Text

Forwards Final Draft Tech Specs & Technical Changes W/ Justifications.Tech Specs Consistent W/Fsar.Ser & as-built Facility W/Exceptions,Editorial Changes & Corrections Noted on Encl marked-up Pages
	ML20214W345
Person / Time
Site:	South Texas
Issue date:	06/05/1987
From:	Vaughn G; HOUSTON LIGHTING & POWER CO.
To:	; NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
References
	ST-HL-AE-2232, NUDOCS 8706150222
	Download: ML20214W345 (146)
	v • d • e

The Light Company n~<,,, w,w - m H ,-- .. .w ,,memn June 5, 1987 ST-HL-AE-2232 File No.: C9.06 10CFR50.36 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555 South Texas Project Unit 1 Docket No. STN 50-498 Certification of Final Draft Technical Specifications

Reference:

May 18, 1987 letter (ST-AE-HL-91263) from Mr. Robert L. Perch to Mr. J. H. Goldberg This letter is provided to certify to the best of my knowledge that the Final Draft Technical Specifications for South Texas Unit 1 are consistent with the Final Safety Analysis Report (FSAR), Safety Evaluation Report (SER) and the as-built facility with the exceptions, editorial changes and corrections noted on the attached marked pages. The basis for this certification is as follows.

In the fall of 1985, a three party review group was formed from individuals representing Houston Lighting & Power (H14P), Bechtel and Westinghouse to develop Technical Specifications (Tech Specs) for South Texas Unit 1. Each section of the Standard Technical Specifications (STS),

NUREC-0452, Draft Revision 5 was marked accordingly to reflect the South Texas three safety train design to reflect plant specific design features such as rapid refueling, four Auxiliary Feedwater (AFW) Pumps, incontainment storage pool, Qualified Display Processing System (QDPS), etc., and to provide flexibility in areas where there exists 100 percent redundancy between the three trains. Each section of the STS was marked and documented by the three party review group and submitted to HIAP Engineering, HL&P Operations, Bechtel and Westinghouse for review and approval. A draft Technical Specifications submittal oased on the results of the three party review was transmitted to the Staff in January 1986. A revision was submitted in July 1986 which included plant specific data not available at the time of the original submitted design changes, results of further analysis, and extended surveillance frequencies and out of service times for the reactor protection system instrumencation based on Westinghouse WCAP-10271.

8706150222 870605 I PDR ADOCM 05000498 A PDR l L3/NRC/eq h \s

Ilouston Lighting & Power Company ST-HL-AE-2232 File No.: G9.06 Page 2 In November 1986, Enercon Energy Services was contracted by HL&P to perform an independent review of the South Texas Technical Specifications as of July 1986. Enercon's objective was to assure that the specifications were consistent with the plant licensing basis and represent the current documented plant design. The specifications were reviewed against the FSAR, SER, applicable system descriptions, design calculations and analysis, appropriate correspondence, the Final Environmental Statement (FES), Environmental Report (ER) and the Offsite Dose Calculation Manual (ODCM) as well. Each problem or inconsistency identified during the review was recorded on a computerized Technical Specification Review Punchlist.

Resolution of the items identified by Enercon was completed subsequent to the issuance of Proof & Review Technical Specifications by NRC in February 1987. Enercon characterized the items identified as isolated errors and further stated that no evidence of programmatic deficiencies in the formalization of the STPEGS Technical Specifications was found during the review process.

Additionally, HL&P undertook a review of short term Technical Specification improvements, which had been granted for nuclear plants, such as Seabrook, Perry, Vogtle and McGuire, to identify whether similar relief could be granted for South Texas. Based upon the results of this review, each improvement was addressed on a plant specific basis for South Texas and submitted to the Staff for approval.

Changes to the Technical Specifications resulting from the preoperational testing of the as-built facility have been identified for resolution. Some issues have previously been resolved while the remaining issues are addressed in the attached pages. Bechtel and HL&P reviewed the Technical Specifications against the FSAR and provided additional changes to the FSAR to reflect current analysis and design calculations. These FSAR changes will be provided in an upcoming amendment to the FSAR as well as submitted via separate letter for the Staff's immediate use. These changes do not affect certification of the South Texas Technical Specifications and are provided as a result of FSAR, SER, and Technical Specification consistency review. Westinghouse has certified to HL&P that the Technical Specification values within the Westinghouse scope are derived from the analyses and evaluations included in the South Texas Project FSAR submitted pursuant to 10CFR50.34 and in accordance with the Westinghouse Quality Assurance Plan (WCAP-8370/7800).

Regarding the items identified in the Staff's May 18, 1986 letter which required FSAR revisions, information or revised values to support certification of Technical Specifications, the following submittals were provided:

L3/NRC/eq I

)

Ilouston Lighting & Power Company ST-HL-AE-2232 File No.: G9.06 Page 3

1. Reactor Trip Setpoints and Engineered Safety Feature Setpoints were provided on May 8, 1987 via letter ST-HL-AE-2169;

2. Balance of Plant Setpoints were provided on May 14, 1987 via letter ST-HL-AE-2174;

3. Battery charger capacities were provided to the Staff on June 3, 1987 via letter ST-HL-AE-2183; and

4. The Reactor Coolant System Temperature average (Tavg) will be provided via letter ST-HL-AE-2192.

If you have any questions regarding the certification of Final Draft Technical Specifications for South Texas, please contact me or Ms. Frostie A.

White at (512) 972-7921 or (512) 972-7985, respectively.

Y G. E. Vaughn Vice President Nuclear Plant Operations FAW/ljm Attachments: 1. Technical Changes with Justification

2. Final Draft Technical Specifications, Specification 3.4.6.1 ACTION

3. Final Draft Technical Specifications, Specification 3/4.3.3.1

4. Final Draft Technical Specifications, Surveillance Requirement 4.5.1.1.c

5. Final Draft Technical Specifications, Specification 4.6.1.3.e

6. Editorials & Typographical Changes L3/NRC/eq

~

l Houston Lighting & Power Company ST-HL-AE-2232 File No.: G9.06 Page 4 cc:

Regional Administrator, Region IV M.B. Lee /J.E. Malaski Nuclear Regulatory Commission City of Austin 611 Ryan Plaza Drive, Suite 1000 P.O. Box 1088 Arlington, TX 76011 Austin, TX 78767-8814 N. Prasad Kadambi, Project Manager A. von Rosenberg/M.T. Hardt U.S. Nuclear Regulatory Commission City Public Service Board 7920 Norfolk Avenue P.O. Box 1771 Bethesda, MD 20814 San Antonio, TX 78296 Robert L. Perch, Project Manager Advisory Committee on Reactor Safeguards U.S. Nuclear Regulatory Commission U.S. Nuclear Regulatory Commission 7920 Norfolk Avenue 1717 H Street Bethesda, MD 20814 Washington, DC 20555 Dan R. Carpenter George Smith Senior Resident Inspector / Operations Westinghouse Electric Corp.

c/o U.S. Nuclear Regulatory Monroeville Nuclear Center Commission Northern Pike P.O. Box 910 Monroeville, PA 15146 Bay City, TX. 7741*

J.D. Shiffer

. Claude E. Johnson Vice President Senior Resident Inspector /STP Nuclear Power Generation c/o U.S. Nuclear Regulatory Pacific Gas & Electric Company Commission 77 Beale Street, Room 1445 P.O. Box 910 San Francisco, CA 94160 Bay City, TX 77414 Dr. W.R. Corcoran M.D. Schwarz, Jr., Esquire 21 Broadleaf Circle Baker & Botts Windsor, CT 06095 One Shell Plaza Houston, TX 77002 J.R. Newman, Esquire Newman & Holtzinger, P.C.

1615 L Street, N.W.

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

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION In the Matter )

)

Houston Lighting & Power ) Docket Nos. 50-498 Company, et al., ) 50-499

)

South Texas Project )

Units 1 and 2 )

AFFIDAVIT G. E. Vaughn being duly sworn, hereby deposes and says that he is Vice President, Nuclear Plant Operations, of Houston Lighting & Power Company; that he is duly authorized to sign and file with the Nuclear Regulatory Commission the attached Certification of Final Draft Technical Specifications; is familiar with the content thereof; and that the matters set forth therein are true and correct to the best of his knowledge and belief.

r1R</

G. E. Vaughn /'

Vice President Nuclear Plant Operations STATE OF TEXAS )

)

COUNTY OF MATAGORDA )

Subscribed and sworn to before me, a otary Public in and for Matagorda County, Texas this g day of turLQ , 1987.

OC4 . M Notary Public in and for the State of Texas My commission expires:

lo 88 l

u

q l

Attechmsnt 1 ST-HL-AE-2232 Page 1 of 39 Technical Changes with Justification

1. Revisions reflect incorporation of Containment Phase "B" Isolation signal. These changes were previously submitted to the NRC via letter ST-HL-AE-2182 dated May 19, 1987.

2. The F-15 interlock is logically equivalent to NOT P-10. Table 3.3-1, l ACTION 8 for P-10 requires that with less than the Minimum Number of i Channels OPERABLE, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months determine by observation of the I associated permissive annunciator windows that the interlock is in its l required state. Since P-15 and NOT P-10 are logically the same, the required ACTION for Table 3.3-3, 9d should be 21,

3. The "**" note for 5f, Table 3.3-4 is inappropriate since it applies to time constants for steam pressure.

4. A change in feedwater isolation valve closure time from 5 to 10 seconds is reflected. The need for this change has been identified in recent preoperational tests, and the justification for this change will be I

provided as soon as possible in a separate letter to the Staff.

5. At STP, the interlock from power range neutron flux to excessive cooldown protection is designated P-15. It is logically equivalent to NOT P-10, as shown on FSAR Figure 7.2-4. The July 24, 1985 letter from H. R.

Denton (NRC) to L. D. Butterfield (Westinghouse Owners Group) provided tha NRC markup of the W Standard Technical Specifications for WCAP-10271 impacts and the rationale for these revisions. In that letter, NRC indicated that the ACOT for P-10 should be revised from M(8) to R, where Note 8 had said "with power greater than or equal to the Interlock Setpoint the required ANALOG CHANNEL OPERATIONAL TEST (ACOT) shall consist of verifying that the interlock is in the required state by observing the permissive annunciator window". The revision to at least once per 18 months (R) is justified by :'ote 4 in the " Notes on the format of marked-up W STS" of that letter. We believe that the same rationale noted for P-10 in RTS is applicable to P-15 in the ESFAS, and have therefore revined the ACOT for P-15 in Table 4.3-2 from M(3) to R, and deleted Note 3.

6. The "***" note was added for clarity and consistency with Technical Specification 3.6.2.3 where it is indicated that there must be OPERABLE a minimum of two units in two groups and one unit in the third group giving a total of 5 RCFCs. If the "***" is not added, then it is necessary to make the Total Number of Channels "5" and the Minimum Channels Operable "3". Please note that this portion of the Remote Shutdown Table 3.3-9, as agreed upon by the Staff on May 4 and 5, 1987, does not have to meet single failure criteria. Additionally, note that only "3" RCFCs ce needed to cool containment during transients.

7. These flowrate monitors have no trip alarm or interlock setpoints.

Therefore, the ACOT is not applicable.

i I

L3/NRC/eq J

Attrchm3nt 1 ST-HL-AE-2232 Page 2 of 39

8. Valve CV0215 for Units 1 & 2 has been locked in place. The handwheel has been removed and stored. A riser clamp, pipe and plate was fabricated and welded into place with weld points on the riser clamp and another on the pipe. Also, the clamp was secured together with a padlock.

9. It will be necessary to add a footnote to LCO 3.5.6c since the RHR Suction Isolation Valves are being power locked out in order to meet 10CFR50, Appendix R requirements. This issue has been discussed with the Staff on May 14, 1987.

10. A new fluence curve is provided for immediate use. See HL&P letter ST-HL-AE-2230 dated June 3,1987 for justification and associated FSAR changes,

11. The total relieving capacity is the product of the individual valve capacity and the total number of valves (1,032,645 lbs/hr/ valve x 20 valves - 20.65E6 lbs/hr). The total secondary steam flow is 16.94E6 lbs/hr as shown on FSAR figure 10.1-1.

12. The value of P-8 changed in WCAP-11273 Rev. O, South Texas Setpoint Methodology, from 48% to 40% Rated Thermal Power as well as 90% to 92%

Nominal full loop flow.

13. Editorial; reflects recent organization changes.

14. The total number of channels has been clarified to show 1/ loop-4 loops.

The minimum channels OPERABLE has been revised to permit one channel to be inoperable. These revisions are necessary so that operation may continue with less than 1 Tavg-low channel per loop, provided certain conditions are met.

15. As indicated in FSAR Figure 10.2-8. there are six turbine reheat stop and intercept valves.

16. There appears to be a typographical error in the standard since the notes apply to the mark ups provided.

17. The changes provided are necessary in order to make Specification 3.8.3.2 consistent with the Mode 5 and 6 Specification 3.8.1.2. Note that Specification 3.8.1.2 requires two diesel generators to be OPERABLE.

Accordingly, two trains of emergency power are required.

18. The Shutdown Margin for South Texas, as shown in Figure 3.1-1, is variable and the Bases should reflect this factor.

19. South Texas specific design.

20. According to Standard Review Plan 6.2.4, sealed closed isolation valves include mechanical devices to seal or lock the valve closed or to prevent power from being supplied to the valve operator. HL&P cannot determine a reason to require the valves to be sealed closed.

l L3/NRC/eq l

~ ATTACHMENT /

' ST AAY '

p M' M l a i n n A P "

ii mL UIUW

. LIMITING SAFETY SYSTEM SETTINGS BASES Pressurizer Pressure In each of the pressurizer pressure channels, there are two independent bistables, each with its own trip setting to provide for a High and Low Pressure trip thus limiting the pressure range in which reactor operation is permitted.

The Low Setpoint trip protects against low pressure which could lead to DNB by tripping the reactor in the event of a loss of reactor coolant pressure.

On decreasing power, the Low Setpoint trip is automatically blocked by P-7 (a power level of approximately 10% of RATED THERMAL POWER with turbine impulse chamber pressure at approximately 10% of full power equivalent); and on increasing power, automatically reinstated by P-7.

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

Pressurizer Water Level The Pressurizer High Water Level trip is provided to prevent water relief through the pressurizer safety valves. On decreasing power, the Pressurizer High Water Level trip is automatically blocked by P-7 (a power level of approxi-0, mately 10% of RATED THERMAL POWER with a turbine impulse chamber pressure at approximately 10% of full power equivalent); and on increasing power, auto-matically reinstated by P-7.

Reactor Coolant Flow The Low Reactor Coolant Flow trips provide core protection to prevent DNB by mitigating the consequences of a loss of flow resulting from the loss of one or more reactor coolant pumps. grMdf 98.%

On increasing power above P-7 (a power level of approximately 10% of b

RATED THERMAL POWER or a turbine impulse amber pressure at approximately 10%

of full power equivalent), an au matic Reactor trip will occur if the flow in more than one loop drops below 0 o nominal full loop flow. Above P-8 (a power level of approximately 4 of RATED THERMAL POWER) an automatic Reactor trip will occur if the flow in any single loop drops below 0 of. nominal full loop flow. Conversely, on decreasing power between P-8 nd the P-7, an automatic Reactor trip will ccur on low reactor coolant flo in more than one loop, and below P-7 the trip function is automatically locked.

Steam Generator Water Level foYo

1PYOW Y f l0 The Steam Generator Water Level Low-Low trip protects the reactor from loss of heat sink in the event of a sustained steam /feedwater flow mismatch 3 resulting from loss of normal feedwater. The specified Setpoint provides

,) allowances, for starting delays of the Auxiliary Feedwater System.

SOUTH TEXAS.- UNIT 1 B 2-6_

TABLE 3.3-3 (Continued) :

2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION M MINIMUM 5 TOTAL NO. CHANNELS CHANNELS ' APPLICABLE 7 FUNCTIONAL UNIT -

0F CHANNELS TO TRIP OPERABLE MODES ACTION c .. .i.

5

-4

3. ContainmentIsolation(Continued) '..

H b. ' Containment Ventilation Isolation

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

2) Actuation Relays *** 3 2 3 1, 2,'3, 4 18 g 3) SafetyInjection*** See Item 1. above for all Safety Injection initiating functions and

requirements.

w ,

$ 4) RCB Purge Radioactivity-High 2 1 2 1,2,3,4,5 y ,6[ ' 18

5) Containment Spray- See Item 2. above for Containment Spray manual initiating functions Manual Initiation and requirements.

6) Phase "A" Isolation- See Item 3.a. above for Phase "A" Isolation manual initiating Manual Isolation function and requirements.

ps,a 4

r 1C

4. , Steam Line Isolation 3%%

c) d "T1 ammme 1 a.- Manual Initiation

.b

., 9 n

1) Individual- 2/ steam line 1/ steam line 2/ operating 1,2,3 24 Sd steam line ga

2) System 2 1 2 1,2,3 23

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

I

_ ATTACHMENT /

. ST HL At 2R3.2 i._PAGE r OF 39

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v. TABLE 3.3-3 (Continued) 8 Ec ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION

t x

3; MINIMUM i

. TOTAL NO. CHANNELS CHANNELS APPLICABLE c- FUNCTIONAL UNIT . OF CHANNELS TO TRIP OPERABLE MODES ACTION 4 5

5. Turbine Trip and i

." Feedwater Isolation (Continued)

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

loop each loop or T

avg

- Low 1/ loop-q g,,p 1/ loop in any jggp 1/ loop -

g, 3,

, M , 2, 3 15* k

{

Y e. Safety Injection See Item 1. for all Safety Injection initiating M functions and requirements.

l

f. T avg

-L w c incident with Reactor Trip (P-4)** 4 (1/ loop) 2 3 1,2,3 20*

6. Auxiliary Feedwater

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

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

c. Actuation Relays 3 2 3 1,2,3 22 $%

d. Stm. Gen. Water Level--

Low-Low e

m Start Motor- o (ro h Driven Pumps and Turbine-4/stm. gen. 2/stm. gen.

in any stm.

3/sta. gen.

in each 1,2,3 20* $$s M P Driven Pump gen. stm. gen.

e. Safety Injection See Item 1. above for all Safety Injection initiating functions and requirements.

f. Loss of Power (Motor See Item 8. below for all Loss of Power initiating functions and i Driven Pumps Only) requirements.

g ,

3 ..

TABLE 3.3-3 (Continued)

[ ,

j.

h ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION m -

x s

MINIMUM i

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

5 .

9. Engineered Safety features

" Actuation System Interlocks (Continued)

c. Reactor Trip, P-4 2 1 2 1,2,3 23

d. Power Range Neutron 4 2 3 1,2,3 4 8 a l.

' ~

Flux Input to Excessive , _

Cooldown Protection, P-15 .

p 10. Control Room Ventilation T a. Manual Initiation 3(1/ train) 2(1/ train) 3(1/ train) All 27

b. Safety Injection See Item 1. above for all Safety Injection initiating ,

functions and requirements, Automatic Actuation Logic 3 2 3 All 27 c.

and Actuation Relays Control Room Intake Air 2 1 2 All 28*

d.

Radioactivity - High I

e. Loss'of Power See Item 8. above for all Loss of Power initiating functions -

and requirements.

11. FHB HVAC im@M Ag ^c

_3(1/ train) 29, 30 o

a. Manual Initiation 3(1/ train) 2(1/ train) 1, 2, 3, 4 or with irradiated fuel in spent A

]M$ s fuel pool

m TABLE 3.3-4 8

g ENGINEERED SAFETY FEATURES ACTUATION SYSTEN INSTRUMENTATION TRIP SETPOINTS N TOTAL SENSOR ERROR R FUNCTIONAL UNIT -

ALLOWANCE (TA) Z (S) TRIP SETPOINT ALLOWABLE VALUE

1. Safety Injection (React,or Trip, ,

4-5 Feedwater Isolation, Control '.

H Room Emergency Ventilation, Start '

M Standby Diesel Generators, Reactor Containment Ccoling Fans gg Essential Cooling Water) 4

a. Nanual Initiation N.A. N.A. N.A. N.A. N.A.

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

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

d. Containment Pressure--High 1 3.6 0.71 2.0 $ 3.0 psig 5 4.0 psig

e. Pressurizer Pressure--Low 13.1 10.71 2.0 1 1850 psigN 1 1842 psin#8

f. Compensated Steam Line 13.6 10.71 2.0 1 735 psig Pressure-Low 3 714.7 m cg m>I

g. Compensated T -Low-Low 4.5 i

0.5 1.0 COLD 1 532*F 1 528*F*j"}

(interlocked with P-15) # pa a

'w~

2. Containment Spray f'

a. Nanual Initiation N.A. N.A. N.A. N.A. N.A. , I

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

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

d. Containment Pressure--High-3 3.6 0.71 2.0 $ .5 psig 5 0.5 psig _ 1 FINAL. DRAFT

TABLE 3.3-4 (Continu d) '

i~

z ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS N TOTAL SENSOR ERROR ALLOWABLE VALUE ALLOWANCE (TA)- Z (S) TRIP SETPOINT hFUNCTIONALUNIT -

n.

e- 3. Containment Isolation

i5

a. Phase "A" Isolation s

N.A. N.A. N.A. N.A. N.A.

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

2) Automatic Actuatio'n Logic N.A.

N.A. N.A. N.A. N.A.

3) Actuation Relays N.A. .

i 4) SafetyInjection SeeItem1.aboveforallSafetyInjectionTripSetpointsandAllowable R Values.

y b. Containment Ventilation Isolation _

N.A. N.A. N.A. N.A. N.A.

1) Automatic Actuation Logic . ,

N.A. N.A. N.A. N.A.

2) Actuation Relays N.A. .

I 3) SafetyInjection See Item 1. above for all safety Injection Trip Setpoints and Allowable Values. .

1.3x10

<6.4x104==y1

4) RCB Purge 3.1x10 ' 1.8x10 4

Radioactivity-High pCi/cc pCi/cc pCi/cc <b104 p /cc pti/cc - -

g p

A 5) Containment Spray - See Item 2. above for Containment Spray manual initiation Trip s 3$h p

/1\ Manual Initiation Setpoints and Allowable Values. gg y Ip# 6) Phase "A" Isolation - See Item 3.a. above for Phase "A" Isolation manual initiation NI aug g 15 % Manual Initiation Trip Setpoints and Allowable Values. u -e

4. Steam Line Isolation '

h T N.A. N.A.

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

N.A{ -e m

- - - - - _ . - _ m

e. . _

, \

t l I -

Wsex.y eg' i

d

c. I'hase "B" ~C5cl*+3*n '

') Amh n ke AeG K v.A. M.A. W. A. n.A. V, A .

, Lgic.

2) hckush w RsI9s 91.h. W.A. s W h. W. A. N. A.

3) C.wkn J Pn ss = -- 3. c, 0.71 2. o <

Wrgh-3 41.s ig .to. r tg '

A) Ca M - d Spen - See 1% 3.. above. k- CmMn~sd p S y m4 ; nil!& Trip

%st ksiink SeipoMs ud Alto.asIsle % lass. , ,

t E

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mit_.g 3 ;

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4(s 9.

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en

p D 9 f TABLE 3.3-4 (Continued)

' i x ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION. TRIP SETPOINTS a SENSOR ERROR TOTAL W ALLOWANCE (TA) Z (S)

TRIP SETPOINT ALLOWABLE VALUE

$; FUNCTIONAL UNIT e

o g, l l e 5. Turbine Tr.ip and Feedyater .., j y Isolation (Continued) *

H 4.0 3.19 0.6 >91.8% of > 90.9% of RCS Flow-Low loop design loop design flow **** flow ****

or ,O 4.5 L 36 0.8

> 571.1*F

>574lF T,yg-Low

e. Safety Injection See Item 1 above for all Safety Injection Trip R* Setpoints and Allowable Values. .0 w > 571.1F 4.5 1.36 0.8 >574lF. l h f. T,yg-Low Coincident with Reacto'r Trip (P-4) W - .

(Feedwater Isolation Only)

6. Auxiliary Feedwater N.A. N.A. N.A. H.A.

a. Manual Initiatton N.A. ,

N.A. N.A.

b. Automatic Actuation Logic 'N.A.

N.A. N.A. N.A. N.A. Nk. I c'. Actuation Relays !

12.75 2.0+0.2# > 33.0% of > 31.5% of 3 oi$ 2"d

d. Steam Generator Water 15.0 Harrow range Harrow ran -N Level--Low-Low' instrument d'$ @

instrument ::

span. span. Qm wh a%s f

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

FINAL. DRAFT

L. .. - ............................................................._....._.._.

. ST.HL AE. JAM FsfEEm)240,inat un TABLE 3.3-5 M gae p*E h3 ; ,, ENGINEERED SAFETY FEATURES RESPONSE TIM INITIATION SIGNAL AND FUNCTION _

RESPONSE TIME IN SECONDS . .

1. Manual' Initiation

a. Safety Injection (ECCS) . N.A. ,

b. Containment Spray N.A.

c. Phase "A" Isolation N.A 1 N. A.

d Phase "S

Isola.+ ion N.A.

eg. Containment Ventilation Isolation f g. Steam Line Isolation - N.A.

Feedwater Isolation N.A.

gf.

Auxiliary Feedwater N.A. .

h 5(..

Essential Cooling Water N.A.

Q.

jf. Reactor Containment Fan Coolers N.A.

Control Room Ventilation N.A.

K/.

Reactor Trip N.A.

1 )l.

N.A.

m/. Start Diesel Generator

2. Containment Pressure--High-1

a. Safety Injection (ECCS) 1 27I1)/12(5)

1) Reactor Trip < 2(3)

2) Feedwater Isolation k

3) Phase "A" Isolation < 33 (1)/23(2) 23(1)/13(2)

4) Containslinesnt Ventilation Isolation -

4- s e inek

5) Auxiliary Feedwater 1 60

6) Essential Cooling Water < 62II)/52(2)

7) Reactor Containment Fan Coolers 38(1)/28(2)

8) Control Room Ventilation 72(1)/62(2)

9) Start Standby Diesel Generators 5 12

- i Q

SOUTH TEXAS - UNIT 1 3/4 3-37 ,

e

~ ..... ._ .... _ . _ , __ ... ..__...... ............._ _ _ _ _ _ . .

ATTACHMENT '

. ST.HL AE J//S.2 Fm@EF./3LL%,,,,

NAL UKAti TABLE 3.3-5 (Continued) .__

~

- - ENGINEERED SAFETY FEATURES RESPONSE TIMES - -

~

RESPONSE TIME IN SECONDS INITIATING SIGNAL AND FUNCTION

3. Pressurizer Pressure--Low ,

a. Safety Injection (ECCS) I 1 27(1)/12(5)

1) Reactor Trip < 2(3)

2) Feedwater Isolation k* Y

3) Phase "A" Isolation s 33(1)/23(2)

Containment Ventilation Isolation - N.A.

4)

5) ' Auxiliary Feedwater s 60

6) Essential Cooling Water < 62(1)/52(2) ~

7) Reactor Containment Fan Coolers 38(1)/2'8(2)

8) Control Room Ventilation 1 2(1)/62(2) 7

9) Start Standby Diesel Generators 1 12

4. Compensated TCOLD ~l "

a. Safety Injection (ECCS) N.A.

Reactor Trip N.A.

Qi 1)

N.A.

N 2) Feedwater Isolation Phase "A" Isolation N.A.

3)

Containment Ventilation Isolation H.A.

4)

Auxiliary Feedwater N.A.

5) 1 Essential Cooling Water H.A.

6)

Reactor Containment Fan Coolers H.A.

7)

Control Room Ventilation N.A.

8)

Start Diesel. Generators N.A.

9)

b. Steam Line Isolation .H.A.

5. Compensated Steam Line Pressure--Low

a. Safety Injection (.ECCS) 1 22(4)/12(5)

Reactor Trip < 2(3) 1)

2)- Feedwater Isolation M 9

3) Phase "A" Isolation 1 33(1)/23(2)

C'ontainment Ventilation Isolation N.A.

l 4)

Auxiliary Feedwater < 60 .

5)

6) Essential Cooling Water k62(1)/52(2)

< 7) Reactor Containment Fan Coolers 1 38(1)/28(2)

SOUTH TEXAS - UNIT 1 3/4 3-38

TABLE 3.3-5 (Co.itinued)

n. ! ATTACHMENT /

F ENGINEERED SAFETY FEATURES RESPONSE TIMES , ST HL AE-Sp6A

";- PAGE F/ OF M

~

INITIATING SIGNAL AND FUNCTION RESPONSE TIME'IN SECONDS

5. Compensated Steam Line Pressure--Low (Continued) '-

8) Control Room Ventilation ' < 72(1)/62(2)

9) Start Diesel Generators 7 12

b. Steam Line' Isolation " 8(3)

6. Containment Pressure--High-3

-- Containment Spray .

< 30(1)/20(2) b Phas'e B" Isola _ tion y( ,g (a) i Containment Pressure--High-2

7. -

Steam Line Isolation 1 7(3)

8. Steam Line Pressure - Negative Rate--High Steam Line Isolation N.A.

9. Steam Generator Water Level--High-High

a. Turbine Trip 1 3(3)

b. Feedwater Isolation i

10. Steam Generator Water Level--Low-Low

a. Motor-Driven Auxiliary Feedwater Pumps 1 60 -

b. Turbine-Driven Auxiliary -

Feedwater Pump 1 60

11. RWST Level--Low-Low Coincident with Safety Injection Automatic Switchover to Containment Sump 1 32(2) _

12. Loss of Power

a. 4.16 kV ESF Bus Undervoltage - < 12

~

.(Loss of Voltage)

b. 4.16 kV ESF Bus Undervoltage < 49 (Tolerable. Degraded Voltage Coincident with Safety Injection)

~

h SOUTH TEXAS - UNIT 1 3/4 3-39

7 _____;__________________

TABLE 3.3-5 (Continued)

FINAL. DRAFT ATTACHMENT I "L ^E-ENGINEERED SAFETY FEATURES RESPONSE TIMES gel 5 0 INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONOS s a. . Loss of Poeur (Con +inu sd)

c. 4.16 kV ESF Bus Undervoltage ,

-< 65 (Sustained Degraded Voltage)

13. RCB Purge Radioactivity-High

a. Containment Ventilation Isolation (9h.chkD < 73(2) b.c % m.4 w h % -g % cq g z ,g(q Compensated Teold--' * ~

14.

a. Turtiine Trip N.A.

b. Feedwater Isolation N.A.

15. Feedwater Flow - High Coincident with 2 of 4 Loops Having Either Reactor Coolant Flow - Low or T,yg - Low

a. Turbine Trip - Reactor Trip N.A.

b. Feedwater Isolation N.A.

16. T ,yg - Low Coincident with Reactor Trip

(, Feedwater Isolation N.A.

17. Control Room Intake Air Radioactivity - High Control Room Ventilation 1 78(2)

18. Spent Fuel Pool Exhaust Radioactivity - High FHB HVAC Emergency Startup i 42(2) e 4

L l

SOUTH TEXAS - UNIT 1 3/4 3-40 ,

f f )

l i

i TABLE 4.3-2 (Centinued) '

$ 1 ENGINEERED' SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION !

SURVEILLANCE REQUIREMENTS :

_9 x TRIP i DIGITAL OR y

ANALOG ACTUATING MODES j Q ' CHANNEL DEVICE MASTER SLAVE FOR WHICH ;

$ a CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY- RELAp, SURVEILLANCE i F

, CHANNEL ' LOGIC TEST TEST TEST., IS REQUIRED CHECK CALIBRATION TFST TEST

c. FUNCTIONAL UNIT ., '

. I h 3. Containment Isolat' ion (Continued) j i

e

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

4')'RCB' Purge H N.A. N.A. N.A. N.A. 1,2,3,4,5*,6*l Radioactivity-High S R ;

1 i

5) Containment Spray - .See Item 2. above for Containment Spray manual initiation Surveillance C

, gp/p , Manual Initiation Requirements.

j Item 3.a. above for Phase "A" Isolation manual initiation R 6) Phase "A" Isolation- See Surveillance Requirements. ,

j

Manual Initiation . ;

w r k 4. Steam'Line Is'oiation . ,

I N.A. R N.A. H.A. N.A. 1,2,3 i

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

N.A. M(1) M(6) Q 1, 2, 3 i

b. Automatic Ac,tuation N.A. N.A N.A Logic and Actuation j Relays I N.A.

N.A. N.A. N.A. 3 R H c Steam Line Pressure- S

. Negative: Rate-High }

d. Containment Pressure - S R H N.A. N.A. N.A. N.A. 1,2,3 Sh

->2 High-2 7I

e. Compensated Steam Line S R H N.A. N.A. N.A. N.A. 1,2,3 %k bb:

Pressure-Low 44N k' '

M N.A. N.A. N.A. N.A. 1,2,3 S R

f. Compensated TCOLD I Low-Low (interlocked ,!

with P-15)

I

?Y.

ATTACHMENT /

+ ~

+ . ST HL AE- 2232

~

PAGE I?OF &T

~ '

n' n' R N

- a.' -- .__

~ .

f 4 x- $ -

~

i e 3

4 2 4 .i '

i Y T ')

l C 4 4 '

A'! -

lif M -i ,

1 d i Y

= y I

b.

% R y I k I U w

k N d e

- . .n

? $ al w

= w i 4 w nWa

.C I l

. i s u+

dN l

%5 .-

k.,.r j..' ..

Ige f y di Ja ee e 2 9 d .

~'

( %. . 'A vi TABLE 4.3-2 (Continued)

O ,

b I' ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION

- SURVEILLANCE REQUIREMENTS "x, '

5; DIGITAL OR TRIP

. ANALOG ACTUATING .

MODES e

c . CHANNEL DEVICE MASTER SLAVE FOR WHICH

~ '

CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY . RELAY, SURVEILLANCE 55 CHANNE'L .*

CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST'i IS REQUIRED FUNCTIONAL UNIT w

8. Loss of Power (Continued)

b. 4.16 kV ESF Bus N.A. R N.A. M N.A. N.A. N.A. 1,2,3,4 Undervoltane (Tolerable Degraded Voltage ,

Coincident with SI)

N.A. M N.A. N . /.. N.A. 1,2,3,4 R

c. 4.16 kV ESF Bus N.A. R Undervoltage (Sustained

" Degraded Voltage)

~s

. 9. Engineered Safety '

Features Actuation , ,

System Interlocks N.A. N.A. N.A. N.A. N.A. 1,2,3

a. Pressurizer R H Pressure, P-11 N.A. -

R H N.A. N.A. N.A. N.A. 1,2,3

b. Low-Low T,yg, P-12

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

R

d. Power Range Neutron N.A. R(2) -M(-3-)- /\ N.A. N.A. N.A. N.A.il',2,3!

mM Flux Input to /5\ .

((gir'"5, Excessive Cooldown

' Protection, P-15 gz

10. Control Room Ventilation ['

a. Manual Initiation N.A. N.A. N.A. R N.A. N.A. N.A. ,

All !

~

FINAL DRAFT

TABLE 4.3-2 (Continued) .

8 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS 5

z y DIGITAL OR TRIP ANALOG ACTUATING MODES

[$; CHANNEL DEVICE MASTER SLAVE FOR WHICH CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY, SURVEILLANCE

, CHANNEL .

FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST, IS REQUIRED c- ,

z '

Q 11. FHB HVAC (Continued)

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

d. Spent Fuel Pool S R M N.A. N.A. N.A. N.A. With Exhaust Radio- irradiated fuel in activity-High spent fuel pool.

m u) >

$ TABLE NOTATION $p

'd

$ (1) Each train shall be tested at least every 62 days on a STAGGERED TEST BASIS. Mg (2) Neutron detectors may be excluded from CHANNEL CALIBRATION.

Mby r

(3) IWith Rated Thermal Power greater than or equal to the P-15 interlock setpoint, the ANALOG CHANNE @e

_p --e OPERATIONAL TEST shall consist of verifying that the P-15 interlock is in the required state by P 5 Lobservino the permissive annunicator windowf (4) Except rel e s K807, K814, K829 (Train B only), K831, K845, K852 and K854 (Trains B and C only) which shall be tested at least once per 18 months during refueling and during each COLD SHUTDOWN exceeding 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months unless they have been tested within the previous 92 days. '"T' l (5) Except relay K815 which shall be tested at indicated interval only when reactor coolant pressure g is above 700 psig.

(6) Each actuation train shall be tested at least every 92 days on a STAGGERED TEST BASIS. Testing of y each actuation train shall include master relay testing of both logic trains. If an ESFAS instru-mentation channel is inoperable due to failure of the Actuation Logic Test and/or Master Relay Test, g increase the surveillance frequency such that each train is tested at least every 62 days on a STAGGERED TEST BASIS unless the failure can be determined by performance of an engineering evaluation to be a single random failure.

During CORE ALTERATIONS or movement of irradiated fuel within containment.

~

i TABLE 3.3-9 (Continued)

REMOTE SHUTDOWN SYSTEM

=

TOTAL NO. MINIMUM g TRANSFER SWITCH CONTROLS OF CHANNELS g TRANSFER SWITCHES AND LOCATIONS LOCATIONS CHANNELS OPERABLE v' ASSOCIATED CONTROLS ,

^

g, c

ZLP-709 (Train C) ,'

12. EAB HVAC Fans ZLP-709 (Train C)

S ZLP-655 (Train C- ZLP-655 (Train C-g (Continued) Battery Room Battery Room and Electrical Penetration and Electrical Space fans) Penetration Space Fans)

13. Ructor Containment Fan ZLP-700 (Train A) ZLP-700 (Train A) 3 *** 2 -X.%E 6 Coolers ZLP-701 (Train B) ZLP-701 (Train B) m ZLP-709 (Train C) g ZLP-709 (Train C)

Y

ASP - Auxiliary Shutdown Panel

- QDPS - Qualified Display Processing System

Must be in the same OPERABLE RCS loop / secondary loop.

1. A total of 50 thermocouples are provided with 25 thermocouples on each of two trains. Quadrants B and D have 6 thermocouples per train each. Quadrants A and C each have 6 thermocouples on one train and 7 thermocouples on the other train. The provisions of ACTION b. are not applicable as long as each quadrant has 4 thermocouples per train OPERABLE. H G**t Ch cbn M q bye. % ene_ RCS C OPERAGLC , WMe. k Cbtr C-00t

] e. ol R< Y CS O D h . Z m

) p r== p$g g

o -

--i

-5s p

~.

TABLE 4.3-9 h RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS 5 ANALOG OR DIGITAL

.N CHANNEL MODES FOR WHICH R CHANNEL SOURCE CHANNEL OPERATIONAL SURVEILLAN(E.

CHECK CHECK CALIBRATION TEST IS REQUIRED INSTRUMENT ,

U 1. GASEOUS WASTE PROCESSING SYSTEM

Explosive Gas Monitoring System N.A. Q(5) M Oxygen Monitor (Process) D

2. Condenser Evacuation System

a. Condenser Air Removal System Discharge Header Noble Gas

Activity Monitor D M R(3) Q(2) m 1 N.A. R -Q S N. q,

q m b. Flow Rate Monitor D O c. Sampler Flow Rate Monitor D N.A. R 'Q

3. Unit Vent

a. Noble Gas Activity Monitor D M R(3) Q(2)

b. Iodine Monitor or ..- D M R(3) Q(2)

Iodine Sampler W N.A. N.A. N.A. l 1

us >

c. Particulate Monitor or D M R(3) Q(2)

Particulate Sampler W N.A. N.A. N.A. @$

b

d. Flow Rate Monitor 'D N.A. R ___Q n. g ,4, 9

pq g 6 -4 j e. Sampler Flow Rate Monitor D N.A. R Q 4 x.

seen

INSTRUMENTATION FINAL DRAFT ATTACHMENT j 3/4.3.4 TURBINE OVERSPEED PROTECTION ST-HL.AE MM p(v' .

PAGE zt. OF 'M LIMITING CONDITION FOR OPERATION 3.3.4 At least one Turbine Overspeed Protection System shall be OPERABLE.

APPICABILITY: MODES 1, 2, and 3.

^

ACTION:

a. With one stop valve or one governor valve per high pressure turbine steam line inoperable and/or with one reheat stop valve or one reheat intercept valve per low pressure turbine steam line inoperable, restore the inoperable valve (s) to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , t or close at least one' valve in the affected steam line(s) or isolate the turbine from the steam supply within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b. With the above required Turbine Overspeed Protection System otherwise inoperable, within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ' isolate the turbine from the steam supply. _

SURVEILLANCE REQUIREMENTS 4.3.4.1 The provisions of Specification 4.0.4-are not applicable.

% 4.3.4.2 The above required Turbine Overspeed Protection System shall be

demonstrated OPERABLE:

a. At least once per 7 days by cycling each of the following valves through at least one complete cycle from the running position:

1) Four high pressure turbine stop valves,

2) Four high pressure turbine governor valves, sfx l 3) 4eur low pressure turbine reheat stop valves, and

4) b low pressure turbine reheat intercept valves.

I

b. At least once per 31 days by direct observation of the movement of each of the above valves through one complete cycle from the running positign, 1 . -

l c. At least.once per 18 months by performance of a CHANNEL CALIBRATION i

on the Turbine Overspeed Protection Systems, and

d. At least once per 40 months by disassembling at least one of each of the above valves and performing a visual and surface inspection of valve seats, disks, and stems and verifying no unaccentable flaws or

, excessive corrosion. If una".:eptable flaws or excessive corrosion are found, all other valves of that type shall be inspectad.

O iv l

l SOUTH TEXAS - UNIT 1 3/4 3-89

REACTOR COOLANT SYSTEM FNAL HAFT ATTACHMENT /

HOT STANDBY .

ST R AE M

,. _ PAGE n OF 7f/

LIMITING CONDITION FOR OPERATION 3.4.1.2 At least two of the reactor coolant loops listed below shall be OPERABLE with valva CV0;i5 lecked close @ and with two reactor coolant loops in operation when the Reactor Trip System breakers are closed and one reactor coolant loop in operation when the Reactor Trip System breakers are open:"

a. Reactor Coolant Loop A and its associated steam generator and reactor coolant pump,.

b. Reactor Coolant Loop B and its associated steam generator and reactor coolant pump,

c. Reactor Coolant Loop C and its associated steam generator and reactor coolant pump, and

d. Reactor Coolant Loop D and its associated steam generator and reactor coolant pump.

APPLICABILITY: MODE 3.

ACTION:

a. With less than the above required reactor coolant loops OPERABLE, restore the required loops to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in HOT SHUT 00WN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b. With only one reactor coolant loop in operation and the Reactor Trip System breakers in the closed position, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months open the Reactor Trip System breakers.

c. With no reactor coolant loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required reactor coolant loop to operation.

SURVEILLANCE REOUIREMENTS 4.4.1.2.1 At least the above required reactor coolant pumps, if not in operation, shall be determined OPERABLE once per 7 days by verifying correct breaker. alignments and indicated power availability.

4.4.1.2.2 The required steam generators shall be determined OPERABLE by verifying secondary side water level to be greater than or equal to 10% narrow range at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

4.4.1.2.3 The 'equired reactor coolant loops with valve CV021S / luded du:>e shall be verified in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

All reactor coolant pumps may be deenergized for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months provided:

(1) no operations are permitted that would cause dilution of the Reactor Coolant System boron concentration, and (2) corr. outlet temperature is I maintained at least 10*F below saturation temperature. ;

SOUTH TEXAS - UNIT 1 3/4 4-2

REACTOR COOLANT SYSTEM FINAL DRAFT ATTACHMENT /

COLD SHUTDOWN - LOOPS NOT FILLED . ST HL AE #SSA

. ._P. AGE 2K OF 39 LIMITING CONDITION FOR OPERATION 3.4.1.4.2 At least two residual heat removal (RHR) loops shall be OPERABLE

and at least one RHR loop shall be in operation.**

APPLICABILITY: MODE 5 with reactor coolant loops not filled.

ACTION:

a. With less than the above required RHR loops OPERABLE, immediately initiate corrective action to return the required RHR loops to OPERABLE status as soon as possible.

b. With no RHR loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required RHR loop to operation.

SURVEILLANCE REOUIREMENTS 4.4.1.4.2.1 At least one RHR loop shall be determined to be in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

4.4.1.4.2.2 Valves FCV-1108, FCV-111B, CV0201A, and CV0221 shall be verified closed and secured in position by mechanical stops or removal of air or electrical power at least once per 31 days.

e a

Two RHR loops may be inoperable for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing provided the other RHR loop is OPERABLE and in operation.

- The RHR pump may be deenergized for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months provided: (1) no opera-tions are permitted that would cause dilution of the Reactor Coolant System boron concentration, and (2) core outlet temperature is maintained at least 10*F below saturation temperature.

SOUTH TEXAS - UNIT 1 3/4 4-6

E ATTACHMENT /

ST.HL.AE 237 --

[ PAGE 7s0F 3 'l EMERGENCY CORE COOLING SYSTEMS FINAL DRAFT

[-- . ~

3/4.5.6 REUDUAL HEAT REMOVAL (RHR) SYSTEM LIMITING CONDITION FOR OPERATION 3.5.6 Three independent Residual Heat Removal (RHR) loops shall be OPERABLE with each loop comprised of:

a. One OPERABLE RHR pump,

b. .One OPERABLE RHR heat exchanger, and

c. One OPERABLE flowpath capable of taking suction from its associated RCS hot leg and dischargir.g to its associated RCS cold leg.* 9 APPLICABILITY: MODES 1, 2 and 3.

ACTION:

a. With one RHR loop inoperable, restore the required loop to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b. With two RHR loops inoperable, restore at least two RHR loops 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 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 .

c. With three RHR loops inoperable, immediately initiate corrective action to restore at least one RHR loop to OPERABLE status as soon as possible. -

SURVEILLANCE REQUIREMENTS 4.5.6.1 hch RHR loop shall be demonstrated OPERABLE pursuant to the require-l ments of Specification 4.0.5.

4.5'6.2 Atleastonceper18monthsb[

fe.rifying automatic isolation and interlock action of the RHR system from-the Reactor Coolant System to ensure that:

a. With a simulated or actual Reactor Coolant Systes pressure signal greater than or equal to 350 psig, the interlocks prevent the valves from being opened, and

b. With a simulated or actual Reactor Coolant System pressure signal less than or equal to 700 psig, the interlocks will cause the valves to automatically close.

Zg b d C a.2 mW- u , ,1, e d c, yg

' T t+ w nw suupeu SOUTH TEXAS - UNIT 1 3/4 5-10

CONTAINMENT SYSTEMS FINAL HAFT ATTACHMENT '

CONTAINMENT 3/ENTILATION SYSTEM . ST.HL AE 223#-

PAGE w0F W

[ f LNG CONDITIOM FOR OPERATION 3.6.1.7 Each containment purge supply and exhaust isolation valve shall be OPERABLE and:

a. Each 48-inch containment shutdown purge supply and exhaust isola-tion valve shall be closed and sealed closed, and

b. The 18-inch supplementary containment purge supply and exhaust isolation valves shall be sealed &losed to the maximum extent practi- 30 cable but may be open for supplementary purge system operation for pressure control, for ALARA and respirable air quality considerations for personnel entry and for surveillance tests that require the valves to be open.

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

ACTION:

a. With a 48-inch containment purge supply and/or exhaust isolation valve open or not sealed closed, close and/or seal close that valve or isolate the penetration (s) within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , 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 .

b. With the 18-inch supplementary containment purge supply and/or exhaust isolation valve (s) open for reasons other than given in Specification 'l.6.1.7.b. above, close the open 18-inch valve (s) or isolate the penetration (s) within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , 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 .

c. With a containment purge supply and/or exhaust isolation valve (s) having a measured leakage rate in excess of the limits of Specifi-cations 4.6.1.7.2 and/or 4.6.1.7.3, restore the inoperable valve (s) to OPERABLE status or isolate the penetrations so that the measured leakage rate does not exceed the limits of Specifications 4.6.1.7.2

-an'd/or 4.6.1.7.3 within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , othentise 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 SHlITDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SOUTH TEXAS - UNIT 1 3/4 6-12

R I

ATTACHMENT

. ST.HL AE AM A J CONTAINMENT SYSTEMS PAGE 200F 39 SURVEILLANCE'RE0DIREMENTS 4.6.1.7.1 Each 48-inch co ainment purge supply and exhaust isolation valve shall be verified to be closedanpciosedatleastonceper31 days.

4.6.1.7.2 At least once per 6 months on a STAGGERED TEST BASIS, the inboard and outboard isolation valves with resilient material seals in each sealed closed 48-inch containment purge supply and exhaust penetration shall be demonstrated OPERABLE by verifying that the mea:ured leakage rate is less than 0.05 L, when pressurized to P,.

4.6.1.7.3 At least once per 3 months each 18-inch supplementary containment purge supply and exhaust isolation valve with resilient material seals shall be demonstrated OPERABLE by verifying that the measured leakage rate is less than 0.01 L, when pressurized to Pa*

4.6.1.7.4 At least once per 31 days each 18-inch supplementary contaigent purge supply and exhaust isolation valve shall be verified to be e n L closed RO or open in accordance with Specification 3.6.1.7.b.

SOUTH TEXAS - UNIT 1 3/4 6-13

FINAL DRAFT CONTAINMENT SYSTEMS ATTACHMENT /

hR

~

. ST HL-AE 2239 SURVEILLANCERf00kiEMENTS(Continuedi TGE 7AOF M-4.6.3.2 Each' isolation valve -speci'ied r. Tab 1I3.5 hall be' demonstrated OPERABLE during the COLD SHUTDOWN or REFUELING MODE at least once per 18 months by: ,

a. Verifying that on a Phase "A" Isolation test signal, each I #

thase "A". isolation valve actuates to its isolation position,

b. Verifying that on a Containment Ventilation Isolation test signal, each purge and exhaust valve actuates to its isolation positione; a.nd ,

4.6.3.3 The isolation time of each power-operated or automatic valve shall be determined to be within its limit when tested pursuant to Specification 4.0.5. -

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+h + a ~ Ph= " S" cam P ase "B" isola-6m vaJvt

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SOUTH TEXAS - UNIT 1 3/4 6-19 l

ATTACHMENT l

. ST.HL AE- 2 232--

"EA PLANT SYSTEMS 3/4.7.3 COMPONENT COOLING WATER SYSTEM LIMITING CONDITION FOR OPERATION

'Thn, 3.7.3 At kahthree independent component cooling water loops shall be OPERABLE.

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

ACTION:

With only two component cooling water loops OPERABLE, restore at least three loops to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REOUIREMENTS Eo.ch 4.7.3 atka5tthre2componentcoolingwaterloop/shallbedemonstratedOPERABLE:

3

a. At least once per 31 days by verifying that each valve outside con-tainment (manual, power-operated, or automatic) servicing safety-related equipment that is not locked, sealed, or otherwise secured in position is in its correct position; and

.b b. At least once per 18 months during shutdown, by verifying that:

1 1) Each automatic valve servicing safety-related equipment or isolating the non-nuclear safety portion of the system actuates to its correct position on a Safety Injection, Loss of Offsite c i dW&ngor low Surge Tank test signal, as applicable,

2) Each Component Cooling Water System pump starts automatically onaSafetyInjectionorLossofOffsitePowertestsignal,and t

3) The surge tank level instrumentation which provides automatic isolation of the non nuckfsafety-related portion of the sys-tem is d:=: tat:d OPERABLE by performance of a CHANNEL CALIBRATION test. A

c. By veri.fying that each valve inside containment (manual, power-operated, or automatic) servicing safety-related equipment that is not locked, sealed, or otherwise secured in position is in its cor-rect position prior to entering MODE 4 following each COLD SHUTDOWN of greater than 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months if not performed within the previous 31 days.

SOUTH TEXAS - UNIT 1 3/4 7-12

ATTACHMENT i

. ST.HL AE 238 Fhrn?datlWim AL UKAtl n

() ,. _ TABLE 4.8-2 BATTERY SURVEILLANCE REQUIREMENTS CATEGORY A ll) CATEGORY B(2)

PARAMETER LIMITS FOR EACH LIMITS FOR EACH ALLOWABLEC3)

DESIGNATED PILOT . CONNECTED CELL VALUE FOR EACH CELL CONNECTED CELL Electrolyte > Minimum level > Minimum level Above top of Level indication mark, indication mark, plates, and < 1/4" above and < 1/4" above and not maximum level maximum level overflowing indication mark indication mark Float Voltage > 2.13 volts >2.13voltsf > 2.07 volts k Not more than -

b 0.020 below the average of all Specific > 1.200 M > 1.195 connected cells Gravity (4)

Average of all Average of all kq j connected cells connected cells

> 1.205 >1.195h i TABLE h0TATIONS (1) For any Category A parameter (s) outside the limit (s) shown, the battery may be considered OPERABLE provided that within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 'all Category B measurements are taken and found to be within their allowable' values, and provided all Category A and B parameter (s) are restored to within limits within the next 6 days.

(2) For any Category B parameter (s) outside the limit (s) shown, the battery may be considered OPERABLE provided that the Category B parameters are within.their allowable values and provided that Category'B parameters (s) are restored to within limits within 7 days.

(3) Any Category 8 parameter not within its allowable value indicates an inoperable battery.

(4) Corrected for electrolyte temperature and level.

(5) Or battery charging current is less than 2 ampi, when on charge.

(6) Corrected for average electrolyte temperature.

SOUTH TEXAS - UNIT 1 3/4 8-12

ATTACHMENT '

. ST.HL AE ##Y PAGEJLOF 39-FINAL DRAFT ELECTRICAL POWER SYSTEMS ONSITE POWER DISTRIBUTION SHUT 00WN LIMITING CONDITION FOR OPERATION

,3.8.3.2 As a minimum, the following electrical busses shall be energized in the specified manner:

Two

a. Either Tr:fr A cr Tr:t- C% f A.C. ESF W : censisting ;f en[4160-volt ESF buscandvtwof80-volt A.C. ESF load centers, 5.muhd * **h d

b. -Twe4120-volt A.C. vital distribution panels :=i: ting of either DFOGi and OP1201, u. DFG02 anu DP1200 energized from their associated inverter connected to its respective D.C. bus, and

% h suecat46

c. Either Channal inrChanna1M125-voltD.C. bus [3 energized from W A ir associated battery banks.

APPLICABILITY MODES 5 and 6.

ACTION:

With any of the above required electrical busses not energized in the required manner, immediately suspend all operations involving CORE ALTERATIONS, positive reactivity changes, or movement of irradiated fuel, initiate corrective action to energize the required electrical busses in the specified manner as soon as possible, and within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , depressurize and vent the RCS through at least

! a 2.0 square inch vent.

I SURVEILLANCE REOUIREMENTS 4.8.3.2 The specified busses shall be detennined energized in the required manner at least once per 7 days by verifying correct breaker alignment and indicated voltage on the busses.

l SOUTH TEXAS - UNIT 1 3/4 8-16

\

ATTACHMENT /

. ST HL AE #23 A --l

, PAGE30F 39 3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION LIMITING CONDITION FOR OPERATION 3.9.1 The boron concentration of all filled portions of the Reactor Coolant System and the refueling canal shall be maintained uniform and sufficient to ensure that the more restrictive of the following reactivity conditions is met; either:

a. A K,ff of 0.95 or less, or

b. A boron concentration of greater than or equal to 2500 ppm.

APPLICABILITY: MODE 6.*

ACTION:

With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS or positive reactivity changes and 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 its equivalent until K,ff is reduced to less than or equal to 0.95 or the boron concentration is restored to greater than or equal to 2500 ppa, whichever is the more restrictive.

SURVEILLANCE REOUIREMENTS 4.9.1.1 The more restrictive of the above two reactiv,ity conditions shall be determined prior to:

a. Removing or unbolting the reactor vessel head, and

b. Withdrawal of any full-length control rod in excess of 3 feet from its fully inserted position within the reactor vessel.

4.9.1.2 The boron concentration of the Reactor Coolant System and the refueling canal shall be d,etermined by chemical analysis at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

4.9.1.3 Val'ves FCV-1108, FCV-111B, CV0201A, and CV0221 shall be g l verified closed and secured in position by mechanical stops or by removal of air or electrical power at least once per 31 days.

The reactor shall be maintained in MODE 6 whenever fuel is in the reactor vessel with the vessel head closure bolts less than fully tensioned or with the head removed.

SOUTH TEXAS - UNIT 1 3/4 9-1

.l ATTACHMENT /

. ST.HL.AE--L7.0 I nd3.DF.21_ i 3/4.1 REACTIVITY CONTROL SYSTEMS n

I fLASEs 3/4.1.1 B0 RATION CONTROL 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 suberitical from all operating conditions, (2) the reactivity transients asso- l ciated 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.

SHUTOOWN MARGIN requiremerits vary throughout core life as a function of fuel depletion,.RCS boron concentration, an1 RCS T,yg. In MODES 1 and 2, the most restrictive condition occurs at E0L, with T,yg at no load operating tecperature, and is associated with a postulated steam line break. accident and i resulting uncontrolled RCS cooldown. In the analysis of this accident, a e mirimum SHUTDOWN MARGIN of 1.75% ak/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 u~ sing Hafnium control rods (Ref. FSAR Table 4.3-3). A :xdin;;1y,'t. -

the ""T007: ".ARCIM7cquirement '^r #0ESir.d ? is heed "per thi: 'i=iti n w.e L un &Jri a consistent with i:AR 5;fety walysi;mastiud In MODES g i 3, 4, and 5, the most restrictive condition occurs at BOL, when the boron concen-tration is the greatest. In these modes, the required SHUTDOWN MARGIN is ,

composed of a constant requirement and a variable requirement, which is a ^

e 5

function of the RCS boron concentration. Th: ;wat;nt ::"T00' 7; "ARGU; . mqu h e- -

c.taf 1.25 AE/k i; based en an un cntrolled RCS coeldewn f.e o 5 t;;nlin; ;.,

b ei eccident, es .es-thm sesc for "00:0 1 ond 2.".- The variable SHUTDOWN 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.

Jhe 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 cociant pump was operating. If at least one reactor coolant pump is not operat-ing in MODE 3 or 4, then the SHUTDOWN MARGIN requirements for MODE 5 shall apply.

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; accordingly, verification of MTC values at conditions other than those explicitly stated will require extrapolation to those conditions in orcer to permit an accurate comparison.

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 analyses to nominal operating' conditions. These corrections involved subtracting the incremental change in the MDC associated with a core 50tJTH TEXAS - UNIT 1 B 3/4 1-1

REACTIVITY CONTROL SYSTEMS FINAL HAFT ATTACHMENT /

(~) BASES ST HL AE-M3A rmac. w vi y B0 RATION SYSTEMS (Continued) 1 The boron capability required below 200*F is sufficient to provide a variable SHUTOOWN MARGIN based on the results of a boron dilution accident analysis where the SHUTOOWN MARGIN is varied as a function of RCS boron concentration after xenon decay and cooldown from 200*F to 140*F. This condition requires either 2900 gallons of 7000 ppm borated water from the boric acid storage system or 122,000 gallons of 2500 ppm borated water from the RWST for MODE 5 and 33,000 gallons of 2500 ppm borated water from the RWST for MODE 6.

The contained water volume limits incluce allowance for water not available because of discharge line location and other physical characteristics.

The limits on contained water volume and boron concentration of the RWST also ensure a pH value of between 7.5 and 10.0 for the solution recirculated within containment after a LOCA. This pH band minimizes the evolution of iodine and minimizes the effect of chloride and caustic stress corrosion on mechanical systems and components. -

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

3/4.1.3 MOVABLE CONTROL ASSEMBLIES The specifications of this section ensure that: (1) acceptable power distri-bution limits are maintained, (2) the minimum SHUTDOWN MARGIN is maintained, and (3) the potential effects of rod misalignment on associated accident analyses are limited. OPERABILITY of the control rod position indicators is required to determine control rod positions and thereby ensure compliance with the control ;

rod alignment and insertion limits. Verification that the Digital Rod Position Indicator agrees with the demanded position within i 12 steps at 24, 48, 120, /\

and 259 steps withdrawn for the Control Banks and 18, g f and 259 steps with- /_1L\

drawn for the Shutdown Banks provides assurances that N Digital Rod Position Indicator is operating correctly over the full range of indication. Since the Digital Rod Position Indicatio stem does not indicate the actual shutdown rod position between 18 steps and steps, only points in the indicated ranges I are picked for verification of agreement with demanded position.

, The ACTION statements which permit limited variations from the basic l

requirements'are accompanied by additional restrictions which ensure that the original design criteria are met. Misalignment of a rod requires measurement of peaking factors and a restriction in THERMAL POWER. These restrictions pro-t vide assurance of fuel rod integrity during continued operation. In addition, those safety analyses affected by a misaligned rod are reevaluated to confirm that the results remain valid during future operation.

i The maximum rod drop time restriction is consistent with the assumed rod drop time used in the safety analyses. Measurement with T avg greater than or l

equal to 561*F and with all reactor coolant pumps operating ensures that the measured drop times will be representative of insertion times experienced during a Reactor trip at operating conditions.

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

f ATTACHMENT /

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, FAST NEUTRON FLUENCE (E>1MeV) AS A FUNCTION OF FULL POWER SERVICE LIFE l

SOUTH TEXAS - UNIT 1 B 3/4 4-10 l

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. ST.HL.AE 4234 PAGEStjoF39 l

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SOUTH TEXAS - UNIT 1 8 3/4 4-11

TTACHMENT /

. ST HL AE J7332-3/4.7 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).

co. c r g The specified valve lift settings and relieving 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 x 108 lbs/h which is 122% of the total secondary steam flow of . x 106 lbs at 100% RATED THERMAL POWER. A minimum of two OPERABLE safety valves per steam generator ensures that sufficient relieving capacity is S . y 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 reduction 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 V = Maximum number of inoperable safety valves per steam line, 109 = Power Range Neutron Flux-High Trip Setpoint for four icop operation, X '= Total relieving capacity of all safety valves per steam line in lbs/ hour, and Y = Maximum relieving capacity of any one safety valve in lbs/ hour SOUTH TEXAS - UNIT 1 B 3/4 7-1 l

ATTACHMEgT /,

ilNWANRDAET InhL usuu I ACC NISTRATIVE CONTROLS 6.4 TRAINING -

6.4.1 A retraining and replacement training program for the unit staff shall be r.aintained under the direction of the Training Manager and shall meet or exceed the requirements and recommendations of Section 5.5 of ANSI N18.1-1971 anc Appendix A of 10 CFR Part 55 and the supplemental requirements specified in Sec. ions A and C of Enclosure 1 of the March 28, 1980 NRC letter to all licensees, anc shall include familiarization with relevant industry operational experience.

6.5 REVIEW AND AUDIT 6.5.1 PLANT OPERATIONS REVIEW COMMITTEE (PORC)

FUN: TION 6.5.1.1 The PORC shall function to advise the Plant Manager on all matters related to nuclear safety.

COM30SITION 6.5.1.2 The PORC shall be composed of the:

Member: Plant Eper"tentnt h d h"M Ok 6 Member: ^cocte. Cpe,atic : ":n::^"ewa Opo h bra cf(

Member: -TeJo.. cal E;';' rt " n:; puA eyA (Yb~ge r Member: Maintenance Manager Member: Ch;. ice'. Cec,c a ns anu A..;13 m .tn:; r Member: heLu ...d Caf;;; i .;ees Men:; n Member: Operations QA Manager The PORC Chairman shall be appointed in writing from among these members by the Plant Manager.

ALTERNATES 6.5.1.3 All alternate members shall be appointed in writing by the Plant Manager to serve on a temporary basis; however, no more than two alternates sha'l participate as voting members in PORC activities at any one time.

MEETING FREQUENCY 6.5.1.4 The PORC shall meet at least once per calendar month and as convened by the PORC Chairman or his designated alternate.

QUORUM 6.5.1.5 The quorum of the PORC necessary for the performance of the PORC responsibility and authority provisions of these Technical Specifications shall consist of the Chairman or his designated alternate, the Operations QA Marager or his designated alternate, and e ree other members including alternates. -tee 50 LITH TEXAS - UNIT 1 6-7

Attachment 2 ST-HL-AE-2232 Page 1 of 2 FINAL DRAFT TECHNICAL SPECIFICATIONS SPECIFICATION 3.4.6.1 ACTION As in icated in Final Draft Technical Specification 3.4.6.1, the RCS leakage detection systems at STP are the following:

a) Containment Atmosphere Radiation Monitor - Noble Gas Channel, b) Containment Normal Sump Level and Flow Monitoring System, and c) Containment Atmosphere Radiation Monitor - Particulate Channel.

The Containment Atmosphere Radiation Monitor has three channels: noble gas, iodine and particulate, all on the same skid, with one sample pump, one microprocessor and the associated tubing, valves and instrumentation It is anticipated that performance of the DJ' 41 Channel Operational Test (required by Specification 4.3.3.1) will e .pproximately 2 to 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months .

Therefore, performance of the Digital Channel ational Test places the plant in Technical Specification 3.0.3 and dc .ot provide sufficient time for correction of any problems found during the test. The net result will be a rapid shutdown of the plant in lieu of an opportunity to correct the problem.

If any of the common equipment (such as the sample pump) becomes inoperable, items (a) and (c) would be inoperable. Again, in this instance, the plant would be required to be Technical Specification 3.0.3.

The sample pump manufacturer recommends pump replacement at 1-1/2 year intervals. Replacement of a sample pump takes approximately 2 to 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months to accomplish. With the monitor located outside the containment, drawing its sample from inside through containment isolation valves, closure of the containment isolation valves can cause stresses to be placed on the sample pump and other monitor components.

Based on the above considerations, and also considering the importance of monitoring for RCS leakage, STP believes that an appropriate action when both channels are inoperable is as follows:

With (a) and (c) of the above required Leakage Detection Systems inoperable:

i

1) Restore either monitoring System ((a) or (c)) to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , and i

2) Obtain and analyze a grab sample of the containment atmosphere for gaseous and particulate radioactivity at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , and

3) Perform a Reactor Coolant System water inventory balance at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

L3/NRC/eq l

Attachment 2 ST-HL-AE-2232 Page 2 of 2 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 .

The markup below provides the actions that STP believes are appropriate.

Actions (a) and (b) are identical to the Action presented in the Final Draft Technical Specifications. Action (c) provides an appropriate repair time for common equipment on the containment atmosphere radiation monitor skid and still provides appropriate measures to detect RCPB leakage during the repair time.

HL&P has reviewed Technical Specifications and FSARs of other plants to identify those with a similar design for RCS leakage instrumentation.

Recently licensed plants with similar designs include Byron and Shearon Harris. The attached markup reflects the Byron Technical Specification wording; the Shearon Harris Technical Specification wording is equivalent.

REVISED ACTION SPECIFICATION 3.4.6.1

a. With (a) or (c) of the above required Leakage Detection Systems inoperable, operation may continue for up to 30 days provided grab samples of the containment atmosphere are obtained and analyzed for gaseous and particulate radioactivity at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when the required Caseous or Particulate Radioactivity Monitoring System is inoperable; 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 .

b. With (b) of the above required Leakage Detection Systems inoperable 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 ,

c. With (a) and (c) of the above required Leakage Detection Systems inoperable:

1) Restore either Monitoring System ((a) or (c)) to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months and

2) Obtain and analyze a grab sample of the containment atmosphere for gaseous and particulate radioactivity at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , and

3) Perform a Reactor Coolant System water inventory balance at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

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 .

L3/NRC/eq

Attachment 3 ST-HL-AE-2232 Page 1 of 4 FINAL DRAFT TECHNICAL SPECIFICATIONS SPECIFICATION 3/4.3.3.1 Submittals made by HL&P for Technical Specification 3/4.3.3.1, Radiation Monitoring for Plant Operations, have consistently shown Item 1.a (Containment Atmosphere Radioactivity - High) as deleted, not applicable for the STP design.

The Standard Technical Specification combines entries of 2 channels minimum operable (with 1 required to trip / alarm), the alarm setpoint being shown in mR/hr and the action statement saying that with less than the minimum operable requirement, operation may continue provided the containment purge valves are closed. The Standard Technical Specification written in this manner implies that two area monitors are required operable and that either one of these monitors reading above the setpoint results in automatic containment purgo valve closure.

The STP design for containment purge valve automatic closure has been the subject of extensive review, with the design modifications and final design being presented in HL&P letter to NRC ST-HL.AE-2141, dated May 8, 1987. It is our understanding that this design is acceptable to the staff. This final design does not include atmosphere radiation monitor input to the containment ventilation isolation signal. Thus Item 1.a has been shown as deleted and not applicable to the STP design in our submittals.

The Final Draft Technical Specifications Table 3.3-6 requirements show that a radiation monitor for containment atmosphere radioactivity shall be operable in all modes, and if the monitor is inoperable, operation may continue for up to 30 days provided containment atmosphere samples are taken and analyzed every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . The concerns that STP has about these requirements are discussed below.

The containment atmosphere monitor has three channels - noble gas, particulate and iodine. Item 1.a of Table 3.3-6 does not indicate which channel (s) are required by this specification. The noble gas and particulate channels are required by Items 1.b.1) and 2). It is believed, based upon discussions with the NRC staff (see below), that the intent of this requirement would be satisfied by use of the noble gas channel.

The Technical Specification is applicable in all Modes. The containment atmosphere monitor particulate and noble gas channels are required operable in modes 1 through 4 for RCS leakage detection. The action required when the monitor is inoperable in these modes is identical to that required by Specification 3.4.6.1, and thus presents no particular concern. However, the requirement for operability in all modes adds the requirement in Modes 5 and 6, which are considered to be non operational modes. Thus, when the monitor is inoperable in Mode 5 or 6, the statement that operation may continue for up to 30 days provided ... would seem to be a non requirement. We are at present unable to determine what the action statement actually requires in Modes 5 and 6.

L3/NRC/eq

Attachment 3 ST-llL AE 2232 Page 2 of 4 The concerns presented by the NRC staff during a telephone conference on April 29, 1987, regarding operability of the containment atmosphere monitor, included monitoring to establish that concentrations are acceptable for initiation of purging and monitoring for personnel protection.

Based upon the concerns identified in the interpretation cf the Final Draft Technical Specification requirements noted above, the existing Technical Specification requirement to either use the Table 3.3 4 ESEAS aetpoint for the RCB purge radiation monitors or to establish an appropriate setpoint based on the ODCM methodology, which would require knowledge of the current containment atmosphere concentrations, and the requirement for plant procedures to address 4 monitoring for personnel radiation protection, llL&P again requests that Item

, 1.a be deleted. Should this request not be granted, llL&P then requests a clarification to the table to address the above concerns.

f J

l l

i L3/NRC/eq

1 '

l

( ) ( ,

v v v

. TABLE 3.3-6 h RADIATION MONITORING INSTRUMENTATION FOR PLANT OPERATIONS, 2

-. ~

MINIMUM

. CHANNELS CHANNELS APPLICABLE ALARM / TRIP

-0 FUNCTIONAL UNIT -, TO TRIP / ALARM OPERABLE MODES SETPOINT ACTION %.

c .'

5-e

1. Contain=ent -

" N.A. 31

a. Containment Atmosphere N.A 1 All Radioactivity-High '

b. RCS Leakage Detection

1) Particulate Radioactivity N.A. 1 l'; 2, 3, 4 H . A'. 34

2) Caseous Radioactivity N.A. I 1,2,3,4 N.A. 34 g , .

~

HB m

-a 2 u>I M ;P. 3:=

c AQ k%

b

[L m~

h

+

m TABLE 4.3-3 8

y RADIATION MONITORING INSTRUMENTATION FOR PLANT

_. OPERATIONS SURVEILLANCE REQUIREMENTS O

3; DIGITAL

, CHANNEL MODES FOR WHICH c . CHANNEL CHANNEL OPERATIONAL SURVEILLANCE 6-Z;

. FUNCTIOMAL UNIT CHECK CALIBRATION TEST IS REQUIRED .

~ Containment J/ ,

A Ca. Containment Atmosphere Radioactivity-High_ S R H 11

-4

4. RCS Leakage Detection pig

r. o

1) Farticulate Radio- S R M 1,2,3,4 hM5 g activity yLje

2) Gaseous Radioactivity z " -4 y S R M 1,2,3,4 , gg w

=1

T l

Attachment 4 ST HL AE 2232 Page 1 of 3 FINAL DRAFT TECHNICAL SPECIFICATIONS SURVEILLANCE REQUIREMENT 4.5.1.1.c The design of the controls and power supply for the accumulator discharge isolation valves has followed the recommendations of Branch Technical Positions (BTP) ICSB 4 " Requirements on Motor Operated Valves in the ECCS Accumulator Lines" and ICSB 18 " Application of the Single Failure Criterion to Manually controlled Electrically Operated Valves".

l ICSB 18 sets forth the NRC position that use of power removal to valves is authorized in order to satisfy the single failure criterion. In this caso, the failuro of an accumulator discharge isolation valvo is not considered in the ECCS single failure analysis, on the basis that power is locked out to the valva motor operator. When it in desired to use power lockout to satisfy the single failure critorion, the design requirements given in BTP ICSB 18 includes the requirement that electrical power can be restored to the valves from the control room (BTP ICSB 18, Section B.3, item a).

The application of the two BTPs for the accumulator dischargo isolation valves is described in FSAR Sections 6.3.5.5.1 and 7.6.3. The logic diagram for those valves is shown on FSAR Figure 7.6-3, Sheets 1 and 2. The design has been evaluated in the South Texas Project SER, with some discussion in Section 6.3 and extensive discussion in Section 7.6.2.2. The staff has found the STP design to be acceptable, as indicated in SER Sections 6.3 and 7.6.2.2.

The hot log isolation valves also have been evaluated for purposes of meeting the single failure criterion, with the result that power lockout has been provided for those valvos as well. The power lockout for these valvos is discussed in FSAR Sections 6.3.5.5.2 and 7.6.7. The logic diagram for these valves is shown on FSAR Figures 7.6 10. The design has been evaluated in the i STP SER, in Section 6.3.2, as acceptable for meeting the single failure criterion based on power lockout.

As can be soon from referenced FSAR sections and figures, power lockout i

for the accumulator discharge isolation valves and the hot leg isolation i valvos is identical, and allows restoration of electrical power to the valves from the control room. As can also be seen on the referenced figures, position indicating lights are provided on 'the control board for the motor breaker position, for use in verifying removal of power from the valve operator, i

! Final Draf t Technical Specification Surveillance Requirement 4.5.1.1.c for the accumulator discharge isointion valves requires " verifying that power to the isolation valve operator is disconnected by removal of the breaker from the circuit." This requirement has been interpreted to mean that the MCC breaker must be physically removed, would require ro. installation for the valve oporator to be supplied power. This reinstallation must be dono at the MCC, Removal of the breaker physiently from the MCC renders the power lockout switch at the main control room useless, which raises a human factors concern.

L3/NRC/eq

Attachment 4 ST llL AE 2230 Page 2 of 3 Final Draft Technical Specification Surveillance Requirement 4.5.2.a for the ECCS hot leg isolation valves requires verifying that the "valvos are in the indicated positions with power to the valve operators removed". This requirement allows use of the system design for power removal from the valve operator, retaining the ability to repower the operator with actions strictly from the control room.

On the basis that STP has provided an accepted design for meeting single failure criterion and provisions of BTP ICSB 18 requiring power restoration capability from the control room, and the design similarity of the accumulator discharge isolation valvo and the hot log isolation valve control logic, STP believes that the surveillance requirements for the two applications in regard to power lockout should bo identical. Accordingly, we submit that Surveillance Requirement 4.5.1.1.c should read as follows:

At least once per 31 days when the RCS pressure is above 1000 psig by verifying that power to the isolation valve operator is removed.

L3/NRC/eq

FINAL DRAFT EMERGENCY CORE COOLING SYSTEMS I ATTACHMENT V b -

-~.- . ST HL AEJRSA 3RVEILLANCLREQUIRE!iEHIS JContinngdi

,_r er 5 o d

c. At least once per 31 days when the RCS pressure is above'1000 psig by verifying that power to the isolation valve operatorYe. is 44wea A--

. nest +d t,j cc v:! c f the breaker-f rc: the-eI

~ "" > moved . ,

d. At least once per 18 months by verifying that each accumulator isola-tion valve opens automatically under each of the following conditions:

1) When an actual.or a simulated RCS pressure signal exceeds the P-11 (Pressurizer Pressure Block of Safety injection) Setpoint,

.and

2) Upon receipt of a Safety Injection test signal.

4.5.1.2 Each accumulator water level and pressure channel shall be demon-strated OPERABLE:

a. At least once per 31 days by the performance of an ANALOG CHANNEL OPERATIONAL TEST, and

b. At least once per 18 months by the performance of a CHANNEL CALIBRATION.

{

4 9

9

( MW' SOUTH TEXAS - UNIT 1 3/4 5-2 6

I l l

l Attechment 5 !

I ST ilL AE-2232 Page 1 of 2 FINAL DRAFT TECllNICAL SPECIFICATION ,

SPECIFICATION 4.6.1.3.e Technical Specifications Surveillance requirement 4.6.1.3.e has been revised based on the pressure differential derived from the guaranteed minimum seal inflation pressure and the minimum required seal inflation pressure. The '

seal inflation, under normal operation, is provided by the plant air supply I system, and it is prescribed to be an inflation pressure within 90 to 95 psig subject to a maximum of 110 psig with a guaranteed minimum of 90 psig. The minimum inflation pressure required for the seals to retain the containment DBA pressure of 37.5 psig is 65 psig. Therefore, in the event of loss of plant air supply, the allowable pressure loss for the inflatable seals is 25 .

psi (90 65 psig). The prescribed acceptance criterion for pressure loss during the seal pressure decay test is set at 1.5 psi / day. That daily rate, when extended to a period of 15 days, amounts to a pressure loss of 22.5 psi which is within the allowable value of 25 pai.

The minimum required inflation pressure of 65 psig corresponding to a containment pressure of 37.5 psig has been confirmed by tests performed on both doors of the personnel airlock.

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t L3/NRC/eq i

FINAL DRAH -

CONTAINMENT SYSTEMS :TACHMENT 5

.st.AE >> M

' 4 0F L ;

SURVEILLMCE REOUIREMENTS 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 c' losing, 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 of 37.5 psig;

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

37.5 p'sig, and verifying the overall air lock leakage rate is within *

' 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, b d. By verifying at least once per 7 days that the instrument air pres-sure in the header to the personnel airlock seals is > 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 . psi (

from 90 psig within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

rmn mwN , ,.

I+ 6

$ M k *The provisions of Specification 4.0.2 are not appilcable.

- This represents an 1xemption to Appendix J, paragraph !!!.0.2 of 10 CFR Part 50.

50lJ111 TEXAS - UNIT 1 3/4 6 6

ATTACHMENT G

. ST.HL AE. n 32 PAGF.1 0F 11 l

ATTACitMENT 6 l i

EDITORIAL.S & TYP0CitArillCAL CilANCES 1.1/NitC/cq

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS f

SECTION PAGE 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX OIFFERENCE...........,,......................... 3/4 2-1 '

FIGURE 3.2-1 AXIAL FLUX DIFFERENCE LIMITS AS A FUNCTION OF RATED THERMAL P0WER...................................... 3/4 2-4 3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR - F (Z)..................... 3/4 2-5 9

FIGURE 3.2-2 K(Z) - NORMAL.IZED F (Z) q AS A FUNCTION OF CORE HEIGHT. 3/4 2-6 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 DND 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

-( REQUIREMENTS............................................. 3/4 3-11 3/4.3.2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION...........,.............................. 3/4 3-16 TABLE 3.3-3 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION.......................................... 3/4 3-1B TABLE 3.3-4 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETP0!NTS........................... 3/4 3-29 TABLE 3.3-5 [NGINEERED SAFETY FEATURES RESPONSE TIMES............. 3/4 3-37 TABLE 4.3-2 [NGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVE!LLANCE REQUIREMENTS................ 3/4 3 42 3/4.3.3 MONITORING INSTRUMENTATION Radiation Monitoring for Plant Operations................ 3/4 3-50 TABLE 3.3- RADIATION HONITORING INSTRUMENTATION

'0R PLANT OPERATIONS..................................... 3/4 3-51 TABLE 4.3-3 RADIATION HONITORING INSTRUMENTATION FOR PLANT OPERATIONS SURVEILLANCE REQUIREMENTS..................... 3/4 3-53 l Novable Incore Detectors................................. 3/4 3-54 l Seismic Instrumentation.................................. 3/4 3 55 TABLE 3.3-7 SE!SMIC HONITORING INSTRUMENTATION.................... 3/4 3 56

(

1 SOUTH TEXAS - UNIT 1 Y u -.. - - - - - . _ - - . - - - _ - - - - _ - _ -- J

ATTACHMENT 6

. ST.HL.AE 212 PAGE 3 OF 91 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 SHUTOOWN MONITORING INSTRUMENTATION SURVEILLANCEREQUIREMENTS................................ 3/4 3-66 Accident Monitoring Instrumentation...................... 3/4 3-67 TABLE 3.3-10 ACCIDENT HONITORING INSTRUMENTATION.................. 3/4 3-68 TABLE 4.3-7 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE RE W ENTS............................................. 3/4 3-73 N $r k 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 -

TABLE 3.3 12 RADIDACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION 3/4 3-80 TABLE 4.3-8 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS................ 3/4 3 82 Radioactive Gaseous Effluent MonitorinD Instrumentation.. 3/4 3 84 TABLE 3.3-13 RADI0 ACTIVE GASEOUS EFFLUENT HONITORING 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 OVERSPEE0 PROTECTION.............................. 3/4 3 89

(

500111 TEXAS - UNIT 1 vt J

w _ ..-.. - - - - -_- .. . - _ _ , - .- - - - - - - - - - - - - - - - - - - - - - - -

ATTACHMENT o

. sT.HL AE. 2m PAGE 9 OF 'll INDEX

_ LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS

{>

SECTION PAGE 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION Sta rtup and Powe r Ope ration. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 4-1 llot Standry.............................................. 3/4 4-2 Hot Shutdown............................................. 3/4 4-3 Cold Shutdown - Loops F111ed............................. 3/4 4-5 Cold Snutdown - Loops Not F111ed. . . . . . . . .'. . . . . . . . . . . . . . . . 3/4 4-6 3/4.4.2 _ SAFETY VALVES

' +-6 h u t d ow n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 4-7 4.-Operating ............................................. 3/4 4-8 3/4.4.3 P H L 5 sviiTZ - R . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 4-9 3/4.4.4 RELIEF VALVES............................................ 3/4 4-10 3/4.4.5 STEAM GENERATORS......................................... 3/4 4-12 TABLE 4.4-1 HINIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTED DURING INSERVICE INSPECT!0N............................. 3/4 4-17

. . . . . TAB LE 4. 4-2. STEAM. GENERATOR . TUDE INSPECTION... .. . . . . . . .. . ..... . . . . .. . 3/4 4-18 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems................................ 3/4 4-19 Operational Leakage...................................... 3/4 4-20 TABLE 3.4-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES...... 3/4 4-22 3/4.4.7 CilEMISTRY................................................ 3/4 4-23 TABLE 3.4 2 REACTOR COOLANT SYSTEM CllEMISTRY LIMITS. . . . . . . . . . . . . . . 3/4 4-24 TABLE 4.4 3 REACTOR COOLANT SYSTEM CilEMISTRY LIMITS SURVEILLANCE REQUIREMENTS............................................. 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 L1 HIT VERSUS PERCENT OF RATED TiiERMAL POWER WITil Tile REACTOR COOLANT SPECIf!C ACTIVITY >l pct / gram DOSE EQUIVALENT I 131.................................... 3/44-20 TABLE 4.4-4 REACTOR COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRA1.................................................. 3/4 4-29 3/4.4.9 PRESSURE /IEMPERATURE LIMITS Reactor Coolant System................................... 3/4 4-31 fA FIGURE 3.4-2 REACTOR COOLANT SYSTEM llEATUP LIMITATIONS -

APPLICABLE UP TO 32 EFPY................................. 3/4 4 32 500111 TEXAS - UNIT 1 vil

ATTACHMEN 6 se INDEX FINAL DRAFT

( LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE FIGURE 3.4-3 REACTOR COOLANT SYSTEM C00LDOWN LIMITATIONS -

APPLICABLE UP TO 32 EFPY................................. 3/4 4-33 .

TA8LE 4.4-5 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM -

WITH0RAWAL SCHE 0VLE...................................... 3/4 4-34 P r e s s u r i z e r . . . . . . . . . . . . '. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 4-35 Overpressure Protection Systems.......................... 3/4 4-36 FIGURE 3.4-4 NOMINAL MAXIMUM ALLOWABLE PORV SETPOINT FOR THE COLO OVERPRESSURE SYSTEM . . . . . . . . . . . . . . . . . . . . . . . . 3/4 4-37 3/4.4.10 STRUCTURAL INTEGRITY..................................... 3/4 4-39 3/4.4.11 REACTOR VESSEL llEAD 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,yg GREATER THAN OR EQUAL TO 350'F....3/4 5-3 3/4.5.3 ECCS SUBSYSTEMS - T avg LESS TilAN 350'F. . . . . . . . . . . . . . . . . . . 3/4 5-6 ECS Subsyst , u. -avg T w ,,,,ThanLess o5 qual to 200'F. . . . . . . 3/4 5-8 C V4. 5. 'l (m.S.w,a ems u se.

E 3/4.5.5 REFUELING WATER STORAGE TANK............................. 3/4 5-9 3/4.5.6 RESIDUAL llEAT REMOVAL (RHR) SYSTEM ...................... 3/4 5-10 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CCNTAINHENT 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 A i r Te mp e ra tu re . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 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 Spray Additive System.................................... 3/4 6-15 Containment Cooling System............................... 3/4 6 17 k

500Til TEXAS UNIT 1 vill J

ATTACHMENT G ST.HL.AE 2232 PActfg_OFH INDEX h LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.6.3 CONTAINMENT ISOLATION VALVES............................. 3/4 6-18 3/4.6.4 COMBUSTIBLE GAS CONTROL Hydrogen Analyzers....................................... 3/4 6-20

, Electric Hydrogen Recombiners............................ 3/4 6-21 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE Safety Va1ves............................................ 3/4 7-1 TABLE 3.7-1 MAXIMUM ALLOWABLE POWER RANGE NEUTRON FLUX HIGH SETPOINT WITH INOPERABLE STEAM LINE SAFETY VALVES DURING 4 LOOP 0PERATION......................................... 3/4 7-2 TABLE 3.7-2 STEAM LINE SAFETY VALVES PER L00P..................... 3/4 7-3 Auxiliary Feedwater System............................... 3/4 7-4 Auxiliary Feedwater Storage Tank......................... 3/4 7-6 Specific Activity........................................ 3/4 7-7

(. TAB LE. 4. 7-1 SECONDARY COOLANT.. SYSTEM SPECIFIC.ACTIVITE SAMPLE...

AN D ANALY S I S P R0G RM. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

3/4 7-8 Main Steam Line Isolation Va1ves......................... 3/4 7-9 Atmospheric Steam Relief Valves ......................... 3/4 7-10 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION.......... 3/4 7-11 3/4.7.3 COMPONENT COOLING WATER SYSTEM........................... 3/4 7-12 3/4.7.4 ESSENTI AL COOLING WATER SYSTEM. . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 7-13 3/4.7.5 ULTIMATE HEAT SINK....................................... 3/4 7-14 3/4.7.6 (This specification number is not used.)

3/4.7.7 CONTROL ROOM MAKEUP AND CLEANUP f!LTRATION SYSTEM........ 3/4 7-16 3/4.7.0 FUEL HANDLING BUILDING (FHB) EXHAUST AIR SYSTEM.......... 3/4 7-19 3/4.7.9 5NUBBERS................................................. 3/4 7-21 FIGURE 4.7-1 SAMPLE PLAN 2) FOR SNUBBEI FUNCTIONAL TEST. 3/4 7...26 3/4.7.10 SEALED SOURCE CONTAMINATION.............................. 3/4 7-27 3/4.7.11 (This' specification number is not used.)

3/4.7.12 (This specification number is not used.)

3/4.7.13 AREA 1[MPERATURE MONITORING.............................. 3/4 7-31

(. TABLE 3.76)AREATEMPERATUREMONITORING........................... 3/4 7-32 3/4.7.14 l ESSENTIAL CHILLED 1lATER SYSTEM .......................... 3/4 7-33 Qp ph,y 500ill TEXAS - UNIT 1 O ix

ATTACHMENT G

. ST4fL.AE. 22 32

, PACE ,7 QF 0

( LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.9.11 WATER LEVEL - STORAGE P00 AS Spent Fuel Pool ......................................... 3/4 9-12

In-Containment Storage Pool ............................. 3/4 9-13 3/4.9.12 FUEL HANDLING BUILDING EXHAUST AIR SYSTEM ............... 3/4 9-14 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUT 00WN HARGIN.......................................... 3/4 10-1 3/4.10.2 GROUP HEIGHT, INSERTION, AND POWER DISTRIBUTION LIMITS... 3/4 10-2 3/4.10.3 PHYSICS TESTS............................................ 3/4 10-3 3/4.10.4 REACTOR COOLANT L00PS.................................... 3/4 10-4 3/4.10.5 POSITION INDICATION SYSTEM - SHUTD0WN.................... 3/4 10-5 3/4.11 RADI0 ACTIVE EFFLUENTS .

3/4.11.1 LIQUID EFFLUENTS 3/4 11-1

([ Concentration............................................

0ose..................................................... 3/4 11-2 Liquid Waste Processing System........................... 3/4 11-3 Liquid Holdup Tanks...................................... 3/4 11-4 3/4.11.2 GASEOUS EFFLUENTS Dose Rate................................................ 3/4 11-5 Dose - Noble Gases....................................... 3/4 11-6 Dose - Iodine-131, Iodine-133, Tritium, and Radioactive Material in Particulate Form............................. 3/4 11-7 Gaseous Waste Processing System.......................... 3/4 11-8 Explosive Gas Hixture.................................... 3/4 11-9 Gas Storage Tanks........................................ 3/4 11-10 3/4.11.3 SOLID RADI0 ACTIVE WASTES................................. 3/4 11-11 3/4.11.4 TOTAL 00SE............................................... 3/4 11-)f'l3 SOUTH TEXAS - UNIT 1 xi

ATTACHMENT G

. ST.HL AE 22 3 2-h 1I HL U b

yt. LOF 91 INDEX f BASES SECTION PAGE 3/4.4.9 PRESSURE / TEMPERATURE LIMITS............................... B 3/4 4-6 TABLE B 3/4.4-1 REACTOR VESSEL TOUGHNESS.s........................ B 3/4 4-9 FIGURE B 3/4.4-1 FAST NEUTRON FLUENCE (E>1MeV) AS A FUNCTION OF FULL POWER SERVICE LIFE.................................. B 3/4 4-10 FIGURE B 3/4.4-2 EFFECT OF FLUENCE AND COPPER CONTENT ON SHIFT OF RT FOR REACTOR VESSELS EXPOSED TO 550*F............ B 3/4 4-11 HDT 3/4.4.10 STRUCTURAL INTEGRITY..................................... B 3/4 4-15 3/4.4.11 REACTOR VESSEL HEAD VENTS................................ B 3/4 4-15 .

3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS.............................................. B 3/4 5-1 3/4.5.2 and 3/4 .3 ECCS SUBSYSTEMS............................... B 3/4 5-1 V4. S.4 Mat us 3/4.5.5 REFUELIN WATER STORAGE TANK.............................. B 3/4 5-2 3/4.5.6 RESIDUAL HEAT REMOVAL (RHR) SYSTEM ....................... B 3/4 5-2

( 3/4.6 CONTAINMENT SYSTEMS B 3/4 6-1 3/4.6.1 PRIMARY CONTAINMENT.......................................

3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS...................... ,

B 3/4 6-3 3/4.6.3 CONTAINMENT ISOLATION VALVES.............................. B 3/4 6-4 3/4.6.4 COMBUSTIBLE GAS CONTR0L................................... B 3/4 6-4 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE............................................. B 3/4 7-1 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION........... B 3/4 7-3 3/4.7.3 COMPONENT COOLING WATER SYSTEM............................ B 3/4 7-3 3/4.7.4 ESSENTIAL COOLING WATER SYSTEM............................ B 3/4 7-3 3/4.7.5 U LT I MAT E H EAT S I N K. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 7-3 3/4.7.6 (Not used) 3/4.7.7 CONTROL ROOM MAKEUP AND CLEANUP FILTRATION SYSTEM......... B 3/4 7-4 3/4.7.8 FUELliANDLINGBUILDINGEXMAUSTAIRSYSTEM................. B 3/4 7-4 3/4.7.9 SNUBBERS.................................................. B 3/4 7-4 l

50lfTH TEXAS - UNIT 1 xiv O.

I ATTACHMENT 6

. ST HL AE. z z 32.

, MQE 'l 0F 9I INDEX

( BASES SECTION PAGE 3/4.7.10 SEALED SOURCE CONTAMINATION............................... B 3/4 7-6 3/4.7.11 (Not used) ,

3/4.7.12 (Not used)

.344_.] 13 AREA TEMPERATURE MONITORING.

B 3/4 -6 3/4."J .19 E ss5WTT4t ~6{IEI.Il> W ATEN. Tern. ... .W. ... ... . .. . . . . . . . . . 4. .1.(,

3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1, 3/4.8.2, and 3/4.8.3 A.C. SOURCES, D.C. SOURCES, and ONSITE POWER DISTRIBUTION................................. B 3/4 8-1 .

3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES................... B 3/4 8-3 3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION....................................... B 3/4 9-1 3/4.9.2 INSTRUMENTATION........................................... B 3/4 9-1 3/4.9.3 DECAY TIME................................................ B 3/4 9-1 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS......................... B 3/4 9-1

{ 3/4~.9.5 COMMUNICATIONS..........................................r. B 3/4 9-1 3/4.9.6 REFUELING MACHINE......................................... B 3/4 9-2 3/4.9.7 CRANE TRAVEL - FUEL HANDLING BUI LDING. . . . . . . . . . . . . . . . . . . . . B 3/4 9-2 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION............. B 3/4 9-2 3 4.9.9 ............. B 3/4 9-2 ONTAINMENT VENTILATION ISOLATION QYS -

3/4.9.10 and 3/4.9.11 WATER LEVEL -4 ACTOR VESSEL and

..... B 3/4 9 2: g g ............

3/4.9.12 FUEL HANDLING BUILDING EXHAUST AIR SYSTEM ................ _D-3 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN MARGIN........................................... B 3/4 10-1 3/4.10.2 GROUP HEIGHT, INSERTION, AND POWER DISTRIBUTION LIMITS.... B 3/4 10-1 3/4.10.3 PHYSICS TESTS............................................. B 3/4 10-1 3/4.10.4 REACTOR COOLANT L00PS..................................... B 3/4 10-1 3/4.10.5 POSIT' ION INDICATION SYSTEM - SHUTD0WN..................... B 3/4 10-1

(

SOUTH TEXAS - UNIT 1 xv

ATTACHMENT 6

== INDEX FINAL DRAFT (a . . DESIGN FEATURES PAGE SECTION 5.1 SITE '

5-1 5.1.1 EXCLUSION AREA.................................................

5-1 5.1.2 LOW POPULATION Z0NE............................................

5.1.3 _ MAP DFFINING UNRESTRICTED AREAS AND SITE BOUNDARY FOR GASE0US AND LIQUID EFFLUENTS......................

5-1

@ DI

5. 2 CONTAINMENT 5-1 5.2.1 CONFIGURATION..................................................

TEMPERATURE................................ 5-1 5.2.2 DESIGN PRESSURE AND 5-2 FIGURE 5.1-1 EXCLUSION AREA..........................................

5-3 FIGURE 5.1-2 LOW POPU LATION Z0NE. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

FIGURE 5.1-3 RESTRICTED AREA AND SITE BOUNDARY FOR RADI0 ACTIVE GASEOUS EFFLUENTS.......................................

5-4 FIGURE 5.1-4 RESTRICTED AREA AND SITE BOUNDARY FOR RADI0 ACTIVE 5-5 LIQUID EFFLUENTS .......................................

5.3 REACTOR CORE 5-6 5.3.1 FUEL ASSEMBLIES................................................

5-6 5.3.2 CONTROL R00 ASSEMBLIES......................................'...

5.4 REACTOR COOLANT SYSTEM TEMPERATURE................................ 5-6 5.4.1 DESIGN PRESSURE AND 5-6 5.4.2 V0LUME.........................................................

5-6 5.5 METEOROLOGICAL TOWER LOCATION. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

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

5-7 5.6.2 DRAINAGE.......................................................

5-7 5.6.3 CAPACITY.......................................................

5-7 5.7 COMPONENT CYCLIC OR TRANSIENT LIMIT. . . . . . . . . . . . . . . . . . . . . . . . . . .

5-8

( TABLE 5.7-1 COMPONENT CYCLIC OR TRANSIENT LIMITS. . . . . . . . . . . . . . . . . .

SOUTH TEXAS - UNIT 1 xvii

ATTACHMENT 6

== INDEX FINAL DRAFT 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............................................. 6-2611 Annual Radiological Environmental Operating Report......... 6-17 Semiannual Radioactive Effluent Release Report............. 6-18 Monthly Operating Reports.................................. 6-20 Radial Peaking Factor Limit Report......................... 6-20 6.9.2 SPECIAL REP 0RTS............................................ 6-20 6.10 RECORD RETENTION........................................... 6-21 l

i 1

SOUTH TEXAS - UNIT 1 xix l

l t

ATTACHMENT 6 SAFETY LIMITS OF 41

(

(, 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 criteria and associated Code requirements.

310r The entire RCS is hydrotested at 125% (&HU)-psig) of design pressure, to .

demonstrate integrity prior to initial operation. ,

~

0 SOUTH TEXAS - UNIT 1 B 2-2

ATTACHMENT 6

. ST.HL.AE. z.tsa PAGE 13 OF 91 LIMITING SAFETY SYsitM dtis mud BASES Undervoltage and Underfrequency - Reactor Coolant Pump Buses The Undervoltage and Underfrequency Reactor Coolant Pump Bus trips pro- ,

vide core protection against DNB as a result of complete loss of forced coolant flow. The specified Setpoints assure a Reactor trip signal is generated before the Low Flow Trip Setpoint is reached. Time delays are incorporated in the Underfrequency and Undervoltage trips to prevent spurious Reactor trips from momentary electrical power transients. For undervoltage, the delay is set so that the time required for a signal to reach the Reactor trip breakers following the simultaneous trip of two or more reactor coolant pump bus circuit breakers shall not exceed 1.2 seconds. For underfrequency, the delay is set so that the time required for a signal to reach the Reactor trip breakers after the Underfrequency Trip Setpoint is reached shall not exceed 0.3 second.

On decreasing power, the Undervoltage and Underfrequency Reactor Coolant Pump Bus trips are automatically blocked by P-7 (a power level of approximately 10%

of RATED THERMAL POWER with a turbine impulse chamber pressure at approximately 10% of full power equivalent); and on increasing power, reinstated automatically by P-7.

Turbine Trip A Turbine trip initiates a Reactor trip. On decreasing power, the Reactor trip from the Turbine trip is automatically blocked by P-9 (a power level of

. approximately 50% of RATED THERMAL POWER); and on increasing power, reinstated automatically by P-9.

Safety Injection Input from ESFAS If a Reactor trip has not already been generated by the Reactor Trip System instrumentation, the ESFAS automatic actuation logic channels will initiate a Reactor trip upon any signal which initiates a Safety Injection. The ESFAS instrumentation channels which initiate a Safety Injection signal are shown in Table 3.3-3.

Reactor Trip System Interlocks The Reactor Trip System interlocks perform the following fu ions:

P-6 On increasing power, P-6 allows the manual block of he ource Range trip (i.e., prevents premature block of Source Rar.ge ri and deenergizes the high voltage to the detectors) On dec g power, Source Range Level trips are automatically reactivated and high voltage restored.

P-7 On increasing power, P-7 automatically enables Reactor trips on low flow in more than one reactor coolant loop, reactor coolant pump bus undervoltage and underfrequency, pressurizer low pressure, and pressurizer high level. On decreasing power, the above listed trips

.( are automatically blocked.

SOUTH TEXAS - UNIT 1 B 2-7

" ATTACHMENT 6 ST.HL AE. 2232 PAGE W OF 9L REACTIVITY CONTROL SYSTEMS f ROD DROP TIME J ,_

LIMITING CONDITION FOR OPERATION 3.1.3.4 The individual full-length (shutdown and control) rod drop time from ,

the fully withdrawn position shall be less than or equal to 2.8 seconds from beginning of decay of stationary gripper coil voltage to dashpot entry with:

a. T,yg greater than or equal to 561*F, and

b. All reactor coolant pumps operating.

APPLICABILITY: MODES 1 and 2.

ACTION:

With the drop time of any full-length rod determined to exceed 4 the above limit, restore the rod drop time to within the above limit prior to proceeding to MODE 1 or 2.

SURVEILLANCE REOUIREMENTS 4.1.3.4 The rod drop time of full-length rods shall be demonstrated through

( measurement prior to reactor criticality:

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

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. . . . .

S b

b m

SOUTH TEXAS - UNIT 1 3/4 1-21

ma ATTACHMENT 6 FINAL HAFT 3/4.2 POWER DISTRIBUTION LIMITS '

3/4.2.1 AXIAL-FLUX DIFFERENCE L_IMITING CONDITION FOR OPERATION )

3.2.1 The indicated AXIAL FLUX DIFFERENCE.(AFD) shall be maintained within the ,

following target band (flux difference units) about the target flux difference:

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

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

The indicated AFD may deviate outside the above required target band at greater than. or equal to 50% but less than 90% of RATED THERMAL POWER provided the indi-cated AFD is within the Acceptable Operation Limits of Figure 3.2-1 and the cumu-lative 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 above required target band and with THERMAL POWER greater than or equal to 90% of RATED THERMAL POWER, within 15 minutes either:

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

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

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

1: ' THERMAL POWER to less than 50% of RATED THERMAL POWER within 36 minutes, and

2. The Power Range Neutron Flux * ** - High Setpoin o less than or N 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 . -

See Special Test Exceptions Specification 3.10.2.

- Surveillance testing of the Power Range Neutron Flux Channel may be performed pursuant to Specification 4.3.1.1 provided the indicated AFD is maintained within -

the Acceptable Operation Limits of Figure 3.2-1. A total of 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months operation C may be accumulated with the AFD outside of the above required target band during testing without penalty deviation.

SOUTH TEXAS - UNIT 1 3/4 2-1

1 ATTACHMENT 6

. ST HL AE.zz sz PAQE t(,0F 91 POWER DISTRIBUTION LIMITS 3/4.2.2 HEAT _ FLUX HOT CHANNEL FACTOR - Fg LIMITING CONDITION FOR OPERATION 3.2.2 F 9(Z) shall be limited by the foll,owing relationships: ,

F0 (Z) 5 2.50 [K(Z)] for P > 0.5 P

F9 (Z) 1 5.0 [K(Z)] for P i 0.5

, and Where: P _ RATED THERMAL POWER THERMAL POWER K(Z) = the function obtained from Figure 3.2-2 for a given core height location.

APPLICABILITY: MODE 1.

ACTION:

With F (Z) exceeding its limit: -

9

a. -Cerply .;ith the-fo44 cuing .*CT!#

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

the limit within 15 minutes and simila grl reduce the Power Range Neutron Flux-High Trip Setpoinig)within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; POWER OPERATION 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 Setpoin Tibbeen reduced at least 1%

for each 1% F (Z) exceeds the ' limit.

9 k ha.s

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 providert F (Z) is demonstrated through incore mapping to be 9

. withinitslimit.

? . .-

I; SOUTH TEXAS - UNIT 1 3/4 2-5

ATTACHMENT 6

. ST.HL.AE 2232 PAGE..t10F 91 POWER DISTRIBUTION LIMITS h 3/4.2.4 QUADRANT POWER TILT RATIO LIMITING CONDITION FOR OPERATION 3.2.4 The QUADRANT POWER TILT RATIO shall not exceed 1.02. ,

APPLICABILITY: MODE 1, above 50% of RATED THERMAL 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 THERMAL 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

-X Setpoin 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 thereafter, verify that Fq (Z) (by xy evaluation) and Fh are within their limits by performing Surveil-F 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.

SURVEILLANCE REOUIREMENTS

{

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

a. Calculating 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 TILT RATIO shall be determined to be within the limit when above 75% of RATED THERMAL POWER with one Power Range channel ,

inoperable by using the movable incore detectors to confirm indicated QUADRANT l 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:

a. Us'ing the four pairs of symmetric thimble locations, or. <

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

l b *See Special Test Exceptions Specification 3.10.2. )

SOUTH TEXAS - UNIT 1 3/4 2-10

O D ]'

vs TABLE 3.3-1 g

x w REACTOR TRIP SYSTEM INSTRUMENTATION O MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE 0F CHANNELS TO TRIP OPERABLE MODES ACTION c-FUNCTIONAL UNIT 5 \- e 1, 2 Manual Reactor Trip , 1 2 1 .

1. 2 '

2 1 2 3*, 4*, 5* 10 .

2. Power Range, Neutron Flux

a. High Setpoint 4 2 3 1, 2 2#

b. Low Setpoint 4 2 3 1###, 2 2#

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

3h Ri>

High Positive Rate 9 g >m E:

A.4 Power Range, Neutron luxgA. 4 2 3 1, 2 2# eke Y High Negative Rate f, M *e m

1###, 2 -3

5. Intermediate Range, Neutron Flux 2 1 2

6. Source Range, Neutron Flux Startup 2 1 2 2## 4 a.

b. Shutdown . 2 1 2 3*, 4*, 5* 10 Extended Range, Neutron Flux 2 0 2 3, 4, 5 4 7.

8. Overtemperature AT 4 2 3 1, 2 6#

9. Overpower AT 4 2 3 1, 2 6# $

10. Pressurizer Pressure--Low 4 2 3 1 6#

(Interlocked with P-7) M P

11. Pressurizer Pressure--High 4 2 3 1, 2 6#

l 6#

C:3

12. Pressurizer Water Level--High 4 2 3 1 (Interlocked with P-7) 3

O ' .p q y,

TABLE 4.3-1 (Centinued) 8 g REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS TRIP h ACTUATING' MODES FOR

g -

ANALOG CHANNEL DEVICE WHICH CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEIL W CE c .

TEST TEST LOGIC TEST IS REQUIRED

' CHECK CALIBRATION 5 FUNCTIONAL UNIT ., '

~

11. Pressurizer Pressure N.A. 1, 2

--High S R Q(17) N.A.

12. Pressurizer Water 34%

Level--High S R Q(17) N.A. N.A. 1 gi3;!

.n I

13. Reactor Coolant Flow Q(17,18) N.A. N. A. @

> @%gg

--Low S R

- aw H y '"e

14. Steam Generator Water N.A. 1, 2 Q(17,18) N.A.

y Level--Low-Low S R

~

15. Undervoltage - Reactor Coolant Pumps N.A. R N.A. Q(17) N.A. I

16. Underfrequency -

Reactor Coolant N.A. 1 Pumps N.A. R N.A. Q(17)

17. Turbine Trip

a. Low Emergency Trip < N. A.< R< N. Ar S/U(1,10)4-N. A+1 g Fluid Pressure Turbine Stop Valve N. An R< N. A.' S/U(1, 10) 4-- N. A+1 b.

Closure N. A.; R' N.A.*1, 2

18. Safety Injection Input from < N. A. < N. A.%

ESFAS

p. :O %

TABLE 3.3-3 g

h ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION MINIMUM y, -

TOTAL NO. CHANNELS CHANNELS APPLICABLE OF CHANNELS TO TRIP OPERABLE MODES ACTION 4 e FUNCTIONAL UNIT- .

z U 1. Safety Injection.(Reactor

Trip, Feedwater Isolation, Control Room Emergency Ventilation, Start Standby Diesel Generators, Reactor i Containment Ceeli..; Ter.;,"- Fan Coolers and Essential Cooling Water). s$h 1,2,3,4 19 MM p59 Manual Initiation 2 1 2 w

a. omg A b. Automatic Actuation %hE ow*

Logic 2 1 2 1,2,3,4. 14 Y ,

"* l co

c. Actuation Relays 3 2 3 ' 1, 2, 3, 4 14 l l

d. Containment 3 2 2 1,2,3,4 15* l Pressure--High-1

e. Pressurizer 4 2 3 1, 2, 3# 20* I Pressure--Low l

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

any steam line in each steam Line Pressure-Low line m amen

g. -

3/ loop 2/ loop in 2/ loop in 1####, 2, 3# 15*

Compensated TCOLD Low-Low (interlocked any loop each loop with P-15)

O

q m A ;

}

(

TABLE 3.3-3 (Continued) l g

k h ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION MINIMUM N APPLICABLE :

g TOTAL NO.

CHANNELS CHANNELS MODES ACTION l FUNCTIONAL UNIT 0F CHANNELS TO TRIP OPERABLE 4 %. !

c '; ;

5

4. Steam Line Isolation (Continued) -

Steam Line Pressure - r g c. 15*

~

Negative Rate--High 3/ steam line 2/ steam line 2/ steam line 3#M '

any steam in each steam line line j 3 2 2 1,2,3 15* ; ;

d. Containment Pressure - '

High-2 l }$$

e. Compensat'ed Steam Line 3/ steam line 2/ steam line any steam line 2/ steam line in each steam 1, 2, 3# 15* AM w Pressure - Low #p I$

line 1

} f. Compensated TCOLD 3/ loop 2/ loop any 2/ loop in each loop 1H N , 2, 3#

' ~

15*

] *"4 H Low-Low (interlocked loop with P-15)

5. Turbine Trip and Feedwater Isolation

a. Automatic Actuation 2 1 2 1,2,3 25 l Logic and Actuation ~

Relays

b. Steam Generator 4/sti. gen. 2/stm. gen. 3/stm. gen. 1, 2, 3 20*

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

c. Compensated TC LD p

1 Low (interlocked 3/ loop 2/ loop in 2/ loop in 1##f, 2, 3# 15* y with P-15) any loop each loop

d. Feedwater Flow - High 3/sta. gen. 2/stm. gen. 2/sta. gen. INN, 2, 3 15* @

coincident with either in any stm. in each sta.

. of the following in gen. gen.

2 of 4 loops: (;ng,7;,c y ,

w%.P-ts) .

C Q O Y a o., bip*. .

m TABLE 3.3-3 (Continued) [ lins. ft 8 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION

= i

-4 MINIMUM O CHANNELS APPLICABLE R TOTAL NO. CHANNELS 0F CHANNELS TO TRIP OPERABLE MODES ACTION

, FUNCTIONAL UNIT -

'a 6 c '

. 7. Automatic Switchover b '

Containment Sump **** .

w

a. Automatic Actuation 3-1/ train 1/ train 1/ train 1,2,3,4 19 Logic and Actuation Relays h> ,

b. RWST Level--Low-Low 3-1/ train 1/ train 1/ train 1, 2., 3, 4 19 :n;:9K>'"i i

Coincident With: See Item 1. above for all Safety Injection initiating functions i g Safety Injection and requirements. M N[

2 l i

Y 8. Loss of Power , l n

. a. 4.16 kV ESF Bus Under- 4/ bus 2/ bus 3/ bus 1, 2, 3, 4 20* l voltage-Loss of Voltage

b. 4.16 kV ESF Bus Under-voltage-Tolerable i Degraded Voltage 20*

q i Coincident with SI 4/ bus 2/ bus 3/ bus 1, 2, 3, 4 ummmm.

t

c. 4.16 kV EsF Bus Under- -

voltage - Sustained P Degraded Voltage 4/ bus 2/ bus 3/ bus 1, 2, 3, 4 20*

9. Engineered Safety Features E Actuatior. System Interlocks ~

a. Pressurizer Pressure, 3 2 2 1,2,3 21 P-11 4 2 3 1,2,3 21

b. Low-Low T,yg, P-12

l ATTACHMENT G n ST-HL-AE. z z 3 r.

TABLE 3.3-4 (Continued) '

! PAGE-73 OF 9/

h ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS

$ TOTAL SENSOR ERROR

_4 ALLOWANCE (TA) (S) TRIP SETPOINT ALLOWABLE VALUE R FUNCTIONAL UNIT Z_

. 4. Steam Line Isolatioh,(Continued) c -

'. N.A.

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

i'i b.

and Actuation Relays w

c. Steam Line Pressure - 2.6 0.5 0 $ 100 psi 5 126.3 psi"*

Negative Rate--High

d. Containment Pressure - 3.6 0.71 2.0 $ 3.0 psig 5 4.0 psig High-2

e. Compensated Steam Line 13.6 10.71 2.0 1 735 psig 1 714.7 psig*

, w Pressure - Low

f. Compensated T -

4.5 0.5 1.0 1 532*F 1 528 F***

COLD w Low-Low (interlocked with P-15)

5. Turbine Trip and Feedwater Isolation N.A. N.A. N.A. N.A. N.A.

a. Automatic Actuation Logic .

and Actuation Relays

b. Steam Generator Water 4.5 2.35 2.0+0.2# < 87.5% of < 88.9% of Level--High-High (P-14) Earrow range harrow range instrument instrument span. span.

c. Compensated T -L w 4.5 0.5 1.0 1 538*F 3 534*F***

COLD (interlocked with P-15)

d. Feedwater Flow-High6 der 7. 2 2.76 4.0 $ 30.0% Flow $ 32.2% Flow Coincident With Either of the Following in 2 of 4 Loops:

f"'% T' O --- . - .

ATTACHMENT %

t

. ST.HL-AE tzs:

TABLE 3.3-4 (Continued) M @ T- - - - -

l m l 8 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS N TOTAL SENSOR ERROR -

y (S) TRIP SETPOINT ALLOWABLE VALUE '

ALLOWANCE (TA) Z Q FUNCTIONAL UNIT ,

?, 6. AuxiharY 1%dwa+v- (con +mued) See Item 8. below for all Loss of Power Trip g.

f. Loss of Power-(Motor .

Driven Pumps Only) Setpoints and Allowable values. ,

c . . '

5

-' 7. Automatic Switchover to '

H . Containment Sump N.A. N.A. N.A. N.A. !

a. Automatic Actuation Lpgic N.A. '

and Actuation Relays 4 1.21 2.0 > 11% > 9.1%

b. RWST Level--Low-Low 5.0 Coincident With: See Item 1. above for all Safety Injection Trip Setpoints and Allowable i Safety Injection Values. .

I w ' ' '

A 8. Loss of Power

N.A. N.A. > 3107 volts > 2979 volts I

a. 4.16 kV ESF Bus Undervoltage N.A. with a < 1.75 with a < 1.93 w (Loss of Voltage) '

second time ' second time !

delay. delay. j N.A. N.A. > 3835 volts > 3786 volts

b. 4.16 kV ESF Bus Undervoltage N.A. Uith a < 35 sith a < 39 (Tolerable Degraded Voltage second Ilme second time Coincident with SI) delay. delay. !

N.A. N.A. > 3835 volts 3786 volts c .- 4.16 kV 5SF Bus Undervoltage H.A. with a < 50 Gith a < 55 i (Sustained Degraded Voltage) second lime second Ilme delay. delay. l 4

FINAL. DRAFT t

e

~ ~-

p. _

TABLE 3.3-4 (Continued) .

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUNENTATION TRIP SETPOINTS 1 '

TOTAL SENSOR ERROR a

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

g -

e 11. FHB HVAC.(Continued) ,

$ b. Automatic Actuation N.A. N.A. N.A. .N.A. *.'N.A. l

Logic and Actuation {

H Relays j

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

d. Spent Fuel Pool Exhausi- 3.1x10 4 1.8x10 4 1.3x10 4 <5.0x10 4 <6.4x10 4 pCi/cc pCi/cc pCi/cc pCi/cc pCi/cc h itoacirit+ g Sh 3 4s .

(d

. h- -

W$N Z Mi >P

?h9

# "e tllllllll5

n .N 9 g TABLE 4.3-2 (C ntinued) l ATTACHMENT G S

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION - ST-HL-AE- 2 2 32-SURVEILLANCE REQUIREMENTS PAGE.4 OF 9/

a 92 DIGITAL OR TRIP ,

$; ANALOG ACTUATING MODES !

,_ CHANNEL DEVICE MASTER SLAVE FOR WHICH !

c- CHANNEL CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAX . SURVEILLANCE !

j'i FUNCTIONAL UNIT .'. CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST. , IS REQUIRED i m .

- 5. Turbine Trip and Feedwater Isolation !

i

a. Automatic Actuation N.A. N.A. N.A. N.A. M(1) M(6) Q(4) 1, 2, 3 i Logic and Actuation -

Relays ,

l

b. Steam Generator Water S R H N.A. N.A. N.A. N.A. 1,2,3 w Level-High-High (P-14) w c. Compensated T COLD

-L w S R M N.A. N.A. N.A. N.A. 1, 2, 3

$ d. Feedwater Flow-High S R M N.A. N.A. N.A. N.A. 1,2,3 Coincident with either of the following in 0 **d**hd d e P-I N 2 of 4 loops: Reactor .

Coolant Flow-Low or T,yg-Low

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

f. T,yg-Low Coincident S R M N.A. N.A. N.A. N.A. 1,2,3 with Reactor Trip M )

(Feedwater Isolation Only)

6. Auxiliary Feedwater

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

FINAL HAFT

- - - - - a

L_.~~ ~ -

. . . . . . . . . . . . . . . . . . . . . . . . .. . 3 GE TABLE 3.3-7 SEISMIC MONITORING INSTRUMENTATION

{ ' ~~

MINIMUM _

HEASUREMENT INSTRUMENTS INSTRUMENTS AND SENSOR LOCATIONS RANGE OPERABLE

1. Triaxial Time-History Accelerometers ***l

a. Free Field 13g 1

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

c. Outside Face Containment Shell i3g 1 (Reactor Containment Building El. 68'0")- ,

d. Steam Gene'rator Upper Lateral Support 13g 1 (Reactor Containment Building El. 66'7 ")

e. Fuel Handling Building Foundation 13g 1 (Fuel Handling Building El. -29'0")

f. Mechanical Electrical Auxiliary Building 13g 1 (Mechanical Electrical Auxiliary Building El. 35'0")

2. Triaxial Peak Accelerographs ,

a. Spent Fuel Pool Heat Exchanger 13g 1 f- (Inlet Line, Fuel Handling Building -

El. 64'5k")

b. Reactor Vessel 13g 1 (Reactor Containment Building El. 68'0")

c. Cold Leg of RC Piping 13g 1 (Reactor Containment Building El. g) *

3. Self-Contained Triaxial Accelerograph 13g 1

.(At Reactor Containment Building Foundation3 Tendon Gallery El. -36'9")

4. Triaxial Seismic Switch ** # 0.03 to 3g 1*

5. Triaxial Seismic Trigger ** ## 0.003 to 0.3g 1*

6. Response $pectrum Analyzer ** 1 to 32 Hz 1*

7. Magnetic Tapeiecorders** 0.1 to 33 Hz 6

8. Playback System ** N.A. 1

Vith reactor' control room indication and alarm y

- At seismic monitoring panel in Control Room (linit 1)

- - Accelerometer data is gathered and analyzed by the Response Spectrum Analyzer (Item 6).

Triaxial seismic switch is set at the OBE acceleration level of 0.05g 1

{ horizontal and 0.033g vertical.

M Triaxial seismic trigger is set at 0.02g all axes.

! SOUTH TEXAS - UNIT 1 3/4 3-56

EE!!"Bli

> FINAL DRAFT TABLE 4.3-4 b SEISMIC MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS

~

ANALOG ' ' ~

CHANNEL CHANNEL CHANNEL OPERATIONAL CHECK CAllBRATION TEST INSTRUMENTS AND SENSOR LOCATIONS ,

o

1. Triaxial Time-History Accelerometers ***

M R SA

a. Free Field -

b. Containment Bldg. Foundation H R SA (TendonGalleryQl.-36'9")

M R SA

c. Outside Face Containment Shell (Reactor Containment Building El. 6B'0") '

d. Steam Generator Upper Lateral Support M R SA (Reactor Containment Building El. 66'75")

e. Fuel Handling Building Foundation M R SA (Fuel Handling Building El. -29'0")

f. Mechanical Electrical Auxiliary M R SA Building (Mechanical Electrical Auxiliary Building El. 35'0")

2. Triaxial Peak Accelerographs

(' a. Spent Fuel Pool Heat Exchanger (Inlet Line fuel Handling Building El.64'51")y N.A. R N.A.

N.A. R N.A.

b. Reactor Vessel (Reactor Containment Building El. 68'0")

c. Cold Leg of RC Piping N.A. R N.A.

(Reactor Containment Building -

El. @ 34'a" SA

2. Self-Contained Triaxial Accelerograph M R (At Reactor Containment Building ,

Foundation3 Tendon Gallery El. -36'9")

M R SA

4. Triaxial' Seismic Switch * **

5. Triaxial' S$1smic Trigger * ** M R SA M- R SA

6. Response Spect' rum Analyzer * **

M R SA

7. Magnetic Tape Recorders **

S. Playback System ** ,

M ,

R- N.A.

i

With reactor control room indication and alann

- At seismic monitoring panel in Control Room, nit 1 . ,

{-. *** Accelerometer data is gathered and analyzed by tTe Response Spectrum Analyzer (Item 6).

SOUTH TEXAS - UNIT 1 -

3/4 3-57 1

ATTACHMENT 4

. ST.HL.AE. 2 2 3 2. E IIMlas p INSTRUMENTATION PAGE 23 OF 91

') REMOTE SHUTDOWR SYSTEM LIMITING _ CONDITION FOR OPERATION 3.3.3.5 The Remote Shutdown System transfer switches, power, controls and monitoring instrumentation channels shown in Table 3.3-9 shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3.

ACTION:

t

a. With the number of OPERABLE remote shutdown monitoring channelg g transfer switches, power or control circuits less than the Min 10m Channels OPERABLE as required by Table 3.3-9, restore the inoperable

' channel (s) to OPERABLE status within 7 days, or be in HOT. SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ,

b. With the number of OPERABLE remote shutdown monitoring channel $ x transfer switches, pow'er or control circuits less than the TotaT Number of Channels as required by Table 3.3-9, within 60 days restore the inoperable chanr.el(s) to OPERABLE status or, pursuant to Specification 6.9.2, submit.a Special Report that defines the corrective action to be taken.

c. The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE REOUIREMENTS 4.3.3.5.1 Each remote shutdown monitoring inst'rumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK and CHANNEL CAllBRATION operations at the frequencies shown in Table 4.3-6.

4.3.3.5.2 Each Remote Shutdown System transfer switch, powsr and control circuit including the actuated components, shall be demonstrated OPERABLE at least once per 18 months.

) .

SOUTH TEXAS - UNIT 1 3/4 3-61

TABLE 3.3-9 8

h REMOTE SHUTDOWN SYSTEM h TOTAL NO. MINIMUM g READOUT OF CHANNELS

, INSTRUMENT LOCATION CHANNELS OPERABLE e . \.

5 1. Neutron Plux - Extended Range

~ a. Startup Rate ASP *-QDPS** 2 X2

b. Flux Level ASP-QDPS 2 E 1

2. Reactor Trip Breaker ASP-QDPS 1/ trip breaker 1/ trip breaker Indication Reactor Trip Switchgear o u> 2-R

3. Reactor Coolant Temperature-Wide Range

$$4 m ;= g 1/ loop - 3 loops #

oM3 h a. Hot Leg ASP-QDPS 4-1/ loop o ;;G m~z

b. Cold Leg ASP-QDPS 4-1/ loop 1/ loop - 3 loops # fU

4. Reactor Coolant Pressure- ASP-QDPS 3 2 Wide Range / Extended Range 1 5. Pressurizer Water Level ASP-QDPS 4 2

6. Steam Line Pressure ASP-QDPS 4-1/ steam line 1/ steam line -

3 steam lines #

7. Steam Generator Water Level- ASP-QDPS 4-1/ steam 1/ steam generator -

Wide Range generator 3 steam generators #

~

O

=1

O O j O :.

i i

TABLE 3.3-9 (Continued) j g -

I c .

i REMOTE SifdT00WN SYSTEM E 5

-e '

TOTAL NO. MINIMUM

'x" CHANNELS !

? READOUT LOCATION

~0F CHANNELS OPERABLE j INSTRUMENT- .

g, :

e 1/ steam generatpr - i i'i 8. Auxiliary Feedwater Flow Rate ASP-QDPS 4-1/ steam

generator 3 steam generators # l w - ,

9. Auxiliary Feedwater Storage ASP-QDPS 3 2 l Tank Water Level l

10. Core Exit Thermocouples ASP-QDPS ## 4 thermocouples/

core quadrant h

m ;= g i

to :

j TOTA 1. NO. MINIMUM [wz$

'Nz j CONTROLS OF CHANNELS R TRANSFER SWITCHES AND

ASSOCIATED CONTROLS TRANSFER SWITCH LOCATIONS LOCATIONS

^

CHANNELS OPERABLE f I" !

i Y 4' '

2# i S 1. Steam Generator PORVs ZLP-6 3 (Train A) ASP i

ZLP-654 (Train B) l ZLP-655 (Train C) . ,

ASP (Train D) j ASP 2 1

2. Reactor Head Vent Throttle ZLP-700 (Train A)

ZLP-701 (Train B) l Valves

3. Reactor Head Vent Isolat' ion ZLP-700 (Train A) ASP 2 po'r 1 paIv-i Valves ZLP-701 (Train B) I i

ZLP-653 (Train A-AFW Pump) ASP 4 2#

4. AFW Pumps and Valves ZLP-700 (Train A-AFW Valves) .

ZLP-654 (Train B-AFW Pump)

. ZLP-701 (Train B-AFW Valves)

ZLP-655 (Train C-AFW Pump) M ZLP-709 (Train C-AFW Valves)

F ASP (Train D)

C q

h

TABLE 3.3-9 (Continued) - !

c I REMGTE SHUTDOWN SYSTEM j i$ MINIMUM TOTAL NO l e

CONTROLS 0F CHANNELS M

g , RANSFER T SWITCHES AND TRANSFER SWITCH OPERABLE l LOCATIONS LOCATIONS CHANNELS '

ASSOCIATED CONTROLS e g. :

1, l ASP 2 g 5. Centrifugal Charging Pumps ZLP-653 (Train A) *

, j g . ZLP-655 (Train C)

" ASP 2 e j.

6. Boric Acid. Transfer Pumps ZLP-653 (Train A)

ZLP-655 (Train C) l 1

m ASP 2 1

7. Pressurizer PORVs and Block ZLP-700 (Train A) f Valves ZLP-701 (Train g O ,

wh wa 8.

AdcumulatorDischarge ZLP-653 (Train A) ASP 3 3 %

9 z ZLP-654 (Train B)

Isolation Valves and Power 5

$ Lockouts ZLP-655 (Train C) 7>h m z .

9. Letdown Stop Valves. ZLP-700 (Train A) ASP 2' .. 1 9d "a

ZLP-709 (Train C)

~

ZLP-653 (Trait'A) 3 2 f 10.CCNYumps'andHeatExchanger ZLP-653 (Train A) a ZLP-654 (Train B) ZLP-654 (Train B)

Outlet Valves ZLP-655 (Train C)

ZLP-655 (Train C) a

~

ZLP-653 (Train A) 3 2 ZLP-653 (Train A)

11. ECW Pumps

.- 7tP-654 (Train D) 71P-654 (Train B) %

==

)

ZLP-655 (Train C)

,~ ZLP-655 (Train C) E 12.EkBHVACFans ZLP-700 (Train A) ZLP-700 (Train A) 3 2 p ZLP-653 (Train A- ZLP-653 (Train A- y l Battery Room and Battery Room Electrical Penetration and Electrical Space Fans) Penetration Space Fans)

ZLP-701 (Train B) ZLP-701 (Train B)

ZLP-654 (Train B- ZLP-654 (Train B-Battery Room and Battery Room Electrical Penetration and Electrical .

Space Fans) Penetration Space Fans) ,

e

.O O N TABLE 3.3-10 (Continued) .i 2 ACCIDENT MONITORING INSTRUMENTATION w

Q TOTAL MINIMUM M NO. OF ' CHANNELS

, INSTRUMENT

. CHANNELS OPERABLE ACTION 6

e - .

$ 13. Containment Water., Level '

(Narrow Range) -

2 1 36 .

w

14. Containment Water Level (Wide Range) 3 1 37

15. Core Exit Thermocouples ** 4 thereoccuples/ core 42 quadrant

16. Steam Lina Radiation Monitor 1/ steam line 1/ steam line ,

40 R 17. Containment ;Ja c.x Y High w

Range Radiation Monitor 2 1 39 mh O>I d 18. Reactor Vessel Water Level (RWL) 2* 1* -41 o PM

19. Neutron Flux (Extended Range) 2 1 36 m u(

E

20. Containment Hydrogen Concentration 2 1 36

21. Containment Pressure (Extended Range) 2 1 36

22. Steam Generator Blowdown Radiation Monitor 1/ blowdown line 1/ blowdown line 40

23. Neutron Flux - Startup Rate (Extended Range) *2 1 36 9

mimumme

A channel is eight sensors in a probe. A channel is OPERABLE if four or mora sensors, one or more E in the upper section and three or more in the lower section, are OPERABLE.

- A total of 50 thermocouples are provided with 25 thermocouples on each of two trains. Quadrants 8 y== mms and D have 6 thermocouples per train each. Quadrants A and C each have 6 thermoco @les on one train and 7 thermocouples on the other train. No ACTION is required as long as each quadrant has 4 g thermocouples per train OPERABLE.

21

- _ -_--_h

D- O %

TABLE 4.3-7 ,

5 5

z ACCIDENT MONIT0iiING INSTRUMENTATION SURVEILLANCE REQUIREMENTS g CHANNEL CHANNEL g

u, INSTRUMENT CHECK CALIBRATION e

c 1. Containment Pressyre

M R ', ,

2. Reactor Coolant Outlet Temperature - THOT (Wide Range) M R

[

3. Reactor Coolant Inlet Temperature - TCOLD (Wide Range) M R

4. Reactor Coolant Pressure - Wide Range and Extended Range M R gg3 a , -<

5. Pressurizer Water Level M R M< r,n g

, 6. Steam Line Pressure M R Q;g

} ou w

, 7. Steam Generator Water Level - Narrow Range M R t' c b

8. Steam Generator Water Level - Wide Range M R

9. Refueling Water Storage Tank Water Level M R

10. Auxiliary Feedwater Storage Tank Water Level M R --

11. Auxiliary Feedwater Flow M R

12. Reactor Coolant System Subcooling Margin Monitor M R -

13. Containment Water' Level (Harrow Range) M R Z

14. Containment Water Level (Wide Range) M R P

15. Core Exit Thermocouples M R O

16. Steam Line Radiation Monitor M R 1,. Cont., _ t m s e,g, R.ng. Rae,ation Mon,to, M R a

9

o ATTACHMENT (,

INSTRUMENTATION

== FINAL DRAFT

{ RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.10 The radioactive liquid effluent monitoring instrumentation channels ,

shown in Table 3.3-12 shall be OPERABLE with their Alarm / Trip Setpoints set to ensure that the limits of Specification 3.11.1.1 are not exceeded. The Alarm /

Trip Setpoints of these channels shall be determined and adjusted in accordance with the methodology and parameters in the OFFSITE DOSE CALCULATION MANUAL (ODCM).

APPLICABILITY: At all times.

ACTION: ,

a. With a radioactive liquid effluent monitoring instrumentation channel Alarm / Trip Setpoint less conservative than required by the above specification, immediately suspend the release of radioactive liquid effluents monitored by the affected channel, or declare the channel inoperable,

b. With less than the minimum number of radioactive liquid effluent monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 3.3-12. Restore the inoperable instrumentation to OPERABLE status within the time specified in the ACTION, or explain in the

'C, next Semiannual Radioactive Effluent Release Report pursuant to Specification 6.9.1.4 why this inoperability was not corrected within the time specified,

c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REOUIREMENTS 4.3.3.10 Each radioactive liquid effluent monitoring instrumentation' channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, at the frequenciesCHANNEL CALIBRATION, shown in Table 4.3-8.and DIGITAL CHANNEL OPERATIONAL TESI 4:k

. l

.. i e

SOUTH TEXAS - UNIT 1 3/4 3-79

" ATTACHMENT G

- FINAL DRAR TABLE 3.3-13 (Continued)

( .

TABLE NOTATIONS

At all times.

- During GASEOUS WASTE PROCESSING SYSTEM operation. ,

ACTION STATEMENTS ACTION 47 - (Not used)

ACTION 48 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided the flow rate is estimated at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

ACTION 49 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided grab samples are taken at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and these samples are analyzed for radioactivity within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

ACTION 50 - (Not used)

ACTION 51 - With the number of channels OPERABLE less than required by the C. Minimum Channels OPERABLE requirement, operation of this GASEOUS WASTE PROCESSING SYSTEM may continue provided grab samples are collected at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and analyzed within the following 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

ACTION 52 - (Not used)

ACTION 53 - With the number of channels CPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via the affected pathway may continue for up to 30 days provided samples are continuously collected with auxiliary sampling equipment as required ir T2'- '.11-?-

L ;n part A of +ka. oocnt..

e SOUTH TEXAS - UNIT 1 3/4 3-86

REACTOP COOLANT SYSTEM f 3/4.4.3 PRESSURIZER L'IMITING C.ONDITION FOR OPERATION 3.4.3 The pressurizer shall be PERABLE with a water volume of less than or equal to 1816 cubic feet, and ft leasVtwo groups of pressurizer heaters supplied by ESF power each having a capacity of at least 175 kW.

APPLICABILITY: MODES 1, 2, and 3. .

ACTION:

a. With only one group pressurizer heaters supplied by ESF power OPERABLE, restore 6t least7two groups to OPERABLE t,tatus within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hotrs and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b. With the pressurizer otheivise inoperable, be in at least HOT STANDBY with the Reactor Tcip System breakers open within 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 RE0VIREMENTS 4.4.3.1 The pressurizer water volume shall be determined 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 .

4.4.3.2 The capacity of each of,the above required groups of pressurizer heaters supplied by ESF power shall be verified by energizing the heaters and reasuring circuit current at least once per 92 days.

e SOUTH TEXAS - UNIT 1 3/4 4-9

ATTACHMENT '

. ST.HL.<AE 2 2 3 Z PAGE 37 OF Ol \

REACTOR COOLANT SYSTEM 3/4.4.4 REllEF VALVES LIMITING CONDITION F0D OPERATION j 3.4.4 All power-operated relief valves (PORVs) and their associated block valves shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3. ,

1 ACTION:

a. With one or more PORV(s) inoperable, because of excessive seat leakage, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months either restore the PORV(s) to OPERABLE status or close the associated block valve (s); 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 ,

b. With one PORV inoperable due to causes other than excessive seat leakage, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months either restore the PORV to OPERABLE status or close the associated block valve and remove power from the block valve; restore the PORV to OPERAB E status within the following 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in HOT STANDBY in the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD 5(

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

c. With both PORV(s) inoperable due to causes other than excessive seat leakage, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months either restore each of the PORV(s) to OPERABLE status or close their associated block valve (s) and remove power from the block valve (s) and be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

d. With one or more block valve (s) inoperable, within 1 hour:

(1) restore the block valve (s) to OPERABLE status, or close the block valve (s) and remove power from the block valve (s), or close the PORV and remove power from the PORV; and (2) apply the ACTION b.

or c. above, as appropriate, for the isolated PORV(s).

e. The provisions of Specification 3.0.4 are not applicable.

SOUTH TEXAS - UNIT 1 3/4 4-10

. S HL A 23 PAGE M OF M REACTOR COOLANT SYSTEM

('. OPERATIONALLEiKAGE SURVEILLANCE REOUIREMENTS (Continued) 4.4.6.2.1 Reactor Coolant System leakages shall be demonstrated to be within each of the above limits by:

yov mon

a. Monitoring the containment atmosphere 4 gaseous radioactivity and par-ticulate radioactivity channels at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ;

b. Monitoring the containment normal sump inventory and discharge at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ;

c. Performance of a Reactor Coolant System water inventory balance at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; and

d. Monitoring the Reactor Head Fl.ange Leakoff System at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

4.4.6.2.2 Each Reactor Coolant System Pressure Isolation Valve specified in Table 3.4-1 shall be demonstrated OPERABLE by verifying leakage to be within its limit:

a. At least once per 18 months,

b. Prior to entering MODE 2 whenever the plant has been in COLD SHUTOOWN for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or more and if leakage testing has not been performed in the previous 9 months,

c. Prior to returning the . valve to~ service following maintenance,

~

repair or replacement work on the valve, and

d. Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following valve actuation due to automatic or manual action or flow through the valve except for valves XRH0060 A,B,C and XRH0061 A,B,C.

e. As outlined in the ASME Code,Section XI, paragraph IW-3427(b).

The provisions of Specification 4.0.4 are not applicable 'for entry into MODE 3 or 4. - '

4 SOUTH TEXAS - UNIT 1 3/4 4-21

ATTACHMENT G

==m flNAL DRAFl TABLE 3.4-1 .-

. REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES

-VALVE NUMBER FUNCTION XSI0007 A, B, C HHSI Cold Leg Injection Check Valves .

(RCS Loops 1, 2, 3)

XSI0009 A, B, C HHSI Hot leg Recirculation Check Valves (RCS Loops 1, 2, 3)

~

XSI0010 A, B, C LHSI/HHSI Hot Leg Recirculation Check Valves (RCS Loops-1, 2, 3)'

XRH0020 A, B, C LHSI Hot Leg Recirculation Check Valves (RCS Loo s 1, 2, 3) lRH XRH0032 A, B, C LHSF, Cold Leg Injection Check Valves '

(RCS Loops 1, 2, 3)

' RHR XSI0038 A, B, C LHSI/HHSI/x,ccumu/

A lator Cold Leg Injection Check Valves (RCS Loops 1, 2, 3)

XSI0046 A, B, C Accumulator Cold Leg Injection Check Valves (RCS Loops 1, 2, 3)

XRH0060 A, B, C RHR Suction Isolation Valves (RCS Loops 1, 2, 3)

XRH0061 A, B, C RHR Suction Isolation Valves (RCS Loops 1, 2, 3)

~

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SOUTH TEXAS - UNIT 1- 3/4 4-22

7. , ~ . , . . . .

ATTACHMENT G N FINAL DRAFT REACTOR COOLANT SYSTEM 3/4.4.8 SPECIFIC ACTIVITY LIMITING CONDITION FOR OPERATION 3.4.8 The specific activity of the reactor coolant shall be limited to:

a. Less than or equal to 1 microcurie per gram DOSE EQUIVALENT I-131, and

b. Less than or equal to.100 6 microcuries per gram of gross radioactivity.

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

ACTION:

MODES 1, 2 and 3*:

a. With the specific activity of the reactor coolant greater than 1 microcurie per gram DOSE EQUIVALENT I-131 for more than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months during one continuous time interval, or exceeding the limit line shown on Figure 3.4-1, be in at least HOT STANDBY with Tavg less than 500*F within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; and

b. With_the gross specific activity of the reactor coolant greater than

,( 100/E microcuries per gram, be in at least HOT STANDBY with T avg less than 500*F within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

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

With the specific activity of the reactor coolant greater than 1 microcuri gram DOSE EQUIVALENT I-131 or greater than 100 6 micro-Curies per ram erform the sampling and analysis requirements of Item 4.a) )(

of Table 4. til the specific activity of the reactor coolant is restored to within its limits.

l i

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With T,yg greater than or equal to 500*F.

SOUTH TEXAS - UNIT 1 3/4 4-26

ATTACHMENT (,

. ST HL AE Z2 34 PAGE G OF 91 i

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ATTACHMENT (,

. ST HL.AE- 2132-PAGE % OF 91 MATERIAL PROPERTY BASIS CONTROLQ NG MATERIAL - RV RT NOT INITIAL: lO'F

% INTEMME)4DIATE SHELL R-1606-3 RT NDT AFTER 32 EFPY COPPER CONTENT: CONSERVATIVELY l/4, gi.F ASSUMED AS 0.IO WTX 3/4T. 64*F

% L CURVE APF)JCABLE FOR HEATUP RATES UP TO IOO* /HR FOR THE SERVICE PERIOD UP TO 32 EFPY AND CONTAINS MARGINS OF IO'F AND 60 PSIG FOR POSSIBLE INSTRUMENT ERRORS 3000 . . ,

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.- O O 100 200 300 400 INDICATED TEMPERATURE (*F)

FIGURE 3.4-2 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS - APPLICABLE UP TO 32 EFPY SOUTH TEXAS - UNIT 1 3/4 4-32

ATTACHMENT G

. ST HL AE.2232. rEkl JGE '/90F 9/

MATERIAL PROPERTY BASIS CONTROLLgG HATERIAL - RV RT NDT INITIAL: 1O*F

/ INTERMENpIATE LHELL R-1606-3 RT NOT AFTER 32 EFPY X C6PPER CONTENTp CONSERVATIVELY 1/4, 91*F j ASSUMED AS 0.10 WT% 3/4T, 64*F !

L SINGLE CURVE APP,1 CABLE FOR COOLDOWN RATES UP TO IOO*/HR FOR THE SERVICE PERIOD UP TO 32 EFPY. AND CONTAINS MARGINS OF IO*F AND 60 PSIG FOR POSSIBLE INSTRUMENT ERRORS 3000 .

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FIGURE 3.4-3 REACTOR COOLANT SYSTEM C00LOOWN LIMITATIONS - APPLICABLE UP TO 32 EFPY SOUTH TEXAS - UNIT 1 3/4 4-33

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FIGURE 3.4-4 NOMINAL MAXIMUM ALLOWABLE PORV SETPOINT FOR THE COLD OVERPRESSURE SYSTEM SOUTH TEXA5 - UNIT 1 3/4 4-37

s.___-- .........._._..................,.. -

ATTACHMENT G

. ST HL AE 223E PB Aa i

PAGEJLOF 91 REACTOR COOLANT SYSTEM

(- - OVERPRESSUREjROTECTIONSYSTEMS

_ SURVEILLANCE REOUIREMENTS 4.4.9.3.1 Each PORV shall be demonstrated OPERABLE by:

a. Performance of an ANALOG CHANNEL OPERATIONAL TEST on the PORV actuation channel, but excluding valve operation, within 31 days prior to entering a condition in which the PORV is required OPERABLE and at least once per 31 days thereafter when the PORV is required OPERABLE; ,

b. Performance of a CHANNEL CALIBRATION on the PORV actuation channel '

at least once per 18 months; and block.

c. Verifying the PORV i:: htien valve is open at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months when the PORY is being used for overpressure protection.

4.4.9.3.2 The RCS vent (s) shall be verified to be open at least once per~

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months

when the vent (s) is being used for overpressure protection. .

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f *Except when the vent pathway'is provided with a valve which is locked, sealed,

( or otherwise secured in the open position, then verify these valves open at least once per 31 days.

SOUTH TEXAS - UNIT 1 3/4 4-38

ATTACHMENT 6 ST HL AE- 2232 PAGE W OF 9/

REACTOR COOLANT SYSTEM 3/4.4.11 REACIOR-VESSEL HEAD VENTS LIMITING CONDITION FOR OPERATION 3.4.11 Two reactor vessel head vent paths /each consisting of two vent valves and a control valve powered from emergency busses shall be OPERABLE and closed.

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

ACTION:

a. With one of the above reactor vessel head vent paths inoperable,

'STARTUP and/or POWER OPERATION may continue provided the inoperable vent patg is maintained closed with power removed from the valve actuator4 of all the vent valves in the inoperable vent path; restore )(

the inoperable vent path to OPERABLE status within 30 days, or, be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b. With two reactor vessel head vent paths inoperablehaintain the - )(

inoperableventpath}closedwithpowerremovedfromthevalveactua- )(

tors of all the vent valves in the inoperable vent paths, and restore at least one of the vent paths to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or

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

SURVEILLANCE REOUIREMENTS 4.4.11 Each reactor vessel head vent path shall be demonstrated OPERABLE at least once per 18 months by:

a. Verifying all manual isolation valves in each vent path are locked in the open position,

b. Cycling each vent valve through at least one complete cycle of full travel from the control room, and

~

c. Verifying flow through the reactor vessel head vent paths during venting.

b SOUTH TEXAS - UNIT 1 3/4 4-40

ATTACHMENT C.

. ST.HL AE. 2 2 32.

PAGE47 0F 9/ ,

3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUEILAT_0RS LIMITING CONDITION FOR OPERATION 3.5.1 Each Safety Injection System accumulator shall be OPERABLE with:

a. The isolation valve open and power removed,

b. A contained borated water volume of between 8800 and 9100 gallons,

c. A boron concentration of between 2400 and 2600 ppa, and

d. A nitrogen cover pressure of between 590 and 670 psig.

APPLICABILITY: MODES 1, 2, and 3*.

ACTION:

a. With one accumulator inoperable, except as a result of a closed isolation valve or the boron concentration outside the required limits, restore the inoperable accumulator to OPERABLE status with I hour or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce pressurizer pressure to less than 1000 psig within the follow-ing 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . ,

b. With one accumulator inoperable due to the isolation valve being closed, either open the isolation valve within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT STANDBY wRh 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce pressurizer pressure to less than 1000 psig withiri the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

c. With the boron concentration of one accumulator outside the required limit, restore the baron concentration to within the required limits within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be Tn at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce pressurizer pressure to less than 1000 psig within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURV ILLANCE REOUIREMENTS 4.5.1.1 Each accumulator shall be demonstrated OPERABLE:

a. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by:

1) Verifying, by the absence of alarms, the contained barated water volume and nitrogen cover pressure in the tanks, and

2) Verifying that each accumulator isolation valve is open.

b. At least once per 31 days and within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after each solution volume increase of greater than or equal to 1% of tank volume by verifying the boron concentration of the accumulator solution; and

" Pressurizer pressure above 1000 psig.

SOUTH TEXAS - UNIT 1 3/4 5-1

ATTACHMENT G ST.HL.AE. 2 23 E PAGE V9 OF 9/

EMERGENCY CORE COOLING SYSTEMS

(

_SURVEILLANCERf0ViREMENTS 4.5.2 Each ECCS subsystem shall be demonstrated OPERABLE:

a. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying that the following valves ,

are in the indicated positions with power to the valve operators removed:

Valve Number Valve Function Valve Position XSId0008A,B,C High Head Hot Leg Closed Recirculation Isolation

.XR 019 A,B,C Low Head Hot leg Closed -

Recirculation Isolation

b. At least once per 31 days by:

1) Verifying that the ECCS piping is full of water by venting the ECCS pump casings and accessible discharge piping high points, and

2) Verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise

( c.

secured in position, is in its correct position.

By a visual inspection which verifies that no loose debris (rags, trash, clothing, etc.) is present in the containment which could be transported to the containment sump and cause restriction of the pump suctions during LOCA conditions. This visual inspection shall be performed:

1) For all accessible areas of the containment. prior to establish-ing CONTAINMENT INTEGRITY, and

2) Of the areas affected within containment at the completion of each containment entry when CONTAINMENT INTEGRITY is established.

d. At least once per 18 months by a visual inspection of the contain-ment _shmp and verifying that the subsystem suction inlets are not restricted by debris and that the sump components (tessh racks, screens,' etc.) show no evidence of structural distress or abnormal corrosion.

(

SOUTH TEXAS - UNIT 1 3/4 5-4

s ..... .... ......... ............. . . ...... ....... .

ATTACHMENT c

. ST HL AE. 22 n FIM AI n r'T PACEjs> 0F 91 g g{ g EMERGENCY CORE COOLING SYSTEMS

(' ,

SURVEILLANCE Rf00lREMENTS (Continued) , ,

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

1) Verifying that each automatjc valve in the flow path actuates e to its correct position on ,S:fety Ir.j::tisc actnetien and ^--aM Automatic Switchover to Containment Sumpq) test signal , and

2) Verifying that each of the following pumps start automatically '

uponreceiptofaSafetyInjectionactuatigtestsignal:

a) High Head Safety Injection pump, and b) Low Head Safety Injection pump. ,

f. By verifying that each of the following pumps develops the indicated differential pressure on recirculation flow when tested pursuant to Specification 4.0.5:

1) High Head Safety Injection pump 1 1480 psid,-and

2) Low Head Safety Injection pump 1 286 psid.

g. By performing a flow test, during shutdown, following completion of

( modifications to the ECCS subsystems that alter the subsystem flow characteristics and verifying that:

1) For H'igh. Head Safety Injection pump lines, with the High Head Safety Injection pump running, the pump flow rate is greater than 1440 gpm and less than 1600 gpm.

.2). . For Low Head Safety Injection pump lines, with the Low Head Safety Injection pump running, the pump flow rate is greater than 2570 gpm and less than 2800 gpm.

e

S -

e e

We-7' . .

h SOUTH TEXAS - UNIT 1 3/4 5-5

ATTACHMENT 6

. ST.HL AE. 2nt pg n PAGE r) 0F 9/

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

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

ACTION:

Without primary CONTAINMENT INTEGRITY, restore CONTAINMENT INTEGRITY within 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 SHUTOOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REOUIREMENTS 4.6.1.1 Primary CONTAINMENT INTEGRITY shall be demonstrated:

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 3.6.3;

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

c. After closing of each penetration subject to Type B testing, except the containment air locks, if opened following a Type A or B test, by leak rate testing the seal with gas at a pressure not less than P,, 37.5 psig, and verifying that when the measured leakage rate for these seals is added to the leakage rates determined pursuant to Specification 4.6.1.2d. for all other Type B and C penetrations, the combined leakage rate is less than 0.60 L,.

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 SHUT 00WN except that such verification need not be performed more of ten than once per 92 days.

SOUTH TEXAS - UNIT 1 3/4 6-1

u.-.-..........,._ _

ATTACHMENT G

. ST.HL AE. 3 2 n FEM &l PAGE 520F 9/

{

CONTAINMENT SYSTEMS C .

SURVEILLANCE RE00litEMENTS (Continued)

b. If any periodic Type A test fails to meet 0.75 L,, the test schedule for subsequent Type A tests shall be reviewed and approved by the ,

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:

Confirms the accuracy of the test by verifying that the supple-1) mental test result, L c, is in accordance with the appropriate following equation:

lLc - R ,, 4 ,)l 10 25 L, where L,, is the measured Type A test leakage end 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; C. and

3) Requires that the rate at which gas is injected into the contain-ment 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,, 37.5 psig, at intervals no greater than 24 months except for tests involving:

1) Air locks,

2) Purge supply and exhaust isolation valves with resilient material seals, o.nd

3) ienetrations using continuous Leakage Monitoring Systems #

e. Air locks shall be tested and demonstrated OPERABLE by the require-ments 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 4.6.1.7.3 or 4.6.1.7.4, as applicable;

(

SOUTH TEXAS - UNIT 1 3/4 6-3

mm ATTACHMENT G CONTAINMENT SYSTEMS FINAL DRAFI

(, .

SURVEILLANCE TE0bfREMENTS (Continued)

b. Performing tendon detensioning, inspections, and material tests on a previously stressed tendon from each group (inverted U and hoop).

A randomly selected tendon from each group shall be completely detensioned in order to identify broken or damaged wires and deter-mining that over the entire length of the removed wire cor strand? *---

that:

1) The tenden wires have not undergone corrosion, cracks, or physical damage in excess of that allowed by ASTM A421-77.

2) There are not changes in the presence or physical appearance of the sheathing filler grease, and '

3) A minimum tensile strength of 240,000 psi (guaranteed ultimate strength of the tendon material) for at least three wire samples (one from each end and one at mid-length) cut from each removed wire. Failure of any one of the wire samples to meet the mini-sum tensile strength test is evidence of abnormal degradation of the containment structure,

c. Performing tendon retensioning of those tendons detensioned for

inspection to their observed lift-off force with a tolerance limit

(.

of +6%. During retensioning of these tendons, the changes in load and elongation should be measured simultaneously at 20%, 60%, and 100%

of the maximum jacking force. If the elongation corresponding to a specific load differs by more than 5% from that recorded during installation, an investigation should be made to ensure that the difference is not related to wire failures or slip of wires in anchorages;

d. Assuring the observed lift-off stresses exceed the average minimum design value given below, which are adjusted to account for elastic and time dependent losses; and Inverted U 126 ksi Hoop: Cylinder 128 ksi

. Dome 123 ksi

e. Verifying the OPERABILITY of the sheathing filler grease by:

1) No voids in excess of 5% of the net duct volume 1,

2) . Minimum grease coverage exists for the different parts of the anchorage system, and

3) The chemical properties of the filler material are within the tolerance limits as specified by the manufacturer.

SOUTH TEXAS - UNIT 1 3/4 6-10

3.............._....._._._-

i E ATTACHMENT G ST.HL AE. 2332 -l PAGE(VOF 91 CONTAINMENT SYSTEMS b.,

SPRAYADDITIdESYSTEM LIMITING CONDITION FOR OPERATION 3.6.2.2 The Spray Additive System shall be OPERABLE with: ,

a. Three spray additive tanks each containing a volume of between 1061 and 1342 gallons of between 30 and 32% by weight NaOH solution, and Three spray additive eductors each capable of adding NaOH solution b.

from its associated' spray additive tank to its Containment Spray System pump flow. .

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

ACTION:

With the Spray Additive System inoperable, restore the system to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; restore the Spray Additive System to OPERABLE status within the

" next 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REOUIREMENTS C~ 4.6.2.2 The Spray Additive System shall be demonstrated OPERABLE:

a. At least once per 31 days by verifying thht each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position;

b. At least once per 6 months by:

1) Verifying the contained solution volume in each spray additive tank, and

2) Verifying the concentration of the NaOH solution by chemical analysis.

c. ,At.Teast once per 18 months during shutdown, by verifying that each automatic valve in the flow path actuates to its correct position on a Containment Pressure High 3 test signal; and L

SOUTH TEXAS - UNIT 1 3/4 6-15

0 ATTACHMENT G

. ST.HL AE 2232-

. PAGE srOF 91 , .

CONTAINMENT SYSTEMS .

SURVEILLANCE RE0VIREMENTS (Continued) . . -

d. At least once per 5 years by verifying:

1) Each eductor suction flow rate is greater than or equal to 30 gpm -

using the RWST as the test source to the eductor inlet, and under the following conditions:

a) CS pump suction pressure -is > 15 psig, .

or as applicsblo 15 b) Valve [CS0019A, @, M3C ee in the full open position, and Pg c) CS pump recirculation flow rate to the RWST is.800 gpm i 100 gpm.

2) The lines between the spray additive tank and the eductors are not blocked by verifying flow.

. ,4 e

e em

, i SOUTH TEXAS - UNIT 1 3/4 6-16

o ATTACHMENT G

. ST HL AE. 2 2.32.

_,PAGEyb OF 9/

CONTAINMENT SYSTEMS a

3/4.6.3 CONTA]NMENT ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3.6.3 The ccatainment isolation valves shall be OPERABLE with isolation times

1ess than or equal to the rec,uired isolation times.

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

ACTION:

With one or more of the isolation valve (s) inoperable, maintain at least one isolation valve OPERABLE in each affected penetration that is open and:

a. Restore the inoperable valve (s) to OPERABLE status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , or

b. Isolate each affected penetration within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months by use of at least one deactivated automatic valve secured in the isolation position, or

c. Isolate each affected penetration within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months by use of at least one closed manual valve or blind flange, or

( d. 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 SHUT 00WN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REOUIREMENTS 4.6.3.1 The isolation valves shall be demonstrated OPERABLE prior to returning the valve to service after maintenance, repair or replacement work is performed on the valve or its associated actuator, control or power . circuit by perform-ance of a cycling test, and verification of isolation time.

TM.+ on pq B/4 fo~l9

. s ho wld ha. moved kova

, de fi.ll py

~

C SOUTH TEXAS - UNIT 1 3/4 6-18 i

ATTACHMENT G

. ST.HL AE. :z 32.

PAGE r7OF 91 CONTAINMENT SYSTEMS _ _ _

FINAL DRAFT SURVEILLANCERfDukREMENTS(Continued) 4.6.3.2 Each isolation valve specified in Teble 3.5 all be demonstrated N OPERABLE during the COLD SHUTDOWN or REFUELING MODE at least once per 18 months by:

a. n Isolation test signal, each Phase ' g that Verifyi,A on a Phase isolation valve "A" actuates to its isolation position; and

b. Verifying that on a Containmont Ventilation Isolation test signal, each purge and exhaust valve actuates to its isolation position.

4.6.3.3 The isolation time of each power-operated or automatic valve shall be determined to be within its limit when tested pursuant to Specification 4.0.5.

SOUTH TEXAS - UNIT 1 3/4 6-19

_ _ _ _ )'

ATTACHMENT C.

. ST.HL.AE. 2 2 32. lM &l J20E.sg 0F 9/ gg 3/4.7 PLANT SYSTEMS 3/4.7.1 TURMN$ CYCLE SAFETY VALVES LIMITING CONDITION FOR OPERATION 3.7.1.1 All main steam line Code safety valves associated with each steam

. generator shall be OPERABLE with lift settings as specified in Table 3.7-2.

APPLICABILITY: MODES 1, 2, and 3.

ACTION:

a. With four reactor coolant loops and associated steam generators in operation and with one or more main steam line Code safety valve inoperable, operation in MODES 1, 2, and 3 may proceed provide that within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , either the inoperable valve is restored to OPER BLE status or the Power Range Neutron Flux High Trip Setpoint is reduced per Table 3.7-1; 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 .

b. The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE REOUIREMENTS 4.7.1.1 No additional requirements other than those required by Specification 4.0.5.

SOUTH TEXAS - UNIT 1 3/4 7-1

ATTACHMENT G k

[E OF 91 PLANT SYSTEMS

(.

AUXILIARY FEf09ATER SYSTEM LIMITING _ CONDITION FOR OPERATION 3.7.1.2 At least four independent steam generator auxiliary feedwater pumps '

and associated flow paths shall be OPERABLE with:

a. Three motor-driven auxiliary feedwater pumps, each capable of being powered from separate emergency busses, and

b. One steam turbine-driven auxiliary feedwater pump capable of being powered from an OPERABLE steam supply system.

APPLICABILITY: MODES 1, 2, and 3.

ACTION: ,

a. With the Train A motor-driven auxiliary feedwater pump inoperable, initiate corrective actions to restore the pump to OPERABLE status as soon as possible. The provisions of Specification 4.0.4 are not applicable,

b. With any of the following combinations of auxiliary feedwater pumps inoperable:

1) Train B or Train C motor-driven pump,

(. 2) Train 0 turbine-driven pump and any one motor-driven pump, Train A and either Train B or Train C motor-driven pump, or 3)

4) Train 0 turbine-driven Nap estore the affected auxiliary feedwater pump (s) to OPERABLE status X

/within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANOBY within the and in HOT SHUTOOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

c. With Train B and Train C motor driven pumps, or any three auxiliary feedwater pumps inoperable, be in at least HOT STANOBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTOOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

d. With four auxiliary feedwater pumps inoperable, immediately initiate corrective action tu restore at least one auxiliary feedwater pump to OPERABLE status as soon as possible.

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 540 gpm;

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 540 gpm when the secondary steam supply pressure is greater than 1000 psig. The provisions of Specification 4.0.4 are not appItcable for entry into MODE 3; SOUTH TEXAS - UNIT 1 3/4 7-4

ATTACHMENT G

. ST.HL AE. 2 23L Pact M 0f.91 PLANT SYSTEMS AUXILIARY FEE 0 WATER 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 S18,000' gallons of water.

APDLICABILITY: 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 next 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 . .

l 23YEILLAliCLEEEIREliElil5 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" tM tsa' !! th:

l verifying the c,ontained n ,m m , m . . ., . .. .. water volume is within its limits.+%a

..m n 4 . - ,-1..... -._,2 5

f O

s e

E SOUTH TEXAS - UNIT 1 3/4 7 6

ATTACHMENT 4 ST.HL AE. 22 32.

PLANT SYSTEMS

[. 3/4.7.4 ESSENTIAL COOLING WATER SYSTEM LIMITING CONDITION FOR OPERATION TNL 3.7.4 hdese4 three 3

independent essenti'l cooling a water loops shall be OPERABLE.

APPLICABILITY: H0 DES 1, 2, 3, and 4.

ACTION:

With only two essential cooling water loop OPERABLE, restore at least three loops to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTOOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE RE0llIREMENTS Each 4.7.4 et % t th, g essential cooling water loops shall be demonstrated OPERABLE:

a. At least once per 31 days by verifying that each valva (manual, power-operated, or automatic) servicing safety-related equipment that is not locked, sealed, or otherwise secured in position is in its correct position;

(* b. At least once per 18 months during shutdown, by verifying that:

1) Each automatic valve servicing safety-related equipment actuates to its correct position on a Safety Injection, ECW pump start, sig-screen wash booster nals, as applicable, pump start and essential chiller start {put-

2) Each Essential Cooling Water pump starts automatically on a SafetyInjectionoraLossofPowertestsignal,and 0Ff5ft

3) Each screen wash booster pump and the traveling screen start automatically on a Safety injection test signal.

s e

SOUTH TEXAS - UNIT 1 3/4 7-13

m_._..._.. . . .

ATTACHMENT 4 ST.HL AE. 2 2 n PAGE a OF 91 PLANT SYSTEMS l 3, C .

BygkVtCE RttulfDENTS (ContinuedL

c. At least once per 18 months or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire, or chemical release in any ventilation zone communi- ,

cating with the system by:

$ v Verifyingthatthemakeupandcleanupsystengsatisfdethe X 1) in-place penetration and bypass leakage testing acceptance criteria of less than 0.05% for HEPA filter banks and 0.10% for charcoal adsorbei banks and uses the test procedure guidance in Regulatory Positions C.S.a. C.S.e, and C.S.d of Regulatory Guide 1.52, Revision 2, March 1978, and the system flow rate is 6000 cfm i 10% for the cleanup units and 1000 cfm i 10% for the ,

makeup units;

2) Verifying, within 31 days after removal, that a laboratory analysis of a representative carbon sample obtained in accor-dance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revi-sion 2, March 1978, for a methyl iodide penetration of less than 1.0% when tested at a temperature of 30'C and a relative humidity of 70%; and

(.' 3) Verifying a system flow rate of 6000 cfm i 10% for the cleanup units and 1000 cfm i 10% for the makeup units during system operation when tested in accordance with AN5! N510-1980.

d. Af ter every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of charcoal adsorber operation, by verifying, within 31 days after removal, that a laboiatory analysis of a repre-sentative carbon sample obtained in accordance with Regulatory Position C.6 b of Regulatory Guide 1.52. Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978, for a methyl iodido penetration of less than 1.0% when tested at a temperature of 30*C and a relative humidity of 70%;

e. At least once per 18 months by:

1)' Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 6.1 inches Water Gauge for the makeup units and 6.0 inches Water Gauge for the cleanup units while operating the system at a flow rate of 6000 cfm i 10% for the cleanup units and 1000 cfm i 17% for the makeup units;

2) Verifying that on a control' room emergency ventilation test signal (High Radiation and/or Safety injection test signal), the system automatically switches into a recirculation and makeup

(% air filtration mode of operation with flow through the HEPA filters and charcoal adsorber banks of the cleanup and makeup units; SOUTH TEXA5 - UNIT 1 3/4 7 17'

ATTACHMENT G

- ST.HL.AE. 22 n o

PAGE 63 0F 91 PLANT SYSTEMS C

SUMEIM.AMCE RQU REMEMILIContinued) ._

Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6.1 of Regulatory Guide 1.52, Revi-sion 2, March 1978, for a methyl iodide penetration of less '

than 1.0% when tested at a temperature of 30'C and a relative humidity of 70%; and .pm w, of +he,

+hre.e.

3) Verifying a system flow rate of 29,000 cfm i 10% durin system operation with two of the three exhaust booster and main exhaust fans operating when tested in accordance with ANSI N510-1980.

All combinations of two exhaust booster fans and two main exhaust fans shall be tested.

c. Af ter every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of charcoal adsorber operation, by. verifying, within 31 days after removal, that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978, for a methyl iodide penetration of less than 1.0% when tested at a temperature of 30*C and a relative humidity of 70%;

d. At least once per 18 months by:

1) Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 6 inches Water Gauge while operating the system at a flow rate of 29,000 cfm i 10%,

2) Verifying that the system starts on High Radiation and Safety Injection test signalsp.ond dirac.ts .40 +hrou3b ha HEPA filfers and 3)' Verifying that the system maintains the FHB at a negative Ch g* *J pressure of greater than or equal to 1/8 inch Water Gauge %>

relative to the outside atmosphere, and * " -

4) Verifying that the heaters dissipate 50 1 5 kW when tested in accordance with ANSI N510-1980.

e. Af ter each complete or partial replacement of a HEPA filter bank, by verifying that the HEPA filter bank satisfies the in-place pene-tration and bypass leakage testing acceptance criteria of less than 0.05%'in accordance with ANSI N510-1980 for a 00P test aerosol while operating the system at a flow rate of 29,000 cfm i 10%; and

f. After each complete or partial replacement of a charcoal adsorber bank, by verifying that the charcoal adsorber bank satisfies the in-place penetration and bypass leakage testing acceptance criteria of less than 0.10% in accordance with ANSI N510-1980 for a halogenated hydrocarbon refrigerant test gas while operating the system at a flow rate of 29,000 cfm i 10%.

k SOUTH TEXAS - UNIT 1 3/4 7 20

ATTACHMENT G p

. STcHL AE 22 32- ed PAGLW OF 9 / l 0p D TABLE 3.7 AREA TEMPERATURE MONITORING .

AREA TEMPERATURE LIMIT (*F)

1. Relay Room . < 78 (Electrical Auxiliary Building E1. 35'0")

2. Switchgear Rooms < 85 (Electrical Auxiliary Building E1. 10'0",

35'0",60'0") .

3. Electrical Penetration Spaces -

< 103 (Electrical Auxiliary Building E1. 10'0",

35'0",60'0") ,

4. Safety Injection and Containment Spray 1 101

. Pump Cubicles (Fuel Handling Building E1. -29'0")

5. Component Cooling Water Pump Cubicles < 112 (Mechanical Auxiliary Building E1. 10'0")

6. Centrifugal Charging Pump Cubicles < 132 (Mechanical Auxiliary Building E1. 10'0")

C' 7. Hydrogen Analyzer Room -

< 102 (Mechanical Auxiliary Building E1. 60'0")

8. Boric Acid Transfer Pump Cubicles < 101 (Mechanical Auxiliary Building El. 10'0")

9. Standby Diesel Generator Rooms -

< 101*

(Diesel Generator Building E1. 25'0")

10. Essential Cooling Water Pump Rooms < 101 (Essential Cooling Water Cdicir Intake., THrhe+vre.

El. 34'0")

11. Isolation Valve Cubicles < 101 (Isolation 1/alve Cubicle E1.10' 0")

12. Qualified Display Processin System Rooms 1 94**

(Electrical Auxiliary Build ng El. 0")

10

Temperature limit is < 120*F when testing the standby diesel generator pursuant to Surveillaiice Requirement 4.8.1.1.2.e.7).

k ** Measurement inside QDPS auxiliary processing cabinets.

SOUTH TEXAS - UNIT 1 3/4 7-32

ATTACHMENT 4

- ST.HL AE.2332 PAGE ur0F 91 PLANT SYSTEMS ,

(I 3/4.7.14 ESSENTIAL CHILLED WATER SYSTEM LIMITING CONDITION FOR OPERATION Tha.

3.7.14 At least three 3 independent Essential Chilled Water System loops shall ,

be OPERABLE.

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

ACTION:

With only two Essential Chilled Water System loops OPERABLE, restore three loops to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6. hours and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . .

SURVEILLANCE REOUIREMENTS 4.7.14 The Essential Chilled Water System shall be demonstrated OPERABLE by:

a. Performance of surveillances as required by Specification 4.0.5, and

b. At least once per 18 months by demonstrating that the system starts automatically on a Safety Injectiong ; tettier, signal.

( -tast e

e swe I

e s

, , 88 *

. , ( *$3 SOUTH TEXAS - UNIT 1 3/4 7-33

ATTACHMENT G

- ST.HL.AE. 2 2 32 PAGE 440F 9/

( 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1 A.C.. SOURCES OPERATING LIMITING CONDITION FOR OPERATION 3.8.1.1 As a minimum, the following A.C. electrical power sources shall be OPERABLE:

a. Two physically indepe.ndent circuits between the offsite transmission network and the onsite Class IE Distribution System **, and

b. Three separate and independent standby diesel generators, each with

,a separate fuel tank containing a minimum volume of 60,500 gallons .

a of fuel.

APPLICABILITY: MODES'1, 2,13, and 4.

ACTION:

a. With one offsite circuit.of the above-required A.C. electrical power sources inoperable, demonstrate the OPERABILITY of the remaining A.C.

r sourceO by performing Surveillance.. Requirement.4.8.1.1.1.a within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter. Demonstrate the OPERABILITY of each standby diesei generator that has not been suc-C"

"" ' d "i'h'" '"' ' " t 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by performing Survesiiance Requirement 4.8.1.1.2.a.2) for each such standby diesel generator, separately, within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Restore the offsite circuit to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWM within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

b. With a standby diesel generator inoperable, demonstrate the OPERABILITY of the above-required A.C. offsite sources by performing Surveillance Requirement 4.8.1.1.1.a within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter. If the standby diesel generator became in-operable due to any cause other than preplannned preventive main-tenance or testing, demonstrate the OPERABILITY of the remaining l ,

OPERABLE standby diesel generators by performing Surveillance Require-

. ment 4.8.1.1.2.a.2) and for each such standby diesel generator, I separately, within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .* Restore the inoperable standby diesel i l 4 generator to OPERABLE states within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT !

l SHUTDOWN within the next 12' hours and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . ;

l With#one o'ffsite circuit and on'e sta'ndby diesel generator of the above

c. ~

required A.C. electrica'l power sources inoperable, demonstrate the OPERABILITY of the remaining A.C. sources by performing Specifica- l l

tion 4.8.1.1.la.-within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months there- ,

after; and if the -standby diesel generator became inoperable due to '

' ' *This test is required to be completed regardless- of when the inoperable standby i{ diescl generator is restored to OPERABILITY.

- Loss of one 13.8 kV Standby bus to 4.16 kV F.SF bus line constitutes loss of one i

offsite source. Loss of two 13.8 kV Standby busses to 4.16 kV ESF bus lines constitutes loss of two offsite sources.

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

ATTACHMENT G

. ST44L AE. 22 32 PAGEG1 0F 91 ELECTRICAL POWER SYSTEMS D.C. SOURCES SHUTDOWN LIMITING CONDITION FOR OPERATION 3.8.2.2 As a minimum, either Channel I or Channel IV 125-volt battery bank and

% % + two associated chargers shall be OPERABLE.

APPLICABILITY: MODES 5 and 6.

ACTION:

With the required battery bank and/or charger (s) inoperable, immediately sus-pend all operations involving CORE ALTERATIONS, positive reactivity changes, or movement of irradiated fuel; initiate corrective action to restore the required battery bank and/or chargers to OPERABLE status as soon as possible, and within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , depressurize and vent the Reactor Coolant System through a 2.0 square inch vent.

SURVEILLANCE REOUIREMENTS 4.8.2.2 The above required 125-volt battery bank and chargers shall be demonstrated OPERABLE in accordance with Specification 4.8.2.1.

l l

l 1

1 SOUTH TEXAS - UNIT 1 3/4 8-13

.m .._,-,.v-_., _ . _ . . . y

ATTACHMENT G

. ST HL AE. 2as2.

PAGEI,7 OF91 _

ELECTRICAL POWER SYSTEMS p':

3/4.8.4 ELECTRIC L- EQUIPMENT PROTECTIVE DEVICES CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES I

LIMITING CONDITION FOR OPERATION 3.8.4.1 For each containment penetration ^prov ed with a penetration conductor overcurrent protective device (s), each device s shall be OPERABLE. -

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

ACTION: ,

With one or more of the containment penetration conductor overcurrent protective device (s) given in lable 3.8-ginoperable:

4

a. Restore the protective device (s) to OPERABLE status or deenergize the circuit (s) by tripping the associated backup circuit breaker or racking out or removing the inoperable circuit breaker within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , declare the affected system or component inoperable, and verify the backup circuit breaker to be tripped or the inoper-able circuit breaker racked out or removed at least once per 7 days p thereafter; the provisions of Specification 3.0.4 are not applicable

\- . to overcurrent devices in circuits which have their-backup circuit breakers tripped, their inoperable circuit breakers racked'ou or removed, or

b. 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 REOUIREMENT,L__

4.8.4.1 Protective devices required to be OPERABLE as containment penetration !

conductor overcurrent protective devices shall be demonstrated OPERABLE:

1

a. At least once per 18 months:

1)- By verifying that the medium voltage 13.8 kV circuit breakers are OPERABLE by selecting, on a rotating basis, at least 10% of the circuit breakers of 4acMvoltage level, and performing the following: -A .

a) A CHANNEL CALIBRATION of the associated protective relays, b) An integrated system functional test which includes simulated automatic actuation of the system and verifying that each relay and associated circuit breakers and control circuits

( function as designed, and '

1 SOUTH TEXAS - UNIT 1 3/4 8-17 l

ATTACHMENT G

. ST HL-AE- 2 2 31 Pl El PAGE GS OF 91 REFUELING OPERATIONS 3/4.9.11 WATER LEVEL - STORAGE POOLS g

SPENT FUEL POOL LIMITING CONDITION FOR OPERATION .

3.9.11.1 At least 23 feet of water shall be maintained over the top of irradiated fuel assemblies seated in the storage racks.

APPLICABILITY: Whenever irradiated fuel assemblies are in the spent fuel pool.

ACTION:

a, With the requirements of the above specification not satisfied, .

suspend all movement of fuel assemblies and crane operations with -

loads in the fuel storage areas and restore the water level to within its limit within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REOUIREMENTS 4.9.11.1 The water level in the spent fuel pool shall be determined to be at least i

. (['9 its minimum required depth at least once per 7 days when irradiated fuel assemblies are in the spent fuel pool.

aa*

9 e

s SOUTH TEXAS - UNIT 1 3/4 9-12

ATTACHMENT 4

. ST.HL-AE- 3 2 32- -

REFUELING OPERATIONS (1

EURVEILLANCE RIOUIBEMENTS (Continued)

b. At -least once per 18 months and (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire, or chemical release in any ventilation zone ,

communicating with the system by:

1) Verifying that the cleanup system satisfies the in place penetration and bypass leakage testing acceptance criteria of less than 0.05% for HEPA filter banks and 0.10% for char-coal adsorber banks and uses the test procedure guidance in Regulatory Positions C.5.a. C.5.c, and C.S.d of Regulatory Guide 1.52, Revision 2, March 1978, and the system flow rate is 29,000 cfm i 10%; .

2) Verifying, within 31 days after removal, that a lab' oratory analysis of a representative carbon sample obtained in accor-dance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978, for a methyl iodide penetration of less than 1.0%

when tested at a temperature of,30*C and a relative humidity of 70%; and f,m3 he of h

/ evu_.

(,' 3) Verifying a system flow rate of 29,000 cfm i 10%Iduring system s operationwithtwoofthethreeexhaustbooster/andgnffnexhaust fans operating when tested in accordance with ANSI N510-1980.

All combinations of two exhaust booster fans and two main exhaust fans shall be tested.

c. After every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of charcoal adsorber operation by verifying, within 31 days after removal, that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978, for a methyl iodide penetration of less than 1.0% when tested at a temperature of 30 C and a relative humidity of 70%.

d. At least once per 18 months by:

1) Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 6 inches Water Gauge while operating the system at a flow rate of 29,000 cfm i 10%,

2) Verifying that on a High Radiation test signal, the system automatically starts (unless already operating) and directs its exhaust flow through the HEPA filters and charcoal adsorber banks, b.

SOUTH TEXAS - UNIT 1 3/4 9-15

!EFFi -

FINAL DRAFT SPECIAL TEST EXCEPTIONS

("a '

3/4.10.4 RE5CTORCOOLANTLOOPS LIMITING CONDITION FOR OPERATION 3.10.4 The limitations of Specification 3.4.1.1 may be suspended during the ,

performance of STARTUP and PHYSICS TESTS ~provided:

a. The THERMAL POWER does not exceed the P-7 Interlock Setpoint, and l l

b. The Reactor Trip Setpoints on the OPERABLE Intermediate and Power Range channels are set less than or equal to 25% of RATED THERMAL POWER.

APPLICABILITY: During operation below the P-7 Interlock Setpoint.

ACTION:

With the THERMAL POWER greater than the P-7 Interlock Setpoint, immediately open the Reactor trip breakers.

SURVEILLANCE REOUIREMENTS

. 4.10.4.1 The THERMAL POWER shall be determined to be less than P-7 Interlock Setpoint at least once per hour during STARTUP and PHYSICS TESTS.

4.10.4.2 Each Intermediate and Power Range channel and P-7-Interlock shall be subjected to an ANALOG CHANNEL OPERATIONAL TEST within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months prior to initiating STARTUP and PHYSICS TESTS.

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e C

SOUTH TEXAS - UNIT 1 3/4 10-4

ATTACHMENT (c

. ST.HL AE H32.

. PAGE'7A OF 'll RADIOACTIVE EFFLUENTS b GASEOUS WASTE PROCESSING SYSTEM LIMITING COND TION FOR OPERATION 3.11.2.4 The GASE0US WASTE PROCESSING SYSTEM shall be OPERABLE and appropriate '

portions of this system shall be used to reduce releases of radioactivity when the projected doses in 31 days due to gaseous effluent releases, from each unit, to areas at and beyond the SITE BOUNDARY,(see Figure 5.1-3) would exceed:

a. 0.2 mrad to air from gamma radiation, or

b. 0.4 mrad to air from beta radiation, or

c. 0.3 mrem to any organ of a MEMBER OF THE PUBLIC. ,

APPLICABILITY: At all times.

ACTION:

a. With radioa,ctive gaseous waste being discharged without treatment and in excess of the above limits, prepare and submit to the Commission .within 30 days, pursuant to Specification 6.9.2, a Special Report that includes the following information:

1. " Identification of any inoperable equipment or subsystems, and

{- the reason for the inoperability,

2. Action (s) taken to restore the inoperable equipment to OPERABLE status, and

3. Summary description of action (s) taken to prevent a recurrence,

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REOUIREMENTS 4.11.2.4.1 Doses due to gaseous releases from each unit to areas at and beyond the SITE BOUNDARY shall be projected at least once per 31 days in accordance with-the methodology and parameters in the ODCM when GASEOUS y

WASTE PROCESSING 5YSTEM is not being fully utilized.

Lh 4.11.2.4.2 The installed GASEOUS WASTE PROCESSING SYSTEM shall be considered OPERABLE by meeting Specifications 3.11.2.1 and 3.11.2.2 or 3.11.2.3.

SOUTH TEXAS - UNIT 1 3/4 11-8

ATTACHMENT 4

. ST.HL.AE 2 2 32 PAGE13 0F O RADI0 ACTIVE EFFLUENTS EXPLOSIVE GAS MIXTURE LIMITING CONDITION FOR OPERATION 3.11.2.5 The concentration of oxygen in the GASE0US WASTE PROCESSING SYSTEMyshall be limited to less than or equal to 3% by volume. x Mrtlef APPLICABILITY: At all times.

ACTION: g

a. With the concentration of oxygen in the GASEOUS WASTE PROCESSING SYSTEM exceeding the limit, restore the concentration to within the X limit within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REOUIREMENTS 4.11.2.5 The concentration of oxygen in the GASE0US WASTE PROCESSING SYSTEM shall be determined to be within the above limits by continuously monitorir:g )(

the waste gases the GASEOUS WASTE PROCESSING SYSTEM with the oxygen monitor required OPERABLE by Table 3.3-13 of Specification 3.3.3.11.

EA h in SOUTH TEXAS - UNIT 1 3/4 11-9

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ATTACHMENT 4 ST-HL.AE- M 3 L PAGE 9# 0F H i REACTIVITY CONTROL SYSTEMS

(. BASES MODERATOR TEMPERATURE COEFFICIENT (Continued) 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 a transformed into the limiting MTC value -4.0 x 10 4 ok/k/*F. The MTC value o f -3.1 x 10 4 Ak/k/*F 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 to the limiting MTC value of -4.0 x 10 4 Ak/k/*F.

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 Systeni average temperature less than 561*F. This

. M ~ limitation (1) the moderator temperature coefficient is within$isanalyzedrequiredtemperature tonensure: range, (2) the trip instrumentation is within its normal operating range, (3) the pressurizer is capr'le of being in an OPERABLE status with a steam bubble, and (4) the reactor vessel is above its minimum RT NDT temperature. .

I 3/4.1.2 B0 RATION SYSTEMS-C The Boron Injection System ensures that negative reactivity control is available during each mode of facility operation. The components required to r perform this function include: (1) borated water sources, (2) charging pumps, l

(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 1.75% Ak/k after xenon decay and cooldown to 200'F. The maximum expected boration capability requirement occurs at EOL from full power equilibrium xenon conditions and requires 27,000 gallons of 7000 ppa borated water from the boric acid storage system or 458,000 gallons of 2500 ppe borated water from the refueling water storage l tank -(RWST).- The RWST volume is an ECCS requirement and is more than adequate for the required boration capability. 4],e p.&/ so oxx With the RCS temperature below 350'F, one acceptable without single failure consideratio/oron/njection3ptr n on the basis of the stable is reactivity condition of the reactor and the additional restrictions prohib,iting

' CORE ALTERATIONS and positive reactivity changes in the event the singleforon

/njection g g gones inoperable.

The limitation for a maximum of one charging pump to be OPERABLE and the

. Surveillance Requirement to verify all charging pumps except the required OPERABLE pump to be inoperable below 350'F provides assurance that a masis addi-

{' tion pressure transient can be relieved by the operation of a single PORV.

SOUTH TEXAS - UNIT 1 B 3/4 1-2

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

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m o ATTACHMENT 4 REACTIVITY CONTROL SYSTEMS FINAL HAFT h BASES BORATION SYSTEMS (Continued) bornhe The, M n capability required below 200*F is sufficient to provide a variable SHUTOOWN MARGIN based on the results of a boron dilution accident analysis where ,

the SHUTDOWN MARGIN is varied as a function of RCS boron concentration after xenon decay and cooldown from 200*F to 140*F. This condition requires either 2900 gallons of 7000 ppm borated water from the boric acid storage system or 122,000 gallons of 2500 ppm borated water from the RWST for MODE 5 and 33,000 gallons of 2500 ppm borated water from the RWST for MODE 6.

The contained water volume limits include allowance for water not available because of discharge line location and other physical characteristics.

The' limits on contained water volume and boron concentration.of the RWST also ensure a pH value of between 7.5 and 10.0 for the solution recirculated within containment after a LOCA. This pH band minimizes the evolution of iodine and minimizes the effect of chloride and caustic stress corrosion on mechanical systems and components.

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

3/4.1.3 MOVABLE CONTROL ASSEMBLIES The specifications of this section ensure that: (1) acceptable power distri-bution limits are maintained, (2) the minimum SHUTDOWN MARGIN is maintained, and (3) the potential effects of rod misalignment on associated accident analyses are limited. OPERABILITY of the control rod position indicators is required to determine control rod positions and thereby ensure compliance with the control rod alignment and insertion limits. Verification that the Digital Rod Position Indicator agrees with the demanded position within i 12 steps at 24, 48, 120, and 259 steps withdrawn for the Control Banks and'18, 210, and 259 steps with-drawn for the Shutdown Banks provides assuranc.es that the Digital Rod Position Indicator is operating correctly over the full range of indication. Since the

~

Digital Rod Position Indication System do'es not indicate the actual shutdown rod position between M steps and 210 steps, only points in the indicated ranges are picked for vetification of agreement with demanded position.

The ACTI0tpstatements which permit limited variations from the basic requirements are accompanied by additional restrictions which ensure that the original design criteria are met. Misalignment of a red requires measurement l of peaking factors and a restriction in THERMAL POWER. These restrictions pro-I vide assurance of fuel rod integrity during continued operation. In addition, those safety analyses affected by a misaligned rod are reevaluated to confirm that the results remain valid during future operation.

The maximum rod drop time restriction is consistent with the assumed rod drop time used in the safety analyses. Measurement with T,yg greater than or equal to 561*F and with all reactor coolant pumps operating ensures that the

{ measured drop times will be representative of insertion times experienced during a Reactor trip at operating conditions.

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

ATTACHMENT G

- ST-HL AE 2232.

POWER DISTRIBUTION LIMITS Em M BASES -

HEAT FLUX HOT CHANNEL FACTOR and NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR (Continued)

When an qF measurement is taken, an allowance for both experimental error 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 RTPq F limit for RATED THERMAL POWER (F ) as provided in the Radial Peaking xy Factor Limit Report per Specification 6.9.1.6 was determined from expected power control manuevers over the full range of burnup conditions in the core.

The 12-hour periodic surveillance of indicated RCS flow is sufficient to N" U detect only flow degradation which could lead to operation outside the require- hy ents of Specification 3.2.3# M 03/y.a.s-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 ONB 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.

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1 1

SOUTH TEXAS - UNIT 1 B 3/4 2-5'

ATTACHMENT G ST.HL AE 2 232 i

L_ PAGE 77 OF91 POWER DISTRIBUTION LIMITS k .. -

BASES ,

QUADRANT POWER TILT RATIO (Continued)

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 sisaligned control rod. In the event such action ::t h ."does not correct the tilt, the margin for uncertainty on F isg reinstated by reducing the maximum allowed power by 3% for each percent of tilt in excess of 1.

For purposes of r.onitoring 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 cf four symmetric th'imbles. The two sets of four symmetric thimbles is a unique set of eight detector locations. Th_ese 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 g,q initial FSAR assumptions and have been analytically demonstrated adequate to maintain a minimum DNBR of 1.30 throughout each analyzed transient. The value of 598 F and the indicated pressurizer pressure value of indicated T,yg 2201 psig are provided assuming that the readings from four channels will be averaged before comparing with the required limit. The flow requirement (395,000 gpm) includes a measurement uncertainty of 3.5%.

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.

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f i SOUTH TEXAS - UNIT 1 B 3/4 2-6

ATTACHMENT G

. ST.HL. AE- 22 32- -

PAGE77 0F il INSTRUMENTATION BASES _

REACTOR TRIP SiSTEM and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION (Continued) is the difference, in percent span, between the trip setpoint and the value used in the analysis for the actuation. R or Rack Error is the "as measured" deviation, in the percent span, for the affected channel from the specified Trip Setpoint. S or Sensor Error is either the "as measured" deviation of the sensor from its calibration point or the value specified in Table 3.3-4, in percent span, from the analysis assumptions. Use of Equation 2.2-1 allows for a sensor drift factor, an increased rack drift factor, and provides a threshold value for REPORTABLE EVENTS.

The methodology to derive the Trip Setpoints is based upon combining all of the uncertainties in the channels. Inherent to the determination of the Trip Setpoints are the magnitudes of these channel uncertainties. Sensor and rack instrumentation utilized in these channels are expected to be capable of cperating within the allowances of these uncertainty magnitudes. Rack drift in excess of the Allowable Value exhibits the behavior that the rack has not met its allowance. Being-that there is a small statistical chance that this will happen, an infrequent excessive drift is expected. Rack or sensor drift,.

in excess of the allowance that is more than occasional, may be indicative of more serious problems and should warrant further investigation, b The measurement of response time at the specified frequencies provides assurance that the Reactor trip and the Engineered Safety Features actuation associated with each channel is completed within the time limit assumed in the safety analyses. No credit was taken in the analyses for those channels with response times indicated as not applicable. Response time may be demonstrated by any series of sequential, overlapping, or total channel test measurements provided that such tests demonstrate the total channel response time as defined.

Sensor response time verification may be demonstrated by either: (1) in place, onsite, or offsite test measurements, or (2) utilizing replacement sensors with certified response times. .

The Engineered Safety Features Actuation System senses selected plant parameters and determines whether or not predetermined limits are being exceeded.

If they are, the signals are combined into logic matrices sensitive to combina-tions indicative,of various accidents, events, and transients. Once the required logic tombination is completed, the system sends actuation signals to those Engineered Safety Features components whose aggregate function best serves the requirements of the ' condition. As an example, the following actions may be initiated by the Engineered Safety Features Actuation System to mitigate the consequences of a steam line break or loss-of-coolant accident: (1) Safety Injection pumps. start, (2) Reactor trip, (3) feedwater isolation, (4) startup of the standby diesel generators, (5) containment spray pumps start and auto-matic valves position, (6) containment isolation, (7) steas line isolation, (8) Turbine trip, (9) a~uxiliary feedwater pumps start and automatic valves position, (10) reactor containment fan coolers start H e ? W 1 9 M W-e

{ y (11) essential cooling water pumps start and a,utomatic valves position, SOUTH TEXAS - UNIT 1 B 3/4 3-2

f o ATTACHMENT 6

. ST.HL AE.2z32. l PAGE 79 0F Tl INSTRUMENTATION RASES REACTOR TRIP S'YSTEM and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION (Continued)

(12) Control Room N hti c = Y. Ventilation Systems start, and (13) component cooling water pumps start and automatic valves position. ,

The Engineered Safety Features Actuation System interlocks perform the following functions: Via P-l(o <

closes main feedwater I P-4 Reactortripped-ActuatesTurbinetrip[theopeningofthesain valves on T,yg below Setpoint, prevents I

feedwater valves which were closed by a Safety Injection or High Steam Generator Water Level or Excessive Cooldown Protection signal, ;

allows Safety Injection block so that components can be reset or ;

tripped, and actuates P-15.

Reactor not tripped prevents manual block of Safety' Injection.

P-11 On increasing pressurizer pressure, P-11 automatically reinstates des -h n*n*I ) Safety Injection cooldown actuation signals, on low and opens the pressurizer pressureisolation accumulator discharge or excessive igax f sta al'"' sge, f g valves. On decreasing pressure, P-11 allows the manual block of h*M'** '".*0

$3M) Safety Injection cooldown actuation signals,+and on lowsteam enables pressurizer pressure line isolation ornegative on high excessive d***"

steam line pressure rate.

P-12 On increasing reactor coolant loop temperature, P-12 automatically

@- provides an arming signal to the Steam Dump Systes. On. decreasing ,

reactor coolant loop temperature, P-12 automatically removes the arming signal f g t y tgam Dump Sy g.g P-14 steam generator wate evel, P-14 automatically trips Onincreasing[feedwaterpumps,and the turbine 5/ 11.feedwater . isolation valves and

>Ptit; feedwater control valve,5rtigti=c-P-15 When the reactor is tripped-(P-4) or Nhen below the power range neutron flux setpoint, P-15 is present and' allows. Safety Injection actuation on Low-Low Tcold and allows feedwater isolation and turbine trip from Low Compensated Tcold or high feedw'af.er flow. g g, g 3/4.3.3 MONITORING INSTRUMENTATION N" ['

3/4.3.3.1. RA'DIATION MONITORING FOR PLANT OPERATIONS The OPERA 5ILITY of the radiation monitoring instrumentation for plant operations ensures that: (1) the associated action will be initiated when the radiation level monitored by each channel or combination thereof reaches its Setpoint, (2) the specified coincidence logic is maintained, and (3) suffi-cient redundancy is maintained to permit a channel to be out of service for testing or maintenance. The radiation monitors for plant operations. sense radiation levels in selected plant systems and locations and determine whether or not predetermined limits are being exceeded. If they are, the signals are combined into logic matrices sensitive to combinations indicative of various

( accidents and abnormal conditions. Once the required logic combination is completed, the system sends actuation signals to initiate alarus or automatic isolation action and actuation of Emergency Exhaust or Ventilation Systems.

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

ATTACHMENT 6 REACTOR COOLANT SYSTEM

== FINAL HAFT

' ~

BASES CHEMISTRY (Continued) the chemistry within the Steady-State Limits provides adequate corrosion protection to ensure the structural integrity of the Reactor Coolant System over the life of the plant. The associated effects of exceeding the oxygen, chloride, and fluoride limits are time and temperature dependent. Corrosion studies show that operation may be continued with contaminant concentration levels in excess of the Steady-State Limits, up to the Transient Limits, for the specified limited time intervals without having a significant effect on the structural integrity of the Reactor Coolant System. The time interval permitting continued operation within the restrictions of the Transient Limits provides time for taking corrective actions to restore the contaminant concen-trations to within the Steady-State Limits.

The Surveillance Requirements provide adequate assurance that concentrations in excess of the limits will be detected in sufficient time to take corrective action.

3/4.4.8 SPECIFIC ACTIVITY The limitations on the specific activity of the reactor coolant ensure that the resulting 2-hour doses at the SITE BOUNDARY will not exceed an appropriately small fraction of 10 CFR Part 100 dose guideline values following a steam generator tube rupture accident in conjunction with an assumed steady-state reactor-to-secondary steam generator leakage rate of 1 gpm. The values for the limits on specific activity represent limits based upon a parametric evaluation by the NRC of typical site locations. These values are conservative in that specific site parameters of the STPEGS site, such as SITE BOUNDARY location and meteorological conditions, were not considered in this evaluation.

The ACTION statement permitting POWER OPERATION te continue for limited time periods with the reactor coolant's specific activity greater than 1 microcurie / gram DOSE EQUIVALENT I-131, but within the allowable limit shown on Figure 3.4-1, accomodates possible iodine spiking phenomenon which may occur following changes in THERMAL POWER.

The sample . analysis for determining the gross specific activity and I can exclude the radiciodines because of the low reactor coolant limit of 1 microcurie /

gram DOSE EQUIVALENT I-131, and because, if the limit is exceeded, the I radioiodine level is to be determined every 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . If the gross specific activity level and radiciodine level in the reactor coolant were at their limits, the radioiodine contribution would be approximately 1%. In a release of reactor coolant with a typical mixture of radioactivity, the actual radio-iodine contribution would probably be about 20%. The exclusion of radio-nuclides with half-lives less than 0 minutes from these determinations has lb l

SOUTH TEXAS - UNIT 1 B 3/4 4-5

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ATTACHMENT A

. ST.HL.AE. 22 32 ma PAGEti OF9)

REACTOR COOLANT SYSTEM BASES SPECIFIC ACTIVITY (Continued) $ '

s. The first consideration is the difficulty to been made for several rea identify short-lived r d onuclides in a sample that requires a significant time to collect, tr port, and analyze. The second consideration is the predictable delay ime between the postulated release of radioactivity from i the reactor coo nt to its release to the environment and transport to the SITE BOUND , which is relatable to at least 30 minutes decay time. The choice of 0 inutes for the half-life cutoff was'made because of the nuclear characteris ics of the typical reactor coolant radioactivity. The radionuclides in the. typical reactor coolant have half-lives of less than 4 minutes or '

half-lives of greater than 14 minutes, which lows a distinction between the .

radionuclides above and below a half-life of 0 minutes. For these reasons the radionuclides that are excluded from conside tion are expected to decay

.o very low levels before they could be transporte f om the reactor coolant to the SITE BOUNDARY under any. accident condition. y3 Based upon the above considerations for excluding certain radionuclides from the sample analysis, the allowable time of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months between sample taking and comp.leting the initial analysis is based upon a typical time necessary to perform the sampling, transport the sample, and perform the analysis of about 90 minutes. Af ter 90 minutes, the gross count should be made in a reproducible

{ geometry of sample and counter having reproducible beta or gamma self-shielding properties. The counter should be reset to a reproducible efficiency versus energy. It is not necessary to identify specific nuclides. The radiochemical determination of nuclides should be based on multiple counting of the sample within typical counting basis following sampling of less than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , about 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , about 1 day, about I week, and about 1 month.

Reducing T to less than 500*F prevents the release of activity should asteamgeneratN9 tube rupture since the saturation pressure of the reactor coolant is below the lift pressure of the atmospheric steam relief valves. .

The Surveillance Requirements provide adequate assurance that excessive specific -

activity levels in the reactor coolant will be detected in sufficient time to take corrective action. A reduction in frequency of isotopic analyses following power changes may be permissible if justified by the data obtained.

3/4.4.9 PRESSURE / TEMPERATURE LIMITS The temperature and pressure changes during heatup and cooldown_are

-limited to be consistent with the requirements given in the ASME Boiler and Pressure Vessel Code,Section III, Appendix G:

1. The reactor coolant temperature and pressure and system heatup and cooldown.

rates (with the exception of the pressurizer) shall be limited in accordance with Figures 3.4-2 and 3.4-3 for the service period specified thereon:

a. Allowable combinations of pressure and temperature for specific temperature change rates are below and to the right of the limit lines shown. Limit lines for cooldown rates between those presented may be obtained by interpolation; and SOUTH TEXAS - UNIT 1 B 3/4 4-6 ..

ATTACHMENT G

. ST.HL AE 2234 PAGET2 OF 91 l REACTOR COOLANT SYSTEM BASES m _

LOW TEMPERATUR'E OVERPRESSURE PROTECTION (Continued) safety injection pumps while in MODE 5 and MODE 6 with the reactor vessel head on. All but one high head safety injection pump are required to be locked out in MODE 4. Technical Specifications also require lockout of all but one charging pump while in MODES 4, 5, and 6 with the reactor vessel head installed and disallow start of an RCP if secondary temperature is more than 50*F above primary temperature.

The Maximum Allowed PORV Setpoint for the COMS will be updated based on the results of examinations of reactor vessel material irradiation surveillance specimens performed as required by 10 CFR Part 50, Appendix H, and in accordance with the schedule in Table 4.4-5.

3/4.4.10 STRUCTURAL INTEGRITY The inservice inspection and testing programs for ASME Code Class 1, 2, and 3 components ensure that the structural integrity and operational readiness of these components will be maintained at ari'heceptable level throughout the life of the plant. These programs are in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and_ applicable Addenda as required by 10 CFR 50.55a(g) except where specific written relief has been granted by the Commission pursuant to 10 CFR 50.55a(g)(6)(i).

Components of the Reactor Coolant System were designed to provide access to permit inservice inspections in~ accordance with Section XI of the ASME Boiler and Pressure Vessel Code, 1974 Edition.and Addenda through Winter 1975.

3/4.4.11 REACTOR VESSEL HEAD VENTS Reactor vessel head ventsare provided to exhaust noncondensible gases and/or steam from the Reactor Coolant System that could inhibit natural circulation core cooling. ' The'0PERABILITY of et leastJtwo reactor vessel head vent paths ensures'that thE capab'ility exists perform this function.

The valve redundancy of.the reactor vessel head vent paths serves to mini- ,

size the probability of inadvertent or, irreversible actuation while ensuring that a single fa.ilure of a vent' valve, power supply, or :fontrol system does not ,

prevent isolatibn of the vent path.

The function, capabilities, and testing requirements of the reactor vessel head vents are consistent with the requirements of Item II.B.1 of NUREG-0737,

" Clarification of TMI Action Plan Requirements," November 1980.

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ATTACHMENT 6

. ST.HL.AE. 22 R

...P. AGE 33 0F 9 I w

3/4.5 EMERGENCY CORE COOLING SYSTEMS

{ BASES ,

3/4.5.1 ACCUMULATORS The OPERABILITY of each Reactor Coolant System (RCS) accumulator ensures that a sufficient volume of borated water will be immediately forced into the ,

reactor core through three cold legs in the event the RCS pressure falls below the pressure of the accumulators. This initial surge of water into the core provides the initial cooling mechanism during large RCS pipe ruptures.

The limits on accumulator volume represent a spread about an average value used in the safety analys.is and have been demonstrated by sensitivity studies to vary the peak clad temperature by less than 20*F. The limit on ;

accumulator pressure ensures that the assumptions used for accumulator injec- ,

tion in the safety analysis are met. '

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l The accumulator power operated isolation valves are considered to be I

" operating bypasses" in the context of IEEE Std. 279-1971, which r'e quires that bypasses of a protective function be removed automatically whenever permissive conditions are not met. In addition, as these accumulator isolation valves fail to meet single failure criteria, removal of power to the valves is required.

The limits for operation with an accumulator inoperable for any reason except an isolation valve closed minirizes the time exposure of the plant to a'LOCA event occurring concurrent with failure of an additional accumulator which may result in unacceptable peak cladding temperatures. If a closed isolation valve cannot be opened within one hour, the full capability of one

.C.' accumulator is not avai1able and prompt action is required to place the reactor in a mode where this capability is not required.

3/4.5.2 and 3/4.5.3 ECCS SUBSYSTEMS The OPERABILITY of three independent ECCS subsystems ensures that sufficient emergency core cooling capability will be available in the event of a LOCA assuming the loss of one subsystem through any. single failure consideration.

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

assumed to discharge completely through the postulated break in the RCS loop.

Thus, three trains are required to satisfy the single failure criterion. Note that the centrifugal charging pumps the RHR pumps are not used in the in-jectionphaseof,theECCS. EachECCSsubsystemandtheRHRpumpsgrovide long-term core. cooling capability in the recirculation mode during the accident recovery period. pytyy of ECcS] And he.at excha.rvytrS When the RCS temperature is below 350*F, the ECCS requirements are balanced between the limitations imposed by the low temperature overpressure protection and the requirements necessary to mitigate the consequences of a LOCA below 350*F. At these temperatures, single failure considerations are not required because of the stable reactivity condition of the reactor and the limited core cooling requirements. Only a single Low Head Safety Injection pump is required to mitigate the effects of a large-break LOCA in this mode. However, two are provided to accommodate the possibility that the break occurs.in a loop con-

(. taining one of the Low' Head pumps. Low Head Safety Injection pumps are not required inoperable below 350 F because their shutoff head is too low to impact the low temperature overpressure protection limits.

~

SOUTH TEXAS - UNIT 1 B 3/4 5-1

ATTACHMENT G

, ST.HL.AE- 2232 PAGE 24 0F91 -

EMERGENCY CORE COOLING SYSTEMS p BASES _

1 l

ECCS SUBSYSTEMS (Continued)

Below 200 F (MODE 5) no ECCS pumps are required, so the High Head Safety Injection pumps are locked out to prevent cold overpressure.

The Surveillance Requirements provided to ensure OPERABILITY of each component ensure that, at a minimum, the assumptions used in the safety analyses are met and that subsystem OPERABILITY is maintained. Surveillance Requirements for flow testing provide assurance that proper ECCS flows will be gngagnedi,nfh,eeventofaLOCA.

3/4.5.5 REFUELING WATER STORAGE TANK The OPERABILITY of the refueling water storage tank (RWST) as part of the ECCS ensures that a sufficient supply of borated water is available for injection .

by the ECCS in the event of a LOCA or a steamline break. The limits on RWST minimum volume and boron concentration ensure that: (1) sufficien't water is available within containment to permit recirculation cooling flow to the core, (2) the reactor will remain subcritical in the cold condition (68*F to 212 F) following a small break LOCA assuming complete mixing of the RWST, RCS, Spray Additive Tank, Containment Spray System and ECCS water volumes with all con-trol rods inserted except the most reactive control rod assembly (ARI-1),

(3) the reactor will remain subcritical in cold condition following a large break LOCA (break flow area > 3.0 ft )2assuming complete mixing of the RWST, RCS, Spray Additive Tank, Containment Spray System and ECCS water volumes

{., and other sources of water that may eventually reside in the sump post-LOCA with all control rods assumed to be out (ARO), and (4) long term subcriticality -

following a steamline break assuming ARI-1 and preclude fuel failure.

The maximum allowable value for the RWST boron concentration forms the basis for determining the time (post-LOCA) at which operator action is required to switch over the ECCS to hot leg recirculation in order to avoid precipita-tion of the soluble boron.

The contained water volume limit includes an allowance for water not usable because of tank discharge line location or other physical characteristics.

The limits on contained water volume and boron concentration of the RWST also ensure a pH value of between 7.5 and 10.0 for the solution recirculated within containment after a LOCA. This pH band minimizes the evolution of iodine and minimizes the effect of chloride and caustic stress corrosion on mechanical syste,ms and components.

3/4.5.6 RESIDtfAL HEAT REMOVAL (RHR) SYSTEM ,

The OPERABILITY of the RHR system ensures adequate heat remov'al capabili-ties for Long-Term Core Cooling in the event of a small-break loss-of-coolant accident (LOCA), an isolatable LOCA, or a secondary break in MODES 1, 2, and 3.

The limits on the OPERABILITY of the RHR system ensure that at least one RHR loop is available for cooling including single active failure criteria.

The surveillances ensure that RHR system isolation valves close upon an overpressure protection system signal.

SOUTH TEXAS - UNIT 1 B 3/4 5-2'

CONTAINMENT SYSTEMS W

" ATTACHMENT G FIEL DRAR

( '

BASES

~

f I

3/4.6.1.5 AIR TEMPERATURE The limitations on containment average air temperature ensure that the over- ,

all containment average air temperature does not exceed the initial temperature condition assumed in the safety analysis for a LOCA or steam line break a,ccident. Measurements shall be made by. fixed instruments, prior to determin- !

ing the average air temperature. l 3/4.6.1.6 CONTAINMENT STRUCTUR'AL 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 with-stand the maximum pressure of 37.5 psig in the event of a LOCA or steam line break accident. The measurement of containment tendon lift-off force, the tensile tests of.the tendon wires er strandy, the visual examination of ten-dons, anchorages and exposed interior and) exterior surfaces of the containment, and the Type A, leakage test are sufficien to demonstrate this capability.

The Surveillance Requirements for demonstrating the containment's structural integrity are in compliance with the recommendations of proposed Regulatory

s. Guide 1.35, " Inservice Surveillance of Ungrouted Tendons in Prestressed Concrete

< Containment Structures," April 1979, and proposed Regulatory Guide 1.35.1, " Deter-mining Prestressing Forces for Inspection of Prestressed Concrete Containments,"

April 1979.

The required Special Reports from any engineering evaluation of containment abnormalities shall include a description of the tendon condition, the condition of the concrete (especially at tendon anchorages), the inspection procedures, the tolerances on cracking, the results of the engineering evaluation, and the correc-tive actions taken.

3/4.6.1.7 CONTAINMENT VENTILATION SYSTEM The 48-inch containment purge supply and exhaust isolation valves are required to be sealed closed during plant operations since these valves have not been demonstrated capable of closing during a LOCA or steam line break accident.

Maintaining.these valves sealed closed during plant operation ensures that exces-sive quantities of radioactive materials will not be released via the Containment Purge System. To provide assurance that these containment valves cannot be inad-vertently opened, the valves are sealed closed in accordance with Standard Review 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-k SOUTH TEXAS - ONIT 1 B 3/4 6-2

~

E ATTACHMENT 4 ES PLANT SYSTEMS-f BASES e'-

3/4.7.1.2 AUXILIARY FEEDWATER SYSTEM The OPERABILITY of the Auxiliary Feedwater System ensures that the Reactor ,

Coolant System can be cooled down to less than 350*F from normal operating conditions in the event of a total loss-of-offsite power.

Each auxiliary feedwater pump is capable of delivering a total feedwater flow of 540 gpm at a pressure of 1324 psig to the entrance of the steam generators. This capacity is sufficient to ensure that adequate feedwater flow is available to remove decay heat and reduce the Reactor Coolant System temperature to less than 350*F when the Residual Heat Removal System may be placed in.to operation. The AFW pumps are tested using the test line back to ,

the AFST and tk at:=ti: :=ta,h u.7the AFW isolation valves closed to prevent injection of cold water into the steam generators. The STPEGS isola-tion 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 pos on.

1:EEDWRTE.R 3/4.7.1.3 AUXILIARY, 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 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'imall fraction of 10 CFR Part 100 dose guideline values in the event of a steam line rupture.

This dose also includes the effects of a coincident 1 gpm primary-to-secondary tube leak in the steam generator of the affected steam line. These values are consistent with the assumptions used in the safety analyses.

3/4.7.1.5 MAIN STEAM LINE ISOLATION VALVES The OPERABILITY of the main steam line isolation valves ensures that no more than one steam generator will blow down in the event of a steam line ,

rupture. This restriction is required to: (1) minimize the positive reac-tivity effects of the Reactor Coolant System cooldown associated with the blowdown, and (2) limit the pressure rise within containment in the event the steam line rupture occurs within containment. The OPERABILITY of the main l steam isolation valves within the closure times of the Surveillance Require- l ments are consistent with the assumptions used in the safety analyses.

l l

SOUTH TEXAS - UNIT 1 B 3/4 7-2 i

ATTACHMENT G PLANT SYSTEMS GE 'l 0F p -

BASES -

SNUBBERS (Continued) associated installation and maintenance records (newly installed snubbers, seal replaced, spring replaced, in high radiation area, in high temperature area, etc.). The requirement to monitor the snubber service life is included to ensure that the snubbers periodically undergo a performance evaluation in view of their age and operating conditions. These records will provide statistical bases for future consideration of snubber service life.

3/4.7.10 SEALED SOURCE CONTAMINATION The limitations on removable contamination for sources requiring leak '

testing, including alpha emitters, is based on 10 CFR 70.39(a)(3) limits for plutonium. This limitation will ensure that leakage from Byproduct, Source, and Special Nuclear Material sources will not exceed allowable intake values.

Sealed sources are classified into three groups according to their use, with Surveillance Requirements commensurate with the probability of damage to a source in that group. Those sources which are frequently handled are required to be tested more often than those which are not. Sealed sources which are continuously enclosed within a shielded mechanism (i.e. , sealed sources within radiation monitoring or boron measuring devices) are considered to be stored and need not be tested unless they are removed from the shielded mechanism.

{

3/4.7.11 (Not Used) 3/4.7.12 (Not Used) 3/4.7.13 AREA TEMPERATURE MONITORING The area temperature limitations ensure that safety-related equipment will not be subjected to temperatures in excess of their environmental qualification temperatures. Exposure to excessive temperatures may degrade equipment and can cause a loss of its OPERABILITY. The temperature limits include an allowance for instrument error of i 3*F maximum.

a/4.7.I4 ESSEMTl AL chi u.1E.o W ATE R SYSTEr%

Tka. of'ER.A hl t T'( .y &_ E.ssen+iaJ CRilled Ma.ttr S3stam e.nsuye.s

+ hat sufficient cochn co.pa6+y is, ava.ilabic. -fw- conhnue.d ogen' of safe +y-YvJafed 'pnet cluvin$ norma.1 and ac.ci de.d cedi+ ions. Tha. undo.n+ cochn3 capacA/ of +fds syst e(#

assurn'ing a_ 'i s cow s is+an+ w'tM. n.e.

assampdens sugle, u se.d. in-filure.,,

W. sa feg a.najyse.s .

k SOUTH TEXAS - UNIT 1 B 3/4 7-6

ATTACHMENT 4

. ST.HL AE 223?-

REFUELING OPERATIONS c" -

BASES -

3/4.9.6 REFUELING MACHINE The OPERABILITY requirements for the refueling machine and auxiliary ,

hoist ensure that: (1) the refueling machine and auxiliary hoist will be used for movement of drive rods and fuel assemblies, (2) the refueling machine has sufficient load capacity to lift a drive rod or fuel assembly, and (3) the core internals and reactor vessel are protected from excessive lifting force in the event they are inadvertently engaged during lifting operations.

3/4.9.7 CRANE TRAVEL - FUEL HANDLING BUILDING The restriction on movement of loads in excess of the nominal weight of a ,

fuel.and control rod assembly and associated handling tool over other fuel assemblies in the storage pool, unless handled by the single-failure proof main hoist of the FHB 15-ton crane, ensures that in the event this load is dropped:

(1) the activity release will be limited to that contained in a single fuel assembly, and (2) any possible distortion of fuel in the storage racks will not result in a critical array. This. assumption is consistent with the activity release assumed in the safety analyses.

3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION The requirement that at least one residual heat removal (RHR) loop be in

{'- operation ensures that: (1) sufficient cooling capacity is available to remove decay heat and maintain the water in the reactor vessel below 140'F as required during the REFUELIhG MODE, and (2) sufficient coolant circulation is maintained through the core to minimize the effect of a boron dilution incident and prevent boron stratification.

Therequirementtoh$vetwoRHRloopsOPERABLEwhenthereislessthan 23 feet of water above the reactor vessel flange ensures that a single failure of the operating RHR loop.will not result in a complete loss of residual heat removal capability. With.the reactor vessel head removed and at least 23 feet cf water above the reactor pressure vessel flange, a large heat sink is avail-able for core cooling. Thus, in the event of a failure of the operating RHR loop, adequate time is provided to initiate emergency procedures to cool the core.

3/4.9.9 CONTAINMENT VENTILATION ISOLATION SYSTEM 2# The dPERABILITY of this system ensures that the containment ='. r.F pMgpenetrations will be automatically isolated upon detection of high radiation levels U + + "- ~ '"-- " " The OPERABILITY of this system is required to restrict the release of radioactive material from the containment atmosphere to the environment.

Q -Hu. pq e.xhaus+

b SOUTH TEXAS - UNIT 1 B 3/4 9-2

ATTACHMENT G ST.HL AE-32R i PAGE 29 0F 9I REFUELING OPERATIONS ee an

( gggg ,

l

( REFQEUNG cAv1TY AMD 5ToRAGE Poot.S 4

3/4.9.10 and 3/4.9.11 WATER LEVEL -4EAGTOR-VE33Ei .nd S?CNT T'J CL r^^L The restrictions on minimum water level ensure that sufficient water a depth is available to remove 99% of the assumed 10% iodine gap activity released from the rupture of an irradiated fuel assembly. The minimum water depth is consistent with the assumptions of the safety analysis.

3/4.9.12 FUEL HANDLING BUILDING EXHAUST AIR SYSTEM The listtations on'the Fuel Handling Building Exhaust Air System ensure that all radioactive material released from an irradiated fuel assembly will be filtered through the HEPA filters and charcoal adsorber prior to discharge -

to the atmosphere. Operation of the system with the heaters operating for at i

least 10 continuous hours in a 31-day period is sufficient to reduce the build-up of moisture on the adsorbers and HEPA filters. The OPERABILITY of this system and the resulting iodine removal capacity are consistent with the assump-tions of the safety analyses. ANSI N510-1980 will be used as a procedural guide for surveillance testing.

C" i

l L

SOUTH TEXAS - UNIT 1 B 3/4 9-3

ATTACHMENT 4 l

. ST.HL.AE 3Z32 PAGE 90 OF 9 l RADIOACTIVE EFFLUENTS C '

BASES This specification applies to the release of radioactive materials in gaseous effluents from each unit at the site.

rooWE- ,

3/4.11.2.3 DOSE - IODINE-131 M 133 TRITIUM, AND RADI0 ACTIVE MATERIAL IN PARTICULATE FORM This specification is provided to implement the requirosants of Sections II.C III.A and IV.A of Ap endik I, 10 CFR Part 50. The Limiting Conditions for Operation are the guides ' set forth in Section II.C of Appendix I. The ACTION statements provide the required operating. flexibility and at the same time implement the guides set forth in Section IV. A of AppcMix I to assure that the releases of radioactive materials in gaseous effluents to UNRESTRICTED AREAS will be kept "as low as is reasonably achievable." The OOCH calculational methods specified in the Surveillance Requirements implement the requirements in Section III. A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data, such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The 00CM calculational methodology and parameters for calculating the doses due to the actual release rates of the subject materials are consistent with the methodology provided in Regulatory Guide 1.109, " Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50,

.{ Appendix I," Revision 1, October 1977 and Regulatory Guide 1.111 " Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors," Revision 1, July 1977. These equa-tions also provide for determining the actual doses based upon the historical average atmospheric conditions. The release rate specifications for Iodine-131 Iodine-133, tritium, and radionuclides in particulate form with half-lives greater than 8 days are dependent upon the existing radionuclide pathways to man, in the areas at and beyond the SITE BOUNDARY. The pathways that were examined in the development of these calculations were: (1) indlyidual inhalation of air-borne radionuclides, (2) deposition of radionuclides onto green leafy vegetation with subsequent consumption by man, (3) deposition onto grassy areas where milk animals and meat producing animals graze with consumption of the milk and meat by man, and (4) deposition on the ground with subsequent exposure to man.

This specification applies to the release of radioactive materials in gaseous effluents from each unit at the site.

3/4.11.2.4 GASEOUS WASTE PROCESSING SYSTEM The OPERABILITY of the GASEOUS WASTE PROCESSING SYSTEM ensures that the systems will be available for use whenever gaseous effluents require treatment prior to release to the environment. The requirement that the appropriate por-tions of these systems be used df'when specified provides reasonable assurance N that the releases of radioactive materials in gaseous effluents will be kept "as low as is reasonably achievable". This specification implements the re-( quirements of 10 CFR 50.36a, General Design Criterion 60 of Appendix A to 10

% CFR Part 50 and the design objectives given in Section II.D of Appendix I to 10 CFR Part 50. The specified limits governing the use of appropriate portions i

SOUTH TEXAS - UNIT 1 B 3/4 11-4

TABLE 5.7-1 8 COMPONENT CYCLIC OR TRANSIENT LIMITS 5

1

-4 CYCLIC OR DESIGN CYCLE h

COMPONENT TRANSIENT LIMIT OR TRANSIENT

6 g Reactor Coolant System 200 heatup cycles at i 100 F/h Heatup cycle - T,y9 from $ 200 F q and 200 cooldown cycles at to > 550*F. ,

~< 100*F/h.

CooTdown cycle - T avg from

~

> 550 F to 1 200*F 200 pressurizer cooldown cycles Pressurizer cooldown cycle at 5 200*F/h. temperatures from > 650 F to g$g i 200*F. g 80 loss of load cycles, without immediate Turbine or Reactor trip.

> 15% of RATED THERMAL POWER to D% of RATED THERMAL POWER.

Nk Qgg

.o w -i 40 cycles of loss-of-offsite loss-of-offsite A.C. electrical 'Ne

[ A.C. electrical power. ESF Electrical System.

80 cycles of loss of flow in one Loss of only one reactor reactor coolant loop. coolant pump.

l 400 Reactor trip cycles. 100% to 0% of RATED THERMAL POWER. -

10 auxiliary spray Spray water temperature differential actuation cycles. > 621*f.

200 leak tests. Pressurized to > 2485 psig.

10 hydrostatic pressure tests. Pressurized to > 3 psig.

Secondary Coolant System 1 steam line break. 8reak in a > 6-inch steam line. E 10 hydrostatic pressure tests. Pressurized to > 1600 psig.

m

ST-HL-AE-2232, Forwards Final Draft Tech Specs & Technical Changes W/ Justifications.Tech Specs Consistent W/Fsar.Ser & as-built Facility W/Exceptions,Editorial Changes & Corrections Noted on Encl marked-up Pages

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