ML20028C057

Amend 19 to License DPR-77,changing Tech Specs to Accommodate Cycle 2 Reload Operations

ML20028C057
Person / Time
Site:	Sequoyah
Issue date:	12/23/1982
From:	Adensam E Office of Nuclear Reactor Regulation
To:
Shared Package
ML20028C058	List: ML20028C057 ML20028C059 ML20028C060
References
NUDOCS 8301050489
Download: ML20028C057 (26)

Text

1 4

TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-327 SEQUOYAH NUCLEAR PLANT, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE

/vuendment No. 19-License No. DPR-77 1.

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

A.

The application for amendment to the Sequoyah Nuclear Plant, Unit 1 (the facility) Facility Operating License No. DPR-77 filed by the Tennessee Valley Authority (licensee), dated September 17, 1982, com-plies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's regulations as set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the license, as amended, the provisions of the Act, and the rules and regulations of the Com-mission; C.

There is reasonable assurance (1) 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 Commission'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 hereby amended by page changes to the Appendix A Technical Specifications as indicated in the attachments to this license amend-ment and paragraph 2.C.(2) of Facility Operating License No. DPR-77 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 19,-are hereby incorporated into the license.

i '8301050499 821223

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navn p OFFIClAL RECORD COPY osomien aas.

NRC FORM 318 (440} NRCM ONO

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• The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license anendment is effective as of its date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION 1

i Elinor G. Adensam, Chief Licensing Branch No. 4 Division of Licensing

Attachment:

Appendix A Technical Specification Changes Date of Issuance: December 23, 1982 s

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Attachment to License Amendment No. 19 Facility Operating License No.-DPR-77 Docket No. 50-327

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Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages.

The revised pages are identified by Amendment number and contain vert.ical lines indicating the area of change.

Anended Page 2-2 2-7 2-8 2-9 2-10 B 2-1 3/4 2-1

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REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS i

S!

E FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES-b 21.

Turbine Imp 01se Chamber Pressure -

< 10% Turbine Impulse

< 11% Turbine Impulse (P-13) Input to Low Power Reactor Trips Pressure' Equivalent Pressure Equivalent Block P-7 22.

Power Range Neutron Flux - (P-8) Low

< 35% of RATED

< 36% of RATED Reactor Coolant Loop Flow, and Reactor THERMAL POWER THERMAL POWER Trip 23.

Powdr Range Neutron Flux - (P-10) -

> 10% of RATED

> 9% of RATED Enable Block.of Source, Intermediate, THERMAL POWER THERMAL POWER and Power Range (low setpoint) Reactor Trips 24.

Reactor Trip P-4 Not Applicable Not Applicable 25.

Power Range Neutron Flux - (P-9) -

< 50% of RATED

< 51% of RATED Blocks Reactor Trip for Turbine THERMAL POWER THERMAL POWER Trip Below 50% Rated Power NOTATION I

I NOTE 1:

Overtemperature AT (

) < AT, {K) - K2 (1 + T 6)[T(

)-T'] + K (P-P') - f (AI)}

2 3

j 1+tS 1+T6 I*T0 j

3 4

i jf

= Lag comp,ensator on measured AT where:

F l.

T j

= Time constants utilized in the lag compensator for AT *1 = 2 secs.

3 AT

= Indicated AT at RATED THERMAL POWER E

K

-< 1.15 1

K

= 0.011 2

C

r TABLE 2.2-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS NOTATION (Continued) g NOTE 1:

(Continued)

Z 1+T5 2

I*I3 The function generated by the lead-lag controller for T,yg dynamic compensation

=

3 Time constants utilized in the lead-lag controller for T,yg, T2 = 33 secs.,

r'&T

=

2 3

3 = 4 secs.

1 T

Average temperature F

=

1+TS Lag compensator on measured T,yg

=

4 4

Time constant utilized in the measured T,yg lag compensator, T = 2 secs.

T

=

4 T'

5 578.2 F (Nominal T,yg at RATED THERMAL POWER)

K

=

0.00055 3

P

= -Pressurizer pressure, psig P'

= 2235.psig (Nominal RCS operating pressure)

Laplace transform operator '(sec-I)

S

=

and f (AI) is a function of the indic~ated difference between top and bottom detectors 3

y of the power-range nuclear ion chambers;.with gains to be selected based on measured g

instrument response during plant startup tests such that:

a.

2 (i) arephrcenkRATEDTHERMALPOWERinthetopandbotlo(AI)=0(whereqborereIpectively, for q q between - 29 percent and + 5 percent f and q 5

m halves of the is total THERMAL POWER in percent of RATED THERMAL POWER).

and qt + 9b 0

e

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TABLE 2.2-1 (Continued) v.

h REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS S! '

!!E NOTATION (Continued)

E NOTE 1:

(Continued) s (ii) for each percent that the magnitude of (q q

exceeds -29 percent, the AT trip set-pointshallbeautomaticallyreducedbyIf50pNr)centofitsvalueatRATEDTHERMALPOWER.

(iii) for each percent that the magnitude of (q q

exceeds +5 percent, the AT trip set-pointshallbeautomaticallyreducedbyOf86pNr)centofitsvalueatRATEDTHERMALPOWER.

I S

I NOTE 2:

Overpower AT (

) 5 AT, {K4 5 (1 + t S)(

) T -K U(

) - T9 - f fAI)}

-K S

6 2

]

1+tS 1+T5 1+TS j

3 4

4 Where:

as defined in Note 1 s

=

3 as defined in Note 1 T

=

j AT, as defined.in Note 1

=

K F 1.087 4

K

= 0.02/*F for-increasing average temperature and 0 for decreasing average 5

temperature i

Tb 5

,h, The function generated by.the rate-lag controller for T,yg dynamic

=

1+rS5 compensation 5

5 4

r TABLE 2.2-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS Y

z NOTATION (Continued) e E

U NOTE 2:

(Continued) 5 Time constant utilized in the rate-lag controller for T,yg, 15

• I0 secs.

1

=

I as defined in Note 1

=

j,73 as defined in Note 1 T

=

4 0.0011 for T > T" and K = 0 for T $ T" K

=

6 6

g as defined in Note 1 T

=

T" Indicated T,yg at RATED THERMAL POWER (Calibration temperature for

=

AT instrumentation, 5 578.2*F)

S.

as defined in Note 1

=

f (AI) 0 for all AI

=

2 NOTE 3: The channel's maximum trip'setpoint shall not exceed its computed trip point by more than l

{

2 percent.

1 a

A

4 t

2.1 SAFETY LIMITS BASES 2.1.1 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 coefficient.

DNB is not a directly measurable parameter during operation and therefore THERMAL POWER and Reactor Coolant Temperature and Pressure have been related to DNB through the W-3 correlation.

The W-3 DNB correlation has been developed to predict the DNB flux and the location of DNB for axially uniform and non-uniform heat flux distributions.

The local DNB he's flux ratio, DNBR, defined as the ratio of the heat flux that would cause DNb o.

, articular core location to the local heat flux, is indicative of the margin to LNB.

v The minimum value of the DNBR during steady state operation, normal operational transients, and anticipated transients is limited to 1.30.

This value corresponds to a 95 percent probability at a 95 percent confidence level that DNB will not occur and is chosen as an appropriate margin to DNB'for all operating conditions.

I i

The curves of Figures 2.1-1 and 2.1-2 show the loci of point's of THERMAL POWER, Reactor Coolant System pressure and average temperature for which the minimum DNBR is no less than 1.30, or the average enthalpy at the vessel exit is equal to the enthalpy of' saturated liquid.

The'curvesarebasedonanenthalpyhotchannelfactor,Fh,of1.55and a reference cosine with a peak of 1.55 for axial power shape.

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

F

= 1.55 [1+ 0.3 (1-P)]

where P is the fra.ction of RATED THERMAL POWER l

SEQUOYAH - UNIT 1 B 2-1 Amendment No. 19

3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE (AFD)

LIMITING CONDITION FOR OPERATION 3.2.1 The indicated. AXIAL FLUX DIFFERENCE (AFD) shall be maintained within the allowed operational space defined by Figure 3.2-1.

APPLICABILITY:

MODE 1 above 50% RATED THERMAL POWER

• ACTION:

With the indicated AXIAL FLUX DIFFERENCE outside of the Figure 3.2-1 a.

limits; 1.

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

Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 30 minutes and reduce the Power Range Neutron Flux-High Trip setpoints to less than or equal to 55 percent 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 />.

b.

THERMAL POWER shall not be increased above 50% of RATED THERMAL

=

POWER unless the indicated AFD is within the Figure 3.2-1 limits.

s

(.

SEQUOYAH - UNIT 1 3/4 2-1 Amendment No. 19 m

POWER DISTRIBUTION LIMITS t

SURVEILLANCE REQUIREMENTS 4.2.1.1 The indicated AXIAL FLUX DIFFERENCE'shall be determined to be within its limits during POWER OPERATION above 50% of RATED THERMAL POWER by:

a.

Monitoring the indicated AFD for each OPERABLE excore channel:

1.

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 <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after restoring the AFD Monitor Alarm to OPERABLE status.

b.

Monitoring and logging the indicated AXIAL FLUX DIFFERENCE 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 AXIAL FLUX DIFFERENCE Monitor Alarm is inoperable.

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

e 4.2.1.2 The indicated AFD shall be considered outside of its limits when at least 2 OPERABLE excore channels are indicating the AFD to be outside the limits.

..g f

SEQUOYAH - UNIT 1 3/4 2-2 Amendment No. 19

POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS-(Continued) i 1

This page left blank intentionally.

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SEQUOYAH - UNIT 1 3/4 2-3 Amendment No 19

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-50

-40

-30

-20

-10 0

10 20 30 40 50 Flux Difference (t.I):

FIGURE 3.2-1 AXIAL FLUX DIFTE9ENCE LIMITS AS A FUNCTION OF RATED THERMAL POW (TYPICAL EXAMPLE)

I SEQUOYAH UNIT 1 3/4 2-4 Amendment No. 19,

a :.

?) t'

> (.

1.

t

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' POWER DISTRIBUTION LIMITS 3/4.2.2 HEATFLUXHOTCHANNELFACTOR-Fg LIMITING CONDITION FOR OPERATION 3.2.2 F (Z) shall be limited by the following relationships:

q F (Z) 5 [2.237] [K(Z)] for P > 0.5 0

P F (Z) 5 [2.237] [K(Z)] for P 5 0.5 9

0.5 THERMAL POWER where P =

RATED THERMAL POWER and K(Z) is the function obtained from Figure 3.2-2 'or a given core height location.

APPLICABILITY:

MODE 1 ACTION:

With F (Z) exceeding its limit:

q a.

Reduce THERMAL POWER at least 1% for each 1% F (Z) exceeds the limit q

within 15 minutes and similarly reduce the Power Range Neutron Flux-High Trip Setpoints within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; 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 Delta T Trip Setpoints (value of K ) have been reduced at least 1% (in AT span) for each 1% F (Z) 4 q

exceeds the limit.

b.

Identify and correct the cause of the out of limit condition prior to increasing THERMAL POWER; THERMAL POWER may then be increased provided F (Z) is demonstrated through incore mapping to be within~

9 its limit.

SURVEILLANCE REQUIREMENTS 4.2.2.1 The provisions of Specification 4.0.4 are not applicable.

SEQUOYAH - UNIT 1 3/4 2-5 Amendment No.19

9 POWER DISTRIBUTION LIMITS

+

SURVEILLANCE REQUIREMENTS (Continued) 4.2.2.2 F (z) shall be evaluated to determi'ne if F (Z) is within.its q

q limit by:

a.

Using the movable incore detectors to obtain a power distribu-tion map at any THERMAL POWER greater than 5% of RATED THERMAL POWER.

b.

Increasing the measured Fq g) component of the power distribution map by 3 percent to account for manufacturing tolerances and further increasing the value by 5% to account for measurement uncertainties.

c.

Satisfying the following relationship:

F "(z) < 2.237 x K(z) for P > 0.5 Q

P x W(z)

FQ (z) < W(z) x 0.5x K(z) for P < 0.5 M

2.237 M

where F (z) is the measured F (z) increased by the allowances for q

q manufacturing tolerances and measurement uncertainty, F limit is q

the F limit, K(z) is given in Figure 3.2-2, P is the relative q

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 6.9.1.14.

N d.

Measuring Fq (z) according to the following schedule:

1.

Upon achieving equilibrium conditions after exceeding by 10 percent or more of RATED THERMAL POWER, the THERMAL POWER at which F (z) was last determined,* or q

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.

1 SEQUOYAH - UNIT 1 3/4 2-6 Amendment No. 19

5 POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued) e.

With measurements indicating M(z)\\

[

I F

maximum k

K(z) over z has increased since the previous determinatin of F M(z) either of the following actions shall be taken:

9 N(z) shall be increased by 2 percent over that specified in 1.

Fq 4.2.2.2.c, or N(z) shall be measured at least once per 7 effective full 2.

Fq power days until 2 successive maps indicate that F" (z) maximum is not increasing.

K(z) over z

(

j f.

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

1.

Calculate the percent F (z) exceeds its limit by the following q

expression:

F "(z) x W(z) aximum n

-1 x 100 for P > 0.5 1 over z 2.237 x K(z) 4

'I Fg (z) x W(z)

-1 fx100 for P < 0.5 maximum ver z 2

7 x K(z)

/

2.

Either of the following actions shall be taken:

a.

Place the core in an equilibrium condition where the limit in 4.2.2.2.c is satisfied.

P~ower level may then be increased provided the AFD limits of Figure 3.2-1 are reduced 1% AFD for each percent F (z) exceeded its limit, q

or b.

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

q SEQUOYAH - L,.T 1 3/4 2-7 Amendment No. 19

POWER DISTRIBUTION-LIMITS SURVEILLANCE REQUIREMENTS (Continued) g.

The limits-specified in 4.2.2.2.c, 4.2.2.2.e, and 4.2.2.2.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.

4.2.2.3 When F (z) is measured for reasons other than meeting the requirements q

of Specification 4.2.2.2 an overall measured F (z) shall be obtained from a q

power distribution map and increased by 3 percent to account for manufacturing J

tolerances for further increased by 5 percent to account for measurement uncertainty.

v f

t l

SEQUOYAH - UNIT 1 3/4 2-8 Amendment No.

19 4

,,.n.

- - - + - - -.

.=

9 POWER DISTRIBUTION LIMITS 3/4.2.3 RCS FLOWRATE AND R LIMITING CONDITION FOR OPERATION 3.2.3 The combination of indicated Reactor Coolant System (RCS) total flow rate and R), R2 shall be maintained _within the regions of allowable operation shown on Figure 3.2-3 for 4 loop operation:

Where:

N a*

R

=

1 1.49 [1.0 + 0.3 (1.0 - P)]

b.

R 2

[1 - BP (Bu)]

p THERMAL POWER RATED THERMAL POWER '

d.

F

=

Measured values of F H

H btained by using the movable incore detectors to obtain a power distribution map.

ThemeasuredvaluesofFhshallbeusedtocalculate R since Figure 3.2-3 includes measurement uncertainties of 3.5% for flow and 4% for incore measurement of F and e.

RBP (Bu) =

Rod Bow Penalty as a function of region average burnup as shown in Figure 3.2-4, where a region is defined as those assemblies with the same loading date (reloads) or enrichment (first core).

l APPLICABILITY: MODE 1 ACTION:

With the combination of RCS total flow rate and R, R utside the regions j

2 of acceptable operation shown on Figure 3.2-3:

a.

Within 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />s:

1 l

1.

Either restore the combination of RCS total flow rate and R, R to within the above limits, or j

2 2.

Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER and reduce the Power Range Neutron Flux - High trip setpoint 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 />.

t SEQUOYAH - UNIT 1 3/4 2-10 Am.endment No. 19

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48

,,,,,,g g,,,,,,,ui w a n u n u t u n +o i n n i e.o i n n e o u + n e i + +:

. MEASUREMENT UNCERTAINTIES OF

[

3.5% FOR FLOW AND 4% FOR INCORE e

MEASUREMENT OF Fyg ARE INCLUDED

.E IN THIS FIGURE.'

46 ACCEPTABLE-OPERATION REGION FOR R,ONLY 2

. UNACCEPTABLE j g

OPERATION O

REGION i

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e g

42

ACCEPTABLE OPERATION g

g y-P.EGION FOR i

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R&R (1.029,40.6$)

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' UNACC$PTABLE f

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l OPE R ATION '

REGION 36 0.90 0.92 0.94 0.96 0.98 1.00 1.02 1.04 1.06 R, = F5g/1.49[1.0 + 0 311.0 - P)]

R = R /[1 - RBP(Bu)]

2 j

to E

FIGURE 3.2-3 RCS Total Flowrate Versus R and R2 - Four Loops in Operation 3

.U C

l INSTRUMENTATION MOVABLE INCORE EETECTORS LIMITING CONDITION FOR OPERATION 3.3.3.2 The movable incore detection system shall be OPERABLE with:

a.

At least 75% of the detector thimbles, b.

A minimum of 2 detector thimbles per core quadrant, and c.

Sufficient movable detectors, drive, and readout equipment to map these thimbles.

APPLICABILITY: When the movable incore detection system is used for:

a.

Recalibration of the excore neutron flux detection system, b.

Monitoring the QUADRANT POWER TILT RATIO, or c.

Measurement of F and F (Z).

g q

ACTION:

With the movable incore detection system inoperable, do not use the system for w

the above applicable monitoring or calibration functions.

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.3.3.2 The movable incore detection system shall be demonstrated OPERABLE by normalizing each detector output when required for:

a.

Recalibration of the excore neutron flux detection system, or.

b.

Monitoring the QUADRANT POWER TILT RATIO, or c.

Measurement of F and F (Z).

H q

v SEQUOYAH - UNIT 1 3/4 3-43 Amendment No.19

..-_-__A---_.----..--_--..----------------.-----.------

t 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:

(a) maintaining the calculated DNBR in the core at or above design during normal operation and_in short term transients, and (b) limiting the fission gas release, fuel pellet temperature and cladding mechanical properties 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 t

these specifications are as follows:

l F (Z)

Heat Flux Hot Channel Factor, is defined as the maximum local 0

heat flux on the surface of a fuel rod'at core elevation Z divided by the average fuel rod heat flux, allowing for manufacturing l

tolerances on fuel pellets and rods.

N i

F Nuclear Enthalpy Rise Hot Channel Factor is defined as the ratio of the iNegraloflinearpoweralongtherodwiththehighestintegratedpowerto e

l the average rod power.

3/4.2.1 AXIAL FLUX DIFFERENCE (AFD)

The limits on AXIAL FLUX DIFFERENCE assure that the F (Z) upper bound q

l envelope of 2.237 times the normalized exial peaking factor is not exceeded l

l during either normal operation or in the event of xenon redistribution follow-ing power chang'es.

i l

Provisions for monitoring the AFD on an automatic basis are derived from the plant process computer through the AFD Monitor Alarm.

The computer deter-mines the one minute average of each of the OPERABLE excore detector outputs and provides an alarm message immediately if the AFD for at least 2 of 4 or 2 l

of 3 OPERABLE excore channels are outside the allowed AI-Power operating space l

and the THERMAL POWER is greater than 50 percent of RATED THERMAL POWER.

i 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR, RCS FLOWRATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR The limits on heat flux hot channel factor, RCS flowrate, 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.

SEQUOYAH - UNIT 1 B 3/4 2-1 Amendment No.19 i

POWER DISTRIBUTION LIMITS BASES 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 insure that the limits.are maintained provided:

a.

Control rods in a single group move together with no individual rod insertion differing by more than + 13 steps from the group demand position.

b.

Control rod groups are sequenced with overlapping groups as described in Specification 3.1.3.6.

c.

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

d.

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

F will b'e maintained within its limits provided conditions a. through H

d. above are maintained.

As noted on Figures 3.2-3 and 3.2-4, RCS' flow.

and F may be " traded off" against one another to ensure that the calculated-DNBR will not be below the design _DNBR value.

The relaxation of F as a H

function of THERMAL POWER allows changes in the radial power shape for all permissible rod insertion limits.

WhenRCSflowrateandFharemeasured,noadditionalallowancesare necessary prior to comparison with the limits of Figures 3.2-3 and 3.2-4.

Measurement errors of. 3.5 percent for RCS total flow rate and 4 percent for Fh have been allowed for in determination of the design DNBR value.

R), as calculated in Specification 3.2.3 and used in Figure'3.2-3, accounts for F less than'or equal to 1 49. This value is the value used in the H

various safety analyses where F influences parameters other than DNBR, e.g.

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

R, as defined, allows-for the inclusion of a penalty for Rod Bow on DNBR 2

only.

Thus, knowing the "as measured" values of F and RCS flow allow for H

" trade off" in excess of R equal to 1.0 for the purpose of offsetting the Rod

. Bow DNBR penalty.

SEQUOYAH - UNIT 1 B 3/4 2-2 Amendment No. 19

_____________A___.___________

+

POWER DISTRIBUTION LIMITS f

t

/

THIS FIGURE DELETED 1

s 1

v 1

i Figure B 3/4.2-1 TYPICAL INDICATED AXIAL FLUX DIFFERENCE VERSUS THERMAL POWER SEQUOYAH'- UNIT 1 B 3/4 2-3 Amendment No. 19

POWER DISTRIBUTION LIMITS BASES The penalties applied to F to account for Rod Bow (Figure 3.2-4) as a H

function of burnup are, consistent with those described in Mr. John F. Stolz's (NRC) letter to T. M. Anderson (Westinghouse) dated April 5, 1979 and W 8691 Rev. 1 (partial rod bow test data).

When an F reasurement is taken, both experimental error and manufacturing q

tolerance must be allowed for.

5 percent is the appropriate allowance for a full core map taken with the incore detector flux mapping system and 3 percent is the appropriate allowance for manufacturing tolerance.

M The hot channel factor Fq (z) is measured periodically-and increased by a cycle and height dependent power factor, W(z), to provide assurance that the limit on the hot channel factor, F (z), is met.

W(z) accounts for the effects q

of normal operation transients and was determined from expected power control maneuvers over the full range of burnup conditions in the core.

The W(z) function for normal operation is provided in the Peaking Factor Limit Report per Specification 6.9.1.14.

3/4.2.4 QUADRANT POWER TILT RATIO The quadrant power tilt ratio limit. assures that the radial power distri-bution satisfies the design values used in the power capability analysis.

Radial power distribution measurements are made during startup testing and periodically during power operation.

The two hour time allowance for operation with a tilt condition greater than 1.02 but less than-1.09 is provided to allow identification and cor-rection of a drcpped or misaligned rod.

In the event such action does not correct the tilt, the margin for uncertainty on F is reinstated by reducing n

the power by 3 percent from RATED THERMAL POWER f5r each percent of tilt in excess of 1.0.

(,

SEQUOYAH - UNIT 1 b 3/4 2-4 Amendment No. 19

t POWER DISTRIBUTION LIMITS BASES o

f 3/4.2.5 DNB PARAMETERS The limits on the DNB related parameters assure that each of the para-meters are maintained within the normal steady state envelope of operation-assumed in the transient and accident analyses. The limits are consistent with the initial FSAR assumptions and have been analytically demonstrated adequate to maintain a minimum DNBR of 1.3 throughout each analyzed transient.

The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> periodic surveillance of these parameters through instrument readout is sufficient to ensure that the parameters are restored within their limits following load changes and other expected transient operation.

f 1,

l

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(

i l

6 l

l.

SEQUOYAH - UNIT 1 8 3/4 2-5 Amendment No. 19

. ADMINISTRATIVE CONTROLS e.

An unplanned offsite release of 1) more than 1 curie of radicactive material in liquid effluents, 2) more than 150 curies of noble gas in, gaseous effluents, or 3) more than 0.05 curies of radioiodine in gaseous effluents. The report of an unplanned offsite release of radioactive material shall. include the -following information:

1.

A description of the event and equipment involved.

2.

Cause(s) for the unplanned release.

3.

Actions taken to prevent recurrence.

4.

Consequences of the unplanned release.

f.

Measured levels of radioactivity in an environmental sampling medium determined to exceed the reporting level values of Table 3.12-2 when averaged over any calendar quarter sampling period.

RADIAL PEAKING FACTOR LIMIT REPORT 6.9.1.14 The W(z) function for normal operation shall be provided to the Direc-tor, Nuclear Reactor Regulation, Attention, Chief of the Core Performance Branch, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555 at least 60 days prior to cycle initial criticality.

In the event that these values would be submitted at some other time during core life, it will be submitted 60 days prior to the date the values would become effective unless otherwise exempted by the Commission.

Any information.needed to suport W(z) will be by request from the NRC and need not be included in this report.

SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the Director of the Office of Inspection and Enforcement Regional Office within the time period specified for each' report.

6.10 RECORD RETENTION 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.

6.10.1 The following records shall be. retained for at least five 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 OCCURRENCES submitted to the Commission.

d.

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

e.

Records of changes made to the procedures required by Specification 6.8.1 and 6.8.4.

(/

f.

Records of radioactive shipments.

g.

Records of sealed source and fission detector leak tests and results.-

h.

Records of annual physical inventory of all sealed source material of record.

SEQUOYAH. UNIT 1 6-22 Amendment No. 19

-