NL-18-1181, Relief Requests VEGP-ISI-RR-05 and VEGP-ISI-RR-06, SNC Response to NRC Request for Additional Information

Relief Requests VEGP-ISI-RR-05 and VEGP-ISI-RR-06, SNC Response to NRC Request for Additional Information

ML18290A709
Person / Time
Site:	Vogtle
Issue date:	10/17/2018
From:	Gayheart C Southern Nuclear Operating Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NL-18-1181
Download: ML18290A709 (8)

Text

~ Southern Nuclear Cheryl A. Gayheart Regula lory Alfa1rs 01rector 3535 Colonnade Parkway Birmingham. AL 35243 205 992 5316 tel 205 992 7601 tax OCT 1 7 2018 cagayhea @southernco.com Docket Nos.: 50-424 NL-18-1181 50-425 U.S. Nuclear Regulatory Commission ATIN: Document Control Desk Washington, D. C. 20555-0001 Vogtle Electric Generating Plant- Unit 1 and 2 Relief Requests VEGP-ISI-RR-05 and VEGP-ISI-RR-06 SNC Response to NRC Request for Additional Information Ladies and Gentlemen:

By letter dated May 24, 2018 (Agencywide Documents Access and Management System Accession Number ML18144B012), Southern Nuclear Operating Company (SNC) requested relief from certain requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section XI. These relief requests, VEGP-ISI-RR-05 and VEGP-ISI-RR-06, pertain to examination coverage of Class 2 and Risk Informed welds achieved during the third inservice inspection interval at the Vogtle Electric Generating Plant (VEGP), Units 1 and 2.

By email dated August 30, 2018, the U.S. Nuclear Regulatory Commission (NRC) staff notified SNC that additional information is needed for the staff to complete their review. to this letter provides the SNC response to the NRC request for additional information (RAI).

This letter contains no NRC commitments. If you have any questions, please contact Jamie Coleman at 205.992.6611.

Respectfully submitted, Cheryl A. a eart Director, Regulatory Affairs Southern Nuclear Operating Company

U. S. Nuclear Regulatory Commission NL-18-1181 Page 2 CAG/kgl/sm

Enclosures:

1. SNC Response to NRC Request for Additional Information (RAJ) - Relief Request VEGP-ISI-RR-05

2. SNC Response to NRC Request for Additional Information (RAI)- Relief Request VEGP-ISI-RR-06 cc: Regional Administrator, Region II NRR Project Manager- Vogtle 1 & 2 Senior Resident Inspector- Vogtle 1 & 2 RType: CVC7000

Vogtle Electric Generating Plant- Unit 1 and 2 Relief Requests VEGP-ISI-RR-05 and VEGP-ISI-RR-06 SNC Response to NRC Request for Additional Information Enclosure 1 SNC Response to NRC Request for Additional Information (RAJ)

Relief Request VEGP-ISI-RR-05 to NL-18-1181 SNC Response to NRC Request for Additional Information (RAI)

Relief Request VEGP-ISI-RR-05 NRC RAI1 Table RR-5 lists an achieved coverage of 32.5 percent for the Vogtle, Unit 1, boron injection tank head-to-vessel shell weld. This lack of required coverage appears to be due primarily to ultrasonic examination from only one side of the weld. The reported coverage achieved during the second lSI interval (ADAMS Accession No. ML003762772) for the same component was 57 percent. Please provide a justification for the reduced inspection coverage for this component while all other components have the same or increased inspection coverage.

SNC Response to RAI 1 During the preparation of the initial submittal, the coverage for 11204-V6-001-W02 was reevaluated based upon the datasheets. This revaluation conservatively determined the coverage to be 32.5% using a graphical method based upon a sketch provided in the datasheets. The reevaluation should have used the coverage percentage numbers listed in the datasheets from the 2014 inspection. A secondary review, performed during the RAI processes, determined the actual coverage to be 63.7%.

NRC RAI2 Under Item 6, "Proposed Alternative and Basis for Use," the licensee states that an acceptable level of quality and safety will be maintained; however, no supporting argument was provided to justify that reasonable assurance of structural integrity or leak tightness continues to exist. Please provide the following:

a. A confirmation that, for each affected component weld having less than 61 percent coverage, the peak stress in the ultrasonic testing (UT) missed region is likely to be similar to or less than the peak stress in the UT covered region. This confirmation should be based on stress reports for the affected components.

b. Past plant-specific and fleet operating experience, if available, regarding leakage and UT indications detected from all affected component welds that required flaw evaluations or repairs.

SNC Response to RAI 2

a. For the 32" Steam Generator Main Steam Nozzle, the design peak stress ratio is slightly higher above the area of limited scanning. However, the ratios are of similar magnitude and the representative of the entire component in the inspection region of interest. To give added confidence of structural integrity, a supplemental magnetic particle (MT) examination was performed on these welds. This examination attained 100% coverage. No recordable indications were found with this examination.

For the 6" Boron Injection Tank (BIT) Inlet and Outlet Nozzles, a review of the BIT design report (PJ-750154252) indicated the peak stress intensities on the inside surface of the nozzle are lower in the missed region than the covered region.

E1-1 to NL-18-1181 SNC Response to NRC Request for Additional Information (RAI)

Relief Request VEGP-ISI-RR-05 For the Residual Heat Removal Heat Exchanger, the Westinghouse Design Report 033001 was reviewed. No local stress evaluations were performed for these nozzles since they were within the Westinghouse Engineering Specification (G-679150) allowable values for loading on the nozzles. The actual peak stress on the nozzles is less than 50% of the allowable stress on the nozzles. To give added confidence of structural integrity, a supplemental surface examination (liquid penetrant) examination was performed on these welds. This examination attained 100% coverage. No recordable indications were found with this examination.

b. In reviewing the SNC fleet operating experience, (VEGP - Units 1 and 2 and Farley Nuclear Plant - Units 1 and 2) no leakage or indications that require flaw evaluations or repairs have been found from the Category C-A, Item Nos. C1.20 and Category C-8, Item No. C2.21 for inspections performed during the previous interval.

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Vogtle Electric Generating Plant - Unit 1 and 2 Relief Requests VEGP-ISI-RR-05 and VEGP-ISI-RR-06 SNC Response to NRC Request for Additional Information Enclosure 2 SNC Response to NRC Request for Additional Information (RAI)

Relief Request VEGP-ISI*RR-06 to NL-18-1181 SNC Response to NRC Request for Additional Information (RAI)

Relief Request VEGP-ISI-RR-06 NRC RAI1 Please specify the piping system (e.g., reactor coolant system, residual heat removal, safety injection) to which each weld listed in Table RR-6 belongs.

SNC Response to RAI 1 Weld Identification Weld ID Number Piping_ System 11208-007-1-RB Chemical and Volume Control 11208-007-2-RB Chemical and Volume Control 11201-030-38-RB Reactor Coolant 11201-031-2-RB Reactor Coolant 11201-042-2-RB Reactor Coolant 11201-046-2-RB Reactor Coolant 11201-051-2-RB Reactor Coolant 11204-126-15-RB Safety Injection 11204-024-13-RB Safety Injection 11204-023-21-RB Safety Injection 11204-076-42-RB Safety Injection 11204-078-1 O-RB Safety Injection 11204-021-26-RB Safety Injection 11204-021-27-RB Safety_ Injection 11204-021-28-RB Safety Injection 11204-025-20-RB Safety Injection 11208-012-5-RB Chemical and Volume Control 11301-001-6-RB Main Steam 21208-007-1-RB Chemical and Volume Control 21208-007-2-RB Chemical and Volume Control 21204-124-15-RB Safety_ Injection 21204-024-15-RB Safety Injection 21204-023-21-RB Safety Injection 21201-030-19-RB Safety Injection 21204-021-26-RB Safety Injection 21204-021-17-RB Safety Injection 21204-024-16-RB Safety Injection 21204-025-22-RB Safety_ Injection NRC RAI2 Given the susceptibility of Category R-A, Item No. R1.11, welds to thermal fatigue, the reduced coverage obtained, and for assurance of structural integrity of the unexamined volume of the weld, please provide the available.cumulative fatigue usage factor for the Class 1 welds.

E2-1 to NL-18-1181 SNC Response to NRC Request for Additional Information (RAI)

Relief Request VEGP-ISI-RR-06 SNC Response to RAI 2 The Thermal Fatigue Program tracks cyclic and transient occurrences at key locations via Fatigue Pro to ensure that components are maintained within the design limits and projects fatigue usage to 60 years. The latest Fatigue Pro analysis is Calculation FP-VOG-321, which provides the fatigue usage at the nozzle locations for these lines. The tracked Fatigue Pro nozzle locations that are applicable to the reduced coverage welds are listed in the table below, which summarizes the current cumulative usage.

c umu Iaf1ve Fa1gue f usage Factor T abl e Weld ID Number Line Function  % of Allow Cumulative Usage 11208-007-1-RB Alternate Charging Line 44.8%

11208-007-2-RB Alternate Charging Line 44.8%

11201-030-38-RB Pressurizer Spray Line 16.8%

11201-031-2-RB High Head Sl Loop 1 21.3%

1120 1-042-2-RB High Head Sl Loop 2 21.3%

11201-046-2-RB High Head Sl Loop 3 21.3%

11201-051-2-RB High Head Sl Loop 4 21.3%

21208-007-1-RB Alternate Charging Line 53.6%

21208-007 RB Alternate Charging Line 53.6%

NRC RAI3 Has the system leakage test performed in accordance with IWB-5000 for Class 1 welds, or IWC-5000 for Class 2 welds, identified any through-wall leaks? If so, please discuss the root cause, the corrective actions, and the extent of condition evaluation.

SNC Response to RAI 3 A review was performed on the NIS-1 reports going back to the first refueling outages for Unit 1 and Unit 2. One through wall-through wall leak was identified during the Class 1 pressure test in 2R8 (Unit 2, Refueling Outage 8).

The leak was located on the leak-off line (3/8 inch diameter) of valve 2HV8701 B. Design Change Package (DCP) OO-V2N0017 installed this piping during 2R8. During the pressure test performed for post maintenance testing, the leak was found on the modified piping.

The rest of the piping affected by the DCP were inspected during the Class 1 pressure test with no leakage. In order to correct this leakage condition, the piping installed by the DCP was removed and capped.

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