ML13056A413
| ML13056A413 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 02/15/2013 |
| From: | Lane N Dominion, Virginia Electric & Power Co (VEPCO) |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| 13-055 | |
| Download: ML13056A413 (9) | |
Text
VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261 February 15, 2013 United States Nuclear Regulatory Commission Serial No.13-055 Attention: Document Control Desk SPS-LIC/CGL RO Washington, D.C. 20555 Docket Nos. 50-280/281 License No. DPR-32/37 VIRGINIA ELECTRIC AND POWER COMPANY (DOMINION)
SURRY POWER STATION UNITS I AND 2 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION RELATED TO ASME SECTION XI INSERVICE INSPECTION (ISI) PROGRAM REQUEST FOR ALTERNATIVE - IMPLEMENTATION OF EXTENDED REACTOR VESSEL INSERVICE INSPECTION INTERVAL RELIEF REQUESTS CMP-007 AND CMP-009 By a letter dated April 25, 2012 (Serial No.12-267), Virginia Electric and Power Company (Dominion) submitted Relief Requests CMP-007 and CMP-009 for Surry Units 1 and 2, respectively. These relief requests proposed an alternative to the requirement of IWB-2412, Inspection Program B, which requires examination of identified reactor vessel (RV) pressure retaining welds once each ten year interval. Pursuant to 10 CFR 50.55a(a)(3)(i), an alternate inspection interval of 20 years was requested.
On January 17, 2013, the NRC requested additional information regarding Relief Requests CMP-007 and CMP-009. The response to the request for additional information is provided in the attachment.
As indicated in our April 25, 2012 letter, Dominion requests approval of Relief Requests CMP-007 and CMP-009 by May 30, 2013 to support performance of the fourth 10-year ISI interval examination of the Surry Units 1 and 2 RVs in 2023 and 2024 in lieu of 2013 and 2014, respectively.
If you have any questions or require additional information, please contact Mrs. Candee G. Lovett at (757) 365-2178.
Sincerely, N. L. Lane Site Vice President - Surry Power Station
Attachment:
Response to Request for Additional Information Regarding Relief Requests CMP-007 for Surry Unit 1 and CMP-009 for Surry Unit 2 Commitments made by this letter: None
Serial No.13-055 Docket Nos. 50-280/281 Page 2 of 2 cc: U.S. Nuclear Regulatory Commission, Region II Marquis One Tower 245 Peachtree Center Avenue NE, Suite 1200 Atlanta, Georgia 30303-1257 State Health Commissioner Virginia Department of Health James Madison Building - 7th floor 109 Governor Street, Suite 730 Richmond, Virginia 23219 NRC Senior Resident Inspector Surry Power Station Ms. K. R. Cotton, NRC Project Manager - Surry U. S. Nuclear Regulatory Commission One White Flint North Mail Stop 08 G9A 11555 Rockville Pike Rockville, Maryland 20852 Dr. V. Sreenivas, NRC Project Manager - North Anna U. S. Nuclear Regulatory Commission One White Flint North Mail Stop 08 G9A 11555 Rockville Pike Rockville, Maryland 20852 Mr. R. A. Smith Authorized Nuclear Inspector Surry Power Station
Serial No.13-055 Docket Nos. 50-280/281 Attachment Response to Request for Additional Information Regarding Relief Requests CMP-007 and CMP-009 Virginia Electric and Power Company (Dominion)
Surry Power Station Units I and 2
Serial No.13-055 Docket Nos. 50-280/281 Attachment Page 1 of 6 REQUEST FOR ADDITIONAL INFORMATION SURRY POWER STATION, UNITS NO. I AND 2 INSERVICE INSPECTION RELIEF REQUEST CMP-007 AND CMP-009 REACTOR VESSEL BELTLINE WELD EXAM EXTENSION (TAC NOs. ME8573 AND ME8574)
In accordance with 10 CFR 50.55a(a)(3)(i) and the NRC safety evaluation approving the use of WCAP-16168-NP-A, Revision 2 (WCAP), the staff needs the following information to complete the review:
RAI 1
NUREG-1 874 and WCAP-1 6168 were developed to consider the structural integrity of PWR reactor vessels, taking into account the plates, axial welds, circumferential welds, and forgings that make up the potentially limiting materials for the structural integrity of the reactor vessel.
The staff notes that the RVID2 database includes a nozzle shell forging, heat
- 122V109VA1, for Surry, Unit 1 and another, heat #123V303VA1, for Unit 2. These forgings and their associated welds do not appear in Table 3 of Attachments 1 and 2.
The staff requests that the licensee provide a rollout diagram of the potentially limiting reactor vessel materials (those exposed to a neutron fluence of greater than 1E+17 n/cm 2 for 48 EFPY) showing the planar dimensions and thickness of each plate/weld/forging for each unit along with representative fluence maps. With the use of the rollout diagram and fluence map, justify why the nozzle shell forgings and their associated circumferential welds are not included to represent the structural integrity of Surry, Units 1 and 2 in Table 3 of Attachments 1 and 2.
Response to RAI 1:
The nozzle shell forging, heats #122V109VA1 and #123V303VA1 for Surry Units 1 and 2, respectively, and the adjacent nozzle-to-intermediate shell circumferential welds were not originally included in the evaluation as these materials are not immediately adjacent to the active core in both units. However, due to the close proximity of these materials in both units to the top of the active core, Dominion agrees that the nozzle shell forgings, heat
- 122V109VA1 and #123V303VA1 for Surry Units 1 and 2, respectively, and the adjacent nozzle-to-intermediate shell circumferential welds should be included as part of the RV weld 10-year ISI interval extension evaluation. These materials have been included in previous reactor vessel integrity analyses for Surry Units 1 and 2. The tables below are updated versions of Table 3 provided in the April 25, 2013 letter for Surry Units 1 and 2.
Serial No.13-055 Docket Nos. 50-280/281 Attachment Page 2 of 6 The through-wall cracking frequency (TWCF) for the nozzle shell forgings (TWCF 95.FO) was calculated for each unit. These results were added to the total TWCF (TWCF95-TOTAL). For Surry Unit 1, the magnitude of TWCF 95-FO was several orders of magnitude below TVWCF95-TOTAL and, therefore, did not change the previously reported TWCF value. Thus, TWCF95.TOTAL for Surry Unit 1 remains 1.41E-08 events per year. For Surry Unit 2, the magnitude of TWCFg95 FO was similar to that calculated for Unit 1; however, TVVCF95-TOTAL for Unit 2 was also of a similar magnitude. Therefore, the Surry Unit 2 TWCF95-TOTAL has been updated and increased from 1.17E-12 to 1.28E-12 events per year.
The nozzle to intermediate shell circumferential welds were also incorporated into the evaluation; however, these welds were not the controlling circumferential weld material for either unit and, therefore, inclusion of these welds did not change the TWCF for circumferential welds (TWCF 95.cw) for Surry Units 1 and 2.
Regarding the NRC's request for a rollout diagram of the potentially limiting reactor vessel materials, Dominion does not possess the information required to define the region of the RV with a neutron fluence of greater than 1E+17 n/cm2 at end-of-license, and additional analyses would be required to determine the extent of the fluence region within the RV that is greater than 1E+17 n/cm 2 at end-of'license. The RAI indicates that the purpose for this diagram would be to provide justification as to why the nozzle shell forgings and their associated circumferential welds were not included in this evaluation. Inclusion of these materials in evaluation and updated results are being provided (versus a rollout diagram) in response to RAI 1.
In conclusion, the total TWCFs for the Surry Units 1 and 2 reactor vessels, including the nozzle shell forgings and circumferential welds, remain bounded by the WCAP-16168-NP TWCF acceptance criteria of 1.76E-08 events per year for Westinghouse plants.
Serial No.13-055 Docket Nos. 50-280/281 Attachment Page 3 of 6 Table 3:
Details of TWCF Calculation for Surry Unit I at 48 Effective Full Power Years (EFPY)
Inputs Nozzle Shell Twa11 [inches]: 9.283 N/A Reactor Coolant System Temperature, TRCS[°F]:
Beltline Twa,1 [inches]: 8.209 Fluence No Region and Material Heat CuM1 ) Ni(i) R.G. CF11) RTNDT(u) [1019 Component No. [w4%l [wt%] 1.99 [OF] [OF] Neutron/cm 2 Description Pos. Ee> 1.0 MeV]
1 Inter. Shell Long. SA-1494/8T1554 0.16 0.57 1.1 143.9 -5 0.897 Weld L3 2 Inter. Shell Long. SA-1494/8T1554 0.16 0.57 1.1 143.9 -5 0.897 Weld L4 3 Lower Shell Long. SA-1494/8T1554 0.16 0.57 1.1 143.9 -5 0.897 Weld Li 4 Lower Shell Long. SA-1526/299L44 0.34 0.68 1.1 220.6 -7 0.897 Weld L2 Nozzle To Inter. Shell 5__ Circ. WeldW06 J726/25017 0.33 0.10 1.1 152.0 0 0.775 6 Inter. To Lower Shell SA-1585/72445 0.22 0.54 1.1 158.0 -5 4.51 Circ. Weld W05 SA-1650/72445 7 Nozzle Shell Forging 122V109VA1 0.11 0.74 1.1 76.1 40 0.775 8 Intermediate Shell C4326-1 0.11 0.55 1.1 73.5 10 4.51 9 Intermediate Shell C4326-2 0.11 0.55 1.1 73.5 0 4.51 10 Lower Shell C4415-1 0.11 0.50 2.1 85.0 20 4.51 11 Lower Shell C4415-2 0.11 0.50 1.1 73.0 0 4.51 Outputs Methodology Used to Calculate AT 30 : Regulatory Guide 1.99, Revision 2(2)
Controlling Fluence Material Region No. RTMAx- [1019 FF
(~~[09Flenc AT30o WF 5 X (RegonmNo. xx [R] Neutron/cm 2, (Flue [OF] TWCF9.xx (FromE > 1.0 MeV] Factor)
Above)
Limiting Axial Weld - AW 4 666.55 0.897 0.970 213.88 6.26E-09 Limiting Plate - PL 10 597.10 4.51 1.382 117.43 1.51E-12 Limiting Forging - FO 7 570.33 0.775 0.928 70.66 1.60E-13 Circumferential Weld- 6 672.96 4.51 1.382 218.29 6.05E-13 CW I__II_
TWCFg5.TOTAL(1AwTWCF9S.AW + gPL + aFoTWCF95-FO + acwTWCF 95cw):
aPLTWCF 95 1.41 E-08 (1) WCAP-15130, Revision 1 (2) NRC Regulatory Guide 1.99, Revision 2
Serial No.13-055 Docket Nos. 50-280/281 Attachment Page 4 of 6 Table 3:
Details of TWCF Calculation for Surry Unit 2 at 48 Effective Full Power Years (EFPY)
Inputs Nozzle Shell Twa11 [inches]: 9.283 8.209 Reactor Coolant System Temperature, TRcs[°F]: N/A N eztle Twall[inches]:
Beltline Twa,, [inches]: 8.209 Fluence No Region and Material Heat CuM1) Ni( 1) R.G. CF(1) RTNDT(u) [1019 Component No. [wt%] [wt%] 1.99 [OF] (l)[OFI Neutron/cm 2 Description Pos. E > 1.0 MeV]
1 Inter. Shell Long. SA-1585/72445 0.22 0.54 1.1 158.0 -5 0.940 2 Inter.Weld ShellL3Long. 02 2 Weld L4(2) SA-1585/72445 0.22 0.54 1.1 158.0(2) -5 0.940 3 Lower Shell Long. WF-4/8T1762 0.19 0.57 1.1 152.4 -5 0.940 Weld Li 4 Lower Shell Long.(2 LoWeld L2(2) WF-4/8T1762 0.19 0.57 1.1 152.4(2) -5 0.940 5 Nozzle To Inter. Shell J737/4275 0.35 0.10 1.1 160.5 0 0.632 Circ. Weld W06 6 Inter. To Lower Shell R3008/0227 0.19 0.55 1.1 149.3 0 4.50 Circ. Weld W05 7 Nozzle Shell Forging 123V303VA1 0.11 0.72 1.1 75.8 30 0.632 8 Intermediate Shell C4331-2 0.12 0.60 1.1 83.0 -10 4.50 9 Intermediate Shell C4339-2 0.11 0.54 1.1 73.4 -20 4.50 10 Lower Shell C4208-2 0.15 0.55 1.1 107.3 -30 4.50 11 Lower Shell C4339-1 0.11 0.54 1.1 73.4 -10 4.50 Outputs Methodology Used to Calculate AT 30 : Regulatory Guide 1.99, Revision 2(3)
Controlling Fluence Material o 19 FF Region No. [OR] Neutron/cm2 (Fluence [TF] TVCF95 .xX (From [ Netrn/m Factor) [°F]
Above) E> 1.0 MeV]
Limiting Axial Weld - AW 1 and 2 609.93 0.940 0.983 155.26 0.OOE+00 Limiting Plate - PL 10 577.86 4.50 1.381 148.19 3.08E-13 Limiting Forging - FO 7 555.72 0.632 0.871 66.05 4.29E-14 Circumferential Weld - 6 665.87 4.50 1.381 206.20 1.80E-13 CW I I I TVVCF95.TOTAL(cAwTVWCF9 5_AW + aPLTWCF 95.PL + aFOTVVCF95-FO + acwTWCF95.cw): 1.28E-12 (1) WCAP-15130, Revision 1 (2) Weld contains two different materials. The material with the most limiting properties was used for this evaluation.
(3) NRC Regulatory Guide 1.99, Revision 2
Serial No.13-055 Docket Nos. 50-280/281 Attachment Page 5 of 6
RAI 2
Regarding observed indications from the most recent ( 3 rd) inservice inspection (ISI) interval examinations for Unit 2 as documented in Table 2 of Proposed Alternative CMP-009, Attachment (2), clearly state the location and size of the four indications that were found in the near-surface region of the reactor pressure vessel beltline area. Were these indications observed in the Is and/or 2 ISI interval inspections? Did the size of the indications change during the course of the three inspections? If there was a change in the size of an indication, can that change be attributed to improved inspection procedures?
Response to RAI 2:
Size and location of the indications found in the Surry Unit 2 RPV beltline region are as follows:
The weld #1-03 (Intermediate to Lower Shell Circ. Weld) centerline elevation is at 231.66" relative to the RV flange, and the weld width is 2.4". The upper/lower limits of the weld at the ID are 230.46" to 232.86". The referenced flaws are circumferentially oriented.
Indication #1 is located in the upper fusion zone of weld #1-03 with the center at a vessel theta position of 218.90. This indication is 0.6" in length, 0.125" in through-wall extent (2a dimension), and is embedded with an 'S' dimension of 0.51" (as measured from the cladding-to-base-metal interface).
Although recorded as in the examination volume of weld #1-03, indication #2 is actually located in weld#1-07, the intersecting Intermediate Long Seam at 2250 with a measured center position at a vessel theta position of 226.50. This indication is 1.1" in length, 0.125" in through-wall extent (2a dimension), and is embedded with an 'S' dimension of 0.78" (as measured from the cladding-to-base-metal interface).
Indication #3 is located in the lower fusion zone of weld #1-03 with the center at a vessel theta position of 181.30. This indication is 1.1" in length, 0.125" in through-wall extent (2a dimension), and is embedded with an 'S' dimension of 0.21" (as measured from the cladding-to-base-metal interface).
Indication #4 is located near the centerline of weld #1-03 with the center at a vessel theta position of 95.70. This indication is 0.6" in length, 0.125" in through-wall extent (2a dimension), and is embedded with an 'S' dimension of 0.78" (as measured from the cladding-to-base-metal interface).
Serial No.13-055 Docket Nos. 50-280/281 Attachment Page 6 of 6 These four indications were observed in the 3 rd ISI interval inspection only, though the characteristics of the flaws do not support concluding the indications to be service-induced.
It is likely that the indications were present but not recorded during the 1 st and 2 nd ISI interval inspections.
The 3 rd ISI interval examination was an Appendix VIII (PDI) examination with a higher sensitivity than either the 1 st or 2 nd IS, interval Section XI/Regulatory Guide 1.150 examinations. For both the 1 st and 2 nd ISI interval examination, calibration was performed with the responses from 0.125" diameter side drilled holes normalized to 80% full screen (DAC). Recording criteria was, for the vessel inner 25% thickness, at 20% DAC or 16% full screen height (FSH). For the 3 rd IS, interval PDI examination, calibration was performed with the responses from 0.063" diameter side drilled holes set to 80% full screen and then increased by 12dB. The recording criteria was to record all valid flaws. The relative gain difference between the two examinations, considering both the difference in calibration reflector size and the addition of 12dB for the PDI examinations, was approximately 18 dB or 8:1. The recorded maximum amplitudes from the 3 rd interval exams of each of the indications in question, along with the probable responses during the 1 st and 2 nd ISI interval examinations, are summarized as follows:
Indication 3rd Interval measured 1 't/2nd Interval probable
- amplitude responses 1 42% FSH -5%FSH 2 75% FSH -9%FSH 3 27% FSH -3%FSH 4 42% FSH -5%FSH Note that all of these probable responses from the 1 st and 2 nd ISI interval examination are well below the recording level of 16%FSH (20%DAC).
In summary, although it is likely that these four indications were present but not recorded during the 1 st and 2 nd ISI interval inspections, the indications were observed only in the 3 rd ISI interval inspection, which was performed with a higher sensitivity examination technique.