L-12-225, Supplemental Reply to Request for Additional Information for the Review of License Renewal Application (TAC No. ME4640) and License Renewal Application Amendment No. 27

Supplemental Reply to Request for Additional Information for the Review of License Renewal Application (TAC No. ME4640) and License Renewal Application Amendment No. 27

ML12191A037
Person / Time
Site:	Davis Besse
Issue date:	07/05/2012
From:	Byrd K FirstEnergy Nuclear Operating Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
L-12-225, TAC ME4640
Download: ML12191A037 (10)

Text

FENOC S5501 Davis-Besse Nuclear Power Station N. State Route 2 FirstEnergyNuclear OperatingCompany Oak Harbor,Ohio 43449 July 5, 2012 L-1 2-225 10 CFR 54 ATTN: Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555-0001

SUBJECT:

Davis-Besse Nuclear Power Station, Unit No. 1 Docket No. 50-346, License Number NPF-3 Supplemental Reply to Request for Additional Information for the Review of the Davis-Besse Nuclear Power Station, Unit No. 1, License Renewal Application (TAC No. ME4640) and License Renewal Application Amendment No. 27 By letter dated August 27, 2010 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML102450565), FirstEnergy Nuclear Operating Company (FENOC) submitted an application pursuant to Title 10 of the Code of Federal Regulations, Part 54 for renewal of Operating License NPF-3 for the Davis-Besse Nuclear Power Station, Unit No. 1 (Davis-Besse). During a telephone conference with the NRC held on June 21, 2012, the NRC requested clarification regarding NRC request for additional information (RAI) 4.7.5.1-1.

The Attachment provides the FENOC supplemental reply to NRC RAI 4.7.5.1-1.

A discussion of the NRC request is shown in bold text followed by the FENOC response.

The Enclosure provides Amendment No. 27 to the Davis-Besse 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.

I declare under penalty of perjury that the foregoing is true and correct. Executed on July 5 ,2012.

Sincerely, Kendall W. Enrd Director, Site Engineering

Davis-Besse Nuclear Power Station, Unit No. 1 L-12-225 Page 2

Attachment:

Supplemental Reply to Request for Additional Information for the Review of the Davis-Besse Nuclear Power Station, Unit No. 1 (Davis-Besse), License Renewal Application, Section 4.7.5.1

Enclosure:

Amendment No. 27 to the Davis-Besse License Renewal Application cc: NRC DLR Project Manager (2 copies)

NRC Region III Administrator cc: w/o Attachment or Enclosure NRC DLR Director NRR DORL Project Manager NRC Resident Inspector Utility Radiological Safety Board

Attachment L-12-225 Supplemental Reply to Request for Additional Information for the Review of the Davis-Besse Nuclear Power Station, Unit No. 1 (Davis-Besse),

License Renewal Application, Section 4.7.5.1 Page 1 of 2 Section 4.7.5.1 Question RAI 4.7.5.1-1 Supplement The Nuclear Regulatory Commission (NRC) initiated a telephone conference call with FirstEnergy Nuclear Operating Company (FENOC) on June 21, 2012, to discuss time-limited aging analyses (TLAAs) associated with the Reactor Coolant System (RCS) Loop 1 cold leg drain line weld overlay as documented in Davis-Besse License Renewal Application (LRA) Section 4.7.5.1, "Reactor Coolant System Loop 1 Cold Leg Drain Line Weld Overlay Repair."

The NRC noted that in response to RAI 4.7.5.1-1, FENOC cited a summary calculation package that was prepared to document the design and analysis of the Davis-Besse reactor coolant pump 1-1 inlet cold leg drain line nozzle-to-elbow weld overlay. This summary calculation package was submitted by FENOC letter dated May 22, 2006 (ML061440282). In this package were summaries of an American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code)Section III evaluation (Calculation Number DB-06Q-304, Rev. 1) and a fatigue crack growth analysis (Calculation Number DB-06Q-307, Rev. 0). LRA Section 4.7.5.1 only identified the ASME Code Section III evaluation as a TLAA.

The NRC's position is that the fatigue crack growth analysis is also a TLAA requiring disposition for license renewal.

FENOC agreed to provide a supplemental response to RAI 4.7.5.1-1 to disposition the TLAA associated with the fatigue crack growth analysis.

RESPONSE RAI 4.7.5.1-1 SUPPLEMENT With respect to the potential for flaw growth, the reactor coolant pump 1-1 inlet cold leg drain line nozzle-to-elbow weld overlay is designed as a standard overlay (full structural) assuming a 360-degree flaw through the original pipe wall. As such, no credit is taken for any of the original pipe wall. The overlay material is Alloy 52, which is resistant to stress-corrosion cracking, and as such, flaw growth into the overlay by this mechanism is not expected. The presence of compressive residual stresses on the inside of the component after the overlay application also mitigates stress-corrosion cracking and minimizes fatigue crack growth into the overlay.

Attachment L-12-225 Page 2 of 2 A fatigue crack growth analysis [Structural Integrity Associates Calculation DB-06Q-307, Rev. 0, "Predicting Crack Growth for the DB Unit 1 RCP 1-1 Cold Leg Drain Nozzle With Design Weld Overlay," May 18, 2006] was performed to demonstrate that flaws equal to, or greater than, the maximum flaw sizes that could have escaped detection during the performance of the ultrasonic examinations would not grow unacceptably in the nozzle, so as to undermine the basis for the weld overlay. The dissimilar metal weld (DMW) contained an axial indication in the nozzle weld butter material (Alloy 182) for which no qualified depth sizing was performed. However, supplemental examinations confirmed that the indication was not present in the outer two-thirds of the wall thickness. Therefore, a flaw depth of one-third of the wall thickness was assumed for the axial and circumferential crack growth evaluation. Stress intensity factors (K) versus flaw depth were computed for three paths through the original DMW and butter, for both axial and circumferential cracks (six cases). For all six crack growth cases, no fatigue or PWSCC growth was predicted, as both Kmax and Kmin were negative for an assumed initial flaw size of one-third of the original base metal thickness.

Plant design cycles multiplied by a factor of 1.5 were used as an input to the structural weld overlay fatigue crack growth analysis. Therefore, the fatigue crack growth analysis is a time-limited aging analysis that requires disposition for license renewal. FENOC performed a comparison of the design cycles (original design cycles multiplied by a factor of 1.5) that were used in the fatigue crack growth analysis to the 60-year projected cycles provided in LRA Table 4.3-1, "60-Year Projected Cycles," and determined that the analyzed cycles bound the 60-year projected cycles. Therefore, the fatigue crack growth analysis associated with the RCS Loop 1 cold leg drain structural weld overlay remains valid for the period of extended operation in accordance with 10 CFR 54.21(c)(1)(i).

LRA Table 4.1-1, "Time-Limited Aging Analyses," and Sections 4.7.5.1 and A.2.6.1, both titled "Reactor Coolant System Loop 1 Cold Leg Drain Line Weld Overlay Repair,"

are revised consistent with this response.

See the Enclosure to this letter for the revision to the Davis-Besse LRA.

Enclosure Davis-Besse Nuclear Power Station, Unit No. 1 (Davis-Besse)

Letter L-12-225 Amendment No. 27 to the Davis-Besse License Renewal Application Page 1 of 6 License Renewal Application Sections Affected Table 4.1-1 Section 4.7.5.1 Section A.2.6.1 The Enclosure identifies the change to the License Renewal Application (LRA) by Affected 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 linedeut and added text underlined.

Enclosure L-12-225 Page 2 Affected LRA Section LRA Page No. Affected Paragraph and Sentence Table 4.1-1 Page 4.1-4 "RCS Loop 1 Cold Leg drain line weld overlay repair" row, "54.21(c)(1) Paragraph" column In the supplemental response to request for additional information (RAI) 4.7.5.1-1, the "RCS Loop 1 Cold Leg drain line weld overlay repair' row, "54.21(c)(1) Paragraph" column of LRA Table 4.1-1, "Time-Limited Aging Analyses," is revised as follows:

Table 4.1-1 Time-Limited Aging Analyses Results of TLAA Evaluation by Category 54.21P(r)(1) LRA Paragraph

( Section Other Plant-Specific Time-Limited Aging Analyses 4.7 RCS Loop 1 Cold Leg drain line weld overlay repair (i) and (Ni) 4.7.5.1

Enclosure L-1 2-225 Page 3 Affected LRA Section LRA Page No. Affected Paragraph and Sentence 4.7.5.1 Pages 4.7-5 Paragraph 2, and new Disposition and 4.7-6 In the supplemental response to RAI 4.7.5.1-1, LRA Section 4.7.5.1, "Reactor Coolant System Loop 1 Cold Leg Drain Line Weld Overlay Repair," is revised to read as follows:

4.7.5.1 Reactor Coolant System Loop 1 Cold Leg Drain Line Weld Overlay Repair FENOC performed a full structural overlay repair for an axial indication found on the Reactor Coolant System Loop 1 cold leg drain line during the Cycle 14 refueling outage. The structural weld overlay of the cold leg drain nozzle was designed consistent with the requirements of ASME Section XI; Code Case N-504-2; non-mandatory Appendix Q; and was supplemented by additional design considerations specific to the unique nature of the geometry and materials of the cold leg drain nozzle-to-elbow weld.

Fatigue Crack Growth Analysis The overl'ay is desiged as a fustll ict.-Ir,,, l verlay that assume& the as fou,,

"Wa P;upayatesto a 4 burl tHwUgyI ioa;f 69 Gutw G~G Fat"601; F~aP @IFt a crack groqwth analysis of the as found flaw. Thus there is no time dependenc /

in the weid over ay de 'gn With respect to the potential for flaw growth, the reactor coolant pump 1-1 inlet cold leq drain line nozzle-to-elbow weld overlay is designed as a standard overlay (full structural)assuming a 360-degree flaw through the originalpipe wall.

As such, no credit is taken for any of the oriqinalpipe wall. The overlay material is Alloy 52, which is resistant to stress-corrosion cracking, and as such, flaw growth into the overlay by this mechanism is not expected. The presence of compressive residual stresses on the inside of the component after the overlay application also mitigates stress-corrosioncracking and minimizes fatigue crack Qrowth into the overlay.

A fatigue crack growth analysis was performed to demonstrate that flaws equal to, or greater than, the maximum flaw sizes that could have escaped detection durinq the nerformance of the ultrasonic examinations would not grow unacceptably in the nozzle, so as to undermine the basis for the weld overlay.

The dissimilarmetal weld (DMW) contained an axial indication in the nozzle weld butter material (Alloy 182) for which no gualified denth sizinq was oerformed.

Enclosure L-1 2-225 Page 4 However, supplemental examinations confirmed that the indication was not present in the outer two-thirds of the wall thickness. Therefore, a flaw depth of one-third of the wall thickness was assumed for the axial and circumferential crack growth evaluation. Stress intensity factors (K) versus flaw depth were computed for three paths through the oriqinalDMW and butter, for both axial and circumferential cracks (six cases). For all six crack growth cases, no fatique or PWSCC growth was predicted, as both Kmax and Kmin were negative for an assumed initial flaw size of one-third of the originalbase metal thickness.

Plant desiqn cycles multiplied by a factor of 1.5 were used as an input to the structuralweld overlay fatique crack growth analysis. Therefore, the fatique crack growth analysis is a time-limited aging analysis that requires disposition for license renewal. FENOC performed a comparison of the design cycles (original design cycles multiplied by a factor of 1.5) that were used in the fatique crack growth analysis to the 60-year proiected cycles provided in LRA Table 4.3-1 and determined that the analyzed cycles bound the 60-year projected cycles.

Therefore, the fatigue crack growth analysis associated with the RCS Loop 1 cold leq drain structural weld overlay remains valid for the period of extended operation.

Disposition: 10 CFR 54.21(c)(1)(i) The fatique crack growth analysis associatedwith the RCS Loop I cold leg drain structural weld overlay remains valid for the period of extended operation.

Fatique Analysis The fatigue analysis for the repaired configuration conservatively estimated cycles for 60 years at 1.5 times the original design cycles. Because this analysis is based on a specific number of cycles, it is a TLAA. The Fatigue Monitoring Program manages the effects of fatigue on the reactor coolant system drain line weld overlay repair by counting the thermal cycles incurred through the period of extended operation.

Disposition: 10CFR54.21(c)(1)(iii) The effects of fatigue on the reactor coolant system cold leg drain line nozzle weld overlay repair will be managed for the period of extended operation by the Fatigue Monitoring Program.

Enclosure L-1 2-225 Page 5 Affected LRA Section LRA Page No. Affected Paragraph and Sentence A.2.6.1 Page A-46 Paragraph 2 In the supplemental response to RAI 4.7.5.1-1, LRA Section A.2.6.1, "Reactor Coolant System Loop 1 Cold Leg Drain Line Weld Overlay Repair," is revised to read as follows:

A.2.6.1 Reactor Coolant System Loop 1 Cold Leg Drain Line Weld Overlay Repair A full structural overlay repair was performed for an axial indication found on the Reactor Coolant System Loop 1 cold leg drain line during the Cycle 14 refueling outage. The structural weld overlay of the cold leg drain nozzle was designed consistent with the requirements of ASME Section XI; Code Case N-504-2; Non-mandatory Appendix Q; and was supplemented by additional design considerations specific to the cold leg drain nozzle-to-elbow weld.

Fatique Crack Growth Analysis The overlay is designed as a full Str,*tUrla verlay that assumes the as foun, flaw propagateis to a 100% through wall 360 degroe Crack rather-than performin a crack growth analysis of the as found flaw. Thus there is nig time dependenc in the weld overlay deskign.-

With respect to the potential for flaw growth, the reactor coolant pump 1-1 inlet cold leq drain line nozzle-to-elbow weld overlay is designed as a standard overlay (full structural)assuming a 360-degree flaw through the originalpipe wall.

As such, no credit is taken for any of the original pipe wall. The overlay material is Alloy 52, which is resistant to stress-corrosion cracking, and as such, flaw growth into the overlay by this mechanism is not expected. The presence of compressive residual stresses on the inside of the component after the overlay application also mitigates stress-corrosioncracking and minimizes fatigue crack growth into the overlay.

A fatigue crack growth analysis was performed to demonstrate that flaws equal to, or greater than, the maximum flaw sizes that could have escaped detection during the performance of the ultrasonic examinations would not grow unacceptably in the nozzle, so as to undermine the basis for the weld overlay.

The dissimilarmetal weld (DMW) contained an axial indication in the nozzle weld butter material (Alloy 182) for which no qualified depth sizinq was performed.

However, supplemental examinations confirmed that the indication was not present in the outer two-thirds of the wall thickness. Therefore, a flaw depth of

Enclosure L-1 2-225 Page 6 one-third of the wall thickness was assumed for the axial and circumferential crack growth evaluation. Stress intensity factors (K) versus flaw depth were computed for three paths through the originalDMW and butter, for both axial and circumferential cracks (six cases). For all six crack growth cases, no fatique or PWSCC growth was predicted, as both Krax and Krn were negative for an assumed initial flaw size of one-third of the oriqinalbase metal thickness.

Plant desiqn cycles multiplied by a factor of 1.5 were used as an input to the structuralweld overlay fatique crack growth analysis. Therefore, the fatique crack growth analysis is a time-limited aging analysis that requires disposition for license renewal. FENOC performed a comparison of the design cycles (original design cycles multiplied by a factor of 1.5) that were used in the fatique crack growth analysis to the 60-year projected cycles provided in LRA Table 4.3-1 and determined that the analyzed cycles bound the 60-year projected cycles.

Therefore, the fatigue crack growth analysis associated with the RCS Loop 1 cold leq drain structural weld overlay remains valid for the period of extended operationin accordance with 10 CFR 54.21(c)(1)(i).

FatiqueAnalysis The fatigue analysis estimated cycles for 60 years based on the original design cycles. Because this analysis is based on a specific number of cycles, it is considered a TLAA. All cumulative usage factors for the reactor coolant pump drain line weld overlay are less than 1.0.

The effects of fatigue on the reactor coolant pump drain line weld overlay repair will be managed by the Fatigue Monitoring Program for the period of extended operation in accordance with 10 CFR 54.21(c)(1 )(iii).