ML13135A637

Confirmatory Action Letter, Notification of Commitment Action Completion

ML13135A637
Person / Time
Site:	North Anna
Issue date:	05/13/2013
From:	Grecheck E Virginia Electric & Power Co (VEPCO)
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML13135A637 (22)

Text

VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261 May 13, 2013 U.S. Nuclear Regulatory Commission Serial No.: 13-143 Attention: Document Control Desk NL&OS/ETS RO Washington, DC 20555 Docket Nos.: 50-338 50-339 License Nos.: NPF-4 NPF-7 VIRGINIA ELECTRIC AND POWER COMPANY (DOMINION)

NORTH ANNA POWER STATION UNITS 1 AND 2 CONFIRMATORY ACTION LETTER NOTIFICATON OF COMMITMENT ACTION COMPLETION By letter dated November 7, 2011 (Serial No. 11-520D), Dominion submitted a compilation of long-term actions to be performed in addressing the August 23, 2011 earthquake.

These long-term actions were reviewed and confirmed by the NRC in a Confirmatory Action Letter (CAL) dated November 11, 2011. As discussed with the NRC staff after the CAL was issued, Dominion agreed to provide periodic status updates for the long-term actions identified in the CAL. Update letters were submitted on May 8, 2012 and January 23, 2013 to provide status of the CAL long-term action commitments. to this letter provides an updated status of each long-term action commitment and Attachment 2 includes summary information regarding the completion of CAL Items 1, 5, 7, 8, and 9.

With the completion of these remaining items, all of the ten (10) long-term action commitments have now been addressed.

If you have any questions or require additional information, please contact Mr. Thomas Shaub at (804) 273-2763.

Sincerely, E. S. Grecheck Vice President - Nuclear Engineering and Development Commitments made in this letter: None

Attachment:

1. Status for Long-Term Action Commitments

2. Summary Information from CAL Action Items 11 ý5- '5`7

Serial Number 13-143 Docket Nos. 50-338/339 Page 2 of 2 cc: Regional Administrator 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 Ms. M. Khanna Branch Chief - Division of Operating Reactor Licensing U. S. Nuclear Regulatory Commission One White Flint North Mail Stop 08 B1A 11555 Rockville Pike Rockville, MD 20852-2738 Mr. R. J. Pascarelli Branch Chief - Division of Operating Reactor Licensing U. S. Nuclear Regulatory Commission One White Flint North Mail Stop 09 D3 11555 Rockville Pike Rockville, MD 20852-2738 Dr. V. Sreenivas 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.

Glen Allen, Virginia 23060

Serial No.13-143 Docket Nos. 50-338/339 Attachment 1 Status of Long-Term Action Commitments Virginia Electric and Power Company (Dominion)

North Anna Power Station Units 1 and 2

Serial No.13-143 Docket Nos. 50-338/339 Attachment 1 Page 1 of 4 Long-Term Action Commitment List North Anna Power Station Units I and 2 Item Commitment Scheduled No. Completion Date 1 Dominion will perform long-term evaluations of plant structures, systems and components in accordance with RG 1.167/EPRI NP-6695, Section 6.3. Any anomalies identified during the evaluations will be entered into the corrective action system and evaluated for extent of condition.

Status: Complete April 30, 2013 In-structure Response Spectra have been developed from the recorded time (Complete) histories of the August 23, 2011 earthquake for the site structures housing safety related SSCs and the evaluations of the SSCs have been completed and a summary included in the UFSAR (see Item 3.b). A summary of the results of the evaluation for structures,systems, and components (SSC) is included in Attachment 2 of this letter.

2 Dominion will develop a plan to characterize the seismic source and any special ground motion effects due to the relative locations of the fault and the site and will update the NRC accordingly.

Status: The action item for plan development has been completed. (Ongoing activities to implement the plan will be completed as indicatedbelow.)

The Dominion plan to characterizethe seismic source and any special ground motion effects due to the relative locations of the fault and site is presented below.

The earthquake catalog containedwithin the Central and Eastern US Seismic Source Characterization(CEUS-SSC) model and databasehas been updated to include the last three years of seismicity data (through mid-December 2011). This update includes the M5.8 earthquake in Louisa County Virginia. The updated earthquake catalog was issued March 8, 2012.

Dominion has obtained data from the USGS and the Virginia Seismologic Observatory at Virginia Tech on the main shock and aftershocks resulting from the earthquake. The aftershocks define a planarfeature in the subsurface that March 31, 2012 appears to be a previously unknown fault. (Complete)

1. Dominion is conducting a geologic reconnaissanceof the relatively linear projection of the aftershock hypocenters to the ground surface and other features of significance that may be identified through review of LiDAR imagery obtained for a defined area of interest and field work. The objective of the LiDAR imagery review and field work is to document the presence or absence of any evidence for surface faulting due to the earthquake. The LiDAR survey review was completed during December2011 and February 2012, and data evaluation is in progress. The field reconnaissancewas guided by using maps that combine geologic data with LiDAR imagery. The field reconnaissancewas completed on April 23, 2012. The final report was completed in May 2012.

2. Separately, the USGS has indicatedplans to perform a geophysical.survey for the August 23, 2011 event during the spring of 2012. Dominion will review the data obtained from that survey and incorporateany new findings into our overall evaluation of the event.

3. The data collected from LiDAR imagery, field reconnaissanceand geologic mapping review did not provide evidence of surface fault expression the August 23, 2011 event might have caused, or longer term, largerscale deformation related to seismic activity. Though readily observable geological deformation evidence was not found during the field reconnaissance,seismic

Serial No.13-143 Docket Nos. 50-338/339 Attachment 1 Page 2 of 4 Long-Term Action Commitment List North Anna Power Station Units I and 2 Item Commitment Scheduled No. Completion Date geophysical data obtained from the numerous aftershocks revealed that the August 23, 2011 event likely originatedalong a previously unmapped fault estimated to be approximately 6.2 miles in length with a N 280 to 300 E strike and 450 to 510 SE dip angle. Additional investigative work by the USGS has not been completed and will be reviewed when it becomes available.

3 Dominion will revise the North Anna Updated Final Safety Analysis Report to document.

a. The recorded August 23, 2011 seismic event.

Status: Complete April 30, 2012 Section 2.7 of the UFSAR has been revised to include a description of the (Complete)

August 23, 2011 seismic event.

b. A summary of the evaluation and results of the seismic analysis of the recorded event completed per RG 1.167/EPRI NP-6695.

Status: Complete April 30, 2013 Section 3.7.8 of the UFSAR has been created to include a description the (Complete) evaluationsof plant structures,systems and components in accordance with RG 1. 167/EPRI NP-6695, Section 6.3. A summary of the assessmentis included in Attachment 2 of this letter.

c. Design controls on seismic margin management (i.e., the Seismic Margin Management Plan). April 30, 2012 Status: Complete April ete)

Section 3.7 of the UFSAR has been revised to include a description of the (Complete)

Seismic Margin Management Plan.

d. Incorporation of Regulatory Guide (RG) 1.166, "Pre-Earthquake Planning and Immediate Nuclear Power Plant Operator Post-Earthquake Actions," and RG 1.167, "Restart of a Nuclear Power Plant Shut Down by a Seismic Event," December 31, into the North Anna licensing basis. 2012 Status: Complete (Complete)

Section 3.7.4 of the UFSAR has been revised to include Dominion's commitment to RG 1.166 and RG 1.167.

4 Dominion will implement a design change to replace the existing seismic equipment and Main Control Room indication with upgraded and enhanced seismic monitoring instrumentation equipment, which includes installation of a permanent, free field seismic monitor. Permanent, seismically qualified backup power to a new Seismic Monitoring Panel will also be installed including a backup battery power supply. The seismic instrumentation will be installed and maintained in accordance December 31, with RG 1.12, Rev 2, Nuclear Power Plant Instrumentation for Earthquakes. The 2012 project will also install seismic recording instrumentation at the station Independent (Complete)

Spent Fuel Storage Installation (ISFSI) pad.

Status: Complete New seismic instruments and associated controls/indicationwere installed in the Unit 1 containmentand main control room during the spring 2012 refueling outage

Serial No.13-143 Docket Nos. 50-338/339 Attachment 1 Page 3 of 4 Long-Term Action Commitment List North Anna Power Station Units 1 and 2 Item Commitment Scheduled No. Completion Date and made functional. The remainderof the instruments (auxiliary building, ISFSI and free field) were installed and consideredfunctional in November 2012.

5 Dominion will perform a re-evaluation of the plant equipment identified in the IPEEE review with HCLPF capacity <0.3g, which will include an assessment of potential improvements. March 31, 2013 Status: Complete (Complete)

The re-revaluationis completed. A summary of the assessment is included in Attachment 2 of this letter.

6 Dominion will develop a plan with the NSSS vendor consisting of additional evaluations or inspections, as warranted, to assure long term reliability of the reactor internals.

Status: Complete Dominion Engineeringperformed a systematic review of the previous evaluations, NRC correspondence, and inspections of the reactor vessel internals (RVI) that were performed following the August 23, 2011 earthquake. Data and rationale for future plannedprogrammaticinspections were considered and included input from February 29, 2012 Westinghouse, the NSSS vendor. Results of the prior activities were summarized (Complete) and evaluated regardingtheirdemonstration of RVI integrity. As a result of the evaluation, the RVI remain capable of reliably performing their function in the future. Furthermore,RVI inspections performed during the spring 2012 Unit I refueling outage did not identify any seismically induced issues. Therefore, it was concluded that developing a plan consisting of additional inspections and evaluations of the RVI beyond those already requiredby existing programs is not warrantedfor assuring the long term reliabilityof the RVI.

7 Dominion will perform a comparison of the calculated load from the August 23, 2011 earthquake and the existing leak-before-break (LBB) analysis and submit the results. March 31, 2013 Status: Complete (Complete)

The comparison shows that the existing licensing basis LBB analysis remains valid.

A summary of the load comparisonis included in Attachment 2 to this letter.

8 Dominion will perform inspections at North Anna Power Station in accordance with In accordance the latest MRP-227 revision approved by the NRC. with the MRP-227 Status: Complete inspection plan, which will be The inspection plan has been finalized and is included in Attachment 2 to this letter. provided by March 31, 2013 (Complete) 9 Dominion will re-evaluate the Time-Limiting Aging Analyses (TLAAs) that include seismic inputs to either: 1) quantitatively demonstrate that the TLAAs are still bounding, or 2) re-analyze the TLAAs, based on the August 23, 2011 earthquake. March 31, 2013 Status: Complete Comparisonsindicate that the conclusions of the seismic-induced fatigue related TLAAs of components reviewed for the License Renewal Programremain valid. A summary of the results is included in Attachment 2 to this letter.

Serial No.13-143 Docket Nos. 50-338/339 Attachment 1 Page 4 of 4 Long-Term Action Commitment List North Anna Power Station Units I and 2 Item Commitment Scheduled No. Completion Date 10 Dominion will implement a long term Seismic Margin Management Plan to address the impact of the August 23, 2011 earthquake. Specifically, to ensure adequate seismic margins are maintained for plant SSCs, Dominion will revise the design change process for North Anna Power Station to require explicit evaluation of plant modifications for the effects of the August 23, 2011 earthquake using In-Structure Response Spectra (ISRS) for the event for the Containment, Auxiliary Building, and other buildings containing safety related SSCs developed based on actual time-histories recorded during the event. In support of future plant design changes, the evaluation of plant SSCs will require design verification and code compliance with December 31, the stresses, loads, accelerations, and displacements generated from the analysis 2011 with the design basis ISRS and the analysis with the ISRS for the August 23, 2011 (Complete) earthquake, whichever are higher.

Status: Complete The NuclearDesign Control Programhas been revised to incorporate controls that will ensure that seismic margin is maintained at North Anna Power Station for plant modifications performed after the August 23, 2011 earthquake.

Dominion will provide 30-days prior written notification to the NRC if it determines the Seismic Margin Management Plan is no longer required based on the implementation of comparable actions as part of the resolution of the NRC's N/A recommendations emanating from the Fukushima Daiichi Task Force and/or Generic Issue (GI)-199.

The shaded items were previously reported in letters dated May 8, 2012 (Serial 12-235) and January 23, 2013 (Serial No.13-007).

Serial No.13-143 Docket Nos. 50-338/339 Attachment 2 Summary Information for Long-Term Action Commitments (Items 1, 5, 7, 8, and 9)

Virginia Electric and Power Company (Dominion)

North Anna Power Station Units I and 2

Serial No.13-143 Docket Nos. 50-338/339 Attachment 2 Page 1 of 14 CAL Item 1 - Long Term Evaluation of Plant Structures, Systems and Components (SCCs)

Discussion The long term evaluations, performed in accordance with RG 1.167 / EPRI NP-6695, consisted of the following steps:

" Calculation of seismic loads, i.e., structural loads and in-structure response spectra (ISRS) derived from the recorded time-histories of the August 23, 2011 M5.8 earthquake

" Comparison of seismic loads of the August 23, 2011 event with design basis seismic loads

, Seismic re-evaluation of representative structures, ASME Class 1 piping systems and equipment where the calculated loads may have exceeded the design basis loads ISRS for each of the safety-related structures were developed using the time-histories recorded at the containment basemat (elevation 216') as the starting point.

Soil-structure interaction (SSI) analyses using the best estimate shear wave velocity profiles were performed for the soil founded structures and fixed base analyses were performed for the rock-founded structures. For each structure and floor elevation, ISRS were compared to the existing ISRS from the Design Basis Earthquake (DBE). The August 23, 2011 spectra exceeded the Operating Basis Earthquake (OBE) and/or DBE spectra at most locations in certain isolated frequency bandwidths.

Selected safety-related structures, representative of the structure types that exist at North Anna, including reinforced concrete and structural steel structures, were analyzed to determine potential areas of high load demand. Methods used in these evaluations included direct comparison of generated building response spectra, equivalent static coefficient analysis, as well as the response spectrum method. The analyses showed that the calculated forces and moments at various locations in these structures were well below their allowable capacities based on ultimate strength design. No area of abnormally high load demand or anomaly was identified.

As required per EPRI NP-6695 guidelines, representative ASME Class 1 piping systems were selected for re-evaluation using the spectra from the August 23, 2011 earthquake.

The basis for selecting piping systems included systems with high design stress and where the spectral acceleration was exceeded at contributing natural frequencies. Of a total of about 120 Class 1 piping models, 13 representative models (or about 10% of the total number of piping models) were selected for re-evaluation. These piping models were analyzed for design pressure, dead weight, and the applicable response spectra and building displacements due to the August 23, 2011 earthquake. For comparison purposes, the OBE and DBE results were extracted from the existing analysis of record.

The calculated stresses from the spectra corresponding to the August 23, 2011

Serial No.13-143 Docket Nos. 50-338/339 Attachment 2 Page 2 of 14 earthquake were significantly below the ASME Section III, Level C allowable. Thus, the results indicated that the piping components had adequate margin to withstand the August 23, 2011 earthquake and no gross deformation in any of these piping systems was expected. The resultant seismic moments in fatigue sensitive locations were tabulated and compared with the resultant seismic moments used in the existing fatigue evaluations. The comparison indicated that the August 23, 2011 earthquake had insignificant effects on component fatigue.

Loads on equipment nozzles and branch line nozzles were compared with the loads used in qualification of equipment nozzles and reactor coolant loop (RCL) branch nozzles. The August 23, 2011 earthquake did not affect the structural integrity of the nozzles. For the pipe supports where the seismic load during August 23, 2011 was higher than the seismic load used in the support design, other operating loads in combination with the loads from the August 23, 2011 earthquake were reviewed. The combined loads resulting from thermal expansion, deadweight, seismic inertia and seismic anchor movements on the pipe supports were compared with the loads used in the support design. The results showed that the structural integrity of the pipe supports was not affected by the August 23, 2011 earthquake.

Representative sample of safety-related equipment was selected for reevaluation using the ISRS developed for the August 23, 2011 earthquake. These included: (a) major equipment and supports in the Reactor Building, (b) mechanical and electrical equipment, and (c) tanks and heat exchangers. The selection of the sample of equipment items was based on: (a) items representative of various classes or types of equipment, (b) items in each of the safety-related structures, and (c) the low capacity items with a high-confidence-of-low-probability-of-failure (HCLPF) capacity less than 0.3g. Major equipment, such as the Reactor Pressure Vessel, Steam Generator, Reactor Coolant Pump and Pressurizer were reviewed by comparing the interface (nozzles, support attachments) loads. The loads from the August 23, 2011 earthquake were less than those from a design basis earthquake. No major equipment or their supports were identified with excessive stress or deformation requiring further evaluation, inspections or testing.

The sample population of electrical/mechanical equipment, and tanks and heat exchangers included equipment items qualified by analysis as well as via shake-table testing. An approximate 10% sample of the equipment was selected for evaluation using the ground spectra or ISRS, as appropriate, from the August 23, 2011 earthquake. For the sample of electrical equipment qualified by shake-tests, the ISRS calculated from the recorded motions of the August 23, 2011 earthquake were adequately enveloped by the corresponding site-specific or generic test response spectra. For equipment qualified by analysis, with two exceptions, the evaluation of the sample components including their anchorages showed that the stresses remained within the allowable values. The exceptions included two groups of low HCLPF capacity components - the 120V Vital AC bus cabinets and the refueling water storage tank (RWST) for both units. Using the response spectra derived from the August 23, 2011 earthquake, the shear-tension interaction of the anchorage of the vital bus

Serial No.13-143 Docket Nos. 50-338/339 Attachment 2 Page 3 of 14 cabinets was calculated to be greater than unity from a conservative static analysis with peak spectral accelerations, and the overturning moment of the RWST exceeded the capacity of the anchorage of this tank. These exceptions were entered into the corrective action system and the anchorages of these components were inspected. No deformation or anomalies were found in these inspections and it is concluded that the August 23, 2011 earthquake did not cause any damage to these components or their anchorages.

Conclusion These long-term evaluations corroborate the results of extensive plant inspections and functional tests that were performed in support of the plant restart effort where no physical or functional damage was observed in safety-related structures, systems and components.

Serial No.13-143 Docket Nos. 50-338/339 Attachment 2 Page 4 of 14 CAL Item 5 - Re-evaluation of SSCs with HCLPF < 0.3g Discussion During the IPEEE review for North Anna Unit 1 and 2, thirteen groups of structures, systems, and components (SSCs) were determined to have HCLPF capacities less than 0.3g. The HCLPF capacities for these components were analyzed using the EPRI seismic margin assessment (SMA) approach per EPRI NP-6041-SL for various applicable failure modes using in-structure response spectra (ISRS) developed from a Review Level Earthquake (RLE) with NUREG/CR-0098 median centered spectral shape anchored to 0.3g peak ground acceleration (PGA) to meet the requirements of NRC GL 88-20 Supplements 4 and 5, in accordance with NUREG 1407.

Following the August 23, 2011 M5.8 Mineral, VA earthquake, a thorough walkdown inspection was performed by seismic review teams (SRT), comprised of at least two Seismic Qualification User Group (SQUG) trained seismic capability engineers (SCEs),

for all the low HCLPF components except for the 4160V Emergency Busses, whose seismic capacity was governed by relay chatter which is not a walk down attribute.

Walkdown findings concluded that none of the components showed any sign of damage or anomaly due to the earthquake. Therefore, even though the IPEEE calculations indicated that these components have calculated HCLPF capacities < 0.3g, each component withstood the August 23, 2011 earthquake extremely well.

Conclusion Dominion reevaluated the HCLPF calculations for each of the components with a HCLPF < 0.3g. This review indicated that, in some cases, there was significant conservatism in the previous calculations; therefore, more realistic analyses were performed. Four groups of SSCs, representing 14 items of equipment (including the Emergency Condensate Storage Tanks, the 4160V Emergency Busses, the Reactor Trip Breaker cabinets, and the Component Cooling Pumps) are now shown to have a HCLPF capacity greater than 0.3g for the RLE. For other components, where immediate improvements to the HCLPF analyses could not be made, potential conservatisms in those analyses were noted, as well as any potential physical modifications that might improve seismic capacity. However, no additional re-analysis or physical modifications of these remaining components are planned at this time. The extent of this re-evaluation is considered to be a reasonable approach because: (a) the NUREG/CR-0098 based RLE used in the IPEEE evaluations is not site-specific; (b) risk contribution of these low HCLPF components to the seismic safety of the plant is not specifically known; some of the low HCLPF components identified during IPEEE are likely to be not risk-significant based on past seismic PRA experience; and (c) Dominion plans to review the need for potential improvements to these components in the near future as part of a seismic PRA using site-specific seismic hazard curves to address the NRC Fukushima Near-Term Task Force Recommendation 2.1.

Serial No.13-143 Docket Nos. 50-338/339 Attachment 2 Page 5 of 14 CAL Item 7 - Leak-Before-Break Comparison

Background

Consideration of leak-before-break (LBB) for the main Reactor Coolant Loop (RCL) for North Anna was analyzed and documented in Westinghouse Report WCAP-11163 (Ref.

1) including supplement 1 to WCAP-11163 (Ref. 2). Review of the loads and material properties of the RCL resulted in the identification of one load critical location and four other toughness critical weld locations for the LBB analysis. The welds at load critical location and toughness critical locations are identified in Figure 1. The layout of the RCL in Units 1 and 2 are essentially the same. The weld numbering in Figure 1 is applicable to all six loops (Three loops for Unit 1 and three loops for Unit 2).

Results of Comparison Dominion performed a comparison of the Calculated Load from the August 23, 2011 earthquake and loads used in the LBB analysis. Comparison of the loads is delineated in the following table.

Loads during 8123/11 Earthquake Loads Used in LBB Evaluation.

Weld # *::

esultnt: *R....ultant:

• ***W *: eld# Resultant

.(WCAP-: Axial Load Bending: Axial Load Rua 11163) . ... '.(Kips)

.. , IMomenti Momet (Kps)Bending (Ki(inKips) ::. Moment,

.... (in-K ps)f...

1 1849 23255 1871 26332 2 1821 22212 1839 23244 3 1727 19513 1728 20730 4 1623 7183 1607 10884 5 1432 6888 1616 12466 Remarks

1. Weld # 1 is a load critical location and Weld # 2, 3, 4, and 5 are toughness critical locations evaluated in WCAP-11163.

2. Normal loads remained the same as those evaluated earlier and are not listed here. The seismic loads are combined with the normal load by absolute summation.

3. Loads during the August 23, 2011 earthquake remained less than the loads used in LBB evaluation at all locations except the axial load at Weld # 4, which exceeded the previously evaluated axial load by about 1%. However, the resultant bending moment during the August 23, 2011 earthquake was only 7,183 in-kips compared to 10,884 in-kips used in the evaluation in WCAP-11163. Review of the original evaluation confirmed that the LBB analysis include at least a margin of 10 on leak rate, 2 on crack size and 1.4 on load.

Serial No.13-143 Docket Nos. 50-338/339 Attachment 2 Page 6 of 14 Figure-1 Schematic Diagram of Primary Loop Showing Weld Locations HOTLEG GENERATOR STEAM .

HOTrLEG Towerstufle:20.2*F; Pr5essre: 2.259 Psi CIOSSOVEG LEG Tepe.'ature:655.5*F;Pressure:2.221psi COLDME T..p.rat.re-sss.s5; Press- 2.30S9-1 Weld Location 1 is the load critical location in each loop of Unit 1 and 2. The Weld Locations 2, 3, 4, and 5 are toughness critical locations as listed below (Ref. 2, Page B-5):

Weld # 2 Unit 2 Loop 3 Weld # 4 Unit 1 Loop 1 & 2 Weld # 3 Unit 1 Loop 1 & 2 Weld # 5 Unit 1 Loop 2 Unit 2 Loop 1 & 2 The postulated leakage size flaws were evaluated for bounding loads. Since the normal operating load is not altered by the earthquake loading, the 10 gpm leakage size flaws determined in Ref. 1 & 2 remain the same.

Conclusion In no case did bending moment exceed assumed values in the original LLB evaluation.

References (1) Westinghouse Report WCAP-1 1163, "Technical Bases For Eliminating Large Primary Loop Pipe Rupture as a Structural Design Basis For North Anna Units 1 & 2, August 1986." (Westinghouse Proprietary Class 2 )

(2) Supplement 1 to Westinghouse Report WCAP-1 1163, "Additional Information in Support of the Technical Justification For Eliminating Large Primary Loop Pipe Rupture as the Structural Design Basis For North Anna Units 1 & 2," January 1988.

(Westinghouse Proprietary Class 2)

Serial No.13-143 Docket Nos. 50-338/339 Attachment 2 Page 7 of 14 CAL Item 8 - MRP-227 Inspection Plan CAL Item 8 commits to the performance of inspection in accordance with the latest material reliability program (MRP)-227 approved by the NRC within two refueling outages following entry into the period of extended operation. Consistent with the commitment of CAL Item 8, Dominion is providing for your information, the inspection plan for the reactor vessel internals, which conforms to MRP-227-A. Based on the results of the inspection of the North Anna Power Station (NAPS) Unit 2 reactor vessel internals performed post earthquake (fall of 2011), no additional inspections or evaluations beyond the planned MRP-227 inspection are deemed necessary. The plan identifies the scope and schedule of the NAPS examinations to meet the requirements of MRP-227-A. Unit 1 and 2 reactor internal components are in line with the Westinghouse design requirements of the material reliability program. Specific evaluations and any analyses required will be performed as needed to support the scheduled examinations and the applicable Licensee Action Items identified in the December 16, 2011 Safety Evaluation for MRP-227. At this time, there are no expected deviations from the recommendations of MRP-227-A, as approved by the NRC. The inspection scope is preliminary and may vary within the limits of the approved version of MRP-227 applicable at the time the examinations are required.

The inspection plan for the reactor vessel internals is included in the following table.

Serial No.13-143 Docket Nos. 50-338/319 Attachment 2 Page 8 of 14 MRP-227 Inspection Plan Primary Components 2 Required Examination Item 1 Examination Method Coverage and Acceptance Expansion Link 4 Schedule For Initial ExaminationsS5 Criteria_

Control Rod Guide Tube Visual examination (VT-3). 20% examination of the number of None Unit 1 - N1R25 (Fall 2016)

(CRGT) Assembly Examine for Loss of Material CRGT assemblies, with all guide Unit 2 - N2R26 (Spring 2019)

Guide plates (cards) (Wear). cards within each selected CRGT assembly examined.

N1.AUG 57.009 The specific relevant condition is N2.AUG 57.009 wear that could lead to loss of Each subsequent ISI interval in control rod alignment and impede accordance with MRP-227-A or most control assembly insertion, current revision.

Control Rod Guide Tube Enhanced visual examination 100% of outer (accessible) CRGT Bottom-mounted Unit 1 - N1 R25 (Fall 2016)

Assembly (EVT-1) for Cracking (SCC, lower flange weld surfaces and instrumentation (BMI) Unit 2 - N2R26 (Spring 2019)

Lower flange welds Fatigue) adjacent base metal on the column bodies, Aging Management (IE and individual periphery CRGT Lower support N1.AUG 57.010 TE). To determine the assemblies, column bodies (cast)

N2.AUG 57.010 presence of crack-like surface A minimum of 75% of the total Upper core plate flaws in flange welds. identified sample population must Lower support Each subsequent ISI interval in be examined, forging/casting accordance with MRP-227-A or most The specific relevant condition is a current revision.

detectable crack-like surface indication.

Core Barrel Assembly Enhanced visual examination 100% of one side of the Lower support Unit 1 - N1R25 (Fall 2016)

Upper core barrel flange (EVT-1) for Cracking (SCC) accessible surfaces of the column bodies (non Unit 2 - N2R26 (Spring 2019) weld The specific relevant condition selected weld and adjacent base cast) is a detectable crack-like metal. A minimum of 75% of the Core barrel outlet Each subsequent ISI interval in N1 .AUG 57.001 surface indication, total weld length. nozzle welds accordance with MRP-227-A or most N2.AUG 57.001 Cracking (SCC) current revision.

Core Barrel Assembly Enhanced visual examination 100% of one side of the Upper and lower core Unit 1 - N1R25 (Fall 2016)

Upper core barrel cylinder (EVT-1) for Cracking (SCC, accessible surfaces of the barrel cylinder axial Unit 2 - N2R26 (Spring 2019) girth weld IASCC, Fatigue) selected weld and adjacent base welds The specific relevant condition metal. A minimum of 75% of the Each subsequent ISI interval in N1.AUG 57.002 is a detectable crack-like total weld length. accordance with MRP-227-A or most N2.AUG 57.002 surface indication, current revision.

Serial No.13-143 Docket Nos. 50-338/33.9 Attachment 2 Page 9 of 14 Required Examination Item 1 Examination Method 2 Coverage and Acceptance 3 Expansion Link4 Schedule For Initial Examinations5 Criteria Core Barrel Assembly Enhanced visual examination 100% of one side of the Upper and lower core Unit 1 - N1R26 (Spring 2018)

Lower core barrel cylinder (EVT-1) for Cracking (SCC, accessible surfaces of the barrel cylinder axial Unit 2 - N2R27 (Fall 2020) girth weld IASCC, Fatigue) selected weld and adjacent base welds The specific relevant condition metal. A minimum of 75% of the Each subsequent ISI interval in N1.AUG 57.003 is a detectable crack-like total weld length. accordance with MRP-227-A or most N2.AUG 57.003 surface indication. current revision.

Core Barrel Assembly Enhanced visual examination 100% of one side of the None Unit 1 - N1R26 (Spring 2018)

Lower core barrel to (EVT-1) for Cracking (SCC, accessible surfaces of the Unit 2 - N2R27 (Fall 2020) support plate weld Fatigue) selected weld and adjacent base The specific relevant condition metal. A minimum of 75% of the Each subsequent ISI interval in N1.AUG 57.004 is a detectable crack-like total weld length. accordance with MRP-227-A or most N2.AUG 57.004 surface indication. current revision.

Baffle-Former Assembly Visual examination (VT-3) for Bolts and locking devices on high None Unit 1 - N1R25 (Fall 2016)

Baffle-edge bolts Cracking (IASCC, Fatigue) fluence seams. 100% of Unit 2 - N2R26 (Spring 2019) that results in: components accessible from core N1.AUG 57.005

• Lost or broken locking side N2.AUG 57.005 devices A minimum of 75% of the total

• Failed or missing bolts population.

• Protrusion of bolt heads Aging Management (IE and Each subsequent ISI interval in ISR) Void swelling is accordance with MRP-227-A or most managed through current revision.

management of void swelling on the entire baffle-former assembly.

Baffle-Former Assembly Volumetric examination (UT) 100% of accessible bolts. Heads Lower support Unit 1 - N1R25 (Fall 2016)

Baffle-former bolts for Cracking (IASCC, Fatigue) accessible from the core side. UT column bolts, Barrel- Unit 2 - N2R26 (Spring 2019)

Aging Management (IE and accessibility may be affected by former bolts N1.AUG 57.006 ISR) Void swelling is complexity of head and locking N2.AUG 57.006 managed through device designs.

management of void swelling A minimum of 75% of the total Each subsequent ISI interval in on the entire baffle-former population. accordance with MRP-227-A or most

_assembly. current revision.

Serial No.13-143 Docket Nos. 50-338/339 Attachment .2 Page 10 of 14 Required Examination Item1 Examination Method 2 Coverage and Acceptance 3 Expansion Link4 Schedule For Initial Examinations5 Criteria Baffle-Former Assembly Visual examination (VT-3) for Core side surface as indicated. None Unit 1 - N1 R25 (Fall 2016)

Assembly Distortion (Void Swelling), or Unit 2 - N2R26 (Spring 2019)

(Includes: Baffle plates, Cracking (IASCC) that results baffle edge bolts and in:

indirect effects of void

• Abnormal interaction with swelling in former plates) fuel assemblies

• Gaps along high fluence N1.AUG 57.007 baffle joint Each subsequent ISI interval in N2.AUG 57.007 e Vertical displacement of accordance with MRP-227-A or most baffle plates near high current revision.

fluence joint

• Broken or damaged edge bolt locking systems along high fluence baffle joint Alignment and Direct measurement of spring Measurements should be taken at None Unit 1 - N1R25 (Fall 2016)

Interfacing Components height. If the first set of several points around the Unit 2 - N2R26 (Spring 2019)

Internals hold down spring measurements is not sufficient circumference of the spring, with a to determine life, spring height statistically adequate number of N1.AUG 48.001 measurements must be taken measurements at each point to N2.AUG 48.001 during the next two outages, minimize uncertainty.

in order to extrapolate the expected spring height to 60 Each subsequent ISI interval in years. To identify Distortion accordance with MRP-227-A or most (Loss of Load) current revision.

Thermal Shield Visual examination (VT-3) for 100% of thermal shield flexures. None Unit 1 - N1R26 (Spring 2018)

Assembly Cracking (Fatigue) or Loss of Unit 2 - N2R27 (Fall 2020)

Thermal shield flexures Material (Wear) that results in thermal shield flexures Each subsequent ISI interval in N1.AUG 57.008 excessive wear, fracture, or accordance with MRP-227-A or most N2.AUG 57.008 complete separation current revision.

Serial No.13-143 Docket Nos. 50-338/33.9 Attachment 2 Page 11 of 14 Existing Programs Components

__________________2 AccuieptneCieidEpaso ExaminationsCvrg ik o Item1 Examination Method 2 Required Examination Coverage Expansion Link Schedule For Initial Examinationss Core Barrel Assembly Visual examination (VT-3) to All accessible surfaces at In accordance with Unit 1 - N1R26 (Spring 2018)

Core barrel flange determine general condition specified frequency. ASME Code Section Unit 2 - N2R27 (Fall 2020) for excessive wear. Xl, Category B-N-3 N1i.B13.70.001 ASME Code Section Xl, Category Each subsequent ISI interval in N2.B13.70.001 B-N-3 accordance with approved Section Xl program.

Upper Internals Visual examination (VT-3) All accessible surfaces at In accordance with Unit 1 - N1R26 (Spring 2018)

Assembly specified frequency. ASME Code Section Unit 2 - N2R27 (Fall 2020)

Upper support ring or skirt XI, Category B-N-3 ASME Code Section Xl, Category Each subsequent ISI interval in N1.B13.70.001 B-N-3 accordance with approved Section Xl N2.B313.70.001 program.

Lower Internals Visual (VT-3) examination of All accessible surfaces at In accordance with Unit 1 - N1R26 (Spring 2018)

Assembly the lower core plates to detect specified frequency. ASME Code Section Unit 2 - N2R27 (Fall 2020)

Lower core plate evidence of distortion and/or Xl, Category B-N-3 loss of bolt integrity. ASME Code Section Xl, Category Each subsequent ISI interval in N1i.B13.70.001 B-N-3 accordance with approved Section Xl N2.B13.70.001 program.

Bottom Mounted Surface examination (ET) Eddy current surface examination Condition Report Unit 1 - Each RFO*

Instrumentation System as defined in plant response to (CR) response Unit 2 - Each RFO*

Flux thimble tubes IEB 88-09.

1-PT-210.4 NUREG-1801, Rev. 1

• Scheduled in accordance with 2-PT-210.4 NAPS TRM 3.3.1 program requirements.

Alignment and Visual examination (VT-3) All accessible surfaces at In accordance with Unit 1 - N1R26 (Spring 2018)

Interfacing Components specified frequency. ASME Code Section Unit 2 - N2R27 (Fall 2020)

Clevis insert bolts XI, Category B-N-3 ASME Code Section Xl, Category Each subsequent ISI interval in N 1.B13.70.001 B-N-3 accordance with approved Section Xl N2.B13.70.001 program.

Serial No.13-143 Docket Nos. 50-338/33,9 Attachment 2 Page 12 of 14

_____Requir____ eptEamnceio Criteriage Exaso ink Item' Examination Method 2 Required Examination Coverage Expansion Link4 Schedule For Initial Examinations 5 Alignment and Visual examination (VT-3) All accessible surfaces at In accordance with Unit 1 - N1 R26 (Spring 2018)

Interfacing Components specified frequency. ASME Code Section Unit 2 - N2R27 (Fall 2020)

Upper core plate Xl, Category B-N-3 alignment pins ASME Code Section Xl, Category B-N-3 Each subsequent ISI interval in N1 .B13.70.001 accordance with approved Section XI N2.B133.70.001 program.

Guide Tube Support Visual examination (VT-3) All accessible surfaces at In accordance with Unit 1 - N1 R26 (Spring 2018)

Pins specified frequency. ASME Code Section Unit 2 - N2R27 (Fall 2020)

Split Pins Xl, Category B-N-3 ASME Code Section XI, Category Each subsequent ISI interval in N1 .B13.70.001 B-N-3 accordance with approved Section Xl N2.B133.70.001 program.

TABLE NOTES

1. The "Item" is the MRP-227-A description and the Inservice Inspection Schedule Summary Number used to schedule and track completion of that particular inspection item.

2. The "Examination Method" as identified in MRP-227-A and the Effect (Mechanism) the examination is to identify.

3. The "Coverage and Acceptance Criteria" as identified in MRP-227-A (Table 4-3) and the associated notes along with the Examination Acceptance Criteria identified in MRP-227-A (Table 5-3).

4. The "Expansion Links" as identified in MRP-227-A (Table 4-3).

5. The schedule is preliminary and may vary within the limits of the approved version of MRP-227 applicable at the time of examination. Subsequent examinations are in accordance with the current understanding of the NAPS aging management program and the current understanding of MRP-227-A, and are subject to future revision in accordance with the program.

Serial No.13-143 Docket Nos. 50-338/339 Attachment 2 Page 13 of 14 CAL Item 9 - Time-Limiting Aging Analyses Discussion Dominion quantitatively demonstrated that the Time Limited Aging Analyses (TLAAs) are still bounding, after the effects of the August 23, 2011 earthquake was considered. The only possible influence of an earthquake on a TLAA can be component fatigue in an extreme situation, where cyclic seismic induced stresses are a significant contributor to the component's usage factor. Therefore, attention was focused on seismic induced component fatigue. The ASME Class 1 components were reviewed by Westinghouse in October 2000 for TLAAs and the review is documented in a License Renewal Technical Report LR-1010/LR-2010 (Reference 2). Fatigue-related TLAAs for the components identified in Reference 2 were reviewed to identify significance of the contribution of seismic loading to determine the effects, if any, of the August 23, 2011 earthquake. The following components were reviewed in detail:

" Steam generator (SG),

" Pressurizer,

• Reactor Pressure Vessel (RPV),

" Certain Locations of Control Rod Drive Mechanism (CRDM),

" Reactor Coolant Pump (RCP), and

" Reactor Coolant Loop (RCL) Piping including the Branch Nozzles.

The results of the review showed that the August 23, 2011 earthquake had insignificant effect on component fatigue. The review was supplemented by analyses where required.

In addition, several representative reactor coolant branch lines, with components that have high predicted seismic stress and high usage factor were reanalyzed for the August 23, 2011 earthquake in order to determine the changes if any, to the fatigue usage considered in the TLAA of these piping components. In all, about 10% of the ASME Section'lll, Class 1 stress analysis models (13 Models out of 120) were reanalyzed. The results of the analyses showed that the level of combined stresses in these lines during the August 23, 2011 earthquake were less than the Code allowable stresses for the OBE condition. For this level of stress, NRC Reg. Guide 1.167 does not require an explicit fatigue reevaluation. Therefore, it can be concluded that the fatigue in the piping components due to the August 23, 2011 earthquake is not significant. As a result, the August 23, 2011 earthquake did not influence the TLAA in these components. However, in order to quantitatively verify the influence of the August 23, 2011 earthquake on TLAA, the resultant moments at fatigue sensitive locations due to this earthquake were compared with the resultant OBE moments used in the fatigue analysis. The comparison showed that at most of the locations resultant moments due to the August 23, 2011 earthquake were less than the resultant moments due to the previously analyzed OBE loading at fatigue sensitive locations. For a few cases, the

Serial No.13-143 Docket Nos. 50-338/339 Attachment 2 Page 14 of 14 resultant seismic moments due to the August 23, 2011 earthquake exceeded the resultant OBE moments used in the component qualification. Those locations were evaluated in detail and it was determined that the component fatigue was not significantly affected. Therefore, conclusions about the TLAAs were not altered.

Conclusion ASME Section Iii, Class 1 fatigue sensitive equipment and piping for North Anna Power Station were reviewed for the effect of the August 23, 2011 earthquake. The result of the review showed that the August 23, 2011 earthquake had insignificant effect on the component fatigue.

Since the only potential effect on the TLAA is seismically induced component fatigue, conclusions of the TLAAs prepared for the plant operating license extension remain valid after the August 23, 2011 earthquake.

References

1. NRC Letter CAL No. NRR-2011-002, "Confirmatory Action Letter Regarding North Anna Power Station Units 1 and 2, Long -Term Commitments to Address Exceeding Design Basis Seismic Event (TAC Nos. ME7254 and ME7255)," November 11, 2011.

2. Technical Report: LR-1010/LR-2010, "Westinghouse Report - Identification of Surry and North Anna Power Stations Time-Limited-Aging-Analyses for Westinghouse Supplied Components," 1/10/2001.