PLA-6383, Response to Request for Additional Information for the Review of the Units 1 & 2, License Renewal Application (LRA) Sections B.2.1, B.2.5, B.2.7, B.2.9, and B.2.10
| ML082110399 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 07/14/2008 |
| From: | Mckinney B Susquehanna |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| PLA-6383 | |
| Download: ML082110399 (23) | |
Text
Brltt T. McKinney Sr. Vice President & Chief Nuclear Officer PPL Susquehanna, LLC 769 Salem Boulevard Berwick, PA 18603 Tel. 570.542.3149 Fax 570.542.1504 btmckinney@pplweb.com P
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JU 1 4 2008 U. S. Nuclear Regulatory Commission Document Control Desk Mail Stop OP 1-17 Washington, DC 20555 SUSQUEHANNA STEAM ELECTRIC STATION REQUEST FOR ADDITIONAL INFORMATION FOR THE REVIEW OF THE SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 AND 2, LICENSE RENEWAL APPLICATION (LRA)
SECTIONS B.2.1, B.2.5, B.2.7, B.2.9, and B.2.10 PLA-6383 Docket Nos. 50-387 and 50-388
References:
"Application for Renewed Operating License Numbers NPF-14 and NPF-22,"
dated September 13, 2006.
"Request for Additional Information for the Review of the Susquehanna Steam Electric Station, Units 1 and 2 License Renewal Application, "dated June 12, 2008.
In accordance with the requirements of 10 CFR 50, 51, and 54, PPL requested the renewal of the operating licenses for the Susquehanna Steam Electric Station (SSES)
Units 1 and 2 in Reference 1.
Reference 2 is a request for additional information (RAI) related to License Renewal Application (LRA) Sections B.2.1, B.2.5, B.2.7, B.2.9, and B.2.10. The enclosure to this letter provides the additional requested information.
There are no new regulatory commitments contained herein as a result of the attached amendments to the SSES LRA.
If you have any questions, please contact Mr. Duane L Filchner at (610) 774-7819.
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Document Control Desk PLA-6383 I declare, under penalty of perjury, that the foregoing is true and correct.
Executed on: 22z1I B. T. McKinney
Enclosure:
PPL Responses to NRC's Request for Additional Information (RAI)
Copy: NRC Region I Ms. E. H. Gettys, NRC Project Manager, License Renewal, Safety Mr. R. Janati, DEP/BRP Mr. F. W. Jaxheimer, NRC Sr. Resident Inspector Mr. A. L. Stuyvenberg, NRC Project Manager, License Renewal, Environmental
Enclosure to PLA-6383 PPL Responses to NRC's Request for Additional Information (RAI)
Enclosure to PLA-6383 Page 1 of 20 RAI B.2.1-1 The staff has noted that the "scope of program," "parameters monitored/inspected," or "detection of aging effects" program element discussions in license renewal (LR) basis document (LRPD-05, Attachment 1.9) for aging management program (AMP) B.2. 1, Inservice Inspection (ISI) Program, does not identify the parameters or aging effects that the program manages. Identify the parameters, aging effects or aging mechanisms that are within the scope of AMP B.2. 1, ISI Program, and for which the AMP is credited.
PPL Response:
The Inservice Inspection (ISI) Program is designed to maintain structural integrity and ensure that aging effects will be discovered and addressed before the loss of intended function of the component. The ISI Program provides identification and measurement of crack initiation and growth and loss of material due to corrosion by following the examination and inspection requirements specified in ASME Section XI Tables IWB-2500-1, IWC-2500-1, or IWD-2500-1, respectively, for Class 1, 2, or 3 components.
Approval for any exceptions to the ASME Code requirements is requested from the NRC via a Relief Request or an Exemption Request.
During the period of extended operation, theISI Program is credited to manage the following aging effects/mechanisms for components within the reactor coolant system, including the reactor vessel and reactor vessel internals:
" Cracking due to stress corrosion cracking, intergranular stress corrosion cracking, irradiation-assisted stress corrosion cracking
" Loss of material due to general, pitting, and crevice corrosion
" Loss of fracture toughness due to thermal aging embrittlement of cast austenitic stainless steel components (see Note)
Note: Loss of fracture toughness due to thermal aging embrittlement of cast austenitic stainless steel components is managed via the detection of cracking and the monitoring of crack growth. Subsequent engineering evaluation of any detected cracking in a cast austenitic stainless steel component, performed as required by the corrective action program, will address the potential effect of any loss of fracture toughness.
RAI B.2.1-2 The staff has noted that the LR basis document for AMP B.2. 1, ISI Program indicated that the criteria in particular NRC-approved Boiling Water Reactor Vessels and Internals Project (BWRVIP) reports may be used in lieu of applicable ASME Code Section XI ISI requirements for ASME Code Class 1, 2, or 3 components. Clarify whether or not
Enclosure to PLA-6383 Page 2 of 20 proposals to use NRC-approved BWRVIP guideline criteria in lieu of applicable ASME Code Section XI requirements will be submitted for staff approval.
PPL Response:
All proposals to use NRC-approved BWRVIP guideline criteria in lieu of applicable ASME Code Section XI requirements will be submitted for staff approval as part of each Ten-year ISI Inspection Plan, in accordance with 10 CFR 50.55a.
RAI B.2.1-3 The staff has noted that the "corrective actions" program element discussion in the LR basis document for the ISI Program indicates that the corrective actions for the program will be implemented through implementation of the applicant's 10 CFR Part 50, Appendix B, Quality Assurance Program. Corrective actions for ASME Code Class components are required, through 10 CFR 50.55a, to be implemented in accordance with applicable corrective action provisions in the ASME Code Section XI Article IWB-3000, or its subarticles, paragraphs, or subparagraphs, or in ASME Code Cases that endorsed for use (through reference in 10 CFR 50.55a) in the latest NRC-issued version of Regulatory Guide 1.147. Clarify how the implementation of the Susquehanna Steam Electric Station (SSES) 10 CFR Part 50, Appendix B, Quality Assurance Program will ensure that the corrective actions for ASME Code Class 1, 2, or 3 components will be implemented in accordance with applicable corrective actions in ASME Code Section XI Article IWB-3000, or its subarticles, paragraphs, or subparagraphs; in NRC-approved ASME Code Cases that are endorsed for use in the latest NRC-issued version of Regulatory Guide 1.147; or through the NRC's relief request process that is defined in 10 CFR 50.55a.
PPL Response:
The SSES ISI program and governing procedures specify compliance with ASME Section XI corrective actions for defects found in Class 1, 2, or 3 components. This includes use of the acceptance standards in the applicable sections of IWB-3000. The approved SSES Ten-Year ISI Inspection Plan describes the use of NRC-approved ASME Code Cases that are endorsed for use in the latest NRC-issued version of Regulatory Guide 1.147. When alternative standards are necessary, NRC approval is obtained through the NRC Relief Request process.
The SSES QA program specifies audits of the ISI program every two years following established auditing procedures. These audits are conducted in accordance with assessment basis documents that provide guidelines specific to the topic that is assessed.
The audit guideline for the ISI program explicitly addresses compliance with 10 CFR 50.55a and related regulatory requirements and commitments.
Enclosure to PLA-6383 Page 3 of 20 RAI B.2.1-4 The staff has noted that the "operating experience" program element discussion in the LR basis document for the ISI Program does not specify any SSES-specific or generic operating experience that the applicant felt was relevant to the AMP. However, the staff has noted that the LR binder for the ISI Program included two (2) condition reports related to the detection of circumferential cracking in the SSES, Unit 1, N2J recirculation outlet nozzle safe-end weld and in the SSES, Unit 1, N1 B recirculation inlet nozzle safe-end weld. Although the staff has verified that applicant has taken appropriate correction actions for these circumferential crack indications (i.e., implementation of weld overlay repairs of the nozzles), the staff is of the opinion that the operating experience should have been mentioned in as applicable operating experience for the "operating experience" program element because a complete failure of the circumferential weld would have resulted in a loss of coolant accident for the facility. Thus, this is important operating experience for SSES.
Part A. The staff requests that the license renewal application (LRA) be amended to list these circumferential crack safe-end nozzle events as relevant operating experience for the "operating experience" program element in AMP B.2. 1, "Inservice Inspection (ISI)
Program."
Part B. Identify the particular weld overlay methodology (along with its reference basis) that was used for the repairs of these safe-end nozzle indications, and clarify whether the overlay methodology required a flaw tolerance evaluation of the flaw indications, and if so, whether the analysis is a time-limited aging analysis (TLAA) for the application.
Justify your basis for concluding that the flaw tolerance analysis is or is not a TLAA if the weld overlay methodology required a flaw tolerance analysis.
PPL Response:
Part A The SSES LRA is modified by adding the following information to the Operating Experience program element of Section B.2.1.
B.2.1 Inservice Inspection (ISI) Program The discussion under Operating Experience in Section B.2.1 (LRA page B-20) is revised by addition (bold italics).
Operating Experience During the Unit I refueling outage in the Spring of 2004, the ultrasonic inspection of the NIB and N2J Recirculation Nozzle-Safe End (NOZ-SE) welds revealed indications which were determined to be flaws. The NOZ-SE welds are dissimilar metal welds that
Enclosure to PLA-6383 Page 4 of 20 join the carbon steel nozzles to the stainless steel safe-ends. The characteristics of these indications were typical of stress corrosion cracking in Alloy 82/182 weld material.
The flaws were repaired using full structural weld overlays, based on the Standard Weld Overlay defined in NUREG-0313, Revision 2. The weld overlay design was based on the requirements of ASME Section XI, IWB-3640 and Code Case N-504-2. The weld overlays were applied using Inconel 52, a material highly resistant to IGSCC.
Subsequent inspections performed in 2008 indicated no cracking in the weld overlays at the NIB and N2JNOZ-SE weld locations.
Part B The flaws were repaired using full structural weld overlays that were designed to bound all cracking conditions in the nozzle-safe end (NOZ-SE) weld area, using the Standard Weld Overlay defined in NUREG-0313, Revision 2. The weld overlay design was based on the requirements of ASME Section XI, IWB-3640 and Code Case N-504-2. The weld overlay design conservatively assumed the flaws were through-wall and extended entirely around the pipe. No credit was taken for any remaining ligament in the original NOZ-SE welds. The weld overlays were applied using Inconel 52, a material highly resistant to IGSCC.
The overlay welding methods result in compressive loads on the area beneath the overlay, thereby limiting the potential for further growth of the existing flaws. The overlay design also included a crack initiation and growth analysis, which demonstrated that the overlay will have a very low susceptibility for crack initiation and growth during the life of repair, due to the high IGSCC-resistance of the Inconel 52 alloy used in the overlay.
Post-repair inspections assured the quality of the repair, and ongoing inspection requirements for the overlay and the underlying base material will identify any future degradation.
The ASME Code required no flaw tolerance evaluations to be performed as part of the design basis for these repairs. There were no design basis analyses performed for the weld overlay repairs that constitute a TLAA.
RAI B.2.1-5 The staff has noted that, in the "detection of aging effects" program element discussion for the ISI Program, the applicant takes exception to the recommended "detection of aging effect" criteria in the generic aging lessons learned (GALL) AMP XI.Ml, "ASME Code Section XI, Subsection IWB, IWC, and IWD," and proposes to credit a risk-informed ISI methodology for the required examinations of particular ASME Code Class welds. Chapter 1 of the GALL Report, Revision 1, Volume 2, makes the following
Enclosure to PLA-6383 Page 5 of 20 statement on the applicability of current Code reliefs for the period of extended operation:
"The NRC Director of the Office of Nuclear Reactor Regulation may approve licensee proposed alternatives to the ASME Code in accordance with the provisions of 10 CFR 50.55a(a)(3). These NRC approved ASME Code alternative requirements may have an associated applicability time limit. The applicability time limits associated with the approved alternatives do not extend beyond the current license term. If an applicant seeks relief from specific requirements of 10 CFR 50.55a and Section XI of the ASME Code for the period of extended operation, the applicant will need to re-apply for relief through the 10 CFR 50.55a relief request process once the operating license for the facility has been renewed.
The staff noted that the risk-informed ISI (RI-ISI) program for SSES, Units 1 and 2, was approved in an NRC-issued safety evaluation (SE) dated September 28, 2005 (ML051990330). The staff also noted that the RI-ISI program relief request was only approved for the 3rd 10-Year ISI Interval for SSES, Units 1 and 2, and that RI-ISI has yet to be proposed and approved for any of the 10-Year ISI intervals that are within the scope of the periods of extended operation for the SSES units. If you plan to use RI-ISI for the 4 th 10-Year ISI Interval and subsequent intervals, the staff requests that you commit to request relief for use of RI-ISI within 12 months before the start of each interval.
PPL Response:
The license renewal application is revised to remove the statement that identifies the use of risk-informed ISI methodology (RI-ISI) as an exception to NUREG-1801, AMP XI.M1. Since the use of RI-ISI at SSES must be approved pursuant to 10 CFR 50.55a(a)(3), it is not considered to be an exception to NUREG-1 801.
Per discussion with NRC staff (J. Medoff) on July 8, 2008, since PPL has removed the use of RI-ISI as an exception to NUREG-1801, PPL does not need to make an LRA commitment to request relief for the use of RI-ISI in future intervals. PPL is already committed to seek approval for the use of RI-ISI in accordance with 10 CFR 50.55 a(a)(3).
B.2.1 Inservice Inspection (ISI) Program The discussion under NUREG-1801 Consistency in Section B.2.1 (LRA pages B-19 and B-20) is revised by addition (bold italics) and deletion (stri4kethrough).
Enclosure to PLA-6383 Page 6 of 20 NUREG-1801 Consistency The SSES Inservice Inspection (ISI) Program is an existing program that is consistent with the 10 elements of an effective aging management program as described in NUREG-1801,Section XI.M1, "ASME Section XI Inservice Inspectionj." with an exeepteioil-Exceptions to NUREG-1801 D......
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Table B-2 Consistency of SSES Aging Management Programs with NUREG-1801 Table B-2 Entry for ISI Program (LRA page B-15) is revised by addition (bold italics) and deletion (sti4kethfough).
Table B-2 Consistency of SSES Aging Management Programs with NUREG-1801 (continued)
New /
Consistent Exceptions Plant-Enhancement Program Name Existing with NUREG-to NUREG-Specific Required 1801 1801 Inservice Inspection Existing Yes Y-es..
(ISI) Program RAI B.2.5-1 The staff has noted the LR binder for AMP B.2.5 BWR Feedwater Program, includes SSES Letter No. PLA-6078, "Susquehanna Steam Electric Station Unit 1 Fourteenth Refueling Outage Owners Activity Report," dated June 21, 2006, and GE Nuclear Energy Ultrasonic Testing (UT) Examination Summary Sheet No. 1-B3.90.0017, "Project Susquehanna Unit 1 - R&IO14, Weld ID No. N4A (NoZ-SC3), Reactor Pressure Vessel Weld," dated March 3, 2006. In these documents, SSES provides the results of augmented examinations that had been performed on the feedwater (FW) nozzles during the last refueling and inspection outages for SSES, Unit I and SSES, Unit 2 (i.e., RIO Ul 14RIO for SSES, Unit 1 and RIO U213RIO for SSES, Unit 2). The staff has noted
11 Enclosure to PLA-6383 Page 7 of 20 that the augmented UT examinations of the SSES Unit 1 N4A FW nozzle indicated the presence of eight (8) recordable flaw indications in the nozzle that had been dispositioned as being acceptable for further service by the ASME Code Section XI IWB-3000 requirements. However, the staff also noted that these inspection results were not mentioned in the "operating experience" program element for the AMP as relevant operating experience for this AMP.
Since these flaw indications were detected as a result of implementing the BWR Feedwater Nozzle Program and since the indications were recordable under the applicant's ISI implementation and nondestructive examination procedures, the staff requests that SSES amend LRA Section B.2.5 to mention these indications as relevant operating experience for the BWR Feedwater Nozzle Program. The staff also requests that LRA Section B.2.5 be amended to provide a basis for leaving these eight (8) flaw indications in service and the basis for reinspecting these indications in the future in accordance with the BWR Feedwater Nozzle Program, as implemented in accordance with the augmented ISI provisions in AMP B.2.1, ISI Program.
PPL Response:
The SSES LRA is modified by adding the following information to the Operating Experience discussion of Section B.2.5.
B.2.5 BWR Feedwater Nozzle Program The discussion under the Operating Experience program element in Section B.2.5 (LRA pages B-26 and B-27) is revised by addition (bold italics) and deletion Operating Experience The original design of the reactor vessel for SSES did not include cladding of the feedwater nozzles, but did include an adequate feedwater flow controller and routing of RWCU return flow to reduce thermal cycles during times of low feedwater flow. In addition, the original feedwater sparger thermal sleeves were replaced with an improved design. Pre-service examinations of the six Unit 1 feedwater nozzles and inner radii were conducted with no indications found. Subsequent inspections of the Unit 1 and Unit 2 feedwater nozzles have resulted in no only one recordable indications. Consistent with industry operating experience and corresponding NRC-approved recommendations, the inspection frequencyfor the feedwater nozzles is once per ten-year interval, was changed to once per-interzval dur-ing the twelfth Unit 1 refueling ouitage-.
During the fourteenth Unit 1 refueling outage in March 2006, all critical regions of the six Unit 1 feedwater nozzles were ultrasonically (UT) inspected as part of the SSES ISI
Enclosure to PLA-6383 Page 8 of 20 Program. No recordable indications were detected in five of the six nozzles. The UT results for Nozzle N4A indicated one recordable flaw and seven other indications that were too small to characterize as flaws. The one recordable flaw was evaluated against the criteria in ASME Section XI Table IWB 3510-1. It was determined to be acceptable for continued service, since the flaw size was less than half of that allowed by IWB-3510. This flaw indication did not represent a noticeable change from the previous inspection results. Since the flaw indication is within the acceptance criterion established in ASME Section XI, no change in the inspection frequency for the N4A or any other feedwater nozzle at SSES is required by the ISI Program or ASME Section XI.
During the thirteenth Unit 2 refueling outage in March 2007, all critical regions of the six Unit 2 feedwater nozzles were ultrasonically (UT) inspected as part of the SSES ISI Program. No recordable indications were detected in any of the six nozzles.
SSES operating experience, consistent with industry operating experience, shows that the BWR Feedwater Nozzle Program has been effective in managing aging effects in that no unacceptable feedwater nozzle cracking has been observed at SSES. Therefore, continued implementation of the program provides reasonable assurance that effects of aging will be managed so that the feedwater nozzles can perform their intended function consistent with the current licensing basis during the period of extended operation.
RAI B.2.5-2 SSES provides its updated final safety analysis report (UFSAR) Supplement summary description for the BWR Feedwater Nozzle Program in LRA Section A. 1.2.6, which includes Commitment No. 5. The staff has noted an inconsistency in the application.
Specifically, the staff has noted that the AMP description for AMP B.2.5, "BWR Feedwater Nozzle Program," states that the UT methodology for the augmented inspections of the FW nozzles will be implemented in accordance with the recommendations of GE Topical Report No. GENE-523-A71-0594. In contrast, the staff has also noted that the UFSAR Supplement summary description for this AMP indicates that the augmented UT inspections of the nozzles will be implemented in accordance with the recommendations in applicable BWRVIP guidelines. Clarify (with justification) which basis and methodology will be used for performing the augmented UT examinations of the SSES FW nozzles during the period of extended operation.
PPL Response:
The SSES BWR Feedwater Nozzle Program is part of the SSES ISI Program. The ISI requirements for the feedwater nozzles are in accordance with ASME Section XI, Subsection IWB, Table 2500-1, and NRC-approved BWR Owners Group Topical Report, GENE-523-A71-0594, Revision 1, which provides guidance for inspecting the feedwater nozzle bore region using UT methodologies. This is consistent with NUREG-1801
Enclosure to PLA-6383 Page 9 of 20 Section XI.M5. The SSES BWR Feedwater Nozzle Program is committed to following the GENE-523-A71-0594 guidelines during the period of extended operation.
The description of the SSES BWR Feedwater Program in the SSES LRA is clarified by removing reference to BWRVIP guidelines.
The discussion in LRA Section A. 1.2.6 (page A-6 of the LRA) and the Program Description of LRA Section B.2.5 (page B-26 of the LRA) are revised by deletions (st-ikethe g
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A.1.2.6 BWR Feedwater Nozzle Program The BWR Feedwater Nozzle Program is an existing program that manages cracking of the feedwater nozzles. The ByWR Feedwater Nozzle Pr.gram
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accordane ith ASME Section X! and Boiling Water Reac to Vesqsqelal Interna*lls Project (BM IP) guidelin~es-.
The program includes (a) enhanced inservice inspection in accordance with the requirements of the ASME Boiler and Pressure Vessel Code,Section XI, Subsection IWB, Table IWB 2500-1 (edition and addenda described in A. 1.2.23) and the recommendations of report GE-NE-523-A71-0594, and (b) system modifications (completed on the spargers prior to initial startup) to mitigate cracking. The program specifies periodic ultrasonic inspection of critical regions of the feedwater nozzles.
B.2.5 BWR Feedwater Nozzle Program Program Description The purpose of the BWR Feedwater Nozzle Program is to manage cracking of the feedwater nozzles. The BWXR Feedwater Nozzle Pro gramisinaccor-dance with ASME Section X!
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This program includes (a) enhanced inservice inspection in accordance with the requirements of the ASME Boiler and Pressure Vessel Code,Section XI, Subsection WB, Table IWB 2500-1 (edition and addenda described in B.2. 1) and the recommendations of report GE-NE-523-A71-0594, and (b) system modifications (completed on the spargers prior to initial startup) to mitigate cracking. The program specifies periodic ultrasonic inspection of critical regions of the feedwater nozzles.
Enclosure to PLA-6383 Page 10 of 20 RAI B.2.7-1 The staff requests the following information with respect to the "preventative actions" program element for AMP B.2.7, BWR Stress Corrosion Cracking Program.
Part A. The "preventative actions" program element for the BWR Stress Corrosion Cracking Program indicates that two welds scheduled for stress relief did not receive a post-weld heat treatment consistent with the NRC Generic Letter (GL) 88-01/
NIUREG-0313 recommendations and that the welds were deemed unacceptable for stress relief credit, as stated in the NRC's SE on the SSES response to GL 88-01. Discuss whether there is any established link between the findings identified in the NRC's SE on the applicant's response to GL 88-01 and the circumferential stress corrosion cracking-induced flaw indications that have been detected in the SSES, Unit 1, N2J recirculation outlet nozzle safe-end weld and in the SSES, Unit 1, NIB recirculation inlet nozzle safe-end weld. Specifically, identify whether these safe-end nozzle welds were among the Class 1 stainless steel piping welds that were scheduled for induction heat stress relief treatments and whether the N2J and NIB nozzle safe-end welds were the welds that had not received the recommended post weld heat treatments that are part of this stress relief process.
Part B. Identify the dates for initiation of hydrogen water chemistry at SSES, Unit 1 and SSES, Unit 2.
PPL Response:
Part A The discussion in the license renewal basis document for the "preventive actions" program element for the BWR Stress Corrosion Cracking Program incorrectly stated that there are "two SI-treated welds that were not given post weld heat treatment." The correct statement is that there are "two SI-treated welds that were not completely ultrasonically examined post-SI."
The two welds in question are identified in the PPL letter to the NRC, PLA-3263, dated October 2, 1989, as DCA1081-FW-5 and DCA1 102-FW-6. These welds are piping welds on the Unit 1 Residual Heat Removal System, not the SSES Unit 1 NIB and N2J recirculation nozzle-safe end welds. And, these piping welds did, in fact, have the Induction Heating Stress Improvement Process (IHSI) performed within two years of commercial operation, consistent with the NRC Generic Letter (GL) 88-01/NUREG-0313 recommendations. However, the post-IHSI ultrasonic examination (UT) of the welds could not be performed, as required by NUREG-0313, due to the weld configuration. In PLA-3263, PPL classified these two welds as IGSCC Category G and committed to inspect the welds during the next refueling outage. In the NRC's SE on the SSES response to GL 88-01, it was the classification of these two welds as IGSCC Category G
Enclosure to PLA-6383 Page 11 of 20 that the NRC found to be unacceptable. Subsequently, PPL inspected these welds during the Unit 1 5 th refueling outage in 1990, and the welds are now classified as IGSCC Category B.
The SSES Unit 1 N1B and N2J nozzle-safe end welds did not have IHSI within two years of commercial operation. As these are dissimilar metal welds, IHSI is not an appropriate stress improvement method. Instead, these welds had the Mechanical Stress Improvement Process (MSIP) applied after approximately ten years of commercial operation.
There is no link between the findings identified in the NRC's SE on the PPL response to GL 88-01 and the flaw indications detected in the SSES Unit 1 NIB and N2J recirculation nozzle safe-end welds.
Part B Hydrogen water chemistry was implemented for SSES Unit 1 in January 1999 and for SSES Unit 2 in August 1999.
RAI B.2.7-2 The staff requests the following information with respect to the "detection of aging effects," "monitoring and trending," and "acceptance criteria" program elements for AMP B.2.7, BWR Stress Corrosion Cracking. The basis document for AMP B.2.7, BWR Stress Corrosion Cracking, implies that the guidelines in BWRVIP-75-A will only be used as a basis for sample expansion if flaw indications are detected on ASME Code Class 1 stainless steel welds. However, the guidelines in BWRVIP-75-A are NRC-approved inspection and flaw evaluation guidelines.
Part A. Clarify whether the updated NRC-approved guidelines in Topical Report BWRVIP-75A would be used as an option for performing other aspects of the augmented ISI Program for these ASME Code Class 1 stainless steel pipe welds.
Part B. Clarify whether the flaw acceptance criteria in NRC-approved Topical Report BWRVIP-75A or in NRC-approved Topical Report BWRVIP-14 will be used for the acceptance criteria of any crack indications that might be detected in these ASME Code Class I stainless steel pipe welds.
PPL Response:
Part A The NRC-approved guidelines in BWRVIP-75-A are used in the augmented ISI Program for IGSCC in ASME Code Class 1 stainless steel pipe welds, known as AUG2. PPL
Enclosure to PLA-6383 Page 12 of 20 follows the inspection criteria of BWRVIP-75-A to schedule the required inspections and for sample expansion if flaw indications are detected under the AUG2 program. Since BWRVIP-75-A only addresses inspection criteria and schedules, PPL will not use it for any other aspects of the program.
Part B Flaw evaluation and acceptance criteria is in accordance with the ASME Code,Section XI, IWB-3640, as specified in NUREG-0313, Revision 2. PPL is committed to follow all requirements of NUREG-0313, Revision 2, except for the inspection criteria and schedule.
PPL does not use BWRVIP-75-A for flaw acceptance criteria, since the report contains no flaw acceptance criteria guidance. The NRC-approved BWRVIP-14 addresses crack growth evaluation of flawed BWR shroud welds and other stainless steel internals. As part of the ASME Code flaw evaluation, a crack growth analysis is required. While PPL may use certain data and evaluation methods from BWRVIP-14 in a crack growth analysis, the evaluation and acceptance criteria will be in accordance with the ASME Code,Section XI, IWB-3640.
RAI B.2.9-1 AMP B.2.9, BWR Vessel Internals Program includes an enhancement to perform augmented inspections of the top guide grid beam and beam-to-beam crevice slots during the period of extended operation. The enhancement commits to enhanced VT-I visual examinations of 5% of these top guide locations within 6 years of entering the period of extended operation with an additional 5% of these locations to be completed within 12 years of entering the period of extended operation. The staff has noted that this enhancement is consistent with the recommendations in the "scope of program" program element in GALL AMP XI.M9, "BWR Vessel Internals." However, the GALL recommendation is predicated on the criteria that cracking has not yet been detected in these top guide locations. Clarify whether SSES has performed any inspections of the SSES top guide grid beam and beam-to-beam crevice slot locations to date, and if so, whether SSES has detected and recorded the occurrence of any flaw indications (cracks) in these locations to date.
PPL Response:
The Unit 1 top guide beam and beam-to-beam crevice slot locations were inspected in 2004 during the thirteenth refueling outage. Twelve fuel cell locations were examined using the VT-3 inspection method. The Unit 1 top guide was inspected again in 2008.
At that time, the top guide beam and beam-to-beam crevice slots at one cell location were inspected using the EVT-1 inspection method. No recordable indications were found in either Unit 1 inspection.
Enclosure to PLA-6383 Page 13 of 20 The Unit 2 top guide beam and beam-to-beam crevice slot locations at twenty-one fuel cell locations were inspected in the Unit 2 eleventh refueling outage in 2003 using the VT-3 inspection method. In 2007, four additional fuel cell locations in the Unit 2 vessel were inspected using the VT-3 inspection method. No recordable indications were found in either Unit 2 inspection.
RAI B.2.9-2 The staff noted that the applicant has not identified any relevant SSES-specific or generic operating experience in "operating experience" program element discussion for AMP B.2.9, BWR Vessel Internals Program. The staff has noted, however, that the LR basis binder for this AMP does include several Condition Reports/Action Requests that reported the occurrence of flaw indications (cracks) in the core spray sparger brackets, core shroud circumferential welds and some of jet pump assembly components (i.e., jet pump restrainers, wedges, and rods). The staff also observed that the applicant has dispositioned these flaw indications as acceptable (i.e., "As-Is") for further service without the need for repair or replacement of the components at this time. Provide your basis why the flaw indications in the core spray sparger brackets, core shroud welds, and jet pump assembly components have not been identified as relevant operating experience for the AMP B.2.9, BWR Vessel Internals Program, and provide your basis for leaving the flaws in these components in service (i.e., acceptable "As-Is") without repair or replacement of the impacted components. State, with a technical justification, what the inspection frequencies and sample sizes will be for re-inspecting these reactor vessel internal (RVI) components during the period of extended operation.
PPL Response:
PPL has identified flaw indications in the core spray sparger brackets, core shroud welds, and jet pump assembly components. This operating experience will be added to the LRA. All identified flaws that have been allowed to remain in-service (i.e., acceptable "As-Is") have been evaluated in accordance with the applicable BWRVIP documents.
During the period of extended operation the core spray sparger brackets, core shroud welds, and jet pump assembly components will be inspected and evaluated using inspection frequencies, sample sizes, and evaluation criteria in accordance with the applicable BWRVIP document (BWRVIP-76 for the core shroud, BWRVIP-I8-A and BWRVIP-76 for the core spray sparger brackets, and BWRVIP-41 for the jet pump assemblies). The exact frequency and sample size will depend on the results of the inspections conducted prior to entering the period of extended operation.
The SSES LRA is modified to add the relevant information to the Operating Experience program element of Section B.2.9.
Enclosure to PLA-6383 Page 14 of 20 The discussion under the Operating Experience program element in Section B.2.9 (LRA pages B-35) is revised by addition (bold italics) and deletion (str4kethreugh).
Operating Experience SSES operating experience is consistent with industry experience; a large number of examinations are being performed, and an occasional indication is being found and resolved. No aging mechanisms not already addressed have been discovered. Examples of how the B WR Vessel Internals Program has detected and managed aging effects of reactor internal components are provided below.
Core Shroud Welds The first indication of cracking in the core shroud horizontal welds was found in Unit 1 during the 1995 refueling outage. Indications in the Unit 2 shroud were also first observed in 1995. Subsequent inspections of the shroud horizontal and vertical welds have been performed in 2000, 2004, and 2006for Unit 1, and in 1999, 2003, and 2005for Unit 2. Most of the horizontal welds in both units exhibit some degree of cracking. The vertical welds have exhibited only one relatively small vertical crack in the Unit 1 shroud. To date, the flaws detected have been evaluated using the methods and criteria defined in BWRVIP-76. All of the welds with flaws were evaluated, and it was determined that they would remain structurally adequate until the next inspection.
Currently, PPL plans to reinspect all of the Unit 2 shroud welds in 2009 and all of the Unit 1 shroud welds in 2010. Results of these inspections will be used to determine the next required inspection for each weld in accordance with the guidance provided in B WR VIP-76.
Core Spray Sparmers Brackets In 1996, a flaw indication was detected in one of the sparger brackets (SB4) in Unit 1. In 1997, a similar flaw was detected in the Unit 2 core sprayer bracket.
The affected bracket welds were evaluated to be acceptable for continued operation, supported by a structural analysis. The Unit 1 flaw was re-inspected in 2004 using improved UT technology to size the flaw. Similarly, the Unit 2 core spray sparger brackets were inspected in 2005 and flaws were found at three locations. Both the Unit 1 and Unit 2 core spray sparger bracket flaws were determined to be in the shroud plate base metal (heat affected zone of the weld) and were evaluated using the flaw evaluation guidance and criteria provided in B WR VIP-76. The Unit 1 and 2 core spray sparger brackets were determined to have adequate structural margin. The core spray sparger brackets are currently inspected every outage in accordance with B WR VIP A. The Unit 1 brackets were inspected in 2006 and 2008, and the Unit 2
Enclosure to PLA-6383 Page 15 of 20 brackets were inspected in 2007. No growth of the flaws identified in 2004 on Unit 1 and in 2005 for Unit 2 has been observed in the subsequent inspections.
Jet Pump Wedges and Set Screws The jet pump components and welds have been inspected on a regular basis since plant startup. The jet pump holddown beams on all Unit 1 and Unit 2jet pumps were replaced in 1993 and 1994, respectively, in response to industry experience.
In 2001, excessive jet pump wedge wear and set screw gaps were observed on the Unit 2jetpumps. Similar observations were made in 2002 on Unit 1. In 2003 (for Unit 2) and 2004 (for Unit 1), modifications were installed, including machining labyrinth seals in 20 jetpump inlet mixers (to reduce flow induced vibration), replacing several wedges, and machining several restrainer bracket pads. Subsequent inspections have revealed additional minor wedge and rod wear. These components will continue to be monitored in accordance with B WR VIP-41, and repairs or modifications made as required to ensure the jet pumps are properly supported.
The extensive industry operating experience with the BWRVIP Program to date provides assurance that the program is effective in managing effects of aging so that components crediting these programs can perform their intended function consistent with the current licensing basis during the period of extended operation.
INPO conducted a BWRVIP vessel and internals review visits at SSES in 2003, 2004, and 2007. during September-13 16, 2004. The visits included a review of inspections, flaw analysis and repairs, mitigation activities and leakage detection and equipment monitoring. INPO r.esu.lts are documented in an SSES self assessment r.epot.... INPO The assessment team concluded that the BWRVIP program was being effectively implemented.
The SSES BWR Vessel Internals Program includes provisions to adopt any BWRVIP guideline changes in the future. This assures that operating experience from the BWR fleet will continue to be incorporated into the SSES program.
RAI B.2.10-1 Part A. The "scope of program" program element for AMP B.2.10, Thermal Aging and Neutron Irradiation Embrittlement of Cast Austenitic Stainless Steel Program, states that the cast austenitic stainless steel (CASS) RVI components will be screened for their susceptibility to loss of fracture toughness by thermal aging embrittlement and neutron irradiation embrittlement. However, the program element does not establish which NRC-approved guideline(s) or basis document(s) that will be used to screen the
Enclosure to PLA-6383 Page 16 of 20 CASS RVI components for susceptibility to these aging phenomena. Clarify the NRC-approved guideline(s) or basis document(s) that will be used to screen the CASS RVI components for susceptibility to the aging phenomena of thermal aging embrittlement and neutron irradiation embrittlement.
Part B. The staff have noted an inconsistency in AMP B.2. 10 in that the "scope of program" program element for the AMP indicates that SSES will use the material properties such as casting method, molybdenum content, and ferrite content as its basis for screening the susceptibility of the CASS RVI components to both thermal aging embrittlement and neutron irradiation embrittlement, whereas the "parameters monitored/inspected" program element description for the AMP states that the screening would be based on both neutron fluence levels and component material properties (including alloying content criteria). Although screening for thermal aging may be on component material properties (i.e., casting method, 6-ferrite content, and Molybdenum content), screening for neutron irradiation embrittlement must be based on an established NRC-approved integrated neutron flux threshold (i.e, neutron fluence) for CASS materials. The staff requests that the inconsistency between the "scope of program" and the "parameters monitored/inspected" program element descriptions be resolved and that SSES identify the specific parameter criteria that will be used to screen the CASS RVI components for reduction of fracture toughness by both thermal aging embrittlement and neutron irradiation embrittlement.
PPL Response:
Part A The screening of reactor vessel internals components for susceptibility to thermal aging will be done per the guidance in the May 19, 2000, letter from Christopher Grimes (NRC) to D. J. Walters (NEI), "Thermal Aging Embrittlement of Cast Austenitic Steel Components," and in EPRI Technical Report 100976, "Evaluation of Thermal Aging Embrittlement for Cast Austenitic Steel Components," January 2001.
The screening of reactor vessel internals components for susceptibility to neutron embrittlement will use the threshold of neutron fluence greater than 1 E+ 17 n/cm2 (E>lMev), as specified in NUREG-1801, Revision 1,Section XI.M13.
Part B The "scope of the program" and the "parameters monitored/inspected" sections of AMP B.2. 10 are revised to be consistent with each other and to identify the screening criteria for thermal aging and neutron embrittlement.
Enclosure to PLA-6383 Page 17 of 20 B.2.10 Thermal Aging and Neutron Embrittlement of Cast Austenitic Stainless Steel (CASS) Program The discussion under Aging Management Program Elements in Section B.2. 10 (LRA pages B-36 and B-37) is revised by addition (bold italics) and deletion (Skethfough).
Aging Management Program Elements The results of an evaluation of each program element are provided below.
Scope of Program The SSES Thermal Aging and Neutron Embrittlement of Cast Austenitic Stainless Steel (CASS) Program will screen reactor vessel internals components to determine which components are susceptible to reduction of fracture toughness-4ue-to the combination of thermal aging anad neut~ron embr-ittlement on the basis of casting method, molybdenuim content, and ferrite content. Screening for thermal aging will be based on casting method, molybdenum content, and ferrite content, in accordance with the criteria found in the May 19, 2000, letter from Christopher Grimes (NRC) to D. J. Walters (NEI), "Thermal Aging Embrittlement of Cast Austenitic Steel Components," and in EPRI Technical Report 100976, "Evaluation of Thermal Aging Embrittlement for Cast Austenitic Steel Components," January 2001.
Screening for neutron embrittlement will use the fluence threshold of lE+1 7 n/cm2 (E>lMev).
" Parameters Monitored or Inspected The SSES program will screen components as discussed under Scope of Program.
for-susceptibility based on neutron fluence and component material pr-operties fellowing the gu-idelines in NUREGCR 4 513, Revision 1. Those components screened as susceptible to Reduction of Fracture Toughness (either due to thermal aging or neutron embrittlement) will require inspection unless it is determined by component-specific evaluations that inspection is not required. The component-specific evaluation will include a mechanical loading assessment to determine the maximum tensile loading on the component. If the loading is low enough to preclude fracture, then supplemental inspection of the component is not required.
Those components evaluated to require inspection will be inspected by augmentation of the Inservice Inspection (ISI) Program as discussed under Detection ofAging Effects.
Enclosure to PLA-6383 Page 18 of 20 RAI B.2.10-2 Part A. The "detection of aging effects" program element for AMP B.2. 10, Thermal Aging and Neutron Irradiation Embrittlement of Cast Austenitic Stainless Steel Program, states that the applicant may use UT as one of the inspection techniques that are used to detect for cracking in these CASS components. Current state of the art UT inspection methods have not yet been qualified as being capable of detecting cracks in CASS materials because the CASS microstructures create significant noise signals (from refraction of the UT beams) that may mask UT signals coming from any relevant flaw indications in the CASS materials. Clarify whether there are any NRC-approved state of the art UT techniques that are capable of detecting cracks in CASS material microstructures, and if not, clarify which alternate NRC-approved inspection technique/method will be implemented to monitor for cracking in these materials.
Part B. The "detection of aging effects' program element in GALL AMP, XI.M13, "Thermal Aging and Neutron Irradiation Embrittlement of Cast Austenitic Stainless Steel," (in part) states that component specific flaw tolerance evaluations meeting specific criteria may be used as an alternative approach for managing reduction of fracture toughness in CASS RVI components. Provide your basis why the "detection of aging effects" or "monitoring or trending" program elements for the AMP B.2. 10 did not credit a supplemental flaw tolerance analysis as an alternative basis for managing reduction of fracture toughness in these CASS RVI components.
PPL Response:
Part A PPL is not aware of any ultrasonic inspection (UT) techniques currently approved by the NRC for detecting cracking in CASS components. The statements made in the LRA were intended to preserve the option to include new examination techniques, such as UT, only if they are developed and approved in the future. At present, the only NRC-approved inspection is the enhanced visual examination (EVT-1), as recommended by NUREG-1801,Section XI.M13. PPL has revised (see below) the last sentence of the "detection of aging effects" discussion in LRA Section B.2. 10 to clarify the NRC-approved inspection method to be implemented.
Part B PPL did not credit a supplemental flaw tolerance evaluation because the CASS reactor vessel internals covered by this program are not reactor coolant pressure boundary components, and, consequently, a classic critical flaw size analysis is not directly applicable.
Once the susceptible components are identified, PPL may perform a component-specific evaluation as discussed in the "detection of aging effects" program element in GALL
Enclosure to PLA-6383 Page 19 of 20 AMP, XI.M1 3. The performance of this component-specific evaluation is added to the LRA Section B.2.10.
B.2.10 Thermal Aging and Neutron Embrittlement of Cast Austenitic Stainless Steel (CASS) Program The discussion under Aging Management Progriam Elements in Section B.2.10 (LRA page B-37) is revised by addition (bold italics) and deletion (Si*kethfeg).
Aging Management Program Elements The results of an evaluation of each program element are provided below.
Detection of Aging Effects The SSES Thermal Aging and Neutron Embrittlement of Cast Austenitic Stainless Steel (CASS) Program will first screen components as discussed under Scope of Program3. then evaluate these For those components screened as susceptible to Reduction of Fracture Toughness, a component-specific evaluation may be performed to determine if supplemental inspection of the component is required, as discussed under Parameters Monitored or Inspected. Those components or portions of components evaluated to be limiting from the standpoint of thermal aging susceptibility, neutron fluence, and cracking susceptibility will be inspected by augmenting the Inservice Inspection (ISI) Program. Supplemental inspections will be added to the 10-year ISI Program Plan for the first 10 years of the period of extended operation. Examination techniques will comply with the requirements of NUREG-1801 Section XI.M1 3. As determined necessary, enhanced visualnondestie examinations (including visual, ultrasonic, and surface techniques) will be performed by qualified personnel following procedures consistent with Section XI of ASME B&PV Code and 10CFR50, Appendix B.
RAI B.2.10-3 The "scope of program" program element for AMP B.2.9, BWR Vessel Internals Program, states (in part) that the program is credited for limited management of loss of material and reduction of fracture toughness in the RVIs components at SSES. SSES's primary program for managing loss of fracture toughness (the mechanisms are by thermal aging and neutron irradiation embrittlement) in the CASS RVI components is AMP B.2. 10, Thermal Aging and Neutron Irradiation Embrittement of Cast Austenitic Stainless Steel Program. The staff has noted, however, that the applicant's Thermal Aging and Neutron Irradiation Embrittement of Cast Austenitic Stainless Steel Program does not identify the option to credit AMP B.2.9 for reduction of fracture toughness as an exception to either the "scope of program" or "parameters monitored/inspected" program
Enclosure to PLA-6383 Page 20 of 20 elements in GALL AMP XI.M 13, "Thermal Aging and Neutron Irradiation Embrittement of Cast Austenitic Stainless Steel."
The staff does not take issue with this alternative aging management approach. However, if SSES is crediting the BWRVIP as a option for managing reduction of fracture toughness in CASS RVI components, the staff requests that SSES amend the LRA to:
(1) identify this as an exception to the scope of program" program element in GALL AMP XI.M 13, "Thermal Aging and Neutron Irradiation Embrittlement of Cast Austenitic Stainless Steel," (2) identify the NRC-approved BWRVIP-based guideline reports that will be credited and used to manage reduction of fracture toughness in the CASS RVI components at SSES, and (3) assess the need to amend FSAR Supplement Section A. 1.2.48 and Commitment No. 10 on LRA Table A-I to reflect that the BWRVIP and applicable NRC-approved BWRVIP inspection and flaw evaluation guidelines may be used as an alternative basis for managing loss of fracture toughness in CASS RVI components.
PPL Response:
As shown in LRA Table 3.1.2-2, the BWR Vessel Internals Program is credited for managing reduction of fracture toughness for components made of either stainless steel (non-cast) or nickel based alloy. The BWR Vessel Internals Program is not credited for managing reduction of fracture toughness for any cast austenitic stainless steel (CASS) vessel internal components. As shown in LRA Table 3.1.2-2, the Thermal Aging and Neutron Embrittlement of Cast Austenitic Stainless Steel (CASS) Program is credited for managing reduction of fracture toughness for all CASS vessel internals. Therefore, there is no exception to the GALL AMP XI.M13, "Thermal Aging and Neutron Irradiation Embrittlement of Cast Austenitic Stainless Steel (CASS)."