L-11-317, Reply to Request for Additional Information for the Review of License Renewal Application (TAC No. ME4640) and Amendment No. 20

Reply to Request for Additional Information for the Review of License Renewal Application (TAC No. ME4640) and Amendment No. 20

ML11298A097
Person / Time
Site:	Davis Besse
Issue date:	10/21/2011
From:	Allen B FirstEnergy Nuclear Operating Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
L-11-317, TAC ME4640
Download: ML11298A097 (62)

Text

FENOC FEAIO 5501 North State Route 2 FirstEnergy Nuclear Operating Company Oak Harbor, Ohio 43449 Bany S. Allen 419-321-7676 Vice President - Nuclear Fax: 419-321-7582 October 21, 2011 L-11-317 10 CFR 54 ATTN: Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555-0001

SUBJECT:

Davis-Besse Nuclear Power Station, Unit No. 1 Docket No. 50-346, License Number NPF-3 Reply to Request for Additional Information for the Review of the Davis-Besse Nuclear Power Station, Unit No. 1, License Renewal Application (TAC No. ME4640) and License Renewal Application Amendment No. 20 By letter dated August 27, 2010 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML102450565), FirstEnergy Nuclear Operating Company (FENOC) submitted an application pursuant to Title 10 of the Code of Federal Regulations, Part 54 for renewal of Operating License NPF-3 for the Davis-Besse Nuclear Power Station, Unit No. 1 (DBNPS). By letter dated September 22, 2011 (ADAMS Accession No. ML11256A149), from Inspector questions during the Region III Inspection Procedure IP-71002 License Renewal Inspection follow-up held the week of August 22, 2011, and by telephone conference calls held on September 16 and October 5, 2011, the Nuclear Regulatory Commission (NRC) requested additional information to complete its review of the License Renewal Application (LRA).

The Attachment provides the FENOC reply to the NRC requests for additional information (RAIs). The NRC request is shown in bold text followed by the FENOC response. The Enclosure provides Amendment No. 20 to the DBNPS LRA.

A 14 S

Davis-Besse Nuclear Power Station, Unit No. 1 L-11-317 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 October ii',2011.

Sincerely, Barry S. Allen

Attachment:

Reply to Request for Additional Information for the Review of the Davis-Besse Nuclear Power Station, Unit No. 1 (DBNPS), License Renewal Application, Sections 2.3.3, 3.1.2.2, 3.3.2.2, 4.1 and B.2.39

Enclosure:

Amendment No. 20 to the DBNPS License Renewal Application cc:

NRC DLR Project Manager NRC Region III Administrator cc:

w/o Attachment or Enclosure NRC DLR Director NRR DORL Project Manager NRC Resident Inspector Utility Radiological Safety Board

Attachment L-11-317 Reply to Request for Additional Information for the Review of the Davis-Besse Nuclear Power Station, Unit No. 1 (DBNPS), License Renewal Application, Sections 2.3.3, 3.1.2.2, 3.3.2.2, 4.1 and B.2.39 Page 1 of 11 Section 3.1.2.2 Question RAI 3.1.2.2.16-1

Background:

By its letter dated August 17, 2011, the applicant addressed its review results on cracking due to primary water stress corrosion cracking (PWSCC) of steam generator nickel alloy tube-to-tubesheet welds in response to the discussion held in a teleconference call dated July 13, 2011.

In the letter, the applicant stated that upon further review after the conference call with the U.S. Nuclear Regulatory Commission (NRC), it determined that the tube-to-tubesheet welds (Alloy 600 welds) for its steam generators do not have a license renewal intended function and therefore, are not subject to an aging management review. The applicant also stated that the steam generators are Babcock & Wilcox Model 177-FA, once-through design and the tubes and the tubesheets of the steam generators form the pressure boundary between the fluid in the secondary system and the reactor coolant system. The applicant further stated that as provided in Updated Safety Analysis Report (USAR) Section 5.5.2.3, the tubes are expanded (to a partial depth) into the tubesheet and the tubes are seal welded to the tubesheet near the tube ends. In addition, the applicant stated that the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code, Section Xl, Division 1, 1995 Edition with 1996 Addenda, IWA-9000 defines a seal weld as a nonstructural weld intended to prevent leakage, where the strength is provided by a separate means. The applicant stated that the "separate means" in this case is the tube-to-tubesheet expansion joint which forms the pressure boundary and that the tube-to-tubesheet welds are seal welds and therefore, are not part of the pressure boundary.

Issue:

The applicant stated the tube-to-tubesheet welds (Alloy 600 welds) for its steam generators do not have a license renewal intended function and therefore, are not subject to an aging management review. The applicant also stated that the tube-to-tubesheet welds are seal welds and therefore, are not part of the pressure boundary. However, the staff noted that the reactor coolant pressure boundary should provide structural and leak-tight integrity. Furthermore, the applicant's statement that the tube-to-tubesheet welds are intended to prevent leakage indicates that these welds perform the intended function of the reactor coolant

Attachment L-11-317 Page 2 of 11 pressure boundary. Therefore, the staff found a need to confirm whether or not the design analysis of the applicant's once-through steam generators, which was used to establish the current licensing basis (CLB), concludes the following: the interference fits between the tubes and the tubesheets are sufficient to ensure the structural and leak-tight integrity of the tube-to-tubesheet joints, without a need for crediting the tube-to-tubesheet welds.

Request:

1)

Confirm whether or not the design analysis, which was used to establish the CLB, concludes that the interference fits are sufficient to ensure the structural and leak-tight integrity of the tube-to-tubesheet joints, without a need for crediting the tube-to-tubesheet welds.

If the design analysis concludes that the interference fits are sufficient to ensure the structural and leak-tight integrity, provide the technical basis of the conclusion and list the reference(s) addressing the technical basis.

2)

If the design analysis, which was used to establish the CLB, credits the tube-to-tubesheet welds for ensuring the structural and leak-tight integrity of the tube-to-tubesheet joints, describe how cracking due to PWSCC will be managed for the steam generator tube-to-tubesheet welds.

RESPONSE RAI 3.1.2.2.16-1 The response titled "Supplemental Response - steam generator aging management review tube-to-tubesheet weld" provided in FENOC letter dated August 17, 2011 (ML11231A966), is replaced in its entirety by the following response.

1)

Although the steam generator tube-to-tubesheet weld is classified as a seal weld, FENOC has confirmed that the design analyses used to establish the current licensing basis credits both the interference fit (between the tube and tubesheet) and the tube-to-tubesheet weld for structural and leak-tight integrity. Therefore, the tube-to-tubesheet welds have a License Renewal intended function of "Pressure boundary" and are subject to an aging management review.

LRA Table 3.1.2-4, "Aging Management Review Results - Steam Generators" is revised to include the aging management review results. In addition, LRA Table 2.3.1-4, "Steam Generator Components Subject to Aging Management Review," is revised to list the tube-to-tubesheet weld with an intended function of "Pressure boundary."

2)

Cracking due to PWSCC will be managed for the steam generator tube-to-tubesheet welds (Alloy 600) by a combination of the PWR Water Chemistry

Attachment L-11-317 Page 3 of 11 Program and the Steam Generator Tube Integrity Program. The PWR Water Chemistry Program controls peak levels of various contaminants (e.g., dissolved oxygen, chlorides, fluorides, and sulfates) below the system-specific limits that can accelerate cracking for nickel-alloy components. The Steam Generator Tube Integrity Program will be enhanced to include enhanced visual (EVT-1 or equivalent) examinations to monitor for cracking of the steam generator tube-to-tubesheet welds.

A review of Davis-Besse operating experience has not identified any instances of cracking of the steam generator tube-to-tubesheet welds (Alloy 600). Therefore, the weld inspection sample size will include 20 percent of the subject weld population or a maximum of 25, whichever is less. In this case the maximum of 25 applies since the weld population for the two steam generators is greater than 60,000. The sample size is consistent with other NUREG-1 801 programs where the inspection is designed to provide assurance that aging is not occurring. Welds included in the inspection sample will be scheduled for examination in each 10-year period that occurs during the period of extended operation. Should the steam generators be replaced in the future with a design such that the tube-to-tubesheet welds are fabricated of Alloy 690-TT material, the examinations will no longer be required.

LRA Section A.1.38, Section B.2.38, Table A-1 and Table B-2 are revised consistent with this response.

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

Table 2.3.3 Question RAI 2.3.3.18-4

Background:

In its response to RAI 2.3.3.18-3 dated August 17, 2011, the applicant provided the following information:

1)

The letdown coolers performed acceptably from initial startup in 1978 until 1991, when plant personnel detected contamination in the component cooling water (CCW) system, and replaced both letdown coolers in 1993.

Then, in 2009, plant personnel identified a small, active reactor coolant leak, and again replaced both letdown coolers in 2010.

Attachment L-11-317 Page 4 of 11

2)

A failure analysis had not been performed on the leaking letdown coolers to determine the specific leak location or to verify the failure mechanism because of high radiation dose rates associated with that effort.

SRP-LR Section A.1.2.3.4, "Detection of Aging Effects," states that nuclear power plants are licensed using the principles of redundancy, and diversity, and that degraded components reduce the reliability of the systems, challenge safety systems, and contribute to plant risk. The SRP-LR continues by stating that the effects of aging on a component should be managed to ensure its availability to perform its intended function(s) as designed when called upon, and notes that a program based solely on detecting component failure should not be considered as an effective aging management program for license renewal.

Issue:

Based on the information provided in this recent response, as well as the information provided in response to RAI 2.3.3.18-2 for the same issue, the staff did not consider that the applicant has provided sufficient bases to justify the replacement frequency of every seventh refueling outage (approximately 14 years) for the letdown coolers in the makeup and purification system.

The bases for the staff's position are as follows:

a)

The applicant established the replacement frequency based on a qualified life, which was empirically derived using two plant-specific data points of 13 and 16 years, after identifying reactor coolant leakage into the component cooling water system.

b)

The applicant has not determined the flaw location, performed flaw sizing, or verified flaw characteristics to allow prediction of flaw stability or growth rate. Without having this information, operation of the letdown cooler with ongoing leakage is risking a failure, which would challenge the pressure relief capability of the component cooling water system and the isolation function of the valves in the makeup and purification system.

c)

While past operating experience (although limited) may have shown that the flaw was stable for some period of time, the replacement frequency determination did not appear to consider normal operational pressure transients that the letdown coolers would be expected to experience.

d)

The letdown cooler replacement frequency appears to be based on overall calendar time and not actual operational time, considering both refueling and extended outages.

Attachment L-11-317 Page 5 of 11 Request:

Provide a letdown cooler replacement frequency that includes adequate margin to initiation of tube leakage and provide the basis for the margin, or propose an aging management program that will adequately manage these components that are within the scope of license renewal.

RESPONSE RAI 2.3.3.18-4 The responses to RAI 3.3.2.2.4-1 and RAI 2.3.3.18-2 submitted in FENOC letter dated June 3, 2011 (ML11159A132), and the response to RAI 2.3.3.18-3 submitted in FENOC letter dated August 17, 2011 (ML11231A966), are revised as described below.

The letdown coolers (DB-E25-1 and 2) and the seal return coolers (DB-E26-1 & 2) in the Makeup and Purification System consist of stainless steel heat exchanger components exposed to treated borated water greater than 600C (> 1400F). Cracking due to stress corrosion cracking (SCC) in stainless steel heat exchanger components that are exposed to treated borated water greater than 600C (>1 400F) is managed by the Pressurized Water Reactor (PWR) Water Chemistry Program. The PWR Water Chemistry Program manages cracking through periodic monitoring and control of contaminants. One-Time Inspection will provide verification of the effectiveness of the PWR Water Chemistry Program to manage cracking.

The coolers are in continuous service and not subject to cyclic loading, eddy current testing of tubes for managing cyclic loading is therefore not applicable. The temperature and radioactivity monitoring of shell side water is performed by installed instrumentation.

FENOC withdraws license renewal future Commitment 25 of LRA Table A-I.

LRA Section 2.3.3.18, Table 2.3.3-18, Section 3.3.2.1.18, Section 3.3.2.2.4.1, Table 3.3.1, Table 3.3.2-18 and Table A-1 are revised consistent with this response.

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

Attachment L-11-317 Page 6 of 11 Table 3.3.2.2 Question RAI 3.3.2.2.10.4-1

Background:

SRP-LR Table 3.3-1, item 26 states that loss of material due to pitting and crevice corrosion could occur for copper alloy piping, piping components, and piping elements exposed to lubricating oil. The SRP-LR recommends GALL AMP XI.M39, "Lubricating Oil Analysis," to manage the aging effect and further evaluation of a program to verify the effectiveness of the Lubricating Oil Analysis Program, such as XI.M32, "One-Time Inspection," because control of contaminants within the lubricating oil may not have always been adequate to preclude corrosion.

In LRA Tables 3.3.2-14, 3.3.2-18, 3.3.2-30, 3.4.2-1, and 3.4.2-4, the applicant referenced LRA Table 3.3.1, item 3.3.1-26 and generic note I for copper alloy components exposed to lubricating oil and stated that the components have no aging effects requiring management. For these items, the applicant further cited plant-specific notes which state that the components are made of copper alloy with less than 15 percent zinc and are not in contact with a more cathodic metal; therefore, the components have no aging effects requiring management.

Issue:

It is unclear to the staff why the applicant claims that components of copper alloy with less than 15 percent zinc exposed to lubricating oil have no aging effects requiring management. The staff noted that components of copper alloy with less than 15 percent zinc are less susceptible to loss of material than other copper alloys, but that the presence of contaminants (e.g., water) in lubricating oil can create an environment conducive to loss of material, regardless of whether or not the component is in contact with a more cathodic metal.

Request:

Explain why components of copper alloy with less than 15 percent zinc exposed to lubricating oil have no aging effects requiring management or provide an appropriate AMP to manage loss of material.

RESPONSE RAI 3.3.2.2.10.4-1 The LRA is revised to identify loss of material due to pitting and crevice corrosion as an aging effect requiring management for copper alloy components with less than 15 percent zinc exposed to lubricating oil. The Lubricating Oil Analysis Program will be used to manage the aging effect of loss of material. The One-Time Inspection will

Attachment L-11-317 Page 7 of 11 provide verification of the effectiveness of the Lubricating Oil Analysis program to manage loss of material.

LRA Section 3.3.2.2.10.4, Tables 3.3.1, 3.3.2-1, 3.3.2-14, 3.3.2-18, and 3.3.2-30, Table 3.3.2 Plant-Specific Notes, Section 3.4.2.2.7.3, Tables 3.4.1, 3.4.2-1 and 3.4.2-4, and Table 3.4.2 Plant-Specific Notes are revised consistent with this response.

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

Section 4.1 Supplemental Question RAI 4.1-2 The NRC initiated a telephone conference call with FENOC on September 16, 2011, to discuss the FENOC response to RAI 4.1-2 submitted by FENOC letter dated August 17, 2011 (ML11231A966). In the response, FENOC stated that the fracture toughness of the cast austenitic stainless steel is not time-dependent as the analysis used a lower bound fracture toughness of 139 ksi'lin that bounds the saturated fracture toughness of the Davis-Besse material. The NRC staff's concern is that the applicant's basis may be predicated on charpy or thermal aging data that are not up-to-date or conservative when compared to the most recent data for the state of the industry.

It is not clear to the staff whether the assumption that "the lower bound fracture toughness of 139 ksi'lin that bounds the saturated fracture toughness of the applicant's materials" remains valid.

To address the NRC staff's concern that the applicant's basis may be predicated on charpy or thermal aging data that are not up-to-date or conservative when compared to the most recent data for the state of the industry, FENOC agreed to compare the thermal aging data used in the ASME Code Case N-481 Evaluation to the most-recent industry data (i.e., NUREG/CR-4513, Rev. 1, "Estimation of Fracture Toughness of Cast Stainless Steels During Thermal Aging in LWR Systems," and NUREG/CR-6428, "Effects of Thermal Aging on Fracture Toughness and Charpy-Impact Strength of Stainless Steel Pipe Welds"), and provide the results in a supplemental response to RAI 4.1-2.

Attachment L-11-317 Page 8 of 11 SUPPLEMENTAL RESPONSE RAI 4.1-2 The fracture toughness of the cast austenitic stainless steel is not time dependent as the Davis-Besse ASME Code Case N-481 evaluation [LRA Reference 4.8-18 (SIR-99-040, Revision 1)] used a lower bound fracture toughness value of 139 ksi'/in that bounds the saturation fracture toughness of the Davis-Besse material.

The saturation fracture toughness was determined using the methodology outlined in NUREG/CP-01 19, Vol. 2, pp. 151-178, "Proceedings of the U.S. Nuclear Regulatory Commission, 1 9 th Water Reactor Safety Information Meeting held at Bethesda, MD, October 28-30, 1991," and considering all available certified material test reports (CMTRs) for the base material and welds of the Davis-Besse reactor coolant pump casings. The saturation fracture toughness value of 139 ksi'lin was the minimum calculated for all the CMTRs considered in the evaluation. This minimum saturation fracture toughness value has since been calculated using NUREG/CR-4513, Revision 1, "Estimation of Fracture Toughness of Cast Stainless Steels During Thermal Aging in LWR Systems," the most recent publication on this subject. Using the methodology and correlation in this latest NUREG results in the same minimum saturation fracture toughness value for the pump casings.

The fracture toughness for welds considering thermal aging has also been presented in NUREG/CR-6428, "Effects of Thermal Aging on Fracture Toughness and Charpy-Impact Strength of Stainless Steel Pipe Welds." A conservative Jic fracture toughness value of 40 KJ/m2 based on the absolute minimum of all available data is provided in this document for aged stainless steel welds; this Jic fracture toughness value translates to 80 ksi'/in. This conservative fracture toughness value still bounds the calculated total applied stress intensity factors calculated in Table 4-5 of SIR-99-040, Revision 1, indicating that the conclusions of SIR-99-040, Revision 1, are unchanged even if the methodology outlined in NUREG-CR-6428, "Effects of Thermal Aging on Fracture Toughness and Charpy-Impact Strength of Stainless Steel Pipe Welds," is used for the Davis-Besse pump casing welds.

New LRA Sections 4.7.6 and A.2.7.5, previously submitted by FENOC letter dated August 17, 2011 (ML11231A966), are revised consistent with the above response.

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

Attachment L-11-317 Page 9 of 11 Section B.2.39 Supplemental Question RAI B.2.39-9 The NRC staff initiated a telephone conference with FENOC on October 5, 2011, to discuss the FENOC response to RAI B.2.39-9 submitted in FENOC letter dated September 16, 2011 (ML11264A059), regarding operating experience with borated water leakage from the reactor Refueling Canal. The NRC staff requested additional information about three subjects:

1.

Details of the FENOC response to the year 2003 report, "Engineering Assessment Report - Refueling Canal Leakage," recommendations.

2.

The structural integrity of concrete in containment affected by the borated water leakage, including why it is acceptable to wait until year 2014 to again verify structural integrity of the concrete.

3.

The rate of the borated water leakage when the Refueling Canal is filled.

SUPPLEMENTAL RESPONSE RAI B.2.39-9

1.

In the year 2003, FENOC initiated two Condition Reports for evaluation of the recommendations included in the 2003 assessment report, "Engineering Assessment Report - Refueling Canal Leakage." For the Refueling Canal leakage, the initial focus of the Condition Report corrective actions was to continue the use of non-destructive examination methods for identification of the leak locations and to repair the leaks. After the corrective actions were assigned, evaluation of industry operating experience had shown that non-destructive examination methods such as vacuum box testing and liquid penetrant examinations were marginally effective. Therefore, resources were used instead on sealing the potential leaks at welds previously identified as suspect in the 2003 assessment report.

In the year 2005, an epoxy coating was applied to areas suspected of having leaks. In the year 2008, a drawing review identified some grafoil washers that may need to be replaced. A work request was initiated to replace the grafoil washers, but that request has not yet been implemented. In the year 2010, FENOC determined that the epoxy coating was ineffective because the epoxy coating did not stay bonded to the Refueling Canal liner. Therefore, a new plan had to be developed to mitigate the leakage. From March through May of 2010, staged fills of the Refueling Canal were conducted to try to narrow-down the Refueling Canal liner areas that could have sources of leakage.

Attachment L-11-317 Page 10 of 11

2.

The 2003 assessment report documented that Refueling Canal leakage had negligible impact on the structural integrity of the concrete structures in containment. The leakage was documented at least four years prior to 2003 and occurred only when the Refueling Canal was filled. Since 2003, periodic visual inspections of concrete surface areas affected by the leakage have shown no additional evidence of further damage to rebar or concrete. Therefore, it is acceptable to wait until the year 2014 to perform additional structural integrity verifications in addition to periodic visual inspections. This conclusion is based on plant-specific and industry operating experience with the effects of borated water leakage on concrete and reinforcing steel.

The plant-specific operating experience included non-destructive testing and petrographic examination of core drills from the east/west tunnel performed in the year 2002 that revealed no significant degradation of the concrete or reinforcing steel. In addition, information about related research and industry operating experience exists that documents the relatively minor effect that borated water leakage, through concrete, has on concrete and carbon steel reinforcing bars. For example, EPRI Report 1019168, "Boric Acid Attack of Concrete and Reinforcing Steel in PWR Fuel Handling Buildings," noted that "reinforcing steel degradation" from boric acid "is minimal." Also, industry operating experience shows that, since the year 1993, another plant has had substantially more refueling canal leakage (i.e., 3-to-7 gallons per minute) than Davis-Besse. As of 2010, with continuing leakage of 3-to-7 gallons per minute when the refueling canal contains water, the industry operating experience states that there had been no known degradation related to the refueling canal leakage.

3.

During the current Davis-Besse Cycle 17 mid-cycle outage, the rate of Refueling Canal leakage is estimated at less than 0.2 gallons per minute. This estimate is based on monitoring the volume of Refueling Canal makeup water adjusted by other identified leakage and evaporation losses.

Supplemental Question RAI OIN-380 During the NRC Region III Inspection Procedure (IP) 71002, "License Renewal Inspection," held the week of August 22, 2011, the NRC staff requested that the Davis-Besse License Renewal Structures Monitoring Program include enhancements to the descriptions of the Parameters Monitored or Inspected, the Detection of Aging Effects and the Acceptance Criteria. The FENOC License Renewal Project created Open Item Number (OIN)-380 to track the request, described as follows:

1.

Include the following enhancement to the Parameters Monitored or Inspected program element for the Structures Monitoring Program: Elastomeric

Attachment L-11-317 Page 11 of 11 vibration isolators and structural sealants are monitored for cracking, loss of material, and hardening.

2.

Include the following enhancement to the Detection of Aging Effects program element for the Structures Monitoring Program: Visual inspection of elastomeric vibration isolation elements will be supplemented by feel to detect hardening if the vibration isolation function is suspect.

3.

Include the following enhancements to the Acceptance Criteria program element for the Structures Monitoring Program: Loose bolts and nuts and cracked high strength bolts are not acceptable unless accepted by engineering evaluation. Structural sealants are acceptable if the observed loss of material, cracking, and hardening will not result in loss of sealing.

Elastomeric vibration isolation elements are acceptable if there is no loss of material, cracking, or hardening that could lead to the reduction or loss of isolation function.

SUPPLEMENTAL RESPONSE RAI OIN-380 LRA Section B.2.39, "Structures Monitoring Program," and Table A-I, "Davis-Besse License Renewal Commitments," license renewal future Commitment 20, are revised to include program procedure enhancements and new license renewal future commitments for:

1.

Parameters Monitored or Inspected for the Structures Monitoring Program:

Elastomeric vibration isolators and structural sealants are monitored for cracking, loss of material, and hardening.

2.

Detection of Aging Effects for the Structures Monitoring Program: Visual inspection of elastomeric vibration isolation elements will be supplemented by feel to detect hardening if the vibration isolation function is suspect.

3.

Acceptance Criteria for the Structures Monitoring Program: Loose bolts and nuts and cracked high strength bolts are not acceptable unless accepted by engineering evaluation. Structural sealants are acceptable if the observed loss of material, cracking, and hardening will not result in loss of sealing. Elastomeric vibration isolation elements are acceptable if there is no loss of material, cracking, or hardening that could lead to the reduction or loss of isolation function.

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

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

Letter L-11-317 Amendment No. 20 to the DBNPS License Renewal Application Page 1 of 49 License Renewal Application Sections Affected Section 2 Table 2.3.1-4 Section 2.3.3.18 Table 2.3.3-18 Section 3 Section 3.1.2.2.16.1 Table 3.1.1 Table 3.1.2-3 Table 3.1.2-4 Section 3.3.2.1.18 Section 3.3.2.2.4.1 Section 3.3.2.2.10.4 Table 3.3.1 Table 3.3.2-1 Table 3.3.2-14 Table 3.3.2-18 Table 3.3.2-30 Table 3.3.2 P-S Notes Section 3.4.2.2.7.3 Table 3.4.1 Table 3.4.2-1 Table 3.4.2-4 Table 3.4.2 P-S Notes Section 4 Section 4.7.6 Appendix A Section A.1.38 Section A.2.7.5 Table A-1 Appendix B Table B-2 Section B.2.38 Section B.2.39 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 finede o and added text underlined.

Enclosure L-11-317 Page 2 of 49 Affected LRA Section LRA Page No.

Page 2.3-21 Affected Paragraph and Sentence Table 2.3.1-4 New Row In response to RAI 3,1.2.2.16-1, LRA Table 2.3.1-4, "Steam Generators Components Subject to Aging Management Review," is revised to include a new row which reads as follows:

C Intended Function Component Type (as defined in Table 2.0-1)

Primary Side; Tube-to-tubesheet Weld Pressure boundary

Enclosure L-11-317 Page 3 of 49 Affected LRA Section LRA Page No.

Page 2.3-105 Section 2.3.3.18 Affected Paragraph and Sentence Components Subject to AMR subsection, 2 nd bulleted item In response to RAI 2.3.3.18-4, LRA Section 2.3.3.18, "Steam Makeup and Purification System," subsection "Components Subject to AMR," the second bulleted item, is deleted as follows:

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Page 2.3-106 Affected Paragraph and Sentence Table 2.3.3-18 2 New Rows In response to RAI 2.3.3.18-4, LRA Table 2.3.3-18, "Makeup and Purification System Components Subject to Aging Management Review," is revised to include two new rows which read as follows:

Intended Function Component Type (as defined in Table 2.0-1)

Heat Exchanger (channel, shell, tubesheet) -Pressure bounda Letdown coolers (DB-E25-1 & 2)

_ressureboundar Heat Exchanger (tubes) - Letdown coolers Heat transfer (DB-E25-1 & 2)

Pressure boundary

Enclosure L-11-317 Page 4 of 49 Affected LRA Section LRA Page No.

Affected Para-graDh and Sentence 3.1.2.2.16.1 Page 3.1-11 New [last] sentence In response to RAI 3.1.2.2.16-1, a new sentence is added to the end of LRA Section 3.1.2.2.16.1, "Stainless Steel or Nickel-Alloy Steam Generator Components - Reactor Coolant," and the section is revised to read:

3.1.2.2.16.1 Stainless Steel or Nickel-Alloy Steam Generator Components -

Reactor Coolant Cracking due to SCC could occur on the primary coolant side of stainless steel, stainless steel clad, and nickel-alloy clad components. Cracking due to SCC (including PWSCC) on the primary coolant side of Davis-Besse stainless steel, stainless steel clad, and nickel-alloy clad components is managed by the Inservice Inspection Program, Nickel-Alloy Management Program and PWR Water Chemistry Program. Cracking due to SCC (including PWSCC) in the nickel alloy steam generator tube-to-tubesheet welds is managed by the Steam Generator Tube Integrity Program and PWR Water Chemistry Program.

Enclosure L-11-317 Page 5 of 49 Affected LRA Section LRA Page No.

Page 3.1-26 Table 3.1.1 Affected Paragraph and Sentence Row 3.1.1-35 "Discussion" column, 2 nd paragraph, last sentence Text in "Discussion" column of Row 3.1.1-35 is revised based on the response to RAI 3.1.2.2.16-1, and 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-1 801," reads 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 Further Item Aging Effect/

Aging Management Evaluation Discussion Number Component/Commodity Mechanism Programs Recommended 3.1.1-35 Steel with stainless steel or Cracking due to Inservice Inspection No, but licensee Consistent with NUREG-1 801.

nickel alloy cladding primary side stress corrosion (IWB, IWC, and IWD) commitment components; steam generator cracking and primary and Water Chemistry needs to be Cracking due to SCC (including upper and lower heads, water stress and for nickel alloy, confirmed PWSCC) in steel steam tubesheets and tube-to-tube corrosion cracking FSAR supplement generator components with sheet welds commitment to stainless steel or nickel alloy itmeapplicable cladding is managed by the implement aInservice Inspection Program plant commitments to (1) and PWR Water Chemistry NRC Orders, Bulletins Program. Davis Bosse has ne and Generic Letters ni..ki..1,..........

thol associated with nickel associated withanickel rofr to th is item. Cracking due to alloys and (2) staff-SCC (including PWSCC) in the accepted industry nickel alloy steam generator guidelines.

tube-to-tubesheet welds is managed by the Steam Generator Tube Integrity

Enclosure L-11-317 Page 6 of 49 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 Further Item Aging Effect/

Aging Management Further Number Component/Commodity Mechanism Programs Evaluation Discussion Recommended Program and PWR Water Chemistry Program.

Further evaluation is documented in Section 3.1.2.2.16.1.

Enclosure L-11-317 Page 7 of 49 Affected LRA Section LRA Pane No.

Page 3.1-163 Affected Paraaragh and Sentence Table 3.1.2-3 8 New Rows Eight new rows were added to LRA Table 3.1.2-3, "Aging Management Review Results - Reactor Coolant System and Reactor Coolant Pressure Boundary," and provided in FENOC letter L-1 1-292 dated October 7, 2011, in response to Supplemental RAI Table 3.1.2-3. However, the table title provided in that response was incorrectly shown as "Aging Management Review Results - Decay Heat Removal and Low Pressure Injection System." The eight new rows added to Table 3.1.2-3 are provided with the corrected table title as follows:

Table 3.1.2-3 Aging Management Review Results - Reactor Coolant System and Reactor Coolant Pressure Boundary NUREG-Row Component Intended Material Environment Aging EfcAging

1801, Table 1 No.

Type Function(s)

Requiring Management Volume Item Notes Management Program 2VIteme Im 2 Item Piping <4 inches RV Pressure Nickel Borated Crackin-flange reactor coolant ratingu TLAA IV.C2-25 3.1.1-08 A

akaaelnge boundary

Alloy, (t

I aiu leakage line (Internal) tap weld Piping <4 inches RV Pressure Nickel Borated Cracking -

C incflange boundary Al reactor coolant Flaw Growth Inservice Inspection IV.C2-26 3.1.1-62 0102 leakage line (Internal) 0103 tap weld Piping <4 inches RV Pressure Nickel Borated Cracking -

flange boundary Alloy reactor coolant PWSCC, Inservice Inspection IV. C2-13 3.1.1-31 A

leakage line (Internal)

SCC/IGA taW weld I

I I

I

Enclosure L-11-317 Page 8 of 49 Table 3.1.2-3 Aging Management Review Results - Reactor Coolant System and Reactor Coolant Pressure Boundary Aging Effect Aging NUREG-Row Component Intended Material Environment Requiring Management

1801, Table I Notes No.

Type Function(s)

Malagenent Manam Volume Item Management Program 2 Item Piping <4 inches RV Pressure Nickel Borated Cracking-Nickel-Allo A

- Lne reactor coolant PWSCC, IV.C2-13 3.1.1-31 A

leakage line (Internal)

SCC/IGA tan weld Piping <4 inches RV Pressure Nickel Borated Cracking-PWR Water fa-ePesr Nikl reactor coolant PWSCC, IV.Wte C2-13 3.1.1-31A flane boundary Allo ra Chemistry leakage line dary (Internal)

SCC/IGA Chemistr tpweld Piping <4 inches RV Pressure Nickel Borated Cracking-Small Bore Class C

flange reactor coolant PWSCC, SI C2-13 3.1.1-31 leakage line bounda Alloy (Internal)

SCC/IGA Piping Inspection tao weld Piping <4 inches RV Pressure Nickel Borated Loss of PWR Water flange reactor coolant IV. C2-15 3.1.1-83 A

leakaae line bounda Alloy (Internal) tap weld Piping <4 Air with inches RV Pressure Nickel borated water None None IVE3 3.1.1-86 A

leakage line boundary Alloy leakage 0103 tan weld (External)

Enclosure L-11-317 Page 9 of 49 Affected LRA Section LRA Paae No.

Page 3.1-185 Affected Paragraph and Sentence Table 3.1.2-4 4 New Rows In response to RAI 3.1.2.2.16-1, four new rows are added to LRA Table 3.1.2-4, "Aging Management Review Results - Steam Generators," to read as follows:

Table 3.1.2-4 Aging Management Review Results - Steam Generators Aging Effect NUREG-Row Component Intended Material Environment Requiring Aging Management

1801, Table I Notes No.

Type Function(s)

Management Program Volume 2 Item Item Primary Side:

Tube-to-Pressure Nickel Borated reactor Cracking -

TLAA ID2-15 3.1.1-06 C

tubesheet boundary Alloy coolant FatTue-0101 Weld Primary Side:

Tube-to-Pressure Nickel Borated reactor Cracking -

C Water D2-4 3.1.1-35 A

tubesheet boundary Alloy coolant SCCIGA Chemistry I0101 Weld Primary Side:

Tube-to-Pressure Nickel Borated reactor Cracking -

Steam Generator E

tubesheet boundary Alloy coolant PWSCC,1 Tube Integrity IV.D24 3.1.1-35 0101 Weld SCC/IGA Primary Side:

Tube-to-Pressure Nickel Borated reactor Loss of PWR Water CC2-15 3.1.1-83 0

tubesheet boundary Alloy coolant Material Chemistry 0101 Weld

Enclosure L-11-317 Page 10 of 49 Affected LRA Section LRA Page No.

Page 3.3-24 Affected Paragraph and Sentence "Environments" subsection, new bulleted item 3.3.2.1.18 In response to RAI 2.3.3.18-4 related to the letdown coolers, the "Environments" subsection of LRA Section 3.3.2.1.18, "Makeup and Purification System," is revised to include a new environment in the "Environments" subsection as follows:

9 Closed cycle coolina water > 600C (> 1400F)

Enclosure L-11-317 Page 11 of 49 Affected LRA Section LRA Pacie No.

Affected Paragraph and Sentence 3.3.2.2.4.1 Page 3.3-40 Entire section In response to RAI 2.3.3.18-4 related to the letdown coolers, LRA Section 3.3.2.2.4.1, "Stainless Steel PWR Nonregenerative Heat Exchanger Components - Treated Borated Water Greater Than 60 0C (> 1400F),"

previously revised by FENOC letter dated June 3, 2011 (ML11159A132), is revised to read as follows:

3.3.2.2.4.1 Stainless Steel PWR Nonregenerative Heat Exchanger Components - Treated Borated Water Greater Than 600C

(> 1400F)

Cracking due to stress corrosion cracking and cyclic loading could occur in stainless steel pressurized water reactor (PWR) nonregenerative heat exchanger components exposed to treated borated water greater than 600C (> 1400F) in the chemical and volume control system. At Davis Besse, the Au,,liary Sytems do nOt contain stainless stee! nonregenerative heat exchanger-component that are exposed to treated borted water greater than 600C (> 1 400F) and subjecGt agn maagement review;, theeADfor, this item is not applicable to Davis-Besso.

At Davis-Besse, the letdown coolers (DB-E25-1 and 2) and the seal return coolers (DB-E26-1 and 2) in the Makeup and Purification System consist of stainless steel heat exchanger components exposed to treated borated water greater than 600C (> 1400F). Cracking due to stress corrosion cracking (SCC) in stainless steel heat exchanger components that are exposed to treated borated water greater than 600C (>1400F) is managed by the PWR Water Chemistry Program. The PWR Water Chemistry Program manages cracking through periodic monitorinq and control of contaminants. One-Time Inspection will provide verification of the effectiveness of the PWR Water Chemistry Program to manage cracking. The coolers are in continuous service and not subiect to cyclic loading: therefore, eddy current testing of the tubes to manage cyclic loading is not applicable. Temperature and radioactivity monitoring of shell side water is performed by installed instrumentation.

Enclosure L-11-317 Page 12 of 49 Affected LRA Section LRA Page No.

Affected Paragraph and Sentence 3.3.2.2.10.4 Page 3.3-45 1 st and 2 nd sentences In response to RAI 3.3.2.2.10.4-1, the first two sentences of LRA Section 3.3.2.2.10.4, "Copper Alloy Piping, Piping Components, and Piping Elements-Lubricating Oil," are revised to read as follows:

3.3.2.2.10.4 Copper Alloy Piping, Piping Components, and Piping Elements - Lubricating Off Loss of material due to pitting and crevice corrosion could occur for copper alloy piping, piping components, and piping elements exposed to lubricating oil. Loss of material due to pitting and crevice corrosion for Davis-Besse copper alloy piping, piping components, and piping elements with a zinc content g.eater. than

-1%-that are exposed to lubricating oil is managed by the Lubricating Oil Analysis Program. Loss of material for copper alloy heat exchanger components wih a zinc con.tent greater than 15% that are exposed to lubricating oil is also managed by the Lubricating Oil Analysis Program. The Lubricating Oil Analysis Program manages loss of material through periodic monitoring and control of contaminants, including water. The One-Time Inspection will provide verification of the effectiveness of the Lubricating Oil Analysis Program to manage loss of material.

Enclosure L-11-317 Page 13 of 49 Affected LRA Section LRA Pane No.

Page 3.3-51 Affected Paragraph and Sentence Row 3.3.1-07, "Discussion" column Table 3.3.1 In response to RAI 2.3.3.18-4, the text in the "Discussion" column of row 3.3.1-07 of LRA Table 3.3.1, "Summary of Aging Management Programs for Auxiliary Systems Evaluated in Chapter VII of NUREG-1 801,"

previously revised by FENOC letter dated June 3, 2011 (ML11159A132), 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 Further Item Aging Effect/

Aging Management Evaluation Discussion Number Component/Commodity Mechanism Programs Recommended 3.3.1-07 Stainless steel non-regenerative Cracking due to Water Chemistry and a Yes, plant Not-applieeb heat exchanger components stress corrosion plant-specific specific The 4 1 iliar, Systems d& no exposed to treated borated cracking and cyclic verification program. An

.ont.in t.

ste.

water >60'C (>140TF) loading acceptable verification tl.a heat e..han...

program is to include

t. At to temperature and t

b radioactivity monitoring

(.-1 and s.bject tai o f th e s h e ll s id e w a te r,

(> 1 4 0.

a n t e

a n d e d d y c u rre n t re v ie w testing of tubes.

Consistent with NUREG-1801.

Cracking due to SCC for stainless steel heat exchanger components in the Auxiliary Systems that are exposed to treated borated water > 600C

(> 1400F) is managed by the PWR Water Chemistry Program.

The One-Time Inspection will provide verification of the

Enclosure L-11-317 Page 14 of 49 Table 3.3.1 Summary of Aging Management Programs for Auxiliary Systems Evaluated in Chapter VII of NUREG-1801 I

F Further Item CC dit Aging Effect/

Aging Management Evaluation Discussion Number omponenommoiy Mechanism Programs E

uRecommended effectiveness of the PWR Water Chemistry Program to manage cracking.

Cracking due to cyclic loading is not applicable since these components are continuously in service and not subiect to cyclic loading.

Temperature and radioactivity monitoring of shell side water is performed by installed instrumentation.

Further evaluation is documented in Section 3.3.2.2.4.1.

Enclosure L-11-317 Page 15 of 49 Affected LRA Section LRA Page No.

Page 3.3-67 Affected Paragraph and Sentence Row 3.3.1-26, "Discussion" column Table 3.3.1 In response to RAI 3.3.2.2.10.4-1, the text in the "Discussion" column of row 3.3.1-26 of 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 Item Aging Effect/

Aging Management Further Number Component/Commodity Mechanism Programs Recommended 3.3.1-26 Copper alloy piping, piping Loss of material Lubricating Oil Analysis Yes, detection of Consistent with NUREG-1801.

components, and piping due to pitting and and One-Time aging effects is elements exposed to lubricating crevice corrosion Inspection to be evaluated Loss of material in copper alloy oil piping, piping components, and piping elements exposed to lubricating oil is managed by the Lubricating Oil Analysis Program if the zinc, content is greator than 4-" %. The One-Time Inspection will provide verification of the effectiveness of the Lubricating Oil Analysis Program to manage loss of material.

This item is also applied to copper alloy heat exchanger components 14ith zinc co-ntent greater than 15% that are

Enclosure L-11-317 Page 16 of 49 Table 3.3.1 Summary of Aging Management Programs for Auxiliary Systems Evaluated in Chapter VII of NUREG-1801 Further Item Aging Effect/

Aging Management Euatis s

Numer Component/Commodity Mehns rgasEvaluation Discussion Number oMechanism Programs Recommended exposed to lubricating oil.

LOSS oA MI.aterial duo to p#ttig and crevic vorso as not ide-netifi-d a8 an aging ect requiring mana gement for Gopp9r alloyppnpipipn coponents, and piping !eoents with a zinc g-ntenAt -aRe

-8,1 thanR.15%V-that are-expesed to lubricating ol Further evaluation is documented in Section 3.3.2.2.10.4.

Enclosure L-11-317 Page 17 of 49 Affected LRA Section LRA Pane No.

Page 3.3-140 Affected Paragraph and Sentence Table 3.3.2-1 Row 119; and, 1 New Row In response to RAI 3.3.2.2.10.4-1, row 119 of LRA Table 3.3.2-1, "Aging Management Review Results -

Auxiliary Building HVAC System," is revised, and a new row is added, to read as follows:

Table 3.3.2-1 Aging Management Review Results - Auxiliary Building HVAC System A

gEffet Aging NUREG-Row Component Intended Material Environment gginge

1801, Table I No.

Type Function(s)

Requiring Management Volume Item Notes Management Program 2 Item None Nn 4-11 uig Pressure Copper Lubricating oilNoeAng4 119 Tubing boundary Alloy (Internal)

Loss of Lubricating Oil VII.C1-8 3.3.1-26 0302 material Analysis A

Tubin~i Pressure Copper Lubricating oil Loss of One-Time V11C1-8 3.3.1-26 A

Tboundary Alloy (Internal) material Inspection I

Enclosure L-11-317 Page 18 of 49 Affected LRA Section LRA Page No.

Page 3.3-337 Affected Paragraph and Sentence Table 3.3.2-14 Row 215; and, I New Row In response to RAI 3.3.2.2.10.4-1, row 215 of LRA Table 3.3.2-14, "Aging Management Review Results -

Fire Protection System," previously revised by FENOC letter dated September 16, 2011 (ML11264A059), is revised, and a new row is added, to read as follows:

Table 3.3.2-14 Aging Management Review Results - Fire Protection System Row Component Intended Aging Effect Aging NUREG1 Ro Type Intended Material Environment Requiring Management 180e Table I Notes No.

Type Function~s)

ManagementPrga Volume Item Management Program 2 Item Heat Exchanger A/ene A

4-215 (tubes) -

Pressure Copper Lubricating oil Loss of Lubricating Oil VII.G-1 1 3.3.1-26 0802 Gear boundary Alloy (External) material Analysis C

Housing Oil Cooler Heat Exchanger (tubes) -

Pressure Copper Lubricating oil Loss of One-Time Gear boundary Alloy (External) material Inspection Housing Oil

-Cooler

Enclosure L-11-317 Page 19 of 49 Affected LRA Section LRA Pane No.

Pages 3.3-368 thru 3.3-397 Affected Paragraph and Sentence Rows 24, 25, 50, 51, 58 and 59; and, 24 New Rows Table 3.3.2-18 In response to RAI 2.3.3.18-4, rows 24, 25, 50, 51, 58 and 59, previously revised by FENOC letter dated June 3, 2011 (ML11159A132), are revised, and 24 new rows are added to LRA 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 Row Component Intended Aging Effect Aging NUREG-Row Copoen Fnctiondd Material Environment Requiring Management Volume Item Notes Management Program 2 Item Item Heat Treated Exchanger borated water 24 (channel) -

Pressure Stainless

> 600C Cracking One-Time VII.E1-42 3.3.1-90 E

Seal return boundary Steel

(>14 0 F)

Inspection 5

07 0315 coolers (DB-(Internal)

E26-1 & 2)

(Internal)

Heat Treated Exchanger borated water (channel) -

Pressure Stainless

> 60rC Cracking PWR Water VII.E0-20 3.3.1-90 4,

25 Seal return boundary Steel

> 100 F Chemistry 5

07 A

coolers (DB-(internal)

E26-1 & 2) 1 1 n te r

Enclosure L-11-317 Page 20 of 49 Table 3.3.2-18 Aging Management Review Results - Makeup and Purification System Row Component Intended Aging Effect Aging NUREG-Row Cmponen Fnctiond Material Environment Requiring Management

1801, Table 1 Notes No.

Type Function(s)

Maaeet PormVolume Item Management Program 2 Item Heat Treated Exchanger borated water 0

(tubes) -

Pressure Stainless

> 6000 Cracking One-Time VII.EI-20 3.3.1-90 E

Seal return boundary Steel

(> 1400F)

Inspection 5

07 0315 coolers (DB-(Internal)

E26-1 & 2)

(Internal)

Heat Treated Exchanger borated water (tubes) -

Pressure Stainless o

water PWR Water VII.EI-20 3.3.1-90 C

Seal return boundary Steel

(> 1400F)

Chemistry 5

07 A

coolers (DB-(Internal)

E26-1 & 2)

(Internal)

Heat Treated Exchanger borated water (tubesheet) -

Pressure Stainless b

water One-Time VlI.EI 3.3.1-90 E

Seal return boundary Steel

(> 1400F)

Inspection 5

07 0315 coolers (DB-(Internal)

E26-1 & 2)

(Internal Heat Treated Exchanger borated water 59 (tubesheet) -

Pressure Stainless

> 60CPWR Water VIEl-20 3.3.1-90 Seal return boundary Steel

(> 1400F)

Chemistry 5

7 A

coolers (DB-(Internal)

E26-1 & 2)

(Internal)

Heat Treated Exchanger borated water (channel) -

Pressure Stainless borte w r

One-Time VII-5 3.3.1-07 E

Letdown boundary Steel

> 60C Cracking Inspection 0-coolers (DB-(Interal)

E25-1 & 2)

(Internal)

Enclosure L-11-317 Page 21 of 49 Table 3.3.2-18 Aging Management Review Results - Makeup and Purification System Aging Effect Aging NUREG-Row Component Intended AigEfc gn 81 al Row Type Intended Material Environment Requiring Management

1801, Table I Notes No.Management Program Volume Item 2 Item Heat Treated Exchanger borated water (channel) -

Pressure Stainless borte w r

PWR Water VII-5 3.3.1-07 A

Letdown boundary Steel

(> 140F)

Chemistry coolersD-(Internal)

E25-1 & 2)

(Internal)

Heat Treated Exchanger borated water E

(channel) -

Pressure Stainless wae Loss of One-Time VII.EI-17 3.3.1-91 0315 Letdown boundary Steel

(> 140F) material Inspection coolers (DB-(Internal)

E25-1 & 2)

(Internal_

Heat Treated Exchanger borated water (channel) -

Pressure Stainless

> 60C Loss of PWR Water VII.El-17 3.3.1-91 C__

Letdown boundary Steel

(> 140F) material Chemistry 0329 coolers (DB-(Internal)

E25-1 & 2)

(Internal)

Heat Exchanger Air with (channel) -

Pressure Stainless borated water None None VIIJ-16 3.3.1-99 C

Letdown boundary Steel leakage coolers (DB-(External)

E25-1 & 2)

Heat Exchanger Air-indoor (channel) -

Pressure Stainless Arido uncontrolled None None VII.J-15 3.3.1-94 C

Letdown boundary Steel (External) coolers (DB-E25-1 & 2)

Enclosure L-11-317 Page 22 of 49 Table 3.3.2-18 Aging Management Review Results - Makeup and Purification System NUREG-Row Component Intended Aging Effect Aging

1801, Table 1 Notes No.

Type Function(s)

Material Environment Requiring Management Volume Item Management Program 2 Item Heat Closed cycle Exchangercoiqwt (shell)

Pressure Steel cooin water Loss of Closed Cooling VIEI-6 3.3.1-48 B

Letdown boundary

(> 140C) material Water Chemistry coolers (DB-(Internal)

I E25-1 & 2)

Heat Exchanger Air with (shell)-

Pressure borated water Loss of Boric Acid V/0-10 3.3.1-89 A

Letdown boundary leakage material Corrosion coolers (DB-(External)

E25-1 & 2)

Heat Exchanger Air-indoor (shell)

Pressure Air-indoor Loss of External Surfaces V1I.1-8 3.3.1-58 A

Letdown boundary (External) material Monitoring coolers (DB-E25-1 & 2)

Heat Treated Exchanger borated water (tubes) -

Heat transfer Stainless

> 60C Reduction in One-Time NIA NIA H

Letdown Steel

(> 140F) heat transfer Inspection 0315 coolers (DB-(Internal)

E25-1 & 2)

(Internal)

Heat Treated Exchanger borated water (tubes) -

Heat transfer Stainless b

wae Reduction in PWR Water N/A N/A H

Letdown Steel

(> 140F) heat transfer Chemistry N

coolers (DB-(Internal)

_ E25-1&2)

&Inte rnal

Enclosure L-1 1-317 Page 23 of 49 Table 3.3.2-18 Aging Management Review Results - Makeup and Purification System NUREG-Row Component Intended Aging Effect Aging

1801, Table I No.

Type Function(s)

Material Environment Requiring Management Volume Item Notes Management Program 2 Item Heat Closed cycle Exchanger cooling water (tubes) -

Heat transfer Stainless

> 60C Reduction in Closed Coolin VIIE3-5 3.3.1-52 B

Letdown heat transfer Water Chemistry 0329 coolers (DB-(External)

E25-1 & 2)

(External)

Heat Treated Exchanger borated water (tubes) -

Pressure Stainless b

water One-Time E

Letdown boundary Steel

> 60C Cracking VII.El-3.3.1-07 315 (10)Inspection

_315 coolers (DB-(internal)

E25-1 & 2)

Heat Treated Exchanger borated water (tubes) -

Pressure Stainless b

water PWR Water VI1.E-5 3.3.1-07 A

Letdown boundary Steel

(> 140F)

Chemistry coolers (DB-(Internal)

E25-1 & 2)

(Internal)

Heat Treated Exchanger E

(tubes) -

Pressure Stainless borated water Loss of One-Time V11.E-17 3.3.1-91 0315 Letdown boundary Steel

(> 140F) material Inspection coolers (DB-(Internal)

E25-1 & 2)

(Internal)

Heat Treated Exchanger Presur Si Loss of PWR Water C

(tubes) -

Pressure Stainless

> 60C V.0-17 3.3.1-91__

Letdown boundary Steel

(> 140F) material Chemistry 0329 coolers (DB-(Internal)

E25-1 & 2)

(internal)

Enclosure L-11-317 Page 24 of 49 Table 3.3.2-18 Aging Management Review Results - Makeup and Purification System Row Component Intended Aging Effect Aging NUREG-Row Copnet Fnctiondd Material Environment Requiring Management

1801, Table I Notes No.

Type Function(s)

MaaeetPormVolume Item Management Program 2 Item Heat Closed cycle Exchanger coolin water (tubes) -

Pressure Stainless

> 6cC Crawatr

,Closed Coolin VII.C2-11 3.3.1-46 D

Letdown boundary Steel

(> 140F)

Water Chemistry coolers (DB-(Exteral)

E25-1 & 2)

(External)

Heat Closed cycle Ehner)

P rn cooling water Loss of Closed Cooing B

(tubes) -

Pressure Stainless

> 60C maeilWtrCe

't V11.C2-10 3.3.1-50 B

Letdown bounda Steel

(> 140) material Water Chemistry 0329 coolers (DB-(Extrna)

E25-1 & 2)

(External)

Heat Treated Exchange borated water (tubesheet) -

Pressure Stainless

> 60C Cracking One-Time VIEEI-5 3.3.1-07 0

Letdown boundary Steel

(> 140F)

Inspection 0315 coolers (DB-(Interal)

E25-1 & 2)

(Internal)

Heat Treated Exchanger borated water (tubesheet) -

Pressure Stainless

> 60C Cracking PWR Water VII-5 3.3.1-07 A

Letdown boundary Steel

(> 140F)

Chemistry coolers (DB-(Inteal)

E25-1 & 2)

(Internal_

Heat Treated Exchanger borated water E

(tLubesheet) -

Pressure Stainless wae Loss of One-Time VII.EI-17 3.3.1-91 0315 LetdoWn boundary Steel

(> 140F) material Inspection 0329 coolers (DB-(Internal)

E2 -

_1 & 2)

Enclosure L-11-317 Page 25 of 49 Table 3.3.2-18 Aging Management Review Results - Makeup and Purification System NUREG-Row Component Intended Material Environment Aging EftAging

1801, Table I Notes No.

Type Function(s)

Management Program Volume Item 2 Item Heat Treated Exchanger borated water (tubesheet) -

Pressure Stainless

> 60C Loss of PWR Water VICE17 3.3.1-91 Letdown boundary Steel

(>140F) material Chemistry 0329 coolers (DB-(Internal)

E25-1 & 2)

(Internal)

Heat Closed cycle Exchanger (tubesheet) -

Pressure Stainless cooling water Closed Cooling

> 60C Cracking VII. C2-1 1 3.3.1-46 D

Letdown boundary Steel

(> 140F)

Water Chemistry coolers (DB-(External)

E25-1 & 2)

(External)

Heat Closed cycle Exchanger (tubesheet) -

Pressure Stainless cooling water Loss of Closed Cooling V11.C2-10 3.3.1-50 B

Letdown boundary Steel

(> 140F) material Water Chemistry 0329 coolers (DB-(External)

E25-1 & 2)

(Ex l

1 1

1

Enclosure L-11-317 Page 26 of 49 Affected LRA Section LRA Page No.

Page 3.3-391 Affected Paragraph and Sentence Table 3.3.2-18 Row 156; and, 1 New Row In response to RAI 3.3.2.2.10.4-1, row 156 of LRA Table 3.3.2-18, "Aging Management Review Results -

Makeup and Purification System," is revised, and a new row is added, to read as follows:

Table 3.3.2-18 Aging Management Review Results - Makeup and Purification System Aging Effect Aging NUREG-Row Component Intended Aging Magint

1801, Table 1 Notes No.

Type Function(s)

Material Environment Requiring Management Volume Item Management Program 2 Item Pressure Copper Lubricating oil4 156 Valve Body boundary Alloy (Internal)

Loss of Lubricating Oil VII.E1-12 3.3.1-26 0302 material Anal}sis A

Pressure Copper Lubricating oil Loss of One-Time VII.El-12 3.3.1-26 A

Valve Body boundary Alloy (Internal) material Inspection I

A

Enclosure L-11-317 Page 27 of 49 Affected LRA Section LRA Page No.

Page 3.3-527 Affected Paragraph and Sentence Table 3.3.2-30 Row 144; and, 1 New Row In response to RAI 3.3.2.2.10.4-1, row 144 of LRA Table 3.3.2-30, "Aging Management Review Results -

Station Blackout Diesel Generator System," is revised, and a new row is added, to read as follows:

Enclosure L-11-317 Page 28 of 49 Affected LRA Section Table 3.3.2 Plant-Specific Notes LRA Page No.

Page 3.3-547 Affected Paragraph and Sentence Row 0302 In response to RAI 3.3.2.2.10.4-1, row 0302 of Table 3.3.2, "Plant-Specific Notes," is no longer used, and is revised as follows:

Plant-Specific Notes:

0302 This m÷twr;il is coppor alley-< 15% Z-and-is not in

ontact 1t;h ; mrq.athodic motal,tha*r*aor, theram non agin o-',,t, rauiring managment in the lubricating oil environment.

Not used.

Enclosure L-11-317 Page 29 of 49 Affected LRA Section LRA Page No.

Affected Paragraph and Sentence 3.4.2.2.7.3 Page 3.4-10 2 nd sentence; and, New [last] sentence In response to RAI 3.3.2.2.10.4-1, the second sentence of LRA Section 3.4.2.2.7.3, "Copper Alloy Piping, Piping Components, Piping Elements -

Lubricating Oil," is revised, and a new last sentence is added to better align with LRA Table 3.4.1, "Summary of Aging Management Programs for Steam and Power Conversion Systems Evaluated in Chapter VIII of NUREG-1 801,"

row 3.4.1-18. LRA Section 3.4.2.2.7.3 reads as follows:

3.4.2.2.7.3 Copper Alloy Piping, Piping Components, Piping Elements -

Lubricating Oil Loss of material due to pitting and crevice corrosion could occur for copper alloy piping, piping components, and piping elements exposed to lubricating oil. At Davis Besse, loss of material due to pitting and crevice corrosion, and sele leaGhing, for copper alloy (copper "o-aft>

,5% Zn) piping, piping components, and piping elements that are exposed to lubricating oil in the Steam and Power Conversion Systems is managed by the Lubricating Oil Analysis Program. The Lubricating Oil Analysis Program manages loss of material through periodic monitoring and control of contaminants, including water. The One-Time Inspection will provide verification of the effectiveness of the Lubricating Oil Analysis Program to manage loss of material. This item is also applied to copper alloy (copper alloy > 15% Zn) heat exchanger components that are exposed to lubricating oil in the Steam and Power Conversion Systems. This item is also applied to loss of material due to selective leaching for copper alloy (copper alloy > 15% Zn) comnonents that are exposed to lubricatina oil.

Enclosure L-11-317 Page 30 of 49 Affected LRA Section LRA Page No.

Page 3.4-24 Affected Paragraph and Sentence Row 3.4.1-18, "Discussion" column Table 3.4.1 In response to RAI 3.3.2.2.10.4-1, the text in the "Discussion" column of row 3.4.1-18 of LRA Table 3.4.1, "Summary of Aging Management Programs for Steam and Power Conversion Systems Evaluated in Chapter VIII of NUREG-1801," is revised to read as follows:

Table 3.4.1 Summary of Aging Management Programs for Steam and Power Conversion Systems Evaluated in Chapter VIII of NUREG-1801 Further Item Aging Effect/

Aging Management Euatis s

Number Component/Commodity Mechanism Programs Recommended 3.4.1-18 Copper alloy piping, piping Loss of material Lubricating Oil Analysis Yes, detection of Consistent with NUREG-1 801.

components, and piping due to pitting and and One-Time aging effects is Loss of material due to pitting elements exposed to lubricating crevice corrosion Inspection to be evaluated and crevice corrosion in copper oil alloy (.OPPe. a"y > 15% Zn) piping, piping components, and piping elements that are exposed to lubricating oil is managed by the Lubricating Oil Analysis Program. The One-Time Inspection will provide verification of the effectiveness of the Lubricating Oil Analysis Program to manage loss of material.

Les of mat*Fi;l due to pitting and cre.ic xin rzio P

~

not

Enclosure L-11-317 Page 31 of 49 Table 3.4.1 Summary of Aging Management Programs for Steam and Power Conversion Systems Evaluated in Chapter VIII of NUREG-1801 Further Item Aging Effect/

Aging Management Euatis s

Numer Component/Commodity Mehns rgasEvaluation Discussion Number oMechanism Programs Recommended

dengtifed -s -n q6-n 4

of.',t c

er 311UY p*ipg, piOp*g compoqnents, ond piping tha.n 15% thait we posed to luFiGatin~g4A This item is also applied to copper alloy (copper alloy > 15%

Zn) heat exchanger components that are exposed to lubricating oil. This item is also applied to loss of material due to selective leaching for copper alloy (copper alloy > 15% Zn) components that are exposed to lubricating oil.

Further evaluation is documented in Section 3.4.2.2.7.3.

Enclosure L-11-317 Page 32 of 49 Affected LRA Section LRA Page No.

Page 3.4-46 Affected Paragraph and Sentence Table 3.4.2-1 Row 36; and, 1 New Row In response to RAI 3.3.2.2.10.4-1, row 36 of LRA Table 3.4.2-1, "Aging Management Review Results -

Auxiliary Feedwater System," is revised, and a new row is added, to read as follows:

Table 3.4.2-1 Aging Management Review Results - Auxiliary Feedwater System A

gEffet Aging NUREG-Row Component Intended Aging EfectAgint

1801, Table I Notes No.

Type Function(s)

Material Environment Requiring Management Volume Item Management Program 2 Item Heat exhngrAene Nen-/-

exchanger Pressure Copper Lubricating oil Loss oLra 36 (tubes) -

boundary Alloy (External) mLoss of Lubricating 0 VI1-8 3.3.1-26 aCs AFW pump maeral___y oil coolers Heat exchanger Pressure Copper Lubricating oil Loss of One-Time (tubes) boundary Alloy (External) material Inspection VII.CI-8 3.3.1-26 C

AFWo uo p oil coolers

Enclosure L-11-317 Page 33 of 49 Affected LRA Section LRA Pane No.

Page 3.4-85 Affected Para-graph and Sentence Table 3.4.2-4 Row 26; and, 1 New Row In response to RAI 3.3.2.2.10.4-1, row 26 of LRA Table 3.4.2-4, "Aging Management Review Results - Main Steam System," is revised, and a new row is added, to read as follows:

Table 3.4.2-4 Aging Management Review Results - Main Steam System NUREG-Row Component Intended Material Environment Aging EfcAging

1801, Table I No.

Type Function(s)

Requiring Management Volume Item Notes Management Program 2 Item Heat exchanger (tubes) -

Pressure Copper Lubricating oil Lapao Lurcig Oil al 3.4.1 A-1 26 AFW pump boundary Alloy (External)

Lossaof Lubricatinsq0 VIIIG-19 3.4.1-18 turbine material Analys C

bearing lube I oil cooler Heat exchanger (tubes) -

Pressure Copper Lubricating oil Loss of One-Time AFWpump boundary Alloy (External) material Inspection VIII.G-19 3.4.1-18 C

turbine bearing lube oil cooler

Enclosure L-11-317 Page 34 of 49 Affected LRA Section LRA Page No.

Affected Paragraph and Sentence Table 3.4.2 Page 3.4-111 Row 0413 Plant-Specific Notes In response to RAI 3.3.2.2.10.4-1, row 0413 of Table 3.4.2, "Plant-Specific Notes," is no longer used, and is revised as follows:

Plant-Specific Notes:

0413 This M-We3! iS coppeFr,,ey 15% Zn and is net in centact with a more cathodic metal; therAfora, thoro are o

';gg eff*o*t No i

nagement in the lubridatin.g enkrnment.

Not used.

Enclosure L-11-317 Page 35 of 49 Affected LRA Section LRA Page No.

Affected Paragraph and Sentence 4.7.6 Page 4.7-6 2 nd Paragraph, sub-item #1; 2 New Paragraphs (Nos. 3 & 4); and, 5 th Paragraph, 1st sentence In response to Supplemental RAI 4.1-2, LRA Section 4.7.6, "ASME Code Case N-481 Evaluation," is revised to read as follows:

4.7.6 ASME CODE CASE N-481 EVALUATION The reactor coolant pumps (RCPs) are the only ASME Code Class 1 pumps installed at Davis-Besse. The pump casings are constructed of cast austenitic stainless steel. The applicable ASME Code for the current Third Ten-Year Inspection Interval for Davis-Besse is ASME Section XI, 1995 Edition, through the 1996Addenda, as modified by 10CFR50.55a or relief granted in accordance with 10 CFR 50.55a. Examination Category B-L-1 of this Code year requires volumetric examination of pump casing welds. ASME Code Case N-481, "Alternative Examination Requirements for Cast Austenitic Pump Casings,"

provides an alternative to the volumetric examination requirement. This code case allows the replacement of volumetric examinations of primary loop pump casings with fracture mechanics-based integrity evaluation (Item (d) of the code case) supplemented by specific visual examinations. Davis-Besse has invoked the use of Code Case N-481 in place of the volumetric examination requirements of Code Category B-L-1. The NRC has accepted Code Case N-481 for use in inservice inspection programs.

Code Case N-481 requires an evaluation to demonstrate the safety and serviceability of the pump casings. The evaluation for the Davis-Besse RCPs required by Code Case N-481 is documented in Structural Integrity Associates (SIA) report SIR-99-040 [Reference 4.8-18]. This evaluation assumed a quarter thickness flaw, with length six times its depth, and showed that the flaw will remain stable considering the stresses and material properties of the pump casing. To determine stability of the postulated flaw, a fracture mechanics evaluation was performed that included a fatigue crack growth analysis to demonstrate that a small initial assumed flaw (10 percent through-wall),

corresponding to the acceptance standards of ASME Code, Section Xl, Subarticle IWB-3500, would not grow to quarter thickness during plant life.

There are two potential time-dependencies in the Code Case N-481 evaluation.

Enclosure L-11-317 Page 36 of 49

1.

The fracture toughness of the cast austenitic stainless steel is not time dependent as the Davis-Besse ASME Code Case N-481 analysis used a lower bound fracture toughness of 139 ksi'lin that bounds the sak*ated saturation fracture toughness of the Davis-Besse material.

2.

The fatigue crack growth analysis is based on design cycles for a 40 year plant life and therefore, is a TLAA requiring analysis and disposition for license renewal.

With respect to Item No. I above, the saturation fracture toughness was determined usinq the methodology outlined in NUREG/CP-0119, Volume 2, pages 151-178, "Proceedings of the U.S. Nuclear Regulatory Commission, 19'h Water Reactor Safety Information Meeting held at Bethesda, MD, October 28-30, 1991," and considering all available certified material test reports (CMTRs) for the base material and welds of the Davis-Besse RCP casings. The saturation fracture toughness value of 139 ksi'lin was the minimum calculated for all the CMTRs considered in the evaluation. This minimum saturation fracture toughness value has since been calculated using NUREG/CR-4513, Revision 1.

"Estimation of Fracture Toughness of Cast Stainless Steels During Thermal Aging in LWR Systems." Usinq the methodology and correlation in this NUREG results in the same minimum saturation fracture touqhness value for the pump casings.

The fracture toughness for welds considering thermal aging has also been presented in NUREG/CR-6428, "Effects of Thermal Aging on Fracture Toughness and Charpy-Impact Strength of Stainless Steel Pipe Welds." A conservative Ji1 fracture toughness value of 40 KJ/m2 based on the absolute minimum of all available data is provided in this document for aged stainless steel welds; this Jc fracture toughness value translates to 80 ksi*/in. This conservative fracture toughness value still bounds the calculated total applied stress intensity factors calculated in Table 4-5 of SIR-99-040, Revision 1, indicating that the conclusions of SIR-99-040, Revision 1, are unchanged even if the methodology outlined in NUREG-CR-6428 is used for the Davis-Besse pump casing welds.

With respect to Item No. 2 above, -Tthe fatigue crack growth analysis assumed an initial flaw size corresponding to the acceptance standards of ASME Code Section XI and considered all the significant plant transients. This analysis examined the design cycles and determined there were 240 cycles that were significant to flaw growth in the RCPs. Then 2000 cycles were conservatively analyzed, and flaw growth (initial 10 percent assumed through-wall had grown only to 15 percent through-wall) remained well below the quarter thickness postulated flaw. The analyzed cycles of 2000 bound the 60-year projected cycles shown in LRA Table 4.3-1 and therefore, the fatigue crack growth TLAA

Enclosure L-11-317 Page 37 of 49 associated with the ASME Code Case N-481 evaluation will remain valid for the period of extended operation.

Disposition:

10 CFR 54.21(c)(1)(i)

The fatigue crack growth TLAA associated with ASME Code Case N-481 evaluation will remain valid through the period of extended operation.

Enclosure L-11-317 Page 38 of 49 Affected LRA Section LRA Page No.

Affected Paragraph and Sentence A.1.38 Pages A-24 &

New 3 rd Paragraph A-25 In response to RAI 3.1.2.2.16-1, a new third paragraph is inserted into LRA Section A.1.38, "Steam Generator Tube Integrity Program," to read as follows:

In addition, cracking due to PWSCC is managed for the steam generator tube-to-tubesheet welds (Alloy 600) by a combination of the PWR Water Chemistry Pro-gram and the Steam Generator Tube Integrity Pro-gram. The PWR Water Chemistry Program controls peak levels of various contaminants (e.g., dissolved oxygen, chlorides, fluorides, and sulfates) below the system-specific limits that can accelerate cracking for nickel-alloy components. The Steam Generator Tube Integrity Program includes enhanced visual (EVT-1 or equivalent) examinations to monitor for cracking of the steam generator tube-to-tubesheet welds. The weld inspection sample size includes 20 percent of the subiect weld population or a maximum of 25, whichever is less. In this case the maximum of 25 applies since the weld population for the two steam generators is greater than 60,000. Welds included in the inspection sample are scheduled for examination in each 10-year period that occurs during the period of extended operation. Unacceptable inspection findings shall be evaluated by the Corrective Action Program using criteria in accordance with Section X1 of the ASME Code. Should the steam generators be replaced in the future with a desiqn such that the tube-to-tubesheet welds are fabricated of Alloy 690-TT material, the examinations will no longer be reguired.

Enclosure L-11-317 Page 39 of 49 Affected LRA Section LRA Page No.

Affected Paragraph and Sentence A.2.7.5 Page A-50 2 nd Paragraph, sub-item #1; 2 New Paragraphs (Nos. 3 & 4); and, 5 th Paragraph, Ist sentence In response to Supplemental RAI 4.1-2, LRA Section A.2.7.5, "ASME Code Case N-481 Evaluation," is revised to read as follows:

A.2.7.5 ASME Code Case N-481 Evaluation The reactor coolant pumps (RCPs) are the only ASME Code Class 1 pumps installed at Davis-Besse. The pump casings are constructed of cast austenitic stainless steel. The applicable ASME Code for the current Third Ten-Year Inspection Interval for Davis-Besse is ASME Section XI, 1995 Edition, through the 1996Addenda, as modified by 10CFR50.55a or relief granted in accordance with 10 CFR 50.55a. Examination Category B-L-1 of this Code year requires volumetric examination of pump casing welds. ASME Code Case N-481, "Alternative Examination Requirements for Cast Austenitic Pump Casings,"

provides an alternative to the volumetric examination requirement. This code case allows the replacement of volumetric examinations of primary loop pump casings with fracture mechanics-based integrity evaluation (Item (d) of the code case) supplemented by specific visual examinations. Davis-Besse has invoked the use of Code Case N-481 in place of the volumetric examination requirements of Code Category B-L-1. The NRC has accepted Code Case N-481 for use in inservice inspection programs.

Code Case N-481 requires an evaluation to demonstrate the safety and serviceability of the pump casings. The evaluation for the Davis-Besse RCPs required by Code Case N-481 is documented in Structural Integrity Associates (SIA) report SIR-99-040 [Reference A.2-18]. This evaluation assumed a quarter thickness flaw, with length six times its depth, and showed that the flaw will remain stable considering the stresses and material properties of the pump casing. To determine stability of the postulated flaw, a fracture mechanics evaluation was performed that included a fatigue crack growth analysis to demonstrate that a small initial assumed flaw (10 percent through-wall),

corresponding to the acceptance standards of ASME Code, Section Xl, Subarticle IWB-3500, would not grow to quarter thickness during plant life. There are two potential time-dependencies in the Code Case N-481 evaluation.

1.

The fracture toughness of the cast austenitic stainless steel is not time dependent as the Davis-Besse ASME Code Case N-481 analysis used a lower bound fracture toughness of 139 ksiin that bounds the saturate saturation fracture toughness of the Davis-Besse material.

Enclosure L-11-317 Page 40 of 49

2.

The fatigue crack growth analysis is based on design cycles for a 40 year plant life and therefore, is a TLAA requiring analysis and disposition for license renewal.

With respect to Item No. I above, the saturation fracture toughness was determined usinq the methodoloqy outlined in NUREG/CP-0119, Volume 2, pages 151-178, "Proceedings of the U.S. Nuclear Regulatory Commission, 19'n Water Reactor Safety Information Meeting held at Bethesda, MD.

October 28-30, 1991," and considering all available certified material test reports (CMTRs) for the base material and welds of the Davis-Besse RCP casings. The saturation fracture toughness value of 139 ksi'/in was the minimum calculated for all the CMTRs considered in the evaluation. This minimum saturation fracture toughness value has since been calculated using NUREG/CR-4513, Revision 1.

"Estimation of Fracture Toughness of Cast Stainless Steels During Thermal Aging in L WR Systems." Using the methodology and correlation in this NUREG results in the same minimum saturation fracture toughness value for the pump casings.

The fracture toughness for welds considering thermal aging has also been presented in NUREG/CR-6428, "Effects of Thermal Aging on Fracture Toughness and Charpy-Impact Strength of Stainless Steel Pipe Welds." A conservative Ji, fracture toughness value of 40 KJim2 based on the absolute minimum of all available data is provided in this document for aaed stainless steel welds: this Ji, fracture touqhness value translates to 80 ksilin. This conservative fracture toughness value still bounds the calculated total applied stress intensity factors calculated in Table 4-5 of SIR-99-040, Revision 1, indicating that the conclusions of SIR-99-040, Revision 1, are unchanged even if the methodology outlined in NUREG-CR-6428 is used for the Davis-Besse pump casing welds.

With respect to Item No. 2 above, -Tthe fatigue crack growth analysis assumed an initial flaw size corresponding to the acceptance standards of ASME Code Section XI and considered all the significant plant transients. This analysis examined the design cycles and determined there were 240 cycles that were significant to flaw growth in the RCPs. Then 2000 cycles were conservatively analyzed, and flaw growth (initial 10 percent assumed through-wall had grown only to 15 percent through-wall) remained well below the quarter thickness postulated flaw. The analyzed cycles of 2000 bound the 60-year projected cycles shown in LRA Table 4.3-1 and therefore, the fatigue crack growth TLAA associated with the ASME Code Case N-481 evaluation will remain valid for the period of extended operation in accordance with 10 CFR 54.21 (c)(1 )(i).

Enclosure L-11-317 Page 41 of 49 Affected LRA Section Table A-1 LRA Page No.

Page A-65 Affected Paragraph and Sentence Commitment No. 20 In response to Supplemental RAI OIN-380 regarding Structures Monitoring Program enhancements, license renewal future Commitment 20 in LRA Table A-I, "Davis-Besse License Renewal Commitments," is revised to include three new bulleted commitments as follows:

Table A-1 Davis-Besse License Renewal Commitments Item IRelated LRA ComitmntImplementation Source Section No,/

Number Commitment Schedule comments Comments 20 0 Monitor elastomeric vibration isolators and structural sealants Prior to LRA A.1.39 for cracking, loss of material, and hardening.

April 22, 2017 and B.2.39 Supplement visual inspection of elastomeric vibration isolation elements by feel to detect hardening if the vibration isolation FENOC Responses to function is suspect.

Letters NRC RAIs

• Identify that:

L-11-153, L-1 1-237L B.2.39-4, o loose bolts and nuts and cracked high strength bolts are not L-1 1-92, B.2.39-5, acceptable unless accepted by engineering evaluation; and B.2.39-6 and o structural sealants are acceptable if the observed loss of L-11-317 B.2.39-7 material, cracking, and hardening will not result in loss of from sealing; and, NRC Letter dated o elastomeric vibration isolation elements are acceptable if April 5, 2011, there is no loss of material, cracking, or hardening that could lead to the reduction or loss of isolation function.

RAIs B.2.39-11

Enclosure L-11-317 Page 42 of 49 Table A-1 Davis-Besse License Renewal Commitments Item Implementation Related LRA Number Commitment Schedule Source Section No./

Comments and 3.5.2.3.12-4 from NRC Letter dated July 21, 2011, Supplemental RAI B.2.39-11 from telecom held with the NRC on September 13, and Supplemental RAI OIN-380 from Region Ill IP-71002 Inspection

Enclosure L-11-317 Page 43 of 49 Affected LRA Section LRA Page No.

Affected Para-graph and Sentence Table A-1 Page A-69 Commitment No. 25 In response to RAI 2.3.3.18-4, license renewal future Commitment No. 25 is no longer needed and is revised to read "Not used," as follows:

Table A-1 Davis-Besse License Renewal Commitments Item IImplementation Related LRA Number Commitment Sheme Source Section No./

Comments 25 FEO cmit o rate a pro ventive m~aintnanco tackt Ap4I-22,-20-1-7 LA2.33.44 poiodically roplaco tho Iot-dow coolors (D9 E21 1 & 2) at a sot Ntuensed Not used.

Enclosure L-11-317 Page 44 of 49 Affected LRA Section LRA Page No.

Affected Paragraph and Sentence Table A-1 Page A-69 New Commitment No. 25 In response to RAI 3.1.2.2.16-1, new license renewal future Commitment No. 25, revised to read "Not used" in response to RAI 2.3.3.18-4, above, is revised to include a new license renewal future commitment as follows:

Table A-1 Davis-Besse License Renewal Commitments e

IRelated LRA Item Commitment Implementation Source Section No./

Number Schedule Comments 25 Enhance the Steam Generator Tube Integrity Program to:

Prior to LRA A. 1.38 Include enhanced visual (EVT-1 or equivalent) examinations to April 22, 2017 and B.2.38 monitor for cracking of the steam generator tube-to-tubesheet welds (Alloy 600). The weld inspection sample size will include 20 percent of the subiect weld population or a maximum of 25, FENOC Response to whichever is less. In this case the maximum of 25 applies since Letter NRC RAI the weld population for the two steam generators is greater than L-11-317 3.1.2.2.16-1 60, 000. Welds included in the inspection sample will be from scheduled for examination in each 10-year period that occurs NRC Letter during the period of extended operation. Unacceptable dated inspection findings will be evaluated by the Corrective Action September 22, Program using criteria in accordance with Section X1 of the 2011 ASME Code. Should the steam generators be replaced in the future with a desiqn such that the tube-to-tubesheet welds are fabricated of Alloy 690-TT material, the examinations will no

Enclosure L-11-317 Page 45 of 49 Table A-1 Davis-Besse License Renewal Commitments Related LRA Item Commitment Implementation Source Section No./

Number Schedule Cmet Comments Ion er be required.

ANet-1Ued7

Enclosure L-11-317 Page 46 of 49 Affected LRA Section LRA Page No.

Affected Paragraph and Sentence Table B-2 Page B-22 1 Row In response to 3.1.2.2.16-1, the "Steam Generator Tube Integrity Program" row of Table B-2, "Consistency of Davis-Besse Aging Management Programs with NUREG-1801," is revised to read as follows:

Table B-2 Consistency of Davis-Besse Aging Management Programs with NUREG-1801 (continued)

Consistent Consistent wt New /

with NUth 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-11-317 Page 47 of 49 Affected LRA Section LRA Page No.

Affected Paragraph and Sentence B.2.38 Page B-151 Program Description subsection, new 4 th paragraph; NUREG-1801 Consistency subsection, revised sentence; and, Enhancements subsection, new enhancements In response to RAI 3.1.2.2.16-1, LRA Section B.2.38, "Steam Generator Tube Integrity Program," a new fourth paragraph is added to subsection "Program Description," the "NUREG-1801 Consistency" subsection is revised, and new enhancements are added to the "Enhancements" subsection, to read as follows:

B.2.38 STEAM GENERATOR TUBE INTEGRITY PROGRAM Program Description In addition, cracking due to PWSCC will be managed for the steam generator tube-to-tubesheet welds (Alloy 600) by a combination of the PWR Water Chemistry Program and the Steam Generator Tube Integrity Program. The PWR Water Chemistry Program controls peak levels of various contaminants (e.g.,

dissolved oxygen, chlorides, fluorides, and sulfates) below the system-specific limits that can accelerate crackinq for nickel-alloy components. The Steam Generator Tube Integrity Program will include enhanced visual (EVT-1 or equivalent) examinations to monitor for cracking of the steam generator tube-to-tubesheet welds. The weld inspection sample size will include 20 percent of the subiect weld population or a maximum of 25, whichever is less. In this case the maximum of 25 applies since the weld population for the two steam generators is greater than 60,000. Welds included in the inspection sample will be scheduled for examination in each 10-year period that occurs during the period of extended operation. Unacceptable inspection findings will be evaluated by the Corrective Action Program using criteria in accordance with Section XI of the ASME Code.

Should the steam generators be replaced in the future with a design such that the tube-to-tubesheet welds are fabricated of Alloy 690-TT material, the examinations will no longer be required.

NUREG-1801 Consistency The Steam Generator Tube Integrity Program is an existing Davis-Besse program that, is-with enhancement, will be consistent with the 10 elements of an

Enclosure L-11-317 Page 48 of 49 effective aging management program as described in NUREG-1801,Section XI.M19, "Steam Generator Tube Integrity."

Enhancements The followinq enhancement will be implemented in the identified program elements prior to the period of extended operation.

• Scope, Parameters Monitored or Inspected, Detection of Aging Effects, Acceptance Criteria The Steam Generator Tube Integrity Pro-gram will include enhanced visual (EVT-1 or equivalent) examinations to monitor for crackinq of the steam generator tube-to-tubesheet welds (Alloy 600). The weld inspection sample size will include 20 percent of the subiect weld population or a maximum of 25, whichever is less. In this case the maximum of 25 applies since the weld population for the two steam generators is greater than 60,000. Welds included in the inspection sample will be scheduled for examination in each 10-year period that occurs durinq the period of extended operation. Unacceptable inspection findings will be evaluated by the Corrective Action Program using criteria in accordance with Section X1 of the ASME Code. Should the steam generators be replaced in the future with a design such that the tube-to-tubesheet welds are fabricated of Alloy 690-TT material, the examinations will no Ionaer be reguired.

Enclosure L-11-317 Page 49 of 49 Affected LRA Section LRA Page No.

Affected Paragraph and Sentence B.2.39 Page B-156 Enhancements subsection, 3 new enhancements In response to Supplemental RAI OIN-380 regarding Structures Monitoring Program enhancements, LRA Section B.2.39, "Structures Monitoring Program,"

three new enhancements are added to the "Enhancements" subsection, to read as follows:

• Parameters Monitored or Inspected The program procedure will be enhanced to require that elastomeric vibration isolators and structural sealants are monitored for cracking, loss of material, and hardening.

• Detection of Aging Effects The pro-gram procedure will be enhanced to require that visual inspection of elastomeric vibration isolators will be supplemented by feel to detect hardening if the vibration isolation function is suspect.

" Acceptance Criteria The program procedure will be enhanced to state that:

o loose bolts and nuts, and cracked high strength bolts are not acceptable unless accepted by engineering evaluation:

o structural sealants are acceptable if the observed loss of material, cracking, and hardening will not result in loss of sealing: and, o elastomeric vibration isolation elements are acceptable if there is no loss of material, cracking, or hardening that could lead to the reduction or loss of isolation function.