AEP-NRC-2022-03, Final Supplemental Response to NRC Generic Letter 2004-02

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Final Supplemental Response to NRC Generic Letter 2004-02
ML22024A167
Person / Time
Site: Cook  American Electric Power icon.png
Issue date: 01/20/2022
From: Lies Q
American Electric Power, Indiana Michigan Power Co
To:
Document Control Desk, Office of Nuclear Reactor Regulation
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ML22024A166 List:
References
AEP-NRC-2022-03
Download: ML22024A167 (23)
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t:m INDIANA Indiana Michigan Power MICHIGAN Cook Nuclear Plant POWER* One Cook Place Bridgman, Ml 49106 A unit ofAmerican Electric Power lndianaMichiganPower.com January 20, 2022 AEP-NRC-2022-03 10 CFR 50.54(f)

Docket Nos.: 50-315 50-316 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001 Donald C. Cook Nuclear Plant, Unit 1 and Unit 2 Final Supplemental Response to NRC Generic Letter 2004-02 The purpose of this submittal is to provide the Indiana Michigan Power Company (l&M), the licensee for Donald C. Cook Nuclear Plant (CNP) Units 1 and 2, final supplemental response to Generic Letter (GL) 2004-02, dated September 13, 2004 (ML042360586), "Potential Impact of Debris Blockage on Emergency Recirculation during Design Basis Accidents at Pressurized-Water Reactors."

On May 15, 2013, l&M submitted a letter (ML13137A046) of intent per SECY-12-0093, "Closure Options for Generic Safety Issue - 191, Assessment of Debris Accumulation on Pressurized-Water Reactor Sump Performance" indicating l&M would pursue Closure Option 2 - Deterministic of the SECY recommendations (refinements to evaluation methods and acceptance criteria). The final outstanding issue for CNP with respect to GL 2004-02 is the in-vessel downstream effects evaluation, which addresses that long-term core cooling can be adequately maintained for all postulated accident scenarios that require sump recirculation.

l&M has completed an in-vessel downstream effects evaluation for CNP Units 1 and 2 that is documented in Enclosure 2 to this letter. This evaluation satisfies the GSl-191 commitment identified in the May 15, 2013 (ML13137A046), Closure Option letter and updated in the October 27, 2016 (ML16302A395) letter. l&M also has an open commitment from May 19, 2011 (ML11147A072), to include the effects of Temp-Mat in the in-vessel effects analysis. The Temp-Mat material is included in the calculated debris amounts used in the evaluation contained in the enclosure to this letter.

Enclosure 1 to this letter contains the affirmation statement. Enclosure 2 contains the CNP non-proprietary final response to GL 2004-02. Enclosure 3 to this letter contains the CNP proprietary final response to GL 2004-02. Enclosure 4 to this letter contains the Westinghouse affidavit for withholding proprietary information.

PROPRIETARY INFORMATION Enclosure 3 to this letter contains proprietary information.

Withhold from public disclosure under 10 CFR 2.390.

Upon removal of Enclosure 3 this Letter is decontrolled.

U. S. Nuclear Regulatory Commission AEP-NRC-2022-03 Page 2 contains information proprietary to Westinghouse Electric Company LLC

("Westinghouse"), and it is supported by an Affidavit signed by Westinghouse (Enclosure 4), the owner of the information. The Affidavit sets forth the basis on which the information may be withheld from public disclosure by the Nuclear Regulatory Commission ("Commission") and addresses with specificity the considerations listed in paragraph (b)(4) of Section 2.390 of the Commission's regulations. Accordingly, it is respectfully requested that the information which is proprietary to Westinghouse be withheld from public disclosure in accordance with 10 CFR Section 2.390 of the Commission's regulations.

There are no new regulatory commitments made in this letter. Should you have any questions, please contact Mr. Michael K. Scarpello, Regulatory Affairs Director, at (269) 466-2649.

Sincerely, Q.t~Jt Site Vice President Indiana Michigan Power Company DLW/mll

Enclosures:

1. Affirmation
2. Final Supplemental Response to GL 2004-02 (Non-Proprietary)
3. Final Supplemental Response to GL 2004-02 (Proprietary)
4. Affidavit of Withholding Pursuant to 10 CFR 2.390, Westinghouse Electric Company c: R. J. Ancona - MPSC EGLE - RMD/RPS J. B. Giessner - NRC Region Ill NRC Resident Inspector R. M. Sistevaris - AEP Ft. Wayne, w/o enclosures J. E. Walcutt - AEP Ft. Wayne, w/o enclosures S. P. Wall - NRC Washington, D.C.

J. Williamson - AEP Ft. Wayne, w/o enclosures PROPRIETARY INFORMATION Enclosure 3 to this letter contains proprietary information.

Withhold from public disclosure under 10 CFR 2.390.

Upon removal of Enclosure 3 this Letter is decontrolled.

Enclosure 1 to AEP-NRC-2022-03 AFFIRMATION I, Q. Shane Lies, being duly sworn, state that I am the Site Vice President of Indiana Michigan Power Company (l&M), that I am authorized to sign and file this request with the U. S. Nuclear Regulatory Commission on behalf of l&M, and that the statements made and the matters set forth herein pertaining to l&M are true and correct to the best of my knowledge, information, and belief.

Indiana Michigan Power Company Site Vice President SWORN TO AND SUBSCRIBED BEFORE ME THIS d0-M DAY OF ~O uci.ry 2022 ddiary~

My Commission Expires

--?>":-c"t/?:

.I . ----

Enclosure 2 to AEP-NRC-2022-03 D.C. Cook Final Supplemental Response to GL 2004-04 Non - Proprietary to AEP-NRC-2022-03 Page 1 Table of Contents 1.0 Overall Compliance 1.1 Overview of D.C. Cook Resolution to GL 2004-02 1.2 Correspondence Background 1.3 General Plant System Description 1.4 General Description of Containment Recirculation Sump Strainers 2.0 General Description and Schedule for Corrective Actions 3.0 Specific Information for Review Areas 3.n Downstream Effects - Fuel and Vessel 3.o Chemical Effects 3.p Licensing Basis 4.0 References Non - Proprietary to AEP-NRC-2022-03 Page2 1.0 Overall Compliance NRC Issue:

Provide information requested in GL 2004-02, "Requested Information." Item 2(a) regarding compliance with regulations. That is, provide confirmation that the [Emergency Core Cooling System (ECCS)] EGGS and [Containment Spray System (CSS)] GSS recirculation functions under debris loading conditions are or will be in compliance with the regulatory requirements listed in the Applicable Regulatory Requirements section of this generic letter. This submittal should address the configuration of the plant that will exist once all modifications required for regulatory compliance have been made and this licensing basis has been updated to reflect the results of the analysis described above.

Indiana Michigan Power Company <l&M) Response:

In accordance with SECY-12-0093 (Reference 1) and as identified in l&M letter to the Nuclear Regulatory Commission (NRC) dated May 15, 2013 (Reference 2), D.C. Cook Nuclear Plant (CNP) Units 1 and 2 elected to pursue GSl-191 Closure Option 2 - Deterministic and identified in-vessel downstream effects as the last outstanding issue. Topical Report (TR) WCAP-17788-P, Rev. 1 (Reference 3) provides evaluation methods and results to address in-vessel downstream effects. As discussed in NRC "Technical Evaluation Report of In-Vessel Debris Effects,"

(Reference 9), the NRC staff has performed a detailed review of WCAP-17788-P. Although the NRC staff did not issue a Safety Evaluation for WCAP-17788, as discussed further in "U.S.

Nuclear Regulatory Commission Staff Review Guidance for In-Vessel Downstream Effects Supporting Review of Generic Letter 2004-02 Responses" (Reference 10), the staff expects that many of the methods developed in the TR can be used by pressurized-water reactor (PWR) licensees to demonstrate adequate long-term core cooling (LTCC). Completion of the analyses demonstrate compliance with 10 CFR 50.46, "Acceptance criteria for emergency core cooling systems for light-water nuclear power plants," (b)(5), "Long-term cooling," as it relates to in-vessel downstream debris effects for CNP.

1.1 Overview of CNP Resolution to GL 2004-02 On February 29, 2008 (Reference 4), August 29, 2008 (Reference 5), and May 26, 2010 (Reference 6), l&M submitted Supplemental Responses to GL 2004-02 for CNP which summarized changes that were implemented to support the resolution of GSl-191. Additionally, on May 19, 2011 (Reference 14), l&M submitted a letter to the NRC about the discovery of Temp-Mat fibrous insulation in the Unit 1 and Unit 2 containments. On December 18, 2013 (Reference 16), l&M submitted the disposition of the Temp-Mat fibrous insulation.

The changes that were implemented and identified in the CNP Supplemental Responses remain valid and no further changes are being made to address in-vessel downstream effects. The Temp-Mat was largely removed from the Unit 1 and Unit 2 containments. The remaining Temp-Mat is included in the fibrous debris source term included in this evaluation.

Non - Proprietary to AEP-NRC-2022-03 Page3 The CNP Supplemental Responses to GL 2004-02 also identified some of the conservatisms in the approach to resolving GSl-191. The conservatisms identified in the CNP Supplemental Responses remain valid and some of the conservatisms in the in-vessel downstream effects analysis are as follows:

  • The amount of fibrous debris calculated to penetrate the sump strainer is based on the Nuclear Energy Institute (NEI) clean plant criteria. Compared to strainer penetration testing results, the 45% bypass fraction used in the clean plant criteria is conservative and results in more fibrous debris penetrating the sump strainer than would be expected.
  • Industry fuel assembly testing found that a low particulate-to-fiber mass ratio ( 1: 1) resulted in the most limiting fibrous debris loads for the limiting hot leg cases. At higher particulate-to-fiber ratios, the fiber limit increases. Since CNP has a small quantity of fibrous debris and larger quantities of particulate debris, particulate-to-fiber ratio is expected to be higher than 1: 1. This increase in particulate-to-fiber ratio would be expected to increase the allowable fiber load per fuel assembly above the WCAP-17788-P limit.
  • The assumed 30 pounds of latent fiber is conservatively high. Plant walk-downs showed that there is very little latent fibrous debris in containment. The walk-downs showed a bounding value of 12 pounds of debris could be used for both units.
  • The debris generation calculation assumed that the Temp-Mat exists in 1.5-inch (in) strips between sections of Calcium Silicate (Cal-Sil). In reality, the gaps between Cal-Sil pieces are much smaller. Considering that the Cal-Sil was installed by skilled tradespeople, it is reasonable to assume that the gaps between pieces would be less than 0. 75 in.
  • All of the Temp-Mat within the Zone of Influence (201) is assumed to be destroyed as fines.

In reality, the Loss of Coolant Accident (LOCA) jet would blow some of the pieces off the pipe intact. The intact pieces would still be subject to erosion from spray and from the pool, but the available amount of fines would be reduced.

  • The as-manufactured density of Temp-Mat is used to calculate the mass of Temp-Mat debris.

Temp-Mat is a blanketed fibrous insulation. Removing the blanket and cutting the insulation into strips to stuff between pieces of Cal-Sil would likely result in a reduction in its density, which would reduce the actual mass of Temp-Mat debris.

  • As fiber accumulates on the strainers, the filtering efficiency of the strainers increases and the penetration fraction therefore decreases. Using a constant penetration fraction is conservative compared to a time-dependent penetration fraction that asymptotically approaches zero.
  • No credit is taken for the diversion of any fibrous debris through the containment spray (CTS) system and back into containment, where it could be held up, settle, or be filtered out as it passed through the Recirculation Sump Strainers again.
  • No credit is taken for diversion of any fibrous debris through alternate flow paths to the core.

All fibrous debris that reaches the reactor vessel is assumed to arrive at the core inlet.

Non - Proprietary to AEP-NRC-2022-03 Page4 1.2 Correspondence Background The following provides a listing of correspondence issued by the NRC or submitted by l&M for CNP, on GL 2004-02:

Generic Letter 2004-02 Correspondences ADAMS Accession Document Date Document Number September 13, 2004 ML042360586 NRC GL 2004-02 March 4, 2005 ML050750069 First Response to GL 2004-02 August31,2005 ML052510512 Information Requested by September 1, 2005 December 19, 2005 ML060030459 Revision to Commitments Februarv 9, 2006 ML060370547 First NRC Reauest for Additional Information (RAI)

June 27, 2006 ML061860257 First RAI Response Julv 27, 2006 ML061860257 Completion Date Extension Request July 28, 2006 ML062020768 NRC Acceptance of Extension Request December 19, 2006 ML063610088 Revision to Commitments June 27, 2007 ML072000387 Second Response to GL 2004-02 June 27, 2007 ML071910354 License Amendment Request (LAR) to Revise Technical Specifications (TS) for GL 2004-02 August 8, 2007 ML072210017 NRC RAls for LAR September 21, 2007 ML072750687 LAR RAI Response October 18, 2007 ML072780605 Issuance of Amendments November 21, 2007 ML073110389 NRC Revised Content Guide December 7, 2007 ML073470140 Completion Date Extension Reauest December 26, 2007 ML073540189 NRC Acceptance of Extension Request Februarv 29, 2008 ML080770394 Suoolemental Response to GL 2004-02 August29,2008 ML082520025 Final Response to GL 2004-02 June 18, 2009 ML091490421 NRC RAls May 26, 2010 ML101540527 Updated Final Response to GL 2004-02 July 27, 2010 ML101960128 Staff Comments on Suoolemental Responses May 19, 2011 ML11147A072 Disposition of Temp-Mat Fibrous Insulation Mav 15, 2013 ML13137A046 Path Forward for Resolution of GSl-191 December 18, 2013 ML13358A009 Final Disposition of Temp-Mat Fibrous Insulation October 27, 2016 ML16302A395 Commitment Schedule Change 1.3 General Plant System Description The CNP ice condenser containment consists of four uniquely defined and separated volumes:

1) upper containment, 2).ice condenser, 3) lower containment, and 4) reactor cavity. Refer to of Reference 4, Figures A4-2 through A4-10 for illustrations of various views of lower containment and a plan view of upper containment.

The upper containment area (Reference 4 - Figure A4-2), which does not contain any high energy piping, is physically separated from the lower compartment by the divider barrier and the ice condenser.

The ice condenser forms an approximate 300° arc around containment between the containment wall and the crane wall. The ice condenser has 24 paired doors in the lower containment area that will open following a pipe break allowing for suppression of the initial pressure surge in containment. There are also doors just above the ice bed and at the top of the ice condenser section to allow steam and non-condensable gases to vent to the upper containment volume.

Non - Proprietary to AEP-NRC-2022-03 Page 5 The lower containment volume contains both the loop compartment inside the crane wall and the annulus area between the crane wall and the containment wall. The crane wall that separates these two regions is 3 feet (ft) thick. There are ventilation openings in the crane wall which are above the maximum flood elevation of containment. These ventilation openings provide for the supply of cooled air to the loop compartment from the containment lower ventilation units. Within the loop compartment above nominal elevation 648 ft are the steam generator and pressurizer enclosures. These enclosures utilize the crane wall as one part of the enclosure with cylindrical concrete walls forming the rest of the enclosures. Each of these enclosures has a concrete roof, the top of which is at-nominal elevation 695 ft. The cylindrical wall sections and the roof comprise the portion of the divider barrier separating the lower containment from the upper containment.

The loop compartment is surrounded on its outside perimeter by the crane wall. The primary shield wall and refueling cavity walls are on the inside perimeter of the loop compartment. The nominal distance from the primary shield wall to the crane wall varies from 22 ft to 23 ft. The nominal distance from the crane wall to the containment wall is 13 ft.

The final volume, the reactor cavity, is the volume that is below (the lower reactor cavity), above (the upper reactor cavity), and around the reactor vessel (annular area). The upper reactor cavity is bounded by the primary shield wall, the vertical bulkheads, and the control rod drive mechanism missile shields. The primary communication path between the lower reactor cavity and lower containment via the overflow wall exists only after water level in either of the volumes exceeds the 610 ft elevation. This level is approximately 11.2 ft above the lower containment floor (where the recirculation sump strainers are located) and approximately 42.3 ft above the lower reactor cavity floor. A secondary communication path exists between the loop compartment and the lower reactor cavity. This path is through the sleeves in the primary shield wall that contain the hardware to position the ex-core nuclear instrumentation in the operating or maintenance positions. The containment liner is attached to the exterior containment wall concrete.

1.4 General Description of Containment Recirculation Sump Strainers As stated in CNP Supplemental Response dated February 29, 2008 (Reference 4), the CNP containment recirculation sump strainers are pocket type strainers designed and manufactured by Control Components Inc. (CCI). The CCI strainer integrally combines the screen and trash rack functions for the recirculation sump. The reinforced front of the new CCI strainer functions as the trash rack. Recessed pockets provide the strainer function. The pockets are made of perforated stainless steel plate. The nominal size of the strainer openings is 1/12 in (2.1 mm).

Each pocket is integrally connected and reinforced into larger units. A CCI main strainer was installed at the face of both the Unit 1 and Unit 2 recirculation sumps.

A single new remote strainer assembly is located in the Annulus area in Quadrant 2 between Azimuth 105° and 120°. The strainer is also designed and manufactured by CCI. The remote strainer is also a pocket type design. The strainer is anchored to the containment floor at El. 598'-9 3/8". The physical design of the remote strainer is customized to fit the space available.

The design maintains the same general shape of the pockets and the same bypass criteria as the main strainer. The pockets are made of perforated stainless steel plate. The nominal size of the strainer openings is 2.1 mm. Each pocket is integrally connected and reinforced into larger units.

The water from the containment inside the crane wall is able to flow into the annulus to supply water to the remote strainer. The water from the containment spray in the annulus as well as the water that flows into the annulus through the flood up overflow wall openings is strained by the Non - Proprietary to AEP-NRC-2022-03 Page6 remote strainer and ducted by a waterway into the recirculation sump. The waterway enters the recirculation sump through an opening in the crane wall.

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

Containment Recirculation Sump Strainer Surface Area Strainer Surface Area (ft2)

Unit 1 Main Strainer 900 ft 2 Unit 2 Main Strainer 900 ft 2 Unit 1 Remote Strainer 1072 ft2 Unit 2 Remote Strainer 1072 ft2 2.0 General Description and Schedule for Corrective Actions NRC Issue:

Provide a general description of actions taken or planned, and dates for each. For actions planned beyond December 31, 2007, reference approved extension requests or explain how regulatory requirements will be met as per nRequested lnformationn Item 2(b). That is provide a general description of and implementation schedule for all corrective actions, including any plant modifications, that you identified while responding to this generic letter.

Efforts to implement the identified actions should be initiated no later than the first refueling outage starting after April 1, 2006. All actions should be completed by December 31, 2007.

Provide justification for not implementing the identified actions during the first refueling outage starting after April 1, 2006. If all corrective actions will not be completed by December 31, 2007, describe how the regulatory requirements discussed in the Applicable Regulatory Requirements section will be met until the corrective actions are completed.

l&M Response:

l&M has performed analyses to determine the susceptibility of the ECCS and CTS recirculation functions for CNP to the adverse effects of post-accident debris blockage and operation with debris-laden fluids. These analyses conform, to the greatest extent practical, to the NEI 04-07 methodology (Reference 8) as approved by the NRC Safety Evaluation (SE) dated December 6, 2004 (Reference 15). As of the submittal of this evaluation, l&M has completed the following GL 2004-02 actions, analyses and modifications:

  • Replaced simple geometry strainers with complex geometry strainers having a filtering surface area of 1972 ft 2 and nominal 2.1 - 2.4 mm circular openings
  • Modified recirculation sump vents to ensure debris would not affect vent function, and to ensure debris larger than strainer openings could not enter the sump via vent
  • Added new safety-related level instruments inside the recirculation sump to provide indication and alarm in the control room in the event of a low water level inside the sump
  • Installed debris interceptors in multiple locations to prevent debris from impeding the flow of water through containment or the function of level instruments Non - Proprietary to AEP-NRC-2022-03 Page 7
  • Isolated the lower containment sump from the recirculation sump to prevent debris from traveling from the former to the latter
  • Removed internals from Containment Equalization (CEQ) fan room drain lines to ensure drainage of CTS water from upper containment to lower containment
  • Removed Cal-Sil (Calcium Silicate) and fiberglass insulation and numerous tags and labels from containment
  • Performed strainer head loss and chemical effects testing
  • Performed latent debris sampling and characterization, including other debris sources
  • Completed debris generation and debris transport analyses
  • Completed ex-vessel downstream effects analysis
  • Completed Net Positive Suction Head (NPSH) analysis
  • Established programmatic and procedural changes to maintain acceptable configuration
  • Performed evaluation of In-vessel Downstream Effects - including Temp-Mat debris l&M has no outstanding corrective actions associated with GL 2004-02 for CNP Unit 1 and Unit 2.

3.0 Specific Information for Review Areas As shown in CNP responses dated February 29, 2008 (Reference 4), August 29, 2008 (Reference 5), and May 26, 2010 (Reference 6), CNP has addressed review areas 3.a through 3.m, and only the outstanding review areas 3.n through 3.p are addressed in this submittal.

3.n Downstream Effects - Fuel and Vessel NRC Issue:

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

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

Non - Proprietary to AEP-NRC-2022-03 Page 8 l&M Response:

TR WCAP-17788-P, Rev. 1 (Reference 3) provides evaluation methods and results to address in-vessel downstream effects. As discussed in NRC "Technical Evaluation Report of In-Vessel Debris Effects," (Reference 9), the NRC staff has performed a detailed review of WCAP-17788-P.

Although the NRC staff did not issue a Safety Evaluation for WCAP-17788-P, as discussed further in "U.S. Nuclear Regulatory Commission Staff Review Guidance for In-Vessel Downstream Effects Supporting Review of Generic Letter 2004-02 Responses" (Reference 10), the staff expects that many of the methods developed in the TR may be used by PWR licensees to demonstrate adequate LTCC. l&M used methods and analytical results developed in WCAP-17788-P, Rev. 1 to address in-vessel downstream debris effects for CNP and has evaluated the applicability of the methods and analytical results from WCAP-17788-P, Rev. 1 for CNP.

3.n.1 Sump Strainer Fiber Penetration l&M had previously completed a plant specific strainer bypass test, documented in Reference 5.

Due to questions regarding the reliability of the testing method, l&M has elected not to utilize the test results. l&M has 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 CNP Unit 1 and Unit 2. The clean plant criteria, as applied to in-vessel effects, utilize a fiber penetration (bypass) fraction of 45% and a debris transport fraction of 75%.

Based on the clean plant criteria values for the debris transport fraction to the strainer (T) and the fiber penetration fraction (P), the following CNP Unit 1 and Unit 2 specific in-vessel debris load is determined using:

B M*T*P

-=---

FA N Where:

g/FA grams of fiber per fuel assembly M mass of fibrous debris (latent + generated from worst-case break) [grams]

T transport fraction to the strainer P strainer penetration fraction N Number of fuel assemblies The mass of latent fibrous debris for CNP is conservatively assumed to be 30 lbm. The worst case amount offibrous debris generated for CNP is 7.6 ft3 of Temp-mat. The density of Temp-mat is 11.8 lbm/ft3 which results in a worst case generation of approximately 90 lbm. The total mass of fibrous debris is therefore 120 lbm.

g (120 lbm x 453.6 g/lbm)

  • 0.75
  • 0.45 FA= 193 = 95 g/FA This is the CNP specific in-vessel fiber load that will be compared to the applicable WCAP-17788-P, Rev. 1 in-vessel debris acceptance criterion, which assumes that all fibrous debris calculated to penetrate the strainer will reach the reactor core.

Non - Proprietary to AEP-NRC-2022-03 Page 9 3.n.2 Applicability to WCAP-17788 Methods and Analysis Results CNP Unit 1 and Unit 2 are Westinghouse 4-loop Pressurized Water Reactors with a converted upflow barrel/baffle configuration. Per Section 3.0 of the NRC Staff Review Guidance (Reference 10), it is necessary to confirm that CNP Unit 1 and Unit 2 are within the key parameters of the WCAP-17788-P, Rev. 1 methods and analysis. Each of the key parameters is discussed below.

3.n.3 Fuel Design CNP Unit 1 uses Westinghouse 15x15 Upgrade Fuel Assembly (UFA).

CNP Unit 2 uses Westinghouse 17x17 Optimized Fuel Assembly (OFA).

3.n.4 WCAP-17788 debris limit The Proprietary total in-vessel (core inlet and heated core) fibrous debris limit contained in Section 6.5 of WCAP-17788-P Volume 1, Rev. 1 applies to CNP Unit 1 and Unit 2.

3.n.5 Methodology used to calculate the fibrous debris amounts As described in Section 3.n.1 of this submittal, CNP assumes that all fibrous debris calculated to penetrate the strainer reaches the reactor vessel. This consists of latent debris and accident generated Temp-Mat debris.

3.n.6 Confirm maximum combined amount of fiber that may arrive at the core inlet and heated core for hot leg break is below the WCAP-17788 fiber limit As shown in the sump strainer fiber penetration section, the CNP maximum amount of fiber calculated to potentially reach the reactor vessel is 95 g/FA, which is less than the proprietary in-vessel fibrous debris limit provided in Section 6.5 of WCAP-17788-P Volume 1, Rev. 1.

3.n.7 Confirmation that the core inlet fiber amount is less than the WCAP-17788-P, Rev. 1 threshold CNP is a Westinghouse 4-loop design with Westinghouse fuel. The applicable WCAP-17788-P, Rev. 1 core inlet fiber threshold is provided in Table 6-3 of WCAP-17788-P, Rev. 1. The core inlet fiber amount for CNP is calculated to be 95 g/FA. While this exceeds the core inlet fiber limit, it is less than the total in-vessel debris limit. The analyses in WCAP-17788-P as well as the earlier work documented in WCAP-16793, conservatively assumed that debris would collect uniformly at the core inlet because that results in the greatest head-loss for a given amount of debris.

However, the debris bed is realistically expected to collect non-uniformly. As a result, the amount of debris required to completely block the core inlet would be greater than that assumed in the analyses (Reference 9). Therefore, the core inlet fiber limit is not a critical parameter and the current LTCC analyses remain applicable. The total core fiber limit in WCAP-17788-P can be used to justify sufficient margin for long term core cooling.

In the unlikely scenario where the core inlet did become completely blocked, alternate flow paths (AFP) will allow flow to reach the core. l&M calculated a plant specific AFP resistance, which is Non - Proprietary to AEP-NRC-2022-03 Page 10 similar to the generic resistances calculated in WCAP-17788-P. The CNP AFPs would be effective to maintain core cooling.

3.n.8 Confirmation that the earliest sump switchover (SSO) time is 20 minutes or greater The earliest possible SSO time for CNP is 23.14 minutes. This value bounds both Unit 1 and Unit 2.

3.n.9 Predicted chemical precipitation timing from WCAP-17788-P. Rev. 1. Volume 5 testing and the specific test group considered to be representative of the plant Chemical precipitation timing is dependent on the plant buffer, sump pool pH, volume and temperature, and debris types and quantities. Table 1 summarizes the key chemical precipitation parameters and values for CNP and compares them to test group 26 from WCAP-17788-P, Rev.

1, Volume 5. Based on the comparison in Table 1, test group 26 is representative of CNP 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 />.

Table 1 Key Parameter Values for Chemical Precipitation Timing Parameter CNP Value Test Group 26 Value Buffer NaOH NaOH pH 8.91 8.91 Minimum Sump Volume (ft3 ) 83,995 46,565 Max Sump Pool Temperature 190 190 (OF)

CalSil (ft3 ) 35.579 35.862 E-glass (ft3 ) 12.5 12.5 Silica (ft3 ) 0.1 0.4 Mineral Wool (ft3 ) 0 0 Al Silicate (ft3 ) 0 0 Concrete (ft3 ) 7,489.88 7,500 lnteram (ft3 ) 0 0 Al (ft2 ) 8,107.32 8,107.32 Galvanized Steel (ft2 ) 71,162.62 174,974 Non - Proprietary to AEP-NRC-2022-03 Page 11 3.n.10 Confirmation that chemical effects will not occur earlier than latest time to implement boric acid precipitation (BAP) mitigation measures As described in UFSAR Section 6.2.2, CNP performs injection realignment to mitigate the potential for boric acid precipitation no later than 7 .5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> following an accident, which is less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

3.n.11 WCAP-17788 tb1ock value for the RCS design category CNP Units 1 and 2 are Westinghouse 4-loop PWRs with a converted upflow barrel/baffle configuration. Based on WCAP-17788-P, Rev. 1, Volume 1, Table 6-1, tb1ock for CNP Unit 1 and Unit 2 is 143 minutes.

3.n.12 Confirmation that chemical effects do not occur prior to tb1ock The earliest time of chemical precipitation for CNP was determined to be 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 tb1ock value of 143 minutes.

3.n.13 Plant rated thermal power compared to the analyzed power level for the RCS design category CNP Unit 1 has a rated thermal power of 3304 MW1. CNP Unit 2 has a rated thermal power of 3468 MW1. CNP Unit 1 and Unit 2 are Westinghouse 4-loop PWRs with a converted upflow barrel/baffle configuration and the applicable analyzed thermal power is 3658 MW1 as provided in WCAP-17788-P, Rev. 1, Volume 4, Table 6-1. The CNP rated thermal power is less than the analyzed power; therefore, this parameter is bounded by the WCAP-17788-P, Rev. 1 alternate flow path analysis.

3.n.14 Plant alternate flow path (AFP) resistance compared to the analyzed AFP resistance for the plant RCS design category CNP Units 1 and 2 are Westinghouse converted upflow barrel/baffle plants. The Proprietary analyzed AFP resistance is provided in Table 6-1 of WCAP-17788-P Volume 4, Rev. 1. The Proprietary CNP-specific AFP resistance provided in Table RAl-4.2-24 of WCAP-17788-P Volume 4, Rev. 1 reflects the plants in a downflow plant configuration. A CNP-specific AFP resistance of

[ ]a,c was calculated for the barrel/baffle region to reflect the converted upflow plant configuration. As expected, the CNP specific AFP resistance is a similar magnitude as other Westinghouse 4-loop converted upflow plants provided in Table RAl-4.2-24 of WCAP-17788-P Volume 4, Rev. 1. The CNP AFP resistance is bounded by the resistance applied to the AFP analysis.

3.n.15 Consistency between the minimum ECCS flow per FA assumed in the AFP analyses and that at the plant CNP Units 1 and 2 are Westinghouse converted upflow barrel/baffle plant. The AFP analysis for Westinghouse upflow plants analyzed a range of ECCS recirculation flow rates from 8 - 40 gpm/FA, as shown in Table 6-1 of WCAP-17788-P Volume 4, Rev. 1. The minimum CNP ECCS recirculation flow rate is 15.5 gpm/FA, and the maximum ECCS recirculation flow rate is 39.4 gpm/FA. These flow are within the range of ECCS recirculation flow rates considered in the AFP analysis.

Non - Proprietary to AEP-NRC-2022-03 Page 12 3.n.16 Summary The comparison of key parameters used in the WCAP-17788-P AFP analysis to the CNP specific values is summarized in Table 2. Based on these comparisons CNP Unit 1 and Unit 2 are bounded by the key parameters and the WCAP-17788-P methods and results are applicable.

Table 2 Key Parameter Values for In-Vessel Debris Effects WCAP-17788 Parameter CNP Value Evaluation Value Maximum Total In- Maximum in-vessel fiber load is Volume 1 Vessel Fiber Load 95 less than WCAP-17788 limit.

Section 6.5 (g/FA)

Non-uniform debris Maximum Core Inlet Volume 1 95 accumulation at the core inlet is Fiber Load (g/FA) Table 6-3 credited to justify LTCC.

Later switchover time results in a lower decay heat at the time Minimum Sump 20 23.14 of debris arrival, reducing the Switchover Time (min) potential for debris induced core uncovery and heatup.

Potential for complete core inlet Minimum Chemical blockage due to chemical 2.4 (tblock) 24 (t:hem)

Precipitate Time (hr) product generation would occur much later than assumed.

Latest hot leg switchover occurs Maximum Hot Leg 24 (t:hem) 7.5 well before the earliest potential Switchover Time (hr) chemical product generation.

Rated Thermal Power U1 - 3304 Lower rated thermal power 3658 results in lower decay heat.

(MW,) U2- 3468 CNP specific AFP resistance is less than the Maximum AFP Volume 4 value calculated analyzed value, which increases Resistance Table 6-1 for the upflow the effectiveness of the AFP.

conversion Minimum ECCS Maximum debris bed resistance Recirculation Flow 8 15.5 at the core inlet occurs at lower (gpm/FA) flow rates.

Non - Proprietary to AEP-NRC-2022-03 Page 13 3.o Chemical Effects NRC Issue:

The objective of the chemical effects section is to evaluate the effect that chemical precipitates have on head loss and core cooling.

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

l&M Response:

l&M has completed the chemical effects analysis of the sump strainers for CNP Units 1 and 2 which was submitted in Supplemental Response dated February 29, 2008 (ML080770394), and subsequent responses submitted on August 29. 2008 (ML082520025), and May 26, 2010 (ML101540527). The CNP sump strainer chemical effects analysis is unchanged.

The CNP in-vessel chemical effects analysis is described in Sections*3.n.9 through 3.n.12.

3.p Licensing Basis NRC Issue:

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

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

l&M Response:

The CNP Unit 1 and Unit 2 licensing basis will be changed in accordance with the requirements of 10 CFR 50.71(e) to incorporate the GL 2004-02 response including the values of analyzed debris limits needed for TSTF-567 following receipt of NRC acceptance of this final supplemental response to the GL. No licensing actions or exemptions were needed to support changes to the plant licensing basis.

The CNP Unit 1 and Unit 2 UFSAR will be updated upon receipt of the NRC closure of GL 2004-02 for CNP Unit 1 and Unit 2.

4. References
1. SECY-12-0093 (ML121310648), "Closure Options for Generic Safety Issue - 191, Assessment of Debris Accumulation on Pressurized-Water Reactor Sump Performance,"

July 9, 2012.

Non - Proprietary to AEP-NRC-2022-03 Page 14

2. Letter from J.P. Gebbie (l&M) to US NRC (ML13137A046, AEP-NRC-2013-45), Donald C.

Cook Nuclear Plant Unit 1 and 2 Path Forward for Resolution of GSl-191, May 15, 2013.

3. WCAP-17788-P, Revision 1, Comprehensive Analysis and Test Program for GSl-191 Closure (PA-SEE-1090)," December 2019.
4. Letter from M.A. Peifer (l&M) to US NRC (ML080770394, AEP:NRC:8054-02), "Donald C.

Cook Nuclear Plant Units 1 and 2 Supplemental Response to Nuclear Regulatory Commission Generic Letter 2004-02: Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized Water Reactors,"

Feb 29, 2008.

5. Letter from L.J. Weber (l&M) to US NRC (ML082520025, AEP-NRC-2008-17), "Donald C.

Cook Nuclear Plant Units 1 and 2 Final Response to Nuclear Regulatory Commission Generic Letter 2004-02: Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized Water Reactors and Associated Request for Additional Information," Aug 29, 2008.

6. Letter from J.P. Gebbie (l&M) to US NRC (ML101540527, AEP-NRC-2010-39), "Donald C. Cook Nuclear Plant Units 1 and 2 Updated Final Response to Nuclear Regulatory Commission Generic Letter 2004-02: Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized Water Reactors and a June 2009 Request for Additional Information," May 26, 2010.
7. Letter from J.N. Jensen (l&M) to US NRC (ML061860257, AEP:NRC:6054-05), "Donald C. Cook Nuclear Plant Units 1 and 2 Update to Response to Nuclear Regulatory Commission Generic Letter 2004-02: Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized Water Reactors,"

June 27, 2006.

8. NEI 04-07, "Pressurized Water Reactor Sump Performance Evaluation Methodology, December 2004.
9. Letter from V.G. Cusumano US NRC to M. Gavrilas US NRC (ML19178A252), "Technical Evaluation Report of In-Vessel Debris Effects," June 2019.
10. Letter from V.G. Cusumano US NRC to J.R. Marshall US NRC (ML19228A011), "U.S.

Nuclear Regulatory Commission Staff Review Guidance for In-Vessel Downstream Effects Supporting Review of Generic Letter 2004-02 Responses," September 2019.

11. WCAP-16793, Revision 2, Evaluation of Long-Term Cooling Considering Particulate, Fibrous and Chemical Debris in the Recirculation Fluid," July 2013.
12. TSTF-567, Revision 1 (ML17214A813), "Add Containment Sump TS to Address GSl-191 Issues," August 2017.
13. NRC Generic Letter 2004-02 (ML042360586), "Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized Water Reactors,"

September 13, 2004.

Non - Proprietary to AEP-NRC-2022-03 Page 15

14. Letter from J.P. Gebbie (l&M) to US NRC (ML11147A072, AEP-NRC-2011-31 ), "Donald C.

Cook Nuclear Plant Units 1 and 2 Disposition of Temp-Mat Fibrous Insulation in Unit 1 and Unit 2 Containments with Respect to Generic Letter 2004-02 Concerns," May 19, 2011.

15. Letter from S.C. Black US NRC to A.R. Pietrangelo NEI (ML043280631), "Pressurized Water Reactor Containment Sump Evaluation Methodology," December 6, 2004.
16. Letter from J.P. Gebbie (l&M) to US NRC (ML13358A009, AEP-NRC-2013-94), "Donald C. Cook Nuclear Plant Units 1 and 2 Final Disposition of Temp-Mat Fibrous Insulation in Unit 1 and Unit 2 Containments," December 18, 2013.

Non - Proprietary

Enclosure 4 to AEP-NRC-2022-03 Affidavit of Withholding Pursuant to 10 CFR 2.390, Westinghouse Electric Company Non - Proprietary

Westinghouse Non-Proprietary Class 3 AFFIDAVIT CA W-22-004 Pagel of3 Commonwealth of Pennsylvania:

County of Butler:

(1) I, Camille Zozula, Manager, Regulatory Compliance and Corporate Licensing, have been specifically delegated and authorized to apply for withholding and execute this Affidavit on behalf of Westinghouse Electric Company LLC (Westinghouse).

(2) I am requesting the proprietary portions of AEP-NRC-2022-03 be withheld from public disclosure under 10 CFR 2.390.

(3) I have personal knowledge of the criteria and procedures utilized by Westinghouse in designating information as a trade secret, privileged, or as confidential commercial or financial information.

(4) Pursuant to 10 CFR 2.390, the following is furnished for consideration by the Commission in determining whether the information sought to be withheld from public disclosure should be withheld.

(i) The information sought to be withheld from public disclosure is owned and has been held in confidence by Westinghouse and is not customarily disclosed to the public.

(ii) The information sought to be withheld is being transmitted to the Commission in confidence and, to Westinghouse's knowledge, is not available in public sources.

(iii) Westinghouse notes that a showing of substantial harm is no longer an applicable criterion for analyzing whether a document should be withheld from public disclosure. Nevertheless, public disclosure of this proprietary information is likely to cause substantial harm to the competitive position of Westinghouse because it would enhance the ability of competitors to provide similar technical evaluation justifications and licensing defense services for commercial power reactors without commensurate expenses. Also, public disclosure of the information would enable others to use the information to meet NRC requirements for licensing documentation without purchasing the right to use the information.

Westinghouse Non-Proprietary Class 3 AFFIDAVIT CAW-22-004 Page 2 of3 (5) Westinghouse has policies in place to identify proprietary information. Under that system, information is held in confidence if it falls in one or more of several types, the release of which might result in the loss of an existing or potential competitive advantage, as follows:

(a) The information reveals the distinguishing aspects of a process (or component, structure, tool, method, etc.) where prevention of its use by any of Westinghouse's competitors without license from Westinghouse constitutes a competitive economic advantage over other companies.

(b) It consists of supporting data, including test data, relative to a process (or component, structure, tool, method, etc.), the application of which data secures a competitive economic advantage (e.g., by optimization or improved marketability).

(c) Its use by a competitor would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing a similar product.

(d) It reveals cost or price information, production capacities, budget levels, or commercial strategies of Westinghouse, its customers or suppliers.

( e) It reveals aspects of past, present, or future Westinghouse or customer funded development plans and programs of potential commercial value to Westinghouse.

(f) It contains patentable ideas, for which patent protection may be desirable.

(6) The attached documents are bracketed and marked to indicate the bases for withholding. The justification for withholding is indicated in both versions by means of lower-case letters (a) through (f) located as a superscript immediately following the brackets enclosing each item of information being identified as proprietary or in the margin opposite such information. These lower-case letters refer to the types of information Westinghouse customarily holds in confidence identified in Sections (5)(a) through (f) of this Affidavit.

Westinghouse Non-Proprietary Class 3 AFFIDAVIT CAW-22-004 Page 3 of3 I declare that the averments of fact set forth in this Affidavit are true and correct to the best of my knowledge, information, and belief. I declare under penalty of perjury that the foregoing is true and correct.

Executed on: 1/18/2022

~l=Oz::r Camille Zozula

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