View source for PLA-6397, Request for Additional Information for the Review of Units 1 and 2, License Renewal Application (LRA) Sections B.3.1, 4.3, and 4.7

{{Adams
| number = ML082250450
| issue date = 08/01/2008
| title = Request for Additional Information for the Review of Units 1 and 2, License Renewal Application (LRA) Sections B.3.1, 4.3, and 4.7
| author name = Mckinney B
| author affiliation = PPL Susquehanna, LLC
| addressee name = 
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000387, 05000388
| license number = NPF-014, NPF-022
| contact person = 
| case reference number = PLA-6397
| document type = Letter type:PLA
| page count = 31
| project = 
| stage = Request
}}

=Text=
{{#Wiki_filter:Brltt T. McKinney                  PPL Susquehanna, LLC Sr. Vice President & Chief Nuclear Officer                    769 Salem Boulevard Berwick, PA 18603 Tel. 570.542.3149 Fax 570.542.1504 btmckinney@pplweb.com U. S. Nuclear Regulatory Commission Document Control Desk Mail Stop OP1-17 Washington, DC 20555 SUSQUEHANNA STEAM ELECTRIC STATION REQUEST FOR ADDITIONAL INFORMATION FOR THE REVIEW OF THE SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 AND 2, LICENSE RENEWAL APPLICATION (LRA)
SECTIONS B.3.1, 4.3, AND 4.7                                                              Docket Nos. 50-387 PLA-6397                                                                                            and 50-388 References.      1) PLA-61 10, Mr. B. T. McKinney (PPL)to Document ControlDesk (USNRC),
                      "Applicationfor Renewed OperatingLicense Numbers NPF-14 and NPF-22,"
dated September 13, 2006.
: 2) Letterfrom Ms. E. H. Gettys (USNRC) to Mr. B. T McKinney (PPL),
                      "Request for Additional Informationfor the Review of the Susquehanna Steam Electric Station, Units 1 and 2 License Renewal Application, "dated July 3, 2008.
In accordance with the requirements of 10 CFR 50, 51, and 54, PPL requested the renewal of the operating licenses for the Susquehanna Steam Electric Station (SSES)
Units 1 and 2 in Reference 1.
Reference 2 is a request for additional information (RAI) related to License Renewal Application (LRA) Sections B.3.1, 4.3, and 4.7. The enclosure to this letter provides the additional requested information.
There are no new regulatory commitments contained herein as a result of the attached RAI responses. However, License Renewal Commitments #9 and #43 are revised in response to RAIs 4.3-4 and B.3.1-4, respectively.
If you have any questions, please contact Mr. Duane L Filchner at (610) 774-7819.
                                                                                                       ,&#xfd;ao "j2A

Document Control Desk PLA-6397 I declare, under penalty of perjury, that the foregoing is true and correct.
Executed on: rA/ A,-,         4 B. T. McKinney

==Enclosure:==
PPL Responses to NRC's Request for Additional Information (RAI)
Copy: NRC Region I Ms. E. H. Gettys, NRC Project Manager, License Renewal, Safety Mr. R. Janati, DEP/BRP Mr. F. W. Jaxheimer, NRC Sr. Resident Inspector Mr. A. L. Stuyvenberg, NRC Project Manager, License Renewal, Environmental

Enclosure to PLA-6397 PPL Responses to NRC's Request for Additional Information (RAI)

Enclosure to PLA-6397 Page 1 of 28 RAI B.3.1-1:
Under the Program Description of the Fatigue Monitoring Program (FMP), B.3. 1, the LRA states that the aging management program (AMP) monitors and tracks the number and severity of critical thermal and pressure transients for the selected reactor coolant system components. Describe how this tracking and monitoring is accomplished.
PPL Response:
The SSES Fatigue Monitoring Program (FMP) is implemented by an approved engineering procedure. The procedure provides instruction for the engineering evaluation and computer software analysis of data collected from the plant process computer system to ensure a consistent approach to monitoring fatigue usage at critical plant component locations over the operating life of the plant. Key aspects of the FMP are identification of the plant transients that should be monitored, selection of the critical component locations, and selection of the monitoring method (i.e., cycle-based fatigue or stress-based fatigue).
SSES uses a combination of engineering evaluation and computer software analysis of collected plant data to detect plant event occurrences involving critical thermal and pressure transients on selected plant components. Computer software analysis is also used to provide an assessment of the fatigue usage at monitored locations. The SSES FMP utilizes the EPRI FatiguePro software for the analysis of plant data. Details of the SSES FMP development and the methodology for monitoring plant transient events are provided below.
The first step in the development of the FMP was to determine which plant events should be monitored (i.e., the events that induce fatigue on plant components). An evaluation reviewed all plant events defined in plant design basis documents such as the reactor vessel thermal cycle drawing, reactor vessel component stress reports, the FSAR, the original plant Technical Specifications, and the ASME Class 1 piping system design specifications and associated stress reports to determine the list of fatigue-critical events to be monitored. The events that can be automatically counted by the FatiguePro software analysis of specific plant data (e.g., temperatures, pressures, flow rates, and valve positions) were also identified. Event cycles that are automatically counted by the SSES FMP are listed below:
Automatically Counted Events
* Design Hydrostatic Test
* Startup
* Shutdown
* Turbine Roll and Increase to Rated Power
* Hot Standby
* Vessel Flooding

Enclosure to PLA-6397 Page 2 of 28
* Loss of Feedwater Heaters, Partial Feedwater Heater Bypass
* Loss of Feedwater Heaters, Turbine Trip with 25% Steam Bypass
* SCRAM (includes Turbine Generator Trip, Feedwater On, Isolation Valves Stay Open, and All Other)
* Blowdown (includes Single Relief or Safety Valve Blowdown, and Automatic Blowdown)
* Loss of Feedwater Pumps, Isolation Valves Close
* Rated Power Normal Operation
* Reduction to 0% Power
* Minor Power Reduction
* SRV Actuation
* Zero Load
* Loss of RWCU at Rated Power
* RWCU Shutdown - No Fatigue
* Recirculation Pump Trip
* RWCU Re-start with Idle Recirculation
* Bottom Head Stratification Relief Those events that cannot be automatically counted are identified by the "manual" review of plant data and operating logs. The event cycles that are manually counted by the SSES FMP are listed below:
Manually Counted Events
* Boltup
* Unbolt
* Boltup-to-Unbolt
* Reactor Overpressure with Delayed SCRAM, Feedwater Stays On, Isolation Valves Stay Open
* Improper Start of Cold Recirculation Loop
* Sudden Start of Pump in Cold Recirculation Loop
* Improper Start with Recirculation Pumps Off and Drain Shut Off
* Pipe Rupture and Blowdown
* Loss of AC Power, Natural Recirculation Restart
* Design Seismic at Steady State
* Maximum Earthquake at Steady State o Standby Liquid Control (SLC) Injection
* Core Spray Injection
* Low Pressure Coolant Injection (LPCI)
* High Pressure Coolant Injection (HPCI)
* Reactor Core Isolation Cooling (RCIC) Injection
* Operating Basis Earthquake + Safety Relief Valve Blowdown (OBE+SRV)

Enclosure to PLA-6397 Page 3 of 28 The SSES FMP cycle counting assumes every event has a severity equal to that assumed in the design basis. Currently, no partial cycles are counted by the SSES FMP. For example, if the plant is starting up from 100'F and the reactor temperature exceeds 225&deg;F, but then immediately begins to cool down, one "Startup" cycle will be counted by the FMP. This is a conservative approach, since the design basis "Startup" cycle consists of a reactor temperature starting at 100&deg;F and increasing to 551 'F.
The second step in the development of the FMP was to determine the critical plant component locations to monitor. A screening criteria was employed, whereby any reactor pressure vessel (RPV) location with a 40-year design basis cumulative fatigue usage factor (CUF) of greater than 0.40 was identified for monitoring. For the ASME Class 1 piping systems, locations with fatigue usage ratios (i.e., design CUF/allowable CUF) of greater than 0.40 were considered. From the identified locations in each piping system, limiting locations (i.e., those with the highest design CUF) were selected for monitoring in both the "break" and (where applicable) "no break" zones. More recently, additional locations were added to the FMP to address environmentally-assisted fatigue, in accordance with NUREG/CR-6260. The RPV components and Class 1 piping systems currently being monitored by the FMP are identified in the SSES LRA Table 4.3-2 The two methods employed by the FMP software for computing fatigue usage at fatigue-critical locations are cycle-based fatigue (CBF) and stress-based fatigue (SBF). CBF uses an event-pairing method that emulates the ASME Class 1 fatigue calculation of the governing design stress reports. The FMP software uses the fatigue tables from the stress reports and substitutes the actual number of counted event cycles in place of the number of cycles assumed for the original 40-year design. As transient events are counted and accumulated, the calculated CBF usage increases accordingly.
SBF computes fatigue based on the analysis of the computed stress history of a component location. The development of SBF monitoring involves a two-step process:
(1) the determination of the local thermal-hydraulic conditions at the specific component location, based on available plant data, and (2) the determination of peak stresses at the component location under those thermal-hydraulic conditions. As the alternating peak stresses (highs and lows) accumulate, the calculated SBF usage increases accordingly.
The SSES FMP employs SBF monitoring for the Feedwater nozzle forgings, Feedwater nozzle safe ends, and the CRD penetrations. The SBF methodology employed for each of these components was benchmarked against the relevant design basis stress report for each component to ensure valid FMP fatigue results. All other monitored locations, as identified in LRA Table 4.3-2, are monitored by the CBF methodology.

Enclosure to PLA-6397 Page 4 of 28 RAI B.3.1-2:
The GALL Program X.M1 recommends under the program element "parameters monitored/inspected," that the program monitors all plant transients that cause cyclic strains which are significant to the fatigue usage factor. Page 4.3-7 of Section 4.3 of the LRA states that PPL monitors the design transients listed in Table 4.3-1 using FatiguePro. Confirm whether or not FatiguePro is used for stress-based monitoring of components, and if so list the components that are stress-based monitored. In addition, confirm whether or not FatiguePro is a part of the FMP and if so, please describe the role FatiguePro has in the Fatigue Monitoring Program.
PPL Response:
The FatiguePro software is used for stress-based fatigue (SBF) monitoring of the Feedwater nozzle forgings, Feedwater nozzle safe ends, and the CRD penetrations.
As stated in the response to RAI B.3.1 -1, the EPRI FatiguePro software is used by, and is a significant part of, the SSES Fatigue Monitoring Program (FMP). The FatiguePro software is used to identify the occurrence of specific plant events. Events that cannot be identified by FatiguePro analysis of plant data are manually entered into the FatiguePro database. The FatiguePro database then provides a complete count of all fatigue-critical plant event cycles that have occurred at SSES. In addition to providing cycle counting and tracking of the fatigue-critical plant events, FatiguePro calculates the cumulative usage factor (CUF) for fatigue at limiting plant component locations, as detailed in the response to RAI B.3.1-1.
RAI B.3.1-3:
The GALL Program X.M1 recommends under the program element "preventative actions," that components be monitored to provide adequate margin against fatigue cracking due to anticipated cyclic strains. Under the "Preventive Actions" and the Monitoring and Trending program element enhancement, the LRA states that additional actions may be taken when sufficient fatigue accumulation has occurred if determined necessary to address fatigue-related concerns. Describe what constitutes as sufficient fatigue accumulation, describe the criteria used to determine if further actions are required, and provide the periodicity of the updates of cumulative usage factor (CUF)s.
PPL Response:
PPL has revised LRA commitment #43 which enhances the SSES Fatigue Monitoring Program (FMP), as detailed in the response to RAI B.3.1-4 below. This enhancement specifies certain actions to be taken when a monitored location reaches a sufficient fatigue accumulation.

Enclosure to PLA-6397 Page 5 of 28 The definition of "sufficient fatigue accumulation" is component specific and will be determined for the FMP-monitored locations as part of the implementation of the enhancement. A specific "action level CUF" for each monitored location will be determined through an engineering evaluation, considering the current CUF, the 40-year design basis CUF, the design basis allowable CUF (typically 1.0 or 0.1), any other applicable fatigue design criteria (e.g., high energy line break), and the projected rate of fatigue accumulation for each monitored component location. The "action level CUF" will be conservatively set to a value that will allow for not less than 4 years of additional plant operation before the location's CUF would reach the allowable CUF.
When the "action level CUF" is reached, the FMP will require an action request to be generated. The action request will require further engineering evaluation to resolve the issue or determine further actions.
The SSES FMP will ensure that a fatigue monitoring update calculation is performed at least once per fuel cycle (every 2 years) for each unit.
RAI B.3.1-4:
The program elements, "preventative actions" and "acceptance criteria" of the GALL Program X.M1 relates to maintaining the fatigue usage factor below the design limit and monitoring the plant transients. It does not appear that SSES proposed enhancement to the program elements "preventative actions" and "acceptance criteria" made on LRA page B-149 are applicable to the GALL Program element recommendations. Provide your justification on how the enhancement to address environmental effects is applicable under preventive actions and acceptance criteria of this AMP.
PPL Response:
The description of the required enhancements to the Fatigue Monitoring Program in the SSES LRA Appendix B, Section B.3.1, is revised to clarify the enhancements to the program, including those to address the environmental effects. Also, Commitment #43 in LRA Table A-I is revised to provide consistent statements for the enhancements to the program.
B.3.1 Fatigue Monitoring Program The Required Enhancements of B.3.1 (pages B-149-151 of the LRA) are revised by addition (bold italics) and deletion (,i-kethfaug-h) as follows:
Pr-evisions will be made the Fatigue Monitor-ing Program to address environmena effects on fatigue at specified locations in accor-dance with NUREG/1CR 6260.

Enclosure to PLA-6397 Page 6 of 28
*Preventivc Actions, Monitor-ing and Trcnding-valve fatigue analyses and other-fatigue rnelatiedd TL&#xfd;AA, such as flued head analyses and.high energy line bra vlainwhen suifficient fatigu accutmulation has occurred, to determine if additional actions are required to address fatigue related concerns.
monitringand per-iodic updates to current and pr-ojected CURs for- limiting the all1ow a ble. before the en-m     thepr-ioed of extne opeation. The actions will do include one or-mor~e of the floig 1.Further-r-efinemenit of the fatigue analyses to lower-the CUFs to less than the allowable;
: 2. Repair-of the affected components;
: 3. Replacemfent of the affected components; the-NRG.
NIJREG/C;R 6260 limiting locations piori to enter-ing the period of extende opeatin.SSES 'will implement one or-mor-e of the followinig if fatigue usage, includinig environmental eff-ects, is pr-ojected to exceed the allowable value of 1.0
: 1. Furfther-r-efinement of the fatigue analyses to lower-the CUFs to less than the allowable; 2     Rpnnir-of the aff-ented    mnet.
  -,    fl~~fl---
It    nepiaeuu*ment of tne anieeeud cumyi     cnUT; A                  1 4I. Ianagemcnt by an inspection program that has been r-eviewed and appr-oved-by the NRG.
*Acccptancc Cr-itcria The Fatigue Monitoring Program does not have specific acceptance cr-iter-ia that.
requirerective action if any CUFs are proejected to exceeed their-allowable values prior-to the end of the licensed period. The Fatigue Monitor-ing Program will be enhanced prior-to the period of extended operation such that corrective actioni is procedur-ally requir-ed should any CUF projections be unacceptable.
The Fatigue Moniitor-ing Proegr-am will incelude a reqireent to irmmediately notify

Enclosure to PLA-6397 Page 7 of 28 The Fatigue Moenitor-ing Program projects allow-able G1UIFs to the end Of the licensed period. if any projections exceed allowabics, the program will require action to be taken, such as a Condition Report being wr-itten to require r-esolution Resolution will consider-mor-e frequent monitoring of the data until anothei solution (reanalysis, component r-eplacement, etc.) is implemented-.
* Preventive Actions-The FatigueMonitoringProgram will be enhanced to ensure that the fatigue usage at all monitored locations, including those locations that accountfor the effect of the reactor water environment, is managed such that an adequate margin againstfatigue cracking is maintained.
PPL will implement one or more of thefollowing actions, iffatigue usage at a monitored location, includingany location that accountsfor the effect of the reactor water environment, is projected to reach the design basis limitprior to the end of the periodof extended operation:
: 1. Furtherrefinement of thefatigue analyses to lower the CUFs to less than the allowable;
: 2. Repair of the affected components;
: 3. Replacement of the affected components;
: 4. Managementby an inspectionprogram that has been reviewed and approved by the NRC.
" ParametersMonitored/Inspected-The FatigueMonitoringProgram will be enhanced to include the review of Class 1 valve fatigue analyses and otherfatigue-relatedTLAA, such asflued head analyses and high energy line break evaluations, when sufficientfatigue accumulation has occurred,to determine if additionalactions are required to addressfatigue-relatedconcerns.
* Monitoring and Trending-The FatigueMonitoringProgram will be enhanced to includefatigue monitoringof the additionallocations required to bound the limiting locations applicableto SSES, as identified in NUREG/CR-6260.
" Acceptance Criteria-The FatigueMonitoringProgram will be enhanced to establish monitoring criteriato ensure that the fatigue usage at all monitored locations, including those locations that accountfor the effect of the reactorwater environment, is

Enclosure to PLA-6397 Page 8 of 28 managedsuch that design basis limits are not exceeded during the period of extended operation. The FatigueMonitoringProgram will define specific fatigue usage valuesfor all monitored locations that, if reached, will require further action. These fatigue usage values shall be conservatively set to values that will allowfor not less than 4 years of additionalplant operationbefore the actualfatigue usage at any location would reach the design basis limit. Upon reachingthe defined usage at a location, the FatigueMonitoringProgram will require an action request to be generated. The action request will require further engineeringevaluation to resolve the issue.
Table A-1       SSES License Renewal Commitments The commitment for the SSES Fatigue Monitoring Program in Table A-i, Item 43 (LRA pages A-50 through A-52) is revised by addition (bold italics) and deletion (stfike4fe*gh*        ) as follows:
Table A-1 SSES License Renewal Commitments FSAR              Enhancement Supplement                  or Item Number                                    Commitment                           Location Location      Implementation (LRA App. A)            Schedule
: 43) Fatigue            Existing program is credited with the following                 A.-1.32         Prior to the period Monitoring         enhancements:                                                    AAI.I4         of extended Program                  Provisions will be made in the Fatigue Monitoring         A.35            operation.
Program to validate that components which have            A.1.2.49 satisfied ASME Section III, Paragraph N-415.1 requirements (i.e., RPV nozzles N6A, N6B, and N7) continue to satisfy these requirements prior to and during the period of extended operation, thereby allowing fatigue to be continued to be addressed under N-415.1.
P-rovisions will be made the Fatigue Mon~eitoig Program to address enviro-enta1 effects an fatigie at spe.ified locations in a...rda.n.e.       .ith NUP~G/CR6260.
The Fatigue Monitoring Pr-ogr-am will incluidee r-equir-ements to- review; Cl-ass 1 valve fatiguae analyses and other- fatigue rel atead T-1--- such as flueed head- -analysesand high ener-g' line bretakl reauatir~ons when~ sufficent fatigue a...umulatic has occuufed, to detenjne if ad-dition-al actions ar-e requir-ed to address fatiguae related concerns.

Enclosure to PLA-6397 Page 9 of 28 Table A-i SSES License Renewal Commitments FSAR          Enhancement Supplement             or Commitment                                   Location                 o  n Item Number (LRa ApA      Implementation (LRA App. A)        Schedule
* f    o,
* requir-emenfts for-cont~ued mofltor-ing and per-odic updates tc cunfent and proejected CFS &For lifflting locations. The proegram 'will include an approeach to addr es s CUPs th;A.1It   will exe ed- the -allowabl1edh  bedfore the end of the pedoid of extended operation+.
A requir-ements to monitor- bMAG/CR 6260 limiting locations prior-to ented-ng the period of extendeAd oper-ation. SSES 'will implaemen an approeach to address CU~s proejected to exceed the allowable before fatig~e, inceluding envro44a effects, he llowblevale0o 1.0. Should PPL&#xfd; exceds select the optionf to m~anage niomnal) assisted fatigue duding the period of extended
                *operation, details of the aging man~agement proegram such as scope, qualification, method, and frequency
                 -will be proevided to the:bW&#xfd;C prior-to the period of extended operation.
The Fatige Moffltod-ng Proegram does not have spec~ific acceptance cd-ted-a that r-equir-e conecetive action if any C=U:s are proeject-ed to- exce-Aed th eir MAWAfallwabe values prior-to the end of the licensed pedoed. The Fatigue Moffltod-ng Proegr-am 'will be&#xfd; enhanced pd~or to the pedoid of extended operationl suc-h that coecfftive action is proceedur ally, r-equired should an", CUF proejections be uniacceptable.
requirement to      iffiediately notify mnanagemnent if any, compoentp CUE is approeacfing its' design limith 0   T-he Fatigue Moffltod-ng Proagram proEjects allowabl-e CUmps to the end of the licensed pediod. if any&#xfd; proejections exceed allowables, the proegram will requir-e action to be taken, such as a Condition Repod4 being written to require r-esolution.
Resolution 'will consider- more frequent monitod-ng of the data until another solution (r-eanalysis, comfponent r-eplacement, aet.) is imnplemnented.
The, F2atigue Moffltod-ng Proegram tracks             itransiets and

Enclosure to PLA-6397 Page 10 of 28 Table A-1 SSES License Renewal Commitments FSAR              Enhancement Supplement                 or Commitment                                 Location            mleti Item Number                                                                                 Location      Implementation (LRA App. A)           Schedule ealeulates-,                   'uuAte fatigue usage factors (CUTs)-.9I any pfrojction exceeds allowable, then appr.eop*ate acltionflf is taken-pfl uni-Ml- the solut~ion r~esults in the proej ation The FatigueMonitoringProgram will be enhancedto ensure that the fatigue usage at all monitored locations,including those locations that accountfor the effect of the reactorwater environment,is managed such that an adequate margin againstfatigue cracking is maintained.
PPL will implement one or more of the following actions, iffatigue usage at a monitoredlocation, including any location that accountsfor the effect of the reactorwater environment, is projectedto reach the design basis limit prior to the end of the period of extended operation:
: 1. Furtherrefinement of thefatigue analyses to lower the CUFs to less than the allowable;
: 2. Repairof the affected components;
: 3. Replacement of the affected components;
: 4. Management by an inspectionprogram that has been reviewed and approved by the NRC.
* The FatigueMonitoringProgram will be enhanced to include the review of Class 1 valve fatigue analyses and otherfatigue-relatedTLAA, such asflued head analyses and high energy line break evaluations,when sufficient fatigue accumulationhas occurred, to determine if additionalactions are requiredto addressfatigue-relatedconcerns.
* The FatigueMonitoringProgram will be enhancedto includefatigue monitoring of the additionallocations requiredto bound the limiting locationsapplicableto SSES, as identified in NUREG/CR-6260.

Enclosure to PLA-6397 Page 11 of 28 Table A-1 SSES License Renewal Commitments FSAR           Enhancement Supplement             or Item Number                              Commitment                          Location Implementation (LRA App. A)        Schedule The FatigueMonitoring Programwill be enhanced to establish monitoring criteriato ensure that the fatigue usage at all monitored locations, including those locations that account for the effect of the reactorwater environment, is managed such that design basis limits are not exceeded during the periodof extended operation.
The FatigueMonitoring Program will define specificfatigue usage valuesfor all monitored locations that, if reached,will requirefurther action. Thesefatigue usage values shall be conservatively set to values that will allowfor not less than 4years of additionalplant operation before the actualfatigue usage at any location would reach the design basis limit. Upon reachingthe defined usage at a location, the Fatigue MonitoringProgramwill require an action request to be generated. The action request will requirefurther engineeringevaluation to resolve the issue.
RAI B.3.1-5:
Consistent with the GALL Program X.M1, the FMP, B.3. 1, will provide for periodic updates of the fatigue usage calculations. However, the LRA is not clear on the actions that will be taken if the updated calculations project higher than the allowable limits CUF values. State the exact actions that will be taken if the FMP projects higher than the allowable limit for the CUF values. Identify the procedure that would be used for these actions.
PPL Response:
When an "action level CUF" (as defined in the response to RAI B.3.1-3) is reached, an action request (AR) will be generated to require further engineering evaluation of the monitored component location. The evaluation will ultimately resolve the issue or determines the next action(s) to be taken to resolve the issue. These actions will be specified by the engineering procedure that implements the FMP.

Enclosure to PLA-6397 Page 12 of 28 The engineering evaluation will consist of the following actions:
: 1. Confirm the results indicating that an "action level CUF" has been reached. If the results are found to be incorrect, and the CUF is confirmed to be below the action level, document the findings and continue normal monitoring. If the results are confirmed, proceed to the next action (2).
: 2. Investigate possible refinements to the design basis fatigue analysis that may reduce the CUF by removing excess conservatism. Refinements may include the use of complex analysis techniques (e.g., 3-dimensional finite element analysis),
revised transient definitions (e.g., realistic heatup and cooldown rates, instead of step changes), and replacement of full design cycles with partial cycles (when possible). If there is a high probability of successfully lowering the CUF with a refined analysis, proceed to the next action (3).
: 3. Refine the fatigue analysis to lower the CUF to a value that provides acceptable margin for continued operation.
When the CUF cannot be analytically lowered to a value that provides acceptable margin for continued operation, there are three remaining options: 1) repair the component, 2) replace the component, or 3) manage the component through inspection and flaw tolerance evaluation.
Management of a high fatigue location is achieved through periodic non-destructive examination (NDE) coupled with a flaw tolerance evaluation. The flaw tolerance evaluation provides the basis for the frequency of the NDE. Repair techniques would typically only be used after a flaw has been identified in the component, or when a flaw tolerance evaluation cannot justify continued operation. Component replacement is the least likely option to be implemented, due to the potential complexity and cost.
However, at some point, replacement may be the only viable option.
Any corrective actions performed upon ASME Class 1 components will be in accordance with the SSES corrective action process and conform to the PPL Operational Quality Assurance Program, which satisfies 10 CFR 50 Appendix B. Weld repair, weld overlay, and complete replacement of ASME Code components would be performed in accordance with the applicable ASME Code sections. NRC approval for the use of any repair techniques that are not approved by the ASME Code would be obtained, in accordance with 10 CFR 50.5 5a.

Enclosure to PLA-6397 Page 13 of 28 RAI B.3.1-6:
The GALL Program X.M1 recommends that industry experience is reviewed as part of the program and any applicable experience should be considered to be incorporated into the FMP. The "operating experience" program element of the FMP, B.3.1 indicates that industry experience has been factored into the SSES FMP but did not list or describe the applicable operating experience that was reviewed. Please provide the list of documents reviewed by SSES in considering the industry experience on metal fatigue and provide the corresponding follow-up actions taken by SSES.
PPL Response:
Documents Reviewed as part of the Fatigue Monitoring Program
: 1. NRC Bulletin 79-13, Revision 2, "Cracking in Feedwater System Piping,"
June 1979.
: 2. NUREG-0619, "BWR Feedwater Nozzle and Control Rod Drive Return Line Nozzle Cracking," U.S. Nuclear Regulatory Commission, November 1980.
: 3. U.S. Nuclear Regulatory Commission Generic Safety Issue 78, "Monitoring of Fatigue Transient Limits for Reactor Coolant System."
: 4. U.S. Nuclear Regulatory Commission Generic Safety Issue 166, "Adequacy of Fatigue Life of Metal Components."
: 5. U.S. Nuclear Regulatory Commission Generic Safety Issue 190, "Fatigue Evaluation of Metal Components for 60-Year Plant Life."
: 6. NRC Bulletin 88-08, "Thermal Stresses in Piping Connected to Reactor Coolant Systems," June 22, 1988, Supplement 3, April 11, 1989.
: 7. NRC Bulletin 88-11, "Pressurizer Surge Line Thermal Stratification,"
December 10, 1988.
: 8. NRC Bulletin 91-38, "Thermal Stratification in Feedwater System Piping,"
June 13, 1991.
: 9. Policy Issue SECY-95-245, "Completion of the Fatigue Action Plan,"
September 25, 1995.
: 10. EPRI Report No. NP-5 835, "FatiguePro: An On-Line Fatigue Usage Transient Monitoring System for Nuclear Power Plants," April 1988.

Enclosure to PLA-6397 Page 14 of 28
: 11. EPRI Report No. NP-6170-M, "FatiguePro: On-Line Fatigue Monitoring System: Demonstration at the Quad Cities BWR," January 1989.
: 12. EPRI Report No. TR- 107448, "FatiguePro, Version 2: Fatigue Monitoring Software," December 1997.
: 13. EPRI Report No. TR- 103843, Revision 1, "BWR Primary Coolant Pressure Boundary License Renewal Industry Report," July 1994.
: 14. EPRI Report No. TR-100281, "User's Manual for FatiguePro Fatigue Monitoring System, Version 3.0," 2001.
: 15. EPRI Report No. TR- 103844, Revision 1, "PWR Primary Coolant Pressure Boundary License Renewal Industry Report," July 1994.
: 16. NEI 95-10, "Industry Guideline for Implementing the Requirements of 10 CFR Part 54 - The License Renewal Rule."
: 17. NUREG- 1801, "Generic Aging Lessons Learned (GALL) Report," Volumes 1 and 2, Revision 1.
: 18. NUREG/CR-6260, "Application of NUREG/CR-5999 Interim Fatigue Curves to Selected Nuclear Power Plant Components," March, 1995.
: 19. Harris Nuclear Plant License Renewal Application, November, 16, 2006.
: 20. Vermont Yankee Nuclear Power Station License Renewal Application, January 27, 2006.
: 21. Monticello License Renewal Application, March 24, 2005.
From the time of plant design and construction, PPL was aware of the issue of metal fatigue from NRC documents such as Bulletin 79-13 and NUREG-0619 (reviewed documents 1 and 2 above). Design changes were implemented for the feedwater thermal sleeve/sparger and the CRD return line nozzle prior to plant startup based on the information in these documents.
SSES was an early industry participant in using fatigue monitoring of reactor coolant pressure boundary components. As an example, reviewed documents 3 through 11, above, were instrumental in PPL's development of the first SSES FatiguePro system in 1989.

Enclosure to PLA-6397 Page 15 of 28 EPRI documents related to FatiguePro, such as reviewed documents 10 through 15, above, were useful to PPL's on-going development and improvement of the SSES FatiguePro system and the fatigue monitoring program (FMP). For License Renewal, documents such as reviewed documents 16 through 21, above, provided guidance for the enhancements to the FMP to meet the'requirements'of License Renewal.
In addition to the industry documents listed above, operating experience was gained through participation in industry peer groups, such as the FatiguePro Users group and NEI License Renewal working groups.
RAI B.3.1-7:
Section 4.3.3 provides the result of CUFs including environmental effects for NUREG/CR-6260 locations:
a) Provide the details of the management of environmentally-assisted fatigue components during the period of extended operation, including elements to the AMP such as scope, qualification, method and frequency.
b) For all locations where 60-year environmental CUF is below 1.0, clarify whether any of these values were calculated using the Green's Function Methodology and if so please describe the details of how Green's Function was used to calculate the CUF values.
PPL Response:
Part a):
All eleven locations in LRA Table 4.3-3 will be managed for environmentally-assisted fatigue by the SSES Fatigue Monitoring Program (FMP). The assessment of the environmentally-assisted fatigue at these eleven locations will be performed in accordance with NUREG/CR-6583 for carbon and low alloy steels and NUREG/CR-5704 for austenitic stainless steels. The cumulative usage factor (CUF), as corrected for environmentally-assisted fatigue, at these eleven locations will be monitored and evaluated as part of each fatigue monitoring update calculation that is performed at least once per fuel cycle (every 2 years) for each unit. And, as described in the response to RAI B.3.1-3, the use of projected CUFs will allow for corrective actions to be initiated at least 4 years before any monitored location's CUF would reach the allowable CUF.
Part b):
Four of the eleven locations in LRA Table 4.3-3 have 60-year environmental CUFs below 1.0. Two of those four locations, the feedwater nozzle forging and the feedwater nozzle safe end, are monitored by the FatiguePro SBF methodology, which includes the

Enclosure to PLA-6397 Page 16 of 28 use of Green's Functions.
The standard method by which Green's Functions are employed in the calculation of CUFs by the FatiguePro software is detailed in Section 3 of the EPRI Report NP-5835M, "FATIGUEPRO: On-Line Fatigue Usage Transient Monitoring System," May 1988.
The Green's Functions for the SSES feedwater nozzle locations were developed for the purpose of computing thermal stresses based on actual flow, temperature, and pressure conditions computed at the local nozzle locations. Using a detailed finite element model of the SSES feedwater nozzle and safe end, the Green's Functions were developed for the applicable feedwater flow conditions for the nozzle forging (inner blend radius) and safe end locations. The thermal stress intensity, Stherma], is then determined by the following equation:
Sthermal = fGrloc
* T1 oc
* dt where:         Grloc    = appropriate Green's Function for the location and flow condition (psi/&deg;F)
Tloc= the feedwater fluid temperature in the nozzle minus the reactor coolant temperature (OF) dt       = integration with respect to time, t (seconds)
The thermal stress intensity is then added to the other applicable stress intensities (e.g.,
pressure, piping thermal moment, cladding thermal stress, SRV load, and top-to-bottom stratification) to determine the total stress intensity, which is then used in the determination of the fatigue usage.
RAI B.3.1-8:
10 CFR 54.21(d) states that the FSAR supplement for the facility must contain a summary description of the programs and activities for managing the effects of aging.
Provide the FSAR supplement for Fatigue Monitoring Program.
PPL Response:
The LRA is amended to provide a FSAR supplement section for the Fatigue Monitoring Program to the LRA Appendix A.

Enclosure to PLA-6397 Page 17 of 28 Appendix A - Table of Contents
 > The Table of Contents for Appendix A (page A-3) is revised by addition (bold italics) as follows:
A. 1.2.48       Thermal Aging and Neutron Embrittlement of Cast Austenitic Stainless Steel (CASS) Program ..........................................                       21 A.1.2.49        FatigueMonitoringProgram                   ................      .     ............ 21 Appendix A LRA Appendix A (page A-21) is revised by addition (bold italics) of the following new section:
A.1.2.49        FatigueMonitoringProgram The FatigueMonitoringProgrammanagesfatiguefor all Class 1 components, including the reactorpressure vessel. In ordernot to exceed the design basis limit on fatigue usage, the aging managementprogram monitors and tracks the number and severity of criticalthermal andpressure transientsand calculates thefatigue usagefor the limiting locations of the reactor coolantpressure boundary.
Priorto the period of extended operation the program will be enhanced by addingthe following requiredactions:
    " The program will verify that components which have satisfiedASME Section III, ParagraphN-415.1 requirements(i.e., RPV nozzles N6A, N6B, and N7) continue to satisfy these requirementsprior to and during the period of extended operation,thereby allowingfatigue to be continued to be addressedunder N-415.1.
    "    The program will review Class 1 valve fatigue analyses and otherfatigue-related TLAA, such asflued head analyses and high energy line break evaluations,
        .when sufficientfatigue accumulation has occurred,to determine if additional actions are requiredto addressfatigue-relatedconcerns.
* The program will define specific fatigue usage values for all monitored locations, includingthose locationsthat accountfor the effect of the reactor water environment, that, if reached,will requirefurther action. These fatigue usage values shall be conservatively set to values that will allowfor not less than 4 years of additionalplantoperationbefore the actualfatigue usage at any location would reach the design basis limit. Upon reaching the defined usage at

Enclosure to PLA-6397 Page 18 of 28 a location, the program will require an action request to be generated. The action request will requirefurther engineeringevaluation to resolve the issue.
The program will implement one or more of the following actions, iffatigue usage at a monitored location, includingany location that accountsfor the effect of the reactor water environment, is projected to reach the design basis limitpriorto the end of the period of extended operation:
: 1. Furtherrefinement of the fatigue analyses to lower the CUFs to less than the allowable;
: 2. Repair of the affected components;
: 3. Replacement of the affected components;
: 4. Management by an inspection program that has been reviewed and approvedby the NRC.
RAI 4.3-1:
Table 4.3-1 of the LRA provides the reactor design transients and 60-year cycle projections:
a) For each transient listed, provide the cycles accrued to date and explain how the 60-year projections values were calculated.
b) Table 3.9-1 of FSAR provides different number of design cycles (than the LRA) for loss of feedwater heaters and pre-op blowdown. Please explain the discrepancy.
c) Scram transient is comprised of turbine generator trip and other scrams, explain why these scrams are grouped together and projected together in the LRA, including how scram transient conditions bound all scrams that may be experienced by the plant.
d) Loss of feedwater heaters and partial feedwater heater bypass are combined as a single transient in the LRA, provide the basis for this, including the effect on feedwater nozzle CUF analysis.
PPL Response:
Part a):
For each transient listed in LRA Table 4.3-1, the table below provides the total number of cycles accrued as of the date shown, 12/31/2002 for Unit 1 and 4/13/2003 for Unit 2.
The 60-year projections were performed using a forward projection method that accounts for the "learning curve" effects of early plant operation by using trending from the last ten years of plant operation, as follows:

Enclosure to PLA-6397 Page 19 of 28 N 60 = Niatest + [ (Niatest - Nprior) / Timeintera. ]
* Timefuture where:           N 60          =   projected number of cycles for 60 years.
Nlatest       = number of cycles as of the latest valid cycle count for the entire operational history.
Nprior        = number of cycles at ten years prior to the latest valid cycle count for the entire operational history.
Timeinterval = elapsed number of days (covering approximately ten years) between the latest and prior cycle counts.
Timefiture = number of days from latest cycle count to the end of the 60-year operating period.
Reactor Design Transients and Current Cycle Count Current Cycle Count Transient                          No. of Design          Unit I          Unit 2 Normal, Upset, and Test Conditions                     Cycles            As of           As of 12131/2002      4/13/2003 Bolt Up                                                       123                20              15 Design Hydrostatic Test                                       130                22              15 Startup (100&deg;F/hr Heatup Rate)(1 )                            117                70              59 Loss of Feedwater Heaters,                                     70                14              20 Partial Feedwater Heater Bypass 50% Safe Shutdown Event at Rated Operating                    10(2)               0               0 Conditions Scram:
: a. Turbine Generator Trip, Feedwater On,                   180                39              14 Isolation Valves Stay Open
: b. Other Scrams Reduction to 0% power, hot standby with main condenser available, shutdown (100&deg;F/hr                       111                69              59 cooldown)(1)
Unbolt                                                        123                19              14 Blowdown                                                        9                 0               0 Natural Circulation Startup                                     3                 0               0 Loss of AC Power, Natural Circulation Restart                   5                 0               0

Enclosure to PLA-6397 Page 20 of 28 Part b):
The number of cycles for the Loss of Feedwater Heaters event in FSAR Table 3.9-1 is listed as a total of 80. The reactor vessel thermal cycle diagram, which defines the transients considered in the SSES RPV design stress and fatigue analyses, identifies two loss of feedwater heating events: (1) Partial Feedwater Heater Bypass (70 cycles) and (2)
Loss of Feedwater Heaters- Turbine Trip with 100% Steam Bypass (10 cycles). The latter event does not apply to SSES, since the SSES bypass system is only capable of 25% bypass flow. Therefore, Table 4.3-1 only lists the number of cycles (70) for the Partial Feedwater Heater Bypass event.
Table 3.9-1 of the FSAR lists 10 Pre-op Blowdown events. None of these' events were experienced by either SSES RPV in the pre-operational phase (prior to the initial plant nuclear startup), and this event is not applicable following initial startup. Therefore, the Pre-op Blowdown events are not tracked by the Fatigue Monitoring Program, and they are not listed in LRA Table 4.3-1. Instead, LRA Table 4.3-1 lists the nine (9) Blowdown events, consisting of eight (8) Single Relief or Safety Valve Blowdown events and one (1) Automatic Blowdown event, as detailed on the reactor vessel thermal cycle diagram.
Part c):
The total number of cycles for the Scram events in Table 4.3-1 is 180. The FSAR Table 3.9-1 separates scrams into two categories: Turbine Generator Trips (40 cycles) and Other Scrams (140 cycles). As detailed on the reactor vessel thermal cycle diagram, Turbine Generator Trips have a slight pressure and temperature spike at the beginning of the event that is not included in the Other Scrams. The FatiguePro software is configured to assume the most limiting conditions for all scram events that are counted. Therefore, the fatigue effects for each scram that is counted are conservatively assessed as Turbine Generator Trips, regardless of the actual initiating cause.
Part d):
The response to Part b) of this question explains why only Partial Feedwater Heater Bypass events are counted by the SSES Fatigue Monitoring Program. As discussed in the response to RAI B.3.1-2, the feedwater nozzle forging and safe end locations are monitored by the stress-based fatigue (SBF) methodology in FatiguePro. As such, the number of cycles is not a factor in the determination of the fatigue usage. Instead, actual plant conditions, which directly affect the locations, are continuously evaluated, and the incremental effect on fatigue usage is determined from the computed stress history.

Enclosure to PLA-6397 Page 21 of 28 RAI 4.3-2:
Table 4.3-2 of the LRA provides the Fatigue Usage for Limiting Reactor Coolant Pressure Boundary Locations. Provide a complete list of design transients that were used to calculate the 60-year CUF projections for the components listed, and describe the details of how they are monitored.
PPL Response:
The complete list of design transients (or events) that were considered in the calculation of the 60-year CUF projections for the limiting Reactor Coolant Pressure Boundary locations are provided in the response to RAI B.3.1-1.
The SSES Fatigue Monitoring Program (FMP) utilizes the EPRI FatiguePro software to identify and track (i.e., monitor) the occurrence of the monitored plant events. Through the engineering evaluation and FatiguePro analysis of collected plant data, plant event occurrences are identified. Certain plant events are automatically counted by FatiguePro.
This is accomplished by the logical interpretation of specific plant data (e.g.,
temperatures, pressures, flow rates, and valve positions). While virtually every plant event could be automatically identified, there are some plant events that are easier to identify and track manually. The manually counted events are identified by engineering (i.e., manual) review of plant data and operating logs. Once identified, the manual events are recorded in the FatiguePro database for tracking and subsequent use in the fatigue usage calculations.
RAI 4.3-3:
Table 4.3-3 indicates that the reactor pressure vessel (RPV) (Shell at Shroud Support)
CUF, including the environmental effects, is projected to exceed 1.0 prior to the period of extended operation. List the transients and corresponding usage factors used for the evaluation. Also, provide the environmental factors used for the evaluation.
PPL Response:
The complete list of design transients (or events) that were considered in the calculation of the 60-year environmental CUF projection for the limiting reactor vessel location (Shell at Shroud Support) is provided in the response to RAI B.3.1-1.
The environmental fatigue multiplier, Fen, methodology was used to perform the environmental fatigue calculations for the locations identified in LRA Table 4.3-3. The methodology, as provided in NUREG/CR-6583 (for carbon/low alloy steels) and NUREG/CR-5704 (for stainless steels), was applied, as appropriate, for the material of the component at each location.

Enclosure to PLA-6397 Page 22 of 28 The environmental multiplier, Fen, was determined for each location as an overall value to bound all operating conditions over the 60-year operating period. The multipliers were determined, as follows:
Fen = (Fen-HWC
* PHWC) + (Fen-NWC
* PNWC) where:   Fen     =   Overall environmental multiplier FenHWC =    Hydrogen Water Chemistry (HWC) multiplier PHWC    =   Percentage of operating life with HWC Fen-NWC =   Normal Water Chemistry (NWC) multiplier PNWC    = Percentage of operating life with NWC The environmental fatigue is then determined as:
Uenv = (U) (Fen) where:           U    =   60-Year projected CUF Uenv =   60-Year environmental CUF The Unit 1 Reactor Vessel location had a projected 60-year CUF of 0.3121. The application of the bounding environmental multiplier, Fen, of 13.04, results in a 60-year environmental CUF of 4.070, as reported in LRA Table 4.3-3.
The Unit 2 Reactor Vessel location had a projected 60-year CUF of 0.2970. The application of the bounding environmental multiplier, Fen, of 12.44, results in a 60-year environmental CUF of 3.695, as reported in LRA Table 4.3-3.
RAI 4.3.-4:
BWIR Vessel Internals Program is credited to manage the effects of aging for the reactor vessel internals. However, this AMP only inspects the top guide for the first twelve years of period of extended operation. Top guide is subject to irradiation assisted stress corrosion cracking, state how this aging effect will be managed for the remainder for the period of extended operation.
PPL Response:
During the period of extended operation, the aging of the top guide will be managed by inspections conducted as part of the SSES BWR Vessel Internals Program. PPL is

Enclosure to PLA-6397 Page 23 of 28 committed to follow the inspections and flaw evaluations for reactor internals issued by the BWR Vessel and Internals Project (BWRVIP). In December of 2007, the BWRVIP issued BWRVIP- 183, "Top Guide Grid Beam Inspection and Flaw Evaluation Guidelines," which specifies that the inspections for the top guide must be performed on at least 10% of the grid beam cells containing control rod drives/blades every twelve years with at least 5% performed within the first six years of each twelve year interval.
PPL has incorporated the inspection frequency and sample size of BWRVIP-183 into the BWR Vessel Internals Program, and PPL will continue to use this guidance through the end of the period of extended operation. Thus, inspections will be performed on at least 10% of the top guide locations every twelve years during the period of extended operation. If flaws are found, evaluations will be performed and, if necessary, the scope will be expanded in accordance with the guidance provided in BWRVIP- 183.
Since the incorporation of BWRVIP-183 into the SSES BWR Vessel Internals Program makes the program consistent with NUREG-1801, XI.M9, the enhancement to the BWR Vessel Internals Program described in the SSES LRA is not needed. The SSES LRA is amended to remove the enhancement from LRA Section A. 1.2.10, LRA Table A-I (Commitment #9), LRA Section B.2.9, and LRA Table B-2.
A.1.2.10       BWR Vessel Internals Program The discussion of the SSES BWR Vessel Internals Program in Section A.1.2.10 (LRA page A-8) is revised by deletion (str-kethfo*gh) as follows:
Prior-to the period of extended oper-ation, the BVAR Vessel Internals Pr-ogr-am will be enhianced treuespecific enhanced visuial examinations of top guide location subjected to high neutron fluence.
Table A-1      SSES License Renewal Commitments The commitment for the SSES BWR Vessel Internals Program in Table A-1, Item 9 (LRA A-35) is revised by addition (bold italics) and deletion (s+i*kedffeeigh) as follows:
Table A-1 SSES License Renewal Commitments FSAR             Enhancement Supplement                 o Commitment                           Location                  or Item Number                                                                        LocationImplementation (LRA App. A)            Schedule
: 9) BWR Vessel        Existing program is credited. with the fellowing             A.1.2.10     Prior-to the period a Internals Program           [
euhaneemepA:
                           .- r-equir.emens to inspe.t five per.ent of the top Inlu.de guaide locations within six years after entering the extended ope..ation.
Ongoing

Enclosure to PLA-6397 Page 24 of 28 Table A-1 SSES License Renewal Commitments FSAR              Enhancement Supplement                   or Commitment                                       Location            m        n Item Number                                                                                                          LocationImplementation (LRA App. A)                 Schedule period of extenided Oper-aticn anid an additicnfal fi;'c percentA of the tcp guijde laIon wAithin twelve year-s after entering the pericd Af extended cperaticn. The top guide lecations to be inspeted are those stubjetto neut-Aron4fluffence levels that wexeed the 1ASGC thfeshoA]d of 5.OE+20 Pn'cm2. The inspectionfs shall be performed udsing the enihanced Visual infSpection technique, EV'T 1.
The exten-;t o-f th-e                          its fr-equencey will be fxf~ainand-b-ased-A-n; a ten percentA sample of the total population-,
                              .vhieh includes all grid beamn and beam to beam ereviee slets,.
B.2.9           BWR Vessel Internals Program The discussion under the NUREG-1801 Consistency and Required Enhancements in LRA Section B.2.9 (LRA page B-34) is revised by addition (bold italics) and deletion (stikethfeugh) as follows:
NUREG-1801 Consistency The BWR Vessel Internals Program is an existing SSES program that, with enhance*ment-,
will.be that is consistent with the 10 elements of an effective aging management program as described in NUREG-1801, Section XI.M9, "BWR Vessel Internals."
Exceptions to NUREG-1801 None.
Required Enhancements None.
1    1             ~1        ~                    -:z ]1 1__~1 ti1'SIL~JLIIt~               I          XIA~~~t.Fkt4.LA~JkktIk implemented in the identified pr-ogram elem~enit The proegr-am will include r-equir-ements to inspect five perceent of the top guide locationgs within six year-s after-enter-ing the period of extended operation and an
                     * *o, o       * *,"                                                         ",1 "   ,       1           P',    *  "
acidiltional live percent otr tfl top, guiae locations witain twelve years alter-enterng

Enclosure to PLA-6397 Page 25 of 28 subjet to neutron..           fluence levels that exceed    .ASCC the            theshold of S.**.20 I   - -  "l- i'-t T'i" 1     l ..-  s       _-.11     -   - -  - -   - t- -n       -  -
visa --    1--    i-..
          -EcfNnlue, -11    "     1. Thne extent     01     mhe examination and its ffegueney will be based on a ten percenft sample of the total population, which ineludes all gr-id beam and h'iwm- to hpenm cr                     Jpelkt Table B-2 Consistency of SSES Aging Management Programs with NUREG-1801
> LRA Table B-2 (LRA page B-14) is revised by addition (bold italics) and deletion (str-kethiough) as follows:
Table B-2 Consistency of SSES Aging Management Programs with NUREG-1801 New I            Consistent          Exceptions       Plant-        Enhancement Program Name                  Existing          with NUREG-           to NUREG-       Specific         Required 1801                1801            -
BWR Vessel ID Attachment Welds                  Existing                Yes                 --....
Program BWR Vessel Internals Program                Existing                Yes                 ....                          Yes--
BWR Water Chemistry Program                 Existing                Yes                  ......
RAI 4.3.-5:
For Section 4.3.4, provide the technical basis for this statement "fatigue usage is typically much higher on the associated piping systems."
PPL Response:
The statement "fatigue usage is typically much higher on the associated piping systems" is related to fatigue on ASME Class 1 valves in the reactor coolant pressure boundary.
Class 1 valves are not analyzed as part of the Class 1 fatigue analysis of the piping systems in which the valves are installed. Class 1 valves have individual design stress reports, typically prepared by the valve manufacturer, which address fatigue, in accordance with the design specifications for the valves.
As stated in LRA Section 4.3.4, a review of a representative sample of Class 1 valve design stress reports determined that the fatigue analyses were very conservative, and the predicted fatigue usage values were low (less than 0.1). Therefore, the basis for the statement "fatigue usage is typically much higher on the associated piping systems" is

Enclosure to PLA-6397 Page 26 of 28 that typical valve CUFs are less than 0.1, while the design CUFs presented in Table 4.3-2 for the piping system components are greater than 0.1, except in "no-break" zones.
RAI 4.3.-6:
For Section 4.3.5, indicate whether any of the numbers of projected cycles for piping and in-line components were derived from evaluation of "partial cycle" transients, where partial cycles are transient cycles that do not experience the full-temperature design cycles. If yes, provide the equation used to calculate the number of projected cycles.
PPL Response:
The SSES Fatigue Monitoring Program (FMP) assumes that every event occurs at the severity equal to that assumed in the design basis (i.e., full cycle). No partial cycles are considered by the FMP.
The 60-year cycle projections for the non-Class 1 piping and in-line components addressed in LRA Section 4.3.5 are identical to the FMP 60-year projection of full cycles.
Therefore, the numbers of projected cycles for non-Class 1 components were not derived from the evaluation of any "partial cycle" transients.
RAI 4.3.-7:
The FSAR supplement of the LRA does not disposition the TLAA for metal fatigue analysis nor the HELB analysis according to 10 CFR 54.21 (c)(i), (ii), or (iii). Explain why the disposition of these TLAAs is not necessary.
PPL Response:
The TLAA dispositions in accordance with 10 CFR 54.21 (c)(1) for the metal fatigue and the HELB analyses are provided in LRA Sections 4.3, 4.6, and 4.7. For clarity, the LRA is amended, as shown below, to include statements of the disposition of these TLAA according 10 CFR 54.21 (c)(1) in the FSAR supplement provided in Appendix A.
A.1.3 Evaluation of Time-Limited Aging Analyses The discussions of the fatigue-related TLAA evaluations in Section A. 1.3 (on LRA pages A-25 through A-32) are revised by the addition of statements to the end of each identified section of Appendix A (addition in bold italics).

Enclosure to PLA-6397 Page 27 of 28 A.1.3.2.1 Reactor Pressure Vessel Fatigue Analysis The FatigueMonitoringProgram is creditedfor managingthe effects offatigue during the period of extended operation. Therefore, the TLAA associatedwith reactor pressure vessel fatigue are dispositionedin accordancewith 10 CFR 54.21(c)(1)(iii).
A.1.3.2.2 Reactor Vessel Internals Fatigue Analyses Structuralevaluations have demonstratedthatfatigue usage will remain within design limits to the end of the period of extended operation. Also, the BWR Vessel Internals Programis creditedfor managingthe aging effects of the reactor vessel internals during the periodof extended operation. Therefore, the TLAA associatedwith fatigue of the reactor vessel internals are dispositionedin accordancewith 10 CFR 54.21(c)(1)(ii) and 10 CFR 54.21(c)(1)(iii).
A.1.3.2.3 Effects of Reactor Coolant Environment on Fatigue Life of Components and Piping (Generic Safety Issue 190)
The FatigueMonitoringProgramis creditedfor managingthe effects of the reactor coolantenvironmental effects on fatigue during the period of extended operation.
Therefore, the TLAA associatedwith environmentally-assistedfatigue has been dispositionedin accordancewith 10 CFR 54.21(c)(1)(iii).
A.1.3.2.4 Reactor Coolant Pressure Boundary Piping and Component Fatigue Analyses The FatigueMonitoringProgram is creditedfor managingthe effects offatigue during the period of extended operation. Therefore, the TLAA associatedwith fatigue of the reactorcoolantpressure boundarypiping and components have been dispositionedin accordancewith 10 CFR 54.21(c)(1)(iii).
A.1.3.3 Non-Class 1 Component Fatigue Analyses The TLAA associatedwith thefatigue of non-Class 1 components have been dispositionedin accordancewith 10 CFR 54.21(c)(1)(i).
A.1.3.5.1 ASME Class MC Components The TLAA associatedwith thermal cycles on the ASME Class MC components have been dispositionedin accordance with 10 CFR 54.21(c)(1)(i).

Enclosure to PLA-6397 Page 28 of 28 A.1.3.5.2 Downcomer Vents and Safety Relief Valve Discharge Piping The TLAA associatedwith stress cycles on the downcomer vents and safety relief valve dischargepiping have been dispositionedin accordancewith 10 CFR 54.21 (c)(1)(i).
A.1.3.5.3 Safety Relief Valve Quenchers The TLAA associatedwith stress cycles on the safety relief valve quenchers have been dispositionedin accordance with 10 CFR 54.21(c)(1)(i).
A.1.3.6.2 High Energy Line Break Cumulative Fatigue Usage Factors The FatigueMonitoringProgram is creditedfor managing the effects offatigue during the period of extended operation. Therefore, the TLAA associatedwith high energy line break cumulativefatigue have been dispositioned in accordancewith 10 CFR 54.21 (c) (1) (iii).
RAI 4.7.2-1:
Please clarify whether there are any Class 1 high energy piping locations with a CUF value less than 0. 1 by the current design basis where the CUF may exceed 0.1 during the period of extended operation.
PPL Response:
There are locations with a CUF value of less than 0.1 by the current design basis where the CUF may exceed 0.1 during the period of extended operation.
PPLhas committed to implement an enhancement to the SSES Fatigue Monitoring Program (FMP), as stated in LRA Table A-i, Item 43, to review the high energy line break evaluations, when sufficient fatigue accumulation (as defined in the response to RAI B.3.1-3) has occurred. The enhancement to the FMP will ensure that the high energy piping continues to meet the applicable criteria for the "Protection Against Dynamic Effects Associated with the Postulated Rupture of Piping," as specified in FSAR Section 3.6, during the period of extended operation.}}