ML102220373
| ML102220373 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 08/26/2010 |
| From: | Gettys E License Renewal Projects Branch 1 |
| To: | Gambhir S Energy Northwest |
| Gettys E, NRR/DLR, 415-4029 | |
| References | |
| TAC ME3058 | |
| Download: ML102220373 (18) | |
Text
UNITED NUCLEAR REGULATORY WASHINGTON, D.C. 20555-0001 August26,2010 Mr. S.K. Gambhir Vice President Technical Services Columbia Generating Station Energy Northwest MD PE04 P.O. Box 968 Richland, WA 99352-0968 REQUEST FOR ADDITIONAL INFORMATION FOR THE REVIEW OF THE COLUMBIA GENERATING STATION, LICENSE RENEWAL APPLICATION FOR FATIGUE MONITORING PROGRAM, TIME-LIMITED AGING ANALYSIS EXEMPTIONS, METAL FATIGUE TIME-LIMITED AGING ANALYSIS, CUMULATIVE FATIGUE DAMAGE, CAST AUSTENITIC STAINLESS STEEL, AND STRUCTURAL (TAC NO ME3058)
Dear Mr. Gambhir:
By letter dated January 19, 2010, Energy Northwest submitted an application pursuant to Title 10 of the Code of Federal Regulations Part 54 (10 CFR Part 54), to renew operating license NPF-21 for Columbia Generating Station, for review by the U.S. Nuclear Regulatory Commission (NRC or the 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. Further requests for additional information may be issued in the future. Items in the enclosure were discussed with Abbas Mostala 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 301-415-4029 or bye-mail at evelyn.gettys@nrc.gov.
Sincerely, Evel Gettys, Project Manager Projects Branch 1 Division of License Renewal Office of Nuclear Reactor Regulation Docket No. 50-397
Enclosure:
As stated cc w/encl: Distribution via listserv REQUEST FOR ADDITIONAL INFORMATION FOR THE REVIEW OF THE GENERATING STATION, LICENSE RENEWAL APPLICATION FOR FATIGUE PROGRAM, TIME-LIMITED AGING ANALYSIS EXEMPTIONS, METAL FATIGUE LIMITED AGING ANALYSIS, CUMULATIVE FATIGUE DAMAGE, CAST STAINLESS STEEL, AND STRUCTURAL (TAC NO RAI B.2.24-02 Background License renewal application Section B.2.24 identifies the Columbia Fatigue Monitoring Program as an existing plant monitoring program that. with enhancement, will be consistent with the program element criteria in Generic Aging Lessons Learned (GALL) aging management program (AMP) X.M1, "Metal Fatigue of Reactor Coolant Pressure Boundary." The scope of the AMP includes both nuclear steam supply system (NSSS) and non-NSSS components and transients in Updated Final Safety Analysis Report (UFSAR) Section 3.9 that are required to be pursuant to tracking requirements in Technical Specification 5.5.5. The program uses plant transient cycle counting activities to ensure that the cumulative number occurrences (cycles) for each transient will remain below the analyzed number of cycles for the transient, or else to ensure that appropriate corrective action or actions will be taken if the design cycle occurrence limit on a given transient is approached or if the limit on a given component CUF value is approached.
AMP B.2.2.4 states that the program will be enhanced to account for the impact of environmental effects on the cumulative usage factor (CUF) values for component locations that are within the scope of the applicant's environmentally-assisted fatigue analysis (refer to LRA Section 4.3.5). LRA Table 4.3-2 provides the 60-year cycle projections (cycle occurrence projections) for Columbia Generating Station (Columbia) design basis transients.
Issue LRA Section 4.3 indicates that the scope of AMP B.2.24 includes both those transients that are within the scope of UFSAR Section 3.9 and additional transients that are outside the scope of the transients listed in UFSAR Section 3.9. However, LRA Section B.2.24 does not identify which UFSAR Section 3.9 based transients and transients outside of the scope of UFSAR Section 3.9 are within the scope of AMP B.2.24 (Le. "scope of program" element or the "parameters monitored/inspected" element.)
Request Identify all UFSAR-defined and non-UFSAR defined transients (either directly or by reference to transients in applicable sections in LRA Section 4.3 or in UFSAR Sections 3 or 5) that are within the "scope of program" and "parameters monitored/inspected" program elements.
Justify any differences between the transients that are within the scope of the program and those that are defined for Columbia in UFSAR Section 3.9 or that are given and analyzed for in LRA Table 1 or 4.3-2. Clarify and justify whether or not operaHonal basis earthquake (OBE) transients need to be within the scope of the Fatigue Monitoring Program. RAI B.2.24-03 Background LRA Section B.2.24 identifies the Columbia Fatigue Monitoring Program as an existing plant monitoring program that, with enhancement, will be consistent with the program element criteria
-in GALL AMP X.M1, "Metal Fatigue of Reactor Coolant Pressure Boundary." The scope of the AMP includes both NSSS and non-NSSS components and transients in UFSAR Section 3.9 that are required to be pursuant to tracking requirements in Technical Specification 5.5.5. The program uses plant transient cycle counting activities to ensure that the cumulative number occurrences (cycles) for each transient will remain below the analyzed num ber of cycles for the transient, or else to ensure that appropriate corrective action or actions will be taken if the design cycle occurrence limit on a given transient is approached or if the limit on a given component CUF value is approached.
AMP B.2.2.4 states that the program will be enhanced to account for the impact of environmental effects on the CUF values for component locations that are within the scope of the applicant's environmentally-assisted fatigue analysis (refer to LRA Section 4.3.5). LRA Table 4.3-2 provides the 60-year cycle projections (cycle occurrence projections) for Columbia design basis transients.
Issue The transient cycle projection data in LRA Table 4.3-2 indicate the cycle counts for some design basis transients may exceed their design limits prior to the expiration of the period of extended operation or even prior to the expiration of the current operating period. Thus, the relationship of the cycle projection data in LRA Table 4.3-2 and the program elements for AMP B.2.2.4 does not clarify whether corrective actions on the cycle counting limits would need to be implemented as part of the enhance program that will be implemented during the period of extended operation or under the existing program that is being implemented during the current period of operation.
Request For those transients in LRA Table 4.3-2 that are projected to exceed their design basis limits, clarify whether the design basis limit for the transient is projected to be exceeded during the current licensed operating period for the facility or during the period of extended operation.
Clarify whether corrective actions on cycle counting will need to be implemented under the existing protocols for the program (I.e., under the program that is currently being implemented for the current licensed operating period) or under the elements of the enhanced program that will be implemented during the period of extended operation.
If a cumulative number of occurrences for a given transient are projected to exceed the allowable during the current operating period, clarify whether this has been brought to the attention of the appropriate plant engineering department for potential disposition under the program's existing (current) program element activities and criteria.
RAI 8.2.24-04 Background LRA Section 6.2.24 identifies the Columbia Fatigue Monitoring Program as an existing plant monitoring program that, with enhancement, will be consistent with the program element criteria in GALL AMP X.M1, "Metal Fatigue of Reactor Coolant Pressure Boundary." The scope of the AMP includes both NSSS and non-NSSS components and transients in UFSAR Section 3.9 that are required to be pursuant to tracking requirements in Technical Specification 5.5.5. The program uses plant transient cycle counting activities to ensure that the cumulative number occurrences (cycles) for each transient will remain below the analyzed number of cycles for the
-transient, or else to ensure that appropriate corrective action or actions will be taken if the design cycle occurrence limit on a given transient is approached or if the limit on a given component CUF value is approached.
AMP B.2.2.4 states that the program will be enhanced to account for the impact of environmental effects on the CUF values for component locations that are within the scope of the applicant's environmentally-assisted fatigue analysis (refer to LRA Section 4.3.5). LRA Table 4.3-2 provides the 60-year cycle projections (cycle occurrence projections) for Columbia design basis transients.
Issue The loading cycles due to OBE at rated operating conditions have been excluded from LRA Table 4.3-3 which gives the analyzed cycles and projected cycles for future analyses.
Footnote "d" of LRA Table 4.3-1 states that the OBE event includes 50 peak OBE cycles for NSSS piping and 10 peak OBE cycles for other NSSS equipment and equipment.
It further states that 50 peak OBE cycles are postulated for all balance of plant piping and components.
This information is also included in relevant subsections of FSAR Section 3.9.1. Request Explain why the loading cycles due to OBE event at rated operating conditions have been excluded from the analyzed cycles and projected cycles for future analyses listed in LRA Table 4.3-3. RAJ 8.2.24-05 The "acceptance criteria" program element for AMP B.2.24 is used to ensure that the number of transient occurrences for a given plant transient will remain below design limit for transient, as defined in Section 3.9 of the UFSAR, or else that appropriate corrective action be taken if the limit on transient occurrences or a components CUF value is LRA Section B.2.24 does not provide any details regarding the action limits that are set design basis transient cycle counting activities or on CUF monitoring activities, or the actions that will be implemented if an action limit of cycle counting or CUF monitoring reached. The staff has noted that the time-limited aging analysis (TLAA) in LRA Section sets the design basis allowable on normal CUF values and environmentally-adjusted values for NUREG/CR-6260 equivalent or bounding locations to a value of 1.0 but sets design basis allowable for high energy line break locations to a value of Identify the "action limits" that are within the scope of the program's "acceptance program element. Specifically, define the "action limit or limits" that will be used for program's cycle counting activities and for the program's CUF monitoring activities (1) design basis CUF values for Class 1 components and any non-Class 1 components evaluated to Class 1 component CUF requirements, (2) environmentally-assisted CUF for the program's NUREG/CR-6260 equivalent or bounding locations, and (3) for Class 1 components that are within the scope of the applicant high energy line break analyses for Class 1
-components.
Clarify those corrective actions that will, with certainty be implemented if an action limit on cycle counting or CUF monitoring is reached, and those additional corrective options that may be implemented in addition to the mandatory corrective actions for the AMP if an action limit on cycle counting or CUF monitoring is reached. RAI Cumulative Fatigue Damage AMR Background LRA Sections 3.2.2.2.1, 3.3.2.2.1, and 3.4.2.2.1 address cumulative fatigue damage in engineered safety features (ESF) systems, auxiliary (AUX) systems, and steam and power conversion (SPC) systems, respectively.
LRA Sections 3.2.2.2.1, 3.3.2.2.1, and 3.4.2.2.1 identify that the TLAA for non-Class 1 components is addressed separately in LRA Section 4.3.4. The GALL Report includes the aging management review (AMR) items on management of cumulative fatigue damage. Examples of these GALL AMR items are as follows: GALL AMR item V.D2-32, for management of cumulative fatigue damage in the piping, piping components and piping elements of the emergency core cooling systems GALL AMR items VII.E3-14 and VII.E3-17, for management of cumulative fatigue damage in the piping, piping components, piping elements, and heat exchanger components of the reactor water cleanup system GALL AMR item VIII.B2-5 and VIII.D2-6, for management of cumulative fatigue damage in the piping, piping components, piping elements of the main steam and feedwater systems Issue LRA Section 4.3.4 states that the non-Class 1 AMRs for Columbia determined piping locations susceptible to fatigue. The staff noted that none of these locations or components are identified in the LRA and AMR line items. LRA Table 3.2.1 Item 3.2.1-01, LRA Table 3.3.1 Item 3.3.1-02, and LRA Table 3.4.1 Item 3.4-01 identify that cumulative fatigue damage is an aging effect requiring management (AERM) for applicable non-Class 1 piping, piping components, piping elements, and in some cases for applicable heat exchanger components.
The staff has noted that, although, LRA Sections 3.2.2.2.1, 3.3.2.2.1, and 3.4.2.2.1, states that fatigue TLAA analyses are required to be evaluated in accordance with 10 CFR 54.21 (c), the LRA does not include any applicable AMR items for non-Class 1 piping, piping components, piping elements, and in some cases for applicable heat exchanger components managed for cumulative fatigue damage. Request Justify why LRA Tables 3.2.2-1 -3.2.2-5 for ESF subsystem components, LRA Tables 3.3.2-1 3.3.2-44 for AUX subsystem components, and in LRA Tables 3.4.2-1 -3.4.2-7 for SPC subsystem components do not include any AMR items related to TLAA for managing cumulative fatigue damage in the steel and stainless steel piping, piping components, and piping elements, and possibly in applicable heat exchanger components.
-5 RAI4.1-1 Background LRA Section 4.1.3 states that pursuant to 10 CFR 54.21 (c)(2), an applicant for license renewal must provide: (1) a listing of plant-specific exemptions granted to 10 CFR 50.12 that are in effect and based on a TLAA, and (2) an evaluation of these exemptions to justify their continuation for the period of extended operation.
The applicant stated that the current licensing basis documentation, identified in Section 4.1.1, was reviewed and there were no exemptions identified that are based on a TLAA. Issue Columbia facility operating license, No. NPF-21, issued December 20, 1993, states in part, "Exemptions from certain requirements of Appendices G, H, and J to 10 CFR Part 50, are described in the Safety Evaluation Report. These exemptions are authorized by law and will not endanger life or property or the common defense and security and are otherwise in the public interest.
Therefore these exemptions are hereby granted pursuant to 10 CFR 50.12. With the granting of this exemption the facility will operate, to the extent authorized herein, in conformity with the application, as amended, the provisions of the Act, and the rules and regulations of the Commission." The staff is unable to determine if exemptions to the requirements of Appendices G, H, and J to 10 CFR Part 50 exist or whether these exemptions are still in effect and are based on a TLAA that will be needed for the period of extended operation.
Request Clarify the exemptions to the requirements of Appendices G, H, and J to 10 CFR Part 50 and clarify whether these exemptions are still in effect and whether they are based on a TLAA. If they are in effect and based on a TLAA, justify continuation of the exemptions for the period of extended operation.
RAI4.3-01 Background LRA Table 4.3-3 provides the design basis CUF values of record for the limiting reactor pressure vessel (RPV) locations.
UFSAR Section 3.9 and UFSAR Table 3.9-2a provides the UFSAR design basis CUFs for these component locations.
Issue The staff has noted that some of these CUF values reported for RPV components in LRA Table 4.3-3 are consistent with those listed for the components in Columbia UFSAR Table 3.9-2a and some are not. The staff has noted that the CUF value listed for feedwater (FW) nozzle safe end in LRA Table 4.3-3 is 0.696 while the UFSAR Table 3.9-2a lists the value as 0.966. It is not clear if some of the CUF of record have been reanalyzed, and if they were, it is not clear whether lower CUF were obtained by decreasing the projected number of load cycles and/or decreasing the severity of the transient.
Also, if severity of the transient were decreased, it is not clear whether the revised transients were verified by plant-specific stress-based fatigue monitoring.
Request Provide your basis why LRA Table 4.3-3 lists a different CUF value for the FW nozzle safe end from that reported for the component in UFSAR Table 3.9-2a. Specifically, justify why the CUF
-6 value listed for the FW nozzle safe end in LRA Table 4.3-3 is 0.696 while the UFSAR Table 2a lists the value as 0.966. If the CUF value listed for the FW nozzle in LRA Table 4.3-3 represents the most updated design basis value for the component, reference the document in the CLB that provides the design basis CUF of record for the FW nozzle component and clarify the changes in the CUF calculation assumptions or bases that resulted in a 0.27 drop in the design basis CUF value for the components (Le. from a value of 0.966 to a value 0.696). RAI4.3-02 Background Several Boiling Water Reactor Vessel Internals Project (BWRVIP) documents credited for Columbia license renewal have NRC safety evaluation reports (SERs) with associated license renewal applicant action items (AAls). The applicant provides a plant-specific response for each of these AAls is provided in Appendix C of the LRA. The applicant's response to AAI No. 4 from BWRVIP-47-A, BWR Lower Plenum Inspection and Flaw Evaluation Guidelines states, "Due to fatigue of the subject safety-related components, applicants referencing the BWRVIP-47 report for LR should identify and evaluate the projected CUF as a potential TLAA issue." In LRA Table C-4, the applicant also states that the TLAAs identified for the lower plenum are the CUF analyses and results for the control rod drive (CRD) housings, CRD stub tubes, and incore housing penetrations, and that these are addressed in LRA Section 4.3.1 (Table 4.3-3). Issue The staff confirmed that the CUF values for the CRD housings and CRD stub tubes are included in the fatigue analyses of the RPV components and the values are listed in LRA Table 4.3-3. However, neither LRA Table 4.3-3 nor LRA Table 4.3-4 provides any CUF value for the incore housing penetrations.
Request Clarify whether the CUF value for stub tubes listed in LRA Table 4.3-3 is the CUF value that is identified in the LRA for incore housings.
If CUF value in LRA Table 4.3-3 for the stub tubes is not the CUF value for the incore housings, identify what the design basis CUF value of record is for the incore housing penetrations and reference the design basis document of record that provides the design CUF value for this component.
RAI4.3-03 Background LRA Section 4.3.2.2 states that in August 2000. Columbia operated for a period of time with the recirculation pumps in an unbalanced mode (i.e .. the running speeds for the pumps differed by more than 50%). The LRA states that the effect of the flow imbalance resulted in a 0.0035 increase in the CUF value for the plant's jet pumps. The LRA also indicates that inspections of the jet pumps in 2001 identified gaps in the jet pump set screws, and that a fatigue analysis of the jet pump risers through operations as of end of cycle 16 indicated an additional 0.119 increase in the CUF value for jet pumps 1 and 6 (including risers 1 and 2 for jet pump 1 and risers 5 and 6 for jet pump 6) due to the gaps in the component configuration.
In addition, the LRA indicates that in 2005, the applicant installed clamps on the jet pump mixer and diffuser
-7 areas in order to minimize flow-induced vibrations caused by leakage at the mixer-to-diffuser slip joint interface.
LRA Section 4.3.2.2 credits the cycle counting activities to disposition the CUF TLAA for the jet pump assemblies in accordance 10 CFR 54.21 (c )(1 )(iii). In LRA Section 4.3.2.2, the applicant states that Columbia uses the BWRVIP to inspect for cracking and gaps (changes in configuration) in applicable jet pump assembly components.
Issue The Columbia Reactor Vessel Internals Program includes inspections of jet pump assembly components consistent with the inspection criteria of Report BWRVIP-41.
The scope of inspections in BWRVIP-41 include inspections of the jet pump risers, riser braces, and mixers for cracking and changes in configuration (including set screw gaps in applicable jet pump components).
LRA Section 4.3.2.2 already states that the Columbia Reactor Vessel Internals Program is credited with inspections of the applicable jet pump assembly components.
Request Provide your basis for using cycle counting under the Metal Fatigue of Reactor Coolant Pressure Boundary Program to disposition the CUF value for the jet pump assembly components, and why it would not be more appropriate to credit the inspections of the jet pump assembly components under Columbia Reactor Vessel Internals Programs as the basis for dispositioning the TLAA for these components and for managing cumulative fatigue damage/cracking by fatigue in these components in accordance with 10 CFR 54.21(c)(1)(iii).
RAI4.3-04 Background The UFSAR Table 3.9-1 lists the design basis transients and design limits for these transients that are applicable to the Columbia NSSS components (including RPV assembly components, core supportlRPV internal components, and Class 1 reactor coolant pressure boundary piping components) and to non-NSSS (balance of plant) components, with the those for the CRDs and the CRD housings and incore housings.
UFSAR Section 3.9.1.1 provides the design basis transients that are applicable to the CRDs and design limits for these transients, and UFSAR Section 3.9.1.2 provides the design transients that are applicable to the CRD housings and incore housings.
Issue The staff confirmed that LRA Tables 4.3-1 and 4.3-1 accurately reflect the design basis transients that are listed in UFSAR Table 3.9-1 as being applicable to the NSSS components and non-NSSS components, and design limits for these transients.
However, the staff has noted that the LRA does not give corresponding tables for the transients and design limits that are applicable to the CRD nozzles and to the CRD housings and incore housings, as given in UFSAR Sections 3.9.1.1 and 3.9.1.2, respectively.
Request For the NSSS and non-NSSS components that were included in LRA Tables 4.3-1 and 4.3-2, as based on the design basis transients and design limits for the transients in UFSAR Table 3.9-1.
-Provide your basis for not including design basis transient cycle and 60-year design basis cycle projection tables in LRA Section 4.3 for the CRDs that are based on the design basis transients and design limits for these transients in UFSAR Section 3.9.1.1.1. Provide your basis for not including design basis transient cycle and 60-year design basis cycle projection tables in LRA Section 4.3 for the CRD housings and incore housings that are based on the design basis transients and design limits for these transients in UFSAR Section 3.9.1.1.2.
RAI4.3-05 Background LRA Section 4.3.5 summarizes the evaluation of the CUF analyses that comprise the "Effects of Reactor Coolant Environment on Fatigue Life of Components and Piping," for the period of extended operation.
The applicant stated that although environmentally-assisted fatigue evaluations are not part of the existing design basis, these evaluations were performed for the 60-year operation period using the projected cycles from the Fatigue Monitoring Program, and the methodology described in Standard Review Plan-License Renewal (SRP-LR) Sections 4.3.2.2 and 4.3.3.2. The applicant further stated that the original fatigue usage calculations were reviewed, and the transient groupings and load pairs used in those analyses were carried over to the environmentally-assisted fatigue analyses.
The applicant added that for each load pair, an environmental correction factor (Fen) was calculated, and environmentally-assisted CUF factor for the load pair was obtained by multiplying the design basis CUF factor for the component factor by Fen factor that was determined for the component.
The environmentally assisted CUF for the location was obtained by summing the individual environmentally assisted usage factors for each load pair. The environmentally assisted CUF and the minimum and maximum Fen for any load pair for the 14 locations in six components are shown in Table 4.3-6. Issue The design basis "CUF in air" values listed for the limiting environmental fatigue components in LRA Table 4.3-6 are different from and usually lower than the design basis CUF values listed for the same component in either LRA Table 4.3-3 or 4.3-5. Specifically, LRA Table 4.3-6 Footnote 2 states that the "Revised CUF in air" value for the component in LRA Table 4.3-6 is the "CUF of record previously identified in Table 4.3-3 and Table 4.3-5." This implies that the design basis CUF value listed in LRA Table 4.3-6 for a given component should be exactly the same as the design basis CUF value that is reported for the component in either LRA Table 4.3-3 or 4.3-5. Request For each component location listed in LRA Table 4.3-6, provide the basis why LRA Table 4.3-6 reports design basis CUF values for the component locations that are different from (and are usually lower than) the CUF values reported for the component locations in either LRA Table 4.3-3 or 4.3-5, particularly when Footnote 2 states that the "Revised CUF in air" value for the component in LRA Table 4.3-6 is the "CUF of record previously identified in Table 4.3-3 and Table 4.3-5." For each environmentally-assisted fatigue location given in LRA Table 4.3-6, clarify which design basis CUF value for each of components (Le., the value as reported in Table 4.3-3 or 4.3-5 or the value reported in LRA Table 4.3-6) is the design basis value of
-record for component and reference the Columbia document of record that establishes the value as the current design basis CUF value for the component.
Clarify that changes that were made in the CUF calculation for a component if the design basis CUF value reported in LRA Table 4.3-6 represents the most up-to-date design basis value of record. RAI4.3-06 Background The design load cycles used in the fatigue analyses of reactor vessel internals are the same as those for RPV components.
The design load cycles are listed in LRA Table 4.3-1 and the projected cycles to 60 years are presented in LRA Table 4.3-2. In addition.
the CUFs for the limiting reactor vessel support structures and vessel internals are summarized in LRA Table 4. These CUF values, except for the jet pump riser brace, are based on the original design cycles for 40-year operation.
The CUF for the jet pump riser brace has been conservatively projected to 60-year operation.
Issue The staff has noted that the Columbia environmentally-assisted fatigue analysis does not always apply the NUREG/CR-6260 methodology to the RPV or Class 1 piping components that have the highest design basis CUF values of record. For example, in the environmental assessment in LRA Table 4.3-6 identifies that the CRD tube and CRD housing were selected as the representative locations for the RPV shell and lower head, and that the design basis CUF values for these component locations were 0.083 and 0.196, respectively.
However. LRA Table 4.3-3 lists the shroud support (0.399), main steam nozzle shell (0.47), or low-pressure coolant injection (LPCI) thermal sleeve (0.430) as all have existing design basis CUF values that are greater than those reported for the CRD tubes and CRD housings in LRA Table 4.3-3. Request Provide your basis for selecting the RPV and Class 1 piping locations that were chosen as the environmentally-assisted fatigue analysis locations for the LRA, as given in LRA Table 4.3-6. Provide your basis for not selecting the core shroud supports, main stem shell nozzles, and LPCI nozzle thermal sleeves as additional environmentally-assisted fatigue assessment locations.
RAI4.3-07 Background LRA Section 4.3.5 indicates that an effective Fen methodology that is based on a time-weighted average of normal water chemistry (NWC) and hydrogen water chemistry (HWC) operations over a cumulative 60 year operating period (Le., 20.9 years at NWC and 39.1 years and HWC, respectively) was used to determine environmentally assisted Fen factors that were used in the environmentally-assisted fatigue calculations (Le. for the environmental CUF calculations).
A footnote in LRA Table 4.3-6 indicates that the dissolved oxygen in the FW nozzle safe end and nozzle to shell junction locations was assumed to be 150 ppb.
-Issue According to BWRVIP-130 "BWR Vessel and Internals Project BWR Water Chemistry Guidelines
-2004 Revision," the operating range for dissolved oxygen is30-200 ppb for NWC and 30-100 ppb for HWC. However, LRA Section 4.3.5 does not give any details regarding the dissolved oxygen concentration values for implementation of NWC and HWC conditions that were derived and applied to the Fen calculation methodology, or the basis for deriving the dissolved oxygen values. Request For each component location listed in LRA Table 4.3-6 (other than the FW nozzle), provide the dissolved oxygen concentration inputs under implementation of NWC and HWC operating conditions that were used in the calculation of the Fen values for the components, and clarify how these dissolved oxygen inputs were derived and why they are considered to be conservative for application to the Fen methodology.
RAI4.3-08 Background LRA Section 4.3.5 indicates that an effective Fen methodology that is based on a time-weighted average of NWC and HWC operations over a cumulative 60 year operating period (Le., 20.9 years at NWC and 39.1 years and HWC, respectively) was used to determine environmentally assisted Fen factors that were used in the environmentally-assisted fatigue calculations (i.e. for the environmental CUF calculations).
A footnote in LRA Table 4.3-6 indicates that the dissolved oxygen in the FW nozzle safe end and nozzle to shell junction locations was assumed to be 150 ppb. Issue The LRA does not provide the basis for assuming a 150 ppb dissolved oxygen concentration value for the FW nozzle. The staff presumes that the 150 ppb dissolved oxygen concentration value for the FW nozzle is the value under implementation of NWC; however, this is not specifically evident from the contents of the LRA. Request Justify your basis for assuming a 150 ppb dissolved oxygen concentration value for the FW nozzles and clarify whether this value represents the value for operations under NWC conditions or HWC conditions.
RAI3.1.2.1-X1 Background Columbia LRA Table 3.1.1, item 3.1.1-55 addresses cast austenitic stainless steel (CASS) Class 1 pump casings and valve bodies and bonnets exposed to reactor coolant> 250°C, which are subject to loss of fracture toughness due to thermal aging embrittlement.
The LRA item also indicates that the applicant uses the ASME Section XI Inservice Inspection, Subsections IWB, IWC and IWD Program to manage the aging effect for the CASS Class 1 components.
The LRA item further states that reduction of fracture toughness for CASS valve bodies less than 4 inches is included in this item and managed by the Small Bore Class 1 Piping Inspection
-11 Program. In LRA Table 3.1.2-3, the AMR line item with Row Number 134 also indicates that reduction of fracture toughness in CASS "Valve Bodies < 4 inches" exposed to reactor coolant is managed by the Small Bore Class 1 Piping Inspection Program. Issue In comparison with the applicant's use of the Small Bore Class 1 Piping Inspection Program, the GALL Report, under item IV.C1-3, recommends the ASME Section Xllnservice Inspection, Subsections IWB, IWC and IWD Program to manage loss of fracture toughness due to thermal aging embrittlement for CASS Class 1 pump casings and valve bodies and bonnets exposed to reactor coolant (>250°C).
Therefore, the staff found the need to further clarify whether the applicant's aging management approach is consistent with the GALL Report as claimed in LRA Table 3.1.1, item 3.1.1-55.
The staff also noted that the 2001 edition of the ASME Code Section XI with 2002 and 2003 addenda, referenced by GALL AMP XI.M1, "ASME Section XI Inservice Inspection, Subsections IWB, IWC and IWD," requires that valve body welds in valves less than normal pipe size (NPS) 4 should be examined using surface examination in accordance with Table IWB-2500-1 (Examination Category B-M-1, Item No. B12.30). The staff finds that the ASME Code Section XI requirement is applicable to the applicant's line item Row No. 134 in LRA Table 3.1.2-3 because the valve bodies are less than 4 inches. Therefore, the staff found the need to clarify whether the CASS valves are less than 4 inches, under the line item with Row No. 134 in Table 3.1.2-3, includes a weld (including weld repair) in the valve bodies and, if a weld is included in the valve bodies, to clarify whether periodic inspections are performed for the valve body weld in a consistent manner with the ASME Section XI Table IWB-2500-1 requirements and the recommendation of the GALL Report. The staff also found the need to further clarify why the applicant's aging management for the CASS valves less than 4 inches is adequate to manage reduction in fracture toughness.
Request Clarify whether the CASS Class 1 valves less than 4 inches, under LRA Table 3.1.2-3 line item Row No. 134, have a valve body weld including weld repair and clarify, if the CASS valves have a valve body weld, whether periodic inspections are performed on the valves in accordance with the ASME Code Section XI requirements. Taking into consideration the requirement for the surface examination of the valve body weld described in ASME Code Section XI Table IWB-2500-1, clarify the justification why the Small Bore Class 1 Piping Inspection Program, which uses a one-time inspection rather than periodic inspections and is different from the ASME Section XI Inservice Inspection Program recommended in the GALL Report, is adequate to manage reduction of fracture toughness for the CASS Class 1 valves less than 4 inches. RAI3.1.2.1-X2 Background Columbia LRA Table 3.1.1, item 3.1.1-48 addresses steel and stainless steel Class 1 piping, fittings and branch connections less than NPS 4 exposed to reactor coolant, which are subject
-to cracking due to stress corrosion cracking, intergranular stress corrosion cracking and thermal and mechanical loading. The LRA item also indicates that the applicant uses the BWR Water Chemistry Program and Small Bore Class 1 Piping Inspection Program to manage the aging effect. In LRA Table 3.1.2-3, AMR line items with Row Numbers 130 and 131 also indicate that the BWR Water Chemistry Program and Small Bore Class 1 Piping Inspection Program are used to manage cracking due to stress corrosion cracking and intergranular attack for cast CASS valve bodies less than 4 inches exposed to reactor coolant. Issue In comparison with the applicant's use of the BWR Water Chemistry Program and Small Bore Class 1 Piping Inspection Program, the GALL Report, under item IV.C1-1, recommends the ASME Section XI Inservice Inspection.
Subsections IWB, IWC and IWD Program, Water Chemistry Program and One-Time Inspection of ASME Code Class 1 Small-bore Piping Program to manage the cracking due to stress corrosion cracking, intergranular stress corrosion cracking and thermal and mechanical loading. Therefore.
the staff found the need to further clarify whether the applicant's aging management approach, which does not use the ASME Section XI Inservice Inspection, Subsections IWB. IWC and IWD Program, is consistent with the GALL Report as claimed in LRA Table 3.1.1, item 3.1.1-48.
The 2001 edition of the ASME Code Section XI with 2002 and 2003 addenda, referenced by GALL AMP XI.M1, "ASME Section Xllnservice Inspection, Subsections IWB, IWC and IWD," requires that valve body welds in valve bodies less than NPS 4 should be examined using surface examination in accordance with Table IWB-2500-1 (Examination Category B-M-1, Item No. B12.30). The staff finds that this ASME Code Section XI requirement is applicable to the applicant's line items with Row Numbers 130 and 131 in LRA Table 3.1.2-3 because the valve bodies are less than 4 inches. Therefore, the staff found the need to clarify whether the applicant's aging management for the CASS valves less than 4 inches includes a weld in the valve bodies including weld repair and, if a weld is included in the valve bodies, to clarify whether surface examinations are performed for the valve body weld in a consistent manner with the ASME Section XI Table IWB-2500-1 requirements and the recommendation of the GALL Report. The staff also finds the need to further clarify why the applicant's aging management for the CASS valves less than 4 inches is adequate to manage the cracking due to stress corrosion cracking, intergranular stress corrosion cracking and thermal and mechanical loading. Request Clarify whether the CASS valves less than 4 inches, under LRA Table 3.1.2-3 line items with Row Numbers 130 and 131 and LRA Table 3.1.1 item 3.1.1-48, have a valve body weld including weld repair and clarify. if the CASS valves have a valve body weld, whether periodic inspections are performed on the valves in accordance with the ASME Code Section XI requirements. Taking into consideration the surface examination requirement for the valve body weld described in ASME Code Section XI Table IWB-2500-1, clarify the justification why the Small Bore Class 1 Piping Inspection Program, which is based on a one-time inspection rather than periodic inspections and is different from the ASME Section XI Inservice Inspection Program recommended in the GALL Report, is adequate to manage the
-cracking due to stress corrosion cracking, intergranular stress corrosion cracking and thermal and mechanical loading. RAI 3.1.2.3-02 Background In LRA Table 3.1.2-3, AMR line item with Row Number 129 indicates that CASS valve bodies less than 4 inches exposed to reactor coolant is subject to an aging effect of "Cracking
-Flaw Growth" and the aging effect is managed by the Small Bore Class 1 Piping Inspection Program. The LRA line item cites generic note H indicating that the aging effect is not addressed in the GALL Report for this component, material and environment combination.
Issue The staff noted that the aging effect, "Cracking
-Flaw Growth," suggests the possibility that a flaw already exists in the CASS valve bodies. However, LRA Section B.2.49 indicates that the Small Bore Class 1 Piping Inspection Program use a one-time inspection approach rather than periodic inspections.
Therefore, the staff found the need to clarify whether the components have an existing flaw. And, if a pre-existing flaw exists the staff would like the applicant to clarify why the Small Bore Class 1 Piping Inspection Program, which is based on a one-time inspection rather than periodic inspections, is adequate to manage "Cracking
-Flaw Growth." Request Clarify whether the components have an existing flaw. If a flaw exists in the components, clarify what aging mechanism(s) caused the flaw and clarify why the Small Bore Class 1 Piping Inspection Program, which is based on a time inspection rather than periodic inspections, is adequate to manage "Cracking Flaw Growth" for the components.
RAI 3.3.2.3.1-01 Background Several GALL Report line items (e.g. II I.A6-6 , III.A6-7, and III.A6-8) discuss aging effects of concrete in a water-flowing environment and suggest AMPs to manage the effects. Issue LRA Table 3.3.2-1 states that there are no AERMs for concrete piping of the Circulating Water System that is exposed to a raw water (internal) environment and that an AMP is not required.
The staff is unclear why an AMP is not credited for management of aging of concrete piping components associated with this line item since they are subjected to a water-flowing environment.
Several aging effects exist for concrete exposed to a flowing water environment, as discussed in the GALL Report line items cited above. Request Explain how aging of these components will be managed, or justify why an AMP is not required for the concrete piping exposed to a raw water (internal) environment.
RAI3.3.2.3.1-02 Background Several GALL Report line items (e.g. III.A6-2, III.A6-3, and III.A6-4) discuss aging effects of concrete in a soil environment and provide AMPs to manage the effects. Issue LRA Table 3.3.2-1 states that there are no AERMs for concrete piping of the Circulating Water System that is exposed to a soil (external) environment and an AMP is not required.
The staff is unclear why an AMP is not credited for management of aging of concrete piping components associated with this line item since they are subjected to a soil (external) environment.
Several aging effects exist for concrete exposed to a soil environment, as discussed in the GALL Report line items cited above. Request Explain how aging of these components will be managed, or justify why an AMP is not required for the concrete piping exposed to a soil (external) environment.
RAI 3.5.2.2.2-01 Industry standards identified in the GALL Report Structures Monitoring AMP suggest a five inspection interval for structures exposed to a natural environment, structures inside containment, continuous fluid-exposed structures, and structures retaining fluid or pressure, a ten year inspection interval for below-grade structures and structures in a controlled LRA Section 3.5.2.2.2.1 states that the Structures Monitoring Program is credited for management of affected concrete structures and structural components even if the AMR did identify aging effects requiring management; however, the LRA does not discuss the interval under the Structures Monitoring Explain in more detail the inspection interval for the structures in the scope of license If the inspection interval exceeds the recommendations in the GALL Report, explain the for extending the interval; include relevant operating experience in the RAJ 3.5.2.3.13-01 Background In GALL Report AMP XI.S6, "Structures Monitoring Program," program elements 3 and 4 state that for each structure/aging effect combination the specific parameters monitored or inspected are selected to ensure that the aging degradation leading to loss of intended function will be detected and quantified before there is a loss of intended function.
Issue Table 3.5.2-13 states that calcium silicate and fiberglass insulation materials exposed to outdoor or air-indoor environments have no AERMs and an AMP is not required.
The LRA also states that the components have an intended function of providing structural or functional support to non-safety related equipment whose failure could prevent satisfactory accomplishment of required safety functions.
The staff is unclear why an AMP is not credited for managing aging of these. Request Explain how aging of these components will be managed, or justify why an AMP is not required for the calcium silicate and fiberglass insulation materials exposed to air-outdoor or air-indoor environments.
August 26, 2010 Mr. S.K. Gambhir Vice President Technical Services Columbia Generating Station Energy Northwest MD PE04 P.O. Box 968 Richland, WA 99352-0968 REQUEST FOR ADDITIONAL INFORMATION FOR THE REVIEW OF THE COLUMBIA GENERATING STATION, LICENSE RENEWAL APPLICATION FOR FATIGUE MONITORING PROGRAM, TIME-LIMITED AGING ANALYSIS EXEMPTIONS, METAL FATIGUE TIME-LIMITED AGING ANALYSIS, CUMULATIVE FATIGUE DAMAGE, CAST AUSTENITIC STAINLESS STEEL, AND STRUCTURAL (TAC NO ME3058)
Dear Mr. Gambhir:
By letter dated January 19, 2010, Energy Northwest submitted an application pursuant to Title 10 of the Code of Federal Regulations Part 54 (10 CFR Part 54), to renew operating license NPF-21 for Columbia Generating Station, for review by the U.S. Nuclear Regulatory Commission (NRC or the 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. Further requests for additional information may be issued in the future. Items in the enclosure were discussed with Abbas Mostala 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 301-415-4029 or bye-mail at evelyn.gettys@nrc.gov.
Sincerely, IRA BPham fori Evelyn Gettys, Project Manager Projects Branch 1 Division of License Renewal Office of Nuclear Reactor Regulation Docket No. 50-397
Enclosure:
As stated cc w/encl: Distribution via Listserv DISTRIBUTION:
See next page ADAMS Accession No' ML102220373
.. OFFICE LA:DLR PM:RPB1 :DLR BC: RPB1 :DLR PM:RPB1:DLR NAME IKing EGettys BPham EGettys (BPham for) DATE 08/12/10 08/25/10 08/26/10 08/26/10 OFFICIAL RECORD COPY Letter to S.K. Gambhir from Evelyn Gettys dated August 26,2010 REQUEST FOR ADDITIONAL INFORMATION FOR THE REVIEW OF THE COLUMBIA GENERATING STATION, LICENSE RENEWAL APPLICATION FOR FATIGUE MONITORING PROGRAM, TLM EXEMPTIONS, METAL FATIGUE TLM, CUMULATIVE FATIGUE DAMAGE, CASS, AND STRUCTURAL (TAC NO ME3058) DISTRIBUTION:
HARD COPY: DLR RF E-MAIL: PUBLIC RidsNrrDlr Resource RidsNrrDlrRpb1 Resource RidsNrrDlrRpb2 Resource RidsNrrDlrRarb Resource RidsNrrDlrRapb Resource RidsNrrDlrRasb Resource RidsNrrDlrRerb Resource RidsNrrDlrRpob Resource EGettys DDoyle FLyon AMostala WWalker, RIV RCohen, RIV LSubin,OGC