ML050960502

Summary of Facility Changes, Tests, and Experiments for 2004

ML050960502
Person / Time
Site:	North Anna
Issue date:	03/29/2005
From:	Jennifer Davis Virginia Electric & Power Co (VEPCO)
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
05-152
Download: ML050960502 (6)

Text

VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261 March 29, 2005 United States Nuclear Regulatory Commission Serial No.05-152 Attention: Document Control Desk NAPS/JRP Washington, D. C. 20555 Docket Nos. 50-338 50-339 License Nos. NPF-4 NPF-7 Gentlemen:

VIRGINIA ELECTRIC AND POWER COMPANY NORTH ANNA POWER STATION UNITS 1 AND 2

SUMMARY

OF FACILITY CHANGES. TESTS AND EXPERIMENTS Pursuant to 10 CFR 50.59(d)(2), enclosed, is a summary description of Facility Changes, Tests and Experiments identified in Regulatory Evaluations implemented at the North Anna Power Station during 2004.

Ifyou have any questions, please contact us.

Enclosure - Attachment cc: U. S. Nuclear Regulatory Commission Region II Atlanta Federal Center 61 Forsyth St., SW, Suite 23T85 Atlanta, Georgia 30303 Mr. J. T. Reece NRC Senior Resident Inspector North Anna Power Station

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Attachment 1 NORTH ANNA UNITS 1 & 2 2004 -10 CFR 50.59 CHANGES, TESTS & EXPERIMENTS Regulatory Evaluation Number: 91-SE-MOD-009 Document Evaluated: DCP-03-135 - RCP SEAL WATER FLOW TRANSMITTER MANIFOLD REPLACEMENT, NAPS - UNIT 2 Brief

Description:

Kerotest 5-valve manifolds replaced with Whitey 3 valve manifolds.

Reason for Change: Kerotest 5-valve manifolds obsolete.

Summary: Whitey 3-valve manifolds exceeded the original design requirements (i.e. purchased Safety-related) and were therefore more reliable. They were seismically qualified in accordance with Engineering Calc. SEO-1 655, and will be seismically mounted. Their function will not change and the associated instrumentation will not be affected. Post maintenance testing will verify leak tightness to ensure the possibility for malfunctions or accidents do not exist. The subject replacement manifolds can be isolated and the consequences of a failure will be that the associated transmitter / indicator will be inoperable.

Regulatory Evaluation Number: 00-SE-MOD-11 Document Evaluated: DCP 00-112, LOOP ISOLATION VALVE - REMOVAL OF DISC PRESSURIZATION Brief

Description:

Upper disc pressurization connection valves and flanges removed from reactor coolant loop stop valves.

Reason for Change: Reactor Coolant loop stop valve disc pressurization has been discontinued and valve / flange assemblies are no longer being used.

Summary: Reactor coolant loop stop valves are double disc motor operated gate valves that are only used during refueling outages for maintenance operations. The disc pressurization connections were used to pressurize the internal volume between the discs to reduce seat leakage during loop maintenance activities. A valve and blind flange assembly made-up the disc pressurization connection and are part of the reactor coolant pressure boundary. The disc pressurization connections are no longer used since makeup of the refueling cavity and the capacity of the primary drain transfer tank pumps are sufficient to process the volume of leakage. Removing the valve and flanges and capping the lines will decrease the probability of a leak due to high cycle fatigue failure. Since the modification is to piping components only, the change would not increase the probability of a malfunction or create a different type of equipment malfunction. The change does not affect or change the Operating License or Technical Specifications.

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Regulatory Evaluation Number: 01-SE-OT-04 Document Evaluated: DCP-01 -806 AND 01 -807 - CONTAINMENT INTEGRITY RE-ANALYSIS

- NAPS UNIT 1 AND UNIT 2 Brief

Description:

Incorporate revised LOCA design inputs related to containment initial conditions and heat removal systems into the UFSAR containment integrity accident analysis description Reason for Change: Revised design inputs were developed to provide a complete, more robust accident analysis.

Summary: This evaluation involved revisions to the containment integrity analysis for North Anna Units 1 and 2 as a result of TS Change Request #385, Containment Integrity Re-Analysis. The analysis included LOCA containment integrity and safeguards pumps NPSH analyses with the Stone & Webster Engineering Corporation (SWEC) LOCTIC computer code, which is also the basis for the existing licensing basis containment integrity analyses. The main steam line break containment integrity analysis was also evaluated. The containment response to the design basis LOCA was analyzed with revised design inputs to address findings from internal design basis review teams and items from industry and internal operating experience.

One of the more significant changes is the incorporation of instrumentation uncertainty in areas of the analysis where nominal response had previously been assumed. Some plant instrumentation changes must be made in order to reduce uncertainties to acceptable values.

The new containment analysis basis was documented in technical report NE-1257, Rev.0. The main objective of the re-analyses was to explicitly include all revised design inputs in the LOCTIC simulations.

Regulatory Evaluation Number: 03-SE-MOD-01 Document Evaluated: DCP 02-011 - REACTOR VESSEL HEAD REPLACEMENT - UNIT 2 Brief

Description:

Replacement of North Anna Unit 2 reactor vessel closure head (RVCH) with a similar RVCH designed and fabricated by Framatome (FANP).

Reason for Change: Dominion elected to replace the NA2 RVCH in lieu of repairing it after the presence of boric acid indicated that the RCS pressure boundary had possibly been breached.

Summary: This evaluation endorsed the DCP-02-011 replacement of the Unit 2 reactor vessel closure head (RVCH) with a similar RVCH designed and fabricated by Framatome (FANP) for a French Utility (EDF). It noted that the replacement RVCH:

• Met or exceeded all code requirements for fabrication, installation and testing of this classification of component. FANP Doc. 51-5023183, RV Closure Head Report of Reconciliation, documented that all components met the code and owner's requirements as required by ASME B&PV Code, Section Xl, therefore the replacement RV closure head would not introduce an increase in the frequency of occurrence of an accident previously evaluated in the UFSAR.

• Modifications due to design differences did not affect code requirements for the components, therefore North Anna Unit 2 remained within the current Operating License Page 2 of 5

and did not require revisions to North Anna TS, TS Bases or the Technical Requirements Manual. Similarly, modifications did not change flow rates or temperatures affecting the core design calculational models or core reactivity, therefore, core design calculations were unaffected.

• Dimensional and material differences with the original closure head did not reduce the capability or change the functionality of any SSC as described in the North Anna UFSAR and would have a negligible impact on plant operation according to Westinghouse Document VRA-03-12, Effect of Head Weight Change on (normal / seismic / LOCA) Loads.

• Computer codes were revised in the UFSAR to meet applicable code and owner's requirements, however, the evaluation methodologies remained essentially the same and generated equivalent results. Therefore the activity did not impact the design basis or the NRC accepted safety analysis for North Anna Unit-2.

• Replacement RVCH materials from the French design code RCC-M were determined to be equivalent to ASME Section II material specifications. Similarly, the materials used to fabricate the replacement RVCH will have the appropriate certified material test reports (CMTRs) included in the North Anna Unit-2 Replacement Reactor Vessel Closure Head ASME Data Report.

• New Code Cases used to justify some of the new materials being used in the replacement RVCH were incorporated into the UFSAR.

Regulatory Evaluation Number: 03-SE-MOD-03 Document Evaluated: DCP 02-014 - REACTOR VESSEL HEAD REPLACEMENT - UNIT 1 Brief

Description:

Replacement of North Anna Unit 1 reactor vessel closure head (RVCH) with a similar RVCH designed and fabricated by Framatome (FANP).

Reason for Change: Dominion elected to replace the North Anna Unit 1 RVCH in lieu of repairing it after the presence of boric acid indicated that the RCS pressure boundary had possibly been breached.

Summary: This evaluation endorsed the DCP-02-014 replacement of the Unit 1 reactor vessel closure head (RVCH) with a similar RVCH designed and fabricated by Framatome (FANP) for a French Utility (EDF). It noted that the replacement RVCH:

• Met or exceeded all code requirements for fabrication, installation and testing of this classification of component. FANP Doc. 51-5025362, RV Closure Head Report of Reconciliation, documented that all components met the code and owner's requirements as required by ASME B&PV Code, Section Xl, therefore the replacement RV closure head would not introduce an increase in the frequency of occurrence of an accident previously evaluated in the UFSAR.

• Modifications due to design differences did not affect code requirements for the components, therefore North Anna Unit 1 remained within the current Operating License and did not require revisions to North Anna TS, TS Bases or the Technical Requirements Manual.

Similarly, modifications did not change flow rates or temperatures affecting the core design calculational models or core reactivity, therefore, core design calculations were unaffected.

• Dimensional and material differences with the original closure head did not reduce the Page 3 of 5

capability or change the functionality of any SSC as described in the North Anna UFSAR and would have a negligible impact on plant operation according to Westinghouse Document VRA-03-12, Effect of Head Weight Change on (normal / seismic / LOCA) Loads.

• Computer codes were revised in the UFSAR to meet applicable code and owner's requirements, however, the evaluation methodologies remained essentially the same and generated equivalent results. Therefore the activity did not impact the design basis or the NRC accepted safety analysis for North Anna Unit-1.

• Modifications did not change flow rates or temperatures affecting the core design calculational models or core reactivity, therefore, N1 C17 core design calculations were unaffected.

• Replacement RVCH materials from the French design code RCC-M were determined to be equivalent to ASME Section II material specifications. Similarly, the materials used to fabricate the replacement RVCH will have the appropriate certified material test reports (CMTRs) included in the North Anna Unit-1 Replacement Reactor Vessel Closure Head ASME Data Report.

• New Code Cases used to justify some of the new materials being used in the replacement RVCH were incorporated into the UFSAR.

Regulatory Evaluation Number: 04-SE-OT-01 Document Evaluated: FN 2003-052 - UFSAR SECTION 18.3.2.4, ENVIRONMENTALLY ASSISTED FATIGUE Brief

Description:

Incorporate Reactor Coolant environmental effects in the calculation of Si Accumulator and Charging Line nozzle cumulative usage factors (CUF) for Safety Injection (SI) and Charging Line nozzles connected to the Reactor Coolant System (RCS).

Reason for Change: Incorporate: a) enhanced, NRC-approved, Westinghouse methodology that justified reductions in SI System inspections and b) reference to NUREG/CR 5704 and 6717 requirements.

Summary: This evaluation endorsed revisions to North Anna UFSAR Section 18.3.2.4 & Table 18-1, based on the completion of an engineering evaluation of the potential effect of environmentally assisted fatigue on safety injection accumulator and charging line nozzles. It determined that the change:

• Was limited to aging management documentation identified in UFSAR Chapter 18.

• Precluded the occurrence of a SBLOCA or a LBLOCA by predicting through engineering evaluation the absence of fatigue failure that could adversely affect the integrity of SI and charging line nozzles connected to the RCS System.

• Did not affect the cumulative usage factors (CUF) for the SI accumulator and charging line nozzles developed by Westinghouse [LTR-PAFM-03-30] since they were less than the ASME Code limit, therefore the need for additional, enhanced inspections were unjustified.

• Did not affect plant design or involve a test or experiment.

• Did not affect Plant TSs, although, descriptions provided in the UFSAR were revised to indicate utilization of enhanced methodology for calculating CUFs.

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Regulatory Evaluation Number: 04-SE-OT-02 Document Evaluated: FN-2003-035 - Implementation of the Core Management System (CMS) and Relaxed Power Distribution Control Methodology into UFSAR Chapters 4 & 15 Brief

Description:

UFSAR Sections 4.3,15.1 and 15.3 revised to incorporate reference to Topical Report VEP-FRD-42 Rev 2.1 -A / VEP-NE-1 Rev 0.1 -A and the Studsvik Core Management System (CMS)

Reason for Change: UFSAR revised to be consistent with recently revised Topical Reports:

VEP-FRD-42, Rev. 2.1-A and VEP-NE-1, Rev.0.1 -A.

Summary: This evaluation endorsed revisions to the North Anna Unit 1 & 2 UFSAR Sections 4.3, 15.1 and 15.3. Specifically, it endorsed incorporation of reference to: a) Topical Report VEP-FRD-42 Rev 2.1-A which supported the NRC-approved use of the Studsvik Core Management System (CMS) in the Core reload design methodology, b) VEP-NE-1 Rev 0.1 -A, which supported the NRC-approved use of CMS in the Relaxed Power Distribution Control Methodology and c) Core Operating Limits Report Rev 2, North Anna 1 Cycle 17 Pattern EZ and Core Operating Limits Report Rev 3, North Anna 2 Cycle 16 Pattern TP.

Incorporating the above references did not constitute a departure from the method of evaluation described in the UFSAR since: a) Technical Reports NE-1 328, Rev 0, and NE-1 375, Rev 0, supported the revisions and b) the methodologies described in VEP-FRD-42, Rev. 2.1-A and VEP-NE-1 Rev. 0.1-A were NRC-approved.

Regulatory Evaluation Number: 04-SE-OT-03 Document Evaluated: FN-2004-045 - Increased Fq uncertainty applied to Framatome (AREVA) fuel Brief

Description:

UFSAR Section 4.4 updated to reflect the approved overall peaking factor uncertainty.

Reason for Change: NRC-approved change in methodology for addressing fuel assembly rod bow effects.

Summary: This evaluation endorsed a change to UFSAR Section 4.4, relative to the increased Fq uncertainty applied to Framatome (AREVA) fuel. Specifically, the UFSAR change was part of interim compensatory actions required by a nonconformance in AREVA's approved methodology for addressing fuel assembly rod bow effects. The evaluation confirmed that the activity did not result in a departure from a method of evaluation described in the UFSAR (ISFSI SAR) used in establishing the design bases or in the safety analyses. Additionally, it noted that the NRC approved the use of the increased Fq uncertainty for application to AREVA fuel and directed that the UFSAR be updated to reflect the approved overall peaking factor uncertainty.

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