1CAN122301, Responses to Request for Additional Information for Request for Alternative for Implementation of Extended Reactor Vessel Inservice Inspection Interval (ANO1-ISl-037)
| ML23348A384 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 12/14/2023 |
| From: | Pyle S Entergy Operations |
| To: | Office of Nuclear Reactor Regulation, Document Control Desk |
| Shared Package | |
| ML23348A383 | List: |
| References | |
| 1CAN122301 | |
| Download: ML23348A384 (1) | |
Text
NOTICE: Enclosure 2 to this letter contains Proprietary Information to be withheld from public disclosure per 10 CFR 2.390. Upon separation from Enclosure 2, this letter is DECONTROLLED.
- ) entergy Stephenie Pyle Director Nuclear Licensing 601-368-5516 1CAN122301 December 14, 2023 10 CFR 50.55a ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001
Subject:
Responses to Request for Additional Information for Request for Alternative for Implementation of Extended Reactor Vessel lnservice Inspection Interval (ANO1-ISl-037)
Arkansas Nuclear One, Unit 1 NRC Docket No. 50-313 Renewed Facility Operating License No. DPR-51
References:
- 1) Entergy letter to the NRC, "Request for Alternative for Implementation of Extended Reactor Vessel lnservice Inspection Interval (ANO1-ISl-037)",
(1 CAN062301 ), (ML23159A269), dated June 8, 2023
- 2) NRC email to Entergy "ANO-1 Final RAI RE: Alternative Request ANO1-ISl-037 (EPID L-2023-LLR-0028)," (1CNA112301), dated November 13, 2023 Pursuant to 10 CFR 50.55a(z)(1 ), Entergy Operations, Inc., (Entergy) requested (Reference 1) the U.S. Nuclear Regulatory Commission (NRC) approval to extend the lnservice Inspection (ISi) interval for the Arkansas Nuclear One, Unit 1 (ANO-1) reactor pressure vessel (RPV) weld examinations from 2027 to 2034. Entergy proposes to implement an alternative to the requirement of American Society of Mechanical Engineers (ASME)Section XI, IWB 2411, Inspection Program, that volumetric examination of RPV Examination Categories B-A and B-D be performed once each 10-year ISi interval.
The NRC staff has reviewed the request and determined that additional information is required to complete their review (Reference 2).
The Requests for Additional Information (RAls) and the associated responses are provided in.
Entergy Operations, Inc. 1340 Echelon Parkway, Jackson, MS 39213
1 CAN122301 Page 2 of 3 provides the updated basis for the proposed alternative. Some information provided in Enclosure 2 is considered proprietary to Framatome and request it to be withheld from public disclosure in accordance with 10 CFR 2.390 of the Commission's regulations. The proprietary information is identified by text enclosed within double bolded brackets ((Example)). The non-proprietary version is provided in Enclosure 3.
Per the NRC's Safety Evaluation Report (SER), the results of WCAP-16168-NP-A, Revision 3, "Risk-Informed Extension of the Reactor Vessel In-service Inspection Interval," [ADAMS Accession Number ML11306A084], dated October 2011, can only be used on the Babcock &
Wilcox (B&W)-designed ANO-1 after the assumption that an equivalent of 12 heat-up/ cool-down cycles per year of operation can be validated to bound all of its design basis transients that contribute significantly to fatigue crack growth. For completeness, Enclosure 4, which provides the summary of the analysis validating this assumption, is provided with no changes made from the information provided in Reference 1, Enclosure 3.
This information is supported by an affidavit, signed by Philip A. Opsal, Manager, Product Licensing, for Framatome Inc. (Framatome, 3315 Old Forest Road, Lynchburg, VA 24501), the owner of the information. The affidavit sets forth the basis by which the information may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in paragraph (a)(4) of 10 CFR 2.390 of the Commission's regulations. The affidavit is included in Enclosure 5.
This letter contains no new regulatory commitments.
If there are any questions or if additional information is needed, please contact Riley Keele, Manager, Regulatory Assurance, at (479) 858-7826.
Respectfully, Ste Ph en,1 e D;g;tally s;gned by Stephen;e Pyle ON: cn=Stephenie Pyle, c=US, o=Entergy, ou=Director, Fleet P I Regulatory Assurance, email=spyle@entergy.com y e Date: 2023.12.14 17:37:12 -06'00' Stephenie Pyle SP/bka
Enclosures:
- 1.
Responses to Request for Additional Information
- 2.
Request for Alternative ANO1-ISl-037 (PROPRIETARY)
- 3.
Request for Alternative ANO1-ISl-037 (NON-PROPRIETARY)
- 4.
Equivalent Fatigue Crack Growth for Arkansas Nuclear One Unit 1 Beltline Shell Location (Framatome Document 86-9352400-000)
- 5.
1 CAN122301 Page 3 of 3 cc:
NRC Region IV Regional Administrator NRC Senior Resident Inspector - Arkansas Nuclear One NRC Project Manager - Arkansas Nuclear One
ENCLOSURE 1 1CAN122301 RESPONSES TO REQUEST FOR ADDITIONAL INFORMATION
1 CAN122301 Page 1 of 3 RESPONSES TO REQUEST FOR ADDITIONAL INFORMATION By Reference 1, Entergy Operations, Inc., (Entergy) requested the U.S. Nuclear Regulatory Commission (NRC) approval to extend the inservice inspection interval for the Arkansas Nuclear One, Unit 1 (ANO-1) reactor pressure vessel (RPV) weld examinations from 2027 to 2034. Entergy proposes to implement an alternative to the requirement of American Society of Mechanical Engineers (ASME)Section XI, IWB 2411, Inspection Program, that volumetric examination of RPV Examination categories B-A and B-0 be performed once each 10-year lnservice Inspection (ISi) interval.
The NRC staff has reviewed the request and determined that additional information was required to complete their review (Reference 2).
Below are the Requests for Additional Information (RAls) and the associated responses.
NRC REOUESJ {BAl-1}
Issue In Section 5 of Enclosure 2 (Reference 1) of the alternative request, the licensee stated that the volumetric examinations of the subject Reactor Vessel (RV) welds for fifth inservice inspection, currently scheduled for 2027 are proposed to be performed in 2036. The NRC staff noted that that this proposed rescheduled inspection date of 2036 occurs after the May 20, 2034, expiration date of Renewed Facility Operating License No. DPR-51 for ANO-1 (ML053130314).
In addition, the cover letter for the request states, in part, that "Entergy requests approval of the requested alternative to the end of the current renewed ANO-1 Operating License (May 2034 ). "
However, in Section 6, "Duration of Proposed Alternative," of Enclosure 2 to the request states, "This request is applicable to the ANO-1 inservice inspection program for the fifth and sixth 1 O year inspection intervals."
Request Update the proposed rescheduled inspection date of 2036 to a date prior to the expiration date of May 20, 2034, of the Renewed Facility Operating License No. DPR-51 for ANO-1 throughout the document.
ENTERGY'S RESPONSE (RAl-1)
The inspection schedule has been updated to prior to May 20, 2034, see Enclosure 2 of this letter.
1 CAN122301 Page 2 of 3 NRC REQUEST {RAl-2}
Issue The licensee's proposed alternative is based on the methodology in NRC-approved topical report WCAP-16168-NP-A, Revision 3, "Risk-Informed Extension of the Reactor Vessel lnservice Inspection Interval" (ML11306A084 ). For Item (6) in Section 3.4 of the NRC staff's safety evaluation (SE) for WCAP-16168-NP-A, Revision 3, dated July 26, 2011, the NRC staff stated that "Licensees seeking second or additional interval extensions shall provide the information and analyses requested in Section (e) of 10 CFR 50.61 a." It is not clear whether the proposed alternative in ANO1-ISl-037 is the first interval extension request for ANO-1 associated with WCAP-16168-NP-A, Revision 3, and accordingly, that Item (6) in Section 3.4 of the NRC staff's SE dated July 26, 2011, does not apply to ANO-1.
Request Confirm whether the proposed alternative in ANO1-ISl-037 is the first interval extension request for ANO-1 associated with WCAP-16168-NP-A, Revision 3, and that therefore, Item (6) in Section 3.4 of the NRC staff's Safety Evaluation (SE) dated July 26, 2011, does not apply to ANO-1.
ENTERGY'S RESPONSE (RAl-2)
It is confirmed that the proposed alternative in ANO1-ISl-037 is the first interval extension request for ANO-1 associated with WCAP-16168-NP-A, Revision 3, and that therefore, Item (6) in Section 3.4 of the NRC staff's SE dated July 26, 2011, does not apply to ANO-1.
NRC REQUEST {BAl-3}
It is not clear whether the fluence values in Table 3 of Enclosure 2 to the submittal are at the clad-to-base metal interface of the RV, since the values are slightly lower than those in the referenced source (i.e., Areva Report ANP-3300, Revision 0, "Arkansas Nuclear One (ANO)
Unit 1 Pressure-Temperature Limits at 54 EFPY [Effective Full Power Years]," dated June 2014 (ML14241A241 )).
Request Clarify whether the fluence values in Table 3 of Enclosure 2 to the submittal are the at the clad-to-base metal interface of the RV.
ENTERGY'S RESPONSE (RAl-3)
The fluence values in Table 3 of Enclosure 2 were calculated at the inside wetted surface.
However, as discussed in the response of RAI 4, the fluence values have been updated in the relief request to the fluence values provided in ANP-3300, Revision 1 to be consistent with the site's current licensing basis. The differences in the fluence does not change the overall
1 CAN122301 Page 3 of 3 conclusions of the Through-Wall Cracking Frequency (TWCF) calculations since there is large margin between the pilot plant and ANO-1 TWCF by an order of magnitude of 4.
NRC REQUEST {RAl-4}
Issue The NRC staff noted that the referenced source for the information in Table 3 of Enclosure 2 to the submittal is Areva Report ANP-3300, Revision 0. The NRC staff further noted that Areva Report ANP-3300, Revision 0, was revised in November 2014 to ANP-3300, Revision 1 (ML14330A250), and therefore, the revised report should be the current licensing basis (CLB) analysis of record. ANP-3300, Revision 1, is the technical basis in a license amendment request regarding the ANO-1 pressure-temperature (P-T) limits for 54 EFPY (ML14330A249), which the NRC staff approved by Amendment No. 254, dated April 24, 2015 (ML15096A324). Lastly, the NRC staff noted that Table 3 of Enclosure 2 to the submittal contains some input values (e.g.,
the initial reference temperature values) that are different than those reported in Table 3-1 of ANP-3300, Revision 1.
Request Update the information in Table 3 of Enclosure 2 to the submittal such that all input values reflect those from the CLB analysis of record in ANP-3300, Revision 1, and that the results are recalculated, as necessary.
ENTERGY'S RESPONSE (RAl-4)
The reference to ANP-3300 has been updated to Revision 1, which is the CLB analysis of record for ANO Unit 1. The input values (e.g., the initial reference temperature values) have been updated based on ANP-3300, Revision 1. In addition, the fluence values have been updated to the wetted surface fluence values from ANP-3300, Revision 1. The through-wall cracking frequency has been recalculated using the updated initial reference temperature values and fluence values and has been provided in Enclosure 2 of this letter. The differences in the initial reference temperature values and fluence values does not change the overall conclusions of the TWCF calculations since there is large margin between the pilot plant and ANO-1 TWCF by an order of magnitude of 4.
References
- 1. Entergy letter to the NRC, "Request for Alternative for Implementation of Extended Reactor Vessel lnservice Inspection Interval (ANO1-ISl-037)", (1 CAN062301 ), (ML23159A269),
dated June 8, 2023
- 2. NRC email to Entergy "ANO-1 Final RAI RE: Alternative Request ANO1-ISl-037 (EPID L-2023-LLR-0028)," (1CNA112301), dated November 13, 2023
ENCLSURE 3 1CAN122301 REQUEST FOR ALTERNATIVE ANO1-ISl-037 (NON-PROPRIETARY)
1 CAN122301 Page 1 of 8 Request for Alternative for Implementation of Extended Reactor Vessel lnservice Interval (ANO-ISI-037)
- 1. ASME Code Component(s) Affected The affected component is the Arkansas Nuclear One (ANO) Unit 1 reactor vessel (RV),
specifically, the following American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (BPV) Code,Section XI (Reference 1) examination categories and item numbers covering examinations of the RV. These examination categories and item numbers are from IWB-2500 and Table IWB-2500-1 of the ASME BPV Code,Section XI.
Category B-A welds are defined as "Pressure Retaining Welds in Reactor Vessel." Category B-D welds are defined as "Full Penetration Welded Nozzles in Vessels."
Examination Category Item No.
Description B-A B1.10 Shell Welds B-A B1.11 Circumferential Shell Welds B-A B1.12 Longitudinal Shell Welds B-A B1.20 Head Welds B-A B1.21 Circumferential Head Welds B-A B1.30 Shell-to-Flange Weld B-A B1.50 Repair Welds B-A B1.51 Beltline Welds B-D B3.90 Nozzle-to-Vessel Welds B-D B3.100 Nozzle Inside Radius Section (Throughout this request, the above examination categories are referred to as "the subject examinations" and the ASME BPV Code,Section XI, is referred to as "the Code.")
2. Applicable Code Edition and Addenda
ASME Code Section XI, "Rules for lnservice Inspection of Nuclear Power Plant Components," 2007 Edition though 2008 Addenda (Reference 1 ).
3. Applicable Code Requirement
IWB-2411, Inspection Program, requires volumetric examination of essentially 100% of reactor vessel pressure-retaining welds identified in Table IWB-2500-1 once each 10-year interval. The fifth 10-year inservice inspection (ISi) interval for ANO Unit 1 is scheduled to end on May 30, 2027. The applicable Code for the sixth 10-year ISi interval will be selected in accordance with the requirements of 10 CFR 50.55a.
1 CAN122301 Page 2 of 8
4. Reason for Request
An alternative is requested from the requirement of the IWB-2411 Inspection Program, that volumetric examination of essentially 100% of reactor vessel pressure-retaining Examination Category B-A and B-D welds be performed once each 10-year interval. Extension of the interval between examinations of Category B-A and B-D welds from 10 years to up to 20 years will result in a reduction in man-rem exposure and examination costs.
5. Proposed Alternative and Basis for Use
Entergy proposes not to perform the ASME Code required volumetric examination of the ANO Unit 1 reactor vessel full penetration pressure-retaining Examination Category B-A and B-D welds for the fifth inservice inspection, currently scheduled for 2027. Entergy will perform the fifth ASME Code required volumetric examination of the ANO Unit 1 reactor vessel full penetration pressure-retaining Examination Category B-A and B-D welds in the sixth inservice inspection interval prior to May 20, 2034.
The proposed inspection date is a deviation from the latest revised implementation plan, OG-10-238 (Reference 2), since the implementation plan reflects the next inspection being performed in 2028 for ANO Unit 1. The impact to the implementation plan in OG-10-238 would increase the number of inspections in 2034 (from five to six) and decrease the number of inspections in 2028 (from five to four). Based on Figure 3 and Figure 4 of OG-10-238, this proposed inspection schedule is considered to have a minor impact on the future inspection plan and the distribution of inspections over time.
In accordance with 10 CFR 50.55a(z)(1 ), an alternate inspection interval is requested on the basis that the current interval can be revised with negligible change in risk by satisfying the risk criteria specified in Regulatory Guide 1.174 (Reference 3). The methodology used to conduct this analysis is based on that defined in the study WCAP-16168-NP-A, Revision 3, "Risk-Informed Extension of the Reactor Vessel In-service Inspection Interval" (Reference 4 ).
This study focuses on risk assessments of materials within the beltline and extended beltline regions of the RV wall. The results of the calculations for ANO Unit 1 were compared to those obtained from the Babcock and Wilcox (B&W) pilot plant evaluated in WCAP-16168-NP-A, Revision 3. Appendix A of WCAP-16168-NP-A identifies the parameters to be compared.
Demonstrating that the parameters for ANO Unit 1 are bounded by the results of the B&W pilot plant qualifies ANO Unit 1 for an ISi interval extension.
1 CAN122301 Page 3 of 8 Table 1 below lists the critical parameters investigated in WCAP-16168-NP-A and compares the results of the B&W pilot plant to those of ANO Unit 1. Tables 2 and 3 provide additional information that was requested by the NRC and included in Appendix A of Reference 4.
Table 1:
Critical Parameters for the Application of Bounding Analysis for AN0-1 Pilot Plant Plant-Specific Additional Parameter Evaluation Basis Basis Required?
Dominant Pressurized Thermal NRC PTS Risk PTS Generalization Shock (PTS) Transients in the NRC Study (Reference 5)
Study No PTS Risk Study are Applicable (Reference 6) 1.08E-10 Events Through-Wall Cracking Frequency 4.42E-07 Events per per year No (TWCF) year (Reference 4)
(Calculated per Reference 4)
Yes (as 12 heatup/cooldown Bounded by required by Frequency and Severity of Design cycles per year 12 heatup/cooldown Reference 4 Basis Transients and (Reference 4) cycles per year summarized in Reference 10)
Cladding Layers (Single/Multiple)
Single Layer Single No (Reference 4)
Layer
1 CAN122301 Page 4 of 8 Table 2 below provides a summary of the latest reactor vessel inspection for ANO Unit 1 and an evaluation of the recorded indications. This information confirms that satisfactory examinations have been performed on the ANO Unit 1 reactor vessel.
Table 2: Additional Information Pertaining to Reactor Vessel Inspection for ANO Unit 1 The latest RV ISi for ANO Unit 1 was conducted in accordance with the requirements of Appendix VIII of the ASME Code,Section XI, 2001 Inspection methodology:
Number of Past Inspections:
Number of indications found:
Proposed inspection schedule for balance of plant life:
Edition through 2003 (Reference 11 ). Examinations of Category B-A and B-D welds were performed to the acceptance standards of Section XI, Appendix VIII, 2001 Edition through the 2003 as modified by 10 CFR 50.55a(b)(2)(xiv, xv and xvi). Future inservice inspections will continue to be performed to ASME Section XI, Appendix VIII methodology.
Four 10-Year inservice inspections and a pre-service inspection have been performed.
There was one indication identified in the beltline and extended beltline regions during the most recently completed inservice inspection. This subsurface indication is located in the lower shell longitudinal weld (Item 12/13 in Table 3). The indication is acceptable per Table IWB-3510-1 of Section XI of the ASME Code. The indication is not within the inner 1/10th or 1 inch of the reactor vessel thickness; therefore, it is inherently acceptable per the requirements of the Alternate PTS Rule, 10 CFR 50.61a (Reference 7).
The fourth 10-year inspection was the first ISi examination that detected the one indication described above. There is no site-specific flaw growth data since this indication was evaluated as acceptable per ASME Section XI Table IWB-3510-1.
The fifth inservice inspection is scheduled for 2027. This inspection will instead be performed prior to May 20, 2034. The proposed inspection date is a deviation from the latest revised implementation plan, OG-10-238 (Reference 2), since the implementation plan reflects the next inspection being performed in 2028 for ANO Unit 1. The impact to the implementation plan in OG-10-238 would increase the number of inspections in 2034 (from five to six) and decrease the number of inspections in 2028 (from five to four). Based on Figure 3 and Figure 4 of OG-10-238, this proposed inspection schedule is considered to have a minor impact on the future inspection plan and the distribution of inspections over time.
1 CAN122301 Page 5 of 8 Table 3 summarizes the inputs and outputs for the calculation of through-wall cracking frequency (TWCF).
Table 3: Details of TWCF Calculation for ANO Unit 1 at 54 Effective Full Power Years (EFPY) lnputs11I Above LNBF taper T wa 11 [inches]:
~elow LN~t-taper I wall [inches]:
Material Heat Copper Nickel R.G.1.99 Chemistry RTNDT(u)
No.
Region and Component Description No.
[weight%]
[weight%]
Position Factor [°F]
[OF]
Identification 1
Lower Nozzle Beltline Forging (LNBF)
AYN 131 0.03 0.70 1.1 20.0 27.5 2
Upper Shell Plate 1 C5120-2 0.17 0.55 1.1 122.75 1
3 Upper Shell Plate 2 C5114-2 0.15 0.52 1.1 105.6 10 4
Lower Shell Plate 1 C5120-1 0.17 0.55 1.1 122.75 1
5 Lower Shell Plate 2 C5114-1 0.15 0.52 1.1 105.6 30 6
Dutchman Forging (Transition Piece) 125W609V
((
))
((
))
1.1
((
))
((
))
A1 7
Lower Nozzle Beltline Forging to Upper Shell WF-182-1 0.24 0.63 1.1 177.95
-84.2 Circumferential Weld 8
Upper Shell to Lower Shell Circumferential WF-112 0.27 0.59 1.1 182.55
-98.0 Weld 9
Lower Shell to Dutchman Forging (Transition SA-1788
((
))
((
))
1.1
((
))
((
))
Piece) Weld 10 Upper Shell Longitudinal Weld 1 WF-18 0.19 0.57 1.1 167.0
-48.6 11 Upper Shell Longitudinal Weld 2 WF-18 0.19 0.57 1.1 167.0
-48.6 12 Lower Shell Longitudinal Weld 1 WF-18 0.19 0.57 1.1 167.0
-48.6 13 Lower Shell Longitudinal Weld 2 WF-18 0.19 0.57 1.1 167.0
-48.6 Outputs Methodology Used to Calculate.6T30: Regulatory Guide 1.99, Revision 2 (Reference 8)
Controlling RTMAX-XX Fluence FF Material axx
[OR]
[Neutron/cm2, (Fluence
.6T30 [°F]
Region No.
E >1.0 MeV]
Factor)
Limiting Axial Weld - AW 5
2.5000 599.65 1.16E+19 1.0414 109.98 Limiting Plate - PL 5
2.5000 603.64 1.33E+19 1.0793 113.97 Limiting Forging - FO 1
2.5000 508.28 1.22E+19 1.0554 21.11 Limiting Circumferential Weld - CW 5
2.5000 602.98 1.30E+19 1.0730 113.31 TWCF95-TOTAL(aAWTWCF95-AW + aPLTWCF95-PL + aFOTWCF95-FO + aCWTWCF95-CW):
12 8.44 Fluence
[Neutron/cm2, E > 1.0 MeV]
1.22E+19 1.35E+19 1.33E+19 1.61E+17 1.22E+19 1.30E+19 1.61E+17 1.08E+19 1.16E+19 TWCF95-XX 0.000E+00 4.302E-11 5.590E-15 0.000E+00 1.08E-10 Note 1: Material properties and fluence were based on ANP-3300, Revision 1 (Reference 9). Fluence projections and material properties not included in Reference 9 were provided by Framatome.
1 CAN122301 Page 6 of 8
6. Duration of Proposed Alternative
This request is applicable to the ANO Unit 1 inservice inspection program for the fifth and sixth 10-year inspection intervals.
7. Precedents
"Surry Power Station Units 1 and 2 - Relief Implementing Extended Reactor Vessel Inspection Interval (TAC Nos. ME8573 and ME8574)," dated April 30, 2013, Agencywide Document Access and Management System (ADAMS) Accession Number ML13106A140.
"Vogtle Electric Generating Plant, Units 1 and 2 - Request for Alternatives VEGP-ISI-AL T-05 and VEGP-ISI-AL T-06 (TAC Nos. MF2596 and MF2597)," dated March 20, 2014, ADAMS Accession Number ML14030A570.
"Catawba Nuclear Station Units 1 and 2: Proposed Relief Request 13-CN-003, Request for Alternative to the Requirement of IWB-2500, Table IWB-2500-1, Category B-A and Category B-D for Reactor Pressure Vessel Welds (TAC Nos. MF1922 and MF1923),"
dated March 26, 2014, ADAMS Accession Number ML14079A546.
"Sequoyah Nuclear Plant, Units 1 and 2 - Requests for Alternatives 13-ISl-1 and 13-ISl-2 to Extend the Reactor Vessel Weld lnservice Inspection Interval (TAC Nos. MF2900 and MF2901 )," dated August 1, 2014, ADAMS Accession Number ML14188B920.
"Byron Station, Unit No. 1 - Relief from Requirements of the ASME Code to Extend the Reactor Vessel lnservice Inspection Interval (TAC No. MF3596)," dated December 10, 2014, ADAMS Accession Number ML14303A506.
"Wolf Creek Generating Station - Request for Relief Nos. I3R-08 and I3R-09 for the Third 10-Year lnservice Inspection Program Interval (TAC Nos. MF3321 and MF3322)," dated December 10, 2014, ADAMS Accession Number ML14321A864.
"Callaway Plant, Unit 1 - Request for Relief I3R-17, Alternative to ASME Code Requirements Which Extends the Reactor Vessel Inspection Interval from 10 to 20 Years (TAC No. MF3876)," dated February 10, 2015, ADAMS Accession Number ML15035A148.
"Braidwood Station, Units 1 and 2 - Request for Relief from the Requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) (CAC Nos. MF8191 and MF8192)," dated March 15, 2017, ADAMS Accession Number ML17054C255.
"South Texas Project, Units 1 and 2 - Relief from the Requirements of the ASME Code Regarding the Third 10-Year lnservice Inspection Program Interval (EPID L-2018-LLR-0010)," dated July 24, 2018, ADAMS Accession Number ML18177A425.
"Donald C. Cook Nuclear Plant, Unit No. 1 - Approval of Alternative to the ASME Code Regarding Reactor Vessel Weld Examination - Relief Request ISIR-4-08 (EPID L-2018-LLR-0106)," dated October 26, 2018, ADAMS Accession Number ML18284A310.
"R. E. Ginna Nuclear Power Plant-Issuance of Relief Request ISl-18 Regarding Fifth 10-year lnservice Inspection Program Interval (EPID L-2018-LLR-0104 )," dated April 22, 2019, ADAMS Accession Number ML19100A004.
1 CAN122301 Page 7 of 8 "Point Beach Nuclear Plant, Units 1 and 2 -Approval of Relief Requests 1-RR-13 and 2-RR-13 Regarding Extension of Inspection Interval for Reactor Pressure Welds from 10 to 20 years (EPID L-2019-LLR-0060)," dated March 4, 2020, ADAMS Accession Number ML20036F261.
"St. Lucie Plant, Unit 2 -Authorization of RR#15 Regarding Extension of ASME Requirements Related to Reactor Pressure Vessel Weld Examinations from 10 to 20 Years (EPID L-2020-LLR- 0283)," dated September 30, 2021, ADAMS Accession Number ML21236A131.
"Oconee Nuclear Station, Units 1, 2, and 3 - Authorization and Safety Evaluation for Alternative Reactor Pressure Vessel lnservice Inspection Intervals (EPID L-2021-LLR-0004 )," dated November 19, 2021, ADAMS Accession Number ML21281A141.
"Turkey Point Nuclear Generating Unit Nos. 3 and 4 - Authorization of Relief Request Nos. 8 and 9 Regarding Extension of Inspection Interval for Reactor Pressure Vessel Welds (EPID L-2021-LLR- 0038)," dated May 10, 2022, ADAMS Accession Number ML22123A192.
- 8. References
- 1. ASME Boiler and Pressure Vessel Code,Section XI, "Rules for lnservice Inspection of Nuclear Power Plant Components" 2007 Edition with 2008 Addenda, ASME International.
- 2. OG-10-238, "Revision to the Revised Plan for Plant Specific Implementation of Extended lnservice Inspection Interval per WCAP-16168-NP, Revision 1, "Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval."
PA-MSC-0120," July 12, 2010 [ADAMS Accession Number ML11153A033].
- 3. NRC Regulatory Guide 1.17 4, Revision 1, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," U.S. Nuclear Regulatory Commission, November 2002 [ADAMS Accession Number ML023240437].
- 4. Westinghouse Report, WCAP-16168-NP-A, Revision 3, "Risk-Informed Extension of the Reactor Vessel In-service Inspection Interval," October 2011 [ADAMS Accession Number ML11306A084].
- 5. NUREG-1874, "Recommended Screening Limits for Pressurized Thermal Shock (PTS)," U.S. Nuclear Regulatory Commission, March 2010 [ADAMS Accession Number ML15222A848].
- 6. NRC Letter Report, "Generalization of Plant-Specific Pressurized Thermal Shock (PTS) Risk Results to Additional Plants," U.S. Nuclear Regulatory Commission, December 14, 2004 [ADAMS Accession Number ML042880482].
- 7. Code of Federal Regulations, 10 CFR Part 50.61 a, "Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events," U.S. Nuclear Regulatory Commission, Washington D.C., Federal Register, Volume 75, No. 1, dated January 4, 2010 and No. 22 with corrections to part (g) dated February 3, 2010, March 8, 2010, and November 26, 2010.
1 CAN122301 Page 8 of 8
- 8. NRC Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials, " U.S. Nuclear Regulatory Commission, May 1988 [ADAMS Accession Number ML003740284].
- 9. Areva Report ANP-3300, Revision 1, "Arkansas Nuclear One (ANO) Unit 1 Pressure-Temperature Limits at 54 EFPY," November 2014 [ADAMS Accession Number ML14330A250].
- 10. Framatome Document, 86-9352400-000, "Equivalent Fatigue Crack Growth for Arkansas Nuclear One Unit 1 Beltline Shell Location."
- 11. ASME Boiler and Pressure Vessel Code,Section XI, "Rules for lnservice Inspection of Nuclear Power Plant Components" 2001 Edition with 2003 Addenda, ASME International.
ENCLOSURE 4 1CAN122301 EQUIVALENT FATIGUE CRACK GROWTH FOR ARKANSAS NUCLEAR ONE UNIT 1 BELTLINE SHELL LOCATION (FRAMATOME DOCUMENT 86-9352400-000)
1CAN122301 Page 1 of 10 Controlled Document 0402-01-F01 (Rev_ 021, 03/1 2/2018) framatome CALCULATION
SUMMARY
SHEET (CSS)
Document No_
86 9352400 000 Safety Related: GJYes D No Title Equivalent Fatigue Crack Growth for Arkansas Nuclear One Unit 1 Beltline Shell Location PURPOSE AND
SUMMARY
OF RESULTS:
Purpose:
Per the NRC's Safety Evaluation Report (SER), the results of WCAP-16168-NP-A, Revision 3 (Reference [1]) can only be used on B&W-designed Arkansas Nuclear One Unit 1 (ANO-1) after the assumption that an equivalent of 12 heat-up I cool-down cycles per year of operation can be validated to bound all of its design basis transients that contribute significantly to fatigue crack growth_ This document provides the summary of the analysis validating this assumption_
Summary of Results: Based on the results of the analysis summarized herein, the equivalent fatigue crack growth (Table 4--2) using 12 equivalent heat-up and cool-down cycles per year is larger than the detailed transient fatigue crack growth from the detailed design transients (Table 4-1 );
therefore, the assumption of WCAP-16168-NP-A, Reference [1], for Arkansas Nuclear One Unit 1 that 12 equivalent heat-up and cool-down cycles bound the fatigue crack growth from all the other test, normal / upset and emergency I faulted service level transients is valid.
The maximum RT MAX-Fo for ANO-1 forging= 52.3°F {512.0°R).
If the computer software used herein Is not the latest version per the EASI 11st, AP 0402-01 requires that Justification be provided.
THE FOLLOWING COMPUTER CODES HAVE BEEN USED IN THIS DOCUMENT:
CODENERSIONIREV CODENERSION/REV N/A THE DOCUMENT CONTAINS ASSUMPTIONS THAT SHALL BE VERIFIED PRIOR TO USE Yes
[8:1 No Page 1 of 10
1CAN122301 Page 2 of 10 Controlled Document framatome 0402-01 -F01 (Rev. 021, 03/12/2018)
Document No. 86-9352400-000 Equivalent Fatigue Crack Growth for Arkansas Nuclear One Unit 1 BelUine Shell Location Review Method: ~
Design Review (Detailed Check)
D Alternate Calculation Does this document establish design or technical requirements?
DYES
~NO Does this document contain Customer Required Format?
DYES
~
NO Signature Block Name and Title P/R/A/M Signature and (printed or typed)
LP/LR Martin Kolar, MKOLAR p
Principal Engineer 12/1/2022 Luziana Matte, LR MATTE Advisory Engineer 1211/2022 R
Ryan Hosler, RS HOSLER Supervisory Engineer 12/1/2022 A
Notes: P/R/A designates Preparer (P), Reviewer (R), Approver (A);
LP/LR designates Lead Preparer (LP), Lead Reviewer (LR);
M designates Mentor (M)
Date Pages/Sections Prepared/Reviewed/Approved All Pages / All Sections All Pages / All Sections All Pages / All Sections In preparing, reviewing and approving revisions, the lead preparer/reviewer/approver shall use 'All' or 'All except
' in the pages/sections reviewed/approved. 'All' or
- All except _
' means that the changes and the effect of the changes on the entire document have bee.n prepared/reviewed/approved. It does not mean that the lead preparer/reviewer/approver ha~ prepared/reviewed/approved all the pages of the document.
With Approver pennission, calculations may be revised without using the latest CSS form_ This deviation is permitted when expediency and/or cost are a factor_ Approver shall add a comment in the right-most column that acknowledges and justifies this deviation.
Project Manager Approval of Customer References and/or Customer Formatting (N/A if not applicable)
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1CAN122301 Page 3 of 10 Controlled Document framatome 0402-01-F01 (Rev. 021, 03/l2/2018)
Document No. 86-9352400-000 Equivalent Fatigue Crack Growth for Arkansas Nuclear One Unit 1 Beltline Shell Location Record of Revision Revision Pages/Sections/Paragraphs No.
Changed Brief Description / Change Authorization 000 All Pages / All Sections Initial Issue of the Document Page 3
1CAN122301 Page 4 of 10 Controlled Document framatome Document No. 86-9352400-000 Equivalent Fatigue Crack Growth for Arkansas Nuclear One Unit 1 BelUine Shell Location Table of Contents Page SIGNATURE BLOCK.................................
..... 2 RECORD OF REVISION...........................................................................
................... 3 LIST OF TABLES............................................................................................................................... 5
'1.0 2.0 3.0 PURPOSE................................................................................................................................ 6 ASSUMPTIONS.......................................................................................................................... 6 2.1 Unverified Assumptions.......................... ************************************ ***************.................................. 6 2.2 Justified Assumptions. **********************************************************************************************************************6 INPUTS..................................................................................................................................... 7 3.1 Temperature / Pressure Transients.................................................................................................. 7 4.0 RESULTS...................................................................................................................................... 8 4.1 Transient Fatigue Crack Growth Summary...................................................................................... 8 4.2 Equivalent Crack Growth...................................................................
- 8 4.3 Transient Crack Growth Contribution Ratio................................................................................ 8
5.0 CONCLUSION
.............................................................................................................................. 9 5.1 Fatigue Crack Growth....................................................................................................................... 9 5.2 RT MAX Temperature........................................................................................................................... 9
6.0 REFERENCES
........................................................................................................... 10 Page4
1CAN122301 Page 5 of 10 framatome Controlled Document Document No. 86-9352400-000 Equivalent Fatigue Crack Growth for Arkansas Nuclear One Unit 1 Beltline Shell Location List of Tables Page Table 3-1: List of Transient Events Considered for AN0-1....................................................................... 7 Table 4-1: Transient Fatigue Crack Growth Summary............................................................................ 8 Table 4-2: Equivalent Fatigue Crack Growth........................................................................................... 8 Table 4-3: Crack Growth Relative Contribution.......................................................................... 9 Page5
1CAN122301 Page 6 of 10 framatome Controlled Document Document No_ 86-9352400-000 Equivalent Fatigue Crack Growth for Arkansas Nuclear One Unit 1 Beltline Shell Location 1.0 PURPOSE This document is part of the Risk-Informed Extension of Reactor Vessel In-Service Inspection Interval PWR Owners Group project PA-MSC-0943, B&W Site Specific Fatigue Crack Growth Evaluation_
Westinghouse Electric Company (WEC) performed a Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval analysis documented in WCAP-16168-P-A, Revision 3 and US NRC approved this work in 20 11 (Reference [ 1)) _
Per the RC's Safety Evaluation Report (SER), the results of WCAP-16168-NP-A, Revision 3 (Reference [11) can only be used on B&W designed Arkansas Nuclear One Unit 1 if the following requirements outlined in Appendix B (items 3 and 4) of Reference [1] are addressed:
Item 3:
Licensees must verify that the fatigue crack growth of 12 heat-up/cool-down transients per year bound the fatigue crack growth for all of its design basis transients_
Licensees must identify tl1e design basis transients that contribute to significant fatigue crack growth_
Item 4:
If the subject plant has reactor vessel forgings that are susceptible to underclad cracking with RTMAX-FO values exceeding 240°F, then the WCAP analyses are not applicable_ The licensees must submit a plant-specific evaluation for any extension to the 10-year inspection interval for ASME Code,Section XI, Category B-A and B-D RPV welds_
This document is a su=ary of an analysis validating item 3 and item 4 for a 20 year in-service inspection (ISi) interval_
This current su=ary document is a non-proprietary document in support of a relief request being prepared by Westinghouse_
2.0 ASSUMPTIONS 2.1 Unverified Assumptions o 1mverified assumption was made during the preparation of this document 2.2 Justified Assumptions L For the purpose of this document, based on the chemical composition, ilie cladding weld material is taken as 18Cr-8Ni (fype 304 stainless steel)_
2_
Due to the similarities in the base metal material chemical compositions, the reactor vessel shell matenal is assumed as low-alloy steel SA-508 Class 2 for the entire region of interest Page 6
1CAN122301 Page 7 of 10 Controlled Document framatome Document No. 86-9352400-000 Equivalent Fatigue Crack Growth for Arkansas Nuclear One Unit 1 Beltline Shell Location 3.0 INPUTS 3.1 Temperature I Pressure Transients The inside surface of the reactor vessel is subjected to transient loads in the form of primary coolant cold leg temperatures and pressures as defined by the reactor coolant system functional specificatioIL The number of applicable transient cycles corresponds to 20 years of operation. In order to form complete stress cycles, these individual transients are combined into transient groups.
These combined transients are used to calculate the cyclic variations of stress intensity factor that produce fatigue crack growth over the life of the plant.
Table 3-1 lists transient events considered in this evaluation.
Table 3-1: List of Transient Events Considered for AN0-1 Transient Transient Name / Description Category Number 11 l A Heat-up from 0% to 8% Power normal 18 Cool-down from 8% to 0% Power normal lC Technical Specification Cooldown emergency 2A Power Change from 0% to 15%
normal 28 Power Change from 15% to 0%
normal 3
Power Loading from 8% to 100%
normal 4
Power Unloading from 100% to 8%
normal s
10% Step load Increase normal 6
10% Step load Decrease normal 7
Step load Reduction from 100% to 8% Power upset SA Reactor Trip - Type A (Loss of RC Flow) upset 88 Reactor Trip - Type B (Turbine Trip) upset SC Reactor Trip - Type C (Loss of MFW Flow) upset 9
Rapid Depressurization upset 10 Change of Flow upset 11(2)
Rod Withdrawal Accident upset 12 Hydro-test test 14 Control Rod Drop upset 15(2) loss of Station Power uoset 16 Steam Line Failure faulted 17Al2J loss of Feedwater to One Steam Generator upset 178(2)
Stuck Open Turbine Bypass Valve emergency 21(2) loss of Coolant faulted 22A High Pressure Injection Test normal 228 Core Flood Check Valve Test normal 23 Steam Generator Filling, Draining, Flushing and Cleaning normal 24 RCP Restart with Voids in the RCS emergency 25 Refill of Hot, Dry, Depressurized SG upset 26 Pressurizer Bypass Spray Interruption/Restoration test 27 Natural Circulation Cooldown Note {1): Transients that clo not contribute to the growth of the postulated flaw are not considered.
Note {2): part of Reactor Trip group.
The calculation of final crack size follows methodology of Section A,5200, Reference [2].
Crack growth rates are based on Section A-4300 of Reference [2].
- Main f eed Water, RC{S)
- Reactor Coolant (System), RCP
- Reactor Coolant Pump SG
- Steam Generator Page 7
1CAN122301 Page 8 of 10 Con rolled Document framatome Document No. 86-9352400-000 Equivalent Fatigue Crack Growth for Arkansas Nuclear One Unit 1 Beltline Shell Location 4.0 RESULTS 4.1 Transient Fatigue Crack Growth Summary The initial (3.7% of the shell wall thickness including cladding) and final (for 20 years of operation) flaw depths, considering all the design transients listed under Table 3-1 are listed in Table 4-1. The initial flaw size bounds the as-found flaw from the latest in-service inspection (ISI) for AN0-1. The analysis considers a flaw length over depth aspect ratio Ila = 10 and the flaw aspect ratio is maintained during the entire fatigue crack growih.
Table 4-1 : Transient Fatigue Crack Growth Summary 4.2 Equivalent Crack Growth time (year) 0 20 a
[inch]
0.3191 0.3272 The result of the calculation shown in this Section is an acceptance criterion used to compare with the transient fatigue crack growth calculated in Section 4.1. The final crack size is calculated for 20 years of operation with 12 equivalent heat-up / cool-down cycles per one year.
Table 4-2: Equivalent Fatigue Crack Growth time
[year]
cycles 0
0 a
[inch]
0.3191 20 240 0.3273 4.3 Transient Crack Growth Contribution Ratio This section presents results from the detailed transient fatigue crack growth calculation summarized in Table 4-1 for 20 years of operation. The lines in Table 4-3 are sorted by their relative contribution to total fatigue crack growth.
Table 4-3 lists the order of the transients from the highest to the lowest contributors. The highest contributor to crack propagation is from the power change transient, followed by power loading transient, and followed by heat-up / cool-down transient, which together contribute to nearly 97% of the total crack growth and this fatigue crack growth occurs in water environment.
Remaining transients are considered negligible with a total contribution of approximately 3% of the total crack growth.
Page 8
1CAN122301 Page 9 of 10 Control ed Document framatome Document No. 86-9352400-000 Equivalent Fatigue Crack Growth for Arkansas Nuclear One Unit 1 Beltline Shell Location Table 4-3: Crack Growth Relative Contribution S.O CONCLUSION transient name power change power loading heat-up/ cool-down rod withdrawal flow change power unloading trip step load increase/ decrease feedwater loss 5.1 Fatigue Crack Growth participation
% ratio
~97%
The equivalent fatigue crack growth (Table 4-2) using 12 equivalent heat-up / cool-down cycles per year is larger than the fatigue crack growth from the detailed design transients (fable 4-1 ).
Therefore, the requirements outlined in Item 3 of Appendix B of WCAP-1 6168-P-A, Refrrence [1], for Arkansas Nuclear One Unit 1 are met.
Furthermore, power change, power loading and heat-up / cool-down transients were idmtified as major contributors to the fatigue crack growth.
This fulfills the intent of Item 3, Appe-ndix B of WCAP-16168-P-A, Refrrmce [l] for Arkansas uclear One Unit 1.
5.2 RT MAX Temperature The maximum RT~IAX-FO value for ANO-1 forgmgs = 52.3°F (512.0°R). This folfills the intent of Item 4, Appendix B ofWCAP-16168-NP-A, Reference[!] for Arkansas Nuclear One Unit 1.
Page 9
1CAN122301 Page 10 of 10 framatome Controlled Document Document No. 86-9352400-000 Equivalent Fatigue Crack Growth for Arkansas Nuclear One Unit 1 Beltline Shell Location
6.0 REFERENCES
- 1.
Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval, WCAP-16168-NP-A, Revision 3, October 201 1, Westinghouse Electric Company LLC
- 2.
ASME Boiler and Pressure Vessel Code,Section XI, Rules for Inservice Inspection ofNuclear Power Plant Components, 2013 Edition Page 10
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