ML13092A132

Summary of Facility Changes, Tests and Experiments

ML13092A132
Person / Time
Site:	North Anna
Issue date:	03/18/2013
From:	Gerald Bichof Virginia Electric & Power Co (VEPCO)
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
13-123
Download: ML13092A132 (16)

Text

VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261 March 18, 2013 United States Nuclear Regulatory Commission Serial No.13-123 Attention: Document Control Desk NAPS/JHL Washington, D. C. 20555 Docket Nos. 50-338, 339 License Nos. NPF-4, NPF-7 Gentlemen:

VIRGINIA ELECTRIC AND POWER COMPANY (DOMINION)

NORTH ANNA POWER STATION UNITS I AND 2

SUMMARY

OF FACILITY CHANGES, TESTS AND EXPERIMENTS Pursuant to 10 CFR 50.59(d)(2), a report containing a brief description of any changes, tests, and experiments, including a summary of the evaluation of each, must be submitted to the NRC, at intervals not to exceed 24 months. Attachment 1 provides a summary description of Facility Changes, Tests and Experiments identified in 10 CFR 50.59 Evaluations implemented at the North Anna Power Station during 2012. provides Commitment Change Evaluation Summaries that were completed during 2012.

If you have any questions, please contact Page Kemp at (540) 894-2295.

Very truly yours, Gerald T. Bischo*--~'

Site Vice President Attachments

1. 10 CFR 50.59 Summary Description of Facility Changes, Tests and Experiments

2. Commitment Change Evaluation Summaries cc: Regional Administrator United States 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

ATTACHMENT I 10 CFR 50.59

SUMMARY

DESCRIPTION OF FACILITY CHANGES, TESTS AND EXPERIMENTS NORTH ANNA POWER STATION UNITS 1 AND 2 VIRGINIA ELECTRIC AND POWER COMPANY (DOMINION)

,If NORTH ANNA UNITS I AND 2 10 CFR 50.59

SUMMARY

DESCRIPTION OF FACILITY CHANGES, TESTS AND EXPERIMENTS 10 CFR 50.59 EVALUATION: 12-SE-MOD-01 Document Evaluated: Design Change 07-016, Fire Protection and Domestic Water System Modifications Brief

Description:

The activity consists of changes to the Fire Protection System (FPS) and Domestic Water System (DWS) piping. These design changes will replace the existing piping west of the flood protection dike (west of the Unit 2 service and turbine buildings) with new FPS and DWS piping installed in the top ten feet of the dike. After the installation of the new FPS and DWS piping in the dike, compaction of the backfill on the dike around the piping will meet the same requirements used for original construction of the dike.

Reason for Change: This change will replace the existing FPS and DWS piping around the 4 units originally proposed for the North Anna Power Station (NAPS) site with smaller loops around only Units 1 and 2 that will perform the same overall function for supply of FPS and DWS water. The existing piping loops west of the flood protection dike have to be removed for the excavation for the new Unit 3.

Summary: This aspect of the proposed changes did not increase the frequency or the consequences of an accident previously evaluated in the UFSAR. There are no accidents evaluated in the UFSAR that consider failure of the dike. This aspect of the proposed changes also did not increase the likelihood of occurrence or the consequences of a malfunction of an SSC important to safety. The UFSAR does not evaluate a malfunction of the dike, and the proposed activities would not result in a malfunction of the dike. This aspect also did not result in a different accident or malfunction than any previously evaluated In the UFSAR because failure of the DWS or FPS piping would not cause a failure of the dike to provide flood protection during a probable maximum flood (PMF) event and therefore a different accident or malfunction would not be created. This aspect of the proposed changes also is not associated with a design basis limit for a fission product barrier, and does not depart from a method of evaluation described In the UFSAR.

In summary, if the changes to the FPS and DWS piping are implemented, the safety related design function for flood protection by the dike as documented in the UFSAR remains unchanged. Specifically, the ability of the dike to provide flood protection will remain as previously reviewed and approved by the NRC.

.I 10 CFR 50.59 EVALUATION: 12-SE-OT-01 Document Evaluated: Engineering Technical Evaluation (ETE)-NAF-2011-0173, Revision 0, Implementation of the Analyses Supporting the Westinghouse 17x17 RFA-2 Fuel at North Anna Units I and 2, and Updated Final Safety Analysis Report Change Request (UCR) 2010-007 Brief

Description:

This 10 CFR 50.59 evaluation is being performed to support the transition to RFA-2 fuel at North Anna.

Reason for Change: ETE-NAF-2011-0173 implements the analyses supporting the RFA-2 fuel transition at North Anna. UCR-2010-007 is a UFSAR change request that encompasses the changes related to the RFA-2 fuel transition at North Anna. Three License Amendment Requests (LARs) were submitted to the NRC for approval in support of the RFA-2 transition at North Anna. The LAR for Optimized ZIRLO has been approved and the LARs for VIPRE-D and Westinghouse's Best Estimate Large Break LOCA evaluation were being tracked and were subsequently approved by the NRC.

Summary: This 10 CFR 50.59 evaluation was performed for the RFA-2 fuel transition at North Anna. This evaluation was performed to address (14) items that screened in.

The 14 items that screened in were:

(1) Methodology change for fuel assembly lift force analysis (2) Methodology change for Cladding stress limit (3) Methodology change for stress analysis supporting WABAs (4) Reduction in margin for hydrodynamic flow instability (5) Reanalysis of the main steamline break safety analysis (6) Reanalysis of the rod ejection safety analysis (7) Methodology change due to the use of MULTIFLEX 3.0 (8) RCCA drop time administrative limit (9) Use of ANSYS vs. WECAN as a method of evaluation in Reactor Vessel Internals structural analysis (10) Reactor vessel structural reanalysis (11) Change in pressurizer surge line thermal stratification profile (12) RCP horsepower and thrust loads (13) Reactor vessel head assembly reanalysis (14) Methodology change for evaluation of DNB transition core penalties for Advanced Mark-BW fuel NRC review and approval is not required prior to implementing the changes associated with the analysis supporting the RFA-2 fuel transition at North Anna. The fourteen (14) items listed in this 10CFR 50.59 evaluation have been shown to not impact the associated design function, exceed/alter a design basis limit for a fission product barrier (DBLFPB), or result in a departure in a method of evaluation as described in the UFSAR. All other changes and analysis performed for the RFA-2 fuel transition either

were approved by the NRC or screened out. Therefore, the RFA-2 fuel transition can be performed under the provisions of 10CFR 50.59.

Additionally, this evaluation supports the UFSAR change request that encompasses the changes related to the RFA-2 fuel transition at North Anna.

10 CFR 50.59 EVALUATION: 12-SE-OT-02 Document Evaluated: Technical Requirement (TR) 3.9.3, Manipulator Crane Brief

Description:

Technical Requirement 3.9.3, Manipulator Crane, states "The manipulator crane and auxiliary hoist shall be used for movement of control rods or fuel assemblies and shall be FUNCTIONAL. The auxiliary hoist shall have a load indicator and shall not be used to lift loads in excess of 600 pounds." A temporary revison to TR 3.9.3 is being performed to allow movement of potentially damaged fuel assemblies with a crane other than the manipulator crane.

Reason for Change: TR 3.9.3, Manipiulator Crane, is temporary revised to allow removal of potentially damaged fuel assemblies from the North Anna Unit 1 core /

containment with a crane other than the manipulator crane.

Summary: On March 18, 2012, while lifting the North Anna Unit 1 reactor vessel upper internals, the lift was stopped when it was identified that fuel assembly 0X5 at core location M13 was attached to the upper internals package and was coming out of the core as the package was lifted. Under separate documentation, the fuel assembly will be stabilized and disconnected from the upper internals, and the internals moved to the stand. At that point the fuel assembly will be able to be moved to the upender and to the spent fuel pit for further examination. Photos of fuel assembly 0X5 indicate potential damage to the top nozzle which may preclude use of normal fuel movement tooling to remove it from the core (manipulator crane). Furthermore, there exists some uncertainty that if the manipulator crane could latch 0X5, it may not be able to unlatch it, which would preclude retrieving the fuel assembly from the crane mast. Use of special tooling to grasp the fuel assembly is required, and therefore the manipulator crane cannot be used.

Design requirements of the fuel handling system, and specifically the manipulator crane, are detailed in UFSAR Section 9.1.4. The following applicable fuel handling equipment requirements were assembled from section 9.1.4 are being added to the bases of Technical Requirement 3.9.3 as requirements for any crane used to move potentially damaged fuel out of the reactor core and out of containment:

1. The load carrying capability of the crane must be verified greater than five times the expected load,

2. When handling fuel assemblies, administrative or electronic means to stop fuel movement must be used if indications of assembly binding are encountered,

3. When handling fuel assemblies, administrative or electronic means to prevent interaction with permanent structures must be used,

4. When handling fuel assemblies, administrative or electronic means to maintain a minimum required depth of water shielding (7 feet) must be used, and

5. The handling equipment used must be capable of maintaining the load engaged and suspended under design basis earthquake conditions.

Given that the TR will require that any crane used to manipulate fuel will meet the same requirements as the manipulator crane, the proposed change has been demonstrated to satisfy 10 CFR 50.59 criteria 1 through 6 (criteria 7 and 8 are not applicable).

Therefore, prior NRC review and approval is not required.

10 CFR 50.59 EVALUATION: 12-SE-OT-03 Document Evaluated: Engineering Technical Evaluation (ETE)-NAF-2012-0083, Revision 0, Assessment of AREVA Advanced Mark-BW Fuel at North Anna in Response to NSAL 11-2 Brief

Description:

This 10 CFR 50.59 evaluation has been performed for an update to the structural analysis of the AREVA Advanced Mark-BW fuel product. Two items that screened in for evaluation are the increased amount of grid deformation under the combined LOCA/seismic loadings and a change to the method of combining the LOCA and seismic time histories.

Reason for Change: The activity being evaluated is the implementation of an updated structural analysis basis for the AREVA Advanced Mark-BW fuel product at North Anna.

The structural analysis basis for the AREVA Advanced Mark-BW fuel design is being updated in response to the generation of new seismic/LOCA core plate motions by Westinghouse in support of the RFA-2 fuel transition at North Anna.

Summary: The increased amount of grid deformation under combined LOCA/seismic loadings is limited to the core periphery. The analysis for the amount of grid deformation under combined LOCA/seismic loading is used to verify that the core remains amenable to cooling and that the control rods are able to insert. The increased grid deformation is limited to the core periphery and has been evaluated to ensure the design limits continue to be met.

The change in the method of combining the LOCA and seismic time histories from that presented in BAW-1033PA to that used in the updated analysis results in a predicted amount of grid deformation that is essentially the same. The seismic loadings due to the time histories contribute less than 0.5% to the combined loading. Since the LOCA loads dominate the combined response, the change in the method of combining the time histories results in essentially the same amount of grid deformation that is predicted when the square-root-of-sum-of-squares is used to combine the loads. As such, this change does not result in departure from a method of evaluation described in the Safety Analysis Report used in establishing the design bases or in the safety analyses.

NRC review and approval is not required prior to implementing the changes associated with the structural analysis of the AREVA Advanced Mark-BW fuel at North Anna. The two items listed in this 10 CFR 50.59 evaluation have been shown to not impact the associated design function, exceed/alter the design basis limit for a fission product barrier, or result in a departure in a method of evaluation as described in the UFSAR.

10 CFR 50.59 EVALUATION: 12-SE-PROC-01 Documents Evaluated: 1-GOP-4.25, Fuel Assembly Recovery Process, 0-OP-4.59, Operation of the Special Fuel Assembly Cruciform Handling Tool, and Engineering Technical Evaluation (ETE)-NA-2012-0024, Fuel Assembly Capture Device, Dislodge Tool, and Cruciform Handling Tool Evaluation Brief

Description:

1-GOP-4.25 and 0-OP-4.59 have been developed to capture fuel assembly 0X5, dislodge it from the upper internals, and move it into storage in the Spent Fuel Pit. Development and qualification of required tooling is documented in ETE-NA-2012-0024.

Reason for Change: On March 18, 2012, while lifting the North Anna Unit 1 reactor vessel upper internals, the lift was stopped when it was identified that fuel assembly 0X5 at core location M13 was attached to the upper internals package and was coming out of the core as the package was lifted. The fuel assembly will be stabilized and disconnected from the upper internals, and the internals moved to the stand. At that point the fuel assembly will be able to be moved to the upender and to the spent fuel pit for further examination. Photos of fuel assembly 0X5 indicate potential damage to the top nozzle which may preclude use of normal fuel movement tooling to remove it from the core with the manipulator crane. Furthermore, there exists some uncertainty that if the manipulator crane could latch 0X5, it may not be able to unlatch it, which would preclude retrieving the fuel assembly from the crane mast by normal means. Use of special tooling to grasp the fuel assembly is required, and therefore the manipulator cannot be used. TRM 3.9.3 has been revised to allow fuel movement in containment with a crane other than the manipulator crane. 1-GOP-4.25 and 0-OP-4.59 have been developed to capture fuel assembly 0X5, remove it from the upper internals, and move it to storage in the Spent Fuel Pit. Development and qualification of required tooling is documented in ETE-NA-2012-0024.

Summary: On March 18, 2012, while lifting the North Anna Unit 1 reactor vessel upper internals, the lift was stopped when it was identified that fuel assembly 0X5 at core location M13 was attached to the upper internals package and was coming out of the core as the package was lifted. The fuel assembly will be stabilized and disconnected from the upper internals, and the internals moved to the stand. At that point the fuel assembly will be able to be moved to the upender and to the spent fuel pit for further examination. Photos of fuel assembly 0X5 indicate potential damage to the top nozzle which may preclude use of normal fuel movement tooling to remove it from the core with the manipulator crane.

Design requirements of the fuel handling system, and specifically the manipulator crane, are detailed in UFSAR Section 9.1.4. The following applicable fuel handling equipment requirements were assembled from section 9.1.4 are being added to the bases of Technical Requirement 3.9.3 as requirements for any crane used to move potentially damaged fuel out of the reactor core and out of containment:

1. The load carrying capability of the crane must be verified greater than five times the expected load,

2. When handling fuel assemblies, administrative or electronic means to stop fuel movement must be used if indications of assembly binding are encountered,

3. When handling fuel assemblies, administrative or electronic means to prevent interaction with permanent structures must be used,

4. When handling fuel assemblies, administrative or electronic means to maintain a minimum required depth of water shielding (7 feet) must be used, and

5. The handling equipment used must be capable of maintaining the load engaged and suspended under design basis earthquake conditions.

The evolution will move the fuel assembly from a condition where a fuel handling accident (dropped assembly) is possible (attached to the upper internals via the friction interaction of a guide pin and the associated fuel assembly top nozzle) to a safe condition (captured and moved to the pent fuel pit.

ETE-NA-2012-0024 documents the design of the tools used in 1-GOP-4.25 and 0-OP-4.59. The tools have been designed to support the maximum loads expected without dropping the assembly.

Therefore, prior NRC review and approval of the proposed change is not required.

ATTACHMENT 2 COMMITMENT CHANGE EVALUATION SUMMARIES NORTH ANNA POWER STATION UNITS I AND 2 VIRGINIA ELECTRIC AND POWER COMPANY (DOMINION)

Commitment Change Evaluation Summaries Original Commitment

Description:

On March 18, 2020, the NRC issued Bulletin 2002-01, Reactor Pressure Vessel Head Degradation and Reactor Coolant Pressure Boundary Integrity. In a letter to the NRC dated May 16, 2002, Dominion provided a description of the boric acid corrosion control program. In a November 21, 2002 NRC request for additional information, the NRC requested additional information associated with the boric acid corrosion control program.

Dominion responded to the request for information is a letter dated January 31, 2003.

The following commitment was made in that letter: Bare metal visual (BMV) exams of locations highly susceptible to Primary Water Stress Corrosion Cracking (PWSCC) will be performed every outage. Listed among those locations was Steam Generator (SG) primary loop nozzle Alloy 82/182 butter welds at North Anna Unit 1.

Revised Commitment

Description:

Reduce the frequency of BMV exams of North Anna Unit 1 SG cold leg nozzles from every refueling outage to once per interval.

Justification for the Commitment Change:

Inspection requirement for Class 1 Pressurized Water Reactor (PWR) Components containing Alloy 600/82/182 are listed in Code Case N-722-1 and codified in 10CFR50.55a. Inspection Item B15.115 for SG cold leg nozzle-to-pipe connections lists a requirement to do a BMV exam once per interval. More susceptible hot leg locations are required to be inspected every refueling outage per Inspection Item B15.110. In addition to a once per interval BMV exam, these locations are inspected volumetrically every second inspection period not to exceed 7 years in accordance with Code Case N-770-1 as implemented by 10CFR50.55a. Due to lower temperatures of cold legs, any crack that does initiate will grow slowly and will be detected before the flaw becomes safety significant. Previous BMV exams for SG cold leg nozzle locations revealed no evidence of pressure boundary leakage or corrosion. Therefore, due to the history of satisfactory inspections and the reduced temperature and subsequently reduced susceptibility to PWSCC of SG cold leg nozzles, it is reasonable to reduce the frequency of BMV exams to follow Code requirements.

Original Commitment

Description:

Virginia Electric and Power Company letter dated February 18, 1988 and NRC Safety Evaluation Report dated May 26, 1988 specify the Accident Mitigation System Actuation Circuitry (AMSAC) system will be tested end-to-end (i.e., input to AMSAC through output of AMSAC) during each refueling outage.

Revised Commitment

Description:

Since testing was performed during the most recent forced outage due to the earthquake, the testing scheduled during the next refueling outage is not required. This testing will resume during the next North Anna Unit 1 refueling outage in September 2013.

Justification for the Commitment Change:

This change does not negatively impact the ability of the AMSAC system to perform its safety function. The testing that was performed in Spetember 2011 adequately demonstrated the ability of the system to perform its safety function. Also, the logic testing that occurred on January 4, 2012 demonstrated that the AMSAC system had no functionality issues and is fully capable of performing as specified. The change also does not negatively impact the ability of personnel to ensure the AMSAC system is capable of performing its function. The AMSAC system logic is tested on a quarterly basis. This logic check ensures that all aspects of the AMSAC logic in functioning properly. The only difference between the quarterly testing and the full outage testing is a testing of the equipment that is initiated by the AMSAC system. However, the testing performed in September 2011 verified this equipment.

A review of the history of AMSAC testing shows that in the last 10 years there have been no equipment issues with the testing performed. There have also been no issues in that same timeframe with the quarterly testing. There have been a total of 13 equipment issues with the AMSAC system since 1999 (9 on Unit 1 and 4 on Unit 2).

None of these issues occurred during testing (quarterly or during a refueling outage). A review of these issues finds that none of the equipment issues would have prevented the AMSAC system from performing its intended function.

Based on the historical testing information, a successful quarterly test and end-to-end test in the past 5 months, that is no need to perform another test during the upcoming Unit 1 refueling outage in March 2012. Testing will be performed during the next Unit 1 refueling outage scheduled for September 2013.

Original Commitment

Description:

NRC Inspection Reports Nos. 50-338/88-02 and 50-339/88-02 document violations of the ASME Xl requirements for the testing of valves. Specifically, a number of air operated valve stroke times exceeded the stroke time limits or failed to exhibit the required change of valve stem or disk position during stroke time testing. Part of the cause of the solenoid operated valve (SOV) failures that contributed to these violations was that contaminants from the Instrument Air (IA) system were accumulating inside the SOV's and causing them to stick. One of the recommendations by ASCO to alleviate the problem with the sticking SOV's was to increase the frequency of testing of SOV's to every month instead of every 3 months as required by the ASME Code.

Based on this recommendation by the manufacturer, the Reply to the Notice of Violation from NRC Inspection Reports Nos. 50-338/88-02 and 50-339/88-02 committed to stroking SOV actuated cold shutdown valves or those valves previously identified as cold shutdown valves which could be tested during normal operations, every month.

Revised Commitment

Description:

Perform quarterly stroke time testing of Inservice Testing (IST) valves with ASCO solenoid operated valves (SOVs) in accordance with the ASME Code for Operation and Mainetence (OM) 2004 Edition.

Justification for the Commitment Change:

Engineering Technical Evaluation ETE-NA-2011-0100 evaluated changing the commitment to stroke IST valves with ASCO SOVs from a monthly to a quarterly frequency. The evaluation indicated that the change in the testing frequency was justified based on upgrades that were made to the IA System which improved air quality. Specifically, Design Change 89-04-3 was implemented to upgrade the instrument air system. The Sullair reciprocating compressors were replaced with Atlas Copco oil free, rotary screw air compressors. The air dryers were replaced with heatless regenerative type air dryers, which provide better water removal. The dew point and filtration capabilities of the new air dryers and filters meet the filtration and dew point requirements of ISA-7.3. Implementation of this upgraded IA system was completed in 1992.

The problem with the sticking SOVs associated with the air-operated valves (AOVs) have been eliminated and IST stroke time failures have been drastically reduced. It is justified to return to the quarterly frequency specified in the ASME OM Code as required by the North Anna IST Program for Pumps and Valves.

Original Commitment

Description:

Virginia Electric and Power Company (VEPCO) letter dated January 12, 1979 provided additional information regarding the second level of undervoltage protection. This letter stated (regarding the Degraded Voltage relay setting) that "This 90% voltage level sepoint represents the minimum voltage at which safety-related motors can operate within their nameplate values and offers considerable margin above the approximate 85% potential value at which the loads would be operated outside their rated service factor."

NRC letter dated March 20, 1979, entitled North Anna Power Station Units 1 and 2 -

Protection from Degraded Grid Voltage Condition/Interaction of Offsite and Onsite Power Systems, indicated that the design pertaining to protection from a degraded voltage condition was acceptable.

Revised Commitment

Description:

Although not a commitment, the revision to the licensing basis (Electrical Engineering, Calculation EE-0373, Rev. 0) justifies some 460 volt motors and motor control centers (MCCs) capability to perform their intended safety function with less than 90% terminal voltage during a degraded voltage condition.

Justification for the Commitment Change:

Calculation EE-0373, Rev. 0 dated April 1991, concluded that the existing degraded voltage relay setpoint is adequate. While the setpoint does not ensure greater than 90% rated terminal voltage for 460 volt motors, it does provide greater than 85% at the terminals of these loads with only a few exceptions that were evaluated as acceptable.

The 4000 volt motors are ensured adequate voltage with the existing setpoint for a sustained degraded voltage condition. Additional reviews performed during the 2012 NRC Component Design Basis Inspection regarding thermal overload sizing for continuous duty motors considering a sustained degraded voltage condition also supports equipment operability as was concluded by Calculation EE-0373, Rev. 0.

Based on the additional reviews and Calculation EE-0373, Rev. 0 conclusions regarding the adequacy of the degraded voltage setting, Engineering considers the degraded voltage setpoint and associated instrumentation capable of performing its design function but non-conforming based on the statement in the January 12, 1979 VEPCO letter.

The Quench Spray and Inside Recirculation Spray pump flows have been previously evaluated at the expected 480 volt bus voltages. Generic Letter 89-10 motor operated valves have been previously evaluated at the expected 480 volt motor control center voltages to ensure that sufficient torque is developed. Continuous duty 460 volt motors supplied from the motor control centers have margin to permit operation at 85 to 90 percent of rated voltage. Motors are purchased with a 1.15 service factor and are sized to operate below rated horsepower. Motor ambient temperatures are below maximum ratings for most conditions. Safety related motors are typically purchased to start at 70% voltage and therefore have sufficient torque to run at degrade voltage conditions.

The increased current and motor winding heating that occur from operation at 85 to 90 percent voltage can result in reduced motor life. This reduction is offset by many years of operation at reduced current as the offsite power transformer(s) automatic load tap changers normally maintain higher than normal voltages. While it is normal industry practice to calculate equipment voltages based on 4160 volt bus voltage just above the degraded voltage relay setting, it is relevant to note that separate GDC-17 voltage profile analyses cantained in Calculation EE-0008 do not predict sustained degraded voltages.