Text

Reply to Request for Additional Information for the Review of License Renewal Application and License Renewal Application No. 15
	ML11264A059
Person / Time
Site:	Davis Besse
Issue date:	09/16/2011
From:	Allen B; FirstEnergy Nuclear Operating Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	L-11-252, TAC ME4640
	Download: ML11264A059 (144)
	v • d • e

FENOC 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 September 16, 2011 L-1 1-252 10 CFR 54 ATTN: Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555-0001

SUBJECT:

Davis-Besse Nuclear Power Station, Unit No. 1 Docket No. 50-346, License Number NPF-3 Reply to Request for Additional Information for the Review of the Davis-Besse Nuclear Power Station, Unit No. 1, License Renewal Application (TAC No. ME4640), and License Renewal Application Amendment No. 15 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 July 11, 2011 (ML11174A191), July 21, 2011 (ML11195A020), and August 11, 2011 (ML11216A236),

the Nuclear Regulatory Commission (NRC) requested additional information to complete its review of the License Renewal Application (LRA).

The content and submittal date of this letter was discussed during telephone conferences with Mr. Samuel Cuadrado de Jesus, NRC Project Manager, and, due to the expanded scope of the letter, the submittal date was deferred to a mutually agreeable submittal date of September 16, 2011. The Attachment provides the FENOC reply to NRC requests for additional information (RAIs), in the order listed as follows:

4 of 4 RAIs in NRC letter dated July 11,2011 (ML11174A191)

- Includes RAIs B.2.32-1; B.2.32-2; B.2.32-3; and, 3.1.2.2-3 2 of 13 RAIs in NRC letter dated July 21, 2011 (ML11195A020)

- Includes RAIs B.2.22-5 and B.2.39-9 2 of 3 RAIs in NRC letter dated August 11,2011 (ML11216A236)

- Includes RAts B.2.34-2 and 3.1.2.2-3 5 Supplemental RAI responses Includes supplemental responses for RAIs 3.2.2.2.3.6-2 and 3.3.2.2.5-2, and for questions regarding makeup pump casing inspections, abandoned equipment, and LRA Table 3.3.2-14 A _

Davis-Besse Nuclear Power Station, Unit No. 1 L-1 1-252 Page 2 The NRC request is shown in bold text in the Attachment followed by the FENOC response. The Enclosure provides Amendment No. 15 to the DBNPS LRA.

The responses to the remaining 11 of 13 RAIs in NRC letter dated July 21, 2011 (ML11195A020), were provided in FENOC letters dated August 17, 2011 (7 of 13 RAIs; ML11231A966), and August 26, 2011 (4 of 13 RAIs; ML11242A166).

The response to the remaining RAI (RAI 4.3.2.3.2-1 -(Supplement) regarding Class 1 valve fatigue analyses) in NRC letter dated August 11, 2011 (ML11216A236), is on-hold pending response from a vendor regarding the requested analyses.

The 5 supplemental RAIs originated during telephone conference calls with the NRC held on August 22, 29, and September 7, 2011, as described in the Attachment.

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 September __, 2011.

Sincerely, Barry S. llen

Attachment:

Reply to Request for Additional Information for the Review of the Davis-Besse Nuclear Power Station, Unit No. 1 (DBNPS), License Renewal Application, Sections B.2.22, B.2.32, B.2.34, B.2.39 and 3.1.2

Enclosure:

Amendment No. 15 to the DBNPS License Renewal Application cc: NRC DLR Project Manager NRC Region III Administrator

Davis-Besse Nuclear Power Station, Unit No. 1 L-1 1-252 Page 3 cc: w/o Attachment or Enclosure NRC DLR Director NRR DORL Project Manager NRC Resident Inspector Utility Radiological Safety Board

Attachment L-1 1-252 Reply to Request for Additional Information for the Review of the Davis-Besse Nuclear Power Station, Unit No. 1 (DBNPS), License Renewal Application, Sections B.2.22, B.2.32, B.2.34, B.2.39 and 3.1.2 Page 1 of 33 Section B.2.32 Question RAI B.2.32-1 B3ackgqround:

By letter dated December 31, 2008, the Electric Power Research Institute submitted Materials Reliability Program (MRP) Report 1016596 (MRP-227),

Revision (Rev.) 0, "Pressurized Water Reactor (PWR) Internals Inspection and Evaluation Guidelines," for U.S. Nuclear Regulatory Commission (NRC or the staff) review and approval.

The staff has reviewed MRP-227, Rev. 0, and determined that its guidance will provide acceptable levels of quality and safety with respect to inspection and evaluation (I&E) of reactor vessel (RV) internal components in pressurized water reactor (PWR) vessels supplied by Westinghouse, Babcock and Wilcox, and Combustion Engineering. MRP-227, Rev. 0 provides I&E guidelines for implementation by license renewal applicants and licensees with renewed operating licenses in their plant-specific aging management programs (AMPs) for PWR RV internal components. However the MRP-227, Rev. 0, guidelines will be amended, as specified in the conditions and limitations identified in Section 4.1 of the staffs Safety Evaluation (SE) concerning MRP-227, Rev. 0, which is currently under development.

Section 4.2 of the SE for MRP-227, Rev. 0, identifies applicant/licensee plant-specific action items that will need to be addressed on a plant-specific basis by license renewal applicants or licensees with renewed operating licenses.

These action items address topics related to the plant-specific implementation of MRP-227, as amended by the staffs SE, that could not be effectively addressed on a generic basis in MRP-227, as amended by the staffs SE.

Applicant action item 7 from Section 4.2 of the staff's SE for MRP-227, Rev. 0, requires that license renewal applicants submit a plant-specific application to implement a new AMP for the RV internal components that is based upon the implementation of the MRP-227 guidelines, as amended by the staffs SE for MRP-227, and that applicants' AMP submittals shall include the specific information identified in items (1) through (5) in Section 3.5.1 of the SE for MRP-227, Rev. 0.

Attachment L-1 1-252 Page 2 of 33 Issue:

FirstEnergy Nuclear Operating Company (the applicant) has provided a plant-specific AMP submittal for the RV internals as part of the Davis-Besse LRA.

The AMP submittal is provided in LRA Section B.2.32, "PWR Reactor Vessel Internals Program."

As stated in applicant action item 7 from Section 4.2 of the staffs SE for MRP-227, Rev. 0, the applicant's PWR RV internals AMP submittal shall include the following information identified in Section 3.5.1 of the SE for MRP-227:

(1) An AMP for the facility shall address the 10 program elements as defined in Generic Aging Lessons Learned (GALL) Report, NUREG-1801, Rev. 2, December 2010, Chapter Xl, AMP XI.M16A, "PWR Vessel Internals." The staff notes that LRA Section B.2.32 states that the Davis-Besse RV internals AMP is evaluated against the 10 elements described in Appendix A.1, Section A.1.2.3 of NUREG-1800, "Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants." LRA Section B.2.32 does not specifically address the 10 program elements defined in GALL AMP XI.MI6A from Rev. 2 of the GALL.

(2) To ensure the MRP-227, Rev. 0, guidelines, as amended by the staff's SE, and the plant-specific action items identified in the staff's SE will be carried out by applicants/licensees, applicants/licensees are to submit an inspection plan for staff review and approval consistent with the licensing basis for the plant. The applicant has not identified any plant-specific inspection plan for ensuring the implementation of MRP-227, Rev. 0, guidelines and plant-specific action items, as identified in the staff's SE for MRP-227.

(3) Applicants for license renewal referencing MRP-227, Rev. 0, for aging management of the RV internal components shall ensure that the programs and activities specified in MRP-227, Rev. 0, as amended by the staff's SE, are summarily described in the Final Safety Analysis Report (FSAR) supplement.

(4) If the plant's current licensing basis includes specific inspection or analysis requirements for RV internal components in either the operating license for the facility or in the facility's technical specifications (TS's),

and these requirements are more comprehensive than the corresponding I&E guidelines in the MRP-227 report, as amended by the staff's SE, the requirements of the applicable license conditions or TSs take precedence over the corresponding I&E guidelines of the MRP report.

(5) Applicants/licensees who implement MRP-227 must evaluate the current licensing basis for their facilities to determine if they have plant-specific

Attachment L-1 1-252 Page 3 of 33 time-limited aging analyses (TLAAs), applicable to RV internal components, which must be addressed. These TLAAs shall be submitted to the NRC for review along with the submittal for an AMP implementing MRP-227 guidelines, as modified by the staffs SE. The applicant has identified a TLAA applicable to the RV internal components. This TLAA addresses the reduction in fracture toughness for the RV internal components and is described in LRA Section 4.2.7.

The staff notes that, in order for plant-specific PWR RV internals AMPs to be consistent with GALL AMP XI.M16A from Rev. 2 of the GALL, applicants'/licensees' PWR RV internals AMP submittals must address all applicable plant-specific/vendor-specific action items established in Section 4.2 of the staff's SE for MRP-227.

Request:

The staff requests that the applicant revise the Davis-Besse PWR RV internals AMP description provided in LRA Section B.2.32 and the corresponding Updated Safety Analysis Report (USAR) Supplement (LRA Section A.1.32) to address each of the five plant-specific AMP information requirements identified in Section 3.5.1 of the SE for MRP-227 (as required by applicant action item 7 from Section 4.2 of the staffs SE for MRP-227) as follows:

(1) LRA Section B.2.32 should be revised to address the 10 elements of an acceptable AMP described in GALL AMP XI.M16A (GALL, Rev. 2, December 2010). Specifically, the AMP's description of the 10 elements should be revised/supplemented as follows to ensure consistency with the 10 elements in GALL AMP XI.M16A:

Element 1, Scope of Program: With respect to program scope, LRA Section B.2.32 must be supplemented to address the applicable plant-specific and vendor-specific license renewal applicant action items (LRAAIs) on the MRP-227 methodology identified in Section 4.2 of the staffs SE for MRP-227, including all the programs and activities discussed in the LRAAI responses credited for aging management of reactor vessel internal (RVI) components. The staff notes that the LRAAIs are identified in Section 4.2 of the staffs SE on MRP-227.

Element 2, Preventive Actions: The applicant's description of preventive actions in LRA Section B.2.32 is not consistent with Element 2 from GALL AMP XI.M16A. Please revise/supplement the description of preventive actions in LRA Section B.2.32 to ensure that it is consistent with Element 2 from GALL AMP XI.M16A.

Attachment L-1 1-252 Page 4 of 33 Elements 3 through 10: Please revise/supplement the description of these program elements in LRA Section B.2.32, as necessary, to ensure that they are consistent with GALL AMP XI.M16A in Rev. 2 of NUREG-1801.

In addition to the above, please revise the subsection in LRA Section B.2.32, titled "NUREG-1801 Consistency," to state that the corresponding AMP is described in NUREG-1801, Rev. 2, GALL Chapter Xl, AMP XI.M16A and that the Davis-Besse PWR Reactor Vessel Internals Program is evaluated against the 10 elements described in GALL AMP XI.M16A.

(2) The applicant has not identified any plant-specific inspection plan in LRA Section B.2.32 for ensuring the implementation of MRP-227, Rev. 0, guidelines and plant-specific action items, as identified in the staffs SE for MRP-227. It is necessary for the applicant to either: (1) provide the plant-specific inspection plan based on the final approved version of MRP-227 and the applicant's responses to the plant-specific action items identified in the staffs SE for MRP-227; or, (2) provide a specific license renewal commitment to submit the plant-specific inspection plan to the NRC for review and approval no later than two years after issuance of the renewed operating license or two prior to the beginning of the period of extended operation, whichever is earlier.

The staff recognizes that it may not be feasible to submit an adequate inspection plan as part of this RAI response. Therefore, please supplement LRA Sections B.2.32 and A.1.32 to state that a plant-specific inspection plan for ensuring the implementation of MRP-227 program guidelines, as amended by the staffs SE for MRP-227, and Davis-Besse's responses to the plant-specific action items, as identified in the staffs SE, will be submitted for NRC staff review and approval. Also, please include a specific Davis-Besse License Renewal Commitment to submit for NRC review and approval the required plant-specific inspection plan that is based on the final approved version of MRP-227 (as modified by the staff's SE) and Davis-Besse's responses to all applicable plant-specific action items identified in the staffs SE for MRP-227. This license renewal commitment shall require the submittal of this inspection plan for NRC review and approval no later than two years after issuance of the renewed operating license or two years prior to the beginning of the period of extended operation, whichever is earlier.

(3) LRA Section A.1.32 provides the USAR supplement summary description for the Davis-Besse PWR RV Internals Program. The USAR supplement summary description for this program references the I&E guidelines of MRP-227 and states that the PWR RV Internals Program will be revised, as necessary, to incorporate the final recommendations and requirements published in the staff-approved version of MRP-227. Please include a statement in LRA Section A.1.32 indicating that the PWR RV Internals

Attachment L-11-252 Page 5 of 33 Program will address all plant-specific action items applicable to Davis-Besse that are established in Section 4.2 of the staffs SE for MRP-227.

(4) The staff found that the Davis-Besse current licensing basis TSs require that the core support shield assembly (CSS) vent valve components be inspected on a 24-month cycle. However, MRP-227, Rev. 0, Table 4-1, "B&W plants Primary components," states that the CSS vent valve discs, top retaining ring, bottom retaining ring, disc shaft, and hinge pin will be examined using VT-3 visual methods during the next 10-year in-service inspection (ISI) interval and that subsequent VT-3 examinations will occur on the plant's 10-year ISI interval. As stated above for this item, the requirements of the applicable license conditions or TSs take precedence over the corresponding I&E guidelines of the MRP report. Therefore, based on the above discrepancy between the TS requirements and MRP guidelines for examination frequency for the CSS vent valve components, please provide the information and LRA revisions necessary for resolving this discrepancy. Specifically, either: (1) identify the applicable TS requirement as the governing inspection requirement for the CSS vent valve components, and accordingly, revise the PWR RVI AMP description in LRA Section B.2.32 to state that these TS requirements take precedence over the applicable MRP-227 I&E guidelines for the CSS vent valve components; or (2) submit the necessary TS changes, including technical justification for the proposed TS changes, to implement the less comprehensive MRP-227 I&E guidelines for the CSS vent valve components, in accordance with Title 10 of the Code of Federal Regulations (10 CFR) 54.22.

Similarly to the situation described above for the CSS vent valve components, please provide additional information concerning any other discrepancies between Davis-Besse current licensing basis requirements and the applicable MRP-227 guidelines for I&E of RVI components, and provide the information and LRA revisions necessary for resolving these discrepancies. Specifically, either; (1) identify the applicable TS requirement as the governing requirement for the component, and accordingly, revise the PWR RVI AMP description in LRA Section B.2.32 to state that the applicable TS requirements take precedence over the applicable MRP-227 I&E guidelines for the component; or, (2) submit the necessary TS changes, including technical justification for the proposed TS changes, to implement the less comprehensive MRP-227 I&E guidelines for the component, in accordance with 10 CFR 54.22.

(5) The applicant has identified one TLAA applicable to the RV internal components. This TLAA addresses the reduction in fracture toughness for the stainless steel (SS) RV internal components and is described in LRA Section 4.2.7. The description of this TLAA in LRA Section 4.2.7 is limited to the effects of neutron embrittlement on the SS RV internal

Attachment L-1 1-252 Page 6 of 33 components' deformation limits and the corresponding ability of the SS RV internal components to absorb local strain at the regions of maximum stress intensity. Please supplement the PWR RV internals AMP description in LRA Section B.2.32 to state how this TLAA will be managed for the period of extended operation (i.e., that reduction in fracture toughness for RV internals will be managed in accordance with the implementation of the MRP-227 guidelines, as modified by the staff's SE, including all activities associated with Davis-Besse's responses to plant-specific action items identified in the staff's SE for MRP-227.)

RESPONSE RAI B.2.32-1 The Davis-Besse PWR Reactor Vessel Internals Program provided in LRA Section B.2.32 and the corresponding Updated Safety Analysis Report (USAR)

Supplement (LRA Section A.1.32) are revised to address each of the five plant-specific AMP information requirements identified in Section 3.5.1 of the safety evaluation for MRP-227. Highlights of the changes are as follows:

(1) LRA Section B.2.32 is revised to address the 10 elements of an acceptable aging management program (AMP) described in NUREG-1801, Rev. 2,Section XI.M16A. The "Scope of Program" element is revised to address license renewal applicant action items (LRAAIs). The "Preventive Actions" element and program elements 3 through 10 are revised for consistency with NUREG-1801, Rev. 2. The NUREG-1801 consistency statement is revised to identify the PWR Reactor Vessel Internals Program as a new Davis-Besse program that will be consistent with the 10 elements of an effective aging management program as described in NUREG-1801, Rev. 2,Section XI.M16A, "PWR Vessel Internals."

(2) LRA Sections A.1.32 and B.2.32 are revised to state that a plant-specific inspection plan for ensuring the implementation of MRP-227 program guidelines, as amended by the safety evaluation for MRP-227, and the FENOC responses to the plant-specific action items, as identified in Section 4.2 of the safety evaluation for MRP-227, will be submitted for NRC review and approval.

In addition, in LRA Table A-i, a specific Davis-Besse license renewal future commitment is provided to submit for NRC review and approval the required plant-specific inspection plan that is based on the final approved version of MRP-227 (as amended by the staff's safety evaluation) and the FENOC responses to the applicable plant-specific action items identified in the staff's safety evaluation for MRP-227. The license renewal future commitment requires the submittal to be no later than two years after issuance of the renewed operating license or two years prior to the beginning of the period of extended operation, whichever is earlier.

Attachment L-1 1-252 Page 7 of 33 (3) LRA Sections A.1.32 and B.2.32 are revised to state that the Davis-Besse PWR Reactor Vessel Internals Program will address all plant-specific action items applicable to Davis-Besse that are established in Section 4.2 of the safety evaluation for MRP-227.

(4) LRA Sections A.1.32 and B.2.32 are revised to provide the following.

MRP-227 I&E guidelines require a visual (VT-3) examination of the core support shield (CSS) vent valve retaining rings and disc shaft for every 10 year Inservice Inspection Interval. In addition, Davis-Besse Technical Specification 5.5.4 requires testing of the CSS vent valves every 24 months to verify by visual inspection that the valve body and valve disc exhibit no abnormal degradation, verify the valve is not stuck in an open position, and verify by manual actuation that the valve is fully open when a force of <400 lbs is applied vertically upward. The technical specification inspection will continue to be performed at the prescribed frequency of 24 months. The MRP-227 required visual (VT-3) examination will also be performed at the prescribed frequency of every 10 year Inservice Inspection Interval.

(5) LRA Sections A.1.32 and B.2.32 are revised to state that the program includes management of the time-limited aging analysis (TLAA) for reduction in fracture toughness of the reactor vessel internals. This TLAA will be managed in accordance with the implementation of the MRP-227 guidelines, as amended by the MRP-227 safety evaluation, including all activities associated with the FENOC responses to plant-specific action items identified in the Section 4.2 of the safety evaluation.

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

Question RAI B.2.32-2 According to Section A.1.4 in MRP-1175, "Materials Reliability Program: PWR Internal Aging Degradation Mechanism Screening Threshold Values,"

susceptibility to stress corrosion cracking (SCC) in nickel-based Alloy X-750 PWR RV internal components depends on the type of heat treatment that is performed on the alloy. High temperature heat treatment (HTH) processes that are used on Alloy X-750 components offer better resistance to SCC than the other age hardened heat treatment processes. The type of heat treatment applied to Alloy X-750 PWR RV internal components is a critical parameter for ensuring that the Davis-Besse PWR RV internals AMP will adequately manage the effects of aging due to SCC for the Alloy X-750 components. Please provide information related to the type of heat treatment process that was used for the Alloy X-750 RV internal components at Davis-Besse.

Attachment L-11-252 Page 8 of 33 Please state whether there are any RV internal components fabricated from Alloy X-750 with heat treatment other than HTH, as described above. For any such non-HTH X-750 RV internal components, discuss how the effects of aging due to SCC will be managed, outside the scope of the I&E guidelines described in MRP-227.

RESPONSE RAI B.2.32-2 There are two types of components fabricated from Alloy X-750 inside the Davis-Besse reactor vessel internals:

" Alloy X-750 replacement bolts and the associated Alloy X-750 compression collars were installed in 1984 and 1990. These Alloy X-750 replacement bolts and compression collars are the HTH condition.

Alloy X-750 dowels fabricated to the Aeronautical Material Specifications (AMS)-5667F specification were installed during original plant construction. The AMS-5667F heat treatment requirement listed below is also called the AH condition (Reference AMS-5667F, 1960). These Alloy X-750 dowels are in this heat treatment condition:

o Equalize heat treatment at 1625+/-25F for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , followed by air cooling o Precipitation heat treatment at 1300+/-25F for 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months , followed by air cooling The above two types of Alloy X-750 components inside the Davis-Besse reactor vessel internals are further discussed below.

1. The Alloy X-750 HTH bolts and associated compression collars are listed below:

a. Replacement upper core barrel (UCB) bolts and compression collars

b. Replacement lower core barrel (LCB) bolts and compression collars

c. Replacement lower thermal shield (LTS) bolts and compression collars

d. Replacement surveillance holder tube (SSHT) bolts and compression collars The above Alloy X-750 HTH bolts are categorized either as "primary" or "expansion" in MRP-227-Rev. 0 due to SCC concern. Because the compression collars are under compressive stress, SCC is not applicable to the compression collar. Therefore, the Alloy X-750 HTH compression collars are categorized as "No Additional Measures" in MRP-227-Rev. 0.

Attachment L-1 1-252 Page 9 of 33

2. Alloy X-750 dowels fabricated to the AMS-5667F are used for the following applications:

a. Some Alloy X-750 dowels were used to aid assembly process such as to align the components before they were joined by bolting. Once assembled, these dowels were completely enclosed. Because these Alloy X-750 dowels do not have any design function, they were categorized as "No Additional Measures" in MRP-227-Rev. 0. For example, the Alloy X-750 core barrel-to-former plate dowels were used to align the former plates with the core barrel cylinder at the top and bottom former plate level (16 dowels at each level).

After the former plates were bolted to the core barrel cylinder, the core barrel-to-former plate dowels no longer had a design function.

b. Alloy X-750 dowels are used in the lower grid fuel assembly support pads.

These dowels have a design function to join the pads to the lower grid rib section in conjunction with Type 304 screws and pad-to-lower grid welds.

Based on the FMECA (Failure Modes, Effects, and Criticality Analysis), the Alloy X-750 dowels for the lower grid fuel assembly support pads are categorized as "Not A" for irradiation embrittlement. All other aging degradation mechanisms including SCC are categorized as "A," (i.e., below the screening criteria and therefore, deemed not susceptible to the age-related degradation mechanism). SCC is categorized as "A" because the stress level is below the MRP-175 SCC screening level for Alloy X-750.

Therefore, these Alloy X-750 dowels are categorized as "expansion" in MRP-227-Rev. 0 due to irradiation embrittlement. In addition to irradiation embrittlement concern, these Alloy X-750 dowels used nickel-base locking welds; the locking welds are categorized as "expansion" in MRP-227-Rev. 0 due to SCC and irradiation embrittlement concerns.

c. Alloy X-750 dowels are used in the upper grid fuel assembly support pads.

These dowels have a design function to join the pads to the upper grid rib section in conjunction with Type 304 screws. Unlike the Alloy X-750 dowels for the lower grid fuel assembly support pads, the Alloy X-750 dowels for the upper grid fuel assembly support pads do not have any locking welds and are completely enclosed by the fuel pads (i.e., inaccessible). This design feature is unique to Davis-Besse. Based on the FMECA (Failure Modes, Effects, and Criticality Analysis), the Alloy X-750 dowels for the upper grid fuel assembly support pads are categorized as Category "A" for all eight aging degradation mechanisms including SCC. SCC is categorized as "A" because the stress level is below the MRP-1 75 SCC screening level for Alloy X-750. Therefore, the Alloy X-750 dowels for the upper grid fuel assembly support pads are categorized as "No Additional Measures" in MRP-227-Rev. 0.

Attachment L-1 1-252 Page 10 of 33 Question RAI B.2.32-3 During the extended period of operation, cast austenitic stainless steel (CASS)

PWR RV internal components are susceptible to a reduction in fracture toughness due to the combined effects of neutron embrittlement and thermal embrittlement, and the potential for irradiation-assisted stress corrosion cracking (IASCC). The synergistic effects of neutron embrittlement and thermal embrittlement may lead to the potential for failure of CASS RV internal components under some design basis loading conditions. Please explain how the Davis-Besse PWR RV internals AMP, as described in LRA Section B.2.32, will account for the reduction in fracture toughness due to the synergistic effects of neutron embrittlement and thermal embrittlement when evaluating CASS components to determine susceptibility to reduction in fracture toughness. The staff notes, in particular, that CASS RV internal components should be initially screened based on casting method, ferrite content, and molybdenum to determine if the components are susceptible to thermal embrittlement, and that components deemed susceptible to thermal embrittlement based on the above screening criteria should receive either supplemental examinations or a component-specific evaluation to ascertain the reduction in fracture toughness due to the synergistic effects of neutron embrittlement and thermal embrittlement.

RESPONSE RAI B.2.32-3 CASS reactor vessel internal components were initially screened based on casting method, percent ferrite, and molybdenum content to determine if the components were susceptible to thermal embrittlement. Components that were determined to be susceptible to thermal embrittlement and that are expected to exceed a fluence of 1 x 1017 n/cm 2 for 60-years of operation are the incore monitoring instrumentation (IMI) guide tube spiders and control rod guide tube (CRGT) spacer castings.

As provided in Section 3.3.7 of the NRC safety evaluation for MRP-227, the MRP identified that some cast austenitic stainless steel (CASS) RVI components require a plant-specific analysis to demonstrate that their structural integrity and functionality are maintained during the extended period of operation. For Babcock & Wilcox (B&W) designed plants, the safety evaluation provided that the applicants/licensees shall develop a plant-specific analysis for the IMI guide tube spiders and CRGT spacer castings to demonstrate that these components will maintain their functions during the period of extended operation and that these analyses should consider the possible loss of fracture toughness in these components due to thermal and irradiation embrittlement.

Section 4.2.7 of the NRC safety evaluation for MRP-227 provides an action item for plant-specific evaluation of CASS materials. For B&W design plants, this action item requires the development of plant-specific analyses to demonstrate that IMI guide tube spiders and CRGT spacer castings will maintain their functionality during the period of

Attachment L-1 1-252 Page 11 of 33 extended operation. These analyses should also consider the possible loss of fracture toughness in these components due to thermal and irradiation embrittlement.

As revised by RAI B.2.32-1, above, the PWR Reactor Vessel Internals Program will address all plant-specific action items applicable to Davis-Besse that are established in Section 4.2 of the NRC safety evaluation for MRP-227, and a commitment is provided to ensure that these action items will be submitted for NRC review and approval.

Section 3.1.2 Question RAI 3.1.2.2-3 (from NRC letter dated July 11, 2011 (ML11174A191))

LRA Table 3.1.2-2, "Aging Management Review Results - Reactor Vessel Internals," does not address certain RVI components that are listed in the Aging Management Review (AMR) line items found in Chapter IV, Section B4, "Reactor Vessel Internals (PWR) - Babcock and Wilcox," of the GALL Report, Rev. 2.

GALL, Chapter IV, Section B4. Item No. IV.B4.RP-245 - "Core barrel assembly:

(a) upper thermal shield bolts; (b) surveillance specimen holder tube bolts (Davis-Besse, only); (c) surveillance specimen tube holder studs, and nuts (Crystal River Unit 3, only)." For this line item, the staff is only concerned with (b), "surveillance specimen holder tube bolts (Davis-Besse, only)."

GALL, Chapter IV, Section B4, Item No. IV.B4.RP-253 - "Core support shield (CSS) assembly: (a) CSS cast outlet nozzles (Oconee Unit 3 and Davis-Besse, only); (b) CSS vent valve discs." For this line item, the staff is only concerned with (a), "CSS cast outlet nozzles (Oconee Unit 3 and Davis-Besse, only)."

GALL, Chapter IV, Section B4. Item Nos. IV.B4.RP-259 (Incore Monitoring Instrumentation Guide Tube Assembly), IV.B4.RP-260 (Lower Grid Assembly),

IV.B4.RP-262 (Lower Grid Assembly), IV.B4.RP-261 (Lower Grid Assembly) -

Each of these line items identifies specific welds in the subject assembly.

These specific welds are not identified in the listing of components for the corresponding assemblies in LRA Table 3.1.2-2.

If LRA Table 3.1.2-2 addresses the subject components, as identified above, please state the Table 3.1.2-2 Row No. where these components are listed.

Otherwise, please supplement LRA Table 3.1.2-2 to address these specific components (for which the staff identified a concern), as stated in the GALL.

Rev. 2 AMR line items, and identify the Davis-Besse aging management programs applicable to the management of aging effects for these RV internal components.

Attachment L-1 1-252 Page 12 of 33 RESPONSE RAI 3.1.2.2-3 (from NRC letter dated July 11,2011 (ML11174A191))

In response to RAI 3.1.2.2-2 submitted by FENOC letter dated July 22, 2011 (ML11208C274), LRA Table 3.1.2-2, "Aging Management Review Results- Reactor Vessel Internals," was replaced in its entirety. The revised table included item numbers IV.B4.RP-245, IV.B4.RP-253, IV.B4.RP-259, IV.B4.RP-260, IV.B4.RP-261, and IV.1B4.RP-262 of NUREG-1801, "Generic Aging Lessons Learned (GALL) Report,"

Revision 2. LRA Table 3.1.2-2 is being replaced in its entirety in response to RAI 3.1.2.2-3 from NRC letter dated August 11,2011 (ML11216A236), below, and is contained in the Enclosure to this letter. The line item numbers identified above are also included in the revised table.

Section B.2.22 Question RAI B.2.22-5

Background:

By letter dated May 24, 2011, FirstEnergy Nuclear Operating Company (FENOC or the applicant) responded to a staff request for additional information (RAI)

B.2.22-1 regarding the presence of standing water in the containment annulus pocket region and observed areas of corrosion on the containment exterior surface. In its response to the RAI, the applicant stated that FENOC plans to perform non-destructive testing (NDT) at a minimum of three representative locations. The applicant also stated that FENOC plans to inspect and maintain the accessible materials in the annulus sand pocket area. The applicant further stated that FENOC conducted thorough evaluation of the containment vessel corrosion in July of 2002. The report concluded that integrity of the containment vessel will be maintained with negligible additional corrosion in the future.

Issue:

IWE 1241 (a) requires augmented examination of the containment interior and exterior surface areas that are subject to accelerated corrosion with no or minimal corrosion allowance or areas where absence or repeated loss of protective coatings has resulted in substantial corrosion or pitting. Typical locations of such areas are those exposed to standing water, repeated wetting and drying, persistent leakage, and those with geometrics suitable for water accumulation, condensation, and microbiological attack. Such areas may include penetration sleeves, surface wetted during refueling, concrete-to-steel shell or liner interfaces, embedment zones, leak chase channels, drain areas, or sump liners.

Attachment L-1 1-252 Page 13 of 33 During the audit, the staff found that there is a history of ground water infiltration into the annulus area, and reviewed documentation (CR-72660, dated April 2, 2010) that indicated presence of standing water in the annulus sand pocket region. In addition, this condition report documented corrosion and peeling of the coating on the exterior surface of the containment shell, and deterioration of the coating applied to the top of the sand pocket. IWE Table IWE-2500-1, Examination Category E-C requires 100% UT measurement of the area designated for augmented examination during each inspection period until the areas examined remain essentially unchanged for three consecutive inspection periods. In the RAI response, the applicant did not explain why a one-time NDT examination at three locations in sand pocket region that is about 300-400 feet long, prior to period of extended operation, is appropriate in lieu of IWE-1241(a) and Table IWE-2500-1 requirements. In addition, the applicant did not provide specific details for inspecting and maintaining the accessible materials in the annulus sand pocket area.

Request:

1. Provide technical justification for not following the requirements of IWE-1241(a) and Table IWE-2500-1 for performing UT examination of 100%

of the area designated for augmented examination during each inspection period until the area remains essentially unchanged for three consecutive inspection periods. The staff is concerned that one-time NDT examination at three locations in sand pocket region that is about 300-400 feet long, prior to period of extended operation, may not be able to detect and establish a trend of the potential degradation of the steel containment over the long term.

2. Provide details and schedule of specific actions FENOC has planned to minimize water seepage into the sand pocket region.

3. Provide specific details and requirements for inspection, maintenance, and repair of the annulus sand pocket accessible and inaccessible areas, including the replacement of deteriorated grout and coating.

RESPONSE RAI B.2.22-5

1.Section XI of American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (ASME Code) 1995 Edition through 1996 Addenda, IWE-1232(a) states that "[p]ortions of Class MC containment vessels, parts, and appurtenances that are embedded in concrete or otherwise made inaccessible during construction of the vessel or as a result of vessel repair, modifications, or replacement are exempted from examination...."

Attachment L-1 1-252 Page 14 of 33 The provisions for using the IWE-1232(a) exemption are satisfied and therefore, ASME Code Section Xl, Subsection IWE-1241 is not applicable to the subject inaccessible surfaces.

However, 10 CFR 50.55a(b)(2)(ix)(A) requires that for Class MC applications, the licensee shall evaluate the acceptability of inaccessible areas where conditions exist in accessible areas that could indicate the presence of or result in degradation to such inaccessible areas. For each inaccessible area identified, the licensee shall provide the following in the Inservice Inspection (ISI) Summary Report.

(1) A description of the type and estimated extent of degradation, and the conditions that led to the degradation; (2) An evaluation of each area, and the result of the evaluation, and; (3) A description of necessary corrective actions.

An evaluation of the acceptability of the inaccessible area adjacent to the sand pocket region was performed in accordance with 10 CFR 50.55a(b)(2)(ix)(A) and reported to the NRC in FENOC letter, "Inservice Inspection Summary Report of Cycle 13 and 13th Refueling Outage Activities for the Davis-Besse Nuclear Power Station," dated June 23, 2004 (ADAMS Accession Number ML041180104). The letter (ML041180104) reported that water seepage in the sand pocket region near the Containment Vessel base in the annulus between the Containment Vessel and the Shield Building had occurred. Light corrosion stains on the protective coating on the Containment Vessel exterior surface in the sand pocket region had been noted. Later, in 2010, water seepage and areas of minor surface corrosion were reported in the sand pocket region by Condition Report (CR) 10-72660.

The evaluation performed by CR 10-72660 reported that no loss of base metal appeared to have occurred in the areas with minor surface corrosion and, based on a comparison with photographs, the surface corrosion had remained essentially unchanged since the Cycle 14 refueling outage. The Investigation Summary for CR 10-72660 includes an evaluation of the areas of minor surface corrosion. The Investigation Summary evaluation for CR 10-72660 noted that it appears that due to the environmental conditions present at the bottom of the Annulus Sand Pocket Region (damp with oxygen present), the coating is beginning to degrade. Based on the water sample data taken from the Cycle 16 refueling outage, an average iron content of 4.94 parts per million was found in the water sampled from the Annulus Sand Pocket Region. That value is relatively low and is not a sign of any significant corrosion of the Containment Vessel. Therefore, the evaluation reported in Letter ML041180104 bounds the conditions described in CR 10-72660.

Attachment L-1 1-252 Page 15 of 33

2. At this time, FENOC plans to continue to monitor the sand pocket region for aging degradation. A protective coating has been previously applied to the Containment Vessel in the accessible affected areas. Pooling of the ground water against the containment vessel surface is minimized by annulus drains and by grout installed with a slope to direct water away from the Containment Vessel toward the Shield Building side of the annulus. No specific actions are planned by FENOC to further minimize water seepage into the sand pocket region. Undefined pathways of groundwater seepage locations below the surface of the annulus sand pocket region and the inaccessibility of the Shield Building foundation (more than 40 feet below grade and adjacent to the Auxiliary Building foundation) preclude practical repairs for full mitigation of the groundwater leakage. In accordance with the FENOC Corrective Action Program, FENOC plans to revisit this approach after each of the Containment Vessel inspections described in the Response to Item 3, or if the quantity of seepage or chemistry of the groundwater seepage indicate that further efforts to minimize the seepage are required.

3. FENOC plans to perform visual inspections of 100% of the accessible areas of the wetted outer surface of the Containment Vessel in the sand pocket region during each refueling outage. The visual inspections are planned to include accessible dry areas of the outer surface of the Containment Vessel and the areas above the grout-to-steel interface up to Elevation 567' + 3", - 1". Visual inspections of inaccessible surface areas of the Containment Vessel in the sand pocket region are also planned to be performed when such areas are made accessible for ultrasonic testing (UT) thickness examination. FENOC plans to record and evaluate indications of pitting or MIC, if found during inspections.

As practical access to the area is limited to refueling outage periods, water samples are planned to be taken during refueling outages for chemical analyses whenever sufficient water volumes are available in the sand pocket region. The number of sampled water volumes are expected to be determined by the number of water volumes observed and the size of those water volumes.

If sufficient water is available, the samples are planned to be analyzed for pH, chlorides, iron and sulfates. The water samples may be taken at different times during each outage. If sufficient water is not available for the full set of analyses, engineering judgment should be used to determine the priority of the chemical analyses to be performed. If the concentration of chlorides is determined to be greater than 250 parts per million (ppm), the sand pocket region is planned to be treated or washed or some combination thereof to reduce the measured chloride concentrations to less than 250 ppm.

LRA Table A-i, "Davis-Besse License Renewal Commitments," license renewal future Commitment 35, is revised to update the nondestructive testing (NDT) plan for the steel Containment Vessel in the sand pocket region. The Containment Vessel is planned to be examined at least twice by taking UT

Attachment L-1 1-252 Page 16 of 33 thickness measurements from the outer surface in accordance with the following criteria:

a) At five areas of previously identified groundwater in-leakage.

b) A minimum of three vertical grid locations at 12 inches nominal horizontal spacing are planned to be examined at each of the above areas.

c) At each of the above locations, the vessel is planned to be examined at a minimum of three elevations: 1) at approximately three inches below the existing grout-to-vessel interface level in the sand pocket region; 2) at the existing grout-to-vessel interface level in the sand pocket region; and, 3) at approximately three inches above the existing grout-to-vessel interface level in the sand pocket region.

d) The first examination is planned to be performed in 2014 and a second examination is planned to be performed by 2025 (scheduled for the Cycle 23 refueling outage).

The UT thickness readings are planned to be compared to minimum ASME Code vessel thickness requirements and the results obtained during previous UT thickness examinations of the Containment Vessel.

The need for maintenance or repair of the Containment Vessel is planned to be determined based on the results and evaluation of the examinations.

The capability for detecting a potential degradation trend of the steel Containment Vessel over the long term will not be limited by the results obtained during the 2014 UT thickness examination and the scheduled Cycle 23 UT thickness examination. When the second scheduled UT thickness examination is performed, the examination is planned for areas exposed to a damp environment for approximately 30 years. The vessel nominal plate thickness is known and there are earlier UT thickness examination results for comparison.

Therefore, the 2014 and the scheduled Cycle 23 examination results, combined with the other known information, should provide sufficient information for detection of a trend of the potential degradation of the steel Containment Vessel over the longer term. The 2014 and scheduled Cycle 23 examination results are planned to be documented using the work order system. Adverse conditions are also planned to be documented and evaluated in accordance with the FENOC Corrective Action Program for an evaluation of potential degradation of the steel Containment Vessel thickness over the longer term. The need for additional examinations and the appropriate frequency of examination are planned to be determined in accordance with the Corrective Action Program, if such examinations are needed.

During each refueling outage, 100% of the accessible grout in the sand pocket region is planned to be visually inspected for deteriorated grout (e.g., missing or damaged). Accessible grout is defined by FENOC as the normally exposed

Attachment L-1 1-252 Page 17 of 33 surface of the grout in the sand pocket region. Inaccessible grout is defined by FENOC as the grout below the normally exposed grout surface in the sand pocket region. Areas exposed by removal of the grout for UT thickness examination are also planned to receive visual inspection. Descriptions of deteriorated grout areas are planned to be entered in the FENOC Corrective Action Program for evaluation and corrective actions to address the conditions.

Visual inspection of the containment vessel coating, accessible from the annulus, is included in the existing Maintenance Rule structures evaluation procedure that is being enhanced for the License Renewal Structures Monitoring Program. As an example, CR 10-72660 was initiated on March 4, 2010, to document vessel coating condition findings of the Maintenance Rule Shield Building (Annulus) evaluation completed on May 3, 2010. In order to ensure further coatings and base material degradation does not occur, Order 200411438 was created to rework the coating applied to the vessel in the areas degraded during the Cycle 17 refueling outage.

Corrective Action #3 for CR 10-72660 was assigned to Outage Management to track Order 200411438 to completion in the Cycle 17 refueling outage, currently scheduled for Spring 2012.

CR 10-72660 also reported an observation of minor peeling of clear coating on the containment vessel. The Investigation Summary for CR 10-72660 includes an evaluation of the peeling of the clear coating. The Investigation Summary evaluation noted that the minor peeling will not have any effect on the protective coating system applied to the Containment Vessel. That issue is planned to be addressed by the corrections made in Order 200411438 scheduled for completion in the Cycle 17 refueling outage.

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

Section B.2.39 Question RAI B.2.39-9

Background:

By letter dated May 24, 2011, the applicant responded to a staff RAI B.2.39-1 regarding operating experience with borated water leakage from the reactor refueling cavity. In the response the applicant stated that in 2003 it had been determined that there was no concern with the structural integrity of the affected structures. The applicant further stated that the leakage was still occurring and

Attachment L-1 1-252 Page 18 of 33 committed to "continue to reduce or mitigate the refueling canal leaks inside containment prior to entering the period of extended operation" (Commitment 33).

Issue:

1. The applicant stated that the leakage has continued during outages since the initial investigation in 2003; however, the applicant did not explain why the 2003 investigation results remained applicable after eight years, or provide a plan to reconfirm the integrity of the affected structures prior to entering the period of extended operation. In addition, it is not clear from the RAI response, if all recommendations of the 2003 report, "Engineering Assessment Report-Refueling Canal Leakage," have been implemented.

2. Commitment 33, to reduce or mitigate the leakage, is vague and does not contain any information on possible mitigation techniques or timeframes.

The applicant also did not address what actions would be taken if they were unable to stop the leakage.

3. The response did not address any of the components, such as structural supports, that may be affected by the borated water leakage.

Request:

1. Demonstrate that the structural integrity of the affected structures is adequate prior to the period of extended operation. Either clearly explain why the results of the initial investigation will remain valid until the leakage is stopped, or commit to a plan to confirm the integrity of affected structures prior to the period of extended operation (e.g. concrete core bores, concrete removal to inspect reinforcing steel, etc.).

2. Confirm if all the recommendations of the 2003 report, "Engineering Assessment Report-Refueling Canal Leakage," have been implemented or justify why they are unnecessary.

3. Provide more information on Commitment 33. As a minimum this should include probable corrective actions and a preliminary schedule. The response should also include what actions will be taken if the leakage cannot be stopped, when the actions would occur, and why the proposed timeframe is acceptable.

4. Describe the detail and types of inspections, and repairs that have been performed or planned to manage aging of the components (pipe supports, conduit supports, piping etc.) that are exposed to borated water leakage and exhibit signs of corrosion, as identified in condition reports during the 2010 refueling outage.

Attachment L-1 1-252 Page 19 of 33 RESPONSE RAI B.2.39-9

1. Deposits of borated water were indentified on the walls of the core flood pipe tunnel (east/west tunnel) in 1998 as documented in Potential Condition Adverse to Quality Report (PCAQR) 98-0538. The non-destructive testing and petrographic examination of core drills from the east/west tunnel in 2003 found no significant degradation of the concrete or reinforcing steel. However, to ensure that the 2003 refueling canal leakage investigation results bound the effects to the concrete or reinforcing steel, additional investigations are planned to be conducted. LRA Table A-I, "Davis-Besse License Renewal Commitments," Commitment 33, is revised to include the FENOC plans to confirm that the integrity of affected structures in containment is adequate prior to the period of extended operation (e.g., concrete core bores, concrete removal to inspect reinforcing steel, etc.).

FENOC currently plans to perform the following activities by the end of 2014 to confirm the integrity of affected structures:

" Perform a core bore in the south wall of the east/west section of the core flood pipe tunnel; this area has been previously identified as a leakage path since at least 1998.

Perform a visual examination of the concrete. Test the concrete core bore sample for compressive strength and subject it to petrographic examination to assess degradation, if any, resulting from borated water exposure.

" Expose the reinforcing steel for examination. Inspect reinforcing steel for corrosion, and, if corrosion exists on the reinforcing steel and there is sufficient material present, then collect corrosion samples for evaluation.

Measure reinforcing steel wastage if wastage is observed, and measure corrosion buildup and evaluate concrete cracking if a buildup of corrosion products has caused cracking of the concrete.

" Enter degradation identified as a result of testing and examination into the Corrective Action Program and evaluate for impact on structural integrity of affected structures.

Conduct the evaluations in accordance with the requirements of the current licensing basis Maintenance Rule Program.

If leakage from the refueling canal has not been eliminated by the beginning of the period of extended operation, then FENOC plans to evaluate the concrete again, in a similar manner, within 6 years after entering the period of extended operation, and every 10 years thereafter. However, for these later evaluations, FENOC plans to use American Concrete Institute (ACI) Report 349.3R-02 as a reference for acceptance criteria for specific inspection and testing results. The overall acceptance criterion for identified degradation to concrete and

Attachment L-1 1-252 Page 20 of 33 reinforcing steel is that the affected structures in containment exposed to refueling canal leakage will continue to perform their intended functions during the period of extended operation.

FENOC plans to conduct additional inspections, if warranted, based on the inspection or testing results of the interaction between refueling canal leakage and concrete. FENOC plans to discontinue testing when no indications of refueling canal leakage are present. If refueling canal leakage resumes after it has been stopped, FENOC plans to resume the described evaluation of concrete and reinforcing steel.

2. The recommendations from the 2003 "Engineering Assessment Report -

Refueling Canal Leakage" have not all been completed. FENOC considered only one of the recommendations to be unnecessary. Each recommendation is discussed below:

a. "Complete the Refuel Canal non-destructive inspections and examinations to identify the remaining potential leakage paths on the liner plate." This recommendation has not been completed. Although attempts have been made to identify the remaining potential leakage paths on the liner plate, the leakage path identification has not been successful. Industry operating experience has shown that non-destructive examination methods such as vacuum box and liquid penetrant examinations are marginally effective. Therefore, resources were spent on coating the leaks at welds previously identified as suspect in the 2003 "Engineering Assessment Report - Refueling Canal Leakage". However, these attempts to eliminate the leakage have been unsuccessful to date. License Renewal Commitment 33 addresses the continuation of the implementation of this recommendation.

b. "Repair the Refuel Canal liner plate at all areas of leakage. This repair should occur within the next interval of flooding the canal with water, or within approximately 2 years." This recommendation has not been implemented because all areas of leakage have not been indentified, as noted in Response 2.a. License Renewal Commitment 33 addresses the continuation of the implementation of this recommendation.

c. "Repair and seal the concrete cracks and cold joints that have shown signs of leakage or are currently showing signs of leakage. This repair will prevent possible ongoing corrosion of the rebar and concrete degradation near the crack or joint. It is recommended that this repair be performed after the liner leaks are sealed to prevent leakage from occurring in different locations." This recommendation has not been implemented because the liner leaks have not been sealed, as noted in Response 2.b. License Renewal Commitment 33 addresses the planned implementation of this recommendation.

Attachment L-1 1-252 Page 21 of 33

d. "If measures are not taken to identify and repair the liner leaks and to seal the concrete cracks and cold joints showing leakage within approximately two years or before the next interval of flooding the canal with water, then perform full visual and non-destructive examinations, including electric half cell and UPV on the concrete structure, to reverify the condition of the structure." This recommendation has not been implemented. In lieu of full visual and non-destructive examination, the accessible areas of leakage exposure continue to be visually examined and evaluated. License renewal future Commitment 33 addresses the plan to implement an alternate to this recommendation.

e. "Update Davis-Besse Calculation VC05/B01-02 to reflect the evaluation indicated in Appendix K with respect to the excessive rebar cover (reduced effective depth "d")." The calculation has been updated.

f. "Evaluate the diagonal crack discovered on the west wall in the Incore Instrumentation Tunnel." The engineering evaluation of the diagonal crack in the Incore Instrumentation Tunnel has been completed and found to be acceptable.

3. FENOC plans to select and utilize a leak detection method to pin point the leakage area in the refueling canal by the end of 2014. FENOC also plans to evaluate temporary and permanent repair methods to stop or significantly reduce the refueling canal leakage. FENOC currently plans to implement the selected repair plan by the end of 2016.

If the leakage cannot be stopped or if the leakage resumes, the actions that are planned to occur are described in the response to RAI Item 1.

The proposed timeframe is acceptable based on plant-specific operating experience with the effects of borated water leakage on concrete and reinforcing steel. The plant-specific operating experience included non-destructive testing and petrographic examination of core drills from the east/west tunnel in 2003 that found no significant degradation of the concrete or reinforcing steel. In addition, information about related research and industry operating experience exists that documents the relatively minor effect that borated water leakage, through concrete, has on concrete and carbon steel reinforcing steel. For example, such information was discussed in the Salem License Renewal Safety Evaluation Report (SER) (ML110900295), Section 3.0.3.2.15, Indian Point License Renewal SER (ML093170671), Section 3.03.2.15-2, and the Prairie Island License Renewal SER (ML092890209),

Section 3.0.3.2.17. As noted in Response 1, FENOC commits to a plan to confirm that the integrity of affected structures is adequate prior to the period of extended operation. Also as noted in Response 1, FENOC commits to a plan to repeat such confirmation until there are no indications of continued or renewed refueling canal leakage.

Attachment L-11-252 Page 22 of 33

4. Components (pipe supports, conduit supports, piping etc.) that are exposed to refueling canal borated water leakage and exhibit signs of corrosion, as identified in condition reports during the 2010 refueling outage, were remediated in accordance with the FENOC Boric Acid Corrosion Control (BACC) procedures. FENOC procedure NOP-ER-2001, "Boric Acid Corrosion Control Program" provides detailed requirements for conducting and documenting the inspections and for evaluating the results of the inspections.

Procedure NOP-ER-2001 includes requirements for removing boric acid deposits and performing an "as-left" inspection. Inspection Procedure EN-DP-01 501, "Boric Acid Corrosion Control," requires BACC recurring inspections of the specific areas where the borated water leakage was found.

The procedure also requires that the inspection findings be recorded on a BACC screening form. Evaluation of the inspection results includes verification that there will be no significant corrosion of components before the next inspection is conducted.

In the Incore Instrumentation area, the incore tube supports, unistruts and electrical conduit that were exposed to refueling canal borated water were evaluated as carbon steel. In the East/West Tunnel area, pipe supports and an electrical box that were exposed to refueling canal borated water were evaluated as carbon steel. Based on Revision 1 of the EPRI Boric Acid Corrosion Guidebook, the incore tube supports, unistruts, electrical conduit, pipe supports and electrical box were evaluated as having no more than 0.002 inches of material loss over the next fuel cycle. That amount of material loss due to corrosion before the next inspection was considered to be negligible. Therefore, the components that had exhibited signs of corrosion were accepted as is.

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

Section B.2.34 Question RAI B.2.34-2

Background:

In its response to RAI B.2.34-1, the applicant stated that according to the certificate of material test report (CMTR) for the reactor head closure studs, the actual measured yield strength varied from 151 to 159 ksi, and the tensile strength varied from 166 to 171 ksi. The applicant also stated that its reactor head stud material is SA-540, Grade B-23 and that as provided in Regulatory Guide (RG) 1.65, "Materials and Inspections for Reactor Vessel Closure Studs," this material when tempered to a maximum tensile strength of 170 ksi, is relatively

Attachment L-1 1-252 Page 23 of 33 immune to stress corrosion cracking (SCC). The applicant proposes to enhance 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.

The "preventive actions" program element of GALL AMP XI.M3, "Reactor Head Closure Stud Bolting," references the guidance in RG 1.65 and NUREG-1339, "Resolution of Generic Safety Issue 29: Bolting Degradation or Failure in Nuclear Power Plants."

Issue:

LRA Section B.2.34 states that the Reactor Head Closure Program is an existing program that, with enhancements, will be consistent with GALL AMP XI.M3. All of the applicant's reactor head closure studs were fabricated from material with measured yield strength above 150 ksi and some of the furnished materials have a measured tensile strength above 170 ksi. The staff noted that this is an exception to the "preventive actions" program element of GALL AMP XI.M3, which recommends using bolting material for closure studs with actual measured yield strength less than 150 ksi to reduce susceptibility to SCC.

Request:

1) Revise the appropriate sections of the LRA to reflect the use of reactor head closure studs with measured yield strength above 150 ksi as an exception to GALL AMP XI.M3.

2) Address the exception to the "preventive actions" element for using closure stud material with greater susceptibility to SCC. Justify the adequacy of the Reactor Head Closure Program to manage cracking due to SCC of high-strength bolting material. As part of the justification, describe how the program manages the potential exposure of closure bolting to borated water and other potential contaminants that may initiate SCC of the reactor head closure bolting studs and components.

RESPONSE RAI B.2.34-2 The Reactor Head Closure Studs Program is revised to include an exception to the "preventive actions" element as follows.

NUREG-1801 Section XI.M3 recommends use of bolting material for closure studs that have an actual measured yield strength of less than 150 kilo-pounds per square inch (ksi). However, the Davis-Besse reactor head closure studs have an actual measured yield strength of greater than 150 ksi. Justification for the adequacy of the Reactor Head Closure Program to manage cracking due to

Attachment L-1 1-252 Page 24 of 33 stress corrosion cracking of high-strength bolting material (i.e., yield strength of greater than 150 ksi) is as follows.

The Reactor Head Closure Studs Program inspections are implemented by the Inservice Inspection (ISI) Program. The ISI 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 Xl, Subsection IWB (1995 Edition through the 1996 Addenda) and approved ASME Code Cases. The extent and frequency of these examinations provide for timely detection of cracks. To manage cracking, each reactor head closure stud is volumetrically examined once per each 10-year Inservice Inspection Interval.

In addition, Davis-Besse has not experienced cracking of the reactor head closure studs.

When the reactor head closure studs are removed from the reactor vessel flange during refueling outages, the studs, nuts and washers are stored in protective racks after removal and the reactor vessel flange holes are plugged with watertight plugs during cavity flooding. These methods assure the holes, studs, nuts, and washers are protected from borated water and other potential contaminants during cavity flooding. In addition, the visible portions of the studs are inspected for boric acid corrosion prior to removal.

The appropriate sections of the LRA are revised to reflect the above exception to NUREG-1801,Section XI.M3.

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

Section 3.1.2 Question RAI 3.1.2.2-3 (from NRC letter dated August 11, 2011 (ML11216A236))

Background:

In Request 3 of RAI 3.1.2.2-2 issued by letter dated June 20, 2011) the staff requested that the applicant describe the functional groups for the following two components that are addressed in LRA Table 3.1.2-2: (1) core support assembly (CSA) vent valve body, and (2) plenum cylinder reinforcing plate. The staff also requested that if existent, the applicant describe their link relationships (such as primary/expansion link) with other components. In addition, the applicant was requested to describe the inspection method, including the inspection frequency, for the components and the technical basis for the applicant's aging management methods.

Attachment L-1 1-252 Page 25 of 33 In its response dated July 22, 2011, the applicant stated that in Topical Report MRP-227, "Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines," the reactor internals were assigned to one of the following four functional groups: Primary, Expansion, Existing Programs, and No Additional Measures components. The applicant also stated that the link relationships are consistent with that provided in Tables 4-1 and 4-4 of MRP-227, Rev. 0. The applicant further stated that 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. In comparison, the revised LRA Table 3.1.2-2 in response to RAI 3.1.2.2-2 does not include an AMR item to manage loss of fracture toughness of the cast austenitic stainless steel (CASS) CSA vent valve body and plenum cylinder reinforcing plate.

In its review, the staff noted that GALL Report, Rev. 2, item IV.B4.RP-382 recommends GALL AMP XI.M1, "ASME [American Society of Mechanical Engineers Boiler and Pressure Vessel Code] Section Xl Inservice Inspection, Subsections IWB, IWC, and IWD," to manage cracking or loss of material due to wear of core support structure components; however, the LRA does not address this item. The staff also noted that Section 5.4.4 of the applicant's Technical Specifications requires that it should be verified by visual inspection every 24 months that the vent valve body exhibits no abnormal degradation. In addition, the staff noted that Section 3.2.3, Table 3-2 and Section 4 of Topical Report BAW-2248A, "Demonstration of the Management of Aging Effects for the Reactor Vessel Internals," indicate that reduction of fracture toughness due to thermal aging embrittlement is applicable to reactor vessel internal vent valve bodies.

In its review, the staff also noted that the revised LRA Table 3.1.2-2 submitted by letter dated July 22, 2011, does not address the following GALL Report Rev. 2 items:

(1) items IV.B4.RP-236 and IV.B4.RP-237 for the components with no additional measures and (2) items IV.B4.RP-238 and IV.B4.RP-239 for the inaccessible locations of the reactor vessel internals.

Issue:

In its response to RAI 3.1.2.2-2, the applicant indicated that the applicant's aging management methods for the plenum cylinder reinforcing plate and vent valve body are described in MRP-227 Tables 4-1 and 4-4. However, the staff noted that MRP-227 Tables 4-1 and 4-4 referenced in the applicant's response do not clearly address information regarding: (1) the functional groups, (2) the link relationships, or (3) the inspection method, including the frequency, specified for the CSA vent valve body and plenum cylinder reinforcing plate. In addition, the

Attachment L-1 1-252 Page 26 of 33 revised LRA Table 3.1.2-2 in response to RAI 3.1.2.2-2 does not address an AMR line item to manage loss of fracture toughness of these CASS components.

In its review, the staff also found a need to clarify the following items: (1) why LRA Table 3.1.2-2 does not address GALL Report, Rev. 2, items IV.B4. RP-382, IV.B4.RP-236, IV.B4.RP-237, IV.B4.RP-238 and IV.B4.RP-239, (2) whether or not GALL Report, Rev. 2, item IV.B4.RP-382 is applicable to the plenum cylinder reinforcing plate and vent valve body, and (3) why LRA Table 3.1.2-2 does not address an AMR item for aging management of loss of fracture toughness of the vent valve body even though applicant's Technical Specifications require visual inspections of the component to ensure no abnormal degradation and Topical Report BAW-2248A indicates that reduction of fracture toughness due to thermal aging embrittlement is applicable to reactor vessel internal vent valve bodies.

Request:

1. Provide the justification as to why LRA Table 3.1.2-2 does not address the following GALL Report items for the components with no additional measures and inaccessible areas: GALL Report items IV.B4.RP-236, IV.B4.RP-237, IV.B4.RP-238 and IV.B4.RP-239 In addition, describe the applicant's operating experience to clarify whether or not the accessible areas of the applicant's components have indicated aging effects that need management.

2. Provide the justification as to why LRA Table 3.1.2-2 does not address GALL Report, Rev. 2, item IV.B4.RP-382 that recommends GALL AMP XI.M1, "ASME Section Xl Inservice Inspection, Subsections IWB, IWC, and IWD," to manage cracking or loss of material of core support structure. In addition, clarify whether or not this item for the core support structure is applicable to the plenum cylinder reinforcing plate and vent valve body.

3. Provide the justification as to why LRA Table 3.1.2-2 does not address an AMR item to manage loss of fracture toughness of the CASS vent valve body even though applicant's Technical Specifications require visual inspections of the component to ensure no abnormal degradation and Topical Report BAW-2248A indicates that reduction of facture toughness is applicable to the internal valve bodies.

4. Provide the information regarding: (1) the functional groups, (2) the link relationships (if existent), and (3) the inspection method including the frequency used to manage loss of fracture toughness of the CSA vent valve body and plenum cylinder reinforcing plate. As part of the response, provide the technical basis to demonstrate that these aging management methods are adequate to manage loss of fracture toughness of the components.

Attachment L-1 1-252 Page 27 of 33 If the functional group of the components is Existing Programs or No Additional Measures group, provide the method and frequency of the existing inspections specified for the CASS components.

RESPONSE RAI 3.1.2.2-3 (from NRC letter dated August 11, 2011 (ML11216A236))

1. LRA Table 3.1.2-2, "Aging Management Review Results - Reactor Vessel Internals," is revised to include GALL Report items IV.B4.RP-236, IV.B4.RP-237, IV.B4.RP-238 and IV.B4.RP-239.

As part of the Inservice Inspection Program, a visual (VT-3) examination of the reactor vessel removable core support structure accessible surfaces is conducted once per Inservice Inspection interval in accordance with ASME Section XI, Table IWB-2500-1, Examination Category B-N-3. These inspections have not identified any unacceptable indications.

2. LRA Table 3.1.2-2 is revised to include GALL Report item IV.B4.RP-382. As discussed above, the ASME Section XI Category B-N-3 examination is applicable to the accessible surfaces of the reactor vessel internals. The accessible surfaces of the plenum cylinder reinforcing plate and vent valve body are included in this inspection.

3. As part of the development of the MRP-227 program, the CASS reactor vessel internal components were initially screened based on casting method, percent ferrite and molybdenum content to determine if the components were susceptible to thermal embrittlement. EPRI Report 1018292, "Materials Reliability Program: Screening, Categorization, and Ranking of B&W-Designed PWR Internals Component Items (MRP-189-Rev. 1)," provides the results of this screening. As documented in MRP-1 89, the vent valve body was deemed not to be susceptible to any of the eight age-related degradation mechanisms (includes loss of fracture toughness due to thermal embrittlement) and therefore, was categorized as "A" (i.e., below the screening criteria of the age-related degradation mechanisms considered by the MRP-227 Program).

In addition, Technical Specification 5.5.4 requires testing of the vent valve to verify by visual inspection that the valve body and valve disc exhibit no abnormal degradation, verify the valve is not stuck in an open position, and verify by manual actuation that the valve is fully open when a force of < 400 lbs is applied vertically upward. This testing and inspection is not credited by MRP-227 for managing loss of fracture toughness of the vent valve body.

Also, BAW-2248A, "Demonstration of the Management of Aging Effects for the Reactor Vessel Internals," dated March 2000, pre-dates the MRP-227 program development and provides that for CASS components, thermal embrittlement is considered a potential aging effect. BAW-2248A only identified potential aging

Attachment L-1 1-252 Page 28 of 33 effects and did not perform screening based on casting method, percent ferrite, and molybdenum content to determine if the components were susceptible to thermal embrittlement. As provided above, screening was provided under the MRP-227 program and the vent valve body was deemed not to be susceptible to loss of fracture toughness due to thermal embrittlement.

4. As documented in MRP-189, the vent valve body was deemed not to be susceptible to any of the eight age-related degradation mechanisms (includes loss of fracture toughness due to thermal embrittlement) and therefore, was categorized as "A" and placed into the "No Additional Measures" inspection category.

As documented in the MRP response to NRC RAI 4-15 (ML103160381), the plenum cylinder reinforcement castings (Davis-Besse only) were not included in any MRP or PWROG evaluations. In the response it was stated that, "It is possible that this item could be dispositioned by reviewing the materials records and determining the ferrite content to be below the MRP-1 75 screening criteria, which would also classify it as 'No Additional Measures.'

AREVA NP conducted a records search for selected B&W internals castings.

As related to the Davis-Besse plenum cylinder reinforcement castings, the report provided the following information:

"The plenum cylinder reinforcement castings at DB-1 were unknown, therefore were not evaluated by MRP-189 or MRP-227 (Rev. 0). The two plenum cylinder reinforcement castings are near the two CSS outlet nozzles. The plenum cylinder LOCA boss or LOCA bars are an integralpart of the reinforcement castings. This casting item is unique to the DB-1 and was uncovered during the present records search. The plenum cylinder reinforcement for the other six B&W units is of a different design and does not use casting.

The heat chemical composition has been found for the two DB-1 plenum cylinder reinforcement castings. The ferrite content is below the 20% screening value for thermal aging embrittlement of CF-8 casting. Since the DB- I plenum cylinder reinforcement castings would have been placed in the Category "A" items during the screeningprocess in MRP-189 and in MRP-227 (Rev. 0), no additionalinspection requirements or evaluations would be needed."

As provided in MRP-227, no inspections, except for those specified in ASME Code, Section Xl, are required for the "No Additional Measures" components.

As part of the Davis-Besse Inservice Inspection Program, a visual VT-3 examination of the reactor vessel removable core support structure is conducted once per Inservice Inspection interval in accordance with ASME Section Xl, Table IWB-2500-1, Examination Category B-N-3. Included in the

Attachment L-1 1-252 Page 29 of 33 VT-3 examination are the accessible surfaces of the vent valve bodies and the plenum cylinder reinforcement castings.

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

Supplemental RAIs Supplemental Question RAI 3.2.2.2.3.6-2 The NRC initiated a telephone conference call with FENOC on August 22, 2011 to discuss FENOC's 8/17/2011 response letter.

The teleconference began with the NRC staff stating the following concern about the FENOC's response to RAI 3.2.2.2.3.6-2.

LRA Table 3.2.2-2, row 20 stainless steel piping components exposed to moist air are being managed for cracking by the One-Time Inspection Program. The AMR items cite generic note H. The AMR item also cites plant-specific note 0202, which states that the One-Time Inspection is being used to confirm the absence of aging effects or that aging is slow acting so as to not affect the subject component's intended function during the period of extended operation.

By letter dated May 2, 2011, the staff issued RAI 3.2.2.1.26-1 requesting that the applicant justify its use of the One-Time Inspection Program for managing these aging effects. In its response of June 3 the applicant stated that this item was deleted. However, the staff noted there was no evidence of this in the applicant letter dated May 24, 2011.

On this item there is no resolution if a proper periodic program will be used above the air water interface to manage this aging. RAI 3.2.2.2.3.6-2, was issued requesting resolution of this component in the Containment Spray System.

In its response dated August 13, 2011, the applicant did not state for Table 3.2.2-2, row 20, stainless steel piping in moist air (internal) exposed to cracking Containment Spray System, whether that item was a) retained in the One-Time Inspection program, b) documented for deletion or c)updated with an aging management program. The staff finds the applicant's response not acceptable because the resolution concerning this line item is incomplete.

Attachment L-1 1-252 Page 30 of 33 SUPPLEMENTAL RESPONSE RAI 3.2.2.2.3.6-2 LRA Table 3.2.2-2, "Aging Management Review Results - Containment Spray System,"

row 20, is not deleted and should not be deleted. Cracking of the subject stainless steel piping components exposed to moist air will be managed by the Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Program.

An extent of condition review determined the same material, environment, and aging effect combination is also found in LRA Table 3.3.2-4, "Aging Management Review Results - Boron Recovery System," row 159, and LRA Table 3.3.2-5, "Aging Management Review Results - Chemical Addition System," row 60. The LRA is revised to credit the Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Program as the aging management program for these rows.

In FENOC Letter L-11-238 dated August 17, 2011, the response to RAI 3.2.2.2.3.6-2 identified changes to the LRA to credit the Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Program for managing loss of material of components in a moist air environment. As part of the extent of condition review, additional LRA changes are made to credit the Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Program in lieu of a One-Time Inspection for managing the loss of material of components in a moist air environment.

Furthermore, the LRA is revised to define the moist air (internal) environment to encompass both the air-water interface and the air environment above the interface. In conclusion, the Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Program manages loss of material (except for selective leaching) and cracking for all in scope components subject to a moist air environment.

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

Supplemental Question RAI 3.3.2.2.5-2 The NRC initiated a telephone conference call with FENOC on August 29, 2011, to discuss the FENOC response to RAI 3.3.2.2.5-2 submitted in FENOC letter dated August 17, 2011 (ML11231A966), regarding aging management of elastomers.

NRC requested that the "Collection, Drainage and Treatment Components Inspection Program," "Scope of Program" program element should be updated to include the information that was previously added to the Program Description, and the "Acceptance Criteria" program element should be revised to include acceptance criteria for elastomeric components added to scope.

Attachment L-1 1-252 Page 31 of 33 SUPPLEMENTAL RESPONSE RAI 3.3.2.2.5-2 LRA Section B.2.9, "Collection, Drainage and Treatment Components Inspection Program," aging management program elements "Scope of Program" and "Acceptance Criteria," are revised to include the appropriate information as requested during the telephone conference.

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

Supplemental Question Makeup Pump Casing Inspections The NRC initiated a telephone conference call with FENOC on September 7, 2011 to discuss the FENOC response to Supplemental RAI Makeup Pump Casing Inspections submitted in FENOC letter dated August 26, 2011 (ML11242A166).

FENOC plans to manage cracking due to cyclic loading of the stainless steel high-pressure makeup pumps (DB-P37-1 and 2) in the Makeup and Purification System using enhanced visual (VT-1 or equivalent) and/or volumetric (RT or UT) inspections. However, the designation for enhanced visual exam is EVT-1.

FENOC used the term EVT-1 correctly in the table on p 35 of the enclosure when revising the "detection of aging effects" element of the AMP, but used the wrong designation in the commitment and "scope" program element. NRC stated that it is an important distinction to make, because a VT-1 exam may not be adequate to detect cracking.

SUPPLEMENTAL RESPONSE Makeup Pump Casing Inspections FENOC agrees that the designation for enhanced visual exam is EVT-1. LRA Table A-I, "Davis-Besse License Renewal Commitments," license renewal future Commitment 13, and LRA Section B.2.30, "One-Time Inspection," are revised to include the following:

"... enhanced visual (EVT-1 or equivalent) or surface examination (magnetic particle, liquid penetrant), or volumetric (RT or UT) inspections..."

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

Supplemental Question - Abandoned Equipment The NRC initiated a telephone conference call with FENOC on September 7, 2011, to discuss the FENOC supplemental response to RAI 2.1-3 (submitted in FENOC

Attachment L-1 1-252 Page 32 of 33 letter dated August 17, 2011 (ML11231A966)) regarding abandoned equipment.

The NRC staff requested discussion on the FENOC implementation schedule regarding actions to address abandoned equipment. Following discussion of the FENOC plan, NRC staff agreed that the proposed plan was acceptable and requested that the plan be submitted in the next RAI response letter. Based on the proposed schedule for the plans to address abandoned equipment, it was mutually agreed that license renewal future Commitment 26 on this topic was no longer necessary and would be deleted; NRC plans to treat this issue as an Open Item in the License Renewal Safety Evaluation Report for Davis-Besse.

SUPPLEMENTAL RESPONSE - Abandoned Equipment LRA Table A-1, "Davis-Besse License Renewal Commitments," is revised to delete Commitment 26 regarding abandoned equipment.

FENOC plans to perform the following actions by February 15, 2012, to ensure abandoned equipment is identified, isolated and drained:

1. Determine the scope of abandoned equipment - includes review of Piping &

Instrumentation Diagrams (P&IDs), plant walkdowns, and review of the Shift Operations Management System (eSOMS) clearance database.

2. Determine the status of abandoned equipment - includes review of system status files and the eSOMS database for as-left valve positions, walkdowns to validate valve position status, and ultrasonic testing to confirm that abandoned piping is drained.

3. Place abandoned equipment in a configuration that will not impact safety-related equipment - create and implement Operations Evolution Orders to isolate and drain abandoned systems with fluids, and create and implement Document Change Requests as necessary to correct the configuration of the plant as shown on plant drawings.

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

Supplemental Question - LRA Table 3.3.2-14 The NRC initiated a telephone conference call with FENOC on September 7, 2011, to discuss the replacement of LRA Table 3.3.2-14, "Aging Management Review Results - Fire Protection System," in FENOC letter dated August 26, 2011 (ML11242A166). The NRC staff noted that, following a line-by-line comparison to LRA Table 3.3.2-14, the fire pump diesel engine rows appeared to be missing in the replacement table.

Attachment L-1 1-252 Page 33 of 33 SUPPLEMENTAL RESPONSE - LRA Table 3.3.2-14 FENOC agrees that Table 3.3.2-14 requires revision to include the missing fire pump diesel engine rows that were inadvertently not included in the previous response. LRA Table 3.3.2-14, "Aging Management Review Results - Fire Protection System,"

submitted in FENOC letter dated August 26, 2011 (ML1I1242A1 66), is replaced in its entirety to include the Fire Protection System components and the fire pump diesel engine and associated components.

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

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

Letter L-1 1-252 Amendment No. 15 to the DBNPS License Renewal Application Page 1 of 108 License Renewal Application Sections Affected Section 3.1 Table 3.3.2-26 Table 3.1.1 Table 3.3.2-27 Table 3.1.2-1 Table 3.3.2-30 Table 3.1.2-2 Table 3.3.2 PS Notes Table 3.1.2 PS Notes Section 3.4 Section 3.2 Table 3.4.2-2 Table 3.2.1 Table 3.4.2 PS Notes Table 3.2.2-2 Table 3.2.2-4 Appendix A Table 3.2.2 PS Notes Section A. 1.30 Section A. 1.32 Section 3.3 Table A-1 Section 3.3.2.1.3 Table 3.3.2-2 Appendix B Table 3.3.2-3 Table B-1 Table 3.3.2-4 Table B-2 Table 3.3.2-5 Section B.2.9 Table 3.3.2-7 Section B.2.30 Table 3.3.2-11 Section B.2.32 Table 3.3.2-12 Section B.2.34 Table 3.3.2-14 The Enclosure identifies the change to the License Renewal Application (LRA) by Affected LRA Section, LRA Page No., and Affected Paragraph and Sentence. The count for the affected paragraph, sentence, bullet, etc. starts at the beginning of the affected Section or at the top of the affected page, as appropriate. Below each section the reason for the change is identified, and the sentence affected is printed in italics with deleted text !e,-d out and added text underlined.

Enclosure L-11-252 Page 2 of 108 Affected LRA Section LRA Page No. Affected Paragraph and Sentence Table 3.1.1 Page 3.1-37 Row 3.1.1-71 "Discussion" column Text in "Discussion" column is revised based on the response to RAI B.2.34-2. 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 ," and now reads as follows:

Table 3.1.1 Summary of Aging Management Programs for Reactor Vessel, Internals, Reactor Coolant System and Reactor Coolant Pressure Boundary, and Steam Generators Evaluated in Chapter IVof NUREG-1801 Further Item Aging Effect/ Aging Management Euatiss Number ComponentlCommodity Mechanism Programs Recommended 3.1.1-71 High-strength low alloy steel Cracking due to Reactor Head Closure No Consistent with NUREG-1801L closure head stud assembly stress corrosion Studs with exceptions.

exposed to air with reactor cracking; loss of Cracking and loss of material for coolant leakage materal due to wear the reactor vessel head closure studs are managed by the Reactor Head Closure Studs Program.

Enclosure L-1 1-252 Page 3 of 108 Affected LRA Section LRA Page No. Affected Paragraph and Sentence Table 3.1.2-1 Page 3.1-45 Rows 9 and 10 Notes The Notes for rows 9 and 10 of LRA Table 3.1.2-1, "Aging Management Review Results - Reactor Pressure Vessel," are revised based on the response to RAI B.2.34-2, and Table 3.1.2-1 reads as follows:

Table 3.1.2-1 Aging Management Review Results - Reactor Pressure Vessel dAging AgnEffect fetAging Management NUREG-1801, Table 1 oe Row Component Intended Material Environment Requiring Agn a a e et 1 0 , T be1 Notes No. Type Function(s) Management Program Volume 2 Item Item Closure Air with 9Pressure borated water Reactor Head V.A-2 3.1.1-71 B-and Washers boundary Steel (External)

Closure Air with Pressure borated water Loss of Reactor Head IV.A2-3 3.1.1 10 Studs, Nuts, boundary Steel leakage material Closure Studs B and Washers (External)I

Enclosure L-11-252 Page 4 of 108 Affected LRA Section LRA Page No. Affected Paragraph and Sentence Table 3.1.2-2 Pages 3.1-60 Entire Table thru 3.1-121 Table 3.1.2 Plant- Page 3.1-187 3 New Notes Specific Notes In response to RAI 3.1.2.2-3 from NRC letter dated August 11,2011 (ML11216A236), LRA Table 3.1.2-2, "Aging Management Review Results - Reactor Vessel Internals," is replaced in its entirety, and new Plant-Specific Notes are added to the table. LRA Table 3.1.2-2 and Table 3.1.2 Plant-Specific Notes read as follows:

Table 3.1.2-2 Aging Management Review Results - Reactor Vessel Internals Row Intended Aging Effect Aging NUREG-1v801 Table No. Component Type Function(s) Material Environment Requiring Management Rev. 1 Item te Notes Management Program (Rev. 2 Item) Item Plenum Cover Asseml Plenum Cover Rib Pads Stainless Borated Reactor IV__B4-1 Loss of PWR Reactor IV.84 15

- (primarycomponent Suppo Steel Coolant with material- wear Vessel Internals (IV.B4.R 3.1.1-63 A with no expansion Neutron Fluence 251) components)

Plenum Cover Support Flange Stainless Borated Reactor IV.B4-15 Loss of PWR Reactor IV.B4-15 (primarycomponent SUot Steel Coolant with materal - wear Vessel Internals (IV.B4.RP- 3.1.1-63 A with no expansion Neutron Fluence 251) components)

Alloy X-750 Dowels-to- PWR Reactor Plenum Cover BoratedReactor Vessel Internals None 3 Bottom Flange We Support Nickel Coolant with Cracking - SCC (IV.B4.RP- None A (no additional ____ Neutron Fluence PWR Water 236) measures component) Chemistry

Enclosure L-1 1-252 Page 5 of 108 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 m Notes Management Program (Rev. 2 Item) Item ControlRod Guide Tube (CRGT) Assembly CRGT Spacer Casting (expansion component Cast with Primarycomponent Austenitic Borated Reactor Reduction in PWR Reactor IV.B4-4 4 link of CSS Cast Outlet Suppot Stainless Coolant with fracture Vessel Internals 242)

Nozzles, CSS Vent Steel Neutron Fluence toughness 242)

Valve Discs or IMI Guide Tube Spiders)

CRGT Rod Guide Tubes 5 Stainless Borated Reactor Loss of

_______(IV.B4.RP-PWR Reactor None None A 5 (noadditional Suppor Steel Neutron Fluence material- wear Vessel Internals 237) measures component _

CRGT Rod Guide Sectors 6 Supp Stainless Steel Boratedwith Coolant Reactor Loss of - wear Colatwih(IV.B4.RP-material VesselReactor PWR Internals None 237).R- None on A_

(no additional Steel Neutron Fluence wr e r measures comoonent)

Core Suport Shield CSS Assembi CSS Top Flange (primary component Stainless Borated Reactor Loss of PWR Reactor IV.B4-15

- with no expansion Suppo Steel Nuolun c material- wear Vessel Internals 251) components) Neutron w53.

Fluence

Enclosure L-11-252 Page 6 of 108 Table 3.1.2-2 Aging Management Review Results - Reactor Vessel Internals Row Component Type No.

Upper Core Barrel (UCB) Bolts (original bolts) and their locking devices (primarycomponent 8 with expansion components of UTS Bolts and theirlocking devices, LTS Bolts and their locking devices, and SSHT Bolts and I their locking devices)

Enclosure L-11-252 Page 7 of 108 Table 3.1.2-2 Aging Management Review Results - Reactor Vessel Internals Row Intended Aging Effect Aging NUREG-1801 Table 1 Function(s) Material Environment Requiring Management Rev. 1 Item Notes No. Component Type Management Program (Rev. 2 Item) Item PWR Reactor Upper Core Barrel (ICB) Bolts Bolt: Cracking -Vessel Internals IV.B4-20 A (repacee bolts) aoltd Cracn -(IV.B4.RP- 3.1.1-37 0.

(replacementbolts) and Bolt: SCC PWR Water 248) 0114 their locking devices Nickel Chemistry (primarycomponent Alloy Borated Reactor Bolt:

9 with expansion Suppo Coolant with Cumulative TLAA IV.B4-37 3.1.1-05 A components of UTS Locking Neutron Fluence fatigue damage (IV.B4.R-53)

Bolts and theirlocking Devices: - fatigue devices, LTS Bolts and Stainless Locking their locking devices, Steel Device: PWR Reactor IV.B4-01 and SSHT Bolts and Vessel Internals (IV.B4.RP- 3.1.1-22 C theirlocking devices) Loss of 243)

'______material - wear CSS Cast Outlet Nozzles Cast BoratedReactor Reduction in IV.B4-21 10 (primarycomponent Austenitic Coolant with fracture V Reactos Intrnals 253 with expansion Su__o__Stainless Neutron Fluence to3hness) components as follows: Steel CRGT Spacer Casting)

CSS Vent Valve Top Retaining Ring and Bottom Retaining Ring Stainless Borated Reactor Reduction in PWR Reactor IV.B4-16 11o Coolant with fracture (IV.B4.RP- 3.1.1-22 A (primarycomponent Steel Neutron Fluence toughness Vessel Internals 252) with no expansion components)

Enclosure L-11-252 Page 8 of 108 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 CSS Vent Valve Discs (primary component SUPO Cast Borated Reactor Reduction in PWR Reactor IV.B4-21 12 with expansion Flow Austenitic Coolant with fracture Vessel Internals (IV.B4.RP- 3.1.1-80 A component of CRGT -ow Stainless Neutron Fluence toughness 253)

_Spacer Casting) Control Steel CSS Vent Valve Disc IV.B4-16 Shaft (IV.B4.RP-Stainless Borated Reactor Reduction in PWR Reactor (IV.B4.RP 13 (primary component Suppot Steel Coolant with fracture Vessel Intemals 252); 3.1.1-22 A with no expansion Neutron Fluence toughness (IV.B4.RP-com ponents) 239)

Core BarrelAssembly Core Barrel Cylinder (including vertical and IV.B4-12 circumferentialseam Borated Reactor Reduction in (IV.B4.RP-14 welds)1 wed UDot SUDDO Steel Stainless Coolant with fracture VeselInernls______11-2_

PWR Reactor 250 3.1.1-22 A (expansion component Neutron Fluence toughness Vessel Internals (IV.B4.RP-with primary component 239) link of Baffle Plates)

PWR Reactor Alloy X-750 Vessel Internals None Core Barrel-to-Former Borated Reactor Cracking - SCC (IV.B4.RP- None A 15 Plate Dowel Suppo Nickel Coolant with PWR Water 236) 5- Alloy Coln ihChemistry___

(no additional Neutron Fluence Reduction in None measures component) fracture PWR Reactor (IV.B4.RP- None A

"_ _Vessel toughness Internals 2 27

Enclosure L-1 1-252 Page 9 of 108 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)

Alloy X-750 PWR Reactor Dowel-to-Core Barrel Borated Reactor Vessel Internals None 16 Cylinder Fillet Welds Suppo Alloy N Coolant with Cracking - SCC (IV.B4.RP- None A (no additional oy Neutron Fluence PWR Water 236) measures component) Chemistry Thermal Shield Upper Cracking-Restraint Cap Stainles Borated Reactor PWR R t None 17 Screws (Not Exposed) Support Steel Coolant with LOSS Of e aco (IV.B4.RP-237) None A (no additiona Neutron Fluence material- wear Vessel Internals measures component) Loss of preload Baffle Plates (primarycomponent with expansion components of Core Stainless Borated Reactor Reduction in PWR Reactor IV.B4-12 18 Barrel Cylinder, Suppo Steel Coolant with fracture Vessel Intemals (IV.B4.RP- 3.1.1-22 A including vertical and Neutron Fluence toughness 249) circumferentialseam welds, and Former Plates)

FormerPlates IV.B4-12 19 (expansion componen (epnincmoetSanes Stainles Coolant Reactor Boratedwith Reduction in fracture PWR Reactor (IV.B4.RP-20;3112 19 with primary component Support Steel Vessel Internals 25_ 3.1.1-22 A link of Baffle Plates) Neutron Fluence toughness (IV.B4.RP-239)1

Enclosure L-1 1-252 Page 10 of 108 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. I Item Item No.Funtio~s)Management Program (Rev. 2 Item) Ie PWR Reactor IV.B4-07 Cracking - Vessel Internals (IV.B4.RP-Core Barrel-to-Former 244); 3.1.1-30 A (CBF) Bolts Borated ReactorIASCC PWR Water (IV.B4.RP-2.0 (expansion component Suppor Stainless Steel Coolant with Coln ihIV.B4-01 Chemistry 238) with primary component Neutron Fluence Reduction in (IV.B4.RP-link of FB Bolts) PWR Reactor fracture Vessel Internals 243): 3.1.1-22 A toughness (IV.B4.RP-239)

PWR Reactor Baffle-to-Former (FB Cracking - Vessel Internals IV.B4-07 Bolets (C (IV.B4.RP- 3.1.1-30 A Bolts IASCC PWR Water 241)

(primary21component _____Coolant Stainless Boratedwith Reactor Chemistry with expansion SuIvort Steel Neutron FluenceR components Bolts and CBF of Bolts)

BB Reduction fracture in PWR___Reactor_ IV.B4-01 Vessel Internals (IV.B4.RP- 3.1.1-22 A touqhness 240)

PWR Reactor Vessel Internals None Baffle-to-Baffle (BB) Cracking fatinq (IVB4.RP- None A Bolts - internal Borated Reactor PWR Water 375)

Stainless BrtdRaorChemistry (expansion component Suor Steel Coolant with 22 with primary component Neutron Fluence Reduction in IV.B4-01 link of FB Bolts) rection PWR Reactor fracture (IV.B4.RP- 3.1.1-22 A toughness Vessel Internals 243)

Enclosure L-1 1-252 Page 11 of 108 Table 3.1.2-2 Aging Management Review Results - Reactor Vessel Internals Row Intended Aging Effect Aging NUREG-1801 Table I 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 ratinge (IV.B4.RP- None A aPWR Water 375)

Chemistry Baffle-to-Baffle (BB) PWR Reactor IV.B4-07 Bolts - external tVesselCracking -

Borated Reactor Internals VeslItras (IV.B4.RP

(.BRP Stainless '244): 3.1.1-30 A (expansion component S Steel Coolant with IASCC PWR Water (IV.B4.RP-23 with primary component Neutron Fluence link of FB Bolts) Chemistry 238)

Reduction in IV.B4-01 fracture toughness PWR VesselReactor Internals (IV.B4.RP-243): 3.1.1-22 A (IV.B4.RP-239)

Accessible Locking PWR Reactor Device and Locking Cracking - Vessel Internals IV.B4-07 Weld (FB Bolts and IASCC (IV.B4.RP- 3.1.1-30 A InternalBB Bolts) PWR Water 241)

(primary component Borated Reactor ChemistryStines 24 with expansion Suppo Staes Coolant with components of Neutron Fluence Reduction in IV.B4-01 Inaccessible Lockinq DeieadLcigfracture PWR ReactorReductioni VseInmas (IV.B4.RP- 3.1.1-22 A Device and Locking toughness Vessel Internals 240)

Weld (CBF Bolts and External BB Bolts))

Enclosure L-1 1-252 Page 12 of 108 Table 3.1.2-2 Aging Management Review Results - Reactor Vessel Internals Row Intended Aging Effect Aging NUREG-1801 Table N Function(s) Material Environment Requiring Management Rev. 1 Item Notes No. Component Type Item Management Program (Rev. 2 Item)

Inaccessible Locking PWR Reactor IV.B4-07 Device and Locking Cracking - Vessel Internals (IV.B4.RP-Weld (CBF Bolts and IASCC 244); 3.1.1-30 A External BB Bolts) PWR Water (IV.B4.RP-25 (expansion component Stainless Borated Coolant Reactor with Chemistry 238) 2 with primarycomponent Support Steel Nuolun c IV.B4-01 link of Accessible Reduction in PWR Reactor (IV.B4. RP-Locking Device and fracture Vessel Interna 243) 3.1.1-22 A Locking Weld (FB Bolts toughness (IV.B4.RP-and Internal BB Bolts)) 239)

Lower Core Barrel PWR Reactor (LCB) Bolts (original) Bolt: Cracking - Vessel Internals IV.B4-13 A and their locking SCC (IV.B4.RP- 3.1.1-37 0114 devices PWR Water 247)

____________Chemistry (primarycomponent Stainless Borated Reactor 26 with expansion Suppo Steael Coolant with components of UTS Neutron Fluence Devic Bolts and their locking Device: PWR Reacto IV.B4-01 devices, LTS Bolts and Loss of Vessel Internals (IV.B4.RP- 3.1.1-22 C their locking devices, material- wear 243) and SSHT Bolts and their locking devices) I

Enclosure L-1 1-252 Page 13 of 108 Table 3.1.2-2 Aging Management Review Results - Reactor Vessel Internals Row Intended Aging Effect Aging NUREG-1801 Table 1 Function(s) Material Environment Requiring Management Rev. 2 Item Notes No. Component Type Management Program (Rev. 2 Item) Item PWR Reactor Lower Core Barrel (LCB) Bolts Bolt: Cracking - Vessel Internals IV.B4-13 A (LCBepBolacement) ad thCr g -(IV.B4.RP- 3.1.1-37 A (replacement) and their Bolt: SCC PWR Water 247) 0114 locking devices Nickel Chemist (primarycomponent Alloy Borated Reactor Bolt:

27 with expansion SuDpo Coolant with Cumulative TLAA IV.B4-37 3.1.1-05 A components of UTS Locki Neutron Fluence fatigue damage (IV.B4.R-53)

Bolts and theirlocking Devices: - fatigue devices, L TS Bolts and Stainless Lockinq their locking devices, Steel Device: PWR Reactor IV.B4-01 and SSHT Bolts and Vessel Internals (IV.B4.RP- 3.1.1-22 C their locking devices) Loss of 243) material- wear Upper Thermal Shield PWR Reactor (UTS) Bolts and their Bolt: Cracking - Vessel Internals IV.B4-13 locking devices SCC (IV.B4.RP- 3.1.1-37 A PWR Water 245)

(expansioncomponent Stainless Borated Reactor Chemistrv with Primarycomponent 28 link of UCB Bolts and U Steel Coolant with their locking devices, Neutron Fluence Locking LCB Bolts and their Device: PWR Reactor (IV.B4.RP- 3.1.1-22 C locking devices, and FD Loss of Vessel Internals 243)

Bolts and their locking material- wear devices)

Enclosure L-1 1-252 Page 14 of 108 Table 3.1.2-2 Aging Management Review Results - Reactor Vessel Internals Row Intended Aging Effect Aging NUREG-1801 Table 1 Function(s) Material Environment Requiring Management Rev. 2 Item Notes No. Component Type (Rev. 2 Item) Item Management Program Surveillance Specimen PWR Reactor Holder Tube (SSHT) Bolt: Cracking - Vessel Internals IV.B4-13 Bolts and theirlocking (IV.B4.RP- 3.1.1-37 A devices Bolt: SCC PWR Water 245)

Nickel Chemistry (expansion componen Alloy Borated Reactor Bolt:

29 with pmary component upport Coolant with Cumulative TLAA IV.B4-37 3.1.1-05 A wihPiaycmoetL ovck%

i ng Neutron Fluence fatigue damage (IV.B4.R-53) link of UCB Bolts and ______

link of UCB Bolts and Devices: - fatioue their locking devices, Stainless Locking LCB Bolts and their Steel Device: PWR Reactor IV.B4-01 locking devices, and FD (IV.B4.RP- 3.1.1-22 C Bolts and their locking Loss of Vessel Internals 243) devices) material- wear Upper GridAssembly Alloy X-750 Dowel-to- PWR Reactor Upper GBd Rib Section Borated Reactor Vessel Internals None 30 Bottom Flange Welds Support Nickel Coolantwith Cracking - SCC (IV.B4.RP- None A (no additional Alloy Neutron Fluence PWR Water 236) measures component) Chemistry

Enclosure L-1 1-252 Page 15 of 108 Table 3.1.2-2 Aging Management Review Results - Reactor Vessel Internals Row Intended Aging Effect Aging NUREG-1801 Table I No. Component Type Function(s) Material Environment Requiring Management Rev. 1 Item Item Notes Management Program (Rev. 2 Item) Item Lower GridAssembly Lower Fuel Assembly Support Pads:

Pad, Pad-to-Rib Section Weld, Cap Screw and associated Locking Weld, and Stainless Alloy X-750 Dowel Steel, Borated Reactor Reduction in PWR Reactor IV.B4-31 31 (expansion componen Suppo Nickel Coolant with fracture Vessel Internals (IV.B4.RP- 3.1.1-22 A with Primary component Alloy Neutron Fluence toughness 260) link of IMI Guide Tube (Dowel)

Spiders and IMI Guide Tube Spider-to-Lower Grid Rib Section Welds)

Lower Fuel Assembly PWR Reactor Support Pads: Vessel Internals IV.B4-32 Alloy X-750 Dowel Crackinq - SCC (IV.B4.RP- 3.1.1-37 A Locking Weld PWR Water 262 (expansion component Borated Reactor Chemistickel 22 with primary component Support Alloy Coolant with link of IMI Guide Tube Neutron Fluence Spiders and IMI Guide Reduction in PWR Reactor IV.B4-31 Tube Spider-to-Lower fracture (IV.B4.RP- 3.1.1-22 A Grid Rib Section toughness 260)

Welds)

Enclosure L-1 1-252 Page 16 of 108 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 Lower Grid Assembly:

Alloy X-750 Dowel-to-Lower FuelAssembly Support Pad Welds PWR Reactor Borated Vessel Internals IV.B4-32 33 (expansion component Su ort Nickel Coolant Reactor with Cracking - SCC (IV.B4.RP- 3.1.1-37 A with primary component - - Alloy eto lecePRWtr 22 link of Lower Grid Neutron Fluence PWR Water 262)

Assembly: Alloy X-750 Chemistry 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 34 with expansion Supo Ao Coolant with Cracking - SCC (IV.B4.RP- 3.1.1-37 A components of Alloy X- Neutron Fluence PWR Water 261) 750 Dowel-to-Lower Chemistry FuelAssembly Support Pad Welds)

Alloy X-750 Dowel-to- PWR Reactor Lower Grid Borated Reactor Vessel Internals None 3A5 Shell Forging Welds Support Alloy Coolant with Cracking - SCC (IV.B4.RP- None A (no additional Allo Neutron Fluence PWR Water 236) measures component) Chemistry

Enclosure L-1 1-252 Page 17 of 108 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 Notes Management Program (Rev. 2 Item) Item PWR Reactor Alloy X- 750 Dowel-to- Cracking - Vessel Internals None Lower Grid Rib Borated Reactor SCC, IASCC (IV.B4.RP- None A 36 Section Welds __o Nickel Borat Rato PWR Water 236)

Coolant with Chemistry (no additional Alloy Neutron Fluence Reduction in None measures component) fracture PWR Reactor (IV.B4.RP- None A toughness Vessel Internals 237)

Lower Grid Rib-to-Shell Cracking -

Forgina Stainles Borated Reactor fatiue None 37 ScrewsSSteel Supo Steel Coolantwith Loss of PWR Internals VesselReactor 237).P-(IV.B4.R Nne None A A

(no additional Neutron Fluence material- wear measures component) Loss of preload Lower Grid Support Cracking -

Post Pipe Cap BoratedReactor fatgue None SteelVesseIntenals(IV.B4.RP- None None A 38 Screws (no additional Suuppo tainless Coolant with Steel Neutron Fluence Loss of - wear material VesselReactor PWR Internals measures component) Loss of preload

Enclosure L-1 1-252 Page 18 of 108 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 No._Functions)Management Management Rev. 1 Item Notes Program (Rev. 2 Item) Item PWR Reactor Lower Thermal Shield Bolt: Cracking - Vessel Internals IV.B4-32 (LTS) Bolts and their ' (IV.B4.RP- 3.1.1-37 A locking devices Bolt: SCC PWR Water 246)

Nickel Chemistry (expansion component Alloy with primary component Borated Reactor BV4t7 link of UCB Bolts

ýL andt BoltsNeutron Lc Coolantwith Cumulative TLAA IV.B4-37 -05 A Fluence fatigue

- fatiguedamage (VB.5_ 31-5 their locking devices, Devices:

LCB Bolts and their Stainless Locking locking devices, and FD Steel Device: Reactor VB4-01 IV.

Bolts and their locking WR (IV.B4.RP- 3.1.1-22 C devices) Loss of Vessel Internals 243) 1 material- wear I Flow DistributorAssembly Flow Distributor(FD) PWR Reactor Bolts and their lockin Bolt: Cracking - Vessel Internals IV.B4-25 devices SCC (IV.B4.RP- 3.1.1-37 A SCCPWR Water 256) 01 (primarycomponent with expansion Borated Reactor Chemistry 40 components of UTS Suppo Steel Coolant with Bolts and their locking Neutron Fluence Locking devices, LTS Bolts and Device: PWR Reactor IV.B4-01 their lockinq devices, Loss of (IV.B4.RP- 3.1.1-22 C Vessel Interals 243) and SSHT Bolts and material- wear their locking devices)

Enclosure L-1 1-252 Page 19 of 108 Table 3.1.2-2 Aging Management Review Results - Reactor Vessel Internals Incore MAonitorinq Instrumentation(MIV Guide Tube Assembl' IMI Guide Tube Spiders (primarycomponent with expansion components of CRGT SpacerCasting and Cast Lower Fuel Assembly Borated Reactor Reduction in IV.B4-28 Austenitic PWR Reactor 42 Support Pads: Pad, Stainless Coolant with fracture Vessel Internals (IV.B4.RP- 3.1.1-801A Pad-to-Rib Section Neutron Fluence toughness 258)

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

Enclosure L-1 1-252 Page 20 of 108 Table 3.1.2-2 Aging Management Review Results - Reactor Vessel Internals Row Component Type No.

IMI Guide Tube Spider-to-Lower Grid Rib Section Welds (primarycomponent with expansion components of CRGT Spacer Castingand 43 Lower Fuel Assembly Support Pads: Pad, Pad-to-Rib Section Weld, Cap Screw and associatedLocking Weld, Alloy X-750 Dowel and Alloy X-750 Dowel Locking Weld)

Core Support Structure Stainless Steel, Reactor Vessel Nickel Borated Reactor Loss of . IV.B4-42 A 44 Internals Suppo ast Coolantwh material- wear Inspection 3.1.1-63 0115 (accessiblesurfaces) Austenitic Neutron Fluence 382)

Stainless Steel Borated Reactor Reduction in None 45 Internals Suppo Steel Coolant with fracture TLAA (IV.B4.RP- None 0 Neutron Fluence toughness 376) 0116

Enclosure L-1 1-252 Page 21 of 108 Plant-Specific Notes:

0114 Flow distributor(FD) bolts were reassiqnedas a primary component per MRP-227, Rev. 0 as amended by the safety evaluation.

Due to this change, the FD bolts are not an expansion component for the UCB and LCB bolts.

0115 Applicable to accessible surfaces of the removable Core Support Structure. For components in the "No Additional Measures"group that are classified as core supportstructures, the inservice inspection requirementsof the ASME Code Section X1, Subsection IWB, Examination Category B-N-3 must continue to be met.

0116 PWR Reactor Vessel Intemals Programmanages the TLAA associatedwith the reductionin fracture toughness for RV internals.

LRA Section 4.2.7 addresses the subiect TLAA.

Enclosure L-1 1-252 Page 22 of 108 Affected LRA Section LRA Page No. Affected Paragraph and Sentence Table 3.2.1 Page 3.2-40 Row 3.2.1-53 "Discussion" Text in "Discussion" column is revised based on the response to Supplemental RAI 3.2.2.2.3.6-2. LRA Table 3.2.1, "Summary of Aging Management Programs for Engineered Safety Systems Evaluated in Chapter VII of NUREG-1801," reads as follows:

Table 3.2.1 Summary of Aging Management Programs for Engineered Safety Features Systems Evaluated in Chapter V of NUREG-1 801 Further Evaluation Item Aging Effect/ Aging Management Number Component/Commodit Mechanism Programs Ecommended Recommended 3.2.1-53 Stainless steel, copper alloy, and None None NA - No AEM or Consistent with NUREG-1 801.

nickel alloy piping, components, piping and piping AMP No N aging gn effects fet requiring eurn elements exposed to air - indoor management were identified for uncontrolled (external) any stainless steel, copper alloy, or nickel alloy piping, piping components, or piping elements that are exposed to air-indoor uncontrolled (external).

This item is also applied to stainless steel and copper alloy heat exchanger components, and to stainless steel tanks, that are exposed to an air-indoor uncontrolled (external).

This item is also applied to internalsurfaces of stainless steel and copper alloy piping components, and to stain!ovo

____ank&7 that are exposed to

Enclosure L-1 1-252 Page 23 of 108 Table 3.2.1 Summary of Aging Management Programs for Engineered Safety Features Systems Evaluated in Chapter V of NUREG-1801 Further Discussion Item Aging Effect/ Aging Management Evaluation Number Component/Commodity Mechanism Programs Ecommendes Recommended an air-indooruncontrolled (internal)where it has been demonstratedthat the internal environment is the same as the externalenvironment.

Enclosure L-11-252 Page 24 of 108 Affected LRA Section LRA Page No. Affected Paragraph and Sentence Table 3.2.2-2 Page 3.2-53 Row 20 In response to Supplemental RAI 3.2.2.2.3.6-2, row 20 of LRA Table 3.2.2-2, "Aging Management Review Results - Containment Spray System," is revised as follows:

Table 3.2.2-2 Aging Management Review Results - Containment Spray System dAging Effect NUREG-Row Component Intended Material Environment Requiring Aging Management 1801, Table 1 Notes No. Type Function(s) Management Program Volume Item 2 Item One Tim~e In~pection Pressure Stainless Moist air Inspection of Internal H 20 Piping boundary Steel (Internal) Cracking Surfaces in N/A N/A Miscellaneous Piping and Ductinq

Enclosure L-1 1-252 Page 25 of 108 Affected LRA Section LRA Page No. Affected ParaaraDh and Sentence Table 3.2.2-4 Pages 3.2-86 Rows 113 & 119 and 3.2-87 In response to Supplemental RAI 3.2.2.2.3.6-2, rows 112 and 118 of LRA Table 3.2.2-4, "Aging Management Review Results - Decay Heat Removal and Low Pressure Injection System," are no longer necessary and are marked as "Not used." LRA Table 3.2.2-4 is revised as follows:

Table 3.2.2-4 Aging Management Review Results - Decay Heat Removal and Low Pressure Injection System Aging Effect NUREG-Row Component Intended Material Environment Requiring Aging Management 1801, Table 1 Notes No. Type Function(s) Management Program Volume 2 Item Item T11 R1442 4-Not used Tank InagrQ tankAir ;lnde 11 D -4 bona se nGentreged NAtne Nona 4F1 32 0201 118 (DTu92) _--

Not used

Enclosure L-1 1-252 Page 26 of 108 Affected LRA Section LRA Page No. Affected Paragraph and Sentence Table 3.2.2 Page 3.2-118 Row 0202 Plant-Specific Notes In response to Supplemental RAI 3.2.2.2.3.6-2, row 0202 of Table 3.2.2, "Plant-Specific Notes," is no longer used, and is revised as follows:

Plant-Specific Notes:

0202 The One Time In.pection w# confi, .

fGo s..bject to a "oist

.mponents air (nter n.." environment, the absence of agi.g effecta o that aging is slow aeting o as to not afet the sUbject component's intended fUncton dvg the period of etended operation,

..hcverifies the effectiveness of aging management proegrmsg credited above and belW thsinefae Not used

Enclosure L-1 1-252 Page 27 of 108 Affected LRA Section LRA Page No. Affected Paragraph and Sentence 3.3.2.1.3 Page 3.3-7 "Environments" subsection In response to Supplemental RAI 3.2.2.2.3.6-2, the "Environments" subsection of LRA Section 3.3.2.1.3, "Auxiliary Steam and Station Heating System," is revised to read as follows:

Environments Subject mechanical components of the Auxiliary Steam and Station Heating System are exposed to the following normal operating environments:

Air with borated water leakage

Air with steam or water leakage

Air-indoor uncontrolled

Closed cycle cooling water > 600C (> 1400 F)

Condensation

Moist air

" Steam

Enclosure L-1 1-252 Page 28 of 108 Affected LRA Section LRA Page No. Affected Paragraph and Sentence Table 3.3.2-2 Pages 3.3-154 Rows 52, 55, 58, 61, 64, & 67 thru 3.3-156 In response to Supplemental RAI 3.2.2.2.3.6-2, rows 52, 55, 58, 61, 64 and 67 of LRA Table 3.3.2-2, "Aging Management Review Results - Auxiliary Building Chilled Water System," are revised as follows:

Table 3.3.2-2 Aging Management Review Results - Auxiliary Building Chilled Water System Aging Effect NUREG-Row Component Intended Material Environment Requiring Aging Management 1801, Table I Notes No. Type Function(s) Management Program Volume Item 2 Item Tank-AiF 52 . St-ee/ 1eee-ef Ext V41. .3. -35 Not used Onei Time InspeGctkn Tank - Air Structural Steel Moist air (internal) Loss of material Inspection Surfaces inof Internal VII.G-23 3.3.1-71 E 55 separator integrity (Miscellaneous Piping and Ductino Air/,-,tdnccr~rfcce 58 used4) Uneeteled Not used

Enclosure L-1 1-252 Page 29 of 108 Table 3.3.2-2 Aging Management Review Results - Auxiliary Building Chilled Water System Aging Effect NUREG-Row Component Intended Material Environment Requiring Table Notes No. Type Function(s) Management Program Volume Item 2 Item one Time3 ln~pectien Tank -calkp Inspection of Internal Chemical pot Structural Steel Moist air Loss of Surfaces in VII.G-23 3.3.1-71 E 61 feeder T154) (DB- integrity (internal) materialMiscellaneous PipinQ and Ductinq 64 -T--) steel UMGe/rned 141 ,3.34-5 Not used One Tkrno !%specti#n Tank-Expansion Structural Moist air Loss of Inspection of Intemal txpank (Dio itegrictu Steel Mointemai) matesalSurfaces in VII.G-23 3.3.1-71 E 67 tank (D13- integrity (interal) material Miscellaneous T88) PipinQ and Ductinq

Enclosure L-1 1-252 Page 30 of 108 Affected LRA Section LRA Page No. Affected Paragraph and Sentence Table 3.3.2-3 Pages 3.3-170 Rows 61 & 64 and 3.3-171 In response to Supplemental RAI 3.2.2.2.3.6-2, rows 61 and 64 of LRA Table 3.3.2-3, "Aging Management Review Results - Auxiliary Steam and Station Heating Systems," are revised as follows:

Table 3.3.2-3 Aging Management Review Results - Auxiliary Steam and Station Heating Systems dAging Effect NUREG-Row Component Intended Material Environment Requiring Aging Management 1801, Table 1 Notes No. Type Function(s) Management Program Volume Item 2 Item 61 .. .. Speum Air--(In4ernao) One Tirm !n~poctien ...1W:2-24 &3.4-7-4

('D& P27 1 neg~ m;;tead 0342 Not used Pump Casing One Tire lnspoqtiqn

- Evaporator package Structural Moist air Loss of Inspection of Intemal E 64integrity (internal) material Surfaces in VII.G-23 3.3.1-71 drain pump Miscellaneous (DB-P275-1 Pipinq and Ductina

&2) 1 1

Enclosure L-1 1-252 Page 31 of 108 Affected LRA Section LRA Page No. Affected Paragraph and Sentence Table 3.3.2-4 Page 3.3-211 Row 159 In response to Supplemental RAI 3.2.2.2.3.6-2, row 159 of LRA Table 3.3.2-4, "Aging Management Review Results - Boron Recovery System," is revised as follows:

Table 3.3.2-4 Aging Management Review Results - Boron Recovery System dAging Effect NUREG-Row Component Intended Material Environment Requiring Aging Management 1801, Table I Notes No. Type Function(s) Management Program Volume 2 Item Item one Timeo in~pertion Tank -

159 Concentrates soaetn Structural inert Stainless Ste Moist air (Itra) Cracking Inspectioninof nte/al Surfaces N/A N/A H

-

storage tank integrity Steel (internal) Miscellaneous (DB-T16) Piping and Ductinq

Enclosure L-11-252 Page 32 of 108 Affected LRA Section LRA Page No. Affected Paragraph and Sentence Table 3.3.2-5 Page 3.3-225 Row 60 In response to Supplemental RAI 3.2.2.2.3.6-2, row 60 of LRA Table 3.3.2-5, "Aging Management Review Results - Chemical Addition System," is revised as follows:

Table 3.3.2-5 Aging Management Review Results - Chemical Addition System Row Component Rown Intended Material Environment Aging EcomponentMaageeInten1dealed Effect Requiring Aging Management TNUREG-1801, Table 1 Notes No. Type Function(s) Management Program Volume 2 Item Item 0n3 Time inSpectien Tank - Boric Inspection of Interal acid addition Pressure Stainless Moist air Cns in of Itn H 60 tanks (DB- boundary Steel (Internal) Cracking Suraces in N/A N/A033 T7-1 & 2) T7-1 & 2)Pipingq and Ductincq Miscellaneous

Enclosure L-1 1-252 Page 33 of 108 Affected LRA Section LRA Page No. Affected Paragraph and Sentence Table 3.3.2-7 Page 3.3-245 Row 77 In response to Supplemental RAI 3.2.2.2.3.6-2, row 77 of LRA Table 3.3.2-7, "Aging Management Review Results - Component Cooling Water System," is no longer needed, and is revised as follows:

Table 3.3.2-7 Aging Management Review Results - Component Cooling Water System Aging Effect NUREG-Row Component Intended Material Environment Requiring Aging Management 1801, Table I Notes Volume Item 2 No. Type Function(s) Management Program Item chemieal 4 77 fed9F-(DB

-T-/4-)

tUWa integty 2P-Q; Rtpo Ar ne UGGt*-Jed A None 1 3.3.4-94 -4; 0307 Not used

Enclosure L-1 1-252 Page 34 of 108 Affected LRA Section LRA Page No. Affected Paragraph and Sentence Table 3.3.2-11 Page 3.3-276 Row 28 and previously added row In response to Supplemental RAI 3.2.2.2.3.6-2, row 28 and a previously added level gage row of LRA Table 3.3.2-11, "Aging Management Review Results - Demineralized Water Storage System," are no longer needed and are revised as follows:

Enclosure L-1 1-252 Page 35 of 108 Affected LRA Section LRA Page No. Affected Paragraph and Sentence Table 3.3.2-12 Pages 3.3-302 Rows 158, 161 and a previously and 3.3-303 added row In response to Supplemental RAI 3.2.2.2.3.6-2, rows 158, 161, and a previously added level gage row of LRA Table 3.3.2-12, "Aging Management Review Results - Emergency Diesel Generators System," are revised as follows:

Table 3.3.2-12 Aging Management Review Results - Emergency Diesel Generators System dAging Effect NUREG-Row Component Intended Material Environment Requiring Aging Management 1801, Table 1 Notes No. Type Function(s) Management Program Volume 2 Item Item Tank-Jaoket 44t 158 tank (9f ...... fos Ext-! Steef W 3.3. i--

T-121 & 2) bo.nda. (R.... material Mo.ke.in0307 Not used one T44me Inaoerti*n Tank-Jacket water Pressure Moist air Loss of Inspection of Internal E 161 expansion Steel (Internal) material Surfaces in VII.H2-21 3.3.1-71 tank (DB- Miscellaneous T121-1 &2) Piping and Ductinq

Enclosure L-11-252 Page 36 of 108 Table 3.3.2-12 Aging Management Review Results - Emergency Diesel Generators System Ro Cmonn ItnddAging Effect Aging Management NUREG-1801, Table 1 Row Component Intended Material Environment Requiring Agn aae et 10 , T be1 Notes No. Type Function(s) Management Program Volume 2 Item Item Ai in;oe~

L-eveel Gage bN-.edne 1411 f- .0307 (tetedna

Enclosure L-1 1-252 Page 37 of 108 Affected LRA Section LRA Page No. Affected Paragraph and Sentence Table 3.3.2-14 Page 3.3-313 Entire Table The NRC initiated a telephone conference call with FENOC on September 7, 2011, to discuss the replacement of LRA Table 3.3.2-14, "Aging Management Review Results - Fire Protection System," in FENOC letter dated August 26, 2011 (ML11242A166). The NRC staff noted that, following a line-by-line comparison to LRA Table 3.3.2-14, the fire pump diesel engine rows appeared to be missing in the replacement table. LRA Table 3.3.2-14, "Aging Management Review Results - Fire Protection System," submitted in FENOC letter dated August 26, 2011 (ML11242A166), is replaced in its entirety to include the Fire Protection System components and the fire pump diesel engine and associated components, and to include changes in response to Supplemental RAI 3.2.2.2.3.6-2, and reads as follows:

Enclosure L-11-252 Page 38 of 108 Table 3.3.2-14 Aging Management Review Results - Fire Protection System

Enclosure L-11-252 Page 39 of 108 Table 3.3.2-14 Aging Management Review Results - Fire Protection System Row Component Intended AigEfc Aging Effect Aging gn NUREG-81 al Row Cmponen Fnctiond Material Environment Requiring Management 1801, Table 1 Notes No. Type Function(s) Management Program Volume Item 2 Item Collection, Pressure Raw water Drainage, and G 12 Bolting boundary Steel (External) Cracking Treatment N/A N/A 0324 Components Inspection Collection, 13 Bolting Pressure Presure Steel Raw water Loss of Drainage, and E 13 B g boundary (Treatment VI1.C1-19 3.3.1-76 0324 S l (External) material Components Inspection Pressure Raw water Loss of 14 Bolting boundary Steel (External) preload Bolting Integrity NA NA H 15 Bolting Pressure Steel Soil (External) Loss of Buried Piping and boundary material Tanks Inspection VII.G-25 3.3.1-19 C Air with steam 16 Bolting integrity Steel lewater Cracking Bolting Integrity VII.1-3 3.3.1-41 B Snteructurl Ste leakage (External)

Air with steam Structural or water Loss of 17 Bolting integrity Steel leakage material Bolting Integrity VII.I-6 3.3.1-42 B (External)

Structural Loss of 18 Bolting integrity Steel uncontrolled materialBolting Integrity VII.-4 3.3.1-43 B (External) material Structural Loss of 19 Bolting integrity Steel uncontrolled (External) preloadBolting Integrity V.-5 3.3.1-45 B pr d I I I I

Enclosure L-11-252 Page 40 of 108 Table 3.3.2-14 Aging Management Review Results - Fire Protection System Aging Effect Aging NUREG-Row Component Intended Mring Magint 1801, Table 1 Notes No. Type Function(s) Material Environment Requiring Management Volume Item Management Program 2 Item Heat Exchanger (channel) - Air-indoor 20 Fire Water Pressure Steel uncontrolled Loss of External Surfaces VIIG-5 3.3.1-59 A Storage Tank boundary (External) material Monitoring Heat Exchanger (DB-E52)

Heat Exchanger (channel) -

Fire Water Pressure Raw water Loss of Fire Water VII.G-24 3.3.1-68 C 21 Storage Tank boundary Steel (Internal) material Heat Exchanger (DB-E52)

Heat Exchanger (shell) - Fire Water Pressure Steam Loss of One-Time E 22 Storage Tank boundary (Internal) material Inspection 0315 Heat Exchanger (DB-E52)

Enclosure L-11-252 Page 41 of 108 Table 3.3.2-14 Aging Management Review Results - Fire Protection System Row Component Intended Aging Effect Aging NUREG-Row Cmponen Fnctiond Material Environment Requiring Management 1801, Table 1 Notes No. Type Function(s) Maaeet Program Volume Item Management P2 Item Heat Exchanger (shell) - Fire 23 Water Pressure Steel Steam Loss of PWR Water VIII.B1-8 3.4.1-37 C Storage Tank boundary (Internal) material Chemistry Heat Exchanger (DB-E52)

Heat Exchanger (shell) - Fire Air-indoor 24 Water SogeTn Pressure bonay Steel Air-indoor uncontrolled Loss of External Surfaces mtraMoirngVII.G-5 3.3.1-59 A Storage Tank boundary (External) material Monitoring Heat Exchanger (DB-E52)

Heat Exchanger Collection, (tubes) - Fire Drainage, and Water Stainless Raw water Reduction in Draina and Storage Tank Heat transfer Steel (Internal) heat transfer Components Heat Inspectnts Exchanger Inspection (DB-E52)

Enclosure L-1 1-252 Page 42 of 108 Table 3.3.2-14 Aging Management Review Results - Fire Protection System Aging Effect Aging NUREG-Row Component Intended Aging Magint 1801, Table 1 Notes No. Type Function(s) Material Environment Requiring Management Management Program Volume Item 2 Item Heat Exchanger (tubes) - Fire StoWater Storage Tank Heat transfer Stainless Steel Steam (External) Reduction in heat transfer PWR Water Chemistry N/A N/A G Heat Exchanger (DB-E52)

Heat Exchanger (tubes) - Fire Water Stainless Steam Reduction in One-Time G Storage Tank Steel (External) heat transfer Inspection 0315 Heat Exchanger (DB-E52)

Heat Exchanger (tubesheet) -

Fire Water Pressure Raw water Loss of 28 Storage Tank boundary Steel (Inter (internal) materialFire material Water VII.G-24 3.3.1-68 C Heat Exchanger (DB-E52)

Enclosure L-1 1-252 Page 43 of 108 Table 3.3.2-14 Aging Management Review Results - Fire Protection System Aging Effect Aging NUREG-Row Component Intended Material Environment Requiring Management 1801, Table I Notes No. Type Function(s) Malagenent Manam Volume Item Management Program 2 Item Heat Exchanger (tubesheet) -

29 Fire Water Pressure Steel Steam Loss of Steam.Lossof3One-Tim One-Time E Storage Tank boundary (External) material Inspection 0315 Heat Exchanger (DB-E52)

Heat Exchanger (tubesheet) -

30 Fire Water Pressure Steam Loss of PWR Water VIII.B1-8 3.4.1-37 C Storage Tank boundary Steel (External) material Chemistry Heat Exchanger I___(DB-E52) 31 Hydrant Pressure Gray Cast Raw water Loss of Fire Water VII.G-24 3.3.1-68 A boundary Iron (Internal) material 32 Hydrant Pressure Gray Iron Cast Raw water (Internal) Loss of material Selective InspectionLeaching VII.G-14 3.3.1-85 A boundary 33 Pressure Gray Cast Air-outdoor Loss of External Surfaces VII.I-9 3.3.1-58 A Hydrant boundary Iron (External) material Monitoring 34 Hydrant Hydrant Pressure boundary Gray Iron Cast Soil (External) Loss of material Buried Inspection Tanks Piping and VII.G-25 3.3.1-19 A

Enclosure L-1 1-252 Page 44 of 108 Table 3.3.2-14 Aging Management Review Results - Fire Protection System Aging Effect Aging NUREG-Row Component Intended Magint 1801, Table 1 Notesing No. Type Function(s) Material Environment Requiring Management Volume Item Management Program 2 Item Pressure Gray Cast Loss of Selective Leaching VII.G-15 3.3.1-85 A 35 Hydrant Prssboundary I Soil (External) material Inspection OiiePressure Raw water Loss of 36 Orifice boundary Steel (Inter (Internal) materialFire Water material VII.G-24 3.3.1-68 A Pressure Air-indoor Loss of material External Surfaces 37 Orifice 37 boundaryOrifice Steel uncontrolled (External) materialonitorin Monitoring3.3.1-58 A Raw water Loss of 38 Orifice Throttling Steel (Internal) material Fire Water VII.G-24 3.3.1-68 A Pressure Copper Raw water Loss of FireWater VIIG-12 3.3.1-70 A 39 Piping boundary Alloy (Internal) material Air with 40 Piping Pressure Copper borated water None None VIIJ-5 3.3.1-99 A Piping boundary Alloy leakage (External)

Air-indoor 41 Piping Pressure Copper uncontrolled None None VIII.I-2 3.4.1-41 A boundary 41 Piping Alloy (External)

Pressure Gray Cast Raw water Loss of Fire Water VIIG-24 3.3.1-68 A 42 Piping boundary Iron (Internal) material 43 PipingPressure Gray Cast Raw water Loss of Selective Leaching VILG.14 3.3.1-85 A Piping boundary Iron (Internal) material Inspection

Enclosure L-1 1-252 Page 45 of 108 Table 3.3.2-14 Aging Management Review Results - Fire Protection System Aging Effect Aging NUREG-Row Component Intended Material Environment Requiring Management 1801, Table 1 Notes No. Type Function(s) Malagenent Manam Volume Item Management Program 2 Item 44 Piping Pressure Gray Cast Air-outdoor Loss of External Surfaces VII.I-9 3.3.1-58 A boundary Iron (External) material Monitoring 45 Pipin Pressure Gray Cast Soil (External) Loss of Buried Piping and VII.G-25 3.3.1-19 A g boundary Iron material Tanks Inspection Pressure Gray Cast Soil (External) Loss of Selective Leaching VII.G-15 3.3.1-85 A 46 Piping boundary Iron material Inspection Pressure Stainless Air-indoorC 47 Piping Steel 47 Pipgboundary uncontrolled (Internal) None None VII.J-15 3.3.1-94 0301 Air with 48 Piping borated None None VIIJ-16 3.3.1-99 A Pressure boundary Stainless Steel leakage water (External)

Pressure Stainless Air-indoor 49 Piping Steel uncontrolled None None VII.J-15 3.3.1-94 A gboundary (External)

Air with 50 Piping Pressure Steel borated water Loss of Boric Acid VII.I-10 3.3.1-89 A boundary leakage material Corrosion (External) 51 Piping Pressure boundary Steel Ai-nor uncontrolled Loss of material External Surfaces Monitoring3.3.1-58 C 0301 (Internal)

Pressure Air-outdoor Loss of External Surfaces C 52 Piping boundary Steel (Internal) material Monitoring VII.I-9 3.3.1-58 0301

Enclosure L-1 1-252 Page 46 of 108 Table 3.3.2-14 Aging Management Review Results - Fire Protection System Row Component Intended Aging Effect Aging NUREG-Ro opnn nedd Material Environment Requiring Management 1801, Table 1 Notes No. Type Function(s) Management Program Volume Item 2 Item 53 Piping Pressure boundary Steel Raw water (Internal) Cracking TIA N/A H N/A 0339 Piping Pressure Raw water Loss of boundary Steel (Internal) material Fire Water VII.G-24 3.3.1-68 A 55 Piping Pressure boundary Steel Air-indoor uncontrolled uncoteroalle Loss of External MonitoringSurfaces mtraMoirngVI1.1-8 material 3.3.1-58 A 55 Ppingbounary(External) 56 Piping Pressure Air-outdoor Loss of External Surfaces VII.I-9 3.3.1-58 A boundary Steel (External) material Monitoring 57 Piping Pressure Concrete None None VIIJ-21 3.3.1-96 A 7Piping boundary Steel (External)

Pressure Raw water Loss of A 58 Piping boundary Steel (External) material Fire Water VII.G-24 3.3.1-68 0323 PrsueLoss of Buried Piping and 59 Piping Pressure Steel Soil (External) material Tanks Inspection VII.G-25 3.3.1-19 A boundarymAi Structural Copper Air-indoorA 60 Piping integrity Alloy uncontrolled None None VIII.I-2 3.4.1-41 0301 60_ Pipingintegrity Alloy (Internal) 0301 Structural Copper Air-indoor 61 Piping integrity Strityral Alloy loy uncontrolled (External) None None VIII.I-2 3.4.1-41 A

Enclosure L-1 1-252 Page 47 of 108 Table 3.3.2-14 Aging Management Review Results - Fire Protection System Row Component Intended AigEfc Aging Effect Aging gn NUREG-81 al Row Cmponen Fnctiond Material Environment Requiring Management 1801, Table 1 Notes No. Type Function(s) Maaeet PormVolume Item Management Program 2 Item Structural Loss of External Surfaces A 62 Piping integrity Steel uncontrolled material Monitoring3.3.1-58 0301 (Internal) mtrl Mnoi 63 Piping Structural Steel Raw water Loss of Fire Water VIIG-24 3.3.1-68 A iping integrity (Internal) material 64 Piping Structural integrity Steel Air-indoor uncontrolled (External) Loss of mtraMoirngVII.I-8 material External MonitoringSurfaces 3.3.1-58 A Pump Casing Fire Diesel(DB- Pressure Gray Cast Moist air Loss of Selective Leaching N/A N/A H 65 -Pump boundary Iron (Internal) material Inspection 0321 P5-2)

Pump Casing Inspection of 66 - Diesel Fire Pressure Gray Cast Moist air Loss of Internal Surfaces in VII.G-23 3.3.1-71 E Pump (DB- boundary Iron (Internal) material Miscellaneous P5-2) Piping and Ducting Pump Casing 67 - Diesel Fire Pressure Gray Cast Raw water Loss of Fire Water VII.G-24 3.3.1-68 A Pump (DB- boundary Iron (Internal) material P5-2)

Pump Casing 68 -Pump Diesel Fire Pressure Gray Iron Cast (Internal)

Raw water Loss of Selective InspectionLeaching (DB- boundary material VII.G-14 3.3.1-85 A P5-2)

Pump Casing Pup asngAi-idor Air-indoor Loss of External Surfaces Diesel(DB-

-Pump Fire Pressure boundary Gray Iron Cast uncontrolled (External) LVII.I-8 material Monitoring 3.3.1-58 A P5-2) (External)

Enclosure L-1 1-252 Page 48 of 108 Table 3.3.2-14 Aging Management Review Results - Fire Protection System Component Intended Aging Effect Aging NUREG-Row Row Cmponen Fnctiond Material Environment Requiring Management 1801, Table 1 Notes No. Type Function(s) Management Program Volume Item 2 Item Pump Casing 70 - Diesel Fire Pressure Gray Cast Moist air Loss of Selective Leaching N/A N/A G Pump (DB- boundary Iron (External) material Inspection 0321 P5-2)

Pump Casing Inspection of 71 - Diesel Fire Pressure Gray Cast Moist air Loss of Internal Surfaces in VII.G-23 3.3.1-71 E Pump (DB- boundary Iron (External) material Miscellaneous 0322 P5-2) Piping and Ducting Pump Casing 72 -Pump Diesel(DB-Fire Pressure boundary Gray Iron Cast Raw water (External) Loss of material Fire Water VII.G-24 3.3.1-68 A P5-2)

Pump Casing 73 - Diesel Fire Pressure Gray Cast Raw water Loss of Selective Leaching VII.G-14 3.3.1-85 A Pump (DB- boundary Iron (External) material Inspection P5-2)

Pump Casing 74 - Electric Fire Pressure Gray Cast Raw water Loss of Fire Water VIIG-24 3.3.1-68 A Pump (DB- boundary Iron (Internal) material P5-1)

Pump Casing 75 - Electric Fire Pressure Gray Cast Raw water Loss of Selective Leaching VII.G-14 3.3.1-85 A Pump (DB- boundary Iron (Internal) material Inspection P5-1)

Pump Casing Air-indoor Gray Cast uncontrolled Loss of External Surfaces VII.I-8 3.3.1-58 A 76 -Pump (DB-Fire Electric Pressure boundary Iron (External) material Monitoring P5-1)

Enclosure L-1 1-252 Page 49 of 108 Table 3.3.2-14 Aging Management Review Results - Fire Protection System Row Component Intended AigEfc Aging Effect Aging gn NUREG-81 al Row Type Intended Material Environment Requiring Management 1801, Table I Notes No. Type Function(s) Management Maaeet Program PormVolume 2VIteme Im Item 2 Item Pump Casing

- Fire Water 77 Storage Tank Pressure Gray Cast Raw water Loss of FireWater VIG-24 3.3.1-68 A Recirculation boundary Iron (Internal) material Pump (DB-P1 14)

Pump Casing

- Fire Water 78 Storage Tank Pressure Gray Cast Raw water Loss of Selective Leaching VII.G-14 3.3.1-85 A Recirculation boundary Iron (Internal) material Inspection Pump (DB-Pl 14)

Pump Casing

- Fire Water Air-indoor 79 Storage Tank Pressure Gray Cast uncontrolled Loss of External Surfaces Recirculation boundary Iron material Monitoring Pump (DB- (External)

P114)

Pressure Copper Air-indoor A 80 Spray Nozzle Alloy > uncontrolled None None VIII.I-2 3.4.1-41 0301 80 Seboundary 15% Zn (Internal)

Inspection of CorPressure Coy Air-outdoor Cracking Internal Surfaces in N/A N/A G boundary 15% Zn (Internal) Miscellaneous 15%_ _n Piping and Ducting Inspection of 82 Spray Nozzle Pressure Copper Air-outdoor Loss of Internal Surfaces in N/A N/A G Zn y15% (Internal) material Miscellaneous 82ISprayNozzleboundary I 15% n I Piping and Ducting

Enclosure L-1 1-252 Page 50 of 108 Table 3.3.2-14 Aging Management Review Results - Fire Protection System Row Component Intended Aging Effect Aging NUREG-Row Cmponen Fnctiond Material Environment Requiring Management 1801, Table I Notes No. Type Function(s) Maaeet Management PormVolume Program 2 Item Item 83 Spray Nozzle Pressure Copper Air-outdoor Loss of Selective Leaching boundary Alloy> (Internal) material Inspection N/A N/A G 15% Zn 84 Spray Nozzle boundary Alloy > (internal) material Fire Water VII.G-12 3.3.1-70 A 15% Zn 85 Spray Nozzle Pressure Copper Raw water Loss of Selective Leaching boundary Alloy > (Inter Loss Inspectio n VII.G-13 3.3.1-84 A 15% Zn (Internal) material Inspection Air with 86 Spray Nozzle Pressure boundary Copper Alloy > borated water lekg Loss of Boric Acid aeilCroinVI1.1-12 3.3.1-88 A bonay 15% Zn leakage material Corrosion 15%_ _n (External)

Pressure Copper Air-indoor 87 Spray Nozzle boundary Alloy> uncontrolled None None VIII.I-2 3.4.1-41 A 15% Zn (External) 88 Spray Nozzle Pressure boundary Copper Alloy > Air-outdoor (External) Cracking External MonitoringSurfaces N/A N/A G 15% Zn Pressure Copper Air-outdoor Loss of External Surfaces 89 Spray Nozzle boundary Alloy > (External) material Monitoring N/A N/A G 15% Zn Pressure Copper Air-outdoor Loss of Selective Leaching 90 Spray Nozzle boundary 15% Zn (External) material Inspection N/A N/A G Copper Air-indoor A 91 Spray Nozzle Spray Alloy > uncontrolled None None VIII.I-2 3.4.1-41 0301 15% Zn (Internal)

Enclosure L-1 1-252 Page 51 of 108 Table 3.3.2-14 Aging Management Review Results - Fire Protection System Row Component Intended Aging Effect Aging NUREG-No. Type Function(s) Material Environment Requiring Management V m Tem Notes Management Program Volume Item 2 Item Inspection of Spray Copper Alloy >

Air-outdoor (Itra)

Cracking Internal Surfaces in CaknMiclneuN/NA N/A N/A G G

92 Spray Nozzle 15% Zn (Internal) Miscellaneous 15%_Zn Piping and Ducting Inspection of Copper Air-outdoor Loss of Internal Surfaces in N/A N/A G 15% Zn (Internal) material Miscellaneous Piping and Ducting Copper Air-outdoor Loss of Selective Leaching N/A N/A G 94 Spray Nozzle Spray Alloy > (internal) material Inspection 15% Zn Copper Raw water Loss of 95 Spray Nozzle Spray Alloy > (Internal) material Fire Water VII.G-12 3.3.1-70 A 15% Zn Copper Raw water Loss of Selective Leaching VII.G-13 3.3.1-84 A 96 Spray Nozzle Spray Alloy > (Internal) material Inspection 15% Zn Copper Air with water borated Loss of Boric Acid VII.l-12 3.3.1-88 A 97 Spray Nozzle Spray Alloy > laaemtra orso I.-2 3318 15% Zn leakage material Corrosion (External)

Copper Air-indoor 98 Spray Nozzle Spray Alloy > uncontrolled None None VIII.I-2 3.4.1-41 A 15% Zn (External)

Copper Air-outdoor External Surfaces 99 Spray Nozzle Spray Alloy >

15% Zn (External) Cracking Monitoring N/A N/A G

Enclosure L-11-252 Page 52 of 108 Table 3.3.2-14 Aging Management Review Results - Fire Protection System Row Component Intended Aging Effect Aging NUREG-Row Type Intended Material Environment Requiring Management 1801, Table No. Type Function(s) Management Maaeet Program PormVolume 2VIteme Im 1 Item Notes 2 Item Copper Air-outdoor Loss of External Surfaces 100 Spray Nozzle Spray Alloy > (External) material Monitoring N/A N/A G 15% Zn Copper Air-outdoor Loss of Selective Leaching N/A N/A G 101 Spray Nozzle Spray Alloy > (External) material Inspection 15% Zn Structural Copper Air-indoor A 102 Spray Nozzle integrity Alloy > uncontrolled None None VIII.I-2 3.4.1-41 0301 15% Zn (Internal) 103 Spray Nozzle Structural integrity Copper Alloy > Raw water (Internal) Loss of material Fire Water VII.G-12 3.3.1-70 A 15% Zn Structural Copper Raw water Loss of Selective Leaching 104 Spray Nozzle integrity 15% Zn (Internal) material Inspection Structural Copper Air-indoor 105 Spray Nozzle integrity Alloy > uncontrolled None None VIII.I-2 3.4.1-41 A 15% Zn (External) 106 Strainer Pressure Gray Iron Cast Raw water (Internal) Loss of material FireWater VIIG-24 3.3.1-68 A (body) boundary 107 Strainer Pressure Gray Cast Raw water Loss of Selective Leaching VII.G-14 3.3.1-85 A (body) boundary Iron (Internal) material Inspection Air with 108 Strainer (body) Pressure boundary Gray Iron Cast borated leakage water Loss of material Boric Acid Corrosion VII.-l10 3.3.1-89 A (External)

Enclosure L-11-252 Page 53 of 108 Table 3.3.2-14 Aging Management Review Results - Fire Protection System Row Component Intended AigEfc Aging Effect Aging gn NUREG-81 al Row Cmponen Fnctiond Material Environment Requiring Management 1801, Table 1 Notes No. Type Function(s) Maaeet Program Volume Item Management Program 2 Item 3.3.1-58 A MonitoringSurfaces Strainer Pressure Gray Cast Air-indoor Loss of External VII.I-8 (body) boundary Iron (External)ial 110 Strainer Pressure Steel Raw water Loss of Fire Water VIIG-24 3.3.1-68 A (body) boundary (Internal) material Air with Strainer Pressure Steel borated water Loss of Boric Acid VII.I-10 3.3.1-89 A (body) boundary leakage material Corrosion (External)

Strainer Pressure Air-indoor Loss of External Surfaces 112 (body) boundary Steel uncontrolled material Monitoring VII.I-8 3.3.1-58 A (External)

Strainer Copper Raw water Loss of 113 (screen) Filtration Alloy > (External) material Fire Water VII.G-12 3.3.1-70 A 15% Zn Strainer Copper Raw water Loss of Selective Leaching 114 (screen) Filtration >

Alloy Zn (External) material Inspection VII.G-13 3.3.1-84 A 15%

115 Strainer Filtration Stainless Raw water Loss of Fire Water VII.G-19 3.3.1-69 A (screen) Steel (External) material Tank- Fire Inspection of 116 Water Pressure Steel Moist air Loss of Internal Surfaces in VII.G-23 3.3.1-71 E Storage Tank boundary (Internal) material Miscellaneous (DB-T81) Piping and Ducting

Enclosure L-11-252 Page 54 of 108 Table 3.3.2-14 Aging Management Review Results - Fire Protection System Row Component Intended Aging Effect Aging NUREG-No. Type Function(s) Material Environment Requiring Management V m Tem Notes Management Program Volume Item 2 Item Tank- Fire Water Pressure Raw water Loss of 117 Steel (Inter materialFire Water VII.G-24 3.3.1-68 C Storage Tank boundary (intemal) material (DB-T81)

Tank - Fire 118 Water Pressure Steel Air-outdoor Loss of Aboveground Steel VI.HI-11 3.3.1-40 B Storage Tank boundary (External) material Tanks Inspection 0333 (DB-T81)

Tank - Fire 119 Water Storage Tank Pressure boundary Steel Air-outdoor (External) Loss of material External MonitoringSurfaces VII.l-9 3.3.1-58 A (DB-T81)

Tank - Copper Air-indoor 120 Retard Alloy > uncontrolled None None VIII.1-2 3.4.1-41 0 Chamber boundary 15% Zn (Internal)

Copper Air Arwt with Tank -

12 1Rtank

- Pressure Copper borated water Loss of Boric Acid VII.l-12 3.3.1-88 C 121 Retard boundary leakage material Corrosion Chamber u15% Zn (External)

Tank - Pressure Copper Air-indoor 122 Retard boundary Alloy > uncontrolled None None VIII.l-2 3.4.1-41 C Chamber 15% Zn (External)

Tank - Pressure Gray Cast Air-indoor Loss of External Surfaces C 123 Retard b uncontrolled material Monitoring 3.3.1-58 301 0VII.-8 Chamber oundary Iron (Internal) material _ Monitoring_0301 Tank - Air with 124 Retard Pressure Gray Cast borated water Loss of Boric Acid VII.I-10 3.3.1-89 A Chamber boundary Iron leakage material Corrosion I__________ I(External)

_ I I I _ I _I _I

Enclosure L-11-252 Page 55 of 108 Table 3.3.2-14 Aging Management Review Results - Fire Protection System Row Component Intended Aging Effect Aging NUREG-Row Cmponen Fnctiond Material Environment Requiring Management 1801, Table 1 Notes No. Type Function(s) Maaeet PormVolume Item Management Program 2 Item Tank - Pressure Gray Cast Air-indoor Loss of External Surfaces 125 Retard bounre y Iron uncontrolled material Monitoring VII.I-8 3.3.1-58 A Chamber oundary Iron (External) 126 Tubing Pressure Copper Raw water Loss of Fire Water VII.G-12 3.3.1-70 A boundary Alloy (Internal) material Air with 127 Tubing Pressure boundary Copper Alloy borated leakage water None None VIIJ-5 3.3.1-99 A (External)

Air-indoor 128 Tubing Pressure Copper uncontrolled None None VIII.1-2 3.4.1-41 A gboundary Alloy (External)

Pressure Loss of External Surfaces C 129 Tubing boundary Steel uncontrolled material Monitoring3.31-58 0301 (Internal)

Pressure Raw water Loss of 130 Tubing boundary Steel (Internal) material FireWater VII.G-24 3.3.1-68 A Air with 131 Tubi Pressure Ste borated water Loss of Boric Acid VII.l-10 3.3.1-89 A uing boundary ee leakage material Corrosion (External) 132 Tubing Pressure boundary Steel Air-indoor uncontrolled Loss of material External Surfaces Monitoring3.3.1-58 A (External) mtrl Mnoi Structural Loss of External Surfaces C 133 Tubing integrity Steel uncontrolled (Internal) material maeIal Monitoring3.3.1-58 MonItoring 0301

Enclosure L-1 1-252 Page 56 of 108 Table 3.3.2-14 Aging Management Review Results - Fire Protection System Aging Effect Aging NUREG-Row Component Intended Aging Magint 1801, Table I Notes No. Type Function(s) Material Environment Requiring Management Volume Item Management Program 2 Item Structural Raw water Loss of 134 Tubing integrity Steel (Internal) material Fire Water VII.G-24 3.3.1-68 A Structural Air-indoor Loss of External Surfaces 135 Tubing integrity Steel uncontrolled material Monitoring VII.I-8 3.3.1-58 A (External) materialMonitoring 136 Valve Body Pressure Copper Air-indoor uncontrolled None None VIII.I-2 3.4.1-41 A OY boundary Alloy (Internal) 0301 Inspection of 137 Valve Body Pressure Copper Air-outdoor Loss of Internal Surfaces in N/A N/A G boundary Alloy (Internal) material Miscellaneous Piping and Ducting 138 Valve Body Pressure Copper Raw water Loss of Fire Water VI.G-12 3.3.1-70 A boundary Alloy (Internal) material Air with 139 Valve Body Pressure Copper Alloy borated leakage water None None VII-5 3.3.1-99 A boundary (External)

Pressure Copper Air-indoor 140 Valve Body uncontrolled None None VIII.I-2 3.4.1-41 A 140 ValveBody boundary Alloy (External) 141 Valve Body Pressure Copper Alloy Air-outdoor (External) Loss of material External MonitoringSurfaces N/A N/A G boundary Pressure Copper Air-indoor 142 Valve Body Alloy > uncontrolled None None VIII.I-2 3.4.1-41 A Sboundary 15% Zn (Internal) I I I I I _ I

Enclosure L-1 1-252 Page 57 of 108 Table 3.3.2-14 Aging Management Review Results - Fire Protection System Row Component Intended Aging Effect Aging NUREG-Row Type Intended Material Environment Requiring Management 1801, Table No. Type Function(s) Management Maaeet Program PormVolume 2VIteme tm 1 Item Notes 2 Item Prsue Copper Air-outdoor CaknMIselnl Inspection of Sraneos i 143 Valve Body Pressure Alloy > itr Cracking Internal Surfaces in N/A N/A G boundary 15% Zn (internal)Miclaeu 15% ___ZnPiping and Ducting Inspection of Pressure Copper Air-outdoor Loss of Internal Surfaces in 1% Zn

%boundary(Internal) material Miscellaneous N/A N/A G 15%__ZnPiping and Ducting Pressure Copper Air-outdoor Loss of Selective Leaching 145 14PVury alveBody bounary Alloy >

15% Zn (Internal) material Inspection N/A N/A G Pressure Copper Raw water Loss of 146 Valve Body boundary Alloy > (internal) material Fire Water VIIG-12 3.3.1-70 A 15% Zn 147 Valve Body Pressure Copper Raw water Loss of Selective InspectionLeaching VII.G-13 3.3.1-84 A boundary Alloy > (Internal) material 15% Zn Pressure Copper Air-indoor 148 Valve Body Alloy > uncontrolled None None VIII.I-2 3.4.1-41 A boundary 15% Zn (External)

Pressure Copper Air-outdoor External Surfaces 149y boundary Alloy >

15% Zn (External) Cracking Monitoring N/A N/A G Pressure Copper Air-outdoor Loss of External Surfaces 150 Valve 150Body boundary alveBody bounary 15% Zn 15% > (External) material Monitoring N/A N/A G Pressure Copper Air-outdoor Loss of Selective Leaching 151 Valve Body boundary Alloy > (External) material Inspection N/A N/A G 15% Zn

Enclosure L-1 1-252 Page 58 of 108 Table 3.3.2-14 Aging Management Review Results - Fire Protection System Row Component Intended Aging Effect Aging NUREG-Row Cmpnen Fnctiondd Material Environment Requiring Management 1801, Table 1 Notes No. Type Function(s) Maaeet PormVolume Item Management Program 2 Item Pressure Gray Cast Air-indoor Loss of External Surfaces C 152 Valve Body boundary Iron (Internal) material Monitoring VII.I-8 3.3.1-58 0301 153 Valve Pressure Gray Cast Air-outdoor Loss of External Surfaces VII.I-9 3.3.1-58 C boundary Iron (Internal) material Monitoring 0301 154 Valve Body Pressure Gray Cast Raw water Loss of Fire Water VIIG-24 3.3.1-68 A boundary Iron (Internal) material 155 Valve Body Pressure Gray Cast Raw water Loss of Selective Leaching VIIG-14 3.3.1-85 A boundary Iron (Internal) material Inspection Air with 156 Valve Body Pressure Gray VII.l-10 3.3.1-89 boundary Iron Cast borated leakage water Loss of material Boric Acid Corrosion A (External)

Pressure G Ct Air-indoor Loss of External Surfaces 157 Valve Body 157ray boundary Iron Cast uncontrolled (External) material Monitoring materialMonitoring VII.I-8 3.3.1-58 A 158 Valve Body Pressure Gray Iron Cast Air-outdoor Loss of External boundary (External) material MonitoringSurfaces VII.l-9 3.3.1-58 A 159 Valve Body Pressure Gray Iron Cast boundary Soil (External) Loss of Buried Piping and VII.G-25 3.3.1-19 A material Tanks Inspection 160 Valve Body Pressure Gray Cast Soil (External) Loss of Selective Leaching VII.G-15 3.3.1-85 A boundary Iron material Inspection

Enclosure L-1 1-252 Page 59 of 108 Table 3.3.2-14 Aging Management Review Results - Fire Protection System Row Component Intended Aging Effect Aging NUREG-No. Type Function(s) Material Environment Requiring Management Volume Item Notes Management Program 2 Item Item 161 Valve Body Pressure boundary Steel Air-indoor uncontrolled (internal) Loss of mtraMoi material External Surfaces rngVII.I-8 Monitoring 3 3.3.1-58 0301 162 Valve Body Pressure Steel Raw water N/A N/A H boundary (Internal) Cracking TLAA 0339 Pressure Raw water Loss of 163 Valve Body boure Steel (Internal) material Fire Water VII.G-24 3.3.1-68 A boundary(itra) meil Air with 164 Valve Body Pressure boundary Steel borated leakage water Loss of material Boric Acid Corrosion VII.-10 3.3.1-89 A (External) 165 Valve Body Pressure boundary Steel Air-indoor uncontrolled (External) Loss of mtraMoirngVII.I-8 material External MonitoringSurfaces 3.3.1-58 A Structural Copper Air-indoor 166 Valve Body i A uncontrolled None None VIII.I-2 3.4.1-41 A integrity Alloy (Internal) 0301 Structural Copper Raw water Loss of Fire Water VII.G-12 3.3.1-70 A 167 Valve Body integrity Alloy (Internal) material Air-indoor 168 Valve Body Structural Copper uncontrolled None None VIII.I-2 3.4.1-41 A yintegrity Alloy (External)

Air-indoor 00 169 Valve Body Structural integrity Gray Iron Cast (Intemal) uncontrolled Loss of matrrial External MonitoringSurfaces VII.-8 3.3.1-58 A (internal

Enclosure L-1 1-252 Page 60 of 108 Table 3.3.2-14 Aging Management Review Results - Fire Protection System Row Component Intended Aging Effect Aging NUREG-No. Type Function(s) Material Environment Requiring Management V m Tem Notes Management Program Volume Item 2 Item 170 Valve Body Structural Gray integrity Iron Cast (Internal)

Raw water Loss of material Fire Water VIIG-24 3.3.1-68 A 171 Valve Body Structural Gray Cast Raw water Loss of Selective Leaching VIIG-14 3.3.1-85 A integrity Iron (Internal) material Inspection 172 Valve Body Structural integrity Gray Iron Cast Air-indoor uncontrolled Loss of External Surfaces material Monitoring

____________(External)

Fire Pump Diesel Engine and Associated Components Air with steam Pressure Steel or water Cracking Bolting Integrity VII.I-3 3.3.1-41 B 173 Bolting boundary leakage (External)

Air with steam Pressure Steel or water Loss of Bolting Integrity VII.1-6 3.3.1-42 B 174 Bolting boundary leakage material (External)

Air-indoor Loss of Pressure Steel uncontrolled materialBolting Integrity V.-4 3.3.1-43 B 175 Bolting boundary (External) material Air-indoor Loss of 176 Bolting Pressure Steel uncontrolled preload Bolting Integrity VII.I-5 3.3.1-45 B boundary 176Bolting (External) preload 177 Bolting Pressure Steel Air-outdoor Loss of Bolting Integrity VII.I-1 3.3.1-43 B boundary (External) material

Enclosure L-1 1-252 Page 61 of 108 Table 3.3.2-14 Aging Management Review Results - Fire Protection System Row Component Intended AigEfc Aging Effect Aging gn NUREG-81 al Row Cmponen Fnctiond Material Environment Requiring Management 1801, Table 1 Notes No. Type Function(s) Management Program Volume Item

______________ 2 Item 178 Bolting Pressure Steel Air-outdoor Loss of Bolting Integrity N/A N/A H boundary (External) preload Compressor Pressure Air-indoor A 179 Casing- Aluminum uncontrolled None None V.F-2 3.2.1-50 0307 Turbocharger boundary (Internal)

Compressor Pressure Air-indoor 180 Casing - Aluminum uncontrolled None None V.F-2 3.2.1-50 A Turbocharger boundary (External)

Compressor Pressure Air-indoor Loss of External Surfaces C 181 Casing - boundary Steel uncontrolled V1I.1-8 3.3.1-58 0307 Turbocharger- material Monitoring(Int3.15 00 Inspection of Compressor ureessure Diesel exhaust Loss of Internal Surfaces in VII.H2-2 3.3.1-18 E 182 Casing - boundary Steel (Internal) material Miscellaneous Turbocharger Piping and Ducting Compressor Pressure Air-indoor Loss of External Surfaces 183 Casing - boundary Steel uncontrolled VII.I-8 3.3.1-58 A Turbocharger (External) material Monitoring Collection, 18 ile Bd Pressure Raw water SelTreatment Loss of Drainage, and VII.G-24 3.3.1-68E 184 Filter Body boundary Steel (Internal) material Components Inspection 1 5Filter Body Pressure boundary Steel Air-indoor uncontrolled Loss of mtraMoi ExternalrngVII.1-8 Surfaces 3.3.1-58 C 0307 (internal) material Monitoring

Enclosure L-1 1-252 Page 62 of 108 Table 3.3.2-14 Aging Management Review Results - Fire Protection System Row Component Intended AigEfc Aging Effect Aging gn NUREG-81 al Row Cmponen Fnctiond Material Environment Requiring Management 1801, Table I Notes No. Type Function(s) Maaeet PormVolume Item Management Program 2 Item Inspection of Filter Body Pressure Steel Lubricating oil Loss of Internal Surfaces in VII.H2.20 3.3.1-14 E 186 boundary (Internal) material Miscellaneous 0325 Piping and Ducting Pressure Air-indoor Loss of External Surfaces 187 Filter Body boundary Steel uncontrolled L o ter nalSri ae VII.1-8 3.3.1-58 A (External) material Monitoring Flexible Pressure Air-indoor Hardening and External Surfaces 188 Connection boundary uncontrolled loss of MonitoringVIF1-7 3.3.1-11 E Conetin bondr _________ (Internal) strength Moniorin Inspection of Air-indoor Hardening and Internal Surfaces in 189 Flexible Pressure Elastomer uncontrolled loss of Miscellaneous VII.F1-7 3.3.1-11 E Connection boundary (Internal) strength Piping and Ducting Air-indoor Inspection of 190 Flexible Pressure Elastomer Andoor Loss of Internal Surfaces in VIIF16 3.3.1-34 E Connection boundary (interoal) material Miscellaneous Piping and Ducting 191 Flexible Connection Pressure boundary Elastomer Fuel oil (Internal) None None N/A N/A F 192 Flexible Connection Pressure boundary Elastomer Lubricating (Internal) oil None None N/A N/A F Collection, Flexible Pressure Raw water Hardening and Drainage, and 193 Flexib Peure Elastomer Rwater loss of Treatment VII.C1-1 3.3.1-75 E Connection boundary (Interal) strength Components I Inspection

Enclosure L-1 1-252 Page 63 of 108 Table 3.3.2-14 Aging Management Review Results - Fire Protection System Aging Effect Aging NUREG-Row Component Intended Material Environment Requiring Management 1801, Table 1 Notes No. Type Function(s) Malagenent Management Program Program Volume 2 Item Item Collection, Flexible Pressure Raw water Loss of Drainage, and 194 Connection boundary Elastomer (Internal) material Treatment VII.C1-2 3.3.1-75 E Components Inspection Flexible Pressure Air-indoor Hardening and External Surfaces 195 Connection boundary uncontrolled loss of MonitoringVIF1-7 3.3.1-11 (External) strength Inspection of Air-indoor Hardening and Internal Surfaces in 196 Flexible Pressure Elastomer uncontrolled loss of Miscellaneous Piping and Ducting Connection boundary (External) strength Flexible Pressure Air-indoor Loss of External Surfaces 197 Connection boundary Elastomer uncontrolled material Monitoring3.3.1-34 (External) materialMonitoring Inspection of Flexible Pressure Stainless Diesel exhaust Internal Surfaces in VII.H2-1 3.3.1-06 E 198 Connection boundary Steel (Internal) Cracking Miscellaneous Piping and Ducting Flexible Pressure Stainless Diesel exhaust H Connection boundary Steel (Internal) 0337 Inspection of 200 Flexible Pressure Stainless Diesel exhaust Loss of Internal Surfaces in VII.H2-2 3.3.1-18 E Connection boundary Steel (Internal) material Miscellaneous Piping and Ducting 201 Flexible Pressure Stainless Fuel oil Loss of material Fuel Oil Chemistry VIIG-17 3.3.1-32 B Connection boundary Steel (Internal)

Enclosure L-11-252 Page 64 of 108 Table 3.3.2-14 Aging Management Review Results - Fire Protection System Aging Effect Aging NUREG-Row Component Intended Material Environment Requiring Management 1801, Table I Notes No. Type Function(s) Malagenent Manam Volume Item Management Program 2 Item 202 Flexible Pressure Stainless Fuel oil Loss of One-Time VII.G-17 3.3.1-32 A Connection boundary Steel (Internal) material Inspection Collection, Flexible Pressure Stainless Raw water Loss of Drainage, and 203 C Treatment VII.G-19 3.3.1-69 E Connection boundary Steel (Internal) material Components Inspection Flexible Pressure Stainless Air-indoor 204 Flexion Connection boury boundary Steel Steel uncontrolled (Etra)____ None None VII.J-15 3.3.1-94 A (External) 205 Gear Pressure Gray Cast Lubricating oil Loss of Lubricating Oil VILG-22 3.3.1-14 C Housing boundary Iron (Internal) material Analysis 0304 206 Gear Pressure Gray Cast Lubricating oil Loss of One-Time VII.G-22 3.3.1-14 C Housing boundary Iron (Internal) material Inspection Gear Pressure Gray Cast Air-indoor Loss of External Surfaces 207 Gear Peur y Cast uncontrolled material Monitoring3.3.1-58 A Housing boundary Iron (External)

Heat 208 (shell) - Gear Pressure Lubricating oil Loss of Lubricating Oil N/A N/A G Echangear boundary Aluminum (Internal) material Analysis housing oil budr cooler Heat 209 (shell)-ger Pressure Lubricating oil Loss of One-Time N/A N/A G Echangear boundary Aluminum (Internal) material Inspection housing oil cooler_______ ______ _______ ________ __________ ___ __

Enclosure L-11-252 Page 65 of 108 Table 3.3.2-14 Aging Management Review Results - Fire Protection System Aging Effect Aging NUREG-Row Component Intended Aging Magint 1801, Table 1 Notes No. Type Function(s) Material Environment Requiring Management Management Volume Item Program 2 Item Heat Exchanger Pressure Air-indoor 210 (shell) - Gear Aluminum uncontrolled None None V.F-2 3.2.1-50 C housing oil boundary (External) cooler Heat Collection, Exchanger Copper Raw water Reduction in Drainage, and 211 (tubes) - Heat transfer Coy (Inter heationan Treatment VII.C1-6 3.3.1-83 E Gear housing Alloy (Internal) heat transfer Components oil cooler Inspection Heat Exchanger Lubricatin oil Reduction in Lubricating Oil 212 (tubes) - Heat transfer Copper Lubrnal)g heationansfer ricatis V.A-12 3.2.1-09 A Gear housing Alloy (External) heat transfer Analysis oil cooler Heat Exchanger Copper Lubricating oil Reduction in One-Time 213 (tubes) - Heat transfer Coy Lubrnal heationan Inection V.A-12 3.2.1-09 A Gear housing Alloy (External) heat transfer Inspection oil cooler Heat Collection, Exchanger Pressure Copper Raw water Loss of Drainage, and 214 (tubes) - boundary Alloy (internal) material Treatment VII.G-12 3.3.1-70 E Gear housing Components oil cooler Inspection Heat Exchanger Pressure Copper Lubricating oil N 215 (tubes) - boundary Alloy (External) None None VII.G-11 3.3.1-26 0302 Gear housing

___oil cooler I______ I____ I________________I_______

Enclosure L-1 1-252 Page 66 of 108 Table 3.3.2-14 Aging Management Review Results - Fire Protection System Row Component Intended Aging Effect Aging NUREG-Row Type Intended Material Environment Requiring Management 1801, Table 1 Notes No. Type Function(s) Program Volume Item 2 Item Heat Collection, Copper of Drainage, and 216 Exchanger t(shell) r Pressure boundary Alloy 15% Zn> Raw(water Lossmo Treatment Components VII.G-12 3.3.1-70 E Inspection Heat Heat Copper 217 Exchanger Pressure Alloy > Raw water Loss of material Selective Inspection Leaching VII.C1-4 3.3.1-84 A (shell) - boundary 15% Zn (Internal)

Radiator Heat Hean Copper Air-indoor 218 Exchanger Pressure Alloy > uncontrolled None None V.F-3 3.2.1-53 C (shell) - boundary 15% Zn (External)

Radiator Heat Collection, he r Copper Drainage, and 219 Exchanger (tubes) - Heat transfer Alloy 15% Zn> Raw(water (Intemal) Reduction heat in transfer Treatment Components VII.C1-6 3.3.1-83 E Radiator15 nCmoet Inspection Heat Collection, Exchanger Copper Raw water Reduction in Drainage, and 220 (tubes)- Heat transfer Alloy > (Exter heationan Treatment VII.C1-6 3.3.1-83 E (tue 15% Zn (External) heat transfer Components Radiator Inspection Heat Collection, 221 Exchanger PCopper Prsunr Alloyp> Raw water Loss m of Drainage, Treatment and VII.G-12 3.3.1-70 E (tubes) - boundary 15% Zn Components Inspection

Enclosure L-1 1-252 Page 67 of 108 Table 3.3.2-14 Aging Management Review Results - Fire Protection System Row Component Intended Aging Effect Aging NUREG-Row Type Intended Material Environment Requiring Management 1801, Table 1 Notes No. Type Function(s) Management Maaeet Program PormVolume 2VIteme Im Item 2 Item Heat Collection, Copper e Raw water Loss of Drainage, and 222 Exchangure Alloyp> (xtern m Treatment VII.G-12 3.3.1-70 E i(tubes)

- boundaro 15% Zn Components Radiator Inspection Pressure Diesel exhaust TLAA N/A N/A H 0337 223 Piping boundary Steel (internal) Cracking Inspection of 224 Piping Pressure Steel Diesel exhaust Loss of Internal Surfaces in VIIH2-2 3.3.1-18 E boundary (Internal) material Miscellaneous Piping and Ducting 225 Piping Pressure Steel Lubricating oil Loss of Lubricating Oil VII.G-22 3.3.1-14 A boundary (Internal) material Analysis PrsueLubricating oil Loss of One-Time 226 Piping Pressure Steel Lurctn i oso n-ieVII.G-22 3.3.1-14 A boundary (Internal) material Inspection 227 Piping Pressure boundary Steel Air-indoor uncontrolled (External) Loss of mtraMoirngVII.I-8 material External MonitoringSurfaces 3.3.1-58 A 228 Piping Pressure Steel Air-outdoor Loss of External Surfaces VII.l-9 3.3.1-58 A boundary (External) material Monitoring Silencer Pressure Steel DieselCracking TLAA N/A N/A H (exhaust) boundary (Internal) 0337

Enclosure L-1 1-252 Page 68 of 108 Table 3.3.2-14 Aging Management Review Results - Fire Protection System Aging Effect Aging NUREG-Row Component Intended Material Environment Requiring Management 1801, Table 1 Notes No. Type Function(s) Malagenent Manam Volume Item Management Program 2 Item Inspection of 230 Silencer Pressure Diesel exhaust Loss of Internal Surfaces in VII.H2-2 3.3.1-18 E (exhaust) boundary Steel (Internal) material Miscellaneous I_ Piping and Ducting 231 Silencer Pressure SelVII.I-9 Air-outdoor Loss of External Surfaces 3.3.1-58 A (exhaust) boundary (External) material Monitoring 232 Tubing Pressure Stainless Lubricating oil Loss of Lubricating Oil VII.H2-17 3.3.1-33 A boundary Steel (External) material Analysis 233 Tubing Pressure Stainless Lubricating oil Loss of One-Time VI1.12-17 3.3.1-33 A boundary Steel (External) material Inspection Pressure Stainless Air-indoor 234 Tubing Steel uncontrolled None None VII.J-15 3.3.1-94 A g boundary (External) 235 Tubing Pressure Steel Fuel oil (Internal) Loss of material Fuel Oil Chemistry VII.H2-24 3.3.1-20 B boundary Pressure Fuel oil Loss of One-Time Inspection VII.H2-24 3.3.1-20 A 236 Tubing boundary Steel (Internal) material 237 Tubing Pressure Steel Lubricating oil Loss of Lubricating Oil VII.G-22 3.3.1-14 A boundary (Internal) material Analysis 238 Tubing Pressure Steel (Internal) oil Lubricating Loss of material One-Time Inspection VII.G-22 3.3.1-14 A boundary

Enclosure L-11-252 Page 69 of 108 Table 3.3.2-14 Aging Management Review Results - Fire Protection System Row Component Intended Aging Effect Aging NUREG-Row Cmponen Fnctiond Material Environment Requiring Management 1801, Table I Notes No. Type Function(s) Maaeet Management PormVolume Program 2 Item Item Collection, Pressure Raw water Loss of Drainage, and 239 Tubing boundary Steel (Intemal) material Treatment VIsG-24 3.3.1-68 E bounaryIntenal)Components Inspection 240 Tubing Pressure boundary Steel Air-indoor uncontrolled (External) Loss of mtraMoirngVII.I-8 material External MonitoringSurfaces 3.3.1-58 A Collection, 24Pressure aveBd Sel.Treatment Raw water Loss of Drainage, and VII.G-24 3.3.1-68E 241 Valve Body boundary Steel (Internal) material Components Inspection Collection, Pressure Copper Raw water Loss of Drainage, and 242 Valve Body Alloy > (internal) material Treatment VII.G-12 3.3.1-70 E boundary 15% Zn Components Inspection Pressure Copper Raw water Loss of Selective Leaching boundary Alloy > (Internal) material Inspection VI1.Cl-4 3.3.1-84 C 243__ VvBo bodr 15% Zn Pressure Copper Air-indoor 244 Valve Body Alloy > uncontrolled None None V.F-3 3.2.1-53 C boundary 15% Zn (External) 245 Valve Body Pressure Lubricating oil Loss of Lubricating Oil VIIG-22 3.3.1-14 A boundary Steel (Internal) material Analysis

Enclosure L-11-252 Page 70 of 108 Table 3.3.2-14 Aging Management Review Results - Fire Protection System Aging Effect Aging NUREG-Row Component Intended Material Environment R Mt 1801, Table 1 Notes No. Type Function(s) equiring anagement Volume Item Management Program 2 Item Pressure Steel Lubricating oil Loss of One-Time 246 Valve Body boundary (Internal) material Inspection 247 Valve Body Pressure boundary Steel Air-indoor uncontrolled Loss of mtraMoirngVII.I-8 External Surfaces 3.3.1-58 A (External) material Monitoring

Enclosure L-1 1-252 Page 71 of 108 Affected LRA Section LRA Page No. Affected Paragraph and Sentence Table 3.3.2-26 Pages 3.3-473 Rows 72, 73, 79, 80 & 83 thru 3.3-475 In response to Supplemental RAI 3.2.2.2.3.6-2, rows 72, 73, 79, 80 and 83 of LRA Table 3.3.2-26, "Aging Management Review Results - Service Water System," are revised as follows:

Table 3.3.2-26 Aging Management Review Results - Service Water System dAging Effect NUREG-Row Component Intended Material Environment Agn fetAging Agn Management Requiring a a e et 1801, 10 , Table T be1 1 oe Notes No. Type Function(s) Management Program Volume 2 Item Item P40RP Ga~inR pun MSU~eAi-(B nde es6 ef Extamal Surfaceesr 72 P480) Steel .....

!.u..

teiei

!)ll.ed mtra VU! 0307-Not used One Timelnspetion Pump Casing Inspection of Internal

- Dilution Pressure Moist air Loss of in E ESues pump P180) (DB- boundary (Internal) material sce s PMiscellaneous -VI-1.1-P 180) Piping and Ductinc

Enclosure L-1 1-252 Page 72 of 108 Table 3.3.2-26 Aging Management Review Results - Service Water System Aging Effect NUREG-Row Component Intended Material Environment Requiring Aging Management 1801, Table 1 Notes No. Type Function(s) Management Program Volume 2 Item Item

-sep9 4e water-pump esse ...... Et... S--

79 (D9 PS ,-2. Stee UR*e/lQed V Not used one Timeo lnspotien Pump Casing

- Service Pressure Moist air Loss of Inspection of Internal E 80 water pump boundary Steel (Internal) material Surfaces in VII.G-23 3.3.1-71 (DB-P3-1, 2, Miscellaneous

&3) Piping and Ductina One Timeo Inspection Pump Casing

- Service Pressure Moist air Loss of Inspection of Intemal G 83 water pump boundary Steel (Internal) material Surfaces in N/A N/A (DB-P3-1, 2, Miscellaneous

&3) Piping and Ductinq

Enclosure L-1 1-252 Page 73 of 108 Affected LRA Section LRA Page No. Affected Paragraph and Sentence Table 3.3.2-27 Page 3.3-488 Row 38 In response to Supplemental RAI 3.2.2.2.3.6-2, row 38 of LRA Table 3.3.2-27, "Aging Management Review Results

- Spent Fuel Pool Cooling and Cleanup System," is revised as follows:

Table 3.3.2-27 Aging Management Review Results - Spent Fuel Pool Cooling and Cleanup System nAging Effect NUREG-Agen et E leing Aging Management 1801, Table 1 Notes No. Type Function(s) Material Environment Requiring Program Volume 2 Item Management Item one TPrn in~pectien Structural Stainless Moist air Loss of Inspection of Internal G 38 Piping integrity Steel (Internal) material Surfaces in N/A N/A 2 Miscellaneous PipinQ and Ductincq

Enclosure L-11-252 Page 74 of 108 Affected LRA Section LRA Page No. Affected Paragraph and Sentence Table 3.3.2-30 Page 3.3-524 Rows 117 & 120 In response to Supplemental RAI 3.2.2.2.3.6-2, rows 117 and 120 of LRA Table 3.3.2-30, "Aging Management Review Results - Station Blackout Diesel Generator System," are revised as follows:

Table 3.3.2-30 Aging Management Review Results - Station Blackout Diesel Generator System dAging Effect NUREG-Row Component Intended Material Environment Requiring Aging Management 1801, Table 1 Notes No. Type Function(s) Management Program Volume Item 2 Item Jacket wate 117 tank Surface in VII.H2-21 3.3.1-71 Tank- be~nda' mateiaJ Atneean Not used One Tirna iR8poctia Tank - Inspection of internal 10Jacket water expansion Pressure boundary Sel Moist air (Internal) Loss of material Sufcsi i03-13.

Miscellaneous 1.-22 ..- 1E tank Piping and Ductinq

Enclosure L-1 1-252 Page 75 of 108 Affected LRA Section LRA Page No. Affected Paragraph and Sentence Table 3.3.2 Page 3.3-548 Row 0313 Plant-Specific Notes In response to Supplemental RAI 3.2.2.2.3.6-2, row 0313 of Table 3.3.2, "Plant-Specific Notes," is no longer used, and is revised as follows:

Plant-Specific Notes:

0313 The One Time !nspeGtion wi!! con&.'M, for components sUbject to a moist air environmentat the air water inte.face, the absenco o aging ffectS OF that aging i6 slow acting so as to not o uaffct the SUbject Gomponent1' intnPdQd functions dring the p940 extondoed-operation. The aging effectsG above14- and-be-low the _air wapter intedaoce are m~anaged, a6 necossairyL, by thoir repectv Not used

Enclosure L-1 1-252 Page 76 of 108 Affected LRA Section LRA Page No. Affected Paragraph and Sentence Table 3.4.2-2 Pages 3.4-54 Rows 8 & 11 and 3.4-55 In response to Supplemental RAI 3.2.2.2.3.6-2, rows 8 and 11 of LRA Table 3.4.2-2, "Aging Management Review Results - Condensate Storage System," are revised as follows:

Table 3.4.2-2 Aging Management Review Results - Condensate Storage System Aging Effect NUREG-Row Component Intended Material Environment Requiring Aging Management 1801, Table 1 Notes No. Type Function(s) Management Program Volume Item 2 Item One Time !nRpoct@on Tank -

Condensate Pressure Moist air Loss of Inspection of Internal G 8 storage tanks boundary Steel (Internal) material Surfaces in N/A N/A (DB-T31-1 & Miscellaneous

2) Piping and Ducting Gend4ensate sto~Etoma tan Afacde 11 B4...... Stee/ U -- ## - 3.4. -2 Ntuemase Not used

Enclosure L-11-252 Page 77 of 108 Affected LRA Section LRA Page No. Affected Paragraph and Sentence Table 3.4.2 Page 3.4-111 Row 0404 Plant-Specific Notes In response to Supplemental RAI 3.2.2.2.3.6-2, row 0404 of Table 3.4.2, "Plant-Specific Notes," is no longer used, and is revised as follows:

Plant-Specific Notes:

0404 The Ono T4'ne iR8pectin WXl GOnfRm, for cOMPOnents subject to a moeist air-environment at the air water- inteiface, the absenceo aging effectS or-that aging is slow acting so as not to affect the FuUbject components' inended fu-nctions di ring the pwerid of extended operation. The aging effectS above and belowth ai wte inteffa-e-are mnanagedý, as necegsar, by theirrespectv Not used

Enclosure L-1 1-252 Page 78 of 108 Affected LRA Section LRA Page No. Affected Paragraph and Sentence A.1.30 Page A-20 Second paragraph, last sentence In response to Supplemental RAI 3.2.2.2.3.6-2, the second paragraph of LRA Section A.1.30, "One-Time Inspection," is revised as follows:

A.1.30 ONE-TIME INSPECTION One-Time Inspection also provides assurance that aging which has not yet manifested itself is indeed not occurring, or that the age-related degradation is so insignificant that an aging management program is not warranted. ne.Gcti.en at air wa1togr interfacois providos.onfirm.ation fuFhor that deg.adati. n is n o,,uFng at leoations. where a potential exists fr Gonti to duo1R tio al-ternaP.RtO 14ettWn and drying.

Enclosure L-1 1-252 Page 79 of 108 Affected LRA Section LRA Page No. Affected Paragraph and Sentence A.1.32 Page A-21 Entire section In response to RAI B.2.32-1, LRA Section A.1.32, "PWR Reactor Vessel Internals Program," is replaced in its entirety. LRA Section A.1.32 reads as follows:

A.1.32 PWR REACTOR VESSEL INTERNALS PROGRAM The PWR Reactor Vessel Internals Program relies on implementation of the Electric Power Research Institute (EPRI) Report No. 1016596, "Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227)," and EPRI Report No. 1016609, "Materials Reliability Program: Inspection Standard for PWR Internals (MRP-228)," to manage the aging effects on the reactor vessel internal (RVI) components.

This program is used to manage the effects of age-related degradation mechanisms that are applicable in general to the PWR RVI components at Davis-Besse, a Babcock & Wilcox (B&W) designed plant. These aging effects include (a) various forms of cracking, including stress corrosion cracking (SCC),

which also encompasses primary water stress corrosion cracking (PWSCC),

irradiation-assisted stress corrosion cracking (IASCC), or cracking due to fatigue/cyclical loading; (b) loss of material induced by wear; (c) loss of fracture toughness due to either thermal aging or neutron irradiation embrittlement; and (d) loss of preload due to thermal and irradiation-enhanced stress relaxation or creep. In addition, the program includes management of the time-limited aging analysis (TLAA) identified in License Renewal Application (LRA) Section A.2.2.7 for reduction in fracture toughness of the reactor vessel internals. This TLAA will be managed in accordance with the implementation of the MRP-227 guidelines, as amended by the MRP-227 safety evaluation, including all activities associated with Davis-Besse's responses to plant-specific action items identified in Section 4.2 of the safety evaluation.

The program applies the guidance in MRP-227, Rev. 0, as amended by the safety evaluation for inspecting, evaluating, and, if applicable, dispositioning non-conforming RVI components at Davis-Besse. The program conforms to the definition of a sampling-based condition monitoring program, as defined by the Branch Technical Position RSLB-1, with periodic examinations and other inspections of highly-affected internals locations. These examinations provide reasonable assurance that the effects of age-related degradation mechanisms will be managed during the period of extended operation. The program includes expanding periodic examinations and other inspections if the extent of the degradation effects exceeds the expected levels.

Enclosure L-1 1-252 Page 80 of 108 The MRP-227 guidance for selecting RVI components for inclusion in the inspection sample is based on a four-step ranking process. Through this process, the reactor internals were assigned to one of the following four groups: Primary, Expansion, Existing Programs, and No Additional Measures components.

Definitions of each group are provided in GALL Chapter IX.B.

The result of this four-step sample selection process is a set of Primary Internals Component locations for each of the three plant designs (Westinghouse, Combustion Engineering and Babcock & Wilcox) that are expected to show the leading indications of the degradation effects, with another set of Expansion Internals Component locations that are specified to expand the sample should the indications be more severe than anticipated. The degradation effects in a third set of internals locations are deemed to be adequately managed by Existing Programs. A fourth set of internals locations are deemed to require no additional measures. As a result, the program typically identifies 5 to 15 percent of the RVI locations as Primary Component locations for inspections, with another 7 to 10 percent of the RVI locations to be inspected as Expansion Components, as warranted by the evaluation of the inspection results. Another 5 to 15 percent of the internals locations are covered by Existing Programs, with the remainder requiring no additional measures. This process thus uses appropriate component functionality criteria, age-related degradation susceptibility criteria, and failure consequence criteria to identify the components that will be inspected under the program in a manner that conforms to the sampling criteria for sampling-based condition monitoring programs in Section A.1.2.3.4 of NRC Branch Position RLSB-1. Consequently, the sample selection process is adequate to assure that the intended function(s) of the PWR reactor internal components are maintained during the period of extended operation.

No existing generic industry programs contain the specificity considered sufficient for monitoring the aging effects addressed by the MRP-227 guidelines for B&W plants. Therefore, no components for B&W plants were placed into the Existing Programs group.

MRP-227 I&E guidelines require a visual (VT-3) examination of the core support shield (CSS) vent valve retaining rings and disc shaft for every 10 year Inservice Inspection Interval. In addition, Davis-Besse Technical Specification 5.5.4 requires testing of the CSS vent valves every 24 months to verify by visual inspection that the valve body and valve disc exhibit no abnormal degradation, verify the valve is not stuck in an open position, and verify by manual actuation that the valve is fully open when a force of <400 lbs is applied vertically upward.

The technical specification inspection will continue to be performed at the prescribed frequency of 24 months. The MRP-227 required visual (VT-3) examination will also be performed at the prescribed frequency of every 10 year Inservice Inspection Interval.

Enclosure L-1 1-252 Page 81 of 108 The program's use of visual examination methods in MRP-227 for detection of relevant conditions (and the absence of relevant conditions as a visual examination acceptance criterion) is consistent with the ASME Code,Section XI rules for visual examination. However, the program's adoption of the MRP-227 guidance for visual examinations goes beyond the ASME Code,Section XI visual examination criteria because additional guidance is incorporated into MRP-227 to clarify how the particular visual examination methods will be used to detect relevant conditions and describes in more detail how the visual techniques relate to the specific RVI components and how to detect their applicable age-related degradation effects.

The technical basis for detecting relevant conditions using volumetric ultrasonic testing (UT) inspection techniques can be found in MRP-228, where the review of existing bolting UT examination technical justifications has demonstrated the indication detection capability of at least two vendors, and where vendor technical justification is a requirement prior to any additional bolting examinations. Specifically, the capability of program's UT volumetric methods to detect loss of integrity of PWR internals bolts, pins, and fasteners, such as baffle-former bolting in B&W and Westinghouse units, has been well demonstrated by operating experience. In addition, the program's adoption of the MRP-227 guidance and process incorporates the UT criteria in MRP-228, which calls for the technical justifications that are needed for volumetric examination method demonstrations, required by the ASME Code,Section V.

The program also includes future industry operating experience as incorporated in periodic revisions to MRP-227. The program thus provides reasonable assurance for the long-term integrity and safe operation of reactor internals in all commercial operating U.S. PWR nuclear power plants.

Age-related degradation in the reactor internals is managed through an integrated program. Specific features of the integrated program are listed in the following ten program elements. Degradation due to changes in material properties (e.g., loss of fracture toughness) was considered in the determination of inspection recommendations and is managed by the requirement to use appropriately degraded properties in the evaluation of identified defects. The integrated program is implemented by the applicant through an inspection plan.

The Davis-Besse PWR Reactor Vessel Internals Program will address all plant-specific action items applicable to Davis-Besse that are established in Section 4.2 of the safety evaluation for MRP-227. In addition, a plant-specific inspection plan for ensuring the implementation of MRP-227 program guidelines, as amended by the safety evaluation for MRP-227, and Davis-Besse's responses to the plant-specific action items, as identified in Section 4.2 of the safety evaluation for MRP-227, will be submitted for NRC review and approval.

Enclosure L-1 1-252 Page 82 of 108 Affected LRA Section LRA Page No. Affected Paragraph and Sentence Table A-1 Page A-63 Commitment 13 Based on the response to Supplemental Question - Makeup Pump Casing Inspections, license renewal future Commitment 13 in LRA Table A-i, "Davis-Besse License Renewal Commitments," is revised to read as follows:

Table A-1 Davis-Besse License Renewal Commitments e IRelated LRA Item Commitment Implementation Source Section No./

Number Schedule Cmet Comments 13 Implement the One-Time Inspection as described in LRA Section Prior to LRA A.1.30 B.2.30. Enhance the One-Time Inspection to: April 22, 2017 and B.2.30 Include enhanced visual (EVT-1 or equivalent) or surface FENOC Response to examination (magneticparticle, liquid penetrant), or volumetric Letters NRC RAI (RT or UT) inspections enhanced visual (VT I or eguivaionq L-11-166, 3.3.2.2.4.3-I and/or volumotric (RT or UD inspections to detect and aRd from characterizecracking due to cyclic loading of the stainless L-11-237, NRC Letter steel makeup pump casings (DB-P37-1 and 2) of the Makeup and dated and PurificationSystem. The one-time inspections will provide L-11-252 May 2, 2011 verification of the absence of cracking due to cyclic loading, and Supplemental Question -

Makeup Pump Casing Inspections

Enclosure L-11-252 Page 83 of 108 Affected LRA Section LRA Page No. Affected Paragraph and Sentence Table A-1 Page A-63 Commitment 15 In response to RAI B.2.32-1, LRA Table A-1, "Davis-Besse License Renewal Commitments," license renewal future Commitment 15 is revised to address submittal of a reactor vessel internals inspection plan.

LRA Table A-i, Commitment 15, is revised to read as follows:

Table A-1 Davis-Besse License Renewal Commitments T

Implementation Source Related LRA Section No./

Item Number Commitment Schedule comments Comments 15 Ro*,So the PWR Rorat, VelS InteMnaS Progam, aS n..... Sa..r. Followin N.- LRA A.1 .32 tO inGnrporato the final rFcm..R.ndations and roqui.rm.nts a5 and B.2.32 published in MRP-227 AU.RP227 n In association with the PWR Reactor Vessel Internals Program,a guide F plant-specific inspection plan for ensuring the implementation of as FENOC Response to MRP-227 program guidelines, as amended by the safety evaluation Letter NRC RAI for MRP-227, and Davis-Besse's responses to the plant-specific Priorto L-11-252 B.2.32-1 from action items, as identified in Section 4.2 of the safety evaluation for April 22, 2015

NRC Letter MRP-227, will be submitted for NRC review and approval, dated

NOTE: The inspection plan will be submitted no laterthan two July 11, 2011 years after issuance of the renewed operatinglicense or two years prior to the beginning of the period of extended operation (April 22, 2015), whichever is earlier.

Enclosure L-1 1-252 Page 84 of 108 Affected LRA Section LRA Page No. Affected Paragraph and Sentence Table A-1 Page A-69 Commitment 26 The NRC initiated a telephone conference call with FENOC on September 7, 2011, to discuss the FENOC supplemental response to RAI 2.1-3 (submitted in FENOC letter dated August 17, 2011 (ML11231A966)) regarding abandoned equipment. The NRC staff requested discussion on the FENOC implementation schedule regarding actions to address abandoned equipment. Following discussion of the FENOC plan, NRC staff agreed that the proposed plan was acceptable and requested that the plan be submitted in the next RAI response letter. Based on the proposed schedule for the plans to address abandoned equipment, it was mutually agreed that license renewal future Commitment 26 on this topic was no longer necessary and would be deleted; NRC plans to treat this issue as an Open Item in the License Renewal Safety Evaluation Report for Davis Besse. LRA Table A-i, "Davis-Besse License Renewal Commitments," license renewal future Commitment 26, is revised to read as follows:

Table A-1 Davis-Besse License Renewal Commitments Item Implementation Source Related LRA Number Commitment ScheduleSection Comments No.

26 Ensure that abandoned equipmeont is identifed, and either-isolate Prior tA FENO~ SpplementaJ and drained.Or inclded ,..ithin the... c...pe of .cens re.newal

. and Deeer-,-m , LetrF k-i Fe6pese-t isubje.t to a-gin mn t-raview. 24-2 I-4-1-288 NRl RA.I NRG LeteF Not used 4eig 204*,4,4

Enclosure L-11-252 Page 85 of 108 Affected LRA Section LRA Page No. Affected Paragraph and Sentence Table A-1 Page A-69 Commitment 33 License renewal future Commitment 33 regarding examination of the Containment Vessel in the sand pocket region is replaced in its entirety based on the response to RAI B.2.39-9, and LRA Table A-I, "Davis-Besse License Renewal Commitments," Commitment 33, is revised to read as follows:

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

Number Schedule Cmet Comments 33 Phase 1 Phase 1: FENOC Response to Perform the following actions to reduce or mitigate the refueling Action 1 prior to Letter NRC RAI canal leaks inside containment: December 31, 2014 L-11-252 B.2.39-9 from

1. Select and implement a leak detection method to locate the NRC Letter leaka-ge area. dated Action 2 prior to July 27, 2011

2. Evaluate temporaryand permanent repairmethods to stop December 31, 2016 or siqnificantly reduce the leaka-ge, and implement a repairplan.

Phase 2 Perform the followinq actions to evaluate the impact of refuelin Phase 2:

canal leaks on concrete and reinforcinqsteel structures. Action I priorto Discontinue core bores, testing and reinforcing steel inspections December 31, 2014 when indicationsof refueling canal leakage are no longer present:

1. Perform a core bore in the south wall of the east-west

Enclosure L-11-252 Page 86 of 108 Table A-1 Davis-Besse License Renewal Commitments Item IImplementation Related LRA Iter Commitment ISheme Source Section No./

NumberSch ScheduleComments section of the core flood pipe tunnel Action 2 prior to

a. Assess borated water degradationof the concrete by December 31, 2023 testing the core bore sample for compressive strength and by petrographicexamination, and evaluate the results. Action 3 - Ongoing

b. Conduct a visual examination of the concrete and reinforcing steel to identify aging effects (e.g.,

concrete degradationor steel corrosion). Enter identified aging effects into the FENOC Corrective Action Programand evaluate in accordance with the requirements of the currentlicensing basis Maintenance Rule Program.

2. If leakage from the refueling canal has not been eliminated or resumes by the beginning of the period of extended operation, then evaluate the concrete structuresin a manner similarto the way that they were evaluated under Phase 2, Action 1. However, use acceptance criteriafrom the American Concrete Institute (ACI) Report 349.3R for the evaluation.

3. If leakage from the refueling canal has not been eliminated or resumes during the period of extended operation, then evaluate the concrete structures again in a manner similarto the way that they were evaluated under Phase 2. Action 2.

Perform evaluations every ten years until the end of the nerind of ertended nnerntinn_

. . . . of . extended_____

. . . . .. ______ I_________

Enclosure L-1 1-252 Page 87 of 108 Affected LRA Section LRA Page No. Affected Paragraph and Sentence Table A-1 Page A-69 Commitment 35 License renewal future Commitment 35 regarding examination of the Containment Vessel in the sand pocket region is replaced in its entirety based on the response to RAI B.2.22-5, and LRA Table A-I, "Davis-Besse License Renewal Commitments," Commitment 35, is revised to read as follows:

Table A-1 Davis-Besse License Renewal Commitments Item TImplementation Source Related LRA Section No./

Number Commitment Schedule Comments 35 Perform the following actions for each of two examinations Phase 1 priorto FENOC Response to (Phase I and Phase 2) of the Containment Vessel in the sand December 31, Letter NRC RAI pocket region: 2014 L-11-252 B.2.22-5 from NRC Letter dated

. Perform nondestructive examination (NDE) of the and Phase 2 prior to July 21. 2011 Containment Vessel from the outer surface at five areas of December 31, previously-identified groundwaterin-leakage.

o Examine the vessel at a minimum of three vertical grid 2025 locations at 12" nominal horizontal spacing at each area.

Examine the Containment Vessel at a minimum of three elevations:

1. approximately 3 inches below the existing grout-to-vessel interface in the sand pocket region;

2. at the existing grout-to-vessel interface level in the sand pocket region: and,

3. approximately 3 inches above the existing

Enclosure L-1 1-252 Page 88 of 108 Table A-1 Davis-Besse License Renewal Commitments CImplementation Related LRA Item Commitment Sheme Source Section No./

Number Comments grout-to-vessel interface in the sand pocket region.

Compare the ultrasonictesting (UT) thickness readings to minimum ASME Code vessel thickness requirements and to the results obtained during previous UT thickness examinationsof the Containment Vessel. Determine the need for maintenance or repairof the Containment Vessel based on the results and evaluation of the examinations.

Document the results of each of the two examinations in the work order system. Document and evaluate adverse conditions in accordancewith the FENOC Corrective Action Programfor an evaluation of potential degradationof the steel Containment Vessel thickness over the longer term.

Enclosure L-11-252 Page 89 of 108 Affected LRA Section LRA Page No. Affected Paragraph and Sentence Table A-1 Page A-69 Commitment 36 License renewal future Commitment 36 regarding inspection of the Containment Vessel in the sand pocket region is replaced in its entirety based on the response to RAI B.2.22-5, and LRA Table A-1, "Davis-Besse License Renewal Commitments," Commitment 36, is revised to read as follows:

Table A-1 Davis-Besse License Renewal Commitments e IRelated LRA Item Commitment Implementation Source Section No./

Number Schedule Cmet Comments 36 Perform the following actions related to the Containment Vessel Ongoing FENOC Response to sand pocket region each refueling outage: Letter NRC RAI L- 11-252 B. 2.22-5 from

Perform visual inspection of 100 percent of the accessible areas NRC Letter dated of the wetted outer surface of the Containment Vessel in the Julyd21 2011 sand pocket region.

Perform visual inspection of accessible dry areasof the outer surface of the Containment Vessel in the sand pocket region and the areas above the grout-to-steel interface up to Elevation 566' + 3", - I".

" Perform visual inspection for deterioration(e.g., missing or damaged grout) of accessible grout in the sand pocket area.

" Perform opportunistic visual inspections of inaccessible areas of the Containment Vessel in the sand pocket region when such areasare made accessible.

" Perform opportunisticvisual inspections for deterioration(e.g.,

Enclosure L-11-252 Page 90 of 108 Table A-1 Davis-Besse License Renewal Commitments Item Number Commitment To Implementation Source Related LRA Section No./

Comments missing or damaged grout)of inaccessible groutin the sand Docket region when such areas are made accessible.

Inaccessiblegrout is the grout below the normally-exposed surface of the grout in the sand Docket area.

Address issues of pitting. microbiologically-influenced corrosion (MIC), degradedgrout, moisture barrieror sealant identified during the inspections using the FENOC Corrective Action Program.

Sample the water in the sand Docket region when sufficient volumes are available. The number of sampled water volumes will be determined by the number of water volumes observed and the size of those water volumes. Analyze the sample(s) for pH, chlorides, iron and sulfates. Treat or wash (ora combination thereof) the sand Docket area to reduce measured chloride concentrationsto less than 250 parts per million (ppm) if the concentration of chloridesin a sample exceeds 250 ppim.

Note: Water samples may be taken at different times during each outage. Engineeringiudqment may be used to determine the priority of the chemical analyses to be performed if sufficient water is not available in a given sample for all analyses.

Enclosure L-1 1-252 Page 91 of 108 Affected LRA Section LRA Parge No. Affected ParagraDh and Sentence Table B-1 Page B-12 Row XI.M16 In response to RAI B.2.32-1, row XI.M16 of Table B-i, "Correlation of NUREG-1801 and Davis-Besse Aging Management Programs," is revised to read as follows:

Table B-1 Correlation of NUREG-1801 and Davis-Besse Aging Management Programs (continued)

Number NUREG-1801 Program Corresponding Davis-Besse AMP X1 A4- PWR Vessel Internals Plant specii" aging management program ir dInted fG:

XI.M16A (NUREG-1801, Rev. 2) r"anagemRnt aginguPWR Reactor Vessel Internals Program (See Section B.2.32).

Enclosure L-1 1-252 Page 92 of 108 Affected LRA Section LRA Page No. Affected Paragraph and Sentence Table B-2 Page B-22 2 Rows (as listed)

In response to RAI B.2.32-1, the "PWR Reactor Vessel Internals Program" row of Table B-2, "Consistency of Davis-Besse Aging Management Programs with NUREG-1801 ," is revised to read as follows:

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

Consistent Consistent wt New /

Existing with NUEG Plant- Enhancement NUREG- 1801 with Specific Required 1801 181wt Exceptions PWR Reactor Vessel Internals Program New Yes --

Section B.2.32 In response to RAI B.2.34-2, the "Reactor Head Closure Studs Program" row of Table B-2 now shows a program exception, and is revised to read as follows:

Reactor Head Closure Studs Program Existing - Yes -- Yes Section B.2.34

Enclosure L-1 1-252 Page 93 of 108 Affected LRA Section LRA Page No. Affected Paragraph and Sentence B.2.9 Pages B-47 Scope of Program, 1 st paragraph, thru B-49 last sentence; and Acceptance Criteria, 2 "dparagraph In response to Supplemental RAI 3.3.2.2.5-2, the "Scope of Program," first paragraph, and "Acceptance Criteria," second paragraph, of LRA Section B.2.9, "Collection, Drainage, and Treatment Components Inspection Program," are revised to read as follows:

B.2.9 COLLECTION, DRAINAGE, AND TREATMENT COMPONENTS INSPECTION PROGRAM Aging Management Program Elements Scope of Program The scope of the Collection, Drainage, and Treatment Components Inspection Program includes visual inspections of the internal surfaces of copper alloy (including copper alloy > 15% Zn), gray cast iron, stainless steel (including cast austenitic stainless steel), and steel components exposed to untreated water, in collection, drainage, or treatment service, that are not covered by other aging management programs. 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. The scope of the program also includes visual inspection and physical (manipulation or prodding) examination of the internal surfaces of flexible elastomeric connections exposed to untreated (raw) water service.

Acceptance Criteria iFor metallic components, unacceptable inspection findings will include visible evidence of cracking, loss of material,or reduction in heat transferdue to fouling that could lead to loss of component intended function during the period of extended operation. Forelastomeric components, unacceptable inspection findinqs will include visual evidence of surface deqradation, such as crackinq, loss of materialor discoloration,or physical evidence of hardeninqand loss of strenqth identified throuqh manipulation or Droddinq.

Enclosure L-1 1-252 Page 94 of 108 Affected LRA Section LRA Page No. Affected Paragraph and Sentence B.2.30 Pages B-120 "Program Description," 3 rd paragraph, and B-124 last sentence Aging Management Program Element:

"Operating Experience," 1 st paragraph, last sentence In response to Supplemental RAI 3.2.2.2.3.6-2, the Program Description, third paragraph, and the Operating Experience program element, first paragraph, of LRA Section B.2.30, "One-Time Inspection," are revised as follows:

B.2.30 ONE-TIME INSPECTION Program Description One-Time Inspection will provide assurance that aging which has not yet manifested itself is indeed not occurring, or that the age-related degradation is so insignificant that an aging management program is not warranted. An.seotien-at air water- intod-acea provides fuwher confirmation that do gradastion i-s not ocrinat loations whoVe a petentia! exists for Gontaminants to concentrato due to altorxnate wettingan yn.

Aging Management Program Elements 0 Operating Experience Operating experience for select components and environments within the scope of One-Time Inspection was evaluated to ensure use of a one-time inspection was appropriate. Review of Davis-Besse operating experience did not identify any instances of degradation that were caused by an ineffective chemistry program. As such, One-Time Inspection is credited to verify the effectiveness. Review of Davis Bosse .p..ating pe did not id.-ntAf ayaig effects that wora aftributod to-air: water inted~acos (water- line a*.l,. /As such, One Time .n pe.tion is croditd to voify that aging is not occurin or i.s oc.U.g very slwly at air water- intedas and gcmponen funtio i mantane thoug te period of extended operation.

Enclosure L-1 1-252 Page 95 of 108 Affected LRA Section LRA Page No. Affected Paragraph and Sentence B.2.30 Page B-121 "Scope" subsection of the "Enhancements" section, last paragraph Based on the response to Supplemental Question - Makeup Pump Casing Inspections, the last paragraph of the "Scope" subsection of the "Enhancements" section of Section B.2.30 is revised as follows:

The One-Time Inspection will also include enhanced visual (EVT-1 or equivalent) or surface examination (maqnetic particle,liquid penetrant), or volumetric (RT or UT) inspections ""hanc.d v.iuaI (VT '1Or ouialnyt) and"/o volumetric (RT or UT) ing"p",,,n to detect and characterizecracking due to cyclic loading of the stainless steel makeup pump casings (DB-P37-1 and 2) of the Makeup and PurificationSystem. The one-time inspections will provide verification of the absence of cracking due to cyclic loading.

Enclosure L-1 1-252 Page 96 of 108 Affected LRA Section LRA Page No. Affected Paragraph and Sentence B.2.32 Pages B-129 Entire section thru B-133 In response to RAI B.2.32-1, LRA Section B.2.32, "PWR Reactor Vessel Internals Program," subsections "Program Description," is replaced in its entirety, and includes the details for each of the 10 Aging Management Program Elements for clarity. LRA Section B.2.32 reads as follows:

B.2.32 PWR REACTOR VESSEL INTERNALS PROGRAM Program Description The PWR Reactor Vessel Internals Program relies on implementation of the Electric Power Research Institute (EPRI) Report No. 1016596, "Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227)," and EPRI Report No. 1016609, "Materials Reliability Program: Inspection Standard for PWR Internals (MRP-228)," to manage the aging effects on the reactor vessel internal (RVI) components.

This program is used to manage the effects of age-related degradation mechanisms that are applicable in general to the PWR RVI components at Davis-Besse, a Babcock & Wilcox (B&W) designed plant. These aging effects include (a) various forms of cracking, including stress corrosion cracking (SCC),

which also encompasses primary water stress corrosion cracking (PWSCC),

irradiation-assisted stress corrosion cracking (IASCC), or cracking due to fatigue/cyclical loading; (b) loss of material induced by wear; (c) loss of fracture toughness due to either thermal aging or neutron irradiation embrittlement; and (d) loss of preload due to thermal and irradiation-enhanced stress relaxation or creep. In addition, the program includes management of the time-limited aging analysis (TLAA) identified in License Renewal Application (LRA) Section 4.2.7 for reduction in fracture toughness of the reactor vessel internals. This TLAA will be managed in accordance with the implementation of the MRP-227 guidelines, as amended by the MRP-227 safety evaluation, including all activities associated with Davis-Besse's responses to plant-specific action items identified in Section 4.2 of the safety evaluation.

The program applies the guidance in MRP-227, Rev. 0, as amended by the safety evaluation for inspecting, evaluating, and, if applicable, dispositioning non-conforming RVI components at Davis-Besse. The program conforms to the definition of a sampling-based condition monitoring program, as defined by the Branch Technical Position RSLB-1, with periodic examinations and other

Enclosure L-1 1-252 Page 97 of 108 inspections of highly-affected internals locations. These examinations provide reasonable assurance that the effects of age-related degradation mechanisms will be managed during the period of extended operation. The program includes expanding periodic examinations and other inspections if the extent of the degradation effects exceeds the expected levels.

The MRP-227 guidance for selecting RVI components for inclusion in the inspection sample is based on a four-step ranking process. Through this process, the reactor internals were assigned to one of the following four groups: Primary, Expansion, Existing Programs, and No Additional Measures components.

Definitions of each group are provided in GALL Chapter IX.B.

The result of this four-step sample selection process is a set of Primary Internals Component locations for each of the three plant designs (Westinghouse, Combustion Engineering and Babcock & Wilcox) that are expected to show the leading indications of the degradation effects, with another set of Expansion Internals Component locations that are specified to expand the sample should the indications be more severe than anticipated. The degradation effects in a third set of internals locations are deemed to be adequately managed by Existing Programs. A fourth set of internals locations are deemed to require no additional measures. As a result, the program typically identifies 5 to 15 percent of the RVI locations as Primary Component locations for inspections, with another 7 to 10 percent of the RVI locations to be inspected as Expansion Components, as warranted by the evaluation of the inspection results. Another 5 to 15 percent of the internals locations are covered by Existing Programs, with the remainder requiring no additional measures. This process thus uses appropriate component functionality criteria, age-related degradation susceptibility criteria, and failure consequence criteria to identify the components that will be inspected under the program in a manner that conforms to the sampling criteria for sampling-based condition monitoring programs in Section A.1.2.3.4 of NRC Branch Position RLSB-1. Consequently, the sample selection process is adequate to assure that the intended function(s) of the PWR reactor internal components are maintained during the period of extended operation.

No existing generic industry programs contain the specificity considered sufficient for monitoring the aging effects addressed by the MRP-227 guidelines for B&W plants. Therefore, no components for B&W plants were placed into the Existing Programs group.

The program's use of visual examination methods in MRP-227 for detection of relevant conditions (and the absence of relevant conditions as a visual examination acceptance criterion) is consistent with the ASME Code, Section Xl rules for visual examination. However, the program's adoption of the MRP-227 guidance for visual examinations goes beyond the ASME Code, Section Xl visual examination criteria because additional guidance is incorporated into MRP-227 to

Enclosure L-1 1-252 Page 98 of 108 clarify how the particular visual examination methods will be used to detect relevant conditions and describes in more detail how the visual techniques relate to the specific RVI components and how to detect their applicable age-related degradation effects.

The technical basis for detecting relevant conditions using volumetric ultrasonic testing (UT) inspection techniques can be found in MRP-228, where the review of existing bolting UT examination technical justifications has demonstrated the indication detection capability of at least two vendors, and where vendor technical justification is a requirement prior to any additional bolting examinations. Specifically, the capability of program's UT volumetric methods to detect loss of integrity of PWR internals bolts, pins, and fasteners, such as baffle-former bolting in B&W and Westinghouse units, has been well demonstrated by operating experience. In addition, the program's adoption of the MRP-227 guidance and process incorporates the UT criteria in MRP-228, which calls for the technical justifications that are needed for volumetric examination method demonstrations, required by the ASME Code,Section V.

The program also includes future industry operating experience as incorporated in periodic revisions to MRP-227. The program thus provides reasonable assurance for the long-term integrity and safe operation of reactor internals in all commercial operating U.S. PWR nuclear power plants.

Age-related degradation in the reactor internals is managed through an integrated program. Specific features of the integrated program are listed in the following ten program elements. Degradation due to changes in material properties (e.g., loss of fracture toughness) was considered in the determination of inspection recommendations and is managed by the requirement to use appropriately degraded properties in the evaluation of identified defects. The integrated program is implemented by the applicant through an inspection plan.

The Davis-Besse PWR Reactor Vessel Internals Program will address all plant-specific action items applicable to Davis-Besse that are established in Section 4.2 of the safety evaluation for MRP-227. In addition, a plant-specific inspection plan for ensuring the implementation of MRP-227 program guidelines, as amended by the safety evaluation for MRP-227, and Davis-Besse's responses to the plant-specific action items, as identified in Section 4.2 of the safety evaluation for MRP-227, will be submitted for NRC review and approval.

NUREG-1801 Consistency The PWR Reactor Vessel Internals Program is a new Davis-Besse program that will be consistent with the 10 elements of an effective aging management program as described in NUREG-1801, Rev. 2,Section XI.M16A, "PWR Vessel Internals." The results of an evaluation for each element are provided below.

Enclosure L-1 1-252 Page 99 of 108 Exceptions to NUREG-1801 None.

Enhancements None.

Aging Management Program Elements The results of an evaluation of each program element are provided below.

Scope The scope of the program includes all RVI components at Davis-Besse, which is built to a B&W NSSS design. The scope of the program applies the methodology and guidance in the most recently NRC-endorsed version of MRP-227, which provides augmented inspection and flaw evaluation methodology for assuring the functional integrity of safety-related internals in commercial operating U.S. PWR nuclear power plants designed by B&W, CE, and Westinghouse. The scope of components considered for inspection under MRP-227 guidance includes core support structures (typically denoted as Examination Category B-N-3 by the ASME Code, Section Xl), those RVI components that serve an intended license renewal safety function pursuant to criteria in 10 CFR 54.4(a)(1), and other RVI components whose failure could prevent satisfactory accomplishment of any of the functions identified in 10 CFR 54.4(a)(1)(i), (ii), or (iii). The scope of the program does not include consumable items, such as fuel assemblies, reactivity control assemblies, and nuclear instrumentation, because these components are not typically within the scope of the components that are required to be subject to an aging management review (AMR), as defined by the criteria set in 10 CFR 54.21(a)(1). The scope of the program also does not include welded attachments to the internal surface of the reactor vessel because these components are considered to be ASME Code Class 1 appurtenances to the reactor vessel and are adequately managed in accordance with an applicant's aging management program (AMP) that corresponds to GALL AMP XI.M1, "ASME Code, Section Xl Inservice Inspection, Subsections IWB, IWC, and IWD."

In addition, the scope of the program includes management of the time-limited aging analysis (TLAA) identified in LRA Section 4.2.7 for reduction in fracture toughness of the reactor vessel internals. This TLAA will be managed in accordance with the implementation of the MRP-227 guidelines, as amended by the MRP-227 safety evaluation, including all activities associated with Davis-Besse's responses to plant-specific action items identified in the Section 4.2 of the safety evaluation.

Enclosure L-1 1-252 Page 100 of 108 The scope of the program includes the response bases to applicable license renewal applicant action items (LRAAIs) on the MRP-227 methodology, and any additional programs, actions, or activities that are discussed in these LRAAI responses and credited for aging management of the applicant's RVI components. The LRAAIs are identified in the staffs safety evaluation on MRP-227 and include applicable action items on meeting those assumptions that formed the basis of the MRP's augmented inspection and flaw evaluation methodology (as discussed in Section 2.4 of MRP-227), and NSSS vendor-specific or plant-specific LRAAIs as well. Davis-Besse's responses to the plant-specific action items, as identified in Section 4.2 of the safety evaluation for MRP-227, will be submitted for NRC review and approval.

The guidance in Section 2.4 of MRP-227 specifies applicability limitations to base-loaded plants and the fuel loading management assumptions upon which the functionality analyses were based. General assumptions used in the analysis include:

1) 30 years of operation with high leakage core loading patterns (fresh fuel assemblies loaded in peripheral locations) followed by implementation of a low-leakage fuel management strategy for the remaining 30 years of operation;

2) base load operation, i.e., typically operates at fixed power levels and does not usually vary power on a calendar or load demand schedule; and

3) no design changes beyond those identified in general industry guidance or recommended by the original vendors.

Davis-Besse had approximately 13 years of operation with fresh fuel assemblies at peripheral locations. Cycle 15 has implemented a new failure resistant fuel design in high vulnerability locations. The core design for Davis Besse is within the assumption of MRP-227. Davis-Besse is a base load plant and has incorporated no design changes beyond those identified in general industry guidance or recommended by the original vendors.

Preventive Actions The guidance in MRP-227 relies on PWR water chemistry control to prevent or mitigate aging effects that can be induced by corrosive aging mechanisms (e.g., loss of material induced by general, pitting corrosion, crevice corrosion, or stress corrosion cracking or any of its forms [SCC, PWSCC, or IASCC]).

Reactor coolant water chemistry is monitored and maintained in accordance with the PWR Water Chemistry Program. The PWR Water Chemistry Program is an existing Davis-Besse program that is consistent with the 10 elements of an effective aging management program as described in NUREG-1801,Section XI.M2, "Water Chemistry."

Enclosure L-1 1-252 Page 101 of 108

Parameters Monitored or Inspected The program manages the following age-related degradation effects and mechanisms that are applicable in general to the RVI components at the facility: (a) cracking induced by SCC, PWSCC, IASCC, or fatigue/cyclical loading; (b) loss of material induced by wear; (c) loss of fracture toughness induced by either thermal aging or neutron irradiation embrittlement; (d) changes in dimension due to void swelling and irradiation growth, distortion, or deflection; and (e) loss of preload caused by thermal and irradiation-enhanced stress relaxation or creep. For the management of cracking, the program monitors for evidence of surface breaking linear discontinuities if a visual inspection technique is used as the non-destruction examination (NDE) method, or for relevant flaw presentation signals if a volumetric UT method is used as the NDE method. For the management of loss of material, the program monitors for gross or abnormal surface conditions that may be indicative of loss of material occurring in the components. For the management of loss of preload, the program monitors for gross surface conditions that may be indicative of loosening in applicable bolted, fastened, keyed, or pinned connections. The program does not directly monitor for loss of fracture toughness that is induced by thermal aging or neutron irradiation embrittlement, or by void swelling and irradiation growth; instead, the impact of loss of fracture toughness on component integrity is indirectly managed by using visual or volumetric examination techniques to monitor for cracking in the components and by applying applicable reduced fracture toughness properties in the flaw evaluations if cracking is detected in the components and is extensive enough to warrant a supplemental flaw growth or flaw tolerance evaluation under the MRP-227 guidance or ASME Code, Section Xl requirements. The program uses physical measurements to monitor for any dimensional changes due to void swelling, irradiation growth, distortion, or deflection.

Specifically, the program implements the parameters monitored/inspected criteria for B&W designed Primary Components in Table 4-1 of MRP-227.

Additionally, the program implements the parameters monitored/inspected criteria for B&W designed Expansion Components in Table 4-4 of MRP-227.

No existing generic industry programs contain the specificity considered sufficient for monitoring the aging effects addressed by the MRP-227 guidelines for B&W plants. Therefore, no components for B&W plants were placed into the Existing Programs group. No inspections, except for those specified in ASME Code, Section Xl, are required for components that are identified as requiring "No Additional Measures," in accordance with the analyses reported in MRP-227. As part of the Davis-Besse Inservice Inspection Program, a visual VT-3 examination of the reactor vessel removable core support structure is conducted once per Inservice Inspection

Enclosure L-1 1-252 Page 102 of 108 interval in accordance with ASME Section Xl, Table IWB-2500-1, Examination Category B-N-3.

MRP-227 I&E guidelines require a visual (VT-3) examination of the core support shield (CSS) vent valve retaining rings and disc shaft for every 10 year Inservice Inspection Interval. In addition, Davis-Besse Technical Specification 5.5.4 requires testing of the CSS vent valves every 24 months to verify by visual inspection that the valve body and valve disc exhibit no abnormal degradation, verify the valve is not stuck in an open position, and verify by manual actuation that the valve is fully open when a force of

-400 lbs is applied vertically upward. The technical specification inspection will continue to be performed at the prescribed frequency of 24 months. The MRP-227 required visual (VT-3) examination will also be performed at the prescribed frequency of every 10 year Inservice Inspection Interval.

Detection of Aging Effects The detection of aging effects is covered in two places: (a) the guidance in Section 4 of MRP-227 provides an introductory discussion and justification of the examination methods selected for detecting the aging effects of interest; and (b) standards for examination methods, procedures, and personnel are provided in a companion document, MRP-228. In all cases, well-established methods were selected. These methods include volumetric UT examination methods for detecting flaws in bolting, physical measurements for detecting changes in dimension, and various visual (VT-3, VT-1, and EVT-1) examinations for detecting effects ranging from general conditions to detection and sizing of surface-breaking discontinuities. Surface examinations may also be used as an alternative to visual examinations for detection and sizing of surface-breaking discontinuities.

Cracking caused by SCC, IASCC, and fatigue is monitored/inspected by either VT-1 or EVT-1 examination (for internals other than bolting) or by volumetric UT examination (bolting). The VT-3 visual methods may be applied for the detection of cracking only when the flaw tolerance of the component or affected assembly, as evaluated for reduced fracture toughness properties, is known and has been shown to be tolerant of easily detected large flaws, even under reduced fracture toughness conditions. In addition, VT-3 examinations are used to monitor/inspect for loss of material induced by wear and for general aging conditions, such as gross distortion caused by void swelling and irradiation growth or by gross effects of loss of preload caused by thermal and irradiation-enhanced stress relaxation and creep.

In addition, the program adopts the recommended guidance in MRP-227 for defining the Expansion criteria that need to be applied to inspections of Primary Components and Existing Requirement Components and for expanding the examinations to include additional Expansion Components. As

Enclosure L-1 1-252 Page 103 of 108 a result, inspections performed on the RVI components are performed consistent with the inspection frequency and sampling bases for Primary Components and Expansion Components in MRP-227, which have been demonstrated to be in conformance with the inspection criteria, sampling basis criteria, and sample Expansion criteria in Section A.1.2.3.4 of NRC Branch Position RLSB-1.

Specifically, the program implements the parameters monitored/inspected criteria and bases for inspecting the relevant parameter conditions for B&W designed Primary Components in Table 4-1 of MRP-227 and for B&W designed Expansion Components in Table 4-4 of MRP-227.

As provided in Section 4.1.3 of the MRP-227 safety evaluation, the flow distributor-to-shell forging bolts (also known as the flow distributor bolts) in B&W designed plants were added to the "Primary" inspection category. The safety evaluation provides that the examination method shall be volumetric examination (UT), the examination coverage for these components shall conform to the criteria as described in Section 3.3.1 of the safety evaluation, and the re-examination frequency shall be on a 10-year interval similar to other "Primary" inspection category components. For B&W designed plants, no other additional components were added to the "Primary" inspection category and no additional components were added to the "Expansion" inspection category.

In addition, in some cases (as defined in MRP-227), physical measurements are used as supplemental techniques to manage for the gross effects of wear, loss of preload due to stress relaxation, or for changes in dimension due to void swelling, deflection or distortion. The physical measurements methods applied in accordance with this program includes Section 4.3.1 of MRP-227 that describes the physical measurements needed for the B&W internals core clamping items. In addition, Table 4-1 provides the required examination method and examination coverage and Table 5-1 provides the acceptance criteria for the physical measurements.

Monitoring and Trending The program requires that all inspections shall be documented for future review; defects shall be documented in accordance with the Davis-Besse corrective action program.

In addition, the program requires that a summary report of all inspections and monitoring, items requiring evaluation, and new repairs shall be submitted to the MRP Program Manager within 120 days of the completion of an outage during which PWR internals within the scope of MRP-227 are examined.

Enclosure L-1 1-252 Page 104 of 108 Section 6 of MRP-227 will not be used by FENOC for evaluating examination results that do not meet the acceptance criteria identified in Section 5 of MRP-227. Rather, FENOC plans to use WCAP-17096-NP, Revision 2 as the framework to develop those generic and plant-specific evaluations triggered by findings in the RVI examinations. As provided in the safety evaluation for MRP-227, Rev. 0, the NRC staff is currently reviewing WCAP-17096-NP, Revision 2.

Acceptance Criteria Section 5 of MRP-227 provides specific examination acceptance criteria for the Primary and Expansion Component examinations. For components addressed by examinations referenced to ASME Code, Section Xl, the IWB-3500 acceptance criteria apply. For other components covered by Existing Programs, the examination acceptance criteria are described within the Existing Program reference document.

The guidance in MRP-227 contains three types of examination acceptance criteria:

For visual examination (and surface examination as an alternative to visual examination), the examination acceptance criterion is the absence of any of the specific, descriptive relevant conditions; in addition, there are requirements to record and disposition surface breaking indications that are detected and sized for length by VT-1/EVT-1 examinations;

" For volumetric examination, the examination acceptance criterion is the capability for reliable detection of indications in bolting, as demonstrated in the examination Technical Justification; in addition, there are requirements for system-level assessment of bolted or pinned assemblies with unacceptable volumetric (UT) examination indications that exceed specified limits; and,

" For physical measurements, the examination acceptance criterion for the acceptable tolerance in the measured differential height from the top of the plenum rib pads to the vessel seating surface in B&W plants are given in Table 5-1 of MRP-227.

Section 6 of MRP-227 will not be used by FENOC for evaluating examination results that do not meet the acceptance criteria identified in Section 5 of MRP-227. Rather, FENOC plans to use WCAP-17096-NP, Revision 2 as the framework to develop those generic and plant-specific evaluations triggered by findings in the RVI examinations. As provided in the safety evaluation for MRP-227, Rev. 0, the NRC staff is currently reviewing WCAP-17096-NP, Revision 2.

Enclosure L-1 1-252 Page 105 of 108

" Corrective Actions This element is common to Davis-Besse programs and activities that are credited with aging management during the period of extended operation and is discussed in Section B.1.3.

Confirmation Process This element is common to Davis-Besse programs and activities that are credited with aging management during the period of extended operation and is discussed in Section B.1.3.

" Administrative Controls This element is common to Davis-Besse programs and activities that are credited with aging management during the period of extended operation and is discussed in Section B .1.3.

" Operating Experience Relatively few incidents of PWR internals aging degradation have been reported in operating U.S. commercial PWR plants. However, a considerable amount of PWR internals aging degradation has been observed in European PWRs, with emphasis on cracking of baffle-former bolting. For this reason, the U.S. PWR owners and operators began a program a decade ago to inspect the baffle-former bolting in order to determine whether similar problems might be expected in U.S. plants. A benefit of this decision was the experience gained with the UT examination techniques used in the inspections. In addition, the industry began substantial laboratory testing projects in order to gather the materials data necessary to support future inspections and evaluations. Another item with existing or suspected material degradation concerns that has been identified for PWR components is cracking in some high-strength bolting. This condition has been corrected primarily through bolt replacement with less susceptible material and improved control of pre-load.

Stress corrosion cracking (SCC) has occurred in Alloy A-286 internals bolting in B&W units, this included Davis-Besse. The Alloy A-286 bolt failures in B&W PWR internals were subjected to a comprehensive failure analysis that is documented in BAW-1843PA, "The B&W Owners Group Evaluation of Internal Bolting Concerns in 177FA Plants," dated January 1986. BAW-1843PA was reviewed and approved by the NRC. This failure analysis addressed probable cause of the cracking, assessment of likelihood and consequences of joint failure, and replacement bolt design. The recommended replacement bolts were Alloy X-750 HTH bolts that are less susceptible to SCC and have overall excellent material properties.

Enclosure L-1 1-252 Page 106 of 108 Davis-Besse has replaced the majority of the Alloy A-286 bolts for the reactor vessel internals (upper core barrel, lower core barrel, lower thermal shield and surveillance specimen holder tubes) with Alloy X-750 HTH bolts. To satisfy a needed action under NEI 03-08 protocol, Davis-Besse performed UT examinations of 100% of all upper core barrel bolts during the cycle 16 refueling outage. This inspection did not identify any unacceptable indications.

As part of the Inservice Inspection Program, a visual (VT-3) examination of the reactor vessel removable core support structure is conducted once per Inservice Inspection interval in accordance with ASME Section Xl, Table IWB 2500 1, Examination Category B-N-3. These inspections have not identified any unacceptable indications.

FENOC participates in the industry programs for investigating and managing aging effects on reactor vessel internals. Through its participation in EPRI MRP activities, FENOC will continue to benefit from the reporting of reactor vessel internals inspection information, and will share its own internals inspection results with the industry, as appropriate.

Conclusion The PWR Reactor Vessel Internals Program provides reasonable assurance that cracking, including stress corrosion cracking (SCC), which also encompasses primary water stress corrosion cracking (PWSCC), irradiation-assisted stress corrosion cracking (IASCC), or cracking due to fatigue/cyclical loading; loss of material induced by wear; loss of fracture toughness due to either thermal aging or neutron irradiation embrittlement; and loss of preload due to thermal and irradiation-enhanced stress relaxation or creep of the subject reactor vessel internals components will be adequately managed so that intended functions of components within the scope of license renewal are maintained consistent with the current licensing basis for the period of extended operation.

Enclosure L-11-252 Page 107 of 108 Affected LRA Section LRA Page No. Affected Paragraph and Sentence B.2.34 Page B-137 NUREG-1801 Consistency - first paragraph Exceptions to NUREG-1801 - entire section In response to RAI B.2.34-2, LRA Section B.2.34, "Reactor Head Closure Studs Program," subsections "NUREG-1801 Consistency," and "Exceptions," are revised to read as follows:

NUREG-1801 Consistency The Reactor Head Closure Studs Program is an existing Davis-Besse program that, with enhancement, will be consistent with the 10 elements of an effective aging management program as described in NUREG-1801,Section XI.M3, "ReactorHead Closure Studs," with the followinq exceptions.

Exceptions to NUREG-1801 None.

Program Elements Affected:

Preventive Actions NUREG-1801 Section XI.M3 recommends use of bolting materialfor closure studs that have an actualmeasured yield strength of less than 150 kilo-pounds per square inch (ksi). However, the Davis-Besse reactor head closure studs have an actual measured yield strength of greater than 150 ksi. Justificationfor the adequacy of the ReactorHead Closure Programto manage cracking due to stress corrosion cracking of high-strengthbolting material(i.e., yield strength of greaterthan 150 ksi) is as follows.

The ReactorHead Closure Studs Program inspections are implemented by the Inservice Inspection (ISI) Program. The ISI Programprovides for examination of the reactorvessel stud assembliesin accordance with the examination and inspection requirementsspecified in the ASME B&PV Code, Section X1, Subsection IWB (1995 Edition through the 1996 Addenda), and approvedASME Code Cases. The extent and frequency of these examinations provide for timely detection of cracks. To manage cracking, each reactorhead closure stud is volumetrically examined once

Enclosure L-1 1-252 Page 108 of 108 per each 10-year Inservice Inspection Interval. In addition, Davis-Besse has not experienced cracking of the reactorhead closure studs.

When the reactorhead closure studs are removed from the reactorvessel flange during refueling outages, the studs, nuts and washers are stored in protective racks after removal and the reactorvessel flange holes are plugged with watertiqhtplugs during cavity flooding. These methods assure the holes, studs, nuts, and washers are protected from borated water and other potential contaminantsduring cavity flooding. In addition, the visible portions of the studs are inspected for boric acid corrosion prior to removal.

v t e Letter Sequence Response to RAI
TAC:ME4640, Steam Generator Tube Integrity (Approved, Closed)
Initiation Request, Request, Request, Request Acceptance... Supplement, Supplement, Supplement, Supplement, Supplement Administration Withholding Request Acceptance Meeting Questions 28 February 2011 17 February 2011 17 March 2011 21 March 2011 18 March 2011 5 April 2011 30 March 2011 20 April 2011 2 May 2011 19 May 2011 20 June 2011 11 July 2011 12 July 2011 21 July 2011 21 July 2011 27 July 2011 11 August 2011 12 September 2011 9 September 2011 22 September 2011 11 October 2011 8 November 2011 22 November 2011 2 December 2011 2 December 2011 2 December 2011 27 December 2011 13 December 2011 21 December 2011 29 December 2011 12 July 2011 12 July 2011 12 July 2011 23 January 2012 30 January 2012 19 January 2012 1 February 2012 9 February 2012 27 January 2012 21 February 2012 22 February 2012 2 March 2012 27 February 2012 22 February 2012 24 February 2012 16 February 2012 27 February 2012 15 May 2012 28 February 2012 2 April 2012 ... further results Answers 18 March 2011 23 March 2011 23 March 2011 15 April 2011 20 April 2011 5 May 2011 24 May 2011 3 June 2011 24 June 2011 17 June 2011 22 July 2011 26 August 2011 17 August 2011 16 September 2011 7 October 2011 21 October 2011 9 November 2011 23 November 2011 5 April 2012 25 May 2012 14 June 2012 5 July 2012 11 July 2012 16 August 2012 16 August 2012 7 September 2012 2 November 2012 20 November 2012 20 December 2012 21 January 2013 12 February 2013 28 February 2013 14 March 2013 19 April 2013 21 May 2013 17 May 2013 4 June 2013 31 January 2014 11 March 2014 3 July 2014 29 July 2014 16 September 2014 28 October 2014 28 January 2015 15 April 2011 29 April 2011 9 March 2012 25 May 2012 Results Approval, Approval, Approval, Approval, Approval, Approval, Approval, Approval, Approval, Approval, Approval, Approval Other: L-11-114, Drawing No. M-042B, Rev. 35, Sampling System Sh. 2 Local Grab Samples, L-11-131, Drawing No. LR-OS041A2, Revision 1, Operational Schematic Emergency Diesel Generator Systems., L-11-334, Reply to Request Additional Information for the Review of the License Renewal Application Amendment No. 21, L-12-337, Review of the Safety Evaluation Report with Open Items Related to the License Renewal, L-12-444, Submittal of Contractor Equivalent Margins Assessments for Reactor Vessel Welds (Nonproprietary Versions), L-12-456, Notification of Closure of Commitments Related to the Review of the License Renewal Application, L-13-257, Notification of Completion of a License Renewal Commitment Related to the Review of License Renewal Application TAC No. ME4640) and License Renewal Application Amendment No. 45), L-13-330, License Renewal Application Amendment No. 46 - Annual Update, L-13-341, Review of the Safety Evaluation Report Related to the License Renewal of Davis-Besse Nuclear Power Station, L-14-085, License Renewal Application (TAC No. ME4640) Amendment No. 48, L-14-206, License Renewal Application Amendment No. 50 - Annual Update, L-15-120, Notification of Completion of License Renewal Commitments Related to the Review of the Davis-Besse Nuclear Power Station, Unit No. 1, License Renewal Application and License Renewal Application Amendment No. 55, L-15-139, License Renewal Reactor Vessel Internals Inspection Plan, L-15-214, License Renewal Application Amendment No. 59 - Annual Update, L-15-309, License Renewal Application Amendment No. 60, L-15-310, C-CSS-099.20-069, Rev 0, Shield Building Laminar Cracking Limits., ML102450565, ML111050091, ML11110A089, ML11110A091, ML11110A092, ML11110A093, ML11110A094, ML11110A095, ML11110A105, ML11110A106, ML11110A107, ML11122A014, ML11126A017, ML11126A018, ML11126A019, ML11126A020, ML11126A021, ML11126A022, ML11126A023, ML11126A024, ML11126A025, ML11126A026, ML11126A032, ML11126A033, ML11126A034, ML11126A035, ML11126A036, ML11126A037, ML11126A038, ML11126A039, ML11126A040, ML11126A041, ML11126A042, ML11126A043... further results
MONTHYEAR2011-02-25 [Table View]

L-11-252, Reply to Request for Additional Information for the Review of License Renewal Application and License Renewal Application No. 15

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