ML15222B151

ENT000674 - NL-08-014, Letter from Fred R. Dacimo, Entergy, to NRC Document Control Desk, Clarifications to Reactor Vessel Surveillance Program and Neutron Embrittlement Time-Limited Aging Analyses and Audit Item #105; and Revision to Licen

ML15222B151
Person / Time
Site:	Indian Point
Issue date:	01/17/2008
From:	Entergy Nuclear Operations
To:	Atomic Safety and Licensing Board Panel
SECY RAS
References
RAS 28138, ASLBP 07-858-03-LR-BD01, 50-247-LR, 50-286-LR
Download: ML15222B151 (44)

Text

ENT000674 Submitted: August 10, 2015 Enterav Nuclear Northeast Indlan Point Energy Center 450 Broadway GSB P 0 Box 249 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 t o Reactor Vessel Surveillance Proqram and Neutron Embrittlement Time-Limited Aqinq Analyses and Audit Item #105; and Revision t o License Renewal Reaulatorv 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 1I , 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 L~censeRenewal Appltcation (LRA)

Envtronmental 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, Dac~moto Document Control Desk "Amendment 1 to Lrcense Renewal Appl~cai~on (LRA)"

(NL-07-153)

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

Dear Sir or Madam:

In the referenced letters. Enterav Nuclear Ooerations. Inc. (Enterovl ..

-<. aoolied for renewal of the Indian Point Energy center opGiating license for Unit 2 and 3 and responded to staff questions reaardina Reactor Vessel Surveillance Program and Reactor Neutron Embrittlement Time-

~imitedAging Analyses. Per telecom between the NRC staff and Entergy on December 4, 2007, Entergy agreed to clarify the RAI responses submitted in Reference 6. provides additional clarification to address staff questions regarding Reactor Vessel Surveillance Program and Reactor Neutron Embrittlement Time-Limited Aging Analyses. 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

/- / 7 - o f .

Sincerely,

%r Fc Fred R. Dacimo ~ t ~ f ~ k w Vice President License Renewal

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

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

2. Audit Item #I05 Clarification (This revision supersedes the revision submitted in letter NL-07-153dated 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 Proaram and Reactor Neutron Embrittlement Time-Limited Asina Analvses 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 50-285

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 Surveiliance 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).

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 9.190Ec18 n/cmz. 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 af 21 45 degree vessel location Cycle flux [n/cm2-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-16251, "Analysis of Capsule X from Entergy's Indian Point 3 Reactor Vessel Radiation Suweillance Program". At the end of Cycle 14, Indian Point 3 had operated for 19.3 EFPY with a calculated fluence of 6.56E+18 n!cm2. 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/cm2-s]

15 9.63E+9 16 9.78E-i.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 82803-3 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 82803-3, 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 revisionlclarificationis 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 NUREGJCR-5799) to effectively plot a parallel curve, Few Seven sets of surveillance data were reviewed, and a correction factor for each surveillance point was determined. The M l e w e s t correction factor (giving ihe

NL-08-014 Attachment 1 Docket Nos. 50-247 & 50-286 Page 3 of 21 highest O/O 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:

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 USE USE Factor Plate AO512-2 T 0.263 12.00% 24.05% -7.54 Plate A05 12-2 T 0.263 16.00% 24.05% -2.05 Plate A05 12-2 Y 0.692 25.00% 30.24% 3.28 Plate A0512-2 i' 1.04 22.00% 33.31% -2.20 Plate A05 12-2 Z 1.04 18.00% 33.31% -6.17 Plate A05 12-2 X 0.874 23.00% 31.96% -0.25 Plate A0512-2 X 0.874 24.00% 31.96% 0.78 Ustng a correctton factof of 3.28 the 48 EFPY USE for lower shell plate B2803-3 as shown tn the LRA IS revtsed. Refer to the attached LRA revlslon for changes to LRA Table 4.2-2.

With the additional surveillance data, the chemistry factor for lower shell plate 82803-3 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 ART, 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 ENDFIBVI 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 ENDFIBVI). WCAP-15629, Revision Iindicates 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 ARTNDTvalue caused by irradiation and the RTpsi 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 ARTNDTcaused by irradiation for the lower shell plate 82803-3 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-IA, 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 guidan required by 10 CFR50.67 (C)

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 RTp7s value for plate B2803-3 in IP3 will exceed the PTS screening criterion. Identify the flux reduction program initiated by the applicant to prevent the RTp7s 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 82803-3 in IP3 is projected to exceed the PTS screening criterion.

Response for RA14.2.5-2 A response to the RAI was provided in Reference 6. Per teiecom 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 at 21 I R A 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-irradiatedCvUSE information, is from %e-W.D2 PWCAP-15629, Revision 1, WCAP-16251. Analvsis of Cawsule X from Enterav's lndian Point 3 Reactor Vessel Radiation Surveillance Proaram, and WCAP-15805, Analvsis of H.B. Robinson Unit 2 Cawsule X. The projected 48 EFPY peak beltline fluence level at the cladlbase 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+19 n/cm2. The resulting projected 48 EFPY CvUSE drop and resulting %t CvUSE are shown in Table 4.2-1.

One intermediate shell plate (82002-3) and one lower shell plate (82003-1) 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 42,4 48.3 ft-lb for intermediate shell plate 82002-3. An equivalent margins analysis performed in WCAP-13587, Rev. Idemonstrated 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 4?74 48.3 ft-lb for intermediate shell plate 82002-3 is above the 43 ft-lbs minimum acceptable USE for 4 loop plants determined in WCAP-13587 Rev. I.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 10CFR54.21(c)(l)(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 projected48 EFPYbelfiine fluenGand the beitliner&gion chemlstryarid surveiiIaice data including the un-irradiated CvUSE information as summarized in WCAP-16251, Analvsis of Caosule X from Enterav's lndian Point 3 Reactor Vessel Radiation Surveillance Proaram. The projected 48 EFPY peak beltline fluence level at the cladibase metal interface of 1.560Ec19 n/cm2was conservatively applied to all beftline materials. The 48 EFPY %t fluence level of 9.298E+18 n/cm2was calculated in accordance with Regulatovj Guide 1.99, Equation (3) based on a vesse! thickness of 8.625". The resulting projecied 48 EFPY CvUSE drop and resulting '/nt CvUSE are displayed in Table 4.2-2. All plate and weld beltline materials exceed 50 ft-lb at 48 EFPY a&+w

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

. . with the exceotion of the lower shell olate 82803-3 with a oredicted USE of 49.8 ft-lbs. As noted above for IP2. an equivalent marains analysis performed in WCAP-13587. Rev. Idemonstrated that the minimum acce~table USE for reactor vessel late material in 4 looo olants such as IP3 is 43 ft-lbs. Therefore, the IP3 lower shell olate 82803-3 USE value of 49.8 ft-lbs is acceotable. This determination is consistent with NUREG-1800, Section 4.2.2.1.1.2, and with the NRC Safetv Evaluation Reoort of acceotable 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)(l)(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 cladibase metal interface of 1.560E+19 nicm2was applied to all beltline materials. The resulting projected 48 EFPY R T ~ Tare s shown in Table 4.2-4. All projected FtTp~svalues are within the established screening criteria for 48 EFPY with the exception of plate B2803-3, which exceeds the screening criterion by &Q 9.5O F . Values of R T N ~for T the IP3 beltline materials at % T and 3 ?

T are summarized in Table 4.2-6.

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

Reactor Vessel Fluence Un-Material Vessel %Drop Location (Beltline 114T %Cu irradiated ldent in USE Identification) 48 EFPY USE 48 EFPY Inlermedtate shell

--- 1 B2802-1 Intermediate shell

/ lntermed~ateshell / 82802-3 Lower shell 1 82803-1 I-- T Lower shell 8 2 8 m - 2 Lower shell 82803-3 Linde 348009

-7,~

Lower shell axial 1092 Linde 1092 348009

-al C =

al 3E al.;

0

>m m

,-0 c

So-

3

=2

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

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)

Reactar Vessel Location (Beltline

NL-08-014 Attachment 1 Docket Nos. 50-247 & 50-286 Page 17 of 21 Table 4.2-5 IP2 Adjusted Reference Temperature at 48 Effective Full-Power Years (EFPY) 314 T Chemistry 114 T Neutro 48 Un- 314 T 48 EFPY Reactor Vessel Factor Neutron 114 T n EFPY Material Heat irradiated Fluenc 114 T Locatton (Beltline WGAP- Fluence Fluence Fluenc 314 T Number RTN~ (10'~ e RTNDT Identification) #xa Factor e Factor RTNDT ReK-l (" F, n/cm2) (1019 ("F) f"F) n/crn2) 7 lnterrnedlate shell 62002-1 1.036 0.404 0.748 169.1 136 3 lnterrnedlate shell / 62002-3 Lower shell 1 62003-2 lnterrned~ateshell 2-042 axla1 welds NBIC Lower shell axlal 3-042 AfB welds

NL-08-014 Attachment I 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)

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) 114 T 314 T Neutro Neutro 48 48 Un- 314 T Reactor Vessel Chemistr n 114 T 114 T n EFPY EPPY Material Heat irradiated Fluenc Location (Beltline y Factor Fluenc Fluence ARTNOT Fluenc 114 T 3/4 T

,dent Number RTNOT e Identifjcation) iwtm e Factor (" F) e Factor RTNDT RTN~T

(" F) (10" (10 (" F) (' F) n/cm2) nlcm2)

/ Intermediate shell / 82802-1 1 8-5394-2 / 137.0 1 5.0 / 0.930 / 0.980 1 134.2 1 0.330 / 0.695 / 95.2 / 173.2 / 134.2 1 Lower shell 82803-1 lntermed~ateshell / 2-042 /34BOO9~22l13 1-56 !0.930!0.980 /216.70.330!0.@5!~3-8--

1 / / / /

axial welds Lower shell axla1 13-0; 348009 221.3 -56 0.930 0.980 216.7 0.330 0.695 153 8 226.2 welds Intermediate to 9-042 13253 189.1 lower shell clrculnferent~al

NL-08-074 Attachment 1 Docket Nos. 50-247 & 59-286 Page 20 of 21 Table 4.2-6 IP3 Adjusted Reference Temperature at 48 Effective Full-Power Years (EFPY)

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-13587, 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 values are therefore acceptable since the lowest projected USE level for the beltline plate material through the period of extended operation of W $ft-lb , for intermediate shell plate 82002-3 is above the 43 ft-lbs minimum acceptable USE for four-loop plants determined in WCAP-13587 Rev. I.

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 114 T iiuence vaiues. Tine projected 48 EFPY peak beitline fluence level was conservatively applied to all beltline materials.

One lower shell olate (82803-3) has a ~iolecteduooer shelf enerav level below 50 ft-lb durina the ~ e r l 0 dof extended ooeratron. The CvUSE for all remalnina olate and weld beltline materials meets the acceotable value of areater than 50 ft-lb at 48 EFPY.

An eauivalent marains analvsis performed in WCAP-13587, Rev. 1. demonstrated that the minimum acceptable USE for reactor vessel oiate material in four-looo olants is 43 ftlbs. In the safetv assessment of WCAP-13587, the NRC concluded the reoort demonstrated marains of safetv eauivalent to those of the ASME code for beltline olate and foiainq materials. The USE value is therefore acce~tablesince the oroiected USE level throuah the period of extended ooeration of 49.8 ft-lb for lower shell plate 82003-3 is above the 43 ft-lbs minimum acceptable USE for four-looo olants determined in WCAP-13587 Rev. 1.

ATTACHMENT 2 TO NL-08-014 Audit Item #I05 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 ltem 105 Clarification The LRA amendment for Audit ltem 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 I I I

3. Parameters Monitored or lnswected

4. Detection of Aging Effects W& Revise applicable procedures to inspect the internal surface of the foam based fire suppression tanks. Acceptance 1

criteria will be enhanced to verify no

6. Acceptance Criteria 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 Requlatorv 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 SECTlOl I AUDIT ITEN 3 Enhance the Abovegroind Steel Tanks Progiam for IP2: NL-07-039 A.2.1 .I 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 first ten years of the period of extended operation. IP3:

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 materials for low susceptibility to SCC and clarify the 2013 B.1.2 prohibition on use of lubricants containing MoS2for bolting. IP3: NL-07-153 Audit items 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 lP2: NL-07-039 A.2.1.5 Program for IP2 and 1P3 as described in LRA Section September 28, A.3.1.5 B.1.6. 013 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 2015 Inspection.

NL-08-014 Attachment 3 Docket Nos. 50-247 & 50-286 Page 2 of 16 COMMITMENT IMPLEMENTAT101 SCHEDULE IP2:

Enhance the Diesel Fuel Monitoring Program to September 28, include cleaning and inspection of the IP2 GT-1 gas 2013 turbine fuel oil storage tanks, IP2 and IP3 EDG fuel oil day tanks, IP2 SBOIAppendix R diesel generator fuel IP3:

oil day tank, and IP3 Appendix R fuel oil storage tank December 12, and day tank once every ten years.

2015 Enhance the Diesel Fuel Monitoring Program to include quarterly sampling and analysis of the IP2 SBOIAppendix 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 10mgIl. 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: ED fuel oil storage tanks, EDG fuel oil day tanks, SBOIAppendix 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; lP3: 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 direc 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 COMMITMENT IMPLEMENTATIOI SCHEDULE LRA SECTIOI IP2:

Enhance the External Surfaces Monitoring Program September 28, A.3.1.10 for IP2 and IP3 to include periodic inspections of 2013 systems in scope and subject to aging management review for license renewal in accordance with 10 CFR 54,4(a)(l) and (a)(3), Inspections shall include areas December 12, surrounding the subject systems to identify hazards to 015 those systems. lnspections 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).

llP2: NL-07-039 A.2.1 .I1 Enhance the Fatigue hlonitoring Program for IP2 to A.3.1.11 monitor steady state cycles and feedwater cycles or 8.1.12, perform an evaluation to determine monitoring is not NL-07-153 Audit Item required. Review the number of allowed events and 164 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 COMMITMENT IIMPLEMENTATION SCHEDULE I SOURCE RELATED LRA SECTION 1P2:

Enhance the Fire Protection Program to inspect 28, A.3.1.12 external surfaces of the IP3 RCP otl collectton 8.1.13 systems for loss of material each refueling cycle.

Enhance the Fire Protection Program to explicitly state that the IP2 and IP3 diesel fire pump engine sub-systems (including the fuel supply line) shall be 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 C02fire 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 SOURCE RELATED LRA SECTIC f AUDIT ITEl A.2.1.13 A.3.1.13 8.1.14 Acceptance criteria will be revised to verify no Audit Item unacceptable signs of degradation.

105, 106 Enhance the Fire Water Program to replace all or test a sample of IP2 and IP3 sprinkler heads required for 10 CFR 50.48 using guidance of NFPA 25 (2002 edition), Section 5.3.1 .I .Ibefore 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 time!y 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.

NL-08-014 Attachment 3 Docket Nos. 50-247 & 50-286 Page 6 of 16 COMMITMENT IMPLEMENTATION SOURCE RELATED SCHEDULE LRA SECTIOE 9 Enhance the Flux Thimble Tube Inspection Program for IP2 and IP3 to implement comparisons to wear rates identified in WCAP-12866. Include provisions to compare data to the previous performances and I perform evaluations regarding change to test frequency and scope.

Enhance the Flux Thimble Tube lnspection 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 lnspection 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 IMPLEMENTATIO SCHEDULE LRA SECTlOr b

IP2:

Enhance the Heat Exchanger Monitoring Program for September 28, A.3.1.16 IP2 and IP3 to include the following heat exchangers 013 in the scope of the program.

Safety injection pump lube oil heat exchangers lP3:

RHR heat exchangers m

b;er 12.

RHR pump seal coolers Non-regenerative heat exchangers Charging pump seal water heat exchangers Charging pump fluid drive coolers I

Charging pump crankcase oil coolers Spent fuel pit heat exchangers I

Secondary system steam generator sample coolers Waste gas compressor heat exchangers II 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. Estabfi-acceptance criteria for-heat - --

exchangers visually inspected to include no unacceptable signs of degradation.

NL-08-014 Attachment 3 Docket Nos. 50-247 & 50-286 Page 8 of 16 COMMITMENT IMPLEMENTATION SOURCE RELATED SCHEDULE LRA SECTIOI I AUDIT ITEN IP2: NL-07-039 A.2.1.17 Enhance the IS1 Program for IP2 and IP3 to provide September 28, A.3.1.17 periodic visual inspections to confirm the absence of 2013 B.1.18 aging effects for lubrite sliding supports used in the NL-07-153 Audit item steam generator and reactor coolant pump support IP3: 59 systems.

December 12, 11~3:

December 12.

I, I 2015 Enhance the Metal-Enclosed Bus Inspection Program IP2:

to add IP2 480V bus associated with substation A to September 28, the scope of bus inspected. 2013 I5 NL-07-153 Audit Item Enhance the Metal-Enclosed Bus lnspection Program IP3: 124 for IP2 and IP3 to visually inspect the external surface December 12, Audit ltem of MEB enclosure assemblies for loss of material at 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 lnspection 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 Implement the Non-EQ Bolted Cable Connections 8.1.22.

NL-08-014 Attachment 3 Docket Nos. 50-247 & 50-286 Page 9 of 16 Section 6.1.24.

the corresponding program described in NUREG-LRA Section 6.1.25.

the corresponding program described in NUREG-

NL-08-014 Attachment 3 Docket Nos. 50-247 & 50-286 Page 10 of 16 COMMITMENT IMPLEMENTATIOI SOURCE RELATED SCHEDULE LRA SECTlOl IAUDIT ITEhi IP2: A.2.1.25 Enhance the Oil Analysis Program for lP2 to sample September 28, A.3.1.25 and analyze lubricating oil used in the SBOiAppendix 013 0.1.26 R diesel generator consistent with oil analysis for other site diesel generators.

IP3:

Enhance the Oil Analysis Program for IP2 and IP3 to December 12, sample and analyze generator seal oil and turbine 2015 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:

Implement the One-Time lnspection Program for IP2 September 28, and IP3 as described in LRA Section 8.1.27.

2013 8.1.27 This new program will be implemented consistent with Audit item the corresponding program described in NUREG- IP3: 173 1801.Section XI.M32. One-Time Ins~ection. December 12.

2015 IP2: NL-07-039 A.2.1.27 20 Implement the One-Time Inspection -Small Bore September 28, A.3.1.27 Piping Program for IP2 and IP3 as described in LRA 8.1.28 Section 8.1.28.

NL-07-153 Audit item 173 Maintenance Program for IP2 and IP3 as necessary to assurethat the effects ofaging will tie managed such that applicable components will continue to perform their intended functions consistent with the

NL-08-014 Attachment 3 Docket Nos. 50-247 & 50-286 Page 1Iof 16 I # i COMMITMENT llMPLEMENTATIONi SOURCE I RELATED I SCHEDULE LRA SECTION f AUDIT ITEM IP2: NL-07-039 A.2.1.31 22 Enhance the Reactor Vessel Surveillance Program for September 28, A.3.1.31 IP2 and IP3 revising the specimen capsule withdrawal 2013 8.1.32 schedules to draw and test a standby capsule to cover the peak reactor vessel fluence expected through the end of the period of extended operation.

Enhance the Reactor Vessel Surveillance Program for IP2 and IP3 to require that tested and untested specimens from all capsules pulled from the reactor vessel are maintained in storage.

IP2: NL-07-039 A.2.1.32 Implement the Selective Leaching Program for IP2 September 28, A.3.1.32 and IP3 as described in LRA Section 8.1.33.

2013 8.1.33 ,

This new program will be implemented consistent with the corresponding program described in NUREG- IP3:

1801,Section XI.M33 Selective Leachina - of Materials. December 12.

2015 iP2:

Enhance the Steam Generator Integrity Program for IP2 and IP3 to require that the results of the condition September 28, 2013 monitoring assessment are compared to the operational assessment performed for the prior IP3:

operating cycle with differences evaluated.

December 12, Enhance the Structures Monitoring Program to IP2: NL-07-039 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)

Audit item 86 Appendix R diesel generator switchgear and Audit item city water storage tank foundation 88 condensate storage tanks foundation (IP3)

Audit Item 87 fire pumphouse (IP2) fire protection pumphouse (IP3) fire water storage tank foundations (IP213) gas turbine 1 fuel storage tank foundation outage building-elevated

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

1. COMMITMENT 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) service water pipe chase (IP213) service water valve pit (IP3) superheater stack transformer/switchyard support structures (IP2) waste holdup tank pits (IP213)

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)

HVAC duct supports 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 wilt also inspect inaccessible concrete areas in environments where observed conditions in accessible areas expose& to 'the same environineny, 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 identify cracking and change in material properties

NL-08-014 Attachment 3 Docket Nos. 50-247 & 50-286 Page 13 of 16

1. COMMITMENT IMPLEMENTATION SOURCE RELATED SCHEDULE LRA SECTION IAUDIT 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 giound 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 noimally submerged concrete portions of the intake structures at least once every 5 years.

Implement the Thermal Aging Embrittlement of Cast IP2: NL-07-039 Austenitic Stainless Steel (CASS) Program for IP2 September 28, A.3.1.36 and IP3 as described in LRA Section B.1.37. 013 8.1.37 NL-07-153 Audit item This new program will be implemented consistent with IP3: 173 the corresponding program described in NUREG- December 12, 1801, Sect~onXI.Ml2, Thermal Aging Embrittlement 2015 of Cast Austenitic Stainless Steel (CASS) Program.

Implement the Thermal Aging and Neutron Irradiation IP2: NL-07-039 Embrittlement of Cast Austenitic Stainless Steel September 28, A.3.1.37 (CASS) Program for IP2 and IP3 as descr~bedin LRA 2013 8.1.38 Section 8.1.38. NL-07-153 Audit item lP3.

.. -. 173 This new program will be implemented consistent with December 12, the corresponding program described in NUREG- 2015 1801 Section XI.Ml3, Thermal Aging and Neutron Embrittlement of Cast Austenitic Stainless Steel (CASS) Program.

NL-08-014 Attachment 3 Docket Nos. 50-247 & 50-286 Page 14 of 16 COMMITMENT IMPLEMENTATlOl SCHEDULE LRA SECTlOl IP2:

Enhance the Water Chemistry Control - Closed September 28, A.3.1.39 Cooling Water Program to maintain water chemistry of 2013 the IP2 SBOJAppendix R diesel generator cooling system per EPRl 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 EPRl guidelines.

IrL.

Enhance the Water Chemistry Control - Primary and Secondary Program for IP2 to test sulfates monthly in September 28, 2013 the RWST with a limit of 4 5 0 ppb.

IP2:

For aging management of the reactor vessel internals, September 28, IPEC will (1) participate in the industry programs for investigating and managing aging effects on reactor 2011 internals; (2) evaluate and implement the results of IP3:

the industry programs as applicable to the reactor internals; and (3) upon completion of these programs, 12, 013 but not less than 24 months before entering the period of extended operation, subm~tan inspection plan for reactor internals to the NRC for review and approval.

IP2:

Additional P-T curves will be submitted as required September 28, per 10 CFR 50, Appendix G prior to the period of 2013 extended operation as part of the Reactor Vessel Surveillance Program.

IP3:

December 12.

2015 As required by 10 CFR 50.61 (b)(4), IP3 will submit a IP3:

plant-spectfic safety analysis for plate 82803-3 to the December 12, NRC three years prior to reaching the RTpsi 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.

NL-08-014 Attachment 3 Docket Nos. 50-247 & 50-286 Page 15 of 16

1. COMMITMENT IMPLEMENTATIOI SOURCE RELATED SCHEDULE LRA SECTlOl I AUDIT ITEFI IP2: NL-07-039 A.2.2.2.3 33 At least 2 years prior to entering the period of A.3.2.2.3 September 28, extended operation, for the locations identified in LRA 011 4.3.3 Table 4.3-13 (IP2) and LRA Table 4.3-14 (IP3), IP2 NL-07-153 Audit item and IP3 will implement one or more of the following:

IP3: 146 (1) Refine the fatigue analyses to determine valid CUFs less than 1 when accounting for the effects of reactor water environment. This ~ncludesapplying the appropriate Fen factors to valid CUFs determined in accordance with one of the following:

1. For locations, including NUREGICR-6260 locations, with existing fatigue analysis valid for the period of extended operation, use the existing CUF to deteimine the environmentally adjusted CUF. I

2. In addition to the NUREGICR-6260 locations, more limiting plant-specificlocations 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 analvsis using an NRC-approvedversion of the ASME code or htiC-approieo allernat ve (e g . , hRC-approved code case, may be padc~nlcolo aeteimlne a 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 o p t ~ o nto manage the aging effects due to environmental-assisted fatigue during the period of extended operation, deta~lsof the aging management program such as scope, qualification, method, and frequency will be submitted to the NRC at least 2 years prior 1 to the period of extended operation.

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

1. COMMITMENT IMPLEMENTATION SOURCE RELATED SCHEDULE LRA SECTION 1 AUDIT ITEM 34 April 30, 2008 NL-07-078 2.1.1.3.5 lP2 SBO i Appendix R diesel generator will be installed and operational by April 30, 2008. This committed change to the facility meets the requirements of 10 CFR 50.59(~)(1)and, therefore, a license amendment pursuant to 10 CFR 50.90 is not required.