NL-08-014, Clarifications to Reactor Vessel Surveillance Program and Neutron Embrittlement Time-Limited Aging Analyses and Audit Item 105; and Revision to License Renewal Regulatory Commitment List

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Clarifications to Reactor Vessel Surveillance Program and Neutron Embrittlement Time-Limited Aging Analyses and Audit Item #105; and Revision to License Renewal Regulatory Commitment List
ML080250027
Person / Time
Site:  Entergy icon.png
Issue date: 01/17/2008
From: Dacimo F
Entergy Nuclear Northeast, Entergy Nuclear Operations
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
NL-08-014
Download: ML080250027 (44)
v • d • e


Text

Enterpy Nuclear Northeast Indian Point Energy Center 450 Broadway, GSB P.O. Box 249 SEntergy Buchanan, NY 10511-0249 Tel (914) 788-2055 Fred Dacimo Vice President License Renewal January 17, 2008 Re: Indian Point Units 2 & 3 Docket Nos. 50-247 & 50-286 NL-08-014 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001

SUBJECT:

Entergy Nuclear Operations Inc.

Indian Point Nuclear Generating Unit Nos. 2 & 3 Docket Nos. 50-247 and 50-286 Clarifications to Reactor Vessel Surveillance Program and Neutron Embrittlement Time-Limited Aging Analyses and Audit Item #105; and Revision to License Renewal Regulatory Commitment List

REFERENCES:

1. Entergy Letter dated April 23, 2007, F. R. Dacimo to Document Control Desk, "License Renewal Application" (NL-07-039)
2. Entergy Letter dated April 23, 2007, F. R. Dacimo to Document Control Desk, "License Renewal Application Boundary Drawings" (NL-07-040)
3. Entergy Letter dated April 23, 2007, F. R. Dacimo to Document Control Desk, "License Renewal Application Environmental Report References" (NL-07-041)
4. Entergy Letter dated October 11, 2007, F. R, Dacimo to Document Control Desk, "License Renewal Application (LRA)" (NL-07-124)
5. Entergy Letter November 14, 2007, F. R, Dacimo to Document Control Desk, "Supplement to License Renewal Application (LRA)

Environmental Report References" (NL-07-133)

6. Entergy Letter dated November 28, 2007, F.R. Dacimo to Document Control Desk, "Reply to Request for Additional Information Regarding License Renewal Application" (NL-07-140)
7. Entergy Letter December 18, 2007, F. R, Dacimo to Document Control Desk, "Amendment 1 to License Renewal Application (LRA)"

(NL-07-153)

/UI~4~

NL-08-014 Docket Nos. 50-247 & 50-286 Page 2 of 3

Dear Sir or Madam:

In the referenced letters, Entergy Nuclear Operations, Inc. (Entergy) applied for renewal of the Indian Point Energy Center operating license for Unit 2 and 3 and responded to staff questions regarding Reactor Vessel Surveillance Program and Reactor Neutron Embrittlement Time-Limited Aging Analyses. Per telecom between the NRC staff and Entergy on December 4, 2007, Entergy agreed to clarify the RAI responses submitted ,in Reference 6.

Attachment 1 provides additional clarification to address staff questions regarding Reactor Vessel Surveillance Program and Reactor Neutron Embrittlement Time-Limited Aging Analyses.

Attachment 2 provides clarification to Audit Item #105. Attachment 3 consists of a revision to the list of regulatory commitments associated with the LRA.

If you have any questions, or require additional information, please contact Mr. Robert Walpole at 914-734-6710.

I declare under penalty of perjury that the foregoing is true and correct. Executed on Sin7-or Sincerely, Fred R. Dacimo e4' &le e-.d oi, Vice President License Renewal

/

NL-08-014 Docket Nos. 50-247 & 50-286 Page 3 of 3 Attachments:

1. Reactor Vessel Surveillance Program and Reactor Neutron Embrittlement Time-Limited Aging Analyses RAI Clarifications (This clarification supplements submittal in letter NL-07-140 dated 11-28-2007)
2. Audit Item #105 Clarification (This revision supersedes the revision submitted in letter NL-07-153 dated 12-18-2007)
3. List of Regulatory Commitments, Revision 2 (This revision supersedes the revision submitted in letter NL-07-153 dated 12-18-2007) cc: Mr. Samuel J. Collins, Regional Administrator, NRC Region I Mr. Kenneth Chang, NRC Branch Chief, Engineering Review Branch I Mr. Bo M. Pham, NRC Environmental Project Manager Mr. John Boska, NRR Senior Project Manager Mr. Paul Eddy, New York State Department of Public Service NRC Resident Inspector's Office Mr. Paul D. Tonko, President, New York State Energy, Research, & Development Authority

ATTACHMENT 1 TO NL-08-014 Reactor Vessel Surveillance Program and Reactor Neutron Embrittlement Time-Limited Aging Analyses RAI Clarifications (This clarification supplements submittal in letter NL-07-140 dated 11-28-2007)

ENTERGY NUCLEAR OPERATIONS, INC.

INDIAN POINT NUCLEAR GENERATING UNIT NOS. 2 & 3 DOCKET NOS. 50-247 AND 50-286

NL-08-014 Attachment 1 Docket Nos. 50-247 & 50-286 Page 1 of 21 INDIAN POINT NUCLEAR GENERATING UNIT NOS. 2 AND 3 LICENSE RENEWAL APPLICATION (LRA)

REQUESTS FOR ADDITIONAL INFORMATION (RAI)

Entergy responded in letter NL-07-140, Reply to Request for Additional Information Regarding License Renewal Application, dated November 28, 2007 to staff questions regarding Reactor Vessel Surveillance Program and Reactor Neutron Embrittlement Time-Limited Aging Analyses.

Per telecom between the NRC staff and Entergy on December 4, 2007, Entergy agreed to clarify the RAI responses (ML073450327).

RAI 4.2.1-1 The Charpy Upper-Shelf Energy (USE) and Pressurized Thermal Shock analyses utilize the neutron fluence at 48 effective full power years (EFPY) to represent the neutron fluence for the reactor vessels at the end of the period of extended operation.

A) What were the EFPY achieved for each unit prior to the last refueling outage? What capacity factors and neutron flux were assumed for each unit from the last refueling outage to the end of the period of extended operation to result in 48 EFPY at the end of the period of extended operation? Explain why these capacity factors and neutron flux values are applicable for determining the neutron fluence for the reactor vessels at the end of the period of extended operation.

B) How will future capacity factors, neutron flux and neutron fluence values be monitored to ensure 48 EFPY values bound the actual conditions of the reactor vessels at the end of the period of extended operation?

Response for RAI 4.2.1-1 A response to the RAI was provided in Reference 6. Per telecom on December 4, 2007 (ML073450327) the following clarification is provided:

For IP2, the calculated fluence received by the vessel at 21.8 EFPY (end of cycle 17) is 9.190E+18 n/cm 2. The expected neutron flux corresponding to the licensed reactor power rating for Cycle 18 through the period of extended operation is as follows.

NL-08-014 Attachment 1 Docket Nos. 50-247 & 50-286 Page 2 of 21 45 degree vessel location Cycle flux [n/cm 2-s]

18 1.14E+10 19 1.16E+10 Future cycles to 48 EFPY 1.22E+10 The IP3 expected neutron flux corresponding to the licensed reactor power rating from Cycle 14 through the period of extended operation is documented in Table 6-2 of WCAP-1 6251, "Analysis of Capsule X from Entergy's Indian Point 3 Reactor Vessel Radiation Surveillance Program". At the end of Cycle 14, Indian Point 3 had operated for 19.3 EFPY with a calculated fluence of 6.86E+18 n/cm 2. Using WCAP-16251 Table 6-2, the expected neutron flux corresponding to the licensed reactor power rating for Cycle 15 through the period of extended operation is as follows.

45 degree vessel location Cycle flux [n/cm 2-s]

15 9.63E+9 16 9.78E+9 Future cycles to 48 EFPY 9.78E+9 RAI 4.2.2-1 Table 4.2-2 in the LRA indicates that the percentage drop in Charpy USE for plate B2803-3 is 21.3 percent at 48 EFPY. The percentage drop in Charpy USE for plate B2803-3 was determined using its surveillance data, in accordance with Position 2.2 of Regulatory Guide (RG) 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials." Provide the analysis that was used to determine the percentage drop in Charpy USE for plate B2803-3, include all surveillance data (unirradiated and irradiated Charpy USE and surveillance capsule neutron fluence) and references for the surveillance data.

Response for RAI 4.2.2-1 A response to the RAI was provided in Reference 6. Per telecom on December 4, 2007 (ML073450327) the following revision/clarification is provided:

The second paragraph of our original response is being revised as follows:

"However, to maximize accuracy, IPEC used a spreadsheet and the equations for the RG 1.99 Figure 2 curves (available in NUREG/CR-5799) to effectively plot a parallel curve. Feuf Seven sets of surveillance data were reviewed, and a correction factor for each surveillance point was determined. The hiahestleweet correction factor (giving the

NL-08-014 Attachment 1 Docket Nos. 50-247 & 50-286 Page 3 of 21 highest % drop in USE) was then used in the formula to predict the 48 EFPY % drop in USE."

The last paragraph and table are being deleted as follows:

1----I .........

The .. W......o.. Tw...

... . ........ v..: *nwcgr",y database ;:-) was

.... ...... ... .w IuIep hp Aptrrpinrp th*p r-prrtiap, ... rhqr-Lead Meae~red inu Heat4 Capsue Faeter Vni;F 4TE P4/ae ,405422 S.-74 0-.34 O424 &.52 67 58 4.24464 P4a# A0542- -Y S-74 0204 67 &7 1.20649 Mae A0512- z S.46 0.-24 0.52 405 82 1.336 P~ae ,406422 S.46 4-.04 0.-24 67 56 "1.2 014 Clarification for RAI 4.2.2-1 is as follows:

The surveillance data from WCAP-16251, "Analysis of Capsule X from Entergy's Indian Point 3 Reactor Vessel Radiation Surveillance Program" as shown below was used to determine the correction factor.

Measured Predicted Fluence,  % drop in  % drop in Correction Type Heat ID Capsule 10E19 %Cu %Ni USE USE Factor Plate A0512-2 T 0.263 0.24 0.52 12.00% 24.05% -7.54 Plate A0512-2 T 0.263 0.24 0.52 16.00% 24.05% -2.05 Plate A0512-2 Y 0.692 0.24 0.52 25.00% 30.24% 3.28 Plate A0512-2 z 1.04 0.24 0.52 22.00% 33.31% -2.20 Plate A0512-2 z 1.04 0.24 0.52 18.00% 33.31% -6.17 Plate A0512-2 x 0.874 0.24 0.52 23.00% 31.96% -0.25 Plate A0512-2 x 0.874 0.24 0.52 24.00% 31.96% 0.78 Using a correction factor of 3.28 the 48 EFPY USE for lower shell plate B2803-3 as shown in the LRA is revised. Refer to the attached LRA revision for changes to LRA Table 4.2-2.

With the additional surveillance data, the chemistry factor for lower shell plate B2803-3 for 48 EFPY shown in LRA Tables 4.2-4 and 4.2-6 requires revision based on Table D-1 of WCAP-16251. Refer to the LRA revision attachment for changes to these tables.

NL-08-014 Attachment 1 Docket Nos. 50-247 & 50-286 Page 4 of 21 RAI 4.2.5-1 A) Table 4.2-3 in the LRA indicates that the ARTNDT value caused by irradiation for the intermediate shell axial welds and the lower shell axial welds in IP2 were determined using surveillance data reported in WCAP-15629, Revision 1, "Indian Point Unit 2 Heatup and Cooldown Limit Curves for Normal Operation and PTLR Support Documentation." This WCAP has surveillance data from IP2, IP3, and H.B. Robinson, Unit 2. The IP2 fluences were calculated using approved methodologies (WCAP-15557-RO, "Qualification of the Westinghouse Pressure Vessel Neutron Fluence Evaluation Methodology," and WCAP-14040-NP-A, Revision 2, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves") that are based on RG 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," March 2001, (This RG requires the use of ENDF/BVI for determining neutron cross-sections which are included in the BUGLE-96 cross-section file). In addition, there is excellent agreement between calculated and corresponding measured values. The IP3 capsule analyses also used approved methods and cross sections, thus, they are acceptable. The H.B. Robinson calculations are reported in WCAP-14044, 'Westinghouse Surveillance Capsule Neutron Fluence Re-evaluation," that was issued in 1994 (before the issuance of RG 1.190 and the availability of BUGLE-96 and ENDF/BVI). WCAP-1 5629, Revision 1 indicates that 15%

was added to the values reported in WCAP-14044. Explain why 15% was added to the values reported in WCAP-14044. Provide neutron fluence values derived using a methodology that adheres to the guidance in RG 1.190. If the revised analysis results in a change in neutron fluence for the H.B. Robinson, Unit 2 surveillance capsules, provide the ARTNDT value caused by irradiation and the RTPTS value for the intermediate shell axial welds and the lower shell'axial welds in IP2 and provide the surveillance data analysis required by 10 CFR 50.61 (c)(2)(i).

B) Table 4.2-4 in the LRA indicates that the ARTNDT caused by irradiation for the lower shell plate B2803-3 in IP3 was determined using surveillance data reported by the licensee's response to Generic Letter (GL) 92-01, "Reactor Vessel Structural Integrity." This surveillance data was reported in Attachment I to a September 4, 1998, letter from J. Knubel (New York Power Authority). As discussed in RAI 4.2.5-1A, the surveillance data from IP3 is also reported in WCAP-15629, Revision 1. The neutron fluence values for the IP3 surveillance capsule that are reported in WCAP-15629, Revision 1 and in the September 4, 1998, letter have different values. The applicant is requested to revise the PTS analyses using neutron fluence values for the surveillance capsules that are determined using the guidance in RG 1.190 and to provide the surveillance data analysis required by 10 CFR 50.61 (c)(2)(1).

NL-08-014 Attachment 1 Docket Nos. 50-247 & 50-286 Page 5 of 21 Response for RAI 4.2.5-1 A response to the RAI was provided in Reference 6. Per telecom on December 4, 2007 (ML073450327) the original response is being revised in its entirety. The following is the revised RAI 4.2.5-1 response:

Part A Response for RAI 4.2.5-1 The latest surveillance data for H. B. Robinson Unit 2 (HBR2) is shown in WCAP-15805, "Analysis of H.B. Robinson Unit 2 Capsule X", and the latest IP3 data is shown in WCAP-16251, "Analysis of Capsule X from Entergy~s Indian Point 3 Reactor Vessel Radiation Surveillance Program". Both reports use the guidance of RG 1.190. Combining the IP2, IP3, and HBR2 surveillance data results in revised IP2 intermediate shell and lower shell axial weld chemistry factors. Refer to the attached LRA revision for changes to LRA Tables 4.2-3 and 4.2-5.

An incorrect margin term was discovered in LRA Table 4.2-3 while preparing this RAI clarification. Since intermediate shell plate B2002-2 used RG 1.99 position 2.1, the correct margin term is 17.0. Refer to the attached LRA revision for changes to LRA Tables 4.2-3 and 4.2-5.

The surveillance data for IP2 intermediate shell plates B2002-1, B2002-2, and B2002-3 was updated by WCAP-15629 which resulted in changes to 48 EFPY USE values for these plates.

Refer to the attached LRA revision for changes to LRA Table 4.2-1.

Part B Response for RAI 4.2.5-1 The IP3 48 EFPY neutron fluence values reported in LRA Table 4.2-4 are shown in Table 6-2 of WCAP 16251, "Analysis of Capsule X from Entergy's Indian Point 3 Reactor Vessel Radiation Surveillance Program" which follows the guidance of RG 1.190.

RAI 4.2.5-2 10 CFR 50.61 (b)(4) indicates that each pressurized water nuclear power reactor for which the analysis required by PTS rule indicates that if there is no reasonably practicable flux reduction program to prevent the RTPTS value from exceeding the PTS screening criteria based on the neutron fluence at the expiration date of the operating license, the licensee shall submit a safety analysis to determine what, if any, modifications to equipment, systems, and operation are necessary to prevent potential failure of the reactor vessel as a result of postulated PTS events if continued operation beyond the screening criteria is allowed. The analysis must be submitted at least three years before the RTPTS value is projected to exceed the PTS screening criteria.

NL-08-014 Attachment 1 Docket Nos. 50-247 & 50-286 Page 6 of 21 Section 4.2.5 in the LRA indicates that the RTpTs value for plate 82803-3 in IP3 will exceed the PTS screening criterion. Identify the flux reduction program initiated by the applicant to prevent the RTPTS value for plate B2803-3 in IP3 from exceeding the PTS screening criterion. Based on the information provided in response to RAI 4.2.5-1 (B) and RAI 4.2.1-1, identify when the RTPTS value for plate B2803-3 in IP3 is projected to exceed the PTS screening criterion.

Response for RAI 4.2.5-2 A response to the RAI was provided in Reference 6. Per telecom on December 4, 2007 (ML073450327) the following clarification is provided:

The response to RAI 4.2.5-2 is not affected by the clarified response to RAI 4.2.5-1.

NL-08-014 Attachment 1 Docket Nos. 50-247 & 50-286 Page 7 of 21 LRA Revisions Per telecom on December 4, 2007 (ML073450327) the following LRA revisions are provided:

LRA Section 4.2.2, Charpy Upper-Shelf Energy, Unit 2, is revised as follows.

The upper shelf energy (USE) values have been determined based on the maximum projected 48 EFPY beltline fluence shown in Section 4.2.1. The beltline region chemistry and surveillance data, including the un-irradiated CvUSE information, is from the-RVI-D2 databasc and clGarificd in WCAP-1 5629, Revision 1, WCAP-1 6251, Analysis of Capsule X from Entergy's Indian Point 3 Reactor Vessel Radiation Surveillance Program, and WCAP-15805, Analysis of H.B. Robinson Unit 2 Capsule X. The projected 48 EFPY peak beltline fluence level at the clad/base metal interface of 1.906E+19 n/cm2 was applied to all beltline materials except axial welds where the expected peak fluence is 1.295E+1 9 n/cm 2 . The resulting projected 48 EFPY CvUSE drop and resulting 1t CvUSE are shown in Table 4.2-1.

One intermediate shell plate (B2002-3) and one lower shell plate (B2003-1) have projected upper shelf energy levels that fall below 50 ft-lb during the period of extended operation. All remaining plate and weld beltline materials exceed 50 ft-lb at 48 EFPY.

10 CFR Part 50, Appendix G, Section IV.A.1 requires licensees to take further corrective actions for cases where the 50 ft-lbs end-of-life USE criterion cannot be met (e.g., when the EOL USE falls below the USE value criterion specified in a previously NRC-approved EMA).

As noted in Table 4.2-1, the lowest projected USE level for the IP2 beltline plate material through the period of extended operation is 474 48.3 ft-lb for intermediate shell plate B2002-3. An equivalent margins analysis performed in WCAP-1 3587, Rev. 1 demonstrated that the minimum acceptable USE for reactor vessel plate material in 4 loop plants such as IP2 is 43 ft-lbs. In the safety assessment of WCAP-13587, the NRC concluded the report demonstrated margins of safety equivalent to those of the ASME code for beltline plate and forging materials. The IP2 USE values are therefore acceptable since the IP2 lowest projected USE level for the IP2 beltline plate material through the period of extended operation of 474 48.3 ft-lb for intermediate shell plate B2002-3 is above the 43 ft-lbs minimum acceptable USE for 4 loop plants determined in WCAP-13587 Rev. 1. This determination is consistent with NUREG-1800, Section 4.2.2.1.1.2, and with the NRC Safety Evaluation Report of acceptable USE for H. B Robinson Unit 2 as documented in NUREG-1785. The TLAA for USE is projected through the period of extended operation in accordance with 1 OCFR54.21 (c)(1)(ii).

LRA Section 4.2.2, Charpy Upper Shelf Energy, Unit 3, is revised as follows.

The IPEC Unit 3 upper shelf energy values have been determined based on the maximum projected 48 EFPY beltline fluence and the beltline region chemistry and surveillance data including the un-irradiated CvUSE information as summarized in the RVQD2 databasc WCAP-1 6251, Analysis of Capsule X from Enterqy's Indian Point 3 Reactor Vessel Radiation Surveillance Program. The projected 48 EFPY peak beltline fluence level at the clad/base metal interface of 1.560E+1 9 n/cm 2 was conservatively applied to all beltline materials. The 48 EFPY 'At fluence level of 9.298E+1 8 n/cm 2 was calculated in accordance with Regulatory Guide 1.99, Equation (3) based on a vessel thickness of 8.625". The resulting projected 48 EFPY CvUSE drop and resulting 1/t CvUSE are displayed in Table 4.2-2. All plate and weld beltline materials exceed 50 ft-lb at 48 EFPY aid aR

NL-08-014 Attachment 1 Docket Nos. 50-247 & 50-286 Page 8 of 21

... utalent m.argins analys*s as not rc.uircd with the exception of the lower shell plate B2803-3 with a predicted USE of 49.8 ft-lbs. As noted above for IP2, an equivalent margins analysis performed in WCAP-1 3587, Rev. 1 demonstrated that the minimum acceptable USE for reactor vessel plate material in 4 loop plants such as IP3 is 43 ft-lbs. Therefore, the IP3 lower shell plate B2803-3 USE value of 49.8 ft-lbs is acceptable. This determination is consistent with NUREG-1 800, Section 4.2.2.1.1.2, and with the NRC Safety Evaluation Report of acceptable USE for H. B Robinson Unit 2 as documented in NUREG-1785. The TLAA for USE is projected through the period of extended operation in accordance with 10CFR54.21 (c)(1)(ii)

LRA Section 4.2.5, Pressurized Thermal Shock, Unit 3, first paragraph is revised as follows.

The projected 48 EFPY peak beltline fluence level at the clad/base metal interface of 1.560E+19 n/cm 2 was applied to all beltline materials. The resulting projected 48 EFPY RTPTs are shown in Table 4.2-4. All projected RTPTS values are within the established screening criteria for 48 EFPY with the exception of plate B2803-3, which exceeds the screening criterion by " 9._5 F. Values of RTNDT for the IP3 beltline materials at 1/4 T and 34 T are summarized in Table 4.2-6.

NL-08-014 Attachment 1 Docket Nos. 50-247 & 50-286 Page 9 of 21 Table 4.2-1 IP2 Charpy Upper-Shelf Energy Data for 48 Effective Full-Power Years (EFPY)

Fluence 48 Reactor Vessel Material Material Vessel Fluence Un-  % Drop EFPY RG 1.99 Location (Beitline Ident Type Heat # Clad/BM 1/4T %Cu irradiate in USE USE at Position Identification) 48 EFPY 48EFPY dUSE 1/4 T Intermediate shell B2002-1 A302BM B-4688-2 1.906E+19 1.136E+19 0.190 70 21,4% 6&2 2.2 21.8% 54.7 Intermediate shell B2002-2 A302BM B-4701-2 1.906E+19 1.136E+19 0.170 73 2.8% 5&4 2.2 24.3% 55.3 Intermediate shell B2002-3 A302BM B-4922-1 1.906E+19 1.136E+19 0.250 74 36.0% 4-74 2.2 34.8% 48.3 Lowershell B2003-1 A302BM B-4791-1 1.906E+19 1.136E+19 0.200 71 29.9% 49.8 1.2 Lower shell B2003-2 A302BM B-4782-1 1.906E+19 1.136E+19 0.190 88 28.9% 62.6 1.2 Intermediate shell 2-042 Linde W5214 1.295E+19 7.72E+18 0.213 121 422% 6Q,9 2.2 axial welds A/B/C 1092 49.8% 60.8 Lower shell axial 3-042 A/B Linde W5214 1.295E+19 7.72E+18 0.213 121 422% 6 2.2 welds 1092 49.8% 60.8

NL-08-014 Attachment 1 Docket'Nos. 50-247 & 50-286 Page 10 of 21 Table 4.2-1 IP2 Charpy Upper-Shelf Energy Data for 48 Effective Full-Power Years (EFPY)

Reactor Vessel Location (Beltline Identification)

Intermediate to lower shell circumferential weld

NL-08-014 Attachment 1 Docket Nos. 50-247 & 50-286 Page 11 of 21 Table 4.2-2 IP3 Charpy Upper-Shelf Energy Data for 48 Effective Full-Power Years (EFPY)

Fn 48 Reactor Vessel Mence Fluence Un-  % Drop EFPY RG 1.99 MaeilVessel Mtra %Cup irrad1at9 %4 Location (Beltline Ident Te Heat # 1/4T %Cu irradiated in USE USE Position Identification) 48CEFPY 48 EFPY USE at 1/4 T

Intermediate shell B2802-1 A302BM B-5394-2 1.560E+1 9 9.298E+1 8 0.200 102 28.5% 72.9 1.2 Intermediate shell B2802-2 A302BM: A-0516-2 1.560E+19 9.298E+18 0.220 97 30.5% 67.4 1.2 Intermediate shell B2802-3 A302BM B-5391-2 1.560E+19 9.298E+18 0.200 95 28.5% 67.9 1.2 Lowershell B2803-1 A302BM A-0495-2 1.560E+19 9.298E+18 0.190 72 27.5% 52.2 1.2 Lower shell B2803-2 A302BM C-1 397-3 1i.560E+1 9 9.298E+1 8 0.220 94 30.5% 65.4 1.2 Lower shell B2803-3 A302BM A-0512-2 1.560E+1 9 9.298E+1 8 0.240 68 21.3% 5&3 2.2 26.8% 49.8 Intermediate shell 2-042 Linde 34B009 1.560E+19 9.298E+18 0.192 112 32.6% 75.5 1.2 axial welds 1092 Lower shell axial 3-042 Linde 34B009 1.560E+19 9.298E+18 0.192 112 32.6% 75.5 1.2 welds 1092

NL-08-014 Attachment 1 Docket Nos. 50-247 & 50-286 Page 12 of 21 Table 4.2-2 IP3 Charpy Upper-Shelf Energy Data for 48 Effective Full-Power Years (EFPY) 48 Reactor Vessel Material Materil Vessel Fluence Fluence Un-  % Drop EFPY RG 1.99 Location (Beltline Ident Type Heat # 1/4T %Cu irradiated in USE USE Position Identification) 48 EFPY 48EFPY USE at 1/4 T

Intermediate to lower 9-042 Linde 13253 1.560E+1 9 9.298E+1 8 0.221 111 35.5% 71.6 1.2 shell circumferential 1092 weld

NL-08-014 Attachment 1 Docket Nos. 50-247 & 50-286 Page 13 of 21 Table 4.2-3 IP2 Pressurized Thermal Shock Data for 48 Effective Full-Power Years (EFPY)

Fluence Chemistry Un- 48 Reactor Vessel Vessel Factor irradiated ARTNDT Margin EFPY Method Location Vsel Material Material Heat %Cu %Ni Clad/BM Fluence A RTNDT (FF) (°F) RTpTs 1G Identification) Ident Type Number (1019 Factor 1-,5639 1.99 n/cm 2) Rev-i- (°F) (°F)

Intermediate shell B2002-1 A302BM B-4688-2 0.190 0.650 1.906 1.176 114.0 34.0 134.1 17.0 185.1 2.1 Intermediate shell B2002-2 A302BM B-4701-2 0.170 0.460 1.906 1.176 118.2 21.0 139.1 34-0 194.4 2.1 17.0 177.1 Intermediate shell B2002-3 A302BM B-4922-1 0.250 0.600 1.906 1.176 181.9 21.0 214.0 17.0 252.0 2.1 Lower shell B2003-1 A302BM B-4791-1 0.200 0.660 1.906 1.176 152.00 20.0 178.8 34.0 232.8 1.1 Lower shell B2003-2 A302BM B-4782-1 0.190 0.480 1.906 1.176 128.80 -20.0 151.5 34.0 165.5 1.1 Intermediate shell 2-042 Linde W5214 0.213 1.007 1.295 1.072 264.7 -56 273.0 44.0 261.0 2.1 axial welds A/B/C 1092 251.8 269.9 257.9 Lower shell axial 3-042 Linde W5214 0.213 1.007 1.295 1.072 2-54.7 -56 273.0 44.0 261.0 2.1 welds A/B 1092 251.8 269.9 257.9

NL-08-014 Attachment 1 Docket Nos. 50-247 & 50-286 Page 14 of 21 Table 4.2-3 IP2 Pressurized Thermal Shock Data for 48 Effective Full-Power Years (EFPY)

Fluence Chemistry Un- 48 aVessel Factor irradiated ARTNDT Margin EFPY RG Location Vsel Material Material Heat %Cu %Ni Clad/BM Fluence A- FG Identification) Ident Type Number (1019 Factor 15622 RTNDT ('F) (°F) RTpTs 1.99

(°F) (°F) n/cm 2) Rev Intermediate to 9-042 Linde 34B009 0.192 1.007 1.906 1.176 220.9 -56 259.9 65.5 269.4 1.1 lower shell 1092 circumferential weld

NL-08-014 Attachment 1 Docket Nos. 50-247 & 50-286 Page 15 of 21 Table 4.2-4 IP3 Pressurized Thermal Shock Data for 48 Effective Full-Power Years (EFPY)

Fluence Un- 48 Reactor Vessel Vessel Fluenc Chemistry URTNDT irradiate Margin EFPY Method Location (Beltline Material Material Heat %Cu %Ni Clad/BM e Factor RG Identification) Ident Type Number (1019 Factor at--9-01 d RTNDT (°F) ('F) RTpTs 1.99

(°F) (°F) n/cm 2)

Intermediate shell B2802-1 A302BM B-5394-2 0.200 0.50 1.560 1.123 137.0 5.0 153.8 34.0 192.8 1.1 0

Intermediate shell B2802-2 A302BM A-0516-2 0.220 0.53 1.560 1.123 151.6 -4.0 170.2 34.0 200.2 1.1 0

Intermediate shell B2802-3 A302BM B-5391-2 0.200 0.49 1.560 1.123 135.8 17.0 152.5 34.0 203.5 1.1 0

Lower shell B2803-1 A302BM A-0495-2 0.190 0.47 1.560 1.123 127.7 49.0 143.4 34.0 226.4 1.1 0

Lower shell B2803-2 A302BM C-1397-3 0.220 0.52 1.560 1.123 150.2 -5.0 168.7 34.0 197.7 1.1 0

Lowershell B2803-3 A302BM A-0512-2 0.240 0.52 1.560 1.123 16.2 74.0 48&. 17.0 27. 2.1 0 167.9 188.5 279.5 Intermediate shell 2-042 Linde 34B009 0.192 1.00 1.560 1.123 221.3 -56 248.5 65.5 258.0 1.1 axial welds 1092 7 Lower shell axial 3-042 Linde 34B009 0.192 1.00 1.560 1.123 221.3 -56 248.5 65.5 258.0 1.1 welds 1092 7

NL-08-014 Attachment 1 Docket Nos. 50-247 & 50-286 Page 16 of 21 Table 4.2-4, IP3 Pressurized Thermal Shock Data for 48 Effective Full-Power Years (EFPY)

Fluence Un- 48 Reactor Vessel Vessel Fluenc Chemistry irradiate ARTNMT Margin EFPY RG Location (Beltline Material Material Heat %Cu %Ni Clad/BM e Factor dRTNDT (aF) (°F) Rp Identification) Ident Type Number (101i Factor Q d RTNDT ('F) (oF) RTPTS 1.99

(°F) (oF) n/cm')

Intermediate to 9-042 Linde 13253 0.221 0.73 1.560 1.123 189.1 -54 212.3 56.0 214.3 1.1 lower shell 1092 2 circumferential weld

NL-08-014 Attachment 1 Docket Nos. 50-247 & 50-286 Page 17 of 21 Table 4.2-5 1P2 Adjusted Reference Temperature at 48 Effective Full-Power Years (EFPY) 3/4 T Chemistry Un- 1/4 T Neutro 3/4 T 48 EFPY 48 Reactor Vessel Material Heat Factor irradiated Neutron 1/4 T 1/4 T n Fluenc 3/4 T 1/4 T EFPY Location (Beitline WCA1- Fluence 2 Fluence ARTNDT Fluenc ARTNDT 3/4T Identification) Ident Number - (

(0 F F) nc (1019 Factor (°F) e 11 e Factor ( ( F) (F)

°F RTNDT RTD rI/cm2) (10 19 (oF) n/cm 2)

Intermediate shell B2002-1 B-4688-2 114.0 34.0 1.136 1.036 118.1 0.404 0.748 85.3 169.1 136.3 Intermediate shell B2002-2 B-4701-2 118.2 21.0 1.136 1.036 122.4 0.404 0.748 88.5 1:77 4 143.5 160.4 126.5 Intermediate shell B2002-3 B-4922-1 181.9 21.0 1.136 1.036 188.4 0.404 0.748 136.1 226.4 174.1 Lowershell B2003-1 B-4791-1 152.00 20.0 1.136 1.036 157.4 0.404 0.748 113.8 211.4 167.8 Lower shell B2003-2 B-4782-1 128.80 -20.0 1.136 1.036 133.4 0.404 0.748 96.4 147.4 110.4 Intermediate shell 2-042 W5214 254.7- -56 0.772 0.927 236.2 0.274 0.647 4-641.9 224.2 152.9 axial welds. A/B/C 251.8 233.5 163.0 221.5 151.0 Lower shell axial 3-042 A/B W5214 254.7 -56 0.772 0.927 236.2 0.274 0.647 164.9 224.2 152.9 welds 251.8 233.5 163.0 221.5 151.0

NL-08-014 Attachment -1 Docket Nos. 50-247 & 50-286 Page 18 of 21 Table 4.2-5 IP2 Adjusted Reference Temperature at 48 Effective Full-Power Years (EFPY)

Reactor Vessel Location (Beltline Identification)

NL-08-014 Attachment 1 Docket Nos. 50-247 & 50-286 Page 19 of 21 Table 4.2-6 IP3 Adjusted Reference Temperature at 48 Effective Full-Power Years (EFPY) 1/4 T 3/4 T Neutro 3/4 T 48 48 Un- Neutro Reactor Vessel Material Heat Chemistr irradiated n 1/4 T 1/4 T n Fluenc 3/4 T EFPY EFPY Ident Number y Factor RT Fluenc Fluence ARTNDT Fluenc e ARTNDT 1/4 T 3/4 T Identification) FRV)I e Factor F) e F) RTNDT RTNDT (0 F) (1019 (1019 Factor (a

(0 F) (oF) n/cm2 ) n/cm 2)

Intermediate shell B2802-1 B-5394-2 137.0 5.0 0.930 0.980 134.2 0.330 0.695 95.2 173.2 134.2 Intermediate shell B2802-2 A-0516-2 151.6 -4.0 0.930 0.980 148.5 0.330 0.695 105.4 178.5 135.4 Intermediate shell B2802-3 B-5391-2 135.8 17.0 0.930 0.980 133.0 0.330 0.695 94.4 184.0 145.4 Lower shell B2803-1 A-0495-2 127.7 49.0 0.930 0.980 125.1 0.330 0.695 88.8 208.1 171.8 Lower shell B2803-2 C-1397-3 150.2 -5.0 0.930 0.980 147.1 0.330 0.695 104.4 176.1 133.4 Lower shell B2803-3 A-0512-2 1-68.2 74.0 0.930 0.980 164.9 0.330 0.695 1469 255. 207 167.9 - 164.5 116.7 255.5 207.7 Intermediate shell 2-042 34B009 221.3 -56 0.930 0.980 216.7 0.330 0.695 153.8 226.2 163.3 axial welds Lower shell axial 3-042 34B009 221.3 -56 0.930 0.980 216.7 0.330 0.695 153.8 226.2 163.3 welds Intermediate to 9-042 13253 189.1 -54 0.930 0.980 185.2 0.330 0.695 131.4 187.2 133.4 lower shell circumferential

NL-08-014 Attachment 1 Docket Nos. 50-247 & 50-286 Page 20 of 21 Table 4.2-6 IP3 Adjusted Reference Temperature at 48 Effective Full-Power Years (EFPY) 1/4 T 3/4 T Neutro 3/448 48 Un- Neutro Reactor Vessel Chemistr n 1/4 T 1/4 T n Fluenc 3/4 T EFPY EFPY Location (Beltline y Factor Fluenc Fluence ARTNDT Fluenc e ARTNDT 1/4 T 3/4 T Identification) Ident Number 141102 RTNDT e Factor (o F) e Ft e F) F RTNDT RTNDT (o F) (1019 (10,19 Factor (o F) oF) 2 n/cm ) n/cm )

weld

NL-08-014 Attachment 1 Docket Nos. 50-247 & 50-286 Page 21 of 21 LRA Section A.2.2.1.3, Charpy Upper-Shelf Energy, third paragraph, is revised as follows.

An equivalent margins analysis performed in WCAP-1 3587, Rev. 1, demonstrated that the minimum acceptable USE for reactor vessel plate material in four-loop plants is 43 ftlbs. In the safety assessment of WCAP-1 3587, the NRC concluded the report demonstrated margins of safety equivalent to those of the ASME code for beltline plate and forging materials. The USE values are therefore acceptable since the lowest projected USE level for the beltline plate material through the period of extended operation of 47448.3 ft-lb for intermediate shell plate B2002-3 is above the 43 ft-lbs minimum acceptable USE for four-loop plants determined in WCAP-1 3587 Rev. 1.

LRA Section A.3.2.1.3, Charpy Upper-Shelf Energy, is revised as follows.

The predictions for percent drop in CvUSE at 48 EFPY are based on chemistry data, unirradiated CvUSE data, and 1/4 T fluence values. The projected 48 EFPY peak beltline fluence level was conservatively applied to all beltline materials.

All plato and- .. o'ld. bo nomatoials meoteth rtequirement of exceeding a CUSE valuof ft lb at 48 EIRYV One lower shell plate (B2803-3) has a proiected upper shelf enerqy level below 50 ft-lb during the period of extended operation. The CvUSE for all remaining plate and weld beltline materials meets the acceptable value of greater than 50 ft-lb at 48 EFPY.

An equivalent margins analysis performed in WCAP-1 3587, Rev. 1, demonstrated that the minimum acceptable USE for reactor vessel plate material in four-loop plants is 43 ftlbs. In the safety assessment of WCAP-13587, the NRC concluded the report demonstrated margins of safety equivalent to those of the ASME code for beltline plate and forging materials. The USE value is therefore acceptable since the proiected USE level through the period of extended operation of 49.8 ft-lb for lower shell plate B2003-3 is above the 43 ft-lbs minimum acceptable USE for four-loop plants determined in WCAP-1 3587 Rev. 1.

ATTACHMENT 2 TO NL-08-014 Audit Item #105 Clarification (This clarification supersedes the revision submitted in letter NL-07-153 dated 12-18-2007)

ENTERGY NUCLEAR OPERATIONS, INC.

INDIAN POINT NUCLEAR GENERATING UNIT NOS. 2 & 3 DOCKET NOS. 50-247 AND 50-286

NL-08-014 Attachment 2 Docket Nos. 50-247 & 50-286 Page 1 of 1 INDIAN POINT NUCLEAR GENERATING UNIT NOS. 2 AND 3 LICENSE RENEWAL APPLICATION (LRA)

AUDIT ITEM CLARIFICATION Audit Item 105 Clarification The LRA amendment for Audit Item 105 communicated in letter NL-07-153, dated December 18, 2007, is replaced with the following.

LRA Section B.1.14, Fire Water System, Enhancements, is revised as follows.

The following enhancements will be implemented prior to the period of extended operation.

Attributes Affected Enhancements

3. Parameters Monitored or 4--. Revise applicable procedures to Inspected inspect the internal surface of the foam
4. Detection of Aging Effects based fire suppression tanks. Acceptance
6. Acceptance Criteria criteria will be enhanced to verify no significant corrosion.

LRA Section A.2.1.13, Fire Water System Program, fourth paragraph, is revised to add the following.

  • Revise applicable procedures to inspect the internal surface of the foam-based fire suppression tanks. Acceptance criteria will be enhanced to verify no significant corrosion.

ATTACHMENT 3 TO NL-08-014 List of Regulatory Commitments, Revision 2 (This revision supersedes the revision submitted in letter NL-07-153, dated 12-18-2007)

ENTERGY NUCLEAR OPERATIONS, INC.

INDIAN POINT NUCLEAR GENERATING UNIT NOS. 2 & 3 DOCKET NOS. 50-247 and 50-286

NL-08-014 Attachment 3 Docket Nos. 50-247 & 50-286 Page 1 of 16 List of Regulatory Commitments Rev. 2 The following table identifies those actions committed to by Entergy in this document.

Any other statements in this submittal are provided for information purposes and are not considered to be regulatory commitments.

  1. COMMITMENT IMPLEMENTATION SOURCE RELATED SCHEDULE LRA SECTION

/ AUDIT ITEM Enhance the Aboveground Steel Tanks Program for IP2: NL-07-039 A.2.1.1 IP2 and IP3 to perform thickness measurements of September 28, A.3.1.1 the bottom surfaces of the condensate storage tanks, 013 B.1.1 city water tank, and fire water tanks once during the iP3:

first ten years of the period of extended operation. December 12, Enhance the Aboveground Steel Tanks Program for 2015 IP2 and IP3 to require trending of thickness measurements when material loss is detected.

2 Enhance the Bolting Integrity Program for IP2 and IP3 IP2: NL-07-039 A.2.1.2 to clarify that actual yield strength is used in selecting September 28, A.3.1.2 013 B.1.2 materials for low susceptibility to SCC and clarify the prohibition on use of lubricants containing MoS 2 for IP3: NL-07-153 Audit Items bolting. I December 12, 201,241, The Bolting Integrity Program manages loss of 2015 270 preload and loss of material for all external bolting.

3 Implement the Buried Piping and Tanks Inspection IP2: NL-07-039 A.2.1.5 Program for IP2 and IP3 as described in LRA Section September 28, A.3.1.5 B.1.6. 2013 B.1.6 NL-07-153 Audit Item This new program will be implemented consistent with IP3: 173 the corresponding program' described in NUREG- December 12, 1801 Section XI.M34, Buried Piping and Tanks 015 Inspection. _

NL-08-014 Attachment 3 Docket Nos. 50-247 & 50-286 Page 2 of 16

  1. COMMITMENT IMPLEMENTATION SOURCE RELATED SCHEDULE LRA SECTION I 4/ AUDIT ITEM 4 Enhance the Diesel Fuel Monitoring Program to IP2: NL-07-039 A.2.1-.8 include cleaning and inspection of the IP2 GT-1 gas September 28, A.3.1.8 turbine fuel oil storage tanks, IP2 and IP3 EDG fuel oil 013 Adtie day tanks, IP2 SBO/Appendix R diesel generator fuel IP3: 128, 129, oil day tank, and IP3 Appendix R fuel oil storage tank DP Decembere112, 12132 and day tank once every ten years. 2015 Enhance the Diesel Fuel Monitoring Program to include quarterly sampling and analysis of the IP2 SBO/Appendix R diesel generator fuel oil day tank, IP2 security diesel fuel oil day tank, and IP3 Appendix R fuel oil storage tank. Particulates, water and sediment checks will be performed on the samples.

Filterable solids acceptance criterion will be less than or equal to 10mg/l. Water and sediment acceptance criterion will be less than or equal to 0.05%.

Enhance the Diesel Fuel Monitoring Program to include thickness measurement of the bottom surface of the following tanks once every ten years. IP2: EDG fuel oil storage tanks, EDG fuel oil day tanks, SBO/Appendix R diesel generator fuel oil day tank, GT-1 gas turbine fuel oil storage tanks, and diesel fire pump fuel oil storage tank; IP3: EDG fuel oil day tanks, Appendix R fuel oil storage tank, and diesel fire pump fuel oil storage tank.

Enhance the Diesel Fuel Monitoring Program to change the analysis for water and particulates to a quarterly frequency for the following tanks. IP2: GT-1 gas turbine fuel oil storage tanks and diesel fire pump fuel oil storage tank; IP3: Appendix R fuel oil day tank and diesel fire pump fuel oil storage tank.

Enhance the Diesel Fuel Monitoring Program to specify acceptance criteria for thickness measurements of the fuel oil storage tanks within the scope of the program.

Enhance the Diesel Fuel Monitoring Program to direct samples be taken near the tank bottom and include direction to remove water when detected.

Enhance the Diesel Fuel Monitoring Program to direct the addition of chemicals including biocide when the presence of biological activity is confirmed.

NL-08-014 Attachment 3 Docket Nos. 50-247 & 50-286 Page 3 of 16

  1. COMMITMENT IMPLEMENTATION SOURCE RELATED SCHEDULE LRA SECTION

/ AUDIT ITEM 5 Enhance the External Surfaces Monitoring Program IP2: NL-07-039 A.2.1.10 for IP2 and IP3 to include periodic inspections of September 28, A.3.1.10 2013 B.1.11 systems in scope and subject to aging management review for license renewal in accordance with 10 CFR P&

54.4(a)(1) and (a)(3). Inspections shall include areas surrounding the subject systems to identify hazards to De r1 those systems. Inspections of nearby systems that could impact the subject systems will include SSCs that are in scope and subject to aging management review for license renewal in accordance with 10 CFR 54.4(a)(2).

IP2: NL-07-039 A.2.1.1,1 6 Enhance the Fatigue Monitoring Program for IP2 to Spt b 28, A.3.1.11 monitor steady state cycles and feedwater cycles or September 28, A.3.1.11 perform an evaluation to determine monitoring is not 013 8.1.12, required. Review the number of allowed events and 1Ai resolve discrepancies between reference documents and monitoring procedures.

Enhance the Fatigue Monitoring Program for IP3 to IP3:

include all the transients identified. Assure all fatigue December 12, analysis transients are included with the lowest 2015 limiting numbers. Update the number of design transients accumulated to date.

NL-08-014 Attachment 3 Docket Nos. 50-247 & 50-286 Page 4 of 16

  1. COMMITMENT IMPLEMENTATION SOURCE RELATED SCHEDULE LRA SECTION IIP2: I__

NL-07-039 I/AUDIT ITEM A.2.1.12 7 Enhance the Fire Protection Program to inspect September 28, A.3.1.12 external surfaces of the IP3 RCP oil collection 2013 B.1.13 systems for loss of material each refueling cycle.

Enhance the Fire Protection Program to explicitly IP3:

state that the IP2 and IP3 diesel fire pump engine December 12, sub-systems (including the fuel supply line) shall be 2015 observed while the pump is running. Acceptance criteria will be revised to verify that the diesel engine does not exhibit signs of degradation while running; such as fuel oil, lube oil, coolant, or exhaust gas leakage.

Enhance the Fire Protection Program to specify that the IP2 and IP3 diesel fire pump engine carbon steel exhaust components are inspected for evidence of corrosion and cracking at least once each operating cycle.

Enhance the Fire Protection Program for IP3 to visually inspect the cable spreading room, 480V switchgear room, and EDG room C02 fire suppression system for signs of degradation, such as corrosion and mechanical damage at least once every six months.

NL-08-014 Attachment 3 Docket Nos. 50-247 & 50-286 Page 5 of 16

  1. COMMITMENT IMPLEMENTATION SOURCE RELATED*

SCHEDULE LRA SECTION

/ AUDIT ITEM 8 Enhance the Fire Water Program to include inspection IP2: NL-07-039 A.2.1.13 September 28, A.3.1.13 of IP2 and IP3 hose reels for evidence of corrosion.

be revised to verify no 2013 B.1.14 Acceptance criteria will IP3: 105, 106 unacceptable signs of degradation.

Enhance the Fire Water Program to replace all or test December 12, NL-08-014 a sample of IP2 and IP3 sprinkler heads required for 2015 10 CFR 50.48 using guidance of NFPA 25 (2002 edition), Section 5.3.1.1.1 before the end of the 50-year sprinkler head service life and at 10-year intervals thereafter during the extended period of operation to ensure that signs of degradation, such as corrosion, are detected in a timely manner.

Enhance the Fire Water Program to perform wall thickness evaluations of IP2 and IP3 fire protection piping on system components using non-intrusive techniques (e.g., volumetric testing) to identify evidence of loss of material due to corrosion. These inspections will be performed before the end of the current operating term and at intervals thereafter during the period of extended operation. Results of the initial evaluations will be used to determine the appropriate inspection interval to ensure aging effects are identified prior to loss of intended function.

Enhance the Fire Water Program to inspect the internal surface of foam based fire suppression tanks.

Acceptance criteria will be enhanced to verify no I significant corrosion. I

NL-08-014 Attachment 3 Docket Nos. 50-247 & 50-286 Page 6 of 16

  1. COMMITMENT IMPLEMENTATION SOURCE RELATED SCHEDULE LRA SECTION

/ AUDIT ITEM IP2: NL-07-039 A.2.1.15 9 Enhance the Flux Thimble Tube Inspection Program September 28, A.3.1.15 for IP2 and IP3 to implement comparisons to wear 2013 B.1.16 rates identified in WCAP-12866. Include provisions to compare data to the previous performances and IP3:

perform evaluations regarding change to test December 12, frequency and scope.

2015 Enhance the Flux Thimble Tube Inspection Program for IP2 and IP3 to specify the acceptance criteria as outlined in WCAP-12866 or other plant-specific values based on evaluation of previous test results.

Enhance the Flux Thimble Tube Inspection Program for IP2 and IP3 to direct evaluation and performance of corrective actions based on tubes that exceed or are projected to exceed the acceptance criteria. Also stipulate that flux thimble tubes that cannot be inspected over the tube length and cannot be shown by analysis to be satisfactory for continued service, must be removed from service to ensure the integrity of the reactor coolant system pressure boundary.

NL-08-014 Attachment 3 Docket Nos. 50-247 & 50-286 Page 7 of 16 COMMITMENT IMPLEMENTATION SOURCE RELATED SCHEDULE LRA SECTION I . _ I_ I I/ AUDIT ITEM 10 Enhance the Heat Exchanger Monitoring Program for IP2: NL-07-039 A.2.1.16 IP2 and IP3 to include the following heat exchangers September 28, A.3.1.16 in the scope of the program. 2013 B.1.17, NL-07-153 Audit Item

  • Safety injection pump lube oil heat exchangers IP3: 52 SRHR heat exchangers December 12, 2015
  • RHR pump seal coolers
  • Non-regenerative heat exchangers
  • Charging pump seal water heat exchangers
  • Charging pump fluid drive coolers
  • Charging pump crankcase oil coolers
  • Spent fuel pit heat exchangers
  • Waste gas compressor heat exchangers
  • SBO/Appendix R diesel jacket water heat exchanger (IP2 only)

Enhance the Heat Exchanger Monitoring Program for IP2 and IP3 to perform visual inspection on heat exchangers where non-destructive examination, such as eddy current inspection, is not possible due to heat exchanger design limitations.,

Enhance the Heat Exchanger Monitoring Program for IP2 and IP3 to include consideration of material-environment combinations when determining sample population of heat exchangers.

Enhance the Heat Exchanger Monitoring Program for IP2 and IP3 to establish minimum tube wall thickness for the new heat exchangers identified in the scope of the program. Establish acceptance criteria for heat exchangers visually inspected to include no unacceptable siqns of degradation.

NL-08-014 Attachment 3 Docket Nos. 50-247 & 50-286 Page 8 of 16

  1. COMMITMENT IMPLEMENTATION SOURCE RELATED SCHEDULE LRA SECTION

/ AUDIT ITEM IP2: NL-07-039 A.2.1.17 A.2.1.17 11 Enhance the ISI Program for IP2 and IP3 to provide periodic visual inspections to confirm the absence of September 28, A.3.1.17 aging effects for lubrite sliding supports used in the 013 B.1.18 NL-07-153 Audit 59item steam generator and reactor coolant pump support NP3:

systems. December 12, 2015 12 Enhance the Masonry Wall Program for IP2 and IP3 1P2: NL-07-039 A.2.1.18 to specify that the IP1 intake structure is included in September 28, A.3.1.18 the program. 2013 B.1.19 IP3:

December 12, 2015 13 Enhance the Metal-Enclosed Bus Inspection Program IP2: NL-07-039 A.2.1.19 to add IP2 480V bus associated with substation A to September 28, A.3.1.19 the scope of bus inspected. 2013 B.1.20 NL-07-153 Audit Item Enhance the Metal-Enclosed Bus Inspection Program IP3: 124 for IP2 and IP3 to visually inspect the external surface December 12, Audit Item of MEB enclosure assemblies for loss of material at 2015 133 least once every 10 years. The first inspection will occur prior to the period of extended operation and the acceptance criterion will be no significant loss of material.

Enhance the Metal-Enclosed Bus Inspection Program for IP2 and IP3 to inspect bolted connections at least once every five years if performed visually or at least once every ten years using quantitative measurements such as thermography or contact resistance measurements. The first inspection will occur prior to the period of extended operation.

The plant will process a change to applicable site procedure to remove the reference to "re-torquing" connections for phase bus maintenance and bolted connection maintenance.

14 Implement the Non-EQ Bolted Cable Connections IP2: NL-07-039 A.2.1.21 Program for IP2 and IP3 as described in LRA Section September 28, A.3.1.21 B.1.22. 2013 B.1.22 IP3:

December 12, 12015

NL-08-014 Attachment 3 Docket Nos. 50-247 & 50-286 Page 9 of 16

  1. COMMITMENT IMPLEMENTATION SOURCE RELATED SCHEDULE LRA SECTION

/ AUDIT ITEM 15 Implement the Non-EQ Inaccessible Medium-Voltage P2: NL-07-039 A.2.1.22 Cable SectionProgram B.1.23. for IP2 and IP3 as described in LRA September 2013 28, A.3.1.22 B.1 .23 NL-07-153 Audit item This new program will be implemented consistent with IP3: 173 the corresponding program described in NUREG- December 12, 1801 Section XI.E3, Inaccessible Medium-Voltage 2015 Cables Not Subject To 10 CFR 50.49 Environmental Qualification Requirements.

IP2: NL-07-039 A.2.1 .23 16 Implement the Non-EQ Instrumentation Circuits Test A.2.1.23 Review Program for IP2 and IP3 as described in LRA September 28, A.3.1.23 Section B. .24. 2013 B. 1.24 NL-07-153 Audit item This new program will be implemented consistent with IP3: 173 the corresponding program described in NUREG- December 12, 1801 Section XI.E2, Electrical Cables and 015 Connections Not Subject to 10 CFR 50.49 Environmental Qualification Requirements Used in Instrumentation Circuits.

17 Implement the Non-EQ Insulated Cables and eP2: NL-07-039 A.2.1.24 Connections Program for IP2 and IP3 as described in eptember 28, A.3.1.24 LRA Section B.1.25. 2013 B.1.25 NL-07-153 Audit item This new program will be implemented consistent with IP3: 173 the corresponding program described in NUREG- December 12, 1801 Section XI.E1, Electrical Cables and 2015 Connections Not Subject to 10 CFR 50.49 Environmental Qualification Requirements.

NL-08-014 Attachment 3 Docket Nos. 50-247 & 50-286 Page 10 of 16

  1. COMMITMENT IMPLEMENTATION SOURCE RELATED SCHEDULE LRA SECTION I AUDIT ITEM 18 Enhance the Oil Analysis Program for IP2 to sample ISeptember P2: 28, NL-07-039 A.2.1 .25 A.3.1.25 and analyze lubricating oil used in the SBO/Appendix 2013 B.1.26 R diesel generator consistent with oil analysis for 013 B.1.26 other site diesel generators. IP3:

Enhance the Oil Analysis Programfor IP2 and IP3 to December 12, sample and analyze generator seal oil and turbine 015 hydraulic control oil.

Enhance the Oil Analysis Program for IP2 and IP3 to formalize preliminary oil screening for water and particulates and laboratory analyses including defined acceptance criteria for all components included in the scope of this program. The program will specify corrective actions in the event acceptance criteria are not met.

Enhance the Oil Analysis Program for IP2 and IP3 to formalize trending of preliminary oil screening results as well as data provided from independent laboratories.

IP2: NL-07-039 A.2.1.26 19 Implement the One-Time Inspection Program for IP2 Sptb 28, A.3.1.26 and IP3 as described in LRA Section B.1.27. September 28, A.3.1.26 2013 B. 1.27 This new program will be implemented consistent with NL-07-153 Audit item the corresponding program described in NUREG- IP3: 173 1801,Section XI.M32, One-Time Inspection. December 12, 2015 20 Implement the One-Time Inspection - Small Bore IP2: NL-07-039 A.2.1.27 Piping Program for IP2 and IP3 as described in LRA September 28, A.3.1.27 Section B.1.28. 013 8.1.28 NL-07-153 Audit item This new program will be implemented consistent with IP3: 173 the corresponding program described in NUREG- December 12, 1801,Section XI.M35, One-Time Inspection of ASME 2015 Code Class I Small-Bore Piping.

21 Enhance the Periodic Surveillance and Preventive IP2: NL-07-039 A.2.1.28 Maintenance Program for IP2 and IP3 as necessary September 28, A.3.1.28 to assure that the effects of aging will be managed such that applicable components will continue to perform their intended functions consistent with the D e r1 current licensing basis through the period of extended D m 1 operation. 015

NL-08-014 Attachment 3 Docket Nos. 50-247 & 50-286 Page 11 of 16

  1. COMMITMENT IMPLEMENTATION SOURCE RELATED SCHEDULE LRA SECTION I AUDIT ITEM Enhance the Reactor Vessel Surveillance Program for 1P2: NL-07-039 A.2.1.31

.22 September 28, A.3.1.31 IP2 and IP3 revising the specimen capsule withdrawal 013 B.1.32 schedules to draw and test a standby capsule to cover the peak reactor vessel fluence expected IP3:

through the end of the period of extended operation. December 12, Enhance the Reactor Vessel Surveillance Program for 2015 IP2 and IP3 to require that tested and untested specimens from all capsules pulled from the reactor I vessel are maintained in storage.

IP2: NL-07-039 A.2.1.32 23 Implement the Selective Leaching Program for IP2 September 28, A.3.1.32 and IP3 as described in LRA Section B.1.33.

2013 B. 1.33 This new program will be implemented consistent with NL-07-153 Audit item the corresponding program described in NUREG- IP3: 173 1801,Section XI.M33 Selective Leaching of Materials. December 12, 2015 IP2: NL-07-039 A.2.1 .34 24 Enhance the Steam Generator Integrity Program for Ip2: 28, A.3.1.34 IP2 and IP3 to require that the results of the condition September 28, B.1.35 monitoring assessment are compared to the operational assessment performed for the prior IP3:

operating cycle with differences evaluated. December 12, 015 Enhance the Structures Monitoring Program to P:N0709 IP2: NL-07-039 A213 A.2.1.35 25 explicitly specify that the following structures are September 28, A.3.1.35 included in the program. 2013 B.1.36

  • Appendix R diesel generator foundation (IP3) NL-07-153
  • Appendix R diesel generator fuel oil tank vault IP3: Audit item (IP3) December 12, 86
  • Appendix R diesel generator switchgear and 2015 enclosure (IP3) Audit item
  • city water storage tank foundation 88
  • condensate storage tanks foundation (IP3) Audit Item
  • containment access facility and annex (IP3) 87
  • discharge canal (IP2/3)
  • fire pumphodse (IP2)
  • fire protection pumphouse (IP3)
  • fire water storage tank foundations (IP2/3)
  • gas turbine 1 fuel storage tank foundation
  • maintenance and outage building-elevated passageway (IP2)

NL-08-014 Attachment 3 Docket Nos. 50-247 & 50-286 Page 12 of 16

  1. COMMITMENT IMPLEMENTATION SOURCE RELATED SCHEDULE LRA SECTION

_ /

IAUDIT ITEM

  • new station security building (IP2)
  • nuclear service building (IP1)
  • primary water storage tank foundation (IP3)
  • refueling water storage tank foundation (IP3)
  • security access and office building (IP3)
  • superheater stack
  • transformer/switchyard support structures (IP2)
  • waste holdup tank pits (IP2/3)

Enhance the Structures Monitoring Program for IP2 and IP3 to clarify that in addition to structural steel and concrete, the following commodities (including their anchorages) are inspected for each structure as applicable.

  • cable trays and supports
  • concrete portion of reactor vessel supports
  • conduits and supports
  • cranes, rails and girders
  • equipment pads and foundations

" fire proofing (pyrocrete)

  • jib cranes

" manholes and duct banks

" manways, hatches and hatch covers

  • monorails
  • new fuel storage racks
  • sumps, sump screens, strainers and flow barriers Enhance the Structures Monitoring Program for IP2 and IP3 to inspect inaccessible concrete areas that are exposed by excavation for any reason. IP2 and IP3 will also inspect inaccessible concrete areas in environments where observed conditions in accessible areas exposed to the same environment indicate that significant concrete degradation is occurring.

Enhance the Structures Monitoring Program for IP2 and IP3 to perform inspections of elastomers (seals, gaskets, seismic joint filler, and roof elastomers) to identifv crackina and chanae in material orooerties

NL-08-014 Attachment 3 Docket Nos. 50-247 & 50-286 Page 13 of 16 COMMITMENT IMPLEMENTATION SOURCE RELATED SCHEDULE LRA SECTION

/ AUDIT ITEM and for inspection of aluminum vents and louvers to identify loss of material.

Enhance the Structures Monitoring Program for IP2 and IP3 to perform an engineering evaluation of groundwater samples to assess aggressiveness of groundwater to concrete on a periodic basis (at least once every five years). IPEC will obtain samples from at least 5 wells that are representative of the ground water surrounding below-grade site structures.

Samples will be monitored for sulfates, pH and chlorides.

Enhance the Structures Monitoring Program for IP2 and IP3 to perform inspection of normally submerged concrete portions of the intake structures at least once every 5 years.

26 Implement the Thermal Aging Embrittlement of Cast IP2: NL-07-039 A.2.1.36 Austenitic Stainless Steel (CASS) Program for IP2 September 28, A.3.1.36 and IP3 as described in LRA Section B.1.37. 013 B. 1.37 NL-07-153ý Audit item This new program will be implemented consistent with IP3: 173 the corresponding program described in NUREG- December 12, 1801,Section XI.M12, Thermal Aging Embrittlement 2015 of Cast Austenitic Stainless Steel, (CASS) Program.

27 Implement the Thermal Aging and Neutron Irradiation IP2: NL-07-039 A.2.1.37 Embrittlement of Cast Austenitic Stainless Steel September 28, A.3.1.37 2013 B. 1.38 (CASS) Program for IP2 and IP3 as8..38.NL-07-1 Sectin described in LRA 013 53 8u1.38 Audit item Section B.1.38. IP3: 173 This new program will be implemented consistent with December 12, the corresponding program described in NUREG- 2015 1801 Section XI.M13, Thermal Aging and Neutron Embrittlement of Cast Austenitic Stainless Steel (CASS) Program.

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/AUDIT ITEM IP2: NL-07-039 A.2.1.39 28 Enhance the Water Chemistry Control - Closed Sptb 28, A.3.1.39 Cooling Water Program to maintain water chemistry of September 28, A.314.39 the IP2 SBO/Appendix R diesel generator cooling system per EPRI guidelines. IP3:

Enhance the Water Chemistry Control - Closed December 12, Cooling Water Program to maintain the IP2 and IP3 2015 security generator cooling water system pH within limits specified by EPRI guidelines.

29 Enhance the Water Chemistry Control - Primary and IP2: NL-07-039 A.2.1.40 Secondary Program for IP2 to test sulfates monthly in September 28, .01.41 the RWST with a limit of <150 ppb. 013 30 For aging management of the reactor vessel internals, P2: NL-07-039 A.2.1.41 IPEC will (1) participate in the industry programs for September 28, A.3.1.41 investigating and managing aging effects on reactor 011 internals; (2) evaluate and implement the results of IP3:

the industry programs as applicable to the reactor December 12, internals; and (3) upon completion of these programs, 2013 but not less than 24 months before entering the period of extended operation, submit an inspection plan for reactor internals to the NRC for review and. approval.

31 Additional P-T curves will be submitted as required IP2: NL-07-039 A.2.2.1.2 per 10 CFR 50, Appendix G prior to the period of September 28, A.3.2.1.2 extended operation as part of the Reactor Vessel 013 4.2.3 Surveillance Program. IP3:

December 12, 2015 32 As required by 10 CFR 50.61 (b)(4), IP3 will submit a IP3: NL-07-039 A.3.2.1.4 plant-specific safety analysis for plate B2803-3 to the December 12, 4.2.5 NRC three years prior to reaching the RTPTS 2015 screening criterion. Alternatively, the site may choose to implement the revised PTS (10 CFR 50.61) rule when approved, which would permit use of Regulatory Guide 1.99, Revision 3.

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/ AUDIT ITEM IP2: NL-07-039 A.2.2.2.3 33 At least 2 years prior to entering the period of September 28, A.3.2.2.3 extended operation, for the locations identified in LRA 2011 4.3.3 Table 4.3-13 (IP2) and LRA Table 4.3-14 (IP3), IP2 and IP3 will implement one or more of the following: NL-07-153 Audit item IP3: 146 (1) Refine the fatigue analyses to determine valid December 12, CUFs less than 1 when accounting for the effects of 2013 reactor water environment. This includes applying the appropriate Fen factors to valid CUFs determined in accordance with one of the following:

1. For locations, including NUREG/CR-6260 locations, with existing fatigue analysis valid for the period of extended operation, use the existing CUF to determine the environmentally adjusted CUE.
2. In addition to the NUREG/CR-6260 locations, more limiting plant-specific locations with a valid CUF may be evaluated. In particular, the pressurizer lower shell will be reviewed to ensure the surge nozzle remains the limiting component.
3. Representative CUF values from other plants, adjusted to or enveloping the IPEC plant specific external loads may be used if demonstrated applicable to IPEC.
4. An analysis using an NRC-approved version of the ASME code or NRC-approved alternative (e.g., NRC-approved code case) may be performed to determine a valid CUF.

(2) Manage the effects of aging due to fatigue at the affected locations by an inspection program that has been reviewed and approved by the NRC (e.g., periodic non-destructive examination of the affected locations at inspection intervals to be determined by a method acceptable to the NRC).

(3) Repair or' replace the affected locations before exceeding a CUF of 1.0.

Should IPEC select the option to manage the aging effects due to environmental-assisted fatigue during the period of extended operation, details of the aging management program such as scope, qualification, method, and frequency will be submitted to the NRC at least 2 years prior to the period of extended operation.

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/ AUDIT ITEM 34 1P2 SBO / Appendix R diesel generator will be April 30, 2008 NL-07-078 2.1.1.3.5 installed and operational by April 30, 2008. This committed change to the facility meets the requirements of 10 CFR 50.59(c)(1) and, therefore, a license amendment pursuant to 10 CFR 50.90 is not required.

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