RC-17-0140, Transmittal of 10 CFR 50.59 Biennial Report

Transmittal of 10 CFR 50.59 Biennial Report

ML17297A498
Person / Time
Site:	Summer
Issue date:	10/23/2017
From:	Lippard G South Carolina Electric & Gas Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
LTD-324, RC-17-0140
Download: ML17297A498 (8)

Text

George Lippard Vice President, Nuclear Operations 803.345.4810 A SCANA COMPANY October 23, 2017 Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555

Subject:

VIRGIL C. SUMMER NUCLEAR STATION (VCSNS) UNIT 1 DOCKET NO. 50-395 OPERATING LICENSE NO. NPF-12 10 CFR 50.59 BIENNIAL REPORT South Carolina Electric & Gas Company (SCE&G) hereby submits the Twenty-Fifth VCSNS Report pursuant to 10 CFR 50.59(d)(2).

This report contains a brief description and summary of the evaluations performed to support the changes and modifications made to the facility in accordance with 10 CFR 50.59(c) (Attachment).

This report covers the time frame from October 1, 2015 to September 30, 2017. The report also include one evaluation (2015-0001) that was inadvertently omitted in the previous October 1, 2013 to September 30, 2015 biennial report. CR-17-05367 was written to capture the omission.

If you have any questions or require additional information, please contact Bruce Thompson at (803) 931-5042.

WCM/TDG/ts 0 CFR 50.59 Summary of Evaluations and Changes c: K. B. Marsh S. A. Byrne J. B. Archie N. S. Cams J. H. Hamilton G. J. Lindamood W. M. Cherry C. Haney S. A. Williams NRC Resident Inspector K. M. Sutton NSRC RTS (LTD-324)

File (818.02-8, RR 8450)

PRSF (RC-17-0140)

V. C. Summer Nuclear Station

• P. 0. Box 88

• Jenkinsville, South Carolina

• 29065

• F (803) 941-9776

• www.sceg.com

Document Control Desk Attachment I LTD 324, RR-8450 RC-17-0140 Page 1 of 7 10 CFR 50.59 Summary of Changes and Evaluations 50.59 Log Parent Document Change Description Evaluation Summary No.

2013-0005 ECR-71781, ECR-71781 supports removal of The 50.59 Applicability Determination (Revision 1) RMA-2 Guidance Technical Specification Relocation concluded that the proposed changes (TSR) TSR-1069 operability criteria for ECR-71781 required a 10CFR50.59 for RMA-2 based on definition of review. All 50.59 Screen questions were Technical Specification Task Force answered NO except Question 111.4 (TSTF) -513. Operability guidance relating to revising a FSAR described for RMA-2 has been developed evaluation methodology. The full under design calculation DC00030- 10CRF50.59 evaluation, however, 058, Revision 1. The calculation concluded that the revisions do not serves to: represent a departure from a method of evaluation described in the FSAR since

1. More formally document their use is conservative, thus leading to the FSAR evaluations for the overall conclusion that the proposed the buildup of reactor changes for ECR-71781 can be building activity due to a implemented without prior NRC one gpm leak using the approval.

licensing/design basis methods and RCS source terms.

2. Determine limits on the alarm setpoints that ensure that the TSTF definition of operability is met when using licensing /design basis assumptions.

ECR-71781 also updates the FSAR (FSAR Revision Notice (RN) 13-11) as appropriate, and enters DC00030-058 into records.

Revision 1 of this 50.59 Evaluation corrects a typographical error involving the incorrect ECR number (ECR-71825) being cited, as identified in CR-16-05762. The correct number is ECR-71781.

Affected pages of the evaluation are corrected. The 50.59 Screen does not contain this error.

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2015-0001 ECR-50567S, The condition of concern was a The 50.59 Screen reviewed the scope XVB03107A/B possible water hammer when the of the proposed design change (ECR Closing Time reactor building cooling units 505675) and concluded most elements Watchdog (RBCUs) were aligned with the screen-out. The discussion found in Permissive service water (SW) system for Question 3 concluded two elements of cooling and a Safety Injection the proposed change should be concurrent with a Loss of Offsite screened-in for consideration in a 50.59 Power (SI + LOOP) was to occur. Evaluation:

Because of the high elevation of the RBCUs and XVB03107A/B- (1) Replacing a manual action with an SW, an SI + LOOP, while SW is automatic action. The present design aligned to the RBCUs causes a requires Operations to recognize a slow gravity drain of the SW piping. closure of XVB03107A/B-SW and When the SW booster pumps manually place the MCB Start/Stop (SWBPs) is sequenced on, the switch for Service Water Booster Pump subsequent water flow through the (XPP0045A/B) in "pull to lock" before it piping will rapidly collapse the is sequenced on the 1 E busses to vacuum bubble potentially causing prevent a possible water hammer of the a water hammer. Analysis has RBCU SW pipe. This manual action shown that the resulting pipe was replaced with automatic circuitry to stresses exceed the ASME Code lockout XPP0045AIB when allowables. However, the analysis XVB03107A/B-SW is slow to close.

shows that the forces are not severe enough to result in pipe (2) Adding a manual action. The rupture or containment breach. proposed design change required Therefore, the purpose of this Operations to initialize the water activity is to regain ASME Code hammer automatic protective feature in margin for the SW piping the XPP0045A/B breaker control circuit.

associated with the RBCUs to This new manual action is required protect against a possible water anytime the control circuit is returned to hammer resulting from a service.

postulated slow closure of XVB03107A/B-SW. The proposed activity added a manual action and replaces a manual action The previous design attempted to with an automatic action.

address the delayed closure of XVB03107A/B-SW via procedural Each of the eight evaluation questions controls. This approach required a were answered "No," concluding the manual operator action, which has proposed activity can be implemented proven to be ineffective. The as designed without prior NRC approval proposed activity modifies the per 10CFR50.59.

control circuit of the SWBP breaker by interlocking it with XVB03107A/B-SW to prevent SWBP start if the valve is not in the fully closed position in less than 10+/-1 seconds. Analysis has shown that as long as XVB03107A/B-SW

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closes within 11 seconds, the resulting water hammer forces do not exceed ASME Code stress allowables.

To preserve the ASME Code margin for the Service Water (SW) pipe, this activity modified the start circuit of the SW Booster Pump (XPP0045A/B) to inhibit an engineered safety features loading sequencer in the event XVB03107A/B-SW experience a postulated slow closure in the presence of an SI signal concurrent with a LOOP.

2015-0004 ECR-50846K, Weld Perform an overlay weld repair as Since the WCAP process to be used is Repair Contingency detailed in WCAP-15987-P a deviation from the usual ASME XI for RV Head Revision 2-P-A "Technical Basis methodology of repairing these flaws, Inspections (RF-22) for Embedded Flaw for Repair of the 50.59 screening question Reactor Vessel Head Nozzles." VC concerning methods of analysis Summer performed reactor vessel (Question 4) is answered "Yes". This head inspections during Refuel different process method required an Outage 22. evaluation. The answers to all other screening questions are "No". The During examination, Primary Water method of welding described in WCAP-Stress Corrosion Cracking 15987-P Revision 2-P-A has been indications were found in the approved for use at Westinghouse Control Rod Drive Mechanism plants per a NRC SER approved in (CRDM) penetrations. These flaws December, 2003, provided that the were repaired prior to entering plant fulfills the criteria for use as Mode 5. defined in the WCAP.

There are 2 applicable criteria for a plant to use this WCAP.

1. The plant must be of Westinghouse or Combustion Engineering design.

a. VC Summer is a 3-loop Westinghouse NSSS Reactor and thus meets this condition.

2. Failure Effects Analysis of the

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found flaws must support the ability to use the Westinghouse repair process.

a. This analysis shall be completed before entering Mode 5.

Analysis shall support the use of WCAP-15987-P Revision P-A for repair. This analysis has been completed and is documented in WCAP-17758-NP "Technical Basis for Westinghouse Embedded Flaw Repair for V.C. Summer Unit 1 Reactor Vessel Head Penetration Nozzles and Attachment Welds."

Because VC Summer met these conditions and the repair method has been approved by NRC, the proposed repair activity was implemented without obtaining a License Amendment. This is consistent with the Relief Request 4-05, "Alternative Weld Repair for Reactor Vessel Upper Head Penetrations" approved by the NRC for the RF-21 repairs. The WCAP repair process bounds flaws found in the J-groove weld, the CRDM Tube outside diameter (OD), and the J-weld to CRDM Tube OD interface.

2016-0001 ECR-50695E, EFW Due to low Emergency Feedwater The evaluation of the proposed Flow Margin (EFW) pump margins (currently changes find that this design and Improvement 1%-2%), ECR50695E modified the licensing change can be implemented EFW system to gain EFW Pump without prior NRC approval. There is no margins by installing flow-limiting increase in the frequency of occurrence Venturis, automatic recirculation- of FSAR evaluated events. There is flow control (ARC) valves for motor less than a minimal increase in the driven (MD) pumps and by likelihood of structures, systems and increasing the operating speed of components malfunctions as evaluated the turbine driven (TD) pump. in the FSAR: due to the new ARC

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ECR50695E altered the EFW valves and now crediting any two EFW system and add both passive and pumps for FLB, CDF increased by 7E-8 active components, which and LERF by 3E-9. These increases eliminated the need for EFW flow reflect the trade-off of the current isolation to the faulted steam reliance on the EFW flow control valves generator (S/G) during/following a (FCV) to promptly close on FLB, as postulated Feed Line Break (FLB). opposed to the FCVs not needing to Reliance on manual operator close on FLB, but two EFW pumps action to isolate EFW to the faulted must run and the associated new S/G for containment integrity and component, MD ARC valve, must for satisfactory Environmental perform its active safety function of Quality zone conditions for high automatically closing-off its miniflow energy line breaks outside path. These changes are deemed to containment will remain as constitute less than a minimal increase currently described in the FSAR in the likelihood of a malfunction of the (10.4.9.3). The overall result of the EFW system, due to the small modification was to gain nominal CDF/LERF changes. There is no pump-head margin of increase in the consequences of FSAR approximately 7% for MD pumps evaluated events. The evaluation finds and 14% for the TD pump, which that there is no potential for creating a meets the goal of gaining at least new type of event/accident. The 5% pump-head margin. evaluation finds that there is no impact on fission product barriers, as EFW flows are met for all Chapter 15 events, containment pressure and temperature analyses result in lower peak pressure and no change in peak temperature.

Also, no new analytical methods were utilized in the design efforts for this proposed modification.

2016-0002 ECR-50585U, The following two portions of the The modifications made as a result of Safety Related "B" and "C" Chiller Replacement this activity conform fully to the current Chiller under ECR-50585W and X were licensing basis for the plant. All Replacement screened IN as requiring an functions described in FSAR section Activities - A Evaluation per 10CFR50.59: 9.4.7.2.4 were performed identically by Chiller the replacement chillers.

Replacement A. The elimination of a manual action to throttle of Service Water The redesign of the new chillers to to the chillers, which was operate down to a lower load eliminates previously performed manually to the need for operator action to throttle prevent the chillers from tripping Service Water. The elimination of this under conditions of low Service manual action to throttle is a slight Water temperature and low load, change, but is an improvement, and is was made unnecessary by the acceptable under the current licensing design of the replacement chillers. basis.

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B. The use of a Triconex digital The previous chiller controls use analog Class 1E control system on the controllers and relay logics, whereas, replacement chillers. following this modification, the control system for each of these chillers consist of a Triconex PLC Class 1E digital controller. The detailed evaluation was performed in accordance with the NRC-endorsed NEI 01-01 (TR-102348) guidelines, demonstrates that:

a) Triconex digital control system Topical Report 7286-545-1-A has been accepted by NRC in its SER dated December 12, 2001 for safety-related use in nuclear power plants; b) All plant-specific conditions of approval specified in the NRC SER have been satisfied for the replacement chiller application, as discussed in detail in Appendix 2A; c) The potential for software common mode failure has been carefully considered. The Triconex plans, procedures, QA and V&V for the Triconex operating system software are extensive and robust, and have been approved in the NRC SER. The application software has been developed by NLI under its plans, procedures, QA and V&V, and tested extensively at the factory. The Triconex PLC hardware and software is assigned a Software Integrity Level of 4 due to the PLC being used in a safety related application. NLI developed the user code per IEEE1012 guidance for the design, development, testing and verification of safety related software to ensure the quality of the user generated software. On this basis, the potential for a software common mode failure that could disable all chillers is concluded to be highly unlikely, no more likely

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than the potential for common cause failures in the original chiller equipment; and, d) This activity was performed under 10CFR50.59.

2017-0001 ECR-50897, This project installs the In accordance with Section 5.2.3 of the Replacement replacement reactor vessel closure 10CFR50.59 Resource Manual Reactor Service head identified in ECR-50868 (Revision 3), "If only the question Structure- along with design/analysis, concerning methods of analysis is Integrated Head fabrication and installation of the answered "yes", then only the eighth Assembly (IHA) IHA. The activity that is being question in the 10 CFR50.59 addressed under this 10CFR50.59 evaluation, which concerns a departure Evaluation is limited to the change from a method of analysis, is required to in a single element of the analysis be answered when performing the methodology associated with 10CFR50.59 evaluation." The evaluating and demonstrating the elimination of Missile 5 as a change in design basis function of the missile an element of an FSAR described shields, as well as, a change in evaluation methodology was assessed FSAR described methodology for to determine that there is no departure response spectra analysis of NSSS from an existing FSAR described equipment. methodology. The analysis conclusions of Chapter 15.4.6 for a rupture of a control rod drive mechanism housing (rod cluster control assembly ejection) is unchanged.

The change in FSAR described evaluation methodology for seismic response spectra analysis used for the IHA components was assessed to determine that the change provides conservative results compared to the existing described analysis and the methodology utilized has been previously approved by the NRC for this intended use. Therefore, there is no departure from a FSAR described methodology. The activities included within the scope of this evaluation are determined as permitted to be

'performed under 10CFR50.59 without obtaining a License Amendment or prior NRC approval.