ML15058A034

Summary of Facility Changes, Tests and Experiments

ML15058A034
Person / Time
Site:	North Anna
Issue date:	02/23/2015
From:	Gerald Bichof Virginia Electric & Power Co (VEPCO)
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
15-053
Download: ML15058A034 (12)

Text

VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261 February 23, 2015 United States Nuclear Regulatory Commission Serial No.15-053 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 1 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 2014. provides a Commitment Change Evaluation Summary that was completed.

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

Very truly yours, Gerald T. Bischof Site Vice President Attachments

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

2. Commitment Change Evaluation Summary 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

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ATTACHMENT 1 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)

. f NORTH ANNA UNITS 1 AND 2 10 CFR 50.59

SUMMARY

DESCRIPTION OF FACILITY CHANGES, TESTS AND EXPERIMENTS 10 CFR 50.59 EVALUATION: 14-SE-MOD-01 Document Evaluated: EVAL-ENG-NAF-N2C24, Reload Safety Evaluation North Anna 2 Cycle 24 Pattern CRY Brief

Description:

The proposed activity that is the subject of this evaluation is the implementation of the new Westinghouse cladding corrosion model as a part EVAL-ENG-RSE-N2C24, Rev. 0. This methodology is documented in WCAP-12610-P-A, Addendum 2-A and was approved by the NRC on July 18, 2013.

Reason for Change: The proposed activity that is the subject of this evaluation is the implementation of WCAP-12610-P-A, Addendum 2-A, "Westinghouse Clad Corrosion Model for ZIRLO and Optimized ZIRLO" as a replacement for the existing Westinghouse cladding corrosion model. This new model is used in the North Anna 2 Cycle 24 fuel rod design analysis performed by Westinghouse. A 10 CFR 50.59 evaluation was required since the method documented in WCAP-1 2610-P-A Addendum 2-A is replacing the current method cited in the North Anna Updated Final Safety Analysis Report (UFSAR). The current fuel rod design methodology cited in the UFSAR for cladding corrosion is documented in the original issue WCAP-12610 and Addendum 1-A.

Summar: A 10 CFR 50.59 evaluation was conducted based on the application of the new methodology. The evaluation concluded that the proposed activity does not constitute a departure from a method of evaluation as the NRC has approved this method for use in the intended application for which it is applied. In the Safety Evaluation Report (SER) for WCAP-12610 Addendum 2-A, the NRC found it acceptable that Westinghouse apply this method to fuel rod design analysis for ZIRLO and Optimized ZIRLO based fuel. As the Westinghouse RFA-2 design currently employed at North Anna uses Optimized ZIRLO, the intended application is fulfilled.

The Westinghouse fuel rod design analysis for North Anna 2 Cycle 24 has been conducted in accordance with the methodologies as approved by the NRC.

10 CFR 50.59 EVALUATION: 14-SE-MOD-02 Document Evaluated: Design Change NA-14-00076, Abandonment of Unit 2 Core Exit Thermocouples Brief

Description:

Incore thermocouples 2-RC-TE-T4 in core location E4 and 2-RC-TE-T41 in core location H13 have failed and are now inoperable. Repair or replacement of these thermocouples is not a practical solution at this time.

Reason for Change: Incore thermocouples 2-RC-TE-T4 in core location E4 and 2-RC-TE-T41 in core location H13 have failed and are now inoperable. Repair or replacement of these thermocouples is not a practical solution at this time. The failed thermocouples will be abandoned in place, as sufficient instruments are still available to provide the required indication of core exit coolant temperature.

Summary: This change removed two Core Exit Thermocouples permanently from service. This change was allowed for the following reasons:

1. The change did not violate the Technical Specification limit for thermocouple operability per quadrant.

2. The thermocouples provided no control function or affected any safety systems.

3. The thermocouples were used for indication only to evaluate core conditions. While the removal of these thermocouples from service does reduce the amount of thermocouples available for monitoring, sufficient thermocouples remain for the Operator to monitor core exit temperature.

4. The change did not affect any safety system, safety analyses, or design basis described in the UFSAR.

5. The pressure boundary integrity of the RCS is maintained.

Based on the above discussion, this change did not result in a more than minimal adverse impact.

10 CFR 50.59 EVALUATION: 14-SE-MOD-03 Document Evaluated: Design Change NA-14-00006, Unit 1 Steam Generator Blowdown Sodium Analyzer Upgrade Brief

Description:

The sodium analyzers used in the Unit 1 Secondary Sampling System's On-Line Chemistry Monitoring System (OLCMS) have started to deteriorate and replacement parts are no longer manufactured. Therefore, in order to maintain compliance with Updated Final Safety Analysis Report (UFSAR) Section 18.2.5, the sodium analyzer sampling system needs to be replaced.

Reason for Change: Unit 1 Secondary Sampling System's On-Line Chemistry Monitoring System (OLCMS) operate Swan Model 2114 sodium analyzers to detect condenser tube leaks and monitor pressure boundary integrity. Over time, functions of the sodium analyzers have started to deteriorate and are operating at their lower analysis limits. Replacement parts for the sodium analyzers are no longer manufactured; therefore, in order to maintain compliance with UFSAR Section 18.2.5, the Swan Model 2114 sodium analyzer sampling system need to be replaced with the Swan Analytic AMI Soditrace sodium analyzers. Swan Analytic AMI Soditrace sodium analyzers will improve the sodium detection capabilities with better accuracy and measuring range, which is beneficial for preventing secondary side corrosion.

Swan AMI Soditrace sodium analyzers, are wider than the current sodium analyzers, especially with the covers over them. Therefore, in order to install these new sodium analyzers, modifications to Unit l's OLCMS Panel 2 (01-EI-CB-385M2) are necessary.

Modifications involve the removal of the Steam Generator Blowdown Sample Chloride Analyzers, abandoning in place two recorders, and sparing a few Flow Indicating Control Valves (FICVs).

Summary: In order to make room for the new sodium analyzers, Steam Generator Blowdown Chloride Analyzers will be removed from the OLCMS. Without the Steam Generator Blowdown Chloride Analyzers (1-SS-CLDA-1 08, 1-SS-CLDA-1 09, and 1-SS-CLDA-110) on the OLCMS, an Ion Chromatography analyzer located in the Chemistry Lab will be utilized to monitor Steam Generator chloride levels. The Ion Chromatography analyzer is designed to analyze grab samples, not continuous analyzing. Therefore, the Design Basis described in UFSAR Section 9.3.2.1, Sampling System - Design Basis, for the Steam Generator Blowdown chloride monitoring frequency will be modified to delete continuous monitoring.

During the 10 CFR 50.59 Screen, it was determined a 10 CFR 50.59 Evaluation was required to determine whether an adverse change would occur by deleting chloride from the Steam Generator continuous monitoring parameters from the Sampling System's Design Basis.

Currently, UFSAR Section 9.3.2, Sampling System, states the sampling system is to provide a means of obtaining representative primary and secondary liquid during both

normal and post accident operation. "Certain samples are continuously monitored, such as the steam generatorblowdown for radioactivity,pH, conductivity, chloride, and sodium, and the condensate pump discharge for conductivity, pH, and dissolved oxygen and sodium." Chloride was added to the UFSAR in 2000 to reflect the 1984 Steam Generator Maintenance agreement with Westinghouse. Since installation of the OLCMS, additional industry guidance has been issued for the preservation of Pressurized Water Reactor Steam Generators by NEI and EPRI. Dominion's Secondary Chemistry Program is committed to meeting the NEI 97-06 and EPRI guidelines. This commitment is reflected in UFSAR Section 18.2.5 and the North Anna Renewed Licenses.

In accordance with, EPRI's "PWR Secondary Water Chemistry Guidelines", chloride monitoring frequency requires daily sampling. To reflect EPRI's guidelines, changing the chloride monitoring frequency from continuous to daily has a negligible impact on the likelihood of a malfunction occurring to the steam generators, as previously evaluated in the SAR. Daily sampling is sufficient for maintaining an effective monitoring program.

Therefore, the Unit 1 Steam Generator Blowdown Chloride Analyzers will be removed from the OLCMS panel. UFSAR Figure 9.3-2 needs to be revised to reflect the removal of Unit 1 Steam Generator Blowdown Chloride analyzers from the OLCMS and chloride will be removed the steam generator blowdown continuous monitoring, listed in UFSAR Section 9.3.2.1, Sampling System Design Basis. UFSAR Change Request number NAPS-UCR-2014-036 has been initiated.

10 CFR 50.59 EVALUATION: 14-SE-OT-01 Document Evaluated: Engineering Technical Evaluation (ETE)-NAF-2014-0043, Implementation of Revised North Anna Containment Response Analysis for Resolution of Corrective Actions for Westinghouse NSAL-1 1-5 and NPSH Required Brief

Description:

This Engineering Technical Evaluation (ETE) implements changes to the North Anna Power Station (NAPS) Updated Final Safety Analysis Report (UFSAR)

Chapter 6 Loss of Coolant Accident (LOCA) containment analyses, including

" Containment Peak Pressure (CPP)

" Containment Pressure/Temperature Depressurization time (CDT)

" Net Positive Suction Head (NPSH) for the inside recirculation spray (IRS),

outside recirculation spray (ORS) and low head safety injection (LHSI) pumps.

Reason for Change: These analyses were revised for the following reasons:

• LOCA Mass and Energy (M&E) break flow data was modified to correct several vendor errors related to NSAL-11-5 or identified during the resolution process.

Revised M&E data was obtained for the double-ended hot leg guillotine (DEHLG) break blowdown phase and for the limiting double-ended pump suction guillotine (DEPSG) break blowdown and reflood phases. The revised M&E data also incorporates other modifications intended to recover margin.

• NPSH required (NPSHR) for the IRS and ORS pumps was revised to include temperature dependency of the pumped fluid. Previously, a constant value was assumed.

" Casing Cooling Tank (CCT) assumed fluid injection volume to the ORS pump suction was reduced to provide additional margin to the low-low CCT level isolation setpoint. This setpoint was recently increased to preclude CCT level vortex concerns, and it is desired to recover some of the fluid volume margin that was lost as a result.

• Maximum safety-related pump flows (IRS, Quench Spray (QS), Low Head Safety Injection (LHSI), High Head Safety Injection (HHSI), Service Water (SW)) were modified to include the effect of Emergency Diesel Generator (EDG) over-frequency/over-voltage (fN). The SW flow rates also consider increases associated with throttling of valves in the SW system.

It is noted that the UFSAR Chapter 6 containment analyses for main steamline break (MSLB) were not changed by this activity.

Summary: ETE-NAF-2014-0043 implemented revisions to the North Anna Power Station Units 1 and 2 containment analyses described in UFSAR Chapter 6. The changes involved modifications to analysis design inputs that are allowed within the

NRC approved analysis methodology. There were no changes to the UFSAR analytical methods.

The UFSAR described functions potentially affected by this activity included the function of the containment structure to contain the release of radioactive fluids and fission products following a LOCA by remaining within the pressure design limit and the pressure envelope assumed in the dose analyses. In addition, the function of the ESF pumps (IRS, ORS, and LHSI) to maintain adequate NPSH and associated required performance could be affected. This included the function of the Casing Cooling system to support operation of the ORS pumps and provided added cooling.

The LOCA containment analyses using the GOTHIC and Westinghouse M&E methods were revised and the results were similar to the current UFSAR reported results and all acceptance criteria continued to be met. A 10 CFR 50.59 Evaluation was required because it was necessary to revise a UFSAR safety analysis to demonstrate that all required safety functions and design requirements continued to be met. In addition, although the results remained very similar to current Analysis of Record (AOR) values, some results associated with the acceptance criteria required further evaluation. A summary of the evaluation is provided below.

The activity did not result in more than a minimal increase in the frequency of occurrence of an accident previously evaluated in the SAR because the activity only revised the UFSAR LOCA containment analyses, which assumes that a LOCA has already occurred.

The activity did not result in more than a minimal increase in the likelihood of occurrence or consequences of a malfunction of a structure, system or componet (SSC important to safety previously evaluated in the SAR because the activity made conservative assumptions relative to single failures and the loss-of-offsite power and continues to ensure adequate NPSH margins for the IRS, ORS and LHSI pumps such that required performance was maintained throughout the LOCA event.

The activity did not result in more than a minimal increase in the consequences of an accident previously evaluated in the SAR because the activity revised the UFSAR Chapter 6 LOCA containment analyses and demonstrated similar results for all acceptance criteria associated with containment pressure and IRS, ORS, and LHSI pump NPSH margins. In addition, there was no effect on the radiological dose analyses.

The activity did not create the possibility for an accident of a different type than any previously evaluated in the SAR because the activity revised the UFSAR LOCA containment analyses and only involved changes to the initial conditions and plant response for a LOCA event. This activity did not introduce any new failure modes and all containment analysis acceptance criteria continued to be met.

The activity did not create a possibility for a malfunction of a SSC important to safety with a different result than any previously evaluated in the SAR because the activity revised the NAPS LOCA containment safety analyses and demonstrated similar results for containment peak pressure, depressurization response, and IRS, ORS, and LHSI pump NPSH margins. No new failure modes were introduced by this activity and all acceptance criteria continued to be met.

The activity did not result in a design basis limit for a fission product barrier described in the SAR being exceeded or altered because this activity revised the NAPS LOCA containment safety analyses and yielded similar results for containment peak pressure and depressurization response to that previously reported and continued to be bounded by the pressure profile assumed in the radiological dose analyses. The dose analyses were not affected by this activity.

The activity did not result in a departure from a method of evaluation described in the SAR used in establishing the design bases or in the safety analyses because this activity revised the containment analyses and used the same UFSAR Chapter 6 analysis methodologies that were used for the current AOR analyses. All design inputs and model options were reviewed to ensure that there were no changes to any element of the methods used in the revised analyses.

ATTACHMENT 2 COMMITMENT CHANGE EVALUATION

SUMMARY

NORTH ANNA POWER STATION UNITS 1 AND 2 VIRGINIA ELECTRIC AND POWER COMPANY (DOMINION)

Commitment Change Evaluation Summary Original Commitment

Description:

NUREG-0053, Safety Evaluation Report Related to Operation of North Anna Power Station Units 1 and 2, Supplement 10, Section 8.3.2 discusses Diesel Generator Reliability. NUREG-0053, Supplement 10, Section 8.3.2, in part, states "The reliability of the installed diesel generators has been demonstrated by performance of the preoperational testing specified in Regulatory Guide (RG) 1.108 "Periodic Testing of Diesel Generator Units Used as Onsite Electric Power Systems at Nuclear Power Plants."

This includes performance of 69 consecutive start and load tests with zero failures, and a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> full-load-carrying capability test. A continuing demonstration of reliability will be obtained by inclusion in the Technical Specifications of the periodic testing provision of Regulatory Guide 1.108." RG 1.108 specifies pertinent testing of emergency diesel generators (EDGs) at least once every 18 months.

Revised Commitment

Description:

RG 1.108 requires EDG testing during the plant preoperational test program and at least once every 18 months. RG 1.9 replaced RG 1.108 in 1993. RG 1.9 also requires certain EDG testing during a refueling outage. Surveillance Test Interval Evaluation STI-N12-2014-002 evaluated the acceptability of extending the frequency for performing surveillance requirement (SR) 3.8.1.10, 3.8.1.12, 3.8.1.15 and 3.8.1.17, associated with ESF/LOOP actuation logic testing, from 18 months to 36 months in accordance with the Surveillance Frequency Control Program.

Justification for the Commitment Change:

1. RG 1,108 specifies pertinent testing of EDGs at least once every 18 months.

Regulatory Guide 1.9, "Application and Testing of Safety-Related Diesel Generators at Nuclear Power Plants," Revision 3 integrated into a single regulatory guide (RG) pertinent guidance previously addressed in Revision 2 of Regulatory Guide 1.9, Revision 1 of Regulatory Guide 1.108 and Generic Letter 84-15, and it endorses, as appropriate, guidelines set forth in IEEE Std 387-1984.

Regulatory Guides 1.108 and 1.9 specify pertinent testing of diesel generator units at least once every 18 months (refueling frequency).

2. Technical Specification (TS) 5.5.17 stipulates the requirements for the Surveillance Frequency Control Program (SFCP). TS 5.5.17 requires that changes to the frequencies listed in the SFCP shall be made in accordance with NEI 04-10, "Risk-Informed Method for Control of Surveillance Frequencies," Revision 1. The requirements of NEI 04-10, Revision 1 have been incorporated into procedures CM-AA-STI-101, Risk Informed Technical Specification Surveillance Frequency Control Program and CM-NA-STI-101, Technical Specification Surveillance Test Interval

(STI) List. In accordance with the NEI 04-10 and associated implementing procedures, changes to surveillance frequencies shall be made based on Maintenance Rule information, commitment review, review of the surveillance test history, equipment reliability/unavailability review, operating experience review, vendor recommended maintenance frequency, codes and standards review, other qualitative assessments, impact on defense-in depth protection, phased implementation requirements, proposed surrogate monitoring recommendations and PRA analysis.

3. Surveillance Test Interval Evaluation STI-N12-2014-002 acceptably evaluated extending the frequency for performing TS Surveillance Requirement (SR) 3.8.1.10, 3.8.1.12, 3.8.1.15 and 3.8.1.17, associated with ESF/LOOP actuation logic testing, from 18 months to 36 months in accordance with the Surveillance Frequency Control Program. The change to the surveillance frequency was, both quantitatively and qualitatively, determined to have no impact on system health, EDG design function, or continued safe operation of the plant.