W3F1-2024-0009, Proposed Alternative WF3-RR-24-02 for Examinations of Steam Generator Pressure-Retaining Welds and Full Penetration Welded Nozzles
| ML24078A376 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 03/18/2024 |
| From: | Couture P Entergy Operations |
| To: | Office of Nuclear Reactor Regulation, Document Control Desk |
| References | |
| W3F1-2024-0009 | |
| Download: ML24078A376 (1) | |
Text
Phil Couture Senior Manager Fleet Regulatory Assurance - Licensing 601-368-5102
March 18, 2024
ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001
Subject:
Proposed Alternative WF3-RR-24-02 for Examinations of Steam Generator Pressure-Retaining Welds and Full Penetration Welded Nozzles
Waterford Steam Electric Station, Unit 3 NRC Docket No. 50-382 Renewed Facility Operating License No. NPF-38
In accordance with 10 CFR 50.55a(z)(1), Entergy Operations, Inc. (Entergy) hereby requests Nuclear Regulatory Commission (NRC) approval of a proposed alternative for Waterford Steam Electric Station, Unit 3 (WF3). Specifically, the proposed alternative concerns American Society of Mechanical Engineers (ASME) Class 1, Examination Category B-B, Pressure-Retaining Welds in Vessels Other Than Reactor Vessels, Item Number B2.40, Steam Generators (Primary Side), Tubesheet-to-Head Weld, Class 2, Examination Category C-A, Pressure-Retaining Welds in Pressure Vessels, Item Number C1.20, Head Circumferential Welds, Class 2, Examination Category C-A, Item Number C1.30, Tubesheet-to-Shell Welds, Class 2, Examination Category C-B, (Pressure Retaining Nozzle Welds in Pressure Vessels,Section XI, Division 1), Item Number C2.21, Nozzle-to-Shell (Nozzle-to-Head or Nozzle-to-Nozzle) Welds, and Class 2, Examination Category C-B, Item Number C2.22, Nozzle Inside Radius Sections.
The proposed alternative is to defer the ISI examinations for select Item Nos. for the Steam Generators (SG) at WF3 from the current American Society of Mechanical Engineers (ASME)
Code,Section XI, Division 1 10-year requirement to the end of currently licensed operating life, which is scheduled to end on December 18, 2044. This equates to an extension of 27 years, 18 days from the end of the third inservice inspection interval (November 30, 2017) at which time all ASME Code,Section XI, Division 1 requirements were satisfied. Entergy requests authorization to use the proposed alternative pursuant to 10 CFR 50.55a(z)(1) on the basis that the alternative provides an acceptable level of quality and safety.
The Enclosure to this letter provides the Alternative Request WF3-RR-24-02 with Enclosure, Attachment 1 providing the WF3 Plant-Specific Applicability and Enclosure, Attachment 2 providing the Results of Industry Survey.
Entergy Operations, Inc., 1340 Echelon Parkway, Jackson, MS 39213 W3F1-2024-0009 Page 2 of 2
Entergy requests approval of the proposed Alternative Request by February 28, 2025, to support the WF3 Refueling Outage (RF26). The proposed changes would be implemented during RF26.
This letter contains no new regulatory commitments.
Should you have any questions or require additional information, please contact me at 601-368-5102.
Respectfully,
Phil Couture
PC/chm
Enclosure:
Alternative Request WF3-RR-24-02
Attachments to
Enclosure:
- 1. Plant-Specific Applicability for WF3
- 2. Results of Industry Survey
cc: NRC Region IV Regional Administrator NRC Senior Resident Inspector - WF3 NRC Project Manager - WF3
Enclosure
Alternative Request WF3-RR-24-02
W3F1-2024-0009 Enclosure Page 1 of 18
TABLE OF CONTENTS
1.0 ASME CODE COMPONENTS AFFECTED...................................................................... 2 2.0 APPLICABLE CODE EDITION AND ADDENDA.............................................................. 3 3.0 APPLICABLE CODE REQUIREMENT............................................................................. 3 4.0 REASON FOR REQUEST................................................................................................ 4 5.0 PROPOSED ALTERNATIVE AND BASIS FOR USE....................................................... 4 5.1 Technical Basis.............................................................................................................. 5 5.1.1 Applicability of the Degradation Mechanism Evaluation in References [9.1] and
[9.2] to WF3............................................................................................................ 5 5.1.2 Applicability of the Stress Analysis in References [9.1] and [9.2] to WF3............... 5 5.1.3 Applicability of the Flaw Tolerance Evaluation in References [9.1] and [9.2] to WF3
................................................................................................................................ 7 5.1.4 Limiting PSI/ISI Scenario......................................................................................... 8 5.1.5 FW Nozzle Inside Radius Section........................................................................... 8 5.1.6 FW Nozzle-to-Shell Weld........................................................................................ 9 5.1.7 Remainder of WF3 SG Welds............................................................................... 10 5.1.8 Inspection History.................................................................................................. 12 5.1.9 Industry Survey..................................................................................................... 12 5.1.10 Performance Monitoring........................................................................................ 12 5.1.11 Conclusion............................................................................................................. 12 6.0 DURATION OF PROPOSED ALTERNATIVE................................................................. 13 7.0 PRECEDENTS................................................................................................................ 13 7.1 Vogtle Units 1 and 2..................................................................................................... 13 7.2 Millstone Unit 2............................................................................................................. 13 7.3 Other Approved Actions Related to Inspections of SG Welds and Components......... 14 7.4 Similar Topical Reports for Relief for Class 1 Nozzles................................................. 14 8.0 ACRONYMS................................................................................................................... 15
9.0 REFERENCES
................................................................................................................ 15 10.0 ATTACHMENTS............................................................................................................. 18
W3F1-2024-0009 Enclosure Page 2 of 18
ALTERNATIVE REQUEST WF3-RR-24-02
1.0 ASME CODE COMPONENTS AFFECTED American Society of Mechanical Engineers (ASME)
Code Class: Class 1 and Class 2
Description:
Steam generator (SG) pressure-retaining welds and full penetration welded nozzles (nozzle-to-shell welds and inside radius sections)
Examination Category: Class 1, Category B-B, pressure-retaining welds in vessels other than reactor vessels Class 2, Category C-A, pressure-retaining welds in pressure vessels Class 2, Category C-B (Pressure Retaining Nozzle Welds in Pressure Vessels,Section XI, Division 1)
Item Numbers: B2.40 - Steam generators (primary side), tubesheet-to-head weld C1.20 - Head circumferential welds C1.30 - Tubesheet-to-shell welds C2.21 - Nozzle-to-shell (nozzle-to-head or nozzle-to-nozzle) welds C2.22 - Nozzle inside radius sections WaterfordUnit3 ASMECategory ASMEItemNo. ComponentID ComponentDescription B-B B2.40 03-076 Steam Generator (Primary Side)
Tubesheet-to-Head Weld B-B B2.40 04-076 Steam Generator (Primary Side)
Tubesheet-to-Head Weld C-A C1.20 03-068 Pressure Vessels Head Circumferential Welds C-A C1.20 04-068 Pressure Vessels Head Circumferential Welds C-A C1.30 03-075 Pressure Vessels Tubesheet-to-Shell Weld C-A C1.30 04-075 Pressure Vessels Tubesheet-to-Shell Weld Nozzles Without Reinforcing Plate in C-B C2.21 03-073 Vessels > 1/2 in. (13 mm) Nominal Thickness Nozzle-to-Shell (Nozzle to Head or Nozzle to Nozzle) Weld W3F1-2024-0009 Enclosure Page 3 of 18
WaterfordUnit3 ASMECategory ASMEItemNo. ComponentID ComponentDescription Nozzles Without Reinforcing Plate in C-B C2.21 04-073 Vessels > 1/2 in. (13 mm) Nominal Thickness Nozzle-to-Shell (Nozzle to Head or Nozzle to Nozzle) Weld Nozzles Without Reinforcing Plate in C-B C2.22 03-074 Vessels > 1/2 in. (13 mm) Nominal Thickness Nozzle Inside Radius Section Nozzles Without Reinforcing Plate in C-B C2.22 04-074 Vessels > 1/2 in. (13 mm) Nominal Thickness Nozzle Inside Radius Section
2.0 APPLICABLE CODE EDITION AND ADDENDA The fourth 10-year inservice inspection (ISI) interval Code of record for WF3 is the 2007 Edition of American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code,Section XI through the 2008 Addenda, "Rules for Inservice Inspection of Nuclear Power Plant Components."
3.0 APPLICABLE CODE REQUIREMENT ASME Section XI IWB-2500(a), Table IWB-2500-1, examination Category B-B and IWC-2500(a), Table IWC-2500-1, Examination Categories C-A and C-B require examination of the following Item Nos.:
Item No. B2.40 - Volumetric examination of essentially 100% of the weld length of all welds during the first Section XI inspection interval. For successive inspection intervals the examination may be limited to one vessel among the group of vessels performing a similar function. The examination volume is shown in Figure IWB-2500-6.
Item No. C1.20 - Volumetric examination of essentially 100% of the weld length of the head-to-shell weld during each Section XI inspection interval. In the case of multiple vessels of similar design, size, and service (such as steam generators, heat exchangers), the required examinations may be limited to one vessel or distributed among the vessels. The examination volume is shown in Figure IWC--2500-1.
Item No. C1.30 - Volumetric examination of essentially 100% of the weld length of the tubesheet-to-shell welds during each Section XI inspection interval. In the case of multiple vessels of similar design, size, and service (such as steam generators, heat exchangers), the required examinations may be limited to one vessel or distributed among the vessels. The examination volume is shown in Figure IWC-2500-2.
Item No. C2.21 - Volumetric and surface examination of all nozzle welds at terminal ends of piping runs during each Section XI inspection interval. In the case of multiple vessels of similar design, size, and service (such as
W3F1-2024-0009 Enclosure Page 4 of 18
steam generators, heat exchangers), the required examinations may be limited to one vessel or distributed among the vessels. The examination area and volume are shown in Figures IWC-2500-4(a),
(b), or (d).
Item No. C2.22 - Volumetric examination of all nozzle inside radius sections at terminal ends of piping runs during each Section XI inspection interval. In the case of multiple vessels of similar design, size, and service (such as steam generators, heat exchangers), the required examinations may be limited to one vessel or distributed among the vessels. The examination volume is shown in Figures IWC-2500-4(a), (b), or (d).
4.0 REASON FOR REQUEST The Electric Power Research Institute (EPRI) performed assessments in References
[9.1] and [9.2] of the bases for the ASME Code,Section XI examination requirements specified for the above listed ASME Code,Section XI, Division 1 examination categories for steam generator (SG) welds and components. The assessments include a survey of inspection results from 74 domestic and international nuclear units and flaw tolerance evaluations using probabilistic fracture mechanics (PFM) and deterministic fracture mechanics (DFM). The Reference [9.1] and [9.2] reports concluded that the current ASME Code,Section XI ISI examinations can be deferred for some time with no impact to plant safety. Based on the conclusions of the two EPRI reports supplemented by plant-specific evaluations contained herein, Entergy is requesting an ISI examination deferral for the subject welds. The Reference [9.1] and [9.2] reports were developed consistent with the recommendations provided in EPRIs White Paper on suggested content for PFM submittals [9.12] and NRC Regulatory Guide 1.245 for PFM submittals and associated technical basis [9.13, 9.14].
5.0 PROPOSED ALTERNATIVE AND BASIS FOR USE For WF3, Entergy is requesting an inspection alternative to the examination requirements of ASME Code,Section XI, Tables IWB-2500-1 and IWC-2500-1, for the following examination categories and item numbers:
ASME Item No. Description Category
B-B B2.40 Steam generators (primary side), tubesheet-to-head weld C-A C1.20 Head circumferential welds C-A C1.30 Tubesheet-to-shell welds C-B C2.21 Nozzle-to-shell (nozzle-to-head or nozzle-to-nozzle) welds C-B C2.22 Nozzle inside radius sections
In 2012 (second period of the third inspection interval) the WF3 SGs were replaced. The new SG welds and components received the required PSI examinations prior to service W3F1-2024-0009 Enclosure Page 5 of 18
followed by ISI examinations through the first period of the current fourth inspection interval.
The proposed alternative is to defer the ISI examinations for these Item Nos. for the SGs at WF3 from the current ASME Code,Section XI, Division 1 10-year requirement to the end of currently licensed operating life, which is scheduled to end on December 18, 2044. This equates to an extension of 27 years, 18 days from the end of the third inservice inspection interval (November 30, 2017) at which time all ASME Code,Section XI, Division 1 requirements were satisfied.
5.1 Technical Basis A summary of the key aspects of the technical basis for this request is summarized below. The applicability of the technical basis to WF3 is shown in Attachment 1.
5.1.1 Applicability of the Degradation Mechanism Evaluation in References [9.1] and [9.2] to WF3 An evaluation of degradation mechanisms that could potentially impact the reliability of the SG welds and components was performed in References [9.1] and [9.2]. The degradation mechanisms that were evaluated included stress corrosion cracking (SCC),
environmental assisted fatigue (EAF), microbi ologically 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 known active degradation mechanisms identified that significantly affect the long-term structural integrity of the SG welds and components covered in this request. This observation was acknowledged by the NRC in Section 3.8, page 6, second paragraph of the Reference [9.16] Safety Evaluation (SE) for Vogtle Units 1 & 2 and Section 2.0, page 3, second paragraph of the Reference [9.18] SE for Millstone Unit 2. As shown in Attachment 1, the materials and operating conditions for the plants considered in this Request for Alternative are similar to those in the References [9.1] and [9.2] and therefore, the conclusions of these Reports apply to the plants in this Request for Alternative. The fatigue-related mechanisms were considered in the PFM and DFM evaluations in References [9.1] and [9.2].
As part of the technical basis in References [9.1] and [9.2], a comprehensive industry survey involving 74 PWR units was conducted to determine the degradation history of these components. The survey reviewed exam ination results from the start of plant operation. Most of these plants have operated for over 30 years and in some cases over 40 years. The results showed that no examinations identified any unknown degradation mechanisms (i.e., mechanisms other than those listed above). Based on this exhaustive industry survey, it is concluded that although the emergence of an unknown degradation mechanism cannot be completely ruled out, the possibility of the occurrence of such an unknown degradation mechanism is highly unlikely.
5.1.2 Applicability of the Stress Analysis in References [9.1] and [9.2] to WF3 Finite element analyses (FEA) were performed in References [9.1] and [9.2] to determine the stresses in the SG welds and components covered in this request. The finite element models used in References [9.1] and [9.2] are consistent with the configuration of WF3 and therefore no new FEA model is required for the stress analysis of these plants. The analysis in References [9.1] and [9.2] was performed using W3F1-2024-0009 Enclosure Page 6 of 18
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 WF3 is demonstrated in Attachment 1 and confirms that all plant-specific requirements are met. In particular, the key geometric parameters used in the Reference [9.1] and [9.2] stress analyses are compared to those of WF3 in Tables 1 and 2:
Table 1 SG Vessel Dimensions Plant Primary Primary Primary Secondary Secondary Secondary Lower Lower Lower Upper Shell Upper Upper Head ID Head Thk Head ID Shell Thk Shell (in) (in) Ri/t (in) (in) Ri/t EPRI Report (Table 4-2 of [9.2]) 155.33 6.94 11.2 230.87 4.91 23.5 WF3 163.62(1) 7.25(1)(2) 11.3 253.12(1) 5.5(1) 23.01
Notes:
- 1. Reference [9.21].
- 2. Calculated using inner and outer radii.
Table 2 SG Nozzle Dimensions Plant FW Nzl ID FW Nzl Thk FW Nzl MS Nzl ID MS Nzl Thk MS Nzl (in) (in) Ri/t (in) (in) Ri/t EPRI Report (Figures 4-9 and 16.5 6 1.38 22.25 4.53 2.46 4-10 of [9.1])
WF3 16.12(1) 9.63(2) 0.84 Note 3 Note 3 Note 3 Notes:
- 1. Reference [9.22].
- 3. N/A as Entergy is not requesting relief from examination of any MS nozzle components.
As discussed in Sections 4.3.3 and 4.6 of Reference [9.1] and noted by the NRC in Section 3.8.3.1, page 9, third paragraph of the SER for Vogtle [9.16], the dominant stress is the pressure stress. Therefore, the variation in the R i/t ratio determined in Tables 1 and 2 can be used to scale up the stresses of the Reference [9.1] and [9.2]
reports to obtain the plant-specific stresses for each unit and component. From Tables 1 and 2, the stress ratio (Ri/t) of WF3 (relative to the that used in the EPRI report) are as follows:
Primary lower head stress ratio: (11.3/11.2) = 1.01 (applicable to primary side welds but conservatively assumed applicable to the rest of the SG welds)
Secondary upper shell stress ratio: (23.01/23.5) = 0.98 (applicable to FW nozzle-to-shell welds)
W3F1-2024-0009 Enclosure Page 7 of 18
FW nozzle stress ratio: (0.84/1.38) = 0.61 (applicable to FW inside radius sections)
In the selection of the transients in Section 5 of References [9.1] and [9.2] and the subsequent stress analyses in Section 7, test conditions beyond a system leakage test were not considered since pressure tests at WF3 are performed at normal operating conditions. No hydrostatic testing had been performed at WF3 since the unit went into operation.
In Reference [9.2], clad residual stress was not considered for the primary side welds.
In a previous NRC RAI (Reference [9.19], RAI 3c), the NRC raised this issue. In response to the RAI (Reference [9.20], RAI Response 3.c), an evaluation was performed which showed that the clad residual stress has no significant impact on the conclusions of Reference [9.2] and this was found acceptable by the NRC in Section 5.3 of Reference [9.18].
5.1.3 Applicability of the Flaw Tolerance Evaluation in References [9.1] and [9.2] to WF3 Flaw tolerance evaluations were performed in References [9.1] and [9.2] consisting of probabilistic fracture mechanics (PFM) evaluations and confirmatory deterministic fracture mechanics (DFM) evaluations. The results of the PFM analyses indicate that, after a preservice inspection (PSI) followed by subsequent inservice inspections (ISI),
the U.S. Nuclear Regulatory Commissions (NRCs) safety goal of 1x10 -6 failures per year is met.
The PFM analysis in Reference [9.1] was performed using the PRobabilistic OptiMization of I nSp Ection ( PROMISE) Version 1.0 software, developed by Structural Integrity Associates. As part of the NRCs review of Southern Nuclears alternative request, the NRC performed an audit of the PROMISE Version 1.0 software as discussed in the NRCs audit plan dated May 14, 2020 (ML20128J311). The PFM analysis in Reference [9.2] was performed using the PROMISE Version 2.0 software which has not been audited by the NRC. The only technical difference between the two versions is that in PROMISE Version 1.0, the user-specified examination coverage is applied to all inspections, whereas in PROMISE Version 2.0, the examination coverage can be specified by the user uniquely for each inspection. In both Versions 1.0 and 2.0, the software assumes 100% coverage for the PSI examination.
In Section 8.2.2.2 of Reference [9.1] and Section 8.3.2.2 of Reference [9.2], a nozzle flaw density of 0.001 flaws per nozzle was assumed for the nozzle inside radius sections. In Section 3.8.5 of the SE for Vogtle in Reference [9.16], the NRC indicated that a nozzle flaw density of 0.1 flaws per nozzle should have been used. Sensitivity studies performed in Section 8.2.4.3.4 in Reference [9.2] indicated that by changing the number of flaws in the nozzle inside radius sections from 0.001 to 0.1, the probabilities of leak and rupture increased by two orders of magnitude but were still significantly below the acceptance criterion of 1x10 -6 per year. A comparison of the PSI/ISI scenarios used in the sensitivity studies performed in References [9.1] and [9.2] to that at WF3 is provided below. Note that the assumption below of a 30-year ISI deferral is conservative compared to the end of currently licensed operating life for each plant.
For the WF3 replacement SGs installed in 2012 (second period of the third inspection interval), PSI examinations have been performed followed by ISI examinations for one complete 10-year interval(s) following SG replacement. The PSI/ISI scenario considered W3F1-2024-0009 Enclosure Page 8 of 18
is therefore PSI plus one 10-year ISI examinations to be followed by two 30-year ISI deferrals (PSI+10+40+70).
5.1.4 Limiting PSI/ISI Scenario From Reference [9.1], the limiting component for Item Nos. C2.21 and C2.22 is the FW nozzle. Therefore, for WF3 three separate evaluations are performed with the following limiting PSI/ISI scenarios:
- 1. FW nozzle inside radius sections (PSI+10+40+70)
The above limiting PSI/ISI scenarios for WF3 were not specifically considered in the Reference [9.1] and [9.2] PFM evaluations in combination with key variables, as evaluated by the NRC in Section 4.0 (page 6) of the Reference [9.16] Safety Evaluation.
Therefore, the following additional plant-specific evaluations were performed with the limiting PSI/ISI scenarios shown above.
5.1.5 FW Nozzle Inside Radius Section From Reference [9.1], the critical location for the inside radius section is FW nozzle Case ID FEW-P1N. An evaluation similar to that shown in Table 8-28 of Reference
[9.1] was performed for this location assuming a nozzle flaw density of 0.1, a fracture toughness of 200 ksi in and a standard deviation 5 ksi in as recommended by the NRC in Reference [9.14]. A stress multiplier of 1.7 was applied. This stress multiplier was chosen to result in probability of rupture or probability of leakage close to the acceptance criteria after 80 years. The results of the evaluation, using PROMISE Version 1.0, are summarized in Table 3 and show that after 80 years of plant operation the probabilities of rupture and leakage are below the acceptance criterion of 1.0x10 -6.
Table 3 Sensitivity to Combined Effects of Fracture Toughness, Stress, and Nozzle Flaw Density for 80 Years for Westinghouse (bounds CE design) FW Nozzle Inside Radius Section (Case ID FEW-P1N from Reference [9.1])
Probability per Year for Combined Case KIC = 200 ksiin.
Time SD = 5 ksiin.
(yr) Stress Multiplier = 1.7 Nozzle Flaw Density = 0.1 PSI+10+40+70 Rupture Leak 10 3.30E08 1.72E07
20 2.25E08 1.06E07 30 4.97E08 2.03E07
40 2.59E07 5.54E07 W3F1-2024-0009 Enclosure Page 9 of 18
Probability per Year for Combined Case KIC = 200 ksiin.
Time SD = 5 ksiin.
(yr) Stress Multiplier = 1.7 Nozzle Flaw Density = 0.1 PSI+10+40+70 Rupture Leak 50 2.09E07 4.46E07
60 1.76E07 3.82E07 70 1.55E07 3.43E07
80 1.36E07 3.00E07
5.1.6 FW Nozzle-to-Shell Weld For the FW nozzle-to-shell weld, Table 8-16 of Reference [9.1] indicates that the critical Case ID is FEW-P3A. For the evaluation, a nozzle flaw density of 1 flaw per nozzle was assumed. A fracture toughness of 200 ksi in and standard deviation 5 ksiin were also used. A stress multiplier of 1.2 was applied such that probability of rupture or probability of leakage are close to the acceptance criteria after 80 years. The results of the evaluation, using PROMISE Version 1.0, are summarized in Table 4 and show that after 80 years of plant operation the probabilities of rupture and leakage are below the acceptance criterion of 1.0x10 -6.
Table 4 Sensitivity to Combined Effects of Fracture Toughness, Stress, and Nozzle Flaw Density for 80 Years for Westinghouse (bounds CE design) FW Nozzle-to-Shell Weld (Case ID FEW-P3A from Reference [9.1])
Probability per Year for Combined Case KIC = 200 ksiin.
Time SD = 5 ksiin.
(year) Stress Multiplier = 1.2 Nozzle Flaw Density = 1 PSI+10+40+70 Rupture Leak 10 1.00E08 1.00E08 20 5.00E09 5.00E09
30 3.33E09 5.33E08 40 2.50E09 9.20E07
50 2.00E09 7.40E07 W3F1-2024-0009 Enclosure Page 10 of 18
Probability per Year for Combined Case KIC = 200 ksiin.
Time SD = 5 ksiin.
(year) Stress Multiplier = 1.2 Nozzle Flaw Density = 1 PSI+10+40+70 Rupture Leak 60 1.67E09 6.45E07
70 1.43E09 6.09E07 80 1.25E09 5.34E07
5.1.7 Remainder of WF3 SG Welds For the remaining SG welds, Table 8-32 of Reference [9.2] indicates that the critical Case ID is SGPTH-P4A. This case was evaluated for the limiting inspection scenario of PSI+10+40+70, a flaw density of 1.0 flaw per weld, a fracture toughness of 200 ksi in and a standard deviation 5 ksiin. A relatively high with a stress multiplier of 1.8 was applied. The results of the evaluation, using PROMISE Version 2.0, are summarized in Table 5 and show that after 80 years of plant operation the probabilities of rupture and leakage are well below the acceptance criterion of 1.0x10 -6.
Table 5 Sensitivity to Combined Effects of Fracture Toughness, Stress, and Nozzle Flaw Density for 80 Years for the Remaining SG Welds (CE or B&W)
(Case ID SGPTH-P4A from Reference [9.2])
Probability per Year for Combined Case KIC = 200 ksiin.
Time SD = 5 ksiin.
(year) Stress Multiplier = 1.8 Nozzle Flaw Density = 1 PSI+10+40+70 Rupture Leak 10 1.00E08 1.00E08 20 5.00E09 5.00E09
30 2.33E08 3.33E09 40 2.33E07 2.50E09
50 1.86E07 2.00E09 60 1.70E07 1.67E09 70 1.71E07 1.43E09 W3F1-2024-0009 Enclosure Page 11 of 18
Probability per Year for Combined Case KIC = 200 ksiin.
Time SD = 5 ksiin.
(year) Stress Multiplier = 1.8 Nozzle Flaw Density = 1 PSI+10+40+70 Rupture Leak 80 1.50E07 1.25E09
The plant-specific PFM evaluation presented above for WF3 indicates that with conservative inputs of the critical parameters, the probabilities of rupture and leakage are well below the acceptance criterion of 1.0x10 -6 failures per year. The stress multipliers applied in Tables 3 through 5 are greater than the plant-specific stress ratios determined previously from the geometrical data in Tables 1 and 2 and therefore the stresses and fracture mechanics evaluations in the References [9.1] and [9.2] EPRI reports are conservative in application to WF3. It should also be noted that the evaluation incorporates conservative assumptions with regard to the PSI/ISI scenarios.
Furthermore, the evaluation was performed for 30 years, which is longer than the deferral being sought by Entergy in this Request for Alternative.
An evaluation was performed to show acceptability of the low K IC values at the beginning and ending of the heatup/cooldown transient for the FW and MS nozzles to address Item No. 2.e.iii during the NRC audit of PROMISE [9.25]. The evaluation was performed using an RTNDT value of 60oF, the maximum allowed by BTP 5-3 [9.26]. The RT NDT value of 60oF bounds the value of -5 oF in Attachment 2 for the SG materials at WF3. The evaluation showed acceptable results for the limiting Case IDs from the Reference [9.1]
EPRI report. This was found acceptable by the NRC [9.27]. A similar evaluation was performed for the remainder of the SG welds in Reference [9.28] using the limiting Case ID from the Reference [9.2] EPRI report to address NRC RAI-6 in Reference [9.29]. In this evaluation, the limiting RTNDT value of 60oF was used and acceptable results were also obtained.
The PFM evaluations documented in References [9.1] and [9.2] and the plant-specific evaluations above used a Section XI, Appendix VIII-based probability of detection (POD) curve in the PFM evaluation because most ISI examinations of major plant Class 1 and Class 2 components are performed using Appendix VIII procedures. However, for Class 2 components, the use of Appendix VIII procedures is plant-specific. In the case of WF3, ASME Code,Section V procedures may be used for at least some Class 2 components (e.g., the FW nozzle inside radius section examinations). Based on the observations made by the NRC in Section 3.8.8.2, page 21 of the Vogtle SE [9.16], the use of the ASME Code,Section XI, Appendix VIII based POD curve for inspections based on ASME Code,Section V procedures would have minimal impact of the PFM results since the POD curve is not one of the parameters that significantly affects the PFM results.
The DFM evaluations in Table 8-31 of Reference [9.1] and Table 8-3 of Reference [9.2]
provide verification of the above PFM results for WF3 by demonstrating that it takes approximately 80 years for a postulated flaw with an initial depth equal to ASME Code, W3F1-2024-0009 Enclosure Page 12 of 18
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.
5.1.8 Inspection History As described in Section 8.2.4.1.1 of Reference [9.1] and Section 8.3.4.1 of Reference
[9.2], preservice examination (PSI) refers to the collective examinations required by ASME Code,Section III during fabrication and any ASME Code,Section XI examinations performed prior to service. The Section III fabrication examinations required for these components were robust and any Section XI preservice examinations further contributed to thorough initial examinations.
Inspection history for WF3 (including examinations performed to date, examination findings, inspection coverage, and relief requests) is presented in Attachment 1. As shown in the attachment, all examinations received > 90% coverage, and no flaws that exceeded the ASME Code,Section XI acceptance standards were identified during any examinations.
5.1.9 Industry Survey The inspection history for these components as obtained from an industry survey is presented in Attachment 2. The results of the survey indicate that these components are very flaw tolerant.
5.1.10 Performance Monitoring To provide additional defense in depth to the Request for Alternative, Entergy will adopt a performance monitoring plan during the requested deferral period. The components listed below have received a PSI examination and at least one ISI examination during the third 10-year interval. Since a 27-year, 18-day deferral is being requested for these component examinations, Entergy would like to propose performing the examinations (previously planned for the fourth 10-year interval) prior to end of currently licensed operating life, which is scheduled to end on December 18, 2044.
Components: 04-076,04-068, 04-073,04-074 5.1.11 Conclusion It is concluded that the SG pressure-retaining welds and full penetration welded nozzles are very flaw tolerant. PFM and DFM evaluations performed as part of the technical basis reports [9.1] and [9.2], supplemented by plant-specific evaluations performed as part of this Request for Alternative, demonstrate that using conservative PSI/ISI inspection scenarios for all plants, the NRC safety goal of 1.0x10 -6 failures per reactor year is met with considerable margins. Plant-specific applicability of the technical basis to WF3 is demonstrated in Attachment 1. The requested ISI deferrals provide an acceptable level of quality and safety in lieu of the current ASME Code,Section XI 10-year inspection frequency.
Operating and examination experience demonstrates that these components have performed with very high reliability, mainly due to their robust design. Attachment 1 shows the examination history for the SG welds examined in the two most recent 10-year inspection intervals.
In addition to the required PSI examinations for these SG welds and components, WF3 has performed multiple ISI examinations through the current 10-year inspection interval.
W3F1-2024-0009 Enclosure Page 13 of 18
No flaws that exceeded the ASME Code,Section XI acceptance standards were identified during any examinations, as shown in Attachment 1.
Finally, as discussed in Reference [9.3], for situations where no active degradation mechanism is present, it was concluded that subsequent ISI examinations do not provide additional value after PSI has been performed and the inspection volumes have been confirmed to be free of defects.
Therefore, Entergy requests the NRC grant this proposed alternative in accordance with 10 CFR 50.55a(z)(1).
6.0 DURATION OF PROPOSED ALTERNATIVE The proposed alternative is to defer the ISI examinations for these Item Nos. for the SGs at WF3 from the current ASME Code,Section XI, Division 1 10-year requirement to the end of currently licensed operating life, which is scheduled to end on December 18, 2044. This equates to an extension of 27 years, 18 days from the end of the third inservice inspection interval (November 30, 2017) at which time all ASME Code,Section XI, Division 1 requirements were satisfied.
7.0 PRECEDENTS 7.1 Vogtle Units 1 and 2 The following previous submittal has been made by Southern Nuclear to provide relief from the ASME Code,Section XI Examination Category C-B (Item Nos. C2.21 and C2.22) surface and volumetric examinations based on the Reference [9.1] technical basis report:
Southern Nuclear letter to NRC - "Vogtle Electric Generating Plant, Units 1 & 2 Proposed Inservice Inspection Alternative VEGP-ISI-ALT-04-04 Version 2.0,"
(ML20253A311), dated September 9, 2020 [9.15].
The NRC issued a safety evaluation of the Southern Nuclear request for alternative on January 11, 2021.
NRC letter to Southern Nuclear - "Vogtle Electric Generating Plant, Units 1 & 2 -
Relief Request for Proposed Inservice Inspection Alternative VEGP-ISI-ALT-04-04 to the Requirements of ASME Code," (ML20352A155),
dated January11, 2021 [9.16].
7.2 Millstone Unit 2 The following previous submittal has been made by Dominion Energy to provide relief from the ASME Section XI Examination Category B-B (Item No. B2.40) and Category C-A (Item Nos. C1.10, C1.20 and C1.30) surface and volumetric examinations based on the Reference [9.2] technical basis report:
Dominion Energy letter to the US NRC - "Dominion Energy Nuclear Connecticut, Inc. Millstone Power Station Unit 2 Alternative Request RR-05 Inspection Interval Extension for Steam Generator Pressure-Retaining Welds and Full-Penetration Welded Nozzles," (ML20198M682), dated July 15, 2020 [9.17].
W3F1-2024-0009 Enclosure Page 14 of 18
The NRC issued a safety evaluation of the Do minion Energy request for alternative on July 16, 2021.
NRC letter to Dominion Energy - "Millstone Power Station Unit 2 - Authorization and Safety Evaluation for Alternative Request No. RR-05-06," (ML21167A355),
dated July 16, 2021 [9.18].
7.3 Other Approved Actions Related to Inspections of SG Welds and Components In addition, the following is a list of approved actions (including relief requests and topical reports) related to inspections of SG welds and components:
NRC letter to Northeast Nuclear Energy Company - "Safety Evaluation of the Relief Request Associated with the First and Second 10-Year Interval of the Inservice Inspection (ISI) Plan, Millstone Nuclear Power Station, Unit 3,"
(ML003730922), dated July 24, 2000 [9.30].
NRC letter to Southern Nuclear - "Second 10-Year Interval Inservice Inspection Program Plan Requests for Relief 13, 14, 15, 21 and 33 for Vogtle Electric Generating Plant, Units 1 and 2," (ML011640178), dated June 20, 2001 [9.31].
NRC letter to CP&L - "Shearon Harris Nuclear Power Plant Unit 1 - Request for Relief 2R1-019, 2R1-020, 2R1-021, 2R1-022, 2R2-009, 2R2-010, and 2R2-011 for the Second Ten-Year Interval Inservice Inspection Program Plan,"
(ML093561419), dated January 7, 2010 [9.32].
NRC letter to Dominion Nuclear - "Millstone Power Plant Unit No. 2 - Issuance of Relief Requests RR-89-69 Through RR-89-78 Regarding Third 10-Year Interval Inservice Inspection Program Plan," (ML120541062), dated March 12, 2012 [9.33].
NRC letter to PG&E - "Diablo Canyon Plant, Units 1 and 2 - Relief Request; NDE 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, Inservice Inspection Program (CAC Nos.
MF6646 and MF6647)," (ML15337A021), dated December 8, 2015 [9.34].
7.4 Similar Topical Reports for Relief for Class 1 Nozzles In addition, there are precedents related to similar topical reports that justify relief for Class 1 nozzles:
Based on studies presented in Reference [9.4], the NRC approved extending PWR reactor vessel nozzle-to-shell welds from 10 to 20 years in Reference [9.5].
Based on work performed in BWRVIP-108 [9.6] and BWRVIP-241 [9.8], the NRC approved the reduction of BWR vessel FW nozzle-to-shell weld examinations (Item No. B3.90 for BWRs from 100% to a 25% sample of each nozzle type every 10 years) in References [9.7] and [9.9]. The work performed in BWRVIP-108 and BWRVIP-241 provided the technical basis for ASME Code Case N-702 [9.10], which has been conditionally approved by the NRC in Revision 19 of Regulatory Guide 1.147 [9.11].
W3F1-2024-0009 Enclosure Page 15 of 18
8.0 ACRONYMS ASME American Society of Mechanical Engineers B&W Babcock and Wilcox BWR Boiling Water Reactor BWRVIP Boiling Water Reactor Vessel and Internals Program CE Combustion Engineering CFR Code of Federal Regulations DFM Deterministic fracture mechanics EAF Environmentally assisted fatigue EPRI Electric Power Research Institute FAC Flow accelerated corrosion FEA Finite element analysis FW Feedwater ISI Inservice Inspection MIC Microbiologically influenced corrosion MS Main Steam NPS Nominal pipe size NRC Nuclear Regulatory Commission NSSS Nuclear steam supply system O.D. Outside diameter POD Probability of detection PFM Probabilistic fracture mechanics PSI Preservice inspection PWR Pressurized Water Reactor SCC Stress corrosion cracking SG Steam Generator WEC Westinghouse Electric Company
9.0 REFERENCES
9.1 EPRI Report - 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 W3F1-2024-0009 Enclosure Page 16 of 18
9.2 EPRI Report - Technical Bases for Inspection Requirements for PWR Steam Generator Class 1 Nozzle-to-Vessel Welds and Class 1 and Class 2 Vessel Head, Shell, Tubesheet-to-Head and Tubesheet-to-Shell Welds. EPRI, Palo Alto, CA: 2019. 3002015906.
9.3 American Society of Mechanical Engineers (ASME) - Risk-Based Inspection:
Development of Guidelines, Volume 2-Part 1 and Volume 2-Part 2, Light Water Reactor (LWR) Nuclear Power Plant Components. CRTD-Vols. 20-2 and 20-4, ASME Research Task Force on Risk-Based Inspection Guidelines, Washington, D.C., 1992 and 1998.
9.4 WCAP-16168-NP-A, Rev. 3 - B. A. Bishop, C. Boggess, N. Palm, "Risk-Informed extension of the Reactor Vessel In-Service Inspection Interval," dated October 2011.
9.5 NRC to PWROG - "Revised Safety Evaluat ion 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," (ML111600295 and ML111600303), dated July 26, 2011.
9.6 EPRI - 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.
9.7 NRC letter to BWRVIP - 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)," (ML073600374), dated December 19, 2007.
9.8 EPRI - BWRVIP-241: BWR Vessels and Internals Project, Probabilistic Fracture Mechanics Evaluation for the Boiling Water Reactor Nozzle-to-Shell Welds and Nozzle Blend Radii, EPRI, Palo Alto, CA 2010. 1021005.
9.9 NRC letter to Southern Nuclear - 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)," (ML13071A240 and ML13071A233), dated April 19, 2013.
9.10 ASME - 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.
9.11 NRC - Regulatory Guide 1.147, Revision 18, "Inservice Inspection Code Case Acceptability, ASME Code Section XI, Division 1," dated March 2017.
9.12 EPRI letter to NRC - BWR Vessel & Internals Project (BWRVIP) Memo No.
2019-016, "White Paper on Suggested Content for PFM Submittals to the NRC,"
(ML19241A545), dated February 27, 2019.
9.13 NRC - Regulatory Guide 1.245, Revision 0, "Preparing Probabilistic Fracture Mechanics Submittals," dated January 2022.
W3F1-2024-0009 Enclosure Page 17 of 18
9.14 NRC - Report NUREG/CR-7278, "Technical Basis for the use of Probabilistic Fracture Mechanics in Regulatory Applications," dated January 2022.
9.15 Southern Nuclear (SNC) letter to NRC - "Vogtle Electric Generating Plant, Units 1
& 2 Proposed Inservice Inspection Alter native VEGP-ISI-ALT-04-04 Version 2.0,"
(ML20253A311), dated September 9, 2020.
9.16 NRC letter to SNC - "Vogtle Electric Generating Plant, Units 1 & 2 - Relief Request for Proposed Inservice Inspection Alternative VEGP-ISI-ALT-04-04 to the Requirements of the ASME Code," (ML20352A155), dated January 11, 2021.
9.17 Dominion Entergy letter to NRC - "Millstone Power Station Unit 2 Alternative Request RR-05 Inspection Interval Extension for Steam Generator Pressure-Retaining Welds and Full-Penetration Welded Nozzles,"
(ML20198M682), dated July 15, 2020.
9.18 NRC letter to Dominion Entergy - "Millstone Power Station Unit 2 - Authorization and Safety Evaluation for Alternative Request No. RR-05-06," (ML21167A355),
dated July 16, 2021.
9.19 NRC Email to Dominion Entergy - "Millstone Unit 2 - Request for Additional Information - Alternative Request RR-05-06 Inspection Interval Extension for SG Pressure Retaining Welds and Full-Penetration Welded Nozzles,"
(ML21034A576), dated February 3, 2021.
9.20 Dominion Entergy letter to NRC - "Millstone Power Station Unit 2 - Response to Request for Additional Information for Alternative Request RR-05 Inspection Interval Extension for Steam Generator Pressure-Retaining Welds and Full-Penetration Welded Nozzles," (ML21081A136), dated March 19, 2021.
9.21 Entergy - Waterford 3 S.E.S. Drawing No. 5817-13788, Sheet 1, "Replacement Steam Generator W3 General Arrangement," Revision 0 (PROPRIETARY).
9.22 Entergy - Waterford 3 S.E.S. Drawing No. 5817-13787, Sheet 1, "Replacement Steam Generator Waterford 3 Outline," Revision 0 (PROPRIETARY).
9.23 Not used.
9.24 Not used.
9.25 NRC letter to SNC - "Vogtle Electric Generating Plant, Units 1 and 2 - Audit Plan for Relief Request Inservice Inspection Alternative VEGP-ISI-ALT-04-04,"
(ML20128J311), dated May 14, 2020.
9.26 NRC - NUREG-0800 - Chapter 5, Branch Technical Position (BTP) 5-3, Revision 2, Fracture Toughness Requirements, dated March 2007.
9.27 SNC Letter to NRC - "Vogtle Electric Generating Plant, Units 1 and 2 - Audit Report for the PROMISE Version 1.0 Probabilistic Fracture Mechanics Software Used in Relief Request VEGP-ISI-ALT-04-04," (ML20258A002), dated December 10, 2020.
9.28 Constellation letter to NRC - RS-22-084, "Response to Request for Additional Information - Proposed Alternative for Examinations of Examination Categories B-B, B-D, and C-A Steam Generator Pressure Retaining Welds and Full Penetration Welded Nozzles," (ML22168A005), dated June 17, 2022.
W3F1-2024-0009 Enclosure Page 18 of 18
9.29 NRC Email to Constellation - "Draft RAIs for Requests for Alternatives I4R-17, I4R-23, ISI-05-018, I6R-10," (ML22129A013), dated May 6, 2022.
9.30 NRC letter to Northeast Nuclear Energy Company - "Safety Evaluation of Relief Request Associated with the First and Second 10-Year Interval of the Inservice Inspection (ISI) Plan, Millstone Nuclear Power Station, Unit 3," (ML003730922),
dated July 24, 2000.
9.31 NRC letter to Southern Nuclear - "Second 10-Year Interval Inservice Inspection Program Plan Requests for Relief 13, 14, 15, 21 and 33 for Vogtle Electric Generating Plant, Units 1 and 2," (ML011640178), dated June 20, 2001.
9.32 NRC letter to CP&L - "Shearon Harris Nuclear Power Plant Unit 1 - Request for Relief 2R1-019, 2R1-020, 2R1-021, 2R1-022, 2R2-009, 2R2-010, and 2R2-011 for the Second Ten-Year Interval Inservice Inspection Program Plan,"
(ML093561419), dated January 7, 2010.
9.33 NRC letter to Dominion Nuclear - "Millstone Power Plant Unit No. 2 - Issuance of Relief Requests RR-89-69 Through RR-89-78 Regarding Third 10-Year Interval Inservice Inspection plan," (ML120541062), dated March 12, 2012.
9.34 NRC letter to PG&E - "Diablo Canyon Plant, Units 1 and 2 - Relief Request; NDE 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, Inservice Inspection Program," (ML15337A021), dated December 8, 2015.
10.0 ATTACHMENTS
- 1. Plant-Specific Applicability for WF3
- 2. Results of Industry Survey
Enclosure, Attachment 1
Plant-Specific Applicability for WF3
W3F1-2024-0009 Enclosure, Attachment 1 Page 1 of 11
PLANT-SPECIFIC APPLICABILITY FOR WF3
Section 9 of References [1-1] and [1-2] 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 WF3 is provided in Table 1-1 and indicates that all plant-specific requirements are met. Therefore, the results and conclusions of the EPRI reports are applicable to WF3.
Table 1-1 Applicability of References [1-1] and [1-2] Representative Analyses to WF3 Items No. B2.40 (SG Primary Side Shell Welds)
Category Requirement from Reference [1-1] Applicability to WF3 General The Loss of Power transient (involving auxiliary The WF3 RSGs have not experienced Requirements feedwater being introduced into a hot SG that a loss of power transient resulting in has been boiled dry following blackout, resulting unheated auxiliary feedwater being in thermal shock of a portion of the vessel) is introduced into a hot SG that has been not considered in this evaluation due to its rarity. boiled dry following blackout, resulting In the event that such a significant thermal in thermal shock of any portion of the event occurs at a plant, its impact on the KIC vessel.
(material fracture toughness) value may require more frequent examinations and other plant actions outside the scope of this reports guidance.
The materials of the SG vessel heads and The WF3 RSG vessel heads and tubesheet must be low alloy ferritic steels which tubesheet are fabricated from SA-508 conform to the requirements of ASME Code, Grade 3 Class 2 material (per Section XI, Appendix G, Paragraph G-2110. Reference [1-3]). The RTNDT value for the material is -5°F, which is bounded by that used in the EPRI report.
SA-508 Grade 3 Class 2 material is a low alloy ferritic steel which conforms to the requirements of ASME Code,Section XI, Appendix G, Paragraph G-2110.
Specific The weld configurations must conform to those The WF3 tubesheet-to-head weld Requirements shown in Figures 1-1 and Figure 1-2 of configuration is shown in Figure 1-2 Reference [1-1]. and shows conformance with Figure 1-2 of Reference [1-1].
W3F1-2024-0009 Enclosure, Attachment 1 Page 2 of 11
Category Requirement from Reference [1-1] Applicability to WF3 The SG vessel dimensions must be within 10% Using the ID and thickness values from of the upper and lower bounds of the values Table 1 in the main section of this provided in the table in Section 9.4.3 of request for alternative, the WF3 SG Reference [1-1]. OD vessel dimensions are as follows:
SG Lower Head OD = 178.12" SG Upper Shell OD = 264.12" The dimensions are within 10% of those specified in Table 9-2 in Section 9.4.3 of Reference [1-1] for CE plants.
The component must experience transients and As shown in Table 1-2, the WF3 cycles bounded by those shown in Table 5-7 of number of cycles projected to occur Reference [1-1] over a 60-year operating life. over a 60-year operating life are significantly lower than those shown in Table 5-7 of Reference [1-1].
Items No. C1.20 and C1.30 (SG Secondary Side Shell Welds)
Category Requirement from Reference [1-1] Applicability to WF3 General The Loss of Power transient (involving auxiliary WF3 has not experienced a loss of Requirements feedwater being introduced into a hot SG that power transient resulting in unheated has been boiled dry following blackout, resulting auxiliary feedwater being introduced in thermal shock of a portion of the vessel) is into a hot SG that has been boiled dry not considered in this evaluation due to its rarity. following blackout, resulting in thermal In the event that such a significant thermal shock of any portion of the vessel.
event occurs at a plant, its impact on the KIC (material fracture toughness) value may require more frequent examinations and other plant actions outside the scope of this reports guidance.
The materials of the SG vessel shell and The WF3 RSG vessel heads and tubesheet must be low alloy ferritic steels which tubesheet are fabricated from SA-508 conform to the requirements of ASME Code, Grade 3 Class 2 material (per Section XI, Appendix G, Paragraph G-2110. Reference [1-3]). The RTNDT value for the material is -5°F, which is bounded by that used in the EPRI report.
SA-508 Grade 3 Class 2 material is a low alloy ferritic steel which conforms to the requirements of ASME Code,Section XI, Appendix G, Paragraph G-2110.
Specific The weld configurations must conform to those The WF3 weld configurations are Requirements shown in Figure 1-7 and Figure 1-8 of shown in Figures 1-3 and 1-4 and Reference [1-1]. conform to Figures 1-7 and 1-8 of Reference [1-1].
W3F1-2024-0009 Enclosure, Attachment 1 Page 3 of 11
Category Requirement from Reference [1-1] Applicability to WF3 The SG vessel dimensions must be within 10% Using the ID and thickness values from of the upper and lower bounds of the values Table 1 in the main section of this provided in the table in Section 9.4.4 of request for alternative, the WF3 SG Reference [1-1]. OD vessel dimensions are as follows:
SG Lower Head OD = 178.12" SG Upper Shell OD = 264.12" The dimensions are within 10% of those specified in Table 9-3 in Section 9.4.4 of Reference [1-1] for CE plants.
The component must experience transients and As shown in Table 1-4, the WF3 cycles bounded by those shown in Table 5-9 of number of cycles projected to occur Reference [1-1] over a 60-year operating life. over a 60-year operating life are significantly lower than those shown in Table 5-9 of Reference [1-1].
Items Nos. C2.21 and C2.22 (FW Nozzle to Shell Welds and Inside Radius Sections)
Category Requirement from Reference [1-2] Applicability to WF3 General The nozzle-to-shell weld shall be one of the The WF3 FW nozzle-to-shell weld is Requirements configurations shown in Figure 1-1 or Figure 1-2 shown in Figure 1-5 and is of Reference [1-2]. representative of the configuration shown in Figure 1-2 of Reference [1-2].
The MS nozzle is N/A since WF3 is not requesting relief from any MS nozzle examinations.
The materials of the SG shell, FW nozzles, and The WF3 RSG shell and FW nozzles MS nozzles must be low alloy ferritic steels are fabricated from SA-508 Grade 3 which conform to the requirements of ASME Class 2 material (per Reference [1-4]).
Code,Section XI, Appendix G, Paragraph The RTNDT value for the material G-2110. is -5°F, which is bounded by that used in the EPRI report.
SA-508 Grade 3 Class 2 material is a low alloy ferritic steel which conforms to the requirements of ASME Code,Section XI, Appendix G, Paragraph G-2110.
The MS nozzle material is N/A since WF3 is not requesting relief from any MS nozzle examinations.
The SG must not experience more than the As shown in Table 1-5, the WF3 SGs number of all transients shown in Table 5-5 of are not projected to experience more Reference [1-2] over a 60-year operating life. than the number of transients shown in Table 5-5 of Reference [1-2] over a 60-year operating life.
W3F1-2024-0009 Enclosure, Attachment 1 Page 4 of 11
Category Requirement from Reference [1-2] Applicability to WF3 SG Feedwater The piping attached to the FW nozzle must be The piping attached to the WF3 FW Nozzle 14-inch to 18-inch NPS. nozzle is 14-inch to 18-inch NPS (per Reference [1-4]).
The FW nozzle design must have an integrally The WF3 FW nozzle design has an attached thermal sleeve. integrally attached thermal sleeve.
Auxiliary feedwater nozzles connected directly N/A for WF3.
to the SG are not covered in this evaluation.
SG Main Steam For Westinghouse and CE SGs, the piping N/A since WF3 is not requesting relief Nozzle attached to the SG main steam nozzle must be from any MS nozzle examinations.
28-inch to 36-inch NPS.
For B&W SGs, the piping attached to the main N/A for WF3.
steam nozzle must be 22-inch to 26-inch NPS.
The SG must have one main steam nozzle that N/A since WF3 is not requesting relief exits the top dome of the SG. For B&W plants, from any MS nozzle examinations.
there may be more than one main steam nozzle; it will exit the side of the SG.
The main steam nozzle shall not significantly N/A since WF3 is not requesting relief protrude into the SG (e.g., see Figure 4-7 of from any MS nozzle examinations.
Reference [1-2]) or have a unique nozzle weld configuration (e.g., see Figure 4-6 of Reference
[1-2]).
W3F1-2024-0009 Enclosure, Attachment 1 Page 5 of 11
Table 1-2 WF3 Data for Thermal Transients for Stress Analysis of the PWR SG Primary-Side Head Welds (Comparison to Table 5-7 of Reference [1-1])
Number of Cycles for 60 WF3 60-Year Transient(1) Years from Table 5-7 of Projection Reference [1-1]
Heatup / Cooldown 300 144 / 144(2)
Plant Loading /
Unloading 5,000 536(3)
Reactor Trip 360 187(2)
Notes:
- 1. Table 5-7 of Reference [1-1] also includes allowable transient temperatures and pressures. From previous experience with C-E plants, these values are with 2% of the values used in the EPRI report. This is acceptable based on the large margins in the evaluation as demonstrated by the low plant-specific stress ratio compared to the maximum allowed stress ratio.
- 2. Table 4.3-1 (page 4.3-4 of Reference [1-5].
- 3. Reference [1-7].
W3F1-2024-0009 Enclosure, Attachment 1 Page 6 of 11
Table 1-3 WF3 Data for Thermal Transients for Stress Analysis of the PWR SG Inlet Nozzle-to-Vessel Welds (Item No. B3.130)
(Comparison to Table 5-8 of Reference [1-1])
Number of Cycles for 60 WF3 60-Year Transient(1) Years from Table 5-7 of Projection Reference [1-1]
Heatup / Cooldown 300 144 / 144(2)
Plant Loading /
Unloading 5000 536(3)
Reactor Trip 360 187(2)
Notes:
- 1. Table 5-7 of Reference [1-1] also includes allowable transient temperatures and pressures. From previous experience with C-E plants, these values are with 2% of the values used in the EPRI report. This is acceptable based on the large margins in the evaluation as demonstrated by the low plant-specific stress ratio compared to the maximum allowed stress ratio.
- 2. Table 4.3-1 (page 4.3-4) of Reference [1-5].
- 3. Reference [1-7].
W3F1-2024-0009 Enclosure, Attachment 1 Page 7 of 11
Table 1-4 WF3 Data for Thermal Transients for Stress Analysis of the PWR SG Secondary-Side Vessel Welds (Comparison to Table 5-9 of Reference [1-1])
Number of Cycles for 60 WF3 60-Year Transient(1) Years from Table 5-9 of Projection Reference [1-1]
Heatup / Cooldown 300 144 / 144(2)
Plant Loading /
Unloading 5000 536(3)
Reactor Trip 360 187(2)
Notes:
- 1. Table 5-9 of Reference [1-1] also includes allowable transient temperatures and pressures. From previous experience with C-E plants, these values are with 2% of the values used in the EPRI report. This is acceptable based on the large margins in the evaluation as demonstrated by the low plant-specific stress ratio compared to the maximum allowed stress ratio.
- 2. Table 4.3-1 (page 4.3-4) of Reference [1-5].
- 3. Reference [1-7].
W3F1-2024-0009 Enclosure, Attachment 1 Page 8 of 11
Table 1-5 WF3 Data for Thermal Transients Applicable to PWR SG Feedwater and Main Steam Nozzles (Comparison to Table 5-5 of Reference [1-2])
Number of Cycles Transient for 60 Years from WF3 60-Year Table 5-5 Projection Reference [1-2]
Heatup / Cooldown 300 144 / 144(1)
Plant Loading 5000 536(4)
Plant Unloading 5000 536(4)
Loss of Load 360 187(1)(2)
Loss of Power 60 27(3)
Notes:
- 1. Table 4.3-1 (page 4.3-4) of Reference [1-5].
- 2. Loss of Load + Reactor Trip.
- 3. Based on a Licensee Event Reports (LER) search of transient events involving Emergency Feedwater (EFW) injection.
- 4. Reference [1-7].
Table 1-6 WF3 Inspection History Item Component Exam Interval/Period/Outage Exam Coverage Relief No. ID Date Results(1) Request C2.21 04-073 04/30/17 3rd/3rd/1RF21 NRI 99.31% No C2.22 04-074 04/28/17 3rd/3rd/1RF21 NRI 100% No C1.30 04-075 05/10/17 3rd/3rd/1RF21 NRI 97.65% No C1.30 04-075 10/06/20 4th/1st/1RF23 NRI 97.65% No C1.20 04-068 04/29/17 3rd/3rd/1RF21 NRI 99.58% No B2.40 04-076 05/04/17 3rd/3rd/1RF21 NRI 97.12% No Notes:
- 1. NRI = no recordable indications.
W3F1-2024-0009 Enclosure, Attachment 1 Page 9 of 11
Figure 1-1 WF3 Steam Generator Layout [1-4]
Figure 1-2 WF3 Item No. B2.40 Weld Configuration [1-6]
W3F1-2024-0009 Enclosure, Attachment 1 Page 10 of 11
Figure 1-3 WF3 Item No. C1.20 Weld Configuration [1-3]
Figure 1-4 WF3 Item No. C1.30 Weld Configuration [1-6]
Figure 1-5 WF3 Feedwater Nozzle Configuration [1-3]
W3F1-2024-0009 Enclosure, Attachment 1 Page 11 of 11
References 1-1. EPRI - Technical Bases for Inspection Requirements for PWR Steam Generator Class 1 Nozzle-to-Vessel Welds and Class 1 and Class 2 Vessel Head, Shell, Tubesheet-to-Head and Tubesheet-to-Shell Welds. EPRI, Palo Alto, CA: 2019.
3002015906.
1-2. EPRI - 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.
1-3. Entergy - Waterford 3 S.E.S. Drawing No. 5817-13788, Sheet 1, "Replacement Steam Generator W3 General Arrangement," Revision 0 (PROPRIETARY).
1-4. Entergy - Waterford 3 S.E.S. Drawing No. 5817-13787, Sheet 1, "Replacement Steam Generator Waterford 3 Outline," Revision 0 (PROPRIETARY).
1-5. Entergy - "Waterford Steam Electric Station, Unit 3, License Renewal Application,"
(ML16088A331 thru A335), dated March 23, 2016.
1-6. Entergy - Waterford 3 S.E.S. Drawing No. 5817-13788, Sheet 2, "Replacement Steam Generator W3 General Arrangement," Revision 0 (PROPRIETARY).
1-7. Structural Integrity (SI) - Calculation No. 2200654.302, "Waterford Unit 3 (WSES) Plant Loading/Unloading and Pressurizer Insurge/Outsurge Transients," Revision 0.
Enclosure, Attachment 2
Results of Industry Survey
W3F1-2024-0009 Enclosure, Attachment 2 Page 1 of 4
RESULTS OF INDUSTRY SURVEY
Overall Industry Inspection Summary for Code Items B2.31, B2.32, B2.40, B3.130, C1.10, C1.20, and C1.30 The results of an industry survey of past inspections of SG nozzle-to-shell welds, inside radius sections and shell welds are summarized in Reference [2-1]. Table 2-1 provides a summary of the combined survey results for Item Nos. B2.31, B2.32 (see Table Note 3), B.240, B3.130, C1.10, C1.20, and C1.30. The results of the industry survey identified numerous steam generator (SG) examinations being performed with no service-induced flaws being detected.
Performing these examinations adversely impact outage activities including worker exposure, personnel safety, and radwaste. A total of 74 domestic and international boiling water reactor (BWR) and pressurized water reactor (PWR) units responded to the survey and provided information representing all PWR plant designs currently in operation in the United States. 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 1374 examinations for the components of the affected Item Nos.
were conducted, with 1148 of these specifical ly for PWR components. The majority of PWR examinations were performed on SG welds.
A relatively small number of flaws were identified during these examinations which required flaw evaluation. None of these flaws were found to be service-induced. For Item No. B2.40, examinations at two units at a single plant site identified multiple flaws exceeding the acceptance criteria of ASME Code Section XI; however, these were determined to be subsurface-embedded fabrication flaws and non-service-induced (see Table Note 1). For Item No. C1.20, two PWR units reported flaws exceeding the acceptance criteria of ASME Code,Section XI. In the first unit, a single flaw was identified, and was evaluated as an inner diameter surface imperfection. Reference [2-3] indicates that this was a spot indication with no measurable through-wall depth. This indication is therefore not considered to be service-induced but rather fabrication-related. A flaw evaluation per IWC-3600 was performed for this flaw and it was found to be acceptable for continued operation. In the second unit, multiple flaws were identified (see Table Note 2). As discussed in References [2-4] and [2-5],
these flaws were most likely subsurface weld defects typical of thick vessel welds and not service-induced. A flaw evaluation for IWC-3600 was performed for these flaws and they were found to be acceptable for continued operation.
W3F1-2024-0009 Enclosure, Attachment 2 Page 2 of 4
Table 2-1 Summary of Survey Results for SG Nozzle-to-Shell, Inside Radius Section, and Shell Weld Components Item No. No. of Examinations No. of Reportable Indications BWR PWR Total BWR PWR Total B2.31 0 30 30 0 0 0 B2.32 0 13 13 0 0 0 (Note 3)
B2.40 0 183 183 0 Note 1 Note 1 B3.130 0 135 135 0 0 0 C1.10 140 305 445 0 0 0 C1.20 54 319 373 0 Note 2 Note 2 C1.30 32 163 195 0 0 0 Totals 226 1,148 1,374 0 Notes 1 Notes 1 and 2 and 2 Notes:
- 1. Two PWR W-2 Loop units at a single plant reported multiple subsurface embedded fabrication flaws.
- 2. A single PWR W-2 Loop unit reported multiple flaws [2-4, 2-5].
- 3. Item No. B2.32 was evaluated in the Reference [2-1] technical basis and included in the industry survey, but is not contained in the scope of this alternative request.
W3F1-2024-0009 Enclosure, Attachment 2 Page 3 of 4
Overall Industry Inspection Summary for Code Items C2.21, C2.22, and C2.32 The results of an industry survey of past inspections of SG main steam (MS) and feedwater (FW) nozzles are summarized in Reference [2-2]. Table 2-2 provides a summary of the combined survey results for Item Nos. C2.22, C2.21, and C2.32 (see Table Note 1). The results identify that SG MS and FW Nozzle-to-Shell Welds and Nozzle Inside Radius Section examinations being performed with no service-induced flaws being detected. Performing these 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 NSSS vendors (i.e., B&W, CE, and Westinghouse). A total of 727 examinations for Item Nos. C2.21, C2.22, and C2.32 (see Table Note 1) components were conducted, with 563 of these specifically for PWR components. The majority of the PWR examinations were performed on SG 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 2-2 Summary of Survey Results for SG Main Steam and Feedwater Nozzle Components Number Number of Number of Plant Type of Units Examinations Reportable Indications BWR 27 164 0 PWR 47 563 2 Totals 74 727 (Note 1) 2 Notes:
- 1. Item No. C2.32 was evaluated in the Reference [2-2] technical basis and included in the industry survey but is not contained in the scope of this alternative request.
W3F1-2024-0009 Enclosure, Attachment 2 Page 4 of 4
References 2-1. EPRI - Technical Bases for Inspection Requirements for PWR Steam Generator Class 1 Nozzle-to-Vessel Welds and Class 1 and Class 2 Vessel Head, Shell, Tubesheet-to-Head and Tubesheet-to-Shell Welds. EPRI, Palo Alto, CA: 2019.
3002015906.
2-2. EPRI - 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.
2-3. Exelon letter to NRC - "Byron Station Unit 2 90-Day Inservice Inspection Report for Interval 3, Period 3, (B2R17)," (ML13217A093), dated July 29, 2013.
2-4. Nuclear Management Company (NMC) letter to NRC - "Prairie Island Nuclear Generating Plant, Unit 1 Inservice Inspection Summary Report, Interval 3, Period 3 Refueling Outage Dates 1-19-2001 to 2-25-2001 Cycle 20 / 05-26-99 to 02-25-2001,"
(ML011550346), dated May 29, 2001.
2-5. NMC letter to NRC - "Response to Opportunity for Comment on Task Interface Agreement (TIA) 2003-01, 'Application of ASME Code Section XI, IWB-2430 Requirements Associated with Scope of Volumetric Weld Expansion at the Prairie Island Nuclear Generating Plant,'" (ML031040553), dated April 4, 2003.