ML13281A569
| ML13281A569 | |
| Person / Time | |
|---|---|
| Site: | Byron, Braidwood  |
| Issue date: | 12/12/2013 |
| From: | Daily J License Renewal Projects Branch 1 |
| To: | Gallagher M Exelon Generation Co |
| Daily J, 415-3873 | |
| References | |
| TAC MF1879, TAC MF1880, TAC MF1882, TAC NF1881 | |
| Download: ML13281A569 (18) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 December 12,2013 Mr. Michael P. Gallagher Vice President, License Renewal Projects Exelon Generation Company, LLC 200 Exelon Way Kennett Square, PA 19348
SUBJECT:
REQUESTS FOR ADDITIONAL INFORMATION FOR THE REVIEW OF THE BYRON NUCLEAR STATION, UNITS 1 AND 2, AND BRAIDWOOD NUCLEAR STATION, UNITS 1 AND 2, LICENSE RENEWAL APPLICATION- AGING MANAGEMENT, SET 4 (TAC NOS. MF1879, MF1880, MF1881, AND MF1882)
Dear Mr. Gallagher:
By letter dated May 29, 2013, Exelon Generation Company, LLC, submitted an application pursuant to Title 10 of the Code of Federal Regulations Part 54, to renew operating licenses NPF-37, NPF-66, NPF-72, and NPF-77 for Byron Nuclear Station, Units 1 and 2, and Braidwood Nuclear Station, Units 1 and 2, respectively, for review by the U.S. Nuclear Regulatory Commission staff. The staff is reviewing the information contained in the license renewal application and has identified, in the enclosure, areas where additional information is needed to complete the review.
These requests for additional information were discussed with John Hufnagel, and a mutually agreeable date for the response is within 30 days from the date of this letter. If you have any questions, please contact me at 301-415-3873 or by e-mail at 10hn.daily@ nrc.gov.
Sincerely, JvL~
John W. Daily, Sr. Project Manager Projects Branch 1 Division of License Renewal Office of Nuclear Reactor Regulation Docket Nos. 50-454, 50-455, 50-456. and 50-457
Enclosure:
As stated cc w/encl: Listserv
December 12, 2013 Mr. M1chael P. Gallagher Vice President, License Renewal Projects Exelon Generation Company, LLC 200 Exelon Way Kennett Square, PA 19348
SUBJECT:
REQUESTS FOR ADDITIONAL INFORMATION FOR THE REVIEW OF THE BYRON NUCLEAR STATION, UNITS 1 AND 2, AND BRAIDWOOD NUCLEAR STATION, UNITS 1 AND 2, LICENSE RENEWAL APPLICATION- AGING MANGEMENT, SET 4 (TAG NOS. MF1879, MF1880, MF1881, AND MF1882)
Dear Mr. Gallagher:
By letter dated May 29, 2013, Exelon Generation Company, LLC, submitted an application pursuant to Title 10 of the Code of Federal Regulations Part 54, to renew operating licenses NPF-37, NPF-66, NPF-72, and NPF-77 for Byron Nuclear Station, Units 1 and 2, and Braidwood Nuclear Station, Units 1 and 2, respectively, for review by the U.S. Nuclear Regulatory Commission staff. The staff is reviewing the information contained in the license renewal application and has identified, in the enclosure, areas where additional information is needed to complete the review.
These requests for additional information were discussed with John Hufnagel, and a mutually agreeable date for the response is within 30 days from the date of this letter. If you have any questions, please contact me at 301-415-3873 or by e-mail at john.daily@nrc.gov.
Sincerely, IRA/
John W. Daily, Sr. Project Manager Projects Branch 1 Division of License Renewal Office of Nuclear Reactor Regulation Docket Nos. 50-454, 50-455, 50-456, and 50-457
Enclosure:
As stated cc w/encl: Listserv DISTRIBUTION:
See next page via email
Letter to M. P. Gallagher from John W. Daily dated December 12, 2013
SUBJECT:
REQUESTS FOR ADDITIONAL INFORMATION FOR THE REVIEW OF THE BYRON NUCLEAR STATION, UNITS t AND 2, AND BRAIDWOOD NUCLEAR STATION, UNITS t AND 2, LICENSE RENEWAL APPLICATION -AGING MANGEMENT, SET 4 (TAG NOS. MFt879, MFt880, MFt88t, AND MFt882)
DISTRIBUTION:
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PUBLIC RidsNrrDir Resource RidsNrrDirRpb1 Resource RidsNrrDirRpb2 Resource RidsNrrDraApla Resource RidsOgcMaiiCenter JDaily LRobinson DMclntyre, OPA JWiebe, DORL EDuncan, Rill JBenjamin, Rill AGarmoe, Rill SRI, Byron Station, Rill JRobbins, Rill VMitlyng, Rill PChandrathll, Rill
BYRON NUCLEAR STATION, UNITS 1 AND 2 AND BRAIDWOOD NUCLEAR STATION, UNITS 1 AND 2 REQUEST FOR ADDITIONAL INFORMATION AGING MANAGEMENT, SET 4 (TAG NOS. MF1879, MF1880, MF1881, AND MF1882)
RAI B.2.1.22-1, ASME Code Class 1 Socket weld failure (036)
Applicability: Byron Nuclear Station (Byron), Units 1 and 2
Background:
Generic Aging Lessons Learned (GALL) Report aging management program (AMP) XI.M35 states under the "detection of aging effects" program element that the one-time inspection program does not apply to plants that have experienced cracking in ASME Code Class 1 small-bore piping due to stress corrosion, cyclical (including thermal, mechanical, and vibration fatigue) loading, or thermal stratification and thermal turbulence. License renewal application (LRA) Section 8.2.1.22 states that Byron and Braidwood Nuclear Station (Braidwood) have not experienced this type of cracking.
Issue:
The applicant documented plant-specific operating experience in the LRA section which states that, in 1998, Byron, Unit 1, experienced a failure of an ASME Code Class 1 socket weld that attached an elbow to a pipe on a safety injection system line. The applicant attributed the failure to a fabrication flaw.
Based on the limited information provided at the audit, the staff determined that the failure could have been caused by vibration fatigue.
Request:
Provide information in terms of metallurgical analysis to support the conclusion that the failure was caused by a fabrication flaw.
If failure of ASME Code Class 1 small-bore piping is identified due to vibration fatigue, provide a plant-specific program that includes periodic inspections, otherwise explain why the one-time inspection program will adequately manage cracking RAI 8.2.1.22-2, Small-bore piping weld sample populations (036)
Applicability: Byron and Braidwood Nuclear Station (Braidwood), all units
Background:
The LRA states that AMP B.2.1.22, "One-Time Inspection of ASME Code Class 1 Small-Bore Piping" is an existing program that is consistent with the program elements in GALL Report AMP XI.M35, **one-Time Inspection of ASME Code Class 1 Small-Bore Piping."
ENCLOSURE
GALL Report AMP XI.M35 states under the "detection of aging effects" program element that
"[t]his inspection should be performed at a sufficient number of locations to ensure an adequate sample. This number, or sample size, is based on susceptibility, inspectability, dose considerations, operating experience, and limiting locations of the total population of ASME Code Class 1 small-bore piping locations."
Issue:
The LRA section does not provide the weld populations. It is not clear to the staff how the inspection sample(s) would be selected.
Request:
Provide the population of in-scope small-bore piping welds for each weld type (i.e., butt welds and socket welds) at each unit.
RAI 8.2.1.6-1, Control rod assembly CASS components (014)
Applicability: Byron and Braidwood Stations
Background:
LRA Section 8.2.1.6 describes the applicant's Thermal Aging Embrittlement of Cast Austenitic Stainless Steel (CASS) Program. LRA Section 8.2.1.6 states that this program is a new condition monitoring program that provides assurance that reactor coolant pressure boundary CASS components (i.e., Class 1 piping and control rod assembly pressure boundary components) susceptible to thermal aging embrittlement meet the specified intended functions.
The aging management review results in LRA Table 3.1.2-2 identify both CASS and non-cast stainless steel as the materials used to fabricate the control rod assembly components, which include latch housing, rod travel housing, cap, and control rod drive mechanism adapter.
However, the LRA does not provide any more specific information on the materials used to fabricate these different components of the control rod assembly (pressure boundary).
By contrast, Section 15.4.8.1.1, "Design Precautions and Protection" of the applicant's updated final safety analysis report (UFSAR) states that the latch mechanism housing and rod travel housing are each a single length of forged Type 304 stainless steel.
Issue.
Given the lack of specific information on the materials of fabrication for the different control rod assembly pressure boundary components, the staff needs to clarify which components of the control rod assembly are made of CASS, so that CASS components are appropriately identified in the scope of the applicant's program.
Request:
Clarify which components of the control rod assembly are made of CASS so that CASS components are appropriately identified in the scope of the applicant's program.
RAI 8.2.1.6-2, Susceptible CASS components with ferrite content greater than 25 percent (014)
Applicability: Byron and Braidwood Stations
Background:
LRA Section 8.2.1.6 states that the applicant's Thermal Aging Embrittlement of Cast Austenitic Stainless Steel (CASS) Program is consistent with GALL Report AMP XI.M12, "Thermal Aging Embrittlement of Cast Austenitic Stainless Steel Program (CASS) " GALL Report AMP XI. M12 states that a flaw tolerance evaluation for components with a ferrite content up to 25 percent is performed according to the principles associated with ASME Code,Section XI, IWB-3640 for submerged arc welds. The GALL Report also states that a flaw tolerance evaluation for piping with greater than 25 percent ferrite is performed on a case-by-case basis by using the applicant's fracture toughness data.
Issue:
The LRA does not address whether the applicant has any susceptible CASS components with a ferrite content greater than 25 percent. In addition, the LRA does not clearly address whether the flaw tolerance evaluation for susceptible CASS components with greater than 25 percent ferrite will be performed on a case-by-case basis in the applicant's program. The staff also needs additional information regarding high-ferrite CASS components and flaw tolerance evaluation for the components.
Request 1 Clarify whether the applicant has any susceptible CASS components with a ferrite content greater than 25 percent.
- 2. If susceptible CASS components with a ferrite content greater than 25 percent are present, provide the following information for the CASS components: (1) component name, (2) casting method and material grade (e.g., centrifugally cast CFB), (3) ferrite contents based on a method consistent with GALL Report AMP XI.M12 and, if existent, actual measurements, (4) clarification as to whether applicant's flaw tolerance evaluation will be performed on a case-by-case basis using relevant fracture toughness data, and (5) applicant's methodology to be used in the flaw tolerance evaluation and the technical basis for the methodology.
RAI 8.2.1.6-3, Operating Experience for control rod assembly and reactor coolant fitting components made of CASS (014)
Applicability: Byron and Braidwood Stations
Background:
LRA Section 8.2.1.6 describes the applicant's Thermal Aging Embrittlement of Cast Austenitic Stainless Steel (CASS) Program. LRA Section 8.2.1.6 states that this program manages loss of fracture toughness due to thermal aging embrittlement of reactor coolant pressure boundary CASS components (i.e., Class 1 piping and control rod assembly components).
Issue:
LRA Section 8.2.1.6 includes a section to discuss operating experience related to the applicant's program. However, the staff noted that this operating experience discussion does not provide operating experience that is specific to the control rod assembly components and reactor coolant fittings made of CASS. The staff needs this information in order to ensure that any previous operating experience related to these CASS components does not identify a need to enhance the applicant's program.
Request:
Provide operating experience specific to the CASS control rod assembly components and reactor coolant fittings, including relevant inspection results.
RAI 8.3.1.1-1, Monitoring of additional plant design transients for fatigue (055)
Applicability: Byron and Braidwood Stations
Background:
The Fatigue Monitoring Program is an existing program that has been monitoring and tracking transients since initial plant startup. LRA Section 8.3.1.1 identifies an enhancement to the "scope of program," "preventive actions," "parameters monitored or inspected," and "acceptance criteria" program elements of the Fatigue Monitoring Program. Specifically, it states that the program will "[m]onitor and track additional plant transients that are significant contributors to component fatigue usage."
GALL Report X.M1, "Fatigue Monitoring," states that the program should monitor all plant design transients that contribute significantly to the fatigue usage factor.
Issue:
For the additional plant transients that will be monitored and tracked as a result of this enhancement, it is unclear to the staff how the applicant ensures that the accumulated transients from initial plant startup will be appropriately accounted for, both in number of cycles and severity, prior to the program/procedure enhancement.
"5" Request:
- 1. Describe the methodology that will be used to identify the additional plant transients that contribute significantly to the fatigue usage factor and justify that it will ensure the program monitors all plant design transients that contribute significantly to the fatigue usage factor.
2_ For the additional transients that will be monitored and tracked, explain how the existing program ensures that cycles from initial plant startup will be appropriately captured, both cycle counts and severity, by the Fatigue Monitoring Program upon entering the period of extended operation.
- 3. Justify that this baseline cycle count will appropriately account for each of the additional transients such that fatigue will be managed during the period of extended operation.
RAI 8.3.1.1-2, Analyses other than ASME Code Section Ill fatigue analyses in the Fatigue Monitoring Program (055)
Applicability: Byron and Braidwood Stations
Background:
LRA Section B.3.1.1 states that the Fatigue Monitoring Program is an existing program that manages cumulative fatigue damage by monitoring and tracking transients. The LRA states that the program will be used to ensure the validity of the ASME Code Section Ill fatigue analyses.
LRA Section 4.7.4 states a Class 1 fracture mechanics analysis was performed on the Byron and Braidwood, Units 1 and 2, residual heat removal heat exchangers tube side inlet and outlet nozzles. LRA Section 4.7.6 states a flaw growth analysis was performed on the Byron, Unit 2, pressurizer seismic restraint lug. LRA Section 4.7.7 states that a crack growth analysis was performed on the Braidwood, Unit 2, feedwater pipe elbow. LRA Section 4.7.8 describes crack growth rate analyses on primary system components. The LRA states that these analyses were performed consistent with ASME Code Section XI and identifies them as plant-specific time-limited aging analyses (TLAAs). The LRA dispositions these TLAAs in accordance with 10 CFR 54.21 (c)(1 )(iii) and credits the Fatigue Monitoring Program to monitor the transient cycles to ensure the transient inputs in the fracture mechanics or fatigue crack growth analyses supporting the flaw evaluations will not be exceeded during the period of extended operation.
Issue:
LRA Section B.3.1.1 states the Fatigue Monitoring Program will be used to monitor transients for ASME Code Section Ill fatigue analyses. It is unclear to the staff whether analyses other than ASME Code Section Ill fatigue analyses are within the scope of the Fatigue Monitoring Program. The flaw evaluations in LRA Sections 4.7.4, 4.7.6, 4.7.7, and 4.7.8 credit the Fatigue Monitoring Program to monitor transient cycles.
Request:
- 1. Clarify whether analyses other than ASME Code Section Ill fatigue analyses are within the scope of the Fatigue Monitoring Program. If so, identify all non-AS ME Code Section Ill
fatigue analyses that credit the Fatigue Monitoring Program to ensure the transient inputs to the analyses will not be exceeded.
- 2. For all non-ASME Code Section Ill fatigue analyses identified in the response above, justify that it is appropriate to credit the Fatigue Monitoring Program to disposition these analyses in accordance with 10 CFR 54.21(c)(1)(iii) such that effects of aging on the intended function(s) of these components will be adequately managed for the period of extended operation.
- 3. Update the program elements of the Fatigue Monitoring Program in LRA Section 8.3.1 1 and UFSAR Section A3.1.1 as necessary, based on the above requests.
RAI 8.2.1.7-1, FMECA for A/LA items 2 and 7 in LRA Appendix C (017)
Applicability: Byron and Braidwood Stations
Background:
LRA Appendix C discusses the applicant's responses to applicanUlicensee action item (A/LA I)
No. 2 and No. 7 of MRP-227 -A. In its responses, the applicant identified two components that were fabricated of different materials than those considered in MRP-227-A. Specifically, the applicant indicated that the upper instrumentation conduit and supports: brackets, clamps, terminal blocks and conduit straps at Bryon and Braidwood were fabricated from CASS rather than forged 304 stainless steel. The applicant stated that due to the material difference in these components a failure mode, effects, and criticality analysis (FMECA) was performed For the upper instrumentation conduit and supports: brackets, clamps, terminal blocks, and conduit straps, installed at Byron and Braidwood, the FMECA determined that, with the inclusion of loss of fracture toughness due to thermal aging embrittlement as a degradation mechanism, the components remained in the "No Additional Measures" inspection category.
Issue:
The staff noted that the details and basis for the applicant's FMECA conclusion were not provided for the upper instrumentation conduit and supports: brackets, clamps, terminal blocks, and conduit straps. The staff noted that this information is necessary to assess whether the applicant will implement an adequate aging management strategy for these components.
The staff also noted that the applicant's response to A/LA I No. 2 focused on how thermal embrittlement was assessed in the FMECA process, but did not provide a discussion on how irradiation embrittlement was considered. It is not clear to the staff if or how irradiation embrittlement was considered in the applicant's FMECA for the upper instrumentation conduit and supports: brackets, clamps, terminal blocks, and conduit straps, insta!led at Byron and Braidwood.
Request Describe in detail the FMECA performed for these components when considering loss of fracture toughness due to thermal and irradiation embrittlement, and justify the conclusion that components were ranked as Category A components, which equates to the "No Additional Measures" inspection category.
RAI 8.2.1.7-2, MRP-227A Analyses for Byron upper support plate assemblies (017)
Applicability: Byron, Units 1 and 2
Background:
LRA Appendix C discusses the applicant's responses to A/LAI No. 2 and No. 7. In its responses, the applicant identified two components that were fabricated of different materials than those considered in MRP-227-A. Specifically, the applicant indicated that the upper support plate assembly- upper support plate, flange and upper support ring at Byron, Units 1 and 2, was fabricated from CASS rather than forged Type 304 stainless steeL The applicant stated that due to the material difference in these components a FMECA was performed.
For the upper support plate assembly: upper support plate, flange, and upper support ring or skirt installed in Byron, Units 1 and 2, the FMECA determined that the upper support plate was "Non-Category A;" thus, further evaluation is required for plant-specific disposition. The applicant explained in its response to A/LAI No. 2 that, based on the certified material test reports (CMTRs) and use of guidance in NRC letter "License Renewal Issue No. 98-0030, Thermal Aging Embrittlement of Cast Austenitic Stainless Steel Components," dated May 19, 2000, the single piece castings, which include the upper support plate at Byron, Units 1 and 2, are not susceptible to loss of fracture toughness due to thermal aging embrittlement. As a result, the applicant determined that the upper support plate was categorized as a "No Additional Measures" component consistent with its original categorization in MRP-227 -A Issue:
The staff noted that the details and bases for the applicant's FMECA and susceptibility analysis conclusion for thermal aging embrittlement were not provided for the upper support plate assembly, which are necessary to assess whether the applicant will implement an adequate aging management strategy.
The staff also noted that the applicant's response to A/LAI No. 2 focused on how thermal aging embrittlement was assessed in the FMECA process, but did not provide a discussion on how irradiation embrittlement was considered. It is not clear to the staff if or how irradiation embrittlement was considered in the applicant's FMECA for the upper support plate assembly:
upper support plate, flange, and upper support ring or skirt installed in Byron, Units 1 and 2.
Request:
Applicable to the upper support plate assembly: upper support plate, flange, and upper support ring or skirt installed in Byron, Units 1 and 2:
- 1. Describe and justify how loss of fracture toughness due to irradiation embrittlement was considered in the FMECA. If irradiation embrittlement was not considered in the FMECA, justify why it was not considered.
- 2. Describe in detail and justify the susceptibility evaluation performed for the upper support plate that utilized the CMTRs and guidance in the NRC letter dated May 19, 2000, to determine that the single piece castings for the upper support plate assembly are not susceptible to thermal aging embrittlement. The justification should
address, but is not limited to, the method in which delta ferrite content was calculated, the calculated delta ferrite content based on the CMTRs, and the consideration of niobium, if applicable_
RAI 8.2.1.7 -4, Evaluating RVI components with existing CUFs for effects of reactor water environment (017)
Applicability: Byron and Braidwood Stations
Background:
LRA Appendix C discusses the applicant's response to NLAI No. 8, Item No. 5. The applicant's response states that the Fatigue Monitoring Program will be enhanced to evaluate the effects of the reactor coolant system water environment on the reactor vessel internal components with existing fatigue cumulative usage factor (CUF) analyses to satisfy the evaluation requirements ASME Code, Section Ill. Subsections NG-2160 and NG-3121. In addition, during its audit, the staff noted that the "scope of program" program element in the applicant's program basis document for the Fatigue Monitoring Program indicates that the resulting CUFen values will not exceed 1.0 for these evaluations.
Issue:
The staff noted that, based on the applicant's response to NLAI No.8, Item No. 5, it is not clear how the applicant is addressing effects of the reactor coolant system water environment on the reactor vessel internal components with existing fatigue CUF analyses. The applicant did not identify the specific approach or method by which the Fatigue Monitoring Program will evaluate the reactor vessel internal components with existing fatigue CUF analyses to address the effects of reactor coolant system water environment.
Request:
- 1. Indicate the reactor vessel internals (RVI) components with existing CUF analyses for which the Fatigue Monitoring Program will evaluate the effects of reactor coolant system water environment and provide the associated material type and CUF value for each component.
- 2. Describe and justify the approach and method that will be used to address the effects of reactor coolant system water environment on the reactor vessel internal components with existing fatigue CUF analyses.
RAI 8.2.1.7-5, Managing aging effects beyond those mentioned in MRP-227-A (017)
Applicability: Byron and Braidwood Stations
Background:
LRA Appendix C provides the pressurized-water reactor (PWR) Vessel Internals Inspection Plan that is outlined in Tables A through D:
- Table A specifies the vessel internal components classified as Primary components and is based on MRP-227-A, Table 4.3.
- Table B specifies the vessel internal components classified as Expansion components and is based on MRP-227-A, Table 4.6.
- Table C specifies the examination acceptance and expansion criteria and is based on MRP-227-A, Table 5.3.
- Table 0 specifies the components that are classified as Existing Program components.
The staff noted that, although LRA Appendix C, Tables A and B, are based on MRP-227-A, they include the management of aging effects that were not identified in MRP-227-A, Tables 4.3 and 4.6. In addition, the staff noted that LRA Appendix C, Table C, provides the "examination acceptance criteria," "expansion criteria," and "additional examination acceptance criteria" for Primary and Expansion components, but only for those aging effects that were identified and evaluated in MRP-227-A, Tables 4.3 and 4.6 Issue:
For example, Table 4-3 of MRP-227-A identifies that the control rod guide tube assembly: guide plates (cards) are managed for loss of material due to wear as a "Primary" component.
However, Table A of LRA Appendix C, identifies that the control rod guide tube assembly:
guide plates (cards) are managed for loss of material, cracking, loss of fracture toughness and changes in dimensions The staff noted that this is only an example and is not the only instance in which the applicant proposed the management of aging effects beyond those discussed in MRP-227 -A. Since the applicant has identified aging effects that were not addressed in MRP-227-A, Tables 4.3 and 4.6, the staff noted that the program may not currently include suitable inspections and proper acceptance and examination criteria to manage these additional aging effects.
The applicant's proposal to manage these additional aging effects not addressed in MRP-227 -A is conservative; however, the staff noted that in order for the applicant's program to adequately manage these additional aging effects it is necessary for the program and inspection plan to establish the appropriate inspection, acceptance and examination criteria.
Request:
For those additional effects that are not addressed in MRP-227-A but are outlined in the Vessel Internals Inspection Plan, establish and justify that appropriate inspections will be performed to adequately manage these additional aging effects. Specifically, consider in the justification that a proper inspection technique is used and appropriate examination acceptance criteria, expansion criteria and additional examination acceptance criteria are established for a particular aging effect.
RAI 8.2.1.7-6, Aging management for RVI clevis insert bolts (017)
Applicability: Byron and Braidwood Stations
Background:
LRA Table 3.1.2-3, Reactor Vessel Internals, indicates that the clevis insert bolts are nickel alloy and that cracking will be managed by the PWR Vessel Internals Program. In addition, the staff noted that TableD in LRA Appendix C indicates that the clevis insert bolts are managed by part of the inspections performed in accordance with ASME Code Section XI, Category 8-N-3.
Appendix A to MRP-227-A indicates that failures of Alloy X-750, precipitation-hardenable nickel-chromium alloy, clevis insert bolts were reported by one Westinghouse-designed plant in 2010 and were suspected to be a result of primary water stress cracking corrosion.
The staff noted that the only aging mechanism requiring management by MRP*227 *A for the clevis insert bolts is wear, and the bolts are categorized as an "Existing Programs" component.
Thus, under MRP*227 *A, the clevis insert bolts will be inspected in accordance with the ASME Code Section Xllnservice Inspection Program to manage the effects due to wear only.
Issue:
The staff noted that the ASME Code Section XI specifies a VT*3 visual inspection for the clevis insert bolts, which may not be adequate to detect cracking before bolt failure occurs. In addition, it is not clear to the staff whether this operating experience is applicable to the applicant and whether the applicant's PWR Vessel Internals AMP will need to be modified to account for this operating experience.
Request
- 1. Specify the fabrication material, including any applicable heat treatment, for the clevis insert bolts at Byron and Braidwood, Units 1 and 2.
- 2. Discuss and justify whether the operating experience associated with cracking of the clevis insert bolts (discussed above) is applicable to Byron and Braidwood Stations, Units 1 and 2.
o If applicable, discuss and justify how your PWR Vessel Internals Program will be augmented to require an inspection of the clevis insert bolts capable of detecting cracking. If the PWR Vessel Internals Program will not be augmented, provide a technical justification for the adequacy of the existing VT*3 visual inspection to detect cracking before it results in clevis insert bolt failure.
RAI 8.2.1.5*1, Missing operating experience from LRA concerning loss of material due to wear in CRDM nozzles (011)
Applicability: Byron and Braidwood Stations
Background:
LRA Section 8.2. 1.5 states that the Cracking of Nickel-Alloy Components and Loss of Material Due to Boric Acid-Induced Corrosion in Reactor Coolant Pressure Boundary Components Program is consistent with GALL Report AMP XI.M11 B, which includes the examinations of ASME Code Case N-729-1 as required by 10 CFR Part 50.55a. During the audit, the staff noted that the applicant performed ultrasonic testing (UT) examination of the CRDM nozzles at Byron, Unit 1, in 2011, in accordance with ASME Code Case N-729-1. During the UT examination, the applicant found that CRDM nozzle nos. 4 and 8 experienced wear as a result of interactions between the CRDM nozzle thermal sleeve centering pads and the CRDM nozzle (wall).
In addition, the staff noted that the applicant's operating experience also indicates that loss of UT data occurred above the J-groove welds on these penetration nozzles because water couplant was not able to make up the gap between the UT blade probe and the CRDM nozzle in these wear areas. The applicant's operating experience further indicates that it was not possible to determine the exact thickness values of the CRDM nozzles in the wear area since the zero-degree UT probe, which could measure the nozzle thickness, could not receive a UT signal due to the water couplant issue described above.
Issue:
LRA Section B.2.1.5 does not describe this operating experience, which indicates a loss of material in the CRDM nozzles due to wear caused by the thermal sleeve centering pads.
Therefore, the staff needs additional information to confirm that this loss of matenal of CRDM nozzles due to wear by the centering pads will be adequately managed for the period of extended operation. In addition, the staff needs to clarify whether and how the applicant resolved the water couplant issue (i.e., loss of UT data due to the absence of water couplant in the gap between the UT probe and the CRDM nozzle near these wear areas during the UT examination)
Request:
- 1. Provide the following baseline information related to the observed wear indications of the CRDM penetration nozzles due to the interactions between the CRDM nozzle thermal sleeve centering pads and the CRDM nozzle walls.
a) The total number of the CRDM penetration nozzles for each unit, and how many head penetration nozzles have been found to have these wear indications for each unit of the Byron and Braidwood Stations.
b) The maximum depth of the wear indications, if measured, in comparison with the CRDM nozzle wall thickness, for each unit.
c) Clarification of whether these wear indications of the nozzles (from the centering pads) are located at reactor coolant pressure boundary locations.
d) The applicant's acceptance basis for the continued operation with the wear indications, including the acceptable wear depth that was determined in the applicant's analysis.
e) Clarification of whether all wear indications are located in the volume or extent of the examination specified in the program (e.g., the volumetric examination of ASME Code Case N-729-1).
- 2. Clarify whether this wear may occur for other types of reactor vessel head nozzles (e.g.,
reactor vessel level indication system penetration nozzles). If so, provide information in response to Request 1 as applied to the other types of reactor vessel head nozzles.
- 3. Describe how the loss of material due to this wear of reactor vessel head penetration nozzles will be monitored and managed. As part of the response, describe the inspection method, scope, and frequency of the examinations for managing loss of material due to wear of these nozzles.
- 4. Clarify whether and how the water couplant issue was resolved (i.e., loss of UT data due to the absence of water couplant in the gap between the UT probe and the reactor vessel head nozzle near the wear locations during the UT examination). As part of the response, describe the extent of loss of the UT data {e.g., the percentage of the UT examination volume that could not be appropriately examined for cracking and loss of material).
If the issue has not been resolved, provide additional information to justify why loss of UT data near the wear locations is acceptable in managing cracking and wear of the reactor vessel head nozzles for the period of extended operation.
- 5. Identify all program enhancements and additional aging management review items as necessary for aging management. In addition, ensure that the LRA is consistent with the applicant's response.
RAI 8.2.1.5-2, Loss of material from thermal sleeves of reactor vessel head nozzles (011)
Applicability: Byron and Braidwood Stations
Background:
As discussed in RAI 8.2.1.5-1, during the audit, the staff noted that the applicant performed UT examinations of the CRDM nozzles at Byron, Unit 1, in 2011, in accordance with ASME Code Case N-729-1. The UT examination found that CRDM nozzle nos. 4 and 8 experienced wear as a result of the interactions between CRDM nozzles and CRDM nozzle thermal sleeves. LRA Table 3.1.2-2 also indicates that the thermal sleeves of reactor vessel head nozzles are subject to loss of material due to wear.
In addition, the staff noted that the thermal sleeves of reactor vessel head nozzles perform the following functions which significantly contribute to safety: {1) shielding the nozzles from thermal transients, (2) providing a lead-in for the rod cluster control assembly (RCCA) drive rods into the CRDM nozzles, and (3) protecting the RCCA drive rods from the head cooling spray cross flow in the reactor vessel upper head plenum region.
Issue.
The staff needs additional information to clarify how the applicant will monitor and manage loss of material due to wear of the reactor vessel head nozzle thermal sleeves.
Request:
- 1. Describe for each unit which reactor vessel head nozzles have a thermal sleeve that is subject to loss of material due to wear.
- 2. In addition, clarify how loss of material due to wear will be monitored and managed for these thermal sleeves. As part of the response, describe the inspection method, scope, and frequency of the examinations for managing the loss of material for the reactor vessel head thermal sleeves.
RAI 8.2.1.19-1, Reactor vessel surveillance withdrawal schedules (032)
Applicability: Byron and Braidwood Stations
Background:
Appendix H to 10 CFR Part 50 states, "Surveillance specimen capsules must be located near the inside vessel wall in the beltline region so that the specimen irradiation history duplicates, to the extent practicable within the physical constraints of the system, the neutron spectrum, temperature history, and maximum neutron fluence experienced by the reactor vessel inner surface."
GALL Report AMP XI.M31, "Reactor Vessel Surveillance Program," recommends that one capsule be withdrawn at an outage in which the capsule receives a neutron fluence of between one and two times the peak reactor vessel wall neutron fluence at the end of the period of extended operation and be tested in accordance with ASTM E185-82, "Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels."
LRA Section 8.2.1.19 describes the applicant's Reactor Vessel Surveillance Program and states:
There were six (6) specimen capsules installed in each Byron and Braidwood Stations (BBS) RPV prior to plant start-up. The capsules contain representative RPV material specimens, neutron dosimeters, and thermal monitors (eutectic alloy). All six (6) specimen capsules have been withdrawn from each of the BBS RPVs. Three (3) specimen capsules from each RPV were tested and the remaining three (3) untested specimen capsules from each RPV are currently stored in the spent fuel pool. Of the three (3) untested specimen capsules from each RPV at least one (1) untested specimen capsule has been irradiated in excess of the projected peak neutron fluence of the associated RPV at the end of the period of extended operation. Capsules that have been withdrawn will be tested as necessary to fulfill the surveillance capsule recommendations contained in ASTM 185-82 as required by 10 CFR Part 50, Appendix H.
By letter dated November 11, 2011 (ADAMS Accession No. ML113050427), Exelon provided additional information regarding the reactor vessel material surveillance program to support a license amendment request dated June 23, 2011 (ADAMS Accession No. ML111790030), for a
measurement uncertainty recapture (MUR) power uprate for Byron and Braidwood. The reactor vessel surveillance capsule withdrawal schedules for Byron and Braidwood are contained in the pressure-temperature limits report (PTLR) for each unit (ADAMS Accession Nos.
ML070680370, ML070240261, and ML071070447 for Braidwood, Units 1 and 2, Byron, Unit 1, and Byron Unit 2, respectively). The neutron fluence values in the PTLRs are consistent with neutron fluence values from the most recent neutron fluence surveillance capsule reports contained in Table 1 Table 1. Neutron Fluence Values for Surveillance Capsule Reports/PTLRs and MUR RAI Response Submittal Dated November 1, 2011 Fast Neutron Fluence (E > 1.0 MeV)
Plant, Unit Capsule ID Capsule Submittal Report/PTLR 11/01/2011 (n/cm 2) (n/cm 2)
Braidwood, 1 w 209x1o 19 19Bx1o 19 Braidwood, 2 w 2.25 X 10 19 2.07xto 19
- ' Byron, 1 w 243x1o 19 226xto 19 Byron, 2 X 230xto 19 2.18xto 19 Issue:
Appendix H to 10 CFR Part 50 provides requirements for reactor vessel surveillance programs.
Changes to the surveillance program require NRC approval prior to implementation and must monitor changes in the fracture toughness properties resulting from the maximum neutron fluence experienced by the ferritic materials in the reactor vessel beltline. In addition, since no exceptions are identified in LRA Appendix B Section B.2.1.19 regarding the surveillance capsule withdrawal schedule, the submittal should be consistent with the GALL Report AMP XI.M31 for testing surveillance specimens from a capsule that has been exposed to a neutron fluence of between one and two times the peak reactor vessel wall neutron fluence at the end of the period of extended operation.
Request:
- 1. Provide an updated surveillance capsule withdrawal schedule for each unit including, but not limited to: identification of the capsule and associated neutron fluence value which will provide test results consistent with the GALL Report recommendation of a neutron fluence exposure of between one and two times the peak reactor vessel wall neutron fluence at the end of the period of extended operation, and identification of a date for the submittal of each summary technical report.
- 2. The neutron fluence values in Table 1 are not consistent. Specifically, the neutron fluence values in the most recently submitted surveillance capsule report for each Byron and Braidwood unit, which are identical to the neutron fluence values in the PTLR surveillance
capsule withdrawal schedules, differ from the values contained in the November 1, 2011 submittal. Provide a basis for the change in neutron fluence values for each unit.