ML20267A516
| ML20267A516 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 09/23/2020 |
| From: | Mahesh Chawla NRC/NRR/DORL/LPL4 |
| To: | Ronnie Garcia Energy Northwest |
| References | |
| L-2020-LLR-0068 | |
| Download: ML20267A516 (11) | |
Text
From: Chawla, Mahesh Sent: Wednesday, September 23, 2020 12:30 PM To: Garcia, Richard M.
Cc: Tsao, John; Dijamco, David; Gonzalez, Hipo; Dixon-Herrity, Jennifer
Subject:
Columbia Generating Station - Final - Request for Additional Information -
Fourth Ten-Year Interval Inservice Inspection (ISI) Program Relief Request 4ISI EPID L-2020-LLR-0068 Attachments: Columbia RAI 9-23-2020_Final.docx
Dear Mr. Garcia,
By letter dated April 22, 2020, (Agencywide Documents and Access Management System (ADAMS)
Accession No. ML20114E234), Energy Northwest (the licensee) requested relief from certain requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, Table IWB-2500-1 for the inservice inspection (ISI) program at the Columbia Generating Station (CGS). Pursuant to 10 CFR 50.55a(z)(1), the licensee submitted Relief Request 4ISI-09 for the performance of alternate examinations of the reactor vessel feedwater nozzle-to-shell welds and feedwater nozzle inner radii on the basis that the proposed alternative would provide an acceptable level of quality and safety.
On September 9, 2020, we transmitted the draft request for additional information (RAI), which was discussed in a clarification call with your staff on September 22, 2020. Based on the discussion during yesterdays call, the NRC staff has revised the RAI which is attached here as the final RAI. During the call you agreed to provide your response on the docket within 30 days of the receipt of this email. We appreciated your participation in the call and look forward to receiving your supplemental information.
Thanking you, Sincerely, Mahesh Chawla, Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission ph: 301-415-8371 ODORL/LPL4/PM DORL/LPL4/BC JDixon-NMChawla Herrity D9/23/2020 9/23/2020
Hearing Identifier: NRR_DRMA Email Number: 801 Mail Envelope Properties (DM8PR09MB6853943451DE34435FA73681F1380)
Subject:
Columbia Generating Station - Final - Request for Additional Information - Fourth Ten-Year Interval Inservice Inspection (ISI) Program Relief Request 4ISI EPID L-2020-LLR-0068 Sent Date: 9/23/2020 12:30:01 PM Received Date: 9/23/2020 12:30:01 PM From: Chawla, Mahesh Created By: Mahesh.Chawla@nrc.gov Recipients:
"Tsao, John" <John.Tsao@nrc.gov>
Tracking Status: None "Dijamco, David" <David.Dijamco@nrc.gov>
Tracking Status: None "Gonzalez, Hipo" <Hipolito.Gonzalez@nrc.gov>
Tracking Status: None "Dixon-Herrity, Jennifer" <Jennifer.Dixon-Herrity@nrc.gov>
Tracking Status: None "Garcia, Richard M." <rmgarcia@energy-northwest.com>
Tracking Status: None Post Office: DM8PR09MB6853.namprd09.prod.outlook.com Files Size Date & Time MESSAGE 1597 9/23/2020 12:30:01 PM Columbia RAI 9-23-2020_Final.docx 50198 Options Priority: Normal Return Notification: No Reply Requested: No Sensitivity: Normal Expiration Date:
1 REQUEST FOR ADDITIONAL INFORMATION BY THE OFFICE NUCLEAR REGULATION RELIEF REQUEST 4ISI-09 ALTERNATE EXAMINATION OF REACTOR VESSEL FEEDWATER NOZZLES AND NOZZLE-TO-SHELL WELDS COLUMBIA GENERATING STATION ENERGY NORTHWEST DOCKET NO. 50-397 EPID L-2020-LLR-0068
Background
By letter dated April 22, 2020, (Agencywide Documents and Access Management System (ADAMS) Accession No. ML20114E234), Energy Northwest (the licensee) requested relief from certain requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, Table IWB-2500-1 for the inservice inspection (ISI) program at the Columbia Generating Station (CGS).
Pursuant to 10 CFR 50.55a(z)(1), the licensee submitted Relief Request 4ISI-09 for the performance of alternate examinations of the reactor vessel feedwater nozzle-to-shell welds and feedwater nozzle inner radii on the basis that the proposed alternative would provide an acceptable level of quality and safety.
Regulatory Basis Adherence to Section XI of the ASME Code is mandated by 10 CFR 50.55a(g)(4), which states, in part, that ASME Code Class 1, 2, and 3 components will meet the requirements, except the design and access provisions and the pre-service examination requirements, set forth in the ASME Code,Section XI.
The regulations in 10 CFR 50.55a(g) require that the ISI of ASME Code Class 1, 2, and 3 components be performed in accordance with Section XI of the ASME Code and applicable addenda. The ASME Code,Section XI, requires that all reactor vessel nozzles to be inspected during each 10-year ISI interval. The volumes in each nozzle required to be inspected are 100 percent of the nozzle-to-vessel shell weld volume and 100 percent of the nozzle inner radius section volume, as shown in the applicable figure in Figures IWB-2500-7(a) through (d) "Nozzle in Shell or Head," of the ASME Code,Section XI.
Request for Additional Information To complete its review, the Nuclear Regulatory Commission (NRC) requests the following additional information.
- 1. Inspection RAI 1.1 Issue:
The NRC staff noted that the relief request does not discuss the inspection strategy if an indication is detected in a feedwater nozzle or in a nozzle-to-shell weld in the future, whether an expansion (extent of condition) inspection will be performed.
2 Request:
Discuss the expansion (extent of condition) inspection if an indication is detected in a feedwater nozzle or in a nozzle-to-shell weld in the future. If no expansion inspection will be performed, provide justification.
RAI 1.2 Issue:
On Page 5 of the relief request, the licensee states that it reviewed the most recent examination results for the subject components and reported that no recordable indications in the feedwater nozzle inner radii or nozzle-to-shell welds. The licensee further stated that all the examinations had greater than 99% examination coverage. It is not clear to the NRC staff whether all six feedwater nozzles were examined, what examination method(s) were used, and whether the 99% examination coverage is applicable to all the six feedwater nozzles and associated welds.
Request:
a) Confirm that all six-feedwater nozzle inner radii and associated nozzle-to-shell welds were inspected in the most recent examination.
b) Discuss any other examination method used to inspect the subject components besides the ultrasonic testing.
c) Discuss whether the 99% examination coverage is applicable to all six feedwater nozzle radii and nozzle-to-shell welds.
- 2. Deterministic Stress Analysis RAI 2.1 Issue:
First paragraph on Page 5 of the relief request states that the cladding on the feedwater nozzle inner radius has been removed. However, Enclosure 2 to the licensees April 22, 2020 letter does not specifically mention that cladding is not modeled on the nozzle radius in the finite element analysis. Also, the relief request does not include a drawing to show location of the cladding on the reactor vessel shell that near the nozzle inner radius. The NRC staff notes that the location of cladding is significant because thermal expansion of cladding is different from that of the reactor vessel shell.
Request:
(a) Confirm that the finite element model in Enclosure 2, Figure 3 does not include cladding on the inner surfaces of feedwater nozzles and associated nozzle-to-shell welds. If there is no cladding on the nozzle-to-shell welds, justify that the clad stress due to the thermal expansion difference between the clad and the reactor vessel shell of the adjacent cladded inner surface shown in Figure 3 has negligible impact on stresses in the nozzle-to-shell welds.
(b) Provide a sketch to show the distance from the feedwater nozzle or nozzle-to-shell weld to the reactor vessel shell that has no cladding.
(c) Confirm that the finite element model of the feedwater nozzle radius and nozzle-to-shell weld is consistent with the actual field configuration.
3 RAI 2.2 Issue:
The reactor vessel nozzles analyzed in BWRVP-108-A, 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 Blend Radii, and BWRVIP-241-A, BWR Vessel and Internals Project Probabilistic Fracture Mechanics Evaluation for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii, are represented as a 360-degree nozzle configuration in the finite element model. However, the finite element model in , Figure 3 shows a quarter of the feedwater nozzle. The NRC staff noted that the azimuthal locations (i.e., 0, 90, 180, or 270 degrees) of a feedwater nozzle may experience different stresses. In addition, the finite element model in Figure 3 shows that the reactor shell on which the feedwater is attached is also a quarter-size panel.
Request:
a) Discuss the adequacy of the quarter size feedwater nozzle model in Enclosure 2, Figure 3 to generate the appropriate stress distributions as compared to the 360-degree full nozzle model.
b) Clarify how would the stresses extracted from the quarter-sized finite element model in Enclosure 2, Figure 3 represent the appropriate stresses to calculate the probability of failure when the full 360-degree nozzle is not represented in the finite element model.
c) Clarify whether the quarter sized reactor shell modeled in Figure 3 will provide accurate stress distribution in the feedwater nozzle radius region and nozzle-to-shell welds.
RAI 2.3 Issue: , Figure 4 is labeled as the applied pressure and boundary conditions. However, the NRC staff is not clear exactly what are the applied pressure and boundary conditions in Figure 4.
Request:
Clarify what are the pressure and boundary conditions that are applied to the finite element model as shown in Figure 4.
RAI 2.4 Issue:
Section 3.0 in Enclosure 2 describes applied loadings. However, it appears that the applied loading from the feedwater pipe imposed on the feedwater nozzle was not included in the finite element analysis. The NRC staff notes that the loading from the feedwater pipe may cause stresses on the feedwater nozzle and, therefore, should be considered in the stress analysis of the feedwater nozzle.
Request:
Discuss whether the forces and moments generated from the feedwater pipe are included in the stress analysis of the feedwater nozzle and nozzle-to-shell weld in the finite element analysis. If not, provide justification.
4 RAI 2.5 Issue: , Tables 1 to 4, provide transient definitions for various events. Enclosure 2 states that the thermal transient cycles were predicted for 60 years of operation and that it follows the methodology used in BWRVIP-108-A and BWRVIP-241. However, it is not clear to the NRC staff the exact source of the transient cycles and definitions in the stress analysis.
Request:
Discuss whether the thermal cycles and transients used in the stress analysis of the feedwater nozzle radius and nozzle-to-shell weld in Enclosure 2 come from the plant-specific licensing basis, or from the generic transients as shown in BWRVIP-108-A and BWRVIP-241. If the generic thermal cycles and transients in these BWRVIP reports were used, discuss whether they bound the plant-specific transient data at CGS.
RAI 2.6 Issue: , Section 3.1.2 states that only cyclic loads such as thermal transient and pressure are included in the stress analysis and that deadweight, which does not cycle, is not needed.
The NRC noted that non-cyclic loads such as deadweight and residual stress are still needed for cyclic fatigue crack growth because they raise the mean stress which affects the stress distribution in the nozzle and weld.
Request:
a) Discuss why deadweight load and residual stress are not included in the finite element model to calculate stresses.
b) Discuss why seismic loads are not included in the finite element stress analysis.
c) Discuss whether there are any other loadings applied to the feedwater nozzle and nozzle-to-shell welds besides the pressure load and thermal load in the stress analysis.
RAI 2.7 Issue:
NRCs safety evaluations for BWRVIP-108-A and BWRVIP-241 require that the maximum reactor vessel heatup/cooldown rate be limited to less than 115 ºF/hour.
Request:
Confirm that CGS will satisfy this condition.
RAI 2.8 Issue:
The NRC staff compared the thermal transient cycles used in the feedwater nozzle analysis as shown in Table 5 in Enclosure 2 to the thermal transient cycles used in the recirculation outlet nozzle at CGS as shown in Table 5-5 of BWRVIP-241. The NRC staff noted that the thermal transients in Table 5-5 of BWRVIP-241 are for 40 years of operation whereas the thermal transients in Enclosure 2 are for the 60-year plant life. The NRC further noted that Table 5 in does include more transient categories than that of Table 5-5 in BWRVIP-241.
Nevertheless, the NRC staff identified the following three discrepancies between Table 5 in and Table 5-5 in BWRVIP-241. (1) Table 5 in Enclosure 2 does not identify the
5 Scram transient. It does have a other scram category but with only 90 cycles. The Scram category in Table 5-5 in BWRVIP-241 indicates 180 cycles. (2) The natural recirculation startup transient shows 3 cycles in both Table 5 in Enclosure 2 and Table 5-5 in BWRVIP-241.
However, it seems that Table 5 should have more cycles than that of Table 5-5 because Table 5 is for the 60-year plant life whereas Table 5-5 is for 40 years. (3) Table 5-5 in BWRVIP-241 has a loss of feedwater pump transient whereas Table 5 in Enclosure 2 does not. In addition, the NRC staff notes that Section 6.1 of BWRVIP-05, BWR Vessel and Internals Project: BWR Reactor Pressure Vessel Shell Weld Inspection Recommendations (BWRVIP-05), states that loss of feedwater heaters will affect feedwater nozzles in terms of fatigue. Rapid cycling fatigue was found to occur as a result of mixing of relatively colder water with the hotter reactor water, which was addressed by modifications and design changes to the feedwater nozzles.
Request:
(a) Explain these three discrepancies between the thermal transient cycles in Table 5 in Enclosure 2 and Table 5-5 of BWRVIP-241.
(b) Clarify why Enclosure 2, Table 5 does not include the loss of feedwater heaters transient.
RAI 2.9 Issue:
Section 4.1.3 of BWRVIP-108 discusses various pressure loads where Load Case 1 represents the end cap pressure load which is applied on the reactor vessel and on the nozzle. Load Case 2 is where an axial load of 1 kips is applied to the safe end of the nozzle. Load Case 3 is the in-plane bending moment applied at the nozzle (perpendicular to the RPV centerline). Load Case 4 is the out-of-plane bending moment (parallel to RPV centerline). Section 4 in Enclosure 2 appears to model these load cases such as the end cap pressure load. However, it is not clear whether Enclosure 2 modeled the in-plane and out-of-plane bending moments on the feedwater nozzle.
Request:
Clarify whether these four load cases are applicable to the CGS feedwater nozzle. If they are applicable, discuss whether these four load cases are included in the finite element analysis for the CGS feedwater nozzle.
RAI 2.10 Issue:
It seems that the deterministic fracture mechanics (DFM) evaluation and some its results are not discussed in Enclosure 2. The NRC staff noted that the DFM evaluation should be part of the probabilistic fracture mechanics (PFM) evaluation to demonstrate the acceptability of reducing inspection of number of the feedwater nozzles and nozzle-to-shell welds.
Request:
(a) Discuss the equations, input values, and calculations for the DFM evaluation, including initial flaw size, crack growth, final flaw size, applied stress intensity factor, and material fracture toughness that were used for CGS.
(b) Discuss how the axial flaw and circumferential flaw are modeled in the nozzle radius and nozzle-to-shell weld in the DFM evaluation for CGS. Specifically, discuss the initial flaw size and the direction of the axial and circumferential flaw propagation in the nozzle radius and nozzle-to-shell weld.
6 (c) The results of the DFM evaluation at the end of the plant life should be in terms of (c1) the final size of the postulated axial and circumferential flaws in the nozzle inner radius and nozzle-to-shell weld, and (c2) the applied stress intensity factor (Ki) of the final flaw size as compared to fracture toughness of the material (KIC).
(d) Clarify whether the weld residual stress was calculated in the deterministic stress analysis in Enclosure 2, Section 4.0. If not, provide justification.
(e) Use the equations in the DFM evaluation and the algorithm flow diagram in the PFM evaluation to describe two runs (two realizations) of Monte Carlo simulation to show how the PFM calculation is performed from the flaw initiation, flaw growth, to the probability of failure.
- 3. Probabilistic Fracture Mechanics Analysis RAI 3.1 Issue:
The NRC staff notes that Figure 8-1 of BWRVIP-05 provides an overview of the PFM analysis methodology. Figure 8-11 of BWRVIP-05 provides a flow diagram of the computer code VIPER. states that it follows the PFM methodology in BWRVIP-05, BWRVIP-108-A, and BWRVIP-241-A. However, it is not clear to the NRC staff exactly how the PFM evaluation in was performed.
Request:
(a) Discuss whether the PFM evaluation in Enclosure 2 is similar to the flow diagrams in BWRVIP-05. If there are differences, provide a flow diagram that explains the methodology in the PFM evaluation in Enclosure 2.
(b) Discuss how the stress paths in the stress analysis are used to derive the applied stress intensity factors.
RAI 3.2 Issue:
The licensee stated that its feedwater nozzle analysis was based on BWRVIP-108 and BWRVIP-241. Enclosure 2 uses the VIPERNOZ computer code to perform the PFM evaluation.
The NRC staff noted that both BWRVIP reports do not analyze feedwater nozzles and associated nozzle-to-shell welds. Therefore, it is not clear how the two BWRVIP reports are applicable to the CGS feedwater nozzle analysis.
Request:
Discuss whether the version of VIPERNOZ used in the PFM evaluation for CGS is the same version as was used in the two BWRVIP reports. If not, provide the differences and discuss the impact of these differences on the CGS feedwater nozzle analysis.
RAI 3.3 Issue: does not discuss how the flaw growth due to stress corrosion cracking (SCC) is combined with the flaw growth due to fatigue to calculate the probability of failure of the feedwater nozzle and nozzle-to-shell weld.
7 Request:
Clarify how the SCC flaw growth is added to the flaw growth due to fatigue in the calculation of probability of failure of the feedwater nozzle and nozzle-to-shell weld.
RAI 3.4 Issue: , Section 5.1, Item 6 indicates that KIC of the feedwater nozzle is 200 ksiin. The NRC staff noted that KIC of 200 ksiin is applicable to material at high temperature. KIC at lower temperature would be lower than 200 ksiin.
Request:
Discuss whether the temperature at the feedwater nozzle and nozzle-to-shell weld stays at sufficient high temperature at the time of maximum total applied load to qualify for the use of 200 ksiin.
RAI 3.5 Issue: , Section 5.0 states that for the nozzle-to-shell weld, either a circumferential or an axial crack, depending on weld orientation, can initiate due to either component fabrication (i.e.
considering only welding process) or stress corrosion cracking. It is generally known that welds get repaired during construction of nuclear plants. The operating experience has shown that construction repair creates weld residual stresses which could increase the probability of flaw initiation and growth. Enclosure 2, Section 5.0 does not discuss whether the nozzle-to-shell weld had been repaired during the original construction Request:
Discuss whether any of the feedwater nozzle-to-shell welds was repaired at the time of the construction. If yes, discuss the flaw size that was repaired and discuss whether the weld residual stress analysis includes such repaired flaw. If not included, provide justification.
RAI 3.6 Issue:
Section 5.2.2.3 in Enclosure 2 states that the stress corrosion cracking (SCC) initiation data was based on cast stainless steel. The reactor vessel and feedwater nozzle are made of low alloy steel; therefore, the SCC initiation law should be based on the cracking data in low alloy steel.
Request:
Clarify why the SCC initiation data for cast stainless steel was used for the crack growth in the feedwater nozzle and nozzle-to-shell weld even though both components are not made of cast stainless steel.
RAI 3.7 Issue: , Section 5.2.2.5 discusses the comparison between the fatigue crack growth data that were used in the CGS analysis and the fatigue crack growth law in the ASME Code,Section XI in a reactor water environment. The licensee stated that its fatigue crack growth data
8 show a reasonable comparison; however, the fatigue growth law in the ASME Code,Section XI is more conservative than that in the EPRI report on growth rate at high K (Kmax - Kmin).
Request:
Discuss whether the fatigue crack growth curves used in the CGS PFM evaluation is adequate when the fatigue crack growth curve at high K (Kmax - Kmin) in the ASME Code,Section XI is more conservative than the fatigue crack growth curves used in the PFM evaluation.
RAI 3.8 Issue:
It appears that the POD in Enclosure 2, Table 10 is applied to both the feedwater nozzle radius and nozzle-to-shell welds. The NRC staff noted that the nozzle radius and nozzle-to-shell weld are of different shape and thickness. The ultrasonic examination coverage may be different between the feedwater nozzle radius and the nozzle-to-shell weld.
Request:
Justify why the same POD is applicable to both components.
RAI 3.9 Issue: does not describe the interface between the probability of detection (POD) and the inspection of 25% of the six feedwater nozzles.
Request:
Discuss how the POD is combined with the 25% inspection (i.e., inspecting 25% of the six feedwater nozzles) is analyzed to reach the probably of failure.
RAI 3.10 Issue:
The NRC staff notes that with the proposed 25% of inspection of the six feedwater nozzles, there will be feedwater nozzles that will not be inspected for the remainder of the plant life. In this scenario, the PFM evaluation would need to assume that there is no inspection for those feedwater nozzles and calculate the probability of failure accordingly (i.e., 0% inspection, not 25% inspection).
Request:
Discuss whether this is a plausible scenario for the PFM evaluation. If yes, discuss the impact of this scenario on the final probability of failure.
RAI 3.11 Issue: does not provide failure criteria for the feedwater nozzle and nozzle-to-shell weld in the PFM evaluation. Without the failure criteria, it is not clear how the probability of failure is estimated.
9 Request:
Clarify what are the criteria (in terms of nozzle leakage or fracture) for the failure to occur at the feedwater nozzle inner radius and associated nozzle-to-shell weld in the PFM evaluation and whether the criteria are consistent with that of BWRVIP-108 and BWRVIP-241.
RAI 3.12 Issue: , Section 6 states that the probability of failure (PoF) for the CGS feedwater nozzles or the nozzle-to-shell welds at 25% inspection in 60 years for the normal operation condition and LTOP event are less than 1.67E-8 and 1.67E-11, respectively. However, these two PoF estimated are much less than the PoF estimated for the recirculation nozzles for a 25%
inspection in 40 years in BWRVIP-108 and BWRVIP-241. For example, Table 4, Case 5 in the BWRVIP-108 Supplement dated September 13, 2007 (ADAMS Accession No. ML072600173, proprietary) shows a higher PoF for the recirculation nozzles than the CGS feedwater nozzles.
Also, the PoF estimated for the CGS feedwater nozzles during the LTOP event in Enclosure 2 is less than the PoF estimated for the CGS recirculation nozzle as shown in Table 5-9 in BWRVIP-241. The NRC staff notes that the PoF for the CGS feedwater nozzles is calculated for 60 years of operation whereas the PoF for the recirculation nozzles is calculated for 40 years in BWRVIP-108 and BWRVIP-241. In general, the feedwater nozzles should have experienced more transients than the recirculation nozzles. In addition, the CGS feedwater nozzles having more applicable transients as shown in Enclosure 2, Table 5 than the corresponding nozzles analyzed in the September 13, 2007 supplement of BWRVIP-108 and Table 5-9 of BWRVIP-241. Therefore, it appears that the CGS feedwater nozzle PoF should be higher, not lower, than the recirculation nozzle PoF.
Request:
Justify why the CGS feedwater nozzles in Enclosure 2 have lower PoF than the recirculation nozzles in the BWRVIP-108 supplement and BWRVIP-241.