Text

Forwards,For Info & Use,Comments & Justifications for Changes Proposed in Proof & Review Tech Specs
	ML20207T186
Person / Time
Site:	South Texas
Issue date:	03/16/1987
From:	Wisenburg M; HOUSTON LIGHTING & POWER CO.
To:	; NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
	ST-HL-AE-1984, NUDOCS 8703230291
	Download: ML20207T186 (220)
	v • d • e

The Light Company ii->n u iioi,x x & i- i. a im im ii_,,,. i_ ,,mi

. gi33 m.wii March 16, 1987 ST-HL-AE-1984 File No.: G9.06 10CFR50.36 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555 South Texas Project Unit 1 Docket No. STN 50-498 Proof and Review Technical Specification Comments

References:

ST-HL-AE-1802 dated February 24, 1987 ST-HL-AE-1867 dated January 7, 1987 ST-HL-AE-1862 dated January 27, 1987 ST-HL-AE-1882 dated January 13, 1987 ST-HL-AE-1883 dated January 13, 1987 ST-HL-AE-1897 dated January 28, 1987

ST-HL-AE-1901 dated February 2, 1987 ST-HL-AE-1930 dated February 24, 1987 ST-HL-AE-1923 dated March 13, 1987 Attached for your information and use are Houston Lighting & Power's comments and justifications for changes proposed in the following Proof and Review Technical Specifications (Tech Specs):

1) Definitions

2) 2.0 Safety Limits and Limiting Safety System Settings

3) Safety Limit Bases

4) 3/4.1 Reactivity Control Systems

5) 3/4.2 Power Distribution Limits

6) 3/4.3 Instrumentation

7) 3/4.4 Reactor Coolant System

8) 3/4.5 Emergency Core Cooling Systems

9) 3/4.6 Containment Systems

10) 3/4.7 Plant Systems

11) 3/4.8 Electrical Power Systems

12) 3/4.9 Refueling Operations

13) 3/4.11 Radioactive Effluents

14) BASES

15) 5.0 Design Features

16) 6.0 Administrative Controls 8703230291 870316 L3/NRC/cm PDR A

ADOCK 05000498 PDR

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liouston 1.ighting & Power Company ST-HL-AE-1984 File No.: G9.06 Page 2 Where justification was previously provided and is still under NRC review, the letter and date of the previous justification are provided as technical justification for the proposed change in Proof and Review.

Please note that Veritrak transmitter uncertainty and RTD response time are open issues with Westinghouse and resolution of these issues, as they affect Tech Specs, will be provided in a subsequent letter. Also, Instrumentation

~ items were discussed during the March 9, 1987 meeting with NRC staff. Wording of items related to that meeting is provided in the attached comments.

If you should have any questions on this matter, please contact Ms.

Frostie A. White at (512) 972-7985.

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M. R. W enburg Deputy P oject Ma gr FAW/ljm Attachments: 1) Definitions

2) 2.0 Safety Limits and Limiting Safety System Settings

3) Safety Limit Bases

4) 3/4.1 Reactivity Control Systems

5) 3/4.2 Power Distribution Limits

6) 3/4.3 Instrumentation

7) 3/4.4 Reactor Coolant System

8) 3/4.5 Emergency Core Cooling Systems

9) 3/4.6 Containment Systems

10) 3/4.7 Plant Systems

11) 3/4.8 Electrical Power Systems

12) 3/4.9 Refueling Operations

13) 3/4.11 Radioactive Effluents

14) BASES

15) 5.0 Design Features

16) 6.0 Administrative Controls L3/NRC/cm

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flouston 1.ighting & Power Company ST-HL-AE-1984' File No.: G9.06' Page 3 cc:

Regional Administrator, Region.IV. M.B. Lee /J.E. Malaski Nuclear Regulatory Commission City of Austin 611 Ryan Plaza Drive, Suite 1000 P.O. Box 1088 Arlington, TX -76011 Austin,1DC ' 78767-8814

'N, Prasad Kadambi, Project Manager M.T. Hardt/A. von Rosenberg U.S. Nuclear Regulatory Commission City Public Service Board 7920 Norfolk Avenue P.O. Box 1771-Bethesda,.MD 20814 San Antonio, TX 78296 Robert L. Perch, Project Manager Advisory Committee on Reactor Safeguards' U.S. Nuclear Regulatory Commission U.S. Nuclear Regulatory Commission 7920 Norfolk Avenue 1717 H Street Bethesda, MD- 20814 Washington, DC 20555 Dan R. Carpenter-Senior Resident Inspector / Operations e/o U.S. Nuclear Regulatory .

Commission P.O. Box 910 Bay City, TX' 77414' Claude E. Johnson Senior Resident Inspector /STP c/o U.S. Nuclear Regulatory Commission P.O. Box 910 Bay City, TX 77414 M.D. Schwarz , Jr. , Esquire Baker & Botts One Shell Plaza Houston, TX 77002 J.R. Newman, Esquire Newman & Holtzinger, P.C.

1615 L Street, N.W.

Washington, DC 20036 T.V. Shockley/R.L. Range Central Power & Light Company P. O. Box 2121 Corpus Christi, TX 78403 L3/NRC/cm

fl L ATTACHMENT / !

ST-HL-AE /9#'/

,2AGLJ.OF3 Definitions A. Definition 1.7, page 1 Item a 2) was changed to reflect deletion of Containment Isolation Valves as justi-fled in letter ST-HL-AE-1923.

B. Definition 1.10, page 1 Ch'anged to allow using data base manipulation in cases where injection of simulated data is not possible. The justification for this change was provided in ST-HL-AE-1901.

C Definition 1.11, Page 1 Change to Regulatory Guide 1.109 to specify South Texas commitment as identified in FSAR Table 3.12-1. South Texas is committed to this Regulatory Guide.

l L3/NRC/cm

ATTACHMENT /

. ST-HL-AE /## ,

l PAQROj(DQ 3 Dmfpll[

DEFINITIONS CONTAINMENT INTEGRITY 1.7 . CONTAINMENT INTEGRITY shall exist when:

a.

All penetrations are either: required to be closed during accident conditions 1)

Capable isolation of being valve closedorby an OPERABLE containment automatic system, 2)

Closed by manual valves, blind flanges, or deactivated automatic valves secured in their closed positions, except as provided in Tem e & 1 et Specification 3.6.3.

b.

All equipment hatches are closed and sealed, c.

Each3.6.1.3, tion air-lock is in compliance with the requirements of Specifica-

, d.

The containment tion 3.6.1.2, and leakage rates are within the limits of Specifica-e.

bellows, or 0-rings) is OPERABLE.The sealing mechanism asso O CONTROLLED LEAKAGE b

1.8 coolant pump seals. CONTROLLED LEAKAGE shall be that seal water flow suppli' CORE A_LTERATIONS 1.9 within the vessel.the reactor pressure vessel with the vessel head r Suspension of CORE ALTERATION shall . t preclude completion of movement of a component to a safe conservative position.

DIGITAL CHANNEL OPERATIONAL TEST 1.10 or A DIGITAL CHANNEL OPERATIONAL TEST shall xercisina consist ofthe digital computer hardware using data base manipulation in.iecting simulated

~ process %w dat@to verify eMakle OPERABILITY of alarm, interlock, and/or trip functions.

DOSE EQUIVALENT I-131 1.11 which alone would produce the same thyroid dose as the mixture of I-131, I-132, I-133, I-134, and I-135 actually present. .

dose conversion factors used for this calculation shall be those listed inThe thyroid O. '

SOUTH TEXAS - UNIT 1 1-2

ATTACHMENT /

- ST HL-AE /41V F90AM AFr,5 Vin.g; rsnov. ...

y -- 3-DEFINITIONS DOSE EQUIVALENT I-131 (Continued) 9kd Tw;,_li-I af T ne Micuiatim 04++. nee Factur-s fort hmFgyggg, R-r ';ttee.+

(kchn %es 09 G.de. \.\o9 "Cabdon of- Awd b:aet to Gn Rorn E % &s k e- % b sc. ob %& hanz v.6 gOcEn f W o, E - AVERAGE DISINTEGRATION ENERGY

@p.nd4_r

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1.12 E shall be the average (weighted in proportion to the concentration of each radionuclide in the sample) of the sum of the average beta and gamma energies per disintegration (MeV/d) for the radionuclides in the sample.

ENGINEERED SAFETY FEATURES RESPONSE TIME 1.13 The ENGINEERED SAFETY FEATURES (ESF) RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ESF Actuation Setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e.,-the valves travel to their required positions, pump '

discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays where applicable.

FREQUENCY NOTATION 1.14 The FREQUENCY NOTATION specified for the performance of Surveillance

, Requirements shall correspond to the intervals defined in Table 1.1.

GASEOUS WASTE PROCESSING SYSTEM 1.15 A GASEOUS WASTE PROCESSING SYSTEM shall be any system designed and installed to reduce radioactive gaseous effluents by collecting Reactor Coolant System offgases from the Reactor Coolant System and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment.

l IDENTIFIED LEAKAGE l

j 1.16 IDENTIFIED LEAKAGE shall be:

a. Leakage (except CONTROLLED LEAKAGE) into closed systems, such as pump seal or valve packing leaks that are captured and conducted to a sump or collecting tank, or

b. . Leakage irito the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of Leakage Detection Systems or not to be PRESSURE BOUNDARY LEAKAGE, or

c. Reactor Coolant System leakage through a steam generator to the Secondary Coolant System.

O SOUTH TEXAS - UNIT 1 1-3

- ~. --

- ATTACHMENT 42 f~^. - ST-HL-AE /7ft' PAGE / OF 9 2.0 Safety Limits and Limiting Safety System Settings

'A. Section 2.1.2,.page 2-1: The safety limit is 110% of the RCS design

. pressure.(2500 psia). In gauga units this value-is 2735 psig. As this value is'used in other Westinghouse plant technical specifications for plants with identical design pressure limits, we believe-this cor.stitutes' a generic change. The proof and review value represents a technically insignificant difference; however, it does create an additional plant specific parameter that operators must learn.

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B. Table 2.2-1, page 2-4: (1) Specific values provided based upon the Westinghouse WCAP 11273 which is the Setpoint Methodology for South Texas.

(2) The. Loop Design Flow is 95,400 as shown on FSAR page 5.1-1, Amendment 57.

page 2-5: (1) Specific values provided; see (1) above.

(2) The turbine stop valve closure trip setpoint and allowable value are provided; agreed upon during the March 9, 1987 meeting with NRC Staff.

page 2-6: Specific values provided; see 2-4 (1) above.

page 2-7: South Texas does not have a RTD manifold.

page 2-8: Specific value provided; see 2-4 (1) above, page 2-9: Setpoint methodology nomenclature and brackets on T equation.

page 2-10: Specific value provided; see 2-4 (1) above.

L3/NRC/cm

. ST-HL AE l98 2.0 SAFETY LIMITS AND LIMTTING SAFETY SYSTEM SETTINGS . PAGE..iLOF7 2.1 SAFETY LIMITS 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 Figure 2.1-1.

APPLICABILITY: MODES 1 and 2.

ACTION:

Uhene W r the point defined by the combination of the highest operating loop average temperature and THERMAL POWER has exceeded the appropriate pressurizer pressure line, be in HOT STANDBY within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , and comply with the require-ments of Specification 6.7.1.

REACTOR COOLANT SYSTEM PRESSURE

-27W 2.1.2 The Reactor Coolant System pressure shall not exceed 2343 psig.

APPLICABILITY: MODES 1, 2, 3, 4, and 5.

ACTION:

O MODES 1 and 2:

213r Whenever the Reactor Coolant System pressure has exceeded 2733 psig, be ,

in HOT SlANDBY with the Reactor Coolant System pressure'within its limit within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , and comply with the requirements of Specification 6.7.1.

MODES 3, 4 and 5:

2 93 s '

Whenever the Reactor Coolant System pressure has exceeded 2EP.Ppsig, reduce the Reactor Coolant System pressure to within its limit within 5 minutes, and comply with the requirements of Specification 6.7.1.

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SOUTH TEXA5 - UNIT 1 2-1

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TABLE 2.2-1 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS SENSOR M TOTAL ERROR

% FUNCTIONAL UNIT ALLOWANCE (TA) Z (S) TRIP SETPOINT ALLOWABLE VALUE j i 1. Manual Reactor Trip N.A. N.A. N.A. N.A. N.A. ;

E 2. Power Range, Neutron Flux 3 l O a. High Setpoint 7.5 4.56 0 $109% of RTP** $111.J% of RTP**

e-

b. Low Setpoint 8.3 4.56 0 125% of RTP** 127.3% of RTP**

3. Power Range, Neutron Flux, 1.6 0.50 0 <5% of RTP** with <6.3% of RTP** with High Positive Rate i time constant i tin.a constant l >2 seconds

>2 seconds

4. Power Range, Neutron Flux, 1.6 0.50 0 15% of RTP** with <6.3% of RTP** with om>

High Negative Rate a time constant i time constant 74 l

>2 seconds >2 seconds mh to 5. Intermediate Range, 17.0 8.41- 0 -<25% of RTP** <30=9% of RTP** mk 1 Neutron Flux 31-1 Q i 6. Source Range, Neutron Flux 17.0 10.01, 0 <105 cps ^<1.4 x 105 cps 4 7(g k

7. Overtemperature AT 6.8 4.f 32 3D U +Lo ~See Note 1 See Note 2

8. Overpower AT 5.5 1.40 G # 1.( See Note 3 See Note 4 IR Q

9. Pressurizer Pressure-Low 3.1 0.71 & & % D >1870 psig >M%-2 psig

10. Pressurizer Pressure-High 3.1 0.71 332.0{2380psig < psig

11. Pressurizer Water Level-High 8/.0 2. g E:S g. 0592% of instrument k9 of instrument span span g

12. Reactor Coolant Flow-Low 2.5 agi- 0.6 >90% of loop >89.g%ofloop

.1.33 Besign flow

Besign flow

go

Loop design flow = gpm

- RTP = RATED THERMAL POWER THIS PAGE OPEN PENDING RECEIPT OF 3 INFORMATLJ 30MTHE APPLICMIT 4

O O O TABLE 2.2-1 (Continued) g

% REACTOR TRIP SYSTEM INSTRUNENTATION TRIP SETPOINTS g TOTAL SENSOR g FUNCTIONAL UNIT ALLOWANCE (TA)

ERROR Z (S) TRIP SETPOINT ALLOWABLE VALUE 7 13. Steam Generator Water 15.0 12 48 1:5 c Level Low-Low .75 233% of narrow 231.#)l% of narrow

$ 2 0+d.2 range instrument range instrument span span l 14. Undervoltage - Reactor 10. 0.3 Coolant Pumps 0. O 110,Q2B volts volts 1

i 15. Underfrequency - Reactor 3.4 Coolant Pumps 0.0/ 0.0 ->57.2 Hz >57.1 Hz

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16. Turbine Trip

a. Low Emergency Trip Fluid 232.1 rp 100.8 131.3 11245.8 psig 211f4.5psig Pressure m

mha

b. Turbine Stop Valve N.A. N.A. N.A. f

4. % c\csed 9aly closej Qg Closure -8 5 egen og g

17. Safety Injection Input N.A. N.A. N.A. N.A.

from ESFAS N.A.

3 8,

90

D W

8 A

A V

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TABLE 2.2-1 (Continued) g REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS 5

z SENSOR

--i TOTAL ERROR h FUNCTIONAL UNIT ALLOWANCE (TA) Z (S) TRIP SETPOINT ALLOWABLE VALUE 7 18. Reactor Trip System c Interlocks z

Z a. Intermediate Range N.A. N.A. N.A. ll x 10 30 amp s Neutron Flux, P-6 )6 x 10 88 amp

b. Low Power Reactor Trips Block, P-7 4

1) P-10 input N.A. N.A. N.A. <10% of RTP** $12. of RTP**

2) P-13 input l

ep m

N.A. N.A. N.A. <10% RTP** Turbine Tmpulse Pressure Equivalent

<12. RTP** Turbine Impulse Pressure hh

%h>

' Equivalent a>9 3 F

c. Power Range Neutron Flux, P-8 N.A. N.A. N.A. 148% of RTP** $50. $ of RTP** ME D E

l d. Power Pe,9e Neutron N.A. N.A. N.A.

3 Mh Flux, P-9 $50% of RTP** $62./%ofRTP**

e. Power Range Neutron N.A. N.A. N.A Flux, P-10 210% of RTP** 27. ) of RTP**

f. Turbine Impulse Chamber N.A. N.A. N.A. <10% RTP** Turbine <12. 3 Pressure, P-13 RTP** Turbine Impulse Pressure Impulse Pressure O Equivalent Equivalent '

19. Reactor Trip Breakers N.A.

s go

N.A. N.A N.A. N.A. 2

20. Automatic Trip and Interlock N.A.

Logic N.A. N.A. N.A. N.A. $

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- RTP = RATED THERMAL POWER 8

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O (3 O V Al TABLE 2.2-1 (Continued)

, TABLE NOTATIONS E

H

NOTE 1: OVERTEMPERATURE AT P 4E l

'AT 1 (1 + TaS} i o

-K 2 y d

T(j, - 1] + K3 (f - P' ) - f, (F) }

v, e

RCS g Where: AT = Measured AT by IE31Ai3EEEE5tzt Iristrumentation; 1

= Lead-lag compensator on measured AT; ti, T2 = Time constants utilirrd in lead-lag compensator for AT, t i = 8 sec, T2 = 3 sec; yf 3

Lag compensator on measured AT; y 13

Timeconstant[utilizedinthelagcompensatorforAT,13 = 0 sec; Jhg AT, = Indicated AT at RATED THERMAL POWER; Sh> g j'F E Ki = 1.08; Q_g -o g K2 = 0.0185/ F; *du 1 t' l =

} The function dynamic generated by the lead-lag compensator for T,yg compensation; T4, is = = 33 sec, Time constants utilized in the lead-lug compensator for T3yg, 14 Is = 4 sec; 5

T = Average temperature, F; i 8,r 1

=

Lag compensator on measured Tavg; 9 2

= Tinie constant utilized in the measured T ag 1 g compensator To = 0 sec; m

Tc a

~8

O O O TABLE 2.2-1 (Continued) o" TABLE NOTATIONS (Continued)-

5

[ NOTE 1: (Continued) m j

q 5

m T' -< 593.0*F (Nominal Tavg at RATED THERMAL POWER);

i K3 = 0.000857/psig; j E j Z P = Pressurizer pressure, psig; i ~

P' = 2235 psig (Nominal RCS operating pressure);

5 = Laplace transform operator, sec 1; j i and f (AI) is a function of the indicated difference between top and bottom detectors of the

' power range neutron ion chambers; with gains to be selected based on measured instrument response during plant startup tests such that:

m !$$

i o'o (1) For(qt 4hbetween-39%and+10%,f(AI)=0,whereqt and qb are percent RATED THERMAL 'Mh>}

t g

POWER in the top and bottom halves of the core respectively, and qt + q bis total THERMAL oim

POWER in percent of RATED THERMAL POWER; m mz qH M

l (2) Foreachpercentthatthemagnitudeof(qt 9}b exceeds-39%,theATTripSetpointshall be automatically reduced by 1.55% of its value at RATED THERMAL POWER; and (3) Foreachpercentthatthemagnitudeof(qt qhexceeds+10%,theATTripSetpointshall be automatically reduced by 1.52% of its value at RATED THERMAL POWER.

4 R

! NOTE 2: The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than O i

I 5% AT span.

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2.0% R*

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i 8

m i

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O O O TABLE 2.2-1 (Continued) y TABLE NOTATIONS (Continued)

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4 h NOTE 3: OVERPOWER AT E S) ( 1 ) ( t,S ) ( 1 ) ( 1 ' #

-K 3 (y e

AT ((1 + 2 ti1 + T 5) (1 + 0 (K4 IaS) i T7 3) (y Tsg) T - Ks )-T{-f(N)I (y Tsg) 2 E

Z w Where: AT = As defined in Note 1, 1 = As defined in Note 1, It. T2 = As defined in Note 1,

= As defined in Note 1, 1 + taS . _ . _ . _

T m

m u ><

+4 Ta = As defined in Note 1,

>h M

>I AT, = As defined in Note 1, g%

o

'i g z K4 = 1.08,

%"b

! K3 = 0.02/*F for increasing average temperature and 0 for decreasing average temperature, b

ld

=

3[7 3 The function generated by the eate-lag compensa..cr for T,yg dynamic compensation, i

3 lead d 17 =

Time constants utilized in the rate-lag compensator for Tavg, ty = 10 ,sec, o

= As defined in Note 1, 1 + tc5 m_

to = As defined in Note 1, Q

c>

O

O O O i

TABLE 2.2-1 (Continued)

$ TABLE NOTATIONS (Continued.j m

y NOTE 3: -(Continued) 2 y Ks =

0.00135/ F for T > T" and Ks = 0 for T 5 T", l E T = As defined in Note 1,

-e w T" =

Indicated T,yg at RATED THERMAL POWER (Calibration temperature for AT instrumentation, 5 593.0 F),

S = As defined in Note 1, and f 2(AI) = 0 for all AI.

7 NOTE 4: The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than Es 3.j/%ATspan.

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ATTACHMENT 2

. ST HL-AE- l124

~ SAFETY LIMIT BASES A. B2-1: FSAR Section 4.2.2.1 describes that at South Texas the W-3 R Grid correlation is being used.

B. B2-2: See previous justification on page 2-1.

C. B2-5: Response time for overtemperature delta-T does not need to be addressed in this section. Response times for reactor trip-

-signals are provided in Section 3/4.3 (Table 3'.3-2).

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L3/NRC/cm

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. PAGFJ ST HL-AEOF %'d.

rR00FT Mtvitw COPY 2.1 SAFETY LIMITS 7

(7/ BASES 2.1.1 REACTOR CORE The restrictions of this Safety Limit prevent overheating of the fuel and possible products cladding to the reactorperforation coolant. which would result in the release of fission 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 slightlyabovethecoolantsaturationtemperkture.

RGoa &

Operation above the upper boundary of the nucleate boiling regime could result in excessive cladding temperatures becaus of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient. DNB is not a directly measurab parameter during operation and therefore THERMAL POWER and reactor coolant em erature and pressure have been related to DNB through the W-3 correlation. The W-3 DNB correlation has been developed to predict the DNB flux and the location of DNB for axially uniform and nonuniform heat flux distributions.

is" defined as the ratio of the heat fluxThe local that DNBcause would heat flux ratioat DNB (DNBR) a particular core location to the local heat flux and is indicative of the margin to DNB.

The minimum value of the DNBR during steady-state operation, normal operational transients, and anticipated transients is limited to 1.30. This j value corresponds to a 95% probability at a 95% confidence level that DNB will not occur and is chosen as an appropriate margin to DNB for all operating conditions.

The curves of Figure 2.1-1 show the loci of points of THERMAL POWER, Reactor Coolant System pressure and average temperature for which the minimum DNBR is no less to the enthalpy ofthan 1.30, liquid.

saturated or the average enthalpy at the vessel exit is equal These curves are based on an enthalpy hot channel factor, F H, f 1.52 and a reference cosine with a peak of 1.61 for axial power shape. An allowance is included for an increase in F g at reduced power based on the expression:

F H = 1.52 W 0.3 (1 %

Where P is the fractiun 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 f insertion assuming the axial power imbalance is within the limits of the 2 (AI) function of the Overtemperature trip.

When the axial power imbalance -

is not within the tolerance, the axial power imbalance effect on the Over-temperature with AT trips core Safety will reduce the Setpoints to provide protection consistent Limits.

3 SOUTH TEXAS - UNIT 1 B 2-1

ATTACHMENT J

. ST.HL.AE 19f 4 PAGE J_QF 4 rnour & MtvitW COPY SAFETY LIMITS g-BASES 2.1.2 REACTOR COOLANT SYSTEM PRESSURE The restriction of this Safety Limit protects the integrity of the Reactor Coolant System (RCS) from overpressurization and thereby prevents the release of radionuclides contained in the reactor coolant from reaching the containment atmosphere.

2'l W The reactor vessel, pressurizer, and the RCS piping, valves, and fittings are designed to Section III of the ASME Code for Nuclear Power Plants which permits a maximum transient pressure of 110% (2233 psig) of design pressure.

The Safety Limit of 2733 psig is therefore consistent with the design criteria and associated Code requirements.

%llo The entire RCS is hydrotested at 125% (3107 psig) of design pressure, to demonstrate integrity prior to initial operation.

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O SOUTH TEXAS - UNIT 1 8 2-2

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4TTACHMENT 3

. ST.HL-AE. IW/

PAGE_ X QF_V___ __

riTUUr 4 titVitVV UUt'T LIMITING SAFETY SYSTEM SETTINGS '

BASES Intermediate and Source Rance. Neutron Flux The Intermediate and Source Range, Neutron Flux trips provide core protection during reactor startup to mitigate the consequences of an uncon-trolled condition. rod cluster control assembly bank withdrawal from a subcritical of the Power Range, Neutron Flux channels.These trips provide redundant protection initiate a Reactor trip at about 105 The Source Range channels will when P-6 becomes active. counts per second unless manually blocked above P-10. The Source Range channels are automatically blocked The Intermediate Range channels will initiate a Reactor trip at a current level equivalent to approximately 25% of RATED THERMAL POWER unless manually blocked when P-10 becomes active.

Overtemperature AT The Overtemperature AT trip cogbinations of pressure, power, provides core protection to prevent DN8 for all coolant temperature, and axial power distribu-tion, provided that the from the core to the temperature detectors transient is slow with respect to piping transit delays

(..~ ? . .. 2), and pressure is within the range between point is automatically varied with: (1) coolant the Pressurizer High and Low Pressure trips. The Set-temperature to correct for temperature-induced changes in density and heat capacity of water and includes O dynamic compensation for piping dela detectors, (2) pressurizer pressure,ysand from(3)the corepower axial to thedistribution.

loop temperature With normal axial power distribution, this Reactor trip limit is always below the core Safety Limit as shown in Figure 2.1-1. If axial peaks are greater than design, as indicated by the difference between top and bottom power range nuclear detectors, the Reactor trip is automatically reduced according to the notations in Table 2.2-1.

Overpower AT The Overpower AT trip provides assurance of fuel integrity (e.g., no fuel pellet melting and less than 1% cladding strain) under all possible overpower conditions, limits the required range for Overtemperature AT trip, and providesvaried is automatically a backup with:to the High Neutron Flux trip. The Setpoint (1) coolant temperature to correct for temperature-induced changes in density and heat capacity of water, and (2) rate of change of temperatere for dynamic compensation for piping delays from the core to the loop temperature detectors, to ensure that the allowable heat generation rate (kW/ft) is not exceeded. The Overpower AT trip provides

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

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SOUTH TEXAS - UNIT 1 B 2-5

ATTACHMENT '/

- - ST-HL AE . Wf PAGE / OF /d 3/4.1 Reactivity Control Systems A. 3/4 1-1, -la: In letter ST-HL-AE-1897, HL&P provided proposed Tech Specs based upon Boron Dilution Analysis. Subsequent discus-sions with.the NRC Staff have revealed that it will be necessary to include the Shutdown Margin versus RCS Critical Boron Concentration' curves in the Tech Specs. As such, the curve for Modes 1, 2, 3 and 4,' Figure 3.1.1-1, is added to the LCO and Surveillance Requirements,-accord-ingly.

B. 3/4 1-3, -3a: See previous comment. The curve in Figure'3.1.1-2 is for MODE 5.

-C. 3/4 1-8: (1) LCO 3.1.2-2a adds a gravity feed connection and is

' justified in letter ST-HL-AE-1897.

(2) The ACTION on 3.1.2.2 reflects Boron D11ution Analysis as described in letter ST-HL-AE-1897.

D. 3/4 1-9: The "**" note was added and justified in letter ST-HL-AE-1897.

E. 3/4 1-10: Adding reference to Figure 3.1.1-2 in the ACTION to 3.1.24 for Boron Dilution as described in letter ST-HL-AE-1897.

F. 3/4 1-12: Adding reference to Figure 3.1.1-2 in the ACTION to LCO 3.1.2.6 reflects Boron Dilution Analysis; see letter ST-HL-AE-1897.

G. 3/4 1-22: Camera-Ready Figure provided via letter ST-HL-AE-1930.

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ATTACHMENI y

. ST.HL.AE /f/1 PAGE .1 OF /d i) ROOF & REVIEW COPY 3/4.1 REACTIVITY CONTROL SYSTEMS

() 3/4.1.1 B0 RATION CONTROL SHUTDOWN MARGIN - T,y GREATER THAN 200*F LIMITING CONDITION FOR OPERATION 3.1.1.1 The SHUTDOWN MARGIN shall be greater than or equal to ! M n k /k -the. lM APPLICABILITY: MODES 1, 2*, 3, and 4.

ACTION:

-6e rpre) hmd With the SHUTDOWN MARGIN less than .L f5i A W , immediately initiate and con-tinue boration at greater than or equal to 30 gpm of a solution containing greater than or equal to 7000 ppm boron or equivalent until the required SHUTDOWN MARGIN is restored.

SURVEILLANCE REOUIREMENTS 4.1.1.1.1 The SHUTDOWN MARGIN shall be determined to be greater than or equal to lE E-/ hike.repted lists %A in Eyee. a. i .t-.1:

a. Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after detection of an inoperable control rod (s) and

.\ at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter while the rod (s) is inoperable.

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

b. When in MODE 2 with K,ff less than 1, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior to achieving reactor criticality by verifying that the predicted critical control rod position is within the limits of Specification 3.1.3.6;

., c. Prior to initial operation above 5% RATED THERMAL POWER after each fuel loading, by consideration of the factors of Specifica-tion 4.1.1.1.1(J below, with the control banks at the maximum inser-tion limit of Specification 3.1.3.6; and O' *See Special Test Exceptions Specification 3.10.1.

SOUTH TEXAS - UNIT 1 3/4 1-1

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ATTACHMENT V ST-HL-AE 1984 DBfGft- 5 GE,1, ---,

u = =www w utbygLgy W fg REACTIVITY CONTROL SYSTEMS SHUTDOWN MARGIN - T,yg LESS THAN OR EQUAL TO 200 F ,

+

LIMITING CONDITION FOR OPERATION 3.1.1.2 The SHUTDOWN MARGIN shall be greater than or equal to H s a b U M APPLICABILITY: MODE 5.

'" O ACTION:

h repwed hwsY With the SHUTDOWN MARGIN less than # =, immediately initiate and continue boration at greater than or equal to 30 gpm of a solution containing greater than or equal to 7000 ppm boron or equivalent until the required SHUTDOWN MARGIN is restored.

SURVEILLANCE REQUIREMENTS 4.1.1.2 The SHUTDOWN MARGIN shall be determined to be greater than or equal to E n 4 & % t speb b h gn pip n.,3,g,[.7,

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

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

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

1) Reactor Coolant System boron concentration,

2) Control rod position,

3) Reactor Coolant System average temperature,

4) Fuel burnup based on gross thermal energy generation,

5) Xenon concentration, and

6) Samarium concentration.

O SOUTH TEXAS - UNIT 1 3/4 1-3

ATTACHMENT V '

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l Versus RCS Critical Boron Concentration 24.1 la 1

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. ST HL.AE. rif4 ~

PAGE 4 OF /c c

muur & REVIEW COPY REACTIVITY CONTROL SYSTEMS FLOW PATHS - OPERATING LIMITING CONDITION FOR OPERATION l-3.1.2.2: At least two of the following three boron injection flow paths shall be OPERABLE:

i

a. The flow path from the Boric Acid Storage System via.a boric acid !

. transfer (RCS),and pumpvand a charging pump to the Reactor Coolant System

y-Te*d come.dion

b. ;

Two pumps to the RCS.

flow paths from the refueling water storage tank via charging i APPLICABILITY: MODES 1, 2, and 3.*

ACTION:

With only one of the above required boron injection flow paths to the RCS OPERABLE, restore at least two boron injection flow paths to the RCS to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY and borated to a SHUTDOWN MARGIN equivalent to at least E i s n M9; @ ithin thel next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; restore at least two flow paths to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . .

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SURVEILLANCF REQUIREMENTS h Rpre. s.).\-1 4.1.2.2 At least two of the above required flow paths shall be demonstrated OPERABLE:

a. At least once per 7 days by verifying that the temperature of the ;

heat traced portion of the flow path from the boric acid tanks is l greater than or equal to 65'F when it is a required water source;

' b. At least once per 31 days by verifying that each valve (manual,

- power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position;

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

d. At least once per 18 months by verifying that the flow path required by Specification 3.1.2.2a. delivers at least 30 gpm to the RCS.

The provisions of Specifications 3.0.4 and 4.0.4 are aot applicable for entry into MODE 3 for the charging pump declared inoperable pursuant to Specifica-tion 4.1.2.3.2 provided that the charging pump is rrstored to OPERABLE status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or prior to the temperature of one or more of the RCS cold legs exceeding 375'F, whichever comes first.

SOUTH TEXAS - UNIT 1 3/4 1-8

AftACHMENT #

ST.HL AE N/ . '

_P. AGE.S OF /0 PROOF & REVIEW COPY REACTIVITY CONTROL SYSTEMS

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[V CHARGING PUMP - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.2.3 One charging pump in the boron injection flow path required by Specification 3.1.2.1 shall be OPERABLE and capable of being powered from an OPERABLE emergency power source.

M APPLICABILITY: MODES 4, 5, and 6.

ACTION:

With no charging pump OPERABLE or capable of being powered from an OPERABLE emergency power source, suspend all operations involving CORE ALTERATIONS or positive reactivity changes.

SURVEILLANCE REQUIREMENTS 4.1.2.3.1 The above required charging pump shall be demonstrated OPERABLE by verifying, on recirculation flow, that a differential pressure across the pump of greater than or equal to 2300 psid is developed when tested pursuant to Specification 4.0.5.

p 4.1.2.3.2 All charging pumps, excluding the above required OPERABLE pump, V shall be demonstrated inoperable

at least once per 31 days, except when the reactor vessel head is removed, by verifying that the motor circuit breakers are secured in the open position.

l l

l l *An inoperable pump may be energized for testing provided the discharge of the pump has been isolated from the RCS by a closed isolation valve with power A removed from the valve operator, or by a manual isolation valve secured in the l

V closed position.

c9anaE.enn h rs dYu d" A c w 1"#' h'"Z"m m2 3%;r d dwe skdLbt wedare a me s ==e 4 %."5 d SOUTH TEXAS - UNIT 1 3/4 1-9 %, %q % m w. h l

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ATTACHMENT '/

W/

.If^Cd1 P ST HL-Ag-(fyl$,y 8 copy REACTIVITY CONTROL SYSTEMS

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/N V CHARGING PUMPS - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.4 At least two charging pumps shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3.*

ACTION: 4 g;Wh sph%eb in Eye U I estore at least two charging pumps to WithonlyonechargingpumpOPERABLE,(inatleastHOTSTANDBYandboratedtoa OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be M!UTDOWN MARGIN equivalent to at leastlli sw = _E F within the next E hours; restore at least two charging pumps to OPERABLE status within the next T eys or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SU WL RAANCE REOUIREMENTS l

4.1.2.4 At least two charging pumps shall be demonstrated OPERABLE by verifying, on recirculation flow, that a differential pressure across each pump of greater than or equal to 2300 psid is developed when tested pursuant to Specification 4.0.5. ,

O i

The prov'isions of Specification 3.0.4 and 4.0.4 are not applicable for entry into MODE 3 for the charging pumps declared inoperable pursuant to Specifica-tion 4.1.2.3.2 provided that the charging pump is restored to OPERABLE status O

V within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or prior to the temperature of one or more of the RCS cold legs exceeding 375*F, whichever comes first.

l l

SOUTH TEXAS - UNIT 1 3/4 1-10

ATTACHMENT 4

. ST HL.AE- @/

nR&GL T,C L /

' nuve a r r.vmvf wrr REACTIVITY CONTROL SYSTEMS hv 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 Specification 3.1.2.2:

a. A Boric Acid Storage System with:

1) A minimum contained borated water volume of 27,000 gallons,

2) A minimum boron concentration of 7000 ppm, and

3) A minimum solution temperature of 65 F.

b. The refueling water storage tank (RWST) with:

1) A minimum contained boratea water volume of 458,000 gallons, and

2) A minimum boron concentration of 2500 ppm.

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

ACTION: & tW ksps died M Wyre 11. l-2.

O)

\

m a. With the Boric Acid Storage System inoperable and being used as one of the above required orated water sourc- , restore the system to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or 9 in at least HOT STANDBY within the nex 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and borated to a SHUTDOWN MARGIN equivalent to at least 1% +"- ;

2^^O restore the Boric Acid Storage System to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b. With the RWST inoperable, restore the tank to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT STANDBY within the next

,, 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

h m

SOUTH TEXAS - UNIT 1 3/4 1-12

ATTACHMENT V ST.HL A5. l9FV;o PMg)np5ppylpg ggpy O +

O i

i THIS PAGE OPEN PDIDING RECEIPT OF INFORMATION FROM THE APPLICANT FIGURE 3.1-1 ROD BANK INSERTION LIMITS VERSUS THERMAL POWER FOUR LOOP OPERATION SOUTH TEXAS - UNIT 1 3/4 1-22

ATTACHMENT 5

- ST44L-AE J7+1 -

PAGE /-OF C

'~~~

3/4.2 Power Distribution Limits-

.A. 3/4 2-1, 2-2: Deleted "**" note to ACTION b.2 and inserted under;the ~

APPLICABILITY'since the note should apply when surveil -

lance testing is allowed with the~AFD outside of the target band.

B. 3/4:2-4: . Figure 3.2-1 was submitted via letter-ST-HL-AE-1930.

C. 3/4 2-9: 11e Nuclear Enthalpy Rise Hot ' Channel Factor equation' was changed to reflect FSAR Section 4.3.2.2.6, page'4.3-17.

The constant 1.52 was reduced by 4%. measurement uncertain-ty as defined in the Bases and surveillance 4.2.3.2.

D. 3/4:2-11i (1)' The DNB parameters are provided with RCS flow using 3.5%

uncertainty, which is based upon accident analysis.

.(2)- There is no Table 3.2-1 since the modification to 3.2.5 was incorporated.

L3/NRC/cm I.

ATTACHMENTS ST-HL AE lff/

PA g d J){ #

nuvr ai ncystyy cun 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLbX DIFFERENCE ,

LIMITING CONDITION FOL OPERATION 3.2.1 The indicated AXIAL FLUX DIFFERENCE (AFD) shall be maintained within the following target band (flux difference units) about the target flux difference:

a. 15% for core average accumulated burnup of less than or equal to 3000 MWD /MTU; and

b. + 3%, -12% for core average accumulated burnup of greater than 3000 MWD /MTU.

The indicated AFD may deviate outside the above required target band at greater than or equal to 50% but less than 90% of RATED THERMAL POWER provided the indi-cated AFD is w* thin the Acceptable Operation Limits of Figure 3.2-1 and the cumu-lative penalty deviation time does not exceed 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months during the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

The indicated AFD may deviate outside the above required target band at greater than 15% but less than 50% of RATED THERMAL POWER provided the cumulative penalty deviation time does not exceed I hour during the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

APPLICABILITY: MODE 1, above 15% of RATED THERMAL POWER.*e M ACTION:

a. With the indicated AFD outside of the above required target band and with THERMAL POWER greater than or equal to 90% of RATED THERMAL POWER, within 15 minutes either:

1. Restore the indicated AFD to within the target band limits, or

2. Reduce THERMAL POWER to less than 90% of RATED THERMAL POWER.

b. With the indicated AFD outside of the above required target band for more than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months of cumulative penalty deviation time during the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or outside the Acceptable Operation Limits of l Figure 3.2-1 and with THERMAL POWER less than 90% but equal to or greater than 50% of RATED THERMAL POWER, reduce:

. 1. THERMAL POWER to less than 50% of RATED THERMAL POWER within 30 minutes, and i

See Spic;al Test Exceptions Specification 3.10.2.

- Surveillance testing of the Power Range Neutron Flux Channel may be performed pursuant to Specification 4.3.1.1 provided the indicated AFD is maintained within the Acceptable Operation Limits of Figure 3.2-1. A total of 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months operation may be accumulated with the AFD outside of the above required target band during testing without penalty deviation.

i SOUTH TEXAS - UNIT 1 3/4 2-1 L _ _ . _- .- ____ _ - - - . - - - - _ __ -_

~

ATTACHMENT 5

. ST-HL-AE #ff PAGE S OF 6 rNUUt & REVIEW COPY POWER DISTRIBUTION LIMITS O LIMITING CONDITION FOR OPERATION -

ACTION (Continued)

~

2. The Power Range Neutron Flux

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

c. With the indicated AFD outside of the above required target band for morc than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months of cumulative penalty deviation time during the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and with THERMAL POWER less than 50% but greater than 15% of RATED THERMAL POWER, the THERMAL POWER shall not be increased equal to or greater than 50% of RATED THERMAL POWER until the indicated AFD is within the above required target band.

SURVEILLANCE REOUIREMENTS 4.2.1.1 The indicated AFO shall be determined to be within its-limits during-POWER OPERATION above 15% of RATED THERMAL POWER by:

3. Monitoring the indicated AFD for each OPERABLE excore channel:

1) At least once per 7 days when the AFD Monitor Alarm is OPERABLE, and O 2) At least once per hour for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after restoring the AFD Monitor Alarm to OPERABLE status,

b. Monitoring and logging the indicated AFD for each OPERABLE excore channel at least once per hour for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and at least once per 30 minutes thereafter, when the AFD Monitor Alarm is inoperable. The logged values of the indicated AFD shall be assumed to exist during the interval preceding each logging.

l- 4.2.1.2 The indicated AFD shall be considered outside of its target band when j two or more OPERABLE excore channels are indicating the AFD to be outside the target band. Penalty deviation outside of the above required target band shall l' be accumulated on a time basis of:

! a. One minute penalty deviation for each 1 minute of POWER OPERATION outside of the target band at THERMAL POWER levels equal to or above 50% of RATED THERMAL POWER, and

b. One-half minute penalty deviation for each 1 minute of POWER OPERATION outside of the target band at THERMAL POWER levels between 15% and 50% of RATED THERMAL POWER.

l

4.2.1.3 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 Specification 4.0.4 are not applicable.

4.2.1.4 The target flux difference shall be updated at least once per O 31 Effective Full Power Days by either determining the target flux difference SOUTH TEXAS - UNIT 1 3/4 2-2

, ATTACHMENT 5 !

. ST.HL AE / fly PAG E F OF 6 YMUX & REVIEW COPY l

o -

l i

l l

l O

THIS PAGE OPEN PENDING RECEIPT OF r ii60RMATl0N FROM THE APPLICANT i

FIGURE 3.2-1 1 AXIAL FLUX DIFFERENCE LIMITS AS A FUNCTION OF RATED THERMAL POWER SOUTH TEXAS - UNIT 1 3/4 2-4

IN A RFVIFW cnpy ATTACHMENTS POWER DISTRIBUTION LIMITS HL-A y GE p 3/4.2.3 NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR V

LIMITING CONDITION FOR OPERATION N #6 3 3.2.3 Fg shall be less than 1.52[1.0+07(1-)]

THERMAL POWER Where:

P _ RATED THERMAL POWER APPLICABILITY: MODE 1.

ACTION:

With F H

exceeding its limit:

a. Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months reduce the THERMAL POWER to the level where the LIMITING CONDITION FOR OPERATION is satisfied.

b. Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER above the limit required by AClION a.,

above;THERMALPOWERmaythenbeincreased,providedFhisdemon-strated through incore mapping to be within its limit.

SURVEILLANCE REOUIREMENTS 4.2.3.1 The provisions of Specification 4.04 are not applicable.

4.2.3.2 FfH shall be demonstrated to be within its limit prior to operation above 75% RATED THERMAL POWER after each fuel loading and at least once per 31 EFPD thereafter by:

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

b. Using the measured value of F which does not include an allowance H

for measurement uncertainty.

O '

SOUTH TEXAS - UNIT 1 3/4 2-9 i

l PROOF & REVIEW COPY l

POWER DISTRIBUTION LIMITS ATTACHMENT 5

. ST-HL-AE /7f 3/4.2.5 DNB PARAMETERS PAGE 4 OF LIMITING CONDITION FOR OPERATION 3.2.5 The following DNB-related parameters shall be maintained within the limits following:

577. U

a. Reactor Coolant System T < (L 19) F ava,6f 19

b. Pressurizer Pressure, > itaisk) psig*

375,000

c. Reactor Coolant System Flow, > (IIIE!D gpm**

APPLICABILITY: MODE 1.

ACTION:

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

SURVEILLANCE REOUIREMENTS 4.2.5.1 Each of the parameters a 741- 1 2-1 shall be verified to be within its limits at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . Additionally, RCS flow shall be demon-strated within its limit prior to operation above 75% RTP after each fuel load-ing. The provisions of Specification 4.0.4 are not applicable for verification that RCS flow is within its limit.

4.2.5.2 The RCS flow rate indic.cors shall be subjected to a channel calibra-tion at least once per 18 months.

4.2.5.3 The RCS total flow rate shall be determined by precision heat balance measurements at least once per 18 months. Within 7 days prior to performing the precision heat balance flow rreasurement, the instrumentation used for per-forming the precision heat balance shall be calibrated.

THIS PAGE0 PEN PENDING RECElPT OF INFORMATION FROM THE APPLICANT

Limit not applicable during either a Thermal Power ramp in excess of 5% of RTP per minute or a Thermal Power step in excess of 10% RTP.

- Includes a 3.5% flow measurement uncertainty.

SOUTH TEXAS - UNIT 1 3/4 2-11

ATTACHMENT 4

- ST HL AE- /9/v

,_PAGE / OF #4 3/4.3 Instrumentation A. Table 3.3-1, pages 3/4 3-2 thru 3/4 3-8:

page 3-2: Added Extended Range, Neutron Flux as described in

) letter ST-HL-AE-1867. HL&P reviewed the use of 1 ACTION 5 instead of ACTION 4 for Extended Range, j Neutron Flux and found that ACTION 5 is inappropriate since opening of the Reactor Trip breakers does not apply to the subject monitors. ACTION 5 is covered by ACTIONS 4 and 10.

page 3-3: Note (1) should be added to items 13a, b and 14 to reflect that the same bistable is used for both reactor trip and ESFAS. Note (1) requires ACTION defined in ESEAS Specification 3.3.2 which is more restrictive. See also letter ST-HL-AE-1867.

page 3-4: Renumbered.

page 3-5: Renumbered.

page 3-7: ACTION 5 is not used in Table 3.3-1.

page 3-8: ACTION 11 is not used in Table 3.3-1.

B. Table 3.3-2 pages 3/4 3-9-10: Overtemperature 21 T and Overpower A T response times are 6.5 seconds as indicated in SER Supplement 2, page 4-2.

C. Table 4.3-1 pages 3/4 3-11-15: Addition of Extended Range, Neutron Flux. This was discussed during the March 9, 1987 meeting with NRC Staff; see letter ST-HL-AE-1867.

D. page 3/4 3-12: Note 19 was added to Undervoltage (UV) and Underfrequency (UF) reactor coolant pumps.

The STS requires a monthly TADOT test for the UV and UF relays. The definition of TADOT requires that setpoint accuracy be verified; this is the same as a channel ;

(relay) calibration. At South Texas, there are 8 solid state relay (UV and UF) Agastat timers per train that would have to be verified every month. We believe that this is excessive and would place the plant in a configuration that could lead to inadver-tent MODE 2 actuations. This was discussed during the March 9, 1987 meeting with the Staff.

E. page 3/4 3-13: Renumbered.

L3/NRC/cm

i ATTACHMENT 4 -

ST.HL-AEs9PV PAGE .7 OFra F. page 3/4 3-14: (1)- Table Note 5 deleted reference to manufacturer's curves. As explained in the March 9 meeting, HL&P is not provided with manufacturer's data.

(2) Table Note 9 deleted quarterly surveillance on the Boron Dilution Alarm. As agreed'in the March 9 meeting, the Boron Dilution Alarm Setpoint was added to Table Note 12.

G. page 3/4 3-15: (1) Reworded Note 12, as agreed upon during the March 9 meeting, and included the Boron Dilution Alarm setpoint from Note 5.

H. page 3/4 3-17: Justification for rewording of 4.3.2.2 was provided in letter ST-HL-AE-1901.

I. page 3/4 3-18: The logic, as shown in FSAR Figure 7.2-9, shows that Compensated Tcold-Low-Low for Safety Injection is the same as Item 5d.

This was also discussed in the March 9, 1987 meeting.

J. page 3/4 3-20: Justification for changes to 3b.1) and 5) were provided in letter ST-HL-AE-1882.

K. page 3/4 3-21: (1) ;astification for addition of MSIV closure l Logic (Items 4d, e, f) was provided in letter ST-HL-AE-1882. Note that "###" was added to Compensated Teold Low-Low (4f) for 3/4 3-18 above.

(2) Deleted Sa; this was discussed also on March 9. South Texas does not have Manual Initiation of Feedwater Isolation. Note that this was approved for Vogtle.

i L. page 3/4 3-22: Note that Sg. has been changed to delete

" coincident with Reactor Trip (P-4)".

There are 4 (1 per loop) low Tavg. signals but only 1 of these is combined with the P4 interlock on 1 train. Changed for clarity.

M. page 3/4 3-23: ACTION 19 added to 7a and 7b. Standard Tech Specs have 2 out of 4 to actuate, while South Texas has 1 out of 1, three times to actuate. This was discussed with the NRC Staff on March 9.

N. pages 3/4 3-24-25: (1) Item 9c was changed for clarity. Note that this was approved at Vogtle and discussed during the March 9 meeting.

(2) The P-15 ACTION was changed to 21 to reflect the use of Interlock ACTION statement.

L3/NRC/cm I

ATTACHMENT 4 ST HL AE NF JAGEJ_OFP.2 (3) ACTION added to 10 and 11 were provided and justified in letter ST-HL-AE-1901.

O. page 3/4 3-26: Addition of Table Note "##" and "+" were justified in letters ST-HL-AE-1882 and ST-HL-AE-1901.

P. page 3/4 3-27: ACTIONS added to reflect Items 10 and 11.

See justification in letter ST-HL-AE-1901.

Q. Table 3.3-4 pages 3/4 3-28-35: (1) Specific values provided throughout based upon W WCAP 11273. See also justification provided in letter ST-HL-AE-1897.

(2) The "*" and "***" notes were changed to show "<" to be consistent with W WCAP 11273.

R. Table 3.3-5 pages 3/4 3-34-40: (1) Specific values provided for the Table.

(2) Table notations were changed to be consistent with trip delays defined in FSAR Chapter 15.0 and Tables 15.0-4 and 15.0-5. Westinghouse has verified via letter that this is correct.

S. Table 4.3-2 pages 3/4 3-41-48: (1) Added MODE 4 to Item 1C; discussed and agreed upon during March 9 meeting.

(2) Changed Item 3c 1) to reflect MSIV closure as justified in letter ST-HL-AE-1882.

(3) See previcus comment on 3/4 3-22 for Item 5f; note that (5) was added since Tavg-Low actuation occurs on Feedwater Isolation only.

(4) Surveillance Requirements are provided for 3b 4) as justified in letter ST-HL-AE-1901.

(5) Items 4d, 4e, and 4f were justified in letter ST-HL-AE-1882.

(6) For items 8a, 8b and 8c, see page 3/4 3-12 justification for addition of note (6).

(7) Added notes (2) and (3) for Item 9d L3/NRC/cm

=4

" ATTACHMENT 6

. ST HL Al 17V4 pant 6' Of 7 M' Z. Table 4.3.5~ page 3/4 3-59: See previous comme _1."--

AA. Tables 3.3-9 and 4.3-6 pages 3/4 3-60-63: The STP design has been found acceptable by the NRC as indicated in the SER for one train of equipment being available for safe shutdown, so the minimum channels OPERABLE is given in the Table as 1,.unless there is another consideration. These changes were discussed during the March 9, 1987 meeting with NRC Staff.

1. At STP, more than one train of equipment is transferable away from the control room. Therefore, HL&P considers that ACTIONS (a) and (b) are appropriate for the transfer switches and controls as well as the monitoring instrumentation. Action (c) has been deleted.

2. The instrumentation listed in Table 3.3-9 has been verified against the information in the FSAR and the FHAR,.

resulting in the deletion of pressur-izer pressure and the change in the number of (fire-protected) channels of steamline pressure. Minimum channels needed operable are also provided.

3. The required equipment for safe shutdown is also listed, with_ indica-tion of the location of transfer switches and controls, total number available at STP and the minimum needed for safe shutdown from outside-the control room.

4. The channel calibration requirement for the extended range neutron flux instruments has been changed from-monthly to once per 18 months, to be consistent with the requirement for these same instruments in the Reactor Trip Instrumentation specification and the accident monitoring specification.

BB. Tables 3.3-10 and 4.3-7, pages 3/4 3-64-68 1. Because of the various ACTIONS required, depending upon the total number of channels monitoring the same parameter, the ACTION statements for each channel have been transferred to the table. Places where tLo actions L3/NRC/cm

ATTACHMENT l

. ST HL AE 1789 ,

PAGE4 0F E3-

'for the P-15 permissive. A Channel Calibration'is required to be per-formed on an,18 month basis and an ANALOG CHANNEL OPERATIONAL TEST to be performed on a monthly basis. Due to the design,-this would require lifting of leads and installing jumpers. In

.an effort to minimize lifted leads and jumpers, relaxed surveillances 'are proposed. The proposed surveillances will require' normal testing when it-can be done without lifting leads and

' installing jumpers, and also provides for periodic verification of the functional status of the bistable by visual means. The surveillance is consistent with that allowed for the P-10 permissive in the Reactor Trip Instrumentation, and is considered logically equivalent to NOT (P-10).

(8) Item 10d and 11d change provided in letter ST-HL-AE-1882.

(9) Table Note (1) changed to 92 days to be consistent with ESEAS Response Time Testing.

(10) See Item applicability for remainder of notes for justification.

T.. Table 3.3-6, pages 3/4 3-50-51: Changes provided in this Table were provided in ST-HL-AE-1901. A separate letter will be provided which requests deletion of criticality monitors as granted under 10CFR70.24 for the SNM-1972 license.

U. Table 4.3-3 page 3/4 3-52: See previous comments.

V. 4.3.3.2 page 3/4 3-53: Editorial.

W. Table 3.3-7 page 3/4 3-55: (1) Specific locations provided.

(2) OBE Setting provided for Items 4 and 5 based on the March 12, 1987 teleconference with NRC Staff.

(3) Added note to Item 6 on Response Spectrum Analyzer per NRC request to indicate where accelerometer data is gathered and analyzed.

't X. Table 4.3-4 page 3/4 3-56: Made consistent with Table 3.3-7.

Y. Table 3.3-8 page 3/4 3-58: Change feet to meters to be consistent with South Texas procedures.

=

A

ATTACHMENT 4

"g ' '

iST HL AE- Mr </

- r PAGE (, OF PA t - -

required differ from the standard are H noted below..

2. When 4 channels'are provided and'l is minimum operable, the first channel-inoperable requires. restoration within.

3 months and the second channel inoperable-requires restoration within-31 days. (See alsolitem-6 below.)

3. - When 3 channels are provided and l'is minimum operable, if only 2 are operable,' the third channel must be

. . restored operable within 31 days.

.(See also item 6 below.)

~

4. Based upon the STP Emergency Operating-

. Procedures, the radiation monitors listed in Table 3.3-10 are Type A variables needed for accident _diagno-sis and must be available after an accident. ACTIONS S and 6 reflect the need for these monitors, and provide appropriate time to repair the equip-ment.

5. Action requirements for the reactor 1 . vessel water level instrumentation are shown as ACTION 7. These actions are considered appropriate based upon the need for reactor vessel water level monitoring and the equipment provided,

. parts of which (the probes) are mounted on the head and extend into the upper head region of the reactor vessel.' These actions have been previously used at the Palo Verde units.

6. At STP, the Qualified Display Processing System (QDPS) is used to display the Regulatory Guide (R.G.)

1.97 Category 1 parameters (on a fully qualified basis). The QDPS checks the validity of each sensor's input to it, and calculates a group value using a redundant sensor algorithm (RSA).

That algorithm provides a group value and a quality tag. If all the inputs were found valid and within an appro-priate range of each other, the quality tag is good. If any input is bad, then the quality tag is poor.

The action statements reflect that action to restore all inputs to operable status is appropriate L3/NRC/cm

5 ATTACHMENT 4,

- ST-HLeAE- /9F y PAGE 7 0F PJL whenever the RSA returns a poor quality tag. Quicker actions to restore. sensors to-operable status are then developed as fewer sensors are available. - Action within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months is required when less than one sensor is operable, i.e., when the RSA assigns a.

- quality tag of bad.

7. For the core exit temperature sensors, p the first action required is based j; upon the NUREG-0737 requirement that the primary and backup displays use a minimum of 4 thermocouples per quad-rant. This has been interpreted on L STP as being 4 thermocouples per f quadrant on each train.

I The RSA for core exit temperature assigns a good quality tag when at least 6 thermocouples per quadrant ,

have valid QDPS inputs. Thus, in !

keeping with the philosophy in item 6 above, STP will require action to restore thermocouples to operable status when less than 6 thennocouples per quadrant have valid inputs to QDPS. When less than 4 thermocouples per quadrant are valid inputs, the I l

quality tag assigned is bad. This number is then used as the ninimum I channels operable, with action re- .

quired within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

CC. 3/4 3-70, 71: (1) Deletion of Fire Protection is described in the justification provided via letter ST-HL-AE-1867.

(2) Table 3.3-11 is deleted;~see previous comment.

DD. 4.3.3.10, page 3/4 3-73: Justification provided in letter ST-HL-AE-1901.

EE. Table 3.3-12, page 3/4 3-74: Justification provided ir letter ST-HL-AE-1901.

FF. Table 4.3-8, pages 3/4 3-76, 77: Justification provided in letter ST-HL-AE-1901.

GG. Table 3.3-13, page 3/4 3-79: Justification provided in letter ST-HL-AE-1901.

HH. Table 4.3-9, pages 3/4 3-81, 82: Justification provided in letter ST-HL-AE-1901.

L3/NRC/cm

O

O n v

TABLE 3.3-1

$ REACTOR TRIP SYSTEM INSTRUMENTATION

$ MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE Q .

OF CHANNELS TO TRIP OPERABLE MODES ACTION B'; FUNCTIONAL UNIT i

2 1, 2 1 l

1. Manual Reactor Trip 2 1 2 1 2 3*, 4*, 5* 10 5

-e

" 2. Power Range, Neutron Flux l 2 3 1, 2 2#

a. High Setpoint 4 1###, 2 2#

Low Setpoint 4 2 3 b.

3. Power Range, Neutron Flux 4 2 3 1,'2 2#

High Positive Rate 4 2 3 1, 2 2#

Power Range, Neutron Flux,

4. I R High Negative Rate i Intermediate Range, Neutron Flux 2 1 2 1###, 2 3 V 5.

m

6. Source Range, Neutron Flux 2 1 2 2## 4 l a. Startup 1, i , 5 -

l in Shutseen

9: 7 3*, 4*, 5*

3-10 2 1 2 b.r. Shutdown o 3 , y, g 4/

'l Edewbl %e., A!aEron M 2. .t 4 2 3 1, 2 6#

FJ. Overtemperiture AT 4 2 3 1, 2 6#

Overpower AT 7 f.

2 3 1 6#

3$h 3 Pressurizer Pressure--Low 4 10 4 @g>S ood- @'p 1, 2 6# r s Pressurizer Pressure--High 4 2 3 11 Jd. % s,nE g2 A 21. Pressurizer Water Level--High 4 2 3 1 6# %7 Q w g 8

a

s TABLE 3.3-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION

_., MINIMUM i

0 . TOTAL NO. CHANNELS CHANNELS APPLICABLE

! R FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE M0ES - ACTION i 3 c- If. Reactor Coolant Flow--Low 5

a. Single Loop (Above P-8) 3/ loop 2/ loop in 2/ loop in 1 6# (1) any oper- each oper-ating loop ating loop

b. Two Loops (Above P-7 and 3/ loop 2/ loop in 2/ loop 1 6# 0) below P-8). two oper- each oper--

ating loops ating loop 1/. Steam Generator Water 4/sta. gen. 2/sta. gen. 3/sta. gen. 1, 2 6# f1)

Level--Low-Low in any oper- each oper-R ating stm. ating sta.

gen. gen.

If.Undervoltage--ReactorCoolant 4-1/ bus 2 3 1 6# N Pumps to If. Underfrequency--Reactor Coolant Pumps 4-1/ bus 2 3 1 6#

1. Turbine Trip I a. Low Emergency Trip. Fluid l Pressure 3 2 2 1 6#

$$h

! b. Turbine Stop Valve Closure 4 2 3 1 6# .

4>@ f,i i

o"R

, z o

4 N H RO j .b t o m Q

1 3o i

O O O

! TABLE 3.3-1 (Continued)

$ REACTOR TRIP SYSTEM INSTRUMENTATION

$ MINIMUM y

g TOTAL NO. CHANNELS CHANNELS APPLICABLE R FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION d 1. Safety Injection Input izi from ESFAS 2 1 2 1, 2 9

-A H 1. Reactor Trip System Interlocks ,

a. Intermediate Range Neutron Flux, P-6 2 1 2 2## 8

b. Low Power Reactor l Trips Block, P-7 l P-10 Input 4 2 3 1 8 or l R P-13 Input 2 1 2 1 8 ,

=

y c. Power Range Neutron

Flux, P-8 4 2 3 1 8

d. Power Range Neutron 4 2 3 1 8 Flux, P-9

e. Power Range Neutron Flux, P-10 4 2 3 1,2 8

f. Turbine Impulse Chamber Pressure, P-13 2 1 2 1 8 :

i $}>

2039' Reactor Trip Breakers 2 1 2 2

1, 2 3*, 4*, 5*

9, 12 10 m?

[Q r

2 1 g g

& Nz

' DJ -*

No a

a

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l, e Jk_

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= ek J

O N O

I 90 T 1 C

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LM 13 P

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T A

T

) N d E S e

u M LP U EI n R NR i T NT 11 t S A n N HO o I CT C

M O (1 3

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S Y

S S 3 .L P OE E I NN L R N B T LA 22 A AH T R TC O O T TF C O A

E R

k c

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T I c N i U t a

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^

ATTACHMENT ST-HL-A /W pans. QFid.._.

-. = m.vrtw UUPy 1

TABLE 3.3-1 (Continued)

ACTION STATEMENTS (Continued)

ACTION 3 - With the number of channels OPERABLE one less than the Minimum Channels OPERABLE requirement and with the THERMAL POWER level:

l a. Below the P-6-(Intermediate Range Neutron Flux Interlock) l Setpoint, restore the inoperable channel to OPERABLE !

status prior to increasing THERMAL POWER above the P-6 '

Setpoint, and

b. Above the P-6 (Intermediate Range Neutron Flux Interlock)

Setpoint but below 10% of RATED THERMAL POWER, restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above 10% of RATED THERMAL POWER.

ACTION 4 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, suspend all operations involving positive reactivity changes.

ACTION 5 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel NoTdsQ to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or open the Reactor Trip System breakers, suspend all operations involving positive reactivity changes and verify Valves are closed and secured in position within the newt ha v ACTION 6 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:

a. The inoperable channel is placed in the tripped condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and

b. The Minimum Channels OPERABLE requirement is met; however, the inoperable channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months

,, for surveillance testing of other channels per Specification 4.3.1.1.

ACTION 7 - (Not Used)

ACTION 8 - With less than the Minimum Number of Channels OPERABLE, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months determine by observation of the associated permissive annunciator window (s) that the interlock is in its required state for the existing plant condition, or apply Specification 3.0.3.

ACTION 9 - With the number of OPERABLE channels one less than the Minimum Channels OPERA 8LE requirement, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; however, one channel may be bypassed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing per Specification 4.3.1.1, provided the other channel is OPERABLE.

SOUTH TEXAS - UNIT 1 3/4 3-7

mar ee o e m,mu _,

T a "Ad M M*Z " '

ST HL-AE / 97 q TABLE 3.3-1 (Continued) PAGE /SOF F.2 gO ACTION STATEMENTS (Continued)

ACTION 10 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or open the Reactor Trip

. System breakers within the next hour.

ACTION 11 "With the number of OPERABLE channels less than the Total Num er

of Channels, operation may continue provided the inoperable i u;.hannels are placed in the tripped condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ACTION 12 - With one of the diverse trip features (undervoltage or shunt trip attachment) inoperable, restore it to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or declare the breaker inoperable and apply ACTION 9. The breaker shall not be bypassed while one of the diverse trip features is inoperable except for the time required for performing maintenance to restore the breaker to OPERABLE status.

O s

i e #

O SOUTH TEXAS - UNIT 1 3/4 3-8

O O O

! TABLE 3.3-2

$ REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES N FUNCTIONAL UNIT RESPONSE TIME B; ,

. 1. Manual Reactor Trip N.A.

2. Power Range, Neutron Flux 5 0.5 second*

3. Power Range, Neutron Flux, High Positive Rate N.A.

4. Power Range, Neutron Flux, High Negative Rate 5 0.5 second*

5. Intermediate Range, Neutron Flux N.A.

R

6. Source Range, Neutron Flux N.A.

7. Ate nded n. 4.

Y PJ. Overtemper%e.,

ature AT rJM Ex. <#vgeconds*

w 9 A. Overpower AT 5 /'3seconds

10 3. Pressurizer Pressure--Low < 2 seconds il 10. Pressurizer Pressure--High 5 2 seconds 1231. Pressurizer Water Level--High Itat, S asewnd s IN

~e.m ,

Neutron detectors are exempt from response time testing. Response time of the neutron flux signal portion % E of the channel shall be measured from detector output or input of first electronic component in channel. o 5, a

a

O '

O O.

TABLE 3.3-2 (Continued)

$ REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES 52 N FUNCTIONAL UNIT RESPONSE TIME 5; 3,

. 17 Rea: tor Coolant Flow--Low C

5

a. Single Loop (Above P-8) < 1 second

b. Two Loops (Above P-7 and below P-8) 7 1 second w .q -

lj. Steam Generator Water Level--Low-Low < 2 seconds S

If. Undervoltage - Reactor Coolant Pumps 5 1.5 seconds

'a lji. Underfrequency - Reactor Coolant Pumps 5 0.6 second 9

w if. Turbine Trip k

w c. Low Emergency Trip Fluid Pressure N.A.

.'. b. Turbine Stop Valve Closure N.A.

f 1/. Safety Injection Input from ESF N.A.

9 IJ. Reactor Trip System Interlocks N.A.

30 J57. Reactor Trip Breakers N.A.

l.

2p. Automatic Trip and Interlock Logic N.A.

s41 i

%s%-

r ps o c!

2!

3 p n.

(O O NJ TABLE 4.3-1

$ REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS Y

y TRIP R ANALOG ACTUATING MODES FOR

$; CHANNEL DEVICE WHICH

, CHANNEL CHANNEL OPERATI0lAL OPERATIONAL ACTUATION SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST IS REQUIRED g

] 1.

2.

Manual Reactor Trip Power Range, Neutron Flux N.A. N.A. N.A. R O'h N.A. 1, 2, 3*, 4*, 5*

a. High Setpoint S D(2, 4), Q(17) N.A. N.A. 1, 2 M(3, 4),

Q(4, 6),

R(4, 5)

b. Low Setpoint S R(4) S/U(1) N.A. N.A. 1***, 2

$ 3. Power Range, Neutron Flux, N.A. R(4) Q(17) N.A. N.A. 1, 2 w High Positive Rate O

4. Power Range, Neutron Flux, N.A. R(4) Q(17) N.A. N.A. 1, 2 High Negative Rate

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

6. Source Range, Neutron Flux S R(4, 5, ljk) S/U(1),Q(9)(17) N.A. N.A. 2**, 3, 4, 5 1 EsteMdflagM A pu s e.Cy) &o a)(gj) g. 6, N. A, 3, 9. 5 Overtemperature AT S R Q(17) N.A. N.A. 1, 2 77 7

p'. Overpower AT S R Q(17) N.A. N.A. 1, 2 $a$$

C' g

g -4 jo

p. Pressurizer Pressure--Low S R Q(17) N.A. N.A. 1 S '" C o

_L 9' UW ID. Pressurizer Pressure--High Q(17) N.A. N.A. 1, 2 uO m If. Pressurizer Water Level--High S

S R

R Q(17) N.A. N.A. 1 Q

D{

P* n

?

k If.ReactorCoolantFlow--Low S R Q(17flg) N.A. N.A. 1 g A

ex t) b O

, TABLE 4.3-1 (Continued)

$ REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS 5$

y TRIP R ANALOG ACTUATING MODES FOR

$; CHANNEL DEVICE WHICH

, CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE c- FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST IS REQUIRE 0

. Steane Generator Water Level-- S R Q(17,18) N.A. N.A. 1, 2

Low-Low I

Ig. Undervoltage - Reactor Coolant N.A. R N.A. Q(17,)1) N.A. 1 Pumps

1. Underfrequency - Reactor N.A. R N.A. Q(1719) N.A. 1 Coolant Pumps sa 1

} If. Turbine Trip Y a. Low Fluid Oil Pressure N.A. R N.A. S/U(1,10) N.A. 1 U b. Turbine Stop Valve N.A. RN. A. N.A. S/U(1, 10) N.A. 1 Closure 1

1/. Safety Injection Input from N.A. N.A. N.A. R N.A. 1, 2 ESF

1. Reactor Trip System Interlocks

a. Intermediate Range Neutron Flux, P-6 N.A. R(4) R N.A. N.A. 2**

b. Low Power Reactor Trips Block, P-7 N.A. R(4) R N.A. N.A. 1 MN g

c. Power Range Neutron 5 Flux, P-8 N.A. R(4) R N.A. N.A. 1 N> IM*

d. Power Range Neutron M Flux, P-9 N.A. R(4) R N.A. N.A. 1 M'z r1 jov 0 3:

b 1

~

(. ,

\ v v

' TABLE 4.3-1 (Continued)

$ REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIRENENTS TRIP y

g ANALOG ACTUATING MODES FOR CHANNEL DEVICE WHICH R OPERATIONAL ACTUATION SURVEILLANCE CHANNEL CHANNEL OPERATIONAL c FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST IS REQUIRED 1h.ReactorTripSystemInterlocks(Continued)

e. Power Range Neutron Flux, P-10 N.A. R(4) R N.A. N.A. 1, 2

f. Turbine Impulse Chamber Pressure, P-13 N.A. R R N.A. N.A. 1 70 g Reactor Trip Breaker N.A. N.A. N.A. M(7, 11) N.A. 1, 2, 3*, 4*, 5*

m D '

m 29 Automatic Trip and Interlock N.A. N.A. N.A. N.A. M(7) 1, 2, 3* , 4* , 5*

J. Logic w v Reactor Trip Bypass Breaker N.A. N.A. N.A. M(15),R(16) N.A. 1, 2, 3*, 4*, 5*

2/.

M$N e 0%t ,i,,

rn s k%-1 5

8 a

ATTACHMENT 4

. ST.HL AE /9f/ ;

P E /f_ OF P R rnwr a NtVIEW COPY !

TABLE 4.3-1 (Continued)

( TABLE NOTATIONS

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

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

- - Below P-10 (Low Setpoint Power Range Neutron Flux Interlock) Setpoint.

(1) If not performed in previous 31 days.

(2) Comparison of calorimetric to excore power indication above 15% of RATED THERMAL POWER. Adjust excore channel gains consistent with calorimetric power if absolute difference is greater than 2%. The provisions of Specification 4.0.4 are not applicable to entry into MODE 2 or 1.

(3) Single point comparison of incore to excore AXIAL FLUX DIFFERENCE above 15% of RATED THERMAL POWER. Recalibrate if the absolute difference is greater than or equal to 3%. The provisions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1.

(4) Neutron detectors may be excluded from CHANNEL CALIBRATION.

(5) Detector plateau curves shall be obtained, and evaluated and r preu Q

V tn == = '=M :: '+ date:. For the Intermediate Range and Power Range Neutron Flux channels the provisions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1.

(6) Incore - Excore Calibrationgf&bove 78% at RATES THEI#tAL Pottit. The provisions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1.

(7) Each train shall be tested at least every 62 days on a STAGGERED TEST BASIS.

,,(8) (Not Used) [

(9) 4*, and 5* shall also include Quarterly verificationsurveillance that permissives in MODES P-6 3[*

P-10 are in their required state for existing y observation of the permissive ann _unciator windo fw. plantsurveillance Quarterly conditionsshall include verification gr tne Boron D1lution Alarm Setpoint of less than or equal to (an Vncrease of twice the count rate within a 10-minute period).

O SOUTH TEXAS - UNIT 1 3/4 3-14

ATTACHMENT 4 ST.HL AE /97Y FAGEaoOF ra.

PROOF & REVIEW COPY TABLE 4.3-1 (Continued)

TABLE NOTATIONS (Continued)

(10) Setpoint verification is not applicable.

(11) The TRIP ACTUATING DEVICE OPERATIONAL TEST shall independently verify the OPERABILITY of the undervoltage and shunt trip attachments of the Reactor Trip Breakers.

(12)'At least once per 18 months during shutdown, verify that on a simulate Boron Dilution Doubling test signal the normal CVCS discharge valves '

close and the centrifugal charging pumps suction valves from the RWST ope within [30] secondsj se ( n s g g a m (13) (Not used)

(14) The TRIP ACTUATING DEVICE OPERATIONAL TEST shall independently verify the OPERABILITY of the undervoltage and shunt trip circuits for the Manual Reactor Trip Function. The test shall also verify the OPERABILITY of the Bypass Breaker trip circuit (s).

(15) Local manual shunt trip prior to placing breaker in service.

(16) Automatic undervoltage trip.

(17) Each channel shall be tested at least every 92 days on a STAGGERED TEST BASIS.

(18) The surveillance frequency and/or MODES specified for these channels in Table 4.3-2 are more restrictive and, therefore, applicable.

69 VeMeden -nd al j ushert e44tipskobt 6 **"["ed-

,,L3ert: ,y A gego j TL

'"? OPE 2461LiTY 6.ron%

" d 'ccuracie s , Mon .

&A W be, yg,ge,}

(b,,e,(Mpinh du., eJehei

'" b te b c-O SOUTH TEXAS - UNIT 1 3/4 3-15

ATTACHMENT 4 ST.HL.AE. /9N M k EU COIY INSTRUMENTATION

(')

SURVEILLANCE REQUIREMENTS 4.3.2.1 Each ESFAS instrumentation channel and interlock and the automatic actuation logic and relays shall be demonstrated OPERABLE by performance of the ESFAS Instrumentation Surveillance Requirements specified in Table 4.3-2.

4.3.2.2 The ENGINEERED SAFETY FEATURES RESPONSE TIME of each ESFAS function shall be demonstrated to be within the limit at least once per 18 months._

Each test shall include at least one train so thatSall trains are tested at Gast once per 36 months and one channel per function so that all channels are tested at least once per N times 18 months where N is the total number J of redundant channels in a specific ESFAS function as shown in the " Tot fal (No. of Channels" column of Table 3.3-3. f Pr N* *I #"

8ach%

each \oka kedh 4isYeSitd km ts %.%A 6st once pr E54\eask mm%s *"CC tand c')*

c_6ne\ pt bc%s so Ld:- J c6nc\s *ce. MNeh. ok \ cat. e%

P'" 4 he s 19 monM ukere N A % htd numbe.v ef rebndad c6nels in a. speedic. estas bhen as rhown in %. wnta uo.of

""t h U c.olu.mn ok Yade.1.~3-3, O

O SOUTH TEXAS - UNIT 1 3/4 3-17

O O O

! TABLE 3.3-3

$ ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION R MINIMJM

$; TOTAL NO. CHANNELS CHANNELS APPLICABLE

, FUNCTIONAL UNIT ,

OF CHANNELS TO TRIP OPERABLE MODES ACTION b

1. Safety Injection (Reactor Trip, Feedwater Isolation,

Control Room Emergency ventilation, Start Standby Diesel Generators, Reactor Containment Cooling Fans, and Essential Cooling Water).

a. Manual Initiation 2 1 2 1,2,3,4 19 w

1 b. Automatic Actuation w Logic 2 1 2 1,2,3,4 14

" c. Actuation Relays 3 2 3 1,2,3,4 14

d. Containment 3 2 2 1,2,3,4 15*

Pressure--High-1

e. Pressurizer 4 2 3 1, 2, 3# 20*

Pressure--Low

f. Compensated Steam 3/ steam line 2/ steam line 2/ steam line 1, 2, 3# 15* . , . ,

Line Pressure-Low any steam line .,

D

g. Compensated T Low-Low COLD 3/ loop 2/ loop any loop 2/ loop 1, 2, 3# 15* j

! g y

E ni 55 P

e M:

~i

i

'v) i (

v)

TABLE 3.3-3 (Continued) m 8 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION 5

h MINIMUM g TOTAL NO. CHANNELS CHANNELS APPLICABLE

, FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION

3. Containment Isolation (Continued)

H b. Containment Ventilation s.( M cedainnJI s.g aul mmud bse[6 p "ise\dden. te iternords' .

1) Manual Initiation .2 . *- ~# i,7, 4 4 2

2) Automatic Actuation Logic 2 1 2 1,2,3,4 18

$ 3) Actuation RelaysM 3 2 3 1,2,3,4 18

[ 4) SafetyInjection44 See Item 1. above for all Safety Injection initiating functions and o requireiaents.

5) RCB Purge ##< W Radioactivity-High 2 1 2 i / 4, a 18

4. Steam Line Isolation

a. Manual Initiation

1) Individual 2/ steam line 1/ steam line 2/ operating 1, 2, 3 24 steam line

2) System 2 1 2 1,2,3 23 8%59

'1

r." E:

b. Automatic Actuation 2 1 2 1,2,3 22 9' E Logic and Actuation =

Relays m

s }>

a a

O O O

! TABLE 3.3-3 (Continued)

$ ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION 5

-e Q MINIMUM g TOTAL NO. CHANNELS CHANNELS APPLICABLE

, FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION

4. Steam Line Isolation (Continued)

H

c. Steam Line Pressure -

Negative Rate--High 3/ steam line 2/ steam line 2/ steam line 3### 15*

d- 4% M N mte.-sg 2, 3 any steam

,, 3, 3 ,g 4 C-hw Line Preme._ R 1ine 2

4. Gr e%Ted h.g% 3/sh a En a htcamline'm aq a/ct- /s 4

5. Turbine Trip and t,

3 /baf 5%"be ed hbei Ae. I,2,3%

m Feedwater Isolation 16 ,

A liap sh cey loop 2/f.epcn cA y 1 32.34

$ r Mannat In*tiation "F 7 -lt

[ a ,tr. Automatic Actuation 2 1 2 1,2,3 25

- Logic and Actuation Relays b y. Steam Generator 4/sta. gen. 2/sta. gen. 3/sta. gen. 1, 2, 3 20*

Water Level-- in any oper- in each High-High (P-14) ating sta. gen. operating stm. gen.

c_ Af. Compensated T ~

cold Low 3/ loop 2/ loop in 2/ loop in 1####, 2, 3# @15*R 'om>

any loop each loop y}

d A. Feedwater Flow - High 3/sta. gen. 2/sta. gen. 2/sta. gen. 1####, 2, 3 15* !MS coincident with either 4 h in any stm. in each sta. h

, oM following in 2 of 4 loops: gen. gen. p 'g$

- Uwss t

I 8

-7

O ;

O O TABLE 3.3-3 (Continued)

$ ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION h MINIMUM M TOTAL NO. CHANNELS CHANNELS APPLICABLE

, FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION 5 5. Turbine Trip and Feedwater Isolation (Continued)

]

Reactor Coolant Flow-Low 3/ loop 2/ loop in any 2/ loop in 1####, 2, 3 d 15*

loop each loop or T,yg - Low 1/ loop 1/ loop in any 1/ loop 1####, 2, 15*

m loop 1

m f. Safety Injection See Item 1. for all Safety Injection initiating a functions and requirements.

~

x1

g. T,,g-Low h h f a Reactor Tiip-(W" 4(1/ loop)f 2 3 1, 2, 3i 20* i

6. Auxiliary Feedwater Manual Initiation

a. 1/ pump 1/ pump 1/ pump 1, 2, 3 26

b. Automatic Actuation Logic 2 1 2 1,2,3 22

c. Actuation Relays 3 1,2,3 3 2 22 -

gg

d. Sta. Gen. Water Level-- g r., g

, ob Low-Low g E Start Motor- 7>' I 2 pi E Driven Pumps 4/sta. gen. 2/sta. gen. 3/sta. gen. 1, 2, 3 20* rr y D and Turbine-Driven Pump in any stm.

gen.

in each stm. gen.

% b**

E

e. Safety Injection See Item 1. above for all Safe'ty Injection initiating functions and 8o requirements.

~

TABLE 3.3-3 (Continued) g ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION 5

x w MINIMlM R TOTAL NO. CHANNELS CHANNELS APPLICA8LE 3; FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE M00ES ACTION e 7. Automatic Switchover to Containment Sump ****

{

a. Automatic Actuation Logic and Actuation 3 3 3 1, 2, 3, 4 W6 Relays

b. RWST Level--Low-Low 3-1/ pump 1/ pump 1/ pump 1, 2, 3, 4 4713 Coincident With:

w Safety Injection See Ites 1. above for all Safety Injection initiating functions 1 and requirements.

w 8. Loss of Power

a. 4.16 kV ESF Bus Under- 4/ bus 2/ bus 3/ bus 1, 2, 3, 4 20*

voltage-Loss of Voltage

b. 4.16 kV ESF Bus Under-voltage-Tolerable Degraded Voltage Coincident with SI 4/ bus 2/ bus 3/ bus 1, 2, 3, 4 20*

c. 4.16 kV ESF Bus Under- ,

voltage - Sustained Degraded Voltage 4/ bus 2/ bus 3/ bus 1, 2, 3, 4 20* ; mg>h

t ro

9. Engineered Safety Features j hj Actuation System Interlocks

g N 4

a. Pressurizer Pressure, P-11 3 2 2 1,2,3 21 j p.g 1

rI

b. 4 2 3 1,2,3 21 Low-Low T,yg, P-12 f N

f) O ~h V V (C TABLE 3.3-3 (Continued)

$ ENGINEERED SAFETY FEATURES ACTUATION SYS1EM INSTRtmENTATION 5

O NININUN R TOTAL NO. CHANNELS CHANNELS APPLICABLE

, FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE NODES ACTION

$ 9. Engineered Safety Features Actuation System Interlocks (Continued)

]

c. Reactor Trip, P-4 4i2/ train 51 2 1,2,3 23

d. Power Range Neutron 4 2 3 1,2,3 % ~J 1 Flux Input to Excessive Cooldown Protection, P-15 m 10. Control Room Ventilation k

w a. Manual Initiation 3(1/ train) 2(1/ train) 3(1/ train) All d O

b. Safety Injection See Item 1. above for all Safety Injection initiating ftx tions and requirements.

c. Automatic Actuation Log ~s.

. 3 2 3 All 21

.4 Actuation Relays

d. Control Room Intake Air 2 1 2 Radioactivity - High All 294

11. FHB HVAC

a. Manual Initiation 3(1/ train) 2(1/ train) 3(1/ train) M 1, 2, 3, 4 or with irradiated 2} E I

fuel in spent 5 fuel pool %

k$

U t w

! i!

n

P

O O O

! TA8LE 3.3-3 (Continued)

$ ENGINEERED SAFETY FEATURES ACTUATION SYSTEN INSTRUNENTATION h NINIltM 3 TOTAL NO. CHANNELS CHANNELS APPLICABLE

, FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE NDDES ACTION

11. FHB HVAC (Continued)

" b. Automatic Actuation 3 2 3 1, 2, 3, 4 or Logic and Actuation with irradiated 2*)

Relays fuel in spent fuel pool

c. Safety Injection See Ites 1. above for all Safety Injection initiating functions and requirements.

$ d. Spent Fuel Pool Exhaust Radioactivity - High 2 1 2 With irradiated 30+

m fuel in spent 4 fuel pool w

NA E45

As 24" 8

A

ATTACHMENT 4

. ST HL AE.Bt y PAGE.7G F .P2 PROOF & REVIEW COPY TABLE 3.3-3 (Continued)

O) t TABLE NOTATIONS

The provisions of Specification 3.0.4 are not applicable. ,,

- Feedwater Isolation only.

- - Function is actuated by either actuation train A or actuation train B'.'

Actuation train C is not used for this function.

- - - Automatic switchover to containment sump is accomplished for each train using the corresponding RWST level transmitter.

Trip function may be blocked in this MODE below the P-11 (Pressurizer Pressure Interlock) Setpoint.

M Ar% cone m:reuves or encument d irradideL\ u!& Con % mere

1. 1. Trip fuhetion automatically blocked above P-11 and may be blocked below P-11whenSufgigrg gg-M. ,.. ..._. . is not blocked.

1. 1. 1. Trip function is blocked in MODE 1 above the P-15 (Excessive Cooldown Protection) setpoint.

+ % probbens 09 Sf*h 3 ms J.6 3 = 10 7 are. not appIcad, ACTION STA'ij,iNTS ACTION 14 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, be in at least HOT STANDBY O within 6 5c'i~ and in COL. SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months ; lo ,ver, one channel may be bypassed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for survisilla 1 testing per Specification 4.3.2.1, provided the other che al is OP;RABLE.

ACTION 15 - With the nuraer of OPERABLE channels one less than the Total Number of Channels, operation may proceed until performance of the next required ANALOG CHANNEL OPERATIONAL TEST provided the inoperable channel is placed in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

,, ACTION 16 - (Not Used)

ACTION 17 - With the number of OPERABLE channels one less than the Total Number of Channels, operation may proceed provided the inoperable channel is placed in the bypassed condition and the Minimum Channels OPERABLE requirement is met. One additional channel may be bypassed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing per Specification 4.3.2.1.

ACTION 18 - With less than the Minimum Channels OPERABLE requirement, operation may continue provided the containment purge supply and exhaust valves are maintained closed.

ACTION 19 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel

(\ to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTOOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SOUTH TEXAS - UNIT 1 3/4 3-26

ms

. ST HL< AE, ME V n nouss.a. t

- - - ' - - - ~ - '

. ST.HL.AE /4c/

PAGE3o0F PA PAGE f 0F /4 ACTIO!! 27: MCDES 1,2,3,4: With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in at least BOT STAllDEY 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 .

MODES 5 and 6: With the number of OPERABLE channels less than the Minimum Channels CPERABLE requirement, restore the '

inoperable Channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or initiate and maintain operation of the Control Room Envelope .

Ventilation System (ac 100E capacity) in the recirculation and makeup filtration mode.

ACTIoll 28: MCDES 1,2,3,4: With the nucher of OPERABLE channels less than the Minimum Channels OPERABLE requirement, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months isolate the Control Room Envelope and maintain operation of the ventilation system in the filtered recirculation mode.

MCDES 5 and 6: With the number of OPERABLE channels less than the Minimum Channels OPERABLE requirement, within I hour initiate and maintain operation of the Control Room Envelope Ventilation System (at 100E capacity) in the recirculation and makeup filtration mode.

ACTIO!! 29: MCDES 1,2,3,4: With the nu=ber of CPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to CPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or either ,

initiate and maintain operation of the FHB exhaust air filtration system (at 100" capacity) or be in at least HOT STA!iDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWII within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

With irradiated fuel in the spent fuel pools With the number of OPERABLE channels less than the Minimum Channels OPERABLE requirement, fuel movement within the spent fuel pool or crane operation with loads over the spent fuel pool may proceed provided the FHB exhaust air filtration system is in operation and discharging through at least one train of IIEPA filters and charcoal adsorbers.

ACTI0li 30: With irradiated fuel in the spent fuel pool With the number of OPERABLE channels less than the Minimum Channels OPERABLE requirement, fuel movement within the spent fuel pool or crane operation with loads over the apent fuel pool may proceed provided the FHB exhaust air filtration system is in operation and discharging through at least one train of itEPA filters and charcoal adsorbers.

. t SoM TC%h M b A

O O O TABLE 3.3-4

$ ENGINEERED SAFETY FEATURES ACTUATION SYSTEN INSTRUENTATION TRIP SETPOINTS iA

--e R SENSOR R TOTAL ERROR

, FUNCTIONAL UNIT ALLOWANCE (TA) Z (S) TRIP SETPOINT ALLOWABLE VALUE

1. Safety Injection (Reactor Trip, Feedwater Isolation, Control Room Emergency Ventilation, Start Stan6y Diesel Generators, Reactor Containment Cooling Fans, and Essential Cooling Water)

a. Manual Initiation N.A. N.A. N.A. N.A. N.A.

m b. Automatic Actuation Logic N.A. M.A. M.A. M.A. N.A. .

2 s~

w :. Actuation Relays N.A. N.A. N.A. M. A. N.A.

.d L

60 2, 3 4. 0

~

e.E

d. Containment Pressure--High 1 3.75 0.71 A.0 $ 4.0 psig 5 X.8 psig t' g$

a.0 n o*g=

e. Pressurizer Pressure-Low 13.1 10.71 h5 1 1850 psig 1 1839 psig Q ',$

f. Compensated Steam Line 13.6 is rig IW 2.o ES 1 735 psig 9; y,9 1 @ .F psig*

yh Pressure-Low

v. S I. O 528

g. Compensated T 1 COW -Low-L w 0.5 M 1 532*F 153k4*F

2. Containment Spray

a. Manual Initiation N.A. N.A. N.A. N.A. N.A. @

o

b. Automatic Actuation Logic M.A. N.A. N.A. M.A. N.A. %

Qo

c. Actuation Relays M.A. M.A. N.A. N.A. N.A. m

19. 5 2.S 0

Q

d. Containment Pressure--High-3 Itf13. /, 0.71 in 2.0 $ [ " M3 psig i [ m n ] psig -

TH S :).$i Pi1 PENDING RECEIPT OF 8 j in.

. .,i .10M THE A?AiCANT 3

r f' O(~~h TABLE 3.3-4 (Continued)

E C

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUNENTATION TRI

~

I SENSOR g FUNCTIONAL UNIT TOTAL ERROR 5 ALLOWANCE (TA) Z _ (S)

_ TRIP SETPOINT ALLOWABLE VALUE

3. Containment Isolation E a. Phase "A" Isolation 2

- 1) Manual Initiation N.A. N.A. N.A. N.A. N.A.

2) Automatic Actuation Logic N.A. N.A. M..*.. ~ N.A. N.A.

3) Actuation Relays N.A. N.A. N.A. N.A. ::. .a,,

4) Safety Injection See Item 1. above for all Safety Injection Trip Setpoints and Allowable Values.

b.

Y Containment Ventilation Isolation g 1) Manual Initiation M.A. N.A. N.A. N.A. N.A.

2) Automatic Actuation N.A. N.A. N.A.

Logic and Actuation N.A. N.A.

Relays

3) Safety Injection See Item 1. above for all Safety Injection Trip Setpoints and Allowable Values.

4) RCB Purge .

_y -9 Radioactivity-High 34 xiefclu- I "'O Ml4 i' y ,E"N <5 x 10 4 pc/cc <6.4 x 10 4 pC/c$$>

4. Steam Line Isolation p

a. Manual Initiation -

N.A. N.A. N.A. @s3 N.A. N.A. + -4 >M C9%

po 5?

s THIS PAGE OPEN PDIDING RECDPT OF ! i!

INFORMATION FROMTHE APPUCANT ;g

O O O

! TABLE 3.3-4 (Continued) h ENGINEERED SAFETY FEATURES ACTUATION SYSTEN INSTRIM NTATION TRIP SETPOINTS 5!

_. SENSOR O TOTAL ERROR

$; FUNCTIONAL UNIT ALLOWANCE (TA) Z_ (S) TRIP SETPOINT ALLOWABLE VALUE h 4. Steam Line Isolation (Continued) m

b. Automatic Actuation Logic M.A. M.A. M.A. N.A. N.A.

~

and Actuation Relays

c. c % ,,,d % _ w 4 a. 3 . I, o. 7 I .2. 6 i 3 0 95g 490 P51p

d. Steam Line Pressure - 5 efr a.(, 0.5 0 < 100 ps' s 44 < 12Eb5 psi Negative Rate--High .4 83 .S

c. ce.geA 5% w -6 s3.6 to.'I I 20 .h 735'go

& b iy bj- --- Sue itas=:1. abuse for s** Satsty :-j : ;-

.17,$r.qp ,Y.g W

w W .24hi Toa-LouLew s'. s' c. g g. o k S32.cY*d4 Z s~22. O 'Fq 1 5. Turbine Trip and Feedwater w Isolation a

o

Mumz1 N M2. N-1. E A. M. 182.

a. 4r. Automatic Actuation Logic N.A. M.A. N.A. N.A. N.A.

b .e: Actuation Relays N.A. N.A. N.A. N.A. N.A.

to 2.ss 2 oh1 90.0 92.3

c. t. Steam Generator Water f.0 3 .38 2F $ ST.0K of < 88;8K of narrow Level-High-High (P-14) narrow range range instrument instrument span.

Span. >

d. e. Compensated T4 LLow 0.5 1.)l0 2 538'F 1 F*** M 1.1 .2.9 to <t. o 32.9. q 1 e7 Fee &eter Flow-High Coincident With:

f.4 -1:fr -3:8 5 30.0K Flow < 3EP5E Flow w

{>$

E5 E<2 N

h

!E a

l

, m -

( ) ) i l

v' n ,' \_

~

TABLE 3.3-4 (Continued) c ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTR'JMENTATION TRIP SETPOINTS 3E

_. SENSGR R TOTAL ERROR 3 FUNCTIONAL UNIT ALLOWANCE (TA) Z (S) TRIP SETPOINT ALLOWABLE VALUE e 5. Turbine Trip and Feedwater 5 Isolation (Continued)

-* 133 3

~ RCS Flow-Low 2.5 2:::I 0.6 3 90.0% Flow 189.f% Flow or

$2 $bE T,,g-Low 4.45 1.Jo .1/rc.% 1 F 1 336 2*F

g. Safety Injection See Item 1 above for all Safety Injection Trip w Setpoints and Allcwable Values.

2 o s7V 570 2

h. .4:::G 48 1.Jo -1::Sc g 1 638*F 1 5322*F m T,,g-Lew Cu i.~ i e- t with

" Reau= Te ?;, Mt

6. Auxiliary Feedwater

a. Manual Initiation N.A. N.A. N.A. N.A. N.A.

b. Automatic Actuation Logic N.A. N.A. N.A. N.A. N.A.

c. Actuation Relays N.A. N.A. N.A. N.A. N.A.

7s lom'A [

d. Steam Generator Water 15.0 12.18 4:::5 1 33.0% of 1 31.J% of narrow g narrow range range instrument g j$>

Level-Low-Low instrument span. c,g span. p*ghgg

Qsg8F E

e. Safety Injection See Item 1. above for all Safety Injection Trip Setpoints and Allowable Values. ,% +

5

Elf 3E > EEd% Coincident With: Safety Injection See Item 1. above for all Safety Injection Trip Setpoints and Allowable. Values. 5:' 8. Loss of Power a w a. 4.16 kV ESF Bus Undervoltage N. A. N.A. N.A. > 3107 > 2979 volts J, (Loss of Voltage) volts with sith a < 1.93 a < 1.75 second time second time delay. delay. 3235 3730

b. 4.16 kV ESF Bus Undervoltage M.A. M.A. N.A. > 3321 > 3830 golts (Tolerable Degraded Voltage with ty, volts time time Coincident with SI) delaysr i 395 with delaya yl 5339 sec. for sec. for atass er trip h trip wt*8CSI; w*M ST; h CSS Euc= fur h N D 3%M 1835 3?1/, d vI

c. 4.16 kV ESF Bus Undervoltage N.A. N.A. N.A. > 3821 volts > 35 M volts ! PE (Sustained Degraded Voltage)

Gith delaysh -c time Gith = 30. delays$ time c 31 g IsE pQd n - .- i: pp .:__ .:+ = - - - : : THIS PAGE OPEN PENDING RECElPT OF m; ' h; " ' g INFORMATION FROMTHE APPUCANT {,joSecfor {,ysSecfor p p (p (m O ) ' )

Table 3.3-4 (Continued)

$ ENGINEERED SAFETY FEATURES ACTUATION SYSTEN INSTRUNENTATION TRIP SETPOINTS 5 - SENSOR R TOTAL ERROR 3; FLNCTIONAL UNIT ALLOWANCE (TA) Z (5) TRIP SETPOINT ALLOWABLE VALUE h 9. Engineered Safety Features g Actuation System Interlocks ~ IP' I .O

a. Pressurizer Pressure, P-11 N.A. N.A. N.A. $ @l%5$ psig 1[L%di psig SGS 559. S'

b. Lew-Low T ,g, P-12 N.A. N.A. N.A. 1 E353}*F 3 [550.6]*F

c. Reactor Trip, P-4 N.A. N.A. N.A. M.A. N.A.

12.3%

d. Power Range Neutron M.A. N.A. N.A. < 10% Rated < ME Rated w Flux Input to Thermal Power Thermal Power A Excessive Cooldown ,

w Protection, P-15 ! iL I " 10. Control Room Ventilation

a. Manual Initiation M.A. N.A. N.A. M.A. N.A.

b. Safety Injection See Ites 1. above for all Safety Injection Trip Setpoints and Allowable Values.

c. Automatic Actuation Logic N.A. N.A. N.A. N.A. N.A.

and Actuation Relays .._l

d. Control Rcom Intake Air -r Radioactivity - High WIO c/c, s .L 2 x N c k 14 v 5 h fcc G. l sdhic, g. exio$.7

11. FHB HVAC %S

a. Manual Initiatien N.A. N.A. M.A. N.A. N.A. y

&cn e T::!S PAGE OPEN PENDING RECEIPT OF lievid tMION FROM THE APPUCANT 3 ! TABLE 3.3-4 (Continued) $ ENGINEERED SAFETY FEATURES ACTUATION SYSTEN INSTRUNENTATION TRIP SETPOINTS [ SENSOR Q TOTAL ERROR 3 FUNCTIONAL UNIT ALLOWANCE (TA) Z (S) TRIP SETPOINT ALLOWABLE VALUE h 11. FHB HVAC (Continued) 5 -* b. Automatic Actuation M.A. N. A. N.A. N.A. N.A. - Logic and Actuation Relays

c. Safety Injection See Item 1. above for all Safety Injection Trip l

Setpoints and Allowable Values.

d. Spent Fuel Pool i q Exhaust Radioactivity-High 11 xcj c;f g.tx 6 p g i, w -

5 S.0x10 4 pC/cc $ 6.4x10 4 pC/cc Y m! i y> THIS PAGE OPEN PDIDING RECEIPT OF l k1 INFORMAT10N FR0tlTHE APPUCANT e e"ili 2 ,, s z f1 leg Qd J) < D

E E

a ATTACHMENT 6 . ST HL AE /91'/ PMSff _OF_ SR rnwr & KtVltW COPY TABLE 3.3-4 (Continued) ( ) i TABLE NOTATIONS V 1

Time constants utilized in e lead-lag controller for Steam Line Pressure-Low are is > 50 seconds and 12, 5 seconds. CHANNEL CALIBRATION shall ensure that these time con ants are adjusted to these values.

- The time constant utilized in the rate-lag controller for Steam Line Pressure-Negative Rate-High is less than or equal to 50 seconds. CHANNEL CALIBRATION shall ensure that this time constant is adjusted to this value.

- - Time constants utilized in t e lead-lag controller for Compensated T COLD are t > 12 seconds and 12 > 3 seconds. CHANNEL CALIBRATION shall ensure Ihat these timercons ants are adjusted to these values.

Ov .e v SOUTH TEXAS - UNIT 1 3/4 3-25 ATTACHMENT 4 . ST.HL.AE. N P PAGED 9 0F FA rnuut & REVIEW COPY TABLE 3.3-5 ENGINEEREDSAFETYFEATURESRESPONSETINES INITIATION SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS

1. Manual Initiation PsPb

a. Safety Injection (ECCS)*-

3 N.A.

b. conConta w n,.inment Spray N.A.

nt

c. 3 Phase "A" Isolation N.A.

d. Containment Ventilation Isolation N.A.

e. Steam Line Isolation N.A.

f. Feedwater Isolation N.A.

g. Auxiliary Feedwater Pepo N.A.

h. Essential Cooling Water N.A.

i. Reactor Containment Fan Coolers N.A.

j. Control RoomNNIN Emgmy Sva W N.A.

k. Reactor Trip N.A.

-1. Stri 01:::1 Cr.:r;t;r; %. A. -

2. ContainmentPressureg-ggh-1 ,,

.h;0SafetyInjection,(SCF 5) 1h7hl)/ A. f) Reactor Trip 1 edwater Isolation < 3) .3 2) c wm.n* - 4 3) , Phase "A" Isolation < [173(p/[273r n a 3(s)/A31) C 6 $) Containment Ventilation Isolation [20 AS' / 8 3

45) Auxiliary Feedwater Pep 3 [$0(("/[1G]# 62

? 5) Essential Cooling Water Sys+2rn [32[1I/ [O[ 6 8-8 7') Reactor Containment Fan Coolers -[55["/[40[ 35Y48 W

98) Control Room [I:I9: Q N.A.

l -;0 lJ) Ste,tSh;;lCenereterg ' [1n) __ 1 I s l THIS PAGE OPEN PENDING RECEIPT OF INFORMATION FROM THE APPLICANT . l l i SOUTH TEXAS - UNIT 1 3/4 3-36 7 .- e i ITTACHMENT (, , ST HL AE /99Y PAGE ^/o 0F Ea _-

PROOF & REVIEW COPY TABLE 3.3-5 (Continued)

ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS

3. Pressurizer Pressure--Lo.s

)(. ) Safety Injection SP 1 7f(1)/h(5) .1 2) Reactor Trip < 3 7) Feedwater Isolation 3) 4 J) be 'h Isolation h {17["/[ i;(U 33 43( } 5g) Containment Ventilation Isolation d [2;] i[ Auj(& d' A

6F) Auxiliary Feedwater 9mps 7 fi) Essential Cooling Water Sp+% sh0((1I[32)(2) i [47 /

c,; g,)g/ g

67) Reactor Containment Fan Coolers noblem 3'l D-](1)/[46- ](2)

A7 y., 9 %) Control RoomgI;;isti;r, E m e, e,ncy N.A. o .;y "i.iL 0:e;;l Ca.r & : W f [10]-- g 4. (Il c.a ovmpensar.ea iCOLD""'"" h' C ety Injection CCSF _ 12](5) A J) ctor Trip < [2] 2:: w 3 f) G""'"'"* Feedw r Isolation - [7]C3) h 4 )) 4 Phase "A" lation M 1 [17](2)/[27)(1) ca. o l 5 g) Containment Ven ation Isolation 1 [25](1)/[10](2) g l l 4 $) Auxiliar}y- dwater ps +a 1 [60] < [32](2)fg47)(1) a$ Q-wp 7 fi) Esspen181CoolingWater 8/) actor Containment Fan Coolers [55)(1)/[40)(2) $ ;<g Control Room [I D N ;I ! 9 - 0*s 7r y7 N.A. o u_ ., 5) Start Diesel Generators 1 ] % jE il ) Steam Line Isolation <[ ] 4g CompensatedSteamLingressure--Low l g.)SafetyInjection(EC$)- 3 1[12p5)jg;;fn gadj,,pO l .2 X) Reactor Trip F . 3 J) co1edwaternm en+ Isolation 1((2f) - 7h3 - 4 J) gPhase "A" Isolation 1-[173(n/[iij rn. 33(0/43(4) / 5f) Containment Ventilation Isolation -i [25 [ 1 / 193 N'A' (, J) Auxiliary Feedwater Peps < [- 7 g) Essential cooling Water Systm [ [::p21jp47f11 d 5 0 [ (al N/Sl@) 6 /) Reactor Containment Fan Coolers, 1 [W)(1)/[40-](2) 57 av l SOUTH TEXAS - UNIT 1 3/4 3-37 l ~ ATTACHMENT 4 ST HL AE /* F PAGE v/ OF PA PROOF & REVIEW COP TABLE 3.3-5 (Continued) ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME'IN SECONDS 4X Compensated Steam Line g g grlow (Continued) 1,8) Control Room 3 ?:: h tkr. Emqq N.A. l0$ 5t..t ".... Cc r:te : b4 i [1 & )K., Id) Steam Line Isolation I3) 1 [43 8 8 ,k- Containment Pressure--High-3

~ Containment Spray 1 05] /G7] 30 /de

/e X Containment Pressure--High-2 7 Steam Line Isolation 1 [43(3) 7 X. SteamLinePressure[NegativeRate--High 3 Steam Line Isolation 1 [93(3) 8E Steam Generator Water Level--High-High g)

a. Turbine Trip 1 EE-53 3

b. Feedwater Isolation sj7f3)

7 J8' Steam Generator Water Level--Low-Low

a. Motor-Driven Auxiliary *y Feedwater Pumps id600

b. Turbine-Driven Auxiliary p- >-

Feedwater Pump 13600 /0 )1' RWST Level--Low-Lew CoihcIM wM Safely L efb e f -- [ Automatic Switchover to Containment . . , Sump --&A 8-- h 3a(_i) . l . Coincident with Containment Sump Level--High and Safety Injection (Automatic Switchover to Containmen ' g g l, ,, _ _ n t umD) 16 .rv a i LWJ ' ' ll }/.

Loss of Power

a. 4.16 kV ESF Bus Undervoltage 1-E19F 14.

(Loss of Voltage)

b. 4.16 kV ESF Bus Undervoltage < 47

~ ' (Tolerable Degraded Voltage Coincident with Safety Injection) ~ THIS PAGE OPEN PENDING RECEIPT OF INFORMATION FROM THE APPUCANT SOUTH TEXAS - UNIT 1 3/4 3-38 ATTACHMENT c l ST HL AE /97y PAGE yaOF E PROOF & REVIEW COPY TABLE 3.3-5 (Continued) ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS

c. 4.16 kV ESF Bus Undervoltage -< 65 (Sustained Degraded Voltage)

Puge. is. g Containment ^ir cr= Radioactivity-High Containment Ventilation Isolation -(L?ter}- h 'l 3

14. Low ensated T cold --' *

a. Turb e Trip < 2.5

b. Feedwate solation h7.0(3)

15. High Feedwater Flow High ncident with 2 of 4 Loops Having E1 Reactor Coolant A Flow - Low or T,yg -

a. Turbine Tri - Reactor T < 2.5

b. Feedw r Isolation [7.0(3)

16. T Low Coincident with Reactor Trip Feedwater Isolation < 7.0(3) 13 g Control Room Intake Air Radioactivity - High Control Room Ventilation Emergency Startup -(Latc ) f: 1 8 g Spent Fuel Pool Exhaust Radioactivity - High FHB HVAC Emergency Startup 'Later) - 6 42.

THIS PAGE OPEN PENDING RECEIPT OF INFORMATION FROM THE APPUCANT SOUTH TEXAS - UNIT 1 3/4 3-39 ATTACHMENT 4. ST HL AE /9PV PAGLWDFfa? lEW COPY TABLE 3.3-5 (Continued) TABLE NOTATIONS (1) Diesel generator starting and sequence loading delays included. (2) Diesel generator starting =d n;e ee ! delay not included e Offsite power available. yle buce.g J6fy (3) .u_____..., ...r..____ ... .u. 9.+ aern 6 w+ g ,gn3ona an M e,

11 3 "5'"d*d-(4) Diesel generator starting and sequence loading delay included. Low Head Safety Injection pumps not included.

l (5) Diesel generator starting =d n;== halb; delay ( not included regee, Low Head Safety Injection pumps not included. bb delg g ioci d.d . .e 6 m . t S0lfTH TEXAS - UNIT 1 3/4 3-40 O ' O O TABLE 4.3-2 $y ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS h TRIP $; ANALOG ACTUATING MODES 4 , CHANNEL DEVICE MASTER SLAVE FOR WHICH c- CHANNEL CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE g FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST IS REQUIRED H 1. Safety Injection (Reactor Trip, Feedwater Isolation, Control Room Emergency Ventilation, Start Standby i Diesel Generators, Reactor Containment Fan Coolers, and Essential Cooling Water) $ a. Manual Initiation N.A. N.A. N.A. R N.A. N.A. N.A. 1,2,3,4 4 b. Automatic Actuation N.A. N.A. N.A. N.A. M(1) M(1) Q 1, 2, 3, 4 ~ Logic and Actuation Relays

c. Containment Pressure- S R M N.A. N.A. N.A. N.A. 1, 2, 3, d/

High-1 I d. Pressurizer Pressure S R M N.A. N. A. N.A. N.A. 1,2,3 Low

e. Compensated Steam Line S R M N.A. N.A. N.A. N.A. 1, 2, 3 Pressure-Low

f. Compensated T COLD S R H N.A. N.A. N.A. N.A. 1, 2, c ,'

Low - Low m, g C - i l l _ _ l

O "

O O TABLE 4.3-2 (Continued) l un i 8 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION l p SURVEILLANCE REQUIREMENTS a

W TRIP

$; ANALOG ACTUATING MODES , CHANNEL DEVICE MASTER SLAVE FOR WHICH c CHANNEL CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE LOGIC TEST TEST ! }* FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST TEST IS REQUIRED l

2. Containment Spray

a. Manual Initiation N.A. N.A. N.A. R N.A. N.A. N.A. 1, 2, 3, 4

b. Automatic Actuation N.A. N.A. N.A. N.A. M(1) M(1) Q 1,2,3,4 Logic and Actuation .

Relays $ c. Containment Pressure- S R M N.A. N.A. N.A. N.A. 1,2,3 w High-3 1 I N

3. Containment Isolation i

! a. Phase "A" Isolation

1) Manual Initiation N.A. N.A. N.A. R N.A. N.A. N.A 1,2,3,4

2) Automatic Actuation N.A. N.A. N.A. N.A. M(1) M(1) Q 1,2,3,4 Logic and Actuation Relays

3) Safety Injection See Item 1. above for all Safety Injection Surveillance Requirements.

L,/.ContainmentVentilationIsolation

1) Manual Initiation IRC ICk IC8t

M- M # l' 2'

See A ms M and 3.a.1 % n.,M W&b sm6 bee. re- Eernents. ! PROOF & REVIEW COPY 1 ATTACHMENT G-ST HL AE /PP V PAGE s<r0F AR O O O ! TABLE 4.3-2 (Continued) $ ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION 5! SURVEILLANCE REQUIREMENTS I Q Mr4L ett.- TRIP R ANALOG ACTUATING MODES i CHANNEL DEVICE MASTER SLAVE FOR WHICH~ c CHANNEL CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE 1 FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST IS REQUIRED '

h /f /ontainment Mentilation

/ solation (Continued)

2) Automatic Actuation N.A. N.A. N.A. N.A. M(1) M(1) Q 1,2,3,4 Logic and Actuation Relays

3) Safety Injection See Item 1. above for all Safety Injection Surveillance Requirements.

w 4) RCB Purge i i w Radioactivity-High S R M N.8. M.4. M. 4. N.4. S($ , L(y)

4. Steam Line Isolation

a. Manual Initiation N.A. N.A. N.A. R N.A. N.A. N.A. 1,2,3

b. Automatic Actuation N.A. N.A N.A N.A. M(1) M(1) Q 1,2,3

Logic and Actuation

. Relays

c. Steam Line Pressure- S R M N.A. N.A. N.A. N.A. 3 i Negative Rate-High .

I m j d%fety !=jectica See ICBBC1. atsse-fw att Safety.injectien N Neq=c:M Go i d. Coabn,oed prmm- s q m N. 4. Pl. A. N. 4, N. A, I, 2,3 i +@ a. cescdo A Steam g, gg

e. Codw L p< e -e. - Lo u . -

o EGPsd L.ow - Loua Mc4 - 5 (L (A N. A - N. A , N. A., N . h. Is *7, 3 ATTACHMENT 6 . ST.HL-AE /9# '/ ___ PM E M OE E2 - O O O TABLE 4.3-2 (Continued) $ ENGINEERED SAFETY FEA1URES ACTUATION SYSTEM INSTRUMENTATION l M SURVEILLANCE REQUIREMENTS N TRIP I R ANALOG ACTUATING MODES , CHANNEL DEVICE MASTER SLAVE FOR WHICH c- CHANNEL CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE g FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST IS REQUIRED

5. Turbine Trip and Feedwater Isolation i a. Automatic Actuation N.A. N.A. N.A. N.A. M(1) M(1) Q 1, 2, 3 e-i Logic and Actuation l Relays w b. Steam Generator Water S R M N.A. N.A. N.A. N.A. 1,2,,3 e 1 Level-High-High (P-ly) j c. Compensated T cold -L w S R M N.A. N.A. N.A. N.A. 1, 2, 3

d. Feedwater Flow-High S R M N.A. N.A. N.A. N.A. 1, 2, 3 Coincident with either of the following in 2 of 4 loops: Reactor i Coolant Flow-Low or T,yg-Low

e. Safety Injection See Item 1. above for all Safety Injection Surveillance Requirements.

f. T -Lob:?1?.L..e S R H N.A. N.A. N.A. N.A. 1,2,3 E avg witte Asactor -Trid gn no

6. Auxiliary Feedwater m

a. Manual Initiation N.A. N.A. N.A. R N.A. N.A. N.A. 1,2,3 E

b. Automatic Actuation N.A. N.A N.A. N.A. M(1) 1,2,3 o M(1) Q and Actuation Relays @

' ATTACHMENT c -' . ST-HL-AE '999

fM.Erf.or, pA

O O b (")s ( V ! TABLE 4.3-2 (Continued) $ ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION $ SURVEILLANCE REQUIREMENTS R TRIP $l ANALOG ACTUATING MODES i CHANNEL DEVICE MASTER SLAVE FOR WHICH c: CHANNEL CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST IS REQUIRED

6. Auxiliary Feedwater (Continued)

c. Steam Generator Water S R M N.A. N.A. N.A N.A 1,2,3 Level-Low-Low

d. Safety Injection See Item 1. above for all Safety Injection Surveillance Requirements.

w 7. Automatic Switchover to A Containment Sump

a. Automatic Actuation N.A. N.A. N.A. N.A. M(1) M(1) Q 1,2,3,4

Logic and Actuation Relays

b. RWST Level-Low-Low S R M N.A. N.A. N.A. N.A 1, 2,.3, 4 Coincident With Safety Injection See Item 1. above for all Safety Injection Surveillance Requirements.

8. Loss of Power

a. 4.16 kV ESF Bus N.A. R N.A M [6) N.A. N.A. N.A. 1, 2, 3, 4 Undervoltage (Loss of Voltage) 3 l, b,

, 4, = %,ih! cm Q 'G 2 % - -4 8 a ( m m TABLE 4.3-2 (Continued) $ ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION if SURVEILLANCE REQUIREMENTS N TRIP R ANALOG ACTUATING MODES , CHANNEL DEVICE MASTER SLAVE FOR WHICH c- CHANNEL CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE g FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST IS REQUIRED ~

8. Loss of Power (Continued)

b. 4.16 kV ESF Bus N.A. R N.A. M(h) N.A. N.A. N.A. 1,2,3,4 Undervoltage (Tolerable Degraded Voltage Coincident with SI) w c. 4.16 kV ESF Bus N.A. R N.A. M (6) N.A. N.A. N.A. 1,2,3,4 h Undervoltage (Sustained w Degraded Voltage)

A

9. Engineered Safety Features Actuation System Interlocks

a. Pressurizer N.A. R M N.A. N.A. N.A. N.A. 1, 2, 3 Pressure, P-11

b. Low-Low T,yg, P-12 N.A. R M N.A. N.A. N.A. N.A. 1,2,3

c. Reactor Trip, P-4 N.A. N.A N.A. R N.A. N.A. N.A. 1,2,3

d. Power Range Neutron N.A. R (:D M (3) N.A. N.A. N.A. N.A. 1, 2, Flux Input to ;y$

Excessive Cooldown Protection, P-15 og

10. Control Room Ventilation k '.j?'

E

a. Manual Initiation N.A. N.A. N.A. R N.A. N.A. N.A. All $%e

O~' ' O O TABLE 4.3-2 (Continued) i $ ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION I ****r SURVEILLANCE REQUIREMENTS -4 0 Dia rm. .p TRIP R ANALOG ACTUATING MODES i CHANNEL DEVICE MASTER SLAVE FOR WHICH e CHANNEL CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST IS REQUIRED ]

10. Control Room Ventilation (Continued)

b. Automatic Actuation N.A. N.A. N.A. N.A. M(1) N.A. N.A. All Logic and Actuation s+.-

Relays

c. Safety Injection See Item 1. above for all Safety Injection Surveillance Requirements.

w A d. Control Room Intake Air w Radioactivity-High b R N 4 A. N. 4. g. A, y, Q, a Lt. A " 11. FHB HVAC

a. Manual Initiation N.A. N.A. N.A. R N.A. N.A. N.A. 1,2,3,4,

or with irradiated fuel in the spent fuel pool

b. Automatic Actuation N.A. N.A. N.A. N.A. M(1) N.A. N.A. 1,'2, 3, 4,

Logic and Actuation or with Relays -g irradiated

p3o fuel in the 4 '

gig ro p Q spent fuel pool THIS PAGE OPEN PDIDING RECEIPT OF }5$ = vakf' INFORMATION FROM1HE APPUCANT ;;i'i , 3 i N . _ __ O O O TABLE 4.3-2 (Continued) ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION

j SURVEILLANCE REQUIREMENTS w

Q w eirs u TRIP R ANALOG ACTUATING MODES CHANNEL DEVICE MASTER SLAVE FOR WHICH c CHANNEL CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE g FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST IS REQUIRED

11. FHB HVAC (Continued)

c. Safety Injection See Item 1. above for all Safety Injection Surveillance Requirements,

d. Spent Fuel Pool Exhaust Radio- S R M N . b. N. 4, N . S. Y b. u& irrghekd activity-High ge\ <n es g

%\ poo ) y <[ TABLE NOTATION (1) Each train shall be tested at least every 2 days on a STAGGERED TEST BASIS. (4 Nedrm Mchs <wq be oubAe1 b c64 CAbkon. LO to% pemer Drecter openenmeL An or eU14o 4ne. 'mtet\oct. set ' i red s+*e % me<e,ska\\ cosist c0 venymp Tasr %h o c k is sw % h%%\poht, %. re "f;t , ) u ge.aou c,,mneae, ga,m. Dw Cc % AcEnATioNs o megeged o4 'sreddd (\ g",nh cont 4nmed. bn nehu k e< r w\ k es m w& ma oag+Det 4 +,.e sys,6 not ry <ed. .g~ ; g ,THIS PAGE OPEN PENDING RECEIPT OF og5 o i INFORMATION FROM THE APPUCANT gg ; ,amE := sz g SOYo Y 18 , A O O O TABLE-3.3-6 $ RADIATION MONITORING INSTRUMENTATION FOR PLANT OPERATIONS i$ -4 Q MINIMUM $; CHANNELS CHANNELS APPLICABLE ALARM / TRIP , FUNCTIONAL UNIT TO TRIP / ALARM OPERABLE MODES SETPOINT ACTION C l'i -4

1. Containment

a. " -- tM - _..t stummtere t- 2-- Att- < [;] ." ?q EMie:_ t ? t- -:::-3-i

~ a A. RCS Leakage Detection

1) Particulate Radioactivity N.A. 1 1,2,3,4 N.A. 29

2) Gaseous Radioactivity N.A. I 1, 2, 3, 4 N.A. 29 w -

2. Purge and Exhaust Ventilation g a. Particulate Radioactivity 1 2 All

26

b. Gaseous Radioactivity 1 2 All

26

3. Fuel Storage Pool Areas

a. Radioactivity-High Gaseous Radioactivity 1 2 **

5 [2] mR/h 27

b. Criticality-Radiation Level 1 2 ***

$ 15 mR/h 28 j$> l c)

4. Control Room

[ ;

a. Air Intake-Radiation Level 1/ intake 2/ intake All 5 [2] mR/h 27 Q %

eo { b. Control Room Atmosphere 1 2 All < [2] mR/h 27 D *e c- ::o ( Radiation-High QI 3 8 A yT PROOF & REVIEW COPY TABLE 3.3-6 (Continued) ATTACHMENT G HL TABLE NOTATIONS gE}AE / p\J $ ? ,wd set ' 'f S p--i r:H q 3_11 7. 1- r y - -t:.

With i--+# eta fust in the fuel siemj pcurl -

" hth fust in the fusi sie ;-W actes. ACTION STATEMENTS ACTION 26 - With less than the Minimum Channels OPERABLE requirement, operation may continue provided the containment purge and M GLED Lexhaust valves are maintained closed.J ACTION 27 - YiththenumberofOPERABLEchannelsonelessthantheMinimum Channels OPERABLE requirement, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months isolate the Control Room Emergency Ventilation System and initiate operation g gD of the Control Room Emergency Ventilation System in thea recirculation mode.J ACTION 28 - With less than the Minimum Channels OPERABLE requirement, opera-tion may continue for up to 30 days provided an appropriate NoT uSED portable continuous monitor with the same Alarm Setpoint is provided in the fuel storage pool area. Restore the inoperable monitors to OPERABLE status within 30 days or suspend all operations involving fuel movement in the fuel storage pool w ACTION 29 - Must satisfy the ACTION requirement for Specification 3.4.6.1. O SOUTH TEXAS - UNIT 1 3/4 3-51 k' O O O

TABLE 4.3-3

$ RADIATION MONITORING INSTRUMENTATION FOR PLANT M OPERATIONS SURVEILLANCE REQUIREMENTS .. - %$d ANAt06 R , $; CHANNEL MODES FOR WHICH 4 . CHANNEL CHANNEL OPERATIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST IS REQUIRED ! g Z 1. Containment

Nt it=:phn Ecdica-. .i-ity "' & -

11:- 1 %It

Ap'. RCS Leakage Detection

1) Particulate Radio- S R M 1,2,3,4 activity

! 2) Gaseous Radioactivity S R M 1,2,3,4 Y r

2. Purge and Exhaust Ventilation

] g a. Particulate Radioactivity S R M All

b. Gaseous Radioactivity S R M All

3. Fuel Storage Pool Areas l a. Radioactivity-High-j Gaseous Radioactivity S R M

b. Criticality-Radiation Level S R M _____.

4 4. Control Room $*-4h

a. Air Intake Radiation Level S R M All ]

b. Control Room Atmosphere PE j'3 m E @9 i k Radiation-High S R M All

* -4! m W

i 1 TABLE NOTATIONS

With fuel in the fuel storage pool area. - :

Q With irradiated fuel in the fuel storage pool areas. ,

g !

b x, - .-. - -.- .. -- _ - -_ ~ -. __ - - .. - DDAAC D nrturnu nom. INSTRUMENTATION ATT46HWERY'5"" "" . ST HL AE. /% y PAGE 55' OF T3-MOVABLE INCORE DETECTORS , LIMITING CONDITION FOR OPERATION 3.3.3.2 The Movable Incore Detection System shall be OPERABLE with:

a. At least 75% of the detector thimbles,

b. A minimum of two detector thimbles per core quadrant, and

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

APPLICABILITY: When the Movable Incore Detection System is used for:

a. Recalibration of the Excore Neutron Flux Detection System, or

b. Monitoring the QUADRANT POWER TILT RATIO, or N

c. Measurement of FAH, xy.

Q(f)andF ACTION: . With the Movable Incore Detection System inoperable, do not use the system for the above applicable monitoring or calibration functions. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. i 1 SURVEILLANCE REQUIREMENTS 4.3.3.2 The Movable Incore Detection System shall be demonstrated OPERABLE at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by normalizing each detector output when required for:

a. Recalibration of the Excore Neutron Flux Detection System, or

b. Monitoring the QUADRANT POWER TILT RATIO, or N A

c. Measurement of FAH,

Q(Z) and Fxy, V

l O SOUTH TEXAS - UNIT 1 3/4 3-53 PROOF & Rru!nu cory TABLE 3.3-7 ATTACHMENT 4 . ST HL AE IM'l O setsatc ao"itoat"o tastauae"Tatto" "^o' " " # ' MINIMUM MEASUREMENT INSTRUMENTS INSTRUMENTS AND SENSOR LOCATIONS RANGE OPERABLE

1. Triaxial Time-History Accelerometers 44

a. Free Field 13g 1

b. Containment Bldg. Foundation 13g 1 (Tendon Gallery, E1. -36'9")

c. Outside Face Containment Shell 13g 1 b b c d innerg . 68'0") ,,m ,p,r

d. h e [ Upper Lateral Support 13g 1 A % m h q 1. 66'71")-%%M s

e. fM8/ Foundation D G\blwr 13g 1

(El. -29'0") ,j. gyg ggg-13g 1

f. MEAS mW&bd (El. 35'0") g ,gL6) g &J\&l g g

.( 2. Triaxial Peak Accelerographs

a. Spent Fuel Pool NX 44 S'bTv 13g 1 (Inlet Line QEl. 64'51") s

b. Reactor Vessel LA bity %\L 13g 1 R=6 cort.l f c.Co17LegofRCPiping 13g 1 Ach o,wed LiAmy 3.Self)ContainedTriaxialAccelerograph 13g 1 (At(RC8' Foundation Tendon Gallery, EL. -396 9")

o.ofh aca39+

4. Triaxial Seismic Switch ** WOS te 39 1*

1 o.on eg + i

5. Triaxial Seismic Trigger ** GdG; w 0.Jg 1*

6. Response Spectrum Analyzer ** 1 m M 4 nA 1*

7. Magnetic Tape Recorders ** 0.1 to 33 Hz 6

8. Playback System W N.A. 1

With reactor control room indication and alarm d **At seismic monitoring panel in Control Room, Unit 1 44 Auebwh AM. L b ed. e b d ad suhmam\prcxdti)' l b h ruIeElve

+ n. se+ pain t-SOUTH EXAS - UNIT 1 3/4 3-55 l l .. ._ -_ . _ _ _ . .- ATTACHMENT 4 . STML.AE 14tf PA4100Ff prvirw mov p -- ... TABLE 4.3-4 O SE1SM1c M0 110atso taStauMearar10" SuaverLLAace aEcuraEseNTS ANALOG CHANNEL CHANNEL CHANNEL OPERATIONAL INSTRUMENTS AND SENSOR LOCATIONS CHECK CALIBRATION TEST

1. Triaxial Time-History Accelerometersx g

a. Free Field M R SA

b. Containment Bldg. Foundation M R SA (Tendon Gallery, E1. -36'9")

c. Outside Face Containment Shell M R SA

%g-ut JE1. 68'{} d.3dtgUpperLateralSupport M R SA bs4co%d (E1. 66'7 ") 'M 'l e. AtEAFoundation (41. -29'0") ' '} g,., M R SA

f. Mtt8 A6*.\ mew,a (b.t..,7 s)

M R ~ SA -- (E1. 35'0")

2. Triaxial Peak Accelerographs

a. Spent Fuel Pool WX4.& N.A.

' EA' f2 R N.A. (Inlet Line MSQ. gD

b. Reactor Vessel ) N.A. R N.A.

Abggt&,n,d(RCSVEl.68'0") '7c.ColdLegofRCPiping N.A. R N.A. S M R SA

3. % elf-Contained RES FoundationTriaxial TendonAccelerograph Gallery, q'gh,h El. -3g'9")

4. Triaxial Seismic Switch * ** M R SA

5. Triaxial Seismic Trigger * ** M R SA

"" 6. Response Spectrum Analyzer * ** M R SA

7. Magnetic Tape Recorders ** M R SA

8. Playback System *# M R % N4

With reactor control room indication and alarm

- At seismic monitoring panel in Control Room, Unit 1 '

At Acc lerembda h ph A 4 uc\)ed% %. mpech ==heedibD.

SOUTH TEXAS - UNIT 1 3/4 3-56 PROOF & REVIEW COPY TABLE 3.3-8 l ATTACHMENT 6 g ST.HL.AE / 95 4 METEOROLOGICAL MONITORING INSTRUMENTATION ' MINIMUM INSTRUMENT LOCATION OPERABLE

1. Wind Speed

a. Primary meteorological Nominal Elev. 33d10m 1 tower or backup mete-orological tower

b. Primary meteorological Nominal Elev. Isr 60m 1 tower

2. Wind Direction

a. Primary meteorological Nominal Elev. W 10 m 1 tower or backup mete-orological tower

b. Primary meteorological Nominal Elev. 32 60m 1 tower

3. Air Temperature - AT

a. Primary meteorological Nominal Elev. E lO m 1 tower

b. Primary meteorological Nominal Elev. W loCM 1 tower s

+ 0 SOUTN TEXAS - L' NIT 1 3/4 3-58 PROOF & REVIEW COPY ATTACHMENT (, TABLE 4.3-5 ST HL AE 199 4 METEOROLOGICAL MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL CHANNEL INSTRUMENT CALIBRATION CHECK _

1. Wind Speed

a. Nominal Elev. W 10m D SA

b. Nominal Elev. 23t'(,0 % D SA

2. Wind Direction

a. Nominal Elev. 3t 10 m D SA

b. Nominal Elev. IS1' loom D SA

3. Air Temperature - AT

a. Nominal Elev. -3Sd-lOm D SA

b. Nominal Elev. THZ'60rn D SA I

O SOUTH TEXAS - UNIT 1 3/4 3-59 PROOF & REVIFW cnov INSTRUMENTATION I ATTACHMENT 6 i ST HL-AE 1911 ( REMOTE SHUTDOWN SYSTEM i PAGE (,0 OF P LIMITING CONDITION FOR OPERATION 3.3.3.5 The Remote Shutdown System transfer switches, power, controls and monitoring instrumentation channels shown in Table 3.3-9 shall be OPERABLE. APPLICABILITY: MODES 1, 2, and 3. ACTION: bd, sJ%b, ge,ceAe\s or

a. WiththenumberofOPERABLEremoteshutdownlmonitoringchannelsless than the Minimum Channels OPERABLE as required by Table 3.3-9, restore the inoperable channel (s) to OPERABLE status within 7 days, or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , b dwr w h p e , e b b,or

b. WiththenumberofOPERABLEremoteshutdowndmonitoringchannelsless than the Total Number of Channels as required by Table 3.3-9, within 60 days restore the inoperable channel (s) to OPERABLE status or, pursuant to Specification 6.9.2, submit a Special Report that defines the corrective action to be taken, c<- With e r mo e R mot S td wn Sys m ran fer swi hes p e, or c ntr ci cui s i pe ab e, es re he ino era e s it ( /

ci ui s) OP RAB sa s th 7 ay , o be in T A BY O thi the ex 12 ur . cg. The provisions of Specification 3.0.4 are not applicable. SURVEILLANCE REQUIREMENTS 4.3.3.5.1 Each remote shutdown monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3-6. ., 4.3.3.5.2 Each Remote Shutdown System transfer switch, power and control circuit including the actuated components, shall be demonstrated OPERABLE at least once per 18 months. O SOUTH TEXAS - UNIT 1 3/4 3-60 O O O TABLE 3.3-9 c REMOTE SHUTDOWN SYSTEM M _., TOTAL NO. MINIMUM R READOUT OF CHANNELS INSTRUMENT LOCATION CHANNELS OPERABLE

1. Neutron Flux --ExbAd

a. Startup Rate ASP *-QDPS** 2 1

b. " Range r%6e\ ASP-QDPS 2 1

2. Reactor Trip Breaker Indication ASP-QDPS 1/ trip breaker 1/ trip breaker Reactor Trip Switchgear

3. Reactor Coolant _ Temperature-w Wide Range cn

a. Hot Leg 4-1/ loop .s. /lep ASP-QDPS 'a. \ oops #

b. Cold Leg ASP-QDPS 4-1/ loop .i./ w p - 2 \oo p tF

4. Reactor Coolant Pressure-Wide Range / Extended Range ASP-QDPS 3 1 t -- es =

x Pressawe dif -Q:;f-i -4:- s ,8. Pressurizer Water Level ASP-QDPS 4 1

/,J. Steam Line Pressure ASP-QDPS .i g/ steam line 1/ steam line- a lines #. ik'" $ g 7p. Steam Generator Water Level- { gg Wide Range ASP-QDPS 4-1/ steam 2 %= wmg generator 9%J 3% -e -t Y* V m 3: E 3 O O ' (3) G TABLE 3.3-9 (Continued) E REMOTE SHUTDOWN SYSTEM w TOTAL NO. MINIMUM R READOUT OF CHANNELS $ INSTRUMENT LOCATION CHANNELS OPERABLE c 9. Auxiliary Feedwater Flow Rate ASP-QDPS 4-1/ steam 24 5 a generator " 10. Auxiliary Feedwater Storage Tank Water Level ASP-QDPS 3 1 [RANSFERSWITCHES g SWITCHLOCATION) CONTROL CIRCUITS SWITCH LOCATIO l 3 , T

ASP - Auxiliary Shutdown Panel 8 **QDPS - Qualified Display Processing System

- m.e w . % s-- e,m aa. m wpA 4 4 i

! THIS PAGE OPEN PSDMG REC 8PT OF l INFORMATION FROM THE APPLICANT

%n3 g

~ E,hz i 8 b[M A $ '* e m o I E q n ,,,,v,,,,,._,,, 9 ST HL AE p q9 , ' GE t,3 0F 9% . - 1 1 $ 5 m= $ n , b*n - ~ ~ m - N N N n x o $$ + N m t N W n m N m m g n m ba n E Eo N w z o. M $ I3

k k 4 L g T

S 4 4 q4 4 a 1 % & ^ n a5 48 o % o5 3. q 4 e ^

-g e (2 3a, 4 x ^ E " v v c. o a s ( R I E 3 S e; w W d E s e g . I

O s ! 1 i a 3 i a 4 9 i

.** O v c s .6 d vi -4.- i % k I a i 2 L1 , 4l ? eYH s g y +g i K ,i .b' )- -l "j e < g = ,s l l )i + d b f } & dr + a y 5 y 5 -!33 I r4a, J ?

d. 2i 4 i Is 33

- mt 8 a d 3 1 3 m i i E

b .

<4 m # w a w % e a :s y e SOUTH TEXAS - UNIT 1 3/4 3-62A . ' ATTACHMENT 6 . ST HL AE /414 PAGE iAOF D. J (. TABLE 3.3-9 (Continued) ! NOTES TO TABLE (1) Train A ZLP-653 !

ZLP-700 AW Valves Train B ZLP-654 ZLP-701 AW Valves I 4 ;

Train C ZLP 655 , ZLP-709 A W Valves j I Train D ASP (2) Train A ZLP-700 - Train B ZLP-701 (3) Train A ZLP-653

Train C ZLP-655 j (4) Train A ZLP-653 i

Train B ZLP 654 Train C ZLP-655 (5)' Train B ZLP-654 4 Train C ZLP-655 (6) Train A ZLP-700 ZLP-653 Battery Room and Electrical Penetration Space Fans Train B ZLP-701 ZLP-654 Battery Room and Electrical Penetration Space Fans , Train C ZLP-709 ZLP-655 Battery Room and Electrical Penetration Space Fans i (7) Train A ZLP-700 i Train B ZLP-701 1 Train C ZLP-709 (8) Train A ZLP-700 Train C ZLP-709 i i i SOUTH TEXAS - UNIT 1 3/4 3-62C l L3/NRC/cm 4 l ,-- -,--. ,, ..-,___,v_.. -.,-__-.---,--.,,,--,.-_---_....-v,-,. .,----_-..----v.,-,~.,m, TABLE 4.3-6 m REMOTE SHUTDOWN MONITORING INSTRUMENTATION l 8 SURVEILLANCE REQUIREMENTS _. CHANNEL CHANNEL O INSTRUMENT CHECK CALIBRATION l 5; , 1. Neutron Flux C

a. Startup Rate M g&

i'i.

b. Extended Range M AA

2. Reactor Trip Breaker Indication M N.A.

3. Reactor Coolant Temperature-Wide Range w a. Hot Leg M R i

l w b. Cold leg M R m w I I

4. Reactor Coolant Pressure-Wide Range / Extended Range M R i

-k Fr-ssurim r Pawssure 4t:- t 5 A. Pressurizer Water Level M R l

64. Steam Line Pressure M R

74. Steam Generator Water Level-Wide Range M R fg. Auxiliary Feedwater Flow Rate.

$$D l M R @gy 1 720. Auxiliary Feedwater Storage e%E '^ 8 Tank Water Level M R % ,n,Qg , % dM p F*e , I _ I B PROOF & REVIEW COPY , INSTRUMENTATION ATTACHMENT 6 ! t HL AE. /qt 7 ' ACCIDENT MONITORING INSTRUMENTATION '1_% OF D-LIMITING CONDITION FOR OPERATION 3.3.3.6 The accident.monito-ing instrumentation channels shown in Table 3.3-10 shall be OPERABLE. 1 APPLICABILITY: MODES 1, 2, and 3. ACTION: b de co Tak L3-10 ;

a. With the number of OPERABLE accident monitoring instrumentation channels less than the Total fa ber n of Chanaels shown in Table 3.3-10, restore the inoperable channel (s) to OPERABLE status within 7 days, or be in at least HOT STANDBY within the next 6 hjutrL_a_nd in at least unT _ SHUTDOWN within the followin A hour 1J

! b. ith the number of OPERABLE accident monitoring instrumentation channels except the unit vent-high range noble gas monitor, the steam - relief-high range radiation monitor, the containment atmosphere-high range radiation monitor, and the reactor coolant radiation level monitor less than the Minimum Channels OPERABLE requirements of Table 3.3-10, restore the inoperable channel (s) to OPERABLE status j within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in at least HOT STANDBY within the next j 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in at least HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

c. With the number of OPERABLE channels for the unit vent-high range noble gas monitor, or the steam relief-high range radiation monitor or the containment atmosphere-high range radiation monitor, or the ppqwr Smijreactor coded pediation level monitor less than required by the Minimum Channels OPERABLE requirements, initiate an alternate method of monitoring the appropriate parameter (s), within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , and either restore the inoperable channel (s) to OPERABLE status within 7 7 days or prepare and submit a Special Report to the Commission, pur ,

suant to Specification 6.9.2, within 14 days that provides actions

taken, cause of the inoperability, and the plans and schedule for

[estoring the channels to OPERABLE status. , . The provisions of Specification 3.0.4 are not applicable. j SURVEILLANCE REQUIREMENTS j 4.3.3.6 Each accident monitoring instrumentation channel shall be demonstrated , OPERABLE by performance of the CHANNEL CHECK and CHANNEL CALIBRATION at the frequencies shown in Table 4.3-7. o ; SOUTH TEXAS - UNIT 1 3/4 3-64 s ex ( , J L.j TABLE 3.3-10 ACCIDENT MONITORING INSTRUMENTATION 2 w TOTAL MINIMUM Q NO. OF CHANNELS R INSTRUMENT CHANNELS OPERABLE A CTO'd a c- 1. Containment Pressure 4 1 q 5

2. Reactor Coolant Outlet Temperature - TH0T (Wide Range) 1/ loop 1/ loop

] 1

3. Reactor Coolant Inlet Temperature - TCOLD (Wide Range) 1/ loop 1/ loop 1

4. Reactor Coolant Pressure [ Wide RangDA 3 1 3

5. Pressurizer Water Level 4 1 q w 6. Steam Line Pressure 4/ steam generator 1/ steam generator q k

w 7. Steam Generator Water Level - Narrow Range 4/ steam generator 1/ steam generator 4

8. Steam Generator Water Level - Wide Range 1/ steam generator 1/ steam generator l

9. Refueling Water Storage Tank Water Level 3 1 3

10. Auxiliary Feedwater Storage Tank Water Level 3 1 3

11. Auxiliary Feedwater Flow 1/ steam generator 1/ steam generator g

12. Reactor Coolant System Subcooling Margin fonito 2 1 2

13. Containment Water Level (Narrow Range) 2 1 2

14. Containment Water Level (Wide Range) 3 1 3 "j]

15. Core Exit -T!=mccoupies TageDe_ X%------,--:== '/// core quadrant p

16. Steam Line Radiation Mar.itsr Levd 1/ steam line 1/ steam line 4 Qgg

17. Containment @ mospher - High Range Radiation b leve l 2 1 M 8

'e O ' O O TABLE 3.3-10 (Continued) $ ACCIDENT MONITORING INSTRUMENTATION N w TOTAL MINIMUM R NO. OF CHANNELS M INSTRUMENT CHANNELS OPERABLE &cTON e g 18. Reactor Vessel Water Level (RVWL) 2* 1* 1 w

19. Neutron Flux (Extended Range) 2 1 2.

20. Containment Hydrogen Concentration 2 1 2,

21. Containment Pressure (Extended Range) 2 1 2

22. Steam Generator Blowdown Radiation Fiusiit.vi Leve_\

1/ blowdown line 1/ blowdown line b w 23. Neutron Flux - Startup Rate (Extended Range) 2 1 2 1 Y $ *A channel is eight sensors in a probe. A channel is OPERABLE if four or more sensors, one or more in the upper section and three or more in the lower section, are OPERABLE. M Prwe. -h.%\ e4 To %eemocoq\es n pideb , t*N R Otmecoy o weemoc l et eneach E Do NS . dludeds G wM D 9 h e-ecoq\t \es. p% o A , 6r h am emeA.c h imi&s A - 4 t e 4 w e. 6 % % = no 96 ovs e,m. +e and % ~ewiu e.< w eeenam. wrica ,, ceptred ,o gon3 m s .4 I la % y b:ng 1 m: 5E p.q-F $NP l C D b l i ATTACHMENT b ST HL-AE NW I PAGE69 OF fd , _ TAB 1.E 3.3-10 (Continued) Y ACTION STATDENTS i 1 \ ACTION 1 - With the number of OPERAILE channels . lass than the Minimum ' Channels Operable requirement, restore at least one inoperable channel to OPERA 31.E status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , or be in 807 SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . - ACTION 2 - a. With the number of OPERABLE channels one less than the Total Number of Channels requirements, restore one inoperable channel to OPERABLE status within 7 days, or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b. With the number of OPERABLE channels less than the Minimum Channels Operable requirements, restore at least one inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , or be in NOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . -

ACTION 3 - a. With the number of CPERABLE channels one less than the Total Number of Channels requirements, restore the inoperable channel to CPERABLE status within 31 days, or be in at least HOT .

$HUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ,

b. With the number of OPERABLE channels two less than' the Total Numbe'r of Channels requirement, restore at least one inoperable 4

channel to OPERA 8LE status within 7 days, or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . c.WiththenumberofOPERABLEchannelslessthantheitinimumChan-nels Operable r'equirement, restore at least one inoperable chan-nel to OPERA 8LE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in HOT SHUTOOWN with j the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . ACTION 4 . a. With the number of OPERABLE channels one less than the Total Number of Channels requirements, restore the inoperable channel to OPERABLE status within 90 days, or be in a.t least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

k. With the number of OPERABLE channels two less than the Total Number of Channels requirements, restore the inoperable channel to OPERA 8LE status within 31 days,'or be in at least NOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

l c.. With the number of OPERABLE channels three less than the Total Number of Channels requirement, restore at least one-inoperable channel to OPERA 8LE status within 7 days, or be in at least NOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

d. With the number of OPERABLE channels less than the Minimum Chan-nels Operable requirement, restore at least one inoperable chan-nel 1,o OPERABLE status with 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in NOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . .

, 3/4 3-CJ./l , : a , - .-. - - - .. - - - l ATTACHMENT e,T HL AE- /9f9 PAGEy OF F2 - TABLE 3.3 to(Continued) ACTION _$TATEMENTS ACTION 5 . a. With the number of OPERABLE channels one less than the Number of Channels requirements, restore one inoperable channel to OPERABLE status within 7 days, or be in at least NOT SHUTOOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b. With the number of OPERABLE channels less than the Minimu Channels Operable requirements, restore at least one inoperable channel to OPERABLE status within '12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , or be in HOT SHUTOOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

ACTION 6 '- With the number of OPERABLE channels less than the Minimum Chan-nels Operabia requirements, restore at least one inoperable chan-nel to CPERA8LE statud within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , or be in HOT SHUTDOWN 'within the next,12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . / 7 ACTION 7I - a. With the number of OPERABLE Channels one less than ~ the Required Number of Channels, either restore , the system to 0PERABLE status within 7 days if repairs are feasible without shutting down-or - prepare and submit a Special Report to the' Comission pursuant to Specification 6.9.2 within 30 days following the event outlining the action taken, the cause of the inoperability and the plaris and . schedule for restoring the system to OPERABLE status. . D

b. With the number of OPERABLE Channels one less than the Minimum Channels OPERABLE in Table 3.3-10, 5 either restore the inoperable channel (s) to -

OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months if repairs are feasible without shutting down or: ,

1. Initiate an alternate method of monitoring the reactor vessel inventory; *

2. Prepare and submit a Special Report to the Comnission pursuant to Specification 6.9.2 within 30 days i

following the event outlining'the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status; and

3. Restore the system to OPERGLE status at the next .,

scheduled refueling. . m 9 3- u, a / a ~' ATTACHMENT 4 . ST-HL AE /9F4 PAGE % 'OF T2. TABLE 3.3-10 (Continued) ACTION STATEMENTS ACTION 8 - a. With the number of OPERABLE channels.less than 4 thermocouples per quadrant per train, restore these thermocouples to OPERABLE status within 31 days, or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b. With the number of OPERABLE channels less than 6 thermocouples per quadrant, restore these thermocouples to OPERABLE status within 7 days, or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

c. With the number of OPERABLE channels less thr.n 4 thermocouples per quadrant, restore these thermocouples to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

s:,y '[ n F ? j Jj * , i s a' ' ,f ./ , I, ,' t s a l*

r. .

O _ s - r r T l 3/4 3-66C L3/NRC/cm ', I. d * ' (- ,y %.,) ! TABLE 4.3-7 m 8 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS -4 CHANNEL CHANNEL 92 INSTRUMENT CHECK CALIBRATION a c 1. Containment Pressure M j 'R

2. Reactor Coolant Outlet Temperature - THOT (Wide Range) M R

[

3. Reactor Coolant Inlet Temperature - TCOLD (Wide Range) M R Reactor Coolant Pressure M A

4. M R s

5. Pressurizer Water Level M R g 6. Steam Line Pressure M R a

y 7. Steam Generator Water Level - Narrow Range M R r

8. Steam Generator Water Level - Wide Range M '

R

9. Refueling Water Storage Tank Water Level M _

R

10. Auxiliary Feedwater Storage Tank Water Level M R

11. Auxiliary Feedwater Flow M R

12. Reactor Coolant System Subcooling Margin M R

13. Containment Water Level (Narrow Range) M R :j >

14. Containment Water Level (Wide Range) M R g.AA og r-

15. Core Exit The:- mW TeywitIn M R Q

16. Steam Line Radiation M2- : ter-- \. eye { M R h

17. Containment High Range Radiation M . "r Leve} M R o

O O O - O ! TABLE 4.3-7 (Continued) ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIRENENTS a CHANNEL CHANNEL INSTRUMENT CHECK CALIBRATION i 18. Reactor Vessel Water Level (RWL) M R

19. Neutron Flux (Extended Range) M R

20. Containment Hydrogen Concentration M R

21. Containment Pressure (Extended Range) M R

22. Steam Generator Blowdown Radiation Msintur. Lauel M R

23. Neutron Flux - Startup Rate (Extended Range) M R R.

Y a> 99 3 8 y.b n e = E 8 3 PROOF & REVIEW COPY INSTRUMENTATION Mk ATTACHMENT (r - ST HL AE 19N b) v FIRE DETECTION INSTRUMENTATION PAGE 14 OF 9% LIMITING CONDITION FOR OPERATION 3.3.3.8 As a minimum, the fire detection instrumentation for each fire detec-tion zone shown in Table 3.3-11 shall be OPERABLE. APPLICABILITY: Whenever equipment protected by the fire detection instrument is required to be OPERABLE. ACTION:

a. With any, but not more than one-half the total in any fire zone, Function A fire detection instruments shown in Table 3.3-11 inoperable, restore the inoperable instrument (s) to OPERABLE status within 14 days or within the next I hour establish a fire watch patrol to inspect the zone (s) with the inoperable instrument (s) at least once per hour, unless the instrument (s) is located inside the containment, then inspect that containment zone at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months (or monitor the containment air temperature at least once per hour at the locations listed in Specification 4.6.1.6).

b. With more than one-half of the Function A fire detection instruments in any fire zone shown in Table 3.3-11 inoperable, or with any Function B fire detection instruments shown in Table 3.3-11 inoperable, (s or with any two or more adjacent fire detection instruments shown in Table 3.3-11 incperable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months establish a fire watch patrol to inspect the zone (s) with the inoperable instrument (s) at least once per hour, unless the instrument (s) is located inside the.contain-ment, then inspect that containment zone at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months (or monitor the containment air temperature at least once per hour at the locations listed in Specification 4.6.1.6).

c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.3.3.8.1 Each of the above required fire detection instruments which are accessible during plant operation shall be demonstrated OPERABLE at least once per 6 months by performance of a TRIP ACTUATING DEVICE OPERATIONAL TEST. Fire detectors which are not accessible during plant operation shall be demonstrated OPERABLE by the performance of a TRIP ACTUATING DEVICE OPERATIONAL TEST during each COLD SHUTDOWN exceeding 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months unless performed in the previous 6 months. 4.3.3.8.2 The NFPA Standard 72D supervised circuits supervision associated with the detector alarms of each of the above required fire detection instruments shall be demonstrated OPERABLE at least once per 6 months. 4.3.3.8.3 The nonsupervised circuits, associated with detector alarms, between O the instrument and the control room shall be demonstrated OPERABLE at least once per 31 days. SOUTH TEXAS - UNIT 1 3/4 3-70 3dk PROOF A psview ennv ATTACHMENTT ~ ' TABLE 3.3-11 - ST.HL-AE 1939 PAGE 75 OF p .O FIRE DETECTION INSTRUMENTS V TuiAL NUMBER INSTRUMENT LOCATION OF INSTRUMENTS * [ Illustrative] HEAT FLAME SMOKE

1. Containment **

a. Zone 1 Elevation

b. Zone 2 Elevation

2. Control Room

3. Cable Spreading

a. Zone 1 Elevation

b. Zone 2 Elevation

4. Computer Room

5. Switchgear Room

6. Remote Shutdown Panels

7. Station Battery Rooms

8. Turbine C a. Zone 1 Elevation '

b. Zone 2 Elevation

9. Diesel Generator

a. Zone 1 Elevation

b. Zone 2 Elevation

10. Safety-Related Pumps l

a. Zone 1 Elevation l b. Zone 2 Elevation

11. Fuel Storage

a. Zone 1 Elevation

b. Zone 2 Elevation l

l [ List all detectors in areas required to ensure the OPERABILITY of safety-related equipment.]

(x/y): x is number of Function A (early warning fire detection

! and notification only) instruments. y is number of Function B (actuation of Fire Suppression Systems and early warning and notification) instruments.

- The fire detection instruments located within the containment are not required to be OPERABLE during the performance of Type A containment leakage rate tests.

SOUTH TEXAS - UNIT 1 3/4 3-71 l l PRntir a nei -. - ~ ATTACHEEMT' W UUVY INSTRUMENTATION . ST-HL-AE 14fy RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION , 3.3.3.10 The radioactive liquid effluent monitoring instrumentation channels i shown in Table 3.3-12 shall be OPERABLE with their Alarm / Trip Setpoints set to ensure that the limits of Specification 3.11.1.1 are not exceeded. The Alarm / , Trip Setpoints of these channels shall be determined and adjusted in accordance with the methodology and parameters in the OFFSITE DOSE CALCULATION MANUAL (ODCM). APPLICABILITY: At all times. ACTION:

a. With a radioactive liquid effluent monitoring instrumentation channel Alarm / Trip Setpoint less conservative than required by the above l

specification, immediately suspend the release of radioactive liquid effluents monitored by the affected channel, or declare the channel inoperable.

b. With less than the minimum number of radioactive ~ liquid effluent monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 3.3-12. Restore the inoperable instrementation to OPERABLE status within the time specified in the ACTION, or explain in the next Semiannual Radioactive Effluent Release Report pursuant to Specification 6.9.1.4 why this inoperability was not corrected within the time specified.

c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.3.3.10 Each radioactive liquid effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATIONa and DIGITAL CHANNEL OPERATIONAL TEST, as applicable, at the frequencies shown in / Table 4.3-8. I ANeto6 Ane\ opend:Gwd.%D 1 t 4 i l O SOUTH TEXAS - UNIT 1 3/4 3-73 o f% b b. - TABLE 3.3-12 1 g RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION l N w g MINIMUM g CHANNELS , INSTRUMENT OPERABLE ACTION C j'i 1. Radioactivity Monitors Providing Alarm and

Automatic Termination of Release H

a. Liquid Waste Processing Discharge Monitor 1 35

b. Condcasate Polisher Di: charge "0nitcr + *

c. Tur^

u ine/Generai.u. Ouilding Drai- "aniter P W w 2. Flow Rate Measurement Devices A

w a. Liquid Waste Processing Discharge Line 1 38

b. Condcasate PelisMng Disch:rge Line- i, l

$ND I Elis r m.o g E La a-42-v ec g= 2 O

U M

O ~ O O TABLE 4.3-8 $ RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS Art 410 Lv 0 - DIGITAL M CHANNEL . CHANNEL SOURCE CHANNEL OPERATIONAL c INSTRUNENT CHECK CHECK CALIBRATION TEST

Radioactivity Monitors Providing 1.

Alarm and Automatic Termination of Release

a. l.iquid Waste Processing Discharge Monitor D P R(3) Q(1)

-h- em pstetusy W nnnitrir -dh- W 1t(3) W w -ra T = inc/^....7:t:r hiidi.., C. :. ". ; iter $- *- EC37 -QCI) D w 2. Flow Rate Measurement Devices ~s

Liquid Waste Processing Discharge Line

a. D(4) N.A. R -Q- N O tr tsunmrate- Pei i:: *- G s e _i - f tie Upt1 it;&. t 4

--4 --4 @N as 1

mg o*m 9 33k N 8

_=

i C3 ~ ~ i c> l t C > 3 ATTACHMENT 4 ST-HL AE- 14d pane 3 6JFTd PR0ur & REVlEW COPY TABLE 4.3-8 (Continued) TABLE NOTATIONS l (1) The DIGITAL CHANNEL OPERATIONAL TEST shall also demonstrate that automatic isolation of this pathway and control room alarm annunciation occur if any of the following conditions exists:

a. Instrument indicates measured levels above the Alarm / Trip Setpoint, or

b. Monitor failure,

t. Imi.r......^. ;;.."..;?., n-e-. ::t i eperei- -;&.

(2) Dhe DIGITAL CHANNEL OPERATIONAL TEST shall also demonstrate that control room alarm annunciation occurs if any of the following conditions exists: g3g a. Instrument indicates measured levels above the Alarm Setpoint, or

b. Circuit failure, or

c. Instrument indicates a downscale failure, or td. Instrument controls not set in operate mod (3) The initial CHANNEL CALIBRATION shall be performed using one or more of p the reference standards certified by the National Bureau of Standards (NBS)

( or using standards that have be'en obtained from suppliers that participate in measurement assurance activities with NBS. These standards shall permit calibrating the system over its intended range of energy and measurement range. For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration shall be used. (4) CHANNEL CHECK shall consist of verifying indication of flow Juring periods of release. CHANNEL CHECK shall be made at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months on days on which continuous, periodic, or batch releases are made. THIS PAGE OPEN PENDING RECEIPT OF INFORMATION FROM THE APPLICANT O SOUTH TEXAS - UNIT 1 3/4 3-77 O ' O O TABLE 3.3-13 $ RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION h MINIMUM CHANNELS g INSTRUMENT OPERABLE APPLICABILITY ACTION h z

1. GASEOUS WASTE PROCESSING SYSTEM Explosive Gas Monitoring System F* 0xygen Monitor (Process) 1 ** 49

2. Condenser Evacuation System

a. Condenser Air Removal System Discharge Header Noble Gas Activity Monitor 1

47

$ b. Flow Rate Monitor 1

46

c. Sampler Flow Rate Monitor 1

46

3. Unit Vent

a. Noble Gas Activity Monitor 1

47

b.

Iodine Em;im 6 )e, 7 1 S1

c. Particulate Eno;;ur bL 4 1 51 m u)

d. Flow Rate Monitor

1 46 f h>

< r. o

e. Sampler Flow Rate Monitor 1 46 5 $y a

a ( i n 4 , E a s O ' O Q TABLE 4.3-9 $ RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS Y _, A nto G 6r p DIGITAL R CHANNEL MODES FOR WHICH , CHANNEL SOURCE CHANNEL OPERATIONAL SURVEILLANCE c- INSTRUMENT CHECK CHECK CALIBRATION TEST IS REQUIRE 0 i'i

1. GASEOUS WASTE PROCESSING SYSTEM P Explosive Gas Monitoring System Oxygen Monitor (Process) D N.A. Q M

2. Condenser Evacuation System

a. Consenser Air Removal System Discharge Header Noble Gas 3 1 m

1 Activity Monitor D M R(/) Q(f)

b. Flow Rate Monitor D N.A. R Q

c. Sampler Flow Rate Monitor D N.A. R Q

3. Unit Vent

.2. 3 *

a. Noble Gas Activity Monitor D M R(J) Q(J)

b. Iodine NantturSQer -ft V -$t n. ( h u. 4 M. A.

9W -r q, 74. 9 , N. A, * -

c. Particulate Man ++ w S % gley w

d. Flow Rate Monitor D N.A. R Q g g

e. Sampler Flow Rate Monitor D N.A. R Q M5 3 Ro l N @

s ? PROOF & REVIEW COPY TABLE 4.3-9 (Continued) 6 i ATTACHMENT i ST.HL.AE /9W TABLE NOTATIONS PAGE 9 OF N i At all times. M s.edebese g During WA5TE GA5 E LDU% SYSTEM operation. (1) I DIGITAL CHANNEL OPERATIONAL TEST shall also demonstrate that automatic isolation of this pathway and control room alarm annunciation occurs if any of the following conditions exists:

a. InstrumentindicatesmeasuredlevelsabovetheAlarm/TripSetpoint,or/

b. Monitor failure, or Alcr LRE D

c. Instrument indicates a downscale failure, or

( Instrument controls not set in operate mode. (2) The DIGITAL CHANNEL OPERATIONAL TEST shall also demonstrate that control room alarm annunciation occurs if any of the following conditions exists:

a. Instrument indicates measured levels above the Alarm Setpoint, or

b. Monitor failure,

c. 'i ++ xt :-3rt : mrt- sut tw up=u.eg mode- (CF %di

ly).

(3) The initial CHANNEL CALIBRATION shall be performed using one or more of the reference standards certified by the National Bureau of Standards (NBS) or using standards that have been obtained from suppliers that participate in measurement assurance activities with NBS. These standards shall permit calibrating the system over its intended range of energy and measurement range. For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration shall be used, l (4) "The CHANNEL CALIBRATION shall include the use of standard gas samples l containing a nominal: I

a. One volume percent hydrogen, balance nitrogen, and

b. Four volume percent hydrogen, balance nitrogen.

(5) The CHANNEL CALIBRATION shall include the use of standard gas samples containing a nominal- 1 1

a. One volume percent oxygen, balance nitrogen, and l NWumb

b. Four volume percent oxygen, balance nitrogen THIS PAGE OPEN PENDING RECEIPT OF O INFORMATION FROM THE APPLICANT SOUTH TEXAS - UNIT 1 3/4 3-82

ATTACHMENT q . ST HL AL $4 PAGE_i OFJ3 _ 3/4.4. Reactor Coolant System A. 4.4.1.2.2, page 3/4 4-2: Replaced 17% with 10% narrow range in accordance with letter ST-HL-AE-1901 (secondary side water level). B, 4.4.1.3.2, page 3/4 4-4: See previous comment. C. 3.4.1.4.1.b, page 3/4 4-5: See previous comment. D. 3.4.6.1 ACTION, page 3/4 4-19: Replace ACTION with Insert on page 3/4 4-19A in accordance with letter ST-HL-AE-1901. E. 3.4.6.2.e*, page 3/4 4-20: Note added because it is not possible to test valves at full system pressure prior to entry into Mode 4 since full system pressure cannot be attained until Mode 4. F. 4.4.6.2.1.a, page 3/4 4-21: Revised to reflect STP Specific Nomenclature. G. 4.4.6.2.1.c, page 3/4 4-21: (1) Deleted as the result of adding note (see E above) to page 3/4 4-20. (2) STP Specific Data provided. H. 4.4.6.2.1.d, page 3/4 4-21: STP Specific Data provided. I. 4.4.6.2.2.d, page 3/4 4-21: The design of the RHR suction piping at STP is such that in order to test one RHR suction isolation valve, the other must be opened. As a result, it is impossible to finish testing the valves as required by surveillance 4.4.6.2.2.d. Furthermore, it is our position that the surveillance is written primarily to address check valves, not MOVs. Although this surveillance provides positive assurance that the check valves have seated, valve position indica-tion provides the same degree of confidence on the RHR MOVs. This exception is consis-tent with that already approved for the Callaway Plant. J. 3.4.8. ACTION a, page 3/4 4-26: Removed the reference to " time" since Table 3.4-1 (page 3/4 4-22) no longer has time associated with it. K. 3.4.9.2.c, page 3/4 4-35: STP specific values provided and justified in letter ST-HL-AE-1867. L. Figure 3.4-4, page 3/4 4-37: Stated figure submitted in letter ST-HL-AE-1930. L3/NRC/cm ATTACHMENT ? - ST.HkAE 19t4 .j$4ct GL OF 13 M. 3.4.11, page 3/4 4-40: Change " Reactor Coolsnt System vent path" to " Reactor Vessel Head vent path" to reflect STP nomenclature. N. 3.4.11.a b, and c, page 3/4 4-40: Deleted to reflect STP design. O. 3.4.11. ACTION a and b, page 3/4 4-40: Modified LCO to reflect STP design. P. 4.4.11.1, page 3/4 4-40 Deleted and modified since LCO requires valves to be closed. L3/NRC/cm pon,e o ~ ATTA'CM(EqVitW COPY REACTOR COOLANT SYSTEM . ST HL AEWY ,q HOT STANDBY ( PAGE 3 0F - ~ 13 ~ ~ - U LIMITING CONDITION FOR OPERATION 3.4.1.2 At least two of the reactor coolant loops listed below shall be OPERABLE with two reactor coolant loops in operation when the Reactor Trip System breakers are closed and one reactor coolant loop in operation when the Reactor Trip System breakers are open:*

a. Reactor Coolant Loop A and its associated steam generator and reactor coolant pump,

b. Reactor Coolant Loop B and its associated steam generator and reactor coolant pump,

c. Reactor Coolant Loop C and its associated steam generator and reactor coolant pump, and

d. Reactor Coolant Loop D and its associated steam generator and reactor coolant pump. --

APPLICABILITY: MODE 3. INFORMATION FROM THE APPUCANT ACTION:

a. With less than the above required reactor coolant loops OPERABLE, restore the required loops to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be .

in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b. With only one reactor coolant loop in operation and the Reactor Trip System breakers in the closed position, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months open the Reactor Trip System breakers.

c. With no reactor coolant loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required reactor coolant loop to operation.

SURVEILLANCE REOUIREMENTS 4.4.1.2.1 At least the above required reactor coolant pumps, if not in operation, shall be determined OPERABLE once per 7 days by verifying correct breaker alignments and indicated power availability. 4.4.1.2.2 The required steam generators shall be determined OPERABLE by verifying secondary side water level to be greater than or equal to [EG at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . IC% %t tem mgt, 4.4.1.2.3 The required reactor coolant loops shall be verified in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

All reactor coolant pumps may be deenergized for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months provided:

(1) no operations are permitted that would cause dilution of the Reactor Coolant System boron concentration, and (2) core outlet temperature is maintained at least 10 F below saturation temperature. SOUTH TEXAS - UNIT 1 3/4 4-2 . ST.HL AE lit 1 PAGEtl OF f3 REACTOR COOLANT SYSTEM p HOT SHUTDOWN LJ LIMITING CONDITION FOR OPERATION 1 l 4.4.1.3.1 The required reactor coolant pump (s) and/or RHR pump (s), if not in 1 operation, shall be determined OPERABLE once per 7 days by verifying correct , breaker alignments and indicated power availability. . 4.4.1.3.2 The required steam generator (s) shall be determined OPERABLE by verifying secondary side water level to be greater than or equal to EHSE at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . io7ewerwrangt 4.4.1.3.3 At least one reactor coolant loop, or one RHR loop with valve CV0-198 limited to 125 gpm shall be verified in operation and circulating reactor cool-ant at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . O ililS PAGE OPEN PENDING RECEIPT OF INFORMATION FROM THE APPUCANT 4 1 O SOUTH TEXAS - UNIT 1 3/4 4-4 ATTACHMENT -) ST.HL AO 19N PAGE f OL /3__ REACTOR COOLANT SYSTEM F & REVIEW COPY [] v COLD SHUTDOWN - LOOPS FILLED LIMITING CONDITION FOR OPERATION 3.4.1.4.1 At least one residual heat removal (RHR) loop shall be OPERABLE and in operation *, and either:

a. One additional RHR loop shall be OPERABLE **, or

b. The secondary side water level of at least two steam generators shall be greater than EE25. Iofo natteus ranga.

APPLICABILITY: MODE 5 with reactor coolant loops filled ***. ACTION:

a. With two of the RHR loops inoperable and with less than the required steam generator water level, immediately initiate corrective action to return one of the inoperable RHR loops to OPERABLE status or restore the required steam generator water level as soon as possible.

b. With no RHR loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required RHR loop to operation.

L) SURVEILLANCE REOUIREMENTS 4.4.1.4.1.1 The secondary side water level of at least two steam generators when required shall be determined to be within limits at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . 4.4.1.4.1.2 At least one RHR loop shall be determined to be in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

The RHR pump may be deenergized for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months provided: (1) no operations are permitted that would cause dilution of the Reactor Coolant System boron concentration, and (2) core outlet temperature is maintained at least 10 F below saturation temperature.

- Two RHR loops may be inoperable for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing provided the other RHR loop is OPERABLE and in operation.

- - A reactor coolant pump shall not be started with one or more of the Reactor Coolant System cold leg temperatures less than or equal to 350 F unless the secondary water temperature of each steam generator is less than 50 F p above each of the Reactor Coolant System cold leg temperatures.

U THIS PAGE OPEN PENDING RECEIPT OF SOUTH TEXAS - UNIT 1 3/4 4-5 INFORMATION FROM THE APPLICANT ATTACHMENT 7 ST.HL.AE mi PAGE 4 OF T3 PHOOF& REVlEW COPY REACTOR COOLANT SYSTEM /7 3/4.4.6 O REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION SYSTEMS LIMITING CONDITION FOR OPERATION 3.4.6.1 The be OPERABLE: following Reactor Coolant System Leakage Detection Systems shall a. The Containment Atmosphere Gaseous Radioactivity Monitoring Sy b. The Containment Normal Sump Level and Flow Monitoring System, and c. The Containment System. Atmosphere Particulate Radioactivity Monitoring APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: See, Ned hg e. 3 W With only two of the above required Leakage Detection Systems OPERABLE , ment atmosphere are obtained and analyzed at least o required Gaseous or Particulate Radioactive Monitoring System is inoperab otherwise, be in at least HOT STANOBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTOOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .J SURVEILLANCE RE00_IREMENTS 4.4.6.1 The Leakage Detection Systems shall be demonstrated OPERABLE by: c. Containment Atmosphere Gaseous and Particulate Monitoring Systems OPERATIONAL TEST at the frequencies specified i b. Containment Normal Sump Level and Flow Monitoring System performance of CHANNEL CALIBRATION at least once per 18 months. e O SOUTH TEXAS - UNIT 1 3/4 4-19 ~ ' ' ATTACHMENT 7 C ST HL AE. IM - ' ~ " ,PAGE r1 OF /3 -- hGVl5e~.0 RCrood St 'Ec.nn cAnc~' 3. +. G. / a. With a. or c. of the above required Leakage Detection Systems inoperable, ooeration may continue for up to 30 days provided grab samples of the containment a*Josonere are obtained and analyzed for gaseous and particulate radioactivity at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when the required Gaseous or Particulata Radioactivity Monitoring System is inoperable; otherwise, be in at least the HOT following 30 STANOBY hours. within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDO b. With b, of the above required Leakage Detaction Systems incoerable be in at least HOT following STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOW 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . - With a. anc . of the above recuired Leakage Detaction Systems inoceracle: 1) Rest:rt eit::er Monitoring Systam (a. or :.) :: OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months anc

2)

Obtain and maly:e a grab samole of the containment atmosonere for gaseous and T., articulate racicactivity at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , and 3) Perf:rm per a Reactor Coolant System watar inventary balance at least once S hours. - Otherwise, be in at least HOT STANOBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTOCWN within :ne following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . i 1 I l PROOF & REVIEW cnPY ATTACHMENT 7 REACTOR COOLANT SYSTEM i ST.HL AE.19fY

PAGE f OF /3

~ OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION 3.4.6.2 Reactor Coolant System leakage shall be limited to:

a. No PRESSURE BOUNDARY LEAKAGE,

b. 1 gpm UNIDENTIFIED LEAKAGE,

c. 1 gpm total reactor-to-secondary leakage through all steam generators and 500 gallons per day through any one steam generator,

d. 10 gpm IDENTIFIED LEAKAGE from the Reactor Coolant System, and

e. 0.5 gpm leakage per nominal inch of valve size up to a maximum of 5 gpm at a Reactor Coolant System pressure of 2235 1 20 psig from any Reactor Cool Table 3.4-1.gntSystemPressureIsolationValvespecifiedin APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION: ( O a. With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT STANDBY within 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 .

b. With any Reactor Coolant System leakage greater than'any one of the above limits, excluding PRESSURE BOUNDARY LEAKAGE and leakage from Reactor Coolant System Pressure Isolation Valves, reduce the leakage

rate to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at least HOT STANDBY i 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 .

c. With any Reactor Coolant System Pressure Isolation Valve leakage i ,, greater than the above limit, isolate the high pressure portion of the affected system from the low pressure portion within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months by i

use of at least two closed manual or deactivated automatic valves, 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 l SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . L W n 22Sf t 4 4 f(tillAXE1 (15 4 te o 1S0 f15 to .uss obal h t.Aabe le&p &p % 4 Mk, ph4\) pop Ami&d b test psuh(depewe l

ss=gL\f ene- r puer.

SOUTH TEXAS - UNIT 1 3/4 4-20 no~ , . _ _ _ REACTOR COOLANT SYSTEM ATTAUMDR pCVitW UUPT ^ OPERATIONAL LEAKAGE .E F G) SURVEILLANCE REQUIREMENTS (Continued) 4.4.6.2.1 each of the above limits by: Reactor Coolant System leakages n sha

a. s g& ;h d Monitoring the [ ntainment M mosphere gaseous /or{ jd radieed4v4ty g at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ; particulate b.

Monitoring least once per 12 thehours; containment normal sump inventory and dis c. Measurement seals when the Reactor of the CONTROLLED Coolant System pressure LEAKAGE is to the r 2235 1 20 psig at least once per 31 days with the modulating valve The fully open provisions of Specification 4.0.4 are not applicable for entry MODE 3 or 4; c.aK Performance least once per 72 of a Reactor hours; and Coolant System water inventory at ba d_ /. Monitorin 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ,g the Reactor Head Flange Leakoff System at least once 4.4.6.2.2 pd Table its limit: 3.4-1 shall be demonstrated OPERABLE by n a. At least once per 18 months, b. Prior to entering MODE 2 whenever the plant has been in COLD SHUTDOWN for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or more and if leakage testing has not performed in the previous 9 months, c. Prior to returning the valve to service following maintenance i repair or replacement work on the valve, and , d. Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following valve actuation due to automatic or ma e

e. action or flow through the valvecu pT* v&es w3 ,

As outlined in the ASME Code,Section XI, paragraph IW-3427(b . The or 4. provisions of Specification 4.0.4 are not applicable for entry in O b V SOUTH TEXAS - UNIT 1 i 3/4 4-21 ~, , _ , _ , -- - - - - - - _ - , - . - - , ~ - . , , - . - , - . . , , - - - - n.. --n,. _ _ _ . -, - , -- , - . . m PROOF & Eview onno ATTACHMENT 7 REACTOR COOLANT SYSTEM l ST.HL AE. l'h PAGE 10 OF /3 3 3/4.4.8 SPECIFIC ACTIVITY [O LIMITING CONDITION FOR OPERATION 3.4.8 The specific activity of the reactor coolant shall be limited to:

a. Less than or equal to 1 microcurie per gram DOSE EQUIVALENT I-131, and

b. Less than or equal to 100/I microcuries per gram of gross radioactivity.

APPLICABILITY: MODES 1, 2, 3, 4, and 5. ACTION: MODES 1, 2 and 3*:

a. With the specific activity of the reactor coolant greater than 1 microcurie per gram DOSE EQUIVALENT I-131 for more than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months during one continuous time interval, or exceeding the M limit line shown on Figure 3.4-1, be in at least HOT STANDBY with T avg less than 500*F within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; and

b. With.the gross specific activity of the reactor coolant greater than p 100/E microcuries per gram, be in at least HOT STANDBY with T avg less than 500'F within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , MODES 1, 2, 3, 4, and 5:

With the specific activity of the reactor coolant greater than 3 microcurie per gram DOSE EQUIVALENT I-131 or greater than 100/E micro-Curies per gram perform the sampling and analysis requirements of Item 4.a) of Table 4.4-4 until the specific activity of the reactor coolant is restored to within its limits. "With T,yg greater than or equal to 500'F. SOUTH TEXAS - UNIT 1 3/4 4-26 ATTACHMENT 'l ST HL AE 1%4 PAGE If OF I3 REACTOR COOLANT SYSTEM NM t /'~3 j PRESSURIZER LIMITING CONDITION FOR OPERATION 3.4.9.2 The pressurizer temperature shall be limited to:

a. A maximum heatup of 100'F in any 1-hour period,

b. A maximum cooldown of 200*F in any 1-hour period, and ne k v

c. A maximumAspraf water temperature differential of EEEl'F.

APPLICABILITY: At all times. ACTION: With the pressurizer temperature limits in excess of any of the above limits, restore the temperature to within the limits within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the pressurizer; determine that the pressurizer remains acceptable for continued operation 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 reduce the pressurizer pressure to less than 500 psig , within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . O) (, SURVEILLANCE REQUIREMENTS 4.4.9.2 The pressurizer temperatures shall be determined to be within the limits at least once per 30 minutes during system heatup or cooldown. The spray water temperature differential shall be determined to be within the limit at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during auxiliary spray operation. O SOUTH TEXAS - UNIT 1 3/4 4-35 ATTACHMENT 9 ST HL AE tilY RGEP AF 13 f9M a N;VitW UUPT O O . THIS PAGE OPEN PENDING RECEIPT OF INFORMATION FROMTHE APPLICANT A FIGURE 3.4-4 NOMlHAL MAXIMUM ALLOWABLE PORV SETPOINT FOR THE COLD OVERPRESSURE SYSTEM SOUTH TEXAS - UNIT 1 3/4 4 37 ATTACHMENT rj ST.HL AE l%V ,pMCd,.ll QF L 3 . REACTOR COOLANT SYSTEM "~ = nevrtW COPY 3/4.4.11 REACTOR COOLANT SYSTEM VENTS Q(% LIMITING CONDITION FOR OPERATION

5624 %o vesse\ R..A each at Wt 600 3.4.11 R=MPR=ne Reactor C;;i=a 5y : = vent pathwonsisting of trauzS. vent valves (s) - [on-] nin a veke powered from emergency busses shall be OPERABLE and closed ei -- h ef=the fei N : .g luutiem

a. Pm t _ -::-^' '- - 9 ,

b. P w ; w ste;. = c]. and C. [WeeUnr inniani s stem y ninn n o i n t },

APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: %. \ bd

a. With one of the above Reactor Cecied SyEtse vent paths inoperable, STARTUP and/or POWER OPERATION may continue provided the inoperable vent path is maintained closed with power remove from the valve actuator of all the vent valvescana biock valve in the inoperable vent path; restore the inoperable vent path to OPERABLE status within 30 days, or, be in HOT STANDBY within 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 .

, O bon % ul Head With in er-moee Reactor Costant System vent paths inoperable; U/ b. , maintain the inoperable vent path :losed with power removed from the valve actuators of all the vent valves and block valves in the inoperable vent paths, and restore at least [two] of the vent paths to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in HOT STANDBY within 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 REQUIREMENTS b11.1 Each Reactor Coolant System vent path block valve not required to be closed by ACTION a or b., above, shall be demonstrated OPERABLE at least 'once per 92 days by operating the valve through one complete cycle of full travel from the control room.f ssse\ Hed 4.4.11.fi Each Reactor Coolant SFItEm vent path shall be demonstrated - OPERABLE at least once per 18 months by:

a. Verifying all manual isolation valves in each vent path are locked in the open position,

b. Cycling each vent valve through at least one complete cycle of full travel from the control room, and M e\ Head

c. Verifying flow through the Reactor Coutant: System vent paths during O venting.

SOUTH TEXAS - UNIT 1 3/4 4-40 ATTACHMENT J . ST HL AE ith PACE 1 0F f_ _. 3/4.5 Emergency Core Cooling Systems l A. 3/4 5-1: (1) NRC Staff accepted STP markup in Proof and Review; however, based upon justification provided in letter ST-llL-AE-1867, ACTION a.should require 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and this ACTION applies for water volume and/or nitrogen pressure, not boron concentration. l (2) Justification for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months provided in letter ST-ilL AE-1923. B. 3/4 5-2: Delete ANALOC CllANNEL OPERATIONAL TEST. This has been approved at other plants. Typically, ACOTs require testing of bistables; this is not appropriate for alarm testing. Channel calibration should be adequate. C. 3/4 5-5: STP specific data provided. D. 3/4 5-6: Justification provided in ST-HL-AE-1897. k L3/NRC/cm C._______________________________________________________ ATTACHMENT 8 . ST.HL-AE- M PAGE @ OF 5 3/4.5 EMERGENCY CORE COOLING SYSTEMS & GEW COPY 3/4.5.1 ACCUMULATORS LIMITING CONDITION FOR OPERATION 3.5.1 Each Safety Injection System accumulator shall be OPERABLE with:

a. The isolation valve open and power removed,

b. A contained borated water volume of between 8800 and 9100 gallons,

c. A boron concentration of between 2400 and 2600 ppm, and

d. A nitrogen cover pressure of between 590 and 670 prig.

APPLICABILITY: MODES 1, 2, and 3*. ACTION: g oQ 4/,

a. With one accumulator inoperable except as a result of a closed isolation valve or the baron 6 outside the required limits, restore the inoperable accumulator to OPERABLE status with

[a 4 - @rhour or be in at least HOT STANOBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce pressurizer pressure to less than 1000 psig within the follow-ing 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . ! b. With one accumulator inoperable due to the isolation valve being i closed, either open the isolation valve within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at } 1 east HOT STANDBY with 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce pressurizer pressure to

less than 1000 psig within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

j c. With the boron concentration of one accumulator outside the required limit, restore the boron concentration to within the required limits within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months i and reduce pressurizer pressure to less than 1000 psig within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . l

- SURVEILLANCE RFOUIREMENTS l

4.5.1.1 Each accumulator shall be demonstrated OPERABLE:

a. At least once per hours by:

o Verifying, by the absence of alarms, the contained borated 1) j water volume and nitrogen cover-pressure in the tanks, and

2) Verifying that each accumulator isolation valve is open.

! b. At least once per 31 days and within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after each solution { volume increase of greater than or equal to 1% of tank volume by i verifying the boron concentration of the accumulator solution; and

Pressurizer pressure above 1000 psig.

SOUTH TEXAS - UNIT 1 3/4 5-1 ATTACHMENT f . ST.HL AL WW PAGE 3 OF f + _ EMERGENCY CORE COOLING SYSTEMS PROOF & REVIEW COPY { t , (/ SURVEILLANCE REQUIREMENTS (Continued) s # c. At least once per 31 days when the RCS pressure n 4Gv[ 1000 /ps by verifying that power to the isolation valve r/perator .is discon ^ nected by removal:of the breaker from the circuit. ., d. c; - J, . , _ s At least once,per:18 months by verifying tha* o

1)

When an actual or a simulated Rt.s pressuie signal exceeds the P-11 and (Pressurizer Pressdr~e Glock of Safety Injection) Setpoint, 2) Upon receipt of a Safety Injection test signal. 4.5.1.2 strated OPERABL Each accumulator water level and pressure ~ channel shall be demon- - # At least N m Nm p 21 %, and ty-the p ri= =ce at au ANAt06 GMANN64, mMfst TM 'l A+ Qt least once per 18 months 4 the performance of a CHANNEL CALIBRATION, Q o ' P G i .l Q . ? SOUTH TEXAS - UNIT 1 '3/4 5-2 _,--w-,ww----wmr ,-------e,- em -- -,,-r- r - - - - - - - - - - - , - - - - , - - - , - - ,-,w-c-- - - ----y --,---,w.,- -, , ATTACHMENT F ST-HL-AE M [ PAGE 4 OF f EMERGENCY CORE COOLING SYSTEMS PROOF & REVIEW COPY ,a \ ] f. SURVEILLANCE REOUIREMENTS (Contguedl j e. At least once pir-18 fronths, during shutdown, by: ./ 1) .- Verifying that each automatic valve in the flow path actuates a to~iu currect position on (Safety Injection actuation and f' Automatic iwitchover to Containment Sump) test signals, and 2) 'icritying that each of the following pumps start automatically upon receipt of a Safety Injection actuation test signal: a) High Head Safety Injection pump, and b) Low Head Safety Injection pump. f. By verifying that each of the following pumps develops the indicated differential pressure Specification 4.0.5: on recirculation flow when tested pursuant to

1) High Head Safety Injection pump 1 1445 psid, and

2) Low Head Safety Injection pump 2 283 psid.

g. C) By performing a flow test, during shutdown, following completion of 'v modifications tc the ECCS subsystems that alter the subsystem flow characteristics and verifying that: 4 1) For High Head Safety Injection pump lines, with the High Head 4 Safety Injection pump running, the pump flow rate is greater than g gpm and less than it,c0 gpm.

2) For to:(Head Safety Injection pump lines, with the Low Head Safety Injection pump running, the pump flow rate is greater than 22 gpm and less than 2PCO gpm.

i THS PAGE OPEN PENDING RECEIPT OF jinV0RMATION FROM THE APPUCANT 9 9 SOUTH TEXAS - UNIT 1 3/4 5-5 .= - . . . . . , _ _ _ .. ST.HL AE.l9N g PAGE r OF C ~ EMERGEhtY CORE COOLING SYSTEMS PROOF & REVIEW COP 3/4.5.3 ECCS SUBSYSTEMS - T,yg LESS THAN 350'F LIMITING CONDITION FOR OPERATION i 3.5.3.1 As a minimum, the following ECCS components shall be OPERABLE:

a. Two OPERABLE High Head Safety Injection pumps,*

b. Two OPERABLE Low Head Safety Injection pumps and their associated RHR heat exchangers, and

c. Two OPERABLE flow paths capable of taking suction from the refueling water storage tank upon being manually realigned and transferring suction to the containment sump during the recirculation phase of operation through a High Head Safety Injection pump and into the Reactor' Coolant System and through a Low Head Safety Injection pump and its respective RHR heat exchanger into the Reactor Coolant System.

APPLICABILITY: MODE 4. ACTION:

a. With less than the above-required ECCS components OPERABLE because st. ,

the ' ep=4'"ty of stther the .Migh 14 sad. 33* sty injul':+ pumps- or-O %e ttow pathe f*om the fT9ee%4ng water sespege tank, restore at least the required ECCS components to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in COLD SHUTDOWN within the next 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months . )r'. Ethlessthantheabove-requiredECCScomponentsOPERABLEbecauseo the inoperability of either the residual heat removal heat exchangers l l or the Low Head Safety Injection pumps, restore at least the required ECCS components to OPERABLE status or maintain the Reactor Coolant i System T less than 350*F by use of alternate heat raanval method ., b[' - In the event the ECCS is actuated and injects water into the Reactor Coolant System, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 90 days describ-m fng the circumstances of the actuation and the total accumulated , actuation cycles to date. The current value of the usage factor for each affected Safety Injection nozzle shall be provided in this Special Report whenever its value exceeds 0.70, i.

A maximum of one High Head Safety Injection pump shall be OPERABLE within O 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after entering MODE 4 from MODE 3 or prior to the temperature of one or more of the RCS cold legs decreasing below 325'F, whichever comes first.

I SOUTH TEXAS - UNIT 1 3/4 5-6 t I ? ATTACHMENT 9 ST.HL-AE #4 PAGE / OP 6 n 3/4.6 Containment Systems A. 4.6.1.6.1, page 3/4 6-10: Changes made to this page reflect STP design and are justified in letter ST-HL-AE-1883. B. 4.6.2.2.d.1, page 3/4 6-16: (1) Added minimum flow of 30 gpm based upon preoperational testing data. (2) Deleted " throttled to gpm at" to reflect STP design of using pump suction pressure. C. 4.6.2.3.a.2, page 3/4 6-17: Provided STP specific data. D. 4.6.4.1, page 3/4 6-32: Delete surveillance a and move b to the body. The STP Hydrogen Analyzers analyze specifically for hydrogen; i.e., the measuring cell will not generate an output current unless hydrogen is present in the sample gas. Therefore, the instrument has an absolute zero, and no zero gas is required to standardize the analyzer. The measuring cell has the ability.of. accurately responding te the presence of hydrogen, irrespective of flowrate. The analyzer is calibrated using atmospheric-air as a span gas, providing one of the three ranges of analysis encompass the hydrogen content of air. Actual calibra-tion is accomplished by withdrawing the cell probe from its holder, thus exposing the cell to the surrounding atmosphere. Therefore, the changes made reflect STP design. Note the Hydrogen Analyzer Manual was provided to the NRC Staff on March 9, 1987 for review. E. 4.6.4.2.a, page 3/4 6-33: Specific data provided. L3/NRC/cm r l ATTACHMENT 9 . ST HL-AE 198+ i pdiME o2,gth, , _ _ i -w u ncvirYY LUr'T CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) i

b. Performing tendon detensioning, inspections, and material tests on a previously stressed tendon from each group (inverted U and hoop).

A randomly selected tendon from each group shall be completely detensioned in order to identify broken or damaged wires and deter-mining that over the entire length of the removed wire or strand i that: be evfC Me. enL-

1) The ten n wires "r dr-uk or da hh e d g egrrgsion, cracks, W15dmM %

g saldamage'p cwe @ GAME. d h mYsg fe.

2) There are not changes in the presence or physical appearance of ;

the sheathing filler grease, and l

3) A minimum tensile strength of 240,000 psi (guaranteed ultimate strength of the tendon material) for at least three wire or N samples (one from each end and one at mid-length) cut from each removed wire c: : t: ;:2. Failure of any one of the wire on=stesad samples to meet the minimum tensile strength test is evidence of abnormal degradation of the containment structure.

g gg ;%

c. ee-Performing tendon retensioning of those tendons etensioned % for hue

(" inspection to their observed lift-off force with a tolerance limit of +6%. During retensioning of these tendons, he. changes in load and elongation should be measured simultaneously ^ - ' ' r ef thee- appm- %tdy eMy 3+-Havais of furce n. -re zese ami Stur seau g frrwe If the elong& tion corresponding to a specific load ditters by more than 5% from that recorded during installation, an inrt ,tigation should be made to ensure that the difference is not related to wire failures or slip of wires in anchorages;

d. Assuring the observed lift-off stresses exceed the average minimum design value given below, which are adjusted to account for elastic =nNme-4e % losses; and

!!,9 Inverted U $1391 ksi Cyl hder $3423itd Hoo@toma bertd (1R]M .tsi N.

e. Verifying the OPERABILITY of the sheathing filler grease by:

1) No voids in excess of 5% of the net duct volume

) E) Minimum grease coverage exists for the different parts of the anchorage cystem, and 2#) The chemical properties of the filler material are within the tolerance limits as specified by the manufacturer. O SOUTH TEXAS - UNIT 1 3/4 6-10 ATTACHMENT 9 - . ST HL-AE L984 PAGE 3 OF 6 PROOF & REVIEW COPY CONTAINMENT SYSTEMS CONTAINMENT COOLING SYSTEM LIMITING CONDITION FOR OPERATION

d. At least once per 5 years by verifying: g

1) Each eductor flow rate is greater than or equal to [ fgpm using the RWST as the test source th:ett4d ts [ ]:gpm % 4e the eductor inlet, and

2) The lines between the spray additive tank and the eductors are not blocked by verifying flow.

O THIS PAGEOPEN PENDING RECElPT OF INFORMATION FROMTHE APPLICANT I

SOUTH TEXAS - UNIT 1 3/4 6-16 ATTACHMENT 9 ST HL AE.1984- 1 pjns 4 nr 6___ j _ CONTAINMENT SYSTEMS T00F& REVIEW COP 7-U) CONTAINMENT COOLING SYSTEM H5S PAGEOPEN PWNG REC q LIMITINGCONDITIONFOROPERATION INFORMATION FROMTHEAPPUCANT 3.6.2.3 shall be OPERABLE with a minimum of two unit third group. i unit in the _ APPLICABILITY: MODES 1, 2, 3, and E. ACTION: a. With one group of the above required Reactor Containment F inoperable and all Containment Spray Systems OPERABLE an Coolers ; inoperable group of RCFC to OPERABLE status , restore the or be in within 7 d at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . . b. With two groups of the above required Reactor Containment inoperable and all Containment Spray Systems OPERABLE an Coolers , restore at least one group of RCFC to OPERABLE status ewithin in 72 h at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . RCFC to OPERABLE status within 7 days of initial at loss o least HOT STANDBY within the next 6 within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . hours an c. With one group of the above required Reactor Containment F inoperable and one Containment Spray System inoperable n Coolers inoperable Spray System to OPERABLE status ewithin , restore the in 72 h at least within HOT30STANDBY the following hours. within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months a Restore the inoperable group of containment cooling fans to OPERABLE status within 7 day loss or be in at least HOT STANDBYn within COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . n the SURVEIU ANCE REOUIRFMFNTS_ - 4.6.2.3 OPERABLE: Each group of Reactor Containment Fan Coolerse shall be d a. At least once per 31 dajs by: 1) Starting each non-operating fan group from the control room verifying and that each fan group operates for at least, and 15 minutes 2) @ gpm to each cooler. Verifying a cooling water flow rate o . b. At least once per 18 months by verifying that each automatically on a Safety Injection test signal. artsfan group s SOUTH TEXAS - UNIT 1 3/4 6-17 ATTACHMENT 9 . ST.WAE 1984 Pant 5 OFfo CONTAINMENT SYSTEMS 3/4.6.4 PROOF & REVIEW COP COM8USTIBLE GAS CONTROL HYDROGEN ANALYZERS LIMITING CONDITION FOR OPERATION 3.6.4.1 Two independent containment hydrogen analyzers sha APPLICABILITY: . MODES 1 and 2. ACTION: a. With one hydrogen analyzer inoperable, restore the inoper a eleast analyzer bl to STAN OPERABLE next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , status within 30 days or be in at HOT within the b. With both hydrogen analyzers inoperable, restore at least one analyzer to e OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . be withinin theat leas SURVEILLANCE REQUIREMENTS 4.6.4.1 Each hydrogen analyzer shall be demonstrated OPERAB of a CHANNEL CHECK at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , TEST at least once per 31 days, and at least once per 92 an r ormance ANA days o O_ L TEST BASIS by performing % a CHANNEL CALIBRATION a STAGGERED s containing using s 7t;;s ninae pweeat hydw, tatance-P+, and A our volume percent hydrogen, W_nttvogen.unk% blwe, b SA-I ( O 6 SOUTH TEXAS - UNIT 1 i 3/4 6-32 ATTACHMENT 9 ST HL-AE.1964- . PAGE G OF G - . - _ CONTAINMENT SYSTEMS i ELECTRIC HYDROGEN RECOMBINERS .O PROOF & REVIEW COPY: ' \ LIMITING CONDITION FOR OPERATION i 3.6.4.2 _ Two independent Hydrogen Recombiner Systems shall be OP . APPLICABILITY: MODES 1 and 2. ACTI0N:_ With one Hydrogen Recombiner System inoperable, restore the i to OPERABLE next S hours. status within 30 days or be in at least noperable HOTsystem STANDBY within the SURVEILLANCE REQUIRFMENTS 4.6.4.2 " Each Hydrogen Recombiner System shall be demonstrated O a. At least once per 6 months by verifying, during a Hydrogen System functional test, that the miningp er increases,to ter sheathy tp than or equal to Q70 erature Upon reaching Fwithin09 for 2 minutes an 0Q*F,increasethepower tting to max minutes. um power equalto@60gkW,andverify that the power meter reads greater than o O b. cY 6 At least once per 18 months by: ' 1) Performing a CHANNEL tion and control circuits, CALIBRATION of all recombiner in 2) Verifying through a visual examination that there is no (i.e., loose wirinevidence of abnormal conditions within the re foreign materials,getc.),and or structural connections, deposits of 3) l performing a resistance to ground test follow required functional test. heater phase shall be greater than or equal to 10,000 oh . THIS PAGE OPEN PENDING RECEIPT OF INFORMATION FROMTHE APPLICANT - O . SOUTH TEXAS - UNIT 1 3/4 6-33 - ATTACHMENT.lO ST HL AE 1984 of.GE t OF 24 3/4.7 Plant Systems A. 3/4_7-4: STP specific data provided. B. 3/4 7-5: Not all valves in the AFW flow path position are " fully open". C. 3/4 7-6: Deletion is appropriate inasmuch as there is now a Tech Spec for RHR with regard to long term core cooling. See RHR TS 3/4.5.6. See Table 1-6 in SER Supplement 2, page 1-10, and License Condition Number 2, which are resolved. D. 3/4 7-11: (1) Editorial. (2) Added " accessible" since some valves that are in containment may not be accessible once per 31 days. E. 3/4 7-13: Justification provided in letter ST-HL-AE-1882. Follow up letter provided via ST-HL-AE-1956 dated March 9, 1987. F. 3/4 7-17: STP specific value provided. G. 3/4 7-18: South Texas preoperational tests have shown that the flowrate is 31,000 cfm. H. 3/4 7-19: (1) Item d 3) should be 1/8 inch water gauge. (2) Item d 2) - clarified that you get a High Radiation and/or Safety Injection test signal. I. 3/4 7-28 to 7-41: Deletion of Fire Protection was provided in letter ST-HL-AE-1867. J. 3/4 7-43: Table 3.7-6 provided. L3/NRC/cm ATTACHMENT 10 ST.MLAE'190# PLANT SYSTEMS - NI - N CUPT 4 AUXILIARY FEE 0 WATER SYSTEM THIS PAGE OPEN PENDING RECEIPT OF LIMITING CONDITION FOR OPERATION p g

3. 7.1. 2 and associated flow paths shall be OPERABLE with:At least fou a.

Three motor-driven auxiliary feedwater powered from separate emergency busses, and pumps, each capable of being b. One steam turbine-driven auxiliary feedwater pump capable of b powered from an OPERABLE steam supply system. APPLICABILITY: MODES 1, 2, and 3. ACTION: . a. With the Train A motor-driven auxiliary feedwate as soon as possible. applicable. The provisions of Specification 4.0.4 are not b. With any of the following combinations of auxiliary feedwater pump inoperable: 1)

2) Train B or Train C motor-driven pump,

3) Train D turbine-driven pump and any one motor-driven pum

4) Train A and either Train B or Train C motor-driven pump,p, Train 0 turbine-driven pump or Restore the affected auxiliary feedwater pump (s) to OPERABLE status within 72SHUTDOWN and in HOT hours orwithin be the in following at least HOT STANDBY within the 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

c. With. Train B and Train C motor driven pumps, o and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . ,. d. l With four auxiliary feedwater pumps inoperable, immediately initiate to OPERABLE status as soon as possible. corrective action to i SURVEILLANCE REOUIREMENTS 4.7.1.2.1 Each auxiliary feedwater pump shall be demonstrated OPERABLE: a. At least once per 31 days on a STAGGERED TEST BASIS by: 1) Verifying that each motor-driven pump develops a discharge i pressure of greater than or equal to N59psig at a flow of I greater than or equal to R gpm; - 2) Verifying that the steam turbine-driven pump develops a discharge pressure of greater than or equal to /W4psig at, a flow of greater than or equal to E gpm when the secondhry steam supply pressure is greater than psig. The provisions of Specification 4.0.4 are not applica e for entry into MODE 3; SOUTH TEXAS - UNIT 1 3/4 7-4 ATTACHMENT 10 ST-HL AE- l984 OAGE 3 OF 24 PLANT SYSTEMS PROOF & REVIEW COPY SURVEILLANCE REQUIREMENTS (Continued)

3) Verifying that each non-automatic valve in the flow path that is not locked, sealed, or otherwise secured in position is in '

its correct position; and l

4) Verifying that each automatic valve in the flow path is in the ,

coned' fsth open position whenever the Auxiliary Feedwater System is l placed in automatic control or when above 10% RATED THERMAL POWER.

b. At least once per 18 months during shutdown by:

1) Verifying that each automatic valv'e in the flow path actuates to its correct position upon receipt of an Auxiliary Feedwater Actuation test signal, and

2) Verifying that each auxiliary feedwater pump starts as designed automatically upon receipt of an Auxiliary Feedwater Actuation test signal.

3) Verifying that each auxiliary feedwater flow regulating valve limits the flow to each steam generator between 550 gpm and 675 gpm. 98 4.7.1.2.2 An auxiliary feedwater flow path to each steam generator shall be demonstrated OPERABLE following each COLD SHUTDOWN of greater than 30 days prior to entering MODE 2 by verifying normal flow to each steam generator.

e f O SOUTH TEXAS - UNIT 1 3/4 7-5 \' ATTACHMENT to ' 1984 ! . pynt ~ ST.HL 4- AR'nV 2k \ PLANT SYSTEMS 1 PROOF & REVIEW COPY AUXILIARY FEEDWATER STORAGE TANK LIMITING CONDITION FOR OPERATION i 3.7.1.3 I contained water volume of at leastThe auxiliary feedwater storage t 518,000 gallons of water. i APPLICABILITY: MODES 1, 2, and 3. ACTION: \ With the AFST inoperable, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months either: a. Restore the AFST to OPERABLE status or be in at leas STANDBY the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , withinor the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTD i [b. ,' Demonstrate the OPERABILITY of the temporary hose connection to the AFST as a backup supply to th _ auxiliary feedwater pumps and restore the AFST to OPERABLE j st within 7 days or be in at least HOT STANDBY within the n and in HOT SHUTDOWN within the following 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months sy SURVEILLANCE REQUIREMENTS 4.7.1.3.1 verifying the contained water volume is within its li supply source for the auxiliary feedwater pumps. 4.7.1.3.2 the AFST shall be demonstrated OPERABLE at least y [ method dependent upon alternate source] whenever the Essentialystem Cooling Wate is the supply source for the auxiliary feedwater pumg 1 THIS PAGEOPEN PENDING RECEIPT OF INFORMATION FROM THE APPUCANT l SOUTH TEXAS - UNIT 1 3/4 7-6 l ATTACHMENT lo ST.HL-AE 1984-e pg4 _ PLANT SYSTEMS ) 3/4.7.3 COMPONENT COOLING WATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.3 At least three independent component cooling water loops shall be OPERABLE. APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: With only two component cooling water looitf0PERABLE, restore at least three loops to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . SURVEILLANCE REOUIREMENTS 4.7.3 At least three component cooling water loops shall be demonstrated OPERABLE: occe.suMe.

a. At least once per 31 days by verifying that eachAvalve (manual, power-operated, or automatic) servicing safety-related equipment that is not locked, sealed, or otherwise secured in position is in its correct position; and -

b. At least once per 18 months during shutdown, by verifying that:

, 1) Each automatic valve servicing safety-related equipment or isolating the non-nuclear safety portion of the system actuates to its correct position on a Safety Injection, Loss of Offsite Power or Low Surge Tank test signal, as applicable,

2) Each Component Cooling Water System pump starts automatically on a Safety Injection or Loss of Offsite Power test signal, and

3) The surge tank level instrumentation which provides automatic isolation of the non-nuclear safety-related portion of the sys-tem is demonstrated OPERABLE by performance of a CHANNEL CALIBRATION test.

1 O t- . SOUTH TEXAS - UNIT 1 3/4 7-11 ATTACHMENT ST.wAE 19B PAGL4 E i PLANT SYSTEMS rNUOF & REVIEW COPY 3/4.7.5 ULTIMATE HEAT SINK O V LIMITING CONDITION FOR OPERATION 3.7.5 a. The ultimate heat sink shall be OPERABLE with: USGS datum, andA minimum water level at or above eleva n Sea Level,

b. An =Eu,=id;d. Coo % m%Ye.

===_ waterAtemperature of less than or equal 99 to WF APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: g With the r+ ; '&het.tMo_-t: ut t'Ar Ep _admNe_ J ' ' -; ;"n echote T 4b CPERABLE. h M HOT STANOBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and.in COLD oSHUTDOWN owing 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . within SURVEILLANCE REQUIREMENTS 4.7.5 within their. limits.24 'Are. hours by verifying the evel to be e per::rg wa O i O SOUTH TEXAS - UNIT 1 3/4 7-13 l l [ STTACHMENT ici ST-HL-AE 1984 gE , 1QF 24 i nuur et rttVitW COPY PLANT SYSTEMS V SURVEILLANCE REOUIREMENTS (Continued)

3) Verifying that the system maintains the control room envelope at a positive pressure of greater than or equal to 1/8 inch Water Gauge at less than or equal to a pressurization flow of 2000 cfm relative to adjacent areas during system operation;

4) Verifying that the makeup filter unit heaters dissipate 4.51 0.45 kW when tested in accordance with ANSI N510-1980; and

5) Verifying that on a High Toxic Gas test signal, the system automatically switches into a recirculation mode of operation by isolating the normal supply and exhaust flow within 2646 seconds.

f. After each complete or partial replacement of a HEPA filtar bank, by verifying that the cleanup system satisfies the in pine per.atration and bypass leakage testing acceptance criteria of less than 1.0% in accordance with ANSI N510-1980 for a 00P test aerosol while operating the system at a flow rate of 6000 cfm i 10% for the cleanup units and 1000 cfm i 10% for the makeup units; and

g. After each complete or partial replacement of a charcoal adsorber bank, by verifying that the cleanup system satisfies the in place

,.,T penetration and bypass leakage testing acceptance criteria of less (O than 1.0% in accordance with ANSI N510-1980 for a halogenated hydro-carbon refrigerant test gas while operating the sy:; tem at a flow rate of 6000 cfm i 10% for the cleanup units and 1000 cfm i 10% for the makeup units. O SOUTH TEXAS - UNIT 1 3/4 7-17 ATTACHMENT 10 I , ST HL.AE.1984 ' PAGE B 0F '2+ _. _ PR005 fiiEVIEW COPP PLANT SYSTEMS n U 3/4.7.8 FUEL HANDLING BUILDING (FHB) EXHAUST AIR SYSTEM LIMITING CONDITION FOR OPERATION 3.7.8 The FHB Exhaust Air System comprised of the following components shall be OPERABLE:

a. Two independent exhaust air filter trains,

b. Three independent exhaust booster fans,

c. Three independent main exhaust fans, and j

d. Associated dampers.

APPLICABILITY: MODES 1, 2, 3, and 4. l ACTION: With less than the above FHB Exhaust Air System components OPERABLE but with at least one FHB exhaust air filter train, two FHB exhaust booster fans, two FHB main exhaust fans and associated dampers OPERABLE, restore the inoperable system to OPERABLE status within 7 days 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 4.7.8 The Fuel Handling Building Exhaust Air System shall be demonstrated OPERABLE:

a. At least once per 31 days on a STAGGERED TEST BASIS by initiating, from the control room, flow through the HEPA filters and charcoal adsorbers and verifying that the system operates for at least 10 continuous hours with the heaters operating with two of the three exhaust booster fans and two of the three main exhaust fans operat-ing to maintain adequate air flow rate;

b. At least once per 18 months or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire, or chemical release in any ventilation zone communi-cating with the system by:

1) Verifying that the cleanup system satisfies the in place penetration and bypass leakage testing acceptance criteria of less than 1.0% and uses the test procedure guidance in Regula-tory Positions C.5.a, C.5.c, and C.S.d of Regulatory Guide 1.52, Revision 2, March 1978, and the system flow rate is 3M000 cfm i 10%; 31,000

2) Verifying, within 31 days after removal, that a laboratory O analysis of a representative carbon sample obtained in accor-

\ dance with Regulatory Position C.6.b of Regulatory Guide 1.52, SOUTH TEXAS - UNIT 1 3/4 7-18 t e-- - . , - ---ne - p ATTACHMENT to ST-HL AE.198+ oAGE 9 OF 14-PRO 0f & NtVitW CUPY PLANT SYSTEMS j SURVEILLANCE REOUIREMENTS (Continued) Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revi-sion 2, March 1978, for a methyl iodide penetration of less than 1.0%; and al,000

3) Verifying a system flow rate of 37f006 cfm i 10% during system operation with two supply fans and two of the three exhaust booster and main exhaust fans operating when tested in accor- .

dance with ANSI N510-1980. All combinations of two exhaust ' booster fans and two main exhaust fans shall be tested. i

c. After every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of charcoal adsorber operation, by verifying, within 31 days after removal, that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978, for a methyl iodide penetration of less than 1.0%;

d. At least once per 18 months by:

1) Verifying that the pressure drop across the combined HEPA

, filters and charcoal adsorber banks is less than 6 inches Water Gauge while operating the system at a flow rate of N 3),000 cfm i 10%, agWLbm eA/oc

2) Verifying that the system starts on agafety Injection test

, signal,

3) Verifying that the system maintains t at a negative pressure of greater than or equal to inch Water Gauge relative to the outside atmosphere, an

4) Verifying that the heaters dissipate 50 t 5 kW when tested in accordancewithANSINg0 g

e. After each complete o pakialreplacementofaHEPAfilterbank, by verifying that the leanup system satisfies the in place pene-tration and bypass leakage testing acceptance criteria of less than

, 1.0% in accordance with ANSI N510-1980 for a DOP test aerosol while operating the system at Q w g e g g cfm i 10%; and

f. After each complete or partia replacemenf. of a charcoal adsorber bank, by verifying that the leanup system satisfies the in place penetration and bypass leakage testing acceptance criteria of less than 1.0% in accordance with ANSI N510-1980 for a halogenated ,

hydrocarbon refrigerant test gas while operating the system at a t i flow rate of iir SG u cfm i 10%. 36,o00 THIS PAGE OPEN PENDING RECEIPT OF INFORMATION FROMTHE APPLICANT O SOUTH TEXAS - UNIT 1 3/4 7-19 M i i /wl IlviLil l to ST-HL AE 8984 l T~ E Io OF 2A be\da PLANT SYSTEMS PROOF & REVIEW COPY ! (^') v 3/4.7.11 FIRE PROTECTION SYSTEMS 1 FIRE PROTECTION WATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.11.1 The Fire Protection Water System shall be OPERABLE with: f

a. At least three fire pumps, each with a capacity of 2500 gpm, with their discharge aligned to the underground piping ring main, j

b. Two separate water supplies, each with a minimum contained volume of 270,000 gallons, and 1

c. An OPERABLE flow path capable of taking suction from the number 1 fire water storage tank and the number 2 fire water storage tank and transferring the water through distribution piping with OPERABLE sectionalizing control or isolation valves to the yard hydrant curb valves, the last valve ahead of the water flow alarm device on each sprinkler or hose standpipe, and the last valve ahead of the deluge valve on each Deluge or Spray System required to be OPERABLE per Specifications 3.7.11.2, 3.7.11.5, and 3.7.11.6.

APPLICABILITY: At all times. N',.) ACTION:

a. With one pump and/or one water supply inoperable, restore the inoper-able equipment to OPERABLE status within 7 days or provide an alter-nate backup pump or supply. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

b. With the Fire Protection Water System otherwise inoperable, establish a backup Fire Protection Water System within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

b) v SOUTH TEXAS - UNIT 1 3/4 7-28 i . ST-HL-AE- 19 84-N jEiy^,6F) ; i PLANT SYSTEMS g () SURVEILLANCE REOUIREMENTS 4.7.11.1.1 l The Fire Protection Water System shasl be demonstrated OPERAB a. At least volume, once per 7 days by verifying the contained water supply b. At least once per 31 days by verifying that each valve (manual, operated, or automatic) in the flow path is in its correct position, c. At least once per 6 months by performance of a system flush, d. At least once per 12 months by cycling each testable valve in the flow path through at least one complete cycle of full travel, t e. At least once per 18 months by performing a system functional test which includes its operating simulated sequence, and: automatic actuation of the system throu 1) Verifying that position, to 1ts correct each automatic valve in the flow path actuates 2) Verifying that each pump develops at least 2500 gpm at a system head of 289 feet, O O 3) plant operation through at least one complete c travel, and 4) Verifying that each fire pump starts sequentially to maintain to 125 psig.the Fire Protection Water System pressure greater tha f. At least once per 3 years by performing a flow test of the system in accordance with Chapter 5, Section 11 of the Fire Protection Handbo 14th Edition, published by the National Fire Protection Association. G , b a SOUTH TEXAS - UNIT 1 3/4 7-29 -. . . .-- - .~..

A'TACHMENT 10 DN

_ PLANT SYSTEMS 0-. mN I u _. . _ -.- / SURVEILLANCE REQUIREMENTS (Continued) 4 4.7.11.1.2 The fire pump diesel engine shall be demonstrated OPERABL a. At least once per.31 days by verifying: 1) The and fuel storage tank contains at least 400 gallons of fuel , 2) The diesel starts from ambient conditions and operates fo least 30 minutes on recirculation flow.

b. 1 At least once per 92 days by verifying tha is within the acceptable limits specified in Table 1 of ASTM when checked for viscosity and water and sediment; and 0975 1977 c.

At least once per 18 months, during shutdown, by subjecti to an inspection in accordance with procedures prepared on in 4.7.11.1.3 with its manufacturer's recommendations for tha c . shall be demonstrated OPERABLE:The fire pump diesel starting 24 a. At least once per 7 days by verifying that: 1) The electrolyte level of each battery is above the plates , and 2) The overall battery voltage is greater than or equal to 24 vo . b. . At least once per 92 days by verifying that the specific gravit appropriate for continued service of the battery, and c. At least once per 18 months by verifying that: 1) The batteries, cell plates, and battery racks 2) The battery-to-battery and terminal connect . I l O ! l SOUTH TEXAS - UNIT 1 3/4 7-30 ATTACHMENT 10 be_\k . ST.HL AE 191M PAGE G OF 24 . _ . . _ . P_LANT SYSTEMS lO V SPRAY AND/OR SPRINKLER SYSTEMS LIMITING CONDITION FOR OPERATION 3.7.11.2 OPERABLE: The Spray and/or Sprinkler Systems a given in T bl _ APPLICABILITY: e 3.7-3 shall be required to le OPERABLE.Whenever equipment protected by ACTION: prinkler System is a. With one or more of the above required Spray inoperable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months establish a conti and/or Sprinkler Systems backup fire suppression equipment for thosenuous fire w systems hourly or components fire watch patrol. could be damaged; for otheare , b. r areas, establish an The provisions of Specifications 3.0 3 and 3 0 4 . are not applicable. SURVEILLANCE RFOUIRFMENTS ~4.7.11.2 demonstrated OPERABLE:Each of the above required Spray a _ T er Systems shall be a. At least once per 31 da operated, or automatic)ys by verifying that each valve (manu

b. in the flow path is in its correct position At least once per 12 months by sa cycling each te t bl c.

flow path through at least one complete cyclin.the e valve At least once per 18 months: e of full travel, 1) automatic actuation of the ncludes system, simulated and:By a) Verifying that the automatic valves in the flo actuate to their correct positions onw apath and fire signal, alarm b) during of full travel. plant operation through estableat least complete cycle O SOUTH TEXAS - UNIT 1 3/4 7-31 : ATTACHMENT to ST-HL-AE 198 4-N PAGE 14OF 24 De.\ R UL, PLANT SYSTEMS PROOF & REVIEW COPY O M ILLANCE REOUIREMENTS (Continued)

2) By a visual inspection of the dry pipe spray and sprinkler headers to verify their integrity; and

3) By a visual inspection of each nozzle's spray area to verify the spray pattern is not obstructed.

d. At least once per 3 years by performing an air flow test through each open head spray / sprinkler header and verifying each open head spray / sprinkler nozzle is unobstructed.

O O SOUTH TEXAS - UNIT 1 3/4 7-32 . . _. . .- -. = - - _ - . - - - - l PROOF & REVIEW COPY Table 3.7-3 ATTACHMENT ID , ST.HL AE W84 PA M 15 0F2.& __ p/ 'v SPRAY AND/0R SPRINKLER SYSTEMS . FIXED WATER SPRAY DELUGE SYSTEM FOR CARBON FILTER UNITS Area 6 Zone 090 Makeup Unit EAB El. 86'-0" ! Area 6 Zone 091 Makeup Unit EAB E1. 86'-0" Area 6 Zone 092 Makeup Unit EAB El. 86'-0" ! Area 35 Zone 324 Exhaust Subsystem FHB El. 30"-0" Area 35 Zone 325 Exhaust Subsystem FHB El. 30"-0" l Area 35 Zone 326 Exhaust Subsystem FHB E1. 42"-6" Area 35 Zone 327 Exhaust Subsystem FHB El. 42"-6" i Area 35 Zone 328 Exhaust Subsystem FHB El. 53"-3" Area 35 Zone 329 Exhaust Subsystem FHB El. 53"-3" Area 63 Zone 230 Containment Carbon Filter Unit 11A RCB El. 52'-0" Area 63 Zone 231 Containment Carbon Filter Unit 11B RCB El. 52'-0" AUTO WET-PIPE SFRINKLER Area 31 2047 Cable Spreading Room Train B Elevation 60'-0", EAB Area 34 2060 Cable Area Elevation 76'-0", EAB Area 33 2018 Cable Area Elevation 48'-0", EAB Area 2 2010 Power Cable Vault Elevation 10'-0", EAB Area 65 Z057 Cable Spreading, Train C Elevation 74'-0", EAB FOAM-WATER SPRINKLER SYSTEM Area 39 2503 Diesel Generator Fuel Oil Storage Tank # 13 Area 40 2504 Diesel Generator Fuel Oil Storage Tank # 12 Area 41 2505 Diesel Generator Fuel Oil Storage Tank # 11 PREACTION SPRINKLER SYSTEM Area 36 2500 Diesel Generator #13 Engine Room Area 37 Z501 Diesel. Generator #12 Engine Room Area 38 Z502 Diesel Generator #11 Engine Room AUTOMATIC WET-PIPE SPRINKLER SYSTEM Area 59 Z801 Fire Pump House Area 60 2802 Fire Pump House Area 61 2803 Fire Pump House i Area 61 Z804 Fire Pump House Area 22 Z133 Pipe Penetration Area Elevation 10"-0", MAB , Area 24 Z102 Hallway Elevation 10'-0", MAB Area 64 Z026 Cable Spreading Room Train A, Elevation 21'-0", EAB Area 3 Z141 Office Area (Room 2198), Elevation 35'-0", EAB Area 3 Z147 Above Suspended Ceiling, Elevation 35'-0", EAB O 1 ! SOUTH TEXAS - UNIT 1 3/4 7-33 ATTACHMENT C i ST-HL.AE.198+ M eA.L g gi4gEVltW GUPY _R PLANT SYSTEMS HALON SYSTEMS O LIMITING CCNDITION FOR OPERATION 3.7.11.4 The following Halon System shall be OPERABLE: Electric Auxiliary Building Area 1, Zone 032 Relay Room El. 35'-0" APPLICABILITY: Whenever equipment protected by the Halon System is required to be OPERABLE. ACTION:

a. With the above required Halon System inoperable, within I hour establish a continuous fire watch with backup fire suppression equipment.

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

M REQUIREME!!TS 4.7.11.4 The above required Halon System shall be demonstrated OPERABLE:

a. At least once per 31 days by verifying that each valve (manual,

' power-operated, or automatic) in the flow path is in its correct position,

b. At least once per 6 months by verifying Halon storage tank weight to be at least 95% of full charge weight [or level] and pressure to be at least 90% of full charge pressure, and 1

c. At least once per 18 months by:

1) Verifying the system, including associated Ventilation System fire dampers and fire door release mechanisms, actuates manually l and automatically, upon receipt of a simulated actuation signal, and

2) Performance of a flow test through headers and nozzles to assure no blockage.

I O SOUTH TEXAS - UNIT 1 3/4 7-34 ATTMCHMENT 10" - ST HL-AE.1984-hM PAGE 17 OF 24 PLANT SYSTEMS PR0OF & REVIEW COPY FIRE H0SE STATIONS LIMITING CONDITION FOR OPERATION 3.7.11.5 The fire hose stations given in Table 3.7-4 shall be OPERABLE. APPLICABILITY: Whenever equipment in the areas protected by the fire hose stations is required to be OPERABLE. ACTION:

a. With one or more of the fire hose stations given in Table 3.7-4 inoperable, provide gated wye (s) on the nearest OPERABLE hose station (s). One outlet of the wye shall be connected to the standard length of hose provided for the hose station. The second outlet of the wye shall be connected to a length of hose sufficient to provide coverage for the area left unprotected by the inoperable hose station. Where it can be demonstrated that the physical routing of the fire hose would result in a recognizable hazard to operating technicians, plant equipment, or the hose itself, the fire hose shall be stored in a roll at the outlet of the OPERABLE hose station. Signs shall be mounted above the gated wye (s) to identify the proper hose to use. The above ACTION requirement shall be accomplished within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months if the inoperable fire hose is the primary means of fire suppression; otherwise route the additional hose within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ,

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REOUIREMENTS 4.7.11.5 Each of the fire hose stations given in Table 3.7-4 shall be demonstrated OPERABLE:

a. At least once per 31 days, by a visual inspection of the fire hose stations accessible during plant operations to assure all required equipment is at the station.

b. At least once per 18 months, by:

1) Visual inspection of the stations not accessible during plant operations to assure all required equipment is at the station,

2) Removing the hose for inspection and re-racking, and

3) Inspecting all gaskets and replacing any degraded gaskets in the couplings.

c. At least once per 3 years, by:

1) Partially opening each hose station valve to verify valve OPERABILITY and no flow blockage, and

2) Conducting a hose hydrostatic test at a pressure of 250 psig or at least 50 psig above maximum fire main operating pressure, whichever is greater.

SOUTH TEXAS - UNIT 1 3/4 7-35 ATTACHMENT fC) l g ST HL-AE 1984 yR,e_t,t PAGE IB OF 2.4 TABLE 3.7-4 , FIRE HOSE STATIONS LOCATION ELEVATION HOSE RACK NUMBER 1.0 Electrical Auxiliary Building Area 02, Zone 006 Electrical Penetration Area 10'-0" 1EAB-BR-5 Area 02, Zone 010 Power Cable Vault 10'-0" 1EAB-BR-1,9 Area 02, Zone 016 Corridor 10'-0" 1EAB-BR-1,-3,-7,-12 Area 66, Zone 025 BVAC-Electrical Area 23'-0" 1EAB-BR-37 Area 64, Zone 026 Cable Spreading Room 21'-0" 1EAB-BR-2,-10,&-Later Area 02, Zone 028 Corridor 23'-0" 1EAB-HC-27 Area 03, Zone 031 Electrical Penetration Area 35'-0" 1EAB-BR-6 Area 03, Zone 036 Corridor 35'-0" 1EAB-HC-4,-8,-10 Area 03, Zone 043 125-v de Distribution Room 35'-0" 1EAB-HC-2, IEAB-BR-39,-40 Area 04, Zone 046 Electrical Penetration Area 60'-0" 1EAB-BR-7 Area 31, Zone 047 Cable Spreading Room 60'-0" 1EAB-BR-3,-11 Area 04, Zone 050 Corridor 60'-0" 1EAB-HC-5,-6,-9,-14 Area 65, Zone 057 Cable Vault 74'-0" 1EAB-BR-4,-12 Area 67, Zone 058 Technical Sup. Center 72'-0" 1EAB-HC-11.-15 Area 34, Zone 060 HVAC-Electrical Area 76'-0" 1EAB-BR-8 Area 06, Zone 063 HVAC 86'-0" 1EAB-HC-31 Area 06, Zone 085 Corridor 86'-0" 1EAB-BR-39 2.0 Mechanical Auxiliary Building O Area 27, Zone 138 BAT Area Corridor

1MAB-HC-16,-28

'\ Area 24, Zone 102 CCW Area-Corridor 10'-0" 1MAB-BR-15,-30 Area 22, Zone 135 Pipe Penetration Area 10'-0" 1MAB-BR-19 Area 27, Zone 142 CCW' Heat Exchangers 26'-0" 1MAB-HC-17,-19 Area 23, Zone 103 RMWST Room 10'-0" 1MAB-BR-36 Area 23, Zone 134 Non-Rad. Pipe Chase 29'-0" 1MAB-BR-16 Area 00, Zone 115 Radwaste Control Room 41'-0" 1MAB-HC-24 Area 03, Zone 116 Piping Penetration Area 41'-0" 1MAB-B R-21,- 23 Area 03, Zone 147 Health Physics Office-Labs , 41'-0" 1MAB-HC-20,-21,-22,-23 Area 23, Zone 131 Filters-Demineralizers 60'-0" 1MAB-BR-34,-35 Area 32, Zone 122 HVAC Equipment Room 60'-0" 1MAB-13,-14,-18,-33 Area 32, Zone 145 Containment Access 60'-0" 1MAB-BR-24 Area 03, Zone 130 Corridor 41'-0" 1MAB-HR-17,-31,-32 Area 03, Zone 130 Corridor 41'-0" 1MAB-HC-18,-19 3.0 Reactor Containment Building (Proaction) Area 63, Zone 202 Operating Floor 68'-0" 1RCB-HC-23,-24,-25,-26 Area 63, Zone 203 Southwest Section 52'-0" 1RCB-HC-22 Area 63, Zone 204 Northwest Section 52'-0" 1RCB-HC-19 Area 63, Zone 205 Northeast Section 52'-0" 1RCB-HC-20 Area 63, Zone 206 Southeast Section 52'-0" 1RCB-HC-21 l SOUTH TEXAS - UNIT 1 3/4 7-36 1 i . ST.HL.AE- 1984-h e_ PAGE 19 OF 24 PROOF & REVIEW COPY TABLE 3.7-4 (Continued) [ V; FIRE HOSE STATIONS LOCATION ELEVATION HOSE RACK NUMBER 3.0 Reactor Containment Building (Preaction cont.) Area 63, Zone 207 Southwest Section 37'-3" 1RCB-HC-1B Area 63, Zone 208 Northwest Section 37'-3" 1RCB-HC-15 Area 63, Zone 209 Northeast Section 37'-3" 1RCB-HC-16 Area 63, Zone 210 Southeast Section 37'-3" 1RCB-HC-17 Area 63, Zone 211 Southwest Section 19'-0" 1RCB-HC-14 Area 63, Zone 212 Northwest Section 19'-0" 1RCB-HC-10 Area 63, Zone 213 Northeast Section 19'-0" 1RCB-HC-11 Area 63, Zone 214 Southeast Section 19'-0" 1RCB-HC-13 Area 63, Zone 215 Southwest Section (-)11'-0" 1RCB-HC-4 Area 63, Zone 216 Northwest Section (-)2'-0" 1RCB-HC-5/-1 Area 63, Zone 217 Northeast Section (-)11'-0" (-)11'-0" 1RCB-HC-2 Area 63, Zone 218 Southeast Section (-)2'-0" 1RCB-HC-7/-3 Area 63, Zone 221 S.G.'s 1A Cubicle (-)11'-0" 19'-0" 1RCB-HC-9 Area 63, Zone 223 S.G.'s IC Cubicle 19'-0" 1RCB-HC-12 4.0 Fuel Handling Building s Area 35, Zone 303 HVAC Equipment Room (-)21'-0" 1FHB-HC-1,-2,-3,-5 x ' to 53'-3" Area 35, Zone 310 Operating Floor 4'-0" to 1FHB-HC-8,-9 68'-0" 1FHB-HR-3,-5 Area 35, Zone 311 Spent Fuel Pool 21'-11" 1FHB-HC-6 to 42'-6"

5. 0 Isolation Valve Cubicle Area 49, Zone 409 Main Steam Line, Valves Train D 58'-6" lIVC-HR-3 Area 49, Zone 408 Main Steam Line, Valves Train A 58'-6" IIVC-HR-4 Area 49, Zone 407 Main Steam Line, Valves Train B 58'-6" IIVC-HR-5 Area 49, Zone 406 Main Steam Line, Valves Train C 58'-6" lIVC-HR-6 Area 49, Zone 405 HVAC Room 10'-0" lIVC-HR-1,-2 6.0 Diesel Generator Building Area 36, Zone 512 "C" Diesel Air Intake 55'-0" 1DGB-HR-5 Filter Room Area 37, Zone 513 "B" Diesel Air Intake 55'-0" 1DGB-HR-3 Filter Room Area 38, Zone 514 "A" Diesel Air Intake 55'-0" 10GB-HR Filter Room Area 42, Zone 506 "C" Diesel Cubicle Stairway 55'-0" 10GB-HR-6 Area 43, Zone 507 "B" Diesel Cubicle Stairway 55'-0" 1DGB-HR-4 Area 44, Zone 508 "A" Diesel Cubicle Stairway 55'-0" 10GB-HR-2 O

SOUTH TEXAS - UNIT 1 3/4 7-37 ~^ ^ ~ r ~-C, ATTACNMENT f6" - . ST HL AE 1984- l E-g \L. yA .- PAGElo OF 2.4 _ _ l PLANT SYSTEMS- ' F & REVIEW COPY YARD FIRE HYDRANTS AND HYDRANT H0SE HOUSE 5 -O - ), LIMITING CONDITION FOR OPERATION- _ 3.7.11.6 The yard fire hydrants and associated hydrant hose houses given in Table 3.7-5 shall be OPERABLE. APPLICABILITY: Whenever equipment in the areas protected by the>jard fire hydrants is required to be OPERABLE. , m ACTION:

a. With one or more of the yard fire hydrants er associdted hydrant hose houses given in Table 3.7-5 inoperable, within I hour have i sufficient additional lengths'of 2 1/2 inch diameter hose located in an adjacent' OPERABLE hydrant hose house te provide $rrvice to the '

i unprotected area (s) if the inoperable fire hydrant or' associated hydrant hose house is the primary means of fire suppresrion; i otherwise, provide the additional hose within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS - . 4.7.11.6 Each of the yard fire hydrants and asshciated hydEdht hose houses given in Table 3.7-5 shall be demonstrated OPERABLE:

a. At least once per 31 days, by visual inspection of 'the hydre.nt tose 3 house to assure all required equipment is at the hose house, -

l

b. At least once per 6 months (once during March, April, or May and once during September, October, or November), by visually inspecting s each yard fire hydrant and verifying that the hydrant barreisii, dry and that the' hydrant is not damaged, and

c. At least once per 12 months by:

1) Conductingahosebyprostatictestatapressureof150ps} gor ,

at least 50 psig above maximum fire main operating pres'sure, I whichever is greater, i !

2) Inspecting all the gaskets and replacing any degraded gaskete <

in the couplings, and

3) Performing a flow check of each hydrant to verify its OPERABILITY. '

i i c l \ l , i a 3/4 7-38 SOUTH TEXAS - UNIT 1 I ATTACHMENT to '; . ST.HL-AE 1984-CE \ D ., o GL1] QF 24-r a TtVitW L'UPY i s b' , TABLE 3.7-5

YhRD FIRE HYDRANTS AND ASSOCIATED HYDRANT HOSE HOUSES 3

LOCATION

HYDRANT NUMBER

/ y; '~ .?> . , ,.~ ' ,. f 4 ~ c .s f I.. r ) {: ? 1';

l ,'-'-

.p A' ' O ,. , *List all Yard Fire Hydrant and Hydrant Hose Houses to ensure the OPERABILITY of sa"ety-related equipment. l / fr - SOUTH TEXAS - UNIT 1 3/4 7-39 e 1 /'/ fj /a 1 4 ATTACHMENT 10- I ' * ' ST HL-AE 1984-m' , i PAGE 22. OF 24-K7 ~' ~ F- s ? '- PLANT SYSTEMS \& PROOF & REVIEW COPY 3/4.7.12 FIRE RATED ASSEMBLIES -LIMITING CONDITION FOR OPERATION O l ' f,! ' 3.7.12 Allfirehatedassemblies'(valls, floor / ceilings,cabletrayenclosures, and other fire barriers) separating safety-related fire areas or separating ,0, portions of redundant systems important to safe shutdown within a fire area , l and all sealing devices in fire rated assembly penetrations (fire doors, fire .k. \ windows, fire dampers, cable, piping,,and ventilation duct penetration. seals) ! 4 shall be OPERABLE. ' APPLICABILITY: At all times. ACTION:

a. With one or more of"the above rquired fire rated assemblies and/or sealing devices inoperable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months either establish a continuous fire watch on at least one side of the affected assembly, or verify the OPERABILITY of fire detectors on at least one side of the inoperable assembly and establish an hourly fire watch. patrol.

a

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS .O= 4.7.12.1 At least once per 18 months the above required fire rated assemblies ,s and penetration sealing devices shall be verified OPERABLE by' performing a visual inspection of:

a. The exposed surfaces of.each fire rated assembly,

b. Each fire window / fire damper and associated hardware, and

c. At least 10% of each type of sealed penetration. If apparent s changes 'in appearance or abnormal degradations are found, a visn'al inspection of an additional 10% of each type of sealed penetration r shall be made. ,This inspection process shall continue until a 10% ,'

sample with no apparent changes in appearance or abnormal degradation' is found. Samples shall be selected.such that each penetration will' l V. be inspected every 15 years. . s T Q s i O - SOUTH TEXAS - UNIT 1 3/4 7-40 s 3 t _ _ . , _ , - . _ _ _ _ _ - -~ ATTACHMENT to ST.HL-AE 1984 %d O PAE2 3 iy 24_ ,,_ , m V W 6 fTtyltyy bury PLANT SYSTEMS O O SURVEILLANCE REQUIREMENTS (Continued) 4.7.12.2 Each of the above required fire doors shall be verified OPERABLE by inspecting the automatic hold-open, release and closing mechanism and latches at least once per 6 months, and by verifying:

a. The OPERABILITY of the fire door supervision system for each electrically supervised fire door by performing a TRIP ACTUATING DEVICE OPERATIONAL TEST at least once per 31 days,

b. That each locked closed fire door is closed at least once per 7 days,

c. That doors with automatic hold-open and release mechanisms are free of obstructions at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , and a functional test is performed at least once per 18 months, and

d. That each unlocked fire door without electrical supervision is closed at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

O O SOUTH TEXAS - UNIT 1 3/4 7-41 I ATTACHMENT 10 l ST HL AE.1984-i, Pd f L 2_4_OF 2 + rnuur & MtVitW UUPy ; TABLE 3.7-6 AREA TEMPERATURE MONITORING ARIA TEMPERATURE LIMIT ( F)

1. t% lay Room B5

2. wikh eor 9 Rooms ios

3. Electrical Penetrofion Spaces ios 4.

Sa9etyIn echen i Containment Spsy 5. R>me C icles g C.ompened Celing Water Rune Cebicles 12 3 ro, centrifeol charging Rmp Cubiete.s 13' 7, Hydrogen Analyper R=m i b

6. Borie. Anid Tronsfer Ramp Cubicles tcrt c), Stondby Diesel Generator Rooms 12 3 iD. essenHot Cooling WderR;mp Rooms 10 7 ti, isolation Volve Cobicles 807 THIS PAGE OPEN PENDING RECEIPT OF O INFORMATION FROM THE APPLICANT O

SOUTH TEXAS - UNIT 1 3/4 7-43 ATTACHMENT // . ST.HL AE /9#Y PAGE / OF 9 ~ ~ ~ ~ ~ - - - " 3/4.- 8 Electrical Power-Systems A. 3/4 8-1: Note added to clarify STP design since STS do not reflect STP AC Power Distribution design. B. 3/4 8-5: (1) Specific data provided. (2) Item 6 c) has been clarified. C. 3/4 8-6: The South Texas ESF load sequencer is based upon time, not intervals. .Therefore, changed to reflect a 0.6 second time delay. D. 3/4 8-8: (1) Change in Surveillance Requirement numbers provided. (2) Added paragraph.to allow " troubleshooting" of diesel generator problems for the purposes of complying to the Regulatory Guide. Note that this paragraph was provided as part of July 1986 submittal and is consistent with STP position on Reg.-Guide 1.108 as described'in FSAR Table 3.12-1 and STS Rev. 4. E. 3/4 8-10: Justification provided in letter ST-HL-AE-1897. F. 3/4 8-17: Justification provided in letter ST-HL-AE-1923 dated March 13, 1987. G. 3/4'8-18: The current Surveillance Requirement specifies that the instantaneous element for molded case circuit breakers be tested by injecting a current equal to i 20% of the pickup - value and verify that the breaker trips. Experience at Palo' Verde has shown that the narrow tolerance of + 20% ~ causes numerous failures and subsequently requires unnec-essary additional testing. Palo Verde's Technical Speci-fications were changed to allow a tolerance of +40%/-25% for a frame size of 250 amps and less, and a tolerance of i 25% for a frame size of 400 amps and greater. STP utilizes molded case circuit breakers similar to those used at Palo Verde and expects to observe similar prob-lems. These recommendations are consistent with those provided in Part 4B of NEMA Standard AB 2-1980. L3/NRC/cm ATTACHMENT B ST.HL AE MM l PAGE otOF f PROOF & REVIEW COPY 3/4.8 ELECTRICAL POWER SYSTEMS U 3/4.8.1 A.C. SOURCES OPERATING LIMITING CONDITION FOR OPERATION 3.8.1.1 As a minimum, the following A.C. electrical power sources shall be OPERABLE:

a. Twophysicallyindependentcircuitsbetweentheoffgtetransmission network and the onsite Class 1E Distribution System, and

b. Three separate and independent standby diesel generators, each with a separate fuel tank containing a minimum volume of 60,500 gallons of fuel.

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

a. With one offsite circuit of the above-required A.C. electrical power sourcespinoperable, demonstrate the OPERABILITY of the remaining A.C.

<> source @ by performing Surveillance Requirement 4.8.1.1.1.a within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter. Demonstrate the 7 OPERABILITY of each standby diesel generator that has not been suc-(d cessfully tested within the past 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by performing Surveillance Requirement 4.8.1.1.2.a.2) for each such standby diesel generator, separately, within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Restore the offsite cir'cuit to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

b. With a standby diesel generator inoperable, demonstrate the OPERABILITY of the above-required A.C. offsite sources by performing Surveillance Requirement 4.8.1.1.1.a within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and at least once l per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter. If the standby diesel generator became in-operable due to any cause other than preplannned preventive main-tenance or testing, demonstrate the OPERABILITY of the remaining i

OPERABLE standby diesel generators by performing Surveillance Require-ment 4.8.1.1.2.a.2) and for each such standby diesel generator, separately, within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .* Restore the inoperable standby diesel l generator to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT l- SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

c. With one offsite circuit and one standby diesel generator of the above required A.C. electrical power sources inoperable, demonstrate the OPERABILITY of the remaining A.C. sources by performing Specifica-tion 4.8.1.1.la. within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months there-after; and if the standby diesel generator became inoperable due to t

This test is required to be completed regardless of when the inoperable standby diesel generator is restored to OPERABILITY.

SOUTH TEXAS - UNIT 1 3/4 8-1 M Less A one. 3,eicv Sbn bas 4o 4. / 6 KV ESF bas. We consNs \ess a b on Me Lou I Do

WGL semes.tLo IMtV sW&1kssepe % dev.9 nr. 6 \Mesce W M \en

ATTACHMENT a ST-HL-AE 19Fy PAGE J OF 9 PROOF & REVIEW COPY ELECTRICAL POWER SYSTEMS O V SURVEILLANCE REQUIREMENTS (Continued)

b. At least once per 31 days and after each operation of the diesel where the period of operation was greater than or equal to I hour by checking for and removing accumulated water from its associated fuel tank; g

c. By sampling new fuel oil in accordance with AS D4057 prior to addition to storage tanks and:

1) By verifying in accordance with the tests specified in ASTM-D975-81 prior to addition to the storage tanks that the sample has:

a) An API Gravity of within 0.3 degrees at 60 F, or a specific gravity of within 0.0016 at 60/60*F, when compared to the supplier's certificate, or an absolute specific gravity at 60/60*F of greater than or equal to 0.83 but less than or equal to 0.89, or an API gravity of greater than or equal to 27 degrees but less than or equal to 39 degrees; b) A kinematic viscosity at 40'C of greater than or equal to 1.9 centistokes, but less than or equal to 4.1 centistokes m if gravity was not determined by comparison with the . supplier's certification; c) A flash point equal to or greater than 125*F; and d) A clear and bright appearance with proper color when tested in accordance with ASTM-D4176-82.

2) By verifying within 30 days of obtaining the sample that the other properties specified in Table 1 of ASTM-D975-81 are met when tested in accordance with ASTM-D975-81 except that the analysis for sulfur may be performed in accordance with ASTM-D1552-79, ASTM-D2622-82, or ASTM-04294-83.

d. At least once every 31 days by obtaining a sample of fuel oil in accordance with ASTM-D2276-78, and verifying that total particulate contamination is less than 10 mg/ liter when checked in accordance with ASTM-D2276-78, Method A;

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

1) Subjecting the diesel to an inspection in accordance with procedurespreparedinconjunctionwithitsmanufacturer's recommendations for this class of standby service;

2) Verifying the generator capability to reject a load of greater than or equal to 785.3 kW while maintaining voltage at O 4160 1 416 volts and frequency at 60 1 4.5 Hz; SOUTH TEXAS - UNIT 1 3/4 8-4

ATTACHMENT 4 1 ST-HL-AE 199y \ u PAGE 9 OF 9 ) PROOF & REVIEW COPY ! ELECTRICAL POWER SYSTEMS C SURVEILLANCE RE0VIREMENTS (Continued)

3) Verifying the generator capability to reject a load of 5500 kW without tripping. The generator voltage shall not exceed 047R0 volts during and following the load rejection; sa42

4) Simulating a loss-of-offsite power by itself, and:

a) Verifying deenergization of the ESF busses and load shedding from the ESF busses, and b) Verifying the diesel starts on the auto-start signal within 10 seconds, energizes the auto-connected shutdown loads through the load sequencer and operates for greater than or equal to 5 minutes while its generator is loaded with the shutdown loads. After energization, the steady-state voltage and frequency of the ESF busses shall be o maintained at 4160 + 416 volts and 60 + 1.2 Hz during gy this test. ~ - Q- <:c El 52 5) Verifying that on a Safety Injection test signal, without loss-of- @g cc a, offsite power, the diesel generator starts on the auto-start signal and operates on standby for greater than or equal to 5 p e< minutes. The generator voltage and frequency shall be 4160 + 416 volts and 60 + 1.2 Hz within 10 seconds after the auto-start () Z_ cm y signal; the steady-state generator voltage anti frequency shall be $g g maintained within these limits during this test; a- o c 6) Simulating a loss-of-offsite power in conjunction with a Safety 5 "c c1. Z Injection test signai, and: a) Verifying deenergization of the ESF busses and load shedding ]h:c cp from the ESF busses; $.EN b) Verifying the diesel starts on the auto-start signal with- "8 - in 10 seconds, energizes the auto-connected ESF (accident) % Z_ loads through the load sequencer and operates for greater than or equal to 5 minutes while its generator is loaded with the ESF loads. After energization, the steady-state voltage and frequency of the ESF busses shall be maintained at 4160 + 416 volts and 60 + 1.2 Hz during this test; Okkk.oAbbSmap c) Verifying that all automatic diesel generator trips, except engine overspeed, generator differential, and low lube oil pressurelare automatically bypassed upon loss of voltage on the ESF bus concurrent with a Safety Injection Actuation signal. O 7) Verifying the standby diesel generator operates for at least U 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . During the first 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months of this test, the diesel SOUTH TEXAS - UNIT 1 3/4 8-5 ATTACHMENT A ST HL-AE- /975 [ AGE .5 0F 9 PROOF & REVIEW COPY ELECTRICAL POWER SYSTEMS p SURVEILLANCE REQUIREMENTS (Continued) generator shall be loaded to greater than or equal to 5935 kW* and during the remaining 22 hours2.546296e-4 days 0.00611 hours 3.637566e-5 weeks 8.371e-6 months of this test, the diesel generator shall be loaded to greater than or equal to 5500 kW. The generator voltage and frequency shall be 4160 + 416 volts and 60 + 1.2 Hz within 10 seconds after the start signal; the steady-state generator voltage and frequency shall be maintained within these limits during this test. Within 5 minutes after com-pleting this 24-hour test, perform Specification 4.8.1.1.2e.6)b);**

8) Verifying that the auto-connected loads to each standby diesel generator do not exceed the 2000-hour rating of 5935 kW;

9) Verifying the standby diesel generator's capability to:

a) Synchronize with the offsite power source while the generator is loaded with its ESF loads upon a simulated restoration of offsite power, b) Transfer its loads to the offsite power source, and c) Be restored to its standby status.

10) Verifying that with the standby diesel generator operating in a Ox test mode, connected to its bus, a simulated Safety Injection signal overrides the test mode by: (1) returning the diesel gen-erator to standby operation, and (2) automatically energizing the ESF loads with offsite power;

% bo% %c

11) Verifying /thattheautomaticloadsequencetimerisOPERABLE with the/, interval tRFtween each load block within @ of its design W 4 \ o.to u c.

12) Verifying that the standby diesel generator emergency stop lock-l out feature prevents diesel generator starting; and l

If future load conditions exceed the 2000-hour rating (5935 kW) of the

. diesel generator, the diesel generator will be tested at the 2-hour rating (6050 kW) thereafter.

- If Specification 4.8.1.1.2e.6)b) is not satisfactorily completed, it is not necessary to repeat the preceding 24-hour test. Instead, the standby diesel generator may be operated at 5500 kW for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or until operating temperature j has stabilized. '

SOUTH TEXAS - UNIT 1 3/4 8-6 ATTACHMENT // ST-HL AE /974 L N0m ElmL nam, I MWT M HLYf LTT Wr3 Table 4.8-1 (J DIESEL GENERATOR TEST SCHEDULE NUMBER OF FAILURES IN NUMBER OF FAILURES IN LAST 20 VALID TESTS

LAST 100 VALID TESTS

TEST FREQUENCY 14 Once per 31 days

$1 1 2** 25 Once per 7 days Z w tT: %ever, Tedn. To ve% cetedon d ke colde1. aOer%e. a%_id %. is dechst% a te b tersce.". h 6'tue.\ ed ne_ 45*' *M qstems w be. . erded , mt mem ,b peebrm b'"V* ShA 4 a cetteth .Q.spcMe 69ex5, pier to dMit i ud%1 %est opbs c. A d.,(t f"T*S 4"yhs st% % Repdary Gde_. O

Criteria for determining number of failures and number of valid tests shall be in accordance with Regulatory Position C.2.e of Regulatory Guide 1.108, but determined on a per diesel generator basis. @iERT)

For the purpose of determining the required test frequency, the previous test ' failure count may be reduced to zero if a complete diesel overhaul to like-new condition is completed, provided that the overhaul, including appropriate post-maintenance operation and testing, is specifically approved by the manu-facturer and if acceptable reliability has been demonstrated. The reliability criterion shall be the successful completion of 14 consecutive tests in a single series. Ten of these tests shall,be in accordance yith the routine Surveillance Requirements 4.8.1.1.2.a.fand 4.8.1.1.2.a.gind four tests in accordance th the 184-day testing requirement of Surveillance Requirements 4.8.1.1.2.a. and 4. 8.1.1. 2. a.f,3 If this criterion is not satisfied during the first series of tests, any alternate criterion to be used to transvalue the failure count to zero requires NRC approval.

- The associated test frequency shall be maintained until seven consecutive failure free demands have been performed and the number of failures in the last 20 valid demands has been reduced to one.

SOUTH TEXAS - UNIT 1 3/4 8-8 '~ ATTACHMENT n - ST HL AE. /9 Py PAGE 7 OF 9 PR0OF & REVIEW COPY ELECTRICAL POWER SYSTEMS ,ms () 3/4.8.2 D.C. SOURCES - l OPERATING l LIMITING CONDITION FOR OPERATION 3.8.2.1 As a minimum, the following D.C. electrical sources shall be OPERABLE:

a. Channel I 125-volt Battery Bank (E1A11), and its two associated chargers,

b. Channel II 125-volt Battery Bank (E1011), and its associated full capacity charger,

c. Channel III 125-volt Battery Bank (ElB11) and its associated full capacity charger, and

d. Channel IV 125-volt Battery. Bank (E1C11) and its two associated chargers.

APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: 7 With one of the required battery banks inoperable, restore the inoperable (d a. battery bank to OPERABLE status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . ehe ' h dea % cQ % %t 6

b. With less than the required number of chargers on any one train OPERABL demonstrate the OPERABILITY of the associated battery bank by performing 4 Surveillance Requirement 4.8.2.1.a.1) within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and at least once per g 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter. If any Category A limit in Table 4.8-2 is not met, y,g )

declare the battery inoperable. Restore the inoperable charger (s) 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 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDGWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . SURVEILLANCE REQUIREMENTS / 4.8.2.1 Each 125-volt battery bank and charger shall be demonstrated OPERABLE:

a. At least once per 7 days by verifying that:

1) The parameters in Table 4.8-2 meet the ' Category A limits, and

2) The total battery terminal voltage is greater than or equal to 129 volts on float charge.

a SOUTH TEXAS - UNIT 1 3/4 8-10 L

ATTACHMENT //

. ST HL AE.1989 i PAGE F OF9 . _ . . _ PROOF & REVIEW COPY . ELECTRICAL POWER SYSTEMS V 3/4.8.3 ONSITE POWER DISTRIBUTION OPERATING LIMITING CONDITION FOR OPERATION 3.8.3.1 The following electrical busses shall be energized in the specified manner:

a. Train A A.C.. ESF Busses consisting of:

1) 4160-Volt ESF Bus # E1A, and

2) 480-Volt ESF Busses # E1A1 and E1A2 from respective load center transformers,

b. Train B A.C. ESF Busses consisting of:

1) 4160-Volt ESF Bus # ElB, and

2) 480-Volt ESF Busses # E181 and E182 from respective load center

, transformers.

c. Train C A.C. ESF Busses consisting of:

1) 4160-Volt ESF Bus # E1C, and

2) 480-Volt ESF Busses # E1C1 and E102 from respective load center transformers.

d. 120-Volt A.C. Vital Distribution Panels DP1201 and DP001 energized from their associated inverters connected to D.C. Bus si E1A11 ,

e. 120-Volt A.C. Vita istribution Panel DP1202 ener associated inverte connectedtoD.C. Bus #E1D11gizedfromits

f. 120-Volt A.C. VitalfDistribution Panel DP1203 energized from its associated inverter connected to D.C. Bus # E1811*,

.I g. 120-Volta.C.VitalDistributionPanelsDP1204andDP002 energized from their associated inverters connected to D.C. Bus # E1C11 ,

h. 125-Volt D.C. Bus EIA11 energized from Battery Bank E1A11,

i. 125-Volt D.C. Bus E1011 energized from Battery Bank E1011,.

j. 125-Volt D.C. Bus E1811 energized from Battery Bank E1811, and

k. 125-Volt D.C. Bus E1C11 energized from Battery Bank E1C11.

The inverters associated with one channel may be disconnected from its D.C.

bus for. up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months as necessary, for the purpose of performing an equaliz-ing charge on its associated battery bank provided: (1) its vital distribu-tion panels are energized, and (2) the vital distribution panels associated with the other battery banks are energized from their associated inverters' and connected to their associated D.C. busses. SOUTH TEXAS - UNIT 1 3/4 8-14 ATTACHMENT n - ST HL-AE /9W PAGE 4 OF 4 PMUUt & NtVIEW COPY ELECTRICAL POWER SYSTEMS CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT P LIMITING CONDITION FOR OPERATION mum 4 3.8.4.1 . 3 A gisan M i M shall be OPERABLE.A11Acdntainment penetration cond APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: With one device or more (s) gi-- - M- of1:4the containment penetration conductor overcurren inoperable: a. the or circuit (s) by tripping the associated backup rackin - 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ,g out or removing the inoperable circuit breaker within declare the affected system or component inoperable, and verify the backup circuit breaker to be tripped or the inoper-able circuit breaker racked out or removed at least once per 7 day thereafter the provisions of Specification 3.0.4 are not applicable to overcurr;ent devices in circuits which have their back i Q breakers removed, ortripped, their inoperable circuit breakers racked out, or b. Be in at least SHUTDOWN within HOT STANDBY the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in SURVEILLANCE REQUIREMENTS i 4.8.4.1 All containment reped gi-+" ;- iebie i.M ihal enetration conductor overcurrent protective devices be demonstrated OPERABLE:

a. At least once per 18 months:

1) 13.9KV By verifying that the medium voltage [4- H M circuit breakers are OPERABLE by selecting, on a rotating basis, at least 10% of following:the circuit breakers of each voltage level, and performi a)

A CHANNEL CALIBRATION of the associated protective relays, b) An integrated system functional test which includes simulated i automatic actuation of the system and verifying that each - i relay and associated circuit breakers and control circuits 1 function as designed, and O - SOUTH TEXAS - UNIT 1 3/4 8-17 ATTACHMENT /4 . ST HL AL MM Pact / nr 4- _ 3.4.9 Refueling Operations A .~ 3/4'9-6: (1) The indicated changes _are necessary to be able to set the main hoist overload trips in accordance with Westinghouse Instruction F-5.3 and to perform the operability test of the auxiliary hoist. (2) The primary overload cutoff setpoint has been deleted. This setpoint is not in the Standardized Technical Speci-fications. B. 3/4 9-7: Justification provided in letter ST-HL-AE-1862. C. 3/4 9-14: Editorial, for consistency. D. 3/4 9-15: See Tech Spec comments on page 3/4 7-18, 19. E. 3/4 9-16: (1) See previous comment. (2) Specific negative pressure provided based upon-preoperational testing. L3/NRC/cm ATTACHMENT /R . ST.HL-AE /9ff PAGE J Df4 rnwr & REVlEW COPY REFUELING OPERATIONS O C 3/4.9.6 REFUELING MACHINE LIMITING CONDITION FOR OPERATION 3.9.6 The refueling machine and auxiliary hoist shall be used for movement of thimble plugs, drive rods or fuel assemblies and shall be OPERABLE with: .a . Therefuelingmachineusedfgrgovementoffuelassemblieshaving:

1) A minimum capacity of 3000 pounds, and

2) An automatic overload cutof att fwith M.ww the -gb

d. +o 320 aMS nri or ey*1=W W .

P a) Primary - less than or equal to 250 pounds above tee indicated suspended weight for wet conditions and less than or equal to 350 pounds above the indicated suspended weight for dry conditions, and b C)' Secondary - less than primary overload cutoff.orf equal to 150 pounds above the ~ b. The auxiliary hoist used for latching and unlatching drive rods Lug and fw tMmHe plug t==+ Hag, c4 eMg h3gtug: 5

1) A minimum capacity of N pounds, and

2) A 1,000 pound load indicator which shall be used to monitor lifting loads for these operations.

APPLICABILITY: During movement of drive rods or fuel assemblies within the reactor vessel. ACTION: With the requirements for crane and/or hoist OPERABILITY not satisfied, suspend use of any inoperable refueling machine and/or auxiliary hoist from operations involving the movement of drive rods and fuel assemblies within the reactor. ; vessel. l l i SURVEILLANCE REQUIREMENTS 4.9.6.1 Each refueling machine used for movement of fuel assemblies within the reactor vessel shall be demonstrated OPERABLE within 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months prior to the start of such operations by performing a load test of at least 12$% Uf the 3300 s=mdmf m***a==He ombrud cutntf and demonstrating an automatic load cutof f when the refueling machine load exceeds the setpoints of Specification 3.9.6a.2). 4.9.6.2 Each auxiliary hoist and associated load indicator used for movement of drive rods within the reactor vessel shall be demonstrated OPERABLE within 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months prior to the start of such operations by performing a load test of O at least 1250. pounds. %O SOUlH TEXAS - UNIT 1 3/4 9-6 ATTACHMENT /g ST.HL AE /9/y A fqE_ _3 O F G REFUELING OPERATIONS NUUt & REVlEW COPY 3/4.9.7 CRANE TRAVEL - FUEL HANDLING BUILDING LIMITING CONDITION FOR OPERATION -2sco 3.9.7 Loads in excess of W pounds shall be prohibited from travel over fuel assemblies in the spent fuel pool except when carried by the single-failure proof 15-ton hoist of the FHB crane. APPLICABILITY: With fuel assemblies in the spent fuel pool. ACTION:

a. With the requirements of the above specification not satisfied, place the crane load in a safe condition.

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS .2500 4.9.7 Loads shall be verified less than or equal to ti400 pounds prior to movement over fuel assemblies in the spent fuel pool unless they are carried by the single-failure proof 15-ton hoist of the FHB crane. O 4 e O SOUTH TEXAS - UNIT 1 3/4 9-7 ATTACHMENT /4 ST HL-AE /981 PAGE V OFlo mUDF & REVIEW COPY REFUELING OPERATIONS 3/4.9.12 FUEL HANDLING BUILDING EXHAUST AIR SYSTEM - M 1NDITION FOR OPERATION ; 3.9.12 The FHB Exhaust Air System comprised of the following components shall be OPERABLE:

a. Two independent exhaust air filter trains, j

b. Three independent exhaust booster fans, !

c. Three independent main exhaust fans, and i

d. Associated dampers.

APPLICABILITY: Whenever irradiated fuel is in the spent fuel pool. ACTION:

a. With less than the above FHB Exhaust Air System components OPERABLE but with at least one FHB exhaust air filter train, two FHB exhaust booster fans, two FHB main exhaust fans, and associated dampers OPERABLE, fuel movement within the spent fuel pool or crane operation with loads over the spent fuel pool may proceed provided the OPERABLE FHB Exhaust Air System components are capable of being powered from O an OPERABLE emergency power source and are in operation and discharg-ing through at Jeast one train of HEPA filters and charcoal adsorbers,

b. With no FHB exhaust air filter train, or less than two FHB exhaust booster fans, or less than two FHB main exhaust fans and associated dam)ers OPERABLE, suspend all operations involving movement of fuel

,[ gg wit 11n thelttorage pool or crane operation with loads over the spent fuel pool until at least one FHB exhaust air filter train, two FHB . exhaust booster fans, two main exhaust fans, and associated dampers are restored to OPERABLE status. ., c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. SURVEILLANCE REOUIREMENTS 4.9.12 The above required FHB Exhaust Air Systems shall be demonstrated OPERABLE:

a. At least once per 31 days on a STAGGERED TEST BASIS by initiating, from the control room, flow through the HEPA filters and charcoal adsorbers and verifying that the system operates for at least 10 continuous hours with the heaters operating with two of the three exhaust booster fans and two of the three main exhaust fans operating to maintain adequate air flow rate; O

SOUTH TEXAS - UNIT 1 3/4 9-14 , I I ATTACHMENT /3 ' ST.HL AE l '/f'/ PAGE S OF/. REFUELING OPERATIONS ~ PROOF & REVIEW COPY O V SURVEILLANCE REQUIREMENTS (Continued)

b. At least once per 18 months or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire, or chemical release in any ventilation zone communicating with the system by:

1) Verifying that the cleanup system satisfies the in place penetration and bypass leakage testing acceptance criteria of less than 1.0% and uses the test procedure guidance in Regulatory Positions C.5.a. C.5.c, and C.5.d of Regulatory Guide 1.52, Revision 2, March 1978, and the system flow rate is P cfm i 10%;

ai,000

2) Verifying, within 31 days after removal, that a laboratory analysis of a representative carbon sample obtained in accor-dance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978, for a methyl iodide penetration of less than 1.0%;

and 1%0 i M,000

3) Verifying a system flow rate of 3M000 cfm i 10% during system -

operation with two supply fans and two of the three exhaust booster and main haust fans operating when tested in accordance with ANSI N510-O' All combinations of two exhaust booster fans and two main exhaust fans shall be tested.

c. After every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of charcoal adsorber operation by verifying, within 31 days after removal, that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978, for a methyl iodide penetration of less than 1.0%.

.- d. At least once per 18 months by:

1) Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 6 inches Water Gauge while operating the system at a flow rate of Sih000 cfm i 10%,

ai,oco

2) Verifying that on a High Radiation test signal, the system automatically starts (unless already operating) and directs its exhaust flow through the HEPA filters and charcoal adsorber banks, l

O i SOUTH TEXAS - UNIT 1 3/4 9-15 .c - g ATTACHMENT G ST-HL-AE MW r.'= at 9_Q?A ne n?FE ntvitW COPY REFUELING OPERATIONS O SURVEILLANCE REQUIREMENTS (Continued)

3) Verifying that the system maintains the spent fuel storage pool area at a negative pressure of greater than or equal to 02$t} y2 inch Water Gauge relative to the outside atmosphere during system operation, 4 VeP+@ing that tta Fi4wr ceeli -; tyypats fatwo-teo Bs m =i'q g_

'd "' 3/v,7

95) Verifying that the heaters dissipate 50 1 5 kW when tested in accordance with ANSI N510-1325.liRD t 90

e. After each complete or partial replacement of a HEPA filter bank, by verifying that the cleanup system satisfies the in place penetration and bypass leakage testing acceptance criteria of less than 1.0% in accordance with ANSI N510- for a DOP test aerosol while operating

, the system at a flow rate of 3M UGO cfm i 10%. aboco

f. After each complete or partial replacement of a charcoal adsorber bank, by verifying that the cleanup system satisfies the in place penetration and bypass leakage testing acceptance criteria of less than 1.0% in accordance with ANSI N510-y for a halogenated hydrocarbon refrigerant test gas whileloperating the system at a O flow rate of 34000 cfm i 10%.

31,000 lwo " THIS PAGE OPEN PCMDING RECEIPT OF INFORMATION FROM NE APPLICANT O SOUTH TEXAS - UNIT 1 3/4 9-16 ATTACHMENT /3 . ST.HL AE /*#v PAGE ( OFJ 3/4.11 Radioactive Effluents A. 3/4'11-7: STP Specific nomenclature. B. 3/4 11-16: Justification provided in letter ST-HL-AE-1901. L3/NRC/cm ATTACHMENT /3 ST.HL AE /9FV - L2 mm, i PnofA@ num alCu, wr i nu RADI0 ACTIVE EFFLUENTS \ j LIQUID HOLDUP TANKS

LIMITING CONDITION FOR OPERATION 3.11.1.4 The quantity of radioactive material contained in each of the following unprotected outdoor tanks shall be limited to less than or equal to 150 Curies, excluding tritium and dissolved or entrained noble gases:

a. Outside temporary tank APPLICABILITY: At all times.

ACTION:

a. With the quantity of radioactive material in any of the above listed tanks exceeding the above limit, immediately suspend all additions of radioactive material to the tank, within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months reduce the tank contents to within the limit, and describe the events leading to this condition in the next Semiannual Radioactive Effluent Release Report, pursuant to Specification 6.9.1.4.

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS (] LJ 4.11.1.4 The quantity of radioactive material contained in each of the above listed tanks shall be determined to be within the above limit'by analyzing a representative sample of the tank's contents at least once per 7 days when - radioactive materials are being added to the tank.

Tanks included in this specification are those outdoor tanks that are not surrounded by liners, dikes, or walls capable of holding the tank contents and that do not have tank overflows and surrounding area drains connected p to the Liquid Radweste Treatment System.

Wh Nemi Q SOUTH TEXAS - UNIT 1 3/4 11-7 ~ttACHMENT A /3 . ST.HL AE /4fV PAGE 3 OF3 RADIOACTIVE EFFLUENTS PROOF & REVIEW COPY GAS STORAGE TANKS LIMITING CONDITION FOR OPERATION 3.11.2.6 The quantity of radioactivity contained in each gas storage tank shall be limited to less than or equal to U a 10' Curies of noble gases (considered as Xe-133 equivalent). l.o x io r APPLICABILITY: At all times. ACTION:

a. With the quantity of radioactive material in any gas storage tank exceeding the above limit, immediately suspend all additions of radioactive material to the tank, within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months reduce the tank contents to within the limit, and describe the events leading to this condition in the next Semiannual Radioactive Effluent Release Report, pursuant to Specification 6.9.1.4.

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS T 4.11.2.6 The quantity of radioactive material contained in each gas storage s ) tank shall be determined to be within the above limit at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when radioactive materials are being added to the tank., O SOUTH TEXAS - UNIT 1 3/4 11-16 7_ _ ATTACHMENT /4/ . ST.HL AE /9fy PAGEl_Of#E__ _ BASES A. B3/4 0-1: Consistent with STP three-train design. B. B3/4 1-3: (1) Boration Bases changed to reflect Boron Dilution Analysis related Tech Specs. (2) RUST pH is 7.5 and 10.0; see SER Supplement 2, page 6-1 and FSAR Section 6.2.2. (3) There are 259 Rod Withdrawal steps; consistent with Tech Spec Section 3/4.1.3. C. B3/4 2-2, 2-4: Changes made to reflect deletion of RCS flow from Tech Spec 3.2.3 and new RCS flow of 395,000 using 3.5% uncertainty. D. B3/4 2 6: Made consistent with Tech Spec 3.2.5; see comments on page 3/4 2-11. E. B3/4 3-5: (1) Change to Accident Monitoring consistent with comments on pages 3/4 3-60-63. (2) Fire Protection justification for deletion provided in letter ST-HL AE 1867. F. B3/4 3 6: Deletion justification for Loose Parts Monitoring provided in letter ST IIL AE-1923. G. B3/4 4-1: The transient pressure is 110% of design at 2500 psia, which is 2735 psig, 11 . B3/4 4-4: Made consistent with Tech Spec 3/4.4.6.2. I. B3/4 4-7: Justification submitted in letter ST llL AE 1867. J. B3/4 4 9: Westinghouse specific data provided. K. B3/4 4-12: Added to reflect COMS Analysis. L. B3/4 4 14: Deleted, since this is not consistent with COMS Analysis and applicable Tech Specs. M. B3/4 4 15: Justification provided in July 1987 Tech Specs as part of original submittal; still under NRC review. N. B3/4 5 1, 5-2: (1) Made consistent with Tech Specs. (2) Deleted standard reference to MODE 4 SBLOCA. This is a Westinghouse generic issue that will not be re-solved until after STP fuel load. Therefore, we are not able to address this issue at this time. (3) See previous comment on pli for B3/4 1 3. L3/NRC/cm ATTACFHAENT I 4- . ST HL Ag. l164- _ Pant 2 OF 24l... . BASES (Continued) O. B3/4 6-1: The outside design pressure, as indicated in FSAR l 6.2.1.1-2, is 3.5 psig. P. B3/4 6-2: (1) Chapter 15 Accident analyses show that the maximum pressure within containment is 37.5 psig. See also FSAR Section 3.8. (2) Deletion provided in letter ST HL AE 1883. Q. 83/4 6-3: See previous comment on B3/4 1 3. , R. B3/4 7-2: Provided as part of original submittal. Added for clarification of STP design. S. B3/4 7-6: Deletion of Fire Protection justified in letter ST-HL-AE-1867. T. B3/4 7 7: Deleted reference to instrument error. The Temperature Monitors have varied instrument errors. U. B3/4 8-1: Statement added to clarify that ambient conditions for South Texas diesels are the hot pre-lube condition. Palo Verde has had a problem with their OI&E Region requiring a cold start on diesels designed to start in hot pre-lube condition. V. B3/4 11-2: (1) STP nomenclature. (2) No potable water source along gradient. W. B3/4 11 5: Nothing is shared at STP. X. B3/4 12 1: Justification submitted in letter ST IIL AE 1921 dated March 16, 1987. L t L3/NRC/cm i ATTACHMENT 14 . ST.HL-AE \984 PAGE 3. 0F 28 muut & REVIEW COPY 3/4.0 APPLICABILITY v BASES The specifications of this section provide the general requirements applicable to each of the Limiting Conditions for Operation and Surveillance Requirements within Section 3/4. In the event of a disagreement between the requirements stated in these Technical Specifications and those stated in an applicable Federal Regulation or Act, the requirements stated in the applicable Federal Regulation or Act shall take precedence and shall be met. 3.0.1 This specification defines the applicability of each specification in terms of defined OPERATIONAL MODES or other specified conditions and is provided to delineate specifically when each specification is applicable. 3.0.2 This specification defines those conditions necessary to constitute compliance with the terms of an individual Limiting Condition for Operation and associated ACTION requirement. 3.0.3 The specification delineates the measures to be taken for those circumstances not directly provided for in the ACTION statements and whose occurrence would violate the intent of a specification. For example, Specifi-cation 3.7.3 requires three independent component cooling water loops to be OPERABLE and provides explicit ACTION requirements if one component cooling water loop is inoperable. Under the requirements of Specification 3.0.3, if two or more of the required component cooling water loops are inoperable, O within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months measures must be initiated to place the unit in at least HOT Mge STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and in at least HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . As a further example, Specification 3.6.2.1 requires b k Containment Spray Systems to be OPERABLE and provides explicit ACTION require-ments if one Spray System is inoperable. Under the requirements of Specifica-tion 3.0.3, if two or more of the required Containment Spray Systems are inoperable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months measures must be initiated to place the unit in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , in at least HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and in COLD SHUTDOWN within the subsequent 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . It is acceptable to initiate and complete a reduction in OPERATIONAL MODES in a shorter time interval than required in the ACTION statement and to add the unused portion of this allowable out-of-service time to that provided for operation in subsequent lower OPERATION MODE (S). Stated allowable out-of-service times are applicable regardless of the OPERATIONAL MODE (S) in which the inoperability is discovered but the times provided for achieving a mode reduction are not applicable if the inoperability is discovered in a mode lower than the applicable mode. For example if the Containment Spray System was discovered to be inoperable while in STARTUP, the ACTION Statement would allow up to 156 hours0.00181 days 0.0433 hours 2.579365e-4 weeks 5.9358e-5 months to achieve COLD SHUTDOWN. If HOT STANDBY is attained in 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months rather than the allowed 78 hours9.027778e-4 days 0.0217 hours 1.289683e-4 weeks 2.9679e-5 months 140 hours would still be available beforetheplantwouldberequiredtobeInCOLDSHUTDOWN However, if this system was discovered to be inoperable while in HOT STANDBY, the 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months provided to achieve HOT STANDBY would not be additive to the time available to achieve COLD SHUTDOWN so that the total allowable time is reduced from i 156 hours0.00181 days 0.0433 hours 2.579365e-4 weeks 5.9358e-5 months to 150 hours0.00174 days 0.0417 hours 2.480159e-4 weeks 5.7075e-5 months . I 3.0.4 This specification provides that entry into an OPERATIONAL MODE or l other specified applicability condition must be made with: (1) the full SOUTH TEXAS - UNIT 1 B 3/4 0-1 - - . _ _ _ , - - _ . _ , . - . _ _ . _ _ _ _ _ _ _ , _ _ _ . . . . _ _ _ _ _ . , _ . , , _ _ , , . _ _ , _ _ - . _ , , , , , , , , _ , . _ _ , _ . . , _ . , _,_,_,,___.__,_.,.-,_._m ATTACHMENT 14 ST HL AE 1984 PAQE 4_ OF 2.8 El MtVitW UUPT REACTIVITY CONTROL SYSTEMS aAEFE u= 4k B0 RATION SYSTEMS (Continued)W=A e+ % "NA A 6 WMoo = The boron apability required below 200'gus mMcN *F isAere. % 5%w sufficient to provide aA SHUTDOWN MARGIN at-2 kaka after xenon decay and cooldown from 200*F to 140*F. This condition requires either 2900 gallons of 7000 ppe borated water from the boric acid storage system or 122,000 gallons of 2500 ppe borated water from the RWST for M00E 5 and 33,000 gallons of 2500 ppe borated water from the RWST for MODE 6. The contained water volume limits include allowance for water not available because of discharge line location and other physical characteristics.

7. .f The limits on contained water / volume and boron concentration of the RWST also ensure a pH value of betweenAk 6 and 10.0 for the solution recirculated within containment after a LOCA. This pH band minimizes the evolution of iodine and minimizes the effect of chloride and caustic stress corrosion on mechanical systems and components.

The OPERABILITY of one Boron Injection System during REFUELING ensures .that this system is available for reactivity control while in MODE 6. 3/4.1.3 MOVABLE CONTROL ASSEMBLIES The specifications of this section ensure that: (1) acceptable power distri-bution limits are maintained, (2) the minimum SHUTDOWN MARGIN is maintained, and (3) the potential effects of rod misalignment on associated' accident analyses are limited. OPERABILITY of the control rod position indicators is required to detemine control rod positions and thereby ensure compliance with the control i rod alignment and insertion limits. Verification that the Digital Rod Position ! Indicator agrees with the demanded position within i 12 steps at 24, 48, 120, ! 267~~snifA238; n steps withdrawn for the Control Banks and 18, 210, and $$$ptops wiUr -W l drawn for the Shutdown Banks provides assurances that the Digital Rod Position Indicator is operating correctly over the full range of indication. Since the l Digital Rod Position Indication System does not indicate the actual shutdown rod i position between 18 steps and 210 steps, only points in the indicated ranges

are picked for verification of agreement with demanded position.

The ACTION statements which permit limited variations from the basic requirements are accompanied by additional restrictions which ensure that the ! original design criteria are met. Misalignment of a rod requires measurement of peaking factors and a restriction in THERMAL POWER. These restrictions pro- ! vide assurance of fuel rod integrity during continued operation. In addition, those safety analyses affected by a misaligned rod are reevaluated to confirm that the results remain valid during future operation, l The maximum rod drop time restriction is consistent with the assumed rod drop time used in the safety analyses. Measurement with T,yg greater than or ' 'a er O a i

sa1 r d ith ii ct r a r *4 a measured drop times will be representative of insertion times experienced during a Reactor trip at operating conditions.

'" t '"- SOUTH TEXAS - UNIT 1 8 3/4 1-3 ATTACHMENT 14 . ST HL AE 1984 65d1B PRMF AGFVim COpv POWER DISTRIBUTION LIMITS O b BASES AXIAL FLUX DIFFERENCE (Continued) Although it is intended that the plant will be operated with the AFD within the target band required by Specification 3.2.1 about the target flux dif ference, during rapid plant THERMAL POWER reductions, control rod motion will cause the AFD to deviate outside of the target band at reduced THERMAL POWER levels. This deviation will not affect the xenon redistribution suffi-ciently to change the envelope of peaking factors which may be reached on a subsequent return to RATED THERMAL POWER (with the AFD within the target band) provided the time duration of the deviation is limited. Accordingly, a 1-hour penalty deviation limit cumulative during the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is provided for operation outside of the target band but within the limits of Figure 3.2-1 while at THERMAL POWER levels between 50% and 90% of RATED THERMAL POWER. For THERMAL POWER levels between 15% and 50% of RATED THERMAL POWER, deviations of the AFD outside of the target band are less significant. The penalty of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months actual time reflects this reduced significance. Provisions for monitoring the AFD on an automatic basis are derived from the plant process computer through the AFD Monitor Alarm. The computer deter-mines the 1-minute average of each of the OPERABLE excore detector outputs and provides an alarm message immediately if the AFD for two or more OPERABLE excore channels are outside the target band and the THERMAL POWER is greater During operation at THERMAL POWER levels O' than 90% of RATED THERMAL POWER. between 50% and 90% and between 15% and 50% RATED THERMAL POWER, the computer outputs an alarm message when the penalty deviation accumulates beyond the limits of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , respectively. Figure B 3/4 2-1 shows a typical monthly target band. 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR, e m AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR

The limits on heat flux hot channel factor, PIT **az zmTF, and nuclear enthalpy rise hot channel factor ensure that: (1) the design limits on peak local power density and minimum DNBR are not exceeded and (2) in the event of l 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 ensure that the limits are maintained provided:

a. Control rods in a single group move together with no individual rod insertion differing by more than i 12 steps, indicated, from the group demand position;

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

SOUTH TEXAS - UNIT 1 B 3/4 2-2 L_ _ _ _ _ T y ._ ATTACHMENT 14 g NIfl COW Is POWER DISTRIBUTION LIMITS e BASES _ ;- HEAT FLUX HOT CHANNEL FACTOR, and RC", FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR (Continued) ,

c. The control rod insertion limits of Specifications 3.1.3.5 and 3.1.3.6 are taaintained; and ,y

d. The axial power distribution, expressed in terms of AXIAL FLUX '

DIFFERENCE, is at.intained within the limits. , FhwillbemaintainedwithinitslimitsprovidedConditico a. ti/ rough'

d. above are maintained. The combination of the RCS flow requirenat N

(ML CQQ3w gpm) ,v..and the requirement on F3g guarantees that the DN8R used in the N safety analysis will be met. The relaxation of I3.H as a function of THERMAL POWER allows changes in the radial pou r cape fee all permissible rod inser-tion limits. The h ,ew!. _ N of 1-32 inch = a 5 cEwwi 4 h design and e4 - tWa4*ty 1m the measured vatue vf N g . TfiniWee, the.msasues4 vat:a *f h should to ince****d tty 45 befoce be.ing cW with the ;"4rM vake e5 ML. Qs ~ The fim re4uirement (S89,600 gm) already includes a me4surement uncer-ainty of 3.51 Therefore, no adjusteent of tha measured flow value is reces- - ' - C -" rybeforecomparingagainsttheflowrequirement[ Fuel rod bowing reduces the value of DNB ratio. Credit is available to offset this reduction in the generic margin. The generic margini, totaling 3.3% DNBR completely offset any rod bow penalties. This margin includes the following:

a. Design limit DNBR of 1.30 vs 1.28,

b. Grid Spacing (Ks ) of 0.059 vs 0.066, and ,

c. Thermal Diffusion Coefficient (for use in modified spacer factor) of 0.059 vs 0.061.

The applicable values of rod bow penalties are explained in FSAR Sec-tion 4.4.2.2.5. 1 .r.nssar: w ke n Fl4 Is rwee d, n o oddiin I a llow a n ce.s ore -c HitQqbW paa c. .na., meraf 4 era r- of 'l 90 a w fe r'

m. n f A/l0LdCCI b in k e cl e.l e m vsr>. 0 4 I k $ % . h c/n;;,,

has feen >>NB(( value. SOUTH TEXAS - UNIT 1 B 3/4 2-4 l ATTACHMENT I4-ST HL AE 1984 PAGE 1 OF 28 OWER DIS--7'4IBUTION LIMITS" V 4 t BASES ( . QUADRANT POWER TILT RATIO (Continued) The 2-hour time allowance for operation with a tilt condition greater than 1.02 is provided to dllow identification and correction of a dropped or misaligned control rod. In the event such action action does not correct the tilt, the margin for uncertainty on Fn is reinstated by reducing the maximum allowed power by 3% for each percent 5f tilt in excess of 1. For purposes of monitoring QUADRANT POWER TILT RATIO when one excore detec!.or is inoperable, the moveable incore detectors are used to confirm that .t'he normalized symmetric power distribution is consistent with the QUADRANT POWER TILT RATIO. The incore detector monitoring is done with a full incore flux map or two sets of four symmetric thimbles. The two sets of four symmetric , thimbles is a unique set,of eight detector locations. These locations are C-8, E-5, E-11, H-3, H-13, L-5, L-11, N-8. 3/4.2..* _, 0NB 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 tnd have been analytically demonstrated adequate to maintain a minimum DNBR of 1.30 throughout each analyzed transient. The (] V inge,ctedT,yg value o { g F and the indicated pressurizer pressure value of [?2243 psi M: Yb ta analytical 1%its- 6# S9S*F and 2266-psig reg en _ & s bkx tive' wi gWwanse at ea;r--^nt uncatdatys as i

q kmA "

D.0 5b?-hourperto Y,MM,"l '"

d (Jtgom gn) csuYvN'"ianceoTQ. ' . " "J""

ese parameters through i strument readout is sufficient to ensure that the parameters are restored within their lirits ic11oving load changes and other expected transient operation. t THIS PAGE OPEN PENDIN3 RECEIPT OF INFORMATION FROM THE APPLICANT O SOUTH TEXAS - UNIT 1 B 3/4 2-6 l . [ ATTACHMENT 14 i . , . ST.HL AE.1984-PAGLEDF,2fg g I UN W HE T f byy gy, y INSTRUMENTM10N O : i RASES l REMOTE SHUTDOWN SYSTEM (Continued) The OPERABILITY of the Remote Shutdown System ensures that a fire will not preclude achieving safe shutdown. The Remote Shutdown System instrumentation, s control, and power circuits and transfer switches necessary to eliminate effects of the fire and allow operation of irstrumentation, control and power circuits required to achieve and maintain a safe shutdown condition are independent of areas where a fire could damage systems normally used to shut down the reactor. This capability is consistent with General Design Criterion 3 and Appendix R " to 10 CFR Part 50. i + ; ~ 4 - 3/4.3.3.6 ACCIDENT MONITORING INSTRUMENTATION The OPERABILITY of the accident monitoring instrumentation ensures that sufficient information is available on selected plant parameters to monitor and assess these variables following an accident. This capability 11s consis-tent with the recommendations of Regulatory Guide 1.97, Revision)l, "Instrumen- ' tation for Light-Water-Cooled Nuclear Power Plants to Assess Plant Conditions . During and Following an Accident," ",a 10" and NUREG-0737, " Clarification of TMI Action Plan Requirements," November 1980. 7/r. iarA .dk /Med in M/ ;

3.3-so con.<e..Js 4 A c.fepy .1 s n6. ,b/% 4 gret, uleef t , test ,p d A i .

3/4.3.3.7 CHEMICAL DETECTION SYSTEMS a J N etw cel b ia m. des *! i* R*N

c ;J. t.9Qe. 2. -

The OPERABILITY of the Chemical Detection Systems ensures that sufficient capability is available to promptly detect and initiate protective action in - the event of an accidental chlorine release. This capability'is required to , protect control room personnel and is consistent with the recommendations of Regulatory Guide 1.95, Revision 1, " Protection of Nuclear Power Plant Control Room Operators Against an Accidental Chlorine Release," January 1977. FIRE DETECTION INSTRUMENTATION {3/4.3.3.8 , (! The OPERABILITY of the fire detection instrumentation ensures that both adequate warning capability is available for prompt detection of fires and that . Fire Suppression Systems, that are actuated by fire detectors, will discharge , extinguishing agents in a timely manner. Prompt detection and suppression of [ fires wf)1 reduce the potential for damage to safety-related equipment and is an integral element in the overall facility Fire Protection Program. 7 Fire detectors that are used to actuate Fire Suppression Systems represent a more critically important component of a plant's Fire Protection Program than detectors that are installed solely for early fire warning and notifica-tion. Consequently, the minimum number of OPERABLE fire detectors must be greater. The loss of detection capability for Fire Suppression Systems, actuated by fire detectors, represents a significant degradation of fire protection for pny area. As a result, the establishment of a fire watch patrol must be {nitiated at an earlier stage than would be warranted for the loss of detector SOUTH TEXAS . UNIT 1 B 3/4 3-5 l -s ATTACHMENT (+ - ST HL-AE. I'/64 PAGE 9 OF 26 FMUUt & NtVitW UUPY i INSTRUMENTATION-BASES RE DETECTION INSTRUMENTATION (Continued) I that provide only early fire warning. The establishment of frequent fire , patrols in the affected areas is required to provide detection capability ( until the inoperable instrumentation is restored to OPERABILITY. 3/4.3.3.9 LOOSE-PART DETECTION SYSTEM l The OPERABILITY of the Loose-Part Detection System ensures that J sufficient capability is available to detect loose metallic parts in the Reactor System and avoid or mitigate damage to Reactor System components. The allowable out-of-service times and surveillance requirements are consistent 1 with the recommendations of Regulatory Guide 1.133, " Loose-Part Detection (Program for the Primary System of Light-Water-Cooled Reactors," May 1981. 3/4.3.3.10 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION The radioactive liquid effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in liquid effluents during actual or potential releases of liquid effluents. The Alarm / Trip Setpoints for these instruments shall be calculated and adjusted in

accordance with the methodology and parameters in the 00CM to ensure that the l alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20. The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50.

3/4.3.3.11 RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION l The radioactive gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous effluents c during actual or potential releases of gaseous effluents. The Alarm / Trip Setpoints F for these instruments shall be calculated and adjusted in accordance with the methodology and parameters in the ODCM to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20. This instrumentation also includes provisions for monitoring (and controlling) the concentrations of potentially explosive gas mixtures in the GASEOUS WASTE PROCESSING SYSTEM. The OPERABILITY and use of this instausentation is consistent with the requirements , of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50. The i sensitivity of any noble gas activity monitors used to show compliance with the , gaseous effluent release requirements of Specification 3.11.2.2 shall be such that concentrations as low as 1 x 10.s pCi/cc are measurable. 3/4.3.4 TURBINE OVERSPEED PROTECTION l i Thi.s specification is provided to ensure that the turbine overspeed ! protection instrumentation and the turbine speed control valves are OPERABLE and will protect the turbine from excessive overspeed. Protection from turbine l excessive overspeed is required since excessive overspeed of the turbine could j generate potentially damaging missiles which could impact and damage safety-related components, equipment, or structures. SOUTH TEXAS - UNIT 1 B 3/4 3-6 l

1. '

\ . _ . . ATTACHMENT I A - ST HL AE 1984 _ PAG [lo cp 23 NUUP 5: R[ydfQ ! 3/4.4' REACTOR COOLANT SYSTEM BASES 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION The plant is designed to operate with all reactor coolant loops in operation and maintain DNBR above 1.30 during all normal operations and anticipated transients. In MODES 1 and 2 with one reactor coolant loop not in operation this specification requires that the plant be in at least HOT ' STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . In MODE 3, two reactor coolant loops provide sufficient heat removal capability for removing core decay heat even in the event of a bank withdrawal accident; however, a single reactor coolant loop provides sufficient heat removal capacity if a bank withdrawal accident can be prevented, i.e., by opening the Reactor Trip System breakers. Single failure considerations require that two loops be OPERABLE at all times. In MODE 4, and in MODE 5 with reactor coolant loops filled, a single l-reactor coolant loop or RHR loop provides sufficient heat removal capability for removing decay heat; but single failure considerations require that at least two loops (either RHR or RCS) be OPERABLE. In MODE 5 with reactor coolant loops not filled, a single RHR loop provides sufficient heat removal capability for removing decay heat; but single failure , l considerations, and the unavailability of the steam generators as a heat l removing component, require that at least two RHR loops be OPERABLE. The boron dilution analysis assumed a common RCS volume, and maximum di- ; lution flow rate for MODES 3 and 4, and a different volume and flow rate for O MODE 5. The MODE 5 conditions assumed limited mixing in the RCS and cooling with the RHR system only. In MODES 3- and 4, it was assumed that at least one reactor coolant pump was operating. If at least one reactor coolant pump is t not operating in MODE 3 or 4, then the maximum possible dilution flow rate must ' be limited to the value assumed for MODE 5. l The operation of one reactor coolant pump (RCP) or one RHR pump provides adequate flow to ensure mixing, prevent stratification and produce gradual reactivity changes during boron concentration reductions in the Reactor Coolant System. The reactivity change rate associated with boron reduction will, therefore, be within the capability of operator recognition and control. The restrictions on starting an RCP with one or more RCS cold legs less than or equal to 350*F are provided to prevent RCS pressure transients, caused by energy additions from the Secondary Coolant System, which could exceed the limits of Appendix G to 10 CFR Part 50. The RCS will be protected against overpressure transients and will not exceed the limits of Appendix G by restricting starting of the RCPs to when the secondary water temperature of each steam generator is less than 50'F above each of the RCS cold leg temperatures. 3/4.4.2 SAFETY VALVES 2'13 5 ThepressurizerCodesafetyvalves)operatetopreventtheRCSfrombeing Each safety valve is designed pressur.ized to relieve 504,950 above its Safety Limit of(2233.psig.lbs per hour of saturated 2575 psi p.,The relief capacity of a single safety valve is adequate In the event to relieve any overpressure condition which could occur during shutdown.tha O SOUTH TEXAS - UNIT 1 B 3/4 4-1 ATTACHMENT I4 - ST.HL AE. I9eA-PAGE 11 OF 23 F & REVIEW CDPT REACTOR COOLANT SYSTEM O BASES OPERATIONAL LEAKAGE (Continued) the presence of any PRESSURE BOUNDARY LEAKAGE requires the unit to be promptly placed in COLD SHUTDOWN. Industry experience has shown that while a limited amount of leakage is expected from the RCS, the unidentified portion of this leakage can be reduced to a threshold value of.less than 1 gpm. This threshold value is sufficiently low to ensure early detection of additional leakage. The total steam generator tube leakage limit of 1 gpm for all steam generators not isolated from the RCS ensures that the dosage contribution from the tube leakage will be limited to a small fraction of 10 CFR Part 100 dose guideline values in the event o* either a steam generator tube rupture or steam line break. The 1 gpm limit is consistent with the assumptions used in the analysis of these accidents. The 500 gpd leakage limit per steam generator ensures that steam generator tube integrity is maintained in the event of a main steam line rupture or under LOCA conditions. The 10 gpm IDENTIFIED LEAKAGE limitation provides allowance for a limited amount of leakage from known sources whose presence will not interfere with the Leakage Detection Systems. the detectiondof UNIDENTIFIED 0 SSP m k@t. parLEAKAGE Wm by\ iwch oS- v4ve si+e. To %uimwn oh Qv O The specified allowed leakagelfrom any RCS pressure isolation valve is sufficiently low to ensure early detection of possible in-series check valve failure. It is apparent that when pressure isolation is provided by two in-series check valves and when failure of one valve in the pair can go undetected for a substantial length of time, verification of valve integrity is required. Since these valves are important in preventing overpressurization and rupture of the ECCS low pressure piping which could result in a LOCA that bypasses containment, these valves should be tested periodically to ensure low probability of gross failure. The Surveillance Requirements for RCS pressure isolation valves provide added assurance of valve integrity thereby reducing the probability of gross valve failure and consequent intersystem LOCA. Leakage from the RCS pressure isolation valve is IDENTIFIED LEAKAGE and will be considered as a portion of the allowed limit. 3/4.4.7 CHEMISTRY

The limitations on Reactor Coolant System chemistry ensure that corrosion of the Reactor Coolant System is minimized and reduces the potential for Reactor Coolant System leakage or failure due to stress corrosion. Maintaining l

O SOUTH TEXAS - UNIT 1 B 3/4 4-4 A. ATTACHMENT I4- - ST HL AE 1984 - PAGE 12. OF 26 PROOF & REVIEW COPY REACTOR COOLANT SYSTEM BASES PRESSURE TEMPERATURE LIMITS (Continued)

b. Figures 3.4-2 and 3.4-3 define limits to assure prevention of non-ductile failure only. For normal operation, other inherent plant characteristics, e.g., pump heat addition and pressurizer heater capacity, may limit the heatup and cooldown rates that can be achieved over certain pressure-temperature ranges.

2. These limit lines shall be calculated periodically using methods provided below,

3. The secondary side of the steam generator must not be pressurized above 200 psig if the temperature of the steam generator is below 70*F,

4. The pressurizer heatup and cooldown rates shall not exceed 100 F/h and 200*F/h, resp.ectively. The spray shall not be used if the temperature difference between the pressurizer and the spray fluid is greater than

/s.21sh/>]4,and i 5. System preservice hydrotests and inservice leak and hydrotests shall be performed at pressures in accordance with the requirements of ASME Boiler and Pressure Vessel Code,Section XI. The fracture toughness properties of the ferritic materials in the reactor vessel are determined in accordance with the NRC Standard Review Plan, A3TM E185-73, and in accordance with additional reactor vessel requirements. These properties are then evaluated in accordance with Appendix G of the 1976 Summer Addenda to Section III of the ASME Boiler and Pressure Vessel Code and the calculation methods described in WCAP-7924-A, " Basis for Heatup and Cooldown Limit Curves," April 1975. Heatup and cooldown limit curves are calculated using the most limiting value of the nil-ductility reference temperature, RTNDT, at the end of l 32 effective full power years (EFPY) of service life. The 32 EFPY service l life period is chosen such that the limiting RT at the 1/4T location in the core region is greater than the RT NDT ofth$DIimitingunitradiatedmaterial. The selection of such a limiting RTNDT assures that all components in the

Reactor Coolant System will be operated conservatively in accordance with applicable Code requirements.

The reactor vessel materials have been tested to determine their initial RTNDT; the results of these tests are shown in Table B 3/4.4-1. Reactor opera-j tion and resultant fast neutron (E greater than 1 MeV) irradiation can cause an increase in the RT NDT. Therefore, an adjusted reference temperature, based upon the fluence, copper content, and phosphorus content of the material in ! question, can be predicted using Figure B 3/4.4-1 and the largest value of ART computed by either Regulatory Guide 1.99, Revision 1, " Effects of NDT SOUTH TEXAS - UNIT 1 B 3/4 4-7 l o O O TABLE B 3/4.4-1 REACTOR VESSEL TOUGHNESS , N j h _Te s t.T h e e b 4 b u RlfalsW A a n BCl-l all o.012 -3o <, co IO Average Upper 2 , - Self Energy j ,_ Normal to

3

-1 50 ft-lb Principal- ~ Principal i T 35 mil Working Working s cobE B Cu RT Component P NOT Temp. NOT Direction Direction Heat: Ikr. Grade / (%) (%) (*F) (*F) (*F) (ft-lb) (ft-lb) i Closure head done R1616-1 A53 CL 1 0.07 0.018 -30 80 20 116 - l Closure head torus R1615-1 A53 CL 1 0.04 0.010 -30 <30 -30 152 - 1 Closure head torus R1615-2 A53 CL 1 0.11 0.012 -30 <30 -30 1% - , mg C19 sure head torus R1615-3 A53 CL 1 0.07 0.011 -40 <20 -40 132 - ! Closure head flange R1602-1 A508 CL 2 0.05 0.007 0- <60 0 109 - cm Vessel flange R1601-1 A508 CL 2 0.02 0.017 -10 <50 -10 160.5_ - ! u> Inlet nozzle R1613-1 A508 CL 2 - 0.009 -10 <50 -10 140 - j A Inlet nozzle R1613-2 A508 CL 2 - 0.013 0 <60 0 130.5 - , a Inlet nozzle R1613-3 A508 CL 2 0.09 0.009 -20 <40 -20 175 - E Inlet nozzle R1613-4 A508 CL 2 - 0.006 20 <80 20 128 - I Outlet nozzle R1614-1 A508 CL 2 - 0.006 10 <70 10 106 - Outlet nozzle R1614-2 A508 CL 2 - 0.006 0 <60 0 114 - j Outlet nozzle R1614-3 A508 CL 2 - 0.009 -30 <30 -30 129 - Outlet nozzle R1614-4 A508,CL 2 - 0.006 10 <70 10 118 - ' Nozzle shell R1607-1 A5338 CL 1 0.08 0.012 0 110 50 89 - Nozzle shell R1607-2 A5338 CL 1 0.08 0.012 -20 110 50 85 - Nozzle shel1 R1607-3 A533B CL 1 0.07 0.010 -50 90 30 82 -

Inter. shell R1606-1 A533B CL 1 0.04 0.009 -40 70 109.5 p2

Inter. shell R1606-2 A533B CL 1 0.04 0.008 -20 60 10 0- 94 130 119 of_

m,g l Inter. shell Lower shell R1606-3 R1622-1 A533B CL 1 A5339 CL 1 0.05 0.05 0.007 0.006 -20 -30 70 30 10 -30 105.5 111 132 143 Qz o*M 't i Lower shell R1622-2 A5333 CL 1 0.07 0.006 -30 30 -30 122- 149 R* l Lower shell R1622-3 A533 3 CL 1 A533B,CL 1 0.05 0.007 -30 30 -30 127 148 ]cc.p$:27 80ttom head torus R1617-1 0.14 0.012 -50 <10 -50 143 - pQ

Bottom head torus R1618-1 A533 CL 1 0.08 0.015 -50 <10 -50 128 l Inter..and lower o).qo se 0.03 0.004 -50 <10 -50 *158 3

shell vert. welds \ g

Inter. and lower em 5%I\ 0.03 0.007 -70 <10 -70. *100 -

] shell girth weld 4

Normal to principal welding direction b i

PROOF & REVIEW COPY REACTOR COOLANT SYSTEM ATTACHMENT (4 .ST HL AE 1984 fG - PACE 14M 28 V; BASES PRESSURE / TEMPERATURE LIMITS (Continued) fracture toughness curve, defined in Appendix G to the ASME Code. The K IR curve is given by the equation: KIR = 26.78 + 1.223 exp [0.0145(T-RTNDT + 160)] (1) Where: K IR is the reference stress intensity factor as a function of the metal temperature T and the metal nil-ductility reference temperature RTNDT. Thus, the governing equation for the heatup-cooldown analysis is defined in Appendix G of the ASME Code as follows: CKyg + kit <KIR (2) Where: KIM = the stress intensity factor caused by membrane (pressure) stress, , K = the stress intensity factor caused by the thermal gradients, It KIR = constant provided by the Code as a function of temperature relative to the RT NDT f the material, ys C = 2.0 for level A and B service limits, and \ C = 1.5 for inservice hydrostatic and leak test operations. , At any time during the heatup or cooldown transient, KIR is determined by the metal temperature at the tip of the postulated flaw, the appropriate value for RTNDT, and the reference fracture toughness curve. The thermal stresses l resulting from temperature gradients through the vessel wall are calculated P and then the corresponding thermal stress intensity factor, KIT, for the reference flaw is computed. From Equation (2) the pressure stress intensity factors are obtainedmand, lP %

Jauedfrom Puvthese s owd-the b- allowable

% cows press k res d are kalculated. d ksed on &rd C00LDOWN 4 ***"# * ^ # " & v i Wud d c,' inMow wn Ww apeimm puW ar. reydred by ioewrtSo, opp. 9, n uerdaxe. t % h s.

L da_ k T e le_ v. y-s, For the calculation of the allowable pressure versus coolant temperature during cooldown, the Code reference flaw is assumed to exist at the inside of the vessel wall. During cooldown, the controlling location of the flaw is always at the inside of the wall because the thermal gradients produce tensile stresses at the inside, which increase with increasing cooldown rates. Allowable pressure-temperature relations are generated for both steady-state and finite cooldown rate situations. From these relations, composite limit curves are constructed for each cooldown rate of interest.

The use of the composite curve in the cooldown analysis is necessary because control of the cooldown procedure is based on measurement of reactor O coolant temperature, whereas the limiting pressure is actually dependent on the material temperature at the tip of the assumed flaw. During cooldown, the SOUTH TEXAS - UNIT 1 B 3/4 4-12 l -. . _- . _ _ = . ATTACHMENT (4 . ST.HL AE.1984 PAGE XDL1? _ rnour a NEvitW COFY REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (Continued) increase with increasing heatup rate, a lower bound curve cannot be defined. Rather, each heatup rate of interest must be analyzed on an individual basis. Following the generation of pressure-temperature curves for both the steady-state and finite heatup rate situations, the final limit curves are produced as follows. A composite curve is constructed based on a point-by-point comparison of the ste.dy-state and finite heatup rate data. At any given temperature, the allowable pressure is taken to be the lesser of the three values taken from the curves under consideration. The use of the composite curve is necessary to set conservative heatup limitations because.it is possible for conditions to exist such that over the course of the heatup ramp the controlling condition switches from the inside to the outside and the pressure limit must at all times be based on analysis of the most critical criterion. i Finally, the composite curves for the heatup rate data and the cooldown rate data are adjusted for possible errors in the pressure and temperature sensing instruments by the values indicated on the respective curves. J Although the pressurizer operates in temperature ranges above those for which there is reason for concern of nonductile failure, operating limits are provided to assure compatibility of operation with the fatigue analysis perfomed in accordance with the ASME Code requirements. LOW TEMPERATURE OVERPRESSURE PROTECTION The OPERABILITY of two PORVs or an RCS vent opening of at least 2.0 square inches ensures that the RCS will be protected from pressure transients which could exceed the limits of Appendix G to 10 CFR Part 50 when one or more of the RCS cold legs are less than or equal to 350'F. Either PORV has adequate relieving capability to protect the RCS from overpressurization when i the transient is limited to either: (1) the start of an idle RCP with the ( secondary water temperature of the steam generator less than or equal to 50*F l above the RCS cold leg temperatures, or (2) the maximum credible mass injection i flow rate due to the startup of a single HHSI pump plus 100 gpm net charging flow, while the RCS is in a water solid condition and the RCS temperature is i between 350*F and 200*F. For RCS temperatures less than 200*F, the maximum overpressure event con-sists of operating a centrifugal charging pump with complete termination of 1etdown and a failure of the charging flow control valve to the full flow l' condition. l' g The Maximum Allowed PORV Setpoint for the Low Temperature Ov the LTOPS assuming various mass input and heat input transients. Operation with a PORV Setpoint less than or equal to the maximum Setpoint ensures that Appendix G criteria will not be violated with consideration for a maximum pressure overshoot beyond the PORV Setpoint which can occur as a result of time delays in signalTo i l processing and valve opening, instrument uncertainties, and single failure. ! ensure that mass and heat input transients more severe than those assumed cannot occur, Technical Specifications require lockout of all but one safety injection pump and all but one centrifugal charging pump while in MODES 4, 5, and 6 with L__--- SOUTH TEXAS - UNIT 1 B 3/4 4-14 ATTACHMENT I4 ST-HL AE 1984 PAGEko 0F2S PROOF & REVIEW COPY REACTOR COOLANT SYSTEM BASES LOW TEMPERATURE OVERPRESSURE PROTECTION (Continued) _ [the reactor erature vessel is more headabove than 50'F installed primaryand disallow start of an RCP if' secondary te temperature. The Maximum Allowed PORV Setpoint for the LTOPS will be updated based on the results of examinations of reactor vessel material irradiation surveillance specimens performed as required by 10 CFR Part 50, Appendix H, and in accordance - (with the schedule in Table 4.4-5. f 3/4.4.10 STRUCTURAL INTEGRITY The inservice inspection and testing programs for ASME Code Class 1, 2, and 3 components ensure that the structural integrity and operational readiness of these components will be maintained at an acceptable level throughout the life of the plant. These programs are in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50.55a(g) except where specific written relief has been granted by the Commission pursuant to 10 CFR 50.55a(g)(6)(i). l Components of the Reactor Coolant System were designed to provide access l to permit inservice inspections in accordance with Section XI of the ASME Boiler and Pressure Vessel Code,1974 Edition and Addenda through Winter 1975. t a O_ 3/4.4.11 REACTOR COOLANT SYSTEM VENTS ' veocl. He vents are provided to exhaust noncondensible gases i Reactor te m nt and/or circulation steam from the Reactor Coolant System that cot.1d inhibit natura core cooling.- vent path >fraL the [rre-:& =arsel t:d], tha E: :_ier C::i::t Ey:;t "i@ - -psQ the Egesar' =r -terr sp-reL ud th-- [i:;;;htiza undspsse higtt ps*Nt3 ensures that the capability exists to perform this function. eme.\ l The valve redundancy of the Reactor Seo%ent 4, Q.:: vent paths serves ' minimize the probability of inadvertent or irreversible actuation while ensuring that a single failure of a vent valve, power supply, or control system does not prevent isolation of the vent path. vene.\ The function, capabilities, and testing requirements of the Reactor Ceelent t+eal5y.w vents are consistent with the requirements of Item II.B.1 of NUREG-0737, " Clarification of TMI Action Plant Requirements," November 1980. Resumer pods serve m ved b %e_ premneer ad ofe. cmeal h S eihMoc p 3 /q, y, y, O

SOUTH TEXAS - UNIT 1 B 3/4 4-15

. _ _ _ . - ~ _ _ _ . _ - . _ . _ _ _ _ , _ _ _ . .- .e . _. . - - _.- . .- . -- ATTACHMENT 14 - ST.HL AE 1984 PAGE M OF 28 PROOF & REVIEW COPY 3/4.5 EMERGENCY CORE COOLING SYSTEMS O BASES 3/4.5.1 ACCUMULATORS The.0PERABILITY of each Reactor Coolant System (RCS) accumulator ensures that a sufficient volume of borated water will be immediately forced into the reactor core through three 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. l i 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 ower to the valves is required. fail to meet single failure criteria, removal ofgo exhr The limits.for operation with an accumulator inoperable for any reason except an isolation valve closed minimizes the ime exposure of the plant to a LOCA event occurring concurrent with failure of an additional accumulator which may result in unacceptable peak claddingj temperatures. -If a closed O isolation valve cannot be imustasy opened the full capability of one accumulatorisnotavailableandpromptacti[onisrequiredtoplacethe reactor in a mode where this capability is not required. 3/4.5.2 and 3/4.5.3 ECCS SUBSYSTEMS The OPERABILITY of three independent ECCS subsystems ensures that sufficient emergency core cooling capability will be available in the event of a LOCA assuming the loss of one subsystem through any single failure consideration. Each 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. One ECCS is assumed to discharge completely through the postulated break in the RCS loop. Note. M m - Thus, three trains are required to satsify the single failure criterion.f thatthecentpachECCSsubsystem I "9 0 ^a , ,. rovides long-term core cooling capability in the recirculation mode during th accident recovery period. awk %.a m 9 9 c4.heh d $e.ts When the RCS temperature is below 350*F, the ECCS requirements are balanced between the limitations imposed by the low temperature overpressure protection and the requirements necessary to mitigate the consequences of a LOCA below 350'F. 'At these temperatures, single failure considerations are not required because of the stable reactivity condition of the reactor and the limited core cooling requirements. Only a single Low Head Safety Injection pump is required ( to mitigate the effects of a large-break LOCA in this mode. However, two are [ -O- provided to accommodate the possibility that the break _ occurs in a loop con- - taining one of the Low Head pumps. Gor a smaII-break LOCA_in MODE 4. a single SOUTH TEXAS - UNIT 1 B 3/4 5-1 . ~e-- a w n--w- - ,-n-*e,--,. ,evn,.,e a,,me ,,ew,-vse ,m,,-,,,-er-r,--- a,--- ,-- ,,,---eae .,e -- - -,m.,,w- -,,---n -w- - - ,-- ,--ee w-,v,- -mn- -w,m ATTACHMENT I4 . ST-HL.AE 1984 P4 % .tB Of 2B rnwr & MtVIEW COPY EMERGENCY CORE COOLING SYSTEMS g h BASES ECCS SUBSYSTEMS (Continued) ow Head Safety Injection pump is sufficient if a High Head Safety Injection . pump can be OPERABLE in 30 minutes. Thus, a second High Head Safety Injection pump is required OPERABLE in 30 minutes to accommodate the possibility that the . initial High Head Safety Injection pump is located in the same RCS loop as the L break. This configuration provides assurance that a mass addition transient will involve only one High Head AS Injection pump, and can be relieved b <the operation of a single P0JR__1 Low Head Safety Injection pumps are not re-quired Inoperaoie petow 350"F because their shutoff head is too low to impact the low temperature overpressure protection limits. Below 200*F (MODE 5) no ECCS pumps are required, so the High Head Safety Injection pumps are locked out to prevent cold overpressure. The Surveillance Requirements provided to ensure OPERABILITY of each component ensure that, at a minimum, the assumptions used in the safety e analyses are met and that subsystem OPERABILITY is maintained. Surveillance Requirements for flow testing provide assurance that proper ECCS flows will be maintained in the event of a LOCA. 3/4.5.5 REFUELING WATER STORAGE TANK l l O d The OPERABILITY of the refueling water storage tank (RWST) as part of the l ECCS ensures that a sufficient supply of borated water is available for injection l by the ECCS in the event of a LOCA. The limits on RWST minimum volume and boron concentration ensure that: (1) sufficient water is available'within containment to permit recirculation cooling flow to the core, and (2) the reactor will remain suberitical in the cold condition following mixing of the RWST and the RCS water volumes with all control rods inserted except for the most reactive control assembly. These assumptions are consistent with the LOCA analyses. ~ The contained water volume limit includes an allowance for water not usable because of tank discharge line location or other physical characteristics. '7

The Ifmits on contained water lume/vo.

and 5' boron concentration of the RWST also ensure a pH value of betweenf8-5 and 10.0 for the solution recirculated within containment after a LOCA. This pH band minimizes the evolution of iodine and minimizes the effect of chloride and caustic stress corrosion on mechanical systems and components. 3/4.5.6 RESIDUAL HEAT REMOVAL (RHR) SYSTEM The OPERABILITY of the RHR system ensures adequate heat removal capabili-ties for Long-Term Core Cooling in the event of a small-break loss-of-coolant accident (LOCA), an isolatable LOCA, or a secondary break in MODES 1, 2, and 3. The limits on the OPERABILITY of the RHR system ensure that at least one RHR loop is available for cooling including single active failure criteria. [) The surveillances ensure that RHR system isolation valves close upon an overpressure protection system signal. 1 SOUTH TEXAS - UNIT 1 B 3/4 5-2 ATTACHMENT 14- - ST.HL AE.1984 PAGE 19 OF 28 PROOF & REVIEW COPY 3/4.6 CONTAINMENT SYSTEMS BASES 3/4.6.1 PRIMARY CONTAINMENT 3/4.6.1.1 CONTAINMENT INTEGRITY Primary CONTAINMENT INTEGRITY ensures that the release of radioactive , materials from the containment atmosphere will be restricted to those leakage l paths and associated leak rates assumed in the safety analyses. This restriction, in conjunction with the leakage rate limitation, will limit the SITE BOUNDARY radiation doses to within the dose guideline values of 10 CFR Part 100 during accident conditions. 3/4.6.1.2 CONTAINMENT LEAKAGE The limitations on containment leakage rates ensure that the total containment leakage volume will not exceed the value assumed in the safety analyses at the peak accident pressure, P,. As an added conservatism, the measured overall integrated leakage rate is further limited to less than or equal to 0.75 L, or 0.75 L t, as applicable, during performance of the periodic l test to account for possible degradation of the containment leakage barriers between leakage tests. The surveillance testing for measuring leakage rates is consistent with the requirements of Appendix J of 10 CFR Part 50. 3/4.6.1.3 CONTAINMENT AIR LOCKS The limitations on closure and leak rate for the. containment air locks are required to meet the restrictions on CONTAINMENT INTEGRITY and containment leak rate. Surveillance testing of the air lock seals provides assurance that the overall air lock leakage will not become excessive due to seal damage during the intervals between air lock leakage tests. , 3/4.6.1.4 INTERNAL PRESSURE 3,6 The limitations on containment internal pressure ensure that: (1) the containment structure is prevented from exceeding its design negative pressure differential with respect to the outside atmosphere of E psig, and (2) the containment peak pressure does not exceed the design pressure of 56.5 psig during LOCA or steam line break conditions. The maximum peak pressure expected to be obtained from a LOCA or steam line break event is 37.5 psig. The:.Mait of 0;#:psig-fee bit?e 7,:'t're een+*4nment pressttte wi=H Masi tk tutti peessure-to-3Wg, which ts tuss-thes design presstree and it n .. i+t-6t with the safety en !y ::. %e 'm%.l poshe. co M mc # prenon J be a Sp C . O SOUTH TEXAS - UNIT 1 8 3/4 6-1 ATTACHMENT -14 - ST.HL AE l964 'd En h C l'T CONTAINMENT SYSTEMS BASES 3/4.6.1.5 AIR TEMPERATURE The limitations on containment average air temperature ensure that the over-all containment average air temperature does not exceed the initial temperature condition assumed in the safety analysis for a LOCA or steam line break accident. Measurements shall be made st aM ttECe4 i=-U: + w-J.m by fixed oe. peri ble instruments, prior to determining the average air temperature. 3/4.6.1.6 CONTAINMENT STRUCTURAL INTEGRITY 'L7. .f This limitation ensures that the structural integrity of the containment will be maintained comparable to the original design standards for the life of the facility. Structural integrit is required to ensure that the containment will with-stand the maximum pressure of 39 psig in the event of a LOCA or steam line break accident. The measurement of containment tendon lift-off force, the tensile tests of the tendon wires or strands, the visual examination of tendons, anchorages and exposed interior and exterior surfaces of the containment, and the Type A leakage test are sufficient to demonstrate this capability. CThe tendon wire on.steend s3grht wtM siso bet 5chjE t M to steess ud hg testsr sad- te accsturztud ree-s-tun tssts to siadde the ten Js :p; ret N c~4*Me and ="&~=-at. ) The Surveillance Requirements for demonstrating the containment's structural integrity are in compliance with the recommendations of proposed Regulatory Guide 1.35, " Inservice Surveillance of Ungrouted Tendons in Prestressed Concrete l Containment Structures," April 1979, and proposed Regulatory Guide 1.35.1, " Deter-mining Prestressing Forces for Inspection of Prestressed Concrete Containments," l April 1979. The required Special Reports from any engineering evaluation of containment abnormalities shall include a description of the tendon condition, the condition of the concrete (especially at tendon anchorages), the inspection procedures, the tolerances on cracking, the results of the engineering evaluation, and the correc-tive actions taken. 3/4.6.1.7 CONTAINMENT VENTILATION SYSTEM The 48-inch containment purge supply and exhaust isolation valves are required to be sealed closed during plant operations since these valves have not been demonstrated capable of closing during a LOCA or steam line break accident. Maintaining these valves sealed closed during plant operation ensures that exces-sive quantities of radioactive materials will not be released via the Containment Purge System. To provide assurance that these containment valves cannot be inad-vertently opened, the valves are sealed closed in accordance with Standard Review I Plan 6.2.4 which includes mechanical devices to seal or lock the valve closed, or ! prevents power from being supplied to the valve operator. The use of the containment purge lines is restricted to the 18-inch purge supply and exhaust isolation valves since, unlike the 48-inch valves, the 18-inch valves are capable of closing during a LOCA or steam line break accident. There-SOUTH TEXAS - UNIT 1 8 3/4 6-2 ATTACHMENT I4 ST HL-AE- 1984 _ PAGE1l OF 10 VNUUF & REVlEW COPY m CONTAINMENT SYSTEMS BASES CONTAINMENT VENTILATION SYSTEM (Continued) fore, the SITE BOUNDARY dose guideline of 10 CFR Part 100 would not be exceeded in the event of an accident during containment PURGING operation. Leakage integrity tests with a maximum allowable leakage rate for containment purge supply and exhaust supply valves will provide early indication of resilient material seal degradation and will allow opportunity for repair before gross leak-age failures could develop. The 0.60 L leakage limit of Specification 3.6.1.2b. shall not be exceeded when the leakage fates determined by the leakage integrity tests of these valves are added to the previously determined total for all valves and penetrations subject to Type B and C tests. 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS 3/4.6.2.1 CONTAINMENT SPRAY SYSTEM The OPERABILITY of the Containment Spray System ensures that containment depressurization and cooling capability will be available in the event of a LOCA or steam line break. The pressure reduction and resultant lower containment leakage rate are consistent with the assumptions used in the safety analyses. p The Containment Spray System and the Containment Cooling System are redundant l l to each other in providing post-accident cooling of the containment atmosphere. l However, the Containment Spray System also provides a mechanism for removing i iodine from the containment atmosphere and therefore the time requirements for restoring an inoperable Spray System to OPERABLE status have been maintained l consistent with that assigned other inoperable ESF equipment. 3/4.6.2.2 SPRAY ADDITIVE SYSTEM 15" The OPERABILITY of the pray Additive System W the Ai_"iary Samp:. Acciria Tank ensures that sufficien NaOH is added to the containment spray and contain-ment sump in the event of a LOCA. The limits on NaOH volume and concentration ensure a pH value of between fkt and 10.0 for the solution recirculated within containment after a LOCA. This pH band minimizes the evolution of iodine and minimizes the effect of chloride and caustic stress corrosion on mechanical systems and components. The contained solution volume Ifmit includes an allowance for solution not usable because of tank discharge line location or other physical characteristics. These assumptions are consistent with the iodine removal efficiency assumed in the safety analyses. l l l O SOUTH TEXAS - UNIT 1 8 3/4 6-3 ST HL-AE 1984-PAGE 22.OF 26 PROOF & #EVIEW COPY PLANT SYSTEMS O BASES 3/4.7.1.2 AUXILIARY FEEDWATER SYSTEM The OPERABILITY of the Auxiliary Feedwater System ensures that the Reactor Coolant System can be cooled down to less than 350'F from normal operating conditions in the event of a total loss-of-offsite power. 9D . Eacjhauxiliaryfeedwaterpumpiscapableofdeliveringatotalfeedwater flow off556 gpm at a pressure of 1324 psig to the entrance of the steam generators. This capacity is sufficient to ensure that adequate feedwater flow is available to remove decay heat and reduce the Reactor Coolant System temperature to less than 350'F when th i plagegi,nto operatJ.on Wyg*fg%e w m8M.E vacdAGE TANK (AFST) 1 ira d .leen on o Am don s 3. - Residual Heat du R ew 3/4.7.1.3 AUXILIARY ST0 414.2 %.& ires %,,. ee 4 be veM44 m %. The OPERABILITY of the auxiliary feedwater storage tank with the minimum j water volume ensures that sufficient water is available to maintain the RCS at HOT STANDBY conditions for 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months with steam discharge to the atmosphere concurrent with total loss-of-offsite power followed by a cooldown to 350*F at 25'F per hour. The contained water volume limit includes an allowance for water not usable because of tank discharge line location or other physical characteristics. 3/4.7.1.4 SPECIFIC ACTIVITY The limitations on Secondary Coolant System specific activity ensure that the resultant offsite radiation dose will be limited to a small fraction of 10 CFR Part 100 dose guideline values in the event of a steam line rupture. This dose also includes the effects of a coincident 1 gpm primary-to-secondary j tube leak in the steam generator of the affected steam line. These values are consistent with the assumptions used in the safety analyses. 3/4.7.1.5 MAIN STEAM LINE ISOLATION VALVES The OPERABILITY of the main steam line isolation valves ensures that no more than one steam generator will blow down in the event of a steam line rupture. This restriction is required to: (1) minimize the positive reac-tivity effects of the Reactor Coolant System cooldown associated with the blowdown, and (2) limit the pressure rise within containment in the event the . steam line rupture occurs within containment. The OPERABILITY of the main steam isolation valves within the closure times of the Surveillance Require-ments are consistent with the assumptions used in the safety analyses. - 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION The limitation on steam generator pressure and temperature ensures that the pressure-induced stresses in the steam generators do not exceed the maximum allowable fracture toughness stress limits. The limitations of 70'F and O 200 psig are based on a steam generator RTNDT of 10'F and are sufficient to prevent brittle fracture. SOUTH TEXAS - UNIT 1 B 3/4 7-2 l ATTACHMENT 14 - ST HL-AE t984 PA E M pF,, R __ ' "vvi at ntytty gggy PLANT SYSTEMS v BASES SEALED SOURCE CONTAMINATION (Continued) source in that group. Those sources which are frequently handled are required to be tested more often than those which are not. Sealed sources which are continuously enclosed within a shielded mechanism (i.e., sealed sources within radiation monitoring or boron measuring devices) are considered to be stored and need not be tested unless they are removed from the shielded mechanism. [3/4.7.11 FIRE PROTECTION SYSTEMS The OPERABILITY of the Fire Protection Systems ensures that adequate fire suppression capability is available to confine and extinguish fires occurring in any portion of the facility where safety-related equipment is located. The Fire Protection System consists of the water system, spray, and/or sprinklers, Halon, fire hose stations, and yard fire hydrants. The collective capability of the Fire Protection Systems is adequate to [@ 6 minimize potential damage to safety-related equipment and is a major element in the facility Fire Protection Program. In the event that portions of the Fire Protection Systems are inoperable, alternate backup fire-fighting equipment is required to be made available in the affected areas until the inoperable equipment is restored to service. When the inoperable fire-fighting equipment is intended for use as a backup means of fire suppression, a longer period of time is allowed to provide an alternate means of fire fighting than if the inoperable equipment is the (' primary means of fire suppression. The Surveillance Requirements provide assurance that the minimum OPERABILITY requirements of the Fire Protection Systems are met. An allowance is made for ensuring a sufficient volume of Halon in the Halon storage tanks by verifying either the weight or the level of the tanks. Level measurements are made by either a U.L. or F.M. approved method. In the event the Fire Protection Water System becomes inoperable, immediate corrective measures must be taken since this system provides the major fire suppression capability of the plant. 3/4.7.12 FIRE RATED ASSEMBLIES O j The functional integrity of the fire rated assemblies and barrier penetra-tions ensures that fires will be confined or adequately retarded from spreading to adjacent portions of the facility. These design features minimize the pos-sibility of a single fire rapidly involving several areas of the facility prior to detection and extinguishing of the fire. The fire barrier penetrations are i a passive element in the facility Fire Protection Program and are subject to periodic inspections. Fire barrier penetrations, including cable penetration barriers, fire doors Qddampersareconsideredfunctionalwhenthevisuallyobservedconditionisthe SOUTH TEXAS - UNIT 1 8 3/4 7-6 ATTACHMENT 14-ST HL-AE 1984 PAGE 14 OF 2.9 PROOF & REVIEW COPY PLANT SYSTEMS BASES FIRE RATED ASSEMBLIES (Continued) same as the as-designed condition. For those fire barrier penetrations that are not in the as-designed condition, an evaluation shall be performed to show that the modification has not degraded the fire rating of the fire barrier penetration. During periods of time when a barrier is not functional, either: (1) a contin-uous fire watch is required to be maintained in the vicinity of the affected barrier, or (2) the fire detectors on at least one side of the affected barrier must be verified OPERABLE and an hourly fire watch patrol established until the barrier is restored to functional status. f f 3/4.7.13 AREA TEMPERATURE MONITORING The area temperature limitations ensure that safety-related equipment will not be subjected to temperatures in excess of their environmental qualification telhperatures. Exposure to excessive temperatures may degrade equipment and can cause a loss of its OPERABILITY. TE temperature timus inc4mie an attewance fur intrement erPwvf,* '4. O THIS PAGE OPEN PENDING RECEIPT OF INFORMATION FROM THE APPUCANT SOUTH TEXAS - UNIT 1 B 3/4 7-7 l ATTACHMENT I4 ST HL-AE 1984 , PAGE 25 OF 28 PROOF & REVlEW COPY 3/4.8 ELECTRICAL POWER SYSTEMS BASES 3/4.8.1, 3/4.8.2, and 3/4.8.3 A.C. SOURCES, D.C. SOURCES, and ONSITE POWER DISTRIBUTION The OPERABILITY of the A.C. and D.C power sources and associated distribu-tion systems during operation ensures that sufficient power will be available to supply the safety-related equipment required for: (1) the safe shutdown of the facility, and (2) the mitigation and control of accident conditions within the facility. The minimum specified independent and redundant A.C. and D.C. power sources and distribution systems satisfy the requirements of General Design Criterion 17 of Appendix A to 10 CFR Part 50. The ACTION requirements specified for the levels of degradation of the power sources provide restriction upon continued facility operation commensurate w ith the level of degradation. The OPERABILITY of the power sources are consistent with the initial condition assumptions of the safety analyses and are based upon maintaining at least two redundant sets of onsite A.C. and D.C. power sources and associated distribution systems OPERABLE during accident conditions coincident with an assumed loss-of-offsite power and single failure of the other onsite A.C. source. The A.C. and D.C. source allowable out-of-service times are based on Regulatory Guide 1.93, " Availability of Electrical Power Sources," December 1974. When one standby diesel generator is inoperable, there is an additional ACTION requirement to verify that all required systems, O subsystems, trains, components and devices, that depend on the remaining OPERABLE standby diesel generators as a source of emergency power, are also OPERABLE, and that the steam-driven auxiliary feedwater pump is OPERABLE. This require-ment is intended to provide assurance that a loss-of-offsite power event will not result in a complete loss of safety function of critical systems-during the period one of the standby diesel generators is inoperable. The term, verify, as used in this context means to administrative 1y check by examining logs or other information to determine if certain components are out-of-service for maintenance or other reasons. It does not mean to perform the Surveillance Requirements needed to demonstrate the OPERABILITY of the component. The OPERABILITY of the minimum specified A.C. and D.C. power sources and associated distribution systems during shutdown and refueling ensures that: (1) the facility can be maintained in the shutdown or refueling condition for extended time periods, and (2) sufficient instrumentation and control capa-I bility is available for monitoring and maintaining the unit status. l The Surveillance Requirements for demonstrating the OPERABILITY of the l diesel generators are in accordance with the recommendations of Regulatory l Guides 1.9, " Selection of Diesel Generator Set Capacity for Standby Power l Supplies," Revision 2, December 1979; 1.108, " Periodic Testing of Diesel l Generator Units Used as Onsite Electric Power Systems at Nuclear Power Plants," l Revision 1, August 1977; and 1.137, " Fuel-Oil Systems for Standby Diesel l Generators," Revision 1, October 1979. The standby diesel generators auxiliary ! systems are designed to circulate warm oil and water through the diesel while l p the diesp1 is not running, to preclude cold ambient starts. L % p es c4 V surve%nce -Wh , cues cudrhons are_ cenAdered % lae the o 4-' Prelded ced on SOUTH TEXAS - UNIT 1 B 3/4 8-1 l l runn4m@JU BQ ~ ST HL AE- t*184-PAGE 26 0F 28 PROOF & REVIEW COPY RADI0 ACTIVE EFFLUENTS BASES DOSE (Continued) This specification applies to the release of radioactive materials in liquid effluents from each unit at the site. %s4c. Meewi 3/4.11.1.3 LIQUID RADWaME - SYSTEM mde Pac m The OPERABILITY of the Liquid R s 2t+ Treatmen%t' System ensures that this system will be available for use whenever liquid effluents require treatment prior to release to the environment. The requirement that the appropriate portions of this system be used when specified provides assurance that the releases of radioactive materials in liquid effluents will be kept "as low as is reasonably achievable". This specification implements the requirements of 10 CFR 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50 and the design objective given in Section II.D of Appendix I to 10 CFR Part 50. The specified limits governing the use of appropriate portions of the Liquid Rsnte L-hnt System were specified as a suitable fraction of the dose design objectives set forth in Section II. A of Appendix I,10 CFR Part 50, for liquid effluents. This specification applies to the release of radioactive materials in wte%u) l liquid effluents from each unit at the site. 3/4.11.1.4 LIQUID HOLDUP TANKS CJ The tanks listed in this specification include all those outdoor radwaste tanks that are not surrounded by liners, dikes, or walls capable of holding the tank contents and that do not have tank overflows and surrounding area drains connected to the Liquid Radwaste Treatment System. Restricting the quantity of radioactive material contained in the specified ', tanks provides assurance that in the event of an uncontrolled release of the tanks' contents, the resulting concentrations would be less than the limits of 10 CFR Part 20, Appendix B, Table II, Column 2, at tiensavest W. Ale water supply and the nearest surface water supply in an UNRESTRICTED AREA. 3/4.11.2 GASEOUS EFFLUENTS J 3/4.11.2.1 DOSE RATE This specification is provided to ensure that the dose at any time at and l beyond the SITE BOUNDARY from gaseous effluents from all units on the site l I will be within the annual dose limits of 10 CFR Part 20 to UNRESTRICTED AREAS. The annual dose limits are the doses associated with the concentrations of 10 CFR Part 20, Appendix B, Table II, Column 1. These limits provide reasonable assurance that radioactive material discharged in gaseous effluents will not I result in the exposure of a MEMBER OF THE PUBLIC in an UNRESTRICTED AREA, either within or outside the SITE BOUNDARY, to annual average concentratiods p exceeding the limits specified in Appendix 8. Table II of 10 CFR Part 20 - V (10 CFR Part 20.106(b)). For MEMBERS OF THE PUBLIC who may at times be within SOUTH TEXAS - UNIT 1 B 3/4 11-2 ATTACHMENT 14 . . ST.HL.AE.1984 PAGE 21 OF 26 _ . RADIOACTIVE EFFLUENTS PROOF & REVIEW COPY . BASES GASEOUS WASTE PROCESSING SYSTEM (Continued) 4 of the system were specified as a suitable fraction of the dose design objec-tives set forth in Section II.B and II.C of Appendix I, 10 CFR Part 50, for gaseous effluents. ~ This specification applies to the release of radioactive materials in gaseous effluents from each unit at the sitej when shared Radwaste Treatment r!Iy~ stems are used by more than one unit on a site, the wastes from all units are mixed for shared treatment; by such mixing, the affluent releases cannot accu-rately be ascribed to a specific unit. An estimate should be made of the con-tributions from each unit based cn input conditions, e.g. , flow rates and radio-activity concentrations, or, if not practicable, the treated effluent releases may be allocated equally to each of the radioactive waste producing units shar-ing the Radwaste Treatment System. For determining conformance to LCOs, these allocations from shared Radwaste Treatment Systems are to be added to the re-leases specifically attributed to each unit to obtain the total releases per tunit. f ~ , 3/4.11.2.5 EXPLOSIVE GAS MIXTURE This specification is provided to ensure that the concentration of poten-l tially explosive gas mixtures contained in the GASE0US WASTE PROCESSING SYSTEM is maintained below the flammability limits of hydrogen and oxygen. The con-l centration of oxygen in the inlet header to the GASEOUS WASTE PROCESSING SYSTEM t is continuously monitored and a high level alarm isolates the' GASEOUS WASTE . PROCESSING SYSTEM. Provision is made to manually purge the system with nitrogen l' and/or isolate the source of oxygen. Maintaining the concentration of hydrogen and oxygen below their flammability limits provides assurance that the releases , of radioactive materials will be controlled in conformance with the requirements l of General Design Criterion 60 of Appendix A to 10 CFR Part 50. 3/4.11.2.6 GAS STORAGE TANKS The tanks included in this specification are those tanks for which the - quantity of radioactivity contained is not limited directly or indirectly by another Technical Specification. Restricting the quantity of radioactivity contained in each gas storage tank provides assurance that in the event of an uncontrolled release of the tank's contents, the resulting whole body exposure to a MEMER 0F THE PUBLIC at the nearest SITE BOUNDARY will not exceed 0.5 rem. This is consistent with Standard Review Plan 11.3, Branch Technical Position ETSB 11-5, " Postulated Radioactive Releases Due to a Waste Gas System Leak or Failure," in NUREG-0800, July 1981. Since only the gamma body dose factor (DFBg ) , is used in the analysis, the Xe-133 equivalent is determined from the DFBj value for Xe-133 as compared to the composite DFB9 for the actual mixture in the tank. !O SOUTH TEXAS - UNIT 1 B 3/4 11-5 t.-... _ . - . . . . . . . ~ . . -.. . _ , , - . . - _ _ .-.r_.- ,m.,_y.w_,m _,.-...m , , , , , , , - - _ , , . ATTACHMENT 14 . ST.HL AE 198A .. RAGE 7tLDE#.. PROOF & REVIEW COPY 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING BASES 3/4.12.1 MONITORING PROGRAM The Radiological Environmental Monitoring Program required by this specification provides representative measurements of radiation and of radio-active materials in those exposure pathways and for those radionuclides that lead to the highest potential radiation exposure of MEMBERS OF THE PUBLIC resulting from the plant operation. This monitoring program implementsSection IV.B.2 of Appendix I to 10 CFR Part 50 and thereby supplements the Radiological Effluent Monitoring Program by verifying that the measurable concentrations of radioactive materials and levels of radiation are not higher than expected on the basis of the effluent measurements and the modeling of the environmental exposure pathways. Guidance for this monitoring program is provided by the Radiological Assessment Branch Technical Position on Environ-mental Monitoring. The initially specified monitoring program will be effective for at least the first 3 years of commercial operation. Following this period, program changes may be initiated based on operational experience. The required detection capabilities for environmental sample analyses are Scp b i=i: > bulated in terms of the lower limits of detection (LLDs). The LLDs required 1121 are considered optimum for routine environmental measurements in ndustrial laboratories. It should be recognized that the LLD is defined as an a priori (before the fact) limit representing the capability of a measure-ment system and not as an a posteriori (after the fact) limit for a particular measurement. Detailed discussion of the LLD, and other detection limits, can be found in HASL Procedures Manual, HASL-300 (revised annually), Currie, L. A. , " Limits for Qualitative Detection and Quantitative Determination - Application to Radiochemistry," Anal. Chem. 40, 586-93 (1968), and Hartwell, J. K. , " Detection Limits for Radioanalytical Counting Techniques" Atlantic Richfield Hanford Company Report ARH-SA-215 (June 1975). 3/4.12.2 LAND USE CENSUS This specification is provided to ensure that changes in the use of areas i at and beyond the SITE BOUNDARY are identified and that modifications to the Radiological Environmental Monitoring Program are made if required by the results of this census. The best information from the door-to-door survey, from aerial survey or from consulting with local agricultural authorities ~ shall be used. This census satisfies the requirements of Section IV.B.3 of Appendix I to 10 CFR Part 50. Restricting the census to gardens of greater l than 50 m2 provides assurance that significant exposure pathways via leafy vegetables will be identified and monitored since a garden of this size is the minimum required to produce the quantity (26 kg/ year) of leafy vegetables assumed -in Regulatory Guide 1.109 for consumption by a child. To determine this minimum garden size, the following assumptions were made: (1) 20% of the garden was used for growing broad leaf vegetation (i.e., similar to lettuce and cabbage), and (2) a vegetation yield of 2 kg/m2, SOUTH TEXAS - UNIT 1 B 3/4 12-1 3 ATTACHMENT /5 ST HL AE /9#V PAGE / OF3 '5.0 Design Features A. 5-1: Specific STP data provided; see FSAR Section 1.2.2.3.1. B. 5-8: Justification provided in letter ST-HL-AE-1867. L3/NRC/cm ATTACHMENT /5 ST.HL-AE /9fV PRODF AWFVirw MDV. 3 -. ... 5.0 DESIGN FEATURES 5.1 SITE EXCLUSION AREA 5.1.1 The Exclusion Area shall be as shown in Figure 5.1-1. LOW POPULATION ZONE 5.1. 2 The Low Population Ione shall be as shown in Figure 5.1-2. MAP DEFINING UNRESTRICTED AREAS AND SITE BOUNDARY FOR RADI0 ACTIVE GASEOUS AND LIQUID EFFLUENTS 5.1.3 Information regarding radioactive gaseous and liquid effluents, which will allow identification of structures and release points as well as defini-tion of UNRESTRICTED AREAS within the SITE BOUNDARY that are accessible to MEMBERS OF THE PUBLIC, shall be as shown in Figures 5.1-3 and 5.1-4. The definition of UNRESTRICTED AREA used in implementing these Technical Speci-fications has been expanded over that in 10 CFR 20.3(a)(17). The UNRESTRICTED AREA boundary may coincide with the Exclusion (fenced) Area boundary, as defined in 10 CFR 100.3(a), but the UNRESTRICTED AREA does not include areas over water bodies. The concept of UNRESTRICTED AREAS, established at or beyond the SITE BOUNDARY, is utilized in the Limiting Conditions for Operation to keep levels of radioactive materials in liquid and gaseous effluents as low as is reason-ably achievable, pursuant to 10 CFR 50.36a. 5.2 CONTAINMENT CONFIGURATION 5.2.1 The reactor containment building is a steel-lined, reinforced concrete building of cylindrical shape, with a dome roof and having the following design features:

a. Nominal inside diameter = 150 feet. in

b. -239.J Nominal inside height = 242:2yfeet,

c. Minimum thickness of concrete walls = 4 feet.

d. Minimum t'hickness of concrete roof =gfeet.

e. Minimum thickness of concrete floor pad = $ feet (A o h%

f. Nominal thickness of steel liner = 3/8 inches. b to n M D .

g. Net free volume = 3.56 x 10s cubic feet.

DESIGN PRESSURE AND TEMPERATURE 5.2.2 The reactor containment building is designed and shall be maintained for a maximum internal pressure of 56.5 psig and a temperature of 286'F. , SOUTH TEXAS - UNIT 1 5-1 K , % 't.} , TABLE 5.7-1 O y COMPONENT CYCLIC OR TRANSIENT LIMITS M 5 CYCLIC OR DESIGN CYCLE y COMPONENT TRANSIENT LIMIT OR TRANSIENT E Reactor Coolant System 200 heatup cycles at < 100*F/h Heatup cycle - T Z and 200 cooldown cycles at avg from 5 200 F to > 550 F. - - 100*F/h. CooTdown cycle - T**9 from > 550*F to 5 200*F 200 pressurizer cooldown cycles Pressurizer cooldown cycle at < 200*F/h. temperatures from > 650 F to 5 200*F. 80 loss of load cycles, without > 15% of RATED THERMAL POWER to immediate Turbine or Reactor trip. 0% of RATED THERMAL POWER. T 40 cycles of loss-of-offsite Loss-of-offsite A.C. electrical A.C. electrical power. ESF Electrical System. 80 cycles of loss of flow in one loss of only one reactor reactor coolant loop. coolant pump. 400 Reactor trip cycles. 100% to 0% of RATED THERMAL POWER. (4c%T p<num rev 103 auxiliaryAspray Spray water temperature differential actuation cycles. > 320*F. Gal 200 leak tests. , gm> Pressurized to > 2485 psig. o 10 :.ydrostatic pressure tests. Pressurized to > 3100 psig. [ %g amQ TTt Q $~t Secondary Coolant System 1 steam line break. Break in a > 6-inch steam line. i rr!

1600 psig, e c 'S 1 i ATTACHMENT & . ST.HL AE /9#V PAGE /-OF 6 6.0 Administrative Controls. L A. 6-1: Deletion justification provided in letter ST-HL-AE-1867. B. 6-8: (1) PORC procedure review is on quality-related station administrative procedures as defined in the Operations QA Program. (2) Addition of Fire Protection Program justified in letter . ST-HL-AE-1867. C. 6-10: STP nomenclature provided. .The NSRB Manager's title is NSRB Director. D. 6-14: (1) Justification for Addition of Fire Protection to 6.8.1 was provided in. letter ST-HL-AE-1867. (2) PORC does not review all procedures pertaining to Appendix A-of Regulatory Cuide 1.33. (3) Justification for Primary Sources Outside of containment is provided in ST-HL-AE-1882. E. 6-20: Change made to reflect Boron Dilution Analysis. ? I t I L3/NRC/cm I n 3 ATTACHMENT /4 . .AE /97'l , &@EVIEW cnPY a e ADMINISTRATIVE CONTROLS 6.1 RESPONSIBILITY 6.1.1 The Plant Manager shall be responsible for overall unit operation and shall delegate in writing the succession to this responsibility during his absence. 6.1.2 The Shift Supervisor (or during his absence from the control room, a designated individual) shall be responsible for the control room command function. A management directive to this effect, signed by the Group Vice President, Nuclear shall be reissued to all station personnel on an annual basis. 6.2 ORGANIZATION 0FFSITE 6.2.1 The offsite organization for unit management and technical support shall be as shown in Figure 6.2-1.

UNIT STAFF 1

6.2.2 The unit organization shall be as shown in Figure 6.2-2 and:

a. Each on-duty shift shall be composed of at least the minimum shift crew composition shown in Table 6.2-1;

b. At least one iicensed Operator shall be in the control room when In addition, while the unit'is in MODE 1, l fuel is in the reactor.

2, 3, or 4, at least one licensed Senior Operator shall be in the control room;

c. A Health Physics Technician

shall be on site when fuel is in the reactor; e

d. All CORE ALTERATIONS shall be observed and directly supervised by either a licensed Senior Operator or licensed Senior Operator Limited

to Fuel Handling who has no other concurrent responsibilities during this operation; A site Fire Brigade of at least five members

shall be maintaine D i

.. site at all times. The Fire Brigade shall not include the Shift Supervisor and the two other members of the minimum shift crew / necessary for safe shutdown of the unit and any personnel requireg/ . Qorotheressentialfunctionsduringafireemergency;andf e.,ff Administrative procedures shall be developed and implemented to limit the working hours of unit staff who perform safety-related functions (e.g., licensed Senior Operators, licensed Operators, health physicists, auxiliary operators, and key maintenance personnel).

The Health Physics Technician 6nd Fire Brigad position may be less than .

the minimum requirements for a period of time not to exceed 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , in order O to accommodate unexpected absence, provided immediate action is taken to fill the required positions. SOUTH TEXAS - UNIT 1 6-1 J ATTACHMENTVG . . ST HL.AE.198y I pAM !.QF G_ s rMUUr & Ktvitwm0PY1 ^ ADMINISTRATIVE CONTROLS _ 4 RESPONSIBILITIES , 6.5.1.6 The PORC shall be responsiblex for: ,

a. Review of all related station administrativ procedures and changes thereto.

b. Review of safety evaluations for (1) procedures, (2) changes to s procedures, structures, components, or systems, and (3) tests or ' '

experiments completed under the provision of 10 CFR 50.59 to verify that such actions did not constitute an unreviewed safety question. ,

c. Review of proposed (1) procedures, (2) changes to procedures, struc- .

tures, components, or systems, and (3) tests or experiments which may involve an unreviewed safety question as defined in 10 CFR 50.59.

d. Review of al'1 programs and procedures required by Specificatio' n 6.8 and changes thereto.

e. ReviewofproposedchangestotheTechdicalSpecificationsorthe Operating License.

f. Review of all REPORTABLE EVENTS. , 1

g. Review of reports of significant operating abnormalities or devia-tions from normal and expected performance of plant equipment or i systems that affect nuclear safety. ,>

h. Review of reports of unanticipated defie.iencies in the tiesign' or

( operation of structures, systems, or components that affect nuclear safety.

i. Review of the Security Plan and implementing procedures and changes thereto.

j. Review of the Emergency Plan and implementing proceduras and changes thereto.

k. Review of the PROCESS CONTROL PROGRAM and implementing procedures and changes thereto. ,

1. Review of the CFFSITE DOSE CALCULATION 'MANUAL and implementing proce-dures and changes thereto. ,

m. Performance of special review, investigations, or analyses and reports thereon as requested by the Plant Manager or the Nuclear Safety Review Board (NSRB).

n. Review of any accidental, unplanned, or uncontrolled radioactive re- ,a lease including the preparation of reports ::overing evaluation, rec-ommendations, and disposition of the correctiva action to prevent recurrence and the forwarding of these report.s to the Plant Manager and to the NSRB.

o. Reports of violations of codes, regulations, orders, Technical Speci-fications, or Operating License requirements having nuclear safety significance or reports of abnormal degradition of systems' designed to contain radioactive material.

o 4 -\ Q V ice M ec M o 9 M. g. Reece&qpM.Ad&nh q~ co %m ,5% ' - SOUTH TEXAS - UNIT 1 6-8 ATTACHMENT /G M ST HL.AE /96 p ,. E vEO'COFi gINISTRATIVECONTROLS V FUNCTION (Continued) The NSRB shall report to end advise the Group Vice President-Nuclear on those s areas of rasponsibility specified in Specifications 6.5.2.7 and 6.5.2.8. C0F i810N 6.5.2.2 The NSRB shall be composed of ttic a.c1be d d ded b "** " Er SS: N RB P+=hh8er Member: General Manager, South Texas Project Member: Deputy Project Manager, Engineering and Licensing Member: Plant Manager Member: General Manager, Nuclear Assurance Member: Manager, Nuclear Engineering ALTERNATES 6.5.2.3 All alternate members shall be appointed in writing by the Group Vice President-Nuclear to serve on a temporary basis; however, no more than two alternates shall participate as voting members in NSRB activities at any one time. CONSULTANTS 6.5.2.4 Consultants saall be utilized as determined by the NSRB Director s to provide expert advice to the NSRB. MEETING FREQUENCY f 6.5.2.5 The NSRB shall meet at least once per calendar quarter during the l initial year of unit operation following fuel loading and at least once per 6 months thereafter. QUORUM r 6.5.2.6 The quorum of the NSRB necessary for the performance of the NSRB l' - - ' review and audit functions of these Technical Specifications shall consist of the Director or his designated alternate and at least four NSRB members l including alternates. No more than a minority of the quorum shall have line l responsibilit;/ for operation of the unit,

i. REVIEW l

l 6.5.2.7 The NSRB shall be responsible for the review of: i l

a. The saf tty saluations for: (1) changes to procedures, equipment, or systems; and (2) tests or experiments completed under the provision of 10 CFR 50.59, to verify that such actions did not constitute an unreviewed safety question;

b. Proposed changes to procedures, equipment, or systems which involve an unreviewed safety question as defined in 10 CFR 50.59; SOUTH TEXAS - UNIT 1 6-10 i

m ________d ATTACHMENT & . l . ST.HL AE /9fy PAGE f OF fo PR0OF & REVIEW COPY ADMINISTRATIVE CONTROLS /G. V SAFETY LIMIT VIOLATION (Continued)

a. The NRC Operations Center shall be notified by telephone as soon as possible and in all cases within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . The Group Vice President-Nuclear and the NSRB shall be notified within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ;

b. A Safety Limit Violation Report shall be prepared. The report shall be reviewed by the PORC. This report shall describe: (1) applicable circumstances preceding the violation, (2) effects of the violation upon facility components, systems, or structures, and (3) corrective action taken to prevent recurrence;

c. The Safety Limit Violation Report shall be submitted to the Commission, the N3B, and the Group Vice President-Nuclear within 14 days of the violation; and

d. Operation of the unit shall not be resumed until authorized by the Commission.

6.8 PROCEDURES AND PROGRAMS 6.8.1 Written procedures shall be established, implemented, and maintained covering the activities referenced below:

a. The applicable procedures recommended in Appendix A of Regulatory Guide 1.33, Revision 2, February 1978; -

i b. The emergency operating procedures required to implement the require-f) d ments of NUREG-0737 and Supplement 1 to NUREG-0737 as stated in Generic Letter No. 82-33; l c. Security Plan implementation;

d. Emergency Plan implementation;

e. PROCESS CONTROL PROGRAM implementation;

f. OFFSITE DOSE CALCULATION MANUAL implementation; and-

g. QualityAssuranceProgramforeffluentandenvironmentalmonitoring'.M

h. R e P wA u.tien b e 6.8.2 Each procedure of Spec $rcow 6.8.1,holex#diion l x ficatiod and etanges thereto shall be reviewed by th POP,C and--sher be e-+ d b5e M4 7 tant-Haacgr prior to implementation and reviewed periodically as W fo .h in Specification 6.5.

6.8.3 The following programs shall be estd.fisi d . implemented, and maintained: l

a. Primary Coolant Sources Outside Containment A program to reduce leakage from those portions of systems outside containment that could contain hi0hly radioactive fluids during a serious transient or accident to as low as practical levels. The ami vd=* w= ul , and. RC$ s ; ' i 0 mhaf[%

systems include the containment spray, Safety Injection,pi=1 %daryn **' The program shall include (* h a d ' the following: W'"3 "^ pww) septm s 1) Preventive maintenance and periodic visual inspection requirements,

and SOUTH TEXAS - UNIT 1 6-14 a

n. s nwmn is y

. ST.HL AE /f Py Bl.QFA PROOF & REVIEW COPY ADMINISTRATIVE CONTROLS f^ d SEMIANNUAL RADI0 ACTIVE EFFLUENT RELEASE REPORT (Continued) The Semiannual Radioactive Effluent Release Reports shall also include the following: an explanation as to why the inoperability of liquid or gaseous effluent monitoring instrumentation was not corrected within the time specified in Specification 3.3.3.10 or 3.3.3.11, respectively; and description of the I events leading to liquid holdup tanks or gas storage tanks exceeding the limits of Specification 3.11.1.4 or 3.11.2.6, respectively. MONTHLY OPERATING REPORTS 6.9.1.5 Routine reports of operating statistics and shutdown experience, including documentation of all challenges to the PORVs 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 Administrator of the Regional Office of the NRC, no later l than the 15th of each month following the calendar month covered by the report. RADIAL PEAKING FACTOR LIMIT REPORT 6.9.1.6 The F xy limits for RATED THERMAL POWER (Fx ) and the =h4=n SHUTUURN MARCIR regnin +x h***d ce buon dH"+4n-a;;! Gat andly is shall be estab-i lished for at least each reload core and shall be maintained available in the l Control Room. The limits shall be established and implemented on a time scale l O consistent with normal procedural changes. C The analytical methods used to generate the Fxy limits end the SHUTDO@MAMfM r p ? : --- nM based on-boren t!Huthn aec-+=nt andysis shall be those previously approved by the NRC.* If changes to these methods are deemed necessary, they will be evaluated in accordance with 10 CFR 50.59 and submitted to the NRC for review and approval prior to their use if the changes are determined to involve l an unreviewed safety question or if such a change would require amendment of j previously submitted documentation. A report containing the F xy limits for all core planes containing Bank "0" con-l ., trol rods and all unrodded core planes and the plot of predicted (F q.PRel) vs Axial Core Height with the limit envelope and the minimum S WT00WR MAR &iN ce-qui: as based en:twon dMution accient an;'y i4- 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. SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the Regional Administrator of the Regional Office of the NRC within the time period specified for each report.

WCAP-8385, " Power Distribution Control and Load Follow Procedures"; WCAP-9272.A, O " Westinghouse Reload Safety Evaluation Methodology"; and NUREG-0797, % fdy-Evstaation itsput- Rahted to the Opeha of- Comanche iteak TTettz+c 5m.lurr Untts:1 and 2;" as d4+cwn in the JEMEny-2&,1984 Ivtter (ST-4L- AE IN),

Sectden Et.3.t SOUTH TEXAS - UNIT 1 6-20 l

ST-HL-AE-1984, Forwards,For Info & Use,Comments & Justifications for Changes Proposed in Proof & Review Tech Specs

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Lag compensator on measured AT; y 13

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ST-HL-AE-1984, Forwards,For Info & Use,Comments & Justifications for Changes Proposed in Proof & Review Tech Specs

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Lag compensator on measured AT; y 13

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