ML22322A158
| ML22322A158 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 11/17/2022 |
| From: | James Holloway Virginia Electric & Power Co (VEPCO) |
| To: | Office of Nuclear Reactor Regulation, Document Control Desk |
| References | |
| 22-114 | |
| Download: ML22322A158 (1) | |
Text
VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261 November 17, 2022 10 CFR 50.55a(z)(1)
U. S. Nuclear Regulatory Commission Serial No.: 22-114 Attention: Document Control Desk NRA/GDM: RO Washington, DC 20555-0001 Docket Nos.: 50-280/281 License Nos.: DPR-32/37 VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION UNITS 1 AND 2 ALTERNATIVE REQUEST TO DEFER ASME CODE SECTION XI INSERVICE INSPECTION EXAMINATIONS FOR PRESSURIZER AND STEAM GENERA TOR PRESSURE-RETAINING WELDS AND FULL PENETRATION WELDED NOZZLES In accordance with 10 CFR 50.55a, "Codes and Standards," paragraph (z)(1 ), Virginia Electric and Power Company (Dominion Energy Virginia) requests Nuclear Regulatory Commission (NRC) approval of a proposed alternative to the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, inservice inspection (ISi) requirements for Surry Power Station (SPS) Units 1 and 2.
Specifically, Table IWB-2500-1, Examination Category B-B, and Table IWC-2500-1, Examination Category C-A and C-B, component examinations will be deferred:
- 1) through the sixth 10-year ISi interval ending on October 13, 2033, for SPS Unit 1, and
- 2) through the remainder of the third period of the fifth 10-year ISi interval through the sixth 10-year ISi interval ending on May 9, 2034, for SPS Unit 2.
The proposed alternative is requested on the basis that it provides an acceptable level of quality and safety in lieu of the current ASME Code Section XI 10-year inspection frequency requirement.
The proposed alternative, which includes a summary of the technical basis for the request, is provided in Attachment 1. The plant-specific applicability of the technical basis to SPS Units 1 and 2 is provided in Attachments 2 and 3. The SPS Units 1 and 2 pressurizer and steam generators inservice inspection history and the inspection history for the applicable components, as obtained from an industry survey, are presented in Attachments 4 and 5, respectively.
Pursuant to 10 CFR 50.55a(z), the proposed alternative requires NRC review and approval before implementation. Dominion Energy Virginia requests NRC approval of this request by April 1, 2023, to support the SPS Unit 2 Spring 2023 refueling outage.
Serial No.: 22-114 Docket Nos.: 50-280/281 Page 2 of 3 If you have any questions or require additional information, please contact Mr. Gary D.
Miller at (804) 273-2771.
Respectfully,
~i1/2~~
James E. Holloway * .
Vice President - Nuclear Engineering & Fleet Support Commitments made in this letter: None Attachments:
- 1. Proposed Alternative to ASME Code Section XI Requirements for lnservice Inspection of Pressurizer and Steam Generator Pressure-Retaining Welds and Full Penetration Welded Nozzles
- 2. Plant-Specific Applicability
- 3. Comparison of lnsurge/Outsurge Transients
- 4. Inspection History
- 5. Results of Industry Survey
Serial No.: 22-114 Docket Nos.: 50-280/281 Page 3 of 3 cc: Regional Administrator, Region II U.S. Nuclear Regulatory Commission Marquis One Tower 245 Peachtree Center Avenue, NE, Suite 1200 Atlanta, Georgia 30303-1257 Mr. L. John Klos NRC Senior Project Manager- Surry Power Station U.S. Nuclear Regulatory Commission One White Flint North, Mail Stop 09 E-3 11555 Rockville Pike Rockville, Maryland 20852-2738 Mr. G. Edward Miller NRC Senior Project Manager - North Anna U.S. Nuclear Regulatory Commission One White Flint North, Mail Stop 09 E-3 11555 Rockville Pike Rockville, MD 20852-2738 NRC Senior Resident Inspector Surry Power Station
Serial No.22-114 Docket Nos.: 50-280/281 ATTACHMENT 1 Proposed Alternative to ASME Code Section XI Requirements for lnservice Inspection of Pressurizer and Steam Generator Pressure-Retaining Welds and Full Penetration Welded Nozzles SURRY POWER STATION UNITS 1 AND 2 VIRGINIA ELECTRIC AND POWER COMPANY (DOMINION ENERGY VIRGINIA)
Serial No.22-114 Docket Nos.: 50-280/281 Attachment 1 Page 1 of 19 Proposed Alternative to ASME Code Section XI Requirements for lnservice Inspection of Pressurizer and Steam Generator Pressure-Retaining Welds and Full Penetration Welded Nozzles Proposed Alternative In Accordance with 10 CFR 50.55a(z)(1)
-- Acceptable Level of Quality and Safety --
1.0 American Society of Mechanical Engineers (ASME) Code Components Affected The ASME Code components affected are Class 1 and 2 pressurizer (PZR) vessel shell-to-head welds and steam generator (SG) pressure-retaining welds and full penetration welded nozzles listed in Table 1. The affected components are identified in Table 2.
Table 1. ASME Code components affected.
Code Class Class 1 and Class 2 PZR vessel shell-to-head welds Description SG pressure-retaining vessel welds / full penetration welded nozzles Class 1, Category B-B, pressure-retaining welds in vessels other than reactor vessels Examination Class 2, Category C-A, pressure-retaining welds in pressure vessels Categories Class 2, Category C-8, pressure-retaining nozzle welds in vessels B2.11 PZR, shell-to-head welds, circumferential B2.12 PZR, shell-to-head welds, longitudinal B2.40 SG (primary side), tubesheet-to-head weld C1.10 Shell circumferential welds Item Numbers C1.20 Head circumferential welds C1.30 Tubesheet-to-shell weld C2.21 Nozzle-to-shell (nozzle-to-head or nozzle-to-nozzle) welds C2.22 Nozzle inside radius sections
Serial No.22-114 Docket Nos.: 50-280/281 Attachment 1 Page 2 of 19 Table 2. Affected component IDs ASME ASME Component ID Component Description Category Item No.
Unit 1 - Pressurizer (01-RC-E-2) 8-8 82.11 1-07 PZR Shell to Upper Head 8-8 82.11 1-08 PZR Shell to Lower Head 8-8 82.12 1-09 PZR Shell Longitudinal Weld - Lower 8-8 82.12 1-15 PZR Shell Longitudinal Weld - Upper Unit 2 - Pressurizer (02~RC-E-2) 8-8 82.11 1-04 PZR Shell to Lower Head 8-8 82.11 1-07 PZR Shell to Upper Head 8-8 82.12 1-02 PZR Shell Longitudinal Weld - Upper 8-8 82.12 1-03 PZR Shell Longitudinal Weld - Lower Unit 1 - 'A' Steam Generator (01-RC-E-1A) 8-8 82.40 1-01 SG Primary Head to Tubesheet C-A C1.10 2-03 SG Shell to Shell C-A C1.10 2-05 SG Shell to Lower Transition Cone C-A C1.10 2-06 SG Shell to Upper Transition Cone C-A C1.20 2-08 SG Shell to Upper Head C-A C1.30 2-02 SG Shell to Tubesheet C-8 C2.21 2-09 SG Shell to Feedwater Nozzle C-8 C2.21 2-10 SG Shell to Main Steam Nozzle C-8 C2.22 1-RC-2-01CNIR SG Shell to Main Steam Nozzle Inside Radius C-8 C2.22 1-RC-2-01 DNIR SG Shell to Feedwater Nozzle Inside Radius Unit 1 - 'B' Steam Generator (01-RC-E-18) 8-8 82.40 1-01 SG Primary Head to Tubesheet C-A C1.10 2-03 SG Shell to Shell C-A C1.10 2-05 SG Shell to Lower Transition Cone C-A C1.10 2-06 SG Shell to Upper Transition Cone C-A C1.20 2-08 SG Shell to Upper Head C-A C1.30 2-02 SG Shell to Tubesheet C-8 C2.21 2-09 SG Shell to Feedwater Nozzle C-8 C2.21 2-10 SG Shell to Main Steam Nozzle C-8 C2.22 1-RC-2-02CNIR SG Shell to Main Steam Nozzle Inside Radius C-8 C2.22 1-RC-2-02DNIR SG Shell to Feedwater Nozzle Inside Radius
Serial No.22-114 Docket Nos.: 50-280/281 Attachment 1 Page 3 of 19 ASME ASME Component ID Component Description Category Item No.
Unit 1 - 'C' Steam Generator (01-RC-E-1C) 8-8 82.40 1-01 SG Primary Head to Tubesheet C-A C1.10 2-03 SG Shell to Shell C-A C1.10 2-05 SG Shell to Lower Transition Cone C-A C1.10 2-06 SG Shell to Upper Transition Cone C-A C1.20 2-08 SG Shell to Upper Head C-A C1.30 2-02 SG Shell to Tubesheet C-8 C2.21 2-09 SG Shell to Feedwater Nozzle C-8 C2.21 2-10 SG Shell to Main Steam Nozzle C-8 C2.22 1-RC-2-03CNIR SG Shell to Main Steam Nozzle Inside Radius C-8 C2.22 1-RC-2-03DNIR SG Shell to Feedwater Nozzle Inside Radius Unit 2- 'A' Steam Generator (02-RC-E-1A) 8-8 82.40 1-01 SG Primary Head to Tubesheet C-A C1.10 2-03 SG Shell to Shell C-A C1.10 2-05 SG Shell to Lower Transition Cone C-A C1.10 2-06 SG Shell to Upper Transition Cone C-A C1.20 2-08 SG Shell to Upper Head C-A C1.30 2-02 SG Shell to Tubesheet C-8 C2.21 2-09 SG Shell to Feedwater Nozzle C-8 C2.21 2-10 SG Shell to Main Steam Nozzle C-8 C2.22 2-RC-2-01 CNIR SG Shell to Main Steam Nozzle Inside Radius C-8 C2.22 2-RC-2-01 DNIR SG Shell to Feedwater Nozzle Inside Radius Unit 2- 'B' Steam Generator (02-RC'.'E-1B) 8-8 82.40 1-01 SG Primary Head to Tubesheet C-A C1.10 2-03 SG Shell to Shell C-A C1.10 2-05 SG Shell to Lower Transition Cone C-A C1.10 2-06 SG Shell to Upper Transition Cone C-A C1.20 2-08 SG Shell to Upper Head C-A C1.30 2-02 SG Shell to Tubesheet C-8 C2.21 2-09 SG Shell to Feedwater Nozzle C-8 C2.21 2-10 SG Shell to Main Steam Nozzle C-8 C2.22 2-RC-2-02CNIR SG Shell to Main Steam Nozzle Inside Radius C-8 C2.22 2-RC-2-02DNIR SG Shell to Feedwater Nozzle Inside Radius
Serial No.22-114 Docket Nos.: 50-280/281 Attachment 1 Page 4 of 19 ASME ASME Component ID Component Description Category Item No.
Unit 2 - 'C' Steam Generator (02-RC-E-1 C)
B-B 82.40 1-01 SG Primary Head to Tubesheet C-A C1.10 2-03 SG Shell to Shell C-A C1.10 2-05 SG Shell to Lower Transition Cone C-A C1.10 2-06 SG Shell to Upper Transition Cone C-A C1.20 2-08 SG Shell to Upper Head C-A C1.30 2-02 SG Shell to Tubesheet C-B C2.21 2-09 SG Shell to Feedwater Nozzle C-B C2.21 2-10 SG Shell to Main Steam Nozzle C-B C2.22 2-RC-2-03CNIR SG Shell to Main Steam Nozzle Inside Radius C-B C2.22 2-RC-2-03DNIR SG Shell to Feedwater Nozzle Inside Radius
2.0 Applicable Code Edition and Addenda
The Code of record for the Surry Power Station (SPS), Units 1 and 2, fifth 10-year inservice inspection (ISi) interval is the ASME Boiler and Pressure Vessel Code (ASME Code),Section XI, "Rules for lnservice Inspection of Nuclear Power Plant Components," 2004 Edition [1-1]. The Unit 1 fifth 10-year ISi interval started on December 14, 2013, and ends on October 13, 2023. The Unit 2 fifth 10-year ISi interval started on May 10, 2014, and ends on May 9, 2024.
3.0 Applicable Code Requirement
The ASME Code,Section XI, paragraph IWB-2500(a), Table IWB-2500-1, Examination Category 8-8 and paragraph IWC-2500(a), Table IWC-2500-1, Examination Categories C-A and C-8 require examination of the following Item Nos.:
82.11 Volumetric examination of essentially 100% of the weld length for both circumferential shell-to-head welds during each inspection interval. The examination volume is shown in Figure IWB-2500-1.
82.12 Volumetric examination of one (1) foot of all longitudinal shell-to-head welds during the first inspection interval and one foot of one (1) weld per head during successive intervals. The examination volume is shown in Figure IWB-2500-2.
Serial No.22-114 Docket Nos.: 50-280/281 Attachment 1 Page 5 of 19 82.40 Volumetric examination of essentially 100% of the weld length of all welds during the first Section XI inspection interval. For successive inspection intervals the examination may be limited to one (1) vessel among the group of vessels performing a similar function. The examination volume is shown in Figure IWB-2500-6.
C1.10 Volumetric examination of essentially 100% of the weld length of the cylindrical-shell-to-conical shell-junction welds and shell (or head)-to-flange welds during each Section XI inspection interval. In the case of multiple vessels of similar design, size, and service (such as SGs, heat exchangers), the required examinations may be limited to one (1) vessel or distributed among the vessels. The examination volume is shown in Figure IWC-2500-1.
C1.20 Volumetric examination of essentially 100% of the weld length of the head-to-shell weld during each Section XI inspection interval. In the case of multiple vessels of similar design, size, and service (such as SGs, heat exchangers), the required examinations may be limited to one (1) vessel or distributed among the vessels. The examination volume is shown in Figure IWC-2500-1.
Volumetric examination of essentially 100% of the weld length of the tubesheet-to-shell welds during each Section XI inspection interval.
In the case of multiple vessels of similar design, size, and service (such as SGs, heat exchangers), the required examinations may be limited to one vessel or distributed among the vessels. The examination volume is shown in Figure IWC-2500-2.
C2.21 Volumetric and surface examination of all nozzle welds at terminal ends of piping runs during each Section XI inspection interval. In the case of multiple vessels of similar design, size, and service (such as SGs, heat exchangers), the required examinations may be limited to one (1) vessel or distributed among the vessels. The examination area and volume are shown in Figures IWC-2500-4(a), (b), or (d).
Volumetric examination of all nozzle inside radius sections at terminal ends of piping runs during each Section XI inspection interval. In the case of multiple vessels of similar design, size, and service (such as SGs, heat exchangers), the required examinations may be limited to one (1) vessel or distributed among the vessels. The examination volume is shown in Figures IWC-2500-4(a}, (b), or (d).
Serial No.22-114 Docket Nos.: 50-280/281 Attachment 1 Page 6 of 19
4.0 Reason for Request
Electric Power Research Institute (EPRI) performed assessments in References
[1-2], [1-3], and [1-4] of the basis for the ASME Code Section XI examination requirements specified for the above listed ASME Code Section XI, Division 1 Examination Categories for PZR and SG welds and components. The assessments include a survey of inspection results from 74 domestic and international nuclear units and flaw tolerance evaluations using probabilistic fracture mechanics (PFM) and deterministic fracture mechanics (DFM). The reports in References [1-2], [1-3], and [1-4], developed consistent with the recommendations provided in EPRl's White Paper on PFM [1-5], concluded that the current ASME Code Section XI inspection interval of ten (10) years can be increased significantly with no impact to plant safety. Based on the conclusions of the three EPRI reports, Dominion Energy Virginia is requesting an alternative to the 10-year inspection interval for the subject welds.
5.0 Proposed Alternative and Basis for Use Dominion Energy Virginia is requesting an alternative to the ASME Code Section XI examination requirements in Tables IWB-2500-1 and IWC-2500-1 for the following Examination Categories and Item Numbers:
ASME Item No. Description Category B-B 82.11 PZR, shell-to-head welds, circumferential 8-B 82.12 PZR, shell-to-head welds, longitudinal B-B 82.40 SGs (primary side), tubesheet-to-head weld C-A C1.10 SGs (secondary side), shell circumferential welds C-A C1.20 SGs (secondary side), head circumferential welds C-A C1.30 SGs (secondary side), tubesheet-to-shell weld SGs (secondary side), nozzle-to-shell (nozzle-to-head or C-B C2.21 nozzle-to-nozzle) welds C-B C2.22 SGs (secondary side), nozzle inside radius sections
Serial No.22-114 Docket Nos.: 50-280/281 Attachment 1 Page 7 of 19 SPS, Units 1 and 2, are currently in the third period of the fifth 10-year ISi interval.
Unit 1 will have completed the required fifth interval ISi examinations of the PZR and SGs at the time of the approval of this request. The ISi examinations of the Unit 2 PZR and SGs for the third period, fifth interval will not be completed by the requested approval date for the proposed alternative.
The proposed alternative is to increase the inspection interval for these examination items from the current ASME Code Section XI 10-year requirement by deferring the PZR and SG examinations for the following:
- 2) For SPS Unit 2, PZR and SGs, for the remainder of the third period of the fifth 10-year ISi interval through the sixth 10-year ISi interval.
A summary of the technical basis for this request is provided below. The applicability of the technical basis to SPS, Units 1 and 2, is demonstrated in Attachments 2, 3, and 4.
A. Degradation Mechanism Evaluation An evaluation of degradation mechanisms that could potentially impact the reliability of the PZR and SG welds and components was performed in References
[1-2], [1-3], and [1-4]. The degradation mechanisms that were evaluated included stress corrosion cracking (SCC), environmental assisted fatigue (EAF),
microbiologically influenced corrosion (MIC), pitting, crevice corrosion, erosion-cavitation, erosion, flow accelerated corrosion (FAC), general corrosion, galvanic corrosion, and mechanical/thermal fatigue. Other than the potential for EAF and mechanical/thermal fatigue, there were no active degradation mechanisms identified that significantly affect the long-term structural integrity of the PZR and SG welds and components covered in this request. Therefore, only these fatigue-related mechanisms considered in the PFM and DFM evaluations in References
[1-2], [1-3], and [1-4] are applicable to the components in this request.
B. Stress Analysis Finite element analyses (FEA) were performed in References [1-2], [1-3], and [1-4] to determine the stresses in the PZR and SG welds and components covered in this request. The finite element models used in References [1-2], [1-3], and [1-4] are consistent with the configurations of SPS, Units 1 and 2, therefore no new FEA model is required for the stress analysis of these units. The analyses were performed using representative pressurized water reactor (PWR) geometries, bounding transients, and typical material properties. The results of the stress
Serial No.22-114 Docket Nos.: 50-280/281 Attachment 1 Page 8 of 19 analyses were used in a flaw tolerance evaluation. The applicability of the FEA analysis to SPS, Units 1 and 2, is demonstrated in Attachments 2 and 3 and confirms that all plant-specific requirements are met. Therefore, the evaluation results and conclusions contained in References [1-2], [1-3], and [1-4] are applicable to SPS, Unit 1 and Unit 2. In particular, the key geometric parameters used in the stress analyses in References [1-2], [1-3], and [1-4] are compared to those of SPS, Units 1 and 2, in Table 3 for the PZR and Tables 4 and 5 for the SGs:
Table 3. PZR shell dimensions Shell Inside Shell/ Clad Shell Shell Diameter (ID) Thickness Ro/t Rlt (inches) (inches)
EPRI Report 84 <1> 3.75 / 0.063 <1> 12.2 <1> 11.2 (Table 4-4 of [1-2])
SPS, Units 1 and 2 84 3.75 / 0.19 12.2 11.2 1
<> Westinghouse PZR dimensions associated with model for lower head.
As noted by the Nuclear Regulatory Commission (NRC) in the Safety Evaluation (SE) 1 [1-6] for Salem Generating Station (Salem), Units 1 and 2, the dominant stress is pressure stress. Therefore, the variation in the Ri/t ratio determined in Tables 3 and 4 can be used to scale up the stresses in Reference [1-2] to obtain the plant-specific stresses for each unit and component.
In the selection of the transients in Section 5 of Reference [1-2], test conditions beyond a system leakage test were not considered since pressure tests for SPS, Units 1 and 2, are performed at normal operating conditions. No hydrostatic testing of the SPS, Units 1 and 2, PZR and SGs has been performed since the plants went into operation.
1 Section 5.1, page 7, fourth paragraph
Serial No.22-114 Docket Nos.: 50-280/281 Attachment 1 Page 9 of 19 Table 4. SG vessel dimensions Primary Secondary Lower Lower Lower Lower Lower Lower Head Head Shell Shell Head Shell ID Thicknes ID Thicknes Ri/t RJt (inches) (inches) (inches) (inches)
EPRI Report 155.87 6.94 11.2 162.45 3.65 22.3 (Table 4-2 of [1-3])
SPS, Units 1 and 2 125.28 5.2 12.0 129.38 3.25 19.9 Table 5. SG nozzle dimensions Feedwater Main Steam Nozzle Nozzle Nozzle Nozzle Nozzle Nozzle ID Thickness ID Thickness RJt Rdt (inches) (inches) (inches) (inches)
EPRI Report 16.5 6 1.38 22.25 4.53 2.46 (Figures 4-9/4-10 of [1-4])
SPS, Units 1 and 2 15.96 5.02 1.58 28.96 (1) 1> A specific Ri/t was not used for the SPS MS nozzle configuration in Reference [1-4]. The SPS MS nozzle consists of a tapered nozzle with a concave taper that resembles a triangle. Due to the differences between the Westinghouse MS nozzle configurations, additional stress modeling was performed to evaluate how the stresses varied through the wall thicknesses compared to the nozzle selected for evaluation. The stress comparison for the MS nozzle design is detailed in Section 7.4 of Reference [1-4].
As discussed in Sections 4.3.3 and 4.6 of Reference [1-4] and noted by the NRC in the SE 2 [1-7] for Vogtle Electric Generating Plant (Vogtle), Units 1 and 2, the dominant stress is pressure stress. Therefore, the variation in the Ri/t ratio determined in Tables 4 and 5 can be used to scale up the stresses in References
[1-3] and [1-4] to obtain the plant-specific stresses for each unit and component.
In the selection of the transients in Section 5 of References [1-3] and [1-4] and the subsequent stress analyses in Section 7, test conditions beyond a system leakage test were not considered since pressure tests for SPS, Units 1 and 2, are 2 Section 3.8.3.1, page 9, third paragraph
Serial No.22-114 Docket Nos.: 50-280/281 Attachment 1 Page 10 of 19 performed at normal operating conditions. No hydrostatic testing of the SPS, Units 1 and 2, PZR and SGs has been performed since the plants went into operation.
In Reference [1-3], clad residual stress was not considered for the primary side welds. This was noted by the NRC in a Request for Additional Information (RAI) for Millstone Power Station, Unit 2 (MPS2). In response to the RAI [1-8], an evaluation was performed which showed that the clad residual stress has no significant impact on the conclusions of Reference [1-3].3 This was found acceptable by the NRC in Section 5.3 of the SE [1-9] for MPS2.
C. Flaw Tolerance Evaluation Flaw tolerance evaluations were performed in References [1-2], [1-3], and [1-4]
consisting of PFM evaluations and confirmatory DFM evaluations. The results of the PFM analyses indicate that, after a preservice inspection (PSI) followed by subsequent inservice inspections, the NRC's safety goal of 1.0x1 o-6 failures per year is met. The PFM analysis in Reference [1-4] was performed using the PRobabilistic OptiMization of lnSpEction (PROMISE), Version 1.0 software developed by Structural Integrity Associates. As part of the NRC's review of Southern Nuclear Operating Company, lnc.'s Alternative Request[1-1 O] forVogtle, the NRC performed an audit [1-22] of the PROMISE, Version 1.0 software. The PFM analysis in References [1-2] and [1-3] was performed using the PROMISE, Version 2.0 software which has not been audited by the NRC. The only technical difference between Version 1.0 and Version 2.0 of the PROMISE software is that a user-specified examination coverage is applied to all inspections in Version 1.0, whereas the examination coverage can be specified by the user uniquely for each inspection in Version 2.0. In both versions of the software, 100% coverage for the PSI examination is assumed. The NRC staff found the use of PROMISE, Version 2.0 acceptable as approved in the Salem SE4 [1-6].
A comparison of the PSI/ISi scenarios used in the sensitivity studies performed in Reference [1-2] to those for the SPS, Unit 1 and 2, PZRs is provided below.
For the SPS Unit 1 PZR, PSI examinations have been performed followed by ISi examinations over five (5) complete 10-year intervals. The inspection schedule scenario for these welds is PSI plus five (5) complete 10-year ISi examinations (PSI+ 10 + 20 + 30 + 40 + 50 Inspection Scenario). For the SPS Unit 2 PZR, PSI examinations have been performed followed by ISi examinations over four (4) complete 10-year intervals. The inspection schedule scenario for these welds is 3 RAI Response 3c 4 Section 3.1, page 5, fourth paragraph
Serial No.22-114 Docket Nos.: 50-280/281 Attachment 1 Page 11 of 19 PSI plus four (4) complete 10-year ISi examinations (PSI + 10 + 20 + 30 + 40 Inspection Scenario). The analyses involve conservative assumptions with regards to the PSI/ISi scenarios. Furthermore, the evaluation was performed for 80 years, which is longer than the alternative proposed by Dominion Energy Virginia for SPS, Units 1 and 2, in this request.
In the PFM evaluations in Reference [1-2], the Pressure Vessel Research Facility User's Facility (PVRUF) initial flaw size distribution was used. This distribution is applicable to thick vessels and not to relatively thin vessels like PZRs. In a RAI [1-11 ], the NRC staff asked PSEG Nuclear, LLC., (PSEG) to justify its application of this distribution to the Salem PZR vessel lower head shell welds. In response to the RAI [1-12], PSEG used various initial flaw size distributions in a sensitivity study which showed that regardless of which distribution was used, the conclusions of Reference [1-2] remain the same. 5 The NRC determined this conclusion was acceptable in its SE [1-6] for Salem, dated April 12, 2021.
The results of the PFM analyses indicate that, after a PSI, no other inspections are required for up to 80 years of plant operation to meet the NRC's safety goal of 1.0x1 o-6 failures per year. For the specific case of SPS, Units 1 and 2, where PSI, followed by at least two (2) 10-year interval inspections have been performed, Table 8-10 of References [1-3] and [1-4] and Table 8-12 of Reference [1-2]
indicates that if the inspection interval is increased to 30 years after these previous inspections, the NRC safety goal is met (with considerable margin) for up to 80 years of plant operation. The DFM evaluations provide verification of the PFM results by demonstrating that it takes approximately 80 years for a postulated flaw with an initial depth equal to the ASME Code Section XI acceptance standards to grow to a depth where the maximum stress intensity factor (K) exceeds the ASME Code,Section XI, allowable fracture toughness.
In Section 8.2.2.2 of Reference [1-4] and Section 8.3.2.2 of Reference [1-3], a nozzle flaw density of 0.001 flaws per nozzle was assumed for the nozzle inside radius sections. In Section 3.8.5 of the SE [1-7] for Vogtle, the NRC indicated that a nozzle flaw density of 0.1 flaws per nozzle should have been used. Sensitivity studies performed in Section 8.2.4.3.4 in Reference [1-3] indicated that by changing the number of flaws in the nozzle inside radius sections from 0.001 to 0.1, the probabilities of leak and rupture increased by two (2) orders of magnitude but were still significantly below the acceptance criterion of 1x1 o- 6 failures per year.
A comparison of the PSI/ISi scenarios used in the sensitivity studies performed in 5 Section 9.1, page 15, last paragraph
Serial No.22-114 Docket Nos.: 50-280/281 Attachment 1 Page 12 of 19 References [1-3] and [1-4] to those for the SPS, Units 1 and 2, SGs is provided below.
For the replaced portion of the SPS Unit 1 SGs, PSI examinations have been performed followed by ISi examinations in four (4) complete 10-year ISi intervals following SG replacement (PSI + 1O + 20 + 30 + 40 ISi Scenario). For the portion of the Unit 1 SGs not replaced, PSI examinations have been performed followed by ISi examinations in five (5) complete 10-year ISi intervals (PSI + 10 + 20 + 30
+ 40 + 50 ISi Scenario).
For the replaced portion of the SPS Unit 2 SGs, PSI examinations have been performed followed by ISi examinations in three (3) complete 10-year ISi intervals following SG replacement (PSI+ 10 + 20 + 30 ISi Scenario). For the portion of the Unit 2 SGs not replaced, PSI examinations have been performed followed by ISi examinations in four (4) complete 10-year ISi intervals (PSI + 10 + 20 + 30 + 40 ISi Scenario).
The analyses involve conservative assumptions with regards to the PSI/ISi scenarios. Furthermore, the evaluation was performed for 80 years, which is longer than the alternative proposed by Dominion Energy Virginia for SPS, Units 1 and 2, in this request.
The PFM evaluations documented in References [1-2], [1-3], and [1-4] used an ASME Code Section XI, Appendix VIII-based probability of detection (POD) curve in the PFM evaluation because most ISi examinations of major Class 1 and Class 2 plant components are performed using Appendix VIII procedures. However, for some PZR components, the use of Appendix VIII procedures is plant specific.
Many plants adopt and use their Appendix VIII procedures for major Class 1 components (such as PZRs) for consistency across all their examinations. In the case of SPS, Units 1 and 2, ASME Code Section V procedures are used for the PZR and SG ultrasonic examinations. As stated in the NRC SEs for Salem 6 [1-6]
and Vogtle 7 [1-7], the use of the ASME Code Section XI, Appendix VIII-based POD curve for inspections based on Section V procedures would have minimal impact on the PFM results since the POD curve is not one of the parameters that significantly affect the PFM results.
The DFM evaluations in Table 8-4 of Reference [1-2], Table 8-3 of Reference [1-3], and Table 8-31 of Reference [1-4] provides verification of the above PFM results for SPS, Units 1 and 2, by demonstrating that it takes approximately 80 years for a postulated flaw with an initial depth equal to ASME Code Section XI 6 Section 9.2, page 15 7 Section 3.8.8.2, page 21
Serial No.22-114 Docket Nos.: 50-280/281 Attachment 1 Page 13 of 19 acceptance standards to grow to a depth where the maximum stress intensity factor (K) exceeds the ASME Code Section XI allowable fracture toughness.
D. Inspection History As described in Section 8.3.4.1 of Reference [1-2], Section 8.3.4.1 of Reference
[1-3], and Section 8.2.4.1.1 of Reference [1-4], PSI refers to the collective examinations required by the ASME Code Section Ill during fabrication and any Section XI examinations performed prior to service. The Section Ill fabrication examinations required for these components were robust and any Section XI PSI examinations further contributed to thorough initial examinations.
The inspection history for SPS, Units 1 and 2, (including examinations performed to-date, examination findings, examination coverage, and relief requests) is provided in Attachment 4.
As shown in the attachment, some of the welds/components have limited examination coverage. Based on the NRC's SE for Salem 8 [1-6], the probability of rupture is not sensitive to examination coverage. Also, as shown in Attachment 4, no flaws that exceeded the ASME Code Section XI acceptance standards were identified during any examinations.
E. Industry Survey The inspection history for these components (as obtained from an industry survey) is presented in Attachment 5. The results of the survey indicate that these components are very flaw tolerant.
F. Conclusion It is concluded that the PZR and SG pressure-retaining welds and full penetration welded nozzles are very flaw tolerant. PFM and DFM evaluations performed as part of the technical basis reports [1-2], [1-3], and [1-4] demonstrate that using conservative PSI/ISi inspection scenarios for all plants, the NRC safety goal of 1.0x1 o-6 failures per reactor year is met with considerable margins. Plant-specific applicability of the technical basis to SPS, Units 1 and 2, is demonstrated in Attachments 2 and 3. The requested inspection interval provides an acceptable 8 Section 10, page 20
Serial No.22-114 Docket Nos.: 50-280/281 Attachment 1 Page 14 of 19 level of quality and safety in lieu of the current ASME Code Section XI 10-year inspection frequency.
Operating and examination history demonstrates that these components have performed with very high reliability, mainly due to their robust design. Attachment 4 shows the examination history for the PZR and SG welds examined during each of the 10-year inspection intervals to date.
All three (3) SPS, Unit 2, SGs were replaced during the third period of the first 10-year ISi interval in 1980. In 1981, during the third period of the first 10-year ISi interval, all three (3) SPS, Unit 1, SGs were replaced. The upper portions of the SGs were reused, as indicated on Figure A3. The new welds in the SGs received the required fabrication acceptance and PSI examinations followed by the required scheduled ISi examinations.
In addition to the required fabrication and PSI examinations, ISi examinations have been performed for both units during the first five (5) 10-year inspection intervals for the subject PZR and SG welds and components, as shown in Attachment 4.
No flaws that exceeded the ASME Code Section XI acceptance standards were identified during any examinations. It is important to note that all other inspection activities, including the system leakage test (Examination Categories B-P and C-H) will continue to be performed in accordance with the Section XI requirements providing further assurance of safety.
Finally, as discussed in Reference [1-13], for situations where no active degradation mechanism is present, it was concluded that subsequent ISi examinations do not provide additional value after PSI has been performed and the inspection volumes examined have been confirmed to be free of defects.
Therefore, Dominion Energy Virginia requests the NRC grant this proposed alternative for SPS, Units 1 and 2, in accordance with 10 CFR 50.55a(z)(1 ).
6.0 Duration of Proposed Alternative Approval of the proposed alternative is requested by April 1, 2023, to support the Spring 2023 refueling outage (RFO). Upon approval, the proposed alternative will remain in effect for the remainder of the current fifth 10-year ISi interval and through the sixth 10-year ISi interval for the PZR welds, and through the sixth 10-year ISi interval for the SG welds and components for both units.
Serial No.22-114 Docket Nos.: 50-280/281 Attachment 1 Page 15 of 19 The current 10-year ISi interval schedule for SPS, Units 1 and 2, is provided below.
Unit ISi Interval Interval Begins Interval. Ends 5th October 14, 2013 October 13, 2023 1 6th October 14, 2023 October 13, 2033 7th October 14, 2033 October 13, 2043 5th May 10, 2014 May 9, 2024 2 6th May 10, 2024 May 9, 2034 7th May 10, 2034 May 9, 2044 7 .0 Precedent The following is a list of approved actions (including relief requests and topical reports) related to inspections of PZR and SG welds and components:
- Letter from J. G. Danna (NRC) to E. Carr (PSEG Nuclear, LLC), "Salem Generating Station Unit Nos. 1 and 2 -Authorization and Safety Evaluation for Alternative Request No. SC-I4R-200 (EPID L-2020-LLR-0103)," dated August 5, 2020. (ADAMS Accession No. ML21145A189)
- Letter from M. T. Markley (NRC) to C. A. Gayheart (Southern Nuclear Operating Company, Inc.), "Vogtle Electric Generating Plant, Units 1 and 2
- Relief Request for Proposed lnservice Inspection Alternative VEGPISI-ALT-04-04 to the Requirements of the ASME Code (EPID L-2020-LLR-0109)," dated January 11, 2021. (ADAMS Accession No. ML20352A155)
- Letter from J. G. Danna (NRC) to D. G. Stoddard (Dominion Energy Nuclear Connecticut, Inc.), "Millstone Power Station Unit 2 Authorization and Safety Evaluation for Alternative Request No. RR-05-06 (EPID L-2020-LLR-0097)," dated July 16, 2021. (ADAMS Accession No. ML21167A355)
- Letter from J. W. Clifford (NRC) to S. E. Scace (Northeast Nuclear Energy Company), "Safety Evaluation of the Relief Request Associated with the First and Second 10-Year Interval of the lnservice Inspection (ISi) Plan, Millstone Nuclear Power Station, Unit 3 (TAC No. MA 5446)," dated July 24, 2000. (ADAMS Accession No. ML003730922)
Serial No.22-114 Docket Nos.: 50-280/281 Attachment 1 Page 16 of 19
- Letter from R. L. Emch (NRC) to J. B. Beasley, Jr. (Southern Nuclear Operating Company, Inc.), "Second 10-Year Interval lnservice Inspection Program Plan Requests for Relief 13, 14, 15, 21 and 33 for Vogtle Electric Generating Plant, Units 1 and 2 (TAC No. MB0603 and MB0604)," dated June 20, 2001. (ADAMS Accession No. ML011640178)
- Letter from T. H. Boyce (NRC) to C. L. Burton (Carolina Power & Light Company), "Shearon Harris Nuclear Power Plant Unit 1 - Request for Relief 2R1-019, 2R1-020, 2R1-021, 2R1-022, 2R2-009, 2R2-010, 2R2-011 for the Second Ten-Year Interval lnservice Inspection Program Plan (TAC Nos.
ME0609, ME0610, ME0611, ME0612, ME0613, ME0614 and ME0615),"
dated January 7, 2010. (ADAMS Accession No. ML093561419)
- Letter from M. Khanna (NRC) to D. A. Heacock (Dominion Nuclear Connecticut, Inc.), "Millstone Power Plant Unit No. 2 - Issuance of Relief Requests RR-89-69 Through RR-89-78 Regarding Third 10-Year Interval lnservice Inspection Plan (TAC Nos. ME5998 Through ME6006)," dated March 12, 2012. (ADAMS Accession No. ML120541062)
- Letter from R. J. Pascarelli (NRC) to E. D. Halpin (Pacific Gas & Electric Company ), "Diablo Canyon Plant, Units 1 and 2 - Relief Request; NOE SG-MS-IR, Main Steam Nozzle Inner Radius Examination Impracticality, Third 10-Year Interval, American Society of Mechanical Engineers Boiler and Pressure Vessel Code,Section XI, lnservice Inspection Program (CAC Nos. MF6646 and MF6647)," dated December 8, 2015. (ADAMS Accession No. ML15337A021)
In addition, there are precedents related to similar topical reports that justify relief for Class 1 nozzles:
- Based on studies presented in Reference [1-14], the NRC approved extending PWR reactor vessel nozzle-to-shell weldsJrom 10 to 20 years in Reference [1-15].
- Based on work performed in Boiling Water Reactor Vessel and Internals Program (BWRVIP)-108 [1-16] and BWRVIP-241 [1-17], the NRC approved the reduction of boiling water reactor (BWR) vessel feedwater nozzle-to-shell weld examinations (Item No. 83.90 for BWRs from 100% to a 25%
sample of each nozzle type every 10 years) in References [1-18] and [1-19]. The work performed in BWRVIP-108 and BWRVIP-241 provided the technical basis for ASME Code Case N-702 [1-20], which has been conditionally approved by the NRC in Revision 20 of Regulatory Guide 1.147[1-21].
Serial No.22-114 Docket Nos.: 50-280/281 Attachment 1 Page 17 of 19 REFERENCES 1-1 The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI, "Rules for lnservice Inspection of Nuclear Power Plant Components," 2004 Edition.
1-2 EPRI Technical Report 3002015905: Technical Bases for Inspection Requirements for PWR Pressurizer Vessel Head, Shell-to-Head and Nozzle-to-Vessel Welds. Palo Alto, California, 2019.
1-3 EPRI Technical Report 3002015906: Technical Bases for Inspection Requirements for PWR Steam Generator Class 1 Nozzle-to-Vessel Welds and Class 1 and Class 2 Vessel Head, Shell, Tubesheet-to-Head and Tubesheet-to-Shell Welds. Palo Alto, California, 2019.
1-4 EPRI Technical Report 3002014590: Technical Bases for Inspection Requirements for PWR Steam Generator Feedwater and Main Steam Nozzle-to-Shell Welds and Inside Radius Sections. Palo Alto, California, 2019.
1-5 N. Palm (EPRI), BWR Vessel & Internals Project (BWRVIP) Memo No. 2019-016, "White Paper on Suggested Content for PFM Submittals to the NRC," February 27, 2019. (ADAMS Accession No. ML19241A545).
1-6 Letter from James G. Danna (NRC) to Eric. Carr (PSEG Nuclear, LLC), "Salem Generating Station Unit Nos. 1 and 2 - Authorization and Safety Evaluation for Alternative Request No. SC-14R-200 (EPID L-2020-LLR-0103)," dated April 12, 2021. (ADAMS Accession No. ML20218A587) 1-7 Letter from Michael T. Markley (NRC) to Cheryl A. Gayheart (Southern Nuclear Operating Company, Inc.), "Vogtle Electric Generating Plant, Units 1 & 2 - Relief Request for Proposed lnservice Inspection Alternative VEGP-ISI-ALT-04-04 to the Requirements of ASME Code (EPID L-2020-LLR-0109)," dated January 11, 2021.
(ADAMS Accession No. ML20352A155) 1-8 Letter from Gerald T. Bischof (Dominion Energy) to the NRC, "Dominion Energy Nuclear Connecticut, Inc. Millstone Power Station Unit 2 Response to Request for Additional Information for Alternative Request RR-05 Inspection Interval Extension for Steam Generator Pressure-Retaining Welds and Full-Penetration Welded Nozzles," dated March 19, 2021. (ADAMS Accession No. ML21034A576) 1-9 Letter from James G. Danna (NRC) to Daniel G. Stoddard (Dominion Energy),
"Millstone Power Station Unit 2 - Authorization and Safety Evaluation for Alternative Request No. RR-05-06 (EPID L-2020-LLR-0097)," dated July 16, 2021.
(ADAMS Accession No. ML21167A355)
Serial No.22-114 Docket Nos.: 50-280/281 Attachment 1 Page 18 of 19 1-10 Letter from C. A. Gayheart (Southern Nuclear Operating Company, Inc.) to the NRC, "Vogtle Electric Generating Plant, Units 1 & 2 Proposed lnservice Inspection Alternative VEGP-ISI-AL T-04-04 Version 2.0," dated September 9, 2020. (ADAMS Accession No. ML20253A311) 1-11 Email Letter from J. Kim (NRC) to P. R. Duke (PSEG Nuclear, LLC), "Salem Generating Station Units 1 and 2.- Final Request for Additional Information Regarding Alternative for Examination of ASME Section XI, Category B-B, Item Number B2.11 And B2.12 (L-2020-LRR-0103)," dated February 11, 2021.
(ADAMS Accession No. ML21043A144) 1-12 Letter from P. R. Duke, Jr. (PSEG Nuclear, LLC) to NRC, "Response to Request for Additional for Proposed Alternative for ASME Section XI, Category B-B, Item Number B2.11 And B2.12," dated April 12, 2021. (ADAMS Accession No. ML21102A024) 1-13 ASME, Risk-Based Inspection: Development of Guidelines, Volume 2-Part 1 and Volume 2-Part 2, Light Water Reactor (LWR) Nuclear Power Plant Components.
CRTD-Vols. 20-2 and 20-4, ASME Research Task Force on Risk-Based Inspection Guidelines, Washington, DC, 1992 and 1998.
1-14 B. A. Bishop, C. Boggess, N. Palm, "Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval," WCAP-16168-NP-A, Rev. 3, October 2011.
1-15 NRC, "Revised Safety Evaluation by the Office of Nuclear Reactor Regulation; Topical Report WCAP-16168-NP-A, Revision 2, 'Risk-Informed Extension of the Reactor Vessel In-service Inspection Interval,' Pressurized Water Reactor Owners Group, Project No. 694," July 26, 2011. (ADAMS Accession No. ML111600303) 1-16 EPRI Technical Report 1003557: BWRVIP-108: BWR Vessels and Internals Project, Technical Basis for the Reduction of Inspection Requirements for the Boiling Water Reactor Nozzle-to-Shell Welds and Nozzle Blend Radii. Palo Alto, California, 2002.
1-17 EPRI Technical Report 1021005: BWRVIP-241: BWR Vessels and Internals Project, Probabilistic Fracture Mechanics Evaluation for the Boiling Water Reactor Nozzle-to-Shell Welds and Nozzle Blend Radii. Palo Alto, California, 2010.
1-18 NRC, Safety Evaluation of Proprietary EPRI Report, "BWR Vessel and Internals Project, Technical Basis for the Reduction of Inspection Requirements for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Inner Radius (BWRVIP-108)," December 19, 2007. (ADAMS Accession No. ML073600374)
Serial No.22-114 Docket Nos.: 50-280/281 Attachment 1 Page 19 of 19 1-19 NRC, Safety Evaluation of Proprietary EPRI Report, "BWR Vessel and Internals Project, Probabilistic Fracture Mechanics Evaluation for the Boiling Water Reactor Nozzle-to-Shell Welds and Nozzle Blend Radii (BWRVIP-241 )," April 19, 2013.
(ADAMS Accession Nos. ML13071A240 and ML13071A233) 1-20 Code Case N-702, "Alternate Requirements for Boiling Water Reactor (BWR)
Nozzle Inner Radius and Nozzle-to-Shell Welds," ASME Code Section XI, Division 1, Approval Date: February 20, 2004.
1-21 NRC Regulatory Guide 1.147, Revision 20, "lnservice Inspection Code Case Acceptability, ASME Code Section XI, Division 1," dated December 2021.
1-22 Letter from John G. Lamb (NRC) to C. A. Gayheart (Southern Nuclear Operating Company, Inc.), "Vogtle Electric Generating Plant, Units 1 & 2 Audit Plan for Relief Request lnservice Inspection Alternative VEGP-ISI-AL T-04-04 (EPID L-2019-LLR-0109)," dated May 14, 2020. (ADAMS Accession No. ML20128J311)
Serial No.22-114 Docket Nos.: 50-280/281 ATTACHMENT 2 Plant-Specific Applicability SURRY POWER STATION UNITS 1 AND 2 VIRGINIA ELECTRIC AND POWER COMPANY (DOMINION ENERGY VIRGINIA)
Serial No.22-114 Docket Nos.: 50-280/281 Attachment 2 Page 1 of 17 Plant-Specific Applicability Section 9 of References [2-1], [2-2], and [2-3] provide requirements that must be demonstrated to apply the representative stress and flaw tolerance analyses to a specific plant. Plant-specific evaluation of these requirements for Surry Power Station (SPS), Units 1 and 2, is provided in Tables A 1 through A6.
Tables A 1, A2, A3, A4, AS, and A6 of this attachment and Tables A? and AB of Attachment 3 indicate that all plant-specific requirements are met for SPS, Units 1 and 2. Therefore, the results and conclusions of the EPRI reports are applicable to SPS, Units 1 Unit 2.
Figures A 1, A2, and A3 show the layout of the pressurizers (PZRs) and steam generators (SGs).
Serial No.22-114 Docket Nos.: 50-280/281 Attachment 2 Page 2 of 17 Table A1. Plant-specific applicability of References [2-1], [2-2], and [2-3] Representative Analyses to SPS, Units 1 and 2, PZR and SG components Applicability toSPS,Units1* and 2 The plant-specific PZR general transients and The number and type of the SPS, Units 1 and 2, general cycles must be bounded by those shown in Table transients are compared to the transients listed in Table 5-6 5-6 for a 60-year operating life. It should be noted of Reference [2-1]. As shown in Table A2, the SPS, Units 1 that the number of cycles were extrapolated to 80 and 2, transients are bounded by the transients listed in years in the evaluations. Table 5-6 of Reference [2-1].
General Requirements The materials of the PZR shell and nozzles must The SPS, Units 1 and 2, PZR upper and lower heads and be low alloy ferritic steels which conform to the nozzles are fabricated from SA-216, Grade WCC, cast requirements of ASME Code,Section XI, carbon steel material. The PZR shells are fabricated of SA-Appendix G, Paragraph G-2110. 302, grade B, carbon steel material.
The limiting RT NOT value for the PZR head SA-216, Grade wee, material is 60°F and is bounded by the 60°F assumption used in Reference [2-1].
The limiting RT NOT value for the PZR shell, SA-302, Grade B, material is 30°F and is bounded by the 60°F assumption used in Reference [2-1].
Serial No.22-114 Docket Nos.: 50-280/281 Attachment 2 Page 3 of 17 The plant-specific PZR upper head and bottom The SPS, Units 1 and 2, PZR upper head and bottom head head weld configurations must conform to those weld configurations conform to those shown in Figure 1-1 shown in Figure 1-1 and Figure 1-2 for Item No. and Figure 1-2 for Item No. 82.11 and Item No. B2.12, 82.11 and Item No. 82.12, respectively. [2-1] respectively. [2-1]
The plant-specific dimensions of the PZR upper The comparison of the SPS, Units 1 and 2, PZR dimensions head and nozzles, shell, lower head, and the with those in Table 9-1 of Reference [2-1] is provided in surge nozzle must be within the range of values Table A3. The comparison shows that the SPS, Units 1 and Specific listed in Table 9-1 of Reference [2-1]. 2, configurations are within the range of values shown in Requirements Table 9-1. [2-1]
The plant-specific lnsurge/Outsurge (1/0) In Attachment 3 of this request, the SPS, Units 1 and 2, 1/0 transient definitions (temperature difference transients are compared to the number and type of between the PZR shell and the PZR surge nozzle transients listed in Table 5-10 of Reference [2-1]. As can be fluid temperature and associated number of seen from Table AB, the SPS, Units 1 and 2, transients are cycles) must be bounded by those shown in Table bounded by those transients listed in Table 5-10 of 5-1 O for a Westinghouse/CE plant of Reference Reference [2-1].
[2-1].
Serial No.22-114 Docket Nos.: 50-280/281 Attachment 2 Page 4 of 17 Category Requirement from Reference [2-2] Applicability to SPS, Units 1 and 2 The Loss of Power transient involving SPS, Units 1 and 2, have not experienced a loss of power unheated auxiliary feedwater (AFW) being transient resulting in unheated AFW being introduced into a introduced into a hot SG that has been boiled hot SG that has been boiled dry following blackout, resulting dry following blackout, resulting in thermal in thermal shock of portion of the vessel.
shock of portion of the vessel) is not considered in this evaluation due to its rarity. If such a significant thermal event occurs at a plant, its impact on the K,c value may require more frequent examinations and other plant actions outside the scope of this report's guidance.
General Requirements The materials of the SG vessel heads and The SPS, Units 1 and 2, SG lower heads are fabricated of tubesheet must be low alloy ferritic steels SA-216, Grade WCC, carbon steel. The tubesheets are which conform to the requirements of ASME fabricated of SA-508, Class 2a material. The SG lower shells Code Section XI, Appendix G, Paragraph G- are fabricated of SA-533, Grade A, Class 2 material.
2110.
The limiting RT NDT value for the SG lower head material is 60°F and is bounded by the 60°F assumption used in Reference [2-2].
The limiting RT NDT value for the SG tubesheet and lower shell materials is 30°F and is bounded by the 60°F assumption used in Reference [2-2].
Serial No.22-114 Docket Nos.: 50-280/281 Attachment 2 Page 5 of 17 The weld configurations must conform to those The SPS, Units 1 and 2, weld configurations conform to shown in Figures 1-1 and 1-2 [2-2]. Figures 1-1 and 1-2 [2-2].
The SG vessel dimensions must be within 10 The SPS, Units 1 and 2, SG vessel dimensions are as percent of the upper and lower bounds of the follows:
values provided in the table in Section 9.4.3 of Reference [2-2]. SG Lower Head SG Upper Shell Specific Outside Diameter Outside Diameter Requirements 135.88 inches 175. 75 inches These dimensions are within 1O percent of those specified in Section 9.4.3, Table 9-2. [2-2]
The component must experience transients As shown in Table A4, there are slight variations on some and cycles bounded by those shown in Table temperature and pressure values between SPS, Units 1 and 5-7 of Reference [2-2] over a 60-year operating 2 and Table 5-7 of Reference [2-2]. The SPS, Units 1 and 2 life. transients and number of cycles projected to occur over an 80-year life are bounded by those shown in Table 5-7 of Reference [2-2].
Serial No.22-114 Docket Nos.: 50-280/281 Attachment 2 Page 6 of 17 The Loss of Power transient (involving SPS, Units 1 and 2, have not experienced a loss of power unheated AFW being introduced into a hot SG transient resulting in unheated AFW being introduced into a that has been boiled dry following blackout, hot SG that has been boiled dry following blackout, resulting resulting in thermal shock of portion of the in thermal shock of portion of the vessel.
vessel) is not considered in this evaluation due to its rarity. If such a significant thermal event occurs at a plant, its impact on the K,c value may require more frequent examinations and other plant actions outside the scope of this report's guidance.
General Requirements The materials of the SG shell, FW nozzles, and The SPS, Units 1 and 2, SG vessel upper heads and shells MS nozzles must be low alloy ferritic steels are fabricated of SA-533, Grade A, class 1 material. The FW which conform to the requirements of ASME and MS nozzles are fabricated of SA-508, Class 2 material.
Code Section XI, Appendix G, Paragraph G-The limiting RTNor value for the SG upper head, upper shell, 2110.
and nozzle materials is 30°F and is bounded by the 60°F assumption used in Reference [2-3].
Serial No.22-114 Docket Nos.: 50-280/281 Attachment 2 Page 7 of 17 Applicability. to SPS, Units 1 and 2 The weld configurations must conform to those The SPS, Units 1 and 2, weld configurations conform to shown in Figures 1-7 and 1-8. [2-2] Figures 1-7 and 1-8. [2-2]
The SG vessel dimensions must be within 10 The SPS, Unit 1 and 2, SG vessel dimensions are as follows:
percent of the upper and lower bounds of the values provided in the table in Section 9.4.4 of SG Lower Head .*
SG Upper Shell Reference [2-2]. Outside.**Diameter Outside Diameter Specific Requirements 135.88 inches 175. 75 inches These dimensions are within 10 percent of those specified in Section 9.4.4, Table 9-3. of Reference [2-2].
The component must experience transients As shown in Table AS, there are slight variations on some and cycles bounded by those shown in Table temperature and pressure values between SPS, Units 1 and 5-9 of Reference [2-2] over a 60-year operating 2, and Table 5-7 of Reference [2-2]. The SPS, Units 1 and life. 2, transients and number of cycles projected to occur over an 80-year life are bounded by those shown in Table 5-7 of Reference [2-2]
Serial No.22-114 Docket Nos.: 50-280/281 Attachment 2 Page 8 of 17 Category Requirementfro1n Reference [2-3]
The nozzle-to-shell weld shall be one of The SPS, Units 1 and 2, FW and MS nozzle-to-shell weld the configurations shown in Figure 1-1 or configurations are representative of the configuration shown in 1-2 of Reference [2-3]. Figure 1-2 of Reference [2-3].
The materials of the SG shell, FWnozzles, The most limiting sensitivity study performed in Reference [2-3]
and MS nozzles must be low alloy ferritic for fracture toughness was performed for the FW nozzles. The steels which conform to the requirements applicable transients used in the analysis are associated with the of ASME Code Section XI, Appendix G, Loss of Load condition and occur at temperatures above 200°F Paragraph G-2110. which result in the larger stress intensity values shown in Table 8-31 of Reference [2-3] for path FEW-P1 N. From Figure 7-32 of Reference [2-3], the maximum stress levels associated with the Loss of Load transient are more severe than stresses associated General with the Heatup and Cooldown transient. It is also noted that a Requirements significant contributor for stresses in these areas is the system pressure, which is lower at the beginning of the Heatup and at the end of the Cooldown transients when the temperatures are less than 200°F, and the stress intensity values reported in Table 8-31 are representative of full system pressure.
The materials for the SPS MS and FW nozzles are SA-508, class 2 and the upper head and shell material is SA-533, grade A, class
- 1. Both SA-508, class 2 and SA-533, grade A, class 1 materials have a minimum yield strength of 50 ksi and are therefore considered to have an RTNor of 30°F as previously noted, which is bounded by the conservative 60°F assumption used in the analysis and the SPS SG FW and SG MS nozzles are considered bounded by the EPRI analysis, Reference [2-3].
Serial No.22-114 Docket Nos.: 50-280/281 Attachment 2 Page 9 of 17 The SG must not experience more than As shown in Table A6, the SPS, Units 1 and 2, SGs are not General the number of all transients shown in projected to experience more than the number of transients Requirements Table 5-5 of Reference [2-3] over a 60- shown in Table 5-5 of Reference [2-3].
year operating life.
The piping attached to the FW nozzle The SPS, Units 1 and 2, FW piping lines are 16 inches.
must be 14-inch to 18-inch nominal pipe size (NPS).
The FW nozzle design must have an The SPS, Units 1 and 2, FW nozzle configuration has an SG Feedwater integrally attached thermal sleeve. integrally attached thermal sleeve.
Nozzle AFW nozzles connected directly to the SG are not covered in this evaluation. N/A For Westinghouse plants, the piping SPS, Units 1 and 2, are 3-loop Westinghouse PWRs.
attached to the SG MS nozzle must be 28-The SPS, Units 1 and 2, MS piping lines are 30-inch NPS.
inch to 36-inch NPS.
The SG must have one (1) MS nozzle that SPS, Units 1 and 2, have one (1) MS nozzle per SG that exits the SG Main Steam exits the top dome of the SG. top dome of each SG.
Nozzles The MS nozzle shall not significantly The SPS, Units 1 and 2, MS nozzle configuration does not protrude into the SG (e.g., see Figure 4-7 protrude significantly into the SG as shown in Figure 4-7 [2-3] and of Reference [2-3]) or have a unique does not have a unique weld configuration as shown in Figure 4-nozzle weld configuration (e.g., see Figure 6 [2-3].
4-6 of Reference [2-3]).
Serial No.22-114 Docket Nos.: 50-280/281 Attachment 2 Page 10 of 17 Table A2. Comparison of SPS, Units 1 and 2, PZR general transients to the transients evaluated in Reference [2-1]
Heatup/Cooldown 300 165 146 Loss of Load <1> 360 202 200
<1> Sum of Reactor Trips, 50% Step Load Decrease with Steam Dump, Loss of Load, Loss of Flow in One RC Loop Only, and Loss of Offsite AC Power Events Table A3. Range of PZR geometric parameters for which the evaluation is applicable in comparison with SPS, Units 1 and 2
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PZR Shell Inside Diameter Must be between 80 and 88 84 NPS of piping or component (e.g., reducer)
Surge Nozzle Must be between 12 and 18 14 attached to nozzle safe-end NPS of piping or component (e.g., reducer)
Safety Nozzle Must be between 4 and 8 6 attached to nozzle safe-end NPS of piping or component (e.g., reducer)
Relief Nozzle Must be between 4 and 8 4 attached to nozzle safe-end NPS of piping or component (e.g., reducer)
Spray Nozzle Must be between 4 and 6 4 attached to nozzle safe-end
Serial No.22-114 Docket Nos.: 50-280/281 Attachment 2 Page 11 of 17 Table A4. SPS, Units 1 and 2 data for thermal transients for stress analysis of the PWR SG primary-side head welds (Comparison to Table 5-7 of Reference [2-2])
Max Min Max Min Max Min Thot Thot Tcold Tcold Press Press 60-Year Transient Cycles (OF) (°F) (OF) (OF) (PSIG) (PSIG)
Heatup/Cooldown Reference [2-2] 545 70 545 70 2235 0 300 Heatup/Cooldown SPS, Units 1 and 2 547 60 547 60 2235 0 165*
Plant Loading/Unloading Reference [2-2] 610 550 550 545 2300 2300 5000 Not typica l operation <1>, also note that the applicable number of cycles for a fast Plant Loading/Unloading SPS, Units 1 and 2 ramp which is limited to <5%/min is less than 1700 for loading and unloading.
Reactor Trip Reference [2-2] 615 530 565 530 2435 1700 360 Reactor Trip SPS, Units 1 and 2 609.1 547 547 536.3 2235 1857.7 147**
- Current licensing basis (CLB) number for Heatup and Cooldown (200 each), 80-year projections are <200 cycles for both SPS Units. Based on the conservative Heatup and Cooldown rates used in [2-2], there is no concern for the slightly lower initial temperature assumption.
- SPS 80-year projections. 147 cycles is the highest cycle count projected of Reactor Trips for Unit 1 or Unit 2. Projected cycles are Unit 1 Reactor Trips at 147 cycles and Unit 2 at 142 cycles. The analysis performed in Reference [2-2] are bounding based on the relativelv hiah rates of chanae assumed.
<1> Load following operation is not typical. Plant Loading/ Unloading transients are not applicable.
Serial No.22-114 Docket Nos.: 50-280/281 Attachment 2 Page 12 of 17 Table A5. SPS, Units 1 and 2, Data for Thermal Transients for Stress analysis of the PWR SG Secondary-Side Vessel Welds (Comparison to Table 5-9 of Reference [2-2])
Heatup/Cooldown Reference [2-2] 545 70 0 300 Heatup/Cooldown SPS, Units 1 and 2 547* 60 1005* 0 165*
Plant Loading/Unloading Reference [2-2] 610 550 2300 2300 5000 Not typical operation <2>, also note that the applicable number of cycles for a fast Plant Loading/Unloading SPS, Units 1 and 2 ramp which is limited to <5%/min is less than 1700 for loading and unloading.
Reactor Trip <3> Reference [2-2] 555 530 1130 1000 360 Reactor Trip SPS, Units 1 and 2 547 508.9 1005 722 400**
- Design temperature is 560°F and design pressure is 1085 psig.
SPS 80-year projections. 165 cycles is the highest cycle count projected of Heatups or Cooldowns for Unit 1 or Unit 2. Projected cycles are Unit 1 Heatup 165 cycles and Cooldown 164 cycles, Unit 2 - Heatup 146 cycles and Cooldown 145 cycles. Based on the conservative heatup and cooldown rates used in [2-2] there is no concern for the slightly lower initial temperature assumption.
- 80-year projections. 147 cycles is the highest cycle count projected of Reactor Trips for Unit 1 or Unit 2. Projected cycles are Unit 1 Reactor Trips 147 cycles and Unit 2 142 cycles. The analysis performed in [2-2] are bounding based on the relatively high rates of change assumed.
<1> Tss = secondary-side temperature
<2> Load following operation is not typical. Plant Loading/Unloading transients are not applicable.
<3> The specific SPS temperatures are based on steam temperature and pressures.
Serial No.22-114 Docket Nos.: 50-280/281 Attachment 2 Page 13 of 17 Table A6. SPS, Units 1 and 2 data for thermal transients applicable to PWR SG FW and SG MS Nozzles (Comparison to Table 5-5 of Reference [2-3])
Heatup/Cooldown 300 165 <1)
Plant Loading Not typical operation <2), also note that the applicable number of cycles for a 5000 fast ramp which is limited to <5%/min is less than 1700 for loading and Plant Unloading unloading.
202 (3)
Loss of Load 360 (Note that the BO-year projected number of Reactor Trips are less than 360 cycles.)
Loss of Power 60 45
<1> 80-year projections. 165 cycles is the highest cycle count projected of Heatups or Cooldowns for Unit 1 or Unit 2. Project cycles are: Unit 1,
= =
Heatup 165 cycles, Unit 1, Cooldown 164 cycles, Unit 2, Heatup 146 cycles, and Unit 2, Cooldown 145 cycles.
<2 > Load following operation not typical. Plant Loading / Unloading transients are not applicable.
<3> Loss of Load is the sum of Reactor Trips, 50% Step Load Decrease with Steam Dump, Loss of Load, Loss of Flow in One RC Loop Only, and Loss of Offsite AC Power Events. 80-year projections. 202 cycles is the highest cycle count projected for Loss of Load for Unit 1 or Unit 2.
Projected cycles are: Unit 1 =202 cycles and Unit 2 =200 cycles.
Serial No.: 22-114 Docket Nos.: 50-280/281 Attachment 2 Page 14 of 17 1-RC-E-2
~;........................... l - J3
~ ........ ~.............
4" :
11 4
I.I) l
~! ~
Ul,
~1- - - - - *~"IPlr"'~--s1Hc-+--__,,_..._ _,,...--
It'.!
N Figure A1. SPS Unit 1 Pressurizer Layout
Serial No.22-114 Docket Nos.: 50-280/281 Attachment 2 Page 15 of 17 2-RC-E-2
- SHELL MATERIAL SA*3e2 MADE 8 CAffBOH STEEL 151 N
SUPPORT SKlfU' ~TEJUFI'..
SA-5 16 Cif!ADE 70 CARBOH STEE'.I..
IMHERSJON HEATER
~u:..~ws WO T£ s,
......,.,,_-....-.....- -~~~~:;::t~='=t:.~;::---~===::::::.~
Figure A2. SPS Unit 2 Pressurizer Layout
Serial No.22-114 Docket Nos.: 50-280/281 Attachment 2 Page 16 of 17 ORIGmAL GEJfER!!TOR
. . . - - -.....----- ~ -ir---iwl l._WEPD*l3*
J~ -F'e;Et:kMUEA kQlll.£
- ).:;-c; 2*t,£ REPLACEMEII GENERATOR TRAHsmoH coNE.
ASMESA.~
GR°:ACL1 i:JB" STEAM I GENERATOR
'IU9e SkfEf -* - - - - - -- LSI O..ASS 2A AS~ SA~598 ... - litc:t:.ASS 1A CI.M-S 2 GENERAL ELEVATJON Figure A3. Typical SPS, Units 1 and 2, Steam Generator Layout
Serial No.22-114 Docket Nos.: 50-280/281 Attachment 2 Page 17 of 17 REFERENCES 2-1 EPRI, Technical Report 3002015905: Technical Bases for Inspection Requirements for PWR Pressurizer Vessel Head, Shell-to-Head and Nozzle-to-Vessel Welds. Palo Alto, California: 2019.
2-2 EPRI, Technical Report 3002015906: Technical Bases for Inspection Requirements for PWR Steam Generator Class 1 Nozzle-to-Vessel Welds and Class 1 and Class 2 Vessel Head, Shell, Tubesheet-to-Head and Tubesheet-to-Shell Welds. Palo Alto, California: 2019.
2-3 EPRI, Technical Report 3002014590: Technical Bases for Inspection Requirements for PWR Steam Generator Feedwater and Main Steam Nozzle-to-Shell Welds and Inside Radius Sections. Palo Alto, California: 2019.
2-4 Letter from P.R. Duke, Jr. (PSEG Nuclear, LLC) to NRC, "Response to Request for Additional for Proposed Alternative for ASME Section XI, Category B-8, Item Number B2.11 And B2.12," dated April 12, 2021. (ADAMS Accession No. ML21102A024) 2-5 Code Case N-890, "Materials Exempted from G-211 0(b) Requirement," ASME Code,Section XI, Division 1. Approval Date: October 16, 2018.
2-6 NRC Regulatory Guide 1.147, Revision 20, "lnservice Inspection Code Case Acceptability, ASME Code,Section XI, Division 1," dated December 2021.
2-7 Procurement Specification 953361, Revision 1, Virginia Electric and Power Company, Surry Power Station Units Numbers 1 and 2, Replacement Steam Generator, June 15, 1977.
Serial No.22-114 Docket Nos.: 50-280/281 ATTACHMENT 3 Comparison of lnsurge/Outsurge Transients to the lnsurge/Outsurge Transients Evaluated in EPRl's Technical Report 3002015905 SURRY POWER STATION UNITS 1 AND 2 VIRGINIA ELECTRIC AND POWER COMPANY (DOMINION ENERGY VIRGINIA)
Serial No.22-114 Docket Nos.: 50-280/281 Attachment 3 Page 1 of 3 Comparison of lnsurge/Outsurge Transients to the lnsurge/Outsurge Transients Evaluated in EPRl's Technical Report 3002015905 Surry Power Station (SPS), Units 1 and 2, lnsurge/Outsurge (1/0) transients are provided in Table A7. The temperature differences (LlTs) identified in Table A7 are combined conservatively by summing all the events into the 320°F /1T bin. Conservatively, the transients in Table A7 reflect 80 years of operation compared to the 60-year analysis summarized in Table 5-10, "Summary of temperature differences and numbers of cycles for lnsurge/Outsurge transients (Westinghouse and CE [Combustion Engineering]
plants)," of Electric Power Research lnstitute's (EPRl's) Technical Report 3002015905, "Technical Bases for Inspection Requirements for PWR Pressurizer Vessel Head, Shell-,
to-Head and Nozzle-to-Vessel Welds" [3-1 ].
A comparison of the SPS, Units 1 and 2, 1/0 transients to the requirements in Reference
[3-1] is shown in Table AB. The results of Table AB indicate that the SPS, Units 1 and 2, 1/0 transients are bounded by those in Reference [3-1].
Serial No.22-114 Docket Nos.: 50-280/281 Attachment 3 Page 2 of 3 Table A7. 80-year 1/0 transients for SPS, Units 1 and 2
- .,,>;;.;;;,\ .;>?.'*>* .,.. i,.,,.**. ., *** ,, **,,< .,.,. r'ur;,,. *,;.,,. .:<,,.,,, **** ,,...... ,. . *.. *.* i ,,::,/ {y§:fs
?:l+,;}\>'.":\\tt;>r,:
tr***
- '- '*"* *._ . :::: *.:.: l\;/ci:.J:iC.fJ\>' * '.II£, ** ;,:;; 1~**,l*.~lllt: t*1t* Jf,,;;,r:J'i */'i',,; J' .. ;uu~, c~.~iY'r7{8'f::Yi \;\ . . . **.*
- ,i,,. ,., -:i; 1:1 i *.*:*>ii, ': *i **
- i /i. :*.. * ;* *: >\:, : * ;n .: *,/i. '{<*<; \c\':':?i :**'i ii y.o,: *; ,; c* ; ... ** . i .. **** i .* ... **** ** ,. !,i.e.*, i,i ,.*:, .., /i. .* ,.
1 HU340 13 2 HU330 5 3 HU320 3 4 HU310 13 5 HU300 13 6 HU280 10 7 HU260 21 8 HU240 15 9 CD330 7 10 CD320 11 11 CD310 11 12 CD300 25 13 CD280 21 14 CD260 7 15 CD240 11 16 MOPHU320 0 17 MOPHU310 0 18 MOPHU300 2 19 MOPHU280 8 20 MOPHU260 73 21 MOPHU240 24 22 MOPCD320 0 23 MOPCD310 0 24 MOPCD300 0 25 MOPCD280 28 26 MOPCD260 26 27 MOPCD240 53
<1> The Transient Name is XXnnn, where XX = HU for 1/0 transients that occur during Heatup events, or CD for 1/0 transients that occur during Cooldown events, and nnn = the temperature difference, 11T, between the PZR fluid temperature and the fluid temperature in the surge nozzle. The transient names that begin with "MOP" indicate Modified Operating Procedures to mitigate the effects of 1/0 transient events in the PZR lower head.
2 SPS does not explicitly monitor the breakdown of the transients shown in this table. Rather, the numbers of cycles shown in this table were used in the governing fatigue evaluations for the PZR surge nozzle, and SPS monitors the environmental fatigue usage factor for the surge nozzle as a part of the SPS Fatigue Management Program.
Serial No.22-114 Docket Nos.: 50-280/281 Attachment 3 Page 3 of 3 Table AB. Comparison of SPS, Units 1 and 2, 1/0 transient temperature differences and numbers of cycles with the 1/0 transient data from EPRI Technical Report 3002015905
[3-1]
(2) 340 13 330 600 12 320 3,000 480 (3) 103 1,500 0
<1> t:,. T is the temperature difference between the PZR fluid temperature and the fluid temperature in the surge nozzle.
<2> SPS, Units 1 and 2, have 13 cycles of 340°F ~ T projected to 80 years of operation, which is not enveloped by the transients considered in Reference [3-1], however the EPRI report dispositions a higher temperature difference with significantly more cycles (200 versus 13), and this will bound the SPS units.
3
<> The number of cycles is conservatively equal to the sum of all events in 80 years for 80 years of operation.
REFERENCES 3-1 EPRI, Technical Report 3002015905: Technical Bases for Inspection Requirements for PWR Pressurizer Vessel Head, Shell-to-Head and Nozzle-to-Vessel Welds. Palo Alto, California: 2019.
Serial No.22-114 Docket Nos.: 50-280/281 ATTACHMENT 4 Inspection History SURRY POWER STATION UNITS 1 AND 2 VIRGINIA ELECTRIC AND POWER COMPANY (DOMINION ENERGY VIRGINIA)
Serial No.22-114 Docket Nos.: 50-280/281 Attachment 4 Page 1 of 12 Inspection History Table A9 provides the inspection history for the Surry Power Station (SPS), Units 1 and 2, pressurizer (PZR) and steam generator (SG) welds and components. Complete examination history from the pre-service inspection (PSI) and the first and second 10-year inservice inspection (ISi) intervals is not included in Table A9. Many of these older records are stored off-site and were not readily accessible for inclusion in this summary. However, the onsite ISi Program records document that these examinations have been completed in accordance with ASME Code,Section XI requirements. In addition, several of the older examination reports that were used to populate Table A9 did not record the examination coverage. The examination coverages documented in Table A9 show that the coverages are consistent for the examinations performed throughout the examination history for that weld or component.
Therefore, the coverages for the more recent examinations can be used to determine the examination coverage for the earlier examination where coverage is not documented.
Table A9. PZR and SG Inspection History Serial No.22-114 Docket Nos.: 50-280/281 Attachment 4 Page 2 of 12 Unit 1: Pressurizer (01-RC-E-2) 1:I?Y)/;;t:.,.;;c,;, . . . '*i.[>it .** * * **
1
- .t-'"!'.* *************;., .**. ::;,,; . . . . **/.; *....* ,.,~.z:*~0.,***.**** * * * * * * * ** * . . '** * * ** * **.* *******-*.:-. ,,. ,,.;\;'*;.! .. :,;:*r:;;;;;r: 0
- g,gy~r,!9!h > :.,.:;.;,,:,.,,,..J,; .... N\,*.**.*;
11 1-07 (0 11 -288.9 ) 2 - 1994 12 I P3 Acceptable 51.0% SR-005 1-07 (0"-288.9") 5 - 2003 131 P3 Acceptable 46.0% PRT-05 1-07 (0"-288.9") 11-2013 141 P3 Acceptable 46.0% LMT-C01 1-08 (276"-0"-83") 5 - 1986 12 I P1 Acceptable Note 1 NIA 1-08 (193"-276") 4 - 1992 12 I P3 Acceptable 96.7% NIA 11 11 82.11 1-08 (83 -97 ) 4 - 1992 12 I P3 Acceptable 100% NIA 1-08 (0"-146") 9 - 1995 13 / P1 Acceptable 99.6% NIA 1-08 (145"-290.25") 3 -1997 13 / P2 Acceptable 98.7% N/A 1-08 (0"-145.25") 11 - 2004 14 I P1 Acceptable 99.5% NIA 1-08 (145 11 -290 11 ) 11-2010 14 I P2 Acceptable 99.5% NIA 1-08 (0"-290") 5 - 2018 151 P2 Acceptable 99.4% NIA 1-09 (0"-12") 5 - 1986 12 I P1 Acceptable 100% NIA 1-09 (8"-12") 4 - 1992 12 I P3 Acceptable 100% NIA 1-09 (0"-12") 9 - 1995 13 I P1 Acceptable 100% NIA 1-09 (6"-12") 3 - 1997 13 I P2 Acceptable 100% NIA 1-09 (0"-6") 11 - 2004 14 I P1 Acceptable 100% NIA 82.12 1-09 (6"-12") 11 - 2010 141 P2 Acceptable 100% NIA 1-09 (0"-12") 5 - 2018 151 P2 Acceptable 100% NIA 1-15 (24"-36") 5 - 1986 12 I P1 Acceptable 100.00% NIA 1-15 (0"-12") 5 - 2003 13 / P3 Acceptable 30.5% PRT-05 1-15 (0"-12") 11-2013 141 P3 Acceptable 30.5% LMT-C01 Note 1: Coverage was not documented.
Table A9. PZR and SG Inspection History (continued) Serial No.22-114 Docket Nos.: 50-280/281 Attachment 4 Page 3 of 12 Unit 2: Pressurizer (02-RC-E-2)
I/:1;,tf+/-tf.iAlt[i)i
- r*,;,,
\
- }::\j/?f/'..{t)'.:~':f~ :*_
-;*, ':**i'i,*::,,,:;., *.:/'*' /*,'* '**.,,,',: **
_;:,: ::;'-c,_,.,".J,
.- *********-* **.- ':£k' t'*;i,'.p.e>r,c:,:.f,:*.S*tr,:C*,.*cs,,,,?*x,.
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- "-*',.?"-_*. *, *.:,;::.,-.,.'..::;"
.~k)I}i;f";*[i[t: l~~~f
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_,,*-1 ,. * * ....
.:, . :0-*,->:..:~ ~:,:-~_fY;\!};.~'.
cC-~\< *::*->.*
. .; /'.**. ** * ., *,;t('f.* **;
--Y :*, r:c<;<*
- ,..,-., *....* ::,;;:,.: <:;,':._::,;::,.,,__c **,-, .. ,. ::.,:c,.:,,:
- '->;* ~"*~*-**
- ~* **
dL;f.;, ,t;.::
l,Y
,_,,,,,,:t*>+*<**,_s,::;,:.s,>t*,
IJ\ .. ,,,,._;- *"' './;' .--:. -* . . .*-. .,., __ .-, .,_.;-
-* :--.-~. . ,,._ *. *,.l'..:~*Ltf:J,: ,:
- /'.;
- ,
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5; E~~iJi.i~~t~gt*: :*:::1/2'*:f>.,;., ;!~'..f{~ii:'I 9Q~~r~'9fti
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1-07 (0"-288.9") 2 - 1995 12 I P3 Acceptable 63.5% SR-011 1-07 (0"-288.9") 9 - 2003 131 P3 Acceptable 82.0% NIA 1-07 (0"-288.9") 5 - 2014 14 I P3 Acceptable 82.0% LMT-C01 1-04 3 - 1979 11 I P2 Acceptable Note 1 NIA 1-04 (0"-96") 10 - 1986 12 I P1 Acceptable Note 1 NIA B2.11 1-04 (192"-290.5") 4 - 1993 12 I P3 Acceptable 99.4% NIA 1-04 (0"-145") 5 - 1996 13 I P1 Acceptable 99.0% NIA 1-04 (145"-291 ") 10 - 1997 13 / P2 Acceptable 99.0% N/A 1-04 (0"-145.125") 5 - 2005 141 P1 Acceptable 98.7% NIA 1-04 (145"-291") 11 - 2009 141 P2 Acceptable 98.7% NIA 1-04 (0"-290") 10 -2015 15 I P1 Acceptable 97.1% NIA 1-02 3 - 1979 111 P2 Acceptable Note 1 NIA 1-02 (0"-12") 2 - 1995 121 P3 Acceptable Note 1 SR-011 1-02 (0"-12") 9 - 2003 131 P3 Acceptable 50.0% NIA 1-02 (0"-12") 5 - 2014 141 P3 Acceptable 50.0% LMT-C01 1-03 3 - 1979 111 P2 Acceptable Note 1 NIA B2.12 1-03 (0"-12") 10 -1986 12 I P1 Acceptable Note 1 NIA 1-03 (0"-6") 5 - 1996 131 P1 Acceptable 100% NIA 1-03 (6"-12") 10 - 1997 131 P2 Acceptable 99% NIA 1-03 (0"-12") 10 - 2006 14 I P1 Acceptable >90.0% NIA 1-03 (0"-12") 11 - 2009 14 I P2 Acceptable 95.0% N/A 1-03 (0"-12") 10-2015 15 I P1 Acceptable 100.0% NIA Note 1: Coverage was not documented.
Table A9. PZR and SG Inspection History (continued) Serial No.22-114 Docket Nos.: 50-280/281 Attachment 4 Page 4 of 12 Unit 1: 'A' Steam Generator (01-RC-E-1A)
- ?*
- .{~ . '
,.,., :*,,,Vi' 'ft ,,,., '*'"),.,,,i,,'Cf '*' *f,;,;:;,;;,
,1;
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- .J**'f ',,',*< ". ;,::,:,*, "':'.'.:~: 1,* ...
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'.c;'i'\C:,, ~*:, ;,* ~t,;c;;
1.:.:; ,,,,.:,:,,..,., .,,,, ;, ;*.*:;;,':
- ,, i' f
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...,')::} ,,,:)".' ,::<?;,,
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),, >*,:;;:-, ' 1
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*/ '( ,., ,,,,.::,.:, ,.,,
""'"\'.<
,, :*;/\;;;:'; };:'.',
,: ri (:'. '
/,c,, /')<;: ,,;,,-:-., <;,":fyi:, '*_:":':.,:.'*' ~:il-'."i'
;..'/'."'/:,:\ I ,,,
.~, ~ ... -., \;;':,:C ,: '
,:,C,:*' >,:,,* .. , '""'
c..* ',,
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PRIMARY SIDE SHELL WELDS 1-01 (0 11 -142 11 ) 9 - 1995 13 / P1 Acceptable 97.4% N/A 11 11 1-01 (142 -284 ) 3 - 1998 13/ P2 Acceptable 98.00% NIA 11 11 1-01 (284 -426.7 ) 4 - 2003 13 / P3 Acceptable 99.6% N/A 11 11 1-01 (0 -142 ) 11 - 2004 14 / P1 Acceptable 99.6% N/A 82.40 11 11 1-01 (142 -284 ) 11 - 2007 14/ P2 Acceptable 94.8% N/A 11 1-01 (284 -426.7") 11-2013 14/ P3 Acceptable 91.4% N/A 1-01 (0"-426.7") 4 - 2015 15 / P1 Acceptable 97.3% N/A SECONDARY SIDE SHELL WELDS 2-03 (0"-142") 9 - 1995 13 / P1 Acceptable 100% N/A 11 11 2-03 {0 -142 ) 11 - 2004 14 / P1 Acceptable 100% N/A 2-03 (291 "-431 ") 10 - 2016 15 / P1 Acceptable 100% N/A 11 2-03 (147"-291 ) 10 - 2016 15 / P1 Acceptable 100% N/A 2-03 (0"-147") 10-2016 15 / P1 Acceptable 100% N/A 2-05 2 - 1994 13 / P1 Acceptable Note 1 N/A C1.10 2-05 (0"-142") 10 -1995 13 / P1 Acceptable 98.9% N/A 2-05 (0"-142 11 ) 5 - 2006 14 / P1 Acceptable 100% N/A 2-06 (370"-522.2") 2 - 1994 13 / P1 Acceptable 98.6% N/A 11 2-06 (0"-184 ) 9 - 1995 13 / P1 Acceptable 98.9% N/A 2-06 (0"-184") 5 - 2006 14 / P1 Acceptable 97.0% N/A 11 11 2-06 (0 -184 ) 4 - 2015 15 / P1 Acceptable 98.9% N/A
Table A9. PZR and SG Inspection History (continued) Serial No. 22-114 Docket Nos.: 50-280/281 Attachment 4 Page 5 of 12 Unit 1: 'A' Steam Generator (01-RC-E-1A) (continued)
. * ,. " .,.> ; "/~~tw}P }\*t;}f;***'i?/:.:::. :, ,.*.* :~~; .*,. . *******
",'.:<,"*.;'.>'ft>'<'*' ,.'{' *" ***** :; .,... ,.** ... *'* .:,,. <2 ..'> ,, ;,., ,, *. . */C/J§*c> t':c./.;.;'" ' ' L_*,i'!,i:,:'.i
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' 1I::*.; ,)fi; .,. ,.,. ,,, **;*)' ..( *,,.<: .*., .. ,.,,,,,.,.:/: ex~:ni111*afi'8falb .,._/'; '\c'.i **;_:,*,*.o,';
){;i'.:..i;J.c:/;,
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l*'i,**. ., c:'*',.:'**c* .. ,*. *,.**,?'r£;,~;;y;,t ::s*,,,,::.*.. '*>. *.,,,,,,,.... *a ., * ,.
,., **.** ,,,..,,, ..""'*':-.c;:;, . .,
',i>:\'.':..;;:,::. *.*; .*,.:,.,, ,., . , .,,,.,
;'.}.!.~ -,;--.- -,,,.~,.,;- *c,_.: . -~~:'.'*-"-'-c*c.: . ,,.*,;::, -., .,.*
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.,. " ..x..;:.,:::; ..... ,.,,;.,::;-c,,.\ .. , **** ...*,**H.c* **.*'.:*******"'**"'***,. r;;r..:,;,::::, tJ'.Q9¥~f!ij~'; -~:?/:
)i'//;',, . ,,,,. ,,,.,,,.:;,::,.;;. :
2-08 (0"-184") 9 - 1995 13 / P1 Acceptable 99.7% N/A C1.20 2-08 (0"-184") 5 - 2006 14 / P1 Acceptable 99.6% N/A 2-02 (0"-142") 9 - 1995 13 / P1 Acceptable 97.8% N/A C1.30 2-02 (0"-142") 11 - 2004 14 / P1 Acceptable 99.0% N/A 2-02 (0"-428") 4 - 2015 15 / P1 Acceptable 94.9% N/A SECONDARY SIDE NOZZLES 2-09 (0"-88.2") 2 - 1994 13 / P1 Acceptable 92.5% N/A 11 11 2-09 (0 -42 ) 9 - 1995 13 / P1 Acceptable 99.6% N/A C2.21 2-09 (0"-16") 11 - 2004 14 / P1 Acceptable 73.2% NIA 2-10 (0"-33") 9 - 1995 13 / P1 Acceptable 100% N/A 2-10 (0°-120°) 5 - 2006 13 / P1 Acceptable 100% N/A 1-RC-2-01 CNIR (0"-33") 9 - 1995 13 / P1 Acceptable 100% N/A 1-RC-2-01 CNIR (0"-60") 5 - 2006 14 / P1 Acceptable 100% N/A C2.22 1-RC-2-01 DNIR (0"-44") 9 - 1995 13 / P1 Acceptable 100% N/A 1-RC-2-01DNIR (0"-16") 11 - 2006 14 / P1 Acceptable 100% N/A Note 1: Coverage was not documented.
Table A9. PZR and SG Inspection History (continued) Serial No. 22-114 Docket Nos.: 50-280/281 Attachment 4 Page 6 of 12 Unit 1: '8' Steam Generator (01-RC-E-1 B) l~t"J"::sr;::tr::i;:;\.*gi;...., *.*. .*...,
.I c1:1:c:Y::>1;ci'. ,ev:>;;;,;.,r,sc.,11,;,1s;,1: f<<i~:c,;t,,, ':;C?J};*J'.;' 1.,,,. * , . ... ...... ., .
***,::-_.,~,-:_ '..'"*,< -1*:'::!':_.,..
----1,-,_. '.**is
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.>,-:*:,:'>t:\.':.."77\1.~:'."<,-
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SECONDARY SIDE SHELL WELDS 2-03 (142"-284") 3 - 1997 13 / P2 Acceptable 100% N/A 2-03 (142"-284") 11 - 2004 14 / P1 Acceptable 100% N/A 2-03 (142"-284") 11-2010 14/ P2 Acceptable 100% NIA 2-05 (142"-284") 3 - 1997 13/ P2 Acceptable 98.7% N/A 2-05 (142"-284") 10-2010 14/ P2 Acceptable >90% N/A C1.10 2-05 (0"-426") 5 - 2018 15 / P2 Acceptable 93.9% N/A 2-06 (185"-370") 6 - 1986 12 / P2 Note 2 Note 1 N/A 11 2-06 (185 -370") 5 - 1988 12 / P2 Note 2 Note 1 N/A 2-06 (184"-368") 3 - 1997 13/ P2 Note 2 99.0% N/A 2-06 (184"-368") 11-2010 14/ P2 Note 2 97.8% N/A 2-06 (184"-368") 4 - 2018 15 / P2 Note 2 98.7% N/A C1.20 2-08 (184"-368") 3- 1997 13 / P2 Acceptable 99.0% N/A 2-02 (148"-284") 3 - 1997 13 / P2 Acceptable 97.3% N/A C1.30 2-02 (148"-284") 10 - 2010 14/ P2 Acceptable 97.3% N/A SECONDARY SIDE NOZZLES 2-09 (16"-33") 3 - 1997 13 / P1 Acceptable 100% N/A 2-09 11 - 2010 14 / P2 Acceptable 96.0% N/A C2.21 2-09 4 - 2018 15 / P2 Acceptable 100% N/A 2-10 (33"-67") 3 - 1997 13 / P1 Acceptable 100% N/A 1-RC-2-02CNIR (33"-67") 3 - 1997 13 / P2 Acceptable 100% N/A 11 11 1-RC-2-02CNIR (16 -33 ) 4- 1997 13 / P2 Acceptable 100% N/A C2.22 1-RC-2-02DNIR 11 - 2010 14/ P2 Acceptable 100% N/A 1-RC-2-01 DNIR 4 - 2018 15 / P2 Acceptable 100% N/A Note 1: Coverage was not documented. Note 2: Recordable subsurface fabrication type indications.
Table A9. PZR and SG Inspection History (continued) Serial No. 22-114 Docket Nos.: 50-280/281 Attachment 4 Page 7 of 12 Unit 1 - 'C' Steam Generator (01-RC-E-1C) __ ,., ,,, ,.,,,,*,,,,,,, ,i, ./i 1?1.1.*
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2-09 (33"-0") 10 -2001 13 / P2 Acceptable 99% N/A C2.21 2-10 (67"-0") 11 - 2001 13/ P2 Acceptable 100% N/A 1-RC-2-03CNIR (33"-67") 5 - 2003 13 / P3 Acceptable 100% N/A C2.22 1-RC-2-03DNIR (33"-0") 10 - 2001 13 / P2 Acceptable 100% N/A Note 1: Coverage was not documented. Note 2: Recordable subsurface fabrication type indications.
Table A9. PZR and SG Inspection History (continued) Serial No. 22-114 Docket Nos.: 50-280/281 Attachment 4 Page 8 of 12 Unit 2: 'A' Steam Generator (02-RC-E-1A)
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PRIMARY SIDE SHELL WELDS 1-01 (0"-142") 2 - 1995 13 / P1 Acceptable 97.1% N/A 1-01 (142"-284") 10 - 2000 13 / P2 Acceptable 98.5% N/A 1-01 (284"-426.7") 9 - 2003 13 / P3 Acceptable 98.7% N/A B2.40 1-01 (0"-142") 4 - 2005 14 / P1 Acceptable 98.9% N/A 1-01 (142"-284") 4 - 2011 14 / P2 Acceptable 99.0% NIA 1-01 (284"-426.9") 11-2012 141 P3 Acceptable 98.8% NIA 1-01 (0"-426") 5 - 2020 15 / P2 Acceptable 99.5% N/A SECONDARY SIDE SHELL WELDS 2-03 (0"-142") 2 - 2005 13 / P1 Acceptable 100% N/A 2-03 (284"-0") 2 - 2005 13 / P1 Acceptable 100% N/A 2-03 (0"-142") 11 - 2006 14 / P1 Acceptable 100% N/A 2-03 (O" -426") 5 - 2020 15 / P2 Acceptable 100% N/A 2-05 (0"-142") 2 - 1995 13 / P1 Acceptable 100% N/A C1.10 2-05 (284"-0") 2 - 1995 13 / P1 Acceptable 100% N/A 2-05 (0"-142") 11 - 2006 14 / P1 Acceptable 100% N/A 2-06 (0"-184") 2 -1995 13 / P1 Acceptable 98.1% N/A 2-06 (0"-184") 11 - 2006 14 / P1 Acceptable >90% N/A 2-06 (0"-184") 10-2015 15 / P1 Acceptable 98.9% NIA 2-08 (0"-184") 2 - 1995 13 / P1 Acceptable 100% N/A C1.20 2-08 (370"-0") 2 - 1995 13 / P1 Acceptable 100% N/A
Table A9. PZR and SG Inspection History (continued) Serial No. 22-114 Docket Nos.: 50-280/281 Attachment 4 Page 9 of 12 Unit 2: 'A' Steam Generator (02-RC-E-1A) (continued)
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- 2-02 (0"-142") 2 - 1995 13 / P1 Acceptable 96.5% N/A C1.30 2-02 (0"-142") 4 - 2005 14 / P1 Acceptable 96.8% N/A 2-02 (0"-428") 5 - 2020 15 I P2 Acceptable 98.1% N/A SECONDAR.YSIDE NOZZLES . .
2-09 2 - 1995 13 / P1 Acceptable 100% N/A C2.21 2-09 11 - 2006 14 / P1 Acceptable 98.7% N/A 2-10 (0"-33") 2 - 1995 13 / P1 Acceptable 100% N/A 1-RC-2-01 CNIR (0"-33") 4 - 2005 13 I P1 Acceptable 100% N/A 1-RC-2-01 CN IR (0"-60") 11 - 2006 14 I P1 Acceptable 100% N/A C2.22 1-RC-2-01 DNIR (0"-16") 4 - 2005 14 I P1 Acceptable 100% N/A 1-RC-2-01 DNIR 11 - 2006 14 / P1 Acceptable 100% N/A
Table A9. PZR and SG Inspection History (continued) Serial No. 22-114 Docket Nos.: 50-280/281 Attachment 4 Page 10 of 12 Unit 2: 'B' Steam Generator (02-RC-E-1 B)
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- ..*.,,,i SECONDARY SIDE SHELL WELDS 2-03 (0"-30") 1 - 1980 11 / P3 Acceptable Note 1 N/A 2-03 (142"-172") 1 - 1980 11 / P3 Acceptable Note 1 N/A 2-03 (284"-341 ") 1 - 1980 11 / P3 Acceptable Note 1 N/A 2-03 (141"-284") 10 - 1997 13/ P2 Acceptable 100% N/A 2-03 (141"-282") 11 -2009 14/ P2 Acceptable 100% N/A 2-05 (141"-282") 10-1997 13 I P2 Acceptable 99.7% N/A C1.10 2-05 (141"-282") 4 - 2011 14/ P2 Acceptable 100% N/A 2-05 11-2015 15 / P1 Acceptable 100% N/A 2-06 5 - 1985 12 / P1 Note 2 92.9% N/A 2-06 (368"-522") 3 - 1993 12 / P3 Acceptable 92.9% N/A 2-06 (184"-368") 4 - 1999 13 / P2 Note 2 97.2% N/A 2-06 (184"-368") 4 - 2011 14/ P2 Note 2 97.8% N/A 2-06 (184"-368") 11-2018 15 / P2 Note 2 98.7% N/A 2-08 (156"-370") 9 - 1988 12 / P2 Acceptable 99.0% N/A C1.20 2-08 (184"-368") 4 - 2000 13 / P2 Acceptable 99.0% N/A 2-08 5 - 2014 14/ P3 Acceptable 99.0% N/A C1.30 2-02 (141 "-284") 10 - 1997 13 / P2 Acceptable 98.3% N/A 2-02 (141"-284") 11 - 2009 14/ P2 Acceptable 98.3% NIA
Table A9. PZR and SG Inspection History (continued) Serial No. 22-114 Docket Nos.: 50-280/281 Attachment 4 Page 11 of 12 Unit 2: 'B' Steam Generator (02-RC-E-1 B) (continued) 2-09 (0"-16") 5 - 2005 14 / P1 Acceptable > 90% N/A 2-09 11-2018 15 / P2 Acceptable 92.1% N/A C2.21 2-10 (0"-33") 5 - 1996 13 / P1 Acceptable 100% N/A 2-10 5 - 2014 14/ P3 Acceptable 92.1% N/A 1-RC-2-02CNIR (33"-67") 4 - 1999 13 / P2 Acceptable 100% N/A 1-RC-2-02CNIR 11 - 2009 14/ P2 Acceptable 100% N/A C2.22 1-RC-2-01 DNIR (16"-33") 4- 1999 13/ P2 Acceptable 100% NIA 1-RC-2-01 DNIR 11 - 2009 14/ P2 Acceptable 100% N/A 1-RC-2-01 DNIR 11-2018 15/ P2 Acceptable 100% N/A Note 1: Coverage was not documented. Note 2: Recordable subsurface fabrication type indications.
Table A9. PZR and SG Inspection History (continued) Serial No. 22-114 Docket Nos.: 50-280/281 Attachment 4 Page 12 of 12 Unit 2: 'C' Steam Generator (02-RC-E-1C) SECONDARY SIDE SHELL WELDS 2-03 (284"-0") 10 - 2003 13/ P2 Acceptable 100% N/A 2-03 (284"-0") 11-2012 14/ P3 Acceptable 100% N/A 2-05 (284"-0") 4 - 2002 13 / P3 Acceptable 100% N/A C1.10 2-05 (284"-0") 11 - 2012 14/ P3 Acceptable 100% N/A 2-06 (368"-522.2") 10 - 2003 13 / P3 Acceptable 99.8% N/A 2-06 (368"-522.2") 5 - 2014 14/ P3 Acceptable 99.8% N/A C1.20 2-08 (368"-0") 10 - 2003 13 / P3 Acceptable 99.6% N/A 2-02 (284"-21 ") 3 - 1993 12 / P3 Acceptable 92.0% N/A C1.30 2-02 (284"-0") 3 - 2002 13 / P3 Acceptable 98.2% N/A 2-02 (284"-0") 11-2012 14 / P3 Acceptable 98.9% N/A
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Serial No. 22-114 Docket Nos.: 50-280/281 ATTACHMENT 5 Results of Industry Survey SURRY POWER STATION UNITS 1 AND 2 VIRGINIA ELECTRIC AND POWER COMPANY (DOMINION ENERGY VIRGINIA)
Serial No. 22-114 Docket Nos.: 50-280/281 Attachment 5 Page 1 of 5 Overall Industry Inspection Summary for Pressurizer Code Item Nos. 82.11, 82.12, 82.21, 82.22, and 83.110 The results of an industry survey of past pressurizer (PZR) weld inspections are summarized in Electric Power Research lnstitute's (EPRl's) Technical Report 3002015905, "Technical Bases for Inspection Requirements for PWR Pressurizer Vessel Head, Shell-to-Head and Nozzle-to-Vessel Welds" [5-1]. A total of 47 domestic and international pressurized water reactor (PWR) units responded to the survey. The survey represented all PWR plant designs currently in operation in the United States, including 2-loop, 3-loop, and 4-loop PWR designs, from each of the PWR nuclear steam supply system (NSSS) vendors (i.e., Babcock and Wilcox (B&W), Combustion Engineering (CE), and Westinghouse). The combined survey results for Item Nos. 82.11, 82.12, 82.21, 82.22, and B3.110 are summarized in Table A 10 below. A total of 1,162 examinations of PWR PZR components were reported by the survey for the affected Item Nos. Of the 1,162 total examinations, only four (4) examinations identified flaws exceeding the acceptance criteria of ASME Code, Section XI. All four (4) flaw indications for Item No. 82.11 occurred at two (2) units of a single plant site. None of these flaws were found to be service induced. Flaw evaluations were performed to show acceptability of these indications and follow-on examinations showed no change in flaw sizes since the original inspections. No other indications were identified in any in-scope components. The results in Table A10 indicate that the number of reportable indications resulting from examinations of the PWR PZR components for the affected Item Nos. is negligible. Therefore, the increase in worker radiation exposure, risk to personnel safety, and production of radwaste resulting from these examinations adversely impacts outage-related activities without a corresponding increase in the level of quality or safety. Table A 10. Summary of survey results for Item Nos. 82.11, 82.12, 82.21, 82.22, and 83.110 Number of Number of Item No. Examinations Reportable Indications 82.11 269 4 (1) 82.12 269 0 82.21 4 0 82.22 30 0 83.110 590 0 1
<> Flaw evaluations were performed to show acceptability of these indications and follow-on examinations showed no change in flaw sizes since the original inspections.
Serial No. 22-114 Docket Nos.: 50-280/281 Attachment 5 Page 2 of 5 Overall Industry Inspection Summary for Steam Generator Code Item Nos. 82.31, 82.32, 82.40, 83.130, C1.10, C1 .20, and C1 .30 The results of an industry survey of past inspections of steam generator (SG) nozzle-to-shell welds, inside radius sections, and shell welds are summarized in EPRl's Technical Report 3002015906, "Technical Bases for Inspection Requirements for PWR Steam Generator Class 1 Nozzle-to-Vessel Welds and Class 1 and Class 2 Vessel Head, Shell, Tubesheet-to-Head and Tubesheet-to-She/1 Welds" [5-2]. A total of 74 domestic and international 8WR and PWR units responded to the survey. The survey represented all PWR plant designs currently in operation in the United States, including 2-loop, 3-loop, and 4-loop PWR designs from each of the PWR NSSS vendors (i.e., 8&W, CE, and Westinghouse). The combined survey results for Item Nos. 82.31, 82.32, 82.40, 83.130, C1.10, C1 .20, and C1 .30 are summarized in Table A 11 below. A total of 1,324 examinations were reported by the survey for the components of the affected Item Nos., with 1,098 of these specifically for PWR components. The majority of the PWR examinations were performed on SG welds. A small number of flaws were identified during these examinations which required flaw evaluation. None of these flaws were found to be service-induced. For Item No. 82.40, examinations at two (2) units at a single plant site identified multiple flaws exceeding the acceptance criteria of ASME Code, Section XI; however, these were determined to be subsurface-embedded fabrication flaws and not service-induced (see Table Note 1). For Item No. C1 .20, two (2) PWR units reported flaws exceeding the acceptance criteria of ASME Code, Section XI. In the first unit, a single flaw was identified and evaluated as an inner diameter surface imperfection. Reference [5-4] indicates that this was a spot indication with no measurable through-wall depth. This indication is therefore not considered to be service-induced but rather fabrication-related. A flaw evaluation performed in accordance with IWC-3600 determined this indication was acceptable for continued operation. In the second unit, multiple flaws were identified (see Table Note 2). As discussed in References [5-5] and [5-6], these flaws were most likely subsurface weld defects typical of thick vessel welds and not service-induced. A flaw evaluation performed in accordance with IWC-3600 determined these flaws to be acceptable for continued operation. The results of the industry survey identified numerous SG examinations being performed with no service-induced flaws being detected. The results in Table A 11 indicate that the number of reportable indications resulting from examinations of the PWR SG components for the affected Item Nos. is negligible. Therefore, the increase in worker radiation
Serial No. 22-114 Docket Nos.: 50-280/281 Attachment 5 Page 3 of 5 exposure, risk to personnel safety, and production of radwaste resulting from these examinations adversely impacts outage-related activities without a corresponding increase in level of quality or safety. Table A11. Summary of survey results for Item Nos. 82.31, 82.32, 82.40, 83.130, C1.10, C1 .20, and C1 .30 82.31 0 30 30 0 0 0 82.32 0 13 13 0 0 0 (1) (1) 82.40 0 183 183 0 83.130 0 135 135 0 0 0 C1.10 140 305 445 0 0 0 (2) (2) C1.20 54 319 373 0 C1.30 32 113 145 0 0 0 (1) (2) (1) (2) Totals 226 1098 1324 0
<1> Two (2) PWR Westinghouse 2-Loop units at a single plant reported multiple subsurface embedded fabrication flaws.
2
< > A single PWR Westinghouse 2-Loop unit reported multiple flaws [5-5 and 5-6].
Serial No. 22-114 Docket Nos.: 50-280/281 Attachment 5 Page 4 of 5 Overall Industry Inspection Summary for Steam Generator Code Item Nos. C2.21, C2.22, and C2.32 The results of an industry survey of past inspections of SG feedwater (FW) and main steam (MS) nozzles are summarized in EPRl's Technical Report 3002014590, "Technical Bases for Inspection Requirements for PWR Steam Generator Feedwater and Main Steam Nozzle-to-Shell Welds and Inside Radius Sections" [5-3]. A total of 74 domestic and international BWR and PWR units responded to the survey. The survey represented all PWR plant designs currently in operation in the United States including two (2)-loop, three (3)-loop, and four (4)-loop PWR designs from each of the PWR NSSS vendors (i.e., B&W, CE, and Westinghouse). The combined survey results for Item Nos. C2.21, C2.22, and C2.32 are summarized in Table A 12 below. A total of 727 examinations for Item Nos. C2.21, C2.22, and C2.32 components were conducted, with 563 of these specifically for PWR components. The majority of the PWR examinations were performed on SG FWand MS nozzles. Only one (1) PWR examination identified two (2) flaws that exceeded ASME Code, Section XI acceptance criteria. The flaws were linear indications of 0.3 inches and 0.5 inches in length and were detected in a MS nozzle-to-shell weld using magnetic particle examination techniques. The indications were dispositioned by light grinding [5-4]. The results of the industry survey identified numerous SG FW and MS nozzle-to-shell welds and nozzle inside radius section examinations being performed with no service-induced flaws being detected. The results in Table A 12 indicate that the number of reportable indications resulting from examinations of the SG FW and MS nozzles for the affected Item Nos. is negligible. Therefore, the increase in worker radiation exposure, risk to personnel safety, and production of radwaste resulting from these examinations adversely impacts outage-related activities without a corresponding increase in level of quality or safety. Table A12. Summary of survey results for Item Nos. C2.21, C2.22, and C2.32
'Ntimb~r*()t> ,*.; (}}c::t~:;:1,I}I'~,m,§~~:.~(J'~i,:,,;/\'i 164 0 PWR 2 Totals 2
Serial No. 22-114 Docket Nos.: 50-280/281 Attachment 5 Page 5 of 5 REFERENCES 5-1 EPRI, Technical Report 3002015905: Technical Bases for Inspection Requirements forPWR Pressurizer Vessel Head, Shell-to-Head and Nozzle-to-Vessel Welds. Palo Alto, California: 2019. 5-2 EPRI, Technical Report 3002015906: Technical Bases for Inspection Requirements for PWR Steam Generator Class 1 Nozzle-to-Vessel Welds and Class 1 and Class 2 Vessel Head, Shell, Tubesheet-to-Head and Tubesheet-to-Shell Welds. Palo Alto, California: 2019. 5-3 EPRI, Technical Report 3002014590: Technical Bases for Inspection Requirements for PWR Steam Generator Feedwater and Main Steam Nozzle-to-Shell Welds and Inside Radius Sections. Palo Alto, California: 2019. 5-4 Letter from F. A. Kearney (Exelon) to NRC, "Byron Station Unit 2 90-Day lnservice Inspection Report for Interval 3, Period 3, (B2R17)," dated July 29, 2013. (ADAMS Accession Number ML13217A093) 5-5 Letter from J. M. Sorensen (Nuclear Management Company, LLC) to NRC, "Unit 1 lnservice Inspection Summary Report, Interval 3, Period 3 Refueling Outage Dates 1-19-2001 to 2-25-2001 Cycle 20 / 05-26-99 to 02-25-2001," dated May 29, 2001. (ADAMS Accession Number ML011550346) 5-6 Letter from J. P. Solymossy (Nuclear Management Company, LLC) to NRC, "Response to Opportunity for Comment on Task Interface Agreement (TIA) 2003-01, 'Application of ASME Code Section XI, IWB-2430 Requirements Associated With Scope of Volumetric Weld Expansion at the Prairie Island Nuclear Generating Plant' (Tac Nos. MB7294 and MB7295)," dated April 4, 2003. (ADAMS Accession Number ML031040553)}}