ML21042B926
| ML21042B926 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 02/11/2021 |
| From: | Mark D. Sartain Virginia Electric & Power Co (VEPCO) |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| 21-033 | |
| Download: ML21042B926 (8) | |
Text
VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261 February 11, 2021 10 CFR 50.55a(z)(1)
United States Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555-0001 VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION UNITS 1 AND 2 Serial No.:
NRA/GDM:
Docket Nos.:
License Nos.:
ASME SECTION XI INSERVICE INSPECTION PROGRAM 21-033 R2 50-280/281 DPR 32/37 REQUEST FOR NRC APPROVAL OF PROPOSED ALTERNATIVE REQUESTS S1-I5-ISl-05 AND S2-I5-ISl-06 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION By letter dated October 22, 2020, Virginia Electric and Power Company (Dominion Energy Virginia) requested NRC approval of proposed alternative requests S1-I5-ISl-05 and S2-I5-ISl-06 for Surry Power Station (Surry) Units 1 and 2, respectively. The proposed alternatives would eliminate the volumetric examination requirement 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," dated March 2016.
On January 19, 2021, the NRC staff requested additional information to facilitate the completion of their technical review.
The requested information is provided in the attachment.
If you have any questions or require additional information, please contact Mr. Gary D. Miller at (804) 273-2771.
Respectfully,
~
Mark D. Sartain Vice President - Nuclear Engineering and Fleet Support Commitments made in this letter: None
Attachment:
Response to NRC Request for Additional Information, Proposed Alternative Requests S 1-I5-ISl-05 and S2-I5-ISl-06
cc:
U.S. Nuclear Regulatory Commission Region II Marquis One Tower 245 Peachtree Center Ave., NE Suite 1200 Atlanta, Georgia 30303-1257 Mr. Vaughn Thomas NRG Project Manager - Surry U. S. Nuclear Regulatory Commission One White Flint North Mail Stop 04 F12 11555 Rockville Pike Rockville, Maryland 20852-2738 Mr. G. Edward Miller Serial No.21-033 Docket Nos. 50-280/281 RAI Response - Proposed Alternative Request Page 2 of 2 NRG Senior Project Manager - North Anna U. S. Nuclear Regulatory Commission One White Flint North Mail Stop 09 E3 11555 Rockville Pike Rockville, Maryland 20852-2738 NRG Senior Resident Inspector Surry Power Station
Attachment Serial No.21-033 Docket Nos. 50-280/281 RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION PROPOSED ALTERNATIVE REQUESTS 51-15-151-05 AND 52-15-151-06 Virginia Electric and Power Company
{Dominion Energy Virginia)
Surry Power Station Units 1 and 2
Serial No.21-033 Docket Nos. 50-280/281 Attachment Response to Request for Additional Information Request for NRC Approval of Proposed Alternatives 51-15-151-05 and 52-15-151-06 Surry Power Station Units 1 and 2 NRC Comment:
Background
By letter dated October 22, 2020, Dominion Energy Virginia, (DEV or the licensee) requested an alternative from the requirements of the American Society of the Mechanical Engineers (ASME) Boiler and Pressure Vessel for Surry Power Station (Surry) Units 1 and 2. The licensee proposed alternatives, S1-l5-ISl-05 and S2-l5-ISl-06, which would eliminate the volumetric examination of the reactor pressure vessel (RPV) threads in flange (i.e., Category B-G-1 examinations) during the ASME Code, Fifth 10-year inservice inspection (ISi) interval at Surry.
Regulatory Basis Title 10 of the Code of Federal Regulations (10 CFR) Part 50.55a(z) establishes a process for licensees to propose alternatives to codes and standard requirements. Specifically, as requested by the licensee, DEV must demonstrate that the proposed alternative would provide an acceptable level of quality and safety. The following requests for additional information (RAls) are needed to reach a conclusion of acceptable level of quality and safety.
NRC Question Number 1 The licensee provided the following statement, in part, under the "Flaw Tolerance Evaluation" section of the submittal.
"As seen from the stress intensity factor (K) calculation documented in Table 6-1 Reference 1 (reproduced in Table 2 above), the maximum K is 19.8 ksi--vfn. The allowable K calculated in Section 6.2.2\\ of the report is 69.6 ksi--vfn, significantly higher than the calculated value. Assuming an RPV flange with 60 studs originally and one inoperable stud, the increase in K is approximately 1.7% resulting in a maximum K of approximately 20.14 ksi--vfn which is still significantly less than the allowable value."
The NRG staff has concern that the above example does not appear to be applicable to Surry. As provided in Table 1 of the submittal, the licensee shows that Surry, Unit 1 and Unit 2, both have 58 studs. There is no mentioning of any inoperable studs for either Unit. The staff relies on a clear understanding of the RPVflange parameters (i.e., number of studs, stud hole diameter, etc.) and any factors affecting operability in order to complete an adequate review.
Page 1 of 5
Please provide clarification regarding the state of the RPV flange:
Serial No.21-033 Docket Nos. 50-280/281 Attachment Clarify whether Surry, Unit 1 and 2, have any inoperable studs. If so, discuss the reason for its inoperability.
Dominion Energy Virginia Response Each Surry Unit 1 and Unit 2 RPV flange contains 58 studs. Ultrasonic examination reports for the most recent ISi (lnservice Inspection) examinations performed on the reactor head closure studs were reviewed for both units. The exams were performed during the spring 2015 for Unit 1 and the fall 2015 for Unit 2. This review confirmed there are no inoperable studs in either unit. The references to the RPV containing 54 or 60 studs or an inoperable stud are input for the bounding stress evaluations performed by EPRI.
Explain why an RPV flange of 60 studs with one inoperable stud is a bounding analysis for Surry, Unit 1 and 2.
Dominion Energy Virginia Response When calculating bounding preload under the "Stress Analysis" section of the Surry Proposed Alternative, the EPRI Ppreload calculation uses the smallest number of studs [54],
smallest stud size [6 in.] and largest Reactor Pressure Vessel (RPV) inside diameter (ID)
[173 in.] for determining the bounding value. The Ppreload calculation also uses the highest internal pressure from the EPRI survey of 2500 psia as seen with pressurized water reactors (PWRs) such as the Surry units. This gives the largest allowable preload for the United States (US) fleet at 42,338 psi.
Ppreload = C. P. ID2 = 1.1
- 2500
- 1732 = 42 338 psi S*D2 54-62 I
Where:
Ppreload = Preload pressure to be applied on modeled bolt (psi)
P
= Internal pressure (psi)
ID
= Largest inside diameter of RPV (in.)
C
= Bolt-up contingencies (+10%)
S
= Least number of studs D
= Smallest stud diameter (in.)
Page 2 of 5
Serial No. 21 -033 Docket Nos. 50-280/281 Attachment The Surry units have a larger number of operable studs [58], equal stud size [6 in.] and smaller RPV ID [154.6 in. ID at bolt hole] and are bounded by the above calculation as shown below:
Ppreload = 1.1 x 2500 psi x (155 in.)2
= 31,642 psi 58 X (6 in.)2 which is less than 42,338 psi.
In the second part of the EPRI analysis, the allowable flaw size based on ASME Code Section XI IWB-3500 is determined based on a worst-case stress analysis. As stated in the EPRI report, the limiting stress intensity factor for the US fleet occurs with the greatest number of bolts [60], largest bolt diameter [7 in.] and largest RPV diameter. The larger and more numerous bolts results in less flange material between bolt holes, and the larger RPV diameter results in higher pressure and thermal stresses. The Surry reactor vessel head flanges have fewer number of bolts [58], smaller bolt diameter and smaller diameters (radius, R) as shown in red in the figure below compared to the model used in the EPRI analysis dimensions shown in black.
Bolt Hole Center Line R95.94" I
R85.7" R86.5" R77.3" 17.0" R83.75" R74.5" R85.69" R78.5" lZ.D" 7.0" 6.0" 1~
14.7" 10.75" Page 3 of 5 8.5"
Serial No.21-033 Docket Nos. 50-280/281 Attachment The EPRI analysis was performed to bound the United States nuclear reactor fleet. The analysis uses the least number of bolts [54] to determine maximum allowable preload and the greatest number of bolts [60] with one inoperable stud to determine maximum allowable stress intensity at the bolt hole. Surry has no inoperable studs; therefore, the stress intensity is more uniform across the bolt holes than assuming one inoperable stud.
As discussed here, the analysis performed and summarized in the EPRI report is bounding for the Surry reactor flanges with [58] bolt holes and no inoperable studs.
The Shearon Harris unit has identical design dimensions as the Surry dimensions provided in Table 1: "Comparison of SPS to Bounding Values Used in Analysis Minimum RPV," of the submittal. The similar Shearon Harris alternative request was approved as documented in ADAMS Accession No. ML17331A086.
The response to Question Number 2 provides plant-specific details for worst-case allowable stress intensity and fracture toughness at a bolt-up temperature lower than the operating temperature, confirming that the maximum predicted stress intensity in the reactor vessel flange area will remain below the limiting plant-specific allowable stress intensity at preload.
NRC Question Number 2 Additional detail is needed regarding the lower bound fracture toughness (Kie) at preload conditions. A critical combination of applied Ki and Kie may occur at lower temperatures, such as the temperature at preload conditions since fracture toughness decreases with temperature, and therefore, the most limiting fracture toughness may exist at preload conditions.
Please provide a plant-specific evaluation that shows that the allowable fracture toughness of the flange is higher than the most limiting Ki value of 17.4 ksi-,Jn, for a preload condition, as provided in Table 2 of the submittal.
Dominion Energy Virginia Response Based on the Surry Technical Specifications limits for Heatup and Cooldown limits, the limiting RT NOT value for the Surry Unit 1 and 2 Reactor Vessel Flange region is 10°F. The Surry Unit 1 and Unit 2 bolt-up temperature is based on an administrative lower limit bolt-up temperature for the Reactor Vessel Head of 60° F. In response to the NRG request, the lower bounding Plain Strain Fracture Toughness, Kie, applicable to the Reactor Vessel Flange region for Surry Units 1 and 2 is determined at 60°F using the reference curve equation provided in ASME Section XI (Article A-4200 and Appendix G, section G-2110, 2004 Edition).
Kie= 33.2+(20.734 )e(.o 2x(T-RT NOT))
Kie = 33.2+(20. 734 )e(.o2x(5o-1 o)) = 89.56 ksivf n Page 4 of 5
Serial No.21-033 Docket Nos. 50-280/281 Attachment Then for a postulated flaw per ASME Section XI, Subsection IWB-3612, the maximum (limiting) stress intensity for normal operating conditions is K, < K1c/-v'10. The limiting stress intensity for the Surry Units 1 and 2 Reactor Vessel Flange region is calculated below.
K, < K,c / -v'10 = 89.56 ksin / -v'10 = 28.32 ksin This result is greater than the maximum preload stress intensity of 17.4 ksin presented in Table 2 of the alternative request and based on EPRI report 3002007626, "Nondestructive Evaluation: Reactor Pressure Vessel Threads in Flange Examination Requirements" (ADAMS Accession No. ML16221A068).
Page 5 of 5