Text

Technical Specifications for Amendment No. 167, Corrected Pages
	ML031180636
Person / Time
Site:	Kewaunee
Issue date:	04/24/2003
From:	; NRC/NRR/DLPM/LPD3
To:	;
	Lamb J, NRR/DLPM, 415-1446
References
	TAC MB5717, TAC MB5718
	Download: ML031180636 (58)
	v • d • e

TABLE OF CONTENTS TECHNICAL SPECIFICATIONS APPENDIX A Section Title Page 1.0 Definitions ..... .1.0-1 1.0.a Quadrant-to-Average Power Tilt Ratio . 1.0-1 1.0.b Safety limits ....... , 1.0-1 Limiting Safety System Settings . 1.0-1 1.0.c 1.0-1 1.0.d Limiting Conditions for Operation .

Operable - Operability . 1.0-1 1.0.e 1.0.f Operating . 1.0-1 Containment System Integrity . 1.0-2 1.0.g Protective Instrumentation Logic . 1.0-2 1.0.h 1.0-3 1.0.i Instrumentation Surveillance .

1.0j Modes . 1.0-4 Reactor Critical . 1.0-4 1.0.k 1.0-4 1.0.1 Refueling Operation .

1.0.m Rated Power . 1.0-4 1.0.n Reportable Event . 1.0-4 Radiological Effluents . 1.0-5 1.0.o 1.0-6 1.0.p Dose Equivalent 1-131 ....................

1.0.q Core Operating Limits Report .1.0-6 1.0.r Shutdown Margin .1.0-6

. 2.1-1 2.0 Safety Limits and Limiting Safety System Settings ..

2.1 Safety Limits, Reactor Core .. . 2.1-1 Safety Limit, Reactor Coolant System Pressure .. . 2.2-1 2.2 2.3 Limiting Safety System Settings, Protective Instrumentation .. 2.3-1 2.3.a Reactor Trip Settings . 2.3-1 2.3.a.1 Nuclear Flux . 2.3-1 2.3.a.2 Pressurizer . 2.3-1 2.3.a.3 Reactor Coolant Temperature . 2.3-2 2.3.a.4 Reactor Coolant Flow . 2.3-3 2.3.a.5 Steam Generators . 2.3-3 2.3.a.6 ReactorTrip Interlocks .2.3-4 2.3.a.7 Other Trips. 2.3-4 3.0 Limiting Conditions for Operation . .3.0-1 3.1 Reactor Coolant System . .3.1-1 3.1.a Operational Components .. 3.1-1 3.1.a.1 Reactor Coolant Pumps . 3.1-1 3.1.a.2 Decay Heat Removal Capability . 3.1-1 3.1.a.3 Pressurizer Safety Valves . 3.1-3 3.1.a.4 Pressure Isolation Valves . 3.1-4 3.1.a.5 Pressurizer PORV and PORV Block Valves . 3.1-4 3.1.a.6 Pressurizer Heaters . 3.1-5 3.1.a.7 Reactor Coolant Vent System . 3.1-5 3.1.b Heatup & Cooldown Limit Curves for Normal Operation ...... 3.1-6 3.1.c Maximum Coolant Activity .. 3.1-7 3.1 .d Leakage of Reactor Coolant .. 3.1-8 3.1.e Maximum Reactor Coolant Oxygen, Chloride and Fluoride Concentration .. 3.1-9 3.1 .f Minimum Conditions for Criticality . . 3.1-10 TSi Amendment No. 165

Title Page Section 3.2 Chemical and Volume Control System ...................................... 3.2-1 3.3 Engineered Safety Features and Auxiliary'Systems . ...........................3.3-1 3.3.a Accumulators ....................... 3.3-1 Emergency Core Cooling System ....................... 3.3-2 3.3.b Containment Cooling Systems ...................... 3.3-4 3.3.c 3.3-6 3.3.d Component Cooling System .......................

Service Water System ....................... 3.3-7 3.3.e 3.4 Steam and Power Conversion System .. ......................... 3.4-1 Main Steam Safety Valves ........................... 3.4-1 3.4.a 3 4-2 3.4.b Auxiliary Feedwater System ...........................

3.4.c Condensate Storage Tank ........................... 3.4-4 Secondary Activity Limits ........................... 3.4-5 3.4.d

........................... 3.5-1 3.5 Instrumentation System

........................... 36-1 3.6 Containment System

........................... 3.7-1 3.7 Auxiliary Electrical Systems 3.8-1 3.8 Refueling Operations ...........................

3.9 Deleted 3.10 Control Rod and Power Distribution Limits . ............................ 3.10-1 3.10.a Shutdown Reactivity.............................. 3.10-1 Power Distribution Limits .............................. 3.10-1 3.10.b Quadrant Power Tilt Limits .............................. 3.10-4 3.10.c 3.10-4 3.10.d Rod Insertion Limits ..............................

3.10.e Rod Misalignment Limitations .............................. 3.10-5 3.1 0.f Inoperable Rod Position Indicator Channels ............................ 3.10-5 Inoperable Rod Limitations .............................. 3.10-6 3.1 0.g 3.10-6 3.10.h Rod Drop Time ..............................

3.10.i Rod Position Deviation Monitor .............................. 3.10-6 3.1 0j Quadrant Power Tilt Monitor ................................3.10-6 Core Average Temperature .............................. 3.10-6 3.10.k Reactor Coolant System Pressure .............................. 3.10-6 3.10.1 3.10.m Reactor Coolant Flow .............................. 3.10-7 3.10.n DNBR Parameters .............................. 3.10-7 Core Surveillance Instrumentation ................................... 3.11-1 3.11 3.12-1 3.12 Control Room Post-Accident Recirculation System . .........................

Shock Suppressors (Snubbers) ................................. .. 3.14-1 3.14

.................. 4.0-1 4.0 Surveillance Requirements 4.1-1 4.1 Operational Safety Review . . ...............................

. .................................. 4.2-1 4.2 ASME Code Class In-service Inspection and Testing 4.2.a ASME Code Class 1, 2,3, and MC Components and Supports . . . . ............... 4.2-1 4.2.b Steam Generator Tubes . . . ................ 4.2-2 4.2.b.1 Steam Generator Sample Selection and Inspection . 4.2-3 4.2.b.2 Steam Generator Tube Sample Selection and Inspection .............. 4.2-3 4.2.b.3 Inspection Frequency .............. 4.2-4 4.2.b.4 Plugging Limit Criteria .............. 4.2-5 4.2.b.5 Deleted 4.2.b.6 Deleted 4.2.b.7 Reports .4.2-5 4.3 Deleted TS ii Amendment No. 165

Section Title Paqe 6.0 Administrative Controls . . . . . 6.1-1 6.1 Responsibility................................................................................................6.1-1 6.2 Organization ............................... 6.2-1 6.2.a Off-Site Staff ......................... 6.2-1 6.2.b Facility Staff ......................... 6.2-1 6.2.c Organizational Changes .. ....................... 6.2-1 6.3 Plant Staff Qualifications . . . ...................... 6.3-1 6.4 Training ............................ 6.4-1 6.5 Deleted ... ...................... 6.5-1 - 6.5-6 6.6 Deleted ............................ 6.6-1 6.7 Safety Limit Violation . . . ...................... 6.7-1 6.8 Procedures ........................... 6.8-1 6.9 Reporting Requirements . . . ...................... 6.9-1 6.9.a Routine Reports ..................................... 6.9-1 6.9.a.1 Startup Report ...................................... 6.9-1 6.9.a.2 Annual Reporting Requirements .......................... 6.9-1 6.9.a.3 Monthly Operating Report .................................... 6.9-3 6.9.a.4 Core Operating Limits Report ............................. 6.9-3 6.9.b Unique Reporting Requirements ....................................... 6.9-5 6.9.b.1 Annual Radiological Environmental Monitoring Report ..................................... 6.9-5 6.9.b.2 Radioactive Effluent Release Report . ...............

9-5 6.9.b.3 Special Reports ..................................... 6.9-5 6.10 Record Retention . ............................ 6.10-1 6.11 Radiation Protection Program . . ........................... 6.11-1 6.12 System Integrity ... .......................... 6.12-1 6.13 High Radiation Area . . ........................... 6.13-1 6.14 Deleted ............................... 6.14-1 6.15 Secondary Water Chemistry . . ........................... 6.15-1 6.16 Radiological Effluents . . ........................... 6.16-1 6.17 Process Control Program (PCP) . . ........................... 6.17-1 6.18 Offsite Dose Calculation Manual (ODCM) ............................. 6.18-1 6.19 Major Changes to Radioactive Liquid, Gaseous and Solid Waste Treatment Systems . . ..........................6.19-1 6.20 Containment Leakage Rate Testing Program . ...................................... 6.20-1 6.21 Technical Specifications (TS) Bases Control Program ............................... 6.21-1 7/8.0 Deleted TS iv Amendment No. 165

LIST OF FIGURES FIGURE TITLE 2.1-1 .. Deleted 3.1-1 . . Heatup Limitation Curves Applicable for Periods Up to 331'1 Effective Full-Power Years 3.1-2 .. Cooldown Limitation Curves Applicable for Periods Up to 33111 Effective Full-Power Years 3.1-3 . . Dose Equivalent 1-131 Reactor Coolant Specific Activity Limit Versus Percent of Rated Thermal Power 3.1-4 . . Deleted 3.10-1 . . Deleted 3.10-2 . . Deleted 3.10-3 . . Deleted 3.10-4 . . Deleted 3.10-5 .. Deleted 3.10-6 .. Deleted 4.2-1 . . Deleted 5.4-1 .. .. Minimum Required Fuel Assembly Bumup as a Function of Nominal Initial Enrichment to Permit Storage in the Tranfer Canal Note:

l Although the curves were developed for 33 EFPY, they are limited to 28 EFPY (corresponding to the end of cycle 28) by WPSC Letter NRC-99-017.

TS vi Amendment No. 165

j. MODES OOLANT , OFTEMP FISSION MODE MOEIj REACTIVITY Ak/k T.

1 POWER%

REFUELING <-5%

140 -0 COLD SHUTDOWN * -1%

200 -Q INTERMEDIATE (1) > 200 < 540 -0 SHUTDOWN _______

HOT SHUTDOWN (1) > 540 -0 HOT STANDBY < 0.25% -Toper <2 OPERATING < 0.25% -Toper 2 2 LOW POWER PHYSICS (To be specified by specific tests)

TESTING _ _ _ _ _ _ _ _ _ _ _

(1) Refer to the required SHUTDOWN MARGIN as specified in the Core Operating Limits Report.

k. REACTOR CRITICAL The reactor is said to be critical when the neutron chain reaction is self-sustaining.

I. REFUELING OPERATION REFUELING OPERATION is anyoperation involving movementof reactorvessel internal components (those that could affect the reactivity of the core) within the containmentwhen the vessel head Is unbolted or removed.

m. RATED POWER RATED POWER is the steady-state reactor core output of 1,650 MWt.

n. REPORTABLE EVENT A REPORTABLE EVENT is defined as any of those conditions specified in 10 CFR 50.73 TS 1.0-4 Amendment No. 165

p. DOSE EQUIVALENT 1-131 DOSE EQUIVALENT 1-131 is that concentration of 1-131 (p Cilgram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of 1-131,1-132,1-133, 1-134 a nd 1-135 a ctually p resent. T he t hyrold d ose conversion factors used for this calculation shall be as listed and calculated with the methodology established in Table Ill of TID-1 4844, "Calculation of Distance Factors for Power and Test Reactor Sites."

DOSE CONVERSION FACTOR ISOTOPE 1.0000 1-131 0.0361 1-132 0.2703 1-133 0.0169 1-134 0.0838 1-135

q. CORE OPERATING LIMITS REPORT (COLR) for The COLR is the unit specific document that provides cycle-specific parameter limits for the current reload cycle. These cycle specific parameter limits shall be determined each reload cycle in accordance with Specification 6.9.a.4. Plant operation within these limits is addressed in individual Specifications.

r. SHUTDOWN MARGIN (SDM)

SDM shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming:

1. All rod cluster control assemblies (RCCAs) are fully inserted except for the single RCCA of highest reactivity worth, which is assumed to be fully withdrawn. However, with all RCCAs verified fully Inserted by two independent means (TS 3.1 0.e), it is not necessary to account for a stuck RCCA in the SDM calculation. With any RCCA not capable of being fully inserted, the reactivityworth of the RCCA must be accounted for in the determination of SDM, and

2. In the OPERATING and HOT STANDBY MODES, the fuel and moderator temperatures are changed to the nominal zero power design temperature.

TS 1.0-6 Amendment No. 165

2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS - REACTOR CORE APUL ABkIfl System pressure Applies to the limiting combination of thermal power, Reactor Coolant STANDBY MODES.

and coolant temperature during the OPERATING and HOT 0-RJECTIVE To maintain the integrity of the fuel cladding.

SPFCIFICATION and coolant

a. The combination of RATED POWER level, coolant pressure, SAFETY LIMIT temperature shall not exceed the limits specified in the COLR. The Coolant System is exceeded if the point defined by the combination of Reactor appropriate pressure average temperature and power level is at any time above the line.

21.14 for the

b. The departure from nucleate boiling ratio (DNBR) shall be maintained HTP DNB correlation oF

c. The peak fuel centerline temperature shall be maintained < 4700 TS 2.1-1 Amendment No. 165

TS FIGURE 2.1-1 DELETED Amendment No. 165

3. Reactor Coolant Temperature A. Overtemperature AT <*ATo [ KiK 2(T-T') +rj 1 + K 3(PP') f (AI)]

where ATo = Indicated AT at RATED POWER, % l T = Average temperature, 0 F T= [*]OF l P - Pressurizer pressure, psig P = [*] psig K1 = [1] l K2 = [*]

K3 1*

I*]sec.

12 = [*]sec. l f(AI) = An even function of the indicated difference between top and bottom detectors of the power-range nuclear ion chambers. Selected gains are l based on measured instrument response during plant startup tests, where qt and qb are the percent power in the top and bottom halves of the core respectively, and qt + qb is total core power in percent of RATED POWER, such that:

1. Forqt-qbwithinl1*],[*]%.f(A1)=O°-

2. For each percent that the magnitude of qt - qb exceeds [*] % the AT trip setpoint shall be automatically reduced by an equivalent of [*] % l of RATED POWER. r

3. For each percent that the magnitude of qt- qb exceed -[*] % the AT trip setpoint shall be automatically reduced by an equivalent of [*J %

of RATED POWER.

Note: [*1 As specified In the COLR TS 2.3-2 Am~endment No. 165

B. Overpower

+I T-K rjS 6 (T-T')-f (AI)]

AT*5AT4o ATS Kn4 -Ks

~o[ - 5 rlS where ATo = Indicated AT at RATED POWER, % l T = Average Temperature, 'F T = [*]IF I K4 [*]M K5 2 [*] for increasing T; [*] for decreasing T l K6 2*forT>T;[*]forT<T l X = *]3sec. l f(AI) = As defined above Note: [*] As specified in the COLR

4. Reactor Coolant Flow A. Low reactor coolant flow per loop 2 90% of normal indicated flow as measured by elbow taps.

B. Reactor coolant pump motor breaker open

1. Low frequency setpoint 2 55.0 Hz

2. Low voltage setpoint 2Ž75% of normal voltage

5. Steam Generators Low-low steam generator water level 2 5% of narrow range instrument span.

TS 2.3-3 Amendment No. 165

3.1 REACTOR COOLANT SYSTEM APPLICABILITY Applies to the OPERATING status of the Reactor Coolant System (RCS).

OBJECTIVE To specify those LIMITING CONDITIONS FOR OPERATION of the Reactor Coolant System which must be met to ensure safe reactor operation.

SPECIFICATIONS

a. Operational Components

1. Reactor Coolant Pumps A. At least one reactor coolant pump or one residual heat removal pump shall be in operation when a reduction is made in the boron concentration of the reactor coolant.

B. When the reactor is in the OPERATING mode, except for low power tests, both reactor coolant pumps shall be in operation.

C. A reactor coolant pump shall not be started with one or more of the RCS cold leg temperatures s 2000F unless the secondary water temperature of each steam generator is < 1000 F above each of the RCS cold leg temperatures.

2. Decay Heat Removal Capability A. At least two of the following four heat sinks shall be OPERABLE whenever the average reactor coolant temperature is < 3500 F but > 200 0F.

1. Steam Generator 1A

2. Steam Generator lB

3. Residual Heat Removal Train A

4. Residual Heat Removal Train B If less than the above number of required heat sinks are OPERABLE, then corrective action shall be taken immediately to restore the minimum number to the OPERABLE status.

TS 3.1-1 Amendment No. 165

B. Two residual heat removal trains shall be OPERABLE whenever the average reactor coolant temperature is

200OF and irradiated fuel is in the reactor, except when in the REFUELING MODE with the minimum water level above the l top of the vessel flange 2 23 feet, one train may be inoperable for maintenance.

1. Each residual heat removal train shall be comprised of:

a) One OPERABLE residual heat removal pump l b) One OPERABLE residual heat removal heat exchanger l c) An OPERABLE flow path consisting of all valves and piping associated l with the above train of components and required to remove decay heat from the core during normal shutdown situations. This flow path shall be capable of taking suction from the appropriate Reactor Coolant System hot leg and returning to the Reactor Coolant System.

2. Ifone residual heat removal train is inoperable, then corrective action shall be taken immediately to return itto the OPERABLE status.

TS 3.1-2 Amendment No. 165

4. Pressure Isolation Valves A. All pressure isolation valves listed in Table TS 3.1-2 shall be functional as a pressure isolation device during OPERATING and HOT STANDBY MODES, except as specified In 3.1.a.4.B. Valve leakage shall not exceed the amounts indicated.

B. In the event that integrity of any pressure isolation valve as specified in Table TS 3.1-2 cannot be demonstrated, reactor operation may continue, provided that at least two valves in each high pressure line having a non-functional valve are in. and remain in, the mode corresponding to the isolated condition.()

C. IfTS 3.1 .a.4.A and TS 3.1 .a.4.B cannot be met, then an orderly shutdown shall be initiated and the reactorshall be in the HOT SHUTDOWN condition within the next 4 hours0.167 days 0.0238 weeks 0.00548 months , the INTERMEDIATE SHUTDOWN condition in the next 6 hours0.25 days 0.0357 weeks 0.00822 months and the COLD SHUTDOWN condition within the next 24 hours1 days 0.143 weeks 0.0329 months .

5. Pressurizer Power-Operated Relief Valves (PORV) and PORV Block Valves A. Two PORVs and theirassociated block valves shall be OPERABLE during HOT STANDBY and OPERATING modes.

1. With one or both PORVs inoperable because of excessive seat leakage, within one hour either restore the PORV(s) to OPERABLE status or close the associated block valve(s) with power maintained to the block valve(s);

otherwise, action shall be initiated to:

- Achieve HOT STANDBYwithin 6 hours0.25 days 0.0357 weeks 0.00822 months

- Achieve HOT SHUTDOWN within the following 6 hours0.25 days 0.0357 weeks 0.00822 months

2. With one PORV inoperable due to causes other than excessive seat leakage, within one hour either restore the PORV to OPERABLE status or close its associated block valve and remove power from the block valve.

Restore the PORV to OPERABLE status within the following 72 hours3 days 0.429 weeks 0.0986 months or action shall be initiated to:

- Achieve HOT STANDBY within 6 hours0.25 days 0.0357 weeks 0.00822 months

- Achieve HOT SHUTDOWN within the following 6 hours0.25 days 0.0357 weeks 0.00822 months

3. With both PORVs inoperable due to causes other than excessive seat leakage, within one hour either restore at least one PORV to OPERABLE status or close its associated block valve and remove power from the block valve and

- Achieve HOT STANDBYmwithin 6 hours0.25 days 0.0357 weeks 0.00822 months

- Achieve HOT SHUTDOWN within the following 6 hours0.25 days 0.0357 weeks 0.00822 months

( Manual valves shall be locked in the closed position. Motor operated valves shall be placed

in the closed position with their power breakers locked out.

TS 3.1-4 Amendment No. 165

4. With one block valve inoperable, within one hour restore the block valve to OPERABLE status or place its associated PORV in manual control. Restore the block valve to OPERABLE status within 72 hours3 days 0.429 weeks 0.0986 months ; otherwise action shall be Initiated to:

- Achieve HOT STANDBY within 6 hours0.25 days 0.0357 weeks 0.00822 months

- Achieve HOT SHUTDOWN within the following 6 hours0.25 days 0.0357 weeks 0.00822 months

5. With both block valves inoperable, within one hour restore the block valves to OPERABLE status or place their a ssociated P ORVs i n manual control.

Restore at least one block valve to OPERABLE status within the next hour; otherwise, action shall be initiated to:

- Achieve HOT STANDBY within 6 hours0.25 days 0.0357 weeks 0.00822 months

- Achieve HOT SHUTDOWN within the following 6 hours0.25 days 0.0357 weeks 0.00822 months

6. Pressurizer Heaters A. At least one group of pressurizer heaters shall have an emergency powersupply available when the average RCS temperature is > 3501F.

7. Reactor Coolant Vent System A. A reactor coolant vent path from both the reactor vessel head and pressurizer steam space shall be OPERABLE and closed prior to the average RCS temperature being heated > 2000F except as specified in TS 3.1.a.7.B a nd TS 3.1.a.7.C below.

B. When the average RCS temperature is > 200 0F, any one of the following conditions of inoperability may exist:

1. Both of the parallel vent valves in the reactorvessel vent path are inoperable.

2. Both of the parallel vent valves in the pressurizer vent path are inoperable.

If OPERABILITY is not restored within 30 days, then within one hour action shall be initiated to:

- Achieve HOT STANDBY within 6 hours0.25 days 0.0357 weeks 0.00822 months

- Achieve HOT SHUTDOWN within the following 6 hours0.25 days 0.0357 weeks 0.00822 months

- Achieve COLD SHUTDOWN within an additional 36 hours1.5 days 0.214 weeks 0.0493 months C. If no Reactor Coolant System vent paths are OPERABLE, then restore at least one vent path to OPERABLE status within 72 hours3 days 0.429 weeks 0.0986 months . If OPERABILITY is not restored within 72 hours3 days 0.429 weeks 0.0986 months , then within one hour action shall be initiated to:

- Achieve HOT STANDBY within 6 hours0.25 days 0.0357 weeks 0.00822 months

- Achieve HOT SHUTDOWN within the following 6 hours0.25 days 0.0357 weeks 0.00822 months

- Achieve COLD SHUTDOWN within an additional 36 hours1.5 days 0.214 weeks 0.0493 months TS 3.1-5 Amendment No. 165

b. Heatup and Cooldown Limit Curves for Normal Operation

1. The reactor coolant temperature and pressure and system heatup and cooldown rates (with the exception of the pressurizer) shall be limited in accordance with Figures TS 3.1-1 and TS 3.1-2. Figures TS 3.1-1 and TS 3.1-2 are applicable forthe service period of up to 33(1) effective full-power years.

A. Allowable combinations of pressure and temperature for specific temperature change rates are below and to the right of the limit lines shown. Limit lines for cooldown rates between those presented may be obtained by interpolation.

B. Figures TS 3.1-1 and TS 3.1-2 define limits to assure prevention of non-ductile failure only. For normal operation other inherent plant characteristics, e.g., pump heat addition and pressurizer heater capacity may limit the heatup and cooldown rates that can be achieved over certain pressure-temperature ranges.

C. The isothermal curve in Figure TS 3.1-2 defines limits to assure prevention of non-ductile failure applicable to low temperature overpressurization events only.

Application of this curve is limited to evaluation of LTOP events whenever one or more of the RCS cold leg temperatures are less than or equal to the LTOP enabling temperature of 2000F.

2. The secondary side of the steam generator must not be pressurized> 200 psig ifthe temperature of the steam generator is < 700 F. 1

3. The pressurizer cooldown and heatup rates shall not exceed 200OF/hr and 100°F/hr, respectively. The spray shall not be used if the temperature difference between the pressurizer and the spray fluid is > 3200F.

4. The overpressure protection system for low temperature operation shall be OPERABLE whenever one or more of the RCS cold leg temperatures are < 2000F, and the reactor vessel head is Installed. The system shall be considered OPERABLE when at least one of the following conditions is satisfied:

A. The overpressure relief valve on the Residual Heat Removal System (RHR 33-1) shall have a set pressure of c 500 psig and shall be aligned to the RCS by maintaining valves RHR 1A, 1B. 2A, and 2B open.

1. With one flow path inoperable, the valves in the parallel flow path shall be verified open with the associated motor breakers for the valves locked in the off position. Restore the inoperable flow path within five days or complete depressurization and venting of the RCS through a 2 6.4 square inch vent within an additional eight hours.

2. With both flow paths or RHR 33-1 inoperable, complete depressurization and venting of the RCS through at least a 6.4 square inch vent pathway within eight hours.

t Although the curves were developed for 33 EFPY, they are limited to 28 EFPY (corresponding to the end of cycle 28) by WPSC Letter NRC-99-017.

TS 3.1-6 Amendment No. 165

B. A vent pathway shall be provided with an effective flow cross section 2 6.4 square inches.

1. When low temperature overpressure protection is provided via a vent pathway, verifytheventpathwayatleastonce per31 days when the pathway is provided by a valve(s) that is locked, sealed, or otherwise secured in the open position. Ifthe vent path is provided by any other means, then verify the vent pathway every 12 hours0.5 days 0.0714 weeks 0.0164 months .

c. Maximum Coolant Activity

1. The specific activity of the reactor coolant shall be limited to:

A. 50.20 pCi/gram DOSE EQUIVALENT 1-131, and B. 9- - gross radioactivity due to nuclides with half-lives > 30 minutes Ecc excluding tritium (E is the average sum of the beta and gamma energies In Mev per disintegration) whenever the reactor is critical or the average coolant temperature is > 5000 F.

2. Ifthe reactor is critical or the average temperature is > 5000 F:

A. With the specific activity of the reactor coolant > 0.20 PCi/gram DOSE EQUIVALENT 1-131 for more than 48 hours2 days 0.286 weeks 0.0658 months during one continuous time interval, or exceeding the limit shown on Figure TS 3.1-3, be in at least INTERMEDIATE SHUTDOWN with an average coolant temperature of < 5000 F within six hours.

91 pCi B. With the specific activity of the reactor coolant >-9 -- i of gross radioactivity, be in at least INTERMEDIATE SHUTDOWN with an average coolant temperature < 5000 F within six hours.

C. With the specific activity of the reactor coolant > 0.20 pCi/gram DOSE EQUIVALENT 1-131 or> 9 ci perform the sample and analysis requirements E cc of Table TS 4.1-2, item 1.f, once everyfour hoursuntil restored towithin its limits.

3. Annual reporting requirements are identified in TS 6.9.a.2.D.

TS 3.1-7 Amendment No. 165

d. Leakage of Reactor Coolant

1. Any Reactor Coolant System leakage indication in excess of 1 gpm shall be the subject of an investigation and evaluation initiated within four hours of the indication.

Any indicated leak shall be considered to be a real leak until it is determined that no unsafe condition exists. If the Reactor Coolant System leakage exceeds 1 gpm and the source of leakage Is not Identified within 12 hours0.5 days 0.0714 weeks 0.0164 months , then the reactor shall be placed in the HOT SHUTDOWN condition utilizing normal operating procedures. If the source of leakage exceeds I gpm and is not identified within 48 hours2 days 0.286 weeks 0.0658 months , then the reactor shall be placed in the COLD SHUTDOWN condition utilizing normal operating procedures.

2. Reactor coolant-to-secondary leakage through the steam generator tubes shall be limited to 150 gallons per day through any one steam generator. With tube leakage greaterthan the above limit, reduce the leakage rate within four hours or be in COLD SHUTDOWN within the next 36 hours1.5 days 0.214 weeks 0.0493 months .

3. If the sources of leakage other than that in 3.1.d.2 have been identified and it is evaluated that continued operation is safe, then operation of the reactor with a total Reactor Coolant System leakage rate not exceeding 10 gpm shall be permitted. If leakage exceeds 10 gpm, then the reactor shall be placed in the HOT SHUTDOWN condition within 12 hours0.5 days 0.0714 weeks 0.0164 months utilizing normal operating procedures. If the leakage exceeds 10 gpm for 24 hours1 days 0.143 weeks 0.0329 months , the reactor shall be placed in the COLD SHUTDOWN condition utilizing normal operating procedures.

4. If any reactor coolant leakage exists through a non-isolable fault in a Reactor Coolant System component (exterior wall of the reactorvessel, piping, valve body, relief valve leaks, pressurizer, steam generator head, or pump seal leakoff), then the reactor shall be shut down; and cooldown to the COLD SHUTDOWN condition shall be initiated within 24 hours1 days 0.143 weeks 0.0329 months of detection.

5. When the reactor is critical and above 2% power, two reactor coolant leak detection systems of different operating principles shall be in operation with one of the two systems sensitive to radioactivity. Either system may be out of operation for up to 12 hours0.5 days 0.0714 weeks 0.0164 months provided at least one system is OPERABLE.

TS 3.1-8 Amendment No. 165

e. Maximum Reactor Coolant Oxygen, Chloride and Fluoride Concentration

1. Concentrations of contaminants in the reactor coolant shall not exceed the following limits when the reactor coolant temperature is > 250'F.

CONTAMINANT NORMAL STEADY-STATE TRANSIENT lA. Oxygen 0.10 1.00 B. Chloride 0.15 1.50 iC. Fluoride 0.15 1.50

2. Ifany of the normal steady-state operating limits as specified in TS 3.1 .e.1 above are exceeded, or if it is anticipated that they may be exceeded, then corrective action shall be taken immediately.

3. If the concentrations of any of the contaminants cannot be controlled within the transient limits of TS 3.1 .e.1 above or returned to the normal steady-state limitwithin 24 hours1 days 0.143 weeks 0.0329 months , then the reactor shall be brought to the COLD SHUTDOWN condition, utilizing normal operating procedures, and the cause shall be ascertained and corrected. The reactor may be restarted and operation resumed If the maximum concentration of any of the contaminants did not exceed the permitted transient values. Otherwise a safety review by the Plant Operations Review Committee shall be made before starting.

4. Concentrations of contaminants in the reactor coolant shall not exceed the following 0

maximum limits when the reactor coolant temperature is < 250 F.

NORMAL TRANSIENT CONTAMINANT CONCENTRATION (ppm) LIMITS (ppm)

B. Chloride 0.15 1.50 C. Fluoride 0.15 1.50

5. If the transient limits of TS 3.1.e.4 are exceeded or the concentrations cannot be returned to normal values within 48 hours2 days 0.286 weeks 0.0658 months , then the reactor shall be brought to the COLD SHUTDOWN condition and the cause shall be ascertained and corrected.

6. To meet TS 3.1 .e.1 and TS 3.1 .e.4 above, reactor coolant pump operation shall 0be permitted for short periods, provided the coolant temperature does not exceed 250 F.

TS 3.1-9 Amendment No. 165

f. Minimum Conditions for Criticality

1. The reactor shall not be brought to a critical condition until the pressure-temperature state Isto the right of the criticality limit line shown in Figure TS 3.1-1.

2. The reactor shall be maintained subcritical by at least 1% AkIk until normal water level is established In the pressurizer.

3. When the reactor is critical the moderator temperature coefficient shall be as specified in the COLR, except during LOW POWER PHYSICS TESTING. The maximum upper moderator temperature coefficient limit shall be S 5 pcm/oF for power levels < 60% RATED POWER and < 0 pcrmI 0 F for power levels > 60% RATED POWER.

4. Ifthe limits of 3.1 .f.3 cannot be met, then power operation may continue provided the following actions are taken:

A. Within 24 hours1 days 0.143 weeks 0.0329 months , develop and maintain administrative control rod withdrawal limits sufficient to restore the moderator temperature coefficient to within the limits specified in TS 3.1 .f.3. These withdrawal limits shall be in addition to the insertion limits specified in TS 3.10.d.

B. If the actions specified in TS 3.1.f.4.A are not satisfied, then be in HOT STANDBY within the next 6 hours0.25 days 0.0357 weeks 0.00822 months .

TS 3.1-1 0 Amendmeht No. 165

3.8 REFUELING OPERATIONS APPLICABILITY Applies to operating limitations during REFUELING OPERATIONS.

OBJECTIVE To ensure that no incident occurs during REFUELING OPERATIONS that would affect public health and safety.

SPECIFICATION

a. During REFUELING OPERATIONS:

1. Containment Closure

a. The equipment hatch shall be closed and at least one door in each personnel air lock shall be capable of being closed ( in 30 minutes or less. In addition, at least one door in each personnel air lock shall be closed when the reactor vessel head or upper internals are lifted.

b. Each line that penetrates containment and which provides a direct air path from containment atmosphere to the outside atmosphere shall have a closed isolation valve or an operable automatic isolation valve.

2. Radiation levels in fuel handling areas, the containment and the spent fuel storage pool shall be monitored continuously.

3. The reactor will be subcritical for 148 hours6.167 days 0.881 weeks 0.203 months prior to movement of its Irradiated fuel assemblies. Core subcritical neutron flux shall be continuously monitored by at least two neutron monitors, each with continuous visual indication in the control room and one with audible indication in the containment whenever core geometry is being changed. When core geometry is not being changed at least one neutron flux monitor shall be in service.

4. At least one residual heat removal pump shall be OPERABLE.

5. When there is fuel in the reactor, a minimum boron concentration as specified Inthe COLRshall be maintained intheReactorCoolantSystemduring reactorvesselhead [

removal or while loading and unloading fuel from the reactor. The required boron concentration shall be verified by chemical analysis daily.

(' Administrative controls ensure that:

Appropriate personnel are aware that both personnel air lock doors are open,

A specified individual(s) is designated and available to close the air lock following a required evacuation of containment, and

Any obstruction(s) (e.g., cables and hoses) that could prevent closure of an open air lock can be quickly removed.

TS 3.8-1 Amendment No. 165

6. Direct communication between the control room and the operating floor of the containment shall be available whenever changes in core geometry are taking place.

7. Heavy loads, greater than the weight of a fuel assembly, will not be transported over or placed in either spent fuel pool when spent fuel is stored in that pool. Placement of additional fuel storage racks is permitted, however, these racks may not traverse directly above spent fuel stored in the pools.

8. The containment ventilation and purge system, including the capability to initiate automatic containment ventilation isolation, shall be tested and verified to be operable immediately prior to and daily during REFUELING OPERATIONS.

9. a. The spent fuel pool sweep system, including the charcoal adsorbers, shall be operating during fuel handling and when any load is carried over the pool if irradiated fuel in the pool has decayed less than 30 days. If the spent fuel pool sweep system, including the charcoal adsorber, Is not operating when required, fuel movement shall not be started (any fuel assembly movement in progress may be completed).

b. Performance Requirements

1. The results of the in-place cold DOP and halogenated hydrocarbon tests at design flows on HEPA filters and charcoal adsorber banks shall show

>99% DOP removal and >99% halogenated hydrocarbon removal.

2. The results of laboratory carbon sample analysis from spent fuel pool sweep system carbon shall show >95% radioactive methyl iodide removal0 when tested in accordance with ASTM D3803-89 at conditions of 30 C and 95% RH.

3. Fans shall operate within +10% of design flow when tested.

10. The minimum water level above the vessel flange shall be maintained at 23 feet.

11. A dead-load test shall be successfully performed on both the fuel handling and manipulator cranes before fuel movement begins. The load assumed by the cranes for this test must be equal to or greater than the maximum load to be assumed by the cranes during the REFUELING OPERATIONS. A thorough visual inspection of the cranes shall be made after the dead-load test and prior to fuel handling.

12. A licensed senior reactor operator will be on-site and designated in charge of the REFUELING OPERATIONS.

b. If any of the specified limiting conditions for REFUELING OPERATIONS are not met, refueling of the reactorshall cease. W ork shall be initiated to correct the violated conditions so that the specified limits are met, and no operations which may increase the reactivity of the core shall be performed.

TS 3.8-2

Amendmejt No.' 165

3.10 CONTROL ROD AND POWER DISTRIBUTION LIMITS APPLICABILITY Applies to the limits on core fission power distributions and to the limits on control rod operations.

OBJECTIVE To ensure: 1) core subcriticality after reactor trip, 2) acceptable core power distribution during power operation in order to maintain fuel integrity in normal operation transients associated with faults of moderate frequency, supplemented by automatic protection and by administrative procedures, and to maintain the design basis initial conditions for limiting faults, and 3) limited potential reactivity insertions caused by hypothetical control rod ejection.

SPECIFICATION

a. Shutdown Reactivity When the reactor is subcritical prior to reactor startup, the SHUTDOWN MARGIN shall be at least that as specified in the COLR

b. Power Distribution Limits

1. At all imes, except during Low Power Physics Tests, the hot channel factors defined in the basis must meet the following limits:

A. FoN(Z) Limits shall be as specified in the COLR.

B. F&HN Limits shall be as specified in the COLR.

2. If. for any measured hot channel factor, the relationships of FoN(Z) and FMN specified in the COLR are not true, then reactor power shall be reduced by a fractional amountof the design power to a value for which the relationships are true, and the high neutron flux trip setpoint shall be reduced by the same fractional amount. If subsequent incore mapping cannot, within a 24-hour period, demonstrate that the hot channel factors are met, then the overpower AT and overtemperature AT trip setpoints shall be similarly reduced.

3. Following initial loading and at regular effective full-power monthly intervals thereafter, power distribution maps using the movable detection system shall be made to confirm that the hot channel factor limits of TS 3.1 O.b.1 are satisfied.

4. The measured FoEO(Z) hot channel factors under equilibrium conditions shall satisfy the relationship for the central axial 80% of the core as specified in the COLR.

TS 3.10-1 Amendment No. 165

5. Power distribution maps using the movable detector system shall be made to confirm the relationship of FoEQ(Z) specified in the COLR according to the following schedules with allowances for a 25% grace period:

A. During the target flux difference determination or once per effective full-power monthly interval, whichever occurs first.

B. Upon achieving equilibrium conditions after reaching a thermal power level > 10%

higher than the power level at which the last power distribution measurement was performed in accordance with TS 3.10.b.5.A.

C. If a power distribution map indicates an increase in peak pin power, FAHN, of 2% or more, due to exposure, when compared to the last power distribution map, then either of the following actions shall be taken:

i. FaM(Z) shall be increased by the penalty factor specified in the COLR for comparison to the relationship of F0 EC(Z) specified in the COLR, or Hi. FoQE(Z) shall be measured by power distribution maps using the incore movable detector system at least once every seven effective full-power days until a power distribution map indicates that the peak pin power, FAHN, is not increasing with exposure when compared to the last power distribution map.

6. If, for a measured F EQ, the relationships of F0 EO(Z) specified in the COLR are not satisfied and the relationships of FQN(Z) and F N specified in the COLR are satisfied, then within 12 hours0.5 days 0.0714 weeks 0.0164 months take one of the following actions:

A. Take corrective actions to improve the power distribution and upon achieving equilibrium conditions measure the target flux difference and verify that the relationships of FoEO(Z) specified in the COLR are satisfied, or B. Reduce reactor power and the high neutron flux trip setpoint by 1% for each percent that the left hand sides of the relationships of F0EQ(Z) specified in the COLR exceed the limits specified in the right hand sides. Reactor power may subsequently be increased provided that a power distribution map verifies that the relationships of F0EQ(Z) specified in the COLR are satisfied with at least 1% of margin for each percent of power level to be increased.

7. The reference equilibrium indicated axial flux difference as a function of power level (called the target flux difference) shall be measured at least once per full-power month.

8. The indicated axial flux difference shall be considered outside of the limits of TS 3.1 0.b.9 through TS 3.10.b.12 when more than one of the OPERABLE excore channels are indicating the axial flux difference to be outside a limit.

9. Except during physics tests, during excore detector calibration and except as modified by TS 3.10.b.10 through TS 3.10.b.12, the indicated axial flux difference shall be maintained within the target band about the target flux difference as specified in the COLR.

TS 3.10-2 Amendment No. 165

10. At a power level> 90% of rated power, if the indicated axial flux difference deviates from its target band, the flux difference shall be returned to the target band within 15 minutes or reactor power shall be reduced to a level no greater than 90% of rated power.

11. At power levels > 50% and

90% of rated power A. The indicated axial flux difference may deviate from the target band, specified in the COLR, for a maximum of one hour (cumulative) in any 24-hour period provided the flux difference does not exceed an envelope specified in the COLR. If the cumulative time exceeds one hour, then the reactor power shall be reduced to

50% of rated thermal power within 30 minutes and the high neutron flux setpoint reduced to < 55% of rated power.

If the indicated axial flux difference exceeds the outer envelope specified in the COLR, then the reactor power shall be reduced to *50% of rated thermal power within 30 minutes and the high neutron flux setpoint reduced to <55% of rated power.

B. A power increase to a level > 90% of rated power is contingent upon the indicated axial flux difference being within its target band.

12. At a power level no greater than 50% of rated power A. The indicated axial flux difference may deviate from its target band.

B. A power increase to a level > 50% of rated power is contingent upon the indicated axial flux difference not being outside its target band for more than two hours (cumulative) of the preceding 24-hour period.

One half of the time the indicated axial flux difference is out of its target band, up to 50% of rated power is to be counted as contributing to the one hour cumulative maximum the flux difference may deviate from its target band at a power level < 90%

of rated power.

13. Alarms shall normally be used to indicate nonconformance with the flux difference requirement of TS 3.1 0.b.1 0 orthe flux difference time requirement of TS 3.1 0.b.11 .A. If the alarms are temporarily out of service, then the axial flux difference shall be logged, and conformance with the limits assessed, every hour for the first 24 hours1 days 0.143 weeks 0.0329 months , and half-hourly thereafter.

TS 3.10-3 Amendment No. 165

c. Quadrant Power Tilt Limits

1. Except for physics tests, whenever the indicated quadrant power tilt ratio> 1.02, one of the following actions shall be taken within two hours: l A. Eliminate the tilt.

B. Restrict maximum core power level 2% for every 1% of indicated power tilt ratio

> 1.0.

2. Ifthe tilt condition is not eliminated after 24 hours1 days 0.143 weeks 0.0329 months , then reduce power to 50% or lower. l

3. Except for Low Power Physics Tests, if the indicated quadrant tilt is > 1.09 and there is simultaneous indication of a misaligned rod:

A. Restrict maximum core power level by 2% of rated values for every 1% of indicated power tilt ratio > 1.0.

B. Ifthe tilt condition is not eliminated within 12 hours0.5 days 0.0714 weeks 0.0164 months , then the reactor shall be brought l to a minimum load condition (530 Mwe).

4. If the indicated quadrant tilt is > 1.09 and there is no simultaneous indication of rod misalignment, then the reactor shall immediately be brought to a no load condition

(<5% reactor power).

d. Rod Insertion Limits

1. The shutdown rods shall be withdrawn to within the limits, specified in the COLR, when l the reactor is critical or approaching criticality.

2. The control banks shall be limited in physical insertion; insertion limits are specified in the COLR. Ifany one of the control bank insertion limits is not met:

A. Within one hour, initiate boration to restore control bank insertion to within the limits l specified in the COLR, and I B. Restore control bank insertion to within the limits specified in the COLR within l two hours of exceeding the insertion limits. I C. If any one of the conditions of TS 3.1 0.d.2.A or TS 3.1 0.d.2.B cannot be met, then within one hour action shall be initiated to: l

- Achieve HOT STANDBY within 6 hours0.25 days 0.0357 weeks 0.00822 months

- Achieve HOT SHUTDOWN within the following 6 hours0.25 days 0.0357 weeks 0.00822 months

3. Insertion limit does not apply during physics tests or during periodic exercise of individual rods. However, the shutdown margin, as specified in the COLR, must be maintained I except for the Low Power Physics Test to measure control rod worth and shutdown margin. For this test, the reactor may be critical with all but one high worth rod inserted.

TS 3.10-4 Amendment No. 165

e. Rod Misalignment Limitations This specification defines allowable limits for misaligned rod cluster control assemblies. In TS 3.1 O.e.1 and TS 3.1 O.e.2, the magnitude, in steps, of an indicated rod misalignment may be determined by comparison of the respective bank demand step counter to the analog individual rod position indicator, the rod position as noted on the plant process computer, or through the conditioning module output voltage via a correlation of rod position vs. voltage.

Rod misalignment limitations do not apply during physics testing.

1. When reactor power Is >85% of rating, the rod cluster control a ssembly s hall be maintained within i 12 steps from their respective banks. If a rod cluster control assembly is misaligned from its bank by more than + 12 steps when reactor power is

Ž85%, then the rod will be realigned or the core power peaking factors shall be determined within four hours, and TS 3.1 0.b applied. If peaking factors are not determined within four hours, the reactor power shall be reduced to < 85% of rating.

2. When reactor power is < 85% but Ž 50% of rating, the rod cluster control assemblies shall be maintained within i 24 steps from their respective banks. If a rod clustercontrol assembly is misaligned from its bank by more than + 24 steps when reactor power is

< 85% but > 50%, the rod will be realigned or the core power peaking factors shall be determined within four hours, and TS 3.10.b applied. If the peaking factors are not determined within four hours, the reactor power shall be reduced to < 50% of rating.

3. And, in addition to TS 3.10.e.1 and TS 3.10.e.2, if the misaligned rod cluster control assembly is not realigned within eight hours, the rod shall be declared inoperable.

f. Inoperable Rod Position Indicator Channels

1. If a rod position indicator channel is out of service, then:

A. For operation between 50% and 100% of rating, the position of the rod cluster control shall be checked indirectly by core instrumentation (excore detector and/or thermocouples and/or movable incore detectors) at least once per eight hours, or subsequent to rod motion exceeding a total displacement of 24 steps, whichever occurs first.

B. During operation < 50% of rating, no special monitoring is required.

2. Notmorethan one rod position indicator channel pergroup nortwo rod position indicator channels per bank shall be permitted to be inoperable at any time.

3. If a rod cluster control assembly having a rod position indicator channel out of service is found to be misaligned from TS 3.10.f.1.A, then TS 3.1 O.e will be applied.

TS 3.10-5 Amendment No. 165

9. Inoperable Rod Limitations

1. An inoperable rod is a rod which does not trip or which is declared inoperable under TS 3.10.e orTS 3.10.h.

2. Not more than one inoperable full length rod shall be allowed at any time.

3. If reactor operation is continued with one inoperable full length rod, the potential ejected rod worth and associated transient power distribution peaking factors shall be determined by analysis within 30 days unless the rod is made OPERABLE earlier. The analysis shall include due allowance for nonuniforrn fuel depletion in the neighborhood of the inoperable rod. If the analysis results in a more limiting hypothetical transient than the cases reported in the safety analysis, the plant power level shall be reduced to an analytically determined part power level which is consistent with the safety analysis.

h. Rod Drop Time At OPERATING temperature and full flow, the drop time of each full length rod cluster control shall be no greater than 1.8 seconds from loss of stationary gripper coil voltage to dashpot entry. If drop time is > 1.8 seconds, the rod shall be declared inoperable.

i. Rod Position Deviation Monitor If the rod position deviation monitor is inoperable, Individual rod positions shall be logged at least once per eight hours after a load change > 10% of rated power or after> 24 steps of control rod motion.

j. Quadrant Power Tilt Monitor If one or both of the quadrant power tilt monitors is inoperable, individual upper and lower excore detector calibrated outputs and the quadrant tilt shall be logged once per shift and after a load change > 10% of rated power or after > 24 steps of control rod motion. The monitors shall be set to alarm at 2% tilt ratio.

k. Core Average Temperature During steady-state power operation, Ta.,e shall be maintained within the limits specified in the COLR, except as provided by TS 3.1 0.n.

I. Reactor Coolant System Pressure During steady-state power operation, Reactor Coolant System pressure shall be maintained within the limits specified in the COLR, except as provided by TS 3.1 0.n.

TS3.10-6 Amendment No. 165

m. Reactor Coolant Flow

1. During steady-state power operation, reactor coolant flow rate shall be 2 93,000 gallons per minute average per loop and greater than or equal to the limit specified inthe COLR.

Ifreactor coolant flow rate is not within the limits as specified in the COLR, action shallI be taken in accordance with TS 3.10.n.

2. Compliance with this flow requirement shall be demonstrated by verifying the reactor coolant flow during initial power escalation following each REFUELING, between 70%

and 95% power with plant parameters as constant as practical.

n. DNBR Parameters if, during poweroperation anyof the conditions of TS 3.10.k, TS 3.10.1, orTS 3.10.m.1 are not met, restore the parameter Intwo hours or less to within limits or reduce power to

< 5% of thermal rated powerwithin an additional six hours. Following analysis, thermal power may be raised not to exceed a power level analyzed to maintain a DNBR greater than the minimum DNBR limit.

TS 3.10-7 Amendment No. 165

TS FIGURE 3.10-1 DELETED Amendment No. 1-65

TS FIGURE 3.10-2 DELETED Amendment No. 165

TS FIGURE 3.10-3 DELETED Amendment No. 165

TS FIGURE 3.10-4 DELETED Amendment No. 165

TS FIGURE 3.10-5 DELETED Amendment No. 165

TS FIGURE 3.10-6 DELETED Amendment No. 165

3. Monthly OPERATING Report Routine reports of OPERATING statistics and shutdown experience shall be submitted on a monthly basis to the Document Control Desk, U.S. Nuclear Regulatory Commission, Washington, D.C., 20555, with a copy to the appropriate Regional Office, to be submitted by the fifteenth of each month following the calendar month covered by the report.

4. Core Operating Limits Report (COLR)

A. Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:

(1) TS 2.1 Reactor Core Safety Limit (2) TS 2.3.a.3.A Overtemperature AT Setpoint (3) TS 2.3.a.3.B Overpower AT Setpoint (4) TS 3.1.f.3 Moderator Temperature Coefficient (MTC)

(5) TS 3.8.a.5 Refueling Boron Concentration (6) TS 3.10.a Shutdown Margin (7) TS 3.10.b.1.A FoNZ) Limits (8) TS 3.10.b.1.B F^H Limits (9) TS 3.10.b.4 FoEO(Z) Limits (10) TS 3.10.b.5.C.i FoEO(Z) penalty (11) TS 3.10.b.9 Axial Flux Difference Target Band (12) TS 3.10.b.11.A Axial Flux Difference Envelope (13) TS 3.10.d.1 Shutdown Bank Insertion Limits (14) TS 3.10.d.2 Control Bank Insertion Limits (15) TS 3.10.k Core Average Temperature (16) TS 3.10.1 Reactor Coolant System Pressure (17) TS 3.10.m.1 Reactor Coolant Flow 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) Safety Evaluation by the Office of Nuclear Reactor Regulation on "Qualifications of Reactor Physics Methods For Application To Kewaunee" Report, dated August 21, 1979, report date September 29, 1978 (2) Kewaunee Nuclear Power Plant - Review For Kewaunee Reload Safety Evaluation Methods Topical Report WPSRSEM-NP, Revision 3 (TAC No. MB0306) dated September 10, 2001.

TS 6.9-3 Amendment No. 165

(3) Nissley, M.E. et, al., "Westinghouse Large-Break LOCA Best-Estimate Methodology," WCAP-10924-P-A, Volume 1, Revision 1, Addendum 4, March 1991, Volume 1: Model Description and Validation; Addendum 4: Model Revisions.

(4) N. Lee et al., "Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code," WCAP-10054-P-A (Proprietary) and WCAP-10081-NP-A (Non-Proprietary), dated August 1985.

(5) C.M. Thompson, et al., "Addendum to the Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code: Safety Injection into the Broken Loop and COSI Condesation Model,"

WCAP-10054-P-A, Addendum 2, Revision I (Proprietary) and WCAP-10081-NP (Non-Proprietary), dated July 1997.

(6) XN-NF-82-06 (P)(A) Revision 1 and Supplements 2, 4, and 5, "Qualification of Exxon Nuclear Fuel for Extended Burnup, Exxon Nuclear Company, dated October 1986.

(7) ANF-88-133 (P)(A) and Supplement 1, "Qualification of Advanced Nuclear Fuels' PWR Design Meghodology for Rod Burnups of 62 GWd/MTU," Advanced Nuclear Fuels Corporation, dated December 1991.

(8) EMF-92-116 (P)(A) Revision 0, "Generic Mechanical Design Criteria for PWR Fuel Designs," Siemens Power Corporation, dated February 1999.

(9) XN-NF-77-57, Exxon Nuclear Power Distribution Control for Pressurized Water Reactors, Phase II, dated January 1978, and Supplement 2, dated October 1981.

(10) WCAP-8745-P-A, 'Design Basis for the Thermal Overpower Delta-T and Thermal Overtemperature Delta-T Trip Function",

dated September 1986.

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 upon issuance for each reload cycle to the NRC.

TS 6.94

Amendment .No. 165

b. Unique Reporting Requirements

1. Annual Radiological Environmental Monitoring Report A. Routine Radiological Environmental Monitoring Reports covering the operation of the unit during the previous calendar year shall be submitted prior to May 1 of each year. The report shall include summaries, interpretations, and analysis of trends of the results of the Radiological Environmental Monitoring Program for the reporting period. The material provided shall be consistent with the OFF-SITE DOSE CALCULATION MANUAL (ODCM) and Sections IV.B.2, IV.B.3, and IV.C of Appendix I to 10 CFR Part 50.

2. Radioactive Effluent Release Report Routine Radioactive Effluent Release Reports covering the operation of the unit for the previous calendar year shall be submitted by May I of each year. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit. The material provided shall be consistent with the objectives outlined in the ODCM and the PCP, and in conformance with 10 CFR 50.36a and Section IV.B.1 of Appendix I to 10 CFR Part 50.

3. Special Reports A. Special reports may be required covering inspections, test and maintenance activities. These special reports are determined on an individual basis for each unit and their preparation and submittal are designated in the Technical Specifications.

(1) Special reports shall be submitted to the Director of the NRC Regional Office listed in Appendix D. 10 CFR Part 20, with a copy to the Director, Office of Inspection and Enforcement, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555 within the time period specified for each report.

TS 6.9-5 Amendment NO. 165

BASIS - Safety Limits-Reactor Core (TS 2.1) I To maintain the integrity of the fuel cladding and prevent fission product release, it is necessary to prevent overheating of the cladding under all OPERATING conditions. This is accomplished by operating the hot regions of the core within the nucleate boiling regime of heat transfer, wherein the heat transfer coefficient is very large and the clad surface temperature is only a few degrees Fahrenheit above the coolant saturation temperature. The upper boundary of the nucleate boiling regime is termed departure from nucleate boiling (DNB) and at this point there is a sharp reduction of the heat transfer coefficient, which would result in high clad temperatures and the possibility of clad failure. DNB is not, however, an observable parameter during reactor operation. Therefore, the observable parameters of RATED POWER, reactor coolant temperature and pressure have been related to DNB through a DNB correlation. The DNB correlation has been developed to predict the DNB heat flux and the location of the DNB for axially uniform and non-uniform heat flux distributions. The local DNB ratio (DNBR), defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux, is indicative of the margin to DNB. The minimum value of the DNBR, during steady-state operation, normal operational transients, and anticipated transients is limited to the DNBR limit. This minimum DNBR corresponds to a 95%

probability at a 95% confidence level that DNB will not occur and is chosen as an appropriate margin to DNB for all OPERATING conditions.

The SAFETY LIMIT curves as provided in the Core Operating Report Limits Report which show the allowable power level decreasing with increasing temperature at selected pressures for constant flow (two loop operation) represent the loci of points of thermal power, coolant system average temperature, and coolant system pressure for which either the DNBR is equal to the DNBR limit or the average enthalpy at the exit of the core is equal to the saturation value. At low pressures or high temperatures the average enthalpy at the exit of the core reaches saturation before the DNBR ratio reaches the DNBR limit and thus, this limit is conservative with respect to maintaining clad integrity.

The area where clad integrity is ensured is below these lines.

The curves are based on the nuclear hot channel factor limits of as specified in the COLR.

These limiting hot channel factors are higher than those calculated at full power for the range from all control rods fully withdrawn to maximum allowable control rod insertion. The control rod insertion limits are given in TS 3.1 O.d. Slightly higher hot channel factors could occur at lower power levels because additional control rods are in the core. However, the control rod insertion limits as specified in the COLR ensure that the DNBR is always greater at partial power than at full power.

The Reactor Control and PROTECTION SYSTEM is designed to prevent any anticipated combination of transient conditions that would result in a DNBR less than the DNBR limit.

Two departure from nucleate boiling ratio (DNBR) correlations used in the safety analyses: the high thermal performance (HTP) DNBR correlation and the W-3 DNBR correlation. The HTP correlation applies to FRA-ANP fuel with HTP spacers. The W-3 correlation is used for the analysis of non-HTP fuel designs and for all fuel designs at low pressure and temperature conditions (e.g., the conditions analyzed during a main steam line break accident). Both DNBR correlations have been qualified and approved for application to Kewaunee.

TS B2.1-1 Amendment No. 165

RAISIS - L imiting Safety Sysipm Spftingg - ProtartiVia Instrtimnntntion (IS 2-3)

Nuclear Flux The source range high flux reactor trip prevents a startup accident from subcritical conditions from proceeding into the power range. Any setpoint within its range would prevent an excursion from proceeding to the point at which significant thermal power is generated.

The power range reactor trip Iow setpoint provides protection in the power range for a power excursion beginning from low power. This trip was used in the safety analysis.0)

The power range reactor trip high setpoint protects the reactor core against reactivity excursions which are too rapid to be protected by temperature and pressure protective circuitry. The prescribed setpoint, with allowance for errors, is consistent with the trip point assumed in the accident analysis.'2)

Two sustained rate protective trip functions have been incorporated in the Reactor PROTECTION SYSTEM. The positive sustained rate trip provides protection against hypothetical rod ejection accident. T he negative sustained rate trip provides protection for the core (low DNBR) in the event two or more rod control cluster assemblies (RCCAs) fall into the core. The circuits are independent and ensure immediate reactor trip independent of the initial OPERATING state of the reactor. These trip functions are the LIMITING SAFETY SYSTEM actions employed in the accident analysis.

PresudIner The high and low pressure trips limit the pressure range in which reactor operation is permitted.

The high pressurizer pressure trip setting is lower than the set pressure forthe safety valves (2485 psig) such that the reactor Is tripped before the safety valves actuate. The low pressurizer pressure trip causes a reactor trip in the unlikely event of a loss-of-coolant accident.° The high pressurizer water level trip protects the pressurizer safety valves against water relief. The specified setpoint allows margin for instrument error (2) and transient level overshoot before the reactor trips.

Rearctor Conlnnt Thmperatt Ire The overtemperature AT reactor trip provides core protection against DNB for all combinations of pressure, power, coolant temperature, and axial power distribution, provided only that: 1) the transient is slow with respect to piping transit'delays from the core to the temperature detectors (about 2 seconds), and 2) pressure is within the range between the high and low pressure reactor trips. With normal axial power distribution, the reactor trip limit, with allowance fore rrors(2 is always below the core SAFETY LIMITS shown in the Core Operating Limits Report. Ifaxial peaks are greater than design, as indicated by differences between top and bottom power range nuclear detectors, the reactor trip limit is automatically reduced.

(")USAR Section 14.1.1

{2) USAR Section 14.0 (3) USAR Section 14.3.1 TS B2.3-1 Amendment NO. 165

BASIS - Reactor Coolant System (TS 3.1.a) I Reactor Coolant Pumps (TS 3.1.a.1)

When the boron concentration of the Reactor Coolant System is to be reduced, the process must be uniform to prevent sudden reactivity changes in the reactor. Mixing of the reactor coolant will be sufficient to maintain a uniform boron concentration if at least one reactor coolant pump or one residual heat removal pump is running while the change is taking place. The residual heat removal pump will circulate the equivalent of the primary system volume in approximately one-half hour.

Part one of the specification requires that both reactor coolant pumps be OPERATING when the reactor is in power operation to provide core cooling. Planned power operation with one loop out-of-service is not allowed in the present design because the system does not meet the single failure (locked rotor) criteria requirement for this MODE of operation. The flow provided in each case in part one will keep Departure from Nucleate Boiling Ratio (DNBR) well above 1.30.

Therefore, cladding damage and release of fission products to the reactor coolant will not occur.

One pump operation is not permitted except for tests. Upon loss of one pump below 10% full power, the core power shall be reduced to a level below the maximum power determined for zero power testing. Natural circulation can remove decay heat up to 10% power. Above 10% power, an automatic reactor trip will occur if flow from either pump is lost. 1 )

The RCS will be protected against exceeding the design basis of the Low Temperature Overpressure Protection (LTOP) System by restricting the starting of a Reactor Coolant Pump (RXCP) to when the secondary watertemperature of each SG is < I 00OF above each RCS cold leg temperature. The restriction on starting a reactor coolant pump (RXCP) when one or more RCS cold leg temperatures is s 2001F is provided to prevent a RCS pressure transient, caused by an energy addition from the secondary system, which could exceed the design basis of the LTOP System.

Decay Heat Removal Capabilities (TS 3.1.a.2)

When the average reactor coolant temperature is

3500 F a combination of the available heat sinks is sufficient to remove the decay heat and provide the necessary redundancy to meet the single failure criterion.

When the average reactor coolant temperature is

2000 F, the plant is in a COLD SHUTDOWN condition and there Is a negligible amount of sensible heat energy stored in the Reactor Coolant System. Should one residual heat removal train become inoperable under these conditions, the remaining train is capable of removing all of the decay heat being generated.

') USAR Section 7.2.2 TS B3.1-1 Amendment No. 165

The requirement forat least one train of residual heat removal when in the REFUELING MODE is to ensure sufficient cooling capacity is available to remove decay heat and maintain the water in the reactor vessel < 1400 F. The requirement to have two trains of residual heat removal OPERABLE when there is < 23 feet of water above the reactorvessel flange ensures that a single failure will not result in complete loss-of-heat removal capabilities. With the reactor vessel head removed and at least23feetofwaterabove the vessel flange, a large heatsinkis available. In the eventofa failure of the OPERABLE train, additional time is available to initiate alternate core cooling procedures.

Pressurizer Safety Valves (TS 3.1.a.3)

Each of the pressurizersafetyvalves is designed to relieve 325,000 lbs perhourofsaturated steam at its setpoint. Below 3500 F and 350 psig, the Residual Heat Removal System can remove decay heat and thereby control system temperature and pressure. Ifno residual heat were removed by any of the means available, then the amount of steam which could be generated at safety valve relief pressure would be less than half the valves' capacity. One valve therefore provides adequate protection against overpressurization.

Pressure Isolation Valves (TS 3.1 .a.41 The basis for the pressure isolation valves is discussed in the Reactor Safety Study (RSS),

WASH-1400, and identifies an intersystem loss-of-coolant accident in a PWR which is a significant contributorto riskfromcore melt accidents (EVENTV). The designexamined inthe RSScontained two in-series check valves isolating the high pressure Primary Coolant System from the Low Pressure Injection System (LPIS) piping. The scenario which leads to the EVENT V accident is initiated by the failure of these check valves to function as a pressure Isolation barrier. This causes an overpressurization and rupture of the LPIS low pressure piping which results in a LOCA that bypasses containment.(2)

PORVs and PORV Block Valves (TS 3.1 .a.5)

The pressurizer power-operated relief valves (PORVs) operate as part of the Pressurizer Pressure Control System. They are intended to relieve RCS pressure below the setting of the code safety valves. These relief valves have remotely operated block valves to provide a positive shutoff capability should a PORV become inoperable.

The pressurizer PORVs and associated block valves must be OPERABLE to provide an alternate means of mitigating a design basis steam generator tube rupture. Thus, an inoperable PORV (for reasons other than seat leakage) or block valve is not permitted in the HOT STANDBY and OPERATING MODES for periods of more than 72 hours3 days 0.429 weeks 0.0986 months .

(2) Order for Modification of License dated 4/20181 TS B3.1-2 Amendment No. 165.

Pressurizer Heaters (TS 3.1.a.6)

Pressurizer heaters are vital elements in the operation of the pressurizer which is necessary to maintain system pressure. Loss of energy to the heaters would result in the inability to maintain system pressure via heat addition to the pressurizer. Hot functional testst 3) have indicated that one group of heaters is required to overcome ambient heat losses. Placing heaters necessary to overcome ambient heat losses on emergency power will ensure the ability to maintain pressurizer pressure. Surveillance tests are performed to ensure heater OPERABILITY. I Reactor Coolant Vent System (TS 3.1.a.7)

The function of the High Point Vent System is to vent noncondensible gases from the high points of the RCS to ensure that core cooling during natural circulation will not be inhibited. The OPERABILITY of at least one vent path from both the reactor vessel head and pressurizer steam space ensures the capability exists to perform this function.

The vent path from the reactor vessel head and the vent path from the pressurizer each contain two independently emergency powered, energize to open, valves in parallel and connect to a common header that discharges either to the containment atmosphere or to the pressurizer relief tank. The lines to the containment atmosphere and pressurizer relief tank each contain an independently emergency powered, energize to open, isolation valve. This redundancy provides protection from the failure of a single vent path valve rendering an entire vent path inoperable.

A flow restriction orifice in each vent path limits the flow from an Inadvertent actuation of the vent system to less than the flow capacity of one charging pump.(4)

° Hot functional test (PT-RC-31)

(4) Letter from E. R. Mathews to S. A. Varga dated 5121/82 TS B3.1-3 Amendment No. 165

From equation (3.1 b-1) the variables that affect the heatup and cooldown analysis can be readily identified. Kim is the stress intensity factor due to membrane (pressure) stress. Kit is the thermal (bending) stress Intensityfactor and accounts forthe linearly varying stress in the vessel wall due to thermal gradients. During heatup Kit is negative on the inside and positive on the outer surface of the vessel wall. The signs are reversed for cooldown and, therefore, an ID or an OD one quarter thickness surface flaw is postulated in whichever location is more limiting. KIR Is dependent on irradiation and temperature and, therefore, the fluence profile through the reactor vessel wall and the rates of heatup and cooldown are important. The heatup and cooldown limit curves have been developed by combining the most conservative pressure temperature limits derived by using material properties of the intermediate forging, closure head flange, and beltline circumferential weld to form a single set of composite curves. Details of the procedure used to account for these variables are explained in the following text.

Following the generation of pressure-temperature curves for both the steady-state (zero rate of change of temperature) and finite heatup rate situations, the final limit curves are produced in the following fashion. First, a composite curve is constructed based on a point-by-point comparison of the steady-state and finite heatup rate data for each of the limiting materials. At any given temperature, the allowable pressure is taken to be the lesser of the values taken from the curves under consideration. The composite curve is then adjusted to allow for possible errors in the pressure and temperature sensing Instruments Including the pressure difference between the gage and beltline weld.

The use of the composite curve is mandatory in setting heatup limitations because it is possible for conditions to exist such that over the course of the heatup ramp the controlling analysis switches from the OD to the ID location. The pressure limit must, at all times, be based on the most conservative case.

The cooldown analysis proceeds in the same fashion as that for heatup with the exception that the controlling location is always at the ID. The thermal gradients induced during cooldown tend to produce tensile stresses at the ID location and compressive stresses at the OD position. Thus, the IDflaw is clearly the worst case.

As in the case of heatup, allowable pressure-temperature relations are generated for both steady-state and finite cooldown rate situations for each of the limiting materials. Composite limit curves are then constructed for each cooldown rate of interest. Again, adjustments are made to l account for pressure and temperature instrumentation error.

The use of the composite curve in the cooldown analysis is necessary because system control is based on a measurement of reactor coolant temperature, whereas the limiting pressure iscalculated using the material temperature at the tip of the assumed reference flaw. During cooldown, the 1/4T vessel location is at a higher temperature than the fluid adjacent to the vessel ID. This condition, of course, is not true for the steady-state situation. It follows that the AT induced during cooldown results in a calculated higher KIR for finite cooldown rates than for steady-state under certain conditions.

TS B3.1-5 Amendment No. 165

Pressurizer Limits (TS 3.1.b.3)

Although the pressurizer operates at temperature ranges above those forwhich there is reason for concern about brittle fracture, OPERATING limits are provided to ensure compatibility of operation with the fatigue analysis performed in accordance with Code requirements. In-plant testing and calculations have shown that a pressurizer heatup rate of I000F/hr cannot be achieved with the installed equipment.

Low Temperature Overpressure Protection (TS 3.1.b.4)

The Low Temperature Overpressure Protection System must be OPERABLE during startup and shutdown conditions below the enable temperature (i.e., low temperature) as defined in Branch Technical Position RSB 5-2 as modified by ASME Boiler and Pressure Vessel Code Case N-514.

Based on the Kewaunee Appendix G LTOP protection pressure-temperature limits calculated through 3311] effective full-power years, the LTOP System must be OPERABLE whenever one or more of the RCS cold leg temperatures are

2001F and the head is on the reactor vessel. The LTOP system is considered OPERABLE when all four valves on the RHR suction piping (valves RHR-1A, I B, 2A, 2B) are open and valve RHR-33-1, the LTOP valve, is able to relieve RCS overpressure events without violating Figure TS 3.1-2.

The set pressure specified in TS 3.1 .b.4 includes consideration for the opening pressure tolerance of +/-3% (i 15 psig) as defined in ASME Boiler and Pressure Vessel Code,Section III, Division 1, Subsection NC: Class 2 Components for Safety Relief Valves. The analysis of pressure transient conditions has demonstrated acceptable relieving capability at the uppertolerance limit of 515 psig.

If one train of RHR suction piping to RHR 33-1 is isolated, then the valves and valve breakers in the other train shall be verified open, and the isolated flowpath must be restored within five days. If the isolated flowpath cannot be restored within five days, then the RCS must be depressurized and vented through at least a 6.4 square inch vent within an additional eight hours.

If both trains of RHR suction are isolated or valve RHR 33-1 is inoperable, then the system can still be considered OPERABLE if an alternate vent path is provided which has the same or greater effective flow cross section as the LTOP safety valve (26.4 square inches). If vent path isprovided by physical openings in the RCS pressure boundary (e.g., removal of pressurizer safety valves or steam generator manways), then the vent path is considered secured in the open position.

Note

[1] Although the curves were developed for 33 EFPY, they are limited to 28 EFPY (corresponding to the end of cycle 28) by WPSC Letter NRC-99-017.

TS B3.1-7 Amendment No. .165

Leakage of Reactor Coolant (TS 3.1.d)116)

TS (TS 3.1.d.1)

Leakage from the Reactor Coolant System is collected in the containment or by the other closed systems. These closed systems are: the Steam and Feedwater System, the Waste Disposal System and the Component Cooling System. Assuming the existence of the maximum allowable activity in the reactor coolant, the rate of I gpm unidentified leakage would not exceed the limits of 10 CFR Part 20. This is shown as follows:

If the reactor coolant activity is 911E pCi/cc (E = average beta plus gamma energy per disintegration in Mev) and 1 gpm of leakage is assumed to be discharged through the air elector, or through the Component Cooling System vent line, then the yearly whole body dose resulting from this activity at the SITE BOUNDARY, using an annual average X/Q = 2.0 x 10 6sec/m3 , is 0.09 rem/yr, compared with the 10 CFR Part 20 limits of 0.1 rem/yr.

With the limiting reactor coolant activity and assuming Initiation of a 1 gpm leak from the Reactor Coolant System to the Component Cooling System, the radiation monitor in the component cooling pump inlet header would annunciate in the control room. Operators would then investigate the source of the leak and take actions necessary to isolate it. Should the leak result in a continuous discharge to the atmosphere via the component cooling surge tank and waste holdup tank, the resultant dose rate at the SITE BOUNDARY would be 0.09 rem/yr as given above.

Leakage directly into the containment indicates the possibility of a breach in the coolant envelope.

The limitation of I gpm for an unidentified source of leakage is sufficiently above the minimum detectable leak rate to provide a reliable indication of leakage, and is well below the capacity of one charging pump (60 gpm).

Twelve hours of operation before placing the reactor in the HOT SHUTDOWN condition are required to provide adequate time for determining whether the leak is into the containment or into one of the closed systems and to identify the leakage source.

TS 3.1.d.2 Limiting the leakage through any single steam generator to 150 gpd ensures that tube integrity is maintained during a design basis main steam line break or loss-of-coolant accident. Remaining within this leakage rate provides reasonable assurance that no single tube-flaw will sufficiently enlarge to create a steam generator tube rupture as a result of stresses caused by a Loss of Coolant Accident (LOCA) or a main steam line break accident within the time allowed for detection of the accident condition and resulting commencement of plant shutdown. This operational leakage rate is less than the condition assumed in design basis safety analyses and conforms to industry standards established by the Nuclear Energy Institute through its NEI 97-06, "Generic Steam Generator Program Guidelines."

(16) USAR Sections 6.5, 11.2.3, 14.2.4 TS B3.1-9 Amendment No. 165

TS 3.1.d.3 When the source of leakage has been identified, the situation can be evaluated to determine if operation can safely continue. This evaluation will be performed bythe plant operating staffand will be documented in writing and approved by either the Plant Manager or his designated alternate.

Under these conditions, an allowable Reactor Coolant System leak rate of 10 gpm has been established. This explained leak rate of 10 gpm is within the capacity of one charging pump as well as being equal to the capacity of the Steam Generator Blowdown Treatment System.

TS 3.1.d.4 The provision pertaining to a non-isolable fault in a Reactor Coolant System component is not intended to coversteam generator tube leaks, valve bonnets, packings, instrument fittings, orsimilar primary system boundaries not indicative of major component exterior wall leakage.

TS 3.1.d.5 Ifleakage is to the containment, it may be identified by one or more of the following methods:

A. The containment air particulate monitor is sensitive to low leak rates. The rates of reactor coolant leakage to which the instrument Is sensitive are dependent upon the presence of corrosion product activity.

B. The containment radiogas monitor is less sensitive and is used as a backup to the air particulate monitor. The sensitivity range of the instrument is approximately 2 gpm to

> 10 gpm.

C. Humidity detection provides a backup to A. and B. The sensitivity range of the instrumentation is from approximately 2 gpm to 10 gpm.

D. A leakage detection system is provided which determines leakage losses from all water and steam systems within the containment. This system collects and measures moisture condensed from the containment atmosphere by fancoils of the Containment Air Cooling System and thus provides a dependable and accurate means of measuring integrated total leakage, including leaks from the cooling coils themselves which are part of the containment boundary. The fancoil units drain to the containment sump, and all leakage collected by the containment sump will be pumped to the waste holdup tank. Pump running time will be monitored in the control room to indicate the quantity of leakage accumulated.

If leakage is to another closed system it will be detected by the area and process radiation monitors and/or inventory control.

TS B3.1-10 Amendment No. 165

Maximum Reactor Coolant Oxycen, Chloride and Fluoride Concentration (TS 3.1.e)

By maintaining the oxygen, chloride and fluoride concentrations in the reactor coolant below the limits as specified in TS 3.1.e.1 and TS 3.1.e.4, the integrity of the Reactor Coolant System is ensured under all OPERATING conditions. (17)

Ifthese limits are exceeded, measures can be taken to correct the condition, e.g., replacement of ion exchange resin or adjustment of the hydrogen concentration in the volume control tank.0')

Because of the time-dependent nature of any adverse effects arising from oxygen, chloride, and fluoride concentration in excess of the limits, it is unnecessary to shut down immediately since the condition can be corrected. Thus, the time periods for corrective action to restore concentrations within the limits have been established. Ifthe corrective action has not been effective at the end of the time period, reactor cooldown will be initiated and corrective action will continue.

The effects of contaminants in the reactor coolant are temperature dependent. The reactor may be restarted and operation resumed if the maximum concentration of any of the contaminants did not exceed the permitted transient values; otherwise a safety review by the Plant Operations Review Committee is required before startup.

Minimum Conditions for Criticality (TS 3.11f)

During the early part of the fuel cycle, the moderator temperature coefficient may be calculated to be positive at* 60% RATED POWER. The moderatorcoefficientwill be most positive at the beginning of life of the fuel cycle, when the boron concentration in the coolant Is greatest. Later in the fuel cycle, the boron concentrations in the coolant will be lower and the moderator coefficients eitherwill be less positive or will be negativeP)

The requirementthat the reactoris not to be made critical except as specified in TS 3.1.f.1 provides increased assurance thatthe properrelationship between reactorcoolant pressure and temperature will be maintained during system heatup and pressurization whenever the reactor vessel is in the nil-ductility temperature range. Heatup to this temperature will be accomplished by operating the reactor coolant pumps and by the pressurizer heaters.

The shutdown margin specified in the COLR precludes the possibility of accidental criticality as a result of an increase in moderator temperature or a decrease in coolant pressure.'>9 The requirement that the pressurizer is partly voided when the reactor is < 1% subcritical ensures that the Reactor Coolant System will not be solid when criticality is achieved.

(17) USAR Section 4.2 1 USAR Section 9.2

' USAR Section 3.2.1 TS B3.1-11

Amendment No. 165

The requirement that the reactor is not to be made critical when the moderator coefficient is greater than the value specified in the COLR has been imposed to prevent any unexpected powerexcursion during normal operation as a result of either an increase in moderator temperature or a decrease in coolant pressure. The moderator temperature coefficient limits are required to maintain plant operation within the assumptions contained in the USAR analyses. Having an initial moderator temperature coefficient no greater than the value specified in the COLR provides reasonable assurance that the moderator temperature coefficient will be negative at 60% rated thermal power.

The moderator temperature coefficient requirement is waived during low power physics tests to permit measurement of reactor moderator coefficient and other physics design parameters of interest. During physics tests, special OPERATING precautions will be taken. In addition, the strong negative Doppler coefficientd 20 ) and the small integrated Ak/k would limit the magnitude of a power excursion resulting from a reduction in moderator density.

Suitable physics measurements of moderator coefficients of reactivity will be made as part of the startup testing program to verify analytical predictions.

Analysis has shown that maintaining the moderator temperature coefficient at criticality less than or equal to the value specified in the COLR will ensure that a negative coefficient will exist at 60%

power. Current safety analysis supports OPERATING up to 60% power with a moderator temperature coefficient less than or equal to the value specified in the COLR. At power levels greater than 60%, a negative moderator temperature coefficient must exist.

The calculated hot full power (HFP) moderator temperature coefficient will be more negative than the value specified in the COLR for at least 95% of a cycle's time at HFP to ensure the limitations associated with and anticipated transient without scram (ATWS) event are not exceeded. NRC approved methods 2 0 )(21 )will be used to determine the lowest expected HFP moderator temperature coefficient for the 5% of HFP cycle time with the highest boron concentration. The cycle time at HFP is the maximum number of days that the cycle could be at HFP based on the design calculation of cycle length. The cycle time at HFP can also be expressed in terms of burnup by converting the maximum number of days atfull power to an equivalent bumup. Ifthis HFP moderatortemperature coefficient is more negative than the value specified Inthe COLR, then the ATWS design limit will be met for 95% of the cycle's time at HFP. Ifthis HFP moderator temperature coefficient design limit is still not met after excluding the 5% of the cycle bumup with the highest boron concentration, then the core loading must be revised.

The results of this design limit consideration will be reported in the Reload Safety Evaluation Report.

(20) USAR Section 3.2.1 (20) `NRC Safety Evaluation Report for Qualification of Reactor Physics, Methods for Application to Kewaunee," dated October22, 1979.

(21) "NRC Safety Evaluation Report f6r the Reload Safety Evaluation Methods for Application to Kewaunee," dated April 11, 1988.

TS B3.1-12 Amendment No. 165

Inthe event that the limits as provided in the COLR are not met, administrative rod withdrawal limits shall be developed to prevent further increases in temperature with a moderator temperature coefficient that is outside analyzed conditions. In this case, the calculated HFP moderator temperature coefficient will be made less negative by the same amount the hot zero power moderator temperature coefficient exceeded the limit as provided in the COLR. This will be accomplished by developing and implementing administrative control rod withdrawal limits to achieve a m oderator t emperature c oefficient within t he I imits f or HFP moderator temperature coefficient.

Due to the control rod insertion limits of TS 3.1 O.d and potentially developed control rod withdrawal limits, it is possible to have a band for control rod location at a given power level. The withdrawal limits are not required if TS 3.1.f.3 is satisfied or if the reactor is subcritical.

If after 24 hours1 days 0.143 weeks 0.0329 months , withdrawal limits sufficient to restore the moderator temperature coefficient to within the limits as provided in the COLR are not developed, then the plant shall be taken to HOT STANDBY until the moderator temperature coefficient is within the limits as specified in the COLR.

The reactor is allowed to return to criticality whenever TS 3.1 . is satisfied.

TS B3.1-13 Amendment No. 165

BASIS - Refueling Operations (TS 3.8) I The equipment and general procedures to be utilized during REFUELING OPERATIONS are discussed in the USAR. Detailed Instructions, the above specified precautions, and the design of the fuel handling equipment incorporating built-in interlocks and safety features, provide assurance that no incident occurs during the REFUELING OPERATIONS thatwould resultin a hazard to public health and safety.!' Whenever changes are not being made in core geometry, one flux monitor is sufficient. This permits maintenance of the instrumentation. Continuous monitoring of radiation levels (TS 3.8.a.2) and neutron flux provides immediate indication of an unsafe condition. The residual heat removal pump is used to maintain a uniform boron concentration.

A minimum shutdown margin of greater than or equal to 5% AkMk must be maintained in the core.

The boron concentration as specified in the COLR is sufficient to ensure an adequate margin of safety. The specification for REFUELING OPERATIONS shutdown margin is based on a dilution during refueling accident.(21 With an initial shutdown margin of 5% Ak/k, under the postulated accident conditions, it will take longer than 30 minutes for the reactor to go critical. This is ample time for the operator to recognize the audible high count rate signal, and isolate the reactor makeup water system. Periodic checks of refueling waterboron concentration ensure that proper shutdown margin is maintained. Specification 3.8.a.6 allows the control room operator to inform the manipulator operator of any impending unsafe condition detected from the main control board indicators during fuel movement.

Interlocks are utilized during REFUELING OPERATIONS to ensure safe handling. Only one assembly at a time can be handled. The fuel handling hoist is dead weight tested prior to use to assure propercrane operation. Itwill notbe possible to liftorcarryheavyobjects overthespentfuel pool when fuel is stored therein through interlocks and administrative procedures. Placement of additional spent fuel racks will be controlled by detailed procedures to prevent traverse directly above spent fuel.

The one hundred forty-eight hour decay time following plant shutdown is consistent with the spent fuel pool cooling analysis and also bounds the assumption used In the dose calculation for the fuel handling accident. The requirement for the spent fuel pool sweep system, including charcoal adsorbers, to be operating when spent fuel movement is being made provides added assurance that the off-site doses will be within acceptable limits in the event of a fuel handling accident. The spent fuel pool sweep system is designed to sweep the atmosphere above the refueling pool and release to the Auxiliary Building vent during fuel handling operations. Normally, the charcoal adsorbers are bypassed but for purification operation, the bypass dampers are closed routing the air flow through the charcoal adsorbers. Ifthe dampers do not close tightly, bypass leakage could exist to negate the usefulness of the charcoal adsorber. Ifthe spent fuel pool sweep system is found not to be operating, fuel handling within the Auxiliary Building will be terminated until the system can be restored to the operating condition.

The bypass dampers are integral to the filter housing. The test of the bypass leakage around the charcoal adsorbers will include the leakage through these dampers.

(1)USAR Section 9.5.2 (2) USAR Section 14.1 TS 83.8.-I Amendment No. 165

High efficiency particulate absolute (HEPA) filters are installed before the charcoal adsorbers to prevent clogging of the iodine adsorbers. The charcoal adsorbers are installed to reduce the potential radiolodine releases to the atmosphere. Bypass leakage forthe charcoal adsorbers and particulate removal efficiency for HEPA filters are determined by halogenated hydrocarbon and DOP, respectively. The laboratory carbon sample test results indicate a radioactive methyl iodide removal efficiency under test conditions which are more severe than accident conditions.

Operation of the fans significantly differentfrom the design flowwill change the removal efficiencyof the HEPA filters and charcoal adsorbers. Ifthe performances are as specified, the calculated doses would be less than the guidelines stated in 10 CFR Part 100 for the accidents analyzed.

The spent fuel pool sweep system will be operated for the first month after reactor is shutdown for refueling during fuel handling and crane operations with loads over the pool. The potential consequences of a postulated fuel handling accident without the system are a verysmall fraction of the guidelines of 10 CFR Part 100 after one month decay of the spent fuel. Heavy loads greater than one fuel assembly are not allowed over the spent fuel.

In-place testing procedures will be established utilizing applicable sections of ANSI N510 - 1975 standard as a procedural guideline only.

A fuel handling accident in containment does not cause containment pressurization. One containment door in each personnel air lock can be closed following containment personnel evacuation and the containment ventilation and purge system has the capabilityto initiate automatic containment ventilation isolation to terminate a release path to the atmosphere.

The presence of a licensed senior reactor operator at the site and designated in charge provides qualified supervision of the REFUELING OPERATIONS during changes in core geometry.

Accident analysis assumes a charcoal adsorber efficiency of 90%. 3) To ensure the charcoal adsorbers maintain that efficiency throughout the operating cycle, a safety factor of 2 is used.

Therefore, if accident analysis assumes a charcoal adsorber efficiency of 90%, this equates to a methyl iodide penetration of 10%. If a safety factorof 2 is assumed, the methyl iodide penetration is reduced to 5%. Thus, the acceptance criteria of 95% efficient will be used for the charcoal adsorbers.

Although committing to ASTM D3803-89, it was recognized that ASTM D3803-89 Standard references MilitaryStandards MIL-F-51068D, Filter, Particulate High Efficiency, Fire Resistant, and MIL-F-51079A, Filter, Medium Fire Resistant, High Efficiency. These specifications have been revised and the latest revisions are, MIL-F-51 068F and MIL-F-51 079D. These revisions have been canceled and superseded byASME AG-I, Code on NuclearAir and Gas Treatment. ASME AG-i is an acceptable substitution. Consequently, other referenced standards can be substituted if the new standard or methodology Is shown to provide equivalent or superior performance to those referenced in ASTM D3803-89.

(3)USAR TABLE 14.3-8, "Major Assumptions for Design Basis LOCA Analysis" TS 83.8-2 Amendment No. 1i6-5

BASIS-Control Rod and Power Distribution Limits (TS 3.10P) I Shutdown Reactivity (TS 3.1Q.a)

Trip shutdown reactivity is provided consistent with plant safety analysis assumptions. To maintain the required trip reactivity, the rod insertion limits as specified in the COLR must be observed. In addition, for HOT SHUTDOWN conditions, the shutdown margin as specified in the COLR must be provided for protection against the steam line break accident.

Rod insertion limits are used to ensure adequate trip reactivity, to ensure meeting power distribution limits, and to limit the consequences of a hypothetical rod ejection accident.

The exception to the rod insertion limits in TS 3.10.d.3 is to allow the measurement of the worth of all rods. This measurement is a part of the Reactor Physics Test Program performed at the startup of each cycle. Rod worth measurements augment the normal fuel cycle design calculations and place the knowledge of shutdown capability on a firm experimental as well as analytical basis.

Operation with abnormal rod configuration during low power and zero power testing is permitted because of the brief period of the test and because special precautions are taken during the test.

TS 1.0.r, "Shutdown Margin," states the definition of shutdown margin as used in the technical specifications. As a part of this definition is a statement which removes the assumption that the highest reactivity worth rod cluster control assembly (RCCA) is fully withdrawn. This includes the verification that all RCCA's are fully inserted by two independent means. Although not fully independent, this requirement refers to Indications which are independent. These independent means include such indicators as the control board individual rod position indicators or the rod position as indicated on the plant process computer system (PPCS) or the condition monitors referenced in TS 3.10.e.

Power Distribution Control (TS 3.10.b)

Criteria Criteria have been chosen for Condition I and It events as a design basis for fuel performance related to fission gas release, pellet temperature, and cladding mechanical properties. First, the peak value of linear power density must not exceed the value assumed in the accident analyses. The peak linear power density is chosen to ensure peak clad temperature during a postulated large break loss-of-coolant accident is less than the 2200° F limit. Second, the minimum DNBR in the core must not be less than the DNBR limit in normal operation or during Condition I or II transient events.

TS B3.10-1 Amendment No. 165

FpN(Z), Height Dependent Nuclear Flux Hot Channel Factor F0N(Z), Height Dependent Nuclear Flux Hot Channel Factor, is defined as the maximum local linear power density in the core at core elevation Z divided by the core average linear power density, assuming nominal fuel rod dimensions.

FOEO(Z) is the measured FaN(Z) obtained at equilibrium conditions during the target flux determination.

An upper bound envelope for FQN(Z) as specified in the COLR has been determined from extensive analyses considering all OPERATING maneuvers consistent with the Technical Specifications on power distribution control as given in TS 3.10. The results of the loss-of-coolant accident analyses based on this upper bound envelope indicate the peak clad temperatures remain less than the 22000 F limit.

The F0N(Z) limits as specified in the COLR are derived from the LOCA analyses. The LOCA analyses are performed for FRA-ANP heavy fuel and for FRA-ANP standard fuel.

When a FaN(Z) measurement is taken, both experimental error and manufacturing tolerance must be allowed for. Five percent is the appropriate allowance for a full core map taken with the movable incore detector flux mapping system and 3% is the appropriate allowance for manufacturing tolerance.

FQN(Z) is arbitrarily limited for P c0.5 (except for low power physics tests).

FeHN Nuclear Enthalpv Rise Hot Channel Factor FwN, Nuclear Enthalpy Rise Hot Channel Factor, is defined as the ratio of the maximum integral of linear, power along a fuel rod to the core average integral fuel rod power.

It should be noted that F6HN is based on an integral and Is used as such in DNBR calculations. Local heat fluxes are obtained by using hot channel and adjacent channel explicit power shapes which take into account variations in horizontal (x-y) power shapes throughout the core. Thus, the horizontal power shape at the point of maximum heat flux is not necessarily directly related to FH N.

The FaHN limit is determined from safety analyses of the limiting DNBR transient events. The safety analyses are performed for FRA-ANP heavy fuel and for FRA-ANP standard fuel. In these analyses, the important operational parameters are selected to minimize DNBR. The results of the safety analyses must demonstrate that minimum DNBR is greater than the DNBR limit for a fuel rod operating at the FAHN limit.

The use of FaHN in TS 3.1 0.b.5.C is to monitor "upbum" which is defined as an increase in FaHN with exposure. Since this is not to be confused with observed changes in peak power resulting from such phenomena as xenon redistribution, control rod movement, power level changes, or changes in the number of instrumented thimbles recorded, an allowance of 2% is used to account for such changes.

TS B3.10-2 Amendment No. 165

I Rod Bow Effects No penalty for rod bow effects needs to be included in TS 3.10.b.1 for FRA-ANP fuel.(')

Surveillance Measurements of the hot channel factors are required as part of startup physics tests, at least each full power month of operation, and whenever abnormal power distribution conditions require a reduction of core power to a level based on measured hot channel factors. The incore map taken following initial loading provides confirmation of the basicnucleardesign bases including properfuel loading pattems. The periodic monthly incore mapping provides additional assurance that the nuclear design bases remain inviolate and identifies operational anomalies which would otherwise affect these bases.

For normal operation, it is not necessary to measure these quantities. Instead it has been determined that, provided certain conditions are observed, the hot channel factor limits will be met.

These conditions are as follows:

1. Control rods In a single bank move together with no individual rod insertion differing by more than an indicated 12 steps from the bank demand position where reactor power is 285%, or an indicated 24 steps when reactor power is < 85%.

2. Control rod banks are sequenced with overlapping banks as specified in the COLR.

3. The control bank insertion limits as specified in the COLR are not violated, except as allowed by TS 3.1 0.d.2.

4. Axial power distribution control specifications which are given Interms of flux difference control and control bank insertion limits are observed. Flux difference refers to the difference in signals between the top and bottom halves of two-section excore neutron detectors. The flux difference is a measure of the axial offset which is defined as the difference in normalized power between the top and bottom halves of the core.

The specifications for axial powerdistribution control referred to above are designed to minimize the effects of xenon redistribution on the axial power distribution during load-follow maneuvers.( 2 )

Conformance with TS 3.10.b.9 through TS 3.1 0.b.12 ensures the F0N upper bound envelope is not exceeded and xenon distributions will not develop which at a later time would cause greater local power peaking.

At the beginning of cycle, power escalation may proceed without the constraints of TS 3.1 0.b.5 since the startup test program provides adequate surveillance to ensure peaking factor limits. Target flux difference surveillance is initiated afterachieving equilibrium conditions for sustained operation.

')N. E. Hoppe, "Mechanical Design Report Supplement for Kewaunee High Burnup (49 GWDIMTU)

Fuel Assemblies," XN-NF-84-28(P), Exxon Nuclear Company, July 1984.

(2)XN-NF-77-57 Exxon Nuclear Power Distribution Control for Pressurized Water Reactor, Phase II, January 1978.

TS B3.10-3 Amendment No. 165

The target (or reference) value of flux difference is determined as follows. A t anytime that equilibrium xenon conditions have been established, the indicated flux difference is determined from the nuclear instrumentation. This value, divided by the fraction of full power at which the core was OPERATING is the full power value of the target flux difference. Values for all other core power levels are obtained by multiplying the full power value by the fractional power. Since the indicated equilibrium value was noted, no allowances for excore detector error are necessary and indicated deviations as specified in the COLR are permitted from the indicated reference value.

Strict control of the flux difference (and rod position) is not as necessary during part power operation. This is because xenon distribution control at part power is not as significant as the control at full power and allowance has been made in predicting the heat flux peaking factors for less strict control at part power. Strict control of the flux difference is not possible during certain physics tests or during required, periodic, excore calibrations which require larger flux differences than permitted. Therefore, the specifications on power distribution control are not applied during physics tests or excore calibrations; this is acceptable due to the low probability of a significant accident occurring during these operations.

In some instances of rapid plant power reduction automatic rod motion will cause the flux difference to deviate from the target band when the reduced power level is reached. This does not necessarily affect the xenon distribution sufficiently to change the envelope of peaking factors which can be reached on a subsequent return to full power within the target band; however, to simplify the specification, a limitation of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months in any period of 24 hours1 days 0.143 weeks 0.0329 months is placed on operation outside the band.

This ensures that the resulting xenon distributions are not significantly different from those resulting from operation within the target band.

The instantaneous consequences of being outside the band, provided rod insertion limits are observed, is not worse than a 10% increment In peaking factor for flux difference envelope as specified in the COLR. Therefore, while the deviation exists the power level is limited to 90% or lower depending on the indicated flux difference without additional core monitoring. If, for any reason, flux difference is not controlled within the target band as specified in the COLR for as long a period as one hour, then xenon distributions may be significantly changed and operation at 50% is required to protect against potentially more severe consequences of some accidents unless incore monitoring is initiated.

As discussed above, the essence of the procedure is to maintain the xenon distribution in the core as close to the equilibrium full power condition as possible. This is accomplished by using the boron system to position the full length control rods to produce the required indicated flux difference.

For Condition IIevents the core is protected from overpower and a minimum DNBR less than the DNBR limit by an automatic Protection System. Compliance with the specification is assumed as a precondition for Condition II transients; however, operator error and equipment malfunctions are separately assumed to lead to the cause of the transients considered.

TS B3.10-4 Amendment No. .165

Quadrant Power Tilt Limits (TS 3.10.cI The radial power distribution within the core must satisfy the design values assumed for calculation of power capability. Radial power distributions are measured as part of the startup physics testing and are periodically measured at a monthly or greaterfrequency. These measurements are taken to assure that the radial power distribution with any quarter core radial power asymmetry conditions are consistent with the assumptions used in power capability analyses.

The quadrant tilt power deviation alarm is used to indicate a sudden or unexpected change from the radial power distribution mentioned above. The 2% tilt alarm setpoint represents a minimum practical value consistent with instrumentation errors and operating procedures. This symmetry level is sufficient to detect significant misalignment of control rods. Misalignment of control rods is considered to be the most likely cause of radial power asymmetry. The requirement for verifying rod position once each shift is imposed to preclude rod misalignment which would cause a tilt condition less than the 2% alarm level. This monitoring Is required by TS 4.1.

The two hour time interval in TS 3.1 0.c is considered ample to identify a dropped or misaligned rod.

If the tilt condition cannot be eliminated within the two hour time allowance, additional time would be needed to investigate the cause of the tilt condition. The measurements would include a full core power distribution map using the movable detector system. For a tilt ratio > 1.02 but < 1.09, an additional 22 hours0.917 days 0.131 weeks 0.0301 months time interval is authorized to accomplish these measurements. However, to assure that the peak core power Is maintained below limiting values, a reduction of reactor powerof 2% for each 1%of indicated tilt is required. Power distribution measurements have Indicated that the core radial power peaking would not exceed a two-to-one relationship with the indicated tilt from the excore nuclear detector system for the worst rod misalignment. If a tilt ratio of > 1.02 but < 1.09 cannot be eliminated after 24 hours1 days 0.143 weeks 0.0329 months , then the reactor power level will be reduced to <50%.

If a misaligned rod has caused a tilt ratio > 1.09, then the core power shall be reduced by 2% of rated value for every 1%of indicated power tilt ratio > 1.0. If after eight hours the rod has not been realigned, then the rod shall be declared inoperable in accordance with TS 3.1 0.e, and action shall be taken in accordance with TS 3.1 0.g. If the tilt condition cannot be eliminated after 12 hours0.5 days 0.0714 weeks 0.0164 months , then the reactor shall be brought to a minimum load condition; i.e., electric power <30 MW. Ifthe cause of the tilt condition has been identified and is in the process of being corrected, then the generator may remain connected to the grid.

If the tilt ratio is > 1.09, and it is not due to a misaligned rod, then the reactor shall be brought to ano load condition (i.e., reactor powers 5%) for investigation by flux mapping Although the reactor may be maintained critical for flux mapping, the generator must be disconnected from the grid since the cause of the tilt condition is not known, or it cannot be readily corrected.

Rod Insertion Limits (TS 3.10.d)

The allowed completion time of two hours for restoring the control banks to within the insertion limits provides an acceptable time forevaluation and repairing minor problems without allowing the plant to remain in an unacceptable condition for an extended period of time.

Operation beyond the rod insertion limits is allowed for a short-time period in order to take conservative action because the simultaneous occurrence of either a LOCA, loss-of-flow accident, ejected rod accident, or other accident during this short time period, together with an inadequate power distribution or reactivity capability, has an acceptably low probability.

TS B3.10-5 Amendment No. 165

The tme limits of six hours to achieve HOT STANDBY and an additional six hours to achieve HOT SHUTDOWN allow for a safe and orderly shutdown sequence and are consistent with most of the remainder of the Technical Specifications.

Rod MisaliQnment Limitations (TS 3.10.e0 During normal power operation it is desirable to maintain the rods in alignment with their respective banks to provide consistency with the assumption of the safety analyses, to maintain symmetric neutron flux and power distribution profiles, to provide assurance that peaking factors are within acceptable limits and to assure adequate shutdown margin.

Analyses have been performed which indicate that the above objectives will be met if the rods are aligned within the limits of TS 3.1 0.e. A relaxation in those limits for power levels < 85% is allowable because of the increased margin in peaking factors and available shutdown margin obtained while OPERATING at lower power levels. This increased flexibility is desirable to account for the nonlinearity inherent in the rod position indication system and for the effects of temperature and power as seen on the rod position indication system.

Rod position measurement is performed through the effects of the rod drive shaft metal on the output voltage of a series of vertically stacked coils located above the head of the reactor pressure vessel. The rod position can be determined by the analog individual rod position indicators (IRPI),

the plant process computer which receives a voltage input from the conditioning module, or through the conditioning module output voltage via a correlation of rod position vs. voltage.

The plant process computer converts the output voltage signal from each IRPI conditioning module to an equivalent position (in steps) through a curve fitting process, which may include the latest actual voltage-to-position rod calibration curve.

The rod position as determined by any of these methods can then be compared to the bank demand position which is indicated on the group step counters to determine the existence and magnitude of a rod misalignment. This comparison is performed automatically by the plant process computer.

The rod deviation monitor on the annunciator panel is activated (or reactivated) if the two position signals for any rod as detected by the process computer deviate by more than a predetermined value. The value of this setpoint is set to warn the operatorwhen the Technical Specification limits are exceeded.

The rod position indicator system is calibrated once per REFUELING cycle and forms the basis of the correlation of rod position vs. voltage. This calibration is typically performed at HOT SHUTDOWN conditions prior to initial operations forthatcycle. Upon reaching full powerconditions and verifying that the rods are aligned with their respective banks, the rod position indication may be adjusted to compensate for the effects of the power ascension. After this adjustment is performed, the calibration of the rod position indicator channel is checked at an intermediate and low level to confirm that the calibration is not adversely affected by the adjustment.

TS B3.10-6

Anendmenii No. 165

Inoperable Rod Position Indicator Channels (TS 3.101)

The rod position indicator channel is sufficiently accurate to detect a rod +/- 12 steps away from its demand position. If the rod position indicator channel is not OPERABLE, then the operatorwill be l fullyaware of the inoperability of the channel, and special surveillance of core power tilt indications, using established procedures and relying on excore nuclear detectors, andlor movable incore detectors, will be used to verify power distribution symmetry.

Inoperable Rod Limitations (TS 3.1 0.q)

One inoperable control rod is acceptable provided the potential consequences of accidents are not worse than the cases analyzed in the safety analysis report. A 30-day period is provided for the reanalysis of all accidents sensitive to the changed initial condition.

Rod Drop Time (TS 3.1 0.h)

The required drop time to dashpot entry is consistent with safety analysis.

Core Average Temperature (TS 3.1 O.k)

The core average temperature limit is consistent with the safety analysis.

Reactor Coolant System Pressure (TS 3.10.1)

The reactor coolant system pressure limit is consistent with the safety analysis.

Reactor Coolant Flow (TS 3.10.m)

The reactor coolant flow limit is consistent with the safety analysis.

DNBR Parameters (TS 3.10.n)

The DNBR related safety analyses make assumptions on reactortemperature, pressure, and flow.

Inthe event one of these parameters does not meet the TS 3.10.k, TS 3.10.1 orTS 3.1 0.m limits, an analysis can be performed to determine a power level at which the MDNBR limit is satisfied.

TS B3.10-7 Amendment No. 165

v t e Letter Sequence Acceptance Review
TAC:MB5717, Delete SR Frequencies Based on Inoperable Alarms (Approved, Closed) TAC:MB5718, Delete SR Frequencies Based on Inoperable Alarms (Approved, Closed)
Initiation Request, Request, Request, Request, Request Acceptance, Acceptance, Acceptance Administration Questions 31 October 2002 3 July 2002 23 August 2002 21 November 2002 21 November 2002 21 January 2003 Answers 27 February 2003 27 February 2003 Results Approval, Approval, Approval, Approval Other: ML022280368, ML023120333, ML023300476, ML023440297, ML023650134, ML030700456, ML031130505
MONTHYEAR2002-09-05 [Table View]

ML031180636

Text

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