ML23108A136
| ML23108A136 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 04/24/2023 |
| From: | Jennifer Dixon-Herrity Plant Licensing Branch IV |
| To: | Entergy Operations |
| Wengert T | |
| References | |
| EPID L-2023-LLR-0000 | |
| Download: ML23108A136 (1) | |
Text
April 24, 2023 ARKANSAS NUCLEAR ONE, UNIT 2 - APPROVAL OF REQUEST FOR HALF-NOZZLE REPAIR OF REACTOR VESSEL CLOSURE HEAD PENETRATION NO. 46 (EPID L-2023-LLR-0000)
LICENSEE INFORMATION Licensee:
Entergy Operations, Inc.
Licensee Address:
ANO Site Vice President Arkansas Nuclear One Entergy Operations, Inc.
N-TSB-58 1448 S.R. 333 Russellville, AR 72802 Plant Name and Unit:
Arkansas Nuclear One, Unit 2 (ANO-2)
Docket No.:
50-368 APPLICATION INFORMATION Submittal Date: January 20, 2023 Submittal Agencywide Documents Access and Management System (ADAMS) Accession No.: ML23020A940 (package)
Supplement Dates: March 29, 2023 (2 letters)
Supplement ADAMS Accession Nos.: ML23088A198 and ML23088A214 (package)
Licensee Proposed Alternative No. or Identifier: Relief Request ANO2-RR-23-001 Applicable Provision: Title 10 of the Code of Federal Regulations (10 CFR),
paragraph 50.55a(z)(1), Acceptable level of quality and safety Applicable Code Requirements: American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, paragraph IWB-3420 and subparagraph IWB-3132.3; ASME Code,Section III, paragraph NB-5245 and subparagraph NB-5330(b); ASME Code Case N-638-7, Similar and Dissimilar Metal Welding Using Ambient Temperature Machine GTAW [Gas Tungsten Arc Welding] Temper Bead Technique,Section XI, Division 1, paragraphs 1(g) and 2(b).
Applicable Code Edition and Addenda: 2007 Edition through 2008 Addenda of ASME Code Section XI; 1992 Edition of ASME Code,Section III Brief Description of the Proposed Alternative and Basis:
As discussed in a previous proposed alternative dated November 5, 2021 (ML21309A007),
Entergy Operations, Inc. (the licensee) performed ultrasonic testing (UT) of (ANO-2 reactor vessel closure head (RVCH) penetration nozzles in Refueling Outage 2R28 (fall 2021) in accordance with Item No. B4.20 of ASME Code Case N-729-6, Alternative Examination Requirements for PWR [Pressurized-Water Reactor] Reactor Vessel Upper Heads with Nozzles Having Pressure-Retaining Partial-Penetration Welds,Section XI, Division 1. During the examinations, the licensee detected an axial, planar indication in control element drive mechanism (CEDM) Nozzle No. 46.
To perform a repair of the penetration nozzle, the licensee submitted the previous proposed alternative dated November 5, 2021, which contained alternative approaches to certain requirements of the ASME Code. On November 9, 2021 (ML21314A121), the NRC staff verbally authorized the use of the previous proposed alternative regarding inside diameter temper bead (IDTB) welding repair for one operating cycle (Operating Cycle 29). The authorization of the previous proposed alternative is documented in the U.S. Nuclear Regulatory Commission (NRC) staffs safety evaluation dated April 7, 2022 (ML22073A095). The proposed repair was performed during Refueling Outage 2R28. In the previous proposed alternative dated November 5, 2021, the licensee also committed to submit a revised request to extend the design life of the repaired nozzle.
Accordingly, the licensee submitted the subject proposed alternative dated January 20, 2023, to extend the design life of the repaired nozzle until the end of the current license term that expires on July 17, 2038 (i.e., after 60 years of plant operation). The proposed alternative includes fracture mechanics evaluations on the triple point anomaly of the repair weld and the as-left flaw in the original J-groove weld to demonstrate that the postulated crack growth from 2021 to 2038 (17-year evaluation period from the repair to the end of the current license term) does not affect the integrity of the repaired nozzle.
The licensees submittal includes linear elastic fracture mechanics (LEFM) analyses on the triple point anomaly. The licensee also performed the LEFM and elastic plastic fracture mechanics (EPFM) analyses on the as-left flaw in the original J-groove weld.
REGULATORY EVALUATION The NRC regulations in 10 CFR 50.55a(z), Alternatives to codes and standards requirements, state that [a]lternatives to the requirements of paragraphs (b) through (h) of [10 CFR 50.55a] or portions thereof may be used when authorized by the Director, Office of Nuclear Reactor Regulation. A proposed alternative must be submitted and authorized prior to implementation.
The licensee must demonstrate that its request meets one of two criteria: (1) the proposed alternative would provide an acceptable level of quality and safety in accordance with paragraph (z)(1); or (2) compliance with the specified requirements of this section would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety in accordance with paragraph (z)(2).
The licensee submitted the request on the basis that the proposed alternative would provide an acceptable level of quality and safety in accordance with 10 CFR 50.55a(z)(1). Based on the above regulations, the NRC staff finds that regulatory authority exists to authorize an alternative to ASME Code,Section III and Section XI, as requested by the licensee.
TECHNICAL EVALUATION The April 7, 2022, safety evaluation documented the following topics regarding the nozzle repair and its acceptability: (1) repair welding approaches, including the use of ASME Code Case N-638-7, use of peening on the final layer of the repair weld, and simulation of post-weld heat treatment in the welding process qualification; (2) acceptance examination of the weld modification; (3) preservice and inservice inspections; and (4) corrosion evaluation.
The current proposed alternative for the remainder of the current license term includes the repair welding approaches, acceptance examination of the weld modification, preservice and inservice inspections and corrosion evaluation, consistent with those of the previous proposed alternative dated November 5, 2021, that was subsequently authorized by the NRC staff.
1.0 REPAIR WELD TRIPLE POINT ANOMALY ANALYSIS ASME Code, Section Ill, NB-5330(b) states that indications characterized as cracks, lack of fusion, or incomplete penetration are unacceptable regardless of length. With respect to this design code requirement, the proposed alternative addressed the repair weld triple point anomaly. The licensee stated that the triple point anomaly is an artifact of ambient IDTB welding at the triple point. As illustrated in figure 2 of the proposed alternative, there are two triple points in the proposed weld repair. The upper triple point is the point in the repair weld where the low alloy steel RVCH base material, the Alloy 600 nozzle, and the Alloy 52M repair weld intersect.
The lower triple point is the point in the repair weld where the low alloy steel RVCH base material, the Alloy 690 replacement nozzle, and the Alloy 52M repair weld intersect.
The licensee explained that the anomaly consists of an irregularly shaped very small void and that mockup testing verified that the anomalies are common and do not exceed 0.10 inches in through-wall extent. The NRC staff finds that the locations of the triple point anomaly evaluation and the associated flaw sizes were adequately determined based on the weld configuration and mockup testing results.
The licensee performed LEFM and limit load analyses for the repaired nozzle considering the flaws at the weld anomaly triple point locations. In the analyses, the crack growth was also postulated for 17 years of operation between 2021 and 2038 (i.e., after 60 years of plant operation) to address the life-of-repair analyses.
The licensees analyses postulated circumferential and axial flaws at the outside surface of the nozzle repair weld. A 360-degree continuous circumferential flaw, lying in a horizontal plane, was considered to be a conservative representation of crack-like defects that would exist in the repair weld triple point anomaly. The circumferential flaw was subjected to axial stresses in the nozzle. An axially oriented semi-circular outside surface flaw was also considered since it would lie in a plane normal to the higher circumferential stresses. Both of these flaws were postulated to propagate across the nozzle wall thickness from the outside diameter to the inside diameter of the nozzle for the upper and lower triple points.
The licensees analyses also postulated that the flaw growth extended down from the upper triple point or up from the lower triple point along the outside surface of the repair weld between the upper and lower triple points. Specifically, a cylindrically oriented flaw was postulated to lie along this interface, subjected to radial stresses with respect to the nozzle.
The NRC staff finds that the postulations of the circumferential, axial and cylindrical flaws and their growth paths are comprehensive and reasonable based on potential crack initiations at the triple anomaly points and their orientations with respect to the nozzle and repair weld configurations.
In the crack growth analyses, the licensee calculated the fatigue crack growth in RVCH low alloy steel material in air or reactor water environments, as applicable, in accordance with Article A-4300 of ASME Code,Section XI, 2007 Edition through 2008 Addenda. Specifically, the embedded weld anomaly at the upper triple point used crack growth rates in air. The embedded weld anomaly at the lower triple point used the crack growth rates for material exposed to light-water reactor environments, if those rates are bounding for the crack growth rates for the air environment. If the fatigue crack growth rate in reactor water environments is lower than that in air, the licensee used the fatigue crack growth rate in the air environment in accordance with ASME Code Section XI, A-4300(b)(2).
For the Alloy 600/690 nozzle and Alloy 52M weld materials either in air (for the upper triple point) or exposed to reactor water environments (for the lower triple point), the licensee used a fatigue crack growth rate twice that of Alloy 600 material to bound the Alloy 600/690 nozzle and Alloy 52M weld materials. This approach is based on the fatigue crack growth data in NUREG/CR-6907, Crack Growth Rates of Nickel Alloy Welds in a PWR Environment (ML061720302), indicating that the fatigue crack growth rates in nickel alloy welds are higher than the growth rate for Alloy 600 by a factor of 2.
The NRC staff finds that the fatigue crack growth calculations are acceptable because the calculations accounted for the environmental effects of the reactor coolant on the fatigue crack growth rates for the fabrication materials of the repaired nozzle in accordance with the provisions in ASME Code,Section XI, nonmandatory appendix A and the fatigue crack growth data in NUREG/CR-6907.
In addition, the licensees analyses included the crack growth due to primary water stress corrosion cracking (PWSCC) for the lower triple point of the repair weld, as it is subject to a primary water environment. The NRC staff finds that the postulation of PWSCC initiation and crack growth for the lower triple point is a conservative approach because: (1) the exposure of the lower triple point to reactor coolant is limited due to the tight gap between the nozzle outer surface and the RVCH bore surface below the lower triple point, and (2) the environmental effects due to the access of reactor coolant to the lower triple point are expected to further decrease as corrosion products fill the tight gap during the operation.
Following the crack growth calculations, the licensee used the acceptance criteria of the ASME Code to determine the acceptability of the final end-of-life flaw sizes. For the cylindrical flaw in the low alloy steel RVCH, the licensee used the acceptance criteria in ASME Code,Section XI, IWB-3613. Specifically, IWB-3613 addresses the acceptance criteria and the associated flaw evaluations for shell regions near structural discontinuities. In the LEFM-based flaw evaluations, the end-of-life flaw is acceptable if the applied stress intensity factor for the end-of-life flaw dimensions meets the acceptance criteria in IWB-3613.
For the low alloy steel material of the RVCH, the licensee also indicated that the maximum metal reference nil-ductility temperature (RTNDT) is 10 degrees Fahrenheit (°F), as documented in table 3-2 of Westinghouse Report WCAP-18169-NP, Revision 1, Arkansas Nuclear One Unit 2 Heatup and Cooldown Limit Curves for Normal Operation, June 2018 (ML18215A178).
The fracture toughness of the low alloy steel material was determined based on the operating temperature and RTNDT value in accordance with figure A-4200-1 of ASME Code,Section XI, 2007 Edition through 2008 Addenda.
In relation to the triple point anomaly analysis, the licensee further identified a minimum fluid temperature for certain transients as an analysis limitation to ensure a sufficient fracture toughness for the low alloy steel material. In one of the supplements dated March 29, 2023 (ML23088A198), the licensee indicated that, even though the hydrostatic and leak test transients related to the analysis limitations were included in the triple point anomaly and as-left J-groove weld flaw analyses, these test transients are not required to be considered for fatigue analysis.
In relation to this topic, the NRC staff noted that ANO-2 technical specification (TS) Limiting Condition for Operation (LCO) 3.4.9.1 and TS figure 3.4-2C specify the pressure-temperature limits for the hydrostatic and leak tests (ML053130317). In addition, Westinghouse Report WCAP-18169-NP, Revision 1, describes the technical bases of the pressure-temperature limits.
These pressure-temperature limits were approved in ANO-2 License Amendment No. 311, dated November 27, 2018 (ML18298A012).
Specifically, table 8-3 in WCAP-18169-NP, Revision 1, indicates that the minimum test temperature at a test pressure of 2083 pounds per square inch - gauge (psig) is 165°F for these test transients. The table also indicates that a test pressure higher than 2083 psig requires a test temperature higher than 165°F. Considering that the test pressure for the hydrostatic and leak tests is higher than 2083 psig at ANO-2, the NRC staff finds that the test temperature for these tests is higher than 165°F. Therefore, the NRC staff finds that the pressure-temperature limits in the TSs ensure that the analysis limitation related to the triple anomaly analysis and minimum fluid temperature is met.
For the axial and circumferential flaws in the Alloy 52M repair weld, the licensee used the acceptance criteria in ASME Code,Section XI, IWB-3642. Specifically, IWB-3642 states that piping containing flaws exceeding the acceptance standards of IWB-3514.1 may be evaluated using analytical procedures described in ASME Code,Section XI, nonmandatory appendix C and is acceptable for continued service during the evaluated time period when the critical flaw parameters satisfy the criteria in nonmandatory appendix C.
In accordance with nonmandatory appendix C of ASME Code,Section XI, C-4230, the licensee relied on the flaw evaluation procedures of C-4210 for flaws in the nickel-chromium-iron (Ni-Cr-Fe) (austenitic) weld metal. Based on figure C-4210-1 of nonmandatory appendix C, the licensee used C-5000 for the postulated flaws in Ni-Cr-Fe weld material that uses non-flux welds. Specifically, the licensee performed limit load evaluations for the axial and circumferential flaws in accordance with C-5400 and C-5300, respectively.
The NRC staff finds that the licensees acceptance criteria used in the triple point anomaly and crack growth analyses are acceptable because the licensee used the relevant flaw evaluation provisions and acceptance criteria of the ASME Code for the end-of-life flaws in the low alloy steel RVHC and Alloy 52M repair weld. The NRC staff also finds that the licensees analyses demonstrated that the postulated flaw growth meets the acceptance criteria of the ASME Code.
As discussed above, the NRC staff finds that the licensees justification for the triple point anomalies is acceptable for the remainder of the current operating license because of the following: (1) the locations of the triple point anomaly evaluation and the associated flaw sizes were adequately determined based on the weld configuration and mockup testing results; (2) the fatigue crack growth analyses accounted for the environmental effects of the reactor coolant on the crack growth rates in accordance with the provisions in ASME Code,Section XI, nonmandatory appendix A, and the crack growth data in NUREG/CR-6907; (3) the time period for the postulated crack growth properly covered 17 years of operation since the nozzle repair, between 2021 and 2038; (4) the analyses further included the potential crack growth due to PWSCC in the repair weld for the lower triple point, which is a conservative approach given that the access of reactor coolant to the lower triple point is limited due to the tight gap between the nozzle outer surface and RVCH bore surface and corrosion products filling the tight gap during operation; and (5) the licensee demonstrated that the postulated flaw growth meets the acceptance criteria of the ASME Code.
2.0 AS-LEFT J-GROOVE WELD FLAW EVALUATION The licensee addressed the analytical evaluation of the as-left J-groove weld flaw. The provisions of IWB-3600 of ASME Code,Section XI for analytical flaw evaluation assume that cracks are fully characterized in accordance with IWB-3420. The licensee indicated that there are no qualified UT examination techniques available for examining the original nozzle-to-RVCH J-groove welds. Due to the impracticality to characterize the flaw geometry that may exist in the J-groove weld, the licensee conservatively assumed that the as-left condition of the J-groove weld may involve the extension of the flaw through the entire Alloy 82/182 J-groove weld including the weld buttering.
The licensee explained that the preferential direction for cracking would be radial as a result of the dominant hoop stresses in the J-groove weld. Accordingly, the licensee assumed that an radial-axial crack in the J-groove weld would propagate into the low alloy steel by fatigue crack growth under cyclic loading because the low alloy steel is resistant to PWSCC in reactor coolant environments.
The NRC staff finds that the licensees approach and assumptions for the as-left J-groove weld flaw evaluation are acceptable because of the following: (1) since no UT examination technique is currently available to characterize the flaw size in the J-groove weld, the licensee conservatively postulated that the initial flaw extends throughout the entire J-groove weld; (2) the postulation of the radial-axial crack growth direction is reasonable because the hoop stress is dominant near the nozzle and the radial-axial crack orientation is consistent with the licensees nondestructive examination results; and (3) the low alloy steel RVCH is resistant to PWSCC in reactor coolant environments such that fatigue crack growth is the limiting crack growth mechanism for the RVCH in the flaw evaluation.
The fatigue crack growth analyses also considered the conservative initial flaws and the associated crack growth. In the analysis, the fatigue crack growth rates were calculated in accordance with ASME Code,Section XI, nonmandatory appendix A, A-4300. The licensee also considered the environmental effects on the crack growth rates in accordance with the provisions of A-4300. The initial flaw was assumed to grow for 17 years of operation between 2021 and 2038 (i.e., 60 years of operation) to address the life-of-repair analyses.
The NRC staff finds that the licensees approach for the calculation of fatigue crack growth rates is acceptable because: (1) the approach is consistent with the provisions of the ASME Code for
estimating the fatigue crack growth rates in the low alloy steel material, (2) the ASME Code provisions accounted for the environmental effects of the reactor coolant on the fatigue crack growth rates, and (3) the postulated crack growth properly covered 17 years of operation since the repair.
After crack growth was calculated, the licensee evaluated the postulated flaws in accordance with guidance in ASME Code Case N-749, Alternative Acceptance Criteria for Flaws in Ferritic Steel Components Operating in the Upper Shelf Temperature Range,Section XI, Division 1, as conditioned by Regulatory Guide (RG) 1.147, Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1, Revision 20 (ML21181A222). ASME Code Case N-749 is an alternative approach to the flaw evaluation provisions in the ASME Code,Section XI.
In relation to the as-left J-groove weld flaw evaluation, the licensee identified a minimum fluid temperature for hydrostatic and leak test transients as an analysis limitation to ensure the EPFM can be applied to the fracture mechanics evaluations for the transients. As previously discussed, ANO-2 TS LCO 3.4.9.1 and TS figure 3.4-2C specify the pressure-temperature limits for the hydrostatic and leak tests. In addition, Westinghouse Report WCAP-18169-NP, Revision 1 provides the technical bases for these limits.
Specifically, table 8-3 in WCAP-18169-NP, Revision 1 indicates that the minimum test temperature at a test pressure of 2083 psig is 165°F for these test transients. The table also indicates that a test pressure higher than 2083 psig requires a test temperature higher than 165°F. Considering that the test pressure for the hydrostatic and leak tests is higher than 2083 psig at ANO-2, the NRC staff finds that the test temperature for these tests is higher than 165°F.
Accordingly, the NRC staff finds that the pressure-temperature limits in the TSs ensure the analysis limitation related to the as-left flaw evaluation and minimum fluid temperature is met.
The licensee determined which fracture mechanics approach (i.e., LEFM or EPFM approach) should be used for the acceptance of postulated end-of-life flaws depending on the limiting load cases of the applicable transients in accordance with the guidance in ASME Code Case N-749, as conditioned in RG 1.147, Revision 20. For the LEFM analysis, the licensee used the acceptance criteria described in ASME Code,Section XI, IWB-3613. For the EPFM analysis, the licensee used the acceptance criteria specified in ASME Code Case N-749, sections 3.1 and 3.2, which address the acceptance criteria based on limited ductile crack extension and the acceptance criteria based on limited ductile crack extension and stability, respectively.
ASME Code Case N-749 states, in part, that the end-of-evaluation-period flaw is acceptable for continued operation if the acceptance criteria of section 3.1 are satisfied. If the acceptance criteria of section 3.1 on limited ductile crack extension are not satisfied, the acceptance criteria of section 3.2 on crack extension and flaw stability are required to be met.
The NRC staff finds that the licensees acceptance criteria and the associated flaw evaluations are acceptable because they are consistent with the ASME Code and Code Case provisions as approved by the NRC staff.
As discussed above, the NRC staff finds that the as-left J-groove weld flaw evaluation is acceptable because of the following: (1) the postulation of the axial-radial crack growth direction is reasonable based on the hoop stress, which is dominant near the repaired nozzle, and the UT examination results for the flaw; (2) the fatigue crack growth rates were calculated in accordance with ASME Code,Section XI, nonmandatory appendix A, A-4300; (3) the fatigue crack growth analysis accounted for the environmental effects of the reactor coolant on the
crack growth rates; (4) the time period for the postulated crack growth properly covered 17 years of operation since the repair, between 2021 and 2038; (5) the licensee determined the relevant fracture mechanics approach (i.e., LEFM or EPFM approach) for the flaw evaluation in accordance with ASME Code Case N-749, as conditioned in RG 1.147, Revision 20; (6) the acceptance criteria are consistent with ASME Code,Section XI, IWB-3613 for LEFM analysis and ASME Code Case N-749, as conditioned in RG 1.147, Revision 20, for EPFM analysis; and (7) the licensee demonstrated that the postulated flaw growth meets the acceptance criteria of the ASME Code and Code Case provisions with the conditions mandated by the NRC.
In addition, the NRC staff finds that the repair welding approaches, acceptance examination, preservice and inservice inspections, and the corrosion evaluation described in this proposed alternative remain acceptable to the NRC staff for the remainder of the current license term that expires on July 17, 2038, consistent with the previous verbal authorization dated November 9, 2021, and the NRC staff safety evaluation dated April 7, 2022.
CONCLUSION As set forth above, the NRC staff determines that the proposed alternative, as described in the licensees letter dated January 20, 2023, as supplemented by two letters dated March 29, 2023, for the use of the repaired ANO-2 RVCH penetration nozzle No. 46 is acceptable on the basis that the proposed alternative provides an acceptable level of quality and safety. The NRC staff finds that the proposed alternative will provide reasonable assurance of the structural integrity and leak tightness of the nozzle for the remainder of the current license term. Accordingly, the NRC staff concludes that the licensee has adequately addressed the regulatory requirements in 10 CFR 50.55a(z)(1). Therefore, the NRC staff authorizes the use of the proposed alternative until the end of the current license term at ANO-2 (July 17, 2038).
All other requirements in ASME Code,Section III and Section XI for which relief or an alternative was not specifically requested and approved as part of this subject request remain applicable, including third-party review by the Authorized Nuclear Inservice Inspector.
Principal Contributor: S. Min Date: April 24, 2023 Jennifer L. Dixon-Herrity, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation cc: Listserv Jennifer L.
Dixon-Herrity Digitally signed by Jennifer L. Dixon-Herrity Date: 2023.04.24 14:45:30 -04'00'
- concurrence via email OFFICE NRR/DORL/LPL4/PM* NRR/DORL/LPL4/LA* NRR/DNRL/NPHB/BC* NRR/DORL/LPL4/BC*
NAME TWengert PBlechman MMitchell JDixon-Herrity DATE 4/19/2023 4/18/2023 4/6/2023 4/24/2023