ML16358A526

Amendment 18 to License Renewal Application

ML16358A526
Person / Time
Site:	Indian Point
Issue date:	12/15/2016
From:	Vitale A J Entergy Nuclear Northeast
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NL-16-138
Download: ML16358A526 (14)

Text

• *===* Entergx NL-16-138 December 15, 2016 U.S. Nuclear Regulatory Commission Document Control Desk 11545 Rockville Pike, TWFN-2 F1 Rockville, MD 20852-2738

SUBJECT:

Amendment 18 to License Renewal Application (LRA) Indian Point Nuclear Generating Unit Nos. 2 and 3 Docket Nos. 50-247 and 50-286 License Nos. DPR-26 and DPR-64 Entergy Nuclear Northeast Indian Point Energy Center 450 Broadway, GSB P.O. Box 249 Buchanan, NY 10511-0249 Tel (914) 254-2055 Anthony J Vitale Site Vice President

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 dated November 14, 2007, F. R. Dacimo to Document Control Desk, "Supplement to License Renewal Application (LRA) Environmental Report References" (NL-07-133)

Dear Sir or Madam:

In accordance with 10 CFR 54.21 (b), each year following submittal of the license renewal application and at least 3 months before scheduled completion of the NRC review, an amendment to the renewal application must be submitted that identifies any change to the CLB of the facility that materially affects the contents of the license renewal application (LRA), including the FSAR supplement.

In the referenced letters, Entergy Nuclear Operations, Inc. applied for renewal of the Indian Point Nuclear Generating Units 2 and 3 operating licenses.

This letter provides Amendment 18 of the Indian Poiryt Units 2 and 3 license renewal application.

There are no new commitments being made in this submittal.

Should you have any questions concerning this report, please contact Mr. Robert W. Walpole, Licensing Manager, at (914) 254-6710.

NL-16-138 Dockets 50-247 and 50-286 Page 2 of 2 . de;lare !under penalty of perjury that the foregoing is true and correct. Executed on

'2016. Sincerely, AJV/rl Attachment

1. Annual Update Amendment cc: Mr. Daniel H. Dorman, Regional Administrator, NRC Region I Mr. William Burton, NRC Senior Project Manager, Division of License Renewal Mr. Douglas Pickett, NRR Senior Project Manager Ms. Bridget Frymire, New York State Department of Public Service Mr. John B. Rhodes, President and CEO NYSERDA NRC Resident Inspector's Office ATTACHMENT 1 TO NL-16-138 ANNUAL AMENDMENT ENTERGY NUCLEAR OPERATIONS, INC. INDIAN POINT NUCLEAR GENERATING UNIT NOS. 2 and 3 DOCKET NO. 50-247 and 50-286 NL-16-138 Dockets 50-247 and 50-286 Attachment 1 Page 1 of 11 INDIAN POINT NUCLEAR GENERATING UNIT Nos. 2 AND 3 LICENSE RENEWAL APPLICATION ANNUAL AMENDMENT In accordance with 1 O CFR 54.21 (b), each year following submittal of the license renewal application and at least 3 months before scheduled completion of the NRC review, an amendment to the renewal application must be submitted that identifies any change to the CLB of the facility that materially affects the contents of the license renewal application (LRA), including the FSAR supplement.

This attachment is the required annual amendment to the LRA. Amendment 18 is based on a review of documents potentially affecting the CLB during the periods of September 1, 2015 through August 31, 2016. The review concluded that certain sections of the LRA are affected by changes to CLB documents and other related LRA reviews. The table below summarizes the changes listing the affected system (if applicable), an explanation of the change (including effect on the LRA), and the affected LRA section. Affected LRA Sections Change IP3 -EC 46625 Permanently repaired pipe (line # 1033) with a pressure retaining clamp and using CSI Pipe Wrap-Ply material (carbon fiber) applied on the exterior circumference of the pipe and clamp. IP2-EC 58208 Install metal flex hoses on IP2 hydrogen cooler vent and drain lines and upgraded threaded pipe nipple from SCH 40 to SCH 80 IP2-EC 58594 Re-route chlorine piping for the north service water bay LRA Section Affected Table 3.3.2-17-IP3 Table 2.3.3-19-39-IP2 Table 3.3.2-19-39-IP2 Table 3.3.2-19-44-IP2 I NL-16-138 Dockets 50-247 and 50-286 Attachment 1 Change IP3 -EC 59472 & NL-15-014

& RA-15-115 LRA Section 4.2.3 states "Technical Specifications contain pressure/temperature limits valid through 34 EFPY ... ". This EC extends the PIT limits to a lifetime burnup of 37 EFPY. LRA Section 4.2.3 is revised to eliminate "valid through 34 EFPY." The validity period for the PIT limits is stated in the Technical Specification section containing the limits. The period of validity can change during normal operations as governed by 10 CFR 50.61 requirements.

There is no need to state in the LRA the EFPY for which the limits are applicable.

It is sufficient to indicate that the limits, and the period for which the limits are valid, are found in the Technical Specifications.

IP3 -EC 60109 Evaluation to change the Unit 3 FSAR hot penetration cooling criteria using prior Unit 2 Safety Evaluation 94-238-EV and calculation GCC-00021 CR-I P2-2016-5826 Correction of 2014 IPEC Annual Update(IP-RPT-14-LRD01 Rev 0) related to EC 42090 Page 2of11 I LRA Section Affected I Section 4.2.3 Section 3.5.2.2.1.3 Section 3.5.2.2.2.3 Section 2.3.3.4 Section 2.3.2.5 Table 3.2.2-5-IP2 NL-16-138 Dockets 50-247 and 50-286 Attachment 1 Change Appendix A UFSAR Appendix B Aging Management Programs and Activities Typographical corrections (revise 'cumulative use factor' to 'cumulative usage factor' to agree with Supplement 2 of the Safety Evaluation Report, November 2014) Page 3of11 LRA Section Affected A.2.1.11 8.1.12 NL-16-138 Dockets 50-247 and 50-286 Attachment 1 Page 4of11 IPEC LRA changes are shown below. (Changes are shown as strikethroughs for deletions and underlines for additions)

Table 3.3.2-17-IP3 City Water Summary of Aging Management Review Table 3.3.2-17-IP3:

City Water Component Intended Aging Effect Aging NU REG-Table 1 Material Environment Requiring Management 1801 Vol. Notes Type Function Management Programs 2 Item Item ----Pressure E 5Qjng Fiberglass

--bounda!Y Soil (ext) None None 3.5.2.2.1.3 Reduction of Strength and Modulus of Concrete Structures due to Elevated Temperature For Unit 3 containment during normal operation, areas are maintained below a bulk average temperature of 130°F. Piping penetrations through the containment cylinder wall associated with pipes carrying hot fluid are cooled using air-to-air heat exchangers and the pipes are insulated to maintain the temperature in the adjoining concrete below 2GG250°F.

NUREG-1801 allows for concrete temperatures higher than 200°F for local areas if tests or calculations are provided to evaluate the reduction in strength.

Concrete associated with the Unit 3 hot piping penetrations has been evaluated and determined acceptable at temperatures up to 250°F.

NL-16-138 Dockets 50-247 and 50-286 Attachment 1 3.5.2.2.2.3 Reduction of Strength and Modulus of Concrete Structures due to ElevatedTemperature Page 5of11 For reduction of strength and modulus of concrete structures due to elevated temperatures for Groups 1-5, NUREG-1801 recommends a plant-specific AMP and further evaluation if the general temperature is greater than 150°F or if the local temperature is greater than 200°F. During normal operation, bulk average temperature of Groups 1-5 concrete elements is maintained below 150°F and local temperatures remain below 200250°F.

NUREG-1801 allows for concrete temperatures higher than 200°F for local areas if tests or calculations are provided to evaluate the reduction in strength.

Concrete associated with the hot piping Unit 3 penetrations has been evaluated and determined acceptable at temperatures up to 250°F. Table 2.3.3-19-39-IP2 Service Water System Nonsafety-related Components Potentially Affecting Safety Functions Subject to Aging Management Review Component Type Intended Function Flex joint Pressure boundary NL-16-138 Dockets 50-247 and 50-286 Attachment 1 Table 3.3.2-19-39-IP2 Service Water System Nonsafety-Related Components Potentially Affecting Safety Functions Summary of Aging Management Review Page 6of11 Table 3.3.2-19-39-IP2:

Service Water System Component Intended Aging Effect Aging NU REG-Table 1 Type Function Material Environment Requiring Management 1801 Vol. Item Management Programs 2 Item Vll.F1-1 Flex joint Pressure Stainless Condensation Loss of External Surfaces 3.3.1-27 boundar:y steel (ext) material Monitoring (A-09) Notes £ NL-16-138 Dockets 50-247 and 50-286 Attachment 1 Page 7of11 Table 3.3.2-19-44-IP2 Chlorination System Nonsafety-Related Components Potentially Affecting Safety Functions Summary of Aging Management Review Table 3.3.2-19-44-IP2:

Chlorination System Component Intended Aging Effect Aging NU REG-Type Function Material Environment Requiring Management 1801 Vol. Management Programs 2 Item Piping Pressure Plastic Air-indoor None None boundary (ext) ---Piping Pressure Plastic Treated water boundary (int) None None ---Valve body Pressure Plastic Air-indoor boundary (ext) None None ---Valve body Pressure Plastic Treated water boundary (int) None None ---4.2.3 Pressure-Temperature Limits Table 1 Item ------------Notes ------------Technical Specifications contain pressure/temperature limits valid through 34EFPY including the effects of the stretch power uprate. At present, plate 82803-3 with an initial RT NDT of 7 4 °F restricts operation (P-T) in the 150-250°F range. Resolution to the P-T operating window is a current operating term issue and will be resolved three years prior to reaching the RT PTS screening criterion per 10 CFR 50.61 requirements.

NL-16-138 Dockets 50-247 and 50-286 Attachment 1 Page 8of11 The P-T limit curves will continue to be updated, as required by Appendix G of 1 O CFR Part 50 or as operational needs dictate. This . updating will assure that the operational limits remain valid through the period of extended operation.

Additional P-T limit analysis is not required at this time. Maintaining the P-T limit curves in accordance with Appendix G of 10 CFR 50 assures that the effects of aging on the intended function(s) will be adequately managed for the period of extended operation consistent with 10 CFR 54.21 (c)(1 )(iii). 2.3.2.5 Containment Penetrations System Descrjption Containment penetrations have the following intended function for 1 OCFR54.4( a)( 1 ).

• Prevent release of radioactivity to outside environment.

Containment penetrations have no intended functions for 10CFR54.4(a)(2).

or (a)(3). Containment penetrations have the following intended funstion for 10CFR54.4(a)(3).

• Provide optional eapabilitv.

located in the CCR to bypass the existing manual containment isolation valves allO'vving the instrument air flow to reaeh the Fan Mouse Safe Shut Down Panel in ease of fire inside containment to support the Alternate Safe Shutdmun System U\SSS).

NL-16-138 Dockets 50-247 and 50-286 Attachment 1 Table 3.2.2-5-IP2 Containment Penetrations Summary of Aging Management Review Table 3.2.2-5-IP2:

Containment Penetrations Componen Intended Aging Effect Aging tType Function Material Environment Requiring Management Management Programs PFess1:1Fe Stai Riess c,* , <IF tFeateEI Nooe Nooe ee1:1RElapt steel fffi!t Val>.ie eeEl>r PFess1:1Fe Stai Riess A" r<IF tFeateEI Nooe Nooe ee1:1RElaPr steel fiffi1 Page 9of11 NUREG-1801 Vol. 2 Item Table 1 Item Notes --G --G 2.3.3.4 Compressed Air System Description NL-16-138 Dockets 50-247 and 50-286 Attachment 1 The compressed air systems include the instrument air (IA) and station air (SA) subsystems.

Instrument Air The IA system has the following intended functions for 10 CFR 54.4(a)(3).

Page 10of11

• Provide a source of compressed gas for pneumatically operated components for the Appendix R event (10 CFR 50.48). *Enable charging pump 21 to be run at high speed without the need for operator actions outside the CCR in response to a fire for the Appendix R event (1 O CFR 50.48).

• Provide optional capability.

located in the CCR. to bypass the existing manual containment isolation valves allowing the instrument air flow to reach the Fan House Safe Shut Down Panel in case of fire inside containment to support the Alternate Safe Shutdown System (ASSS). *Support safe shutdown in the event of a fire in the auxiliary feed pump room (10 CFR 50.48) (see Section 2.3.4.5).

A.2.1.11 Fatigue Monitoring Program The Fatigue Monitoring Program is an existing program that tracks the number of critical thermal and pressure transients for selected reactor coolant system components.

The program ensures the validity of analyses that explicitly analyzed a specified number of fatigue transients by assuring that the actual effective number of transients does not exceed the analyzed number of transients.

The program provides for update of the fatigue usage calculations to maintain a cumulative use usage factor fCUFl of< 1.0 for the period of extended operation.

For the locations identified in Section A.2.2.2.46, updated calculations will account for the effects of the reactor water environment.

8.1.12 Fatigue Monitoring Program Description NL-16-138 Dockets 50-247 and 50-286 Attachment 1 Page 11 of 11 The Fatigue Monitoring Program is an existing program that tracks the number of critical thermal and pressure transients for selected reactor coolant system components.

The program ensures the validity of analyses that explicitly analyzed a specified number of fatigue transients by assuring that the actual effective number of transients does not exceed the analyzed number of transients.

The program provides for update of the fatigue usage calculations to maintain a cumulative usage factor {CU Fl of < 1.0 for the period of extended operation.

For the locations identified in LRA Table 4.3-13 (IP2) and LRA Table 4.3-14 (IP3}, updated calculations will account for the effects of the reactor water environment.

These calculation updates are governed by Entergy's 1 O CFR 50 Appendix B Quality Assurance (QA) program and include design input verification and independent reviews ensuring that valid assumptions, transients, cycles, external loadings, analysis methods, and environmental fatigue life correction factors will be used in the fatigue analyses.