L-08-287, Supplement to Reply to Request for Additional Information for the Review of the License Renewal Application and License Renewal Application Amendment No. 25

Supplement to Reply to Request for Additional Information for the Review of the License Renewal Application and License Renewal Application Amendment No. 25

ML082800177
Person / Time
Site:	Beaver Valley
Issue date:	10/02/2008
From:	Brosi R FirstEnergy Nuclear Generation Corp
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
L-08-287, TAC MD6593, TAC MD6594
Download: ML082800177 (22)

Text

SrENOCf Beaver Valley Power Station P.O. Box 4

- _0% Shippingport,PA 15077 FirstEnergyNuclear Operating Company Roy K. Brosi 724-682-5234 Director,Site Performance Improvement Fax: 724-643-8069 October 2, 2008 L-08-287 10 CFR 54 ATTN: Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555-0001

SUBJECT:

Beaver Valley Power Station, Unit Nos. 1 and 2 BV-1 Docket No. 50-334, License No. DPR-66 BV-2 Docket No. 50-412, License No. NPF-73 Supplement to Reply to Request for Additional Information for the Review of the Beaver Valley Power Station, Units 1 and 2, License Renewal Application (TAC Nos. MD6593 and MD6594) and License Renewal Application Amendment No. 25 Reference 1 provided the FirstEnergy Nuclear Operating Company (FENOC) License Renewal Application (LRA) for the Beaver Valley Power Station (BVPS). Reference 2 provided the FENOC reply to a U.S. Nuclear Regulatory Commission (NRC) request for additional information regarding BVPS license renewal time-limited aging analyses (TLAA) related to metal fatigue in Sections 4.3 and B.2.27 of the BVPS LRA.

During conference calls between FENOC and the NRC on August 28, 2008 and September 4, 2008, related to the FENOC reply in Reference 2, the NRC staff asked for supplements to the responses to the NRC requests for additional information (RAIs) for RAIs 4.3-1, 4.3-3, 4.3-11 and 4.3-16 to clarify the information submitted. This letter provides the FENOC supplemented response to NRC RAIs 4.3-1, 4.3-3, 4.3-11 and 4.3-16. This letter also provides Amendment No. 25 to the BVPS LRA, including revised license renewal future commitments, based on changes resulting from the FENOC supplemental responses to the NRC RAIs.

The Attachment provides the FENOC replies to the request for supplemental information. The Enclosure provides Amendment No. 25 to the BVPS LRA.

There are no regulatory commitments contained in this letter. If there are any questions or if additional information is required, please contact Mr. Clifford I. Custer, Fleet License Renewal Project Manager, at 724-682-7139.

Beaver Valley Power Station, Unit Nos. 1 and 2 L-08-287 Page 2 I declare under penalty of perjury that the foregoing is true and correct. Executed on October I2- , 2008.

Sincerely, Roy K. Brosi

References:

1. FENOC Letter L-07-113, "License Renewal Application," August 27, 2007.

2. FENOC Letter L-08-209, "Reply to Request for Additional Information for the Review of the Beaver Valley Power Station, Units 1 and 2, License Renewal Application (TAC Nos. MD6593 and MD6594), and License Renewal Application Amendment No. 15," July 11, 2008.

Attachment:

Supplement to the Response to Request for Additional Information Regarding Beaver Valley Power Station, Units 1 and 2, License Renewal Application, Section 4.3

Enclosure:

Amendment No. 25 to the BVPS License Renewal Application cc: Mr. K. L. Howard, NRC DLR Project Manager Mr. S. J. Collins, NRC Region I Administrator Mr. J. E. Richmond, NRC Region I DRS/EB1 cc: w/o Attachment or Enclosure Mr. B. E. Holian, NRC DLR Director Mr. D. L. Werkheiser, NRC Senior Resident Inspector Ms. N. S. Morgan, NRC DORL Project Manager Mr. D. J. Allard, PA BRP/DEP Director Mr. L. E. Ryan, PA BRP/DEP

ATTACHMENT L-08-287 Supplement to the Response to Request for Additional Information Regarding Beaver Valley Power Station, Units 1 and 2, License Renewal Application, Section 4.3 Page 1 of 9 RAI-4.3-01 NRC Follow-up Questions (Conference Call August 28, 2008):

New Question 1 Line I of the initial response (page 20 of FENOC Letter L-08-209) lists dates that seem to conflict with statements regarding monitoring start dates that appear in the referenced letters. The letters state that data collected began before the dates cited in the RAI response. FENOC stated during the conference call that data collection did start before the dates listed in the response, but that data collection was for the establishment of a baseline, and once the baseline was established FENOC deemed the monitoring began on the dates listed in the response. Please provide clarification.

RESPONSE for New Question 1 Thermocouple data collection for the establishment of a baseline was commenced on the monitoring start dates that appear in the referenced letters (References 1 and 2).

Therefore, the 1 st paragraph is revised as follows:

Collection of thermocouple monitoring data commenced in June 1989 (startup from the first refueling) for Unit 2 and in December 1989 (startup from the seventh refueling) for Unit 1, this data collection was suspended in 2002.

New Question 2 The initial response states that" .... renewed thermocouple monitoring may be required for some lines." The phrase "may be required" is too vague. FENOC stated during the conference call that the requirement for monitoring will be determined and tracked via MRP-146.

Please provide a commitment added to Appendix A stating that monitoring will be done in accordance with MRP-146.

Attachment L-08-287 Page 2 of 9 RESPONSE for New Question 2 FENOC provides a future License Renewal commitment to implement. "needed actions" of MRP-146 (Reference 4) as follows:

FENOC will implement "needed actions" of MRP-146. These actions include screening, detailed analysis, inspections and temperature monitoring in accordance with the guidelines of MRP 146. FENOC has completed screening of the BVPS RCS branch lines.

References:

1. Sieber, John D. (BVPS), Letter to NRC, "Beaver Valley Power Station Unit No. 2, Docket No. 50-412, License No. NFP-73, NRC Bulletin 88-08," 7/14/1989 (NRC PDR Ascension Number 8907240226)

2. Sieber, John D. (BVPS), Letter to NRC, "Beaver Valley Power Station Unit No. 1, Docket No. 50-334, License No. DPR-66, NRC Bulletin 88-08," 2/7/1990 (NRC PDR Ascension Number 9002150239)

3. EPRI Technical Report 1000701, Interim Thermal Fatigue Management Guideline (MRP-24)," January 2001

4. EPRI Technical Report 1011955, "Management of Thermal Fatigue in Normally Stagnant Non-Isolable Reactor Coolant System Branch Lines (MRP-146),"

June 2005 See the Enclosure to this letter for the revision to the BVPS LRA.

Attachment L-08-287 Page 3 of 9 RAI 4.3-03 NRC Follow-up Questions (Conference Call September 4, 2008):

NUREG/CR-6934 is not endorsed by the NRC. If FENOC selects the analysis option that uses general methodology as described in NUREG/CR-6934 the following statement is required: "This option will require NRC review and approval." The previously provided commitment will need to be changed. Noting FENOC's present course for data reduction on the number of transients (Unit I Pressurizer Surge line Hot Leg Nozzle), that FENOC has reasonable expectation to be successful (60-year cumulative usage factor (CUF) with EAF considerations

< 1.0), that FENOC has two paths they are pursuing (data reduction and fracture mechanics), and that when FENOC is done they will submit those results, please state that regarding (c)(1)(iii), FENOC will do one of the three options (with the above-mentioned NRC review and approval) and add it to the Metal Fatigue program as an enhancement. Then it will not be an open item.

RESPONSE RAI 4.3-03 NRC Follow-up Questions The response to the original RAI 4.3-03 from Letter L-08-209 is replaced in its entirety with the following:

a. Please provide the schedule for refining the analysis for the environmental assisted fatigue (EAF) of the NUREG/CR-6260 (Reference 1) locations in which the cumulative usage factor includes environmental effects (Uenv) exceed the design code allowable value of 1.0.

The refined analyses for Unit 1 Charging System Nozzle, Unit 2 Charging System Nozzle, Unit 2 Safety Injection System Nozzle, and Unit 2 Residual Heat Removal System Piping are completed. For these NUREG/CR-6260 locations, the refined analyses resulted in cumulative usage factors including environmental effects (Uenv) of less than the design code allowable (i.e., Uenv -- 1.0). A summary of how the calculations were performed is provided in item b of this response.

LRA Table 4.3-1 and Section 4.3.3.3.3 (including the associated statements in LRA Appendix A, Sections A.2.3.3.2 and A.3.3.3.3) are revised to address the results of the refined analyses. See the Enclosure to this letter for the revision to the BVPS LRA.

At two locations (Unit 1 pressurizer surge line to hot leg nozzle and Unit 2 pressurizer surge line to hot leg nozzle), Uenv exceeded the design code allowable limit of 1.0. The refined analyses including other actions to manage the environmental-assisted fatigue for the Unit 1 pressurizer surge line to hot leg nozzle and the Unit 2 pressurizer surge line to hot leg nozzle will be managed by the Metal Fatigue of Reactor Coolant Pressure Boundary Program (B.2.27). Previously, in FENOC Letter L-08-209, an enhancement was added to the Metal Fatigue of Reactor Coolant Pressure Boundary Program (B.2.27) and associated commitments

Attachment L-08-287 Page 4 of 9 provided in LRA Appendix A (Table A.4-1, Item Number 25 and Table A.5-1, Item Number 26 as follows:

"Add a requirement that fatigue will be managed for the NUREG/CR-6260 locations. This requirement will provide that management is accomplished by one or more of the following.

1. Further refinement of the fatigue analyses to lower the predicted CUFs to less than 1.0;

2. Management of fatigue at the affected locations by an inspection program that has been reviewed and approved by the NRC (e.g., periodic non-destructive examination of the affected locations at inspection intervals to be determined by a method acceptable to the NRC); or,

3. Repair or replacement of the affected locations."

In addition, the response to RAI-4.3-11 is revised in this letter to provide that a minimum of 50 cycles of OBE (5 events of 10 cycles each) are considered in each design analysis calculating CUF for NUREG/CR-6260 locations.

Therefore, the Regulatory Commitment for the refined analyses including the alternative analysis (fracture mechanics analysis) and use of a minimum of 50 cycles of OBE previously provided in Attachment 2 to FENOC Letter L-08-209 is withdrawn.

b. Please explain how the calculations for the fatigue life correction factor (Fen),

used to express the effects of the reactor coolant environment, will be performed. Specifically, how the transient pairs will be considered in the calculations.

Unit 1 Surge For the surge line hot leg nozzle, reactor water Line to Hot Leg environmental effects were evaluated by calculating Fen Nozzle: factors on fatigue usage using the general methodology in NUREG/CR-5704 (Reference 3) for stainless steel.

According to this method, fatigue usage is calculated with environmental fatigue correction factors on each load pair incremental usage. See Enclosure B, page 18 of FENOC Letter L-08-209 that provided the Westinghouse input to this RAI response.

Attachment L-08-287 Page 5 of 9 Unit 1 Charging The B31.1 analysis for the Unit 1 charging system was System Nozzle: modified to meet the requirements of ASME Ill, Class 1. The design transients for the corresponding Unit 2 piping are judged to be representative of the transients experienced by Unit 1. Design numbers for the CVCS transients were modified in accordance with operating experience at Unit 1.

An appropriate Fen was calculated in accordance with the guidance of NUREG/CR-5704 for stainless steel. The design CUF was multiplied by the calculated Fen.

Unit 2 Surge For the surge line hot leg nozzle, reactor water Line to Hot Leg environmental effects were evaluated by calculating Fen Nozzle: factors on fatigue usage using the general methodology in NUREG/CR-5704 for stainless steel. According to this method, fatigue usage is calculated with environmental fatigue correction factors on each load pair incremental usage. See Enclosure B, page 18 of FENOC Letter L-08-209 that provided the Westinghouse input to this RAI response.

Unit 2 Charging The analysis of record for the Unit 2 charging piping was System Nozzle: revised to incorporate new and revised thermal transients reflecting the operating experience at BVPS Unit 2. In addition, analytical conservatism was reduced to address the effects of environmentally assisted fatigue (EAF). All original design transients continue to be used without reduction for projected cycles. A design CUF was calculated. An appropriate Fen was calculated in accordance with the guidance of NUREG/CR-5704 for stainless steel. The design CUF was multiplied by the calculated Fen.

Unit 2 Safety A supplemental design analysis was performed for the SI Injection nozzle location as defined by NUREG/CR-6260. The System Nozzle: original design transients were used; however, the cycles for some transients were reduced to a bounding number. A design CUF was calculated. An appropriate Fen was calculated in accordance with the guidance of NUREG/CR-5704 for stainless steel. The design CUF was multiplied by the calculated Fen.

Attachment L-08-287 Page 6 of 9 Unit 2 Residual A supplemental design analysis was performed for the RHR Heat Removal system piping location as defined by NUREG/CR-6260. The System Piping: original design transients were used without reduction for projected cycles. A design CUF was calculated. An appropriate Fen was calculated in accordance with the guidance of NUREG/CR-5704 for stainless steel. The design CUF was multiplied by the calculated Fen.

c. Please describe the criteria and methodology that will be performed for the additional analyses in calculating the CUF, including environmental effects, for the components where the CUF exceeds the design code allowable value of 1.0.

Unit 1 Surge The surge line hot leg nozzle fatigue analyses were Line to Hot Leg performed according to the detailed methods of ASME Code Nozzle: Section III, NB-3200, as permitted by the NB-3600 piping design section. The NB-3200 evaluation was performed using program WESTEMS TM . See Enclosure 1, page 19 of FENOC Letter L-08-209 that provided the Westinghouse input to this RAI response. The method used to evaluate the effects of reactor water environment on the ASME fatigue usage is addressed in part b of this response.

Refined analysis is in progress as described in part a, above. While it is anticipated that the refined analysis will be successful, as an alternative a fracture mechanics analysis will be performed in accordance with the existing Metal Fatigue of Reactor Coolant Pressure Boundary Program.

Unit 1 Charging Using the method described in item b above, the reanalysis System Nozzle: resulted in a cumulative usage factors including environmental effects (Uenv) of less than the design code allowable (i.e., Uenv - 1.0). No additional analysis is required.

Attachment L-08-287 Page 7 of 9 Unit 2 Surge The surge line hot leg nozzle fatigue analyses were Line to Hot Leg performed according to the detailed methods of ASME Code Nozzle: Section III, NB-3200, as permitted by the NB-3600 piping design section. The NB-3200 evaluation was performed using program WESTEMS TM . See Enclosure 1, page 19 of FENOC Letter L-08-209 that provided the Westinghouse input to this RAI response. The method used to evaluate the effects of reactor water environment on the ASME fatigue usage is addressed in part b of this response.

Refined analysis is in progress as described in part a, above. It is anticipated that the refined analysis will be successful.

Unit 2 Charging Using the method described in item b above, the reanalysis System Nozzle: resulted in a cumulative usage factors including environmental effects (Uenv) of less than the design code allowable (i.e., Uenv - 1.0). No additional analysis is required.

Unit 2 Safety Using the method described in item b above, the reanalysis Injection resulted in a cumulative usage factors including System Nozzle: environmental effects (Uenv) of less than the design code allowable (i.e., Uenv - 1.0). No additional analysis is required.

Unit 2 Residual Using the method described in item b above, the reanalysis Heat Removal resulted in a cumulative usage factors including System Piping: environmental effects (Uenv) of less than the design code allowable (i.e., Uenv - 1.0). No additional analysis is required.

References:

1. NUREG/CR-6260, "Application of NUREG/CR-5999 Interim Fatigue Curves to Selected Nuclear Power Plant Components," February 1995

2. NUREG/CR-6934, "Fatigue Crack Flaw Tolerance in Nuclear Power Plant Piping

- A Basis for Improvements to ASME Code Section Xl Appendix L," May 2007

3. NUREG/CR-5704, "Effects of LWR Coolant Environments on Fatigue Design Curves of Austenitic Stainless Steels," March 1999

4. WESTEMS M T Integrated Diagnostics and Monitoring System See the Enclosure to this letter for the revision to the BVPS LRA.

Attachment L-08-287 Page 8 of 9 RAI-4.3-11 NRC Follow-up Question (Conference Call September 4, 2008):

Regarding operating basis earthquake (OBE) cycles, FENOC is using 50 OBE cycles (which is the current licensing basis), but also says that, if necessary, FENOC will use less than 50 cycles. Once you have an OBE you're going to have a certain number of cycles. You can't parse that. Do you need to reduce the OBE cycles? If not, please restate your response.

RESPONSE RAI-4.3-11 NRC Follow-up Question The Regulatory Commitment referenced in the original response to RAI 4.3-11 has been withdrawn. In order to remove the reference to that commitment and reply to the follow-up question above, the response to the original RAI 4.3-11 from Letter L-08-209 is replaced in its entirety with the following:

The 60-year projected operational cycles for operational basis earthquakes (OBE) is not provided in LRA Table 4.3-2. Please explain how many OBE occurrences or stress cycles will be included in the 60-year EAF.

A minimum of 50 cycles of OBE (5 events of 10 cycles each) is utilized in each design analysis calculating CUF for NUREG/CR-6260 locations.

Attachment L-08-287 Page 9 of 9 RAI-4.3-16 NRC Follow-up Question (Conference Call August 28, 2008):

The LRA discussed the heat-up and cooldown pressurizer transients. RAI 4.3-16 requested the associated histograms. No pressurizer cooldown histogram was provided. Why?

RESPONSE RAI-4.3-16 NRC Follow-up Question There is no independent transient that is tracked for pressurizer cooldown.

The LRA Section 4.3.4, 3rd paragraph, is revised in its entirety as follows:

Because plant performance has improved with time, the first option typically results in a more accurate projection, while the second option provides a more conservative number of thermal cycles. With the exception of the Unit 1 plant heatup and cooldown, and Unit 1 reactor trip transients, the extrapolation for all transients was completed using the second option. For the Unit 1 plant heatup and cooldown, the projected cycles were determined using the first option. For the Unit 1 reactor trip transient, the first option was also chosen, but then biased with additional reactor trips as the unit approaches end-of-life. Accrued operational cycles are based on initial operations for Unit 1 of 1975 and Unit 2 of 1986, and use a current plant life as of October 2003.

Therefore, the operating lifetimes used for the evaluations were 28 and 17 years for Unit 1 and Unit 2, respectively. The results of the transient cycle extrapolation are presented in Table 4.3-2.

See the Enclosure to this letter for the revision to the BVPS LRA.

ENCLOSURE Beaver Valley Power Station (BVPS), Unit Nos. I and 2 Letter L-08-287 Amendment No. 25 to the BVPS License Renewal Application Page 1 of 11 License Renewal Application Sections Affected Table A.4-1 Table A.5-1 Table 4.3-1 Section 4.3.3.3.3 Section A.2.3.3.2 Section A.3.3.3.3 Section 4.3.4 The Enclosure identifies the correction by Affected License Renewal Application (LRA) Section, LRA Page No., and Affected Paragraph and Sentence. The count for the affected paragraph, sentence, bullet, etc. starts at the beginning of the affected Section or at the top of the affected page, as appropriate. Below each section the reason for the change is identified, and the sentence affected is printed in italics with deleted text fined out and added text underlined.

Enclosure L-08-287 Page 2 of 11 Affected Paragraph Affected LRA Section LRA Page No. and Sentence Table A.4-1 Page A.4-9 New Item Number 31 LRA Table A.4-1, "Unit 1 License Renewal Commitments," requires a new license renewal future commitment to implement "needed actions" of MRP-146. New Item Number 31 is created, and LRA Table A.4-1 is revised to read as follows:

Table A.4-1, cont.

ImplementationRead Related LRAA Item ItmCommitment Ipeme Source Section No./

No. ComtetSchedule Cmet Comments 31 Implement "neededactions"of MRP-146. FENOC will FENOC None These actions include screening, detailed perform detailed Letter analysis, inspections and temperature evaluations L-08-287 monitoring in accordancewith the (analysis, guidelines of MRP-146. FENOC has inspections completed screening of the B VPS RCS and/or branch lines. monitoring)in accordancewith MRP-146 schedule requirements,or as established by the MRP committee.

Enclosure L-08-287 Page 3 of 11 Affected Paragraph Affected LRA Section LRA Page No. and Sentence Table A.5-1 Page A.5-10 New Item Number 32 LRA Table A.5-1, "Unit 2 License Renewal Commitments," requires a new license renewal future commitment to implement "needed actions" of MRP-146. New Item Number 32 is created, and LRA Table A.5-1 is revised to read as follows:

Table A.5-1, cont.

ImplementationRead Related LRA A

Item Ite Commitment Impeme Source Section No./

No. Schedule Cmet Comments 32 Implement "neededactions"of MRP-146. FENOC will FENOC None These actions include screening, detailed perform detailed Letter analysis, inspections and temperature evaluations L-08-287 monitoringin accordance with the (analysis, guidelines of MRP-146. FENOC has inspections completed screeningof the BVPS RCS and/or branch lines. monitoring)in accordancewith MRP-146 schedule requirements, or as established by the MRP committee.

Enclosure L-08-287 Page 4 of 11 Affected Paragraph Affected LRA Section LRA Page No. and Sentence Table 4.3-1 Page 4.3-12 & 13 Entire table As described in the amended response to RAI 4.3-03, Table 4.3-1 is revised to address the results of the refined analyses and to round the Uenv values to three decimal places. Table 4.3-1 is revised to read as follows:

Location Material Design CUF INUREG/GR Environmental (U 60 ) Multiplier CUF (Uenv)

UNIT I Reactor Vessel Shell 2.53 0:2568 and Lower Head. Low Alloy Steel 00102 0.026 Reactor Vessel Inlet Low Alloy Steel 0.0663 2.53 0467-9 Nozzle 0.168 Reactor Vessel Outlet Low Alloy Steel 0.0508 2.53 04286 Nozzle 0.129 Surge Line Hot Leg Stainless Steel Rosa00 I1201 Nozzle 0.4155 10.18 4.231 Charging System Nozzle Stainless Steel 0.127- 45.3 4,95 0.0998 2.86 0.285 Safety Injection System Stainless Steel 0.0121 15.35 0486i Nozzle 0.186 Residual Heat Removal Stainless Steel 0.0087 15.35 0.433 System Tee 0.134 UNIT 2 Reactor Vessel Shell 2.53 0.0044 and Lower Head Low Alloy Steel 0.0016 0.004 Reactor Vessel Inlet Low Alloy Steel 0.0891 2.53 0.2265 Nozzle 0.226 Reactor Vessel Outlet Low Alloy Steel 0.0601 2.53 0.62 Nozzle 0.152 Surge Line Hot Leg Stainless Steel 0,93 46536 442 Nozzle 0.4995 9.7 4.844 Charging System Nozzle Stainless Steel 0,6 15.35 4,.3 0.0301 0.462 Safety Injection System Stainless Steel 0044 463 0.22 Nozzle 0.3586 2.715 0.974 Residual Heat Removal Stainless Steel 1.030 i4.3 i5.8i System Piping 0.3889 2.55 0.992

a. Proicctcd 60 ycaFrcycles are exp~ected to exceed the design cycles by 50 percent. To acco)unt for thc mncrcased r,;;lcs. th dP1sian tal-aue

-sao

--Ft; ) was hnrese SO nU orc-nt.

Enclosure L-08-287 Page 5 of 11 Affected Paragraph Affected LRA Section LRA Page No. and Sentence Section 4.3.3.3.3 Page 4.3-13 & 14 Entire Section The following supersedes the FENOC letter L-08-209 LRA changes shown for Section 4.3.3.3.3 (Enclosure A, pages 10 and 11). Section 4.3.3.3 is replaced in its entirety to read as follows:

At two locations (Unit 1 pressurizersurge line to hot leg nozzle and Unit 2 pressurizersurge line to hot leg nozzle), Uenv exceeded the desiqn code allowable limit of 1.0. Forthese two locations, BVPS will implement one or more of the following as required by the Metal Fatiqueof Reactor Coolant PressureBoundary Program (Section B. 2.2 7):

1. Furtherrefinement of the fatique analyses to lower the predicted CUFs to less than 1.0:

2. Management of fatigue at the affected locations by an inspection program that has been reviewed and approved by the NRC (e.g.,

periodic non-destructive examination of the affected locations at inspection intervals to be determined by a method acceptable to the NRC); or,

3. Repair or replacementof the affected locations.

The Uenv at the otherNUREG/CR-6260 locations (Unit I reactor vessel shell and lower head, reactorvessel inlet and outlet nozzles, charging system nozzle, safety iniection nozzle and RHR system tee: Unit 2 reactorvessel shell and lower head, reactorvessel inlet and outlet nozzles, chargingsystem nozzle, safety iniection system nozzle and RHR system piping), have been demonstrated to remain less than the desiqn code allowable limit of 1.0 for the period of extended operation.

As discussed in Section 4.3.1. since 60-year projected operationalcycles were used in determining that the desiqn fatigue analyses remain valid for the period of extended operation, the Metal Fatique of Reactor Coolant Pressure Boundary Programmust continue to be used to validate the assumptions used in the evaluations. Therefore, the TLAAs associatedwith the NUREG/CR-6260 locations have been dispositionedin accordancewith 10 CFR 54.21(c)(1)(iii)

Enclosure L-08-287 Page 6 of 11 At several locatiens (Unit 1 preissurizer-surge line and c~harging systcm nozzleT Unit 2 pressurizer surge line, charging system nozzle, and RHR system piping),-i".excee---de d the de sign code ahie wabl/e !imit o f 1.0. FerF these-l i-otnS, BVPS wi, implement one orm*o r.e of the following as re.ui. d by the Metal Fatigue of Reactor Coolanpt PrFessure Boundary ProGgram (Section

1. Further refinement of the faiue analys-es to lower-the pr-edicted CU s to less than -.0ý,-

2. Management of fatigue at the affected loc-ations by an inspection program that has been reviewed and approved by the NRC (e.g.,

periodicG non destructive examination of the affected locations at insection intervals to be determined by a method acceptable toth

3. Repi orrpacement of the affected locations.-

program suc-h as scope, qua fification, mqethodý, and frequenc-y will be submitted to the NRC prior to the period of extended operation. Therefore, the TLAAS associated with the Unit I pressuie sug line and charging system nozzle,-

and the Unit 2 pressuie sure lnhagng syste nozzle, and RHR system piping have been disposfitine..d inm accordance with 10 CFR The CU#s, including environmental faigue at theother liMiting locations (Un 1 reactor vessel shel and lower head, reactor vessel Inlet and outVet nozzles, safety injecion nozzle and RHR system tee;-- Unit 2reaGto* vessel shel and lower head, reactor v.e-ssel inet and outlet nozzles-, and safet injection nozzle), have been demonstraqted to remgain less than the design cod-e alowable lmit of 1. for-the period of extended operation. Therefore, the TL=As associated with these other locations have been dispositioned in accordance with 10 CFR 54.21(cJ(1)(fi,).

Enclosure L-08-287 Page 7 of 11 Affected Paragraph Affected LRA Section LRA Page No. and Sentence Section A.2.3.3.2 Page A.2-10 Last 3 paragraphs of section The following supersedes the FENOC letter L-08-209 LRA changes shown for Section A.2.3.3.2 (Enclosure A, pages 22 and 23). The last three paragraphs of Section A.2.3.3.2 are replaced in their entirety to read as follows:

At the pressurizersurge line to hot leg nozzle, Uenv exceeded the desiqn code allowable limit of 1.0. For this location, BVPS will implement one or more of the following as required by the Metal Fatigue of Reactor Coolant Pressure Boundary Program:

1. Furtherrefinement of the fatigue analyses to lower the predicted CUFs to less than 1.0;

2. Management of fatigue at the affected locations by an inspection program that has been reviewed and approved by the NRC (e.g.,

periodic non-destructive examination of the affected locations at inspection intervals to be determined by a method acceptable to the NRC): or,

3. Repair or replacement of the affected locations.

The Uenv at the otherNUREG/CR-6260 locations (reactorvessel shell and lower head, reactorvessel inlet and outlet nozzles, charging system nozzle, safety injection nozzle and RHR system tee), have been demonstrated to remain less than the desiqn code allowable limit of 1.0 for the period of extended operation.

As discussed in Section A.2.3. 1, since 60-year proiectedoperationalcycles were used in determining that the design fatique analyses remain valid for the period of extended operation, the Metal Fatigue of Reactor Coolant Pressure Boundary Programmust continue to be used to validate the assumptions used in the evaluations. Therefore, the TLAAs associatedwith the NUREG/CR-6260 locations have been dispositionedin accordancewith 10 CFR 54.21(c)(1)(iii).

Enclosure L-08-287 Page 8 of 11 At two locations (prsuie surge line and charging system nozzeU

... eeded the design code al.owable imit of 1.0. For these locations, ,VPSw implement one or mer- of the following as required by the Metal Fatigue o.

Reactor Coolant Prcessure Boundary Proegram:.

1. Fughe refinement of the fatigue analys-es to lower the predicted CG!

to less than 1.0;-

2. Management of fatigue at the affeIted loVations by an i#nspetion progrm that has boon reviewed and approved by the NfRC (e.g.,

periodic non dlestructive examination of the affected locations at inspotion intefvals to be dete~mined by a method aG~eptable toth

3. Repi orrpacement of the affectedl locations. Should BVPS selectth optioto anage envirnmentally assisted fatigue during the period of extenpd-ed operation, details of the aging management programq, such a-s scope, qualiation, method, and requency, will be submitted to NRC prior to the period of eLended operation. Theredfoe, the pressurizer surge line and charging sysm nozzle TLAAs have be disposgitined in accordance with 10 CFR 54.2 1(e)(1)Rii.-

The cumulative usage factorS ineluding environmental faiue at the othe; nozzles-, safet injection nozzle and RH4R system tee) have been demonstrated to remain less than the design code allowable limit Of 1.0 for the period of extended operation. Therefore, the fatigue TLA~s associated wt those locations have been dispoisitioned in accordance with 10 CFR 54.21(G)1IW

Enclosure L-08-287 Page 9 of 11 Affected Paragraph Affected LRA Section LRA Page No. and Sentence Section A.3.3.3.3 Page A.3-13 Last 3 paragraphs of section The following supersedes the FENOC letter L-08-209 LRA changes shown for Section A.3.3.3.3 (Enclosure A, pages 31 and 32). The last three paragraphs of Section A.3.3.3.3 are replaced in their entirety to read as follows:

At the pressurizersurge line to hot leg nozzle, Uenv exceeded the design code allowable limit of 1.0. Forthis location, B VPS will implement one or more of the following as reguired by the Metal Fatique of Reactor Coolant Pressure Boundary Program:

1. Furtherrefinement of the fatique analyses to lower the predicted CUFs to less than 1.0.

2. Management of fatique at the affected locations by an inspection program that has been reviewed and approved by the NRC (e.g.,

periodic non-destructive examination of the affected locations at inspection intervals to be determined by a method acceptable to the NRC). or,

3. Repair or replacement of the affected locations.

The Uenv at the otherNUREG/CR-6260 locations (reactorvessel shell and lower head, reactorvessel inlet and outlet nozzles, chargingsystem nozzle, safety injection system nozzle and RHR system piping), have been demonstratedto remain less than the design code allowable limit of 1.0 for the period of extended operation.

As discussed in Section A.3.3. 1. since 60-year projected operationalcycles were used in determining that the design fatique analyses remain valid for the period of extended operation,the Metal Fatigue of ReactorCoolant Pressure Boundary Programmust continue to be used to validate the assumptions used in the evaluations. Therefore, the TLAAs associatedwith the NUREG/CR-6260 locations have been dispositionedin accordancewith 10 CFR 54.21(c)(1)(iii).

Enclosure L-08-287 Page 10 of 11 At thrce locations (p~re-ssuizer-surge line, charging system nozzle, and RR sstem piping), II " exc-eedmed the design codfe allwable limit of 1.0. F-or thesce locations, BPS ill im-plemeInt one or mrc,, of the following as required by the Metal Fatigue of Reactor-Coolanlt Pressure Boundary Program:.-

1. Further refinement of the fatigue analys-es to lower-the predicted CUFs to less than 1.0O;

2. Management of fAtigue at the affected Iocations by an inspection pregram that has been reviewed and approrved by the NRC (e.g.,

periodic non destructive examination of the affected locations at insection intewals to be determined by a method acceptable toth

3. Repair Or replacement of the affected locations.-

Should BVPS select the option to manage tally assisted tue during the period of extended operation, details of the agn aagement program, such as scope, qualification, method, and frqcc ilbe submitted to the NRC prior te the period of extended operation. Thereýfoe, th-e TLAAs associated with the pressurizer surge line, charging system nozzle, and RHR systm piping have been dispositioned in accordance With 10 CFR The cumulative usage factors including environmental fatigue at the other nozzles, and safet injection nozzle), have been demonstrated to remain less than the deSign code allowable limit of 1.0 fer the period of extended operation. Therefore, the TL4As associated with theise locations have been4 disepsitioned in accordance with 10 CFR 54.211(G) (W)i.

Enclosure L-08-287 Page 11 of 11 Affected Paragraph Affected LRA Section LRA Page No. and Sentence Section 4.3.4 Page 4.3-14 &15 Third paragraph Section 4.3.4 requires revision because no independent transient is tracked for pressurizer cooldown. The third paragraph is modified to read as follows:

Because plant performance has improved with time, the first option typically results in a more accurate projection, while the second option provides a more conservative number of thermal cycles. With the exception of the Unit I plant heatup and cooldown, pressurizrGcd"*own, and Unit I reactortrip transients, the extrapolationfor all transients was completed using the second option. For the Unit I plant heatup and cooldown and for p..ssu.izeF

,. .ldown, the projected cycles were determined using the first option. For the Unit I reactor trip transient,the first option was also chosen, but then biased with additional reactortrips as the unit approachesend-of-life. Accrued operationalcycles are based on initial operationsfor Unit I of 1975 and Unit 2 of 1986, and use a current plant life as of October 2003. Therefore, the operating lifetimes used for the evaluations were 28 and 17 years for Unit 1 and Unit 2, respectively.

The results of the transientcycle extrapolationare presented in Table 4.3-2.