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Technical Specification Pages Re 4.85 Percent Stretch Power Uprate & Relocation of Cycle-Specific Parameters
	ML050870384
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Site:	Indian Point
Issue date:	03/24/2005
From:	; Office of Nuclear Reactor Regulation
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r Definitions 1.1 1.1 Definitions MODE (continued)

OPERABLE-OPERABILITY PHYSICS TESTS vessel head closure bolt tensioning specified in Table 1.1-1 with fuel in the reactor vessel.

A system, subsystem, train, component, or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified safety function(s) and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication, and other auxiliary equipment that are required for the system, subsystem, train, component, or device to perform its specified safety function(s) are also capable of performing their related support function(s).

PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation.

These tests are:

a. Described in FSAR Chapter 13, Initial Tests and Operations;

b. Authorized under the provisions of 10 CFR 50.59; or

c. Otherwise approved by the Nuclear Regulatory Commission.

QUADRANT POWER TILT RATIO (QPTR)

RATED THERMAL POWER (RTP)

QPTR shall be the ratio of the maximum upper excore detector calibrated output to the average of the upper excore detector calibrated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater.

RTP shall be a total reactor core heat transfer rate to the reactor coolant of 3216 MWt.

I (continued)

INDIAN POINT 3 1.1 - 5 Amendment 225

SLs 2.0 2.1 SLs 2.1.1 Reactor Core SLs In MODES 1 and 2, the combination of THERMAL POWER, Reactor Vessel inlet temperature, and pressurizer pressure shall not exceed the limits specified in the COLR; and the following SLs shall not be exceeded:

2.1.1.1 The departure from nucleate boiling ratio (DNBR) shall be maintained > 1.17 for the WRB-1 DNB correlations.

2.1.1.2 The peak fuel centerline temperature shall be maintained < 5080 0F, decreasing by 580F per 10,000 MWD/MTU of burnup.

2.1.2 RCS Pressure SL In MODES 1, 2, 3, 4, 5, and in MODE 6 when the reactor vessel head is on, the RCS pressure shall be maintained < 2735 psig.

2.2 SL Violations 2.2.1 If SL 2.1.1 is violated, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

2.2.2 If SL 2.1.2 is violated:

2.2.2.1 In MODE 1 or 2, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

2.2.2.2 In MODE 3, 4, 5, or 6, restore compliance within 5 minutes.

INDIAN POINT 3

2. 0-1 Amendment 22.5

i

-f, RPS Instrumentation 3.3.1 Table 3.3.1-1 (page 1 of 8)

Reactor Protectioa-System Instrumentation FUNCTION APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS

1.

Manual Reactor Trip 1,2 3(a) 4 (a) 5(a)

REQUIRED CHANNELS 2

2 CONDITIONS B

C SURVEILLANCE REQUIREMENTS SR 3.3.1.14 SR 3.3.1.14 ALLOWABLE VALUE NA NA

2.

Power Range Neutron Flux

a. High 1,2 D

SR 3.3.1.1 SR 3.3.1.2 SR 3.3.1.7 SR 3.3.1.11

111% RTP

b. Low 4a)

E SR 3.3.1.1 SR 3.3.1.8 SR 3.3.1.11 F

SR 3.3.1.1 SR 3.3.1.8 SR 3.3.1.11

25% RTP NA

3.

Intermediate Range Neutron Flux 1 (b), 2(c) 1 (continued)

(a)

(b)

(c) a)

With Rod Control System capable of rod withdrawal and one or more rods not fully inserted.

Below the P-10 (Power Range Neutron Flux) interlocks.

Above the P-6 (Intermediate Range Neutron Flux) Interlocks.

Only 3 channels required during Mode 2 Physics Tests, LCO 3.1.8 INDIAN POINT 3 3.3.1 -1 3 Amendment 225

r RPS Instrumentation 3.3.1 Table 3.3.1-1 (page 3 of 8)

Reactor Protection System Instrumentation APPLICABLE FUNCTION MODES OR OTHER SPECIFIED REQUIRED SURVEILLANCE CONDITIONS CHANNELS CONDITIONS REQUIREMENTS ALLOWABLE VALUE

7.

Pressurizer Pressure

a. Low 1(e) 4 H

SR 3.3.1.1 21900 psig SR 3.3.1.7 SR 3.3.1.10 SR 3.3.1.1

b. High 1,2 3

E SR 3.3.1.7 52400 psig SR 3.3.1.1 0

8.

Pressurizer 1

3 H

SR 3.3.1.1 s97%

Water Level -

SR 3.3.1.7 High SR 3.3.1.10

9.

Reactor Coolant 1(e) 3 per loop H

SR 3.3.1.1 290%

Flow - Low SR 3.3.1.7 SR 3.3.1.10 (continued)

I (e) Above the P-7 (Low Power Reactor Trips Block) interlock.

INDIAN POINT 3 3.3.1 -1 5 Amendment 225

RPS Instrumentation 3.3.1 Table 3.3.1-1 (page 7 of 8)

Reactor Protection System Instrumentation Note 1: Overtemrerature AT The Overtemperature AT Function Allowable Value shall not exceed the following Trip Setpoint by more than 2.8% of AT span:

AT'ATo{ K,-K2 (I +

[ T-2T] +SK,- P)-fi(A1)}

Where:

AT is measured RCS AT, 'F.

ATo is the indicated AT at RTP, OF.

s is the Laplace transform operator, sec'.

T is the measured RCS average temperature, IF.

T' is the nominal T.,9 at RTP, * [ * ]'F.

P is the measured pressurizer pressure, psig P' is the nominal RCS operating pressure, < [ * ] psig KI SI*]

K 2>[*))/F K32[*]/psig To 2 [*]sec r2 <

sec f1(Al)=

[*]{[]

+ (qt - qb)}

when qt - qb

-]%

RTP 0% of RTP when -[ * ]% RTP < qt - qb

[ *]% RTP

-1 * ] {(qt - qb) - I

}

when qt - qb >1[ * ]% RTP Where qt and qb are percent RTP in the upper and lower halves of the core, respectively, and qt + qb is the total THERMAL POWER in percent RTP.

The values denoted with [ * ] are specified in the COLR.

INDIAN POINT 3 3.3.1 -1 9 Amendment 225

RPS Instrumentation 3.3.1 Table 3.3.1-1 (page 8 of 8)

Reactor Protection System Instrumentatiop Note 2: Overpower AT The Overpower AT Function Allowable Value shall not exceed the following Trip Setpoint by more than 1.8% of AT span:

A T - A To Key Ks (r 3 S T -K6 (T -T")-

(A I) t (I+ 13 S) 6--

2

/

Where:

AT is measured RCS AT, IF.

ATO is the indicated AT at RTP, OF.

s is the Laplace transform operator, sec'l.

T is the measured RCS average temperature, "F.

T" is the nominal T.,, at RTP, [*]'F.

K4s [*1 K5Ž [*]/'FforincreasingTa.,

K6Ž [*]/'FwhenT>T

[* ] /OF for decreasing Tavg

[*] /'F when T

T

,r3

[ J sec f2 (AI) =[*]

The values denoted with [* ] are specified in the COLR.

INDIAN POINT 3 3.3.1-20 Amendment 225

r ESFAS Instrumentation 3.3.2 Table 3.3.2-1 (page 1 of 6)

Engineered Safety Feature Actuation System Instrumentation APPLICABLE MODES OR FUNCTION OTHER SPECIFIED REQUIRED SURVEILLANCE CONDITIONS CHANNELS CONDITIONS REQUIREMENTS ALLOWABLE VALUE

1.

Safety Injection

a.

Manual Initiation

b.

Automatic Actuation Logic and Actuation Relays

c.

Containment Pressure-Hi

d.

Pressurizer Pressure-Low

e.

High Differential Pressure Between Steam Lines

f.

High Steam Flow in Two Steam Lines Coincident with Tavg Low 1,2,3,4 1,2,3,4 1,2,3 1,2,3 }

1,2,3 1,2(d),3(d) 1,2(d),3(d) 2 2 trains 3

3 3 per steam line 2 per steam line 1 per loop B

SR 3.3.2.6 C

SR 3.3.2.2 SR 3.3.2.3 SR 3.3.2.5 NA NA

4.80 psig D

SR SR SR 3.3.2.1 3.3.2.4 3.3.2.7 D

SR 3.3.2.1 SR 3.3.2.4 SR 3.3.2.7 D

SR 3.3.2.1 SR 3.3.2.4 SR 3.3.2.7 21710 psig I

NA (c)

Ž540.50 F (continued)

(a) Not used (b) Above the Pressurizer Pressure interlock.

(c) Less than or equal to turbine first stage pressure corresponding to 54% full steam flow below 20% load, and increasing linearly from 54% full steam flow at 20% load to 120% full steam flow at 100% load, and corresponding to 120% full steam flow above 100% load. Time delay for SI *6 seconds.

(d)

Except when all MSIVs are closed.

INDIAN POINT 3 2

3.3.2-8 Amendment 225

ESFAS Instrumentation 3.3.2 Table 3.3.2-1 (page 4 of 6)

Engineered Safety Feature Actuation System Instrumentation APPLICABLE MODES OR FUNCTION OTHER SPECIFIED REQUIRED SURVEILLANCE CONDITIONS CHANNELS CONDITIONS REQUIREMENTS ALLOWABLE VALUE

4. Steam Line Isolation

a. Manual Initiation

b. Automatic Actuation Logic and Actuation Relays

c. Containment Pressure (Hi-Hi)

d. High Steam Flow in Two Steam Lines Coincident with Tevg-Low 1,2 (d), 3 (d) 12 2(d)3 (d) 1 2(d) 1 2(d)

P3 d) 1 2(d)

Pd) '

2 per steam line 2 trains 2 sets of 3 2 per steam line 1 per loop F

SR 3.3.2.6 G

SR 3.3.2.2 SR 3.3.2.3 SR 3.3.2.5 NA NA S24 psig E

SR SR SR D

SR SR SR 3.3.2.1 3.3.2.4 3.3.2.7 3.3.2.1 3.3.2.4 3.3.2.7 3.3.2.1 3.3.2.4 3.3.2.7 (c) 2540.50F

e. High Steam Flow in Two Steam Lines Coincident with Steam Line Pressure-Low 1 2(d) 3 (d) 1 2 (d) pled) '

2 per steam line 1 per steam line D

SR 3.3.2.1 SR 3.3.2.4 SR 3.3.2.7 D

SR 3.3.2.1 SR 3.3.2.4 SR 3.3.2.7 (c) 2500 psig (c)

Less than or equal to turbine first stage pressure corresponding to 54% full steam flow below 20% load, and increasing linearly from 54% full steam flow at 20% load to 120% full steam flow at 100% load, and corresponding to 120% full steam flow above 100% load. Time delay for SI S6 seconds.

(d)

Except when all MSIVs are closed.

INDIAN POINT 3 3.3.2-11 Amendment 225

r RCS Pressure, Temperature, and Flow Limits 3.4.1 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.1 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits LCO 3.4.1 RCS DNB parameters for pressurizer pressure, RCS average temperature, and RCS total flow rate shall be within the limits specified below:

a.

Pressurizer pressure is greater than or equal to the limit specified in the COLR;

b.

RCS average loop temperature is less than or equal to the limit specified in the COLR; and

c.

RCS total flow rate 2 354,400 gpm and greater than or equal to the limit specified in the COLR.

APPLICABILITY:

MODE 1.

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

Pressurizer pressure limit does not apply during:

a.

THERMAL POWER ramp > 5% RTP per minute; or

b.

THERMAL POWER step > 10% RTP.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more RCS A.1 Restore RCS DNB 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months DNB parameters not parameter(s) to within limits.

within limits.

B. Required action and B.1 Be in MODE 2.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met.

INDIAN POINT 3 3.4.1-1

.Amendment 225

I -

RCS Pressure, Temperature, and Flow Limits 3.4.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.1.1 Verify pressurizer pressure is greater 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months than or equal to the limit specified in the COLR.

SR 3.4.1.2 Verify RCS average loop temperature is 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months less than or equal to the limit specified in the COLR.

SR 3.4.1.3 Verify RCS total flow rate is 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 2 354,400 gpm and greater than or equal to the limit specified in the COLR.

SR 3.4.1.4

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

Not required to be performed until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after 2 90% RTP.

Verify by precision heat balance that 24 months RCS total flow rate is 2 354,400 gpm and greater than or equal to the limit specified in the COLR.

INDIAN POINT 3

3. 4.1-2 Amendment 225

r Pressurizer 3.4.9 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.9 Pressurizer LCO 3.4.9 The pressurizer shall be OPERABLE with:

a.

Pressurizer water level 5 54.3% in MODES 1 and 2 or < 90% in MODE 3; and

b.

Two groups of pressurizer heaters OPERABLE with the capacity of each group 2 150 kW and capable of being powered from an emergency power supply.

APPLICABILITY: MODES 1, 2, and 3.

I ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Pressurizer water A.1 Be in MODE 3 with 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months level not within reactor trip limit.

breakers open.

AND A.2 Be in MODE 4.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months B. One required group B.1 Restore required 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months of pressurizer group of heaters inoperable.

pressurizer heaters to OPERABLE status.

C. Required Action and C.1 Be in MODE 3.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of AND Condition B not met.

C.2 Be in MODE 4.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months INDIAN POINT 3

3. 4.9-1 Amendment 225

Pressurizer 3.4.9 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.9.1 Verify pressurizer water level is 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months

< 54.3% in MODES 1 and 2 OR < 90% in MODE 3.

SR 3.4.9.2 Verify capacity of each required group 24 months of pressurizer heaters is 2 150 kW.

INDIAN POINT 3

3. 4.9-2 Amendment 225

MSSVs 3.7.1 Table 3.7.1-1 (page 1 of 1)

OPERABLE Main Steam Safety Valves versus Applicable Neutron Flux Trip Setpoint in Percent of RATED THERMAL POWER MINIMUM NUMBER OF MSSVs APPLICABLE Neutron Flux Trip PER STEAM GENERATOR Setpoint REQUIRED OPERABLE

(% RTP) 4

<S 57 3

S 38 2

< 20 INDIAN POINT 3 3.7.1-3 Amendment 225.

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.15 Containment Leakage Rate Testing Program (continued) cooler unit when pressurized at > 1.1 Pa.

This limit protects the internal recirculation pumps from flooding during the 12-month period of post accident recirculation.

The provisions of SR 3.0.3 are applicable to the Containment Leakage Rate Testing Program.

Nothing in these Technical Specifications shall be construed to modify the testing Frequencies required by 10CFR50, Appendix J.

The calculated peak containment internal pressure for the design basis loss of coolant accident, Pa, is 42.0 psig.

The containment design pressure is 47 psig.

The maximum allowably primary containment leakage rate, La, at Pa, shall be 0.1% of primary containment air weight per day.

INDIAN POINT 3 5.0 - 31 Amendment 225

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

1. Specification 2.1, Safety Limits (SL);

2. Specification 3.1.1, Shutdown Margin;

3. Specification 3.1.3, Moderator Temperature Coefficient;

4. Specification 3.1.5, Shutdown Bank Insertion Limits;

5. Specification 3.1.6, Control Bank Insertion Limits;

6. Specification 3.2.1, Heat Flux Hot Channel Factor (FQ(Z));

7. Specification 3.2.2, Nuclear Enthalpy Rise Hot Channel Factor;

8. Specification 3.2.3, AXIAL FLUX DIFFERENCE (AFD);

9. Specification 3.3.1, Reactor Protection System Instrumentation;

10. Specification 3.4.1, RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits; and

11. Specification 3.9.1, Boron Concentration.

b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

1.

WCAP-9272-P-A, "WESTINGHOUSE RELOAD SAFETY EVALUATION METHODOLOGY," July 1985 (W Proprietary). (Specifications 3.1.5, Shutdown Bank Insertion Limits, 3.1.6, Control Bank Insertion Limits, and 3.2.2, Nuclear Enthalpy Rise Hot Channel Factor);

2a.

WCAP-8385, "POWER DISTRIBUTION CONTROL AND LOAD FOLLOWING PROCEDURES, TOPICAL REPORT," September 1974 (W Proprietary). (Specification 3.2.3, Axial Flux Difference (AFD) (Constant Axial Offset Control);

2b. T. M. Anderson to K. Kneil (Chief of Core Performance Branch, NRC) January 31, 1980 --

Attachment:

Operation and Safety Analysis Aspects of an Improved Load Follow Package. (Specification 3.2.3, Axial Flux Difference (AFD) (Constant Axial Offset Control));

2c. NUREG-0800, Standard Review Plan, U.S. Nuclear Regulatory Commission,Section 4.3,Nuclear Design,July 1981. Branch (continued)

INDIAN POINT 3 5.0 -

34 Amendment 225

r Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

Position CPB 4.3-1, Westinghouse Constant Axial Offset Control (CAOC), Rev. 2, July 1981.

(Specification 3.2.3, Axial Flux Difference (AFD) (Constant Axial Offset Control));

3a.

WCAP-12945-P-A, Volume 1 (Revision 2) and Volumes 2 through 5 (Revision 1), "Code Qualification Document for Best-Estimate Loss-of-Coolant-Accident Analysis,"

March 1998 (Westinghouse Proprietary);

3b. WCAP-11397-P-A, "Revised Thermal Design Procedure,"

April 1989 (Specification 2.1, Safety Limits (SL)) and Specification 3.4.1, (RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits);

3c. WCAP-8745-P-A, "Design Bases for the Thermal Overpower AT and Thermal Overtemperature AT Trip Functions,"

September 1986 (Specification 2.1, Safety Limits (SL));

3d.

WCAP-10054-P-A, "SMALL BREAK ECCS EVALUATION MODEL USING NOTRUMP CODE," (W Proprietary). (Specification 3.2.1, Heat Flux Hot Channel Factor (FQ(Z));

3e. WCAP-10054-P-A, Addendum 2, Revision 1, "Addendum to the Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code; Safety Injection into the Broken Loop and Cosi Condensation Model," July 1997 (Specification 3.2.1, Heat Flux Hot Channel Factor (FQ(Z)));

3f. WCAP-10079-P-A, "NOTRUMP NODAL TRANSIENT SMALL BREAK AND GENERAL NETWORK CODE," (W Proprietary). (Specification 3.2.1, Heat Flux Hot Channel Factor (FQ(Z))); and 3g.

WCAP-12610, "VANTAGE+ Fuel Assembly Report,"

(W Proprietary). (Specification 3.2.1, Heat Flux Hot Channel Factor).

c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d. The COLR, including any midcycle revisions or supplements, shall be provided for each reload cycle to the NRC.

5.6.6 NOT USED (continued)

INDIAN POINT 3

5. 0 - 35 Amendment 225

ML050870384

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