L-11-334, Reply to Request Additional Information for the Review of the License Renewal Application Amendment No. 21

Reply to Request Additional Information for the Review of the License Renewal Application Amendment No. 21

ML11306A066
Person / Time
Site:	Davis Besse
Issue date:	10/31/2011
From:	Allen B FirstEnergy Nuclear Operating Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
L-11-334, TAC ME4640
Download: ML11306A066 (31)

Text

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FENOC Withhold From Public Disclosure Under 10 CFR 2.390 5501 North State Route 2 FirstEnergyNuclear OperatingCompany Oak Harbor,Ohio 43449 Barry S. Allen 419-321-7676 Vice President - Nuclear Fax: 419-321-7582 October 31, 2011 L-1 1-334 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 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. 21 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 (DBNPS). By letter dated October 11, 2011 (ADAMS Accession No. ML11271A147), the Nuclear Regulatory Commission (NRC) requested additional information to complete its review of the License Renewal Application (LRA).

The Attachment provides the FENOC responses to three of the four NRC requests for additional information (RAIs) in the letter dated October 11, 2011. Specifically, the Attachment includes responses for RAIs B.2.40-3, 4.7.3-1 and 4.7.4-1. The response to RAI 4.7.5.2-2 is still under review and is planned to be provided to the NRC within the requested 30-day response time. The NRC request is shown in bold text in the Attachment followed by the FENOC response. Enclosure A provides Amendment No. 21 to the DBNPS LRA.

In response to RAI 4.7.3-1, Enclosure C provides AREVA NP Inc. (AREVA NP)

Calculation No. 32-9124893-001, "DB-1 Pressurized Thermal Shock (PTS) Analysis for 32 and 52 EFPY." The calculation documents the analysis of reactor vessel integrity Withhold From Public Disclosure Under 10 CFR 2.390 When separated from Enclosure C, this document is decontrolled.

(UL

Withhold From Public Disclosure Under 10 CFR 2.390 Davis-Besse Nuclear Power Station, Unit No. 1 L-1 1-334 Page 2 under specific pressurized thermal shock conditions. The AREVA NP reactor vessel integrity analysis contains proprietary information that is to be withheld from public disclosure pursuant to 10 CFR 2.390. A nonproprietary version of the calculation does not exist. Enclosure B provides the AREVA NP affidavit to support the disclosure request for the proprietary reactor vessel integrity analysis.

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 October 3(1_, 2011.

Sincerely, Barry S. Aen

Attachment:

Reply to Request for Additional Information for the Review of the Davis-Besse Nuclear Power Station, Unit No. 1 (DBNPS), License Renewal Application, Sections B.2.40, 4.7.3 and 4.7.4

Enclosures:

A. Amendment No. 21 to the DBNPS License Renewal Application B. Affidavit for Calculation No. 32-9124893-001, "DB-1 Pressurized Thermal Shock (PTS) Analysis for 32 and 52 EFPY" C. Calculation No. 32-9124893-001, "DB-1 Pressurized Thermal Shock (PTS)

Analysis for 32 and 52 EFPY" (Proprietary) cc: NRC DLR Project Manager (Attachment and Enclosure A)

NRC Region III Administrator (Attachment and Enclosure A)

Withhold From Public Disclosure Under 10 CFR 2.390 When separated from Enclosure C, this document is decontrolled.

Withhold From Public Disclosure Under 10 CFR 2.390 Davis-Besse Nuclear Power Station, Unit No. 1 L-11-334 Page 3 cc: w/o Attachment or Enclosures NRC DLR Director NRR DORL Project Manager NRC Resident Inspector Utility Radiological Safety Board Withhold From Public Disclosure Under 10 CFR 2.390 When separated from Enclosure C, this document is decontrolled.

Attachment L-1 1-334 Reply to Request for Additional Information for the Review of the Davis-Besse Nuclear Power Station, Unit No. 1 (DBNPS), License Renewal Application, Sections B.2.40, 4.7.3 and 4.7.4 Page 1 of 8 Section B.2.40 Question RAI B.2.40-3 Follow-Up to RAI B.2.40-2

Background:

By letter dated August 17, 2011, the applicant responded to a staff RAI regarding operating experience with degradation of the north embankment of the safety-related portion of the intake canal. In the response the applicant committed to ensure that an investigation of the embankment degradation would be completed prior to the period of extended operation. The applicant further committed to evaluate the results and complete needed repairs or modifications of the embankment prior to the period of extended operation.

Issue:

Although the applicant committed to completing long-term evaluation plans, no information was provided about the plan, such as schedule, scope, or acceptance criteria.

Request:

Provide details about the embankment investigation. The response should include scheduling information, activities planned and completed to date, and probable corrective actions. The response should provide technical justification for the timeliness of the repairs, including an explanation why prior to the period of extended operation is an acceptable deadline for completing the repairs.

RESPONSE RAI B.2.40-3 Follow-Up to RAI B.2.40-2 Long-term plans have been developed to complete the investigation and needed repairs or modification of the degraded portion of the safety-related Intake Canal embankment.

1. Activities completed

a. Performed preventive maintenance for the Forebay and Intake Canal Dikes.

The preventive maintenance included a detailed survey and underwater inspection of the dikes. The area inspected and surveyed included the portion

Attachment L-1 1-334 Page 2 of 8 of the Intake Canal that serves as the ultimate heat sink. The data from the survey and inspection were evaluated by a geotechnical engineering consultant. The degraded area of the intake canal embankment was included in the survey, inspection and evaluation. The geotechnical consultant provided FENOC with a written report on the evaluation, which indicated that the as-found condition of the canal is less conservative than the as-analyzed condition. The dike slopes are non-conservative, and the volume and surface area of the ultimate heat sink are less than the current design values. FENOC entered the conditions described into the Corrective Action Program for evaluation and resolution. FENOC performed a Prompt Operability Determination for the degraded Intake Canal. The Prompt Operability Determination showed that there was sufficient conservatism in the canal design to ensure that the ultimate heat sink could continue to perform its design functions.

Corrective actions for the identified ultimate heat sink conditions include resolution of the discrepancies between the as-found and as-analyzed ultimate heat sink. The resolution may include calculation alterations or repairs to the ultimate heat sink. There is also a corrective action to close out the Prompt Operability Determination upon completion of the resolution of the discrepancies. The current due date for close-out of the Prompt Operability Determination is December 18, 2011.

b. Performed four core bores into the degraded area of the dike for evaluation of the existing condition of the degraded area.

c. Installed two slope inclinometers (slope stability monitoring devices), in the degraded area of the dike to verify the long-term stability of the dike.

2. Activities Planned and Scheduling Information - The actions and dates described below are based on currently available information and anticipated time durations.

a. Geotechnical engineering consultant submittal of the core bore evaluation report is due by November 11,2011.

b. Monitor slope inclinometer devices until April 30, 2012, to determine if there is any ongoing movement in the affected areas of the dike.

c. Geotechnical engineering consultant to submit a report detailing results of the canal dike studies and the recommended remedial actions by May 30, 2012.

d. FENOC and third party review of the geotechnical report and proposed remedial actions (including any resolution of comments) is planned to be completed by July 30, 2012.

Attachment L-1 1-334 Page 3 of 8

e. FENOC management approval of the project design and implementation for the remediation of the dike is planned to be completed by August 30, 2012.

f. The engineering change package for the remediation work is planned to be issued by November 31, 2012.

g. Completion of implementation of the remediation work is expected by December 30, 2014.

3. The probable remediation method would likely include excavation of the affected areas, replacement of material in affected areas of the dike with designed fill material, and completion of engineering documentation supporting the acceptability of the resulting factor of safety for the canal dike.

4. Technical Justification for the timeliness of the corrective actions The technical justification for the timeliness of the corrective actions is based on the Prompt Operability Determination. The Prompt Operability Determination included engineering analyses that showed that reduction in ultimate heat sink surface area and volume is acceptable because incorporation of existing volume and surface area margin showed that the acceptance criteria with respect to maximum fluid temperature, service water pump net positive suction head and service water pump submergence are satisfied. Also, the Prompt Operability Determination included a review of the seismic stability of the intake canal dikes.

The Prompt Operability Determination showed that there is reasonable assurance that the intake canal dikes, in their current as-surveyed condition, remain capable of performing their design function of bounding water to serve as the ultimate heat sink during an accident (seismic event). No compensatory actions are required.

Also, the integrity of the degraded area of the embankment is currently being ensured by the evaluations and corrective actions initiated earlier in accordance with the FENOC Corrective Action Program. The integrity of the embankment was initially addressed in 2007. A slope stability study was conducted by an independent geotechnical engineering consultant as a corrective action. The engineering evaluation of the slope stability study concluded that the affected area is localized to the areas identified and that the current condition is insignificant to the original design of the intake canal.

In 2011, the geotechnical engineering consultant again evaluated the degraded area of the embankment and concluded that the slope movement has stabilized.

The preventive maintenance described in Item 1.a above is scheduled to be performed on a three-year interval.

Attachment L-1 1-334 Page 4 of 8 Therefore, based on the Prompt Operability Determination, the actions performed, and the planned and ongoing activities, the proposed schedule for implementing the planned remedial actions is timely while allowing sufficient time to complete the complex corrective actions described in Item 2 above.

For license renewal, "prior to the period of extended operation" is an acceptable deadline for repairs because completion of remediation work by that deadline ensures that, at the beginning of the period of extended operation, the condition of the dike will be aligned with the current licensing basis, without the previously identified degradation.

Section 4.7.3 Question RAI 4.7.3-1

Background:

License renewal application (LRA) Section 4.7.3 discusses a fracture mechanics analysis for evaluating the integrity of the reactor vessel (RV) during the pressurized thermal shock (PTS) event associated with low-temperature (35 0 F) water injection from the borated water storage tank (BWST) following a small steam line break. LRA Section 4.7.3 states that the current licensing basis (CLB) analysis for this event is addressed in the Davis-Besse Updated Safety Analysis Report (USAR), Section 5.2 and that the subject analysis was revised to consider the period of extended operation (52 EFPY).

Issue:

The staff reviewed USAR Section 5.2 and could not locate the CLB analysis for evaluating RV integrity under the subject PTS conditions. Furthermore, the staff found no references in LRA Section 4.8 for reports documenting the analysis of RV integrity under these PTS conditions for the period of extended operation, based on the 52 EFPY reference temperature for PTS (RTPTS) values.

Request:

1. State the USAR section and page number where the summary of the CLB analysis of the subject PTS event is located. If a summary of the CLB analysis is not located in the USAR, please state where it can be found.

Attachment L-1 1-334 Page 5 of 8

2. Provide the reports documenting the projected 52 EFPY analysis of RV integrity under the subject PTS conditions.

RESPONSE RAI 4.7.3-1

1. The analysis to demonstrate that the reactor vessel can safely accommodate the rapid temperature change associated with the postulated operation of the Emergency Core Cooling System (ECCS) at the end of the reactor vessel's design life is addressed in Section 5.2 of the Davis-Besse Updated Safety Analysis Report (USAR), pages 5.2-2 and 5.2-3.

An initial pressurized thermal shock (PTS) analysis of the reactor vessel utilized a non-conservative temperature for postulated operation of the ECCS. The Technical Specifications allowable minimum temperature for the Borated Water Storage Tank (BWST) is 35 0F. The PTS scenario was reanalyzed using 35 0F BWST water temperature and reactor vessel material properties that equate to 52 effective full power years (EFPY). The results show that the reactor vessel will retain its structural integrity during the analyzed event beyond the analyzed lifetime of 52 EFPY.

USAR Section 5.2 was revised May 26, 2011, under an approved USAR change notice, to include the following information:

USAR Section 5.2, Item #4 on page 5.2-2, is replaced with the following:

The reactor vessel was analyzed to demonstrate that it can safely accommodate the rapid temperature change associated with the postulated operation of the Emergency Core Cooling System (ECCS) at the end of the vessel's design life. Reference (XX) evaluates the reactor vessel integrity during a small steam line break, which creates a pressurized thermal shock condition. This transient generates the greatest level of stress in the reactor vessel. The analysis assumes that the water initially injected from the Borated Water Storage Tank (BWST) is at 35 0 F. The results show that the integrity of the reactor vessel is not violated.

The reactor vessel's compliance with 10CFR50.61, "Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events," has also been assessed in accordance with the methods specified in that regulation. The Pressure and Temperature Limits Report (PTLR) reports the required RTPTS values for the reactor vessel beltline welds and forgings. The vessel meets the 1 OCFR50.61 screening for the most limiting weld and forging

Attachment L-1 1-334 Page 6 of 8 material at a fluence equal to 52 Effective Full Power Years of Operation (EFPY).

USAR Section 5.2, page 5.2 the following new paragraph is inserted between the "Brittle Fracture" and "Fatigue" paragraphs:

Reference (XX) documents the thermal shock analysis analyzed to ensure the reactor vessel's integrity will be maintained throughout the life of the plant. The analysis was performed with a vessel fluence equal to 52 Effective Full Power Years (EFPY) of operation. The results demonstrate that the vessel will not fail. The vessel's compliance with 10CFR50.61 is reported in the Pressure and Temperature Limits Report (PTLR). The results show that the vessel is well below the applicable 10CFR50.61 screening criteria for the most limiting forging and weld material at 52 EFPY.

USAR Section 5.2, page 5.2 the first full paragraph on the page, revised by Revision 27 to the USAR, is deleted as follows:

The analysis for pressurized, thermal hock GaRn be found in ARBVA ANP document 5116810800 (Rof.25). Thiso docu...m..rmt p.roides a qualitative assessment that the RCS and tho RoactorF Vessel will maintain integrity when 35 0F water i6 injected. This docum.nt provides analYsis for all components; of the RCS and deteFMrmie ta the rcFa.*Or

.o... ^

is tho limniting component. A .. se..atiyo e.timate limits operation to, 25 F=P=YD.The document also indic;ates that the 25 EFPY is overly conereative, and the amount of oper-able EFPY could beinOreased with further study,,.

USAR Section 5.7 includes the following new reference:

Reference (XX) - AREVA NP Calculation 32-9124893-001, "DB-1 Pressurized Thermal Shock (PTS) Analysis for 32 and 52 EFPY,"

12/14/09.

The analysis information provided above will be incorporated into the USAR in the next update in accordance with 10 CFR 50.71(e). The USAR change documentation is available for onsite review.

2. The fracture mechanics analysis for evaluating the integrity of the reactor vessel during the PTS event associated with low-temperature (35°F) water injection from the BWST following a small steam line break is documented in AREVA NP Calculation 32-9124893-001, "DB-1 Pressurized Thermal Shock (PTS) Analysis for 32 and 52 EFPY," dated December 14, 2009.

Attachment L-1 1-334 Page 7 of 8 See Enclosure C to this letter for a copy of AREVA NP Calculation 32-9124893-001. This calculation document contains information that is classified by AREVA NP as proprietary, and requested to be withheld from public disclosure pursuant to 10 CFR 2.390. See Enclosure B to this letter for an AREVA NP Affidavit requesting the withholding of the proprietary information in Enclosure C from pubic disclosure in accordance with 10 CFR 2.390.

Section 4.7.4 Question RAI 4.7.4-1 (NRC Letter dated October 11, 2011)

Background:

By letter dated June 3, 2011, the applicant provided Amendment 8 to Davis-Besse LRA. LRA Amendment 8 revised the disposition for the analysis of the HPI/Makeup Nozzle Thermal Sleeves in LRA Section 4.7.4 from "10 CFR 54.21 (c)(1)(iii)" to "Not a TLAA." As an explanation for the revised disposition, LRA Section 4.7.4, as amended, now states that "[b]ased on the

[USAR Supplement] commitment [to replace the subject thermals sleeves], the HPI/Makeup nozzle thermal sleeves are short-lived (not 40-year) parts and therefore this analysis is not a TLAA." Similarly, LRA Amendment 8 revised the corresponding USAR Supplement section in LRA Section A.2.7.4 to reflect the changed disposition. LRA Section A.2.7.4, as revised by LRA Amendment 8, now states that, "[b]ased on the commitment [to replace the subject thermal sleeves],

the HPI/makeup nozzle thermal sleeves are short lived (not 40-year) parts and therefore this analysis is not a TLAA." Finally, LRA Section 4.1, Table 4.1-1, was amended per LRA Amendment 8 to state that the evaluation of the subject thermal sleeves is "Not a TLAA."

Issue:

The staff determined that aging of the subject thermal sleeves, as discussed in LRA Section 4.7.4. should be identified as a time-limited aging analysis (TLAA) in LRA Sections 4.1, 4.7.4, and the USAR Supplement, because the aging mechanism is time dependent (i.e., it is dependent on the number of transient cycles incurred), and the staff cannot accept future commitments to replace components as a means for disposition of the currently-installed components undergoing time-dependent aging processes, without a TLAA of the currently installed components.

Attachment L-1 1-334 Page 8 of 8 Request:

Based on the above, the staff requests that the applicant amend LRA Sections 4.1, 4.7.4, and A.2.7.4 to identify HPI/makeup thermal sleeve aging as a TLAA. The staff also requests that the applicant select an appropriate disposition under Title 10 of the Code of Federal Regulations (10 CFR 54.21(c)(1)). If the applicant proposes a 10 CFR 54.21(c)(1)(iii) disposition for this analysis, then the staff requests that the applicant amend LRA Sections 4.7.4 and A.2.7.4 to propose an appropriate aging management program (AMP) for managing the effects of aging on the intended function of the thermal sleeves. Any AMP identified in LRA Sections 4.7.4 and A.2.7.4 for a 10 CFR 54.21(c)(1)(iii) disposition of this analysis should ensure that the effects of aging on the subject thermal sleeves are appropriately managed for the period of extended operation.

RESPONSE RAI 4.7.4-1 (NRC Letter dated October 11, 2011)

LRA Sections 4.1 (Table 4.1-1), 4.7.4 and A.2.7.4 are revised to identify HPI/makeup nozzle thermal sleeve aging as a TLAA; the effects of cracking on the makeup nozzle thermal sleeve will be managed by the Inservice Inspection Program through the period of extended operation in accordance with 10 CFR 54.21 (c)(1)(iii). In addition, LRA Table A-I, "Davis-Besse License Renewal Commitments," is revised to delete HPI nozzle thermal sleeves from license renewal future Commitment 23. Also, LRA Section B.2.24, "Inservice Inspection Program," is revised to include the augmented examination of the HPI/makeup nozzle thermal sleeve.

See Enclosure A to this letter for the revision to the DBNPS LRA.

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

Letter L-1 1-334 Amendment No. 21 to the DBNPS License Renewal Application Page 1 of 16 License Renewal Application Sections Affected Table 3.3.2-31 Table 3.3.2 Plant-Specific Notes Table 3.4.2-1 Table 3.4.2-3 Table 3.4.2-4 Table 3.4.2 Plant-Specific Notes Table 4.1-1 Section 4.7.4 Section A.2.7.4 Table A-1 Section B.2.24 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 italicswith deleted text lined-ut and added text underlined.

Enclosure A L-1 1-334 Page 2 of 16 Affected LRA Section LRA Page No. Affected ParaqraDh and Sentence Table 3.3.2-31 Pages 3.3-533 Rows 20, 25, 31, 37, 44, 52, 57, 61, 65 and 70 through 540 Errata: During RAI response development, FENOC identified that LRA Table 3.3.2 Plant-Specific Note 0334, previously introduced in FENOC RAI response letter dated April 29, 2011 (ML11126A016) had been inadvertently duplicated in FENOC letter dated May 24, 2011 (ML11151A090) and applied to rows 20, 25, 31, 37, 44, 52, 57, 61, 65 and 70 of LRA Table 3.3.2-31, "Aging Management Review Results - Station Plumbing, Drains, and Sumps System," in that same letter (May 24, 2011). Plant-Specific Note 0334 is renumbered to 0342 (see "Table 3.3.2 Plant-Specific Notes" revision, below), and the 10 rows from LRA Table 3.3.2-31 are revised to include the new Note number, as follows:

Table 3.3.2-31 Aging Management Review Results - Station Plumbing, Drains, and Sumps System Aging Effect NUREG-Row Component Intended Material Environment Requiring Aging Management 1801, Table I Notes No. Type Function(s) Management Program Volume 2 Item Item Inspection of Internal E 20 Piping Pressure Stainless Moist air Loss of Surfaces in V.11-29 3.2.1-08 0332 boundary Steel (Internal) material Miscellaneous Piping 0334 and Ducting 0342 Inspection of Internal E 25 Piping Pressure Moist air Loss of Surfaces in VIIG-23 3.3.1-71 0334 boundary Steel (Internal) material Miscellaneous Piping 0342 and Ducting Inspection of Internal E 31 Piping Structural Gray Cast Moist air Loss of Surfaces in VILG-23 3.3.1-71 0332 integrity Iron (Internal) material Miscellaneous Piping 0334 and Ducting 0342

Enclosure A L-1 1-334 Page 3 of 16 Table 3.3.2-31 Aging Management Review Results - Station Plumbing, Drains, and Sumps System Aging Effect NUREG-Row Component Intended Material Environment Requiring Aging Management 1801, Table 1 Notes No. Type Function(s) Management Program Volume 2 Item Item Inspection of Internal E Structural Stainless Moist air Loss of Surfaces in VD129 3.2.1-08 0332 37 Piping integrity Steel (Internal) material Miscellaneous Piping 08 34 and Ducting 0342 Inspection of Internal E 44 Piping Structural Steel Moist air Loss of Surfaces in VII.G-23 3.3.1-71 034 integrity (Internal) material Miscellaneous Piping 0342 and Ducting Inspection of Internal E Structural Stainless Moist air Loss of Surfaces in 0332 52 Tubing integrity Steel (Internal) material Miscellaneous Piping V.D1-29 3.2.1-08 and Ducting 0342 Inspection of Internal E Pressure Stainless Moist air Loss of Surfaces in V.D1-29 32108 0332 57 Valve Body boundary Steel (Internal) material Miscellaneous Piping 0334 and Ducting 0342 Inspection of Internal E 61 Valve Body Moist air Loss of Surfaces in VII.G-23 3.3.1-71 0334 boundary Steel (Internal) material Miscellaneous Piping 0342 and Ducting Inspection of Internal E 65 Valve Body Structural Stainless Moist air Loss of Surfaces in 0332 integrity Steel (Internal) material Miscellaneous Piping V.D1-29 3.2.1-08 334 and Ducting 0342

Enclosure A L-1 1-334 Page 4 of 16 Table 3.3.2-31 Aging Management Review Results - Station Plumbing, Drains, and Sumps System ffectNUREG-Aging Aging Management Ro Cmonn ItnddAging Effect NURE1, Table 1 Row Component Intended Material Environment Requiring 1801, Notes No. Type Function(s) Management Program Volume 2 Item Item Inspection of Internal E 70 Valve Body Structural Steel Moist air Loss of Surfaces in VII.G-23 3.3.1-71 O34 Sintegrity (Internal) material Miscellaneous Piping 0342 and Ducting

Enclosure A L-11-334 Page 5 of 16 Affected LRA Section LRA Page No. Affected Paragraph and Sentence Table 3.3.2 Page 3.3-549 Note 0334 Plant-Specific Notes Errata: During RAI response development, FENOC identified that LRA Table 3.3.2 Plant-Specific Note 0334, previously introduced in FENOC RAI response letter dated April 29, 2011 (ML11126A016) had been inadvertently duplicated in FENOC letter dated May 24, 2011 (ML11151AO90). LRA Table 3.3.2 Plant-Specific Note 0334 (from letter dated May 24, 2011 (ML11151A090)), is renumbered as follows:

Plant-Specific Notes:

0224 The Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Program will manage loss of material for 0342 station plumbing, drains, and sumps system components that are open to ambient outdoor or indoor air and subject to frequent wetting and drying, referred to as a "Moist air (Internal)" environment.

Enclosure A L-11-334 Page 6 of 16 Affected LRA Section LRA Page No. Affected Paragraph and Sentence Table 3.4.2-1 Page 3.4-53 1 Row Errata: During RAI response development, FENOC identified that LRA Table 3.4.2 Plant-Specific Note 0411 had been inadvertently duplicated in FENOC letter dated June 3, 2011 (ML11159A132), and applied to a newly added row of LRA Table 3.4.2-1, "Aging Management Review Results - Auxiliary Feedwater System," in that same letter. Plant-Specific Note 0411 is renumbered to 0415 (see "Table 3.4.2 Plant-Specific Notes" revision, below), and the new row of LRA Table 3.4.2-1 is revised to include the new Note number, as follows:

Enclosure A L-11-334 Page 7 of 16 Affected LRA Section LRA Page No. Affected Paragraph and Sentence Table 3.4.2-3 Page 3.4-80 3 Rows Errata: During RAI response development, FENOC identified that LRA Table 3.4.2 Plant-Specific Note 0410 had been inadvertently duplicated in FENOC letter dated June 3, 2011 (ML11159A132), and applied to three newly added rows of LRA Table 3.4.2-3, "Aging Management Review Results - Main Feedwater System," in that same letter. Plant-Specific Note 0410 is renumbered to 0414 (see "Table 3.4.2 Plant-Specific Notes" revision, below), and the new rows of LRA Table 3.4.2-3 are revised to include the new Note number, as follows:

Table 3.4.2-3 Aging Management Review Results - Main Feedwater System Row Component Intended Aging Effect Aging NUREG-Ro opoet Inedd Material Environment Reurn Magent 1801, Table No. Type Function(s) equiring Management anagement Program Volume 2 Item Item I Notes Treated water A Piping Pressure Steel > 60C Cracking TLAA VIII.D1-7 3.4.1-01 044

-boundary (> 140F) 0414 (Internal)

Treated water A Pressure Stainless >

Tubing boundary Steel (>60C 140F) Cracking TLAA VII.E3-14 3.3.1-02 0440 (Internal) 0414 Treated water A Valve Body Pressure Steel > 60C Cracking TLAA VIII.D1-7 3.4.1-01 040 boundary (> 140F) 0414 (Internal)

Enclosure A L-1 1-334 Page 8 of 16 Affected LRA Section LRA Page No. Affected ParaoraDh and Sentence Table 3.4.2-4 Page 3.4-109 12 Rows Errata: During RAI response development, FENOC identified that LRA Table 3.4.2 Plant-Specific Note 0410 had been inadvertently duplicated in FENOC letter dated June 3, 2011 (MLI 11 59A1 32), and applied to 12 newly added rows of LRA Table 3.4.2-4, "Aging Management Review Results - Main Steam System," in that same letter. Plant-Specific Note 0410 is renumbered to 0414 (see "Table 3.4.2 Plant-Specific Notes" revision, below), and the new rows of LRA Table 3.4.2-4 are revised to include the new Note number, as follows:

Table 3.4.2-4 Aging Management Review Results - Main Steam System Row Component Intended Aging Effect Aging NUREG-Row Copnet Fnctiondd Material Environment Requiring Management 1801, Table 1 Notes No. Type Function(s) Maaeet PormVolume Item Management Program 2 Item 3 A Pressure Steam VIII.B1-Piping boundary Steel (Internal) Cracking TLAA 10 3.4.1-01 4 0414 Treated water A Piping Pressure Steel > 60C (> Cracking TLA VIII.B1- 3.4.1-01 04-boundary 140F) 10 0414 (interal) _4 Steam V11I.131-A A Structural Piping integrity Steel (Internal) Cracking TLAA 10 3.4.1-01 0440 i (0414 Steam VllI.B1-A A Pressure Trap Body boundary Steel (internal) Cracking TLAA 10 3.4.1-01 044 0414

Enclosure A L-1 1-334 Page 9 of 16 Table 3.4.2-4 Aging Management Review Results - Main Steam System Aging Effect Aging NUREG-Row Component Intended Material Environment Requiring Management 1801, Table 1 Notes No. Type Function(s) Malagenent Manam Volume Item Management Program 2 Item Pressure Stainless Steam A

-- Tubing boundary Steel (Internal) Cracking TLAA VII.E3-14 3.3.1-02 044 0414 Treated water A Tubing Pressure boundary Stainless Steel

> 60C (>

140F)

Cracking TLAA VIIE3-14 3.3.1-02 044 0414 (Internal)

Structural Stainless Steam A Tubing integrity Steel (Intenal) Cracking TLAA VII.E3-14 3.3.1-02 0440 0414 Treated water A Tubing

-- Structural Stainless > 60C (> Cracking TLAA VIIE3-14 3.3.1-02 044 integrity Steel 140F) 0414

~0414 (Internal)

Pressure Stainless Steam A Valve Body boundary Steel (Internal) Cracking TLAA VII.E3-14 3.3.1-02 0440 0414 1 A Pressure Steam VlII.B1-Valve Body boundary Steel (Internal) Cracking TLAA 10 3.4.1-01 0414 0414 Treated water A Valve Body Pressure Steel > 60C (> Cracking TLA VIII.B1- 3.4.1-01 0440 boundary 140F) 10 0414 (internal)

Steam VIII.BI-A A Structural Valve Body integrity Steel (internal) t Cracking

( l TLAA 110 3.4.1-01 0440 0414

Enclosure A L-1 1-334 Page 10 of 16 Affected LRA Section LRA Page No. Affected Paragraph and Sentence Table 3.4.2 Page 3.4-111 Notes 0410 and 0411 Plant-Specific Notes Errata: During RAI response development, FENOC identified that LRA Table 3.4.2 Plant-Specific Notes 0410 and 0411 had been inadvertently duplicated in FENOC letter dated June 3, 2011 (ML11159A132). LRA Table 3.4.2 Plant-Specific Notes 0410 and 0411 are renumbered as follows:

Plant-Specific Notes:

0440 Fatigue TLAA is evaluated in LRA Section 4.3.3.1, for piping and (in-line) piping components.

0414 0444 Fatigue TLAA is evaluated in LRA Section 4.3.3.1, for piping (including in the Auxiliary Feedwater System near the 0415 steam generators).

Enclosure A L-1 1-334 Page 11 of 16 Affected LRA Section LRA Page No. Affected Paraaraoh and Sentence Table 4.1-1 Page 4.1-4 "High Pressure Injection / Makeup Nozzle Thermal Sleeves - life prediction" row In response to RAI 4.7.4-1, the "High Pressure Injection / Makeup Nozzle Thermal Sleeves - life prediction" row of LRA Table 4.1-1, "Time-Limited Aging Analyses," previously revised by FENOC letter dated June 3, 2011 (ML11159A132), is revised as follows:

Table 4.1-1 Time-Limited Aging Analyses Results of TLAA Evaluation by Category 54.21(c)(1) LRA Paragraph Section Other'Plant-Specific Time-Limited Aging Analyses _________.._ 4.7 High Pressure Injection / Makeup Nozzle Thermal Sleeves - life Aet-a-TL4A 4.7.4 prediction (/I)

Enclosure A L-1 1-334 Page 12 of 16 Affected LRA Section LRA Page No. Affected Paragraph and Sentence 4.7.4 Page 4.7-5 2 nd paragraph, and "Disposition" subsection In response to RAI 4.7.4-1, the second paragraph and the "Disposition" subsection of LRA Section 4.7.4, "High Pressure Injection / Makeup Nozzle Thermal Sleeves," previously revised by FENOC letter dated June 3, 2011 (MLI 1159A132), are replaced in their entirety to read as follows:

After re-routinq of the makeup flow throuqh HPI nozzle A-2, the thermal sleeve for this nozzle has since been replaced during the Cycle 13 refueling outage that ended in March 2004. In addition, the Inservice Inspection Program was revised to require an augmented VT-1 visual examination of the makeup nozzle thermal sleeve once every other refueling outage commencing with the Cycle 15 refueling outage. Therefore, the effects of cracking on the makeup nozzle thermal sleeve will be managed by the Inservice Inspection Program through the period of extended operation.

Disposition: 10 CFR 54.21(c)(1)(iii) The effects of cracking on the makeup nozzle thermal sleeve will be managed by the Inservice Inspection Program through the period of extended operation.

Enclosure A L-11-334 Page 13 of 16 Affected LRA Section LRA Paqe No. Affected Paragraph and Sentence A.2.7.4 Page A-49 2 nd and 3 rd paragraphs In response to RAI 4.7.4-1, the second and third paragraphs of LRA Section A.2.7.4, "High Pressure Injection / Makeup Nozzle Thermal Sleeves," previously revised by FENOC letter dated June 3, 2011 (ML11159A132), are replaced in their entirety to read as follows:

After re-routing of the makeup flow through HPI nozzle A-2. the thermal sleeve for this nozzle has since been replacedduring the Cycle 13 refueling outage that ended in March 2004. In addition, the Inservice Inspection Program was revised to require an augmented VT-1 visual examination of the makeup nozzle thermal sleeve once every other refueling outage commencing with the Cycle 15 refueling outage.

The effects of cracking on the makeup nozzle thermal sleeve will be managed by the Inservice Inspection Program through the period of extended operation in accordance with 10 CFR 54.21(c)(1)(iii).

Enclosure A L-1 1-334 Page 14 of 16 Affected LRA Section LRA Page No. Affected Paragraph and Sentence Table A-1 Page A-69 Commitment No. 23 In response to RAI 4.7.4-1, license renewal future Commitment No. 23 in LRA Table A-i, "Davis-Besse License Renewal Commitments," previously revised in FENOC letter dated June 17, 2011 (ML11172A389),

is revised to read as follows:

Table A-1 Davis-Besse License Renewal Commitments Item Number Commitment Implementation Schedule I u-Source jRelat LRA Section No./

Comments 23 In association with the TLAA for effects of environmentally assisted Prior to LRA A.2.3.4.2 fatigue of the high pressureinjection (HPI) nozzle safe end April 22, 2017 and 4274 including the associatedAlloy 82/182 weld (weld that connects the safe end to the nozzle), and cracking of the HP!nmakoup nozzle FENOC Responses to thermal sleeve,-replace the HPI nozzle safe end including the Letters NRC RAls associatedAlloy 82/182 weld, and tho thermal *!*eov* for all four L-11-107, 4.7.4-1 from HPI nozzles priorto the period of extended operation.The Fatigue L-11-203 NRC Letter Monitoring Program will evaluate the environmental effects and and dated manage cumulative fatigue damage for the replacement high L-11-334 April 15, 2011, pressure injection (HPI) nozzle safe ends and associated welds. 4.3-18 from NRC Letter dated June 17, 2011, and 4.7.4-1 from NRC Letter

Enclosure A L-1 1-334 Page 15 of 16 Table A-1 Davis-Besse License Renewal Commitments Related LRA Item Implementation Source Section No./

Number Commitment Schedule comments Comments dated October 11, 2011

Enclosure A L-1 1-334 Page 16 of 16 Affected LRA Section LRA Page No. Affected Paragraph and Sentence B.2.24 Page B-151 Program Description, 3 rd paragraph In response to RAI 4.7.4-1, LRA Section B.2.24, "Inservice Inspection Program,"

subsection "Program Description," third paragraph, is revised to read as follows:

The Inservice Inspection Program has been augmented to include commitments made to the regulatoryauthoritiesbeyond the ASME Code, Section Xl. Examples include the augmented examination of auxiliary feedwater header components, high pressure injection (IHPI) ASME Class 1 piping welds, HPI/makeup nozzle thermal sleeve, and decay heat removal ASME Class 1 pipe to valve welds.

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

Letter L-1 1-334 Affidavit for Calculation No. 32 - 9124893 - 001, "DB-1 Pressurized Thermal Shock (PTS) Analysis for 32 and 52 EFPY" 3 pages follow

AFFIDAVIT COMMONWEALTH OF VIRGINIA )

) ss.

CITY OF LYNCHBURG )

1. My name is Gayle F. Elliott. I am Manager, Product Licensing, for AREVA NP Inc. (AREVA NP) and as such I am authorized to execute this Affidavit.

2. I am familiar with the criteria applied by AREVA NP to determine whether certain AREVA NP information is proprietary. I am familiar with the policies established by AREVA NP to ensure the proper application of these criteria.

3. I am familiar with the AREVA NP information contained in 32-9124893-001 entitled, "DB-1 Pressurized Thermal Shock (PTS) Analysis for 32 and 52 EFPY," dated December 2009 and referred to herein as "Document." Information contained in this Document has been classified by AREVA NP as proprietary in accordance with the policies established by AREVA NP for the control and protection of proprietary and confidential information.

4. This Document contains information of a proprietary and confidential nature and is of the type customarily held in confidence by AREVA NP and not made available to the public. Based on my experience, I am aware that other companies regard information of the kind contained in this Document as proprietary and confidential.

5. This Document has been made available to the U.S. Nuclear Regulatory Commission in confidence with the request that the information contained in this Document be withheld from public disclosure. The request for withholding of proprietary information is made in accordance with 10 CFR 2.390. The information for which withholding from disclosure is requested qualifies under 10 CFR 2.390(a)(4)

"Trade secrets and commercial or financial information."

6. The following criteria are customarily applied by AREVA NP to determine whether information should be classified as proprietary:

(a) The information reveals details of AREVA NP's research and development plans and programs or their results.

(b) Use of the information by a competitor would permit the competitor to significantly reduce its expenditures, in time or resources, to design, produce, or market a similar product or service.

(c) The information includes test data or analytical techniques concerning a process, methodology, or component, the application of which results in a competitive advantage for AREVA NP.

(d) The information reveals certain distinguishing aspects of a process, methodology, or component, the exclusive use of which provides a competitive advantage for AREVA NP in product optimization or marketability.

(e) The information is vital to a competitive advantage held by AREVA NP, would be helpful to competitors to AREVA NP, and would likely cause substantial harm to the competitive position of AREVA NP.

The information in this Document is considered proprietary for the reasons set forth in paragraphs 6(b) and 6(c) above.

7. In accordance with AREVA NP's policies governing the protection and control of information, proprietary information contained in this Document has been made available, on a limited basis, to others outside AREVA NP only as required and under suitable agreement providing for nondisclosure and limited use of the information.

8. AREVA NP policy requires that proprietary information be kept in a secured file or area and distributed on a need-to-know basis.

9. The foregoing statements are true and correct to the best of my knowledge, information, and belief.

SUBSCRIBED before me this day of _ _ _ _ ,2011.

Sherry L. McFaden NOTARY PUBLIC, COMMONWEALTH OF VIRGINIA MY COMMISSION EXPIRES: 10/31/14 Reg. # 7079129 SHERRY MCFADEN L.Public Notafq '

Commonwealth of Virg nia [

7079129 MY Commission Expires Oct 31, 2014