ML22160A477
| ML22160A477 | |
| Person / Time | |
|---|---|
| Site: | Summer  |
| Issue date: | 06/09/2022 |
| From: | Lawrence D Dominion Energy South Carolina |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| 22-115 | |
| Download: ML22160A477 (53) | |
Text
Dominion Energy South Carolina, Inc.
5000 Dominion Boulevard, Glen Allen. VA 23060 DominionEnergy.com June 9, 2022 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20555-0001 DOMINION ENERGY SOUTH CAROLINA, INC. (DESC)
VIRGIL C. SUMMER NUCLEAR STATION UNIT 1 Dominion Energy" Serial No.:
22-115 NRA/ENC:
RO Docket No.:
50-395 License No.:
NPF-12 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), Dominion Energy South Carolina (DESC) requests Nuclear Regulatory Commission (NRG) approval of a proposed alternative to the inservice inspection (ISi) requirements for American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code)Section XI, Table IWB-2500-1, Examination Category B-8 and B-D and Table IWC-2500-1, Examination Category C-A and C-8, component examinations for Virgil C.
Summer Nuclear Station Unit 1 (VCSNS). Specifically, DESC requests to defer examinations for the remainder of the current fourth 10-year ISi interval through the fifth 10-year ISi interval ending on December 31, 2033. 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 VCSNS is provided in Attachments 2 and 3. The VCSNS 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 NRG review and approval before implementation. DESC requests NRG approval of this request by April 1, 2023, to support the VCSNS Spring 2023 refueling outage.
Serial No.: 22-115 Docket No.: 50-395 Page 2 of 3 If you have any questions or require additional information, please contact Ms. Erica N.
Combs at (804)-273-3386.
Sincerely, Douglas C. Lawrence Vice President - Nuclear Engineering & Fleet Support Dominion Energy South Carolina 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 Commitments made in this letter: None
cc:
Regional Administrator, Region II U. S. Nuclear Regulatory Commission Marquis One Tower 245 Peachtree Center Ave., NE, Suite 1200 Atlanta, Georgia 30303-1257 Mr. G. Edward Miller NRG Senior Project Manager - Virgil C. Summer Nuclear Station U.S. Nuclear Regulatory Commission Mail Stop 09 E-3 One White Flint North 11555 Rockville Pike Rockville, Maryland 20852-2738 NRG Senior Resident Inspector Virgil C. Summer Nuclear Station Mr. G. J. Lindamood Santee Cooper - Nuclear Coordinator 1 Riverwood Drive Moncks Corner, South Carolina 29461 Serial No.: 22-115 Docket No.: 50-395 Page 3 of 3
ATTACHMENT 1 Serial No.: 22-115 Docket No.: 50-395 Proposed Alternative to ASME Code Section XI Requirements for lnservice Inspection of Pressurizer and Steam Generator Pressure-Retaining Welds and Full Penetration Welded Nozzles VIRGIL C. SUMMER NUCLEAR STATION UNIT 1 (VCSNS)
DOMINION ENERGY SOUTH CAROLINA, INC. (DESC)
Serial No.: 22-115 Docket No.: 50-395 Page 1 of 17 Proposed Alternative to ASME Code Section XI Requirements for lnservice Inspection of Pressurizer and Steam Generator Pressure-Retaining Welds and Full Penetration Welded Nozzles In accordance with 10 CFR 50. 55a(z)(1 ), the proposed alternative provides an 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 full penetration welded nozzles, 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 and full penetration welded nozzles Description SG pressure-retaining vessel welds and full penetration welded nozzles Class 1, Category 8-8, pressure-retaining welds in vessels other than reactor vessels Examination Class 1, Category 8-D, full penetration welded nozzles in vessels Categories Class 2, Category C-A, pressure-retaining welds in pressure vessels Class 2, Category C-8, pressure-retaining nozzle welds in vessels 82.11 PZR, shell-to-head welds, circumferential 82.12 PZR, shell-to-head welds, longitudinal 82.40 SG (primary side), tubesheet-to-head weld 83.110 PZR, nozzle-to-vessel welds Item C1.10 Shell circumferential welds Numbers C1.20 Head circumferential welds C1.30 Tubesheet-to-shell weld C2.21 Nozzle-to-shell (nozzle-to-head/nozzle-to-nozzle) welds C2.22 Nozzle inside radius sections
Serial No.: 22-115 Docket No.: 50-395 Page 2 of 17 Table 2. Affected component IDs ASME ASME Component ID Component Description Category Item No.
Pressurizer 8-8 82.11 1-2100A-1 PZR Shell to Lower Head 8-8 82.11 1-2100A-4 PZR Shell to Upper Head 8-8 82.12 1-2100A-5 PZR Shell Longitudinal Weld - Lower 8-8 82.12 1-2100A-7 PZR Shell Longitudinal Weld - Upper 8-D 83.110 1-2100A-8 PZR Surge Line Nozzle to Vessel Weld 8-D 83.110 1-21008-9 PZR Spray Line Nozzle to Vessel Weld 8-D 83.110 1-21008-10 PZR 'A' Safety Line Nozzle to Vessel Weld 8-D 83.110 1-21008-11 PZR '8' Safety Line Nozzle to Vessel Weld 8-D 83.110 1-21008-12 PZR 'C' Safety Line Nozzle to Vessel Weld 8-D 83.110 1-21008-13 PZR Relief Line Nozzle to Vessel Weld Steam Generator 'A' 8-8 82.40 1-3100-14A SG Primary Head to Tubesheet C-A C1.10 2-1100-17A SG Shell to Lower Transition Cone C-A C1.10 2-1100-18A SG Shell to Upper Transition Cone C-A C1.20 2-1100-20A SG Shell to Upper Head C-A C1.30 2-1100-15A SG Shell to Tubesheet C-8 C2.21 2-1100-23A SG Shell to Feedwater Nozzle C-8 C2.22 2-1100-231 R-A SG Shell to Feedwater Nozzle Inner Radius Steam Generator 'B' 8-8 82.40 1-3100-148 SG Primary Head to Tubesheet C-A C1.10 2-1100-178 SG Shell to Lower Transition Cone C-A C1.10 2-1100-188 SG Shell to Upper Transition Cone C-A C1.20 2-1100-208 SG Shell to Upper Head C-A C1.30 2-1100-158 SG Shell to Tubesheet C-8 C2.21 2-1100-238 SG Shell to Feedwater Nozzle C-8 C2.22 2-1100-23IR-8 SG Shell to Feedwater Nozzle Inner Radius Steam Generator 'C' 8-8 82.40 1-3100-14C SG Primary Head to Tubesheet C-A C1.10 2-1100-17C SG Shell to Lower Transition Cone C-A C1.10 2-1100-18C SG Shell to Upper Transition Cone C-A C1.20 2-1100-20C SG Shell to Upper Head C-A C1.30 2-1100-15C SG Shell to Tubesheet C-8 C2.21 2-1100-23C SG Shell to Feedwater Nozzle C-8 C2.22 2-1100-231 R-C SG Shell to Feedwater Nozzle Inner Radius
Serial No.: 22-115 Docket No.: 50-395 Page 3 of 17 2.0
Applicable Code Edition and Addenda
The Code of record for the Virgil C. Summer Nuclear Station Unit 1 (VCSNS) fourth 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," 2007 Edition with 2008 Addenda [1-1]. The VCSNS fourth 10-year ISi interval started on January 1, 2014, and ends on December 31, 2023.
3.0
Applicable Code Requirement
The ASME Code,Section XI, IWB-2500(a), Table IWB-2500-1, Examination Categories B-8 and B-O and IWC-2500(a), Table IWC-2500-1, Examination Categories C-A and C-8 require examination of the following Item Nos.:
82.11 B2.12 82.40 83.110 C1.10 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.
Volumetric examination of one (1) foot of all longitudinal shell-to-head welds during the first inspection interval and one (1) foot of one (1) weld per head during successive intervals. The examination volume is shown in Figure IWB-2500-2.
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.
Volumetric examination of all full penetration nozzle-to-vessel welds during each inspection interval. The examination volume is shown in Figures IWB-2500-?(a), (b), and (c).
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 C1.30 C2.21 C2.22 Serial No.: 22-115 Docket No.: 50-395 Page 4 of 17 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 (1) vessel or distributed among the vessels. The examination volume is shown in Figure IWC-2500-2.
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).
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 7 4 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
Serial No.: 22-115 Docket No.: 50-395 Page 5 of 17 to plant safety. Based on the conclusions of the three EPRI reports, Dominion Energy South Carolina, Inc. (DESC) is requesting an alternative to the 10-year inspection interval for the subject welds.
5.0 Proposed Alternative and Basis for Use DESC 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 Nos.:
ASME Item No.
Description Category 8-8 82.11 PZR, shell-to-head welds, circumferential 8-8 82.12 PZR, shell-to-head welds, longitudinal 8-8 82.40 SG (primary side), tubesheet-to-head weld 8-D 83.110 PZR, nozzle-to-vessel welds C-A C1.10 Shell circumferential welds C-A C1.20 Head circumferential welds C-A C1.30 Tubesheet-to-shell weld C-8 C2.21 Nozzle-to-shell (nozzle-to-head or nozzle-to-nozzle) welds C-8 C2.22 Nozzle inside radius sections VCSNS is currently in the third period of the fourth 10-year ISi interval. Some welds have not yet received an ISi examination for the fourth interval. The proposed alternative increases the inspection interval for these examination items from the current ASME Code,Section XI 10-year requirement thereby deferring examinations for the remainder of the current fourth 10-year ISi interval through the fifth 10-year ISi interval ending on December 31, 2033. During the sixth 10-year ISi interval starting on January 1, 2034, examinations will be done in accordance with ASME Code,Section XI ISi 10-year requirements. The sixth 10-year ISi interval ends on December 31, 2043.
A summary of the technical basis for this request is provided below. The applicability of the technical basis to VCSNS is demonstrated in Attachments 2, 3, and 4.
A. Degradation Mechanism Evaluation Serial No.: 22-115 Docket No.: 50-395 Page 6 of 17 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 (FAG), 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 those 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 for VCSNS, therefore no new FEA model is required for the stress analysis of VCSNS. The analyses were performed using representative pressurized water reactor (PWR) geometries, bounding transients, and typical material properties. The results of the stress analyses were used in a flaw tolerance evaluation. The applicability of the FEA analysis to VCSNS 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 VCSNS. 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 VCSNS, in Tables 3 and 4 for the PZR and Tables 5 and 6 for the SGs.
Table 3. PZR shell dimensions Shell Inside Shell/ Clad Shell Shell Diameter (ID)
Thickness Roft Rift (inches)
(inches)
EPRI Report 84 <1>
3.75 / 0.063 <1>
12.2 <1>
11.2 (Table 4-4 of [1-2])
VCSNS 84 3.75 / 0.19 12.2 11.2
<1> Westinghouse PZR dimensions associated with model for lower head.
Table 4. PZR nozzle dimensions Surge Nozzle ID Thickness Rift (inches)
(inches)
EPRI Report 12.44 (l) 3.27 (l) 1.9 (l)
(Table 4-5 of (1-2])
VCSNS 12.44 3.28 1.9 Serial No.: 22-115 Docket No.: 50-395 Page 7 of 17 Safety I Relief Nozzle ID Thickness Rift (inches)
(inches) 5.625 <2>
1.19<2>
2.363 <2>
5.89 2.57 1.15
<1> Westinghouse PZR nozzle dimensions associated with model for lower head.
<2> Combustion Engineering (CE) PZR nozzle dimensions associated with model for upper 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 VCSNS are performed at normal operating conditions. No hydrostatic testing of the VCSNS PZR or SGs has been performed since the plant began operation.
Table 5. SG vessel dimensions Primary Primary Secondary Secondary Lower Lower Primary Lower Lower Secondary Head Head Lower Shell Shell Lower ID Thickness Head ID Thickness Shell
/Clad Ri/t R1/t (inches)
(inches)
(inches)
(inches)
EPRIReport 155.87 6.94 / 0.27 11.2 162.45 3.65 22.3 (Table 4-2 of [1-3])
VCSNS 125.57 5.26 / 0.22 11.9 128.80 3.35 19.22 1 Section 5.1, page 7, fourth paragraph
Table 6. SG nozzle dimensions FW Nozzle ID (inches)
EPRI Report 16.5 (Figure 4-10 of [1-4])
VCSNS 16.5 FWNozzle Thickness (inches) 6 4.75 Serial No.: 22-115 Docket No.: 50-395 Page 8 of 17 FWNozzle Ri/t 1.38 1.74 As discussed in Sections 4.3.3 and 4.6 of Reference [1-4] and noted by the NRC in the SE2 [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 5 and 6 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 VCSNS are performed at normal operating conditions. No hydrostatic testing of the VCSNS PZR or SGs has been performed since the plant began operation.
In Reference [1-3], clad residual stress was not considered for the primary side welds. This was noted by the NRC in a RAI for Millstone Power Station, Unit 2 (MPS2). In response to the RAI [1-8], an evaluation was performed that showed 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 ISi, the NRC's safety goal of 1.ox10-5 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 [1-10] for Vogtle, the NRC performed an audit [1-22] of the PROMISE, Version 1.0 software. The PFM 2 Section 3.8.3.1, page 9, third paragraph 3 RAI Response No. 3c
Serial No.: 22-115 Docket No.: 50-395 Page 9 of 17 analysis in References [1-2] and [1-3] was performed using the PROMISE, Version 2.0 software, which has not been audited by the NRG. The only technical difference between Version 1.0 and Version 2.0 of the PROMISE software is that the 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 NRG staff found the use of the 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 VGSNS PZR is provided below.
For the VGSNS PZR, PSI examinations have been performed followed by ISi examinations over three (3) complete 10-year intervals. The inspection schedule scenario for these welds is PSI plus three (3) 10-year ISi examinations (PSI + 10
+ 20 + 30 Inspection Scenario). Most of the required fourth interval examinations have been completed at the time of this request. 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 DESG for VGSNS 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 NRG staff asked PSEG Nuclear, LLG., (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 NRG 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 NRG's safety goal of 1.0x1 o-6 failures per year. For the specific case of VGSNS, 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 NRG 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 4 Section 3.1, page 5, fourth paragraph 5 Section 9.1, page 15, last paragraph
Serial No.: 22-115 Docket No.: 50-395 Page 10 of 17 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 of 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 1.0x1 o-6 failures per year. A comparison of the PSI/ISi scenarios used in the sensitivity studies performed in References [1-3] and [1-4] to those for the VCSNS SGs is provided below.
In 1994, all three (3) VCSNS SGs were replaced. For the replaced SGs, PSI examinations have been performed in the first period of the second 10-year interval followed by ISi examinations for the second and third 10-year ISi intervals. The inspection scenario for these welds and components is PSI plus two (2) complete 10-year ISi examinations for the replaced SGs (PSI + 10 + 20 Inspection Scenario). It should be noted that most of the required fourth interval examinations have been completed at the time of this request. 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 DESC for VCSNS 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 ofVCSNS, the Section V procedures are used for the PZR and SG ultrasonic examinations. As stated in the NRC SEs for Salem6 [1-6] and Vogtle7 [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 6 Section 9.2, page 15 7 Section 3.8.8.2, page 21
Serial No.: 22-115 Docket No.: 50-395 Page 11 of 17 since the POD curve is not one of the parameters that significantly affects 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] provide verification of the above PFM results for VCSNS 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.
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 VCSNS (including examinations performed to-date, examination findings, examination coverage, and relief requests) is provided in.
As shown in the attachment, some of the welds/components have limited examination coverage, however all coverage is greater than 50%. This examination coverage was determined to be acceptable by the NRG in Section 10 of the SE [1-6] for Salem. 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 Serial No.: 22-115 Docket No.: 50-395 Page 12 of 17 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 1.0x 1 o-6 failures per reactor year is met with considerable margins. Plant-specific applicability of the technical basis to VCSNS is demonstrated in Attachments 2 and
- 3. The requested inspection interval provides an acceptable 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) SGs were replaced during the second ISi interval in 1994. 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 during the first three (3) inspection intervals for the subject PZR and SG welds and components at VCSNS, 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, DESC requests the NRC grant this proposed alternative for VCSNS in accordance with 10 CFR 50.55a(z)(1).
6.0 Duration of Proposed Alternative Serial No.: 22-115 Docket No.: 50-395 Page 13 of 17 Approval of the proposed alternative is requested by April 1, 2023, to support the VCSNS Spring 2023 refueling outage (RFO). Upon approval, the proposed alternative will cover the remainder of the current fourth 10-year ISi interval through the end of the fifth 10-year ISi interval ending on December 31, 2033.
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-AL T-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)
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)
Serial No.: 22-115 Docket No.: 50-395 Page 14 of 17 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 Energy 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 welds from 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. B3.90 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].
REFERENCES Serial No.: 22-115 Docket No.: 50-395 Page 15 of 17 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," 2007 Edition with 2008 Addenda.
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-She/1 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-She/1 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 (NRG) to Eric 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 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-AL T-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 NRG, "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.: 22-115 Docket No.: 50-395 Page 16 of 17 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 the 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, D.C., 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, Revision 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.: 22-115 Docket No.: 50-395 Page 17 of 17 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)
ATTACHMENT 2 Plant-Specific Applicability Serial No.: 22-115 Docket No.: 50-395 VIRGIL C. SUMMER NUCLEAR STATION UNIT 1 (VCSNS)
DOMINION ENERGY SOUTH CAROLINA, INC. (DESC)
Plant-Specific Applicability Serial No.: 22-115 Docket No.: 50-395 Page 1 of 15 Section 9 of References [2-1], [2-2], and [2-3] provide requirements that must be demonstrated in order to apply the representative stress and flaw tolerance analyses to a specific plant. Plant-specific evaluation of these requirements for VCSNS is provided in Tables A1 through A7.
Tables A 1 through A7 indicate that all plant-specific requirements are met for VCSNS.
Therefore, the results and conclusions of the Electric Power Research Institute (EPRI) reports are applicable to VCSNS. Figures A 1 and A2 show the layout of the pressurizer (PZR) and steam generators (SGs).
Table A1. Plant-specific applicability of References [2-1], [2-2], and [2-3] representative analyses to the VCSNS PZR and SG components PZR Shell-to-Head Welds (Circumferential and Longitudinal) and Nozzle-to-Shell Welds Category General Requirements ITEM Nos. B2.11, B2.12, AND B3.110 Requirement from Reference [2-1]
The plant-specific PZR general transients and cycles must be bounded by those shown in Table 5-6 for a 60- year operating life. It should be noted that the number of cycles were extrapolated to 80 years in the evaluations.
The materials of the PZR shell and nozzles must be low alloy ferritic steels which conform to the requirements of ASME Code,Section XI, Appendix G, Paragraph G-2110.
Applicability to VCSNS The number and type of the VCSNS general transients are compared to the transients listed in Table 5-6 of Reference [2-1]. As shown in Tables A2 and A3, the VCSNS transients are bounded by the transients listed in Table 5-6 of Reference [2-1].
The VCSNS PZR upper and lower heads and shells are fabricated of SA-533, grade A, class 2 material.
The nozzle forgings are fabricated of SA-508, class 2 material.
The materials for the PZR conform to the requirements of ASME Code,Section XI, Division 1, Appendix G, Paragraph G-2110.
Category Specific Requirements Requirement from Reference [2-1]
Serial No.: 22-115 Docket No.: 50-395 Page 2 of 15 Applicability to VCSNS The plant-specific PZR upper head The VCSNS PZR upper head and and bottom head weld configurations bottom head weld configurations must conform to those shown in conform to those shown in Figure 1-1 Figure 1-1 (Item No. B2.11 ), Figure 1-(Item No. B2.11 ), Figure 1-2 (Item No.
2 (Item No. B2.12) and Figures 1-4 B2.12) and Figures 1-4 and 1-5 (Item and 1-5 (Item No.
B3.110) of No. B3.110) of Reference [2-1].
Reference [2-1 ].
The plant-specific dimensions of the The comparison of the VCSNS PZR PZR upper head and nozzles, shell, dimensions with those in Table 9-1 of lower head, and the surge nozzle Reference [2-1] is provided in Table must be within the range of values A4. The comparison shows that the listed in Table 9-1 of Reference [2-1].
VCSNS configurations are within the The plant-specific lnsurge/Outsurge transient definitions (temperature difference between the PZR shell and the PZR surge nozzle fluid temperature and associated number of cycles) must be bounded by those shown in Table 5-10 for a
Westinghouse/CE plant, of Reference
[2-1 ].
range of values shown in Table 9-1 of Reference [2-1].
In Attachment 3 of this request, the VCSNS lnsurge/Outsurge transients are compared to the number and type of transients listed in Table 5-10 of Reference [2-1]. As can be seen from, Table A9, the VCSNS transients are bounded by those transients listed in Table 5-10 of Reference [2-1 ].
Serial No.: 22-115 Docket No.: 50-395 Page 3 of 15 SG Primary Side Tubesheet-To-Head Welds ITEM No. 82.40 Category Requirement from Reference [2-2]
Applicability to VCSNS The Loss of Power transient (involving VCSNS has not experienced a loss of unheated auxiliary feedwater (AFW) power transient resulting in unheated being introduced into a hot SG that AFW being introduced into a hot SG has been boiled dry following that has been boiled dry following blackout, resulting in thermal shock of blackout, resulting in thermal shock of portion of the vessel) is not portion of the vessel.
considered in this evaluation due to its rarity. If such a significant thermal event occurs at a plant, its impact on the K,c (material fracture toughness)
General value may require more frequent Requirements examinations and other plant actions outside the scope of this report's guidance.
The materials of the SG vessel heads and tubesheet must be low alloy ferritic steels which conform to the requirements of ASME Code,Section XI, Appendix G, Paragraph G-2110.
- heads, tubesheet, shell, and nozzles are fabricated of SA-508, class 3a material. This material conforms to the requirements of ASME Code,Section XI, Appendix G, Paragraph G-2110.
The weld configurations must conform The VCSNS weld configurations to those shown in Figures 1-1 and conform to Figure 1-1 and Figure 1-2 Specific Requirements Figure 1-2 of Reference [2-2]
of Reference [2-2].
The SG vessel dimensions must be within 10 percent of the upper and lower bounds of the values provided in the table in Section 9.4.3 of Reference [2-2].
The VCSNS SG vessel dimensions are as follows:
Diameter SG Lower Head 136.08 inches SG Upper Shell 176.26 inches These dimensions are within 10 percent of those specified in Table 9-2 in Section 9.4.3 of Reference [2-2].
Category Specific Requirements Requirement from Reference [2-2]
The component must experience transients and cycles bounded by those shown in Table 5-7 of Reference [2-2]
over a 60-year operating life.
Serial No.: 22-115 Docket No.: 50-395 Page 4 of 15 Applicability to VCSNS As shown in Table A5, the VCSNS transients and number of cycles projected to occur over a 60-year life are bound by those shown in Table 5-7 of Reference [2-2]. There is slight variation in some of the temperature and pressure parameters between VCSNS and Table 5-7 of Reference
[2-2]. Due to the conservative Heatup and Cooldown rates used in Reference [2-2], the slight variation in the Heatup and Cool down temperatures is not a concern. In
- addition, the VCSNS 60-year projected Heatup and Cooldown cycles is 114 which is significantly less than the 300 cycles evaluated in Reference [2-2].
Likewise, the evaluation performed in Reference [2-2] uses conservative rates of change in temperature and pressure for reactor trip transients, and the slight variation in the reactor trip parameters between VCSNS and Table 5-7 of Reference [2-2] is not a concern.
- Further, the VCSNS 60-year projected reactor trip cycles is 147 which is less than half of the 360 cycles evaluated in Reference [2-2].
VCSNS considers three (3) reactor trip transients including:
- Case A - No Cooldown;
- Case 8 - Cooldown, no SI; and
- Case C - Cooldown, with SI.
As seen in Table A5, Case C is the worse-case transient, and the VCSNS 60-year projected reactor trip for Case C cycles is only 5.
Category General Requirements Specific Requirements Serial No.: 22-115 Docket No.: 50-395 Page 5 of 15 SG Secondary Side Shell Welds ITEMS Nos. C1.10, C1.20 AND C1.30 Requirement from Reference [2-1]
The Loss of Power transient (involving unheated AFW being introduced into a hot SG that has been boiled dry following blackout, resulting in thermal 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 Kie (material fracture toughness) value may require more frequent examinations and other plant actions outside the scope of this report's guidance.
Applicability to VCSNS VCSNS has not experienced a loss of power transient resulting in unheated AFW being introduced into a hot SG that has been boiled dry following blackout, resulting in thermal shock of portion of the vessel.
The materials of the SG shell, The VCSNS SG vessel
- heads, feedwater (FW) nozzles, and main steam (MS) nozzles must be low alloy ferritic steels which conform to the requirements of ASME Code,Section XI, Appendix G, Paragraph G-2110.
tubesheet, shell, and nozzles are fabricated of SA-508, class 3a material. This material conforms to the requirements of ASME Code,Section XI, Appendix G, Paragraph G-2110.
The weld configurations must conform The VCSNS weld configurations to those shown in Figure 1-7 and conform to Figure 1-7 and Figure 1-8 Figure 1-8 of Reference [2-2].
of Reference [2-2].
The SG vessel dimensions must be within 10 percent of the upper and lower bounds of the values provided in the table in Section 9.4.4 of Reference [2-2].
The VCSNS SG vessel dimensions are as follows:
Diameter SG Lower Head 136.08 inches SG Upper Shell 176.26 inches These dimensions are within 10 percent of those specified in Table 9-3 in Section 9.4.4 of Reference [2-2].
Category Requirement from Reference [2-1]
The component must experience transients and cycles bounded by those shown in Table 5-9 of Reference [2-2]
over a 60-year operating life.
Serial No.: 22-115 Docket No.: 50-395 Page 6 of 15 Applicability to VCSNS As shown in Table A6, the VCSNS transients and number of cycles projected to occur over a 60-year life are bound by those shown in Table 5-9 of Reference [2-2]. There is slight variation in some of the temperature parameters between VCSNS and Table 5-9 of Reference [2-2]. Due to the conservative Heatup and Cooldown rates used in Reference [2-2], the slight variation in the Heatup and Cooldown temperatures is not a concern. In addition, the VCSNS 60-year projected Heatup and Cooldown cycles is 114 which is significantly less than the 300 cycles evaluated in Reference [2-2].
Likewise, the evaluation performed in Reference [2-2] uses conservative rates of change in temperature and pressure for reactor trip transients, and the slight variation in the reactor trip temperatures between VCSNS and Table 5-9 of Reference [2-2] is not a concern.
- Further, the VCSNS 60-year projected reactor trip cycles is 147 which is less than half of the 360 cycles evaluated in Reference [2-2]. VCSNS considers three (3) reactor trip transients including:
Case A - No Cooldown; Case B - Cooldown, no SI; and Case C-Cooldown, with SI.
The reactor trip temperature and pressure parameters in Table A6 are for Case C which is the worse-case transient, and the VCSNS 60-year projected reactor trip for Case C cycles is only 5.
Category Serial No.: 22-115 Docket No.: 50-395 Page 7 of 15 SG FW Nozzle-to-Shell Welds and Inside Radius Sections ITEMS Nos. C2.21 AND C2.22 Requirement from Reference [2-3]
Applicability to VCSNS The nozzle-to-shell weld shall be one The VCSNS FW nozzles are of the configurations shown in Figure representative of the configuration 1-1 or Figure 1-2 of Reference [2-3].
shown in Figure 1-2 of Reference [2-3].
The materials of the SG shell and FW The VCSNS FW nozzles and vessel nozzles must be low alloy ferritic General steels which conform to the shell are fabricated of SA-508, class 3a material. This material conforms to the requirements of ASME Code,Section XI, Appendix G, Paragraph G-2110.
Requirements requirements of ASME Code,Section XI, Appendix G, Paragraph G-2110.
SGFW Nozzle The SG must not experience more than the number of all transients shown in Table 5-5 of Reference [2-3]
over a 60-year operating life.
As shown in Table A?, the VCSNS SG is not projected to experience more than the number of the transients shown in Table 5-5 of Reference [2-3].
The piping attached to the FW nozzle The VCSNS FW piping lines are 16 must be 14-inch to 18-inch NPS.
inches.
The FW nozzle design must have an The VCSNS FW nozzle configuration integrally attached thermal sleeve.
has an integrally attached thermal sleeve.
AFW nozzles connected directly to the SG are not covered in this evaluation.
Serial No.: 22-115 Docket No.: 50-395 Page 8 of 15 Table A2. VCSNS PZR general transients applicable to this request VCSNS Transient Name Up to 2018 60-Year Projected Maximum Cycles (Controlling Limit)
Heatup@ <100°F/hr 68 114 200 Cooldown @ <100°F/hr 66 110 200 Reactor Trips <1>
88 147 400 50% Step Load Decrease w/
58 97 200 Steam Dump Loss of Load 17 29 80 Loss of Flow 2
4 80 in One RC Loop Only Loss of Offsite AC Power 7
12 40
<1> Reactor Trip Transients include Case A-No Cooldown; Case B - Cooldown with no SI; and Case C
- Cooldown with SI.
Table A3. Comparison of VCSNS PZR general transients to the transients evaluated in Reference [2-1]
Number of Cycles VCSNS Transient for 60 Years 60-Year Projections from Reference [2-1]
Heatup/Cooldown 300 114 Loss of Load <1>
360 289
<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.
Serial No.: 22-115 Docket No.: 50-395 Page 9 of 15 Table A4. Range of PZR geometric parameters for which the evaluation is applicable in comparison with VCSNS Geometric Parameter Westinghouse VCSNS Component Plant Dimensions (inches)
(inches)
(inches)
PZR Shell Inside diameter Must be between 84 80 and 88 NPS of piping or component Must be between Surge Nozzle (e.g., reducer) 12 and 18 14 attached to nozzle safe-end NPS of piping or component Must be between Safety/Relief Nozzle (e.g., reducer) 4 and 8 6
attached to nozzle safe-end NPS of piping or component Must be between Spray Nozzle (e.g., reducer) 4 and 6 4
attached to nozzle safe-end
Serial No.: 22-115 Docket No.: 50-395 Page 10 of 15 Table AS. VCSNS 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 Thot Min Thot Max Tcold Min Tcold Max Min Press Press 60-Year Transient Cycles (OF)
(OF)
(OF)
(OF)
(PS/G)
(PSIG)
Reference [2-2]
545 70 545 70 2235 0
300 Heatup/Cooldown VCSNS 557 70 557 70 2235 0
200 Reference [2-2]
610 550 550 545 2300 2300 5000 Plant Loading / Unloading VCSNS Not typical operation, not counted <1l Reactor Trip Reference [2-2]
615 530 565 530 2435 1700 360 Reactor Trip <2l - Case A VCSNS 621.9 563.9 552.6 557.6 2235 1960 230 Reactor Trip <2l - Case B VCSNS 621.9 531.9 552.6 517.6 2235 1600 160 Reactor Trip <2l - Case C VCSNS 621.9 457.9 552.6 452.6 2235 1485 10
<1l Load following operation is not typical.
<2l Reactor Trip Transients include Case A-No Cooldown; Case B-Cooldown with no SI; and Case C - Cooldown, with SI. Transient is assumed to begin at 100% full power.
Serial No.: 22-115 Docket No.: 50-395 Page 11 of 15 Table A6. VCSNS data for thermal transients for stress analysis of the PWR SG secondary-side vessel welds (Comparison to Table 5-9 of Reference [2-2])
Max Tss Min Tss Max Press Min Press Transient
(°F)
(OF)
(PSIG)
(PS/G)
Reference [2-2]
545 70 1000 0
Heatup/Cooldown VCSNS 557 70 951 0
Reference [2-2]
545 540 1000 1000 Plant Loading/Unloading VCSNS Not typical operation, not counted (1>
Reactor Trip Reference [2-2]
555 530 1130 Reactor Trip (2>
VCSNS 550.4 460.4 951 (3)
(1l Load following operation not typical.
<2) Reactor Trip Transient Case C - Cooldown, with SI, is considered since it is the worse-case transient.
<3) SG secondary-side pressure is not an analyzed design transient parameter for Reactor Trip Transients.
1000 951 (3)
GO-Year Cycles 300 200 5000 360 400
Serial No.: 22-115 Docket No.: 50-395 Page 12 of 15 Table A7. VCSNS data for thermal transients applicable to pressurized water reactor SG FW (Comparison to Table 5-5 of Reference [2-3])
Transient Cycles from Table 5-5 VCSNS VCSNS of Reference [2-3]
60-year Projected Cycles 60-year Allowable Cycles Heatup/Cooldown 300 58 200 Plant Loading 5000 Not typical operation (1>
Plant Unloading Loss of Load 360 NIA (2>
N/A (2>
Loss of Power 60 N/A (2>
N/A (2>
(1) Load following operation not typical.
(2) Loss of Load and Loss of Power Transients are not evaluated. Upon Loss of Load and Loss of Power Transients, flow to the FW nozzle is at nominal temperature before 60 seconds. After 60 seconds, flow is to the emergency FW nozzle. Therefore, VCSNS will not experience Loss of Load or Loss of Power Transients on the FW nozzle.
ti i l q
1-21008-11 1-2100A-5 1-2100A-1 Serial No.: 22-115 Docket No.: 50-395 Page 13 of 15 1-21008-9 1-21008-12 1-21008-13 Figure A1. VCSNS Pressurizer Layout
t--~~.;............---- ----I I
I I I
I +
1 I
I 1
j I
-~----4!!!!'='~~-
2-1100-20 2-1100-23 2-1100-18 2-1100-17 2-1100-15 1-3100-14 Serial No.: 22-115 Docket No.: 50-395 Page 14 of 15 Figure A2. VCSNS Steam Generator Layout
REFERENCES Serial No.: 22-115 Docket No.: 50-395 Page 15 of 15 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.
ATTACHMENT 3 Comparison of lnsurge/Outsurge Transients Serial No.: 22-115 Docket No.: 50-395 VIRGIL C. SUMMER NUCLEAR STATION UNIT 1 (VCSNS)
DOMINION ENERGY SOUTH CAROLINA, INC. (DESC)
Serial No.: 22-115 Docket No.: 50-395 Page 1 of 2 Comparison of lnsurge/Outsurge Transients VCSNS lnsurge/Outsurge (1/0) transients are provided in Table AS. The temperature differences (/iTs) identified in Table AS are combined conservatively by summing all the events into the 320°F /iT bin. It should be noted that the transients in Table AS reflect 35 years of operation. For comparison with Table 5-10 of Reference [3-1], they are extrapolated to 60 years by multiplying by 1. 7. With this conservative treatment of the 1/0 transients, the comparison of VCSNS 1/0 transients to the requirements in Reference [3-1] is shown in Table A9. The results of Table A9 indicate that the VCSNS 1/0 transients are bounded by those in Reference [3-1].
Table AB. 35-year 1/0 transients for VCSNS Number Transient Name <1>
35-Year Cycles 1
HU340 0
2 HU330 0
3 HU320 0
4 HU310 0
5 HU300 2
6 HU280 7
7 HU260 37 8
HU240 8
9 CD340 1
10 CD330 0
11 CD320 0
12 CD310 1
13 CD300 3
14 CD280 13 15 CD260 32 16 CD240 3
<1> The Transient Name is XX.nnn, where XX = HU for lnsurge/Outsurge transients that occur during Heatup events, or CD for 1/0 transients that occur during Cooldown events, and nnn = the temperature difference, LH, between the PZR fluid temperature and the fluid temperature in the surge nozzle.
Serial No.: 22-115 Docket No.: 50-395 Page 2 of 2 Table A9. Comparison of VCSNS 1/0 transient temperature differences and numbers of cycles with the 1/0 transient date from Reference [3-1]
60-Year Number of Cycles VCSNS AT Cycles Projected to From Reference [3-1]
60 Years of Operation (OF) (1) 330 600 0
320 3,000 182 (2)(3) 103 1,500 0
<1> b. T is the temperature difference between the PZR fluid temperature and the fluid temperature in the surge nozzle.
<2> The number of cycles is conseNatively equal to the sum of all events in Table A8 analyzed for 35 years (107 cycles), increased by a factor of 1. 7 to reflect 60 years of operation.
<3> Transient CD340 was projected under the 320°F b. T bin. A 340°F b. T is an anomaly and is not expected to occur in the future.
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.
ATTACHMENT 4 Inspection History Serial No.: 22-115 Docket No.: 50-395 VIRGIL C. SUMMER NUCLEAR STATION UNIT 1 (VCSNS)
DOMINION ENERGY SOUTH CAROLINA, INC. (DESC)
Inspection History Serial No.: 22-115 Docket No.: 50-395 Page 1 of 6 Table A 10 provides the inspection history for the VCSNS pressurizer (PZR) and steam generator (SG) welds and components for the pre-service inspection (PSI) and the first, second, third, and fourth 10-year inservice inspection (ISi) intervals. Several of the older examination reports that were used to populate Table A 10 did not record the examination coverage. The examination coverages documented in Table A 10 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 A 10. PZR and SG inspection history Pressurizer Inspection History Item No.
Component ID Examination Interval l Period Examination Date Results 1-2100A-1 1982 PSI Acceptable 1-2100A-1 4 - 1990 11 I P2 Acceptable 1-2100A-1 9 -1994 12 I P1 Acceptable 1-2100A-1 4 - 2011 13 I P3 Acceptable 82.11 1-2100A-4 1982 PSI Acceptable 1-2100A-4 10 - 1991 11 I P3 Acceptable 1-2100A-4 10 - 1997 12 I P1 Acceptable 1-2100A-4 5 - 2005 13 I P1 Acceptable 1-2100A-4 4 - 2014 14 I P1 Acceptable 1-2100A-5 1982 PSI Acceptable 1-2100A-5 4 - 1990 11 I P2 Acceptable 1-2100A-5 5 - 2002 12 I P3 Acceptable 1-2100A-5 4 - 2011 13 I P3 Acceptable 82.12 1-2100A-7 1982 PSI Acceptable 1-2100A-7 10 - 1991 11 I P3 Acceptable 1-2100A-7 10 - 1997 12 I P1 Acceptable 1-2100A-7 5 - 2005 13 I P1 Acceptable 1-2100A-7 4 - 2014 14 I P1 Acceptable Examination Coverage 100%
100%
98.3%
98.3%
100%
87%
98%
98%
97%
100%
100%
100%
100%
100%
100%
100%
100%
100%
Serial No.: 22-115 Docket No.: 50-395 Page 2 of 6 Relief Request NIA NIA NIA NIA NIA NIA NIA NIA NIA NIA NIA NIA NIA NIA NIA NIA NIA NIA
Pressurizer Inspection History Item No.
Component ID Examination Interval / Period Examination Date Results 1-2100A-8 1982 PSI Acceptable 1-2100A-8 10 -1985 11 I P1 Acceptable 1-2100A-8 10 -1997 12 / P1 Acceptable 1-2100A-8 10 -2012 13 I P3 Acceptable 1-2100B-9 1982 PSI Acceptable 1-2100B-9 10 -1985 11 I P1 Acceptable 1-2100B-9 4 -1996 12 / P1 Acceptable 1-2100B-9 5 -2005 13 I P1 Acceptable 1-2100B-9 4 - 2014 14 I P1 Acceptable 1-2100B-10 1982 PSI Acceptable 1-2100B-10 10-1991 11 I P3 Acceptable B3.110 1-2100B-10 4 -1996 12 / P1 Acceptable 1-2100B-10 5 - 2005 13 / P1 Acceptable 1-2100B-10 4 - 2014 14 I P1 Acceptable 1-2100B-11 1982 PSI Acceptable 1-2100B-11 10-1991 111 P3 Acceptable 1-2100B-11 4 - 1996 12 I P1 Acceptable 1-2100B-11 5 - 2005 13 I P1 Acceptable 1-2100B-11 4 - 2014 14 I P1 Acceptable 1-2100B-12 1982 PSI Acceptable 1-2100B-12 10-1991 11 I P3 Acceptable 1-2100B-12 4 -1996 12 I P1 Acceptable 1-2100B-12 5 - 2005 13 I P1 Acceptable Examination Coverage Note 1 Note 1 Note 1 75.5%
Note 1 Note 1 Note 1 92%
70.2%
Note 1 Note 1 Note 1 92%
72.3%
Note 1 Note 1 Note 1 92%
72.3%
Note 1 Note 1 Note 1 92%
Serial No.: 22-115 Docket No.: 50-395 Page 3 of 6 Relief Request N/A N/A N/A RR-11I-10 NIA N/A N/A N/A TBD NIA NIA NIA NIA TBD N/A N/A NIA NIA TBD NIA NIA N/A NIA
Pressurizer Inspection History Item No.
Component ID Examination Interval / Period Examination Date Results 1-21008-12 4 - 2014 14 / P1 Acceptable 1-21008-13 1982 PSI Acceptable 1-21008-13 10 -1985 11 / P1 Acceptable 1-21008-13 4 - 1996 12 / P1 Acceptable 1-21008-13 5 - 2005 13 / P1 Acceptable 1-21008-13 4 - 2014 14 / P1 Acceptable Note 1: Coverage was not documented.
Examination Coverage 72.3%
Note 1 Note 1 Note 1 92%
72.3%
Serial No.: 22-115 Docket No.: 50-395 Page 4 of 6 Relief Request T8D N/A NIA NIA N/A T8D
Steam Generator Inspection History Item No.
Component ID Examination Interval l Period Examination Date Results SG Primary Side Shell Welds 1-3100-14A 7 - 1994 121 P1 PSI Acceptable 1-3100-148 7 - 1994 121 P1 PSI Acceptable 1-3100-14C 8 - 1994 12 / P1 PSI Acceptable 82.40 1-3100-14C 11 - 2003 12 / P3 Acceptable 1-3100-14C 11 - 2009 131 P2 Acceptable 1-3100-14C 4 - 2020 141 P2 Acceptable SG Secondary Side Shell Welds 2-1100-17A 7 - 1994 12 / P1 PSI Acceptable 2-1100-178 7 - 1994 121 P1 PSI Acceptable 2-1100-17C 7 - 1994 12 / P1 PSI Acceptable 2-1100-17C 4 - 1999 12/ P2 Acceptable 2-1100-17C 11 - 2009 131 P2 Acceptable C1.10 2-1100-18A 7 - 1994 121 P1 PSI Acceptable 2-1100-188 7 - 1994 12 / P1 PSI Acceptable 2-1100-18C 7 - 1994 121 P1 PSI Acceptable 2-1100-18C 4 - 1999 12/ P2 Acceptable 2-1100-18C 11 - 2009 13 / P2 Acceptable 2-1100-18C 4 - 2020 14/ P3 Acceptable Examination Coverage 100%
100%
100%
100%
100%
100%
Note 1 Note 1 Note 1 Note 1 98%
100%
100%
100%
100%
100%
100%
Serial No.: 22-115 Docket No.: 50-395 Page 5 of 6 Relief Request NIA NIA NIA N/A NIA N/A NIA NIA N/A N/A N/A NIA NIA NIA N/A NIA NIA
Item No.
Component ID 2-1100-20A 2-1100-208 2-1100-20C C1.20 2-1100-20C 2-1100-20C 2-1100-20C 2-1100-15A 2-1100-158 C1.30 2-1100-15C 2-1100-15C 2-1100-15C SG Secondary Side Nozzles 2-1100-23A 2-1100-238 C2.21 2-1100-23C 2-1100-23C 2-1100-23C 2-1100-231R-A 2-1100-231R-8 C2.22 2-1100-231R-C 2-1100-231R-C 2-1100-231R-C Note 1: Coverage was not documented Steam Generator Inspection History Examination Interval/ Period Examination Date Results 7 - 1994 12 I P1 PSI Acceptable 7 - 1994 12 I P1 PSI Acceptable 7 - 1994 12 I P1 PSI Acceptable 5 - 2002 12 I P3 Acceptable 10 - 2012 131 P3 Acceptable 4 - 2020 141 P3 Acceptable 7 - 1994 12 I P1 PSI Acceptable 7 - 1994 12 / P1 PSI Acceptable 7 - 1994 12 I P1 PSI Acceptable 11 - 2003 12 I P3 Acceptable 10 - 2012 13 I P3 Acceptable 7 - 1994 12 I P1 PSI Acceptable 7 - 1994 12 I P1 PSI Acceptable 7 - 1994 12 I P1 PSI Acceptable 10 - 2003 12 I P3 Acceptable 10-2012 13 I P3 Acceptable 7 - 1994 12 I P1 PSI Acceptable 7 - 1994 12 / P1 PSI Acceptable 7 - 1994 12 I P1 PSI Acceptable 10 - 2003 12 I P3 Acceptable 10 - 2012 13 I P3 Acceptable Examination Coverage 100%
100%
100%
100%
100%
99.5%
Note 1 Note 1 Note 1 94.4%
94.4%
Note 1 Note 1 Note 1 100%
100%
100%
100%
100%
100%
100%
Serial No.: 22-115 Docket No.: 50-395 Page 6 of 6 Relief Request NIA NIA NIA NIA NIA NIA NIA NIA NIA NIA NIA NIA NIA N/A N/A NIA NIA NIA N/A NIA NIA
ATTACHMENT 5 Results of Industry Survey Serial No.: 22-115 Docket No.: 50-395 VIRGIL C. SUMMER NUCLEAR STATION UNIT 1 (VCSNS)
DOMINION ENERGY SOUTH CAROLINA, INC. (DESC)
Overall Industry Inspection Summary for Serial No.: 22-115 Docket No.: 50-395 Page 1 of 5 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 two (2)-loop, three (3)-loop, and four (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. B2.11, B2.12, B2.21, B2.22, and B3.110 are summarized in Table A 11 below. A total of 1,162 examinations of PWR PZR components were reported by the survey for the affected Item Nos. Of the 1,169 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. B2.11 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 A 11 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.
TableA11. Summaryofsurveyresultsforltem Nos. B2.11, B2.12, B2.21, B2.22and 83.110 Item No.
Number of Examinations Number of 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.
Overall Industry Inspection Summary for Serial No.: 22-115 Docket No.: 50-395 Page 2 of 5 Steam Generator Code Items 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 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. B2.31, B2.32 (see Table Note 3), B2.40, B3.130, C1.10, C1.20, and C1.30 are summarized in Table A 12 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. B2.40, 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 [5-2] 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 12 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 exposure, risk to personnel safety, and production of radwaste resulting from these
Serial No.: 22-115 Docket No.: 50-395 Page 3 of 5 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. 82.31, 82.32, 82.40, 83.130, C1.10, C1.20, and C1.30 Number of Number of Examinations Reportable Indications Item No.
BWR PWR Total BWR PWR Total B2.31 0
30 30 0
0 0
B2.32 C3l 0
13 13 0
0 0
B2.40 0
183 183 0
(1)
(1)
B3.130 0
135 135 0
0 0
C1.10 140 305 445 0
0 0
C1.20 54 319 373 0
(2)
(2)
C1.30 32 113 145 0
0 0
Totals 226 1098 1324 0
(1) (2)
(1) (2)
<1> Two (2) PWR W-2 Loop units at a single plant reported multiple subsurface embedded fabrication flaws.
<2> A single PWR W-2 Loop unit reported multiple flaws [5-4, 5-5].
<3> Item No. 82.32 was evaluated in the Reference [5-1] technical basis and included in the industry survey but is not contained in the scope of this alternative request.
Overall Industry Inspection Summary for Serial No.: 22-115 Docket No.: 50-395 Page 4 of 5 Steam Generator Code Items 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 7 4 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 A13 below (see Table Note 1). A total of 727 examinations for Item Nos. C2.21, C2.22, and C2.32 (see Table Note 1) components were conducted, with 563 of these specifically for PWR components. The majority of the PWR examinations were performed on SG FW and MS nozzles. Only one (1) PWR examination identified two (2) flaws that exceeded the 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 13 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 A13. Summary of survey results for Item Nos. C2.21, C2.22, and C2.32 Plant Type Number of Number of Number of Units Examinations Reportable Indications BWR 27 164 0
PWR 47 563 2
Totals 74 727 <1>
2
<1> Item No. C2.32 was evaluated in the Reference [5-2] technical basis and included in the industry survey but is not contained in the scope of this alternative request.
REFERENCES Serial No.: 22-115 Docket No.: 50-395 Page 5 of 5 5-1 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.
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-She/1 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)