ML14045A248

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Issuance of Amendment Pressure-Temperature Limit Curves and Low Temperature Over Pressure Requirements
ML14045A248
Person / Time
Site: Indian Point 
Issue date: 03/05/2014
From: Pickett D
Plant Licensing Branch 1
To:
Entergy Nuclear Operations
Pickett D
References
TAC MF0634, FOIA/PA-2016-0148
Download: ML14045A248 (29)
v • d • e


Text

{{#Wiki_filter:UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 Vice President, Operations Entergy Nuclear Operations, Inc. Indian Point Energy Center 450 Broadway, GSB P.O. Box 249 Buchanan, NY 10511-0249 March 5, 2014

SUBJECT:

INDIAN POINT NUCLEAR GENERATING UNIT NO. 2-ISSUANCE OF AMENDMENT RE: PRESSURE-TEMPERATURE LIMIT CURVES AND LOW TEMPERATURE OVER PRESSURE REQUIREMENTS (TAC NO. MF0634)

Dear Sir or Madam:

The Commission has issued the enclosed Amendment No. 274 to Facility Operating License No. DPR-26 for the Indian Point Nuclear Generating Unit No. 2. The amendment consists of changes to the Technical Specifications (TSs) in response to your application dated February 6, 2013, as supplemented on July 9, 2013, October 3, 2013, and February 24, 2014. The amendment changes the TSs by revising the reactor heatup and cooldown curves (also referred to as pressure-temperature limits) and low temperature overpressure protection requirements to cover a lifetime burnup of 48 Effective Full Power Years (EFPY), which is an increase from the current value of 29.2 EFPY. A copy of the related Safety Evaluation is enclosed. A Notice of Issuance will be included in the Commission's next regular biweekly Federal Register notice. Docket No. 50-247

Enclosures:

1. Amendment No. 274 to DPR-26
2. Safety Evaluation cc w/encls: Distribution via Listserv Sincerely, Douglas V. Pickett, Senior Project Manager Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 ENTERGY NUCLEAR INDIAN POINT 2. LLC ENTERGY NUCLEAR OPERATIONS. INC. DOCKET NO. 50-247 INDIAN POINT NUCLEAR GENERATING UNIT NO.2 AMENDMENT TO FACILITY OPERATING LICENSE AND TECHNICAL SPECIFICATIONS

1. The Nuclear Regulatory Commission (the Commission) has found that:

Amendment No. 27 4 License No. DPR-26 A. The application for amendment by Entergy Nuclear Operations, Inc. (Entergy, or the licensee) dated February 6, 2013, as supplemented on July 9, 2013, October 3, 2013, and February 24, 2014, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-26 is hereby amended to read as follows: (2) Technical Specifications The Technical Specifications contained in Appendices A, Band C, as revised through Amendment No. 274, are hereby incorporated in the license. ENO shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of the date of its issuance and shall be implemented within 60 days.

Attachment:

Changes to the License and Technical Specifications Date of Issuance: March 5, 2014 FOR THE NUCLEAR REGULATORY COMMISSION Benjamin G. Beasley, Chief Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

ATTACHMENT TO LICENSE AMENDMENT NO. 274 FACILITY OPERATING LICENSE NO. DPR-26 DOCKET NO. 50-247 Replace the following page of the License with the attached revised page. The revised page is identified by amendment number and contains marginal lines indicating the areas of change. Remove Page Insert Page 3 3 Replace the following pages of the Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change. Remove Pages 3.4.3-3 3.4.3-4 3.4.6-1 3.4.7-1 3.4.10-1 3.4.12-1 3.4.12-3 3.4.12-5 3.4.12-6 3.4.12-7 3.4.12-8 3.4.12-9 3.4.12-10 3.4.12-11 3.4.12-12 3.4.12-13 Insert Pages 3.4.3-3 3.4.3-4 3.4.6-1 3.4.7-1 3.4.10-1 3.4.12-1 3.4.12-3 3.4.12-5 3.4.12-6 3.4.12-7 3.4.12-8 3.4.12-9 3.4.12-10 3.4.12-11 3.4.12-12 3.4.12-13

(4) (5) instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; ENO pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; ENO pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility and Indian Point Nuclear Generating Unit No. 3 (IP3). Arndt. 42 10-17-78 Arndt. 220 09-06-01 C. This amended license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter 1: Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below: (1) Maximum Power Level ENO is authorized to operate the facility at steady state reactor core power levels not in excess of 3216 megawatts thermal. (2) Technical Specifications Arndt. 241 10-27-2004 The Technical Specifications contained in Appendices A, B, and C, as revised through Amendment No. 27 4, are hereby incorporated in the license. ENO shall operate the facility in accordance with the Technical Specifications. (3) The following conditions relate to the amendment approving the conversion to Improved Standard Technical Specifications:

1.

This amendment authorizes the relocation of certain Technical Specification requirements and detailed information to licensee-controlled documents as described in TableR, "Relocated Technical Specifications from the CTS," and Table LA, "Removed Details and Less Restrictive Administrative Changes to the CTS" attached to the NRC staff's Safety Evaluation enclosed with this amendment. The relocation of requirements and detailed information shall be completed on or before the implementation of this amendment. Amendment No. 27 4

2500 2000 Cl .e: Q)... 1500

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  • Applicable to 48 EFPY and during vacuum fill

"~I - No Allowance for Instrument Error or Pressure Bias Test ~ -Heatup Rate - 60 *F/Hr I - Minimum bolt-up temperature remains applicable during Umit -Heatup Rate-100 *F/Hr vacuum fill J I Unacceptable Operation 1 1 ~ (To Left or Above Applicable Curve) I ? If ~ I Acceptable Operation I (To Right or Below Applicable Curve) ~ Bolt-up Temperature I 60 °F 50 100 150 200 250 300 350 400 450 500 RCS Temperature (OF) Figure 3.4.3-1 : Heatup Limitations for the Reactor Coolant System (RCS) and Hydrostatic and lnservice Leak Testing Limitations for the RCS. 550 INDIAN POINT 2 3.4.3-3 Amendment No. 274

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")( ta 500 J I Unacceptable Operation , I I (To Left or Above Applicable Curve) ¥ ~, ~ Bolt-up Temperature I 60 ' F 0 0 50 100 150 200 250 300 RCS PIT Limits 3.4.3 cooldown Rate - Steady State -cooldown Rate - 20 ' F/Hr cooldown Rate - 40 ' F/Hr cooldown Rate - 60 ' F/Hr f-- cooldown Rate-100 ' F/Hr I Acceptable Operation (To Right or Below Applicable Curve) 350 400 450 500 550 RCS Temperature (°F) Figure 3.4.3-2: Cooldown Limitations for the RCS (including RCS cooldown following RCS inservice leak and hydrostatic testing) INDIAN POINT 2 3.4.3-4 Amendment No. 274

1 RCS Loops - MODE 4 3.4.6 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.6 RCS Loops - MODE 4 LCO 3.4.6 APPLICABILITY: ACTIONS CONDITION Two loops consisting of any combination of RCS loops and residual heat removal (RHR) loops shall be OPERABLE, and one loop shall be in operation. -NOTES-

1.

All reactor coolant pumps (RCPs) and RHR pumps may be removed from operation for~ 1 hour per 8 hour period provided:

a.

No operations are permitted that would cause introduction into the RCS of any coolant with a boron concentration less than that required to meet the SDM of LCO 3.1.1, and

b.

Core outlet temperature is maintained at least 1 0°F below saturation temperature.

2.

No RCP shall be started with any RCS cold leg temperature ~ 288°F unless the requirements for RCP starting in LCO 3.4.12 are met. MODE 4. REQUIRED ACTION COMPLETION TIME A. One required loop inoperable. A.1 Initiate action to restore a second loop to OPERABLE status. Immediately A.2 INDIAN POINT 2 -NOTE-Only required if RHR loop is OPERABLE. Be in MODE 5. 3.4.6-1 24 hours Amendment No. 27 4

RCS Loops - MODE 5, Loops Filled 3.4.7 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.7 RCS Loops-MODE 5, Loops Filled LCO 3.4.7 APPLICABILITY: INDIAN POINT 2 One residual heat removal (RHR) loop shall be OPERABLE and in operation, and either:

a.

The non-operating RHR loop shall be OPERABLE or

b.

The secondary side water level of at least two steam generators (SGs) shall be~ 0% narrow range. -NOTES-

1.

The RHR pump of the loop in operation may be removed from operation for~ 1 hour per 8 hour period provided:

a.

No operations are permitted that would cause introduction into the RCS of any coolant with a boron concentration less than that required to meet the SDM of LCO 3.1.1; and

b.

Core outlet temperature is maintained at least 1 oaF below saturation temperature.

2.

One required RHR loop may be inoperable for up to 2 hours for surveillance testing provided that the other RHR loop is OPERABLE and in operation.

3.

No reactor coolant pump shall be started unless the requirements for RCP starting in LCO 3.4.12 are met.

4.

All RHR loops may be removed from operation during planned

  • --------~~~-~~E-~~-~-Q~.§-~-~~~~--~!-~~~~-~~~-~-!3_g_~-l~~R-~~-~~-~P-~!~!~~~: ________.

MODE 5 with RCS Loops Filled 3.4.7-1 Amendment No. 27 4

Pressurizer Safety Valves 3.4.1 0 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.1 0 Pressurizer Safety Valves LCO 3.4.10 Three pressurizer safety valves shall be OPERABLE with lift settings set ~ 2460 psig and ::; 2510 psig. APPLICABILITY: MODES 1, 2, and 3, MODE 4 with all RCS cold leg temperatures > 288°F. -NOTE-The lift settings are not required to be within the LCO limits during MODES 3 and 4 for the purpose of setting the pressurizer safety valves under ambient (hot) conditions. This exception is allowed for 54 hours following entry into MODE 3 provided a preliminary cold setting was made .P_~i~~-!<?_~~-~!~P-*--------------------------------------------------------------------------------

  • ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

One pressurizer safety A.1 Restore valve to 15 minutes valve inoperable. OPERABLE status. B. Required Action and B.1 Be in MODE 3. 6 hours associated Completion Time not met. AND OR B.2 Be in MODE 4 with any 24 hours Two or more pressurizer RCS cold leg temperature

L TOP Applicability safety valves inoperable.

temperature specified in LCO 3.4.12. INDIAN POINT 2 3.4.1 0- 1 Amendment No. 274

LTOP 3.4.12 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.12 Low Temperature Overpressure Protection (L TOP) LCO 3.4.12 APPLICABILITY: INDIAN POINT 2 L TOP shall be OPERABLE in accordance with one of the options in Table 3.4.12-1 and the accumulators shall be isolated. -NOTES-

1.

Accumulator isolation is only required when accumulator pressure is greater than or equal to the maximum RCS pressure for the coldest existing RCS cold leg temperature allowed by the PIT limit curves provided in Figure 3.4.12-1.

2.

If conditions require the use of High Head Safety Injection (HHSI) pumps in the event of loss of RCS inventory, the pumps can be made capable of injecting into the RCS.

3.

One HHSI pump may be made capable of injecting into the RCS as needed to support abnormal operations such as emergency boration or response to loss of RHR cooling.

4.

SR 3.4.12.8 shall be met prior to starting a reactor coolant pump (RCP) if no other RCPs are in operation. MODE 4 when any RCS cold leg temperature is~ 288°F, MODE 5, MODE 6 when the reactor vessel head is on. 3.4.12-1 Amendment No. 27 4

ACTIONS (continued) CONDITION C. Required Action and C.1 associated Completion Time of Condition B not met. OR C.2 D. One required PORV D.1 inoperable. E. Two required PORVs E.1 inoperable. OR Required Action and associated Completion Time of Condition A or D AND not met. E.2 OR L TOP inoperable for any reason other than Condition A, B, Cor D. INDIAN POINT 2 REQUIRED ACTION Increase RCS cold leg temperature to > 288°F. Depressurize affected accumulator to less than the maximum RCS pressure for existing cold leg temperature allowed in Figure 3.4.12-1. Restore required PORV to OPERABLE status. Initiate action to reduce the number of HHSI pumps and charging pumps capable of injecting into the RCS consistent with Table 3.4.12-1. Depressurize RCS and establish RCS vent required by Table 3.4.12-1 for existing plant conditions. 3.4.12-3 LTOP 3.4.12 COMPLETION TIME 12 hours 12 hours 7 days Immediately 8 hours Amendment No. 27 4

SURVEILLANCE REQUIREMENTS (continued) SR 3.4.12.4 SR 3.4.12.5 SR 3.4.12.6 SURVEILLANCE -NOTE-Not required to be met when Table 3.4.12-1, Option A, B, C, D or E is met. Verify required RCS vent meets the following:

a.
b.
2.00 square inches (or 1 PORV blocked fully open) when required by Table 3.4.12-1, Option F;
5.00 square inches (or 2 PORVs blocked fully open) when required by Table 3.4.12-1, Option G or H; or
c.
5.00 square inches when required by Table 3.4.12-1, Option I.

Verify PORV block valve is open for each required PORV. -NOTE-Not required to be performed until 12 hours after decreasing RCS cold leg temperature to~ 288°F. Perform a COT on each required PORV, excluding actuation. LTOP 3.4.12 FREQUENCY 24 hours for unlocked open vent valve(s) 31 days for locked open vent valve(s) 72 hours 31 days SR 3.4.12.7 Perform CHANNEL CALIBRATION for each required 24 months PORV actuation channel. I NOlAN POl NT 2 3.4.12 - 5 Amendment No. 27 4

SR 3.4.12.8 INDIAN POINT 2 SURVEILLANCE LTOP 3.4.12 FREQUENCY Verify the conditions in one of the following options is Within 30 minutes satisfied prior to starting any RCP: prior to starting A. 1 Temperature of all steam generators (SGs) ~ RCS temperature. 8.1 Two PORVs with lift settings within Figure 3.4.12-1 limits are OPERABLE; and 8.2 Temperature of all SGs :5 40°F higher than the RCS temperature; and 8.3 RCS temperature is as shown in Table on Figure 3.4.12-1; and 8.4 Pressurizer level ;:::: 30% and :5 85% of span. OR C.1 Temperature of all SGs :5 40°F higher than the RCS temperature; and C.2 RCS pressure, temperature and pressurizer level within limits specified in Figure 3.4.12-5 for RCP pump start with SGs :5 40°F higher than the RCS temperature. OR 0.1 Temperature of all SGs :5 1 00°F higher than the RCS temperature; and 0.2 RCS Pressure, temperature and pressurizer level within limits specified in Figure 3.4.12-6 for RCP pump start with SGs :5 1 00°F higher than the RCS temperature. 3.4.12-6 any RCP if no other RCP is operating Amendment No. 27 4

Table 3.4.12-1 Options for LTOP OPERABILITY LTOP Relief Capability or Maximum Injection Option Vent Size Capability HHSI Charging Pumps Pumps A. 2 OPERABLE PORVs with 0

53 setpoint specified in Figure 3.4.12-1 B.

2 OPERABLE PORVs with

51
52 setpoint specified in Figure 3.4.12-1 C.

None 0

5 1 D.

None 0

52 E.

None 0

53 F.

~ 2 square inch vent

5 1
53 (1 PORV blocked open)

G. ~ 5 square inch vent

52
53 (2 PORVs blocked open)

H. ~ 5 square inch vent

53
52 (2 PORVs blocked open)

I. ~ 5 square inch vent

53
53 INDIAN POINT 2 3.4.12-7 LTOP 3.4.12 Restrictions on RCS Temperature, Pressure and Pressurizer Level None None As Specified in Figure 3.4.12-2 As Specified in Figure 3.4.12-3 As Specified in Figure 3.4.12-4.

None None None None Amendment No. 27 4

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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 274 TO FACILITY OPERATING LICENSE NO. DPR-26 ENTERGY NUCLEAR OPERATIONS. INC. INDIAN POINT NUCLEAR GENERATING UNIT NO. 2 DOCKET NO. 50-247

1.0 INTRODUCTION

By letter dated February 6, 2013 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML13052A018), as supplemented by letters dated July 9, 2013 (ADAMS ML13197A218), October 3, 2013 (ADAMS ML13281A498), and February 24, 2014, Entergy Nuclear Operations, Inc. (Entergy, or the licensee) submitted a request for changes to the Indian Point Nuclear Generating Unit No. 2 (IP2) Technical Specifications (TSs). The amendment changes the TSs by revising the reactor heatup and cooldown curves (also referred to as pressure-temperature (P-T) limits) and low temperature overpressure protection (L TOP) requirements to cover a lifetime burnup of 48 Effective Full Power Years (EFPY), which is an increase from the current value of 29.2 EFPY. The supplements dated July 9, 2013, October 3, 2013, and February 24, 2014, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the Nuclear Regulatory Commission (NRC) staff's original proposed no significant hazards consideration.

2.0 REGULATORY EVALUATION

The following explains the applicability of General Design Criteria (GDC) for IP2. The construction permit for IP2 was issued by the Atomic Energy Commission (AEC) on October 14, 1966, and the operating license was issued on September 28, 1973. The plant GDC are discussed in the Updated Final Safety Analysis Report (UFSAR) Chapter 1.3, "General Design Criteria," with more details given in the applicable UFSAR sections. The AEC published the final rule that added Title 10 of the Code of Federal Regulations (1 0 CFR) Part 50, Appendix A, "General Design Criteria for Nuclear Power Plants," in the Federal Register (36 FR 3255) on February 20, 1971, with the rule effective on May 21, 1971. In accordance with an NRC staff requirements memorandum from S. J. Chilk to J. M. Taylor, "SECY-92-223-Resolution of Deviations Identified During the Systematic Evaluation Program," dated September 18, 1992 (ADAMS Accession No. ML003763736), the Commission decided not to apply the Appendix A GDC to plants with construction permits issued prior to May 21, 1971. Therefore, the GDC which constitute the licensing bases for IP2 are those in the UFSAR. As discussed in the UFSAR, the licensee for IP2 has made some changes to the facility over the life of the unit that has committed to some of the GDCs from 10 CFR Part 50, Appendix A. The extent to which the Appendix A GDC have been invoked can be found in specific sections of the UFSAR and in other IP2 licensing basis documentation, such as license amendments. The NRC has established requirements in Part 50 of Title 10 of the Code of Federal Regulations ( 10 CFR Part 50) to protect the integrity of the reactor coolant pressure boundary in nuclear power plants. The staff evaluates the acceptability of a facility's proposed P-T limits based on the following NRC regulations and guidance: Appendix G, "Fracture Toughness Requirements," to 10 CFR Part 50; Appendix H, "Reactor Vessel Material Surveillance Program Requirements," to 10 CFR Part 50; Regulatory Guide (RG) 1.99, Revision 2 (Rev. 2), "Radiation Embrittlement of Reactor Vessel Materials;" Generic Letter (GL) 92-01, Rev. 1, "Reactor Vessel Structural Integrity;" GL 92-01, Rev. 1, Supplement 1, "Reactor Vessel Structural Integrity;" and Standard Review Plan (SRP) Section 5.3.2, "Pressure-Temperature Limits, Upper-Shelf Energy, and Pressurized Thermal Shock" (NUREG-0800- ADAMS Accession No. ML062130599). Appendix G to 10 CFR Part 50 also provides minimum temperature requirements that must be considered in the development of the P-T limit curves and requires that facility P-T limits for the reactor pressure vessel (RPV) be at least as conservative as those obtained by applying the linear elastic fracture mechanics methodology of Appendix G to Section XI of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code). RG 1.99, Rev. 2, contains methodologies for determining the increase in transition temperature and the decrease in upper-shelf energy resulting from neutron radiation. GL 92-01, Rev. 1 requested that licensees submit the RPV data for their plants to the staff for review, and GL 92-01, Rev. 1, Supplement 1, requested that licensees provide and assess data from other licensees that could affect their RPV integrity evaluations. The most recent version of Appendix G to Section XI of the ASME Code which has been endorsed in 10 CFR 50.55a, and therefore by reference in 10 CFR Part 50, Appendix G, is the 2008 Edition of the ASME Code. This edition of Appendix G to Section XI of the ASME Code incorporates the provisions of ASME Code Case N-588, "Attenuation to Reference Flaw Orientation of Appendix G for Circumferential Welds in Reactor Vessels," and ASME Code Case N-640, "Alternative Reference Fracture Toughness for Development of P-T Limit Curves." Additionally, Appendix G to 10 CFR Part 50 imposes minimum head flange temperatures when system pressure is at or above 20 percent of the preservice hydrostatic test pressure. SRP Section 5.3.2 provides an acceptable method of determining the P-T limit curves for ferritic materials in the beltline of the RPV based on the linear elastic fracture mechanics methodology of Appendix G to Section XI of the ASME Code. The basic parameter of this methodology is the stress intensity factor K1, which is a function of the stress state and flaw configuration. ASME Code, Section XI, Appendix G requires a safety factor of 2.0 on stress intensities resulting from reactor pressure during normal and transient operating conditions, and a safety factor of 1.5 on these stress intensities for hydrostatic testing curves. The flaw postulated in the ASME Code, Section XI, Appendix G has a depth that is equal to 1/4 of the RPV beltline thickness (1/4T) and a length equal to 1.5 times the RPV beltline thickness. The critical locations in the RPV beltline region for calculating heatup and cooldown P-T limit curves are the 1/4T and 3/4 thickness (3/4T) locations, which correspond to the maximum depth of the postulated inside surface and outside surface defects, respectively. The methodology found in Appendix G to Section XI of the ASME Code requires that licensees determine the adjusted reference temperature (ART or adjusted RT NoT) by evaluating material property changes due to neutron radiation. The ART is defined as the sum of the initial (unirradiated) reference temperature (initial RT NoT), the mean value of the adjustment in reference temperature caused by irradiation (~RT NoT) and a margin (M) term. The ~RT NOT is a product of a chemistry factor (CF) and a fluence factor. The CF is dependent upon the amount of copper and nickel in the material and may be determined from tables in RG 1. 99, Rev. 2, or from surveillance data. The fluence factor is dependent upon the neutron fluence at the maximum postulated flaw depth. The margin term is dependent upon whether the initial RT NoT is a plant-specific or a generic value and whether the CF was determined using the tables in RG 1.99, Rev. 2, or surveillance data. The margin term is used to account for uncertainties in the values of the initial RT NoT, the copper and nickel contents, the neutron fluence and the calculational procedures. RG 1.99, Rev. 2, describes the methodology to be used in calculating the margin term. The licensee states that the calculations are consistent with the guidance contained in Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence" (RG 1.190). RG 1.190 describes methods and assumptions acceptable to the NRC staff for determining the pressure vessel neutron fluence with respect to the General Design Criteria contained in Appendix A to 10 CFR 50. In consideration of the guidance set forth in RG 1.190, GDC 14, 30, and 31 are applicable. GDC 14, "Reactor Coolant Pressure Boundary," requires the design, fabrication, erection, and testing of the reactor coolant pressure boundary so as to have an extremely low probability of abnormal leakage, of rapidly propagating failure, and of gross rupture. GDC 30, "Quality of Reactor Coolant Pressure Boundary," requires, among other things, that components comprising the reactor coolant pressure boundary be designed, fabricated, erected, and tested to the highest quality standards practical. GDC 31, "Fracture Prevention of Reactor Coolant Pressure Boundary," pertains to the design of the reactor coolant pressure boundary, stating: The reactor coolant pressure boundary shall be designed with sufficient margin to assure that when stressed under operating, maintenance, testing, and postulated accident conditions, ( 1) the boundary behaves in a non brittle manner and (2) the probability of rapidly propagating fracture is minimized. The design shall reflect consideration of service temperatures and other conditions of the boundary material under operating maintenance, testing and postulated accident conditions and the uncertainties in determining (1) material properties, (2) the effects of irradiation on material properties, (3) residual, steady state and transient stresses, and (4) size of flaws.

3.0 TECHNICAL EVALUATION

3.1 Licensee's Evaluation The revised P-T limits are based on the methodology of WCAP-14040-NP-A, Revision 4, "Methodology used to Develop Cold Overpressure Mitigating System Setpoints and RCS [Reactor Coolant System] Heatup and Cooldown Limit Curves," (henceforth the Westinghouse methodology). The Westinghouse methodology is an approved generic methodology for generating P-T limits and Cold Overpressure Mitigation System setpoints based on the plant specific ARTs. The Westinghouse methodology provides a discussion of several P-T limit curve generation approaches and equations to generate the necessary P-T limit curves. The Westinghouse methodology was implemented via analysis contained in WCAP-16752-NP (ADAMS Accession No. ML090760605) that was previously submitted as the basis for the P-T limits and L TOP settings for 29.2 EFPY by Entergy's letter dated March 5, 2009 (ADAMS ML090760649). The application was approved by the NRC staff in TS Amendment No. 262 dated August 17, 2009 (ADAMS ML092150447). For the limiting RPV beltline materials, the licensee identified the intermediate shell plate 8-2002-3 for the axial-flaw methodology and circumferential weld wire heat number 348009 for the circumferential-flaw methodology. ART values were calculated for 48 EFPY. The parameters used to determine the licensee's ART values for the limiting materials at the one-quarter of the RPV wall thickness (1/4T) and three-quarter (3/4T) locations for 48 EFPY are shown in Tables 2-1 and 2-4 of WCAP-16752-NP-A. 3.2 Staff Evaluation 3.2.1 ART Value and P-T Limit Curves The NRC staff assessed the validity of the licensee's proposed curves. The staff first performed independent calculation of the ART values using the methodology of RG 1.99, Rev. 2. Based on these calculations the staff verified that the licensee's limiting material in the axial-flaw case is Intermediate Shell Plate 8-2002-3, and that the limiting material in the circumferential-flaw case is Circumferential Weld 348009. The staff's calculated ART values were in good agreement with the licensee's calculated ART of 237 oF and 187 oF for the 1 /4T and 3/4T locations in 8-2002-3 material, and 238 oF and 175 oF for the 1/4T and 3/4T locations in 348009 material. The NRC staff evaluated the licensee's proposed P-T limit curves for acceptability by performing independent calculations using the methodologies of Appendix G of Section XI of the ASME Code and 10 CFR Part 50, Appendix G. The licensee stated that the proposed P-T limit curves were based on the methodologies of Appendix G of Section XI of the ASME Code, 1998 Edition with the 2000 Addenda, which utilizes an alternative reference fracture toughness (KIC) curve instead of the K1a fracture toughness curve for RPV materials in determining the P-T limit curves. NRC Regulatory Issues Summary 2004-04, "Use of Code Cases, N-588, N-640, and N-641 in Developing Pressure-Temperature Operating Limits," dated April5, 2004 (ADAMS Accession No. ML040920323), states that the ASME Code, Section XI, Appendix G, 1998 Edition with 2000 Addenda, may be used without the need for an exemption. The use of the K1c fracture toughness curve is acceptable for evaluating the potential for crack initiation without imposing unnecessary conservatism. The K1c curve appropriately implements the use of static initiation fracture toughness behavior to evaluate the controlled heatup and cooldown process of an RPV. The proposed P-T limit curves in WCAP-16752-NP-A were created at the same time as the previously approved 29.2 EPFY curves using the approved Westinghouse methodology. Limiting heatup curves used heatup rates of 60 °F/hr and 100 °F/hr applicable for 48 EFPY. The limiting cooldown curves used cooldown rates of 0, 20, 40, 60, and 100 oF/hr applicable for 48 EFPY. The NRC staff determined that the minimum temperature requirements of Table 1 of Appendix G to 10 CFR Part 50 were properly implemented in the P-T limit curves. The NRC staff requested that the licensee address two topics. First, the staff requested that the licensee confirm that all materials in the RPV having received a cumulative neutron fluence in excess of 1 x 1017 n/cm2 (E > 1.0 MeV) were bounded by the proposed P-T limit curves. The licensee responded in their October 3, 2013, letter in detail. Regarding the expanded beltline, the licensee identified all materials adjacent to the previously identified beltline materials and provided the fluence estimates for these locations. The materials/locations identified included the inlet nozzles, outlet nozzles, nozzle shell plates, nozzle shell longitudinal welds, nozzle shell to intermediate shell circumferential weld, and the lower shell to lower vessel head circumferential weld. Of these, all but the nozzles themselves exceeded the fluence threshold for inclusion as beltline materials. The licensee presented ART calculations for all of these materials establishing that none of these extended beltline materials would approach the beltline materials with regards to limiting the P-T limit curves. Second, the NRC staff requested that the licensee clarify how all RPV materials (beltline and non-beltline) and the lowest service temperature of all ferritic reactor coolant pressure boundary materials were addressed in the P-T limit curves. The licensee responded by identifying the extended beltline materials, reactor vessel inlet and outlet nozzles, replacement steam generator (SGs), and original pressurizer as the components subject to this request. For these components the licensee analyzed the material and stress state to establish that these components were not limiting. The licensee provided calculations establishing that the extended beltline materials did not displace the bounding material for the RPV. The staff reviewed these calculations and their inputs and confirmed that these materials would not be limiting. For the nozzles, the licensee generated nozzle specific P-T limit curves noting that these were bounded by the submitted curves. The NRC staff confirmed the inputs and basis used to generate these curves. Regarding the replacement SGs, the licensee reviewed two critical locations in the SGs, the sheet to channel head junction and the primary nozzle knuckle region. For the sheet to channel head junction the licensee produced an estimated P-T limit curve and identified that it was bounded by the submitted curves. For the knuckle region the licensee established that the stresses in that region were bounded by the RPV outlet nozzle, and given that the knuckle region possessed superior toughness, established that the knuckle region would be bounded by the nozzle. The staff reviewed this information and confirmed that the inputs and basis used to generate these conclusions provided reasonable assurance that the SGs were bounded by the submitted P-T limit curves. Regarding the pressurizer, the licensee noted that the pressurizer was designed to withstand in-and outs urge events. The licensee outlined the analysis for surge events and concluded that the pressurizer had substantial margin beyond the P-T limit curves. The staff reviewed this information and confirmed that the pressurizer would not impact the submitted P-T limit curves. Therefore, the NRC staff verified that the licensee's proposed P-T limits are in accordance with Appendix G to Section XI of the ASME Code and satisfy the requirements of Appendix G to 10 CFR Part 50. 3.2.2 Neutron Fluence The fluence values cited in the application are identical to those supplied in the original submittal of WCAP-16752-NP. WCAP-16752-NP indicates that the methodology used to develop the PT curves was WCAP-14040-NP-A, Revision 4. This NRC-approved methodology includes a description of the core source and transport calculations used to determine the reactor vessel neutron fluence. As noted in the NRC Safety Evaluation approving WCAP-14040-NP-A, the fluence methods were accepted by the staff based on their adherence to the guidance in RG 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," dated March 2001. The WCAP also indicates that the fluence calculations account for a recently implemented stretch power uprate, and the neutron source projection is based on a current cycle design multiplied by 1.05 to increase margin for future cycle variations. Since the calculations were performed using NRC-approved, RG 1.190-adherent methodology and future cycles include a conservative multiplier to account for cycle to cycle variation, the NRC staff determined that the fluence calculations are acceptable for the purpose of developing P-T limit curves for IP2. 3.2.3 LTOP The LTOP system provided by the pressurizer power operated relief valves (PORVs) works to prevent RCS overpressurization below certain temperatures, thus maintaining reactor coolant pressure boundary integrity. The L TOP analysis yields limiting conditions for operation that constitute L TOP system alignments for the period of applicability. The NRC staff reviewed the LTOP analysis using the guidance contained in Branch Technical Position (BTP) 5-2, "Overpressurization Protection of Pressurized Water Reactors While Operating at Low Temperatures," of NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants." The licensee has followed the methods used in the development of the current P-T limits and LTOP requirements (for 29.2 EFPY). These methods were reviewed and accepted by NRC staff in the SE forTS Amendment 262 noted in Section 3.1 of this SE. They are also adherent to the NRC-approved licensing topical report WCAP-14040-NP-A. The revised L TOP system setpoint and associated L TOP system curves reflect the increase in neutron fluence for a service life increase from 29.2 EFPY to 48 EFPY. The PORV opening setpoint is implemented as a variable setpoint for the L TOP system instrumentation. The NRC staff has accepted the use of ASME Code Case N-588, "Alternative to Reference Flaw Orientation of Appendix G for Circumferential Welds in Reactor Vessels," and N-640, "Alternative Reference for Fracture Toughness for Development of P-T Limit Curves," as allowed by ASME Code, Section XI, Appendix G, for determination of L TOP system setpoints. The L TOP arming temperature must be at least RT Nor + 50 oF at the beltline location (1/4T or 3/4T) that is controlling in the Appendix G limit calculations. The NRC staff calculated that in accordance with ASME Code, the IP2 L TOP system arming temperature (Tenable) can be as low as 288 °F, confirming the licensee's analysis.

3. 3 Conclusion Based on the NRC staff's review of the information provided in the application and supplementary materials, the staff concludes that the proposed IP2 P-T limits and LTOP requirements meet the requirements of 10 CFR Part 50, Appendix G; ASME Code, Appendix G; and the other required documents specified above. Therefore, the revised P-T limits and LTOP requirements are acceptable as part of the IP2 licensing bases.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the New York State official was notified of the proposed issuance of the amendment. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and changes surveillance requirements. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration, and there has been no public comment on such finding (78 FR 19750). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.

6.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above that (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) there is reasonable assurance that such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public. Principal Contributors: C. Fairbanks, NRR B. Parks, NRR Date: March 5, 2014

March 5, 2014 Vice President, Operations Entergy Nuclear Operations, Inc. Indian Point Energy Center 450 Broadway, GSB P.O. Box 249 Buchanan, NY 10511-0249

SUBJECT:

INDIAN POINT NUCLEAR GENERATING UNIT NO.2-ISSUANCE OF AMENDMENT RE: PRESSURE-TEMPERATURE LIMIT CURVES AND LOW TEMPERATURE OVER PRESSURE REQUIREMENTS (TAC NO. MF0634)

Dear Sir or Madam:

The Commission has issued the enclosed Amendment No. 27 4 to Facility Operating License No. DPR-26 for the Indian Point Nuclear Generating Unit No. 2. The amendment consists of changes to the Technical Specifications (TSs) in response to your application dated February 6, 2013, as supplemented on July 9, 2013, October 3, 2013, and February 24, 2014. The amendment changes the TSs by revising the reactor heatup and cooldown curves (also referred to as pressure-temperature limits) and low temperature overpressure protection requirements to cover a lifetime burnup of 48 Effective Full Power Years (EFPY), which is an increase from the current value of 29.2 EFPY. A copy of the related Safety Evaluation is enclosed. A Notice of Issuance will be included in the Commission's next regular biweekly Federal Register notice. Docket No. 50-247

Enclosures:

1. Amendment No. 27 4 to DPR-26
2. Safety Evaluation cc w/encls: Distribution via Listserv DISTRIBUTION:

PUBLIC LPL 1-1 RIF RidsNrrDorl RidsNrrDorllpl1-1 CFairbanks, EVIB RidsNrr DssSrxb RidsNrrDssStsb RidsNrrPMindianPoint RidsNrrLAKGoldstein SParks, SRXB ADAMS ACCESSION NO.: ML14045A248 OFFICE LPL 1-1/PM LPL 1-1/LA NAME DPickett KGoldstein DATE 02/24/2014 02/19/2014 Sincerely, IRA/ Douglas V. Pickett, Senior Project Manager Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation RidsNrrDe Evib RidsRgn1 Mail Center RidsAcrsAcnw MaiiCTR RidsNrrDoriDpr ABurritt, R1

  • via memo EVIB/BC*

OGC LPL 1-1/BC SRosenberg DRoth BBeasley 02/11/2014 02/21/2014 03/05/2014 OFFICIAL RECORD COPY}}

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