ML24163A300
| ML24163A300 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 08/08/2024 |
| From: | Markley M Plant Licensing Branch II |
| To: | Carr E Virginia Electric & Power Co (VEPCO) |
| Miller G | |
| References | |
| EPID L-2024-LLR-0022 | |
| Download: ML24163A300 (12) | |
Text
August 8, 2024 Eric S. Carr President and Chief Nuclear Officer Innsbrook Technical Center 5000 Dominion Blvd.
Glen Allen, VA 23060-6711
SUBJECT:
NORTH ANNA POWER STATION, UNIT NO. 1 - RE: RELIEF REQUEST FOR PROPOSED ALTERNATIVE N1-I5-NDE-007 INSERVICE INSPECTION FOR REQUIRED REACTOR COOLANT PUMP CASING INSPECTION (EPID L-2024-LLR-0022)
Dear Eric Carr:
By letter dated March 22, 2024, as supplemented by letter dated March 25, 2024, Virginia Electric and Power Company (Dominion Energy Virginia), submitted Relief Request for proposed alternative N1-I5-NDE-007 for the North Anna Power Station (NAPS), Unit 1, to the U.S. Nuclear Regulatory Commission (NRC) for review and approval, pursuant to the requirements of Title 10 of the Code of Federal Regulations, Section 50.55a(z)(2).
Specifically, the proposed alternative would allow deferral of inspection of an additional reactor coolant pump from the spring 2024 refueling outage to the fall 2025 refueling outage. Dominion Energy Virginia submitted alternative request N1-I5-NDE-007 on the basis of hardship or unusual difficulty without a compensating increase in the level of quality and safety.
On March 26, 2024, the NRC staff verbally authorized the use of alternative request N1-I5-007 for NAPS, Unit 1. This letter provides the full written safety evaluation of the alternative request.
All other American Society of Mechanical Engineers Code,Section XI requirements for which relief was not specifically requested and approved remain applicable, including third-party review by the Authorized Nuclear Inservice Inspector.
If you have any questions, please contact the project manager, Ed Miller at 301-415-2481 or via email at Ed.Miller@nrc.gov.
Sincerely, Michael Markley, Chief Plant Licensing Branch 2-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-338
Enclosure:
Safety Evaluation cc: Listserv MICHAEL MARKLEY Digitally signed by MICHAEL MARKLEY Date: 2024.08.08 14:10:42 -04'00'
Enclosure SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION PROPOSED INSERVICE INSPECTION ALTERNATIVE N1-I5-NDE-007 VIRGINIA ELECTRIC AND POWER COMPANY NORTH ANNA POWER STATION, UNIT 1 DOCKET NO. 50-338
1.0 INTRODUCTION
By letter dated March 22, 2024 (Agencywide Documents and Access Management System (ADAMS) Accession No. ML24082A274), as supplemented by letter dated March 25, 2024 (ML24086A429), 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 (NAPS), Unit 1.
The proposed alternative would allow deferral of inspection of additional reactor coolant pump (RCP) from the spring 2024 refueling outage to the fall 2025 refueling outage. Pursuant to Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Section 10 CFR 50.55a(z)(2), the licensee submitted for U. S. Nuclear Regulatory Commission (NRC) review and approval Alternative Request N1-I5-NDE-007 on the basis that conformance with the ASME Code would constitute a hardship or unusual difficulty without a compensating increase in the level of quality and safety. The proposed alternative and the associated analysis, ETE-NA-2024-0033, are shown in Attachments 1 and 2 to the March 22, 2024, letter, respectively.
On March 26, 2024 (ML24087A010),1 the NRC staff verbally authorized the use of Alternative Request N1-I5-NDE-007 for North Anna, Unit 1, through restart from the fall 2025 refueling outage, before the start of Cycle 32.
2.0 REGULATORY EVALUATION
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, states, in part, that ASME Code Class 1, 2, and 3 components (including supports) must 1 In the email documenting the verbal approval, the ADAMS Accession No. for the March 25, 2024, supplement is identified as ML24086A089. Due to a document issue, the supplement was resubmitted to the Document Control Desk and, as a result, was given a new Accession No. of ML24086A429.
meet the requirements, except the design and access provisions and the preservice examination requirements, set forth in the Section XI of additions and addenda of the ASME BPV [Boiler and Pressure Vessel] Code 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 initial 120-month inspection interval [first 10-year ISI interval] and subsequent intervals must comply with the requirements in the latest edition and addenda of Section XI of the ASME 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 NRC, if (1) the proposed alternatives would provide an acceptable level of quality and 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 TECHNICAL EVALUATION
3.1 Licensee's Proposed Alternative
=
Applicable Code Edition and Addenda===
The applicable Code for the NAPS Unit 1 Fifth 10-Year ISI Interval and ISI Program is the ASME BPV,Section XI, 2013 Edition with no Addenda. The NAPS Unit 1 Fifth 10-Year ISI interval started May 1, 2019, and ends April 30, 2029.
ASME Code Components Affected The affected components are the NAPS, Unit, 1 RCP casings. NAPS, Unit 1, has three RCPs and their identification numbers are 1-RC-P-1A, 1-RC-P-1B, and 1-RC-P-1C (hereafter refer to as 1A RCP, 1B RCP, and 1C RCP).
The RCP casings are inspected under Examination Category B-L-2, Item Number B 12.20, in accordance with IWB-2500 and Table IWB-2500-1 of the ASME Code,Section XI.
ASME Code Requirement for Which Alternative Is Requested The provisions of the ASME Code,Section XI, IWB-2430, Additional Examinations, requires examinations that reveal flaws or relevant conditions exceeding the acceptance standards of Table IWB-3410-1 be extended to include additional examinations during the current outage.
Licensee's Proposed Alternative to the ASME Code In lieu of examining of casing of an additional RCP during the current 2024 refueling outage in accordance with IWB-2430(a)(1)(a) of the ASME Code,Section XI, the licensee proposed to perform the additional examination of an RCP casing during the next refueling outage. The licensee stated that the next scheduled pump replacement is currently planned for the next refueling outage N1R31 in fall 2025.
The licensee stated that the limiting depth of the indentations in the 1A RCP casing are conservatively estimated to exceed the limiting flaw depth of 0.3 inches in the acceptance
standards of the ASME Code,Section XI, Table 3519-2.2. The licensee evaluated the degradation in accordance with IWB-3142.4 to demonstrate acceptability for continued service as shown in Attachment 2 to the letter dated March 22, 2024.
The licensee stated that the 1A RCP evaluation in Attachment 2 can be applied to 1B and 1C RCPs. The licensee further stated that although the cause evaluation is not complete, the likely cause of the defects in the 1A RCP casing is cap screw fragment migration into the gap between the diffuser adaptor and casing and long-term fretting. The licensee explained that cap screw impression of half the width was a conservative assumption considering the gap dimensions and the lack of a large static load to cause debris. The only reasonable mechanism to cause a deeper flaw than analyzed is fretting, which does not involve plastic deformation. The licensee stated that it assumed a single, excessive static load that would initiate plastic deformation and cracking, which was still found to be acceptable. The licensee further stated that the bounding nature of the analysis covers potential conditions in all three RCPs that have similar age and operating time. The licensee concluded that this approach provides reasonable assurance that the other pump casings would have structural integrity, supporting the proposed extension of performing the additional examination to the next refueling outage.
The licensee requested to complete the additional examination of an RCP casing as required by IWB-2430 of the ASME Code,Section XI during the fall 2025 refueling outage, which is before the end of the current 5th ISI Interval, which ends on April 30, 2029.
Reason for Proposed Alternative The licensee requested an alternative from the requirements of IWB-2430 associated with relevant indications found during the VT-3 visual examination of the 1A RCP casing per Examination Category B-L-2, Item B12.20 of Table IWB-2500-1 of the ASME Code,Section XI.
The licensee examined the RCP casing on March 16, 2024, during NAPS, Unit 1, R30 refueling outage as part of a pump replacement activity. The licensee stated that this examination is only required when the pump is disassembled for maintenance or repair and is an examination of the internal pressure boundary including all pressure boundary surfaces made accessible for examination by disassembly. The licensee replaced 1A RCP with a refurbished pump as part of the NAPS Subsequent License Renewal Project. The licensee stated that the required VT-3 visual examination was performed on 1A RCP casing to identify (a) corrosion or erosion that reduces the pressure retaining wall thickness by more than 10 percent; (b) wear of mating surfaces that may lead to loss of function or leakage; or (c) crack-like surface flaws developed in service or grown in size beyond that recorded during preservice visual examination. The licensee stated that during the remote visual examination, it identified multiple instances of structural deformation near the bottom of the casing. The deformations appear to be impressions in the casing made by loose parts from the degraded cap screws used to attach the diffuser adapter to the turning vane diffuser. The licensee estimated the limiting depth of the indentations in the pump casing to exceed the limiting flaw depth of 0.3 inches in the acceptance standards of ASME Section XI, Table 3519-2.2.
Duration of Proposed Alternative The licensee requested to complete the additional examination of an RCP casing as required by IWB-2430 of the ASME Code,Section XI during the fall 2025 refueling outage, which is before the end of the current 5th ISI Interval, which ends on April 30, 2029.
3.2
NRC Staff Evaluation
The NRC staff evaluated the proposed alternative to confirm that (1) the extension of one fuel cycle to examine the designated RCP in fall 2025 will not affect its operability, (2) the degraded 1A RCP casing is acceptable for continued service to the end of 80 years as the licensee does not plan to replace the degraded casing, (3) NAPS, Unit 1, has defense-in-depth measures to stop the pump operation in case the pumps degrade during normal operation, and (4) the hardship is appropriately justified.
As discussed below, the NRC staff considers the following key topics: affected components, degradation characterization, operating experience, structural integrity analysis, defense-in-depth measures, and hardship. In the March 25, 2024, supplement, the licensee provided additional information to address the NRC staffs request for additional information (RAI) regarding these topics.
3.2.1 Affected Components The next scheduled pump replacement and examination is currently planned for the next refueling outage, N1R31, fall 2025. In response to RAI Nos 1 and 2, the licensee clarified that the 1B RCP is scheduled to be inspected/refurbished during the fall 2025 refueling outage and the additional pump casing examination will be completed before Cycle 32 startup. The licensee stated that if there are unforeseen circumstances during the next operating cycle, the 1C RCP may be inspected/refurbished in the fall 2025 refueling outage in lieu of the 1B RCP.
The licensee indicated that the initial inspection/refurbishment scope involves a decontamination cycle, disassembly, cleaning, and inspection on a removed (used) RCP. The 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. The licensee clarified that pump casings are not part of the pump refurbishment project and are not planned to be replaced. Based on the above, the NRC staff finds that the licensee has sufficiently described the specific RCP that will be inspected in the 2025 refueling outage and the scope of the inspection.
3.2.2 Degradation Characterization The NRC staff requested additional information regarding the possibility of growth of the impression/dent. In its response to RAI No. 10, the licensee stated that the suspected cause of the surface damage seen on the 1A RCP casing is fretting wear. The licensee stated that fretting wear does not have the potential to increase once the source of the fretting is removed, which has been done for the 1A RCP. The licensee further stated that the rate of degradation due to fretting is also noted to be very low, and not a structural concern for one cycle of operation. The licensee indicated that fretting wear would not generate cracking but can create the irregular surfaces in the 1A RCP casing. The licensee stated that its evaluation also shows that even for an extreme condition, such as a large one-time applied static load that could potentially 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 inches postulated in the licensees structural analysis, which will be discussed further in this safety evaluation (SE).
The NRC staff requested additional information about 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.). In its response to RAI No. 11, the licensee stated that the root cause evaluation for the cap screw failures on 1A RCP is in the incipient stages. The cap screws that were installed in 1982 were installed on all 3 pumps with procedures, guidance, and parts recommended and provided by Westinghouse, the pump manufacturer. The licensee further stated that its review of historical information indicates that this issue appears to be unique to the 1A RCP only. In 1982, the licensee removed all 3 pumps for refurbishment and the 1A RCP was the only one with failed cap screws. The licensee further stated that the Condition Report history shows that only 1A pump has had issues being put onto the backseat in 2012 and had issues with the transition off of backseat in 2013. The licensee explained that issues with back-seating the 93A RCPs could be an indicator that there could be potential issues with the diffuser adapter.
The licensee stated that back-seating of the RCP 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. It provides the reactor coolant system maintenance boundary for reactor coolant pump seal replacement activities. Based on the lack of backseat issues on the 1B and 1C RCPs, the NRC staff finds that the diffuser adapter cap screws are not likely to be similarly loose in the 1B and 1C RCPs.
3.2.3 Other Degradation Mechanisms The licensee stated that the original and replacement cap screw materials do not pose a concern for galvanic corrosion and the pump casing material of 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 licensee explained that its evaluation performed in ETE-NA-2024-033 assumes conservative 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 1A RCP casing. The licensee stated that its evaluation is bounding for all Unit 1 RCP casings for damage due to failure of the diffuser adaptor cap screws.
3.2.4 Structural Integrity Analysis To demonstrate structural integrity of a component, licensees can perform a fracture mechanics analysis to show that a degraded component with a flaw is acceptable for the remainder of the plant license duration. In addition to such an analysis, licensees can also perform a flaw tolerance evaluation such as a leak-before-break (LBB) analysis showing that the component with a postulated 100 percent through wall crack will not lead to catastrophic failure. The licensee performed a leak-before-break assessment and a fracture mechanics analysis for the RCP at North Anna as discussed below.
3.2.4.1 Leak-Before-Break Assessment The licensees analysis ETE-NA-2024-0033 discusses a flaw evaluation of the pump casing based on a LBB analysis. The licensee demonstrated that a postulated through-wall crack in the pump casing remains stable based on a detailed fracture mechanics methodology and that a gross failure of the casing is not feasible. In response to RAI No.7, the licensee stated that its LBB analysis was intended to show additional defense-in-depth due to the robust design and materials of the pump casing. The licensee stated that the through-wall crack size is the leakage
crack size required to result in 10 gallons per minute (gpm), which is 10 times the leak detection capabilities of 1 gpm for North Anna, Units 1 and 2. The licensee described selected parts of WCAP-13045 that demonstrated that all the through-wall leakage crack lengths far exceed the extent of the non-through wall 1/4 thickness flaws, with structural integrity being maintained. The licensee stated that circumferential cracks in the inlet and outlet piping to pump nozzle locations were controlling compared to all locations in the pump casing evaluated. The licensee stated 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. The NRC staff notes that a factor of 10 exists between the leakage flaw size and the minimum leak detection rate for North Anna, Units 1 and 2, per the guidance of Regulatory Guide 1.45, Guidance on Monitoring and Responding to Reactor Coolant System Leakage, (ML073200271).
The NRC staff determined that the 1A pump casing does satisfy the margins for the leakage crack size as compared to the critical crack size and to the leak detection rate satisfy the margins in Standard Review Plan 3.6.3, Leak Before Break Evaluation Procedures, (ML063600396) of NUREG-0800, Standard Review Plan for the Review of Safety Analysis Report for Nuclear Power Plants: LWR Edition. Therefore, the NRC staff finds that the LBB assessment demonstrates that, if a 100 percent through wall crack exists in a component, the leakage from the crack will alert the operator to take timely corrective action prior to a catastrophic failure.
3.2.4.2 Fracture Mechanics Analysis The structural integrity of the pump casing is maintained when its material fracture toughness value, JIC, is more than the driving stresses from applied loading represented by Japplied, such that the flaw will not propagate uncontrollably. The flaw may grow but slowly and in a stable fashion.
The crack propagation in the 1A pump is based on the fatigue degradation mechanism (fatigue crack growth). The NRC staff notes that the material fracture toughness may reduced because of embrittlement as the material ages. As such, the NRC staff evaluated whether the licensees material fracture toughness is acceptable for 80 years of plant life.
The licensees analysis ETE-NA-2024-0033 references WCAP-13045 as part of 1A pump analysis. The NRC staff notes that WCAP-13045 is valid only for 40-year, not 80-year of plant life. In response to RAI No. 8, the licensee noted that NRC staff 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).
The licensee stated that the NAPS initial and subsequent license renewal applications provide plant-specific RCP casing analysis showing that the observed damage identified on the internal surface of the 1A pump casing remains bounded by the fatigue crack growth and fracture mechanics evaluations performed in WCAP-13045, which has been reconciled with the updated analysis performed in PWROG-17033. Section 4.7.6 of NAPSs SLRA references PWROG-17033 for the RCP analysis. Thus, any postulated cracks in the areas found to have a degraded surface condition on 1A RCP 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.
For the analysis of the 1A pump, in response to RAI No. 8, the licensee stated that the area of the degraded 1A RCP casing has no observable cracking. The licensee explained that 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 licensee stated 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. The licensee also concluded 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 inches that was used in the bounding fatigue and fracture mechanics analysis performed in WCAP-13045.
The licensee stated that it applied a fatigue strength reduction factor of 4.0 to the surface of the 1A pump casing in the areas of degradation to show that even with a reduction in strength the integrity of the pump casing will still be acceptable. The licensee stated that based on the hoop stress levels of 12,500 pounds per square inch (psi) reported in Figure 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.
The licensee stated that this stress remains less than the maximum stress level for path 4 as shown in Figure 9-2 of WCAP-13045 at the pump outlet nozzle, which is the limiting location evaluated in WCAP-13045 and in North Annas initial and subsequent license renewal applications. The licensee explained that the estimated increase in stress in the 1A pump casing is only for the localized area at the damaged surface. The licensee stated that 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 pump casing thickness), as compared to the bounding axial crack location that was shown to maintain structural integrity for an 80-year service life.
The licensee stated that the global axial stress levels at the postulated circumferential crack evaluated in WCAP-13045 are only slightly higher than the global stresses in the casing wall near the 1A pump degraded area. As such, the inner surface ledge where the actual damage has occurred on the 1A pump casing experiences global compressive stresses in the axial direction. The licensee explained that even if small areas of local tensile stress exist at or near the surface of the damaged areas of the 1A pump casing, the overall crack driving force would be inadequate to propagate any small flaws into the area of global tensile stress. Therefore, the area of local damage on the inner 1A pump casing surface is bounded by the limiting circumferential and axial flaws of WCAP-13045.
The NRC staff finds that based on the comparison of local and through-thickness global stress levels and the applied crack driving force, Japplied around degradation on the 1A pump casing remains below the bounding applied driving force calculated in the generic analysis performed in WCAP-13045 and PWROG-17033. The NRC staff notes that because the limiting flaws in the generic analyses are acceptable for 80 years as shown in PRWOG-17033 and the 1A RCP casings Japplied is lower than Japplied of the limiting flaws in the generic analyses, then 1A RCP is acceptable for 80 years of operation.
In addition to the fracture mechanics analysis discussed above, the licensee further stated that the actual material fracture toughness of the NAPS RCP casings exceeds the material fracture toughness in WCAP-13045 performed for Westinghouse Model 93A pump casings. The licensee stated that the material fracture toughness 958 in-lbf/in2 for North Anna RCP casings is 27.7% higher than the generic value 750 in-lbf/in2 (958 in-lbf/in2 / 750 in-lbf/in2 = 1.277). The licensee further stated that the tearing modulus (which is used to predict the onset of flaw growth instability) for the NAPS RCP casings was also shown to have a 5% increase in margin as compared to the generic value reported in WCAP-13045. The NRC staff finds that, based on
tearing modulus and material fracture toughness, NAPS RCP casing material has sufficient resistance to limit the instable flaw growth.
Based on the above, the NRC staff determined that the structural integrity of 1A RCP casing will be maintained to the end of 80 years because (1) the plant-specific material fracture toughness of 1A RCP casing is higher than the generic value used in WCAP-13045, (2) the stresses at the degraded area will not likely to challenge the pump integrity, (3) the degradation is caused by fretting such that the crack will not likely occur, and (4) even if occurring in the degraded area, crack(s) will not likely to propagate significantly to challenge the pump integrity.
The NRC staff further determined that (1) although the licensees fracture mechanics analysis was performed for the 1A RCP casing, the results are applicable to the 1B and 1C RCPs, and (2) the RCP analyses in Section 4.7.6 of the North Anna initial and subsequent license renewal applications are not significantly affected by the indented surface at the 1A RCP casing.
3.2.5 Defense-in-Depth Measures The proposed alternative states that if all the adapter cap screws had failed, significant operation degradation of the RCP would not have resulted. The proposed alternative further states that a loose adapter would initially drop reactor coolant system loop flow about 0.2 percent, which is much less than the existing flow margin of approximately 5 percent above core thermal design flow. In addition, the automatic low flow reactor trip would prevent operation below core thermal design flow. In RAI No. 6, the NRC staff questioned whether other defense-in-depth measures could trip the RCPs to protect RCPs, the reactor vessel, and reactor coolant system components. In response, the licensee stated that in addition to automatic reactor protection system features, the operators have the following two RCP trip criteria (1) any seal stage p less than 25 pounds per square inch differential, or (2) reactor coolant system pressure less than 240 pounds per square inch guage. The licensee stated that other RCP trip criteria include timing of hot and cold leg isolation valves for RCP, limits on pump starting current, vibration limits, temperature limits on bearing and pump stator winding, and limits on loss of seal injection and component cooling to RCP thermal barrier. Therefore, the NRC staff finds that RCP trip setpoints and various sensors have provided adequate defense-in-depth measures to protect the RCPs, reactor vessel and reactor coolant system and therefore are acceptable.
3.2.6 Hardship The licensee stated that the process for complete disassembly and removal of an RCP to allow complete inspection of the pump casing presents a hardship without a compensating increase in quality and safety because of significant increase in outage duration, dose, effects on critical equipment, and challenges to nuclear and personnel safety. Based on the previous RCP pump refurbishment projects (Unit 2 1A pump in fall 2023, Unit 1, 1A pump in spring 2024), the expected dose associated with this evolution averages 5.5 rem [Roentgen Equivalent Man]. The licensee stated that the removal of an RCP would also require the use of critical spare parts such as seal faces (currently in an industry-wide parts shortage due to closure of primary manufacturing facility), seal package soft goods, and pump main flange gaskets.
Based on the dose and equipment issues discussed above, the NRC staff finds that the licensee has provided a reasonable justification for the hardship that would be incurred if required to perform inspection of an additional RCP casing in the spring 2024 refueling outage.
As such, the NRC staff finds that complying with the specified ASME Code requirements would
result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
3.3 Summary Based on the above evaluation, the NRC staff determined that (1) the structural integrity of 1A RCP casing will be maintained for the remaining life of the plant license based on fracture mechanics analysis and stress analysis, (2) the structural integrity of 1B RCP and 1C RCP will be maintained, (3) the defense-in-depth measures will protect the RCPs, reactor vessel and reactor coolant system, (4) the hardship that would be incurred if additional examination is performed during the spring 2024 refueling outage is reasonable, and (4) the risk of the degraded 1A RCP casing to the safe operation and public health and safety is low.
The NRC staff concludes that the licensee has provided reasonable assurance that the structural integrity of the three RCPs will be maintained for the remainder of the plant license.
4.0 CONCLUSION
As evaluated in this SE, the NRC staff finds that the proposed alternative in Alternative Request N1-I5-NDE-007 will provide reasonable assurance of the structural integrity of the 1A, 1B, and 1C RCP casings. The NRC staff determines that complying with the specified ASME Code requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. Accordingly, the NRC staff concludes that the licensee has adequately addressed all the regulatory requirements set forth in 10 CFR 50.55a(z)(2). Therefore, the NRC authorizes the use of Alternative Request N1-I5-NDE-007 at North Anna, Unit 1, through restart from the fall 2025 refueling outage, before the start of Cycle 32.
All other ASME BPV Code,Section XI, requirements for which an alternative was not specifically requested and authorized remain applicable, including third-party review by the Authorized Nuclear Inservice Inspector.
Principal Contributor: J. Tsao, NRR Date: August 8, 2024
- Via SE Input OFFICE NRR/DORL/LPL2-1/PM NRR/DORL/LPL2-1/LA NRR/DORL/LPL2-1/LA NAME GEMiller KGoldstein (KEntz for)
KGoldstein DATE 06/11/2024 06/13/2024 08/08/2024 OFFICE NRR/DNRL/NVIB/BC NRR/DORL/LPL2-1/BC NRR/DORL/LPL2-1/PM NAME ABuford*
MMarkley GEMiller DATE 7/11/2024 08/06/2024 08/06/2024