented design controls to ensure that plant changes will not adversely affect strainer performance. The licensee also stated that controls of maintenance activities have been implemented to prevent

those activities from affecting strainer performance. These controls include housekeeping and foreign material exclusion programs. The licensee stated that the coatings program had been improved to reduce the amount of coatings that may become debris under LOCA conditions.

The licensee has also replaced problematic insulation types with less problematic materials and maintains RMI on a significant portion of piping that could be impacted by a LOCA jet.

In its November 30, 2020 submittal, the licensee listed insulation modifications since its September 28, 2010 RAI response. These modifications included replacing Temp-Mat fibrous insulation with RMI on reactor vessel inlet and outlet nozzles for Unit 1, removing additional portions of calcium silicate insulation from the RCS loop and pressurizer cubicles for Unit 1, replacing NUKON fibrous insulation blankets on the Unit 2 pressurizer PORV inlet piping with jacketed Thermal Wrap insulation (secured with Sure-Hold banding), replacing Min-K insulation on the Unit 2 steam generator level instrumentation tubing with jacketed Thermal Wrap insulation, and removing additional portions of calcium silicate insulation from the RCS loop and pressurizer cubicles of Unit 2.

The licensee also stated that the NaOH chemical addition systems at both Units were replaced with baskets of powdered NaTB mounted to the lower containment floor. Use of NaTB reduces the maximum pH of the recirculation pool and containment sprays when compared to NaOH, thereby reducing the corrosion rate of susceptible materials, such as metallic aluminum. The reductions in both pH and chemical precipitate loads were provided in tables in the submittal.

The licensee further stated that Unit 2 has six manway hatches and that 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 ZOI of the limited offset reactor vessel nozzle break.

NRC STAFF CONCLUSION:

The licensee used staff approved guidance to address the debris source term areas. In addition, the licensee has removed problematic materials from containment to reduce potential effects on sump strainers. After reviewing the licensees supplemental response, the NRC staff concludes that the area had been addressed in a conservative manner. Therefore, the NRC staff concludes that the debris source term evaluation for Beaver Valley Units 1 and 2 is acceptable. The NRC staff considers this item closed for GL 2004-02.

12.0 SCREEN MODIFICATION PACKAGE The objective of the screen modification package section is to provide a basic description of the sump screen modification.

NRC STAFF REVIEW:

The NRC staff review is based on documentation provided by the licensee through September 28, 2010. The NRC found that the licensee provided the requested information regarding the modification to the sump strainers. The license stated that a passive, safety-related strainer assembly was installed in Units 1 and 2 to replace the existing strainers. The new Unit 1 and 2

containment sump strainers have an effective flow area of 3,086 ft2 and 3,396 ft2, respectively.

The Unit 1 strainer is designed and fabricated by Components Control Incorporated (CCI). The Unit 2 strainer is designed by Enercon and fabricated by Transco. Also, the Unit 2 strainer is equipped with a bypass eliminator to reduce debris bypassing the strainer. The staff finds the description of the strainer modification acceptable.

NRC STAFF CONCLUSION:

The licensee provided the information requested for this area. Based on the review of all areas associated with Generic Letter 2004-02, the strainers are designed appropriately. Therefore, the NRC staff concludes that the screen modification package evaluation for Beaver Valley Units 1 and 2 is acceptable. The NRC staff considers this item closed for GL 2004-02.

13.0 SUMP STRUCTURAL ANALYSIS The objective of the sump structural analysis section is to verify the structural adequacy of the sump strainer including seismic loads and loads due to differential pressure, missiles, and jet forces.

NRC STAFF REVIEW:

FENOCs structural analysis results show that the components making up the replacement assemblies meet the code allowable stress requirements of the American Society of Mechanical Engineers Boiler & Pressure Vessel Code (B&PVC),Section III, Subsection NF. Regarding the possibility of dynamic loading effects on the Beaver Valley Power Station Unit 1 and Unit 2 replacement strainers, the licensee stated that the new sump strainer is located on the bottom floor of the containment and entirely outside of the crane wall and that high energy systems, such as feedwater, main steam, steam generator blowdown and reactor coolant piping are isolated from the sump by major structural features such as walls and floors. The licensee further stated that these structural features will act as barriers that will withstand loadings caused by missile impact, jet forces, and pipe whip impact forces.

NRC STAFF CONCLUSION:

The licensee used staff approved guidance and codes to address the strainer structural analysis areas. After reviewing the licensees supplemental response, the NRC staff concluded that the area had been addressed in a conservative manner. Therefore, the NRC staff concludes that the sump structural analysis for Beaver Valley Units 1 and 2 is acceptable. The NRC staff considers this item closed for GL 2004-02.

14.0 UPSTREAM EFFECTS The objective of the upstream effects assessment is to evaluate the flow paths upstream of the containment sump for holdup of inventory, which could reduce flow to the sump.

NRC STAFF REVIEW:

The NRC staff review is based on documentation provided by the licensee through September 28, 2010. The NRC found that the licensee followed approved guidance and made generally conservative assumptions in the performance of the upstream effects evaluation.

The licensee stated that an evaluation of flowpaths necessary to return water to the recirculation sump strainer was performed. This evaluation was performed to identify obstructions or volumes that could result in the holdup of water not previously considered. The licensee stated that with the exception of water held up in the fuel transfer canal, all other water returns to the sump. The licensee stated that flowpaths have sufficiently large openings to prevent the holdup of significant quantities of water that could affect the containment sump water level analysis.

During the analysis, one potential holdup point was identified. The new drain for the reactor cavity was designed with a cruciform personnel exclusion device. Due to the location of this device and the turbulence in the vicinity of the drain hole, it is possible that large pieces of debris could be transported and trapped by the exclusion device. The design has been enhanced such that the device was removed from Unit 1 during the fall 2007 refueling outage and was removed from Unit 2 during the spring 2008 refueling outage.

The major hold up volumes for both Units is the refueling canal. As discussed in the NPSH section, the NRC determined that the sump levels were determined appropriately, including these holdups.

NRC STAFF CONCLUSIONS:

The licensee used staff approved guidance and codes to perform the upstream effects evaluation. After reviewing the licensees supplemental response, the NRC staff concludes that the area had been addressed in a conservative manner. Therefore, the NRC staff concludes that the upstream analysis for Beaver Valley Units 1 and 2 is acceptable. The NRC staff considers this item closed for GL 2004-02.

15.0 DOWNSTREAM EFFECTS - COMPONENTS AND SYSTEMS The objective of the downstream effects, components and systems section is to evaluate the effects of debris carried downstream of the containment sump screen on the function of the ECCS and CSS in terms of potential wear of components and blockage of flow streams.

NRC STAFF REVIEW:

The NRC staff review is based on documentation provided by the licensee through September 28, 2010. The NRC found that the licensee followed approved guidance for the performance of the downstream effects for components and systems evaluation.

The licensee stated that it has performed the ex-vessel downstream effects evaluations for Beaver Valley Units 1 and 2 following the guidance in WCAP-16406-P, Revision 1, and the NRC SE of that document. The licensee also stated that as a result of the evaluation, the HPSI cold leg throttle valves were replaced at Beaver Valley Unit 1. At Unit 2, the HPSI throttle valves

were modified to ensure adequate function. The licensee determined that the evaluation results indicate that no unacceptable component wear of the ECCS and RSS flow paths will occur, and therefore inadequate core or containment cooling will not result due to the effects of the debris.

Because the licensee performed ex-vessel downstream effects calculations and analyses in accordance with the NRC recognized methods prescribed in WCAP-16406-P, Revision 1 and the associated NRC SE, including limitations and conditions, the staff concludes that the downstream effects of debris laden recirculated sump fluid on ex-vessel downstream components and systems has been adequately addressed at Beaver Valley Units 1 and 2.

NRC STAFF CONCLUSION:

Because the licensee performed ex-vessel downstream effects calculations and analyses in accordance with the NRC recognized methods prescribed in WCAP-16406-P, Revision 1 and the associated NRC SE, including limitations and conditions, the staff concludes that the downstream effects of debris laden recirculated sump fluid on ex-vessel downstream components and systems has been adequately addressed at Beaver Valley Units 1 and 2. The NRC staff considers this item closed for GL 2004-02.

16.0 DOWNSTREAM EFFECTS - FUEL AND VESSEL 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 LTCC.

NRC STAFF REVIEW:

The licensee provided its in-vessel evaluation in its November 30, 2020 supplemental response.

In the supplement, the licensee discussed strainer penetration testing that was performed to quantify the amount of fibrous debris that could reach the reactor vessel after a LOCA. The licensee also provided information regarding the assurance of LTCC for both Units 1 and 2.

The NRC staff identified potential issues with the licensees strainer penetration testing methodology and requested additional information in its March 23, 2021 RAI. The licensee provided its RAI response on May 27, 2021. The licensee stated that the strainer penetration testing was completed separately in 2008 for each unit. The testing used strainer modules identical to those installed in the plant. The fiber that penetrated the strainer was captured in filter bags. The licensee provided a description of the test flume, the debris used in the penetration testing, the debris introduction practices, and the transport efficiency during testing.

For Unit 1, the licensee provided the flow rates used during testing and stated that because of high differential pressure across the filter bags, the design flow rate could not be maintained during the testing. The licensee stated that the reduced flow rate had minimal effect because a higher flow rate increases the filtering efficiency on the strainer due to bed compression, the design flow rate was increased by 30 percent, and that the full series of tests included a period of testing at the representative maximum flow rate during which samples were taken and no fibers identified in the water. The NRC staff requested that the licensee provide additional information regarding these statements partially because they are contrary to other test observations and also because the description of the flow rates used was difficult to understand.

The licensee provided additional details regarding the testing and also provided information that

corrected the penetration amounts for flow based on empirical data from penetration testing performed at various flow rates. The licensee stated that the bounding break for fiber penetration for Unit 1 is the pressurizer safety valve inlet piping break and that the test flow rate was bounding of this break. For the LBLOCA, double-ended hot-leg break case, the plant flow rate is higher than the tested value by 11.3 percent for the long-term test flow rate. The NRC staff noted that the difference between the flow rate at which the debris was added and the plant flow rate is significantly greater, about 45.8 percent. To account for the most conservative plant flow rate, the staff calculated that the test rate should be increased by 61 percent. To account for the difference in the flow values used by the licensee, the staff performed calculations to ensure that the licensees underestimation of flow rate did not significantly affect the overall conclusions regarding fiber penetration for Unit 1. The staff used empirical data as a basis to estimate the increase in penetration that would occur if the flow rate during the test was increased by 61 percent. The staff calculation predicts that the maximum amount of fiber that could reach the Unit 1 reactor vessel is 21.5 grams/fuel assembly (g/FA). See the discussion of acceptable in-vessel fiber amounts below.

For Unit 2, the licensee stated that the flow rate for the majority of the test was maintained at the design flow rate except during debris additions when the flow was reduced to about half of the design flow rate. The licensee stated that this methodology provided acceptable results. The NRC staff position is that the flow rate should be maintained at the design flow rate during all phases of testing but considered that this deviation would not affect the results enough to cause a failure of the acceptance criteria for in-vessel fiber amount. The NRC staff performed a simplified calculation based on the Unit 1 calculation described above and estimated that the amount of fiber arriving in the core for Unit 2 would not exceed 12.3 g/FA. See the discussion regarding margins below.

For Unit 2, the licensee included some small pieces of fibrous debris. This is not typical for penetration testing. The licensee stated that the test debris ratio of fine to small sizes was greater than the ratio that would occur in the plant and was therefore conservative. The staff concluded that using more fine fiber than was expected to transport to the strainer in the plant is conservative. The licensee stated that all fine fiber was added to the test and provided an opportunity to penetrate before the small pieces were added. The staff concluded that this debris addition sequence is acceptable because fine fiber is more likely to reach the strainer and more likely to arrive earlier than small pieces. This conclusion is based on realistic physical behavior of fiber and the fact that this addition sequence provides an opportunity for the fine fiber to penetrate the strainer without being filtered by the small pieces which are less likely to penetrate. The licensee also corrected the penetration results to account for the inclusion of small pieces by subtracting the amount of small pieces from the total upstream fiber source term when calculating the penetration percentage. The NRC staff concluded that this correction was appropriate because it excludes small pieces, which are unlikely to penetrate, from the penetration calculation. Based on the above, the NRC staff concluded that the penetration test methodology was acceptable with the exception of the flow rate. See the discussion of flow rate in the paragraph above.

The licensee reported the amounts of fiber predicted to reach the core based on the testing.

For Unit 1, the in-vessel fiber amounts were calculated to be 4.8 g/FA for the large hot-leg break and 13.4 g/FA for the pressurizer safety valve inlet line break, which is the limiting case. The Unit 2 limiting in-vessel fiber amount was calculated to be 7.2 g/FA for a large hot-leg break.

Both units have low fiber debris source terms and maintain significant margin to the established fiber acceptance values for in-vessel debris. The discrepancies identified in testing were accounted for in the licensees response to staff RAIs as described above. In addition, the NRC staff performed calculations that ensure that flow rate discrepancies in the testing would not result in the in-vessel fiber acceptance limits being exceeded with significant margin. Even though the testing and analysis was not performed fully per accepted practices, the NRC staff concluded that the licensee demonstrated adequate margin for in-vessel fiber amounts such that there is reasonable assurance that the testing demonstrates that debris in the reactor vessel will not inhibit LTCC. In addition, the plant maintains margins to other key parameters that provide additional assurance that LTCC will be maintained.

With respect to reactor vessel related chemical effects, the licensee identified the Volume 5 WCAP-17788-P test groups that represented the post-LOCA environments for each unit.

Beaver Valley Unit 1 was represented by WCAP-17788-P Test Group 27. The NRC staff verified the test group parameters reasonably represent Beaver Valley Unit 1. The licensees November 30, 2020 letter stated that longer drain times observed for Test Group 27 taken at 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or earlier were attributed to particulate material rather than chemical products, and that chemical precipitation occurs at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> as noted in WCAP-17788-P. The NRC staff agrees that chemical precipitates were detected at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in Test Group 27. The NRC staff assessed the Test Group 27 out of bag filtering data, which is hard to interpret due to the effects from various particulate (e.g., calcium silicate or microporous insulation) on the test filter. Based on its independent review of the data, the NRC staff cannot rule out that some precipitates may have been present at 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> in some tests (IBOB 27-01 and OB27-02), since these tests had drain times 6 times and 2.5 times longer than their respective 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> sample drain times. The staff notes that the PWROG conclusion that no precipitation occurred at 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> is based on additional SEM analysis of the filter cake and ICP analysis (to quantify aluminum on the filter). For the staffs in-vessel chemical effects evaluation, however, the staff conservatively assumed precipitates formed at 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

Even assuming precipitation occurred at 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, the Beaver Valley Unit 1 chemical effects analysis is acceptable from an in-vessel perspective for the following reasons. The Beaver Valley Unit 1 representative autoclave Test Group 27 used approximately 2.4 times the scaled aluminum surface area than in the plant, which means precipitation in the WCAP-17788-P autoclave tests would occur much sooner than in a plant post-LOCA environment. This adds additional conservatism to the NRC assumption on precipitation timing. From the in-vessel perspective, Beaver Valley Unit 1 identified that both Tblock (less than 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />) and hot leg recirculation time (6.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> after LOCA) are less than the minimum time to chemical effects. Therefore, even if chemical precipitates formed at the earliest time assumed by NRC staff, the precipitates would not impact long term core cooling since alternate flow paths would be available before that time.

The licensee identified WCAP-17788-P Test Group 17 as representative for Beaver Valley Unit

2. The NRC staff verified the test group parameters reasonably represented Beaver Valley Unit

2. The licensee stated that the minimum time to chemical precipitation for Beaver Valley Unit 2 is 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and the staff agrees that the WCAP-17788-P data supports the licensees assessment. The licensee also identified that both Tblock (less than 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />) and the time until hot leg circulation mode (approximately 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />) occur well before the time for chemical

precipitate formation. Therefore, the staff concludes that even if chemical precipitates form following a LOCA, they would not impact long term core cooling since alternate flow paths would be available before that time.

The licensee evaluated the key parameters as recommended by NRC staff guidance and WCAP-17788-P. The submittal stated that the key parameters for both units including in-vessel fiber load, sump switchover time, chemical precipitate time, hot-leg switchover time, rated thermal power, alternate flowpath (AFP) resistance, and ECCS flow are bounded or within the limits suggested by the guidance. The only exception noted is that the Unit 2 AFP resistance is greater than that analyzed in WCAP-17788-P. For this parameter, the licensee stated that when scaled for a lower plant thermal power the AFP resistance is acceptable for the analysis.

This issue is discussed below. Other issues identified by the staff, but found to be acceptable, are discussed above in this section.

For Beaver Valley Unit 2, the plant is a 3-loop Westinghouse plant with a standard upflow barrel/baffle configuration. The unadjusted barrel/baffle form loss coefficient is higher for Beaver Valley Unit 2 than the WCOBRA/TRAC upflow plant that was modeled in WCAP-17788-P. However, the upflow plant was a four-loop Westinghouse model with a much higher thermal power rating than Beaver Valley Unit 2 (3,658 megawatts thermal analyzed vs.

2,900 megawatts thermal rated). In its response to WCAP-17788-P, Volume 4, RAI 4.2, Westinghouse indicated that the AFP resistance can be scaled by power level, because A plant with a higher power will require more flow through the AFP to adequately remove decay heat compared to a plant with lower power. Therefore, the licensees statement that the Beaver Valley Unit 2 AFP resistance, when adjusted for rated thermal power, is less than the analyzed AFP resistance, is acceptable.

NRC STAFF CONCLUSION:

The licensees supplement demonstrates that both Beaver Valley units are well represented by the analyses in WCAP-17788-P, Volume 4, and as such, the NRC staff concludes that the licensee has demonstrated the capability to provide adequate LTCC in the unlikely event the entire core inlet becomes blocked with debris.

For the IVDEs review area, the licensee has provided sufficient information such that the NRC staff has reasonable assurance that the subject review area has been addressed conservatively or prototypically. Therefore, the NRC staff concludes that the licensees evaluation of this area is acceptable. Based on the information provided by the licensee, the NRC staff considers this area closed for GL 2004-02.

17.0 CHEMICAL EFFECTS:

The objective of the chemical effects section is to evaluate the effect that chemical precipitates have on strainer head loss and core cooling. Chemical effects impact to the reactor vessel is evaluated separately in Section 16.0, Downstream Effects - Fuel and Vessel.

NRC STAFF REVIEW:

The NRC staff chemical effects review is based on documentation provided by the licensee as detailed in the table in Section 1.0 Introduction above. The reference documents used for this review include NRC Staff Review Guidance Regarding Generic Letter 2004-02 Closure in the Area of Plant Specific Chemical Effects Evaluations, dated March 2008 (ADAMS Accession No. ML080380214) and Argonne National Laboratory (ANL) Technical Letter Report on Evaluation of WCAP aluminum hydroxide surrogate stability at elevated pH (ADAMS Accession No. ML090480294).

The licensees November 30, 2020 supplemental response indicates the NaOH spray system for post-LOCA pool pH control in Unit 1 was retired. Therefore, both Beaver Valley Units 1 and 2 use NaTB powder dissolution to adjust the post-LOCA pool pH. The switch to NaTB from NaOH results in lower post-LOCA pH and less aluminum based chemical precipitates. The licensee replaced their pre-existing sump strainers with a passive, safety-related strainer assembly for both Units 1 and 2. The Unit 1 original sump strainers were replaced with CCI pocket strainers. The Unit 2 sump strainer was replaced with Enercon top hat strainer design.

The new Units 1 and 2 containment sump strainers have an effective flow area of 3,086 ft2 and 3,396 ft2, respectively.

The licensees plant-specific debris generation and transport analyses determined that the debris sources for Unit 1 included aluminum, calcium silicate, fiberglass insulation (latent fiber),

Temp-Mat, Min-K and concrete. The debris sources for Unit 2 included aluminum (submerged/

not submerged), calcium silicate, fiberglass insulation (latent fiber), Temp-Mat, thermal wrap, Min-K, Microtherm and concrete.

RAI REVIEW There were some RAIs regarding the chemical effects area throughout this review. The licensee responded to GL 2004-02 in letters dated March 4, July 22, and September 6, 2005 (ADAMS Accession Nos. ML050680211, ML052080167, and ML052510411). The licensee submitted a second response letter on September 5, 2005 (ADAMS Accession No. ML052510411). At that time, the licensee had no specific head loss term established for chemical effects (due to the lack of test information).

The NRC staff had concerns with the licensees overall strategy to evaluate potential chemical effects and therefore sent an RAI on February 9, 2006 (ADAMS Accession No. ML060380342).

The licensee response on December 20, 2007 (ADAMS Accession No. ML073620201) stated that chemical effects testing for Units 1 and 2 were completed by Alion Science and Technology (Alion) in November 2007; however, the evaluation of the results will not be completed until the end of 2007. The licensee was planning to fully assess the chemical effects testing that had been performed to date, formalize the results, and develop an action plan by February 29, 2008.

Therefore, the licensee requested an extension to fully assess the chemical effects testing and downstream effects analyses and develop required corrective actions by February 29, 2008.

The NRC approved the extension request on December 27, 2007 (ADAMS Accession No. ML073600373).

The licensee provided a revised supplemental response to GL 2004-02 on February 29, 2008 (ADAMS Accession No. ML080660597). The response provided results from integrated chemical effects tests done by Alion at the VUEZ facility in Slovakia. Based on NRC staff observations of testing at VUEZ (ADAMS Accession No. ML073450430) and subsequent

discussions, the NRC staff found that the approach was not technically justifiable (ADAMS Accession No. ML080870246). Therefore, the licensee reconstructed the testing.

The licensee superseded its February 29, 2008 response with an October 29, 2008 (ADAMS Accession No. ML083080094) supplemental response. The licensees supplement provided the completed chemical precipitate tank test for Unit 1 and a corrective action plan for Unit 2 (to retest following the same approach as Unit 1). At the time of this supplemental response letter, Unit 2 testing was scheduled for fall 2008. Therefore, portions of the Unit 2 responses were not complete.

The licensee provided their Integrated Chemical Effects testing results and their evaluation for Unit 2 on June 30, 2009 (ADAMS Accession No. ML091830390). The June 2009 supplement also stated that additional Temp-Mat insulation was identified on the reactor vessel inlet and outlet nozzles for Unit 1. The licensee received an extension from the NRC to take corrective actions following the startup from the fall 2010 refueling outage (ADAMS Accession Nos.

ML091250180 and ML091240030). At the time, the NRC staff had no additional questions related to chemical effects. However, there was an open item for the plant to address additional insulation that was not bounded by their analysis at that time. Therefore, on February 18, 2010 (ADAMS Accession No. ML100290318), the NRC staff requested the licensee to confirm that the chemical effects analysis for Unit 1 was not affected by the additional material discovered in containment.

The licensee responded on September 28, 2010 (ADAMS Accession No. ML102770023). The NRC staff evaluated the response and found that the previous chemical effects analysis provided by the licensee was valid, since the additional insulation material discovered during refueling outage 1R19 for Unit 1 (ADAMS Accession No. ML091250180) was subsequently removed and replaced with RMI during the fall 2010 refueling outage (ADAMS Accession No. ML121140300).

Overall Strainer Chemical Effects Methodology The licensees overall strainer chemical effects approach relied on head loss testing that added pre-mixed chemical precipitates to the test loop after a fiber and particulate debris bed had formed and stabilized on the test strainer. The approach was based on the WCAP-16530-NP-A precipitate methodology while also crediting delayed precipitation based on aluminum solubility. No deviations were taken from the WCAP base model spreadsheet for calculating quantity of chemical precipitate. In addition, the quantity of precipitates calculated with WCAP-16530-NP-A for Units 1 and 2 assumed containment sprays would be operated for the full 30 day post-LOCA mission time.

In its November 30, 2020 supplemental response, the licensee credited delayed chemical precipitation until the sump temperatures decreased significantly. The Beaver Valley Unit 1 GL 2004-02 analysis assumes precipitation occurs at approximately 8.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> post-LOCA or 150 °F. The Beaver Valley Unit 2 GL 2004-02 analysis assumes precipitation occurs at approximately 16.7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> post-LOCA or 140 °F.

Testing:

The retest for Units 1 and 2 was performed in accordance with the March 2008 NRC staff Review Guidance regarding Generic Letter 2004-02 Closure in the Area of Plant-Specific Chemical Effects Evaluations (ADAMS Accession No. ML080380214).

The retesting for Unit 1 was completed in the spring of 2008 by Alion at their facility in Warrenville, IL. A total of nine tests were performed for the strainer array. The specific debris mixture used included fibrous insulation debris, particulate debris, and chemical precipitates.

Debris and chemical precipitates generated during a postulated LOCA were considered to arrive at the strainer at the time of recirculation initiation, for establishing minimum NPSH margins.

The generated debris was considered to fully transport to the strainer for head loss testing.

There was no credit for near field settling; all debris loads were added over a sparger in the tank. The chemical precipitates were premixed in another tank and then added to the test tank in batches to provide chemical loads corresponding to the precipitate generation cases identified in the test plan. Test 6 was determined to be the target test representing planned insulation modifications. This test (Test 6) was performed for the purpose of determining the head loss of the final insulation configuration. Additional quantities of sodium aluminum silicate and aluminum oxyhydroxide equivalent to an additional 10 percent each were also added to allow for increased margin in head loss testing. The test case bounded the debris and chemical quantities for all breaks including loop, reactor nozzle, and surge line breaks.

Testing for Unit 2 was completed in the fall of 2008 by Alion at their Warrenville, IL facility.

Testing performed was based on the configuration after planned insulation modifications. A total of seven tests were performed for the strainer array. The specific debris mixture used included fibrous insulation debris, particulate debris, and WCAP-16530-NP-A predicted chemical precipitates. The chemical precipitates were premixed in another tank and then added to the test tank in batches to provide chemical loads corresponding to the precipitate generation cases identified in the test plan. Near-field settlement was not credited.

The final configuration for Unit 2 was represented by two tests, Test 1A and Test 5. Test 1A represented the loads from loop break and surge line break. Test 5 represented the head loss for the reactor vessel nozzle break and bounds the maximum allowable Microtherm debris load. Tests also included additional quantities of sodium aluminum silicate and aluminum oxyhydroxide equivalent to an additional 10 percent of each to allow for increased margin in head loss testing. The resultant head loss values indicated that a partial thin bed was potentially formed, resulting from the additional contribution of the particulates and chemical precipitates. However, these head loss values were low enough to accommodate this potential partial thin bed formation. In its November 30, 2020 submittal, the licensee identified an additional break that needed to be considered for Unit 2. This was the PRV line break. For additional details regarding this break and NRC concerns with the licensees original evaluation of its debris, see the debris generation section of this summary. Tests bounded the debris and chemical quantities for all breaks including the loop, reactor vessel nozzle, and surge line break loads.

In addition, the licensee included additional information (September 28, 2010 ADAMS Accession No. ML102770023) used to determine the chemical effects loading for both units.

The licensee used the ANL aluminum solubility correlation approach from Technical Letter Report on the Evaluation of WCAP aluminum hydroxide surrogate stability at elevated pH to credit delay in the onset of chemical effects (ADAMS Accession No. ML090480294) for both units. The licensee explained the reasoning for delaying the onset of chemical effects for both

units was to provide a more realistic application of the strainer head loss test results. In addition, the licensee noted that the basis for the delay in the onset of chemical effects for both units was temperature. References to the post-LOCA timing were made to simplify the post-accident timeline of events, and based on the plant specific maximum temperature, minimum cooling scenario for each unit.

For Unit 1, the licensee used the aluminum solubility limit per the ANL aluminum solubility correlation to show that aluminum will remain in solution at temperatures above or equal to 150 °F with a minimum sump pH of 7.8. The licensee explained that the application of head loss due to chemical effects can be delayed for up to 8.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for the maximum temperature case after the initiation of a LOCA when the sump temperature at Unit 1 has fallen to 150 °F.

The licensee also explained that for other breaks that resulted in more rapid cooling, the use of 150 °F for the application of chemical effects head loss would still be conservative since other breaks would result in the sump temperature falling to 150 °F in less than 8.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. This would result in less dissolved aluminum and silicon in comparison to the maximum temperature case.

For Unit 2, the licensee used the aluminum solubility limit per the ANL aluminum solubility correlation to show that aluminum will remain in solution at temperatures above or equal to 140 °F at a minimum sump pH of 7. The licensee also explained that the application of head loss due to chemical effects can be delayed up to 16.7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> after the initiation of a LOCA (for the maximum temperature profile case) when the sump temperature at Unit 2 has fallen to 140 °F. The licensee explained that for other breaks that resulted in more rapid cooling, the use of 140 °F for the application of chemical effects head loss would still be conservative. This is valid because other breaks would result in the sump temperature falling to 140 °F in less than 16.7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />, which will result in less dissolved aluminum and silicon in comparison to the maximum temperature case.

NRC staff concludes that the licensees supplemental chemical effects information related to delayed precipitate formation is acceptable because the licensee:

• Used WCAP-16530-NP-A methodology for both units chemical product generation predictions with no refinements to the base model. The NRC staff has previously reviewed and accepted this methodology as documented in the SE for WCAP-16530-NP-A.

• Applied a 2X multiplier to the initial WCAP-16530-NP-A aluminum release rate per the NRC SE guidance.

• Used temperature profile assumptions that maximized the aluminum release when calculating how much aluminum could be present in solution.

• Showed that the ANL equation would not predict precipitation at the temperatures when precipitation was assumed when using the minimum possible sump pH values to minimize solubility. For the Beavery Valley Units 1 and 2 post-LOCA pH values, the ANL equation was shown to be conservative by autoclave testing.

NRC STAFF CONCLUSION:

For the chemical effects area, the licensee has provided sufficient information such that the NRC staff has reasonable assurance that chemical effects have been addressed conservatively or prototypically for Beaver Valley, Units 1 and 2. Therefore, the NRC staff concludes that the

chemical effects evaluation for Beaver Valley, Units 1 and 2 is acceptable. The NRC staff considers this area closed for GL 2004-02.

18.0 LICENSING BASIS The objective of the licensing basis section is to provide information regarding any changes to the plant licensing basis due to the changes associated with GL 2004-02.

To achieve sufficient water level to cover the containment sump strainers following a containment pressurization event, the start signal for the RSS pumps was changed. License Amendment Nos. 280 (dated October 5, 2007) and 164 (dated March 11, 2008), that change Technical Specifications to reflect the new RSS pump start signal, have been implemented for Units 1 and 2 respectively.

The Unit 1 licensing basis presently credits containment overpressure to meet NPSH requirements. This credit assists in maintaining NPSH margins. License Amendment No. 167 for Unit 2, issued March 26, 2009, authorized changes to the licensing basis as described in the Unit 2 updated final safety analysis report (UFSAR) regarding the method of calculating the net positive suction head available to the RSS pumps by crediting containment over pressure. The Unit 2 licensing basis was revised April 7, 2009 and also credits containment overpressure to meet RSS pump NPSH requirements.

The licensee replaced the NaOH additive system with a passive NaTB system. While not required to meet the head loss margins described in this response, this change will result in a lower sump water pH and reduce the quantity of chemical precipitates. The NRC issued License Amendment No. 168 by letter dated April 16, 2009 to change the pH buffer from NaOH to NaTB prior to achieving Mode 4 during startup from the Unit 2 fall 2009 refueling outage.

The Units 1 and 2 UFSARs have been updated to reflect installation of the new sump strainers and the new RSS pump start signal. Other facility and procedure changes made to support compliance with GL 2004-02 requirements will be incorporated into the Units 1 or 2 UFSAR, as appropriate, in accordance with 10 CFR 50.71(e). The licensee intends to develop and issue other licensing bases changes following NRC review and approval of the response to GL 2004-02.

The licensee committed to change the FSAR in accordance with 10 CFR 50.71(e) to reflect the changes to the plant in support of the resolution to GL 2004-02. In addition, the licensee stated that changes would be made to the FSAR describing the new licensing basis to reflect the revised debris loading as it affects ECCS sump strainer performance and in-vessel effects, including the following:

• Break Selection

• Debris Generation

• Latent Debris

• Debris Transport

• Head Loss

• Additional Design Considerations

NRC STAFF CONCLUSION:

For this review area the licensee has provided information, such that the NRC staff has reasonable assurance that the subject review area has been addressed conservatively or prototypically. Based on the licensees commitment, the NRC has confidence that the licensee will affect the appropriate changes to the Beaver Valley FSAR, in accordance with 10 CFR 50.71(e), that will reflect the changes to the licensing basis as a result of corrective actions made to address GL 2004-02. Therefore, the NRC considers this item closed for GL 2004-02.

19.0 CONCLUSION

The NRC staff has performed a thorough review of the licensees responses and RAI supplements to GL 2004-02. The NRC staff conclusions are documented above. Based on the above evaluations the NRC staff finds the licensee has provided adequate information as requested by GL 2004-02.

The stated purpose of GL 2004-02 was focused on demonstrating compliance with 10 CFR 50.46. Specifically, the GL requested addressees to perform an evaluation of the ECCS and CSS recirculation and, if necessary, take additional action to ensure system function in light of the potential for debris to adversely affect LTCC. The NRC staff finds the information provided by the licensee demonstrates that debris will not inhibit the ECCS or CSS performance following a postulated LOCA. Therefore, the ability of the systems to perform their safety functions, to assure adequate LTCC following a design-basis accident, as required by 10 CFR 50.46, has been demonstrated.

Therefore, the NRC staff finds the licensees responses to GL 2004-02 are adequate and considers GL 2004-02 closed for Beaver Valley.