Text

Amend 7 to License NPF-63,revising Tech Specs to Change Control Rod Bank-D Configuration,Increasing Radial & Total Peaking Factors & Introducing Boron Dilution/Sliding Shutdown Margin
	ML18005A549
Person / Time
Site:	Harris
Issue date:	08/16/1988
From:	Adensam E; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML18005A551	List: ML18005A549;
References
	NPF-63-A-007 NUDOCS 8808220191
	Download: ML18005A549 (56)
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UNITEDSTATES NUCLEAR REGULATORY COMMISStON wAsMINQTON,D. c.~

CAROLINA POWER 8

LIGHT COMPANY et al.

OOCKET NO. 50-400 SHEARON HARRIS NUCLEAR POWER PLANT UNIT 1 AMENOMENT TO FACILITY OPERATING LICENSE Amendment No. 7 License No.

NPF-63 l.

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

A.

The application for amendment by Carolina Power 8 Light Company, (the licensee),

dated February 1 and amended February 8, 1988, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; 8.

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

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comnission'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 2.

E.

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

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.

NPF-63 is hereby amended to read as follows:

8808220i9f 8808i6 PDR ADQCK 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.

7

, are hereby incorporated into this license.

Carolina Power 8 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 COMMISSION

/~

Elinor G.

Adensam, Director Project Directorate II-1 Division of Reactor Projects I/II

Attachment:

Changes to the Technical Specifications Date of Issuance:

August 16, 1988

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ATTACHMENT TO LICENSE AMENDMENT NO.

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 revised areas are indicated by marginal lines.

Remove Pa es iv vxiii 2-2 2-8 8 2-1 3/4 1-1 3/4 1-2 3/4 1-3 3/4 1-8 3/4 1-10 3/4 1-11 3/4 1-12 3/4 1-14 3/4 1-21 3/4 1-22 3/4 2-1 3/4 2-2 3/4 2-3 3/4 2-4 3/4 2-5 3/4 2-6 3/4 2-8 3/4 2-9 3/4 3-11 3/4 3-14 8 3/4 1>>1 8 3/4 1-2 8 3/4 1-3 8 3/4 2-1 8 3/4 2-2 8 3/4 2-3 8 3/4 2-4 5-6 6-24 Insert Pa es iv vxiii 2-2 2>>8 8 2-1 3/4 1-1 3/4 1-2 3/4 1-3 3/4 1-3a 3/4 1-8 3/4 1-10 3/4 1-11 3/4 1-12 3/4 1-14 3/4 1-21 3/4 1-22 3/4 2-1 3/4 2-2 3/4 2-3 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 3-11 3/4 3-14 3/4 3-14a 8 3/4 1-1 8 3/4 1-la 8 3/4 1-2 8 3/4 1-3 8 3/4 2-1 8 3/4 2-2 8 3/4 2-2a 8 3/4 2-3 8 3/4 2-4 5-6 5-6a 6-24 6-24a

3 '/4+0 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE UIREMENTS SECTION

/4 '

APPLICABILITYo~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

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

~ ~ ~ ~ ~

3 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4 ~ 1 ~ 1 BORATION CONTROL Shutdown Margin - MODES 1 and 2., ~ ~ ~ ~ ~. ~ ~ ~ ~.. ~. ~ ~ ~..

~ ~. ~.

Shutdown Margin - MODES 3, 4, and 5 '

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

~...

FIGURE 3'-1 SHUTDOWN MARGIN VERSUS RCS BORON CONCENTRATION MODES 3f 4f AND 5 ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

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

Moderator Temperature Coefficient ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

~ ~ ~.. ~ ~ ~ ~. ~

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

3/4. 1 ~ 2 BORATION SYSTEMS Flow Path Shu'tdown

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

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

Flow Paths Operat1ng..... "....................... "".

Charg1ng Pump - Shutdown.

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

~ ~ ~. ~ ~ ~ ~. ~

Charg1ng Pumps - Opera t1ng.."""""""""."..."".

Borated Water Source - Shutdown..

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

~ ~.. ~....

Borated Water Sources - Operating"."".."... "... "."

3/4 '

~ 3 MOVABLE COHTROL ASSEMBLIES Group Hexghtiiii ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

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

TABLE 3.1-1 ACCIDEHT ANALYSES REQUIRING REEVALUATION IH THE EVENT OF AN INOPERABLE RODo ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

~ ~ ~ ~ ~ ~ ~ ~ ~

- Position Indication Systems - Operating." " ~ ~ ~" ~" ~

Position Indication System - Shutdown.....

~........

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Rod Drop Time o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ "o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

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Shutdown Rod Insert1on L1m1t... ~""~""~"""~ ~"""~

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Control Rod Insert1on L1m1ts. ~... ~ ~ ~ ~ ~ ~ ~.." ~ ~. ~ ~ ~ ~ ~. ~ ~"

FIGURE 3.1-2 ROD GROUP IHSERTION LIMITS VERSUS THERMAL POWERS THREE LOOP OPERATIONo ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

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

PAGE 3/4 0-1 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 LV Amendment No.

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LIMITIMGCONDITIONS FOR OPERATION AND SURVEILLANCE RE UIREMENTS SECTION 3/4 '

POMER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE......

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

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

FIGURE 3.2-1 AXIAL FLUX DIFFERENCE LIMITS AS A PUNCTION OP RATED THERMAL POWER FOR RAOC ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

~ ~ ~ ~

3/4 ~ 2 ~ 2 HEAT FLUX HOT CHANNEL FACTOR FQ(Z )

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

FIGURE 3.2-2 K(Z) - LOCAL AXIAL PENALTY FUNCTION POR P (Z) ~. ~ ~ ~ ~. ~

3/4.2.3 RCS FLOM RATE AND NUCLEAR ENTtMPY RISE HOT CHANNEL FACTORe ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

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

~ ~ ~ ~ ~

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

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

3/4 ' '

DNB PARAMETERS ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

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

~ ~ ~

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 INSTRUMENTATION 3/4'.1 REACTOR TRIP SYSTEM INSTRUMENTATION............... ~ ~ ~... ~

TABLE 3'-1 REACTOR TRIP SYSTEM INSTRUMENTATION......... ~ ~ ~ ~ ~ ~ ~ ~. ~

TABLE 3.3-2 REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES'

TABLE 4.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

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

3/4.3.2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION+ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

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

~

TABLE 3'-3 ENGINEERED SAFETY FEATURES ACTUATIOM SYSTEM INSTRUMENTATIONo~ o ~ ~ ~ ~ ~ ~ ~ ~ oo ~ ~ ~ oo ~ I ~ ooo ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~

TABLE 3'-4 EHCINEERED SAFETY PEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTSeo

~ ~ ~ ooo ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

~ ~ ~

TABLE 3 ~ 3 5 EMCIHEERED SAPETY PEATURES

RESPONSE

TIMES ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

TABLE 4,3-2 EHCINEERED SAFETY FEATURES ACTUATION SYSTEM IHSTRUMEHTATIOM SURVEILLANCE REQUIRBKNTS ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ s ~

3/4 ~3. 3 MOHITORING INSTRUMENTATION Radiation Monitoring for Plant Operations" "". ~."~" ~ ~

3/4 3-1 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/a 3-5o SHEARON HARRIS - UNIT 1 Amendment No.

7

3.0/4 '

BASES SECTION 3/4o0 APPLICABILITYoo.oo~ o ~ ooo.o

~ ~ ~ ~ oooooooooooooooo

~ ~ ~ oooo ~ ~ ~ ~ o

~

o'/4.1 REACTIVITY CONTROL SYSTEMS 3/4.F 1 BORATION CONTROL.........

~ ~ ~..

~ ~. ~. ~ ~...

~ o ~ ~ ~ ~..

~ ~. ~ ~. ~. ~.

3/4 ' '

BORATION SYSTEMS ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

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

3/4.1.3 MOVABLE CONTROL ASSEMBLIESoo ~.. ~.....

~. ~ ~ ~. ~ ~ ~.. ~. ~ ~,

~ ~...

3/4 '

POSER DISTRIBUTION LIMITS ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

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

3/4.2.1 AXIAL FLUX DIFFERENCE...............

~...

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

3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR, AND RCS PlAN RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL PACTORo ~ ~ ~ ~ ~ ~ ~

PIGURE B 3/4 '

1 (DELETED)

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

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

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

~ ~. ~ o.....

~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~...

3/ao2.5 DNB PARAMETERS'...

~ ~ ~ ~.

. ~ ~ ~..

~ ~......

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

.oooo.

3/4o3 INSTRUMENTATION 3/4.3 '

and 3/4'.2 REACTOR TRIP SYSTEM AND ENGINEERED SAPETY FEATURES ACTUATION SYSTEM IHSTRUMENTATIONo~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

3/4.3.3 MONITORING INSTRUMENTATION.~... ~... ~......

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

~ ~ ~.

3/4.3.4 TURBINE OVERSPEED PROTECTION.......

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

3/4.4 REACTOR COOLANT SYSTEM PAGE B 3/4 0-1 B 3/4 1-1 B 3/4 1-2 B 3/4 1-3 B 3/4 2-1 B 3/4 2-1 B 3/4 2-2 s F14 2-3 B 3/4 2 5 B 3/4 2-6 B 3/4 3-1 B 3/4 3-3 B 3/4 3"6 3/a.4.1 3/4.4.2 3/4o4.3 3/a.a.a 3/4.4.5 3/4.4.6 3/4.4.7 3/4.4.8 3/a.a.9 REACTOR COOLANT LOOPS AHD COOLANT CIRCULATIONo~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

SAFETY VALVES ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

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

PRESSURIZERo

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

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

~ ~ ~ ~

RELIEF VALVES ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

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

STEAM GEHERATORSo ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

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

REACTOR COOLANT SYSTEM LEAKAGE>> ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

CHEMISTRYo ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

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

~ ~ ~ ~ ~

SPECIPIC ACTIVITYo ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~

oo. ~ ~ ~ o ~ ~ ~ oo ~ ~ ~ oo ~ ~ oo PRESSURE/TEMPERATURE LIMITS ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

B 3/4 4-1 B 3/4 4-1 B 3/4 4-2 B 3/4 4-2 B 3/4 4-2 B 3/a a-3 B 3/4 4-4 B 3/4 4-5 B 3/4 4-6 SHEARON HARRIS - UNIT 1 XI.11 Amendment No.

7

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570 630 C'

620 610 590 570 0.0 0.1 OZ 00 OA 0.5 0.5 0.7 0.8 0.9 1.0 1.1 1.2 FRACTION OF RATED THERMAL POWER FIGURE 2.1-1 REACTOR CORE SAFETY LIMITS THREE LOOPS IN OPERATION SHEARON Y~IS - UNIT 1

2-2 Amendment No.

7

TABLE 2.2-1 (Continued)

TABLE NOTATIONS NOTE 1:

(Continued)

M I

Q

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

~ 0.000828/psig;

~ Pressurizer

pressure, psig; pl

~ 2235 psig (Nominal RCS operating pressure);

~ Laplace transform operator, s

3 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 teits such that:

(1)

For qt - qb between

-25Z and + 7.0X, fl (AI) 0, where qt and qb are percent RATED THERHAL POMER in the top and bottom halves of the core respectively, and qt + qb is total THERHAL POMER in percent of RATED THERHAL POWER; O

NOTE 2:

(2)

For each percent that the magnitude of qt - qb exceeds

-25X, the hT Trip Setpoint shall be automatically reduced by 2.36Z of its value at RATED THERHAL POMER; and (3)

For each percent that the magnitude of qt - qb exceeds

+7.0Z, the AT Trip Setpoint shall be automatically reduced by 1.70X of its value at RATED THERHAL POWER.

The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than 1.9X 4T span.

2 ~ 1 SAFETY LIMITS BASES 2'.1 REACTOR CORE The restrictions of this Safety Limit prevent overheating of the fuel and pos-sible cladding perforation which would resuLt in the release of fission prod-ucts 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 nu<<

cleate boiling (DNB) and the resultant sharp reduction in heat transfer coeffi-cient.

DNB is not a directly measurable parameter during operation and there-fore THERMAL POWER and reactor coolant temperature and prcssure have been re-lated to DNB through thc W-3 correlation.

The W-3 DNB correlation has been developed to predict the DNB flux and thc location of DNB for axially uniform and nonuniform heat flux distributions'he local DNB heat flux ratio (DNBR) is defined as the ratio of the calcul.ated heat flux that would cause DNB at a particular core location to the actual local heat flux and is indicative of the margin to DNB.

The minimum value of the DNBR during steady-state operation~

normaL operational transients, and anticipated transients is limited to 1.30'his value corre" sponds to a 95Z probability at a 95Z confidence level that DNB will not occur and is chosen as an appropriate margin to DNB for all operating conditions'he curves of Figure 2.1-1 show the loci of points of THERMAL POWER, Reactor Coolant System pressure and average temperature for which thc minimum DNBR is no less than 1 ~ 30, or the average enthalpy at the vessel exit is equal to the enthalpy of saturated liquid.

These curves are based on an enthalpy hot channel factor, F<H, of 1.55 and a N

reference cosine with a peak of 1.55 for axial power shape.

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

FAH ~ 1 ~ 55 (1 + 0' (1-P)j 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 allowabl.e control rod insertion assuming the axiaL power imbalance is within the Limits of the fl (hI) function of thc 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 2-1 Amendment No.

7

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3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1 ~ 1 BORATION CONTROL SHUTDOWN MARGIN - MODES 1 AND 2 LIMITING CONDITION FOR OPERATION 3.1.1.1 The SHUTDOWN MARGIN shall be greater than or equal to 1770 pcm for 3-loop operation.

APPLICABILITY!

MODES 1 and 2*.

ACTION:

With the SHUTDOWN MARGIN less than 1770 pcm, imediately initiate and continue boration at greater than or equal to 30 gpm of a solution containing greater than or equal to 7000 ppm boron or equivalent until the required SHUTDOWN MARGIN is restored.

SURVEILLANCE RE UIREMENTS 4.1.1.1.1 The SHUTDOWN MARGIN shall be determined to be greater than or equal to 1770 pcm'.

Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after detection of an inoperable control rod(s) and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter while the rod(s) is inoperable.

If the inoperable control rod is immovable or untrippable, the above required SHUTDOWN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the immovable or untrippable control rod(s);

b.

When in MODE 1 or MODE 2 with K ff greater than or equal to 1 at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by verifying that control bank withdrawal is within the limits of Specification 3.1.3.6; c.

Within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior to achieving reactor criticality by verifying that the "predicted critical control rod position is within the limits of Specification 3.1 ~3.6; d.

Prior to initial operation above 5X RATED THERMAL POWER after each fuel loading, by consideration of the factors below, with the control banks at the maximum insertion limit of Specification 3.1.3 'N and

See Special Test Exceptions Specification 3.10.1 ~

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

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REACTIVITY CONTROL SYSTEMS SURVEILLANCE RE UIREMENTS (Continued) 1)

2) 3)

4) 5)

6)

Reactor Coolant System boron concentration, Control rod position, Reactor Coolant System average temperature, Fuel burnup based on gross thermal energy generation, Xenon concentration, and Samarium concentration.

4'.lan 2 The overall core reactivity balance shall be compared to predicted values to demonstrate agreement within +1000 pcm at least once per 31 Effective Full Power Days (EFPD).

This comparison shall consider at least those factors stated in Specification 4.l.l.l.le., above.

The predicted reactivity values shall be adjusted (normal. ized) to correspond to the actual core conditions prior to exceeding a fuel burnup of 60 EFPD after each fuel loading. If later experience shows adjustment is desirable at approximately 60 EFPD, the adjustment is permissible.

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

7

REACTIVITY CONTROL SYSTEMS SHUTDOWN MARGIN - MODES 3

4 AND 5 LIMITING CONDITION FOR OPERATION 3.1.1.2 The SHUTDOWN MARGIN shall be greater than or equal to the limit shown in Figure 3.1-1.

APPLICABILITY:

MODES 3, 4, AND 5.

ACTION:

With the SHUTDOWN MARGIN less than the required value iaaaediately initiate and continue boration at greater than or equal to 30 gpm of a solution containing greater than or equal to 7000 ppm bo~on or equivaLent until the required SHUTDOWN MARGIN is restored.

SURVEILLANCE RE UIREMENTS 4.1.1.2 The SHUTDOWN MARGIN shalL be determined to be greater than or equal to the required value:

a ~

Wi'thin 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after detection of an inoperable control rod(s) and at Least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter while the rod(s) is inoperable.

If the inoperabLe control rod is iamovable or untrippable, the SHUTDOWN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the iaaaovable or untrippable control rod(s);

and b.

At Least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by consideration of the following factors.')

2) 3)

4) 5)

6)

Reactor CooLant System boron concentration, Control rod position, Reactor Coolant System average temperature, Fuel burnup based on gross thermaL energy generation, Xenon concentration, and Samarium concentration.

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

7

8000 7000 (2200, <60) 6000

. 5000 Cl M

4000

+~pe 3000 O

20M M

R E 4 AT lEhST NE P IN ERA N

AND

,1770 OO,22

)

1000 0

400 800 1200

=

1600 2000 REQUIRED RCS BORON CONCENTRATION (ppm)

(BURNUP DEPENDENT)

+ Applicable to Mode 4, with or without RCP's in operation FIGURE 3.1-1 SHUTDOWN MARGIN VERSUS RCS BORON CONCENTRATION MODES 3, 4, AND 5

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REACTIVITY CONTROL SYSTEMS FLOW PATHS - OPERATINC LIMITING CONDITION FOR OPERATION 3.1.2.2 At least two of the following three boron injection flow paths shall be OPERABLE:

a.

The flow path from the boric acid tank via a boric acid transfer pump and a charging/safety injection pump to the Reactor Coolant System (RCS),

and b.

Two flow paths from the refueling water storage tank via charging/

safety injection pumps to the RCS.

APPLICABILITY:

MODES 1, 2, and 3 ~

ACTION:

With only one of the above required boron injection flow paths to the RCS OPERABLE, restore at least two boron injection flow paths to the RCS to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY and borated to a SHUTDOWN MARGIN as required by Figure 3'-1 at 200'F within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; restore at least two flow paths to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE RE UIREMENTS 4.1.2.2 At least two of the above required flow paths shaLL be demonstrated OPERABLE:

a ~

At Least once per 7 days by verifying that the temperature of the flow path between the boric acid tank and the charging/safety injec-tion pump suction header tank is greater than or equal to 65'F when a flow path from the boric acid tank is used; b.

At least once per 31 days by verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked,

sealed, or otherwise secured in position, is in its correct position; c.

At least once per 18 months during shutdown by verifying that each automatic valve in the flow path actuates to its correct position on a safety injection test signal; and d.

At least once per 18 months by verifying that the flow path required by Specification 3.1.2.2a.

delivers at least 30 gpm to the RCS.

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

7

REACTIVITY CONTROL SYSTEMS CHARGING PUMPS - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.4 At least two charging/safety injection pumps shall be OPERABLE.

APPLICABILITY:

MODES 1, 2, and 3.

ACTION:

With only one charging/safety injection pump OPERABLE, restore at least two charging/safety injection pumps to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY and borated to a SHUTDOWN MARGIN as required by Figure 3.1-1 at 200'F within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ', restore at least two charging/safety injection pumps to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE RE UIREMENTS 4.1.2.4 At least two charging/safety injection pumps shall be demonstrated OPERABLE by verifying, on recirculation flow or in service supplying flow to the Reactor Coolant System and reactor coolant pump seals, that a differential pressure across each pump of greater than or equal to 2446 psid is developed when tested pursuant to Specification 4.0.5.

SHEARON HARRIS - UNIT 1 3/4 1-10 Amendment No.

7

REACTIVITY CONTROL SYSTEMS BORATED WATER SOURCE - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.2.5 As a minimum, one of the following borated water sources shall be OPERABLE'.

A boric acid tank with:

1 ~

A minimum contained borated water volume of 7100 gallons, which is equivalent to 17X indicated level, 2.

A boron concentration of between 7000 and 7750

ppm, and 3.

A minimum solution temperatureof 65'F.

b.

The refueling water storage tank (RWST) with:

1.

A minimum contained borated water volume of 106,000 gallons, which is equivalent to 12X indicated level, 2.

A boron concentration of between 2000 and 2200

ppm, and 3.

A minimum solution temperature of 40'F.

APPLICABILITY:

MODES 5 and 6.

ACTION:

With no borated water source OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes'URVEILLANCE RE UIREMENTS 4.1.2.5 The above required borated water source shall be demonstrated OPERABLE:

a.

At least once per 7 days by.'.

. Verifying the boron concentration of the water, 2.

Verifying the contained borated water volume, and 3.

Verifying the boric acid tank so'l.ution temperature when it is the source of borated water.

b.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the RWST temperature when it is the source of borated water and the outside air temperature is less than 40'F.

SHEARON HARRIS - UNIT 1

3/4 1-11 Amendment No.

7

REACTIVITY CONTROL SYSTEMS BORATED WATER SOURCES - OPERATING LIMITING CONDITION FOR OPERATION 3 '.2.6 As a minimum, the following borated water source(s) shall be OPERABLE as required by Specification 3.1.2.2:

a.

A boric acid tank with:

1 ~

A minimum contained borated water volume of 21,400 gallons'hich is equivalent to 60Z indicated level.

2.

A boron concentration of between 7000 and 7750

ppm, and 3.

A minimum solution temperature of 65'F.

b.

The refueling water storage tank (RWST) with:

A minimum contained borated water volume of 436,000 gallons, which is equivalent to 92Z indicated level.

2.

A boron concentration of between 2000 and 2200

ppm, 3 ~

A minimum solution temperature of 40'F, and 4.

A maximum solution temperature of 125'F.

APPLICABILITY:

MODES 1, 2, 3, and 4.

ACTION:

a.

With the boric acid tank inoperable and being used as one of the above required, borated water sources, restore the boric acid tank to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and borated to a SHUTDOWN MARGIN as required by Figure 3'-l at 200'F; restore the boric acid tank to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months '.

With the RWST inoperable, restore the tank to OPERABLE status within 1 ho'ur or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SHEARON HARRIS UNIT 1 3/4 1-12 Amendment No.

7

J

~

~ i

~

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 zods inoperable due to being immovable as a zesult 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 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months '.

With more than one rod misaligned from the gzoup step counter demand

. position by more than

+

12 steps (indicated position),

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

c ~

With moze 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 houz's, be in HOT STANDBY within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

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 'an

+ 12 steps (indicated position),

POWER OPERATION may continue provided that within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months '.

1.

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 of Figure 3.1-2.

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 S days', this zeevaluation shall confirm that the previously analyxed 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.

7

I I

A A

~

~ ~

~

REACTIVITY CONTROL SYSTEMS CONTROL ROD INSERTION LIMITS LIMITINC CONDITION FOR OPERATION 3.1.3.6 The control banks shall be limited in physical insertion as shown in Figure 3'-2.

APPLICABILITY:

MODES 1+ and 2* ~<.

ACTION:

With 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 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , or b.

C ~

Reduce THERMAL POWER within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to less than or equal to that fraction of RATED THERMAL POWER which is allowed by the bank posi-tion using Figure 3.1-2, or Be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

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 0.00333 hours 1.984127e-5 weeks 4.566e-6 months except during time intervals when the rod insertion limit monitor is inoperable, then verify the individual rod positions at least once per 4 bourse'See Special Test Exceptions Specifications 3.10.2 and 3.10.3 ~

~ith Keff greater than or equal to l.

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

7

K C 18S) 160 a

140 a

120 100 A

Q 60 (0,1 20 (0,

0.00

0. 0 ONO 0~

OAO 0.50 0.60 0.70 O.BO 0.90 1.00 FRACllON OF RATED THERMAL POWER FIGURE 3.1-2

- ROD GROUP INSERTION LIMITS VERSUS THERMAL POV/ER THREE-LOOP OPERATION SHEARON HARRIS - UNIT 1

3/4 1-22 Amendment No.

7

I I

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~

3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE LIMITING CONDITION FOR OPERATION 3.2 '

The indicated AXIAL FLUX DIFFERENCE (AFD) shall be maintained within.

a.

the acceptable operational space defined by Figure 3.2"1 for ReLaxed Axial Offset Control (RAOC) operation, or b.

within a

+ 3 percent target band about the target AFD during Base Load operation.

APPLICABILITY:

MODE 1 above 50Z of RATED THERMAL POWER+.

ACTION:

a.

For RAOC operation with the indicated AFD outside of the Figure 3.2-1 limits, either'.

1.

Restore the indicated AFD to within the Figure 3.2-1 limits within 15 minutes, or 2.

Reduce. THERMAL POWER to less than 50Z of RATED THERMAL POWER within 30 minutes and reduce the Power Range Neutron Flux "

High Trip setpoints to less than or equal to 55Z of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

b.

C ~

For Base Load operation above APL with 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 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 50Z of RATED THERMAL POWER unless the indicated AFD is within the Figure 3.2-1 Limits.-

+See Special Test Exception 3.10.2

~APL is the minimum allowable power level for Base Load operation and will be provided in the Peaking Factor Limit Report per Specification 6.9.1.6.

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

7

I I

I

~

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:

l.

At least once per 7 days when the AFD Monitor Alarm is OPERABLE, and 2.

At least once per hour for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after restoring the AFD Monitoring Alarm to OPERABLE status.

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 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 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'he 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.

7

~ I

~

E

~

I Page 3/4 2-3 Intentionally Left Blank SHEARON HARRIS - UNIT l 3/4 2-3 Amendment No.

7

120 110 100

(-31

)

90 70 8

5 20 10 50 ~

.30

-20

-10 0

10 20 30 40 50.

AXIAL FLUX DIFFERENCE (s)

FIGURE 3.2-1 AXIAL FLUX DIFFERENCE LIMITS AS A FUNCTION OF RATED THERMAL POWER FOR RAOC SHEARON HARRIS - UNIT 1

3/4 2-~

Amendment No.

7

POWER DISTRIBUTION LIMITS 3/4 ' '

HEAT FLUX HOT CHANNEL FACTOR - F (Z)

LIMITING CONDITION FOR OPERATION 3'.2 F~(Z) shall be limited by the following relationships:

Fq(Z) < 2+32 [K(Z)] FOR P > 0' P

FQ(Z)

C (4.64) [K(Z)] FOR P

C 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 lo ACTION!

With F~(Z) exceeding its limit:

a.

Reduce THERMAL POWER at least 1Z for each 1Z F~(Z) exceeds the limit within 15 minutes and similarly reduce the Power Range Neutron Flux-High Trip Setpoints within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; POWER OPERATION may proceed for up to a total of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; subsequent POWER OPERATION may proceed provided the Overpower hT Trip Setpoints have been reduced at least 1Z for each 1Z F~(Z) exceeds the limit.

b.

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 PO'WER may then be increased provided F~(Z) is demonstrated through incore mapping to be within its limit.

SHEARON HARRIS - UNIT 1 3/4 2-5 Amendment No.

7

~

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, F~(Z) shall be evaluated to determine if it is within its limit by:

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 Fz(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'erify the requirements of Specification 3.2.2 are satisfied.

c.

Satisfying the following relationship'.

F M(Z) < 2 32 P x W Z F

(Z) ( 2.32 x K(Z) for P ( 0 '

x 0.5 where F (Z) is the measured F (Z) increased by the allowances for manufacturing tolerances and sIeasurement uncertainty, 2.32 is the F~

limit, K(Z) is given in Figure 3'-2, P is the fraction 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 Peaking Factor Limit Report as per Specification 6.9.1 ~ 6.

d.

Measuring F~ (Z) according to the following schedule'.

1.

Upon achieving equilibrium conditions after exceeding by 10X or more of RATED THERMAL POWER, the THERMAL POWER at which F~(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

I

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I

POWER DISTRIBUTION LIMITS SURVEILLANCE RE UIREMENTS (Continued) e.

With measurements indicating maximum FM (z) has increased since the previous determination of F~ (Z) either of the following actions shall be taken.')

P (Z) shall be increased by 2Z over that specified in S ecification 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 FM (z) is not increasing.

E.

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:

max 1mum F(Z) x W(Z) x K(z) 2'2 P

~

x 100 for P ) 0.5 maximum F (Z) x W(Z) x K(Z) 2'2 0.5 x100 for P(05 2)

One of the following actions shall be taken:

a)

Within 15 minutes, control the AFD to within new AFD limits which are determined by reducing the AFD limits of 3.2-1 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 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , 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 Eor Base Load operation are satisfied and enter Base Load operation.

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

7

POWER DISTRIBUTION LIMITS SURVEILLANCE RE UIREMENTS (Continued) g.

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 15Z, inclusive.

2.

Upper core region from 85 to 100Z, 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 '

Eor at least the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Maintain Base Load operation surveillance (AFD within + 3Z of target Elux difEerence) 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 g) surveillance is maintained pursuant to Specification 4.2.2.4.

APL is defined as:

APL

= minimum [

x 100Z BL

~

2'2 x K(Z)

Fq(Z) x W(Z)BL where.'F (Z) is the measured F (Z) increased by the allowances for manufacturing tolerances and measurement uncertainty.

The F limit is 2.32.

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

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

The function is given in the Peaking Factor Limit Report as per Specification 6.9.1.6.

b.

During 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 F~(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 above APL b.

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

Verify the requirements of Specification 3.2.2 are satisfied.

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

7

POWER DISTRIBUTION LIMITS SURVEILLANCE RE UIREMENTS (Continued)

C ~

Satisfying the foLLowing relationship.

F (Z)

W Z) for P

> APL M

2.32 x K(Z)

ND BL d ~

where.'

(Z) is the measured F~(Z).

The F~ Limit is 2.32.

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

P is the fraction of RATED THERMAL POWER.

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

operation.

This function is given in the Peaking Factor Limit Report as per Specification 6.9.1.6.

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

determination according to the following schedule.'.

Prior to entering Base Load operation after satisfying Section 4.2.2.3 unless a ful,l core'Lux 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 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to mapping, and e.

2.

At least once per 31 effective full power days.

With measurements indicating F(z) maximum

[

)

]

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

1.

F (Z) shall be increased by 2 percent over that specified in 4.2.2.4.c, or 2 ~

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

EFPD until 2 saccessive maps indicate that F (z) maximum [

]

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 F (Z), or SHEARON HARRIS - UNIT 1 3/4 2-7c Amendment No.

7

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) x K(Z) 2.32 P

g.

The limits specified in 4.2.2.4.c, 4'.2.4.e, and 4'.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 overalI measured F (Z) shall be obtained from a power distribution map and increased by 3Z 3o account for manufacturing tolerances and further increased by 5Z to account for measurement uncertainty.

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

7

1.25 1.00 N

075 0.50 TOTAL Fq 2.32 CORE HEIGHT 0.000 S.OOO 10.800 12.000 KQZ 1.000 0.840 0.647 0.25 0.00 6

8 CORE HEIGHT (FT) 10 FIGURE 3.2-2 K(Z) LOCAL AXIAL PENALTY FUNCTION FOR Fq (Z)

POWER DISTRIBUTION LIMITS 3/4 ~ 2 ~ 3 RCS FMW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR LIMITINC CONDITION FOR OPERATION 3'.3 The indicated Reactor Coolant System (RCS) total flow rate and F<H shall be maintained as follows:

a.

Measured RCS flow rate

> 292,800 gpm x (1.0 + Cl), and b.

Measured F<H < 1.49 [1.0 + 0.3(1.0-P)]

Where:

P

~ THERMAL POWER

, and RATED THERMAL POWER hH N

Measured values of F<H obtained by using the movable incore detectors tI obtain a power distribution map, and the measured values of F<M shall be used for comparison above since the 1.49 value above accounts for a 4Z allowance on incore measurement of F<H.

N Cl ~

Measurement uncertainty for core flow as described in the Bases.

APPLICABILITY:

MODE l.

ACTIONS With RCS total flow rate or F<H outside the above'imits:

a.

Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months either:

l.

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 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SHEARON HARRIS - UNIT 1 3/4 2-9 Amendment No.

7

TABLE 4.3-1 REACTOR TRIP SYSTEH IHSTRUHENTATIOH SURVEILLANCE RE UIREHEHTS g

FUNCTIONAL UNIT CHAHHEL CHECK CHANNEL CALISBATIOH AHAMG CHAHHEL OPERATIOHAL TEST TRIP ACTUATING DEVICE OPERATIONAL TEST ACTUATION LOGIC TEST HODES FOR WHICH SURVEILLANCE IS RE IRED 1.

Manual Reactor Trip 2.

Power Range, Heutron Flux a.

High Setpoint b.

Low Setpoint H.ho H.h.

D(2, 4),

f(3, 4),

ff(4, 6)

R(4, S)

R(4)

H.h.

q(iS)

S/U(1)

R(12)

H.A.

H,h, H.h.

H.h.

1, 2, 3+,

4* 5*

1, 2

3.

Power Range, Neutron Flux>>

H A.

High Positive Rate 4.

Power Range, Neutron Flux, H.h.

High Negative Rate ST Intermediate

Range, Heutron Flux R(4)

R(4)

R(4, S) q(is) q(is)

S/u(1)

H.h.

Hah H.ho H.A, H.ho H.A.

1, 2

O 6.

Source

Range, Neutron Flux S

7 ~

Overtemperature hT 8.

Overpower hT S

9 ~

Presaurixer Pressure Low S

10.

Presaurixer Pressure High S

R(4, s)

R(11)

S/U(1),

Q(8, 15)

Q(15)

Q(15) q(is)

H.h.

H.h, H A H.h.

H.h H.h.

H.h.

Hobo Ho A.

H.h.

2~>>

3>> 4>>

5 1>> 2 1,

2 i (16) 1, 2

TABLE 4.3-1 (Continued)

TABLE NOTATIONS

'"'When the Reactor Trip System breakers are closed and the Control Rod Drive System is capable of rod withdrawal.

"-Below P-6 (Intermediate Range Neutron Flux Interlock) Setpoint.

~+"-Below P"10 (Low Setpoint Power Range Neutron Flux InterLock) Setpoint.

PEach 31 Effective Full Power Days.

HEach 92 Effective Full Power Days.

(1) If not performed in previous 31 days.

(2)

Comparison of calorimetric to excore power indication above 15Z of RATED THERMAL POWER.

Adjust excore channel gains consistent with calorimetric power if absolute difference is greater than 2Z.

The provisions of Specification 4.0.4 are not applicable to entry into MODE 2 or l.

(3)

Single point comparison of incore to excore,AXIAL FLUX DIFFERENCE above 15X of RATED THERMAL POWER.

Recalibrate if the absolute difference is greater than or equal to 3X.

The provisions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1 ~

(4)

Neutron detectors may be excluded from CHANNEL CALIBRATION.

(5)

Detector plateau curves shall be obtained, and evaluated and compared to manufacturer's data.

For the Intermediate Range and Power Range Neutron Flux channels the provisions of Specification 4.0.4 are not applicable for entry into MODE 2 or l.

(6)

Incore Excore Calibration, above 75Z of RATED THERMAL POWER.

The provisions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1 ~

(7)

Each train shall be tested at least every 62 days on a STAGGERED TEST BASIS'8)

Quarterly surveillance in MODES 3*, 4*, and 5* shall also include verification that permissives P-6 and P-10 are in their required state for existing plant conditions by observation of the permissive annunciator window.

(9)

Setpoint verification is not applicable.

(10) The TRIP ACTUATING DEVICE OPERATIONAL TEST shall independently verify the OPERABILITY of the undervoltage and shunt trip attachments of the reactor trip breakers.

SHEARON HARRIS - UNIT 1

3/4 3-14 Amendment No.

7

TABLE 4.3-1 (Continued)

TABLE NOTATIONS (Continued)

(11)

CHANNEL CALIBRATLON shalL incLude the RTD bypass Loops Elow rate.

(12) Verify that appropriate signals reach the undervoltage and shunt trip

relays, Eor both the main and bypass

breakers, Erom the manual reactor trip switch.

SHEARON HARRIS - UNIT 1

3/4 3"14a Amendment No.

7

3/4e 1 REACTIVITY CONTROL SYSTEMS BASES 3/4. 1 ~ 1 BORATION CONTROL 3/4.1 ~ 1 ~ 1 and 3/4.1.1 ~ 2 SHUTDOWN MARGIN A sufficient SHUTDOWN MARGIN ensures that:

(1) the reactor can be made sub-critical from all operating conditions, (2) the reactivity transients associated with postulated accident conditions are controllable within acceptable limits, and (3) the reactor will be maintained sufficiently subcritical to preclude inadvertent'criticality in the shutdown condition.

SHUTDOWN MARGIN requirements vary throughout core life as a function of fuel depletion, RCS boron concentration, and RCS Tav.

In MODES 1 and 2 the most restrictive condition occurs at EOL, with T v It no load operating temperature, and is associated with a postufaied steam line break accident and resulting uncontrolled RCS cooldown.

In the analysis of this accident, a

minimum SHUTDOWN MARGIN of 1770 pcm is required to control thc reactivity transient.

Accordingly, the SHUTDOWN MARGIN requirement is based upon this limiting condition and is consistent with FSAR safety analysis assumptions In MODES 3, 4, and 5, the most restrictive condition occurs at BOL, when the boron concentration is the greatest.

In these

modes, the required SHUTDOWN MARGIN is composed of a constant requirement and a variable requirement, which is a function of the RCS boron concentration.

The constant SHUTDOWN MARGIN requirement is based on an uncontrolled RCS cooldown from a steamline break

accident, as is the case for MODES 1 and 2.

The variable SHUTDOWN MARCIN requirement is based on the results of boron dilution accident analyses, where the SHUTDOWN MARGIN is varied as a function of RCS boron concentration, to guarantee a minimum of 15 minutes for operator action prior to a loss of SHUTDOWN MARGIN+

Figure 3.1-1 must be used with a curve giving the required shutdown boron concentrations for various temperatures as a function of care burnup.

This cycle dependent relationship is provided for each cycle in the plant Curve Book.

From the Curve Book, a required boron concentration that will provide adequate SHUTDOWN MARCIN can be determined and this concentration may be used to enter Figure 3.1-1 to determine the specific required SHUTDOWN MARGIN for that condition, The boron dilution analysis assumed a coaean RCS volume and dilution flow rate for MODES 3 and 4, which differed from the volume and flow rate assumed for MODE 5 analysis'he MODE 5 conditions assumed limited mixing in the RCS and cooling with the RHR system only.

In MODES 3 and 4, it was assumed that at least one reactor coolant pump was operating.

If at least, one reactor coolant pump is not operating in MODE 4, then the SHUTDOWN MARGIN requirements for MODE 5 shall apply, provided that the dilution flow rate assumed in the MODE 5 Boron Dilution analysis is not exceeded.

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

7

REACTIVITY CONTROL SYSTEMS BASES 3/4.1.1.3 MODERATOR TEMPERATURE COEFFICIENT The limitations on moderator temperature coefficient (MTC) are provided to ensure that the value of this coefficient remains within the limiting condition assumed in the FSAR accident and transient analyses.

The MTC values of this specification are applicable to a specific set of plant conditions; i.e., the positive limit is based on core conditions for all rods withdrawn, BOL, hot zero THERMAL POWER, and the negative limit is based on core conditions for all rods withdrawn,

EOL, RATED THERMAL POWER.

Accordingly, veri-fication of MTC values at conditions other than those explicitly stated will require extrapolation to those conditions in order to permit an accurate comparison.

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

7

l

~0

~ ~,

~

REACTIVITY CONTROL SYSTEMS BASKS MODERATOR TEMPERATURE COEFFICIENT (Continued)

The most negative HTC, value equivalent to the most positive moderator density coefficient (MDC), vas obtained by incrementally correcting the.HDC used in the FSAR analyses to nominal operating conditions'hese corrections involved subtracting the incremental change in the MDC associated with a core condition of all rods inserted (most positive MDC) to an all rods vithdravn condition

and, a conversion for the rate of change of moderator density vith temperature at RATED THERMAL POWER conditions'his value of the HDC vas then transformed into the limiting MTC value -42 pcm/'F.

The MTC value of -33 pcm/'F represents a conservative value (with corrections for burnup and soluble boron) at a core condition of 300 ppm equilibrium boron concentration and is obtained by making these corrections to the limiting MTC val.ue of -42 pcm/'F.

The Surveillance Requirements for measurement of the MTC at the beginning and near the end of the fuel cycl.e are adequate to confirm that the MTC remains within its limits since this coefficient changes slowly due principally to the reduction in RCS boron concentration associated with fuel burnup.

3/4.1.1.4 HI¹HUM TEMPERATURE FOR CRITICALITY This specification ensures that the reactor vill not be made critical with the Reactor Coolant System average temperature Les@ then 551'F.

Thi's Limitation is required to ensure:

(1) the moderator temperature coefficient is vithin it enalyxed temperature

range, (2) the trip instrumentation is within its normaL operating

range, (3) the pressurixer is capable of being in an OPERABLE status with a steam bubble, and (4} the reactor vesseL is above its minimum RTNDT temperature.

3/4.1.2 BORATION SYSTEMS The Boron Injection System ensures that negative reactivity control is available during each mode of facility operation.

The components required to perform this function include.'(1) borated water sources, (2) charging/safety injection

pumps, (3) separate flow paths, (4) boric acid transfer

pumps, and (5) an emergency pover supply from OPERABLE diesel generators.

With the RCS average temperature above 350'F, a minimum of two boron injectioa flow paths are required to ensure single functional capabiLity in the event an assumed failure renders one of the flow paths inoperable.

The boration cape" bility of either flow path is'ufficient to provide the required SHUTDOWN MARGIN as defined by Specification 3/4.1.1.2 after xenon decay and cooldown to 200'F.

The maximum expected boration capability requirement occurs at BOL from full pover equilibrium xenon conditions and requires 21,400 gallons of 7000 ppm berated water be rpaintained" in the boric acid storage tanks or 436,000 gallons of 2000>>2200 ppm borated 'water be maintained in the refueling vater storage tank (RWST).

With the RCS temperature below 350'F, one boron injection flov path is accept" abLe without single faiLure consideration on the basis of the stable reactivity SHEARON HARRIS - UNIT 1 B 3/4 1-2 Amendment No.

7

4 E

~gyp

~

I

c ~

REACTIVITY CONTROL SYSTEMS BASES BORATION SYSTEMS (Continued) condition of the reactor and the additional restrictions prohibiting CORE ALTERATIONS and positive reactivity changes in the event the single boron injection flow path becomes inoperable.

The limitation for a maximum of one charging/safety injection pump (CSIP) to be OPERABLE and the Surveillance Requirement to verify all CSIPs except the required OPERABLE pump to be inoperable below 335'F provides assurance that a

mass addition pressure transient can be relieved by the operation of a single PORV.

The boron capability required below 200'F is sufficient to provide the required SHUTDOWN MARGIN as defined by Specification 3/4.1.1.2 after xenon decay and cooldown from 200'F to 140'F.

This condition requires either 7100 gallons of 7000 ppm borated water be maintained in the boric acid storage tanks or 106,000 gallons of 2000-2200 ppm borated water be maintained in the RWST.

The gallons given above are the amounts that need to be maintained in the tank in the various circumstances.

To get the specified value, each value had added to it an allowance for the unusable volume of water in the tank, allowances for other identified needs, and an allowance for possible instrument error.

In addition, for human factors purposes, the percent indicated levels were then raised to either the next whole percent or the next even percent and the gallon figures rounded off.

This makes the LCO values conservative to the analyzed values.

The specified percent level and gallons differ by less than 0.3X.

The limits on contained water volume and boron concentration of the RWST also ensure a

pH value of between 8.5 and 11.0 for the solution recirculated within containment after a

LOCA.

This pH band minimizes the evolution of iodine and minimizes the effect of chloride and caustic stress corrosion on mechanical systems and components.

The BAT minimum temperature of 65'F ensures that boron solubility is maintained for concentrations of at least the 7750 ppm limit.

The RWST minimum temperature is consistent with the STS value and is based upon other considerations since solubility is not an issue at the specified concentration levels.

The RWST high temperature was selected to be consistent with analytical assumptions for containment heat load.

The OPERABILITY of one Boron Injection System during REFUELING ensures that this system is available for reactivity control while in MODE 6..

3/4.1.3 MOVABLE CONTROL ASSEMBLIES The specifications of'his section ensure'hat!

(1) acceptable power distri-bution limits are maintained, (2) the minimum SHUTDOWN MARGIN is maintained,.

and (3) the potential effects of rod misalignment on associated accident analyses are limited.

OPERABILITY of the control rod position indicators is required to determine control rod positions and thereby ensure compliance with the control rod alignment and insertion limits.

SHEARON HARRIS - UNIT 1 B 3/4 1-3

1

3/4.2 POMER 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 1.30 during normaL operation and in short-term transients, and (2) Limiting the fission gas release, fuel pellet temperature, and cladding mechanical proper-ties to within assumed design criteria.

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.'Q(Z)

F~H N

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 2.32 times the normalired axial peaking factor xs not exceeded I

during either normal operation or in the event of xenon redistribution following power changes.

Target 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 POMER 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 HNER 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.

7

1

'I

~

r-

POWER DISTRIBUTION LIHITS BASES AXIAL FLUX DIFFERENCE (Continued)

At pover levels below APL

, the limits on AFD are defined by Figure 3.2-1, i.e., that defined by the RAOC operating procedure and limits These limits were calculated in a manner such that expected operational transientsf eag 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 vill not result in significant xenon redistribution auth that the envelope of peaking gators would change sufficiently to prevent operation in the vicinity of the APL pover level.

At power levels greater than APL

, two modes of operation are permissible')

DAOC, the AFD limits of which are defined by Pigura 3.2-1, and 2) gene Load operation, which is defined as the maintenanae of tha APD wit/a a 32 band about a target value.

The DAOC sparging proaedure above APL is tha same as that defined for operation below APL

~

Hovever, it is possible vhen 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 allov operation at thc maximum permissible value, the Base Load operating procedure restricts the indioated APD to Ih relatively smaIf target band and power swings (AFD target band of + 3Z, APL

< power

< APL or 100Z Rated Thermal Pover, whichever is lover).

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 there is no residual xenon redistribution impact from past operation on thI Base Load operation, a 24-hour vaiting 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 alloved by the Base Load procedure.

After the vaiting

period, extended Base Load operation is permissible.

The computer determines the one-minute average of each of thc OPERABLE excore detector outputs and provides an alarm message imnediately if the AFD for tvo or more OPERABLE excore channels are:

1) outside the allovcd dI pover operating space (for RAOC operation), or 2) outside the acceptable AFD target band (for Base Load operation)

~

These alarms are active when power g greater than!

1) 50Z 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 vhich operation outside of the target band is allowed.

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

7

POWER DISTRIBUTION LIMITS BASES AXIAL FLUX DIFFERENCE (Continued) 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 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.

7

FIGURE B 3/4 2-1 DELETED SHEARON HARRIS - UNIT 1

B 3/4 2-3 Amendment No.

7

POWER DISTRIBUTION LIMITS BASES HEAT FLUX HOT CHANNEL FACTOR AND RCS FLOW RATE AND NUCLEAR DlTHALPY RISE HOT CHANNEL FACTOR Continued c.

The control rod insertion limits of Specifications 3.1.3 '

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.

a(ove are maintained.

The combinations of the Rch floe rennirrment 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 THEBMAL POWER allows changes in the radial power shape for all permissible rod insertion Limits.

FzH is evaluated as being less than og equal to 1.49.

This value is used in t8e various accident analyses

~here F<~ influences parameters other than

DNBR, e.g.,

peak clad temperature, and thus x,s the maximum "as measured" value alLowed.

Fuel rod bowing reduces the value of DNB ratio.

Credit is available to offset this reduction in the generic margin.

The generic margins, totaling 9.1Z DNBR completely offset any rod bow penalties'he applicablc value of rod bow and any other penalties is presented in FSAR Section 4.4.2.2.4 '.

This margin includes the following:

a.

b.

C ~

d.

Design limit DNBR of 1.30 vs 1.28, Grid Spacing (Ks) of 0.046 vs 0.059, Thermal Diffusion Coefficient of 0.038 vs 0'59, DNBR Multiplier of 0.86 vs 0.88, and Pitch reduction.

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

An allowance of 5Z is appropriate for a full-core map taken with the Incore Detector Flux Mapping System, and a 3Z allowance is appropriate for manufacturing tolerance.

The hot channel factor F~(Z) is measured periodically and increased by a cycLe and height dependent power factor appropriate to either RAOC or Base Load operation, W(Z) or W(Z)~Li to provide assurance that the limit on the hot channeL factor, F~(Z), xs 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 Peaking Factor Limit Report per Specification 6.9.1.6.

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

7

DESIGN FEATURES DESIGN PRESSURE AND TEMPERATURE 5.2.2 The containment buiLding is designed and shall be maintained Eor a maximum internal pressure of 45.0 ps'ig and a peak air.temperature oE 380'F.

5.3 REACTOR CORE FUEL ASSEMBLIES 5.3.1 The coze shall contain 157 fueL assemblies vith each Euel assembly normally containing 264 fuel rods clad with Zircaloy"4 except that Limited substitution oE fuel rods by EiLLer rods consisting of Zircaloy-4, stainless

steel, or by vacancies may be made in fuel assembLies iE justified by a cycle-speciEic evaluation.

ShouLd more than a totaL of 30 fuel rods or more than 10 fuel rods in any one assembly be replaced per refueling a Special Report describing the number of rods z'eplaced vill be submitted to the Commission, pursuant to Specification 6.9.2, vithin 30 days after cycle startup.

Each fuel. rod shalL have a nominal active fuel length of 144 inches'he initial core Loading shall have a maximum enrichment of 3.5 vcight percent U-235 'eload Eucl shaLL be similar in physical design to the initiaL core loading and shall have a maximum enrichment of 4.2 veight percent U>>235.

CONTROL ROD ASSEMBLIES 5.3.2 The core shall contain 52 shutdown and control rod assembLies.

The shutdown and rod assemblies shall contain a nominal 142 inches of absorber materials The nominal values of absorber material shall bc 80Z silver, 15X indium, and 5Z cadmium, or 95Z hafnium vith thc remainder zirconium.

ALL control rods shall be clad vith stainless steel tubing.

5.4 REACTOR COOLANT SYSTEM DESIGN PRESSURE AND TEMPERATURE 5.4.1 The Reactor Cool.ant System is designed and shall be maintained:

a.

In accordance vith the Code requirements specified in Section 5.2 of the FSAR, vith allovance Eor normaL degradation pursuant to the.

applicable Surveillance Requirements, b.

For a pressure of 2485 psig, and c.

For a temperature oE 650'F, except for the pressurizer vhich is 680'F ~

VOLUME 5.4.2 The total water and steam volume of the Reactor Coolant System is 9410

+ 100 cubic feet at a nominaL T v of 588.8'F.

SHEARON HARRIS - UNIT ]

5-6 Amendment No.

I

~

5.5 METEOROLOGICAL TONER LOCATION 5.F 1 The meteorological station shall be located as shown on Figure 5.1-1 ~

SHEARON HARRIS " UNtf' 5-6a Amendment No. 5, 7

ADMINISTRATIVECONTROLS PEAKINC FACTOR LIMIT REPORT 6.9Q.6 The W(Z) Functions for RAOC and Base Load operation end 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 thc NRC for review end 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 (es 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 APL will be by request

from the NRC and need not be included in this report.

SPECIAL REPORTS 6'.2 Special reports shall be submitted to the Regional Administrator 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; 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>

WCAP-10216

~Rela~etion of Constant Axial 0 Technical Specification."

SHEARON HARRIS - UNIT 1 6-24 Amendment No.

7

ADMINISTRATIVE CONTROLS PEAKING FACTOR LIMIT REPORT (Continued) 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.

7

ML18005A549

Text

Attachment:

RESPONSE

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