ML24086A429
| ML24086A429 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 03/25/2024 |
| From: | James Holloway Virginia Electric & Power Co (VEPCO) |
| To: | Office of Nuclear Reactor Regulation, Document Control Desk |
| References | |
| 24-142A | |
| Download: ML24086A429 (1) | |
Text
VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261 March 25, 2024 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555 VIRGINIA ELECTRIC AND POWER COMPANY NORTH ANNA POWER STATION UNIT 1 Serial No.
NRA/JHH Docket No.
License No. 24-142A RO 50-338 NPF-4 RESPONSE TO REQUEST FOR ADDIT IONAL INFORMATION ASSOCIATED WITH ALTERNATIVE REQUEST N1-I5-NDE-007 REQUEST TO DEFER ADDITIONAL REQUIRED REACTOR COOLANT PUMP CASING INSPECTION By letter dated March 22, 2024 (ADAMS Accession No. ML24082A274), Virginia Electric and Power Company (Dominion Energy Virginia) submitted Alternative Request N 1-I5-NDE-007 to request Nuclear Regulatory Commission (NRG) approval for a proposed alternative to the examination requirements of the American Society of Mechanical Engineers Boiler and Pres sure Vessel Code (ASME Code),Section XI, Subsection IWB-2430, for North Anna Power Station (NAPS) Unit 1. The proposed alternative would defer the inspection of a reactor coolant pump casing from the Spring 2024 refueling outage to the Fall 2025 refueling outage, before the start of Cycle 32.
By email dated March 23, 2024, the NRG Project Manager for NAPS provided a draft request for additional information (RAI) for the NRG staff to complete its review of the proposed alternative request. The NRG RAI and the associated Dominion Energy Virginia response are provided in the attachment.
This letter updates the request date for verbal approval to March 27, 2024.
Should you have any questions regarding this submittal, please contact Julie Hough at (804) 273-3586.
Sincerely, r James E. Holloway Vice President Nuclear Engineering & Fleet Support
Attachment:
Response to Request for Additional Information Associated with Alternative Request N1-I5-NDE-007, Request to Defer Additional Reactor Coolant Pump Casing Inspection Commitments made in this letter: None cc: U. S. Nuclear Regulatory Commission, Region II Marquis One Tower 245 Peachtree Center Avenue, NE, Suite 1200 Atlanta, Georgia 30303-1257 Mr. G. E. Miller NRC Senior Project Manager -North Anna Power Station U.S. Nuclear Regulatory Commission One White Flint North Mail Stop 9E3 11555 Rockville Pike Rockville, Maryland 20852-2738 NRC Senior Resident Inspector North Anna Power Station Serial No. 24-142A Docket No. 50-338 Page 2 of 2 ATTACHMENT Serial No. 24-142A Docket No. 50-338 RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION ASSOCIATED WITH ALTERNATIVE REQUEST N1-I5-NDE-007, REQUEST TO DEFER ADDITIONAL REACTOR COOLANT PUMP CASING INSPECTION VIRGINIA ELECTRIC AND POWER COMPANY (DOMINION ENERGY VIRGINIA)
NORTH ANNA POWER STATION UNIT 1 Serial No. 24-142A Alternative Request N1-I5-NDE-007 RAI Response Attachment, Page 1 of 12 RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION Alternative Request N1-I5-NDE-007 to Defer Additional Reactor Coolant Pump Casing Inspection
- 1. 0 Introduction By letter dated March 22, 2024 (Agencywi de Documents and Access Management System (ADAMS) Accession No. ML24082A274), Virginia Electric and Power Company (Dominion Energy Virginia, the licensee) requested relief from the examination requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, Subsection IWB-2430, for North Anna Power Station Unit 1. The proposed alternative would move the inspection of a reactor coolant pump from the Spring 2024 refueling outage to the Fall 2025 refueling outage.
Pursuant to Title 10, Code of Federal Regulations, Part 50, 10 CFR 50.55a(z)(2), the licensee submitted for Nuclear Regulatory Commission (NRG) review and approval Alternative Request N1-/5-NDE-007 for Norlh Anna, Unit 1.
The NRG staff requests the following additional information (RAJ) to complete its review of the alternative request.
2.0 REGULATORYEVALUATION
The inservice inspection (ISi) of ASME Code Class 1, 2, and 3 components is to be performed in accordance with Section XI of the ASME Code and applicable editions and addenda as required by 10 CFR 50.55a(g), "Preservice and inservice inspection requirements."
Pursuant to 10 CFR 50.55a(g)(4), "lnservice inspection standards requirement for operating plants," ASME Code Class 1, 2, and 3 components (including supports) must meet the requirements, except the design and access provisions and the preservice examination requirements, set forth in the ASME Code,Section XI, to the extent practical within the limitations of design, geometry, and materials of construction of the components. The regulations require that inservice examination of components and system pressure tests conducted during the first 10-year ISi interval and subsequent intervals comply with the requirements in the latest edition and addenda of Section XI of theASME Code incorporated by reference in 10 CFR 50.55a(a)(1)(ii), 18 months prior to the start of the 120-month interval, subject to the conditions listed in 10 CFR 50.55a(b)(2).
Pursuant to 10 CFR 50.55a(z), "Alternatives to codes and standards requirements,"
alternatives to the requirements of paragraph (g) may be used, when authorized by the NRG, if (1) the proposed alternatives would provide an acceptable level of quality and Serial No. 24-142A Alternative Request N1-I5-NDE-007 RAI Response Attachment, Page 2 of 12 safety, or (2) compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
3.0 Request for Information RAJ Question No. 1 Section 5 of Attachment 1 to the March 22, 2024 submittal states that " ... This request proposes an alternative to performing the additional examination of an RCP casing during the current outage in accordance with IWB-2430(a)(1)(a) to performing the additional examination of an RCP casing during the next refueling outage. The next scheduled pump replacement is currently planned for the next refueling outage, N1 R31 (Fall 2025) ... ":
(1) Clarify whether the additional examination will be performed on either 1-RC P-18 RCP or 1-RC-P-1C RCP for the Fall 2025 refueling outage.
(2) Some parls of the submittal mention the pump refurbishment. Clarify whether pump replacement means the refurbishment of hydraulic parls of the RCP such as impeller, turning vane diffuser, diffuser adapter, and casing.
(3) Discuss whether pump casing will be replaced as parl of periodic pump refurbishment.
Dominion Energy Virginia Response to RAI Question No. 1 1.Based on the current pump refurbishment schedule, 1-RC-P-1 B is scheduled to be inspected/refurbished during the Fall 2025 refueling outage. However, if there are unforeseen circumstances during the next operating cycle that result in 1-RC-P-1C inspection/refurbishment in the Fall 2025 refueling outage, this will still meet the code requirement and proposed alter native to inspect another pump casing.
2.Based on the refurbishment strategy, the initial scope involves a decontamination cycle, disassembly, cleaning, and inspection on a removed (used) RCP. Base scope includes the replacement of the line shaft components (shaft, journal bearing, impeller), thermal barrier to main flange and thermal barrier to casing gaskets, main flange bolts, in addition to upgrades on small bore piping welds, turning vane bolts, and diffuser adapter cap screws. Any additional scope or replacement of additional subcomponents is dependent on the material condition of the component(s) and evaluation of any other potential remediation/repair options.
3.Pump casings are not part of the pump refurbishment project and are not planned to be replaced.
RAJ Question No. 2 Serial No. 24-142A Alternative Request N1-I5-NDE-007 RAI Response Attachment, Page 3 of 12 Section 6 of Attachment 1 to the March 22, 2024 submittal states that " ... The proposed alternative is requested to complete the additional examination of an RCP casing as required by IWB-2430 during the fall 2025 refueling outage, which is before the end of the current 5th ISi Interval, which ends on April 30, 2029 ... "
Clarify the exact duration of the proposed alternative, i.e., whether the proposed alternative is requested for the duration up to the fall 2025 refueling outage or up to the end of the 5th ISi interval.
Dominion Energy Virginia Response to RAI Question No. 2 The proposed alternative is to complete the additional examination of an RCP casing as required by IWB-2430 during the fall 2025 refueling outage. The additional casing examination will be completed before Cycle 32 startup.
RAJ Question No. 3 By Jetter dated August 24, 2020 (ML20246G703), the licensee submitted the subsequent license renewal application (SLRA) for Norlh Anna Units 1 and 2. Section
- 4. 7. 6 of the SLRA discusses time limited aging analysis of reactor coolant pump inspection respect of ASME Code Case N-481. As a result of the degradation at 1-RC P-1 A RCP:
(1) discuss how the SLRA will be supplemented to include the subject operating experience.
(2) discuss whether Section 4. 7. 6 in the Norlh Anna SLRA is still valid or it needs to be revised/updated.
Dominion Energy Virginia Response to RAI Question No. 3 The proposed alternative to complete the additional inspection during the Fall 2025 outage is within the current 60 year operating license. Dominion has confirmed that Section 4.7.6 of the North Anna SLRA remains valid, and that no updates are required.
This conclusion is based on the technical justification provided in the initial alternative request, supplemented by the clarifying information provided in response to RAI questions 8, 9 & 10 below.
RAJ Question No. 4 Page 4 of Attachment 2, Engineering Evaluation ETE-NA-2024-0033, last sentence of the first paragraph states that " ... It should also be noted that 1-RC-P-1B and 1-RC-P-1C were removed from their casings in 1982 following the identification of the broken Serial No. 24-142A Alternative Request N1-I5-NDE-007 RAI Response Attachment, Page 4 of 12 diffuser cap screws on 1-RC-P-1 A and none of the cap screws on either pump had failed but were replaced as a precaution JAW with Westinghouse Recommendations ... ":
(1) Discuss what year was the cap screws replaced on 1-RC-P-1 B and 1-RC-P-1 C pumps (i.e., were the cap screws replaced in recent years or in 1982?).
(2)If the cap screws were replaced in 1982, discuss the likelihood of failure of the cap screws in 1-RC-P-1 B or 1-RC-P-1 C pumps prior to the 2025 refueling outage.
Dominion Energy Virginia Response to RAI Question No. 4 1.The 1-RC-P-1 B and 1-RC-P-1 C Diffuser Adapter Cap Screws (DACS) were replaced in 1982, at the same time as 1-RC-P-1A DACS were replaced.
2.The failure mode of the 1-RC-P-1A diffuser adapter cap screws (DACS) that were removed in 2024 is under investig ation at this time. However, Dominion's review of historical information indicates this issue appears to be unique to 1-RC-P-1A. In 1982, all 3 pumps were removed and 1-RC-P-1A was the only pump with failed DACS. Additionally, a review of Condition Report history shows that only 1-RC-P-1A has had issues when being put onto the backseat in 2012, and also had issues with the transition off of backseat in 2013. Back-seating of the reactor coolant pump is uncoupling the pump from the motor and lowering the shaft to the end of the axial travel. The backseat is the interface between the shaft journal step and the thermal barrier bevel. This provides the reactor coolant system maintenance boundary for reactor coolant pump seal replacement activities. Issues with back-seating the Model 93A RCPs could be an indicator of potential issues with the Diffuser Adapter per Westinghouse lnfoGram 95008A.
It should be noted that the cause evalu ation and pump performance history described here is preliminary.
Based on the lack of CR's identifying back seating issues, the likelihood of 1-RC-P-1 Band 1-RC-P-1C having the same condition is considered low.
RAJ Question No. 5 Page 5 of Attachment 2, Engineering Evaluation ETE-NA-2024-0033, states that " ... The interior of the casing, including the areas where all of the indications were found, does not participate in a measurable way in the hydraulic performance of the pump ... " The NRG staff understands that RCPs have vibration limits. Discuss whether excessive vibration at RCP 1-RC-P-1A were detected by the vibration sensor or abnormal occurrence as a result of loose parts during the plant operation.
Serial No. 24-142A Alternative Request N1-I5-NDE-007 RAI Response Attachment, Page 5 of 12 Dominion Energy Virginia Response to RAI Question No. 5 Based on a review of Operator Rounds Trend between 2010 and 2024, the vibration amplitudes on 1-RC-P-1A have been in the acceptable range. Reactor Coolant Pump vibrations are recorded by Control Room Operators daily. The 1-RC-P-1A vibrations did not reach the Alert level of 15 mils at any time since 2010. If the vibration levels reach Alert levels, Operators would submit a condition report for further investigation. The Operators are directed to trip the affected reactor coolant pump if vibrations exceed the Danger level of 20 mils. An alarm on the Main Control Room Annunciator Panel would alert the Operators of both the Alert and Danger vibration levels. No loose parts monitoring alarms were received as a result of this condition due to the physical separation of the sensors and the location of the RCPs.
RAJ Question No. 6 Page 5 of Attachment 2, states that " .. .if all of the adapter socket head cap screws had failed, significant operation degradation of the reactor coolant pump would not have resulted ... A loose adapter would initially drop loop flow about 0.2% which is much less than the existing flow margin of approximately 5% above core thermal design flow. In addition, the automatic low flow reactor trip would prevent operation below core thermal design flow ... "
The NRG staff understands that a loose adapter would reduce the reactor coolant system loop flow. If the loop flow is lower than the core thermal design flow, the reactor would trip which would cause the RCP to trip. As such, there is an automatic low flow reactor trip sensor to protect the reactor operation. Also, there is a RCP vibration limit to trip RCPs. Discuss any other defense-in- depth measures that would trip the RCPs to protect RCPs, the reactor vessel, and reactor coolant system components.
Dominion Energy Virginia Response to RAI Question No. 6 In addition to automatic reactor protection system features, the operators have the following RCP trip criteria:
RCP Seal Trip Criteria:
- Any Seal Stage p less than 25 psid
- RCS pressure less than 240 psig Other RCP Trip Criteria
- Hot and Cold Leg Isolation Valves for RCP being started open in coincidence with RCP loop flows NOT increasing within 30 seconds after closing breaker
- RCP starting current NOT decreasing within 30 seconds after breaker closure
- RCP proximity vibration greater than 20 mils (Alert 15 mils)
- RCP seismic vibration greater than 5 mils (Alert 3 mils)
Serial No. 24-142A Alternative Request N1-I5-NDE-007 RAI Response Attachment, Page 6 of 12
- RCP Motor Bearing temperatures greater than 195 °F
- RCP Lower Seal Water Bearing (Pump Bearing) temperature greater than 225 °F
- RCP Stator Winding temperature greater than 300 °F
- Loss of Seal Injection AND Component Cooling to RCP Thermal Barrier Based on the operating history of the NAPS reactor coolant pumps and industry OE related to dropped diffuser adapters, it is unlikely that any of these criteria would be met with a dropped diffuser adapter.
RAJ Question No. 7 Page 8 of Attachment 2, second paragraph discusses a Leak Before Break (LBB) analysis of the pump casing in WCAP-11517. The licensee demonstrated that a postulated through-wall crack in the pump casing remains stable based on a detailed fracture mechanics evaluation and that a gross failure of the casing is not feasible.
With respect to that, discuss the following:
(1) The size of the through wall crack based on applied loading; (2) The critical crack size of the pump casing; (3) The leakage crack size if different from the through wall crack size; (4) The reactor coolant system leakage detection capabilities (e.g., what is the smallest leak rate the sensors can detect); and (5)Whether the fracture mechanics analysis has satisfied (1) the margin of 2 between the crack size and the critical crack size and (2) margin of 10 for the leakage crack size as compared to the leak rate of the leakage detection system as specified in NRC's Standard Review Plan 3.6.3.
Dominion Energy Virginia Response to RAI Question No. 7 1, 3, & 4. Dominion did not perform a new detailed fracture mechanics evaluation to assess leak-before-break (LBB) of the pump casing. The analysis of the pum p casing indents relies on WCAP-13045 and subsequent evaluations to demo nstrate acceptability for 80-year plant life, as documented in RAl-8 response. Reference to WCAP-11517 was intended to show additional defense-in-depth due to the robust design and materials of the pump casing. The following discussion is extracted from WCAP-11517 without including any proprietary information.
The through-wall crack size is the leakage crack size required to result in 10 gpm, which is 10 times the leak detection capabilities of 1 gpm for the Westi nghouse Design NSSS systems that is appl icable to North Anna Units 1 & 2. LBB paths 2, 3
& 6 are shown in WCAP-11517, Figure 12-1 and are similar areas to those evaluated to Code Case N-481 in WCAP-13045 (N-481 paths 1, 2 and 4) for the Model 93A pump casings.
Serial No. 24-142A Alternative Request N1-I5-NDE-007 RAI Response Attachment, Page 7 of 12 The leakage crack sizes determined in WCAP-11517 for LBB paths 2, 3 and 6 are shown in Table 12-4. All the reported through-wall leakage crack lengths far exceed the extent of the non-through wall 1 /4 thickness flaws that were demonstrated to be acceptable in WCAP-13045, with structural integrity being maintained. It is noted that circumferential cracks in the inlet and outlet piping to nozzle locations (WCAP-11517 LBB paths 1 and 4) were controlling compared to all locations in the pump casing evaluated in WCAP-11517 for LBB, including LBB paths 2, 3, and 6.
2.WCAP-11517 did not calculate a critical flaw size. However, based on a review of Table 12-5 from WCAP-11517, the results for the faulted loading condition with a flaw length of twice the leakage flaw length at 10 gpm meet the stability criteria for all assumed paths with significant margins.
5.Based on a review of the above responses, it is concluded that the margin of 2 on the critical flaw size is met given that significant margin exists for through-wall flaws that are twice as long as the calculated 10 gpm leakage flaw sizes. A factor of 10 exists between the leakage flaw size and the minimum leak detection rate for North Anna Units 1 & 2, per the requirements of Regulatory Guide 1.45, Guidance on Monitoring and Responding to Reactor Coolant System Leakage.
RAJ Question No. 8 Page 8 of Attachment 2 discusses WCAP-13045 for the pump casing analysis. The NRG staff noted that WCAP-13045 is valid only for 40-year, not BO-year of plant life.
The licensee also referenced WCAP-15555 which applies to the pump casing analysis involving the initial license renewal period up to 60 years of plant life. The licensee has applied to renew its operating licensee up to 80 years for North Anna units. The NRG staff noted that it has approved generic use of topical report PWROG-17033, Revision 1, "Update for Subsequent License Renewal: WCAP-13045, 'Compliance to ASME Code Case N-481 of the Primary Loop Pump Casings of Westinghouse Type Nuclear Steam Supply Systems,"' for the application of subsequent license renewal up to 80 years (ML19319A188). Based on the licensee's conclusion on page 13 of Attachment 2, it appears that 1-RC-P-1 A pump casing will not be replaced or repaired for the remaining life of the plant.
(1) Demonstrate by analysis and/or inspection that the degraded casing will maintain its structural integrity to the end of BO-year plant life, i.e., the degraded areas of 1-RC-P-1 A pump casing will not grow to challenge the structural integrity of the pump casing.
(2) Discuss whether RCP 1-RC-P-1 A is acceptable for use for the BO-year plant life per PWROG-17033, Revision 1.
Serial No. 24-142A Alternative Request N1-I5-NDE-007 RAI Response Attachment, Page 8 of 12 Dominion Energy Virginia Response to RAI Question No. 8 WCAP-13045 is a generic evaluation performed by Westinghouse for the industry to provide the initial basis for elimination of volumetric examination requirements for RCP casings. The NRC has accepted this evaluation for North Anna Power Station Units 1 &
2 by approval of WCAP-13126 in 1991 for the initial 40-Year License, and in WCAP-15555 in 2000, as part of the 60-Year License Renewal Application for North Anna Power Stations Units 1 & 2. Although the BO-Year License application has not been approved for North Anna at the time of this alternative request, all RAls associated with the North Anna 80-Year SLR application regarding the updated Code Case N-481 evaluation, as documented in WCAP-18503 for North Anna Units 1 & 2, have been resolved and accepted by the NRC. Note that the work performed in WCAP-18503 is plant specific and applies the methodology specified in WCAP-13045 and PWROG-17033.
As stated in ETE-NA-2024-0033 the area of the degraded casing that was evaluated has no observable cracking. Based on the potential that a one-time load application was the cause of the observed pump casing surface degradation, Dominion performed the conservative structural evaluation prov ided in ETE-NA-2024-0033 as a bounding evaluation. The most likely cause of the degradation has since been postulated to be a fretting mechanism, which would not have the potential to introduce surface or subsurface cracking that a one-time load application could produce. The evaluation performed in ETE-NA-2024-0033 concludes that if a one-time load application was the cause of the noted pump casing inner surface damage, it would be bounding compared to fretting damage. ETE-NA-2024-0033 also concludes that any surface or subsurface micro-cracking in the pump casing that could result from a one-time load application would not have resulted in cracks that exceed the initial crack depth assumption of 0.3" which was used in the bounding fatigue and fracture mechanics analysis performed in WCAP-13045. In addition, a conservative fatigue strength reduction factor of 4.0 can be applied to the surface of the pump casing on 1-RC-P-1A in the noted areas of degradation. Based on the hoop stress levels of 12,500 psi reported in Fig. 8-12 of WCAP-13045 for this area of the pump casing, the maximum equivalent stress is (4.0)(12,500 psi) = 50,000 psi. This remains less than the maximum stress level for path 4 at the pump outlet nozzle (Ref. Fig. 8-16 of WCAP-13045 for the maximum hoop stress levels at the path 4 location shown in Fig. 9-2), which is the limiting location evaluated in WCAP-13045, WCAP-17555, and WCAP-18503. The stress level predicted by use of a conservative fatigue strength reduction factor of 4.0 is slightly less than stress levels at the discharge nozzle evaluation path 4, which is the limiting location in the fatigue and fracture evaluation performed in WCAP-13045, WCAP-15555, and WCAP-18503. However, the conservatively estimated increase in stress in the 1-RC-P-1A pump casing is only for the localized area at the damaged surface. The global through-thickness stress levels which impact the fatigue crack growth rates and dominate the fracture evaluation remain significantly lower in this area of the pump casing, through the applicable 1/4 thickness (for at least 1/2 of the casing thickness), as Serial No. 24-142A Alternative Request N1-I5-NDE-007 RAI Response Attachment, Page 9 of 12 compared to the bounding axial crack location 4 that was shown to maintain struc tural integrity for an 80-year service life.
Although the global axial stress levels at postulated circumferential crack path 2 evaluated in WCAP-13045 are only slightly higher than the global stresses in the casing wall near the area of degradation being evaluated for 1-RC-P-1A, the inner surface ledge where the actual damage has occurred on the 1-RC-P-1A pump casing experiences global compressive stresses in the axial direction (Ref. Fig. 8-11 of WCAP-13045). Thus, even if small areas of local tensile stress exist at or near the surface of the damaged areas of the 1-RC-P-1A pump casing, the overall crack driving force would be inadequate to propagate any small flaws into the area of global tensile stress.
Therefore, Dominion concluded that the area of local damage on the inner pump casing surface on 1-RC-P-1A is bounded by the evaluation performed in WCAP-13045 for path 2, which assumes a circumferential flaw orientation, and path 4, which assum es an axial flaw orientation.
Therefore, it is the position of Dominion that the observed damage identified on the internal surface of the pump casing for 1-RC-P-1A at North Anna Unit 1 remains bounded by the fatigue crack growth and fracture mechanics evaluations performed in WCAP-13045 and reconciled for North Anna pump casings in WCAP-13126 (for the initial 40-year service life), in WCAP-15555 (for the current 60-year operating license period), and in WCAP-18503 (for an 80-year service life). WCAP-18503 reconciled the updated analysis performed in PWROG-17033, which extended the N-481 pump casing evaluations to 80-years and was approved by the NRC, for North Anna Units 1 & 2.
Thus, any postulated cracks in the areas found to have a degraded surface condition on North Anna Unit 1 1-RC-P-1A Reactor Coolant Pump casing will not grow to a size that would challenge the structural integrity of the pump casing, out to a service life of 80 years.
It is also noted that the analysis performed for North Anna Units 1 & 2 in WCAP-15555 for 60-years of operation and in WCAP-18503 for 80-Years of operation were refined for North Anna specific loading and operati onal cycles, compared to the analysis that was generically performed for the industry in WCAP-13045. The results of the refined analysis performed in WCAP-15555 and WCAP-18503 determined that J,c values for North Anna Units 1 & 2 exceed the J,c values reported in WCAP-13045. WCAP-13045 documented a generically bounding J,c of 750 in-lbf/in2. However, reanalysis performed in WCAP-15555 determined that for North Anna Power Station Units 1 & 2 the actual J,c value is at least 958 in-lbf/in2. The N-481 analysis performed in WCAP-15555 was extended to 80-years of service for the North Anna Units 1 & 2 SLR application, documented in WCAP-18503, as no loading conditions or transients have changed, and the 60-year cycles were not exceeded for the 80-Year projections. The results documented in WCAP-18503, for an 80-year service life of North Anna Units 1 & 2 has Serial No. 24-142A Alternative Request N 1-I5-NDE-007 RAI Response Attachment, Page 10 of 12 been accepted by the NRC as previousl y noted and remain bounding for all North Anna Unit 1 & 2 Reactor Coolant Pump casings.
RAJ Question No. 9 Page 10 of Attachment 2 states that the limiting material fracture toughness, JIG= 750 in-/bflin2. With respect to that:
(1) Discuss the Japplied value based on the actual crack driving force.
(2) Discuss whether the JIG= 750 in-lbflin2 is applicable to 80 years.
Dominion Energy Virginia Response to RAI Question No. 9 1.The local stress levels in the area of the surface damage on the 1-RC-P-1 A pump casing were estimated using a conservative fatigue strength reduction factor of 4.0, as discussed in the response to RAl-8. These estimated local stresses and the global hoop stress levels in this area of the pump casing reported in WCAP-13045 (Ref. Figures 8-12 and 8-14) were compared to stress levels reported in WCAP-13045 for the limiting location analyzed for the 93A pump casings (path 4). Based on this comparison of local and through-thickness global stress levels and the applied stress intensity, Japplied (crack driving force) in the area of noted internal surface degradation on the 1-RC-P-1A casing can be concluded to remain well below the bounding applied stress intensity calculated in the generic analysis performed in WCAP-13045 and PWROG-17033. In addition, refined analysis performed in support of the approved 60-Year License Renewal Application (WCAP-15555) and for the 80-Year SLR application for North Anna Units 1 & 2 (WCAP-18503) determined that the actual Fracture Toughness of the North Anna RCP pump casings exceeds the Fracture Toughness reported in the generic N-481 evaluations (WCAP-13045) performed for Westing house Model 93A pump casings. The additional margin in Fracture Toughness is at least (958 in-lbf/in2 / 750 in-lbf/in2) =
1.277 (27.7%), and the Tearing Modulus was also shown to have a 5% increase in margin for the North Anna pump casings compared to the generic value reported in WCAP-13045.
2.See response for RAl-9 (1 ). The reported value of Fracture Toughness, J1c = 750 in lbf/in2 in WCAP-13045 is a bounding value used by Westinghouse for the initial generic industry Code Case N-481 evaluation. The North Anna Specific value of J1c is at least 958 in-lbf/in2 and applies through the extended operating period of 80 years.
RAJ Question No. 10 Pages 11 and 12 of Attachment 2 discuss fracture mechanics analysis of the impression/dent on the pump casing wall. The NRG staff noted that the licensee's Serial No. 24-142A Alternative Request N 1-I5-N DE-007 RAI Response Attachment, Page 11 of 12 analysis does not address the possibility of growth of the impression/ dent. Discuss whether the impression areas would grow in depth and width to a size to challenge the structural integrity of the pump casing at the end of 80 years of operation.
Dominion Energy Virginia Response to RAI Question No. 10 The suspected cause of the surface damage seen on the casing of North Anna Reactor Coolant Pump 1-RC-P-1A is fretting wear. Fretting wear does not have the potential to increase once the source of the fretting is removed, which has been done for 1-RC-P-1A. The rate of degradation due to fretting is also noted to be very low, and not a structural concern for one cycle of operation. Additionally, fretting wear would not generate cracking, but can create the irreg ular surfaces found in the recent pump casing inspections of 1-RC-P-1A. As noted in response to RAl-8 above, even considering a conservative fatigue strength reduction factor of 4.0 for this location of the pump casing, the stress levels remain below the stress levels of the bounding location of the Model 93A pump casings that was shown to maintain structural integrity out to 80 years of service in WCAP-18503. A flaw size of 0.3" was postulated in the generic Code Case N-481 analysis performed in WCAP-13045, which is conservative for North Anna Units 1 & 2 Reactor Coolant Pump casings. The conservative evaluation performed in ETE-NA-2024-0033 also dem onstrates that even for an extreme condition, such as a large onetime applied static load that could potent ially result in surface or subsurface micro-cracking of the indented areas, the expected crack size generated by such a load excursion would not exceed the flaw size of 0 .3".
It is also noted that the original and replacement cap screw materials do not pose a concern for galvanic corrosion and the pump casing material (Austenitic Stainless Steel) is not susceptible to general corrosion. Thus, there are no corrosion related degradation mechanisms that could result in additional degradation of the pump casing.
The evaluation performed in ETE-NA -2024-033 makes very conservative assumptions regarding the depth of the surface damage in the pump casing and is also bounding for fretting damage which is the suspected cause of the surface damage observed on the pump casing of 1-RC-P-1A. Thus, the evaluation is bounding for all North Anna Reactor Unit 1 Coolant Pump casings for damage due to failure of the diffuser adaptor cap screws.
RA/ Question No. 11 Discuss Dominion Energy efforts to address the potential for common-cause failure between pumps in terms of hardware (e.g., procurement material batches, dedication, etc.) and work practices (e.g., procedures, training, common work crews, quality control, etc.). Please include whether there is any additional information or observations of RCPs 1 B and 1 C that can inform whether a similar condition exists in those pumps.
Serial No. 24-142A Alternative Request N1-I5-NDE-007 RAI Response Attachment, Page 12 of 12 Dominion Energy Virginia Response to RAI Question No. 11 The cause evaluation for the DACS failures on 1-RC-P-1A is in the incipient stages. The DACS that were installed in 1982 were installed on all 3 pumps with procedures, guidance, and parts recommended and provided by the pump OEM (Westinghouse).
Dominion's review of historical information indicates this issue appears to be unique to 1-RC-P-1A. In 1982, all 3 pumps were removed and 1-RC-P-1A was the only pump with failed DACS. Additionally, a review of Condition Report history shows that only 1-RC-P-1A has had issues being put onto the backseat in 2012 and had issues with the transition off of backseat in 2013. Issues with back-seating the 93A RCPs could be an indicator that there could be potential issues with the Diffuser Adapter per Westinghouse lnfoGram 95008A.
RAJ Question No. 12 The NRG staff noted what appears to be a minor typographic error on the first line on Pages 2 and 3 of the Attachment 1 which states "Proposed Alternative in Accordance with 10 CFR 50.55a(a)(3)(ii)." Please confirm whether "Proposed Alternative in Accordance with 10 CFR 50. 55a(z)(2)" was the intended citation.
Dominion Energy Virginia Response to RAI Question No. 12 Dominion Energy Virginia confirms that "Proposed Alternative in Accordance with 10 CFR 50.55a(z)(2)" was the intended citation.