ML21260A161

Closeout of Generic Letter 2004 02, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors

ML21260A161
Person / Time
Site:	Farley
Issue date:	09/24/2021
From:	Stephanie Devlin-Gill Plant Licensing Branch II
To:	Gayheart C Southern Nuclear Operating Co
Devlin-Gill S, NRR/DORL/LPL2-1, 415-5301
References
EPID L-2017-LRC-0000
Download: ML21260A161 (33)

Text

September 24, 2021 Ms. Cheryl A. Gayheart Regulatory Affairs Director Southern Nuclear Operating Co., Inc.

3535 Colonnade Parkway Birmingham, AL 35243

SUBJECT:

JOSEPH M. FARLEY NUCLEAR PLANT, UNITS 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)

Dear Ms. Gayheart:

The U.S. Nuclear Regulatory Commission (NRC) issued Generic Letter (GL) 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors (Agencywide Documents Access and Management System (ADAMS) Accession No. ML042360586), dated September 13, 2004, requesting that licensees address the issues raised by Generic Safety Issue (GSI)-191, Assessment of Debris Accumulation on PWR [Pressurized Water Reactor] Sump Performance.

By letter dated May 16, 2013 (ADAMS Accession No. ML13137A131), Southern Nuclear Operating Company, Inc. (the licensee) stated that they will pursue Option 2 (deterministic) for the closure of GSI 191 and GL 2004 02 for Joseph M. Farley Nuclear Plant, Units 1 and 2 (Farley).

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 GSI-191, the NRC staff issued a technical evaluation report on in-vessel downstream effects (ADAMS Accession Nos. ML19178A252 and ML19073A044 (not publicly available, proprietary information)).

Following the closure of GSI-191, the NRC staff also issued the review guidance for in-vessel downstream effects, NRC Staff Review Guidance for In-Vessel Downstream Effects Supporting Review of Generic Letter 2004-02 Responses (ADAMS Accession No. ML19228A011), to support review of the GL 2004-02 responses.

The NRC staff has reviewed the licensees responses and supplements associated with GL 2004-02. Based on the 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 Title 10 of the Code of Federal Regulations (10 CFR) Section 50.46. Specifically, GL 2004-02 requested addressees to perform an evaluation of the emergency core cooling system (ECCS) and containment spray system (CSS) recirculation and, if necessary, take additional action to ensure system function considering the potential for debris to adversely affect long-term core cooling.

C. Gayheart 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 long-term core cooling following a design-basis accident, as required by 10 CFR 50.46, has been demonstrated.

Based on its review, the NRC staff finds the licensees responses to GL 2004-02 are adequate and considers GL 2004-02 closed for Farley.

Enclosed is the summary of the NRC staff's review. If you have any questions, please contact me at 301-415-5301 or via e-mail at Stephanie.Devlin-Gill@nrc.gov.

Sincerely,

/RA/

Stephanie Devlin-Gill, Project Manager Plant Licensing Branch II-I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50 348 and 50 364

Enclosure:

NRC Staff Review of GL 2004-02 for Farley cc: Listserv

U.S. NUCLEAR REGULATORY COMMISSION STAFF REVIEW OF THE DOCUMENTATION PROVIDED BY SOUTHERN NUCLEAR OPERATING COMPANY, INC.

FOR JOSEPH M. FARLEY NUCLEAR PLANT, UNITS 1 AND 2 DOCKET NOS. 50 348 AND 50 364 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 every 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 core cooling (LTCC) following a LOCA is a basic safety function for nuclear reactors. The recirculation sump provides a water source to the ECCS in a pressurized-water reactor (PWR) 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 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 Enclosure

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.

The 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 deterministic rule requires maintaining 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 staff 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.

The 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. The request was based on the identified potential susceptibility of PWR recirculation sump screens to debris blockage during design-basis accidents (DBAs) 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. 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. ML041550661), 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 staff 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 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):

x 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).

x 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).

x 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:

x corrective actions for GL 2004-02, x break selection, x debris generation/zone of influence (ZOI) (excluding coatings),

x debris characteristics, x latent debris, x debris transport, x head loss and vortexing, x NPSH, x coatings evaluation, x debris source term, x screen modification package, x sump structural analysis, x upstream effects, x downstream effects - components and systems, x downstream effects - fuel and vessel, x chemical effects, and x 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:

x 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 (g/FA).

x 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 of risk informed analysis to demonstrate adequate long term core cooling, e.g., South Texas Project (ADAMS Package Accession No. ML17019A001).

x 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 (SRM-SECY-0093) 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. ML13137A131), Southern Nuclear Operating Company, Inc. (the licensee) stated that they will pursue Option 2 (deterministic) for the closure of GSI-191 and GL 2004-02 for Joseph M. Farley Nuclear Plant, Units 1 and 2 (Farley).

On July 23, 2019 (ADAMS Package Accession No. ML19203A303), GI-191 was closed. It was determined that the technical issues identified in GI-191 were now well understood and therefore GI-191 could be closed. Prior to and in support of closing the GI, NRC 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 GI, NRC 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).

Table 1 of this document lists the documentation provided by the licensee in response to GL 2004-02:

Table 1: Responses to GL 2004-02 DOCUMENT DATE ACCESSION NUMBER DOCUMENT February 25, 2005 ML050610168 Initial Response to GL August 31, 2005 ML052430746 Supplemental Information February 9, 2006 ML060370387 1st NRC RAI February 28, 2008 ML080660654 Supplemental Information April 29, 2008 ML081210452 Supplemental Information December 17, 2008 ML083530543 Supplemental Information March 9, 2009 ML090620117 2nd NRC RAI July 27, 2009 ML092380647 Licensee Response to RAI March 30, 2010 ML100900004 Supplemental Information May 5, 2010 ML101230051 NRC Partial Closure Letter May 16, 2013 ML13137A131 Closure Option March 23, 2021 ML21082A264 Final Response The NRC staff reviewed the information provided by the licensee in response to GL 2004-02 and all request for additional information (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 The 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 Farley in support of the resolution of GL 2004-02:

x Replaced original containment sump strainers with nominal 1/8 inch diameter openings with seven horizontal stacked disk strainers and one vertical stacked disk strainer (with nominal 3/32 inch diameter openings). Unit 1 has the only vertical stacked strainer installed on the B-Train containment spray pump suction. Unit 1 and Unit 2 have a total of approximately 2,783 square feet (ft2) and 2,827 ft2 of perforated plate surface area, respectively.

x Completed detailed structural analysis of strainers.

x Implemented the removal of the refueling cavity drain covers during modes requiring ECCS operability.

x Installed debris interceptors inside containment for both units. No credit is taken in the analysis for the resulting reduced debris transport.

x Installed ECCS high head branch flow line orifices and changed the associated throttle valves to ensure that adequate clearance in the valve will prevent debris from plugging the flowpath.

x Completed ECCS throttle valves wear and blockage evaluation.

x Completed hydraulic model of the ECCS system.

x Completed latent debris sampling and characterization, including other debris sources, e.g., labels, etc.

x Completed debris generation and debris transport analyses.

x Completed ex-vessel downstream effects analysis.

x Completed downstream wear and blockage analysis.

x Completed NPSH analysis.

x Completed head loss analysis.

x Completed strainer bypass (penetration) testing.

x Completed vendors strainer head loss testing.

x Completed chemical effects testing (bench top and head loss testing)

x Established procedural and program controls to ensure materials used in the containments will not result in an increase of the debris loading beyond the analyzed values. This includes controls for containment coatings, labels, and insulation.

x Proceduralized actions to ensure that the post LOCA ECCS sump levels are maximized.

Based on the information provided by the licensee, the NRC staff considers this item closed for GL 2004-02.

3.0 BREAK SELECTION The objective of the break selection process is to identify the break size and location 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 the ruptured piping.

3.1 NRC Staff Review The NRC staff review is based on documentation provided by the licensee through April 29, 2008.

The licensee provided the content guide specified information. The licensee used a discrete approach to evaluate break locations rather than the Guidance Report (GR)/SE (ADAMS Accession No. ML050550156), which suggested systematic evaluation at 5-foot intervals along the piping. The break locations analyzed were chosen to maximize the amounts and types of debris generated. Therefore, the breaks analyzed were in the largest pipes near large debris sources such as a steam generator, reactor coolant pumps, pressurizer surge lines, and near walls and floors. Breaks located where debris would be more easily transported to the strainers were also considered. The licensee described three breaks in the primary loop piping, one on the 31 inch ID Loop C intermediate (crossover) leg near the base of the steam generator, one on the Loop B intermediate leg near the base of the steam generator, and one on the 27.5 inch ID Loop A cold-leg near the reactor coolant pump discharge. The break on the Loop B intermediate leg also affects the pressurizer surge line insulation and is thus the limiting break for insulation debris and is on the loop with the shorter path to the strainers than breaks in the other two RCS loops. The licensee indicated that the maximum insulation debris location is combined with the maximum coating debris location.

Secondary (main feedwater and main steam) piping breaks were not considered since the associated accident analyses do not credit ECCS recirculation.

The licensee stated that the majority of insulation inside the Farley containments is Transco reflective metal insulation (RMI) or Mirror RMI with a small amount of TempMat (fibrous) insulation on the steam generator instrument lines and reactor vessel bottom head. There is also Armaflex closed cell foam insulation on chilled water (service water and component cooling water) lines. Armaflex is very low density and will float if dislodged regardless of size.

The RCS line breaks at the reactor vessel were considered but not evaluated. The insulation that could be generated by breaks in the reactor cavity is bounded generation from breaks in the lines outside the reactor cavity. Much of the debris generated in the cavity would be contained within the reactor cavity and would not transport to the strainer.

3.2 NRC Staff Conclusion

For this 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 break selection evaluation for Farley is acceptable. Based on the information provided by the licensee, the NRC staff considers this area closed for GL 2004-02 for Farley.

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.

4.1 NRC Staff Review The initial NRC staff review is based on documentation provided by the licensee through April 29, 2008.

The licensee provided the content guide. The licensee used the GR Section 4.2.2.1.1 ZOI refinement of debris-specific spherical ZOI.

The licensee assumed the approved methodology (NEI 04-07 guidance report (ADAMS Accession Nos. ML050550138) and associated NRC staff SE (GR/SE) (ADAMS Accession No. ML050550156), default ZOI values of 2D for Transco RMI and 28.6D for Mirror RMI. The licensee indicated that there was a small amount of Temp-Mat fibrous insulation in each containment on steam generator instrumentation lines and on the reactor bottom head. All the Temp-Mat identified was assumed (one cubic foot identified as conservative estimate) to contribute to debris available for transport from all postulated breaks so no specific ZOI was assumed for this insulation.

The licensee stated that labels, tags, stickers, placards, and other miscellaneous or foreign materials were evaluated via walk down, which determined a total area of 36.4 ft2. Twice that area (72.86 ft2) was assumed for determining sacrificial area for strainer testing. Approved guidance for miscellaneous debris allows 75 percent of the total miscellaneous debris area to be used as sacrificial area. Based on these inputs 54.65 ft2 was used as sacrificial area in the analysis.

4.2 NRC Staff Conclusion

For the debris generation/ZOI review area, the licensee provided information such that the NRC staff has reasonable assurance that the subject review area has been addressed conservatively or prototypically. The NRC staff noted that the RMI debris is unlikely to contribute to head loss and the amount of fibrous debris is very small. Therefore, the NRC staff concludes that the debris generation/ZOI evaluation for Farley 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.

5.1 NRC Staff Review The initial NRC staff review is based on documentation provided by the licensee through July 27, 2009.

The licensee used industry standard debris sizing, densities, and destruction properties for most debris. The licensee assumed that all fibrous debris was rendered into fine debris, which is conservative. The licensee made an assumption that large pieces of RMI have a tumbling velocity of 0.20 ft/sec, which was lower than the 0.28 ft/sec tumbling velocity for small pieces.

The licensee assumed that a significant quantity of the qualified coatings debris would fail as chips based on the assumption that a filtering debris bed would not form on the strainer. The NRC staff noted that testing indicates that a filtering bed will form. Therefore, the staff stated that coatings should be added as particulate unless plant specific data is available that shows that the coatings fail in some other manner.

The NRC issued an RAI requesting the licensee provide the amount of qualified coatings assumed to fail as chips and the amount assumed to fail as particulate. This RAI was captured in one of the coatings RAIs and eliminated from the debris characteristics area. The licensee provided the information in its July 27, 2009 RAI response. Based on the matrix of tests that examined cases including chips, particulate, and mixtures of the two, as shown in Table 2 below, the NRC staff considered the information provided to be reasonable.

Table 2: Test Matrix and Test Result Summary Test TempMat Transco Silicon Coating Inorganic Temp Q HL (lbm) (lbm) Carbide Chips Zinc (°F) (gpm) (in.

(lbm) (lbm) (lbm) H2O) 1-0F-100SC-2X 0.24 0.89 123.1 0 47.9 59 239 19.2 63 198 18.3 1A-50F-OSC-2X 0.24 0.89 18.95 52.1 47.9 68 234 20.8 70 198 18.6 2-50F-50SC-2X 0.24 0.89 71.1 52.1 47.9 60 239 33.0 65 196 41.5 3-100F-OSC-2X 0.24 0.89 18.92 104.2 47.9 62 103 186

5.2 NRC Staff Conclusion

For the debris characteristics review area, the licensee 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 debris characteristics evaluation for Farley is acceptable. The NRC staff considers this item closed for GL 2004-02.

6.0 LATENT DEBRIS The objective of the latent debris evaluation 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.

6.1 NRC Staff Review The NRC staff review is based on documentation provided by the licensee through April 29, 2008.

The licensee provided a conservative estimate of 200 pounds-mass (lbm) of latent debris based upon a methodology that follows the recommendations of NEI and NRC. This estimate is based on a methodology of measurement and extrapolation that provides a calculated result of 125 lbm.

The latent debris mass, including dust, particulate, and lint was evaluated using 12 area types with 4 samples taken per area type. The accuracy of the individual mass measurements was 0.01 gram. The uncertainty of +/- 0.01 gram leads to an acceptably small uncertainty in the computed latent debris mass. The scale-up of sample area to containment surface areas was performed in accordance with NRC and NEI guidance. The 90 percent confidence limit of the mean value was conservatively used for each surface type. Of the 200 lbm, 15 percent was taken as fiber, and 85 percent as particulate. The NRC staff found this acceptable based on NEI and SE guidance.

The licensee performed a foreign material walk-down for Unit 2 and tabulated the areas of labels, tags, stickers, placards, and other materials. The licensee stated that the walk-down identified unqualified labels and tags equivalent to 36.4 square feet (ft2). This was doubled to 72.86 ft2 of area for conservatism. Overlap allowance was made and 75 percent of this area, or 54.56 ft2, was taken as the sacrificial area, as is consistent with NRC guidance. This area was split between the residual heat removal (RHR) (31.13 ft2) and CSS screens (23.52 ft2).

6.2 NRC Staff Conclusion

For the latent debris review area, the licensee 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 latent debris evaluation for Farley 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.

7.1 NRC Staff Review The NRC staff review is based on documentation provided by the licensee through April 29, 2008.

The licensee stated that its transport analysis considered the analytical steps of blow-down, wash-down, pool fill up, and recirculation transport. A computational fluid dynamics (CFD) analysis of pool flow patterns during recirculation was used to simulate transport during the recirculation phase. All fibrous debris was considered to be generated as fines. All debris except for RMI was assumed to be deposited on the containment floor. The NRC staff noted that these assumptions are conservative. The licensee conservatively assumed that all fine debris transported to the strainer. Because all fibrous debris was generated as fines, no erosion

evaluation was needed. The licensee installed debris interceptors but took no credit for them in the transport evaluation. The licensees transport evaluation was simplified and conservative.

The NRC staff noted that one issue that was not considered in the transport evaluation was the potential for one strainer to collect more debris than another. This is possible because each CSS and RHR pump has its own strainer and the strainers are arranged circumferentially in the annular space between the shield wall and the outer containment wall. Depending on which pumps are operating, the break location, the location of wash-down debris, etc., one strainer could attract more debris than another. However, due to conservatisms in the assumed debris amounts and characteristics of the debris, the NRC staff determined that this issue is insignificant.

Further margins were noted by the NRC with respect to the timing for the generation and transport of potential chemical precipitates. The NRC staff also noted that the NPSH margins for Farley are significant and increase as the sump temperature decreases.

7.2 NRC Staff Conclusion

For this review area, the licensee has provided information such that the NRC staff has reasonable assurance that the debris transport has been addressed conservatively or prototypically. Therefore, the NRC staff concludes that the debris transport evaluation for Farley is acceptable. Therefore, the NRC staff considers this area 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.

8.1 NRC Staff Review The NRC staff review is based on documentation provided by the licensee through July 27, 2009.

The licensee installed a separate General Electric (GE) strainer for each RHR and CSS pump for operation on recirculation. The NRC staff noted that although this scheme provides excellent redundancy and reduces flow through each strainer, the strainers installed at Farley are relatively small varying from 389 ft2 to 878 ft2. Continuum Dynamics Inc. (CDI) performed testing of the strainer under various debris loads. The NRC staff noted that Farley is a very low fiber plant with only 1 cubic foot (ft3) of fiber predicted to become debris in addition to the latent fiber in the containment. The licensee conservatively assumed 12.5 ft3 of latent fiber when containment surveys indicated that significantly less is actually present.

All of the strainers except the Unit 1 CSS B-train strainer are horizontally-stacked two-module strainers. The Unit 1 B-Train CSS strainer is a vertically-stacked three-module strainer. The strainer disk plates are perforated with 3/32-in holes and covered with woven wire mesh cloth.

The maximum flow rate is 4,500 gallons per minute (gpm) for each RHR pump and 3,400 gpm for each CSS pump. The strainer areas and maximum screen approach velocities are provided below in Tables 3 and 4, respectively.

Table 3: Strainer Areas, ft2 Unit 1 Unit 2 System Train A Train B Train A Train B RHR 878 878 878 878 CSS 638 389 638 433 Total 1,516 1,267 1,516 1,301 Unit Total 2,783 2,827 Table 4: Maximum Screen Approach Velocities, ft/s Unit 1 Unit 2 System Train A Train B Train A Train B RHR 0.0114 0.0114 0.0114 0.0114 CSS 0.0119 0.0195 0.0119 0.0175 For large-break LOCAs, the strainers are expected to be fully submerged at initiation of recirculation with at least 6 inches of submergence. The minimum sump pool water level is calculated to be 54 inches above the sump floor. However, the submergence was not calculated for small-break LOCAs (SBLOCAs). The licensees rationale was based on the lesser quantities of anticipated debris that would be generated for a SBLOCA. The NRC asked an RAI to ensure that this issue was addressed by the licensee. The licensee stated that the RHR strainer could have the water about 5.5 inches below the top of the strainer at the initiation of RHR recirculation during a SBLOCA. The strainer would be fully submerged in a short time after RHR is swapped to recirculation because containment spray continues to inject from the refueling water storage tank (RWST). Testing at about 11 times the plant flow rate found that air ingestion was observable with the water level about 9 inches below the top of the strainer.

Based on testing and the time it would take to form a debris bed on the strainer, the NRC staff determined that operation of the strainer under the limiting SBLOCA conditions would be acceptable.

The test debris loads assumed single train operation and that the debris reaching each strainer was proportional to the flow through the strainer. The smallest CSS strainer area was used for scaling of test parameters.

The licensee completed testing at CDI with a prototype strainer with chemical precipitates that resulted in a maximum head loss of 55.3-in (4.6 ft) at a temperature of 96.3 degrees Fahrenheit

(°F). The maximum head loss without the chemicals was 3-in (0.25 ft) at 93.8 °F.

The clean strainer head losses ranged from 0.25 ft to 2.3 ft. Therefore, the total head losses must be determined independently for the four individual strainers for each unit.

The licensee assumed that at least some of the qualified coatings would fail as chips instead of particulate. This issue could affect head loss test results. It is discussed in the coatings section.

On March 9, 2009, the NRC staff issued 12 RAIs for this area. The NRC staff asked RAIs to determine if the test program was conducted in a manner that would result in realistic or conservative head loss values. The RAI issues relate to debris preparation, debris addition, debris characteristics, post test data extrapolation for temperature and velocity, and extrapolation to ECCS mission times or a final value.

On July 27, 2009, the licensee responded to the RAIs. The licensee provided details of the test program and the use of the test data such that the staff was able to determine that the testing and use of data from the testing was performed appropriately, with a few exceptions discussed below. RAI 4, which questioned the debris addition procedures used during testing and verification of no settling or agglomeration of debris was not answered adequately. Also, during review of the RAI responses, additional questions arose regarding the inputs and assumptions for the head loss testing. The new issues that were identified are discussed below.

During the review of the response to RAI 1, the NRC staff could not determine the basis that the licensee used to evaluate various postulated conditions following a LOCA. The staff review found that the licensee credits delayed precipitation of chemical effects, but the conditions under which chemical effects are accounted for were not specified. The response to RAI 1 seemed to build a reasonable case for adequate NPSH margin, but the staff could not determine the conditions that were evaluated by the licensee and whether these conditions bounded the potential plant conditions. It was not clear whether extrapolations to temperatures other than the tested condition were made in the calculations.

During the review of the responses to RAIs 6 and 7, the NRC staff could not determine the basis for the flow rates and debris loading used during the final chemical effects head loss testing.

The debris loads specified for the testing nearly bounded the potential for all debris to transport to one RHR strainer. However, staff could not determine that the amount of debris in the test bounded the potential for debris transport to the smaller CSS strainers. It also appeared that the flow rate through the test strainer was less than would be expected when scaling for the smallest CSS strainer.

Senior NRC staff performed additional reviews of this area to determine whether the licensee had provided information that ensured reasonable assurance of system function under design conditions. The senior staff noted that the effects of chemical precipitates are clearly time dependent and would not add to strainer head loss until the sump pool had cooled significantly providing large NPSH margins. The NRC staff also identified that many conservatisms in the licensees analyses can be considered on a holistic basis to offset concerns related to head loss and vortex testing. There are significant conservatisms associated with assuming 100 percent transport of fine debris; and the licensee assumed the limiting coatings debris coincident with the limiting insulation debris (two separate break locations which cannot occur simultaneously).

The NRC staff noted the following additional conservatisms: (1) treating 15 percent of latent debris as fiber (a bounding number which is typically 7 percent or less); (2) considering 200 lbm quantity of latent debris, which is significantly greater than what plant-specific sampling indicated; (3) assuming 100 percent transport of miscellaneous debris to the strainer resulting in a very conservative sacrificial area during testing, assuming all fibrous debris is fines and 100 percent transports to the strainer; and, (4) installation of debris interceptors that were not credited in the analyses. The staff finds that these conservatisms are sufficient to offset uncertainties noted above.

Additionally, the staff considered the licensee margins provided on page 3 of the licensees February 28, 2008 submittal associated with testing the most limiting strainer to be significant:

CSS and Residual Heat Removal (RHR) strainer testing was performed using the most limiting strainer size for FNP [Farley Nuclear Plant]. The limiting strainer size for FNP is the CSS strainer. The RHR strainers are significantly larger and have lower hydraulic approach velocities. CSS strainer performance is required for a much shorter time

frame than RHR strainers. In effect, more limiting conditions were tested than are needed for adequate sump performance.

Regarding the CSS, the NRC staff considered that this strainer will not be in service full time, and additionally, the presence of redundant and independent strainers represents a design strength that mitigates challenges of passive failures, while still providing reasonable assurance that design basis conditions will not challenge NPSH margins for the spray pumps for the duration of their mission time.

8.2 NRC Staff Conclusion

For the head loss and vortexing area, the licensee has provided information such that the NRC staff has reasonable assurance that the strainer head loss and potential for air ingestion has been addressed conservatively or prototypically. Therefore, the NRC staff concludes that the head loss and vortexing evaluation for Farley is acceptable. The NRC staff considers this area 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.

9.1 NRC Staff Review The NRC staff review is based on documentation provided by the licensee through April 29, 2008.

The licensee generally provided the content guide specified information. The licensee used a standard industry practice methodology for calculating NPSH margin using a combination of realistic and conservative assumptions.

Farleys ECCS and CSS consist of 3 centrifugal charging pumps (CCPs) (high head safety injection function, normally 2 pumps operating in safety injection mode) and two trains each having a RHR pump (low head safety injection function), and a CSS pump. There is also a safety injection accumulator tank for each of the three RCS loop cold-legs. During the injection phase, all pumps start and take suction from the RWST with the CCP and RHR pumps discharging into the loop cold-legs and the CSS pumps to the containment spray ring headers.

The safety injection accumulator tanks also discharge into the RCS cold-legs when RCS pressure is low enough. When the RWST level drops to 12.5 feet, a low-level alarm signal occurs, and the operators manually transfer RHR pump suction to the containment sump and the CCP suction to the RHR pump discharge. The RHR pumps discharge through the RHR heat exchangers providing the core cooling. The CSS pumps continue to discharge water from the RWST to the containment until the RWST decreases to about 4.5 feet, at which time the operators manually transfer suction to the containment sump. At about 7.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> into the event, operators realign the RHR discharge from the RCS cold-legs to the RCS hot-legs. The CCPs continue to discharge to the RCS cold-legs.

The licensee indicated that the runout flows (system runout flow limited by orifices and throttle valves) for RHR pumps are 4,500 gpm, and for the CSS pumps are 3,400 gpm. The maximum ECCS recirculation flow is 15,800 gpm. However, each pump takes suction from a separate strainer and has its own NPSH margin. The licensee indicated that the CSS pump maximum

flow was determined using detailed hydraulic network software models with boundary conditions and input values chosen to predict conservatively high flow rates. No specific software program used or source document for friction or other flow losses was identified. This was consistent with the prior level of detail in the licensees updated final safety analysis report (UFSAR). The CSS pump curve in the UFSAR showed that the 3,400 gpm flow is a system runout flow as it was close to pump runout flow and about what would be expected to maintain the 40 pounds per square inch differential design pressure differential at the spray nozzles.

The transient curve in the UFSAR shows a maximum sump temperature of about 255 °F at the start of recirculation, 225 °F at 5,000 seconds after the event, and about 198°F at 10,000 seconds after the event. The minimum water level above an active strainer surface is 6 inches. However, a postulated, but not analyzed relatively small break at the top of the pressurizer (highest point in the RCS thus resulting in more water holdup from the recirculation pool to refill the RCS) would result in some limited duration (minutes) condition of incomplete strainer coverage at the start of ECCS recirculation until the CSS pumps deliver enough additional water from the RWST. The licensees rationale for not completely analyzing this condition was that the largest break at the top of the pressurizer would be a 6 inch pipe and would result in a very small fraction of the debris generated by the primary loop pipe breaks.

Also, the RHR pump flows from the recirculation strainers would be less for the smaller breaks and the continued operation of the CSS pumps transferring water from the RWST would cover the strainers within a few minutes, before significant debris buildup could occur on the strainers.

The NRC staff requested additional information regarding strainer submergence. This is discussed in the head loss and vortexing section.

The licensee stated that the NPSH required values were based on the industry standard practice criterion of 3 percent degradation in pump head. Containment accident pressure is not credited in the NPSH calculations.

The licensee showed the following minimum pump NPSH margins to occur at the initiation of recirculation. CSS transfer occurs several minutes after RHR pump transfer and RHR pump NPSH margin will be increasing by that time with the additional water the CSS pumps transfer from the RWST:

Unit 1 RHR Pumps: 0.98 feet Unit 1 CSS A Pump: 2.63 feet Unit 1 CSS B Pump: 3.04 feet Unit 2 RHR Pumps: 0.48 feet Unit 2 CSS A Pump: 2.40 feet Unit 2 CSS B Pump: 0.57 feet The assumptions for calculation of minimum sump water level include:

x ECCS and CSS recirculation occurs instantaneously at the appropriate RWST level alarms x RWST, safety injection accumulators, and the RCS contain at their design minimum water masses at the start of the event x Initial containment conditions assume a conservatively minimized atmosphere water content x RCS refloods conservatively above the break location and water temperature drops to sump temperature increasing density/mass in RCS

x Containment atmospheric water holdup is maximized x The reactor cavity (beneath reactor) fills up to the loop nozzles x Maximum allowed ECCS and CSS leakage outside containment x Normally empty portions of CSS and RHR piping are filled x CSS water in transit from spray nozzles to containment floor/surfaces is considered x Reactor cavity waste sump fills with coolant x Water is held up in containment surface films The licensee did not identify any single failure assumption but did state that each of the two RHR and two CSS pumps have their own separate suction strainers and that only one train (one RHR and one CSS pump) is assumed to operate, drawing all transported debris to those two most limiting strainers.

9.2 NRC Staff Conclusion

For the NPSH area, the licensee has provided information such that the NRC staff has reasonable assurance that it has been addressed conservatively or prototypically. Therefore, the NRC staff concludes that the NPSH evaluation for Farley is acceptable. The NRC staff considers this area 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.

10.1 NRC Staff Review The NRC staff review is based on documentation provided by the licensee through July 27, 2009.

The licensees coatings ZOI was reduced to 4D based on WCAP-16568-P, Jet Impingement Testing to Determine the Zone of Influence (ZOI) for DBA-Qualified/Acceptable Coatings, (ADAMS Accession No. ML061990594). Coatings in the ZOI failed as chips to simulate the plants worst condition, and the transport fraction was 100 percent, based upon CFD and NUREG/CR-6916, Hydraulic Transport of Coating Debris, (ADAMS Accession No. ML070220061). The failure of coatings as chips in the ZOI is justified by the NRC guidance, since the fiber loading is below thin bed quantities.

All unqualified coatings in containment are assumed to fail as 40 percent chips and 60 percent particulate. It was first assumed that they would all fail as chips. After CFD analysis, any of the chips that were expected to settle were modeled as particulate to maintain the 100 percent transport fraction.

Farley is a low fiber plant. In order to maximize head loss, the maximum amount of fiber and coatings debris of all the breaks were used for the head loss testing. This combination of fiber and coatings debris came from two different breaks and the occurrence of these two breaks would be unlikely to happen in an accident and is beyond the plant design basis. From the head loss test with this combination, a thin bed was observed. From the review guidance, if

there is a thin bed present, all coating debris should be treated as particulate and transport to the sump. If no thin bed is formed, then paint chips should be used in testing.

On March 9, 2009, the NRC staff asked 2 RAIs for this area. In the first RAI, the NRC staff pointed out a contradiction within the licensees February 28, 2008 submittal regarding the transport fraction for qualified coatings. On July 27, 2009, the licensee responded that the the previously reported 100 percent transport fraction for qualified coatings was in error. In the second RAI, the NRC staff requested information regarding the amounts of qualified and unqualified coatings assumed to fail as chips and the amounts of qualified and unqualified coatings assumed to fail as particulate used in the strainer qualification tests. The staff also requested the licensee to provide justification for treating qualified and unqualified coating debris as chips, if chips were used in the testing, given that the licensees supplemental response indicated that a thin bed is expected to form during strainer operation. In its response, the licensee clarified that the fiber loading is below thin bed quantities, and that the use of paint chips in the head loss testing resulted in higher head loss results. The NRC staff reviewed the responses and found them acceptable.

The NRC staff noted that the surrogate material used for testing is acceptable to the staff and that the licensees coating assessment program met expectations.

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0.2 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 Farley 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.

11.1 NRC Staff Review The NRC staff review is based on documentation provided by the licensee through July 27, 2009.

The licensee identified the 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, with the exception of how latent debris control will be accomplished. The NRC staff stated that it was not clear if a containment cleanliness program exists (cleaning practices and containment cleanliness surveillance procedures during refueling outages). This aspect is significant for Farley since this plants debris load is mostly RMI.

Procedure Foreign Material Exclusion Program, establishes the administrative controls and personnel responsibilities for the Foreign Material Exclusion (FME) program. This procedure places emphasis on the FME program and controls. The procedure describes methods for controlling and accounting for material, tools, parts, and other foreign material to preclude their uncontrolled introduction into an open or breached system during work activities. Additionally, procedure, Containment Inspection (General), provides detailed guidance for containment

inspection to ensure no loose debris (rags, trash, clothing, etc.) is present in the containment which could be transported to the containment sump and cause restriction of pump suctions during LOCA conditions. This procedure contains an extensive checklist detailing all areas of containment that must be inspected for cleanliness prior to plant startup after each outage.

Procedure, Containment Inspection (Post Maintenance), establishes guidance to inventory and control items carried into containment during non-outage entries. This procedure ensures that no loose debris (rags, trash, clothing, etc.) is present in the containment which could be transported to the containment sump and cause restriction of pump suctions during LOCA conditions.

The final safety analysis report (FSAR) is reviewed during preparation for each design change.

The FSAR was updated to reflect the analytical assumptions and numerical inputs of the analysis supporting the modifications made in response to GL 2004-02. The licensee stated that an enhancement to the engineering guidance procedure will be made as part of the design change process. This procedure provides screening guidelines for performing design change activities. This enhancement will provide guidance for reviewing the impact of a proposed change on potential effects on sump operation. The specific areas that will be addressed are:

insulation, coatings, inactive volumes, and labels inside containment; structural changes (i.e.,

choke points) inside containment; and downstream effects (piping components downstream of the ECCS sump strainers). Inclusion of this guidance in the engineering guidance procedure will ensure that design changes consider these attributes during the design process.

Maintenance activities, including temporary changes are subject to the provisions of 10 CFR 50.65(a)(4), as well as Farleys TSs. The licensees fleet procedures also provide guidance such as the 50.59 Review Process Procedure, which provides details and guidance on maintenance activities and temporary alterations, the On-line Work Control Process Procedure, which establishes the administrative controls for performing on-line maintenance of structures, systems, components (SSC) in order to enhance overall plant safety and reliability, and the Temporary Configuration Changes (TCC) Procedure, which establishes the overall requirements for TCC.



On March 9, 2009, the NRC staff issued an RAI requesting the licensee to explain how its containment cleanliness and FME programs assure that latent debris in containment will be controlled and monitored to be maintained below the amounts and characterizations assumed in the ECCS strainer design. The staff also asked if latent debris sampling will become an ongoing program. On July 27, 2009, the licensee responded to the RAI. In its response, the licensee stated that an enhanced containment cleaning program will be performed on a three-outage basis. This enhanced cleaning program will focus on removal of latent debris.

Latent debris in containment consists of dirt, hair, and other particles or fiber type debris that generally is carried into containment on personnel/equipment during outage maintenance periods. The licensee stated that the amount of latent debris inside Farley containment has been sampled and the results have been provided as input in the containment sump screen design calculations. The values used in these calculations included a substantial margin. The licensee stated that this enhanced cleaning program, performed on a three-outage basis, provides reasonable assurance that latent debris in containment will remain below the conservative values used in the containment sump screen design calculations.



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1.2 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 debris source term evaluation for Farley 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.

12.1 NRC Staff Review The NRC staff review is based on documentation provided by the licensee through April 29, 2008.

The licensee provided a description of the major features of the sump screen design modification and a list of modifications for both Farley units. The licensee contracted with GE to provide sump strainers that meet the requirements of GL 2004-02 (seven horizontal stacked disk strainers and one vertical stacked disk strainer). The strainers were installed in both units' RHR and CSS suction points. Unit 1 has the only vertical stacked strainer installed on the B-Train CSS suction. The strainers are located outside the bio-wall between the bio-wall and containment outside wall. This location protects the strainers from missile impacts. In addition, debris interceptors were installed inside the shield wall to minimize the transport of debris to the strainers. These interceptors are not credited in this analysis. To remove a possible holdup source, the refueling cavity drain covers are removed prior to the unit returning to power after a refueling outage. The safety injection throttle valves have been replaced with new valves that allow for the valves internal opening to be greater than the strainer hole size.

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2.2 NRC Staff Conclusion

For the screen modification package review area, the licensee provided screen location, configuration, and construction information such that the NRC staff has confidence in the design of the strainer. Therefore, the NRC staff concludes that the screen modification package information provided for Farley 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.

13.1 NRC Staff Review The NRC staff review is based on documentation provided by the licensee through April 29, 2008.

The licensees evaluation performed for the new sump strainer assembly demonstrates compliance with the American Society of Mechanical Engineers (ASME) Boiler and Pressure

Vessel Code,Section III, 1989 Edition. The various components are designed in accordance with the corresponding subsections of the code (e.g. NC, ND, and NF).

The licensee stated that a variety of structural analysis techniques and methods were employed to qualify the replacement sump strainers and associated piping. ANSYS computer software was used to develop finite element analysis models. The models were subjected to loadings associated with differential pressure, dead weight, debris weight, seismic loading, hydrodynamic effects, and thermal loading combinations as appropriate. These loads are consistent with the guidance of NEI 04-07. With respect to seismic considerations, the strainers and piping configurations were found to be predominantly in the rigid range for natural frequency. For this reason, the zero period acceleration values were typically employed on the models as equivalent static forces. In instances where the configuration was found to be below the rigid range, the licensee employed a multi-mode factor increase to the corresponding acceleration or (in one case) performed a complete response spectrum dynamic analysis. In all analytical evaluations, the stresses induced in each of the strainer and piping components was stated to be within the allowable limits of the governing criteria.

To address the possibility of dynamic loadings due to a HELB, the licensee stated that the loadings were precluded by the location of the sump strainers.

The strainers were not subjected to a reverse loading condition associated with back flushing because the licensee is not taking credit for this operation as a mitigating measure.

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3.2 NRC Staff Conclusion

For this review area, the licensee has provided information such that the NRC staff has reasonable assurance that overall the subject review area has been addressed conservatively or prototypically. Therefore, the NRC staff concludes that the sump structural analysis evaluation for Farley 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.

14.1 NRC Staff Review The NRC staff review is based on documentation provided by the licensee through July 27, 2009.

Evaluations of containment along with review of the CFD model indicate no significant areas will become blocked with debris and hold up water during the sump recirculation phase. The area of the refueling cavity, which is the area around the reactor head that is flooded prior to fuel movement, is the only significant area in containment that can retain water.

The location of the postulated limiting LOCA is inside the secondary shield wall in the lower elevations of the containment. The flow path from this break area to the sump strainers is primarily through two labyrinth egress points through the shield wall. These walkways provide a large, clear flow path from inside the shield wall to the strainer area. There are also smaller openings through the shield wall for pipes, but these are much smaller than the walkways and

any restriction of these would have minimal effect on the overall flow path from inside the shield wall to the strainers.

Containment spray wash down has a clear path to the containment sump area. Large sections of the floor on each level in containment are covered with grating that allows the water to pass.

Water that falls into the refueling cavity exits via the cavity drains to the sump.

Visual inspections were performed on each unit to identify potential choke point for water flowing to the sumps. The licensee identified and modified the reactor cavity drain covers to be removed prior to the unit returning to power after each refueling.

There are no curbs that provide water volume holdup in the containments. The installed debris interceptors are 25 inches high, which is well below the calculated minimum water level of 54 inches. Complete blockage of these will not significantly impede flow to the screens.

The refueling cavity drains were identified as a point that required modification. The licensee modified the reactor cavity drain covers to be removed after each refueling. This provides a large clear flow path that cannot be easily blocked with debris. Since Farley is mostly a RMI plant, any RMI that is blown into the cavity would tend not to be flat and thus would not be expected to block both drains. The covers on these drains are removed prior to entry into Mode 4 from Mode 5. The drains are 6 inch pipes with approximately 8 inch inlets. There are two drains located approximately 12 feet apart and are in the vicinity of the containment fuel handling up-ender frame, which would tend to prevent any large debris from landing on the drains. The limiting break occurs under the operating deck and inside the secondary shield wall. Debris generated from this break would have to travel a tortuous path to reach the refueling cavity. Therefore, the clogging of the reactor cavity drains is not postulated.

The drains into the area under the reactor (reactor cavity) could become blocked. There is no detrimental impact of this blockage as it would inhibit loss of water from the active ECCS sump to an inactive area beneath the vessel. The ECCS sump level analysis assumes this area floods during the event.

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4.2 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 upstream effects evaluation for Farley 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.

15.1 NRC Staff Review The NRC staff review is based on documentation provided by the licensee through July 27, 2009.

The licensee provided a detailed description of the methods used to evaluate the downstream effects of debris that bypass the ECCS sump strainers. Generally, the licensee followed the methods of WCAP-16406-P, Evaluation of Downstream Sump Debris Effects in Support of GSI-191, Revision 1. Westinghouse performed the ex-vessel downstream-effects evaluations for the licensee.

The licensee evaluated the downstream impact of sump debris on the performance of the ECCS and CSS following a LOCA. The effects of debris ingested through the containment sump strainer during the recirculation mode of the ECCS and CSS include erosive wear, abrasion, and potential blockage of flow paths. The smallest clearance found for the heat exchangers, orifices, and spray nozzles in the recirculation flow path is 0.375 inches (3/8) for the containment spray nozzles. The licensee stated that with a sump strainer hole size of 0.09375 inch (3/32), no blockage of the ECCS flow paths is expected. The instrumentation tubing was also evaluated for potential blockage of the sensing lines and the licensee concluded that the transverse velocity past the tubing would be sufficient to prevent debris settlement into the lines; therefore, no blockage is expected to occur. The reactor vessel level instrumentation system was also evaluated, and the licensee concluded that no effect on its performance is expected due to the debris.

The licensee evaluated the heat exchangers, orifices, and spray-nozzles for the effects of erosive wear with a debris concentration of 764.47 parts per million (ppm) over the mission time of 30 days and concluded that the erosive wear on these components would be insufficient to affect the system performance. The wear evaluation on the ECCS and CSS equipment was performed using the wear models developed in WCAP-16406-P, Revision 1. The licensee evaluated pumps for the effect of debris ingestion through the sump strainer. Three aspects of operability (hydraulic performance, mechanical shaft seal assembly performance, and mechanical performance (vibration) of the pump), were evaluated. The licensee determined that the hydraulic and mechanical performances of the pump would not be affected by the recirculating sump debris. The mechanical shaft seal assembly performance evaluation resulted in a recommendation to replace the RHR pumps carbon/graphite backup seal bushings with a more wear resistant material, such as bronze. However, this recommendation was not implemented because Farley has an Engineered Safety Feature atmospheric filtration system in its auxiliary building and this action was not required. The licensee stated that evaluations of the system valves showed that the minimum recirculation flow rates would be adequate to preclude debris sedimentation in all cases. The licensee stated that all of the valves that could be subject to blockage, pass the plugging criteria at their current disc positions and all of the valves that are subject to erosion, pass the acceptance criteria for the mission time of 30 days.

The licensee replaced both units ECCS branch flow throttle valves. Three of the 12 replacement valves on Unit 1 were determined to have clearances of approximately 106 percent of the strainer hole size, deviating from the 110 percent criteria in WCAP-16406. The remaining nine replacement valves all have clearances greater than 110 percent of the strainer hole size.

The licensee performed a plant-specific evaluation of the clearance deviation and concluded that deformable debris that may pass through the replacement sump strainer at Farley Unit 1 will also pass through the three high pressure safety injection valves and will not cause blockage of these valves.

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5.2 NRC Staff Conclusion

For the ex-vessel downstream effects 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.

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.

16.1 NRC Staff Review The NRC staff review is based on documentation provided by the licensee through March 23, 2021.

The SNC applied the NEI clean plant criteria to determine the amount of fibrous debris penetrating the sump strainers for use in the downstream in-vessel debris analysis for Farley.

The clean plant criteria, as applied to in-vessel effects, use a fiber penetration (bypass) fraction of 45 percent and debris transport fraction of 75 percent. Based on the clean plant criteria values, the licensee computed a Farley specific in-vessel debris load. The mass of latent fibrous debris for Farley is assumed to be 30 lbm. The worst-case amount of fibrous debris generated is 1 ft3 of Temp-Mat. The density of Temp-Mat is 11.8 lbm/ft3, which results in a worst-case fibrous debris generation of 11.8 lbm. Based on that information, the licensee calculated a 40.8 g/FA in-vessel fiber load.

The licensee provided a discussion for the assumed 45 percent fiber penetration bypass and 75 percent debris transport fractions. The licensee used information from the Vogtle responses to justify the fiber penetration bypass fraction. The licensee provided a comparison table of the critical parameters for the sump strainer bypass testing.

The licensee stated that the Vogtle bypass testing discussion shows that the prompt fiber penetration (bypass) fraction is less than 45 percent with no fiber bed on the strainer and then quickly decays as the fiber bed forms. The total fibrous debris load at the Farley strainers assuming 75 percent transport is 14,220 grams. The prompt bypass decays to between 2 to 3 percent at this fiber loading. The licensee stated that reviewing the Vogtle bypass testing confirms that the prompt bypass fraction makes up the majority of the fiber entering the RCS and that fiber shedding through the sump strainers is only a minor contributor. Therefore, the licensee claimed that the use of a constant 45 percent bypass fraction for Farley is conservative and justified.

The licensee stated that the debris transport fraction of 75 percent from the NEI clean plant criteria is applied to Farley. Farley is a very low fiber plant with containment insulation primarily consisting of RMI. The licensee claimed that the latent fiber considered in the calculation of the in-vessel debris is conservatively high at 30 lbm. The licensee stated that plant walkdowns indicated that the bounding amount of latent fiber is less than 19 lbm. The licensee also stated that the Temp-Mat insulation considered in the calculations is maximized at 11.8 lbm even though the location of the insulation prevents all of the insulation from being a credible debris

source from a pipe rupture. Thus, there is significant conservatism in the amount of fiber considered in determining the in-vessel debris load.

The licensee stated that the NEI clean plant criteria discusses numerous points supporting that the use of the 75 percent debris transport fraction is reasonable. The Farley containment is highly compartmentalized resulting in expected debris hold-up in numerous inactive volumes.

Certain areas of containment may collect debris from a pipe break that may not allow the debris to transport to the sump. Debris interceptors are installed, which would likely preclude some debris from reaching the strainers. The effectiveness of settling and spray washout mechanisms is not known and will only account for some latent debris capture/transport. The majority of the fine latent debris will likely be left in place due to condensation drainage; directly sprayed surfaces will have the majority of fine debris removed but the retention of that debris is uncertain.

The licensee concluded that considering these aspects, a 75 percent debris transport fraction is a reasonable assumption for the in-vessel fibrous debris load.

The licensee stated that Farley is a Westinghouse 3-loop PWR with an upflow barrel/baffle configuration. Given the staff guidance, it is necessary to confirm that Farley is within the key parameters of the WCAP-17788-P, Revision 1 Comprehensive Analysis and Test Program for GSI-191 Closure (ADAMS Package Accession No. ML20010F181) methods and analysis.

Farley uses Westinghouse 17x17 OFA fuel.

The proprietary total in-vessel fibrous debris limit contained in Section 6.5 of WCAP-17788 applies to Farley. The 40.8 g/FA in-vessel fiber amount is based on the assumption that all fibrous debris calculated to penetrate the strainer reaches the reactor vessel. The maximum amount of fiber calculated to reach the reactor vessel at Farley is less than the in-vessel fibrous debris limit in WCAP-17788.

The licensee stated that the applicable WCAP-17788 core inlet fiber threshold is provided in Table 6-3 of WCAP-17788-P. The core inlet fiber amount for Farley is 40.8 g/FA, which is less than the applicable WCAP-17788-P core inlet fiber threshold.

The licensee stated that as described in the UFSAR Section 6.3.2.2.7, the earliest possible sump switchover time for Farley is 20 minutes.

The licensee stated that chemical precipitation timing is dependent on the plant buffer, sump pool pH, volume and temperature, and debris types and quantities. The licensee provided a table summarizing the key chemical precipitation parameters and values for Farley and compared them to test group 43 from WCAP-17788-P, Volume 5. The staff notes that the pH of 10.5 shown in the licensees table is unusually high for a plant with a trisodium phosphate (TSP) buffer. This was recognized by Westinghouse and the WCAP-17788 Volume 5 tests for Test Group 43 which were performed at a pH of 9. Based on the comparison, test group 43 is representative of Farley and the predicted chemical precipitation timing (tchem) is 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The licensee stated that the test group 43 aluminum surface area is 16,079.8 ft2. However, using the Farley minimum sump volume and test group 43 autoclave data, the equivalent aluminum is scaled up to 17,049.18 ft2. The licensee stated that using this value, the Farley quantity of aluminum is bounded by the WCAP testing. The licensee noted that the plant quantity includes aluminum sources which are either protected from containment spray or are located such that the containment spray cannot be transported to the active sump. Therefore, approximately

214 ft2 of additional aluminum margin is available in the Farley value of 16,099.98 ft2. The NRC staff reviewed the Test Group 43 data in Volume 5 of WCAP-17788-P and confirmed that the data supports a chemical precipitation timing of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

The licensee stated the amount of galvanized steel and insulation material does not impact tchem; however, the amount of these materials should not exceed the maximum tested for each buffer type. The licensee stated that the amount of insulation debris from test group 43 is bounded by the Farley values. The amount of galvanized steel at Farley is bounded by the amount tested for test groups 33 and 34, which both use TSP as a sump buffer and consider greater than 55,000 ft2 of galvanized steel. Therefore, the licensee stated that the Farley quantity of galvanized steel is acceptable.

The licensee stated that based on the above comparison, test group 43 is representative of Farley.

The licensee stated that as described in the UFSAR Section 6.3.2.2.7, injection realignment to mitigate the potential for boric acid precipitation is performed no later than 7.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />, which is less than predicted chemical precipitation time of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

The licensee stated that tblock for Farley is 143 minutes. The earliest time of chemical precipitation for Farley was 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, which is greater than the applicable tblock of 143 minutes.

Farley has a rated thermal power of 2,821 megawatts thermal. The applicable analyzed thermal power is 3,658 as provided in WCAP-17788-P, Volume 4, Table 6-1. Since the Farley thermal power is less than the analyzed power, the parameter is bounded by the WCAP-17788 alternate flow path analysis.

The licensee stated that the proprietary analyzed alternate flow path (AFP) resistance is provided in Table 6-1 of WCAP-17788-P, Rev. 1, Volume 4. The proprietary Farley specific AFP resistance is provided in Table RAI-4.2-24 and is less than the value analyzed in WCAP-17788-P, Rev. 1, Volume 4. Since the Farley AFP resistance is less than the analyzed value, the Farley AFP resistance is bounded by the resistance applied to the AFP analysis.

The licensee stated that the AFP analysis for Westinghouse upflow plants analyzed a range of ECCS recirculation flow rates from 8 - 40 gpm/FA. The licensee stated that the Farley ECCS recirculation flow rate corresponding to the worst-case GSI-191 hot-leg break scenario is 8.9 gpm/FA which is within the range of ECCS recirculation flow rates considered in the AFP analysis.

The licensee concluded that based on the comparison of key parameters used in the WCAP-17788 AFP analysis to the Farley specific values, Farley is bounded by the key parameters and the WCAP-17788 methods and results are applicable.

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6.2 NRC Staff Conclusion

For the in-vessel downstream effects review area, the licensee has provided sufficient information such that the NRC staff has reasonable assurance that the subject review area has overall been addressed conservatively or prototypically. The licensee used approved guidance to calculate the debris amounts reaching the reactor vessel and provided additional information to justify that these values are applicable to Farley. Chemical effects testing documented in WCAP-17788-P showed that precipitates would not form before 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, long after the time for

hot leg switch over (HLSO). The licensee also demonstrated that the key parameter plant-specific values specified in NRC staff guidance are bounded by the analysis values from WCAP-17788. 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 head loss and core cooling. The licensee evaluation of in-vessel chemical effects is discussed in Section 16 DOWNSTREAM EFFECTS - FUEL AND VESSEL.

17.1 NRC Staff Review The NRC staff review is based on the documentation provided by the licensee in GL 2004-02 supplemental responses dated February 28, 2008, April 29, 2008, July 27, 2009, and March 30, 2010. The ADAMS Accession numbers for these letters are provided in Section 1.0, INTRODUCTION above. The reference documents used for this review include content guidance for chemical effects provided in the March 31, 2008, NRC Staff SE of WCAP-16530-NP-A, WCAP-16530-NP-A, Evaluation of Post-Accident Chemical Effects in Containment Sump Fluids to Support GSI-191 (ADAMS Accession No. ML081150379) and the March 28, 2008 guidance, NRC Staff Review Guidance Regarding Generic Letter 2004-02 Closure in the Area of Plant-Specific Chemical Effects Evaluations (ADAMS Accession No. ML080380214).

Farley uses -TSP for post-LOCA pool pH control. The licensee has replaced the original sump strainers with GE stacked disk strainers. The strainer assembly for each unit consists of both RHR and CSS strainers.

The licensees plant-specific debris generation and transport analyses determined that the debris sources for Farley include Transco RMI, Mirror RMI, Armaflex, and TempMat Fiber.

17.2 Chemical Precipitate Calculation The licensees chemical effects test for Farley, as detailed in the February 28, 2008 letter (ADAMS Accession No. ML080660654), was based on the WCAP-16530-NP methodology (ADAMS Accession No. ML081150379). The WCAP methodology involves determining the chemical precipitate load, preparing the calculated amount of precipitates, and adding the pre-mixed precipitates to the test loop after a debris bed is formed on the test strainer. The NRC staff has accepted the WCAP-16530-NP approach for calculating the quantity of chemical precipitate to add to strainer testing.

The licensee applied the following assumptions to the WCAP-16530-NP methodology for Farley:

1. A maximum post-LOCA pool pH was conservatively calculated for a large break LOCA by assuming plant parameters that yielded a maximum pH. The assumed maximum sump pH value of 8.6 results in a maximum aluminum dissolution rate.

2. The licensee assumed that metallic aluminum, Temp-Mat and concrete would contribute to chemical effects. All plant aluminum inventory was assumed to corrode even if it is not submerged or exposed to containment sprays. Aluminum that dissolves and is

transported to an inactive pool (i.e., would not reach the sump pool) following a LOCA was also included in the chemical precipitate quantity.

3. The licensee did not credit inhibition of aluminum corrosion from phosphate (TSP buffer) in the source term calculation.

4. The licensee did not credit long term solubility of aluminum in the sump pool.

The results of the WCAP-16530-NP calculation predicted that the expected precipitates were sodium aluminum silicate, aluminum oxy-hydroxide, and calcium phosphate. The licensees results using plant-specific inputs showed that the expected maximum precipitate amounts were 0.7 lbs. of calcium phosphate, 7.2 lbs. of sodium aluminum silicate and 988.3 lbs. of aluminum oxy-hydroxide.

17.3 Testing Farley plant-specific chemical effects head loss was determined by two sets of chemical effects testing:

x Bench-top testing - to determine the temperature at which aluminum oxy-hydroxide precipitates would form. This temperature was established at 140 °F. These tests were performed at the VUEZ facility.

x Integrated chemical effects head loss testing - to determine the debris induced head loss (including chemical precipitates) by scaling the plants debris load across a strainer module. The quantity of chemical precipitate added to the prototype tank testing was scaled from the full plant predicted values obtained from WCAP-16530-NP-A. Strainer head loss testing was performed at the CDI facility in Ewing, NJ.

The NRC staff observed chemical effects testing at both the VUEZ facility in Slovakia and the CDI facility in NJ. Summaries of NRC staff visits are available in ADAMS (ADAMS Accession Nos. ML073450430 and ML070670071).

17.4 Previous RAI Review Following the review of the chemical effects evaluation detailed in the licensees April 29, 2008 supplemental response (ADAMS Accession No. ML081210452), the NRC staff identified that additional information was needed to determine if the testing was performed in an acceptable manner. Therefore, on March 9, 2009, the NRC staff issued RAIs (RAIs 16-18) (ADAMS Accession No. ML090620117). On July 27, 2009 (ADAMS Accession No. ML092380647), the licensee responded to the March 9, 2009 RAIs. The RAIs and the licensees responses to the RAIs are discussed below:

For RAI 16, the NRC staff referenced the value given for the mass of aluminum oxy-hydroxide in different licensee submittals. The licensee stated in its February 28, 2008 letter that the mass of aluminum oxy-hydroxide was 729 lbs.; however, in its April 29, 2008 letter, the licensee stated that the mass value was 988 lbs. Therefore, the NRC staff requested the licensee explain the discrepancy in these values and identify which amount of aluminum oxy-hydroxide was used as a basis for the chemical effects head loss testing. The NRC staff also requested the licensee to further discuss the interpretation of the test results if less than the predicted amount of chemical precipitate was used in head loss testing.

In response to RAI 16, the licensee stated that subsequent to the submission of the February 28, 2008 letter, it conducted a different review of the chemical effects quantities assuming longer containment spray run times. The licensee explained that the review resulted

in a greater aluminum oxy-hydroxide quantity and the greater of the two values was used for head loss testing.

The NRC staff evaluated the licensees response to RAI 16 and found it acceptable since the licensee added additional conservatism to their chemical effects test by assuming longer containment spray run time and more aluminum oxy-hydroxide.

For RAI 17, the NRC staff referenced the licensees WCAP-16530-NP chemical spreadsheet results which predicted that most of the precipitates would come from aluminum oxy-hydroxide.

The licensees April 29, 2008 letter provided the one-hour precipitate settlement data for calcium phosphate and sodium aluminum silicate; but not for aluminum oxy-hydroxide. Therefore, the NRC staff requested the licensee provide the one-hour precipitate settlement data for the aluminum oxy-hydroxide precipitate used in head loss testing.

In response to RAI 17, the licensee stated the aluminum oxy-hydroxide turbidity measurement during testing indicated 98 percent of the solution in the graduated cylinder was turbid at one-hour. This measurement meets the acceptance criteria stated within WCAP-16530-NP-A.

The NRC staff evaluated the licensees response to RAI 17 and found it acceptable since the licensee provided the settlement data used during testing for aluminum oxy-hydroxide precipitate and this data meets the acceptance criteria from WCAP-16530-NP-A.

For RAI 18, the NRC staff requested the licensee provide experimental data that would support the assumptions used for bench-top testing performed by Alion in which aluminum-based precipitates will not form at temperatures above 140 °F.

In response to RAI 18, the licensee stated that the plant-specific bench-top experiments performed by Alion identified a visible precipitate occurring on or about day 17 while temperatures were reduced from 200 °F to 140 °F. At day 17, the sump temperatures were calculated to be approximately 130 °F. The licensee also stated that the post-LOCA sump temperatures are reduced below 200 °F within about 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> following a LOCA event. In addition, the licensee provided utility test data from another licensee which provided flow loop testing done with a similar sump pH as Farley. For more details, refer to the licensees July 27, 2009 RAI 18 response.

The NRC staff evaluated the licensees response to RAI 18 and found it unacceptable since the licensee did not provide detailed information concerning post-LOCA pH range and the assumptions used to calculate the dissolved aluminum concentration. Without this information, the NRC staff was unable to reach reasonable assurance that the chemical effects area was adequately addressed.

Following the review of the July 27, 2009 RAI responses, the NRC staff requested a teleconference (held on February 23, 2010) with the licensee to discuss some of the responses addressing the chemical effects testing. On March 30, 2010, the licensee provided additional information for RAI-18 (ADAMS Accession No. ML100900004). The NRC staff reviewed the licensees March 30, 2010 response and concludes that the licensee adequately addressed the chemical effects area for the following reasons:

x The licensees calculation of the post-LOCA sump pool aluminum concentration is conservative. The licensee assumed the maximum post-LOCA pH which increases the aluminum release. The licensee assumed a conservative containment spray duration

of 12.9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> instead of the projected post-LOCA spray duration of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. All plant aluminum was assumed to corrode even if it is not wetted or there is no water path from the aluminum to the containment sump pool. Finally, the licensee takes no credit for aluminum corrosion inhibition by TSP. These assumptions resulted in a conservative calculated aluminum concentration of 51 ppm.

x The licensees integrated chemical effects head loss tests with a full WCAP-16530-NP-A chemical precipitate load show that sufficient NPSH margin exists below a temperature of approximately 200 °F. The NRC staff estimated the aluminum solubility limit using plant specific parameters using the Argonne National Laboratory solubility equation (see Equation 4 of the Argonne National Laboratory Technical Letter Report on Aluminum Solubility in Boron Containing Solutions as a Function Of pH And Temperature (ADAMS Accession No. ML091610696)). Using a pH in the middle of the plant specific range, instead of the maximum pH that was used to maximize aluminum release (since lower pH reduces solubility), the NRC staff calculated a solubility temperature limit of approximately 150 °F for a 50 ppm aluminum concentration. The licensee demonstrated sufficient NPSH margin below 200 °F. Therefore, the NRC staff would not expect precipitation to occur until there is adequate NPSH margin to account for all dissolved aluminum precipitating and reaching the sump strainer.

The NRC staff finds the overall chemical effects evaluation for Farley acceptable since an NRC approved WCAP-16530-NP methodology was used to calculate the plant-specific chemical precipitate and to prepare the chemical precipitate load that was added into the licensees integrated chemical effects head loss test. The licensee used the WCAP-16530-NP-A base model without any refinements, which is acceptable to the NRC staff. The licensee also addressed all RAIs.

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7.5 NRC Staff Conclusion

For the chemical effects review area, the licensee has provided sufficient information such that the NRC staff has reasonable assurance that the chemical effects review area has been addressed conservatively or prototypically for Farley. 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.

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. The NRC staff review is based on documentation provided by the licensee through March 23, 2021.

The licensee stated that the Farley licensing basis was changed in accordance with the requirements of 10 CFR 50.71. Farleys FSAR, Appendix 6D describes the new ECCS screens installed to address GL 2004-02. A description of the new Unit 1 and 2 screens, including size, assembly details, and figures was added. Also included was:

x A summary approach used to size the new screens using the guidance of NEI 04-07 and the containment walk down used to confirm installed installation x A description of pipe break characterization, debris generation, latent debris accumulation, and debris transport to the containment sump.

x RHR and CSS pumps head losses as a result of debris accumulation, including the vortexing analysis.

x The sump structural analysis, including a description of the passive screen.

x The upstream effects of debris accumulation, downstream effects associated with any debris bypass, and chemical effects testing.

x Tables for debris generation ZOI, LOCA generated insulation debris inside ZOI, debris generated from coatings based on 4D ZOI, latent and foreign material debris used in the analysis, and summary of debris generated and transported to the screen modules.

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8.1 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. 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 DBA, 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 Joseph M. Farley Nuclear Plant, Units 1 and 2.

Principal Contributors: S. Smith A. Russel Date: September 24, 2021

ML21260A161 OFFICE NRR/DORL/LPL2-1/PM NRR/DORL/LPL2-1/LA NRR/DSS/STSB/BC(A)

NAME SDevlin-Gill KGoldstein NJordan DATE 09/17/2021 09/23/2021 07/28/2021 OFFICE NRR/DNRL/NCSG NRR/DORL/LPL2- NRR/DORL/LPL2-1/PM 1/BC(A)

NAME SBloom GEMiller (for MMarkley) SDevlin-Gill DATE 09/17/2021 09/24/2021 09/24/2021