Text

Responses to NRC Audit Review Questions for Essential Raw Cooling Water and Component Cooling Water System License Amendment Request
	ML15243A044
Person / Time
Site:	Watts Bar
Issue date:	08/28/2015
From:	James Shea; Tennessee Valley Authority
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	CNL-15-170
	Download: ML15243A044 (67)
	v • d • e

Tennessee Valley Authority, 1101 Market Street, Chattanooga, Tennessee 37402 CNL-15-170 August 28, 2015 10 CFR 50.90 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001 Watts Bar Nuclear Plant, Unit 1 Facility Operating License No. NFP-90 NRC Docket No. 50-390

Subject:

Responses to NRC Audit Review Questions for Watts Bar Nuclear Plant Unit 1 Essential Raw Cooling Water and Component Cooling Water System License Amendment Request

References:

1. Letter from TVA to NRC, Watts Bar Nuclear Plant Unit 1 - Application to Revise Technical Specifications for Component Cooling Water and Essential Raw Cooling Water to Support Dual Unit Operation (TS-WBN-15-13), dated June 17, 2015 [ML15170A474]

2. Email from NRC to TVA, Preliminary Draft RAIs Associated with Proposed WBN 1 ERCW and CCS Technical Specifications LAR, dated July 2, 2015

3. Letter from NRC to TVA, Watts Bar Nuclear Plant, Unit 1 - Supplemental Information Needed for Acceptance of Requested Licensing Action Regarding Application to Add Technical Specifications to Support Dual-Unit Operations (TAC No. MF6376), dated July 9, 2015 [ML15187A403]

4. Letter from TVA to NRC, Responses to NRC Acceptance Review Questions for Watts Bar Nuclear Plant Unit 1 Essential Raw Cooling Water and Component Cooling Water System License Amendment Request (TAC No. MF6376), dated July 14, 2015 [ML15197A357]

By letter dated June 17, 2015, Tennessee Valley Authority (TVA) submitted a request for a change to Facility Operating License No. NPF-90 for Watts Bar Nuclear Plant (WBN) Unit 1 (Reference 1). The proposed change would create new Technical Specifications (TS) 3.7.16, Component Cooling System (CCS) - Shutdown, and TS 3.7.17, Essential Raw Cooling Water (ERCW) System - Shutdown, to support dual unit operation of WBN Units 1 and 2. By email dated July 2, 2015, the Nuclear Regulatory Commission (NRC) provided requests for additional information (RAI) on the proposed WBN Unit 1 license amendment (Reference 2). By letter dated July 9, 2015, the NRC requested supplemental information associated with the proposed WBN Unit 1 license amendment (Reference 3). By letter dated July 14, 2015 (Reference 4),

TVA submitted the requested supplemental information and responses to the NRC acceptance review questions, including proposed changes to TS 3.7.16 and TS 3.7.17.

U.S. Nuclear Regulatory Commission CNL-15-170 Page 2 August 28, 2015 Following submittal of the requested supplemental information and responses to the NRC RAIs, the NRC indicated that sufficient information was provided by TVA to support the NRC review of the proposed license amendment request (LAR). However, to facilitate a more efficient and timely interaction between the NRC and TVA, the NRC decided to perform an audit of the proposed LAR in the NRC White Flint offices located in Rockville, MD during the weeks of July 27 to July 31, 2015, August 3 to August 7, 2015, and August 25 to August 28, 2015.

During the audit, the NRC and TVA discussed numerous questions related to the LAR.

The enclosure provides the TVA responses to the NRC audit review questions. As a result of the TVA responses to the NRC audit review questions, changes are required to TS 3.7.16, TS 3.7.17, and the associated Bases. To address the NRC concern regarding the availability of Reactor Coolant System loops within the initial seven hours after reactor shut down, a change to additional TSs may be required. In addition, the enclosure includes wording additions for Final Safety Analysis Report (FSAR) Chapters 6, 9 and 10 to clarify ice bed sublimation assumptions and water sources available to the AFW System. The proposed changes to the TS will be submitted in a license amendment request by September 11, 2015. The FSAR changes will be incorporated in WBN Unit 2 FSAR Amendment 114.

Consistent with the standards set forth in Title 10 of the Code of Federal Regulations (10 CFR) 50.92(c), TVA has determined that the responses, as provided in this letter, do not affect the no significant hazards considerations associated with the proposed license amendment to add TS 3.7.16 and TS 3.7.17 previously provided in Reference 1. TVA has further determined that the proposed amendment still qualifies for a categorical exclusion from environmental review pursuant to the provisions of 10 CFR 51.22(c)(9). Additionally, in accordance with 10 CFR 50.91(b)(1), TVA is sending a copy of this letter and the enclosure to the Tennessee Department of Environment and Conservation.

There are no new regulatory commitments associated with this letter. Please direct any questions concerning this matter to Gordon Arent at (423) 365-2004.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 28th day of August 2015.

Respectfully, Digitally signed by J. W. Shea J. W. Shea DN: cn=J. W. Shea, o=Tennessee Valley Authority, ou=Nuclear Licensing, email=jwshea@tva.gov, c=US Date: 2015.08.28 22:26:28 -04'00' J. W. Shea Vice President, Nuclear Licensing Enclosure cc: See Page 3

U.S. Nuclear Regulatory Commission CNL-15-170 Page 3 August 28, 2015

Enclosure:

Responses to NRC Audit Review Questions for WBN Unit 1 ERCW and CCS License Amendment Request cc (Enclosure):

U.S. Nuclear Regulatory Commission, Region II NRC Senior Resident Inspector - Watts Bar Nuclear Plant, Unit 1 NRC Project Manager - Watts Bar Nuclear Plant, Unit 1 Director - Division of Radiological Health - Tennessee State Department of Environment and Conservation

ENCLOSURE Responses to NRC Audit Review Questions for WBN Unit 1 ERCW and CCS License Amendment Request

Background

By letter dated June 17, 2015, Tennessee Valley Authority (TVA) submitted a request for a change to Facility Operating License No. NPF-90 for Watts Bar Nuclear Plant (WBN) Unit 1 (Reference 1). The proposed change would create new Technical Specifications (TS) 3.7.16, Component Cooling System (CCS) - Shutdown, and TS 3.7.17, Essential Raw Cooling Water (ERCW) System - Shutdown, to support dual unit operation of WBN Units 1 and 2. By email dated July 2, 2015, the Nuclear Regulatory Commission (NRC) provided requests for additional information (RAI) on the proposed WBN Unit 1 license amendment (Reference 2). By letter dated July 9, 2015, the NRC requested supplemental information associated with the proposed WBN Unit 1 license amendment (Reference 3). By letter dated July 14, 2015 (Reference 4),

TVA submitted the requested supplemental information and responses to the NRC acceptance review questions, including proposed changes to TS 3.7.16 and TS 3.7.17.

Following submittal of the requested supplemental information and responses to the NRC RAIs, the NRC indicated that sufficient information was provided by TVA to support the NRC review of the proposed license amendment request (LAR). However, to facilitate a more efficient and timely interaction between the NRC and TVA, the NRC decided to perform an audit of the proposed LAR in the NRC White Flint offices located in Rockville, MD during the weeks of July 27 to July 31, 2015, August 3 to August 7, 2015, and August 25 to August 28, 2015.

During the audit, the NRC and TVA discussed numerous questions related to the LAR.

This enclosure provides the TVA responses to the NRC audit review questions. As a result of the TVA responses to the NRC audit review questions, changes are required to TS 3.7.16, TS 3.7.17, and the associated Bases. To address the NRC concern regarding the availability of Reactor Coolant System loops within the initial seven hours after reactor shut down, a change to additional TSs may be required. In addition, this enclosure includes wording additions for Final Safety Analysis Report (FSAR) Chapters 6, 9 and 10 to clarify ice bed sublimation assumptions and water sources available to the AFW System. The proposed changes to the TS will be submitted in a license amendment request by September 11, 2015. The FSAR changes will be incorporated in WBN Unit 2 FSAR Amendment 114.

Page 1 of 64

ENCLOSURE Responses to NRC Audit Review Questions for WBN Unit 1 ERCW and CCS License Amendment Request References

1. Letter from TVA to NRC, Watts Bar Nuclear Plant Unit 1 - Application to Revise Technical Specifications for Component Cooling Water and Essential Raw Cooling Water to Support Dual Unit Operation (TS-WBN-15-13), dated June 17, 2015

[ML15170A474]

2. Email from NRC to TVA, Preliminary Draft RAIs Associated with Proposed WBN 1 ERCW and CCS Technical Specifications LAR, dated July 2, 2015

3. Letter from NRC to TVA, Watts Bar Nuclear Plant, Unit 1 - Supplemental Information Needed for Acceptance of Requested Licensing Action Regarding Application to Add Technical Specifications to Support Dual Unit Operations (TAC No. MF6376), dated July 9, 2015 [ML15187A403]

4. Letter from TVA to NRC, Responses to NRC Acceptance Review Questions for Watts Bar Nuclear Plant Unit 1 Essential Raw Cooling Water and Component Cooling Water System License Amendment Request (TAC No. MF6376), dated July 14, 2015

[ML15197A357]

Page 2 of 64

ENCLOSURE Responses to NRC Audit Review Questions for WBN Unit 1 ERCW and CCS License Amendment Request Item NRC Question/Request TVA Response/Dated Posted No. Date Posted BOP - 1 With one Component Cooling Water System (CCS) or Essential Raw Cooling Water (ERCW) System train inoperable, a loss of redundancy has occurred. However, the TS 3.0.4 does not allow Mode changes when capability to mitigate an accident in one unit and cooldown the other unit (or maintain the applicable LCOs for that Mode are not met. The other unit in a cooldown condition) is maintained by the remaining operable CCS and problem with possible Mode change from 4 to 3 after ERCW trains. Therefore, a Mode change from 5 to 4 or from 4 to 3 is not anticipated.

shutdown is that equipment may have been taken out of service that is required to meet the LCOs of During a normal shutdown, decay heat removal is via the reactor coolant system (RCS)

TS that are required for Mode 3. A similar statement loops until sometime after the unit has been cooled down to Residual Heat Removal can be made for a Mode change from 5 to 4. The (RHR) System entry conditions (Tcold < 350ºF). Therefore, as LCO 3.7.16 and LCO 3.7.17 proposed TS make no provision for suspending or become Applicable (entry into Mode 4), the RCS loops are still operable. At this point, stopping the process of making more equipment LCO 3.7.16 requires an additional CCS Train B pump powered from and aligned to the inoperable that otherwise would be required to CCS Train B header, and LCO 3.7.17 requires one additional ERCW pump be capable of

1. support Mode change, if the proposed LCO is not being powered from and aligned to each ERCW train. However, the requirement of met. Explain why the proposed TSs do not include LCO 3.4.6, RCS-Loops - MODE 4, is still being met by the two operable RCS loops.

an Action to stop making more equipment inoperable that would be required for the Mode change when If the requirement of either LCO 3.7.16 or LCO 3.7.17 is not met, maintaining the unit in the proposed TS LCOs are not met, or make Mode 4 with decay heat removal from the RCS loops is preferred, given the additional provisions to correct the issues. methods available to remove decay heat (i.e., RCS loops). However, if TS Required Actions require entry into Mode 5, the remaining operable RHR loop is sufficient to cooldown the unit to and maintain it in Mode 5, even with a concurrent loss of coolant Date Posted: 07/31/15 accident (LOCA) in the other unit.

Therefore, it is unnecessary for TS to provide provisions for suspending or stopping the process of making more equipment inoperable.

BOP - 2 The Actions of LCO 3.7.16 and LCO 3.7.17 are predicated on the preference to maintain the unit in a condition with multiple methods of decay heat removal available, i.e., maintain TS 3.7.7 and TS 3.7.8 have provision to enter LCO the unit in Mode 4 with two RCS loops operable in addition to the remaining operable 3.4.6 with one train inoperable because ERCW and RHR loop. This action precludes entry into the LCO 3.4.6 Actions, as LCO 3.4.6 is met CCS are support systems for decay heat with two operable RCS loops and one RCS loop in operation. However, if it is necessary

2. removal. Otherwise TS 3.0.6 could allow LCO 3.4.6 to place the unit in Mode 5 to comply with TS Required Actions, LCO 3.7.16 and to not be entered. Standard TS are similarly worded. LCO 3.7.17 Actions require the unit to be placed in Mode 5 in 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

It appears that proposed TS 3.7.16 and 3.7.17 actions for one train inoperable should similarly have If the Action to verify two RCS loops operable and one RCS loop in operation cannot be provisions for entering TS 3.4.6. TS 3.4.6 would lead met, no actions are specified. Therefore, LCO 3.0.3 applies, requiring the unit to be to different action than what TS 3.7.16 and TS 3.7.17 placed in MODE 5 in 37 hours4.282407e-4 days 0.0103 hours 6.117725e-5 weeks 1.40785e-5 months . With one ERCW train inoperable and Required Actions Page 3 of 64

ENCLOSURE Responses to NRC Audit Review Questions for WBN Unit 1 ERCW and CCS License Amendment Request Item NRC Question/Request TVA Response/Dated Posted No. Date Posted currently propose for 1 loop inoperable. Explain why require the unit to be placed in MODE 5; Condition A applies, requiring the unit to be the proposed TS differ from 3.7.7 and 3.7.8 in this placed in MODE 5 in 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

respect, or correct the issue.

These Actions are conservative to the Required Actions of LCO 3.7.7, LCO 3.7.8, and Date Posted: 07/31/15 LCO 3.4.6 when there are two operable RCS loops, and are consistent with the requirements of LCO 3.7.7, LCO 3.7.8, and LCO 3.4.6 when there are no operable RCS loops and one inoperable RHR loop.

BOP - 3 Similar to the response to Question #2, decay heat removal is via the RCS loops until sometime after the unit has been cooled down to RHR entry conditions (Tcold < 350ºF). As TS 3.7.16 and TS 3.7.17 have required action to LCO 3.7.16 and LCO 3.7.17 become applicable, the requirement of LCO 3.4.6 is still verify Tavg > 200 F. But the bases for TS 3.4.6 says being met by the two operable RCS loops.

with one RHR train available, it would be safer to be in Mode 5 because if the remaining train of decay If the requirements of either LCO 3.7.16 or LCO 3.7.17 are not met, maintaining the unit in heat was lost, the loss would occur at a lower Mode 4 with decay heat removal from the RCS loops is preferred, given the additional temperature. Explain this difference between methods available to remove decay heat. This action precludes entry into the LCO 3.4.6 TS 3.4.6 and proposed TS 3.7.16 and 3.7.17. Actions, as LCO 3.4.6 is met with two operable RCS loops and one RCS loop in operation. However, if it is necessary to place the unit in Mode 5 to comply with TS Date Posted: 07/31/15 Required Actions, LCO 3.7.16 and LCO 3.7.17 Actions require the unit to be placed in Mode 5 in 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

3.

If the Action to verify two RCS loops operable and one RCS loop in operation cannot be met, no actions are specified. Therefore, LCO 3.0.3 applies, requiring the unit to be placed in MODE 5 in 37 hours4.282407e-4 days 0.0103 hours 6.117725e-5 weeks 1.40785e-5 months . With one ERCW train inoperable and Required Actions require the unit to be placed in MODE 5, Condition A applies, requiring the unit to be placed in MODE 5 in 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

These Actions are conservative to the Required Actions of LCO 3.7.7, LCO 3.7.8, and LCO 3.4.6 when there are two operable RCS loops, and are consistent with the requirements of LCO 3.7.7, LCO 3.7.8, and LCO 3.4.6 when there are no operable RCS loops and one inoperable RHR loop.

BOP - 4 LCO 3.7.16 and LCO 3.7.17 provide requirements in addition to those of LCO 3.7.7 and LCO 3.7.8. However, the additional requirements of LCO 3.7.17 are not required for DG 4.

TS 3.7.8 for ERCW has a provision for entering TS operability. There is sufficient flow to the diesel generators (DGs) from ERCW without a 3.8.1 for emergency diesel generators made third ERCW pump in each train to support DG Operability. Although the requirements of Page 4 of 64

ENCLOSURE Responses to NRC Audit Review Questions for WBN Unit 1 ERCW and CCS License Amendment Request Item NRC Question/Request TVA Response/Dated Posted No. Date Posted inoperable by ERCW when one ERCW loop is LCO 3.7.17 may not be met (i.e., a third pump capable of being aligned to each ERCW inoperable. Discuss why TS 3.7.17 does not have the Train) the requirements of LCO 3.7.8 are still met. If the requirements of LCO 3.7.8 are same provision. not met, the Actions of LCO 3.7.8 include the requirement to enter the Conditions and Required Actions of LCO 3.8.1 for DGs made inoperable by ERCW.

Date Posted: 07/31/15 BOP - 5 With two ERCW trains inoperable for LCO 3.7.17, there may be insufficient ERCW flow available to place the non-accident unit in Mode 5 during the mitigation of a LOCA on the TS 3.7.17 has a Note for LCO 3.0.3 to suspend Mode other unit. Otherwise, the preference is to maintain the unit in a condition with multiple change. For a scenario of two ERCW trains methods of decay heat removal available, i.e., maintain the unit in Mode 4 with two RCS inoperable per TS 3.7.17, but at least one ERCW loops operable.

5.

train operable per TS 3.7.8, provide discussion and justification why it is safer to stay in Mode 4 and not continue cooldown to Mode 5 for this TS condition.

Date Posted: 07/31/15 BOP - 6 Similar to the response to Question #2, decay heat removal is via the RCS loops until sometime after the unit has been cooled down to RHR entry conditions (Tcold < 350ºF). As In TS 3.7.16 and TS 3.7.17 for one train inoperable, LCO 3.7.16 and LCO 3.7.17 become applicable, the requirement of LCO 3.4.6 is still the TS Actions make a distinction between a normal being met by the two operable RCS loops.

shutdown and a TS required shutdown. Explain the distinction in shutdown requirements between a If the requirements of either LCO 3.7.16 or LCO 3.7.17 are not met, Condition B requires normal shutdown and a TS required shutdown for TS that the unit be maintained in Mode 4 (with decay heat removal from the RCS loops).

3.7.16 and 3.7.17 Maintaining the unit in Mode 4 with additional methods of decay heat removal available minimizes the likelihood of a situation where the decay heat and residual heat of the unit Date Posted: 07/31/15 exceeds the capability of the available RHR loop resulting in the possibility of an 6.

unintentional Mode change. If the Required Actions and Completion Times of Condition B are not met, no actions are specified. Therefore, LCO 3.0.3 applies, requiring the unit to be placed in Mode 5 in 37 hours4.282407e-4 days 0.0103 hours 6.117725e-5 weeks 1.40785e-5 months . With one CCS train inoperable and Required Actions require the unit to be placed in Mode 5, Condition A applies, requiring the unit to be placed in Mode 5 in 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

If TS Required Actions require entry into Mode 5, the remaining operable RHR loop is sufficient to cooldown the unit to and maintain it in Mode 5, even with a concurrent LOCA in the other unit.

Page 5 of 64

ENCLOSURE Responses to NRC Audit Review Questions for WBN Unit 1 ERCW and CCS License Amendment Request Item NRC Question/Request TVA Response/Dated Posted No. Date Posted BOP - 7 See responses to BOP-10 and BOP-11.

TVA has made a general statement that there is adequate clean water on site to support a 48 hour5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months cooldown with aux feedwater since:

x Unit 1 CST @ 395,000 gallons (TS SR 3.7.6.1 requires 200,000 gallons) x Unit 2 CST @ 395,000 gallons (TS SR 3.7.6.1 (Rev J) requires 200,000 gallons) x Demin tank @ 500,000 gallons x Aux feedwater storage tank @ 500,000 gallons (added for FLEX mitigating strategies)

TVAs response in the supplement included appeared to credit cross-tie between the Unit 1 and Unit 2 CSTs as well as reliance on water stored in the FLEX

7. tank. It is not clear how these clean water sources are currently identified and credited within the licensing basis to support normal hot standby cooling functions, as this does not appear to be described in the FSAR, and other licensing basis related references, as noted below:

From Watts Bar FSAR:

9.2.6.3 The ERCW system pool quality feedwater will be used during an extreme emergency when safety is the prime consideration and steam generator cleanliness is of secondary importance.

10.4.9.2 Since the ERCW system supplies poor quality water, it is not used except in emergencies when the condensate supply is unavailable.

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ENCLOSURE Responses to NRC Audit Review Questions for WBN Unit 1 ERCW and CCS License Amendment Request Item NRC Question/Request TVA Response/Dated Posted No. Date Posted From SE in NUREG 0847, Supp 23 (June 2011):

TVA's proposed clarification to the FSAR is acceptable to the NRC staff. Because the CSTs are credited only for the SBO event under 10 CFR 50.63, and TVA does not plan to share CSTs between the units during plant operation, the staff concludes that TVA satisfies GDC 5 regarding the CSTs. Confirmation by the staff of TVA's change to FSAR Section 10.4.9 to reflect TVA's intention to operate with each CST isolated from the other is Open Item 62 (Appendix HH).

Date Posted: 07/31/15 BOP - 8 TVA notes the AFW inlet temperatures are affected by the component cooling heat exchangers and the containment spray heat exchangers as well as miscellaneous loads.

Because the AFW takeoff is at the end of the ERCW The component cooling heat exchanger analysis demonstrates that ERCW discharge system, after heat removal from the safety related temperature may exceed 120ºF but these are cases where AFW is not required. However, heat exchangers, confirm ERCW to AFW will still be an examination of the containment spray heat exchanger discharge layout with respect to at the assumed maximum allowable temperature to AFW suction indicates cases where AFW could exceed 120ºF when AFW is required.

satisfy Chapter 15 requirements (80-120°F).

The following arrangements are noted:

Date Posted: 07/31/15 x Motor driven AFW 1A-A/TDAFW 1A-S is upstream of CCS heat exchanger 1A but 8.

downstream of CCS heat exchanger 2A.

x Motor driven AFW 1B-B/TDAFW 1A-S is upstream of CCS heat exchanger 2B but downstream of CCS heat exchanger 1B.

x Motor driven AFW 2A-A/TDAFW 2A-S is upstream of CCS heat exchanger 1A but downstream of CCS heat exchanger 2A.

x Motor driven AFW 2B-B/TDAFW 2A-S is upstream of CCS heat exchanger 2B but downstream of CCS heat exchanger 1B.

If a LOCA was postulated in unit 1:

Page 7 of 64

ENCLOSURE Responses to NRC Audit Review Questions for WBN Unit 1 ERCW and CCS License Amendment Request Item NRC Question/Request TVA Response/Dated Posted No. Date Posted x MDAFW 1A-A would be bounded by 120ºF and MDAFW 2A-A would be bounded by 120ºF.

x MDAFW 1B-B would be bounded by 130ºF and MDAFW 2B-B would be bounded by 130ºF.

If a LOCA was postulated in unit 2:

x MDAFW 1A-A would be bounded by 130ºF and MDAFW 2A-A would be bounded by 130ºF.

x MDAFW 1B-B would be bounded by 120ºF and MDAFW 2B-B would be bounded by 120ºF.

Therefore, there are four cases where the AFW inlet temperature might exceed 120ºF.

Impacts would be as follows:

LBLOCA The large break LOCA is described in FSAR Section 15.4. For the case where the AFW is on the same unit as the large break LOCA (U1/MDAFW 1B-B and U2/MDAFW 2A-A) the AFW is only used to initially fill the steam generators. The large break LOCA does not credit the continued use of AFW. This filling would occur prior to switchover to containment spray recirculation and therefore would happen prior to the temperature reaching 130ºF at the discharge of the containment spray heat exchanger. The large break LOCA peak clad temperature and the containment peak pressure analysis would not be impacted.

For the case where the large break LOCA was in the opposite unit (U1/MDAFW 2B-B and U2/MDAFW 1A-A), the AFW would be used to shutdown the non-LOCA unit and may be at a higher ERCW discharge temperature. This would result in slightly warmer water for unit shutdown and would have no deleterious effect on the non-LOCA unit.

SBLOCA The small break LOCA is described in FSAR Section 15.3. Unlike the large break LOCA, the small LOCA credits the continued use of AFW for cooling the LOCA unit. Break sizes for small LOCAs examined in the FSAR range from 2 inches to over 8 inches in diameter.

It can be observed from FSAR Table 15.3-2 and supporting calculations that for all but the smallest break size, the transient has peaked prior to switchover to the containment sump recirculation and therefore would not be impacted. The smallest break peaks later in the Page 8 of 64

ENCLOSURE Responses to NRC Audit Review Questions for WBN Unit 1 ERCW and CCS License Amendment Request Item NRC Question/Request TVA Response/Dated Posted No. Date Posted transient but at a much lower peak clad temperature. Because the only impact of the higher AFW temperature is a slightly increased enthalpy for the liquid entering the generator and because the majority of the energy is removed by vaporization (inlet enthalpy of 81 Btu/lb versus 91 Btu/lb for AFW compared to steam at approximately 1183 Btu/lb), a less than 1% change in enthalpy would be insignificant to the smallest LOCA.

The smallest LOCA peaks at 1009ºF for Unit 2 with considerable margin to the limiting peak clad temperature of 2200ºF and margin to the limiting small break LOCA peak clad temperature of 1183.9ºF.

NPSH A NPSH analysis indicates AFW will perform acceptably at 130ºF with NPSH margin.

Condition Report (CR) 1072659 has been initiated to document this issue. A copy of the CR has been uploaded to the Sharepoint Site. The AFW system description, FSAR, and associated calculations will be revised to note cases where the AFW maximum temperature of 120ºF may be exceeded.

BOP - 9 When the opposite unit has been shutdown for a period of time, the additional CCS and ERCW pump requirements of LCO 3.7.16 and LCO 3.7.17 are not required to ensure Describe the ERCW and CCS analysis as it pertains adequate decay heat removal by the RHR System. However, there may be some to a LOCA in one unit while in MODE 4 and a scenarios when the opposite unit has been shutdown for greater than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months that the controlled shutdown of the other Unit as it enters heat removal capacity of the RHR System is insufficient without the CCS and ERCW MODE 4 or 5. System requirements of LCO 3.7.16 and LCO 3.7.17 being applicable.

Date Posted: 07/31/15 Therefore, TVA will remove Applicability Note b, so that the Applicability of LCO 3.7.16 and LCO 3.7.17 in Modes 4 and 5 is dependent on whether the associated unit has been shutdown for less than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

9.

This response supersedes the response provided to Acceptance Review Question #2 provided in TVA letter dated July 14, 2015.

The following events are required to be supported by the CCS and ERCW configurations proposed in TS 3.7.16 and TS 3.7.17.

The CCS shall be designed to remove heat from potentially or normally radioactive heat loads during any mode of normal operation, and incidents of moderate frequency. In addition, the CCS shall be designed to remove heat from the RHR HXs and various pump Page 9 of 64

ENCLOSURE Responses to NRC Audit Review Questions for WBN Unit 1 ERCW and CCS License Amendment Request Item NRC Question/Request TVA Response/Dated Posted No. Date Posted seal and/or lube oil coolers during infrequent incidents, and limiting faults. The CCS is required to mitigate the consequences of Design Basis Events (DBEs). The required DBEs and associated safety functions for the CCS are in WB-DC-40-64.

EVENTS IN WB-DC-40-64 THAT CREDIT CCS Fire Operating Basis Earthquake Safe Shutdown Earthquake Tornado Combustible Gases Inside Containment Control Room Evacuation Internally Generated Missiles General High Energy Line Break Heavy Load Drop Small Break LOCA Large Break LOCA Steam Generator Tube Rupture Rupture of a Control Rod Drive Mechanism Housing Waste Gas Decay Tank Rupture Fuel Handling Accident Loss of External Electrical Load and/or Turbine Trip Loss of Offsite Power Main Steam Line Break Main Feedwater Line Rupture Event Accidental Depressurization of Main Steam System Loss of Normal Feedwater Excess Heat Removal Due to Feedwater System Malfunction Moderate Energy Line Break Partial Loss of Forced Reactor Coolant Flow Single Reactor Coolant Pump Locked Rotor or Shaft Break Complete Loss of Forced Reactor Coolant Flow Excessive Load Increase Incident Accidental Depressurization of The Reactor Coolant System Inadvertent Safety Injection Operation - Power Operation Uncontrolled RCCA Bank Withdrawal From a Subcritical or Hot Zero Power Condition Uncontrolled RCCA Bank Withdrawal At Power Page 10 of 64

ENCLOSURE Responses to NRC Audit Review Questions for WBN Unit 1 ERCW and CCS License Amendment Request Item NRC Question/Request TVA Response/Dated Posted No. Date Posted Single RCCA Withdrawal At Full Power RCCA Misalignment Uncontrolled Boron Dilution Improper Fuel Assembly Loading Anticipated Transient Without Scram Failure of Nonsafety-Related Control Systems as an Initiating Event Minor Secondary System Pipe Breaks Loss of All AC Power (Station Blackout)

Loss of RHR During Mid-Loop Operations The ERCW System is required to mitigate the consequences of plant Design Basis Events described in WB-DC-40-64. It performs a Primary Safety Function by providing cooling and makeup for essential safety-related plant equipment and components in response to adverse plant operating conditions which impose safety-related performance requirements on the systems being served.

EVENTS IN WB-DC-40-64 THAT CREDIT ERCW Fire Design Basis Flood Operating Basis Earthquake Safe Shutdown Earthquake Tornado Combustible Gases Inside Containment Control Room Evacuation Internally Generated Missiles General High Energy Line Break Heavy Load Drop Small Break LOCA Large Break LOCA Steam Generator Tube Rupture Rupture of a Control Rod Drive Mechanism Housing Waste Gas Decay Tank Rupture Fuel Handling Accident Loss of External Electrical Load and/or Turbine Trip Loss of Offsite Power Main Steam Line Break Page 11 of 64

ENCLOSURE Responses to NRC Audit Review Questions for WBN Unit 1 ERCW and CCS License Amendment Request Item NRC Question/Request TVA Response/Dated Posted No. Date Posted Main Feedwater Line Rupture Event Accidental Depressurization of Main Steam System Loss of Normal Feedwater Excess Heat Removal Due to Feedwater System Malfunction Moderate Energy Line Break Partial Loss of Forced Reactor Coolant Flow Single Reactor Coolant Pump Locked Rotor or Shaft Break Complete Loss of Forced Reactor Coolant Flow Excessive Load Increase Incident Accidental Depressurization of the Reactor Coolant System Inadvertent Safety Injection Operation - Power Operation Uncontrolled RCCA Bank Withdrawal From a Subcritical or Hot Zero Power Condition Uncontrolled RCCA Bank Withdrawal at Power Single RCCA Withdrawal at Full Power RCCA Misalignment Uncontrolled Boron Dilution Improper Fuel Assembly Loading Anticipated Transient Without Scram Failure of Nonsafety-Related Control Systems as an Initiating Event Minor Secondary System Pipe Breaks Loss of All AC Power (Station Blackout)

Loss of RHR During Mid-Loop Operations BOP - 10 a. ERCW is the safety-related source of water to AFW. Whenever the steam generators are relied on for heat removal, the switchover from the CST to ERCW is required to be Follow-up to NRC acceptance review Question 3 - operable.

letter date July 14, 2015. The summary of that question was related to maintaining in Mode 3 or b. TVA considers the current Applicability of LCO 3.3.2, Table 3.3.2-1, Item 6.f, Auxiliary Mode 4 with decay heat being removed through the Feedwater Pumps Train A and B Suction Transfer on Suction Pressure - Low,

10. steam generators for at least 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months as one of the appropriate as written.

options for managing a unit shutdown and for TVA to address the use of available and approved clean TVA will develop a TRM to control this function while a change is being evaluated for a water sources, ERCW, and the CST in accordance generic approach with PWROG. Procedure guidance will be provided in the cooldown with the approved licensing basis. procedure (GO-6).

From the TVA response: Draft TRM has been posted to SharePoint site.

Page 12 of 64

ENCLOSURE Responses to NRC Audit Review Questions for WBN Unit 1 ERCW and CCS License Amendment Request Item NRC Question/Request TVA Response/Dated Posted No. Date Posted The safety-related water supply for AFW is ERCW.

The AFW suction source automatically switches from the CST to ERCW when a low pressure condition exists in the AFW pump suction piping from the CST.

The switchover to ERCW will occur whenever AFW is in service to assure heat removal through the steam generators if the low pressure condition exists. This assures the safety function of decay heat removal is accomplished.

TS Table 3.3.2-1, Engineered Safety Feature Actuation System Instrumentation. Item 6.f describes the auxiliary feedwater pump suction transfer on suction low. The applicable Mode for this item is Modes 1, 2, 3.

TS 3.7.5, Auxiliary feedwater system, applicable is Modes 1, 2, and 3 plus Mode 4 when steam generator is relied upon for heat removal.

TS 3.7.6 Condensate storage tank, applicability is Modes 1, 2, and 3 plus Mode 4 when steam generator is relied upon for heat removal.

BTS 3.7.6 states that as the preferred water source to satisfy accident analysis assumptions, the CST must contain sufficient cooling water to remove decay heat for 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months following a reactor trip from 100.6% RTP, and then to cool down the RCS to RHR entry conditions, assuming a coincident loss of offsite power and the most adverse single failure. In doing this, it must retain sufficient water to ensure adequate net positive suction head for the AFW pumps during cooldown, as well as account for any losses from the steam driven AFW pump turbine, or before isolating Page 13 of 64

ENCLOSURE Responses to NRC Audit Review Questions for WBN Unit 1 ERCW and CCS License Amendment Request Item NRC Question/Request TVA Response/Dated Posted No. Date Posted AFW to a broken line. The CST level required is equivalent to a usable volume of 200,000 gallons, which is based on holding the unit in MODE 3 for 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , followed by a cooldown to RHR entry conditions at 50 F/hour. This basis is established in Reference 4 and exceeds the volume required by the accident analysis.

UFSAR 9.2.6.3, Safety Evaluation, states that the condensate storage tanks are the preferred source of clean water supply for the auxiliary feedwater pumps and a storage reservoir for secondary system water.

The tanks are not an engineered safety feature. The engineered safety feature water source for the auxiliary feedwater system is the ERCW system (Safety Class 2b). Either tank is isolable, but auxiliary feedwater can be obtained from both tanks. The ERCW system pool quality feedwater will be used during an extreme emergency when safety is the prime consideration and steam generator cleanliness is of secondary importance.

NRC Questions:

a. Since the Unit 1 CST water is limited in volume to support operations beyond 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months and the proposed changes state that AFW operations is now needed for operations out to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , describe the bases for the CST to ERCW automatic switchover to support Mode 4 operations.

b. Based on the response to part a, describe the necessary changes to the UFSAR, TS, and TS Bases.

Date Posted: 08/03/15 Page 14 of 64

ENCLOSURE Responses to NRC Audit Review Questions for WBN Unit 1 ERCW and CCS License Amendment Request Item NRC Question/Request TVA Response/Dated Posted No. Date Posted BOP - 11 a. The Condensate Storage Tank (CST) is sized to provide seven hours of clean water (200,000 gallons). This assumes maintaining operation in Mode 3 for two hours and Follow-up to NRC acceptance review Question 3 - then cooling down for five hours at 50 degrees per hour. At seven hours post trip, the letter date July 14, 2015. The summary of that required amount for AFW flow is 175 gpm. An extrapolation of this amount for the question was related to maintaining in Mode 3 or remaining 41 hours4.74537e-4 days 0.0114 hours 6.779101e-5 weeks 1.56005e-5 months would result in the conclusion that approximately 430,500 gallons Mode 4 with decay heat being removed through the of clean water would be required to feed AFW.

steam generators for at least 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months as one of the options for managing a unit shutdown and for TVA to This estimate is very conservative in that maintaining a stable level requires much less address the use of available and approved clean water than cooling down and also in the fact that this level does not account for the water sources, ERCW, and the CST in accordance slow decay of AFW flow required over the period. As an example, at the 48 hour5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months point, with the approved licensing basis. only approximately 110 gpm is required to satisfy the AFW demand.

From the TVA response: b. Normal makeup to the CST is provided from vendor operated equipment. Clean water is produced and added to the Demineralized Water Storage Tank (DWST). The There is adequate clean water to support a unit capacity of the normal makeup system is sufficient to maintain CST inventory for being maintained on AFW for 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . The capacity extended continued operation in Mode 3 or 4.

of each of the two CSTs is 395,000 gallons and the normal maximum volume in the CSTs is As an actual example, in 2014, WBN was maintained in Mode 3 for the period from July 11.

approximately 385,000 gallons. Review of operational 13 at 1937 to July 15 at 0503. During the first 27 hours3.125e-4 days 0.0075 hours 4.464286e-5 weeks 1.02735e-5 months , the normal makeup system to data for the past five years shows that the WBN Unit the CST maintained CST level at the normal level of 325,000 gallons while supplying 1 CST has been maintained at approximately the required AFW flow in Mode 3. The level did decrease during the final five hours of 330,000 gallons. Because AFW is not the only Mode 3 operation, but this decrease was due to increased water usage associated with system that uses CST water, a standpipe is provided feed and bleed on the condensate system to establish secondary parameters to in the tank to assure that a minimum of 200,000 support plant startup.

gallons of water is available for the sole use of AFW.

Thus, the site maintains approximately 130,000 Thus, for a normal extended operation in Mode 4, such as one in which the plant gallons of water in the CST above the TS limit. cannot continue into Mode 5 due to inability to establish the required CCS and ERCW Normal make-up to the CST comes from the alignments to support entry into Mode 5, the normal DWST makeup to the CST will be Demineralized Water Storage (DWST) Tank and the used to replenish the CST inventory with clean water.

Make-up Water Treatment Plant (MWTP). The DWST tank has a capacity of 500,000 gallons and the level Should augmentation to the normal DWST makeup method be required, another has historically been maintained between 65 and 90 historical example is provided. On February 23, 2012 at 0235, it was discovered that percent full. There have been instances, including the normal makeup source was lost, forcing the site to bring in portable equipment to one earlier this year, where WBN Unit 1 was replenish the CSTs. The equipment was in place and operating on the same day at maintained in Mode 3 for more than two days using 2105 or approximately 18.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months following the loss. The portable trailers are capable the DWST and the MWTP. of providing 200 gpm to the CST. The procedure 0-SOI-59.01 Section 8.4, provides Page 15 of 64

ENCLOSURE Responses to NRC Audit Review Questions for WBN Unit 1 ERCW and CCS License Amendment Request Item NRC Question/Request TVA Response/Dated Posted No. Date Posted guidance for placing portable trailers in-service.

WBN recently added the Auxiliary Feedwater Storage Tank (AFWST) as part of the FLEX In all cases, the following guidance is provided in the Annunciator Response mitigating strategies. This tank has a capacity of Instructions (ARI). The low level alarm for the CSTs comes in at 210,000 gallons. The 500,000 gallons and is an immediately available response to this alarm is contained in procedure 1/2-ARI-36-42. Below is the wording source of clean water. The tank was designed to be associated with low level response for Unit 1. The Unit 2 instruction is identical with the seismically robust and to withstand the effects of exception of switching the designated tanks (A for B and B for A).

tornados. The AFWST supply piping is normally isolated by air operated valves (AOVs) from the Unit [3] IF level is low, THEN, REFER TO Tech Specs (LCO 3.7.6), and INITIATE makeup to 1 and Unit 2 condensate piping that supply the CST A from one of the following sources as listed in preferred order:

suction for the AFW pumps. The AOVs open on a low [3.1] From CST B per SOI-2&3.01, CONDENSATE AND FEEDWATER SYSTEM pressure signal from the upstream condensate [3.2] From DI Water Storage Tank per SOI-59.01, DEMINERALIZED WATER piping, a loss of AC power, or a loss of control air. SYSTEM.

Water can be transferred from the DWST to the AFWST using hoses and pumps that are maintained Makeup from the opposite unit CST is specified as the first response, as the normal by the FLEX program if power cannot be provided to makeup to the CST is from the Demin Water System. Receipt of this alarm will not the DWST booster pumps. The two CSTs have a normally be expected if the normal method remains available as the capacity of the cross tie that when opened provides an additional Demin makeup system exceeds the requirements for AFW flow.

approximately 330,000 gallons of clean water for the case of a LOCA on one unit and the other unit being The CST LoLo level alarm comes in at approximately 11,600 gallons. The actions shutdown. specified for this alarm are.

NRC Questions: [1] REDUCE demand from CST, if possible.

a. Provide the total expected water volumes [2] MAKEUP to CST A at maximum possible rate.

of clean water to support operations of [3] IF AFW Pumps are running, THEN the AFW system for the 48 hour5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months MONITOR the following:

durations. x AFW Storage Tank (AFWST) level decrease.

b. Provide the procedure steps (alarm x AFW Pump Suction Valves for swap to ERCW Discharge Header suction.

responses, AOPs, EOPs, etc) that direct [4] REFER TO Tech Specs (LCO 3.7.6).

operators to supplement the Unit 1 CST clean water supply from the new Flex These actions acknowledge that if continued AFW demand is required and adequate tank, opposite units CST and DWST (in clean water makeup can not be established, the crew should ensure that the safety front of the ERCW automatic switchover). related supply to AFW (ERCW) properly aligns to supply AFW needs when required.

c. Provide access to the design change package for the addition of the new Flex e. The following changes will be addressed in the FSAR.

tank. This should include the 50.59 Page 16 of 64

ENCLOSURE Responses to NRC Audit Review Questions for WBN Unit 1 ERCW and CCS License Amendment Request Item NRC Question/Request TVA Response/Dated Posted No. Date Posted reviews performed, P&ID drawings, 9.2.6.1 Design Bases piping isometrics, piping safety classification and new AOVs logic. The condensate storage facilities are designed to serve as a receiver of water from

d. Describe how the new Flex tank the main condenser high level dump and to provide treated water for makeup to the automatic switchover and ERCW main condenser while reserving a minimum amount for the auxiliary feedwater automatic switchover interact. system. This amount is required to hold the plant for two hours after a Design Basis Specifically, the set-point differences Event (DBE) and 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months to cool RCS from no-load hot standby at 50°F per hour to should be described. the point at which the residual heat removal system can take over.

e. Based on the response above, describe any necessary changes to the UFSAR, When the CSTs are intact and offsite power is available, the inventory available in TS, and TS Bases based on this new the CSTs plus makeup from the make-up water plant and the demineralized water required water volume. storage tank (FSAR Section 10.4), is capable of supplying clean water to support maintaining the plant on auxiliary feedwater for longer than seven hours without the Date Posted: 08/03/15 need to transfer the AFW pump suction to ERCW. No credit is taken for this additional water in the design and safety evaluations of condensate storage or AFW.

The condensate storage tanks are not an engineered safety feature and are not seismically qualified. The supply from the make-up water tank and the demineralized water storage tank and associated piping are not engineered safety features and are not seismically qualified. The storage tanks supply the preferred source of water to the auxiliary feedwater system, but the engineered safety feature source is the ERCW System (Safety Class 2b).

9.2.6.3 Safety Evaluation The condensate storage tanks are the preferred source of clean water supply for the auxiliary feedwater pumps and a storage reservoir for secondary system water. The tanks are not an engineered safety feature. The engineered safety feature water source for the auxiliary feedwater system is the ERCW system (Safety Class 2b).

Either tank is isolable, but auxiliary feedwater for either unit can be obtained from both tanks. This will be done only if necessary since each condensate storage tank normally contains auxiliary feedwater for just one unit.

The ERCW system pool quality feedwater will be used during events when safety is the prime consideration and steam generator cleanliness is of secondary importance.

Piping connected to the condensate storage tanks is routed through a heated tunnel Page 17 of 64

ENCLOSURE Responses to NRC Audit Review Questions for WBN Unit 1 ERCW and CCS License Amendment Request Item NRC Question/Request TVA Response/Dated Posted No. Date Posted under the tanks. Ice formation in the tanks during a period of prolonged low temperatures can be prevented, if necessary, by recirculation of water through the condensate transfer pump. The tank and its connecting piping can accommodate water whose temperature is in the range of 40°F to 130°F.

10.4.9.2 System Description The two reactor units have separate AFW systems, as shown in Figure 10.4-21.

Each system has two electric motor-driven pumps and one turbine-driven pump.

Each of the electric pumps serves two steam generators; the turbine pump serves all four. All three pumps supporting a unit automatically deliver rated flow within one minute upon a trip of both turbine-driven main feedwater pumps, loss of offsite power, an AMSAC signal, a safety injection signal or low-low steam generator water level. The motor driven pumps (MDPs) start on a two-out-of-three low-low level signal in any steam generator and the turbine driven pump starts on a two-out-of-three low-low level signal in any two steam generators. Each pump supplies sufficient water for evaporative heat removal to prevent operation of the primary system relief valves or the uncovering of the core. The operator has the capability to open an additional recirculation line on the MDPs when there is low decay heat required to be removed from the SG. These lines contain a normally closed valve that closes on an accident signal. The valve is operable after the accident signal, but if an additional accident signal occurs, the valve would be reclosed. This ensures that the forward flow requirements to remove decay heat have been satisfied. Significant pump design parameters are given in Table 10.4-1.

The preferred sources of water for all auxiliary feedwater pumps are the two 395,000 gallon condensate storage tanks. A minimum of 200,000 gallons in each tank is reserved for the AFW Systems by means of a standpipe through which other systems are supplied. The two CSTs are normally isolated from each other, with one CST dedicated to each unit. The AFW safety analyses take no credit for the ability to crosstie the CSTs. As an unlimited backup water supply for each unit, a separate ERCW system header feeds each motor-driven pump. The turbine-driven pump can receive backup water from either ERCW header. The ERCW supply is automatically (or remote-manually) initiated on a two-out-of-three low pressure signal in the AFW system suction lines. Pump protection during the automatic transfer to the ERCW supplies is assured by providing sufficient suction head and flow to the pumps and is Page 18 of 64

ENCLOSURE Responses to NRC Audit Review Questions for WBN Unit 1 ERCW and CCS License Amendment Request Item NRC Question/Request TVA Response/Dated Posted No. Date Posted verified by system analysis. Since the ERCW system supplies poor quality water, it is not used except when the condensate supply is unavailable.

In addition, the high pressure fire protection (HPFP) system which is cross-connected to the discharge of each motor driven AFW pump can be aligned to supply unlimited raw water directly to the steam generators, in the unlikely event of a flood above plant grade. Water from the HPFP system is supplied by four high pressure, vertical turbine, motor-driven, Seismic Category I pumps conforming to the requirements of ASME B&PV Code Section III, Class 3 with each having a rating of 1590 gpm at 300 feet head. These pumps are installed in the Seismic Category I Intake Pumping Station with motors above the maximum possible flood level. Each pump is capable of supplying 100% of the auxiliary feedwater demands for both units during a flood above plant grade. The four pumps are supplied from normal and emergency power with two pumps assigned to each of the two emergency power trains. Each pair of pumps on the same power train takes suction from a common sump which receives water through a settling baffle arrangement for all normal, and flood reservoir levels.

Generated CR 1075753 to address interactions from a GDC-5 perspective for Unit 1 and Unit 2 CSTs and AFW storage tank. This CR has been posted to SharePoint.

BOP-12 Tube plugging for the safety related heat exchangers are controlled by issued design output. These values may be found in the System Descriptions for System 70, Component What controls are in place to revisit flow calculations Cooling System (5%), System 72, Containment Spray System (10%), and System 74, to ensure assumed HX tube plugging and fouling Residual Heat Removal System (5%).

12.

factors reflect actual HX degradation or required plugging? TVAs system descriptions are design output under the 10 CFR 50 Appendix B, Quality Assurance Program.

BOP - 13 a) CCS HX C peak heat removal rate for mitigation of LOCA on one unit with loss of offsite power and loss of Train A:

The licensee is requested to provide the following

13. information: Heat Load Item With the loss of Train A: (Btu/hr) a) What is the peak heat removal rate demand RHR Heat Exchanger 54,800,000 on the CCS C heat exchanger to mitigate a Page 19 of 64

ENCLOSURE Responses to NRC Audit Review Questions for WBN Unit 1 ERCW and CCS License Amendment Request Item NRC Question/Request TVA Response/Dated Posted No. Date Posted design basis accident (loss of coolant Centrifugal Charging Pump 66,760 accident with a loss of offsite power and RHR Pump 100,000 failure of Train A)? And what is the corresponding required ERCW flow rate and Safety Injection Pump 46,000 CCS flow rate to meet this peak heat removal Radiation Monitor 0 demand?

Containment Spray Pump 14,746 b) What is the concurrent ERCW flow rate Total 55,027,506 required to be sent to the Containment Spray Heat Exchanger 1B or 2B? Train B ERCW Flows are NOT separated by unit. The total required Train B ERCW flow to the CCS HX C for one unit in LOCA-Recirculation and the other unit in Hot c) What is the concurrent required ERCW flow Shutdown for less than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months is provided below.

rate to the operating EDGs?

CCS ERCW Flow d) What is the concurrent required ERCW flow Item Flow to CCS HX rate to the CCS heat exchanger A or B as a (gpm) C (gpm) heat sink for the spent fuel pools? RHR Heat Exchanger 5,000 9,200 e) What is the concurrent required ERCW flow Centrifugal Charging Pump 28 rate to the other safety related ERCW loads RHR Pump 10 for the LOCA unit? Safety Injection Pump 15 Radiation Monitor 6 Containment Spray Pump 2 Total 5,061 b) For one Containment Spray Heat Exchanger: 5,200 gpm c) For two DGs: 2,600 gpm d) CCS HX A: 1,370 gpm e) Concurrent ERCW flow rates to other safety-related loads for LOCA unit:

Page 20 of 64

ENCLOSURE Responses to NRC Audit Review Questions for WBN Unit 1 ERCW and CCS License Amendment Request Item NRC Question/Request TVA Response/Dated Posted No. Date Posted ERCW Item Flow (gpm)

Electric Board Room Chiller (Unit Common) 300.0 Main Control Room Chiller (Unit Common) 240.0 Shutdown Board Room Chiller (Unit 560.0 Common)

Auxiliary Control Air System Compressor 3.5 (Unit Common) 2 X ERCW Pump Cooling (Unit Common) 12.0 2 X ERCW Pump PreLube (Unit Common) 1.6 ERCW Screen Wash (Unit Common) 10.0 2 X ERCW Strainer (Unit Common) 900.0 AFW & Boric Acid Transfer Pump Area 60.0 Cooler (Unit Common)

AFW & Component Cooling System Area 102.0 Cooler (Unit Common)

Emergency Gas Treatment System Room 10.0 Cooler (Unit Common)

Spent Fuel Pool & Thermal Barrier Booster 29.0 Pumps Area Cooler (Unit Common)

Containment Spray Pump Room Cooler 28.0 Centrifugal Charging Pump Room Cooler 25.0 Elevation 692 Penetration Room Cooler 12.0 Elevation 713 Penetration Room Cooler 11.0 Elevation 737 Penetration Room Cooler 12.0 Pipe Chase Room Cooler 15.0 RHR Pump Room Cooler 19.0 Safety Injection Pump Room Cooler 22.0 Total 2,372.1 BOP 1 Clarification Question Spent fuel pool cooling is provided by the 1B-B CCS pump through the A CCS HX which

14. is realigned later in the event. The flow for the spent fuel pool is therefore not included in The total CCS flow is stated to be 5061 GPM to each the C CCS HX flow.

Page 21 of 64

ENCLOSURE Responses to NRC Audit Review Questions for WBN Unit 1 ERCW and CCS License Amendment Request Item NRC Question/Request TVA Response/Dated Posted No. Date Posted unit in BOP-12 and BOP-13.

TVA will submit a license amendment request to maintain Auxiliary Feedwater capability Is additional CCS flow to CCS HX A or B required for in support of TS 3.4.6 Loops Operable requirements for 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months .

Spent Fuel Pool cooling and how much? What will then be the required total CCS flow from the CCS B Train when in the plant conditions described in BOP-12 and BOP-13?

BOP - 14 a) CCS HX C peak heat removal rate for unit in Mode 4 with loss of Train A:

The licensee is requested to provide the following Heat Load Item information: (Btu/hr)

With a loss of Train A: RHR HX 89,265,200 a) What is the peak heat removal rate demand Centrifugal Charging Pump 66,760 on the CCS C heat exchanger to maintain a RHR Pump 100,000 unit in Mode 4 assuming the unit achieved Safety Injection Pump 0 Mode 4 in the minimum amount of time after Containment Spray Pump 0 shutdown? And what is the corresponding Radiation Monitor 0 ERCW flow rate and CCS flow rate to meet Total 89,431,960 this peak heat removal rate assuming steam generators are not in use for heat removal? ERCW flow rate and CCS flow rate to meet peak heat removal rate assuming steam 15.

generators are not in use for heat removal:

b) What is the concurrent required ERCW flow rate to the other safety related ERCW loads Note: Even though certain non-accident unit pumps may not be running (no heat for the Mode 4 unit? load), they may still receive CCS cooling water flow.

CCS ERCW Flow Item Flow to CCS HX (gpm) C (gpm)

RHR Heat Exchanger 5,000 9,200 Centrifugal Charging Pump 28 RHR Pump 10 Safety Injection Pump 15 Radiation Monitor 6 Page 22 of 64

ENCLOSURE Responses to NRC Audit Review Questions for WBN Unit 1 ERCW and CCS License Amendment Request Item NRC Question/Request TVA Response/Dated Posted No. Date Posted Containment Spray Pump 2 Total 5,061 b) Concurrent required ERCW flow rate to the other safety related ERCW loads for the Mode 4 unit:

ERCW Item Flow (gpm)

ERCW Pump Cooling (Unit Common) 6.0 ERCW Pump PreLube (Unit Common) 0.8 ERCW Screen Wash (Unit Common) 10.0 Centrifugal Charging Pump Room Cooler 25.0 Elevation 692 Penetration Room Cooler 12.0 Elevation 713 Penetration Room Cooler 11.0 Elevation 737 Penetration Room Cooler 12.0 Pipe Chase Room Cooler 15.0 RHR Pump Room Cooler 19.0 2 X Upper Containment Vent Cooler (non-safety load)* 46.0 2 X Lower Containment Vent Cooler (non-safety Load)* 612.0 2 X Control Rod Drive Mechanism Cooler (non-safety load)* 248.0 2 X Reactor Coolant Pump Motor Air Cooler (non-safety load)* 220.0 Reactor Building Instrument Room Chiller (non-safety load) 30.0 Total 1,266.8

These non-safety containment coolers are sized for normal power operation.

During Hot Shutdown (Mode 4) and Cold Shutdown (Mode 5), their cooling loads (and flow requirements) are significantly reduced.

BOP 14 - 1 TVA will submit a license amendment request to maintain Auxiliary Feedwater capability Please clarify the simultaneous validity of the in support of TS 3.4.6 Loops Operable requirements for 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months .

following statements as submitted by TVA and

16. respond to the following questions.

x Page E1-5 of the June 17, 2015, letter states that the RHR system is normally placed in service Page 23 of 64

ENCLOSURE Responses to NRC Audit Review Questions for WBN Unit 1 ERCW and CCS License Amendment Request Item NRC Question/Request TVA Response/Dated Posted No. Date Posted four hours after reactor shutdown.

x Page E1-4 of the July 14 submittal states that it will be approximately 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months before RHR is placed in service.

x TVA response to NRC Audit Review Questions BOP-14 states that the peak heat removal rate from the RHR heat Exchanger for a unit in Mode 4 with a loss of Train A is 89,265,200 BTU/hr.

A) Is 89,265,200 BTU/hr applicable to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months after shutdown of the non-accident unit?

B) With RHR in service 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after shutdown of the non-accident unit, how is removal of residual heat from the non-accident unit assured between hours 4 and 7 after shutdown?

BOP - 15 a) CCS HX A or B peak heat removal rate for mitigation of LOCA on one unit with loss of offsite power and loss of Train B:

The licensee is requested to provide the following information: Heat Load Item With a loss of Train B: (Btu/hr) a) What is the peak heat removal rate demand RHR Heat Exchanger 54,800,000 on the CCS A or B heat exchanger to Centrifugal Charging Pump 66,760

17. mitigate a design basis accident (loss of RHR Pump 100,000 coolant accident with a loss of offsite power Safety Injection Pump 46,000 and failure of Train B)? And what is the Containment Spray Pump 14,746 corresponding required ERCW flow rate and Seal Water Heat Exchanger 941,000 CCS flow rate to meet this peak heat removal Non-Regenerative Letdown Heat demand? 0 Exchanger

Sample Heat Exchanger A 0 b) What is the concurrent ERCW flow rate Page 24 of 64

ENCLOSURE Responses to NRC Audit Review Questions for WBN Unit 1 ERCW and CCS License Amendment Request Item NRC Question/Request TVA Response/Dated Posted No. Date Posted required to be sent to the Containment Spray Sample Heat Exchanger B 0 Heat Exchanger 1A or 2A? Sample Heat Exchanger C 0 Hot Sample Chiller Package 0 c) What is the concurrent required ERCW flow Radiation Monitor 0 rate to the operating EDGs? Waste Gas Compressor 135,135 Total 56,103,641 d) What is the concurrent required ERCW flow rate to the CCS heat exchanger A or B as a

There is no heat load on the Non-Regenerative Letdown Heat Exchanger heat sink for the spent fuel pools? during LOCA conditions; however, the flow control valve fails OPEN on loss of air or loss of power so there may be up to 1,000 gpm of flow going through it.

e) What is the concurrent required ERCW flow rate to the other safety related ERCW loads Corresponding required ERCW flow rate and CCS flow rate to meet peak heat for the LOCA unit? removal demand:

CCS Flow ERCW Item (gpm) (gpm)

RHR Heat Exchanger 5,000 4,000 Centrifugal Charging Pump 28 RHR Pump 10 Safety Injection Pump 15 Containment Spray Pump 2 Seal Water Heat Exchanger 200 Non-Regenerative Letdown Heat Exchanger 1,000 Sample Heat Exchanger A 20 Sample Heat Exchanger B 28 Sample Heat Exchanger C 20 Hot Sample Chiller Package 22 Radiation Monitor 6 Waste Gas Compressor 50 Spent Fuel Pool Hx** ~2,000 Total 8,401

There is no heat load on the Non-Regenerative Letdown Heat Exchanger during LOCA conditions; however, the flow control valve fails OPEN on loss of air or loss of power so there may be up to 1,000 gpm of flow going through it.

Page 25 of 64

ENCLOSURE Responses to NRC Audit Review Questions for WBN Unit 1 ERCW and CCS License Amendment Request Item NRC Question/Request TVA Response/Dated Posted No. Date Posted

- CCS Train 2A only (flow exists however it is not required) b) For one Containment Spray Heat Exchanger: 5,200 gpm c) For two DGs: 2,600 gpm d) This is not separately available. The required ERCW flow is included in the non-accident CCS Heat Exchanger flow, i.e., 5,050 gpm for > 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months ; 7,100 gpm for < 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

e) Concurrent required ERCW flow rate to the other safety related ERCW loads for the LOCA unit:

ERCW Item Flow (gpm)

Electric Board Room Chiller (Unit Common) 300.0 Main Control Room Chiller (Unit Common) 240.0 Shutdown Board Room Chiller (Unit Common) 560.0 Auxiliary Control Air System Compressor (Unit Common) 3.5 2 X ERCW Pump Cooling (Unit Common) 12.0 2 X ERCW Pump PreLube (Unit Common) 1.6 ERCW Screen Wash (Unit Common) 10.0 2 X ERCW Strainer (Unit Common) 900.0 AFW & Boric Acid Transfer Pump Area Cooler (Unit Common) 60.0 AFW & Component Cooling System Area Cooler (Unit Common) 102.0 Emergency Gas Treatment System Room Cooler (Unit Common) 10.0 Spent Fuel Pool & Thermal Barrier Booster Pumps Area Cooler (Unit 29.0 Common)

Containment Spray Pump Room Cooler 28.0 Centrifugal Charging Pump Room Cooler 25.0 Elevation 692 Penetration Room Cooler 12.0 Elevation 713 Penetration Room Cooler 11.0 Elevation 737 Penetration Room Cooler 12.0 Pipe Chase Room Cooler 15.0 Page 26 of 64

ENCLOSURE Responses to NRC Audit Review Questions for WBN Unit 1 ERCW and CCS License Amendment Request Item NRC Question/Request TVA Response/Dated Posted No. Date Posted RHR Pump Room Cooler 19.0 Safety Injection Pump Room Cooler 22.0 Total 2,372.1 BOP - 16 a) CCS HX A or B peak heat removal rate to maintain one unit in Mode 4 with loss of Train B:

The licensee is requested to provide the following information: Heat Load Item With a loss of Train B: (Btu/hr) a) What is the peak heat removal rate demand RHR Heat Exchanger 89,265,200 on the CCS A or B heat exchanger to Centrifugal Charging Pump 66,760 maintain a unit in Mode 4 assuming the unit RHR Pump 100,000 achieved Mode 4 in the minimum amount of Safety Injection Pump 0 time after shutdown? And what is the Containment Spray Pump 0 corresponding ERCW flow rate and CCS flow Seal Water Heat Exchanger 517,000 rate to meet this peak heat removal rate Non-Regenerative Letdown Heat Exchanger 2,530,000 assuming steam generators are not in use for Sample Heat Exchanger A 0 heat removal?

Sample Heat Exchanger B 0 Sample Heat Exchanger C 0

18. b) What is the concurrent required ERCW flow Hot Sample Chiller Package 0 rate to the other safety related ERCW loads for the Mode 4 unit? Radiation Monitor 0 Waste Gas Compressor 0 Spent Fuel Pool Heat Exchanger 21,000,000 Total 113,478,960 Corresponding ERCW flow rate and CCS flow rate to meet peak heat removal rate assuming steam generators are not in use for heat removal:

CCS ERCW Item Flow Flow (gpm) (gpm)

RHR Heat Exchanger 5,000 7,100 Centrifugal Charging Pump 28 RHR Pump 10 Page 27 of 64

ENCLOSURE Responses to NRC Audit Review Questions for WBN Unit 1 ERCW and CCS License Amendment Request Item NRC Question/Request TVA Response/Dated Posted No. Date Posted Safety Injection Pump 15 Containment Spray Pump 2 Seal Water Heat Exchanger 200 Non-Regenerative Letdown Heat Exchanger 1,000 Sample Heat Exchanger A 20 Sample Heat Exchanger B 28 Sample Heat Exchanger C 20 Hot Sample Chiller Package 22 Radiation Monitor 6 Waste Gas Compressor 50 Spent Fuel Pool Heat Exchanger 2,000 Total 8,401 b) Concurrent required ERCW flow rate to the other safety related ERCW loads for the Mode 4 unit:

ERCW Item Flow (gpm)

ERCW Pump Cooling (Unit Common) 6.0 ERCW Pump PreLube (Unit Common) 0.8 ERCW Screen Wash (Unit Common) 10.0 Centrifugal Charging Pump Room Cooler 25.0 Elevation 692 Penetration Room Cooler 12.0 Elevation 713 Penetration Room Cooler 11.0 Elevation 737 Penetration Room Cooler 10.0 Pipe Chase Room Cooler 15.0 RHR Pump Room Cooler 19.0 2 X Upper Containment Vent Cooler (non-safety load)* 46.0 2 X Lower Containment Vent Cooler (non-safety Load)* 612.0 2 X Control Rod Drive Mechanism Cooler (non-safety load)* 248.0 2 X Reactor Coolant Pump Motor Air Cooler (non-safety load)* 220.0 Reactor Building Instrument Room Chiller (non-safety load) 30.0 Total 1,264.8 Page 28 of 64

ENCLOSURE Responses to NRC Audit Review Questions for WBN Unit 1 ERCW and CCS License Amendment Request Item NRC Question/Request TVA Response/Dated Posted No. Date Posted

These non-safety containment coolers are sized for normal power operation.

During Hot Shutdown (Mode 4) and Cold Shutdown (Mode 5), their cooling loads (and flow requirements) are significantly reduced.

BOP - 17 The 48 hour5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months time-delay was selected due to limitations in the CCS and the ERCW System to simultaneously mitigate a design basis LOCA on one unit and remove core decay heat The licensee has stated on page E1-10 of the June from the non-accident unit when the non-accident unit was relying on RHR to remove core 17, 2015, submittal: decay heat. After 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , the non-accident unit core decay heat is sufficiently low

(~ 56.7 MBtu/hr) that the CCS and the ERCW System can support both the accident and once Unit 1 has been shutdown for 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months non-accident units with any single active failure. The actual time delays vary depending or more, the total ERCW heat removal and on system availability and the single active failure postulated between CCS and the

19. thus, flow requirements, drop below the ERCW System. The most limiting time delay is 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months due to the availability of only one flowrate provided by two ERCW pumps. CCS pump aligned to CCS HX C.

Provide the reasons for the 48 hour5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months time by identifying the heat load on the ERCW system from the shutdown unit and from the LOCA unit at time 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months and explain how two ERCW pumps provides adequate flow at this time.

BOP - 18 The scenario postulates a case where a single failure does not remove a complete train of electrical power and poses the question whether a second ERCW pump would be The licensee has stated on page E1-11 of the June required on a single train to supply sufficient flow. This scenario has not been explicitly 17, 2015, submittal: analyzed. However, for this event, three ERCW pumps would be available for cooldown, with two ERCW pumps on one train and one ERCW pump on the other train. In addition, The requirement to have two ERCW pumps at least three CCS HXs are available with multiple CCS pumps. Qualitatively, the running on one DG is required for the following table compares the two scenarios:

scenario of a LOCA on one unit and the other

20. unit cooled by RHR within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months of 3 ERCW Pumps on 3 ERCW Pumps Item shutdown. The single failure of a loss of a 1 Train on 2 Trains train of power must also occur to require two ERCW Pumps ERCW pumps to be loaded on a single DG. 3 3 Available Other single failures including the loss of a High Flow Velocities Low Flow Velocities DG or a 6.9 kV shutdown board will not ERCW Flow

/ Resistance / Resistance require two ERCW pumps to be loaded on a CCS HXs Available 1 or 2 3 single DG.

Page 29 of 64

ENCLOSURE Responses to NRC Audit Review Questions for WBN Unit 1 ERCW and CCS License Amendment Request Item NRC Question/Request TVA Response/Dated Posted No. Date Posted For the above scenario with a loss of a single Train A CCS Pump 3 maximum 3 or 4 DG or Train A ERCW pump, where one Train A Available ERCW pump is running and two Train B ERCW High Flow Velocities Lower Velocity for pumps are running: CCS Flow for Train B Train B RHR HXs Available 2 maximum 3 or 4 Has this scenario been analyzed to demonstrate that SFP HXs Available 0 for Train B 2 a second ERCW pump on a DG will not be required as stated? If so, describe the assumptions and By comparison, it can be seen that the failure of one DG or one ERCW pump results in results of this analysis. more equipment available than in the limiting case of a loss of an entire train of 6.9 kV shutdown boards. Therefore, conditions are more favorable for cooldown.

BOP - 19 The adequacy of two ERCW pumps per train and one CCS pump per train to removal all decay heat from both units for the scenarios provided by the NRC cannot be assured The licensee has stated on page E1-4 of the July 14, under all worst case conditions. Therefore, the Applicability Note of TS 3.7.16 and 2015 letter: TS 3.7.17 will be revised by the removal of part b regarding the condition of the opposite unit. The proposed changes to TS 3.7.16 and TS 3.7.17 will be provided in a separate When the assumptions include a loss of offsite submittal.

power and the loss of Train A power, two CCS pumps need to be aligned to the CCS Train B In light of this information and the removal of TS 3.7.16 and TS 3.7.17 Applicability Note b, header and in operation when RHR is in service on the third paragraph of the response to NRC Acceptance Review Question 2 provided in both units and both units have been shutdown for TVA letter to NRC, dated July 14, 2015, page E1-4, is revised as follows:

less than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

When the assumptions include a loss of offsite power and the loss of Train A power,

21. TVA agrees that the submittal did not provide two CCS pumps need to be aligned to the CCS Train B header and in operation when much discussion of the non-accident case because RHR is in service on both units and both units have either unit has been shutdown for the LOCA plus shutdown case is more limiting. less than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

There is discussion on pages E1-10 and E1-11 of the license amendment with respect to required ERCW pumps.

The licensee has stated, in effect, if one unit has been shutdown for 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or greater and the other unit has reached Mode 4 at the earliest opportunity, then TS 3.7.7 and TS 3.7.8 are sufficient without proposed TS 3.7.16 and TS 3.7.17.

Page 30 of 64

ENCLOSURE Responses to NRC Audit Review Questions for WBN Unit 1 ERCW and CCS License Amendment Request Item NRC Question/Request TVA Response/Dated Posted No. Date Posted Accordingly, the staff requests TVA to address the heat removal requirements for Unit 2 shutdown at 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months and Unit 1 having just reached Mode 4 (approximately 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months after shutdown) and the following scenarios:

1) DBA Unit 2 (LOCA/LOOP loss of Train A or B)

Are the one CCS pump and 2 ERCW pumps for Train A or Train B sufficient to mitigate the DBA and maintain cool down on Unit1?

Identify associated heat loads and CCS and ERCW flowrates.

2) DBA (LOOP/Loss of Train A or B)

Are the one CCS pump and 2 ERCW pumps for Train A or Train B sufficient to maintain cool down on both units? Identify associated heat loads and CCS and ERCW flowrates.

BOP - 20 The values in FSAR Table 5.5-8 represent the design point of the RHR Heat Exchangers. This is just one set of conditions under which the RHR Heat Exchangers can The response to BOP-13 part (a) listed the heat load operate. With one (1) RHR Heat Exchanger removing a core decay heat of ~ 89.4 on the RHR heat exchanger as 89,265,200 Btu/hr MBtu/hr, one set of flow and temperature conditions for Loss of Train A and Loss of Train with a CCS flow rate of 5000 gpm. FSAR Table 5.5-8 B, each, are as follows:

(RESIDUAL HEAT REMOVAL SYSTEM COMPONENT DATA) lists the design heat removal Loss of Train A Loss of Train B capacity of the RHR heat exchanger as 37,400,000 Parameter EPMJN010890, EPMJN010890,

22. Btu/hr with a CCS design flow of approximately 5000 Table C7.7.89 Table C7.5.9 GPM. CCS Inlet 108 105 Temperature (°F)

Explain how the RHR HX will transfer 89,265,200 CCS Outlet 144 141 Btu/hr as listed in BOP-13, including inlet and outlet Temperature (°F) temperatures on the shell and tubes sides and CCS CCS Flow (Mlbm/hr) 2.48 2.48 and RC flow rates. CCS Flow (gpm) 5,000 5,000 RHR Inlet 239 247 Temperature (°F)

Page 31 of 64

ENCLOSURE Responses to NRC Audit Review Questions for WBN Unit 1 ERCW and CCS License Amendment Request Item NRC Question/Request TVA Response/Dated Posted No. Date Posted RHR Outlet 153 148 Temperature (°F)

RHR Flow (Mlbm/hr) 1.05 0.90 RHR Flow (gpm) 2,146 1,828 Core Decay Heat 89.4 89.4 (MBtu/hr)

UA (MBtu/(hr - °F)) 1.48 1.43 During actual plant operations, the CCS flow is set to approximately 5,000 gpm and the operators control the rate of cooldown and CCS outlet temperature from the RHR Heat Exchanger by manually throttling the RHR flow rate. All parameters (temperature, pressure, flow rate) are within the RHR Heat Exchanger design conditions as shown on the Vendor datasheets.

BOP - 21

a. Based on Westinghouse analysis sensitivity runs using a UA of 2.0, the containment TVA Calculation EPMJN010890 Revision 19, pressure change is minimal (11.73 to 11.76 psig). TVA will docket the sensitivity Performance of CCS Heat Exchanger Appendix analysis in support of the UA used in the TVA Calculation EPMJN010890 Revision 19, C, Table C7.7.69, and Appendix E, Tables E1 and Performance of CCS Heat Exchanger Appendix C, Table C7.7.69, and Appendix E, E2. Tables E1 and E2.

(a) The total heat load on the CCS heat exchanger The purpose of Appendix E was to develop a representative Heat Exchanger UA (HX) C stated in Table C7.7.69 is 144.72 MBtu/hr value for use by Westinghouse in containment analyses. Since the Watts Bar C HX (sum of 89.4 MBtu/hr shutdown unit load, is shared between the two units, Westinghouse needed to know what the effective 54.8 MBtu/hr LOCA unit load, and 0.53 MBtu/hr part of the heat exchanger was supporting the unit having the LOCA. Appendix E miscellaneous loads) is more limiting than the apportions the C HX UA value according to the percent mass flow going to the heat 23.

total heat load of 112.03 MBtu/hr given in Table exchanger from each unit.

E1. Explain why the data in Table E1 is used to calculate the UA for the virtual CCS HXs for Appendix E1 was started prior to the project to add a third ERCW pump to support the containment analysis and shutdown cooling Hot Shutdown / LOCA-Recirculation mode of operation within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months of a unit analysis. shutdown. The flows and heat loads reflect two ERCW pumps in operation and after 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months from shutdown. Heat Exchanger UA values are a function of heat (b) In Table E2, the CCS virtual HX assigned for exchanger geometry and flow and are not dependent on heat duty on the device.

containment analysis has UA = 3.17 MBtu/hrºF Using the lower two ERCW pump flow of 7,125 gpm would produce lower UA values which is consistent with the value used in than the use of the higher 9,200 MBtu/hr flow from three ERCW pumps. Note that the Westinghouse containment analysis report in LOCA containment analysis must support both two and three ERCW pumps in-Reference 1. What is the UA for the CCS virtual service.

HX assigned for shutdown cooling and how is it Page 32 of 64

ENCLOSURE Responses to NRC Audit Review Questions for WBN Unit 1 ERCW and CCS License Amendment Request Item NRC Question/Request TVA Response/Dated Posted No. Date Posted calculated based on the real CCS HX UA = During discussions with NRC reviewers, a question was posed as to the conservatism 6.44 MBtu/hr-ºF? of using a mass flow based allotment of UA. A counter proposal was made that the allotment should be made based on the heat load from each unit.

(c) Provide the analysis that shows that the calculated value of shutdown unit virtual HX CCS The spreadsheets at the end of this enclosure provide a comparison between the two UA (based on UA = 3.17 MBtu/hr-ºF for the CCS methods, using 9,200 gpm total ERCW flow. These spreadsheets will be added to LOCA virtual HX and 6.44 MBtu/hr-ºF for the real Calculation EPMJN010890, Appendix E.

CCS HX) can handle the shutdown cooling load of 89.4 MBtu/hr plus 0.53 MBtu/hr miscellaneous c. Response to a and b will address c. Results of a and b will also be provided.

loads.

b. Add data sheets reflecting virtual CCS HX performance based on CCS Flow and CCS (d) Table C7.7.69 states the RHR HX heat load for heat load. Include sensitivity analysis based on ERCW flow based on heat load.

the shutdown unit is 89.4 MBtu/hr. At what time during the shutdown transient does this load Using the larger 9,200 gpm ERCW flow the real HX UA is 6.82 MBtu/hr-F. With the occur? HX apportioned by Flow, the shutdown cooling HX UA is 3.35 (the same as the LOCA unit). With the HX apportioned by Heat Load, the shutdown cooling HX UA is (e) Table E1 states RHR HX heat load for shutdown 4.12 MBtu/hr-F. It is calculated the same as above for the LOCA unit.

unit s 56.7 MBtu/hr. At what time during the shutdown transient does this load occur? c. This is demonstrated by Table C7.7.69 and is reproduced in the first table above for a total ERCW flow of 9,200 gpm. All ERCW and CCS temperatures are acceptable.

(f) Tables C7.7.69 and E1 states the RHR HX heat Note: the miscellaneous heat load and flow gets its own virtual HX with UA load for the LOCA unit is 54.8 MBtu/hr. apportioned by either Flow or Heat Load.

Please confirm this is the heat load at the d. 7 - hours initiation of the RHR sprays assumed to start operating at 3600 second from the LOCA e. 48 - hours initiation.

f. In TVAs calculation, the 54.8 MBtu/hr is assumed from initiation of recirculation mode (g) At what time does the LOCA occur in relation to (~40 minutes assuming the loss of train event).

the initiation of the shutdown transient assumed in Tables C7.7.69 and E1? g. For GDC-5 analysis, the LOCA occurs 7-hours after the initiation of a shutdown on the non-accident unit.

(h) Tables C7.7.69 and E1, miscellaneous heat load 0.53 MBtu/hr is for which unit. Confirm whether h. 0.53 MBtu/hr is the combined miscellaneous heat load for two units. The this is a combined miscellaneous load for both miscellaneous loads are accounted for using an additional virtual heat exchanger.

units, and if so, how much is imposed on each Page 33 of 64

ENCLOSURE Responses to NRC Audit Review Questions for WBN Unit 1 ERCW and CCS License Amendment Request Item NRC Question/Request TVA Response/Dated Posted No. Date Posted CCS virtual HX. i. The limiting temperature is 146°F CCS temperature at the outlet of the RHR HX (analyzed pipe stress limit). The Operating Modes calculation EPMJK022988 shows (i) What is the maximum operating temperature of that CCS temperatures through the RHR HX should not exceed 146ºF. The stress the CCS fluid under the condition of the most calculation is N3-70-04A.

limiting heat load on the RHR HX C?

j. See EPMJN010890, Section 7.1, Sheet 31 and 32.

(j) In Tables C7.7.69, E1, and E2, explain the difference between the design and actual values k. See EPMJN010890, Section 6.1 for ho and hi for the CCS and RHR HXs.

l. See EPMJN010890, Section 7.0.2, Equation 10 and Section 7.3, Sheets 41 and 42.

(k) Refer to Section 6.1.4 and 6.1.5 of the calculation which states the HX tubes were changed from 90-10 copper nickel to stainless steel. Confirm that the CCS U and UA given in Tables C7.7.69 and E1 are based on the as-built HX material thermal conductivity, tube thickness, worst fouling resistance and tube plugging.

(l) In Tables C7.7.69, E1, and E2, what is represented by F, R, and S and the HX correction factor r, and p?

BOP - 22 As documented in MDQ00006720080341 Appendices 11 (two ERCW Pumps) and 17 (three ERCW Pumps), MULTIFLOW demonstrates that the ERCW pumps develop In BOP-13-16, the licensee listed the design heat sufficient head to supply all required users during loss of downstream dam conditions.

loads and corresponding required flow rates for CCS and ERCW during a DBA and loss of Train A or Train Calculation EPM-WUC-072489 demonstrates that three ERCW pumps have required B. One of the design basis conditions is a loss of NPSHA for loss of downstream dam conditions.

24.

downstream dam.

Describe and justify how the ERCW design accounts for a loss of downstream dam during a DBA with the equipment and system lineups that are specified in both LCOs of TS 3.7.8 and proposed TS 3.7.17.

ELEC 3 Response Summary from TVA Letter to NRC dated August 3, 2015:

25.

As documented in NUREG-0847, "Safety Evaluation Page 34 of 64

ENCLOSURE Responses to NRC Audit Review Questions for WBN Unit 1 ERCW and CCS License Amendment Request Item NRC Question/Request TVA Response/Dated Posted No. Date Posted Report Related to the Operation of Watts Bar Nuclear Q1 Plant, Unit 2," Supplement 22, (SSER 22) published With offsite power available, there is no change to the licensing basis documented in February 2011, the licensing basis of Watts Bar SSER 22. Changes to the cases in the calculations for the dual unit shutdown with offsite Nuclear Units is: power available were not revised in association with the June 17, 2015 LAR.

1. Dual-unit trip as a result of an abnormal The SI pump, Containment Spray pump, and AFW pump combined horsepower is operational occurrence approximately 1660 horsepower. This is approximately twice the horsepower requirement

2. Accident in one unit and concurrent shutdown of of an ERCW pump. Since these three pumps are not running for the GDC-5 case, there is the second unit (with and without offsite power) considerable margin compared to the limiting case thus demonstrating that the start of a

3. Accident in one unit and spurious engineered second ERCW pump on a non-accident shutdown board is acceptable and bounded by safety feature actuation in the other unit (with and the LOCA/inadvertent SI case.

without offsite power)

The evaluations of CSSTs A and B are included in WBN calculation The license amendment request (LAR) for Watts Bar EDQ00099920070002, AC Auxiliary Power System Analysis. TVA submitted a response Nuclear Plant (WBN) Unit 1, dated June 17, 2015 to NRC Open Items from SSER 22 on April 6, 2011 for WBN Unit 2. This submittal (Agencywide Documents Access and Management described the margin studies done for all four CSSTs. The margin studies that TVA System (ADAMS) Accession No. ML15170A474) provided in the submittal were discussed in SSER 24 in relation to the closure of SSER proposes realignment of Component Cooling System Open Items 27 and 28. Because the LOCA on one unit with an inadvertent SI on the other pumps and Essential Raw Cooling Water (ERCW) unit results in a higher load than the scenario discussed in the LAR, additional margin Pumps to support heat removal capability. studies were not required.

The staff is also reviewing a LAR submitted by letter Q2 dated August 1, 2013 (ADAMS Accession No. The DG loading for the first 20 minutes is the base case loading described in the WBN ML13220A103), to modify limiting conditions for Calculation EDQ00099920080014, Diesel Generator Loading Analysis. The CCS pumps operation for Technical Specification 3.8.1, AC are assumed to start and run in the base case, so the proposed amendment related to Sources - Operating, for the available maintenance CCS does not represent a change from an electrical standpoint. The loading of a second feeder for Common Station Service Transformers ERCW pump on an individual DG occurs no earlier than 40 minutes after DG start. This is (CSSTs) A and B. why the base case applies for the first 20 minutes.

The staff is requesting clarification on loading of Table 3 of the June 17, 2015 submittal represents the bounding cases after 20 minutes.

onsite and offsite power systems and has determined The values in the table provide the horsepower assumed for each of the large motor loads that the following additional information is needed to for each available DG and provide the total kW loading on each available DG for the complete the review of the LAR: scenarios in Table 3. Attachment 1 of that letter provided excerpts from the DG loading calculation including tables that summarized the loads on each DG for a variety of cases

1. For the scenarios related to dual unit shutdown including Loss of Offsite Power (LOOP) / LOCA and dual unit cases.

with offsite power system available, please Page 35 of 64

ENCLOSURE Responses to NRC Audit Review Questions for WBN Unit 1 ERCW and CCS License Amendment Request Item NRC Question/Request TVA Response/Dated Posted No. Date Posted provide a summary of the calculations performed Q3 to evaluate capability of offsite power Before a second ERCW pump can be loaded on its DG, the AFW Pump, if running, will be transformers CSST A and B for the licensing stopped and the main control room hand switch placed in pull-to-lock. This action assures basis documented in SSER 22. Provide the that the AFW pump will not inadvertently start to preclude overloading the DG. TVA impact of changes proposed in the June 17, 2015, currently plans to include these actions in the same procedure that starts the second LAR on the licensing basis documented in SSER ERCW pump. The actions will be placed in a step that precedes the start of the second

22. ERCW pump. This is the only load shed assumed in the DG loading analysis.

2. Table 3 in Enclosure 1 of the WBN Unit 1, LAR dated June 17, 2015, contains the Summary of Steady-State Diesel Generator (DG) Loading with 3 ERCW Pumps (>20 minutes) only.

Please clarify whether the electrical system loadings considered in Table 3 of the LAR is bounding for all the scenarios without offsite power addressed in SSER 22 summarized above. For the scenarios related to shutdown using onsite power systems, please provide details (calculations or explanation) related to large motor loads (Rating and horse power value) considered for the specific scenarios. Provide details of additional kilo-Watt (kW) loading considered in the total kW loading of each DG. Also provide DG loadings during 20 minutes.

3. Please provide details on any load shedding that may be procedurally controlled to preclude overloading the power source(s).

Date Posted: 07/31/15 ELEC - 4 A revision to the DG loading calculation has been completed. Discrepancies between the revised calculation and Enclosure 1, Table 3 of the June 17, 2015 letter have been Resolve the apparent discrepancy between the DG resolved. An updated version of the DG loading calculation has been posted on

26. loading calculation, Appendix N for AFW load on Sharepoint. An updated version of Enclosure 1, Table 3 is included at the end of this LOCA unit and the June 17, 2015 Enclosure 1, Table enclosure.

3 AFW load indicated on the LOCA unit (i.e., 600 hp versus 300 hp).

Page 36 of 64

ENCLOSURE Responses to NRC Audit Review Questions for WBN Unit 1 ERCW and CCS License Amendment Request Item NRC Question/Request TVA Response/Dated Posted No. Date Posted Also update Enclosure 1, Table 3 to reflect other corrections made to DG loading Calculation.

ELEC - 5 A revision to the DG loading calculation is in progress with an outcome that is expected to be favorable. An updated version of the DG loading calculation has been posted on Evaluate the differences in DG loading for a small Sharepoint.

27. break LOCA. Is the large break LOCA assumed in the GDC 5 analysis bounding?

HF - 1 The first cue provided to the operating staff is a procedural step in E-1, Loss of Reactor or Secondary Coolant to check the status of electrical power. Should a complete loss of Describe what cues will be provided to the operator either train of 6.9 kilovolt (kV) shutdown board be detected in this step, the procedure will indicating that new and changed manual actions (as require that additional actions be taken.

described in TVAs response to NRC Acceptance Review Question 5, Item 1) are required. In your The first action will be to determine if the remaining train of power is supplied from offsite response, identify the specific plant condition, or DG source. If the offsite power source is supplying the bus, the operator is directed to annunciator status, associated alarms, and start an additional ERCW pump associated with the shutdown units 6.9 kV shutdown procedure steps that will provide instructions to the board that remains powered. Depending on the train of power lost, the operations staff operator. Further, identify information that is required may be required to start an additional CCS pump.

to inform the operator that these manual actions have been correctly performed, and that they can be Should the source of power remaining be the DGs, the operator is directed to perform terminated. actions in accordance with an appendix to the E-1 procedure. This appendix is handed off 28.

to the shutdown unit.

Date Posted: 07/31/15 The shutdown unit will first determine if RHR cooling is in service. If RHR cooling is NOT in service, the shutdown unit is directed to secure plant cooldown and maintain current plant temperature. An additional action is specified to perform throttling of ERCW cooling flow through the CCS heat exchanger (HX), if required due to the specific loss of power.

If RHR cooling is in service, actions are contained in the Appendix to: place the motor driven AFW (MDAFW) pump hand switch in pull-to-lock (PTL) if the turbine driven AFW (TDAFW) pump is in operation, dispatch an operator to the 6.9 kV shutdown board to place the ERCW bypass switch in the bypass condition and start the ERCW pump when 20 minutes have elapsed.

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ENCLOSURE Responses to NRC Audit Review Questions for WBN Unit 1 ERCW and CCS License Amendment Request Item NRC Question/Request TVA Response/Dated Posted No. Date Posted The main control room (MCR) will be alerted that the bypass switch has been placed in bypass by a MCR annunciator that alarms when the switch is placed in bypass. Should the field operator position the wrong switch, the MCR staff would become aware of this fact upon attempting to start the ERCW pump. The expected indications, breaker position lights, pump amps, discharge pressure and flow, would not be observed. This would cue the MCR staff to request that the field operator verify the correct switch had been operated.

An additional cue is provided in a note contained in the main body procedure response of E-1 that the additional ERCW pump must be placed in service prior to aligning containment sump recirculation on the accident unit.

Once the additional alignments are put in place by E-1, they will remain in place until a determination is made by the Emergency Response Organization (ERO) that conditions no longer require their operation. In this event, the Technical Support Center (TSC) will be staffed and the Shift Manager (SM) will transfer Site Emergency Director duties to the TSC. Since the continued need of the additional alignments will vary depending on the specific conditions at the start of the event, no attempt was made to proceduralize securing the alignments. The SM will consult with the TSC to determine that plant conditions are such that the alignments are no longer required.

HF - 2 In this case, the accident unit will be considered to be in the lead during the event since the emergency operating procedure (EOP) for the accident unit is the driver for actions Describe how the operators of each of two units will required. Upon EOP entry, the accident units Unit Supervisor (US) will perform a crew be informed of the status of the other unit, and how update. Although this update is primarily intended to focus the attention of the particular their actions will be coordinated. Clarify if one of the units crew, the common design of the Watts Bar Nuclear Plant (WBN) MCR allows either two units will be put in lead, and what events will units operating staff to hear a crew update performed on either unit. In addition, the many result in changes to the chain of command. alarms that are received during a LOCA will be immediately noticed by the shutdown unit,

29. so there is no potential that the shutdown unit will not realize that conditions have Date Posted: 07/31/15 degraded on the accident unit.

Upon EOP entry, the standard practice at WBN is to recall the Shift Technical Advisor (STA) and the SM to the MCR, if they are not currently present. This is practiced routinely during operator requalification training by removing all persons from the simulator except for the minimum staffing requirements. A condition is then inserted on the simulator, and the operators remaining in the MCR are required to recall the rest of the staff to the Page 38 of 64

ENCLOSURE Responses to NRC Audit Review Questions for WBN Unit 1 ERCW and CCS License Amendment Request Item NRC Question/Request TVA Response/Dated Posted No. Date Posted simulator to combat the casualty.

The standard chain of command will remain in effect during this event. The SM retains overall oversight and will have overall responsibility for ensuring dual unit activities are adequately prioritized and supported. The STA provides additional oversight and backup that the accident crew is taking appropriate actions based on the plant conditions at the accident unit. The accident unit US directs the crews response in the EOP. The shutdown unit US directs the procedural activities required for shutdown.

The accident unit operating staff will conduct the electrical power monitoring activities directed in procedure E-1. When a complete loss of one train of power is detected, the US will direct the appropriate mitigating steps based on the existing conditions. The duties required of the shutdown unit will be directed by handing off the attachment of E-1 that contains the necessary actions.

In this event, the communication between the units will be provided via US communication or the SM. The nature of this event is that a large number of alarms will occur, so the fact that something unusual has happened to the accident unit will be readily apparent to the shutdown unit. It is predictable that the annunciators, coupled with the observed level of activity on the accident unit, will prompt one of the shutdown units operators to be dispatched to gather information on what is happening on the accident unit. It would also be the expectation that if the shutdown units condition allows, one of the Unit Operators (UOs) from the shutdown unit would function to support mitigating activities on the accident unit.

HF - 3 The Institute of Nuclear Power Operations (INPO) Operating Experience (OE) database and the TVA OE database were searched for industry events associated with ERCW, Describe if TVA identified any relevant, pre-existing CCS and dual unit operations. The list obtained from these searches was reviewed to performance issues associated with procedural determine lessons that might need to be incorporated in the development of this design guidance, training, and operator manual actions change. In addition, WBN benchmarked both Sequoyah and Browns Ferry for lessons

30. related to starting and stopping ERCW and CCW learned on dual unit operations. This benchmarking ultimately resulted in the generic pumps and operating switches on the 6.9 kV post-accident response that was developed in coordinating the WBN actions for this shutdown board or issues associated with dual unit postulated event.

operation at other sites.

The most common identified OE associated with dual unit operation of equipment involved Date Posted: 07/31/15 plant transients due to operation of the wrong unit component. In this case, the changes Page 39 of 64

ENCLOSURE Responses to NRC Audit Review Questions for WBN Unit 1 ERCW and CCS License Amendment Request Item NRC Question/Request TVA Response/Dated Posted No. Date Posted at WBN do not fit this common error mode, in that the ERCW is common unit equipment, so that there is no potential for operating the correct component on the wrong unit. The method chosen for alignment of the second train B CCS pump (normally from the shutdown unit and operated by the shutdown units staff) also provides protection from wrong unit equipment concerns. The CCS pump will only be started on a LOCA accompanied with a loss of train of electrical power. In this case, the B train CCS pump for the accident unit will automatically start, so the chances that the operating unit staff will attempt to manipulate this component are nonexistent.

Another common identified OE involved misoperation of the expected component. This condition is precluded by the WBN design in that the MCR staff will be alerted that the interlock bypass switch has been placed in the correct position by MCR annunciation.

HF - 4 During development of this design change, operations, training and engineering personnel worked to determine the final product that would be installed in the plant. Operations is a Describe any changes that were required of the quorum member of all Design Change Notice (DCN) meetings in order to ensure that Control Room task analysis that was done as part of operations provides input into all design changes implemented in the plant.

TVAs Detailed Control Room Design Review. If no update to the task analysis was necessary, describe No new switches or components were added in the MCR for this issue. Components how task requirements were developed for the operated in the MCR are: 1 CCS pump, 1 ERCW pump, 1 AFW handswitch and ERCW identified new and changed operator actions. flow control valves (FCVs) for CCS HX. These components are already routinely Describe what reasonable or credible potential errors operated by the MCR staff for multiple normal alignments and casualty situations.

associated with the new and changed operator actions were identified during task analysis. A new annunciator designed consistent with current design requirements, MCR

31. standards, and NUREG-0700 requirements has been added to the MCR. The WBN Date Posted: 07/31/15 Annunciator Response Instructions contain the response to this annunciator, consistent with the method contained in the existing annunciator response guidance.

An additional bypass switch has been added to the 6.9 kV shutdown boards to allow starting the third ERCW pump if DG power is all that is available. If required, a Nuclear Assistant Unit Operator (NAUO) will be dispatched to the area to operate the switch. The switch itself is new, but the task of aligning electrical board transfer switches is one that is routinely performed by the NAUOs. The new switches most closely resemble the Appendix R transfer switches that are located on multiple electrical components that are currently operated by NAUOs when required.

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ENCLOSURE Responses to NRC Audit Review Questions for WBN Unit 1 ERCW and CCS License Amendment Request Item NRC Question/Request TVA Response/Dated Posted No. Date Posted The interlock bypass switches are labeled consistent with other plant equipment as required by TI-12.14, Replacement and Upgrade of Plant Component Identification Tagging and Labeling. The positions of the switch are clearly discernable. Attached to this response are several pictures of the new switches to convey location and labeling that is installed in the plant.

When installed, the switches for each electrical board will be checked and independently verified to be in the correct position at least once per month. The switches will be added to the DG standby checklist which is performed after each monthly DG surveillance or for any evolution that has removed the DG from service. This will require 8 NAUOs to view the switches and their positions locally each month. This frequent check, coupled with standard labeling and similarity of this task to those already performed by the NAUOs, gives the station great confidence that switch operation can be successfully performed by the NAUOs.

HF - 5 The increase in workload due to this license amendment is considered well within the existing capabilities of the current required minimum staffing complement. The actions Describe any increase in operator workload that will added are: monitoring status of electrical power and taking the appropriate actions to start occur with the proposed license amendment an ERCW/CCS pump.

Date Posted: 07/31/15 If offsite power is supplied to the remaining electrical power train, additional actions are limited to starting the ERCW pump, and potentially the B train CCS pump. An additional potential action is throttling CCS HX flows if required. These activities can be performed by one individual in a 1 to 2 minute timeframe.

32. If DG power is the only source to the remaining electrical train, actions in addition to the ones above are required. These actions will have minimum impact on the accident unit operating staff as the Attachment directing these actions will be handed off to the shutdown unit to complete. The actions of the attachment to prepare to start an additional ERCW pump will require dispatching an NAUO to the 6.9 kV shutdown board to place the bypass switch in the bypass position. The switch positioning and access to the board room area will require approximately 3 minutes and 35 seconds as outlined in the previously submitted dose assessment. The shutdown unit will then start an additional ERCW pump after 40 minutes has elapsed.

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ENCLOSURE Responses to NRC Audit Review Questions for WBN Unit 1 ERCW and CCS License Amendment Request Item NRC Question/Request TVA Response/Dated Posted No. Date Posted HF - 6 NPG-SPP-09.3, Plant Modifications and Engineering Change Control, outlines the process in which the bypass switches were designed and installed. In this particular Describe the process used to design the ERCW instance, DCN 53785 is being utilized to install the ERCW pump interlock bypass pump interlock bypass switch. In your response, switches.

explain if the switch complies with control room standards and the applicable guidance of NUREG- The bypass switches are not augmented with status lights. As shown on the pictures at 0700, Human-System Interface Design Review the end of this enclosure, the positions of bypass and normal are clearly labeled. When Guidelines. Further, describe how the bypassed and the switch is taken to bypass, an alarm is received in the MCR. Thus, when the MCR non-bypassed states are labeled, and whether they operators dispatch the NAUO to position this switch, the alarm that is received will inform are augmented with status lights showing actual them that the action is complete prior to the communication from the NAUO. Should the valve position NAUO fail to position the switch correctly, the lack of alarm will provide the MCR staff with opportunity to identify the error. Should the NAUO position the wrong switch, the ERCW 33.

Date Posted: 07/31/15 pump will fail to start which will alert the operators that verification is needed to ensure that previous actions were performed correctly.

The switches conform to the requirements for local workstation controls outlined in NUREG-0700. Labeling of the switches is in accordance with labeling requirements of TI-12.14, Replacement and Upgrade of Plant Component Identification Tagging and Labeling. This labeling is consistent with other components that the operating staff manipulates during routine evolutions. Standard abbreviations are used on the switch, are easily recognizable to the station staff and are defined in 0-TI-12.13, Acronyms/Abbreviations Listing for Labeling. The positions required to be selected are clearly marked and align with the instruction that is outlined in procedures. A picture of the switch and labeling provided at the end of this enclosure.

HF - 7 The information contained in this response supersedes the information originally provided to the NRC concerning the procedures that will be developed for this event. TVAs initial TVAs response to NRC Acceptance Review response indicated that E-0 and ES-1.3 would contain the guidance that would be Question 5, Item 2, paragraph b states, in part: The implemented post LOCA. After initial drafts were reviewed, it was determined that this final decision about what procedures will be affected guidance more appropriately belonged in the EOP associated with a LOCA (E-1).

by this license amendment request is part of the 34.

impact review that occurs once the submittal is A list of procedures that will ultimately be modified based on the changes needed to approved. implement the requirements is included near the end of this enclosure. The previous TVA statement regarding the final decision about what procedures will be affected by this NUREG-1764, Guidance for the Review of Changes license amendment request is part of the impact review that occurs once the submittal is to Human Actions, Revision 1, Section 3.8, approved requires clarification. The procedures will not actually be implemented until the Procedure Design, states, in part: The objective of license amendment request (LAR) process is complete. However, part of the LAR Page 42 of 64

ENCLOSURE Responses to NRC Audit Review Questions for WBN Unit 1 ERCW and CCS License Amendment Request Item NRC Question/Request TVA Response/Dated Posted No. Date Posted this review is to verify that applicable plant process and the accompanying DCN process require that impacts be identified in advance procedures have been appropriately modified, where of the LAR or DCN. Therefore, at this time, TVA has a good understanding of the scope needed, to provide adequate guidance for the of the procedure revisions required. Drafts of the proposed changes have been successful completion of the [Human Actions] HAs, distributed for review to the appropriate personnel.

and that the procedures adequately reflect changes in plant equipment and HAs. The only procedure that has been changed at this time is 0-SI-82.02, Diesel Generator (DG) 1B-B System Operating Instruction. Revision 005 of this instruction was issued to Identify any new, revised, or deleted procedures ensure that the bypass switch on the 6.9 kV shutdown board is checked during the required to support the proposed LAR not previously performance of the DG standby checklist. This change has been field verified to ensure identified in docketed submittals. Provide procedure the guidance in the procedure is consistent with the equipment information in the field.

number, revision, title, and a summary of the actions changed, added, or deleted. The remaining procedures needed to implement this change will be issued as the implementing process requires. For the bypass switches, this will occur on return to Date Posted: 07/31/15 operation of the DCN paperwork following the work to install the switches. For the remainder, it will follow LAR approval.

The GO and EOP procedures will be changed as required to support the issuance of Unit 2 EOPs. This is currently procedurally defined as sometime between completion of Hot Functional Testing and fuel load. Although an actual date can not be identified, the entire Unit 2 EOP network will be required to be in place prior to operation of Unit 2.

The guidance needed to realign lineups required in the CCS and ERCW systems is currently in place. Therefore, no future change to these procedures is required to implement these actions. Although it is possible that validations of the procedures remaining to be issued might identify a note or other informational enhancements that might be required, they would not constitute a condition that would prevent issuing the revised procedures.

HF - 8 Two hard copies of each units TS are available in the MCR. One is maintained at the US station and the other is at the SM desk. TS are also available electronically via the WBN Describe how the operators access and use electronic document storage and retrieval system known as the Business Support Library Technical Specifications (e.g., if they are available (BSL). Access to BSL for documents is routinely performed by the operating staff to print 35.

electronically, as a hard copy, etc.). Describe the copies for performing surveillance instructions and to verify that the hard copies used are interface between the Emergency Operating the current revision of the procedure.

procedures and Technical Specifications, specifically, if the procedures refer the operator to the Technical The Abnormal Operating Instructions (AOI) identify the potential TS that could be Page 43 of 64

ENCLOSURE Responses to NRC Audit Review Questions for WBN Unit 1 ERCW and CCS License Amendment Request Item NRC Question/Request TVA Response/Dated Posted No. Date Posted Specifications impacted by the particular event in a list format. When entry into the AOI procedures occurs, this list and TS will be referenced to identify any TS limitations or required actions.

Date Posted: 07/31/15 This is particularly important in the AOI network, as these are the instructions that are used to combat emergencies in a condition where the plant may maintain operation in Modes 1 and 2.

The emergency operating instructions (EOI) do not contain direct reference to TS information. This network will be utilized following a reactor trip and presuppose a condition adverse to TS has occurred such as loss of primary pressure boundary, complete loss of AC power, steam generator (SG) tube leak, or steam line rupture for which the potential TS limitations or required actions are secondary to ensuring that immediate actions are taken to place or restore the plant to a stable condition. In the case of entry into the EOI network, conforming to the Westinghouse developed Emergency Response Guidelines (ERG) ensures the plant is maintained in the safest condition for the event. It would then become the responsibility of the operating staff and the emergency response organization to identify potential impacts to TS equipment that may influence future actions that are the result of the plant condition.

HF - 9 Training representatives are part of the team that reviews DCN and LAR impacts. In addition, one of the responsibilities of the Operations member reviewing impacts is to Describe the plans and schedules for revising the make a determination if training is impacted and to inform the appropriate training training program, to reflect the changes in the management representative if training is required.

proposed license amendment. Clarify if training will be provided prior to implementation of the proposed The training associated with these changes will be performed in multiple formats.

changes. Relevant DCN impacts to the plant are included each licensed operator requalification (LOR) cycle in the Plant Changes portion. The Plant Changes training covers DCN Date Posted: 07/31/15 changes, procedure changes, relevant industry events and significant corrective action 36.

events.

The training changes required for this change are tracked in the corrective action program. The training needs analysis has been performed. Additions to the Changes Lesson Plan and the lesson plans associated with the ERCW System and CCS will be complete for the next scheduled LOR training cycle. All training will be complete prior to the implementation of the license amendment request.

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ENCLOSURE Responses to NRC Audit Review Questions for WBN Unit 1 ERCW and CCS License Amendment Request Item NRC Question/Request TVA Response/Dated Posted No. Date Posted HF - 10 NPG-SPP-01.2.1, Interim Administration of Site Technical Procedures for Watts Bar 1 and 2, contains the requirements that must be met when developing or revising technical Describe the process used to verify and validate the procedures. This would apply to the SOIs and GOs associated with the change. As ability of TVAs operators to accomplish tasks defined in the procedure review verification checklist (Attachment 3) of this procedure, required for the proposed license amendment. In lieu various activities are performed to ensure that the procedure guidance is developed in a of a description, relevant administrative procedure(s) manner that includes consideration for human factors.

for verification and validation or a verification and validation plan for the proposed change (if The following is a list of some of the items that are required to be verified in procedure developed) may be provided. In your response, development:

clarify if the validation will include a representative sample of operators and whether it will be performed Does the procedure agree with and reference applicable drawings?

with Technical Specification minimum staffing and Can the procedure be correctly performed in the designated sequence?

nominal staffing. Are equipment numbers and nomenclature used in the procedure identifiable to those displayed on the equipment?

Date Posted: 07/31/15 Can equipment identified in the procedure be easily located?

Are the units of measurement used in the procedure to record readings the same as those displayed on the equipment?

37. Have human factors and system interactions been properly considered?

Procedures are walked down after development to ensure that the information provided in the procedure agrees with conditions in the field. Personnel who would normally perform the task are the individuals who are tasked with these walkdowns to ensure that the developed content provides the level of detail that is needed to successfully perform the evolution.

For the emergency operating network, TI-12.11, Emergency Operating Instruction (EOI)

Control, contains a more specific validation process. This instruction defines what validation method should be used based on the change, the persons that should make up the validation team, how the validation is conducted and how the validation is documented.

In all cases involving the EOI network, consideration is taken on whether the task can be performed by the minimum shift compliment. If it is judged that the task would interfere with the ability of minimum shift staffing requirements, then efforts are taken to either redevelop the desired process or increase the required staffing as an interim measure.

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ENCLOSURE Responses to NRC Audit Review Questions for WBN Unit 1 ERCW and CCS License Amendment Request Item NRC Question/Request TVA Response/Dated Posted No. Date Posted HF - 11 TI-12.11, Emergency Operating Instruction (EOI) Control, contains direction that is intended to address this very concern. For the current modification, the EOI network Describe the process used to monitor new and contains the only procedures that this issue might apply to.

changed operator actions to ensure that they remain feasible and reliable over the long term, and are not Limitations presented by having additional persons to perform the CCS and ERCW degraded due to design changes, inadequate alignments prior to entering Mode 4 do not represent a safety concern. If insufficient staff training, or other mechanisms. is available to accomplish these alignments, then entry into Mode 4 will be prohibited.

This represents a station efficiency and outage completion concern and not a safety Date Posted: 07/31/15 concern. In practice, Operations crews transition to super crew alignment approximately two weeks prior to the start of the outage, so sufficient personnel will be available to prevent this from being a concern.

The actions taken in the EOI network require an initial assessment for this very concern.

In addition, any future revisions are required to evaluate whether minimum operator 38.

staffing levels are impacted by the proposed change.

Operator requalification training on EOIs provides a means of periodically verifying the technical adequacy of emergency instructions. Operators and training personnel are responsible for ensuring that problems or discrepancies discovered in EOIs during training are documented using a Condition Report or Procedure Change Request (PCR), as appropriate. Proposed enhancements and suggestions for improvement of EOIs are also encouraged.

Should a future attempt be made to change the operation of the bypass switches, plant processes would identify that this would require a 10 CFR 50.59 review. This review would prevent future manipulations that would have a negative impact on maintaining plant safety.

HF - 12 There will be a total of four bypass switches installed, one on each 6.9 kV shutdown board. Each bypass switch will bypass the interlock that prevents two ERCW Pumps from TVAs response to NRC Acceptance Review being powered from the same shutdown board.

Questions dated July 14, 2015 (ADAMS Accession 39.

Number ML15197A357), Question 5, Item 4 states: Only one annunciator will be added. The ERCW annunciator panel in the MCR is a An interlock bypass switch for ERCW pumps will be common annunciator panel (does not have a separate panel for Unit 1 and Unit 2).

installed as described above and in the license Pictures of the MCR annunciator panel and the window for this alarm are included at the amendment request. An annunciator window will be end of this enclosure.

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ENCLOSURE Responses to NRC Audit Review Questions for WBN Unit 1 ERCW and CCS License Amendment Request Item NRC Question/Request TVA Response/Dated Posted No. Date Posted added in the main control room to show when an interlock bypass switch has been activated. The main function of the alarm is to alert the operations staff should the switch be moved from its normal position during normal operations. (See pictures of switch at the end of Clarify how many interlock bypass switches will be this enclosure). In this case, the ARI will direct that an operator be dispatched to the installed and how many annunciator windows will be 6.9 kV shutdown boards to determine which switch has been moved to the bypass added. Further, clarify if all bypass switches will be position.

identical in design an appearance, with the exception of identification and labeling All bypass switches will be identical in design and appearance, with the exception of identification and labeling, which is unique for each 6.9 kV shutdown board.

HF - 13 This response supersedes the response provided on June 17, 2015. The procedure guidance for aligning either CCS Pump 1B-B or 2B-B previously existed in 0-SOI-70.01, TVAs application dated June 17, 2015 (ADAMS thus no change is required to this instruction for the proposed changes.

Accession Number ML15170A474), Enclosure 1, Section 4.1.2, Postulated GDC 5 Event, states, in part: The following procedures will be affected

40. System Operating Instruction SOI-70.01, Component Cooling Water (CCS) System. This SOI does not require a revision because the steps to realign the CCS Train B pumps are currently in the procedure.

Clarify how SOI-70.01 is affected by the changes proposed in this LAR, if it does not require a revision.

SCVB-RAI-1 (a) The material properties chosen are intended to represent the three most common structural materials in the RCS; stainless steel 304, stainless steel 316, and low alloy Reference 1, Attachment to Enclosure 1, carbon steel.

Westinghouse Summary Report, Section 4.4.1.3, seventh bullet in the summarized assumptions for (b) The ASME Boiler and Pressure Vessel Code,Section II, Part D (Reference 1) mass and energy release analysis states [hereafter referred to as "the ASME BPVC" (Reference 1)] was used as the source of

41. the material property data. Table PRD provides material densities in units of lbm/in3.

Density and specific heat values of 501 lbm/ft2 and The density of 0.29 lbm/in3 (converted to 501 lbm/ft3) is representative of the density 0.145 BTU/lbm-°F, respectively, model a volumetric at a cold state of 70°F for Stainless Steel 304 and Stainless Steel 316. The density heat capacity which bounds the values found in of carbon steel at a cold state of 70°F is listed as 0.28 lbm/in3 (converted to Part D of the ASME boiler pressure vessel code. 484 lbm/ft3). The bulk of the metal mass in the reactor vessel and the steam generators is carbon steel. The density of stainless steel, 501 lbm/ft 3, was (a) Provide the material specification for which the conservatively applied to all steel alloys (note that the steam generator tube material Page 47 of 64

ENCLOSURE Responses to NRC Audit Review Questions for WBN Unit 1 ERCW and CCS License Amendment Request Item NRC Question/Request TVA Response/Dated Posted No. Date Posted above density and specific heat values are is treated separately) in the LOCA Mass and Energy (M&E) calculations to maximize assumed. the metal mass.

(b) Confirm that the assumed density 501 lbm/ft3 (c) The specific heat (as a function of temperature) was determined for each material by bounds the values given in Part D of the ASME using the following information from the ASME BPVC (Reference 1); the equation Section II Boiler and Pressure Vessel code relating thermal conductivity, thermal diffusivity, density, and specific heat from the instead of Part D of ASME boiler pressure Table TCD General Notes, thermal expansion coefficients in Tables TE-1 through vessel code, If not provide more information TE-4, thermal conductivities in Table TCD, and thermal diffusivities in Table TCD. An regarding the source of the assumed value of uncertainty of 10% was added to the calculated specific heat values.

density.

The computer codes that are part of the WCAP-10325-P-A evaluation model are (c) ASME Boiler and Pressure Vessel Code (BPVC), constructed such that a single specific heat value is required. Therefore, a bounding Section II, Part D does not provide specific heat value of 0.145 BTU/lbm°F provides a conservative estimate of the total amount of values. Please state the source document of the metal energy over the temperature range that is relevant in a LOCA M&E calculation ASME specific heat of the RCS metal which is (metal cools down from approximately 600°F to 200°F) for all three of the structural bounded by the assumed specific heat of 0.145 steel alloys.

BTU/lbm-°F given in the above statement.

Reference:

1. An International Code, 2010 ASME Boiler and Pressure Vessel Code, 2010 Edition, July 1, 2010,Section II, Part D, Properties (Customary), Materials, ASME Boiler and Pressure Vessel Committee on Materials, Three Park Avenue, New York, NY, 10016 USA.

SCVB-RAI-2 The difference between 2,750,700 lbs and the analytical value of 2,585,000 lbs considers losses due to sublimation only. The surveillance requirements address uncertainties Reference 1, Enclosure 1, Section 4.0, Conclusions, introduced through weighing.

first paragraph states calculated values of the ice weight from containment analysis 2,585,000 lbs with (a) The Surveillance Requirement value of 2,750,700 pounds contains a sublimation an average ice basket weight of 1,330 lbs. The allowance of six percent. The value does not include a margin for measurement

42. second paragraph states an ice weight specified in uncertainty. The fact that the value does not include the measurement uncertainty is Surveillance Requirement (SR) 3.6.11.2 will be stated in the Technical Specification Bases for SR 3.6.11.2 and 3. Surveillance 2,750,700 lbs with a per basket value of 1415 lbs Instruction 1-SI-61-2, 18 Month Ice Condenser Surveillance states that the specified in SR 3.6.11.2 and SR 3.6.11.3. instrument uncertainty must be added to the SR ice weight value to determine an acceptable basket weight. The allowance for instrument uncertainty is approximately (a) Please confirm that the difference between the +15 lbs.

SR value 2,750,000 lbs and the analytical value Page 48 of 64

ENCLOSURE Responses to NRC Audit Review Questions for WBN Unit 1 ERCW and CCS License Amendment Request Item NRC Question/Request TVA Response/Dated Posted No. Date Posted 2,585,000 lbs considers the sublimation (b) The method used for determining the sublimation rate is from cycle to cycle ice allowance and the measurement uncertainty basket weighing. The as-left basket weight is compared to the as-found basket during surveillance. If so, how much individual weight in the next cycle for the same basket. Overall trends for the ice bed at large margins are provided for sublimation and and on a row group basis are also used to validate the sublimation rate. Historical measurement uncertainty. If not, explain the data for WBN Unit 1 and for Sequoyah Nuclear Plant have shown cycle to cycle basis for the difference between the two values. sublimation rates of around three percent. The selection of six percent is based on engineering judgment to provide a large safety margin.

(b) Please explain the methodology used for calculating the sublimation allowance. What are The FSAR 6.7.14.3 will be updated to reflect the following.

the assumed initial parameters (such as ice temperature, environment temperature in the ice Sublimation - Historical condenser, etc) for the calculation?

The following information was developed during the design and initial operation of the ice condenser system. Actual sublimation rates have been established during the operation of Watts Bar and are discussed in the Section entitled Sublimation - Actual.

The other mechanism that affects the long-term storage of the ice is sublimation.

Sublimation has several effects inside the ice condenser. The geometry of the ice mass changes where sublimation occurs, and the resulting vapor is deposited on a colder surface at another location inside the ice condenser.

In normal cold storage room application, the cooling coil is exposed to the air in the room, and moisture in the air freezes on the coil. If ice is stored in the room, all of ice eventually migrates to the coil (which is defrosted periodically, draining the water outside the room) through a sublimation-mass transfer mechanism. To avoid the mechanism, and maintain a constant mass of ice, the ice condenser is provided with double wall insulation. The annular gap between the insulated walls is provided with a heat sink in the form of a flow of cool, dry air that enters arid and leaves through the insulated panels.

However, a small amount of heat enters the system through the inlet doors, which are not double insulated, and also through the double layer insulation system. The effect of this heat gain on the ice condenser has been examined analytically. An analytical model of the sublimation process has been developed to provide an estimate of the expected sublimation rate as well as identify the significant parameters affecting the sublimation rate. The model developed a relationship identifying the fraction of total heat input which sublimes ice (the rest of the heat raises the temperature of the air, which transports the Page 49 of 64

ENCLOSURE Responses to NRC Audit Review Questions for WBN Unit 1 ERCW and CCS License Amendment Request Item NRC Question/Request TVA Response/Dated Posted No. Date Posted vapor to the cold surface where it freezes).

The sublimation fraction depends on the difference in vapor pressure between warmest and coldest air temperatures within the ice condenser. The sublimation fraction decreases as the T decreases and also as the average ice condenser temperature decreases. For an average temperature of 15°F in the ice condenser compartment, the analytical model predicts a sublimation rate of about 1% of the ice mass sublimed per year per ton (12,000 Btu/hr) of heat gain to the ice storage compartment. The final heat gain calculations identified a heat gain into the ice storage compartment of 1 to 1.5 tons, most of which enters the compartment through the doors.

For the purposes of this report, it is assumed that the reference heat gain for the unit is 1 ton, and therefore, the calculated reference sublimation rate would be 1% of the ice weight per year.

Selected baskets are weighed as indicated in Technical Specifications to verify that the actual sublimation rate has not excessively depleted the ice inventory.

Sublimation - Operational The method used for determining the sublimation rate is from cycle to cycle ice basket weighing. The as-left basket weight is compared to the as-found basket weight in the next cycle for the same basket. Overall trends for the ice bed at large and on a row group basis are also used to validate the sublimation rate. Historical data for Watts Bar Unit 1 and for the Sequoyah Nuclear Plant have shown cycle to cycle sublimation rates of around three percent. The selection of six percent is based on engineering judgment to provide a large safety margin.

The surveillance acceptance criteria contain a sublimation allowance of six percent. The value does not include a margin for measurement uncertainty. Instrument uncertainty must be added to the surveillance requirement ice weight value to determine an acceptable basket weight. The allowance for instrument uncertainty is approximately +15 lbs.

The Ice Bed Temperature is maintained between 15°F and 20°F during plant operation.

The empirical sublimation rates described above are the results of operating in this Page 50 of 64

ENCLOSURE Responses to NRC Audit Review Questions for WBN Unit 1 ERCW and CCS License Amendment Request Item NRC Question/Request TVA Response/Dated Posted No. Date Posted temperature range. Procedures require actions to restore normal operating temperatures when ice bed temperatures reach 23°F.

SCVB-RAI-3 TVA has clarified the referenced FSAR Section (6.7) to designate the current verbiage as historical and provided the empirical operational basis for how sublimation is calculated in FSAR Amendment 113, under heading Sublimation, response to SCVB-RAI-2.

page 6.7-64 states:

(a) and (b) The verbiage in question is part of the historical basis for ice condenser For an average temperature of 15oF in the ice operation. The new operational basis consists of the following. The Ice Bed condenser compartment, the analytical model Temperature is maintained between 15°F and 20°F during plant operation.

predicts a sublimation rate of about 1% of the ice The empirical sublimation rates described above are the results of operating mass sublimed per year per ton (12,000 Btu/hr) of in this temperature range. Procedure (ARI-138-144) requires actions to heat gain to the ice storage compartment. restore normal operating temperatures when ice bed temperatures reach The ice condenser compartment temperature of 15ºF 23°F.

43. specified in the above statement is not consistent with the ice bay air temperature of 27ºF specified in Table 1 of Reference 1, Enclosure 1.

(a) Correct the ice condenser compartment temperature in the above statement from 15ºF to 27oF, or provide justification for the difference.

(b) Recalculate the sublimation rate based on ice compartment air temperature of 27ºF, and provide its impact on the sublimation allowance.

SCVB-RAI-4 The proposed long-term LOCA containment integrity analysis used the reviewed and approved LOTIC-1 code. Internal to the configured LOTIC-1 code is a limit placed on the Reference 1, Enclosure 2, Watts Bar Nuclear Plant stagnation heat transfer coefficient of 72 Btu/hr-ft2-°F.

Unit 2 Revised FSAR Section 6.2.1 Pages, page 6.2.1-8, under heading Structural Heat Removal, This can be seen from the stagnation heat transfer coefficient presented in Equation 54 in 44.

states a Tagami heat transfer coefficient for the lower WCAP-8355-P-A, Long Term Ice Condenser Containment Code LOTIC Code, April containment compartment structures was limited to 1976. When the steam to air ratio is 1.4, Hstag is limited to 72 Btu/hr ft2-°F.

72 Btu/hr-ft2.

Enclosure 1, Attachment Westinghouse Summary Page 51 of 64

ENCLOSURE Responses to NRC Audit Review Questions for WBN Unit 1 ERCW and CCS License Amendment Request Item NRC Question/Request TVA Response/Dated Posted No. Date Posted Report, Section 4.4.3.3, Structural Heat Removal, page 42, mentions Tagami correlations for the heat transfer coefficient for the lower compartment structures were used in the proposed analysis, but does not provide its value. Confirm that the proposed analysis used heat transfer coefficient of 72 Btu/hr-ft2 consistent with the FSAR. In case it is changed, please revise the FSAR and justify if a less conservative (greater) value was used.

No changes were made to the code for this proposed analysis, so the structural heat transfer coefficient limit of 72 Btu/ht-ft2-°F described in the Structural Heat Removal subsection at the end of FSAR Section 6.2.1.3.3 continues to exist.

SCVB-RAI-5 According to the Revision log for SDD N3-61-4001, Westinghouse performed a post-LOCA containment sump boron concentration analysis to address PER 03-006899-000.

Reference 1, Enclosure 3, Watts Bar Nuclear Plant The maximum ice weight used as an input to the Westinghouse Analysis was 3.0 x 106 Unit 2 Revised Pages for TS and TS Bases 3.6.11, lbs. DCN D51416-A revised the SDD to ensure the maximum total ice condenser weight SR 3.6.11.2a weighs samples 144 ice baskets and does not exceed 3.0 x 106 lbs.

verifies each basket contains 1415 lbs of ice. This surveillance allows individual baskets to weigh Westinghouse Letter WAT-D-10850, Section 1.1, concludes the evaluation performed by greater than 1415 lbs. FSAR Amendment 113, Westinghouse determined that the WBN ice basket maximum average loading limits and

45. Section 6.7.6.1, page 6.7-22, specifies maximum total configuration requirements are acceptable based on seismic design allowables.

ice weight 3x106 lbs which results in an individual ice basket weight (3x106/1944) = 1543.2 lbs. The configuration shown is 1/3 of 1944 baskets (648 baskets) at a maximum ice weight including basket of 1809 lbs, another 648 baskets at a maximum ice weight including (a) Confirm that the maximum ice weight of 3x106 basket of 1909 lbs and the final 648 baskets at a maximum ice weight including basket of lbs is based on seismic qualification test results. 2009 lbs for a total of 3,711,096 lbs of ice (including the weight of the baskets). The weight of the ice baskets as 250 lbs for a total empty ice basket weight of 250 x 1944 =

(b) Explain how, during surveillance, it would be 486,000 lbs. Therefore, WBN is seismically analyzed for 3,711,096 - 486,000 =

verified that the individual ice basket weight and 3,225,096 lbs of ice in the ice condenser. As mentioned above, the limiting maximum Page 52 of 64

ENCLOSURE Responses to NRC Audit Review Questions for WBN Unit 1 ERCW and CCS License Amendment Request Item NRC Question/Request TVA Response/Dated Posted No. Date Posted the total ice weight in the ice condenser would value of ice is not based on the seismic qualification of the ice condenser but rather the not exceed the maximum limits. post-LOCA sump boron concentration.

Surveillance requirements provide acceptance criteria not to exceed the 3,000,000 lb total weight based on the sample of ice basket weights as defined by Technical Specifications.

1-SI-61-2, 18 Month Ice Weighing, surveillance.

Additionally, individual ice baskets weights as described above are also controlled in the surveillance procedure as follows.

1) The structural limit for the Ice Condenser per basket in rows 1, 2 & 3 is 1,809 lbs, but 1,794 lbs will be the adjusted weight to allow for M&TE inaccuracy when using the electronic load cells.

2) The structural limit for the Ice Condenser per basket in rows 4, 5 & 6 is 1,909 lbs, but 1,894 lbs will be the adjusted weight to allow for M&TE inaccuracy when using the electronic load cells.

3) The structural limit for the Ice Condenser per basket in rows 7, 8 & 9 is 2,009 lbs, but 1,994 lbs will be the adjusted weight to allow for M&TE inaccuracy when using the electronic load cells.

SCVB-RAI-6 In addition to the design basis, this analysis accounted for the effects of other plant changes of which Westinghouse is aware. These include increased valve stroke time (of Reference 1, Enclosure 1, Attachment, Section +13 seconds) to open the containment spray flow control valves (Reference 1), initial 4.4.1.1 states: condition uncertainties on RCS temperature of +7°F, and 17x17 Robust Fuel Assembly-2 (RFA-2) fuel (which may incorporate tritium-producing burnable absorber rods In addition to the design basis, this analysis (TPBAR)). Also, the evaluation that was provided in Reference 17 with the conclusion accounted for the effects of other plant changes of that a +/- 0.2 Hz variation in the diesel frequency would have a negligible impact on the 46.

which Westinghouse is aware. These include LOCA mass and energy release analysis remains valid. It should be noted that these increased.. items were included for completeness even though they may not be currently implemented at WBN Unit 2.

Please describe all other changes that were incorporated in the mass and energy analysis beside the four changes described in Section 4.4.1.1.

Page 53 of 64

ENCLOSURE Responses to NRC Audit Review Questions for WBN Unit 1 ERCW and CCS License Amendment Request Item NRC Question/Request TVA Response/Dated Posted No. Date Posted SCVB-RAI-7 For the time after initiation of shutdown cooling in the Shutdown Unit, the LOCA is assumed to occur at 7-hours after initiation of the shutdown. The loss of Train A power is Please confirm that Loss of train A concurrent with the limiting case for the containment analysis.

loss of offsite power (LOOP) during which CCS heat exchanger (HX) C carries the heat load of both the LOCA and shutdown unit is the most limiting

47. condition for the CCS fluid temperature, containment analysis parameters (containment peak pressure and temperature) and the shutdown cooling analysis. For the most limiting case of heat loads, specify at what time after initiation of shutdown cooling in the Shutdown Unit the LOCA is assumed to occur in the LOCA Unit.

TS - 2 Per LCO 3.0.2, if the LCO is met or is no longer applicable prior to expiration of the specified Completion Time(s), completion of the Required Actions is not required unless Please provide a more complete explanation of why otherwise stated. Therefore, the Completion Times of LCO 3.7.16 and LCO 3.7.17 Conditions B of both proposed LCO 3.7.16 and Condition B Required Actions continue until the LCO is met or is no longer Applicable.

proposed LCO 3.7.17 have no effective completion, i.e. Once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ? Is this condition proposed The TS Bases will be revised to reflect the follow:

to continue until the LCO expires i.e. 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months

after entry into Mode 3 from Mode 1 or 2 ? The The purpose of the guidance contained in Condition B is to ensure clear direction is given

48. current basis statement for each LCO does not to NOT enter Mode 5, if the additional ERCW and CCS alignments associated with adequately explain why no restoration action is TS 3.7.16 and TS 3.7.17 are not performed. This assumes that the CCS and ERCW needed. System meet the requirements of TS 3.7.7 and TS 3.7.8, respectively, to support continued operation in Mode 4. In this case, with the plant in Mode 4, additional methods Date Posted: 08/25/15 of decay heat removal are available and the potential for an uncontrolled heatup from Mode 5 should the postulated accident occur is avoided. Should additional inoperabilities impact compliance with TS 3.7.7, TS 3.7.8, or TS 3.4.6, the Actions associated with those TSs would prevail.

TS - 4 LCO 3.7.16 / LCO 3.7.17 Applicability Note b was originally proposed to preclude the requirement for additional CCS and ERCW pumps if complying with Required Actions to The supplemental proposal changed the applicability be in Mode 5, since additional failures, such as a loss of Train A 6.9 kV shutdown boards, 49.

statements for the new LCOs 3.7.16 and 3.7.17 does not have to be postulated while in a TS Action. However, as stated in the NRC without explanation in the document. Please further scenario, if the postulated failure is the loss of Train A 6.9 kV shutdown boards, the explain why the applicability statement This LCO is GDC 5 event is still viable and the requirement for the additional CCS and ERCW pumps Page 54 of 64

ENCLOSURE Responses to NRC Audit Review Questions for WBN Unit 1 ERCW and CCS License Amendment Request Item NRC Question/Request TVA Response/Dated Posted No. Date Posted not applicable for either of the following conditions: is still required. Therefore, Applicability Note b was removed, as well as the adoption of

b. When complying with Required Actions to be in TSTF-273.

Mode 5 was essentially made new Condition A in each of the new TS AND Complying with If the requirement of either LCO 3.7.16 or LCO 3.7.17 is not met, maintaining the unit in Required Actions to be in Mode 5. Mode 4 with decay heat removal from the RCS loops is preferred. However, if TS Required Actions require entry into Mode 5, the remaining operable RHR loop is sufficient Date Posted: 08/03/15 to cooldown the unit to and maintain it in Mode 5, even with a concurrent LOCA in the other unit. Therefore, the wording of Conditions A and B provide for these two scenarios.

TS - 5 The following information was provided by TVA in letter dated June 17, 2015, Enclosure 1, Section 4.1.2, Postulated GDC 5 Event, page E1-10:

In a public meeting and audit discussions it has been presented that the additional ERCW pumps required The ERCW System design was based on requiring two ERCW pumps to handle the by proposed LCO 3.7.17 are needed essentially for cooling loads to the UHS for shutting down both units during either normal operation or in increased ERCW flow when the accident unit the event of a LOCA and the shut down of the non-accident unit. It has been determined, switches to recirculation mode to remove heat from for the specific set of scenarios in this evaluation, that three ERCW pumps will be required

50. the accident units Containment Spray heat if a cool down of the non-accident unit using RHR occurs within the first 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months after a exchangers. Please clearly articulate this point in a shutdown. The higher heat loads associated with continuing the cool down of the unit that docketed submittal. has been shut down for less than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , combined with the heat removal requirements of the safety analysis for the DBA LOCA via RHR and containment spray, necessitates Date Posted: 08/03/15 the use of three ERCW pumps during the initial 48 hour5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months time period. This additional cooling capacity is required prior to placing containment spray on recirculation mode.

Once the unit has been shut down for 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or more, the total ERCW heat removal and thus, flow requirements, drop below the flowrate provided by two ERCW pumps.

Page 55 of 64

ENCLOSURE Responses to NRC Audit Review Questions for WBN Unit 1 ERCW and CCS License Amendment Request For integrated HX performance, See EPMJN010890 Table C7.7.69 and for U2 C7.7.89 At time 7-hours, provide two virtual HX models LOCA Unit Apportioned by Flow LOOP & Loss of Train A LOCA Recirc Supports Loss of Train A LOCA Containment Analysis CCS HX C RHR HX RHR HX MISC UA 3.35 1.63 NOT IN N/A F 1.00 0.90 SERVICE m HOT, gpm 5,000.00 3,801.98 m HOT, m#/hr 2.47 1.86 DENSITY (of cold fluid) 61.23 61.95 M cold, gpm 4,524.89 5,000.00 M cold, m#/hr 2.22 2.48 R 0.86 1.22 S -0.73 -1.15 Q (MBtu/hr) 54.80 54.80 t1 (deg. F) 85 103 t2 (deg. F) 110 125 T2 (deg. F) 103 136 T1 (deg. F) 125 166 delta t (deg. F) 25 22 delta T (deg. F) 22 29 HX CORRECTION FACTOR r 0.75 p 0.47 F 0.92 UA ADJUSTMENT m HOT, DESIGN 2.21 1.48 M cold, DESIGN 2.95 2.48 ho, DESIGN 1,172.18 1,386.96 ho, ACTUAL 1,252.52 1,390.11 hi, DESIGN 1,013.34 2,159.37 hi, ACTUAL 807.61 2,592.56 U, ACTUAL 190.42 381.27 UA, ACTUAL 3.35 1.63 Page 56 of 64

ENCLOSURE Responses to NRC Audit Review Questions for WBN Unit 1 ERCW and CCS License Amendment Request LOCA Unit Apportioned by Heat Load LOOP & Loss of Train A LOCA Recirc Supports LoTA LOCA Containment Analysis CCS HX C RHR HX RHR HX MISC UA 2.64 1.63 NOT IN N/A F 1.00 0.90 SERVICE m HOT, gpm 5,000.00 3,801.98 m HOT, m#/hr 2.47 1.86 DENSITY (of cold fluid) 61.23 61.84 M cold, gpm 3,483.59 5,000.00 M cold, m#/hr 1.71 2.48 R 0.62 1.22 S -0.88 -1.15 Q (MBtu/hr) 54.80 54.80 t1 (deg. F) 85 111 t2 (deg. F) 117 133 T2 (deg. F) 111 145 T1 (deg. F) 133 174 delta t (deg. F) 32 22 delta T (deg. F) 22 29 HX CORRECTION FACTOR r 0.75 p 0.47 F 0.92 UA ADJUSTMENT m HOT, DESIGN 1.70 1.48 M cold, DESIGN 2.27 2.48 ho, DESIGN 1,172.18 1,386.96 ho, ACTUAL 1,463.36 1,388.62 hi, DESIGN 1,013.34 2,159.37 hi, ACTUAL 807.61 2,592.56 U, ACTUAL 194.67 381.16 UA, ACTUAL 2.64 1.63 Page 57 of 64

ENCLOSURE Responses to NRC Audit Review Questions for WBN Unit 1 ERCW and CCS License Amendment Request LOOP & Loss of Train A Shutdown Unit Supports LoTA LOCA Containment Analysis CCS HX C RHR HX RHR HX MISC UA 4.12 1.57 NOT IN N/A F 1.00 0.90 SERVICE m HOT, gpm 5,000.00 3,025.23 m HOT, m#/hr 2.46 1.48 DENSITY (of cold fluid) 61.23 61.92 M cold, gpm 5,679.60 5,000.00 M cold, m#/hr 2.79 2.48 R 1.22 1.47 S -0.63 -1.25 Q (MBtu/hr) 89.27 89.27 t1 (deg. F) 85 105 t2 (deg. F) 117 141 T2 (deg. F) 105 156 T1 (deg. F) 141 216 delta t (deg. F) 32 36 delta T (deg. F) 36 60 HX CORRECTION FACTOR r 0.60 p 0.54 F 0.90 UA ADJUSTMENT m HOT, DESIGN 2.78 1.48 M cold, DESIGN 3.70 2.48 ho, DESIGN 1,172.18 1,386.96 ho, ACTUAL 1,089.98 1,389.78 hi, DESIGN 1,013.34 2,159.37 hi, ACTUAL 807.61 2,159.37 U, ACTUAL 186.19 368.36 UA, ACTUAL 4.12 1.57 Page 58 of 64

ENCLOSURE Responses to NRC Audit Review Questions for WBN Unit 1 ERCW and CCS License Amendment Request The respective UAs are 3.35 apportioned by flow and 2.64 apportioned by heat load. A UA of 3.17 was used in the analysis.

A sensitivity run for Unit 2 by Westinghouse showed that a change in UA from 3.17 to 2.00 using an ERCW flow of 3504 gpm to the virtual component cooling heat exchanger resulted in an increase in containment pressure from 11.73 to 11.76 psig; therefore, the use of any of the three values (3.35, 3.17, or 2.64) will produce nearly identical containment pressures. Therefore, the containment design conditions are not exceeded for the two or three pump ERCW cases.

CCW HX UA 3.17 2.00 spray= 20,032 BTU/sec 20,047 BTU/sec At 2718 sec Spray Recirc RHR= 10,253 BTU/sec 9,262 BTU/sec spray= 19,551 BTU/sec 19,641 BTU/sec At 3600 sec RHR Spray Start RHR= 9,128 BTU/sec 8,347 BTU/sec Unit 1 W-COBRA/TRAC (Lotic 1) Results A similar analysis was run on Unit 1 with a UA of 2.00 and an ERCW flow of 3504 gpm to the virtual component cooling heat exchanger. There is no impact to the peak calculated pressure using the new approved W-COBRA/TRAC Mass & Energies when the CCW HX UA is reduced to 2.0. The change in heat rates at the time of spray initiation is shown in the table below.

CCW HX UA 3.17 2.00 spray= 22,798 BTU/sec 22,798 BTU/sec At 2718 sec Spray Recirc RHR= 11,770 BTU/sec 10,612 BTU/sec There was no change in the spray heat removal rate and an approximately 10% reduction in the RHR heat removal rate. The RHR sprays are not credited in the analysis which uses W-COBRA/TRAC Mass & Energies because the calculated pressure will either not exceed 9.5 psig, or will only remain above 9.5 psig for a duration less than 3600 seconds.

Therefore, similarly on U1 the containment design is not challenged and the results are acceptable for two or three ERCW pump cases.

Page 59 of 64

ENCLOSURE Responses to NRC Audit Review Questions for WBN Unit 1 ERCW and CCS License Amendment Request Procedure Type Description 0-SI-82.01, Diesel Generator (DG) 1A-A System Operating Instructions Contains a check of switch position in the diesel standby alignment. This is 0-SI-82.02, Diesel Generator (DG) 1B-B performed after each surveillance run, the most frequent of which is monthly. In 0-SI-82.03, Diesel Generator (DG) 2A-A addition, this check is performed upon return to service of the DG following any 0-SI-82.04, Diesel Generator (DG) 2B-B maintenance activities.

1/2-GO-4, Normal Power Operation General Operating Instructions Contains direction to commence alignments (GO-4 and GO-5).

1/2-GO-5, Unit Shutdown from 30%

1/2-GO-6, Unit Shutdown from HS to CSD Contains direction to ensure alignments are complete prior to entering Mode 4 in GO-6.

1/2-E-1, Loss of Reactor or Secondary Coolant Emergency Operating Instructions Contains direction to place in operation the equipment needed following a LOCA.

Page 60 of 64

ENCLOSURE Responses to NRC Audit Review Questions for WBN Unit 1 ERCW and CCS License Amendment Request MCR ANNUNCIATOR - ERCW PUMP INTERLOCK HANDSWITCH IN BYPASS Page 61 of 64

ENCLOSURE Responses to NRC Audit Review Questions for WBN Unit 1 ERCW and CCS License Amendment Request ERCW PUMP INTERLOCK BYPASS SWITCH EXAMPLE Page 62 of 64

ENCLOSURE Responses to NRC Audit Review Questions for WBN Unit 1 ERCW and CCS License Amendment Request Updated June 17, 2015 letter, Enclosure 1, Table 3 Summary of Steady-State DG Loading with 3 ERCW Pumps (0 mins to end)

Pumps U1 LOCA / U2 Shutdown / U2 LOCA / U1 Shutdown / U1 LOCA / U2 Shutdown / U2 LOCA / U1 Shutdown /

Loss of Train A Loss of Train A Loss of Train B Loss of Train B DG 1A 2A 1B 2B 1A 2A 1B 2B 1A 2A 1B 2B 1A 2A 1B 2B ERCW 805 1610 1610 805 805 1610 1610 805 CCS 378 720 378 720 720*** 378 720*** 378 AFW (motor- 400* ** ** 400* 400* ** ** 400*

driven)

Containment 596 596 596 596 Spray Centrifugal 695 532 532 695 695 532 532 695 Charging SI 460 460 460 460 RHR 440 370 370 440 440 370 370 440 Total / Large 3774 3232 2890 4116 4116 2890 3232 3774 Motor Load (HP)

Pressurizer 500 500 500 500 Heaters (kW)

DG Loading ****

0 - 20 minutes kW 4182 4261 4188 4283 4004 4174 3984 4165 kVA 4851 4797 4745 4930 4622 4684 4492 4781 20 min - 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months kW 4218 4015 3941 4313 4040 3927 3738 4201 kVA 4887 4514 4462 4960 4658 4401 4209 4818 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months - End kW 4067 4015 3941 4163 4148 3927 4015 4033 kVA 4688 4514 4462 4766 4769 4401 4513 4625 Page 63 of 64

ENCLOSURE Responses to NRC Audit Review Questions for WBN Unit 1 ERCW and CCS License Amendment Request Note: Refer to Table 1 for CCS and ERCW pump power alignments.

0 Min - 2 Hrs 600 hp until SGs refilled; thereafter 400 hp (for both LBLOCA and SBLOCA).

- 0 Min - 20 minutes 300 hp; then stopped

- - 378 hp until CCS Pump C-S is manually aligned after 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for spent fuel pool cooling; then 720 hp (360hp each for CCSP 1A and C-S)

- - - Values extracted from Appendix N-1, Pages 1 thru 4 of Diesel Loading Calculation, EDQ00099920080014 R31.

Page 64 of 64

v t e Letter Sequence Request
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