ML17171A230

Response to Request for Additional Information ASME Section XI Inservice Inspection Program Request for Proposed Alternative N1-I4-009 and N2-I4-NDE-004

ML17171A230
Person / Time
Site:	North Anna
Issue date:	06/14/2017
From:	Mark D. Sartain Virginia Electric & Power Co (VEPCO)
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
17-211
Download: ML17171A230 (6)

Text

VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261 June 14, 2017 U. S. Nuclear Regulatory Commission Serial No.17-211 Attention: Document Control Desk NRA/DEA RO Washington, DC 20555-0001 Docket Nos.: 50-338 50-339 License Nos.: NPF-4 NPF-7 VIRGINIA ELECTRIC AND POWER COMPANY NORTH ANNA POWER STATION UNITS 1 AND 2 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION ASME SECTION XI INSERVICE INSPECTION PROGRAM REQUEST FOR PROPOSED ALTERNATIVE N1-14-NDE-009 AND N2-14-NDE-004 In a November 30, 2016 letter (Serial No.16-280), Virginia Electric and Power Company submitted a request for Nuclear Regulatory Commission (NRC) approval of inservice inspection (ISi) alternative N1-14-NDE-009 and N2-14-NDE-004. Specifically, Virginia Electric and Power Company requested to eliminate the ASME Category B-G-1, Item 86.40, Pressure retaining bolting greater than 2-inches, Reactor Vessel - Threads in Flange volumetric examination, in accordance with an industry initiative analyzed in Electric Power Research Institute (EPRI) Report #3002007626, "Nondestructive Evaluation: Reactor Pressure Vessel Threads in Flange Examination Requirements,"

March 2016.

In an email dated May 10, 2017, the Nuclear Regulatory Commission (NRC) transmitted a request for additional information (RAI) related to inservice inspection (ISi) alternative N1-14-NDE-009 and N2-14-NDE-004. The response to the RAI is provided in .

Should you have any questions in regard to this submittal, please contact Ms. Diane E.

Aitken at (804) 273-2694.

Sincerely, i

Mark D. Sartain Vice President - Nuclear Engineering and Fleet Support Commitments made in this letter: None

Serial No.17-211 Docket Nos.: 50-338/339 Response to RAI - Proposed ISi Alternative N1-14-NDE-009 & N2-14-NDE-004 Page 2of2

Attachment:

1. Response to Request for Additional Information for Proposed lnservice Alternative N1-14-NDE-009 and Proposed lnservice Alternative N2-14-NDE-004 Commitments made in this letter: None.

cc: U.S. Nuclear Regulatory Commission Region II Marquis One Tower 245 Peachtree Center Ave., NE, Suite 1200 Atlanta, GA 30303-1257 Ms. K. R. Cotton-Gross NRC Project Manager U.S. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike Mail Stop 08 G-9A Rockville, MD 20852-2738 Mr. James R. Hall NRC Senior Project Manager U.S. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike Mail Stop 08 G-9A Rockville, MD 20852-2738 Mr. J. E. Reasor, Jr.

Old Dominion Electric Cooperative Innsbrook Corporate Center 4201 Dominion Blvd.

Suite 300 Glen Allen, VA 23060 NRC Senior Resident Inspector North Anna Power Station

Serial No.17-211 Docket Nos.: 50-338/339 Page 1 of 4 ATTACHMENT 1 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION PROPOSED INSERVICE INSPECTION ALTERNATIVE N1-14-NDE-009 PROPOSED INSERVICE INSPECTION ALTERNATIVE N2-14-NDE-004 Virginia Electric and Power Company North Anna Power Station Units 1 and 2

Serial No.17-211 Attachment 1 Response to RAI - Proposed ISi Alternative N1-14-NDE-009 & N2-14-NDE-004 Page 2 of 4 Response to Request for Additional Information (RAil Proposed lnservice Alternative N1-14-NDE-009 Proposed lnservice Alternative N2-14-NDE-004

RAI 1

Section 6 of Attachment 1 to the November 16, 2016, submittal states that the proposed alternative is applicable to the fourth ten-year /SI intervals for NAPS 1 and NAPS 2.

However, the second sentence of the submittal Jetter states: "The request is to eliminate the ASME Category B-G-1 , Item 86.40, Pressure retaining bolting greater than 2-inches, Reactor Vessel - Threads in Flange volumetric examination ... " The NRG staff requests the licensee to confirm that the request is to eliminate the volumetric examination onlv for the fourth ten-vear /SI intervals of NAPS 1 and NAPS 2, as stated in Section 6 of Attachment 1 to the submittal.

RAI RESPONSE:

Proposed inservice inspection (ISi) alternative N1-14-NDE-009 and N2-14-NDE-004 to eliminate the volumetric examination is only for the fourth ten-year ISi intervals of North Anna Power Station (NAPS) Unit 1 and NAPS Unit 2, as stated in Section 6 of to letter dated November 30, 2016 submittal that requested approval of the proposed ISi alternative.

RAl2:

The licensee stated on page 3 of Attachment 1 to the submittal that the bolt/stud preload stress was calculated as detailed in the NAPS 1 and NAPS 2 RPV manual.

The licensee showed on the same page the preload equation and the resulting bolt/stud preload stress of 42,338 pounds per square inch (psi). This preload stress value is based on the bounding values shown in Table 1 of Attachment 1 to the submittal. The NRG staff requests the licensee to confirm that the actual bolt/stud preload stress applied to the NAPS 1 and NAPS 2 RPV bolt/studs is less than or equal to 42,338 psi.

Serial No.17-211 Attachment 1 Response to RAI - Proposed ISi Alternative N1-14-NDE-009 & N2-14-NDE-004 Page 3of4 RAI RESPONSE:

Based on the North Anna Unit 1 and Unit 2 Replacement Reactor Vessel Closure Head Stress and Fatigue Analysis Reports, the preload stress in the Reactor Pressure Vessel (RPV) Head closure studs is calculated to be 37.2 ksi. This preload stress value represents the average tensile stress based on -uniformly tensioned studs and is less than the bounding bolt/stud preload stress of 42.3 ksi, calculated in the EPRI Report used as the basis for the original proposed ISi alternative request.

It is noted that stud elongation tolerances are not specifically addressed in the applicable North Anna stress reports, and that a 10% bolt-up contingency factor is used to calculate the bounding bolt/stud preload stress in the EPRI Report. However, adequate margin for elongation tolerances exist such that if this contingency factor were applied, a stud preload stress of (1.1 )x(37.2 ksi) = 40.9 ksi would result, which is less than the bounding bolt/stud preload stress of 42.3 ksi calculated in the EPRI Report.

RAl3:

Table 2 of Attachment 1 to the submittal shows values of applied stress intensity factor (K1) for two load cases, "Preload" (occurs at the temperature the bolt preload is applied) and "Preload + Heatup +Pressure" (occurs at high or operating temperature). However, the licensee provided a comparison of K1with the allowable value (K1d.Y10) only for the "Preload + Heatup + Pressure" case. K1e is defined to be the material fracture toughness of the RPV flange that contains the bolt hole threads. The NRG staff observes that the "Preload" case could be more limiting than the "Preload + Heatup +

Pressure" case because: (1) it expects the value of K1e to be lower at the temperature the bolt preload is applied, and (2) most of the applied K1 comes from the "Preload" case. Therefore, the NRG staff requests the licensee to provide a comparison of applied K1 with K1d.Y10 for the "Preload" case for the NAPS 1 and NAPS 2 threads in RPVflange.

RAI RESPONSE:

A review of material data for Reactor Vessels at North Anna Units 1 & 2 concluded that the applicable RT NDT for areas of the reactor vessel flange is -22°F. Based on the American Society of Mechanical Engineers (ASME)Section XI 2004 Ed. fracture toughness curves and the station procedural limitations on stud tensioning being performed at a minimum of 60°F, the Plane Strain Fracture Toughness, K1c , for the "Preload" case is given as:

Kie= 33.2+(20.734)e<0*02xcso-<-22>>> = 140.087, ksiv'in

Serial No.17-211 Attachment 1 Response to RAI - Proposed ISi Alternative N1-14-NDE-009 & N2-14-NDE-004 Page 4of4

~ f Per ASME Section IWB-3612, the maximum allowable stress intensity is 44.27 ksi~in (K1 < Kic/~10 = 140/~10). This provides adequate margin to the maximum preload stress intensity factor of 17.4 ksi~in, calculated in EPRI Report 3002007626, "Nondestructive Evaluation: Reactor Pressure* Vessel Threads in Flange Examination Requirements," which was used as the basis for the original alternative request.

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