ML18005B116

Amend 15 to License NPF-63,changing Tech Specs to Allow Refueling & Operation W/Vantage 5 Design

ML18005B116
Person / Time
Site:	Harris
Issue date:	10/18/1989
From:	Adensam E Office of Nuclear Reactor Regulation
To:
Shared Package
ML18005B117	List: ML18005B116
References
NPF-63-A-015 NUDOCS 8910250363
Download: ML18005B116 (78)

Text

gp.S <<ECylg

~4

'a s oO t

qO

~+*<<+

UNITED STATES NUCLEAR R EG ULATORY COMMISS ION WASHINGTON, O. C. 20555 CAROLINA POWER

& LIGHT COMPANY et al.

DOCKET NO. 50-400 SHEARON HARRIS NUCLEAR POWER PLANT UNIl' AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.

15 License tIo.

NPF-63 1.

The Nuclear Regulatory COItlnission (the Commission) has found that:

A.

The applicatior for amendment by Carolina Power

& Light Company, (the licensee),

dated April 17,

1989, as supplemented June 29, 1989, July 13, 1989 and October 2, 1989, complies with the standards and requiremerits of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducteo in compliance with the Cowrission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifications, as indicated in the attachment to this license amendment; and paragraph 2.C.(2) of Facility Operating License No.

VPF-63 is hereby amended to read as follows:

8910250363 891018 PDR ADOCK 05000400 P

PNU (2)

Technical S ecifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, and the Environmental Protection Plan contained in Appendix B, both of which are attached

hereto, as revised through Amendment No.

15, are hereby incorporated into this license.

Carolina Power 5 Light Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

This license amendment is effective as of the date of its issuance.

FOR THE NUCLEAR REGULATORY COhINISSION E

Attachment:

Changes to the Technical Specifications Date of Issuance:

October 18, 1989 Original Signed By:

Elinor G. Adensam, Director Project Directorate II-1 Division of Reactor Projects I/II Office of Nuclear Reactor Regulation OFC LA PDH NAhlE :PAnd r PR: PM: PD21: DRPR

)RR W

0 OGC y Chan

0:

21 RPR EA sam

~

WWWWW DATE

~ ro p g

- /y jg n'og OFFICIAL RECORD COPY

7 ~

~

• 4 W ~

V

ATTACHMENT TO LICENSE AMENDMENT NO 15 FACILITY OPERATING LICENSE NO.

NPF-63 DOCKET NO. 50-400 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages.

The revisea areas are indicated by marginal lines.

Remove Pa es Insert Pa es iv iv Xlx 1-2 2-2 2-4 2-5 2-7 2-8 2-9 2-10 B 2-1 B 2-4 B 2-5 B 2-6 xix 1-2 1-2a 2-2 2-4 2-5 2-7 2-8 2-9 2-10 B 2-1 B 2-la B 2-4 B 2-5 B 2-6

Remove Pa es 3/4 1-14 3/4 1-19 3/4 1-20 3/4 1-21 3/4 1-22 3/4 2-1 3/4 2-2 3/4 2-4 3/4 2-5 3/4 2-6 3/4 2-7a 3/4 2-7b 3/4 2-7c 3/4 2-7d 3/4 2-8 3/4 2-9 3/4 2-14 3/4 3-7 3/4 3-34 B 3/4 2-1 B 3/4 2-2 B 3/4 2-2a B 3/4 2-4 B 3/4 2-5 3/4 0/4 3/4 1-19 1-21 3/4 1-22 3/4 2-1 3/4 2-2 3/4 3/P 2-4 2-5 3/4 2-6 3/4 3/4 3/4 3/4 3/4 2 7a 2-'7b 2-7c 2-7d 2-8 3/4 2-9 3/4 2-14 3/4 3

7 3/4 3-34 3/4 3/4 3/4 2-1 2

2-2a B 3/4 2-4 B 3/4 2-5 Insert Pa es 3/4 1-14

Remove Pa es B 3/4 2-6 B 3/4 4-1 6-24 6-24a Inset t Pa es 8 3/4 2-6 B 3/4 4-1 6-24 6-24 a

I Py

~, ~

4A

INDEX 1.0 DEFINITIONS SECTION ACTION ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

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

~ ~ ~

~

~ ~

1.2 ACTUATION LOGIC TESTo ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

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

~ ~ ~

PAGE 1.3 1.4 1.5 1.6 1.7 1.8 ANALOG CHANNEL 'OPERATIONAL TEST AXIAL FLUX DIFFERENCE ~ ~ ~ ~ ~ ~ ~

~ ~ ~

CHANNEL CALIBRATION............

CHANNEL CHECK...." ~.............

CONTAINMENT INTEGRITYo ~ ~ ~

~ '

~ ~ ~

l~

CONTROLLED LEAKAGE~ ~ ~ ~ ~ ~

~ ~ ~ ~ ~ ~ ~

~

~ ~

~

~ ~

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

~

~ ~

1-2 1-2 1.9 1.9a 1.10 1.11 1.12 1.13 1.14 1.15 1.16 1.17 1.18

l. 19 1.20 1.21 1.22 1.23 1.24 1.25 1.26 1.27 1.28 CORE ALTERATION..o.'..ohio..o ~ ooo ~ o..

CORE OPERATING LIMITS REPORT ~ ~ ~ ~ ~ ~

~

DIGITAL CHANNEL OPERATIONAL TEST...

DOSE EQUIVALENT I-131..........

~...

E - AVERAGE DISINTEGRATION ENERGY..

ENGINEERED SAFETY FEATURES

RESPONSE

EXCLUSION AREA BOUNDARY.....

FREQUENCY NOTATION~

~

~ ~ ~ ~ ~ ~ ~ ~

~

~

~ ~

~ ~

~

GASEOUS RADWASTE TREATMENT SYSTEM..

IDENTIFIED LEAKAGE~

~ ~

~ ~ ~ ~ ~ ~ ~ ~ ~

~ ~ ~ ~ ~

MASTER RELAY TEST..............o...

MEMBER(S) OF THE PUBLIC............

OFFSITE DOSE CALCULATION MANUAL....

OPERABLE OPERABILITY.............

OPERATIONAL MODE MODE ~ ~

~

~

~

~

~

~

~

~

~

~

PHYSICS TESTS ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

~

PRESSURE BOUNDARY LEAKAGE..........

PROCESS CONTROL PROGRAM.....

PURGE PURGING'.o..........

QUADRANT POWER TILT RATIO...

RATED THERMAL POWERo

~ ~ ~ ~

~

~ ~ ~

TIME~ ~ ~ ~

~

~ ~

~

~

~

~

~

~

~ o

~ ~

1-2 1-2 1-2 1-2a 1-3 1-3 1-3 1-3 1-3 1-3 1-4 1-4 1-4 1-4 1-4 1-4 1-5 1-5 1-5 1.29 1.30 REACTOR TRIP SYSTEM RESPONSE REPORTABLE EVENT............

SHUTDOWN MARGINo ~ ~ ~ ~ ~ ~

~

~

~

~ ~ ~

TIME ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

~ ~ ~

1-5 1-5 1-5 SHEARON HARRIS UNIT 1 Amendment Noo 15

a 0

e 1

,I

.f

)f r J 4r4 1"

INDEX 3.0/4+0 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE UIREMENTS SECTION PAGE 3/4.0 APPLICABILITY...........~........

~.....

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

~.....

3/4 O-l TABLE 3.1-1 ACCIDENT ANALYSES REQUIRING REEVALUATION IN EVENT OF AN INOPERABLE ROD ~

~ ~ ~ ~ ~ ~

~ ~ ~ ~ o ~ ~ ~ ~ ~ ~

~ ~ ~

THE Position Indication Systems Operating.

~ ~.

~ ~ ~...

~ ~.

~.

~.

~

Positron Indication System - Shutdown....................

V Rod Drop Time.

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

~

~ ~

~

~

~ ~

~

~

Shutdown Rod Insertion Lament.............................

Control Rod Insert@on Limits.............................

FIGURE 3.1-2 (DELETED)............................................

3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1 '

BORATION CONTROL Shutdown Margin - MODES 1 and 2.....................

Shutdown Margin MODES 3, 4, and S.................

FIGURE 3.1-1 SHUTDOWN MARGIN VERSUS RCS BORON CONCENTRATION MODES 3~ 4, AND 5.............

~ ~....

~..

~ ~ ~.....

Moderator Temperature Coefficient.

~........

~...

Minimum Temperature for Criticality... ~ ~.......

3/4.1.2 BORATION SYSTEMS Flow Path Shutdown..............,......................

Flow Paths - Operating...................................

Charging Pump - Shutdown...............

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

Charging Pumps Operating...............................

Borated Water Source Shutdown..........................

Borated Water Sources Operating........................

3/4.1.3 MOVABLE CONTROL ASSEMBLIES Group Height.............................................

3/4 1-1 3/4 1-3 3/4 1-3a 3/4 1-4 3/4 1"6 3/4 1-7 3/4 1-8 3/4 1-9 3/4 1-10 3/4 1-11 3/4 1-12 3/4 1-14 3/4 1-16 3/4 1-17 3/4 1-18 3/4 1-19 3/4 1-20 3/4 1-21 3/4 1-22 SHEARON HARRIS - UNIT 1 1V Amendment No.

J/, ]5

"P

$1

INDEX LIMITING COHDITIONS FOR OPERATION AND SURVEILLANCE RE UIREMEHTS SECTEOM 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR " FQ(Z).

FIGURE 3.2-2 K(Z) - LOCAL AXIAL PENALTY FUNCTIOH FOR FQ(Z).

~..

~ ~.

~

SE HOT CHANNEL 3/4.2.3 RCS FLOW RATE AND NUCLEAR EHTHALPY RI FACTOR ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

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

~ o ~ ~

~

~

3/4.2.4 QUADRANT POWER TILT RATIO. ~ ~.............................

3/4.2.5 NB PARAMETERS...........................................

D 3/4.2.

1 AXIAL FLUX DIFFERENCE....................................

FIGURE 3.2-1 (DELETED)............................................

PAGE 3/4 2-1 3/4 2-4 3/4 2-5 3/4 2-8 3/4 2-9 3/4 2-11 3/4 2-14 3/4.3 1NSTRUMEHTATION 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION......................

3/4 3-1 TABLE 3.3"1 REACTOR TRIP SYSTEM INSTRUMENTATION..................o TABLE 3.3-2 (DELETED)...

TABLE 4.3"1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS..........................................

3/4.3.2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION...................................-..

~...

TABLE 3.3-3 ENGINEERED SAfETY FEATURES ACTUATIOH SYSTEM INSTRUMENTATION.............................,.........

TABLE 3.3-4 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS........................

TABLE 3.3"5 (DELETED).

TABLE 4.3-2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS..........

3/4.3.3 MONITORING INSTRUMENTATION Radiation Monitoring Eor Plant Operations.

~..

~....

~ ~ ~

~

~

~

~

3/4 3-2 3/4 3-9 3/4 3-11 3/4 3-16 3/4 3-18 3/4 3-28 3/4 3-37 3/4 3-41 3/4 3-50 SHEARON HARRIS - UNIT 1

Amendment No. 7,J8,l5

w kll

INDEX ADMINISTRATIVE CONTROLS SECTION PAGE 6.5.3 CORPORATE NUCLEAR SAFETY SECTION Unctlon ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

~ ~

~ ~ ~

~ ~ ~ ~ ~ ~

~ ~

~

~ ~ ~ ~ ~

F 6-11 0rgan1zation................................................

6-11 eV1e>o

~ ~ ~

~

~ ~ o ~

~ ~

~ ~ ~ ~ ~

~

~

~ ~ ~ ~

~ ~ ~ ~

~

~

~

~

~

~

~

~

~ ~

~

~ ~

~

~

~ ~ ~ ~

~ ~ ~

~

~

~

~ ~

R 6-12 Records.............................o.......................

6-13 6.5.4 CORPORATE QUALITY ASSURANCE AUDIT PROGRAM ud 1t s ~ ~

~ ~ ~ ~ ~ ~ ~ ~ ~

~

~ ~

~ ~ ~

~ ~ ~

~ ~

~ ~ ~ ~ ~

~ ~ ~ ~ ~ ~ ~

~ ~ ~

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

~

~

~

~

A 6-14 Records...........................o......o..............o...

6-15 AuthOr1tyoo..o.o.o.....oooo..ooooo...o..ooooo.o.oooo..o

~..o.

6-15 6.5.5 OUTSIDE AGENCY INSPECTION AND AUDIT PROGRAM.......o.........

6-15 6.6 REPORTABLE EVENT ACTION.......................................

6-15 6.7 SAFETY LIMITVIOLATION..............o.............o...........

6-16 6.8 PROCEDURES AND PROGRAMS.......................................

6.9 REPORTING RE UIREMENTS 6.9.1 ROUTINE REPORTS.............................................

Startup Report..............................................

Annual Reports..............................................

Annual Radiological Environmental Operating Report..........

Semiannual Radioactive Effluent Release Report..............

6-16 6-20 6-20 6-20 6-21 6-22 Monthly Operating Reports...................................

6-23 Core Operating L1mits Report.........................

~ ~

~

~

~ ~ ~

6.9.2 SPECIAL REPORTS.............................................

6.10 RECORD RETENTION..........o..................................

6-24 6-24 6-24 6.11 RADIATION PROTECTION PROGRAM.................................

6-26 6.12 HIGH RADIATION AREA.............o.o........."................

6-26 6.13 PROCESS CONTROL PROGRAM (PCP)....-......................

~

. ~

6-27 SHEARON HARRIS UNIT 1 X1X Amendment No.

15

DEFINITIONS CONTAINMENT INTEGRITY 1.7 CONTAINMENT INTEGRITY shall exist when'.

a.

All penetrations required to be closed during accident conditions are either'.

1.

Capable of being closed by an OPERABLE containment automatic isolation valve system, or 2.

Closed by manual valves, blind flanges, or deactivated automatic valves secured in their closed positions, except as provided in Table 3.6-1 of Specification 3.6.3.

b.

All equipment hatches are closed and sealed, c.

Each air lock is in compliance with the requirements of Specification 3.6.1.3, d.

The containment leakage rates are within the limits of Specification 3.6.1.2, and II e.

The sealing mechanism associated with each penetration (e.g.,

welds, bellows, or 0-rings) is OPERABLE.

CONTROLLED LEAKAGE 1.8 CONTROLLED LEAKAGE shall be that seal water flow supplied to the reactor coolant pump seals.

CORE ALTERATION 1.9 CORE ALTERATION shall be the movement or manipulation of any component within the reactor pressure vessel with the vessel head removed and fuel in the vessel.

Suspension of CORE ALTERATION shall not preclude completion of movement of a component to a safe conservative position.

CORE OPERATING LIMITS REPORT 1.9.a The CORE OPERATING LIMITS REPORT is the unit-specific document that provides core operating limits for the current operating reload cycle.

These cycle-specific core operating limits shall be determined for each reload cycle in accordance with Specification 6.9.1.6.

Plant operation within these core operating limits is addressed within the individual specifications'IGITAL CHANNEL OPERATIONAL TEST 1.10 A DIGITAL CHANNEL OPERATIONAL TEST shall consist of exercising the digi-tal computer hardware using data base manipulation to verify OPERABILITY of alarm and/or trip functions'HEARON HARRIS UNIT 1

1-2 Amendment No.

15

P

DEFINITIONS DOSE E UIVALENT I-131 1.11 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microCurie/gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134'nd 1-135 actually present.

The thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-14844, "Calculation of Distance Factors for Power and Test Reactor Sites."

SHEARON HARRIS UNIT 1

1-2a Amendment No.

15

P'.

l' g

iw qPC

668 6se 0

o 628 0

I 6I 6 s~e see

.2

.V

.8 FOMENT tfr action of nominal I FlGURE 2.1-1 REACTOR CORE SAFETY LlhllTS THREE LOOPS IN OPERATlON SHEARON HARRIS " UNIT 1 2-2 Amendment No. /I, 15

TABLE 2.2-1 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT TOTAL ALLOWANCE (TA)

Z SENSOR ERROR (E)

TRIP EETPOINT ALLOWABLE VALUE 1.

Manual Reactor Trip N.A.

N.A.

N.A.

N.A.

N.A.

2.

Power Range, Neutron Flux a.

High Setpoint b.

Low Setpoint 3.

Power Range, Neutron Flux, High Positive Rate 7.5 8.3 1.6 4.56 0

4.56 0

0.5

<109X of RTP~

<25Z of RTP>*

<5Z of RTP~< with a time constant

>2 seconds

<111.1X of RTW*

<27.1X of RTP~

<6.3X of RT&* with a time constant

>2 seconds 4.

Power Range, Neutron Flux, High Negative Rate 5.

Intermediate

Range, Neutron Flux 1.6 17.0 0.5 8.41 0

<5X of RTP~> with a time constant

>2 seconds

<25X of RTW>>

<6.3X of RTP-* with a time constant

>2 seconds

<30.9X of RTP~~

6.

Source

Range, Neutron Flux 7.

Overtemperature AT 8.

Overpower AT 9.

Pressurizer Pressure-Low 10.

Pressurizer Pressure-High 17.0 8.7 4.7 5.0 7.5 10.01 0

<10 cps 1.50 2.21 5.01 1.9 1.5 0.5 See Note 3

>1960 psig

<2385 psig 6.02 Note 5

See Note 1

<1.4 x 10 cps See Note 2

See Note 4

>1946 psig

<2399 psig ll.

Pressurizer Water Level-High 8.0 2.18 1.5

<92X of instru-ment span

<93.8X of instru-ment span

-'-RTP = RATED THERMAL POWER

TABLE 2.2-1 (continued)

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT 12.

Reactor Coolant Flow-Low TOTAL ALLOWANCE (TA) 2.9 1.98 0.6

>90.5X of loop full indicated flow SENSOR ERROR (S)

TRIP SETPOINT ALLOWABLE VALUE

>89.5X of loop full indicated flow 13.

Steam Generator Water Level Low-Low 19.2 18.18 1.5

>38.5X of narrow range instrument span

>38.0X of narrow range instrument span 14.

Steam Generator Water Level-Low Coincident With Steam/Feedwater Flow Mismatch 19.2 20.0 6.68 1.5 3.41 Note 6

>38.5X of narrow range instrument span

<40X of full steam flow at RTP-"-"

>36.8X of narrow range instrument span

<43.1X of full steam flow at RTP"=-'5.

Undervoltage Reactor Coolant Pumps 16.

Underfrequency Reactor Coolant Pumps 17.

Turbine Trip a.

Low Fluid Oil Pressure b.

Turbine Throttle Valve Closure 14.0 5.0 N.A.

N.A.

1.3 3.0 N.A.

N.A.

0.0

>5148 volts 0.0

>57.5 Hz N.A.

>1000 psig N.A.

>1X open

>4920 volts

>57.3 Hz

>950 psig

>1X open 18.

Safety Injection Input from ESF N.A.

N.A.

N.A.

N.A N.A.

-""RTP = RATED THERMAL POWER

TABLE 2.2-1 (Continued)

TABLE NOTATIONS NOTE 1:

OVERTEMPERATURE AT (1 +

~ S) 1 (1+

< S) 1 (1 + x2S) 1 + ~3S (1 + v5S) 1 + v6S Where.'1+

T S

1+ ~2S Tlf T2 Measured AT by RTD Manifold Instrumentation; Lead-lag compensator on measured 4T; Time constants utilized in lead-lag compensator for 4Tf Tl = 8 s, T2 = 3 st 1+ ~3S T3 Kl K2 Lag compensator on measured AT; Time constants utilized in the lag compensator for ATf v3 = 0 s; Indicated AT at RATED THERMAL POWER; 1 ~ 17; 0.0224/'F'l+

v S

1+ t5S T4t T5 1+ v6S The function generated by the lead-lag compensator for Tavg dynamic compensation; Time constants utilized in the lead-lag compensator for Tav f T4 20 s,

=4s 5

avg' Average temperature,

'F; Lag compensator on measured T

avg'ime constant utilized in the measured T

lag compensator, v6 = 0 sf avg

TABLE 2.2-1 (Continued)

TABLE NOTATIONS NOTE 1:

(Continued)

K3 pl

< 588.8'F (Nominal T at RATED THERMAL POWER);

0.001072/psig; Pressurizer

pressure, psig; 2235 psig (Nominal RCS operating pressure);

Laplace transform operator,'

ly and fl (hl) is a function of the indicated difference between top and bottom detectors of the power-range neutron ion chambers>

with gains to be selected based on measured instrument response during plant startup tests such that:

For qt qb between -21.6X and +6.0Z, fl (AI) = 0, where qt and qb are percent RATED THERMAL POWER in the top and bottom halves of the core respectivelyf and qt + qb is total THERMAL POWER in percent of RATED THERMAL POWER; (2)

For each percent that the magnitude of qt qb exceeds

-21.6X, the AT Trip Setpoint shall be automatically reduced by 2.36X of its value at RATED THERMAL POWER; and i

(3)

For each percent that the magnitude of qt qb exceeds

+ 6.0X, the AT Trip Setpoint shall be automatically reduced by 1.57X of its value at RATED THERMAL POWER.

NOTE 2:

The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than 2.1X AT span.

I F

TABLE 2.2-1 (Continued)

TABLE NOTATIONS NOTE 3:

OVERPOWER AT (1+vS)(1)(vS)(1)

(1

+ ASS) (1 + TSS) 4 S

(1 + t)S)

(1

+ TSS)

- f (Al))

(1 + T6S)

Where.'s defined in Note 1,

~1+

v S

1+ T2S T]s T2 1+ T3s T3 As defined in Note 1, As defined in Note 1, As defined in Note 1, As defined in Note 1, As defined in Note 1, Kp 1.079, 0.02/'F for increasing average temperature and 0 for decreasing average temperature, S

1+ T7S T7 1+ T6S The function generated by the rate-lag compensator for T dynamic compensation, Time constants utilized in the rate-lag compensator for T T7 = 10 s, avg's defined in Note 1, As defined in Note 1,

TABLE 2.2-1 (Continued)

TABLE NOTATIONS NOTE 3:

(Continued)

K6 T

0.002/'F for T > T" and K6 = 0 for T < T",

As defined in Note 1, Indicated T

at RATED THERMAL POWER (Calibration temperature for AT instrumentation,

< 588.8'F),

NOTE 4:

S As defined in Note 1, and f2(AI)

=

0 for all hl.

The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than 2.3Z hT span.

NOTE 5:

The sensor error for temperature is 1.9 and 1.1 for pressure.

NOTE 6:

The sensor error for steam flow is 0.9, for feed flow is 1.5, and for steam pressure is 0.75.

t Q

P C

e,4

~ I

'C Qpt

2 '

SAFETY LIMITS BASES 2'.1 REACTOR CORE The restrictions of this safety limit prevent overheating of the fuel and possible cladding perforation which would result in the release of fission products to the reactor coolant.

Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature.

Operation above the upper boundary of the nucleate boiling regime could result in excessive cladding temperatures because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficients DNB is not a directly measurable parameter during operation and therefore THERMAL POWER and Reactor Coolant Temperature and Pressure have been related to DNB.

This relation has been developed to predict the DNB flux and the location of DNB for axially uniform and nonuniform heat flux distributions.

The local DNB heat flux 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 DNB design basis is as follows:

there must be at least a 95 percent probability that the minimum DNBR of the limiting rod during Condition I and II events is greater than or equal to the DNBR limit of the DNB correlation being used (the WRB-1 or WRB-2 correlation in this application).

The correlation DNBR limit is established based on the entire applicable experimental data set such that there is a 95 percent probability with 95 percent confidence that DNB will not occur when the minimum DNBR is at the DNBR limit (1.17 for the WRB-1 or WRB-2 correlation).

In meeting this design basis, uncertainties in plant operating parameters, nuclear and thermal parameters, and fuel fabrication parameters are considered statistically such that there is at least a

95 percent probability with 95 percent confidence level that the minimum DNBR for the limiting rod is greater than or equal to the DNBR limit.

The uncertainties in the above plant parameters are used to determine the plant DNBR uncertainty.

This DNBR uncertainty, combined with the correlation DNBR limit, establishes a design DNBR value which must be met in plant safety analyses using values of input parameters without uncertainties.

In addition, margin has been maintained in the design by meeting safety analysis DNBR limits in performing safety analyses.

The curves of Figure 2.1-1 show the loci of points of THERMAL POWER, Reactor Coolant System pressure and average temperature below which the calculated DNBR is no less than the design DNBR value or the average enthalpy at the vessel exit is less than the enthalpy of saturated liquid.

SHEARON HARRIS - UNIT 1 B 2"1 Amendment No. 7, 15

2.1 SAFETY LIMITS BASES 2.1.1 REACTOR CORE (Continued)

These curves are based on an enthalpy hot channel factor, F

, of 1.62 for 4H LOPAR fuel and 1.65 for VANTAGE 5 fuel and a reference cosine with a peak of 1.55 for axial power shape.

An allowance is included for an increase in calculated F4H at reduced power based on the expression:

F4H = 1.62

[1 + 0.3 (1-P)] for LOPAR fuel, and F4H = 1.65

[1 + 0.35 (1-P)] for VANTAGE 5 fuel Where P is the fraction of RATED THERMAL POWER.

These limiting heat flux conditions are higher than those calculated for the range of all control rods fully withdrawn to the maximum allowable control rod insertion assuming the axial power imbalance is within the Limits of the fl (4I) function of the Overtemperature trip.

When the axial power imbalance is not within the tolerance, the axial power imbalance effect on the Over-temperature 4T trips will reduce the Setpoints to provide protection consistent with core Safety Limits.

SHEARON HARRIS - UNIT 1

B 2-la Amendment No.

15

E W ~

llll l4

~-

~4 5

\\k IP

LIMITING SAFETY SYSTEM SETTINGS BASES Power Ran e, Neutron Flux (Continued)

The Low Setpoint trip may be manually blocked above P-10 (a power level of approximately 10X of RATED THERMAL POWER) and is automatically reinstated below the P-10 Setpoint.

Power Ran e, Neutron Flux Hi h Rates The Power Range Positive Rate trip provides protection against rapid flux increases which are characteristic of a rupture of a control rod drive housing.

Specifically, this trip complements the Power Range Neutron Flux High and Low trips to ensure that the criteria are met for rod ejection from mid-power.

The Power Range Negative Rate trip provides protection for control rod drop accidents.

At high power a single or multiple rod drop accident could cause local flux peaking which could cause an unconservative local DNBR to exist.

The Power Range Negative Rate trip will prevent this from occurring by tripping the reactor.

No credit is taken for operation of the Power Range Negative Rate trip for those control rod drop accidents for which DNBRs will be greater than the design DNBR value.

Intermediate and Source Ran e

Neutron Flux The Intermediate and Source

Range, Neutron Flux trips provide core protection during reactor startup to mitigate the consequences of an uncontrolled rod cluster control assembly bank withdrawal from a subcritical condition.

These trips provide redundant protection to the Low Setpoint trip of the Power

Range, Neutron flux channels.

The Source Range channels will initiate a Reactor trip at about 10 counts per second unless manually blocked when P-6 becomes active.

The Intermediate Range channels will initiate a Reactor trip at a current level equivalent to approximately 25X of RATED THERMAL POWER unless manually blocked when P-10 becomes active.

Overtem erature 4T The Overtemperature 4T trip provides core protection to prevent DNB for all combinations of pressure, power, coolant temperature, and axial power distribu" tion, provided that the transient is slow with respect to piping transit delays from the core to the temperature detectors (about 4 seconds),

and pressure is within the range between the Pressurizer High and Low Pressure trips.

The Set-point is automatically varied with:

(1) coolant temperature to correct for temperature induced changes in density and heat capacity of water and includes dynamic compensation for piping delays from the core to the loop temperature detectors

, (2) pressurizer

pressure, and (3) axial power distribution.

With normal axial power distribution, this Reactor trip limit is always below the core Safety Limit as shown in Figure 2.1-1.

If axial peaks are greater than

design, as indicated by the difference between top and bottom power range nuclear detectors, the Reactor trip is automatically reduced according to the notations in Table 2.2-1.

SHEARON HARRIS - UNIT 1 B 2-4 Amendment No.

15

LIMITING SAFETY SYSTEM SETTINGS BASES Over ower 4T The Overpower 4T trip provides assurance of fuel integrity (e.g.,

no fuel pellet melting and less than 1X cladding strain) under all possible overpower conditions, limits the required range for Overtemperature 4T trip, and provides a backup to the High Neutron Flux trip.

The Setpoint is automatically varied with:

(1) coolant temperature to correct for temperature induced changes in density and heat capacity of water, and (2) rate of change of temperature for dynamic compensation for piping delays from the core to the loop temperature detectors, to ensure that the allowable heat generation rate (kW/ft) is not exceeded.

The Overpower 4T trip provides protection to mitigate the consequences of various size steam breaks as reported in WCAP-9226, "Reactor Core Response to Excessive Secondary Steam Releases."

Pressurizer Pressure In each of the pressurizer pressure

channels, there are two independent bistables, each with its own trip setting to provide for a High and Low Pressure trip thus limiting the pressure range in which reactor operation is permitted.

The Low Setpoint trip protects against low pressure which could lead to DNB by tripping the reactor in the event of a loss of reactor coolant pressure.

On decreasing power the Low Setpoint trip is automatically blocked by the loss of P-7 (a power level of approximately 10X of RATED THERMAL POWER or turbine impulse chamber pressure at approximately 10X of full power equivalent);

and on increasing power, automatically reinstated by P-7.

The High Setpoint trip functions in conjunction with the pressurizer relief and safety valves to protect the Reactor Coolant System against system overpressure.

Pressurizer Water Level The Pressurizer High Water Level trip is provided to prevent water relief through the pressurizer safety valves.

On decreasing power the Pressurizer High Water Level trip is automatically blocked by the loss of P-7 (a power level of approximately 10X of RATED THERMAL POWER or a turbine impulse chamber pressure at approximately 10X of full power equivalent);

and on increasing power, automatically reinstated by P-7

~

Reactor Coolant Flow The Reactor Coolant Low Flow trips provide core protection to prevent DNB by mitigating the consequences of a loss of flow resulting from the loss of one or more reactor coolant pumps.

On increasing power above P-7 (a power level of approximately 10X of RATED THERMAL POWER or a turbine impulse chamber pressure at approximately 10X of full power equivalent),

an automatic Reactor trip will occur if the flow in more than one loop drops below 90.5X of nominal full loop flow.

Above P-8 SHEARON HARRIS UNIT 1

B 2-5 Amendment No.

15

LIMITING SAFETY SYST SETTINGS BASES Reactor Coolant Flow (Continued)

(a power level of approximately 49X of RATED THERMAL POWER) an automatic Reactor trip will occur if the flow in any single loop drops below 90.5X of nominal full loop flow.

Conversely, on decreasing power between P-8 and the P"7 an-automatic Reactor trip will occur on low reactor coolant flow in more than one loop and below P-7 the trip function is automatically blocked.

Steam Generator Water Level The Steam Generator Water Level Low-Low trip protects the reactor from loss of heat sink in the event of a sustained steam/feedwater flow mismatch resulting from loss of normal feedwater.

The specified Setpoint provides allowances for starting delays of the Auxiliary Feedwater System.

Steam/Feedwater Flow Mismatch and Low Steam Generator Water Level The Steam/Feedwater Flow Mismatch in coincidence with a Steam Generator Low Water Level trip is not used in the transient and accident analyses but is included in Table 2.2-1 to ensure the functional capability of the specified trip settings and thereby enhance the overall reliability of the Reactor Trip System.

This trip is redundant to the Steam Generator Water Level Low-Low trip.

The Steam/Feedwater Flow Mismatch portion of this trip is activated when the s)earn flow exceeds the feedwater flow by greater than or equal to 1.627 x 10 lbs/hour.

The Steam Generator Low Water level portion of the trip is activated when the water level drops, below 38.5X, as indicated by the narrow range instrument.

These trip values include sufficient allowance in excess of normal operating values to preclude spurious trips but will initiate a Reactor trip before the steam generators are dry.

Therefore, the required capacity and starting time requirements of the auxiliary feedwater pumps are reduced and the resulting thermal transient on the Reactor Coolant System and steam generators is minimized.

Undervolta e and Underfre uenc Reactor Coolant Pum Buses The Undervoltage and Underfrequency Reactor Coolant Pump Bus trips provide core protection against DNB as a result of complete loss of forced coolant flow.

The specified Setpoints assure a Reactor trip signal is generated before the Low Flow Trip Setpoint is reached.

Time delays are incorporated in the Underfrequency and Undervoltage trips to prevent spurious Reactor trips from momentary electrical power transients.

For undervoltage, the delay is set so that the time required for a signal to reach the Reactor trip breakers following the simultaneous trip of two or more reactor coolant pump bus circuit breakers shall not exceed 1.2 seconds.

For underfrequency, the delay is set so that the time required for a signal to reach the Reactor trip breakers after the Underfrequency Trip Setpoint is reached shall not exceed 0.3 second.

On decreasing power the Undervoltage and Underfrequency Reactor Coolant Pump Bus trips are automatically blocked by the loss of P-7 (a power level of approximately 10% of RATED THERMAL POWER or a turbine impulse chamber pressure SHEARON HARRIS UNIT 1 B 2-6 Amendment No.

15

REACTIVITY CONTROL SYSTEMS 3/4.1.3 MOVABLE CONTROL ASSEMBLIES

. GROUP HEIGHT LIMITING CONDITION FOR OPERATION 3.1.3.1 All shutdown and control rods shall be OPERABLE and positioned within

+

12 steps (indicated position) of their group step counter demand position.

APPLICABILITY:

MODES 1

and 2

ACTION:

a ~

With one or more rods inoperable due to being immovable as a result of excessive friction or mechanical interference or known to be untrippable, determine that the SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is satisfied within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and be in HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

b.

With more than one rod misaligned from the group step counter demand position by more than

+

12 steps (indicated position),

be in HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

C ~

With more than one rod inoperable, due to a rod control urgent failure alarm or obvious electricaL problem in the rod control system existing for greater than 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />, be in HOT STANDBY within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

d ~

With one rod trippable but inoperable due to causes other than addressed by ACTION a.,

above, or misaligned from its group step counter demand height by more than

+

12 steps (indicated position),

POWER OPERATION may continue provided that within 1

hour.'.

The rod is restored to OPERABLE status within the above alignment requirements, or 2.

The rod is declared inoperable and the remainder of the rods in the group with the inoperable rod are aligned to within +

12 steps of the inoperable rod while maintaining the rod sequence and insertion limits on control banks specified in the Core Operating Limits Report.

The THERMAL POWER level shall be restricted pursuant to Specification 3.1.3.6 during subsequent operation, or 3.

The rod is declared inoperable and the SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is satisfied.

POWER OPERATION may then continue provided that:

a)

A reevaluation of each accident analysis of Table 3.1-1 is performed within 5 days; this reevaluation, shall confirm that the previously analyzed results of these accidents See Special Test Exceptions Specifications 3.10.2 and 3.10.3.

SHEARON HARRIS - UNIT 1

3/4 1-14 Amendment No. g, 15

t

'E)

~

~

REACTIVITY CONTROL SYSTEMS ROD DROP TIME LIMITING CONDITION FOR OPERATION 3.1.3.4 The.individual shutdown and control rod drop time from the fully withdrawn position shall be less than or equal to 2.7 seconds from beginning of decay of stationary gripper coil voltage to dashpot entry with:

a.

T greater than or equal to S51'F, and b.

All reactor coolant pumps operating.

APPLICABILITY:

MODES 1 and 2.

ACTION0 With the drop time of any rod determined to exceed the above limit, restore the rod drop time to within the above limit prior to proceed-ing to MODE 1 or 2.

b.

With the rod drop times within limits but determined with two reactor coolant pumps operating, operation may proceed provided THERMAL POWER is restricted to less than or equal to 66X of RATED THERMAL POWER.

SURVEILLANCE RE UIREMENTS 4.1.3.4 The rod drop time of shutdown and control rods shall be demonstrated through measurement prior to reactor criticality.'

~

For all rods following each removal of the reactor vessel

head, b.

For specifically affected individual rods following any maintenance on or modification to the'Control Rod Drive System which could affect the drop time of those specific rods, and C ~

At least once per 18 months.

SHEARON HARRIS - UNIT 1 3/4 1-19 Amendment No. 15

REACTIVITY CONTROL SYSTEMS SHUTDOWN ROD INSERTION LIMIT LIMITING CONDITION FOR OPERATION 3.1.3.5 All shutdown rods shall be fully withdrawn as specified in the Core Operating Limits Report.

APPLICABILITY:

MODES l~ and 2+

ACTION:

I With a maximum of one shutdown rod not fully withdrawn, except for surveillance testing pursuant to Specification 4.1.3.1.2, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> either:

a.

Fully withdraw the rod, or b.

Declare the rod to be inoperable and apply Specification 3.1.3.1.

SURVEILLANCE RE UIREMENTS 4.1.3.5 Each shutdown rod shall be determined to be fully withdrawn:

a ~

Within 15 minutes prior to withdrawal of any rods in Control Bank A, B, C, or D during an approach to reactor criticality, and b.

At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter.

>See Special Test Exceptions Specifications 3.10.2 and 3.10.3.

~~With K ff greater than or equal to l.

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

15

e-

REACTIVITY CONTROL SYSTEMS CONTROL ROD INSERTION LIMITS LIMITING CONDITION FOR OPERATION 3.1.3.6 The control banks shall be limited in physical insertion as specified in the Core Operating Limits Report.

APPLICABILITY:

MODES 1"- and 2" ACTION'ith the control banks inserted beyond the above insertion limits, except for surveillance testing pursuant to Specification 4.1.3.1.2:

a ~

Restore the control banks to within the limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, or b.

Reduce THERMAL POWER within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to less than or equal to that fraction of RATED THERMAL POWER which is allowed by the bank posi-tion using the insertion limits specified in the Core Operating Limits Report, or C ~

Be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

SURVEILLANCE RE UIREMENTS 4.1.3.6 The position of each control bank shall be determined to be within the insertion limits at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> except during time intervals when the rod insertion limit monitor is inoperable, then verify the individual rod positions at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

-"See Special Test Exceptions Specifications 3.10.2 and 3.10.3.

"-With K ff greater than or equal to 1.

SHEARON HARRIS UNIT 1

3/4 1-21 Amendment No. 7, 15

FIGURE 3.1-2 DELETED.

SHEARON HARRIS - UNIT 1 3/e 1-22 Amendment No. jI, 15

I't

~

T ql

J 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2. 1 AXIAL FLUX DIFFERENCE LIMITING CONDITION FOR OPERATION 3.2.1 The indicated AXIAL FLUX DIFFERENCE (AFD) shall be maintained within:

a.

the acceptable operational space as specified in the Core Operating Limits Report for Relaxed Axial Offset Control (RAOC) operation, or b.

within a band about the target AFD during Base Load operation as specified in the Core Operating Limits Report.

APPLICABILITY:

MODE 1 above 50% of RATED THERMAL POWER ACTION:

a ~

b.

1 For RAOC operation with the indicated AFD outside of the Limits specified in the Core Operating Limits Report, either:

1.

Restore the indicated AFD to within the limits specified in the Core Operating Limits Report within 15 minutes, or 2.

Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 30 minutes and reduce the Po~er Range Neutron Flux High Trip setpoints to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

N Dcc>':

For Base Load operation above APL 'ith the indicated AXIAL FLUX DIFFERENCE outside of the applicable target band about the target AFD, either:

1.

Restore the indicated AFD to within the target band limits within 15 minutes, or C ~

2.

Reduce THERMAL POWER to less than APL of RATED THERMAL POWER and discontinue Base Load operation within 30 minutes.

THERMAL POWER shall not be increased above 50% of RATED THERMAL POWER unless the indicated AFD is within the limits specified in the Core Operating Limits Report for RAOC operation.

See Special Test Exception 3.10.2

>'-::APL is the minimum allowable power level for Base Load operation and will be provided in the Core Operating Limits Report per Specification 6.9.1.6.

SHEARON HARRIS UNIT 1

3/4 2-1 Amendment No. g, 15

POWER DISTRIBUTION LIMITS SURVEILLANCE RE UIREMENTS 4.2.1.1 The indicated AFD shall be determined to be within its limits during POWER OPERATION above 50X of RATED THERMAL POWER by:

a.

Monitoring the indicated AFD for each OPERABLE excore channel at least once per 7 days when the AFD Monitor Alarm is OPERABLE, and b.

Monitoring and logging the indicated AFD for each OPERABLE excore channel at least once per hour for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and at least once per 30 minutes thereafter, when the AFD Monitor Alarm is inoperable.

The logged values of the indicated AFD shall be assumed to exist during the interval preceding each logging.

4.2.1.2 The indicated AFD shall be considered outside of its limits when two or more OPERABLE excore channels are indicating the AFD to be outside the limits.

4.2.1.3 When in Base Load operation, the target AFD of each OPERABLE excore channel shall be determined by measurement at least once per 92 Effective Full Power Days.

The provisions of Specification 4.0.4 are not applicable.

4.2.1.4 When in Base Load operation, the target AFD shall be updated at least once per 31 Effective Full Power Days by either determining the target AFD in conjunction with the surveillance requirements of Specification 4.2.1.3 above or by linear interpolation between the most recently measured value and the calculated value at the end of cycle life.

The provisions of Specification 4.0.4 are not applicable.

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

FIGURE 3. 2-1 DELETED SHEARON HARRIS - UNIT 1

3/4 2-4 Amendment No. g, l5

l 4

POWER DISTRIBUTION LIMITS 3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR F (Z)

LIMITING CONDITION FOR OPERAT10N 3.2.2 F~(Z) shall be Limited by the following relationships:

FQ(Z)

< 2.45 [K(Z)] FOR P

> 0.5 P

Fq(Z)

< (4.90) fK(Z) ]

FOR P < 0.5 Where:

P

= THERMAL POWER

, and RATED THERMAL POWER K(Z) = the function obtained from Figure 3.2"2 for a given core height location.

APPLICABILITY:

MODE 1.

ACTION:

With F~(Z) exceeding its limit:

a

~

b.

Reduce THERMAL POWER at least 1X for each 1X F~(Z) exceeds the limit within 15 minutes and similarly reduce the Power Range Neutron Flux-High Trip Setpoints within the next 4

hours; POWER OPERATION may proceed for up to a total of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />; subsequent POWER OPERATION may proceed provided the Overpower AT Trip Setpoints have been reduced at least 1%

for each 1X F~(Z) exceeds the limit.

Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER above the reduced limit re-quired by ACTION a.,

above; THERMAL POWER may then be increased provided F~(Z) is demonstrated through incore mapping to be within its limits SHEARON HARRIS - UNIT 1 3/4 2-5 Amendment No. g, 15

I I

POWER DISTRIBUTION LIMITS SURVEILLANCE RE UIREMENTS 4.2.2.1 The provisions of Specification 4.0.4 are not applicable.

4.2.2.2 For RAOC operation, FQ(Z) shall be evaluated to determine if it is within its limit by:

a.

Using the movable incore detectors to obtain a power distribution map at any THERMAL POWER greater than 5X of RATED THERMAL POWER.

b.

Increasing the measured F<(Z) component of the power distribution map by 3X to account for ihanufacturing tolerances and further increasing the value by 5X to account for measurement uncertainties.

Verify the requirements of Specification 3.2.2 are satisfied.

c ~

Satisfying the following relationship'.

FQ (Z) ( 2.45' K(Z) for P ) 0.5 FQ (Z) ( 2.45 x K(Z) for P ( 0.5 W(Z) x 0.5 where F (Z) is the measured F (Z) increased by the allowances for manufacturing tolerances and measurement uncertainty, 2.45 is the FQ limit, K(Z) is given in Figure 3.2-2, P is the fractio'n of RATED THERMAL POWER, and W(Z) is the cycle dependent function that accounts for power distribution transients encountered during normal operation.

This function is given in the Core Operating Limits Report as per Specification 6.9.1.6.

d.

Measuring F

(Z) according to the following schedule:

Q 1 ~

Upon achieving equilibrium conditions after exceeding by 10X or more of RATED THERMAL POWER, the THERMAL POWER at which FQ(Z) was last determined,~

or 2.

At Least once per 31 Effective Full Power Days, whichever occurs first.

""During power escalation at the beginning of each cycle, power level may be increased until a power level for extended operation has been achieved and a power distribution map obtained.

SHEARON HARRIS - UNIT 1 3/4 2-6 Amendment No. 7, 15

, POWER DISTRIBUTION LIMITS

.SURVEILLANCE RE UIREMENTS (Continued) e.

With measurements indicating FM maximum q (z)

KKKz has increased since the previous determination of F~ (Z) either of the following actions shall be taken:

1)

F~ (Z) shall be increased by 2Z over that specified in Specification 4.2.2.2c, or 2)

F~ (Z) shall be measured at least once per 7 Effective Full Power Days until two successive maps indicate that maximum FM9 (Z) is not increasing.

KKKz f.

With the relationships specified in Specification 4.2.2.2c above not being satisfied:

1)

Calculate the percent F~(Z) exceeds its limit by the following expression:

F"(Z) x W(Z) maximum maximum 2.45 K(Z)

F (Z) x W(Z) x K(Z) 2.45 0.5 x

100 for P ? 0.5 x

100 for P < 0.5 2)

One of the following actions shall be taken.')

Within 15 minutes, control the AFD to within new AFD limits which are determined by reducing the AFD limits specified in the Core Operating Limits Report by 1X AFD for each percent F~(Z) exceeds its limits as determined in Specification 4.2.2.2f.l).

Within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, reset the AFD alarm setpoints to these modified limits, or b)

Comply with the requirements of Specification 3.2.2 for F (Z) exceeding its limit by the percent calculated

above, 0

c)

Verify that the requirements of Specification 4 2.2.3 for Base Load operation are satisfied and enter Base Load operation.

SHEARON HARRIS UNIT 1 3/4 2-7a Amendment No. 7, 15

~

h*

ICI*

hI

POWER DISTRIBUTION LIMITS SURVEILLANCE RE UIREMENTS (Continued) go The limits specified in Specifications 4.2.2.2c, 4.2.2.2e, and 4.2.2.2f above are not applicable in the following core plane regions'.

1.

Lower core region from 0 to 15X, inclusive.

2.

Upper core region from 85 to 100X, inclusive.

4.2.2.3 Base Load operation is permitted at powers above APL if the following conditions are satisfied:

a.

Prior to entering Base Load operation, maintain THERMAL POWER above APL and less than or equal to that allowed by Specification 4.2.2.2 for at least the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Maintain Base Load operation surveillance (AFD within the limits specified in the Core Operating Limits Report) during this time period.

Base Load operation is then permitted

.providing THERMAL POWER is maintained between APL and APL or between APL and 100X (whichever is most limiting) and FQ surveillance is maintained pursuant to Specification 4.2.2.4.

APL is defined as:

BL APLBL

(

2.45 x

K(Z)

]

100X F (Z) x W(Z) where:

FQ(Z) is the measured FQ(Z) increased by the allowances for M

manufacturing tolerances and measurement uncertainty.

The FQ limit is 2.45.

K(Z) is given in Figure 3.2-2.

W(Z)BL is the cycle dependent function that accounts for limited power distribution transients encountered during Base Load operation.

The function is given in the Core Operating Limits Report as per Specification 6.9.1.6.

b.

Durjgg Base Load operation, if the THERMAL POWER is decreased below APL then the conditions of 4.2.2.3.a shall be satisfied before re-entering Base Load operation.

4.2.2.4 During Base Load operation FQ(Z) shall be evaluated to determine if it is within its limit by:

a ~

Using the movable incore detectors

)o obtain a power distribution map at any THERMAL POWER above APL b.

Increasing the measured FQ(Z) component of the power distribution map by 3X to account for manufacturing tolerances and further increasing the value by 5% to account for measurement uncertainties.

Verify the requirements of Specification 3.2.2 are satisfied.

SHEARON HARRIS UNIT 1 3/4 2-7b Amendment No. 1 ~

G'l yt

'E, S<

~ I

POWER DISTRIBUTION LIMITS SURVEILLANCE RE UIREMENTS (Continued)

C ~

Satisfying the following relationship'.

FM(Z)

(

2.45 x

K(Z) f P ) APLND Q

P x

W(Z)BL where:

FQ(Z) is the measured FQ(Z).

The FQ Limit is 2.45.

M K(Z) is given in Figure 3.2-2.

P is the fraction of RATED THERMAL POWER.

W(Z)BL is the cycle dependent function that accounts for limited power distribution transients encountered during normal operation.

This function is given in the Core Operating Limits Report as per Specification 6.9.1.6.

d.

Measuring F (Z) in conjunction with target flux difference Q

determination according to the following schedule.'.

Prior to entering Base Load operation after satisfying Section 4.2.2.3 unless a full core flux map has been taken in the previous 31 EFPD with the relative thermal power having been maintained above APL for the 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to mapping, ND and 2.

At Least once per 31 effective fuLL power days.

e.

With measurements indicating F (Z) maximum [~i K(Z) has increased since the previous determination FQ(Z) either of the following actions shall be taken:

1.

FQ(Z) shall be increased by 2 percent over that specified in M

4.2.2.4.c, or 2.

FQ(Z) shall be measured at least once per 7

EFPD until 2

M successive maps indicate that F

(Z) maximum

[

K(Z)

)

is not increasing.

f.

With the relationship specified in 4.2.2.4.c above not being satisfied, either of the following actions shaLL be taken.

1.

Place the core in an equilibrium condition where the Limit in 4.2.2.2.c is satisfied, and remeasure FQ(Z)

, or M

SHEARON HARRIS - UNIT 1

3/4 2-7c Amendment Ho. 7, 15

POWER DISTRIBUTION LIMITS SURVEILLANCE RE UIREMENTS (Continued) 2.

Comply with the requirements of Specification 3.2.2 for F (Z) exceeding its limit by the percent calculated with the following expression:

F (Z) x W(Z)

[(max. of

(

]

) -1] x 100 for P > APL x

K(Z) 2.45 P

g.

The limits specified in 4.2.2.4.c, 4.2.2.4.e, and 4.2.2.4.f above are not applicable in the following core plane regions:

1.

Lower core region 0 to 15 percent, inclusive.

2 ~

Upper core region 85 to 100 percent, inclusive.

4.2.2.5 When F~(Z) is measured for reasons other than meeting the requirements of Specification 4.2.2.2 an overall measured F (Z) shall be obtained from a.power distribution map and increased by 3X 3o account for manufacturing tolerances and further increased by 5X to account for measurement uncertainty.

SHEARON HARRIS UNIT 1 3/4 2-7d Amendment No. 7, 15

1A O.O O,1 flevatfon Norlaltzed Pea%in Factor 0.0

%.0 12.0

).0 l.o 0.925 O.)

5 b

COfK ElEVAllOH(FEKl)

FIGURF 3.2-2 K(Z) LOCAL AXIAL PENALTY FUNCTION FOR F

(Z)

POWER DISTRIBUTION LIMITS 3/4.2.3 RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR LIMITING CONDITION FOR OPERATION 3.2.3 The indicated Reactor Coolant System (RCS) total flow rate and F<H shall be maintained as follows:

a.

Measured RCS flow rate ) 293,540 gpm x (1.0

+ Cl), and b.

FAH < 1.62 [1.0 + 0.3(l.O-P)] for LOPAR fuel, and F<H < 1.65 [1.0 + 0 '5(1.0-P)] for VANTAGE 5 fuel.

Where.'

= THERMAL POWER

, and RATED THERMAL POWER FAH =

Nuclear enthalpy rise hot channel factor obtained by using the movable incore detectors to obtain a power distribution map, with the measured value of the nuclear enthalpy rise hot N

channel factor (F<H) increased by an allowance of 4Z to account for measurement uncertainty.

Cl

=

Measurement uncertainty for core flow as described in the Bases.

APPLICABILITY:

MODE 1.

ACTION With RCS total flow rate or F<H outside the above limits:

a.

Within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> either'.

Restore RCS total flow rate and F<H to within the above limits, or 2 ~

Reduce THERMAL POWER to.less than 50Z of RATED THERMAL POWER and reduce the Power Range Neutron Flux " High Trip Setpoint to less than or equal to 55Z of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

SHEARON HARRIS UNIf 1

3/4 2-9 Amendment No. jF, 15

L g(

POWER DISTRIBUTION LIMITS 3/4.2 '

DNB PARAMETERS LIMITING CONDITION FOR OPERATION 3.2.5 The following DNB-related parameters shall be maintained within the following limits.'.

Indicated Reactor Coolant System T

594.1'F after addition for instrument uncertainty, and b.

Indicated Pressurizer Pressure

> 2185 psig* after subtraction for instrument uncertainty.

APPLICABILITY:

MODE 1.

ACTION:

With any of the above parameters exceeding its indicated limit, restore the parameter to within its limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to less than 5X of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

SURVEILLANCE RE UIREMENTS 4.2.5 Each of the parameters shown in Specification 3.2.5 shall be verified to be within its limit at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

"'This limit is not applicable during either a Thermal Power Ramp in excess of

+5X Rated Thermal Power per minute or a Thermal Power step change in excess of +10X Rated Thermal Power.

SHEARON HARRIS UNIT 1 3/4 2-14 Amendment No.

15

C'

TABLE 3.3-1 (Continued)

ACTION STATEMENTS (Continued)

ACTION 3 With the number of channels OPERABLE one less than the Minimum Channels OPERABLE requirement and with the THERMAL POWER level:

a ~

Below the P-6 (Intermediate Range Neutron Flux Interlock)

Setpoint, restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above the P-6 Setpoint, and b.

Above the P-6 (Intermediate Range Neutron Flux Interlock)

Setpoint but below 10Z of RATED THERMAL POWER, restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above 10Z of RATED THERMAL POWER.

ACTION 4 With the number of OPERABLE channeLs one less than the Minimum Channels OPERABLE requirement, suspend all operations involving positive reactivity changes.

ACTION 5 a.

With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoper-able channel'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 open the Reactor Trip System breakers, and verify compliance with the shutdown margin requirements of Specification 3.1.1.2 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter.

b.

With no channels OPERABLE, open the Reactor Trip System breakers within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and suspend all operations involving positive reactivity changes.

Verify compliance with the SHUTDOWN MARGIN requirements of Specification 3.1.1.2 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter.

ACTION 6 With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:

a.

The inoperable channel. is placed in the tripped condition within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and b.

The Minimum Channels OPERABLE requirement is met; however, the inoperable channel may be bypassed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for surveillance testing of other channels per Specification 4.3.1.1.

ACTION 7 - With less than the Minimum Number of Channels OPERABLE, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> determine by observation of the associated permissive annunciator window(s) that the interlock is in its required state for the existing plant condition, or apply Specification 3.0.3.

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

15

TABLE 3. 3-4

( Cont inued )

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT 9.

Loss-of-Offsite Power TOTAL ALLOWANCE (TA)

SENSOR ERROR (s)

TRIP SETPOINT ALLOWABLE VALUE a.

6.9 kV Emergency Bus Undervoltage-"Primary N.A.

N.A.

N.A.

> 4830 volts with a < 1.0 second time delay.

> 4692 vol'ts wit~

a time delay

< 1.5 seconds b.

6.9 kV Emergency Bus Undervol tage-Secondary 10.

Engineered Safety Features Actuation System Interlocks N.A.

N.A.

N.A.

> 6420 volts with a 16 second time delay (with Safety Injection).

> 6420 volts with a

< 54.0 second time delay (with-out Safety Injection).

> 6392 volts with a time delay

< 18 seconds (with Safety Injection).

> 6392 volts with a

< 60 second time delay (with-out Safety Injection).

a.

Pressurizer

Pressure, P-11 Not P-ll N.A.

N.A.

N.A N.A.

N.A.

N.A.

> 2000 psig

< 2000 psig

> 1986 psig

< 2014 psig 0

b.

Low-LowT, P-12 N.A.

N.A.

N.A.

> 553 F

> 549.3'F

~

'~

>r

3/4.2 POWER DISTRIBUTION LIMITS BASES The specifications of this section provide assurance of fuel integrity during Condition I (Normal Operation) and II (Incidents of Moderate Frequency) events by:

(1) maintaining the minimum DNBR in the core greater than or equal to the design DNBR value during normal operation and in short-term transients, and (2) limiting the fission gas release, fuel pellet temperature, and cladding mechanical properties to within assumed design criteri'a.

In addition, limiting the peak linear power density during Condition I events provides assurance that the initial conditions assumed for the LOCA analyses are met and the ECCS acceptance criteria limit of 2200'F is not exceeded.

The definitions of certain hot channel and peaking factors as used in these specifications are as follows:

Fq(Z)

N FAH Heat Flux Hot Channel Factor, is defined as the maximum local heat flux on the surface of a fuel rod at core elevation Z divided by the average fuel rod heat flux, allowing for manufacturing tolerances on fuel pellets and rods; Nuclear Enthalpy Rise Hot Channel Factor, is defined as the ratio of the integral of linear power along the rod with the highest integrated power to the average rod power',

3/4.2.1 AXIAL FLUX DIFFERENCE The limits on AXIAL FLUX DIFFERENCE (AFD) assure that the F~(Z) upper bound envelope of 2.45 times the normalized axial peaking factor xs not exceeded during either normal operation or in the event of xenon redistribution following power changes'arget flux difference (TARGET AFD) is determined at equilibrium xenon condi-tions.

The rods may be positioned within the core in accordance with their respective insertion limits and should be inserted near their normal, position for steady-state operation at high power levels.

The value of the target flux difference obtained under these conditions divided by the fraction of RATED THERMAL POWER is the TARGET AFD at RATED THERMAL POWER for the associated core burnup conditions.

TARGET AFD for other THERMAL POWER levels are obtained by multiplying the RATED THERMAL POWER value by the appropriate fractional THERMAL POWER level.

The periodic updating of the target flux difference value is necessary to reflect core burnup considerations.

SHEARON HARRIS UNIT 1

B 3/4 2-1 Amendment No. g, 15

POWER DISTRIBUTION LIMITS BASES AXIAL FLUX DIFFERENCE (Continued)

At power levels below APL

, the limits on AFD are specified in the Core Operating Limits Report, i.e., that defined by the RAOC operating procedure and limits.

These limits were calculated in a manner such that expected operational transients, e.g.,

load follow operations, would not result in the AFD deviating outside of those limits.

However, in the event such a deviation

occurs, the short period of time allowed outside of the limits at reduced power levels will not result in significant xenon redistribution such that the envelope of peaking factors would change sufficiently to prevent operation in the vicinity of the APL power level.

At power levels greater than APL two modes of operation are permissible:

1)

RAOC with fixed AFD limits as a function of reactor power level and 2) Base Load operation which is defined as the maintenance of the AFD within a band about a target value.

Both the fixed AFD limits for RAOC operation and the band for Base Load operation are specified for each reload cycle in the CORE OPERATION LIMITS REPORT per Specification 6.9.1.6.

The RAOC operating procedure above APL is the same as that defined for operation below APL However, it is possible when following extended load following maneuvers that the AFD limits may result in restrictions in the maximum allowed power or AFD in order to guarantee operation with F~(Z) less than its limiting value.

To allow operation at the maximum permissible

value, the Base Load operating procedure restricts the indicated AFD to a relatively small target band and power swings.

For Base Load operation, it is expected that the plant will operate within the target band.

Operation outside of the target band for the short time period allowed will not result in significant xenon redistribution such that the envelope of peaking factors would change sufficiently to prohibit continued operation in the power region defined above.

To assure'here is no residual xenon redistribution impact from past operation on the Base Load operation, a 24-hour waiting period at a power level above APL and allowed by RAOC is necessary.

During this time period, load changes and rod motion are restricted to that allowed by the Base Load procedure.

After the waiting period, extended Base Load operation is permissible.

The computer determines the one-minute average of each of the OPERABLE excore detector outputs and provides an alarm message immediately if the AFD for two or more OPERABLE excore channels are'.

1) outside the allowed DI power operating space (for RAOC operation),

or 2) outside the acceptable AFD target band (for Base Load -operation).

These alarms are active when power is greater than:

1) 50X of RATED THERMAL POWER (for RAOC operation),

or 2) APL (for Base Load operation).

Penalty deviation minutes for Base Load operation are not accumulated based on the short period of time during which operation outside of the target band is allowed.

SHEARON HARRIS UNIT 1 B 3/4 2-2 Amendment No. 7, 15

POWER DISTRIBUTION LIMITS BASES 3/4.2.2 AND 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR The limits on heat flux hot channel factor, RCS flow rate, and nuclear enthalpy rise hot channel factor ensure that:

(1) the design limits on peak local power density and minimum DNBR are not exceeded and (2) in the event of a LOCA the peak fuel clad temperature will not exceed the 2200'F ECCS acceptance criteria limit.

Each of these is measurable but will normally only be determined periodically as specified in Specifications 4.2.2 and 4.2.3.

This periodic surveillance is sufficient to ensure that the limits are maintained provided:

a ~

Control rods in a single group move together with no individual rod insertion differing by more than

+

12 steps, indicated, from the group demand position',

b.

Control rod groups are sequenced with overlapping groups as described in Specification 3.1.3.6; SHEARON HARRIS UNIT 1 B 3/4 2-2a Amendment No. J, 15

t 4e P

4

- J

POWER DISTRIBUTION LIMITS BASES HEAT FLUX HOT CHANNEL FACTOR, AND RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR (Continued) c.

The control rod insertion limits of Specifications 3.1.3.5 and 3.1.3.6 are maintained; and d.

The axial power distribution, expressed in terms of AXIAL FLUX DIFFERENCE, is maintained within the limits.

F H will be maintained within its limits provided Conditions a. through d.

above are maintained.

The combinations of the RCS flow requirement and the measurement of F>H ensures that the calculated DNBR will not be below the design DNBR value.

The relaxation of F>H as a function of THERMAL POWER allows changes in the radial power shape for all permissible rod insertion limits.

F<H is evaluated as being less than or equal to 1.56 for LOPAR fuel and 1.59 N

for VANTAGE 5 fuel.

These values are used in the various accident analyses where F<H influences parameters other than DNBR, e.g.,

peak clad N

temperature, and thus is the maximum "as measured" value allowed.

Margin is maintained between the safety analysis limit DNBR and the design limit DNBR.

This margin is more than sufficient to offset any rod bow penalty and transition core penalty.

When an F~ measurement is taken, an allowance for both experimental error and manufacturing tolerance must, be made.

An allowance of 5% is appropriate for a fulL-core map taken with the Incore Detector Flux Mapping System, and a

3%

allowance is appropriate for manufacturing tolerance.

The hot channel factor F~(Z) is measured periodically and increased by a cycle M

and height dependent power factor appropriate to either RAOC or Base Load operation, W(Z) or W(Z)BL, to provide assurance that the limit on the hot channel factor, F~(Z), is met.

W(Z) accounts for the effects of normal operation transients and was determined from expected power control maneuvers over the full range of burnup conditions in the core.

W(Z)BL accounts for the more restrictive operating limits allowed by Base Load operation which result in less severe transient values.

The W(Z) function for normal operation is provided in the Core Operating Limits Report per Specification 6.9.1.6.

SHEARON HARRIS UNIT 1

B 3/4 2-4 Amendment No. 7, 15

A ~

el,

POWER DISTRIBUTION LIMITS BASES HEAT FLUX HOT CHANNEL FACTOR AND RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR Continued)

A measurement error of 4X for F<H has been allowed for in determination of the N

design DNBR value.

When RCS flow rate is measured, no additional allowance is necessary prior to comparison with the limit of Specification 3.2.3.

A normal RCS flowrate error of 2.1X will be included in Cl, which will be modified as discussed below.

The measurement error for RCS total flow rate is based upon performing a

precision heat balance and using the result to calibrate the RCS flow rate indicators.

Potential fouling of the feedwater venturi which might not be detected could bias the result from the precision heat balance in a non-conservative manner.

Therefore, a penalty of 0.1Z for undetected fouling of the feedwater venturi, raises the nominal flow measurement allowance, Clp

'to 2.2Z for no venturi fouling.

Any fouling which might bias the RCS flow rate measurement greater than 0.1X can be detected by monitoring and trending various plant performance parameters.

If detected, action shall be taken before performing subsequent precision heat balance measurements, i.e., either the effect of the fouling shall be quantified and compensated for in the RCS flow rate measurement or the venturi shall be cleaned to eliminate the fouling.

The 12-hour periodic surveillance of indicated RCS flow is sufficient to detect only flow degradation that could lead to operation outside the accept-able region of operation.

3/4.2.4 UADRANT POWER TILT RATIO The QUADRANT POWER TILT RATIO limit assures that the radial power distribution satisfies the design values used in the power capability analysis.

Radial power distribution measurements are made during STARTUP testing and period-ically during power operation.

The limit of 1.02, at which corrective action is required, provides linear. heat generation rate protection with x-y plane power tilts.

tilt of 1.025 can be tolerated before the margin for uncertainty in depleted.

A limit of 1 ~ 02 was selected to provide an allowance for uncertainty associated with the indicated power tilt.

DNB and A limiting FQ is the The 2-hour time allowance for operation with a tilt condition greater than 1.02 but less than 1.09 is provided to allow identification and correction of a dropped or misaligned control rod.

In the event such action does not correct the tilt, the margin for uncertainty on FQ is reinstated by reducing the maximum allowed power by 3X for each percent of tilt in excess of 1.

SHEARON HARRIS UNIT 1 B 3/4 2-S Amendment No. 15

POWER DISTRIBUTION LIMITS BASES UADRANT POWER TILT RATIO (Continued)

For purposes of monitoring QUADRANT POWER TILT RATIO when one excore detector is inoperable, the movable 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 HRC should be submitted within 30 days in accordance with 10CFR50.4.

3/4.2.5 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'he limits are consistent with the ini-tial FSAR assumptions and have been analytically demonstrated adequate to maintain a minimum DNBR that is equal to or greater than the design DNBR value throughout each analyzed transient.

The indicated T

value and the avg indicated pressurizer pressure value are compared to analytical limits of 594.1'F and 2185 psig, respectively, after an allowance for measurement uncertainty is included.

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. )'3 15

\\

)

4

'44 AC d

~ '

0 el'>>

fg

"IP 3/4.4 REACTOR COOLANT SYSTEM BASES 3/4e4

~ 1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION I

The plant is designed to operate with all reactor coolant loops in operation and maintain DNBR above the design DNBR value during all normal operations and anticipated transients.

In MODES 1 and 2 with one reactor coolant loop not in operation this specification requires that the plant be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

In MODE 3, two reactor coolant loops provide sufficient heat removal capability for removing core decay heat even xn the event of a bank withdrawal accident>

however, a single reactor coolant loop provides sufficient heat removal capacity if a bank withdrawal accident can be prevented, i.e.,

by opening the Reactor Trip System breakers.

Single failure considerations require that two loops be OPERABLE at all times'n MODE 4, and in MODE 5 with reactor coolant loops filled, a single reactor coolant loop or RHR loop provides sufficient heat removal capability for remov-ing decay heat; but single failure considerations require that at least two loops (either RHR or RCS) be OPERABLE.

In MODE 5 with reactor coolant loops not filled, a single RHR loop provides sufficient heat removal capability for removing decay heat; but single failure considerations, and the unavailability of the steam generators as a heat remov" ing component, require that at least two RHR loops be OPERABLE.

The operation of one reactor coolant pump (RCP) or one RHR pump provides ade-quate flow to ensure mixing, prevent stratification and produce gradual re-activity changes during boron concentration reductions in the Reactor Coolant System.

The reactivity change rate associated with boron reduction will, there-

fore, be within the capability of operator recognition and control.

The restrictions on starting an RCP with one or more RCS cold legs less than or equal to 335'F are provided to prevent RCS pressure transients, caused by energy additions from the Secondary Coolant

System, which could exceed the limits of Appendix G to 10 CFR Part 50.

The RCS will be protected against over-pressure transients and will not exceed the limits of Appendix G by restricting starting of the RCPs to when the secondary water temperature of each steam generator is less than 50'F above each of the RCS cold leg temperatures.

3/4.4.2 SAFETY VALVES The pressurizer Code safety valves operate to prevent the RCS from being pres-surized above its Safety Limit of 2735 psig.

Each safety valve is designed to relieve 380,000 lbs per hour of saturated steam at the valve Setpoint.

The relief capacity of a single safety valve is adequate to relieve any overpressure condition which could occur during shutdown.

In the event that no safety valves are OPERABLE> an operating RHR loop, connected to the RCS, provides overpressure relief capability and will prevent RCS overpressurization.

In addition, the Overpressure Protection System provides a diverse means of protection against RCS overpressurization at low temperatures.

SHEARON HARRIS - UNIT 1 B 3/4 4-1 Amendment No. 15

ADMINISTRATIVE CONTROLS 6.9.1.6 CORE OPERATING LIMITS REPORT 6.9. 1.6.1 Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT prior to each reload cycle, or prior to any remaining portion of a reload cycle, for the following.'.

The shutdown rod insertion Limits of Specification 3.1.3.5.

b.

The controL rod insertion limits of Specification 3 '.3.6.

cd The axial. flux difference of Specification 3.2.1.

d.

The surveillance requirements of Specifications 4.2.2.2, 4.2.2.3, and 4.2.2.4.

6.9.1.6.2 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.'.

WCAP-10216-P-A, Relaxation of Constant Axial Offset Control F~

SurveiLlance Technical Specification, 1983.

b.

WCAP-9272-P-A, Westinghouse Reload Safety Evaluation Methodology, 1985.

6.9.1.6.3 The core operating limits shall be determined so that all applicabLe limits (e.g ~, fuel thermaL-mechanical Limits, core thermal-hydraulic limits, nuclear limits such as shutdown margin, and transient and accident anaLysis limits) of the safety analysis are met.

6.9.1.6.4 The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements, shall be provided, upon issuance for each reload cycle, to the NRC Document Control. Desk, with copies to the Regional Administrator and Resident Inspector.

SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the Regional'dministrator of the Regional Office of the NRC 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; SHEARON HARRIS UNIT 1

6-24 Amendment No. /, 7P, 15

)

I aW

ADMINISTRATIVE CONTROLS RECORD RETENTION (Continued) b.

Records and Logs of principal maintenance activities, inspections,

repair, and repLacement of principal items of equipment related to nuclear safety',

c.

All REPORTABLE EVENTS; d.

Records of surveillance activities, inspections, and calibrations required by these Technical Specifications; e.

Records of changes made to the procedures required by Specifica-tion 6.8.1; f.

Records of radioactive shipments; g.

Records of sealed source and fission detector leak tests and results; and SHEARON HARRIS UNIT 1

6-24a Amendment No. /1, 15

nn Pl n

W fL eJ