Text

Amends 61 & 50 to Licenses NPF-76 & NPF-80,respectively Re Changes to TS & UFSAR in Response to Ltr
	ML20069M307
Person / Time
Site:	South Texas
Issue date:	05/27/1994
From:	Lyons J; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20069M310	List: ML20069M307; ML20069M327;
References
	NUDOCS 9406210286
	Download: ML20069M307 (69)
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E UNITED STATES (NMf j NUCLEAR REGULATORY COMMISSION

j WASHINGTON, D.C. 2055%C001 r

HOUSTON LIGHTING & POWER COMPANY CITY PUBLIC SERVICE BOARD OF SAN ANTONIO j

CENTRAL POWER AND LIGHT COMPANY CITY OF AUSTIN. TEXAS DOCKET N0. 50-498

^

SOUTH TEXAS PROJECT. UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 61 License No. NPF-76 1.

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

A.

The application for amendment by Houston Lighting & Power Company *

(HL&P) acting on behalf of itself and for the City Public Service Board of San Antonio (CPS), Central Power and Light Company (CPL), and i

City of Austin, Texas (C0A) (the licenseer) dated May 27, 1993, as supplemented by letter dated April 18, 1994, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, as amended, the provisions of the Act, and the rules and regulations of I

the Commission; C.

There is reasonable assurance:

(i) that the activities authorized by 4

this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this license amendment will not be inimical to the common defense and-security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfie'd.

Houston _ Lighting & Power Company is authorized to act for the City Public Service Board of San Antonio, Central Power and Light Company and City of Austin, Texa's and has exclusive responsibility and control over the physical construction, operation and maintenance of the facility.

9406210286 940527 PDR ADOCK 05000498 P

PDR

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. 2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and Paragraph 2.C.(2) of Facility Operating License No. NPF-76 is hereby,

amended to read as follows:

2.

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No.

61, and the Environmental Protection Plan

' contained in Appendix B, are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Technical 5pecifications and the Environmental Protection Plan.

3.

The license amendment is effective as of its date of issuance and is to be implemented prior to the completion of the Unit 1 RE05 outage.

FOR THE NUCLEAR REGULATORY COMMISSION Pr M

A Jabes E. Lyons, A lcting' Director P/oject Directorate IV-2 Division of Reactor Projects III/IV ffice of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance:

May 27, 1994 m

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UNITED STATES i

NUCLEAR REGULATORY COMMISSION g

j WASHINGTON, o.C. 20555-0001 HOUSTON LIGHTING & POWER COMPANJ CITY PUBLIC SERVICE BOARD OF SAN ANTONIO CENTRAL POWER AND LIGHT COMPANY CITY OF AUSTIN. TEXAS DOCKET NO. 50-499 SOUTH TEXAS PROJECT. UNIT 2 i

AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 50 License No. NPF-80 1.

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

A.

The application for amendment by Houston Lighting & Power Company *

(HL&P) acting on behalf of itself and for the City Public Service Board of San Antonio (CPS), Central Power and Light Company (CPL), and City of Austin, Texas (C0A) (the licensees) dated May 27, 1993, as supplemented by letter dated April 18, 1994, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and. regulations set forth in 10 CFR Chapter I; B.

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

There is reasonable assurance:

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

The issuance of this license amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR fart 51 of the Commission's regulations and all applicuble requirements have been satisfied.

Houston Lighting & Power Company is authorized to act for the City Public Service Board of San Antonio, Central Power and Light Company and City of Austin, Texas and has exclusive responsibility and control over the physical construction, operation and maintenance of the facility.

. 2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and Paragraph 2.C.(2) of Facility Operating License No. NPF-80 is hereby amended to read as follows:

2.

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No.

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

The

' licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

The license amendment is effective as of its date of issuance and is to be implemented prior to the completion of the Unit 1 RE05 outage.

FOR THE NUCLEAR REGULATORY COMMISSION y; b?x i.

W JamesE.Lyons,Actinbs Director Project Directorate IV12 Division of Reactor Projects III/IV Off' ice of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance:

May 27, 1994 4

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ATTACHMENT TO LICENSE AMENDMENT NOS. 61 AND 50 FACIllTY OPERATING LICENSE NOS. NPF-76 AND NPF-80 DOCKET NOS. 50-498 AND 50-499 Replace the following pages of the Appendix A Technical Specifications with the attached pages.

The revised pages are identified by Amendment number and contain marginal lines indicating the areas of change.

The corresponding overleaf pages are also provided to maintain document completeness.

REMOVE INSERT viii viii xvii xvii 2-2 2-2 2-4 2-4 2-5 2-5 2-6 2-6 i

2-7 2-7 2-8 2-8 2-10 2-10 B 2-1 B 2-1 3/4 1-3 3/4 1-3 j

3/4 1-5 3/4 1-5 3/4 1-7a 3/4 1-7a 3/4 2-11 3/4 2-11 3/4 3-29 3/4 3-29 3/4 3-30 3/4 3-30 3/4 3-31 3/4 3-31 3/4 3-32 3/4 3-32 3/4 3-34 3/4 3-34 3/4 3-36 3/4 3-36 3/4 6-1 3/4 6-1 3/4 6-2 3/4 6-2 3/4 6-3 3/4 6-3 3/4 6-4 3/4 6-4 3/4 6-5 3/4 6-5 3/4 6-6 3/4 6-6 3/4 6-8 3/4 6-8 3/4 7-5 3/4 7-5 B 3/4 1-1 B 3/4 1-1 B 3/4 1-2 B 3/4 1-2 B 3/4 2-4 8 3/4 2-4 B 3/4 2-6 B 3/4 2-6 B 3/4 6-1 B 3/4 6-1 B 3/4 6-2 B 3/4 6-2 B 3/4 7-2 B 3/4 7-2 e

.,,._.,.-.-y.

. REMOVE INSERT 5-1 5-1 5-6 5-6 5-7 5-7 5-9 5-9 5-9a 5-16 5-16 5-17

_ _ _. =

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMEN SECTION PAGE 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION Startup and Power O Hot Standby........peration.....:........................

3/4 4-1 Hot Shutdown.....................

3/4 4-2 3/4 4-3 Cold Shutdown - Loops Filled.....................

3/4 4-5 Cold Shutdown - Loops Not Filled.......................

3/4 4-6 3/4.4.2 SAFETY VALVES Shutdown...............

0perating..............

3/4 4-7 3/4 4-8 3/4.4.3 PRESSURIZER..............................................

3/4 4-9 j

3/4.4.4 RELIEF VALVES............................................

3/4 4-10 3/4.4.S STEAM GENERATORS.........................................

3/4 4-12 TABLE 4.4-1 MINIMUM NUMBER OF STEAM GENERATORS TO BE INSP DURING INSERVICE INSPECTION......................ECTED 3/4 4-17 TABLE 4.4-2 STEAM GENERATOR TUBE INSPECTION.......................

3/4 4-18 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems.........

Operational Leakage...............

3/4 4-19 3/4 4-20 TABLE 3.4-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES......

3/4 4-22 3/4.4.7 CHEMISTRY................................................

3/4 4-23 TABLE 3.4-2 REACTOR COOLANT SYSTEM CHEMISTRYLIMITS...............

3/4 4-24 TABLE 4.4-3 REACTOR COOLANT SYSTEM CH REQUIREMENTS................EMISTRY LIMITS SURVEILLANCE 3/4 4-25 3/4.4.8 SPECIFIC ACTIVITY........................................

3/4 4-26 FIGURE 3.4-1 DOSE EQUIVALENT I-131 REACTOR COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE REACTOR COOLANT SPECIFIC ACTIVITY >l pCi/ gram DOSE EQUIVALENT I-131....................................

3/4 4-28 TABLE 4.4-4 REACTOR COOLANT SPECIFIC AC PR0 GRAM..................,.....TIVITY SAMPLE AND ANALYSIS 3/4 4-29 3/4.4.9 PRESSURE / TEMPERATURE LIMITS Reactor Coolant System...................................

3/4 4-31 t

FIGURE 3.4-2 REACTOR COOLANT SYSTEM HEATUP LIMITATI APPLICABLE UP TO 32 EFPY..................ONS 3/4 4-32 I

+

SOUTH TEXAS - UNITS 1 & 2 vii

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION f%G1 FIGURE 3.4-3 REACTOR COOLANT SYSTEM C00LDOWN LIMITATIONS -

APPLICABLE UP TO 32 EFPY 3/4 4-33 TABLE 4.4-5 (This table number not used)............

3/4 4-34 Pressurizer....................

3/4 4-35 Overpressure Protection Systems..........

3/4 4-36 FIGURE 3.4-4 NOMINAL MAXIMUM ALLOWABLE PORV SETPOINT FOR THE COLD OVERPRESSURE SYSTEM 3/4 4-38 3/4.4.10 STRUCTURAL INTEGRITY................

3/4 4-39 3/4.4.11 REACTOR VESSEL HEAD VENTS..............

3/4 4-40 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS....................

3/4 5-1 3/4.5.2 ECCS SUBSYSTEMS - T,y, GREATER THAN OR EQUAL to 350*F.....................

3/4 5-3 3/4.5.3 ECCS SUBSYSTEMS - T,,, LESS THAN 350*F........

3/4 5-6 ECCS SUBSYSTEMS - T,y, LESS THAN OR EQUAL TO 200*F.....................

3/4 5-8 3/4.5.4 (This specification number is not used).......

3/4 5-9 3/4.5.5 REFUELING WATER STORAGE TANK............

3/4 5-10 3/4.5.6 RESIDUAL HEAT REMOVAL (RHR) SYSTEM.........

3/4 5-11 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT Containment Integrity................

3/4 6-1 Containment Leakage.................

3 /4 6-2 Containment Air Locks................

3/4 6-5 Internal Pressure..................

3/4 6-7 Air Temperature...................

3/4 6-8 Containment Structural Integrity..........

3/4 6-9 Containment Ventilation System.

3/4 6-12 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS Containment Spray System..............

3/4 6-14 Recirculation Fluid PH Control System........

3/4 6-15 Containment Cooling System.............

3/4 6-17 SOUTH TEXAS - UNITS 1 & 2 viii Unit 1 - Amendment No. 36,61 Unit 2 - Amendment No. N,50

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SECTION PAGE 5.1 SITE 5.1.1 EXCLUSION AREA......................................

5-1 5.1.2 LOW POPULATION ZONE.................................

5-1 5.1.3 MAP DEFINING UNRESTRICTED AREAS AND SITE B0UNDARY FOR RADI0 ACTIVE GASEOUS AND LIQl:ID EFFLUENTS..........

5-1 5.2 CONTAINMENT 5.2.1 CONFIGURATION.......................................

5-1 l

5.2.2 DESIGN PRESSURE AND TEMPERATURE.....................

5-1 FIGURE 5.1-1 EXCLUSION AREA......................................

5-2 FIGURE 5.1-2 LOW POPULATION ZONE.................................

5-3 FIGURE 5.1-3 UNRESTRICTED AREA AND SITE BOUNDARY FOR RADI0 ACTIVE GASE0US EFFLUENTS.................................

5-4 FIGURE 5.1-4 UNRESTRICTED AREA AND SITE B0UNDARY FOR RADI0 ACTIVE LIQUID EFFLUENTS..................................

5-5 5.3 REACTOR CORE 5.3.1 FUEL ASSEMBLIES.....................................

5-6 5.3.2 CONTROL R0D ASSEMBLIES..............................

5-6 5.4 REACTOR COOLANT SYSTEM 5.4.1 DESIGN PRESSURE AND TEMPERATURE.....................

5-6 i

5.4.2 VOLUME..............................................

5-6 j

5.5 METEOROLOGICAL TOWER LOCATION................................

5-6 5.6 FUEL STORAGE 5.6.1 CRITICALITY.........................................

5-6 I

5.6.2 DRAINAGE............................................

5-9 5.6.3 CAPACITY............................................

5-9 FIGURE 5.6-1 MINIMUM BURNUP FOR CATEGORY 2 FUEL..................

5-10 FIGURE 5.6-2 MINIMUM IFBA CONTENT FOR CATEGORY 2 FUEL............

5-11 l

FIGURE 5.6-3 MINIMUM BURNUP FOR CATEGORY 3 FUEL..................

5-12 FIGURE 5.6-4 MINIMUM BURNUP FOR CATEGORY 4 FUEL..................

5-13 FIGURE 5.6-5 REGION 1 CLOSE PACKED AND CHECKERBOARD-FUEL STORAGE......................................

5-14 FIGURE 5.6-6 REGION 2 CLOSE PACKED AND CHECKERBOARD FUEL STORAGE......................................

5-15 FIGURE 5.6-7 MINIMUM IFBA CONTENT FOR IN-CONTAINMENT RACK FUEL STORAGE......................................

5-16 5.7 COMPONENT CYCLIC OR TRANSIENT LIMIT..........................

5-17 l

SOUTH TEXAS - UNITS 1 & 2 xvii Unit 1 - Amendment No. 4,43,46,61 Unit 2 - Amendment No. 3h45,50 1

i INDEX ADMINISTRATIVE CONTROLS SECTION PAGE 6.1 RESPONSIBILITY................................................

6-1 6.2 ORGANIZATION i

6.2.1 0FFSITE AND ONSITE ORGANIZATIONS............................

6-1 6.2.2 UNIT STAFF................................'..................

6-1 TABLE 6.2-1 MINIMUM SHIFT CREW COMPOSITION-TWO UNITS WITH TWO SEPARATE CONTROL R00MS................................

6-4 6.2.3 INDEPENDENT SAFETY ENGINEERING GROUP (ISEG)

Function...................................................

6-6 Composition................................................

6-6 Responsibilities...........................................

6-6 Records....................................................

6-6 6.2.4 SHIFT TECHNICAL ADVIS0R.....................................

6-6 6.3 (Not Used) 6.4 TRAINING......................................................

6-7 6.5 REVIEW AND AUDIT..............................................

6-7 6.5.1 PLANT OPERATIONS REVIEW COMMITTEE (PORC)

Function................

6-7 Composition................................................

6-7 Alternates.................................................

6-7 Meeting Frequency..........................................

6-7 Quorum.....................................................

6-7 Responsibilities.........../...............................

6-8 Records....................................................

6-9 l

i l

SOUTH TEXAS - UNITS 1 & 2 xviii Unit I - Amendment No. 3,4 l

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o Zua__ SAFETY LIMIIS AND LIMITING SAFETY SYSTEM SETTINGS f,.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:

Whenever 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 2.1. 2 The Reactor Coolant System pressure shall not exceed 2735 psig.

APPLICABILITY:

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

ACTION:

MODES 1 and 2:

Whenever the Reactor Coolant System pressure has exceeded 2735 psig, be 1

in HOT STANDBY 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.

i j

MODES 3, 4 and 5:

Whenever the Reactor Coolant System pressure has exceeded 2735 psig, 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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1.2 FRACTION OF RATED THERMAL POWER FIGURE 2.1-1 REACTOR CORE SAFETY UMIT - FOUR LOOPS IN OPERATION l

SOUTH TEXAS - UNITS 1 & 2 2-2 Unit 1 - Amendment No. 4,61 Unit 2 - Amendment No. 50 l

SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS

2. 2 LIMITING SAFETY SYSTEM SETTINGS REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS 2.2.1 The Reac' tor Trip System Instrumentation and Interlock Setpoints shall be set consistent with the Trip Setpoint values shown in Table 2.2-1.

APPLICABILITY:

As shown for each channel in Table 3.3-1.

ACTION:

With a Reactor Trip System Instrumentation or Interlock Setpoint a.

less conservative than the value shown in the Trip Setpoint column but more conservative than the value shown in the Allowable Value column of Table 2.2-1, adjust the Setpoint consistent with the Trip Setpoint value.

b.

With the Reactor Trip System Instrumentation or Interlock Setpoint less conservative than the value shown in the Allowable Value column of Table 2.2-1, either:

1.

Adjust the Setpoint consistent with the Trip Setpoint value of Table 2.2-1 and determine within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months that Equation 2.2-1 was satisfied for the affected channel, or 2.

Declare the channel inoperable and apply the applicable ACTION statement requirement of Specification 3.3.1 until the channel is restored to OPERABLE status with its Setpoint adjusted consistent with the Trip Setpoint value.

Equation 2.2-1 Z + R + 5 < TA Where:

Z = The value from Column Z of Table 2.2-1 for the affected channel, R = The "as measured" value (in percent span) of rack error for the affected channel, S = Either the "as-measured" value (in percent span) of the sensor error, or the value from Column S (Sensor Error) of Table 2.2-1 for the affected channel, and TA = The value from Columh TA (Total Allowance) of Table 2.2-1 for the affected channel.

SOUTH TEXAS - UNITS 1 & 2 2-3

m TABLE 2.2-1 og REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETFOINTS M

TOTAL SENSOR g

ALLOWANCE ERPLR FUNCTIONAL UNIT (TA)

Z (S)

TRIP SETPOINT ALLOWABLE VALUE E

1.

Manual Reactor Trip N.A.

N.A.

U t

2.

Power Range, Neutron Flux i

H a.

High Setpoint 7.5 6.1 0

$109% of RTP**

<110.7% of RTP**

l l

N b.

Low Setpoint 8.3 6.1 0

525% of RTP**

$27.7% of RTP**

l 3.

Power Range, Neutron Flux, 2.1 0.5 0

<5% of RTP** with

<6.7% of RTP** with l

High Positive Rate a time constant a time constant 12 seconds 12 seconds 4.

Deleted 7

5.

Intermediate Range, 16.7 8.4 0

<25% of RTP**

<31.1% of RTP**

l Neutron Flux 6.

Source Range, Neutron Flux 17.0 10.0 0

$105 cps

$1.4 x 10s cps l

7.

Overtemperature AT 10.7 8.7

1. 5 + 1. 5 #

See Note 1 See Note 2 l

8.

Overpower AT 4.7 2.1 1.5 See Note 3 See Note 4 l

gg 9.

Pressurizer Pressure-Low 5.0 2.3 2.0 11870 psig 11860 psig l

E3 10.

Pressurizer Pressure-High 5.0 2.3

2. 0

$2380 psig

$2390 psig l

7 11.

Pressurizer Water Level-High 7.1 4.3 2.0

$92% of instrument

$94.1% of instrument l

span span 12.

Reactor Coolant Flow-Low 4.0 2.1 0.6

>91.8% of loop

>90.5% of loop l

mm ll Besign flow

design flow

t EE

Loop design flow = 95,400 gpm oo

- RTP = RATED THERMAL POWER

1.5% span for AT; 1.5% span for Pressurizer Pressure i

E%

g E-

e TABLE 2.2-1 (Continued) y, E

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS

-4 TOTAL SENSOR E2 ALLOWANCE ERROR 3; FUNCTIONAL UNIT (TA)

Z (S)

TRIP SETPOINT ALLOWABLE VALUE

13. Steam Generator Water 20.0 15.3 2.0 + 0.2## >33% of narrow

> 30.7% of narrow E

Level Low-low range instrument range instrument gj span span

[]

Coolant Pumps

->10,014 volts

-> 9339 volts l

14.

Undervoltage - Reactor 11.9 0.3 0

15.

Underfrequency - Reactor 3.4 0.0 0

>57.2 Hz

>57.1 Hz l

Coolant Pumps 16.

Turbine Trip m

a.

Low Emergency Trip Fluid 232.1 100.8 131.3

>1245.8 psig

>1114.5 psig Pressure u'

b.

Turbine Stop Valve N.A.

N.A.

< Fully closed Fully closed Closure

![![ 17.

Safety Injection Input N.A.

N.A.

c+ c+

from ESFAS m-I t g g-

1. 2.0% span for Steam Generator Level; 0.2% span for Reference Leg RTDs a a, nc

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TABLE 2.2-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS TOTAL SENSOR a

O ALLOWANCE ERROR M FUNCTIONAL UNIT (TA)

Z (S)

TRIP SETPOINT ALLOWABLE VALUE

[

18.

Reactor Trip System 3

Interlocks a

a.

Intermediate Range N.A N.A.

N.A.

31 x 10 2 amp 26 x 10 11 amp

[

Neutron Flux, P-6 b.

Low Power Reactor Trips Block, P-7

1) P-10 input N.A.

N.A.

$10% of RTP**

$11.7% of RTP**

l

2) P-13 input N.A.

N.A.

<10% RTP** Turbine

< 11.7% RTP** Turbine l

Impulse Pressure Impulse Pressure 7

Equivalent Equivalent m

c.

Power Range Neutron N.A.

N.A.

540% of RTP**

$41.7% of RTP**

l Flux, P-8 cc d.

Power Range Neutron N.A.

N.A.

-<50% of RIP **

--< 51.7% of RTP**

l hh Flux, P-9 e.

Power Range Neutron N.A.

N.A.

N.A 110% of RTP**

18.3% of RTP**

l

[

Flux, P-10 aa f.

Turbine Impulse Chamber N.A.

N.A.

<10% RTP** Turbine

< 11.7% RTP** Turbine l

@S Pressure, P-13 Impulse Pressure Impulse Pressure j$

Equivalent Equivalent gg 19.

Reactor Trip Breakers N.A.

N.A.

N.A N. A.

N.A.

{cn 20.

Automatic Trip and Interlock N.A.

N.A.

Logic

- RTP = RATED THERMAL POWER i

t

ng TABLE 2.2-1 (Continued)

I o5 TABLE NOTATIONS

[ NOTE 1:

OVERTEMPERATURE AT (yf 77) $ AT $1 - K (T ( 1 + T

_ p),gy_g)_f g I

1 AT y7 g

2 1+

3 6

Eq Where:

AT Measured AT by RCS Instrumentation;

=

1 7

lead-lag compensator on measured AT;

=

~

l Time constant utilized in lead-lag compensator for AT, ti = 8 sec, l

ti, I2

=

T2 = 3 sec; 1+T3S Lag compensator on measured AT;

=

Time constant utilized in the lag compensator for AT,13 = 0 sec; Is

=

4 AT Indicated AT at RATED THERMAL POWER;

=

g K

=

i 1.14; E5 K2 0.028 / F;

=

l 1 + fr S y.

3 The function generated by the lead-lag compensator for T

=

m-dynamic compensation; avg kk Time constants utilized in the lead-lag compensator for T,yg, 1

=

14, 13 4=

28 sec, l

RR r3 = 4 sec; gg T

Average temperature, F;

=

22 1

??

Lag compensator on measured T,yg;

=

y.

3 83 Time constant utilized in the measured T,yg lag compensator, is = 0 sec;

=

rs

.m.

4 i,

TABLE 2.2-1 (Continued)

TABLE NOTATIONS (Continued) g NOTE 1:

(Continued) 5 T'

5 593.0 F (Nominal T at RATED THERMAL POWER);

m avg 0.00143 /psig; Ka

=

a Pressurizer pressure, psig; m

P

=

2235 psig (Nominal RCS operating pressure);

o-P'

=

~

Laplace transform operator, sec 1; S

=

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

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

m between -70% and + 8 %, f (AI) = 0, where qt and q are pe'rcent RATED THERMAL l

j, (1) For qt 9b 1

b is total THERMAL POWER in the top and bottom halves of the core respectively, and qt

9b POWER in percent of RATED THERMAL POWER; I

(2) For each percent that the magnitude of q

~9 exceeds -70Y,, the AT Trip Setpoint shall t

b be automatically reduced by 0.0% of its value at RATED THERMAL POWER; and cc 11 l

(3) For each percent that the magnitude of qt qb exceeds +8 %, the AT Trip Setpoint shall m-be automatically reduced by 2.65% of its value at RATED THERMAL' POWER.

i 4

$k mm RR NOTE 2:

The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than i

82 1.6% AT span.

[

1 55 r

ho.

)

~

f

5 a

TABLE 2.2-1 (Continued)

TABLE NOTATIONS (Continued)

NOTE 3:

OVERPOWER AT 4 - Ks fy f 3 fyI Ts3)T-Ks(Tfy1

- T") - f (AI))

b O

5 IO o {K

+

1 13 T7 Teg) 2 5

0 Where:

AT As defined in Note 1,

=

e.

f N

As defined in Note 1,

=

ri, I2 As defined in Note 1,

=

yf As defined in Note 1,

=

3 Y

e 13 As defined in Note 1,

=

AT As defined in Note 1,

=

g K4 1.08,

=

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

=

temperature, I7 3

The function generated by the rate-lag compensator for T dynamic

=

compensation, avg Time Constant utilized in the rate-lag compensator for T T7 7

10 sec, avg, T

=

1 As defined in Note 1

=

1+

5 Is As defined in Note 1,

=

m_...___....._..

r TABLE 2.2-1 (Continued)

TABLE NOTATIONS (Continued) w E NOTE 3:

(Continued)

M l

= 0.002 /"F for T > T and Ks = 0 for T < T",

Ks i

C

'As defined in Note 1, 5

T

=

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

=

~

avg instrumentation, < 593.0 F),

r As defined in Note 1, and S

=

0 for all AI.

f (al)

=

2

" NOTE 4:

The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than 1.9% AT span.

l o

CC 1 3.

on 7

Nw I e tD @

3 ma 33 d r+

i 22??

ui m Oa I

\\

I i

2.1 SMETV LIMITS BASES 2.1.1 REACTOR CORE The restrictions of this Safety Limit prevent overheating of the fuel and possible cladding perforation which would result in the release of fission products to the reactor coolant. Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature.

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

DNB is not a directly measurable parameter during operation and therefor:e THERMAL POWER and reactor coolant temperature and pressure have been related to DNB through the WRB-1 correlation.

The WRB-1 DNB correlation has been developed to predict the DNB flux and the location of DNB for axially uniform and nonuniform heat flux distributions. The local DNB heat flux ratio (DNBR) is defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux and is indicative of the margin to DNB.

The DNB design basis is as follows:

uncertainties in the WRB-1 correlation, plant operating parameters, nuclear and thermal parameters, fuel fabrication parameters, and computer codes are considered statistically such that there is at least a 95 percent probability with a 95 percent confidence level that DNBR will not occur on the most limiting fuel rod during Condition I and 11 events.

This establishes a design DNBR value which must be met in plant safety analyses using values of input parameters without uncertainties.

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

The curves of Figure 2.1-1 show the loci of points of THERMAL POWER, Reactor Coolant System pressure and average temperature below which the i

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

These curves are based on a nuclear enthalpy rise hot channel factor, F"

and a reference cosine with a peak of 1.61 for axial power shape.

alto,wanceisincludedforanincreaseinF",,atreducedpowerbasedonthe An expression:

F",, = P, ( 1 + M, ( l M ]

s g

F*DieCOREOPERATINGLIMITSREPORT(COLR);is th where:

in PF C001;and,is the Power Factor Multiplier for F",, specified in the P is the fraction of RTP.

These limiting heat flux conditions are higher than those calculated for the range of all control rods fully withdrawn to the maximum allowable control rod insertion assuming axial imbalance is within the limits of the f (delta

1) function of the Overtemperature trip. When the axial power imbalance is 3

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

SOUTH TEXAS - UNITS 1 & 2 B 2-1 Unit 1 - Amendment No. 61 Unit 2 - Amendment No. 50

c SAFETY LIMITS l

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.

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% (2735 psig) of design pressure.

The Safety Limit of 2735 psig is therefore consistent with the design

)

l i

criteria and associated Code requirements.

The entire RCS is hydrotested at 125% (3110 psig) of design pressure, to demonstrate integrity prior to initial operation.

SOUTH TEXAS - UNITS 1 & 2 B 2-2

g; REQUIRED SHUTDOWN MARGIN FOR MODES 1 AND 2

~

y 1.30% DELTA RHO M

m s

m O

REQUIRED SHUTDOWN MARGIN FOR MODES 3 AND 4

_C s

F.

m 8.0 g

m o

m g 7,o I

~

4 s.0 -

~~

8 ACCEPTABLE REGION (2400 ei s) 3,o.

y z

d --

4,o._

f

$ g,o y

x[. -.

9 O

f 2.0 p

/

g

/(t R Mr Me @tr

]u :E: 4e [eRE u

1.: o) 00, 1.30) m r-

-i o

{, {

o 400 800 1,200 1,600 2,000 2,400

o RCS CRITICAL BORON CONCENTRATION (PPM)

~-

zo for ARI Minus Most Reactive Stuck Rod 8R aa 4M FIGURE 3.1-1 aE g&

REQUIRED SHUTDOWN MARGIN VERSUS RCS CRITICAL BORON CONCENTRATION Sh

e 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 the limit as shown in Figure 3.1-2.

APPLICABILITY:

MODE 5.

ACTION:

With the SHUTDOWN MARGIN less than the limit as shown in Figure 3.1-2, imme-diately 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.

SHRVEILLANCE RE0VIREMENTS 4.1.1.2 The SHUTDOWN MARGIN shall be determined to be greater than or equal to the limit as shown in Figure 3.1-2:

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 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.

l I

SOUTH TEXAS - UNITS 1 & 2 3/4 1-4

v, a

8

-]

REQUIRED SHUTDOWN MARGIN FOR MODE 5 s

=

1

@ 8.0 5

55 d

E 7.0

~

u, xc 6.0

~

ACCEPTABLE REGION ct4oc 4.s o) o 5.0

, 40

/

M M 3.0 9

T' z

2.0 m

y (o, t.3c )

F 2hM* :t. Mil:M:s:1* X 2 y(60 0,1 30) r-

-i>

0 cc 0

400 800 1200 1600 2000 2400 m-o RCS CRITICAL BORON CONCENTRATION (PPM) i e2

&P FIGURE 3.1-2 aa REQUIRED SHUTDOWN MARGIN VERSUS RCS CRITICAL BORON CONCENTRATION PP 8 _*.

REACTIVITY CONTROL SYSTEMS MODERATOR TEMPERATURE COEFFICIENT LLMITING CONDITION FOR OPERATION 3.1.1. 3 The moderator temperature coefficient (MTC) shall be within the limits specified in the Core Operating Limits Report (COLR).

The maximum upper limit shall be less than or equal to that shown in Figure 3.1-2a.

APPLICABILITY:

Beginning of Life (BOL) limit - MODES 1 and 2* only**.

End of Life (E0L) limit - MODES 1, 2, and 3 only**.

ACTION:

With the MTC more positive than the BOL limit specified in the COLR, a.

operation in MODES 1 and 2 may proceed provided:

1.

Control rod withdrawal limits are established and maintained sufficient to restore the MTC to less positive than the BOL limit specified in the COLR within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

These withdrawal limits shall be in addition to the insertion limits of Specification 3.1.3.6; 2.

The control rods are maintained within the withdrawal limits established above until a subsequent calculation verifies that the MTC has been restored to within its limit for the all rods withdrawn condition; and 3.

A Special Report is prepared and submitted to the Commission, pursuant to Specification 6.9.2, within 10 days, describing the value of the measured MTC, the interim control rod withdrawal limits, and the predicted average core burnup necessary for restoring the positive MTC to within its limit for the all rods withdrawn condition, b.

With the MTC more negative than the EOL limit specified in the COLR, be in HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

With Keff greater than or equal to 1.

- See Special Test Exceptions Specification 3.10.3.

SOUTH TEXAS - UNITS 1 & 2 3/4 1-6 Unit 1 - Amendment No. 27 Unit 2 - Amendment No. 17

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

8 ci x

M 7

w 5

8 UNACCEPTABLE OPERATlON

-I d

i 5

~

E 1-

-4 4

O

]

3 ACCEPTABLE OPERATION j

n i

E i

N 2

o m

i w

O 3

2

~

m i

m 0

i m

v i:

(1) h.

(2) cc h5 (3) 0 10 20 30 40 50 60 70 80 90 100 m-

% OF RATED THERMAL POWER EE Figure 3.1-2a 22 MTC versus Power Level

a. o.

22 aa NOTE: Cycle specific MTC limits are displayed in the COLR.

zz FF tt A

i

REACTIVITY CONTROL SYSTEMS MINIMUM TEMPERATURE FOR CRITICALITY LIMITING CONDITION FOR OPERATION 3.1.1.4 The Reactor Coolant System lowest operating loop temperature (Tavg) shall be greater than or equal to 561 F.

APPLICABILITY:

MODES 1 and 2* **.

ACTION:

With a Reactor Coolant System operating loop temperature (T,yg) less than 561 F, restore T,yg to within its limit within 15 minutes or be in HOT STANDBY within the next 15 minutes.

EURVEILLANCE REOUIREMENTS 4.1.1.4 The Reactor Coolant System temperature (Tavg) shall be determined to be greater than or equal to 561 F:

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

b.

At least once' per 30 minutes when the reactor is critical and the Reactor Coolant System T,yg is less than 571 F with the T,yg-T ref Deviation Alarm not reset.

l i

"With Keff greater than or equal to 1.

- See Special Test Exceptions Specification 3.10.3.

SOUTH TEXAS - UNITS 1 & 2 3/4 1-8

0 POWER DISTRIBUTION LIMITS 3/4.2.5 DNB PARAMETERS LIMITING CONDITION FOR OPERATION 3.2.5 The following DNB-related parameters shall be maintained within the limits following:

Reactor Coolant System T,y, s 598'F a.

b.

Pressurizer Pressure, > 2189 psig*

l c.

Reactor Coolant System Flow, 2 392,300 gpm**

l APPLICABILITY: MODE 1.

ACTION:

With any of the above parameters exceeding its limit, restore the parameter to 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 les: 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 RE0VIREMENTS 4.2.# 1 Each of the parameters shown above shall be verified to be within its limit; at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The provisions of Specification 4.0.4 are l

not applicable for verification that RCS flow is within its limit.

4.2.5.2 The RCS flow rate indicators shall be subjected to a channel calibration 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.

The provisions of Specification 4.0.4 are not applicable.

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 2.8% flow measurement uncertainty.

SOUTH TEXAS - UNITS 1 & 2 3/4 2-11 Unit 1 - Amendment No. 61 Unit 2 - Amendment No. 50

d i

TABLE 3.3-4 g

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS

-4g TOTAL SENSOR ERROR 3 FUNCTIONAL UNIT ALLOWANCE (TA)

Z (S)

TRIP SETPOINT ALLOWABLE VALUE

[

1.

Safety Injection (Reactor Trip, 5

Feedwater Isolation, Control d

Room Emergency Ventilation, Start Standby Diesel Generators, Reactor Containment Fan Coolers, and Essential Cooling Water) m a.

Manual Initiation N.A.

N.A.

b.

Automatic Actuation Logic N.A.

N.A.

c.

Actuation Relays N.A.

N.A.

N. A.

w 4

d.

Containment Pressure--High 1 3.6 0.7 2.0 5 3.0 psig 5 4.0 psig l

e.

Pressurizer Pressure--Low 19.6 17.4 2.0 2 1857 psig 1 1851 psig l

f.

Compensated Steam Line 16.4 12.8 2.0 1 735 psig 2 709 psig*

l cc Pressure-Low

3. 3.

((

2.

Containment Spray a.

Manual Initiation N.A.

N.A.

EE b.

Automatic. Actuation Logic N.A.

N.A.

@S l

c.

Actuation Relays N.A.

N.A.

g yf d.

Containment Pressure--High-3 3.6 0.7 2.0

$ 9.5 psig 5 10.5 psig l

$7 e

TABLE 3.3-4 (Continued) m ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS M

TOTAL SENSOR ERROR h FUNCTIONAL UNIT ALLOWANCE (TA)

Z (S)

TRIP SETPOINT ALLOWABLE VALUE

[

3.

Containment Isolation a.

Phase "A" Isolation

[

1) Manual Initiation N.A.

N.A.

e-

2) Autcmatic Actuation Logic N.A.

N.A.

3) Actuation Relays N.A.

N.A.

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

b.

Containment Ventilation Isolation m

s

1) Automatic Actuation N.A.

N.A.

N. A.

N.A.

y logic 5

2) Actuation Relays N.A.

N.A.

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

4) RCB Purge 3.1x10' 1.8x10-1.3x10

<5x10 4

<6.4x10 4 gg Radioactivity-High pCi/cc pCi/cc pCi/cc pCi/cc pCi/cc 23

5) Containment Spray -

See Item 2. above for Containment Spray manual initiation Trip m-Manual Initiation Setpoints and Allowable Values.

6) Phase "A" Isolation -

See Item 3.a. above for Phase "A" Isolation manual initiation

((

Manual Initiation Trip Setpoints and Allowable Values.

((

c.

Phase "B" Isolation

1) Automatic Actuation N.A.

N.A.

Logic 22 PP

2) Actuation Relays N.A.

N.A.

gE

3) Containment Pressure--

3.6 0.7 2.0 5 9.5 psig i 10.5 psig High-3

4) Containment Spray-See Item 2. above for Containment Spray manual initiation Trip Manual Initiation Setpoints and Allowable Values.

\\

~

TABLE 3.3-4 (Continued) m8g ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS r1, TOTAL SENSOR ERROR g FUNCTIONAL UNIT ALLOWANCE (TA)

Z (S)

TRIP SETPOINT ALLOWABLE VALUE d.

RCP Seal Injection Isolation E

1) Automatic Acutation N.A.

N. A.

N.A.

Q Logic and Activation Relays

[

2) Charging Header 4.6 1.0 2.0

> 560.0 psig

> 495.4 psig Pressure - Low m

Coincident with See Item 3.a. above for Phase "A" Isolation Setpoints and Allowable Phase "A" Isolation Values 4.

Steam Line Isolation a.

Manual Initiation N.A.

N.A.

g Y

b.

Automatic Actuation Logic N.A.

N.A.

O and Actuation Relays c.

Steam Line Pressure -

2.6 0.5 0

< 100 psi

-< 126 psi **

l Negative Rate--High EE d.

Containment Pressure -

3.6 0.7 2.0 32 High-2

-< 3.0 psig

-< 4.0 psig l

e.

Compensated Steam Line 16.4 12.8 2.0 2 735 psig 2 709 psig*

Pressure - Low qAm R R 5.

Turbine Trip and Feedwater g8 Isolation AA 22 a.

Automatic Actuation Logic N.A.

N.A.

PP and Actuation Relays b.

Steam Generator Water 10.8 6.5 2.0+0.2#

< 87.5% of

<a9.8% of l

g Level--High-High (P-14)

Harrow range Harrow range instrument instrument span.

span.

c.

Deleted

TABLE 3.3-4 (Continued) oy ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS M

TOTAL SENSOR ERROR y FUNCTIONAL UNIT ALLOWANCE (TA)

Z (S)

TRIP SETPOINT ALLOWABLE VALUE 5.

Turbine Trip and Feedwater E

Isolation (Continued)

~

[

d.

Deleted

[

e.

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

f.

T

-L w Coincident with 4.5 1.1 0.8 2 574 F 2571.7 F l

avg Reactor Trip (P-4)

(Feedwater Isolation Only) 6.

Auxiliary Feedwater a.

Manual Initiation N.A.

N.A.

b.

Automatic Actuation Logic N.A.

N.A.

c.

Actuation Relays N.A.

N.A.

N. A.

N.A.

sF d.

Steam Generator Water 20.0 15.3 2.0+0.2#

2 33.0% of 2 30.7% of l

2 Level--Low-Low narrow range narrow range m-instrument instrument span.

span.

py e.

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

I&

e8 on O.

8.~

e i

m

- ---m

-. m

TABLE 3.3-4 (Continued)

. v, 8y ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS Uh TOTAL SENSOR ERROR

$; FUNCTIONAL UNIT ALLOWANCE (TA)

Z (S)

TRIP SETPOINT ALLOWABLE VALUE h

6.

Auxiliary Feedwater (Continued) ii Of f.

Loss of Power (Motor See Item 8. below for all Loss of Power Trip Driven Pumps Only)

Setpoints and Allowable Values.

e.

7.

Automatic Switchover to m

Containment Sump a.

Automatic Actuation Logic N.A.

N.A.

and Actuation Relays R'

b.

RWST Level--Low-Low 5.0 1.21 2.0

> 11%

-> 9.1%

Coincident With:

Y Safety Injection See Item 1. above for all Safety Injection Trip Setpoints and Allowable b

Values.

8.

Loss of Power a.

4.16 kV ESF Bus Undervoltage N.A.

N.A.

.> 3107 volts

> 2979 volts (Loss of Voltage).

sith a'< 1.75 Gith a < 1.93 second Eime second time i

delay.

b.

4.16 kV ESF Bus Undervoltage N.A.

N.A.

> 3835 volts

> 3786 volts (Tolerable Degraded Voltage Uith a < 35 Gith a < 39 i

Coincident with.SI) second time second time delay.

delay.

c.

4.16 kV ESF Bus Undervoltage N.A.

N.A.

> 3835 volts

> 3786 volts (Sustained Degraded Voltage)

Rith a < 50 Gith a < 55 second Time second time.

delay.

i

TABLE 3.3-4 (Continued) y ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS m

TOTAL SENSOR ERROR y FUNCTIONAL UNIT ALLOWANCE (TA)

Z (S)

TRIP SETPOINT ALLOWABLE VALUE 9.

Engineered Safety Features E

Actuation System Interlocks a.

Pressurizer Pressure, P-11 N.A.

N.A.

< 1985 psig 51995 psig

[

b.

Low-Low T3yg, P-12 N.A.

N.A.

> 563 F

> 560.7 F c.

Reactor Trip, P-4 N.A.

N.A.

N. A.

10.

Control Room Ventilation

{

a.

Manual Initiation N.A.

N.A.

Y b.

Safety Injection See Item 1. above for all Safety Injection Trip E

Setpoints and Allowable Values.

c.

Automatic Actuation Logic N.A.

N.A.

and Actuation Relays d.

Control Room Intake Air 3.7x10 5 2.2x10 5 1.6x10 5

<6.1x10 5

<7.8x10 5 Radioactivity - High pCi/cc pCi/cc pCi/cc pCi/cc pCi/cc e.

Loss of Power See Item 8. above for all Loss of Power Trip Setpoints and Allowable Values.

i i ll 11.

FHB HVAC aa a.

Manual Initiation N.A.

N.A.

~

\\

TABLE 3.3-4 (Continued) m ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS N

TOTAL SENSOR ERROR 3; FUNCTIONAL UNIT ALLOWANCE (TA)

Z (S)

TRIP SETPOINT ALLOW,k val'JE e

11.

FHB HVAC (Continued) 5d b.

Automatic Actuation N.A.

N.A.

Logic and Actuation e.

Relays m

c.

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

d.

Spent Fuel Pool Exhuast-3.1x10 4 1.8x10 4 1.3x10 4

<5.0x10 4'

<6.4x10 4 Radioactivity - High pCi/cc pCi/cc pCi/cc pCi/cc pCi/cc w1 M

[

\\

4 TABLE 3.3-4 (Continued)

TABLE NOTATIONS Time constants utilized in the lead-lag controller for Steam Line Pressure-Low are 7, 2 50 seconds and 7, s 5 seconds.

CHANNEL CALIBRATION shall ensure that these time constants are adjusted to these values.

- The time constant utilized in the rate-lag controller for Steam Line Pressure-Negative Rate-High is greater than or equal to 50 seconds.

CHANNEL CALIBRATION shall ensure that this time constant is adjusted to this value.

2.0% span for Steam Generator Level; 0.2% span for Reference Leg RTDs.

1. Deleted l

1. 1. This setpoint value may be increased up to the equivalent limits of CDCM Control 3.11.2.1 in accordance with the methodology and parameters c f the ODCM during containment purge or vent for pressure control, ALARA and respirable air quality considerations for personnel entry.

SOUTH TEXAS - UNITS 1 & 2 3/4 3-36 Unit 1 - Amendment No. 4,4,47, 61 Unit 2 - Amendment No. 36,50

1 1

3/4.6 CONTAINMENT SYSTEMS I

3/4.6.1 PRIMARY CONTAINMENT i

j CONTAINMENT INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.1.1 Primary CONTAINMENT INTEGRITY shall be maintained.

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

l ACTION:

l Without primary CONTAINMENT INTEGRITY, restore CONTAINMENT INTEGRITY 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 l

SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

4 l

l SURVEILLANCE RE0VIREMENTS

1 4.6.1.1 Primary CONTAINMENT INTEGRITY shall be demonstrated:

l a.

At least once per 31 days by verifying that all penetrations

not capable of being closed by OPERABLE containment automatic isolation valves and required to be closed during accident conditions are closed by valves, blind flanges, or deactivated automatic valves secured in their positions, except as'provided in Specification j

3.6.3; 4'

b.

By verifying that each containment air lock is in compliance with the requirements of Specification 3.6.1.3; and l

l c.

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

test, by leak rate testing the seal with gas at a pressure not less j

than P,, 41.2 psig, and verifying that when the measured leakage l

5 rate for these seals is added to the leakage rates determined i

pursuant to Specification 4.6.1.2d. for all other Type B and C l

penetrations, the combined leakage rate is less than 0.60 L,.

i 4

Except valves, blind flanges, and deactivated automatic valves which are located inside the containment and are locked, sealed or otherwise secured in the closed position. These penetrations shall be verified closed during each COLD SHUTD0WN except that such verification need not be performed more j

often than once per 92 days.

SOUTH TEXAS - UNITS 1 & 2 3/4 6-1 Unit 1 - Amendment No. 61 Unit 2 - Amendment No. 50 i

i

i CONTAINMENT SYSTEMS CONTAINMENT LEAKAGE LIMITING CONDITION FOR OPERATION i

3.6.1.2 Containment leakage rates shall be limited to:

An overall integrated leakage rate of less than or equal to L,2 a.

0.30% by weight of the containment air per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months at P, 41.

psig.

b.

A combined leakage rate of less than 0.60 L for all penetrations andvalvessubjecttoTypeBandCtests,w$enpressurizedtoP,.

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

ACTION:

With either the measured overall integrated containment leakage rate exceeding 0.75 L subjecI, or the measured combined leakage rate for all penetrations and valves to Types B and C tests exceeding 0.60 L, restore the overall integrated leakage rate to less than 0.75 L andthecombinedleakageratefor all penetrations subject to Type B and C te'sts to less than 0.60 L, prior to increasing the Reactor Coolant System temperature above 200*F.

SURVEILLANCE RE0VIREMENTS 4.6.1.2 The containment leakage rates shall be demonstrated at the following test schedule and shall be determined in conformance with the criteria, methods and provisions specified or endorsed in-Appendix J or 10 CFR Part 50:

Three Type A tests (Overall Integrated Containment Leakage Rate) j a.

shall be conducted at 40 i 10 month intervals during shutdown at a pressure not less than P,, 41.2 psig, during each 10-year service l

period. The third test of each set shall be conducted during the shutdown for the 10-year plant inservice inspection; SOUTH TEXAS - UNITS 1 & 2 3/4 6-2 Unit 1 - Amendment No. M,61 Unit 2 - Amendment No. %,50

A CONTAINMENT SYSTEMS SURVEILLANCE RE0VIREMENTS (Continued) b.

If any periodic Type A test fails to meet 0.75 L the test schedule forsubsequentTypeAtestsshallbereviewedan$,approvedbythe Commission.

If two consecutive Type A tests fail to meet 0.75 L,,

a Type A test shall be performed at least every 18 months until two consecutive Type A tests meet 0.75 L, at which time the above test schedule may be resumed; c.

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

1)

Confirms the accuracy of the test by verifying that the supplemental test result, L,, is in accordance with the following equation:

l L,-(L,, +L,) l 50. 25 L, where L., is the measured Type A test leakage and L, is the superimposed leak; 2)

Has a duration sufficient to establish accurately the change in leakage rate between the Type A test and the supplemental test; and 3)

Requires that the rate at which gas is injected into the containment or bled from the containment during the supplemental test is between 0.75 L, and 1.25 L,.

d.

Type B and C tests shall be conducted with gas at a pressure not less than P 41.2 psig, at intervals no greater than 24 months l

exceptforl,estsinvolving:

1)

Air locks, 2)

Purge supply and exhaust isolation valves with resilient material seals, and 3)

Penetrations using continuous Leakage Monitoring Systems.

Air locks shall be tested and demonstrated OPERABLE by the e.

requirements of Specification 4.6.1.3; f.

Purge supply and exhaust isolation valves with resilient material seals shall be tested and demonstrated OPERABLE by the requirements of Specification 4.6.1.7.2 or 4.6.1.7.3, as applicable; SOUTH TEXAS - UNITS 1 & 2 3/4 6-3 Unit 1 - Amendment No. 4,61 Unit 2 - Amendment No. 50

CONTAINMENT SYSTEMS SURVEILLANCE RE0VIREMENTS (Continued) 9 Leakage from isolation valves that are sealed with fluid from a Seal System may be excluded, subject to the provisions of Appendix J,Section III.C.3, when determining the combined leakage rate provided the Seal System and valves are pressurized to at least 1.10 P, and the seal system capacity is adequate to maintain system pressure for at least 30 days; h.

Type B tests for penetrations employing a continuous Leakage Monitoring System shall be conducted at P, 41.2 psig, at intervals l

no greater than once per 3 years; and i.

The provisions of Specification 4.0.2 are not applicable.

SOUTH TEXAS - UNITS 1 & 2 3/4 6-4 Unit 1 - Amendment No. 61 Unit 2 - Amendment No. 50

CONTAINMENT SYSTEMS CONTAINMENT AIR LOCKS LIMITING CONDITION FOR OPERATION 3.6.1.3 Each containment air lock shall be OPERABLE with:

a.

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

l b.

An overall air lock leakage rate of less than or equal to 0.05 L, at j

P,, 41.2 psig.

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

l ACTION:

a.

With one containment air lock door inoperable:

1.

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

2.

Operation may then continue until performance of the next l

i required overt.1 air lock leakage test provided that the

{

i OPERABLE air lock door is verified to be locked closed at least once per 31 days; 3.

Otherwise, 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 ; and 4.

The provisions of Specification 3.0.4 are not applicable.

b.

With the containment air lock inoperable, except as the result of an inoperable air lock door, maintain at least one air lock door closed; restore the inoperable air lock to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least HOT STANDBY within the next 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 .

l l

SOUTH TEXAS - UNITS 1 & 2 3/4 6-5 Unit 1 - Amendment No. 61 Unit 2 - Amendment No. 50 L

CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS 4.6.1.3 Each containment air lock shall be demonstrated OPERABLE:

a.

Within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months following each closing, except when the air lock is being used for multiple entries, then at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , by verifying seal leakage is less than 0.01 L as determined by precision flow measurements when measured for at least 30 seconds with the volume between the seals at a constant pressure not less than P,;

b.

By conducting overall air lock leakage tests at not less than P,,

41.2 psig, and verifying the overall air lock leakage rate is within l

its limit:

1)

At least once per 6 months,* and 2)

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

c.

At least once per 6 months by verifying that only one door in each air lock can be opened at a time.

d.

By verifying at least once per 7 days that the instrument air pressure in the header to the personnel airlock seals is ;t 90 psig.

e.

By verifying the door seal pneumatic system OPERABLE at least once per 18 months by conducting a seal pneumatic system leak test and verifying that system pressure does not decay more than 1.5 psi from 90 psig minimum within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

The provisions of Specification 4.0.2 are not applicable.

- This represents an exemption to Appendix J. paragraph III.D.2 of 10 CFR Part 50.

SOUTH TEXAS - UNITS 1 & 2 3/4 6-6 Unit 1 - Amendment No. 61 Unit 2 - Amendment No. 50

i i

CONTAINMENT SYSTEMS INTERNAL PRESSURE LIMIIING CONDITION FOR OPERATION 3.6.1.4 Primary containment internal pressure shall be maintained between

-0.1 and +0.3 psig.

APPLICABILITY:

MODES 1, 2, 3, and 4.

ACTION:

With the containment internal pressure outside of the limits above, restore the internal pressure to within the limits 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 .

)

SURVEILLANCE REOUIREMENTS 4.6.1.4 The primary containment internal pressure shall be determined to be within the limits at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

i SOUTH TEXAS - UNITS 1 & 2 3/4 6-7

CONTAINMENT SYSTEMS AJR TEMPERATURE LIMITING CONDITION FOR OPERATION 3.6.1.5 Primary containment average air temperature shall not exceed 110'F.

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

ACTION:

With the containment average air temperature greater than 110*F, reduce the l

average air temperature to within the limit within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-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 SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE RE0VIREMENTS 4.6.1.5 The primary containment average air temperature shall be the arithmetical average of a minimum of four RCFC inlet temperatures and shall be determined at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

SOUTH TEXAS - UNITS 1 & 2 3/4 6-8 Unit 1 - Amendment No. 61 Unit 2 - Amendment No. 50

PLANT SYSTEMS SURVEILLANCE REQUIREMENTS 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 pressure of greater than or equal to 1454 psig at.a flow of greater than or equal to 500 gpm; l

2)

Verifying that the steam turbine-driven pump develops a discharge pressure of greater than or equal to 1454 psig at a flow of greater than or equal to 500 gpm when the secondary

[

steam supply pressure is greater than 1000 psig.

The provisions of Specification 4.0.4 are not applicable for entry 1

into MODE 3; 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 4)

Verifying that each automatic valve in the flow path is in the correct position whenever the Auxiliary Feedwater System is 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 valve 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.

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.

SOUTH TEXAS - UNITS 1 & 2 3/4 7-5 Unit 1 - Amendment No. 68, 61 Unit 2 - Amendment No. 50

{

PLANT SYSTEMS AUXILIARY FEEDWATER STORAGE TANK LIMITING CONDITION FOR OPERATION 3.7.1.3 The auxiliary feedwater storage tank (AFST) shall be OPERABLE with a contained water volume of at least 485,000 gallons of water.

l 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 restore the AFST to OPERABLE status or be in at least HOT STANDBY within the asxt 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS 4.7.1.3 The AFST shall be demonstrated OPERABLE at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by verifying the contained water volume is within its limits.

i SOUTH TELAS - UNITS 1 & 2 3/4 7-6 Unit 1 - Amendment No. 33 Unit 2 - Amendment No. 24

3/4.1 REACTIVITY CONTROL SYSTEMS BASES 3/4.1.1 B0 RATION CONTROL l

3/4.1.1.1 and 3/4.1.1.2 SHUTDOWN MARGIN A sufficient SHUTDOWN MARGIN ensures that:

(1) the reactor can be made subcritical from all operating conditions, (2) the reactivity transients associated with postulated accident conditions are controllable within acceptable limits, and (3) the reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition.

SHUTDOWN MARGIN requirements vary throughout core life as a function of i

fuel depletion, RCS baron concentration, and RCS T,t no load opv ating In MODES ) and 2, the n st restrictive condition occurs at EOL, with T,"s', team line break accident and a

taaperature, and is associated with a postulated resulting uncontrolled RCS cooldown.

In the analysis of this accident, a minimum SHUTDOWN MARGIN OF 1.3% Ak/k is required to control the reactivity transient. The 1.3% Ak/k SHUTDOWN MARGIN is the design basis minimum for the 14-foot fuel using silver-indium-cadmium and/or Hafnium control rods (Ref.

FSAR Table 4.3-3).

Accordingly, the SHUTDOWN MARGIN requirement for MODES 1 and 2 is based upon this limiting condition and is consistent with FSAR safety analysis assumptions.

In MODES 3, 4, and 5, the most restrictive condition occurs at BOL, when the boron concentration is the greatest.

In these modes, the required SHUTDOWN MARGIN is composed of a constant requirement and a variable requirement, which is a function of the RCS boron concentration.

The constant SHUTDOWN MARGIN requirement of 1.3% Ak/k is based on an uncontrolled l

RCS cooldown from a steamline break accident.

The variable SHUlDOWN MARGIN requirement is based on the results of a boron dilution accident analysis, where the SHUTDOWN MARGIN is varied as a function of RCS boron concentration, to guarantee a minimum of 15 minutes for operator action after a boron dilution alarm, prior to a loss of all SHUTDOWN MARGIN.

The boron dilution analysis assumed a common RCS volume, and maximum dilution flow rate for MODES 3 and 4, and a different volume and flow rate for MODE 5.

The MODE 5 conditions assumed limited mixing in the RCS and cooling with the RHR system only.

The MODE 5 SHUTDOWN MARGIN curve (Figure 3.1-2) can be used to provide the required C, in the rapid refueling condition (MODE 5 with AR0). The cycle-specific reload safety analysis verifies this curve to be bounding in this condition.

3/4.1.1.3 MODERATOR TEMPERATURE COEFFICIENT The limitations on moderator temperature coefficient (MTC) are provided to ensure that the value of this coefficient remains within the limiting condition assumed in the FSAR accident and transient analyses.

The MTC values of this specification are applicable to a specific set of plant conditions; according1v, verification of MTC values at conditions other than those explicitly stateu' will require extrapolation to those conditions in order to permit an accurate comparison.

SOUTH TEXAS - UNITS 1 & 2 8 3/4 1-1 Unit 1 - Amendment No. p 61 Unit 2 - Amendment No. 2,50

i l

REACTIVITY CONTROL SYSTEMS BASES MODERATOR TEMPERATURE COEFFICIENT (Continued)

The most negative MTC value, equivalent to the most positive moderator density coefficient (MDC), was obtained by incrementally correcting the MDC used in the FSAR analysis to nominal operating conditions.

These corrections involved:

(1) a conversion of the MDC used in the FSAR analysis to its equivalent MTC, based on the rate of change of moderator density with temperature at RATED THERMAL POWER conditions, and (2) subtracting from this value the largest differences in MTC observed at E0L, all rods withdrawn, RATED THERMAL POWER conditions, and those most adverse conditions of moderator temperature and pressure, rod insertion, axial power skewing, and xenon concentration that can occur in nominal operation and lead to a significantly more negative E0L MTC at RATED THERMAL POWER.

These corrections transformed the MDC values used in the FSAR analysis into the limiting E0L MTC value specified in the CORE OPERATING LIMITS REPORT (COLR). The 300 ppm surveillance MTC value specified in the COLR represents a conservative value (with corrections for burnup and soluble boron) at a core condition of 300 ppm equilibrium boron concentration, and is obtained by making these corrections l

to the limiting MTC value.

l The Surveillance Requirements for measurement of the MTC at the beginning and near the end of the fuel cycle are adequate to confirm that the MTC remains within its limits since this coefficient changes slowly due principally to the reduction in RCS boron concentration associated with fuel I

burnup.

i 3/4.1.1.4 MINIMUM TEMPERATURE FOR CRITICALITY

. This specification ensures that the reactor will not be made critical with the Reactor Coolant System average temperature less than 561*F.

This limitation is required to ensure:

(1) the moderator temperature coefficient is within its analyzed temperature range, (2) the trip in:;trumentation is within its normal operating range, (3) the pressurizer is capable of being in an OPERABLE status with a steam bubble, and (4) the reactor vessel is above its minimum RT temperature.

m 3/4.1.2 B0 RATION SYS.:MS The Boron Injection System ensures that negative reactivity control is available during each mode of facility operation.

The components required to perform this function include:

(1) borated water sources, (2) charging pumps, (3) separate flow paths, (4) boric acid transfer pumps, and (5) an emergency power supply from OPERABLE diesel generators.

With the RCS average temperature above 350*F, a minimum of two boron injection flow paths are required to ensure single functional capability in the event an assumed failure renders one of the flow paths inoperable. The boration capability of either flow path is sufficient to provide a SHUTDOWN MARGIN from expected operating conditions of 1.3% Ak/k after xenon decay and l

cooldown to 200*F.

The maximum expected boration capability requires 27,000 gallons of 7000 ppm borated water from the boric acid storage system or 458,000 gallons of 2800 ppm borated water from the refueling water storage tank (RWST).

The RWST volume is an ECCS requirement and is more than adequate for the required boration capability.

SOUTH TEXAS - UNITS 1 & 2 8 3/4 1-2 Unit 1 - Amendment No. N,M,H,64,61 Unit 2 - Amendment No. W,M,40,43, 50

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i INDICATED AXIAL FLUX OlFFERENCE a

i FIGURE B 3/4.2-1 3

TYPICAL INDICATED AXIAL FLUX DIFFERENCE VERSUS THERMA d

1 4

4 i

SOUTH TEXAS - UNITS 1 & 2 B 3/4 2-3 i

a y-t-

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POWER DISTRIBUTION LIMITS BASES HEAT FLUX HOT CHANNEL FACTOR 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 maintained; and d.

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

F"3 will be maintained within its limits provided Conditions a through

d. above,are maintained.

The combination of the RCS flow requirement (392,300 l

gpm) and the requirement on F"3, guarantees that the DNBR used in the safety analysis will be met.

The relaxation of F"3 as a function of THERMAL POWER allows changes in the radial power shape for,all permissible rod insertion limits.

When F"3 is measured, no additional allowances are necessary prior to comparison with the limit. A measurement error of 4% for F"3, has been allowed for in the determination of the design DNBR value.

Fuel rod bowing reduces the value of DNB ratio. Margin has been maintained between the DNBR value used in the safety analyses and the design limit to offset the rod bow penalty and other penalties which may apply.

1 SOUTH TEXAS - UNITS 1 & 2 B 3/4 2-4 Unit 1 - Amendment No. 61 Unit 2 - Amendment No. 50

POWER DISTRIBUTION LIMITS MSIs HEATFLUXHOTCHANNELFACTORandNUCLEARENTHALPYRhSEHOTCHANNEL FACTOR (Continued)

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

and manufacturing tolerance must be made.

An allowance of 5% is appropriate for a full-core map taken with the Incore Detector Flux Mapping System, and a 3% allowance is appropriate for manufacturing tolerance.

The Radial Peaking Factor, Fxy(Z), is measured periodically to provide assurance that the Hot Channel Factor, F (Z), remains within its limit.

The 9

F limit for RATED THERMAL POWER (F RTP) as provided in the Core Operating xy Limits Report (COLR) per Specification 6.9.1.6 was determined from expected power control manuevers over the full range of burnup conditions in the core.

3/4.2.4 QUADRANT POWER TILT RATIO The QUADRANT POWER TILT RATIO limit assures that the radial power distribu-tion satisfies the design values used in the power capability analysis.

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

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

A limit of 1.02 was selected to provide an allowance for the uncertainty associated with the indicated power tilt.

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

In the event such action does not correct the tilt, the margin for uncertainty on F i

q s reinstated by reducing the maximum allowed power by 3% for each percent of tilt in excess of 1.

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

The incort 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.

3/4.2.5 DNB PARAMETERS The limits on the DNB-related parameters assure that each of the parameters are maintained within the normal steady-state envelope of operation assumed in the transient and accident analyses.

The limits are consistent with the SOUTH TEXAS - UNITS 1 & 2 B 3/4 2-5 Unit 1 - Amendment No. 27 Unit 2 - Amendment No.17

POWER DISTRIBUTION LIMITS BASES 3/4.2.5 DNB PARAMETERS (Continued) initial FSAR assumptions and have been analytically demonstrated adequate to maintain a minimum DNBR of greater than or equal to the design limit throughout each analyzed transient. The T value of 598'F and the pressurizer pressure value of 2198 psig ar,e analytical values.

The readings y

from four channels will be averaged and then adjusted to account for measurement uncertainties before comparing with the required limit.

The flow requirement (392,300 gpm) includes a measurement uncertainty of 2.8%.

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

C 1

l SOUTH TEXAS - UNITS 1 & 2 B 3/4 2-6 Unit 1 - Amendment No. 61 Unit 2 - Amendment No. 50

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 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, (41.2 psig).

As an added l

conservatism, the measured overall integrated leakage rate is further limited to less than or equal to 0.75 L, during performance of the periodic 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 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 3.5 psig, and (2) the containment peak pressure does not exceed the design pressure of Fr.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 41.2 psig (P The limit of 0.3 psig for initial positive containment pressure wil)l limit the total pressure to 41.2 psig, which is less than design pressure and is consistent with the safety analyses.

SOUTH TEXAS - UNITS 1 & 2 B 3/4 6-1 Unit 1 - Amendment No. 61 Unit 2 - Amendment No. 50 0

CONTAINMENT SYSTEMS BASES 3/4.6.1.5 AIR TEMPERATURE The limitations on containment average air temperature ensure that the overall 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 by fixed instruments, prior to determining the average air temperature.

3/4.6.1.6 CONTAINMENT STRUCTURAL INTEGRITY 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 integrity is required to ensure that the containment will withstand the maximum pressure of 41.2 psig (P,) in the event of a LOCA l

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

The Surveillance Requirements for demonstrating the containment's structural integrity are in compliance with the recommendations of Regulatory Guide 1.35, " Inservice Inspection of Ungrouted Tendons in Prestressed Concrete Containment Structures," and proposed Regulatory Guide 1.35.1, " Determining Prestressing Forces for Inspection of Prestressed Concrete Containments,"

April 1979.

l 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 corrective 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 i

accident. Maintaining these valves sealed closed during plant operation ensures that excessive quantities of radioactive materials will not be released via the Containment Purge System.

To provide assurance that these containment valves cannot be inadvertently opened, the vaives are sealed closed in accordance with Standard Review 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 - UNITS 1 & 2 B 3/4 6-2 Unit 1 - Amendment No. 61 Unit 2 - Amendment No. 50

3/4.7 PLANT SYSTEMS BASES 3/4.7.1 TURBINE CYCLE 3/4.7.1.1 SAFETY VALVES The OPERABILITY of the main steam line' Code safety valves ensures that the Secondary System pressure will be limited to within 110% (1413.5 psig) of its design pressure of 1285 psig during the most severe anticipated system operational transient.

The maximum relieving capacity is associated with a Turbine trip from 100% RATED THERMAL POWER coincident with an assumed loss of condenser heat sink (i.e., no steam bypass to the condenser).

The specified valve lift settings and relievinc capacities are in accordance with the requirements of Section III of the ASME Boiler and Pressure Code, 1971 Edition.

The total relieving capacity for all valves on all of the steam lines is 20.65 x 106 lbs/h which is 122% of the total second-ary steam flow of 16.94 x 108 lbs/h at 100% RATED THERMAL POWER.

A minimum of two OPERABLE safety valves per steam generator ensures that sufficient reliev-ing capacity is available for the allowable THERMAL POWER restriction in Table 3.7-1.

STARTUP and/or POWER OPERATION is allowable with safety valves inoperable within the limitations of the ACTION requirements on the basis of the reduc-tion in Secondary Coolant System steam flow and THERMAL POWER required by the reduced Reactor trip settings of the Power Range Neutron Flux channels.

The Reactor Trip Setpoint reductions are derived on the following bases:

sp = (X) - (Y)(V) x (109)

Where:

SP = Reduced Reactor Trip Setpoint in percent of RATED THERMAL

POWER, 3

V =

Maximum number of inoperable safety valves per steam line,

)

109 Power Range Neutron Flux-High Trip Setpoint for four loop

=

operation, X

Total relieving capacity of all safety valves per steam

=

line ir, lbs/ hour, and Y

Maximum relieving capacity of any one safety valve in

=

lbs/ hour SOUTH TEXAS - UNITS 1 & 2 B 3/4 7-1

PLANT SYSTEMS BASES 3/4.7.1.2 AUXILIARY FEE 0 WATER 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.

Each auxiliary feedwater pump is capable of delivering a total feedwater flow of 500 gpm at a pressure of 1363 psig to the entrance of the steam l

1 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 35C'F when the Residual Heat Removal System may be placed into operation. The AFW pumps are tested using the test line back to the AFST and the AFW isolation valves closed to prevent injection of cold water into the steam generators.

The STPEGS isolation valves are active valves required to open on an AFW actuation signal.

Specification 4.7.1.2.1 requires these valves to be verified in the correct position.

3/4.7.1.3 AUXILIARY FEEDWATER STORAGE TANK (AFST)

The OPERABILITY of the auxiliary feedwater storage tank with the minimum 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, main feedwater line break and failure of the AFW flow control valve 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 I gpm primary-to-secondary 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 reactivity 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 Requirements are consistent with the assumptions used in the safety analyses.

SOUTH TEXAS - UNITS 1 & 2 8 3/4 7-2 Unit 1 - Amendment No. 33, 61 Unit 2 - Amendment No. 34, 50

o 4

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 Zone shall be as shown in Figure 5.1-2.

MAP DEFINING UNRESTRICTED AREAS AND SITE BOUNDARY FOR RADI0 ACTIVE GASEOUS AND LIOUID EFFLUENTS 5.1.3 Information regarding radioactive gaseous and liquid effluents, which will allow identification of structures and release points as well as definition of UNRESTRICTED AREAS within the SITE BOUNDARY that are acct:sible to MEMBERS OF THE PUBLIC, shall be as shown in Figures 5.1-3 and 5.1-4.

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 includes areas over water bodies.

The concept of UNRESTRICTED AREAS, established at or beyond the SITE B0UNDARY, is utilized in the Limiting Conditions for Operation to keep levels of radioactive materials in liquid and gaseous effluents as low as is reasonable 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.

b.

Nominal inside height - 241.25 feet.

c.

Minimum thickness of concrete walls = 4 feet.

d.

Minimum thickness of concrete roof = 3 feet.

e.

Minimum thickness of concrete floor mat = 18 feet.

f.

Nominal thickness of steel liner = 3/8 inches.

6 g.

Net free volume = 3.38 x 10 - 3.41 x 10' cubic feet.

l 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 structural temperature of l

286*F.

SOUTH TEXAS - UNITS 1 & 2 5-1 Unit 1 - Amendment No. 57,61 Unit 2 - Amendment No. 46,50

tn OM M

xM ESSENTIAL COOLING POND E

EXCLUSION AREA M

v, PRIMARY METEOROLOGICAL TOWER m

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RR SITE DOUNDARY aa 50U111 1EXAS PROJECT llNITS I &2 s*

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22 oo SOUTH TEXAS PROJECT UNITS 1 & 2 sm.s O

FIGURE 5.1-4 i

UNRESTRICTED AREA. AND SITE DOUNDARY FOR RADIOACTIVE LIQUID EFFLUENTS

m..

.--.m.

m- -.

m.

DESIGN FEATURES 5.3 REACTOR CORE FUEL ASSEMBLIES 5.3.1 The core shall contain 193 fuel assemblies with each fuel assembly containing 264 fuel rods clad with Zircaloy-4.

Each fuel rod shall have a nominal active fuel length of 168 inches.

The initial core loading shall have a maximum enrichment of 3.5 weight percent U-235.

Reload fuel shall be similar in physical design to the initial core loading and shall have a maximum enrichment of 5.0 weight percent U-235.

l CONTROL R0D ASSEMBLIES 5.3.2 The core shall contain 57 full-length control rod assemblies. The full-length control rod assemblies shall contain a nominal 158.9 inches of absorber material.

The absorber material within each assembly shall be silver-indium-cadmium or hafnium. Mixtures of hafnium and silver-indium-cadmium are not permitted within a bank. All control rods shall be clad with stainless steel tubing.

5.4 REACTOR COOLANT SYSTEM DESIGN PRESSURE AND TEMPERATURE 5.4.1 The Reactor Coolant System is designed and shall be maintained:

a.

In accordance with the Code requirements specified in Section 5.2 of the FSAR, with allowance for normal degradation pursuant to the applicable Surveillance Requirements.

b.

For a pressure of 2485 psig, and c.

For a temperature of 650*F, except for the pressurizer which is 680*F.

VOLUME 5.4.2 The total water and steam volume of the Reactor Coolant System is 13,814 i 100 cubic feet at a nominal T,y of 561*F.

5.5 METEOROLOGICAL TOWER LOCATION 5.5.1 The meteorological towers shall be located as shown on Figure 5.1-1.

5.6 FUEL STORAGE 5.6.1 CRITICALITY 5.6.1.1 The spent fuel storage racks are designed and shall be maintained with:

SOUTH TEXAS - UNITS 1 & 2 5-6 Unit 1 - Amendment No. 3 4G 16-43,1 6

7 7 Unit 2 - Amendment No, 2 6r33,50 7

DESIGN FEATURES a.

A k,,, equivalent to less than or equal to 0.95 when flooded with unborated water. This requirement shall be met by storing fuel in the spent fuel storage racks according to Specifications 5.6.1.3, 5.6.1.4, and 5.6.1.5.

Additionally, credit may be taken for the 4

presence of soluble boron in the spent fuel pool water, per Specification 3.9.13, to mitigate the misloading of one or more fuel assemblies, as described in Specification 5.6.1.6.

b.

A nominal 10.95 inches center to center distance between fuel assemblies in Region 1 of the storage racks and a nominal 9.15 inches center to center distance between fuel assemblies in Region 2 of the storage racks.

c.

Neutron absorber (Boraflex) installed between spent fuel assemblies in the storage racks in Region 1 and Region 2.

1 5.6.1.2 Prior to insertion into the spent fuel storage racks, each fuel assembly shall be categorized by reactivity, as discussed below, or be designated as a Category 1 fuel assembly. All fuel enrichment values are initial nominal uranium-235 enrichments.

The reactivity categories are:

CATEGORY 1:

Fuel in Category 1 shall have an initial nominal enrichment of less than or equal to 5.0 w/o.

CATEGORY 2:

Fuel in Category 2 shall meet at least one of the following criteria:

1) a maximum initial nominal enrichment of 4.0 w/o; or, t

2) a minimum burnup as shown on Figure 5.6-1; or, 3) a minimum number of Westinghouse Integral Fuel Burnable Absorber pins, as shown on Figure 5.6-2, or a K,l,ased on a unit assembly of less than or equal to l.445. The fuel assembly K shall be configuration (infinite in de lateral and axial extent) in the reactor core geometry, assuming unborated water at 68*F.

The IFBA rod requirements shown in Figure 5.6-2 are based on a nominal IFBA

(

linear B' loading of 1.57 mg-B' / inch.

For higher IFBA linear B'

loadings, the required number of IFBA rods per assembly may'be reduced by the ratio of

~.

the increased B' loading to the nominal 1.57 mg-B / inch loading.

CATEGORY 3:

Fuel in Category 3 shall have the minimum assembly burnup shown on Figure j

5.6-3.

CATEGORY 4:

Fuel in Category 4 shall have the minimum assembly burnup shown on Figure 5.6-4.

Data points for the curves presented in Figures 5.6-1 through 5.6-4 are presented in tables on the respective figures.

Linear interpolation between table values may be used for intermediate points.

SOUTH TEXAS - UNITS 1 & 2 5-7 Unit 1 - Amendment No. 2,43,61 2

Unit 2 - Amendment No. 33,50 4

DESIGN FEATURES 5.6.1.3 Region 1 racks may be used to store Category 1, 2, 3, and 4 fuel.

Category 1 fuel shall be stored in a checkerboard pattern configuration with Category 3 or 4 fuel, alternating fuel assemblies as shown in Figure 5.6-5.

Category 2, 3, and 4 fuel may be stored in a close packed arrangement.

Empty water cells may be substituted for fuel assemblies in all cases.

5.6.1.4 Region 2 racks may be used to store Category 1, 2, 3, and 4 fuel.

Fuel in Categories 1, 2, and 3, shall be stored in a checkerboard pattern configuration alternating fuel assemblies with empty water cells in a 2 out of 4 pattern, as shown in Figure 5.6-6.

Category 4 fuel may be stored either in a close packed arrangement or in the checkerboard pattern described above.

Empty water cells may be substituted for fuel assemblies in all cases.

5.6.1.5 Storage Configuratien Interface Requirements. The transition schemes described below shall be used at the interface of two storage configuration areas in the spent fuel racks.

Empty water cells may be substituted for fuel assemblies in all cases.

Internal Interfaces in Reaion 1 Racks i

The interface between a closed packed fuel storage area in Region 1 and a checkerboarded storage area also in Region 1 shall be such that either:

1.

Category 3 or 4 fuel assemblies in the checkerboard pattern are carried into the first row of the close packed storage area of fuel, as shown in Figure 5.6-5; or, 2.

at least one row of empty water cells separate a close packed fuel storage area and a checkerboarded storage area.

Internal Interfaces in Reaion 2 Racks 1

The interface between a close packed fuel storage area in Region 2 and a checkerboarded storage area in Region 2 shall be such that either:

1.

there is a one row carryover of alternating empty cells from the checkerboard area into the first row of the close packed area with the remaining cells of the row filled with Category 4 assemblies, as shown in Figure 5.6-6; or, 2.

at least one empty row of cells separattu the checkerboard pattern area and the close packed storage area.

Reaion 1 Close Packed Storace Area Ad.iacent to Reaion 2 Close Packed Area There are no restrictions on the interface between Region 1 close packed storage areas and adjacent close packed storage areas in Region 2.

SOUTH TEXAS - UNITS 1 & 2 5-8 Unit 1 - Amendment No. 43 Unit 2 - Amendment No. 32 h

DESIGN FEATURES Reaion 1 Checkerboard Storaae Area Adjacent to Reaion 2 The interface between a checkerboarded storage area in Region 1 and any Region 2 rack storage area shall be such that either:

1.

the Region 1 checkerboard pattern is carried to the Region 1 boundary, but the last row at-the Region 1 boundary leaves the Category 1 fuel assembly positions vacant; or, 2.

at least one row of empty water cells in either Region 1 or Region 2 racks separate the Region 1 checkerboarded storage area and the Region 2 rack storage area.

Reaion 2 Checkboard Storaae Area Adjacent to Reaion 1 The interface between a checkerboarded storage area in Region 2 and any Region I rack storage area shall be such that at least one row of empty water cells in either Region 1 or Region 2 racks separate the Region 2 checkerboarded storage area and the Region I rack storage area.

If checkerboarded storage areas in both Regions 1 and 2 are adjacent, at least one row of empty water cells in either Region 1 or Region 2 racks shall separate the checkerboarded storage areas in the respective racks.

5.6.1.6 The minimum boron concentration specified by Specification 3.9.13,

" Spent Fuel Pool Minimum Boron Concentration" assures that the rack K,n limit in Specification 5.6.1.1.a will not be violated under the following scenarios:

1.

in Region 1, any misloading of Category 1, 2, 3, and 4 assemblies; or, 2.

in Region 2, the misloading of one Category 1 assembly into the center of a fully loaded checkerboard area also containing Category 1 assemblies; or, 3.

the misloading of a Category 1 assembly in a Region 1 rack adjacent to a Category 1 assembly in a Region 2 rack.

5.6.1.7 The new fuel storage racks are designed and shall be maintained with:

a.

A K,g equivalent to less than or equal to 0.95 when flooded with unborated water and less than or equal to 0.98 when filled with aqueous foam moderation (low density water).

This requirement shall be met by limiting the maximum fuel assembly nominal enrichments to 5.0 w/o or less.

b.

A nominal 21 inches center to center distance between fuel assemblies, l

SOUTH TEXAS - UNITS 1 & 2 5-9 Unit 1 - Amendment No. 43,61 Unit 2 - Amendment No. 33,50

DESIGN FEATURES 5.6.1.8 The In-containment fuel storage racks are designed and shall be maintained with:

A K,,,ted water. equivalent to less than or equal to 0.95 when flooded with a.

unbora This requirement shall be met by satisfying at least one of the following criteria:

1) a maximum initial fuel assembly nominal enrichment to 4.5 w/o or less; 2) a minimum number of Westinghouse's Integral Fuel Burnable Absorbers, as a function of initial nominal assembly enrichment,asshownonFigure5.6-7,oraK,Tebasedona of less than or equal to 1.484.

The fuel assembly X shall unit assembly configuration (infinit$,in the lateral and axial extent) in the reactor core geometry, assuming unborated water at 68'F.

The IFBA rod requirements shown in Figure 5.6-7 are based on a l

nominal IFBA {inear B " loading of 1.57 mg-B' / inch.

For higher IFBA linear B loadings, the required number of IFBA rods per assembly may be reduced by the ratio of the increased B' loading to the nominal 1.57 mg-B / inch loading; or, 3) the fuel assembly is categorized as a Category 2, 3, or 4 assembly, per Specification 5.6.1.2.

b.

A nominal 16 inches center to center distance between fuel assemblies.

DRAINAGE 5.6.2 The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 62 feet-6 inches.

CAPACITY 5.6.3 The spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than 1969 fuel assemblies.

SOUTH TEXAS - UNITS 1 & 2 5-9a Unit 1 - Amendment No. 61 Unit 2 - Amendment No. 50

1 l

Minimum Burnup for Category 2 Fuel 6,000 i

i

..w.

._m.

- -.=-.

initial Enrichment Minimum Bumup (w/o)

(MWD /MTU)

~D 5,000 4,o o

F--2 4.5 2,700

~

C 5.0 5,400 3 4,000 s

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@ 3,000 B

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- Acceptable UNACCEPTABLE

- 2,000 g

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1,000

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es e

ar e M.

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6.g y.4, 0

4.0 4.2 4.4 4.6 4.8 5.0 Initial Enrichment (w/o)

Assemblies with an initial enrichment less than 4.0 w/o are ACCEPTABLE Figure 5.6-1 SOUTH TEXAS - UNITS 1 & 2 5-10 Unit 1 - Amendment No. 43 Unit 2 - Amendment No. 32

1 1

l l

i l

i i

Region 2 Close Packed and Checkerboard Fuel Storage i

i close-packed storage '

A I

1 l

l fY h f Yb h h f/

l W M ff/A" 94M WA 1

VA

'f V

My(Ai 3

l l

m

/

l l

l

)

Y checkerboard storage Y

h f

Category 4 Fuel Assembly Category 1,2, or 3 Fuel Assembly i

Water Cell 1

f Figure 5.6-6 i

SOUTH TEXAS - UNITS 1 & 2 5-15 Unit 1 - Amendment No. 43 i

Unit 2 - Amendment No. 32 i

._,_.-._.....,___._.._m.

Minimum IFBA Content for in-Containment Rack Fuel Storage 40

" 7 "7FBA

[

m

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0 4.5 4.6 4.7 4.8 4.9 5.0 Initial Enrichment (w/o)

Assemblies with an initial enrichment less than 4.5 w/o are ACCEPTABLE Figure 5.6-7 SOUTH TEXAS - UNITS 1 & 2 5-16 Unit 1 - Amendment No. 61 Unit 2 - Amendment No. 50

DESIGN FEATURES 5.7 COMPONENT CYCLIC OR TRANSIENT LIMIT 5.7.1 The components of the reactor coolant system are designed and shall be maintained within limits addressed in the Component Cyclic and Transient Limit Program as required by specification 6.8.3f.

)

SOUTH TEXAS - UNITS 1 & 2 5-17 Unit 1 - Amendment No. 4h46,61 Unit 2 - Amendment No. 3 h35,50

... -

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