ML15226A159

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Issuance of Amendment Changes to Reactor Vessel Heatup and Cooldown Curves and Low Temperature Overpressure Protection System Requirements
ML15226A159
Person / Time
Site: Indian Point Entergy icon.png
Issue date: 09/03/2015
From: Pickett D
Plant Licensing Branch 1
To:
Entergy Nuclear Operations
Pickett D V
References
TAC MF5746, FOIA/PA-2016-0148
Download: ML15226A159 (32)
v • d • e


Text

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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. 3 - ISSUANCE OF AMENDMENT RE: CHANGES TO REACTOR VESSEL HEATUP AND COOLDOWN CURVES AND LOW TEMPERATURE OVERPRESSURE PROTECTION SYSTEM REQUIREMENTS (TAC NO. MF5746)

Dear Sir or Madam:

The Commission has issued the enclosed Amendment No. 258 to Facility Operating License No. DPR-64 for the Indian Point Nuclear Generating Unit No. 3. The amendment consists of changes to the Technical Specifications (TSs) in response to your application dated February 12, 2015, as supplemented by letter dated August 11, 2015.

The amendment revises TS 3.4.3, "RCS [reactor coolant system] Pressure and Temperature (PIT) Limits," and 3.4.12, "Low Temperature Overpressure Protection (LTOP)," to include new RCS PIT limit curves for heatup, cooldown, and pressure test operations and LTOP system setpoints. The proposed PIT limit curves and LTOP system setpoints will be valid for 37 effective full power years of facility operation, which is the accumulated burn up estimated to occur in December 2023 during the period of extended plant operation.

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.

Sincerely,

~vP~

Douglas V. Pickett, Senior Project Manager Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-286

Enclosures:

1. Amendment No. 258 to DPR-64
2. Safety Evaluation cc w/encls: Distribution via Listserv

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 ENTERGY NUCLEAR INDIAN POINT 3. LLC AND ENTERGY NUCLEAR OPERATIONS. INC.

DOCKET NO. 50-286 INDIAN POINT NUCLEAR GENERATING UNIT NO. 3 AMENDMENT TO FACILITY OPERATING LICENSE AND TECHNICAL SPECIFICATIONS Amendment No. 258 License No. DPR-64

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

A. The application for amendment by Entergy Nuclear Operations, Inc. (Entergy, the licensee) dated February 12, 2015, as supplemented by letter dated August 11, 2015, 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.

Enclosure 1

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-64 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendices A, Band C, as revised through Amendment No. 258, 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 30 days.

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:

Changes to the License and Technical Specifications Date of Issuance: September 3, 201 5

ATTACHMENT TO LICENSE AMENDMENT NO. 258 FACILITY OPERATING LICENSE NO. DPR-64 DOCKET NO. 50-286 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 Insert Pages 3.4.3-1 3.4.3-1 3.4.3-2 3.4.3-2 3.4.3-3 3.4.3-3 3.4.3-4 3.4.3-4 3.4.3-5 ------

3.4.12-1 3.4.12-1 3.4.12-2 3.4.12-2 3.4.12-4 3.4.12-4 3.4.12-5 3.4.12-5 3.4.12-8 3.4.12-8 3.4.12-9 3.4.12-9 3.4.12-10 3.4.12-10 3.4.12-11 3.4.12-11 3.4.12-12 ------

(4) ENO pursuant to the Act and 10 CFR Parts 30, 40 and 70, Arndt. 203 to receive, possess, and use in amounts as required any 11/27/00 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; (5) ENO pursuant to the Act and 10 CFR Parts 30 and 70, to Arndt. 203 possess, but not separate, such byproduct and special 11/27/00 nuclear materials as may be produced by the operation of the facility.

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 I: 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; and 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 (100% of rated power).

(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 258 are hereby incorporated in the License. ENO shall operate the facility in accordance with the Technical Specifications.

(3) (DELETED) Arndt. 205 2-27-01 (4) (DELETED) Arndt. 205 2-27-01 D. (DELETED) Amdt.46 2-16-83 E. (DELETED) Amdt.37 5-14-81 F. This amended license is also subject to appropriate conditions by the New York State Department of Environmental Conservation in its letter of May 2, 1975, to Consolidated Edison Company of New York, Inc., granting a Section 401 certification under the Federal Water Pollution Control Act Amendments of 1972.

Amendment No. 258

RCS P/T Limits 3.4.3 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.3 RCS Pressure and Temperature (PIT) Limits LCO 3.4.3 RCS pressure, RCS temperature, and RCS heatup and cooldown rates shall be maintained within the limits specified in Figure 3.4.3-1 and Figure 3.4.3-2.

APPLICABILITY: At all times.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. ---------NOTE--------- A.1 Restore parameter(s) to 30 minutes Required Action A.2 shall within limits.

be completed whenever this Condition is entered. AND A.2 Determine RCS is 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Requirements of LCO not acceptable for continued met in MODE 1, 2, 3, or operation.

4.

B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time of Condition A not AND met.

B.2 Be in MODE 5 with RCS 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> pressure< 500 psig.

(continued)

INDIAN POINT 3 3.4.3-1 Amendment 258

RCS P/T Limits 3.4.3 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. ---------NOTE---------------- C.1 Initiate action to restore Immediately Required Action C.2 shall parameter( s) be completed whenever to within limits.

this Condition is entered.

AND Requirements of LCO not C.2 Determine RCS is Prior to met any time in other acceptable for continued entering MODE 4 than MODE 1, 2, 3, or 4. operation.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.3.1 -------------------NOTE--------------

Only required to be performed during RCS heatup, inservice leak testing, and cooldown operations.

Verify RCS pressure, RCS temperature, and RCS 30 minutes heatup and cooldown rates are within the limits specified in the following:

a. Figure 3.4.3-1 during RCS heatup and during RCS inservice leak testing; and
b. Figure 3.4.3-2 during RCS cooldown.

INDIAN POINT 3 3.4.3-2 Amendment 258

RCS P/T Limits 3.4.3 MATERIAL PROPERTY BASIS LIMITING MATERIAL: Lower Shell Plate B-2803-3 using credible surveillance data, Position 2.1 LIMITING ART VALUES AT 37 EFPY: 1/4T, 245.0°F (Axial Flaw) 3/4T, 198.2°F (Axial Flaw)

I Limit~

<(

VJ 2250 !Leak Test

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s 2000 Acceptable Operation Ill Ill e

a.

Unacceptable Operation E 1750

.s Ill VJ Heatup Rate

~ 1500 60 Deg. F/Hr 0

0

(.)

.s 1250 CV Q) ex:

1000 750 500 250 0 50 100 150 200 250 300 350 400 450 500 550 T cold Temperature (Deg. F)

Figure 3.4.3-1:

Heatup and lnservice Leak Test Limitations for Reactor Coolant System (Without instrument uncertainties)

INDIAN POINT 3 3.4.3-3 Amendment 258

RCS P/T Limits 3.4.3 MATERIAL PROPERTY BASIS LIMITING MATERIAL: Lower Shell Plate B-2803-3 using credible surveillance data, Position 2.1 LIMITING ART VALUES AT 37 EFPY: 1/4T, 245.0°F (Axial Flaw) 3/4T, 198.2°F (Axial Flaw) 2250

<2000 en a.

Q)

1750 Unacceptable Acceptable in Operation in Operation

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a.

~ 1500

-in en

1250 0

0

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0 g 1000 Q) 0:::

750 ates eg. F/Hr teady -state 500 20 0

60 100 250 0 50 100 150 200 250 300 350 400 450 500 550 T cold Temperature (Deg. F)

Figure 3.4.3-2:

Cooldown Limitations for Reactor Coolant System (Without instrument uncertainties)

INDIAN POINT 3 3.4.3-4 Amendment 258

LTOP 3.4.12 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.12 Low Temperature Overpressure Protection (LTOP)

LCO 3.4.12 LTOP shall be OPERABLE with no high head safety injection (HHSI) pumps capable of injecting into the RCS and the accumulator discharge isolation valves closed and de-energized, and either of the following:

Note--------------------

LCO 3.4.12.a and LCO 3.4.12.b are not Applicable when all RCS cold leg temperatures are> 330°F.

a. The Overpressure Protection System (OPS) OPERABLE with two power operated relief valves (PORVs) with lift settings within the limit specified in Figure 3.4.12-1;
b. The RCS depressurized with an RCS vent of :2: 2.00 square inches, or one blocked open PORV with its block valve disabled in the open position.

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 curve in Figure 3.4.12-1.
2. One HHSI pump may be made capable of injecting into the RCS as needed to support emergency boration or to respond to a loss of RHR cooling.
3. One HHSI pump may be made capable of injecting into the RCS for pump testing for a period not to exceed 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

APPLICABILITY: Whenever the RHR System is not isolated from the RCS, MODE 4 when any RCS cold leg temperature is :5 330°F, MODE 5, MODE 6 when the reactor vessel head is on.

INDIAN POINT 3 3.4.12-1 Amendment 258

LTOP 3.4.12 ACTIONS

NOTE --------------------------------------------

LCO 3.0.4.b is not applicable when entering MODE 4.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more HHSI A.1 Initiate action to verify no Immediately pump(s) capable of HHSI pumps are injecting into the RCS. capable of injecting into the RCS.

OR A.2.1 Verify RCS is vented Immediately with opening greater than or equal to one pressurizer code safety valve flange.

AND A.2.2 Verify no more than two Immediately HHSI pumps are capable of injecting into AND the RCS Once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> (continued)

INDIAN POINT 3 3.4.12-2 Amendment 258

LTOP 3.4.12 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME E. Two required PORVs E.1 Depressurize RCS and 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> inoperable. establish RCS vent of

: : 2.00 square inches, or OR one blocked open PORV with its block valve Required Action and disabled in the open associated Completion position.

Time of Condition C or D not met. OR E.2.1 Increase all RCS cold 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> leg temperatures to

> 330°F.

AND E.2.2 Isolate the RHR System 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> from the RCS.

OR E.3 Verify pressurizer level, 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> RCS pressure, and RCS injection capability are AND within limits specified in Figure 3.4.12-2 and Once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Figure 3.4.12-3 for OPS thereafter not OPERABLE.

F. LTOP inoperable for any F.1 Depressurize RCS and 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> reason other than establish RCS vent of Condition A, B, C, D, or  : : : 2.00 square inches, or E. one blocked open PORV with its block valve disabled in the open position.

INDIAN POINT 3 3.4.12-4 Amendment 258

LTOP 3.4.12 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.12.1 Verify no HHSI pumps are capable of injecting into 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> the RCS.

SR 3.4.12.2 Verify each accumulator discharge isolation valve is 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> closed and de-energized; Verify each accumulator pressure is less than the 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> maximum RCS pressure for the coldest existing RCS cold leg temperature allowed by the Prr limit curve in Figure 3.4.12-1.

SR 3.4.12.3 -------------------N()TE----------------------------------------

()nly required to be met when complying with LC()

3.4.12.b.

Verify RCS vent ~ 2.00 square inches, or one blocked 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for open P()RV with its block valve disabled in the open unlocked open position established. vent valve(s) 31 days for locked open vent valve(s)

(continued)

INDIAN POINT 3 3.4.12-5 Amendment 258

SURVEILLANCE REQUIREMENTS continued SURVEILLANCE FREQUENCY SR 3.4.12.9 ------------------N()TES----------------

1. Not required to be met when all RCS cold leg temperatures are> 330°F.
2. Not required to be met if SR 3.4.12.8 is met.

Verify each of the following conditions are satisfied Within 15 minutes prior to starting any RCP: prior to starting any RCP

a. Secondary side water temperature of the hottest steam generator is s:; 50°F above the coldest RCS cold leg temperature; and
b. RCS makeup is less than or equal to RCS losses; and
c. ()verpressure Protection System (()PS) is

()PERABLE; and

d. Pressurizer level is s:; 73%.

Indian Point 3 3.4.12-8 Amendment 258

LTOP 3.4.12 800 750

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c..

- 700

.5

~

c..

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~ 650

..2 1

600

=E 550

1

~

500 450 <

50 100 150 200 250 300 350 0

Auctioneered Low Indicated RCS Temperature ( F)

Figure 3.4.12-1: Maximum Allowable PORV Setpoint for LTOP (OPS), 37 EFPY INDIAN POINT 3 3.4.12-9 Amendment 258

LTOP 3.4.12 100 l'nacceptable 90 Operation

~ so

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~

::; 300 psig
>> 70 s 400 psig 60
  • ~

.......= 50

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..:! s 450 psig iit 40 c

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30 - s-snu psig'----~ -; s 55-0 psig

Acceptable
?; 20 Operation 10 0 -

0 50 100 150 200 250 300 350 400 Indicated Tcold (0 F)

Figure 3.4.12-2: Maximum Allowable Pressurizer Level for LTOPS Inoperable with 1 Charging Pump Capable of Injecting, 37 EFPY INDIAN POINT 3 3.4.12-10 Amendment 258

LTOP 3.4.12 90 l:nacceptable so

  • Operation t 70 s 100 psig
l

~

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N 60 s 200 psig

  • ~ 50 ~J~*- "-

~

-; 40 i ~

~ 30 s soo psig

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~ 20 Acceptable 10 +I Operation 0

0 50 100 150 200 250 300 350 400 Indicated Tcold (°F)

Figure 3.4.12-3: Maximum Allowable Pressurizer Level for LTOPS Inoperable with 3 Charging Pumps Capable of Injecting, 37 EFPY INDIAN POINT 3 3.4.12-11 Amendment 258

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 258 TO FACILITY OPERATING LICENSE NO. DPR-64 ENTERGY NUCLEAR INDIAN POINT 3. LLC AND ENTERGY NUCLEAR OPERATIONS. INC.

DOCKET NO. 50-286 INDIAN POINT NUCLEAR GENERATING UNIT NO. 3

1.0 INTRODUCTION

By letter dated February 12, 2015 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML15061A275}, supplemented by letter dated August 11, 2015 (ADAMS Accession No. ML15230A056}, Entergy Nuclear Operations, Inc. (the licensee) submitted a license amendment request (LAR) to revise the technical specifications (TS) for Indian Point Unit No. 3 (IP3). The LAR would revise TS Sections 3.4.3, "RCS [reactor coolant system] Pressure and Temperature (PIT) Limits," and 3.4.12, "Low Temperature Overpressure Protection (LTOP)," to include new RCS PIT limit curves for heatup, cooldown, and pressure test operations and LTOP system setpoints.

The proposed PIT limit curves would be valid for 37 effective full power years (EFPY) of facility operation, which is the accumulated burnup estimated to occur in December 2023 during the period of the extended plant operating license. The current operating license for IP3 is scheduled to expire on December 12, 2015. On April 23, 2007, the licensee submitted a timely application for license renewal. That application is under Nuclear Regulatory Commission (NRC) review at this time. In accordance with Title 1O of the Code of Federal Regulations (1 O CFR) Section 2.109, if the NRC's review of that application has not concluded by December 12, 2015, the plant would be allowed to continue to operate unit a final determination has been made on the license renewal application.

The supplemental letter dated August 11, 2015, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staffs original proposed no significant hazards consideration.

2.0 REGULATORY EVALUATION

The following explains the applicability of General Design Criteria (GDC) for IP3. The construction permit for IP3 was issued by the Atomic Energy Commission (AEC) on August 13, Enclosure 2

1969, and the operating license was issued on December 12, 1975. 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 10 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 IP3 are those in the UFSAR.

As discussed in the UFSAR, the licensee for IP3 has made some changes to the facility over the life of the unit that 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 IP3 licensing basis documentation, such as license amendments.

The NRC has established requirements in 10 CFR Part 50, Section 50.60, "Acceptance Criteria for Fracture Prevention Measures for Lightwater Nuclear Power Reactors for Normal Operation." These requirements mandate that all light-water nuclear power reactors must meet the fracture toughness requirements for the reactor coolant pressure boundary (RCPB) set forth in 10 CFR Part 50, Appendix G, "Fracture Toughness Requirements, in order to protect the integrity of the RCPB.

The provisions of 10 CFR Part 50, Appendix G require that the PIT limits for an operating light-water nuclear power reactor be at least as conservative as those that would be generated if the methods of Appendix G, "Fracture Toughness Criteria for Protection Against Failure, in Section XI of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) were used to generate the PIT limits. The provisions of 10 CFR Part 50, Appendix G also require that applicable surveillance data from reactor pressure vessel (RPV) material surveillance programs be incorporated into the calculations of plant-specific PIT limits, and that the PIT limits for operating reactors be generated using a method that accounts for the effects of neutron irradiation on the material properties of the RPV beltline materials.

Table 1 of 10 CFR Part 50, Appendix G provides the NRC staff's criteria for meeting the PIT limit requirements of the ASME Code,Section XI, Appendix G, as well as the minimum temperature requirements for the RPV during normal heatup, cooldown, and pressure test operations. In addition, NRC staff regulatory guidance related to PIT limit curves is contained in Regulatory Guide (RG) 1.99, Revision (Rev.) 2, "Radiation Embrittlement of Reactor Vessel Materials, and Standard Review Plan (NUREG-0800) Chapter 5.3.2, "Pressure-Temperature Limits and Pressurized Thermal Shock."

PIT limit curve calculations are based, in part, on the reference nil-ductility temperature (RT NDT) for the material, as specified in the ASME Code,Section XI, Appendix G. 10 CFR Part 50, Appendix G requires that RT NDT values for materials in the RPV beltline region be adjusted to account for the effects of neutron radiation.

On October 14, 2014, the NRG issued Regulatory Issue Summary (RIS) 2014-11, "Information on Licensing Applications for Fracture Toughness Requirements for Ferritic Reactor Coolant Pressure Boundary Components," which clarified that the beltline definition in 10 CFR Part 50, Appendix G is applicable to all reactor vessel ferritic materials with projected neutron fluence values greater than 1x10 17 neutrons/centimeter-squared (n/cm 2 ) with energy greater than one million electron volts (E > 1 MeV), and this fluence threshold remains applicable for the design life as well as throughout the licensed operating period.

RG 1.99, Rev. 2 contains methodologies for calculating the adjusted RT Nor (ART) due to neutron irradiation. The ART is defined as the sum of the initial (unirradiated) reference temperature (initial RTNDT ), the mean value of the adjustment in reference temperature caused by irradiation (L\RT Nor), and a margin term. The L\RT Nor 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 NDT is a plant-specific or a generic value and whether the CF was determined using the tables in RG 1.99, Rev. 2, or from surveillance data. The margin term is used to account for uncertainties in the values of the initial RT Nor, 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.

In March 2001, the NRG staff issued RG 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence." Fluence calculations for use in ART and PIT limit curve analyses are acceptable if they are performed with approved methodologies or with methods which are shown to conform to the guidance in RG 1.190.

Section 182a of the Atomic Energy Act requires applicants for nuclear power plant operating licenses to include TSs as part of the license. The Commission's regulatory requirements related to the content of the TSs are contained in 10 CFR 50.36, "Technical specifications." The TS requirements in 10 CFR 50.36 include the following categories: (1) safety limits, limiting safety systems settings and control settings, (2) limiting conditions for operation, (3) surveillance requirements. (4) design features, and (5) administrative controls.

The LTOP system controls RCS pressure at low temperatures so the integrity of the RCPB is not compromised by violating the PIT limits required by 10 CFR Part 50, Appendix G.

RG 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," describes methods and assumptions acceptable to the NRG staff for determining the pressure vessel neutron fluence with respect to the GDC contained in Appendix A of 10 CFR Part 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 RCPB 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 RCPB 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 RCPB, 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) materials 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

Proposed TS Changes The licensee's application dated February 12, 2015, as supplemented by letter dated August 11, 2015, proposes the following revisions:

and Figure 3.4.3-2, "Cooldown Limitations for Reactor Coolant System," with proposed TS Figures 3.4.3-1 and 3.4.3-2, as shown in Attachment 2 of the August 11, 2015, supplemental letter.

  • Delete existing TS Figure 3.4.3-3, "Hydrostatic and lnservice Leak Testing Limitations for Reactor Coolant System." The RCS inservice leak testing limitations have been added to proposed TS Figure 3.4.3-1. The change in the title of the revised Figure also eliminates reference to hydrostatic tests which are no longer required by ASME Code,Section XI.
  • Revise TS Limiting Condition of Operation (LCO) 3.4.3 to reflect deletion of Figure 3.4.3-3.
  • Revise LCO 3.4.12 from "The RCS depressurized with an RCS vent of~ 2.00 square inches," to "The RCS depressurized with an RCS vent of~ 2.00 square inches, or one blocked open PORV with its block valve disabled in the open position."
  • Revise the Required Actions under Condition A of TS 3.4.12 to better reflect the LCO when one or more high head safety injection (HHSI) pumps are capable of injecting into the RCS.
  • Revise Conditions E.1 and F.1 of TS 3.4.12 from "Depressurize RCS and establish RCS vent of~ 2.00 square inches," to "Depressurize RCS and establish RCS vent of~ 2.00 square inches, or one blocked open PORV with its block valve disabled in the open position."
  • Revise SR 3.4.12.3 from "Verify RCS vent~ 2.00 square inches established," to "Verify RCS vent~ 2.00 square inches, or one blocked open PORV with its block valve disabled in the open position established."
  • Revise SR 3.4.12. 9 from "Verify each of the following conditions are satisfied prior to starting any RCP [reactor coolant pump]: a. Secondary side water temperature of the hottest steam generator is:::; 64°F above the coldest RCS cold leg temperature; and," to "Verify each of the following conditions are satisfied prior to starting any RCP: a.

Secondary side water temperature of the hottest steam generator is :::; 50°F above the coldest RCS cold leg temperature; and" 3.1 Reactor Coolant System PIT Limits Licensee's Evaluation The technical bases for the licensee's revised PIT limits were provided in two enclosures to the LAR: (1) Proprietary Westinghouse Report No. WCAP-17954-P, Rev. 0, "Indian Point Unit 3 Heatup and Cooldown Limit Curves for Normal Operation," December 2014 (ADAMS Accession No. ML15061A278), and (2) Non-Proprietary Westinghouse Report No. WCAP-17954-NP, Rev. 0, "Indian Point Unit 3 Heatup and Cooldown Limit Curves for Normal Operation,"

December 2014 (ADAMS Accession No. ML15061A277). Throughout this Safety Evaluation (SE), reference is made to non-proprietary report WCAP-17954-NP to avoid the need for proprietary markings.

In WCAP-17954-NP, the licensee stated that the proposed 37 EFPY heatup and cooldown PIT limit curves were generated using the most limiting ART values and the NRG-approved methodology documented in Westinghouse Topical Report WCAP-14040-A, Rev. 4, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," May 2004 (ADAMS Accession No. ML050120209). The licensee further stated that the neutron transport evaluation methodologies followed the guidance of RG 1.190, and are consistent with the NRG-approved methodology described in WCAP-14040-A, Rev. 4. According to the licensee, the PIT limit curves were developed based on the limiting ART values for the cylindrical beltline and extended beltline reactor vessel shell materials consistent with the recommendations provided in RIS 2014-11, and the PIT limits developed for the cylindrical beltline shell region bound the PIT limits for the reactor vessel inlet and outlet nozzles for IP3 at 37 EFPY. The licensee's evaluation also included the determination of minimum temperature requirements for the closure flange region, as required by Table 1 of 10 CFR Part 50, Appendix G, based on the limiting RTNOT value for the closure flange region.

The licensee's submittal included information on the methods employed for determining the fracture toughness, Kie, and the applied stress intensity factors due to pressure and thermal stresses, K1P and K1T, respectively, which were used for the PIT limit calculations. For all locations, Kie was established based on the RT NOT for each material consistent with the methodology detailed in the 1998 through the 2000 Addenda Edition of the ASME Code,Section XI, Appendix G. For all beltline portions of the RPV, RTNOT values were adjusted to

account for the effects of neutron embrittlement through 37 EFPY using the procedures in RG 1.99, Rev. 2. For the RPV shell regions, K1P and K1T values were calculated using the formulations specified in WCAP-14040-A, Rev. 4, which are the same as those specified in Paragraph G-2214 of the ASME Code,Section XI, Appendix G. For the inlet and outlet nozzles, K1P and K1T values were determined using component-specific finite element analysis techniques.

The licensee stated that all regions of the RPV were evaluated by considering a postulated flaw with a depth of one-quarter of the RPV wall thickness (1/4T) for cooldown conditions and a depth of three-quarters of the RPV wall thickness (3/4T) for heatup and isothermal conditions, accounting for the specific RPV section thickness (shell thickness, nozzle forging thickness, etc.).

NRC Staff Evaluation

The NRC staff reviewed the licensee's submittal, including WCAP-17954-P and WCAP-17954-NP, to determine whether the licensee's proposed 37 EFPY PIT limit curves are in compliance with the requirements of 10 CFR Part 50, Appendix G, as required by 10 CFR 50.60. The staff verified that the proposed 37 EFPY PIT limits were developed by taking into account all portions of the RPV, including the inlet and outlet nozzles. The staff noted that, based on the evaluation of all regions of the RPV, the bounding PIT limits for 37 EFPY are controlled by the RPV beltline shell region and the minimum temperature criteria based on Table 1 of 10 CFR Part 50, Appendix G.

Evaluation of Licensee's ART Values The NRC staff verified that the licensee's proposed 37 EFPY PIT limits were calculated based on an evaluation of the RPV beltline region, accounting for neutron embrittlement through 37 EFPY.

The licensee's evaluation of the RPV beltline region included PIT limit calculations for two beltline components: the limiting RPV beltline shell material and the inlet/outlet nozzles.

The licensee's ART calculations for the 1/4T and 3/4T locations were provided in Tables 7-2 and 7-3 of WCAP-17954-NP, respectively, including all of the input parameters necessary for calculating the ART values. A summary of those calculations is provided in Table 7-4 of WCAP-17954-NP, and indicates that the limiting beltline shell material is the Lower Shell Plate (Heat No. B-2303-3) using credible surveillance data, which has a 37 EFPY ART of 245°F at the 1/4T location and 198.2°F at the 3/4T location. Based on this information, the NRC staff verified that the 37 EFPY ART values for the limiting beltline materials were calculated correctly using the procedures in RG 1.99, Rev. 2, and that the Lower Shell Plate was limiting.

The NRC staff verified that the initial RT NDT values for the limiting beltline shell materials are consistent with those identified in Table 4.2-6 of the IP3 License Renewal Application (LRA, ADAMS Accession No. ML071210517), and approved by the staff in the IP3 LRA final safety evaluation report, NUREG-1930, "Safety Evaluation Report Related to the License Renewal of Indian Point Nuclear Generating Units Nos. 2 and 3," November 2009 (ADAMS Accession No ML093170671). The CF and margin term values used by the licensee in the LAR for the limiting materials are different than the values documented in the IP3 LRA because of the use of additional credible surveillance data obtained since the time of the IP3 LRA submittal. The licensee's evaluation of these revised parameters is contained in Section 5 of WCAP-17954-NP.

The staff reviewed the licensee's revised CF and margin term evaluations and found them to be reasonable and performed consistently with the provisions of Position 2.1 of RG 1.99 Rev. 2.

Evaluation of Licensee's PIT Limit Curves The NRC staff verified that the licensee's proposed 37 EFPY PIT limits were calculated in accordance with WCAP-14040-A, Rev. 4, and were based on an evaluation of the limiting RPV beltline shell material accounting for neutron embrittlement through 37 EFPY. The staff also verified that the inlet and outlet nozzles were not controlling compared to the PIT limits for the limiting RPV beltline shell material.

For the limiting beltline shell material, the NRC staff performed a set of confirmatory calculations to verify that the licensee's 37 EFPY PIT limits are consistent with WCAP-14040-A, Rev. 4 and ASME Code,Section XI, Appendix G. The staff initially found discrepancies of as much as 25 percent based on the allowable pressure calculated at the lower temperatures of the proposed PIT limits. However, in performing the confirmatory calculations, the staff noted that all of the inputs necessary for duplicating the licensee's PIT limits were not provided in the LAR, so the staff made reasonable assumptions for the missing inputs in order to complete its confirmatory assessment. Upon further review, the staff determined that the discrepancies were caused by differences in the calculated thermal stress intensity factor, K1T, compared to the values documented in Appendix A of WCAP-17954-NP. Using the licensee's K1T values, the staff was able to reproduce the licensee's PIT limits. However, using its own calculated values of K1T, the staff was not able to reproduce the licensee's PIT limits within reasonable agreement.

On April 24, 2015, the NRC staff performed an audit in Westinghouse's office in Rockville, MD with the licensee and representatives of Westinghouse to discuss the licensee's input assumptions. During that meeting, the staff requested the input assumptions used by the licensee for calculating thermal stresses, which in turn are used to calculate K1T. The licensee and Westinghouse explained their calculations of thermal stress in detail and identified previous IP3 documents that detailed the relevant inputs used by the licensee to calculate thermal stress.

Subsequent to the audit, the NRC staff identified that the differences between its confirmatory calculations and the PIT limits documented in WCAP-17954-NP were the result of three items:

i. Method of calculating thermal stress intensity factor. K1T. During the April 24th audit, Westinghouse identified that the IP3 K1T results were based on the use of Equations 2.7-4 and 2.7-5 of WCAP-14040-A, Rev. 4, which rely on polynomial fits of the transient through-wall thermal stress distributions. The NRC staff's confirmatory calculations were based on the use of K1T influence coefficients obtained from linear elastic fracture mechanics (LEFM) finite element modelling, similar to the methods used by the Fracture Analysis of Vessels - Oak Ridge (FAVOR) computer code (Oak Ridge National Laboratory Report No. ORNL/TM-2012/567, "Fracture Analysis of Vessels - Oak Ridge, FAVOR, v12.1, Computer Code: Theory and Implementation of Algorithms, Methods, and Correlations," November 2012, ADAMS Accession No. ML13008A015). Additional staff assessment revealed that these two methods may lead to a significant portion (approximately 40 percent) of the differences in K1T observed in the staff's confirmatory calculations. Furthermore, Equations 2.7-4 and 2.7-5 of WCAP-14040-A, Rev. 4 are

identical to the equations contained in Paragraph G-2214.3 of ASME Code,Section XI, Appendix G, and ASME Code,Section XI, Appendix G does not explicitly mention the use of LEFM influence coefficient methodology. Therefore, the staff performed an assessment of thermal stresses using finite element methods and the equations contained in Paragraph G-2214.3 of ASME Code,Section XI, Appendix G and, together with consideration of Items (ii) and (iii) below, confirmed the thermal stresses used in the licensee's LAR. Based on the results of these investigations, the staff finds the licensee's thermal stress intensity factors to be acceptable.

ii. Method of calculating thermal stresses. During the April 24th audit, Westinghouse identified that the OPERLIM computer code, which is also identified in WCAP-14040-A, Rev. 4, is used to calculate thermal stresses for PIT limit applications, and that OPERLIM uses a finite difference method for solving the one-dimensional transient heat conduction equation (Equation 2.6.1-1 in WCAP-14040-A, Rev. 4). The NRC staff used finite element analysis in its confirmatory evaluation. Both methods provide a means of numerically solving the transient heat conduction equation, and may lead to small (on the order of 10 percent) differences in the resulting thermal stress. The staff finds both methods to be acceptable for use in calculating thermal stresses and finds the differences in results observed between the staff's and licensee's evaluations to be reasonable and generally within the solution accuracy for either method.

iii. Material properties. During the April 24th audit, Westinghouse identified that temperature-dependent material properties consistent with the IP3 Code of Construction were used for the licensee's thermal stress analysis. During its confirmatory evaluation, the NRC staff initially used material properties from Section II, Part D of the 2013 Edition of the ASME Code. The staff noted that there were significant differences in the material property values that led to approximately 50 percent of the differences observed in thermal stresses. During the April 24th audit, Westinghouse directed the staff to submittals to the NRC in the 1997 timeframe that addressed similar questions and findings for Indian Point. Using that information, the staff located a proprietary attachment of supplementary information provided to the NRC by the licensee for Indian Point Unit 2 in July 1997 ("Attachment Ill, Supplementary Information on PIT Limits Item I (Westinghouse Proprietary)," Consolidated Edison Company of New York, Inc., Indian Point Unit No. 2, Docket No. 50-247, July 1997, ADAMS Accession No. ML100351411).

That document listed the temperature-dependent material properties used by the licensee in its PIT limit calculations. During the audit, Westinghouse identified that the material properties used for Indian Point Unit 2 are also applicable for IP3. The NRC previously evaluated observed differences in the material properties as a part of the staff's 1998 SE (NRC Letter to Mr. Paul H. Kinkel, "Issuance of Amendment for Indian Point Nuclear Generating Unit No. 2 (TAC No. M96944)," February 27, 1998, ADAMS Accession No. ML003779041) and found them to be reasonable for use. Therefore, the staff finds the licensee's material properties used to determine thermal stresses to be acceptable.

The NRC staff performed revised confirmatory calculations considering the above three findings. Based on its final confirmatory calculations, the staff determined that the PIT limits for the limiting beltline shell material, Lower Shell Plate (Heat No. B-2303-3), meet the criteria of the

ASME Code,Section XI, Appendix G for heatup, cooldown, and pressure test conditions, as required by 10 CFR Part 50, Appendix G. Therefore, the staff finds the licensee's proposed PIT limits to be acceptable.

Appendix B of WCAP-17954-NP included calculations of the ART values and the applied K1r for the IP3 inlet and outlet nozzles. The NRC staff determined that the KIT calculations for the inlet and outlet nozzles are acceptable because they are based on the appropriate 1/4T postulated flaw, plant-specific nozzle configurations, and bounding pressure and thermal loading conditions specific to IP3. Furthermore, based on additional independent calculations performed by the staff using methods from Oak Ridge National Laboratory Report No. ORNLITM-2010/246, "Stress and Fracture Mechanics Analyses of Boiling Water Reactor and Pressurized Water Reactor Pressure Vessel Nozzles - Revision 1," June 2012 (ADAMS Accession No. ML12181A162), the staff confirmed that the calculations for the inlet and outlet nozzles are consistent with the ASME Code,Section XI, Appendix G, and 10 CFR Part 50, Appendix G, and that the resulting PIT limits are less restrictive than those calculated for the limiting beltline shell material. Therefore, the staff finds the licensee's assessment of the inlet and outlet nozzles to be acceptable.

Based on its review of the licensee's analysis of the limiting beltline shell material and the inlet and outlet nozzles, the NRC staff verified that the licensee's determination that the RPV beltline shell region is limiting with respect to the IP3 PIT limits is acceptable for determining the proposed 37 EFPY PIT limit curves.

Regarding ferritic RCPB components that are not part of the RPV, 10 CFR Part 50, Appendix G, Paragraph IV.A states the following:

The pressure-retaining components of the reactor coolant pressure boundary that are made of ferritic materials must meet the requirements of the ASME Code

[Section Ill, Division 1], supplemented by the additional requirements set forth below [in paragraph IV.A.2, "Pressure-Temperature Limits and Minimum Temperature Requirements"] ...

Therefore, 10 CFR Part 50, Appendix G requires that PIT limits be developed by considering beltline and non-beltline ferritic RCPB components, and that all ferritic RCPB components must meet the applicable ASME Code, Section Ill requirements. For RCPB piping, pumps and valves greater than 2.5 inches in thickness, the relevant ASME Code, Section Ill requirement that will affect the PIT limits is the lowest service temperature (LST) requirement specified in Paragraph NB-2332(b).

The licensee addressed the requirements of Paragraph IV.A of 10 CFR Part 50, Appendix G in Appendix C of WCAP-17954-NP. The licensee noted that that the IP3 RCS does not have ferritic materials in the Class 1 piping, pumps or valves, so the LST requirements of ASME Code, Section Ill, Paragraph NB-2332(b) are not applicable to the IP3 PIT limits. The licensee further noted that the only ferritic RCPB components that are not part of the reactor vessel consist of the Model 44F Replacement Steam Generators (RSGs) and the original 44 Series Pressurizer. Both the RSGs and the pressurizer meet all applicable design requirements of the ASME Code, Section Ill, 1965 Edition through the Summer 1966 Addenda, and have not

undergone neutron embrittlement that would affect Prr limits. Therefore, no further consideration was performed for these components with regards to Prr limits. The NRC staff finds the licensee's assessment of ferritic RCPB components that are not part of the RPV to be acceptable.

Evaluation of RPV Minimum Temperature Requirements In developing the 37 EFPY Prr limits, the licensee utilized the minimum temperature criteria from Table 1 of 10 CFR Part 50, Appendix G. The licensee's minimum temperature analysis was based on the highest RTNor value for the closure flange region, as discussed in Section 6.3 of WCAP-17954-NP. The NRC staff finds the licensee's assessment of RPV minimum temperature requirements to be acceptable.

The NRC staff noted that the minimum boltup temperature was specified as 60°F in Figures 8-1 and 8-2 of WCAP-17954-NP. The minimum boltup temperature was determined consistent with Section 2.10 of WCAP-14040-A, Rev. 4. Therefore, the staff finds the minimum boltup temperature to be acceptable.

NRC Staff Conclusion for Reactor Coolant System Prr Limits Based on the evaluation described above, the NRC staff determined that the licensee's calculated Prr limits for the limiting RPV beltline shell material (Lower Shell Plate B-2803-3) meet the criteria of the ASME Code,Section XI, Appendix G and are in compliance with the fracture toughness requirements of 10 CFR 50.60 and 10 CFR Part 50, Appendix G. Therefore, the staff concludes that the proposed 37 EFPY Prr limit curves are acceptable for incorporation into the IP3TS.

3.2 Low Temperature Overpressure Protection (LTOP) System Licensee's Evaluation The LTOP system controls RCS pressure at low temperatures so the integrity of the RCPB is not compromised in low-temperature modes of operation. TS 3.4.12, "Low Temperature Overpressure Protection (L TOP)," provides overpressure protection by specifying a minimum coolant input capability and an adequate pressure relief capability. The licensee proposes to revise TS 3.4.12 to account for plant operation to 37 EFPY. The revisions include the maximum power-operated relief valve (PORV) opening setpoint and the enable temperature for the LTOP system. These changes reflect the revised LTOP analyses, and are consistent with the new Prr limits. Additionally, the licensee proposes to revise LCO 3.4.12, TS 3.4.12 Conditions E.1 and F.1, and SR 3.4.12.3 from depressurize with an RCS vent of~ 2.00 square inches to add allowance for depressurization with one blocked open PORV with its block valve disabled in the open position.

The licensee's LAR dated February 12, 2015 states the following:

The IP3 TS establishes limits for RCS pressure at low temperatures in order to protect the carbon and low alloy steel RCS components from damage due to

non-ductile crack propagation. These limits include normal heatup and cooldown restrictions, as well as the automatic setpoints and manual restrictions associated with operation of the LTOP system ..... .

WCAP-17954-NP [Revision 0, "Indian Point Unit 3 Heatup and Cooldown Limit Curves for Normal Operation," dated December 2014(ML15061A277)] identifies the methodology by which the figures for the normal plant operation heatup and cooldown curves were generated. These curves are based upon the latest available reactor vessel information and updated calculated fluences, which include the impact of stretch power uprate. The new IP3 heatup and cooldown curves were generated using the "Axial-Flaw" methodology to the 1998 ASME Code,Section XI through the 2000 Addenda (which allows the use of Kie methodology) and the "Circ-Flaw" methodology .....

The licensee's LAR dated August 11, 2015 states the following:

The maximum PORV opening setpoint shown in Figure 3.4.12-1 is implemented as a variable setpoint for the LTOP instrumentation. The revised LTOPs PORV setpoints meet all acceptance criteria for protecting the Appendix G limits, PORV discharge piping limits, and the RCP operation limits and provide increased operating margin compared to the current setpoints.

The TS requirements (i.e., LCO 3.4.12, TS 3.4.12 Condition[s] E [and F], and SR 3.4.12.3) were changed from depressurize with an RCS vent of~ 2.00 square inches to add allowance for depressurization with one blocked open PORV with its block valve disabled in the open position. This was based on the analysis of venting requirements.

The analysis looked at the ability of the blocked open PORV with its block valve disabled open to provide the required venting for the design basis transient. Only the mass input design basis transient (start of three charging pumps) is evaluated for this event since the heat input transient from start of a RCP is not considered credible or the bounding transient. The relief capacity of a blocked open PORV with its block valve disabled in the open position relief capacity is less than the relief capacity of a 2.0 in. 2 vent but it is still greater than the injection capacity of the three charging pumps, the design basis mass injection. Therefore the addition of the allowance for depressurization with one blocked open PORV with its block valve disabled in the open position to LCO 3.4.14, Conditions E.1 and F.1, and SR 3.4.12.3 is acceptable. When a HHSI pump is made operable different vent requirements apply.

NRC Staff Evaluation

The NRC staff reviewed the LTOP pressurizer Power Operated Relief Valves (PORVs) setpoints and the proposed changes to the TS requirements for LCO 3.4.12, TS 3.4.12 Conditions E.1 and F.1, and SR 3.4.12.3 for the allowance for depressurization with one blocked open PORV with its block valve disabled in the open position for IP3. The LTOP setpoints were determined using the Westinghouse methodology described in WCAP-14040-A, Revision 4. The staff reviewed the limiting mass and energy analysis for the LTOP setpoints provided by the licensee. The staff determined that the LTOP setpoints ensure that the

Appendix G requirements will be met, and that the limiting mass and energy analysis for the LTOP setpoints were determined with a level of detail consistent with Section 5.2.2, "Overpressure Protection," of NUREG-0800, "Standard Review Plant for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition," (ML070540076).

The NRC staff reviewed the proposed revisions to TS LCO 3.4.12, TS 3.4.12 Conditions E.1 and F.1, and SR 3.4.12.3 for depressurizing and venting the RCS. These TSs require depressurizing and venting the RCS by a vent of~ 2.00 square inches. The proposed revisions provide an option to block open a PORV with its block valve disabled in the open position. The staff determined that the relief capability of the blocked open PORV and associated block valve is adequate for the design basis mass injection of the three charging pumps. Therefore, the overpressure protection required by LCO 3.4.12 will continue to be met. Based on the considerations discussed above, the staff has determined that the LTOP setpoints and proposed TS revisions for LCO 3.4.12, TS 3.4.12 Condition E.1 and F.1, and SR 3.4.12.3 are acceptable.

The NRC staff reviewed the proposed revisions to the Required Actions to Condition A for TS 3.4.12 when one or more HHSI pumps are capable of injecting into the RCS. The intent of LCO 3.4.12 is to prevent overpressurization by ensuring that either no HHSI pumps are capable of injecting water into the RCS or that the RCS is adequately vented. The staff concluded that the revisions to the Required Actions provide clarification and remain consistent with the LCO by verifying that either the HHSI pumps are not capable of injecting into the RCS or the RCS is vented when one HHSI pump is operable as allowed by Notes 2 and 3 to the LCO. Based on the considerations discussed above, the staff has determined that the proposed revisions to the Required Actions to Condition A of TS 3.4.12 are acceptable.

The NRC staff reviewed the proposed revisions to SR 3.4.12.9 that identify the conditions that must be met prior to starting a RCP. Condition a. currently states "Secondary side water temperature of the hottest steam generator is s 64° F above the coldest RCS cold leg temperature." The b,,T value is determined by analysis and conservatively assumes that RCS water from the hottest steam generator instantaneously flushes with water in the coldest RCS cold leg and transfers sufficient heat to the primary side such that the overall RCS temperature reaches the setpoint where the LTOP protection is disabled. In the analysis, once the RCS temperatures reach 340° F, the fracture toughness requirements for the reactor coolant pressure boundary meet the requirements of Appendix G and LTOP protection is no longer required. The proposed change in b,,T from 64° F to 50° Fis more restrictive because it reduces the maximum possible thermal shock to the RCS, the reactor vessel, and the reactor vessel internal components and it is supported by analysis. Therefore, the staff finds this change to be acceptable. Finally, the licensee proposes to delete Condition e. that references TS Figure 3.4.12.4. TS Figure 3.4.12.4 is being deleted because, as a result of the above analysis, it is no longer necessary. Therefore, the staff finds deletion of Condition e. to be acceptable.

RG 1. 190 Compliance The guidance provided in RG 1.190 indicates that the following attributes comprise an acceptable fluence calculation:

  • A fluence calculation performed using an acceptable methodology
  • Analytic uncertainty analysis identifying possible sources of uncertainty
  • Benchmark comparison to approved results of a test facility
  • Plant-specific qualification by comparison to measured fluence values For input to its PIT Limits, Entergy performed fluence calculations in accordance with Westinghouse Licensing Topical Report, WCAP-17954-NP, Revision 0. The methods used in WCAP-17954-NP, Revision 0, to develop the calculated pressure vessel fluence, are consistent with the NRG-approved methodology described in WCAP-14040-A, Revision 4.

The neutron calculations are performed in a manner consistent with the guidance set forth in RG 1.190. A solution to the Boltzmann transport equation is approximated using the two-dimensional discrete ordinates code (DORT). The licensee uses a cross-section library based on the ENDF/B-Vl.3 nuclear data. Numeric approximations include a P5 Legendre expansion for anisotropic scattering and the modeling uses S16 order of angular quadrature.

These cross-section data and approximations are in accordance with the modeling guidance contained in RG 1.190. Since the licensee used NRG-approved RG 1.190 adherent methods to determine the vessel fluence, the NRC staff has determined that the fluence calculations are acceptable.

Space and energy dependent core power (neutron source) distributions and associated core parameters are treated on a fuel cycle specific basis. Three dimensional flux solutions are constructed using a synthesis of azimuthal, axial, and radial flux. Source distributions include cycle-dependent fuel assembly initial enrichments, burnups, and axial power distributions, which are used to develop spatial and energy dependent core source distributions that are averaged over each fuel cycle. This method accounts for source energy spectral effects by using an appropriate fission split for uranium and plutonium isotopes based on the initial enrichment and burnup history for each fuel assembly. The neutron source and transport calculations, as described above, are performed in accordance with the guidance set forth in RG 1.190. Based on the consistency with the guidance contained in RG 1.190, the NRC staff determined that the source and transport calculations are acceptable.

The NRG-approved methods are supported by an analytic uncertainty analysis in accordance with RG 1.190. The overall calculational uncertainty is less than 20 percent when combining analytic and benchmark uncertainties, which is in accordance with RG 1.190, and hence acceptable. Details of the analytic uncertainty analysis are provided in WCAP-17954-NP, Revision 0.

Guidance in WCAP-17954-NP, Revision O describes the methods qualification using the standard benchmark problems found in RG 1.190. The calculations were compared with the benchmark measurements from the Poolside Critical Assembly (PCA) simulator at the Oak Ridge National Laboratory (ORNL) and the H.B. Robinson power reactor benchmark experiment. The NRC staff determined that these constitute acceptable test facilities, as they are specifically referenced in RG 1.190.

Guidance in WCAP-17954-NP, Revision 0, references a database of pressurized water reactor dosimetry benchmarking specific to IP3. The IP3 unit-specific geometry, a Westinghouse

reactor pressure vessel, is well represented within the database. Plant-specific benchmarking documented in WCAP-17954-NP, Revision 0, confirms that surveillance capsule fluence can be calculated within 20 percent of measured values. The NRC staff has determined, therefore, that appropriate plant-specific benchmarking was performed.

NRC Staff Conclusion for Low Temperature Overpressure Protection System Based on the evaluation described above, the NRC staff has determined that the licensee's LTOP System meets the criteria of RG 1.190 and GDC 14, 30, and 31. Therefore, the staff concludes that the results of the revised LTOP analysis for 37 EFPY are acceptable for incorporation into the IP3 TS.

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 (80 FR 32619). 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: Gary Stevens, Matthew Hardgrove, NRR Date: September 3, 201 5

September 3, 2015 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. 3 - ISSUANCE OF AMENDMENT RE: CHANGES TO REACTOR VESSEL HEATUP AND COOLDOWN CURVES AND LOW TEMPERATURE OVERPRESSURE PROTECTION SYSTEM REQUIREMENTS (TAC NO. MF5746)

Dear Sir or Madam:

The Commission has issued the enclosed Amendment No. 258 to Facility Operating License No. DPR-64 for the Indian Point Nuclear Generating Unit No. 3. The amendment consists of changes to the Technical Specifications (TSs) in response to your application dated February 12, 2015, as supplemented by letter dated August 11, 2015.

The amendment revises TS 3.4.3, "RCS [reactor coolant system] Pressure and Temperature (PIT) Limits," and 3.4.12, "Low Temperature Overpressure Protection (LTOP)," to include new RCS PIT limit curves for heatup, cooldown, and pressure test operations and LTOP system setpoints. The proposed PIT limit curves and LTOP system setpoints will be valid for 37 effective full power years of facility operation, which is the accumulated burnup estimated to occur in December 2023 during the period of extended plant operation.

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.

Sincerely, IRA/

Douglas V. Pickett, Senior Project Manager Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-286

Enclosures:

1. Amendment No. 258 to DPR-64
2. Safety Evaluation cc w/encls: Distribution via Listserv DISTRIBUTION:

PUBLIC RidsNrrDeEvib GStevens, EVIV LPL 1-1 R/F RidsNrrDssStsb MHardgrove, SRXB RidsNrrDorl RidsRgn1 MailCenter RGrover, STSB RidsNrrDorlLpl1-1 RidsNrrPMlndianPoint ABurritt, R1 RidsNrrLAKGoldstein RidsAcrsAcnw_MailCTR RidsNrrDorlDpr RidsNrrDssSrxb ADAMS A ccess1on No.: ML15226A159 *b>y memo d aet d OFFICE DORL/LPLl-1/PM DORL/LPLl-1 /LA DE/EVIB/BC DSS/SRXB/BC NAME DPickett KGoldstein JMcHale* CJackson*

DATE 8/17/2015 8/20/2015 6/16/2015 6/17/2015 OFFICE DSS/STSB/BC OGC (NLO) DORL/LPLl-1 /BC NAME RElliott STurk BBeasley *<

DATE 8/19/2015 8/25/2015 9/03/2015  : .

OFFICIAL RECORD COPY

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