ML19347B105
| ML19347B105 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 12/11/2019 |
| From: | Gayheart C Southern Nuclear Operating Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NL-19-0832, VEGP-ISI-ALT-04-04 | |
| Download: ML19347B105 (22) | |
Text
A Southern Nuclear DEC 11 2019 Docket Nos.: 50-424 50-425 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Cheryl A. Gayheart Reguletory AffBJrs Director 3535 Colonnade Parkway B1nn1ngham, AL 35243 205 992 5316 tel 205 992 7795 fax cagayben@southernco com NL-19-0832 Vogtle Electric Generating Plant, Units 1 & 2 Proposed lnservice Inspection Alternative VEGP-ISI-AL T-04-04 Ladies and Gentlemen:
In accordance with 1 O CFR 50.55a(z)(1 ), Southern Nuclear Operating Company (SNC) hereby requests Nuclear Regulatory Commission (NRG) approval of proposed inservice inspection (ISi) alternative VEGP-ISI-ALT-04-04, Version 1.0. This proposed alternative, shown in Enclosure 1, requests to increase the inspection interval for ASME Section XI, Table IWC-2500-1, exam Category C-8, item number C2.21 and C2.22, exams from 1 O years to 30 years through for the remainder of the 5tti ISi Interval. Enclosure 2 provides a copy of EPRI technical report 3002014590, which contains supporting infonnation to the proposed alternative and is publicly available at epri.com.
NRG review and approval of the proposed alternative is respectfully requested by December 31, 2020.
This letter contains no NRG commitments. If you have any questions, please contact Jamie Coleman at 205.992.6611.
Respectfully submitted, Cheryl ft/.
heart Regulatory Affairs Director CAG/DSP/sm
/
U.S. Nuclear Regulatory Commission NL-19-0832 Page2 :
Proposed Alternative VEGP-ISI-ALT-04-04, Version 1.0, in Accordance with 1 O CFR 50.55a(z)(1) :
EPRI Technical Report 3002014590 (Non-Proprietary) cc:
Regional Administrator NRR Project Manager - Vogtle 1 & 2 Senior Resident Inspector-Vogtle 1 & 2 RType: CVC7000
Vogtle Electric Generating Plant, Units 1 & 2 Proposed lnservice Inspection Alternative VEGP-ISI-AL T-04-04 Proposed Alternative VEGP-ISI-AL T-04-04, Version 1.0, in Accordance with 1 O CFR 50.55a(z)(1) to NL-19-0832 Proposed Alternative VEGP-ISI-ALT-04-04, Version 1.0, in Accordance with 10 CFR 50.55a(z)(1) 1.0 ASME CODE COMPONENTS AFFECTED:
Code Class:
==
Description:==
Examination Category:
Item Numbers:
C ID omponent s:
11201-B6-001-W18 11201-B6-001-W19 11201-B6-002-W18 11201-B6-002-W19 11201-B6-003-W18 11201-B6-003-W19 11201-B6-004-W18 11201-B6-004-W19 11201-86-001-IR04 11201-B6-002-IR04 11201-B6-003-IR04 11201-B6-004-IR04 21201-B6-001-W18 21201-B6-001-W19 21201-B6-002-W18 21201-B6-002-W19 21201-B6-003-W18 21201-B6-003-W19 21201-B6-004-W18 21201-B6-004-W19 21201-B6-001-IR04 21201-B6-002-IR04 21201-B6-003-IR04 21201-B6-004-IR04 Class 2 Nozzle-to-shell welds and inside radius sections C-B (Pressure Retaining Nozzle Welds in Pressure Vessels,Section XI,' Division 1)
C2.21 - Nozzle-to-shell (nozzle-to-head or nozzle-to-nozzle) welds C2.22 - Nozzle inside radius sections 32" STEAM OUTLET NOZZLE TO UPPER HEAD WELD 16" MAIN FEEDWATER NOZZLE TO SHELL WELD 32" STEAM OUTLET NOZZLE TO UPPER HEAD WELD 16" MAIN FEEDWATER NOZZLE TO SHELL WELD 32" STEAM OUTLET NOZZLE TO UPPER HEAD WELD 16" MAIN FEEDWATER NOZZLE TO SHELL WELD 32" STEAM OUTLET NOZZLE TO UPPER HEAD WELD 16" MAIN FEEDWATER NOZZLE TO SHELL WELD MAIN FEEDWATER NOZZLE INNER RADIUS MAIN FEEDWATER NOZZLE INNER RADIUS MAIN FEEDWATER NOZZLE INNER RADIUS MAIN FEEDWATER NOZZLE INNER RADIUS 32" STEAM OUTLET NOZZLE TO UPPER HEAD WELD 1"6" MAIN FEEDWATER NOZZLE TO SHELL WELD 32" STEAM OUTLET NOZZLE TO UPPER HEAD WELD 16" MAIN FEEDWATER NOZZLE TO SHELL WELD 32" STEAM OUTLET NOZZLE TO UPPER HEAD WELD 16" MAIN FEEDWATER NOZZLE TO SHELL WELD 32" STEAM OUTLET NOZZLE TO UPPER HEAD WELD 16" MAIN FEEDWATER NOZZLE TO SHELL WELD MAIN FEEDWATER NOZZLE INNER RADIUS MAIN FEEDWATER NOZZLE INNER RADIUS MAIN FEEDWATER NOZZLE INNER RADIUS MAIN FEEDWATER NOZZLE INNER RADIUS 2.0 REQUESTED APPROVAL DATE:
Approval is requested by December 31, 2020.
3.0 APPLICABLE CODE EDITION AND ADDENDA:
The Fourth lnservice Inspection (ISi) Interval Code of record for Vogtle Units 1 & 2 is the 2007 Edition with 2008 Addenda of ASME Boiler and Pressure Vessel Code,Section XI, "Rules for lnservice Inspection of Nuclear Power Plant Components."
El-1 to NL-19-0832 Proposed Alternative VEGP-ISI-ALT-04-04, Version 1.0, in Accordance with 1 O CFR 50.55a(z)(1) 4.0 APPLICABLE CODE REQUIREMENT:
ASME Section XI IWC-2500(a), Table IWC-2500-1, Examination Category C-8, Item No.
C2.21 requires surface and volumetric examination of all representative steam generator nozzles at terminal ends of piping runs once during each Section XI inspection interval.
ASME Section XI IWC-2500(a), Table IWC-2500-1, Examination Category C-8, Item No.
C2.22 requires volumetric examination of representative steam generator all nozzle at terminal ends of piping runs once during each Section XI inspection interval. The examination areas for Item Nos. C2.21 and C2.22 are shown in Figures IWC-2500-4(a),
(b), and (d).
5.0 REASON FOR REQUEST:
The Electric Power Research Institute (EPRI) performed.an assessment [1] of the basis for the ASME Section XI examination requirements specified for Examination Category C-8 of ASME Section XI, Division 1 for Steam Generator (SG) Main Steam (MS) and Feedwater (FW) Nozzle-to-Shell-Welds and Nozzle Inside Radius Sections. The assessment includes a survey of inspection results from 74 units as well as flaw tolerance evaluations using probabilistic fracture mechanics (PFM) and deter~inistic fracture mechanics (DFM). The Reference [1] report concluded that the current ASME Code Section XI inspection interval of ten years can be increased significantly with no impact to plant safety. It is upon the basis of this conclusion that an alternate inspection interval is being requested. The Reference [1] report was developed consistent with the recommendations provided in EPRl's White Paper on PFM [14].
6.0 PROPOSED ALTERNATIVE AND BASIS FOR USE:
Southern Nuclear Company (SNC) is requesting an inspection alternative.to the examination requirements of ASME Section XI, Table IWC-2500-1, Examination Category C-8, Item Nos. C2.21 and C2.22. The proposed alternative is to increase the inspection interval for these examination items to 30 years (from the current ASME Code Section XI 10-year requirement) for the remainder of the 6th lnservice Inspection (ISi)
Interval. Although the EPRI report [1] supports a longer inspection period, 30 years was selected as a prudent alternative to ensure that one more examination was conducted prior to the end of the current license period for Vogtle Units 1 & 2. A summary of the key aspects of the technical basis for this request are summarized below. The applicability of the technical basis to Vogtle Units 1 & 2 is shown in Appendix A.
Degradation Mechanism Evaluation An evaluation of degradation mechanisms that could potentially impact the reliability of the SG MS and FW Nozzle-to-Shell Welds and Nozzle Inside Radius Sections was performed in Reference [1 ]. Evaluated mechanisms 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 SG MS and FW nozzles.
El-2 to NL-19-0832 Proposed Alternative VEGP-ISI-AL T-04-04, Version 1.0, in Accordance with 1 O CFR 50.55a(z)(1)
Stress Analysis Finite element analysis (FEA) was performed in Reference [1] to detennine the strdsses in the SG MS and FW Nozzle-to-Shell Welds and Nozzle Inside Radius Sections. The analysis was 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 Vogtle Units 1 & 2 is shown in Appendix A and confinns that all plant-specific requirements are met. Therefore, the evaluation results and conclusions of Reference
[1] are applicable to Vogtle Units 1 & 2.
Flaw Tolerance Evaluation Flaw tolerance evaluations were performed in Reference [1] consisting of PFM evaluations and confirmatory DFM evaluations. The results of the PFM analyses indicate that, after a preservice inspection (PSI), no other inspections are required for up to 60 years of plant operation to meet the U.S. Nuclear Regulatory Commission's (NRC's) safety goal of 1 o-a failures per year. For the specific case of Vogtle Units 1 and 2 where PSI followed by three 10-year interval inspections have been perfonned, Table 8-1 O of Reference [1] 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.
Inspection History Plant Vogtle Unit 1 and 2 operating experience (including examinations perfonned to date, examination findings, inspection coverage, and Relief Requests) is presented in Appendix 8. As shown in this Appendix, both the Item No. C2.21 (FW nozzle and MS nozzle) and Item No. C2.22 (MS nozzle) examinations have had limited coverage. Also, as shown in Appendix 8, no flaws that exceeded the ASME Code,Section XI acceptance standards were identified during any examinations.
Industry inspection history for these components (as obtained from an industry survey
[1]) is presented in Appendix C. The results of the survey [1] indicate that these '
components are very flaw tolerant.
Conclusion It is concluded that the SG MS and FW Nozzle-to-Shell Welds and Nozzle Inside Radius Sections are very flaw tolerant. PFM and DFM evaluations perfonned as part of the technical basis [1] demonstrate that, after PSI, no other inspection is required until 60 years to meet the NRC safety goal of 1 o-a failures per reactor year. Plant-specific applicability of the technical basis to Vogtle Units 1 & 2 is demonstrated in Appendix A.
An inspection interval of 30 years provides an acceptable level of quality and safety in El-3
(
to NL-19-0832 Proposed Alternative VEGP-ISI-AL T-04-04, Version 1.0, in Accordance with 1 O CFR 50.55a(z)(1) 7.0 lieu of the ASME Examination Category C-B, Item Nos. C2.21 and C2.22 surface and volumetric examination 10-year inspection frequency.
Operating and examination experience demonstrates that these components have performed with very high reliability, mainly due to their robust design. As shown in Appendix B, to date, SNC has performed 20 inspections of SG MS and FW Nozzle-to-Shell Welds and Nozzle Inside Radius Sections at Vogtle Units 1 & 2. No flaws that exceeded the ASME Code,Section XI acceptance standards were identified during any examinations, as shown in Appendix B. Some of the inspections listed in Appendix B involved limited coverage ranging from 50% to 80%. Section 8.2.5 of Reference [1]
discusses limited coverage and determines that the conclusions of the report are applicable to components with limited coverage. In addition, it is important to note all other inspection actMties, including the system leakage test (Examination Category C-H) conducted each refueling outage, will continue to be performed, providing further assurance of safety.
Finally, as discussed in Reference [2], for situations where no active degradation mechanism is present, it was concluded that subsequent inservice inspections do not provide additional value after PSI has been performed and the inspection volumes have been confirmed to have no flaws that exceeded the ASME Code,Section XI acceptance standards. The Vogtle Units 1 & 2 SG MS and FW nozzles have received the required PSI examinations and 20 follow-on inservice inspections with no flaws that exceeded the ASME Code,Section XI acceptance standards.
Therefore, SNC requests that the NRC authorize this proposed alternative in accordance with 10 CFR 50.55a(z)(1).
DURATION OF PROPOSED ALTERNATIVE:
The proposed Alternative is requested for the remainder of the 4th lnservice Inspection through 6th Inspection (ISi) Interval for Vogtle Units 1 &. 2; currently scheduled to end on 5/30/47.
8.0 PRECEDENT
No previous submittals have been made requesting relief from the ASME Examination Category C-8, Item Nos. C2.21 and C2.22 surface and volumetric examinations on the basis of the Reference [1] technical basis. However, the following is a list of approved Relief Requests related to inspections of SG MS and FW nozzles:
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. (SNOC), "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.
J El-4 to NL-19-0832 Proposed Alternative VEGP-ISI-ALT-04-04, Version 1.0, in Accordance with 1 O CFR 50.55a(z)(1)
Letter from T. H. Boyce (NRC) to C. L. Burton (CP&L), "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 (PG&E), "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 prec?dents related to similar requests for relief for Class 1 nozzles:
Based on studies presented in Reference [3], the NRC approved extending PWR reactor vessel nozzle-to-shell welds from 1 Oto 20 years in Reference [4].
Based on work performed in BWRVIP-108 [5] and BWRVIP-241 [7], the NRC approved the reduction of 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 1 O years) in References [6] and [8]. The work performed in BWRVIP-108 and BWRVIP-241 provided the technical basis for ASME Code Case N-702 [9], which has been conditionally approved by the NRC in Revision 18 of Regulatory Guide 1.147 [1 O].
Finally, there are precedents that used gen~ric industry guidance in a similar approach to the approach requested in this submittal:
Based on EPRI generic analysis, the Vogtle and Farley plants requested an alternative to the Reactor Pressure Vessel Threads in Flange examination
_ requirements of ASME Section XI in References [11] and [12].
NRC relief was granted for the Vogtle and Farley requests for alternatives to the Reactor Pressure Vessel Threads in Flange examination requirements in the reference [13] Safety Evaluation.
El-5 to NL-19-0832 Proposed Alternative VEGP-ISI-AL T-04-04, Version 1.0, in Accordance with 1 O CFR 50.55a(z)(1)
9.0 ACRONYMS
ASME B&W BWR BWRVIP CE CFR DFM EAF EPRI FAC FEA FW ISi MIC MS NPS NRC NSSS PFM PWR sec SG SNC American Society of Mechanical Engineers Babcock and Wilcox Boiling Water Reactor Boiling Water Reactor Vessel and Internals Program Combustion Engineering Code of Federal Regulations Deterministic fracture mechanics Environmentally assisted fatigue Electric Power Research Institute Flow accelerated corrosion Finite element analysis Feedwater lnservice Inspection Microbiologically influenced corrosion Main Steam Nominal pipe size
,Nuclear Regulatory Commission Nuclear steam supply system Probabilistic fracture mechanics Pressurized Water Reactor Stress corrosion cracking Steam Generator Southern Nuclear Company El-6 to NL-19-0832 Proposed Alternative VEGP-ISI-ALT-04-04, Version 1.0, in Accordance with 1 O CFR 50.55a(z)(1)
10.0 REFERENCES
- 1.
Technical Bases for Inspection Requirements for PWR Steam Generator Feedwater and Main Steam Nozzle-to-Shell Welds and Inside Radius Sections. EPRI, Palo Alto, CA: 2019. 3002014590.
- 2.
American Society of Mechanical Engineers, 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.
- 3.
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.
- 4.
US 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.
- 5.
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, EPRI, Palo Alto, CA 2002. 1003557.
- 6.
US 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.
- 7.
BWRVIP-241: BWR Vessels and Internals Project, Probabilistic Fracture Mechanics EvaltJation for the Boiling Water Reactor Nozzle-to-Shell Welds and Nozzle Blend Radii, EPRI, Palo Alto, CA 2010. 1021005.
- 8.
US 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.
- 9.
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.
- 10. U. S. NRC Regulatory Guide 1.147, Revision 18, "lnservice Inspection Code Case Acceptability, ASME Code Section XI, Division 1," dated March 2017.
- 11. Southern Nuclear Company, NL-16-0724, "Vogtle Electric Generating Plant,,Units 1
& 2, Proposed lnservice Inspection Alternative VEGP-ISI-AL T-11, Version 1.0," June 28, 2016, ADAMS Accession No. ML16180A046.
- 12. Southern Nuclear Company, NL-16-0723, "Joseph M. Farley Nuclear Plant, Unit 1, Proposed lnservice Inspection Alternative FNP-ISI-ALT-19, Version 1.0," June 30, 2016, ADAMS Accession No. ML16182A475.
- 13. Michael T. Markley (NRC) to Charles R. Pierce (Southern Nuclear), Vogtle Electric Generating Plant, Units 1 and 2, and Joseph M. Farley Nuclear Plant, Unit 1 -
Alternative to lnservice Inspection Regarding Reactor Pressure Vessel Threads El-7 to NL-19-0832 Proposed Alternative VEGP-ISI-AL T-04-04, Version 1.0, in Accordance with 1 O CFR 50.55a(z)(1) lnflange Inspection (CAC Nos. MF.8061, MFB062, MF8070)," January 26, 2017, ADAMS Accession No. ML ML17006A109.
'White Paper on Suggested Content for PFM Submittals to the NRC," February 27, 2019, ADAMS Accession No. ML19241A545.
- 15. Structural Integrity Associates, Inc. Calculation No, FP-VOG-323, Revision O, "FatiguePro Analysis of Plant Data for Vogtle Units 1 and 2 through 2018".
El-8 to NL-19-0832 Proposed Alternative VEGP-ISI-ALT-04-04, Version 1.0, in Accordance with 10 CFR 50.55a(z)(1)
APPENDIX A VOGTLE UNIT 1 AND UNIT 2 APPLICABILITY El-9 to NL-19-0832 Proposed Alternative VEGP-ISI-ALT-04-04, Version 1.0, in Accordance with 1 O CFR 50.55a(z)(1)
Plant-Specific Applicability Section 9 of Reference [1] provides 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 Vogtle Units 1 & 2 is provided in Table A 1.
Table A 1 indicates that all plant-specific requirements are met for Vogtle Units 1 & 2.
Therefore, the results and conclusions of the EPRI report are applicable to Vogtle Units 1
&2.
Table A 1. Appllcablllty of Reference [1] Representative Analyses to Vogtle Units 1
&2 Category Requirement from Reference Appllcabillty to Vogtle Units 1 & 2
[1]
General The nozzle-to-shell weld shall The Vogtle Units 1 & 2 MS and FW Requirements be one of the configurations nozzle configurations are shown in
- shown in Figure 1-1 or Figure Figures A 1 and A2, and are 1-2 of Reference [1 ].
representative of the configuration shown in Figure 1-1 of Reference [1].
/
The materials of the SG shell, The Vogtle Units 1 & 2 nozzles are FW nozzles, and MS nozzles fabricated of SA-508, Class 2A must be low alloy ferritic steels material, and the SG vessel which conform to the heads/shells are fabricated from SA-requirements of ASME Code, 533, Gr. A, Cl. 2 material. Both of Section XI, Appendix G, these materials conform to the Paragraph G-2110.
requirements of ASME Code,Section XI, Appendix G, Paragraph G-2110.
The number of transients The transient cycles in Table 5-5 of shown in Table 5-5 of Reference [1] meet or exceed the 60-Reference [1] are bounding for year projected cycles for Vogtle Units application over a 60-year 1 and 2 as shown in Table A2 [15].
operating life.
SG Feedwater The piping attached to the FW The Vogtle Units 1 & 2 FW piping Nozzle nozzle must be 14-inch to 18-lines are both 16-inch NPS.
inch NPS.
El-10 to NL-19-0832 Proposed Alternative VEGP-ISI-AL T-04-04, Version 1.0, in Accordance with 1 O CFR 50.55a(z)(1)
Category Requirement from Reference Applicability to Vogtle Units 1 & 2
[1]
The FW nozzle design must The Vogtle Units 1 & 2, FW nozzle have an integrally attached configuration is shown in Figure A 1 thermal sleeve and has an integrally attached thermal sleeve.
I SG Main Steam For Westinghouse and CE Vogtle Units 1 & 2 are Westinghouse Nozzle plants, the piping attached to 4--loop PWRs. The Vogtle Units 1 &
the SG MS nozzle must be 28-2 MS nozzles have 32" to 26" inch to 36-inch NPS.
reducers. The pipe size of the attached reducer to the nozzle end is 32" NPS which satisfies the intent of this requirement.
For B&W SGs, the piping This requirement is not applicable for attached to the main steam Vogtle Units 1 & 2 because they are nozzle must be 22-inch to 26-both Westinghouse 4-loop units.
inch NPS The SG must have one main As shown in Figure A3, Vogtle Units steam nozzle that exits the top 1 & 2 both have one MS nozzle per dome of the SG.
SG that exits the top dome of each SG.
The main steam nozzle shall The Vogtle Units 1 & 2 MS nozzle not significantly protrude into configuration is shown in Figures A2 the SG (e.g., see Figure 4--7 of and A3, and does not protrude Reference [1]) or have a unique significantly into the SG. The Vogtle nozzle weld configuration (e.g.,
Units 1 & 2 MS nozzles are NOT see Figure 4-6 of Reference unique. They are similar to the
[1 ]).
configuration selected for analysis (Figure 4-8 of Reference [1]).
El-11 to NL-19-0832 Proposed Alternative VEGP-181-AL T-04-04, Version 1.0, in Accordance with 1 O CFR 50.55a{z)(1)
SCHD.
12 SC
- IIO 16" SCHED.
120 SHLS ~ll"f:
077" HA~ L io*-c TYP.
~I E C E/IBOAED JO HATCH C m~cAt A nM~~EHA H
~
C
,o.
C 7,00*
~~\\~b CL. 2A I.6:Z" A, BLf:H.0
,.oo* A, BLEND
~
0
~
~
NOZZL[ ---------------
Figure A 1 Vogtle Units 1 & 2 SG Feedwater Nozzle Configuration El-12 C
to NL-19-0832 Proposed Alternative VEGP-ISI-AL T-04-04, Version 1.0, in Accordance with 1 O CFR 50.55a(z)(1) 0 lO e Figure A2 Vogtle Units 1 & 2 SG Main Steam Nozzle Configuration STEAM OlJTlfl NOZZLE------
lffER HEAD ~-
Figure A3 Steam Generator Upper Head El-13 to NL-19-0832 Proposed Alternative VEGP-ISI-ALT-04-04, Version 1.0, in Accordance with 1 O CFR 50.55a(z)(1)
Table A2 Transient Cycles for Vogtle Units 1 and 2 in Comparison to the Requirements in Reference [1]
Transient Cycles From Unit 160-Year Unit 2 60-Y ear Allowable Table 5-5 of Projected Cycles Projected Cycles Cycles From EPRI From Table 3 of From Table 4 of Tables 3 and 4 Report
[15]
[15]
of [15]
3002014590
[1]
Heatup/Cooldown 300 69 16n5<5) 200 Plant Loading<O 5000 164 141 500 Plant Unloading<2) 5000 66 36 500 Loss of Load<3) 360 119 89 760 Loss of Powef 4) 60 3
3 40 Notes:
(1) Transient listed as Plant Loading 0-15% Power in Tables 3 and 4 of [15].
(2) Transient listed as Plant Unloading 0- 15% Power in Tables 3 and 4 of [15].
(3) Loss of Load transient is a bundled to conservatively envelope a combination of several transients listed in Tables 3 and 4 of [15]:
Loss of Load w/o Rx Trip Loss of RC Flow 1 Loop @ Power Large Step Load Decrease Reactor Trip (CD and SI)
Reactor Trip (CD no SI)
Reactor Trip (No Cooldown)
(4) Transient listed as Loss of Offsite Power in Tables 3 and 4 of [15].
(5) Cycles for Heatup and Cooldown, respectively.
El-14 to NL-19-0832 Proposed Alternative VEGP-ISI-AL T-04-04, Version 1.0, in Accordance with 1 O CFR 50.55a(z)(1)
APPENDIX B VOGTLE UNITS 1 & 2 INSPECTION HISTORY El-15
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to NL-19-0832 Proposed Alternative VEGP-ISI-ALT-04-04, Version 1.0, in Accordance with 1 O CFR 50.55a(z)(1)
VOGTLE UNITS 1 & 2 INSPECTION HISTORY Currently, the MS and FW nozzle components for VEGP Units 1 & 2 satisfy all of the inspection requirements of ASME Code,Section XI, 2007 Edition including the 2008 Addenda.
(Note: The first digit in each Component ID depicts the unit for each component.)
MS Nozzle Date Interval/Period Components ID Exam Coverage<3>
Results Item 10/26/88 1st/1st 11201-B6-001-W18 NRI 50%
No.
10/2/97 2nd /181 11201-B6-001-W18 NRI 50%
C2.21 4/1/08 3rd/1st 11201-B6-001-W18 NRI 50%
3/29/14 3rd/3rd 11201-B6-001-W18 NRI 50%
10/9/90 1st/1st 21201-B6-001-W18 Rl(1l 50%
10/19/99 2nd/1st 21201-B6-001,-W18 NRI 50%
10/02/08 3rd/1st 21201-B6-001-W18 NRI 50%
9/24/14 3rd/3rd 21201-B6-001-W18 NRI 50%
(1)
Subsurface Planer flaw acceptable per IWC-3510-1.
(3)
The following relief requests address <90% lnspectton coverage for the 1ot, ~. and 3rd Intervals* RR-29, RR-14, and VEGP-ISI-RR-05.
El-16 to NL-19-0832 Proposed Alternative VEGP-ISI-AL T-04-04, Version 1.0, in Accordance with 1 O CFR 50.55a(z)(1)
FW Nozzle Date Interval/Period Components ID Item 9/30/94 1st/ 3rd 11201-B6-002-W19 No.
3/29/05 20013rd 11201-86-002-W19 C2.21 10/7/09 3rd/1 el 11201-B6-002-W19 10/10/96 1 el/3rd 21201-B6-002-W19 10/1/05 2nd/3rd 21201-B6-002-W19 3/19/10 3rd/3rd 21201-B6-002-W19 Item 10/6/94 1st/3rd 11201-B6-002-IR04 C2.22 3/25/05 2nd/3rd 11201-B6-002-IR04 10/7/09 3rd/1st 11201-B6-002-IR04 9/27/96 1st/3rd 21201-B6-002-IR04 10/1/05 2nd/3rd 21201-86-002-IR04 3/18/10 3rd/2nd 21201-B6-002-IR04 (2)
Subsurface Planer flaw acceptable per IWC-3510-1.
Exam Coverage<3>
Results NRI 50%
NRI 50%
NRI 80%
Rl(2l 50%
Rl(2l 50%
Rl(2l 80%
NRI 100%
NRI 100%
NRI 100%
NRI 100%
NRI 100%
NRI 100%
(3)
The following relief requests address <90% mspecbon coverage for the 181, 2nd, and 3m Intervals: RR-29, RR-14, and VEGP-ISI-RR-05.
El-17 to NL-19-0832 Proposed Alternative VEGP-ISI-AL T-04-04, Version 1.0, in Accordance with 1 O CFR 50.55a(z)(1)
APPENDIX C RESULTS OF INDUSTRY SURVEY El-18 to NL-19-0832 Proposed Alternative VEGP-ISI-ALT-o+o4, Version 1.0, in Accordance with 1 O CFR 50.55a(z)(1)
Overall Industry Inspection Summery The results of an industry survey of past inspections of SG MS and FW nozzles are summarized in Section 3 of Reference [1]. Table C1 provides a summary of the combined survey results for Item Nos. C2.22, C2.21, and c2.32(1l. The results identify that SG MS and FW Nozzle-to-Shell Welds and Nozzle Inside Radius Section examinations adversely impact outage activities including worker exposure, personnel safety, and radwaste. A total of 74 domestic and international BWR and PWR units responded to the survey and provided information representing all PWR plant designs currently in operation in the U.S.
This included 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). A total of 727 examinations for Item Nos. C2.21, C2.22, and C2.32(1l components were conducted, with 563 of these specifically for PWR components. The majority of the PWR examinations were performed on SG MS and FW nozzles. Only one PWR examination identified two (2) flaws that exceeded ASME Code Section XI acceptance criteria. The flaws were linear indications of 0.3" and 0.5" in length and were detected in a MS nozzle-to-shell weld using magnetic particle examination techniques. The indications were dispositioned by light grinding (ADAMS Accession No. ML13217A093).
Table C1 - Summary of Survey Results Number of
- *: ~u~~r;-.... :~ *. Nutnber:of *: *~, -
l;>le.nf T.ij,e
' -*:
- Reporta6ie
. of Units*
- - Exam-inations.: *_-
Indications BWR 27 164 0
PWR 47 563 2
Totals 74 727 2
1 Item No. C2.32 is similar to Item No. C2.21 and was evaluated in the Reference [1] technical basis and included in the industry survey. Vogtle Units 1 & 2 have not perfonned any examinations on Item No. C2.32 components.
El-19