Text

Supplemental Information Regarding the March 12, 2012 Commission Order Modifying Licenses with Regard to Requirements for Mitigating Strategies for Beyond-Design-Basis External Events (Order Number EA-12-049)
	ML16012A145
Person / Time
Site:	North Anna
Issue date:	01/06/2016
From:	Mark D. Sartain; Virginia Electric & Power Co (VEPCO)
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	14-394E, EA-12-049
	Download: ML16012A145 (22)
	v • d • e

VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261 U. S. Nuclear Regulatory Commission January 6, 2016 Serial No.: 14-394E Attention: Document Control Desk NLOS/DEA: R0 Washington, DC 20555-0001 Docket Nos.: 50-338/339 License Nos.: NPF-4/7 VIRGINIA ELECTRIC AND POWER COMPANY (DOMINION)

NORTH ANNA POWER STATION UNITS I AND 2 SUPPLEMENTAL INFORMATION REGARDING THE MARCH 12. 2012 COMMISSION ORDER MODIFYING LICENSES WITH REGARD TO REQUIREMENTS FOR MITIGATING STRATEGIES FOR BEYOND-DESIGN-BASIS EXTERNAL EVENTS (ORDER NUMBER EA-12-049)

On March 12, 2012, the Nuclear Regulatory Commission (NRC) issued Order EA-12-049, "Order to Modify Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events" [the Order]. On May 19, 2015, Dominion notified the NRC staff that North Anna Power Station was in compliance with the Order for both Units 1 and 2 and provided the Final Integrated Plan (FIP) for the North Anna FLEX Mitigation Strategies. Subsequently, a call between Dominion and the NRC staff on August 4, 2015 identified several items for which the NRC staff requested additional information. The attachment to this letter provides the requested information for Interim Staff Evaluation Open Item 3.2.1.2.8 and Safety Evaluation Items #4, #8, #16, and #17.

Additionally, the August 4, 2015, phone call addressed Westinghouse Technical Bulletin 15-1, "Reactor Coolant System Temperature and Pressure Limits for the No. 2 Reactor Coolant Pump Seal." Each of the units at North Anna have replaced two of the three Reactor Coolant Pump seals with low-leakage Flowserve seals. Therefore, Technical Bulletin 15-1 is only applicable to a single Reactor Coolant System loop in each of the two units. Technical Bulletin 15-1 is being addressed through the Dominion corrective action program.

The North Anna FIP will be updated appropriately to reflect the supplemental information contained in the attachment to this letter.

Should you have any questions or require additional information, please contact Ms. Diane E. Aitken at (804)273-2694.

Respectfully, Mark Sartain Vice President - Nuclear Engineering COMMONWEALTH OF VIRGINIA)

COUNTY OF HENRICO)

The foregoing document was acknowledged before me, in and for the County and Commonwealth aforesaid, today by Mr. Mark D. Sartain, who is Vice President - Nuclear Engineering, of Virginia Electric and Power Company. He has affirmed before me that he is duly authorized to execute and file the foregoing document in behalf of that company, and that the statements in the document are true to the best of his knowledge and belief.

Acknowledged before me this .IL". day of ""'* 2016.

My Commission ExPires: ZO.m\tZ~t.

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... *Notary Public Commonwealth of Virginia /6

~Reg. # 7520495 My Commnission Expires January 31, 2O0.jpg*

Serial No. 14-394E Docket Nos. 50-338/339 Order EA-12-049 Supplemental Information Page 2 of 3 Attachments: Supplemental Information Regarding FLEX Mitigating Strategies Commitments contained in this letter:

1) The North Anna FIP will be updated, as appropriate, to reflect the Supplemental Information contained in the attachment to this letter.

cc: U.S. Nuclear Regulatory Commission - Region II Marquis One Tower 245 Peachtree Center Avenue, NE Suite 1200 Atlanta, GA 30303-1257 Dr. V. Sreenivas NRC Project Manager North Anna U.S. Nuclear Regulatory Commission One White Flint North Mail Stop 08 G-9A 11555 Rockville Pike Rockville, MD 20852-2738 Mrs. Lisa M. Regner U.S. Nuclear Regulatory Commission One White Flint North Mail Stop 011 Fl 11555 Rockville Pike Rockville, MD 20852-2738 Mr. Blake A. Purnell U.S. Nuclear Regulatory Commission One White Flint North Mail Stop 012 D20 11555 Rockville Pike Rockville, MD 20852-2738 Mr. Steven R. Jones U.S. Nuclear Regulatory Commission One White Flint North Mail Stop 010 Al 11555 Rockville Pike Rockville, MD 20852-2738 NRC Senior Resident Inspector North Anna Power Station

Serial No. 14-394E Docket Nos. 50-338/339 Order EA-12-049 Supplemental Information Page 3 of 3 Mr. J. E. Reasor, Jr.

Old Dominion Electric Cooperative Innsbrook Corporate Center, Suite 300 4201 Dominion Blvd.

Glen Allen, Virginia 23060

Serial No. 14-394E Docket Nos. 50-338/339 Order EA-12-049 Supplemental Information Attachment Supplemental Information Regarding FLEX Mitigating Strategies Virginia Electric And Power Company (Dominion)

North Anna Units 1 and 2

Serial No. 14-394E Docket Nos. 50-338/339 Order EA-12-049 Supplemental Information Page 1 ofl18 Supplemental Information Regarding FLEX Mitigating Strategies North Anna Units I and 2 On December 8, 2014, Dominion notified the NRC staff that North Anna Unit 2 had met the requirements of Attachment 2 from Order EA-12-049 (Reference 1). On May 19, 2015, Dominion notified the NRC staff that North Anna Power Station was in full compliance with the Order for both Units 1 and 2 and accordingly provided the Final Integrated Plan (FIP) for the North Anna FLEX Mitigation Strategies (Reference 2).

On August 4, 2015, a phone conference was held between Dominion and the NRC staff to discuss various elements of Dominion's responses to Interim Staff Evaluation (ISE) Open Item 3.2.1.2.B and Safety Evaluation (SE) items #4, #8, #16, and #17.

Additionally, Westinghouse Technical Bulletin 15-1 was included in the discussions that occurred during the call.

Many items were resolved/clarified during the phone conference based on previously submitted information. However, the NRC staff identified several items for which additional information was needed to supplement the previously provided responses to ISE 01 3.2.1.2.B and SE items #4, #8, #16, and #17. This attachment identifies the additional information requested and provides Dominion's response as a supplement to the original responses. The North Anna Final Integrated Plan will be updated appropriately to reflect the supplemental information contained in the attachment to this letter.

ISE 01 3.2.1.2.B: Flowserve N-9000 Leakage Rate During the August 4, 2015 phone conversation, the NRC staff requested that the licensee address the following conditions and limitations regarding the draft Revision 1 to the Flowserve seals white paper, namely:

1. Confirm that plant design and planned mitigation strategy are consistent with the information assumed in the calculation performed by Flowserve, which is summarized in Table 1 of the draft white paper.

2. Confirm that the peak cold-leg temperature, prior to the cooldown of the reactor coolant system assumed in Flowserve's analysis, is equivalent to the saturation temperature corresponding to the lowest setpoint for main steam line safety valve lift pressure.

3. In its white paper, Flowserve has generally specified leakage rates in volumetric terms. For converting the specified volumetric flow rates to mass flow rates, licensees should use a density of 621Ibm/ft 3 (approximately 993 kg/in 3 )

throughout the ELAP event. This condition reflects observations made during testing conducted by Flowserve involving a loss of seal cooling, wherein the

Serial No. 14-394E Docket Nos. 50-338/339 Order EA-12-049 Supplemental Information Page 2 of 18 seal leakage mass flow rate remained roughly constant over a range of pressure values.

Dominion Supplemental Response:

1, 2) The information in Table 1 of the draft Revision 1 to the Flowserve seals white paper is consistent with the North Anna plant design and planned mitigation strategy. The parameters in Table 1 are documented in ETE-NAF-2012-0150 (Reference 3) which has been provided to the NRC staff for review. This includes the peak cold-leg temperature which corresponds to the lowest setpoint for the main steam safety valves. Confirmation of this information by Dominion is also documented within Table 1 of the draft Revision 1 of the Flowserve seals white paper.

3) Regarding the use of a cold density of 62 Ibm/ft3 in the Flowserve evaluation for seal leakage, further evaluation results have indicated that the impact in RCP leakage from the Flowserve seals is minimal and does not impact the timeline for the current strate~ly in terms of when to initiate RCS makeup. Thus, a density of 62.4 Ibm/ft° was employed for all conditions. (The density of 62.4 Ibm/ft3 represents the density at standard conditions.) The evaluation resulted in an estimated decrease in the time for RCS makeup initiation of 0.9 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months and is documented in ETE-NAF-2012-0150.

SE #4 - NSAL 14-1. W RCP Seal Leakage:

During the August 4, 2015 phone conversation, the NRC staff requested the following information to supplement the response to SE Item #4 previously provided in of Reference 1.

1) Provide the plant-specific design parameters associated with the seal leakoff line and confirm whether they are bounded by each of the model input parameters in Table 2 of PWROG-14015-P for the appropriate analysis category. If any parameters in Table 2 are not bounded, please provide justification that the generically calculated leakage rate and maximum pressure are applicable.

2) Provide the set pressure and flow area associated with the relief valve on the #1 seal leakoff line common header piping.

3) Provide an estimate of the piping diameter, length, and number and type of components for the seal leakoff line common header piping.

Serial No. 14-394E Docket Nos. 50-338/339 Order EA-12-049 Supplemental Information Page 3 of 18 Dominion Supplemental Response:

Westinghouse OEM RCP shaft leakage seals are currently installed in one RCP in each unit, i.e., Unit 1 "C" RCP and Unit 2 "A" RCP. The remaining RCPs, i.e., Unit I "A" and "B" RCPs and Unit 2 "B" and "C" RCPs, have Flowserve low leakage N-9000 seals as replacements for the Westinghouse OEM seals. The two RCPs with Westinghouse OEM RCP seals are scheduled to be replaced with Flowserve N-9000 seals in upcoming outages. Therefore, this supplemental response is only applicable to the Unit 1 "C" RCP and Unit 2 "A" RCP No. 1 seal leak off lines.

1) The input parameters for Category 3 seal leakoff line configurations in Table 2 of PWROG-14015-P (Reference 9) are similar to the North Anna Westinghouse Reactor Coolant Pump (RCP) No. 1 seal leakoff line configurations. Table 2 essentially contains 3 groups of parameters: 1) piping upstream of the flow element, 2) the flow element, and 3) piping downstream of the flow element.

These parameter groups are discussed below.

The PWROG-14015-P Category 3 flow analysis model for the No. 1 seal line piping upstream of the flow orifices used 2' of 2.067" ID piping with one globe valve and one 900 bend. The North Anna No. 1 seal line piping configurations upstream of the orifice flow elements consist of a small amount of 1 1/2" (1.338" ID) and much more than 2' of 2" (1.687" ID) Sch. 160, Dominion Class 1502 piping with one 2" and one 3/4" globe valve (reduced ends from 1 1/2" pipe),

typically three 2" branch tees, and various amounts of 9Q0 and 450 bends in the flow path. The North Anna No. I seal leakoff line piping configurations upstream of the orifices are bounded by the PWROG-14015-P flow analysis model due to the higher flow resistance associated with the greater amount of piping and components in the North Anna piping configurations. The PWROG-14015-P Category 3 pressure analysis model for the No. 1 seal line piping upstream of the orifice flow elements used 57' of 1.338" ID piping with a generous number of piping components. The PWROG-14015-P pressure analysis model also bounds the primarily 2" (1.687" ID) North Anna No. 1 seal leakoff line piping configurations upstream of the orifices.

The North Anna No. 1 seal leakoff lines have a 0.375" diameter orifice flow elements in the Dominion Class 1502, high pressure portion of the lines, which is the same orifice diameter used for the PWROG-1401 5-P flow analysis model for Category 3 plants. The PWROG-14015-P pressure analysis model for Category 3 plants used a 0.340" diameter orifice. The 0.340" diameter orifice in the PWROG-14015-P Category 3 pressure analysis is bounding for North Anna and other plants with 0.375" orifices.

The PWROG-14015-P Category 3 flow analysis model for the No. 1 seal line piping downstream of the flow orifices, used 6' or 15' of 2.067" ID piping with no piping components. Take-offs for the No. I seal leakoff line piping downstream of

Serial No. 14-394E Docket Nos. 50-338/339 Order EA-12-049 Supplemental Information Page 4 of 18 the orifice flow elements for the Unit 1 "C" RCP and the Unit 2 "A" RCP are detailed in the response to item 3) below. The North Anna No. 1 seal leakoff line piping configurations are bounded by the PWROG-14015-P flow analysis model due to piping components and longer pipe lengths of the actual configurations.

The PWROG-14015-P Category 3 pressure analysis model for the No. 1 seal line piping downstream of the flow orifices used 60' of 1.5" ID piping with no piping components. The piping take-offs detailed in the response to item 3) below show that the North Anna piping configurations are not bounded by the PWROG-14015-P pressure analysis model. Consequently the ELAP maximum pressure in the No. I seal leakoff line could be slightly higher than predicted by the PWROG-14015-P pressure analysis. However, the evaluation documented in ETE-CPR-2015-1005 concluded that the structural integrity of the No. 1 seal leakoff line piping and piping components downstream of the orifices would not be challenged by the ELAP maximum pressures.

2) The set pressure and flow area of the relief valve on the #1 seal leakoff line common header piping are 150 psig and 1.290 in2 (Crosby relief valve - J orifice area), respectively.

3) The following are piping take-offs of the associated No. 1 seal leakoff lines downstream of the orifice flow element, which connect to the seal leakoff line common header piping and branch to the common header relief valve. The take-offs are for the Unit 1 "C" RCP No. 1 seal leakoff line and the Unit 2 "A" RCP No.

1 seal leakoff line. These RCPs are the only two RCPs equipped with Westinghouse/OEM seals. The other four RCPs (two each per unit) are equipped with Flowserve low leakage N-seals.

Piping Take-off: Unit I "C"RCP No. 1 Seal Leakoff Line - Orifice Flow Element to Common Header RV Orifice Flow Element to Pipe Class Break - Class 1502 to Class 153A Item Description Quantity Pipe 3/4/"~ Sch. 160 1 '- 6 15/16" 1/2," Sch. 160 1______ 0' - 33/8" Fittings 1 1/2"x 3/4" Reducing Insert Coupling, 1 6000 lb socket weld

________3/4" Tee (Branch) 6000 lb socket weld 1 Valves 3/4" Globe 1500 lb socket weld ends 1 Pipe Class Break - Class 1502 to Class I153A to Common Header Item Description Quantity Pipe 3/4"Sch. 40 0'- 5 5/8"

_______2" Sch. 40 67'- 8 3/4" Fittings 3/4" x 2" Increaser 3000 lb socket weld I 2" Tee (Run) 3000 lb socket weld 2 2" 900 EL 3000 lb socket weld 6 2" 450 EL 3000 lb socket weld 1

________2"x 3" Increaser Sch. 40 1

Serial No. 14-394E Docket Nos. 50-338/339 Order EA-1 2-049 Supplemental Information Page 5 of 18 Piping Take-off: Unit I "C" RCP No. 1 Seal Leakoff Line - Orifice Flow Element to Common Header RV (Cont.)

___________CommonHeader Item Description Quantity Pipe 3" Sch. 40 18'- 2" Fittings 3" Tee (Run) Sch. 40 1 3" Tee (Branch) Sch. 40 1 3" x 2" Reducer Sch. 40 1 Branch Line from Common Header to RV __________

Item Description Quantity Pipe 2" Sch. 40 11 '- 3 1/2" Fittings 2" 9Q0 EL 3000 lb socket weld 3

________2" 450 EL 3000 lb socket weld 1 Piping Take-off: Unit 2 "A" RCP No. 1 Seal Leakoff Line - Orifice Flow Element to Common Header RV Orifice Flow Element to Pipe Class Break - Class 1502 to Class 153A Item Description Quantity Pipe 3/4" Sch. 160 3-5 I 1/2A" Sch. 160 1' Fittings 1 1/2A" x 3/4" Reducing Insert Coupling, I 6000 lb socket weld 1 1/2" 9Q0 EL 6000 lb socket weld 1 ________

___________ 4" 900 EL 6000 lb socket weld 1 ________

Valves 3/4"4~ Globe 1500 lb socket weld ends 1 ________

Pipe Class Break - Class 1502 to Class 153A to Common Header Item Description Quantity Pipe 3/4" Sch. 40 7'- 11 3/16" 2" Sch. 40 21 '-5" Fittings 3/4" 9Q0 EL 3000 lb socket weld 2 3/4" x 2" Increaser 3000 lb socket weld 1 2" Tee (Run) 3000 lb socket weld 1 2" 9Q0 EL 3000 lb socket weld 5 2" x 3" Increaser Sch. 40 1 CommonHeader_________

Item Description Quantity Pipe 3" Sch. 40 37'- 4" 2" Sch. 40 12'- 7" Fittings -3" Tee (Run) Sch. 40 1 3" Tee (Branch) Sch. 40 2 3" x 2" Reducer Sch. 40 1 Branch Line from Common Header to RV Item Description Quantity Pipe 2" Sch. 401 -9 Fittings -2" 900 EL 3000 lb socket weld 1 2" 450 EL 3000 lb socket weld 1

________2" Tee (Branch) 3000 lb socket weld 1

Serial No. 14-394E Docket Nos. 50-338/339 Order EA-12-049 Supplemental Information Page 6 of 18 SE #8 - Boric Acid Batch Mixing Time:

During the August 4, 2015 phone conversation, the NRC staff requested the following information to supplement the response to SE Item #8 previously provided in of Reference 1. This supplement addresses: 1) the use of a cold density value of 62 Ibm/ft3 for the assessment of leakage from the North Anna Flowserve seals, 2) increased Flowserve seal leakage in a long-term scenario, and 3) the need and timeframe of additional equipment from offsite sources.

Dominion Supplemental Response:

As part of the discussion on the use of the Boric Acid Mixing Tank (BAMT) provided with the response to SE #8 in Attachment 2 of Reference 1, an assessment of Reactor Coolant System (RCS) inventory loss for the North Anna Units was presented. For the Flowserve Reactor Coolant Pump (RCP) N-seal leakage, the assessment used the water density associated with the RCS temperature and pressure applicable while the unit was at the target Steam Generator (SG) pressure of 290 psig. Based on Flowserve testing observations, the leakage from Flowserve N-seals should be evaluated at the cold density of 62 Ibm/ft 3 .

The following revised assessment is based on the new Flowserve N-seal leakage density value:

The rate of RCS inventory loss due to RCP seal leakage for the North Anna units has been revised assuming the current RCP seal configuration for each of the North Anna Units (1 Westinghouse seal and 2 Flowserve N-seals). Based on conservative RCP seal leakage rates for both the Westinghouse and Flowserve seals, the revised Flowserve seal leakge cold density, and the inclusion of a 1 gpm unidentified leakage rate, a total RCS leakage rate (per unit) of 2.25 Ibm/sec was determined with the unit at the target SG pressure of 290 psig. This corresponds to a RCS inventory reduction of 6,750 Ibm over a 50 min period.

Since the RCS injection rate is 45 gpm (i.e., the nominal flow rate of the BDB RCS Injection pump), a minimum addition of 900 gallons of borated water from the BAMT can be delivered every 50 minutes to each unit. The minimum 900 gallon addition per a 50 minute period is based on a 25 minute BAMT batching cycle (20 minutes injecting and 5 minutes valve alignment), alternating one BDB RCS Injection pump between two units. This RCS injection cycle corresponds to the addition of approximately 7,476 Ibm of RCS inventory makeup per unit over the 50 minute period. This makeup mass is

>10% more than the mass lost due to RCP seal leakage in the same time period.

Therefore, conservatively, based on the current configuration of installed RCP seals, RCP seal leak rates, the time necessary to inject batches of borated water using the portable BAMTs and one BDB RCS Injection pump, the RCS inventory for Units 1 and 2 would be increasing at a rate of approximately 726 Ibm every 50 minute period or an average of 871 Ibm/hour.

Serial No. i4-394E Docket Nos. 50-338/339 Order EA-12-049 Supplemental Information Page 7 of 18 Subsequently, an additional portable RCS Injection pump is available from the National SAFER Response Center (NSRC) at approximately 28 hours3.240741e-4 days 0.00778 hours 4.62963e-5 weeks 1.0654e-5 months following the onset of the ELAP event. Assuming only the two BAMTs already in use are available and each of the two units is aligned with one RCS Injection pump and one BAMT, the batch time cycle for each unit would be approximately 30 minutes. The 30 minute batch cycle time is based on 20 minutes to inject the 900 gallons from the BAMT into the RCS followed by a 10 minute period to add boric acid crystals while refilling the tank and agitating the contents prior to and during injection. This ongoing process of using a dedicated BAMT and RCS Injection pump for each unit would be able to inject 1,800 gallons of borated water every hour or 14,950 Ibm/hour. This is well in excess of the leakrate discussed above.

Finally, Flowserve seal leakage is expected to increase after several days at which time the additional RCS Injection pump from the NSRC will have been received and placed into service. The RCS inventory loss due to the increased Flowserve seal leakage would increase to 2.72 Ibm/sec. This leakage rate corresponds to a 9,792 Ibm/hour reduction in RCS inventory. As stated in the previous paragraph, the RCS makeup capability at the time of this increased seal leakage is 14,950 Ibm/hr which easily accomodates the increase in leakage from the Flowserve seals.

SE #16 -W RCP Seal, Seal Leakage Rates:

During an August 4, 2015 phone conversation, the NRC staff requested the following information to supplement the response to SE Item #16 previously provided in of Reference 1.

1) The NRC staff discussed the status of the issue as open generically and discussed the technical rationale, a large part of which is associated with recently revealed test results from Karlstein and Montereau "cold-shock" that show higher leakage than the Montereau "hot-shock" test the PWROG used to benchmark ITCHSEAL. The Staff noted that the existing licensee response did not cover the current state of the issue. The staff noted that, with augmented staffing available, the licensee should be able to provide makeup prior to 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months , which could facilitate resolution. The licensee stated it considers the existing amount of conservatism in its current RCS makeup initiation time sufficient and provided rationale. The staff suggested additional discussion to reach resolution after the staff and licensee have further reflected and refined their positions.

2) The staff considered the magnitude of the interpolation issue to be relatively small and expected it to be bounded by the existing 1.6-hr margin. However, further discussion noted that, during the resolution process for Westinghouse RCP seal leakage issues over the past fall/winter, some staff questions were resolved by referencing the margin (considered large at the time) associated

Serial No. 14-394E Docket Nos. 50-338/339 Order EA-12-049 Supplemental Information Page 8 of 18 with the choice of the RFACT value used in the ITCHSEAL code. However, in light of the leakage rates now known from the Karistein and Montereau "cold-shock" tests, it is appropriate to reconsider the total margin available versus what is necessary to confirm the balance is appropriate. Past discussion between the PWROG and NRC staff indicated different characterizations of the magnitudes of some issues that were resolved by appealing to margin; thus some need for reconciliation of views may be necessary.

The NRC staff has identified three specific areas of potential non-conservatisms in RCP seal leakage assessments which were previously accommodated by the NRC within the margin of the ITCHSEAL code benchmark. Whereas this benchmark margin has been potentially earmarked by the staff to accommodate the unknown issues associated with the cold stock test and the Karlstein test, these areas require reconsideration. The following 'Issues' are paraphrased from NRC staff statements.

Issue 1. PWROG-14027-P, Rev 2 claimed that the effect of linear interpolation of data points for leak rate vs. pressure, which is a non-conservatism, is less than 0.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months . NRC staff did not agree; "There is no basis to conclude this effect is less than 0.5 hr from true functional form.

Presumably it will be possible to track this non-conservatism and address via the demonstration of significant margin associated with the benchmarking work... Note ITCHSEAL is not regarded as capable of handling anything but steady-state calculations. This adds non-conservative uncertainty to the entire calculation process." (These comments do not address PWROG-14027-P, Rev 3, which attempts to quantify the effect of assuming linear interpolation and show that this assumption is bounded by the conservatism in the RCP seal leakage from PWROG-14015-P.)

Issue 2. For cases with high leakage rates, prior to cooldown, it is conservative and in some cases prototypical to assume that RCS approaches a Thot saturation condition earlier (potentially from beginning of event in cases with very high leakage), rather than averaging with the nominal post-trip condition.

The NRC staff believes this issue should be tracked and the non-conservatism should be addressed via the demonstration of significant margin associated with the benchmarking work. Rough staff calculations using ITCHSEAL along with TRACE-calculated plant conditions suggest the effect could be approximately 20-30 minutes for a Category 3 plant with 3 or 4 loops (like NAPS). This is double to triple the impact calculated by Westinghouse.

Issue 3. The RCS pressure may not reach the target pressure of 310 psia very quickly. Rather, the RCS may cool down to about 320 psia over 6 to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , and hold at 320 psia for an extended period. At these conditions, the

Serial No. 14-394E Docket Nos. 50-338/339 Order EA-12-049 Supplemental Information Page 9 of 18 subcooling is at or near the value of 15°F where ITCHSEAL predicts the subcooling effect saturates out (i.e., little difference in leakage with subcooling). Thus, the leakage rate being used appears non-conservative.

The ITCHSEAL calculations show that once the subcooling effect has been saturated out [less than ~.20°F] then increasing the pressure increases leakage. Based on rough values, leakage at cooldown conditions could be underestimated by about 10-15% for Category 1. The effect on other categories is not completely clear. PWROG should estimate the impact and address. A very rough staff estimate suggests that for Category 1, the effect could be on the order of 0.5 hr.

Dominion Supplemental Response:

1) The following discussion reflects Dominion's position regarding the impact of recent test results on Westinghouse seal leakage rates:

North Anna is expected to commence a cooldown prior to two hours and to complete this cooldown by four hours after all AC power is lost. Based on the discussions below, it is concluded that substantial margin exists in the overall analytic approach to determine the FLEX strategy RCS make-up times and, therefore, an RCS makeup time of 16.0 hours0 days 0 hours 0 weeks 0 months is acceptable for compliance with the NRC Order EA-12-049 (Reference 13).

Reactor Coolant System Response Analyses to RCP Seal Leakaae RCS makeup is required at 17.6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months based on the reflux cooling criterion. The boration requirement is non-limiting at 37 hours4.282407e-4 days 0.0103 hours 6.117725e-5 weeks 1.40785e-5 months . The North Anna RCS makeup evaluation (Reference 3) utilized RCP seal leakages based on the results from the Westinghouse ITCHSEAL code as documented in PWROG-14015-P (Reference 9). The basis for the evaluation of the RCS makeup time is the NOTRUMP reference case for the Westinghouse 3-loop plant with the Thor upper head.

The NOTRUMP analyses were initially presented in WCAP-17601 (Reference

5) and WCAP-1 7792 (Reference 6) with further, more design-specific results published in OG-14-60 (Reference 7). Westinghouse determined the times at which the two-phase loop flow rate becomes less than the loop flow rate corresponding to single-phase natural circulation for the Westinghouse reference cases for 2-loop, 3-loop, and 4-loop plants. These reference cases included an initial RCP seal leakage of 21 gpm per RCP that was reflective of a critical flow relationship. The times where the bulk two-phase loop flow rate drops below the single phase flow rate values are 21.4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , 17.2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , and 20.8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months for the two, three and four loop reference cases, respectively. Using the definition for reflux cooling, when SG U-bend flow quality exceeds a value of

Serial No. 14-394E Docket Nos. 50-338/339 Order EA-12-049 Supplemental Information Page 10 of 18 0.1, Westinghouse determined times of 28.1 hours1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , 27.8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months and 17.2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for two, three and four ioop Westinghouse plants, respectively.

The PWROG documented a white paper on NOTRUMP in PWROG-14064-P (Reference 11). The purpose of this white paper was to document the applicability of the NOTRUMP code for the evaluation of the ELAP event and application of its results with regards to criteria for boron mixing and reflux cooling for Westinghouse designed PWRs. From PWROG-14064-P, the comparison of results from the NOTRUMP and TRACE computer codes for the parameters of interest show that the NOTRUMP predicted results agree well or are conservative with respect to the NRC's TRACE predicted results when key input variables and boundary conditions are applied in a consistent manner.

The comparison showed that NOTRUMP provides a conservative estimate of the required time when the primary make-up pumps are required for an FLAP event. Therefore, it is concluded that NOTRUMP is acceptable for simulation of the ELAP event within the criteria for reflux cooling and boron mixing.

Application of the NOTRUMP simulations reference cases requires the implementation of the RCS makeup pump at the times in Table 2.

The NRC provided an endorsement of NOTRUMP for ELAP events with conditions/restrictions (Reference 20). As indicated in Section 3.4 of Reference 3, these conditions/restrictions are met.

Table 2 - RCS Make-up Time for Various Westinghouse Plant Designs Plat Plat Cnfiuraions)

Cnfiuraions) Required RCS Makeup Time Pumpfor 4-loop, TcoId Upper Head 4-loop, That Upper Head 1. or 3-loop, Th17UpperoHea 2-loop, Thot Upper Head 3-loop, TcoId Upper Head 16.4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Assessments for Plant Specific RCP Seal Leakaaqe Response:

Nuclear Safety Advisory Letter 14-1 (NSAL 14-1, References 14 and 15) was issued to identify an issue related to the Westinghouse OEM reactor coolant pump (RCP) seal leakage, as affected by the piping configuration and components of the No. 1 seal leakoff line, during events resulting in the loss of seal cooling. The critical flow relationship based on a RCP seal leak rate of 21 gallons per minute (gpm) for each RCP, as used in WCAP-17601 (Reference 5) and WCAP-17792 (Reference 6), was stated to be not applicable for all plants

Serial No. 14-394E Docket Nos. 50-338/339 Order EA-12-049 Supplemental Information Page 11 of 18 with Westinghouse RCPs because of the various thermal-hydraulic conditions set up by plant specific seal leakoff piping designs.

The PWROG initiated an effort to evaluate the RCP seal leakoff and produced several reports. PWROG-14008-P (Reference 8) documents a categorization of the Westinghouse plants according to the seal leakoff line configuration based on survey results from participating plants. A parametric study was conducted in order to evaluate the significance of the various RCP seal leak-off line hydraulic parameters. The major limitation in the categorization was determined to be the type and diameter of the flow measuring device in each leak-off line.

NAPS Units 1 and 2 were identified as Category 3 plants.

PWROG-14015-P (Reference 9) transmitted the results of the calculation of the seal flow rates for each category of plant identified in PWROG-14008-P. The analysis of maximum flow values for the seal leak-off lines was performed using the Westinghouse two phase flow code ITCHSEAL (previously used in WCAP-10541). The maximum flow analysis used the largest flow element in a given category, along with the minimum component resistances, minimum piping length, and maximum piping diameters based on the information provided in the plant survey. The goal of the maximum flow leak-off line model was to minimize resistance, and therefore maximize the predicted flow when all RCP seal cooling is postulated to be lost. Westinghouse has performed and documented sufficient calculations to confirm the reasonableness of linear interpolation between predicted points, that peak leakage occurs at 1500 psia, and to include minimal subcooling.

Benchmarkina Analysis of RCP Seal Leaka~qe Models:

Westinghouse performed a benchmark of the ITCHSEAL code against the data from the :lectricit6 de France (EDE) hot-shock test performed at the Montereau facility and documented the results in PWROG-14074-P (Reference 12). The hot-shock test data are documented in Appendix B of WCAP-1 0541. The information provided in that document was used to construct a model of the test configuration, including the No. 1 seal leak-off line. ITCHSEAL analysis was performed at five different pressure and temperature conditions from the hot-shock test. This model includes a model 93D pump, a 7 inch aluminum oxide seal, and the No. 1 seal leakoff line that was used at the facility. Based on an investigation of a variety of different factors, it was concluded that a leak-off line orifice exit REACT value of 25 (ITCHSEAL input for fL/D resistance to simulate the pressure drop across the flow measurement orifice) yielded a good match to the Montereau hot shock test measured flow rate. The benchmarking results for the Montereau configuration show a margin of approximately 100% for an orifice, exit REACT value of 0. The benchmarking results also demonstrated that the flow rate results in PWROG-14015-P for the plant categories with orifice (i.e.,

Category 1, 2, 3, and 6) configurations show a margin of 80% to 100% for an orifice exit REACT value of 0. PWROG-14074-P concluded that "While it is not

Serial No. 14-394E Docket Nos. 50-338/339 Order EA-12-049 Supplemental Information Page 12 of 18 possible to precisely quantify the margin available for each plant category and each temperature and pressure condition, the information available supports the conclusion that the (PWROG-14015-P) flow results are conservative and significant margin is available." (It is noted that the NRC staff has reviewed PWROG-14074-P, Revision 0 and has provided informal agreement with the results and conclusions stated therein subject to final review of all RCP seal leakage issues.)

Other Tests The NRC staff has recently become aware of other testing of RCP seals subsequent to their review and the issuance of PWROG-14074-P. The NRC staff notes that the existing North Anna response does not cover these issues.

Dominion has considered the NRC staff concerns with regards to the the other tests. In a PWROG letter dated November 5, 2015 [Reference 17], the PWROG informed the NRC that AREVA, Inc had provided a letter dated November 4, 2015 [Reference 18] in support of the Pressurized Water Reactor Owner's Group (PWROG) to the NRC. This AREVA letter provides a summary of information provided to the NRC staff during an Audit on October 27, 2015. The information characterized the effects of hydrothermal corrosion on reactor coolant pump (RCP) seals. The PWROG letter states that this information has been reviewed and it has been determined that the leakage values in PWROG-14015-P, Revision 2 are bounding for Categories 1, 2, 3, and 6 for plants without a shutdown seal when an early cooldown is performed as recommended in the PWROG Core Cooling Position Paper [Reference 19].

Westinghouse, as sponsored by the PWROG, has issued generic Emergency Response Guidelines (ERGs) that have incorporated the early (first) cooldown.

In addition, the ERGs have also incorporated a second cooldown which would be performed in Loss of all AC Power Situations (e.g., ELAP). The timing of this second cooldown is dependent on meeting boration requirements and isolating or venting the cold leg accumulators. This second cooldown will provide additional beneficial conditions with regards to reducing RCP seal leakage. The ERGs are being incorporated into the site-specific Emergency Operating Procedures (EOPs).

As the information shared at the Audit is the most definitive test data with regards to the 'OEM' equivalent seals at North Anna, Dominion has concluded that the information used herein from PWROG-14015-P, Revision 2 is acceptable to determine FLEX plan RCS make-up times and therefore an RCS makeup time of 16.0 hours0 days 0 hours 0 weeks 0 months is acceptable for compliance with the NRC Order EA-12-049 (Reference 13).

Serial No. 14-394E Docket Nos. 50-338/339 Order EA-12-049 Supplemental Information Page 13 of 18

2) Issue 1 Response:

The North Anna evaluation of the impact of the RCP leakage on the time to initiate RCS makeup used all applicable points from PWROG-14015-P (Reference 9); therefore, the effect of linear interpolation of the leak rate vs.

pressure does not apply to the North Anna evaluation. Dominion concludes that no additional margin is required for this issue for North Anna Units 1 and 2.

Additional support comes from an undocumented evaluation using only three points which resulted in essentially the same time to initiate RCS makeup. This is consistent with the information in PWROG-14015-P (Reference 9) which shows the intermediate points lying approximately on the lines between the three points for a Category 3 seal, and thus, supports the use of interpolation without introducing significant non-conservatisms.

Issue 2 Response:

For this issue, the concern is that the plant may be at a lower pressure than the pressure at peak leakage (i.e., 1500 psia) before the cooldown starts. The North Anna evaluation of the impact of RCP leakage on the time to initiate RCS makeup linearly reduced the pressure from the beginning of the transient to the final target pressure at the end of the cooldown period. This choice was based on the 4-Loop case that showed a nearly linear depressurization in Figure 5.2.2-2 of WCAP-17601 (Reference 5). The cold-leg temperature was held constant at 572°F until the cooldown was initiated. A pressure of 1500 psia is reached in

~1.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months . The ROS first reaches a nearly saturated condition near the beginning of the cooldown at 1250 psia and 568°F. The cooldown with depressurization causes the RCS to be subcooled until reaching the final conditions of 300 psia and 41 5°F. As noted in the response to Issue 3 below, the North Anna evaluation used applicable points below the peak leakage at 1500 psia that have 5°F or less subcooling.

The North Anna evaluation is judged to be representative of the expected cooldown. However, an undocumented, and unrealistic, evaluation was performed that applied the maximum leakage from the start of the event until the cooldown. The reduction in the time to initiate RCS injection was 0.2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months (12 minutes).

In addition, it is noted that the North Anna units have a steam generator (SG) design pressure of 1100 psia as opposed to the reference design pressure values used in the PWROG work which was 1200 psia. Thus, the lower North Anna SG safety valve set-points will dictate an initial TcoId temperature response that is approximately 560°F as opposed to the PWROG-14015-P assumed temperature of 572°F. This will in effect reduce leakage rate in the pre-cooldown time frame.

Serial No. 14-394E Docket Nos. 50-338/339 Order EA-12-049 Supplemental Information Page 14 of 18 The North Anna evaluation appropriately considers the reduction in RCS pressure prior to the cooldown; therefore, Dominion concludes that no additional margin is required for this issue for North Anna Units 1 and 2.

Issue 3 Response:

The North Anna evaluation of the impact of the RCP leakage on the time to initiate RCS makeup used all the applicable points from PWROG-14015-P (Reference 9). The applicable points below the peak leakage at 1500 psia have 5°F or less subcooling. An evaluation of subcooling in Section 6.8 of PWROG-14015-P (Reference 9) showed the difference in leakage to be less than 0.1 gpm for the change from 5°F of sub-cooling to less than 10 F of subcooling.

However, the difference in leakage was shown to be less than 0.2 to 0.3 gpm for the change from 5°F of subcooling to 15°F of subcooling and 2 gpm less for the change to 40°F of subcooling.

The North Anna cooldown will be affected by reducing temperature of the RCS in a linear fashion until the target temperature is reached. The RCS pressure will also decrease in a linear fashion to the target pressure unless the leakage is small. If the RCS pressure is slightly above the target pressure at the end of the cooldown, then the RCS will be slightly subcooled. Hence, the assumed leakage being from a saturated or near saturated condition will be conservative with regards to the actual leakage until the pressure is further reduced. To confirm this, Westinghouse has performed additional ITCHSEAL runs at conditions with higher subcooling near the target temperature with higher pressure. These results show that the leakage is insensitive to the pressure at the lower temperatures.

The North Anna evaluation appropriately considers the cooldown by using the data with minimal subcooling for the temperatures encountered during the cooldown. Therefore, Dominion concludes that no additional margin is required for this issue.

SE #17 - Seal Leakoff Line Overpressurization:

During the August 4, 2015 phone conversation, the NRC staff requested the following information to supplement the response to SE Item #17 previously provided in of Reference 1.

1) The licensee should confirm that all piping and components in the leakoff line upstream of and including the flow measurement orifice can tolerate pressures greater than or equal to RCS design pressure (2500 psia).

2) The previous response stated that all Class 153A valves are downstream of the relief valve; however, the staff noted in drawing 1 1715-FM-095C,

Serial No. 14-394E Docket Nos. 50-338/339 Order EA-1 2-049 Supplemental Information Page 15 of 18 Sheet 2, Block F-5, that there actually appears to be a Class 153A valve that is upstream of the relief valve. (Based on the understanding that choking would occur at the flow orifice, the failure of downstream components should not impact the overall leakage rate assumed in the ELAP analysis.)

Dominion Supplemental Response:

1) The Westinghouse RCP No. I seal leakoff line piping upstream of and including the orifice flow element is Dominion Class 1502 piping with design pressure in excess of plant operating and transient pressures. An evaluation of the margin available between the Class 1502 piping design and postulated event maximum operating pressure / temperature conditions given in NSAL-15-2 (2,030 psig at 572°F) has been performed and has demonstrated that more than adequate margin is available. The evaluation considered the stress analyses of record for both Units 1 and 2 and is documented in engineering evaluation ETE-CPR-2015-1005 (Reference 10). This ETE has been provided to the NRC staff for review.

2) The No. 1 seal leakoff lines downstream of the first isolation valve downstream of the orifice flow element consist of Dominion Class 153A piping/components.

None of the Class 153A piping system components in the No. 1 seal leakoff lines between the orifice flow element and the relief valve were evaluated to be susceptible to loss of structural integrity from over-pressurization resulting from the maximum operating pressure / temperature conditions given in PWROG-14015-P (Reference 9) (i.e., the conditions assumed to be associated with postulated event maximum operating pressure I temperature conditions for an ELAP). The valve not considered in the previous response to SE #17 is a 3/4"4~

drain line located in the relief valve inlet piping, which is in close proximity to the relief valve. As is the case for other Class 153A components, the drain valve has a maximum working pressure rating at the postulated event maximum pressure that is in of excess of the 150 psi relief valve setpoint. Even though the structural integrity of the Class I153A piping and components would not be expected to have their structural integrity challenged during the ELAP postulated event maximum pressure /ltemperature conditions, as stated in NSAL 15-2 (Reference 16), failure of the piping downstream of the orifice flow element is not critical, as RCP seal leakage would not be impacted because the flow (two-phase) chokes at the orifice.

Serial Nos.No. 14-394E Docket 50-338/339 Order EA-12-049 Supplemental Information Page 16 of 18

References:

1. Virginia Electric and Power Company letter to NRC, "North Anna Power Station Unit 2 - Status of Required Actions for EA-12-049 Issuance of Order to Modify Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events," dated December 8, 2014.

2. Virginia Electric and Power Company letter to NRC, "North Anna Power Station Units I and 2 - Compliance Letter and Final Integrated Plan In Response to the March 12, 2012 Commission Order Modifying Licenses With Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events (Order EA-12-049)," dated May 19, 2015.

3. ETE-NAF-2012-0150, Revision 3, "Evaluation of Core Cooling Coping for Extended ,Loss of AC Power (ELAP) and Proposed Input for Dominion's Response to NRC Order EA-12-049 for Dominion Fleet," November 2015.

4. WCAP-10541-P, Revision 2, "Westinghouse Owners Group Report Reactor Coolant Pump Seal Performance Following a Loss of All AC Power,"

November 1986.

5. WCAP-1 7601 -P, Revision 1, "Reactor Coolant System Response to the Extended Loss of AC Power Event for Westinghouse, Combustion Engineering and Babcock & Wilcox NSSS Designs," January 2013.

6. WCAP-17792-P, "Emergency Procedure Development Strategies for the Extended Loss of AC Power Event for all Domestic Pressurized Water Reactor Designs," December 2013.

7. PWROG Letter OG-14-60, "Generic Information to Support Requests for Additional Information in USNRC Reviews of FLEX Overall Integrated Plans with Regard to Reflux Cooling, LTR-LIS-14-79, (PA-ASC-1197)," dated February 13, 2014.

8. PWROG-14008-P, Revision 2, "No. I Seal Flow Rate for Westinghouse Reactor Coolant Pumps Following Loss of All AC Power, Task 1:

Documentation of Plant Configurations," September 2014.

9. PWROG-14015-P, Revision 2, "No. I Seal Flow Rate for Westinghouse Reactor Coolant Pumps Following Loss of All AC Power, Task 2: Determine Seal Flow Rates," April 2015.

Serial No. 14-394E Docket Nos. 50-338/339 Order EA-12-049 Supplemental Information Page 17 of 18

10. ETE-CPR-2015-1005, Rev. 0, "Evaluation of RCP No. I Seal Leakoff Line for Potential Over-Pressure Conditions during an ELAP," October 2015.

11. PWROG-14064-P, Revision 0, "Application of NOTRUMP Code Results for PWRs in Extended Loss of AC Power Circumstances," September 2014.

12. PWROG-14074-P, Revision 0, "No. 1 Seal Flow Rate for Westinghouse Reactor Coolant Pumps Following Loss of All AC Power, Task 8:

Benchmarking the ITCHSEAL Code," April 2015.

13. NRC Order EA-12-049, "Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events," March 12, 2012 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML12054A736).

14. Westinghouse Nuclear Safety Advisory Letter 14-1 (NSAL 14-1), Revision 0, "Impact of Reactor Coolant Pump No. 1 Seal Leakoff Piping on Reactor Coolant Pump Seal Leakage During a Loss of All Seal Cooling," February 10, 2014.

15. Westinghouse Nuclear Safety Advisory Letter 14-1 (NSAL 14-1), Revision 1, "Impact of Reactor Coolant Pump No. 1 Seal Leakoff Piping on Reactor Coolant Pump Seal Leakage During a Loss of All Seal Cooling," September 9, 2014.

16. Westinghouse Nuclear Safety Advisory Letter 15-2 (NSAL 15-2), Revision 0, "Impact of a Break in the Reactor Coolant Pump No. 1 Seal Leak-off Line Piping on Seal Leakage During a Loss of Seal Cooling Event," dated March 23, 2015.

17. Letter from N. J. Stringfellow (PWROG) to USNRC, OG-15-413, "PWR Owners Group, Acknowledgement of AREVA's Transmittal of Summary Information of NRC Staff Audit of October 27, 2015 Regarding RCP Seal Leakage," dated November 5, 2015.

18. Letter from P. Salas (AREVA) to USNRC, NRC:15:043, "Submittal of the AREVA Jeumont (France) Summary Slides Provided in Support of Pressurized Water Reactor Owners Group (PWROG) Audit, October 27, 2015," dated November 4, 2015.

Serial No. 14-394E Docket Nos. 50-338/339 Order EA-12-049 Supplemental Information Page 18 of 18

19. Letter from N. J. Stringfellow (PWROG) to USNRC, OG-13-20, "PWR Owners Group, For Information Only - WCAP 17601-P, Revision I "Reactor Coolant System Response to the Extended Loss of AC Power Event for Westinghouse, Combustion Engineering and Babcock & Wilcox NSSS Designs" and "PWROG Core Cooling Position Paper", Revision 0 (PA-ASC-0916 and PA-PSC-0965)," dated January 30. 2013.

20. Letter from Jack R. Davis (NRC) to J. Stringfellow (PWROG), "Letter to PWROG - NOTRUMP Endorsement of ELAP Events," June 16, 2015.

Note: References 3 and 10 above have previously been provided to the NRC staff on the ePortal and are available for their review.

VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261 U. S. Nuclear Regulatory Commission January 6, 2016 Serial No.: 14-394E Attention: Document Control Desk NLOS/DEA: R0 Washington, DC 20555-0001 Docket Nos.: 50-338/339 License Nos.: NPF-4/7 VIRGINIA ELECTRIC AND POWER COMPANY (DOMINION)

NORTH ANNA POWER STATION UNITS I AND 2 SUPPLEMENTAL INFORMATION REGARDING THE MARCH 12. 2012 COMMISSION ORDER MODIFYING LICENSES WITH REGARD TO REQUIREMENTS FOR MITIGATING STRATEGIES FOR BEYOND-DESIGN-BASIS EXTERNAL EVENTS (ORDER NUMBER EA-12-049)

On March 12, 2012, the Nuclear Regulatory Commission (NRC) issued Order EA-12-049, "Order to Modify Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events" [the Order]. On May 19, 2015, Dominion notified the NRC staff that North Anna Power Station was in compliance with the Order for both Units 1 and 2 and provided the Final Integrated Plan (FIP) for the North Anna FLEX Mitigation Strategies. Subsequently, a call between Dominion and the NRC staff on August 4, 2015 identified several items for which the NRC staff requested additional information. The attachment to this letter provides the requested information for Interim Staff Evaluation Open Item 3.2.1.2.8 and Safety Evaluation Items #4, #8, #16, and #17.

Additionally, the August 4, 2015, phone call addressed Westinghouse Technical Bulletin 15-1, "Reactor Coolant System Temperature and Pressure Limits for the No. 2 Reactor Coolant Pump Seal." Each of the units at North Anna have replaced two of the three Reactor Coolant Pump seals with low-leakage Flowserve seals. Therefore, Technical Bulletin 15-1 is only applicable to a single Reactor Coolant System loop in each of the two units. Technical Bulletin 15-1 is being addressed through the Dominion corrective action program.

The North Anna FIP will be updated appropriately to reflect the supplemental information contained in the attachment to this letter.

Should you have any questions or require additional information, please contact Ms. Diane E. Aitken at (804)273-2694.

Respectfully, Mark Sartain Vice President - Nuclear Engineering COMMONWEALTH OF VIRGINIA)

COUNTY OF HENRICO)

The foregoing document was acknowledged before me, in and for the County and Commonwealth aforesaid, today by Mr. Mark D. Sartain, who is Vice President - Nuclear Engineering, of Virginia Electric and Power Company. He has affirmed before me that he is duly authorized to execute and file the foregoing document in behalf of that company, and that the statements in the document are true to the best of his knowledge and belief.

Acknowledged before me this .IL". day of ""'* 2016.

My Commission ExPires: ZO.m\tZ~t.

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