ML11307A228

Response to Requests for Additional Information, Restart Readiness Determination Plan

ML11307A228
Person / Time
Site:	North Anna
Issue date:	10/31/2011
From:	Price J Virginia Electric & Power Co (VEPCO)
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
11-566E
Download: ML11307A228 (32)

Text

{{#Wiki_filter:VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261 October 31, 2011 10 CFR 100, Appendix A U.S. Nuclear Regulatory Commission Ser Attention: Document Control Desk NLA Washington, DC 20555 Do Lice VIRGINIA ELECTRIC AND POWER COMPANY (DOMINION) NORTH ANNA POWER STATION UNITS I AND 2 ial No.: 11-566E &OS/ETS RO -ket Nos.: 50-338 50-339 ýnse Nos.: NPF-4 NPF-7 RESPONSE TO REQUESTS FOR ADDITIONAL INFORMATION (RAI) RESTART READINESS DETERMINATION PLAN By letters dated September 26, 28, and 30 and October 6, and 13, 2011, the NRC requested additional information regarding Dominion's Restart Readiness Determination Plan for North Anna Power Station following the August 23, 2011 Central Virginia earthquake. By letters dated September 27, 2011 (Serial No. 11-544), October 3, 2011 (Serial Nos. 11-544A and 11-566), October 10, 2011 (Serial Nos. 11-566A and 11-577), October 17, 2011 (Serial No. 11-544B), October 18, 2011 (Serial Nos. 11-566B and 11-577A), October 20, 2011 (Serial No. 11-566C) and October 28, 2011 (Serial Nos. 11-566D and 11-577B), Dominion provided responses to numerous RAIs. Upon reviewing Dominion's responses and as a result of several follow-up conference calls, the NRC has requested additional information to support the review. Accordingly, Dominion is providing responses to recent NRC questions/clarifications in the attachment to this letter. The specific technical review areas and the associated questions being answered are provided below for reference: NRC Request Date NRC Review Branch RAI Questions NRC Conference call EMCB Clarifications 1 through 4 October 26, 2011 NRC Conference call OctCobfere26,2011EMCB-Piping Clarifications 1 through 8 October 26, 2011 October 26, 2011 E-mail EMCB-Seismic Questions 1 through 3 October 26, 2011 E-mail EMCB-Piping/NDE Question 1 October 27, 2011 E-mail EMCB Piping Questions 1 and 2 October 27, 2011 E-mail License Renewal Questions 1 and 2 NRC Conference call Reactor Vessel Clarification of le October 28, 2011 Internals NRC Conference call Clarification Question 3 October 31, 2011 Long Term

Serial Number 11-566E Docket Nos. 50-338/339 Page 2 of 3 If you have any questions or require additional information, please contact Thomas Shaub at (804) 273-2763 or Gary D. Miller at (804) 273-2771. Sincerely, J. A

• re Vic Pr sident - Nuclear Engineering

Attachment:

Response to Request for Additional Information - Restart Readiness Determination Plan Commitments made in this letter.

1. The results of the comparison of the calculated load from the August 23, 2011 earthquake and the existing leak-before-break (LBB) analysis will be submitted by March 31, 2013.

COMMONWEALTH OF VIRGINIA COUNTY OF HENRICO The foregoing document was acknowledged before me, in and for the County and Commonwealth aforesaid, today by J. A. Price 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 61&ý" day of 000X-CL , 2011. M y C o m m is s io n E x p ir e s : 4 ) '0 L". 2 0.1ry ( l Ginger Lynn RutherfordI-otary Public NOTARY PUBLIC Commonwealth of Virginia Reg. # 310847 My Commission Expires 4/30/2015

Serial Number 11-566E Docket Nos. 50-338/339 Page 3 of 3 cc: U.S. Nuclear Regulatory Commission - Region II Marquis One Tower 245 Peachtree Center Ave., NE Suite 1200 Atlanta, Georgia 30303-1257 NRC Senior Resident Inspector North Anna Power Station M. Khanna NRC Branch Chief-Mechanical and Civil Engineering U. S. Nuclear Regulatory Commission One White Flint North Mail Stop 08 G-9E3 11555 Rockville Pike Rockville, MD 20852-2738 R. E. Martin NRC Project Manager U. S. Nuclear Regulatory Commission One White Flint North Mail Stop 08 G-9A 11555 Rockville Pike Rockville, MD 20852-2738 P. G. Boyle NRC Project Manager U. S. Nuclear Regulatory Commission One White Flint North Mail Stop 08 G-9A 11555 Rockville Pike Rockville, MD 20852-2738 J. E. Reasor, Jr. Old Dominion Electric Cooperative Innsbrook Corporate Center 4201 Dominion Blvd. Suite 300 Glen Allen, Virginia 23060

Serial Number 11-566E Docket Nos. 50-338/339 Attachment Response to Request for Additional Information Restart Readiness Determination Plan Virginia Electric and Power Company (Dominion) North Anna Power Station Units 1 and 2

Serial Number 11-566E Docket Nos. 50-338/339 Attachment Page 1 of 27 BACKGROUND By letters dated September 26, 28, and 30 and October 6, and 13, 2011, the NRC requested additional information regarding Dominion's Restart Readiness Determination Plan for North Anna Power Station following-the August 23, 2011 Central Virginia earthquake. By letters dated September 27, 2011 (Serial No. 11-544), October 3, 2011 (Serial Nos. 11-544A and 11-566), October 10, 2011 (Serial Nos. 11-566A and 11-577), October 17, 2011 (Serial No. 11-544B), October 18, 2011 (Serial Nos. 11-566B and 11-577A), October 20, 2011 (Serial No. 11-566C) and October 28, 2011 (Serial Nos. 11-566D and 11-577B), Dominion provided responses to numerous RAIs. Upon reviewing Dominion's responses and as a result of several follow-up conference calls, the NRC has requested additional information to support the review. Accordingly, Dominion is providing responses to recent NRC questions/clarifications. NRC Request for Information The following clarifications were requested in an October 26, 2011 conference call.

1. Provide a discussion of the battery rack seismic margins.

Dominion Response Main Station Batteries Main station batteries were inspected following the August 23, 2011 earthquake. No damage to any of the batteries, racks, and/or anchorage was identified. Main station batteries and locations are listed in the table below. Battery Room Station Battery Location 1-1 1-BY-B-01 SB 294' (Cable Spreading Room) 2-1 2-BY-B-01 SB 294' (Cable Spreading Room) 1-111 1-BY-B-03 SB 294'.(Cable Spreading Room) 2-111 2-BY-B-03 SB 294' (Cable Spreading Room) 1-11 1-BY-B-02 SB 254' (Emergency Switchgear Rm) 2-11 2-BY-B-02 SB 254' (Emergency Switchgear Rm) 1-IV 1-BY-B-04 SB 254' (Emergency Switchgear Rm) 2-IV 2-BY-B-04 SB 254' (Emergency Switchgear Rm) The batteries are supported by two-tier battery racks that have been designed to hold Exide (Enersys) "G" size cells. The racks were seismically evaluated during resolution to USI A-46, "Verification of the Seismic Adequacy of Mechanical and Electrical Equipment in Operating Reactors." The batteries that were installed at the time of the USI A-46 evaluation were subsequently replaced with Yuasa Exide 2GN-23 batteries.

Serial Number 11-566E Docket Nos. 50-338/339 Attachment Page 2 of 27 The Yuasa Exide 2GN-23 batteries were seismically qualified by test as documented in Wyle Test Report 46647-1. The replacement batteries are the same model (2GN-23) and of nearly identical size, weight, fit, and function to the batteries that were evaluated during the A-46 Program. Therefore, the rack evaluations documented during the Seismic Qualification Utility Group (SQUG) reviews are still applicable. A dynamic analysis of two-tier "G" size battery racks for North Anna was performed and is documented in Flight Dynamics International Report No. A-14-85, Revision 1, dated October 4, 1985. A finite element analysis model of the rack was developed using operating loading (weight of battery cells) plus seismic loading. The analysis used peak spectral accelerations corresponding to 0.5% of critical damping for the Operating Basis Earthquake (OBE) condition and 1% for the Design Basis Earthquake condition (DBE). However, consistent with North Anna's licensing basis, the use of 3% damping would have been acceptable. The analysis conservatively used OBE condition allowables as acceptance criteria for structural member stresses for enveloped OBE and DBE level loads, limiting member stresses to 75% of yield. For the DBE condition, stresses may be taken up to 90% of yield. Therefore, based on the conservatively high spectral accelerations used in the rack analysis corresponding to lower damping and the limiting allowable stresses used, the rack has margin to accommodate seismic loading experienced during the August 23, 2011 earthquake. This analysis is applicable for main station battery racks except 2-IV (which is of different model/make). However, as discussed further below the anchorage for the battery racks, including 2-IV were determined to be acceptable. During North Anna's resolution of USI A-46, the station batteries and battery racks were included within the scope and were inspected per the SQUG generic implementing procedure (GIP) by Seismic Capacity Engineers. Per the SQUG GIP, the seismic capacity for batteries on racks (Equipment Class 15) is based on the 1.5 x the GIP Bounding Spectrum. As shown in the figure below, there is large margin in the low frequency range and in higher frequencies above about 16 Hz.

Serial Number 11-566E Docket Nos. 50-338/339 Attachment Page 3 of 27 Service Building, Elevation 294' (5% DBE) compared to 1.5 x SQUG Bounding Spectrum 10.2 08 1.5 xSQUG Bou~nding Spectrum~ SB294 (5% DOE) - East-West 0.294 (5% DBE) - North-South 0 -58294 (5% DOE)- Vert 044-- 020 0.1 1.0 10.0 100.0 Frequency (W.) The anchorage for the main station battery racks has been evaluated by analysis. The racks in Rooms 1-I, 1-Il, and 1-Ill were evaluated by a Stone & Webster (S&W) calculation. That analysis concludes that the anchorage in Rooms 2-1, 2-11, and 2-111 are the same as those in the Unit 1 battery rooms. By SQUG walkdown, it was concluded that the anchorage in Room 1-IV is similar to 1-Il. Therefore, 1-IV is considered acceptable by comparison. A Stone & Webster calculation qualified the anchorage of the two vendor supplied battery racks installed in Room 2-IV. The rack anchorages were qualified using the conservative loads calculated from the Flight Dynamics International analysis. Based on the conservative loads applied for the analysis of the anchorage, which correspond to the higher than design basis spectral accelerations, the main station battery racks remain seismically qualified for the August 23, 2011 earthquake.

Further, during the resolution of Individual Plant Examination of External Events (IPEEE), it was concluded that, based on a comparison of peak spectral accelerations from IPEEE in-structure response spectra for the 294' elevation of the Service Building [for 0.3g Review Level Earthquake (RLE)], the anchorage was considered adequate since the the analysis was performed to enveloping levels.

Therefore, since the anchorage analysis met the IPEEE RLE anchored at 0.3g peak ground acceleration, the rack anchorage is acceptable for the August 23, 2011 earthquake. Emergency Diesel Generator (EDG) Batteries The EDG batteries are located in the EDG rooms, which are at the 275' elevation of the Service Building. The batteries have been qualified by testing. The racks are mounted on the floor. The racks and anchorage were originally qualified by a Stone & Webster

Serial Number 11-566E Docket Nos. 50-338/339 Attachment Page 4 of 27 (S&W) calculation, which has been supplemented by a Dominion calculation to account for replacement model batteries. The S&W analysis used conservative bounding loads for the batteries that enveloped the weights of the replacement batteries. A review of the analysis documented in the S&W calculation indicates that significant conservatisms exist. First, the racks are evaluated for maximum spectral accelerations between 33 Hz and 100 Hz, which are somewhat higher than the zero period accelerations (ZPAs). The frequency of the rack is tabulated to be well into the rigid range (93 Hz); therefore, use of the ZPA would have been appropriate. The analysis very conservatively used maximum peak spectral values between 33 Hz and 100 Hz from the 0.5% OBE and 1% DBE response spectra curves. Consistent with North Anna's licensing basis, the use of 3% damping would have been acceptable. Further, the analysis applied a factor of 1.5 to account for the effect of multi-modal response. Per North Anna's UFSAR, this factor need only be 1.3. However, no multi-mode factor was needed since the first mode frequency is well into the rigid range. Considering the conservatisms in the analysis, there is sufficient margin to conclude that the seismic qualification of the rack is acceptable for the August 23, 2011 earthquake. The above conclusion is corroborated by post-earthquake inspections that did not identify any damage to the batteries, the racks, or the anchorage.

2. EMCB Question 17 clarifications - There appears to be a discrepancy between the UFSAR differential settlement numbers and the numbers included in the response to the question.

Dominion Response In Dominion's initial response to this question, it was reported that up through March 2011, the differential settlement was as high as 0.36 inches, and that post-seismic differential settlement was measured as high as 0.53 inches. These measurements are correct. The Technical Requirements Manual (TRM) Limit for differential settlement is 0.22 feet (2.64 inches). These measurements are correct. The differential settlement discussed in Section 3.8.4.5.4.5 of the UFSAR is the total differential settlement before the service water pipe expansion joints were replaced by a design change in late 2002. When the old expansion joints were removed there was an upward movement of the service water piping to the north of the expansion joints where settlement points SM-17R and SM-18R are located. Survey readings of points SM-17R and SM-18R were performed before and after expansion joint replacement. The upward movement associated with the expansion joint replacement reduced the total differential settlement to a value less than reported in February 2002.

3. EMCB Question 24 clarifications - Expansion anchors - Since two supports were identified with loose anchors/bolts, perform an extent of condition investigation by checking the anchor bolt torque on a sample of supports.

Serial Number 11-566E Docket Nos. 50-338/339 Attachment Page 4 of 27 (S&W) calculation, which has been supplemented by a Dominion calculation to account for replacement model batteries. The S&W analysis used conservative bounding loads for the batteries that enveloped the weights of the replacement batteries. A review of the analysis documented in the S&W calculation indicates that significant conservatisms exist. First, the racks are evaluated for maximum spectral accelerations between 33 Hz and 100 Hz, which are somewhat higher than the zero period accelerations (ZPAs). The frequency of the rack is tabulated to be well into the rigid range (93 Hz); therefore, use of the ZPA would have been appropriate. The analysis very conservatively used maximum peak spectral values between 33 Hz and 100 Hz from the 0.5% OBE and 1% DBE response spectra curves. Consistent with North Anna's licensing basis, the use of 3% damping would have been acceptable. Further, the analysis applied a factor of 1.5 to account for the effect of multi-modal response. Per North Anna's UFSAR, this factor need only be 1.3. However, no multi-mode factor was needed since the first mode frequency is well into the rigid range. Considering the conservatisms in the analysis, there is sufficient margin to conclude that the seismic qualification of the rack is acceptable for the August 23, 2011 earthquake. The above conclusion is corroborated by post-earthquake inspections that did not identify any damage to the batteries, the racks, or the anchorage.

2. EMCB Question 17 clarifications - There appears to be a discrepancy between the UFSAR differential settlement numbers and the numbers included in the response to the question.

Dominion Response In Dominion's initial response to this question, it was reported that up through March 2011, the differential settlement was as high as 0.36 inches, and that post-seismic differential settlement was measured as high as 0.53 inches. These measurements are correct. The Technical Requirements Manual (TRM) Limit for differential settlement is 0.22 feet (2.64 inches). These measurements are correct. The differential settlement discussed in Section 3.8.4.5.4.5 of the UFSAR is the total differential settlement before the service water pipe expansion joints were replaced by a design change in late 2002. When the old expansion joints were removed there was an upward movement of the service water piping to the north of the expansion joints where settlement points SM-17R and SM-18R are located. Survey readings of points SM-17R and SM-18R were performed before and after expansion joint replacement. The upward movement associated with the expansion joint replacement reduced the total differential settlement to a value less than reported in February 2002.

3. EMCB Question 24 clarifications - Expansion anchors - Since two supports were identified with loose anchors/bolts, perform an extent of condition investigation by checking the anchor bolt torque on a sample of supports.

Serial Number 11-566E Docket Nos. 50-338/339 Attachment Page 5 of 27 Dominion Response Three hundred and sixteen (316) anchor bolts on fifty-one (51) supports were torque checked in the Unit 2 Safeguards Building, Auxiliary Building, and Unit 2 Containment. The selected anchor bolts ranged from 1/2" to 1-1/4" in size, and were randomly selected. Of the 316 anchor bolts torque tested all but five (5) passed the test. The five (5) that did not pass were wrench tight, were re-torqued, which confirmed proper grip, and maintained full load carrying capability. The five (5) anchor bolts that did not meet the torque checks were in five (5) different supports. However, the remaining bolts in each support passed the torque check, and the affected support remained tight against the wall, indicating that the five (5) wrench tight bolts were not caused by the August 23, 2011 earthquake. In no case were any supports rendered inoperable. Based on the low number of cycles of strong motion from the August 23, 2011 earthquake, extensive system inspections, and the tightness sampling performed, there is no concern for vibratory damage to expansion anchors.

4. EMCB 1.- What type of inspection was performed on the residual heat removal (RHR) heat exchanger integral support feet?

Dominion Response The examinations of the RHR heat exchanger integral support feet were performed as an ASME Code VT-3 examination and no earthquake damage was identified. The following clarifications were requested in an October 26, 2011 conference call.

1. EMCB Piping 4 - clarification

a. Discuss your approach for the selection of seismically qualified piping and supports designated for the walkdown inspections and please provide a summary of your inspection findings. If your walkdowns did not include all seismically qualified piping and supports, provide a list of piping and supports that were inspected, type of inspection, summary of findings, corrective actions in place and schedule of completion of the corrective actions. For seismically qualified piping and supports that were excluded from this list, provide a justification for exclusion and the consequences if these piping sections or their supports have been damaged. Your summary of inspection findings for piping and pipe supports should contain the documentation depicted in EPRI NP-6695 Section 5.3.5 and Table 5-1 item 4 (piping) sub-items 1 through 6; and item 10 (buried piping) sub-items I through 3.

b. Please address question 4a (above) to separately discuss ASME class I piping and pipe supports.

Serial Number 11-566E Docket Nos. 50-338/339 . Attachment Page 6 of 27

2. Provide percentages of ASME Class 1, 2 and 3 piping that was/was not inspected and discuss how we concluded that the piping was OK.

3. Provide detail regarding the inaccessible piping that was not inspected (e.g.,

approximately how much piping wasn't inspected, in what areas, etc.) and how we concluded that the piping was OK. Dominion Response EMCB piping Question 1 (4a and b), 2 and 3 Comprehensive inspections of both non-safety related and safety related plant piping systems were performed for each system in Unit 1 and Unit 2. This encompassed over eighty (80) systems for Unit 1 (which includes common systems) and over fifty (50) systems for Unit 2. These inspections were performed in accordance with station procedure 0-GEP-30, "Post Seismic Event System Engineering Walkdown," which was developed using the guidance provided in EPRI NP-6695, "Guidelines for Nuclear Plant Response to an Earthquake." While not specifically identified in the inspection procedure, piping system inspections encompassed pipe welds, nozzles, flanges, attachment lugs, couplings, etc. In addition, inspection procedure 0-GEP-30 includes the following specific guidance for performing piping inspections: " Check for snubber damage; i.e., snubbers pulled loose from foundation bolts, leakage of hydraulic fluid and bent piston rods Check for damage at rigid supports; i.e., deformation of support structure, deformation of pipe due to impact to support structure " Check for damage of expansion joints Check for damage or leakage of piping and branch lines " Check for damage to pipe at building joints and interfaces between buildings Inspection results were documented in procedure inspection logs, and discrepancies were entered into the Corrective Action System. The inspections were performed by qualified engineering personnel trained on identifying seismic related damage. The inspections did not identify any physical or functional damage to the piping systems or supports that would render them incapable of performing their intended functions as a result of the August 23, 2011 earthquake. There are no outstanding corrective actions related to earthquake damage for seismically qualified piping and supports. Inaccessible and insulated portions of piping systems were dispositioned based on inspections of associated system components that sustained no damage attributable to the August 23, 2011 earthquake and/or other piping in the same building or structure with similar supports that was inspected with satisfactory results. A summary of areas of ASME Class 1, 2, and 3 piping and supports that were not inspected is discussed below: Class 1 RCS piping that passes through annulus wall (approximately 4 ft/line)

Serial Number 11-566E Docket Nos. 50-338/339 Attachment Page 7 of 27

• Pressurizer spray nozzle and the associated piping inside the Pressurizer Steam Generator (S/G) tubes for 1 B and 1C and 2B S/Gs, (one half of all S/G tubes)

Dominion estimates that <1% of Class 1 piping (not including S/G tubes) was not inspected. Class 2 Buried portion of Reactor Purification (RP) System (approximately 20 ft) Buried portion of Safety Injection (SI) System (approximately 35 ft) Buried portion of Quench Spray (QS) system (approximately 250 ft) Charging (CH) System piping in the Auxiliary Building under lead shielding in overhead (approximately 200 ft - note that the seismic supports for the lead shielding were inspected) Dominion estimates that < 1% of the Class 2 piping was not inspected. Class 3 Buried Casing Cooling System piping (approximately 360 ft) Service Water (SW) sleeved piping to the Main Control Room Air Chillers (approximately 320 ft) Buried SW Piping (approximately 6,650 ft) Underwater SW piping in the SW reservoir SW piping that could be physically inspected without excavation was included. SW tie-in vaults, expansion joint vaults, and the vault beneath the Turbine Building were all entered for inspection. The SW spray arrays have been inspected from the shore of the reservoir with binoculars. In addition, spray array performance is unchanged from its performance prior to the earthquake. Dominion estimated < 1% of the non-buried Class 3 piping was not inspected. The piping/supports that were not inspected were determined to be acceptable as follows: Piping and supports that were inspected, both safety and npn-safety related, had no significant physical or functional earthquake damage. " Areas where piping entered/exited inaccessible locations were inspected (e.g., pipes traversing between buildings, entering the ground or tunnels, etc). Approximately 100 ft of safety related buried piping was uncovered by excavation to allow inspection with no earthquake damage indicated.

Serial Number 11-566E Docket Nos. 50-338/339 Attachment Page 8 of 27 No damage related to the earthquake was identified during piping and support inspections of Class 1, 2, and 3 seismically qualified piping. Dominion letters dated October 10 and 18, 2011 (Serial Nos. 11-577 and 11-577A, respectively) provide an extensive discussion regarding buried piping inspections and testing.

4. Discuss the post-restart (i.e., long-term) actions in place (i.e., those included within the corrective action program) to evaluate seismic qualified piping and pipe supports, including those SSCs discussed on pages 4 and 5 of the September 17, 2011, submittal, to demonstrate compliance with the licensing and design basis and/or applicable design basis code criteria/requirements, when these evaluations consider seismic loading, (OBE and DBE) derived from the August 23, 2011, seismic event, combinedwith all other applicable load cases. Discuss why Appendix F piping re-analysis is not required in accordance with Part 9900 since the DBE was exceeded.

Dominion Response Per the NRC Inspection Manual, a non-conforming condition is defined as: 3.6 Nonconforming Condition: A nonconforming condition is a condition of an SSC that involves a failure to meet the current licensing basis (CLB) or a situation in which quality has been reduced because of factors such as improper design, testing, construction, or modification. The following are examples of nonconforming conditions:

a. An SSC fails to conform to one or more applicable codes or standards (e.g., the CFR, operating license, TSs, UFSAR and or licensing commitments.

b. An as-built or as-modified SSC does not meet the CLB.

c. Operating experience or engineering reviews identify a design inadequacy.

d. Documentation required by the NRC requirements such as 10 CFR 50.49 is unavailable or deficient.

North Anna Power station structures, systems, and components (SSCs) are not considered to be in a non-conforming condition, as defined by NRC Inspection Manual Part 9900. The August 23, 2011 earthquake was a one-time seismic event that exceeded design basis earthquake accelerations in certain frequency bands for very short effective strong motion durations. No safety related SSCs were rendered non-functional and no functional damage was identified during extensive inspection and testing activities conducted in accordance with EPRI NP-6695. Therefore, plant SSCs are presently conforming to the applicable Licensing Basis requirements. Where the Current Licensing Basis is defined in the NRC Inspection Manual as:

Serial Number 11-566E Docket Nos. 50-338/339 Attachment Page 9 of 27 3.1 Current Licensing Basis: The CLB is the set of NRC requirements applicable to a specific

plant, plus a licensee's docketed and currently effective written commitments for ensuring compliance with, and operation within, applicable NRC requirements and the plant-specific design basis, including all modifications and additions to such commitments over the life of the facility operating license.

The set of NRC requirements applicable to a specific plant CLB include:

a. NRC regulations in 10 CFR Parts 2, 19, 21, 26, 30, 40, 50, 51, 54, 55, 70, 72, 73 and 100 and appendices thereto

b. Commission orders

c. License conditions

d. Exemptions

e. technical specifications

f. plant-specific design basis information defined in 10 CFR 50.2 and documented in the most recent UFSAR (as required by 10 CFR 50.71)

g. license commitments remaining in effect that were made in docketed licensing correspondence (such as licensee response to NRC bulletins, Licensee Events Reports, generic letters and enforcement actions)

h. licensee commitment documented in NRC safety evaluations Accordingly, the August 23, 2011 earthquake is being evaluated as one event using the DBE criteria and is being tracked in the corrective action system. Postulating a new event for one-half of this recorded earthquake for purposes of re-evaluating OBE conditions is not warranted.

Additionally, there is no requirement to reanalyze these components to demonstrate compliance with design code criteria, as Dominion does not intend to revise North Anna's current licensing basis as a result of the August 23, 2011 earthquake. EPRI NP-6695, which the NRC has endorsed in RG 1.167, "Long Term Plant Evaluations," states: "The acceptance criteria do not require compliance with allowable stress criteria normally used for design because the applied load is known, the equipment is available for inspection and evaluation and, therefore, structural margins need not be as high as in an original design." While Section 6.3.4.1 of the EPRI guidance does provide that emergency and faulted condition allowable stresses be used for comparison, the intent is to identify areas for increased/augmented inspection. Note that per the guidance, faulted condition allowable limits (Level D) may even be exceeded, and the equipment may still be acceptable provided the recommended augmented inspections and functionality reviews (as clarified per RG 1.167) are completed. In the limited evaluations that have been performed, as described in response to EMCB RAI 1 on Piping in Dominion letter dated October 18, 2011 (Serial No. 11-577A), we have not identified any specific case where stresses equivalent to the Level C Service Limit have been exceeded. The purpose of the EPRI NP-6695 long term evaluations is to provide additional assurance that the plant can continue to operate safely and to ensure plant readiness for another earthquake. Therefore, the post-startup evaluations are valuable and may

Serial Number 11-566E Docket Nos. 50-338/339 Attachment Page 10 of 27 help identify areas for increased scrutiny that should be inspected at the next available opportunity. Dominion is fully confident that North Anna is completely safe for return to operation based on the extensive and comprehensive short term inspections, analyses, and functional tests that have been completed to date.

5. EMCB Piping RAI-5 NAPS Submittal 11-520 states that several long term (post-startup) seismic evaluations and analyses will be performed for North Anna per EPRI NP-6695.

a) The following statement is from EPRI NP-6695, Section 6.3.4.1-1, as corrected by RG 1.167. If the calculated stresses from the actual seismic loading conditions are less than the allowable for emergency conditions (e.g., ASME Code Level C Service Limits or equivalent) or original design bases, the item is considered acceptable, provided the results of inspections and tests (Section 5.3.2) show no damage. Similarly, EPRI NP-6695, Section 6.3.4.1-2 requests comparison of the calculated stresses from the actual seismic loading conditions to the allowable for the emergency and faulted conditions. NAPS UFSAR Table 5.1-13, "LOAD COMBINATIONS AND OPERATING CONDITIONS," shows that the original design basis contains only Normal, Upset and Faulted conditions and does not contain an emergency condition. The above statements (from EPRI NP-6695, Sections 6.3.4. 1-1I and 6.3.4.1-2) in combination with the UFSAR imply that the calculated stresses from the actual seismic loading conditions should be compared to design basis conditions with the appropriate cases (in absence of the emergency condition) being the upset and faulted conditions. Please verify that the calculated stresses from the actual seismic loading conditions shall be compared to the upset and faulted condition allowable values for the purposes of EPRI NP-6695, Sections 6.3.4.1 analyses and evaluations. If not, discuss which alternate allowable values the calculated stresses from the actual seismic loading conditions will be compared against and provide a technical justification for the deviation from the EPRI NP-6695 Report. b) Please discuss your approach in reanalyzing all ASME Code Class I piping for pipe stresses, fatigue usage and pipe support requalification (EPRI requirement as corrected by RG 1.167) and provide the schedule of completion. c) Discuss your approach in reanalyzing seismically qualified non-class I (ASME) piping and pipe supports. If sampling is utilized, please discuss the criteria and methods of sampling that will ensure structural integrity of the remainder population of piping and pipe supports.

Serial Number 11-566E Docket Nos. 50-338/339 Attachment Page 11 of 27 d) Discuss the criteria and methods of sampling for the selection of piping and pipe supports for post-startup analyses that will ensure adequacy of structural integrity for the remainder population of piping and pipe supports. Provide the list of the selected piping and pipe supports to be analyzed and the schedule of completion. Dominion Response Regarding Part (a) of the question, as stated previously in the answer to Question 4 above, there is no requirement to change the plant Licensing Basis. Therefore, Dominion is not required to reanalyze plant SSCs for the experienced earthquake for purposes of demonstrating compliance with design code acceptance criteria. EPRI NP-6695, Sections 6.3.2 and 6.3.3 require seismic "re-evaluation" for the actual event on a sampling basis. Section 6.3.4 also clearly recognizes that acceptance criteria used for re-evaluations need not be in compliance with design codes. Therefore, Dominion will follow the guidance in EPRI NP-6695 and perform sampling analyses of piping systems using in-structure response spectra (ISRS) developed from the recorded time-histories of the August 23, 2011 earthquake. These ISRS will be developed at various elevations of the containment structure and other safety-related structures. The combined loading conditions for DBE, which include loads due to seismic and other phenomena, will be re-evaluated using the ISRS for the August 23, 2011 event and compared to the allowable limits equivalent to ASME Level C Service Limit. If it is concluded that the stress in the pipe and pipe supports remain within the limits of ASME Level C equivalent allowable stresses using ISRS from the August 23, 2011 earthquake, the component will be considered acceptable. If, by analysis, there are exceedances beyond the Level C allowable stresses, the guidance of NP-6695, Section 6.3.4.1 (2) or (3) will be followed to address continued acceptability. Regarding Part (b) of the question, a sampling analysis of Class 1 piping will be performed. Per EPRI NP-6695, Section 6.3.4.1, "Complete seismic reanalysis of all piping is not considered necessary. In general, piping reanalysis should be performed on a sampling basis to verify adequacy of piping and to assess the need for supplemental nondestructive examination of potential high strain areas." Fatigue for ASME Class 1 components will only be reevaluated if it determined that stresses exceed the ASME Level D Service Limit. It is expected that the development of ISRS for various structures and elevations for the August 23, 2011 earthquake will be completed in approximately one year and the re-evaluation of a sample population of Class 1 piping will be completed by April 30, 2013. Part (c) of the question requests justification of the sampling process to be used for non-Class 1 piping and supports. Part (d) of the question requests justification of the sampling process to be used for the remainder of the piping and supports. The sampling process will be in accordance with EPRI NP-6695, Section 6.3.3.

Serial Number 11-566E Docket Nos. 50-338/339 Attachment Page 12 of 27 Again, Dominion is fully confident that North Anna is completely safe for return to operation based on the extensive and comprehensive short term inspections, analyses, and functional tests that have been completed to date. The earthquake had limited damage potential, which is consistent with the low Cumulative Absolute Velocity (CAV) and short effective strong motion duration. No damage to any safety related SSCs was identified based on extensive inspections in accordance with EPRI NP-6695 and functional testing. Therefore, complete reanalysis of piping, pipe supports, or other SSCs is not required for the post startup evaluation. Additionally, there is no requirement to reanalyze these components to demonstrate compliance with design code criteria, as Dominion does not intend to revise North Anna's current design basis earthquake as a result of the August 23, 2011 seismic event. Instead, plant SSCs, including piping and pipe supports, will be re-evaluated for the actual August 23, 2011 earthquake on a sampling basis and for margin management for the long term evaluation program to comply with RG 1.167 and EPRI NP-6695.

6. Piping Question I clarifications Provide additional information regarding the systems sampled, pipe sizes, and configuration. For those few instances where we looked at the combined loads in the few locations and allowable stresses were exceeded, explain why the sampling analysis enveloped DBE and OBE.

Dominion Response Dominion's response to EMCB RAI 1 on Piping in Dominion's letter dated October 18, 2011 (Serial No. 11-577A), identified six piping systems segments that were chosen to conduct a scoping evaluation of the effects of the August 23, 2011 earthquake on analyzed piping. The six piping analysis models were selected on the basis of certain attributes, as identified in that response. The ranges of various attributes such as, pipe size, class, piping layout covering several elevations, operating temperature, and design pressure are listed in the Table below. The models also covered many different types of supports (e.g., snubbers, spring hangers, rigid supports and restraints) besides equipment anchors. It should be noted that these sampling analyses were performed for a scoping assessment purpose only, to assist in planning the long-term evaluations to meet the requirements of Regulatory Guide 1.167 and EPRI NP-6695. The following table lists the systems and relative system parameters.

Serial Number 11-566E Docket Nos. 50-338/339 Attachment Page 13 of 27 Nominal ASMand Designv Ma~x. Operating System Systemn~ Pipe Min Prsue Tm.' Class Prstr enerature: ~ Description siz Elevation '.~ 284.83' to Steam Generator "1A" to Containment Main Steam 32' Class 2 1085 psi 545°FPetrio#7 329.18' Penetration #73 277.83' to Steam Generator "IC" to Containment Feedwater 16" Class 2 3.6' 1100 psi 440°FPetrio#7 303.67' Penetration #77 Portion of Component Cooling Piping from Containment Penetration #26 Component 8" Class 2, 3 225.67' to 125 psi 116'F through IC RCP CC Return Inside Cooling 257.25' Isolation Valve to Anchor Restraint in Containment Portion of Component Cooling Piping Component 230.67' to 125 psi 145°F, 35°F from Excess Letdown Heat Exchanger/ Cooling 3 6" Class 3 247' Neutron Shield Tank Coolers to 18" CC Line in Containment Steam 226' to Steam Generator 1IA" Blowdown Line to Generator 2", 3" Class 2, 3 255.25' 1085 psi 521'F SteamnGent Beowdown Lnt Blowdown Containment Penetration #39 29" pipe @ Reactor 27.5", 614°F Coolant 2 Class 1 256' to 325' 2485 psi 2 1" Reactor Coolant System, Loop "B" Loop 29", 31" 27.5" & 31" @ 547°F A comparison of stresses due to design basis seismic spectra and the recorded spectra of the August 23, 2011 earthquake was made. Based upon the comparison, it is concluded that the stress in the pipe and pipe support remained within ASME Level C equivalent allowable stresses during the August 23, 2011 earthquake. This is consistent with EPRI NP-6695, Section 6.3.4.1-1, as corrected by RG 1.167: "If the calculated stresses from the actual seismic loading conditions are less than the allowable for emergency conditions (e.g., ASME Code Level C Service Limits or equivalent) or original design bases, the items are considered acceptable, provided results of inspections and tests show no damage."

7. Provide a list of the plant modifications that would be performed prior to developing the in-structure response spectra for the EPRI-NP-6695 evaluations.

Dominion Response During a teleconference with NRC on October 26, 2011, relating to the proposed Seismic Margin Management (SMM) program, the NRC asked for a list of design changes that would be implemented prior to having a fully developed program (i.e., informed by location-specific in-structure response spectra (ISRS), which would take approximately one year to complete). The following tables list the design changes that have been implemented for Unit 2 after the August 23, 2011 earthquake, and the design

Serial Number 11-566E Docket Nos. 50-338/339 Attachment Page 14 of 27 changes that will be implemented during the spring 2012 refueling outage for Unit 1 that have the possibility to be impacted by the SMM program. These design changes will be reviewed against the SMM program. Modifications Implemented During 2011 Unit 2 Refueling Outage Design Change Description NA-09-00181 Replacement of 2-QS-MOV-202A/B Valve, Actuator and Motor (JOG Modification) NA-09-00184 Replacement of 2-CH-MOV-2289A and 2-CH-MOV-2289B (JOG Modification) NA-10-00101 EDG Day Tank - Stainless Steel - Equivalency Evaluation NA-1 0-00112 Charging Pump Recirculation MOV and Manual Valve Addition Modification (JOG Modification) NA-10-00121 CRDM Housing Canopy Seal Weld Leakage - Mechanical Clamp Repair NA-10-00123 Copes-Vulcan Model D100 To D1 000 Conversion NA-10-00126 Replacement of 2-SI-MOV-2867A, 2-SI-MOV-2869A and 2-SI-MOV-2869B With New 3" 1500 lb. Velan Valves, and SMB-0 Actuators With Larger Motors (JOG Modification) NA-10-00139 JOG MOV Motor and Gear Ratio Modification on 2-CH-MOV-2381 and Gear Ratio Modifications on 2-SI-MOV-2890A/B/C/D and 2-CH-MOV-2380 (JOG Modification) NA-10-00161 Motor Upgrades For Motor Operated Valves Joint Owner's Group Improvements (JOG Modification) Modifications Planned for 2012 Unit I RFO Design Change Description NA-11-planned Seismic Monitoring Equipment Upgrade NA-11 -planned Correct Line 8"-SI-40-153-Q2 in Contact With Support 1 -RH-R-27.9 NA-1 0-00111 1-CH-MOV-1275A/B/C Reverse Flow and Manual Isolation Addition (JOG Modification) NA-10-00155 1 -SI-MOV-1 867A, 1867B, 1869A, 1869B Valve/Actuator/Motor (JOG Modification) NA-1 1-01025 1-RS-MOV-156A Yoke Reinforcement for Seismic Concern (JOG Modification) NA-11-01047 SG Nozzle Alloy 600 Mitigation Strategy (S/G Hot Leg Nozzle Structural Overlay) DCP 04-166 Install Vent Valves on SW lines to RS HX (8" Pipes; A Headers 1&4) DCP 06-118 1 J/1 J1/1 B3 4160/480V Transformer Replacement Modifications implemented during the Unit 2 refueling outage are being reviewed during the modification closeout process to add a seismic margin evaluation. If it can be judged, based on existing analyses, that the modified system or component has sufficient margin to accommodate the August 23, 2011 earthquake, then the basis for the determination is being included in the design change package documentation. If, however, such judgment cannot be supported without further analysis, a corrective action system entry will be made to perform the evaluation, once the necessary ISRS is available. Likewise, for the modifications that will be implemented during the Unit 1 refueling outage, an evaluation will be completed before the closeout of the design change package, or alternatively a corrective action system entry will be made to ensure the evaluation is tracked to completion.

Serial Number 11-566E Docket Nos. 50-338/339 Attachment Page 15 of 27 The following clarification was requested in an October 31, 2011 conference call. During an October 31, 2001 conference call to discuss the post-restart (i.e., long-term) Seismic Margin Management (SMM) program and how it will be implemented to evaluate plant modifications, the NRC requested clarification of the damping that would be used for our seismic evaluations of plant modifications implemented or planned since August 23, 2011 earthquake. Dominion Response On page 5 of the letter dated October 28, 2011 (Serial No. 11-577B), Dominion stated that 5% of critical damping would be used when performing margin assessments for piping. On further consideration, we will use either ASME Code Case N411 damping value with associated restrictions, or will use 4% damping as provided in RG 1.61 Revision 1 for SSE. These damping values will be used in the design change process for margin assessments when analyzing piping systems for in-structure response spectra (ISRS) developed from the recorded motions of August 23, 2011 earthquake.

8. EMCB Question l Ithe NRC requested the result of the NDE activities planned for North Anna Unit 2 refueling outage.

Dominion Response In letter dated October 10, 2011 (Serial No. 11-577), Dominion stated that the inservice inspection (ISI) examinations were completed for Unit 2 and the results were being evaluated and documented. The following provides the scope and results of the Unit 2 ISI examinations. There were approximately two hundred and fifty (250) examinations completed during the Unit 2 outage. Approximately 150 supports and 70 welds were examined by the following NDE techniques: ultrasonic testing (UT), liquid penetrant (PT), and magnetic particle (MT). These examinations were performed on safety related system (e.g., RCS, SI, Charging, RHR, QS, Component Cooling, Feedwater, etc.). There were no relevant indications identified that were attributable to the August 23, 2011 earthquake. In addition, VT-2 leakage system tests will be performed on applicable systems when the plant reaches normal operating temperature and pressure. The following clarifications were requested in an October 26, 2011 E-mail. EMCB Seismic Clarifications

1. Regarding Dominion's October 18, 2011 response to EMCB Question No. 32, explain the relevance between the statement in the response that begins with the

Serial Number 11-566E Docket Nos. 50-338/339 Attachment Page 16 of 27 second sentence, "it was therefore.., in each of three directions" and question No. 32. Dominion Response The following sentence in our previous response to EMCB question No. 32 was inadvertently omitted from our letter: "The Containment basemat is founded on a granite type hard rock." The two sentences "It was therefore... in each of three directions" are relevant in the context of question No. 32 to demonstrate that the response at the top of the containment basemat is essentially the same as the hard rock under it and there is essentially no soil-structure interaction that would change the seismic response at the top of the basemat. Thus, Dominion concluded that seismic instrumentation located on hard rock (free-field) would have yielded similar time-histories and cumulative absolute velocity (CAV) values as the actual Kinemetrics recorders time-histories, which were located on the Unit 1 Containment basemat and used to calculate the CAV value for the August 23, 2011 earthquake.

2. In the same response, with regards to the incoherency reduction of about 15%,

discuss where this number comes from and further with regards to your statement "to be at most 10% lower than, please explain the source for these numbers. Dominion Response The 15% reduction in In-Structure Response Spectra (ISRS) due to incoherency is a typical, average estimate for frequencies above about 8 HZ and is based on engineering judgment. The primary basis of this judgment is a review of several ISRS curves developed for North Anna Unit 3 for site-specific ground motion response spectra (GMRS) that show comparisons of coherent and incoherent responses. These curves are part of the North Anna 3 combined operating license application (COLA), Revision 3, dated June 2010, which has been submitted to the NRC. Similar reductions have been shown due to incoherency in some of the work done by EPRI. It is noted that the incoherency effect depends on the frequency content of input motion (greater reductions when ground motion spectrum has high frequency content) and the size of the basemat. The statement that the cumulative absolute velocity (CAV) values from a free-field recorder would have been at most 10% higher than the CAV from the recorded time-histories at the containment basemat is also based on engineering judgment. As discussed above, it is assumed that the average reduction in ISRS due to incoherency could be about 15% at higher frequencies. A large part of contribution to CAV would obviously come from area under those segments of the acceleration time-history curve that contain lower frequencies. Area under the higher frequency segments would be small and would therefore give a smaller contribution to CAV. Since there is almost no impact of incoherence in spectral frequencies below about 8 HZ, it is judged that a 10%

Serial Number 11-566E Docket Nos. 50-338/339 Attachment Page 17 of 27 reduction in CAV from the entire time-history (low frequencies and high frequencies together) represents a reasonable estimate. In this context, it is noted that the CAV values calculated from the August 23, 2011 earthquake were below the OBE exceedance threshold of 0.16 g-sec for two orientations and exceeded this threshold by about 8% in the third direction. There was essentially no damage to systems, structures, and components (SSCs), which is consistent with the calculated values of CAV. Even if the projected free field CAV values from the August 23, 2011 earthquake were somewhat higher than those calculated by Dominion, no significant damage to the station would be expected and this would have no impact on the restart readiness of North Anna Power Station.

3. Please explain the difference between the time histories demonstrated in Figures 3.7.9 and 10 in the FSAR and the DBE time histories presented in the Attachment 2 of the October 18, 2011 response.

Dominion Response The time-histories in Figures 3.7.9 and 3.7.10 of the UFSAR were provided by Stone and Webster to describe the solution to a test problem using the STRUDL II computer code, as noted on page 3.7-21 of the North Anna UFSAR. Those two time-histories are the response time-histories at the operating floor level of the containment and are presented in the UFSAR Section 3.7.2.7 under the heading "Validation of Computer Programs". The design basis earthquake (DBE) time histories in Attachment 2 of the October 18, 2011 response are synthetic time-histories that were developed from the DBE ground response spectra for rock-founded structures and were used to perform dynamic analysis of the Containment structure. to develop DBE in-structure response spectra at various elevations. The response spectrum developed from each of the three synthetic time-histories closely match the corresponding DBE ground response spectrum shape, as shown in Attachment 2 of the October 18, 2011 letter (Serial No. 11-577A). The followinq clarifications were requested in an October 26, 2011 E-mail.

1. Piping and NDE - Long Term Question In Dominion letter dated October 18, 2011 (Serial No. 11-566B), page 17, the licensee provided clarification regarding the impact of the recent earthquake on the leak-before-break (LBB) analysis.

The licensee stated that it analyzed a representative reactor coolant loop with the seismic response spectra of the recent earthquake. The licensee found that the seismic loading on the RCS piping based on the recent earthquake is less than the seismic loading due to DBE in the existing LBB analysis. The licensee concluded that the existing LBB analysis remains valid. For the long term action, the staff requests the licensee to submit the analysis that

Serial Number 11-566E Docket Nos. 50-338/339 Attachment Page 18 of 27 demonstrates the existing LBB analysis remains valid by providing a comparison of calculated loads from the recent seismic event to those that were assumed in the existing LBB analysis. If any loads experienced during the seismic event exceed those assumed in the existing LBB analysis, provide an updated LBB analysis using the seismic loads experienced and submit the updated LBB analysis to the NRC. Dominion Response Dominion will update the evaluation performed, as referenced in the above-mentioned response, once the in-structure response spectra (ISRS) at appropriate locations are developed from the recorded time history at the basemat of Containment for the August 23, 2011 earthquake. A comparison of calculated loads from the August 23, 2011 earthquake to those that were assumed in the existing leak-before-break (LBB) analysis will be performed. The results of the comparison of the calculated load from the August 23, 2011 earthquake and the existing leak-before-break (LBB) analysis will be submitted by March 31, 2013. The following clarifications were requested in an October 27, 2011 E-mail EMCB

1. Please discuss your inspection of the reactor vessel supports and any findings of observed damage due to the recent earthquake. Also, discuss the criterion used to determine functionality and structural adequacy of the vessel supports.

Dominion Response A VT-3 inspection was performed by Level II Inspectors on the Unit 1 and Unit 2 Reactor Vessel sliding foot supports during the week of October 17, 2011. The scope of the VT-3 examinations was limited to the sliding foot assembly. In addition, the exposed portions of the Neutron Shield Tank pad and the integrally cast nozzle pad were observed to have no damage or other relevant indications. No evidence of any damage or distress to the supports was observed during the performance of these examinations for both Units. An evaluation of the sliding foot supports was performed. The methodology used in this evaluation was to determine a representative "scaling" ratio for the effect of the August 23, 2011 earthquake, relative to the existing Design Basis Earthquake (DBE) for North Anna. This scaling ratio was selected based on work previously performed by Engineering to determine a representative "scaling" ratio for the effect of the August 23, 2011 earthquake, relative to the existing Design Basis Earthquake (DBE) for North Anna. The functionality and structural integrity of the sliding foot supports was confirmed by evaluations performed by Engineering.

Serial Number 11-566E Docket Nos. 50-338/339 Attachment Page 19 of 27

2. Please discuss the analytical evaluation you plan to perform to confirm conformance with applicable design basis limits for various loading combinations related to the reactor vessel supports. In your discussion, also describe in detail how you obtained applicable seismic loads derived from the recent earthquake.

Dominion Response Reactor pressure vessel (RPV) nozzle support loadings are summarized below for the "net downward" and "net upward" cases respectively. Seismic loadings are increased by a factor of 1.21 to account for the August 23, 2011 earthquake. Previously analyzed loadings are also listed for comparison purposes. The evaluation shows that the seismic loads are bounded. Net Downward Loads Load Condition FH (kIs FV (ip-s Deadweight 284 Seismic (previous) 333 168 a = Seismic (scaled) 403 203 b = Pipe Rupture 1162 1599 Total = DWT + \\![T]' + [b,]2 1230 1896 Previously Analyzed Loads* 1290 1911 Net Upward Loads Load Condition F-kPS) Fy (kips) Deadweight -284 a = Seismic (scaled) 403 203 b = Pipe Rupture 1118 1103 Total = DWT + '[a] 2 + [.b 1188 837 Previously Analyzed Loads* 1238 925

• Includes Pipe Rupture Loads The following clarifications were requested in an October 27, 2011 E-mail License Renewal

1. Given that the August 2011 earthquake resulted in ground motions that exceeded that assumed for your safe shutdown earthquake, please discuss your plans to

Serial Number 11-566E Docket Nos. 50-338/339 Attachment Page 20 of 27 assess the impact of these larger ground motions on the time limited aging analyses (TLAAs) that include earthquake loadings as inputs (e.g., the TLAAs for metal fatigue of ASME Code Class I components, reactor vessel underclad cracking, reactor coolant pump flywheels, fatigue analyses for USAS/ANSI B31.1 components, piping containing subsurface flaws, containment liner plate, and spent fuel pool liner). If a re-analyses of any or all of these TLAAs is not planned, please justify why such re-analyses are not necessary. Discuss your plans for submitting these assessments/justifications to the NRC staff Dominion Response The earthquake that occurred on August 23, 2011 exceeded DBE spectral accelerations in certain frequencies in the North-South (N-S) and Vertical (V) directions. A quantitative comparison of the CAV was performed between the August 23, 2011. earthquake and the DBE. The quantitative comparison, which is provided in Attachment 2 to Dominion letter dated October 17, 2011 (Serial No. 11-577A) notes that, since the CAV values are low, and only the CAV in the North-South direction exceeded the OBE exceedance threshold of 0.16 g-sec by about 8%, no significant damage was expected. This was consistent with the findings from the comprehensive inspections of both safety and non-safety structures, systems and components (SSC) performed and the functional and surveillance tests that have been and are being conducted at North Anna Power Station. Therefore it is concluded that the August 23, 2011 earthquake and the much smaller aftershocks have negligible impact on the time-limited aging analyses (TLAAs) that include earthquake loadings as inputs. As such, the original TLAA conclusions remain unchanged by the earthquake (i.e., there are no changes to the disposition of any TLAA), and additional license renewal related augmented inspections are not required during the period of extended operation. The following additional information is provided to support this conclusion. The evaluations for the following TLAAs were addressed in the response to License Renewal Question 1, which was forwarded to the NRC in Dominion letter dated October 17, 2011 (Serial No. 11-577A): I " Metal Fatigue and the Transient cycle counting program Leak-Before-Break

• Containment liner plate The following additional information is provided for the other TLAAs mentioned in the above request:

Reactor Coolant Pump Flywheel North Anna Safety Evaluation Report (SER) Section 4.7.2 and UFSAR Section 18.3.5.2 address the TLAA on the Reactor Coolant Pump Flywheel. During normal operation, the flywheel possesses sufficient kinetic energy to produce high-energy missiles in the unlikely event of failure. The aging effect of concern is fatigue crack initiation in the

Serial Number 11-566E Docket Nos. 50-338/339 Attachment Page 21 of 27 flywheel bore keyway. This TLAA demonstrated that the flywheel design has a high structural reliability with a very high flaw tolerance and negligible flaw crack growth over a 60-year service life. The flywheel assembly, including any speed-limiting and anti-rotation devices, the shaft, and the bearings, are designed to withstand normal condition, anticipated transients, the design basis loss-of-coolant accident, and the design basis earthquake loads without loss of structural integrity. However, since the recent August 23, 2011 earthquake resulted in the measured exceedance of DBE spectral accelerations in some frequencies in the N-S and V directions, the impact on the flywheel has been considered. Because flywheel overspeed is the critical loading for the TLAA, and the earthquake occurred with the flywheels operating at normal speeds, the event resulted in a less limiting loading on the flywheel than that considered in the TLAA. In addition, no flaws are known to exist in the RCP flywheels and they will continue to be inspected in accordance with our ISI program. Spent Fuel Pool Liner SER Section 4.7.4 and UFSAR Section 18.3.5.4 address the TLAA on the spent fuel pool liner, which is needed to prevent a leak to the environment. A design calculation has been identified which documents that the spent fuel pool design meets the general industry criteria. The calculation includes a thermal fatigue analysis to add a further degree of confidence. The normal thermal cycles occur at each refueling, resulting in 80 cycles for both units in 60 years. Total number of thermal cycles is expected to be 90, which includes normal, upset, emergency, and faulted conditions. The calculations show that the allowable thermal cycles for the spent fuel pool liner for the most severe thermal condition, which includes a loss of cooling, is one hundred (100). One of the faulted conditions considered for evaluation of the spent fuel pool liner is the DBE. Therefore, since the recent August 23, 2011 earthquake resulted in the measured exceedance of DBE spectral accelerations in some frequencies in the N-S and vertical directions, the impact on the spent fuel pool liner was considered. Since the liner has been evaluated for abnormal temperatures up to 170 OF in conjunction with a DBE, and the earthquake occurred with the spent fuel pool at a Service Load temperature of approximately 90 OF, the event resulted in a less limiting loading on the liner than that considered in the TLAA. As previously stated in the response to EMCB Question 10 (which was forwarded to the NRC in Dominion letter dated October 17, 2011 (Serial No. 11-577A,) "The load demand on the spent fuel pool at the time of the August 23, 2011 earthquake, did not exceed the previously designed load combination of DBE with Abnormal Load temperature." The response to that question provided results of inspections and evaluations which have demonstrated the structural adequacy of the spent fuel pool liner following the earthquake. Therefore, the TLAA conclusions remain valid for the period of extended operation.

Serial Number 11-566E Docket Nos. 50-338/339 Attachment Page 22 of 27 Reactor Vessel Underclad Cracking SER Section 4.3 and UFSAR Section 18.3.2.2 address the TLAA pertaining to Reactor Vessel Underclad Cracking. In early 1971, an anomaly was identified in the heat-affected zone of the base metal in a European-manufactured reactor vessel. A generic fracture mechanics evaluation by Westinghouse demonstrated that the growth of underclad cracks during a 40-year plant life would be insignificant. The evaluation was extended to 60 years using fracture mechanics evaluation based on a representative set of design transients. The occurrences were extrapolated to cover 60 years of service life. This 60-year evaluation shows insignificant growth of the underclad cracks. The plant-specific design transients are bounded by the representative set used in the evaluation. The analyses associated with reactor vessel underclad crack growth had been determined to be acceptable through the period of extended operation. However, since the evaluations include seismic inputs, and the August 23, 2011 earthquake resulted in the measured exceedance of DBE spectral accelerations in certain frequencies in the N-S and vertical directions, the impact on the reactor vessel underclad calculations has been considered. For reasons delineated in the first paragraph of this response to this question, it has been concluded that the August 23, 2011 earthquake and the much smaller aftershocks have negligible impact on the time-limited aging analysis (TLAAs) that include earthquake loadings as inputs. Piping Containing Subsurface Indications SER Section 4.7.5 and UFSAR Section 18.3.5.5 address the TLAA pertaining to piping containing subsurface flaws. Calculations have been identified that address piping subsurface indications detected by inspections performed in accordance with ASME Section XI. Section XI provides the acceptance criteria for various flaw orientations, locations and sizes. The calculations determined the number of thermal cycles required for the flaws to reach unacceptable size. Since the number of the cycles experienced by the piping will not exceed those values for sixty years of operation, the analyses were determined to remain valid for the period of extended operation. Since the calculations include seismic inputs, and the recent August 23, 2011 earthquake resulted in the measured exceedance of DBE spectral accelerations in some frequencies in the N-S and vertical directions, the impact on the calculations for piping subsurface indications was considered. For reasons delineated in the first paragraph of this response, it has been concluded that the August 23, 2011 earthquake and the much smaller aftershocks have negligible impact on the time-limited aging analysis (TLAAs) that include earthquake loadings as inputs. In addition, as discussed in Dominion letter dated October 10, 2011, three locations were selected for examination that had pre-existing flaws. These locations were: 1) the Unit 2 Outside Recirculation Spray (RS) suction piping (surface indication), 2) the Unit 2 pressurizer circumferential weld (embedded flaw), and 3) the Unit 2 'B' S/G girth weld

Serial Number 11-566E Docket Nos. 50-338/339 Attachment Page 23 of 27 (embedded flaw). Testing results confirmed that the two embedded flaws were essentially unchanged as a result of the August 23, 2011 earthquake, and the RS surface indication remained non-recordable. These results are directly applicable to Unit 1 as the overall loading environment would have been the same during the earthquake. Therefore, the TLAA conclusions remain valid for the period of extended operation. Conclusion There are no additional evaluations required as a result of the earthquake.

2. Consistent with the operating experience attribute of an aging management program, discuss your plans to address the following and to submit this information to the NRC:

a. Reactor Vessel Internals Program The seismic event of August 2011 exceeded the plant's design basis. The licensee's current reactor vessel internals (RVI) inspection program for the period of extended operation relies on inspection results from RVI components at reactors other than the North Anna units. Given that the August 2011 seismic event exceeded the original design basis, it is not clear that the results of inspections performed of the RVI at another site would be representative of the condition at the North Anna site.

Request: Discuss any planned reactor vessel internals inspection activities for each North Anna unit considering the seismic event of August 2011. Provide a comparison of the activities that Dominion will implement for the reactor vessel internal components during the period of extended operation to those recommended inspections and evaluations for these components in Material Reliability Program (MRP) 227. Justify the basis for any noted deviations or implementation differences from the MRP's recommended inspection and flaw evaluation recommendations. Dominion Response The implementation activities associated with License Renewal Commitment 14 are ongoing for North Anna. As noted, the Surry Unit 1 reactor was selected for the focused one time inspection. The Surry Unit 1 baffle bolts were inspected in Fall 2010 and the focused inspection will be completed in the Spring 2012 prior to the period of extended operation (PEO). Under the commitment, these inspections will form the basis for fulfilling the commitment to evaluate the results and determine any additional inspections required for the Surry and North Anna reactors. Implementation of the additional inspections, per commitment, will be performed during the PEO of Surry Unit 1 and would not likely be performed at North Anna until the timeframe of the PEO.

Serial Number 11-566E Docket Nos. 50-338/339 Attachment Page 24 of 27 The inspections of Surry Unit 1 and evaluation of the inspection results has not yet been performed so the extent of additional inspections cannot be stated at this time. In addition to the activities performed in compliance with the license renewal commitment, Dominion has taken the necessary steps to comply with the Mandatory and Needed requirements of MRP-227 in fulfillment of its responsibilities under NEI 03-08, "Guideline for the Management of Materials Issues." North Anna has programs and plans in place to implement the inspections required by MRP-227. To date, there has been no deviation to MRP-227 Revision 0 requirements at Surry and the recent inspections of the North Anna Unit 2 vessel internals, documented in letters dated October 10 and 18, 2011(Serial Nos. 11-566A and 11-566B), did not identify the need for deviations to that document. As discussed in letter dated September 27, 2011, (Serial No. 11-520A), Dominion is developing a plan with the NSSS vendor consisting of additional evaluations or inspections, as warranted, to assure long term reliability of the reactor internals.

b.

Tank Inspection Activities As described in Section 5.1.1.1(b), there is data to demonstrate that the recent seismic event may have little to no impact on mechanical and electrical systems; however, aboveground tanks are typically not designed with nozzle connections that are as robust as those on a pressure vessel. In addition, the piping that is attached to the nozzle typically has a much shorter run prior to entering the soil than the first anchor downstream or upstream of a pressure vessel. Therefore, these nozzles can be more sensitive to displacements. The staff lacks sufficient information to conclude that these nozzles do not require augmented inspections to assure such displacements did not induce stresses that could exacerbate the effects of aging. For instance, the displacements could have resulted in localized deformation that would subsequently be more likely to corrode due to the increased residual stresses. This postulated deformation may be evident through cracked coatings on the external surface of the tank. For uncoated tanks, the deformation may not be as obvious. In addition, the staff lacks sufficient information to determine if the tanks in the scope of license renewal (and are thus subject to aging management) have internal coatings. If they do have internal coatings, the coating may have been damaged and thus the damage site may become more susceptible to internal corrosion. Request: As a result of the above: " describe the results of aboveground atmospheric storage tank nozzle inspections,

• state whether any in-scope aboveground atmospheric storage tanks have internal coatings,

Serial Number 11-566E Docket Nos. 50-338/339 Attachment Page 25 of 27 if localized nozzle inspections of atmospheric storage tanks have not occurred, state the basis for concluding that there was no impact on the pressure retaining function of these tanks that would necessitate augmented inspections during the period of extended operation and " if internal tank coating inspections have not occurred, state the basis for concluding that there was no impact on tank internal coatings, if applicable, such that augmented inspections during the period of extended operation are not required. Dominion Response Above ground atmospheric storage tanks were included as part of our system inspections performed for plant systems, both safety and non-safety related. These inspections were performed in accordance with station procedure 0-GEP-30, Post Seismic Event System Engineering Walkdown, which specifically identifies inspection attributes for piping and low pressure storage tanks. No nozzle/tank deformation resulting from the earthquake was identified for any of the above ground atmospheric storage tanks and nozzles. No cracking of external coatings was identified related to the earthquake. Additionally, inspection of these tanks are included as part of normal operator rounds and system engineering walkdowns. The Emergency Condensate Storage Tanks and the normal Condensate Storage Tanks have an internal coating. The condition of the Condensate Storage Tanks' external coating is a good indicator of the condition of the interior coatings. Immersion-service coatings are typically designed to have a larger adhesion strength value. Therefore, it is expected that there was no earthquake related cracking of the interior coating. Additionally, regular inspections are performed of the internal coatings as part of our aging management program for inaccessible areas. Submersible cameras are used to perform the inspections. The following is a clarifications to RAIs requested in an NRC Conference call October 28, 2011 call re-gardin-g Reactor Vessel Internals e) Please provide justification which demonstrates that the use of the seismic loads in the current (existing) analyses of record provides reasonable assurance that adequate margin exists between the elastic stress limits and the loads which were induced in the RVIs during the August 23, 2011, seismic event felt at NAPS. Dominion Response The results summarized in Table 1 (provided on page 19 of 20 in the attachment to our letter of October 27, 2011 (Serial No. 11-566D) indicate that the selected interface loads calculated for either the OBE or DBE have a considerable amount of margin compared to the qualified load limits associated with the OBE, with the exception of the lower radial keys. The qualified OBE load limit corresponds to no yield in the components.

Serial Number 11-566E Docket Nos. 50-338/339 Attachment Page 26 of 27 The calculated loads listed for the vessel upper support plate and core support ledge may be doubled and still not exceed the OBE load limits. For the radial keys, a supplemental stress analysis has been performed using calculated loads from a reactor vessel seismic analysis of the August 23, 2011 event. Table 2 presents the maximum primary stresses in the lower radial key from this supplemental analysis for the analyzed load associated with the August 23, 2011 event. Note that the stresses reported are the maximum for any path through the key and these maxima occur along different paths. Therefore, the sum of maximum membrane and maximum bending stresses does not equal the maximum membrane plus bending. Paths are defined along each element division tangent to the core barrel and paths of maximum stress are shown in Figure 1 below. Figure 1: Lower Radial Key ANSYS Model Maxknum Membrane Plus Bending 21,320 psi / -q Maximum Membrane Stress 12,320 psi Maximum Bending Stress 14,000 psi Table 2: Lower Radial Key Primary Stresses Primary Membrane Primary Bending Pm + Pb Stress, Pm (psi) Stress, Pb (psi) (psi) 12,320 14,000 21,320

Serial Number 11-566E Docket Nos. 50-338/339 Attachment Page 27 of 27 At 6000 F, the key material's allowable membrane stress (Sm) is 16,600 psi. For Level A and B conditions (no yield), the margin for primary membrane stress is: Sm 16,600 psi MSmem =1.347 Pm 12,320 psi The margin for primary membrane-plus-bending stress is: 1.5*Sm 24,900 psi MSmem+bend =----------- =---- = 1.168 Pm + Pb 21,320 psi These results demonstrate that 16% margin exists to the no yield stress limit for the lower radial key with the loading resulting from the August 23, 2011 earthquake.}}