W3F1-2008-0069, Final Supplemental Response to NRC Generic Letter 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors.

Final Supplemental Response to NRC Generic Letter 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors.

ML083020551
Person / Time
Site:	Waterford
Issue date:	10/23/2008
From:	Gaudet T Entergy Nuclear South, Entergy Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
GL-04-002, W3F1-2008-0069
Download: ML083020551 (96)

Text

Entergy Nuclear South Entergy Operations, Inc.

17265 River Road Killona, LA 70057-3093 Tel 504-739-6496 EnLI te gy Fax 504-739-6698 tgaudet@entergy.com Timothy J. Gaudet Acting Nuclear Safety Assurance Director Waterford 3 Attachment I Contains 10CFR2.390 Proprietary Information W3F17-2008-0069 October 23, 2008 U.S. Nuclear Regulatory Commission Attn:, Document Control Desk Washington, DC 20555-0001

Subject:

Final Supplemental Response to NRC Generic Letter 2004-02 "Potential Impact of Debris Blockage on Emergency Recirculation during Design Basis Accidents at Pressurized-Water Reactors" Waterford Steam Electric Station, Unit 3 (Waterford 3)

Docket No. 50-382 License No. NPF-38

References:

1. Generic Letter 2004-02, "Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors," dated September 13, 2004

2. Waterford 3 letter, "Request for Extension of: Completion date for Corrective Actions Required by Generic Letter 2004-02," dated November 14, 2007

3. NRC letter, "Approval of Extension Request for Corrective Actions; re:

GL 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation during Design Basis Accident at Pressurized Water Reactors," dated December 12, 2007

4. NRC letter, "Report on Results of Staff Audit of Corrective Actions to Address Generic Letter 2004-02," dated January 28, 2008

5. Waterford 3 letter, "Supplemental Response to NRC Generic Letter 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation during Design Basis Accidents at Pressurized-Water Reactors," dated February 29, 2008

6. Waterford 3 I6tter, "Request for Extension of Completion Date for Resolution of Generic Letter 2004-02," dated May 12, 2008

W3F1 -2008-0069 Page 2

Dear Sir or Madam:

The purpose of this letter is to provide notification of the completion of activities associated with Generic Letter (GL) 2004-02 (Reference 1). The U.S. Nuclear Regulatory Commission (NRC) issued Reference 1 to request that addressees perform an evaluation of the emergency core cooling system (ECCS) and containment spray system (CSS) recirculation functions in light of the information provided in the GL and, if appropriate, take additional actions to ensure system function. Additionally, the GL requested addressees to provide the NRC with a written response in accordance with 10 CFR 50.54(f). The request was based on identified potential susceptibility of the pressurized water reactor (PWR) recirculation sump screens to debris blockage during design basis accidents requiring recirculation operation of ECCS or CSS and on the potential for additional adverse effects due to debris blockage of flowpaths necessary for ECCS and CSS recirculation and containment drainage.

Reference 2 provided Waterford 3's (W-3) summary of actions taken which included the installation of a new sump strainer during the W-3 Fall 2006 refueling outage and the results of the downstream effects evaluation. Additionally, W-3 requested an extension until restart, following refueling outage 15, to complete the analysis needed to achieve compliance with GSI-1 91. The NRC extension was approved in Reference 3. In Reference 4, the NRC reported on staff evaluation of corrective actions to address GL. In Reference 5, W-3 provided the response to open items identified in Reference 4 and the supplemental response per NRC letter to NEI dated November 30, 2007. Additionally, W-3 committed to providing a final supplemental response in Reference 5. In Reference 6, W-3 requested an extension for providing the final supplemental response. This final supplemental response communicates the completion of all actions needed to address the GL. provides the details of the final supplemental response and includes proprietary information. The proprietary information was provided to Entergy in a GE Hitachi Nuclear Energy (GEH) transmittal that is referenced by an affidavit. GEH requests the enclosed proprietary information identified in Attachment 1 be withheld from public disclosure in accordance with the provisions of 10CFR 2.390 and 10CFR 9.17. Attachment 2 is the non-proprietary version of the response. Attachment 3 contains the affidavit for withholding the proprietary information contained in Attachment 1.

This letter contains no regulatory commitments.

Please contact me or Robert J. Murillo, Manager Licensing at (504) 739-6715 if there are any questions regarding this letter.

I declare under penalty of perjury that the foregoing is true and correct. Executed on October 23, 2008.

Siny,

W3F1-2008-0069 Page 3 Attachment(s): 1. Supplemental Response to NRC GL 2004-02 (Proprietary Information)

2. Supplemental Response to NRC GL 2003-02 (Non-Proprietary Information)

3. Affidavit

W3F1 -2008-0069 Page 4 (w/o Attachment 1)(w/Attachments 2 and 3) cc: Mr. Elmo E. Collins, Jr.

Regional Administrator U. S. Nuclear Regulatory Commission Region IV 612 E. Lamar Blvd., Suite 400 Arlington, TX 76011-4125 NRC Senior Resident Inspector Waterford Steam Electric Station Unit 3 P.O. Box 822 Killona, LA 70066-0751 U. S. Nuclear Regulatory Commission Attn: Mr. N. Kalyanam Mail Stop O-07D1 Washington, DC 20555-0001 Wise, Carter, Child & Caraway ATTN: J. Smith P.O. Box 651 Jackson, MS 39205 Winston & Strawn ATTN: N.S. Reynolds 1700 K Street, NW Washington, DC 20006-3817 Morgan, Lewis & Bockius LLP ATTN: T.C. Poindexter 1111 Pennsylvania Avenue, NW Washington, DC 20004

Attachment 2 W3F1-2008-0069 Supplemental Response to NRC GL 2004-02 (Non-Proprietary Information) to W3F1-2008-0069 Page 1 of 87 Table of Contents 1.0 Overall Com pliance .................................................................................................. 3 2.0 General Description of and Schedule for Corrective Actions ................................ 5 3.0 Specific Information Regarding Methodology for Determining Compliance ...... 7 3.a Break Selection ......................................................... . ..

.............................. 7 3.b Debris Generation/Zone of Influence (ZOI) (excluding coatings) ....................... 9 3.c Debris Characteristics .......................................................................................... 11 3.d Latent Debris ............................................................................................................. 13 3.e Debris Transport ................................................................................................... 13 3.f Head Loss and Vortexing ................................................................................... 13 3.g Net Positive Suction Head (NPSH) ............................ ......................................... 13 3.h Coatings Evaluation ............................................................................................ 13 3.i Debris Source Term .............................................................................................. 13 3.j Screen Modification Package .............................................................................. 13 3.k Sum p Structural Analysis ................................................................................... 13 3.1 Upstream Effects ................................................................................................. 13 3.m Downstream Effects - Components and Systems ............................................. 13 3.n Dow nstream Effects - Fuel and Vessel .............................................................. 13 3.0 Chem ical effects ........................................................................................................ 13 3.p Licensing Basis ......................................................................................................... 13 O pen Item s ................................................................................................................................ 13 References ................................................................................................................................. 13 to W3F1-2008-0069 Page 2 of 87 Acronyms AJIT Air Jet Impact Test OPG Ontario Power Generation BWR Boiling Water Reactor PWR Pressurized Water Reactor BWROG Boiling Water Reactor Owners PZR Presurizer Group RAS Recirculation Actuation Signal CFD Computational Fluid Dynamics RC Reactor Coolant CFR Code Federal Regulation RCP Reactor Coolant Pump CS Containment Spray RMI Reflective Metal Insulation CSS Containment Spray System RPV Reactor Pressure Vessel DBA Design Basis Accident RCB Reactor Containment Building DIR Design Input Record RWSP Refueling Water Storage Pool ECCS Emergency Core Cooling System SB LOCA Small Break Loss of Coolant GE General Electric Accident GL Generic Letter SDC Shut-down Cooling GR Guidance Report SE Safety Evaluation HELB High Energy Line Break SER Safety Evaluation Report HPSI High Pressure Safety Injection SG Steam Generator ID Internal Diameter SI Safety Injection IOZ Inorganic Zinc SIS Safety Injection System LB LOCA Large Break Loss of Coolant SIT Safety Injection Tank Accident SS Stainless Steel LOCA Loss of Coolant Accident SSC System, Structure, or Component LPSI Low Pressure Safety Injection TSP Tri-Sodium Phosphate MEl Metal Encapsulated Insulation UFSAR Updated Final Safety Analysis NEI Nuclear Energy Institute Report NRC Nuclear Regulatory Commission WF3 Waterford 3 NPSH Net Positive Suction Head ZOI Zone of Influence

'f-to W3F1-2008-0069 Page 3 of 87 This attachment provides Waterford 3's supplemental response to GL 2004-02 (Reference 1).

The supplemental response follows the format and guidance provided by the NRC in Reference

63. All text from Reference 63 is presented in italic script.

NRC Request, Summary-Level Description The GL supplemental response should begin with a summary-level description of the approach chosen. This summary should identify key aspects of design modifications, process changes, and supporting analyses that the 7icensee believes are relevant or important to the NRC staff's verification that corrective actions to address the GL are adequate. The summary should address significant conservatisms and margins that are used to provide high confidence the issue has been addressedeven with uncertaintiesremaining. Licensees should address commitments and/ordescriptionsof plant programs that support conclusions.

Summary-Level Description for Waterford 3 The key aspects of the approach chosen by Waterford 3 to resolve the concerns identified in GL 2004-02 are as follows:

• Replacement of existing screens for the Safety Injection sump. This increased total surface area of screens from approximately 200 ft 2 to approximately 3699 ft 2.

• Extensive testing and analysis to determine break locations, identify and quantify debris sources, quantify debris transport, determine downstream effects, determine chemical effects, and confirm recirculation function.

• Changes to plant programs, processes, and procedures to limit the introduction of materials into containment that could adversely impact the recirculation function, and establish monitoring programs to ensure containment conditions' will ,continue to support the recirculation function.

• Application of conservative measures to assure adequate margins throughout the actions taken to address the GL 2004-02 concerns.

Additional details regarding these key aspects are provided below:

Plant Modifications Waterford 3 has performed plant modifications as described below to address the concerns identified in GL 2004-02.

• The original box-like SI Sump screen that surrounded the sump itself was removed and replaced with General Electric Energy, Nuclear (GE) Modularized Stacked Disk Strainers.

The total surface area of the new SI Sump strainers is approximately 3699 ft 2 as compared to the original screen with a surface area of approximately 200 ft 2.

Nineteen TSP baskets were relocated to allow easier installation of the new plenum and strainers, or to eliminate interferences with the baskets.

• The screen partition thatseparated the two trains of the SI inlets was replaced with stainless steel grating.

• The low level switch inside the SI Sump was relocated to mount the housing on top of the new screen plenum.

to W3F1-2008-0069 Page 4 of 87

• The tubing for two level transmitters inside the SI Sump was rerouted to penetrate through the plenum in a designed location in order to prevent debris from passing through the penetration opening.

Testing Waterford 3 has completed extensive plant-specific testing of the SI Sump strainer design and the effects of chemical reactions on post-accident strainer head loss.

Conservatisms

• The ZOI for debris generation for all RCS cold leg breaks were based on the 42" diameter of the hot leg piping instead of the 30" diameter of the cold leg piping.

" WCAP-16710-P (Reference 38) showed that the ZOI for stainless steel jacketed NUKON insulation could be reduced to as low as 5D. However, Waterford 3 used a ZOI of 7D.

" Waterford 3 modeled 33% of all fibrous insulation within the ZOI as becoming fines or small debris. Appendix II of the SER for NEI 04-07 (Reference 3) suggested a value of 22% for small and fines debris.

, In the debris transport calculation 2005-05500 (Reference 29), the lift-over-curb velocity used was for a 6" high curb. However, the plenum on which the SI Sump strainers sit is 8" high.

• In all but one of the CFD scenarios in the debris transport calculation 2005-05500 (Reference 29), the approach velocity to the SI Sump strainers did not exceed the lift-over-curb velocity for a 6" curb at any location along the plenum perimeter. In one CFD scenario, less than 25% of the perimeter had velocities in excess of those necessary to lift over a 6" curb. For conservatism, 25% of the small debris is treated as lifting onto the sump strainers.

• All labels and tags are modeled with 100% transport to the screens. The total sacrificial area is calculated as equivalent to 100% of the original single sided surface area with 0%

overlap. As documented in the SER for NEI 04-07 (Reference 3), the sacrificial area could have been reduced to 75% of the total of the original single-sided surface area of the labels and tags.

• For the NPSH calculation ECM07-001 (Reference 51), the flow rates are based on both trains of ECCS and CS operating at pump run-out values. However, the temperature assumptions in containment are based on only one train of CS in operation at design flow, and ECCS at minimum flow.

• Waterford 3 assumed 50% of qualified coatings on the containment liner dome and the liner between elevations 112 and 138 would fail. This value far exceeds the total amount of coating failures throughout containment recorded at Waterford 3 in the most recent refueling outage.

1.0 Overall Compliance NRC Issue 1:

Provide information requested in GL 2004-02, "Requested Information." Item 2(a) regarding compliance with regulations.

GL 2004-02 Requested Information Item 2(a)

Confirmation that the ECCS and CSS recirculation functions under debris loading conditions are or will be in compliance with the regulatory requirements listed in the to W3F1-2008-0069 Page 5 of 87 Applicable Regulatory Requirements section of this GL. 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.

WF3 Response 1:

The recirculation functions for the SIS and the CSS for Waterford 3 are in compliance with the applicable Regulatory Requirements section of the subject generic letter under debris loading conditions upon the completion of Refuel 15 (RF15), the spring 2008 outage for Waterford 3.

The design packages EC-999 and EC-1002 update the Waterford 3 design basis associated with GS1-191 resolution and Generic Letter 2004-02 compliance. These EC's were approved by Entergy on October 23, 2008, along with associated Safety Analysis Report (SAR) changes provided to licensing. This submittal provides the final supplemental response for compliance with the regulatory requirements in GL 2004-02.

2.0 General Description of and Schedule for Corrective Actions NRC Issue 2:

Provide a general description of actions taken or planned, and dates for each. For actions planned beyond December 31, 2007, reference approved extension requests or explain how regulatory requirements will be met as per "Requested Information" Item 2(b). (Note: All requests for extension should be submitted to the NRC as soon as the need becomes clear, preferably not later than October 1, 2007.)

GL 2004-02 Requested Information Item 2(b)

A general description of and implementation schedule for all corrective actions, including any plant modifications, that you identified while responding to this GL. Efforts to implement the identified actions should be initiatedno later than the first refueling outage starting after April 1, 2006. All actions should be completed by December 31, 2007.

Providejustification 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.

WF3 Response 2:

The corrective actions to address the concerns identified in GL 2004-02 at Waterford 3 consist of plant modifications, testing and analysis, changes to plant programs and processes, and changesto the licensing basis. These changes are described below.

Plant Modifications Based on the results of debris generation and transport analyses, modifications to the existing SI Sump screens were required to meet the applicable Regulatory Requirements discussed in GL 2004-02 (Reference 1). The physical changes were performed during the Waterford 3 refueling outage RF14, in the fall of 2006. The changes are listed below:

to W3F1-2008-0069 Page 6 of 87

1. The original box-like SI Sump screen that surrounded the sump itself was removed and replaced with General Electric Energy, Nuclear (GE) Modularized Stacked Disk Strainers. Due to the amount of screen area required to be installed to meet the requirements of GL 2004-02 (Reference 1), a strainer plenum with strainer modules on top was constructed over the SI Sump and over the concrete floor. 2The surface2 area of the new SI Sump strainers was increased from approximately 200 ft. to 3699 ft.

2. The sump partition that separates the two trains of the SI inlets was replaced with stainless steel grating.

3. The housing for the low level switch inside the SI Sump was relocated to mount on top of the new screen plenum.

4. The tubing for two level transmitters inside the SI Sump was rerouted to penetrate through the plenum in a designed location in order to prevent debris from passing through the penetration opening.

5. Nineteen TSP baskets were relocated to allow easier installation of the new plenum and strainers, or to eliminate interferences with the baskets.

Testing and Analyses As approved by NRC approvals of extension requests for Waterford 3 (References 44 and 59),

all testing was completed by the end of February 2008, and analyses were completed by October 23, 2008. The following documents were generated / revised to support the GL 2004-02 actions.

• Debris Generation Calculation 2004-07780 (Reference 28)

• Debris Transport Calculation 2005-05500 (including CFD modeling) (Reference 29)

• Downstream Effects Calculations 2005-02820 and 2005-12840 (References 31 and 32),

• Hydraulic Sizing Report GENE-0000-0053-4416 (Reference 36)

. Water Level Inside Containment Calculation MNQ6-4 (Reference 37)

. ECCS and CS Pump NPSH Analysis ECM07-001 (Reference 51)

• In Vessel Downstream Effects Calculation CN-SEE-I-08-42 (Reference 65)

Plant programs and Processes The program and process changes needed to address the GL 2004-02 concerns were completed by December 31, 2007.

Licensing Basis The licensing basis changes needed to address the GL 2004-02 concerns consist of UFSAR changes related to the plant modifications previously described.

to W3F1-2008-)069 Page 7 of 87 3.0 Specific Information Regarding Methodology for Determining Compliance Backqround Information To facilitate understanding of the methodology for demonstrating compliance with the applicable regulations, Waterford 3 has provided background information regarding the design of the Waterford 3 containment, and the manner in which the GL 2004-02 concerns will be addressed.

Waterford 3 Containment Design Waterford 3 is a Combustion Engineering PWR with a large volume dry containment. Each of the two loops contains two Reactor Coolant Pumps (RCP), one Steam Generator (SG), and associated piping, located within a concrete wall enclosure commonly referred to as a D-ring.

The two RCS piping loops are nearly identical with the exception that one loop includes the PZR and associated piping. The area inside each D-ring is open directly above it. The two D-rings are also open on the bottom to a common open area on the basemat of the plant. The refueling cavity and other concrete walls separate the two loops. The PZR is contained within a separate concrete enclosure.

All postulated pipe break LOCAs for which sump recirculation would be required would take place within the D-rings, in the reactor cavity, or inside the pressurizer cubicle.

3.a Break Selection NRC Issue 3.a:

The objective of the break selection process is to identify the break size and location that present the greatestchallenge to post-accident sump performance.

1. Describe and provide the basis for the break selection criteria used in the evaluation.

2. State whether secondary line breaks were considered in !the evaluation (e.g., main steam and feedwater lines) and briefly explain why or why not.

3. Discuss the basis for reaching the conclusion that the break size(s) and locations chosen present the greatestchallenge to post-accident sump performance.

WF3 Response 3.a.1:

Baseline Break Selection:

A number of breaks were selected in the debris generation calculation 2004-07780 (Reference

28) in order to provide conservative input for the transport calculations. The breaks that were selected are:

1. Break S1 - Hot Leg Piping at SG 1 Nozzle

2. Break S2 - SG 1 Hot Leg Piping at RPV Connection

3. Break S3 - Hot Leg Piping at SG 2 Nozzle

4. Break S4 - Tee between SG 1 Hot Leg and PZR Surge Line (Alternate Break)

5. Break S5 - Cold Leg Piping at RCP 1A Inlet Nozzle

6. Break S6 - Cold Leg Piping at RCP 2B Inlet Nozzle to W3F1-2008-0069 Page 8 of 87

7. Break S7 - PZR Surge Line at PZR Nozzle

8. Break S8 - Cold Leg Piping at RCP 1B Inlet Nozzle

9. Break S9 - Cold Leg Piping at RCP 2A Inlet Nozzle Breaks S1, S2 and S3 are located on the hot leg of the primary piping, which has the largest diameter of the primary piping with a 42-inch diameter, obviously producing the largest ZOI.

Breaks S1 and S3 are placed at the steam generator nozzles in order to capture the most debris. Break S2 is located at the RPV.

Breaks S5 and S6 are located at the RCP inlet nozzles on the cold leg piping, which has the next largest diameter of the primary piping with a 30-inch diameter. For conservatism, the ZOI for the cold leg piping was based on the 42-inch diameter of the hot leg. These breaks are located closer to the-SI Sump, thus providing conservative input for the debris transport and head loss calculations.

Break S4 is located on the PZR surge line at the connection to the SG 1 Hot Leg. In accordance with the alternate break methodology, it is considered to have a 12-inch inner diameter, although the actual ID is 10.126 inches. The debris load from the alternate break was determined; however, the deterministic methodology described in Sections 3, 4 and 5 of the NEI Guidance 04-07 (Reference 2) was used to determine bounding debris generation and transport volumes.

Break S7 is located where the PZR surge line connects'to the PZR inlet nozzle on the bottom of the PZR. This break is located outside the D-ring walls, and nearer the SI Sump location than any of the other breaks. This break does not produce a significant amount of debris; however, it was selected due to the proximity and clear debris path to the SI Sump.

Break 38 is located at the RCP 1B inlet nozzle on the cold leg piping. This break will have the same debris generation as for the S5 break. This break was only analyzed for flow distribution in containment. This was to determine the worst case flow for debris transport. The S8 break did not result in higher debris transport than the other breaks already analyzed.

Break S9 is located at the RCP 2A inlet nozzle on the cold leg piping. This break will have the same debris generation as the S6 break. This break was only analyzed for the flow distribution in containment. This was to determine the worst case flow for debris transport. The S9 break did not result in higher debris transport than the other breaks already analyzed.

SBLOCA events were not included in this evaluation, as the debris load would not be bounding due to the smaller areas covered by the break zones. The majority of the fiber that would be created is inside the D-Rings, and the larger breaks in those areas will create larger quantities of fiber than could be created from the SBLOCA events.

WF3 Response 3.a.2:

Secondary pipe breaks were not considered for this analysis. Based upon a review of the plant UFSAR discussed in the Debris, Generation calculation 2004-07780 (Reference 28),

.containment spray and recirculation are not required for a Main Steam Line Break or a to W3F1-2008-0069 Page 9 of 87 Feedwater Line Break. Additionally, breaks of small lines were not investigated, because.the debris load would not be bounding due to the smaller areas covered by the break zone.

WF3 Response 3.a.3:

The locations of the analyzed breaks are chosen in order to maximize the amount and types of debris generated. To this end, breaks are placed near large equipment, specifically the SGs, RCPs, and PZR. The breaks were also placed near walls and the floor since concrete surfaces have very thick coatings compared to steel surfaces. Finally, breaks were located in areas expected to maximize the transport of debris to the sump strainer.

3.b Debris Generation/Zone of influence (ZOI) (excluding coatings)

NRC Issue 3.b:

The objective of the debris generation/ZOI process is to determine, for each postulated break location: (1) the zone within which the breakjet forces would be sufficient to damage materials and create debris; and (2) the amount of debris generated by the break jet forces.

1. Describe the methodology used to determine the ZOls for generating debris. Identify which debris analyses used approved methodology default values. Fordebris with ZOls not defined in the guidance report/SE, or if using other than default values, discuss method(s) used to determine ZOI and the basis for each.

2. Provide destruction ZOls and the basis for the ZOls for each applicable debris constituent.

3. Identify if destruction testing was conducted to determine ZOls. If such testing has not been previously submitted to the NRC for review or information, describe the test procedure and results with reference to the test report(s).

4. Provide the quantity of each debris type generated for each break location evaluated. If more than four break locations were evaluated, provide data only for the four most limiting locations.

5. Provide total surface area of all signs, placards, tags, tape, and similar miscellaneous materialsin containment.

WF3 Response 3.b.1:

In order to perform the calculation of debris generation within the ZOI, a representative model of the insulation location and volume is utilized in the debris generation calculation 2004-07780 (Reference 28). The model is a Microsoft Excel spreadsheet created from piping isometric drawings and insulation drawings. These drawings were used to develop a 3-dimentional computer model, which was then converted to the Excel spreadsheet model. The spreadsheet determines the amount of insulation within a ZOI centered at coordinates that are input by the user. In this way multiple break locations are able to be evaluated relatively quickly and the user can ensure that conservative and limiting breaks are chosen.

The insulation in containment at Waterford 3 consists of Nukon (canvas encapsulated (unjacketed) or stainless st'eel jacketed), MEI (stainless steel jacketed), Min-K, Microtherm, and Transco RMI. The amount of insulation debris generated is dependent on the proximity of each insulated target to the postulated break.

to W3F1-2008-0069 Page 10 of 87 The SER for NEI 04-07 (Reference 3) recommends a ZOI radius of 17.0D ("D" being the'inside diameter of the pipe break) for both jacketed and unjacketed Nukon Fiber. The SER recommended ZOI radius of 17D is used for unjacketed Nukon Fiber debris generation analysis.

Based on testing contained in Westinghouse report WCAP-1 671 0-P (Reference 38), a reduced ZOI of 7.0D is used for the SS Jacketed Nukon debris generation analysis. NEI Guidance 04-07 and the SER for NEI 04-07 (References 2 and 3) do not address MEI insulation. Based on industry testing contained in Alion report ALION-REP-ENTG-4771-02 (Reference 39), a ZOI of 4.0D is used for the SS Jacketed MEl debris generation analysis. The SER recommended ZOI radius of 28.6D is used for Min-K insulation. The NEI and SER documents do not address the ZOI for Microtherm insulation. Therefore, a ZOI radius of 28:6D is used for Microtherm based on its similarity to Min-K insulation and because the ZOI radius for Min-K is the largest of all the tested materials described in the NEI and SER documents. The SER recommended ZOI radius of 2D is used for RMI debris generation analysis.

As discussed in Section 3d of this response submittal, latent debris and miscellaneous (foreign) materials are also included in the debris generation analysis. The amounts of these types of debris are determined from plant walkdown reports and are presented in their respective section of this response.

WF3 Response 3.b.2:

Debris Type ZOl Basis Jacketed Nukon Fiber Blankets 7 WCAP-1 6710-P (Reference 38)

Unjacketed Nukon Fiber Blankets 17 SER for NEI 04-07 (Reference 3)

ALION-REP-ENTG-4771-02 (Reference 39)

SER for NEI 04-07 M 2(Reference 3)

Microtherm 28.6 2004-07780 (Reference 28)

Transco RMI 2(Rfrne3 SER for NEI 04-07 (Reference 3)

WF3 Response 3.b.3:

Westinghouse report WCAP-16710-P (Reference 38) documents testing performed on jacketed Nukon insulation blankets to determine the proper ZOI. From Section 4 of WCAP-16710-P (Reference 38), "The approach taken to develop this experimental program was to subject the encapsulated... stainless steel jacketed NUKON fiberglass insulation materials to phenomena and processes that accurately simulate those experienced during a postulated LOCA blowdown for a PWR. The conditions of interest are exposure to elevated temperature, pressure and high mass flux. Data collected from and observations from the tests were used as follows:

For the jacketed NUKON fiberglass insulation system, determine an appropriate, technically defensible, realistic material-specific ZOI at which the fiberglass insulation will not experience damage that would require it to be treated as debris."

to W3F1-2008-0069 Page 11 of 87 The objective of the test was to determine the generation of debris of the insulation material that should be considered in post-accident sump performance. The testing consisted of subjecting the jacketed NUKON insulation to a two phase jet originating from a subcooled, high pressure, high temperature reservoir. For the purposes of this testing, debris generation was defined as the "observable release or extrusion of fiberglass insulation material from the fiberglass from the woven fiberglass cloth covered blanket". The results of this testing showed that the ZOI for the SS jacketed NUKON insulation could be reduced to as low as 5D. However, for conservatism, a ZOI of 7D was used.

WF3 Response 3.b.4:

The insulation and coating debris totals for the four most limiting breaks evaluated are presented in the table below.

Table 3b-1: Summary of LOCA Generated Debris Inside the ZOI Debris Type Units $1/$5 S2 $3/$6 S4 Jacketed Nukon Fiber Blankets [ft3] 81.4 177 25.4 0 Unjacketed Nukon Fiber Blankets [ft3] 351.1 113 50W18 331 MEI [ft7] 405.0 3 558.4 40.8 Min-K [ft3] 0.4 0.4 0.4 0.4 Microtherm [ft3] 4.2 4.2 4.2 4.2 Transco RMI 7ft2] 0 8750 0 0 WF3 Response 3.b.5:

As discussed in Section 3d of this response submittal, latent debris and miscellaneous (foreign) materials are also included in the debris generation analysis. An analysis in the Debris Generation calculation 2004-07780 (Reference 28) shows that the debris loading to be used is 250 Ibm.

Labels, tags, stickers, placards and other miscellaneous or foreign materials were evaluated via walkdown. The walkdown results are included in Debris Generation calculation 2004-07780 (Reference 28). Based on this calculation, a sacrificial area of 151 ft 2 of the strainer surface is used for stickers, index cards, placards, tape, glass and other miscellaneous or foreign materials. This total includes only those materials which are not Design Basis Accident (DBA) qualified and does not include any overlap of the materials on the screens.

3.c Debris Characteristics NRC Issue 3.c:

The objective of the debris characteristicsdeterminationprocess is to establish a conservative debris characteristics profile for use in determining the transportability of debris and its contribution to head loss.

1. Provide the assumed size distribution for each type of debris.

2. Provide bulk densities (i.e., including voids between the fibers/particles) and material densities (i.e., the density of the microscopic fibers/particles themselves) for fibrous and particulatedebris.

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3. Provide assumed specific surface areasfor fibrous and particulate debris.

4. Provide the technical basis for any debris characterizationassumptions that deviate from NRC-approved guidance.

WF3 Response 3.c:

The debris sources at Waterford 3 include insulation, coatings, foreign material and latent debris. The characteristics of the insulation debris material are discussed in this section. The characteristics of the other debris types (e.g. coatings, foreign and latent) are included in their respective sections of this response submittal (Sections 3h and 3d).

WF3 Response 3.c.1:

Unjacketed Nukon Insulation Size Distribution:

For unjacketed Nukon insulation with a ZOI of 17D, a size distribution of 8% fines, 25% small pieces, 32% large pieces and 35% intact pieces is used. This distribution is determined based on an analysis of the results of the BWR Owners' Group (BWROG) air-jet impact tests (AJIT) and the Ontario Power Generation (OPG) debris generation tests as described in NUREG/CR-6808 (Reference 18). Based on the results of the AJIT and OPG debris generation tests as presented in Appendix VI of the SER for NEI 04-07 (Reference 3) and Volume 3 of NUREG/CR-6762 (Reference 15), 33% of all fibrous insulation within the ZOI is modeled as becoming fines or small debris. This fraction of fines or small debris is conservatively increased from the value (22%) suggested in Appendix II of the SER based on the OPG debris generation test. Implicit in these values is the assumption that the insulation is uniformly distributed within the ZOI. Due to the fact that the unjacketed Nukon insulation for the applicable breaks is distributed in several locations within each ZOI, uniformity is considered a reasonable approximation. Thus, 67% of all fibrous insulation within the ZOI is modeled as becoming either large debris or remaining intact. To determine the appropriate split between fine and small debris, the results of the AJIT are used. The AJIT indicated that, when insulation was completely destroyed, a maximum of 25% of the insulation was too fine to collect by hand. Thus 25% of the 33% small debris fraction is modeled as becoming fines; i.e. 8% [0.25*0.33] of the fibrous insulation within the ZOI becomes fine debris when destroyed. This implies that 25% [(1-0.25)*0.33] of the fibrous insulation within the ZOI becomes small debris when destroyed. To determine the appropriate split between large and intact debris, the results of the AJIT are also used. Per the SER guidance provided in Appendix VI of the SER for NEI 04-07 (Reference 3), 35% of the fibrous insulation within the ZOI is modeled as intact debris, leaving 32% as large piece debris. Fines that enter the active recirculation pool are considered 100% transportable. Small, large and intact pieces are transported based on velocity data found in various, references. Specifics of debris transport are discussed in Section 3e.

SS Jacketed Nukon Insulation Size Distribution:

For stainless steel jacketed Nukon insulation with a ZOI of 7D a size distribution of 25% fines and 75% small pieces is used. For a ZOI of 7D, the suggested Nukon size distribution contained in Table 3-3 of the SER for NEI 04-07 (Reference 3) is not applicable. Instead the size distribution is determined from Figure I1-1 of the SER, which relates jet pressure toZOI radii, and Figure 11-2 of the SER, which relates jet pressure to the fraction of small debris generated. The data presented in Figure 11-2 comes from the Air Jet Impact tests, which are discussed in many documents related to GSI-191 including NUREG/CR-6808 (Reference 18).

to W3F1-2008-0069 Page 13 of 87 SS Jacketed MEI Insulation Size Distribution:

For stainless steel jacketed MEI insulation with a ZOI of 4D a size distribution of 20% fines and 80% small pieces is used per ALION-REP-ENTG-4771-02 (Reference 39).

Min-K Insulation Size Distribution:

For Min-K insulation a conservative size distribution of 100% fines is used as documented in Table 3-3 of the SER for NEI 04-07 (Reference 3). Fines are the constituent part of the insulation and are considered 100% transportable.

Microtherm Insulation Size Distribution:

For Microtherm insulation a conservative size distribution of 100% fines is used as documented in Table 3-3 of the SER for NEI 04-07 (Reference 3). Fines are the constituent part of the insulation and are considered 100% transportable.

Transco RMI Insulation Size Distribution:

For Transco RMI insulation with a ZOI of 2D, a size distribution of 75% small fines and 25%

large pieces is used. This size distribution is confirmed by Table 3-3 of the SER for NEI 04-07 (Reference 3).

WF3 Response 3.c.2:

Nukon Insulation Density:

Per Table 3-2 of NEI Guidance 04-07 (Reference 2) the bulk density of Nukon insulation is 2.4 Ibm/ft 3. The bulk density of the Nukon insulation installed at Waterford 3 and used in the sump strainer performance testing is also 2.4 Ibm/ft 3 .

MEI Insulation Density:

The metal encapsulated insulation (MEI) is Owens-Corning TIW Type I1. Based on manufacturer data for this insulation the bulk density is 2.4 Ibm/ft3 . The bulk density of the MEI insulation installed at Waterford 3 and used in the sump strainer performance testing is also 2.4 Ibm/ft3.

Min-K Insulation Density:

Per Table 3-2 of NEI Guidance 04-07 (Reference 2) the bulk density of Min-K insulation is 8 to 16 Ibm/ft 3. The bulk density of the Min-K insulation installed at Waterford 3 is 13 Ibm/ft 3 per calculation ECM89-083 (Reference '40). This compares to a bulk density of 14.5 Ibm/ft 3 for the insulation used in the sump strainer performance testing.

Microtherm Insulation Density:

Per Table 3-2 of NEI Guidance 04-07 (Reference 2) the bulk density of Microtherm insulation is 5 to 12 Ibm/ft 3. This compares to a bulk density of 14.5 Ibm/ft 3 of the insulation used in the sump strainer performance testing.

Reflective Metal Insulation Density:

Transco and Mirror RMI are comprised of thin layers of stainless steel foil. Stainless steel has a density of 490 Ibm/ft 3.

to W3F1-2008-0069 Page 14 of 87 WF3 Response 3.c.3:

The Material density and specific surface area (SJ)were only used for preliminary analytically determined head loss values across a debris laden sump screen using the correlation given in NUREG/CR-6224 (Reference 8). Since the head loss across the installed sump screen is determined via testing, these values are not used in the design basis for Waterford 3.

Therefore, these values are not provided as part of this response.

WF3 Response 3.c.4:

Waterford 3 debris, generation, transport, and head loss testing have used the debris characterization assumptions provided in the SER for NEI 04-07 (Reference 3).

3.d Latent Debris NRC Issue 3.d:

The objective of the latent debris evaluation process is to provide a reasonableapproximation of the amount and types of latent debris existing within the containment and its potential impact on sump screen head loss.

1. Provide the methodology used to estimate quantity and composition of latent debris.

2. Provide the basis for assumptions used in the evaluation.

3. Provide results of the latent debris evaluation, including amount of latent debris types and physical data for latent debris as requested for other debris under c. above.

4. Provide amount of sacrificialstrainersurface area allotted to miscellaneous latent debris.

WF3 Response 3.d.1:

Latent debris has been evaluated by containment walkdown as recommended by Section 3.5.2 of NEI Guidance 04-07 (Reference 2) and confirmed by the NRC SER for NEI 04-07 (Reference 3). The walkdown of the Waterford 3 containment was conducted in accordance with the guidance provided by NEI Guidance documents 04-07 (Reference 2) and 02-01 (Reference 20) and the SER for NEI 04-07 (Reference 3). As shown below, four (4) or more samples were collected for all surface types except grating. The additional samples collected for certain surface types increase the statistical accuracy of the evaluation. A listing of the number of each sample type follows.

Number of Samples Collected

• Containment Liner .............. 4 HVAC Duct (Vertical) ...... 4

• Equipment (Horizontal) ....... 4 Pipe (Horizontal) ................ 4

• Equipment (Vertical) ............ 4 Pipe (Vertical) ..... ...... 4

• Floor ..................... I.............. 4 Cable Tray (Horizontal) ........ 4

• W all .................................... 5 Cable Tray (Vertical) ........... 4

• HVAC Duct (Horizontal) ...... 4 Gratings ............................ 0 The weights of the samples collected are used to determine the latent debris mass distribution (g/ft2). Measurements taken are accurate to 0.01 grams. A statistical analysis of the samples is performed in the post-processing of the latent debris walkdown results, which is Attachment to W3F1 -2008-0069 Page 15 of 87 8.8 of the Waterford 3 Debris Generation calculation 2004-07780 (Reference 28). The analysis determined a 90% confidence limit of the mean value for each type of surface based on a normal distribution. The upper limit of the mean value for each surface type is then applied over the entire surface area of that type throughout containment. This analysis lends further confidence and conservatism to the latent debris mass determination.

Labels, tags, stickers, placards and other miscellaneous or foreign materials (including glass) were also evaluated via walkdown. The walkdown results are included as attachment 8.9 of the Waterford 3 Debris Generation calculation 2004-07780 (Reference 28) and are summarized in section 5.5 of the calculation.

WF3 Response 3.d.2:

No samples were available for grating; therefore, grating is conservatively assumed to have the same latent debris loading as the floor.

WF3 Response 3.d.3:

Consistent with the NRC SER for NEI 04-07 (Reference 3), 15% of the latent debris load (by mass) is assumed to be fibrous debris and the other 85% (by mass) is treated as particulate3 debris. Likewise, consistent with the SER for NEI 04-07 (Reference 3), a density of 2.7 g/cm for particulate debris is used. For, latent fibrous debris, a density of 2.4 Ibm/ft3 (bulk density of Nukon per NEI Guidance 04-07, Reference 2) is used in order to conservatively maximize the volume of latent fibrous debris. As the specific surface area of debris is only relevant for head-loss calculations per NUREG/CR-6224 (Reference 8) and head-loss evaluations are being conducted experimentally, the specific surface area of latent debris is not determined.

The results of the latent debris calculation conservatively determined the debris loading to be 250 Ibm.

Miscellaneous latent debris is also discussed in more detail in the following debris transport section 3.e.

Table 3d-1: Latent and Foreign Material Debris Latent and Foreign Material Debris Quantity Latent Debris (Ibm) 250 Fiber (Ibm) 37.5 Particulate (Ibm) 212.5 Foreign Material Debris (ft 2) 151 Stickers, index cards, placards and tape (ft2) .81.

Glass (light bulbs) (ft2 ) 70 WF3 Response 3.d.4:

A sacrificial area of 151 ft2 of the strainer surface is retained for stickers, index cards, placards, tape, glass and other miscellaneous or foreign materials. This total includes only those materials which are not Design Basis Accident (DBA) qualified.

to W3F1-2008-0069 Page 16 of 87 3.e Debris Transport NRC Issue 3.e The objective of the debris transportevaluationprocess is to estimate the fraction of debris that would be transportedfrom debris sources within containment to the sump suction strainers.

1, Describe the methodology used to analyze debris transport during the blowdown, washdown, pool-fill-up, and recirculationphases of an accident.

2, Provide the technical basis for assumptions and methods used in the analysis that deviate from the approved guidance.

3, Identify any computational fluid dynamics codes used to compute debris transport fractions during recirculation and summarize the methodology, modeling assumptions, and results.

4' Provide a summary of, and supporting basis for, any credit taken for debris interceptors.

5, State whether fine debris was assumed to settle and provide basis for any settling credited.

6. Provide the calculated debris transport fractions and the total quantities of each type of debris transportedto the strainers.

WF3 Response 3.e.1:

Debris transport determines the fraction of debris generated that is transported from debris sources (break location) to the sump screen. The debris transport analysis for Waterford 3 is conducted in accordance with both NEI Guidance 04-07 (Reference 2) and the NRC SER for NEI 04-07 (Reference 3). As such, each phase of post-LOCA transport is considered:

blowdown, washdown, pool fill-up and recirculation. A detailed discussion of each transport phase, including information on their effect on overall transport for Waterford 3 follows.

Blowdown and Washdown:

For mostly uncompartmentalized containments such as at Waterford 3, Section 3.6.3.2 of NEI Guidance 04-07 (Reference 2) states that all RMI debris (small and large) is conservatively postulated to fall to the containment floor; i.e. no RMI debris is ejected into the dome. Although NEI 04-07 does not specifically state that all fiber debris is assumed to fall to the containment floor, it is conservatively modeled as such (see Table 3-4 of the SER for NEI 04-07, Reference 3). Similarly, all Min-K, Microtherm, and coating debris is also conservatively modeled as falling to the containment floor. Thus, all LOCA generated debris is conservatively modeled as falling to the floor in the post-accident environment. This is reasonable as large debris should be modeled as falling to the containment floor per Section 3.6.3.2 of NEI Guidance 04-07 (Reference 2) and small debris that could reach the dome would eventually wash down to the active pool. Therefore, since all insulation debris eventually lands on the floor, a detailed blowdown and washdown analysis is not conducted. Rather all insulation debris generated is conservatively placed on the floor immediately and is further transported by pool fill-up and recirculation as discussed in the following sections. Conservatively, qualified coatings are also considered to fall directly to the floor. All other debris types, including unqualified coatings, latent and foreign material debris that are generated from outside the break ZOI are therefore considered to fall directly to the floor.

Pool Fill-up Conservatively, no inactive pools are credited. All debris on the floor prior to pool fill-up remains on the floor in the active pool after pool fill-up. During pool fill-up debris is transported away to W3F1-2008-0069 Page 17 of 87 from the break area and toward the perimeter of containment by the water spilling onto the floor.

Debris is then further transported by recirculation, as discussed in the following section.

Recirculation Debris that reaches the containment pool is subject to transport by the pool flow present during recirculation. In accordance with the NEI Guidance 04-07 and the SER for NEI 04-07 (References 2 and 3) all fine debris that lands in the pool is considered to transport entirely to the sump strainer. The transport of small, large and intact pieces of debris during recirculation is dependent on the velocities present in the containment pool.

Nukon debris transport is investigated and reported in NUREG/CR-6772 (Reference 16).

Transport velocities pertinent to Nukon debris transport at Waterford 3 are taken from this document. The document reports values at which some debris begins to move and at which a majority begins to move. These are referred to herein as the "incipient tumbling" and "bulk transport" velocities. Conservatively, the incipient tumbling velocity is used to determine transport potential. Accordingly, for breaks corresponding to Configuration A in NUREG/CR-6772 (Reference 16), non-fines Nukon pieces are considered to transport at velocities of 0.12 ft/s or greater and small and large Nukon pieces are considered to transport over a 6-inch curb at 0.34 ft/s. For breaks corresponding to Configuration B in NUREG/CR-6772 (Reference 16),

non-fines Nukon pieces are considered to transport at velocities of 0.07 ft/s or greater and small and large Nukon pieces are considered to transport over a 6-inch curb at 0.25 ft/s. For breaks corresponding to Configuration C in NUREG/CR-6772 (Reference 16), non-fines Nukon pieces are considered to transport at velocities of 0.06 ft/s or greater and small and large Nukon pieces are considered to transport over a 6-inch curb at 0.28 ft/s. Intact Nukon fiber blankets and Nukon jacketing are not considered to lift over the curb due to their size.

RMI debris transport is investigated in NUREG/CR-3616 (Reference 7) and NUREG/CR-6772 (Reference 16). Transport velocities pertinent to RMI debris transport at Waterford 3 are taken from these documents. Both documents report values at which some debris begins to move and at which a majority begins to move. These are referred to herein as the "incipient tumbling" and "bulk transport" velocities. Conservatively, the incipient tumbling velocity is used to determine transport potential. Accordingly, for breaks corresponding to Configuration A in NUREG/CR-6772 (Reference 16), non-fines RMI pieces are considered to transport at velocities of 0.28 ft/s or greater and are considered to transport over a 6-inch curb at 0.84 ft/s. For breaks corresponding to Configuration B in NUREG/CR-6772 (Reference 16), non-fines RMI pieces are considered to transport at velocities of 0.41 ft/s or greater and are considered to transport over an 8-inch curb at 0.30 ft/s. For breaks corresponding to Configuration C in NUREG/CR-6772 (Reference 16), non-fines RMI pieces are considered to transport at velocities of 0.20 ft/s or greater and are considered to transport over a 6-inch curb at 1.0 ft/s.

As noted in NUREG/CR-6773 (Reference 17) low density fiberglass debris (such as Nukon and MEI debris) is subject to erosion during Recirculation. The erosion rate used for the small and large Nukon debris pieces is 10% over the 30-day Recirculation mission time as described in ALION-REP-ENT-4536-02 (Reference 41). The erosion rate used for the MEI debris pieces is 90% over the 30-day Recirculation mission time, as described in Appendix II of the.SER for NEI 04-07 (Reference 3).

to W3F1-2008-0069 Page 18 of 87 WF3 Response 3.e.2:

There are no assumptions or methods that deviate from the approved guidance in the SER for NEI 04-07 (Reference 3) in the areas of debris transport, except for the erosion rate of the Nukon insulation. This is justified in an analysis / test documented in the report ALION-REP-ENT-4536-02 (Reference 41).

WF3 Response 3.e.3:

To assist in the determination of recirculation transport fractions, several Computational Fluid Dynamics (CFD) simulations were performed using FluentTM, a commercially available software package. Multiple break locations were investigated by the CFD simulations to determine which scenario would maximize debris transport. Four of the eight simulations conducted were based on the final strainer system design. The simulation results include a series of contour plots of velocity and turbulent kinetic energy (TKE), plots of flow pathlines originating at the break locations and animations of the flow velocities. These results have been combined with information in the GSI-191 literature and plant specific erosion test results to determine the overall transport fractions for small, large and intact pieces of fibrous debris and large pieces of RMI debris (fines are 100% transportable).

WF3 Response 3.e.4:

No debris interceptors were installed at Waterford 3 as part of the GL2004-02 resolution.

WF3 Response 3.e.5:

Credit is taken for the plenum that the strainers sit on. From the CFD results it is determined how much of the plenum perimeter is in areas with flow velocities in excess of the velocity required to lift the debris over a 6-inch obstacle. Since the plenum is approximately 8-inches tall, using the lift-over curb velocity for a 6-inch curb is conservative. The fraction of the curb perimeter in excess of the lift-over curb velocity is applied to the debris pile in the vicinity of the strainer to determine the debris load on the strainer. In all but one of the CFD scenarios the approach velocity does not exceed the lift over curb velocity at any location along the plenum perimeter. In the remaining scenario, less than 25% of the perimeter has velocities in excess of those necessary to lift over a 6-inch curb; however for conservatism, 25% of the small debris is treated as lifting onto the sump strainer.

No credit was taken for settling of fine debris.

WF3 Response 3.e.6:

The amount of debris determined to transport to the sump strainer for the limiting breaks are provided in the following tables.

to W3F17-2008-0069 Page 19 of 87 Insulation:

Table 3.e.6-1: Debris Generated and Transported to Strainer- Break S1 Debris Transport Debris at Ge n r a e Frac t Strainer D eb ris T ranspo rt by Type U nits Generated Fraction Strainer SS Jacketed Nukon [W] 81.4 0.325 26.46 Unjacketed Nukon [ft3] 351.1 0.137 48.1 SS MEI (Fiberglass) 405.0 0.92 372.6 Min-K [ft3] 0.4 1.00 0.4 Microtherm [ft 4.2 1.00 4.2 Transco RMI [ft 2] 0 0.75 0 Qualified Coatings [fta] 13.5 1.00 13.5 Unqualified Coatings [Tj] 31.02

• 21.70 Latent Debris [Ibm] 250 1.00 250 Foreign Materials: [ft 151 1.00 151 Table 3.e.6-2: Debris Generated and Transported to Strainer - Break S3 Debris ' Transport Debris at Debris Transport by Type Units Generated Fraction Strainer SS Jacketed Nukon [W] 25.4 0.325 8.26 Unjacketed Nukon [ft3] 501.8 0.137 68.75 SS MEI (Fiberglass) [ft 3] 558.4 0.92 513.73 Min-K 0.4 1.00 0.4 Microtherm 4.2 1.00 4.2 Transco RMI [ft 1 0 0.75 0 Qualified Coatings [ft3] 13.5 1.00 13.5 Unqualified Coatings [ft] 31.02

• 21.70 Latent Debris [Ibm] 250 1.00 250 Foreign Materials: [ft2] 151 1.00 151 Table 3.e.6-3: Debris Generated and Transported to Strainer - Break S4 Debris Transport Debris at Debris Transport by Type Units Genrae Fract Strainer Generated Fraction Strainer SS Jacketed Nukon [ft3] 0 0.325 0 Unjacketed Nukon [ft3] 331 0.137 45.35 SS MEi (Fiberglass) t3] 40.8 0.92 37.54 Min-K [ft3] 0.4 1.00 0.4 Microtherm [ft] 4.2 1.00 4.2 Transco RMI [ft2] 0 0.75 0 Qualified Coatings [Wt] 13.5 1.00 13.5 Unqualified Coatings [ft 3] 31.02

• 21.70 Latent Debris [Ibm] 250 1.00 250 Foreign Materials: [ft 151 1.00 151 to W3F1-2008-0069 Page 20 of 87 Table 3.e.6-4: Debris Generated and Transported to Strainer - Break S5 Debris .Transport Debris at Gen ra e Fra c t Strainer Debris Transport by Type Units SGenerated Fraction Strainer SS Jacketed Nukon [Wt] 81.4 0.325 26.46 Unjacketed Nukon [ft3] 351.1 0.137 48.1 SS MEI (Fiberglass) [ft 405.0 0.92 372.6 Min-K [ft 0.4 1.00 0.4 Microtherm Ift3] 4.2 1.00 4.2 Transco RMI I[ft] 0 0.75 .0 Qualified Coatings [ft3] 13.5 1.00 13.5 Unqualified Coatings 31.02

• 21.7 Latent Debris [Ibm] 250 1.00 250 Foreign Materials: [ft] 151 1.00 151 Table 3.e.6-5: Total Debris Generated and Transported to Strainer - Break S7 Debris Transport by Type Units Debris Transport Debris at Generated Fraction Strainer SS Jacketed Nukon [fti] 30 0.3925 11.78 Unjacketed Nukon [ft 49 0.1883 9.23 SS MEl (Fiberglass) [ft3] 0 0.93 0 Min-K [ft 0 1.00 0 Microtherm [ft1] 0.6 1.00 0.6 Transco RMI [ft 0 0.775 0 Qualified Coatings [ft3] 0 1.00 0 Unqualified Coatings [ft3] 31.03

• 21.71 Latent Debris [Ibm] 250 1.00 250 Foreign Materials: [fti] 151 1.00 151 Inorganic zinc, coatings within the ZOI, and indeterminate coatings are considered to transport 100% to the sump. The inventory of degraded qualified coatings for use in the sump screen design is the portion which enters the pool on or near the sump strainer, and the portion elsewhere that is in an area with flow velocities high enough to transport that debris.

3.f Head Loss and Vortexing NRC Issue 3. f The objectives of the head loss and vortexing evaluations are to calculate head loss across the sump strainerand to evaluate the susceptibility of the strainerto vortex formation.

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

2. Provide the minimum submergence of the strainer under small-break loss-of-coolant accident (SB LOCA) and large-breakloss-of-coolant accident (LB LOCA) conditions.

3. Provide a summary of the methodology, assumptions and results of the vortexing evaluation. Provide bases for key assumptions.

to W3F1-2008-0069 Page 21 of 87

4. Provide a summary of the methodology, assumptions, and results of prototypical head loss testing for the strainer, including chemical effects. Provide bases for key assumptions.

5. Address the ability of the design to accommodate the maximum volume of debris that is predicted to arrive-at the screen.

6. Address the ability of the screen to resist the formation of a "thin bed" or to accommodate partialthin bed formation.

7. Provide the basis for the strainerdesign maximum head lossl

8. Describe significant margins and conservatisms used in the head loss and vortexing calculationrs.

9. Provide a summary of the methodology, assumptions, bases for the assumptions, and results for the clean strainerhead loss calculation.

10. Provide a summary of the methodology, assumptions, bases for the assumptions, and results for the debris head loss analysis.

11. State whether the sump is partially submerged or vented (i.e., lacks a complete water seal over its entire surface) for any accident scenarios and describe what failure criteria in addition to loss of net positive suction head (NPSH) margin were applied to address potential inability to pass the required flow through the strainer.

12. State whether near-field settling was credited for the head-loss testing and, if so, provide a description of the scaling analysis used to justify near-field credit.

13. State whether temperature/viscositywas used to scale the results of the head loss tests to actual plant conditions. If scaling was used, provide the basis for concluding that boreholes or other differential-pressureinduced effects did not affect the morphology of the test debris bed.

14. State whether containment accident pressure was credited in evaluating whether flashing would occur across the strainersurface, and if so, summarize the methodology.

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0-CD to W3F1-2008-0069 Page 28 of 87 WF3 Response 3.f.2:

The top elevation of the new SI Sump strainers at Waterford 3 is (-)5.935 ft MSL as shown on drawing 5817-13604 (Reference 66). As documented in the response to item 3.g.1, the minimum, water level for the SBLOCA is (-)5.79 ft MSL, and for the LBLOCA is (-)5.29 ft MSL.

This results in the top plate of the strainers being submerged by approximately 2" for a SBLOCA and 8" for a LBLOCA. Note that the top plate of the Waterford 3 sump strainer modules do not have perforated plate, but are solid, so suction does not occur on the top of the strainer.

WF3 Response 3.f.3:

No tests were run specifically for vortexing with specific assumptions. Instead vortexing observations were made as part of the module headloss test program. Module testing was conducted with the water depth over the strainers similar to the plant configuration per test specification 26A6833 (Reference 33). No vortexing or air entrainment was observed during testing. Testing was performed that included observation for vortexing at water levels 4-inches above the strainer (typical of plant minimum sump water level), 3-inches above the top of the strainers, 2-inches above the top of the strainer, and 1-inch above the top of the strainer, for the full range of debris loads. This test program ensures a high degree of confidence that a vortex or other form of air entrainment shall not occur with the Waterford 3 strainer. For added vortex prevention assurance, the original vortex breaker cages have been left in the sump on the intake piping. Pre-operational sump testing confirmed that the cages were effective in preventing both surface and subsurface vortices with flow rates higher than analyzed for Generic Letter 2004-02 resolution.

WF3 Response 3.f.4:

Module testing consists of scaling the plant's debris load and measuring the debris induced head loss across a module of a strainer. These tests determine the head loss characteristics of plant-specific debris as a function of scaled debris load and scaled flow rate.

Six module tests have been performed to analyze the six strainer bounding cases, with constant flow rates and scaling factors, and with varying particulate/fiber ratios, nominal debris bed thicknesses, and low density fiberglass quantities (Test Specification 26A6833 - Reference 33).

The test module is composed of ten 40" X 40" square perforated disks. The test module is mounted on the center of the test pool with the same floor clearance as its simulating strainer.

Water level was maintained 4" above the top of the test article, typical of the plant installation.

The flow rates for the module tests are calculated using Equation 3.f.4-1, which yields the same circumscribed approach velocity for the test module as for the proposed plant strainers.

Because the debris loads evaluated by the module tests result in thick, circumscribed debris beds, the circumscribed area approach velocity is appropriate for this test. The module test flow rate is 367 gpm (for both circumscribed area scaled flow rate and perforated area scaled flow rate).

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Area = Surface Area (ft2 )

The module test debris quantities in the test matrix were calculated using the debris loads provided by the design input (Reference 34) and Equation 3.f.4-2, which yields the same circumscribed debris bed thickness for the module test as in the proposed plant strainer.

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.p lant -- A r e aS a r tfi A r e a circ unscribed cia l "Are a 0 ,,,

p, e p la nt arib ArIea pe'rforated, poant Volume debris.transported.to.sump - Volume of debris that is transported to sump (ft3 )

Pdebris = As-manufactured density of debris (lb/ft3)

Area = Surface Area (ft2 )

The module tests make use of the following assumptions:

o The flow rate is proportional to the circumscribed area of the strainers; o The debris load and flow rate is distributed equally among the strainers; o The debris bed is uniform - same thickness throughout perforated surface; o In the debris load calculation, the circumscribed surface area of a plant installed strainer is the actual circumscribed surface area minus the portion of sacrificial area attributed to the circumscribed area.

The impact of chemical effects on head loss was quantified through plant specific chemical effects testing. This testing included both 30 day integrated testing, and WCAP-16530-NP /

WCAP-16785-NP (References 58 and 53) based testing. See section 3.o of this report for a more thorough explanation.

WF3 Response 3.f.5:

During a LOCA at Waterford 3, the following types of debris may be generated by the high-energy steam and liquid impingement and water wash down/flow (Design Input - Reference 34):

3 o Fibrous Insulation: Nukon Fiber Blankets and Owens-Corning TIW II insulation of 591 ft volume is transported to the sump screen.

o Granular Insulation: Min-K and Microtherm insulation of 4.6 ft 3 volume with all transported to the sump screen.

o Latent Debris: 250 Ibm of latent debris is considered to be 15% by mass of fiber, and 85% particulate. All latent debris is assumed to be transported to the sump screen.

o Qualified and Unqualified Coating: Qualified Coatings (steel and concrete) of 13.5 ft 3 and Unqualified Coatings of 21.7 ft 3 is transported to the sump screen.

o Foreign Materials: The foreign materials (sacrificial area) are assumed to be 151 ft2 in area without taking credit for any overlap of these materials.

Testing was performed with two types of test articles: sectors and modules. [Proprietary Information Removed]

to W3F1-2008-0069 Page 30 of 87 The percentage of transported debris that adhered to each strainer is assumed to be equal to the strainer's percentage of total flow.

WF3 Response 3.f.6:

The nominal debris bed thickness for the tests ranges from 0.12" to 7.1" (Test Specification 26A6833 - Reference 33) in plant installed units with worst case-operating scenario. There is potential for a bed thickness matching the "thin bed" description to be formed during the strainer operation; however, the limiting head loss did not occur with a "thin bed" during Waterford 3 testing. The highest head loss occurs when 100% of the fiber is transported to the strainer, which included sufficient fibrous insulation to fill the strainer gaps and extend beyond the strainer perimeter, forming a "circumscribed" bed.

WF3 Response 3.f.7:

The GE hydraulic suction strainer design methodology is based on plant specific debris head loss testing. Debris head loss correlations were developed using the laboratory test results, scaled to the full plant design conditions.

The head loss is determined by summing up all the head loss components, as follows:

Head Loss= HLel, .is-plant +HL

+JJLde,,_plat, + HLchenical effect where:

Head Loss = maximum head loss of the strainer.

HLdebris plant - debris head loss at plant conditions.

HLclean-plant - clean head loss at plant conditions.

HLpipes&plenum = head loss on pipes and / or plenum.

HLchemicaieffect = head loss due to chemical effect.

WF3 Response 3.f.8:

The assumptions, margins and conservatisms are listed as follows (TDP-0186 - Reference 35):

o The flow rate is proportional to the perforated (sector test) or circumscribed (module) area of the strainers; o The debris load and flow rate are is distributed equally among the strainers; o The debris bed is uniform - same thickness throughout perforated surface; o In the debris load calculation, the circumscribed surface area of a plant installed strainer is the actual circumscribed surface area minus the portion of sacrificial area attributed to the circumscribed area; o 100% of particulate debris transported to the sumps is assumed to adhere to the strainers and contribute to head loss; to W3F1-2008-0069 Page 31 of 87 o All the labels and tags are modeled with 100% transport to the sump screen. The total sacrificial area is calculated by an equivalent to 100% of the original single sided surface area, counting for 0% overlap; o Due to extremely low approach and perforated flow velocities, laminar flow is assumed for debris head loss calculations; o Minimum water level at sump; o All coatings inside the ZOI are assumed to fail as particulate; o Head loss is calculated for indicated low end of sump water temperature and highest ECCS flow rate; o The upper circumscribed surface is assumed to be bounding in terms of air ingestion because air ingestion is evaluated at the top of the module, which is the closest surface to the water level.

WF3 Response 3.f.9:

Clean strainer system head loss is due to clean strainer head loss and plenum head loss.

Plant strainer clean head loss is calculated by scaling the test module clean head loss. Clean strainer head loss is due to the head loss inside the strainer discs, head loss as the flow exits the discs and enters the central cavity, and head loss inside the central cavity. The geometry of the test strainer is similar to that of the plant strainer. It is assumed that clean strainer head loss results primarily due to turbulent flow in the central cavity of the strainer, because the velocity through the perforated plates is relatively low and because water experiences an abrupt turn as it exits the discs and enters the central cavity. For central cavity strainers, assuming the gap width is the same, the scaling factor is based on the square ratio of the flow velocities at the entrance of the central cavity:

2 FlowRatePlantDisc dplant Headloss Clean := Headloss Test.Clean FlowRateTest dTest where:

Head IOSSCIean = plant strainer clean head loss Head IOSSTest.Clean = test strainer clean head loss FlowRatePlantDisc = plant disc flow rate, 34.599 gpm dPlant = plant central cavity diameter, 10.5 inches FlowRateTest = the test flow rate, varied by test dTest = test strainer central cavity diameter, varied by test Clean head loss data measured from the module test is the sum of module clean head loss, connecting pipe entrance head loss and dynamic head because the pressure transducer was installed inside the exiting piping just outside of the test module.

Results of the clean head loss and detail calculation can be found in the sizing report GENE-0000-0053-4416-P (Reference 36).

to W3F1-2008-0069 Page 32 of 87 Plenum head losses are due to the hydraulic losses associated with flow exiting the strainer into the ECCS sumps from the north and east strainers. Plenum losses for the Waterford 3 plenums are calculated in Appendix 2 of GENE-0000-0053-4416-P (Reference 36) to be 0.067 ft.

WF3 Response 3.f. 10:

Because containment sump water temperature following a LOCA is usually considerably greater than the temperature at which the hydraulic tests are run, debris head loss needs to be scaled to plant conditions as follows:

plant water densitYtest H~dbrt

tanttes = H Ldeb,.i _,te ** vis Cos ityp/att vicsty velocity p/att thicknesspi)t*(

H Ldei __,

plant

- vis Cos ity velocity,,,,

(debris, debris -thickness te water- densitypka,t ]

where:

HL = debris head loss through strainer in feet of water.

viscosity = dynamic viscosity of water in Ibm/ft-sec.

waterdensity = density of water in Ibm/ft 3.

velocity = approach velocity in ft/sec.

debris-thickness - nominal debris bed thickness in ft.

Nominal debris bed thickness is calculated as follows:

massfiber debris thickness:

densirtyfibe.

• perfbrated _ area where:

massfiber = mass of fiber debris in Ibm.

densityfiber = as-fabricated density of the fiber debris in Ibm/ft 3 .

perforatedarea = total surface area of the perforated plates in ft2.

The debris bed is assumed to be uniform, same thickness throughout perforated surface.

The sum of the strainer head loss and plenum head loss is tabulated below (GENE-0000-0053-4416-P - Reference 36).

Strainer.Head Plenum Head Total,.Head Loss:

• '"(ft)* : *,, .A ft), '

n-t)

S7-1S-59.2-CS_. 0.468 0.063 0.531 S7-2s8-10,0A-CS 0.463 0.063 0.526 S3-2M-1100-PS:11.,

@120 F 0.454 0.063 0.517 S3-2MM-'1,00PS.*,:

,. :2M-1,0,.F 0.314 0.063 0.377

@21F __ __ __ ___ __ __ __ __ __

to W3F1-2008-0069 Page 33 of 87 Note: The first two tests are sector test and the last two are module test. Design Basis head loss is based on module testing.

WF3 Response 3.f.11:

The strainers operate in fully submerged condition and are not vented to the atmosphere for any accident scenario. In addition to NPSH availability, failure criteria included the presence of vortexing or other forms of air entrainment, or the potential for a single large fiber' bed to blanket multiple strainers (during circumscribed bed formation) and block flow to some strainers.

Vortexing and air entrainment was not observed during several tests that mimicked the full range of plant debris loads, with either a representative water level or conservatively lowered water level. There is enough distance between strainers to preclude one large common fiber bed from obscuring some strainers; the nominal circumscribed debris bed is 7 inches under 100% fiber load conditions, and the closest distance between any two strainers is 32 inches, ensuring that fiber will not bridge between adjoining strainers and that water will have a path to each side of any individual strainer.

WF3 Response 3.f.12:

[Proprietary Information Removed]

The schematic of the pool configuration for the module test is shown in Figure 3.f-1.

[Proprietary Information Removed]

to W3F1-2008-0069 Page 34 of 87 Figure 3.f-lModule Test Setup for Strainer to W3F17-2008-0069 Page 35 of 87 WF3 Response 3.f.13:

Test strainer head loss is scaled based on velocity, viscosity, and bed thickness differences.

Debris head loss and clean strainer head loss are scaled independently.

The debris bed head loss results are scaled using the following equation:

? Plant hiplant -V plant Aplant ) tplant h/test V test (Qtest ttest A

Atest)/

Where:

hi = Debris Bed Head Loss (ft.)

v = Water Viscosity (Ibm/sec-ft)

Q = Sump Flow rate (ft3/s)

A = Perforated Area of strainer(s) (ft) (Does not include top and bottom external surfaces) t Debris bed thickness on perforated area (in.)

Testing was performed at a temperature less than plant temperature. The reduced test temperature results in an increase in viscosity. This difference in viscosity is accounted for by the first term in the equation above. The test head loss is multiplied by the ratio of plant water viscosity to test water viscosity, along with the other terms in the equation, to provide a test head loss that is representative of the plant conditions.

Viscosity scaling was performed for sector tests $7-2S-100-CS and S7-15-59.2-CS, and for module test S3-2M-100-PS. Boreholes were not present in these tests based on the test vendor's report.

Presence of boreholes in the debris bed is apparent in the photographs of the disassembled test sector taken after the test S7-4S-1 3.8A-CS. Viscosity scaling was not applied to this test. This is the 1/8-inch bed case for break S7. Clumps of debris are seen on the debris plate on other areas of the strainer. The strainer used a debris plate, which is intended to mitigate thin bed effects on head loss.

WF3 Response 3.f.14:

Containment accident pressure was not credited in evaluating whether flashing would occur across the strainer surface.

to W3F1-2008-0069 Page 36 of 87 3.g Net Positive Suction Head (NPSH)

NRC Issue 3.q The objective of the NPSH section is to calculate the NPSH margin for the ECCS and CSS pumps that would exist during a loss-of-coolant accident (LOCA) considering a spectrum of break sizes.

1. Provide applicable pump flow rates, the total recirculation sump flow rate, sump temperature(s), and minimum containment water level.

2. Describe the assumptions used in the calculations for the above parameters and the sources/bases of the assumptions.

3. Provide the basis for the required NPSH values, e.g., three percent head drop or other criterion.

4.' Describe how friction and other flow losses are accounted for.

5. Describe the system response scenarios for LBLOCA and SBLOCAs.

6. Describe the operational status for each ECCS and CSS pump before and after the initiation of recirculation.

7. Describe the single failure assumptions relevant to pump operation and sump performance.

8. Describe how the containment sump water level is determined.

9. Provide assumptions that are included in the analysis to ensure a minimum (conservative) water level is used in determining NPSH margin.

10. Describe whether and how the following volumes have been accounted for in pool level calculations: empty spray pipe, water droplets, condensation and holdup on horizontal and vertical surfaces. If any are not accounted for, explain why,

11. Provide assumptions (and their bases) as to what equipment will displace water resulting in higher pool level.

12. Provide assumptions (and their bases) as to what water sources provide pool volume and how much volume is from each source.

13. If credit is taken for containment accident pressure in determining available NPSH, provide description of the calculation of containment accident pressure used in determining the available NPSH.

14. Provide assumptions made which minimize the containment accident pressure and maximize the sump water temperature.

15. Specify whether the containment accident pressure is set at the vapor pressure correspondingto the sump liquid temperature.

16. Provide the NPSH margin results for pumps taking suction from the sump in recirculation mode.

WF3 Response 3.q.1:

Pump Flow Rates Item Injection Flow (pre RAS) Recirculation Flow (post RAS)

HPSI Pump! 985 gpm 985 gpm LPSI Pump 5650 gpm 0 gpm CS Pump 2250 gpm 2250 gpm Per Train Sump 8885 gpm 3235 gpm Table: 3.g-1: Applicable Pump flow Rates to W3F1-2008-0069 Page 37 of 87 Notes: (1) Waterford 3 has a common sump for both ECCS/CS trains (2) Pump flow rates are run-out Values Sump Temperature An analysis was performed to determine a maximum Sump Fluid temperature profile for a period of 30 days post LOCA (Figure 3.g.1-1). The Analysis was performed using the GOTHIC 7.0 program package. Conservative assumptions made in the development of the profile as stated in section 3.g.2.

2H,0.00 250.00 ...-

.I-.-.

V'aporTen p.

III7-Et Iri iii 2*'0.00 I I IA II IIN _ _ I 1 1 111111 1 1 11111 1 1 1 1 11 11 1 1 1 1 1 1 11 2M .00 220.00 1"x 111111 N NS 1

210.00 I 0.00 170.00 1E0.00 SunipTenop 1503.00 _lRecirculation Accuation

'140.00 130.00 120 .00 110.00 1C3.00 10.00 100.00 10000,00 100000.00 1.00 100O000.0 10,000000o.C Time CNec)

Figure 3.g.1-1: Safety Injection Sump Maximum Temperature Profile Time (Sec) Sump Temp.

(degF)

RAS 3253 173 RAS + 30 min 5053 187 Peak 24716 219 LOCA + 1 Day 86400 198 LOCA + 2 Day 172800 188 LOCA + 5 Day 432000 173 LOCA + 10 Day 864000 156 LOCA + 30 Day 2592000 152 Table 3.g.1-2: Temperature Points of Interest to W3F1-2008-0069 Page 38 of 87 Minimum Water Level The minimum water level in containment for a SBLOCA is (-)5.79 ft MSL. The minimum water level in containment for a LBLOCA is (-)5.29' MSL. These values are were determined in calculation MNQ6-4 (Reference 37). The calculation was revised to address items identified during and prior to the NRC GSI-191 audit of Waterford 3. Discussion of the analysis used to determine the minimum water levels can be found in section 3.g.8 of this report.

WF3 Response 3.q.2:

Flow Rate Assumptions

" All pump flow rates are run-out values. This maximizes suction loses which minimizes available NPSH.

" Both trains of ECCS and CS are in operation to maximize flow through common sump screen.

Temperature Assumptions

• Base input data is that used for licensing basis LOCA peak 24 hr pressure analysis with added uncertainties

• Only one train of Containment Spray in operation at design flow

• RWSP at conservative high temperature

• 1 out of 4 Containment Fan Coolers in Operation

• Shutdown Cooling Heat Exchanger parameters selected to minimize heat rejection to Ultimate Heat Sink

• Minimum Safety Injection Flow Minimum Water Level Assumptions

• Assumptions for the Minimum water level analysis are discussed in section 3.g.9 WF3 Response 3.q.3:

The required NPSH values are taken from the vendor certified pump performance curves and associated test data. Waterford 3 calculation ECM07-001 (Reference 51) uses a least squares curve fit polynomial to extrapolate the required NPSH curve data out to the pump's run-out flow rate.

WF3 Response 3.q.4:

Friction loss is being determined using the software program Pipe-FLO. Pipe-FLO performs steady state hydraulic analysis of fluid filled piping system using standard industry approved methods such as defined by Crane Technical Paper No. 410. All system configurations are being modeled for the suction side of the pumps and the system configuration resulting in the smallest NPSH available being used to determine acceptable screen head loss. Vendor supplied flow performance data is being utilized for components such as valves when available.

When vendor information was not available, conservative assumptions are being made using standard data from Crane Technical Paper No. 410. As stated in section 3.g.2, pump run-out flows are being utilized in all analysis to maximize friction loss.

Calculations are being performed for a maximum saturated sump water temperature of 210 degF. This temperature was determined to be the most limiting based on the significant to W3Fl-2008-0069 Page 39 of 87 increase in vapor head below 210 degF which is slightly below the saturation temperature for containment; which is initially at minimum pressure of 14.275 psia prior to the Loss of Coolant Accident (LOCA) event. The increase in vapor head more than compensates for the increase in piping/component losses due to the increase in viscosity of the fluid at temperatures below 210 degF. At temperatures above 210 degF, the increase in vapor pressure is offset by the conservative assumption that containment pressure is equal to the vapor pressure of the sump water which eliminates containment air pressure as a contributor to NPSH available. Higher temperatures also decrease water viscosity which decreases friction losses and further improves NPSH available.

The above methodology is consistent with Regulatory Guide 1.1 in that no credit is being taken for accident over pressure in containment.

WF3 Response 3.q.5:

Safety Injection System Response Scenarios for LBLOCA and SBLOCA The Safety Injection System (SIS) is arranged with two independent redundant trains, each functionally identical to the other and normally aligned to the Refueling Water Storage Pool (RWSP). The SIS is activated by the Safety Injection Actuation Signal (SIAS) which is initiated by either low pressurizer pressure or high containment pressure. The SIAS automatically starts the High Pressure Safety Injection (HPSI) and Low Pressure Safety Injection (LPSI) pumps and opens the motor operated valves (MOVs) that provide a flow path from the discharge of these pumps to the Reactor Coolant System. These MOVs don't actually go full open, but open to a preset position (set by valve position limit switch adjustment) to achieve a balanced flow and to prevent pump run-out. An installed spare HPSI pump is available which can be aligned to replace either of the other two HPSI pumps.

The HPSI system responds to an SIAS by automatically starting the aligned HPSI pumps and opening the cold leg injection flow control valves. If RCS pressure has not fallen below the 1450 PSIG shutoff head of the HPSI pumps, the system operates on recirculation flow until pressure decreases. As pressure decreases; HPSI flow will initiate and continue to increase as pressure falls.

The LPSI system responds by automatically starting the LPSI pumps and opening the cold leg injection flow control valves. The system will operate on recirculation flow until RCS pressure

• drops below the shutoff head of the pumps.

When a low level (10%) is sensed in the RWSP, the recirculation mode is initiated by the Recirculation Actuation Signal (RAS). At this time the HPSI pump suction is diverted to the Safety Injection Sump and the LPSI pumps are stopped.

Simultaneous hot and cold leg injection is used for both small break and large break LOCAs at 2-3 hours after the start of the LOCA and the RCS is not filled. In this mode, the HPSI pumps discharge lines are realigned so that the total injection flow is divided equally between the hot and cold legs. Simultaneous injection into the hot and cold legs is used as the mechanism to prevent the precipitation of boric acid in the reactor vessel following a break that is too large to allow the RCS to refill. Injecting to both sides of the reactor vessel ensures that fluid from the reactor vessel (where the boric acid is being concentrated) flows out the break regardless of the to W3F1-2008-0069 Page 40 of 87 break location and is replenished with a dilute solution of borated water from the other side of the reactor vessel.

Action is taken no sooner than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after the LOCA, since the fluid injected to the hot leg may be entrained in the steam being released from the core. After 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, the core decay heat has dropped sufficiently so that there is insufficient steam velocity to entrain the fluid being injected to the hot leg. Action is taken no later than 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> after the LOCA, in order to ensure that the buildup of boric acid is terminated, well before the potential for boric acid precipitation occurs (approximately 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />). Even though the action is required only for large breaks, it is taken for any LOCA so that the operator need not be required to distinguish between large and small breaks so early in the transient. Simultaneous hot and cold leg injection is not required for small breaks because, for small breaks, the buildup of boric acid is terminated when the RCS is refilled. Once the RCS is refilled, the boric acid is dispersed throughout the RCS via natural circulation.

Hot leg injection is established by closing the HPSI header flow orifice bypass valves and opening the hot leg header isolation valves. The orifices are preset to establish a 50% +/-5% flow balance between the hot and cold leg injection headers while preventing pump run-out conditions.

Long term cooling is initiated when the core is reflooded after a LOCA and is continued until the plant is secured. Two basic modes of long term cooling are available to the operator.

Entry into shutdown cooling (SDC) may be necessary if steam generator heat removal is lost, for certain sized breaks (small breaks). The shutdown cooling system is utilized if certain plant conditions exist.

When possible, the time necessary to refill the RCS and regain control of pressure and inventory depends on break size, break location, RCS cooldown rate and the number of HPSI pumps and charging pumps actuated. With only one HPSI pump actuated, for a break of about 3 inch diameter located on the bottom of the cold leg, it may take as long as 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to refill the RCS. With all injection pumps operable, the time is about 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

Before SDC is operated, RCS activity levels must be determined since the RCS fluid will be circulated outside of the containment building. When high activity is present, circulation outside containment has the potential for release to the environment. If potential for significant releases exists, it may be more desirable to continue cooling with the steam generator. The condensate inventory must be checked to ensure that the supply is sufficient to cool down the plant.

If SDC operation is determined to be appropriate, the SIS is aligned for cold leg injection and the RCS is cooled down and depressurized to allow entry into shutdown cooling.

If SDC operation is not appropriate or if the system is not available, it is desirable to continue RCS heat removal via the steam generators until no further steam is generated.

For large breaks, simultaneous injection provides effective long-term cooling by inducing a flushing flow through the core which will eventually result in a subcooled core. The core cooling is actually provided by the Containment Spray System via the Shutdown Cooling Heat Exchangers. This provides cooling for the Safety Injection Sump water which also provides a water source to the HPSI Pumps.

to W3F1-2008-0069 Page 41 of 87 CSS Response Scenarios to LBLOCA and SBLOOCA The Containment Spray (CS) System consists of two independent and duplicate trains to achieve the required redundancy. One loop operating alone is capable of providing the necessary post-accident heat removal. Each loop contains a CS Pump, a CS Riser Pump, a Shutdown Cooling Heat Exchanger, four spray ring headers, 116 spray nozzles, and the controls and instrumentation necessary to provide for proper system operation.

The CS System is actuated when the SIAS and the High-High Containment Pressure signal are coincident. This generates a Containment Spray Actuation Signal (CSAS) which opens the CS header isolation valves and starts the CS Pumps. The pumps initially take suction from the RWSP through a common header with the SIS pumps and delivers borated water to the spray nozzles located in the top portion of the steel containment. When RAS is initiated the CS pumps continue operation with the suction being taken from the Safety injection Sump.

WF3 Response 3.q.6:

Pump Pre RAS (Injection Post RAS (Recirculation Phase) Phase)

HPSI Operating Operating LPSI Operating Secured CS Operating Operating Table 3.g.6-1: Operational Status of ECCS and CS Pumps WF3 Response 3.q.7:

Only one single failure is being analyzed for NPSH effects. This single failure is a LPSI pump failing to trip upon receiving a Recirculation Actuation Signal. This results in an increased flow through the safety injection sump for an assumed time period of 30 minutes. Upon initiation of the Recirculation Actuation Signal, Operations procedures guide operators to verify that the LPSI pumps have stopped. If the pump has not tripped upon RAS, operators will take appropriate action as necessary to secure the pump.

Failure of a flow control valve will have no effect on the NPSH analysis due to the fact that the analysis assumes pumps are operating at run-out flows.

WF3 Response 3.g.8:

The minimum Safety injection Sump water level is determined by comparing water inventories available to fill the sump with the physical layout of the sump. Conservative assumptions intended to minimize the water level are made concerning available inventories, hold-up mechanisms, and the sump physical layout. Two single worst case (i.e., not time dependent) water levels are determined; one for the SBLOCA scenario and one for the LBLOCA scenario.

The results of the water level analysis can be found in section 3.g.1 while the additional detail on the analysis can be found in sections 3.g.9, .10, .11, & .12. The original minimum water level analysis was revised due to comments during the 2007 NRC GSI-191 Audit at Waterford 3.

to W3F1-2008-0069 Page 42 of 87 WF3 Response 3.q.9:

• Containment sump and Safety Injection sump do not communicate due to clogged drains.

" Containment sump fills to elevation of (+)7.5 ft (point of overflow to safety injection sump).

• Steam volume in containment at max containment temperature and is saturated.

• Refueling cavity assumed to not holdup water due to existence of two 6" floor drains in locations which are unlikely to clog. Drains go directly to Containment floor.

• Safeguards pump are assumed to leak a combined total of 0.5 gpm for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

• Film thickness for condensation assumed to be larger than that determined analytically.

Other assumptions are described in sections 3.g.10, .11, & .12.

WF3 Response 3.q.10:

Empty spray pipe - A portion of the Containment Spray System piping is empty prior to initiation of Containment Spray. This piping is credited for consuming an appropriate portion of the available inventory.

Water droplets - Containment Spray droplets are in transient from the spray nozzles to the containment floor. The volume of spray droplets for two CS trains operating at pump run-out flows are credited for consuming an appropriate portion of the available inventory.

Condensation - Heat sink surfaces condense steam from the atmosphere and develop a condensation film. The volume of the condensation film is credited for consuming an appropriate portion of the available inventory.

Holdup on horizontal and vertical surfaces - Other than condensation on heat sink surfaces, holdup was not considered on other horizontal or vertical surfaces. All concrete flooring above the containment floor has drains or is adjacent to steel grating with no curb to trap water. Floor drains direct water to the containment sump which is assumed to fill completely and overflow into the safety injection sump.

WF3 Response 3.q.11:

Credit was only taken for structural concrete below the flood level for the displacement of water.

Conservatively no credit is taken for the following items:

• Reactor Drain Tank

• Grout Pads /

• TSP Baskets

• Strainer Steel

• RCP support steel

• Structural steel.

WF3 Response 3.q.12:

The following assumptions are made for the sources of water:

0 Technical Specification minimum protected RWSP Volume only injected to W3F1-2008-0069 Page 43 of 87

• RWSP water at maximum Technical Specification temperature

• Safety injection tanks credited for Large Break LOCA only o Water at minimum Technical Specification Level o Water at maximum pre-accident Containment Temperature

• No credit for Reactor Coolant Spillage

• No credit for charging flow from Volume Control Tank or Boric Acid Makeup ranks WF3 Response 3.q.13:

No credit is taken for containment pressure above that present prior to the onset of the accident.

This is consistent with Regulatory Guide 1.1 (Reference 52).

WF3 Response 3.q.14:

As stated in section 3.g.13, no credit has been taken for containment pressure above that present prior to the onset of the accident.

WF3 Response 3.q.15:

As stated in section 3.g.4, NPSH available values are being calculated with a sump temperature of 210 degF with vapor pressure equal to saturation pressure. For the cases that assume a single LPSI pump fails to trip, credit is taken for the sump fluid being sub cooled at 190 degF with respect to containment being at atmospheric pressure.

WF3 Response 3.q.16:

The NPSH margin for the HPSI and CS pumps when taking suction from the sump in recirculation mode is determined in calculation ECM07-001 (Reference 51). The minimum margins during the recirculation mode of operation determined in the calculation are:

Pump Margin w/o Strainer strainer losses (ft) losses (ft)

HPSI A 4.426 0.377 HPSI A/B 3.929 0.377 (A Train)

HPSI B 1.966 0.377 HPSI A/B 3.96 0.377 (B Train)

CS A 5.963 0.377 CS B 5.818 0.377 Table: 3.g.16-1: Recirculation at 210F. No LPSI operating to W3F1-2008-0069 Page 44 of 87 Pump Margin w/o Strainer strainer losses (ft) losses (ft)

HPSI A/B 11.429 0.435 (A Train)

HPSI B 13.257 0.435 LPSI A 9.05 0.435 CS A 13.464 0.435 CS B 17.11 0.435 Table: 3.g.16-2: Recirculation at 190F w/ LPSI failure (Worst case lineup) 3.h Coatings Evaluation NRC Issue 3.h The objective of the coatings evaluation section is to determine the plant-specific ZOI and debris characteristicsfor coatings for use in determining the eventual contribution of coatings to overall head loss at the sump screen.

1. Provide a summary of type(s) of coating systems used in containment, e.g., Carboline CZ 11 InorganicZinc primer,Ameron 90 epoxy finish coat.

2. Describe and provide bases for assumptions made in post-LOCA paint debris transport analysis.

3. Discuss suction strainer head loss testing performed as it relates to both qualified and unqualifiedcoatings and what surrogate material was used to simulate coatings debris.

4. Provide bases for the choice of surrogates.

5. Describe and provide bases for coatings debris generation assumptions. For example, describe how the quantity of paint debris was determined based on ZOI size for qualified and unqualifiedcoatings.

6. Describe what debris characteristics were assumed, i.e., chips, particulate, size distribution and provide bases for the assumptions.

7. Describe any ongoing containment coating condition assessmentprogram.

WF3 Response 3.h.1:

The following types of coating systems are present, or approved to be used, inside Containment, per Waterford 3 Engineering Procedure NOECP-451 and Specification 1564.734 (References 42 and 43).

" Ameron Dimetcote 6(N)

" Ameron Amercoat 66 over Ameron Nu-Klad 11OAA primer

" Ameron Amercoat 66 over Ameron Nu-Klad 114 primer

" Ameron Amercoat 90

" Ameron Amercoat 90 over Ameron Amercoat 66 primer

" Ameron Amercoat 90 over Ameron Amercoat 71 primer

" Ameron Amercoat 90 over Ameron Dimetcote 6(N) primer

" Ameron Amercoat 90 over Ameron Dimetcote E-Z primer

. Ameron Amercoat 90 over Carboline CZ1 1SG primer

" Ameron Amerlock 400 NT to W3F1-2008-0069 Page 45 of 87

" Ameron Dimetcote E-Z

" Carboline 801

" Carboline 890

" Carboline 890 over Carboline Nutec 11 S primer

" Carboline 890 over Carboline Nutec 11 primer

" Carboline 890 over Carboline Nutec 1201 primer

" Carboline Carbo-Zinc 11

" Carboline CZ1 1 SG

" Carboline Nutec 1201 over Carboline Nutec 11 primer

" Carboline Nutec 1201 over Carboline Nutec 11S primer

" Carboline Phenoline 305 over Carboline CZ1 1 SG primer

" Carboline Phenoline 305 over Carboline 191 primer

" Carboline Phenoline 305 over Carboline Phenoline 305 primer

" Tnemec 801

" Unqualified coatings (alkyds, enamels, and epoxies) from various manufacturers.

WF3 Response 3.h.2:

In accordance with the guidance provided by NEI Guidance 04-07 and the SER for NEI 04-07 (References 2 and 3), all qualified coating debris within the ZOI is considered particulate and as such is modeled as transporting to the sump strainer.

Unqualified zinc coatings and indeterminate coatings are considered to fail as particles with 100% transport to the strainer.

50% of the qualified coatings on the containment liner dome and the liner above elevation 112' are assumed to fail. This is a conservative number based on coating failures at Waterford 3 to date. Degraded qualified epoxy coating systems are considered to fail as chips (see response to 3.h.6 below) and are subject to settling in low velocity area of the pool such that only a portion of the generated debris transport to the strainer. Degraded qualified coatings that fall on or near the strainer are considered not to have a chance to settle. Conservatively 20% of degraded qualified coatings are considered to fall on or near the strainer and thus transport to the strainer.

For analysis of the remaining degraded qualified coating transport additional CFD runs were performed. The additional CFD simulations consider break locations farther from the strainer in order to maximize the portion of the pool where flow velocity is too high for settling to occur.

The study in NUREG/CR-6916 (Reference 19) found that the lowest incipient tumbling velocity, the velocity at which the coating chips similar to Waterford 3 debris would move on the floor was 0.264 feet per second for the "curled" 1-to-2-inch chips. Conservatively a transport velocity of 0.2 feet per second is used for all chips with a size greater than 1 /6 4 th inch. Based on the CFD simulations the bounding portion of the cOntainment pool area with a velocity in excess of 0.2 feet per second is determined to be 12.7%. As a judgment, this area is increased to 15% and the failed coatings in the remaining 85% of the pool are considered subject to settling.

WF3 Response 3.h.3 and 3.h.4:

The prior Waterford 3 sector test results indicated that the bounding condition for simulating plant LOCA debris-generation is 100% of the transported fiber and particulate (GENE-0000-0053-4416-P - Reference 36).

to W3F.1 -2008-0069 Page 46 of 87

[Proprietary Information Removed]

This will have a greater effect on head loss than would the heavier latent dirt.

WF3 Response 3.h.5:

In order to determine the amount of qualified coating debris generated at Waterford 3, structural and civil drawings are consulted. The bounding break location is determined from inspection of these drawings, then the total surface area of coated steel and concrete within a 4D ZOI of the break location is calculated. The maximum allowable coating thickness, per the plant coating specification, is then applied to this surface area to determine the total coating debris volume. A spherical ZOI of 4D for qualified coatings was selected based on WCAP-16568-P (Reference 22). This testing concluded that a spherical ZOI of 4D is conservative for the qualified epoxy and the qualified zinc coatings used by Waterford 3.

Unqualified (degraded qualified or indeterminate) coatings are assumed to 100% fail. The unqualified coating debris volume is based on the thickness of similar coatings on other materials in containment.

WF3 Response 3.h.6:

In accordance with the guidance provided in NEI Guidance 04-07 (Reference 2) and the SER for NEI 04-07 (Reference 3), all qualified coating debris within the ZOI, unqualified zinc coating debris, and indeterminate coating debris are treated as particulate and are therefore transported entirely to the sump strainer.

Degraded qualified coatings are. considered to fail as chips with a size distribution per Alion document ALION-REP-TXU-4464-02 (Reference 25) and letter TXX-07156 (Reference 26).

The document ALION-REP-TXU-4464-02 (Reference 25) stated that 49.5% of coating particles were less than 1 /8th inch in size. Letter TXX-07156 (Reference 26) further identifies that 12.375% (25% of 49.5%) of the coating particles are 6 mils (0.006 inches) and 37.125% (75% of 49.5%) are 15.6 mils ( 1/ 6 4th inch). 100% of coating particles with a size less than 1 /6 4 th inch will not settle and will transport to the sump. This quantity amounts to approximately 12.375% of the inventory. The remaining 87.625% of the inventory may settle in favorable flow conditions.

The degraded qualified coating systems at Waterford 3 are compared with the test data using NUREG/CR-6916, ALION-REP-TXU-4464-02, TXX-07156, and CCCL letter dated 9/20/07 (references 19, 25, 26 and 27). The data reported in NUREG/CR-6916 (Reference 19) are for to W3F1-2008-0069 Page 47 of 87 the failure characteristics of many coatings, including epoxy applied as a topcoat over inorganic-zinc. The epoxy top coats applied over a zinc rich primer, Category ZE, are similar to the Carboline Phenoline 305 applied as a topcoat over Carboline Carbo-Zinc 11. The painting system used on the containment dome is Carboline Phenoline 305 applied as a topcoat over Carboline Carbo-Zinc 11 primer. This system is also one of four Service Level 1 paint systems approved for the containment liner. The other systems include Carboline 305 topcoat over Carboline 191 primer, Carboline 801 as a primer and topcoat and Amerlock 400 NT (a one coat system). All are epoxy systems, which are expected to exhibit failure characteristics of Carboline 305. The CCCL letter dated 9/20/07 (Reference 27).confirms that the size distribution presented in ALION-REP-TXU-4464-02 and TXX-07156 (references 25 and 26) is applicable to Carboline Phenoline 305 coatings systems. The CCCL letter dated 9/20/07 (Reference 27) also confirms that 100% of inorganic-zinc coatings will fail as small fines.

WF3 Response 3.h.7:

Waterford 3 performs an inspection of containment coatings each refueling outage. As defined in procedure NOECP-451 (Reference 42), the scope of the coating inspections are coated concrete and steel surfaces inside the SI Sump, the containment liner plates, and approximately 10% of the remaining coated surfaces excluding concrete and insulated piping.

3.i Debris Source Term NRC Issue 3.i 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 recirculationfunctions.

Provide the information requested in GL 04-02 Requested Information Item 2. (0 regarding programmaticcontrols taken to limit debris sources in containment.

GL 2004-02 Requested Information Item 2(f)

A description of the existing or planned programmatic controls that will ensure that potential sources of debris introduced into containment (e.g., insulations, signs, coatings, and foreign materials) will be assessed for potential adverse effects on the ECCS and CSS recirculation functions. Addressees may reference their responses to GL 98-04, A Potential for Degradationof the Emergency Core Cooling System and the Containment Spray System after a Loss-of-Coolant Accident Because of Construction and Protective Coating Deficiencies and Foreign Material in Containment," to the extent that their responses address these specific foreign materialcontrol issues.

In responding to GL 2004 Requested Information Item 2(f), provide the following:

1. A summary of the containment housekeeping programmatic controls in place to control or reduce the latent debris burden. Specifically for RMI/low-fiber plants, provide a description of programmatic controls to maintain the latent debris fiber source term into the future to ensure assumptions and conclusions regardinginability to form a thin bed of fibrous debris remain valid.

to W3F1-2008-0069 Page 48 of 87

2. A summary of the foreign material exclusion programmatic controls in place to control the introduction of foreign material into the containment.

3. A description of how permanent plant changes inside containment are programmatically controlled so as to not change the analytical assumptions and numerical inputs of the licensee analyses supporting the conclusion that the reactorplant remains in compliance with 10 CFR 50.46 and related regulatory requirements.

4. A description of how maintenance activities including associated temporary changes are assessed and managed in accordance with the Maintenance Rule, 10 CFR 50.65.

If any of the following suggested design and operational refinements given in the guidance report (guidance report, Section 5) and SE (SE, Section 5.1) were used, summarize the applicationof the refinements.

5. Recent or planned insulation change-outs in the containment which will reduce the debris burden at the sump strainers

6. Any actions taken to modify existing insulation (e.g., jacketing or banding) to reduce the debris burden at the sump strainers

7. Modifications to equipment or systems conducted to reduce the debris burden at the sump strainers

8. Actions taken to modify or improve the containment coatings program WF3 Response 3.i.1 Entergy's fleet wide FME program procedure provides the requirements and guidance to prevent and control introduction of foreign materials into structures, systems, and components.

Also included within this procedure are steps to take to reestablish and maintain FME areas to prevent foreign material intrusion and to recover/monitor when a loss of FME integrity has occurred.

Housekeeping and foreign material assessments after a plant outage and prior to heat up are performed at the direction of the Waterford 3 operating procedure which provides the requirements and guidance to perform walkdowns of the RCB to assess debris that may represent a risk of blocking the SI recirculation sump screen.

WF3 Response 3.i.2 See Response to Issue 3.i.1.

WF3 Response 3.i.3 The Entergy fleet configuration control procedure controls permanent plant changes inside the RCB so as to not change the analytical assumptions and numerical inputs. A design input consideration was added to the Entergy fleet design input screening procedures to specifically address the SI Sump GL 2004-02 program. Engineers are required to review the impact of a proposed change to determine if there would be an impact to the performance of the ECCS sump. Waterford 3 and/or Entergy fleet procedures require reviews of physical changes in the RCB to address specific areas. The specific areas that are addressed, as a minimum, are:

• Insulation inside containment,

° Coatings inside containment, to W3F1-2008-0069 Page 49 of 87

" Volumes in containment,

" Addition of materials inside containment that may produce chemical effects in the post-LOCA flood pool/environment.

WF3 Response 3.i.4 Temporary changes at Waterford 3 are subject to the same requirements for reviews as for permanent changes. Therefore, the design input and impact screenings will determine if the temporary change should be reviewed for any potential impact on the SI Sump screens.

Entergy fleet procedures also provide guidance such as the 50.59 Review Process procedure, which provides details and guidance on maintenance activities; and the On-Line Work Control Process procedure, which establishes the administrative controls for performing on-line maintenance of SSCs in order to enhance overall plant safety and reliability.

WF3 Response 3.i.5 There are no recent or planned insulation change-outs in the Waterford 3 containment which will reduce the debris burden at the sump strainers.

WF3 Response 3.i.6 No modifications to existing insulation were performed to reduce the debris burden at the sump strainers.

WF3 Response 3.i.7 There were no modifications made to equipment or systems to reduce the debris burden at the sump strainers.

WF3 Response 3.i.8 The coatings procedure for Waterford 3 was revised to provide better instructions to the craft on cleanliness when preparing surfaces.

3.j Screen Modification Package NRC Issue 3.1 The objective of the screen modification package section is to provide a basic description of the sump screen modification.

1. Provide a description of the majorfeatures of the sump screen design modification.

2. Provide a list of any modifications, such as reroute of piping and other components, relocation of supports, addition of whip restraints and missile shields, etc., necessitated by the sump strainermodifications.

WF3 Response 3.i.1 The modification for the SI Sump replaced the original box type screen over the SI Sump. To prevent debris from entering the open sump, the original rectangular box shaped screen had to W3F1-2008-0069 Page 50 of 87 0.078 inch square openings (0.11 inch diagonally) and completely covered the sump inlet. The box screen provided approximately 198.2 ft 2 of available flow area. Inaddition, the box screen was mounted approximately 3.75" above the containment floor which helped to prevent sediment from entering the pit. A divider screen separated the two (SIS) suction lines located in the sump. During RF14, modification ER-W3-2003-0394-001 (Reference 30) installed a passive, safety-related, Nuclear Modularized Stacked Disk Strainer assembly engineered and manufactured by General Electric Energy in place of the original screen. The new strainer arrangement for Waterford 3 consists of 11 strainer modules mounted on top of the plenum mounted over the existing'sump and over the concrete floor to the north and to the east of the sump. The new SI Sump partition that separates the two SIS suction lines is a section of stainless steel grating made of 1" x 1/8" bars separated at 1-3/16". The partition is supported by angles attached to the existing anchor plates. The modification was installed during the 2006 refueling outage.

The effective surface area of the new strainer for each module is 336.3 ft 2, for a total of approximately 3,700 ft2 . There are 11 essentially identical modules mounted on the 8" high

.plenum over the SI Sump. Each module is bolted to the plenum and the plenum is bolted to the containment floor. The plenum prevents debris from entering the system between the modules.

Each module is constructed of 17 stacked perforated disk sets with hole-diameters of 0.093 inch. A disk set is composed of two perforated disks separated from each other by radial fingers and by an outer support, the finger frame. The water enters from the top and bottom disks into the intermediate space, travels towards the center and then axially towards the strainer base. Perforated inner spacer rings separate the disk sets from each other. The modules are located on top of the plenum, approximately 8 inches above the containment floor.

The sump is now totally enclosed by the plenum, preventing material from falling directly into the sump without passing through the strainer assemblies.

The plenum extensions to the north and east side of the sump have internal dimensions 7.25 inches high by 41 inches wide.

The plenum has openings in the top to admit flow of strained water from the modules. The modules are bolted to the plenum, which in turn is bolted to the containment slab. The plenum is made of structural shapes: angles and plates. The strainer design allows for disassembly, replacement of modules, or addition of future modules as needed. The plenum has two access openings to allow access into either side of the sump during outages for inspection and testing.

The access openings are approximately 40" X 40". The access covers are bolted to the plenum to control access. Each module also has an inspection port on top to allow visual inspections inside the module, if necessary.

WF3 Response 3.J.2 A safety related low level switch was relocated due to the installation of the new SIS sump strainer assembly. The switch housing is now mounted on top of the plenum and is on the north end of the sump so that it can detect water at the lowest point of the SIS sump floor. The switch elevation at the new location is identical to the previous elevation and maintains the same function and switch setpoint. This switch is seismically mounted on the top of the plenum. The level switch guard pipe and mounting plate were modified to match the plenum mounting plate and are seismically installed to preclude the potential for a strainer bypass path. The seismic bracing mounted at the sump floor was relocated to align with the guard pipe. The portion of to W3F1-2008-0069 Page 51 of 87 the new guard pipe above the plenum is welded to 150-lb flange and welded to the bolted cover plate. This portion of guard pipe above the plenum has no holes on the pipe wall. The instrument plate-to-plate bolted connection to strainer plenum has a zero gap preventing debris from entering the sump.

Nineteen (19) TSP baskets were relocated to eliminate interference with the new strainer assembly. The new location is at the same elevation of the containment, but at the north end of the building. The relocated TSP baskets are seismically mounted on the concrete floor of containment at elevation -11.0 feet.

The sensors for Reg. Guide 1.97 Type B, Category 1 Level transmitters were temporarily removed from the' mounting plate inside the sump to allow for installation of the new strainer..

After installation of the new strainer was completed the sensors were remounted to the same mounting plate inside the sump. There was no change to the transmitter mounting and the setpoint. The 1/4" diameter capillary tubing and tube track between the level transmitters and the sensors was re-routed. The capillary tubing penetrates the top of the plenum. A plenum opening of 6" x 6" with a W"diameter slotted hole and a cover plate with a slotted hole was provided to allow the re-mounting of the instrument without disconnecting the capillary tubing from the, instrument. The slotted hole design ensures zero gap thereby preventing debris from entering the sump.

3.k Sump Structural Analysis NRC Issue 3.k The objective of the sump structuralanalysis section is to verify the structuraladequacy of the sump strainerincluding seismic loads and loads due to differential pressure, missiles, and jet forces. Provide the information requested in GL 2004-02 Requested Information Item 2(d)(vii).,

GL 2004-02 Requested Information Item 2(d)(vii)

Verification that the strength of the trash racks is adequate to protect the debris screens from missiles and other large debris. The submittal should also provide verification that the trash racks and sump screens are capable of withstanding the loads imposed by expanding jets, missiles, the accumulation of debris, and pressure differentials caused by post-LOCA blockage under flow conditions.

1. Summarize the design inputs, design codes, loads, and load ,combinations utilized for the sump strainerstructuralanalysis.

2. Summarize the structural qualification results and design margins for the various components of the sump strainerstructuralassembly.

• 3. Summarize the evaluations performed for dynamic effects such as pipe whip, jet impingement, and missile impacts associated with high-energy line breaks (as applicable).

4. If a backflushing strategy is credited, provide a summary statement regarding the sump strainerstructuralanalysis considering reverse flow.

WF3 Response 3.k.1 The inputs and loads are discussed in the following paragraphs.

to W3F1-2008-0069 Page 52 of 87 Differential Crush Pressure is the pressure difference across the strainer components. The value is equal to the static pressure outside the strainer minus the static pressure inside the strainer system. The "Design" crush pressure is analogous to design pressure for a pressure vessel; in. that it is > the limiting pressure loss across the equipment specified for hydraulic system design. Crush pressure was applied to the assembly to demonstrate adequacy for pressure loading of the strainer perforated plates, perforated spacer rings and plenum plates including supports.

Equivalent solid plate properties (Poisson's Ratio and Modulus of Elasticity) including a stress multiplier were determined for the disc perforated plates by performing a finite element analysis of perforated plate and solid plate following the guidance contained in the ASME Code Section Ill, Appendix A, Article A-8000. A finite element model of the strainer assembly was then developed with the perforated plates modeled as equivalent solid plates. The strainer disks are modeled as shell and beams while the plenum consists of solid elements. The equivalent solid plate properties are used to model the strainer perforated plates for structural analyses. The equivalent properties for the perforated plate are:

E*= 0.43 E, equivalent modulus of elasticity v* = 0.33, equivalent Poisson's ratio K = 2.33 stress multiplier The equivalent properties were applied to the solid plates in the ANSYS finite element model to simulate the plate perforations.

Modal frequencies in air and in water were determined according to the seismic analysis requirements and were used to determine the seismic accelerations. The load cases required by the design specification were analyzed to determine stresses in all components.

The strainers are designed for a 40-year life. Thermal fatigue was, evaluated qualitatively, and dismissed as insignificant since the normal operation temperature cycle ranges are small (50 0 F) and there is only one LOCA temperature cycle with a range of 199 0 F. Material Properties were based on stainless steel SA 240, Type 304 which is the material of construction. The material properties used were selected from the ASME code and are shown in Table 3.k.1-1.

Table 3.k.1-1 Material Properties Temperature Material / Property Unit 300°F SA-240, Type 304 Elastic modulus' psi 27.OE+6 Coefficient of thermal expansion in/in-0 F 9.OE-6 Poisson's ratio 0.3 Density lb/in3 0.289 Stress Allowable psi 16700 Tie-rod bolt material SA-193, B8 Elastic modulus psi 27.2E+6 Coefficient of thermal expansion in/in-0 F 9.OE-6 Yield Strength psi 22500(1)

(1) Yield strength of SA-193, B8 material at 70F is 30,000 psi.

to W3F1-2008-0069 Page 53 of 87 Load Definitions and Combinations Strainers, Plenums, the Partition and the Sensor, and supports are designed for the loads and load combinations described in this section.

Load Definitions W Strainer Assembly Weight in Air, Normal Plant Operation WD Strainer Assembly Weight in Water + Debris Weight + Hydrodynamic Mass, LOCA TEmax Thermal Expansion in Water, LOCA TEop Thermal Expansion in Air, Normal Plant Operation PO Containment Pressure Pd Containment Design Pressure Pcr Differential Crush Pressure, LOCA OBE1 Operating Basis Earthquake Inertia Loading in Air OBE2 Operating Basis Earthquake Inertia Loading in Water + Debris Mass +

Hydrodynamic Mass SSE1 Safe Shutdown Earthquake Inertia Loading in Air SSE2 Safe Shutdown Earthquake Inertia Loading in Water + Debris Mass

+Hydrodynamic Mass Load Combinations Strainers and Plenums Design = W + TE0 p + P0 + OBE1 Level B = Pd + WD + OBE2 + TEmax + Pcr Support Structures Design = W + TEop Level B = WD + OBE2 + TEmax Level D = WD + SSE2 + TEmax The strainer assembly shall withstand a live load of 250 pounds during outages. This load is negligible compared to operating loads and no specific analysis was performed.

The seismic loads are based on the horizontal and vertical inertial accelerations specified by the seismic response spectrums according to the first mode frequency in water. The design pressures, P0 and Pd, have no impact and add nothing to the load combinations cited above because the strainer system is an open system that is not pressurized by containment pressure but is loaded by crush pressure, Pcr. Hydrodynamic mass values and debris weights, are included where applicable.

Table 3.k.1-2 Mass Properties Dry weight of 1 strainer lb 2,374 Submerged weight of 1 strainer lb 3,560 Dry weight of plenum lb 9,753 to W3F1-2008-0069 Page 54 of 87 Submerged weight of plenum in vertical direction lb 47,788 Submerged weight of plenum in x and z direction lb 12,679 Dry weight entire assembly lb 35,862 Submerged weight of entire assembly in vertical direction lb 86,952 Table 3.k.1-3 Coefficients Used for Seismic Analysis*

OBE lateral 0.25 g OBE vertical 0.20 g SSE lateral 0.38 g SSE vertical 0.30 g

• These accelerations are above the ZPA value, therefore no multiplier is applied The structural response due to OBE & SSE is different depending on whether the equipment is in air or in water.

Loads used in the stress analysis include the weight of the strainer assembly, debris, contained water, crush pressure due to pump operation, and seismic loads. The lateral and vertical inertial accelerations were obtained from the seismic response spectra corresponding to the first mode frequency of the equipment when submerged.

Crush pressure was applied to the strainer plates, spacer rings and plenum plates. The weight of the equipment in water was analyzed as the sum of the weight of the assembly in air, the debris weight and the hydrodynamic mass and contained water. A 1-2g" inertial load in the vertical direction was used to represent the dead weight of the equipment.

WF3 Response 3.k.2 All Qualification information is contained in GE calculation GENE-0000-0054-9349 (Reference 56).

Finite element analyses were performed for all components using ANSYS Version 10 computer program. Stresses from design load combinations are compared with the ASME Code Section Ill, Subsections NC, and ND stress limits. Stress margins for the limiting components were calculated for the Design Condition, Service Level B, and Service Level D Load Combinations.

Table 3.k.2-1 shows selected calculated stresses. The minimum stress margins are shown in Tables 3.k.2-2 (Strainer Components), 3.k.2-3 (Partition Components) and 3.k.2-4 (Sensor Components).

Table 3.k2-1 Stress Summary for Strainer Components Load Combinations & Max Calculated Stresses, ksi Component Design Level B Level D W + OBEI WD + OBE2 + Pcr WD + SSE2 + Pr

-Perforated Plate (1) 1.8 3.5 4.0 Frame & Fingers 3.3 4.7 5.8 Spacers 7.4 10.3 12.5 to W3F1-2008-0069 Page 55 of 87 Strainer Base 1.8 2.3 2.8 Tie Rod 1.1 1.2 1.7 Plenum 9.0 12.8 14.8 2

Allowable Pm Stress( ) 1.0 x S 1.1 x S 2.0 x S Allowable Stress (2) 16.7 18.4 33.4 Notes:

1) Perforated plate includes intensification factor of 2.33

2) For conservatism, the allowable for membrane stress is used, which is the lowest Table 3.k.2-2 Margin Summary for Strainer Components Load Combinations & Margins (1)

Component Design Level B Level D W + OBE1 WD + OBE2 + Pcr WD + SSE2 + Pcr Perforated Plate (1) 8.3 4.3 7.4 Frame & Fingers 4.1 2.9 4.8 Spacers 1.2 0.8 1.7 Strainer Base 8.4 6.9 10.9 Tie Rod 14.6 14.4 18.6 Plenum 0.9 0.4 1.3 Notes:

1) Margin = (Allowable/Calculated) - 1 Table 3.k.2-3 Stress (ksi) Summary for Partition Components Component Design Level B Level D Partition Assembly 1.3 6.4 16.9 Allowable Stress 16.6 18.4 33.4 Partition Margin 11.8 1.9 1.0 Table 3.k.2-4 Stress (ksi) Summary for Sensor Components Component Design Level B Level D Sensor Assembly 0.4 7.4 11.6 Allowable Stress 16.6 18.4 33.4 Sensor Margin 40.5 1.5 1.9 The strainer disk surfaces are covered by a woven wire cloth, which is resistance welded to the perforated plate. In the Waterford 3 application this woven wire mesh is used solely to enhance the debris carrying capability with respect to hydraulic head loss of the disks and no structural credit is taken for its presence; however the mass is included in the analysis. It is necessary to assure that the Woven Wire remains attached to the disk when the disk is subjected to seismic loading and when the disk deflects due to the pressure drop across the disk faces. This assurance is obtained as the result of testing performed by GEH in which the composite of to W3F1-2008-0069 Page 56 of 87 perforated plate and woven wire was deflected over 1" with the wire remaining attached. This deflection is at least an order of magnitude greater than will be experienced in service.

WF3 Response 3.k.3 GENE-0000-0048-9192 (Reference 64) is an evaluation which concluded that the strainer assembly is not subject to pipe whip, jet impingement, or missile impact associated with a HELB.

WF3 Response 3.k.4 A backflushing strategy is not credited in the Waterford 3 analyses.

3.1 Upstream Effects NRC Issue 3.1 The objective of the upstream effects assessment is to evaluate the flowpaths upstream of the containment sump for holdup of inventory, which could reduce flow to and possibly starve the sump.

Provide a summary of the upstream effects evaluation including the information requested in GL 2004-02, "RequestedInformation," Item 2(d)(iv).

GL 2004-02 Requested Information Item 2(d)(iv)

The basis for concluding that the water inventory required to ensure adequate ECCS or CSS recirculation would not be held up or diverted by debris blockage at choke-points in containment recirculationsump return flowpaths.

1. Summarize the evaluation of the flow paths from the postulated break locations and containment spray washdown to identify potential choke points in the flow field upstream of the sump.

2. Summarize measures taken to mitigate potential choke points.

3. Summarize the evaluation of water holdup at installed curbs and/or debris interceptors.

4. Describe how potential blockage of reactor cavity and refueling cavity drains has been evaluated, including likelihood of blockage and amount of expected holdup.

WF3 Resp~onse 3.1.1 The Waterford 3 containment is mostly uncompartmentalized with the exception of the pressurizer room. There are no structures totally surrounding the major components (SG, RCP, etc.) of the RCS. All RCS components with the exception of the pressurizer are within the SG cavities (D-Rings), but the SG cavities are open to each other (on the north side below elevation

-4 ft., open to the annulus below elevation -1 ft and are open to the dome above elevation

+62.25 ft. The PZR is located in a separate room with an opening in the floor that connects to the containment annulus.

Waterford 3 does not have any significant inactive volumes other than the reactor cavity and containment sump. For significant quantities of debris to be trapped in the reactor cavity or to W3F1-2008-0069 Page 57 of 87 containment sump, the break location would have to be at the reactor (within the reactor cavity).

As described in Section 3.e, relatively small quantities of debris that transport to the sump will be created for any break in the reactor cavity itself. For breaks outside the primary shield wall, significant quantities of debris would not be transported to the reactor cavity or containment sump by flowing water during the pool fill-up. This is because the only flow paths from the active pool to the reactor cavity and containment sump at the minimum flood elevation are through several floor drains located in the containment at elevation -11.0 ft that drain to the containment sump.

WF3 Response 3.1.2 As no potential choke points were identified for Waterford 3, no mitigation measures were necessary.

WF3 Response 3.i.3 Waterford 3 does not have any curbs on the -11 ft basemat elevation. Throughout containment, where slabs are adjacent to open areas or grating areas, there are no concrete curbs. For open areas, there are kickboards on the handrails, but these are not flush against the surface of, the concrete and will allow water to flow under and around them.

WF3 Response 3.i.4 Calculation 2005-05500 (Reference 29) documents that the refueling cavity has two 6-inch drain lines (without screens) that drain to the containment floor, and by one 4-inch line that drains to the containment sump. In the event that large debris is propelled over the SG cavity walls into.

the refueling cavity, the debris must land on a drain in order to clog it since the velocities in the cavity are too low to transport a large piece of debris to a drain. During plant operations, the Upper Guide Structure Lift Rig (UGSLR) is stored directly above one of the 6" drains. The UGSLR will prevent any debris larger than 6" from falling directly onto the drain. Any smaller debris that transports to these two 6" drains will pass through the drains since they do not contain screens. Therefore, there will always be drainage available from the refueling cavity to the active pool.

3.m Downstream Effects - Components and Systems NRC Issue 3.m 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.

Provide the information requested in GL 04-02, "Requested Information," Item 2.(d)(v) and 2.(d)(vi) regardingblockage, plugging, and wear at restrictionsand close tolerance locations in the ECCS and CSS downstream of the sump by explaining the basis for concluding that inadequate core or containment cooling would not result due to debris blockage at flow restrictions in the ECCS and CSS flowpaths downstream of the sump screen, (e.g., a HPSI throttle valve, pump bearings and seals, fuel assembly inlet debris screen, or containment spray nozzles). The discussion should consider the adequacy of the sump screen's mesh spacing and state the basis for concluding that adverse gaps or breaches are not present on the screen surface. For GL 2004-02, Item 2(d)(vi) provide verification that the close-tolerance to W3F1-2008-0069 Page 58 of 87 subcomponents in pumps, valves and other ECCS and CSS components are not susceptible to plugging or excessive wear due to extended post-accident operation with debris-laden fluids.

1. GL 2004-02 Requested Information Item 2(d)(v)

The basis for concluding that inadequate core or containment cooling would not result due to debris blockage at flow restrictionsin the ECCS and CSS flowpaths downstream of the sump screen, (e.g., a HPSI throttle valve, pump bearings and seals, fuel assembly inlet debris screen, or containment spray nozzles). The discussion should consider the adequacy of the sump screen's mesh spacing and state the basis for concluding that adverse gaps or breaches are not present on the screen surface.

2. GL 2004-02 Requested Information Item 2(d)(vi)

Verification that the close-tolerance subcomponents in pumps, valves and other ECCS and CSS components are not susceptible, to plugging or excessive wear due to extended post-accidentoperation with debris-ladenfluids.

3. If NRC-approved methods were used (e.g., WCAP-16406-P with accompanying NRC SE) briefly summarize the applicationof the methods.

4. Provide a summary and conclusions of downstream evaluations.

5. Provide a summary of design or operationalchanges made as a result of downstream evaluations.

WF3 Response 3.m.1 As a result of GSI-191, new SI Sump screens were installed in Waterford 3. The nominal perforated plate hole in the screens is 3/32 inches (ER-W3-2003-0394-001 - Reference 30).

Strainer walk downs are performed prior to start up from plant outages to confirm no gaps, breaches, or openings greater than 3/32 inch exist in the strainers.

The ability of the ECCS equipment required to pass debris laden fluid during the recirculation phase after a postulated accident is evaluated in Calculation 2005-02820 (Reference 31). This evaluation determined the ECCS equipment that would be in the post-accident recirculation path and reviewed the dimensions of close-tolerances in this ECCS equipment against the acceptance criteria of 1.1 and 2 times the screen hole size. The HPSI pumps, CS pumps, CS headers solenoid operated valves, HPSI pump recirculation flow orifices, HPSI header throttle valves and RC loop throttle-valves were determined to have minimum flow clearances small enough to require wear evaluations.

To resolve these issues, calculation 2005-12840 (Reference 32) was prepared. This calculation determined that most components in the system would not be blocked by debris. The HPSI pump seal and the CS pump cyclone separator required further evaluation.

Calculation 1062-0802-0015-4 (Reference 67) evaluated the HPSI pump seals for the injection of debris during a LOCA. This evaluation concluded that the potential for mechanical seal failure due to debris blocking axial movement of the rotating seal face is considered low. There is no potential for significant debris-induced wear of the seal faces due to the tight running gap.

The HPSI seals use a priming ring to recalculate water through a heat exchanger and the seal.

Calculation 1062-0015-03 (Reference 68) concluded that the CS pump cyclone separators will continue to provide clean water to the CS pump mechanical seals following a design basis to W3F1-2008-0069 Page 59 of 87 accident. This conclusion is based on a comparison of separator test data to the Waterford 3 separators and debris loading.

WF3 Response 3.m.2 Blockage of components was addressed above; wear of close tolerance components and systems is addressed in this paragraph. Calculation 2005-12840 (Reference 32) primarily addressed component wear; however, it also included instrument lines, relief valves, piston check valves and post accident sampling system components for the potential for blockage due to debris. For equipment addressed by WCAP-16406-P, Revision 1, August 2007, the methods and acceptance criteria were in accordance with the WCAP.

The wear analysis in calculation 2005-12840 concluded that wear was acceptable as it resulted in negligible flow effects based on WCAP-16406-P acceptance criteria with the exception of the HPSI pumps. The analysis only determined that worn condition of the pump and generated performance curves at various time points during the 30 day mission time. The worn performance curves were inputted into the Waterford 3 Long Term Cooling (LTC) Analysis of Record (AOR) to determine acceptability. The result from the LTC AOR evaluation is that adequate core cooling is maintained and the AOR remains valid. The worn condition, wear ring clearances, of the pump were used in a rotor-dynamic analysis to confirm that the pump remains dynamically stable thorough the mission time.

WF3 Response 3.m.3 The methods of WCAP-16406-P were used with interpretations of the November 2007 draft of the SER to the WCAP and with interpretations described during the September 2007 training teleconference. Calculation 2005-12840, Revision 1 used some more detailed methods where additional quantification was required.

Section 5 of WCAP-1 6406-P describes a methodology for calculating debris depletion over time.

TheWCAP also provides values of depletion coefficients by way of example. The WCAP does not provide specific depletion coefficients. Based on flow rates, volumes and settling velocities at Waterford. 3, plant specific depletion coefficients were calculated. These depletion coefficients also credited filtration of particulates as well as fibers on the sump screen where such filtration is supported by plant specific testing.

WCAP-16406-P, Revision 1 provides information on size distribution and settling fraction of coatings. It states that qualified coatings fail as 10 micron particles. This is conservative for pressure drop calculations, but not for downstream calculations. The Waterford 3 specific evaluation used a larger size particle based on vendor information about size of pigments in the coatings. This results in more calculated wear and is conservative. WCAP-16406-P assumes that unqualified coatings larger than 100 microns will settle. The NRC has questioned the "Stoke's Law" models used in such evaluations. The Waterford 3 calculation uses an empirical correlation for friction factor and benchmarks the resulting settling size against NRC-sponsored settling tests. Because the paint chips were all assumed to settle with the widest cross section perpendicular to the direction of settling, the calculation showed a larger settling size for a given paint chip and settling velocity. This results in a conservative, benchmarked, plant-specific settling size for particulates.

to W3F1-2008-0069 Page 60 of 87 A pump curve after wear was calculated for each Waterford 3 ECCS pump rather than utilizing WCAP Figure 8.1-3. The curve in the WCAP is based on a single stage pump with a particular specific speed and does not bound the calculated wear effect for multi-stage high head, low flow pumps likes the High Pressure Safety Injection pump. The more conservative method was used in 2005-12840, Revision 1. WCAP-16406-P recommends a minimum friction factor for maximizing the packing wear.

WCAP-16406-P, Revision 1, Appendix 0, Section 2.3 recommends an assumed friction factor of 0.01 to maximize wear. During the performance of the calculation it was found that the rate of wear, measured as gap.increase, would be maximum when the combination of parameters, friction factor times bearing length divided by clearance, was set equal to 2/3. Since this can be demonstrated mathematically it is no longer necessary to make an assumption about the friction factor in order to maximize the wear.

Entergy understands that Section 7.2 and 8.1.3 of the WCAP and the draft SER mean that if debris laden fluid is piped from the recirculation stream to flush a pump's seal then the primary seal would fail as a direct consequence of the postulated LOCA. That would constitute a common mode failure and all such pump seals would fail concurrently during the recirculation phase of the postulated LOCA. Conversely, if fluid from the recirculation stream is not piped to a pump's seal then there is no credible source of debris to fill the seal chamber and the primary pump seal is not assumed to fail as a direct consequence of the postulated LOCA. Such seals would still be subject to a postulated random failure of the pressure boundary as a moderate or high energy line break. The applicable requirements of SRP 15.6.5 as committed to in the USAR would remain applicable. For future reference, the leakage rate through pump seal one-half hour after a postulated primary seal failure was calculated. This calculation included the affects of wear on the components in the seals that would remain intact after a primary seal failure.

Rounding the inlet to an orifice in conjunction with increasing the orifice diameter decreases the-flow resistance more than just increasing the diameter. In order to account for the effects of rounding the inlet of an orifice by debris, Section 8.4 of WCAP-16406-P, Revision 1 recommended a formula taken from the first edition of Idelchik's "Handbook of Hydraulic Resistance". The first edition, translated from Russian in the 1960's has been updated and the corresponding formula from the third edition of Idelchik's "Handbook of Hydraulic Resistance" is used.

WF3 Response 3.m.4:

Those ECCS components and systems that are required to operate and pass debris laden fluid during the recirculation phase of recovery from a postulated LOCA have been identified. These ECCS components have been evaluated for blockage and wear from debris that would pass through the new containment sumps screens. The ECCS equipment at Waterford 3 would remain capable of passing sufficient flow to the reactor to adequately cool the core during the recirculation phase of a postulated LOCA.

WF3 Response 3.m.5:

At this time no operational changes have been made nor have any been identified for Waterford 3.

to W3F1-2008-0069 Page 61 of 87 3.n Downstream Effects - Fuel and Vessel NRC Issue 3.n The objective of the downstream effects, fuel and vessel section is to evaluate the effects that debris carried downstream of the containment sump screen and into the reactor vessel has on core cooling.

Show that the in-vessel effects evaluation is consistent with, or bounded by, the industry generic guidance (WCAP-16793), as modified by NRC 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.

WF3 Response 3.n The in-vessel effects evaluation was performed in accordance with the guidance in WCAP-16793-NP (Reference 60) and the initial NRC comments provided related to use of that document dated 2/4/08 (Reference 61). This evaluation did not indicate problems with reactor core cooling. The fuel deposit analysis was performed per the WCAP-16793 spreadsheet with conservatively bounding inputs relative to the maximum debris loading conditions for the plant.

This analysis determined that significant margin exists relative to the acceptance criteria, with total deposition thickness of <13 mils, remaining well below the 50-mil maximum value and the maximum clad temperature of <328°F also remaining well below the 800'F acceptance criteria.

The initial NRC comments provided for WCAP-16793 have been withdrawn and the WCAP is currently in revision, although the source of the revision is understood to be related to the fuel blockage analysis, not the fuel deposit methodology. Following the issuance of the revised guidance, further analysis could be necessary.

3.o Chemical effects NRC Issue 3.o 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.

2. Content guidance for chemical effects is provided in Enclosure 3 to a letter from the NRC to NEI dated September 27, 2007 (ADAMS Accession No. ML0726007425).

WF3 Response 3.o Chemical precipitates that form in the post-LOCA containment environment combined with debris do not result in an unacceptable head loss. Head loss due to chemical precipitates and debris is demonstrated by test using WCAP-16530-NP (Reference 58) methods with relatively minor modifications.

to W3F1-2008-0069 Page 62 of 87 The Alion Science and Technology 30-day integrated chemical effects testing identified that calcium phosphate precipitates can form very early post-LOCA due to the very low and retrograde solubility of calcium phosphate (lower solubility at high temperature). The 30-day integrated testing also identified that aluminum based precipitants do not form until the post-LOCA environment has cooled to below 140 degrees F. The prototype testing used these results to sequence the WCAP-16530-NP/16785-NP (References 58 and 53) based precipitates. Head loss calculations used the head loss attributed to calcium phosphate to determine the head loss across the strainer at temperatures greater than 140 degrees F when the NPSH margin is limiting. The 30 day integrated testing and analyses concluded that no aluminum based precipitates would form in the Waterford 3 environmental conditions with a pH less than 8.1; therefore any reduction in the aluminum oxy-hydroxide precipitate is reasonable.

3.p Licensing Basis' NRC Issue 3.p The objective of the licensing basis section is to provide information,regardingany changes to the plant licensing basis due to the sump evaluation or plant modifications.

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.

GL 2004-02 Requested Information Item 2(e)

A general description of and planned schedule for any changes to the plant licensing bases resulting from any analysis or plant modifications made to ensure compliance with the regulator requirements listed in the Applicable Regulatory Requirements section of this GL. Any licensing actions or exemption requests needed to support changes to the plant licensing basis should be included.

WF3 Response 3.P Major changes that have been made to the Waterford 3 Licensing Basis to meet compliance with the Generic Letter include modification to the Safety Injection Sump Strainer, relocation of the Trisodium Phosphate (TSP) baskets and revision to the Safety Injection Sump NPSH parameters.

Updated Final Safety Analysis Report (UFSAR) changes were issued to document the changes in the available head loss for both the HPSI pumps and the CS pumps. Changes were also issued against the FSAR to state that Waterford 3 is in compliance with the requirements of GS1-191 and Generic Letter 2004-02.

to W3F1-2008-0069 Page 63 of 87 Open Items 01.1 The licensee should justify its assumption of a 2D zone of influence for the Waterford 3 metal encapsulatedinsulation fiberglass.

01.1 Response The debris generation calculation 2004-07780 (Reference 28) was revised to use a 4D ZOI for the MEI insulation instead of the 2D ZOI originally used. The justification for this is presented in Alion Report ALION-REP-ENTG-4771-02 (Reference 39). The design of the RMI and MEI insulation cassettes are comparable. The filler material in either one does not contribute to the strength of the insulation system. Therefore, the ZOI for the Transco MEI cassettes should be equal to the ZOI for Transco RMI (2D ZOI). However, to ensure conservatism, and since there is no specific destruction testing that has been performed for the MEI, a 4D ZOI is used for Transco MEI such as used in Waterford 3.

01.2 The licensee should provide comprehensive documentation of the characteristics (macroscopic densities, microscopic densities, and characteristic debris sizes) of the actual plant debris at Waterford 3 and compare these characteristicsto the surrogate debris properties used for head loss testing, justifying any differences.

01.2 Response This is addressed in section 3.h of this supplemental response.

01.3 The licensee should provide an analysis that shows that the coating debris test data credited by Waterford 3 was generated using coating chips that are representative of or bounding with respect to the plant-specific failed coating chips.

Ol.3.Response This is addressed in section 3.h of this supplemental response.

01.4 The licensee should (1) justify that the percentage of debris transporting along the containment floor from the east and west sides of containment is equal to the percentage of flow approaching the sump from the east and west sides of containment and (2) provide a clear definition of the startingpoints for debris transportpaths to avoid contributing to an underestimation of debris transport on the side of containment opposite the break.

01.4 Response The transport calculation 2005-05500 (Reference 29) treats the Waterford 3 containment as uncompartmentalized. All LOCA generated debris is conservatively modeled as falling to the floor.

The flow around each side of containment is used to appropriately apportion the debris to the east or west sides of containment. Immediately after a break occurs, water spills from the break to the floor and the initial water from the break spreads across the floor.

The wave created by the initial water spreading is expected to have a high enough velocity to push debris away from the break, towards the perimeter of containment.

to W3F1-2008-0069 Page 64 of 87 Once the debris has been moved towards the containment perimeter and recirculation flow has been established, the flow distribution around each side of containment will dominate transport.

01.5 The licensee should explain how it has addressed the following four deficiencies in the existing transport analysis for unqualified coating chips: (1) lack of adequate data to justify the assumed size distribution for failed coating chips, (2) improper application of settling data for coating chips with a 400-micron thickness to particle-like coating debris with a 400- micron diameter, (3) lack of justification of the use of an analysis intended for the vertical flow conditions typical of a reactor vessel core inlet plenum for the horizontal flow conditions in the Waterford 3 containment pool, and (4) lack of considerationof the possibility that coating chips that fall into the containment pool in the vicinity of the sump may transportto the sump in suspension in the containment pool prior to settling on the containment floor.

01.5 Response

1) In accordance with NEI Guidance 04-07 (Reference 2) and the SER for NEI 04-07 (Reference 3), all qualified coating debris and unqualified zinc coating debris and indeterminate coating debris are treated as particulate and are therefore transported entirely to the sump strainer. Degraded qualified coatings are considered to fail as chips with a size distribution per ALION-REP-TXU-4464-02 and letter TXX-07156 (References 25 and 26). See Waterford 3 Response 3.h.6 for more details.

2) In calculation 2005-05500 (Reference 29) only degraded qualified coatings are considered to fail as coating chips. Indeterminate coatings are considered to fail as particulates. The only sources of degraded qualified coatings within the Waterford 3 containment are the containment dome and the containment liner. The coating system used on the containment dome is Carboline Phenoline 305 applied as a topcoat over Carboline Carbo-Zinc 11 primer. This system is also one of four Service Level 1 coating systems approved for the containment liner. The other systems include Carboline 305 topcoat over Carboline 191 primer, Carboline 801 as a primer and topcoat and Amerlock 400 NT (a one coat system). In the CCCL Letter dated 9/20/07 (Reference 27), it is confirmed that all are epoxy systems, which are expected to exhibit failure characteristics of Carboline 305. The CCCL Letter dated 9/20/07 (Reference 27) also confirms.that the size distribution presented in ALION-REP-TXU-4464-02 and letter TXX-07156 (References 25 and 26) is applicable to the Carboline Phenoline 305 coatings used at Waterford 3.

NUREG/CR-6916 (Reference 19) presents transport velocities for coatings with the size distributions presented in ALION-REP-TXU-4464-02 and letter TXX-07156 (References 25 and 26). The data reported in NUREG/CR-6916 (Reference 19) are for the failure characteristics of many coatings including epoxy applied as a topcoat over inorganic-zinc. The epoxy top coats applied over a zinc rich primer, Category ZE, are similar to the Carboline Phenoline 305 applied as a topcoat over Carboline Carbo-Zinc 11 used at Waterford 3.

to W3F1-2008-0069 Page 65 of 87

3) Calculation 2005-05500 (Reference 29) was revised and no longer uses the analysis intended for a reactor vessel core inlet plenum to apply to the horizontal flow conditions in the Waterford 3 containment pool. Instead, the calculation utilizes the data in ALION-REP-TXU-4464-02, letter TX-07156, and the CCCL letter dated 9/20/07 (References 25, 26, and 27) for the horizontal flow conditions at .Waterford.

4) The transport calculation 2005-05500 (Reference 29) accounts for coating chips that may land in the containment pool near the strainers and may not have a chance to settle before transporting to the strainer. In the following figure the pool area is shaded and the area considered near the strainer is cross hatched:

i ,. A <,, .K( ., ":::

The cross hatched area is conservatively approximated as 20% of the pool area. As 20%'of coating chips generated are expected enter the pool near the sump strainer, 20% of the -coating chip debris is considered to experience 100% transport to the strainer.

01.6 The licensee should justify that the debris transport fractions in the transportcalculation are representativeof the replacementstrainerconfiguration.

01.6 Response The transport calculation 2005-05500 (Reference 29) now includes several additional Computational Fluid Dynamics (CFD) models. These models were set upto verify the flow conditions in containment with the new strainer modules installed, and the TSP baskets no longer being installed directly adjacent to the strainers. The calculation 2005-05500 (Reference 29) documents the flow velocity, which directly affects the movement of the debris. With these new CFD models, the debris transport fractions are representative of the replacement strainer configuration.

to W3F1-2008-0069 Page 66 of 87 01.7 The licensee should provide results of analysis of the potential-effects of a low-pressure safety injection pump failure to stop on a recirculationactuationsignal.

01.7 Response There is a potential for a LPSI pump to fail to trip for the first 30 minutes of recirculation.

The failure of the LPSI pump to trip would result in a strainer flow rate that is non-conservatively larger than the flow rate that has been analyzed. While the head loss with the LPSI pump in operation was not determined by testing or analysis, GENE-0000-0053-4416-P (Reference 36) documents that the number of plant sump water turnovers in the time that the LPSI pump operates during recirculation time would be small enough that relatively little debris would get to the sump strainers.

Waterford 3 has a minimum sump water volume of about 46,335 cubic feet and a flow rate with the LPSI pumps of 12120 gpm, for a turnover time of 28.6 minutes; the plant water volume would experience approximately one turnover for the 30 minutes of LPSI trip failure. The head loss test performed had a flow rate of 364 gpm and a volume of 196 cubic feet based on a pool size of 123 inches by 72 inches and a water depth of 38.25 inches (Test specification 26A6833 - Reference 33) for a turnover time of four (4) minutes.

A review of the test head loss vs. time curves show the head loss during the first four minutes (one turnover) of testing after debris was added, was significantly lower than the stabilized head loss, i.e. more than one test turnover was required to generate maximum head loss. It can be concluded that the debris bed had not formed within one turnover in the test, nor is the debris bed expected to have formed within one turnover in the plant.

The LPSI pumps will be secured within 30 minutes of the start of recirculation and before the debris bed has formed on the strainer. Therefore, the failure of a LPSI pump to trip upon a RAS signal will not result in any debris bed formation or significant impact on head loss for the sump strainers. Clean strainer head loss is scaled for the added flow in the NPSH determinations. For added assurance that a LPSI failure to trip would not result in failure of the sump or any other pump, an NPSH evaluation has been performed using a sump temperature of 190F, based on sump temperature profiles in section 3.g.

The results of this evaluation can be found in section 3g. 16.

01.8 The licensee should describe how it has implemented prototypically fine fibrous debris preparationin its head loss testing.

01.8 Response For follow-up thin-bed testing, fiber was shredded five times in sequence, resulting in significantly reduced fibrous debris clump size that is more representative of small fines for thin-bed tests. Fiber sizes are generally small clumps of fiber or individual fibers which are representative of eroded fibers..

to W3F1-2008-0069 Page 67 of 87 01.9 The licensee should describe and justify how it has conducted adequate testing to determine thin bed peak head losses.

01.9 Response For follow-up design basis testing, the scope of tests was expanded so that the following fiber thicknesses are included in testing: 0.125", 0.25", 0.5", 0.75", and 1" (gap-filled), to ensure that any localized peaks due to thin-bed effect would be discovered.

01.10 The licensee should provide the results of assessment of the potential for non-prototypical settling and non-prototypical bed formation due to debris agglomeration during partially stirredstrainertesting.

01.10 Response Module testing for the partially stirred circumscribed bed cases was performed by creating a small-scale plant mock-up with one sump strainer module. The mockup included the plenum and one test module with a width and length matching the plant strainer design (40-inches X 40-inches) but with 10 discs rather than the plant design of 17 discs. The plenum height and distance of the strainer above the plenum were each scaled based on the scaled module height. Walls to the back and sides of the test module were set 16-inches from the strainer to match the plant strainer spacing of 32 inches.

The flow rate of the module was increased relative to the flow rate scaled by the circumscribed area of the module to match the average bulk approach velocity created by a plant strainer and thus accurately model the near field effects around the strainer.

The flow rate was then increased by 5% to add margin for expected flow measurement accuracy.

The measured flow rate accuracy for the module test configuration was 1%. The difference in the expected flow measurement accuracy and the actual flow measurement accuracy resulted in a test flow rate that was higher than the scaled plant flow rate. The additional head loss caused by this higher flow rate due to instrument accuracy (1%

actual vs. 5% anticipated) was removed by the scaling methodology. However, the higher flow rate also increased near-field debris transport and debris bed compression.

These effects were not removed and add conservatism.

0l. 11 The licensee should describe and justify how it has resolved the potential for non-'

prototypical flows during module testing due to "solid modeled" trisodium phosphate baskets located nearthe strainermodules.

0111 Response Waterford 3 ran new six (6) new head loss tests with the simulated baskets completely removed from the tank. These newer tests mirror the current installed condition of the SI Sump strainer system with the TSP baskets removed.

to W3F1-2008-0069 Page 68 of 87

01. 12 The licensee should provide the results of evaluation of the potential for and effects of vapor flashing due to strainerhead loss being greaterthan the strainersubmergence.

01.12 Response As stated in section 3.f.2, the sump strainers are submerged by a minimum of about 8 inches for a Large Break LOCA. The maximum head loss determined in section 3.f.10 is 0.517 ft (6.204 inches) with pump run out flows. Sump submergence is greater than maximum head loss therefore flashing will be prevented.

01.13 The licensee should explain how the following four additionalwater holdup mechanisms are modeled in the analysis of minimum containment pool water level: (1) water holdup due to condensation films, (2) water holdup due to spray droplet holdup in the containment atmosphere (as opposed to water vapor holdup in the containment atmosphere), and (3) refill of the reactor pressure vessel with colder and therefore denser water, and (4) the reactor water safety pool water temperature specified to be at the warmer normal operatingcontainment temperature.

01.13 Response Calculation MNQ6-4 (Reference 37) has been revised to include the water holdup due to condensation films, spray droplet holdup in the containment atmosphere, the impact of refilling the reactor coolant system with cooler water, and maximum allowed Refueling Water Storage Pool temperature.

1) Calculation MNQ6-4 determined that a condensation film thickness of 28 mils is appropriate for heat sink surfaces inside containment. However, the film thickness is assumed to be 35 mils. The total containment heat sink area is based on the passive heat sink area used in current containment analysis for Waterford 3.

2) Calculation MNQ6-4 included both water vapor holdup in the containment atmosphere, and water holdup due to spray droplets in the containment atmosphere.

3) Cooling of the RCS will cause the fluid contained within to contract as its density increases. For conservatism, calculation MNQ6-4 does not take credit for any RCS fluid in the water level calculations and assumes that contraction of the RCS is compensated by the lowering of pressurizer and steam generator levels.

4) The temperature of the water in the RWSP is assumed to be at the maximum of 100 F allowed by the Waterford 3 Technical Specifications.

01. 14 The licensee should justify treating unqualified coatings debris characteristics in the same manner as for qualified coatings.

01.14 Response The unqualified coatings at Waterford Unit 3 fall into two categories; degraded qualified coatings and indeterminate coatings. For the indeterminate coatings, the specific coating system applied is not known so these coatings are conservatively treated as 100% small fines and as such experience 100% transport to the sump.

Degraded qualified coatings are found on the containment dome and the containment liner. The painting system used on the containment dome is Carboline Phenoline 305 applied as a topcoat over Carboline Carbo-Zinc 11 primer. This system is also one of to W3F1-2008-0069 Page 69 of 87 four Service Level 1 paint systems approved for the containment liner. The other systems include Carboline 305 topcoat over Carboline 191 primer, Carboline 801 as a primer and topcoat and Amerlock 400 NT (a one coat system). All are epoxy systems, which are expected to exhibit failure characteristics of Carboline 305. In the CCCL letter dated 9/20/07 (Reference 27), it is confirmed that the size distribution presented in ALION-REP-TXU-4464-02 and letter TXX-07156 (References 25 and 26) is applicable to the Carboline Phenoline 305 coatings used at Waterford 3, except for one coating system possibly used on the containment liner.

NUREG/CR-6916 (Reference 19) presents transport velocities for coatings with the size distributions presented in ALION-REP-TXU-4464-02 and letter TXX-07156 (References 25 and 26). The data reported in NUREG/CR-6916 (Reference 19) are for the failure characteristics of many coatings including epoxy applied as a topcoat over inorganic-zinc. The epoxy top coats applied over a zinc rich primer, Category ZE, are similar to the Carboline Phenoline 305 applied as a topcoat over Carboline Carbo-Zinc 11.

Debris characteristics used for the unqualified coatings in containment are based on testing results found in NUREG/CR-6916, ALION-REP-TXU-4462-02, and letter TXX-07156 (References 19, 25, and 26). The debris characteristics used can be found in section 3h of the Generic Letter responses.

01.15 The licensee should summarize how it has addressedthe following aspects of structural analysis for the new strainer.

Part 1 - The licensee should revise the high-energy line break report to provide definitive statements in the conclusions concerning pipe whip and missile impacts to the new strainerassembly and clarificationof the bases for those conclusions.

Part 2 - The licensee should revise the sump strainer design specification to clearly identify the damping to be 2%.

Part 3 - The licensee should correct errors and discrepancies in the sump strainer design specification and stress analysis report.

Part 4 - The licensee should correct the temperature delta in the sump strainer stress analysis report.

Part 5 - The licensee should clarify the sump strainer acceleration values in the sump strainerstress analysis report.

Part 6 - The licensee should correct the values for E*/E and K in the sump strainerstress analysis report.

Part 7 - The licensee should correct the whole strainerand plenum maximum deflection values in the stress analysis report.

Part 8 - The licensee should correct stress limits, safety factor values and certain units used for stress limits in certain tables of the sump strainerstress analysis report.

to W3F1-2008-0069 Page 70 of 87 Part 9 - The licensee should address discrepanciesidentified in the hydrodynamic mass analysis.

01.15 Response - Part 1 Section 3.5 of the UFSAR documents the evaluation of missiles inside and outside the Waterford 3 RCB. Table 3.5-4 of the UFSAR contains a list of the potential missiles inside containment. None of these are located in an area where they could impact the new strainer assembly.

The closest high energy line breaks are located over 20 feet from the closest strainer.

Therefore, there is no possibility of impacting the strainers with a whipping pipe.

The HELB report (GENE-0000-0048-9192) has been revised to provide definitive statements in the conclusions concerning pipe whip and. missile impacts to the new strainer assembly.

0115 Response - Part 2 The curves in the specification have a designation on the lower right of the curve that states these are 2% damping.

01.15 Response - Part 3 The sump strainer design specification, GE Design Specification 26A6870, and the stress analysis report, GENE-0000-0054-9349, have been revised to correct errors and discrepancies.

01.15 Response - Part 4 The design specification (GE 26A6870) and the stress analysis report (GENE-0000-0054-9349) have both been revised to address the temperature delta in the sump strainer stress analysis report.

The questions pertaining to the temperature delta did not result in significant changes since the equipment is constructed of a single material, austenitic stainless steel.

Therefore, there are no significant differential thermal expansions within the structure and no thermal stresses would be developed.

01.15 Response - Part 5 The inertial acceleration values used in the analyses were extracted directly from the Design Envelope spectra contained in the design specification. However, the values reported in the stress report are the ANSYS input values adjusted to account for hydrodynamic mass and debris load to facilitate the ANSYS analyses. Since hydrodynamic mass and debris load are also reported in these same tables, the reader cannot ascertain what was actually used in the analysis. Therefore, the stress report (GENE-0000-0054-9349) was revised to reflect the seismic accelerations specified in the design specification.

01.15 Response - Part 6 The values for E*/E and K in the revision of the stress analysis report reviewed by the staff during the NRC GSI-191 audit at Waterford 3 were incorrect. A review of the displacement numbers showed that the calculation E*/E is 0.43 and K = 2.33. Since the original values used were incorrect, the stresses were reevaluated with the new and to W3F1-2008-0069 Page 71 of 87 correct values. All safety factors will remain above 1.0 even with the increased and correct values. These new and corrected values were included in the latest revision of the stress report, GENE-0000-0054-9349.

01.15 Response - Part 7 The deflection results in the reviewed version of the stress report (GENE-0000-0054-9349) were provided to assist the review in understanding the behavior of the structure, and had no significance beyond that. The latest revision of the stress report (GENE-0000-0054-9349) corrected the labels.

01.15 Response - Part 8 The errors in the stress report (GENE-0000-0054-9349) 'Were revised in the latest revision. There was no adverse impact on the structural adequacy conclusions.

01.15 Response - Part 9 The discrepancies in the hydrodynamic mass analysis have been corrected. The corrections are in the latest revision to GE calculation GENE-0000-0054-9349 (Reference 56). There was no adverse impact on the structural adequacy conclusion.

01. 16 The licensee should summarize how it has evaluated the potential for holdup in the refueling cavity due to falling debris.

01.16 Response Calculation 2005-05500 (Reference 29) documents that the refueling cavity has two 6-inch drain lines (without screens) that drain to the containment floor, and by one 4-inch line that drains to the containment sump. In the event that large debris is propelled over the SG cavity walls into the refueling cavity, the debris must land on a drain in order to clog it since the velocities in the cavity are too low to transport a large piece of debris to a drain. During plant operations, the Upper Guide Structure Lift Rig (UGSLR) is stored directly above one of the 6" drains. The UGSLR will prevent any debris larger than 6" from falling directly on the drain. The diver stairs are located above the other 6" drain and are permanently mounted in the refueling cavity. These stairs will prevent any debris larger than 6" from falling directly onto the drain. Any smaller debris that transports to these two 6" drains will pass through the drains since they do not contain screens. Therefore, there will always be drainage available from the refueling cavity to the active pool.

01.17 The licensee should provide the results of a similitude evaluation for WCAP-16406-P versus conditions at Waterford 3.

01.17 Response In response to the NRC Safety Evaluation for WCAP-16406-P, the Waterford 3 Downstream Effects analysis specifically addresses each of the 31 limitations identified in the Safety Evaluation. The responses to these limitations justify the use of WCAP-16406-P at Waterford 3. Below are the 31 limitations with corresponding responses from calculation 2005-12840 Revision 1 (Reference 32).

to W3F1-2008-0069 Page 72 of 87

1. Where a TR WCAP-16406-P, Revision 1, section or appendix refers to examples, tests, or general technical data, a licensee should compare and verify that the information is applicable to its analysis.

In general, examples were not used for site specific input. The wear equations developed in the WCAP were developed and benchmarked on equipment and with debris similar to that found at Waterford Unit 3.

2. A discussion of EOPs, AOPs, NOPs or other plant-reviewed alternate system line-ups should be included in the overall system and component evaluations as noted in the NRC staff's SE of NEI 04-07, Section 7.3 (Reference 3).

The scope of'equipment to be reviewed for wear and blockage was defined in evaluation 2005-02820, GSI-191 Downstream Effects - Flow Clearances, Revision 0 (Reference 31). . That calculation identified the equipment that could be in the recirculation path following a postulated accident.

3. A licensee using TR WCAP-16406-P, Revision 1, will need to determine its own specific sump debris mixture and sump screen size in order to initiate the evaluation.

Site specific debris generation and transport calculations (Reference 28 and 29) were completed and referenced by this wear calculation as the source of debris.

Screen information was taken from the site specific design documents defining the screens and from site specific debris bypass testing.

4. TR WCAP-16406-P, Revision 1, Section 4.2, provides a general discussion of system and component mission times. It does not define specific times, but indicates that the defined term of operation is plant-specific. As stated in the NRC staffs SE of NEI 04-07, Section 7.3 (Reference 3), each licensee should define and provide adequate basis for the mission time(s) used in its downstream evaluation.

Site specific design and licensing information is used to determine the applicable mission time of equipment evaluated in this calculation.

5. TR WCAP-16406-P, Revision 1, Section 5.8, assumes that the coolant which is not spilled flows into the reactor system and reaches the reactor vessel downcomer. This would be true for most PWR designs except for plants with UPI. Therefore, the methodology of Section 5.8 may not be applicable to plants with UPI and its use should be justified on a plant-specific basis.

The Waterford Unit 3 station utilizes lower plenum injection.

6. TR WCAP-16406-P, Revision 1, Section 5.8, provides equations which a licensee might use to determine particulate concentration in the coolant as a function of time.

Assumptions as to the initial particulate debris concentration are plant-specific and should be determined by the licensee. In addition, model assumptions for ECCS flow rate, the fraction of coolant spilled from the break and the partition of large heavy particles which will settle in the lower plenum and smaller lighter particles which will not settle should be determined and justified by the licensee.

to W3F1-2008-0069 Page 73 of 87 Debris depletion in this calculation is based on plant specific flows, debris types and debris size distributions. The debris depletion methodology is described in Appendix A and the settling size is calculated in Appendix C.

7. TR WCAP-16406-P, Revision 1, Sections 5.8 and 5.9, assumes that debris settling is governed by force balance methods of TR Section 9.2.2 or Stokes Law. The effect of debris and dissolved materials on long-term cooling is being evaluated under TR WCAP-16793-NP (Reference 12). If the results of TR WCAP-16793-NP show that debris settling is not governed by force balance methods of TR Section 9.2.2 or Stokes Law, then the core settling term determined from TR WCAP-16793-NP should be used.

The site specific debris settling size is determined in Appendix C. The methodology uses empirical friction factors based on the debris shape. This methodology is benchmarked against the NRC-sponsored testing of paint chip settling reported in NUREG/CR-6916.

8. TR WCAP-16406-P, Revision 1, Section 7ý2, assumes a mission time of 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> for pump operation. Licensees should confirm that 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> bounds their mission time or provide a basis for the use of a shorter period of required operation.

A 720 hour0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> mission time is used at Waterford Unit 3. This is in accordance with NRC SER to NEI 04-07 and the plant licensing documents.

9. TR WCAP-16406-P, Revision 1, Section 7.2, addresses wear rate evaluation methods for pumps. Two types of wear are discussed: 1) free-flowing abrasive wear and 2) packing-type abrasive wear. Wear within close-tolerance, high-speed components is a complex analysis. The actual abrasive wear phenomena will likely not be either a classic free-flowing or packing wear case, but a combination of the two. Licensees should consider both in their evaluation of their components.

This calculation considers the maximum of the calculated free-flow or packing type wear until a gap of 4 times the original gap is reached. Beyond that point, free-flow wear is modeled.

10. TR WCAP-16406-P, Revision 1, Section 7.2.1.1, addresses debris depletion coefficients. Depletion coefficients are plant-specific values determined from plant-specific calculations, analysis, or bypass testing. Licensees should consider both hot-leg and cold-leg break scenarios to determine the worst case conditions for use in their plant-specific determination of debris depletion coefficient.

Both hot-leg and cold-leg break scenarios are considered in this calculation in the determination of the debris concentrations. The wear calculation is based on the larger of the two debris concentrations.

11. TR WCAP-16406-P, Revision 1, Section 7.3.2.3, recognizes that material hardness has an effect on erosive wear. TR WCAP-16406-P, Revision 1, suggests that "For elastomers, the wear rate is at least one order of magnitude less than steel.

Therefore, for soft-seated valves, divide the estimated wear rate of steel from above equations by 10 per Appendix F." The NRC staff agrees that the wear rates of to W3F1-2008-0069 Page 74 of 87 elastomers are significantly less than for steels. However, the wear coefficient should be determined by use of a suitable reference, not by dividing the steel rate by a factor of 10.

Wear of elastomeric materials is not included in this calculation.

12. TR WCAP-16406-P, Revision 1, Section 8.1.1.2, "Evaluation of ECCS Pumps for Operation with Debris-Laden Water from the Containment Sump," states that "Sufficient time is available to isolate the leakage from the failed pump seal and start operation of an alternate ECCS or CSS train." Also, Section 8.1.3, "Mechanical Shaft Seal Assembly," states: "Should the cooling water to the seal cooler be lost, the additional risk for seal failure is small for the required mission time for these pumps."

These statements refer only to assessing seal leakage in the context of pump operability and 10 CFR Part 100 concerns. A licensee should evaluate leakage in the context of room habitability and room equipment operation and environmental qualification, if the' calculated leakage is outside that which has been previously assumed.

This calculation determines the seal leakage rate that would occur if the primary seal surfaces are assumed to fail. If the failure of the primary seals is a result of debris from the original break, then the primary seal leakage would be part of the design basis and considered in the appropriate flooding and HVAC analyses. However, the seals are not considered to fail as a direct result of the accident. Therefore, except as specifically required by SRP 15.6.5 for offsite dose calculation, a second passive failure in the ECCS system is not part of the design basis of this plant. A postulated passive failure of the primary seal in a single pump would be within the design basis, but would be bounded by a moderate energy line break in the pump room.

13. TR WCAP-16406-P, Revision 1, Section 8.1.3, discusses cyclone separator operation. TR WCAP-16406-P, Revision 1, generically concludes that cyclone separators are not desirable during post-LOCA operation of HHSI pumps. The NRC staff does not agree with this generic statement. If a licensee pump contains a cyclone separator, it should be evaluated within the context of both normal and accident operation. The evaluation of cyclone separators is plant-specific and depends on cyclone separator design and the piping arrangement for a pump's seal injection system.

There are cyclone separators in the ECCS system of this plant. These separators have not been tested. Therefore, further investigation as to the acceptability of these cyclone separators is required.

14. TR WCAP-16406-P, Revision 1, Section 8.1.4, refers to pump vibration evaluations.

The effect of stop/start pump operation is addressed only in the context of clean water operation, as noted in Section 8.1.4.5 of TR WCAP-16406-P, Revision 1. If an ECCS or CSS pump is operated for a period of time and builds up a debris "packing" in the tight clearances, stops and starts again, the wear rates of those areas may be different due to additional packing or imbedding of material on those wear surfaces.

Licensees who use stop/start operation as part of their overall ECCS or CSS operational plan should address this situation in their evaluation.

to W3F1-2008-0069 Page 75 of 87 This calculation utilizes the Archard's method in determining wear. No credit is taken for stai-t/stop operation to reduce the mission time. A packing wear model is used for the duration of the mission time or until a clearance of 4 times the original clearance is achieved.

15. TR WCAP-16406-P, Revision 1, Section 8.1.4, states: "should the multistage ECCS pumps be operated at flow rates below 40% of BEP during the containment recirculation, one or more of the pumps should be secured to bring the flow rate of the remaining pump(s) above this flow rate." The NRC staff does not agree with this statement. System line-ups and pump operation and operating point assessment are the responsibility of the licensee. Licensees must ensure that their ECCS pumps are capable of performing their intended function and the NRC has no requirements as to their operating point during the recirculation phase of a LOCA.

No credit is taken for securing ECCS pumps during the course of a postulated accident. In general shut-off head is used where maximum dP would result in conservative wear calculations. Run-out is used when minimal head or maximum flow would result in conservative calculation results.

16. TR WCAP-16406-P, Revision 1, Section 8.1.5, makes a generic statement that all SI pumps have wear rings that are good "as new" based solely upon "very little service beyond inservice testing." A stronger basis is needed to validate this assumption, if used (e.g., maintenance, test and operational history and/or other supporting data).

This calculation used in-service testing (IST) results to predict wear to the time when the pump degradation would be detected and corrected.

17. TR WCAP-1 6406-P, Revision 1, Section 8.3, identifies criteria for consideration of tube plugging. Licensees should confirm that the fluid velocity going through the heat exchanger is greater than the particle settling velocity and evaluate heat exchanger plugging if the fluid velocity is less than the settling velocity.

Flow velocity in the heat exchangers is calculated and evaluated for settling.

18. TR WCAP-16406-P, Revision 1, Section 8.6, refers to evaluation of instrumentation tubing and system piping. Plugging evaluations of instrument lines may be based on system flow and material settling velocities, but they must consider local velocities and low-flow areas due to specific plant configuration.

Instrument tubing in this analysis is evaluated based on the tubing being at or above the horizontal, rather than on velocity considerations.

19. TR WCAP-16406-P, Revision 1, Sections 8.6.7, 8.6.8, 8.6.9, and 8.6.10 describe, in general terms, the Westinghouse, CE, and B&W RVLIS. TR WCAP-16406-P, Revision 1, recommends that licensees evaluate their specific configuration to confirm that a debris loading due to settlement in the reactor vessel does not effect the operation of its RVLIS. The evaluation of specific RVLIS design and operation is outside the scope of this SE and should be performed in the context of a licensees reactor fuel and vessel evaluations.

to W3F1-2008-0069 Page 76 of 87 Reactor fuel and vessel evaluations are outside the scope of this calculation.

20. TR WCAP-16406-P, Revision 1, Section 8.7, refers to evaluation of system piping.

Plugging evaluations of system piping should be based on system flow and material settling velocities. Licensees should consider the effects of local velocities and low-flow areas due to specific plant configuration. A piping wear evaluation using the free-flowing wear model outlined in Section 7 should be performed for piping systems. The evaluation should consider localized high-velocity and high-turbulence areas. A piping vibration assessment should be performed if areas of plugging or high localized wear are identified.

The wear in high velocity areas such as orifices was calculated. The amount of wear was minimal and would bound the wear to other areas in the piping systems.

Therefore, numeric calculation of wear in general piping areas was not performed.

21. TR WCAP-16406-P, Revision 1, Section 9, addresses reactor internal and fuel blockage evaluations. This SE summarizes seven issues regarding the evaluation of reactor internal and fuel. The PWROG indicated that the methodology presented in TR WCAP-16793-NP (Reference 15) will address the seven issues. Licensees should refer to TR WCAP-16793-NP and the NRC staffs SE of the TR WCAP-16793-NP, in performing their reactor internal and fuel blockage evaluations. The NRC staff has reached no conclusions regarding the information presented in TR WCAP-16406-P, Section 9.

Reactor internal and fuel blockage is outside the scope of this calculation.

22. TR WCAP-16406-P, Revision 1, Table 4.2-1, defines a plant Category based on its Low-Head / Pressure Safety Injection to RCS Hot-Leg Capability. Figure 10.4-2 implies that Category 2 and 4 plants can justify LHSI for hot-leg recirculation.

However, these categories of plants only have one hot-leg injection pathway.

Category 2 and Category 4 plant licensees should confirm that taking credit for the single hot-leg injection pathway for their plant is consistent with their current hot-leg recirculation licensing basis.

Waterford Unit 3 is a Category 1 plant.

23. TR WCAP-16406-P, Revision 1, Appendix F, discusses component wear models.

Prior to using the free-flowing abrasive model for pump wear, the licensee should show.that the benchmarked data is similar to or bounds its plant conditions.

The free flow wear model is used in those pump area that were shown to have low wear in the Davis Besse pump wear testing. The low wear areas were areas where the branch of the flow stream carrying the debris had to turn 180 degrees and travel back toward the center of the pump while the main stream continued outward. These areas act as cyclone separators due to their geometry. The maximum of packing or free flow wear was used in other areas of the pumps.

The pumps gaps at Waterford 3 showed the same geometry as the benchmarked data. The flow paths through the pump gaps were analyzed and the maximum of packing or free flow wear was used when applicable.

to W3F17-2008-0069 Page 77 of 87

24. TR WCAP-16406-P, Revision 1, Appendix H, references American Petroleum Institute (API) Standard 610, Annex 1 eighth edition. This standard is for newly manufactured pumps. Licensees should verify that their pumps are "as good as new" prior to using the analysis methods of API-610. This validation may be in the form of maintenance records, maintenance history, or testing that documents that the as-found condition of their pumps.

This calculation used in-service testing (IST) results to predict wear to the time when the pump degradation would be detected and corrected.

25. TR WCAP-16406-P, Revision 1, Appendix I, provides guidelines for the treatment, categorization and amount of DBA Qualified, DBA Acceptable, Indeterminate, DBA Unqualified, and DBA Unacceptable coatings to be used in a licensee's downstream sump debris evaluation. A technical review of coatings generated during a DBA is not within the scope of this SE. For guidance regarding this subject see the NRC staffs SE of NEI-04-07 (Reference 3) Section 3.4 "Debris Generation."

Debris generation by debris type is not within the scope of this calculation.

26. TR WCAP-16406-P, Revision 1, Appendix J, derives an approach to determining a generic characteristic size of deformable material that will pass through a strainer hole. This approach is only applicable to screens and is not applicable to determining material that will pass through other close tolerance equipment.

Identification of close tolerance passages is not in the scope of this calculation. The criteria used in the phase I review of downstream components was that holes less than twice the screen hole size required further evaluation.

27. TR WCAP-16406-P, Revision 1, Appendix 0, Section 2.2, states that the wear coefficient, K, in the Archard Model is determined from testing. The wear coefficient (K) is more uncertain than the load centering approach and K may vary widely.

Therefore, licensees should provide a clear basis, in their evaluation, for their selection of a wear coefficient.

The wear coefficient used results in calculated wear greater than the amount seen in the Davis-Besse t testing. The materials, debris types and concentrations are comparable. Therefore, the k value presented the WCAP-1 6406-P appears to be the best conservative information available on ECCS pump wear when exposed to insulation and coating debris.

28. TR WCAP-16406-P, Revision 1, Appendix P, provides a method to estimate a packing load for use in Archard's wear model. The method presented was benchmarked for a single situation. Licensees are expected to provide a discussion as to the similarity and applicability to their conditions. The licensee should incorporate its own specific design parameters when using this method.

This calculation utilized the methodology discussed in Appendix 0 of WcAP 16406-P (centering load) for defining loads to be used in the packing wear model.

to W3F1-2008-0069 Page 78 of 87

29. TR WCAP-16406-P, Revision 1, Appendix Q, discusses bounding debris concentrations. Debris concentrations are plant-specific. If 9.02E-5 (mils/hr)/10 PPM is to be used as the free flowing abrasive wear constant, the licensee should show how it is bounding or representative of its plant.

The combined debris concentration of abrasive particulates and fibers used for free-flowing abrasive wear does not exceed 720 ppm. Therefore, the extrapolated wear rate [9.02E-5 (mils/hr)/10 ppm] is not required to be used in this calculation.

30. TR WCAP-16406-P, Revision 1, Appendix R, evaluates a Pacific 11-Stage 2.5" RLIJ pump. The analysis was performed by the PWROG using specific inputs. ECCS pumps with running clearance designs and dimensions significantly different than those covered by the analysis should be subjected to pump-specific analysis to determine the support stiffness based on asymmetric wear. If licensees use the aforementioned example, a similarity evaluation should be performed showing how the example is similar to or bounds their situations.

Themulti-stage HPSI pump was evaluated by finding the stiffness at the uniform increase in clearance equal to 2X as the as-new clearance. The stiffness of the pumps after normal wear and debris induced wear was considered and then calculated. The stiffness of the pump after normal and LOCA asymmetric wear was compared to the allowed stiffness equivalent to a uniform 2X initial clearance to judge the acceptability of the pump.

31. Licensees should compare the design and operating characteristics of the Pacific 2.5" RLIJ 11 to their specific pumps prior to using the results of Appendix S in their component analyses.

As stated in response 30 above, specific stiffness calculations were performed for all applicable pumps using a stiffness corresponding to 2X the as-new clearance as the acceptance criteria.

01.18 The licensee should provide the assumptions, the bases for assumption and the source documents for its downstream evaluation of components and systems.

01.18 Response Following the NRC Audit of Waterford 3's Generic Letter 2004-02 efforts, Waterford 3 re-performed its Downstream Effects evaluations in accordance with WCAP-16406-P and corresponding NRC Safety Evaluation. The preliminary Downstream Effects evaluation reviewed by the NRC during the Audit had not been accepted by Waterford 3 and was only provided by Waterford's vendor at the request of the NRC. Bases and source documents for all assumptions have been provided directly in the analysis or through supporting analyses.

to W3F1-2008-0069 Page 79 of 87 -

01.19 The licensee should provide clearly defined technical bases for the designated mission times for shutdown cooling, high pressure safety injection and containment spray.

01.19 Response A review of the Waterford 3 UFSAR, specifically Table 15.6-18, indicates that the analyzed mission time for the LOCA event is 30 days. During the 30 day mission time, two trains of High Pressure Safety Injection and Containment Spray are conservatively assumed to operate continuously.

01.20 The licensee should develop and justify conservative, b6ounding values for system lineups, fluid flows and system pressures for the downstream effects components and systems analysis.

01.20 Response The downstream analysis in calculation 2005-12840 revision 1 is based on the most conservative system lineup. Lineups were determined using plant operating procedures.

01.21 The licensee should justify the use of design curves, or re-analyze for degraded, actual or modified pump curves for the downstream effects components and systems analysis.

01.21 Response Calculation 2005-12840 revision 1 assumes pump performance is at its IST limits and/or utilizes IST test data to determine the amount of degradation expected over the life of the plant for the wear analysis starting point.

01.22 The licensee should provide the results of analysis of emergency core cooling system (ECCS) air entrainment (apart from vortexing) and the potential for waterhammer and slug flow.

01.22 Response The Waterford 3 ECCS is designed as a water solid system. Post refueling outage, ultrasonic testing is performed at potential void formation points in the ECCS system. If a void is found, the system is flushed eliminating the void. The system. is checked multiple times to verify that all voids are eliminated. Various waterhammer analyses have been performed on the ECCS system throughout the life of the plant. The modifications done to date for Waterford 3 do not affect any existing analysis nor create the potential for new waterhammer events.

01.23 The licensee should re-calculate downstream component wear due to strainer bypass debris and provide the results.

01.23 Response Calculation 2005-12840 Revision 1 has been revised to meet the requirements of WCAP-16406-P and NCR SER requirements. Debris loading used in the evaluation is based on site specific debris generation analysis, transport analysis, and bypass testing.

to W3F1-2008-0069 Page 80 of 87 01.24 The licensee should re-perform its high-pressure safety injection recirculation throttle valve clogging analysis considering the full range of possible recirculation throttle valve positions or failure of the HPSI recirculationthrottle valve to open to its pre-set position, and provide the results.

01.24 Response From Calculation 2005-02820 (Reference 31), the throttle valves are:

" HPSI injection header valves, which are Target Rock 2-inch motor operated globe valves, and

" RC loop hot leg valves, which are Anchor Darling 3-inch motor operated globe valves.

The above motor operated valves are used to balance the flow between the hot leg and cold leg and between the four cold leg injection lines. These valves are throttled from the control room and their flow rates are monitored during a LOCA. These valves can be throttled down to maintain the required flow in the event of excessive wear or opened as excessive resistance due to unexpected clogging is seen. Due to this, no wear calculations are performed.

01.25 The licensee should describe how it has incorporated actions of its operational procedures into the downstream effects evaluation.

01.25 Response The Waterford 3 downstream effect analysis evaluated the ECCS and CS systems in all configurations as defined by operational procedures. The Waterford 3 operational procedures do not require the securing of HPSI or CS pumps at any time, therefore these pumps were analyzed to operate for a full 30 day mission time. At 1-2 hours post RAS, the ECCS system switches from cold leg only injection to simultaneous cold and hot leg injection. Both configurations are considered and addressed in the downstream analysis. Conservative flows' were used for all components to bound any possible operating flow.

01.26 The licensee should justify emergency core cooling system (ECCS) pump wear rings to be "goodas new," or determine a more conservative condition for these rings.

01.26 Response See item 24 under Open Item 17.

01.27 The licensee should provide the results of evaluation of high-pressure safety injection (HPSI) pump stage-to-stage degradation and its effect on pump hydraulic performance, and should provide the results of a pump vibration and rotor dynamics evaluation.

01.27 Response Downstream Effects evaluation 2005-12840 Revision 1 determined that worn condition of the pump and generated performance curves at various time points during the 30 day mission time. The worn performance curves were inputted into the Waterford 3 Long to W3F1-2008-0069 Page 81 of 87 Term Cooling (LTC) Analysis of Record (AOR) to determine acceptability. The result from the LTC AOR evaluation is that adequate core cooling is maintained and the AOR remains valid. The worn condition, wear ring clearances, of the pump were used in a rotor-dynamic analysis to confirm that the pump remains dynamically stable thorough the mission time.

01.28 The licensee should summarize its evaluation of ECCS and CS pump leakage effects in its Safeguards Room.

01.28 Response Waterford 3 currently has no analysis for dose considerations in the Safeguards Room as there are no required operator actions in this area during a LOCA. Control Room dose analysis assumes at least 0.5 gpm total leakage from the Engineered Safety Feature pumps. The downstream analysis performed concluded that degradation of the pump seals is not expected therefore no additional leakage should occur.

01.29 The licensee should summarize how it has determined the effects of settled material at emergency core cooling system (ECCS) low points and integrate these effects into the downstream effects evaluation.

01.29 Response Flow velocities in the ECCS piping system remains relatively high such that settling of material that bypasses the strainers should be minimal and not affect system performance:.

01.30 The licensee should consider the results of the various component wear evaluationsand perform an overall system flow evaluation, and should provide a summary of the results.

01.30 Response The downstream effects calculation 2005-12480 Revision 1 concluded that all components with the exception of the High Pressure Safety Injection Pumps experienced negligible wear based on WCAP-16406-P acceptance criteria., Degraded performance curves were developed for the High Pressure Pumps. These degraded curved were utilized in a Long Term Cooling study to confirm their acceptability in maintain core cooling. Based on the analyses perform concluding that negligible wear occurs on the system components and that acceptability of the degraded High Pressure Safety injection pumps, no overall system flow evaluation was deemed necessary.

01.31 The. licensee should provide the results of an analysis of downstream effects of post-LOCA debris and chemicals on the fuel and vessel.

01.31 Response The in-vessel effects evaluation was performed in accordance with the guidance in WCAP-16793 and the initial NRC comments provided related to use of that document.

This evaluation did not indicate problems with reactor core cooling. The fuel deposit analysis was performed per the WCAP-16793 spreadsheet with conservatively bounding to W3F1-2008-0069 Page 82 of 87 inputs relative to the maximum debris loading conditions for the plant. This analysis determined that significant margin exists relative to the acceptance criteria, with total deposition thickness of <13 mils, remaining well below the 50-mil maximum value and the maximum clad temperature of <328°F also remaining well below the 800'F acceptance criteria. The initial NRC comments provided for WCAP-16793 have been withdrawn and the WCAP is currently in revision, although the source of the revision is understood to be related to the fuel blockage analysis, not the fuel deposit methodology.

Following the issuance of the revised guidance, further analysis could be necessary.

01.32 The licensee should provide the results of resolution of chemical effects at Waterford 3.

01.32 Response Chemical precipitates that form in the post-LOCA containment environment combined with debris do not result in an unacceptable head loss. Head loss due to chemical precipitates and debris is demonstrated by test using WCAP-16530-NP (Reference 58) methods with relatively minor modifications.

The Alion Science and Technology 30-day integrated chemical effects testing identified that calcium phosphate precipitates can form very early post-LOCA due to the very low and retrograde solubility of calcium phosphate (lower solubility at high temperature). The 30-day integrated testing also identified that aluminum based precipitants do not form until the post-LOCA environment has cooled to below 140 degrees F. The prototype testing used these results to sequence the WCAP-16530-NP/16785-NP (References 58 and 53) based precipitates. Head loss calculations used the head loss attributed to calcium phosphate to determine the head loss across the strainer at temperatures greater than 140 degrees F when the NPSH margin is limiting. The 30 day integrated testing and analyses concluded that no aluminum based precipitates would form in the Waterford 3 environmental conditions with a pH less than 8.1; therefore any reduction in the aluminum oxy-hydroxide precipitate is reasonable.

to W3F1-2008-0069 Page 83 of 87 References

1. NRC Generic Letter (GL) 2004-02, "Potential Impact of Debris Blockage on Emergency Recirculation during Design Basis Accidents for Pressurized-Water Reactors," dated September 13, 2004.

2. Nuclear Energy Institute (NEI) document NEI 04-07 Revision 0, December 2004, "Pressurized Water Reactor Sump Performance Evaluation Methodology."

3. Safety Evaluation by the Office of Nuclear Reactor Regulation Related to NRC Generic Letter 2004-02, Nuclear Energy Institute Guidance Report (Proposed Document Number NEI 04-07), "'Pressurized Water Reactor Sump Performance Evaluation Methodology,"

Issued December 6, 2004.

4. Regulatory Guide 1.82, "Water Sources for Long Term Recirculation Cooling, Following a Loss of Coolant Accident," Revision 3, November 2003.

5. NUREG-0800, "U.S. Nuclear Regulatory Commission Standard Review Plan," Section 3.6.2, "Determination of Rupture Locations and Dynamic Effects Associated with the Postulated Rupture of Piping,"Revision 1, July 1981.

6. NUREG/CR-2791, "Methodology for Evaluation of Insulation Debris Effects, Containment Emergency Sump Performance Unresolved Safety Issue A-43," Issued September 1982.

7. NUREG/CR-3616, "Transport and Screen Blockage Characteristics of Reflective Metallic Insulation Materials," January 1984.

8. NUREG/CR-6224, "Parametric Study of the Potential for BWR ECCS Strainer Blockage Due to LOCA Generated Debris, Final Report," Issued October 1995.

9. NUREG/CR-6369, "Drywell Debris Transport Study, Final Report," Volume 1, Issued September 1999.

10. NUREG/CR-6369, "Drywell Debris Transport Study: Experimental Work, Final Report,"

Volume 2, Issued September 1999.

11. NUREG/CR-6369, "Drywell Debris Transport Study: Computational Work, Final Report,"_

Volume 3, Issued September 1999.

12. NUREG/CR-6762, Volume 1, "GSI-191 Technical Assessment: Parametric Evaluations for Pressurized Water Reactor Recirculation Sump Performance," Issued August 2002.

13. NUREG/CR-6762, Volume 2, "GSI-191 Technical Assessment: Summary and Analysis of U.S. Pressurized Water Reactor Industry Survey Responses and Responses to GL 97-04," Issued August 2002.

14. NUREG/CR-6762, Volume 3, "GSI-191 Technical Assessment: Development of Debris Generation Quantities in Support of the Parametric Evaluation," Issued August 2002.

to W3F1-2008-0069 Page 84 of 87

15. NUREG/CR-6762, Volume 4, "GSI-191 Technical Assessment: Development of Debris Transport Fractions in Support of the Parametric Evaluation," Issued August 2002.

16. NUREG/CR-6772, "GSI-191: Separate Effects Characterization of Debris Transport in Water," Issued August 2002.

17. NUREG/CR-6773, "GSI-1 91: Integrated Debris-Transport Tests in Water Using Simulated Containment Floor Geometries," Issued December 2002.

18. NUREG/CR-6808, "Knowledge Base for the Effect of Debris on Pressurized Water Reactor Emergency Core Cooling Sump Performance," Issued February 2003.

i9. NUREG/CR-6916, "Hydraulic Transport of Coating Debris, A Subtask of GSI-191," Issued December 2006.

20. Nuclear Energy Institute (NEI) Document 02-01, "Condition Assessment Guidelines:

Debris Sources Inside PWR Containments," Revision 1.

21. Not used.

22. WCAP-16568-P, "Jet Impingement Testing to Determine the Zone of Influence (ZOI) for DBA-Qualified / Acceptable Coatings," Revision 0.

23. C.D.I. Report 96-06, "Air Jet Impact Testing of Fibrous and Reflective Metallic Insulation,"

Revision A, included in Volume 3 of General Electric Document NEDO-32686-A, "Utility Resolution Guide for ECCS Suction Strainer Blockage."

24. Not used.

25. Alion Document No: ALION-REP-TXU-4464-02, Titled: TXU Paint Chip Characterization, Rev. 0.

26. Letter # TXX-07156 from Mike Blevins, Luminant Generation Company LLC, to the U.S.

Nuclear Regulatory Commission, dated November 8, 2007.

27. Letter from Jon R. Cavallo, Vice President of Corrosion Control Consultants and Labs Inc.

to Charles Feist, dated September 20, 2007.

28. Calculation 2004-07780, "Debris Generation Due to LOCA within Containment for Resolution of GL GSI-191," Revision 3.

29. Calculation 2005-05500, "Post-LOCA Debris Transport, Head Loss Across Safety Injection Sump Screen, and NPSH Evaluation for Resolution of GSI-191," Revision 2.

30. ER-W3-2003-0394-001, "Safety Injection Sump Modification."

31. Calculation 2005-02820, "GSI-191 Downstream Effects - Flow Clearances," Revision 0, dated August 18, 2005.

to W3F1-2008-0069 Page 85 of 87

32. Calculation, 2005-12840, "Evaluation of Downstream Components for Long Term Performance for Resolution of GSI-1 91," Revision 1, dated May 11, 2008.

33. Head Loss Testing of Waterford Unit 3 Safety injection Sump Strainers, 26A6833, Rev 9.

34. S0105 Task Design Input Request (DIR), Rev 5, DRF Object 0000-0079-0092.

35. Hydraulic Sizing and Head Loss Prediction for Suction Strainers (PWRs), TDP-01 86.

36. Safety Injection Pump Passive ECCS Strainer System S0100 Hydraulic Sizing Report, Waterford Unit 3 Nuclear Power Plant, Document No. GENE-0000-0053-4416-P-R4.

37. Calculation MNQ6-4, "Water Levels Inside Containment."

38. WCAP-16710-P, Rev. 0, "Jet Impingement Testing to Determine the Zone of Influence (ZOI) of Min-K and NUKON Insulation for Wolf Creek and Callaway Nuclear Operating Plants."

39. ALION-REP-ENTG-4771-02, Rev. 0, "Waterford 3 Metal Encapsulated Fiberglass Insulation ZOI and Size Distribution Report."

40. Calculation ECM89-083, Rev. 1, "Verification of Gaps at Whip Restraint U-Bolt for Min-K Insulation Required per C1258220, 265936, 266235, and 266371.

41. ALION-REP-ENT-4536-02, Rev. 0, "Waterford Unit 3 Low Density Fiberglass Debris Erosion Testing Report."

42. Procedure NOECP-451, Rev. 1, "Conducting Engineering Inspection of Reactor Containment Building Protective Coatings."

43. Specification 1564.734, Rev. 19, "General Protective Coating for Nuclear Power Plant."

44. NRC Letter from N. Kalyanam to K. Walsh, 12/10/07, "Waterford Steam Electric Station, Unit 3- Approval of Extension Request for Corrective Actions Re. Generic Letter 2004-02, "Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized Water Reactors". (Waterford 3 document # ILN07-01160).

45. Procedure OP-903-027, "Inspection of Containment."

46. Procedure PMC-002-007, "Maintenance and Construction Painting."

47. Procedure UNT-007-006, "Housekeeping."

48. Procedure W4.202, "System and Component Labeling."

49. Procedure EN-MA-118, "Foreign Material Exclusion."

50. Not Used.

to W3F1-2008-0069 Page 86 of 87

51. Calculation ECM07-001, "NPSH Analysis of Safety Injection and Containment Spray Pumps."

52. Regulatory Guide 1.1, "Net Positive Suction Head For Emergency Core Cooling and Containment Heat Removal System Pumps."

53. WCAP 16785-NP, "Evaluation of Additional Inputs to the WCAP-16530-NP Chemical Model."

54. WCAP-16406-P, Rev. 1, "Evaluation of Downstream Sump Debris Effects in Support of GSI-191.

55. Waterford 3 Letter W3F1-2007-0051, November 14, 2007, "Request for Extension of Completion Date for Corrective Action Required by GL 2004-02."

56. GE Report GENE-0000-0054-9349, Waterford 3 Safety Injection Sump Strainer, Plenum, and Sensor Stress Report.

57. NRC Letter from N. Kalyanam to K. Walsh, 1/28/08, "Waterford Steam Electric Station, Unit 3 - Report on Results of Staff Audit of Corrective Actions to Address Generic Letter 2004-02." (Waterford Document ILN03-0015).

58. WCAP-16530-NP, "Evaluation of Post-Accident Chemical Effects in Containment Sump Fluids to Support GSI-1 91."

59. NRC Letter from Thomas G. Hiltz to K. Walsh, May 22, 2008, "Waterford Steam Electric Station, Unit - Generic Letter 2004-02 "Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized Water Reactors" Approval of Extension Request."

60. WCAP-16793-NP, "Evaluation of Long-Term Cooling Considering Particulate, Fibrous and Chemical Debris in the Recirculating Fluid."

61. NRC Letter from William H. Ruland to Anthony R. Pietrangelo, February 4, 2008, "Draft Conditions and Limitations for Use of Westinghouse Topical Report WCAP-16793-NP, Revision 0, "Evaluation of Long-Term Cooling Considering Particulate, Fibrous and Chemical Debris in the Recirculating Fluid."

62. NRC Letter from William H. Ruland to Anthony R. Pietrangelo, dated November 30, 2007, "Supplemental Licensee Responses to Generic Letter 2004-02, "Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized Water Reactors."

63. NRC Letter from William H. Ruland to Anthony R. Pietrangelo, dated November 21, 2007, "Revised Content Guide for Generic Letter 2004-02 Supplemental Responses."

64. GE Report GENE-0000-0048-9192, "Waterford 3 High Energy Line Postulated Pipe Break Evaluation, Containment Sump Strainer."

65. Calculation CN-SEE-I-08-42, "Waterford 3 Nuclear Plant LOCADM."

to W3F1-2008-0069 Page 87 of 87

66. Drawing 5817-13604, Rev. 0, SIS Sump Strainer Interface Control Drawing sheet 1.

67. Calculation 1062-0802-0015-4, "Evaluation of Waterford 3 and ANO 2 HPSI Pump Mechanical Seals."

68. Calculation 1062-0015-03, "Evaluation of Applicability of Debris Laden Test Data to the ECCS Pumps' Cyclone Separators at Waterford 3."

69. Specification 1564.116, Revision 6, "Containment Spray Pumps."

70. Drawing 5817-11683, Revision 1, "Containment Spray Pumps A & B Seal & Water Piping."

71. Vendor Manual TD-C681.0015, "John Crane Installation Instructions for Type 8B-1 Seal, Revision 0."

Attachment 3 W3F1-2008-0069 Affidavit to W3F1-2008-0069 Paqe 1 of 3 GE Hitachi Nuclear Energy Americas LLC AFFIDAVIT 1, Tim E. Abney, state as follows:

(1) 1 am Vice President. Services Licensing, Regulatory Affairs, GE-Hitachi Nuclear Energy Americas LLC ("GEH"), have been delegated the function of reviewing the information described in paragraph (2) which is sought to be withheld, and have been authorized to apply for its withholding.

(2) The infi)rnation sought to be withheld is contained in Enclosure I of GEH's letter, JB08-JXDYR-001, J. Betsill to G. Scott, entitled " GEH Proprietary Mark-ups of Draft Entergy Letter W3FI-2008-0018", dated February 22,22008. GEH proprietary information in Enclosure 1, which is entitled "GEH Proprietary Mark-ups of Draft Entergy Letter W3F I -

2008-0018", is identified, by a dotted underline inside, double square brackets. ((This

.s.en!.c an mp.cYI.fl. In each case, the superscript notation t relers to Paragraph (3) of this affidavit, which provides the basis for the proprietary detennination.

(3) In making this application for withholding of proprietary information of which it is the owner or licensee, GEH relies upon the exemption from disclosure set forth in the Freedom of Infbrmation Act ("FOIA"), 5 USC Sec. 552(b)(4), and the Trade Secrets Act, 18 USC Sec. 1905, and NRC regulations 10 CFR 9.17(a)(4), and 2.390(a)(4) for "trade secrets" (Exemption 4). The material for which exemption from disclosure is here sought also qualify under the narrower definition of "trade secret", within the meanings assigned to those terms for purposes of FOIA Exemption 4 in, respectively, Critical Mass Energy Proiect v. Nuclear Regulatory Commission, 975F2d87l (DC Cir. 1992), and Public Citizen Health Research Group v. FDA 704F2d 1280 (DC Cir. 1983).

(4) Some examples of categories of information which fit into the definition of proprietary information are:

a. Infornnation that discloses a process, method, or apparatus, including supporting data and analyses, where prevention of its use by GEH's competitors without license from GEH constitutes a competitive economic advantage over other companies;

b. Information which, if used 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 of a similar product; C. Inforimnation Which reveals aspects of past, present, or future GEH customer-funded development plans and programs, resulting in potential products to GEII;

d. Information which discloses patentable subject matter for which it may be desirable to obtain patent protection.

The information sought to be withheld is considered to be proprietary for the reasons set forth in paragraphs (4)a. and (4)b. above.

Aft JB09-JX DYR-O 1.do*c Affidvit Page I of 3 to W3F1 -2008-0069 Paae 2 of 3 (5) To address 10 CFR 2.390(b)(4), the information sought to be withheld is being submitted to NRC in conlidence. The information is of a sort customarily held in confidence by GEH, and is in fact so held. The information sought to be withheld has, to the best of my knowledge and beliet, consistently been held in confidence by GEH, no public disclosure has been made, and it is not available in public sources. All disclosures to third parties, including any required transmittals to NRC, have been made, or must be made, pursuant to regulatory provisions or proprietary agreements which provide, for maintenance of the information in confidence. Its initial designation as proprietary information, and the subsequent steps taken to prevent its unauthorized disclosure, are as set forth in paragraphs (6) and (7) following.

(6) Initial approval of proprietary treatment of a document is made by the manager of the originating component, 'the person most likely to be acquainted with the value and sensitivity of the information in relation to industry knowledge, or subject to the terms under which it was licensed to GEH. Access to such documents within GEH is limited on a

ý"need to know" basis.

(7) The procedure for approval of external release of such a document typically requires review by the staff manager, project manager, principal scientist, or other equivalent authority for technical content, competitive effect, and determination of the accuracy of the proprietary designation. Disclosures outside GEH arc limited to regulatory bodies, customers, and potential customers, and their agents, suppliers, and licensees, and others with a legitimate need for the information, and then only in accordance with appropriate regulatory provisions or proprietary agreements.

(8) The intbrmation identified in paragraph (2), above, is classified as proprietary because it contains detailed results of analytical model and method, as well as testing methods, applied to perform evaluations of emergency core cooling system and containment sprays strainers in Boiling Water Reactors ("BWR") and Pressurized Water Reactors. The developmnent and approval of these models and methods was achieved at a significant cost to GEH, on the order of several million dollars.

The development of the evaluation process along with the interpretation and application of the analytical results is derived from the extensive experience database that constitutes a major GEH asset.

(9) Public disclosure of the information sought to be withheld is likely to cause substantial harm to GEH's competitive position and foreclose or reduce the availability of profit-making opportunities. The intbrmation is part of GEH's comprehensive safety and technology base, and its commercial value extends beyond the original development cost.

The value of the technology base goes beyond the extensive physical database and analytical methodology and includes development of the expertise to determine and apply the appropriate, evaluation process. In addition, the technology base includes the, value derived from providing analyses done with NRC-approved methods.

Aft JB08-.IXDYR-1tdoc Affidavit Page 2 of 3 to W3F1 -2008-0069 Page 3 of 3 The research, development, engineering, analytical and NRC review costs comprise a substantial investment of time and money by G EH.

The precise value of the expertise to devise an evaluation process and apply the correct analytical methodology is difficult to quantify, but it clearly is substantial.

GEH's competitive advantage will be lost if its competitors are able to use the results of the GEH experience to normalize or verify their own process or if they are able to claim-an equivalent understanding by demonstrating that they can arrive at the same or similar conclusions.

The value of this information to GEH would be lost if the infornation were disclosed to the public. Making such infonnation available to competitors without their having been required to undertake a similar expenditure of resources would unfairly provide competitors with a windfall, and deprive GEH of the opportunity to exercise its competitive advantage to seek an adequate return on its large investment in developing and obtaining these very valuable analytical tools.

I declare under penalty of perjury that the foregoing affidavit and the matters stated therein are true and correct to the best of my knowledge, information, and belief.

Executed on this 22nd day of February 2008.

N_"i ci /~ ~u Tim E. Abney GE-Hitachi Nuclear Energy Americas LLC Aff J1308-JXDYR-01 .doc Affidavit Page 3 of 3

Text

Entergy Nuclear South Entergy Operations, Inc.

17265 River Road Killona, LA 70057-3093 Tel 504-739-6496 EnLI te gy Fax 504-739-6698 tgaudet@entergy.com Timothy J. Gaudet Acting Nuclear Safety Assurance Director Waterford 3 Attachment I Contains 10CFR2.390 Proprietary Information W3F17-2008-0069 October 23, 2008 U.S. Nuclear Regulatory Commission Attn:, Document Control Desk Washington, DC 20555-0001

Subject:

Final Supplemental Response to NRC Generic Letter 2004-02 "Potential Impact of Debris Blockage on Emergency Recirculation during Design Basis Accidents at Pressurized-Water Reactors" Waterford Steam Electric Station, Unit 3 (Waterford 3)

Docket No. 50-382 License No. NPF-38

References:

1. Generic Letter 2004-02, "Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors," dated September 13, 2004

2. Waterford 3 letter, "Request for Extension of: Completion date for Corrective Actions Required by Generic Letter 2004-02," dated November 14, 2007

3. NRC letter, "Approval of Extension Request for Corrective Actions; re:

GL 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation during Design Basis Accident at Pressurized Water Reactors," dated December 12, 2007

4. NRC letter, "Report on Results of Staff Audit of Corrective Actions to Address Generic Letter 2004-02," dated January 28, 2008

5. Waterford 3 letter, "Supplemental Response to NRC Generic Letter 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation during Design Basis Accidents at Pressurized-Water Reactors," dated February 29, 2008

6. Waterford 3 I6tter, "Request for Extension of Completion Date for Resolution of Generic Letter 2004-02," dated May 12, 2008

W3F1 -2008-0069 Page 2

Dear Sir or Madam:

The purpose of this letter is to provide notification of the completion of activities associated with Generic Letter (GL) 2004-02 (Reference 1). The U.S. Nuclear Regulatory Commission (NRC) issued Reference 1 to request that addressees perform an evaluation of the emergency core cooling system (ECCS) and containment spray system (CSS) recirculation functions in light of the information provided in the GL and, if appropriate, take additional actions to ensure system function. Additionally, the GL requested addressees to provide the NRC with a written response in accordance with 10 CFR 50.54(f). The request was based on identified potential susceptibility of the pressurized water reactor (PWR) recirculation sump screens to debris blockage during design basis accidents requiring recirculation operation of ECCS or CSS and on the potential for additional adverse effects due to debris blockage of flowpaths necessary for ECCS and CSS recirculation and containment drainage.

Reference 2 provided Waterford 3's (W-3) summary of actions taken which included the installation of a new sump strainer during the W-3 Fall 2006 refueling outage and the results of the downstream effects evaluation. Additionally, W-3 requested an extension until restart, following refueling outage 15, to complete the analysis needed to achieve compliance with GSI-1 91. The NRC extension was approved in Reference 3. In Reference 4, the NRC reported on staff evaluation of corrective actions to address GL. In Reference 5, W-3 provided the response to open items identified in Reference 4 and the supplemental response per NRC letter to NEI dated November 30, 2007. Additionally, W-3 committed to providing a final supplemental response in Reference 5. In Reference 6, W-3 requested an extension for providing the final supplemental response. This final supplemental response communicates the completion of all actions needed to address the GL. provides the details of the final supplemental response and includes proprietary information. The proprietary information was provided to Entergy in a GE Hitachi Nuclear Energy (GEH) transmittal that is referenced by an affidavit. GEH requests the enclosed proprietary information identified in Attachment 1 be withheld from public disclosure in accordance with the provisions of 10CFR 2.390 and 10CFR 9.17. Attachment 2 is the non-proprietary version of the response. Attachment 3 contains the affidavit for withholding the proprietary information contained in Attachment 1.

This letter contains no regulatory commitments.

Please contact me or Robert J. Murillo, Manager Licensing at (504) 739-6715 if there are any questions regarding this letter.

I declare under penalty of perjury that the foregoing is true and correct. Executed on October 23, 2008.

Siny,

W3F1-2008-0069 Page 3 Attachment(s): 1. Supplemental Response to NRC GL 2004-02 (Proprietary Information)

2. Supplemental Response to NRC GL 2003-02 (Non-Proprietary Information)

3. Affidavit

W3F1 -2008-0069 Page 4 (w/o Attachment 1)(w/Attachments 2 and 3) cc: Mr. Elmo E. Collins, Jr.

Regional Administrator U. S. Nuclear Regulatory Commission Region IV 612 E. Lamar Blvd., Suite 400 Arlington, TX 76011-4125 NRC Senior Resident Inspector Waterford Steam Electric Station Unit 3 P.O. Box 822 Killona, LA 70066-0751 U. S. Nuclear Regulatory Commission Attn: Mr. N. Kalyanam Mail Stop O-07D1 Washington, DC 20555-0001 Wise, Carter, Child & Caraway ATTN: J. Smith P.O. Box 651 Jackson, MS 39205 Winston & Strawn ATTN: N.S. Reynolds 1700 K Street, NW Washington, DC 20006-3817 Morgan, Lewis & Bockius LLP ATTN: T.C. Poindexter 1111 Pennsylvania Avenue, NW Washington, DC 20004

Attachment 2 W3F1-2008-0069 Supplemental Response to NRC GL 2004-02 (Non-Proprietary Information) to W3F1-2008-0069 Page 1 of 87 Table of Contents 1.0 Overall Com pliance .................................................................................................. 3 2.0 General Description of and Schedule for Corrective Actions ................................ 5 3.0 Specific Information Regarding Methodology for Determining Compliance ...... 7 3.a Break Selection ......................................................... . ..

.............................. 7 3.b Debris Generation/Zone of Influence (ZOI) (excluding coatings) ....................... 9 3.c Debris Characteristics .......................................................................................... 11 3.d Latent Debris ............................................................................................................. 13 3.e Debris Transport ................................................................................................... 13 3.f Head Loss and Vortexing ................................................................................... 13 3.g Net Positive Suction Head (NPSH) ............................ ......................................... 13 3.h Coatings Evaluation ............................................................................................ 13 3.i Debris Source Term .............................................................................................. 13 3.j Screen Modification Package .............................................................................. 13 3.k Sum p Structural Analysis ................................................................................... 13 3.1 Upstream Effects ................................................................................................. 13 3.m Downstream Effects - Components and Systems ............................................. 13 3.n Dow nstream Effects - Fuel and Vessel .............................................................. 13 3.0 Chem ical effects ........................................................................................................ 13 3.p Licensing Basis ......................................................................................................... 13 O pen Item s ................................................................................................................................ 13 References ................................................................................................................................. 13 to W3F1-2008-0069 Page 2 of 87 Acronyms AJIT Air Jet Impact Test OPG Ontario Power Generation BWR Boiling Water Reactor PWR Pressurized Water Reactor BWROG Boiling Water Reactor Owners PZR Presurizer Group RAS Recirculation Actuation Signal CFD Computational Fluid Dynamics RC Reactor Coolant CFR Code Federal Regulation RCP Reactor Coolant Pump CS Containment Spray RMI Reflective Metal Insulation CSS Containment Spray System RPV Reactor Pressure Vessel DBA Design Basis Accident RCB Reactor Containment Building DIR Design Input Record RWSP Refueling Water Storage Pool ECCS Emergency Core Cooling System SB LOCA Small Break Loss of Coolant GE General Electric Accident GL Generic Letter SDC Shut-down Cooling GR Guidance Report SE Safety Evaluation HELB High Energy Line Break SER Safety Evaluation Report HPSI High Pressure Safety Injection SG Steam Generator ID Internal Diameter SI Safety Injection IOZ Inorganic Zinc SIS Safety Injection System LB LOCA Large Break Loss of Coolant SIT Safety Injection Tank Accident SS Stainless Steel LOCA Loss of Coolant Accident SSC System, Structure, or Component LPSI Low Pressure Safety Injection TSP Tri-Sodium Phosphate MEl Metal Encapsulated Insulation UFSAR Updated Final Safety Analysis NEI Nuclear Energy Institute Report NRC Nuclear Regulatory Commission WF3 Waterford 3 NPSH Net Positive Suction Head ZOI Zone of Influence

'f-to W3F1-2008-0069 Page 3 of 87 This attachment provides Waterford 3's supplemental response to GL 2004-02 (Reference 1).

The supplemental response follows the format and guidance provided by the NRC in Reference

63. All text from Reference 63 is presented in italic script.

NRC Request, Summary-Level Description The GL supplemental response should begin with a summary-level description of the approach chosen. This summary should identify key aspects of design modifications, process changes, and supporting analyses that the 7icensee believes are relevant or important to the NRC staff's verification that corrective actions to address the GL are adequate. The summary should address significant conservatisms and margins that are used to provide high confidence the issue has been addressedeven with uncertaintiesremaining. Licensees should address commitments and/ordescriptionsof plant programs that support conclusions.

Summary-Level Description for Waterford 3 The key aspects of the approach chosen by Waterford 3 to resolve the concerns identified in GL 2004-02 are as follows:

• Replacement of existing screens for the Safety Injection sump. This increased total surface area of screens from approximately 200 ft 2 to approximately 3699 ft 2.

• Extensive testing and analysis to determine break locations, identify and quantify debris sources, quantify debris transport, determine downstream effects, determine chemical effects, and confirm recirculation function.

• Changes to plant programs, processes, and procedures to limit the introduction of materials into containment that could adversely impact the recirculation function, and establish monitoring programs to ensure containment conditions' will ,continue to support the recirculation function.

• Application of conservative measures to assure adequate margins throughout the actions taken to address the GL 2004-02 concerns.

Additional details regarding these key aspects are provided below:

Plant Modifications Waterford 3 has performed plant modifications as described below to address the concerns identified in GL 2004-02.

• The original box-like SI Sump screen that surrounded the sump itself was removed and replaced with General Electric Energy, Nuclear (GE) Modularized Stacked Disk Strainers.

The total surface area of the new SI Sump strainers is approximately 3699 ft 2 as compared to the original screen with a surface area of approximately 200 ft 2.

Nineteen TSP baskets were relocated to allow easier installation of the new plenum and strainers, or to eliminate interferences with the baskets.

• The screen partition thatseparated the two trains of the SI inlets was replaced with stainless steel grating.

• The low level switch inside the SI Sump was relocated to mount the housing on top of the new screen plenum.

to W3F1-2008-0069 Page 4 of 87

• The tubing for two level transmitters inside the SI Sump was rerouted to penetrate through the plenum in a designed location in order to prevent debris from passing through the penetration opening.

Testing Waterford 3 has completed extensive plant-specific testing of the SI Sump strainer design and the effects of chemical reactions on post-accident strainer head loss.

Conservatisms

• The ZOI for debris generation for all RCS cold leg breaks were based on the 42" diameter of the hot leg piping instead of the 30" diameter of the cold leg piping.

" WCAP-16710-P (Reference 38) showed that the ZOI for stainless steel jacketed NUKON insulation could be reduced to as low as 5D. However, Waterford 3 used a ZOI of 7D.

" Waterford 3 modeled 33% of all fibrous insulation within the ZOI as becoming fines or small debris. Appendix II of the SER for NEI 04-07 (Reference 3) suggested a value of 22% for small and fines debris.

, In the debris transport calculation 2005-05500 (Reference 29), the lift-over-curb velocity used was for a 6" high curb. However, the plenum on which the SI Sump strainers sit is 8" high.

• In all but one of the CFD scenarios in the debris transport calculation 2005-05500 (Reference 29), the approach velocity to the SI Sump strainers did not exceed the lift-over-curb velocity for a 6" curb at any location along the plenum perimeter. In one CFD scenario, less than 25% of the perimeter had velocities in excess of those necessary to lift over a 6" curb. For conservatism, 25% of the small debris is treated as lifting onto the sump strainers.

• All labels and tags are modeled with 100% transport to the screens. The total sacrificial area is calculated as equivalent to 100% of the original single sided surface area with 0%

overlap. As documented in the SER for NEI 04-07 (Reference 3), the sacrificial area could have been reduced to 75% of the total of the original single-sided surface area of the labels and tags.

• For the NPSH calculation ECM07-001 (Reference 51), the flow rates are based on both trains of ECCS and CS operating at pump run-out values. However, the temperature assumptions in containment are based on only one train of CS in operation at design flow, and ECCS at minimum flow.

• Waterford 3 assumed 50% of qualified coatings on the containment liner dome and the liner between elevations 112 and 138 would fail. This value far exceeds the total amount of coating failures throughout containment recorded at Waterford 3 in the most recent refueling outage.

1.0 Overall Compliance NRC Issue 1:

Provide information requested in GL 2004-02, "Requested Information." Item 2(a) regarding compliance with regulations.

GL 2004-02 Requested Information Item 2(a)

Confirmation that the ECCS and CSS recirculation functions under debris loading conditions are or will be in compliance with the regulatory requirements listed in the to W3F1-2008-0069 Page 5 of 87 Applicable Regulatory Requirements section of this GL. 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.

WF3 Response 1:

The recirculation functions for the SIS and the CSS for Waterford 3 are in compliance with the applicable Regulatory Requirements section of the subject generic letter under debris loading conditions upon the completion of Refuel 15 (RF15), the spring 2008 outage for Waterford 3.

The design packages EC-999 and EC-1002 update the Waterford 3 design basis associated with GS1-191 resolution and Generic Letter 2004-02 compliance. These EC's were approved by Entergy on October 23, 2008, along with associated Safety Analysis Report (SAR) changes provided to licensing. This submittal provides the final supplemental response for compliance with the regulatory requirements in GL 2004-02.

2.0 General Description of and Schedule for Corrective Actions NRC Issue 2:

Provide a general description of actions taken or planned, and dates for each. For actions planned beyond December 31, 2007, reference approved extension requests or explain how regulatory requirements will be met as per "Requested Information" Item 2(b). (Note: All requests for extension should be submitted to the NRC as soon as the need becomes clear, preferably not later than October 1, 2007.)

GL 2004-02 Requested Information Item 2(b)

A general description of and implementation schedule for all corrective actions, including any plant modifications, that you identified while responding to this GL. Efforts to implement the identified actions should be initiatedno later than the first refueling outage starting after April 1, 2006. All actions should be completed by December 31, 2007.

Providejustification 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.

WF3 Response 2:

The corrective actions to address the concerns identified in GL 2004-02 at Waterford 3 consist of plant modifications, testing and analysis, changes to plant programs and processes, and changesto the licensing basis. These changes are described below.

Plant Modifications Based on the results of debris generation and transport analyses, modifications to the existing SI Sump screens were required to meet the applicable Regulatory Requirements discussed in GL 2004-02 (Reference 1). The physical changes were performed during the Waterford 3 refueling outage RF14, in the fall of 2006. The changes are listed below:

to W3F1-2008-0069 Page 6 of 87

1. The original box-like SI Sump screen that surrounded the sump itself was removed and replaced with General Electric Energy, Nuclear (GE) Modularized Stacked Disk Strainers. Due to the amount of screen area required to be installed to meet the requirements of GL 2004-02 (Reference 1), a strainer plenum with strainer modules on top was constructed over the SI Sump and over the concrete floor. 2The surface2 area of the new SI Sump strainers was increased from approximately 200 ft. to 3699 ft.

2. The sump partition that separates the two trains of the SI inlets was replaced with stainless steel grating.

3. The housing for the low level switch inside the SI Sump was relocated to mount on top of the new screen plenum.

4. The tubing for two level transmitters inside the SI Sump was rerouted to penetrate through the plenum in a designed location in order to prevent debris from passing through the penetration opening.

5. Nineteen TSP baskets were relocated to allow easier installation of the new plenum and strainers, or to eliminate interferences with the baskets.

Testing and Analyses As approved by NRC approvals of extension requests for Waterford 3 (References 44 and 59),

all testing was completed by the end of February 2008, and analyses were completed by October 23, 2008. The following documents were generated / revised to support the GL 2004-02 actions.

• Debris Generation Calculation 2004-07780 (Reference 28)

• Debris Transport Calculation 2005-05500 (including CFD modeling) (Reference 29)

• Downstream Effects Calculations 2005-02820 and 2005-12840 (References 31 and 32),

• Hydraulic Sizing Report GENE-0000-0053-4416 (Reference 36)

. Water Level Inside Containment Calculation MNQ6-4 (Reference 37)

. ECCS and CS Pump NPSH Analysis ECM07-001 (Reference 51)

• In Vessel Downstream Effects Calculation CN-SEE-I-08-42 (Reference 65)

Plant programs and Processes The program and process changes needed to address the GL 2004-02 concerns were completed by December 31, 2007.

Licensing Basis The licensing basis changes needed to address the GL 2004-02 concerns consist of UFSAR changes related to the plant modifications previously described.

to W3F1-2008-)069 Page 7 of 87 3.0 Specific Information Regarding Methodology for Determining Compliance Backqround Information To facilitate understanding of the methodology for demonstrating compliance with the applicable regulations, Waterford 3 has provided background information regarding the design of the Waterford 3 containment, and the manner in which the GL 2004-02 concerns will be addressed.

Waterford 3 Containment Design Waterford 3 is a Combustion Engineering PWR with a large volume dry containment. Each of the two loops contains two Reactor Coolant Pumps (RCP), one Steam Generator (SG), and associated piping, located within a concrete wall enclosure commonly referred to as a D-ring.

The two RCS piping loops are nearly identical with the exception that one loop includes the PZR and associated piping. The area inside each D-ring is open directly above it. The two D-rings are also open on the bottom to a common open area on the basemat of the plant. The refueling cavity and other concrete walls separate the two loops. The PZR is contained within a separate concrete enclosure.

All postulated pipe break LOCAs for which sump recirculation would be required would take place within the D-rings, in the reactor cavity, or inside the pressurizer cubicle.

3.a Break Selection NRC Issue 3.a:

The objective of the break selection process is to identify the break size and location that present the greatestchallenge to post-accident sump performance.

1. Describe and provide the basis for the break selection criteria used in the evaluation.

2. State whether secondary line breaks were considered in !the evaluation (e.g., main steam and feedwater lines) and briefly explain why or why not.

3. Discuss the basis for reaching the conclusion that the break size(s) and locations chosen present the greatestchallenge to post-accident sump performance.

WF3 Response 3.a.1:

Baseline Break Selection:

A number of breaks were selected in the debris generation calculation 2004-07780 (Reference

28) in order to provide conservative input for the transport calculations. The breaks that were selected are:

1. Break S1 - Hot Leg Piping at SG 1 Nozzle

2. Break S2 - SG 1 Hot Leg Piping at RPV Connection

3. Break S3 - Hot Leg Piping at SG 2 Nozzle

4. Break S4 - Tee between SG 1 Hot Leg and PZR Surge Line (Alternate Break)

5. Break S5 - Cold Leg Piping at RCP 1A Inlet Nozzle

6. Break S6 - Cold Leg Piping at RCP 2B Inlet Nozzle to W3F1-2008-0069 Page 8 of 87

7. Break S7 - PZR Surge Line at PZR Nozzle

8. Break S8 - Cold Leg Piping at RCP 1B Inlet Nozzle

9. Break S9 - Cold Leg Piping at RCP 2A Inlet Nozzle Breaks S1, S2 and S3 are located on the hot leg of the primary piping, which has the largest diameter of the primary piping with a 42-inch diameter, obviously producing the largest ZOI.

Breaks S1 and S3 are placed at the steam generator nozzles in order to capture the most debris. Break S2 is located at the RPV.

Breaks S5 and S6 are located at the RCP inlet nozzles on the cold leg piping, which has the next largest diameter of the primary piping with a 30-inch diameter. For conservatism, the ZOI for the cold leg piping was based on the 42-inch diameter of the hot leg. These breaks are located closer to the-SI Sump, thus providing conservative input for the debris transport and head loss calculations.

Break S4 is located on the PZR surge line at the connection to the SG 1 Hot Leg. In accordance with the alternate break methodology, it is considered to have a 12-inch inner diameter, although the actual ID is 10.126 inches. The debris load from the alternate break was determined; however, the deterministic methodology described in Sections 3, 4 and 5 of the NEI Guidance 04-07 (Reference 2) was used to determine bounding debris generation and transport volumes.

Break S7 is located where the PZR surge line connects'to the PZR inlet nozzle on the bottom of the PZR. This break is located outside the D-ring walls, and nearer the SI Sump location than any of the other breaks. This break does not produce a significant amount of debris; however, it was selected due to the proximity and clear debris path to the SI Sump.

Break 38 is located at the RCP 1B inlet nozzle on the cold leg piping. This break will have the same debris generation as for the S5 break. This break was only analyzed for flow distribution in containment. This was to determine the worst case flow for debris transport. The S8 break did not result in higher debris transport than the other breaks already analyzed.

Break S9 is located at the RCP 2A inlet nozzle on the cold leg piping. This break will have the same debris generation as the S6 break. This break was only analyzed for the flow distribution in containment. This was to determine the worst case flow for debris transport. The S9 break did not result in higher debris transport than the other breaks already analyzed.

SBLOCA events were not included in this evaluation, as the debris load would not be bounding due to the smaller areas covered by the break zones. The majority of the fiber that would be created is inside the D-Rings, and the larger breaks in those areas will create larger quantities of fiber than could be created from the SBLOCA events.

WF3 Response 3.a.2:

Secondary pipe breaks were not considered for this analysis. Based upon a review of the plant UFSAR discussed in the Debris, Generation calculation 2004-07780 (Reference 28),

.containment spray and recirculation are not required for a Main Steam Line Break or a to W3F1-2008-0069 Page 9 of 87 Feedwater Line Break. Additionally, breaks of small lines were not investigated, because.the debris load would not be bounding due to the smaller areas covered by the break zone.

WF3 Response 3.a.3:

The locations of the analyzed breaks are chosen in order to maximize the amount and types of debris generated. To this end, breaks are placed near large equipment, specifically the SGs, RCPs, and PZR. The breaks were also placed near walls and the floor since concrete surfaces have very thick coatings compared to steel surfaces. Finally, breaks were located in areas expected to maximize the transport of debris to the sump strainer.

3.b Debris Generation/Zone of influence (ZOI) (excluding coatings)

NRC Issue 3.b:

The objective of the debris generation/ZOI process is to determine, for each postulated break location: (1) the zone within which the breakjet forces would be sufficient to damage materials and create debris; and (2) the amount of debris generated by the break jet forces.

1. Describe the methodology used to determine the ZOls for generating debris. Identify which debris analyses used approved methodology default values. Fordebris with ZOls not defined in the guidance report/SE, or if using other than default values, discuss method(s) used to determine ZOI and the basis for each.

2. Provide destruction ZOls and the basis for the ZOls for each applicable debris constituent.

3. Identify if destruction testing was conducted to determine ZOls. If such testing has not been previously submitted to the NRC for review or information, describe the test procedure and results with reference to the test report(s).

4. Provide the quantity of each debris type generated for each break location evaluated. If more than four break locations were evaluated, provide data only for the four most limiting locations.

5. Provide total surface area of all signs, placards, tags, tape, and similar miscellaneous materialsin containment.

WF3 Response 3.b.1:

In order to perform the calculation of debris generation within the ZOI, a representative model of the insulation location and volume is utilized in the debris generation calculation 2004-07780 (Reference 28). The model is a Microsoft Excel spreadsheet created from piping isometric drawings and insulation drawings. These drawings were used to develop a 3-dimentional computer model, which was then converted to the Excel spreadsheet model. The spreadsheet determines the amount of insulation within a ZOI centered at coordinates that are input by the user. In this way multiple break locations are able to be evaluated relatively quickly and the user can ensure that conservative and limiting breaks are chosen.

The insulation in containment at Waterford 3 consists of Nukon (canvas encapsulated (unjacketed) or stainless st'eel jacketed), MEI (stainless steel jacketed), Min-K, Microtherm, and Transco RMI. The amount of insulation debris generated is dependent on the proximity of each insulated target to the postulated break.

to W3F1-2008-0069 Page 10 of 87 The SER for NEI 04-07 (Reference 3) recommends a ZOI radius of 17.0D ("D" being the'inside diameter of the pipe break) for both jacketed and unjacketed Nukon Fiber. The SER recommended ZOI radius of 17D is used for unjacketed Nukon Fiber debris generation analysis.

Based on testing contained in Westinghouse report WCAP-1 671 0-P (Reference 38), a reduced ZOI of 7.0D is used for the SS Jacketed Nukon debris generation analysis. NEI Guidance 04-07 and the SER for NEI 04-07 (References 2 and 3) do not address MEI insulation. Based on industry testing contained in Alion report ALION-REP-ENTG-4771-02 (Reference 39), a ZOI of 4.0D is used for the SS Jacketed MEl debris generation analysis. The SER recommended ZOI radius of 28.6D is used for Min-K insulation. The NEI and SER documents do not address the ZOI for Microtherm insulation. Therefore, a ZOI radius of 28:6D is used for Microtherm based on its similarity to Min-K insulation and because the ZOI radius for Min-K is the largest of all the tested materials described in the NEI and SER documents. The SER recommended ZOI radius of 2D is used for RMI debris generation analysis.

As discussed in Section 3d of this response submittal, latent debris and miscellaneous (foreign) materials are also included in the debris generation analysis. The amounts of these types of debris are determined from plant walkdown reports and are presented in their respective section of this response.

WF3 Response 3.b.2:

Debris Type ZOl Basis Jacketed Nukon Fiber Blankets 7 WCAP-1 6710-P (Reference 38)

Unjacketed Nukon Fiber Blankets 17 SER for NEI 04-07 (Reference 3)

ALION-REP-ENTG-4771-02 (Reference 39)

SER for NEI 04-07 M 2(Reference 3)

Microtherm 28.6 2004-07780 (Reference 28)

Transco RMI 2(Rfrne3 SER for NEI 04-07 (Reference 3)

WF3 Response 3.b.3:

Westinghouse report WCAP-16710-P (Reference 38) documents testing performed on jacketed Nukon insulation blankets to determine the proper ZOI. From Section 4 of WCAP-16710-P (Reference 38), "The approach taken to develop this experimental program was to subject the encapsulated... stainless steel jacketed NUKON fiberglass insulation materials to phenomena and processes that accurately simulate those experienced during a postulated LOCA blowdown for a PWR. The conditions of interest are exposure to elevated temperature, pressure and high mass flux. Data collected from and observations from the tests were used as follows:

For the jacketed NUKON fiberglass insulation system, determine an appropriate, technically defensible, realistic material-specific ZOI at which the fiberglass insulation will not experience damage that would require it to be treated as debris."

to W3F1-2008-0069 Page 11 of 87 The objective of the test was to determine the generation of debris of the insulation material that should be considered in post-accident sump performance. The testing consisted of subjecting the jacketed NUKON insulation to a two phase jet originating from a subcooled, high pressure, high temperature reservoir. For the purposes of this testing, debris generation was defined as the "observable release or extrusion of fiberglass insulation material from the fiberglass from the woven fiberglass cloth covered blanket". The results of this testing showed that the ZOI for the SS jacketed NUKON insulation could be reduced to as low as 5D. However, for conservatism, a ZOI of 7D was used.

WF3 Response 3.b.4:

The insulation and coating debris totals for the four most limiting breaks evaluated are presented in the table below.

Table 3b-1: Summary of LOCA Generated Debris Inside the ZOI Debris Type Units $1/$5 S2 $3/$6 S4 Jacketed Nukon Fiber Blankets [ft3] 81.4 177 25.4 0 Unjacketed Nukon Fiber Blankets [ft3] 351.1 113 50W18 331 MEI [ft7] 405.0 3 558.4 40.8 Min-K [ft3] 0.4 0.4 0.4 0.4 Microtherm [ft3] 4.2 4.2 4.2 4.2 Transco RMI 7ft2] 0 8750 0 0 WF3 Response 3.b.5:

As discussed in Section 3d of this response submittal, latent debris and miscellaneous (foreign) materials are also included in the debris generation analysis. An analysis in the Debris Generation calculation 2004-07780 (Reference 28) shows that the debris loading to be used is 250 Ibm.

Labels, tags, stickers, placards and other miscellaneous or foreign materials were evaluated via walkdown. The walkdown results are included in Debris Generation calculation 2004-07780 (Reference 28). Based on this calculation, a sacrificial area of 151 ft 2 of the strainer surface is used for stickers, index cards, placards, tape, glass and other miscellaneous or foreign materials. This total includes only those materials which are not Design Basis Accident (DBA) qualified and does not include any overlap of the materials on the screens.

3.c Debris Characteristics NRC Issue 3.c:

The objective of the debris characteristicsdeterminationprocess is to establish a conservative debris characteristics profile for use in determining the transportability of debris and its contribution to head loss.

1. Provide the assumed size distribution for each type of debris.

2. Provide bulk densities (i.e., including voids between the fibers/particles) and material densities (i.e., the density of the microscopic fibers/particles themselves) for fibrous and particulatedebris.

to W3F1-2008-0069 Page 12 of 87

3. Provide assumed specific surface areasfor fibrous and particulate debris.

4. Provide the technical basis for any debris characterizationassumptions that deviate from NRC-approved guidance.

WF3 Response 3.c:

The debris sources at Waterford 3 include insulation, coatings, foreign material and latent debris. The characteristics of the insulation debris material are discussed in this section. The characteristics of the other debris types (e.g. coatings, foreign and latent) are included in their respective sections of this response submittal (Sections 3h and 3d).

WF3 Response 3.c.1:

Unjacketed Nukon Insulation Size Distribution:

For unjacketed Nukon insulation with a ZOI of 17D, a size distribution of 8% fines, 25% small pieces, 32% large pieces and 35% intact pieces is used. This distribution is determined based on an analysis of the results of the BWR Owners' Group (BWROG) air-jet impact tests (AJIT) and the Ontario Power Generation (OPG) debris generation tests as described in NUREG/CR-6808 (Reference 18). Based on the results of the AJIT and OPG debris generation tests as presented in Appendix VI of the SER for NEI 04-07 (Reference 3) and Volume 3 of NUREG/CR-6762 (Reference 15), 33% of all fibrous insulation within the ZOI is modeled as becoming fines or small debris. This fraction of fines or small debris is conservatively increased from the value (22%) suggested in Appendix II of the SER based on the OPG debris generation test. Implicit in these values is the assumption that the insulation is uniformly distributed within the ZOI. Due to the fact that the unjacketed Nukon insulation for the applicable breaks is distributed in several locations within each ZOI, uniformity is considered a reasonable approximation. Thus, 67% of all fibrous insulation within the ZOI is modeled as becoming either large debris or remaining intact. To determine the appropriate split between fine and small debris, the results of the AJIT are used. The AJIT indicated that, when insulation was completely destroyed, a maximum of 25% of the insulation was too fine to collect by hand. Thus 25% of the 33% small debris fraction is modeled as becoming fines; i.e. 8% [0.25*0.33] of the fibrous insulation within the ZOI becomes fine debris when destroyed. This implies that 25% [(1-0.25)*0.33] of the fibrous insulation within the ZOI becomes small debris when destroyed. To determine the appropriate split between large and intact debris, the results of the AJIT are also used. Per the SER guidance provided in Appendix VI of the SER for NEI 04-07 (Reference 3), 35% of the fibrous insulation within the ZOI is modeled as intact debris, leaving 32% as large piece debris. Fines that enter the active recirculation pool are considered 100% transportable. Small, large and intact pieces are transported based on velocity data found in various, references. Specifics of debris transport are discussed in Section 3e.

SS Jacketed Nukon Insulation Size Distribution:

For stainless steel jacketed Nukon insulation with a ZOI of 7D a size distribution of 25% fines and 75% small pieces is used. For a ZOI of 7D, the suggested Nukon size distribution contained in Table 3-3 of the SER for NEI 04-07 (Reference 3) is not applicable. Instead the size distribution is determined from Figure I1-1 of the SER, which relates jet pressure toZOI radii, and Figure 11-2 of the SER, which relates jet pressure to the fraction of small debris generated. The data presented in Figure 11-2 comes from the Air Jet Impact tests, which are discussed in many documents related to GSI-191 including NUREG/CR-6808 (Reference 18).

to W3F1-2008-0069 Page 13 of 87 SS Jacketed MEI Insulation Size Distribution:

For stainless steel jacketed MEI insulation with a ZOI of 4D a size distribution of 20% fines and 80% small pieces is used per ALION-REP-ENTG-4771-02 (Reference 39).

Min-K Insulation Size Distribution:

For Min-K insulation a conservative size distribution of 100% fines is used as documented in Table 3-3 of the SER for NEI 04-07 (Reference 3). Fines are the constituent part of the insulation and are considered 100% transportable.

Microtherm Insulation Size Distribution:

For Microtherm insulation a conservative size distribution of 100% fines is used as documented in Table 3-3 of the SER for NEI 04-07 (Reference 3). Fines are the constituent part of the insulation and are considered 100% transportable.

Transco RMI Insulation Size Distribution:

For Transco RMI insulation with a ZOI of 2D, a size distribution of 75% small fines and 25%

large pieces is used. This size distribution is confirmed by Table 3-3 of the SER for NEI 04-07 (Reference 3).

WF3 Response 3.c.2:

Nukon Insulation Density:

Per Table 3-2 of NEI Guidance 04-07 (Reference 2) the bulk density of Nukon insulation is 2.4 Ibm/ft 3. The bulk density of the Nukon insulation installed at Waterford 3 and used in the sump strainer performance testing is also 2.4 Ibm/ft 3 .

MEI Insulation Density:

The metal encapsulated insulation (MEI) is Owens-Corning TIW Type I1. Based on manufacturer data for this insulation the bulk density is 2.4 Ibm/ft3 . The bulk density of the MEI insulation installed at Waterford 3 and used in the sump strainer performance testing is also 2.4 Ibm/ft3.

Min-K Insulation Density:

Per Table 3-2 of NEI Guidance 04-07 (Reference 2) the bulk density of Min-K insulation is 8 to 16 Ibm/ft 3. The bulk density of the Min-K insulation installed at Waterford 3 is 13 Ibm/ft 3 per calculation ECM89-083 (Reference '40). This compares to a bulk density of 14.5 Ibm/ft 3 for the insulation used in the sump strainer performance testing.

Microtherm Insulation Density:

Per Table 3-2 of NEI Guidance 04-07 (Reference 2) the bulk density of Microtherm insulation is 5 to 12 Ibm/ft 3. This compares to a bulk density of 14.5 Ibm/ft 3 of the insulation used in the sump strainer performance testing.

Reflective Metal Insulation Density:

Transco and Mirror RMI are comprised of thin layers of stainless steel foil. Stainless steel has a density of 490 Ibm/ft 3.

to W3F1-2008-0069 Page 14 of 87 WF3 Response 3.c.3:

The Material density and specific surface area (SJ)were only used for preliminary analytically determined head loss values across a debris laden sump screen using the correlation given in NUREG/CR-6224 (Reference 8). Since the head loss across the installed sump screen is determined via testing, these values are not used in the design basis for Waterford 3.

Therefore, these values are not provided as part of this response.

WF3 Response 3.c.4:

Waterford 3 debris, generation, transport, and head loss testing have used the debris characterization assumptions provided in the SER for NEI 04-07 (Reference 3).

3.d Latent Debris NRC Issue 3.d:

The objective of the latent debris evaluation process is to provide a reasonableapproximation of the amount and types of latent debris existing within the containment and its potential impact on sump screen head loss.

1. Provide the methodology used to estimate quantity and composition of latent debris.

2. Provide the basis for assumptions used in the evaluation.

3. Provide results of the latent debris evaluation, including amount of latent debris types and physical data for latent debris as requested for other debris under c. above.

4. Provide amount of sacrificialstrainersurface area allotted to miscellaneous latent debris.

WF3 Response 3.d.1:

Latent debris has been evaluated by containment walkdown as recommended by Section 3.5.2 of NEI Guidance 04-07 (Reference 2) and confirmed by the NRC SER for NEI 04-07 (Reference 3). The walkdown of the Waterford 3 containment was conducted in accordance with the guidance provided by NEI Guidance documents 04-07 (Reference 2) and 02-01 (Reference 20) and the SER for NEI 04-07 (Reference 3). As shown below, four (4) or more samples were collected for all surface types except grating. The additional samples collected for certain surface types increase the statistical accuracy of the evaluation. A listing of the number of each sample type follows.

Number of Samples Collected

• Containment Liner .............. 4 HVAC Duct (Vertical) ...... 4

• Equipment (Horizontal) ....... 4 Pipe (Horizontal) ................ 4

• Equipment (Vertical) ............ 4 Pipe (Vertical) ..... ...... 4

• Floor ..................... I.............. 4 Cable Tray (Horizontal) ........ 4

• W all .................................... 5 Cable Tray (Vertical) ........... 4

• HVAC Duct (Horizontal) ...... 4 Gratings ............................ 0 The weights of the samples collected are used to determine the latent debris mass distribution (g/ft2). Measurements taken are accurate to 0.01 grams. A statistical analysis of the samples is performed in the post-processing of the latent debris walkdown results, which is Attachment to W3F1 -2008-0069 Page 15 of 87 8.8 of the Waterford 3 Debris Generation calculation 2004-07780 (Reference 28). The analysis determined a 90% confidence limit of the mean value for each type of surface based on a normal distribution. The upper limit of the mean value for each surface type is then applied over the entire surface area of that type throughout containment. This analysis lends further confidence and conservatism to the latent debris mass determination.

Labels, tags, stickers, placards and other miscellaneous or foreign materials (including glass) were also evaluated via walkdown. The walkdown results are included as attachment 8.9 of the Waterford 3 Debris Generation calculation 2004-07780 (Reference 28) and are summarized in section 5.5 of the calculation.

WF3 Response 3.d.2:

No samples were available for grating; therefore, grating is conservatively assumed to have the same latent debris loading as the floor.

WF3 Response 3.d.3:

Consistent with the NRC SER for NEI 04-07 (Reference 3), 15% of the latent debris load (by mass) is assumed to be fibrous debris and the other 85% (by mass) is treated as particulate3 debris. Likewise, consistent with the SER for NEI 04-07 (Reference 3), a density of 2.7 g/cm for particulate debris is used. For, latent fibrous debris, a density of 2.4 Ibm/ft3 (bulk density of Nukon per NEI Guidance 04-07, Reference 2) is used in order to conservatively maximize the volume of latent fibrous debris. As the specific surface area of debris is only relevant for head-loss calculations per NUREG/CR-6224 (Reference 8) and head-loss evaluations are being conducted experimentally, the specific surface area of latent debris is not determined.

The results of the latent debris calculation conservatively determined the debris loading to be 250 Ibm.

Miscellaneous latent debris is also discussed in more detail in the following debris transport section 3.e.

Table 3d-1: Latent and Foreign Material Debris Latent and Foreign Material Debris Quantity Latent Debris (Ibm) 250 Fiber (Ibm) 37.5 Particulate (Ibm) 212.5 Foreign Material Debris (ft 2) 151 Stickers, index cards, placards and tape (ft2) .81.

Glass (light bulbs) (ft2 ) 70 WF3 Response 3.d.4:

A sacrificial area of 151 ft2 of the strainer surface is retained for stickers, index cards, placards, tape, glass and other miscellaneous or foreign materials. This total includes only those materials which are not Design Basis Accident (DBA) qualified.

to W3F1-2008-0069 Page 16 of 87 3.e Debris Transport NRC Issue 3.e The objective of the debris transportevaluationprocess is to estimate the fraction of debris that would be transportedfrom debris sources within containment to the sump suction strainers.

1, Describe the methodology used to analyze debris transport during the blowdown, washdown, pool-fill-up, and recirculationphases of an accident.

2, Provide the technical basis for assumptions and methods used in the analysis that deviate from the approved guidance.

3, Identify any computational fluid dynamics codes used to compute debris transport fractions during recirculation and summarize the methodology, modeling assumptions, and results.

4' Provide a summary of, and supporting basis for, any credit taken for debris interceptors.

5, State whether fine debris was assumed to settle and provide basis for any settling credited.

6. Provide the calculated debris transport fractions and the total quantities of each type of debris transportedto the strainers.

WF3 Response 3.e.1:

Debris transport determines the fraction of debris generated that is transported from debris sources (break location) to the sump screen. The debris transport analysis for Waterford 3 is conducted in accordance with both NEI Guidance 04-07 (Reference 2) and the NRC SER for NEI 04-07 (Reference 3). As such, each phase of post-LOCA transport is considered:

blowdown, washdown, pool fill-up and recirculation. A detailed discussion of each transport phase, including information on their effect on overall transport for Waterford 3 follows.

Blowdown and Washdown:

For mostly uncompartmentalized containments such as at Waterford 3, Section 3.6.3.2 of NEI Guidance 04-07 (Reference 2) states that all RMI debris (small and large) is conservatively postulated to fall to the containment floor; i.e. no RMI debris is ejected into the dome. Although NEI 04-07 does not specifically state that all fiber debris is assumed to fall to the containment floor, it is conservatively modeled as such (see Table 3-4 of the SER for NEI 04-07, Reference 3). Similarly, all Min-K, Microtherm, and coating debris is also conservatively modeled as falling to the containment floor. Thus, all LOCA generated debris is conservatively modeled as falling to the floor in the post-accident environment. This is reasonable as large debris should be modeled as falling to the containment floor per Section 3.6.3.2 of NEI Guidance 04-07 (Reference 2) and small debris that could reach the dome would eventually wash down to the active pool. Therefore, since all insulation debris eventually lands on the floor, a detailed blowdown and washdown analysis is not conducted. Rather all insulation debris generated is conservatively placed on the floor immediately and is further transported by pool fill-up and recirculation as discussed in the following sections. Conservatively, qualified coatings are also considered to fall directly to the floor. All other debris types, including unqualified coatings, latent and foreign material debris that are generated from outside the break ZOI are therefore considered to fall directly to the floor.

Pool Fill-up Conservatively, no inactive pools are credited. All debris on the floor prior to pool fill-up remains on the floor in the active pool after pool fill-up. During pool fill-up debris is transported away to W3F1-2008-0069 Page 17 of 87 from the break area and toward the perimeter of containment by the water spilling onto the floor.

Debris is then further transported by recirculation, as discussed in the following section.

Recirculation Debris that reaches the containment pool is subject to transport by the pool flow present during recirculation. In accordance with the NEI Guidance 04-07 and the SER for NEI 04-07 (References 2 and 3) all fine debris that lands in the pool is considered to transport entirely to the sump strainer. The transport of small, large and intact pieces of debris during recirculation is dependent on the velocities present in the containment pool.

Nukon debris transport is investigated and reported in NUREG/CR-6772 (Reference 16).

Transport velocities pertinent to Nukon debris transport at Waterford 3 are taken from this document. The document reports values at which some debris begins to move and at which a majority begins to move. These are referred to herein as the "incipient tumbling" and "bulk transport" velocities. Conservatively, the incipient tumbling velocity is used to determine transport potential. Accordingly, for breaks corresponding to Configuration A in NUREG/CR-6772 (Reference 16), non-fines Nukon pieces are considered to transport at velocities of 0.12 ft/s or greater and small and large Nukon pieces are considered to transport over a 6-inch curb at 0.34 ft/s. For breaks corresponding to Configuration B in NUREG/CR-6772 (Reference 16),

non-fines Nukon pieces are considered to transport at velocities of 0.07 ft/s or greater and small and large Nukon pieces are considered to transport over a 6-inch curb at 0.25 ft/s. For breaks corresponding to Configuration C in NUREG/CR-6772 (Reference 16), non-fines Nukon pieces are considered to transport at velocities of 0.06 ft/s or greater and small and large Nukon pieces are considered to transport over a 6-inch curb at 0.28 ft/s. Intact Nukon fiber blankets and Nukon jacketing are not considered to lift over the curb due to their size.

RMI debris transport is investigated in NUREG/CR-3616 (Reference 7) and NUREG/CR-6772 (Reference 16). Transport velocities pertinent to RMI debris transport at Waterford 3 are taken from these documents. Both documents report values at which some debris begins to move and at which a majority begins to move. These are referred to herein as the "incipient tumbling" and "bulk transport" velocities. Conservatively, the incipient tumbling velocity is used to determine transport potential. Accordingly, for breaks corresponding to Configuration A in NUREG/CR-6772 (Reference 16), non-fines RMI pieces are considered to transport at velocities of 0.28 ft/s or greater and are considered to transport over a 6-inch curb at 0.84 ft/s. For breaks corresponding to Configuration B in NUREG/CR-6772 (Reference 16), non-fines RMI pieces are considered to transport at velocities of 0.41 ft/s or greater and are considered to transport over an 8-inch curb at 0.30 ft/s. For breaks corresponding to Configuration C in NUREG/CR-6772 (Reference 16), non-fines RMI pieces are considered to transport at velocities of 0.20 ft/s or greater and are considered to transport over a 6-inch curb at 1.0 ft/s.

As noted in NUREG/CR-6773 (Reference 17) low density fiberglass debris (such as Nukon and MEI debris) is subject to erosion during Recirculation. The erosion rate used for the small and large Nukon debris pieces is 10% over the 30-day Recirculation mission time as described in ALION-REP-ENT-4536-02 (Reference 41). The erosion rate used for the MEI debris pieces is 90% over the 30-day Recirculation mission time, as described in Appendix II of the.SER for NEI 04-07 (Reference 3).

to W3F1-2008-0069 Page 18 of 87 WF3 Response 3.e.2:

There are no assumptions or methods that deviate from the approved guidance in the SER for NEI 04-07 (Reference 3) in the areas of debris transport, except for the erosion rate of the Nukon insulation. This is justified in an analysis / test documented in the report ALION-REP-ENT-4536-02 (Reference 41).

WF3 Response 3.e.3:

To assist in the determination of recirculation transport fractions, several Computational Fluid Dynamics (CFD) simulations were performed using FluentTM, a commercially available software package. Multiple break locations were investigated by the CFD simulations to determine which scenario would maximize debris transport. Four of the eight simulations conducted were based on the final strainer system design. The simulation results include a series of contour plots of velocity and turbulent kinetic energy (TKE), plots of flow pathlines originating at the break locations and animations of the flow velocities. These results have been combined with information in the GSI-191 literature and plant specific erosion test results to determine the overall transport fractions for small, large and intact pieces of fibrous debris and large pieces of RMI debris (fines are 100% transportable).

WF3 Response 3.e.4:

No debris interceptors were installed at Waterford 3 as part of the GL2004-02 resolution.

WF3 Response 3.e.5:

Credit is taken for the plenum that the strainers sit on. From the CFD results it is determined how much of the plenum perimeter is in areas with flow velocities in excess of the velocity required to lift the debris over a 6-inch obstacle. Since the plenum is approximately 8-inches tall, using the lift-over curb velocity for a 6-inch curb is conservative. The fraction of the curb perimeter in excess of the lift-over curb velocity is applied to the debris pile in the vicinity of the strainer to determine the debris load on the strainer. In all but one of the CFD scenarios the approach velocity does not exceed the lift over curb velocity at any location along the plenum perimeter. In the remaining scenario, less than 25% of the perimeter has velocities in excess of those necessary to lift over a 6-inch curb; however for conservatism, 25% of the small debris is treated as lifting onto the sump strainer.

No credit was taken for settling of fine debris.

WF3 Response 3.e.6:

The amount of debris determined to transport to the sump strainer for the limiting breaks are provided in the following tables.

to W3F17-2008-0069 Page 19 of 87 Insulation:

Table 3.e.6-1: Debris Generated and Transported to Strainer- Break S1 Debris Transport Debris at Ge n r a e Frac t Strainer D eb ris T ranspo rt by Type U nits Generated Fraction Strainer SS Jacketed Nukon [W] 81.4 0.325 26.46 Unjacketed Nukon [ft3] 351.1 0.137 48.1 SS MEI (Fiberglass) 405.0 0.92 372.6 Min-K [ft3] 0.4 1.00 0.4 Microtherm [ft 4.2 1.00 4.2 Transco RMI [ft 2] 0 0.75 0 Qualified Coatings [fta] 13.5 1.00 13.5 Unqualified Coatings [Tj] 31.02

• 21.70 Latent Debris [Ibm] 250 1.00 250 Foreign Materials: [ft 151 1.00 151 Table 3.e.6-2: Debris Generated and Transported to Strainer - Break S3 Debris ' Transport Debris at Debris Transport by Type Units Generated Fraction Strainer SS Jacketed Nukon [W] 25.4 0.325 8.26 Unjacketed Nukon [ft3] 501.8 0.137 68.75 SS MEI (Fiberglass) [ft 3] 558.4 0.92 513.73 Min-K 0.4 1.00 0.4 Microtherm 4.2 1.00 4.2 Transco RMI [ft 1 0 0.75 0 Qualified Coatings [ft3] 13.5 1.00 13.5 Unqualified Coatings [ft] 31.02

• 21.70 Latent Debris [Ibm] 250 1.00 250 Foreign Materials: [ft2] 151 1.00 151 Table 3.e.6-3: Debris Generated and Transported to Strainer - Break S4 Debris Transport Debris at Debris Transport by Type Units Genrae Fract Strainer Generated Fraction Strainer SS Jacketed Nukon [ft3] 0 0.325 0 Unjacketed Nukon [ft3] 331 0.137 45.35 SS MEi (Fiberglass) t3] 40.8 0.92 37.54 Min-K [ft3] 0.4 1.00 0.4 Microtherm [ft] 4.2 1.00 4.2 Transco RMI [ft2] 0 0.75 0 Qualified Coatings [Wt] 13.5 1.00 13.5 Unqualified Coatings [ft 3] 31.02

• 21.70 Latent Debris [Ibm] 250 1.00 250 Foreign Materials: [ft 151 1.00 151 to W3F1-2008-0069 Page 20 of 87 Table 3.e.6-4: Debris Generated and Transported to Strainer - Break S5 Debris .Transport Debris at Gen ra e Fra c t Strainer Debris Transport by Type Units SGenerated Fraction Strainer SS Jacketed Nukon [Wt] 81.4 0.325 26.46 Unjacketed Nukon [ft3] 351.1 0.137 48.1 SS MEI (Fiberglass) [ft 405.0 0.92 372.6 Min-K [ft 0.4 1.00 0.4 Microtherm Ift3] 4.2 1.00 4.2 Transco RMI I[ft] 0 0.75 .0 Qualified Coatings [ft3] 13.5 1.00 13.5 Unqualified Coatings 31.02

• 21.7 Latent Debris [Ibm] 250 1.00 250 Foreign Materials: [ft] 151 1.00 151 Table 3.e.6-5: Total Debris Generated and Transported to Strainer - Break S7 Debris Transport by Type Units Debris Transport Debris at Generated Fraction Strainer SS Jacketed Nukon [fti] 30 0.3925 11.78 Unjacketed Nukon [ft 49 0.1883 9.23 SS MEl (Fiberglass) [ft3] 0 0.93 0 Min-K [ft 0 1.00 0 Microtherm [ft1] 0.6 1.00 0.6 Transco RMI [ft 0 0.775 0 Qualified Coatings [ft3] 0 1.00 0 Unqualified Coatings [ft3] 31.03

• 21.71 Latent Debris [Ibm] 250 1.00 250 Foreign Materials: [fti] 151 1.00 151 Inorganic zinc, coatings within the ZOI, and indeterminate coatings are considered to transport 100% to the sump. The inventory of degraded qualified coatings for use in the sump screen design is the portion which enters the pool on or near the sump strainer, and the portion elsewhere that is in an area with flow velocities high enough to transport that debris.

3.f Head Loss and Vortexing NRC Issue 3. f The objectives of the head loss and vortexing evaluations are to calculate head loss across the sump strainerand to evaluate the susceptibility of the strainerto vortex formation.

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

2. Provide the minimum submergence of the strainer under small-break loss-of-coolant accident (SB LOCA) and large-breakloss-of-coolant accident (LB LOCA) conditions.

3. Provide a summary of the methodology, assumptions and results of the vortexing evaluation. Provide bases for key assumptions.

to W3F1-2008-0069 Page 21 of 87

4. Provide a summary of the methodology, assumptions, and results of prototypical head loss testing for the strainer, including chemical effects. Provide bases for key assumptions.

5. Address the ability of the design to accommodate the maximum volume of debris that is predicted to arrive-at the screen.

6. Address the ability of the screen to resist the formation of a "thin bed" or to accommodate partialthin bed formation.

7. Provide the basis for the strainerdesign maximum head lossl

8. Describe significant margins and conservatisms used in the head loss and vortexing calculationrs.

9. Provide a summary of the methodology, assumptions, bases for the assumptions, and results for the clean strainerhead loss calculation.

10. Provide a summary of the methodology, assumptions, bases for the assumptions, and results for the debris head loss analysis.

11. State whether the sump is partially submerged or vented (i.e., lacks a complete water seal over its entire surface) for any accident scenarios and describe what failure criteria in addition to loss of net positive suction head (NPSH) margin were applied to address potential inability to pass the required flow through the strainer.

12. State whether near-field settling was credited for the head-loss testing and, if so, provide a description of the scaling analysis used to justify near-field credit.

13. State whether temperature/viscositywas used to scale the results of the head loss tests to actual plant conditions. If scaling was used, provide the basis for concluding that boreholes or other differential-pressureinduced effects did not affect the morphology of the test debris bed.

14. State whether containment accident pressure was credited in evaluating whether flashing would occur across the strainersurface, and if so, summarize the methodology.

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The top elevation of the new SI Sump strainers at Waterford 3 is (-)5.935 ft MSL as shown on drawing 5817-13604 (Reference 66). As documented in the response to item 3.g.1, the minimum, water level for the SBLOCA is (-)5.79 ft MSL, and for the LBLOCA is (-)5.29 ft MSL.

This results in the top plate of the strainers being submerged by approximately 2" for a SBLOCA and 8" for a LBLOCA. Note that the top plate of the Waterford 3 sump strainer modules do not have perforated plate, but are solid, so suction does not occur on the top of the strainer.

WF3 Response 3.f.3:

No tests were run specifically for vortexing with specific assumptions. Instead vortexing observations were made as part of the module headloss test program. Module testing was conducted with the water depth over the strainers similar to the plant configuration per test specification 26A6833 (Reference 33). No vortexing or air entrainment was observed during testing. Testing was performed that included observation for vortexing at water levels 4-inches above the strainer (typical of plant minimum sump water level), 3-inches above the top of the strainers, 2-inches above the top of the strainer, and 1-inch above the top of the strainer, for the full range of debris loads. This test program ensures a high degree of confidence that a vortex or other form of air entrainment shall not occur with the Waterford 3 strainer. For added vortex prevention assurance, the original vortex breaker cages have been left in the sump on the intake piping. Pre-operational sump testing confirmed that the cages were effective in preventing both surface and subsurface vortices with flow rates higher than analyzed for Generic Letter 2004-02 resolution.

WF3 Response 3.f.4:

Module testing consists of scaling the plant's debris load and measuring the debris induced head loss across a module of a strainer. These tests determine the head loss characteristics of plant-specific debris as a function of scaled debris load and scaled flow rate.

Six module tests have been performed to analyze the six strainer bounding cases, with constant flow rates and scaling factors, and with varying particulate/fiber ratios, nominal debris bed thicknesses, and low density fiberglass quantities (Test Specification 26A6833 - Reference 33).

The test module is composed of ten 40" X 40" square perforated disks. The test module is mounted on the center of the test pool with the same floor clearance as its simulating strainer.

Water level was maintained 4" above the top of the test article, typical of the plant installation.

The flow rates for the module tests are calculated using Equation 3.f.4-1, which yields the same circumscribed approach velocity for the test module as for the proposed plant strainers.

Because the debris loads evaluated by the module tests result in thick, circumscribed debris beds, the circumscribed area approach velocity is appropriate for this test. The module test flow rate is 367 gpm (for both circumscribed area scaled flow rate and perforated area scaled flow rate).

Areac,... .scribed jest. nod,,e (3.f.4-1) test. inodule QPIOnt A (3.fl4-t A reaci rcumserib.ed.plant - Ar e a sacrificial X circ umscribedp, hmt A reaperforatedplant to W3F1-2008-0069 Page 29 of 87 Q = Flow Rate (gpm)

Area = Surface Area (ft2 )

The module test debris quantities in the test matrix were calculated using the debris loads provided by the design input (Reference 34) and Equation 3.f.4-2, which yields the same circumscribed debris bed thickness for the module test as in the proposed plant strainer.

M~fassebes ,s,,

M Sdbi.tef.no d

. Vo le debri ....... po~ledc.......... P oineerstnpiet.SI XJ Pr ,-is x A r ea ie,......ibed ,est. mod oe (3.f.4-2) bi XArea,,ct .cibed.ilant

.p lant -- A r e aS a r tfi A r e a circ unscribed cia l "Are a 0 ,,,

p, e p la nt arib ArIea pe'rforated, poant Volume debris.transported.to.sump - Volume of debris that is transported to sump (ft3 )

Pdebris = As-manufactured density of debris (lb/ft3)

Area = Surface Area (ft2 )

The module tests make use of the following assumptions:

o The flow rate is proportional to the circumscribed area of the strainers; o The debris load and flow rate is distributed equally among the strainers; o The debris bed is uniform - same thickness throughout perforated surface; o In the debris load calculation, the circumscribed surface area of a plant installed strainer is the actual circumscribed surface area minus the portion of sacrificial area attributed to the circumscribed area.

The impact of chemical effects on head loss was quantified through plant specific chemical effects testing. This testing included both 30 day integrated testing, and WCAP-16530-NP /

WCAP-16785-NP (References 58 and 53) based testing. See section 3.o of this report for a more thorough explanation.

WF3 Response 3.f.5:

During a LOCA at Waterford 3, the following types of debris may be generated by the high-energy steam and liquid impingement and water wash down/flow (Design Input - Reference 34):

3 o Fibrous Insulation: Nukon Fiber Blankets and Owens-Corning TIW II insulation of 591 ft volume is transported to the sump screen.

o Granular Insulation: Min-K and Microtherm insulation of 4.6 ft 3 volume with all transported to the sump screen.

o Latent Debris: 250 Ibm of latent debris is considered to be 15% by mass of fiber, and 85% particulate. All latent debris is assumed to be transported to the sump screen.

o Qualified and Unqualified Coating: Qualified Coatings (steel and concrete) of 13.5 ft 3 and Unqualified Coatings of 21.7 ft 3 is transported to the sump screen.

o Foreign Materials: The foreign materials (sacrificial area) are assumed to be 151 ft2 in area without taking credit for any overlap of these materials.

Testing was performed with two types of test articles: sectors and modules. [Proprietary Information Removed]

to W3F1-2008-0069 Page 30 of 87 The percentage of transported debris that adhered to each strainer is assumed to be equal to the strainer's percentage of total flow.

WF3 Response 3.f.6:

The nominal debris bed thickness for the tests ranges from 0.12" to 7.1" (Test Specification 26A6833 - Reference 33) in plant installed units with worst case-operating scenario. There is potential for a bed thickness matching the "thin bed" description to be formed during the strainer operation; however, the limiting head loss did not occur with a "thin bed" during Waterford 3 testing. The highest head loss occurs when 100% of the fiber is transported to the strainer, which included sufficient fibrous insulation to fill the strainer gaps and extend beyond the strainer perimeter, forming a "circumscribed" bed.

WF3 Response 3.f.7:

The GE hydraulic suction strainer design methodology is based on plant specific debris head loss testing. Debris head loss correlations were developed using the laboratory test results, scaled to the full plant design conditions.

The head loss is determined by summing up all the head loss components, as follows:

Head Loss= HLel, .is-plant +HL

+JJLde,,_plat, + HLchenical effect where:

Head Loss = maximum head loss of the strainer.

HLdebris plant - debris head loss at plant conditions.

HLclean-plant - clean head loss at plant conditions.

HLpipes&plenum = head loss on pipes and / or plenum.

HLchemicaieffect = head loss due to chemical effect.

WF3 Response 3.f.8:

The assumptions, margins and conservatisms are listed as follows (TDP-0186 - Reference 35):

o The flow rate is proportional to the perforated (sector test) or circumscribed (module) area of the strainers; o The debris load and flow rate are is distributed equally among the strainers; o The debris bed is uniform - same thickness throughout perforated surface; o In the debris load calculation, the circumscribed surface area of a plant installed strainer is the actual circumscribed surface area minus the portion of sacrificial area attributed to the circumscribed area; o 100% of particulate debris transported to the sumps is assumed to adhere to the strainers and contribute to head loss; to W3F1-2008-0069 Page 31 of 87 o All the labels and tags are modeled with 100% transport to the sump screen. The total sacrificial area is calculated by an equivalent to 100% of the original single sided surface area, counting for 0% overlap; o Due to extremely low approach and perforated flow velocities, laminar flow is assumed for debris head loss calculations; o Minimum water level at sump; o All coatings inside the ZOI are assumed to fail as particulate; o Head loss is calculated for indicated low end of sump water temperature and highest ECCS flow rate; o The upper circumscribed surface is assumed to be bounding in terms of air ingestion because air ingestion is evaluated at the top of the module, which is the closest surface to the water level.

WF3 Response 3.f.9:

Clean strainer system head loss is due to clean strainer head loss and plenum head loss.

Plant strainer clean head loss is calculated by scaling the test module clean head loss. Clean strainer head loss is due to the head loss inside the strainer discs, head loss as the flow exits the discs and enters the central cavity, and head loss inside the central cavity. The geometry of the test strainer is similar to that of the plant strainer. It is assumed that clean strainer head loss results primarily due to turbulent flow in the central cavity of the strainer, because the velocity through the perforated plates is relatively low and because water experiences an abrupt turn as it exits the discs and enters the central cavity. For central cavity strainers, assuming the gap width is the same, the scaling factor is based on the square ratio of the flow velocities at the entrance of the central cavity:

2 FlowRatePlantDisc dplant Headloss Clean := Headloss Test.Clean FlowRateTest dTest where:

Head IOSSCIean = plant strainer clean head loss Head IOSSTest.Clean = test strainer clean head loss FlowRatePlantDisc = plant disc flow rate, 34.599 gpm dPlant = plant central cavity diameter, 10.5 inches FlowRateTest = the test flow rate, varied by test dTest = test strainer central cavity diameter, varied by test Clean head loss data measured from the module test is the sum of module clean head loss, connecting pipe entrance head loss and dynamic head because the pressure transducer was installed inside the exiting piping just outside of the test module.

Results of the clean head loss and detail calculation can be found in the sizing report GENE-0000-0053-4416-P (Reference 36).

to W3F1-2008-0069 Page 32 of 87 Plenum head losses are due to the hydraulic losses associated with flow exiting the strainer into the ECCS sumps from the north and east strainers. Plenum losses for the Waterford 3 plenums are calculated in Appendix 2 of GENE-0000-0053-4416-P (Reference 36) to be 0.067 ft.

WF3 Response 3.f. 10:

Because containment sump water temperature following a LOCA is usually considerably greater than the temperature at which the hydraulic tests are run, debris head loss needs to be scaled to plant conditions as follows:

plant water densitYtest H~dbrt

tanttes = H Ldeb,.i _,te ** vis Cos ityp/att vicsty velocity p/att thicknesspi)t*(

H Ldei __,

plant

- vis Cos ity velocity,,,,

(debris, debris -thickness te water- densitypka,t ]

where:

HL = debris head loss through strainer in feet of water.

viscosity = dynamic viscosity of water in Ibm/ft-sec.

waterdensity = density of water in Ibm/ft 3.

velocity = approach velocity in ft/sec.

debris-thickness - nominal debris bed thickness in ft.

Nominal debris bed thickness is calculated as follows:

massfiber debris thickness:

densirtyfibe.

• perfbrated _ area where:

massfiber = mass of fiber debris in Ibm.

densityfiber = as-fabricated density of the fiber debris in Ibm/ft 3 .

perforatedarea = total surface area of the perforated plates in ft2.

The debris bed is assumed to be uniform, same thickness throughout perforated surface.

The sum of the strainer head loss and plenum head loss is tabulated below (GENE-0000-0053-4416-P - Reference 36).

Strainer.Head Plenum Head Total,.Head Loss:

• '"(ft)* : *,, .A ft), '

n-t)

S7-1S-59.2-CS_. 0.468 0.063 0.531 S7-2s8-10,0A-CS 0.463 0.063 0.526 S3-2M-1100-PS:11.,

@120 F 0.454 0.063 0.517 S3-2MM-'1,00PS.*,:

,. :2M-1,0,.F 0.314 0.063 0.377

@21F __ __ __ ___ __ __ __ __ __

to W3F1-2008-0069 Page 33 of 87 Note: The first two tests are sector test and the last two are module test. Design Basis head loss is based on module testing.

WF3 Response 3.f.11:

The strainers operate in fully submerged condition and are not vented to the atmosphere for any accident scenario. In addition to NPSH availability, failure criteria included the presence of vortexing or other forms of air entrainment, or the potential for a single large fiber' bed to blanket multiple strainers (during circumscribed bed formation) and block flow to some strainers.

Vortexing and air entrainment was not observed during several tests that mimicked the full range of plant debris loads, with either a representative water level or conservatively lowered water level. There is enough distance between strainers to preclude one large common fiber bed from obscuring some strainers; the nominal circumscribed debris bed is 7 inches under 100% fiber load conditions, and the closest distance between any two strainers is 32 inches, ensuring that fiber will not bridge between adjoining strainers and that water will have a path to each side of any individual strainer.

WF3 Response 3.f.12:

[Proprietary Information Removed]

The schematic of the pool configuration for the module test is shown in Figure 3.f-1.

[Proprietary Information Removed]

to W3F1-2008-0069 Page 34 of 87 Figure 3.f-lModule Test Setup for Strainer to W3F17-2008-0069 Page 35 of 87 WF3 Response 3.f.13:

Test strainer head loss is scaled based on velocity, viscosity, and bed thickness differences.

Debris head loss and clean strainer head loss are scaled independently.

The debris bed head loss results are scaled using the following equation:

? Plant hiplant -V plant Aplant ) tplant h/test V test (Qtest ttest A

Atest)/

Where:

hi = Debris Bed Head Loss (ft.)

v = Water Viscosity (Ibm/sec-ft)

Q = Sump Flow rate (ft3/s)

A = Perforated Area of strainer(s) (ft) (Does not include top and bottom external surfaces) t Debris bed thickness on perforated area (in.)

Testing was performed at a temperature less than plant temperature. The reduced test temperature results in an increase in viscosity. This difference in viscosity is accounted for by the first term in the equation above. The test head loss is multiplied by the ratio of plant water viscosity to test water viscosity, along with the other terms in the equation, to provide a test head loss that is representative of the plant conditions.

Viscosity scaling was performed for sector tests $7-2S-100-CS and S7-15-59.2-CS, and for module test S3-2M-100-PS. Boreholes were not present in these tests based on the test vendor's report.

Presence of boreholes in the debris bed is apparent in the photographs of the disassembled test sector taken after the test S7-4S-1 3.8A-CS. Viscosity scaling was not applied to this test. This is the 1/8-inch bed case for break S7. Clumps of debris are seen on the debris plate on other areas of the strainer. The strainer used a debris plate, which is intended to mitigate thin bed effects on head loss.

WF3 Response 3.f.14:

Containment accident pressure was not credited in evaluating whether flashing would occur across the strainer surface.

to W3F1-2008-0069 Page 36 of 87 3.g Net Positive Suction Head (NPSH)

NRC Issue 3.q The objective of the NPSH section is to calculate the NPSH margin for the ECCS and CSS pumps that would exist during a loss-of-coolant accident (LOCA) considering a spectrum of break sizes.

1. Provide applicable pump flow rates, the total recirculation sump flow rate, sump temperature(s), and minimum containment water level.

2. Describe the assumptions used in the calculations for the above parameters and the sources/bases of the assumptions.

3. Provide the basis for the required NPSH values, e.g., three percent head drop or other criterion.

4.' Describe how friction and other flow losses are accounted for.

5. Describe the system response scenarios for LBLOCA and SBLOCAs.

6. Describe the operational status for each ECCS and CSS pump before and after the initiation of recirculation.

7. Describe the single failure assumptions relevant to pump operation and sump performance.

8. Describe how the containment sump water level is determined.

9. Provide assumptions that are included in the analysis to ensure a minimum (conservative) water level is used in determining NPSH margin.

10. Describe whether and how the following volumes have been accounted for in pool level calculations: empty spray pipe, water droplets, condensation and holdup on horizontal and vertical surfaces. If any are not accounted for, explain why,

11. Provide assumptions (and their bases) as to what equipment will displace water resulting in higher pool level.

12. Provide assumptions (and their bases) as to what water sources provide pool volume and how much volume is from each source.

13. If credit is taken for containment accident pressure in determining available NPSH, provide description of the calculation of containment accident pressure used in determining the available NPSH.

14. Provide assumptions made which minimize the containment accident pressure and maximize the sump water temperature.

15. Specify whether the containment accident pressure is set at the vapor pressure correspondingto the sump liquid temperature.

16. Provide the NPSH margin results for pumps taking suction from the sump in recirculation mode.

WF3 Response 3.q.1:

Pump Flow Rates Item Injection Flow (pre RAS) Recirculation Flow (post RAS)

HPSI Pump! 985 gpm 985 gpm LPSI Pump 5650 gpm 0 gpm CS Pump 2250 gpm 2250 gpm Per Train Sump 8885 gpm 3235 gpm Table: 3.g-1: Applicable Pump flow Rates to W3F1-2008-0069 Page 37 of 87 Notes: (1) Waterford 3 has a common sump for both ECCS/CS trains (2) Pump flow rates are run-out Values Sump Temperature An analysis was performed to determine a maximum Sump Fluid temperature profile for a period of 30 days post LOCA (Figure 3.g.1-1). The Analysis was performed using the GOTHIC 7.0 program package. Conservative assumptions made in the development of the profile as stated in section 3.g.2.

2H,0.00 250.00 ...-

.I-.-.

V'aporTen p.

III7-Et Iri iii 2*'0.00 I I IA II IIN _ _ I 1 1 111111 1 1 11111 1 1 1 1 11 11 1 1 1 1 1 1 11 2M .00 220.00 1"x 111111 N NS 1

210.00 I 0.00 170.00 1E0.00 SunipTenop 1503.00 _lRecirculation Accuation

'140.00 130.00 120 .00 110.00 1C3.00 10.00 100.00 10000,00 100000.00 1.00 100O000.0 10,000000o.C Time CNec)

Figure 3.g.1-1: Safety Injection Sump Maximum Temperature Profile Time (Sec) Sump Temp.

(degF)

RAS 3253 173 RAS + 30 min 5053 187 Peak 24716 219 LOCA + 1 Day 86400 198 LOCA + 2 Day 172800 188 LOCA + 5 Day 432000 173 LOCA + 10 Day 864000 156 LOCA + 30 Day 2592000 152 Table 3.g.1-2: Temperature Points of Interest to W3F1-2008-0069 Page 38 of 87 Minimum Water Level The minimum water level in containment for a SBLOCA is (-)5.79 ft MSL. The minimum water level in containment for a LBLOCA is (-)5.29' MSL. These values are were determined in calculation MNQ6-4 (Reference 37). The calculation was revised to address items identified during and prior to the NRC GSI-191 audit of Waterford 3. Discussion of the analysis used to determine the minimum water levels can be found in section 3.g.8 of this report.

WF3 Response 3.q.2:

Flow Rate Assumptions

" All pump flow rates are run-out values. This maximizes suction loses which minimizes available NPSH.

" Both trains of ECCS and CS are in operation to maximize flow through common sump screen.

Temperature Assumptions

• Base input data is that used for licensing basis LOCA peak 24 hr pressure analysis with added uncertainties

• Only one train of Containment Spray in operation at design flow

• RWSP at conservative high temperature

• 1 out of 4 Containment Fan Coolers in Operation

• Shutdown Cooling Heat Exchanger parameters selected to minimize heat rejection to Ultimate Heat Sink

• Minimum Safety Injection Flow Minimum Water Level Assumptions

• Assumptions for the Minimum water level analysis are discussed in section 3.g.9 WF3 Response 3.q.3:

The required NPSH values are taken from the vendor certified pump performance curves and associated test data. Waterford 3 calculation ECM07-001 (Reference 51) uses a least squares curve fit polynomial to extrapolate the required NPSH curve data out to the pump's run-out flow rate.

WF3 Response 3.q.4:

Friction loss is being determined using the software program Pipe-FLO. Pipe-FLO performs steady state hydraulic analysis of fluid filled piping system using standard industry approved methods such as defined by Crane Technical Paper No. 410. All system configurations are being modeled for the suction side of the pumps and the system configuration resulting in the smallest NPSH available being used to determine acceptable screen head loss. Vendor supplied flow performance data is being utilized for components such as valves when available.

When vendor information was not available, conservative assumptions are being made using standard data from Crane Technical Paper No. 410. As stated in section 3.g.2, pump run-out flows are being utilized in all analysis to maximize friction loss.

Calculations are being performed for a maximum saturated sump water temperature of 210 degF. This temperature was determined to be the most limiting based on the significant to W3Fl-2008-0069 Page 39 of 87 increase in vapor head below 210 degF which is slightly below the saturation temperature for containment; which is initially at minimum pressure of 14.275 psia prior to the Loss of Coolant Accident (LOCA) event. The increase in vapor head more than compensates for the increase in piping/component losses due to the increase in viscosity of the fluid at temperatures below 210 degF. At temperatures above 210 degF, the increase in vapor pressure is offset by the conservative assumption that containment pressure is equal to the vapor pressure of the sump water which eliminates containment air pressure as a contributor to NPSH available. Higher temperatures also decrease water viscosity which decreases friction losses and further improves NPSH available.

The above methodology is consistent with Regulatory Guide 1.1 in that no credit is being taken for accident over pressure in containment.

WF3 Response 3.q.5:

Safety Injection System Response Scenarios for LBLOCA and SBLOCA The Safety Injection System (SIS) is arranged with two independent redundant trains, each functionally identical to the other and normally aligned to the Refueling Water Storage Pool (RWSP). The SIS is activated by the Safety Injection Actuation Signal (SIAS) which is initiated by either low pressurizer pressure or high containment pressure. The SIAS automatically starts the High Pressure Safety Injection (HPSI) and Low Pressure Safety Injection (LPSI) pumps and opens the motor operated valves (MOVs) that provide a flow path from the discharge of these pumps to the Reactor Coolant System. These MOVs don't actually go full open, but open to a preset position (set by valve position limit switch adjustment) to achieve a balanced flow and to prevent pump run-out. An installed spare HPSI pump is available which can be aligned to replace either of the other two HPSI pumps.

The HPSI system responds to an SIAS by automatically starting the aligned HPSI pumps and opening the cold leg injection flow control valves. If RCS pressure has not fallen below the 1450 PSIG shutoff head of the HPSI pumps, the system operates on recirculation flow until pressure decreases. As pressure decreases; HPSI flow will initiate and continue to increase as pressure falls.

The LPSI system responds by automatically starting the LPSI pumps and opening the cold leg injection flow control valves. The system will operate on recirculation flow until RCS pressure

• drops below the shutoff head of the pumps.

When a low level (10%) is sensed in the RWSP, the recirculation mode is initiated by the Recirculation Actuation Signal (RAS). At this time the HPSI pump suction is diverted to the Safety Injection Sump and the LPSI pumps are stopped.

Simultaneous hot and cold leg injection is used for both small break and large break LOCAs at 2-3 hours after the start of the LOCA and the RCS is not filled. In this mode, the HPSI pumps discharge lines are realigned so that the total injection flow is divided equally between the hot and cold legs. Simultaneous injection into the hot and cold legs is used as the mechanism to prevent the precipitation of boric acid in the reactor vessel following a break that is too large to allow the RCS to refill. Injecting to both sides of the reactor vessel ensures that fluid from the reactor vessel (where the boric acid is being concentrated) flows out the break regardless of the to W3F1-2008-0069 Page 40 of 87 break location and is replenished with a dilute solution of borated water from the other side of the reactor vessel.

Action is taken no sooner than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after the LOCA, since the fluid injected to the hot leg may be entrained in the steam being released from the core. After 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, the core decay heat has dropped sufficiently so that there is insufficient steam velocity to entrain the fluid being injected to the hot leg. Action is taken no later than 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> after the LOCA, in order to ensure that the buildup of boric acid is terminated, well before the potential for boric acid precipitation occurs (approximately 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />). Even though the action is required only for large breaks, it is taken for any LOCA so that the operator need not be required to distinguish between large and small breaks so early in the transient. Simultaneous hot and cold leg injection is not required for small breaks because, for small breaks, the buildup of boric acid is terminated when the RCS is refilled. Once the RCS is refilled, the boric acid is dispersed throughout the RCS via natural circulation.

Hot leg injection is established by closing the HPSI header flow orifice bypass valves and opening the hot leg header isolation valves. The orifices are preset to establish a 50% +/-5% flow balance between the hot and cold leg injection headers while preventing pump run-out conditions.

Long term cooling is initiated when the core is reflooded after a LOCA and is continued until the plant is secured. Two basic modes of long term cooling are available to the operator.

Entry into shutdown cooling (SDC) may be necessary if steam generator heat removal is lost, for certain sized breaks (small breaks). The shutdown cooling system is utilized if certain plant conditions exist.

When possible, the time necessary to refill the RCS and regain control of pressure and inventory depends on break size, break location, RCS cooldown rate and the number of HPSI pumps and charging pumps actuated. With only one HPSI pump actuated, for a break of about 3 inch diameter located on the bottom of the cold leg, it may take as long as 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to refill the RCS. With all injection pumps operable, the time is about 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

Before SDC is operated, RCS activity levels must be determined since the RCS fluid will be circulated outside of the containment building. When high activity is present, circulation outside containment has the potential for release to the environment. If potential for significant releases exists, it may be more desirable to continue cooling with the steam generator. The condensate inventory must be checked to ensure that the supply is sufficient to cool down the plant.

If SDC operation is determined to be appropriate, the SIS is aligned for cold leg injection and the RCS is cooled down and depressurized to allow entry into shutdown cooling.

If SDC operation is not appropriate or if the system is not available, it is desirable to continue RCS heat removal via the steam generators until no further steam is generated.

For large breaks, simultaneous injection provides effective long-term cooling by inducing a flushing flow through the core which will eventually result in a subcooled core. The core cooling is actually provided by the Containment Spray System via the Shutdown Cooling Heat Exchangers. This provides cooling for the Safety Injection Sump water which also provides a water source to the HPSI Pumps.

to W3F1-2008-0069 Page 41 of 87 CSS Response Scenarios to LBLOCA and SBLOOCA The Containment Spray (CS) System consists of two independent and duplicate trains to achieve the required redundancy. One loop operating alone is capable of providing the necessary post-accident heat removal. Each loop contains a CS Pump, a CS Riser Pump, a Shutdown Cooling Heat Exchanger, four spray ring headers, 116 spray nozzles, and the controls and instrumentation necessary to provide for proper system operation.

The CS System is actuated when the SIAS and the High-High Containment Pressure signal are coincident. This generates a Containment Spray Actuation Signal (CSAS) which opens the CS header isolation valves and starts the CS Pumps. The pumps initially take suction from the RWSP through a common header with the SIS pumps and delivers borated water to the spray nozzles located in the top portion of the steel containment. When RAS is initiated the CS pumps continue operation with the suction being taken from the Safety injection Sump.

WF3 Response 3.q.6:

Pump Pre RAS (Injection Post RAS (Recirculation Phase) Phase)

HPSI Operating Operating LPSI Operating Secured CS Operating Operating Table 3.g.6-1: Operational Status of ECCS and CS Pumps WF3 Response 3.q.7:

Only one single failure is being analyzed for NPSH effects. This single failure is a LPSI pump failing to trip upon receiving a Recirculation Actuation Signal. This results in an increased flow through the safety injection sump for an assumed time period of 30 minutes. Upon initiation of the Recirculation Actuation Signal, Operations procedures guide operators to verify that the LPSI pumps have stopped. If the pump has not tripped upon RAS, operators will take appropriate action as necessary to secure the pump.

Failure of a flow control valve will have no effect on the NPSH analysis due to the fact that the analysis assumes pumps are operating at run-out flows.

WF3 Response 3.g.8:

The minimum Safety injection Sump water level is determined by comparing water inventories available to fill the sump with the physical layout of the sump. Conservative assumptions intended to minimize the water level are made concerning available inventories, hold-up mechanisms, and the sump physical layout. Two single worst case (i.e., not time dependent) water levels are determined; one for the SBLOCA scenario and one for the LBLOCA scenario.

The results of the water level analysis can be found in section 3.g.1 while the additional detail on the analysis can be found in sections 3.g.9, .10, .11, & .12. The original minimum water level analysis was revised due to comments during the 2007 NRC GSI-191 Audit at Waterford 3.

to W3F1-2008-0069 Page 42 of 87 WF3 Response 3.q.9:

• Containment sump and Safety Injection sump do not communicate due to clogged drains.

" Containment sump fills to elevation of (+)7.5 ft (point of overflow to safety injection sump).

• Steam volume in containment at max containment temperature and is saturated.

• Refueling cavity assumed to not holdup water due to existence of two 6" floor drains in locations which are unlikely to clog. Drains go directly to Containment floor.

• Safeguards pump are assumed to leak a combined total of 0.5 gpm for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

• Film thickness for condensation assumed to be larger than that determined analytically.

Other assumptions are described in sections 3.g.10, .11, & .12.

WF3 Response 3.q.10:

Empty spray pipe - A portion of the Containment Spray System piping is empty prior to initiation of Containment Spray. This piping is credited for consuming an appropriate portion of the available inventory.

Water droplets - Containment Spray droplets are in transient from the spray nozzles to the containment floor. The volume of spray droplets for two CS trains operating at pump run-out flows are credited for consuming an appropriate portion of the available inventory.

Condensation - Heat sink surfaces condense steam from the atmosphere and develop a condensation film. The volume of the condensation film is credited for consuming an appropriate portion of the available inventory.

Holdup on horizontal and vertical surfaces - Other than condensation on heat sink surfaces, holdup was not considered on other horizontal or vertical surfaces. All concrete flooring above the containment floor has drains or is adjacent to steel grating with no curb to trap water. Floor drains direct water to the containment sump which is assumed to fill completely and overflow into the safety injection sump.

WF3 Response 3.q.11:

Credit was only taken for structural concrete below the flood level for the displacement of water.

Conservatively no credit is taken for the following items:

• Reactor Drain Tank

• Grout Pads /

• TSP Baskets

• Strainer Steel

• RCP support steel

• Structural steel.

WF3 Response 3.q.12:

The following assumptions are made for the sources of water:

0 Technical Specification minimum protected RWSP Volume only injected to W3F1-2008-0069 Page 43 of 87

• RWSP water at maximum Technical Specification temperature

• Safety injection tanks credited for Large Break LOCA only o Water at minimum Technical Specification Level o Water at maximum pre-accident Containment Temperature

• No credit for Reactor Coolant Spillage

• No credit for charging flow from Volume Control Tank or Boric Acid Makeup ranks WF3 Response 3.q.13:

No credit is taken for containment pressure above that present prior to the onset of the accident.

This is consistent with Regulatory Guide 1.1 (Reference 52).

WF3 Response 3.q.14:

As stated in section 3.g.13, no credit has been taken for containment pressure above that present prior to the onset of the accident.

WF3 Response 3.q.15:

As stated in section 3.g.4, NPSH available values are being calculated with a sump temperature of 210 degF with vapor pressure equal to saturation pressure. For the cases that assume a single LPSI pump fails to trip, credit is taken for the sump fluid being sub cooled at 190 degF with respect to containment being at atmospheric pressure.

WF3 Response 3.q.16:

The NPSH margin for the HPSI and CS pumps when taking suction from the sump in recirculation mode is determined in calculation ECM07-001 (Reference 51). The minimum margins during the recirculation mode of operation determined in the calculation are:

Pump Margin w/o Strainer strainer losses (ft) losses (ft)

HPSI A 4.426 0.377 HPSI A/B 3.929 0.377 (A Train)

HPSI B 1.966 0.377 HPSI A/B 3.96 0.377 (B Train)

CS A 5.963 0.377 CS B 5.818 0.377 Table: 3.g.16-1: Recirculation at 210F. No LPSI operating to W3F1-2008-0069 Page 44 of 87 Pump Margin w/o Strainer strainer losses (ft) losses (ft)

HPSI A/B 11.429 0.435 (A Train)

HPSI B 13.257 0.435 LPSI A 9.05 0.435 CS A 13.464 0.435 CS B 17.11 0.435 Table: 3.g.16-2: Recirculation at 190F w/ LPSI failure (Worst case lineup) 3.h Coatings Evaluation NRC Issue 3.h The objective of the coatings evaluation section is to determine the plant-specific ZOI and debris characteristicsfor coatings for use in determining the eventual contribution of coatings to overall head loss at the sump screen.

1. Provide a summary of type(s) of coating systems used in containment, e.g., Carboline CZ 11 InorganicZinc primer,Ameron 90 epoxy finish coat.

2. Describe and provide bases for assumptions made in post-LOCA paint debris transport analysis.

3. Discuss suction strainer head loss testing performed as it relates to both qualified and unqualifiedcoatings and what surrogate material was used to simulate coatings debris.

4. Provide bases for the choice of surrogates.

5. Describe and provide bases for coatings debris generation assumptions. For example, describe how the quantity of paint debris was determined based on ZOI size for qualified and unqualifiedcoatings.

6. Describe what debris characteristics were assumed, i.e., chips, particulate, size distribution and provide bases for the assumptions.

7. Describe any ongoing containment coating condition assessmentprogram.

WF3 Response 3.h.1:

The following types of coating systems are present, or approved to be used, inside Containment, per Waterford 3 Engineering Procedure NOECP-451 and Specification 1564.734 (References 42 and 43).

" Ameron Dimetcote 6(N)

" Ameron Amercoat 66 over Ameron Nu-Klad 11OAA primer

" Ameron Amercoat 66 over Ameron Nu-Klad 114 primer

" Ameron Amercoat 90

" Ameron Amercoat 90 over Ameron Amercoat 66 primer

" Ameron Amercoat 90 over Ameron Amercoat 71 primer

" Ameron Amercoat 90 over Ameron Dimetcote 6(N) primer

" Ameron Amercoat 90 over Ameron Dimetcote E-Z primer

. Ameron Amercoat 90 over Carboline CZ1 1SG primer

" Ameron Amerlock 400 NT to W3F1-2008-0069 Page 45 of 87

" Ameron Dimetcote E-Z

" Carboline 801

" Carboline 890

" Carboline 890 over Carboline Nutec 11 S primer

" Carboline 890 over Carboline Nutec 11 primer

" Carboline 890 over Carboline Nutec 1201 primer

" Carboline Carbo-Zinc 11

" Carboline CZ1 1 SG

" Carboline Nutec 1201 over Carboline Nutec 11 primer

" Carboline Nutec 1201 over Carboline Nutec 11S primer

" Carboline Phenoline 305 over Carboline CZ1 1 SG primer

" Carboline Phenoline 305 over Carboline 191 primer

" Carboline Phenoline 305 over Carboline Phenoline 305 primer

" Tnemec 801

" Unqualified coatings (alkyds, enamels, and epoxies) from various manufacturers.

WF3 Response 3.h.2:

In accordance with the guidance provided by NEI Guidance 04-07 and the SER for NEI 04-07 (References 2 and 3), all qualified coating debris within the ZOI is considered particulate and as such is modeled as transporting to the sump strainer.

Unqualified zinc coatings and indeterminate coatings are considered to fail as particles with 100% transport to the strainer.

50% of the qualified coatings on the containment liner dome and the liner above elevation 112' are assumed to fail. This is a conservative number based on coating failures at Waterford 3 to date. Degraded qualified epoxy coating systems are considered to fail as chips (see response to 3.h.6 below) and are subject to settling in low velocity area of the pool such that only a portion of the generated debris transport to the strainer. Degraded qualified coatings that fall on or near the strainer are considered not to have a chance to settle. Conservatively 20% of degraded qualified coatings are considered to fall on or near the strainer and thus transport to the strainer.

For analysis of the remaining degraded qualified coating transport additional CFD runs were performed. The additional CFD simulations consider break locations farther from the strainer in order to maximize the portion of the pool where flow velocity is too high for settling to occur.

The study in NUREG/CR-6916 (Reference 19) found that the lowest incipient tumbling velocity, the velocity at which the coating chips similar to Waterford 3 debris would move on the floor was 0.264 feet per second for the "curled" 1-to-2-inch chips. Conservatively a transport velocity of 0.2 feet per second is used for all chips with a size greater than 1 /6 4 th inch. Based on the CFD simulations the bounding portion of the cOntainment pool area with a velocity in excess of 0.2 feet per second is determined to be 12.7%. As a judgment, this area is increased to 15% and the failed coatings in the remaining 85% of the pool are considered subject to settling.

WF3 Response 3.h.3 and 3.h.4:

The prior Waterford 3 sector test results indicated that the bounding condition for simulating plant LOCA debris-generation is 100% of the transported fiber and particulate (GENE-0000-0053-4416-P - Reference 36).

to W3F.1 -2008-0069 Page 46 of 87

[Proprietary Information Removed]

This will have a greater effect on head loss than would the heavier latent dirt.

WF3 Response 3.h.5:

In order to determine the amount of qualified coating debris generated at Waterford 3, structural and civil drawings are consulted. The bounding break location is determined from inspection of these drawings, then the total surface area of coated steel and concrete within a 4D ZOI of the break location is calculated. The maximum allowable coating thickness, per the plant coating specification, is then applied to this surface area to determine the total coating debris volume. A spherical ZOI of 4D for qualified coatings was selected based on WCAP-16568-P (Reference 22). This testing concluded that a spherical ZOI of 4D is conservative for the qualified epoxy and the qualified zinc coatings used by Waterford 3.

Unqualified (degraded qualified or indeterminate) coatings are assumed to 100% fail. The unqualified coating debris volume is based on the thickness of similar coatings on other materials in containment.

WF3 Response 3.h.6:

In accordance with the guidance provided in NEI Guidance 04-07 (Reference 2) and the SER for NEI 04-07 (Reference 3), all qualified coating debris within the ZOI, unqualified zinc coating debris, and indeterminate coating debris are treated as particulate and are therefore transported entirely to the sump strainer.

Degraded qualified coatings are. considered to fail as chips with a size distribution per Alion document ALION-REP-TXU-4464-02 (Reference 25) and letter TXX-07156 (Reference 26).

The document ALION-REP-TXU-4464-02 (Reference 25) stated that 49.5% of coating particles were less than 1 /8th inch in size. Letter TXX-07156 (Reference 26) further identifies that 12.375% (25% of 49.5%) of the coating particles are 6 mils (0.006 inches) and 37.125% (75% of 49.5%) are 15.6 mils ( 1/ 6 4th inch). 100% of coating particles with a size less than 1 /6 4 th inch will not settle and will transport to the sump. This quantity amounts to approximately 12.375% of the inventory. The remaining 87.625% of the inventory may settle in favorable flow conditions.

The degraded qualified coating systems at Waterford 3 are compared with the test data using NUREG/CR-6916, ALION-REP-TXU-4464-02, TXX-07156, and CCCL letter dated 9/20/07 (references 19, 25, 26 and 27). The data reported in NUREG/CR-6916 (Reference 19) are for to W3F1-2008-0069 Page 47 of 87 the failure characteristics of many coatings, including epoxy applied as a topcoat over inorganic-zinc. The epoxy top coats applied over a zinc rich primer, Category ZE, are similar to the Carboline Phenoline 305 applied as a topcoat over Carboline Carbo-Zinc 11. The painting system used on the containment dome is Carboline Phenoline 305 applied as a topcoat over Carboline Carbo-Zinc 11 primer. This system is also one of four Service Level 1 paint systems approved for the containment liner. The other systems include Carboline 305 topcoat over Carboline 191 primer, Carboline 801 as a primer and topcoat and Amerlock 400 NT (a one coat system). All are epoxy systems, which are expected to exhibit failure characteristics of Carboline 305. The CCCL letter dated 9/20/07 (Reference 27).confirms that the size distribution presented in ALION-REP-TXU-4464-02 and TXX-07156 (references 25 and 26) is applicable to Carboline Phenoline 305 coatings systems. The CCCL letter dated 9/20/07 (Reference 27) also confirms that 100% of inorganic-zinc coatings will fail as small fines.

WF3 Response 3.h.7:

Waterford 3 performs an inspection of containment coatings each refueling outage. As defined in procedure NOECP-451 (Reference 42), the scope of the coating inspections are coated concrete and steel surfaces inside the SI Sump, the containment liner plates, and approximately 10% of the remaining coated surfaces excluding concrete and insulated piping.

3.i Debris Source Term NRC Issue 3.i 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 recirculationfunctions.

Provide the information requested in GL 04-02 Requested Information Item 2. (0 regarding programmaticcontrols taken to limit debris sources in containment.

GL 2004-02 Requested Information Item 2(f)

A description of the existing or planned programmatic controls that will ensure that potential sources of debris introduced into containment (e.g., insulations, signs, coatings, and foreign materials) will be assessed for potential adverse effects on the ECCS and CSS recirculation functions. Addressees may reference their responses to GL 98-04, A Potential for Degradationof the Emergency Core Cooling System and the Containment Spray System after a Loss-of-Coolant Accident Because of Construction and Protective Coating Deficiencies and Foreign Material in Containment," to the extent that their responses address these specific foreign materialcontrol issues.

In responding to GL 2004 Requested Information Item 2(f), provide the following:

1. A summary of the containment housekeeping programmatic controls in place to control or reduce the latent debris burden. Specifically for RMI/low-fiber plants, provide a description of programmatic controls to maintain the latent debris fiber source term into the future to ensure assumptions and conclusions regardinginability to form a thin bed of fibrous debris remain valid.

to W3F1-2008-0069 Page 48 of 87

2. A summary of the foreign material exclusion programmatic controls in place to control the introduction of foreign material into the containment.

3. A description of how permanent plant changes inside containment are programmatically controlled so as to not change the analytical assumptions and numerical inputs of the licensee analyses supporting the conclusion that the reactorplant remains in compliance with 10 CFR 50.46 and related regulatory requirements.

4. A description of how maintenance activities including associated temporary changes are assessed and managed in accordance with the Maintenance Rule, 10 CFR 50.65.

If any of the following suggested design and operational refinements given in the guidance report (guidance report, Section 5) and SE (SE, Section 5.1) were used, summarize the applicationof the refinements.

5. Recent or planned insulation change-outs in the containment which will reduce the debris burden at the sump strainers

6. Any actions taken to modify existing insulation (e.g., jacketing or banding) to reduce the debris burden at the sump strainers

7. Modifications to equipment or systems conducted to reduce the debris burden at the sump strainers

8. Actions taken to modify or improve the containment coatings program WF3 Response 3.i.1 Entergy's fleet wide FME program procedure provides the requirements and guidance to prevent and control introduction of foreign materials into structures, systems, and components.

Also included within this procedure are steps to take to reestablish and maintain FME areas to prevent foreign material intrusion and to recover/monitor when a loss of FME integrity has occurred.

Housekeeping and foreign material assessments after a plant outage and prior to heat up are performed at the direction of the Waterford 3 operating procedure which provides the requirements and guidance to perform walkdowns of the RCB to assess debris that may represent a risk of blocking the SI recirculation sump screen.

WF3 Response 3.i.2 See Response to Issue 3.i.1.

WF3 Response 3.i.3 The Entergy fleet configuration control procedure controls permanent plant changes inside the RCB so as to not change the analytical assumptions and numerical inputs. A design input consideration was added to the Entergy fleet design input screening procedures to specifically address the SI Sump GL 2004-02 program. Engineers are required to review the impact of a proposed change to determine if there would be an impact to the performance of the ECCS sump. Waterford 3 and/or Entergy fleet procedures require reviews of physical changes in the RCB to address specific areas. The specific areas that are addressed, as a minimum, are:

• Insulation inside containment,

° Coatings inside containment, to W3F1-2008-0069 Page 49 of 87

" Volumes in containment,

" Addition of materials inside containment that may produce chemical effects in the post-LOCA flood pool/environment.

WF3 Response 3.i.4 Temporary changes at Waterford 3 are subject to the same requirements for reviews as for permanent changes. Therefore, the design input and impact screenings will determine if the temporary change should be reviewed for any potential impact on the SI Sump screens.

Entergy fleet procedures also provide guidance such as the 50.59 Review Process procedure, which provides details and guidance on maintenance activities; and the On-Line Work Control Process procedure, which establishes the administrative controls for performing on-line maintenance of SSCs in order to enhance overall plant safety and reliability.

WF3 Response 3.i.5 There are no recent or planned insulation change-outs in the Waterford 3 containment which will reduce the debris burden at the sump strainers.

WF3 Response 3.i.6 No modifications to existing insulation were performed to reduce the debris burden at the sump strainers.

WF3 Response 3.i.7 There were no modifications made to equipment or systems to reduce the debris burden at the sump strainers.

WF3 Response 3.i.8 The coatings procedure for Waterford 3 was revised to provide better instructions to the craft on cleanliness when preparing surfaces.

3.j Screen Modification Package NRC Issue 3.1 The objective of the screen modification package section is to provide a basic description of the sump screen modification.

1. Provide a description of the majorfeatures of the sump screen design modification.

2. Provide a list of any modifications, such as reroute of piping and other components, relocation of supports, addition of whip restraints and missile shields, etc., necessitated by the sump strainermodifications.

WF3 Response 3.i.1 The modification for the SI Sump replaced the original box type screen over the SI Sump. To prevent debris from entering the open sump, the original rectangular box shaped screen had to W3F1-2008-0069 Page 50 of 87 0.078 inch square openings (0.11 inch diagonally) and completely covered the sump inlet. The box screen provided approximately 198.2 ft 2 of available flow area. Inaddition, the box screen was mounted approximately 3.75" above the containment floor which helped to prevent sediment from entering the pit. A divider screen separated the two (SIS) suction lines located in the sump. During RF14, modification ER-W3-2003-0394-001 (Reference 30) installed a passive, safety-related, Nuclear Modularized Stacked Disk Strainer assembly engineered and manufactured by General Electric Energy in place of the original screen. The new strainer arrangement for Waterford 3 consists of 11 strainer modules mounted on top of the plenum mounted over the existing'sump and over the concrete floor to the north and to the east of the sump. The new SI Sump partition that separates the two SIS suction lines is a section of stainless steel grating made of 1" x 1/8" bars separated at 1-3/16". The partition is supported by angles attached to the existing anchor plates. The modification was installed during the 2006 refueling outage.

The effective surface area of the new strainer for each module is 336.3 ft 2, for a total of approximately 3,700 ft2 . There are 11 essentially identical modules mounted on the 8" high

.plenum over the SI Sump. Each module is bolted to the plenum and the plenum is bolted to the containment floor. The plenum prevents debris from entering the system between the modules.

Each module is constructed of 17 stacked perforated disk sets with hole-diameters of 0.093 inch. A disk set is composed of two perforated disks separated from each other by radial fingers and by an outer support, the finger frame. The water enters from the top and bottom disks into the intermediate space, travels towards the center and then axially towards the strainer base. Perforated inner spacer rings separate the disk sets from each other. The modules are located on top of the plenum, approximately 8 inches above the containment floor.

The sump is now totally enclosed by the plenum, preventing material from falling directly into the sump without passing through the strainer assemblies.

The plenum extensions to the north and east side of the sump have internal dimensions 7.25 inches high by 41 inches wide.

The plenum has openings in the top to admit flow of strained water from the modules. The modules are bolted to the plenum, which in turn is bolted to the containment slab. The plenum is made of structural shapes: angles and plates. The strainer design allows for disassembly, replacement of modules, or addition of future modules as needed. The plenum has two access openings to allow access into either side of the sump during outages for inspection and testing.

The access openings are approximately 40" X 40". The access covers are bolted to the plenum to control access. Each module also has an inspection port on top to allow visual inspections inside the module, if necessary.

WF3 Response 3.J.2 A safety related low level switch was relocated due to the installation of the new SIS sump strainer assembly. The switch housing is now mounted on top of the plenum and is on the north end of the sump so that it can detect water at the lowest point of the SIS sump floor. The switch elevation at the new location is identical to the previous elevation and maintains the same function and switch setpoint. This switch is seismically mounted on the top of the plenum. The level switch guard pipe and mounting plate were modified to match the plenum mounting plate and are seismically installed to preclude the potential for a strainer bypass path. The seismic bracing mounted at the sump floor was relocated to align with the guard pipe. The portion of to W3F1-2008-0069 Page 51 of 87 the new guard pipe above the plenum is welded to 150-lb flange and welded to the bolted cover plate. This portion of guard pipe above the plenum has no holes on the pipe wall. The instrument plate-to-plate bolted connection to strainer plenum has a zero gap preventing debris from entering the sump.

Nineteen (19) TSP baskets were relocated to eliminate interference with the new strainer assembly. The new location is at the same elevation of the containment, but at the north end of the building. The relocated TSP baskets are seismically mounted on the concrete floor of containment at elevation -11.0 feet.

The sensors for Reg. Guide 1.97 Type B, Category 1 Level transmitters were temporarily removed from the' mounting plate inside the sump to allow for installation of the new strainer..

After installation of the new strainer was completed the sensors were remounted to the same mounting plate inside the sump. There was no change to the transmitter mounting and the setpoint. The 1/4" diameter capillary tubing and tube track between the level transmitters and the sensors was re-routed. The capillary tubing penetrates the top of the plenum. A plenum opening of 6" x 6" with a W"diameter slotted hole and a cover plate with a slotted hole was provided to allow the re-mounting of the instrument without disconnecting the capillary tubing from the, instrument. The slotted hole design ensures zero gap thereby preventing debris from entering the sump.

3.k Sump Structural Analysis NRC Issue 3.k The objective of the sump structuralanalysis section is to verify the structuraladequacy of the sump strainerincluding seismic loads and loads due to differential pressure, missiles, and jet forces. Provide the information requested in GL 2004-02 Requested Information Item 2(d)(vii).,

GL 2004-02 Requested Information Item 2(d)(vii)

Verification that the strength of the trash racks is adequate to protect the debris screens from missiles and other large debris. The submittal should also provide verification that the trash racks and sump screens are capable of withstanding the loads imposed by expanding jets, missiles, the accumulation of debris, and pressure differentials caused by post-LOCA blockage under flow conditions.

1. Summarize the design inputs, design codes, loads, and load ,combinations utilized for the sump strainerstructuralanalysis.

2. Summarize the structural qualification results and design margins for the various components of the sump strainerstructuralassembly.

• 3. Summarize the evaluations performed for dynamic effects such as pipe whip, jet impingement, and missile impacts associated with high-energy line breaks (as applicable).

4. If a backflushing strategy is credited, provide a summary statement regarding the sump strainerstructuralanalysis considering reverse flow.

WF3 Response 3.k.1 The inputs and loads are discussed in the following paragraphs.

to W3F1-2008-0069 Page 52 of 87 Differential Crush Pressure is the pressure difference across the strainer components. The value is equal to the static pressure outside the strainer minus the static pressure inside the strainer system. The "Design" crush pressure is analogous to design pressure for a pressure vessel; in. that it is > the limiting pressure loss across the equipment specified for hydraulic system design. Crush pressure was applied to the assembly to demonstrate adequacy for pressure loading of the strainer perforated plates, perforated spacer rings and plenum plates including supports.

Equivalent solid plate properties (Poisson's Ratio and Modulus of Elasticity) including a stress multiplier were determined for the disc perforated plates by performing a finite element analysis of perforated plate and solid plate following the guidance contained in the ASME Code Section Ill, Appendix A, Article A-8000. A finite element model of the strainer assembly was then developed with the perforated plates modeled as equivalent solid plates. The strainer disks are modeled as shell and beams while the plenum consists of solid elements. The equivalent solid plate properties are used to model the strainer perforated plates for structural analyses. The equivalent properties for the perforated plate are:

E*= 0.43 E, equivalent modulus of elasticity v* = 0.33, equivalent Poisson's ratio K = 2.33 stress multiplier The equivalent properties were applied to the solid plates in the ANSYS finite element model to simulate the plate perforations.

Modal frequencies in air and in water were determined according to the seismic analysis requirements and were used to determine the seismic accelerations. The load cases required by the design specification were analyzed to determine stresses in all components.

The strainers are designed for a 40-year life. Thermal fatigue was, evaluated qualitatively, and dismissed as insignificant since the normal operation temperature cycle ranges are small (50 0 F) and there is only one LOCA temperature cycle with a range of 199 0 F. Material Properties were based on stainless steel SA 240, Type 304 which is the material of construction. The material properties used were selected from the ASME code and are shown in Table 3.k.1-1.

Table 3.k.1-1 Material Properties Temperature Material / Property Unit 300°F SA-240, Type 304 Elastic modulus' psi 27.OE+6 Coefficient of thermal expansion in/in-0 F 9.OE-6 Poisson's ratio 0.3 Density lb/in3 0.289 Stress Allowable psi 16700 Tie-rod bolt material SA-193, B8 Elastic modulus psi 27.2E+6 Coefficient of thermal expansion in/in-0 F 9.OE-6 Yield Strength psi 22500(1)

(1) Yield strength of SA-193, B8 material at 70F is 30,000 psi.

to W3F1-2008-0069 Page 53 of 87 Load Definitions and Combinations Strainers, Plenums, the Partition and the Sensor, and supports are designed for the loads and load combinations described in this section.

Load Definitions W Strainer Assembly Weight in Air, Normal Plant Operation WD Strainer Assembly Weight in Water + Debris Weight + Hydrodynamic Mass, LOCA TEmax Thermal Expansion in Water, LOCA TEop Thermal Expansion in Air, Normal Plant Operation PO Containment Pressure Pd Containment Design Pressure Pcr Differential Crush Pressure, LOCA OBE1 Operating Basis Earthquake Inertia Loading in Air OBE2 Operating Basis Earthquake Inertia Loading in Water + Debris Mass +

Hydrodynamic Mass SSE1 Safe Shutdown Earthquake Inertia Loading in Air SSE2 Safe Shutdown Earthquake Inertia Loading in Water + Debris Mass

+Hydrodynamic Mass Load Combinations Strainers and Plenums Design = W + TE0 p + P0 + OBE1 Level B = Pd + WD + OBE2 + TEmax + Pcr Support Structures Design = W + TEop Level B = WD + OBE2 + TEmax Level D = WD + SSE2 + TEmax The strainer assembly shall withstand a live load of 250 pounds during outages. This load is negligible compared to operating loads and no specific analysis was performed.

The seismic loads are based on the horizontal and vertical inertial accelerations specified by the seismic response spectrums according to the first mode frequency in water. The design pressures, P0 and Pd, have no impact and add nothing to the load combinations cited above because the strainer system is an open system that is not pressurized by containment pressure but is loaded by crush pressure, Pcr. Hydrodynamic mass values and debris weights, are included where applicable.

Table 3.k.1-2 Mass Properties Dry weight of 1 strainer lb 2,374 Submerged weight of 1 strainer lb 3,560 Dry weight of plenum lb 9,753 to W3F1-2008-0069 Page 54 of 87 Submerged weight of plenum in vertical direction lb 47,788 Submerged weight of plenum in x and z direction lb 12,679 Dry weight entire assembly lb 35,862 Submerged weight of entire assembly in vertical direction lb 86,952 Table 3.k.1-3 Coefficients Used for Seismic Analysis*

OBE lateral 0.25 g OBE vertical 0.20 g SSE lateral 0.38 g SSE vertical 0.30 g

• These accelerations are above the ZPA value, therefore no multiplier is applied The structural response due to OBE & SSE is different depending on whether the equipment is in air or in water.

Loads used in the stress analysis include the weight of the strainer assembly, debris, contained water, crush pressure due to pump operation, and seismic loads. The lateral and vertical inertial accelerations were obtained from the seismic response spectra corresponding to the first mode frequency of the equipment when submerged.

Crush pressure was applied to the strainer plates, spacer rings and plenum plates. The weight of the equipment in water was analyzed as the sum of the weight of the assembly in air, the debris weight and the hydrodynamic mass and contained water. A 1-2g" inertial load in the vertical direction was used to represent the dead weight of the equipment.

WF3 Response 3.k.2 All Qualification information is contained in GE calculation GENE-0000-0054-9349 (Reference 56).

Finite element analyses were performed for all components using ANSYS Version 10 computer program. Stresses from design load combinations are compared with the ASME Code Section Ill, Subsections NC, and ND stress limits. Stress margins for the limiting components were calculated for the Design Condition, Service Level B, and Service Level D Load Combinations.

Table 3.k.2-1 shows selected calculated stresses. The minimum stress margins are shown in Tables 3.k.2-2 (Strainer Components), 3.k.2-3 (Partition Components) and 3.k.2-4 (Sensor Components).

Table 3.k2-1 Stress Summary for Strainer Components Load Combinations & Max Calculated Stresses, ksi Component Design Level B Level D W + OBEI WD + OBE2 + Pcr WD + SSE2 + Pr

-Perforated Plate (1) 1.8 3.5 4.0 Frame & Fingers 3.3 4.7 5.8 Spacers 7.4 10.3 12.5 to W3F1-2008-0069 Page 55 of 87 Strainer Base 1.8 2.3 2.8 Tie Rod 1.1 1.2 1.7 Plenum 9.0 12.8 14.8 2

Allowable Pm Stress( ) 1.0 x S 1.1 x S 2.0 x S Allowable Stress (2) 16.7 18.4 33.4 Notes:

1) Perforated plate includes intensification factor of 2.33

2) For conservatism, the allowable for membrane stress is used, which is the lowest Table 3.k.2-2 Margin Summary for Strainer Components Load Combinations & Margins (1)

Component Design Level B Level D W + OBE1 WD + OBE2 + Pcr WD + SSE2 + Pcr Perforated Plate (1) 8.3 4.3 7.4 Frame & Fingers 4.1 2.9 4.8 Spacers 1.2 0.8 1.7 Strainer Base 8.4 6.9 10.9 Tie Rod 14.6 14.4 18.6 Plenum 0.9 0.4 1.3 Notes:

1) Margin = (Allowable/Calculated) - 1 Table 3.k.2-3 Stress (ksi) Summary for Partition Components Component Design Level B Level D Partition Assembly 1.3 6.4 16.9 Allowable Stress 16.6 18.4 33.4 Partition Margin 11.8 1.9 1.0 Table 3.k.2-4 Stress (ksi) Summary for Sensor Components Component Design Level B Level D Sensor Assembly 0.4 7.4 11.6 Allowable Stress 16.6 18.4 33.4 Sensor Margin 40.5 1.5 1.9 The strainer disk surfaces are covered by a woven wire cloth, which is resistance welded to the perforated plate. In the Waterford 3 application this woven wire mesh is used solely to enhance the debris carrying capability with respect to hydraulic head loss of the disks and no structural credit is taken for its presence; however the mass is included in the analysis. It is necessary to assure that the Woven Wire remains attached to the disk when the disk is subjected to seismic loading and when the disk deflects due to the pressure drop across the disk faces. This assurance is obtained as the result of testing performed by GEH in which the composite of to W3F1-2008-0069 Page 56 of 87 perforated plate and woven wire was deflected over 1" with the wire remaining attached. This deflection is at least an order of magnitude greater than will be experienced in service.

WF3 Response 3.k.3 GENE-0000-0048-9192 (Reference 64) is an evaluation which concluded that the strainer assembly is not subject to pipe whip, jet impingement, or missile impact associated with a HELB.

WF3 Response 3.k.4 A backflushing strategy is not credited in the Waterford 3 analyses.

3.1 Upstream Effects NRC Issue 3.1 The objective of the upstream effects assessment is to evaluate the flowpaths upstream of the containment sump for holdup of inventory, which could reduce flow to and possibly starve the sump.

Provide a summary of the upstream effects evaluation including the information requested in GL 2004-02, "RequestedInformation," Item 2(d)(iv).

GL 2004-02 Requested Information Item 2(d)(iv)

The basis for concluding that the water inventory required to ensure adequate ECCS or CSS recirculation would not be held up or diverted by debris blockage at choke-points in containment recirculationsump return flowpaths.

1. Summarize the evaluation of the flow paths from the postulated break locations and containment spray washdown to identify potential choke points in the flow field upstream of the sump.

2. Summarize measures taken to mitigate potential choke points.

3. Summarize the evaluation of water holdup at installed curbs and/or debris interceptors.

4. Describe how potential blockage of reactor cavity and refueling cavity drains has been evaluated, including likelihood of blockage and amount of expected holdup.

WF3 Resp~onse 3.1.1 The Waterford 3 containment is mostly uncompartmentalized with the exception of the pressurizer room. There are no structures totally surrounding the major components (SG, RCP, etc.) of the RCS. All RCS components with the exception of the pressurizer are within the SG cavities (D-Rings), but the SG cavities are open to each other (on the north side below elevation

-4 ft., open to the annulus below elevation -1 ft and are open to the dome above elevation

+62.25 ft. The PZR is located in a separate room with an opening in the floor that connects to the containment annulus.

Waterford 3 does not have any significant inactive volumes other than the reactor cavity and containment sump. For significant quantities of debris to be trapped in the reactor cavity or to W3F1-2008-0069 Page 57 of 87 containment sump, the break location would have to be at the reactor (within the reactor cavity).

As described in Section 3.e, relatively small quantities of debris that transport to the sump will be created for any break in the reactor cavity itself. For breaks outside the primary shield wall, significant quantities of debris would not be transported to the reactor cavity or containment sump by flowing water during the pool fill-up. This is because the only flow paths from the active pool to the reactor cavity and containment sump at the minimum flood elevation are through several floor drains located in the containment at elevation -11.0 ft that drain to the containment sump.

WF3 Response 3.1.2 As no potential choke points were identified for Waterford 3, no mitigation measures were necessary.

WF3 Response 3.i.3 Waterford 3 does not have any curbs on the -11 ft basemat elevation. Throughout containment, where slabs are adjacent to open areas or grating areas, there are no concrete curbs. For open areas, there are kickboards on the handrails, but these are not flush against the surface of, the concrete and will allow water to flow under and around them.

WF3 Response 3.i.4 Calculation 2005-05500 (Reference 29) documents that the refueling cavity has two 6-inch drain lines (without screens) that drain to the containment floor, and by one 4-inch line that drains to the containment sump. In the event that large debris is propelled over the SG cavity walls into.

the refueling cavity, the debris must land on a drain in order to clog it since the velocities in the cavity are too low to transport a large piece of debris to a drain. During plant operations, the Upper Guide Structure Lift Rig (UGSLR) is stored directly above one of the 6" drains. The UGSLR will prevent any debris larger than 6" from falling directly onto the drain. Any smaller debris that transports to these two 6" drains will pass through the drains since they do not contain screens. Therefore, there will always be drainage available from the refueling cavity to the active pool.

3.m Downstream Effects - Components and Systems NRC Issue 3.m 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.

Provide the information requested in GL 04-02, "Requested Information," Item 2.(d)(v) and 2.(d)(vi) regardingblockage, plugging, and wear at restrictionsand close tolerance locations in the ECCS and CSS downstream of the sump by explaining the basis for concluding that inadequate core or containment cooling would not result due to debris blockage at flow restrictions in the ECCS and CSS flowpaths downstream of the sump screen, (e.g., a HPSI throttle valve, pump bearings and seals, fuel assembly inlet debris screen, or containment spray nozzles). The discussion should consider the adequacy of the sump screen's mesh spacing and state the basis for concluding that adverse gaps or breaches are not present on the screen surface. For GL 2004-02, Item 2(d)(vi) provide verification that the close-tolerance to W3F1-2008-0069 Page 58 of 87 subcomponents in pumps, valves and other ECCS and CSS components are not susceptible to plugging or excessive wear due to extended post-accident operation with debris-laden fluids.

1. GL 2004-02 Requested Information Item 2(d)(v)

The basis for concluding that inadequate core or containment cooling would not result due to debris blockage at flow restrictionsin the ECCS and CSS flowpaths downstream of the sump screen, (e.g., a HPSI throttle valve, pump bearings and seals, fuel assembly inlet debris screen, or containment spray nozzles). The discussion should consider the adequacy of the sump screen's mesh spacing and state the basis for concluding that adverse gaps or breaches are not present on the screen surface.

2. GL 2004-02 Requested Information Item 2(d)(vi)

Verification that the close-tolerance subcomponents in pumps, valves and other ECCS and CSS components are not susceptible, to plugging or excessive wear due to extended post-accidentoperation with debris-ladenfluids.

3. If NRC-approved methods were used (e.g., WCAP-16406-P with accompanying NRC SE) briefly summarize the applicationof the methods.

4. Provide a summary and conclusions of downstream evaluations.

5. Provide a summary of design or operationalchanges made as a result of downstream evaluations.

WF3 Response 3.m.1 As a result of GSI-191, new SI Sump screens were installed in Waterford 3. The nominal perforated plate hole in the screens is 3/32 inches (ER-W3-2003-0394-001 - Reference 30).

Strainer walk downs are performed prior to start up from plant outages to confirm no gaps, breaches, or openings greater than 3/32 inch exist in the strainers.

The ability of the ECCS equipment required to pass debris laden fluid during the recirculation phase after a postulated accident is evaluated in Calculation 2005-02820 (Reference 31). This evaluation determined the ECCS equipment that would be in the post-accident recirculation path and reviewed the dimensions of close-tolerances in this ECCS equipment against the acceptance criteria of 1.1 and 2 times the screen hole size. The HPSI pumps, CS pumps, CS headers solenoid operated valves, HPSI pump recirculation flow orifices, HPSI header throttle valves and RC loop throttle-valves were determined to have minimum flow clearances small enough to require wear evaluations.

To resolve these issues, calculation 2005-12840 (Reference 32) was prepared. This calculation determined that most components in the system would not be blocked by debris. The HPSI pump seal and the CS pump cyclone separator required further evaluation.

Calculation 1062-0802-0015-4 (Reference 67) evaluated the HPSI pump seals for the injection of debris during a LOCA. This evaluation concluded that the potential for mechanical seal failure due to debris blocking axial movement of the rotating seal face is considered low. There is no potential for significant debris-induced wear of the seal faces due to the tight running gap.

The HPSI seals use a priming ring to recalculate water through a heat exchanger and the seal.

Calculation 1062-0015-03 (Reference 68) concluded that the CS pump cyclone separators will continue to provide clean water to the CS pump mechanical seals following a design basis to W3F1-2008-0069 Page 59 of 87 accident. This conclusion is based on a comparison of separator test data to the Waterford 3 separators and debris loading.

WF3 Response 3.m.2 Blockage of components was addressed above; wear of close tolerance components and systems is addressed in this paragraph. Calculation 2005-12840 (Reference 32) primarily addressed component wear; however, it also included instrument lines, relief valves, piston check valves and post accident sampling system components for the potential for blockage due to debris. For equipment addressed by WCAP-16406-P, Revision 1, August 2007, the methods and acceptance criteria were in accordance with the WCAP.

The wear analysis in calculation 2005-12840 concluded that wear was acceptable as it resulted in negligible flow effects based on WCAP-16406-P acceptance criteria with the exception of the HPSI pumps. The analysis only determined that worn condition of the pump and generated performance curves at various time points during the 30 day mission time. The worn performance curves were inputted into the Waterford 3 Long Term Cooling (LTC) Analysis of Record (AOR) to determine acceptability. The result from the LTC AOR evaluation is that adequate core cooling is maintained and the AOR remains valid. The worn condition, wear ring clearances, of the pump were used in a rotor-dynamic analysis to confirm that the pump remains dynamically stable thorough the mission time.

WF3 Response 3.m.3 The methods of WCAP-16406-P were used with interpretations of the November 2007 draft of the SER to the WCAP and with interpretations described during the September 2007 training teleconference. Calculation 2005-12840, Revision 1 used some more detailed methods where additional quantification was required.

Section 5 of WCAP-1 6406-P describes a methodology for calculating debris depletion over time.

TheWCAP also provides values of depletion coefficients by way of example. The WCAP does not provide specific depletion coefficients. Based on flow rates, volumes and settling velocities at Waterford. 3, plant specific depletion coefficients were calculated. These depletion coefficients also credited filtration of particulates as well as fibers on the sump screen where such filtration is supported by plant specific testing.

WCAP-16406-P, Revision 1 provides information on size distribution and settling fraction of coatings. It states that qualified coatings fail as 10 micron particles. This is conservative for pressure drop calculations, but not for downstream calculations. The Waterford 3 specific evaluation used a larger size particle based on vendor information about size of pigments in the coatings. This results in more calculated wear and is conservative. WCAP-16406-P assumes that unqualified coatings larger than 100 microns will settle. The NRC has questioned the "Stoke's Law" models used in such evaluations. The Waterford 3 calculation uses an empirical correlation for friction factor and benchmarks the resulting settling size against NRC-sponsored settling tests. Because the paint chips were all assumed to settle with the widest cross section perpendicular to the direction of settling, the calculation showed a larger settling size for a given paint chip and settling velocity. This results in a conservative, benchmarked, plant-specific settling size for particulates.

to W3F1-2008-0069 Page 60 of 87 A pump curve after wear was calculated for each Waterford 3 ECCS pump rather than utilizing WCAP Figure 8.1-3. The curve in the WCAP is based on a single stage pump with a particular specific speed and does not bound the calculated wear effect for multi-stage high head, low flow pumps likes the High Pressure Safety Injection pump. The more conservative method was used in 2005-12840, Revision 1. WCAP-16406-P recommends a minimum friction factor for maximizing the packing wear.

WCAP-16406-P, Revision 1, Appendix 0, Section 2.3 recommends an assumed friction factor of 0.01 to maximize wear. During the performance of the calculation it was found that the rate of wear, measured as gap.increase, would be maximum when the combination of parameters, friction factor times bearing length divided by clearance, was set equal to 2/3. Since this can be demonstrated mathematically it is no longer necessary to make an assumption about the friction factor in order to maximize the wear.

Entergy understands that Section 7.2 and 8.1.3 of the WCAP and the draft SER mean that if debris laden fluid is piped from the recirculation stream to flush a pump's seal then the primary seal would fail as a direct consequence of the postulated LOCA. That would constitute a common mode failure and all such pump seals would fail concurrently during the recirculation phase of the postulated LOCA. Conversely, if fluid from the recirculation stream is not piped to a pump's seal then there is no credible source of debris to fill the seal chamber and the primary pump seal is not assumed to fail as a direct consequence of the postulated LOCA. Such seals would still be subject to a postulated random failure of the pressure boundary as a moderate or high energy line break. The applicable requirements of SRP 15.6.5 as committed to in the USAR would remain applicable. For future reference, the leakage rate through pump seal one-half hour after a postulated primary seal failure was calculated. This calculation included the affects of wear on the components in the seals that would remain intact after a primary seal failure.

Rounding the inlet to an orifice in conjunction with increasing the orifice diameter decreases the-flow resistance more than just increasing the diameter. In order to account for the effects of rounding the inlet of an orifice by debris, Section 8.4 of WCAP-16406-P, Revision 1 recommended a formula taken from the first edition of Idelchik's "Handbook of Hydraulic Resistance". The first edition, translated from Russian in the 1960's has been updated and the corresponding formula from the third edition of Idelchik's "Handbook of Hydraulic Resistance" is used.

WF3 Response 3.m.4:

Those ECCS components and systems that are required to operate and pass debris laden fluid during the recirculation phase of recovery from a postulated LOCA have been identified. These ECCS components have been evaluated for blockage and wear from debris that would pass through the new containment sumps screens. The ECCS equipment at Waterford 3 would remain capable of passing sufficient flow to the reactor to adequately cool the core during the recirculation phase of a postulated LOCA.

WF3 Response 3.m.5:

At this time no operational changes have been made nor have any been identified for Waterford 3.

to W3F1-2008-0069 Page 61 of 87 3.n Downstream Effects - Fuel and Vessel NRC Issue 3.n The objective of the downstream effects, fuel and vessel section is to evaluate the effects that debris carried downstream of the containment sump screen and into the reactor vessel has on core cooling.

Show that the in-vessel effects evaluation is consistent with, or bounded by, the industry generic guidance (WCAP-16793), as modified by NRC 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.

WF3 Response 3.n The in-vessel effects evaluation was performed in accordance with the guidance in WCAP-16793-NP (Reference 60) and the initial NRC comments provided related to use of that document dated 2/4/08 (Reference 61). This evaluation did not indicate problems with reactor core cooling. The fuel deposit analysis was performed per the WCAP-16793 spreadsheet with conservatively bounding inputs relative to the maximum debris loading conditions for the plant.

This analysis determined that significant margin exists relative to the acceptance criteria, with total deposition thickness of <13 mils, remaining well below the 50-mil maximum value and the maximum clad temperature of <328°F also remaining well below the 800'F acceptance criteria.

The initial NRC comments provided for WCAP-16793 have been withdrawn and the WCAP is currently in revision, although the source of the revision is understood to be related to the fuel blockage analysis, not the fuel deposit methodology. Following the issuance of the revised guidance, further analysis could be necessary.

3.o Chemical effects NRC Issue 3.o 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.

2. Content guidance for chemical effects is provided in Enclosure 3 to a letter from the NRC to NEI dated September 27, 2007 (ADAMS Accession No. ML0726007425).

WF3 Response 3.o Chemical precipitates that form in the post-LOCA containment environment combined with debris do not result in an unacceptable head loss. Head loss due to chemical precipitates and debris is demonstrated by test using WCAP-16530-NP (Reference 58) methods with relatively minor modifications.

to W3F1-2008-0069 Page 62 of 87 The Alion Science and Technology 30-day integrated chemical effects testing identified that calcium phosphate precipitates can form very early post-LOCA due to the very low and retrograde solubility of calcium phosphate (lower solubility at high temperature). The 30-day integrated testing also identified that aluminum based precipitants do not form until the post-LOCA environment has cooled to below 140 degrees F. The prototype testing used these results to sequence the WCAP-16530-NP/16785-NP (References 58 and 53) based precipitates. Head loss calculations used the head loss attributed to calcium phosphate to determine the head loss across the strainer at temperatures greater than 140 degrees F when the NPSH margin is limiting. The 30 day integrated testing and analyses concluded that no aluminum based precipitates would form in the Waterford 3 environmental conditions with a pH less than 8.1; therefore any reduction in the aluminum oxy-hydroxide precipitate is reasonable.

3.p Licensing Basis' NRC Issue 3.p The objective of the licensing basis section is to provide information,regardingany changes to the plant licensing basis due to the sump evaluation or plant modifications.

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.

GL 2004-02 Requested Information Item 2(e)

A general description of and planned schedule for any changes to the plant licensing bases resulting from any analysis or plant modifications made to ensure compliance with the regulator requirements listed in the Applicable Regulatory Requirements section of this GL. Any licensing actions or exemption requests needed to support changes to the plant licensing basis should be included.

WF3 Response 3.P Major changes that have been made to the Waterford 3 Licensing Basis to meet compliance with the Generic Letter include modification to the Safety Injection Sump Strainer, relocation of the Trisodium Phosphate (TSP) baskets and revision to the Safety Injection Sump NPSH parameters.

Updated Final Safety Analysis Report (UFSAR) changes were issued to document the changes in the available head loss for both the HPSI pumps and the CS pumps. Changes were also issued against the FSAR to state that Waterford 3 is in compliance with the requirements of GS1-191 and Generic Letter 2004-02.

to W3F1-2008-0069 Page 63 of 87 Open Items 01.1 The licensee should justify its assumption of a 2D zone of influence for the Waterford 3 metal encapsulatedinsulation fiberglass.

01.1 Response The debris generation calculation 2004-07780 (Reference 28) was revised to use a 4D ZOI for the MEI insulation instead of the 2D ZOI originally used. The justification for this is presented in Alion Report ALION-REP-ENTG-4771-02 (Reference 39). The design of the RMI and MEI insulation cassettes are comparable. The filler material in either one does not contribute to the strength of the insulation system. Therefore, the ZOI for the Transco MEI cassettes should be equal to the ZOI for Transco RMI (2D ZOI). However, to ensure conservatism, and since there is no specific destruction testing that has been performed for the MEI, a 4D ZOI is used for Transco MEI such as used in Waterford 3.

01.2 The licensee should provide comprehensive documentation of the characteristics (macroscopic densities, microscopic densities, and characteristic debris sizes) of the actual plant debris at Waterford 3 and compare these characteristicsto the surrogate debris properties used for head loss testing, justifying any differences.

01.2 Response This is addressed in section 3.h of this supplemental response.

01.3 The licensee should provide an analysis that shows that the coating debris test data credited by Waterford 3 was generated using coating chips that are representative of or bounding with respect to the plant-specific failed coating chips.

Ol.3.Response This is addressed in section 3.h of this supplemental response.

01.4 The licensee should (1) justify that the percentage of debris transporting along the containment floor from the east and west sides of containment is equal to the percentage of flow approaching the sump from the east and west sides of containment and (2) provide a clear definition of the startingpoints for debris transportpaths to avoid contributing to an underestimation of debris transport on the side of containment opposite the break.

01.4 Response The transport calculation 2005-05500 (Reference 29) treats the Waterford 3 containment as uncompartmentalized. All LOCA generated debris is conservatively modeled as falling to the floor.

The flow around each side of containment is used to appropriately apportion the debris to the east or west sides of containment. Immediately after a break occurs, water spills from the break to the floor and the initial water from the break spreads across the floor.

The wave created by the initial water spreading is expected to have a high enough velocity to push debris away from the break, towards the perimeter of containment.

to W3F1-2008-0069 Page 64 of 87 Once the debris has been moved towards the containment perimeter and recirculation flow has been established, the flow distribution around each side of containment will dominate transport.

01.5 The licensee should explain how it has addressed the following four deficiencies in the existing transport analysis for unqualified coating chips: (1) lack of adequate data to justify the assumed size distribution for failed coating chips, (2) improper application of settling data for coating chips with a 400-micron thickness to particle-like coating debris with a 400- micron diameter, (3) lack of justification of the use of an analysis intended for the vertical flow conditions typical of a reactor vessel core inlet plenum for the horizontal flow conditions in the Waterford 3 containment pool, and (4) lack of considerationof the possibility that coating chips that fall into the containment pool in the vicinity of the sump may transportto the sump in suspension in the containment pool prior to settling on the containment floor.

01.5 Response

1) In accordance with NEI Guidance 04-07 (Reference 2) and the SER for NEI 04-07 (Reference 3), all qualified coating debris and unqualified zinc coating debris and indeterminate coating debris are treated as particulate and are therefore transported entirely to the sump strainer. Degraded qualified coatings are considered to fail as chips with a size distribution per ALION-REP-TXU-4464-02 and letter TXX-07156 (References 25 and 26). See Waterford 3 Response 3.h.6 for more details.

2) In calculation 2005-05500 (Reference 29) only degraded qualified coatings are considered to fail as coating chips. Indeterminate coatings are considered to fail as particulates. The only sources of degraded qualified coatings within the Waterford 3 containment are the containment dome and the containment liner. The coating system used on the containment dome is Carboline Phenoline 305 applied as a topcoat over Carboline Carbo-Zinc 11 primer. This system is also one of four Service Level 1 coating systems approved for the containment liner. The other systems include Carboline 305 topcoat over Carboline 191 primer, Carboline 801 as a primer and topcoat and Amerlock 400 NT (a one coat system). In the CCCL Letter dated 9/20/07 (Reference 27), it is confirmed that all are epoxy systems, which are expected to exhibit failure characteristics of Carboline 305. The CCCL Letter dated 9/20/07 (Reference 27) also confirms.that the size distribution presented in ALION-REP-TXU-4464-02 and letter TXX-07156 (References 25 and 26) is applicable to the Carboline Phenoline 305 coatings used at Waterford 3.

NUREG/CR-6916 (Reference 19) presents transport velocities for coatings with the size distributions presented in ALION-REP-TXU-4464-02 and letter TXX-07156 (References 25 and 26). The data reported in NUREG/CR-6916 (Reference 19) are for the failure characteristics of many coatings including epoxy applied as a topcoat over inorganic-zinc. The epoxy top coats applied over a zinc rich primer, Category ZE, are similar to the Carboline Phenoline 305 applied as a topcoat over Carboline Carbo-Zinc 11 used at Waterford 3.

to W3F1-2008-0069 Page 65 of 87

3) Calculation 2005-05500 (Reference 29) was revised and no longer uses the analysis intended for a reactor vessel core inlet plenum to apply to the horizontal flow conditions in the Waterford 3 containment pool. Instead, the calculation utilizes the data in ALION-REP-TXU-4464-02, letter TX-07156, and the CCCL letter dated 9/20/07 (References 25, 26, and 27) for the horizontal flow conditions at .Waterford.

4) The transport calculation 2005-05500 (Reference 29) accounts for coating chips that may land in the containment pool near the strainers and may not have a chance to settle before transporting to the strainer. In the following figure the pool area is shaded and the area considered near the strainer is cross hatched:

i ,. A <,, .K( ., ":::

The cross hatched area is conservatively approximated as 20% of the pool area. As 20%'of coating chips generated are expected enter the pool near the sump strainer, 20% of the -coating chip debris is considered to experience 100% transport to the strainer.

01.6 The licensee should justify that the debris transport fractions in the transportcalculation are representativeof the replacementstrainerconfiguration.

01.6 Response The transport calculation 2005-05500 (Reference 29) now includes several additional Computational Fluid Dynamics (CFD) models. These models were set upto verify the flow conditions in containment with the new strainer modules installed, and the TSP baskets no longer being installed directly adjacent to the strainers. The calculation 2005-05500 (Reference 29) documents the flow velocity, which directly affects the movement of the debris. With these new CFD models, the debris transport fractions are representative of the replacement strainer configuration.

to W3F1-2008-0069 Page 66 of 87 01.7 The licensee should provide results of analysis of the potential-effects of a low-pressure safety injection pump failure to stop on a recirculationactuationsignal.

01.7 Response There is a potential for a LPSI pump to fail to trip for the first 30 minutes of recirculation.

The failure of the LPSI pump to trip would result in a strainer flow rate that is non-conservatively larger than the flow rate that has been analyzed. While the head loss with the LPSI pump in operation was not determined by testing or analysis, GENE-0000-0053-4416-P (Reference 36) documents that the number of plant sump water turnovers in the time that the LPSI pump operates during recirculation time would be small enough that relatively little debris would get to the sump strainers.

Waterford 3 has a minimum sump water volume of about 46,335 cubic feet and a flow rate with the LPSI pumps of 12120 gpm, for a turnover time of 28.6 minutes; the plant water volume would experience approximately one turnover for the 30 minutes of LPSI trip failure. The head loss test performed had a flow rate of 364 gpm and a volume of 196 cubic feet based on a pool size of 123 inches by 72 inches and a water depth of 38.25 inches (Test specification 26A6833 - Reference 33) for a turnover time of four (4) minutes.

A review of the test head loss vs. time curves show the head loss during the first four minutes (one turnover) of testing after debris was added, was significantly lower than the stabilized head loss, i.e. more than one test turnover was required to generate maximum head loss. It can be concluded that the debris bed had not formed within one turnover in the test, nor is the debris bed expected to have formed within one turnover in the plant.

The LPSI pumps will be secured within 30 minutes of the start of recirculation and before the debris bed has formed on the strainer. Therefore, the failure of a LPSI pump to trip upon a RAS signal will not result in any debris bed formation or significant impact on head loss for the sump strainers. Clean strainer head loss is scaled for the added flow in the NPSH determinations. For added assurance that a LPSI failure to trip would not result in failure of the sump or any other pump, an NPSH evaluation has been performed using a sump temperature of 190F, based on sump temperature profiles in section 3.g.

The results of this evaluation can be found in section 3g. 16.

01.8 The licensee should describe how it has implemented prototypically fine fibrous debris preparationin its head loss testing.

01.8 Response For follow-up thin-bed testing, fiber was shredded five times in sequence, resulting in significantly reduced fibrous debris clump size that is more representative of small fines for thin-bed tests. Fiber sizes are generally small clumps of fiber or individual fibers which are representative of eroded fibers..

to W3F1-2008-0069 Page 67 of 87 01.9 The licensee should describe and justify how it has conducted adequate testing to determine thin bed peak head losses.

01.9 Response For follow-up design basis testing, the scope of tests was expanded so that the following fiber thicknesses are included in testing: 0.125", 0.25", 0.5", 0.75", and 1" (gap-filled), to ensure that any localized peaks due to thin-bed effect would be discovered.

01.10 The licensee should provide the results of assessment of the potential for non-prototypical settling and non-prototypical bed formation due to debris agglomeration during partially stirredstrainertesting.

01.10 Response Module testing for the partially stirred circumscribed bed cases was performed by creating a small-scale plant mock-up with one sump strainer module. The mockup included the plenum and one test module with a width and length matching the plant strainer design (40-inches X 40-inches) but with 10 discs rather than the plant design of 17 discs. The plenum height and distance of the strainer above the plenum were each scaled based on the scaled module height. Walls to the back and sides of the test module were set 16-inches from the strainer to match the plant strainer spacing of 32 inches.

The flow rate of the module was increased relative to the flow rate scaled by the circumscribed area of the module to match the average bulk approach velocity created by a plant strainer and thus accurately model the near field effects around the strainer.

The flow rate was then increased by 5% to add margin for expected flow measurement accuracy.

The measured flow rate accuracy for the module test configuration was 1%. The difference in the expected flow measurement accuracy and the actual flow measurement accuracy resulted in a test flow rate that was higher than the scaled plant flow rate. The additional head loss caused by this higher flow rate due to instrument accuracy (1%

actual vs. 5% anticipated) was removed by the scaling methodology. However, the higher flow rate also increased near-field debris transport and debris bed compression.

These effects were not removed and add conservatism.

0l. 11 The licensee should describe and justify how it has resolved the potential for non-'

prototypical flows during module testing due to "solid modeled" trisodium phosphate baskets located nearthe strainermodules.

0111 Response Waterford 3 ran new six (6) new head loss tests with the simulated baskets completely removed from the tank. These newer tests mirror the current installed condition of the SI Sump strainer system with the TSP baskets removed.

to W3F1-2008-0069 Page 68 of 87

01. 12 The licensee should provide the results of evaluation of the potential for and effects of vapor flashing due to strainerhead loss being greaterthan the strainersubmergence.

01.12 Response As stated in section 3.f.2, the sump strainers are submerged by a minimum of about 8 inches for a Large Break LOCA. The maximum head loss determined in section 3.f.10 is 0.517 ft (6.204 inches) with pump run out flows. Sump submergence is greater than maximum head loss therefore flashing will be prevented.

01.13 The licensee should explain how the following four additionalwater holdup mechanisms are modeled in the analysis of minimum containment pool water level: (1) water holdup due to condensation films, (2) water holdup due to spray droplet holdup in the containment atmosphere (as opposed to water vapor holdup in the containment atmosphere), and (3) refill of the reactor pressure vessel with colder and therefore denser water, and (4) the reactor water safety pool water temperature specified to be at the warmer normal operatingcontainment temperature.

01.13 Response Calculation MNQ6-4 (Reference 37) has been revised to include the water holdup due to condensation films, spray droplet holdup in the containment atmosphere, the impact of refilling the reactor coolant system with cooler water, and maximum allowed Refueling Water Storage Pool temperature.

1) Calculation MNQ6-4 determined that a condensation film thickness of 28 mils is appropriate for heat sink surfaces inside containment. However, the film thickness is assumed to be 35 mils. The total containment heat sink area is based on the passive heat sink area used in current containment analysis for Waterford 3.

2) Calculation MNQ6-4 included both water vapor holdup in the containment atmosphere, and water holdup due to spray droplets in the containment atmosphere.

3) Cooling of the RCS will cause the fluid contained within to contract as its density increases. For conservatism, calculation MNQ6-4 does not take credit for any RCS fluid in the water level calculations and assumes that contraction of the RCS is compensated by the lowering of pressurizer and steam generator levels.

4) The temperature of the water in the RWSP is assumed to be at the maximum of 100 F allowed by the Waterford 3 Technical Specifications.

01. 14 The licensee should justify treating unqualified coatings debris characteristics in the same manner as for qualified coatings.

01.14 Response The unqualified coatings at Waterford Unit 3 fall into two categories; degraded qualified coatings and indeterminate coatings. For the indeterminate coatings, the specific coating system applied is not known so these coatings are conservatively treated as 100% small fines and as such experience 100% transport to the sump.

Degraded qualified coatings are found on the containment dome and the containment liner. The painting system used on the containment dome is Carboline Phenoline 305 applied as a topcoat over Carboline Carbo-Zinc 11 primer. This system is also one of to W3F1-2008-0069 Page 69 of 87 four Service Level 1 paint systems approved for the containment liner. The other systems include Carboline 305 topcoat over Carboline 191 primer, Carboline 801 as a primer and topcoat and Amerlock 400 NT (a one coat system). All are epoxy systems, which are expected to exhibit failure characteristics of Carboline 305. In the CCCL letter dated 9/20/07 (Reference 27), it is confirmed that the size distribution presented in ALION-REP-TXU-4464-02 and letter TXX-07156 (References 25 and 26) is applicable to the Carboline Phenoline 305 coatings used at Waterford 3, except for one coating system possibly used on the containment liner.

NUREG/CR-6916 (Reference 19) presents transport velocities for coatings with the size distributions presented in ALION-REP-TXU-4464-02 and letter TXX-07156 (References 25 and 26). The data reported in NUREG/CR-6916 (Reference 19) are for the failure characteristics of many coatings including epoxy applied as a topcoat over inorganic-zinc. The epoxy top coats applied over a zinc rich primer, Category ZE, are similar to the Carboline Phenoline 305 applied as a topcoat over Carboline Carbo-Zinc 11.

Debris characteristics used for the unqualified coatings in containment are based on testing results found in NUREG/CR-6916, ALION-REP-TXU-4462-02, and letter TXX-07156 (References 19, 25, and 26). The debris characteristics used can be found in section 3h of the Generic Letter responses.

01.15 The licensee should summarize how it has addressedthe following aspects of structural analysis for the new strainer.

Part 1 - The licensee should revise the high-energy line break report to provide definitive statements in the conclusions concerning pipe whip and missile impacts to the new strainerassembly and clarificationof the bases for those conclusions.

Part 2 - The licensee should revise the sump strainer design specification to clearly identify the damping to be 2%.

Part 3 - The licensee should correct errors and discrepancies in the sump strainer design specification and stress analysis report.

Part 4 - The licensee should correct the temperature delta in the sump strainer stress analysis report.

Part 5 - The licensee should clarify the sump strainer acceleration values in the sump strainerstress analysis report.

Part 6 - The licensee should correct the values for E*/E and K in the sump strainerstress analysis report.

Part 7 - The licensee should correct the whole strainerand plenum maximum deflection values in the stress analysis report.

Part 8 - The licensee should correct stress limits, safety factor values and certain units used for stress limits in certain tables of the sump strainerstress analysis report.

to W3F1-2008-0069 Page 70 of 87 Part 9 - The licensee should address discrepanciesidentified in the hydrodynamic mass analysis.

01.15 Response - Part 1 Section 3.5 of the UFSAR documents the evaluation of missiles inside and outside the Waterford 3 RCB. Table 3.5-4 of the UFSAR contains a list of the potential missiles inside containment. None of these are located in an area where they could impact the new strainer assembly.

The closest high energy line breaks are located over 20 feet from the closest strainer.

Therefore, there is no possibility of impacting the strainers with a whipping pipe.

The HELB report (GENE-0000-0048-9192) has been revised to provide definitive statements in the conclusions concerning pipe whip and. missile impacts to the new strainer assembly.

0115 Response - Part 2 The curves in the specification have a designation on the lower right of the curve that states these are 2% damping.

01.15 Response - Part 3 The sump strainer design specification, GE Design Specification 26A6870, and the stress analysis report, GENE-0000-0054-9349, have been revised to correct errors and discrepancies.

01.15 Response - Part 4 The design specification (GE 26A6870) and the stress analysis report (GENE-0000-0054-9349) have both been revised to address the temperature delta in the sump strainer stress analysis report.

The questions pertaining to the temperature delta did not result in significant changes since the equipment is constructed of a single material, austenitic stainless steel.

Therefore, there are no significant differential thermal expansions within the structure and no thermal stresses would be developed.

01.15 Response - Part 5 The inertial acceleration values used in the analyses were extracted directly from the Design Envelope spectra contained in the design specification. However, the values reported in the stress report are the ANSYS input values adjusted to account for hydrodynamic mass and debris load to facilitate the ANSYS analyses. Since hydrodynamic mass and debris load are also reported in these same tables, the reader cannot ascertain what was actually used in the analysis. Therefore, the stress report (GENE-0000-0054-9349) was revised to reflect the seismic accelerations specified in the design specification.

01.15 Response - Part 6 The values for E*/E and K in the revision of the stress analysis report reviewed by the staff during the NRC GSI-191 audit at Waterford 3 were incorrect. A review of the displacement numbers showed that the calculation E*/E is 0.43 and K = 2.33. Since the original values used were incorrect, the stresses were reevaluated with the new and to W3F1-2008-0069 Page 71 of 87 correct values. All safety factors will remain above 1.0 even with the increased and correct values. These new and corrected values were included in the latest revision of the stress report, GENE-0000-0054-9349.

01.15 Response - Part 7 The deflection results in the reviewed version of the stress report (GENE-0000-0054-9349) were provided to assist the review in understanding the behavior of the structure, and had no significance beyond that. The latest revision of the stress report (GENE-0000-0054-9349) corrected the labels.

01.15 Response - Part 8 The errors in the stress report (GENE-0000-0054-9349) 'Were revised in the latest revision. There was no adverse impact on the structural adequacy conclusions.

01.15 Response - Part 9 The discrepancies in the hydrodynamic mass analysis have been corrected. The corrections are in the latest revision to GE calculation GENE-0000-0054-9349 (Reference 56). There was no adverse impact on the structural adequacy conclusion.

01. 16 The licensee should summarize how it has evaluated the potential for holdup in the refueling cavity due to falling debris.

01.16 Response Calculation 2005-05500 (Reference 29) documents that the refueling cavity has two 6-inch drain lines (without screens) that drain to the containment floor, and by one 4-inch line that drains to the containment sump. In the event that large debris is propelled over the SG cavity walls into the refueling cavity, the debris must land on a drain in order to clog it since the velocities in the cavity are too low to transport a large piece of debris to a drain. During plant operations, the Upper Guide Structure Lift Rig (UGSLR) is stored directly above one of the 6" drains. The UGSLR will prevent any debris larger than 6" from falling directly on the drain. The diver stairs are located above the other 6" drain and are permanently mounted in the refueling cavity. These stairs will prevent any debris larger than 6" from falling directly onto the drain. Any smaller debris that transports to these two 6" drains will pass through the drains since they do not contain screens. Therefore, there will always be drainage available from the refueling cavity to the active pool.

01.17 The licensee should provide the results of a similitude evaluation for WCAP-16406-P versus conditions at Waterford 3.

01.17 Response In response to the NRC Safety Evaluation for WCAP-16406-P, the Waterford 3 Downstream Effects analysis specifically addresses each of the 31 limitations identified in the Safety Evaluation. The responses to these limitations justify the use of WCAP-16406-P at Waterford 3. Below are the 31 limitations with corresponding responses from calculation 2005-12840 Revision 1 (Reference 32).

to W3F1-2008-0069 Page 72 of 87

1. Where a TR WCAP-16406-P, Revision 1, section or appendix refers to examples, tests, or general technical data, a licensee should compare and verify that the information is applicable to its analysis.

In general, examples were not used for site specific input. The wear equations developed in the WCAP were developed and benchmarked on equipment and with debris similar to that found at Waterford Unit 3.

2. A discussion of EOPs, AOPs, NOPs or other plant-reviewed alternate system line-ups should be included in the overall system and component evaluations as noted in the NRC staff's SE of NEI 04-07, Section 7.3 (Reference 3).

The scope of'equipment to be reviewed for wear and blockage was defined in evaluation 2005-02820, GSI-191 Downstream Effects - Flow Clearances, Revision 0 (Reference 31). . That calculation identified the equipment that could be in the recirculation path following a postulated accident.

3. A licensee using TR WCAP-16406-P, Revision 1, will need to determine its own specific sump debris mixture and sump screen size in order to initiate the evaluation.

Site specific debris generation and transport calculations (Reference 28 and 29) were completed and referenced by this wear calculation as the source of debris.

Screen information was taken from the site specific design documents defining the screens and from site specific debris bypass testing.

4. TR WCAP-16406-P, Revision 1, Section 4.2, provides a general discussion of system and component mission times. It does not define specific times, but indicates that the defined term of operation is plant-specific. As stated in the NRC staffs SE of NEI 04-07, Section 7.3 (Reference 3), each licensee should define and provide adequate basis for the mission time(s) used in its downstream evaluation.

Site specific design and licensing information is used to determine the applicable mission time of equipment evaluated in this calculation.

5. TR WCAP-16406-P, Revision 1, Section 5.8, assumes that the coolant which is not spilled flows into the reactor system and reaches the reactor vessel downcomer. This would be true for most PWR designs except for plants with UPI. Therefore, the methodology of Section 5.8 may not be applicable to plants with UPI and its use should be justified on a plant-specific basis.

The Waterford Unit 3 station utilizes lower plenum injection.

6. TR WCAP-16406-P, Revision 1, Section 5.8, provides equations which a licensee might use to determine particulate concentration in the coolant as a function of time.

Assumptions as to the initial particulate debris concentration are plant-specific and should be determined by the licensee. In addition, model assumptions for ECCS flow rate, the fraction of coolant spilled from the break and the partition of large heavy particles which will settle in the lower plenum and smaller lighter particles which will not settle should be determined and justified by the licensee.

to W3F1-2008-0069 Page 73 of 87 Debris depletion in this calculation is based on plant specific flows, debris types and debris size distributions. The debris depletion methodology is described in Appendix A and the settling size is calculated in Appendix C.

7. TR WCAP-16406-P, Revision 1, Sections 5.8 and 5.9, assumes that debris settling is governed by force balance methods of TR Section 9.2.2 or Stokes Law. The effect of debris and dissolved materials on long-term cooling is being evaluated under TR WCAP-16793-NP (Reference 12). If the results of TR WCAP-16793-NP show that debris settling is not governed by force balance methods of TR Section 9.2.2 or Stokes Law, then the core settling term determined from TR WCAP-16793-NP should be used.

The site specific debris settling size is determined in Appendix C. The methodology uses empirical friction factors based on the debris shape. This methodology is benchmarked against the NRC-sponsored testing of paint chip settling reported in NUREG/CR-6916.

8. TR WCAP-16406-P, Revision 1, Section 7ý2, assumes a mission time of 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> for pump operation. Licensees should confirm that 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> bounds their mission time or provide a basis for the use of a shorter period of required operation.

A 720 hour0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> mission time is used at Waterford Unit 3. This is in accordance with NRC SER to NEI 04-07 and the plant licensing documents.

9. TR WCAP-16406-P, Revision 1, Section 7.2, addresses wear rate evaluation methods for pumps. Two types of wear are discussed: 1) free-flowing abrasive wear and 2) packing-type abrasive wear. Wear within close-tolerance, high-speed components is a complex analysis. The actual abrasive wear phenomena will likely not be either a classic free-flowing or packing wear case, but a combination of the two. Licensees should consider both in their evaluation of their components.

This calculation considers the maximum of the calculated free-flow or packing type wear until a gap of 4 times the original gap is reached. Beyond that point, free-flow wear is modeled.

10. TR WCAP-16406-P, Revision 1, Section 7.2.1.1, addresses debris depletion coefficients. Depletion coefficients are plant-specific values determined from plant-specific calculations, analysis, or bypass testing. Licensees should consider both hot-leg and cold-leg break scenarios to determine the worst case conditions for use in their plant-specific determination of debris depletion coefficient.

Both hot-leg and cold-leg break scenarios are considered in this calculation in the determination of the debris concentrations. The wear calculation is based on the larger of the two debris concentrations.

11. TR WCAP-16406-P, Revision 1, Section 7.3.2.3, recognizes that material hardness has an effect on erosive wear. TR WCAP-16406-P, Revision 1, suggests that "For elastomers, the wear rate is at least one order of magnitude less than steel.

Therefore, for soft-seated valves, divide the estimated wear rate of steel from above equations by 10 per Appendix F." The NRC staff agrees that the wear rates of to W3F1-2008-0069 Page 74 of 87 elastomers are significantly less than for steels. However, the wear coefficient should be determined by use of a suitable reference, not by dividing the steel rate by a factor of 10.

Wear of elastomeric materials is not included in this calculation.

12. TR WCAP-16406-P, Revision 1, Section 8.1.1.2, "Evaluation of ECCS Pumps for Operation with Debris-Laden Water from the Containment Sump," states that "Sufficient time is available to isolate the leakage from the failed pump seal and start operation of an alternate ECCS or CSS train." Also, Section 8.1.3, "Mechanical Shaft Seal Assembly," states: "Should the cooling water to the seal cooler be lost, the additional risk for seal failure is small for the required mission time for these pumps."

These statements refer only to assessing seal leakage in the context of pump operability and 10 CFR Part 100 concerns. A licensee should evaluate leakage in the context of room habitability and room equipment operation and environmental qualification, if the' calculated leakage is outside that which has been previously assumed.

This calculation determines the seal leakage rate that would occur if the primary seal surfaces are assumed to fail. If the failure of the primary seals is a result of debris from the original break, then the primary seal leakage would be part of the design basis and considered in the appropriate flooding and HVAC analyses. However, the seals are not considered to fail as a direct result of the accident. Therefore, except as specifically required by SRP 15.6.5 for offsite dose calculation, a second passive failure in the ECCS system is not part of the design basis of this plant. A postulated passive failure of the primary seal in a single pump would be within the design basis, but would be bounded by a moderate energy line break in the pump room.

13. TR WCAP-16406-P, Revision 1, Section 8.1.3, discusses cyclone separator operation. TR WCAP-16406-P, Revision 1, generically concludes that cyclone separators are not desirable during post-LOCA operation of HHSI pumps. The NRC staff does not agree with this generic statement. If a licensee pump contains a cyclone separator, it should be evaluated within the context of both normal and accident operation. The evaluation of cyclone separators is plant-specific and depends on cyclone separator design and the piping arrangement for a pump's seal injection system.

There are cyclone separators in the ECCS system of this plant. These separators have not been tested. Therefore, further investigation as to the acceptability of these cyclone separators is required.

14. TR WCAP-16406-P, Revision 1, Section 8.1.4, refers to pump vibration evaluations.

The effect of stop/start pump operation is addressed only in the context of clean water operation, as noted in Section 8.1.4.5 of TR WCAP-16406-P, Revision 1. If an ECCS or CSS pump is operated for a period of time and builds up a debris "packing" in the tight clearances, stops and starts again, the wear rates of those areas may be different due to additional packing or imbedding of material on those wear surfaces.

Licensees who use stop/start operation as part of their overall ECCS or CSS operational plan should address this situation in their evaluation.

to W3F1-2008-0069 Page 75 of 87 This calculation utilizes the Archard's method in determining wear. No credit is taken for stai-t/stop operation to reduce the mission time. A packing wear model is used for the duration of the mission time or until a clearance of 4 times the original clearance is achieved.

15. TR WCAP-16406-P, Revision 1, Section 8.1.4, states: "should the multistage ECCS pumps be operated at flow rates below 40% of BEP during the containment recirculation, one or more of the pumps should be secured to bring the flow rate of the remaining pump(s) above this flow rate." The NRC staff does not agree with this statement. System line-ups and pump operation and operating point assessment are the responsibility of the licensee. Licensees must ensure that their ECCS pumps are capable of performing their intended function and the NRC has no requirements as to their operating point during the recirculation phase of a LOCA.

No credit is taken for securing ECCS pumps during the course of a postulated accident. In general shut-off head is used where maximum dP would result in conservative wear calculations. Run-out is used when minimal head or maximum flow would result in conservative calculation results.

16. TR WCAP-16406-P, Revision 1, Section 8.1.5, makes a generic statement that all SI pumps have wear rings that are good "as new" based solely upon "very little service beyond inservice testing." A stronger basis is needed to validate this assumption, if used (e.g., maintenance, test and operational history and/or other supporting data).

This calculation used in-service testing (IST) results to predict wear to the time when the pump degradation would be detected and corrected.

17. TR WCAP-1 6406-P, Revision 1, Section 8.3, identifies criteria for consideration of tube plugging. Licensees should confirm that the fluid velocity going through the heat exchanger is greater than the particle settling velocity and evaluate heat exchanger plugging if the fluid velocity is less than the settling velocity.

Flow velocity in the heat exchangers is calculated and evaluated for settling.

18. TR WCAP-16406-P, Revision 1, Section 8.6, refers to evaluation of instrumentation tubing and system piping. Plugging evaluations of instrument lines may be based on system flow and material settling velocities, but they must consider local velocities and low-flow areas due to specific plant configuration.

Instrument tubing in this analysis is evaluated based on the tubing being at or above the horizontal, rather than on velocity considerations.

19. TR WCAP-16406-P, Revision 1, Sections 8.6.7, 8.6.8, 8.6.9, and 8.6.10 describe, in general terms, the Westinghouse, CE, and B&W RVLIS. TR WCAP-16406-P, Revision 1, recommends that licensees evaluate their specific configuration to confirm that a debris loading due to settlement in the reactor vessel does not effect the operation of its RVLIS. The evaluation of specific RVLIS design and operation is outside the scope of this SE and should be performed in the context of a licensees reactor fuel and vessel evaluations.

to W3F1-2008-0069 Page 76 of 87 Reactor fuel and vessel evaluations are outside the scope of this calculation.

20. TR WCAP-16406-P, Revision 1, Section 8.7, refers to evaluation of system piping.

Plugging evaluations of system piping should be based on system flow and material settling velocities. Licensees should consider the effects of local velocities and low-flow areas due to specific plant configuration. A piping wear evaluation using the free-flowing wear model outlined in Section 7 should be performed for piping systems. The evaluation should consider localized high-velocity and high-turbulence areas. A piping vibration assessment should be performed if areas of plugging or high localized wear are identified.

The wear in high velocity areas such as orifices was calculated. The amount of wear was minimal and would bound the wear to other areas in the piping systems.

Therefore, numeric calculation of wear in general piping areas was not performed.

21. TR WCAP-16406-P, Revision 1, Section 9, addresses reactor internal and fuel blockage evaluations. This SE summarizes seven issues regarding the evaluation of reactor internal and fuel. The PWROG indicated that the methodology presented in TR WCAP-16793-NP (Reference 15) will address the seven issues. Licensees should refer to TR WCAP-16793-NP and the NRC staffs SE of the TR WCAP-16793-NP, in performing their reactor internal and fuel blockage evaluations. The NRC staff has reached no conclusions regarding the information presented in TR WCAP-16406-P, Section 9.

Reactor internal and fuel blockage is outside the scope of this calculation.

22. TR WCAP-16406-P, Revision 1, Table 4.2-1, defines a plant Category based on its Low-Head / Pressure Safety Injection to RCS Hot-Leg Capability. Figure 10.4-2 implies that Category 2 and 4 plants can justify LHSI for hot-leg recirculation.

However, these categories of plants only have one hot-leg injection pathway.

Category 2 and Category 4 plant licensees should confirm that taking credit for the single hot-leg injection pathway for their plant is consistent with their current hot-leg recirculation licensing basis.

Waterford Unit 3 is a Category 1 plant.

23. TR WCAP-16406-P, Revision 1, Appendix F, discusses component wear models.

Prior to using the free-flowing abrasive model for pump wear, the licensee should show.that the benchmarked data is similar to or bounds its plant conditions.

The free flow wear model is used in those pump area that were shown to have low wear in the Davis Besse pump wear testing. The low wear areas were areas where the branch of the flow stream carrying the debris had to turn 180 degrees and travel back toward the center of the pump while the main stream continued outward. These areas act as cyclone separators due to their geometry. The maximum of packing or free flow wear was used in other areas of the pumps.

The pumps gaps at Waterford 3 showed the same geometry as the benchmarked data. The flow paths through the pump gaps were analyzed and the maximum of packing or free flow wear was used when applicable.

to W3F17-2008-0069 Page 77 of 87

24. TR WCAP-16406-P, Revision 1, Appendix H, references American Petroleum Institute (API) Standard 610, Annex 1 eighth edition. This standard is for newly manufactured pumps. Licensees should verify that their pumps are "as good as new" prior to using the analysis methods of API-610. This validation may be in the form of maintenance records, maintenance history, or testing that documents that the as-found condition of their pumps.

This calculation used in-service testing (IST) results to predict wear to the time when the pump degradation would be detected and corrected.

25. TR WCAP-16406-P, Revision 1, Appendix I, provides guidelines for the treatment, categorization and amount of DBA Qualified, DBA Acceptable, Indeterminate, DBA Unqualified, and DBA Unacceptable coatings to be used in a licensee's downstream sump debris evaluation. A technical review of coatings generated during a DBA is not within the scope of this SE. For guidance regarding this subject see the NRC staffs SE of NEI-04-07 (Reference 3) Section 3.4 "Debris Generation."

Debris generation by debris type is not within the scope of this calculation.

26. TR WCAP-16406-P, Revision 1, Appendix J, derives an approach to determining a generic characteristic size of deformable material that will pass through a strainer hole. This approach is only applicable to screens and is not applicable to determining material that will pass through other close tolerance equipment.

Identification of close tolerance passages is not in the scope of this calculation. The criteria used in the phase I review of downstream components was that holes less than twice the screen hole size required further evaluation.

27. TR WCAP-16406-P, Revision 1, Appendix 0, Section 2.2, states that the wear coefficient, K, in the Archard Model is determined from testing. The wear coefficient (K) is more uncertain than the load centering approach and K may vary widely.

Therefore, licensees should provide a clear basis, in their evaluation, for their selection of a wear coefficient.

The wear coefficient used results in calculated wear greater than the amount seen in the Davis-Besse t testing. The materials, debris types and concentrations are comparable. Therefore, the k value presented the WCAP-1 6406-P appears to be the best conservative information available on ECCS pump wear when exposed to insulation and coating debris.

28. TR WCAP-16406-P, Revision 1, Appendix P, provides a method to estimate a packing load for use in Archard's wear model. The method presented was benchmarked for a single situation. Licensees are expected to provide a discussion as to the similarity and applicability to their conditions. The licensee should incorporate its own specific design parameters when using this method.

This calculation utilized the methodology discussed in Appendix 0 of WcAP 16406-P (centering load) for defining loads to be used in the packing wear model.

to W3F1-2008-0069 Page 78 of 87

29. TR WCAP-16406-P, Revision 1, Appendix Q, discusses bounding debris concentrations. Debris concentrations are plant-specific. If 9.02E-5 (mils/hr)/10 PPM is to be used as the free flowing abrasive wear constant, the licensee should show how it is bounding or representative of its plant.

The combined debris concentration of abrasive particulates and fibers used for free-flowing abrasive wear does not exceed 720 ppm. Therefore, the extrapolated wear rate [9.02E-5 (mils/hr)/10 ppm] is not required to be used in this calculation.

30. TR WCAP-16406-P, Revision 1, Appendix R, evaluates a Pacific 11-Stage 2.5" RLIJ pump. The analysis was performed by the PWROG using specific inputs. ECCS pumps with running clearance designs and dimensions significantly different than those covered by the analysis should be subjected to pump-specific analysis to determine the support stiffness based on asymmetric wear. If licensees use the aforementioned example, a similarity evaluation should be performed showing how the example is similar to or bounds their situations.

Themulti-stage HPSI pump was evaluated by finding the stiffness at the uniform increase in clearance equal to 2X as the as-new clearance. The stiffness of the pumps after normal wear and debris induced wear was considered and then calculated. The stiffness of the pump after normal and LOCA asymmetric wear was compared to the allowed stiffness equivalent to a uniform 2X initial clearance to judge the acceptability of the pump.

31. Licensees should compare the design and operating characteristics of the Pacific 2.5" RLIJ 11 to their specific pumps prior to using the results of Appendix S in their component analyses.

As stated in response 30 above, specific stiffness calculations were performed for all applicable pumps using a stiffness corresponding to 2X the as-new clearance as the acceptance criteria.

01.18 The licensee should provide the assumptions, the bases for assumption and the source documents for its downstream evaluation of components and systems.

01.18 Response Following the NRC Audit of Waterford 3's Generic Letter 2004-02 efforts, Waterford 3 re-performed its Downstream Effects evaluations in accordance with WCAP-16406-P and corresponding NRC Safety Evaluation. The preliminary Downstream Effects evaluation reviewed by the NRC during the Audit had not been accepted by Waterford 3 and was only provided by Waterford's vendor at the request of the NRC. Bases and source documents for all assumptions have been provided directly in the analysis or through supporting analyses.

to W3F1-2008-0069 Page 79 of 87 -

01.19 The licensee should provide clearly defined technical bases for the designated mission times for shutdown cooling, high pressure safety injection and containment spray.

01.19 Response A review of the Waterford 3 UFSAR, specifically Table 15.6-18, indicates that the analyzed mission time for the LOCA event is 30 days. During the 30 day mission time, two trains of High Pressure Safety Injection and Containment Spray are conservatively assumed to operate continuously.

01.20 The licensee should develop and justify conservative, b6ounding values for system lineups, fluid flows and system pressures for the downstream effects components and systems analysis.

01.20 Response The downstream analysis in calculation 2005-12840 revision 1 is based on the most conservative system lineup. Lineups were determined using plant operating procedures.

01.21 The licensee should justify the use of design curves, or re-analyze for degraded, actual or modified pump curves for the downstream effects components and systems analysis.

01.21 Response Calculation 2005-12840 revision 1 assumes pump performance is at its IST limits and/or utilizes IST test data to determine the amount of degradation expected over the life of the plant for the wear analysis starting point.

01.22 The licensee should provide the results of analysis of emergency core cooling system (ECCS) air entrainment (apart from vortexing) and the potential for waterhammer and slug flow.

01.22 Response The Waterford 3 ECCS is designed as a water solid system. Post refueling outage, ultrasonic testing is performed at potential void formation points in the ECCS system. If a void is found, the system is flushed eliminating the void. The system. is checked multiple times to verify that all voids are eliminated. Various waterhammer analyses have been performed on the ECCS system throughout the life of the plant. The modifications done to date for Waterford 3 do not affect any existing analysis nor create the potential for new waterhammer events.

01.23 The licensee should re-calculate downstream component wear due to strainer bypass debris and provide the results.

01.23 Response Calculation 2005-12840 Revision 1 has been revised to meet the requirements of WCAP-16406-P and NCR SER requirements. Debris loading used in the evaluation is based on site specific debris generation analysis, transport analysis, and bypass testing.

to W3F1-2008-0069 Page 80 of 87 01.24 The licensee should re-perform its high-pressure safety injection recirculation throttle valve clogging analysis considering the full range of possible recirculation throttle valve positions or failure of the HPSI recirculationthrottle valve to open to its pre-set position, and provide the results.

01.24 Response From Calculation 2005-02820 (Reference 31), the throttle valves are:

" HPSI injection header valves, which are Target Rock 2-inch motor operated globe valves, and

" RC loop hot leg valves, which are Anchor Darling 3-inch motor operated globe valves.

The above motor operated valves are used to balance the flow between the hot leg and cold leg and between the four cold leg injection lines. These valves are throttled from the control room and their flow rates are monitored during a LOCA. These valves can be throttled down to maintain the required flow in the event of excessive wear or opened as excessive resistance due to unexpected clogging is seen. Due to this, no wear calculations are performed.

01.25 The licensee should describe how it has incorporated actions of its operational procedures into the downstream effects evaluation.

01.25 Response The Waterford 3 downstream effect analysis evaluated the ECCS and CS systems in all configurations as defined by operational procedures. The Waterford 3 operational procedures do not require the securing of HPSI or CS pumps at any time, therefore these pumps were analyzed to operate for a full 30 day mission time. At 1-2 hours post RAS, the ECCS system switches from cold leg only injection to simultaneous cold and hot leg injection. Both configurations are considered and addressed in the downstream analysis. Conservative flows' were used for all components to bound any possible operating flow.

01.26 The licensee should justify emergency core cooling system (ECCS) pump wear rings to be "goodas new," or determine a more conservative condition for these rings.

01.26 Response See item 24 under Open Item 17.

01.27 The licensee should provide the results of evaluation of high-pressure safety injection (HPSI) pump stage-to-stage degradation and its effect on pump hydraulic performance, and should provide the results of a pump vibration and rotor dynamics evaluation.

01.27 Response Downstream Effects evaluation 2005-12840 Revision 1 determined that worn condition of the pump and generated performance curves at various time points during the 30 day mission time. The worn performance curves were inputted into the Waterford 3 Long to W3F1-2008-0069 Page 81 of 87 Term Cooling (LTC) Analysis of Record (AOR) to determine acceptability. The result from the LTC AOR evaluation is that adequate core cooling is maintained and the AOR remains valid. The worn condition, wear ring clearances, of the pump were used in a rotor-dynamic analysis to confirm that the pump remains dynamically stable thorough the mission time.

01.28 The licensee should summarize its evaluation of ECCS and CS pump leakage effects in its Safeguards Room.

01.28 Response Waterford 3 currently has no analysis for dose considerations in the Safeguards Room as there are no required operator actions in this area during a LOCA. Control Room dose analysis assumes at least 0.5 gpm total leakage from the Engineered Safety Feature pumps. The downstream analysis performed concluded that degradation of the pump seals is not expected therefore no additional leakage should occur.

01.29 The licensee should summarize how it has determined the effects of settled material at emergency core cooling system (ECCS) low points and integrate these effects into the downstream effects evaluation.

01.29 Response Flow velocities in the ECCS piping system remains relatively high such that settling of material that bypasses the strainers should be minimal and not affect system performance:.

01.30 The licensee should consider the results of the various component wear evaluationsand perform an overall system flow evaluation, and should provide a summary of the results.

01.30 Response The downstream effects calculation 2005-12480 Revision 1 concluded that all components with the exception of the High Pressure Safety Injection Pumps experienced negligible wear based on WCAP-16406-P acceptance criteria., Degraded performance curves were developed for the High Pressure Pumps. These degraded curved were utilized in a Long Term Cooling study to confirm their acceptability in maintain core cooling. Based on the analyses perform concluding that negligible wear occurs on the system components and that acceptability of the degraded High Pressure Safety injection pumps, no overall system flow evaluation was deemed necessary.

01.31 The. licensee should provide the results of an analysis of downstream effects of post-LOCA debris and chemicals on the fuel and vessel.

01.31 Response The in-vessel effects evaluation was performed in accordance with the guidance in WCAP-16793 and the initial NRC comments provided related to use of that document.

This evaluation did not indicate problems with reactor core cooling. The fuel deposit analysis was performed per the WCAP-16793 spreadsheet with conservatively bounding to W3F1-2008-0069 Page 82 of 87 inputs relative to the maximum debris loading conditions for the plant. This analysis determined that significant margin exists relative to the acceptance criteria, with total deposition thickness of <13 mils, remaining well below the 50-mil maximum value and the maximum clad temperature of <328°F also remaining well below the 800'F acceptance criteria. The initial NRC comments provided for WCAP-16793 have been withdrawn and the WCAP is currently in revision, although the source of the revision is understood to be related to the fuel blockage analysis, not the fuel deposit methodology.

Following the issuance of the revised guidance, further analysis could be necessary.

01.32 The licensee should provide the results of resolution of chemical effects at Waterford 3.

01.32 Response Chemical precipitates that form in the post-LOCA containment environment combined with debris do not result in an unacceptable head loss. Head loss due to chemical precipitates and debris is demonstrated by test using WCAP-16530-NP (Reference 58) methods with relatively minor modifications.

The Alion Science and Technology 30-day integrated chemical effects testing identified that calcium phosphate precipitates can form very early post-LOCA due to the very low and retrograde solubility of calcium phosphate (lower solubility at high temperature). The 30-day integrated testing also identified that aluminum based precipitants do not form until the post-LOCA environment has cooled to below 140 degrees F. The prototype testing used these results to sequence the WCAP-16530-NP/16785-NP (References 58 and 53) based precipitates. Head loss calculations used the head loss attributed to calcium phosphate to determine the head loss across the strainer at temperatures greater than 140 degrees F when the NPSH margin is limiting. The 30 day integrated testing and analyses concluded that no aluminum based precipitates would form in the Waterford 3 environmental conditions with a pH less than 8.1; therefore any reduction in the aluminum oxy-hydroxide precipitate is reasonable.

to W3F1-2008-0069 Page 83 of 87 References

1. NRC Generic Letter (GL) 2004-02, "Potential Impact of Debris Blockage on Emergency Recirculation during Design Basis Accidents for Pressurized-Water Reactors," dated September 13, 2004.

2. Nuclear Energy Institute (NEI) document NEI 04-07 Revision 0, December 2004, "Pressurized Water Reactor Sump Performance Evaluation Methodology."

3. Safety Evaluation by the Office of Nuclear Reactor Regulation Related to NRC Generic Letter 2004-02, Nuclear Energy Institute Guidance Report (Proposed Document Number NEI 04-07), "'Pressurized Water Reactor Sump Performance Evaluation Methodology,"

Issued December 6, 2004.

4. Regulatory Guide 1.82, "Water Sources for Long Term Recirculation Cooling, Following a Loss of Coolant Accident," Revision 3, November 2003.

5. NUREG-0800, "U.S. Nuclear Regulatory Commission Standard Review Plan," Section 3.6.2, "Determination of Rupture Locations and Dynamic Effects Associated with the Postulated Rupture of Piping,"Revision 1, July 1981.

6. NUREG/CR-2791, "Methodology for Evaluation of Insulation Debris Effects, Containment Emergency Sump Performance Unresolved Safety Issue A-43," Issued September 1982.

7. NUREG/CR-3616, "Transport and Screen Blockage Characteristics of Reflective Metallic Insulation Materials," January 1984.

8. NUREG/CR-6224, "Parametric Study of the Potential for BWR ECCS Strainer Blockage Due to LOCA Generated Debris, Final Report," Issued October 1995.

9. NUREG/CR-6369, "Drywell Debris Transport Study, Final Report," Volume 1, Issued September 1999.

10. NUREG/CR-6369, "Drywell Debris Transport Study: Experimental Work, Final Report,"

Volume 2, Issued September 1999.

11. NUREG/CR-6369, "Drywell Debris Transport Study: Computational Work, Final Report,"_

Volume 3, Issued September 1999.

12. NUREG/CR-6762, Volume 1, "GSI-191 Technical Assessment: Parametric Evaluations for Pressurized Water Reactor Recirculation Sump Performance," Issued August 2002.

13. NUREG/CR-6762, Volume 2, "GSI-191 Technical Assessment: Summary and Analysis of U.S. Pressurized Water Reactor Industry Survey Responses and Responses to GL 97-04," Issued August 2002.

14. NUREG/CR-6762, Volume 3, "GSI-191 Technical Assessment: Development of Debris Generation Quantities in Support of the Parametric Evaluation," Issued August 2002.

to W3F1-2008-0069 Page 84 of 87

15. NUREG/CR-6762, Volume 4, "GSI-191 Technical Assessment: Development of Debris Transport Fractions in Support of the Parametric Evaluation," Issued August 2002.

16. NUREG/CR-6772, "GSI-191: Separate Effects Characterization of Debris Transport in Water," Issued August 2002.

17. NUREG/CR-6773, "GSI-1 91: Integrated Debris-Transport Tests in Water Using Simulated Containment Floor Geometries," Issued December 2002.

18. NUREG/CR-6808, "Knowledge Base for the Effect of Debris on Pressurized Water Reactor Emergency Core Cooling Sump Performance," Issued February 2003.

i9. NUREG/CR-6916, "Hydraulic Transport of Coating Debris, A Subtask of GSI-191," Issued December 2006.

20. Nuclear Energy Institute (NEI) Document 02-01, "Condition Assessment Guidelines:

Debris Sources Inside PWR Containments," Revision 1.

21. Not used.

22. WCAP-16568-P, "Jet Impingement Testing to Determine the Zone of Influence (ZOI) for DBA-Qualified / Acceptable Coatings," Revision 0.

23. C.D.I. Report 96-06, "Air Jet Impact Testing of Fibrous and Reflective Metallic Insulation,"

Revision A, included in Volume 3 of General Electric Document NEDO-32686-A, "Utility Resolution Guide for ECCS Suction Strainer Blockage."

24. Not used.

25. Alion Document No: ALION-REP-TXU-4464-02, Titled: TXU Paint Chip Characterization, Rev. 0.

26. Letter # TXX-07156 from Mike Blevins, Luminant Generation Company LLC, to the U.S.

Nuclear Regulatory Commission, dated November 8, 2007.

27. Letter from Jon R. Cavallo, Vice President of Corrosion Control Consultants and Labs Inc.

to Charles Feist, dated September 20, 2007.

28. Calculation 2004-07780, "Debris Generation Due to LOCA within Containment for Resolution of GL GSI-191," Revision 3.

29. Calculation 2005-05500, "Post-LOCA Debris Transport, Head Loss Across Safety Injection Sump Screen, and NPSH Evaluation for Resolution of GSI-191," Revision 2.

30. ER-W3-2003-0394-001, "Safety Injection Sump Modification."

31. Calculation 2005-02820, "GSI-191 Downstream Effects - Flow Clearances," Revision 0, dated August 18, 2005.

to W3F1-2008-0069 Page 85 of 87

32. Calculation, 2005-12840, "Evaluation of Downstream Components for Long Term Performance for Resolution of GSI-1 91," Revision 1, dated May 11, 2008.

33. Head Loss Testing of Waterford Unit 3 Safety injection Sump Strainers, 26A6833, Rev 9.

34. S0105 Task Design Input Request (DIR), Rev 5, DRF Object 0000-0079-0092.

35. Hydraulic Sizing and Head Loss Prediction for Suction Strainers (PWRs), TDP-01 86.

36. Safety Injection Pump Passive ECCS Strainer System S0100 Hydraulic Sizing Report, Waterford Unit 3 Nuclear Power Plant, Document No. GENE-0000-0053-4416-P-R4.

37. Calculation MNQ6-4, "Water Levels Inside Containment."

38. WCAP-16710-P, Rev. 0, "Jet Impingement Testing to Determine the Zone of Influence (ZOI) of Min-K and NUKON Insulation for Wolf Creek and Callaway Nuclear Operating Plants."

39. ALION-REP-ENTG-4771-02, Rev. 0, "Waterford 3 Metal Encapsulated Fiberglass Insulation ZOI and Size Distribution Report."

40. Calculation ECM89-083, Rev. 1, "Verification of Gaps at Whip Restraint U-Bolt for Min-K Insulation Required per C1258220, 265936, 266235, and 266371.

41. ALION-REP-ENT-4536-02, Rev. 0, "Waterford Unit 3 Low Density Fiberglass Debris Erosion Testing Report."

42. Procedure NOECP-451, Rev. 1, "Conducting Engineering Inspection of Reactor Containment Building Protective Coatings."

43. Specification 1564.734, Rev. 19, "General Protective Coating for Nuclear Power Plant."

44. NRC Letter from N. Kalyanam to K. Walsh, 12/10/07, "Waterford Steam Electric Station, Unit 3- Approval of Extension Request for Corrective Actions Re. Generic Letter 2004-02, "Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized Water Reactors". (Waterford 3 document # ILN07-01160).

45. Procedure OP-903-027, "Inspection of Containment."

46. Procedure PMC-002-007, "Maintenance and Construction Painting."

47. Procedure UNT-007-006, "Housekeeping."

48. Procedure W4.202, "System and Component Labeling."

49. Procedure EN-MA-118, "Foreign Material Exclusion."

50. Not Used.

to W3F1-2008-0069 Page 86 of 87

51. Calculation ECM07-001, "NPSH Analysis of Safety Injection and Containment Spray Pumps."

52. Regulatory Guide 1.1, "Net Positive Suction Head For Emergency Core Cooling and Containment Heat Removal System Pumps."

53. WCAP 16785-NP, "Evaluation of Additional Inputs to the WCAP-16530-NP Chemical Model."

54. WCAP-16406-P, Rev. 1, "Evaluation of Downstream Sump Debris Effects in Support of GSI-191.

55. Waterford 3 Letter W3F1-2007-0051, November 14, 2007, "Request for Extension of Completion Date for Corrective Action Required by GL 2004-02."

56. GE Report GENE-0000-0054-9349, Waterford 3 Safety Injection Sump Strainer, Plenum, and Sensor Stress Report.

57. NRC Letter from N. Kalyanam to K. Walsh, 1/28/08, "Waterford Steam Electric Station, Unit 3 - Report on Results of Staff Audit of Corrective Actions to Address Generic Letter 2004-02." (Waterford Document ILN03-0015).

58. WCAP-16530-NP, "Evaluation of Post-Accident Chemical Effects in Containment Sump Fluids to Support GSI-1 91."

59. NRC Letter from Thomas G. Hiltz to K. Walsh, May 22, 2008, "Waterford Steam Electric Station, Unit - Generic Letter 2004-02 "Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized Water Reactors" Approval of Extension Request."

60. WCAP-16793-NP, "Evaluation of Long-Term Cooling Considering Particulate, Fibrous and Chemical Debris in the Recirculating Fluid."

61. NRC Letter from William H. Ruland to Anthony R. Pietrangelo, February 4, 2008, "Draft Conditions and Limitations for Use of Westinghouse Topical Report WCAP-16793-NP, Revision 0, "Evaluation of Long-Term Cooling Considering Particulate, Fibrous and Chemical Debris in the Recirculating Fluid."

62. NRC Letter from William H. Ruland to Anthony R. Pietrangelo, dated November 30, 2007, "Supplemental Licensee Responses to Generic Letter 2004-02, "Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized Water Reactors."

63. NRC Letter from William H. Ruland to Anthony R. Pietrangelo, dated November 21, 2007, "Revised Content Guide for Generic Letter 2004-02 Supplemental Responses."

64. GE Report GENE-0000-0048-9192, "Waterford 3 High Energy Line Postulated Pipe Break Evaluation, Containment Sump Strainer."

65. Calculation CN-SEE-I-08-42, "Waterford 3 Nuclear Plant LOCADM."

to W3F1-2008-0069 Page 87 of 87

66. Drawing 5817-13604, Rev. 0, SIS Sump Strainer Interface Control Drawing sheet 1.

67. Calculation 1062-0802-0015-4, "Evaluation of Waterford 3 and ANO 2 HPSI Pump Mechanical Seals."

68. Calculation 1062-0015-03, "Evaluation of Applicability of Debris Laden Test Data to the ECCS Pumps' Cyclone Separators at Waterford 3."

69. Specification 1564.116, Revision 6, "Containment Spray Pumps."

70. Drawing 5817-11683, Revision 1, "Containment Spray Pumps A & B Seal & Water Piping."

71. Vendor Manual TD-C681.0015, "John Crane Installation Instructions for Type 8B-1 Seal, Revision 0."

Attachment 3 W3F1-2008-0069 Affidavit to W3F1-2008-0069 Paqe 1 of 3 GE Hitachi Nuclear Energy Americas LLC AFFIDAVIT 1, Tim E. Abney, state as follows:

(1) 1 am Vice President. Services Licensing, Regulatory Affairs, GE-Hitachi Nuclear Energy Americas LLC ("GEH"), have been delegated the function of reviewing the information described in paragraph (2) which is sought to be withheld, and have been authorized to apply for its withholding.

(2) The infi)rnation sought to be withheld is contained in Enclosure I of GEH's letter, JB08-JXDYR-001, J. Betsill to G. Scott, entitled " GEH Proprietary Mark-ups of Draft Entergy Letter W3FI-2008-0018", dated February 22,22008. GEH proprietary information in Enclosure 1, which is entitled "GEH Proprietary Mark-ups of Draft Entergy Letter W3F I -

2008-0018", is identified, by a dotted underline inside, double square brackets. ((This

.s.en!.c an mp.cYI.fl. In each case, the superscript notation t relers to Paragraph (3) of this affidavit, which provides the basis for the proprietary detennination.

(3) In making this application for withholding of proprietary information of which it is the owner or licensee, GEH relies upon the exemption from disclosure set forth in the Freedom of Infbrmation Act ("FOIA"), 5 USC Sec. 552(b)(4), and the Trade Secrets Act, 18 USC Sec. 1905, and NRC regulations 10 CFR 9.17(a)(4), and 2.390(a)(4) for "trade secrets" (Exemption 4). The material for which exemption from disclosure is here sought also qualify under the narrower definition of "trade secret", within the meanings assigned to those terms for purposes of FOIA Exemption 4 in, respectively, Critical Mass Energy Proiect v. Nuclear Regulatory Commission, 975F2d87l (DC Cir. 1992), and Public Citizen Health Research Group v. FDA 704F2d 1280 (DC Cir. 1983).

(4) Some examples of categories of information which fit into the definition of proprietary information are:

a. Infornnation that discloses a process, method, or apparatus, including supporting data and analyses, where prevention of its use by GEH's competitors without license from GEH constitutes a competitive economic advantage over other companies;

b. Information which, if used 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 of a similar product; C. Inforimnation Which reveals aspects of past, present, or future GEH customer-funded development plans and programs, resulting in potential products to GEII;

d. Information which discloses patentable subject matter for which it may be desirable to obtain patent protection.

The information sought to be withheld is considered to be proprietary for the reasons set forth in paragraphs (4)a. and (4)b. above.

Aft JB09-JX DYR-O 1.do*c Affidvit Page I of 3 to W3F1 -2008-0069 Paae 2 of 3 (5) To address 10 CFR 2.390(b)(4), the information sought to be withheld is being submitted to NRC in conlidence. The information is of a sort customarily held in confidence by GEH, and is in fact so held. The information sought to be withheld has, to the best of my knowledge and beliet, consistently been held in confidence by GEH, no public disclosure has been made, and it is not available in public sources. All disclosures to third parties, including any required transmittals to NRC, have been made, or must be made, pursuant to regulatory provisions or proprietary agreements which provide, for maintenance of the information in confidence. Its initial designation as proprietary information, and the subsequent steps taken to prevent its unauthorized disclosure, are as set forth in paragraphs (6) and (7) following.

(6) Initial approval of proprietary treatment of a document is made by the manager of the originating component, 'the person most likely to be acquainted with the value and sensitivity of the information in relation to industry knowledge, or subject to the terms under which it was licensed to GEH. Access to such documents within GEH is limited on a

ý"need to know" basis.

(7) The procedure for approval of external release of such a document typically requires review by the staff manager, project manager, principal scientist, or other equivalent authority for technical content, competitive effect, and determination of the accuracy of the proprietary designation. Disclosures outside GEH arc limited to regulatory bodies, customers, and potential customers, and their agents, suppliers, and licensees, and others with a legitimate need for the information, and then only in accordance with appropriate regulatory provisions or proprietary agreements.

(8) The intbrmation identified in paragraph (2), above, is classified as proprietary because it contains detailed results of analytical model and method, as well as testing methods, applied to perform evaluations of emergency core cooling system and containment sprays strainers in Boiling Water Reactors ("BWR") and Pressurized Water Reactors. The developmnent and approval of these models and methods was achieved at a significant cost to GEH, on the order of several million dollars.

The development of the evaluation process along with the interpretation and application of the analytical results is derived from the extensive experience database that constitutes a major GEH asset.

(9) Public disclosure of the information sought to be withheld is likely to cause substantial harm to GEH's competitive position and foreclose or reduce the availability of profit-making opportunities. The intbrmation is part of GEH's comprehensive safety and technology base, and its commercial value extends beyond the original development cost.

The value of the technology base goes beyond the extensive physical database and analytical methodology and includes development of the expertise to determine and apply the appropriate, evaluation process. In addition, the technology base includes the, value derived from providing analyses done with NRC-approved methods.

Aft JB08-.IXDYR-1tdoc Affidavit Page 2 of 3 to W3F1 -2008-0069 Page 3 of 3 The research, development, engineering, analytical and NRC review costs comprise a substantial investment of time and money by G EH.

The precise value of the expertise to devise an evaluation process and apply the correct analytical methodology is difficult to quantify, but it clearly is substantial.

GEH's competitive advantage will be lost if its competitors are able to use the results of the GEH experience to normalize or verify their own process or if they are able to claim-an equivalent understanding by demonstrating that they can arrive at the same or similar conclusions.

The value of this information to GEH would be lost if the infornation were disclosed to the public. Making such infonnation available to competitors without their having been required to undertake a similar expenditure of resources would unfairly provide competitors with a windfall, and deprive GEH of the opportunity to exercise its competitive advantage to seek an adequate return on its large investment in developing and obtaining these very valuable analytical tools.

I declare under penalty of perjury that the foregoing affidavit and the matters stated therein are true and correct to the best of my knowledge, information, and belief.

Executed on this 22nd day of February 2008.

N_"i ci /~ ~u Tim E. Abney GE-Hitachi Nuclear Energy Americas LLC Aff J1308-JXDYR-01 .doc Affidavit Page 3 of 3