Text

License Renewal Application Amendment No. 50 - Annual Update
	ML14175B381
Person / Time
Site:	Davis Besse
Issue date:	06/23/2014
From:	Lieb R; FirstEnergy Nuclear Operating Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	L-14-206, TAC ME4613, TAC ME4640
	Download: ML14175B381 (92)
	v • d • e

FENOC-5501 North State Route 2 FirstEnergy Nuclear Operating Company Oak Harbor, Ohio 43449 Raymond A. Lieb 419-321-7676 Vice President, Nuclear Fax. 419-321-7582 June 23, 2014 L-14-206 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 License Renewal Application Amendment No. 50 -Annual Update (TAC Nos. ME4640 and ME4613)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).

Each year following submittal of a license renewal application (LRA) and at least 3 months before scheduled completion of the NRC review, 10 CFR 54.21(b) requires an amendment to the renewal application to be submitted identifying any change to the current licensing basis (CLB) of the facility that materially affects the contents of the license renewal application, including the Final Safety Analysis Report (FSAR) supplement.

The Attachment provides a summary of the CLB changes that materially affect the LRA.The Enclosure provides Amendment No. 50 (Annual Update) to the DBNPS LRA as required by 10 CFR 54.21(b), including an update to LRA Appendix E, "Applicant's Environmental Report -Operating License Renewal Stage," Section 3.2,"Refurbishment Activities." A475 Davis-Besse Nuclear Power Station, Unit No. 1 L-14-206 Page 2 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 June _. , 2014.Sincer ly, aymo d A. Lieb

Attachment:

Current Licensing Basis (CLB) and Other Changes that Materially Affect the Davis-Besse Nuclear Power Station, Unit No. 1 (Davis-Besse), License Renewal Application (LRA)

Enclosure:

Amendment No. 50 (Annual Update) to the Davis-Besse License Renewal Application cc: NRC DLR Project Manager NRC DLR Environmental Project Manager NRC Region Ill Administrator cc: w/o Attachment or Enclosure NRC DLR Director NRR DORL Project Manager NRC Resident Inspector Utility Radiological Safety Board Attachment L-14-206 Current Licensing Basis (CLB) and Other Changes that Materially Affect the Davis-Besse Nuclear Power Station, Unit No. 1 (Davis-Besse), License Renewal Application (LRA), Page 1 of 8 FirstEnergy Nuclear Operating Company (FENOC) performed a review of Davis-Besse CLB and other document changes since submittal of the 2013 LRA Annual Update provided by letter dated September 20, 2013 (ML13269A027), and identified the following items that materially affect the contents of the Davis-Besse LRA: I1. Steam Generator Replacement The Davis-Besse original steam generators were replaced during the Cycle 18 refueling outage (Spring 2014). In support of this modification, a portion of the reactor coolant system hot leg piping was replaced.

The replacement steam generators are designed, fabricated, and analyzed with the intent of matching as closely as possible the thermal-hydraulic performance of, and being direct replacements for, the original steam generators.

Several material improvements were made including Alloy 690 TT tubes with Alloy 690 tubesheet cladding.

In addition, the replacement steam generators include design changes, such as a lower bowl flat bottom and a base support stool on a base support platform.Currently, there are no approved tube repair methods for the replacement steam generators; therefore, the component type of tube "sleeve" is not included in the steam generator aging management review results.The following LRA Sections are revised to address the replacement steam generators and new Reactor Coolant System hot leg piping:* Section 2.3.1.4, "Steam Generators"" Table 2.3.1-4, "Steam Generators Components Subject to Aging Management Review"" Section 3.1.2.1.4, "Steam Generators"" Table 3.1-1, "Summary of Aging Management Programs for Reactor Vessel, Internals, Reactor Coolant System and Reactor Coolant Pressure Boundary, and Steam Generators Evaluated in Chapter IV of NUREG-1 801"" Table 3.1.2-4, "Aging Management Review Results -Steam Generators"" Table 4.1-1, "Time-Limited Aging Analyses"" Section 4.3.1.2, "Projected Cycles"" Table 4.3-1, "60-Year Projected Cycles"" Section 4.3.2.2.6, "Once Through Steam Generators (OTSGs)"* Section 4.3.2.2.6.1, "OTSGs Fatigue" Attachment L- 14-206 Page 2 of 8" Section 4.3.2.2.6.2, "OTSGs Tube Sleeves Fatigue"" Section 4.3.2.2.6.3, "OTSGs Auxiliary Feedwater Modification"* Section 4.3.2.2.6.4, "OTSGs Tubes and Tube Stabilizers Flow Induced Vibration"* Section 4.3.2.3, "Class 1 Piping and Valves"* Section 4.7.5.2, "OTSG 1-2 Flaw Evaluations"" Section A.1.38, "Steam Generator Tube Integrity Program"" Section A.2.3.1.5, "Steam Generator Remote Welded Plugs"* Section A.2.3.2.6, "Steam Generator Tube Sleeves Fatigue"" Section A.2.3.2.7," Auxiliary Feedwater Header Modification"" Section A.2.3.2.8, "Steam Generator Tubes and Tube Stabilizers Flow Induced Vibration"" Section A.2.3.2.10, "Once Through Steam Generator"" Section A.2.3.2.1 1, "Class 1 Piping"" Section A.2.6.2, "OTSG 1-2 Flaw Evaluations"" Table A-1, "Davis-Besse License Renewal Commitments," Commitment 25" Table B-2, "Consistency of Davis-Besse Aging Management Programs with NUREG-1801"" Section B.2.38, "Steam Generator Tube Integrity Program"" Appendix E, "Applicant's Environmental Report -Operating License Renewal Stage," Section 3.2, "Refurbishment Activities" 2. Station Blackout Recovery Path A new 345-kV transmission line, the "Hayes Line", was installed in the Davis-Besse switchyard.

Four 345-kV transmission lines now supply electric power to the station switchyard.

The transmission line previously identified in the LRA as the "Ohio Edison Line" is now identified as the "Beaver Tower "C.... line on LRA Figure 2.5-1, "Davis-Besse Station Blackout Recovery Path." The Beaver Tower "C" transmission line was relocated and is now connected between a new 345-kV circuit breaker (81 -B-65) and existing circuit breaker ACB34564.

New foundations were added to the Davis-Besse switchyard to support the new breaker and associated transmission structures.

The following LRA Sections are revised to address the new configuration and structural changes to the Davis-Besse switchyard:

0 Table 2.2-3, "License Renewal Scoping Results for Structures" Attachment L-14-206 Page 3 of 8* Section 2.4, "Scoping and Screening Results: Structures"* Section 2.4.12, "Yard Structures" Section 2.4.12.9, "Station Blackout Component Foundations and Structures in the Yard and Switchyard (including Startup Transformers 01 and 02; Bus-Tie Transformers; 345-kV Switchyard circuit breakers ACB34560, ACB34561, ACB34562, ACB34563, ACB34564, 81-B-66 and 81-B-67; Relay House; Switchyard and Yard Towers for 345-kV distribution; "J" and "K" buses) -Seismic Class I1"" Section 2.5.6.2, "Station Blackout Recovery Path Evaluation Boundaries"* Figure 2.5-1, "Davis-Besse Station Blackout Recovery Path" 3. LRA Commitment 13 Enhancement Deletion -Makeup Pumps The enhancement to the One-Time Inspection in license renewal future Commitment 13 currently reads: Enhance the One- Time Inspection to include enhanced visual (EVT-1 or equivalent) or surface examination (magnetic particle, liquid penetrant), or volumetric (RT or UT) inspections to detect and characterize cracking due to cyclic loading of the stainless steel makeup pump casings (DB-P37-1 and 2) of the Makeup and Purification System. The one-time inspections will provide verification of the absence of cracking due to cyclic loading.License renewal future Commitment 13 is revised to delete the enhancement.

Based on the design and operation of the makeup pumps (DB-P37-1 and 2) in the Davis-Besse Makeup and Purification System, cracking due to cyclic loading of the stainless steel makeup pump casings is not an applicable aging effect for the following reasons: The Makeup and Purification System contains two 100 percent capacity makeup pumps. One makeup pump operates continuously during normal operation to supply approximately 120°F seal injection water to the reactor coolant pumps and makeup water to the Reactor Coolant System (RCS) for the letdown flow from the RCS. With a relatively constant temperature of 120 0 F, thermal (i.e., fatigue) cyclic loading is not a concern for the makeup pump casings.

Attachment L-14-206 Page 4 of 8 The makeup pumps are horizontal, twelve stage, stainless steel, centrifugal, 450 HP electric motor driven pumps. The pumps have two separate casings: 1. An internal casing, which receives water from the suction piping at one end of the pump and passes it through stages 1 through 6 to the center of the pump. From the center of the pump internal casing, the fluid is routed to the other end of the internal pump casing and passed through stages 7 through 12 back to the center of the pump, where it is discharged from the internal casing. This arrangement is designed to equalize the axial thrust on the shaft. The internal casing has no license renewal function.2. An external casing, which receives the water discharged from the center of the internal casing and routes it to the discharge piping.The external casing provides the license renewal pressure boundary function for the pump.If mechanical cyclic loading were present in this pump design, then only the internal casing would be subject to cyclic loading since the pressure increase is developed by the 12 stages in the internal casing. Cracking due to cyclic loading on the internal casing could result in a leak from the internal casing to the external casing, reducing the efficiency of the pump.The external casing (i.e., the pump pressure boundary) is not subject to cyclic loading because it receives the fluid discharged from the 12th stage of the inner casing at a steady pressure, flow rate and temperature.

The makeup pumps are not within the boundary of ASME Boiler & Pressure Vessel Code,Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components." This issue was discussed with the Nuclear Regulatory Commission during a telephone conference call held on March 27, 2014 (ML14099A339).

The following LRA Sections are revised to address deletion of cracking due to cyclic loading as an aging effect for the stainless steel makeup pump casings:* Section 3.3.2.2.4.3, "Stainless Steel PWR High Pressure Pump Casings -Treated Borated Water"" Table 3.3.1, "Summary of Aging Management Programs for Auxiliary Systems Evaluated in Chapter VII of NUREG-1801"" Table 3.3.2-18, "Aging Management Review Results -Makeup and Purification System" Attachment L-1 4-206 Page 5 of 8" Table A-1, "Davis-Besse License Renewal Commitments," Commitment 13" Section B.2.30, "One-Time Inspection" 4. License Renewal Interim Staff Guidance LR-ISG-2012-01 FENOC performed a review of Nuclear Regulatory Commission (NRC) license renewal interim staff guidance (LR-ISG), LR-ISG-2012-01, "Wall Thinning Due to Erosion Mechanisms." Using keywords from the ISG document, FENOC performed a search of the Corrective Action Program database to identify previous issues of erosion, flashing or cavitation at Davis-Besse.

The results of the review demonstrated that issues of cavitation were identified and addressed in the Corrective Action Program by design or operational changes to eliminate the issues. Erosion in raw water systems is being managed by the Open Cycle Cooling Water aging management program. The plant operating experience shows that no additional monitoring program is needed for erosion, flashing or cavitation at this time. FENOC continues to monitor plant and industry operating experience for changes to Davis-Besse aging management programs.5. LRA Commitment 33 (Refueling Canal Leakage Mitigation)-

Phase 1 -Actions 1 & 2, and Phase 2 -Action 1 Closure License renewal future Commitment 33, Phases 1 and 2, currently reads, in part: Phase 1 Perform the following actions to reduce or mitigate the refueling canal leaks inside containment., 1. Select and implement a leak detection method to locate the leakage area.2. Evaluate temporary and permanent repair methods to stop or significantly reduce the leakage, and implement a repair plan.Phase 2 Perform the following actions to evaluate the impact of refueling canal leaks on concrete and reinforcing steel structures.

Discontinue core bores, testing and reinforcing steel inspections when indications of refueling canal leakage are no longer present: 1. Perform a core bore in the south wall of the east-west section of the core flood pipe tunnel.a. Assess borated water degradation of the concrete by testing the core bore sample for compressive strength and by petrographic examination, and evaluate the results.

Attachment L-14-206 Page 6 of 8 b. Conduct a visual examination of the concrete and reinforcing steel to identify aging effects (e.g., concrete degradation or steel corrosion).

Enter identified aging effects into the FENOC Corrective Action Program and evaluate in accordance with the requirements of the current licensing basis Maintenance Rule Program.FENOC evaluated potential leakage pathways and methods to reduce or stop the leakage from the refueling canal. During the recent Cycle 18 refueling outage (Spring 2014), FENOC implemented a permanent refueling canal leak repair method using a two-component silicone rubber based coating material with stainless steel cover plates to encapsulate penetrations (piping, bolting, drain, etc.)through the canal liner. Upon refilling the refueling canal for fuel transfer at the end of the outage, the leakage was reduced from standing water in several areas to minor moist areas of concrete in one specific location, a significant reduction estimated to be almost 100 percent successful.

FENOC obtained a core bore sample that included a section of reinforcing bar (rebar) from the south wall of the east-west section of the core flood pipe tunnel, and submitted the sample to a concrete testing laboratory to assess borated water degradation of the concrete and steel by visual examination, testing the core bore sample for compressive strength, and by petrographic examination.

The core segment was judged to be in good condition, the paste was hard and dense, the bond to the aggregate was tight, and there was no indication of physical degradation resulting from acid attack. Compressive strength testing of the core sample showed values above design values. There was no appreciable deterioration of the rebar sample, indicating that the mechanical integrity of the rebar had not been compromised.

Where the concrete was in physical contact with the rebar, the surface did not exhibit any corrosion.

The remaining surfaces were covered by a superficial oxidation product (rust). No boron was observed.

FENOC engineering evaluated the laboratory results and determined the results were satisfactory, and that no aging effects were identified that required entry into the FENOC Corrective Action Program.LRA Table A-I, "Davis-Besse License Renewal Commitments," is revised to address partial closure of license renewal future Commitment 33, specifically, Phase 1 -Actions 1 & 2 and Phase 2 -Action 1 activities.

6. LRA Commitment 35 (Containment Vessel Exterior Surface Non-Destructive Examination)-Phase 1 Closure License renewal future Commitment 35 currently reads: Perform the following actions for each of two examinations (Phase I and Phase 2)of the Containment Vessel in the sand pocket region:

Attachment L-1 4-206 Page 7 of 8* Perform nondestructive examination (NDE) of the Containment Vessel from the outer surface at five areas of previously identified groundwater in-leakage.

o Examine the vessel at a minimum of three vertical grid locations at 12 inches nominal horizontal spacing at each area. Examine the Containment Vessel at a minimum of three elevations:

1. approximately 3 inches below the existing grout-to-vessel interface in the sand pocket region;2. at the existing grout-to-vessel interface level in the sand pocket region;and, 3. approximately 3 inches above the existing grout to vessel interface in the sand pocket region.* Compare the ultrasonic test (UT) thickness readings to minimum ASME Code vessel thickness requirements and to the results obtained during previous UT examinations of the Containment Vessel. Determine the need for maintenance or repair of the Containment Vessel based on the results and evaluation of the examinations." Document the results of each of the two examinations in the work order system.Document and evaluate adverse conditions in accordance with the FENOC Corrective Action Program for an evaluation of potential degradation of the steel Containment Vessel thickness over the longer term.During the recent Cycle 18 refueling outage (Spring 2014), FENOC performed UT of the outer surface of the Containment Vessel in the sand pocket region as described in Commitment

35. The UT results were consistent with previous UT examinations of the Containment Vessel and were above calculated minimum design ASME Code vessel thickness requirements, with margin. The results were determined to be acceptable, and were documented in the work order system.LRA Table A-i, "Davis-Besse License Renewal Commitments," is revised to address partial closure of license renewal future Commitment 35, specifically, Phase 1 activities.

7. LRA Commitment 39 (Containment Vessel Interior Surface Non-Destructive Examination)--Phase 1 Closure License renewal future Commitment 39 currently reads: Address the potential for borated water degradation of the steel containment vessel through the following actions:* Access the inside surface of the embedded steel containment at a vertical height no greater than 10 inches above bottom dead center. A core bore will be Attachment L-14-206 Page 8 of 8 completed by the end of 2014 (Phase 1). If necessary, a second core bore will be completed by the end of 2020 (Phase 2). If there is evidence of the presence of borated water in contact with the steel containment vessel, conduct non-destructive testing (NDT) to determine what effect, if any, the borated water has had on the steel containment vessel. Based on the results of NDT, perform a study to determine the effect through the period of extended operation of any identified loss of thickness in the steel containment due to exposure to borated water.During the recent Cycle 18 refueling outage (Spring 2014), FENOC performed core bores to access the inside surface of the embedded steel Containment Vessel as described in Commitment

39. There was no visual evidence of the presence of borated water in the concrete or in contact with the steel containment vessel. FENOC performed UT of the Containment Vessel to confirm the thickness of the vessel. The UT thickness measurement at that location was above nominal (i.e., greater than 1.5 inches).FENOC obtained core bore samples, one of which included reinforcing bar (rebar), and submitted the samples to a concrete testing laboratory to assess borated water degradation of the concrete and steel by visual examination, testing the core bore sample for compressive strength, and by petrographic examination.

The core segments were judged to be in good condition.

The paste was hard and dense.The bond to the aggregate was tight. Compressive strength testing of the core sample showed values above design values. There was no appreciable deterioration of the rebar sample, indicating that the mechanical integrity of the rebar had not been compromised.

Where the concrete was in physical contact with the rebar, the surface did not exhibit any corrosion.

The remaining surfaces were covered by a superficial oxidation product (rust). No boron was observed.

FENOC engineering evaluated the laboratory results and determined the results were satisfactory, and that no aging effects were identified that required entry into the FENOC Corrective Action Program.LRA Table A-i, "Davis-Besse License Renewal Commitments," is revised to address partial closure of license renewal future Commitment 39, specifically, Phase 1 activities.

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-14-206 Amendment No. 50 (Annual Update) to the Davis-Besse License Renewal Application Page 1 of 82 License Renewal Application Sections Affected Section 2 Table 2.2-3 Section 2.3.1.4 Table 2.3.1-4 Section 2.4 Section 2.4.12 Section 2.4.12.9 Section 2.5.6.2 Figure 2.5-1 Section 3 Section 3.1.2.1.4 Table 3.1.1 Table 3.1.2-4 Section 3.3.2.2.4.3 Table 3.3.1 Table 3.3.2-18 Table 3.3.2 P-S Notes Section 4 Table 4.1-1 Section 4.3.1.2 Table 4.3-1 Section 4.3.2.2.6 Section 4.3.2.2.6.1 Section 4.3.2.2.6.2 Section 4.3.2.2.6.3 Section 4.3.2.2.6.4 Section 4.3.2.3 Section 4.7.5.2 Appendix A Section A. 1.38 Section A.2.3.1.5 Section A.2.3.2.6 Section A.2.3.2.7 Section A.2.3.2.8 Section A.2.3.2.10 Section A.2.3.2.11 Section A.2.6.2 Table A-1 Appendix B Table B-2 Section B.2.30 Section B.2.38 Appendix E: Environmental Report, Section 3.2 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 fined- '- and added text underlined.

Enclosure L-1 4-206 Page 2 of 82 Affected LRA Section LRA Page No.Affected Paragraph and Sentence Table 2.2-3 2.2-13 1 revised row Based on the addition of the foundation for the new switchyard breaker and reconfiguration of the switchyard, LRA Table 2.2-3, "License Renewal Scoping Results for Structures," is revised to read as follows: Table 2.2-3 License Renewal Scoping Results for Structures Structure Name [In-Scope Screening Results / Section Station Blackout Component Foundations and Structures in the Yard and Switchyard (Startup Transformers 01 and 02, Bus-Tie Transformers, 345-k V Switchyard circuit Yes 2.4.12 breakers A CB34560, ACB34561, ACB34562, ACB34563, ACB34564, 81-B-65,.81-B-66 and 81-B-67, Relay House, and "J" and "K" Buses)

Enclosure L-14-206 Page 3 of 82 Affected LRA Section LRA Page No. Affected Paragraph and Sentence 2.3.1.4 2.3-18 Entire Section Based on the installation of the replacement steam generators, LRA Section 2.3.1.4, "Steam Generators," is replaced in its entirety to read as follows: 2.3.1.4 Steam Generators

System Description

The steam generator (replacement steam generator) is a vertical, straight-tube-and-shell heat exchanger that produces superheated steam at approximately a constant pressure over the power range. Reactor coolant water enters the steam generator at the upper primary head, flows down the Nickel Alloy 690 TT tubes while transferring heat to the secondary shell-side fluid, and leaves through the lower primary head. Steam is generated on the shell side.The hiqh-pressure parts of the unit are the hemispherical heads, the tubesheets, and the straiqht tubes between the tubesheets.

The reactor coolant side has access ports (manways and inspection openings).

The lower bowl is a flat bottom desiqn and thus eliminates the need for a lower primary head drain line. There is a head-to-hot

/eq vent line connection on steam generator 1-2.The shell, the outside of the tubes, and the tubesheets form the boundaries of the steam-producing section of the vessel. Within the shell, the tube bundle is surrounded by a shroud which separates the feedwater inlet (lower annulus between the shell and the shroud) and steam outlet (upper annulus between the shell and the shroud) from the boiling (tube) region. Tube supports hold the tubes in a uniform pattern along their length. Vents, drains, instrumentation nozzles, and access Ports (manways, handholes, and inspection openings) are provided on the shell side of the unit.Reactor coolant enters the steam generator through a nozzle in the upper head, flows down inside the tubes, exits through two outlet nozzles in the lower head, flows to the reactor coolant pumps, and back to the reactor. The main feedwater (MFW) enters each steam generator through a divided circular header and 32 feedwater nozzles. The feedwater nozzles spray the water down into an annulus between the shell and the shroud. During upset or emergency conditions, feedwater may be added through auxiliary feedwater (AFW) nozzles which are located high in the steam generator and discharge directly into the tube bundle.

Enclosure L-14-206 Page 4 of 82 The unit is supported by a base support stool on a base support platform.

This provides for accessibility during outage inspection and maintenance.

Enclosure L-14-206 Page 5 of 82 Affected LRA Section LRA Page No.Affected Paragraph and Sentence Table 2.3.1-4 2.3-21 Entire Table Based on the installation of the replacement steam generators, LRA Table 2.3.1-4, "Steam Generator Components Subject to Aging Management Review," previously revised by FENOC letter dated October 21, 2011 (ML11298A097), is replaced in its entirety to read as follows: Table 2.3.1-4 Steam Generators Components Subject to Aging Management Review Component Type Intended Function C(as defined in Table 2.0-1)oPressure boundary Primary side; manway and inspection opening cover Pressure boundary Primary side; tube Heat transfer Pressure boundary Primary side; tube pluq Pressure boundary Primary side; upper and lower head, inlet and outlet Pressure boundar nozzle Primary side; upper and lower tubesheet Pressure boundary Primary side; tube-to-tubesheet weld Pressure boundary Secondary side; AFW header and riser Pressure boundary Secondary side; AFW nozzle and thermal sleeve Pressure boundary Secondary side; shroud, shroud support ring and lugs Support Secondary side; manway and handhole cover Pressure boundary Secondary side; MFW header support plate and Support gusset Secondary side; MFW header and riser Pressure boundary Secondary side; MFW spray nozzle and thermal Pressure boundar sleeve Secondary side; nozzle (steam outlet, level sensing, Pressure boundar drain and vent nozzle)Secondary side; shell Pressure boundary Enclosure L-14-206 Page 6 of 82 Table 2.3.1-4 (Continued)

Steam Generators Components Subject to Aging Management Review Component Type Intended Function I (as defined in Table 2.0-1)Secondary side; tube suport dplate Support Secondary side; tube support plate spacer Support Secondary side; tube support rod (tie rod) Support Steam generator base support stool and base support Support platform Enclosure L-1 4-206 Page 7 of 82 Affected LRA Section LRA Page No. Affected Paragraph and Sentence 2.4 2.4-1 & 2.4-2 Note Based on the addition of the foundation for the new switchyard breaker and reconfiguration of the switchyard, the "Note" at the end of LRA Section 2.4,"Scoping and Screening Results: Structures," is revised to read as follows: Note: The yard structures evaluated for license renewal include foundations and structural arrangements for the Borated Water Storage Tank (including Trench);Diesel Oil Pump House, Diesel Oil Storage Tank, Emergency Diesel Generator Fuel Oil Storage Tanks; Fire Hydrant Hose Houses; Fire Walls between Bus-Tie Transformers, between Bus-Tie and Startup Transformer 01, and between Auxiliary and Main Transformers; Fire Water Storage Tank; Nitrogen Storage Building; Station Blackout Components and Structures In the Yard and Switchyard (Startup Transformers 01 and 02, Bus-Tie Transformers, 345-kV Switchyard circuit breakers ACB34560, ACB34561, ACB34562, ACB34563, ACB34564, 81-B-65, 81-B-66 and 81-B-67, Relay House, Switchyard and Yard Towers for 345-kV distribution, "J" and "K" buses); Wave Protection Dikes; Duct Banks; Cable Trenches; and Manholes.

Enclosure L-14-206 Page 8 of 82 Affected LRA Section LRA Page No. Affected Paragraph and Sentence 2.4.12 2.4-38 1 1 th Bullet Based on the addition of the foundation for the new switchyard breaker and reconfiguration of the switchyard, the eleventh bullet (Station Blackout Component Foundations and Structures) in the list of Yard Structures in LRA Section 2.4.12, "Yard Structures," is revised to read as follows: Station Blackout Components and Structures in the Yard and Switchyard including Startup Transformers 01 and 02; Bus-Tie Transformers; 345-kV Switchyard circuit breakers ACB34560, ACB34561, ACB34562, ACB34563, ACB34564, 81-B-65, 81-B-66 and 81-B-67; Relay House, Switchyard and Yard Towers for 345-kV distribution, and the 345-kV Switchyard "J" and "K" buses Enclosure L-14-206 Page 9 of 82 Affected LRA Section LRA Page No. Affected Paragraph and Sentence 2.4.12.9 2.4-42 & 43 "Title"; and"Structure Description" Ist and 4 th Paragraphs Based on the addition of the foundation for the new switchyard breaker and reconfiguration of the switchyard, the Section title, and first and fourth paragraphs of LRA Section 2.4.12.9, "Station Blackout Component Foundations and Structures in the Yard and Switchyard (including Startup Transformers 01 and 02; Bus-Tie Transformers; 345-kV Switchyard circuit breakers ACB34560, ACB34561, ACB34562, ACB34563, ACB34564, 81-B-66 and 81-B-67; Relay House; Switchyard and Yard Towers for 345-kV distribution

"J" and "K" buses) -Seismic Class I1," and, although not shown below, the LRA Table of Contents, are revised to read as follows: 2.4.12.9 Station Blackout Component Foundations and Structures in the Yard and Switchyard (including Startup Transformers 01 and 02;Bus-Tie Transformers; 345-kV Switchyard circuit breakers ACB34560, ACB34561, ACB34562, ACB34563L ACB34564,..81-B-65 81-B-66 and 81-B-67; Relay House; Switchyard and Yard Towers for 345-kV distribution; "J" and "K" buses) -Seismic Class II The station blackout component foundations and structures in the yard and switchyard (including Startup Transformers 01 and 02; Bus-Tie Transformers; 345-kV switchyard circuit breakers ACB34560, ACB34561, ACB34562, ACB34563, ACB34564, 81-B-65, 81-B-66 and 81-B-67; Relay House; Switchyard and Yard Towers for 345-kV distribution; "J" and "K" buses) are Seismic Class II structures.

Startup Transformers 01 and 02, Bus-Tie Transformers, and associated breakers (circuit breakers ACB34560, ACB34561, ACB34562, ACB34563, ACB34564, 81-B-65, 81-B-66 and 81-B-67) define the physical boundary that provides an offsite alternating current (AC) source for recovery from a station blackout regulated event.Circuit breakers ACB34560, ACB34561, ACB34562, ACB34563, ACB34564, 81-B-65, 81-B-66 and 81-B-67; the Relay House and "J" and "K" Bus Support Structures are located within the 345-kV Switchyard.

The Relay House is located just east of the switchyard.

Enclosure L-14-206 Page 10 of 82 Affected LRA Section LRA Page No. Affected Paragraph and Sentence 2.5.6.2 2.5-8 2 nd to last paragraph Based on the addition of the new switchyard breaker and the reconfiguration of the Davis-Besse switchyard, the 2 nd to last paragraph of LRA Section 2.5.6.2,"Station Blackout Recovery Path Evaluation Boundaries," is revised to read as follows: Within the switchyard, there are two 345-kV buses -the "J" (East) bus and the"K" (West) bus. The "J" bus is closest to the plant and the "K" bus is located on the farther side of the switchyard, closer to the grid. The current switchyard configuration includes circuit breakers ACB34560, ACB34561, ACB34562, ACB34563, ACB34564, 81-B-65, 81-B-66, and 81-B-67 in a ring bu-GGnfiguFaton.

These circuit breakers and the switchyard buses are within the license renewal evaluation boundary.

This configuration is shown in simplified graphical form in USAR Figure 8.2-2 and in Figure 2.5-1 below.

Enclosure L-14-206 Page 11 of 82 Affected LRA Section LRA Page No. Affected Paragraph and Sentence Figure 2.5-1 2.5-9 Entire Figure Based on the addition of the new switchyard breaker and the reconfiguration of the switchyard, LRA Figure 2.5-1, "Davis-Besse Station Blackout Recovery Path," is replaced in its entirety as follows:[See LRA Figure 2.5-1 on page 12 of 82]

Enclosure L-14-206 Page 12 of 82 Figure 2.5-1 Davis-Besse Station Blackout Recovery Path BUS TIE TRANSFORMER BD'(13.BKV TO 4.16KV)IACB 34563 8-B-66-34561 45J ACB 34564 3456 l81-B-65\ V TO LEMOYNE TO BE SUBSTATION TOWER BI -B-67-1'-~I-TO BAYSHORE SUBSTATION 345KV BUS AVER.C..(WEST BUS)TO HAYES TOWER "C'-NOT IN SCOPE OF LICENSE RENEWAL SBO RECOVERY PATH Enclosure L-14-206 Page 13 of 82 Affected LRA Section LRA Page No. Affected Paragraph and Sentence 3.1.2.1.4 3.1-6 Materials Based on the installation of the replacement steam generators, a new bullet is added under the "Materials" heading of LRA Section 3.1.2.1.4, "Steam Generators," which is revised to read as follows: Materials The materials of construction for subject items of the Steam Generators are: " Nickel alloy* Stainless Steel" Steel" Steel with nickel alloy cladding" Steel with stainless steel backing" Steel with stainless steel cladding Enclosure L-14-206 Page 14 of 82 Affected LRA Section LRA Page No.Affected Paragraph and Sentence Table 3.1.1 3.1-39 Item 3.1.1-74 "Discussion" Based on the installation of the replacement steam generators, the "Discussion" column of Item No. 3.1.1-74 of LRA Table 3.1.1 "Summary of Aging Management Programs for Reactor Vessel, Internals, Reactor Coolant System and Reactor Coolant Pressure Boundary, and Steam Generators Evaluated in Chapter IV of NUREG-1801," is revised to read as follows: Table 3.1.1 Summary of Aging Management Programs for Reactor Vessel, Internals, Reactor Coolant System and Reactor Coolant Pressure Boundary, and Steam Generators Evaluated in Chapter IV of NUREG-1801 Item Aging Effect/ Aging Management Further Number Component/Commodity Mechanism Programs Evaluation Discussion Recommended 3.1.1-74 Chrome plated steel, stainless Cracking due to Steam Generator Tube No Not apliGble., steel, nickel alloy steam stress corrosion Integrity and Water vai,, Boaoo ... .th..ugh generator anti-vibration bars cracking, loss of Chemistry

.............

..t ......d t ..exposed to secondary material due to ...... onQ t............

the.. ...f e e d w a t e r / s t e a m c r e v ic e c o r r o s io n a , h- .. ..........and fretting steam. gener.t.r.

.Loss of material and cracking of the stainless steel tie-rods of the steam generators are managed by a combination of the Steam Generator Tube Integrity Program and the PWR Water Chemistry Program.

Enclosure L-14-206 Page 15 of 82 Affected LRA Section LRA Page No.3.1-164 -185 Affected Paragraph and Sentence Table 3.1.2-4 Entire Table Based on the installation of the replacement steam generators, LRA Table 3.1.2-4, "Aging Management Review Results -Steam Generators," previously revised by FENOC letter dated October 21, 2011 (ML11298A097), is replaced in its entirety to read as follows: Table 3.1.2-4 Aging Management Review Results -Steam Generators Row Component Intended Aging Effect Aging NUREG-Ro opnn nedd Material Environment Reurn Mage nt 1801, Table 1 Notes No. Type Function(s) i equiring anagemen Volume 2 Item Management Program Iteume2 Im Item Air with steam 1 Bolting Pressure Steel or water Crackinqg TLAA .C2-10 3.1.1-07 A boundary leakage Fatique (External)

Air with steam 2 Bolting Pressure Steel or water Cracking -SCC Bolting Integrity IV. C2-7 3.1.1-52 B boundary leakage (External)

Air with steam 3 Boltin Pressure Steel lewater Loss of Material Bolting Integrity V.E-6 3.2.1-22 B_ Boting boundary Steleakage (External)

Air with 4 ______ Pressure Steel borated water Loss of Material Boric Acid ID2-1 3.1.1-58 A Bolting boundary leakage Corrosion (External)

Enclosure L-14-206 Page 16 of 82 Table 3.1.2-4 Aging Management Review Results -Steam Generators Aging Effect Aging NUREG-Row Component Intended Material Environment Requiring Management 1801, Table 1 Notes No. Type Function(s)

Malagenent Manam Volume 2 Item Management Program Item Air with steam 5 Boltin Pressure Steel lewater Loss of Preload Bolting Integrity IV.D2-6 3.1.1-52 B_ Boting bounday Ste leakage (External)

Primary Side; Borated Manway and Steel 6 Inspection Pressure w-SS reactor Cracking -TLAA IV.D2-3 3.1.1-10 A Oeig boundary backin coolantFaiu Ovenin backin (Internal)

Cover Primary Side; Borated anaan Prsue Steel BoaeC Manwayan Pressure wStee reactor Cracking -Flaw Inservice I 7 Inspection boundary W-SS coolant Growth Inspection IC2-26 3.1.1-62 0102 Openin backing (Interal)Cover Primary Side; Borated Manway and Pressure Steel Boae Insection w-SS reactor Cracking -Inservice 8 Inspectio boundary w-SS coolant SCC/IGA Inspection -D2-4 3.1.1-35 C Openin backing (Internal)

Cover Primary Side; Borated Manway and Pressure Steel rorated 9_ Inspectio Presue-S reactor Cracking -PWR Water I.24 3113 9 Inspection boundary wSS coolant SCC/IGA Chemistry V.D2-4 3.1.1-35 C Openin backin (Internal)

Cover Enclosure L-14-206 Page 17 of 82 Table 3.1.2-4 Aging Management Review Results -Steam Generators Agin Effct AingNUREG-Row Component Intended Aging Effect Aging e-1801, Table 1 Notes No. Type Function(s)

Material Environment Requiring Management Nooteeste Management Program Volume 2 Item Item Primary Side: Borated Manway and Pressure Steel reactor PWR Water 10 Inspection w-SS coolant Loss of Material ChemistrIV.C2-15 3.1.1-83 C Openin backing (Inteal)Cover Primary Side: Air with Manway and Pressure Steel borated water Boric Acid 11 Inspection bounda w-SS leakaLOSS of Material D2-1 3.1.1-58 A Openinb backing (External)

Cover r Nickel Borated 12 Primary Side: Pressu reactor Cracking -TLAA IV.D2-15 3.1.1-06 A Tube boundary 690 TT coolant Fatigue (Internal)

Borated Primary Side: Pressure Nickel C 13 im : Pressure Ay reactor Cracking -Flaw Inservice 13 Tube boundary Allo6 coolant Growth Inspection IV.C2-26 3.1.1-62 0102, 690 TT fltmý0103_____________(Internal)

_ _Nickel Borated Cakn ,Primary Side; Pressure Nickel reactor Cracking PWR Water.1 14 Tube boundary Alloy coolant PSCC, Chemistry 690 TTItenl SCC/IGA_ _ _ _ _(Internal)

Nickel Borated Cak g 15 Primary Side; Pressure Nickel reactor Crackin- Steam Generator3.

15 Tube boundar Alloy coolant PWSCC, Tube Integrity IV.D2-14 3..-7_Tuebunay 690 TT (Inte )SCC/IGA Enclosure L-14-206 Page 18 of 82 Table 3.1.2-4 Aging Management Review Results -Steam Generators ponent Intended Aging Effect Aging NUREG-Now Type Function(s)

Material Environment Requiring Management 1801, Table 1 Notes N Management Program Volume 2 Item Item Borated Primary Side; Pressure Nicke reactor Loss of Material IV. C2-15 3.1.1-83 C 16 Tube boundary Alloy coolant Chemistry

-(Internal)

Primary Side; Pressure Nickel Treated water Cracking -PWR Water 17 Tube boundary Alloy (Exteal) SCCIGA Chemistry IV.D2-16 3.1.1-72 A 690 TT Primary Side; Pressure Nickel Treated water Crackinq -Steam Generator 1__8 Tube bounda Alloy (External)

SCC/IGA Tube Integritiy ID2-16 3.1.1-72 A 690 TT Primary Side: Pressure Nickel Treated water Cracking -PWR Water 19 Tube boundary Alloy (External)

SCCIGA Chemistry IV-D2-17 3.1.1-72 A 690 TT Nickel 20 Primary Side; Pressure Alloy Treated water Cracking -Steam Generator IV.D2.17 3.1.1-72 A Tube boundary (External)

SCC/IGA Tube Integrity I________ 690 TT Primary Side; Pressure Nickel Treated water PWR Water 2_ Tube bounda 690 ATT (External)

Denting Chemistry D2-13 3.1.1-75 A 22 Primary Side; Pressure Nickel Treated water Denting Steam Generator Tube boundary Alloy (Extenal)

DTube Inteqrit ID2-13 3.1.1-75 A 690 TT 23 Primary Side; Pressure Nickel Treated water PWR Water 23 Tb onaAlloy (External)

Loss of Material Chemistry IVD2-18 3.1.1-72 A Tube boundary Alo TT Enclosure L-14-206 Page 19 of 82 Table 3.1.2-4 Aging Management Review Results -Steam Generators Row Component Intended Aging Effect Aging NUREG-Row Copnet Fnctiondd Material Environment Requiring Management 1801, Table I Notes No. Type Function(s)

MaaeetPormVolume 2 Item Management Program Item 2.4 Primary Side; Pressure Nickel Treated water Loss of Material Steam Generator IV.D2-18 3.1.1-72 A 690Tube boundar AoTT (External)

Tube Integrity 690kel Primary Side: Nickel Treated water Reduction in PWR Water 25 Tube Heat Transfer Alloy (External)

Heat Transfer Chemistry NA NIA H 690 TT Primary Side Nickel Treated water Reduction in Steam Generator 26 Tube Heat Transfer Alloy (External)

Heat Transfer Tube Integrity NIA NIA H 690 TT Nickel Borated 27 Primary Side; Pressure Alloy reactor Cracking TLAA VD2-15 3.1.1-06 C Tube Plug boundary 690 AOTT coolant Fati-que -0401 (Internal)

Nickel Borated C Primary Side: Pressure reactor Cracking -Flaw Inservice IC2-26 3.1.1-62 0401.Tube Plug boundary Alloy coolant Growth Inspection 0402, (Internal) 0403 Nickel Cracking -29 Primary Side; Pressure Nickl reactor PWSCC, IV.D2-12 3.1.1-73 A Tube Plug boundary Alloy coolant IVD21 Chemist31 0401 690 TT Itenl SCC/IGA Ceity__(Internal)

Nickel Borated Cracking-S Primary Side; Pressure All reactor PWSCC, Steam Generator IV.D2-12 3.1.1-73 A 30 Tube Plug boundary Alloy coolant VD2Tube Integrity 0401 690 TT (Itra)SCC/IGA TueItgiy'_

Enclosure L-1 4-206 Page 20 of 82 Table 3.1.2-4 Aging Management Review Results -Steam Generators Aging Effect Aging NUREG-RowMaterial Environment Requiring Management 1801, Table I Notes No. Type Function(s)

Malagene ing Volume 2 Item Management Program Iteume2 Im Item Borated 31 Primary Side; Pressure Alloy_ reactor Loss of Material PRWtr IV. C2-15 3.1.1-83 --Tube Plug boundary 690 TT coolant Chemistry 0401 (Internal)

Primary Side: Borated Upper and Pressure Steel rated 32 Lower Head, boundaPre w-SS reolant rackiue TLAA IV.D2-3 3.1.1-10 A Inlet and cladding (Internal)

Outlet Nozzle Primary Side: Borated Upper and Pressure Steel reactor Cracking -Flaw Inservice C 33 Lower Head, boundary w-SS coolant Growth Inspection IV.C2-26 3.1.1-62 0102 Inlet and cladding (Internal)

Outlet Nozzle Primary Side; Borated Upper and Pressure Steel reactor Cracking Inservice 34 Lower Head, boundary w-SS coolant SCC/IGA Inspection IV.D2-4 3.1.1-35 A Inlet and cladding (Internal)

Outlet Nozzle Primary Side; Borated Upper and Pressure Steel reactor Crackind PWR Water 35 Lower Head, boundary w-SS coolantD2-4 3.1.1-35 A Inlet and cladding (Internal)

Outlet Nozzle Enclosure L-14-206 Page 21 of 82 Table 3.1.2-4 Aging Management Review Results -Steam Generators Aging Effect Aging NUREG-Row Component Intended Material Environment Requiring Management 1801, Table 1 Notes No. Type Function(s)

MaaeetPormVolume 2 Item Management Program Iteume2 Im Item Primary Side;Upper and Steel Pressure reactor PWR Water 36 Lower Head, boundary w-SS coolant Loss of Material Chemistry IV. C2-15 3.1.1-83 C Inlet and cladding (Internal)

Outlet Nozzle Primary Side; Air with Upper and Steel Arwt 3 Loer ad Pressure St borated water Boric Acid 37 Lower Head, boundary w-SS leakage Loss of Material Corrosion VD2-1 3.1.1-58 A Inlet and cladding (External)

Outlet Nozzle Primary Side; Steel Borated 38 Upper and Pressure w-Nickel reactor Cracking -TLAA IV.D23 3.1.1-10 A Lower boundary Alloy 690 coolant Faiue Tubesheet cladding (Internal)

Primary Side; Steel Borated 39 Upper and Pressure w-Nickel reactor Cracking -Flaw inservice C Lower boundary Alloy 690 coolant Growth Inspection IV.C2-26 3.1.1-62 0102 Tubesheet claddinq (Internal)

Primary Side: Steel Borated Cracking -40 Upper and Pressure w-Nickel reactor PWSCC, Inservice IVD2-4 3.1.1-35 A Lower boundary Alloy 690 coolant SCC/IGA Inspection Tubesheet claddinq (Internal)

Primary Side; Steel Borated Cracking -41 Upper and Pressure w-Nickel reactor PWSCC, Nickel-Alloy VD2-4 3.1.1-35 A Lower boundary Alloy 690 coolant SCC/IGA Management

-3 Tubesheet cladding .(Internal)

Enclosure L-14-206 Page 22 of 82 Table 3.1.2-4 Aging Management Review Results -Steam Generators Row Component Intended Aging Effect Aging NUREG-Row Type Intended Material Environment Requiring Management 1801, Table I Notes Management Program Volume 2 Item Item Primary Side; Steel Borated Cracking -42 Upper and Pressure w-Nickel reactor PWSCC,PWR Water IV.D2-4 3.1.1-35 A Lower boundary Alloy 690 coolant SCC/IGA Chemistr Tubesheet cladding (Internal)

Primary Side; Steel Borated 3 Upper and Pressure w-Nickel reactor Loss of Material PWR Water lV.C2-15 3.1.1-83 C Lower boundary Alloy 690 coolant Chemistry

-Tubesheet cladding (Internal)

Primary Side: Steel Air with 44 Upper and Pressure w-Nickel borated water Loss of Material Boric Acid IV.D2-1 3.1.1-58 A Lower boundary Alloy 690 leakage Corrosion Tubesheet cladding (External)

Primary Side; Steel 5 Upper and Pressure w-Nickel Treated water Los of Material One-Time IV.D2-8 3.1.1-12 C Lower boundary Alloy 690 (External)

Inspection I Tubesheet cladding Primary Side; Steel 46 Upper and Pressure w-Nickel Treated water Loss of Material PWR Water ID2-8 3.1.1-12 C Lower boundary Alloy 690 (External)

Chemistry

_Tubesheet cladding Primary Side; Borated Tube-to- Pressure Nickel reactor CracTLAA D2-15 3.1.1-06 C tubesheet boundary Alloy 690 coolant Fatigue 0101 Weld Enclosure L-14-206 Page 23 of 82 Table 3.1.2-4 Aging Management Review Results -Steam Generators Rw Co nt Intended Aging Effect Aging NUREG-Row Componen Intendd Material Environment Requiring Management 1801, Table 1 Notes No.I Type IFunction(s)

MaaeetPormVolume 2 Item Management Program Item Primary Side; Borated Crackin Itee 48 Tube-to- Pressure Nickel reactor PWSCCi -Water IV.D2-4 3.1.1-35 A tubesheet boundary Alloy 690 coolant SCC/IGA Chemistry 0101 Weld Primary Side; Borated Cracking -Steam Generator IV.D2-4 3.1.1-35 49 Tube-to- Pressure Nickel reactor PWSCC, Stuem Gentegrator 0 tubesheet boundary Alloy 690 coolant SCC/IGA Tube Integrity 0101 Weld Primary Side: Borated 50 Tube-to- Pressure Nickel reactor Loss of Material PWR Water IV. C2-15 3.1.1-83 C 50 tubesheet boundary Alloy 690 coolant Chemistry 0101 Weld Secondary Side; AFW Pressure Steel Treated water Cracking TLAA IV.D2-10 3.1.1-7 C Header, Riser, boundary (Internal)

Fatigue and Nozzle Secondary Side; AFW Pressure Treated water Cracking -Flaw Inservice 5 Header, Riser, boundary (Internal)

Growth Inspection N and Nozzle Secondary Side; AFW Pressure Steel Treated water Loss of Material One-Time IV.D2-8 3.1.1-12 C Header, Riser, boundary (Internal)

Inspection and Nozzle Enclosure L-1 4-206 Page 24 of 82 Table 3.1.2-4 Aging Management Review Results -Steam Generators RowComponent Intended MaterAging Effect Aging NUREG-No. Tpe Intend Material Environment Requiring Management 1 V 2 Notes Management Program volume 2 Item Item Secondary 4 Side: AF Pressure Treated water Loss of Material PWR Water IV.D2-8 3.1.1-12 C Header, Riser, boundary (Internal)

MChemis and Nozzle Secondary Air with Side: AFW Pressure Steel borated water Loss of Material Boric Acid IV.D2-1 3.11-58 A Header, Riser, boundary leakage Corrosion I and Nozzle (External)

Secondary Side; Shroud Treated water Cracking C 56 and Shroud Support Steel (Extednate Cacking TLAA IV.D2-10 3.1.1-7 0101 Support Ring (External)

F 0_0_and Lugs Secondary Side: Shroud Treated water Cracking -Flaw Inservice H 57 and Shroud SURiO Steel (External)

Growth Inspection N/A N/A 0101 Janport Ring and Lugs Secondary Side; Shroud Tretewte One-Time C 58 and Shroud Suppo Steel reated water Loss of Material One-time IV.D2-8 3.1.1-1 2 0101 58 adShrlOu ort Steel (External)

Inspectio 0101 Junport Ring and Lugs Secondary Side; Shroud Treaedate PWR Water C 59 and Shroud SupPOrt Steel Treated water Loss of Material PWRmWatr IV.D2-8 3.1.1-12 0101 Support Ring (External)

Chemistry 0_01 and Luqs Enclosure L-14-206 Page 25 of 82 Table 3.1.2-4 Aging Management Review Results -Steam Generators Row Component Intended Aging Effect Aging NUREG-No. Type Function(s)

Material Environment Requiring Management 1801, Table Notes Management Program Volumem2 Item___________Item Secondary 60 Side: Manway Pressure Treated water Cracking -TLAA ID2-10 3.1.1-7 A and Handhole boundary Steel (Internal)

Fatigue Cover Secondary Side: Manway Pressure Treated water Cracking -Flaw Inservice 6.1 and Handhole boundary Steel (Internal)

Growth Inspection N/A N/A H Cover Secondary Side: Manway Pressure Steel Treated water Loss of Material One-Time IV.D2-8 3.1.1-12 C and Handhole boundary (Internal)

Inspection Cover Secondary 63 Side: Manway Pressure Steel Treated water Loss of Material PWR Water VD2-8 3.1.1-12 C and Handhole boundary (Internal)

Chemistry Cover Secondary Air with 64 Side: Manway Pressure Steel borated water Loss of Material Boric Acid VD2-1 3.1.1-58 A and Handhole boundary leakage Corrosion I Cover (External)

Secondary Air with 65 Header Su:port Steel borated water Cracking -Flaw Inservice Support Plate leakage Growth Inspection N/and Gusset (External)

Secondary Air with Side; MFW 66 Header Suppor Steel borated water Cracking -TLAA IV.C2-10 3.1.1-07 A Support Plate leaka( e Fatigue an___d Gusset (External)

Enclosure L-1 4-206 Page 26 of 82 Table 3.1.2-4 Aging Management Review Results -Steam Generators Row Component Intended Aging Effect Aging NUREG-Ro opnn nedd Material Environment Reurn Mage nt 1801, Table 1 Notes No. Type Function(s) equiring Managemen Volume 2 Item Management Program Item Secondary Air with Side MFW borated water Boric Acid 67 Header Support Steel leakate Loss of Material Corrosion IV.D2-1 3.1.1-58 A Support Plate (ExternaCr and Gusset Secondary 68 Side: MFW Pressure Steel Treated water Crackin -TLAA IV.D2-10 3.1.1-7 C Header and boundary (Internal)

Fatigue Riser Secondary Treated water Cracking -Flaw Inservice 69 Side: MFW Pressure Steel TetdwerNIA NIA H Header and boundary (Internal)

Growth Inspection N Riser Secondary Tetdwater Flow-Accelerated 70 Side; MFW Pressure Steel Treated Loss of Material IV.D2-7 3.1.1-59 C-- Header and boundary (Internal)

Corrosion (FA C)Riser Secondary Tetdwater One-Time Z1 Side; MFW Pressure Steel Treated Loss of Material IV.D2-8 3.1.1-12 C-- Header and boundary (Internal)

Inspection Riser Secondary 72 Side; MFW Pressure Steel Treated water Loss of Material PWR Water IV.D2-8 3.1.1-12 C Header and boundary (Internal)

Chemistry

--Riser Secondary Air with 73 Side: MFW Pressure Steel borated water Loss of Material Boric Acid IV.D2-1 3.1.1-58 A Header and boundary leakage Corrosion

-Riser (External)

Enclosure L-1 4-206 Page 27 of 82 Table 3.1.2-4 Aging Management Review Results -Steam Generators Row Component Intended Aging Effect Aging NUREG-Ro Cmonn Itndd Material Environment Requring Mage nt 1801, Table 1 Notes No. Type Function(s)

Mn Poquig anagemen Volume 2 Item Management Program Item Secondary Side; MFW Nozzle, MFW Nozzle 75 Thermal Pressure Nickel Treated water Cracking -Flaw Ineric HC Sleeve and boundary Alloy 690 (Internal)

GothctN/ 0101 AFW Nozzle Thermal Sleeve Secondary Side; MFW Nozzle, MFW Nozzle 7z Thermal Pressure Nickel Treated water Cracking -Fla Inservice H Sleeve and boundary Alloy 690 (Internal)

Growth Inspection IA 31- 0101 AFW Nozzle Thermal Sleeve Secondary Side; MFW Nozzle, MFW 7._zzlTera Pressure Nickel Treated water Cracking -Inservice IVD- ..-4 C 76 lheevem an boundary Alloy 690 (Internal)

SCC/IGA Inspection IVD- ..-4 0101 AFW Nozzle Thermal 2_Leeve Enclosure L-14-206 Page 28 of 82 Table 3.1.2-4 Aging Management Review Results -Steam Generators Row Component nteded Aging Effect Aging NUREG-No. Type Intended Material Environment Requiring Management 1801, Table 1 Notes N. Type Function(s)

Maaemn Prga Volume 2 Item SManagement Program Item Secondary Side; MFW Nozzle, MFW Nozzle 7_7 Thermal Pressure Nickel Treated water Cracking -PWR Water IV.D2-9 3.1.1-84 C Sleeve and boundary Alloy 690 (Internal)

SCC/IGA Chemistry 0101 AFW Nozzle Thermal Sleeve Secondary Side; MFW Nozzle, MFW Nozzle ____ ___ _________A 78zThe Pressure Nickel Treated water Inservice 3 78 Sleevermand boundary Alloy 690 (Internal)

Loss of Material Inspection VIIIB-1 3.4.1-37 0101 AFW Nozzle Thermal Sleeve Secondary Side; MFW Nozzle, MFW Nozzle _________NozThe Pressure Nickel Treated water PWR Water C 79 ThSleevermand boundary Alloy 690 (Internal)

Loss of Material Chemistr VII-1 3.4.1-37 0101 AFW Nozzle Thermal Sleeve Enclosure L-14-206 Page 29 of 82 Table 3.1.2-4 Aging Management Review Results -Steam Generators Aging ffectAgingREG-Row Component Intended Material Environment Requiring Management 1801, Table 1 Notes No. Type Function(s)

M Management Program Volume 2 Item M a eP rItem Secondary Side; Nozzle 80 (steam outlet, Pressure Steel Treated water Cracking TLAA IV.D2-10 3.1.1-7 A level sensing, boundary (Internal)

Fatigue drain, and vent)Secondary Side; Nozzle (steam outlet, Pressure Steel Treated water Cracking -Flaw Inservice 81 (sea utet resue Ste N/A N/A H level sensing, boundary (Internal)

Growth Inspection

-drain, and_vent)Secondary Side: Nozzle 82 (steam outlet, Pressure Steel Treated water Loss of Material One-Time IV.D2-8 3.1.1-12 C level sensing, boundary (Internal)

Inspection drain, and vent)Secondary Side; Nozzle 83 (steam outlet, Pressure Steel Treated water Loss of Material PWRD2-8 3.1.1-12 C level sensing, boundary (Internal)

Chemistry

-drain, and vent)Secondary Side; Nozzle Air with 84 (steam outlet, Pressure Steel borated water Loss of Material Boric Acid IV.D2-1 3.1.1-58 A level sensing, boundary leakage Corrosion

-drain, and (External) vent) I I Enclosure L-14-206 Page 30 of 82 Table 3.1.2-4 Aging Management Review Results -Steam Generators Aging Effect Aging NUREG-Row Component Intended Material Environment Requiring Management 1801, Table 1 Notes No. Type Function(s)

MaaeetPormVolume 2 IItem Management Program Voteme2 Im Item Secondary Pressure Treated water Cracking -TLAA IV.D2-1O 3.1.1-7 C.Side; Shell boundary (Internal)

Fatgue 6 Secondar Pressure Treated water Cracking -Flaw Inservice 8__6 Side; Shell boundary (Internal)

Growth Inspection N H 87 Secondary Pressure Steel Treated water Loss of Material One-Time IV.D2-8 3.1.1-12 A.Side; Shell boundary (Internal)

Inspection 88 Secondary Pressure Steel Treated water Loss of Material PWR Water IV.D2-8 3.1.1-12 A ,Side; Shell boundary (Internal)

Chemistry

-Air with 89 Secondary Pressure Steel borated water Loss of Material Boric Acid ID2-1 3.1.1-58 A.Side; Shell boundary leakage Corrosion (External)

Secondary Treated water Cracking C 90 Side; Tube Support Steel TExternate Cacking TLAA IV.D2-10 3.1.1-7 0101 Supor Plte(External)

Fatigue 01 Support Plate Secondary Treated water Cracking -Flaw Inservice H 91 Side; Tube Suppoo Steel (External)

Growth Inspection N/A N/A 0101 Support Plate Secondary Treated water Lgamen PWR Water A 92 Side; Tube Support Steel TEatednate Ligaent PWRnWatr IV.D2-11 3.1.1-76 0101 Support Plate (External)

CrackinI Chemistry 0101 Enclosure L-1 4-206 Page 31 of 82 Table 3.1.2-4 Aging Management Review Results -Steam Generators Agin Effct AingNUREG-Row Component Intended Material Environment Aging Effect Aging 1801, Table 1 No. Type Function(s)

Management Program Volume 2 Item Maaeet PormItemL Secondary Side r Treated water LUgament Steam Generator A 93 Sup Suppo Steel (External)

Cracking Tube Integrity ID2-11 3.1.1-76 0101 Support Plate Secondary Treated water One-Time C 94 Side: Tube Support Steel Teatedate Loss of Material One-Time VD2-8 3.1.1-12 0101 SpotPae(External)

Inspe ction 0101 Sun~ortPlate

__SecondaryTrt PWR Water 95 Side: Tube Support Steel Treated water Loss of Material Chemistr IV.D2-8 3.1.1-12 C Support Plate (External)

Chemistry 0101 Secondary Side: Tube Treated water Cracking -TLAA ID210 311-7 C 96 Support Plate Support Steel (External)

Fatigue 0101 Spacer Secondary Side: Tube Steel Treated water Cracking -Flaw Inservice NIA NIA H 97 Support Plate Support (External)

Growth Inspection N 0101 Spacer Secondary Side: Tube Treated water One-Time IV.D2-8 3.1.1-12 C 98 Support Support Steel (External)

Loss of Material Inspection 01-301 I Spacer Secondary 9 Side Tube Support Steel Treated water Loss of Material PWR Water IV.D2-8 3.1.1-12 C.Support Plate (External)

Chemistry 01-301 Spacer Secondary Side; Tube Stainless Treated water Cracking-TLAA V11.E3-14 3.3.1-02 A 100 SupportRo Steel (External)

Fague 0101 S oie Rod0 Enclosure L-14-206 Page 32 of 82 Table 3.1.2-4 Aging Management Review Results -Steam Generators Row Component Intended Aging Effect Aging NUREG-Row Typonet Fntended Material Environment Requiring Management 1801, Table 1 Notes No. Type Function(s)

Management Program Volume 2 Item Item Secondary 101 Side; Tube SUPPO Stainless Treated water Crackin PWR Water CDI-14 3.1.1-74 0 Support Rod Steel (External)

Chemistry 0101 (Tie Rod)Secondary 102 Side; Tube Suppo Stainless Treated water Cracking Steam Generator IV.DI-14 3.1.1-74 C Support Rod Steel (External)

Tube Integrity 0101 (Tie Rod)_____Secondary Side: Tube Stainless Treated water PWR Water C 103 Support Support Steel (External)

Loss of Material Chemistry IV.DI-15 3.1.1-74 0101 (Tie Rod)Secondary Side; Tube Stainless Treated water Steam Generator I-15 3.1.1-74 C 104 Support Support Steel (External)

Loss of Material Tube Integrity 0101 (Tie Rod)Base Support Air with Stool and borated water Cracking -A 105 Base Support S Steel leakage Fatigue TLAA IV.C2-10 3.1.1-07 0101 Platform (External)

Base Support Air with Stool and borated water Cracking -Flaw Inservice H 106 Base Support Support Steel leakage Growth Inspection N/A N/A 0101 Platform (External)

Base Support Air with 107 Stool and Sno Steel borated water Loss of Material Boric Acid IVD21 3.1.1-58 A Base Support leakage Corrosion 0101 Platform (External)

Enclosure L-1 4-206 Page 33 of 82 Affected LRA Section LRA Page No. Affected Paragraph and Sentence 3.3.2.2.4.3 3.3-41 Entire section Based on the design and operation of the makeup pumps, LRA Section 3.3.2.2.4.3, "Stainless Steel PWR High Pressure Pump Casings -Treated Borated Water," is revised to read as follows: 3.3.2.2.4.3 Stainless Steel PWR High Pressure Pump Casings -Treated Borated Water Cracking due to stress corrosion cracking and cyclic loading could occur for the stainless steel pump casing for the PWR high-pressure pumps in the chemical and volume control system. At Davis-Besse, cracking due to stress corrosion cracking and cyclic loading is not identified as an aging effect requiring management for the stainless steel pump casing for the high-pressure pumps in the Makeup and Purification (chemical and volume control) System because the pumps have two casings, an internal casing and an external casing, one pump is in continuous operation, and the casings are is-exposed to treated borated water that is maintained at 120°F or below. Therefore, this item is not applicable to Davis-Besse. due to cy"lic loading of the ,stainless stee! pump .c --g of the high pressure pumps in the Makeup and Purifiagton System 144I1 bh manage by the P-14 Watr ca h ;emistry gram with the One Time inspection providin;verification Of the absence Of cracking due to cyclic 1ading.

Enclosure L-1 4-206 Page 34 of 82 Affected LRA Section LRA Page No.Affected Paragraph and Sentence Row 3.3.1-09, "Discussion" column Table 3.3.1 3.3-52 Based on the design and operation of the makeup pumps, LRA Table 3.3.1, "Summary of Aging Management Programs for Auxiliary Systems Evaluated in Chapter VII of NUREG-1 801 ," is revised to read as follows: Table 3.3.1 Summary of Aging Management Programs for Auxiliary Systems Evaluated in Chapter VII of NUREG-1801 e AFurther Item Component/Commodity Aging Aging Management Evaluation Discussion Number Effect/Mechanism Programs Recommended 3.3.1-09 Stainless steel high-pressure Cracking due to Water Chemistry and a Yes, plant Not applicable pump casing in PWR chemical stress corrosion plant-specific specific Consistent with NUREG 101!.and volume control system cracking and cyclic verification program.loading The AMP is to be Cracking due to SCC and cyclic augmented by verifying loadinQ is not identified as an the absence of cracking aging effect requiring due to stress corrosion management for the stainless cracking and cyclic steel high-pressure pump casings loading. A plant specific in the Makeup and Purification aging management (chemical and volume control)program is to be System.evaluated.

CR~p~9r~ng duo t9 cyclic !ading of the stai, , I sto pFn~p Gaging ot the high proSSUro PUMPS in the Makeup and PUificotion Systor 14i11 -bi Managed by Meo P14 Watoer Chemgistry P4rogram wt the9 09ne TiMo inSpoction Enclosure L-14-206 Page 35 of 82 Table 3.3.1 Summary of Aging Management Programs for Auxiliary Systems Evaluated in Chapter VII of NUREG-1801 I Further Item Aging Aging Management Evaluation Discussion Number Component/Commodity Effect/Mechanism Programs Recommended Ar-vid-ing vorificationq of the-a- -r,v v.s -tu ov leading Further evaluation is documented in Section 3.3.2.2.4.3.

Enclosure L-1 4-206 Page 36 of 82 Affected LRA Section LRA Page No.Affected Paraqraph and Sentence Table 3.3.2-18 3.3-397 Two Rows Based on the design and operation of the makeup pumps, two rows are deleted from Table 3.3.2-18, "Aging Management Review Results -Makeup and Purification System," to read as follows: Table 3.3.2-18 Aging Management Review Results -Makeup and Purification System R Aging Effect Aging NUREG-Row Component Intended Material Environment Requiring Management 1801, Table 1 Notes No.Management Program Volume Item 2 Item Pump GtasJe Theated-Make P4;essure Stagco! 6- beFated WateF GCaGkinOeTieVE!i0 Pkimp-Ga~in_-Makeup P-FeSSUe sae Th33te K4d44.Pup ( ... ,4 .... SteeF ak .............

....... A D4T44- ("In c, na, Enclosure L-14-206 Page 37 of 82 Affected LRA Section Table 3.3.2 Plant-Specific Notes LRA Page No.Affected Paragraph and Sentence 3.3-549 Plant Specific Note 0336 Based on the design and operation of the makeup pumps, Plant Specific Note No. 0336 from Table 3.3.2"Plant-Specific Notes" is no longer applicable and is revised to read "Not used", as follows:

Enclosure L-14-206 Page 38 of 82 Affected LRA Section LRA Page No.Affected Paragraph and Sentence Table 4.1-1 4.1-3 4 Rows Based on the installation of the replacement steam generators, LRA Table 4.1-1,"Time-Limited Aging Analyses," previously revised by FENOC letter dated June 3, 2011 (ML11159A132), is revised to delete 4 rows as follows: Table 4.1-1 Time-Limited Aging Analyses Results of TLAA Evaluation by Category 54.21(c)(1)

LRA Paragraph Section 0OTSGs t'ube s4eoves 4-i.3.2.2.6.2 0OTSGs AFW rnodific~tisn

{-ii-i,,.32...

OTSGs t,-bss and tube stabilizers floW ind-ucod vi-braton ) 4.3.2.2.6.4 OTSG 1 2 flaw evaluationns 4.7.5.

Enclosure L-1 4-206 Page 39 of 82 Affected LRA Section LRA Page No. Affected Paragraph and Sentence 4.3.1.2 4.3-2 3 rd Paragraph (Transient 32)Based on the installation of the replacement steam generators, LRA Section 4.3.1.2 "Projected Cycles," Transient 32 discussion, is deleted in its entirety as follows: Tianisent 32 E-ach r-emote welded plug installed in the onco through steam4 gonerato;(OT-SG~s) is limited to 33 cycles of heatup and coo/down.

The 60 year- cycle for- ome of theise plugs exeeds the design cyle number.Davis Besse monitors thoise cycleis with the Fatigue Moenitoring Program a nd wX! ensure action (either- a reana4ysis of rocord or a plant modifiation) is taken beqfor the design number- of cycleis is reached. Because thee plugs may be r-eanalyed for- other reasonis, Davis Besse will manage fatiue of these plugs Afo the period of extended operation irather- than reanalyze for- the possible additional cycles at this time.

Enclosure L-1 4-206 Page 40 of 82 Affected LRA Section LRA Page No.Affected ParaaraDh and Sentence Table 4.3-1 4.3-6 Transient Nos. 30A, 30B & 32 Based on the installation of the replacement steam generators, Transients 30A, 30B and 32 of LRA Table 4.3-1 "Projected Cycles," previously revised by FENOC letter dated June 3, 2011 (ML11159A132), revised to read "Not used", as follows: Table 4.3-1 60-Year Projected Cycles Program Accrued 60-year Transient Transient Cycles To Projection Cycles Notes# 2/19/2008 Cycles 30 A Not used. Ing 387 076 A Reartor Canlat System hoatup and ,ooldewn is one ..ansio.i;Auilar FedaerBote Nzze14 .y,49J and befting/unbelting of the nozzles is one transient cycle 30 B Not used. 224. 442 075 for T4ansients 30A and 3B.Auiliary Foedwater-Belted Nozzle-1 2 32 Not used. 4-7-5 64 43 The fimiting plug (rem..te welded plug 2A) was i..sta!!d on OTSG Wded Rug (lmitng pug is R. mete ...3/2.03. A Rop-a..t.

r Coo.lant System heatup and tcoldw ih Welded Rlug 2A 7-0 69) (Hoatup/Coeldown) netransient cycle. Tr-an-si@nt 322 isP projected to exceed the numgber- of design cycGles prior- to the end of the period ef extended operation.

Davis Besse manag9s faiue of these9 plug& using the Fatigue Monitoring Pr~ogram-.

Enclosure L-1 4-206 Page 41 of 82 Affected LRA Section LRA Page No. Affected Paragraph and Sentence 4.3.2.2.6 4.3-11 & 12 Entire Section Based on the installation of the replacement steam generators, LRA Section 4.3.2.2.6, "Once Through Steam Generators (OTSGs)," is replaced in its entirety to read as follows: 4.3.2.2.6 Once Through Steam Generators (OTSGs)The steam generator (replacement steam generator) is a vertical, straight-tube-and-shell heat exchanger that produces superheated steam at approximately a constant pressure over the power range. Reactor coolant water enters the steam generator at the upper primary head, flows down the Nickel Alloy 690 TT tubes while transferring heat to the secondary shell-side fluid, and leaves through the lower primary head. Steam is generated on the shell side.The shell, the outside of the tubes, and the tubesheets form the boundaries of the steam-producing section of the vessel Within the shell, the tube bundle is surrounded by a shroud which separates the feedwater inlet (lower annulus between the shell and the shroud) and steam outlet (upper annulus between the shell and the shroud) from the boiling (tube) region. Tube supports hold the tubes in a uniform pattern along their length. Vents, drains, instrumentation nozzles, and access ports (manways, handholes, and inspection openings) are provided on the shell side of the unit.The steam generator fatique analysis is addressed in the subsection below.

Enclosure L- 14-206 Page 42 of 82 Affected LRA Section LRA Page No. Affected Paragraph and Sentence 4.3.2.2.6.1 4.3-12 1 st Paragraph Based on the installation of the replacement steam generators, the 1 st paragraph of LRA Section 4.3.2.2.6.1, "OTSGs Fatigue," is revised to read as follows: 4.3.2.2.6.1 OTSGs Fatigue The primary (tube) and secondary (shell) sides of the once through steam generators are designed to ASME Section III, 1968 Edition through Summe: 1968 Addenda 2001 Edition with 2003 Addenda. The steam generators were analyzed for fatigue by the original equipment manufacturer.

The cumulative usage factors for the limiting primary and secondary side steam generators locations were calculated based on design transients, and are all less than 1.0.-4n addition, the isteam ugenerator Femote weld pluge, have a limited design life of 33 heatup Go ccleIt maintain a fatiue usage of less than 1.0. The number of occurrences of design transients is tracked by the Fatigue Monitoring Program to ensure that action is taken before the design cycles are reached. As such, the effects of aging due to fatigue are managed for the period of extended operation.

Enclosure L-14-206 Page 43 of 82 Affected LRA Section LRA Page No. Affected Para-graph and Sentence 4.3.2.2.6.2 4.3-12 & 13 Entire Section Based on the installation of the replacement steam generators, LRA Section 4.3.2.2.6.2, "OTSGs Tube Sleeves Fatigue," is no longer applicable, and the Section and the LRA Table of Contents (although not shown below) are revised to read "Not used" as follows: 4.3.2.2.6.2 Not used OTSGs Tube Sleeves Fatigue UJSAR Section 5.5.2.3 indicates that steam generator tubes that aro found to b leaking may be plugged Or repaired by m~echanical (rolled) sleeving.

SectionR I~the ASME Code does not provide deSign ruleS for mnechanically roll expando attachments, and theoretical stress analyses are inadequate.

In such cases-, AppeRdix II of ASMIIE Section III permits the use of experim nt stre-ss ---a i.to 66ubstantiate the critical or goeerning st~eros. The structural adequacy of the sleeve attachment to wiVthst-and cyclic leadings was demonstrated by a fatigue test peFr ASMIE Sect;io II1, 11 1500. The sleeve ladinRg transients for the fatigue test were based oR the design transients.

In the pressure cycling peodien of the fatigue test is based OR the Rnumber of sta~tUp cycles for a once through steam generator (360 cycles).Note that the steam generator tube sleeves were tested to 360 sta~tUp cycle t bound all BabcocGk & W~ilcox 177 fuel assembly plants. Davis Besse has only21 staUp cycles allowed Oi USAR Table 5.1 8, and only 128 projected sta1iup cycles in 60 years of operation perF Table 4.3 1. ConRsequently, Davis Besse wl not approach the tested numnber Of cycles for the once through steamR generato tube sleeves duFrig the period Of extended operation, and the T-LAA associate with fatigue testing of the tube sleev~es will rem~ain valid.Disposition:

10CF=R 54.21(c)(1)(i)

The fatigue testing of the once throgh steam generator tube sleeves will remnain v.alid for th peio of extended operation.

Enclosure L-1 4-206 Page 44 of 82 Affected LRA Section LRA Page No. Affected Paragraph and Sentence 4.3.2.2.6.3 4.3-13 & 14 Entire Section Based on the installation of the replacement steam generators, LRA Section 4.3.2.2.6.3, "OTSGs Auxiliary Feedwater Modification," is no longer applicable, and the Section and the LRA Table of Contents (although not shown below) are revised to read "Not used" as follows: 4.3.2.2.6.3 Not used OTSGs Auxi.iary,'

The r header-& iternal to the steamg generators we/re found damaged during the 1082 refueling outage. The repair- installed an exr*nal header On 9eGh team goneratorr, including snme rerouting of piping and , ppoGs. InRcludd in this repair was the evaluation of the eight new holes in the steam generators, the auxiary feedwator thermal sleeves, the iser flange atfachment to Mte shell (shell, thermal sleeve bearing area and studs), and flow induced vibration of the steam generator-tubes.The design of this 1082 modifiation hais beon included in the steam goneratoi stress analysis refer-enced in Section 4.3.2.2.6.1 above. Therefore the fatigue analyseis of the steam generfatr ishell pedrmed as parF of this mo;dification are included in the Ssteam generator-fatigu pr ousy discussed in Sectio 4 .2.2.6..The ana4yasa of the auxali-ry feedwater theFmel sleeve pro'vided a basis for demonstrating that the auxilar-y feedlwater thermal sleeve is capable of w~thstanding 300 cycles of auxiliary feedw-ater-inlection transients.

This analysi was peF-ormed in aGccordance with he requirements of the ASME Code foFr Clas' components.

The riser flange to the steam generator shel! weas also analyed per- ASWE= Code requirem~ents, and was acceptable for- a design life et 875 cycles of auxilir-y feodwator initiation.

Auxiliary feedw'ater-in.tiations, Transients 3a, and 30. in Table 4.3 1, are currently only at 1066.5 and -224.5 cycles Transients 30A andf 30B are to maximum of 389 an.d 442 cycles, respectively, through the period of extended operation.

These 60y ar proleGcton are below the L375 desigc ycleS for the iser fange attachmgent but exceed the 300 design cycles for the auxilaFry feedwater-therm4al sleeve. The num~ber Of occurr~ences Of design transients is tracked by the Fatigue Monitoring Proegramq to ensure that action isy taken befor-e the design cycles are-reachedl.

As such, the effects of agn det fatigue are managed for- the period of extended operation.-

Enclosure L-14-206 Page 45 of 82 Flow induced vibration of the steam generator-tubes with the new feedwater header desI was It was that the stres and deflection With the Rxternal headers was significantly less than the stress and deflection wit th origin4al internal header-s; consequently flow induced vibration was not reanalyed for- this modifiation.

Section 4.3.2.2.6.4 belowL, discu4ssejs the flowA induced vibration analyses of the9 steam generator-tubes Dispoaition:n 160 CFR 54.21 (c)(1)(ii The effects of fatigue en the auxiliary feedwater header- mnodifation will be managed for- the period of extended operation by the Fatigue MonitoringU P4wgtaM.

Enclosure L-1 4-206 Page 46 of 82 Affected LRA Section LRA Page No. Affected Paragraph and Sentence 4.3.2.2.6.4 4.3-14 Entire Section Based on the installation of the replacement steam generators, LRA Section 4.3.2.2.6.4, "OTSGs Tubes and Tube Stabilizers Flow Induced Vibration," is no longer applicable, and the Section and the LRA Table of Contents (although not shown below) are revised to read "Not used" as follows: 4.3.2.2.6.4 Not used OTSGs Tubes and Tube Stabifizer Fls !.w induced Flow induced vibration Of the once thr-ough steam generator tubes has been analyed iseveral timges over- the life of the Davis Besse plant. Th4e latest fo induced .vbration shows that the highest cumulative usage facto f o .any existing tube configuration is 0. 443 for- an up rFepaired tube next to the open lane.Adding 20 years of pperation to this tube incr-eases the cumul4ative usage factor by a factor Of 1. 5 to a 60-year- value of 0. 665, which reman accetable

(ý< 1.0).The cumulative usage factor for the 3/8 inch tube stablieFrs is clcRulated Using both high cycle (flow inducted W'bration) and low cycle (transients) fatigue. As the cumulative usage factor-s are only, 0.12 for the tube to istabilzer-weld and 0. 07 the nail, the flow inducd i n in Of these cumulative usage factos c be by 1.5 60 years&, and the umW "ive usage a-rtors wl K remai belo -4.19.Dispo~sitioen:

10 CFrR 54.21(0)(4)Ni The TLAA assc-iated with the flow indu cod vibration of the steam generator-tubeis and tube stabilzer has been pnroected through the perod Of exteRnded operatiOn.

Enclosure L-1 4-206 Page 47 of 82 Affected LRA Section LRA Page No. Affected Paragraph and Sentence 4.3.2.3 4.3-14 New Paragraph Based on the installation of the replacement steam generators, LRA Section 4.3.2.3, "Class 1 Piping and Valves," is revised to include a new paragraph as follows: 4.3.2.3 Class I Piping and Valves The Davis-Besse reactor coolant system piping, as well as reactor coolant pressure boundary piping in other systems, was designed to American National Standards Institute (ANSI) B31.7 Draft, February 1968 with Errata, June 1968 and also meets the design requirements of ANSI B31.7, 1969 Edition. The B31.7 Piping Code requires evaluation of transient thermal and mechanical load cycles and determination of fatigue usage for Class 1 piping. The reactor head vent and other piping designated as quality group A, B, or C is designed to ASME Section III, 1971 Edition, Class 1, 2 or 3 respectively.

Only quality group D piping is designed to ANSI B31.1. Davis-Besse has no Class 1 piping designed to B31.1.A portion of the reactor coolant system hot leq piping was replaced in support of steam generator replacement in the spring of 2014. Applicable ASME Code of Construction for the replaced hot leq piping isSection III, 2001 Edition with 2003 Addenda.

Enclosure L-1 4-206 Page 48 of 82 Affected LRA Section LRA Page No. Affected Paragraph and Sentence 4.7.5.2 4.7-6 Entire Section Based on the installation of the replacement steam generators, LRA Section 4.7.5.2, "OTSG 1-2 Flaw Evaluations," previously revised by FENOC letter dated June 3, 2011 (ML11159A132), is no longer applicable, and the Section and the LRA Table of Contents (although not shown below) are revised to read "Not used" as follows: 4.7.5.2 Not used OTSG 1 FRaw Evaluati.n.

During the c/yle 5 refuelng outage (May 988) a num.ber. of figw indications were detected in steam gonerator-1 2, both in the shoe# near- the isteamq outlet nozzle and in'the shel welds near te lower- tbesheet-tos he!juncture.

Two of the indications in the shell near- the stoam outlet nozzle were evaluated according to ASME2 Section AL, wt the remaining shell indicationis bounded by those evaluated.

Five of the hndications in the shell welds near the lower tubesheet to shell junctur-e were evaluated-, with th she# weld indications bounded by those ev-a lutd.Simplified evaluation of fatiue cr~ack gro'.w.h, basedl on 240 heatup and cooldwnA cycles, concluded that there would be only slght crack growth, andl the indications were found to be acceptable by ASME- Section AL, !W-9 3612 standlar-ds.

Because these analyses are based on a specifi number Of cycles, they are TLAAs. As ishown in LR Table 4.3 1 , the 60 year- prolected cycles for- heatup and cGoolow are 1 28 and ar bounded by the analyzed numgber of 240. Therefore, the s%7team grTT--eneator flaw growth analyses ill rwenain valid through the period of exte orIo'n.DiSPosition:

10 CF-R 54.21(c,(1)6q The steam4 generator flaw growt analyse-s will4 remnain valid through the periodf of extended operation.-

Enclosure L-1 4-206 Page 49 of 82 Affected LRA Section LRA Page No. Affected Paragraph and Sentence A.1.38 A-24 & 25 Entire Section Based on the installation of the replacement steam generators, LRA Section B.2.38, "Steam Generator Tube Integrity Program," previously revised by FENOC letter dated January 13, 2012 (ML12018A338), is revised as follows: A.1.38 STEAM GENERATOR TUBE INTEGRITY PROGRAM The Steam Generator Tube Integrity Program is credited for aging management of cracking, denting, loss of material, and reduction in heat transfer of the steam generator tubes, as well as cracking of the tube plugs, tube sleeves, and tube support plates.The Steam Generator Tube Integrity Program is a combination condition monitoring and mitigation program. The Steam Generator Tube Integrity Program is based on the Steam Generator Management program, which meets the intent of the guidance in Nuclear Energy Institute (NEI) 97-06, "Steam Generator Program Guidelines." and the requirements of the Technical Specifications.

The Steam Generator Tube Integrity Program also includes secondary-side examinations to assist in verification of tube integrity and the condition of the tube support plates. The program establishes a framework for prevention, inspection, evaluation,a removal from service (plugged) and leakage monitoring measures.In addition, cr-acking due to P14430 is managed for the steam generator tube to tubesheet welds (Alloy 600) by a of the P144P Water Chemis Program and the Steamg Generator Tube Inegit Program. The PWR Wato;Chemgistry Program ontrols peak levl of vtamnants (e.g., doved oxygen, chlorides, flurides, and sulfates below the system specifi limits thai can acc9eleate cracking for nickel alloy components.

The Steam Genorator-Tb Integrity Proegram includeis gross visual inspgection of the steam _generator-tub to tubeshoet welds coupled W""h edd curen insgpection (iebobbin coil 0: rotating coil examinations) of the tue to Imon.itor for cracking and degradation tt the tube to tubesheet weldts (Alloy 600). The gro sis visual ispection of the tube to tubesheet welds are scheduled concu~rrent w~qth e-ddy1 cu~rrent AnspeGtin of the isteam generator-tubeis that are scýheduled in accordance with Davis Bessea Technical Specification 5.5.9. At a m4ium 00% Of the tubes are inspecteda sequential periods of 60 effective full power monthis and therefore, at a mnmm 100% Of the tube to tUbesheet WAe~odS (inclUdeS both the hot leg and cold log Enclosure L-14-206 Page 50 of 82 Welg are insnected at eqefial Peid of 6fl effe~tiveWpoe m,,.,#h The gross 4 isalinpection Of the tuba to tubesheet weldS consiSts of a remo9te WiSu examgination using a m~anipulatr-camqera to obtain a straight-on view of the weld with a visual acuity su4fficient to detect eVidence of degradation.

The gross Visual examinations are peorfbmed by personnel whoe are qualifiedf for- American Societ of MechaniGal Engineers (ASME) code visual examination , .e., ae VT- .or VT- 3 examiner-s) and are knowledgeable in the type of tube-to tube&ho w..elds being eexmined (i.e., fillet welds). A.ceptance citeria for- the gross .sua.Prinsectsi and the eddy cUrient inspectons consist Of no andfoation of cratiag Or relevant eondieonfs ofnegradation.

Should the steamd genertors be replaced in the futuro with a design such that the tubes, tubesheet cladding and tube-to tubeshee wlsaefbrica ted of A#G 690 m~ateria, eniy the i-14.'r 144'vaze-Chem~istry Pro gram wil manage cracking due to 0144 C of the tuIbe toe tubesheet welds and the gross visual inspection wil no longer- be required.Primary-side and secondary-side water chemistry control and foreign material exclusion requirements inhibit degradation.

Eddy current testing and visual inspections are used for the detection of flaws. Condition monitoring compares the inspection results against performance criteria, and an operational assessment ensures that the performance criteria will be met throughout the next operating cycle.

Enclosure L-1 4-206 Page 51 of 82 Affected LRA Section LRA Page No. Affected Paragraph and Sentence A.2.3.1.5 A-37 Entire Section Based on the installation of the replacement steam generators, LRA Section A.2.3.1.5, "Steam Generator Remote Welded Plugs," is no longer applicable, and the Section and the LRA Table of Contents (although not shown below) are revised to read "Not used," as follows: A. 2 31.5 NVot isoodS~aw Gsqrn q a woola W9/dd-Pu E-ach remoete wclded plug installed in the once through steam4 generators is limited to 33 cycles of heatup and cooldown.

The 60 year- cycle proection A3~some of these plugs exceeds the analyed numqber of cycleqs. The num~ber oe occurrences of design transients is tracked by the Fatigue Monitoring Proqgram4 to ensure acio is taken before the dosin cyles are reached. As th ,ffects of aging due to fatigue are managed for the period of extended operation.

The effects of fatigue on Mhe steam4 generator remoete welded plugs K4I be m~anaged by the Fatigue Monitoring Program Afo the pferiod of extended operation in accordance with 10 CFR 54.2 11-c)(1)ii.

Enclosure L-1 4-206 Page 52 of 82 Affected LRA Section LRA Page No. Affected Paragraph and Sentence A.2.3.2.6 A-39 & 40 Entire Section Based on the installation of the replacement steam generators, LRA Section A.2.3.2.6, "Steam Generator Tube Sleeves Fatigue," is no longer applicable, and the Section and the LRA Table of Contents (although not shown below) are revised to read "Not used" as follows: A.2.3.2.6 Not used Steam Gene.rto T'ube Sleeves Fatigue USAR Section 5.5.2.3 indicateg that steamg generater-tubes that are found to b the AS4E= Code does not provide design a~sfFm~aicaly roll ex-panded attachments, and theoretical istress analyseis are inadequate.

In SUch cases, Appendix 4I of ASME Secion Wl permits the use of ex.permental atresis analysi6 to substantite the cr~itial or go .nn -tress. The Structur-al adequacy of the sloeeye attachment to .wIthsan clicoadings was dfemonstrated by a fatigue test with the sleeve loading transients based on the design transients.

Th-e pressure cycling pog~on of the fatigue test for- the steam generator-tube sloe eýis based on 360 sta~lbp cycles to bound all Babcock & WilcoX 177 fuel assem~bl plants. Davis Bgeisse has only 240 stadtup cycles allowed in USAR Table 5.1 8-, and only 128 ata.tup cycles prolected for 60 years of operation.

The fatigue testing of the once throeugh steam generator tube sleeves r-emi valid for the period of extended operation in accor-dance wt 10 CF-R 54.2 1(p)(4)0i).V Enclosure L-14-206 Page 53 of 82 Affected LRA Section LRA Page No. Affected Paragraph and Sentence A.2.3.2.7 A-40 Entire Section Based on the installation of the replacement steam generators, LRA Section A.2.3.2.7, "Auxiliary Feedwater Header Modification," previously revised by FENOC letter dated June 17, 2011 (MLI 11172A389), is no longer applicable, and the Section and the LRA Table of Contents (although not shown below) are revised to read "Not used" as follows: A.2.3.2. 7 Not used Auxi!iaFy Feedwat'r Headcr ModificMaton The Forigia auxilary feedwater- (AF=W) heador-& internal to the isteam genera ter were foundf damaged during the 1982 refueling outage. The r-epair was to insal an exter-nal headoer on each isteam generator-, including some r-eouting of pipiny the 6team generator-stress repodt and include~d in the steam generator faqtigue The auxiliar-y fe ed water thermal sleeve str-esses were also analyed according to the ASME= Code for- Class 1 com4ponents.

The analysis provided a basis fe;demonstrating that the AFW.V thermal sleeve is capable of ',thistanding 40,-000 cycles of auxilir-y feedwator-injection transients-.

in addition, the riser flange attachrnent (auxilar-y feedlwater nozzle flange) to the isteamg generator shel was analzed per- ASME Cede r e .Howeverit was necessar-y to limit the design ilfe to 87pycles Mheatupl/coolown-, boltup/unbolt and AFW initiatien).

Flew induced vibration of the steam generator:

tubes with the new feedwate header- desg wa rviewedf.

it wais conGcluedI that the stress and deflection w ith the etern-alhead/ers was significantly less than the stress andf deflectien With the origial internal headers;-, consequently flow induced vibratieon was not reanalyzed for this m-odification.

Section A. 2.3.2.9 discusses the flow nUe ba analyses of the steamg generator-tubes.The heatuplceoldo wn, boltup,'unbelt and AF=W intiation transients are projected to a m~aximgum of 442 cycles through the period of extended operation.

This iss less than the 7-5 cycles analyed f6o- the riser lange. However-, the nubero occu~rrences of design transients is tracked by the Fatigue Monitoring ProGgram t enasFe that actien is taken before the des ce reahed. As such, the effects of aging due to atigue are mn age f the period of extended operation.

Enclosure L-1 4-206 Page 54 of 82 The effects of fatigue on the auxiliary feedwater-header- modification " be managed by the Fatigue Monitoring Program for the period of extended operation in accor-dance with 10 CFR 54.2 1(cftl)iii%

Enclosure L-14-206 Page 55 of 82 Affected LRA Section LRA Page No. Affected Paragraph and Sentence A.2.3.2.8 A-39 & 40 Entire Section Based on the installation of the replacement steam generators, LRA Section A.2.3.2.8, "Steam Generator Tubes and Tube Stabilizers Flow Induced Vibration," is no longer applicable, and the Section and the LRA Table of Contents (although not shown below) are revised to read "Not used" as follows: A.2.3.2.8 Not used Steam Generator Tubes and Tube Stab!iizeFs Fo'w Induced Vibration ROW n-,U,. d .. bat. o of the steam tubes has been ana-yzed fo. 40 years of operation.

The for an u. repairod tubo has been projected to r-emain below 1.0 qfo 60 years of operation in accordan-e w44ith 10 CE-R The C;UP for- the 3M8 inch tube stabilzers&

is calculated using both high cycle (flow indUced ibration) and low cycle (transients) fatiue. Th4e CUF-a for- the tube stabiliers have boon projected to remain below 1.0 qfo 60 years of operation-.

The analyses associated wit the effects Of flow indIuced vibration on thesteam generator-tubes and tube stabilizr-s have been projected to the end of tepro of extended operation in accordance With 1 0 CF-R 54.2 1 qc,)(1)(4i).

Enclosure L-14-206 Page 56 of 82 Affected LRA Section LRA Page No. Affected Paragraph and Sentence A.2.3.2.10 A-41 Ist and 4 th Sentences Based on the installation of the replacement steam generators, LRA Section A.2.3.2.10, "Once Through Steam Generator," previously revised by FENOC letter dated June 17, 2011 (ML11172A389), is revised to read as follows: A.2.3.2.10 Once Through Steam Generator The primary (tube) and secondary (shell) sides of the once through steam generators are designed to ASME Section III, 1968 through Summe;1968 Addonda 2001 Edition with 2003 Addenda. The steam generators were analyzed for fatigue by the original equipment manufacturer.

The cumulative usage factors for the limiting primary and secondary side steam generators locations were calculated based on design transients, and are all less than 1.0. 4P addition, the steam generator remote weiod plugis have a limited design life of 33

.y.le. to matain a fatgue usage ,f ,ess than 1. 0. The number of occurrences of design transients is tracked by the Fatigue Monitoring Program to ensure that action is taken before the design cycles are reached. As such, the effects of aging due to fatigue are managed for the period of extended operation in accordance with 10 CFR 54.21 (c)(1)(iii).

Enclosure L-1 4-206 Page 57 of 82 Affected LRA Section LRA Page No. Affected Paragraph and Sentence A.2.3.2.11 A-41 New 2 nd Paragraph Based on the installation of the replacement steam generators, LRA Section A.2.3.2.1 1, "Class 1 Piping," previously revised by FENOC letter dated June 17, 2011 (ML11172A389), is revised to include a new 2 nd paragraph, to read as follows: A.2.3.2.11 Class 1 Piping The Davis-Besse reactor coolant system piping, as well as reactor coolant pressure boundary piping in other systems, was designed to American National Standards Institute (ANSI) B31.7 Draft, February 1968 with Errata, June 1968 and also meets the design requirements of ANSI B31.7, 1969 Edition. The B31.7 Piping Code requires evaluation of transient thermal and mechanical load cycles and determination of fatigue usage for Class 1 piping. The reactor head vent and other piping designated as quality group A, B, or C is designed to ASME Section III, 1971 Edition, Class 1, 2 or 3 respectively.

Only quality group D piping is designed to ANSI B31.1.A portion of the reactor coolant system hot leq piping was replaced in support of steam generator replacement in the spring of 2014. Applicable ASME Code of Construction for the replaced hot leq piping isSection III, 2001 Edition with 2003 Addenda.The cumulative usage factors for the Class 1 piping were analyzed based on the design transients, and are all less than 1.0. The number of occurrences of design transients is tracked by the Fatigue Monitoring Program to ensure that action is taken before the design cycles are reached. As such, the effects of aging due to fatigue are managed for the period of extended operation in accordance with 10 CFR 54.21(c)(1)(iii).

Enclosure L-1 4-206 Page 58 of 82 Affected LRA Section LRA Page No. Affected Paragraph and Sentence A.2.6.2 A-46 Entire Section Based on the installation of the replacement steam generators, LRA Section A.2.6.2, "OTSG 1-2 Flaw Evaluations," previously revised by FENOC letter dated June 3, 2011 (ML11159A132), is no longer applicable, and the Section and the LRA Table of Contents (although not shown below) are revised to read "Not used" as follows: A.2.6.2 Not used OTSG 1-2 Flaw E"'aluati.n.

DUring the c;ycle 5 refueling outage (year- 1988) a num~ber- Of flaW indicatin were detected in Steam generato 1 2ý, both in the she#l near the steam4 outlel nozze and n the she,# weds near the lower tubesheet to ,hell juncture.

Two e the indicateios in the shell near the isteamg outlet nozzle were evaluatedf accordiqng to ASME. Sectn X, with the e ghe. indi.ations boun evaluated.

Five of the indc#ations

'in'1 theP shell Welds near- the lower tubesheet to-shelljuncture were evaluate-d, With the rem~akning shell weld indications bounde by those evaluatedl.

Sim9plfied evaluation of fatigue crFack groewth, based on 240 heatup and coolden cycles concluded that there would be only slight crack growth, and the hndications were found to be acceptable by ASME- Section A, !WB 3612 standards.

Becaus-e these analyses are based on a specifi number- of cýycles, the~y are TIM&s The 60-year projected cycles for- heatup and cooldown are 128 and are bounded by the analyed Pnumber- of 240. Therefore, the analyses of the steam generatoi flaw growth wil remain valid through the period of extended operation in accordance with 10 CF=R 54.21(c)(!)N.

Enclosure L-1 4-206 Page 59 of 82 Affected LRA Section LRA Page No.Affected ParaaraDh and Sentence Table A-1 A-63 Commitment 13 Based on the design and operation of the makeup pumps, license renewal future Commitment 13 in LRA Table A-i, "Davis-Besse License Renewal Commitments," is revised to read as follows: Table A-1 Davis-Besse License Renewal Commitments e IRelated LRA Number Commitment Scp enion Source Section No./Comments 13 Implement the One-Time Inspection as described in LRA Section Prior to LRA A.1.30 B.2.30. Enhance the One Time !nepoction to: October 22, 2016 and 1.2.30 inludo cnhancd visuai (EVT- 1 or- equivalent) or- sdaae and examination (ma gnctic paf#clc, l#quid ponctrant), or- velurngtriG (RT or UT) ;nepectien; to detest and ,haact.Fr;e .a ,king- due FENOC Responses to to cycl loading of thc stainless stoel makeup pump .cn. Letters NRC RAIs (DB P37 1 and 2) of the Makeup and Pur-,.ic-a Stem. The L-1 1-153, 3.3.2.2.4.3-1 one time inspections wil. pr.e.de verif aion of the -O L- 11-166, from cracking duo to cycic loadng. L-11-218, NRC Letter L-11-237, dated L-11-252, May 2, 2011, aL11 Supplemental L-13-160 Question -and Makeup Pump L-14-206 Capium Casing Inspections 1

Enclosure L-1 4-206 Page 60 of 82 Enclosure L-14-206 Page 61 of 82 Affected LRA Section LRA Page No.Affected Paraaraoh and Sentence Table A-1 A-69 Commitment No. 25 Based on the installation of the replacement steam generators, LRA Table A-1, "Davis-Besse License Renewal Commitments," license renewal future Commitment No. 25, previously revised by FENOC letter dated January 13, 2012 (ML12018A338), is no longer required, and is revised to read "Not used", as follows: Table A-1 Davis-Besse License Renewal Commitments 1 Related LRA Item Implementation Source Section No.1 Number Commitment Schedule comments Comments 25 Not used.Enhanco the Stoam Genorator-Tube into grity Pr;ogram to Pier4e t RA A.4.3 Includo gross vi'al hnpection of the steam generate Apri 22, 204-7 47 .2.tu ...to ..bo.sheot welds .oupl.d with eddy .urr.nt #n...... FENOC Repense to e. ., bobbin cOl o~r rotating coil eaxaination6) of the tubes to LteNRGRA Mo en.ito~r for cracking and degradation of the tube to tuboshoci4200 welds (Alloy 600). Schodulo the gross visual 4nspotion of he g tube to tub'sho"t welds with eddy currnt ....... insoction of the steamg gonorator-tubos that are schedulod in accordanc ee with Davis Besso Technical Specific-ation 15.55. 9 such that 10-0%2-dt of the tube to tubosheet wolds 64ncludo both the ho lo an 2044mbr-2--

cold leg welds) are insected at sequential periods- 244 60 affective full power- mngAthsG.

PePdor the gross visual insection of the tube to tubeshoot welds thr-ough ramota visua usingamnipulator camerIa to obtain a straight on We of the% we~ldA ýA.lfth 1.--11a aG;~, ty ,91Ac., .;,-,tO f d flt t-&,lrý ltr Enclosure L-14-206 Page 62 of 82 Table A-1 Davis-Besse License Renewal Commitments Item Im Related LRA Nmer Commitment Implementation Source Section No./Number Schedule Cmet Comments of do gr-adatio~n.

P-erfo-Rm the gross Wsalr k~pcctibns uing por-eennel who arc gualkfied for Amoprir-pn Sa;icty Of Mechanical E.gine.. (AS6 E) cSdo : v'i'salexaminat:on

(ý.e., GeAi VT- 1 OF VT- 3 examginers) and arc knowledgeablo in thc type et tubo to tubesheet weld& being examined 1. e., fiPet welds).D_6fine tho acceptanco c:r.toria for th gro.s isa inpections.

and the eddy .....nt ..etins as no indication of cracking 0;rob vant condition of do gradation.

____________

Enclosure L-1 4-206 Page 63 of 82 Affected LRA Section LRA Page No.Affected Paragraph and Sentence Table A-1 A-69 Commitment 33 Based on the completion of Phase 1 -Action 1 and Phase 2 -Action 1 activities during the Cycle 18 refueling outage (Spring 2014), license renewal future Commitment 33 in LRA Table A-i, "Davis-Besse License Renewal Commitments," is revised to read as follows: Table A-1 Davis-Besse License Renewal Commitments Related LRA Item Commitment Implementation Source Section No./Number Schedule Comments 33 Phase 1 Phase 1: FENOC Responses to Perform the following actions to reduce or mitigate the refueling Action 1 Letters NRC RAls canal leaks inside containment:

COMPLETE L-11-252, B.2.39-9 from 1. Select and implement a leak detection method to locate the #9#e-4e and NRC Letter leakage area. Decembor 31, L-13-160 dated 2-044 and July 27, 2011, 2. Evaluate temporary and permanent repair methods to stop L-14-206 and or significantly reduce the leakage, and implement a Action 2 and repair plan. COMPLETE A.1-1 from NRC Letter Q/tber-t-2,dated 6March 26, 2013[continued]

Enclosure L-14-206 Page 64 of 82 Table A-1 Davis-Besse License Renewal Commitments Item Io Related LRA Iter Commitment Implemedion Source Section No./Comments Phase 2 Perform the following actions to evaluate the impact of refueling canal leaks on concrete and reinforcing steel structures.

Discontinue core bores, testing and reinforcing steel inspections when indications of refueling canal leakage are no longer present: 1. Perform a core bore in the south wall of the east-west section of the core flood pipe tunnel.a. Assess borated water degradation of the concrete by testing the core bore sample for compressive strength and by petrographic examination, and evaluate the results.b. Conduct a visual examination of the concrete and reinforcing steel to identify aging effects (e.g., concrete degradation or steel corrosion).

Enter identified aging effects into the FENOC Corrective Action Program and evaluate in accordance with the requirements of the current licensing basis Maintenance Rule Program.2. If leakage from the refueling canal has not been eliminated or resumes by the beginning of the period of extended operation, then evaluate the concrete structures in a manner similar to the way that they were evaluated under Phase 2, Action 1. However, use acceptance criteria from the American Concrete Institute (ACI) Report 349.3R for Phase 2: Action 1 COMPLETE Doccembcw31 2044 Action 2 prior to December 31, 2023 Enclosure L-1 4-206 Page 65 of 82 Table A-1 Davis-Besse License Renewal Commitments Item Implementation jRelated LRA Number Commitment Schedule Source Section No./Comments the evaluation.

3. If leakage from the refueling canal has not been eliminated Action 3 -or resumes during the period of extended operation, then Ongoing evaluate the concrete structures again in a manner similar to the way that they were evaluated under Phase 2, Action 2.Perform evaluations every ten years until the end of the period of extended operation.

Enclosure L-14-206 Page 66 of 82 Affected LRA Section LRA Page No.Affected Paragraph and Sentence Table A-1 A-69 Commitment 35 Based on the completion of Phase 1 activities during the Cycle 18 refueling outage (Spring 2014), license renewal future Commitment 35 in LRA Table A-1, "Davis-Besse License Renewal Commitments," is revised to read as follows: Table A-1 Davis-Besse License Renewal Commitments e IRelated LRA Item Commitment Implementation Source Section No./Comments 35 Perform the following actions for each of two examinations Phase 1 FENOC Response to (Phase 1 and Phase 2) of the Containment Vessel in the sand COMPLETE Letters NRC RAI pocket region: -P,4,F ,t L-11-252 B.2.22-5 from* Perform nondestructive examination (NDE) of the 204-3 and NRC Letter Containment Vessel from the outer surface at five areas of Juy1 2011 previously-identified groundwater in-leakage.

July 21,2011 o Examine the vessel at a minimum of three vertical grid and locations at 12 inches nominal horizontal spacing at each area. Examine the Containment Vessel at a minimum of three elevations:

Phase 2 prior to December 31, 1. approximately 3 inches below the existing grout-to-2025 vessel interface in the sand pocket region;2. at the existing grout-to-vessel interface level in the sand Enclosure L-14-206 Page 67 of 82 Table A-1 Davis-Besse License Renewal Commitments e IRelated LRA Item Commitment Implementation Source Section No./Number Schedule Cmet Comments pocket region; and, 3. approximately 3 inches above the existing grout-to-vessel interface in the sand pocket region." Compare the ultrasonic test (UT) thickness readings to minimum ASME Code vessel thickness requirements and to the results obtained during previous UT examinations of the Containment Vessel. Determine the need for maintenance or repair of the Containment Vessel based on the results and evaluation of the examinations.

Document the results of each of the two examinations in the work order system. Document and evaluate adverse conditions in accordance with the FENOC Corrective Action Program for an evaluation of potential degradation of the steel Containment Vessel thickness over the longer term.

Enclosure L-14-206 Page 68 of 82 Affected LRA Section LRA Page No.Affected Paraqraph and Sentence Table A-1 A-69 Commitment 39 Based on the completion of Phase 1 activities during the Cycle 18 refueling outage (Spring 2014), license renewal future Commitment 39 in LRA Table A-i, "Davis-Besse License Renewal Commitments," is revised to read as follows: Table A-1 Davis-Besse License Renewal Commitments Related LRA Item Commitment Implementation Source Section No./Number Schedule Cmet Comments 39 Address the potential for borated water degradation of the steel Phase 1 FENOC Responses to containment vessel through the following actions: COMPLETE Letters NRC RAIs Access the inside surface of the embedded steel containment at L3 L-1153, B.2.22-2 from a vertical height no greater than 10 inches above bottom dead 2-4 a13 dated center. A core bore will be completed by the end of 2014 L-13-180 April 5,d2011 (Phase 1). If necessary, a second core bore will be completed and and B 2.2fr by the end of 2020 (Phase 2). If there is evidence of the Phase 2 prior to L-14-206 NRC Letter presence of borated water in contact with the steel containment December 31, dated vessel, conduct non-destructive testing (NDT) to determine what 2020 July 2, dated effect, if any, the borated water has had on the steel July 27, 2011 containment vessel. Based on the results of NDT, perform a and study to determine the effect through the period of extended Enclosure L-1 4-206 Page 69 of 82 Table A-1 Davis-Besse License Renewal Commitments Item Imp1ementation Related LRA Number Commitment ImSchedule Source Section No./Comments operation of any identified loss of thickness in the steel Supplemental containment due to exposure to borated water. RAI B.2.22-6 from NRC telecon held on May 9, 2013 Enclosure L-1 4-206 Page 70 of 82 Affected LRA Section LRA Page No.Affected Paragraph and Sentence Table B-2 B-22 1 Row Based on the installation of the replacement steam generators, the "Steam Generator Tube Integrity Program, Section B.2.38" row of LRA Table B-2,"Consistency of Davis-Besse Aging Management Programs with NUREG-1 801," previously revised by FENOC letter dated October 21, 2011 (ML11298A097), is revised to read as follows: Table B-2 Consistency of Davis-Besse Aging Management Programs with NUREG-1801 (continued)

Consistent Consistent wt New I with NUEG Plant- Enhancement Existing NUREG- 1801 with Specific Required 1801 181wt Exceptions Steam Generator Tube Integrity Program Existing Yes -Yes--Section B.2.38 Enclosure L-1 4-206 Page 71 of 82 Affected LRA Section LRA Page No. Affected Paragraph and Sentence B.2.30 B-121 "Scope" subsection of the"Enhancements" section, last paragraph Based on the design and operation of the makeup pumps, the last paragraph of the "Scope" subsection of the "Enhancements" section of LRA Section B.2.30,"One-Time Inspection," previously revised by FENOC letter dated September 16, 2011 (ML11264A059), is deleted as follows: The One Te InSpection W a19l include enhanced visual (EVT 1 0F equivaieno or- sur-ac examnaton (m~agnetic pa~icle, liquid penetrano, O vlume ,tric T or UT) enhanced visual (VT-1 or equivalen) an , or volumetric (PT- or UT) h4&poction8 to doet9ct andf GharacteriZe cracking due- to cyclic loading of the stainless steel m~akeup PUMP casings (DB P37- 1 and 2) of the M~akeu~p and P-urFcation System. The9 ono time in&pectons K4~l provid verification of the absence of cracking due to cyclic loading.Affected LRA Section LRA Page No. Affected Paragraph and Sentence B.2.30 B-1 22 "Scope" subsection of the "Aging Management Program Elements" section, last paragraph Based on the design and operation of the makeup pumps, the last paragraph of the "Scope" subsection of the "Aging Management Program Elements" section of LRA Section B.2.30, "One-Time Inspection," previously revised by FENOC letter dated September 16, 2011 (ML11264A059), is deleted as follows: The Ono TiE Inp&ection wil also include enhanced visual (E=VT- I oF equivaleno or sur-fac exaintin (m~agno tic pa~ice, liquid penetrant4,--or volumetri (RT- or UIT) inspections enhanced visual (VT- e r- equivalent) and4or volumetri (PT- or- UT) inspections to dfetect and ch9aracterize cracking due to cycl#c loading of the stainless steel makeup PUMP casings (D9 P37 1 and 2) ot the Makeup and Purification System. The one time inspectbios wil provide verification of the absence of cracking due to cyclic loading.

Enclosure L-14-206 Page 72 of 82 Affected LRA Section LRA Page No. Affected Paragraph and Sentence B.2.38 B-151 -153 Entire Section Based on the installation of the replacement steam generators, LRA Section B.2.38, "Steam Generator Tube Integrity Program," previously revised by FENOC letter dated January 13, 2012 (ML12018A338), is revised as follows: B.2.38 STEAM GENERATOR TUBE INTEGRITY PROGRAM Program Description The Steam Generator Tube Integrity Program is credited for aging management of cracking, denting, loss of material, and reduction in heat transfer of the steam generator tubes; as well as cracking of tube plugs, tube sleeves, and tube support plates. The Steam Generator Tube Integrity Program is performed as part of the overall Steam Generator Management Program. The Steam Generator Management Program is based on Technical Specification requirements, and is implemented in accordance with NEI 97-06, "Steam Generator Program Guidelines." The Steam Generator Tube Integrity Program also includes secondary-side examinations to assist in verification of tube integrity and the condition of the tube support plates.The Steam Generator Tube Integrity Program is a combination condition monitoring and mitigation program. The Steam Generator Tube Integrity Program manages the effects of aging through a combination of prevention, inspection, evaluation, r-epa r removal from service (pluyQQed) and leakage monitoring.

Preventative measures are intended to inhibit degradation and consist of primary-side and secondary-side water chemistry monitoring and control, and foreign material exclusion requirements.

The Steam Generator Tube Integrity Program provides the requirements for non-destructive examinations for the detection of flaws in tubes, and tube support plates. Degradation assessments identify both potential and existing degradation mechanisms.

Inservice inspections (i.e., eddy current testing and visual inspections) are used for the detection of flaws. Condition monitoring compares the inspection results against performance criteria, and an operational assessment provides a prediction of tube conditions to ensure that the performance criteria will be met throughout the next operating cycle. Primary-to-secondary leakage is continually monitored during operation.

Enclosure L-14-206 Page 73 of 82 In addition, cracking due to PWSCC wil be mganaged for the steamn genorat;tbetotubesheet weldos (Ally 600) by a ombinatin of the P144P Wato, Chemi~stry Program and the Steamg Generator Tu4be Integrity PRorm The PAR Watoer Chemistry Program eontrols peak levels of various oentamginants (e.g., dissolved oxygen, chlorides-, fluordes, and sulfates) below the system 6peciGi limits6 that can acc~eleate cracking for- nickel alloy comgponents.

The Steam GenratoFr Tube Integrity Progra-rm Will include gross visual insct4eion of the isteamg generator-tube-to tubesheet welds coupled with eddy current inspecto (i.e., bobbin cOil or- rotating coil examginations) of the tubes to monitor- for- cracking and degradation of the tube -to tubeisheet weldis (Alloy 600). The gros~s vis4 ýual inspection Of the0 tube to tUbes;hee t Welds wil be scheduled concurrent with eddy current inspection of the steamg generatoF tubes that are scheduled in accordance with Davis Besse Technical Specifiation

5. 5. 8. At a mn~imum, 1 00% of the tubes are nspetedf at sequential periods of 60 effective full power- months and threfoeat a minimum, 100% Of the tube to tubesheet Welds (includes bothth hot leg and cold leg wels) wl be ispected at sequential periods Of 60 effectiv l power- months. The g v inspection of the tube- to tubesheet w wel d nSsit of a m s examination using a manipulato Gamea to obtain a straight on view of the weold with a visual acu~ity sufficient to detect evidence or degradation.

The gross visual xaminations will be pe..med by peronnel are qualifed for American Society Of Mechanical Engineers (ASME) code visual examgination (i.e9., are ce.gifed VT- 1 or VT- 3 examiners) and are knowledgeable in the type of tube -to tubesheet welds being examined (i.e., filet welds%)Acceptance criteria for- the gross Visual inspectiens and the eddy curren inspections,14 wil consi A' R st Of nO indication of cracking or- relevant conditons or degradation.

Should the steam generators be replaced in the future with a design such that the ubes, tbeSheet cladding and ube tubesheet welds are fabricated of Alloy 690 matrial, the PWR WaterI Chemi,.ty Program ill manage cracking due to PWSC" of he tube-to t-besheet welds and the gr-eos visual wll no longer- be requred.NUREG-1801 Consistency The Steam Generator Tube Integrity Program is an existing Davis-Besse program that With will be is consistent with the 10 elements of an effective aging management program as described in NUREG-1801,Section XI.M 19, "Steam Generator Tube Integrity." Exceptions to NUREG-1801:

None.

Enclosure L-14-206 Page 74 of 82 Enhancements:

None.The..owin....g enha..e.tS K4#l be iMp.eMentd in the idet.ed program elernent6 prior to the period of extended operation.

Scopc, Parameters Mon itord or- Inspectedt, Detection of Aging EffectS,-Acceptance Critcria ThevLV/J vlLIJI Sta GnatoF Tube,.int,.4 v! -ow n, wX i4 l.de,,qv#

visual in ....innn Of thn .t .... n nnnatnr *,, t, W_÷,hnohnnt

....S Gn ...... s.ith c-sV GUrret mnpnnct.G,-n Ae. o MOM-hinp ,'uil os- rntateiss,-

c-,, 1 exaof t tubps to mnIiter for- cGh,-,,I in, anG(r .l..... fn-Apn n o f n .. inn ,Meh h wI4s IA gv 6l ).I The wc1eiss W814l in&R eGnnHRi of Aho

tn t nSn Wei& "t! be1. Sopchoc!, ed Gnn ,-on -pernt WAith, or-Is -d s ers-o*nt iR.eGQ Of thO 6tean n si nesn -ntc tibeso tha aFepro nc-M tied sinn anccW;-4P

-.Ann-n ia AitM flns #1 On p Teoonic-ni

.5.7. -..At a e -/V)00 of the tubes are in onocteo at p

orsentan Pe~pI oie& of R e fer -Ati. o e WelI olrn- s-f,,n*h p and thorn for-e at a n-inimm 4--ee-n 1f0% Of the tbe-4 eho*0se~hoght W/iAn#g O/114d-eero bOth the hot Iec- nad GGoi4Inn-160 -lA Wei& wXl be in octn#4 at pnieouenriti Pegg#4--of 60 nffoc-tis~

fs si! r nnan.-s-nths.

The c-smog.visualn in gnnntinn n thn tubeh to tue s~nhet WeInIS4 WX;1 c-nn pe~t Of a FQM-nntn u105 S 1 -etsnnton5sen rnnnpine latrs Ganrnor to obtain a strnirht-nnn tie', of thenel CAn!4 1 th a Wnal~ n nc-s its, pS effn.,t to nlotnct Os .qinnc- of clalnraptieon.

TI-e qrsroc: sWepe en Coc-mntV of h ,lG-hnica

-4I nninon -s- (ASM0A~ -n#4n Wsips en! 'nn-i~natnnie r c-n estfi# VsT- -o1r VsT- 3 nexanqinwnn-p annd i nrn IOOIAIO#4tbhPM the twen o, ti sh to tue eheehot welds#4 Mne4n -rnrs1n Ae., fi/t welds).g A c-c-n tn nc-c-rsterin for the nv-OPPS Ilige ! inonnc-tinr-ann# the ~-QddV Gu-e r9Ps-o6Gi si nno WXnglI c-nnc- lt Of no h~in #410 tOncf nGrnc-linQ tv orroIns tnt c-OR&A9P~

ngOf 9Qanrs- #nti~n Shoe ld the steam-s' oepnnomtorg be s-noiaG9G(n in tho e kitue esns i4h a riopiOnen s-h that thei tubelsp tue ehoght c-ln#ad~n annl tule e-oto-tubes~ohet wel4g a,%-fabric-nted of Ae 01 I mOh1e-nnts,n oni -l tho GII ~nn.Chenrntrst Dmronr -wXil o nc-sna c-rp-L-,rst-duee to- WSG of Meo tubeh to tuesenhnnt l/1-Q.g oand thes wes pp ,,p en inspesogtin WXl! no lonnnvr-M s-oe Fesuirn#Operating Experience During each refueling outage, SG degradation assessments are performed in accordance with the provisions of NEI 97-06 and the EPRI PWR SG examination guidelines.

These industry guidelines are based in part on operating experience and inspection results from other operating PWRs. Degradation assessment topics include SG tube degradation mechanisms, inspection

& expansion Enclosure L-14-206 Page 75 of 82 requirements, tube repair criteria, structural limits, guidelines for testing, and chemical cleaning provisions.

Dais Besso has ident.fied several Of tUbe degradation throeugh edn current eDavBmination.

Causes were eetermined td be mechaRigal equipment degfadatin, which is prmarngy a1; Tfeuation Of tme in operation, tempefratye of operation, and chemistry gendtiorn.

Additional causes were predicted to be prmard ater itess corrosion cracking, stress coion cracking ;SntergrAsemtsaak, denting, and outer diameteri stres acrrosion cracking.t Repairs were made through the Corrective Action Program.As a result ef the Cycie 15 Refueing Outage inspectins, 46 SG tubeis were peugged in pTSG 2 A, bringing the total for that SG to 625 tube&s plgged (4%).Thirty five SG tubes were plugged in OTSCG 1 b, bringing the tota! for that SG t 27-9 tubes plugged (1.8%). As with all previous inspections, the Gondition ofth SGsr (with the degraded tubes plugged) met indural y and egulatorty str-uctura and eakavge integrity guidance, and were expected to meet those criteria foeoisng outage insection.

Steam generator inspection resu are addroessed in the lnse Inspectbion summariy repedts that are submitted to AIRC following each outage.The Davis-Besse original steam -generators were replaced during the Cycle 18 refueling outage (Spring 2014). The station is currently in the first cycle of operation following steam generator replacement, and has no identified tube degradation mechanisms.

Self-Assessments of the program are performed periodically and conclude that it is being effectively implemented, meets FENOC expectations regarding engineering programs, meets current industry requirements and have incorporated many Industry identified beneficial practices.

Davis-Besse has not implemented the alternate repair criteria in GL 95-05, but has amended the technical specifications to be consistent with Technical Specification Task Force r-epot T-ST-F449 Traveler TSTF-510, Revision 2,"Revision to Steam Generator Pro-gram Inspection Freguencies and Tube Sample Selection." Also, this task force traveler includes provisions for selection of tube materials.

One of the tubing materials addressed is thermally treated Alloy 690. which is the tube material used in the Davis-Besse replacement steam.generators.

The Davis-Besse evaluation of IN 2008-07 concluded that the inspection scopes defined in the degradation assessments are appropriate for monitoring cracking in the expansion transition regions as well as at the upper and lower tube ends.

Enclosure L-1 4-206 Page 76 of 82 Using the accepted industry approach to testing and evaluation, and incorporation of pertinent industry operating experience, insures that the steam generator tube integrity program manages the effects of component aging such that the steam generators will continue to perform their intended functions, consistent with the current licensing basis, during the period of extended operation.

Conclusion The Steam Generator Tube Integrity Program has been demonstrated to be capable of managing age-related degradation of the steam generator tubes, tube plugs, htbe ,roe..es, and tube support plates. The Steam Generator Tube Integrity Program provides reasonable assurance that the aging effects will be managed such that components within the scope of this program will continue to perform their intended functions consistent with the current licensing basis for the period of extended operation.

Enclosure L-1 4-206 Page 77 of 82 Affected LRA Section LRA Page No. Affected Paragraph and Sentence Environmental 3.2-1 thru 3.2-5 Entire Section Report Section 3.2 Based on the installation of the replacement steam generators, LRA Appendix E,"Applicant's Environmental Report -Operating License Renewal Stage" (Environmental Report), Section 3.2, "Refurbishment," is revised to acknowledge that the refurbishment activity is complete, to read as follows: 3.2 REFURBISHMENT ACTIVITIES Regulatory Requirement:

10 CFR 51.53(c)(2);

51.53(c)(3)(ii)"The report must contain a description of ... the applicant's plans to modify the facility or its administrative control procedures as described in accordance with§ 54.21. This report must describe in detail the modifications directly affecting the environment or affecting plant effluents that affect the environment....""The environmental report must contain analyses of ... refurbishment activities, if any, associated with license renewal...." FENOC has addressed refurbishment activities in accordance with NRC regulations and complementary information in the NRC GElS for license renewal (NRC 1996). In particular, NRC requirements for the renewal of operating licenses for nuclear power plants include the preparation of an Integrated Plant Assessment (IPA) in accordance with 10 CFR 54.21. The IPA must identify and list systems, structures, and components subject to an aging management review. Items that are subject to aging and might require refurbishment include, for example, the reactor vessel piping, supports, and pump casings, as well as items that are not subject to periodic replacement.

In addition, the GElS (NRC 1996, Section 2.6) provides information on the scope and preparation of refurbishment activities to be evaluated in this environmental report. It describes major refurbishment activities that utilities might perform for license renewal that would necessitate changing administrative control procedures and modifying the facility.

The GElS analysis assumes that an applicant would begin any major refurbishment work shortly after NRC grants a renewed license and would complete the activities during five outages, including one major outage at the end of the 40th year of operation.

The GElS refers to this as the refurbishment period.

Enclosure L-1 4-206 Page 78 of 82 NRC regulations for implementing the National Environmental Policy Act require environmental reports to describe in detail and assess the environmental impacts of refurbishment activities such as planned modifications to systems, structures, and components or plant effluents

[10 CFR 51.53(c)(2)].

NRC regulations at 10 CFR Part 51 do not define "refurbishment," but the GElS provides some examples of refurbishment activities and explains that these are actions that typically take place only once in the life of a nuclear plant, if at all (NRC 1996, Section 2.6.2.6).

Relevant examples of possible refurbishment activities include replacing the turbine and turbine pedestal, steam generator, or reactor coolant system piping when these activities are carried out to ensure safe or more economic operations during the period of extended operations.

The GElS assumes, however, that refurbishment activities would take place during a"refurbishment period"; i.e., within the 10 years prior to current license expiration, over the course of numerous outages, and culminating in a major outage immediately prior to the extended (license renewal) term.FENOC replaced the reactor vessel head in 2011 and plans to ,,ep.aGe the two original steam generators in 2014. FENOC has-determined that the most cost-effective method for long-term management of the replaced reactor vessel head, steam generators, and other large irradiated plant equipment, is to store them on-site in a dedicated storage facility, and then disposition them along with the remaining plant equipment when Davis-Besse is decommissioned.

Therefore, a new permanent structure, the Old Steam Generator Storage Facility was constructed in 2011, which provides approximately 12, 000 square feet of space to house indefinitely the replaced (Midland) reactor vessel head, and lator houso the original steam generators and the Reactor Coolant System hot legs (see below). A permanent multi-story office building was also constructed in 2011 adjacent to the Auxiliary Building to house personnel that supported the replacement activities for the reactor vessel head and W4l-, sippe, ,th.for the steam generators.

The replacement of the reactor vessel head and steam -generators, and the construction of the twe-new permanent structures to support the head replacement project are being performed for and under the current facility operating license. Therefore, the associated environmental impacts are enveloped by the Final Environmental Statement for the current Davis-Besse operating license (NRC 1975).In 2014, FENOC plans t replaced the two original Davis-Besse once-through steam generators with new once-through steam generators, and-pkims-t replaced the Reactor Coolant System hot leg piping in conjunction with the replacement of the steam generators.

Replacement activities a.-_ ,."'- .e ÷" lasted approximately 70-90 days and are-were c.r..ntl planned to be conducted Enclosure L-14-206 Page 79 of 82 during a xtended the Cycle 18 refueling outage in the spring of 2014.FENOC considers the replacement activities associated with the steam generators and the hot leg piping to be license renewal refurbishment activities.

Therefore, the associated environmental impacts are assessed in this ER.Each of the once-through steam generators is a vertically-mounted, straight-tube and shell counter-flow heat exchanger that converts heat from the reactor coolant system into steam to drive the turbine generators and produce electricity.

The exisg-oriinal steam generators are each approximately 75 feet long, have a diameter of approximately 15 feet, and weigh approximately 590 tons. The replacement steam generators wll-be-are dimensionally equivalet -similar to the original steam generators, but weigh only approximately 465 tons each.The approximately 15,500 straight tubes in the original steam generators are 56 feet long and are made of Alloy 600 (inconel) material.

This alloy degrades over time as a result of a variety of corrosion and mechanical factors. Alloy 600 degradation affee~tsaffected both of the steam generators at Davis-Besse.

Accordingly, FENOC has dot,,rmine-that they should be replaced the original steam generators with steam generators that use Alloy 690 tubing material to minimize tube degradation due to Alloy 690's improved resistance to stress corrosion cracking.The replacement steam generators are-be'ig-were manufactured in Cambridge, Ontario, Canada by Babcock and Wilcox Canada, Ltd., and-w&l-be were transported to Davis-Besse.

The steam generators are planne dve-shipped separately, and transport i6 expeGtedinvolved the following methods of transportation and routes:* Rail transport from Cambridge, Ontario, to Port Hamilton, Ontario;Ship transport across Lake Ontario, through the Welland Canal, and across Lake Erie to the Port of Toledo; and,* Rail transport from the Port of Toledo to Davis-Besse.

Babcock and Wilcox Canada, Ltd., is-was responsible for the transportation and delivery of the steam generators to Davis-Besse, and woeld-ensured that all federal, state, and local requirements aFe-were met for associated transportation activities.

Physical modifications to the rail lines may-be-was necessary to transport the replacement steam generators.

After the replacement steam generators arrived at Davis-Besse, FENOC-plans transported the steam generators on a heavy-duty self-propelled modular transporter, and moved them to the 18,000 square foot New Steam Generator Enclosure L-1 4-206 Page 80 of 82 Storage Facility (included in the 80, 000 square feet of facilities described below)constructed at Davis-Besse in 2013.Site planning, construction of temporary facilities, modification of existing buildings, and other preparation activities a.. .plann4 4o'occurred at Davis-Besse prior to removal of the original steam generators from the Containment Vessel.Temporary facilities consisting of approximately 80,000 square feet-are plaped were erected or brouqht on site for additional offices, fabrication and assembly activities, mock-up activities, weld testing, decontamination, warehouse areas, and lay down areas. These temporary facilities consisting of tents and portable trailers would-used portions of existing Davis-Besse structures and facilities (e.g., permanent parking lot, dry cask storage pad), weu LLreqiuired construction of a concrete pad that may remain following the steam generator replacement project, or would-consisted of temporary structures that wouldbe-were completely removed following completion of the project. All temporary facilities were located within the developed industrial areas of the site on previously-disturbed land. and any.peF..n.ent Permanent concrete pads that remain following the replacement project are platned-te be-located within the developed industrial areas of the site on previously-disturbed land.FENOC estimates that the total area disturbed by permanent and temporary construction, decontamination, and laydown activities would be was less than 10 acres, all of which would-be-was on previously-disturbed property within the bounds of the Davis-Besse owner-controlled area. A load-haul path consisting of fill and gravel wou. lkel-be-was constructed for transporting the original steam generators to the permanent Storage Facility.

A minimal amount of fill soil May be-was temporarily required in certain locations along the on-site haul route to ensure the stability of the roads and transporter.

The small amount of disturbed area and implementation of best management practices in accordance with FENOC and site procedures (e.g., watering) would-minimized the amount of fugitive dust generated by refurbishment activities.

To perform the steam generator replacement, FENOC plans created a temporary construction opening approximately 23 feet wide by 31 feet high to-be ore-ated-in the Shield Building and free-standing Containment Vessel. The Shield Building is composed of reinforced concrete walls approximately two and one-half feet thick, and the free standing Containment Vessel is approximately 1.5 inches thick steel. The process of creating the opening would-included aetk46ies s6uh-as-removing concrete, cutting rebar, and cutting and removing a section of the steel Containment Vessel. A hydro-demolition (high pressure water) process or- other mechanical methods are being considered was used to remove the Shield Building concrete, and mechanical methods are being c;onsideed.

were Enclosure L-14-206 Page 81 of 82 used to cut the Containment Vessel opening. After installation of the new steam generators, the openings would bwere sealed and the Containment Vessel and Shield Building returned to their original configurations and integrity.

The two original steam generators would be were drained and cut-away from existing piping and supports.

Steel covers would-be-were seal-welded to the nozzles of main coolant, steam, and feedwater piping openings of the original steam generators to preclude the release of contamination and seal-off internal sections during removal, transport and storage. Loose was removed from the exterior of each original steam generator and a coating would-be-was applied to affix any remaining contamination.

The steam generators wouldwere then be-rigged-out of Containment through the temporary openings.After removal from Containment, the original steam generators-wauld-be were transported on a self-propelled modular transporter to the permanent Storage Facility.

The replacement steam generators wouid-be-were removed from temporary storage and moved by the self-propelled modular transporter to the vicinity of the Davis-Besse Containment, and rigged into place. Installation-woul included construction of supports, connection of piping, and testing of system integrity.

Construction activities would-ikely-resulted in noise levels (primarily from hydro demolition, if u&ed, or other- mechanical means of remoVal) greater than those associated with normal Davis-Besse operation.

Noise from construction activities, however, uld-bewas intermittent and temporary in nature, and would-decreased as the distance from the source- ipGease increased.

The peak period of activity weuld-lkeljy-occurred when the actual removal and replacement of the steam generators take-took place. FENOC,, that a......moy 900 Approximately 1400 additional workers would be were on-site to support the replacement of the steam generators.

Approximately4-300 900 additional temporary workers would-be-were on-site supporting the refueling outage as well, for a peak total of approximately 2,-2OO-2300 additional workers.FE:NOC anticipates that on site On-site storage of diesel fuel and various lubricating oils ay-be-was required during the y90-day steam generator replacement project. FENOC site and company environmental protection procedures (e.g., the Spill Prevention Control and Countermeasure (SPCC) Plan)will- e-were used to control the storage of fuel and oils. Non-hazardous waste generated during the steam generator replacement project and hydro-demolition concrete and demolition debris wbe-was disposed of in accordance with FENOC and site procedures.

Water used in the hydro-demolition process, and other temporary discharges will-be-were addressed in accordance with the Enclosure L-14-206 Page 82 of 82 requirements of the National Pollutant Discharge Elimination System (NPDES)permit.In advance of the steam generator replacement project, FENOC plan-s4e resolved relevant environmental permit requirements (e.g., Ohio Final General Permit for Storm Water Discharge) to ensure compliance.

No significant impacts to bodies of water, ecological resources, cultural resources or land use are a3R-Gipated-occurred in association with the steam generator replacement project because activities are-were Dlanned to be undertaken on previously-disturbed parcels of land, and fugitive dust generation and water run-off-wil-be was managed in accordance with FENOC and site procedures and best-management practices.

In addition, many of the facilities and activities wilbwere short-term and temporary in nature.

v t e Letter Sequence Other
TAC:ME4613, (Approved, Closed) TAC:ME4640, Steam Generator Tube Integrity (Approved, Closed)
Initiation Request, Request, Request Acceptance... Supplement, Supplement, Supplement Administration Questions 15 August 2011 Answers 24 June 2011 Results Approval, Approval, Approval, Approval, Approval, Approval, Approval, Approval, Approval Other: L-11-222, Ohio Department of Natural Resources Office of Coastal Management Concurrence with Federal Consistency Certification Related to the Review of License Renewal Application Environmental Report, L-12-244, Correction of Errors in the License Renewal Application Environmental Report Severe Accident Mitigation Alternatives Analysis, and License Renewal Application Amendment No. 29, L-14-206, License Renewal Application Amendment No. 50 - Annual Update, ML11193A093
MONTHYEAR2011-02-25 [Table View]

L-14-206, License Renewal Application Amendment No. 50 - Annual Update

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SUBJECT:

Attachment:

Enclosure:

System Description

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