ML20195D097

Amend 75 to License NPF-75,amending Tech Specs to Allow Refueling & Operating W/Vantage 5 Improved Fuel Design in Combination W/Westinghouse Low Parasitic Fuel Assemblies Remaining in Core from Cycle 4

ML20195D097
Person / Time
Site:	Summer
Issue date:	10/28/1988
From:	Adensam E Office of Nuclear Reactor Regulation
To:
Shared Package
ML20195D105	List: ML20195D097 ML20195D107
References
NUDOCS 8811040192
Download: ML20195D097 (55)

Text

.

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UNITED STATES 8

NUCLEAR REGULATORY COMMISSION e

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WASHING TON, D. C. 20655

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SOUTH CAROLINA ELECTRIC & GAS CCMPANY SOUTH CAROLINA PUBLIC SERVICE AUTHORITY DOCKET NO. 50-395 VIRGIL C. SUf'JtER NUCLEAR STATION, UNIT NO.1 ANENDMEllT TO FACILITY OPERATING LICENSE Amendment No. 75 License No. NPF-12 1.

The Nuclear Regulatory Ccmission (the Comission) has found that:

A.

The application for amendment by South Carolina Electric & Gas Company (the licensee), dated May 20, 1988, as supplemented June 20,1988, July 8,1988, August 5,1988, September 16,19ES, September 30, 1986, October 11, 1988, October 13, 1988, and October 24, 1988, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; 8.

The f acility will operate in conformity with the application, the provisions ; the Act, and the rules and regulations of the Comission; i

C.

There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; O.

The issuance of this arnendirent will not be inimical to the comon defense and security or to the halth and safety of the public; and e

E.

The issuance of this arnendment is ir. accordance with 10 CFR Part 51 of the eemission's regulations ano all applicable requirerents i

have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifications, as indicated in thc attachment to this license atendment; and paragraph 2.C.(2) of Facility Operating License No. NPF-12 is hereby amended tc read as i311cws:

l G911040192 881028 ADOCK0500g5 PDR e

i

2 (2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 75

, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license.

South Caroline Electric & Gas Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

This amendrent is effective as r;f its date of issuance, and shall be implemented within 30 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION 4 2 >

'N Elinor G. Adensam. D' rector Project Directorate !!-1 Division of Reactor Projects I/II

Attachment:

Changes to the Tecnnical Specifications Date of Issuance:

October 28, 1988 e-er l

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J ATTACHMENT TO LICENSE AMENOMENT NO. 75 TO FACILITY OPERATING LICENSE NO. NPF-12 00CKET NO. 50-395 Replace the following pages of the Appendix "A" Technical Specificatior.s with the enclosed pages.

The revised pages are identified by amendment number and contain vertical lines indicating the areas of change. Corresponding overledf pages are also provided to maintain document completeness.

Remove Pages Insert Pages 2-1 2-1 2-2 2-2 2-5 2-5 2-8 2-8 2-9 2-9 2-10 2-10 B 2-1 8 2-1 8 2-2 B 2-2 B 2-4 B 2-4 3/4 1-3a 3/4 1-3a 3/4 1-4 3/4 1-4 3/4 1-5 3/4 1-5 3/4 1-Sa l

3/4 1-6 3/4 1-6 (overleaf) 3/4 1-11 3/4 1-11 3/4 1-12 3/4 1-12 (overleaf) 3/4 1-19 3/4 1-19

[

3/4 1-20 3/4 1-20 (overleaf) 3/4 P:1 3/4 2-1 3/a 1:2 3/4 ?-2 3/4 2-3 3/4 2-3 3/4 2-4 3/4 2-4 l

3/4 2 5 3/4 2-5

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3/4 2-6 3/4 2-6 3/4 2 64 3/4 2-6b 3/4 2-6c i

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- l Remove Pages Insert Pages 3/4 2-7 3/4 2-7 3/4 2-8 C/4 2-8 3/4 2-9 3/4 2-9 3/4 2-10 3/4 2-10 3/4 2-15 3/42-15(overleaf, 3/4 2-16 3/4 2-16 3/4 3-9 3/4 3-9 3/4 3-10 3/4 3-10 l

3/4 5-1 3/4 5-1 3/4 5-2 3/4 5-2 (overleaf)

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3/4 5-5 3/4 5-5 (overleaf) 3/4 5-6 3/4 5-6 I

3/4 10-1 3/410-1(overleaf) 3/4 10-2 3/4 10-2 7

B 3/4 1-2 B 3/4 1-2 j

B 3/4 1-3 8 3/4 1-3 j

B 3/4 2-1 B 3/4 2-1 l

B 3/4 2-2 8 3/4 2-2 B 3/4 2-3 B 3/4 2-3 l

B 3/4 2-4 8 3/4 2-4 B 3/4 2-5 B 3/4 2-5 l

B 3/4 2-6 l

B 3/4 5-1 B 3/4 S.1 6-18 6-18 i

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2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS i

REACTOR CORE 2.1.1 The combination of THERMAL POWER, pressurizer pressure, and the highest operating loop coolant temperature (Tavg) shall not exceed the limits shown in 3

Figures 2.1-1 for 3 loop operation.

l APPLICABILITY:

MODES 1 and 2.

ACTION:

Whenever the point defined by the corbination of the highest operating loop average temperature and THERMAL POWER has exceeded the appropriate pressurizer pressure line, be in HOT STANOBY within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, and comply with the require-L ments of Specification 6.7.1.

i RfACTOR COOLANT SYSTEM PRESSURE

[

2.1.2 The Reactor Coolant System pressure shall not exceed 2735 psig.

APPLICABILITY:

MODES 1, 2, 3, 4 and 5.

ACTION:

MODES 1 and 2 Whenever the Reactor Coolant System pressure has exceeded 2735 psig, be in HOT STANDBY with the Reactor Coolant System pressure within its limit within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, and comply with the requirements of Specification 6.7.1.

MODES 3, 4 and 5 Whenever the Reactor Coolant System pressure has exceeded 2735 psig, l

reduce the Reactor Coolant System pressure to within its limit within 5 minutes, and comply with the requirements of Specification 6.7.1.

l L

SUMER - UNIT 1 2-1 Amendment No. 75

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'C#EAAf0N 575 l' N )

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10 20 ' 30 40 50 60 70 80 90 100 110 120 POWER (PERCENT)

When operating in the reduced MTP replan of Technical spetifica tion 3.2.3 (figure 3.2 2) the retttittedpower levelmust be tensidered 100 % RTP for this figure.

Figure 2.1-1 Reactor Core Safety Limit - Three Loops in Operation 2-2 Amendment No. 85. 75

t TABLE 2.2-1 m

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS Total l

g Functional Unit Allowance (TA)

Z S_

Trip Setpoint Allowable Value 1

G 1.

Manual Reactor Trip Not Applicable NA NA NA NA y

2.

Power Range, Neutry F(ux 7.5 4.56 0

$109% of RTP 1111.2% of RTP High Setpoint Low Setpoint 8.3 4.56 0

<25% of RTP

<27.2% of RTP 3.

Power Range, Neutron Flux 1.6 0.5 0

< 5?.' RTP with

<6.3% of RTP with High Positive Rate i tt e constant i time constant 12 seconds

>2 seconds 4.

Power Range, Neutron Flux 1.6 0.5 0

<S% af RTP with

<6.3% of RTP with High Negative Rate i time constant a time constant m

J, 12 seconds 12 seconds 5.

Intermediate Range, 17.0 8.4 0

<25% of RTP-

<31% of RTP Neutron Flux

~

6.

Source Range, Neutron Flux 17.0 10.0 0

1105 cps

$1.4 x 105 cps 7.

Overtemperature AT 9.8 7.29 1.9 See note 1 See note 2

& 1.2**

8.

Overpower AT 5.2 2.26 1.9 See note 3 See note 4 9.

Pressurizer Pressure-Low 3.1 0.71 1.5 11870 psig 11859 psig

10. Pressurizer Pressure-High 3.1 0.71 1.5

$2380 psig

<2391 psig

11. Pressurizer Water Level-High 5.0 2.18 1.5 192% of instrument 193.8% of instrument 2

span span

?

12. Loss of Flow 2.5 1.0

1. 5

>90% of loop

>89.2% of loop Besign flow

• Besign flow

• w "toop design flow = %,500 gpm RIP - RATED THERMAL POWER

• • 1.9% span for Delta-T (RIDS) and 1.2% for Pressurizer Pressure.

l

m._.

= _ -

.Z,;

TABLE 2.2-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETFOINTS l

=

l NOTATION b NOTE 1:

OVERTEMPERATURE AT w

~

AT 1 AT, [K -K]

[T - T'] + K (P - P') - f (al)]

2 3

i

't k

Where:

AT

=

Measured AT by RTO Manifold Instrumentation AT,.

1 Indicated AT at RATED THERMAL POWER K

1 1.203 j

i K

1 0.03006 2

f The function generated by the lead-lag controller for T,yg

=

dynamic compensation oo Time constants utilized in lead-leg controller for T,yg, t t i, r2

=

),28 secs.,

I21 4 SECS-T

=

Average temperature. *F T'

1 587.4*F Reference T,yg at RATED THERMAL POWER K3 0.00147 l

P

=

Pressurizer pressure, psig k

P' 1

2235 psig, Nominal RCS operating pressure 1

=

k 5

=

Laplace transform operator, sec 8 A

.F N,

(p

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.we----.,.-,

-+

n-

TABLE 2.2-1 (Continued)

REACTOR TRI? SYSTEM IMSTRUMENTATION TRIP SETPOINTS NOTATION (Continued)

C NOTE 1:

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

II)

I*# 9 9 between - 24 percent and + 4 percent f (a!) = 0 where q and q are percent l

t t b

t b

RATED THERMAL POWER in the top and bottom halves of the core respectively, and qt

• 9 iS b

total THERMAL POWER in percent of RATED THERMAL POWER.

l (ii) for each percent that the magnitude of q g exceeds -24 percent, the AT trip setpoint shall be automatically reduced by 2.27 percent of its value at RATED THERMAL POWER.

(iii) for each percent that the magnitude of q A exceeds +4 percent, the aT trip setpoint t

b shall be automatically reduced by 2.13 percent of its value at RATED THERMAL POWER.

NOTE 2:

The channel's maximum trip setpoint shall not exceed its computed trip point by more than 2.0 percent AT Span.

l NOTE 3:

OVERPOWER AT fg['5 AT $ AT, [K4 - Ks T - Ks [T - T"))

l

{

Where:

AT

=

as defined in Note 1 AT, as defined in Note 1

=

[

K 1

1.0875 4

P 0.02/*F for increasing average temperature and 0 for decreasing average Ks g

temperature l

3 [' S The function generated by the rate-lag controller for T,,g dynamic compensation

=

TABLE 2.2-1 (Continued)

REACTOR TRIP SYSI G INSTRUNENTATION TRIP SETPOINTS NOTATION (Continued)

E

] NOTE 3 (continued)

Time constant utilized in the rate-lag controller for Tavg, 13 > 10 secs.

l

=

Ta g, g,

0.00156/*F for T > T" and K. = 0 for T $ I" l

4 as defined in Note 1 T

=

I" 587.4*F Reference T,,g at RAlED THERMAL POWER as defined in Note 1 S

=

b NOTE 4:

The channel's maximum trip setpoint shall not exceed its computed trip point N more than 2.0 percent AT Span.

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u t

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2.1 SAFETY LIMITS BASES 2.1.1 REACTOR CORE The restrictions of this, hfety Limit prevent overheating of the fuel and possible cladding perforation which would result in the release of fission products to the reactor coolant.

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

Operation above the upper boundary of the nucleate belling regime could result in excessive cladding temperatures because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient.

DNB is not a dir?ctly measurable parameter during operati0n ar,d therefore THERMAL '0WER and keactor Co lant Tamperature and Pressure Nye been r01ated to DNB.

This relation has been developed to predict the DNB flux and the location of DNB for axially uniform and non-uniform heat flux riistributions.

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

The DNB design basis is as follows:

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

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

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

The uncertainties in the above plant parameters are used to dete*mine the plant DNBR uncertainty.

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

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

The curves of Figures 2.1-1 show the loci of points of THiRMAL POWER, Reactor Coolant System pressure and average temperature below which the calculated r

DNBR is no less than the design DNBR value or the average enthalpy at the vessel exit is less than the enthalpy of saturated liquid.

SUMMER - u.41T 1 B 2-1 Amendment No. 75

p SAFETY LIMITS BASES L

REACTOR CORE (Continued)

Thesecurvesarebasedonanenthalpyhotchannelfactor,Ffg,of1.56 (includes measurement uncertainty) and a reference cosine with a peak of 1.55 N

for axial power shape.

An allowance is included for an increase in F at 3g reduced power based on the expression:

Fh = 1.56 [1+ 0.3 (1-P)]

where P is the fraction Of RATED THERMAL POWER These limiting heat flux conditions are higher than those calculated for the range of all control rods fully withdrawn to the maximum allowable control rod insertion assuming the axial power imbalance is within the limits of the fg (delta I) function of the Overtemperature tr y.

When the axial power imbalance is not within the tolerance, the axial power imbalance effect on the Overtemperature delta T trips will reduce the setpoints to provide protection consistent with core safety limits.

2.1.2 REACTOR COOLANT SYSTEM PRESSURE The restriction of this Safety Limit protects the integrity of the Reactor Coolant System from overpressurization and thereby prevents the release of radio-nuclides contained in the reactor coelant from reaching the containment atmosphere.

The reactor pressure vessel and pressurizer are designed to Section III of the ASME Code for Nuclear Power Plants, 1971 Edition which permits a maximum transient pressure of 110% (2735 psig) of design pressure.

The Reactor Coolant System piping, valves and fittings, are also designed to Section III of the ASME Code for Nuclear oower Plants, 1971 Edition which permits a maximum transient pressure of 120% (2985 nsig) of component design pressure.

The Safety Limit of 2735 psig is therefor. a nsistent with the design criteria and associated code requirements.

The entire Reactor Coolant System is hydrotested at 3107 psig 125% of design pressure, 6 demonstrate integrity prior to initial operation.

i I

SUMER - UNIT 1 B 2-2 Amendment No. 75 b

O LIMITING SAFETY SYSTEM SETTINGS BASES REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS (Continued)

The various reactor trip circuits automatically open the reactor trip breakers whenever a condition monitored by the Reactor Protection System reaches a preset or calculated level.

In addition to redundant channels and trains, the design approach provides a Reactor Protection System which monitors numerous system variables, therefore, providing protection system functional diversity.

The Reactor Protection System initiates a turbine trip signal whenever raactor trip is initiated.

This prevents the reactivity insertion tnat would otherwise result from excessive reactor system cooldown and thus avoids unnecessary actuation of the Engineered Safety Features Actuation System.

Manual Reactor Trip l

The Reactor Protection System includes manual reactor trip capability.

Power Ranee, Neutron Flux l

In each of the Power Range Neutron Flux channels there are two independant bistables, each with its own trip setting used for a high and low range trip setting.

The low setpoint trip provides protection during subcritical and low power operations to mitigate the consequences of a power excursion beginning i

from low power, and the high setpoint trip provides protection during power operations to mitigate the consequences of a reactivity excursion from all power levels.

l The low setpoint trip may be manually blocked above P-10 (a power level i

of anproximately 10 percent of RATED THERMA POWER) and is automatically reinstated below the P-10 setpoint.

Power Range, Neutron Flux, High Rates t

The Power Range Pocitive Rate trip provides protection against rapid flux increases which are characteristic of a rupture of a control rod drive housing.

Specifically, this trip complements the Power Range Neutron Flux High and Low i

trips to ensure that the criteria are met for rod ejection from mid power.

I' p

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

Mt high' power, a rod drop accident of a single c.t multiple

[

rods could cause local flux peaking which could cause an unconservative local I

DNBR to exist.

The Power Range Negative Rate trip will prevent this from 1

occurring by tripping the reactor.

No credit is taken for operation of the Power Range Negative Rate trip for those control rod drop accidents for which l

DNBR's will be greater than the limit value.

[

i Intermediate and Source Range, Nuclear Flux

[

The Intermediate and Source Range, Nuclear Flux trips provide reactor core protection during reactor startup to mitigate the consequences of an i

SU M ER - UNIT 1 B 2-4 Amendment Co. 75 i

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O REACTIVITY CONTROL SYSTEMS MODERATOR TEMPERATURE COEFFICIENT LIMITING CONDITION FOR OPERATION 3.1.1. 3 The moderator temperature coefficient (MTC) shall be:

a.

Less positive than the limits shown in Figure 3.1-0, and b.

Less negative than -5.0 x 10 4 delta k/k/'F for the all rous l

withdrawn, end of cycle life (EOL), RATED THERMAL POWER condition.

APPLICARILITY:

Specification 3.1.1.3.a - MODES 1 and 2* only#

Specification 3.1.1.3.b - MODES 1, 2 and 3 only#

ACTION:

a.

With the MTC more positive than the limit of 3.1.1.3.a above, operation in MODES 1 and 2 may proceed provided:

1.

Control rod withdrawal limits are established and maintained sufficient to restore the MTC to less positive than the limits shown in Figure 3.1-0 within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

These withdrawal limits shall be in addition to the insertion limits of Specification 3.1.3.6.

2.

The control rods are maintained within the withdrawal limits established above until a subsequent calculation verifies that the MTC has been restored to within its limit for the all rods withdrawn condition.

3.

In lieu of any other report required by Specification 6.9.1, a Special Report is prepared and submitted to the Commission pursuant to Specification 6.9.2 within 10 days, describing the value of the measured MTC, the interim control rod withdrawal limits and the predicted average core burnup necessary for restoring the positive MTC to within its limit for the all rods witgdrawncondition.

b.

With the MTC more negative than the limit of 3.1.1.3.b above, be in HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

"With K,ff greater than or equal to 1.0

1. See Special Test Exception 3.10.3 SUMMER - UNIT 1 3/4 1-4 Amendment No. 75

O REACTIVITY CONTROL SYSTEMS SURVEILLANOE REQUIREMENTS 4.1.1.3 The MTC shall be determined to be witiin its limits during each fuel cycle as follows:

a.

The MTC shall be measured and comparei to the BOL limit of Specifi-

)

cation 3.1.1.3.a. above, prior to init ial operation above 5% of RATED THERMAL POWER, af ter each fuel loading.

b.

The MTC shall be measured at any THERMAL POWER and compared to

-4.1 x 10 4 delta k/k/*F (all rods withdrawn, RATED THERMAL POWER l

cordition) within 7 EFP0 after reaching an equilibrium boron concen-tretion of 300 ppm.

In the event this comparison indicates the MTC is more negative than -4.1 x 10 4 delta k/k/*F the MTC shall be

[

remeasured, and compared to the EOL MTC limit of specification 3.1.1.3.b, at least once per 14 EFPD during the remainder of the fuel cycle.

.z-

)

SUMMER - UNIT 1 3/4 1-5 Amendment No. 75

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% OF RATfD TNERMAL POWER e

ROURE 3.10 MODERATOR TIMPERATURE COtFFICIENT V5 POWER LIVil.

SUK'iER - UNIT 1 3/4 1-53 Amendment No. 75

O REETIVITY CONTROL SYSTEMS MINIMlm TEMPERATURE FOR CRITICALITY LIMITING CON 0! TION FOR OPERATION 3.1.1.4 The Reactor Coolant Systeo lowest operating loop temperature (T'V9) shall be greater than or equal to 551*F.

APPLICA81LITY: MODES 1 and 2 ACTION:

With a Reactor Coolant Sy', tem operating loop temperature (T *Se) less than 551'F, restore T towithinitslimitwithin15minutesof in H0T STAN08YwithintMSnext.15 minutes.

SURVEILLANCE REQUIREMENTS 4.1.1.4 The R4 actor Coolant System temperature (T**9) shall be determined to be greater than or equal to 551'F:

a.

Within 15 minutes prior to achieving reactor criticality, and b.

At least once per 30 minutes when the reactor is critical and the Reactor Coolant System T is less than 561*F with the T

-T DeviationAlarenotresef.vg avg ref 1

fWith K grotter than or equal to 1.0.

• See spina 1 Test Exception 3.10.3.

I Sure4ER - UNIT 1 3/4 1-6 c

O

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

a.

A boric acid storage system with:

1.

A minimum contained borated water volume of 2700 gallons, 2.

Between 7000 and 7700 ppm of boron, and 3.

A minimum solution temperature of 65'F.

b.

The refueling water storage tank with:

1.

A minimum contained borated water volume of 51,500 gallons, l

2.

A minimum boron concentration cf 2300 ppe, and 3.

A minimum solution temperature of 40*F.

APPLICABILITY:

MODES 5 and 6.

ACTION:

With no borated water source OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes.

SURVEILLANCE REQUIREMENTS 4.1.2.5 The above required borated water source shall be demonstrated OPERABLE:

a.

At least once per 7 days by:

1.

Ve dfying the boron concentration of the water, 2.

Vedfying the contained borated water volume, and 3.

Verifying the boric acid storage tank solution temperature when it is the source of borated water, b.

At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by verifying the RWST temperature when it is the source of borated water and the outside air temperature is l

less than 40'F.

SW M R - UNIT 1 3/4 1-11 Amendment No. fl. 75 1

REACTIVITY CONTROL SYSTEMS BORATED WATER SOURCES - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.6 As a minimum, the following borated water source (s) shall be OPERABLE

~

as required by 3pecification 3.1.2.2:

a.

A boric acid storage system with:

1.

A minimum gontained borated water volume of 13,200 gallons, 2.

Between 700: and 7700 ppe of boron, and 3.

A minimum solution temperature of 65"F.

b.

The refueling water storage tank with:

1.

A minimum contained borated water volume of 453,800 gallons, 2.

A minimum boron concentration of 2300 ppe, and 3.

A minisua solution temperature of 40*F.

APPLICABILITY:

H0 DES 1, 2, 3 and 4.

ACTION:

a.

With the boric acid storage system inoperable and being used as one of the above required berated water sources, restore the storage system to 0PERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at leest HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and borated to a SHUTDOWN MARGIN equivalent to at least 2 percent delta k/k ct 200*F; restore the boric acid storage system to OPERA 8LE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

l b.

With the refueling water storage tank inoperable, restore the tank to OPERA 8LE status within one hour or be in at Isant HOT STAN08Y withtrethe next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SUMMER - UNIT 1 3/4 1-12

1. aendment No. 61 MAS 1. :25c

o 1

REACTIVITY CONTROL SYSTEMS R00 DROP TIME LIMITING CONDITION FOR OPERATION 3.1.3.4 The individual full length (shutdown and control) rod drop time from the fully withdrawn position shall be loss than or equal to 2.7 seconds from l

beginning of decay of stationary gripper coil voltage to dashpot entry with:

I T,yg greater than or equal to 551*F, and a.

b.

All reactor coolant pumps operating.

APPLICABILITY:

MODES 1 and 2.

ACTION:

With the drop time of any full length rod determined to exceed the above limit, i

restore the rod drop time to within the above limit prior to proceeding to MODE 1 or 2.

t SURVEILLANCE REQUIREMENTS t

4.1.3.4 The rod drop time of full length rods shall be demonstrated through measurement prior to reactor criticality:

a.

For all rods following each removai of the reactor vessel head, i

b.

For specifically affected individual rods following any maintenance on or modifiestion to the control rod drive system which could affect the drop time of those specific rods, and c.

At least once per 18 months.

l t

Jr er

(

1 l

\\

l I

i SUMMER. UNIT 1 3/4 1-19 Amendment h'o.

75

[

l r

REACTIVITY CONTROL SYSTEMS SHUT 00W R00 INSERTION LIMIT LIMITING CONDITION FOR OPERATION 3.1.3.5 All shutdown rods shall be fully withdrawn.

ADPLICA81LITY:

MODES 1* and 2*#.

ACTION:

With a maximum of one shutdown rod not fully withdrawn, except for surveillance testing pursuant to Specification 4.1.3.1.2, within one hour either:

a.

Fully withdraw the red, or b.

Declare the rod to be inoperable and apply Specification 3.1.3.1.

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

a.

Within 15 minutes prior to withdrawal of any rods in control banks A, 8, C or D during an approach to reactor criticality, aid b.

At least once per 1.1 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> thereafter.

"See Special Test Exceptions 3.10.2 and 3.10.3.

fWith K,ff greater than or equal to 1.0 c

l SUI 44ER - UNIT 1 3/4 1-20

o I

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

LIMITING CON 0! TION FOR OPERATION 3.2.1 The indicated AXIAL FLUX DIFFERENCE (AFD) shall be maintained within:

a, the allowed operational space defined in the Peaking Factor Limit Report (PFLR) for Relaxed Axial Offset Control (RAOC) operation, or b.

within the target band specified in the PFLR about the target flux

[

difference during base load operation.

APPLICABILITY:

MODE 1 above 50% of RATED THERMAL POWER *.

ACTION:

a.

For RAOC operation with the indicated AFD outside of the applicable limits specified in the PFLR, 1.

Either restore the indicated AFD to within the PFLR specified

+

limits within 15 minutes, or 2.

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

For Base Load operation above APLND** with the indicated AFD outside

[

of the applicable target band about the target flux differences.

f 1.

Either restore the indicated AFD to within the PFLR specified target band within 15 minutes, or ND 2.

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

c.

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

"See Special Test Exception 3.10.2 ND

• • APL is the minimum allowable power level for base load operation and will

[

be provided in the Peaking Factor Limit Report per Specification 6.9.1.11.

SUMER - UNIT 1 3/4 2-1 Amendment No. 75

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

a.

Monitoring the indicated AFD for each OPERABLE excore channel at least once per 7 days when the AFD Monitor Alarm is OPERABLE:

b.

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

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

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

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

The provisions of Specifiestion 4.0.4 are not applicable.

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

The provisions of Specification 4.0.4 are not applicable.

t er i

SUMMER - UNIT 1 3/4 2-2 Amendment No. 75

f This Figure deleted SHER - UNIT 1 3/4 2-3 bendment No. 75 i

POWER DISTRIBUTION LIMITS 3/4.2.2 HEATFLUXHOTCHANNELFACTOR-Fg LIMITING CON 0! TION FOR OPERATION 3.2.2 F (z) shall be limited by the following relationships:

l q

F (z) < (T) (K(z)] for P > 0.5 2.45 l

0 l

F (z) $ (4.9) (K(z)] for P < 0.5 q

l where P _ THERMAL POWER RATED THERMAL POWER I

and K(z) is the functior, obtained from Figure 3.2-1 for a given core height location, l

j f

APPLICABILITY: MODE 1.

ACTION.

With F (Z) exceeding its limit:

9 a.

Reduce THERMAL POWER at least 1% for each 1% F (z) exceeds the 9

limit within 15 minutes and similarly reduce the Power Range Neutron Flux-High Trip Setpoints within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; 00WER l

OPERATION may proceed for up to a total of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />; subsequent l

POWER OPERATION may proceed provided the Overpower delta T Trip l

Setpoints have been reduced at least 1% for each 1% F (z)

I exceeds the limit.

q b.

Identify and correct the cause of the out of limit condition prior i

l to increasing THERMAL POWER above the reduced limit required by a, j

above; THERMAL POWER may then be increased provided F (t) is demon-f I

strated through incore mapping to be within its limitq I

i t

f f

i i

f I

l t

SUM ER - UNIT 1 3/4 2-4 Amendment No. #6, 75 1

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

4.2.2.2 For RAOC operation, F (z) shall be evaluated to determine if F (z) is within its litrit by:

q q

a.

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

b.

Increasing the measured F (z) component of the power distribution q

map by 3% to account for manufacturing tolerances and further increc3ing the value by 5% to account for measurement uncertainties.

Verify the requirements of Specification 3.2.2 are satisfied.

c.

Satisfying the following relationship:

F (z) 1 p245j z) for P > 0.5 F (z) 1 w#

f for P 1 0.5 2

where F (.) is the m asured F (z) increased by the allowances q

for manufacturing tolerances and measurement uncertainty, 2.45 is the F limit, K(z) is given in Figure 3.2-1, P is the relative o

THERML POWER, and W(z) is the cycle dependent funct'on that accounts for power distribution transients encountered during normal operation.

This function is given in the Peaking Factor Limit Report as per Specification 6.9.1.11.

d.

Measuring F (Z) according to the following schedule:

1.

Upon achieving equilibrium conditions after exceeding by 10%

or % re of RATED THERMAL POWER, the THERMAL POWER at which F (a.) was,.last determined,

• or q

2.

At least once per 31 Effective Full Power D3ys, whichever occurs fi:st.

"Ouring 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.

SUP94ER - UNIT 1 3/4 2-5 Amendment No. 75

POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Contidued) e.

With the maximum value of F (z)

K(z) over the core height (z) increasing since the previous determination of F (z) either of the following actions shall be taken:

(1) F (z) shall be increased by 'A over that specified in Specification 4.2.2.2c. or (2) F (z) shall be measured at least once per 7 Effective Full Power Days until two successive maps indicate that the maximum value of F (z)

K(z) over the core height (2) is not increasing, f.

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

(1) Calculate the maximum parcent over the ore height (z) that F (z) exceeds its limit by the following 3xpression:

q (Z)

• W(Z) -1 x 100 for P > 0.5

')

2.45 x K(z)

(,, T a

/

F

- F (z) x W(z)i. g x 100 for P < 0.5

~

T. (z) d 2.45 x K SUMER - UNIT 1 3/4 2-6 Menament No. 3, 75

POWER OISTRIBUTION LIMIT,5 SURVEILLANCE REQtlil"!MENTS (Continued)

(2) One of the following actions shall be taken:

(a) Within 15 minutes, control the AFD to within new AFD limits which are determined by reducing the applicable AFD limits by 1J. AFD for each percent F (z) exceeds its limits as 9

determined in Specification 4.2.2.2f.1).

Within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, reset the AFD alarm setpoints to these modified limits, or (b) Comply with the requirements of Specification 3.2.2 for F (z) axceeding its limit by the percent calculated above, q

or (c) Verify that the requirements of Specification 4.2.2.3 for Base Load operation are satisfied and enter Base Load operation, g.

The limits specified in Specifications 4.2.2.2c., 4.2.2.2e., and 4.2.2.2f. above are not applicable in the following core plane regions:

1.

Lower core region from 0 tn 15%, inclusive.

2.

Upper core region from 85 to 100%, inclusive.

NO 4.2.2.3 Base Load operation is permitted at powers above APL if the following conditions are satisfied:

a.

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

Maintain Base Load operation surveillance (AFD within applicable target band about the target flux difference) during this time period.

Base Load operation is ND thenpegittedprovidingNfHERMALPOWERismaintainedbetweenAPL and APL or between APL and 100% (whichever is most limiting) and F surveillance is maintained pursuant to Specification 4.2.2.4.

OLq APL is, defined as the minimum value of:

~

gpgBL, _ (2.45 x K(z) x 100%

O J

F (z) x W(z)gt over the core height (z) where:

F (z) is the measured F (z) 9 increased by the aH owances for manufacturing tolerances and j

measurement unc(- Linty.

The F limit is 2.45.

K(z) is given in q

Figure 3.2-1.

W(z)BL is the cycle dependent function that accounts for limited power distribution transient encountered during base load operation.

The function is given in the Peaking Factor Limit Report as per Specification 6.9.1.11, SUtHER - UNIT 1 3/4 2-6a Amendment No. 75

POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued) 1 b.

During Base Load operation, if t5e THERMAL POWER is decreased belov ND APL then the conditions of 4.2.2.3.a shall be satisfied before re-entering Base Load operation.

4.2.2.4 During Base Load Operation F (z) shall be evaluated to determine if F (z) is within its limit by:

q q

a.

Using the movable incore detectors to obtain a power dittribution HD map at any THERMAL POWER above APL b.

Increasing the measured F (z) component of the power distribution 9

map by 3% to account for manufacturing tolerances and further increasing the value by 5% to account for measurement uncertainties.

Verify the requirements of Specification 3.2.2 are satisfied, i

c.

Sat'sfying the following relationship:

F (z) 1 z) for P> APLND 2 45 x g

where:

F (z) is the measured F (z).

The F limit is 2.45.

q q

K(z) is given in Figure 3.?-1.

P is the relative THERMAL POWER.,

W(z)m is the cycle dependent function that accounts for limited powe rdistribution transients encountered during normal operation.

This function is given in the Peaking Factor Limit Report as per Specification 6.9.1.11.

d.

Measuring F (z) in conjunction with target flux difference determination according to the following schedule:

1.

Prior to entering BASE LOAD operation after satisfying Sec-tion 4.2.2.3 unless a full core flux map hts been taken in the previous 31EFPDwitNDthe relative thermal ywer having been maip ained above APL for the 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to mapping, and 2.

At jeast 9nce per 31 Effective Full Power Days, e.

With the maximum value of F((z)

K(z) over the core height (z) increasing since the previous determination of F (Z) either of the following actions shall be taken:

J SUM ER - UNIT 1 3/4 2-6b Amendment No. 75 l

POWER DISTRIBUTION LIMITS 5,URVEILLANCE REQUIREMENTS (Continued) 1.

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

2.

F (z) shall be measured at least once per 7 Effective Full Power Days untfl 2 successive maps indicate that the maximum value of Ff(z)

(

Ktz) over the core height (z) is not increasing, j

4 f.

With the relationship specified in 4.2.2.4.c above not being i

satisfied, either of the following actions shall be taken:

1.

Place core in an equilibrium condition where the limit in a.2.2.2.cissatisfied,andremeasureF((z),or l

1 2.

Comply with the requirements of Specifica*. ion 3.2.2 for F (z) exceeding its limit by the maximum percent calculated j

q i

over the core height (z) with the following expression:

) X W(z)gt -

ND

'=

i

-1 (x100forP>APL i

2,,:,4,[ x g(,)

j l

~

4' P

l I

g.

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

1.

Lower core region 0 to 15 percent, inclusive.

)

4 r-r 2.

Upper core region 85 to 100 percent, inclusive.

I t

l 4.2.2.5 When F (z) is measured for reasons other than meeting the requirements 9

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

1 power distribution map and increased by 3% to account for manufacturing i

tolerances and further increased by 5% to account for measurement uncertainty.

t 1

i SUM ER - UNIT 1 3/4 2-6c Amendment No. 75

(

I r

1.2 1.1 1

l l

l 1

j I

.9

.8

.7

?

.6 I,4 I

.5 1

.Y

.3 t

Core -e amt t*O g

,2 00 10 60 10 i

? 2.0 0.925

.1 0

f 0

1 2

3 4

5 6

7 8

9 10 11 12 IOTTOM TOP oF CORE HilGHT, FT.

3, j

FUEL Fugt L

FIGURE 3.21 K(a) f 0.RMAUZED Fg(x) Al A FUNCTION OF CORE HEIGHT

[

k l

i l

i i

cum ER - Oh!T 1 3/4 2-7 Amendment No. 75 r

l

~

l I

POWER DISTRIBUTION LIMITS f

L 3/4.2.3 RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR LIMITING CON 0! TION FOR OPERATION l

I L

3.2.3 The combination of indicated Reactor Coolant System (RCS) total flow I

rate and R shall be maintained within the region of allowable operation shown or. Figure 3.2-2 for 3 loop operation.

g i

Where:

N R = 1.56 [1.0 + 0.3 (1.0 - P)]

a*

l 3

THERMAL POWER i

b*

P =

RATED THERMAL POWER l

FhaMeasuredvaluesofFhobtainedbyusingthemovableincore f

c.

detectors to obtain a power distribution map.

The measured valuesofFfg shall be used to calculate R since Figure 3.2-2 l

l includes measurement uncertainties of 2.1% for flow and 4% for incoremeasurementofFh,and j

f APPLICABILITY:

MODE 1.

a 1

ACTION:

With the combination of RCS total flow rate and R outside c e region of accept-l able operation shown on Figure 3.2 2:

{

{

a.

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

f 1.

Restore the combination of RCS total flow rate and R to withh.

j the above limits, or l

2.

Redur:e THERMAL POWER to less than 50% of RATED THERMAL POWER l

and reduce the Power Range Neutron Flux - High trip setpoint to j

l 1ess than or equal to 55% of RATED THERMAL P/ DER within the

]

next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

[

l b.

Within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of initially being outside the above limits, verify I

through incore flux mapping and RCS total flow rate comparison that the contdnation of R and RCS total flow rate are restored to within t

3 l

the above limits, or reduce THERMAL POWEn to less than 5% of RATED THERMAL PCWER within the next 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

c.

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

[

j to increasing THERMAL POWER above the reduced THERMAL POWER limit j

1 required by ACTION items a.2. and/or b. above; subsequent POWER y

j OPERATION may proceed provided that the combination of R and t

j indicated RCS total flow rate are demonstrated, through incore flux I

mapping and RCS total flow rate comparison, to be within the region t

j of acceptable operation shown on Figure 3.2-2 prior to exceeding the l

i following THERMAL POWER levels:

l SUMER - UNIT 1 3/4 2-8 AmendmentNo.ff,f9.75 f

f I

i

I L

POWER DISTRIBUTION LIMITS f

LIMITING CON 0! TION FOR OPERATION ACTION.

(Continued) j L

1.

A m inal 50% of RATED THERMAL POWER, 2.

A nominal 75% of RATED THERMAL POWER, and i

3.

Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of attaining greater then or equal tv 95% of RATED THERMAL POWER.

2 SURVE!LLANCE REQUIREMENTS 4

4.2.3.1 The provisions of Specification 4.0.4 are not applicable.

f t

4.2.3.2 The combination of indicated RCS total flow rate and R shall be

)

{

determined to be within the region of acceptable operation of Figure 3.2-2:

l

[

t a.

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

b.

At least once per 31 Effective Full Power Days.

I f

4.2.3.3 The indicated RCS total fluw rate shall be verified to be within the i

region of acceptable operation of Figure 3.2-2 at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when

[

l the most recently obtained value of R obtained per Specification 4.2.3.2, is assumed to exist.

4.2.3.4 The RCS total flow rate indicators shall be subjected to a CHANNEL CALIBRATION at least once per 18 months.

i 4.2.3.5 The RCS total flow rate shall be determined by measurement at least once per 18 months, i

.r-I

}

SUMER - UNIT 1 3/4 2-9 Amendment No. A), 75

MEASUREMENT UNCERTAINTIES OF 2.1% FOR PLOW AND 4.0% FOR INCORE MEASUREMENT OF FNAM ARE INCLUDED IN THl1 FIGURE 38 ACCtdfASLE UNACCEPTASL2 OPERATION REGION OPGRAfl0N REG 10N 36 l

I 34

.I

]32 I!

N II 30 I

3 y

(1.00.20 95) m i

(1.00.28 64)

• [l'.

(1 00.28 27) 28 (1 00.28 00) ggg gg7g i

l (1.00.27.79) n...

~ (100.27S) s 26 i

4 24

.9

.95 1

1.05 1.t l

n. m 3., t.se i1.0. o.x1.9))

nott: vn

,. v,4 es,.n. we==m.s so.e, www ma me teme.ees to to 100% of tates WhenmM power 4479) 9., Pagweg J.1 1 l

FIGURE 3.2-2 RCS TOTAL FLOW RATE VS. R THREE LOOP OPERATION r

SU MER - UNIT 1 3/4 2-10 Amendment No.,4), (0, 75

POWER O!STRIBUTION LIMITS 3/42.5 DNt PARAMETERS LIMITING CON 0! TION FOR OPERATION 3.2.5 The following DN8 related parameters shall be maintained within the limits shown on Table 3.2-1:

Reactor Coolant System T,yg.

a.

b.

Presst'rizer Pressure APPLICA8!LITY:

M00E 1.

ACTION:

With any of the above paraasters exceeding its limit, restore the parameter to within its limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to less than 55 of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, i

i SURVE!LLANCE_tEQUIREMENTS i

)

4.2.5 Each of the parameters of Table 3.2-1 shall be verified to be within l

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

er er l

l l

1 SUMMER - UNIT 1 3/4 2-15 Amendrent No. 75

(

- ]

~

l TABLE 3.2-1 m

h DNS PARAE TERS

=

E LIMITS

~

3 Loops In 2 Loops in

~

PARAETER Operation Operation

't It l

Reactor Coolant System 1

< 591.7'f avg Pressurizer Pressure

< 2206 psla*

M

=

7 E,

k

?.

x P

Limit not applicable during either a THENEL POWER ramp in excess of $1 of PATED THEIMAL POWER per minute or a THENEL POWER step in excess of IDE of RATED THEIMAL POWER.

y an These values left blank pending NRC approval of two-loop operation.

l i

• ,-,--,----e-m--

v,enr>.v--w---

-r-,,r se-

--an-e-,

n.--,,,

m---

w

,,w e

-- -~

,s n

ww- -

,- r w

w--

nem

--e,

--e,w n

au

O 1-l u.

TABLE 3.3-2

,3,,

REACTOR TRIP SYSTEM INSTRUE NTATION RESPONSE TIMES ao E

FUNCTIONAL UNIT RESPONSE TIE i

1.

Manual Reactor Trip Not Applicable

~

2.

Power Range, Neutypn (gux

$ 0.5 seconds

• l 3.

Power Range, Neutron Flux, High Positive Rate, Not Applicable 1

4.

Power Range, Neutron Flux, High Negative Rate

< 0.5 seconds

• 5.

Interinediate Range, Neutron Flux Not Applicable R

6.

Source Range, Neutron Flux Not Applic.able l

u E

7.

Overtemperature AT

$ 8.5 seconds

• l 8.

Overpower AT

$ 8.5 seconds

• 9.

Pressurizer Pressure--tow

$ 2.0 sec nds

10. Pressurizer Pressure--High

$ 2.0 seconds 11.

Pressurizer W ter Level--High Not Appilcable k

a 2

S

=

l y

Neutron detectors are exampt from response time testing. Response time of the neutron flux signal portion of the clannel shall be measured from detector output or input of first electronic component in channel.

v.

l

~

TABLE 3.3-2 (Continued)

KAC10R TRIP SYSTEM INSTRUENTATION KSPONSE TIES c

FUNCTIONAL UNIT RESPGNSE TIE 12.

A.

Loss of Flow - Single Loop (Above P-8)

< 1.0 seconds

'l TidoLoops 8.

Loss of Flow (Above P-7 and below P-8)

_ 1.0 seconds

13. Steam Generster idater Level--Low-Low

< 2.0 seconds 14.

Steam /Feessater Flow Mismatch and Low Steam Generator Water Level Not Applicable i

l

{

15. Undervoltage-Reactor Coolant Pumps

< 1.5 seconds

]

16. Underfrequency-Reactor Coolant Pumps

< 0.6 seconds l

o 17.

Turbine Trip A.

Low Fluid Oil Pressure Not Applicable b.

Turbine Stop Valve Closure Not Applicable 18.

Safety Injection Input from ESF Not Applicable 19.

Reactor Trip *.y 7 v Interlocks Not Applicable 20.

Reactor T. p trevars Not Applicable

{

21. Automatic Trip Logic Not Applicable a

I.

1 l

1 1

i

3/4.5 EMERGENCY CORE 2COLING SYSTEMS 3/4.5.1 ACCUMULATORS

~

LIMITING CONDITION FOR OPERATION 3.5.1 Each reactor coolant system accumulator shall be OPERABLE with:

a.

The isolation valve open, b.

A contained borated water voluma of between 7489 and 7685 gallons, c.

A boron concentration of between 2200 and 2500 ppm, and d.

A nitrogen cover pressure of between 600 and 656 psig.

APPLICABILITY:

MODES 1, 2 and 3.*

ACTION:

a.

With one accum'Jiator inoperable, except as a result of a closed isolation valve, restore the inoperable accumulator to OPERABLE status within one hour or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHdTOOWN within the /ollowing 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

b.

With one accumulator inoperable due to the isolation valve being closed, either immediately open the isolation valve or be in at least HOT STANDBY within one hour and in HOT SHUTDOWN within the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

E!1VEILLANCE REQUIREMENTS 1

4.5.1.1 Each accumulator shall be demonstrated OPERABLE:

a.

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

1.

Verifying the contained borated wate volume and nitrogen costr pressure in the tanks, and 2.

Verifying that each eccumulatot isolation valve is open.

"Pressurizer pressure above 1000 psig.

SUMMER - UNIT 1 3/4 5-1 Amendment No.

$J. 75

EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) b.

At least once per 31 days and within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after each s0lution volume increa>2 of greater than or equal to 1% of tank volume ey verifying the boron concentration of tha accumulator solution.

c.

At least once per 31 days when the RCS pressure is abeve 2000 psig by verifying that the isolation valve operator breaker opened at the motor control center and locked in the open position.

d.

At least once per 18 months by verifying that each accumulator isolation valve opens automatically under each of the following conditions:

1.

When an actual or a simulated RCS pressure signal exceeds the P-11 (Pressurizer Pressure Block of Safety Injection) setpoint, 2.

Upon receipt of a safety injection test signal.

er J

SUMMER - UNIT 1 3/4 5 2 Amendment No. 75

EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 2.

A visual inspection of the containment sump and verifying that the subsystem suction inlets are not restricted by debris and that the sump components (trash racks, screens, etc.) show no evidence of structural distress or abnormal corrosion.

e.

At least once per 18 months, during shutdown, by:

1.

Verifying that each automatic valve in the flow P4t

.-Jtes to its correct position on a safety injection actuation and containment sump recirculation test signal.

2.

Verifying that each of the following pumps start automatically upon receipt of a safety injection actuation test signal:

a)

Centrifugal charging pump b)

Residual heat removal pump f.

By verifying that each of the following pumps develops a differential pressure en recirculation flow when tested pursuant to Specifica-tion 4.0.5:

1.

Centrifugal charging pump 1 2472 psi I

2.

Residual heat removal pump 1 128 psi g.

By verifying the correct position of each inechanical position stop for the following ECCS throttle valves:

1.

Within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> following completion of each valv's stroking operation or maintenance on the valve when the ECCS subsystems are required to be OPERABLE.

P.

At least once per 18 months.

HPSI System Valve Number t.

899.6A t.

89968 c.

8996C d.

8994A e.

89948 l

f.

8994C l

g.

8989A l

h.

89898 i.

8989C l

j.

8991A k.

89918 1.

8991C SUMMER - UNIT 1 3/4 5-5 Amendment No. 75 l

l

EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) h.

By performing a flow balance test, during shutdown, following completion of modifications to the ECCS subsystems that alter the subsystem flow characteristics and verifying that:

1)

For centrifugal charging pump lines, with a single pump running and with recirculation flow:

l a)

The sum of the injection line flow rates, excluding the highest flow rate, is greater than or equal to 338 gpm, and b)

The total pump flow rate is less than or equal to 680 gpm.

i.

By performing a flow test, during shutdown, following completion of modifications to the ECCS subsystems that alter the subsystem flow characteristics and verifying that:

1)

For residual heat removal pump lines, with a single pump running the sum of the injection line flow rates is greater than or i

equal to 3663 gpm.

P i

d l

r l

er er SUMMER - UNIT 1 3/4 5-6 Amendment No.

75

+

3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN MARGIN LIMITING CONDITION FOR OPERATION 3.10.1 The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 may be suspended for measurement of control rod worth and shutdown margin provided reactivity equivalent to at least the highest estimated control rod worth is available for trip insertion from OPERABLE control rod (s),

APPLICABILITY:

MODE 2.

ACTION:

a.

With any full length control rod not fully inserted and with less than the above reactivity equivalent available for trip insertion immediately initiate and continue boration at greater than or equal to 30 gpa of a solution containing greater than or equal to 7000 ppe boron or its equivalent until the SHUTDOWN MARGIN required by Specification 3.1.1.1 is

restored, b.

With all full length control rods fully inserted and the reactor subcritical by less than the above reactivity equivalent, immediately initiate and continue boration at greater than or equal to 30 gpm of a solution contain-ing greater than or equal to 7000 ppm boron or its equivalent until the SHUTOOWN MARGIN required by Specification 3.1.1.1 is restored.

SURVEILLANCE REQUIREMENTS 4.10.1.1 The position of each full length rod either partially or fully withdrawn shall be determined at least once per 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

4.10.1.2 Each full length rod not fully inserted shall be demonstrated capable of full insertisa when tripped from at least the 502 withdrawn position within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to reducing the SHUT 00WN MARGIN to less than the limits of Specification 3.4:1.1. '

SUMMER - UNIT 1 3/4 10-1 Amendment No. 75 t

SPECIAL TEST EXCEPTIONS

-3/4.10.2 GROUP HEIGHT. INSERTION AND POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION 3.10.2 The group height, insertion and power distribution limits of Specifi-cations 3.1.3.1, 3.1.3.5, 3.1.3.6, 3.2.1 and 3.2.4 may be suspended during the performance of PHYSICS TESTS provided:

a.

The THERMAL POWER is maintained less than or equal to 85% of RATED THERMAL POWER, and b.

The limits of Specifications 3.2.2 and 3.2.3 are maintained and determined at the frequencies specified in Specification 4.10.2.2

below, i

APPLICABILITY:

MODE 1 ACTION:

With any of the limits of Specifications 3.2.2 or 3.2.3 being exceeded while the requirements of Specifications 3.1. 3.1, 3.1. 3. 5, 3.1. 3. 6, 3. 2.1 and 3. 2. 4 are suspended, either:

a.

Reduce THERMAL POWER sufficient to satisfy the ACTION require-ments of Specifications 3.2.2 and 3.2.3, or b.

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

SURVEILLANCE REQUI_RE,ENTS,.

4.10.2.1 The THERMAL POWER shall be determined to be less than or equal to 85% of RATED THERMAL POWER at least once per hour during PHYSICS TESTS.

4.10.2.2 The Surveillance Requirements of the below listed Specifications (t.

and b.) shall be performed at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during PHYSICS TESTS:

a.

Either Specifitations 4.2.2.2 or 4.2.2.4 and Specification 4.2.2.5.

F b.

Specification 4.2.3.2.

P I

t i

f SUMMER - UNIT 1 3/4 10-2 Amendment No. 75

~

REACTIVITY CONTROL SYSTEMS BASES MODERATOR TEMPERATURE COEFFICIENT (Continued) involved subtracting the incremontal change in the MDC associated with a core condition of all rods inserted (most positive MDC) to an all rods withdrawn condition and, a conversion for the rate of change of moderator density with temperature at RATED THERMAL POWER (.onditions.

This value of the MOC was then transforr.ied into the limiting MTC value -5.0 x 10 4 delta k/k/*F.

The MTC value of -4.1 x 10 4 delta k/k/ F represents a conservative value (with correc-tions for burnup and soluble boron) at a core condition of 300 ppm equilibrium boron concentration and is obtained by making these corrections to the limiting MTC value of -5.0 x 10 4 k/k/ F.

l The surveillance requirements for measurement of the MTC at the beginning and near the end of the fuel cycle are adequate to confirm that the MTC remains within its limits since this coefficient changes slowly due principally to the reduction in RCS l>oron concentration associated with fuel burnup.

3/4.1.1.4 MINIMUM TEMPERATURE FOR CRITICALITY This specification ensures that the reactor will not be made critical with the Reactor Coolant System average temperature less than 551*F.

This limitation is required to ensure 1) the moderator temperature coefficient is within its analynd temperature range, 2) the protective instrumentation is within its nom.41 operating range, 3) 1.he pressurizer is capable of being in an OPERABLE status with a steam bubble, and 4) the reactor pressure vestel is above its minimum RT temperature.

NDT 3/4.1.2 BORATION SYSTEMS L

The baron injection system entures that negative reactivity control is available durin0 each mode of f a.cility operation.

The components re mired to perform this function include 1) borated wator sources, 2) charging p wps,

3) separate flow paths, 4) boric acid transfer pumps, and 5) an emergency power supply from OPERABLE diesel generatora.

With the RCS average temperature above 200'F, a minimum of two boron in-jection flow paths are required to ensure single functional capability in the event an assumed failure renders one of the flow paths inoperable.

The boration capability of either flow path is sufficient to provide the required SHUT 00WN SUMMER - UNIT 1 B 3/4 1-2 Amendment No.

f). 75

REACTIVITY CONTROL SYSTlMS BASES BORATION SYSTEMS (Continued)

MARGIN from expected operating conditions of 1.77% delta k/k or as required by Figure 3.1-3 after xenon decay and cooldown to 200*F.

The maximum expected boration capability requirement occurs from full power equilibrium xenon condi-tions and is satisfied by 13269 gallons of 7000 ppm borated water from the boric acid storage tanks or 98631 gallons of 2300 ppm borated water from the refuel-ing water storage tank.

With the RCS temperature below 200*F, one injection system is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the additional restrictions prohibiting CORD ALTERATIONS and positive reactivity changes in the event the single injection system becomes inoperable.

The limitation for a maximum of one centrifugal charging pump to be OPERABLE and the Surveillance Requirement to verify all charging pumps except the required OPERABLE pump to be inoperable below 275*F provides assurance that a mass addition pressure transient can be relieved by the operation of a single PORV.

The boron capability required below 200*F is sufficient to provide the required SHUT 00WN MARGIN of 1 percent delta k/k or as required by Figure 3.1-3 after xenon decay and cooldown frora 200*F to 140*F.

This condition is satisfied by either 2000 gallons of 7000 ppm borated water from the boric acid storage tanks or 23266 gallons of 2300 ppm borated water from the refueling water storage l

tank.

The contained water volume limits include allowance for water not available l

because of discharge line location and other physical characteristics.

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

1 314.1.3 MOVABLE C0NTROL ASSEMBLIES l

The specifications of this section ensure that (1) acceptable power distribution lim Ks areJmaintained, (2) the minimum SHUTOOWN MARGIN is main-tained, and (3) limit the potential effects of rod misalignment on associated accident analyses.

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

SUMMER - UNIT 1 B 3/4 1-3 Amendment No.

61. 75

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

(1) maintaining the calculated ONBR in the core at or above the design limit during normal operation and in short-term transients, and (2) limiting the fission gas release, fuel pellet temperature, and cladding mechanical properties to within assumed design 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 (z)

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

heat flux on the surface cf a fuel rod at core elevation Z divided by the avnrage fuel rod heat flux, allowing for manufacturing tolerances on fuel pellets ar.d rods; F

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

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

3/4.2.1 AXIAL FLUX DIFFERENCE The limits on AXIAL FLUX DIFFERENCE (AFD) assure that the F (z) upper bound 9

envelope of 2.45 times the normalized axial peaking factor is not exceeded during sither normal operation or in the event of xenon redistribution following power changes.

l The limits on AFD will be provided in the Peaking Factor Limit Report (PFLR) per Technica! Specification 6.9.1.11.

Target flux difference is determined at equilibrium xenon enditions.

The l

full-length rcds leay be positioned within the core in accordance with their l

respective insertion limits and should be inserted near their normal position I

for steady-state oper.stion at high power levels.

The value of the target flux l

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

Target flux differences for other THERMAL POWER levels are obtained by multiplying the RATED THERMAL POWER value by the appropriate fractional THERMAL POWER level.

The periodic updating of the target flux diffe:ence value is necessary to reflect core burnup considerations, i,

i SUMMER UNIT 1 B 3/4 2-1 Amendment No f 6, 75 i

O POWER DISTRIBUTION LIMIT BASES AXIAL FLUX DIFFERENCE (Continued)

NU At power levels below Ahl, the limits on AFD are defined in the PFLR consistent with the Relaxed Axial Offset Control (RA00) operating procedure and limits.

These limits were calculated in a manner such that expected operational transients, e.g., load follow operations, would not result in the AFD deviating outside of those limits.

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

ND At power levels greater than APL

, two modes of operation are permissible; (1) RAOC, the AFD limit of which are defined in the PFLR and (2) Base Load operation, which is defined as the maintenance of the AFD within PFLR specifica-ND tions band abc.ut a target value.

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

To allow operation at the q

maximum permissible power level the Base Load operating procedure restricts the indicated AFD to relatively small target band (as specified in the PFLR) and pcwer swings (APLND < power < APLOL or 100% Rated Thermal Power, whichever is icwer).

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

Operation outside of the target band for che short time period allowed will not result in significant xenon redistribation se:h that the envelope of peaking factors would change sufficiently to prohibit continued operation in the powe* regiun defin M above.

To assure ti.are is no residual xenon redistribt. tion impact from past operation on the Base t. cad ND uperation, a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> waiting period at a power level above APL and allowed by RAOC is necessary.

During this time period load changes and rW notion are restricted to that allowed by the Base load procedure.

After the wait.ing period extended B n e Load operation is permissible.

The computer determines the one minute average of each of the OPERABLE excore detector 00fputs'a'nd provides an alarm message immediately if the AFD for at least 2 of 4 or 2 of 3 OPERABLE excore channels are:

(1) outside the allowed delta-I power operating space (for RAOC operation), or (2; outside the allowed delta-I target band (for Base Load operation).

These alarms are active when power is greater than:

(1) 50% of RATED THERMAL POWER (for RAOC operation), or (2) APLND (for Base Load operation).

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

SUMMER - UNIT 1 B 3/4 2 ",

Amendment No. 75

r-

~

O POWER DISTRIBUTION LIMIT BASES 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR and RCS FLOWRATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR The limits on heat flux hot channel factor, RCS flowrate, and nuclear enthalpj rise hot channel factor ensure that 1) the design limits on peak local power density and minimum DNBR are not exceeded and 2) in the event of a LOCA the peak fuel clad temperature will not exceed the 2200*F ECCS acceptance criteria limit.

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

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

a.

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

b.

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

c.

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

d.

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

F will be maintained within its limits provided conditions a. throdgh H

d. above are maintained.

As noted on Figure 3.2-2, RCS flow rate and F H may be"tradedoff"againstonegnother(i.e.,alowmeasuredRCSflowrateis acceptable if the measured F is also low) to ensure that the calculated DNBR 3g will not be below the design DNBR value.

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

R,ascalculatedin3.2.3andusedinFigure3.2-2,accountsforFh less than or equa M o 1.56.

This value is used in the various accident N

analyses where F influences parameters other than DNBR, e.g., peak clad temperature and thus is the maximum "as measured" value allowed.

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

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

The remaining margin is available for plant design flexibility.

When an F measurement is taken, an allowance for both experimental error g

and manufacturing tolerance must be made.

An allowance of 5% is appropriate for a full core map taken with the incore detector flux mapping system and a 3% allowance is appropriate for manufacturing talerance.

SU)91ER - UNIT 1 B 3h 2-3 Amendment No. 75

POWER DISTRIBUTION LIMIT BASES HEAT FLUX HOT CHANNEL FACTOR and RCS FLOWRATE and NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR (Continued)

The hot channel factor F (z) is measured periodically and increased by a cycle and height dependent power far.i r appropriate to either RAOC or Base load operation, W(z) or W(z)BL, to p avide assurance that the limit on the hot channel factor, F (z) is met. W(z) accounts for the effects of normal opera-q tion transients and was determined from expected power control maneuvers over the full range of *"i y onditions in the core.

W(z)gt accounts for the more F

restrictive operating li ots allowed by Base Load operation which result in less severe transient values The W(z) and W(z)BL functions described above for normal operation are provided in the Peaking Factor Limit Report per Specification 6.9.1.11.

hwrateandFharemeasured,noadditionalallowancesare When RCS necessary pricr to comparison with the limits of Figures 3.2-3.

Measurement N

errors of 2.1% for RCS tot 41 flow rate and 4% for F have been allowed for ON in determining the limits of Figure 3.2-3.

The 12-hour periodic surveillance of indicated RCS flow is sufficient to detect only flow degradation which could lead to operation outside the acceptable region of operation shown on Figure 3.2-3, 3/4.2.4 quaff \\NT POWER TILT RATIO The quadrant power tilt ratio limit assures that the radial power distribu-tion satisfies the design values used in the power capability analysis.

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

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

A limiting tilt of 1.025 can be to!erated before the margin for uncertainty in F is depleted.

The lirrit of 1.02 was selected to provide an allowince for q

the uncertainty asluciated with the indicated power tilt.

The two hour time allowance for operation with a tilt condition greater than 1.02 but lest-than 1.09 is provided to allow identification and correction of a dropped or misaligned control rod.

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

allowed power by 3 percent for each percent of tilt in excess of 1.0.

For purposes of monitoring QUADRANT POWER TILT RATIO when one excore detector is inoperable the movable incore detectors are used to confirm that the normalized symmetric power distribution is consistent with the QUADRANT POWER TILT RATIO.

The incore detector monitoring is done with a full incore l

flux map or two sets of 4 symmetric thimbles.

These locations are C-8, E-5, E-11, H-3, H-13, L-5, L-11, N-8.

l l

l SUMMER - UNIT 1 B 3/4 2-4 Amendment No. 4, 75

POWER DISTRIBUTION LIMIT BASES HEAT FLUX HOT CHANNEL FACTOR and RCS FLOWRATE and NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR (Continued) 3/4.2.5 DNB PARAMETERS The limits on the DN8 related parameters assure that each of the parameters are maintained within the normal steady state envelope of operation assumed in the transient and accident analyses.

The limits are consistent with the initial FSAR assumptions and have been analytically demonstrated adequate to maintain a minimum DNBR of 1.30 throughout each analyzed transient.

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

t

.~

l er l

SUMMER - UNIT 1 B 3/4 2-5 Amendment No. AB 55, f 0.

75 4

r-3/4.5 EMERGENCY CORE COOLING SYSTEMS BASES 3/4.5.1 ACCUMULATORS The OPERABILITY of each Reactor Coolant System (RCS) accumulator ensures that a sufficient volume of borated water will be immediately forced into the reactor core through each of the cold legs in the event the RCS pressure falls below the pressure of the accumulators.

This initial surge of water into the core provides the initial cooling mechanism during large RCS pipe ruptures.

In addition, the borated water serves to limit the maximum power which may be reached during large secondary pipe ruptures.

The limits on accumulator volume, boron concentration and pressure ensure that the assumptions used for accumulator injection in the safety analysis are met.

The accumulator power operated isolation valves are considered to be "operating bypasses" in the context of IEEE Std. 279-1971, which requires that bypasses of a protective function be removed automatically whenever permissive conditions are not met.

In addition, as these accumulator isolation valves fail to meet single failure criteria, removal of power to the valves is required.

The limits for operation with an accumulator inoperable for any reason l

except an isolation valve closed minimizes the time exposure of the plant to a l

LOCA event occurring concurrent with failure of an additional accumulator which may result in unacceptable peak cladding temperatures.

If a closed l

isolation valve cannot be immediately opened, the full capability of one accumulator is not available and prompt action is required to place the reactor in a mode where this capability is not required.

3/4.5.2 and 3/4.5.3 EMERGENCY CORE COOLING SYSTEM (ECCS) SUBSYSTEMS The OPERABILITY of two independent ECCS subsystems ensuros that sufficient emergancy cor<a cooling capability will be available in the event of a LOCA essu.ning the loss of one subsystem through any single failure consideration.

O ther subsystem operating in conjunction with the accumulators is capable of supplying sufficient core cooling to limit the peak cladding temperatures within acceptable limits for all postulated break sizes ranging from the double ended break.of the largest RCS cold leg pipe downward.

In addition, each ECCS subsystem provides long term core cooling capability in the recirculation mode during the accident recovery period.

With the RCS temperature below 350*F, one OPERABLE ECCS subsystem is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the limited core cooling requirements.

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ADMINISTRATIVE CONTROLS i

e.

Type of container (e.g., LSA, Type A, Type B, Large Quantity), and f.

Solidification agent (e.g., cement, urea formaldehyde).

The radioactive effluent release reports shall include unplanned releases from site to unrestricted areas of radioactive materials in gaseous and liquid effluents on a quarterly basis.

The radioactive effluent release reports shall include any changes to the Process Control Program (PCP) made during the reporting period.

MONTHLY OPERATING REPORT 6.9.1.10 Routine reports of operating statistics and shutdown experience, it4-cluding documentation of all challenges to the PORV's or safety valves, shall be submitted on a monthly basis to the Director, Office of Resource Management, U.S. Nuclear Regulatory Commission, Washington, D.C.

20555, with a copy to the Regional Office of Inspection and Enforcement, no later than the 15th of each month following the calendar month covered by the report.

Any changes to the OFFSITE DOSE CALCULATION MANUAL shall be submitted with the Monthly Operating Report within 90 days in which the change (s) was made et fective.

In addition, a report of any major changes to the radioactive waste treatment systems shall be submitted with the Monthly Operating Report for the period in which the evaluation was reviewed and accepted as set forth in 6.5 above.

RADIAL PEAKING FACTOR LIMIT REPORT 6.9.1.11 The AFD limits, the W(z) Functions for RA00 and Base Load operation i

and the value for APLND (as required) shall be established for each reload core and implemented prior to use.

The methodology used to generate the W(z) functions for RAOC and Base load ND Operation and the value for APL shall be those previously reviewed and approved by the NRC.*

If changes to these methods are deemed nocessary they will be evaluated in accordance with 10 CFR 50.59 and submitted to the NRC for 4

review and approv31 prior to their use if the change is determined to involve an unreviewed safety question or if *,uch a change would require amendment of previouslysubmitgddocumentation.

A report containing,the AFD limits, the W(z) functions for RAOC and Base Load operation and the value for APLND (as required) shall be provided to the NRC document control desk with copies to the regional administrator and the resident inspector within 30 days of their implementation.

ND AnyinformationneededtosupportW(z),W(z)$g and APL will be by request from the NRC and need not be included in thi report.

• WCAP-10216 P-A "Relaxation of Constant. Axial Offset Control-F Surveillance q

Technical Specification."

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

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

i l

SUPNER - UNIT 1 6-18a Amendment No.

75