Text

Amends 27 & 17 to Licenses NPF-76 & NPF-80,respectively, Revising TS by Relocating Several cycle-specific Core Operating Limits from TS to Core Operating Limits Repts
	ML20082V255
Person / Time
Site:	South Texas
Issue date:	09/09/1991
From:	Black S; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20082V257	List: ML20082V255; ML20082V258;
References
	NUDOCS 9109230182
	Download: ML20082V255 (41)
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zy T a7 /1 UNITED STATES t -WA i i NUCLEAR REGULATORY COMMISSION

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f WASHINoToN, D.c,20066 g-. _ v IHOUSTON LIGHTING & POWER C;0MPANY CITY PUBLIC SERVICE-BOARD OF SAN ANTONIO

~ CENTRAL POWER AND LIGHT COMPANY 5

i CITY OF AUSTIN, TEXAS DOCKET NO. 50-498-SOUTH TEXAS PROJECT, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 27 License No. NPF-76

, 1; The Nuclear Regulatory Commission _;(the Commission) has found that:

A.

The applications forLamendment 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 September 15, 1989 and January 8..1991, as amended on May 23, 1991, comply.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;-

2 B.'- ; The. facility will operate in: conformity with the application, as amendeo, the provisions of the Act,-and the-rules and regulations of the. Commission; C.

There is reasonable assurance:

(i)'thatthe: 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:complianca-wit _h 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-satisfied.

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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.

9109230182 910909'-

-PDR. ADOCK 05000498 P

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

2.-

Technical Specifications-

)

The Technical Specifications contained in Appendix A, as revised.-

1through Amendment No. 27, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license.

i The licensee shall operate the facility in accordance with the 1

Technical: Specifications and the Environmental Protection Plan.

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3.

The licenseLamendment is effectivo as of its date of issuance.

J FOR THE NUCLEAR REGULATORY COMMISSION

/(

Suzann

. Black, Director.

Project Directorate IV-2 Division of. Reactor Projects - III/IV/V-Office of Nuclear Reactor Regulation'

Attachment:

Changes to the Technical Specifications 1

Date;of Issuance:. September 9, 1991 1

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$*g> *a c q*'o 3 - ( D,/,v 31 UNITED STATES e

i NUCLEAR REGULATORY COMMISSION

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f WASHINGTON. o c. 20r46 cy HOUSTON LIGHTING & POWER COMPANY CITY PUBLIC SERVICE BOARD OF SAN ANTONIO CENTRAL POWER AND LIGHT COMPANY CITY OF AUSTIN, TEXA5 DOCKET NO. 50-499 SOUTH TEXA5 PROJECT, UNIT 2 AMEN 0 MENT TO FACILITY OPERATING LICENSE Amendment No. 17 License No. NPF-80 1.

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

A.

The applications 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 (COA) (the licensees) dated September 15, 1989 and January 8, 1991, as amended on May 23, 1991, comply 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:

(i) that the activities authorized by thi, 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 e' this license amendment will not be inimical +o the common defens and security or to the health and safety of sne public; and E.

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

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.

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2.

Accordingly, the license is' amended _by changes to the Technical Specifi-cations 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.- 17, 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 6

Technical Specifications and the Environmental-Protection Plan.

3.-

The license amendment is~ effective'as of'its date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION f

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.x Suzanne ~C.

Black, Director Project Directorate IV-2

- Division of Reactor Projects - III/IV/V

- Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications-

- Date-of Issuance:

. September 9, 1991 5

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ATTACHMENT TO LICENSE AMENDMENT NOS. 27 AND 17 FACILITY OPERATING LICENSE NGS. 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 vertical lines indicating the areas of change.

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

REMOVE INSERT iv iv v

v xix xix 3/4 1-6 3/4 1-6 3/4 1-7 3/4 1-7 3/4 1-7a 3/4 1-7a 3/4 1-16 3/4 1-16 3/4 1-17 3/4 1-17 3/4 1-22 3/4 1-22 3/4 2-1 3/4 2-1 3/4 2-2 3/4 2-2 3/4 2-4 3/4 2-4 3/4 2-5 3/4 2-5 3/4 2-6 3/4 2-6 3/4 2-7 3/4 2-7 3/4 2-8 3/4 2-8 L

3/4 2-9 3/4 2 9 8 3/4 1 B 3/4 1-2 8 3/4 2-1 B 3/4 2-1 B 3/4 2-2 8 3/4 2-2 8 3/4 2-5 B 3/4 2-5 6 6-20 l 20a l

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INDEX 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS SECTION PAGE 2.1 SAFETY LIMITS 2.1.1 REACTORC0RE..................................................

2-1 2.1.2 REACTOR COOLANT SYSTEM PRESSURE...............................

2-1 FIGURE 2.1-1 REACTOR CORE SAFETY LIMIT - FOUR LOOPS IN OPERATION....

2-2

2. 2 LIMITING SAFETY SYSTEM SETTINGS 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS................

2-3 TABLE 2.2-1 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS......

2-4 BASES SECTION PAGE 2.1 SAFETY LIMITS 2.1.1 REACTOR C0RE.................................................

B 2-1 2.1.2 REACTOR COOLANT SYSTEM PRESSURE..............................

B 2-2 2.2 LIMITING SAFETY SYSTEM SETTINGS 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETP0lNTS................

B 2-3 I

1 SOUTH TEXAS - UNITS 1 & 2 iii

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.0 APPLICABILITY................................................

3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 B0 RATION CONTROL Shutdown Margin - T,yg Greater Than 200 F.................

3/4 1-1 FIGURE 3.1-1 REQUIRED SHUTDOWN MARGIN VERSUS RCS CRITICAL BORON C01 CENTRATION (MODES 1, 2, 3, AND 4)......................

3/4 1-3 Stutdown Margin - T,yg less Than or Equal to 200 F........

3/4 1-4 FIGURE 3.1 2 REQUIRED SHUTDOWN MARGIN VERSUS RCS CRITICAL BORON CONCENTRATION (MODE 5)....................................

3/4 1-5 Moderator Temperature Coefficient.........................

3/4 1-6 FIGURE-3.1-2a BOL MODERATOR TEMPERATURE COEFFICIENT VERSUS POWER....3/4 1-7a Minimum Temperature'for Criticality.......................

3/4 1-8 3/4.1.2 BORATION SYSTEMS Flow Paths - Shutdown.....................................

3/4 1-9 Flow Paths - Operating....................................

3/4 1-10 Charging Pumps - Shutdown.................................

3/4 1-11 Charging Pumps

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

3/4 1-12 Borated Water Sources - Shutdown..........................

3/4 1-13 Borated Water Sources - 0oerating.........................

3/4 1-14 3/4,1.3 MOVABLE CONTROL ASSEMBLIES Group Height..............................................

3/4 1-16 TABLE 3.1-1 ACCIDENT ANALYSES REQUIRING REEVALUATION IN THE EVENT OF AN INOPERABLE FULL-LENGTH R00....................

3/4 1-18 Position Indication Systems - Operating...................

3/4 1-19 l

Position Indication Systems - Shutdown....................

3/4 1-20 Rod Drop Time.............................................

3/4 1-21 Shutdown Rod Insertion Limit..............................

3/4 1-22 Control Rod Insertion Limits.......................

3/4 1-23 FIGURI 3.1-3 (Deleted)

SOUTH TEXAS - UNITS 1 & 2 iv Unit 1 - Amendment No. 9, 27 Unit 2 - Amendment No. I, 17

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE...............................

3/4 2-1 FIGURE 3.2-1 AXIAL FLUX DIFFERENCE LIMITS AS A FUNCTION OF RATED THERMAL P0WER......................................

Deleted l

3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR - F (Z)...................

3/4 2-5 q

FIGURE 3.2-2 K(Z) - NORMALIZED F (Z) AS A FUNCTION OF CORE HEIGHT.

Deleted l

q 3/4.2.3 NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR.................

3/4 2-9 3/4.2.4 QUADRANT POWER TILT RATI0..............................

3/4 2-10 3/4.2.5 DNB PARAMETERS.......................

3/4 2-11 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION......................

3/4 3-1 TABLE 3.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION..................

3/4 3-2 TABLE 3.3-2 REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES....

3/4 3-9 TABLE 4.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMEhis...........................................

3/4 3-11 3/4.3.2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION..........................................

3/4 3-16 TABLE 3.3-3 ENGINEERED SAFETY FEATURc3 ACTUATION SYSTEM INSTRUMENTATION..........................................

3/4 3-18 TABLE 3.3-4 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETP0lNTS...........................

3/4 3-29 TABLE 3.3-5 ENGINEERED SAFETY FEATURES RESPONSE TIMES.............

3/4 3-37 TABLE 4.3-2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS................

3/4 3-42 3/4.3.3 MONITORING INSTRUMENTAT'2N Radiation Monitoring for Plant Operations................

3/4 3-50 TABLE 3.3-6 RADIATION MONITORING INSTRUMFNTATION FOR PLANT 0PERATIONS.....................................

3/4 3-51 TABLE 4.3-3 RADIATION MONITORING INSTRUMENTATION FOR PLANT OPERATIONS SURVEILLANCE REQUIREMENTS.....................

3/4 3-53 Movable Incore Detectors.................................

3/4 3-54 Seismic Instrumentation..........................

3/4 3-55 TABLE 3.3-7 SEISMIC MONITORING INSTRUMENTATION.................

3/4 3-56 SOUTH TEXAS - UNITS 1 & 2 v

Unit 1 - Amendment No. 27 Unit 2 - Amendment No. 17

s INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE TABLE 4.3-4 SEISMIC MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS.............................................

3/4 3-57 Meteorological Instrumentation...........................

3/4 3-58 TABLE 3.3-8 METEOROLOGICAL MONITORING INSTRUMENTATION.............

3/4 3-59 TABLE 4.3-5 METEOROLOGICAL MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS............................................

3/4 3-60 Remote Shutdown System..................................

3/4 3-61 TABLE 3.3-9 REMOTE SHUTDOWN SYSTEM...............................

3/4 3-62 TABLE 4.3-6 REMOTE SHUTDOWN MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS................

3/4 3-66 Accident Monitoring Instrumentation......................

3/4 3-67 TABLE 3.3-10 ACCIDENT MONITORING INSTRUMENTATION 3/4 3-68 TABLE 4.3-7 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS............................................

3/4 3-73 Chemical Detection Systems.............

3/4 3-75 TABLE 3.3-11 (This table number is not used.).....................

3/4 3-77 Radioactive Liquid Effluent Monitoring Instrumentation...

3/4 3-79 TABLE 3.3-12 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION 3/4 3-80 TABLE 4.3-8 _ RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS...............

3/4 3-82 Radioactive Gaseous Effluent Monitoring Instrumentation..

3/4 3-84 TABLE 3.3-13 RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION........................

3/4 3-85 TABLE 4.3-9 RADI0 ACTIVE GASE0US EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS................

3/4 3-87 3/4.3.4 TURBINE OVERSPEED PROTECTION..............................

3/4 3-89 SOUTH TEXA5 - UNITS 1 & 2 vi

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INDEX ADMINISTRATIVE CONTROLS SECTION PAGE 6.5.2 NUCLEAR SAFETY REVIEW BOARD (NSRB)

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

6-9 Composition................................................

6-10 Alternates.................................................

6-10 Consultants................................................

6-10 Meeting Frequency................................

6-10 Quorum.....................................................

6-10 Review.....................................................

6-10 Audits.........................................

6-11 Records....................................................

6-12 6.5.3 TECHNICAL REVIEW AND CONTROL Activities...............................................

6-12 6.6 REPORTABLE EVENT ACTI0N......................................

6-13 6.7 SAFETY LIMIT VIOLATION.......................................

6-13 6.8 PROCEDURES AND PR0 GRAMS,.....................................

6-14 6.9 REPORTING REQUIREMENTS 6.9.1' ROUTINE REP 0RTS............................................

6-16 Startup Report.............................................

6-16 Annual Reports.............................................

6-17 Annual Radiological Environmental Operating Report.........

6-17 Semiannual Radioactive Effluent Release Report.............

6-18 Monthly Operating Reports..................................

6-20

. Core Operating Limits Report...............................

6-20

- C.9.2 SPECIAL REP 0RTS............................................

6-20a l

6.10 RECORD RETENTION............................................

6-21

- SOUTH TEXAS - UNITS 1 & 2 xix Unit 1 - Amendment No. 9, 27 Unit 2 - Amendment No. I, 17

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INDEX ADMINISTRATIVE CONTROLS SECTION PAGE 6.11 RADIATION PROTECTION PR0 GRAM................................

6-22 6.12 HIGH RADIATION AREA...........................

6-22 6.13 PROCESS CONTROL PROGRAM (PCP)...............................

6-23 6.14 0FFSITE DOSE CALCULATION MANUAL (00CM)......................

6-23 6.15 MAJOR CHANGES TO LIQUID, GASEOUS, AND SOLIO RADWASTE TREATMENT SYSTEMS..................................

6-24 i

SOUTH TEXAS - UNITS 1 & 2 xx

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RCS CRITICAL BORON CONCENTRATION (PPM)

FIGURE 3.1-2 REQUIRED SHUTDOWN MARGIN VERSUS RCS CRITICAL BORON CONCENTRATION (MODE 5)

REACTIVITY CONTROL SYSTEMS MODERATOR TEMPERATURE COEFFICIEtjT LIMlllM30ElI10N101LQPlM110N 3.1.1.3 The moderator temperature coefficient (MTC) shall be within 1."Te 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:

a.

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

REACTIVITY CONTROL SYSTEMS SUME1LLANCEEQQ1REMENTS 4.1.1.3 The MTC shall be determined to be within its limits during each fuel cycle as follows:

a.

The MTC shall be measured and compared to the BOL limit specified in the COLR prior to initial operation above 5% of RATED THERMAL POWER, after each fuel loading; and b.

The MTC shall be measured at any THERMAL POWER and compared to the 300 ppm surveillance limit specified in the COLR (all rods withdrawn, RATED THERMAL POWER condition) within 7 EFPD after reaching an equi-librium boron concentration of 300 ppm.

In the event this comparison indicates the MTC is more negative than the 300 ppm surveillance limit specified in the COLR, the MlC shall be remeasured, and compared to the EOL MTC limit specified in the COLR, at least once per 14 EFPD during the remainder of the fuel cycle.

SOUTH TEXAS - UNITS 1 & 2 3/4 1-7 Unit 1 - Amendment No. 27 Unit 2 - Amendment No. 17

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REACTIVITY CONTROL SYSTEMS MINIMUM TEMPERATURE FOR CRITICALITY MjilM CONQ1 HON FOR DEEBal10N 3.1.1.4 The Reactor Coolant System lowest operating loop temperature (T,y9) shall be greater than or equal to 561 F.

APPLICABILITY:

MODES 1 and 2* **.

ACTION:

With a Reactor Coolant System operating loop temperature (1,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.

LURYEILLMCLR(QUlREBEtil; 4.1.1.4 The Reactor Coolant System temperature (1 avg) shall be determined to be greater than or equal to 561 F:

a.

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

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

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

REACTIVITYCONTROLSYSTEM LUMEILLA!iCLSEQutREMENTS 4.1.2.6 Each borated water source shall be demonstrated OPERABLE at least once per 7 days by:

Verifying the boron concentration in the water, a.

b.

Verifying tne contained borated water volume of the water source, and c.

Verifying the Boric Acid Storage System solution temperature when it is the source of borated water.

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SOUTH TEXAS - UNITS 1 & 2 3/4 1-15

REACTIVITY CONTROL SYSTEMS 3/4.1.3 MOVABLE CONTROL ASSEMBLIES GROUP HEIGHT-LIMITING CONDITION FOR OPERATION 3.1.3.1 All full-length shutdown and control rods shall be OPERABLE and positioned within 1 12 steps (indicated position) of their group step counter demand position.

APPLICABILIfY:

MODES 1* and 2*.

ACTION:

With one or more full-length rods inoperable due to being immovable a.

as a result of excessive friction or mechanical interference or known to be untrippable, determine that the SHUTDOWN MARGIN require-ment of Specification 3.1.1.1 is satisfied within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b.

With one full-length rod trippable but inoperable due to causes other than addressed by ACTION a., above, or misaligned from its group step counter demand height by more than i 12 steps (indicated position), POWER OPERATION may continue provided that within I hour:

1.

The rod is restored to OPERABLE status within the above alignment requirements, or 2.

The rod is declared inoperable and the remainder of the rods in the group with the inoperable rod are aligned to within i 12 steps of the inoperable rod while maintaining the rod sequence and insertion limits as specified in the Core Operating Limits Report (COLR).

The THERMAL POWER level shall be restricted pursuant to Specification 3.1.3.6 during subsequent operation, or 3.

The. rod is declared inoperable and the SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is satisfied.

POWER OPERATION may then continue provided that:

a)

-A reevaluation of each accident analysis of Table 3.1-1 is performed within 5 days; this reevaluation shall con-firm that the previously analyzed results of these acci-dents remain valid for the duration of operation under these conditions; b)

The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is determined at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ; j

See Special Test Exceptions Specifications 3.10.2 and 3.10.3.

SOUTH TEXAS - UNITS 1 & 2 3/4 1-16 Unit 1 - Amer.dment No. 27 Unit 2 - Amendment No.17

l REACTIVITY CONTROL SYSTEMS MHil111G_C0tiDU10L10MffRAl10h ACT10'l(Continued) c)

A power d' nribution map is obtained from the movable incore detectors and F (Z) and F are verified to be 9

g within their limits within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; and d)

The THERMAL POWER level is reduced to less than or equal to 75% of RATED THERMAL POWER within the next hour and within the following 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months the High Neutron Flux Trip setpoint is reduced to less than or equal to 85%

of RATED ":iERMAL POWER.

c.

With more than one rod trippable but inoperable due to causes other than addressed by ACTION a. above, POWER OPERATION may continue provided that:

1.

Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , he remainder of the rods in the bank (s) with the inoperable rods are aligned to within ! 12 steps of the inoperable rods while maintaining the rod sequence and insertion limits as specified in the COLR.

The THERMAL POWER level shall l

be restricted pursuant to Specification 3.1.3.6 during subsequent operation and s

2.

The inoperable rods are restored to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

d.

With more than one rod misaligned from its group step counter demand height by more than i 12 steps (indicated position), be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REOUIREMENTS 4.1.3.1.1 The position of each full-length rod shall be determined to be within the group demand limit by verifying the individual rod positions at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months except during time intervals when the rod position deviation monitor is inoperable, then verify the group positions at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

4.1.3.1.2 Each full-length rod not fully inserted in the core shall be determined to be OPERABLE by movement of at least 10 steps in eny one direction at least once per 31 days.

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

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TABLE 3.1-1 ACCIDENT ANALYSES REQUIRING REEVALUATION IN THE EVENT OF AN INOPERABLE FULL-LENGTH ROD Rod Cluster Control Assembly Insertion Characteristics Rod Cluster ontrol Assembly Misalignment

+

Loss of Reactor Coolant from Small Ruptured Pipes or from Cracks in Large Pipes Which Actuates the Emergency Core Cooling System Single Rod Cluster Control Assembly Withdrawal at Full Power Major Reactor Coolant System Pipe Ruptures (Loss-of-Coolant Acciderit)

Major Seco,aary Coolant System Pipe Rupture Rupture of a C'2"rol Rod Drive Mechanism Housing (Rod Cluster Control Assembly Ejection) l SOUTH TEXA5 - UNITS 1 & 2 3/4 1-18

t REACTIVITY CONTROL SYSTEMS ROD DROP 'llME LIMITING @lQll Q S 3 R OPERATION 3.1.3.4 The individual full-length (shutdown and control) rod drop time from the fully withdrawn position shall t,e less than or equal to 2.8 seconds from beginning of decay of stationary gripper coil voltage to dashpot entry with:

T,yg greater than or equal to 561'F, and a.

b.

All reactor coolant pumps operating.

l APPLICABILITY:

MODES I and 2.

ACTION:

I

'With the drop time of any full-length rod determined to exceed the above limit, restore the rod drop time to within the above limit prior to proceeding to MODE 1 or 2.

kBXDLLMLCLELQYlhW1M)

... _ n _ _ _.

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

For all rods following each removal of the reactor essel head, a.

I b.

for specifically affected individual rods following any maintenance on or modification to the Control Rod Drive System which could affect the drop time of those specific rods, and c.

At least once per 18 months, l

l l

SOUTH TEXA5 - UNITS 1 & 2 3/4 1-21

i REACTIVITY CONTROL SYSTEMS SHUTDOWN ROD INSERTION LIMIT LIMITING CONDITION FOR OPERATION

-3.1.3.5 All shutdown rods shall be fully withdrawn, as specified in the Core Operating Limits Report (COLR).

-APPLICABILITY:

MODES 1* und 2* **,

ACTION:

With a maximum of one shutdown rod _not fully withdrawn, except for surveillance testing pu,suant to Specification 4.1.3.12, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months either:

a.

Fully withdraw the rod, or b.

Declare the rod to be inoperable and app *y Specification 3.1.3.1.

$RYEllLAEE RE001REMENTS 4.1.3.5 Each shutdown rod shall be determined te be fully withdrawn:

a.

Within 15 minutes prior to withdrawal of any rods in Control Bank A, B, C, or D during an approach tu reactor criticality, and b.

At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter.

See Special Test Exceptions Specifications-3.10.2 and 3.10.'.

- With K,ff greater than or equal to 1.

SOUTH TEXAS - UNITS 1 & 2 3/4 1-22 Unit 1 - Amendment No. 27 Unit 2 - Amendment No. 17

3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE kitilDMdQMll[0N FOR_0PEMTION 3.2.1 The indicated AXIAL FLUX DIFFERENCE (AFD) shall be maintained within the target band (flux difference units) about the target flux difference as specified in the CORE OPERATING LIMITS REPORT (COLR).

l The indicated AFD may deviate outside the above required target band at greater than or equal to 50% but less than 90% of RATED THERMAL POWER provided the indi-cated AFD is within the Acceptable Operation Limits specified in the COLR and l

the cumulative penalty deviation time does not exceed I hour during the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

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

APPLICABILITY:

MODE 1, above 15% of RATED THERMAL POWER.*

ACTION:

a.

With the indicated AfD outside of the required target band and l

with THERMAL POWER greater than or equal to 90% of RATED THERMAL POWER, within 15 minutes either:

1.

Restore the indicated AFD to within the target band limits, or 2.

Reduce THERMAL POWER to less than 90% of RATED THERMAL POWER, b.

With the indicated AFD outside of the above required target band for more than I hour of cumulative penalty deviation time during the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or outside the Acceptable Operation Limits specified in the COLR and with THERMAL POWER less than 90% but equal to or greater than 50% of RATED THERMAL POWER, reduce:

1.

THERMAL POWER-to less than 50% of RATED THERMAL POWER within 30 minutes, and 2.

The Power Range Neutron Flux * ** - High Setpoint to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

'SeeSpIclalTestExceptionsSpecification3.10.2.

- Surveillance testing of the Powe Range Neutron Flux Channel may be performed pursuant to Specification 4.3.1.1 provided the indicated AFD is maintained within the Acceptable Operation Limits specified in the COLR.

A total of 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months

(

operation may be accumulated with the AFD outside of the above required target band during testing without penalty deviation.

SOUTH TEXAS - UNITS 1 & 2 3/4 2-1 Unit 1 - Amendment No. 9, 27 Unit 2 - Amendment No. I, 17

i POWER DISTRIBUTION LIMITS UMITINGCONDITIONFOROPERATION ACTION (Continued) c.

With the indicated AFD outside of the required target band for more than I hour of cumulative penalty deviation time during the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and with THERMAL POWER less than 50% but greater than 15%

of RATED THERMAL POWER, the THERMAL POWER shall not be increased i

equal to or greater than 50% of RATED THERMAL POWER until the indicated AFD is within the required target band.

l gfgggg(E REOUIREMENTS 4.2.1.1 The indicated AFD shall be determined to be within its limits during POWER OPERATION above 15% of RATED THERMAL POWER by:

a.

Monitoring the indicated AFD for each OPERABLE excore channel:

1)

At least once per 7 days when the AFD Monitor Alarm is OPERABLE, and 2)

At least once per hour for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after restoring the AFD Monitor Alarm to OPERABLE status, b.

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

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

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

Penalty deviation outside of the above required target band shall be accumulated on a time basis of:

a.

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

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

4.2.1.3 The target flux difference of each OPERABLE excore channel shall be determined by measurement at least once per 92 Effective full Power Days.

The provisions of Specification 4.0.4 are not applicable.

4.2.1.4 The target flux difference shall be updated at least once per 31 Effective full Power Days by either determining the target flux difference i

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

POWER DISTRIBUTION LIM 115 kVEYlllLANJ1 3 2 VIE W Mii.R onkinued) pursuant to Specification 4.2.1.3 above or by linear interpolation between the most recently measured value and the predicted value at the end of the cycle life.

The provisions of Specification 4.0.4 are not applicable.

l l

l SOU1H TEXAS - UNITS 1 & 2 3/4 2-3

4 FIGURE 3.2-1 HAS BEEN DELETED 7

SOUTH TEXAS - UNITS 1 & 2-3/4 2-4 Unit 1 - Amendment No. # 27 Unit 2-AmendmentNo.11

4 POWER DISTRIBUTION LIMITS 3/4.2.2 HEATFLUXHOTCHANNELFACTOR-Fg WITINGCONDITIONFOROPERATION 3.2.2 F (Z) shall be limited by the following relationships:

9 RTP F (Z) $ Fn q

P F (Z) $ F0

K(Z) for P 5 0.5 9

0.5 Where

F RTP - the F limit at RATED THERMAL POWER (RTP) g q

specified in the Core Operating Limits Report (COLR).

THERMAL POWER

, and P = RATED THERMAL POWER K(Z) = the normalized F (Z) as a function of core height 9

specified in the 00LR.

APPLICABil.ITY:

MODE 1.

ACTION:

With F (Z) exceeding its limit:

9 a.

Reduce THERMAL POWER at least 1% for each 1% F (Z) exceeds q

the limit within 15 minutes and similarly reduce the Power Range Neutron Flux-High Trip Setpoint within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; POWER OPEkATION may proceed for up to a total of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; subsequent POWER OPERATION may proceed provided the Overpower AT-Trip Setpoint has been reduced at least 1%

for each 1% F (Z) exceeds the limit, q

b.

Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER above the reduced limit re-quired by ACTION a., above; THERMAL POWER may then be increased provided F (Z) is demonstrated through incore mapping to be 9

within its limit.

SOUTH TEXAS - UNITS 1 & 2 3/4 2-5 Unit 1 - Amendment No. 27 Unit 2 - Amendment No. 17

4 FIGURE 3.2-2 HAS BEEN DELETED SOUTH TEXAS - UNITS 1 & 2 3/1 2-6 Unit 1 - Amendment No. 27 Unit 2 - Amendment No.17

POWER DISTRIBUTION LIMITS Id$XEll_Qg(jgBEMENTS 4.2.2.1 The provisions of Specification 4.0.4 are not applicable.

4.2.2,2 f shall be evaluated to determine if F (Z) is within its limit by:

xy q

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

map at any THERMAL POWER greater than 5% of RATED THERMAL POWER, b.

Increasing the measured F component of the power distribution map xy by 3% to account for mtmufacturing tolerances and further increasing the value by 5% to account for measurement uncertainties, C

c.

Comparing the F computed (fxY) obtained in Specification 4.2.2.2b.,

above to:

xy 1)

The F limits for RATED THERMAL POWER (FR P) for the appropriate yy x

measured core planes given in Specification 4.2.2.2e and f.,

below, and 2)

The relationship:

f

=F

[1+PFxy(1-P)],

x l

Where F is the limit for fractional THERMAL POWER operation RTP expressed as a function of F

, PF is the power factor x

xy multiplier for F specified in the COLR, and P is the fraction xy of RATED THERMAL POWER at which F was measured.

xy d.

Remeasuring F according to the following schedule:

xy RTP 1)

When F is greater than the F limit for the appropriate x

measured core plane but less than the F relationship, additional power distribution maps shall be taken dF compared to F and F either:

xy a)

Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after exceeding by 20% of RATED THERMAL C

POWER or greater, the THERMAL POWER at which F*Y was last determined, or b)

At least once per 31 Ef fective Full Power Days (EFPD),

whichever occurs first.

SOUTH TEXAS -. UNITS 1 & 2 3/4 2-7 Unit 1 - Amendment No. 27 Unit 2 - Amendment No. 17

l POWER DISTRIBUTION LIMITS g gyELLLANCE REOUIREMENTS (Continued)

C RTP 2)

When the F is less than or equal to the F limit for the x

x appropriate measured core plane, additional power distribution RTP maps shall be taken and F compared to F and F at least x

once per 31 EFPD.

e.

The F limits used in the Constant Axial Offset Control analysis forRAkEDTHERMALPOWER(FRTP) shall be provided for all core planes x

containing Bank "D" control rods and all unrodded core planes as specified in the COLR per Specification 6.9.1.6; f.

The F limits of Specification 4.2.2.2e., above, are not applicable g

in the following core planes regions as measured in percent of core height from the bottom of the fuel:

1)

Lower core region from 0 to 15%, inclusive, 2)

Upper core region from 85 to 100%, inclusive, 3)

Grid plane regions at 22.4 t 2%, 34.2 1 2%, 46.0 1 2%, 57.8 1 2%,

69.5 1 2% and 81.3 1 2%, inclusive, and 4)

Core plane regions within i 2% of core height (1 3.36 inches) about the bank demand position of the Bank "D" control rods.

g.

With F exceeding F

, the effects of F n F (Z) shall be xy q

ovaluated to determine if F (2) is within its limits.

9 4.2.2.3 'When F (Z) is measured for other than F determinations, an overall 9

xy measured F (Z) shall be obtained from a power distribution map and increased q

by 3% to account for manufacturing tolerances and further increased by 5% to account for measurement uncertainty.

i SOUTH TEXAS - UNITS 1 & 2 3/4 2-8 Unit 1 - Amendment No. 27 Unit 2 - Amendment No.17 -

POWER DISTRIBUTION LIMITS 3/4.2.3 NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR LIMITING _C0!iD1110fLEOR OPERAT103 N

3.2.3 F shall be less than F P(1.0 + PFAH (1-P)]

g AH Where:

F RPT= the FN g

H Limit at RATED THERMAL POWER (RTP) specified in the Core Operating Limits Report (COLR)

Pf3g = the Power Factor Multiplier for F H specified in the COLR.

P=

THERMAL POWER RATED THERMAL POWER APPLICAB!LITY:

MODE 1.

ACTION:

N With F exceeding its limit:

g a.

Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . reduce the THERMAL POWER to the level where the

' LIMITING CONDITION FOR OPERATION is satisfied.

b.

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

N above; THERMAL POWER may then be irr:reased, provided F is demon-strated through incore mapping to be within its limit.

jggF,llt.ABCEJtEWRENENTS 4.2.3.1 The provisions of Specification 4.0,4 are not applicable.

FlH shall be demonstrated to b9 within its limit prior to operation 4.2.3.2

-above 75% RATED THERMAL POWER after each fuel loading and at least once per 31 EFPD thereafter by:

a.

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

Usingthemeasuredvalueo.fFfH which does not include an allowance b.

for measurement uncertainty.

SOUTH TEXAS - UNITS 1 & 2.

3/4 2-9 Unit 1 - Amendment No. 27 l

Unit 2 - Amendment No. 17 l.

POWER DISTRIBU110N LIMITS 3/4.2.4 QUADRANT POWER TIL1 RA110 LIMIT]NG CONDITION FOR OPERATID_N 3.2.4 The QUADRANT POWER TILT RATIO shall not exceed 1.02.

APPLICABILITY:

H0DE 1, above 50% of RATED THERHAL POWER *.

ACTION:

With the QUADRANT POWER TILT RATIO determined to exceed 1.02:

a.

Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months reduce THERHAL POWER at least 3% from RATED THERMAL POWER for each 1% of indicated QUADRANT POWER TILT RATIO in excess of 1 and similarly reduce the Power Range Neutron Flux-High Trip Setpoint within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , b.

Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and every 7 days thereaf ter, verify that F (Z) (by g

F evaluation) and F are within their limits by performing Surveil-xy g

lance Requirements 4.2.2.2 and 4.2.3.2.

THERMAL POWER and setpoint reductions shall then be in accordance with the ACTION statements of Specifications 3.2.2 and 3.2.3.

}yRVEILLANCE REOUIREMENTS 4.2.4.1 The QUADRANT POWER TILT RA110 shall be determined to be within the limit above 50% of RATED THERMAL POWER by:

a.

Osiculating the ratio at least once per 7 days when the alarm is OPERABLE, and b.

Calculating the ratio at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during steady-state operation when the alarm is inoperable.

4.2.4.2 The QUADRANT POWER 11LT RATIO shall be determined to be within the limit'when above 75% of RATED THERHAL POWER with one Power Range channel inoperable by using the movable incore detectors to confirm indicated QUADRAkT POWER TILT RATIO at-least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by either:

i a.

Using the four pairs of symmetric thimble locations, or b.

Using the inovable incore detection system to monitor the QUADRANT POWER TILT RATIO subject to the requirements of Specification 3.3.3.2.

See Special Test Exception. Specification 3.10.2.

SOUTH TEXAS - UNITS 1 & 2 3/4 2-10 1

I

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

0 3/4.1 REACTIVITY CONTROL SYSTEMS BASES 3/4.1.1 BORAT10N CONTROL 3/4.1.1.1 and 3/4.1.1.2 SHUTDOWN MARGIN A suf ficient SHUTDOWN MARGIN ensures that:

subtritical from all operating conditions.

(2) the reactor can be made clated with postulated accident conditions a(2) the reactivity transients asso-limits, and (3) the reactor will be maintained sufficiently suberitical tore controllable preclude inadvertent criticality in the shutdown condition.

SHUTDOWN MARGIN requirements vary throughout core life as a function of fuel depletion, RCS boron concentration, and RCS T,yg.

In MODES I anc 2, the most restrictive' condition occurs at EOL, with T,yg at no load operating temperature, and is associated with a postulated steam line break accident and resulting uncontrolled RCS cooldown..In the analysis of this accident, a minimum SHUTDOWN MARGIN of 1.75% ok/k is required to control the reactivity

. transient.

The 1.75% Ak/k-SHUTDOWN MARGIN is the design basis minimum for the 14 foot fuel using silver-indium-cadmitm and/or Hafnium control rods (Ref. FSAR l

. 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 anal-ysis assumptions.

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

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 1.75% Ak/k is based on an uncontrolled RCS cool-down from a steamline break accident.

The variable SHU1DOWN MARGIN requirement is based on the results of a boron dilution accident analysis, where the SHUT-DOWN MARGIN is varied as a function of RCS boron concentration, to guarantee a mininum of'IS minutes for operator action after a boron dilution alars, 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.

In MODES 3 and 4 it was assumed that at least one-reactor coolant pump was operating.. If at least one reactor coolant pump is not operatin

-shall. apply.g in MODE.3 or 4, then the SHUTDOWN MARGIN requirements for MODE

- 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.

u The MTC values of this specification are applicable to a specific set of plant conditions; accordingly.-verification of MTC values at conditions other than those explicitly stated will require extrapolation to those conditions in order to permit an accurate comparison.

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

Unit os.1, - Amendment No.10

(

- -, -..,,. - +, - - - - - - -

~- - ~ ~ ~ ~ ~* ~ " ~ ~ ~

~

..... s-1.ya _

. - - -~

_ REACTIVITY CONTROL SYSTEMS BfA_SES MODERATOR TEMPERATURE COEFFICIENT (Continued)

~

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

These corrections involved subtracting the incremental change in the MDC associated with a core condition of all rods inserted (most positive MDC) to an all rods withdrawn condition and, a conversion for the rate of change of moderator density with temperature at RATED THERMAL POWER conditions.

This value of the MDC was then transformed into the limiting E0L MTC value specified in the Core Operating Limits Report (COLR).

The 300 ppm surveillance MTC value represents a conser-vative value (with corrections for burnup and soluble boron) at a core condi-tion of 300 ppm equilibrium boron concentration and is obtained by making these corrections to the limiting EOL 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 burnup.

3/4.1.1.4 MINIMUM TEMPERATURE FOR CRITICALITY This specification ensures that the reactor will not be made critical with the Reactor Coolant System average temperature less than 561'F.

This limitation is required to ensure:

(1) the moderator temperature coefficient is within its analyzed temperature range, (2) the trip instrumentation 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.-

NDT 3/4.1.2 BORATION SYSTEMS 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 injec-tion-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 ficw path is sufficient to provide a-SHUTDOWN MARGIN from expected operating conditions of 1.75% Ak/k after xenon decay and cooldown to 200'F.

The maximum expected boration capability requirement occurs at E0L from full power equilibrium xenon conditions and requires 27,000 gallons of 7000 ppm borated water from the boric acid storage system or 458,000 gallons of 2500 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 B 3/4 1-2 Unit 1 - Amendment No. 27 Unit 2 - Amendment No.17

1 1

3/4.2 POWER DISTRIBUTION LIMITS M SE$____,_ _ _,_____ _ _ _ _ ___, _ _ _ _ _ ___,___,_,___,,

The specifications of this section provide assurance of fuel integrity during Condition I (Normal Operation) and 11 (Incidents of Moderate frequency) events by:

(1) maintaining the minimum DNBR in the core greater than or equal to 1.30 during normal operation and in short-term transients, and (2) limiting the fission gas release, fuel pellet temperature, and cladding mechanical pro-perties to within assumed design criteria.

In addition, limiting the peak linear power density during Condition I events provides assurance that the initial conditions assumed for the LOCA analyses are met and the ECC5 accept-ance criteria 11mit of 2200 F is not exceeded.

The definitions of certain hot channel and peaking factors as used in these specifications are as follows:

F (Z)

Heat Flux Hot Channel Factor, is defined as the maximum local heat 9

flux on the surface of a fuel rod at core elevation Z divided by the average fuci rod heat flux, allowing for manufacturing tolerances on fuel pellets and rods; F

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

the integral of linear power along the rod with the highest integrated power to the average rod power; and F*Y(2)

Radia, Peaking Factor, is defined as the ratio of peak power density to average power density in the horizontal plane at core elevatinn 2.

3/4.2.1 AXIAL FLUX DIFFERENCE The limits on AXIAL FLUX DIFFERENCE (AFO) assure that the F (2) upper bound envelope of the F limitspecifiedintheCoreOperatingL9mitsReport (COLR)timesthenorma19zedaxialpeakingfactorisnotexceededduringeither normal operation or in the event of xenon redistribution following power changes.

Target flux difference is determined at equilibrium xenon conditions.

The full-length rods may be positioned within the core in accordance with their respective insertion limits and should be inserted near their normal position for steady-state operation at high power levels.

The value of the target flux difference obtained under these conditions divided by the fraction of RATED THERMAL POWER is the target flux difference at RATED THERMAL POWER for the l

associated core burnup conditions. Target flux differences for other THERMAL L

POWER levels are obtained by multiplying the RATED THERMAL POWER value by the appropriate fractional THERMAL POWER level.

The periodic updating of the target i

flux difference value is necessary to reflect core burnup considerations.

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

-=

4 POWER DISTRIBUTION LIMITS BMES r

AX1AL FLUX DIFFERENCE (Continued)

Although it is intended that the plant will be operated vith the AFD within the target band required by Specification 3.2.1 about the target flux difference, during rapid plant THERMAL POWER reductions, control rod motion will cause the AFD to deviate outside of the target band at reduced THERMAL POWER levels.

This deviation will not affect the xenon redistribution suffi-ciently to change the envelope of peaking factors which may be reached on a subsequent return to RATED THERMAL POWER (with the AFD within the target band) provided the time duration of the deviation is limited.

Accordingly, a 1-hour penalty deviation limit cumulative during the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is provided for operation outside of the target band but withir. the limits specified in the l

COLR while at THERMAL POWER levels between 50% and 90% of RATED THERMAL POWER.

For THERMAL POWER levels between 15% and 50% of RATED THERMAL POWER, deviations of the AFD outside of_the target band are less significant.

The penalty of hours-actual time reflects this reduced significance.

Provisions for monitoring the AFD on an automatic bash are derived from the plant process computer through the AFD Monitor Alarm.

The computer deter-mines the 1-minute average of each of the OPERABLE excore detector outputs and provides an alarm message immediately if the AFD for two or more OPERABLE excore channels are outside the target band and the THERMAL POWER is greater than 90% of RATED THERMAL POWER.

During operation at THERMAL POWER levels between 50% and 90% and between 15% and 50% RATED THERMAL POWER, the computer outputs an alarm message when the penalty deviation accumulates beyond the limits of I hour and 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , respectively.

Figure B 3/4 2-1 shows a typical monthly target band.

3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR and NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR The limits on heat flux hot channel factor and nuclear enthalpy rise hot channel factor ensure that:

(1) the design limits on peak local power density and minimum DNBR are not exceeded and (2) in the event of a LOCA the peak fuel clad temperature will not exceed the 2200'F ECCS acceptance criteria limit.

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

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

Control rods in a single group mote together with no individual rod a.

insertion differing by more than i 12 steps, indicated, from the group demand position; b.

Control rod groups are sequenced with overlapping groups is described in Specification 3.1.3.6; SOUTH TEXAS - UNITS 1 & 2 B 3/4 2-2 Unit 1-AmendmentNo.23 Unit 2 - Amendment No. L

POWER DISTRIBUTION LIMITS BASES HEAT FLUX HOT CHANNEL FACTOR and NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR (Continued)

When an Fq easurement is taken, an allowance for both experimental error m

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 (2), remains within its limit.

The RTPq F

limit for RATED THERMAL POWER (F

) 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 thL indicated power tilt.

The 2-hour time allowance for operation with a tilt condition greater than 1.02 is provided to allow identificatico 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 purpcses of monitoring QUADRANT POWER TILT RATIO when one excore detector is inoperable, the moveable incore detectors are used to confirm that the normalized symmetric power distribution is consistent with the QUADRANT

. POWER TILT RATIO.

The incore detector monitoring is done with a full incore 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 cont.istent with the l

SOUTH TEXAS - UNITS 1 & 2 B 3/4 2-5 Unit 1 - Amendment No. 27 Unit 2 - Amendment No.17

4 4

(ONER DIsfRIBUT!DN LIMll$

WW_x 3/4.2.5 DNBPARAy[TER$(Continued) k.itial 05AR aswmptions and have bua analytically demonstrated adequate to saintain a minimum DNBR of greater t?,.n or equal to the design limit throughout each ana' tyred transient.

The indicated T vplue of 598'F and the indicated pressurizerpressurevalueof2201psiyaff9provided assuming that the readings from four channels will be averaged before comparing with the required limit.

The flow requirement (395,000 gptt) includes a measurement uncertainty of 3.5%.

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

SOUTH TEXA5 - UNITS 1 & 2 B 3/4 2-6 June 27, 199D l

4 A PIN!$TDATIVE CONTROLS SEH] ANNUAL RADI0 ACTIVE EFFL'JENT RELEASE REPORT (Continued) it Appendix B shall be supplemented with three additional categories:

solid wastes (as defined by 10 CFR Part 61), type of container (e.g., LSA, class of Type A, Type B, large Quantity) and SOLIDIFICATION agent or absorbent (e.g.

cement, ures forma!dehyde).

The Semiannual Radioactive Effluent Release Report to be submitted within 60 days af ter January 1 of each year shall include an annual summary of hour meteorological data collected over the previous year.

be either in the form of an hour-by hour listing on magnetic tape of windThis annual or in the fonn of joint frequency distributions of wind spee and atmospheric stability."

This same report shall include an assessment of the radiation doses due to the radioactive liquid and gaseous effluents released from the unit or station during the previous calendar year.

shall also include an assessment of the radiation doses from radioactiveThis same repo liquid and gaseous offluents to MEMBERS OF THE PUBLIC due to their activities inside the SliE BOUNDARY (Figures 5.1-3 and 5.1-4) during the report period.

assumptions used in making these assessments, i.e., specific activity, exposureAll time, and location shall be included in these reports.

conditions concurre,nt with the time of release of radioactive materials inThe meteor gaseous effluents, as det. ermined by sampling frequency and measurement, shall be used for determining the gaseous pathway doses.

The assessment of radiation doses shall be performed in accordance with the methodology and parameters in the OFFSITE DOSE CALCULATION MANUAL (ODCH).

The Semiannual Radioactive Effluent Release Report to be submitted W thin 60 days af ter January 1 of each year shall also include an assessment of radiation doses to the likely most exposed MEMBER OF THE PUBLIC from reactor releases and other nearby uranium fuel cycle sources, including doses from primary ef fluent pathways and direct radiation, for the previous calendar year to show conformance with 40 CFR Part 190 "Environe:. W Radiation Prctection Standards for Nuclear Power Operation.", Acceptable methous for calculating Guide 1.109, Rev. 1, October 1977.the cose contribution from liquid and gaseou The Semiannual Radioactive Effluent Release Reports shall include a list and description of unplanned releases from the site to UNRESTRICTED AREAS of radioactive materials in gaseous and liquid effluents made during the reporting period.

The Semiannyal Radioactive Effluent Release Reports shall include any changes made during the reporting period to the PROCESS CONTROL PROGRAM and the ODCM, pursuant to Specifications 6.13 and 6.14, respectively, as well as any major change to Liquid, Gaseous, or Solid Radwaste Treatment Systems pursuant to Specification 6.15.

It shall also include a listing of new loca-tions for dose calculations and/or environmental monitoring identified by the Land Use Census pursuant to Specification 3.12.2.

"In lieu of submission with the Semiannual Radioactive Effluent Release Report, the licensee has the option of retaining this summary of required meteoro-logical data on site in a file that shall be provided to the NRC upon request.

SOUTH TEXAS - UNITS 1 & 2 6-19

4 ADMINISTRATIVE CONTROLS SEMIANNUAL RADI0 ACTIVE EFFLUENT RELEASE _ REPORT (Continue @

The Semiannual Radioactive Effluent Release Reports shall also include the fellowing:

an explanation as to why the inoperability of liquid or gaseous effluent monitoring instrumentation was not corrected within the time specified in Specification 3.3.3.10 or 3.3.3.11, respectively; and description of the events leading to liquid holdup tanks or gas storage tanks exceeding the limits of Specification 3.11.1.4 or 3.11.2.6, respectively.

MONTHLY OPERATING REPORTS 6.9.1.5 Routine reports of operating statistics and shutdown experience, including documentation of all challenges to the PORVs oc safety valves, shall be submitted on a monthly besir, to the Director, Office of Resource Management, U.S. Nuclear Regulatory Commission Washington, D.C.

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

CORE OPERATING LIMITS REPORT 6.9.1.6.a Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT before each reload cycle, or any part of a reload cycle for the following:

1.

Moderator Temperature Coefficient BOL and r0L limits, and 300 ppm surveillance limit for Specification 3/4.1.1.3, 2.

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

Control Bank Insertion Limits for Specification 3/4.1.3.6, 4.

Axial Flux Difference limits ana target band for Specification 3/4.2.1, 5.

Heat Flux Hot Channel Factor, K(Z), Power factor Multiplier, P

and F

, for Specification 3/4.2.2, and 6.

Nuclear Enthalpy Rise Hot Channel Factor, and Power factor Multiplier for Specification 3/4.2.3.

The CORE OPERATING LIMITS REPORT shall be maintained available in the Control Room.

6.9.1.G b The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC in:

1.

WCAP 9272-P-A, " WESTINGHOUSE RELOAD SAFETY EVALUATION METHODOLOGY", July,1985 (W Proprietary).

(Methodology for Specification 3.1.1.3 - Moderator Temperature Coef ficient, 3.1.3.5 - Shutdown Rod Insertion Limit, 3.1.3.6 -

Control Bank Insertion Limits, 3.2.1 - Axial Flux Difference, 3.2.2 - Heat Flux Hot Channel Factor, and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor.)

SOUTH TEXAS - UNITS 1 & 2 6-20 Unit 1 - Amendment No. 9, 27 Unit 2 - Amendment No. I,17

ADMINISTRATIVE CONTROLS CORE OPERATING LIMl_1_S REPORT (Continued) 2.

WCAP 8385, " POWER DISTRIBUTION AND LOAD FOLLOWING PROCEDURES TOPICAL

-REPORT", September,1974 (W Proprietary).

(Methodology for Specification 3.2.1 - Axial Flux Difference (Constant Axial Offset Control).)

- 3.

Westinghouse letter NS-1MA-2198, T. M. Anderson (Westinghouse) to K. Kniel (Chief of Core Performance Branch, NRC) January 31, 1980 -

Attachment:

Operation and Safety Analysis Aspects of an Improved Load Follow Package.

(Methodology for Specification 3.2.1 - Axial Flux Difference (Constant Axial Offset Control).

Approved by NRC Supplement No. 4 to NUREG-0422 January, 1981 Docket Nos. 50-369 and 50-370.)

4.

NUREG 0800, Standard Review Plan, U.S. Nuclear Regulatory Commission, Section 4.3, Nuclear Design, July, 1981.

Branch Technical Position CPB 4.3-1, Westinghouse Constant Axial Offset Control (CAOC),

Rev. 2 July 1981.

(Methodology for Specification 3.2.1 - Axial flux Difference (Constant Axial Offset Control).)-

5.

WCAP 9220-P-A, Rev. 1, " WESTINGHOUSE ECCS EVALUATION MODEL-1981 VERSION", February 1982 (W Proprietary).

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

6.

WCAP.9561-P-A, ADO. 3 Rev. 1,:"BART A-1:

A COMPUTER CODE FOR THE BEST ESTIMATE ANALYSIS OF REFLOOD TRANSIENTS + SPECIAL REPORT:

THIMBLE MODELING W ECCS EVALUATION MODEL", July, 1986, (W Proprietary).

(Methodolcgy for Specification 3.2.2 - Heat flux Hot Channel factor.)'

6.9.1.6.c The core operating limits shall be determined so that 7 1 applicable 1

limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits..ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met.

6.9.1.6.d The. CORE 0PERATING LIMITS REPORT, including any mid-cycle re,isions or supplements thereto, shall be provided upon issuance, for each

~

reload ~ cycle,-to the-NRC Document Control. Desk, with copies to the Regional Administrator and Resident Inspector.

SPECIAL REPORTS

6.9.2. Special reports shall be submitted to the Regional Administrator of the

-Regional Office of the NRC within the time period specified for each report.

SOUTH TEXAS - UNITS 1 & 2 6-20a Unit 1 - Amendment No. 27 Unit 2 - Amendment No 17

.c

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