ML15058A034
| ML15058A034 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 02/23/2015 |
| From: | Bischof G T 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 COMPANYRICHMOND, VIRGINIA 23261February 23, 2015United States Nuclear Regulatory Commission Serial No. 15-053Attention: Document Control Desk NAPS/JHLWashington, D. C. 20555 Docket Nos. 50-338, 339License Nos. NPF-4, NPF-7Gentlemen:VIRGINIA ELECTRIC AND POWER COMPANY (DOMINION)NORTH ANNA POWER STATION UNITS 1 AND 2SUMMARY OF FACILITY CHANGES, TESTS AND EXPERIMENTSPursuant 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 besubmitted to the NRC, at intervals not to exceed 24 months. Attachment 1 provides asummary description of Facility Changes, Tests and Experiments identified in 10 CFR50.59 Evaluations implemented at the North Anna Power Station during 2014.Attachment 2 provides a Commitment Change Evaluation Summary that wascompleted.If you have any questions, please contact Page Kemp at (540) 894-2295.Very truly yours,Gerald T. BischofSite Vice PresidentAttachments1. 10 CFR 50.59 Summary Description of Facility Changes, Tests and Experiments2. Commitment Change Evaluation Summarycc: Regional AdministratorUnited States Nuclear Regulatory CommissionRegion IIMarquis One Tower245 Peachtree Center Ave., NE, Suite 1200Atlanta, Georgia 30303-1257NRC Senior Resident InspectorNorth Anna Power Station/5Z1 7 ATTACHMENT 110 CFR 50.59 SUMMARY DESCRIPTION OFFACILITY CHANGES, TESTS AND EXPERIMENTSNORTH ANNA POWER STATION UNITS 1 AND 2VIRGINIA ELECTRIC AND POWER COMPANY (DOMINION)
.fNORTH ANNA UNITS 1 AND 210 CFR 50.59 SUMMARY DESCRIPTION OFFACILITY CHANGES, TESTS AND EXPERIMENTS10 CFR 50.59 EVALUATION: 14-SE-MOD-01Document Evaluated: EVAL-ENG-NAF-N2C24, Reload Safety Evaluation North Anna 2Cycle 24 Pattern CRYBrief
Description:
The proposed activity that is the subject of this evaluation is theimplementation 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 theimplementation of WCAP-12610-P-A, Addendum 2-A, "Westinghouse Clad CorrosionModel for ZIRLO and Optimized ZIRLO" as a replacement for the existingWestinghouse cladding corrosion model. This new model is used in the North Anna 2Cycle 24 fuel rod design analysis performed by Westinghouse. A 10 CFR 50.59evaluation was required since the method documented in WCAP-1 2610-P-A Addendum2-A is replacing the current method cited in the North Anna Updated Final SafetyAnalysis Report (UFSAR). The current fuel rod design methodology cited in the UFSARfor cladding corrosion is documented in the original issue WCAP-12610 and Addendum1-A.Summar: A 10 CFR 50.59 evaluation was conducted based on the application of thenew methodology. The evaluation concluded that the proposed activity does notconstitute a departure from a method of evaluation as the NRC has approved thismethod for use in the intended application for which it is applied. In the SafetyEvaluation Report (SER) for WCAP-12610 Addendum 2-A, the NRC found itacceptable that Westinghouse apply this method to fuel rod design analysis for ZIRLOand Optimized ZIRLO based fuel. As the Westinghouse RFA-2 design currentlyemployed at North Anna uses Optimized ZIRLO, the intended application is fulfilled.The Westinghouse fuel rod design analysis for North Anna 2 Cycle 24 has beenconducted in accordance with the methodologies as approved by the NRC.
10 CFR 50.59 EVALUATION: 14-SE-MOD-02Document Evaluated: Design Change NA-14-00076, Abandonment of Unit 2 Core ExitThermocouplesBrief
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 ofthese 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 orreplacement of these thermocouples is not a practical solution at this time. The failedthermocouples will be abandoned in place, as sufficient instruments are still available toprovide the required indication of core exit coolant temperature.Summary: This change removed two Core Exit Thermocouples permanently fromservice. This change was allowed for the following reasons:1. The change did not violate the Technical Specification limit for thermocoupleoperability 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. Whilethe removal of these thermocouples from service does reduce the amount ofthermocouples available for monitoring, sufficient thermocouples remain for theOperator to monitor core exit temperature.4. The change did not affect any safety system, safety analyses, or design basisdescribed 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 minimaladverse impact.
10 CFR 50.59 EVALUATION: 14-SE-MOD-03Document Evaluated: Design Change NA-14-00006, Unit 1 Steam GeneratorBlowdown Sodium Analyzer UpgradeBrief
Description:
The sodium analyzers used in the Unit 1 Secondary SamplingSystem's On-Line Chemistry Monitoring System (OLCMS) have started to deteriorateand replacement parts are no longer manufactured. Therefore, in order to maintaincompliance with Updated Final Safety Analysis Report (UFSAR) Section 18.2.5, thesodium analyzer sampling system needs to be replaced.Reason for Change: Unit 1 Secondary Sampling System's On-Line ChemistryMonitoring System (OLCMS) operate Swan Model 2114 sodium analyzers to detectcondenser tube leaks and monitor pressure boundary integrity. Over time, functions ofthe sodium analyzers have started to deteriorate and are operating at their loweranalysis limits. Replacement parts for the sodium analyzers are no longermanufactured; 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 theSwan Analytic AMI Soditrace sodium analyzers. Swan Analytic AMI Soditrace sodiumanalyzers will improve the sodium detection capabilities with better accuracy andmeasuring 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 sodiumanalyzers, modifications to Unit l's OLCMS Panel 2 (01-EI-CB-385M2) are necessary.Modifications involve the removal of the Steam Generator Blowdown Sample ChlorideAnalyzers, abandoning in place two recorders, and sparing a few Flow IndicatingControl Valves (FICVs).Summary: In order to make room for the new sodium analyzers, Steam GeneratorBlowdown Chloride Analyzers will be removed from the OLCMS. Without the SteamGenerator 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 ChemistryLab will be utilized to monitor Steam Generator chloride levels. The IonChromatography analyzer is designed to analyze grab samples, not continuousanalyzing. Therefore, the Design Basis described in UFSAR Section 9.3.2.1, SamplingSystem -Design Basis, for the Steam Generator Blowdown chloride monitoringfrequency will be modified to delete continuous monitoring.During the 10 CFR 50.59 Screen, it was determined a 10 CFR 50.59 Evaluation wasrequired to determine whether an adverse change would occur by deleting chloridefrom the Steam Generator continuous monitoring parameters from the SamplingSystem's Design Basis.Currently, UFSAR Section 9.3.2, Sampling System, states the sampling system is toprovide 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 generator blowdown for radioactivity, pH, conductivity, chloride, andsodium, and the condensate pump discharge for conductivity, pH, and dissolvedoxygen and sodium." Chloride was added to the UFSAR in 2000 to reflect the 1984Steam Generator Maintenance agreement with Westinghouse. Since installation of theOLCMS, additional industry guidance has been issued for the preservation ofPressurized Water Reactor Steam Generators by NEI and EPRI. Dominion'sSecondary Chemistry Program is committed to meeting the NEI 97-06 and EPRIguidelines. This commitment is reflected in UFSAR Section 18.2.5 and the North AnnaRenewed Licenses.In accordance with, EPRI's "PWR Secondary Water Chemistry Guidelines", chloridemonitoring frequency requires daily sampling. To reflect EPRI's guidelines, changingthe chloride monitoring frequency from continuous to daily has a negligible impact onthe likelihood of a malfunction occurring to the steam generators, as previouslyevaluated in the SAR. Daily sampling is sufficient for maintaining an effectivemonitoring program.Therefore, the Unit 1 Steam Generator Blowdown Chloride Analyzers will be removedfrom the OLCMS panel. UFSAR Figure 9.3-2 needs to be revised to reflect the removalof Unit 1 Steam Generator Blowdown Chloride analyzers from the OLCMS and chloridewill be removed the steam generator blowdown continuous monitoring, listed in UFSARSection 9.3.2.1, Sampling System Design Basis. UFSAR Change Request numberNAPS-UCR-2014-036 has been initiated.
10 CFR 50.59 EVALUATION: 14-SE-OT-01Document Evaluated: Engineering Technical Evaluation (ETE)-NAF-2014-0043,Implementation of Revised North Anna Containment Response Analysis for Resolutionof Corrective Actions for Westinghouse NSAL-1 1-5 and NPSH RequiredBrief
Description:
This Engineering Technical Evaluation (ETE) implements changes tothe 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 severalvendor 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 alsoincorporates other modifications intended to recover margin.* NPSH required (NPSHR) for the IRS and ORS pumps was revised to includetemperature dependency of the pumped fluid. Previously, a constant value wasassumed." Casing Cooling Tank (CCT) assumed fluid injection volume to the ORS pumpsuction was reduced to provide additional margin to the low-low CCT levelisolation setpoint. This setpoint was recently increased to preclude CCT levelvortex concerns, and it is desired to recover some of the fluid volume margin thatwas lost as a result.* Maximum safety-related pump flows (IRS, Quench Spray (QS), Low Head SafetyInjection (LHSI), High Head Safety Injection (HHSI), Service Water (SW)) weremodified to include the effect of Emergency Diesel Generator (EDG)over-frequency/over-voltage (fN). The SW flow rates also consider increasesassociated 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 PowerStation Units 1 and 2 containment analyses described in UFSAR Chapter 6. Thechanges involved modifications to analysis design inputs that are allowed within the NRC approved analysis methodology. There were no changes to the UFSAR analyticalmethods.The UFSAR described functions potentially affected by this activity included thefunction of the containment structure to contain the release of radioactive fluids andfission products following a LOCA by remaining within the pressure design limit and thepressure envelope assumed in the dose analyses. In addition, the function of the ESFpumps (IRS, ORS, and LHSI) to maintain adequate NPSH and associated requiredperformance could be affected. This included the function of the Casing Cooling systemto support operation of the ORS pumps and provided added cooling.The LOCA containment analyses using the GOTHIC and Westinghouse M&E methodswere revised and the results were similar to the current UFSAR reported results and allacceptance criteria continued to be met. A 10 CFR 50.59 Evaluation was requiredbecause it was necessary to revise a UFSAR safety analysis to demonstrate that allrequired 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. Asummary of the evaluation is provided below.The activity did not result in more than a minimal increase in the frequency ofoccurrence of an accident previously evaluated in the SAR because the activity onlyrevised the UFSAR LOCA containment analyses, which assumes that a LOCA hasalready occurred.The activity did not result in more than a minimal increase in the likelihood ofoccurrence or consequences of a malfunction of a structure, system or componet (SSCimportant to safety previously evaluated in the SAR because the activity madeconservative assumptions relative to single failures and the loss-of-offsite power andcontinues to ensure adequate NPSH margins for the IRS, ORS and LHSI pumps suchthat required performance was maintained throughout the LOCA event.The activity did not result in more than a minimal increase in the consequences of anaccident previously evaluated in the SAR because the activity revised the UFSARChapter 6 LOCA containment analyses and demonstrated similar results for allacceptance criteria associated with containment pressure and IRS, ORS, and LHSIpump NPSH margins. In addition, there was no effect on the radiological doseanalyses.The activity did not create the possibility for an accident of a different type than anypreviously evaluated in the SAR because the activity revised the UFSAR LOCAcontainment analyses and only involved changes to the initial conditions and plantresponse for a LOCA event. This activity did not introduce any new failure modes andall containment analysis acceptance criteria continued to be met.
The activity did not create a possibility for a malfunction of a SSC important to safetywith a different result than any previously evaluated in the SAR because the activityrevised the NAPS LOCA containment safety analyses and demonstrated similar resultsfor containment peak pressure, depressurization response, and IRS, ORS, and LHSIpump NPSH margins. No new failure modes were introduced by this activity and allacceptance criteria continued to be met.The activity did not result in a design basis limit for a fission product barrier described inthe SAR being exceeded or altered because this activity revised the NAPS LOCAcontainment safety analyses and yielded similar results for containment peak pressureand depressurization response to that previously reported and continued to be boundedby the pressure profile assumed in the radiological dose analyses. The dose analyseswere not affected by this activity.The activity did not result in a departure from a method of evaluation described in theSAR used in establishing the design bases or in the safety analyses because thisactivity revised the containment analyses and used the same UFSAR Chapter 6analysis methodologies that were used for the current AOR analyses. All design inputsand model options were reviewed to ensure that there were no changes to any elementof the methods used in the revised analyses.
ATTACHMENT 2COMMITMENT CHANGE EVALUATION SUMMARYNORTH ANNA POWER STATION UNITS 1 AND 2VIRGINIA ELECTRIC AND POWER COMPANY (DOMINION)
Commitment Change Evaluation SummaryOriginal Commitment
Description:
NUREG-0053, Safety Evaluation Report Related to Operation of North Anna PowerStation Units 1 and 2, Supplement 10, Section 8.3.2 discusses Diesel GeneratorReliability. NUREG-0053, Supplement 10, Section 8.3.2, in part, states "The reliability ofthe installed diesel generators has been demonstrated by performance of thepreoperational testing specified in Regulatory Guide (RG) 1.108 "Periodic Testing ofDiesel 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 a24 hour full-load-carrying capability test. A continuing demonstration of reliability will beobtained by inclusion in the Technical Specifications of the periodic testing provision ofRegulatory Guide 1.108." RG 1.108 specifies pertinent testing of emergency dieselgenerators (EDGs) at least once every 18 months.Revised Commitment
Description:
RG 1.108 requires EDG testing during the plant preoperational test program and atleast once every 18 months. RG 1.9 replaced RG 1.108 in 1993. RG 1.9 also requirescertain EDG testing during a refueling outage. Surveillance Test Interval EvaluationSTI-N12-2014-002 evaluated the acceptability of extending the frequency forperforming 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 inaccordance 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 Generatorsat 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, asappropriate, guidelines set forth in IEEE Std 387-1984.Regulatory Guides 1.108 and 1.9 specify pertinent testing of diesel generator unitsat least once every 18 months (refueling frequency).2. Technical Specification (TS) 5.5.17 stipulates the requirements for the SurveillanceFrequency Control Program (SFCP). TS 5.5.17 requires that changes to thefrequencies listed in the SFCP shall be made in accordance with NEI 04-10, "Risk-Informed Method for Control of Surveillance Frequencies," Revision 1. Therequirements of NEI 04-10, Revision 1 have been incorporated into procedures CM-AA-STI-101, Risk Informed Technical Specification Surveillance Frequency ControlProgram and CM-NA-STI-101, Technical Specification Surveillance Test Interval (STI) List. In accordance with the NEI 04-10 and associated implementingprocedures, changes to surveillance frequencies shall be made based onMaintenance Rule information, commitment review, review of the surveillance testhistory, equipment reliability/unavailability review, operating experience review,vendor recommended maintenance frequency, codes and standards review, otherqualitative assessments, impact on defense-in depth protection, phasedimplementation requirements, proposed surrogate monitoring recommendations andPRA analysis.3. Surveillance Test Interval Evaluation STI-N12-2014-002 acceptably evaluatedextending 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 FrequencyControl Program. The change to the surveillance frequency was, both quantitativelyand qualitatively, determined to have no impact on system health, EDG designfunction, or continued safe operation of the plant.