L-11-218, Reply to Requests for Additional Information for the Review of the Davis-Besse Nuclear Power Station, Unit No. 1, License Renewal Application and License Renewal Application Amendment No. 12

Reply to Requests for Additional Information for the Review of the Davis-Besse Nuclear Power Station, Unit No. 1, License Renewal Application and License Renewal Application Amendment No. 12

ML11208C274
Person / Time
Site:	Davis Besse
Issue date:	07/22/2011
From:	Allen B FirstEnergy Nuclear Operating Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
L-11-218, TAC ME4640
Download: ML11208C274 (80)

Text

FENOC ""-A% 5501 North State Route 2 FirstEnergyNuclear Operating Company Oak Harbor.Ohio 43449 Barry S. Allen 419-321-7676 Vice President- Nuclear Fax: 419-321-7582 July 22, 2011 L-11-218 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 Requests 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. 12 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 letters dated June 20, 2011 (ML11167A171) and May 2, 2011 (ML111170204), the Nuclear Regulatory Commission (NRC) requested additional information to complete its review of the License Renewal Application (LRA).

Attachment 1 provides the FENOC reply to the NRC request for additional information (RAI) in the letter dated June 20, 2011. RAI B.2.1-2 was discussed during a telephone conference with Mr. Samuel Cuadrado de Jesus, NRC Project Manager, on July 19, 2011, and it was agreed that the RAI response would be withheld pending further review by NRC.

Attachment 2 provides the FENOC reply to three additional NRC RAls. RAI 4.1-1, originally in NRC letter dated May 2, 2011, was discussed with Mr. Samuel Cuadrado de Jesus, NRC Project Manager, on June 14, 2011, and it was mutually agreed to defer the response to a later date; the response to RAI 4.1-1 is included in Attachment 2. Attachment 2 also provides revised FENOC responses to NRC RAI B.2.18-1 and RAI B.2.31-1, as discussed during a telephone conference with the NRC on June 30, 2011.

The Enclosure provides Amendment No. 12 to the DBNPS LRA.

A 4S AILz

Davis-Besse Nuclear Power Station, Unit No. 1 L-1 1-218 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 July ZZ., 2011.

Sincerely, Barry S. Allen Attachments:

1. Reply to Request for Additional Information for the Review of the Davis-Besse Nuclear Power Station, Unit No. I (DBNPS), License Renewal Application, Sections 3.1, A.1 and B.2

2. Reply to Request for Additional Information for the Review of the Davis-Besse Nuclear Power Station, Unit No. 1 (DBNPS), License Renewal Application, Sections 4.1 and B.2

Enclosure:

Amendment No. 12 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 1 L-11-218 Reply to Request for Additional Information for the Review of the Davis-Besse Nuclear Power Station, Unit No. 1, License Renewal Application, Sections 3.1, A.1 and B.2 Page 1 of 18 The following information is provided in response to the NRC request for additional information (RAI) by letter dated June 20, 2011. The NRC request is shown in bold text followed by the FENOC response.

Question RAI B.2.34-1

Background:

The preventive actions program element of Generic Aging Lessons Learned (GALL), Rev. 2, aging management program (AMP) XI.M3, "Reactor Head Closure Stud Bolting," references the guidance outlined in Regulatory Guide (RG) 1.65, Materials and Inspections for Reactor Vessel Closure Studs," and NUREG-1 339, "Resolution of Generic Safety Issue 29: Bolting Degradation or Failure in Nuclear Power Plants." AMP XI.M3 states that one of the preventive measures that can reduce the potential for stress-corrosion cracking includes using bolting material for closure studs that has an actual measured yield strength less than 150 ksi.

During its audit, the U.S. Nuclear Regulatory Commission (NRC or the staff) noted that the FirstEnergy Nuclear Operating Company's (FENOC or the applicant) program basis document for its Reactor Head Closure Studs Program states that the reactor head closure studs and nuts are manufactured from SA-540, Grade 23 material.

Issue:

License renewal application (LRA) Section B.2.34 and the applicant's program basis document do not include the preventive action of using stud materials with an actual measured yield strength level less than 150 ksi. The staff needs to confirm the actual measured yield strength of the applicant's reactor head closure stud material to determine whether the applicant's program is adequate to manage stress-corrosion cracking.

Request:

The staff requests the following information:

1) Clarify whether the actual measured yield strength of the reactor head closure stud material is less than 150 ksi. If the reactor head closure stud material has a measured yield strength level greater than or equal to 150 ksi, justify the adequacy of the AMP to manage stress-corrosion cracking in the high-strength material.

L-11-218 Page 2 of 18

2) Clarify if preventive actions will be added to the Reactor Head Closure Studs Program that would preclude the future use of replacement closure stud bolting fabricated from material with actual measured yield strength greater than or equal to 150 ksi. If not, and in view of the greater susceptibility of the studs for stress-corrosion cracking, describe any preventative actions to avoid exposure of the studs to environments conducive to stress-corrosion cracking. Otherwise, justify why preventative measures to mitigate stress-corrosion cracking of high strength studs will not be required.

RESPONSE RAI B.2.34-1

1. As confirmed by the certificate of material test report (CMTR), the actual measured yield strength ranges from 151 to 159 ksi, and tensile strength ranges from 166 to 171 ksi for the Davis-Besse reactor head closure studs. The Davis-Besse stud material is SA-540 Grade B-23. As provided in Regulatory Guide 1.65, this material when tempered to a maximum tensile strength of 170 ksi, is relatively immune to stress corrosion cracking (SCC). In addition, the Reactor Head Closure Studs Program provides for examination of the reactor vessel stud assemblies in accordance with the examination and inspection requirements specified in the ASME B&PV Code,Section XI, Subsection IWB (1995 Edition through the 1996 Addenda) and approved ASME Code Cases. Specifically, each stud is volumetrically examined once per each 10-year Inservice Inspection Interval. No unacceptable indications were noted in these examinations.

Reactor Head Closure Studs Program preventative measures to mitigate SCC are listed as follows:

a. There are no metal platings applied to the closure studs, nuts, or washers.

b. A manganese-phosphate coating was applied to the studs, nuts and washers during fabrication to act as a rust inhibitor.

c. An enhancement to the program provides for selection of an alternate stable lubricant that is compatible with the fastener material and the environment. A specific precaution against the use of compounds containing sulfur (sulfide), including molybdenum disulfide (MoS2), as a lubricant for the reactor head closure stud assemblies will be included in the program.

2. An enhancement will be added to the Reactor Head Closure Studs Program to preclude the future use of replacement closure stud bolting fabricated from material with actual measured yield strength greater than or equal to 150 ksi except for use of the existing spare reactor head closure stud bolting.

L-11-218 Page 3 of 18 The exception to allow future use of the existing spare reactor head closure stud bolting (2 each) is justified based on Davis-Besse plant-specific operating experience of over 30 years that has not experienced SCC of the reactor head closure stud bolting. The existing spare bolting, if used as a future replacement, would experience less than 30 years of service to the end of the period of extended operation.

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

Question RAI B.2.9-3

Background:

"Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants" (SRP-LR), Section A 1.2.3.4, "detection of aging effects," states that the parameters to be monitored include aspects such as frequency and sample size. This program element states that the basis for the inspection population and sample size should consider the environment, locations most susceptible to the aging effect, and include provisions for expanding the sample size when degradation is detected in the initial sample. The SRP-LR also states that the applicant should provide a justification, including codes and standards referenced, that the technique and frequency are adequate to detect the aging effects before a loss of component's intended function.

The GALL Report recommends the use of AMP XI.M20 "Open-Cycle Cooling Water System" for materials included within the scope of the plant-specific Collection, Drainage and Treatment Components Inspection Program that are exposed to raw water. In LRA Section B.2.9, Collection, Drainage and Treatment Components Inspection Program, under the "detection of aging effects" program element, the applicant stated that if opportunistic inspections have not occurred prior to the period of extended operation then a focused inspection, inclusive of each material in the scope of the program will be performed. The application further states that any evidence of degradation that could lead to loss of a component intended function will be evaluated through the corrective action program to determine the need for subsequent inspections, expansion, and for monitoring and trending the results.

Issue:

Although the systems referenced in the Collection, Drainage and Treatment Components Inspection Program are in some cases different than those listed in the GALL Report for which AMP XI.M20 "Open-Cycle Cooling Water System" is

L-11-218 Page 4 of 18 recommended, it is clear that for the material and environment combination stated in the LRA, a periodic inspection program is recommended.

The staff lacks sufficient information to determine if the inspections conducted for the Collection, Drainage and Treatment Components Inspection Program will be periodic or one-time, or the basis for the inspection frequency, if in the absence of evidence of degradation during a planned inspection, no further inspections are conducted. The staff also lacks sufficient information to find the basis for the inspection size and locations acceptable.

Request:

State the basis for why a one-time inspection would be sufficient for managing the effects of aging for collection, drainage, and treatment components or revise the Collection, Drainage and Treatment Components Inspection Program to ensure that periodic inspections are performed. State the basis for the sample size, the selection factors for the "most susceptible" materials and locations, the frequency to be used during the period of extended operation, and the percentage increase in sample size should degradation be detected.

RESPONSE RAI B.2.9-3 The Collection, Drainage and Treatment Components Inspection Program is revised to perform periodic inspections. These inspections will ensure that the existing environmental conditions are not causing cracking, loss of material, or reduction in heat transfer that could result in a loss of component intended functions.

Periodic inspections will be conducted on a representative sample of piping and components on a 10-year interval, with the first inspection taking place within the 10-year period prior to the period of extended operation.

A representative sample of the system and component population will be inspected. The sample size is 20 percent of the population (defined as components having the same material, environment, and aging-effect combination) or a maximum of 25 components.

The sample population will be determined by engineering evaluation, and, where practical, focused on the (bounding or lead) components considered most susceptible to aging degradation due to time in service, the severity of the operating conditions, and the lowest design margin.

Evidence of degradation that could lead to a loss of component intended function will be documented and evaluated through the Corrective Action Program to determine the need for subsequent inspections, expansion of the sample size, and for monitoring and trending the results. If degradation that could lead to a loss of component intended function is detected, sample size will be increased by 20 percent of the population

L-11-218 Page 5 of 18 (defined as components having the same material, environment, and aging effect combination) or a maximum of 25 components.

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

Question RAI B.2.9-4

Background:

SRP-LR Section A.1.2.3.10, "operating experience," states that "Additionally, an applicant should commit to a review of future plant-specific and industry operating experience for new programs to confirm their effectiveness." In LRA Appendix A.3, Table A-I, License Renewal Commitment List the new Collection, Drainage and Treatment Components Inspection Program does not include a commitment to perform a review of future operating experience to confirm the effectiveness of this program.

Issue:

This program's LRA commitment list is not consistent with the current staff position as stated within the SRP-LR, Revision 2 concerning reviews of future operating experience for new AMPs.

Request:

Revise LRA Appendix A.3, Table A-I, License Renewal Commitment List, item no. 4 for the Collection, Drainage and Treatment Components Inspection Program to include a commitment to perform a future review of operation experience to confirm the effectiveness of this program or justify why such a review is not necessary.

RESPONSE RAI B.2.9-4 By letter dated June 24, 2011 (ML11180A060), in response to RAI B.1.4-1, FENOC provided license renewal future commitment number 43 to "[e]nsure that the current station operating experience review process includes future reviews of plant-specific and industry operating experience to confirm the effectiveness of the license renewal aging management programs, to determine the need for programs to be enhanced, or indicate a need to develop new aging management programs." Therefore, a separate operating experience commitment for the Collection, Drainage and Treatment Components Inspection Program is not necessary.

L-1 1-218 Page 6 of 18 Question RAI B.2.9-5

Background:

SRP-LR Section A.1.2.3.3, "parameters monitored or inspected," states that an applicant should provide a link between the parameter(s) that will be monitored and how the monitoring of these parameters will ensure adequate aging management. In LRA Section B.2.9, "Collection, Drainage and Treatment Components Inspection Program," under the "parameters monitored or inspected" program element, the applicant stated that parameters monitored or inspected are directly related to degradation of the components under review.

Issue:

This program does not provide the details for what parameters, such as wall thickness and surface degradation will be monitored and used to ensure adequate aging management will be completed.

Request:

State what parameters will be linked to detecting the following: (a) loss of material; (b) cracking; and (c) a reduction in heat transfer during the visual inspection. State the basis for detecting loss of material on inaccessible surfaces (e.g., tank bottoms sitting on concrete) using a visual inspection, or revise the program to include volumetric inspections that are capable of adequately managing this aging effect.

RESPONSE RAI B.2.9-5 Inspection parameters for metallic components include the following:

Aging Effect Parameter(s) Monitored Inspection Method Loss of material Surface condition:

" corrosion and material parameters wastage;

• leakage from or onto internal surfaces; or Visual (VT-1 or equivalent)

" worn, flaking, or oxide-coated surfaces Cracking Surface condition, cracks Reduction in heat transfer Tube fouling

L-1 1-218 Page 7 of 18 LRA Section B.2.9 is revised to state that volumetric inspections by qualified personnel will detect a loss of material on inaccessible surfaces (e.g., tank bottoms sifting on concrete).

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

Question RAI B.2.30-1

Background:

The GALL Report Rev. 2, AMP XI.M32, "One-Time Inspection," states in the "detection of aging effects program element" that for components managed by the AMP XI.M2, "Water Chemistry;" AMP XI.M30, "Fuel Oil Chemistry;" and AMP XI.M39, "Lubricating Oil Analysis;" a representative sample size is 20 percent of the population (defined as components having the same material, environment, and aging effect combination) or a maximum of 25 components. LRA AMP B.2.30 states that the sample population will be determined by engineering evaluation, and where practical, will be focused on the (bounding or lead) components considered most susceptible to aging degradation due to time in service, the severity of the operating conditions, and the lowest design margin.

Issue:

Given that the GALL Report. Rev. 2, represents the current staff position on the sample size for the "One-Time Inspection" Program, LRA Section B.2.30 does not provide enough information for the staff to determine if the sample size for this program is consistent with the GALL Report AMP XI.M32.

Request:

State the planned sample size for the One-Time Inspections of the Pressurized-Water Reactor (PWR) Water Chemistry, Fuel Oil Chemistry, and Lubricating Oil Analysis Programs. If the sample size is less than 20 percent of the population (defined as components having the same material, environment, and aging effect combination) or a maximum of 25 components, then state the basis for why the sample size will be representative of aging effects in the systems, and will be sufficient to verify the system-wide effectiveness of the chemistry programs.

RESPONSE RAI B.2.30-1 LRA Section B.2.30 is revised to address the planned sample size for the One-Time Inspections of the components managed by the Pressurized-Water Reactor (PWR)

Water Chemistry, Fuel Oil Chemistry, and Lubricating Oil Analysis Programs. The sample size is 20 percent of the population (defined as components having the same material, environment, and aging effect combination) or a maximum of 25 components.

L-11-218 Page 8 of 18 The sample population will be determined by engineering evaluation, and where practical, will be focused on the (bounding or lead) components considered most susceptible to aging degradation due to time in service, the severity of the operating conditions, and the lowest design margin. The inspections must occur within the 10 year period prior to the period of extended operation to be credited for the program.

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

Question RAI B.2.30-2

Background:

SRP-LR Rev. 2, Table 3.0-1, "FSAR [Final Safety Analysis Report] Supplement for Aging Management of Applicable Systems," states that GALL Report AMP XI.M32, "One-Time Inspection" Program, cannot be used for structures or components with known age-related degradation mechanisms or when the environment in the period of extended operation is not expected to be equivalent to that in the prior 40 years, and that periodic inspections should be proposed in these cases.

SRP-LR Section 3.0.1 states that the FSAR Supplement should also contain a commitment to implement the LRA AMP enhancement prior to the period of extended operation. Title 10 of the Code of Federal Regulations (10 CFR) 54.21(d) states that the FSAR supplement must contain a summary description of the program and the activities for managing the effects of aging. In addition, SRP-LR 3.3.2.4 states that the summary description of the programs and activities for managing the effects of aging for the period of extended operation in the FSAR Supplement should be sufficiently comprehensive such that later changes can be controlled by 10 CFR 50.59, and the description should contain information associated with the bases for determining that aging effects will be managed during the period of extended operation.

In its response to RAIs 3.3.2.2.5-1, 3.3.2.71-2, B.2.8-1, and B.2.18-1, dated May 24, 2011, LRA Section A.1.30, "One-Time Inspection," was revised; however, the change did not include the above wording from SRP-LR Table 3.0-1.

Issue:

The updated (UFSAR) supplement does not reflect change that occurred in Revision 2 to the SRP-LR Table 3.0-1, as stated above. The staff believes that this information is associated with the bases for determining that the aging effects for buried in-scope components will be effectively managed during the period of extended operation. The staff also believes that this information should be explicitly stated in the FSAR supplement to ensure that the licensing basis for the period of extended operation is clear.

L-11-218 Page 9 of 18 Request:

Revise LRA Section A.1.30 to be consistent with and provide the equivalent information as stated within SRP-LR, Rev. 2, Table 3.0-1 GALL Report AMP XI.M32, "One-Time Inspection" Program.

RESPONSE RAI B.2.30-2 LRA Section A.1.30 is revised to state this program cannot be used for structures or components with known age-related degradation mechanisms or when the environment in the period of extended operation is not expected to be equivalent to that in the prior 40 years. Periodic inspections should be proposed in these cases.

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

Question RAI B.2.30-3

Background:

SRP-LR Revision 2, A.1.2.3.10.3 states that, "Additionally, an applicant should commit to a review of future plant-specific and industry operating experience for new programs to confirm their effectiveness."

In LRA Appendix A.3, Table A-I, "License Renewal Commitment List," the new One-Time Inspection Program does not include a commitment to perform a future review of operating experience to confirm the effectiveness of this program.

Issue:

The new One-Time Inspection Program's LRA commitments are not consistent with the current staff position as stated within the SRP-LR, Rev. 2, concerning reviews of future operating experience for new programs.

Request:

Revise LRA Appendix A.3, Table A-I, "License Renewal Commitment List," for the One-Time Inspection Program to include a commitment to perform a future review of operating experience to confirm the effectiveness of this program or state why such a review is not necessary.

L-11-218 Page 10 of 18 RESPONSE RAI B.2.30-3 By letter dated June 24, 2011 (ML11180A060), in response to RAI B.1.4-1, FENOC provided license renewal future commitment number 43 to "[elnsure that the current station operating experience review process includes future reviews of plant-specific and industry operating experience to confirm the effectiveness of the license renewal aging management programs, to determine the need for programs to be enhanced, or indicate a need to develop new aging management programs." Therefore, a separate operating experience commitment for the One-Time Inspection Program is not necessary.

Question RAI B.2.21-6

Background:

SRP-LR Section A 1.2.3.10 states, in part, that for new AMP that have yet to be implemented at an applicant's facility, the programs have not yet generated any operating experience (OE). However, there may be other relevant plant-specific OE at the plant or generic OE in the industry that is relevant to the AMP's program elements even though the OE was not identified as a result of the implementation of the new program. Thus, for new programs, the applicant may need to consider the impact of relevant OE that results from past implementation of its existing AMPs that are existing programs and the impact of relevant generic OE on developing program elements.

As part of RAI B.2.21-1, the staff requested the applicant provide a summary of their evaluation of recently identified industry operating experience and any plant-specific operating experience concerning inaccessible low voltage power cable failures within the scope of license renewal. The staff also requested the applicant provide an evaluation showing how the Non-EQ Inaccessible Medium-Voltage Program test and inspection frequencies, including event driven inspections, incorporate recent industry and plant specific operating experience for both inaccessible low and medium voltage power cable.

Issue:

In its RAI response dated May 5, 2011, the applicant referenced their response to Generic Letter (GL) 2007-01, "Inaccessible or Underground Power Cable Failures that Disable Accident Mitigation Systems or Cause Plant Transients," dated May 8, 2007. The applicant did not provide additional operating experience for inaccessible low and medium voltage power cable subsequent to the applicant's GL response.

L-11-218 Page 11 of 18 Request:

Provide a summary of inaccessible low and medium voltage cable operating experience (both testing and operating) subsequent to your May 8, 2007 response to GL 2007-01.

RESPONSE RAI B.2.21-6 Subsequent to the May 8, 2007, response to GL 2007-01, diagnostic testing continues for both low voltage and medium voltage cables. The cable replacement program for wetted medium voltage cables also continues. For low voltage cables, cables are replaced prior to loss of function based upon diagnostic test results.

Low Voltage Cables The condition of the cable insulation (Polarization Index) is measured utilizing equipment such as the Motor Circuit Evaluation (MCE) to test de-energized motors for resistance-to-ground, capacitance-to-ground, phase-to-phase resistance, and phase-to-phase inductance. The 480 volt cables are diagnostically tested such that changes in cable insulation performance can be identified and the cables can be replaced prior to failure. Reduced insulation resistance has been observed on some 480 volt cables.

The Polarization Index for cable 2PBF1230A, 480 volt feed to EDG fuel oil tank 2 transfer pump motor, is less than desired. The cable is scheduled to be replaced.

The Polarization Index for cable 2PBF1205A, 480 volt feed to auxiliary feed pump vent fan motor, is less than desired. The cable is scheduled to be replaced.

Regarding low voltage cable operating experience, a search of the records in the FENOC Corrective Action Program did not identify any in-service cable failures.

Medium Voltage Cables Cable 3PACD06A, 4,160 volt feed to the service water pump 3 motor, was being replaced as part of the cable replacement program for wetted medium voltage cables. A very-low frequency insulation dissipation factor (tan-delta) test was performed on the old cable prior to removal. At the end of the tan delta test for each phase, test voltage was increased to the "maintenance withstand voltage" of 7,000 volts (at 0.1 Hertz). The cable A phase failed about four minutes into the 15-minute withstand test. The cable B and C phases completed the tan delta and withstand test without incident. The cable was replaced.

Various other medium voltage cables have been tan-delta tested with acceptable results.

L-11-218 Page 12 of 18 Regarding medium voltage cable operating experience, cable BPAD211B, 4,160 volt feed to station transformer 2, failed in service. Cable insulation and jacket damage was clearly visible at the fault locations. The cable was replaced.

Question RAI B.2.21-7 Back-ground:

SRP-LR Section 3.0.1 states, in part, that each LRA will provide an FSAR Supplement which defines the changes to the FSAR that will be made as a condition of a renewed license. The FSAR Supplement defines the AMPs the applicant is crediting to satisfy 10 CFR 54.21 (a)(3). SRP-LR Table 3.0.1 states (along with an inspection performed at least annually and event driven inspections) that the inspection frequency for water collection is established and performed based on plant-specific operating experience with cable wetting or submergence.

As part of RAI B.2.21-1, the staff requested the applicant to explain how DBNPS will manage the effects of aging on inaccessible low voltage power cables within the scope of license renewal with consideration of recent industry operating experience and applicable plant-specific operating experience including an assessment of the program elements and the USAR summary description for the Inaccessible Power Cables Not Subject to 10 CFR 50.49 EQ Requirements Program. The applicant's RAI response indicates that the USAR will be revised to include the change. The applicant's RAI response did revise Commitment No. 11 to include the above change.

Issue As part of the applicant's response to RAI B.2.21-1 the applicant revised the LRA USAR summary description for the Inaccessible Power Cables Not Subject to 10 CFR 50.49 EQ Requirements Program but did not state that the inspection frequency for water collection is established and performed based on plant-specific operating experience with cable wetting or submergence consistent with SRP-LR Table 3.0.1 and GALL AMP XI.E3.

Request:

Explain why the USAR summary description provided in the response to RAI B.2.21-1 does not include the provision that manhole inspection frequencies will be based on plant-specific operating experience consistent with SRP-LR Table 3.0.1 and GALL AMP XI.E3.

L-1 1-218 Page 13 of 18 RESPONSE RAI B.2.21-7 LRA Sections A.1.21 and B.2.21 are revised to state that the inspection frequency for water collection in manholes is established and performed based on plant-specific operating experience with cable wetting or submergence.

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

Question RAI 3.1.2.2-2

Background:

LRA Table 3.1.2-2 indicates the reactor vessel internals components made of cast austenitic stainless steel (CASS) subject to reduction in fracture toughness and managed by the PWR Vessel Internals Program. These CASS components are the following: (1) Incore guide tube assembly spider in the core support assembly (CSA); (2) Plenum control rod guide tube (CRGT). spacer casting; (3) CSA vent valve assembly valve body; and (4) Plenum cylinder reinforcing plate.

LRA Section B.2.32 states that the PWR Vessel Internals Program is based on the examination requirements provided in Electric Power Research Institute (EPRI) Topical Report 1016596, "Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227, Rev. 0),"

along with the implementation guidance described in NEI 03-08. The staff noted that MRP-227, Rev. 0, which is referenced in the GALL Report Rev. 2, categorizes the reactor vessel internals components based on the following functional groups: primary, expansion, existing programs, and no additional measures.

MRP-227 also specifies relevant examination methods and coverage for the expansion group components based on the examination findings of the primary group components.

In addition, GALL Report, Rev. 2, AMP XI.M16A and MRP-227, Rev. 0, Tables 3-1, 4-1 and 4-4 indicate that, in B&W plants. the following CASS vessel internals are the primary group components to be managed for loss of fracture toughness:

(1) core support shield (CSS) cast outlet nozzles, (2) CSS vent valve discs, and (3) incore monitoring instrumentation (IMI) guide tube assembly spiders (accessible top surfaces). MRP-227, Rev. 0, also indicates that these primary group components have link relationships with CRGT spacer castings (accessible surfaces at four screw locations), which are the associated expansion group components.

L-1 1-218 Page 14 of 18 Issue:

The staff noted that, in contrast with MRP-227, Rev. 0, LRA Table 3.1.2-2 does not clearly identify the functional groups and link relationships for the following components: (1) CSS outlet nozzles, (2) CSS vent valve discs, (3) Incore guide tube assembly spiders, and (4) CRGT spacer castings. In addition, LRA Table 3.1.2-2 does not clearly indicate the functional groups and link relationships for the following two components: (1) CSA vent valve body and (2) plenum cylinder reinforcing plate.

Request:

1. Clarify whether or not the CSS outlet nozzles and CSS vent valve discs are made of CASS.

2. Describe the functional groups for the following components: (1) CSS outlet nozzles, (2) CSS vent valve discs and (3) IMI guide tube assembly spiders (accessible top surfaces), and (4) CRGT spacer castings (accessible surfaces at four screw locations). In addition, describe the link relationships for these components (such as primary/expansion link). If the assigned functional groups or links are not consistent with MRP-227, Rev. 0, justify why the inconsistency is acceptable to manage the reduction in fracture toughness of these components.

3. Describe the functional groups for the following two components addressed in LRA Table 3.1.2-2: (1) CSA vent valve body, and (2) plenum cylinder reinforcing plate. If existent, describe their link relationships (such as primary/expansion link) with other components. In addition, describe the assigned inspection method including frequency of the components. Also, provide the technical basis for the assigned component groups, link relationships and inspection method/frequency.

4. Revise LRA Table 3.1.2-2 and other related information in the LRA consistent with the response to this RAI.

RESPONSE RAI 3.1.2.2-2

1. As confirmed by Davis-Besse documentation and as documented in Electric Power Research Institute (EPRI) Topical Report 1016596, "Materials Reliability Program:

Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227-Rev. 0)," the core support shield (CSS) outlet nozzles and CSS vent valve discs are made of cast austenitic stainless steel (CASS).

2. In MRP-227, the reactor internals were assigned to one of the following four functional groups: Primary, Expansion, Existing Programs, and No Additional Measures components. No components for B&W plants were placed into the

L-11-218 Page 15 of 18 Existing Programs group. Reactor vessel internals for which the effects for eight postulated aging mechanisms (stress corrosion cracking, irradiation-assisted stress corrosion cracking, wear, fatigue, thermal aging embrittlement, irradiation embrittlement, void swelling, and the combination of thermal and irradiation-enhanced stress relaxation or irradiation-enhanced creep) are below the screening criteria were placed in the No Additional Measures group and require no further action. Reactor vessel internals that are highly susceptible to the effects of at least one of the eight aging mechanisms were placed in the Primary group and these components require aging management. Reactor vessel internals that are highly or moderately susceptible to the effects of at least one of the eight aging mechanisms, but for which functionality assessment has shown a degree of tolerance to those effects, were placed in the Expansion group. The schedule for implementation of aging management requirements for Expansion components will depend on the findings from the examinations of the Primary components at Davis-Besse. The aging management review results for the primary and expansion components are provided in revised LRA Table 3.1.2-2. In addition, link relationships are provided for the primary and expansion components. These link relationships are consistent with that provided in Tables 4-1 and 4-4 of MRP-227, Rev. 0.

3. See item 2, above. In addition, the inspection frequency and method for the primary and expansion components are provided in Tables 4-1 and 4-4 of MRP-227, Rev. 0.

4. LRA Tables 3.1.1 and 3.1.2-2 are revised consistent with this response. In addition, LRA Sections 2.3.1.2, 3.1.2.1.2, 3.1.2.2.6 and 3.1.2.2.15 are revised consistent with this response.

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

Question RAI B.2.36-4

Background:

The "acceptance criteria" program element of GALL Report (Revision 2) AMP XI.M33, "Selective Leaching," recommends that the acceptance criteria include no visible evidence of selective leaching or no more than a 20 percent decrease in hardness. GALL Report AMP XI.M33 also recommends that for copper alloys with greater than 15 percent zinc, the acceptance criteria is no noticeable change in color from the normal yellow color to the reddish copper color. LRA Section B.2.36 states that the selective leaching inspection will utilize approved inspection techniques to identify selective leaching, and inspection results that identify selective leaching will be entered into the Corrective Action Program.

L-11-218 Page 16 of 18 Issue:

It is not clear to the staff how the GALL Report, Rev. 2, AMP XI.M33 recommendations in the "acceptance criteria" program element are addressed in the applicant's Selective Leaching Inspection Program.

Request:

Describe how the GALL Report, Rev. 2, AMP XI.M33 recommendations in the

acceptance criteria" program element are addressed in the Selective Leaching Inspection Program. If the recommended acceptance criteria are not included, state the basis for not including these acceptance criteria in the Selective Leaching Inspection Program and propose an alternate acceptance criteria that is capable of identifying the aging effects before a loss of intended function.

RESPONSE RAI B.2.36-4 LRA Section B.2.36, "Selective Leaching Inspection," is revised to state that the acceptance criteria are no visible evidence of selective leaching or no more than a 20 percent decrease in hardness. For copper alloys with greater than 15 percent zinc, the criteria also includes no noticeable change in color from the normal yellow color to the reddish copper color.

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

Question RAI A.1.2-1

Background:

SRP-LR 3.3.2.4 states that the summary description of the programs and activities for managing the effects of aging for the period of extended operation in the FSAR Supplement should be sufficiently comprehensive such that later changes can be controlled by 10 CFR 50.59, and the description should contain information associated with the bases for determining that aging effects will be managed during the period of extended operation. In addition, 10 CFR 54.21 (d) states that the FSAR supplement must contain a summary description of the program and the activities for managing the effects of aging.

SRP-LR Tables 3.3-2 and 3.4-2 recommend that the FSAR Supplement for the Aboveground Steel Tanks Program should state that the program includes preventive measures to mitigate corrosion by protecting the external surface of steel components per standard industry practice and with sealant or caulking at the interface of concrete and component, and verification of the effectiveness of

L-11-218 Page 17 of 18 the program by measuring the thickness of the tank bottoms to ensure that significant degradation is not occurring.

Issue:

The USAR supplement, LRA Section A.1.2, does not reflect that the Aboveground Steel Tanks Inspection Program includes the above information.

Request:

Amend the USAR supplement to include statements that the Aboveground Steel Tanks Program includes preventive measures to mitigate corrosion by protecting the external surface of steel components per standard industry practice and with sealant or caulking at the interface of concrete and component, if applicable (see RAI B.2.2-3), and verification of the effectiveness of the program by measuring the thickness of the tank bottoms to ensure that significant degradation is not occurring.

RESPONSE RAI A.1.2-1 LRA Appendix A "Updated Safety Analysis Report Supplement," Section A.1.2, "Aboveground Steel Tanks Inspection Program," is revised to include preventive measures to mitigate corrosion by protecting the external surface of steel components per standard industry practice and with sealant or caulking at the interface of concrete and component, as applicable (see the response to RAI B.2.2-3 in FENOC Letter dated May 24, 2011 (ML11151A090)). LRA Section A.1.2 is also revised to include a statement that the tank bottom inspections will verify the effectiveness of the program by measuring the thickness of the tank bottoms to ensure that significant degradation is not occurring.

The Aboveground Steel Tanks Program manages aging for two steel tanks, the fire water storage tank and the diesel fuel oil storage tank. The fire water storage tank does not have sealant or caulking at the interface edge between the tank and the foundation.

Instead, the tank rests on an oiled sand pad on top of granular fill, which slopes down from the tank center to the outside edge. Therefore, the Aboveground Steel Tanks Program includes steel tank sealant or caulking for the diesel fuel oil storage tank only.

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

L-11-218 Page 18 of 18 Question RAI A.1.6-1

Background:

10 CFR 54.21(d) states that the FSAR supplement must contain a summary description of the program and the activities for managing the effects of aging.

SRP-LR Rev. 2, Section 3.1.2.5 states that the summary description of the programs and activities for managing the effects of aging for the period of extended operation in the FSAR Supplement should be sufficiently comprehensive such that later changes can be controlled by 10 CFR 50.59, and the description should contain information associated with the bases for determining that aging effects will be managed during the period of extended operation.

The SRP-LR Rev. 2, Table 3.0-1, provides an example FSAR Supplement description of GALL AMP XI.M10 "Boric Acid Corrosion," which includes:

(a) visual inspection of external surfaces that are potentially exposed to borated water leakage; (b) timely discovery of leak path and removal of the boric acid residues; (c) assessment of the damage; and (d) follow-up inspection for adequacy. In the USAR Supplement in LRA Section A.1.6, the applicant stated that the Boric Acid Corrosion Program consists of visual inspections.

Issue:

The USAR Supplement does not describe several details of the Boric Acid Corrosion Program that ensure that boric acid corrosion will not lead to degradation on the reactor coolant pressure boundary.

Request:

Revise the USAR supplement to state that the program includes activities associated with discovered evidence of boric acid leakage, including, but not limited to, determination of the principal location of leakage, removal of boric acid residues, and engineering evaluations to establish the impact on the reactor coolant pressure boundary.

RESPONSE RAI A.1.6-1 LRA Section A.1.6 and Section B.2.6 are revised to state that the Boric Acid Corrosion Program includes: (a) visual inspection of external surfaces that are potentially exposed to borated water leakage; (b) timely discovery of leak path and removal of the boric acid residues; (c) assessment of the damage; and (d) follow-up inspection for adequacy.

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

Attachment 2 L-11-218 Reply to Request for Additional Information for the Review of the Davis-Besse Nuclear Power Station, Unit No. 1, License Renewal Application, Sections 4.1 and B.2 Page 1 of 9 The following information is provided to complete the response to an NRC request for additional information (RAI) by letter dated May 2, 2011, and amend previous responses to NRC RAIs as detailed in the responses, below. The NRC request is shown in bold text followed by the FENOC response.

Question RAI 4.1-1 LRA Table 4.1-1 states that the CLB does not include any fatigue analysis for Class 1 valves, which is further discussed in LRA Section 4.3.2.3.2. LRA Section 4.3.2.3.2 states that a review of quality assurance records located the stress reports of record for each of the twelve Class 1 valves with four inch or greater diameter, but no associated fatigue analyses were identified. LRA Section 4.3.2.3.2 also states that "valve bodies were considered robust compared to the piping system in which they were located and fatigue of the attached piping was understood to bound the fatigue of the valve bodies."

USAR Table 5.2-1 identifies that the following Codes are applicable to the design of its Class A or Class 1 valves in the reactor coolant system (RCS):

" Pressurizer safety valves and pressurizer relief valve - 1968 Draft ASME Pump and Valve Code

• Pressurizer Pilot-Operated Relief Pressure Valves - 1974 ASME Section III inclusive of the Summer 1976 Addenda

• Pressurizer Spray Line Isolation Valve - 1986 ASME Section III

• Loop Isolation Valves both 2 % inches and larger and 2 inches in diameter and smaller - 1971 ASME Section III

• Other Class 1 or Class Valves both 2 1/%inches and larger and 2 inches in diameter and smaller - 1971 ASME Section III or later NRC endorsed editions of this code USAR Table 5.1-1b identifies the valves that are included in the reactor coolant pressure boundary (RCPB).

The staff noted that the terminology for valves in USAR Table 5.1-1b does not correlate to the terminology of valves in USAR Table 5.2-1, therefore, it is difficult to confirm the statements in LRA Section 4.3.2.3 without clarifications of the specific valves, including the design code, in the RCS and RCPB.

The staff has the following issues associated with the Class 1 and Class A valves listed in USAR Table 5.2-1 and in USAR Table 5.1-1b.

L-11-218 Page 2 of 9 Issue 1 - USAR Table 5.1-1b identifies several valves that are in the RCPB.

Specifically for the seal injection flow isolation valve, pump seal return isolation valve, letdown cooler inlet valve, HP injection valve, seal return isolation valve, makeup isolation valve, letdown cooler isolation valve, pressure spray control valve, low pressure injection valve and two DH removal outlet valves, the staff is unable to correlate the specific category these valves are classified under. These categories include the following, which are identified in USAR Table 5.2-1: "2 1/2 inch and larger - Loop Isolation Valve," "2 inch and smaller - Loop Isolation Valve," "2% inch and larger - Other Valves," or "2 inch and smaller - Other Valves." In addition, the staff is unable to determine whether the "pressurizer relief isolation valve" or "pressurizer pilot-operated relief valve (PORV)" in USAR Table 5.1-lb correlates to the "pressurizer pilot-operated relief isolation valve" that is listed in USAR Table 5.2-1.

Without a clear correlation between USAR Table 5.1-1b and USAR Table 5.2-1 the staff is unable to verify the specific edition of ASME Section III used for the design of these valves and determine is a fatigue analysis was required by the design code.

Request 1 -

Part A:

1) Identify the edition of ASME Section III used for the design, procurement, and installation of the following valves in USAR Table 5.1-1b: (1) the seal injection flow isolation valve; (2) the pump seal return isolation valve; (3) the letdown cooler inlet valve; (4) the HP injection valve; (5) the seal return isolation valve; (6) the makeup isolation valve; (7) the letdown cooler isolation valve; (8) the pressure spray control valve; (9) pressurizer LP injection valve; and (10) each of the DH removal outlet valves.

2) For each of these valves, justify that an It fatigue analysis was not required in accordance with NB-3545.3 and NB-3550 of the applicable ASME Code Section III edition and the provisions for performing It fatigue analysis in paragraph NB-3553.

3) If an It fatigue analysis was performed as part of the design basis for the specific valve, justify the conclusion that the It fatigue analysis does not need to be identified as a TLAA in accordance with 10 CFR 54.21(c)(1).

Part B:

1) Confirm the description for the "pressurizer relief isolation valve" in USAR Table 5.1-1b correlates to the "relief valve" in USAR Table 5.2-1.

L-11-218 Page 3 of 9

2) Confirm the description for the "pressurizer pilot-operated relief valve (PORV)" in USAR Table 5.1-1 b correlates to the "pressurizer pilot-operated relief isolation valve" in USAR Table 5.2-1.

3) If not, identify the design code that is applicable for the "pressurizer relief isolation valve" and "pressurizer pilot-operated relief valve" in USAR Table 5.1-1b.

Issue 2 (PressurizerSafety Valve and Relief Valve) - USAR Table 5.2-1 indicates that the pressurizer safety valve and relief valve were designed to the 1968 Draft ASME Pump and Valve Code. The staff noted that Sections 452 and 454 of this Code include applicable time-dependent cyclic or fatigue assessment criteria for pumps and valves.

Specifically, Section 454 of the Code includes an It parameter metal fatigue analysis (cycling loading analysis). The staff verified that Section 142 of the 1968 Draft ASME Pump and Valve Code identifies that the requirements in Section 452 and 454 needs to be performed only if the inlet nozzle size for the Class I pump or valve was greater than 4 inches diameter nominal pipe size. Section 410 of this code states that Chapter 4 procedures and analyses (including those in Sections 452 and 454) need to be performed for small bore pumps or valves (i.e. for those pump or valves with inlet nozzles less than or equal to 4 inches in nominal pipe size) if specified by the owner's design specification. The staff noted that it is possible that small bore pumps or valves could be subject to a time-dependent cyclic or fatigue assessment.

Request 2:

• Justify why an It fatigue analysis was not required for the pressurizer safety and relief valves under the provisions of the 1968 Draft ASME Pump and Valve Code as part of the design basis.

• If an It analysis was performed as part of the design basis for these valves, justify why these analyses do not need to be identified as a TLAA in accordance with 10 CFR 54.21(c)(1).

Issue 3 (PressurizerPilot-OperatedRelief Isolation Valve) - USAR Table 5.2-1 indicates that the pressurizer pilot-operated relief isolation valve (PPORIV) was designed to the 1974 edition of ASME Section III, inclusive of the 1976 Summer Addenda. The staff noted that Paragraph NB-3545.3 of this code edition required that the pressure retaining portions of these valves be analyzed for fatigue in accordance with the design rules in NB-3550. This includes the requirements for performing a time-dependent (cycle-dependent) It fatigue analysis described in NB-3553. It is not clear to the staff if an Itfatigue analysis for the PPORIV was performed in accordance with the requirements of ASME Section III, paragraph NB-3553.

L-11-218 Page 4 of 9 Request 3:

" Justify why an It fatigue analysis was not required for the PPORIV in accordance with paragraphs NB-3545.3 and NB-3550 of the 1974 Edition of the ASME Code Section III and the provisions for performing It fatigue analyses in paragraph NB-3553.

" If an It analysis was performed as part of the design basis for the PPORIV, justify the conclusion that the It fatigue analysis for the PPORIV does not need to be identified as a TLAA in accordance with 10 CFR 54.21(c)(1).

Issue 4 (PressurizerSpray Line Isolation Valve) - USAR Table 5.2-1 indicates that the pressurizer spray line isolation valve (PSLIV) was designed to the 1986 edition of ASME Section III, with no applicable Addenda. The staff noted that Paragraph NB-3545.3 of the 1986 code edition required that the pressure retaining portion of this valve be analyzed for fatigue in accordance with the design rules in NB-3550. This includes the requirements for performing a time-dependent (cycle-dependent) It fatigue analysis in NB-3553. It is not clear to the staff if an It fatigue analysis for the PSLIV was performed in accordance with the requirements of ASME Section III, paragraph NB-3553 Request 4:

• Justify why an It fatigue analysis was not required for the PSLIV in accordance with paragraphs NB-3545.3 and NB-3550 of the 1986 Edition of the ASME Code Section III, and the provisions for performing It fatigue analyses in paragraph NB-3553.

" If an It analysis was performed as part of the design basis for the PSLIV, justify the conclusion that the It fatigue analysis for the PSLIV does not need to be identified as a TLAA in accordance with 10 CFR 54.21(c)(1).

RESPONSE RAI 4.1-1 Request 1, Part A

1) The ASME Code requires a fatigue evaluation for Class 1 valves greater than 4 inches diameter nominal pipe size or for valves less than or equal to 4 inches in nominal pipe size if specified by the owner's design specification. The Davis-Besse purchasing specifications did not require a fatigue analysis for Class 1 valves less than or equal to 4 inches in nominal pipe size. Therefore, only valves greater than 4 inches diameter nominal pipe size require a fatigue analysis.

L-11-218 Page 5 of 9 Class 1 valves larger than 4" nominal pipe size listed in USAR Table 5.1-1b are provided below along with the ASME Code Year.

" DHMA and DH1B (LP Injection; outside containment isolation valves) - Draft ASME Pump and Valve Code, November 1968. These valves are installed in a Class 2 line.

" DH 11 and DH12 (DH Removal Outlet: containment isolation valves) - Draft ASME Pump and Valve Code, November 1968 Davis-Besse ASME Class I valves larger than 4" nominal pipe size not listed in USAR Table 5.1-1b are provided below along with the ASME Code Year.

• DH76 and DH77 (LP Injection: stop check inside containment isolation valves)

- ASME Code Year 1971 w/Addenda thru Summer 1971

• DH21 and DH23 (DH Removal Outlet: bypass around containment isolation valves DH1 1 and DH12) - ASME Code Year 1971

" CF28, CF29, CF30 and CF31 (Core Flood: stop check isolation valves) -

ASME Code Year 1971 w/Addenda thru winter 1972

2) The ASME Code Year 1968 requires a fatigue evaluation for valves greater than 4 inches diameter nominal pipe size or for valves less than or equal to 4 inches in nominal pipe size if specified by the owner's design specification. The Davis-Besse purchasing specification did not require a fatigue analysis for Class 1 valves less than or equal to 4 inches in nominal pipe size. Therefore, only valves greater than 4 inches diameter nominal pipe size require a fatigue analysis.

For ASME Code years 1971 or later, a fatigue evaluation is required by NB-3512.2 unless the exemption requirements of NB-3222.4(d) are met for Class 1 valves larger than 4" nominal pipe size. Fatigue analysis was not performed and was not required for Class 1 valves that are 4" and smaller nominal pipe size per NB-3513 and NB-3563.

Piping and instrumentation diagrams (P&IDs) were reviewed to identify Class 1 valves of greater than 4 inches diameter nominal pipe size. There were 12 valves of greater than 4 inches diameter nominal pipe size that were identified as a result of this effort. Those 12 valves are listed above. A search of the Davis-Besse records did not locate fatigue analyses for the subject Class 1 valves. Therefore, a commitment is provided in Appendix A to perform a fatigue evaluation in accordance with the requirements of the ASME Code of record for the Davis-Besse Class 1 valves greater than 4 inches diameter nominal pipe size. The issue of missing records has been documented in the Davis-Besse Corrective Action Program for resolution.

L-11-218 Page 6 of 9

3) A fatigue analysis is required for Class 1 valves larger than 4" nominal pipe size.

Since ASME Code fatigue analyses evaluate an explicit number and type of thermal and pressure transients that are postulated to envelope the number of occurrences possible during the design life of the plant, these fatigue analyses are time-limited aging analyses (TLAAs) and therefore, are required to be evaluated in accordance with 10 CFR 54.21 (c)(1).

The Fatigue Monitoring Program prevents fatigue TLAAs from becoming invalid by assuring that the fatigue usage resulting from actual operational transients does not exceed the Code design limit of 1.0. The program uses the systematic counting of transient cycles and the evaluation of operating data to ensure that the allowable cycle limits are not exceeded, thereby ensuring that component fatigue usage limits are not exceeded. Therefore, the effects of fatigue on Class 1 valves greater than 4 inches diameter nominal pipe size will be managed for the period of extended operation by the Fatigue Monitoring Program in accordance with 10 CFR 54.21 (c)(1)(iii).

A change is required to Section 4.3.2.3.2 of the LRA.

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

Request 1, Part B

1) "Valve - Pressurizer Relief Isolation" (RC11) in USAR Table 5.1-1b correlates to "Pressurizer Pilot Operated Relief Isolation Valve" in USAR Table 5.2-1.

2) "Valve - Pressurizer Pilot-Operated Relief (PORV)" (RC2A) in USAR Table 5.1-1b correlates to "Relief valve" in USAR Table 5.2-1

3) Pressurizer Relief Isolation (RC1 1): ASME III 1974 Edition w/Addenda through Summer 1976 Pressurizer Pilot-Operated Relief (PORV)" (RC2A): Draft ASME Pump and Valve Code, November 1968 Request 2 As provided in USAR Table 5.2-1, the Pressurizer Safety Valves (RC1 3A and RC1 3B) and the Relief Valve (RC2A), also know as the Pressurizer Pilot Operated Relief Valve, were designed to the 1968 Draft ASME Pump and Valve Code. The code requires a fatigue analysis for valves greater than 4 inches diameter nominal pipe size or for valves less than or equal to 4 inches in nominal pipe size if specified by the owner's design specification. The subject Davis-Besse valves are not greater than 4 inches

L-1 1-218 Page 7 of 9 diameter nominal pipe size and the purchasing specification did not require a fatigue analysis. Therefore, a fatigue analysis was not required for the subject valves.

Request 3 As provided in USAR Table 5.2-1, the Pressurizer Pilot-Operated Relief Isolation Valve (RC11) was designed to the 1974 Edition of ASME Ill, with Addenda thru Summer 1976. A fatigue analysis was not required for ASME Section III, Class 1 valves that are 4" and smaller nominal pipe size per NB-3513 and NB-3563. The Pressurizer Pilot-Operated Relief Isolation Valve (RC1 1) is 2 1/2" nominal pipe size and was therefore not required by Code to be evaluated for fatigue, nor was it required by the purchasing specification.

Request 4 As provided in USAR Table 5.2-1, the Pressurizer Spray Line Isolation Valve (RC1 0) was designed to the 1986 Edition of ASME Section III, with no Addenda. A fatigue analysis was not required for ASME Section III, Class 1 valves that are 4" and smaller nominal pipe size per NB-3513 and NB-3563. The Pressurizer Spray Line Isolation Valve (RC10) is 2 1/2" nominal pipe size and was therefore not required by Code to be evaluated for fatigue, nor was it required by the purchasing specification.

Question RAI B.2.18-1 GALL AMP XI.M27, "Fire Water System," states in the "scope of program" element that the Fire Water System Program manages loss of material due to corrosion, MIC or biofouling, and includes flow testing, visual inspections, and non-intrusive examinations to detect these aging effects. LRA Section B.2.18 states that the applicant's Fire Water Program will manage loss of material as well as cracking of susceptible materials. The applicant's program basis documents state that cracking due to stress corrosion cracking of copper alloy (greater than 15 percent zinc) will be managed by the same testing and inspection activities that identify and manage the loss of material. The staff noted that flow tests and visual inspections are not industry-accepted methods to detect cracking.

It is unclear to the staff what technique the applicant plans to use in its Fire Water System Program that will adequately manage cracking of susceptible copper alloy (greater than 15 percent zinc) components.

L-1 1-218 Page 8 of 9 In light of the fact that flow tests and visual inspections are not industry accepted methods to detect cracking, provide additional information regarding the technique to be used to detect cracking of copper alloy (greater than 15 percent zinc) fire water system components.

RESPONSE RAI B.2.18-1 The previous response to this RAI provided in FENOC Letter (ML11151A090) dated May 24, 2011, is replaced in its entirety.

The Fire Protection System Program is revised to remove cracking as an aging effect that is managed. There are no aging management activities in the Fire Protection System Program that will be used to manage cracking of the copper alloy (with greater than 15 percent zinc) components that are exposed to raw water. Stress corrosion cracking or intergranular attack in the identified copper alloys in a raw water environment is only a potential for stations whose operating experience indicates the presence of ammonia or an ammonium salt in raw water. FENOC conducted a review of the Davis-Besse plant specific operating experience for License Renewal. The results of this review showed no evidence of ammonia or an ammonium salt in raw water or cracking of copper alloys in the associated systems.

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

Question RAI B.2.31-1 GALL AMP XI.M20, "Open-Cycle Cooling Water System," states that this program addresses the aging effects of loss of material, fouling due to micro-or macro-organisms, and various corrosion mechanisms generally found in the open cycle cooling water system. The GALL Report AMP does not address cracking, and although it was not identified as an exception or enhancement; the LRA states that copper alloy (with greater than 15 percent zinc) will be managed for cracking by the Open-Cycle Cooling Water Program. The LRA also states that the program consists of inspections, surveillances, and testing to detect and evaluate aging effects including cracking, and it is combined with chemical treatments and cleaning activities to minimize aging effects including cracking.

The LRA does not describe the inspection, surveillance, or testing method(s) that will be used to detect and evaluate cracking of the copper alloy (with greater than 15 percent zinc) components exposed to open cycle cooling water. In addition, the LRA does not describe the chemical treatments and cleaning activities that will be used to minimize cracking.

L-11-218 Page 9 of 9 The staff requests the following information:

1) Describe the aging management activities in the Open-Cycle Cooling Water Program that will be used to manage cracking of the copper alloy (with greater than 15 percent zinc) components with greater than 15 percent zinc that are exposed to raw water.

2) If the Open-Cycle Cooling Water Program will remain, the program used to manage cracking of copper alloy (with greater than 15 percent zinc) components, then the LRA should be updated to reflect this as an exception to GALL AMP XI.M20.

RESPONSE RAI B.2.31-1 The previous response to this RAI provided in FENOC Letter (ML11151A090) dated May 24, 2011, is replaced in its entirety.

The Open-Cycle Cooling Water Program is revised to remove cracking as an aging effect that is managed. There are no aging management activities in the Open-Cycle Cooling Water Program that will be used to manage cracking of the copper alloy (with greater than 15 percent zinc) components that are exposed to raw water. Stress corrosion cracking or intergranular attack in the identified copper alloys in a raw water environment is only a potential for stations whose operating experience indicates the presence of ammonia or an ammonium salt in raw water. FENOC conducted a review of the Davis-Besse plant specific operating experience for License Renewal. The results of this review showed no evidence of ammonia or an ammonium salt in raw water or cracking of copper alloys in the associated systems.

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

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

Amendment No. 12 to the DBNPS License Renewal Application Page 1 of 51 License Renewal Application Sections Affected Section 2.3.1.2 Section A. 1.21 Table 2.3.1-2 Section A. 1.30 Section 3.1.2.1.2 Section A.1.31 Section 3.1.2.2.6 Section A.1.34 Section 3.1.2.2.15 Section A.2.3.2.13 Table 3.1.1 Table A-1 Table 3.1.2-2 Section B.2.6 Table 3.3.2-1 Section B.2.9 Table 3.3.2-14 Section B.2.18 Table 4.1-1 Section B.2.21 Section 4.3.2.3.2 Section B.2.30 Section A.1.2 Section B.2.31 Section A. 1.6 Section B.2.34 Section A. 1.9 Section B.2.36 This document identifies revisions to the License Renewal Application (LRA) by Affected LRA Section, LRA Page No., and Affected Paragraph and Sentence. The count for the affected paragraph, sentence, bullet, etc. starts at the beginning of the affected Section or at the top of the affected page, as appropriate. Below each section the reason for the change is identified, and the sentence affected is printed in italicswith deleted text lined '-t and added text underlined.

LRA Amendment 12 Page 2 of 51 Affected LRA Section LRA Page No. Affected Paragraph and Sentence Section 2.3.1.2 Page 2.3-11 Components Subject to AMR, third paragraph In response to RAI 3.1.2.2-2, the third paragraph of the "Components Subject to AMR" subsection of LRA Section 2.3.1.2, "Reactor Vessel Internals," is revised as follows:

2.3.1.2 Reactor Vessel Internals Components Subiect to AMR Table 2.3.1-2 lists the component types that are subject to AMR and their intended functions.

Table 3.1.2-2, Aging Management Review Results - Reactor Vessel Internals, provides the results of the AMR.

w I Th P

s~urvc'llanco 6peimocrn floldeF tube a66emDiies Ge net pro viaa any s~3,Av0. y I I I I

• Ie function. 6enseguontuy tfh cornqpenont Gees not peForrn an intended runcrien and ;is not ubj-- t to AMR.

The fuel assemblies and control rod assemblies, and incore neutron detectors are not subject to AMR as they are short-lived components whose lifetime will not be affected by the period of extended operation.

LRA Amendment 12 Page 3 of 51 Affected LRA Section LRA Page No. Affected Paraqraph and Sentence Table 2.3.1-2 Page 2.3-12 Two Rows In response to RAI 3.1.2.2-2, two rows of LRA Table 2.3.1-2, "Reactor Vessel Internals Components Subject to Aging Management Review," are revised as follows:

Table 2.3.1-2 Reactor Vessel Internals Components Subject to Aging Management Review CIntended Function Component Type (as defined in Table 2.0-1)

Incore Monitoring Instrumentation Guide Tube Assembly Support Incoroe Guido Tubo Acsemb4y Flow Control Vent Valve Assembly Support

LRA Amendment 12 Page 4 of 51 Affected LRA Section LRA Page No. Affected Paragraph and Sentence 3.1.2.1.2 Page 3.1-3 Subsections:

- Environments

- Aging Effects Requiring Management

- Aging Management Programs In response to RAI 3.1.2.2-2, the "Environments," "Aging Effects Requiring Management," and "Aging Management Programs," subsections of LRA Section 3.1.2.1.2, "Reactor Vessel Internals," are revised to read:

3.1.2.1.2 Reactor Vessel Internals Environments Subject mechanical components of the reactor vessel internals are exposed to the following normal operating environments:

• Berated r-eastor- coolant

• Borated reactor coolant with neutron fluence Aging Effects Requiring Management The following aging effects require management for the subject mechanical components of the reactor vessel internals:

" Change in dimension

• Cracking - due to Fatigue, Flaw Growth, SCC/!GA Stress Corrosion Cracking (SCC), and IrradiationAssisted Stress CorrosionCracking (IASCC)

" Loss of material

" Loss of preload

" Reduction in fracture toughness Aging Management Programs The following aging management programs address the aging effects requiring management for the reactor vessel internals:

. Fatigue Monito.ing Progranm*Oa*,a*u T /AAs)

• PWR Reactor Vessel Internals Program

• PWR Water Chemistry Program

LRA Amendment 12 Page 5 of 51 Affected LRA Section LRA Page No. Affected Paragraph and Sentence 3.1.2.2.6 Page 3.1-9 Last sentence In response to RAI 3.1.2.2-2, the last sentence of LRA Section 3.1.2.2.6, "Loss of Fracture Toughness due to Neutron Irradiation Embrittlement and Void Swelling,"

is revised to read:

3.1.2.2.6 Loss of Fracture Toughness due to Neutron Irradiation Embrittlement and Void Swelling Loss of fracture toughness due to neutron irradiation embrittlement and void swelling could occur in stainless steel and nickel alloy reactor vessel internals components exposed to reactor coolant and neutron flux. At Davis Besse, reduction in fracture toughness due to radiation embrittlement for stainless steel and nickel alloy reactor vessel internals components that are exposed to reactor coolant and neutron flux will be managed by the PWR Reactor Vessel Internals Program. ,urffr for-.hango in dimon.ion duo t,. ve, d w..ing evluation ,-

addresscd 4qn Sectin 3. 1.2.2.15. Void swelling is not identified as an aging effect requiringmanagement for these components.

LRA Amendment 12 Page 6 of 51 Affected LRA Section LRA Page No. Affected Paragraph and Sentence 3.1.2.2.15 Page 3.1-11 Last sentence In response to RAI 3.1.2.2-2, the last sentence of LRA Section 3.1.2.2.15, "Changes in Dimension due to Void Swelling," is revised to read:

3.1.2.2.15 Changes in Dimension due to Void Swelling Changes in dimensions due to void swelling could occur in stainless steel and nickel alloy PWR reactor internal components exposed to reactor coolant.

Changes in dimensions due to void 8wellig for-DavWs Besse stainless stee! an nickel aflly reactor-internals Gonpononts that are exposed to reactor coolant wl be managed by the PWR Roa*ctor Vessel.. ternal Program. Changes in dimensions due to void swelling are not identified as an aging effect requiring management for the reactorvessel internalcomponents.

LRA Amendment 12 Page 7 of 51 Affected LRA Section LRA Page No. Affected Paragraph and Sentence Table 3.1.1 Pages 3.1-19 Rows 3.1.1-22 and 3.1.1-33 and 3.1-24 In response to RAI 3.1.2.2-2, rows 3.1.1-22 and 3.1.1-33 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 IVof NUREG-1801," are revised 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 AFurther Number Component/Commodity Aging Aging Management Evaluation Discussion rEffect/Mechanism Programs Recommended 3.1.1-22 Stainless steel and nickel alloy Loss of fracture FSAR supplement No, but licensee Consistent with NUREG-1801, reactor vessel internals toughness due to commitment to (1) commitment to but a different program is used.

components exposed to reactor neutron irradiation participate in industry be confirmed Reduction in fracture toughness coolant and neutron flux embrittlement, void RVI aging programs (2) due to radiation embrittlement for swelling implement applicable stainless steel and nickel alloy results (3) submit for reactor vessel internals NRC approval > 24 components that are exposed to months before the reactor coolant and neutron flux extended period an RVI will be managed by the PWR inspection plan based on Reactor Vessel Internals industry roraml recommendation. Program.

Change in d-imenpnion dule to 14i swell;ng is addressed in itm Further evaluation is documented in Section 3.1.2.2.6.

LRA Amendment 12 Page 8 of 51 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 Euatis s Item Aging Aging Management Component/Commodity Effect/Mechanism Programs Evaluation Discussion Number Recommended 3.1.1-33 Stainless steel and nickel alloy Changes in FSAR supplement No, but licensee Not applicable.

reactor vessel internals dimensions due to commitment to (1) commitment to Changes in dimensions due to components void swelling participate in industry be confirmed void swelling are not identified as RVI aging programs (2) an aging effect requirina implement applicable management for these results (3) submit for components.

NRC approval > 24 months before the Ch.ng.s in dm.n...ions d,4o, to9 extended period an RVI ,oid ig... w be

, ma.naggd b inspection plan based on the PWR ReactOr Ves!el industry lnt**.*,s P. ogra.

recommendation. Further evaluation is documented in Section 3.1.2.2.15.

LRA Amendment 12 Page 9 of 51 Affected LRA Section LRA Page No. Affected ParaaraDh and Sentence Table 3.1.2-2 Pages 3.1-60 Entire Table through 3.1-121 In response to RAI 3.1.2.2-2, LRA Table 3.1.2-2, "Aging Management Review Results - Reactor Vessel Internals,"

is replaced in its entirety, to read:

Table 3.1.2-2 Aging Management Review Results - Reactor Vessel Internals Row Intended Aging Effect Aging NUREG-1801 Table 1 No. Component Type Function(s) Material Environment Requiring Management Rev. 1 Item Item Notes Management Program (Rev. 2 Item) Item

..PlenumCoverAssembly*

Plenum Cover Rib Pads BrtdRatrIV 41 Stainless Borated Reacto Loss of PWR Reactor IV.84-15 (primarycomponent Support Steel Coolant with material- wear Vessel Internals (IV.B4.RP- 3.1.1-63 E with no expansion Neutron Fluence 251) comnonents)

Plenum Cover Support FlanQe BrtdRatrIV 41 Stainless Borated Reacto Loss of PWR Reactor IV4-15 2 (primarycomponent SuDDo Steel Coolantwith material - wear Vessel Internals (IV.B4.RP- 3.1.1-63 E with no expansion Neutron Fluence 251) components)

LRA Amendment 12 Page 10 of 51 Table 3.1.2-2 Aging Management Review Results - Reactor Vessel Internals Row Intended Aging Effect Aging NUREG-1801 Table 1 No. Component Type Function(s) Material Environment Requiring Management Rev. 1 Item Item Notes Management Program (Rev. 2 Item)

ControlRod Gud*e Tube "R- V Ass CRGT Spacer Casting (expansioncomponent Cast (x

with marn component Ast Borated Reactor Reduction in PWR Reactor IV.B4-4 wihpiaycmoetAustenitic ColatRih ratreeactor- 31.-8 3 link of CSS Cast Outlet Supp Stainless Coolant with fracture Vessel Internals (IV.B4.R 3.1.1 E Nozzles. CSS Vent Steel Neutron Fluence toughness 242)

Valve Discs or IMI Guide Tube Spiders)

Core Supp pitcAhb Assembly." " ".._..

CSS Top Flange (primarycomponent Stainless Borated Reacto Loss of PWR Reactor IV.B4-15 with no expansion Suppo Steel Coolanteutr matenal - wear Vessel Internals 25)E components) Fluence Upper Core Barrel (UCB) Bolts (original bolts) and their locking PWR Reactor devices Stainless Borated Reactor Vessel Internals IV.B4-20 5 (primarycomponent Sup Steel Coolant with Cracking - SCC (IV.B4.RP- 3.1.1-37 E with expansion Neutron Fluence PWR Water 248) components of UTS Chemistry Bolts. LTS Bolts, FD Bolts and SSHT Bolts)

LRA Amendment 12 Page 11 of 51 Table 3.1.2-2 Aging Management Review Results - Reactor Vessel Internals Row Intended Aging Effect Aging NUREG-1801 Table 1 No. Component Type Function(s) Material Environment Requiring Management Rev. I Item Item Notes Management Program (Rev. 2 Item) Item Upper Core Barrel PWR Reactor (UCB) Bolts Vessel Internals IV.B4-20 (replacementbolts) and Cracking - SCC (IV.B4.RP- 3.1.1-37 E their locking devices Borated Reactor PWR Water 248)

Nickel BrtdRaorChemistry with ____

6 (primarycomponent Suppor All Coolant with expansion Allov Neutron Fluence components of UTS Cumulative IV.B4-3 7 Bolts, LTS Bolts, FD fatique damage TLAA (IV.B4.R-53) 3.1.1-05 A Bolts and SSHT Bolts)

CSS Cast Outlet Nozzles Cast Nozzles att Borated Reactor Reduction in IV.B4-21 (primarycomponent Austenitic Coolant with fracture PWR Reactor (IV.B4.RP- 3.1.1-80 E with expansion Stainless Vessel Internals I components as follows: Steel Neutron Fluence toughness 253)

CRG T Spacer Casting)

CSS Vent Valve Top Retaining Ring and Bottom Retaining Ring Stainless Borated Reactor Reduction in PWR Reactor IV.B4-16 8 (primarycomponent SuDDort Steel Coolant with fracture Vessel Internals 522E wtnoepninNeutron with no expansion Fluence toughness VeslItras 252) components)

CSS Vent Valve Discs (piaycmoet SDot Cast _______PW Reco (primary ast Borated Reactor Reduction in IV.B4-21 9 with expansion Coolant with fracture (IV.B4.RP- 3.1.1-80 E component of CRG T Flow Stainless Neutron Fluence toughness Vessel Internals 253)

Spacer Casting) Control Steel

LRA Amendment 12 Page 12 of 51 Table 3.1.2-2 Aging Management Review Results - Reactor Vessel Internals Row Intended Aging Effect Aging NUREG-1801 Table 1 No. Component Type Function(s) Material Environment Requiring Management Rev. 1 Item Item Notes Management Program (Rev. 2 Item)

CSS Vent Valve Disc Shaft Borated Reactor Reduction in IV.B4-16 Stainless Boae eco euto n PWR Reactor I.41 10 (primarycomponent Suppo Steel Coolant with fracture Vessel Internals (IV.B4.RP- 3.1.1-22 E with no expansion Neutron Fluence toughness 252) components)

,. ~Core:BirrelAs'sem Core Barrel Cylinder (includingvertical and circumferentialseam BoratedReactor Reduction in IV.B4-12 welds)1 wed Spot Steel Stainless Coolant with Cooanwth fracture frctreVeseIteal (IV.B4.RP- 3.1.1-22 E (expansion component Neutron Fluence toughness Vessel Internals 250) with primary component link of Baffle Plates)

Baffle Plates (primarycomponent with expansion components of Core Stainless BoratedReactor Reduction in PWR Reactor IV.B4-12 12 Barrel Cylinder, Supor Coolant with fracture (IV.B4.RP- 3.1.1-22 E including verticaland Steel Neutron Fluence toughness Vessel Internals 249) circumferentialseam welds, and Former Plates)

FormerPlates (expansion componen Stainles Borated Reactor Reduction in PWR Reactor IV.B4-12 13 with Primarycomponent Supo Steel Coolantwith fracture Vessel Internals (IV.B4.RP 3.1.1-22 E Neutron Fluence toughness 250) link of Baffle Plates)

LRA Amendment 12 Page 13 of 51 Table 3.1.2-2 Aging Management Review Results - Reactor Vessel Internals Row No. Component Type Intended Aging Effect Aging NUREG-1801 Table 1 Notes Function(s) Material Environment Requiring Management Rev. 1 Item Management Program (Rev. 2 Item) Item PWR Reactor Vessel Internals None Cracking -

racn -(IV.B4.RP- None H PWR Water 375)

Chemistry PWR Reactor Core Barrel-to-Former Vessel Internals IV.B4-07 (CBF) BotsBorated Reactor IASCC (IV.B4.RP- 3.1.1-30 E Stainless BrtdRatrISCPWR Water 244) 14 (expansion component SuDDo Steel Coolant with Chemistry with primary component Neutron Fluence link of FB Bolts) Loss of - wear material Loss of preload PWR Reactor IV.B4-01 Vessel Inteals (IV.B4.RP- 3.1.1-22 E Reduction in 243) fracture touqhness

LRA Amendment 12 Page 14 of 51 Table 3.1.2-2 Aging Management Review Results - Reactor Vessel Internals Row Intended Aging Effect Aging NUREG-1801 Table 1 No. Component Type Function(s) Material Environment Requiring Management Rev. I Item Item Notes Management Program (Rev. 2 Item)

PWR Reactor Cracking - Vessel Internals None (IV.B4.RP- None H PWR Water 375)

Chemistry PWR Reactor Baffle-to-Former (FB) IV.B4-07 Bolts Crackinq - Vessel Internals (oimrvcopoen Sailes Bolts Re ctrackig Borated Reactor IASCC PWR Water (IV.B4.RP-21 3.1.1-30 E (primarycomponent upo Stainless Coolant with PRWtr 21 15 with expansion 15 __________

Suoport Steel

____

Coolant Neutron wChemistry FluenceLosf components of BB Loss of Bolts and CBF Bolts) material - wear Loss of preload PWR Reactor IV.B4-O1 V (IV.B4.RP- 3.1.1-22 E Reducton in Reduction Vessel Internals 240) 20 fracture toughness PWR Reactor Cracking Vessel Internals None rackig ftgePWR Water (IV.B4.RP-375) None H Baffle-to-Baffle (BB) Chemistry Bolts - intemal Borated Reactor Loss of 16 (expansioncomponent SuD Steel Coolant with material - wear with primary component S Neutron Fluence link of FB Bolts) Loss of preload PWR Reactor IV.B4-01 Vessel Internals (IV.B4.RP-Reduction in 243) fracture toughness

LRA Amendment 12 Page 15 of 51 Table 3.1.2-2 Aging Management Review Results - Reactor Vessel Internals Row Intended Aging Effect Aging NUREG-1801 Table 1 No. Component Type Function(s) Material Environment aRequiring Management Rev. IItem Notes No.Funtio~s)Management Program (Rev. 2 Item) Ie PWR Reactor Cracking - Vessel Internals None rackig PWR Water (IV.B4.RP-375) None H Chemistry PWR Reactor 1Baffle-to-Baffle (BB) Cracking - Vessel Internals IV.B4-07 Bolts - external Borated Reactor IASCC (IV.B4.RP- 3.1.1-30 E 17 (expansion component Su Stainless Steel Coolant Coln with ihChemistry______ PWR Water 244) with primary component Neutron Fluence link of FB Bolts) Loss of - wear material Loss of preload PWR Reactor IV.B4-01 Vessel Internals (IV.B4.RP- 3.1.1-22 E Reduction in 243) fracture toughness Accessible Locking PWR Reactor Device and Locking Cracking Vessel Internals IV.B4-07 Weld (FB Bolts and CASCC (IV.B4.RP- 3.1.1-30 E Internal BB Bolts) PWR Water 241)

SBorated Reactor Chemistry Drimary component 18 with expansion SuStateel Coolant with components of Sel Neutron Fluence

______Reduction in PReatr IV.B4-01 Inaccessible Locking fracture PWR Reactor (IV.B4.RP- 3.1.1-22 E Device and Locking toghness Vessel Internals 240)

Weld (CBF Bolts and External BB Bolts))

LRA Amendment 12 Page 16 of 51 Table 3.1.2-2 Aging Management Review Results - Reactor Vessel Internals Row Intended Aging Effect Aging NUREG-1801 Table 1 Notes No. Component Type Function(s) Material Environment Requiring Management Rev. 1 Item eItem Management Program (Rev. 2 Item)

InaccessibleLocking PWR Reactor Device and Locking Cracking - Vessel Internals IV.B4-07 Weld (CBFBolts and Cc (IV.B4.RP- 3.1.1-30 E External BB Bolts) IASCC PWR Water 244) 19 eSu component oortin Stainles Stwiths Coolant Reactor Borated Chemistry (expansion with primary component Suppot Steel Coolant with Neutron Fluence _____ W eco link of Accessible Reduction in IV.B4-01 Locking Device and fracture (IV.B4.RP- 3.1.1-22 E Locking Weld (FB Bolts toughness Vessel Internals 243) and InternalBB Bolts))

Lower Core Barrel (LCB) Bolts (odginal) and their locking PWR Reactor devices Borated Reactor Vessel Internals IV.B4-13 Stainless BoaeRecoVeslItras V.41 20 (Primary component Suppo Steel Coolant with Cracking - SCC (IV.B4.RP- 3.1.1-37 E with expansion Neutron Fluence PWR Water 247) components of UTS Chemistry Bolts. LTS Bolts, FD Bolts and SSHT Bolts)

Lower Core Barrel PWR Reactor (LCB) Bolts Vessel Internals IV.B4-13 (replacement)and their Cracking - SCC (IV.B4.RP- 3.1.1-37 E locking devices Borated Reactor PWR Water 247)

Nickel BrtdRaorChemistry 21 (primary component Suppo Nicke Coolant with with expansion Neutron Fluence components of UTS Cumulative IV.B4-37 Bolts, LTS Bolts, FD fatique damage TLAA (IV.B4.R-53) 3.1.1-05 A Bolts and SSHT Bolts) ftg I

LRA Amendment 12 Page 17 of 51 Table 3.1.2-2 Aging Management Review Results - Reactor Vessel Internals Row Intended Aging Effect Aging NUREG-1801 Table I Component Type Function(s) Material Environment Requiring Management Rev. 1 Item Notes No. Management Program (Rev. 2 Item) Item Upper Thermal Shield (UTS) Bolts PWR Reactor componen Stainles Borated Reactor Vessel Internals IV.B4-13 (expansion Steel Neutron Cracking - SCC PWR Water 245) 22 22 with primary, wexPary component Spot componen t Steel Coolant Fluence with (IV.B4.RP- 3.1.1-37 E link of UCB Bolts and Ch erist5)

LCB Bolts) Chemis Surveillance Specimen PWR Reactor Holder Tube (SSHT) Vessel Internals IV.B4-13 Bolts Borated Reactor Cracking - SCC (IV.B4.RP- 3.1.1-37 E Nickel PWR Water 245) 23 (expansion component U Allo Coolant with Chemistry with primarycomponent Neutron Fluence Cumulative IV.B4-37 link of UCB Bolts and fatique damage TLAA (IV.B4.R-53) 3.1.1-05 A LCB Bolts) - fatique Lower Grid A,,,o,,i' .

Lower Fuel Assembly Support Pads:

Pad,Pad-to-Rib Section Weld, Cap Screw and associated Locking Weld, and StinesI Borated Reactor Reduction in PWR Reactor IV.B4-31 Alloy X-750 Dowel S Nickel Coolant with fracture Vessel Internals 4(expansion component Alloy Neutron Fluence toughness 260) with primary component link of IMI Guide Tube Spiders and IMI Guide Tube Spider-to-Lower Grid Rib Section Welds)

LRA Amendment 12 Page 18 of 51 Table 3.1.2-2 Aging Management Review Results - Reactor Vessel Internals Row Intended Aging Effect Aging NUREG-1801 Table 1 No. Component Type Function(s) Material Environment Requiring Management Rev. I Item Item Notes Management Program (Rev. 2 Item)

Lower FuelAssembly PWR Reactor Support Pads: Vessel Internals IV.B4-32 Alloy X-750 Dowel Cracking - SCC (IV.B4.RP- 3.1.1-37 E Locking Weld PWR Water 262 (expansion component Su Nickel Borated Reactor Chemistry 25 with primary component S Alloy Coolant with link of IMI Guide Tube Neutron Fluence I V.B4-31 Spiders and IMI Guide Reduction in PWR Reactor IV.B4-31 Tube Spider-to-Lower toughness Vessel Internals 260)

GridRib Section t hs2 Welds)

Lower Grid Assembly:

Alloy X-750 Dowel-to-Lower FuelAssemby Support Pad Welds PWR Reactor Nickel Borated Reactor Vessel Internals IV.B4-32 26 (expansion component S Alloy Coolant with Cracking - SCC (IV.B4.RP- 3.1.1-37 E with primao link of Lowercomponent GridCemsr Neutron Fluence PWR Water 262)

Assembly: Alloy X-750 Chemis Dowel-to-Guide Block Welds)

Lower Grid Assembly:

Alloy X-750 Dowel-to-Guide Block Welds PWR Reactor (primarycomponent Nickel Borated Reactor Vessel Internals IV.B4-32 27 with expansion Suppo All Coolant with Cracking - SCC (IV.B4.RP- 3.1.1-37 E components of Alloy X- Neutron Fluence PWR Water 261) 750 Dowel-to-Lower Chemistry FuelAssemby Support Pad Welds)

LRA Amendment 12 Page 19 of 51 Table 3.1.2-2 Aging Management Review Results - Reactor Vessel Internals Row Intended Aging Effect Aging NUREG-1801 Table 1 No. Component Type Function(s) Material Environment Requiring Management Rev. 1 Item Notes Management Program (Rev. 2 Item) Item Lower Thermal Shield Cumulative IV.B4-37 (LTS) Bolts fatique damage TLAA (IV.B4.R-53) 3.1.1-05 A 28 exnsocopnncomponent (expansion Suor Nikl Coolant with PWR Reactor with primary component Su spr Nickel Alloy Borated Coolant Reactor - fatique Vessel Internals IV.B4-32 link of UCB Bolts and Cracking - SCC (IV.B4.RP- 3.1.1-37 E LCB Bolts) PWR Water 246)

Chemistry Flow Distributor Assembli Flow Distributor(FD)

BoltsPWReco Borated Reactor PWR Reactor (expansion component Stainless Vessel Internals IV.B4-25 29 with primary component Steel Coolant with Neutron Fluence Cracking - SCC (IV.B4.RP- 3.1.1-37 E link of UCB Bolts and PWR Water 256)

LCB Bolts) Chemistry

LRA Amendment 12 Page 20 of 51 Table 3.1.2-2 Aging Management Review Results - Reactor Vessel Internals Row Intended Aging Effect Aging NUREG-1801 Table 1 Notes No. Component Type Function(s) Material Environment Requiring Management Rev. 1 Item Management Program (Rev. 2 Item) Item Incore Monitoring !nstrumentation IMI) Guide Tube Assembly' IMI Guide Tube Spiders (primarycomponent with expansion components of CRGT Spacer Casting andCs Lower Fuel Assemba Austenitic Borated Reactor Reduction in PWR Reactor IV.B4-28 3 LSupport Pads:Pad, Supp Stainless Coolant with fracture Vessel Internals (IV.B4.RP- 3.1.1-80 E Pad-to-Rib Section Neutron Fluence toughness 258)

Weld, Cap Screw and Steel associatedLocking Weld, Alloy X-750 Dowel and Alloy X- 750 Dowel Locking Weld)

IMI Guide Tube Spider-to-Lower Grid Rib Section Welds (primarycomponent with expansion components of CRGT Spacer Casting and Stainless Borated Reactor Reduction in PWR Reactor IVoB4-31 31 Lower Fuel Assembly Supr Steel Coolant with fractureVessel Internals (IV.B4.RP- 3.1.1-22 E Support Pads:Pad, Neutron Fluence toughness 259)

Pad-to-Rib Section Weld, Cap Screw and associatedLocking Weld, Alloy X-750 Dowel and Alloy X-750 Dowel Locking Weld)

LRA Amendment 12 Page 21 of 51 Affected LRA Section LRA Page No. Affected Paragraph and Sentence Table 3.3.2-1 Page 3.3-137 Row 99 In the revised response to RAI B.2.31 -1, row 99 of LRA Table 3.3.2-1, "Aging Management Review Results -

Auxiliary Building HVAC System," is revised as follows (note - LRA Table 3.3.2-1 row 99 was previously revised by FENOC Letter dated May 24, 2011 (ML11151A090)):

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

/=/ea4 GREV-9CGppQe water-G0019d P--eSSUr-e R9=.14p14a9w-a ,,* Onn-Time 99 .y Alley- > . . NAM H oe6d6i6R bounda6l ýWeen Not Used

LRA Amendment 12 Page 22 of 51 Affected LRA Section LRA Page No. Affected ParaaraDh and Sentence Table 3.3.2-14 Pages 3.3-323 Rows 77, 85, 88, 100, 131, 191, 197, through 341 199 and 218 In response to RAI B.2.18-1, rows 77, 85, 88, 100, 131, 191, 197, 199 and 218 of LRA Table 3.3.2-14, "Aging Management Review Results - Fire Protection System," are revised as follows (note - LRA Table 3.3.2-14, rows 77, 85, 88, 100 and 131 were previously revised by FENOC Letter dated May 24, 2011 (ML11151A090)):

LRA Amendment 12 Page 23 of 51

LRA Amendment 12 Page 24 of 51 Affected LRA Section LRA Page No. Affected Paragraph and Sentence Table 4.1-1 Page 4.1-4 "Class 1 valves" row In response to RAI 4.1-1, the "Class 1 Valves" row of Table 4.1-1 is revised to read:

Results of TLAA Evaluation by Category 54.21(c)(1) LRA Paragraph Section Metal Fatigue 4.3 Class 1 Fatigue 4.3.2 Class 1 valves Noet a T4A- (ii 4.3.2.3.2

LRA Amendment 12 Page 25 of 51 Affected LRA Section LRA Page No. Affected ParaaraDh and Sentence 4.3.2.3.2 Pages 4.3-16 Entire section and 4.3-17 In response to RAI 4.1-1, LRA Section 4.3.2.3.2, "Class 1 Valves Fatigue," is replaced in its entirety, to read:

4.3.2.3.2 Class I Valves Fatigue The ASME Code requires a fatigue evaluation for Class 1 valves greater than 4 inches diameter nominal pipe size or for valves less than or equal to 4 inches in nominal pipe size if specified by the owner's design specification. The Davis-Besse purchasing specifications did not require a fatigue analysis for Class 1 valves less than or equal to 4 inches in nominal pipe size. Therefore, only valves greater than 4 inches diameter nominal pipe size require a fatigue analysis.

Piping and instrumentation diagrams (P&IDs) were reviewed to identify Class 1 valves of greater than 4 inches diameter nominal pipe size. There were 12 valves of greater than 4 inches diameter nominal pipe size that were identified as a result of this effort. Those 12 valves are listed as follows:

1. Valve ID Size Code Year Description 1 CF-28 2 CF-29 14 1971 w/Addenda Core Flood - core flood tank discharge 3 CF-30 thru Winter 1972 line stop check isolation valve 4 CF-31 5 DH-1 1 12 1968 RCS to Decay Heat - containment 6 DH-12 isolation valve 7 DH-21 RCS to Decay Heat - containment isolation 8 DH-23 8 1971 valve bypass line isolation valve 9 DH-76 10 1971 w/Addenda LP Injection to RCS - stop check isolation 10 DH-77 thru Summer 1971 valve 11 DH1A 10 1968 LP Injection - outside containment 12 DH 1B 1isolation valve A search of the Davis-Besse records did not locate fatigue evaluations for the subject Class 1 valves. Therefore, a commitment is provided in Appendix A to perform a fatigue evaluation in accordance with the requirements of the ASME Code of record for the Davis-Besse Class 1 valves greater than 4 inches diameter nominal pipe size. The issue of missing records has been documented in the Davis-Besse Corrective Action Program for resolution.

LRA Amendment 12 Page 26 of 51 Since ASME Code fatigue analyses evaluate an explicit number and type of thermal and pressure transients that are postulated to envelope the number of occurrences possible during the design life of the plant, these fatigue analyses are time-limited aging analyses (TLAAs) and therefore, are required to be evaluated in accordance with 10 CFR 54.21 (c)(1).

The Fatigue Monitoring Program prevents fatigue TLAAs from becoming invalid by assuring that the fatigue usage resulting from actual operational transients does not exceed the Code design limit of 1.0. The program uses the systematic counting of transient cycles and the evaluation of operating data to ensure that the allowable cycle limits are not exceeded, thereby ensuring that component fatigue usage limits are not exceeded. Therefore, the effects of fatigue on Class 1 valves greater than 4 inches diameter nominal pipe size will be managed for the period of extended operation by the Fatigue Monitoring Program.

Disposition: 10 CFR 54.21(c)(1)(iii) The effects of fatigue on Class 1 valves greater than 4 inches diameter nominal pipe size will be managed for the period of extended operation by the Fatigue Monitoring Program.

LRA Amendment 12 Page 27 of 51 Affected LRA Section LRA Page No. Affected Paragraph and Sentence A.1.2 Page A-10 Entire section In response to RAI A.1.2-1, LRA Section A.1.2, "Aboveground Steel Tanks Inspection Program," is revised to include discussion of preventive measures and program effectiveness verification, and now reads (note - Section A. 1.2 was previously revised by FENOC Letter dated May 24, 2011 (MI1 1151A090)):

A.1.2 ABOVEGROUND STEEL TANKS INSPECTION PROGRAM The Aboveground Steel Tanks Inspection Program manages the effects of corrosion on the external surfaces and inaccessible locations of the steel fire water storage tank, steel diesel fuel oil storage tank, and the stainless steel borated water storage tank. The Aboveground Steel Tanks Inspection Program includes preventive measures to mitigate corrosion by protectinq the external surface of steel components per standard industry practice and with sealant or caulking at the interface of concrete and the diesel fuel oil storage tank. The Aboveground Steel Tanks Inspection Program is a condition monitoring program that consists of periodic visual inspections of tank external surfaces, and volumetric examinations of tank bottoms at least once for each tank within five years after entering the period of extended operation. Additional opportunistic tank bottom inspections will be performed whenever the tanks are drained. The tank bottom inspections will verify the effectiveness of the program by measuring the thickness of the tank bottoms to ensure that significant degradation is not occurring.

LRA Amendment 12 Page 28 of 51 Affected LRA Section LRA Page No. Affected Paracraoh and Sentence A.1.6 Page A-10 Entire section In response to RAI A.1.6-1, LRA Section A.1.6, "Boric Acid Corrosion Program,"

is revised to include discussion of activities performed upon discovery of evidence of boric acid leakage, and now reads:

A.1.6 BORIC ACID CORROSION PROGRAM The Boric Acid Corrosion Program manages the effects of boric acid leakage on the external surfaces of in-scope structures and components potentially exposed to boric acid leakage. The Boric Acid Corrosion Program is a condition monitoring program consisting of visual inspections.

The Boric Acid Corrosion Program manages loss of material due to boric acid corrosion. The program includes provisions to identify, inspect, examine and evaluate leakage, and initiate corrective action. The program relies in part on implementation of recommendations of NRC Generic Letter 88-05, "Boric Acid Corrosion of Carbon Steel Reactor Components in PWR Plants." The Boric Acid Corrosion Program ensures that the pressure boundary integrity and material condition of the subiect structures and components are maintained consistent with the current licensing basis during the period of extended operation.

The Boric Acid Corrosion Program includes: (a) visual inspection of external surfaces that are potentially exposed to borated water leakage: (b) timely discovery of leak path and removal of the boric acid residues; (c) assessment of the damage; and (d) follow-up inspection for adequacy.

LRA Amendment 12 Page 29 of 51 Affected LRA Section LRA Page No. Affected Paragraph and Sentence A.1.9 Page A-1I Entire section In response to RAIs B.2.9-3 and B.2.9-5, LRA Section A.1.9, "Collection, Drainage, and Treatment Components Inspection Program," is replaced in its entirety to read:

A.1.9 COLLECTION, DRAINAGE, AND TREATMENT COMPONENTS INSPECTION PROGRAM The Collection, Drainaqe, and Treatment Components Inspection Program consists of visual and volumetric inspections. This program will be implemented via periodic inspections of a representative sample. These inspections will ensure that the existing environmental conditions in collection, drainage, and treatment service are not causing material degradation that could result in a loss of component intended function during the period of extended operation. Visual inspections will be conducted using visual (VT I or equivalent) inspection methods, capable of detecting loss of material, cracking, or reduction in heat transfer. This pro-gram will also include volumetric inspections of inaccessible surfaces (e.g., tank bottoms sitting on concrete). Inspections will be performed by qualified personnel following procedures consistent with the pertinent ASME code of record and 10 CFR 50, Appendix B.

LRA Amendment 12 Page 30 of 51 Affected LRA Section LRA Paqe No. Affected Para-graph and Sentence A.1.21 Pages A-16 Second paragraph, [new] last and A-17 sentence In response to RAI B.2.21-7, LRA Section A.1.21, "Inaccessible Power Cables Not Subject to 10 CFR 50.49 Environmental Qualification Requirements Program," second paragraph, is revised to include a new sentence to address water collection inspection frequency, and now reads (note - Section A.1.21 was previously revised, in its entirety, by FENOC Letter dated May 5, 2011 (MLI 1131A073)):

A.1.21 INACCESSIBLE POWER CABLES NOT SUBJECT TO 10 CFR 50.49 ENVIRONMENTAL QUALIFICATION REQUIREMENTS PROGRAM At least once every 6 years these cables are tested to provide an indication of the condition of the conductor insulation. The frequency of testing is adjusted based on test results and operating experience. The program also requires periodic inspection of electrical manholes associated with in-scope cables for water accumulation and requires the removal of water from the electrical manholes as necessary. Inspections are performed at least annually and are also performed in response to event-driven occurrences (such as heavy rain or flooding.) The inspection frequency for water collection is established and performed based on plant-specific operating experience with cable wetting or submergence.

LRA Amendment 12 Page 31 of 51 Affected LRA Section LRA Page No. Affected Paraaraoh and Sentence A.1.30 Page A-20 Fourth paragraph In response to RAI B.2.18-1, LRA Section A.1.30, "One-Time Inspection," is revised to read (note - Section A.1.30 was previously revised by FENOC Letter dated May 24, 2011 (ML11151AO90)):

A.1.30 ONE-TIME INSPECTION Th One Tim*e npection includes viUal and volumetri inpotion6 to dotoci and characterize cracking of copper-alloy 145% zinc oXpoed to raw14 wate-r Th u9ne FWUIZ t~InflJUGLLK wvn u 49FWF)Sbee ev:Uwize a6t.whetherIU, and to owa extent;FJL crsckinci h~s rccurrnd Cr~ckina p f Gopper alely >r5% zinc exposed to ra-water-is not addressed by another a . m a gement proegram.

Affected LRA Section LRA Page No. Affected Paragraph and Sentence A.1.30 Page A-20 New [last] paragraph In response to RAI B.2.30-2, a new paragraph is added to the end of LRA Section A. 1.30, "One-Time Inspection," to read (note - Section A. 1.30 was previously revised by FENOC Letter dated May 24, 2011 (ML11151A090)):

A.1.30 ONE-TIME INSPECTION This pro-gram cannot be used for structures or components with known age-related degradation mechanisms or when the environment in the period of extended operation is not expected to be equivalent to that in the prior 40 years.

Periodicinspections should be Proposedin these cases.

LRA Amendment 12 Page 32 of 51 Affected LRA Section LRA Page No. Affected ParagraDh and Sentence A.1.31 Page A-20 Entire section In the revised response to RAI B.2.31-1, the first sentence of the second paragraph of the "Program Description" section of A. 1.31 is revised to delete management of cracking of copper alloy with greater than 15 percent zinc, and now reads (note that this same change to Section A.1.31 was previously provided by FENOC Letter dated May 24, 2011 (ML11151A090), and is repeated here for completeness):

A.1.31 OPEN-CYCLE COOLING WATER PROGRAM The Open-Cycle Cooling Water Programmanages loss of materialdue to crevice, galvanic, general,pitting and microbiologically-influencedcorrosion; and erosion for in-scope components in the Service Water System and components connected to or cooled by the Service Water System (includingthe cooling tower makeup water relative to the CirculatingWater System), along With cracking Of SU.ceptible,material. The program manages fouling due to particulates (e.g.,

corrosion products) and biological material (micro- and macro-organisms) resulting in reduction in heat transfer for heat exchangers (including condensers, coolers, cooling coils, and evaporators) within the scope of the program.

The Open-Cycle Cooling Water Program consists of inspections, surveillances, and testing to detect and evaluate-ao...kig" fouling7 and loss of material, combined with chemical treatments and cleaning activities to minimize .akin..,

fouling7 and loss of material.The program is a combination condition and performance monitoring, and mitigation program that implements the recommendations of NRC Generic Letter 89-13, "Service Water System Problems Affecting Safety-Related Equipment" [Reference A. 1-17] for safety-related equipment in the scope of the program and manages loss of material for in-scope nonsafety-related components that contain service water or cooling tower makeup water.

LRA Amendment 12 Page 33 of 51 Affected LRA Section LRA Page No. Affected Paragraph and Sentence A.1.34 Page A-22 Second paragraph, [new] last sentence In response to RAI B.2.34-1, LRA Section A.1.34, "Reactor Head Closure Studs Program," is revised to include a new sentence at the end of the second paragraph, and now reads:

A.1.34 REACTOR HEAD CLOSURE STUDS PROGRAM The Reactor Head Closure Studs Program includes the preventive measures of NRC Regulatory Guide 1.65, "Materials and Inspection for Reactor Vessel Closure Studs," [Reference A.1-21] to mitigate cracking, including the use of a stable lubricant that is compatible with the fastener material and the environment.

The program provides a specific precaution against the use of compounds containing sulfur (sulfide), including molybdenum disulfide (MoS2), as a lubricant for the reactor head closure stud assemblies. An approved lubricant is applied to the threaded areas of studs and nuts and to the concave and convex faces of the spherical washers during each assembly. There are no metal platin-ls applied to the closure studs, nuts, or washers. A man-qanese-phosphate coatinq was applied to the studs, nuts and washers durinQ fabrication to act as a rust inhibitor and to assist in retaininQ lubricant. The progcram precludes the future use of replacement closure stud boltinQ fabricated from material with actual measured yield strenQth Qreaterthan or equal to 150 ksi except for use of the existing spare reactorhead closure stud bolting.

LRA Amendment 12 Page 34 of 51 Affected LRA Section LRA Page No. Affected Paragraph and Sentence A.2.3.2.13 Page A-41 New Section In response to RAI 4.1-1, new LRA Section A.2.3.2.13, "Class 1 Valves Fatigue,"

is added to LRA Appendix A, to read:

A.2.3.2.13 Class 1 Valves Fatigue The ASME Code requires a fatigue evaluation for Class 1 valves greater than 4 inches diameter nominal pipe size or for valves less than or equal to 4 inches in nominal pipe size if specified by the owner's design specification. The Davis-Besse purchasing specifications did not require a fatigue analysis for Class 1 valves less than or equal to 4 inches in nominal pipe size. Therefore, only valves greater than 4 inches diameter nominal pipe size require a fatigue analysis.

A search of the Davis-Besse records did not locate fatigue evaluations for the subject Class 1 valves. Therefore, a commitment is provided in Table A-1 of this Appendix to perform a fatigue evaluation in accordance with the requirements of the ASME Code of record for the Davis-Besse Class 1 valves greater than 4 inches diameter nominal pipe size.

Since ASME Code fatigue analyses evaluate an explicit number and type of thermal and pressure transients that are postulated to envelope the number of occurrences possible during the design life of the plant, these fatigue analyses are time-limited aging analyses (TLAAs) and therefore, are required to be evaluated in accordance with 10 CFR 54.21(c)(1).

The Fatigue Monitoring Program prevents fatigue TLAAs from becoming invalid by assuring that the fatigue usage resulting from actual operational transients does not exceed the Code design limit of 1.0. The program uses the systematic counting of transient cycles and the evaluation of operating data to ensure that the allowable cycle limits are not exceeded, thereby ensuring that component fatigue usage limits are not exceeded. Therefore, the effects of fatigue on Class 1 valves greater than 4 inches diameter nominal pipe size will be managed for the period of extended operation by the Fatigue Monitoring Program in accordance with 10 CFR 54.21(c)(1)(iii).

LRA Amendment 12 Page 35 of 51 Affected LRA Section LRA Page No. Affected Paragraph and Sentence Table A-1 Page A-63 Commitment No. 13 In the revised response to RAI B.2.18-1, license renewal future Commitment 13 in LRA Table A-I, "Davis-Besse License Renewal Commitments," is revised to delete the enhancement regarding inspection for cracking in copper alloy with greater than 15 percent zinc, and now reads (note - license renewal future Commitment 13 was previously revised, in its entirety, by FENOC Letter dated May 24, 2011 (ML11151A090), and revised again to add a second enhancement, shown below, by FENOC Letter dated June 3, 2011 (MLl 11159A1 32)):

Table A-1 Davis-Besse License Renewal Commitments Item Commitment Implerr Number Sch 13 Implement the One-Time Inspection as described in LRA Section Priior to A.1.30 B.2.30. Enhance the One-Time Inspection to: April *:2,2017 B.2.30 Inc.ude ,-s.-l and volumetric ins.*P9ctOn to dotot and RespeRsetG characteriz cracking o9f copper alloy> 15%, zinc e9XPosed to raW AIRCRA fr.

Water. The One tim~e in&poctons wil pro Vide diect oVidonce as -R2. R-48-4 to wether-, an.d to what axtent, cracking has ccu'rred. Cracking 3.3.2.7-1 2 an Of copper ally 145% zinc expoaed to raw wator i6 not 3.3.2.2.5 addrassed by another aging management proegram. frong NRG LoetteFdate4 Apri-#20; 2011

LRA Amendment 12 Page 36 of 51 Table A-1 Davis-Besse License Renewal Commitments e IRelated LRA Item Number Commitment j Implementation Schedule Source Section No./

Cmet Comments Include visual inspections to detect and characterize FENOC cracking due to cyclic loading of the stainless steel makeup Letter Response to pump casings (DB-P37-1 and 2) of the Makeup and Purification L-11-166 NRC RAI System. The one-time inspections will provide verification of the 3.3.2.2.4.3-1 absence of cracking due to cyclic loading. from NRC Letter dated I May 2, 2011

LRA Amendment 12 Page 37 of 51 Affected LRA Section LRA Page No. Affected Paragraph and Sentence Table A-1 Page A-64 Commitment No. 16 In response to RAI B.2.34-1, license renewal future Commitment 16 in LRA Table A-i, "Davis-Besse License Renewal Commitments," is revised to include a new program enhancement, and now reads:

Table A-1 Davis-Besse License Renewal Commitments Item Ii Related LRA Nmer Number Commitment Implementation Schedule Source Section No./

1.~mm..._Cmet Comments 16 Enhance the Reactor Head Closure Studs Program as follows: Priorto LRA A.1.34

" Select an alternate stable lubricant that is compatible with the April22,2017 B.2.34 fastener material and the environment. A specific precaution against the use of compounds containing sulfur (sulfide),

including molybdenum disulfide (MoS 2), as a lubricant for the reactor head closure stud assemblies will be included in the program.

" Preclude the future use of replacementclosure stud bolting fabricatedfrom materialwith actual measured yield strength FENOC Response to greaterthan or equal to 150 ksi except for use of the existing Letter NRC RAI spare reactorhead closure stud bolting. L-11-218 B.2.34-1 from NRC Letter dated June 20, 2011

LRA Amendment 12 Page 38 of 51 Affected LRA Section LRA Page No. Affected Paragraph and Sentence Table A-1 Page A-69 Commitment No. 44 (RENUMBERED)

License renewal future Commitment No. 41 in LRA Table A-i, "Davis-Besse License Renewal Commitments,"

provided by FENOC Letter dated June 17, 2011 (ML11172A389), is renumbered due to a duplicate commitment number in an earlier letter, and is now Commitment No. 44:

Table A-1 Davis-Besse License Renewal Commitments e IRelated LRA Item Number Commitment j SIpeme Schedule Source Section No./

Cmet Comments 44 The EDG Fuel Oil Storage Tanks (DB-T153-1 and DB-T153-2) and Prior to FENOC Response to the in-scope fuel oil and Service Water buried piping will be April 22, 2017 Letter NRC RAI cathodically protected in accordance with NACE SP0169-2007 or L-1 1-203 B.2.7-1 from NACE RP0285-2002. NRC Letter dated April 20, 2011, as modified per telecon with the NRC held on June 7, 2011

LRA Amendment 12 Page 39 of 51 Affected LRA Section LRA Page No. Affected ParagraDh and Sentence Table A-1 Page A-69 Commitment No. 45 (RENUMBERED)

License renewal future Commitment No. 42 in LRA Table A-1, "Davis-Besse License Renewal Commitments,"

provided by FENOC Letter dated June 17, 2011 (ML11172A389), is renumbered due to a duplicate commitment number in an earlier letter, and is now Commitment No. 45:

Table A-1 Davis-Besse License Renewal Comrr nitments Item Commitment Number 42 Implement the Nuclear Safety-Related Coatings Program as described in LRA Section B.2.42.

45

LRA Amendment 12 Page 40 of 51 Affected LRA Section LRA Page No. Affected Paragraph and Sentence Table A-1 Page A-69 New Commitment No. 46 In response to RAI 4.1-1, new license renewal future Commitment No. 46 is added to LRA Table A-1, "Davis-Besse License Renewal Commitments," to read:

Table A-1 Davis-Besse License Renewal Commitments Implementation Source Section No.I Item Number Commitment Schedule comments Comments 46 FENOC commits to perform a fatigue evaluation in accordance with Priorto LRA 4.3.2.3.2 the requirements of the ASME Code of record for the Davis-Besse April 22, 2015 A.2.3.2.13 Class 1 valves that are greater than 4 inches diameter nominal pipe size. The applicable valve identification numbers are CF28, CF29, FENOC Response to CF30, CF31, DH76, DH77, DH11, DH12, DHMA, DHIB, DH21 and Letter NRC RAI 4.1-1 DH23. L-11-218 from NRC Letter dated May 2. 2011

LRA Amendment 12 Page 41 of 51 Affected LRA Section LRA Page No. Affected Paragraph and Sentence B.2.6 Page B-38 Program Description -

New paragraph In response to RAI A.1.6-1, LRA Section B.2.6, "Boric Acid Corrosion Program,"

is revised to include a new paragraph at the end of the Program Description to describe activities performed upon discovery of evidence of boric acid leakage, which reads:

B.2.6 BORIC ACID CORROSION PROGRAM Program Description The Boric Acid Corrosion Program includes: (a) visual inspection of external surfaces that are potentially exposed to borated water leakage; (b) timely discovery of leak path and removal of the boric acid residues; (c) assessment of the damage: and (d) follow-up inspection for adequacy.

LRA Amendment 12 Page 42 of 51 Affected LRA Section LRA Page No. Affected Paragraph and Sentence B.2.9 Page B-47 Program Description In response to RAIs B.2.9-3 and B.2.9-5, the Program Description for LRA Section B.2.9, "Collection, Drainage, and Treatment Components Inspection Program," is replaced in its entirety to read:

B.2.9 COLLECTION, DRAINAGE, AND TREATMENT COMPONENTS INSPECTION PROGRAM Program Description The Collection, Drainage, and Treatment Components Inspection Pro-gram is a new plant-specific program for Davis-Besse that will consist of visual and volumetric inspections. This program will be implemented via periodic inspections of a representative sample. These inspections will ensure that the existing environmental conditions in collection, drainage, and treatment service are not causing material deqradation that could result in a loss of component intended function durinq the period of extended operation. Visual inspections will be conducted using visual (VT-1 or equivalent) inspection methods, capable of detecting loss of material, cracking, or reduction in heat transfer. This program will also include volumetric inspections of inaccessible surfaces (e.g., tank bottoms sitting on concrete). Inspections will be performed by qualified personnel following procedures consistent with the pertinent ASME code of record and 10 CFR 50, Appendix B. The Collection, Drainage, and Treatment Components Inspection Programis a condition monitoring program.

LRA Amendment 12 Page 43 of 51 Affected LRA Section LRA Page No. Affected ParagraDh and Sentence B.2.9 Pages B-48 Aging Management Program Elements:

and B-49 "Parameters Monitored or Inspected" and "Detection of Aging Effects" In response to RAls B.2.9-3 and B.2.9-5, the Aging Management Program Elements "Parameters Monitored or Inspected" [edited] and "Detection of Aging Effects" [replaced in its entirety] for LRA Section B.2.9, "Collection, Drainage, and Treatment Components Inspection Program," are revised to read:

B.2.9 COLLECTION, DRAINAGE, AND TREATMENT COMPONENTS INSPECTION PROGRAM Aging Management Program Elements

" Parameters Monitored or Inspected Inspections of the surfaces of collection, drainage, treatment, and other miscellaneous components that are exposed to raw (untreated) water, but are not addressed by other aging management programs, will be performed during maintenance and surveillance activities, when the surfaces are accessible for inspection.

Periodic inspections will be conducted on a representative sample of piping and components on a 10-year interval, with the first inspection taking place within the 10 year period prior to the period of extended operation. 4f opportunites for- inspoctien do not arise, tMAn Q focuso6d inSPoctioiW '"lb9 porformed as doScribed for tho DIte--t ion- of ging-E'-ffec-Gts66climenPt belo; 0W.

Parameters monitored or inspected are directly related to degradation of the components under review and include visible evidence of material degradation due to, loss of material (corrosion), as well as due to cracking, of susceptible materials, or reduction in heat transfer (fouling) for susceptible components.

" Detection of Aging Effects The Collection, Drainage, and Treatment Components Inspection Program provides for detection of aging effects prior to the loss of component intended function. Periodicinspections will be conducted on a representativesample of piping and components on a 10-year interval, with the first inspection taking place within the 10-year period prior to the period of extended operation.

LRA Amendment 12 Page 44 of 51 A representative sample of the system and component population will be inspected. The sample size is 20 percent of the population (defined as components having the same material, environment, and aging effect combination) or a maximum of 25 components. The sample population will be determined by engineering evaluation, and, where practical, focused on the (bounding or lead) components considered most susceptible to aging degradation due to time in service, the severity of the operating conditions, and the lowest desiqn margin.

Visual inspections will be conducted using visual (VT-1 or equivalent) inspection methods, capable of detecting loss of material, cracking, or reduction in heat transfer. This program will also include volumetric inspections of inaccessible surfaces. Inspections will be performed by qualified personnel following procedures consistent with the pertinent ASME code of record and 10 CFR 50, Appendix B.

Any evidence of degradation that could lead to a loss of component intended function will be documented and evaluated through the Corrective Action Program to determine the need for subsequent inspections, expansion, and for monitoring and trending the results. If degradationthat could lead to a loss of component intended function is detected, sample size will be increased by 20 percent of the population (defined as components having the same material, environment, and aging effect combination) or a maximum of 25 components.

LRA Amendment 12 Page 45 of 51 Affected LRA Section LRA Page No. Affected Paragraph and Sentence B.2.18 Page B-81 "Program Description" section, second paragraph, first sentence In the revised response to RAI B.2.18-1, the first sentence of the second paragraph of the "Program Description" section of B.2.18, "Fire Water Program,"

is revised to read (note that this same change to Section B.2.18 was previously provided by FENOC Letter dated May 24, 2011 (ML11151A090), and is repeated here for completeness):

B.2.18 FIREWATER PROGRAM Program Description The program is credited with managingloss of material, as we# a crSacking of

... ceptble m"ateriale, for fire water supply and water-basedfire suppression components in the scope of license renewal.

LRA Amendment 12 Page 46 of 51 Affected LRA Section LRA Page No. Affected Paragraph and Sentence B.2.21 Page B-91 Third paragraph, [new] fourth sentence In response to RAI B.2.21-7, LRA Section B.2.21, "Inaccessible Power Cables Not Subject to 10 CFR 50.49 Environmental Qualification Requirements Program," third paragraph, is revised to include a new sentence to address water collection inspection frequency, and now reads (note - Section B.2.21 was previously revised, in its entirety, by FENOC Letter dated May 5, 2011 (ML11131A073)):

B.2.21 INACCESSIBLE POWER CABLES NOT SUBJECT TO 10 CFR 50.49 ENVIRONMENTAL QUALIFICATION REQUIREMENTS PROGRAM Program Description In addition, manholes associated with inaccessible non-EQ power cables will be inspected for water accumulation and the water removed, as necessary. The frequency of inspections for accumulated water will be established and adjusted based on plant-specific inspection results, recognizing that the objective of the inspections, as a preventive action, is to keep the cables infrequently submerged, thereby minimizing their exposure to significant moisture. These inspections for water collection will be conducted at least annually and will also be performed in response to event-driven occurrences (such as heavy rain or flooding). The inspection frequency for water collection is established and performed based on plant-specific operating experience with cable wetting or submergence. The initial inspection will be completed prior to the period of extended operation.

LRA Amendment 12 Page 47 of 51 Affected LRA Section LRA Page No. Affected Paragraph and Sentence B.2.30 Page B-121 Enhancements - "Scope" In the revised response to RAI B.2.18-1, LRA Section B.2.30, "One-Time Inspection," is revised to delete the "Scope" enhancement regarding inspection for cracking in copper alloy with greater than 15 percent zinc, and now reads (note - Section B.2.30 was previously revised, in its entirety, by FENOC Letter dated May 24, 2011 (ML11151A090), and revised again to add a second enhancement, shown below, by FENOC Letter dated June 3, 2011 (ML11159A132)):

B.2.30 ONE-TIME INSPECTION Enhancements The following enhancements, which are plant-specific and in addition to the NUREG-1801,Section XI.M32 elements, will be implemented in the identified program elements prior to entering the period of extended operation.

• Scope The OQne Timol hnsection wil also in do*l visual . andf volumetic s to detoct and characteriZo GAcking Of copper alloy >15% zic' xpo go toT rnaw war IA Th:e one unrnne tnspueun ns wi: pnrV o UnrnT 4vidonce a6 toe Weth*r-, and t what exIent, cFrGing hal GcUrredl.

Qcrak~ia of c~ovor- alloy >15% zn exoed- to raw14 water- isno I I I I *1 addrossod hu anethr-a aoie nq anaaeoment pr-egara.

Ir The One-Time Inspection will also include visual and volumetric 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.

LRA Amendment 12 Page 48 of 51 Affected LRA Section LRA Page No. Affected Paragraph and Sentence B.2.30 Page B-123 Aging Management Program Element:

"Detection of Aging Effects" In response to RAI B.2.30-1, the Aging Management Program Element "Detection of Aging Effects" for LRA Section B.2.30, "One-Time Inspection," is revised to read (note - Section B.2.30 was previously revised, in its entirety, by FENOC Letter dated May 24, 2011 (ML11151A090)):

B.2.30 ONE-TIME INSPECTION Aging Management Program Elements Detection of Aging Effects A representative sample of the system and component population will be inspected using a variety of nondestructive examination methods, including visual inspection, volumetric inspection, and surface inspection techniques.

The sample size for the One-Time Inspections of the Pressurized-Water Reactor (PWR) Water Chemistry, Fuel Oil Chemistry, and Lubricating Oil Analysis Programs is 20 percent of the population (defined as components having the same material, environment, and aging effect combination) or a maximum of 25 components. The sample population will be determined by engineering evaluation, and where practical, will be focused on the (bounding or lead) components considered most susceptible to aging degradation due to time in service, the severity of the operating conditions, and the lowest design margin.

The in.pectio*n will

&b completed ,,th s,,ffii*e*t time to onsu.e that tho agig*

effects which may impact componont inendod fUnctionS early in the period o extended operation will be appropn~ate.'y m~anaged. At the samge time, th-e insections wil be timed to allow the compoenents to agtain sufficient age to ensure that aging effe-t* With kong i a-t,i.o..period can be identied-.

The inspections must occur within the 10-year period prior to the period of extended operation to be credited for the pro-gram.

LRA Amendment 12 Page 49 of 51 Affected LRA Section LRA Page No. Affected Paragraph and Sentence B.2.31 Page B-126 Program Description and and B-128 Conclusion - entire sections In response to RAI B.2.31-1, LRA Section B.2.31, "Open-Cycle Cooling Water Program," Program Description and Conclusion are revised as follows (note that these same changes to Section B.2.31 were previously provided by FENOC Letter dated May 24, 2011 (ML11151A090), and are repeated here for completeness):

B.2.31 OPEN-CYCLE COOLING WATER PROGRAM Program Description The Open-Cycle Cooling Water Program manages loss of material due to crevice, galvanic, general, pitting, and microbiologically-influenced corrosion (MIC), and also due to erosion for components located in the Service Water System, and for components connected to or cooled by the Service Water System, and also in the Circulating Water System. The program manages fouling due to particulates (e.g., corrosion products) and biological material (micro- and macro-organisms) resulting in reduction in heat transfer for heat exchangers within the scope of the program. in additn, the pr'egr'am manages cracking fo; coppor alloy greateFthan 15% zin conWonnt that are coolod by the Ser~wc~e W4ater-Systen+7 The Open-Cycle Cooling Water Programconsists of inspections, surveillances, and testing to detect and evaluate fouling7 and loss of material,and-GaGkigy combined with chemical treatments and cleaning activities to minimize fouling, and loss of material-,an-d .. aki;g. The existing program is a combination condition and performance monitoring and mitigation program that implements the recommendations of Generic Letter 89-13, "Service Water System Problems Affecting Safety-Related Equipment."

Conclusion The Open-Cycle Cooling Water Programhas been demonstrated to be capable of detecting and managing loss of material-,......G., and reduction in heat transferfor susceptible components in raw water environments. The Open-Cycle Cooling Water 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.

LRA Amendment 12 Page 50 of 51 Affected LRA Section LRA Page No. Affected Para-graph and Sentence B.2.34 Page B-138 New Enhancement In response to RAI B.2.34-1, LRA Section B.2.34, "Reactor Head Closure Studs Program," is revised to include a new enhancement, and now reads:

B.2.34 REACTOR HEAD CLOSURE STUDS PROGRAM Enhancements The following enhancements will be implemented in the identified program elements priorto entering the period of extended operation.

• Preventive Actions The Reactor Head Closure Studs pro-gram will preclude the future use of replacement closure stud bolting fabricated from material with actual measured yield strength greaterthan or equal to 150 ksi except for use of the existing spare reactorhead closure stud bolting.

LRA Amendment 12 Page 51 of 51 Affected LRA Section LRA Page No. Affected Para-graph and Sentence B.2.36 Page B-145 Aging Management Program Element:

"Acceptance Criteria" In response to RAI B.2.36-4, the Aging Management Program Element "Acceptance Criteria" for LRA Section B.2.36, "Selective Leaching Inspection," is replaced in its entirety, and now reads:

B.2.36 SELECTIVE LEACHING INSPECTION Aging Management Program Elements Acceptance Criteria The acceptance criteria are no visible evidence of selective leachinq or no more than a 20 percent decrease in hardness. For copper alloys with greater than 15 percent zinc, the criteria also includes no noticeable change in color from the normal yellow color to the reddish copper color.