Text

Amends 223 & 214 to Licenses DPR-77 & DPR-79,respectively, Revising Certain TS & Bases to Allow for Conversion of W Fuel to Framatome Cogema Fuel & Incorporating New License Conditions 2.C.(25) & 2.C.(18)
	ML20137Y883
Person / Time
Site:	Sequoyah
Issue date:	04/21/1997
From:	Hebdon F; NRC (Affiliation Not Assigned)
To:	;
Shared Package
ML20137Y890	List: ML20137Y883; ML20137Y891;
References
	NUDOCS 9704230268
	Download: ML20137Y883 (59)
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UNITED STATES

)

NUCLEAR REGULATORY COMMISSION r

2 WASHINGTON, D.C. 3066H001 TENNE 5SEE VALLEY AUTHORITY DOCKET NO. 50-327 SE000YAH NUCLEAR PLANT. UNIT 1 P

AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 223 License No. DPR-77 1.

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

i A.

The application for~ amendment by Tennessee Valley Authority (the licensee) dated April 4, 19.96, as supplemented by letters dated January 10, February 7, February 13, March 17, March 19, March 20, March 25, April 1, April 6, April 10, April 11, and April 18, 1997, 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; B.

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

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

9704230268 970421 PDR ADOCK 05000327 P

pm a

t-i 2.

Accordingly, the license is amended by changes to the Tech'nical-Specifications as indicated in the attachment to this license amendment.

j Paragraph 2.C.(2) of Facility Operating License No. DPR-77 is hereby amended to read as follows:

-(2) Technical Soecifications i-(

i The Technical Specifications contained ~in Appendices A and B, as j

i revised through Amendment No. 223, are hereby incorporated in-the license.

The licensee shall operate the facility in accordance with

{

the Technical Specifications.

I l

A new License Condition, Paragraph 2.C.(25), is added as follows:

~

(25) Mixed Core DNBR Penalty 5

1 TVA will obtain NRC approval prior to startup for any cycle's core that involves a reduction in the departure from nucleate boiling ratio initial transition core penalty below that value stated in l

TVA's submittal on Framatome fuel conversion dated April 6,1997.

3.

This license amendment is effective as of its date of issuance, to be j

implemented no later than 45 days of its issuance.

j FOR THE NUCLEAR REGULATORY COMMISSION I

gh Frederick J. Hebdon, Director f

i Project Directorate 11-3 i

Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation

Attachment:

1.

Page 12 of license 2.

Changes to the Technical Specifications Date of Issuance: April 21, 1997

Page 12 of the composite License is attached to reflect this change.

(2)

TVA ch311 maintain int: rim omsrgancy support facilities (Technical support Center, operation's Support Center and the Emergency operations Facility) until the final facilities are complete.

l 1.

Relief and Safety Valve Test Recuir;;;nts (Section 22.2, i

II.D.1) a i

TVA shall conform to the results of the EPRI test program.

TVA shall provide documentation for qualifying (a) reactor j

coolant system relief and safety valves, (b) piping and i

supports, and (c) block valves'in accordance with the

{-

review schedule given in SECY 81-491 as approved by the commission.

]

(24) Comoliance with Reculatory Guide 1.97 TVA shall Laplement modifications necessary to comply with Revisior. 2 of Regulatory Guide 1.97, " Instrumentation for Light Water Cooled Nuclear Power Plants to Assess Plant Conditions During and Following An Accident," dated December 1980 by startup from the Unit 2 Cycle 4 refueling outage.

i (25) Mixed Core DNBR Penaltv TVA will obtain NRC approval prior to startup for any cycle's core that involves a reduction in the. departure from nucleate i

boiling ratio initial transition core penalty below that value stated in TVA's submittal on Framatome fuel conversion dated April 6, 1997.

lAE i

j D.

Exemptions from certain requirements of Appendices G and J'to 10 CFR j

Part 50 are described in the office of Nuclear Reactor Regulation's i

Safety Evaluation Report, Supplements No. 1.

These exemptions are authorized by law and will not endanger. life or property or the common defence and security and are otherwise in the public interest.

The exemptions are, therefore, hereby granted. The granting of these g

exemptions are authorized with the issuance of the License for Fuel Loading and Low Power Testing, dated February 29, 1980.

The facility will operate, to the extent authorized herein, Act, and the R77 regulations of the Commission. Additional exemptions are listed in attachment 1.

s E.

Physical Protection The licensee shall fully implement and maintain in effect all provisions of the Commission approved physical security, guard training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revision to 10 CFR 73.55 (51 FR 27817 and 27822) and to the autnority of 10 CFR 50.90 and 10 CFR 50.54(p).

The Safeguards contingency Plan is incorporated into the Physical Security Plan. The plans, which contain Safeguards Information protected under 10 CFR 73.21, are entitled: "Sequoyah Physical Security Plan," with revisions submitted through November 23, 1987; and "Sequoyah Security Personnel Training and Qualification Plan," with revisions submitted through April 16, 1987.' Changes made in accordance with 10 CFR 73.55 shall be implemented in accordance with the schedule set forth therein.

Unit 1 Amendment No. 10, 53, 73, 193, 200, 213, 223 Amendment No. 73 June 24, 1988

_ =

ATTACHMENT TO LICENSE AMENDMENT NO. 223 FACILITY OPERATING LICENSE NO. DPR-77 DOCKET NO. 50-327 Revise the Appendix A Technical Specifications by removing the pages l

identified below and inserting the enclosed pages. The revised pages are identified by the captioned amendment number and contain marginal lines indicating the area of change.

1 REMOVE INSERT 2-5 2-5 l

2-8 2-8 2-9 2-9 2-10 2-10 B 2-1 B 2-1 B 2-4 B 2-4 i

B 2-5 B 2-5' i

3/4 2-5 3/4 2-5 j

3/4 2-6 3/4 2-6 3/4 2-7 3/4 2-7 3/4 2-8 3/4 2-8 3/4 2-10 3/4 2-10 3/4 2-11 3/4 2-11 1

3/4 2-11a 3/4 2-11b 3/4 2-12 3/4 2-12 3/4 2-16 3/4 2-16 3/4 2-17 B 3/4 1-4 B 3/4 1-4 B 3/4 2-1 B 3/4 2-1 B 3/4 2-2 B 3/4 2-2 B 3/4 2-4 8 3/4 2-4 B 3/4 3-2 B 3/4 3-2 6-21 6-21 6-21a e

j

.~..... -

~

TABLE 2.2-1

[

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES

1. Manual Reactor Trip Not Applicable Not Applicable l

2.

Power Range, Neutron Flux Low Setpoint - s 25% of RATED Low Setpol'nt - s 27.4% of RATED

.lR145 THERMAL POWER THERMAL POWER' i

High Setpoint - s 109% of RATED High Setpoint - s 111.4% of RATED lR145 THERMAL POWER THERMAL POWER -

6

3. Power Range, Neutron Flux, s 5% of RATED THERMAL POWER with s 6.3% of RATED THERMAL POWER High Positive Rate a time constant a 2 second with a time constant a 2 second 4.

Power Range, Neutron Flux, s 5% of RATED THERMAL POWER with

.s 6.3% of RATED THERMAL POWER High Negative Rate

'a time constant a 2 second with a time constant a 2 second 5.

Intermediate Range, Neutrori s 25% of RATED THERMAL POWER s 45.20% of RATED THERMAL POWER lR189 Flux

6. Source Range, Neutron Flux s 10 counts per second s 1.45 x 10' counts per second lR199

7. Overtemperature AT See Note 1 See Note 3

8. Overpower AT See Note 2 See Note 4-l 9.

Pressurizer Pressure--Low a 1970 psig a 1964.8 psig 2385 psig s 2390.2 psig' R145

10. Pressurizer Pressure--High s

l 92% of instrument span s 92.7% of instrument span

11. Pressurizer WA

'avel--High s

12. Loss of Flow a 90% of design flow a 89.6% of design flow R225l per loop *

. per loop *

Design flow is 90,045 (87,000 X 1.035) gpm per loop.

l l

i h

t SEQUOYAH - UNIT 1 2-5 Amendmend No. 44, 141, 185, 221,223 I

t A

--e

+- =

se

TABI-E 2. 2-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS NOTATION (Continued)

NOTE 1:

(Continued) t 1+rS t

i The function generated by the lead-lag controller,for T dynamic compensation.

=

1+ rS Time constants utilized in the lead-lag controller for T t

& r

=

1, 2

7 s4 seca,

avg, a 33 secs.,

R215 r

T Average temperature 'F

=

T' s

578.2*F (Nominal T at RATED TRERMAL POWER) avg K

0.00055

=

3 P

=

Pressurizer pressure, psig P'

2235 psig (Nominal RCS operating pressure)

=

S Laplace transform operator (sec )

=

~

and f (AI) is a function of the indicated difference between top and bottom detectors of the power-range nuclear ion chambers; with gains to be selected based on measured instrument response during plant startup tests such that:

(i) for q - g between QTNL* and QTPL* f (AI)

=0 (where q and q b

b l

are percent RATED THERMAL POWER in the top and bottom halves of the core respectively, and q + q is total THERMAL POWER in percent of RATED THERMAL POWER).

b I

4 SEQUOYAH - UNIT 1 2-8 Amendment No. 19, 141, 211, 2?3 L-m

. m m.

m

-mm m

. m.

i TABLE 2.2-1 (Continued)

REACTOR TRIP SYSTEM 'NSTRUMENTATION TRIP SETPOINTS I

NOTATION (Continued)

NOTE 1:

(Continued)

(ii) for each percent that the magnitude of (q - q) exceeds QTNL*, the AT trip set-point shall be automatically reduced by QTNS* of its value at RATED THERMAL POWER.

(iii) for each percent that the magnitude of (q - q) ex eeds QTPL*, the AT trip set-b point shall be automatically reduced by QTPS' of its value at RATED THERMAL PONER.

8 NOTE 2:

Overpower AT (1 + # 8) s AT,{K 4

-K

(

1

)T-K (T-T") - f (AI))

1

+T S

1 + vS S

3

R145, 1+# 8 Where:

4

=

as defined in Note 1 1

1 +7 S5 as defined in Note 1 r

,7

=

5 AT as defined in Note 1

=

o K

s 1.08.7 lR118i 4

K a.0.02/*F for increasing average temperature and 0 for decreasing average lR215; temperature r3S 1 +rS the function generated by the rate-lag controller for T dynamic

=

3 compensation 8V9 R145(

i

QTNL, QTPL, QDIS, and QTPS are specified in the COLR per Specification 6.9.1.14.

l

[

t i

f i

SEQUOYAH - UNIT 1 2-9 Amendment No. 19, 114, 141, 211 223 i

(

TAB 1-E 2.2-1 (Continued)

REACTOR TRIP SYSTEM INSTRtMENTATION TRIP SETPOINTS l

NOTATION.(Continued)

NOTE 2:

(Continued) i Time constant utilized in the rate-lag controller for T r

r = 10 secs.

'D, R215

- K a 0.0011 for T > T" and K = 0 for T s T*

as defined in Note 1 T

=

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

=

avg AT instrumentation, s 578.2*F) as defined in Note 1 S

=

and f ( AI) is m. function of the indicated difference between top and bottom detectors of the power-3 range nuclear ion chambers; with gains to be selected based on measured instrument response during plant startup tests such that:

(i) for g - g between QPNL* and QPPL* f,(AI)

=0 (where g and g are percent RATED THERMAL POWER in the top and bottom halves of the core respectively, and g + g is total THERMAL POWER in percent of RATED THERMAL POWER).

1 (ii) for each percent that the snagnitude of (g - g) exceeds QPNL* the AT trip setpoint shall be automatically reduced by QPNS* of its value at RATED THERMAL POWER.

(iii) for each percent that the magnitude of (g - g) exceeds QPPL* the AT trip setpoint shall be automatically reduced by QPPS* of its value at RATED THERMAL POWER.

NOTE 3:

- The channel's maximum trip setpoint shall not exceed its computed trip point by more than 1.9 percent AT span.

NOTE 4 :

The channel's maximum trip setpoint shall not exceed its computed trip point by more than 1.7 percent AT span.

e

QPNL, QPPL, QPNS, and QPPS are specified in the COLR per Specification 6.9.1.14.

l SEQUOYAH - UNIT 1 2-10 Amendment No. 19, 141, 211, 223

2.1 SAFETY LIMITS BASES 2.1.1 REACTOR CORE l

The restrictions of this safety limit prevent overheating of the fuel and possible cladding perforation which would result in the release of fission i

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.

l Operation above the upper boundary of the nucleate boiling regime could result in excessive cladding temperatures because of the onset of departure i

from nucleate boiling (DNB) and the resultant sharp reduction in heat tr'ansfer coefficient. DNB is not a directly measurabl6 parameter during operation and therefore THERMAL POWER and Reactor Coolant Temperature and Pressure have been related to DNB.

The DNB correlations have been developed to predict the DNB flux and the location of DNB for axially uniform and non-uniform heat flux distributions. The local DNB heat flux ratio, DNBR, defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux, is indicative of the margin to DNB.

]

The DNB design basis is that there must be at least a 95 percent l

probability with 95 percent confidence that DNB will not occur when the minimum lBR-4 DNBR is at the design DNBR limit.

{

In meeting this design basis, uncertainties in plant operating parameters, nuclear and thermal parameters, and fuel fabrication parameters are considered statistically such that there is at least a 95 percent probability at a 95 percent confidence level that the minimum DNBR for the limiting rod is greater than or equal to the DNBR limit. The uncertainties in the above plant parameters are used to determine the plant uncertainty. This DNBR uncertainty,

. combined with the correlation DNBR limit, establishes a design DNBR value which i

.must be met in p1' ant safety analysis using values of input parameters without I

uncertainties.

The curves of Figure 2.1-1 show the loci of points of THERMAL POWER, Reactor Coolant System pressure and average temperature for which the minimum DNBR is no less than the safety analysis DNBR limit, or the average enthalpy at i

the vessal exit is equal to the enthalpy of saturated liquid.

i The cgrves of Figure 2.1-1 are based on an enthalpy rise hot channel factor, F

( minal values have been reduced to include a 4% total rod power uncertainthH, factor), and a reference cosine with a peak of 1.55 for axial power shape. An allowance is included for an increase in l

F{H 8 r* W8 power Msed on de expresskn:

q, = Fgo a.., a-m l

where P =

THERMAL POWER RATED TEERMAL POWER 1.70 - Mark-BW Fuel RTP F

= Nominal Valuer U

1.62 - Westinghouse Fuel i

l SEQUOYAH - UNIT 1 B 2-1 Amendment No. 19, 114, 138, 155,223

e-SAFETY LIMITS BASES 1

Range Channels will initiate a teactor trip at approximately 25 percent of RATED THERMAL POWER unless manually blocked when P-10 becomes active. No credit was taken for operation of the trips associated with either the Intermediate or source Range Channels in the accident analyses; however, their i

functional capability at the specified trip settings is required by this specification to anhance the overall reliability of the Reactor Protection I

system.

Dvertemnerature Delta T The overtemperature Delta T trip provides core protection to prevent DNB for all combinations of pressure, power, coolant temperature, and axial power-distribution,'provided that the transient is slow with respect to transit, thermowell, and RTD response time delays from the core to the temperature detectors (about 8 seconds), and pressure is within the range between the High R145 and Low Pressure reactor trips. This setpoint includes corrections for axial power distribution, changes in density and heat capacity of water with tempera-ture and dynamic compensation for transport, thermowell, and RTD response time delays from the core to RTD output indication. With normal axial power distri-R145 bution, this reactor trip limit is always below the core safety limit as shown in Figure 2.1-1. If axial peaks are greater than design, as indicated by the difference between top and bottom power range nuclear detectors, the reactor trip is automatically reduced according to the notations in Table 2.2-1.

The f ( AI) trip reset term in the overtemperature Delta T trip function i

precludes power distributions that cause the DNB limit to be exceeded during a limiting Condition II event. The negative and positive AI limits at which tre f (AI) term begins to reduce the trip setpoint and the dependence of f (AI) i on THERMAL POWER are determined on a cycle-specific basis using approved j

methodology and are specified in the COLR per Specification 6.9.1.14.

Operation with a reactor coolant loop out of service below the 4 loop P-8 setpoint does not _ require reactor protection system setpoint modification because the P-8 setpoint and associated trip will pretent DNB during 3 loop operation exclusive of the overtemperature Delta T setpoint.'

Delta-T., as used in the overtemperature and Overpower AT trips, represents the 100% RTP value as measured by the plant for each loop.

This normalizes each R145 loop's AT trips to the actual operating conditions existing at the time of measurement, thus forcing the trip to reflect the equivalent full power conditions as assumed in the accident analyses. These differences in RCS loop AT can be due to several factors, e.g., measured RCS loop flows greater than thermal design flow, and slightly asymmetric power distributions between,

quadrants. While RCS loop flows are not expected to change with cycle life, radial power redistribution between quadrants may occur, resulting in small changes in loop specific AT values. Accurate determination of the loop specific AT value should be made quarterly and under steady state conditions BR-11 (i.e., power distributions not affected by Xenon or other transient conditions).

Overnower Delta T The overpower Delta T reactor trip provides assurance of fuel integrity, e.g.,

no melting, under all possible overpower conditions, limits the required range for Overtemperature Delta T protection, and provides a backup to the High Neutron Flux trip. The setpoint includes corrections for changes in axial power distribution, January 2, 1997 SEQUOYAH - UNIT 1 B 2-4 Amendment No. 136, 141,223

~

e i

SAFETY LIMITS 1

RASES 4

density, and heat capacity of water with temperature, and dynamic compensation for transport, thermowell, and RTD response time delays from the core to RTD output indication..The setpoint is automatically reduced according to the notations in Table 2.2-1 to account for adverse axial flux differences.

4 i

The f:(AI) trip reset term in the overpower Delta T trip function i

precludes power distributions that cause the fuel melt limit to be exceeded i

during a limiting Condition II event. The negative and positive AI limits at which the f (AI) term begins to reduce the trip setpoint and the dependence of i

f:( AI) on-THERMAL POWER are determined on a cycle-specific basis using approved I

methodology and are specified in the COLR per Specification 6.9.1.14.

(

The Overpower Delta-T trip provides protection to mitigate the l

consequences of various size steam breaks as reported in WCAP-9226, " Reactor Core Response to Excessive Secondary Steam Releases."

Delta-T., as used'in the overtemperature and Overpower AT trips, represents the 100% RTP value as measured by the plant for each loop. This 1

normalizes each loop's AT trips to the actual operating conditions existing at the time of measurement, thus forcing the trip to reflect the equivalent full i

power conditions as assumed in the accident analyses. These differences in RCS i

loop AT can be due to several factors, e.g., measured RCS loop flows greater than thermal design flow, and slightly asymmetric power distributions between R145 quadrants.

While RCS loop flows are not expected to change with cycle life, radial power redistribution between quadrants may occur, resulting in small l

changes in loop specific AT values. Accurate determination of the loop specific AT value should be made quarterly and under steady state conditons (i.e., power distributions not affected by Xenon or other transient lBR-11 conditions).

i Pressurizer PressuIg 1

I The Pressurizer High and Low Pressure trips are provided to limit the l

pressure range in which reactor operation is permitted. The High Pressure trip is backed up by the pressurizer code safety valves for RCS overpressure 4

i protection, and is therefore set lower than the set pressure for these valver

+

(

.(2485.psig).

The Low Pressure trip provides protection by tripping the reactor 1

l in the event cf a loss of reactor coolant pressure.

2 Pressurizer Water Level The Pressurizer High Water Level trip ensures protection against Reactor l

Coolant System overpressurization by limiting the water level to a volume sufficient to retain a steam bubble and prevent water relief through the pressurizer safety valves. No credit was taken for operation of this trip in 4

i the accident analyses; however, its functional capability at the specified trip setting 1r4 required by this specification to. enhance the overall reliability of 4

the Reactor Protection System.

I, Lggs of Flow 1

The Loss of Flow trips provide core protection to prevent DNB in the event 4

of a 3oss of one or more reactor coolant pumps.

j Above 11 percent of RATED THERMAL POWER, an automatic reactor trip will occur if the flow in any two loops drops below 90% of nominal full loop flow.

4 Above the P-8 interlock, automatic reactor trip will occur if the flow in any R145 single loop drops below 90% of nominal full loop flow. This latter trip will prevent the minimum value of the DNBR from going below the safety analysis DNBR 4

I limit during normal operational transients and anticipated transients when 3 R142 i

j loops are in operation and the Overtemperature Delta T trip set point is aujusted to the value specified for all loops in operation.

R145 i

January 2, 1997 j

SEQUOYAH - UNIT 1 B 2-5 Amendment No. 138, 141, 223 i

1

-u-si'=

y

.p--,,

e y

gr,. -

r wn

....- -. --._ ~... -

I e

1*

POWER DISTRIBUTION LIMITS n

3/4.2.2 NEAT FLUX HOT CHANNEL FACTOR-Fa(X.Y 2) l LIMITING CONDITION FOR OPERATION I

l 3.2.2 F (X,Y, Z) shall be maintained within the acceptable limits specified in g

the COLR:

i APPLICABILITY: MODE 1 ACTION:

l With Fg(X,Y,2) exceeding its limits a.

Reduce THERMAL POWER at least 1% for each 1% Fo(X,Y,Z) exceeds the limit

{

within 15 minutes, and similarly reduce the following:

1. Administratively reduce the allowable power at each point along the AFD limit lines within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , and

2. 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 .

b.

POWER OPERATION may proceed for up to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . Subsequent POWER OPERATION may proceed provided the Overpower Delta T Trip Setpoints (value of K.) have been reduced at least it (in AT span) for each it that Fg (X, Y, Z) exceeds the limit specified in the COLR.

c.

Identify and correct the cause of the out-of-limit condition prior to increasing DIERMAL POWER above the reduced limit required by Action a. and

b., above; THERMAL POWER may then be increased provided Fo(X,Y,2) is demonstrated through incore mapping to be within its limits.

SURVEILLANCE AEQUIREMENTS' 4.2.2.1 The provisions of Specification 4.0.4 are not applicable.

R144 J

SEQUOYAH - UNIT 1 3/4 2-5 Amendment No. 19, 95, 14'0, 155, 223 I

)

i j

)

i POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued) 4.2.2.2 ry(x, y,z) shall be evaluated to determine if Fg(X,Y,Z) is within its limit by:

a. Using the moveable incere detectors to obtain a power distribution map

( rf(X, Y, z) * ) at any THERMAL POWER greater than 5% of RATED THERMAL POWER.

b. Satisfying the following relationship:

&(X, Yo Z) d BONOM(X, Y, Z) where BQNOM(X,Y, Z) ** represents the nominal design increased by an

)

allowance for the expected deviation between the nominal design and the measurement.

]

The BQNOM(X,Y,Z) factors are not applicable in the following core plane regions as measured in percent of core height from the bottom of the fuel

1. Lower core region from 0 to 15%, inclusive.

2. Upper core region from 85 to 100%, inclusive.

I

c. If'the above relationship is not satisfied, then

1. For that location, calculate the t margin to the maximum allowtble design as follows:

5 AFD Margin =

1 BODES (X, Y, Z) x 1005 s

t i (AI) Hargin =

1'

,Cyg,(, y,,

x 100%

s where BQDES (X,Y,2) ** and BCDES (X,Y,2) ** represent the maximum allowable design peaking factors which insure that the licensing criteria will be preserved for operation within Limiting condition for operation limits, and include allowances for the calculational and measurement uncertainties.

No additional uncertainties are required in the following equations for rf(x, y, z), because the limits include uncertainties.

- BQNOM (X,Y,Z), BQDES (X,Y, Z), and BCDES (X,Y, Z) Data bases are provided for input to the plant power distribution analysis computer codes on a cycle specific basis and are determined using the methodology for core limit generation described in the references in Specification 6.9.1.14.

SEQUOYAH - UNIT 1 3/4 2-6 Amendment Nos. 19, 95, 140, 155, 223

POWER DISTRIBiTTION LIMITS SURVEILLANCE REQUIREMENTS (Continued)

2. Find the minimum margin of all locations examined in 4.2.2.2.c.1 above.

AFD min reargin a minimum % margin value of all locations examined.

f (AI) OPAT min margin = minimum t margin value of all locations examined.

3. If the AFD min margin in 4.2.2.2.c.I above is

<0, either the following actions shall be taken, or the action statements for 3.2.2 shall be followed.

1 (a) Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , administratively re' duce the negative AFD limit lines at each power level by:

Reduced AFD * = (AFD* from COLR) + absolute value of (NSLOPE** t x AFD min margin of 4.2.2.2.c.2) l (b) Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , administratively reduce the positive AFD limit-i lines at each power level by:

Reduced AFD * = (AFD* from COLR) - absolute value of (PSLOPE** ' t X AFD min margin) 4.

If the f:( AI) min margin in 4.2.2.2.c.2 above is <0, either the following actions shall be taken, or the action statements for 3.2.2 shall be followed.

(a) Within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , reduce the OPAT negative f (AI) breakpoint limit by:

Reduced OPAT negative f (AI) breakpoint limit = (f (AI) limit of 2

Table 2.2-1) + absolute value of (NSLOPE'"" % x f (bI) min n'argin)

NSLOPE^N and PSLOPE** are the amount of AFD adjustment required to compensate for each 1% that F (X,Y,Z) exceeds the limit provided in the o

COLR per Specification 6.9.1.14.

pgtoyE 8"

and PSLOPE aten are the amounts of the CPAT f ( AI) limit t

adjustment required to compensate for eacn 1% that Fo(X,Y,Z) exceeds the limit provided in the COLR per Specification 6.9.1.14.

i l

SEQUOYAH - UNIT 1 3/4 2-7 Amendment No. 19, 95, 140, 155, 216,.223

4 POWER DISTRIBUTION LIMITS SURVEILIANCE REQUIREMENTS (Continued)

(b)

Within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , reduce the OPAT positive f (AI) breakpoint limit by:

c2 Reduced OPAT positive f (AI) breakpoint limit = (f:( AI) limit of Table 2.2-1) - absolute value of (PsLorzh"** 4 x f,( AI) min marFin) i d.

Measuring py(x,y,g) according to the following schedule i

1.

Upon achievi=r equilibrium conditions after exceeding by l

10 percent or more of RATED THERMAL POWER, the THERMAL POWER at which Fg(X,Y,Z) was last determined,*** or 2.

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

1 e.

With two measurements extrapolated to 31 EFPD beyond.the most recent measurement yielding fy(x, y,g}, sowaw(r, Y,Z), either of the following actions specified shall be taken.

1 1.

rf(x, Y, z) shall be increased over that specified in i

4.2.2.2.a by the appropriate factor specified in the COLR, and 4.2.2.2.c repeated, or I

2.

fy(y,y,z). shall be evaluated according to 4.2.2.2 at or before the time when the margin is projected to result in one of the actions specified in 4.2.2.2.c.3 or 4.2.2.2.c.4.

4.2.2'.3 When Fo(X,Y,Z) is measured for reasons other than meeting the requirements of Specification 4.2.2.2 an overall measured Fg(X,Y,Z). shall be obtained from a power. distribution map, increased by 3% to account for manufacturing tolerances and further increased by 5% to account for measurement uncertainty, and compared to the Fg(X,Y,Z) limit specified in the COLR according to Specification 3.2.2.

pstopg4t*" and PsLorr6(an are the amounts of the CPAT f:(AI) limit adjustment required to compensate tra; each it that Fn(X,Y,Z) exceeds the limit provided in the COLR per Sperification 6.9.1.14.

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

)

i 1

SEQUOYAH - UNIT 1 3/4 2-8 Amendment No. 19, 140, 223 4

9

=

POWER DISTRIBUTION LIMITS 3 /4. 2. 3 NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR - F.u f X. Y) l l

1 LIMITING CONDITION FOR OPERATION 3.2.3 F.u (X, Y) shall be maintained within the limits specified in the COLR.

i APPLICABILITY: MODE 1 ACTION:

With F n(X,Y) exceeding the limit specified in the COLR:

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

i 1.

Restore Fau(X,Y) to within the limit specified in the COLR, or

]

2.

Reduce the allowable THERMAL POWER from RATED THERMAL POWER at i

least RRH*% for each 1% that Fan (X,Y) exceeds the limit, and

b. Within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months either:

.I 1.

Restore Fax (X,Y) to within the limit specified in the COLR, or 1

j 2.

Reduce the Power Range Neutron Flux-High Trip Setpoint in Table 2.2-1 at least RRE*t for each it that F x(X,Y) exceeds that litat t, and

c. Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of initially being outside the limit specified in the i

COLR, either 1.

Restore Fan (X,Y) to within the limit specified in the COLR, or 2.

Verify through incore flux mapping that Fax (X,Y) is restored to 4

within the limit.for the reduced THERMAL POWER allowed by ACTION a.2 or reduce THERMAL POWER to less than 5% of RATED THERME j

POWER within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

]

I i

RRH is the amount of power reduction required to compensate for each 1%

that Fan (X, Y) exceeds the limit provided in the COLR per Specification 6.9.1.14.

SEQUOYAH - UNIT 1 3/4 2-10 Amendment No. 19, 138, 155, 223

t POWER DISTRIBUTION LIMITS ACTION:

(Continued)

d. Within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months of initially being outside the limit. specified in the I

COLR,, reduce the overtemperature Delta T K term in Table 2.2-1 by at least TRH** for each 1% that F,u(X,Y) exceeds the limit, and Identify and correct the cause of the out-of-limit condition prior to e.

increasing THERMAL POWER above the reduced THERMAL POWER limit required by ACTION a.2 and/or b and/or c. and/or d., above:

subsequent POWER OPERATION may proceed provided that F n(X,Y) is i

demonstrated, through incore flux mapping, to be within the above limit prior to exceeding the following THERMAL POWER levels:

1.

A nominal 50% of RATED THERMAL POWER, 2.

A nominal 75% of RATED THERMAL.00WER, and 3.

Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of attaining greater than or equal to 95% of RATED THERMAL POWER.

l t

TRH is the amount of Overtemperature Delta T Ki setpoint reduction 5

required to compensate for each 1% that Fan (X,Y) exceeds the limit 1

provided in the COLR per Specification 6.9.1.14.

i I

l l

l 1

l SEQUOYAH - UNIT 1 3/4 2-11 Amendment No. 138,223 l

i i

a e

POWER DISTRIBUTION LIMITS J

SURVEILLANCE REQUIREMENTS 4.2.3.1 The provisions of Specification 4.0.4 are not applicable.

4.2.3.2 ff,(y, y) shall be evaluated to determine if Fa(X,Y) is within its limit by-Using the movable incore detectors to obtain a power distribution map a.

Ff,(X, Y)

at any THERMAL POWER greater than 5% of RATED THERMAL POWER.

b. Satisfying the following relationship:

FAHR"(X,Y) s BHN0M(X,Y)

Where:

T[,(X, Y)

FAER"(X, Y) =

MAP" / AXIAL (X, Y)

And BHNOM(X,Y)** represents the nominal design increased by an allowance for the expected. deviation between the nominal design and the measurement.

MAP" is the maximum Allowable Peak ** obtained from the measured power distribution.

AXIAL (X,Y) is the axial shape for Fa(X,Y).

c'.

If the above relationship is not satisfied, then' 1.

For the location, calculate the t margin to the maximum allowable design as follows:

1Ts, Margin ={1-Aj x 1006 i

I'

% f (AI)#argin =

1-x 100%

t where BMDES (X,Y) and BRDES (X,Y) ** represent the maximum allowable design peaking factors which insure that the licensing criteria will be preserved for operation within the LCO limits, and include allowances for calculational and measurement uncertainties.

No additional uncertainties are required in the following equations for Ef,(X, Y) and FAHR" (X, Y), because the limits include uncertainties.

j BHNOM (X,Y), MAP", BHDES (X,Y), and BRDES(X,Y) data bases are provided for input to the plant power distribution analysis computer codes on a cycle specific basis and are determined using the methodology for core limit generation described in the references in Specification 6.9.1.14.

SEQUCYAH - UNIT 1 3/4 2-11a Amendment No. 223 p-

1 POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS 4.2.3.1 The provisions of Specification 4.0.4 are not applicable.

4.2.3.2 r*,(x, y) shall be evaluated to determine if Fa(X,Y) is within its limit by:

Using the movable incore detectors to obtain a power distribution map a.

7",(X, Y)

at any THERMAL POWER greater than 5% of RATED THERMAL POWER.-

b. Satisfying the following relationship:

FAHR"(X,Y) s BHNOM(X,Y)

Where:

d8E' D 7Amt*(x, y).

NAP" / AXZAL(X. Y)

And BHNOM(X,Y) ** represents the nominal design increased by an allowance for the expected deviation between the nominal design and the measurement.

M MAP is the taaximum Allowable Peak ** obtained from the measured power distribution.

l AXIAL (X, Y) is the axial shape for Fa(X,Y).

c.

If the above relationship is not satisfied, then 1.

For the location, calculate the t margin to the maximum allowable design as follows:

t Tu Nargin = {1 - {# "fy' x 1006

% / ( AI) Margin =

1-3 x 100%

where BHDES (X,Y) and BRDES (X,Y) ** represent the maximum allowable design peaking fac' tors which insure that the licensing criteria will be preserved for operation within the LCO limits, and include allowances for calculational and measurement uncertainties.

No additional uncertainties are required in the following equations for 7," (x, y} and FAHR"(X, Y), because the limits include uncertainties.

BHNOM(X,Y), MAP", BHDES (X,Y), and BRDES (X,Y) data bases are provided for input to the plant power distribution analysis computer codes on a cycle specific basis and are determined using the methcdology for core limit generation described in the references in Specification 6.9.1.14.

SEQUOYAH - UNIT 1 3/4 2-11a Amendment No. 223

POWER DISTR 2BUTION LIMITS SURVEILIANCE REQUIREMENTS

2.. Find the minimum margin of all locations examined in 4.2.3.2.c.1 above.

Fa min margin = minimum t margin value of all locations examined f (AI) min margin = minimum t margin value of all locations examined 3.

If the Fa min margin in 4.2.3.2.c.2 above is c 0, then within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months reduce the allowable THERMAL POWER f rom RATED THERMAL POWER by RRH*% x most negative margin from 4.2.3.2.c.2 and maintain the requirements of Specification 3.2.3; otherwise the Action statements for 3.2.3 apply.

4.

If the f:(AI) min margin in 4.2.3.2.c.2 above is < 0, then within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months reduce the overtemperature Delta T K1 term in Table 2.2-1 by at least TRH**t x most negative margin from 4.2.3.2.c.2 and maintain the requirements of Specification 3.2.3; otherwise the

. action statements for 3.2.3 apply.

d. With two measurements extrapolated to 31 EFPD beyond the most recent measurement yielding

. FAHR"(X,Y) > BHNOM (X, Y) either of the following actions shall be taken:

1.

4(x, y) shall be increased over that specified in 4.2.3.2.a by the appropriate factor specified in the COLR,' and 4.2.3.2.c.1' repeated, or 2.

4(x, y) shall be evaluated according to 4.2.3.2 at or before the time when the margin is projected to result in the action specified in 4.2.3.2.c.3 or 4.2.3.2.c.4.

4.2.3.3 4(x,n shall be determined to be within its limit by using the incore detectors to obtain a power distribution mapr a.

Prior to operation above 75% of RATED THERMAL POWER after each fuel loading, and

b. At least once per 31 EFPD.

RRH is the amount of power reduction required to compensate for.each it that Fa(X,Y) exceeds the limit provided in the COLR per Specification 6.9.1.14.

TRE is the amount of Overtemperature Delta T Ki setpoint reduction required to compensate for each it that Fa(X,Y) e'ceeds the limit x

provided in the COLR per Specification 6.9.1.14.

SEQUOYAH - UNIT 1 3/4 2-11b Amendment No.223

. =. - - ~. -..

l POWER DISTRIBUTION LIMITS 3/4.2.4 OUADRANT POWER TILT RATIO f

i LIMITING CONDITION FOR OPERATION

[

3.2.4 The QUADRANT POWER TILT RATIO shall not exceed 1.02, i

APPLICABILITY: MODE 1 above 50% of RATED THERMAL POWER

L Mr I

a.

With the QUADRANT POWER TILT RATIO determined to exceed 1.02 but less than or equal to 1.09:

1.

Calculate the QUADRANT POWER TILT RATIO at least once per hour l

until.

a)

Either the QUADRANT POWER TILT RATIO is reduced to within l

j l

its limit, or b)

THERMAL POWER is reduced to less than 50% of RATED THERMAL POWER.

2.

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

a)

Either reduce the QUADRANT POWER TILT RATIO to l

within its limit, or j

v b)

Reduce THERMAL POWER at least 3% from RATED THERMAL POWER i

for each 1% of indicated QUADRANT POWER TILT RATIO in excess of 1.02 and similarly reduce the Power Range Neutron l i

Flux-High Trip Setpoints within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

3.

Ve'rify that the QUADRANT POWER TILT RATIO is within its limit within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months af ter e::ceeding the limit or reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and reduce the Power Range Neutron Flux-High Trip setpoints to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

4.

Identify and correct the cause of the out of limit condition prier to increasing THERMAL POWER; subsequent POWER OPERATION above 50% of RATED THERMAL power may proceed provided that the QUADRANT' POWER TILT RATIO is verified within its limit at least l

once per hour for 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or until verified acceptable at 95%

or greater RATED THERMAL POWER.

l

See Special Test Exception 3.10.2.

l I

l 0

SEQUOYAH - UNIT 1 3/4 2-12 Amendment No.138, 2h2 i

POWER DISTRIBtJrION LIMITS ACTION:.(Continued) b.

With the QUADRANT POWER TILT RATIO determined to exceed 1.09 due to misalignment of either a shutdown or control rod:

1.

Calculate the QUADRANT POWER TILT RATIO at least once per hour until:

a)

Either the QUADRANT POWER TILT RATIO is reduced to within its limit, or b)

THERMAL POWER is reduced to less than 50% of RATED THERMAL POWER.

2.

Reduce THERMAL POWER at least 3% from RATED THERMAL POWER for each 1% of indicated QUADRANT POWER TILT RATIO in excess of 1.-02 l

within 30 minutes.

3.

. Verify that the QUADRANT POWER TILT RATIO is within its limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months after exceeding the limit or reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and reduce the Power Range Neutron Flux-High Trip Setpoints to l

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

4.

Identify and correct the cause of the out of limit condition prior to increaring THERMAL POWER; subsequent POWER OPERATION above 50% of RATO mt POWER may proceed provided that the QUADRANT POWER TILI RATIO is verified within its limit at least once per hour' for 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or until verified acceptable at 95%

or greater RATED THERMAL POWER.

c.

With the QUADRANT POWER TILT RATIO determined to exceed 1.09 due to causes other than the misalignment of either a shutdown or control rod:

1.

Calculate the QUADRANT POWER TILT RATIO at least once per hour until:

a)

Either the QUADRANT POWER TILT RATIO is reduced to within l

its limit, or b)

THERMAL POWER is reduced to less than 50%_of RATED THER.%L POWER.

SEQUOYAH - UNIT 1 3/4 2-13 Amendment No. 138 22$ R142

i TABLE 3.2-1 DNB PARAMETERS LIMITS 4 Loops In PARAMETER Oceratien Reactor Coolant System T, s 583*F Pressurizer Pressure.

m 2220 peia*

h actor' Coolant System Figure 3.2-1 Total Flow

Limit not applicable during either a THERMAL POWER ramp in excess of 5% RATED THERMAL POWER per minute or a THERMAL POWER step in excess of 10% RATED THERMAL POWER, physics test, or performance of surveillance requirement 4.1.1.3.b.

I

?

1 SEQUOYAH - UNIT 1 3/4 2-16 Amendment No. 41, las, 223 i

3 c

1 Figure 3.2-1 Flow vs. Power for 4 Loops in Operation 362000-A 3.5% measurerror.t uncertardy for nowisincludedinINs Agure.

(100.360100) 360000 --

f f

358000 --

1 Acceptable Operation y 356000 -.

Region c.2 e

.=

i E 354000 --

o iZ E

$ 352000 --

x va c

f 350'000 --

U.. acceptable 8

Operati.on j

Region 5 348000 --

345000 --

344000 --

(90.342095) 342000 90 92 94 96 98 100 102 l

Thermal Power Fraction (% of RTP)

SEQUOYAH - UNIT 1 3/4 2-17 Amendment No. 223

e REACTIVITY CONTROL SYSTEMS BASES The control rod insertion limits and shutdown rod insertion limits are l

specified in the COLR per Specification 6.9.1.14.

The ACTION statements which permit limited variations from the basic requirements are accompanied by additional restrictions which ensure that the original design criteria are met.

Misalignment of a red requires measurement of peaking factors and a restriction in THERMAL POWER. These restrictions provide assurance of fuel rod integrity during continued operation.

In addition, those accident analyses affected by a misaligned rod are reevaluated l

to confirm that the results remain valid during future operation.

In the event that a malfunction of the Rod Control System renders control rods R219 immovable, provision is made for continued operation provided:

o The affected control rods remain trippable, and o

The individual control rod alignment limits are met.

In the event that a malfunction of the Rod Control System renders control rod banks immovable during surveillance testing, provision is made for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months of continued operation provided:

o The affected control rod banks remains trippable, o

The individual control rod alignment limits are met, o

A maximum of one control or shutdown bank is inserted no more than 18 steps below the insertion limit, o

No reactor coolant system boron concentration dilution activities or power level increases are allowed, and o

The SHUTDOWN MARGIN requirements are verified every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or upon insertion of controlling bank during the period the insertion limit is not met.

The requirements to preclude Reactor Coolant System.boren concentration dilution, while a control or shutdown bank is below insert limits, will minimize the impact on shutdown margin.

The controlling bank (s), which is normally control Bank D, is excluded from the 72-hour provision since insertion of this bank (s) below the insertion limit is not required for control rod assembly surveillance testing. A controlling bank is defined as any control bank that is less than fully withdrawn as defined in the COLR with the exception of fully withdrawn banks that have been inserted in accordance with Surveillance Requirement 4.1.3.1.2.

This provision excludes the use of the 72-hour allowance for control banks that can be exercised 10 steps in either direction without exceeding the insertion limits.

Checks are performed for each reload core to ensure that bank insertions of.up to 18 steps will not result in power distributions, which violate the DNB criterion for ANS Condition II transients (moderate frequency transients analyzed in Section 15.2 of the UFSAR). Administrative requirements on the initial controlling bank position will ensure that this insertion and an additional controlling bank insertion of five steps or less will not violate the SHUTDOWN MARGIN requirement of Specification 3.1.1.1 during the repair period.

If the controlling bank is inserted more than five steps deeper than its initial position, a calculation will be performed to ensure that the SHUTDOWN MARGIN requirement of Specification 3.1.1.1 te met.

Since no dilution or power level increases are allowed, shutdown marcin will be maintained as long as the controlling bank is far enough above its insertion limit to compensate for the inserted worth of the bank that is beyond its insertion limit.

The 72-hour period for a control rod assembly bank to be inserted below its insertion limit restricts the likelihood of a more severe (i.e., ANS Condition III or IV) accident or transient condition occurring concurrently with the insertion limit violation.

SEQUOYAH - UNIT 1 3 3/4 1-4 Amendment No. 108, 215, 223

. _. - - _.. - =.

i e

l 3/4.2 POWER DIsTainuTION LII4ITS 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 morn.al 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 these specifications are as follows:

Fg(X,Y,Z)

Heat Flux Mot r hannel Factor, is defined as the maximum local l

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.

Nu' lear Enthalpy Rise Hot Channel Factor is defined as the ratio of l

Fw (X,Y) e 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 fAFD)

The limits on AXIAL FLUX DIFFERENCE assure that the 7/q(X,Y,Z) upper bound l

envelope of Fg limit specified in the COLR times the normr.lized axial peaking lR15 factor is not exceeded during either normal operation or in the event of xenon R144 redistribution following power changes.

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

The computer determines the one minute average of each of the OPERABLE excore detector outputs and provides an alarm message immediately if the AFD for at least 2 of 4 or 2 of 3 OPERABLE cxcore channels are outside the allowed AI-Power operating space and the THERMAL POWER is greater than 50 percent of RATED THERMAL POWER.

3/4.2.2 and 3/4.2.3 HEAT FLUX AND NUf' MAR ENTHALPY RISE HOT CHANNEL FACTORS l

The limits on the heat flux hot channel factor and the nuclear enthalpy R142

' rise het channel factor ensure that 1) 1.he 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. The peaking limits are specified in the COLR per Specification 6.9.1.14.

SEQUOYAH - UNIT 1 B 3/4 2-1 Amendment No. 19, 138, 140, 155,223 e

_ ~. _ _ _.. _

i P,QMER DI3TRIBUTION LIMITS RASE.S 2

---m m 1

Each of these hot channel factors is measurable but will normally only be j

determined periodically as specified in Specifications 4.2.2 and 4.2.3.

This i

periodic surveillance is sufficient to insure that the limits are maintained j

provided:

a.

Control rods in a single group move together with no individual rod 4

insertion differing by more than 1 13 steps from the group demand j

position.

L i.

b.

Control rod groups are sequenced with overlapping groups as j

described in Specification 3.1.3.6.

1 c.

The control rod insertion limits of Specifications 3.1.3.5 and l

3.1.3.6 are maintained.

3 1

d.

The axial power distribution, expressed in terms of AXIAL FLUX j

DIFFERENCE, is maintained within the limits.

The Fe(X,Y) limit as a function of THERMAL POWER allows changes in the radial power shape for all peru.issible rod insertion limits.

Fu (X, Y) will be maintained within its limits provided conditions a thru d above, are 1

maintained.

i When an FglX,Y,Z) measurement is taken, an allowance for measurement uncertsAoty is made. An allowance of 5% is appropriate for a full-core map a

taken with the Incore Detector Flux Mapping System, and this allowance is included in the methodology applied to the determination.of the core operating limits as described in the reference in Specification E.9.1.14.

The het channel factors, ry(x, y,z) and rf,(r. r), are measured periodically and compared to the nominal design values to provide a reasonable assurance that the core is operating as' designed and that the limiting criteria will not I

be exceeded for operation within the Technical Specification limits of Sections 2.2 (Limiting Safety System Settings), 3.1.3 (Moveable Control Assemblies), 3.2.1 ( AXIAL FLUX DIFFERENCC), and 3.2.4 (QUADRANT POWER TILT RATIO). An allowance is provided to acrount'for the expected deviation between the calculation and the measurement. It the measurement is above the maximum expected value for that location, at ir assumed to not be operating as designed, and a peaking margin evaluat ton is performed to provide a basis for decreasing the width of the AFD and f t AI) limits, and for reducing THERMAL POWER.

]

SEQUOYAH - UNIT 1 B 3/4 2-2 Amendment No. 19, 138, 155,223

~

.. - - - ~. - _.. ~

a I

POWER DISTRIBUTION LIMITS i

BASES 3/4.2.4 OUADRANT POWER TILT RATIO I

i The QUADRANT POWER TILT RATIO limit assures that no anomaly exists such that the radial power distribution satisfies the design values used in the power capability analysis. Radial power distribution measurements are made during startup testing and periodically during power operation.

The QUADRANT POWER TILT RATIO limit at which corrective action is j

required provides DNB and linear heat generation protection with x-y plane l

power tilts. The QUADRANT POWER TILT RATIO limit is reflected by a

}

corresponding peaking augmentation factor which is included in the generation of the AFD limits.

The 2-hour time allowance for operation with the tilt condition greater-i than 1.02 but less than 1.09, is provided to allow identification and i

correction of a dropped or misaligned control rod.

In the event such action i

does not correct the tilt, the margin for uncertainty on Fg(X,Y,Z) is l

reinstated by reducing the allowable THERMAL POWER by 3 percent for each percent of tilt in excess of 1.02.

t i

3/4.2.5 DNB PARAMETERS i

j-The limits on the DNB related parameters assure that each of the para-j 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 R142

}

to maintain a minimum DNBR of greater than or equal to the safety analysis DNBR limit throughout each analyzed transient.

The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 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 trknsient operation.

The flow parameters indicated in Figure 3.2-1 have been rounded down to bias the analysis in the conservative direction.

SEQUOYAH - UNIT 1 B 3/4 2-4 Amendment No. 19, 138, 155,223

INSTRUMENTATION BASES The measurement of response time at the specified frequencies provides assurance that the protective and ESF action function associated with each channel is completed within the time limit assumed in the accident analyses.

No credit was taken in the analyses for those channels with response times indicated as not applicable in the updated final safety analysis report.

]R194 Response time may be demonstrated by any series of sequential, overlapping or total channel test measurements provided that such tests demonstrate the total channel response time as defined. Sensor response time verification may be demonstrated by either 1) in place, onsite or offsite test measurements or 2) utilizing replacement sensors with certified response times Action 15 of Table 3.3-1, Reactor Trip System Instrumentation, allows the breaker to be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for the purpose of performing maintenance.

The 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is based on a Westinghouse analysis perforr.ed in R58 WCAP-10271, Supplement 1, which determines bypass breaker availability.

The placing of a channel in the trip condition provides the safety BR-9 function of the channel.

If the channel is tripped for testing and no other condition would have indicated inoperability, the channel should not be declared inoperable.

3/4.3.3 MONITORING INSTRUMENTATION 3 /4. 3. 3.1 RADIATION MONITCRING INSTRUMENTATION The OPERABILITY of the radiation monitoring channels ensures that 1) the radiation levels are continually measured in the areas served by the individual channels and 2) the alarm or automatic action is initiated when the radiation level trip setpcin; is exceeded 3/4.3.3.2 MOVABLE INCORE DETEC'IQB1 The OPERABILITY of the movahls incere detectors with the specified minimum complement of equipment ensures that the measurements obtained from use of this system accurately represent the spatial neutron flux distribution of the reactor core. The OPERABILITY of this system is demonstrated by irradiating each detector used and determining the acceptability of its voltage curve.

For the purpose of measuring Fo(X,Y,Z) or Fa (X,Y) a full incere flux map l

is used. Quarter-core flux maps, as defined in WCAP-8648, June 1976, may be used in recalibration of the excore neutron flux detection system, and full incere flux maps or symmetric incere thimbles may be used for monitoring the QUADRANT POWER TILT RATIO when one Power Range Channel is inoperable.

3 /4. 3. 3. 3 SEISMIC INSTRUMENTATION The OPERABILITY of the seismic instrumentation ensures that sufficient capability is available to promptly determine the magnitude of a seismic event and evaluate the response of those features important to safety. This capability is required to permit comparison of the measured response to that used in the SEQUOYAH - UNIT 1 B 3/4 3-2 Amendment No. 54, 190,223 L

e ADMINISTRATIVE CONTROLS MONTHLY REACTOR OPERATING REPORT 6.9.1.10 Routine reports of operating statistics and shutdown experience, including documentation of all challenges to the PORVs er Safety Valves, shall R76 be submitted on a monthly basis no later than the 15th of each month following the calendar month covered by the report.

CORE OPERATING LIMITS REPORT R159 6.9.1.14 Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT before each reload cycle or any remaining part of a reload cycle for the following:

1.

f (AI) limits for overtemperature Delta T Trip setpoints and f (AI) ilimits for overpower Delta T Trip Setpoints for Specification 2.2.1.

l 2.

Hoderator Temperature Coefficient BOL and EOL limits and 300 ppm surveillance limit for Specification 3/4.1.1.3, 3.

Shutdown Bank Insertion Limit for Specification 3/4.1.3.5, 4.

Control Bank Insertion Limits for Specification 3/4.1.3.6, S'.

AXIAL FLUX DIFFERENCE Limits for Specification 3/4.2.1, 6.

Heat Flux Hot Channel Factor and K(z) for Specification 3/4.2.2, and 7.

Nuclear Enthalpy Rise Hot Channel Factor for Specification 3/4.2.3.

6.9.1.14.a The analytical methods used to determine the acre operating limits R159 shall be those previously reviewed and approved by NRC in-1.

BAW-10180P-A, Rev.

1,

'NEMO - NODAL EXPANSION METHOD OPTIMIZED",

i March 1993. (FCF Proprietary)

(Methodology for Specification 3.1.2.3-Moderator Temperature Coefficient) 2.

BAW-10169P-A, "RSG PLANT SAFETY ANALYSIS - B&W SAFETY ANALYSIS METHODOLOGY FOR. RECIRCULATING STEAM GENERATOR PLANTS", October 1989.

(FCF Proprietary)

(Methodology for Specification 3.1.1.3-Moderator Temperature Coefficient) 3.

BAW-10163P-A, Core Operating Limit Methodology for Westinghouse-Designed PWRs, June 1989. (FCF Proprietary)

(Methodology for Specification 2.2.1,

- Limiting Safety System Settings (f ( AI), f:(AI) limits), 3.1.3.S - Shutdown Bank i

Insertion Limits, 3.1.3.6 - Control Bank Insertion Limits, 3/4.2.1 - Axial Flux Difference, 3/4.2.2 - Heat Flux Hot Channel Factor, 3/4.2.3 - Nuclear Enthalpy Rise Hot Channel Factor) 4.

BAW-1016BP-A, Rev. 2, RSG LOCA - B&W Loss of Coolant Accident Evaluation Model for Recirculating Steam Generator Plants, (FCF Proprietary)

(Methodology for Specification 3/4.2.2 - F4at Flux Hot Channel Factor) 5.

BAW-10168P-A, Rev 3, RSG LOCA - B&W Loss of Csolant Accident Evaluation Model for Recirculating Steam Generator Plants, (FCF Proprietary)

(Methodology for Specification 3/4.2.2 - Heat Flux Het Chai. el Factor)

SEQUOYAH - UNIT 1 6-21 Amendment No. 52, 58, 72, 74, 11'.

152, 155, 156, 171, 216,223

}

e

)

ADMINISTRATIVE CONTROLS i

I I

CORE OPERATING LIMITS REPORT (continued) 6.

WCAP-10054-P-A, Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code, August 1985, (E Propriet'ary)

(Methodology for Specification 3/4.2.2 - Heat Flux Hot Channel Factor)

)

7.

WCAP-10266-P-A, Rev. 2, "THE 1981 REVISION OF WESTINGHOUSE EVALUATION MODEL USING BASH CODE", March 1987, (E Proprietary).

(Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor).

6.9.1.14.b The core operating limits shall be determined so that all-applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and.

accident analysis limits) of the safety analysis are met.

6.9.1.14.c THE CORE OPERATING LIMITS REPORT shall be provided within 30 days after cycle start-up (Mode 2) for each reload cycle or within 30 days of R159

{

issuence of any mideycle revision of the NRC Document Control Desk with copies' to the Regional Administrator and Resident Inspector.

SPECIAL REPORTS 6.9.2.1 Special reports shall be submitted within the time period specified j

R76 for each report, in accordance with 10 CFR 50.4.

6.9.2.2 Diesel Generator Keliability Imerovement Procram 1

As a minimum the Reliability Improvement Program report for NRC audit, required by LCO 3.8.1.1, Table 4.8-1, shall include:

(a) a summary of all tests (valid and invalid) that occurred within the time period over which the last 20/100 valid tests were performed R56 (b) analysis of failures and determination of root causes of failures (c) evaluation of each of the recommendations of NUREG/CR-0660, " Enhancement of Onsite Emergency Diesel Generator Reliability in Operating Reactors,"

with respect to their application to the Plant (d) identification of all actions taken or to be taken to 1) correct the root causes of failures defined in b) above and 2) achieve a general improvement of diesel generator reliability (e) the schedule for implementation of'each action from d) above (f) an assessment of the existing reliability of electric power to engineered-saf ety-feature equipment SEQUOYAH - UNIT 1 6-21a Amendment No. 52, 58, 72, 74, j

117, 155,223 1

1 1

c arou p

UNITED STATES g

j NUCLEAR REGULATORY COMMISSION 4,.... + g #g WASHINGTON, D.C. 30666-0001 o

o TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-328 SE000YAH NUCLEAR PLANT. UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendmer.t No. 214 License No. DPR-79 1

1.

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

A.

The application for amendment by Tennessee Valley Authority (the licensee) dated April 4,1996, as supplemented by letters dated January 10, February 7, February 13, March 17, March 19, March 20, March 25, April 1. April 6, April 10, April 11, and April 18, 1997, 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; B.

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

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activitios 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.

e

i*,

2.

Accordingly, the license'is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment.

Paragraph 2.C.(2) of Facility Operating License No. DPR-79 is hereby amended to read as follows:

(2) Technical Soecifist_tions l

The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 214, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

A new License Condition, Paragraph 2.C.(18), is added as follows:

(18) Mixed Core DNBR Penalty TVA will obtain NRC approval prior to startup for any cycle's core that involves a reduction in the departure from nucleate boiling' ratio initial transition core penalty below that value stated in TVA's submittal on Framatome fuel conversion dated April 6,1997.

3.

This license amendme'nt is effective as of its date of issuance, to be implemented no later than 45 days of its issuanc.e.

FOR THE NUCLEAR REGULATORY COMMISSION

^

. Frederick J. Hetkfon, Director Project Directorate II-3

' Division of Reactor Projects -' I/II Office of Nuclear Reactor Regulation

Attachment:

1.

Page 11 of license 2.

Changes-to the Technical Specifications Date of Issuance: April 21,1997

Page 11 of the composite License is attached to reflect this change.

t 9

i 2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment.

7 Paragraph 2.C.(2) of Facility Operating License No. DPR-79 is hereby amended to read as follows:

(2) Technical Soecifications 1

The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 214, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

A new License Condition, Paragraph 2.C.(18), is added as follows:

(18) Mixed Core DNBR Penalty TVA will obtain NRC approval prior to startup for any cycle's core that involves a reduction in the departure from nucleate boiling i

i ratio initial transition core penalty below that value stated in TVA's submittal on Framatome fuel conversion dated April 6,.1997.

3.

This license amendment is effective as of its date of issuance, to be implemented upon installation of FCF fuel in the Unit 2 reactor vessel.

FOR THE NUCLEAR REGULATORY COMMISSION i

?

l Frederick J. Hebdon, Director Project Directorate 11-3 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation

Attachment:

1.

Page 11 of license 2.

Changes to the Technical Specifications Date of Issuance:

Page 11 of the composite License is attached to reflect this change.

4 1

ATTACHMENT TO LICENSE AMENDMENT NO. 214 FACILITY OPERATING LICENSE NO. DPR-79 DOCKET NO. 50-328 Revise the Appendix A Technical Specifications by removing the pages identified below and inserting the enclosed pages.

The revised pages are identified by the captioned amendment number and contain marginal lines indicating the area of change.

REHQy.E INSERT i

2-5 2-5 1

2-10 2-10 2-11 2-11 2

B 2-1 B 2-1 B 2-4 B 2-4 B 2-5 B 2-5 t

3/4 2-4 3/4 2-4 3/4 2-5 3/4 2-5 3/4 2-6 3/4 2-6 3/4.2-6a 3/4 2-6a

'3/4 2-8 3/4 2-8 3/4 2-9 3/4 2-9 3/4 2-9a 3/4 2-9b 3/4 2-10 3/4 2-10 3/4 2-11 3/4 2-11 3/4 2-14 3/4 2-14 3/4 2-15 8 3/4 1-4 8 3/4 1-4 '

B 3/4 2-1 B 3/4 2-1 B 3/4 2-2 B 3/4 2-2 B 3/4 2-4 B 3/4 2-4 8 3/4 3-2 B 3/4 3-2 6-22 6-22 6-22a

m

= _ - _

--..m m

~

TABLE 2.2-1 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT TRIP SETPOINT ALIDWABLE VALUES

1. Manual Reactor Trip Not Applicable Not Applicable 2.

Power Range, Neutron Flux Low Setpoint - s 25% of RATED Low Setpoint - s 27.4% of RATED lR145 THERMAL POWER THERMAL POWER j

High Setpoint - s 109% of RATED High Setpoint - s 111.4% of RATED lR145 THERMAL POWER THERMAL POWER

3. Power Range, Neutron Flux, s 5% of RATED THERMAL POWER with s 6.3% of RATED THERMAL POWER High Positive Rate a time const. ant a 2 second with a time constant a 2 second 4.

Power Range, Neutron Flux, s 5% of RATED THERMAL POWER with

.s 6.3% of RATED THERMAL POWER High Negative Rate

'a time constant a 2 second with a time constant a 2 second 5.

Intermediate Range, Neutrori s 25% of RATED THERMAL POWER s 45.20% of RATED THERMAL POWER lR189 Flux 1.45 x 10' counts per second lR189

6. Source Range, Neutron Flux s 10 counts per second s

7. Overtemperature AT See Note l' See Note 3

8. Overpower AT See Note 2 See Note 4 9.

Pressurizer Pressure--Low a 1970 psig a 1964.8 peig

10. Pressurizer Pressure--High s 2385 psig s 2390.2 psig' R145

11. Pressurizer Water Level--High 92% of instrument span s 92.7% of instrument span i

s

12. Loss of Flow a 90% of design flow a 89.6% of design flow R225l per loop

per loop

i t

Design flow is 90,045 (87,000 X 1.035)

'pm per loop.

l g

l l

SEQUOYAH - UNIT 1 2-5 Amendmend No. 44, 141, 185, 221,223

)

i

i l

TABLE 2.2-1 (Continued)

REAC'1DR TRIP BVws em INSTRINENTATION TRIP SETPOINTS

~

i NOTATION (Continued) t NOTE 1:

(Continued)

Laplace transform operator. sec~

S

=

and f (AI) is a function of the. indicated difference between top and bottom detectors i

of the power-range t.oclear ion chambers; with gains to be selected based on measured

[

~

j instrument response during plant startup tests such that:

(i) for q g between QTML* and QTPL* f (AI)

=0 (where q an,' n are percent RATED THERMAL POWER in the top and bottom halves of the core respectively, and q + g is total THERMAL POWER in percent of RATED THERMAL POWER).

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

-q) exceeds QTML*, the AT trip set -

b point shall be automatically reduced by QTNS* of its value at RATED THERMAL POWER.

(iii) for each percent that the magnitude of (q g) exceeds ' QTPi,*, the AT trip set-f point s' all be automatically reduced by QTPS* of its value at RATED THERMAL POWER.

[

h t

I NOTE 2:

Overpower AT (1 + '48) s AT {K

-K

( b8 ) T -K [T - T"] - f (AI))

O 4

1+T S 1+ r S

t 1+rs

= as defined in Note 1 j

1+T S R132 S

i

QTNL, QTPL, QTNS, and QTPS are specified in the COLR per' Specification 6.9.1.14.

J t

i SEQUOYAH - UNIT 2 2-10 Amendment No. 21, 132,214 l

~.?

1

2.1 SAFETY LIMITS LASES 2.1.1 RFACTOR 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 a

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 r

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.

The DNB correlations have been developed to predict the DNB flux and the location of DNB for axially uniform and non-uniform heat flux distributions.

The local DNB heat flux ratio, DNBR, defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux, is indicative of the margin to DNB.

The DNB design basis is that there must be at least a 95 percent 2

probability with 95 percent confidence that DN3 will not occur when the minimum i

DNBR is at the design DNBR limit.

lBR-5 In meeting this design basis, uncertainties in plant operating parameters, nuclear and thermal parameters, and fuel fabrication parameters are considered statistically such that there is at least a 95 percent probability at a 95 percent confidence level that the minimum DNER for the limiting rod is greater than or equal to the DNBR limit. The uncertainties in the above plant parameters are used to determine the plant uncertainty. This DNBR uncertainty, combined with the correlation DNBR limit, establishes a design DNBR value which must be met in plant safety analysis using values of input parameters without uncertainties.

The curves of Figure 2.1-1 show the loci of points of THERMAL POWER,

.R104 Reactor Coolant System pressure and average temperature for which the minimum DNER is no less than the safety analysis DNBR limit, or the average enthalpy at R130 the vessel exit is equal to the enthalpy of saturated liquid.

The gurves of Figure 2.1-1 are based on an enthalpy rise hot channel factor, F (n minsi values have been reduced to include a 4% total rod power AH,

)

uncertainty factor), and a reference cosine with a peak of 1.55 for axial power R21 shape. An allowance is included for an increase in F{H at reduced power based on the expression:

FfH =FRTP[1+.3 (1-P))

where P =

THERMAL POWER R146 RATED THERMAL POWER 1.70 - Mark-BW Puel RTP F

= Nominal Values AH 1.62 - Westinghouse Puel SEQUOYAH

. UNIT 2 B 2-1 Amendment No. 21, 104, 130, 146, 214 j

l

)

e LIMITING SAFETY SYSTEM SETTINGS BASES Intermediate and Source Rance. Nuclear Flux (Continued) e Range Channels will initiate a reactor trip at approximately 25 perc2nt of RATED THERMAL POWER unless manually blocked when P-10 becomes active. No

=

credit was taken for operation of the trips associated with either the Inter-R129 mediate or Source Range Channels in the accident analyses; however, their functional capability at the specified trip settings is required by this specification to enhance the overall reliability of the Reactor Protection System.

overtemocrature AT i

The Overtemperature Delta T trip provides core protection to prevent DNB for all combinations of pressure, power, coolant temperature, and axial power dis-i tribution, provided that the transient is slow with respect to transit, thermo-well, and RTD response time delays from the core to the temperature detectcrs (about 8 seconds), and pressure is within the range between the High and Low Pressure reactor trips. This setpoint includes corrections for axial power distribution, changes in density and heat capacity of water with temperature R132 and dynamic compensation for transport, thermowell, and RTD response time delays from the core to the RTD output indication. With normal axial power distribution, this reactor trip limit is always below the core safety limit as shown in Figure 2.1-1.

If axial peaks are greater than design, as indicated by the difference between top and bottom power range nuclear detectors, the reactor trip is automatically reduced according to the notat.cccs in Table 2.2-1.

' The f ( AI) trip reset-term in the overtemperature Delta T trip function precludes power distributions that cause the DNB limit to be exceeded during a limiting Condition II event. The negative and positive AI limits at which the f (AI) term begins to reduce the trip setpoint and the dependence of f (AI) on iTHERMAL POWER are determined on a cycle-specific basis using approved j

methodology and are specified in the COLR per Specification 6.9.1.14.

Operation with a reactor coolant loop out of service below the 4 loop P-8 s,etpoint does not require reactor protection system setpoint modification because the P-8 setpcint and associated trip will prevent DNB during 3 loop operation exclusive of the Overtemperature Delta T setpoint.

Delta-T as used in the overtemperat'ure and overpower AT trips, represents the 100 per8e,nt RTP value as measured by the plant for each loop. This normalizes each loop's AT trips to the actual operating conditions existing at the time of measurement, thus forcing the trip to reflect the equivalent full power condi-tions as assumed in the accident analyses. These differences in RCS loop AT can be due to several factors, e.g., measured RCS loop flows greater than ther-mal design flow, and slightly asymmetric power distributions between quadrants.

R132 While RCS loop flows are not expected to change with cycle life, radial power redistribution between quadrants may occur, resulting in small changes in loop specific AT values. Accurate determination of the loop specific AT value should be made when performing Incore/Excore quarterly recalibration and under steady state conditions (i.e., power distributions not affected by xenon or other transient conditions.).

SEQUOYAH - UNIT 2 E 2-4 Amendment No. 129, 132, 214

i 4

9 LIMITING SAFETY SYSTEM SETTINGS BASES Overoower AT The Overpower Delta T reactor trip provides assurance of fuel integrity, e.g.,

no melting, under all possible ovarpower conditions, limits the required range for overtemperature Delta T protection, and provides a backup to the High Neu-

- tron Flux trip. The setpoint includes corrections for changes in axial power distribution, density and heat capacity of water with temperature, and dynamic compensation for transport, thermowell, and RTD response time delays from the core to the RTD output indication. The setpoint is automatically reduced according to the notations in Table 2.2-1 to account for adverse axial flux differences.

The f (AI) trip reset term in the overpower Delta T trip function precludes j

power distributions that cause the fuel melt limit to be exceeded during a limiting condition II event. The negative and positive AI limits at which the f:( AI) term begins to reduce the trip setpoint and the dependence of f:(AI) on THERMAL POWER are determined on a cycle-specific basis using approved methodology and are specified in the COLR per Specification 6.9.1.14.

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

Delta-T,' as used in the overtemperature - and overpower AT trips, represents the 100 per9ent RTP value as measured by the plant for each loop. This normalizes each loop's AT trips to the actual operating conditions existing at the time of measurement, thus forcing the trip to reflect the equivalent full power condi-

~

tions as assumed in the accident analyses. These differences in RCS loop AT can be due to several factors, e.g.,

measured RCS loop flows greater than ther-mal design flow, and slightly asymmetric power distributions between quadrants.

While RCS loop flows are not expected to change with cycle life, radial power redistribution between quadrants may occur, resulting in small changes in loop specific AT values. Accurate determination of the loop specific AT value R132 should be made when performing Incore/Excore quarterly recalibration and under steady state conditions (i.e., power distributions not affected by xenon or other transient conditions.).

i Pressurizer Pressure The Pressurizer High and Low Pressure trips are provided to limit the pressure range in which reactor operation is permitted. The High Pressure trip is backed up by the pressurizer code safety valves for RCS overpressure protection, and is therefore set lower than the set pressure for these valves (24P1 psig). The Low Pressure trip provides protection by tripping _the reactor in the event of a loss of reactor coolant pressure.

]

Pressurizer Water Level The Pressurizer High Water Level trip ensures protection against Reactor Coolant System overpressurization by limiting the water level to a volume sufficient to retain a steam bubble and prevent water relief through the pressurizer safety valves. No credit was taken for operation of this trip in.

the accident analyses; however, its functional capability at the specified trip setting is required by this specification to enhance the overall reliability of '

i the Reactor Protection System.

f SEQUOYAH-UNIT 2 B 2-5 Amendment No. 132, 214 L

e j

-POWER D2STRIBUTION LIMITS 3/4.2.2 NEAT FLUX HOT CHANNEL FACTOR-F (X, Y. Z) l e

4 2

j LIMITING CONDITION FOR OPERATION 3.2.2 Fo (X, Y, Z) shall be maintained within the acceptable limits specified in f

the COLR.

i R ICABILITY: MODE 1 R21 l

ACTION:

With Fg(X,Y,2) exceeding its limit:

Reduce THERMAL POWER at least 1% for each 1% Fg(X,Y,Z) exceeds the limit s.

within 15 minutes, and similarly reduce the following:

1. Administratively reduce the allowable power at each point along the AFD d

limit lines within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , and e

i

2. 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 ; I

b.

POWER OPERATION may proceed for up to 48 ho trs.

Subsequent POWER OPERATION may proceed provided the Overpowe: Delta T Trip Setpoints (value of X.) have been reduced at least it (in AT span) for each it that l

Fo (X, Y, Z) exceeds the limit specified in the COLR.

j l

c.

Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER above the reduced limit required by Action a. and 4

b., above; THERMAL POWER may then be increased provided Fg(X,Y,Z) is demonstrated through incore mapping to be within its limits.

R21 SURVEILLANCE REQUIREMENTS q

4.2.2.1 The provisions of Specification 4.0.4 are not applicable.

,i 1

SEQUOYAH - UNIT 2 3/4 2-4 Amendment Nos. 21, 95, 131, 146, 214

S s

POWER DISTRIBUTION LIMITS

- SURVEILLANCE REQUIREMENTS (Continued) j 4. 2. 2. 2 pg(x, y, z) 'shall be evaluated to determine if Fo(X,Y,Z) is within its limit by:

a. Using the moveable incore detectors to obtain a power distribution map

{

(ry(x, y,z) *)

at any THERMAL POWER greater than 5% of RATED THERMAL POWER.

b.' Satisfying the following relationships i

F5(X, Y, Z) i BONOM(X, Y,2)-

1 where SQNOM(X,Y,Z) ** represents the nominal design increased by an allowance for the expected deviation between the nominal design and the

(

measurement.

I The BQNOM (X, Y, Z) factors are not applicable in the following core plane regions as measured in percent of core height from the bottom of the fuel

)

1. Lower core region from 0 to 15%, inclusive.

2. Upper core region from 85 to loot, inclusive.

-c. If the above relat'ionship is not' satisfied, then a

1. For that. location, calculate the t margin to the maximum allowable design as follows:

6 AFD Margin =

1 ~ gO gg(, 7 g}

XM r

i t

i Ff(X, Y, Z) i t f ( AI) Margin =

1.

SCDES% Y,2) 100t where BQDES (X,Y, Z)" and BCDES (X,Y, Z) ** represent the maximum allowable design peaking factors which insure that the licensing criteria will be preserved for operation within Limiting condition for Operation limits, s

.and include allowances for the calculational and measurement

-uncertainties.

1 4

j No additional uncertainties are required in the following equations for j

Ff(x, Y, r), because the limits include uncertainties.

1 J

- BQNOM (X,Y,Z), BQDES (X, Y, Z), and BCDES(X,Y,Z) Data bases are provided for i

input to the plant power distribution analysis computer codes on a cycle specific basis and are determined using.the methodology for core limit generation described in the references in Specification 6.9.1.14.

SEQUOYAH - UNIT 2 3/4 2-5 Amendmend No. 21, 95, 131, 146, 214 f

4

s d

POWER DISTRIBt7 TION LIMITS a.

l SURVEILLANCE REQUIREMENTS (Continued)

~

2. Find the minimum margin of all locations examined in 4.2.2.2.c.1 above.

]

AFD min margin = minimum t margin value of all locations examined.

f:( AI) OPAT min margin = minimum t margin value of all locations examined.

4

3. If the AFD min margin in 4.2.2.2.c.2 above is

<0, either the following l

actions shall be taken, or the action statements for 3.2.2 shall be followed.

4 i

(a) Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , administratively reduce the negative AFD limit lines at each power level by:

Reduced AFD* = (AFD* from COLR) + absol. A value of

'(NSLOPE'** % x AFD min margin of 4.2.2.2.c.2) 3 i

(b) Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , administratively reduce the positive AFD limit i

lines at each power level by:

Reduced AFD* = (AFD* from COLR) absolute value of (PSLOPE^** t X AFD min. margin)

4. 'If the f ( AI) min margin in 4.2.2.2.c.2 above 14

<0, either the following actions shall be taken, or the action statements for 3.2.2 i

i shall be followed, j

4 (a) Within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , reduce the OPAT negative f (AI) breakpoint limit by:

l I

i Reduced OPAT negative f:(AI) breakpoint limit = (f (AI) limit of Table 2.2-1) + absolute value of (NSLOPE atan" % X f ( AI) min margin) t j

1 I

NSLOPE^N and PSLOPE^N are the amount of AFD adjustment required to

~

compensate for each it.that Fo(X,Y,2) exceeds the limit provided in the COLR per Specification 6.9.1.14.

NSLOPE8" and PSLOPE :(an are the amounts of the OPAT f (AI) limit t

2 adjustment required to compensate for each it that Fo(X,Y,2) exceeds the limit provided in the COLR per Specification 6.9.1.14.

I l

1 1

SEQUOYAH - UNIT 2 3/4 2-6 Amendment No. 21, 95, 131, 146, 206,214 i

= _ _. ---

POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued) u,

(b)

Within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , reduce the OPAT positive f ( AI) breakpoint limit by:

Reduced OPAT positive f (AI) breakpoint limit - (f:( AI) limit of Table 2.2-1) absolute value of (PSLcPf"** t x f,( AI) min margia) d.

Measuring py(x,y,z) according to the following schedules 1.

Upon schieving equilibrium conditions after exceeding by 10 percent or more of RATED THERMAL POWER, the THERMAL POWER at which F (X,Y,Z) was last determined,*** or g

2.

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

e.

With two measurements extrapolated to 31 EFPD beyond the most recent measurement yielding 7l(x,y,z) 30NaH(x, Y,z), either of the following actions specified shall be taken.

1.

7;(x, y,7) shall be increased over that specified in 4.2.2.2.a by the appropriate factor specified in the COLR, and 4.2.2.2.c repeated, or 2.

yg(x, y, g) shall be evaluated according to 4.2.2.2 at or before the time when the margin is projected to result in one of the actions specified in 4.2.2.2.c.3 or 4.2.2.2.c.4.

4.2.2.3 When Fn(X,Y, Z) is measured for reasons other than meeting the requirements of Specification 4.2.2.2 an overall measured F (X,Y,Z) shall be n

obtained from a power distribution map, increased by 3% to account for manufacturing tolerances and further increased by 5% to account for measurement uncertainty, and compared to the Fo(X,Y,Z) limit specified in the COLR according to Specification 3.2.2.

tan are the amounts of the CPAT f (AI) limit ystopgsi'" and PStorga adjustment required to compensate for each 1% that Fo(X,Y,Z) exceeds the limit provided in the COLR per Specification 6.9.1.14.

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 power distribution map obtained.

SEQUOYAH - UNIT 2 3/4 2-6a Amendment No. 21, 95, 131,214

., a i

POWER DISTRIETTTION LIMITS j

3/4c2.3 NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR - F.o(X,Y) i i

LIMITING CONDITION FOR OPERATION I

a e

3.2.3 Fa (X, Y) shall be maintained within the limits specified in the COLR.

APPLICABILITY: MODE 1 i

ACH.QE:

f With Fa(X,Y) exceeding the. limit specified in the COLR:

I l

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

f i

i 1

1.

Restore Fm(X,Y) to within the limit specified in the COLR, or i

2.

Reduce the allowable THERMAL POWER from RATED THERMAL POWER at a

i least RRH*% for each 1% that Fa(X,Y) exceeds the limit, and j

j

b. Within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months either:

1.

Restore Fa(X,Y) to within the limit specified in the COLR, or i

2.

Reduce'the Power Range Neutron Flux-High Trip Setpoint in Table 2.2-1 at least RRH*% for each it that Fa(X,Y) exceeds that limit, 4

and i

1

~

c. Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of initia'lly being outside the limit specified in the J

COLR, either:

i j

1.

Rescore Fa(X,Y) to within the limit specified in the COLR, or i

2.

Verify through incore flux mapping that Fa(X,Y) is restored to

]

within the limit for the reduced THERMAL POWER allowed by ACTION a.2 or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER

]

within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

l

)

RRH is the amount of power reduction required to compensate for each 1%

that Fa(X,Y) exceeds the limit provided in the COLR per Specification 6.9.1.14.

SEQUOYAH - UNIT 2 3/4 2-8 Amendment No. 21, 130, 146, 214

~

e i

t

~,

POWER DISTRIBUTTON 2IMITS e

i ACTION:

(Continued) d.

Within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months of initially being outside the limit specified in the I

COLR, reduce the Overtemperature Delta T K, term in Table 2.2-1 by at j

least TRH** for each 1% that F,w(X,Y) exceeds the limit, and i

e.

Identify and correct the cause of the out-of-limit condition prior to l

3 increasing THERMAL POWER above the reduced THERMAL POWER limit J

required by ACTION a.2 and/or b. and/or c. and/or d.,

above; j

subsequent POWER OPERATION may proceed provided that Fax (X,Y) is demonstrated, through incore flux mapping, to be within the above f

limit prior to exceeding the following THERMAL POWER levels:

]

+

i i

1.

A nominal 50% of RATED THERMAL POWER, l

2.

A nominal 75% of RATED THERMAL POWER, and 4

3.

Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of attaining greater than or equal to 95% of RATED THERMAL POWER.

1 1

1 i

I l

i 1

i

.)

f

?

i i

.i i

1 l

l TRH is the amount of overtemperature Delta T Ki setpoint reduction required to compensate for each 1% that F,u(X,Y) exceeds the limit j

provided in the COLR per Specification 6.9.1.14.

i 4

i i

t i

l 1

4 4

i SEQUOYAH - UNIT 2 3/4 2-9 Amendment No. 130, 214 1

e

..f-POWER DISTRIBUTION L7 M SURVEILLANCE REQUIREMENTS IM b 4.2.3.1 Yhe provisions of Specification 4.0.4 are not applicable.

4.2.3.2 ff,(x, y} shall be evaluated to determine if Fa(X,Y) is within its limit by:

a.

Using the movable incore detectors to obtain a power distribution map Tf,(X, Y)

at any THERMAL POWER greater than 5% of RATED THERMAL POWER.

b.

Satisfying the following relationship:

FAHR"(X,Y) s BHNOM (X, Y) i Where:

TARR"(X, Y) = MAP" / AIIAL(X, Y)

And BENOM(X,Y) ** represents the nominal design increased by an allowance for the expected deviation between the nominal design and the measurement.

MAP" is the maximum Allowable Peak +* obtained from the measured power distribution.

AXIAL (X, Y) is the axial shape for Fa(X,Y).

c.

If the above relationship is not satisfied, then 1.

For the location, calculate the t margin to the maximum allowable design as follows:

R Ts,'Nargin = 1 -

x 1005 A

% f ( AI) Margin =

1-x 100%

3 where BHDES (X,N) and BRDES (X,Y) ** represent the maximum allowable design peaking factors which insure that the licensing criteria will be preserved for operation within the LCO limits, and include allowances for calculational and measurement uncertainties.

No additional uncertainties are required in the following equations for Ef,(X, Y) and FAHR"(X,Y), because the limits include uncertainties.

BHNOM(X,Y), MAP", BHDES (X,Y), and BRDES (X,Y) data bases are provided for input to the plant power distribution analysis computer codes on a cycle specific basis and are determined using the methodology for core limit generation described in the references in Specification 6.9.1.14.

SEQUOYAH - UNIT 2 3/4 2-9a Amendment No. 214

_.. =. __

e POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS l

2.

. Find the minimum margin of all locations examined in 4.2.3.2.c.1 above.

i Fa min margin = minimum % margin value of all locations examined i

f (AI) min margin - minimum t margin value of all locations-i examined 3.

If the Fa min margin in 4.2.3.2.c.2 above is < 0, then within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months reduce the allowable THERMAL POWER from RATED THERMAL l

POWER by RRH*% x most negative margin from 4.2.3.2.c.2 and

't maintain the requirements of Specification 3.2.3; otherwise the l

Action statements for 3.2.3 apply.

[

l*

4.

If the f ( AI) min margin in 4.2.3.2.c.2 above is < 0, then i

within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months reduce the Overtemperature Delta T K1 term in l

Table 2.2-1 by at least TRH**% x most negative margin from j

4.2.3.2.c.2 and maintain the requirements of Specification i

3.2.3; otherwise the action statements for 3.2.3 apply.

d.

With two measurements extrapolated to 31 EFPD beyond the most recent measurement yielding FAHR"(X,Y) > BHNOM(X,Y) either of the following actions shall be taken:

1.

&(X, Y) shall be increased over that specified in 4.2.3.2.a by the appropriate factor specified in the COLR, and 4.2.3.2.c.1 j

' repeated, or 2.

4(x, y) shall be evaluated according to 4.2.3.2 at or before i

the time when the margin is projected to result in the action j

specified in 4.2.3.2.c.3 or 4.2.3.2.c.4.

4.2.3.3 4(x, y) shall be determined to be within its limit by using the incore detectors to obtain a power distribution maps Prior to operation above 75% of RATED THERMAL POWER after each fuel j

a.

loading, and b.

At least once per 31 EFPD.

j RRE is the amount of power reductic. required to compensate for each 1%

that Fa(X,Y) exceeds the limit provided in the COLR per Specification 6.9.1.14.

TRE is the amount of Overtemperature Delta T Ki setpoint reduction required to compensate for each it that Fa(X,Y) exceeds the limit provided in the COLR.per Specification 6.9.1.14.

SEQUOYAH - UNIT 2 3/4 2-9b Amendment No.214 4

e POWER DISTRIBUTION LIMITS 2

3 /4. 2. 4 OUADRANT POWER TILT RATIO 4

LIMITING CONDITION FOR OPERATION a

3.2.4 The QUADRANT POWER TILT RATIO shall not exceed 1.02.

I APPLICABILITY: MODE 1 above 50% of RATED THERMAL POWER

ACTION:

i a.

With the QUADRANT POWER TILT RATIO determined to exceed 1.02 but less than or equal to 1.09:

i 1.

calculate the QUARANT POWER TILT RATIO at least once per hour until either a)

The QUADRANT POWER TILT RATIO is reduced to within its limit, or b)

THERMAL POWER is reduced to less than 50% of RATED THERMAL 1

POWER.

1 2.

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

a)

Reduce the QUADRANT POWER TILT RATIO to within its limit,-

or b)

Reduce THERMAL POWER at least 3% from RATED THERMAL POWER j

for each 1% of indicated QUADRANT POWER TILT RATIO in excess of 1.02 and similarly reduce the Power Range Neutron i

Flux-High Trip Setpoints within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

j 3.

Verify that the QUADRANT POWER TILT RATIO is within its limit within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after exceeding the limit or reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and reduce the Power Range Neutron Flux-High Trip set-4 points to less than or equal to 55% of RATED THERMAL POWER j

within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

d 4.

Identify and correct the cause of the out of limit condition prior to increasing THERMAL POWER; subsequent POWER OPERATION above 50% of RATED THERMAL power may proceed provided that the QUADRANT POWER TILT RATIO is verified within its limit at least once per hour for 13 hours1.50463e-4 days 0.00361 hours 2.149471e-5 weeks 4.9465e-6 months or until verified acceptable at 95%

or greater RATED THERMAL POWER.

isee Special Test Exception 3.10.2.

lR130 SEQUOYAH - UNIT 2 3/4 2-10 Amendment No. 130, 214

l POWER DISTRIBUTION LIMITS ACTION: (Continued) b.

With the QUADPANT POWER TILT RATIO determined to exceed 1.09 due to misalignment of either a shutdown or control rod:

1.

Calculate the QUADRANT POWER TILT RATIO at least once per hour until either:

a)

The QUADRANT POWER TILT RATIO is reduced tio within its limit, or N b)

THERMAL POWER is reduced to less than 50% of RATED THERMAL POWER.

2.

Reduce THERMAL POWER at least 3% 'from RATED THERMAL POWER for each 1% of indicated QUADRANT POWER TILT RATIO in excess of 1.02 within 30 minutes.

3.

Verify that the QUADRANT POWER TILT RATIO is within its limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months after exceeding the limit or reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and reduce the Power Range Neutron Flux-High Trip Setpoints to less than or equal,to 55% of RATED THERMAL POWER within the next l 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

4.

Identify and correct the cause of the out of limit condition prior to increasing THERMAL POWER; subsequent POWER OPERATION above 50% of RATED THERMAL POWER may proceed provided that the QUADRANT POWER TILT RATIO is verified within its limit at least once per hour for 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or until verified acceptable at 95%

or greater RATED THERMAL POWER.

c.

With the QUADRANT POWER TILT RATIO determined to exceed 1.09 due to causes other than the misalignment of either a shutdown or control rod:

1.

Calculate the QUADRANT POWER TILT RATIO at least once per hour until either:

4 a)

The QUADRANT POWER TILT RATIO is reduced to within its limit, or b)

THERMAL POWER is reduced to less than 50% of RATED THERMAL i

POWER.

i

~

.i 1

i SEQUOYAH - UNIT 1 3/4 2-11 Amendment No.13c,21h3a l

~. __

e TAB 7E 3.2-1 I

DNB PARAMETERS i

l LIMITS 1

1 4

I 4 Loops In PARAMETER Oneration i

d Reactor Coolant System T 583*F j

avg Pressurizer Pressure 2,.2220 psia

4 i

Reactor Coolant System Flow Rate Figure 3.2-1 l

l I

q-l 4

4 f

3 I

J l

4

Limit not applicable during either a TERMAL POWER ramp in excess of 5% of RATED THERMAL POWER per minute or a TERMAL POWER step in excess of 10% of 1

RATED THERMAL POWER, physics test, or performance of surveillance.

j requirement 4.1.1.3.b.

l i

4 4

l 1

1 j

SEQUOYAH - UNIT 2 3/4 2-14 Amendment No. 33, 130,214 i

i

)

i

Figure 3.2-1 Flow vs. Power for 4 Loops in Operation 4

)

362000 A 3.5% measwoment uncertanty for ikwr is included in this flgwe.

(100,360100) i 360000 --

l 358000 --

i Acceptable Operation g 356000 --

Region c.

S so E 354000 --

W o

u.

E

$ 352000 --

us

'c' 2

j j

350000 "

Unacceptable 8

Operation j

j Regi6n es l

st: 348000 --

1 i

346000 --

)

4 344000 --

i i

n (90,342095) 342000 l

i 90 92 94 96 98 100 102 Thermal Power Fraction (% of RTP) 4 4

SEQUOYAH - UNIT 2 3/4 2-15 Amendment No. 214

i REACTIVITY CONTRCL SY3TEMS i

l BASES a

+

i l

l MOVEABLE CONTROL ASSEMBLTES (Continued) t The control rod insertion limits and shutdown rod insertion limits are 1

j specified in the COLR for Specification 6.9.1.14.

}

The ACTION statements which permit limited variations from the basic

)

i requirements are accompanied by additional restrictions which ensure that the original design criteria are met. Misalignment of a rod requires measurement of peaking factors and a restriction in THERMAL POWER. These restrictions l

1 provide assurance of fuel rod integrity during continued operation. In addition, those safety analyses.affected by a misaligned rod are reevaluated to confirm that the results remain valid during future operation.

In the event that a malfunction of the Rod Control System renders control rods R205 immovable, provision is made for continued operation provided:

o The affected control rods remain trippable, and o

The individual control rod alignment limits are met.

{

_ In the event that a malfunction'of the Rod Control System renders control rod banks immovable during surveillance testing, provision is made for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months of continued operation provided:

o The affected control rod banks remains tripp'able, o,

The individual control rod alignment limits are met, o

A maximum of one control or shutdown bank is inserted no more than i

la steps below the insertion limit, j

o No reactor coolant system boron concentration dilution activities or power level increases are allowed, and i

o The SHUTDOWN MARGIN requirements are verified every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or i

upon insertion of controlling bank during the period the insertion l

limit is not met.

The requirements to preclude Reactor Coolant System boron concentration 4

dilution, while a control or shutdown bank is below insert limits, will minimize the impact on shutdown margin.

e i

- The controlling bank (s), which is normally Control Bank D, is excluded from the 72-hour provision since insertion of this bank (s) belew the insertion limit is not required for control rod assembly surveillance testing. A controlling bank is defined as any control bank that is less than fully withdrawn as defined in the COLR with the exception of fully withdrawn banks that have been inserted in accordance with Surveillance Requirement 4.1.3.1.2.

This provision excludes the use of the 72-hour allow.' ace for control banks that can be exercised 10 steps in either direction without exceeding the insertion limits.

. Checks are performed for each reload core to ensure that bank insertions of up to 18 steps will not result'in power distributions, which violate the DNB criterion for ANS Condition II transients (moderate frequency transients analyzed in Section 15.2 of the UFSAR). Administrative requirements on the initial controlling bank position will ensure that this insertion and an additional controlling bank insertion of five steps or less will not violate the SHUTDOWN MARGIN requirement of Specification 3.1.1.1 during the repair period.

If the controlling bank is inserted more than five steps deeper than its initial position, a calculation will be performed to ensure that the SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is met.

Since no dilution or power level increases are allowed, shutdown margin will be maintained as long as the controlling bank is far enough above its insertion limit to compensate for the inserted worth of the bank that is beyond its insertion limit.

The 72-hour period for a control rod assembly bank to be inserted below its insertion limit restricts the likelihood of a more severe (i.e., ANS Condition III or IV) accident or transient condition occurring concurrently with the insertion limit violation.

SEQUOYAH - UNIT 2 B 3/4 1-4 Amendment No. 98, 205, 214

e 3/4.2 POWER DISTRIBUTION LIMITS RASES i

i n

l I

EE The specifications of this section provide assurance of fuel integrity R21 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 these specifications are as follows:

F (X,Y,2)

Heat Flux Hot channel Factor, is defined as the maximum local heat flux en the surface of a fuel rod at core elevation z divided R21 by the average fuel rod heat flux, allowing for manufacturing tolerances on fuel pellets and rods.

FAH(X,Y)

Nuclear Enthalpy Rise Hot Channel Factor, is defined as the ratio l

of the integral of linear power along the rod with the highest integrated power to the average rod power.

~

R21 3/4.2.1 AXIAL FLUX DIFFERENCE (AFD)

I The' limits on AXIAL FLUX DIFFERENCE (AFD) assure that the Fo(X,Y,2) upper bound envelope of the Fo limit specified in the COLR times the normalized axial R146 peaking factor is not exceeded during either normal operation or in the event of xenon redistribution following power changes.

R21 Provisions for monitoring the AFD on an automatic basis are derived from the plant process computer through the AFD Monitor Alarm. The compuer deter-mines the one minute average of each of the OPERABLE axcore detector outputs and provides an alarm message immediately if the AFD for at least 2 of 4 or 2 of 3 OPERABLE excere channels are.outside the allowed AI-Power operating space and the THERMAL POWER is greater than 50 percent of RATED THERMAL ?OWER.

3 /4. 2. 2 and 3 /4. 2. 3 HEAT FLUX AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTORS The limits on the heat. flux hot channel factor and the nuclear enthalpy rise hot channel factor ensure that 1) the design limits on peak local power R21 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. The peaking limits are specified in the COLR per Specification 6.9.1.14 SEQUOYAH - UNIT 2 B 3/4 2-1 Amendment No. 21, 130, 131, 146, 214

s 20WER DISTRIBtTTION LIMITS 1

d BASES Each of these hot channel factors is measurable but will normally only be lR130 determined periodically as specified in Specifications 4.2.2 and 4.2.3.

This R21 periodic surveillance is sufficient to insure that the limits are maintained provided:

a.

Control rods in a single group move togetner with to individual rod

)

insertion differing by more than 3,13 steps from the group demand i

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 terms of AXIAL FLUX DIFFERENCE, is maintained within the limits.

i The Fa(X,Y) limit as a function of THERMAL POWER allows changes in the radial power shape for all permissible, rod insertion limits.

Fa (X, Y) will be maintained within its limits provided conditions a thru d above, are maintained.

When an F (X,Y, Z) measurement is taken, an allowance for measurement n

uncertainty is made. An allowance of 5% is appropriate for a full-core map taken with the Incore Detector Flux Mapping System, and this allowance is included in the methodology applied to the detereination of the core operating limits as described in the reference in Specification 6.9.1.14 The hot channel f actors,' rf(X. r.g; and r" (X. y), are measured periodically u

and compared to the nominal design values to provide a reasonable assurance that the core is operating as designed and that the limiting criteria will not be exceeded for operation within the Technical Specification limits of Sections 2.2 (Limiting Safety System Settings), 3.1.3 (Moveable Control Assemblies), 3.2.1 (AXIAL FLUX DIFFERENCE), and 3.2.4 (QUADRANT POWER TILT RATIO). An allowance is provided to account for the expected deviation between the calculation and the measurement.

If the measurement is above the maximum expected value for that location, it is assumed to not be operating as designed, and a peaking margin evaluation is performed to provide a basis for decreasing the width of the AFD and f(AI) limits, and for reducing THERMAL 1

POWER.

4 E

4 l

2 l

SEQUOYAH - UNIT 2 B 3/4 2-2 Amendment No. 21, 130, 146, 214

e POWER DISTRIBUTION LIMITS BASES 3/4.2.4 OUADRANT POWER TILT RATIO The QUADRANT POWER TILT RATIO limit assures that no anomaly exists such that the radial power distribution satisfies the design values uted in the 1

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

The QUADRANT POWER TILT RATIO limit at which corrective action is required provides DNB and linear heat generation protection with x-y plane power tilts.

The QUADRANT POWER TILT RATIO limit is reflected by a -

corresponding peaking augmentation factor which is included in the generation of the AFD limits.

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

In the event such action does not correct the tilt, the margin for uncertainty on Fg(X,Y,Z) is reinstated by reducing the allowable THERMAL POWER by 3 percent for each percent of tilt in excess of 1.02.

3/4.2.5 DNB PARAMETERS R21 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 greater than or equal to the safety analysis DNBR R120 limit throughout each analyzed transient.

The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months periodic surveillance of these parameters through instrument R21 readout is sufficient to ensure that the parameters are restored within their limits following load changes and other expected transient operation.

The flow parameters indicated in Figure 3.2-1 have been rounded down to bias the analysis in the conservative direction.

f SEQUOYAH - UNIT 2

,B 3/4 2-4 Amendment 21, 130, 146,214 1

. - ~

e ADMIN 3STRATIVE CONTROLS pH I

MONTHLY REACTOR OPERATING REPORT 6.9.1.10 Routine reports of operating statistics and shutdown experience, including documentation of all challenges to the PORVs or Safety Valves, shall R64 be submitted on a monthly basis no later than the 15th of each month following the calendar month covered by the report.

1 CORE OPERATING LIMITS REPORT R146 6.9.1.14 Core oper,ating limits shall be established and documented in the CORE i

OPERATING LIMITS REPORT before each reload cycle or any remaining part of a j

reload cycle for the following.

1 1.

f ( AI) limits for overtemperature Delta T Trip Setpoints and f (AI) i 2

limits for overpower Delta T Trip Setpoints for Specification 2.2.1.

l 2.

Moderator Temperature Coefficient BOL and EOL limits and 300 ppm surveillance limit for Specification 3/4.1.1.3, 3.

Shutdown Bank Insertion Limit for Specification 3/4.1.3.5, 4.

Control Bank Insertion Limits for Specificacion 3/4.1.3.6, S.

AXIAL FLUX DIFFERENCE Limits for Specification 3/4.2.1, i

6.

Heat Flux Hot Channel Factor and K(z) for Specification 3/4.2.2, and 7

Nuclear Enthalpy Rise Hot channel Factor.for Specification 3/4.2.?.

6.9.1.14.a The analytical methods used to determine the core operating limits l[R146 shall be those previously reviewed and approved by NRC in:

1 l

1.

BAW-10180P-A, Rev. If "NEMO - NODAL EXPANSION METHOD OPTIMIZED",

)

March 1993. (FCF Proprietary)

(Methodology for Specification 3.1.1.3-Moderator Temperature Coefficient) 2.

BAW-10169P-A, "RSG PLANT SAFETY ANALYSIS - B&W SAFETY ANALYSIS j

METHODOLOGY FOR RECIRCULATING STEAM GENERATOR PLANTS", October 1989.

(FCF Proprietary)

(Methodology for Specification 3.1.1.3-Moderator Temperature Coefficient) 3.

BAW-10163P-A, Core Operating Limit Methodology for Westinghouse-Designed PWRs, June 1989. (FCF Proprietary)

(Methodology for Specification 2.2.1,

- Limiting safety System Settings (f ( AI), f ( AI) limits), 3.1.3.5 - Shutdown Bank 2

Insertion Limits, 3.1.3.6 - Control Bank Insertion Limits, j

3/4.2.1 - Axial Flux Difference, 3/4.2.2 - Heat Flux Hot Channel Factor, 3/4.2.3 - Nuclear Enthalpy Rise Hot Channel Factor) 4.

RAW-1016BP-A, Rev. 2, RSG IOCA - B&W Loss of Coolant Accident Evaluation Model for Recirculating Steam Generator Plants, (FCF Proprietary)

(Methodology for Specification 3/4.2.2 - Heat Flux Hot channel Factor) 5.

BAW-10168P-A, Rev 3, RSG LOCA - B&W Loss of Coolant Accident Evaluation Model for Recirculating Steam Generator Plants, (FCF Proprietary)

(Methodology for Specification 3/4.2.2 - Heat Flux Hot Channel Factor)

SEQUOYAH - UNIT 2 6-22 Amendment No. 44, 50, 64, 66, 107, 134, 142, 146, 161, 206,214

e f

INSTRUMENTATIOM i

j BASES l

REACTOR TRIP SYSTEM AND ENGINEERED SAFETY FEATURE ACTUATION SYSTEM l

INSTRUMENTATION (Continued) i The measurement of response, time at the specified frequencies provides i

assurance that the protective and the engineered safety feature actuation i

associated with each channel is completed within the time limit assumed in the l

accident analyses. No credit wr.0 taken in the analyses for those channels with response times indicated as not applicable in the updated final safety analysis e

j report.

R182 r

Response time may be demonstrated by any series of sequential, overlapping or total channel test measurements provided that such tests 7

demonstrate the total channel response time as defined.

Sensor response time i

verification may be demonstrated by either 1) in place, onsite or offsite test j

measurements or 2) utilizing replacement sensors with certified response times.

t l

Action 15 of Table 3.3-1, Reacter Trip System Instrumentation, allows the

{

breaker to be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for the purpose of performing P46 j

maintenance.

The 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is msed on a Westinghoure analysis performed in i

WCAP-10271, Supplement 1, whi h d$termines bypast breaker availability.

I The placing of a channel in the trip condition provides the safety l-function of the channel.

If the channel is tripped for testing and no other BR-10

{

condition would have indicated inoperability, the channel should not be J

i declared inoperable.

4 l

j 3/4.3.3 MONITORING INSTRUMENTATION l-3/4.3.3.1-RADIATION MCMITORING INSTRUMENTATION i

j The OPERABILITY of the radiation monitoring channels ensures that 1) the radiation levels are continually measured in the areas served by the individual channels and.2) the alarm or automatic action is initiated when the radiation level trip setpoint is exceeded.

3/4.3.3.2 MOVABLE INCORE DETECTORS i

The OPERABILITY of the movable incere detectors with the specified minimum complement of equipment ensures that the measurements obtained from use j

of this system accurately represent the spatial neutron flux distribution of a'

the reactor core.

The OPERABILITY of this system is demonstrated by irradiating each detector used and determining the acceptability of its voltage curve.

f f

For the purpose of measuring F (X,Y,2) or Fa(X,Y) a full incore flux map g

is used.

Quarter-core flux maps, as defined in WCAP-8648, June 1976, may be

{

used in recalibration of the excore neutron flux detection system, and full incore flux maps or symmetric incere thimbles may be used for monitoring the QUADRANT POWER TILT RATIO when one Power Range Channel is inoperable.

j j

3/4.3.3.3 SEISMIC INSTRUMENTATION j

l l

The OPERABILITY of the seismic instrumentation ensures that sufficient capability-is available to promptly determine the magnitude of a seismic event and evaluate the response of those features important to safety. This capability is required to permit comparison of the measured response to that used in the R72 SEQUOYAH - UNIT 2 B 3/4 3-2 Amendment Nos. 46, 72, 182,214 1

e ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPOPT (continued) 6.

WCAP-10054-P-A, Westinghouse Small Break ECCS Evaluation Model Using 4

the NOTRUMP Code, August 1985, (E Proprietary)

(Methodology for Specification 3/4.2.2 - Heat Flux Hot Channel Factor) 7.

WCAP-10266-P-A, Rev. 2, *THE 1981 REVISION OF WESTINGHOUSE EVALUATION MODEL USING RASH CODE", March 1987, (W Proprietary).

(Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor).

6.9.1.14.b The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met.

6.9.1.14.c THE CORE OPERATING LIMITS REPORT shall be provided within 30 days after cycle start-up (Mode 2) for each reload cycle or within 30 days of.

issuance of any midcycle revision of the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.

SPECIAL REPORTS 6.9.2.1 Special reports shall be submitted within the time period specified for each report, in accordance with 10 CFR 50.4.

R64 6.9.2.2 Diesel Generator Reliability morovement Proersm.

As a minimum the Reliability Improvement Program report for NRC audit, required R44 by LCO 3.8.1.1, Table 4.8-1, shall include:

(a) a summary of all tests (valid and invalid) that occurred within the time period over which the last 20/100 valid tests were performed (b) analysis of failures and determination of root causes of failures (c) evaluation of each of the recommendations of NUREG/CR-0660, " Enhancement of Onsite Emergency Diesel Generator Reliability in Operating Reactors,"

with respect to their application to the Plant (d) identification of all actions taken or to be taken to 1) correct the root causes of failures defined in b) above and 2) achieve a general improvement of diesel generator reliability (e).

the schedule for implementation of each action from d) above (f) an assessment of the existing reliability of electric power to engineered-safety-feature equipment 1

1 SEQUOYAH - UNIT 2 6-22a Amendment Nos. 44, 50, 64, 66, 107, 134, 146, 206,;214

ML20137Y883

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