Text

Proposed Tech Specs Supporting SG Replacement
	ML20059G290
Person / Time
Site:	Summer
Issue date:	10/29/1993
From:	; SOUTH CAROLINA ELECTRIC & GAS CO.
To:	;
Shared Package
ML20059G255	List: L-93-036, Licensing Submittal to Support Replacement SG TS Changes for Vsns; ML20059G290; RC-93-0272, Forwards SECL-93-036, Licensing Submittal to Support SG TS Changes for Vsns, Including Revised TS Pages & Proprietary Westinghouse TRs WCAP-13480 & WCAP-13605.Westinghouse TRs Withheld;
References
	NUDOCS 9311080057
	Download: ML20059G290 (35)
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- i Attachment to Document Control Desk Letter TSP 930015 ,

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4 STEAM GENERATOR REPLACEMENT i

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TECHNICAL SPECIFICATION REVISIONS .f i

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e N N 590 N' \ (120 % 590) 580 Acceptable Operaton T

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l (120 % 573) 570 l ' ' '

560

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0 20 40 60 80 100 120 i Power (Percent)

When operating in the reduced RTP region of Technical Specification 3.2.3 the restricted power level j must be considered 100% RTP for this figure. i O

Figure 2.1-1 Reactor Core Safety Limits - Three Loop Operation SMFIR - UNIT 1 2-2 Amendment No. 45, 75

O O O. '

n .

m TABLE 2.2-1 i

R_ FACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS U

Total

[ Functional Unit Allowance (TA) Z S Trip Setpoint Allowable Value

1. Manual Reactor Trip Not Applicable NA NA NA NA

2. Power Range, Neutron Flux High Setpoint 7.5 4.56 0 Low Setpoint $109% of RTP 5111.2% of RTP 8.3 4.56 0 $25% of RTP $27.2% of RTP l 3. Power Range, Heutron Flux 1.6 0.5 0 High Positive Rate 55% of RTP with a time 56 3% of RTP with a time constant 22 seconds constant 22 seconds

4. Deleted 7 5. Intermediate Range, 17.0 8.4 0 525% of RTP Neutron Flux $31% of RTP

6. Source Range, Heutron Flux 17.0 10.0 0 $105 cps $1.4 x 105 cps

7. Overtemperature AT 14.7 12.2 1.5 See note 1 See note 2

& 1.3**

8. Overpower AT 5.1 2.0 1.5 See note 3 See note 4

9. Pressurizer Pressure-Low 3.1 0.71 1.5 21870 psig 21859 psig

10. Pressurizer Pressure-High 6.9 5.0 0.9 52380 psig E $2391 psig k 11. Pressurizer Water Level-High 5.0 2.18 1.5 592% of instrument g 593.8% of instrument

s span span f

[

g?

12. Loss of Flow

' I 2.5 1,48 .6 290% of loop design 188.9% of loop design 4

flow * -

i e

. flow

CP

~

Loop design flow - 94,500 gpm u

P *** RTPs-vRATED THERMAltPOWER 9tw b 1.5% span for Delta-T (RIDS) and 1.3% for Pressurizer Pressure.

+t , ,,

, , , , ,, , # i h e. . . , , t , u 14...

o O O '

E i TABLE 2.2-1 (continued) 7J

, REACTOR TRIP SYSTEM INSTRUNENTATION TRIP SETPOINTS G

M

~

Total '

Functional Unit Allowance (TA) Z S Trip Setpoint Allowable Value

13. Steam Generator Water Level Low-Low Barton Transmitter 7.0 5.1 1.7 127.0% of span 126.1% of span Rosemount Transmitter 7.0 5.1 1.7 127.0% of span 225.7% of span

14. Steam /Feedwater riow Mis- 16.0 13.24 1.5/ <40% of full <42.5% of full Match Coincident With 1.5 steam flow at RTP steam flow at RTP m Steam Generator Water Level

& Low Barton Transmitter 7.0 5.1 1.7 127.0% of span 126.1% of span Rosemount Transmitter 7.0 5.1 1.7 127.0% of span 125.7% of span

15. Undervoltage - Reactor 2.1 1.28 0.23 14830 volts Coolant Pump 14760 l'

16. Underfrequency - Reactor 7.5 0 0.1 157.5 Hz Coolant Pumps 157.1 Hz

17. Turbine Trip A. Low Trip System Pressure NA NA NA 1800 psig 1750 psig B. Turbine Stop Valve NA NA NA gl% open 21% open Closure li RTP - RATED THERMAL POWER F

C-

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en O O O

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TABLE 2.2-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS

- NOTATION NOTE 1: OVERTEMPERATURE AT

~

(1 & s,S) aT s 6T, K, - K, T - T' t K, ( P - P') -fjan Where: AT = Measured AT by RTD Instrumentation AT, 5 Indicated aT at RATED THERMAL POWER K, 5 1.23 l

'f K, t 0.0292/*F l -- !

1 1 + t,S l

=

The funtion generated by the lead-lag controller I + '28 for T avg dynamic compensation 53,t2 =

Time constants utilized in lead-lag controller for Tavg

11.1 28 secs., i 1

2 5 4 secs..

T = Average temperature, "F T' 5 Indicated T avg at RATED THERMAL POWER, 572.0*F s T's 587.4*F l.

K 3 3 0.00161/ psi l P =

Pressurizer pressure. psig o y P' 3 .2235 psig, Noninal RCS operating pressure u ti g S = Laplace transform operator, sec1 PR ..

R

.g y

O O O E

TABLE 2.2-1 (Continued)_

i REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS NOTATION (Continued)

NOTE 1: (Continued) and f (al) is a function of the indicated difference between top and bottom detectors of the power-range nuclea,r ion chambers; with gains to be selected based on measured instrument response during plant startup tests such that:

(1) for gt - 9b between -35 percent and +6 percent f (al) = 0 where qt and qb are percent RATED THERMAL 3 l l

POWER in the top and bottom halves of the core respectively, and yt + 9 b is total THERMAL POWER in l percent of RATED THERMAL POWER.

l (ii) for each percent that the magnitude of qt - 4b exceeds -35 percent, the AT trip setpoint shall be l i $ automatically reduced by 2.46 percent of its value at RATED THERMAL POWER. l (iii) for each percent that the magnitude of qt - 9b exceeds 46 percent, the AT trip setpoint shall be i automatically reduced by 3.29 percent of its value at RATED THERMAL POWER. l NOTE 2: The channel's maximum trip setpoint shall not exceed its computed trip point by more than 2.2 percent AT Span.

NOTE 3: OVERPOWER AT (t 8I I a -

STsaT K-K T-K.i' T-T o 4 5 (1 + t3 S)

, Where: AT =

cs defined in Note 1 F ATo =

as defined in Note 1 3 :s K, 5 1.078 l

K, >

0.02/*F for increasing average temperature and 0 for decreasing average temperature

?

s, s M =

The function generated by the rate-lag controller for Tavg dynamic

(

w j+53 8 compensation >- o e M :. h e i 4 o pma ' e, i'..

.,4 4+

m El -

, TABLE 2.2-1 (Continued) g REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS NOTATION (Continued)

NOTE 3: (continued)

T3 =

Time constant utilized in rate-lag controller for Tavg* 53 2 10 secs.

K 3 2 0.00198/*F for T > T~ and K, = 0 for T $ T~ l T =

as defined in Note 1 y T~ $ Indicated T avg at RATED THERMAL POWER, 572.0*F s T's 587.4*F l E3 5 =

as defined in Note 1 NOTE 4: The channel's maximum trip setpoint shall not exceed its computed trip point by more than 2.3 percent AT Span. l U!I a

na n

;' r i

2.1 SAFETY LIMITS :

BASES 2.1.1 REACTOR CORE '

i The restrictions of this Safety Limit prevent overheating of the fuel and !

possible cladding perforation which would result in the release of fission t 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 ;

from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer i coefficient. DNB is not a directly measurable parameter during operation and i therefore THERMAL POWER and Reactor Coolant Temperature and Pressure have been related to DNB. This relation has been developed to predict the DNB flux and i the location of DNB for axially uniform and non-uniform heat flux distributions. 7 The local DNB heat flux ratio (DNBR) defined as the ratio of the heat flux that l would cause DNB at a particular core location to the local heat flux, is indicative of the margin to DNB. ;

The DNB design basis is as follows: there must be at least a 96 percent f i probability that the minimum DNBR of the limiting rod during Condition I and II i 7

events is greater than or equal to the DNBR limit of the ONB correlation being i used. The correlation DNBR limit is established based on the entire applicable l 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.

l 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 protability with 95 percent ,

confidence level that the minimum DNBR for the limiting rod is greater than or i equal to the DNBR limit. The uncertainties in the above plant parameters are used to determine the plant DNBR uncertainty. This DNBR uncertainty, combined :

with the correlation DNBR limit, establishes a design DNBR value which must be -

met in plant safety analyses using values of input parameters without uncertainties.

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

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

Reactor Coolant System pressure and average temperature below whicn the calculated

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

SUMMER - UNIT 1 B 2-1 Amendment No. 75 !

l

LIMITING SAFETY SYSTEM SETTINGS !

BASES REACTOR TRIP SYSTEP 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 reciundant channels and trains, the design approach provides a Reactor Protection System which monitors numerous system variables, therefore, providing protect'on system functional diversity. The Reactor Protection System initiates a turbine trip signal whenever reactor trip is initiated. This prevents the reactivity insertion that would otherwise result from excessive reactor system cooldown and thus avoids unnecessary actuation of the Engineered Safety Features Actuation System.

Manual Reactor Trio i

The Reactor Protection System includes manual reactor trip capability. i Power Range, Neutron Flux In each of the Power Range Neutron Flux channels there are two independent bistables, each with its own trip setting used for a high and low range trip setting. The low setpoint trip provides protection during subcriticc1 and low 1 power operations to mitigate the consequences of a power excursion beginning i from low power, and the high setpoint trip provides protection during power O operations to mitigate the consequences of a reactivity excursion from all power levels.

i The low setpoint trip may be manually blocked above P-10 (a power level of approximately 10 percent of RATED THERMAL POWER) and is automatically reinstated below the P-10 setpoint. '

Power Ranoe, Neutron Flux, Hich Rates The Power Range Pcsitive Rate trip provides protection against rapid flux increases which are characteristic of a rupture of a control rod drive housing.

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

4 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 l

. O l

t SUMMER - UNIT 1 B 2-4 Amendment No. 75 -

LIMITING SAFETY SYSTEM SETTINGS 4 i

BASES Pressurizer Pressure (Continued)

On decreasing power the low setpoint trip is automatically blocked by P-7 (a power level of approximately 10 percent of RATED THERV L POWER with turbine impulse chamber pressure at approximately 10 percent of full power equivalent);

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

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

Pressurizer Water Level The pressurizer high water level trip is provided to prevent water relief through the pressurizer safety valves. On decreasing power the pressurizer high water level trip is automatically blocked by P-7 (a power level of approximately 10 percent of RATED THERMAL POWER with a turbine impulse chamber pressure at approximately 10 percent of full equivalent); and on increasing power, automatically reinstated by P-7.

Loss of Flow The Loss of Flow trips provide core protection to prevent DNB by mitigating the consequences of a loss of flow resu.ing from the loss of one or more reactor coolant pumps.

On increasing power above P-7 (a power level of approximately 10 percent O of RATED THERMAL POWER or a turbine impulse chamber pressure at approximately 10 percent of full power equivalent), an automatic reactor trip will occur if the flow in more than one loop drops below 90% of nominal full loop flow.

Above P-8 (a power level of approximately 38 percent of RATED THERMAL POWER) an automatic reactor trip will occur if the flow in any single loop drops below 90 percent of nominal full loop flow. Conversely on decreasing power between P-8 and the P-7 an automatic reactor trip will occur on loss of flow in more than one loop and below P-7 the trip function is automatically blocked.

Steam Generator Water Level The steam generator water level low-low trip protects the reactor from loss of heat sink in the event of a sustained steam /feedwater flow mismatch resulting ? rom loss of normal feedwater. The specified setpoint provides allowances for starting delays of the auxiliary feedwater system.

Steam /Feedwater Flow Mismatch and Low Steam Generator Water Level The steam /feedwater flow mismatch in coincidence with a steam generator low water level trip is not used in the transient and accident analyses but is '

included in Table 2.2-1 to ensure the functional capability of the specified trip settings and thereby enhance the overall reliability of the Reactor Protection System. This trip is redundant to the Steam Generator Water Level Low-Low trip. The Steam /Feedwater Flow Mismatch portion of this trip is activated when the steam flow exceeds the feedwater flow by greater than or equal to 40% of full steam flow at RTP. The Steam Generator Low Water level portion of the trip is activated when the water level drops below the low l SUMMER - UNIT 1 B 2-6 Amendment No.

4 FIGURE 3.1-3 E REQUIRED SHUTDOWN MARGIN

, 8 (MODES 3,4, AND 5) x

' 4.5 - _ _

c-z _ ____ _ _ __ _ _ __ _ _ _ _-_ _ __ _ _ l px".

- 4- -

.---_Acceptoble-- - -

-Operation

,, A

_- _- g-j/ as 4

_i

_ g

% J.,5- _

d_/_ _.

c f._y _ ._ _ _

m 3- - - - - - - - -

A-/_ --. . _y___.

/-_Z J, _ _ _ _- - ___

- p_ _

__ c.- y / ms c 2.5 - -

/

/. _ _-

,-z-

-- _ -- =~ __ _ _ _ _ .. _

w n _ ___ _ _ . . .

'5 2- _ --

_--- _- _- - f,

_y

/- - -

._ -. i Si _ _ _

}3,5 f

_y-- _ __ _ _ .

s _

f- _ _ _ _

er _

e as s ,

og }

Unocceploble

_ _ - _ _ _ _ _. . . =

O ro - - __ _

g __ _ __ _ _ _ _ _ _ _ __ ._. _.

a

<=

.__ Operoh.on . - .

g 0.5 - _ _._ _- _- - -

g o .

.o,. O ~ e i i 0 500 1000 1500 2000 250

.a 1- i l: -. 2 RCS Boron Concentmtion (ppm) he..t he,n ni , o." p " 9 9 9 e n 9 p o u

. ? 4 ( ,. .., s, 7

4 6 s

u-_ . _ . _ _ . . . _ _ _ _ _.__m.s.-._s__ m_m ._, ..-e ........r o.m_-e- -~.-._d.e-_-~~.-.w-+-+4.,#.% e,vew_e.-.-...e.~,. .--e ... .- -.. w_ .....uw. . .&w...,.u--. .....>.....-_....$.4...m-...

POWER DISTRIBUTION LIMITS 3/4.2.3 RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR

[ LIMITING CONDITION FOR OPERATION 3.2.3 The combination of indicated Reactor Coolant System (RCS) total flow rate and R shall be maintained within the region of allowable operation as specifiedintheCOREOPERATINGLIMITSREPORT(COLR)figureentitledRCS Tctal Flow Rate Versus R For Three Loop Operation. '

Where:

a. R = FH ,

RTP FaH [1.0 + PFag (1.0 - P)]

b. P =

THERMAL POWER ,

RATED THERMAL POWER N N

c. FtH =

Measured values of FcH obtained by using the movable incore detectors to obtain a power distribution map. The measured N

] values of FaH shall be used to calculate R since the RCS Total Flow Rate Versus R figure in the COLR includes m2asurement uncertainties of 2.1% (includes 0.1% for feedwater venturi N -

fouling) for flow and 4% for incore measurement of FaH. and- '

RTP y

d. FaH =

The FcH limit at RATED THERMAL POWER specified in the COLR. ;

e. PFsH =

The Power Factor Multiplier specified in the COLR. ,

QPIICABILITY: MODE 1.

ACTION:

With the combir:ation of RCS total flow rate and R outside the region of accept-able operation specified in the COLR:

a. Within 2 P s 4ther:

1. Res-the e

+ tN combination of RCS total flow rate and R to within

. limits, or 2.

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

b. Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of initially being outside the above limits, verify through incore flux mapping and RCS total flow rate comparison that

(- the combination of R and RCS total flow rate are restored to within the above limits, 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 . I l

SUMMER - UNIT 1 3/4 2-8 Amendment No. 45,-60,'75,'

63 ,

POWER DISTRIBUTION LIMITS

("%

LIMITING CONDITION FOR OPERATION ACTION: (Continued)

c. Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER above the reduced THERMAL POWER limit required by ACTION items a.2. and/or b. above; subsequent POWER OPERATION may proceed provided that the combination of R and indicated RCS total flow rate are demonstrated, through incore flux mapping and RCS total flow rate comparison, to be within the region -

of acceptable operation specified in the COLR prior to exceeding the following THERMAL POWER levels:

1. A nominal 50% of RATED THERMAL POWER,

2. A nominal 75% of RATED fHERMAL POWER, 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 TiiERMAL POWER.

~

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

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 specified in the COLR.

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

b. At least once per 31 Effective Full Power Days.

4.2.3.3 The indicated RCS total flow rate shall be verified to be within the region of acceptable operation specified in the COLR at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when 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.

4.2.3.5 The RCS total flow rate shall be determined by heat balance measurement at-

>90% RATED THERMAL POWER at least once per 18 months.

O l

SUMMER - UNIT 1 3/4 2-9 Amendment No. 45, 75, 88 s j l

m O O O g TABLE 3.2-1

[

DNB PARAMETERS e

LlHITS 3 Loops In 2 Loops In PARAMETER Operation Operation Indicated Reactor Coolant System Tavg s 589.2*F **

l Indicated Pressurizer Pressure 2 2206 psig* **

k' s.

7 E

l li 'I g .

q

. E i

a-

n. ,

l 5 a

u h Limit not applicable during either a THERMAL POWER ramp in excess of 5% of RATED THERMAL POWER per_ minute i i

or atTHERMALtPOWERistepiin excess of 10% of RATED:THEftMAL+

t POWER., i e .a'6;b.g .ytv 4 .e a .

b.-

, . t. . . .a - , , , -

p ,

6

.g

- These values left blank pending NRC approva' of two-loop' operation. '

H y,ou oi..,ttt=4wt u io!q m ;m uae is t . en .a.,p 4. e o ,

_.._______.__.___,__..._.._.__.___._~-....__.__.._.~..._...__.;_...

- - .._ _-_ .-. - ;. .-_-2.w. _ . _ _._ - . - . _ _ _ _ . . . . . _

POWER DISTRIBUTION LIMIT BASES

O HEAT FLUX HOT CHANNEL FACTOR and RCS FLOWRATE and NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR (Continued) 1 1 The hot channel factor FM(z) is measured periodically and increased by a Q

cycle and height dependent power factor appropriate to either RAOC or Base ;

Load operation, W(z) or W(z)BL, to provide assurance that the limit on the hot '

channel factor, FQ(z) is met. W(z) accounts for the effects of normal opera- ,

tion transients and was determined from expected power control maneuvers over the full range of burnup conditions in the core. W(z)BL accounts for the more restrictive operating limits allowed by Base Load operation which result in less severe transient values. The W(z) and W(z)BL functions described above for i

normal operation are specified in the CORE OPERATING LIMITS REPORT (COLR) per Specification 6.9.1.11.

When RCS flow rate and FN are measured, no additional allowances are LH necessary prior to comparison with the limits of the RCS Total Flow Rate I Versus R figure in the COLR. Measurement errcrs of 2.1% for RCS total flow rate including 0.1% for feedwater venturi fouling and 4% for

, FN have been allowed for in determining the limits of the RCS Total Flow ['

LH Rate Versus R figure in the COLR. t 4

The 12-hour periodic surveillance of indicated RCS flow is sufficient to - !

detect only flow degradation which could lead to operation outside the accept-d

.O able region of operation specified on the RCS Total Flow Rate Versus R figure in the COLR.

3/4.2.4 QUADRANT POWER TILT RATIO i .

' The quadrant power tilt ratio limit assures that the radial power distribu- !

tion satisfies the design values used in the power capability analysis. Radial l power distribution measurements are made during startup testing and periodically during power operation.

i The limit of 1.02, at which corrective action is required, provides DNB '

i and linear heat generation rate protection with x-y plane power tilts. A limiting tilt of 1.025 can be tolerated before the margin for uncertainty in - '

F Q is depleted. The limit of 1.02 was selected to provide an allowance for ~

[

the uncertainty associated with the indicated power tilt. ,

' The two hour time allowance for operation with a tilt condition greater t than 1.02 but less than 1.09 is provided to allow identification and correction ~ !

of a dropped or misaligned control rod. In the event such action does not correct -

l the tilt, the margin for uncertainty on FQ is reinstated by reducing the maximum; 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 l POWER TILT RATIO. The incore detector monitoring is done with a full incore i flux map or two sets of 4 symmetric thimbles. These locations are C-8, E-5, l E-11, H-3, H-13, L-5, L-11 N-8. '

s SUMMER - UNIT 1 B 3/4 2-4 Amendment No. 45, 75, 88

1

i i

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 DNB related parameters assure that each of the parameters !

are maintained within the normal steady state envelope of operation assumed in ;

the transient and accident analyses. The limits are consistent with the initial '

FSAR assumptions and have been analytically demonstrated adequate to maintain ,

a minimum DNBR in the core at or above the design limit throughout each analyzed transient. The maximum indicated Tavg limit of 589.2*F and the minimum.

indicated pressure limit of 2206 psig correspond to analytical limits of 591.4*F and 2185 psig respectively, read from control board indications.

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.

i t

i i

t i

SUMMER - UNIT 1 B 3/4 2-5 Amendment No. 45 -56, 60,- !

75, 88 Correction

V s' w

U

@ TABLE 3.3-1 i

REACTOR TRIP SYSTEM INSTRUMEliTATION d

M

,. MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION

1. Manual Reactor Trip 2 1 2 1, 2 1 2 1 2 3*, 4*, 5* 9

2. Power Range, Neutron Flux A. High Setpoint 4 2 3 1 2 2#

B. Low Setpoint 4 2 3 IIII,2 2I

3. Power Range, Neutron Flux 4 2 3 1, 2 2I g High Positive Rate y 4. Deleted l I-J

5. Intermediate Range, Neutron Flux 2 1 2 1#II, 2 3

6. Source Range Neutron Flux A. Startup 2 1 2 2EI 4 B. Shutdown 2 0 1 3, 4 and 5 5 C. Shutdown 2 1 2 3*,4*,5* 9

7. Overtemperature AT V: Three Loop Operation 3 2 2 1, 2 6I Two Loop Operation **** **** **** **** ****

]

?

g 8. Overpower AT

[ Three Loop Operation 3 2 2 1, 2 6I o Two Loop Operation **** **** **** **** ****

f 9. ,

Pressurizer Pressure-Low 4 ,,f *,,, is o 4 0 . i . ~,

3 2

2 1 6f .

10. Pressurizer Pressure--High 3 2 2 1, 2 6I io.

> . m ais.d b nuta h all.e4 8

r 4 e

O O O s

% TABLE 3.3-2 x

i REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES FUNCTIONAL UNIT RESPONSE TIME

1. Manual Reactor Trip Not Applicable

2. Power Range, Neutron Flux s 0.5 seconds (l)

3. Power Range, Neutron Flux, Not Applicable High Positive Rate

4. Deleted l

S. Intermediate Range, Neutron Flux Not Applicable

6. Source Range, Neutron Flux Not Applicable

7. Overtemperature AT s 8.5 seconds (1)(2)

8. Overpower AT s 8.5 seconds (1)(2)

9. Pressurizer Pressure--Low s 2.0 seconds

10. Pressurizer Pressure--High s 2.0 seconds

11. Pressurizer Water Level--High Not Applicable E

B a

E F.

y (1)#1Neutron detectors are exempt from response time testing. Response time of the neutron flux signal portion of the channel shallebe measured from detector output or input of first electronic component,in channel, g (2) The 8.5 second response tjme,iincludes a 5.0 second delay for the RTDs mounted in thermowells.

.et s s ,.e . . ,

t .

O O O e

j TABLE 4.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS E

tj TRIP ANALOG ACTUATING MODES FOR CHANNEL DEVICE WHICH CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST IS REQUIRED

1. Manual Reactor Trip N.A. N.A. N.A. R(11) N.A. 1, 2, 3 * , 4 * , 5 *

2. Power Range, Neutron Flux High Setpoint S D 2, 4), Q N.A. N.A. 1, 2 M 3, 4),

Q 4, 6),

R R 4, 5)

[, Low Setpoint S R(4) S/U(1) N.A. N.A. 1###, 2 C 3. Power Range, Neutron Flux N.A. R(4) Q N.A. N.A. 1, 2 High Positive Rate

4. Deleted

5. Intermediate Range, S R(4) S/U(1), N.A. N.A. l#88, 2 Neutron Flux

6. Source Range, Neutron Flux S R(4) S/U(1),Q(9) H.A. N.A. 2ff, 3, 4, 5

7. Overtemperature aT S R Q N.A. N.A. 1, 2

8. Overpower aT S R Q N.A. N.A. 1, 2

9. Pressurizer Pressure--Low S R Q N.A. N.A. 1 E 10. Pressurizer Pressure--High S R Q N.A. N.A. 1, 2

11. Pressurizer Water Level--High S R Q H.A. N.A. 1

-- ... .. m ... o n ,, ,, u ... ,,,,,, . - .. > n S ." 12. Loss of Flow S R Q N.A. N.A. 1

,q p U e , r* l ioq , ,. , , 1 6..o ,t i ti t ,

O O O

~

TABLE 3.3-4 (Continuedl E

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS h Total

] Functional Unit Allowance (TA) Z S Trip Setpoint Allowable Value

5. TURBINE TRIP AND FEEDWATER ISOLATION

a. Steam Generator Water Level - High-High Barton Transmitter 20.8 11.4 1.7 579.2% of span 181.0% of span Rosemount Transmitter 20.8 12.4 1.7 579.2% of span 181.0% of span

6. EMERGENCY FEEDWATER y a. Manual NA NA NA NA NA

b. Automatic Actuation Logic NA NA NA NA NA

c. Steam Generator Water Level - Low-Low Barton Transmitter 7.0 5.1 1.7 127.0% of span 126.1% of span Rosemount Transmitter 7.0 5.1 1.7 127.0% of span 125.7% of span

d. & f. Undervoltage-ESF Bus 15760 Volts with 35652 Volts with a a 50.25 second 50.275 second time time delay delay 16576 volts with 36511 Volts with a a <3.0 second <3.3 second time g time delay delay a [

n.

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E-_.m.- _ - _ - _ _ _ _ _ ______,_m_ . - - . , , , . . - , - - , . . ~ - , , , , - - . . . . , m. * . - ,~-_m., , e. - , .. _ . _ _ , _.., . . , - . , , .

A REACTOR COOLANT SYSTEM 3/4.4.5 STEAM GENERATORS LIMITING CONDITION FOR OPERATION 3.4.5 Each steam generator shall be OPERABLE. !

APPLICABILITY: MODES 1, 2, 3 and 4 ACTION:

With one or more steam generators inoperable, restore the inoperable generator (s) to OPERABLE status prior to increasing Tavg above 200*F.

SURVEILLANCE REOUIREMENTS 4.4.5.0 Each steam generator shall be demonstrated OPERABLE by performance of the following augmented inservice inspection program and the requirements of Specification 4.0.5.

4.4.5.1 Steam Generator Sample Selection and Inspection - Each steam generator '

shall be determined OPERABLE during shutdown by selecting and inspecting at least the minimum number of steam generators specified in Table 4.4-1.

O 4.4.5.2 Steam Generator Tube Sample Selection and Inspection - The steam generator tube minimum sample size, inspection result classification, and ther corresponding action required shall be as specified in Table 4.4-2. The inservice inspection of steam generator tubes shall be performed at the fre- ,

quencies specified in Specification 4.4.5.3 and the inspected tubes shall be verified acceptable per the acceptance criteria of Specification 4.4.5.4. The tubes selected for each inservice inspection shall include at least 3% of the total number of tubes in all steam generators. The tubes selected for these inspections shall be selected on a random basis except: [>

a. Where experience in similar plants with similar water chemistry indicates critical areas to be inspected, then at least 50% of the tubes inspected shall be from these critical areas.

b. The first sample of tubes selected for each inservice inspection - i (subsequent to the preservice inspection) of each steam generator shall include. ,

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N O l SUMMER - UNIT 1 3/4 4-11 Amendment No. 59 i

I l

)

REACTOR COOLANT SYSTEM SURVEILLANCE REOUIREMENTS (Continued)

1. All nonplugged tubes that previously had detectable wall penetrations greater than 20%. l,

2. Tubes in those areas where experience has indicated potential problems.

3. A tube inspection (pursuant to Specification 4.4.5.4.a.8) shall be performed on each selected tube. If any selected tube does not permit the passage of the eddy current probe for a tube inspection, this shall be recorded and an adjacent tube shall be selected and subjected to a tube inspection.

c. The tubes selected as the second and third samples (if required by l' Table 4.4-2) during each inservice inspection may be subjected to a partial tube inspection provided:

1. The tubes selected for these samples include the tubes from

t. hose areas of the tube sheet array where tubes with imperfections were previously found.

2. The inspections include those portions of the tubes where imperfections were previously found. ^

The results of each sample inspection shall be classified into one of the following three categories:

Catecorv Inspection Results C-1 tess than 5% of the total tubes inspected are degraded tubes and none of the inspected tubes '

are defective. .

i C-2 One or more tubes, but not more than 1% of the total tubes inspected are defective, or between 5% and 10% of the total tubes inspected are degraded tubes.

C-3 More than 10% of the total tubes inspected are degraded tubes or more than 1% of the inspected ;

tubes are defective.

Note: In all inspections, previously degraded tubes must exhibit ;

significant (greater than 10%) further wall penetrations to be included in the above percentage calculations. i O

SUMMER - UNIT 1 'l/4 4-12 Amendment No. 54, 59, 96 i

REACTOR COOLANT SYSTEM SURVEILLANCE REOUIREMENTS (Continued) !

4.4.5.3 Inspection Frecuencies - The above required inservice inspections of -

steam generator tubes shall be performed at the following frequencies:

a. The first inservice inspection after the steam generator replacement shall be performed after at least 6 Effective Full Power Months from the time of the replacement but within 24 calendar months of initial !

criticality after the steam generator replacement. Subsequent inservice -

inspections shall be performed at intervals of not less than 12 nor

'i more than 24 calendar months after the previous inspection. If two consecutive inspections following service under AVT conditions, not including the preservice inspection, result in all inspection results falling into the C-1 category or if two consecutive inspections demonstrate that previously observed degradation has not continued and no additional degradation has occurred, the inspection interval may be extended to a maximum of once per 40 months.

b. If the results of the inservice inspection of a steam generator

- conducted in accordance with Table 4.4-2 at 40 month intervals fall in Category C-3, the inspection frequency shall be increased to at least once per 20 months. The increase in inspection frequency shall apply until the subsequent inspections satisfy the criteria of -

i Specification 4.4.5.3.a; the interval may then be extended to a maximum of once per 40 months.

c. Additional, unscheduled inservice inspections shall be performed on ;

each steam generator in accordance with the first sample inspection !

specified in Table 4.4-2 during the shutdown subsequent to any of i

the following conditions. ;

, 1. Primary-to-secondary tube leaks (not including leaks i originating from tube-to-tube sheet welds) in excess of the limits of Specification 3.4.6.2.

t

2. A seismic occurrence greater than the (perating Basis Earthquake. - !

3. A loss-of-coolant accident requiring actuation of the engineered safeguards. #

4. A main steam line or feedwater line break. ep

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SUMMER - UNIT 1 3/4 4-13 Amendment No.

I i

REACTOR COOLANT SYSTEM SURVEILLANCE REOUIREMENTS (Continued) 4.4.5.4 Acceptance Criteria

a. As used in this Specification: '

1. Imperfection means an exception to the dimensions, finish or -

contour of a tube from that required by fabrication drawings or !

specifications. Eddy-current testing indications below 20% of the nominal tube wall thickness, if detectable, may be considered as imperfections.

2. Degradation means a service-induced cracking, wastage, wear or - '

general corrosion occurring on either inside or outside of a tube.

3. Degraded Tube means a tube containing imperfections greater than or equal to 20% of the nominal wall thickness caused by degradation.

4. % Degradation means the percentage of the tube wall thickness '

affected or removed by degradation.

5. Defect means an imperfection of such severity that it exceeds the plugging limit. A tube containing a defect is defective.

O 6. Tube Pluggina limit means the imperfection depth at or beyond l

which the tube shall be removed from service by plugging and is equal to 40% of the nominal tube wall thickness.

7. Unserviceable describes the condition of a tube if it leaks or [

contains a defect large enough to affect its structural integrity in the event of an Operating Basis Earthquake, a loss-of-coolant accident, or a steam line or feedwater line break as specified in 4.4.5.3.c, above.

8. Tube Inspection means an inspection of the steam generator tube [ 1 from the point of entry (hot leg side) completely around the U-bend to the top support of the cold leg.

C

c O

SUMMER - UNIT 1 3/4 4-14 Amendment No. 54, 59, 91, 96 !

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SURMER - UNIT 1 3/4 4-14a Amendment No. 59, 91 -

';s

l REACTOR COOLANT SYSTEM i

SURVEILLANCE RE0VIREMENTS (Continued)

9. Preservice Inspection means an inspection of the full length of {j each tube in each steam generator performed by eddy current techniques prior to service to establish a baseline condition ]

of the tubing. This inspection shall be performed after the manuf acturer's field hydrostatic test and prior to initial POWER [i OPERATION using the equipment and techniques expected to be used during subsequent inservice inspections.

b. I The steam generator shall be determined OPERABLE after completing thecorrespondingactions(plugalltubesexceedingtheplugginglimit) required by Table 4.4-2. [,

4.4.5.5 Reports

a. Within 15 days following the completion of each inservice inspection of steam generator tubes, the number of tubes plugged in each steam generator shall be reported to the Commission in a Special l' Report pursuant to Specification 6.9.2.

b. The complete results of the steam generator tube inservice inspection shall be submitted to the Commission in a Special Report pursuant to Specification 6.9.2 within 12 months following the completion of the inspection. This Special Report shall include:

1. Number and extent of tubes inspected. .

2. Location and percent of wall-thickness penetration for each indication of an imperfection.

3. Identification of tubes plugged.

l<

c. Results of steam generator tube inspections which fall into Category C-3 and require prompt notification of the Commission shall be reported pursuant to 10 CFR 50.72(b)2(1) prior to resumption of plant operation. A report pursuant to 10 CFR 50.73(a)2(ii) shall be submitted to provide a description of investigations conducted to deteniine cause of the tube degradation and corrective measures taken ;

to prevent recurrence.

O SUMMER - UNIT 1 3/4 4-15 Amendment No. 35, 54, 59, 87, 91, 96

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.-i; SUMMER - UNIT 1 3/4 4-15a Amendment No.-96 i. >

REACTOR COOLANT SYSTEM BASES 3/4.4.5 STEAM GENERATORS The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS will be main-tained. The program for inservice inspection of steam generator tubes is based on a modification of Regulatory Guide 1.83, Revision 1. Inservice inspection of steam generator tubing is essential in order to maintain surveillance of the conditions of the tubes in the event that there is evidence of mechanical damage or progressive degradation due to design, manufacturing errors, or inservice conditions that lead to corrosion. Inservice inspection of steam generator tubing also provides a means of characterizing the nature and cause of any tube degradation so that corrective measures can be taken.

The plant is expected to be operated in a manner such that the secondary coolant will be maintained within those chemistry limits found to result in negligible corrosion of the steam generator tubes. If the secondary coolant chemistry is not maintained within these limits, localized corrosion may likely result in stress corrosion cracking. The extent of cracking during plant operation would be limited by the limitation of steam generator tube leakage between the primary coolant system and the secondary coolant system (primary-to-secondary leakage = 500 gallons per day per steam generator).

Cracks having a primary-to-secondary leakage less than this limit during operation will have an adequate margin of safety to withstand the loads imposed during normal operation and by postulated accidents. Operating plants have demonstrated that primary-to-secondary leakage of 500 gallons per day per -

steam generator can readily be detected by radiation monitors of steam generator blowdown. Leakage in excess of this limit will require plant shutdown and an unscheduled inspection, during which the leaking tubes will be located and plugged.

[

Wastage-type defects are unlikely with proper chemistry treatment of the secondary coolant. However, even if a defect should develop in service, it -

will be found during scheduled inservice steam generator tube examinations.

Plugging will be required for all tubes with imperfections exceeding 40% of the- [

tube nominal wall thickness. Steam generator tube inspections of operating plants have demonstrated the capability to reliably detect wastage-type -

degradation that has penetrated 20% of the original tube wall thickness. .,

Whenever the results of any steam generator tubing inservice inspection f all into Category C-3, these results will be promptly reported to the Comission -

pursuant to 10 CFR 50.72(b)2(i) prior to resumption of plant operation. Such cases will be considered by the Commission on a case-by-case basis and may result in a requirement for analysis, laboratory examinations, tests, additional eddy-current inspection, and revision of the Technical Specifications, if necessary.

1 1

SUMMER - UNIT 1 B 3/4 4-3 Amendment No 35, 54, i 59, 96 !

l l

DESIGN FEATURES 5.3 REACTOR CORE i

FUEL ASSEMBLIES 5.3.1 The reactor ccre shall contain 157 fuel assemblies with each fuel i assembly normally containing 264 fuel rods with Zircaloy-4 or ZIRLO alloy clad- -

ding, except that limited substitution of fuel rods by filler rods consisting of Zircaloy-4 , ZIRLO alloy, stainless steel, or by vacancies, may be made if - ,

justified by a cycle specific reload analysis. Each fuel rod shall have a nominal active fuel length of 144 inches. The initial core loading shall have a maximum enrichment of 3.2 weight percent U-235. Reload fuel shall be similar in physical design to the initial core loading and shall have a maximum enrich-ment of 4.25 weight percent U-235. -

CONTROL ROD ASSEMBLIES 5.3.2 The reactor core shall contain 48 full length control rod assemblies. .

The full length control rod assemblies shall contain a nominal 142 inches of '

absorber material. The nominal values of absorber material shall be 80 percent silver, 15 percent indium and 5 percent cadmium. All control rods shall be clad with stainless steel tubing. .

, 5.4 REACTOR COOLANT SYSTEM DESIGN PRESSURE AND TEMPERATURE 5.4.1 The reactor coolant system is designed and shall be maintained: !

a. In accordance with the code requirements specified in Section 5.2 ;

of the FSAR, with allowance for normal degradation pursuant to the J

applicable Surveillance Requirements,

b. For a pressure of 2485 psig, and

c. For a temperature of 650*F, except for the pressurizer which is

, 680*F.

VOLUME 3

5.4.2 The total water and steam volune of the reactor coolant system is 9914 100 cubic feet at an indicated Tavg of 587.4"F. "l

, 5.5 METEOROLOGICAL TOWER LOCATION '

5.5.1 The meteorological tower shall be located as shown on Figure 5.1-1.

5'~

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O SUMMER - UNIT 1 5-6 Amendment No. 27, 55,-62,-

74, 105 i

4 4

i 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT CONTAINMENT INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.1.1 Primary CONTAINMENT INTEGRITY shall be maintained. !

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION: '

Without primary CONTAINMENT INTEGRITY, restore CONTAINMENT INTEGRITY within one hour or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . ,

SURVEILLANCE REOUIREMENTS 4.6.1.1 Primary CONTAINMENT INTEGRITY shall be demonstrated:

i

a. At least once per 31 days by verifying that all penetrations

not capable of being closed by OPERABLE containment automatic isolation valves and required to be closed during accident conditions are ,

Od closed by valves, blind flanges, or deactivated automatic valves secured in their positions, except for valves that are open under administrative control as permitted by Specification 3.6.4.

i

b. By verifying that each containment air lock is in compliance with the requirements of Specification 3.6.1.3.

c. After each closing of each penetration subject to Type B testing. I except the containment air locks, if opened following a Type A or B

test, by leak rate testing the seal with gas at Pa (53.5 psig) and [

verifying that when the measured leakage rate for these seals is added to the leakage rates detemined pursuant to Specification 4.6.1.2.d for all other Type B and C penetrations, the combined i leakage rate is less than 0.60 La-l 1

s !

Except valves, blind flanges, and deactivated automatic valves which are located inside the containment and are locked, sealed or otherwise secured in the closed position. These penetrations shall be verified closed during each COLD SHUTDOWN except that such verification need not be performed more often than once per 92 days. !

O SUMMER - UNIT 1 3/4 6-1 Amendment No. 110.

7[ ~

l CONTAINMENT SYSTEMS CONTAINMENT LEAKAGE

~

LIMITING CONDITION FOR OPERATION 3.6.1.2. Containment leakage rates shall be limited to:

a. An overall integrated leakage rate ef: -k Q

[

1. Less than or equal to La, 0.20 percent by weight of the contain-- ? . ,

ment air per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months at Pa, 53.5 psig, or AR?l Ai

2. Less than or equal to Lt . 0.10 percent by weight of the contain g7 ,

ment air per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months at a reduced pressure of P t . 26.8 psig.1.

b. Acombinedleakagerateoflessthan0.60Laforallpenetr valves subject to Type B and C tests, when pressurized to Pa.

/

APPLICABILITY: MODES 1, 2, 3 and 4. I ACTION: ;

7 ;

exceeding 0.75 La or 0.75 L t, as applicable, or (b) with the measured combined.--'Q.,

With either (a) the measured overall integrated containment leakage rate i J

leakage rate for all penetrations and valves subject to Types B and C tests -

exceeding 0.60 La, restore the overall integrated leakage rate to less than or rjQ.(. a l

~

equal to 0.75 La or less than o combined leakage rate for all.penetrations r equal to 0.75 t L . as subject to applicable, Type B and and the to C tests (@%f

.yf p.

less than 0.60 La prior to increasing the Reactor Coolant System temperature - ~w above 200*F. '

i SURVEILLANCE REOUIREMENTS 4.6.1.2 The containment test schedule leakage rates and shall be determined shall be demonstrated in conformance at the with the criteria speci- following i *.'

fied in Appendix J of 10 CFR 50 using the methods and provisions of ANSI N45.4-1972: [M ;

]M

.:cney ;

a. Three Type A tests (Overall Integrated Containment Leakage Rate) shall be conducted at 40 1 10 month intervals

during shutdown at r fg- ";;

either Pa (53.5 psig) or at Pt (26.8 psig) during each 10-year -

l service period. The third test of each set shall be conducted .;

during the shutdown for the 10-year plant inservice inspection. .,

j

.. ,. . . [ '

ag i

A one time extension of the test interval is allowed for the third Type A test r within the first 10-year service period, provided unit shutdown occurs no - -

later than June 1, 1993 and performance of the Type A test occurs prior to = ,

O unit restart following RF7. ],$g,%

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SUlHER - UNIT 1 3/4 6-2 Amendment No.-97 E -

n. e 04;

-y

I CONTAINMENT SYSTEMS 4

SURVEILLANCE REOUIREMENTS (continued)

b. If any periodic Type A test fails to meet either 0.75 La or 0.75 Lt. '

the test schedule for subsequent Type A tests shall be reviewed and approved by the Comission. If two consecutive Type A tests fail to meet either 0.75 La or 0.75 Lt. a Type A test shall be performed at :

least every 18 months until two consecutive Type A tests meet either '

O.75 La or 0.75 Lt at which time the above test schedule may be resumed. ,

c. The accuracy of each Type A test shall be verified by a supplemental test which:

1. Confirms the accuracy of the Type A test by verifying that the !

difference between supplemental and Type A test data is within !

0.25 La, or 0.25 L t. i e

i

, 2. Has a duration sufficient to establish accurately the change in j

leakage rate between the Type A test and the supplemental test.

3. Requires the quantity of gas injected into the containment or ,

bled from the containment during the supplemental test to be i equivalent to at least 25 percent of the total measured leakage '

!O d.

at Pa (53.5 psig) or Pt (26.8 psig).

Type B and C tests shall be conducted with gas at Pa (53.5 psig) at l

l l 1

intervals no greater than 24 months except for tests involving: '

1. Air locks.

I i

2. Purge supply and exhaust isolation valves with resilient k

j [

material seals. '

e. Purge supply and exhaust isolation valves with resilient material [

seals shall be tested and demonstrated OPERABLE per Surveillance Requirement 4.6.1.7.3.

f.

Air locks shall be tested and demonstrated OPERABLE per Surveillance !

Requirement 4.6.1.3. '

t

g. The provisions of Specification 4.C.2 are not applicable.

c h

!O ;

i SUMMER - UNIT 1 3/4 6-3 Amendment No.

1

CONTAINMENT SYSTEMS '

CONTAINMENT AIR LOCKS i

LIMITING CONDITION FOR OPERATION 3.6.1.3 Each reactor building air lock shall be OPERABLE with:

a. Both doors closed except when the air lock is being used for normal transit entry and exit through the containment, then at least one air lock door shall be closed, and

b. An overall air lock leakage rate of less than or equal to 0.10 Laa t Pa, 53.5 psig.

[

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

a. With one reactor building air lock door inoperable:

1.

Maintain at least the OPERABLE air lock door closed and either restore the inoperable air lock door to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or lock the OPERABLE air lock door closed.

2. Operation may then continue until performance of the next

'. O required overall air lock leakage test provided that the OPERABLE air lock door is verified to be 'ocked closed at least once per 31 days.

3. Otherwise, be in at least HOT STANDBY within the next six hours and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

4. The provisions of Specification 3.0.4 are not applicable.

b. With the reactor building air lock inoperable, except as the result ;

of an inoperable air lock door, maintain at least one air lock door-closed; restore the inoperable air lock to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least HOT STANDBY within the next s1x hours and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

O '

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

I

w v

CONTAINMENT SYSTEMS l

SURVEILLANCE REOUIREMENTS 4.6.1.3 Each reactor building air lock shall be demonstrated OPERABLE:

a. Within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months following each closing, except when the air lock is '

being used for multiple entries, then at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , by verifying that the seal leakage rate is less than or equal to 0.01 La when the volume between the door seals is pressurized to greater than or equal to 8.0 psig for at least 3 minutes. i

b. By conducting overall air lock leakage tests at not less than Pa.

53.5 psig, and verifying the overall air lock leakage rate is within its limit: l 1

1. At least once per 6 monthsf, and i

2. Prior to establishing CONTAINMENT INTEGRITY when maintenance has been performed on the air lock that could affect the air lock sealing capability.*

c. At least once per six months by verifying that only one door in each air lock can be opened at a time.

d. At least orce per 6 monthsi, by verifying that the seal leakage rate l is less than or equal to 0.01 La when the volume between the handwheel i shaft seals is pressurized to greater than or equal to 8.0 psig for at least 3 minutes. ;

.i .

^

1 1

4 t

t fThe provisions of Specification 4.0.2 are not applicable.

Exemption to Appendix J of 10 CFR 50.

i ;

o ;

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

i

CONTAINMENT SYSTEMS '

O BASES 3/4.6.1.4 INTERNAL PRESSURE The limitations on reactor building internal pressure ensure that 1) the reactor building structure is prevented from exceeding its design negative pressure differential with respect to the outside atmosphere of 3.5 psig and

2) the reactor building peak pressure does not exceed the design pressure of 57 psig during steam line break conditions. ,

The maximum peak pressure expected to be obtained from a steam line break event is 53.5 psig. The limit of 1.5 psig for initial positive containment pressure will limit the total pressure to 53.5 psig which is less than design pressure and is consistent with the accident analyses. ,

3/4.6.1.5 AIR TEMPERATURE The limitations on reactor building average air temperature ensure that '

the overall containment average air temperature does not e.xceed the initial temperature condition assumed in the accident analysis for a steam line break accident.

3/4.6.1.6 REACTOR BUILDING STRUCTURAL INTEGRITY This limitation ensures that the structural integrity of the containment will be maintained comparable to the origina'. design standards for the life of the facility. Structural integrity is required to ensure that the containment will withstand the maximum pressure of 53.5 psig in the event of a steam line ,

break accident. The measurement of containment tendon lift off force, the l tensile tests of the tendon wires, the visual examination of tendons, anchorages and exposed interior and exterior surfaces of the containment, and the Type A l leakage test are sufficient to demonstrate this capability. '

The tendon lift off forces are evaluated to ensure that 1) the rate of ;

tendon force loss is within predicted limits, and 2) a minimum required prestress level exists in the containment. In order to assess the rate of i force loss, the lift off force for a tendon is compared with the force predicted for the tendon times a reduction factor of 0.95. This resulting force is referred to as the 95% Base Value. The predicted tendon force is equal to the original stressing force minus losses due to elastic shortening of the tendon, stress relaxation of the tendon wires, and creep and shrinkage of the concrete.

The 5% reduction en the predicted force is intended to compensate for both j uncertainties in the prediction techniques for the losses and for inaccuracies i in the lift-off force measurements.

SUMMER - UNIT 1 B 3/4 6-2 Amendment No. 49 :

l

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ML20059G290

Text

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