ML14038A336

Request 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 10

ML14038A336
Person / Time
Site:	Byron, Braidwood
Issue date:	02/26/2014
From:	Daily J Division of License Renewal
To:	Gallagher M Exelon Generation Co
John Daily 301-415-3873
References
TAC MF1879, TAC MF1880, TAC MF1881, TAC MF1882
Download: ML14038A336 (20)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-Q001 February 26, 2014 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 10 (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, John W. Daily, Sr. Project Man ger 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

• ML14038A336

• concurred via email OFFICE LA: DLR/RPB2 PM: DLR/RPB1 BC: DLR/RPB1 PM: DLR/RPB1 NAME I King JDa~y YDiaz-Sanabria JDailv DATE 02/11/14 02/25/14 02/26/14 02/26/14

Letter toM. P. Gallagher from John W. Daily dated February 26, 2014 SUBJECT REQUESTS FOR ADDITIONAL INFORMATION FOR THE REVIEW OF THE BYRON STATION, UNITS 1 AND 2, AND BRAIDWOOD STATION, UNITS 1 AND 2, LICENSE RENEWAL APPLICATION-AGING MANGEMENT-SET 10 (TAC NOS. MF1879, MF1880, MF1881, AND MF1882)

DISTRIBUTION:

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PUBLIC RidsNrrDir Resource RidsNrrDirRpb1 Resource RidsNrrDirRpb2 Resource RidsNrrPMByron Resource RidsNRRPMBraidwood Resource RidsOgcMaiiCenter JDaily LRobinson DMclntyre, OPA EDuncan, Rill JBenjamin, Rill AGarmoe, Rill SRI, Byron Station, Rill JRobbins, Rill VMitlyng, Rill PChandrathil, Rill

BYRON NUCLEAR STATION, UNITS 1 AND 2 AND BRAIDWOOD NUCLEAR STATION, UNITS 1 AND 2 REQUEST FOR ADDITIONAL INFORMATION AGING MANAGEMENT-SET 10 (TAC NOS. MF1879, MF1880, MF1881, AND MF1882)

RAI 4.3.1-1, Transient basis redefinition for 4 new transients (060)

Applicability: Byron Nuclear Station (Byron) and Braidwood Nuclear Station (Braidwood)

Background:

The license renewal application (LRA) states that for Transient 6, "Letdown Flow Shutoff Prompt Return to Service," in LRA Table 4.3.1-2 and 4.3.1-5, the baseline cycles for Byron, Unit 2, and the projected 60-year cycles for all four units exceed the current licensing basis (CLB) cycle limit of 200. The LRA states that the transient was redefined as four differential temperature range transients. The LRA further states that number of baseline and 60-year projected cycles for each of the differential temperature range transients were determined and a reanalysis was performed for the bounding location, which confirmed that the cumulative fatigue usage will remain below 1.0.

Issue:

The staff is unclear on the technical basis for redefining the original transient definition to four new transients. The applicant did not provide the new transient definitions, baseline cycle counts, 60~year projected cycle counts, and CLB cycle limits for the four redefined transients.

Request:

1 Provide the four redefined differential temperature range transient definitions and identify the cycle limits, the baseline cycle counts, and projected cycle counts for each new transient.

2 Describe and justify the basis for redefining the original transient definition.

3. Update the applicable LRA tables to include the redefined transients.

4. Confirm that the Fatigue Monitoring Program, when implemented, will monitor the redefined transient cycles and severities and will require action prior to exceeding design limits. If not, justify that the aging effects due to fatigue will be managed during the period of extended operation for the components impacted by these redefined transients.

RAI 4.3.1-3, Transient 16, regarding recovery of main feedwater flow after isolation (060)

Applicability: Byron and Braidwood ENCLOSURE

Background:

LRA Tables 4.3.1-1 and 4.3.1-5 state that Transient 16, "Recovery of Main Feedwater Flow After Isolation (Unit 1 only)," is applicable to the Unit 1 Steam Generators only for both Byron and Braidwood. The LRA further states that the transient was not evaluated separately because cycles associated with switching between main and auxiliary feedwater flow are implicit in the cycles counted for the other reactor coolant system (RCS) transients.

Issue:

It is unclear to the staff which "other RCS transients" will be monitored since they are implicit in cycles associated with switching between main and auxiliary feedwater flow. The applicant did not provide enough information to describe why monitoring these "other RCS transients" will accurately account for Transient 16. It is also unclear to the staff why Transient 16 is applicable to Byron, Unit 1, and Braidwood, Unit 1, only.

Request:

1. Clarify which RCS transients will be monitored to account for Transient 16. Explain and justify how monitoring these other RCS transients will be adequate so that Transient 16 will not need to be monitored through the period of extended operation.

2. Confirm that these other RCS transients monitored in lieu of Transient 16 are included in applicable LRA tables and will be incorporated into the Fatigue Monitoring Program. If not, revise the LRA to ensure that these other RCS transients are identified in the appropriate LRA tables in LRA Section 4.3 and are included in the Fatigue Monitoring Program.

3.

Explain and justify why Transient 16 is only applicable to Byron, Unit 1, and Braidwood, Unit

1. Clarify if the other RCS transients monitored in lieu of Transient 16 is applicable only to Byron, Unit 1, and Braidwood, Unit 1.

RAI4.3.1-4, Frequency of samples for Sampling Line and Nozzle Transients (060)

Applicability: Byron and Braidwood

Background:

LRA Table 4.3.1-2 and 4.3.1-5 state that Transient 14, "Sampling Line and Nozzles Transients,"

will not be monitored. The LRA states that chemistry samples are taken at a much lower frequency than that which was assumed in the design, resulting in fewer cycles. The LRA further states that samples are no longer taken from the RCS as specified in the design, and are taken instead from the letdown system. The LRA states that samples from the letdown system result in lower temperature differences and lower transient severity.

Issue:

The applicant did not provide enough information regarding the lower frequency and temperature differences when taking chemistry samples from the letdown system instead of the RCS system. The staff is unclear how the lower frequency and temperature differences support the basis for not monitoring Transient 14.

Request

1. Provide the comparison of frequencies at which chemistry samples are taken from the letdown system instead of the RCS system. Explain and justify why this less frequent sampling supports the basis for not monitoring Transient 14.

2. Provide the comparison of temperature differences from taking the chemistry samples from the letdown system instead of the RCS system. Explain and justify why the lower temperature differences support the basis for not monitoring Transient 14.

RAI 4.3.3-1, Metal fatigue analysis information on auxiliary feedwater, emergency diesel generator and other systems {060)

Applicability: Byron and Braidwood

Background:

LRA Section 4.3.3 states that implicit fatigue analyses were evaluated for piping and components designed in accordance with ASME Section Ill, Class 2 and 3, and ANSI 831.1 design rules. LRA Table 4.3.1-3 and Table 4.3.1-6 lists the transients and their 60-year projections for the Class 2 and 3 and ANSI 831.1 piping considered to experience transients associated with the Reactor Coolant System and Auxiliary Systems. The LRA states that these transients were summed to verify that they were less than 7,000 cycles.

LRA Section 4.3.3 further states that an operational review was performed on remaining systems that are affected by different thermal and pressure cycles. The LRA states that the review concluded that the total number of cycles projected for 60 years are significantly less than 7,000 cycles. The LRA states that these remaining systems include the Auxiliary Feedwater, Emergency Diesel Generator, Fire Protection, Heating Water and Heating Steam System, and Service Water Systems. The applicant dispositioned the time-limited aging analysis (TLAA) in accordance with 10 CFR Part 54.21{c)(1)(i) such that the ASME Section Ill, Class 2 and 3, and ANSI 831.1 allowable stress calculations for the remaining systems remain valid for the period of extended operation.

Issue:

The applicant provided the 60-year cycle projections for the transients associated with the Class 2 and 3 and ANSI 831.1 piping in the reactor coolant system and auxiliary systems. However, the applicant did not provide enough information for the auxiliary feedwater, emergency diesel generator, fire protection, heating water and heating steam system, and service water systems.

The staff requires additional clariflcation on the transients and 60-year projections on these remaining systems.

Request:

For each of the following systems: auxiliary feedwater, emergency diesel generator, fire protection, heating water and heating steam system, and service water systems:

a) Provide the transients used in the implicit fatigue analysis b) For each of these transients, provide the current cycle count, projected 60 year count, and justification for the 60-year projections.

c) Justify that the TLAA remains valid for the period of extended operation in accordance with 10 CFR 54 21(c)(1)(i).

RAI 4.3.4-1, Use of data to reduce conservatisms in environmentally-assisted fatigue analyses (060)

Applicability: Byron and Braidwood

Background:

LRA Section 4.3.4 states that "where plant specific data was available, it was incorporated into the analysis to reduce conservatism on an as-needed basis for qualification" with resanalyses for the NUREG/CR-6260, "Application of NUREG/CR-5999 Interim Fatigue Curves to Selected Nuclear Power Plant Components," locations.

LRA Section 4.3.4 also states that "since the analyses are based on design cycles and 60-year cycle projections, monitoring of usage through the period of extended operation is required to ensure these conclusions remain valid."

Issue:

The applicant did not identify what plant-specific data was used and the evaluations in which plant-specific data was used to reduce conservatism. It is also not clear in which analysis the applicant has used 60-year projected cycles.

Request:

1. Identify the environmentally assisted fatigue (EAF) evaluations in which plant-specific data was used. Describe the plant-specific data that was used to reduce conservatism Provide the basis for the use of the plant-specific data in these EAF evaluations.

2. Identify the EAF evaluations, including the specific transients and cycles for each location, in which 60-year projected cycles and/or reduced number of cycles were used.

RAI 4.3.4-2, WESTEMS'" Metal Fatigue calculation methodology (060)

Applicability: Byron and Braidwood

Background:

LRA Section 4.3.4 states that "the WESTEMS' fatigue calculation methodology employs a conservative algorithm for selection of the stress peaks and valleys for use in the ASME fatigue evaluation. In some cases, conservatism may be removed by the analyst using optional program tools, to produce a more accurate final result For any ASME component fatigue evaluation in which the analyst removed conservatism in the peak and valley selection, full documentation of the justification of peak removal was included in the supporting calculations.

Otherwise, the conservatism inherent to the WESTEMS' software was retained for the ASME fatigue evaluations."

Issue:

It is not clear to the staff whether the applicant used these "optional program tools" to remove conservatism to produce a more accurate final result for fatigue evaluations and the applicant's basis for removal of this conservatism.

Request:

1. Clarify whether any of the fatigue evaluation has used these "optional program tools" to remove conservatism.

a) If so, identify all the fatigue evaluations in which these "optional program tools" were used. Provide three examples in which these "optional program tools" were used.

For each example, provide the basis for the removal of conservatism and justify that a more accurate final result was produced.

RAI 4.3.4-3, Determining most limiting cumulative usage factor for metal fatigue transient sections (060)

Applicability: Byron and Braidwood

Background:

LRA Section 4.3.4 states that the Class 1 components were grouped into transient sections, which is defined as a group of sub-components/locations that experience the same transients.

The LRA further states that components that reside in the same transient section can easily be compared with each other to determine the most limiting component (or leading location) which is the location with the highest cumulative usage factor (CUF) value. The differences in stresses experienced by each component in a transient section are generally the result of the material and geometry differences.

Issue:

The staff noted that in order to have a meaningful comparison of CUF values to determine the most limiting component (or leading location) by using the highest CUF value, it is important that the CUFs were assessed similarly (e.g., amount of rigor in calculating CUF) and used the same fatigue curves in ASME Code, Section Ill, Appendix I. It is also not clear whether the applicant considered the differences in component materials when comparing CUF values since material properties may impact the CUF values. The staff noted that through the course of plant operation it is possible that CUF values for specific components were possibly re-evaluated as part of power uprates, generic letters, bulletins, etc. to different editions of ASME Code, Section Ill and with varying levels of rigor when compared to the fatigue evaluations performed for the plant's original design.

Request:

1. Confirm that the CUFs that were compared with each other in a transient section to identify the location with the highest CUF value were assessed similarly (e.g., amount of rigor in calculating CUF) and used the same fatigue curves in ASME Code Section Ill Appendix I to provide a meaningful comparison. If not, provide the basis for ranking or comparing the CUFs to one another to provide an appropriate method for screening and determining a leading/limiting location.

2. Clarify whether CUF values of different material types were compared to one another when determining the leading location(s) within a transient section. If yes, identify the transient section, locations and materials that have been compared and eliminated for consideration of EAF. Justify that this comparison of CUF values between different materials within a transient section for the consideration of EAF is appropriate or valid.

RAI4.3.4-4, Determinations of environmentally-assisted fatigue related to components in various sections but in the same major component or system (060)

Applicability: Byron and Braidwood Stations

Background:

LRA Section 4.3.4 states that components that reside in different transient sections, but are within a common system or piece of major equipment, were also compared to determine leading locations to represent their respective system/equipment.

The LRA also states that often, it is the transients themselves that control which components have the highest usage factors in a given system, and so, within a particular system, those transient sections with the most severe system transients will usually have the components with the highest usage factors. However, the applicant stated that the comparison of components in different transient sections must be performed after the appropriate environmental fatigue correction factor (Fen) is applied to the component usage factor because Fen is dependent on temperature and strain rate and, therefore, can vary for each transient section.

Based on the information in the LRA, it is not clear when the applicant compared components that reside in different transient sections, but are within a common system or piece of major equipment, to determine leading locations to represent their respective system/equipment. It is also not clear what assumptions or factors were considered by the applicant when making this comparison to determine the leading location that resides in different transient sections and the basis for eliminating a location for consideration of EAF.

Request:

1. Identify the locations that were compared from different transient sections, but within a common system or piece of major equipment, and the component that was eliminated. A sample set of locations and eliminated component(s) may be provided; however, a justification of sufficient detail is necessary to explain that this sample set is bounding for the population of compared locations and eliminated component(s).

2.

For each of these situations, provide the basis for the comparison that was made for different transient sections but within a common system or piece of major equipment. In addition, provide the basis for eliminating the component(s) that was eliminated for consideration of EAF. As part of these justifications, specifically address any assumptions, factors or criteria that were applicable when implementing this comparison.

RAI 4.3.4~5, Use of stress based comparisons to remove components or locations from EAF consideration (060)

Applicability: Byron and Braidwood Stations

Background:

LRA Section 4.3.4 states that a stress basis comparison is performed to identify the leading transient section locations. The LRA states that Westinghouse has developed an approach that was applied to Byron and Braidwood, Units 1 and 2, for performing a stress basis comparison for the components included in the screening process.

The applicant stated that the following stress analysis characteristics were considered in determining the limiting locations within a given transient section:

1) Qualification Criteria (ASME Code Section Ill, NB-3200, NB-3600, etc.)

2) Stress Analysis Technique Furthermore, the applicant stated that in order to perform these stress basis comparisons, a hierarchy of stress analysis techniques was developed based on fatigue analysis experience to define the relative complexity of the various techniques.

1) Standard NB-3600 analysis

2) NB-3600 with non-standard mechanical stress indices or stress quantities used in stress formulas

3) NB-3600 with non-standard thermal stress indices or stress quantities used in stress formulas

4) Combination of 2) and 3)

5) NB-3200 Fatigue Analysis Those components with a screening environmentally-adjusted cumulative usage factor (CUFen) of less than 1.0 were removed from the list because they have been calculated using the design basis fatigue usage factors with a maximum Fen based on material.

Issue:

The staff noted that the stress basis comparison described in LRA Section 4.3.4 consists of two aspects: (1) consideration of stress analysis characteristics; and (2) a hierarchy of stress analysis techniques.

The staff noted that it appears the applicant eliminated certain Safety Class 1 reactor pressure boundary locations susceptible to EAF by performing a "stress basis comparison." It is not clear which locations were eliminated or what the technical basis was for removing these locations from consideration of EAF as a leading location.

Request:

1. Confirm whether the use of a stress basis comparison and screening CUFen of less than 1.0 were the only methods for eliminating locations for consideration of EAF. If not, describe and justify any other methods that were used. Include the locations that were eliminated and the associated technical basis.

2. Describe and justify the circumstances and situation when locations were eliminated using a stress basis comparison.

3. Identify the locations that were eliminated as a result of performing this stress basis comparison and provide the basis for eliminating these locations/components. Specifically, address any assumptions, factors, or criteria that were used when eliminating these locations for consideration for EAF. A sample set of locations may be provided; however, a justification of sufficient detail is necessary to explain that this sample set is bounding for the population of eliminated locations.

RAI4.3.4-6, Use of NUREG/CR-6909 and its corresponding fatigue curves in calculating cumulative usage factor values (060)

Applicability: Byron and Braidwood Stations

Background:

LRA Section 4.3.4 states that when performing an EAF evaluation, a plant can either use guidance from NUREG/CR-5704, "Effects of LWR Coolant Environments on Fatigue Design Curves of Austenitic Stainless Steels," for austenitic stainless steels, NUREG/CR-6583, "Effects of LWR Coolant Environments on Fatigue Design Curves of Carbon and Low-Alloy Steels," for carbon and low alloy steels, and NUREG/CR-6909, "Effects of the LWR Coolant Environments on the Fatigue Life of Reactor Materials," for nickel alloy steels, or they can use guidance from NUREGICR-6909 for all materials. Note that if NUREG/CR-6909 is used, the corresponding fatigue curves therein must be considered in calculating the CUF values. This difference must be addressed as part of the EAF screening process. The applicant also indicated that NUREGICR-6909 was used for nickel alloy steels only.

Issue:

As noted in the LRA, when NUREG/CR*6909 is used, the corresponding fatigue curves in the report must be considered in calculating the CUF values. However, it is not clear to the staff how many/if any nickel alloy steel components were eliminated based on the environmentally-adjusted cumulative usage factor (CUFen) screening process described in the LRA or how the applicant accounted for the difference in fatigue curves used in the fatigue analyses and NUREG/CR-6909 as part of the EAF screening process.

Request:

1. Identify the locations that were eliminated by the CUFen screening process, including the CUF and environmental fatigue correction factor (Fen ) values for these components. A sample set of locations may be provided; however, a justification of sufficient detail is necessary to explain that this sample set is bounding for population of eliminated locations.

2. Discuss and justify how the difference in fatigue curves used in the fatigue analyses of these components and NUREG/CR-6909 was addressed as part of the EAF screening process.

3. Provide the design basis CUF and revised CUF as a result of the fatigue curves in NUREG/CR-6909.

RAI4.3.4-7, Derivation methods for determining maximum Fen as used in environmental fatigue calculations (060)

Applicability: Byron and Braidwood Stations

Background:

LRA Section 4.3.4 states that once the stress basis comparison has been performed and the leading transient section locations have been identified, screening environmental correction factors (Fen) are developed for each component so that cumulative usage factors including environmental fatigue, CUFen. can be calculated.

Furthermore, the LRA states that those components with a screening CUFen of less than 1.0 were removed from the list because they have been calculated using the design basis fatigue usage factors with a maximum Fen based on material.

Issue:

It is not clear whether the "maximum Fen" is the maximum calculated from the NUREG reports or whether is the maximum calculated for a particular transient section. The staff noted that if it is the latter, it is important to understand the applicant's assumptions in calculating the maximum Fen based on material for a particular transient section.

Request:

Clarify if the maximum Fen based on the material is the calculated maximum Fen from the applicable NUREG reports or the calculated maximum from a particular transient section.

If the maximum Fen was based on the transient section, identify any assumptions (e.g.,

temperature, sulfur, dissolved oxygen, strain rate) used in calculating the Fen and the basis for these assumptions.

RAI 4.3.7 -1, Revisions to fatigue evaluations to account for thermal stratification (060)

Applicability: Byron and Braidwood

Background:

NRC Bulletin 88-11, "Pressurizer Surge Line Thermal Stratification," issued December 1988, requested utilities to demonstrate that the design requirements of the pressurizer surge line consider the effects for thermal stratification. LRA Section 4.3.7 states the demonstration was an ASME Section Ill fatigue analysis to account for thermal stratification that was identified as a TLAA.

The LRA further states that the original fatigue analyses for the pressurizer surge line included stratification sub-transients. The LRA states that the fatigue evaluations of the components affected by the bulletin were revised to consider transients in Table 4.3.1-1, 4.3.1 2, 4.3.1-4, and 4.3.1-5 and determined that the resulting cumulative fatigue usage will remain below 1.0.

Issue:

The applicant stated that the fatigue evaluations were revised to consider NRC Bulletin 88-11.

However, it is unclear how the fatigue evaluations were revised to account for thermal stratification in applicable components.

Request:

1. Identify the pressurizer surge line stratification sub-transients in the CLB.

2.

Identify which transients were considered when the fatigue evaluations were revised for components affected by Bulletin 88-11

3. Confirm that the Fatigue Monitoring Program, when implemented, will monitor the redefined transient cycles and severities and will require action prior to exceeding design limits. If not, justify that the aging effects due to fatigue will be managed during the period of extended operation for the components impacted by these redefined transients.

RAI4.7.5-1, Reactor coolant pump flywheel fatigue crack growth analyses (067)

Applicability: Byron and Braidwood

Background:

LRA Section 4.7.5 describes the applicant's evaluation of the TLAA for reactor coolant pump flywheel fatigue crack growth. The LRA states that Technical Specification 5.5.7 requires an inspection program for the reactor coolant pump flywheels. Under this program, two of the flywheels receive either an ultrasonic or surface examination at 10-year intervals approximately, and all of the other flywheels receive these examinations at an interval not to exceed 20 years.

Per the LRA, the fatigue crack growth analyses described in WCAP~14535A, "Topical Report on Reactor Coolant Pump Flywheel Inspection Elimination," dated November 1996, and WCAP-15666-A, "Extension of Reactor Coolant Pump Motor Flywheel Examination," dated October 2003, provide the bases for the 10- and 20-year inspection intervals, respectively.

The reactor coolant pump flywheel fatigue crack gro'Nth analyses are based on 6,000 reactor coolant pump starts and stops. Based on a projection that the actual number of reactor coolant pump starts and stops will not exceed this number through 60 years of plant operation, the applicant stated that the analyses remain valid for the period of extended operation. However, the applicant did not provide the results from any of the past flywheel examinations. These results could invalidate the analyses if the examinations show an increase in flaw size or crack growth rate greater than calculated in WCAP-14535A and WCAP-15666-A Request:

Summarize the results of all of the past inservice inspections that were performed on the reactor coolant pump flywheel components. Include the examination dates and describe any flaws that were found. If flaws were detected, quantify any growth and provide a comparison against the crack growth rates from WCAP-14535A and WCAP-15666-A RAI4.7.5-2, Reactor coolant pump flywheel fatigue crack growth TLAA and [updated final safety analysis report (UFSAR)] summary (067)

Applicability: Byron and Braidwood

Background:

The reactor coolant pump flywheel fatigue crack growth TLAA, described in LRA Section 4.7.5, covers two separate analyses. One analysis, based on WCAP-14535A, establishes an approximate 1 0-year inspection frequency for two reactor coolant pump flywheels. The other analysis, based on WCAP-15666-A, establishes a 20-year maximum inspection frequency for all of the other flywheels.

Issue:

LRA Section A.4.7.5 provides the UFSAR supplement summary description for the reactor coolant pump flywheel fatigue crack growth TLAA. However, the summary description does not:

(a) indicate that the TLAA supports the basis for continuation of the 20-year inspection frequency (it only mentions that it supports continuation of the 1 0-year inspection frequency), and (b) identify each reactor coolant pump and specify which WCAP report and corresponding inspection frequency apply to that pump.

Thus, it is not evident as to which of the two WCAP reports apply to the specific reactor coolant pump flywheels at Bryon Units 1 and 2 and Braidwood, Units 1 and 2.

Request:

Revise LRA Section A.4.7.5 to clearly identify each of the reactor coolant pump motor flywheels at Bryon Units 1 and 2 and Braidwood Units 1 and 2 and specify which WCAP report and corresponding inspection frequency apply to each pump mofor's flywheel.

RAt 3.2.1.20-1, Aging Management for Venturi Flow Meters (101)

Applicability: Byron and Braicf\\Nood

Background:

EPRI Report TR-112118, "Nuclear Feedwater Flow Measurement Application Guide," July 1999, indicates that venturi flow meters used to calculate feedwater flow rates in nuclear power plants are susceptible to aging degradation such as fouling and loss of material due to erosion and corrosion (including defouling), all of which can cause flow measurement errors. In its review of LRA Item 3.2.1-20, the staff also noted that LRA Tables 3.2.2-3 and 3.2.2-4 address aging management of loss of material in stainless steel restricting orifices using the One-Time Inspection Program and the Water Chemistry Program. However, the LRA does not address any AMR item to manage aging for venturi flow meters.

During the audit, the staff noted the applicant's operating experience described in AR 00748581, "Feedwater Venturi Fouling Indication," dated March 12, 2008, which states that during normal thermal performance monitoring activities, some signs of feedwater flow venturi fouling were observed for Braidwood, Unit 2, venturis.

Issue:

Flow measurement and calculation errors associated with aging degradation of venturi flow meters can cause safety-related issues such as overpower conditions and can accelerate aging effects of piping and piping components through those overpower conditions. However, the LRA does not describe how the applicant will manage fouling and loss of material for feedwater venturi flow meters.

Request:

Describe how the aging effects of fouling and loss of material for feedwater venturi flow meters will be managed, and revise the LRA consistent with the response. Alternatively, provide adequate justification why managing of these aging effects is not required.

RAI 3.1.1.80-1: Aging Management for Piping and Piping Components of the Reactor Coolant System (101)

Applicability: Byron and Braidwood

Background:

LRA Item 3.1.1-80 addresses cracking due to stress corrosion cracking (SCC) for stainless steel pressurizer relief tank and associated components (non-ASME Section XI components) exposed to treated borated water greater than 140 °F. LRA Item 3.1.1-80 also indicates that the applicant uses the One-Time Inspection Program and Water Chemistry Program to manage this aging effect for the none-ASME Section XI components.

LRA Table 3.1.2-1 (Page 3.1-57) indicates that the applicant also uses LRA Item 3.1.1-80 to manage cracking due to sec for stainless steel piping, piping components, and piping elements of the RCS, which are exposed to reactor coolant These components are associated with generic note C, indicating that these components are different from those which are evaluated in the Generic Aging Lessons Learned (GALL) Report, but applicant's aging management is consistent with the GALL Report for material, environment, aging effect, and aging management program. However, the staff noted that the LRA does not clearly indicate whether these are ASME Code Class components.

Issue:

The LRA does not address why periodic inspections (e.g., ASME Code Section XI examinations) are not used to manage cracking for these piping, piping elements and piping components exposed to reactor coolant.

Request:

1. Clarify whether the subject piping, piping components and piping elements of the reactor coolant system are ASME Code Class components. As part of the response, describe the names and locations of the components in order to demonstrate that the ASME Code examination requirements are not applicable for these components.

2.

If these are ASME Code Class components, justify why periodic inspections (e.g., ASME Code Section XI examinations) are not used to manage cracking for the components which are exposed to reactor coolant.

RAI 3.1.1.81~1: Aging Management for Reactor Coolant Pump Thermal Barrier Heat Exchanger Tubes (101)

Applicability: Byron and Braidwood

Background:

LRA Item 3.1.1 ~81 addresses cracking due to SCC for stainless steel pressurizer spray head exposed to reactor coolant, which is managed by the One~time Inspection Program and Water Chemistry Program. As described in LRA Table 3.1.2-1, the applicant uses LRA Item 3.1.1~81 to manage cracking due to sec for tubes of reactor coolant pump thermal barrier heat exchangers which are exposed to reactor coolant. The LRA does not identify any periodic inspection activities to manage cracking for these heat exchanger tubes.

In addition, the staff noted that UFSAR Section 11.5.2.3.2, "Component Cooling Water Monitors," indicates that applicant's radiation detectors continuously monitor the component cooling system for leakage of reactor coolant from the reactor coolant system and/or the residual heat removal system.

Issue:

The reactor coolant pump thermal barrier heat exchanger tubes form the pressure boundary between the component cooling and reactor coolant systems. However, the LRA does not identify any periodic inspections to manage cracking for these heat exchanger tubes. In addition, the LRA does not address whether applicant's operating experience, including the component cooling water monitoring activity of the UFSAR, confirms that cracking is not occurring in the heat exchanger tubes.

Request:

1. Justify why the LRA does not identify any periodic inspections to manage cracking for the reactor coolant pump thermal barrier heat exchanger tubes. Alternatively, identify periodic inspections to manage cracking for these components. As part of the response, confirm whether or not the heat exchanger tubes are ASME Code Class 1 components.

2. Clarify whether applicant's operating experience confirms that cracking is not occurring in the heat exchanger tubes.

RAI 4. 7.4-1: TlAA for Flaws in RHR Heat Exchanger Nozzles (111)

Applicability: Byron and Braidwood

Background:

LRA Section 4. 7.4 addresses applicant's fracture mechanics analysis for the tube side inlet and outlet nozzles of residual heat removal {RHR) heat exchangers. The LRA identifies this fracture mechanics analysis as a TLAA. The LRA states that indications were detected in Braidwood 2A and 2B RHR heat exchangers during ultrasonic examinations in 1991 and some of these indications exceeded the acceptance standards of the 1983 Edition through the Summer 1983 Addenda of ASME Section XI, Subarticle IWB*3500.

The LRA also states that these flaws were subjected to further evaluation in accordance with ASME Section XI, Subarticle IWB*3600. The LRA further states that even though the RHR heat exchangers are ASME Code Class 2 components, the applicant performed an ASME Code Class 1 fracture mechanics analysis and this evaluation demonstrated that continued operation of Braidwood Unit 2 RHR heat exchangers was acceptable even with these flaws.

In addition, the LRA states that subsequently, ultrasonic examinations were performed on all the RHR heat exchanger tube side inlet and outlet nozzles at Byron and Braidwood Stations, Units 1 and 2. The LRA states that any additional indications exceeding the IWB*3500 acceptance standards were dispositioned with the results of the fracture mechanics analysis. The LRA also states that WCAP*13454, "Fracture Mechanics Evaluation, Byron and Braidwood, Units 1 and 2, Residual Heat Exchanger Tube Side Inlet and Outlet Nozzles," August 1992 (Proprietary), and WCAP~13455, "Fracture Mechanics Evaluation, Byron and Braidwood, Units 1 and 2, Residual Heat Exchanger Tube Side Inlet and Outlet Nozzles," August 1992 (Non-proprietary, ADAMS Accession No. 9208280207), were submitted to the NRC on August 25, 1992 to present the original fracture mechanics methodology for dispositioning the indications which were found at Byron and Braidwood.

In its review of LRA Section 4.7.4, the staff also noted that a relief request by the applicant indicated that an ASME Section XI repair by excavation was completed on the unacceptable flaws of the Braidwood Unit 2 RHR heat exchanger nozzle-to-vessel welds ("Relief from lnservice Inspection Requirements for Residual Heat Removal Heat Exchanger Nozzle-to-Vessel Welds," dated December 12, 1995, ADAMS Accession No. 951219036). The staff also noted that this reference stated that the Braidwood, Unit 2, flaws were fabrication flaws, slag, incomplete fusion and excess porosity. The staff further noted that the above reference did not identify any other unacceptable flaws of the Byron and Braidwood RHR heat exchanger nozzles_

It is unclear to the staff whether there are flaws currently in the nozzles that exceed the acceptance standards of ASME Code Section XI IWB-3500. It is also unclear to the staff whether the fracture mechanics analysis is relied upon to support: (a) continued service with existing flaws in these nozzles, or (b) applicant's relief request for an alternative to the ASME Code inservice inspection method for these nozzles (e.g., performing VT-2 visual examination in place of ultrasonic testing). Finally, the staff needs additional information regarding the existing flaws and flaw growth analysis.

Request:

1 Clarify whether there are flaws currently in the Byron and Braidwood RHR heat exchanger nozzles that exceed the acceptance standards of ASME Code Section XI IWB-3500.

2. Clarify whether the fracture mechanics analysis is relied on to support: (a) continued service of the heat exchanger nozzles with existing flaws, or (b) applicant's relief request for an alternative to the ASME Code inservice inspection method for these nozzles (e.g.,

performing VT-2 visual examination in place of ultrasonic testing)

The staff notes that relief requests for inservice inspections are only valid for the current inservice inspection ten-year interval and are required to be resubmitted for each interval for the period of extended operation if desired_ Given this fact, if the fracture mechanics analysis is relied on to support the use of an alternate inspection method under a relief request process, clarify why the relief request process is not identified as part of the 10 CFR 54.21 (c)(1 )(iii) aging management basis in conjunction with the fracture mechanics analysis.

3.

Provide the following information for the applicant's fracture mechanics analysis: (a) current flaw sizes (i.e., length and depth), orientations (i.e., circumferential and axial) and locations based on the most recent inspection results in comparison with nozzle dimensions, and (b) projected flaw sizes at the end of the period of extended operation As an alternative to (a) and (b), if a bounding-case analysis is applicable to each nozzle, provide the maximum current flaw size and maximum projected flaw size with the associated orientation and location which bound the other flaws for each nozzle. (c) In addition to the information for (a) and (b), describe the acceptance criteria for the flaws and when the most recent volumetric examination was performed on each nozzle. As part of this response, provide the relevant transient names and projected numbers of transient cycles for the fracture mechanics analysis.

RA14.1-1, Absence of a TLAA for Flow-Induced Vibrations (058)

Applicability: Byron and Braidwood

Background:

LRA Table 4.1-1 identifies that the Byron and Braidwood CLBs do not include any flow-induced vibration analyses for reactor vessel internal (RVI) components that conform to the definition of a TLAA in 10 CFR 54.3(a) or would need to be identified as TLAAs in accordance with the requirement in 10 CFR 54.21 (c)(1 ). Consistent with this basis, LRA Section 4.3.5 states that the applicable RVI flow*induced vibration analyses in the CLB are based on stress ranges that are below the ASME Section Ill fatigue endurance limit of 1011 cycles. LRA Section 4.3.5 states that the number of the stress range cycles is not limited over the current operating life and, therefore, the analyses are not based on any time-dependent assumptions defined by the current operating terms and are not classified as TLAAs in accordance with 10 CFR 54.3(a),

Criterion 3.

UFSAR Section 3.9 provides relevant design basis information_ UFSAR Section 3.9.5.2 states that the design of the RVI components is based on the design basis loading conditions for normal operating, upset, emergency, and faulted condition transients that are listed on UFSAR pages 3.9*96 and 3.9*97. The UFSAR indicates that vibratory loads (including those that would occur during postulated operational basis earthquake conditions) are listed as normal operating condition loads for the RVI components. In addition, the UFSAR states that the design basis relies on previous RVI flow*induced vibration models and tests that were performed at the Indian Point Unit 2 and Trojan nuclear power plants and that form the basis for assessing flow*

induced vibrations of the RVI components at the Byron and Braidwood Stations. The UFSAR establishes that these models and tests are summarized in the following Westinghouse technical reports (TR):

WCAP*8317-A, "Prediction of the Flow-Induced Vibration of Reactor Internals by Scale Model Tests," July 1975 (a report that is applicable to Indian Point Unit 2)

WCAP-8780, "Verifications of Neutron Pad and 17 X 17 Guide Tube Designs by Preoperational Tests on the Trojan Unit 1 Plant," May 1976 Issue:

The applicant's basis for claiming that the treatment of RVI flow-induced vibrations does not need to be within the scope of a TLAA is based on the position that the vibratory stress loads for the RVI components are lower than the stress endurance limits for inducing fatigue in components. The staff finds this TLAA-identification basis is acceptable if the stated treatment of RVI vibratory loads was established in either Westinghouse TR No. WCAP-8317 -A or TR No.

WCAP-8780 as approved reports for the current design basis. However, the staff cannot determine, with certainty, whether this type of technical basis was established in either of the referenced WCAP reports.

Request:

Clarify whether TR No. WCAP-8317 -A or TR No. WCAP-8780 establishes the basis in the CLB for concluding that the RVI vibration stress loads are lower than the endurance limit for initiation of high-cycle fatigue. If not, identify and justify the document in the CLB that establishes and is relied upon for this design basis position RAI4.1-2, Absence of a TLAA for Metal Corrosion Allowances (058)

Applicability: Byron and Braidwood

Background:

LRA Table 4.1-1 identifies that the Byron/Braidwood CLBs do not include any component-specific metal corrosion allowance analyses that conform to the definition of a TLAA in 10 CFR 54 3(a) or would need to be identified as TLAAs in accordance with 10 CFR 54.21 (c)(1 ).

Issue:

The UFSAR does make reference to one metal corrosion allowance report. Specifically, UFSAR Section 5.4.2.54 refers to Babcock and Wilcox (B&W) Report No. 222-7720-PROS, Revision 3, "Replacement Steam Generators Secondary Side Corrosion Allowance Values for Design and Analysis." However, the UFSAR does not specifically indicate whether this report is being relied upon as part of the CLBs or design bases for the Byron and Braidwood reactor units.

Request:

Clarify whether B&W Report No. 222-7720-PR05, Revision 3, is relied upon for the CLBs or current design bases of the Byron and Braidwood reactor units. If the report is relied upon for the CLB or current design basis, assess the contents of the report against the six criteria in 10 CFR 54.3(a) for defining an analysis as a TLAA and justify why the metal corrosion allowance analysis in this report would not need to be identified as a TLAA for the secondary sides of the steam generators (SGs) in the Byron and Braidwood units.