W3F1-2015-0026, Response to NRC Natural Circulation Cooldown Analysis Audit Request for Additional Information

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Response to NRC Natural Circulation Cooldown Analysis Audit Request for Additional Information
ML15091A513
Person / Time
Site: Waterford Entergy icon.png
Issue date: 04/01/2015
From: Chisum M
Entergy Operations
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
W3F1-2015-0026
Download: ML15091A513 (23)
v • d • e


Text

s Entergy Operations, Inc.

17265 River Road Hwy 18 Killona, LA 70057 Tel 504 739 6660 Fax 504 739 6678 Michael R. Chisum Site Vice President Waterford 3 W3F1-2015-0026 April 1, 2015 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555

SUBJECT:

Response to NRC Natural Circulation Cooldown Analysis Audit Request for Additional Information Waterford Steam Electric Station, Unit 3 Docket No. 50-382 License No. NPF-38

1. W3F1-2013-0043, Request for Review of Change to Updated Final

REFERENCES:

Safety Analysis Report Clarifying Pressurizer Heaters Function for Natural Circulation at the Onset of a Loss of Offsite Power, dated November 11, 2013 [NRC ADAMS Accession Number ML13316C052].

2. NRC Letter Request for Additional Information Regarding a Change to the Update Final Safety Anlaysis Report Clarifying Pressurizer Heaters Function for Natural Circulation, October 21, 2014 [NRC ADAMS Accession Number ML14246A015].
3. NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light-Water Reactor] Branch Technical Position (BTP) 5-4, Design Requirements of the Residual Heat Removal System, Revision 4, March 2007 [NRC ADAMS Accession Number ML070850123].
4. W3F1-2014-0072, Waterford Steam Electric Station, Unit 3 Response to Request for Additional Information Regarding a Change to the Updated Final Safety Analysis Report Clarifying Pressurizer Heaters Function for Natural Circulation, January 13, 2015 [NRC ADAMS Accession Number ML15013A439].
5. NRC Audit Report,

Subject:

Waterford Steam Electric Station, Unit 3 -

Regulatory Audit Report in Support of License Amendment Request to Change the Updated Final Safety Analysis Report Clarifying Pressurizer Heaters Function for Natural Circulation at the Onset of a Loss of Offstie Power [NRC ADAMS Accession Number ML15071A337].

W3F1-2015-0026 Page 2

Dear Sir or Madam:

On November 11, 2013, Waterford Steam Electric Station, Unit 3 submitted a license amendment request [Reference 1] to clarify how the pressurizer heater function is met for natural circulation at the onset of a loss of offsite power concurrent with the specific single point vulnerability. NRC letter dated October 21, 2014 [Reference 2] requested that Entergy provide additional information to support the review of the license amendment request. Specifically, the NRC staff requested that Entergy provide documentation that demonstrates NRC approval of the natural circulation cooldown analysis performed to comply with NUREG-0800 Branch Technical Position (BTP) 5-4

[Reference 3]. Waterford 3 provided a response to the NRC request on January 13, 2015 [Reference 4] which stated that the current Waterford 3 natural circulation cooldown analysis was documented in a design analysis report (DAR) as part of the replacement steam generator (RSG) project.

The NRC staff performed an audit of the natural circulation cooldown analysis documentation at the offices of Westinghouse Electric Company in Rockville, Maryland on February 11, and 12, 2015. The NRC staff's completed an audit report [Reference 5]

which had a Request for Additional Information (RAI) in order to complete the licensing amendment request review. The NRC audit RAIs are addressed in Attachment 1 to this letter.

This amendment response contains one new commitment (Attachment 4).

If you have any questions or require additional information, please contact John Jarrell, Regulatory Assurance Manager, at 504-739-6685.

I declare under penalty of perjury that the foregoing is true and correct. Executed on April 1, 2015.

Sincerely, MRC/JPJ/wjs Attachments:

1. NRC Natural Circulation Cooldown Analysis Audit Response
2. Draft Final Safety Analysis Report Section 5.4.10.2 Markup
3. Draft Final Safety Analysis Report Section 9.3 Markup
4. List of Regulatory Commitments

W3F1-2015-0026 Page 3 cc: Mr. Marc L. Dapas Regional Administrator U. S. NRC, Region IV RidsRgn4MailCenter@nrc.gov NRC Senior Resident Inspector for Waterford 3 Frances.Ramirez@nrc.gov (SRI)

Chris.Speer@nrc.gov (RI)

NRC Program Manager for Waterford 3 Michael.Orenak@nrc.gov Louisiana Department of Environmental Quality Office of Environmental Compliance Surveillance Division Ji.Wiley@LA.gov

Attachment 1 to W3F1-2015-0026 NRC Natural Circulation Cooldown Analysis Audit Response

W3F1-2015-0026 Attachment 1 Page 1 of 8 NRC Natural Circulation Cooldown Analysis Audit Response On November 11, 2013, Waterford Steam Electric Station, Unit 3 submitted a license amendment request [Reference 1] to clarify how the pressurizer heater function is met for natural circulation at the onset of a loss of offsite power concurrent with the specific single point vulnerability. NRC letter dated October 21, 2014 [Reference 2] requested that Entergy provide additional information to support the review of the license amendment request. Specifically, the NRC staff requested that Entergy provide documentation that demonstrates NRC approval of the natural circulation cooldown analysis performed to comply with NUREG-0800 Branch Technical Position (BTP) 5-4

[Reference 3]. Waterford 3 provided a response to the NRC request on January 13, 2015 [Reference 4] which stated that the current Waterford 3 natural circulation cooldown analysis was documented in a design analysis report (DAR) as part of the replacement steam generator (RSG) project.

The NRC staff performed an audit of the natural circulation cooldown analysis documentation at the offices of Westinghouse Electric Company in Rockville, Maryland on February 11, and 12, 2015. The NRC staff's completed an audit report [Reference 5]

which had a Request for Additional Information (RAI) to complete the licensing amendment request review. Waterford 3 entered the Final Safety Analysis Report related issues identified in the audit report into the site corrective action process [CR-WF3-2015-01673]. The NRC audit RAIs are addressed below.

NRC Audit Report RAI #1 State the limiting failure chosen in the natural circulation cooldown analysis and explain why it is limiting.

Entergy Response #1 The Waterford 3 natural circulation analysis determines the time required to reach cold shutdown conditions and the amount of emergency feedwater inventory required. The limiting single failure is dependent upon what results are being determined. The information below is intended to compile the major active equipment required in the analyses and to provide the references to where the NRC has already reviewed and approved the design with respect to single failure criteria associated with the residual heat removal system.

The limiting single failure with respect to the longest cooldown time is the loss of a DC bus. The loss of a DC bus causes that train emergency diesel generator and atmospheric dump valve control logic to fail. In this scenario, only one train of safety related equipment is available, and in particular only one shutdown cooling system train is available for cooldown from 350°F to 200°F. The transient credits local manual control of the atmospheric dump valve within the four hour hold period prior to cooldown initiation. Thus, the Waterford 3 plant is capable of being cooled to a cold shutdown conditions with only offsite or onsite power available within a reasonable period of time

W3F1-2015-0026 Attachment 1 Page 2 of 8 of 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />. Waterford 3 letter W3F1-2014-0072 [Reference 4] provided information from the NRC review of the extended power uprate natural circulation analysis.

Waterford 3 letter W3F1-2014-0072 [Reference 4] Attachment 2 contains the pertinent section of Waterford 3 letter W3F1-2004-0061 [Reference 6] which explains the loss of DC bus as one of the limiting single failures.

The limiting single failure with respect to emergency feedwater inventory usage is the failure of an atmospheric dump valve. For this single failure, the atmospheric dump valve is permanently unavailable, forcing a cooldown on a single steam generator.

Once on the shutdown cooling system, the cooldown proceeds rapidly, as two trains are available. The analysis demonstrates that sufficient safety related emergency feedwater inventory is available to achieve cold shutdown conditions. FSAR Figures 9.3-8a and 9.3-8b show the cooldown profile for the natural circulation cooldown with a failed atmospheric dump valve. The atmospheric dump valve actuators (FSAR Section 10.3.1) backup supply of motive gas is provided by Safety Class 3, Seismic Category I accumulators and provides a ten hour minimum supply. For this scenario, shutdown cooling entry conditions exceed 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />, thus procedural actions are credited for manually operating the remaining atmospheric dump valve handwheel or lining up backup air supplies for continued operation. The NRC review and approval of using manual action to control the atmospheric dump valve is described in NRC Safety Evaluation for Natural Circulation Cooldown, April 8, 1988 [Reference 7].

The Emergency Diesel Generators (EDGs) are described in FSAR Section 8.3.1

[Reference 1]. FSAR Tables 8.3-6 through 8.3-11 depict the single failure analysis and the failure mode analysis for the Class 1E electric systems. The natural circulation cooldown analyses [Reference 25-27] already address the EDGs as a potential limiting single failure.

The Emergency Feedwater System (EFW) is described in FSAR Section 10.4.9

[Reference 1]. FSAR Table 10.4-14.shows a failure mode and effects analysis for the EFW system. During the original Waterford 3 plant licensing, the NRC reviewed the EFW system with respect to residual heat removal. NUREG-0787 Supplement 5

[Reference 19] Section 5.4.3 page 5-3 states the following:

The EFWS, which is documented in the Waterford Unit 3 FSAR, includes three safety-grade emergency feedwater pumps. Two motor-driven emergency feedwater (EFW) pumps are powered from separate emergency power supplies, and one turbine-driven EFW pump is designed so that it can deliver EFW to either steam generators under the postulated complete loss-of-ac-power conditions. The EFWS is designed to seismic Category I , electric Class IE, and ASME Code Class 2 and 3 requirements. The staff has reviewed the EFWS and concluded that it meets SRP Section 10.4.9 and the system will have a high degree of reliability.

The Charging System (CVC) and Auxilary Pressurizer Spray System (APS) are described in FSAR Section 9.3.4 [Reference 1]. FSAR Table 9.3-15 shows a failure mode and effects analysis for the CVCS and APS. NUREG-0787 Supplement 8

W3F1-2015-0026 Attachment 1 Page 3 of 8

[Reference 21] required additional information associated with CVC and APS single failures. Waterford 3 supplied the additional information and the NRC issued a safety evaluation report associated with the auxiliary pressurizer spray system [Reference 23].

The APS safety evaluation report specifically states the following:

Therefore, the staff concluded that the APSS design of Waterford 3 could satisfy the requirements of BTP RSB 5-1 with the single failure of a charging loop isolation valve.

The Shutdown Cooling System (SDC) is described in FSAR Section 9.3.6 [Reference 1]. FSAR Table 9.3-16 shows failure modes and effects analysis for the SDCS.

NUREG-0787 [Reference 14] Section 5.4.3 page 5-22 states the following:

If onsite electric power is available and offsite electric power is unavailable, the SDCS is capable of cooling the RCS given a single active failure. Each of the two SDCS trains may be isolated independently from the other while allowing the nonisolated 100% capacity train to perform its safety function, which is in compliance with GDC 34.

Ultimate Heat Sink (UHS) is described in UFSAR Section 9.2.5 [Reference 1]. NUREG-0787 [Reference 14] Section 9.2.5 page 9-12 states the following:

The design described above assures that adequate heat removal capability to maintain plant safety is provided by the UHS for all modes of operation including accidents coincident with a single active failure. Thus, the requirements of GDC 44 and the guide lines of Regulatory Guide 1.27 "Ultimate Heat Sink for Nuclear Power Plants" Positions C.1, C.3, and C.4 regarding the UHS ability to maintain proper system temperature under all modes of operation are met.

NRC Audit Report RAI #2 Verify that the licensee is continuing to follow its current licensing basis assumptions for WF3 in a conservative manner.

Entergy Response #2 A review of the Waterford licensing basis was performed to validate that the licensing basis is being maintained conservative with respect to the design basis. The FSAR

[Reference 12], Technical Specifications [Reference 13], NUREG-0787 and Supplements [References 14-22], NRC Waterford 3 Natural Circulation Safety Evaluation Report [Reference 7], and NRC Waterford 3 Extended Power Uprate Safety Evaluation Report [Reference 23] were the major documents reviewed. The design basis information contained in letter W3F1-2004-0061 [Reference 6], Westinghouse Design Analysis Report DAR-PS-03-8 [Reference 25], Westinghouse calculation CN-SEE-II-09-21 [Reference 26] and Westinghouse calculation CN-SEE-II-08-6 [Reference 27] were reviewed.

W3F1-2015-0026 Attachment 1 Page 4 of 8 During the review, it was identified that the natural circulation cooldown analysis

[Reference 26] Table 2.1-2 shows the time to reach shutdown cooling entry condition is 13.9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br />. This scenario is with a single failure of the atmospheric dump valve. The design information conflicts with the current licensing basis information contained in UFSAR Section 9.3.6.3.3 and UFSAR Section 10.3.1. UFSAR Section 9.3.6.3.3 and UFSAR Section 10.3.1 both state that the natural circulation cooldown reaches shutdown cooling entry conditions prior to 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />. This discrepancy was entered into the Waterford 3 corrective action program [CR-WF3-2015-01674]. The NRC Waterford 3 Natural Circulation Cooldown Safety Evaluation [Reference 7] evaluated that the atmospheric dump valve would not be able to operate remotely for the most limiting cooldown time and approved the use of manual action.

With the exception identified above, the Westinghouse analyses continue follow the licensing basis assumptions in a conservative manner.

NRC Audit Report RAI #3 Provide justification for the power level that is applied in the natural circulation cooldown analyses as to being set at 100.5 percent, as provided in the January 13, 2015, supplement, instead of the normally used 102.0 percent.

Entergy Response #3 The Commission published a final rule in the June 1, 2000, Federal Register (Volume 65, Number 106, Rules and Regulations, pages 34913-34921), allowing licensees to justify a smaller margin for power measurement uncertainty. The final rule gave licensees the option of applying a reduced margin between the licensed power level and the assumed power level for the Emergency Core Cooling System (ECCS) evaluation, or maintaining the current margin of 2 percent power. The amended rule gave licensees the opportunity to use a reduced margin if they determine that there is a sufficient benefit. Licensees were allowed to apply the margin to gain benefits from operation at higher power, or the margin could be used to relax ECCS-related TSs (e.g.,

pump flows).

In Waterford 3 letter W3F1-2001-0091 [Reference 8], Waterford 3 requested that the Operating License be amended to reflect an increase in the licensed reactor power level from 3,390 MWt to 3,441 MWt utilizing the new 10CFR50 Appendix K rule requirements. The Waterford 3 power measurement uncertainty was reduced by increasing the feedwater flow measurement accuracy by utilizing high accuracy ultrasonic flow measurement instrumentation.

The NRC approved the Waterford 3 request in NRC Amendment 183 [Reference 9].

NRC Amendment 183 increased licensed power level to 3441 MWt with an associated full power measurement uncertainty of 0.5%. Thus, the Waterford 3 full power measurement uncertainty is no longer the previous standard 2% but is 0.5% which corresponds to an analysis initial power level of 100.5% instead of the 102%.

W3F1-2015-0026 Attachment 1 Page 5 of 8 NRC Audit Report RAI #4 For the two originally proposed UFSAR inserts in Section 5.4.10.2 concerning BTP 5-4, no references for additional information is present. Provide updated UFSAR pages that include text referring the reader back to UFSAR Section 9.3.6 for additional information about the natural circulation cooldown and BTP 5-4.

Entergy Response #4 (Draft Final Safety Analysis Report Section 5.4.10.2 Markup) duplicates letter W3F1-2013-0043 Final Safety Analysis Report (FSAR) markups and adds a reference to FSAR Section 9.3.6.3.3.1. FSAR Section 9.3.6.3.3.1 is supplied in .

NRC Audit Report RAI #5 UFSAR Section 9.3.6 does not explain the use of the CENTS code in the natural circulation cooldown analysis. Provide revised UFSAR pages that both explain the CENTS code applied in the natural circulation cooldown analysis and provide a reference to the latest calculation file applying the CENTS code for natural circulation cooldowns.

Entergy Response #5 contains a new FSAR Section 9.3.6.3.3.1 which provides the natural circulation cooldown analysis details. The following information is added to that FSAR section to address the use of the CENTS code.

This analysis utilizes the CENTS code (refer to FSAR Section 15.0.3.1.6 for the code description) to model the nuclear steam supply system transient.

Using a reference to FSAR Section 15.0.3.1.6 to explain the CENTS code is consistent with Regulatory Guide 1.181 [Reference 10] which endorses NEI 98-03 [Reference 11].

NEI 98-03 Section A4 states that referencing, rather than duplicating, information in the UFSAR can simplify the presentation and maintenance of the UFSAR information.

It was also requested to provide a reference to the latest calculation file applying the CENTS code for natural circulation cooldowns. FSAR Section 9.3.6.3.3.1 was added to describe the natural circulation cooldown analyses. Adding the analysis detail was done instead of adding the specific calculation numbers in the FSAR references. This is consistent with FSAR Chapter 15 accident analyses where the analysis is described but the specific calculation numbers are not listed. This avoids required FSAR changes for each revision of the calculation and alleviates a potential FSAR maintenance issue.

W3F1-2015-0026 Attachment 1 Page 6 of 8 NRC Audit Report RAI #6 No reference is provided in Section 9.3.6 for UFSAR Figures 9.3-8A and 9.3-8B. Provide updated UFSAR pages with a reference in Section 9.3.6 where these figures are mentioned as to the source of the reactor coolant temperature versus time plots.

Entergy Response #6 contains a new FSAR Section 9.3.6.3.3.1 which provides the natural circulation cooldown analysis details. FSAR Figures 9.3-8A and 9.3-8B were included in the subsection for the natural circulation cooldown with a single failure of an atmospheric dump valve.

NRC Audit Report RAI #7 Clarify or change the sentence in UFSAR Section 9.3.6 on page 9.3-48 that states shutdown cooling conditions were reached in less than ten hours, when both hot leg temperatures are reduced to 400F. However, the final hot leg temperature reached at 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> (36,000 seconds) as shown in Figures 9.3-8A and 9.3-8B is approximately 350F which is the stated entry temperature for the RHRS. Thus, this text appears to be in conflict with Figures 9.3-8A and 9.3-8B.

Entergy Response #7 contains a new FSAR Section 9.3.6.3.3.1. The natural circulation cooldown analysis FSAR information was enhanced to provide a better explanation of the analyses. The conflicting information was removed from the FSAR.

W3F1-2015-0026 Attachment 1 Page 7 of 8 REFERENCES

1. W3F1-2013-0043, Request for Review of Change to Updated Final Safety Analysis Report Clarifying Pressurizer Heaters Function for Natural Circulation at the Onset of a Loss of Offsite Power, dated November 11, 2013 [NRC ADAMS Accession Number ML13316C052].
2. NRC Letter Request for Additional Information Regarding a Change to the Update Final Safety Anlaysis Report Clarifying Pressurizer Heaters Function for Natural Circulation, October 21, 2014 [NRC ADAMS Accession Number ML14246A015].
3. NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light-Water Reactor] Branch Technical Position (BTP) 5-4, Design Requirements of the Residual Heat Removal System, Revision 4, March 2007 [NRC ADAMS Accession Number ML070850123].
4. W3F1-2014-0072, Waterford Steam Electric Station, Unit 3 Response to Request for Additional Information Regarding a Change to the Updated Final Safety Analysis Report Clarifying Pressurizer Heaters Function for Natural Circulation, January 13, 2015 [NRC ADAMS Accession Number ML15013A439].
5. NRC Audit Report,

Subject:

Waterford Steam Electric Station, Unit 3 -

Regulatory Audit Report in Support of License Amendment Request to Change the Updated Final Safety Analysis Report Clarifying Pressurizer Heaters Function for Natural Circulation at the Onset of a Loss of Offsite Power, March 18, 2015

[NRC ADAMS Accession Number ML15071A337].

6. W3F1-2004-0061, Supplement to Amendment Request NPF-38-249, Extended Power Uprate Waterford Steam Electric Station, Unit 3, July 28, 2004 [NRC ADAMS Accession Number ML042120475].
7. NRC Safety Evaluation for Natural Circulation Cooldown, April 8, 1988.
8. W3F1-2001-0091, Waterford 3 Appendix K Margin Recovery - Power Uprate Request, September 21, 2001 [NRC ADAMS Accession Number ML012700104].
9. NRC Issuance of Amendment 183 for the Waterford 3 Appendix K Margin Recovery - Power Uprate, March 29, 2002 [NRC ADAMS Accession Number ML020910734].
10. Regulatory Guide 1.181, Content of the Updated Final Safety Analysis Report in Accordance with 10CFR50.71(e), September 1999 [NRC ADAMS Accession Number ML992930009].
11. NEI 98-03 Revision 1, "Guidelines for Updating Final Safety Analysis Reports, June 1999 [NRC ADAMS Accession Number ML003779028].
12. Waterford 3, Final Safety Analysis Report, Revision 308 [Letter W3FI -2014-0062 November 11, 2014].
13. Waterford 3, Technical Specifications, Through Amendment 242.

W3F1-2015-0026 Attachment 1 Page 8 of 8

14. NUREG-0787, Waterford Steam Electric Station Unit 3, Safety Evaluation Report, July 1981.
15. NUREG-0787 Supplement 1, Waterford Steam Electric Station Unit 3, Safety Evaluation Report, October 1981.
16. NUREG-0787 Supplement 2, Waterford Steam Electric Station Unit 3, Safety Evaluation Report, January 1982.
17. NUREG-0787 Supplement 3, Waterford Steam Electric Station Unit 3, Safety Evaluation Report, April 1982.
18. NUREG-0787 Supplement 4, Waterford Steam Electric Station Unit 3, Safety Evaluation Report, October 1982.
19. NUREG-0787 Supplement 5, Waterford Steam Electric Station Unit 3, Safety Evaluation Report, June 1983.
20. NUREG-0787 Supplement 6, Waterford Steam Electric Station Unit 3, Safety Evaluation Report, June 1984.
21. NUREG-0787 Supplement 8, Waterford Steam Electric Station Unit 3, Safety Evaluation Report, December 1984.
22. NUREG-0787 Supplement 10, Waterford Steam Electric Station Unit 3, Safety Evaluation Report, March 1985.
23. NRC Waterford 3 Auxiliary Pressurizer Spray Safety Evaluation Report, April 22, 1986.
24. NRC letter, Waterford Steam Electric Station, Unit 3 - Issuance of Amendment 199 RE: Extended Power Uprate (TAC NO. MC1355), dated April 15, 2005 [NRC ADAMS Accession Number ML051030082].
25. Westinghouse Design Analysis Report DAR-PS-03-8 Revision 2, Waterford 3 Branch Technical Position 5-4Cooldown Report for Replacement Steam Generators.
26. CN-SEE-II-09-21 Revision 1, Natural Circulation Cooldown to 350ºF to Support BTP 5-4 Criteria for Waterford-3 with Replacement Steam Generators.
27. CN-SEE-II-08-6 Revision 1, Shutdown Cooling Analysis for the Waterford-3 RSG Program.

Attachment 2 to W3F1-2015-0026 Draft Final Safety Analysis Report Section 5.4.10.2 Markup Note: This markup is a duplication of letter W3F1-2013-0043 with changes on page 3 of 3 marked.

W3F1-2015-0026 Attachment 2 Page 1 of 3

W3F1-2015-0026 Attachment 2 Page 2 of 3

W3F1-2015-0026 Attachment 2 Page 3 of 3 Insert 2 Part of the closing circuitry to the breakers that provide power to the 480V non-safety switchgear buses 3A32 and 3B32 (that power the Pressurizer Heaters) share a specific common circuit breaker, CVCEBKR014AB-13. CVCEBKR014AB-13 powers the interlock 52z relay, SSDEREL2348-D (SSDEREL2398-D). The interlock 52z relay checks for completion of load stripping on the respective 480V non-safety switchgear buses 3A32(3B32) at the onset of a Loss of Offsite Power. If the load stripping is complete, the interlock 52z relay closes a contact in the closing circuitry to the breakers that provide power to the 480V non-safety switchgear buses 3A32 and 3B32 to allow the breakers to close automatically when the sequencer load block contact in the closing circuitry is closed.

Alternatively, if the specific common circuit breaker, CVCEBKR014AB-13, is Open, then the breakers that provide power to the 480V non-safety switchgear buses 3A32 and 3B32 will not close automatically at the onset of a Loss of Offsite Power. To close the breakers that power each Pressurizer Heater electrical switchgear 3A32(3B32), local manual operator action in the respective train Switchgear room is necessary.

Insert 2A The natural circulation cooldown analysis (refer to FSAR Section 9.3.6.3.3.1), performed to comply with Branch Technical Position 5-4, Design Requirements of the Residual Heat Removal System, does not credit the operation of any pressurizer heaters.

Therefore, the operator action to energize the Pressurizer Heaters is not a time critical operator action.

Insert 3 At the onset of a Loss of Offsite Power concurrent with the specific common circuit breaker, CVCEBKR014AB-13, being Open, the reenergization of the 480V non-safety switchgear buses 3A32 and 3B32 (that power the Pressurizer Heaters) will require action to be performed outside of the Control Room. To close the breakers that power the 480V nonsafety switchgear buses 3A32 and 3B32, local manual operator action in the respective train Switchgear room is necessary. Once each 32 switchgear bus is reenergized, the necessary Pressurizer Heaters powered from that bus can be reenergized from the Control Room.

The natural circulation cooldown analysis (refer to FSAR Section 9.3.6.3.3.1), performed to comply with Branch Technical Position 5-4, Design Requirements of the Residual Heat Removal System, does not credit the operation of any pressurizer heaters.

Therefore, the operator action to close the breakers that power each Pressurizer Heater electrical switchgear 3A32(3B32), located outside of the control room, is not a time critical operator action.

Attachment 3 to W3F1-2015-0026 Draft Final Safety Analysis Report Section 9.3 Markup

W3F1-2015-0026 Attachment 3 Page 1 of 4

W3F1-2015-0026 Attachment 3 Page 2 of 4 INSERT A 9.3.6.3.3.1 Natural Circulation Cooldown Analysis The natural circulation cooldown analyses are accomplished in two phases, the first phase is the initial cooldown to shutdown cooling initiation temperature and pressure, then shutdown cooling system (FSAR Section 9.3.6) operation phase to cool to the reactor coolant system temperature of 200°F. For a loss of offsite power and associated natural circulation cooldown, the initial phase is accomplished through the emergency feedwater system (FSAR Section 10.4.9) and the atmospheric dump valves (FSAR Section 10.3). This equipment is used to reduce the reactor coolant system temperature and pressure to values that permit operation of the shutdown cooling system. This analysis utilizes the CENTS code (refer to FSAR Section 15.0.3.1.6 for the code description) to model the nuclear steam supply system transient. The shutdown cooling system removes core decay heat and provides long-term core cooling following the initial phase of reactor cooldown. These analyses calculate the time to cooldown the plant to cold shutdown conditions and the emergency feedwater inventory required.

The natural circulation cooldown analyses are used to demonstrate compliance with Branch Technical Position (BTP) 5-4. BTP 5-4 delineates the design requirements of the residual heat removal system that was formerly BTP Reactor System Branch (RSB) 5-1. These analyses demonstrate that the following BTP 5-4 paragraph B functional requirements are met.

1. The design shall be such that the reactor can be taken from normal operating conditions to cold shutdown using only safety-grade systems satisfying General Design Criteria 1 through 5.
2. The systems shall have suitable redundancy in components and features, and suitable interconnections, leak detection, and isolation capabilities to assure that for onsite electrical power system operation (assuming offsite power is not available) and for offsite electrical power system operation (assuming onsite power is not available) the system function can be accomplished assuming a single failure.
3. The systems shall be capable of being operated from the control room with either only onsite or only offsite power available. In demonstrating that the systems can perform their function assuming a single failure, limited operator action outside the control room is considered acceptable if suitably justified.
4. The systems shall be capable of bringing the reactor to a cold shutdown condition, with only onsite or offsite power available, within a reasonable period of time following shutdown, assuming the most limiting single failure.

The limiting single failure with respect to emergency feedwater inventory usage is the failure of an atmospheric dump valve. For this single failure, the atmospheric dump valve is permanently unavailable, forcing a cooldown on a single steam generator.

Once on the shutdown cooling system, the cooldown proceeds rapidly, as two trains are

W3F1-2015-0026 Attachment 3 Page 3 of 4 available. The analysis demonstrates that sufficient safety related emergency feedwater inventory is available to achieve cold shutdown conditions. Figures 9.3-8a and 9.3-8b show the cooldown profile for the natural circulation cooldown with a failed atmospheric dump valve. The atmospheric dump valve actuators (FSAR Section 10.3.1) backup supply of motive gas is provided by Safety Class 3, Seismic Category I accumulators and provides a ten hour minimum supply. For this scenario, shutdown cooling entry conditions exceed 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />, thus procedural actions are credited for manually operating the remaining atmospheric dump valve handwheel or lining up backup air supplies for continued operation.

The limiting single failure with respect to the longest cooldown time is the loss of a DC bus. The loss of a DC bus causes that train emergency diesel generator and atmospheric dump valve control logic to fail. In this scenario, only one train of safety related equipment is available, and in particular only one shutdown cooling system train is available for cooldown from 350°F to 200°F. The transient credits local manual control of the atmospheric dump valve within the four hour hold period prior to cooldown initiation. Thus, the Waterford 3 plant is capable of being cooled to a cold shutdown conditions with only offsite or onsite power available within a reasonable period of time of 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />.

CEN-259 (Reference 5) documents the results of a natural circulation cooldown test performed at San Onofre Nuclear Generating Station that is applicable to Waterford 3.

This report shows that adequate boron mixing can be achieved with natural circulation and no letdown and that the cooldown can be achieved without the formation of a void in the upper head. This test was reviewed and approved by the NRC as applicable to Waterford 3 (Reference 6). Thus, the requirements of BTP RSB 5-4 are met.

The natural circulation cooldown analysis does not credit the operation of the pressurizer heaters. Therefore, operator action to energize the pressurizer heaters is not a time critical operator action.

W3F1-2015-0026 Attachment 3 Page 4 of 4

Attachment 4 to W3F1-2015-0026 List of Regulatory Commitments to W3F1-2015-0026 Page 1 of 1 List of Regulatory Commitments This table identifies actions discussed in this letter for which Entergy commits to perform. Any other actions discussed in this submittal are described for the NRCs information and are not commitments.

TYPE SCHEDULED (Check one)

COMMITMENT COMPLETION DATE ONE-TIME CONTINUING (If Required)

ACTION COMPLIANCE The final safety analysis report will be X After NRC updated to reflect W3F1-2015-0026 approval of the changes. license amendment request.

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