ML18005B072

Proposed Tech Specs Re Various Administrative Corrections

ML18005B072
Person / Time
Site:	Harris
Issue date:	09/25/1989
From:	CAROLINA POWER & LIGHT CO.
To:
Shared Package
ML18005B070	List: ML18005B069 ML18005B072
References
NUDOCS 8909290077
Download: ML18005B072 (24)

Text

INDEX

-LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE UIREMENTS SECTION PACE 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE.................................... 3/4 2-1 FIGURE 3.2-1 AXIAL FLUX DIFFERENCE LIMITS AS A FUNCTION OF RATED THERMAL POWER FOR RAOC............................. 3/4 2-4 3/4 2 2 HEAT FLUX HOT CHANNEL FACTOR FQ(Z) 3/4 2-5 FIGURE 3.2-2 K(Z) - LOCAL AXIAL PENALTY FUNCTION FOR FQ(Z).. 3/4 2-8 3/4.2.3 RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 2-9 3/4 ' ' QUADRANT POWER TILT RATIO ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 2-11 3/4.2.5 DNB PARAMETERS.................o..............oo......... 3/4 2-14 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION....... .......... ... ~ ~ 3/4 3-1 TABLE 3.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION..e............... 3/4 3-2 TABLE 3.3-2 (DELETED) ..o'..e...o..o..o.......

~ ~ ~ .....o.o......... 3/4 3-9 TABLE 4.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS..oooo.. ~ .o........oo....... .....o...... 3/4 3-11 3/4.3.2 INSTRUMENTATION.wooing ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION....oo... ~ o.o ~ ....................o.....'.. 3/4 3-16 TABLE 3.3-3 ENGINEERED'SAFETY FEATURES ACTUATION SYSTEM

. o................

~ ... ..o. ...

~ ~ 3/4 3-18 TABLE 3.3-4 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ AN ~ 3/4 3-28 TABLE 3o3-5 (DELETED) ~ ~ .... o..........o.....o.... ..............

~ ~ 3/4 3-37 [

TABLE 4.3-2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS......o..... 3/4 3-41 3/4.3.3 MONITORING INSTRUMENTATION Radiation Monitoring Eor Plant Operations...... ~ ~ ........ 3/4 3-5O SHEARON HARRIS UNIT 1 Amendment No.

SVO9V90077 8eoeaS PDR ADOCN, 05000400 P PNU

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE UIREMENTS SECT10N PAGE 3/4.4.9 PRESSURE/TEMPERATURE LIMITS Reactor Coolant System................................... 3/4 4-33 FIGURE 3.4-2 REACTOR COOLANT SYSTEM COOLDOWN LIMITATIONS APPLICABLE UP TO 4 EFPY.................................. 3/4 4-35 FIGURE 3.4"3 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS APPLICABLE UP TO 4 EFPY........................,......... 3/4 4-36 TABLE 4o4-5 (DELETED) ~ ~ ~ ~ ~ o ~ .oo ~ oo ~ ~ oo ~ ~ o..o. ~ ~ ~ o ~ . ~ .. ~ ooo ~ o ~ . ~ 3/4 4-37 TABLE 4 ~ 4 6 MAXIMUM HEATUP AND COOLDOWN RATES FOR MODES 4 ~ 5 AND 6 (WITH REACTOR VESSEL HEAD ON) ~ ~ ~ ~ ~ ~ ~ ~ ~ .. ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 4-38 P ressurxzer......... ~ ....... ...........................

~ ~ 3/4 4-39 Overpressure Protection Systemic ....... ................ ~ 3/4 4"40 FIGURE 3 '-4 MAXIMUM ALLOWED PORV SETPOINT FOR THE LOW TEMPERATURE OVERPRESSURE SYSTEM+. ~ ~ ~ ~ ~ . ~ ~ ~ ~ ~ ~ ~ .o ~ ~ o ~ ~ ~ . ~ oo ~ ~ ~ .. ~ ~ . ~ ~ ~ 3/4 4-41 3/4.4+10 STRUCTURAL INTEGRITYo ~ ~ o ~ .oo ~ . ~ ~ ~ ~ .o ~ o ~ ~ ~ o.. ~ .oo ~ .~. ~ ~ .. ~ 3/4 4-43 3/4.4.11 REACTOR COOLANT SYSTEM VENTS. ~ ~ ~ ~ ~ .. ..~ ~ ~

'

.... ...~ ~ . ~ ~ ~ . ~ 3/4 4"44 3/4 ' EMERGENCY CORE COOLING SYSTEMS 3/4 'ol ACCUMULATORS ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ 3/4 5-1

.2 ECCS SUBSYSTEMS - Ta GREATER THAN OR EQUAL TO O'.... 3/4 5-3 3/4.5.3 ECCS SUBSYSTEMS T LESS THAN 350'F........o.. ~ ..... ~ ~ 3/4 5-7 3/4 ~ 5 4

~ REFUELING WATER STORAGE TANK@ ~ ~ ~ "~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 5-9 SHEARON HARRIS UNIT 1 v111 Amendment No.

LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE UIREMENTS SECTION PACE 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT'ontainment Integrity.;.................................. 3/4 6-1 Containment Leakage.........'............................. 3/4 6-2 Containment Air Locks....................................

A 3/4 6-4 Internal Pressure........................................ 3/4 6-6 Air Temperature.....................,..................... 3/4 6'-7 Containment Vessel Structural Integrity.................. 3/4 6-8 Containment Ventl.latlon System........ .................. ~ 3/4 6-9 3/4 '.2 DEPRESSURIZATION AND COOLING SYSTEMS Containment Spray System................... . ~ 3/4 6"11 Spray Addltlve s s ~ ~

System....... . ~ ~ ~ ~ ~ ~ .. ..... .

~ ~ ~ ~ .......... 3/4 6-12 Containment Cooling System.. ~ .. ~ 3/4 6-13 3/4.6.3 CONTAINMENT ISOLATION VALVES. .. ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 6-14 TABLE 3o6-1 (DELETED)o ~ ~ o ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ o ~ o ~ ~ ~ .. ~ ~ ~ o ~ ...... ~ o ~ oo ~ o ~ ~ 3/4 6-16 3/4.6.4 COMBUSTIBLE GAS CONTROL Hydrogen Monitors........................................ 3/4 6-30 Electric Hydrogen Recombiners. ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 6-31 3/4.6.5 VACUUM RELIEF SYSTEM.. ~ ......... 3/4 6-32 3/4.7 PLANT SYSTEMS 3/4.7 ' TURBINE CYCLE Safety Valves............................................ 3/4 7-1 TABLE 3.7-1 -MAXIMUM ALLOWABLE POWER RANGE NEUTRON FLUX HIGH SETPOINT WITH INOPERABLE STEAM LINE SAFETY VALVES DURING 3 LOOP OPERATIONo ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 7-2 TABLE 3 '-2 STEAM fINE SAFETY VALVES PER LOOPe ~ ~ ~ ~ ~ ~ ~ ~ ~ ...... .. ~ 3/4 7-3 Auxiliary Feedwater System.....'...............,......... 3/4 7-4 Condensate Storage Tank.................................. 3/4 7-6 Speclllc Actlvlty P

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 7-7 TABLE 4.7-1 SECONDARY COOLANT SYSTEM SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM..o...o ~ ooee ~ ~ oo. ~ ~ ~ ~ o ~ ~ oo. ~ ..o.o.... 3/4 7-8 Main Steam Line Isolation Valves......................... 3/4 7-9 SHEARON HARRIS UNIT 1 lx Amendment No.

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE UIREMENTS SECTION PAGE 3/4.7.2 STEAM GENERATOR PRESSURE/TEMPERATURE LIMITATION.......... 3/4 7-10 3/4.7.3 COMPONENT COOLING WATER SYSTEM...... ~ ~ . ~ ~ ................. 3/4 7-11 3/4.7.4 EMERGENCY SERVICE WATER SYSTEM........................... 3/4 7-12 3/4.7.5 ULTIMATE HEAT SINK....;.................0................ 3/4 7-<3 3/4.7.6 CONTROL ROOM EMERGENCY FILTRATION SYSTEM0................ 3/4 7-14 3/4 ~ 7 ~ 7 REACTOR AUXILIARY BUILDING (RAB) EMERGENCY EXHAUST SYSTEM0 ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ -

3/4 7-17 3/4 '.8 SNUBBERS ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ , ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ .~ ...,, ~ ~ 3/4 7-19 FIGURE 407-1 (DELETED) ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ . ~ ~ ~ ~ ~ 3/4 7-24 [

3/4.7 ' SEALED SOURCE CONTAMINATION0 ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 7-25 3 /4 ' '0 (DELETED) ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 7-27 TABLE 3 '-3 (DELETED) ~ ~ ~ ~ .~~~~~~~~~~ .. ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 7-27 TABLE 3.7-4 (DELETED)00 ~ ~ ~ ~ 0 ~ .~~~0~~ ~ ~ 0 ~ ~ 0 ~ ~ ~ ~ ~ ~ ~ ~ ~ 0 ~ ~ ~ ~ ~ .'0 ~ ~ ~ .0 3/4 7-27 TABLE 3.7-5 (DELETED)00 ~ 0 ~ 00 ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 0 ~ 00 ~ ~ ~ ~ ~ ~ ~ 00 ~ ~ .0. ~ ~ 000 ~ 001 3/4 7-27 3 /4.F 11 (DELETED) ~ ~ ~ ~ 0 ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ . ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ . ~ ~ ~ . ~ ~ .. ~ ~ . ~ 3/4 7-27 3/F 7.12 AREA TEMPERATURE MONITORING. ~ ~ ~ . ~ ~ ~ ~ ~ ~ .. ~ ~ ~ ~ ~ . ~ ~ ~ ~ . ~ ~ ~ ... 3/4 7-28 TABLE 3.7-6 AREA TEMPERATURE MONITORING... ~ ~ ~ ~ ~ .. ...

~ ~ ~ . ~ ~ ~ ....... 3/4 7-29 3/407.13 ESSENTIAL SERVICES CHILLED WATER SYSTEM'.'.... ~ ~ ~ ........ 3/4 7-3O 3/4 ' ELECTRICAL POWER SYSTEMS 3/4.8.1 A.C. SOURCES 0 peratxng000 ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 0 ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 0 ~ ~ ~ 0 ~ ~ ~ 00 ~ 0 ~ 0 ~ 0 ~ 0 ~ 0 ~ ~ 00 3/4 8-1 TABLE 408-1 DIESEL GENERATOR TEST SCHEDULE ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 8-10 ShUtdOWll ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 8"11 3/4 ' ' D~C~ SOURCES 0 Peratl ng1010111010 0 0 010 ~ 0 0 .. 0. 110 0 01 ~ ~ 0 0110 0 010 0 0 011011. 3/4 8"12 TABLE 408-2 BATTERY SURVEILLANCE REQUIREMENTS ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 0. 3/4 8-14 SkLUtdOWll ~ ~ ~ ~ ~ ~ 0 ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 8-15 3/4.8.3 ONSITE POWER DISTRIBUTION Operating.10010.....1.00000.0.1000100010000000000.001. 000 3/4 8"16 S!1UtdOWIl ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 8-18 SHEARON HARRIS - UNIT 1 Amendment No.

~

LIMITING CONDITIONS SECTION 3/4.8.4 TABLE TABLE 3/4o9 3/4.F TABLE 4 3/4 ' '

3/4.9.9 3/4.9 '0 3/4.9.11 3/4.9.12 3/F 10 3/4.10.1 3/4.10 '

3/4.10.3 3/4.10.4 3/4.10.5 3.8-1 (DELETED)........

3.8-2 1

3/4 ~ 9 ~ 3 3/4.9.4 3/4.9 '

3/4 '

3/4.9 '

3/4.9.8

'

'-1

.

SHEARON HARRIS INSTRUMENTATION.~

DECAY TIME ~

CONTAINMENT BUILDING

~o ~

High Water Levels Low Water FOR OPERATION AND SURVEILLANCE RE UIREMENTS ELECTRICAL EQUIPMENT PROTECTIVE DEVICES Containment Penetration Conductor Overcurrent Protective Devices..................,...,.....,...,,...

Motor-Operated Valves Thermal Overload

..

~

~ ~ ~

CONTAINMENT VENTILATION ISOLATION SYSTEM o WATER LEVEL WATER LEVEL -

- REACTOR VESSELo NEW AND SPENT

.........o.......os (DELETED)...... ~ ......o.........o...o..ooo...o.......

REFUELING OPERATIONS BORON CONCENTRATION. ~

FUEL HANDLING BUILDING EMERGENCY EXHAUST SYSTEM.

SPECIAL TEST EXCEPTIONS SHUTDOWN REACTOR COOLANT

.

Level..........................................

MARGINo ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

GROUP HEIGHT~ INSERTION~ AND POWER PHYSICS TESTS ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

POSITION INDICATION SYSTEM

- UNIT 1 LOOPS's

~ ~ ~

~

INDEX

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~

ADMINISTRATIVE CONTROLS TO PREVENT DILUTION DURING REFUELINGo ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

COMMUNICATIONS.. ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

REFUELING MACHINE CRANE TRAVEL

...

.

~ ~

~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

FUEL HANDLING BUILDINGo RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION

~ ~ ~ ~

FUEL POOLS..

~ ~ ~ o ~

SHUTDOWN

.

.

~

~ ~ ..

~

~ ~ ~

~

.

..

~ ~ ~ ~

Protection........

~ ~ ~ ~ ~ ~ ~ ~

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

PENETRATIONS ~ ~ ~ ~ ~ ~

.. ...

.

DISTRIBUTION LIMITS~

~ ~

~ ~ ~

~ ~ ~ ~ ~ ~ ~

...

. ........

~

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

.~~~

. ~ . ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ . ~ ...

~ ~ ~ ~ .os

~

.

~ ~ ~ ~

~ ~ ~ ~

~ ~ o ~ ~

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

.~~~~~

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

.~. ~ ~ ~

..os

.

~ ~ ~ ~ ~ ~ ~ ~ ~

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

..o .o

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

~

Amendment No.

~ ~

...

'AGE 3/4 3/4 3/4 3/4 3/4 3/4 3/4 3/4 3/4 3/4 3/4 3/4 3/4 3/4 3/4

'/4 3/4 3/4 3/4 3/4 3/4 8-19 8-21 8-39 8-40 3/4 9"1 9-2 9-3 9-4 9-5 9"6 9-7 9-8 9"9 9-10 9-11 9-12 9"13 3/4 9-14 10-1 10"2 10-3 10-4 10-5

TABLE 2.2-1 (Continued)

TABLE NOTATIONS NOTE 1: OVERTEMPERATURE 4T 4T ( 1 + T S)

(

1

)

AT (Kl - K2 (1 + S) T 1 T' K3 (P - P')

f1 (AI)}

1 + T2S) 1 + T3S

( )

1 + T5S) 1 + T6S Where: Measured AT by RTD Manifold Instrumentation;

~1+v S Lead-lag compensator on measured AT; 1+ T2S Tlt T2 Time constants utilized in lead-lag compensator for ATt Tl = 8 s, T2 = 3 st Lag compensator on measured AT; 1 + T3S T3 Time constants utilized in the lag compensator for ATt T3 = 0 s, Indicated AT at RATED THERMAL POWER; Kl 1.09; K2 0.0182/'F;

~l+ ~ S The function generated by the lead-lag compensator for T 1 + T5S dynamic compensation; avg 4t Time constants utilized in the lead-lag compensator for T , T4 = 20 s, 5

=4s't avg' Average temperature, 'F; 1 Lag compensator on measured T 1+ ASS avg'ime constant utilized in the measured T lag compensator T 6

= 0 s'vg 'I

TABLE 3.3-2 REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES This table is deleted from Technical Specifications.

The information in this table is controlled by plant procedure PLP-106, Technical Specification Equipment List Program.

PAGE 3/4 3-10 HAS BEEN DELETED.

SHEARON HARRIS UNIT 1 3/4 3-9 Amendment No.

TABLE 3.3-5 ENGINEERED SAFETY FEATURES RESPONSE TIMES This table is deleted from Technical Specifications'he information in this table is controlled by plant procedure PLP-106, Technical Specification Equipment List Program.

PAGES 3/4 3-38 THROUGH 3/4 3-40 HAVE BEEN DELETED.

SHEARON HARRIS UNIT 1 3/4 3-37 Amendment No.

'0 INSTRUMENTATION ACCIDENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.6 The accident monitoring instrumentation channels shown in Table 3.3-10 shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3.

ACTION:

a. With the number of OPERABLE accident monitoring instrumentation channels less than the Total Required Number of Channels shown in Table 3 '-10, except for the pressurizer safety valve position indicator or the sub-cooling margin monitor, restore the inoperable channel(s) to OPERABLE status within 7 days,'r be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in at least HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; or

b. With the number of OPERABLE accident monitoring instrumentation channels, except the radiation monitors, the pressurizer safety valve position indicator, or the sub-cooling margin monitor, less than the Minimum Channels OPERABLE requirements of Table 3 '-10, restore the inoperable channel(s) to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in at least HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; or c~ With the number of OPERABLE channels for the radiation monitors, the pressurizer safety valve position indicator*, or the sub-cooling margin monitored, less than required by the Minimum Channels OPERABLE requirements, initiate the preplanned alternate method of monitoring the appropriate parameter(s), within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, and either restore the inoperable channel(s) to -OPERABLE status within 7 days or prepare and submit a Special Report to the Commission, pursuant to Specification 6.9.2, within the next 14 days, that provides actions taken, cause of the inoperability, and the plans and schedule for restoring the channels to OPERABLE status.

d. The provisions of Specification 3.0.4 are not applicable.

• The alternate method shall be a check of safety valve piping temperatures and evaluation to determine position.

f The alternate method shall be the initiation of the backup method as required by Specification 6 '.4.d.

SHEARON HARRIS - UNIT 1 3/4 3-66 Amendment, No.

TABLE 4.3-8 RADIOACTIVE LI UID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE RE UIREMENTS DIGITAL CHANNEL CHANNEL SOURCE CHANNEL OPERATIONAL INSTRUMENT CHECK CHECK CALIBRATION TEST

1. Radioactivity Monitors Providing Alarm and Automatic Termination of Release

a. Liquid Radwaste Effluent Lines

1) Treated Laundry and Hot Shower R(3) q(1)

Tanks Discharge Monitor

2) Waste Monitor Tanks and R(3) q(1)

Wast'e Evaporator Condensate Tanks Discharge Monitor

3) Secondary Waste Sample Tank P, M(5) R(3) q(1)

Discharge Monitor

b. Turbine Building Floor Drains R(3) q(1)

Effluent Line

c. Outdoor Tank Area Drain Transfer R(3) q(1)

Pump Monitor

2. Radioactivity Monitors Providing Alarm But Not Providing Automatic Termination of Release

a. Normal Service Water System Return From the Waste Processing Building to the Circulating Water System R(3) q(2)

REACTOR COOLANT SYSTEM HOT SHUTDOWN LIMITING CONDITION .FOR OPERATION 3.4.1.3 At least two of the loops listed below shall be OPERABLE and at least one of these loops shall be in operation'.<<

a. Reactor Coolant Loop A and its associated steam generator and reactor coolant pump,~'<<'.

Reactor Coolant Loop B and its associated steam generator and reactor coolant pump,**

c. Reactor Coolant Loop C and its associated steam generator and reactor coolant pump

d. RHR Loop A, or

e. RHR Loop B.

APPLICABILITY: MODE 4o ACTION:

a. With less than the above required loops OPERABLE, immediately initiate corrective action to return the required loops to OPERABLE status as soon as possible) if the remaining OPERABLE loop is an RHR loop, be in COLD SHUTDOWN within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

b. With no loop in operation, suspend all operations involving a reduc-tion in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required loop to operation.

+All reactor coolant pumps and RHR pumps may be deenergired for up to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> provided: (1) no operations are permitted that would cause dilution of the Reactor Coolant System boron concentration, and (2) core outlet temperature is maintained at least 10'F below saturation temperature.

~A reactor coolant pump shall not be started with one or more of the Reactor Coolant System cold leg temperatures less than or equal to 335'F unless the secondary water temperature of each steam generator is less than 50'F above each of the Reactor Coolant System cold leg temperatures.

SHEARON HARRIS - UNIT 1 3/4 4"4 Amendment No.

STEAM GENERATOR SURVEILLANCE RE UIREMENTS (Continued) 4.4.5.4 Acce tance Criteria (Continued)

b. The steam generator shall be determined OPERABLE after completing the corresponding actions (plug all tubes exceeding the plugging limit and all tubes containing through-wall. cracks) required by Tables 4.4-2A and B.

4.4.5.5 ~Re orts

a. Within 15 days following the completion of each inservice inspection of steam generator tubes, the number of tubes plugged in each steam generator shall be reported to the Commission in a Special Report pursuant to Specification 6.9.2;

b. The complete results of the steam generator tube inservice inspection shall be submitted to the Commission in a Special Report pursuant to Specification 6.9.2 within 12 months following the completion of the inspection. This Special Report shall include.'.

Number and extent of tubes inspected,

2. Location and percent of wall-thickness penetration for each indication of an imperfection, and 3 ~ Identification of tubes plugged.

c~ Results of steam generator tube inspections which fall into Category C-3 shall be reported in a Special Report pursuant to Specification 6.9.2 within 30 days and prior to resumption of plant operation. This report shall provide a description of investigations conducted to determine cause of the tube degradation and corrective measures taken to prevent recurrence.

SHEARON HARRIS UNIT 1 3/4 4-17 Amendment No.

EV

~h

REACTOR COOLANT SYSTEM REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION 3.4.9.2 The Reactor Coolant System (except the pressurizer) temperature and pressure shall be limited in accordance with the limit lines shown on Figures 3.4-2 and 3.4-3 during heatup, cooldown, criticality, and inservice leak and hydrostatic testing with:

a. A maximum heatup rate as shown on Table 4.4-6.

b. A maximum cooldown rate as shown on Table 4.4"6.

c. A maximum temperature change of less than or equal to 10'F in any 1-hour period during inservice hydrostatic and leak testing opera-tions above the heatup and cooldown limit curves.

APPLICABILITY: MODES 4, 5, and 6 with reactor vessel head on.

ACTTON:

With any of the above limits exceeded, restore the temperature and/or pressure to within the limit within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the Reactor Coolant System; determine that the Reactor Coolant System remains acceptable for continued operation or maintain the RCS T v and pressure at less than 200'F and 500 psig, respectively.

SURVEILLANCE RE UIREMENTS 4.4.9.2.1 The Reactor Coolant System temperature and pressure shall be determined to be within the limits at least once per 30 minutes during system heatup, cooldown, and inservice leak and hydrostatic testing operations.

t 4.4.9.2.2 Deleted from Technical Specifications. Refer to plant procedure PLP-106, Technical Specification Equipment List Program.

SHEARON HARRIS " UNIT 1 3/4 4-34 Amendment No.

TABLE 4 '-5 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM This table is deleted from Technical Specifications.

The information in this table is controlled by plant procedure PLP-106, Technical Specification Equipment List Program.

SHEARON HARRIS - UNIT 1 3/4 4-37 Amendment No.

CONTAINMENT SYSTEMS CONTAINMENT ISOLATION VALVES SURVEILLANCE RE UIREMENTS (Continued) 4.6.3.2 Each isolation valve shall be demonstrated OPERABLE during the COLD SHUTDOWN or REFUELING MODE at least once per 18 months by'.

a. Verifying that on a Phase "A" Isolation test signal, each Phase "A" isolation valve actuates to its isolation position',

b. Verifying that on a Phase "B" Isolation test signal, each Phase "B" isolation valve actuates to its isolation position; and

c. Verifying that on a Containment Ventilation Isolation test signal, each normal, preentry purge makeup and exhaust, and containment vacuum relief valve actuates to its isolation position, and

d. Verifying that, on a Safety Injection "S" test signal, each containment isolation valve receiving an "S" signal actuates to its isolation position, and

e. Verifying that, on a Main Steam Isolation test signal, each main steam isolation valve actuates to its isolation position, and

f. Verifying that, on a Main Feedwater Isolation test signal, each feedwater isolation valve actuates to its isolation position.

4.6.3.3 The isolation time of each power-operated or automatic valve shall be determined to be within its limit when tested pursuant to Specification 4.0.5.

SHEARON HARRIS - UNIT 1 3/4 6-15 Amendment No ~

TABLE 3.6-1 CONTAINHENT ISOLATION VALVES This table is deleted from Technical Specifications.

The information in this table is controlled by plant procedure PLP-106, Technical Specification Equipment List Program.

PAGES 3/4 6-17 THROUGH 3/4 6-29 HAVE BEEN DELETED.

SHEARON HARRIS " UNIT 1 3/4 6-16 Amendment No.

The inservice inspection program for SNUBBERS is deleted from Technical Specifications and is controlled in plant procedure PLP-106, Technical Specification Equipment List Program.

PAGES 3/4 7-21 THROUGH 3/4 7-23 HAVE BEEN DELETED.

SHEARON HARRIS UNIT 1 3/4 7-20 Amendment No.

FIGURE 4.7-1 SAMPLE PLAN(2) FOR SNUBBER FUNCTIONAL TEST This figure is deleted from Technical Specifications and is controlled in plant procedure PLP-106, Technical Specification Equipment List Program.

SHEARON HARRIS UNIT 1 3/4 7-24 Amendment No.

TABLE 3.8-1 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES This table is deleted from Technical Specifications.

The informa'tion in this table is controlled by plant procedure PLP-106, Technical Specification Equipment List Program.

PAGES 3/4 8-22 THROUGH 3/4 8-38B HAVE BEEN DELETED.

SHEARON HARRIS UNIT 1 3/4 8-21 Amendment No.

TABLE 3o8-2 MOTOR"OPERATED VALVES THERMAL OVERLOAD PROTECTION This table is deleted from Technical Specifications.

The information in this table is controlled by plant procedure PLP-106, Technical Specification Equipment List,Program.

PAGES 3/4 8-41 THROUGH 3/4 8-43 HAVE BEEN DELETED.

SHEARON HARRIS UNIT 1 3/4 8-40 Amendment No.

POWER DISTRIBUTION LIMITS BASES UADRANT POWER TILT RATIO (Continued)

For purposes of monitoring QUADRANT POWER TILT RATIO when one excore detector is inoperable, the moveable incore detectors are used to confirm that the normalized symmetric power distribution is consistent with the QUADRANT POWER TILT RATIO. The incore detector monitoring is done with a full incore flux map or two sets of four symmetric thimbles. The preferred sets of four sym-metric thimbles is a unique set of eight detector locations. These locations are C-8, E-5, E-ll, H-3, H-13, L-5, L-ll, N-8. If other locations must be used, a special report to the NRC should be submitted within 30 days in accordance with 10CFR50.4.

3/4.2 ' DNB PARAMETERS The limits on the DNB-related parameters assure that each of the parameters are maintained within the normal steady-state envelope of operation assumed in the transient and accident analyses. The limits are consistent with the ini-tial FSAR assumptions and have been analytically demonstrated adequate to maintain a minimum DNBR of 1.30 throughout each analyzed transient. The indi-cated T value and the indicated pressurizer pressure value are compared to analytical limits of 592.6'F and 2205 psig, respectively, with allowance for measurement uncertainty.

The 12-hour periodic surveillance of these parameters through instrument read" out is sufficient to ensure that the parameters are restored within their limits following load changes and other expected transient operation.

SHEARON HARRIS - UNIT 1 B 3/4 2-6 Amendment No.

ADMINISTRATIVE CO LS 6.9 REPORTING RE UIREMENTS

.ROUTINE REPORTS 6.9.1 In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the, following reports shall be submitted to the NRC in accordance with 10CFR50.4.

STARTUP REPORT 6.9.1.1 A summary report of plant startup and power escalation testing shall be submitted following: (1) receipt of an Operating License, (2) amendment to the license involving a planned increase in power level, (3) installation of fuel that has a different design or has been manufactured by a different fuel supplier, and (4) modifications that may have significantly altered the nuclear, thermal, or hydraulic performance of the unit.

The Startup Report shall address each of:.the tests identified in the Final Safety Analysis Report and shall include a description of the measured values of the operating conditions or characteristics obtained during the test program and a comparison of these values with design predictions and specifications.

Any corrective actions that were required to obtain satisfactory operation shall also be described. Any additional specific details required in license condi-tions based on other commitments shall be included in this report.

Startup Reports shall be submitted within: (1) 90 days following completion of the Startup Test Program, (2) 90 days following resumption or commencement of commercial power operation, or (3) 9 months following initial criticality, whichever is earliest. If the Startup Report does not cover all three events (i.e., initial criticality, completion of Startup Test Program, and resumption or commencement of commercial operation), supplementary reports shall be sub-mitted at least every 3 months'until all three events have been completed.

ANNUAL REPORTS 6.9.1.2 Annual Reports covering the activities of the unit as described below for the previous calendar year shall be submitted prior to March 1 of each year. The initial report shall be submitted prior to March 1 of the year following initial criticality.

Reports required on an annual basis shall include!

a. A tabulation on an annual basis. of the number of station, utility, and other personnel (including contractors) receiving exposures greater than 100 mrem/yr and their associated man-rem exposure according to work and job functions* (e.g., reactor operations and surveillance, inservice inspection, routine maintenance, special maintenance [describe maintenance], waste processing, and refueling).

The dose assignments to various duty functions may be estimated based on pocket dosimeter, thermoluminescent dosimeter (TLD), or

• This tabulation supplements the requirements of f20.407 of 10 CFR Part 20.

SHEARON HARRIS - UNIT 1 6-20 Amendment No.

ADMINISTRATIVE CONTROLS PEAKING FACTOR LIMIT REPORT 6.9.1.6 The W(Z) Functions for RAOC and Base Load operation and the value for APL (as required) shall be established for each reload core and implemented prior to use.

The methodology used to generate the.. W(Z) functions for RAOC and Base Load operation and the value for APL shall be those previously reviewed and approved by the NRC.~ If changes to these methods are deemed necessary, they will be evaluated in accordance with 10 CFR 50.59 and submitted to the NRC for review and approval prior to their use if the change is determined to involve an unreviewed safety question or if such a change would require amendment of previously submitted documentation.

A report containing the W(Z) function for RAOC and Base Load operation and the value for APL (as required) shall be provided to the NRC in accordance with 10 CFR 50.4 within 30 days after each cycle initial criticality.

Any information needed to support W(Z), W(Z)BL, and APLND will be by request from the NRC and need not be included in this report.

SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the NRC in accordance with 10CFR50.4 within the time period specified for each report.

6.10 RECORD RETENTION 6.10.1 In addition to the applicable record retention requirements of, Title 10, Code of Federal Regulations, the following records shall be retained for at least the minimum period indicated.

6.10.2 The following records shall be retained for at least 5 years:

a ~ Records and logs of unit operation covering time interval at each power level;

b. Records and logs of principal maintenance activities, inspections, repair, and replacement of principal items of equipment related to nuclear safety',

C ~ All REPORTABL'E EVENTS; d ~ Records of surveillance activities, inspections, and calibrations required by these Technical Specifications; WCAP-10216, "Relaxation of Constant Axial Offset Control-F~ Surveillance Technical Specification."

SHEARON HARRIS UNIT 1 6-24 Amendment No.