ST-HL-AE-2182, Responds to Questions Raised in 870505 Meeting W/Nrc Re Util Compliance W/Gdc 56 & 57 & SRP 6.2.4 Concerning Remote Manual Isolation of Primary Containment Isolation Valves Following Dba.Revised Response to Generic Ltr 85-12 Encl
| ML20214S795 | |
| Person / Time | |
|---|---|
| Site: | South Texas  |
| Issue date: | 05/29/1987 |
| From: | Wisenburg M HOUSTON LIGHTING & POWER CO. |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-2.E.4.2, TASK-TM GL-85-12, ST-HL-AE-2182, NUDOCS 8706090424 | |
| Download: ML20214S795 (50) | |
Text
.
The Light hE f Ilouston Lighting & Power P.O. Box 1700 llouston,'lixas 77001 (713) 228-9211 May 29, 1987 ST-HL-AE-2182 File No.: G9.17 J22.3, G9.6 10CFR50 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555 South Texas Project Units 1 and 2 Docket Nos. STN 50-498, STN 50-499 Primary Containment Isolation-Component Cooling Water Supply / Return to Reactor Coolant Pumps, Reactor Coolant Drain Tank Heat Exchanger and Excess Letdown Heat Exchanger On May 5, 1987, at a meeting held in Bethesda, Md between Houston Lighting and Power Company (HL&P) and members of the NRC staff, questions were raised by the staff concerning STP's compliance with General Design Criteria 56 and 57 and Standard Review Plan 6.2.4 pursuant to the remote manual isolation of the primary containment isolation valves for the subject equipment following a design basis accident.
The present component cooling water (CCW) subsystem design contains one inlet and one outlet containment penetration of Class 2 design. The CCW inlet penetration contains two parallel Class 2 motor operated valves located outboard and powered from diverse sources, and one inboard Class 2 check valve. The outlet penetration contains one motor operated and one pneumatic operated Class 2 valve located in parallel on the outboard side, and two motor operated Class 2 valves located in parallel on the inboard side. Each valve is powered from diverse sources. At present, operator action from the control room is required to close each of the penetration isolation valves.
The piping from the inlet penetration isolation valve to the reactor coolant pumps (RCP) supplying CCW for notor/ pump cooling is Class 3 as is the RCP return lines to the outlet penetration isolation valve. The header and branch piping supplying component cooling water to the reactor coolant drain tank heat exchanger and to the excess letdown heat exchanger is Class 3 up to and including the Class 3 isolation valves. Downstream of the Class 3 isolation valves to the heat exchangers, the piping is seismically designed, non-safety. The common header and supply lines to these non-essential heat exchangers is automatically isolated on an SI signal via the Class 3 isolation valves. One isolation valve is located in the common header and one valve is located in each of the branch lines. The CCW return lines for these non-essential heat exchangers are automatically isolated by two Class 3 check valves, in each line.
L1/NRC/pc B706090424 870529 PDR ADDCK 05000498 A PDR yl ht m
s
. ST-HL-AE-2182
. Houston Lighting Ac Power Company File No.: G9.17, J22.3,'G9.6 Page 2 Pursuant to the requirements of the Standard Review Plan and General Design Criteria, and in accordance with NUREG-0737 Item II.E.4.2 - TMI action plan requirements, HL&P will now close the aforementioned containment isolation valves automatically on a Phase B isolation signal. The setpoint for the automatic closure will be approximately 10 psig-containment pressure.
This setpoint ensures that containment isolation is maintained for non-essential systems following a design basis event and mitigates the possibility of containment atmosphere leakage to the environment via the CCW system. This setpoint also allows for RCP operation following a-Phase A isolation signal in accordance with Westinghouse recommendations and TMI operating experience. Annotated FSAR and Technical Specification revisions-regarding the incorporation of the Phase B isolation signal are provided as attachments.
Incorporation.of the Phase B isolation signal also affects our previous response to Generic Letter 85-12. Our revised response to Item B1 of Generic
. Letter 85-12 is provided in Attachment 2.
These modifications will be installed prior to Unit 1. fuel load.
Accordingly, with the installation of this change HL&P believes-that compliance with NRC requirements for containment isolation is complete.
If you should have any questions on this matter,s please contact.Mr.
M. E. Powell at (713) 993-1328.
pL& w M. R. senburg Manager, Engin ring a Licensing LRS/yd
Attachment:
- 1. Annotated FSAR Revisions Concerning the Incorporation of the Phase B Isolation Signal
- 2. Revised Response to Generic Letter 85-12
- 3. Annotated Technical Specification Revisions concerning the Incorporation of the Phase B Isolation Signal L1/NRC/pc
ST-HL-AE-2182 Film No.: C9.17, J22.3, C9.6 Ifouston Lighting & Power Company p g. 3 cc:
Regional Administrator, Region IV M.B. Lee /J.E. Malaski
~ Nuclear Regulatory Commission. City of Austin 611 Ryan Plaza Drive, Suite 1000 P.O. Box 1088 Arlington, TX 76011 Austin, TX 78767-8814 N. Prasad Kadombi, 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 Senior Resident Inspector / Operations e/o U.S. Nuclear Regulatory Commission P.O. Box 910 Bay City, TX 77414 Claude E. Johnson Senior Resident Inspector / Construction c/o U.S. Nuclear Regulatory Commission P.O. Box 910 Bay City, TX 77414 M.D. Schwarz, Jr., Esquire Baker & Botts One Shell Plaza Houston, TX 77002 J.R. Newman,-Esquire Newman & Holtzinger, P.C.
1615 L Street, N.W.
Washington, DC 20036 T.V. Shockley/R.L. Range Central Power & Light company P. O. Box 2121 Corpus C'<.risti, TX 78403 Ll/NRC/pc Revised 2 '3/87
s 4
t 4
f I
t
! ATTACHMENT 1 f
Annotated FSAR Revisions Concerning the Incorporation of.the Phase B Isolation Signal I
i 4
D i
i i
I L1/NRC/pc 1
m w-, 2 <>ks's e-m=+-+--,-wew- +--,--w.--er,~:,-~w-mmw--mvmwnmv~,,aw-wn--w, we e -+
,s J W v ,
TABLE 1.3-2 (Continued) -
SICMIFICANT DESIGN CNANGES References Item FSAR Description of Change Various coincidence logic changes Section 7.2, 7.3 1) SG tow-tow water tevel reactor trip changed from 2/3 to 2/4.
- 2) Turbine stop velves closed signet for reactor trip changed from 3/4 to 2/4.
- 3) So high-high water tevel signet (P 14) for turbine trip and FW isotation changed from 2/3 to 2/4.
1 l 4) Turbine trip siysts modified to meet plant needs.
l _
s --
l 7' Phase B containment isolation Section 7.3 Signal is no longer used to activate any opipment; continued supply g i y clonel of CCW to the RCPs is desirable and isolation is remote manuel for m g these isolation volves. y nenuet override of containment Section 7.3 Manuet override of containment isolation signets is no longer permitted; the isolation signets operator must reset the containment isolation signet to open the valve.
Main steam Line isolation signets Section 7.3 Addition of high ageam pressure rate for MSIV closure below the P-11 57 setpoint.
Feedwater isolation / turbine trip Section 7.3 Addition of SI signal to signets causing FW isolation and turbine trip.
Fuel handling building exhaust Section 7.3 System is actuated fottowing an SI signet to provide filtration of any $$a NVAC system ESFAS teakage in the ESF puup cubicles. mr -$
->I
, 45 mg y Auxillery feedwater initiation Section 7.3 Puups are now also started on loss of-offsite power; however, AFW m
$ signets regulating valves do not open s til SI signet or tow-tow SG water %
S, tevet signet is received. Capability to block the low-low water g.
@ tevet signet added, a 5
-7. The Containment isolation valves are designed to meet leaktightness stan-
). dards consistent with the overall leaktightness of the RCB.
I
- 8. The system is designed with redundancy and physical separation so that no single active failure can result in loss of Containment integrity.
39
, 9. The system is designed to withstand the environmental conditions which accompany an accident without loss of function.
10'. Containment isolation valve closure speeds and leaktightness limit radio-logical effects from exceeding guideline values as established in l 10CFR100.
l^
l 11. Provisions are made for periodically testing the operability and l 1eaktightness of the isolation valves to the extent necessary to ensure that the system will meet its performance requirements in the event of an j accident requiring RCB isolation. Leak rate testing is in accordance t
with 10CFR50, Appendix J.
- 12. The system is designed in accordance with the quality group classifica-tions in RC 1.26, seismic categories in RG 1.29, and the power supply 39 requirement in RG 1.32. I i
l 6.2.4.2 System Design. The signals utilized to actuate the CIS are as fol-lows:
,1% 6 Mah ed holekw
, 1. Phase A Containment isolation j and containment ventilation isolation sig-nals isolate all nonessential lines &d & cCW leas k W.RCPs
- 2. The steam line isolation signal automatically closes all main steam iso- 39 lation valves (MSIVs) to prevent the continuous, uncontrolled blowdown of more than one steam generator (SG) and thereby uncontrolled RCS cooldown. 53
- 3. The main W line isolation signal automatically closes all W isolation valves. The W isolation prevents or mitigates the effects of excessive RCS cooldown.
These isolation signals and details of their derivation are described in Sec-tion 7.3. 39 Figure 6.2.4-1 identifies the types of isolation valves and isolation schemes ,
provided for lines which penetrate the containment Building. This Figure ;
includes: (1) open or closed status under normal operating conditions, as
- well as shutdown or accident conditions, (2) modes of actuation, (3) the sig- l nals to initiate isolation valve closure, and (4) closure time for the isola-tion valves.
The three types of fluid lines penetrating the Containment Building which l39 require Containment isolation valves are:
i Type A - Lines which form part of the reactor coolant pressure boundary.
l Type B - Lines which connect directly with the containment atmosphere. - l41
) Type C - Lines which are part of a closed system: those which are neither part of the RCPB nor connected to the Containment atmosphere. l41 6.2-45 Amendment 56 i
~
a--mk-e- w-w-p-w-*'-=*a.--*M
- ATTACHMENT,1 STP FSAR ST HL AE M l a PAGE % QFA9 Isolation valves are also provided in other fluid lines penetrating the Con-tainment (i.e., SC isolation) shown in Figure 6.2.4-1.
}
6.2.4.2.1 Special Containment Isolation Provisions:
- 1. Essential Systems Certain systems are required to perform a safety function following an accident and the isolation valves for these systems must be opened auto .
natically or remote-manually or remain open for the system to operate.
,Clositig of these valves would defeat their intended purpose. Special provisions for Containment isolation for each of these systems is des-cribed below. h3
- a. _ Low Head Safety Injection (LHSI), High Head Safety Injection (HHSI) and Containment Spray These valves are designed to be operated remote-manually from the control room. The operator is made aware of leakage in these sys-tems by the following provisions after which the valves can to closed. ,
- 1) Sump-le.*e1 alarms are provided for each sump in the pump cut,i-cles. The detection and alarm capability of the sump level instrumentation and design criteria are discussed in Section 9.3.3.
- 2) Increased radiation levels in the pump cubicles as a result of -
leaks will be indicated and alarmed by the Area Radiation Mon-itoring System described in Section 12.3.4. The monitor loca- '~
tions are given in Table 12.3.4-1 (Monitors N1RA-RE-8084, N1RA-RE-8085 and N1RA-RE-8086).
)
In the unlikely event of a tube rupture in one of these HXs, radio-active materials would be entrained in the CCWS. The operator in the control room would be made aware of this condition by the CCWS radiation monitor discussed in Section 11.5. Additional indications of this condition would be given by the CCWS flow indications, the high and low flow alarms, surge tank level indications and tank high level alarms. The isolation valves in this system are designed to be operated remote-manually from the control room. 53 imakage (normal or abnormal) from the CCWS will not result in re-leases of radioactive material. Laakage is collected in building sumps and conveyed to the Liquid Waste Processing System (LWPS) as described in Section 9.3.3.
- c. Auxiliary Feedwater System (AFW)
Laakage from the AFW System will not result in release of radioac-s tive material. The water being conveyed to the SGs is free from radioactive contamination. The isolation valves used for AFW to 6.2-46 Amendment 53
____m__ _ _ _ _ - _ _ _ - _ _ _ _ _ _ _ - _ _ _ _ _
ATTACHMENT '
STP FSAR ST HL AE &l9 PAGE 4 OF J at
} TABLE 6.2.4-2 (Continued)
CONTAINMENT ISOLATION VALVES MAXIMUM ISOLATION TIME PENETRATION VALVE NO. FUNCTION (SECONDS)
CONTAINMENT VENTILATION ISOLATION (Continued)
M-41 HC0010 Containment Normal 60 Purge Exhaust M-80 RA0001 Containment Atmosphere 10 Radiation Monitor M-80 RA0003 Containment Atmosphere 10
- Radiation Monitor
~
M-80 RA0004 Containment Atmosphere 10 Radiation Monitor M-80 RA0006 Containment Atmosphere 10 58 Radiation Monitor
@ThecaT* Ison-4nM PMsc' 6
}[.z37f REMOTE MANUAL M-83 AF0019 A W Supply N.A.
M-94 AF0048 AW Supply N.A.
M-95 AF0065 AW Supply N.A.
M-84 AF0085 A W Supply N.A.
M-33 CC0012 CCW to RHR HX and RHR Pump N.A.
M-34 CC0049 CCW From RHR HX and RHR Pump N.A.
M-34 CC0050 CCW From RHR HX and RHR Pump N.A.
M-26 CC0068 Cooling Water From RCFC N.A.
M-26 CC0069 Cooling From RCFC N.A.
M-35 CC0122 CCW to RHR HX and RHR Pump N.A.
M-36 CC0129 CCW From RER HX and RHR Pump- N.A.
6.2-237e Amendment 58
-iim- ____m______m.__. _ _ _ _ _ _ _ _ - - - - - _ - - _ _ _ _ - -
ATTACHMENT l STP FSAR ST HL AE.pirra.
PAGEe; 0Fe23
) TABLE 6.2.4-2 (Continued)
CONTAINMENT ISOIATION VALVES MAXIMUM IS01ATION TIME PENETRATION VALVE NO. FUNCTION (SECON6S)
~
REMOTE MANUAL (Continued)
M-36 CC0130 CCW to RHR HX and RHR Pump N.A.
M-28 CC0147 Cooling Water From RCFC N.A.
M-28 CC0148 CCW From RCFC N.A.
M-37 CC0182 CCW to RHR HX and RHR Pump N.A.
M-38 CC0189 CCW From RHR HX and RHR Pump N.A.
M-38 CC0190 CCW From RHR HX and RHR Pump N.A. 58 M-24 CC0197 CCW to RCFC N.A.
M-23 CCO208 Cooling Water From RCFC N.A.
M-23 CCO210 CCW From RCFC N.A.
M-39 CC0291 CCW to RCPs t o -h I M-39 CCO318 CCW to RCPs to -N-* .
M-40 4 94---- j CC0403 CCW Frca RCPs to Nrk:
c.2- Z37e M-40 CC0404 CCW From RCPs so h M-40 r00542 CCW From RCPs 10 WA.
M-40 .~/-4493 CCW From RCPs i o WA .
M-1 MS0143 Main Steam Supply N.A.
to AW Turbine M-1 FV-0143 Bypass for MS Supply N.A.
to A W Turbine M-18 XSIOOO4A High Head SI Discharge N.A.
M-14 X.cIOOO4B High Head SI Discharge N.A.
1 6.2-237f Amendment 58 9
ATTACHMENT l
. ST.HL AE 2/ f &
STP FSAR PAGE /o OF A9 TABLE 6.2.6-1 SYSTEMS NOT VENTED DURING TYPE A TESTING Penetration SYSTEM No.* Justification H'i gh Head Safety Injection 10, 14, 18 The system is normally filled with water and oper-ating under post-accident conditions.
Chemical and Volume Control 46, 48, 53 The system maintains the plant in a safe condition during testing.
Low Head Safety Injection 11, 15, 19 The system is normally filled with water and oper-ating under post-accident conditions.
Refueling cavity 55, 76 The piping inside the k to RWST Containment is connected to RHR system by two locked i da;na closed manual valves and is 9 60) / used only during refueling.
The system is filled with water during normal and post-accident conditions.
- Component Cooling Water 23, 24, 25, This is a closed system to Reactor containment 26, 27, 28 inside the Containment Fan Coolers which is normally filled with water and operating under post-accident condi-tions.
I 39 Component Cooling Water 33, 34, 35, This is a closed system to RHR Heat Exchanger 36, 37, 38 inside the Containment which is normally filled with water and operating under l
post-accident conditions.
39, 40 This is a closed system CCWtoRCP/ inside the Containment which is normally filled with water and operating under normal and post-accident l53 conditions.
m pw manyt% Ak- 444r lam"~a hi. J*
p m wz-3 .Jr *-&c.u.n ru
- 1. 3).
- See Figure 6.2.4-1 for further description.
6.2-245 Amendment 53
EXPLANATIONS AND ABBREVIATIONS sum
- 1. ISOLATION VALVE NO: ONLY CONTAINMENT ISOLATION VALVE NUMBERS ARE LISTED THOUGH OTHER VALVES NECESSARY FOR
- PERFORMING LEAK RATE TESTING ARE SHOWN.
} 2. CHK: CHECK VALVE.
- 4. ELEC/HYD: ELECTRO-HYDRAULIC 0PERATOR.
- 5. SELF ACT: SELF ACTUATING.
- 5. RIA: REMOTE MANUAL
- 7. L.0/L.C.: LOCKED OPEN/ LOCKED CLOSED.
- 8. TRAIN: THE POWER TRAIN (A.B.C.OR D)THAT ENABLES THE CLOSURE OF THE VALVE FOR CONTAINMENT ISOLATION.
AN ENTRY O F D INDICATES DC POWER SUPPLIED FROM THE CHANNEL 11 BATTERY SYSTEM.
dh "U
--4 S 9. ISOLATION SIGNAL: THE SIGNAL THAT CAUSES CLOSURE OF THE VALVE TO FACILITATE ISOLATION OF THE CONTAINMENT. 4
- 10. FWl: FEEDWATER ISOLATION SIGNAL kN
%%-4
- 11. PHASE A: PHASE AISOLATION SIGNAL JN sm.PMsr6: fust 6 unne r own
- 12. CVI: CONTAINMENT VENTILATION ISOLATION SIGNAL
- 13. SI: SAFETYINJECTIONSIGNAL
- 14. AFWl: AUXILIARY FEEDWATER INITIATION SIGNAL
- 15. ARSLI: MAIN STEAM LINE ISOLATION SIGNAL SOUTH TEXAS PROJECT
- 15. LENGTH 0F PIPING OUTSIDE CONT: THE LENGTH OF PING OUTSIDE THE CONTAINMENT TO THE OUTSIDE CONTAINMENT UNITS 1 & 2 ISOLATION VALVE.
- 17. GOC: 10 CFR 50. APPENDIX A. GENERAL DESIGN CRITERIA. containeetwT rentTRATious
=
Shast 1 of 100 g 18. IRC/ ORC: INSIDE REACTOR CONTAINMENT /0UTSIDE REACTOR CONTAINMENT.
- 19. N/A: NOT APPLICABLE. % ,u , wn
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,- STP FSAR ATTACHMENT l ST HL AE-als9-
' M :$ PAGE /0 0F M
- 6. Phase B Containment Isolation - Closure of rocess lines,' initiated by l43 Containment HI-3 pressure signal _(process lines do not include Engineered
)
Safety Feature [ESF] lines). ((This signal is available but not used. ) l43
- 7. . Protection System Response Times -
l43
- a. Reactor Trip System (RTS) Response Time - The time delays are defined as the time required for-the reactor trip to be initiated (i.e., the time the rods are free and begin to fall) following a-step change in the variable being monitored from 5 percent below (or
. above) to 5 percent above (or below) the ' rip setpoint.
- b. ESF Actuation System (ESFAS) Response Time - The interval required for the ESF sequence to be initiated subsequent to the point in time that the appropriate variable (s) exceed setpoints. The response time includes sensor / process (analog) and logic (digital) delay.
Times required for standby diesel generator startup and loading (as applicable) and equipment response times are not included in the protection system response times. These times are considered 43 however in the accident analyses discussed in Chapters 6 and 15.
8.- Reproducibility - Scientific Apparatus Manufacturers Association (SAMA)
Standard PMC-20.1-1973, " Process Measurement and Control Terminology,"
defines reproducibility as "the closeness of agreement among repeated measurements of the output for the same value of input, under normal operating conditions over a period of time, approaching from both direc-tions." It includes drift due to environmental effects, hysteresis, -~
long-term drift, and repeatability. Long-term drift (aging of compo-
' nents, etc.) is not an important factor in accuracy requirements since, in general, the drift is not sig..ificant with respect to the time elapsed l between testing. Therefore, long-term drift may be eliminated from this l_ definition. Reproducibility, in most cases, is a part of the definition of accuracy (see below).
i 9. Accuracy - An accuracy statement for a device falls under Note 2 of the SAMA Standard PMC-20.1-1973 definition of accuracy, which means reference accuracy or the accuracy of that device at reference operating condi-tions: " Reference accuracy includes conformity, hysteresis and repeatability." To adequately define the accuracy of a system, the term "repoducibility" is useful as it covers normal operating conditions.
The terms " trip accuracy," " indicated accuracy," etc., then include con- l 43 formity and reproducibility under normal operating conditions. Where the final result does not have to conform to an actual process variable but is related to another value established by testing, conformity may be eliminated, and the term " reproducibility" may be substituted for
" accuracy."
- 10. Normal Operating Conditions - For the STP FSAR, these conditions cover j all normal process temperature and pressure changes. Also included are ambient temperature changes around the transmitter and racks. This -
document does not include any accuracies under post-accident. conditions.
7.1-2 Amendment 43 l
' STP FSAR ATTACFndENT l ST HL AE4/fd PAGE ll EOF;t9 7.1.1.2 Engineered Safety Feature Actuation Systems. The ESF Actuation Systems are those instrumentation and control systems which are needed to j actuate the equipment and systems required to mitigate the consequences of ANS /
Condition II, III, and IV faults (see Chapter 15). The ESF and ESF support systems requiring ESFAS actuation are:
- 1. Standby diesel generators and ESF load sequencers
- 2. Emergency Core Cooling System (Safety Injection System) ,
3.- Main steam line and feedwater (FW) isolation
- 1
- 4. Containment isolation (Phase d Containment Ventilation A fn% e. 6 Isolation) 43
- 5. Containment heat removal (Reactor Containment Fan Coolers and Containment Spray System)
- 6. Electrical Auxiliary Building Main Area HVAC System
- 7. Electrical Penetration Space HVAC System
- 8. Fuel Handling Building HVAC Exhaust Subsystem
- 9. Control Room Envelope HVAC System
- 10. Auxiliary Feedwater System
- 11. Component Cooling Water System f
- 12. Essential Cooling Water System
- 13. Essential Chilled Water System
- 14. Various HVAC equipment as required to support these ESF components and systems The ESF Actuation Systems are discussed in Section 7.3. Design bases for the 43 ESFAS are given in Section 7.1.2.1. A single-line diagram of the Westinghouse ESFAS is shown on Figure 7.1-2. The SSPS cabinet layout is shown on Figure 7.1-3.
Systems supporting the ESFAS are the Class 1E AC Power System and the Class 1E 125 vde Power System. Both power nystems are discussed in Chapter 8.
7.1.1.3 Systems Required for Safe Shutdown. Systems required for safe-shutdown nre defined as those essential for pressure and reactivity control, coolant inventory makeup, and removal of residual hest once the reactor has been brought to a suberitical condition.
Identification of the equipment and systems required for safe shutdown is provided in Section 7.4. .
s t
7.1-4 Amendment 43
.m. . . . . . . . . .
, ATT/IPMENT i ST.HL AE- AIR &
- , STP FSAR PAGEffAOFAS g procedures require immediate operator action based solely on bypass indications.
- 3. Appropriate separation criteria are applied to the design and installa-tion of the system in order to avoid degrocation of the safety-related 43 systems.
- 4. The capability for assuring the operable status of the system is -
provided.
- 5. Bypass indication on a system basis is provided for the following systems:
- a. Safecy Injection Syster, 43
'b. Containment Spray System
- c. Containment Isolation Phase A
- d. Containment Ventilation Isolation 2
- e. Class lE 125 vde and 120 V Vital ac systems f32. l)
- f. Combustible cas Control System (hydrogen recombiners - see Section 6.2.5, and the ' Containment Hydrogen Monitoring System - see Section 7.6.5).
) g. Containment Heat Removal System
- h. Fuel Handling Building HVAC Exhaust subsystem V
- i. Solid State Protection System
- j. Feedwater Isolation
- k. Steamline Isolation
- 1. Auxiliary Feedwater System
- m. Electrical Penetration Space HVAC System
- n. Control Room Envelope and Electrical Auxiliary Building Main Area HVAC Systems L4a, 4 1s.Iss a Pk*ss S '
o.
- 6. The following support systems, when bypassed or rendered inoperable, activate bypass indication for that system plus all supported systems identified in item 5 above: 43
- a. Component Cooling Water System
- b. Essential Cooling Water System *
)~
7.1 17 Ameadment 57 a is lamu E
ATTACHMENT / ;
< ST HL AE G/8&
PAGE 130FA9
, STP FSAR TABLE 7.1-2 (Continued)
)
PLANT COMPARISON
- ENCINEERED SAFETY FEATURES DIFFERENCES FROM ACTUATION SYSTEMS (Continued) COMANCHE PEAK NUCLEAR STATION actuates one control room .
HVAC train. STP has a separate
. control room for each unit.
Each control room has redundant air inlet radiation monitors,~each actuating all three trains of control room HVAC.
- 8. Fuel Handling Building 8. STP uses SI signal or high Exhaust HVAC ESF Actuation radiation signal (from either Signals (Figure 7.3-27) of two redundant Class lE spent fuel pool exhaust monitors) to initiate FHB exhaust filtration.
On Comanche Peak, fuel building exhaust is always filtered; no actuation is required.
DeleItd pe,leled
- 9. P'- r : C;;;;i..s...; I;;1etisa 9. C: ;nJ.. F..i u... thi; 21... 1 Sig;;l (rig;;; '.2 *) 0; isel-i. CC'J t; the rerrter
- lent yo-y.. air J.;; net
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- .;.;.d t;; ;; 23.i y.. ..; i. 44
- tunt.d.) ,
- 10. Excessive Cooldown Feedwater 10. Addition for STP of excessive Isolation Signals (Figures cooldown signals for feedwater 7.2-5, 7.2-7, 7.2-9 and isolation on:
7.2 14) a. [ low primary loop tiow or low T in 2/4 loops +
high IU8]f low + P-15 interlock signal, or
- b. low compensated T P-15interlocksikEkI.+
Comanche Peak does not have these signals,
- 11. Turbine Trip signal From 11. Addition on STP of manual Feedwater Isolation Signals reset capability for the turbine (Figure 7.2-14) trip signal from the combined signal of P-16 or any of the following signals: safety injection or P-14 signal or excessive cooldown feedwater isolation. Comanche Peak does not provide this capability.
T 7.1-34 Amendment 57
ATTACHMENT /
ST HL AE. A fr/-
) -3. Reactor Containment fan coolers, which serve to cool the Containment and limit the potential for release of fission products from the Containment by reducing the pressure following an accident.
4 Component cooling water and essential cooling water pumps and associated valves, which serve as auxiliary heat removal systems. l ,'3 i
- 5. ' Motor-driven auxiliary feedwater pumps and turbine-driven auxiliary l3 feedwater pump and associated valves, which serve to cool the SG's on 143 loss of main feedwater. I
- 6. Containment isolation phase A, whose function is to prevent fission product release, i.e., isolation of lines not essential to reactor protection.
- 7. Steam line isolation to prevent the continuous, uncontrolled blowdown of more than one steam generator and thereby uncontrolled RCS cooldown.
- 8. Main feedwater line is61stion as required to prevent or mitigate the effect of excessive cooldown.
- 9. Standby diesel generators to assure backup supply of power to emergency and supporting systems components.
.g 10. Operation of the Control Room Envelope HVAC System to meet control room
") occupancy requirements following a 14CA. Operation of the EAR Main Area HVAC System to meet equipment environment requirements.
- 11. Containment spray pumps and associated valves, which serve to reduce Containment pressure and temperature (and to remove iodine) following a 14CA or steam line break accident inside the Containment.
- 12. Containment isolation phase B, ::i;;i. fly designed to isolate the Containment following a 1DCA or a steam or feedwater line break withift M the Containment to limit radioactive releases. (5: N -* 8 - -_t i::1; "
M i- -h;:: 3 -Y ' i: n:: r f. (The Containment isolation phase A bp#w. si and the Containment ventilation signal clos y ll lines pene-
, hP* # - trating Containment which are not considered *!!IE::ry for reactor 43 i$5 protection and accident mitigation).
- 13. Containment ventilation isolation, to ensure that all Containment purge lines have been isolated, thus preventing fission product release.
- 14. Operation of the FHB HVAC Exhaust System, to ensure filtration of air exhausted from the cubicles containing the safety injection and containment spray pumps, thus minimizing offsite releases of postulated leakage from these pumps.
- 15. Turbine trip, to prevent excessive cooldown of the RCS. t
- 16. Essential Chilled Water System, to provide chilled water for necessary
_ HVAC systems.
7.3-3 Amendment $1
ATTACHMENT j STP FSAR ST HL-AE Al00-
. PAGE 16 0F ael steam pressure rate), 7.2-8 (ESF actuation), 7.2-9 (excessive cooldown
) protection), 7.2-14 and 7.2-15 (feedwater control and isolation), and 7.2-16 (auxiliary feedwater).
To facilitate ESF actuation testing, six cabinets (two per train) are provided which enable operation, to the maximum practical extent, of safety features loads on a group-by-group basis until actuation of all devices has been checked. Final actuation testing is discussed in detail in Section 7.3.1.2. -
7.3.1.1.4 Final Actuation Circuitry: The Solid-State Protection System supplies the following signals: 43
- 1. Safety injection signal (Table 7.3-5 lists actuated equipment. Typical control logics for actuated equipment are shown on Figures 7.3-2 through 7.3-8.)
- 2. Containment spray signal (Table 7.3-6 lists actuated equipment. %3 Typical control logics for actuated equipment are shown on Figures 7.3-9 and 7.6-14.) ,
- 3. Containment isolation Phase A signal (Table 7.3-7 lists actuated equip- 43 ment. Typical control logics are shown on Figures 7.3-11 through 7.3-13.)
(-131<. 15 -8 la h sb.ek d sydr <d.
- 4. Containment isolation Phase B si ign:1 i. uvi. . n, tr e -9"=*= - ~ qM; mt .y -r ftcJ y c.gnal
,, % -(O.1:
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- s. 5. Containment ventilation isolation signal (Table 7.3-9 lists actuated equipment. Typical control logics are shown on Figures 7.3-16 and 7.3-17.)
- 6. Steam line isolation signal (Table 7.3-10 lists actuated equipment.
Typical control logics are shown on Figures 7.3-18 and 7.3-18A.) 43
- 7. Feedwater isolation signal (Table 7.3-11 lists actuated equipment.
Typical control logics are shown on Figures 7.3-19 and 7.3-20.)
- 8. Auxiliary Feedwater (AW) initiation signal (Table 7.3-15 lists a032.16l45 actuated equipment. Typical control logics are shown on Figures 7.3-21, 7.3 21A and 7.3-218.)
Loads are sequenced onto the three Class 1E ESF buses by the ESF load sequencers, as described in Chapter 8. The design meets the requirements of CDC 35. L3 7.3.1.1.5 Design Bases Information. The functional diagrams presented on Figures 7.2 5 through 7.2-9 and 7.2-14 through 7.2-16 provide a graphic outline of the functional logic associated with requirements for the ESFAS.
Requirements for the ESFAS are given in Chapter 15. Given below is the design bases information required in Institute of Electrical and Electronic Engineers (IEEE) 279-1971.
-)
7.3-5 Amendment 53
STF FSAR ATTACHMENT l ST HL-AE BIG
. PAGE 160FJi TABLE 7.3-2A
')
FUNCTIONS / SYSTEMS ACTUATED BY VESTINCHOUSE ESFAS SIGNALS SAFETY INJECTION SIGNAL CONTAINMENT SPRAY SIGNAL Reactor Trip System Containment Spray System ,
43 Turbine Trip Containment Isolation Phase 6 5(= ::; '
- .J.3 4 e..t)"'
Feedwater Isolation Auxiliary Feedwater System 57 Standby Diesel Generators AUXILIARY FEEDWATER INITIATION SIGNAL Component Cooling Water System Auxiliary Feedwater System 43 Q32.16 Safety Injection System ,.
Steam Generator Blowdown Isolation Essential Cooling Water System Steam Generator Sample Isolation Reactor Containment Fan Coolers Containment Isolation Phase A 43 Containment Ventilation Isolation MAIN STEAMLINE IS01ATION Control Room Envelope HVAC System Steamline Bypass Valve Closure 57 EAS Main Area HVAC System Main Steam Isolation Valve Closure FHB HVAC Exhaust Subsystem ESF Load Sequencers Essential Chilled Water System Electrical Penetration Space HVAC System 43
& s 7.3-32 Amendment 17
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PAGE 19 OF F1 Table 7.3-8 has been deleted.
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- ATTACHMENT l STP FSAR ST4iL-AE Jir PAGEJJ OF 9 The capability for initiating a manual bypass indication and alarm is provided via a system level manual bypass switch to indicate the bypass / inoperable condition to the operator for those components or conditions which are not automatically monitored. 41 Manual bypass / inoperable indication may be set up or removed under administra-tive control. The automatic indication feature of the ESF Status Monitoring System can not be removed by operator action.
Bypass and/or status indication on a system level is provided for'the fol- 2
, lowing safety-related systems: Q32.7
- 1. Solid-State Protection System (bypass / inoperable only)
- 2. Safety Injection System (SIS) (including RHR' system components required 41 for accident mitigation or safe shutdown)
- 3. Containment Spray System (CSS)
'4. Containment Isolation Phase A
- 5. Containment Ventilation Isolation 41
- 6. Class 1E 125 vde and 120 v Vital AC Systems
- 7. Combustible Gas Control System (bypass / inoperable only) 8 Containment Heat Removal System
- 9. Fuel Handling Building (FHB) Heating, Ventilating, and Air Conditioning (HVAC) Exhaust Subsystem
- 10. Electrical Penetration Space HVAC System
- 11. Control Room Envelope and Electrical Auxiliary Building (EAB) Main Area HVAC System 41
- 12. Feedwater Isolation
- 13. Steam Line Isolation
- 14. Auxiliary Feedwater System (AFWS) it. G.,h a L W A 8 The following support systems activate bypass indication of all supported safety systems listed above when they are bypassed or rendered inoperable:
- 1. Component Cooling Water System (CCWS)
- 2. Essential Cooling Water System (ECWS)
~
- 4. Essential Chilled Water System 7.5-4a Amendment 53
4 - -
. AppIndix 7A [DfSI AfoY'M#D (bg. pVFo 0"'1 II.E.4.2 CONTAINMENT ISOLATION DEPENDABILITY ATTACHMENT 1
-} ST HL AE- Gita-Position (NUREG 0737) PAGE JULOF AQ (1) Containment isolation system designs shall comply with the recommenda-tions of Standard Review Plan (SRP) Section 6.2.4 (i.e., that there be diversity in the parameters sensed for the initiation of containment isolation). -
(2) All plant personnel shall give careful consideration to the definition of essential and nonessential systems, identify each system determined to be essential, identify each system determined to be nonessential, describe the basis for selection of each essential system, modify their contain-ment isolation designs accordingly, and report the results of the reeval-untion to the NRC.
(3) All nonessential systems shall be automatically isolated by the contain-ment isolation signal.
(4) The design of control systems for automatic containment isolation valves shall be such that resetting the isolation signal will not result in the automatic reopening of containment isolation valves. Reopening of containment isolation valves shall require deliberate operator action.
36 (5) The containment setpoint pressure that initiates containment isolation for nonessential penetrations must be reduced to the minimum compatible with normal operating conditions.
(6) Containment purge valves that do not satisfy the operability criteria set forth in Branch Technical Position CSB 6-4 or the Staff Interim Position of October 23, 1979, must be sealed closed as defined in SRP 6.2.4, item II.3.f during operational conditions 1, 2, 3, and 4. Furthermore, these valves must be verified to be closed at least every 31 days. (A copy of the Staff Interim Position is enclosed as Attachment 1.)
(7) Containment purge and vent isolation valves must close on a high radia-tion signal.
Clarification (1) The reference to SRP 6.2.4 in position 1 is only to the diversity requirements set forth in that document.
(2) For post-accident situations, each nonessential penetration (except instrument lines) is required to have two isolation barriers in series that meet the requirements of General Design Criteria 54, 55, 56, and 57, as clarified by SRP, Section 6.2.4. Isolation must be performed automat-i ically (i.e., no credit can be given for operator action). Manual valves must be sealed closed, as defined by SRP, Section 6.2.4, to qualify as an isolation barrier. Each automatic isolation valve in a nonessential pen-etration must receive the diverse isolation signals.
~.
l l
l l
7A.II.E.4.2-1 Amendment 40 l '
STP FSAR ATTACHMENT J Appindix 7A ST HL AE Al&
PAGED 30FoM (3) Revision 2 to Regulatory Guide (RG) 1.141 will contain guidance on the classification of essential versus nonessential systems and is due to be )
issued by June 1981. Requirements for operating plants to review their list of essential and nonessential systems will be issued in conjunction with this guide including an appropriate time schedule for completion.
(4) Administrative provisions to close all isolation valves manually before resetting the isolation signals is not an acceptable method of meeting ,
position 4.
(5) Canged reopening of containment isolation valves is not acceptable.
Reopening of isolation valves must be performed on a valve-by-valve basis, or on a line-by-line basis, provided that electrical independence and other single-failure criteria continue to be satisfied.
(6) The containment pressure history during normal operation should be used as a basis for arriving at an appropriate minimum pressure setpoint for initiating containment isolation. The pressure setpoint selected should be far enough above the maximum observed (or expected) pressure inside the containment during normal. operation so that inadvertent containment isolation does not occur during normal operation from instrument drif t or fluctuations due to the accuracy of the pressure sensor. A margin of 1 psi above the maximum expected containment pressure should be adequate t 36 account for instrument error. Any proposed values greater than 1 psi will require detailed justification. Applicants for an operating license and operating plant licensees that have operated less than one year should use pressure history data from similar plants that have operated g more than one year, if possible, to arrive at a minimum containment setpoint pressure.
')
(7) Sealed-closed purge isolation valves shall be under administrative control to_ assure that they cannot be inadvertent 1v opened. Adminis-trative control includes mechanical devices to seal or lock the valve closed, or to prevent power from being supplied to the valve operator.
Checking the valve position light in the control room is an adequate method for verifying every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> that the purge valves are closed.
STP Response (1) The STP Containment isolation signal is generated by diverse parameters.
Containment phase A isolation, steam line isolation, feedwater line ise-
, . lation, and Containment ventilation isolation are all initiated by the i~satetyinpectionsignal. The safety injection signal is initiated by the following parameters:
I WZ
- Containment pressure 1
- pressurizer low pressure l low compensated steamline pressure low-low compensated T-cold concurrent with power below 10% (P-15) manual actuation Containment phase A isolation and steam line isolation may also be initi-ated manually. ,
g g g g g g;uf.KA crpal.aidNekdw A s4A7 h UN=Dw gr.3 J g b e r ,f ~7A.II.Eg4.2-2w u y M en,
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- ATTACHMENT ST-HL AE- alta-l
. STP FSAR P_ AGE Jti 0FM The CCWS is designed to:
1.
}
Provide cooling water to various nuclear plant components during all modes of plant operation. This includes plant equipment required for safe shut-down and ESF equipment required after a postulated DBA.
- 2. Provide an intermediate fluid barrier between potentially radioactive systems and the ECWS to reduce the possibility of leakage of radioactive .
contamination to the outside environment.
- 3. Perform its cooling function following a DBA with offsite or standby power 39 sources, automatically and without operator action, assuming a single active or passive failure.
- 4. Provide cooling water at 60*F to 105'F temperature during normal opera-tion. The maximum temperature during DBA is 120.9'F (refer to Table 39 9.2.5-5 for temperature for the individual scenarios).
- 5. Conform to seismic Category I requirements and safety classifications, as indicated on Figures 9.2.2-1 through 9.2.2-5 and in Table 3.2.A-1,
- 6. Permit periodic inspection of important components and periodic and func-tional testing to assure the integrity and operability of the system. See 39 Sections 3.9.6 and 6.6.
In addition, the CCWS is protected from the effects of tornado loadings, mis-siles, flooding, pipe whip, and jet forces from pipe breaks. See Sections 3.3.2, 3.4.1, 3.5, and 3.6.
9.2.2.2 System Description.
9.2.2.2.1
Description:
The CCWS consists of three separate redundant !
trains, each with a pump, HX, associated piping, and valves, that service two Reactor Containment Fan Coolers (RCFC), Residual Heat Removal (RHR) Heat 55 Exchanger and RHR pump, as shown on Figures 9.2.2-1 through 9.2.2-3. The three trains are connected to a common header which services other equipment as shown in Figures 9.2.2 4 and 9.2.2-5. In addition, a compartmentalized surge tank is used to accommodate the water thermal expansion and contraction, and a chemical addition tank is used to balance the water chemistry (Table 9.2.2-2) of the system.
A CCW HX bypass line is provided to maintain 60*F minimum CCUS temperature.
This line is only used when the ECW temperature is very low.
For heat removal following a DBA, all three CCUS trains will operate if avail-able, but two trains are capable of performing the heat removal function.
Except for the seal water HX, reactor coolant pump (RCP) lube oil coolers and thermal barrier, RCP motor air coolers, RHR pump seal coolers, centrifugal charging pump (CCP) supplementary coolers, CCP lube oil coolers, and positive displacement pump supplementary cooler, the remaining equipment is isolated by A es.which close on an_SI signal.J An SI signal opens the pneumatic valve r (closed during normal operation) to provide cooling water to each RHR HX.
Also an SI signal shifts the cooling water supply to the RCFCs from the -
chilled water system to the CCUS by closing the chilled water and opening the CCWS motor operated supply and return valves.
)
L.- ptw 4 +k. RCf L.k- Al wolus H Gw ~ d *A A "ol'n 4 4AJL '
Amendment 5
/uGAJ v ~@ 14 Gh *4d nr-3 QM.
W STP FSAR q#f h n.x p% s w,yd s J n.
The CCVS design incorporates manual isola ion of cooling water to the RCPs j
') Loss of cooling water to the RCPs due to the operator closing a single 39
- g. W = tar-c; m tad valve is avoided by parallel Containment isolation valves, as shown on Figure 9.2.2-5. @c % blIgic k N
- y M H 8. M w 5 H .-co m A g% 9 E h sig' 9 D. 3 '2' 35aYa/e W .s.W h~a6 p # .'s./*h, /4.i;N "and I$81ation:4 ' To minimize
. nim D tYo E betI t the possibility of leakage, welded construction is used throughout the CCWS where practical. ,
Makeup to the system is automatic. The surge tank level is monitored by the main plant computer and the Qualifed Display Processing System. Opening of the valve providing tank makeup is alarmed by the computer to give an 55 indication of system leakage. If the normal source of makeup (the DWS) fails or is inadequate, the surge tank level instrumentation provides alarms and actuations as discussed in Section 9.2.2.2.1.
Should a large inleakage into the CCWS develop, the level in the surge tank l53 will rise, and the high-high-level condition will be annunciated. If a leaking fluid is radioactive, then the high radiation level will be alarmed.
39 Refer to Section 11.5 for further information on the process radiation monitor. If the level in the surge tank continues to increase, the CCW will 5(
discharge to the CCW sump through the open vent. The vent is designed to 53 accommodate the maximum inleakage due to the rupture of one RCP thermal barrier, which is estimated at 275 gal / min. The increased pressure due to a rupture of the RCP thermal barrier will close two active self-actuated pressure regulated valves located at the CCW outlet from the RCP thermal barrier and isolate the RCS inleakage. In addition, a high flow or high 55 g temperature in the return line from the thermal barrier, indicating failure of
') the RCP thermal barrier pressure-retaining boundary, will close the CCW motor-operated valve downstream of the two pressure regulated valves; an alarm in the control room alerts the operator as well. The portion isolated is.
designed for RCS pressure and temperature.
l 55 The operator, by checking flow and temperature readings against normal values, can locate the affected portion of the system and isolate this portion by closing the appropriate remotely operated or manual valves. Very small leaks will be detected by periodic inspection of the system piping and valves.
The relief valves on the CCW lines to the various HXs are sized to relieve the volumetric expansion occurring if the shell side of the HX is isolated and high-temperature process fluid flows through the HX tubes. 55 ATTACHMENT I ST Ht. AE. e2 i f Q.
PAGEg, OF 49
)
9.2-15 Amendment 56
- 4 %
Table 9.2.2-3 (Continued)
Co e onent Coolins Water System Falture Modes and Effects Analysis PLANT METHOD FAILURE EFFECT DESCRIPTION SAFETY OPERATING FAILURE OF FAILimE ON SYSTEM SAFETY GENERAL REMARKS OF COMPONENT FUNCTION MODE
- MODE (S) DETECTION FUNCTION CAPA5ILITY RCS isolation RCP Close to isolate 16 One valve falls e Position Indic- None: Two valves in CCW Sg ply (Mov 0318, Containment open sting Lights series for isolation.
0291) nonmatty open e ESF Status Non-l 55 itoring Open to provide 1-6 One vetve falls a Position indic- None: Two valves em cooling flow closed stins lights parettet will ensure e ESF Status Mon- water is provided. 53 ,
7 n
itoring y m
RCs isolation close on reverse 16 valve faits open None None: Two valves in 55 l checkvalve(CC0319) flow series (one outside b the RCs).
RCDT and excess Close-isolate non- 16 One valve falls a ESF Status Non- None: Two valves in tetdown heat ESF Load open itoring tights series exchanger isolation a Position indic-valve (Mov 0297, 0392, atleY 0393) normally open
[J log~08 RCDT and excess close on reverse flow 16 One valve faits None None: Two check valves T rn g tetdown heat ex-changer check open in series.
% oo-4 y valves (CC0540, 4-
@ CC0541, CC0763,
- 5. CCO'02) ,
N
- Plant Modes
, 1. Power Operation 3. Not Staney 5. Cold Shutdown
" 2. Start-i , 4. Not Shutdown 6. Refueling
v b .
Table 9.2.2-3 (Continued)
Component Cootina Water System Felture Isodes and Effects Anotysis I
PLAllT IEETN00 FAILURE EFFECT DESCRIPTICII SAFETY CPERATING FAILURE OF FAILURE 011 SYSTEst SAFETY GENERAL REMARES OF COMPOWENT FUIICT1011 MODE
- IquDE(S) DETECT 1011 FUNCT1011 CAPASILITY RCS footetton vetwee Close (to footste RCS) 1-6 One vetve faits e Position Indic- IIone: Three vetves out (nov 0403, 0542, open eting Lights of the four are edespote ]
to footste the RCS.
@40 and y tic e ESF Status ston-ItorIng veIwe FV449Tnonest--
ty open Open to ettow flow 1-6 One vetve faits to e Positten Indic- none: vetves are in per-open sting lights ettet to ensure flow.
? e ESF Status plon-7 Itorins 53
- CCW supply check 1) Close (on 1-6 1) One vetve felts 1) None 1) none - two check
- vetves (CC0327, reverse flow) open vetves in series.
0759, 0363, 0757, 0321, 0756, 0346, 2) Remain open 2) One vetve felts 2) mene 2) mone - seat water Pressure Intes-0758) to RCF close to RCP is provided Ity of the seet thernet barrier by the charging po p for RCP is moln-from the CVCS. telned either by seetweter from the CVCS or m thernet borrier M cooling of the Ot g
en ii ,b5 S.
2 E
p*
ss
- Plant Modes l
- 1. Power Operation 3. Not Stenttry 5. Cold Shutdown
- 2. Start-@ 4. Not Shutdown 6. Refueling
STP FSAR gg gov pcD ATTACHMENT l y M # g uty ST.HL AE 2J f&
Question 010.9 _PA,GE R OF 4t$ \ _.
i The design of your component cooling water system provides a single supply and return line, each supplying cooling water to all four reactor coolant pumps.
These lines contain motor-operated valves for containment isolation. The seals and bearings of the reactor coolant pumps require continuous cooling by the component cooling water system during all modes of operation. Inadvertent closure of any one of the above motor-operated valves would terminate the coolant flow to all of the pumps which potentially may lead to fuel damage, due to a locked rotor. Therefore, it is our position that you design this portion of the component cooling water system so that the following criteria
- are met:
- 1. A single failure in the component cooling water system shall not result in fuel damage or damage to the reactor coolant system pressure boundary caused by an extended loss of cooling to the reactor coolant pumps.
Single failure includes operator error, spurious actuation of motor-oper-ated valves, and loss of component cooling water pumps.
- 2. A moderate energy leakage crack or an accident that is initiated from a failure in the component cooling water system shall not result in exces-sive fuel damage or a breach of the reactor coolant system pressure boundary when an extended loss of cooling to the reactor coolant pumps occurs. A single active failure shall be considered when evaluating the i consequences of the accident. Moderate leakage cracks shculd be deter-mined in accordance with the guidelines of Branch Technical Positionc APCSB 3-1, " Protection Against Postulated Failures in Fluid Systems Outside Containment."
)
To meet the two criteria above, that portion of the component cooling water system which supplies cooling water to the reactor coolant pumps can be designed to non-seismic Category I requirements and Quality D if you demon-strate that the reactor coolant pumps are capable of operating with loss of cooling for longer than 30 minutes without loss of function and the need for operator protective action. And, safety grade instrumentation to detect the loss of component cooling water to the reactor coolant pumps and to alarm the operator in the control room is provided. The entire instrumentation system,
, including audible and visual status indicators for loss of component cooling l water should meet the requirements of IEEE Std 279-1971. Alternately, if it cannot be demonstrated that the reactor coolant pumps will, operate with loss of cooling water for longer that 30 minutes without loss of function or oper-4 ator corrective action, then your entire component cooling water system design l aust meet the following requirements:
i'
- 1. Safety grade instrumentation consistant with the criteria for the protec-tion system shall be provided to initiate automatic prctection of the plant. For this case, the component cooling water supply to the seal and bearing of the pumps may be designed to non-seismic Category I require-ments and Quality Group D; or
- 2. The component coolir.g water supply to the pumps shall be capable of withstanding a single active failure or a moderate energy line crack as
)
[
Q&R 9.2-1 Amendment 39 9
_ _ . - < ,_.__ _,_, . . . _ , , _ _ . _ _ . . . - _ _ _ _ . _ . _ _ _ _ , . _ , _ . . _ . . _ ~ ~ _ _ - . _ , - -
I ATTACHME T )
+ STP FSAR ST.HL AE I f d-PAGEdQ_ cR9 Question (Continued) defined in our Branch Technical Position APCSB 3-1 and be designed to seismic Category I, Quality Group C and ASME Section III, Class 3 requirements.
Response
The Component Cooling Water System (CCWS) is described in Section 9.2.2 of the South Texas Project FSAR. is.t. 6 .;p .4 ,s i. a 4
- 1. As presented in Section 9.2.2.3.1, the containment isolation schemes for the reactor coolant pumps (RCPs) CCW supply and re rn lines assure continuance of CCW flow in the event of a single failure. Single failures considered in system design include spuriousivalve closure and 46 loss of any one CCW pump. This redundancy in CCW containment isolation valving, as shown in Figures 9.2.2-5 and 6.2.4-1 (sheets 41 and 42), also asssures CCW flow isolation when neeerccry ( cr.u;i :: tier, requir:d) again l55 assuming the most limiting single active lure. A failure modes and effects analysis (FMEA) of the CCWS ie resented in Table 9.2.2-3 _
.m , h ,y n./.h Aa* 6 .N is W )
- 2. The RCP CCW supply and return piping and valves are designed to Safety 4 Class 3, seismic Category I requirements outside of containment penetration areas. The piping and valves in the penetration area are designed to Safety Class 2 Seismic Category I requirements. Safety class interface points are shown on Figure 9.2.2-5. A moderate energy line 13 9 crack (postulated in accordance withthe guidelines of BTP APCSB 3-1) in a CCWS supply line to the RCPs would cause a reduction in CCW flow to RCP bearing oil coolers. A rise in bearing oil temperature would be experienced. As described in Section 5.4.1.3.3, each RCP motor bearing 4 contains embedded temperature detectors, which are indicated and alarmed in the control room. Additionally, as discussed in revised Section 9.2.2.5, low CCW flow from the RCP bearing oil coolers is alarmed on a p6 per pump basis to provide additional information to the control room operator. These alarms will alert the operator with information of a b7 degraded CCW flow condition with the RCPs.
Testing has been performed by Westinghouse, which has shown the manufac-turer's recommended maximum bearing operating temperature will be reached 4 in approximately ten (10) minutes upon total loss of CCW flow. Since Westinghouse has demonstrated that the RCPs are capable of operation for ten (10) minutes upon total loss of CCW flow without incurring any damage to the motor bearing and since the operator has redundant control room alarms to warn of degraded CCW flow conditions, ten (10) minutes is a l55 conservative operator response time for this event during normal opera- 46 tion. RCP seal cooling is maintained by seal injection from the charginS 55 system.
i Q&R 9.2 2 Amendment 55
ATTACllMENT 2 Revised Response to Generic Letter 85-12 L1/NRC/pc
Attcchment 2 ST-HL- AE- 2182 File No.: C9.17, J22.3, C9.6 Page 1 of 4 Revised Response to Generic Letter 85-12 South Texas Project Electric Generating Station B. Potential Reactor Coolant Pump Problems
- 1. Requirement Assure that Containment Isolation, including inadvertent isolation, will not cause problems if it occurs for non-LOCA transients and accidents.
- a. Demonstrate that, if water services needed for RCP operations are terminated, they can be restored fast enough once a non-LDCA situation is confirmed to prevent seal damage or failure.
- b. Confirm that containment isolation with continued pump operation will not lead to seal or pump damage or failure.
Response
containment Isolation (including inadvertent isolation) will not cause reactor coolant pump problems during non LOCA transients and accidents because the Containment Isolation Phase A (CIA) signal by itself does not terminate either of the two water supplies to the O.CP seals. Note that a safety injection (SI) signal generates a CIA signal.
During normal operation, seal injection flow to the RCP seals is provided from the Chemical and Volume Control System (CVCS). Seal injection flow is not isolated except when a CIA signal is present concurrent with low charging header pressure. Cooling water to the RCP thermal barrier heat exchanger is provided by the component cooling water (CCWS) system to limit heat transfer from reactor coolant to the RCP internals and to the RCP seals in the event that seal injection flow is lost. Automatic isolation of CCW flow occurs on a Containment Isolation Phase B (CIB) signal or on a low CCW surge tank level signal. The presence of a CIB signal is indicative of a LOCA or steamline break type of event. CIB and Containment Spray are actuated simultaneously either upon reaching the containment pressure HI-3 setpoint, or by manual initiation of the containment spray actuation switches. The CCVS and CVCS are described in FSAR Sections 9.2.2 and 9.3.4 respectively.
- a. The most credible condition which would interrupt both of these sources of cooling water is a loss of offsite power (LOOP), as described in STP FSAR Appendix 7A.II.K.3.25. In this case, the RCP motors are, of course, de energized. The diesel L1/NRC/pc
Attcchment 2 ST HL AE-2182 File No.: C9.17, J22.3, C9.6 Page 2 of 4 generators are automatically started, and seal injection flow and component cooling water to the thermal barrier heat exchanger is automatically restored within seconds (see FSAR fable 8.3 3). Either of these water supplies is adequate to provide seal cooling and prevent seal failure during a loss of offsite power.
Should an accident occur and generate an SI signal (which initiates Containment Isolation Phase A) without a coincident LOOP, seal injection would not be terminated. The design is such that Class 1E equipment is not stripped following an SI; seal injection is thus continued using the Class 1E centrifugal .
charging pump (s). If the containment pressure did not rise '
above the HI 3 setpoint, the CCW flow to the RCPs would also continue without interruption.
Should an accident occur and generate an SI (and thus CIA).with a coincident LOOP, seal injection would be terminated. Since the charging pumps are not resequenced following an SI coincident with a LOOP, the low charging header pressure coincident with a CIA signal would close the seal injection
- containment isolation valves as discussed above. If the containment pressure did not rise above the HI-3 setpoint, the CCW flow to the RCPs would be reinstated within seconds (see FSAR Table 8.3 3).
If a main steam line break inside containment should occur coincident with a LOOP, both sources of cooling water would be isolated. The RCPs would also be tripped automatically on the LOOP. Manual action would then be required to restore Reactor Coolant Pump (RCP) seal cooling. Seal injection flow would be isolated based on CIA coincident with low charging header pressure, as described above. CCW flow would be isolated by a CIB signal when containment pressure reached the HI 3 setpoint.
Westinghouse has confirmed that the RCP seals will remain intact for a period of at least 10 minutes without cooling water while the pumps are not operating. During this period, a charging pump would be manually started. Removal of the low char 81 ng header pressure signal would then allow the seal injection containment isolacion valves to be reopened, thus restoring seal injection flow. Emergency operating procedures will be revised to reflect these steps. Diesel loading is such that char 61 ng pumps may be started on the Train A and C diesel generators at any time. Some load shedding may be required after approximately 2000 hours0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> of operation on the A train only.
L1/NRC/pc
j Attachment 2 ST HL AE 2182 File No.: C9.17, J22.3, C9.6 Page 3 of 4
- b. The essential service systems required for continued RCP operation are CCW and CVCS seal injection. Both of these systems remain open to the RCPs following a CIA signal. The RCP seal injection is isolated only upon a concurrent CIA signal and low charging header pressure. Seal injection flow would be maintained through a CIA without a concurrent LOOP, since charging header pressure would not decrease. The CCW supply to the RCPs is automatically isolated only upon low CCW surge tank level or a CIB signal.
Westinghouse has confirmed that seal integrity will be maintained for a period of at least 10 minutes while the RCP operates with either cooling water source isolated. In case of a loss of seal injection, operating procedures direct that the affected RCP remain in operation while efforts are made to restore seal injection flow in a controlled manner. If the RCP bearing or #1 seal inlet temperature exceeds the established setpoint, then the affected RCP is tripped. Therefore, loss of seal injection flow would not be a reason for RCP trip as long as temperatures are maintained within limits by the coeponent
- cooling water flow to the thermal barrier heat exchangers.
Response to a loss of component cooling water to the oil coolers is to monitor the RCP bearing temperatures while continuing to operate the RCP. RCP trip would be initiated if r operational temperature limits were exceeded. Loss of cooling '
water to the thermal barrier heat exchangers would not be a reason for RCP trip as long as a sufficient flow rate existed from the seal injection region into the RCS, as determined from the injection rate and leakoff rate flow instrunnentation.
Should both sources of cooling water be isolated, the RCP must be tripped within 1 minute.
Procedures for RCP restart following trip provide full consideration of the effects of initiating cooling water to a hot RCP component and the potential thermal stresses which could be induced.
L1/NRC/pc
Attachment 2 ST.HL-AE.2182 File No.: C9.17, J22.3, C9.6 Page 4 of 4 References For Previous STP Correspondence concerning Generic Letter 85 12 A. Thompson,ilush L. Jr. ; "Implemen.stion of TMI Action Item II.K.3.5,
' Automatic Tr19 cf Reactor Coolar.t Pumps' (Generic Letter 85 12)"; NRC letter addressed to All Applicants and Licensees with Westinghouse (W)
Designed Nuclear Steam Supply Systems (NSSS); June 28, 1985; ST IIL AE 90652.
B. Wisenburg, M. R.; " South Texas Project, Units 1 & 2 Docket Nos. STN 50 498, STN 50 499, Partial Response to NRC Conoric Letter 85 12,
' Implementation of TMI Action Item II.K.3.5, Automatic Trip of Reactor Coolant Pumps'"; Letter to Hugh L. Thompson, NRC, from Houston Lighting & Power; Nov. 6, 1985; ST llL AE 1433. (This letter provided responses to sections B and C of CL 85 12.)
C. Wisenburg, M. R. ; " South Texas Project, Units 1 & 2 Docket Nos. STN 50 498, STN 50 499, THI Item II.K.3.5 and Generic Letter 8512, Automatic Trip of Reactor Coolant Pumps"; Letter to Vincent S. Noonan, NRC, from llouston Lighting & Power; Jan. 28, 1986; ST ilL.AE.1598. (This letter provided an interim response to section A of CL 8512.)
D. Kadsabi, N. P. ; " Status of Review of South Texas Project (STP)
Information on Generic Letter 8512"; NRC Letter to J.11. Goldberg, itL&P, l
< April 17, 1986; ST AE IIL.90865. (This letter summarized information I
discussed during a conference call on March 19, 1986.)
E. Wisenburg, M. R.; " South Texas Project, Units 1 & 2. Docket Nos. STN
< 50 498, STN 50 499, Response to ' Status of Review of South Texas Project (STP) Information on Generic Letter 8512'"; Letter to llugh L. Thompson, NRC, from llouston Lighting & Power; May 30, 1986; ST llL AE.1673. (This letter provided clarification to the Staff's understanding, as documented in Reference D, of IIL&P's response to Item B1 of Generic Ittter 8512.)
F. Wisenburg M. R. ; " South Texas Project Units 1 & 2 Docket Nos. STN
$0 498, STN 50 499, Final Response to Section A of NRC Ceneric Letter 8512, ' Implementation of TMI Action Item II.K.3.5 Automatic Trip of Reactor Coolant Pumps'"; Letter to llugh L. Thompson, NRC, from llouston Lighting & Power; September 30, 1986; ST llL.AE 1753. (This letter provided a response to Section A of the Generic letter and addressed the NRC comment relative to overall guidance portfnent to RCP trip.)
L1/NRC/pc
i l
ATTACitMENT 3 Annotated Technical Specification Revisions concerning the Incorporation of the Phase B Isolation Signal e
L1/NRC/pc
TABLE 3.3-3 (Continued)
ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION h TOTAL NO.
MINIMJM CHANNELS APPLICABLE 3 -
0F CHANNELS CHANNELS TO TRIP OPERABLE MODES ACTION
, FUNCTIONAL UNIT -
c * \*
5 3. Containment Isolatjon (Continued) '
--e w 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 R 3) Safety Injection *** See Item 1. above for all Safety Injection initiating functions ani
- requirements.
w
$ 4) RC8 Purge Radioactivity-High 2 1 2 1,2,3,4,5 y y
,6 18
- 5) Containment Spray- See Ites 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 93 Manual Isolation function,and requirements.
Aa 5
- 4. Steam Line Isolation l M
o u)>
- a. Manual Initiation p t"PX
- 1) Individual 2/ steam line 1/ steam line 2/ operating 1, 2, 3 24 -i I steam line o'" g
- 2) System 2 1 2 1,2,3 23
% Y Lt
- b. Automatic Actuation 2 ~
1 2 1,2,3 22 Logic and Actuation Relays h
ATTACHMENT 1
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y ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUNENTATION TRIP SETPOINTS TOTAL SENSOR ERROR N
3 FUNCTIONAL UNIT ALLOWANCE (TA) Z (5) TRIP SETPOINT ALLOWA8LE VAlg Safety Injection (Reac}cr Trip, L-E 1.
5 Feedwater Isolation Control '..
- Room Emergency Ventilation, Start '
" Stan&y Diesel Generators, Reactor Containment Cooling Fans, and Essential Cooling Water)
- a. Manual Initiation M.A. N.A. N.A. N.A. N.A.
- b. Automatic Actuation Logic N.A. H.A. N.A. N.A. N.A.
Actuation Relays M.A. N.A. M.A.
g c. N.A. M.A.
Y d. Containment Pressure--High 1 3.6 0.71 2.0 $ 3.0 psig i 4.0 psig 3 13.1 10.71 2.0 1 1850 psigN 1 1842 psigN
- e. Pressurizer Pressure--Low
- f. Compensated Steam Line 13.6 10.71 2.0 1 735 psig 1 714.7 psig*
Pressure-Low 4.5 0.5 1.0
- g. Compensated T COW-Low-tow 1 532*F 1 528'F***
(interlocked with P-15)
- 2. Containment Spray y< -
- a. Manual Initiation M.A. M.A. N.A. M.A.
- N.A. aIs m-r 1
- b. Automatic Actuation Logic M.A. N.A. N.A. N.A. M.A. ([y>M
?z
- c. Actuation Relays N.A. M.A. M.A. N.A. N.A. A g.s so.s
- d. Containment Pressure--High-3 3.6 0.71 2.0 $ its psig $ 20;S psig FINAL MAFI
TABLE 3.3-4 (Continued) 3 ENGIIEERED SAFETY FEATURES ACTUATION SYSTEM INSTRIMNTATION TRIP SETP0lNTS N TOTAL SENSOR ERROR
~
R FUNCTIDW L LAIT ALLOWANCE (TA) Z (5) TRIP SETPOINT ALLOWABLE VALUE e
=
- 3. Canta* w t Isolation. .
5-U a. Phase "A" Iso'lation w
- 1) Manual Initiation N.A. N.A. N.A. N.A. N.A.
- 2) Automatic Actuation Logic M.A. N.A. N.A. N.A. N.A.
- 3) Actuation Relays N.A. M.A. N.A. M.A. M.A.
- 4) SafetyInjection See Ites 1. above for all Safety Injection Trip Setpoints and Allowable Values.
~
R.
Y b. Containment Ventilation Isolation 8
- 1) Automatic Actuation M. A. N.A. M.A. N.A. N.A.
Logic
- 2) Actuation Relays M.A. M.A. M.A. N.A. N.A.
- 3) SafetyInjection See Item 1. above for all Safety Injection Trip Setpoints and Allowable Values.
- 4) RCS Purge 3.1x10 1.8x10 1.3x10" <5x10 * * <6.4x10 *- T1 Radioactivity-High pCi/cc pCi/cc pCi/cc pCi/cc iici /cc
- 5) Containment Spray - See Item 2. above for Containment Spray manual initiation Trip w>
' Z Manual Initiation Setpoints and Allowable Values.
$iy<-
o
- 6) Phase "A" Isolation - See Item 3.a. above for Phase "A" Isolation manual initiation y g
-resa2r ,
Manual Initiation Trip Setpoints and Allowable Values. . gg
- 4. Steam Line Isolation N
- a. Manual Initiation N.A. N.A. N.A. N.A.
N.A.
i
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TABLE 3.3-5 FINAL NAFT ATTACHMENT g
, ENGINEERED SAFETY FEATURES RESPONSE O TIMES/0 hAC7E INITIATIONSlGNAi.ANDFUNCTION RESPONSE TIME IN SECONDS
- 1. Manual Initiation
- a. Safety Injection (ECCS) N.A.
- b. Containment Spray N.A.
. c. Phase "A" ! solation N.A.
- d. F% u. ' 6" Lola hea A/.4. I e d. Containment Ventilation Isolation N.A. 1
( s. Steam Line Isolation N.A.
- f. Feedwater Isolation N.A.
A g. Auxiliary Feedwater N.A.
t H. Essential Cooling Water N.A.
j t. Reactor Containment Fan Coolers N.A.
A J. Control Room Ventilation N.A.
/ X. Reactor Trip N.A.
a l'. Start Diesel Generator N.A.
- 2. Containment Pressure- High 1
- a. Safety Injection (ECCS) s 27(I)/12(5)
- 1) Reactor Trip i 2(3)
- 2) Feedwater Isolation < 7(3)
- 3) Phase "A" Isolation 33(II/23(2)
- 4) Containment Ventilation Isolation [23(I)/13(2)
L) Auxiliary Feedwater 1 60
- 6) Essential Cooling Water < 62(I)/52(2)
- 7) Reactor Containment Fan Coolers 38(I)/28(2)
- 8) Control Room Ventilation h72(I)/62(2)
- 9) Start Standby Diesel Generators i 12 4
i SOUTH TEXAS - UNIT 1 3/4 3-37 e
~
_ TABLE 3.3-5 (Continued)
FINAL DRAFT
-Ali ACHMENT1 J/8b
, ENGINEERED SAFETY FEATURES RESPONSE PAGE 7 TIMEL ST.HL Al 0 INITIATING $1GNAL AND FUNCTION RESPONSE TIMt IN SECON05
- 5. Compensated Steam Line Pressure--Low (Continued)
- 8) Control Room Ventilation < 72I1)/62(2)
- 9) Start Diesel Generators ~ 17
- b. Steam Line Isolation ' i8 W
- 6. Containment Pressure--High-3 i
- d. Containment Spray i 30(I)/20(2)
- b. Phuc'6' hkhm g pg(')/ f e,N
- 7. Containment Pressure- High-2 Steam Line Isolation I3) 17
- 8. Steam Line Pressure - Negative Rate High Steam Line Isolation N.A.
- 9. Steam Generator Water Level High High
- a. Turbine Trip < 3(3)
- b. Feedwater Isolation i7(3)
- 10. Steam Generator Water Level -Low Lnw
- a. Motor-Driven Auxiliary Feedwater Pumps 1 60
- b. Turbine-Oriven Auxiliary Feedwater Pump i 60
- 11. RWST Level--Low-Low Coincident with Safety Injection Automatic Switchover to Containment Sump 1 32(2)
- 12. Loss of Power
~
.(Less of Voltage)
~
(Tolerable Degraded Voltafle Coincident with Safety In, action)
SOUTH TEXA5 - UNIT 1 3/4 3 39 4
O O
TABLE 4.3-2 (Continued)
E ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION 5 SuavEILLANCE REQUIREFEMIS 1
DIGITAL OR TRIP ~
ANALOG ACTUATING ICOES R DEVICE MASTER SLAVE FOR %dHICH CHANNEL CHAMMEL CHMSIEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELV SURVEILLANCE
. CHMSIEL .
FUNCTI0 mat UNIT CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST, IS REQUIRED g , ,
Q 3. Containment Isolation (Continued) w
- 3) Safety Injection See Ites 1. above for all Safety Injection Surveillance Requirements.
- 4) RCS Purge Radioactivity-High 5 R M N.A. M.A. N.A. N.A. 1,2,3,4,5*,6*
- 5) Contairument Spray - See Iten 2. above for Containment Spray manual initiation Surveillance pr Requirements.
.g. Manual Initiation
!:* 6) Phase "A" Isolation- See Itee 3.a. above for Phase "A* Isolation manual initiation
- Manual Initiation Surveillance Requirements.
u 2 4. Steam Line Isolation
- a. Manual Initiation M. A. M.A. N.A. R N.A. N.A. N.A. 1, 2, 3
- b. Automatic Actuation M.A. N.A N.A N.A. N(1) M(6) Q 1,2,4--
Logic and Actuatien g$3 Relays p~.
E O g
- c. Steam'Line Pressure- S R M N.A. M.A. N.A. M.A. 3 NI Negative Rate-High %gg M N.A. M.A. M.A. N.A. 1, 2, 3' P(y
- d. Containment Pressure - S R High-2
- e. Compensated Steam Line 5 R M N.A. M.A. M.A. M.A. 1, 2, I Pressure-Low
- f. Compensated Tg -
S E
~
M N.A. N.A. N.A. N.A. 1, 2, 1 Low-Lew (inter 1ccked with P-15)
! o ATTACHMENT S l
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CONTAINMENT SYSTEMS FINAL DRAFT
) '
SURVEILLANCE Rf0VIREMENTS (Continued) 4.6.3.2 Each isolation valvesoecified in Table 3.6 all be demonstrated OPERABLE during the COLD SHUTDOWN or REFUELING MODE at least once per 18 months by:
- a. Isolation test signal, each Verify 1,APhase isolation, ngvalve that on a Phase actuates to"A" its isolation position; and
- b. Verifying that on a Containment Ventilation Isolation test signal, each purge and exhaust valve actuates to its isolation positionpd r
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.
} l L c. bih ) ib/ m a., f % '6' L .f.4i.. A sf si, 4 e 4 r % d
- n,Is lies v.lvs asksiw % ih Iset.11.n p.,;};w, O
SOUTH TEXA5 - UNIT 1 3/4 6-19 u________
PLANT SYSTEMS FINAL HAFT ATTACHMEgT LA 3 3/4.7.3 COMPONENT COOLING WATER SYSTEM 0F /0 S
P [E LIMITING CONDITION FOR OPERATION 3.7.3 At least three independent component cooling water loops shall be OPERABLE.
APPLICABILITY: MODES 1, 2, 3, and 4.
ACTION:
With only two component cooling water loops OPERABLE, restore at least three loops to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
SURVEILLANCE REOUIREMENTS 4.7.3 At least three component cooling water loops shall be demonstrated OPERABLE:
- 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. At least once per 18 months during shutdown, by verifying that:
- 1) Each automatic valve servicing safety-related equipment or isolating the non-nuclear safety portion of the system actuates to its correct position on a Safety Injection, Loss of Offsite PowegorLowSurgeTanktestsignal,asapplicable,
$No. 2) Each Component Cooling Water System pump starts automatically onaSafetyInjectionorLossofOffsitePowertestsignal,and 1
- 3) The surge tank level instrumentation which provides automatic isolation of the non-nuclear safety-related portion of the sys-tem is demonstrated OPERABLE by performance of a CHANNEL CALIBRATION test.
- 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-rectpositIonpriortoenteringMODE4followingeachCOLDSHUT00WN of greater than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> if not performed within the previous 31 days.
SOUTH TEXAS - UNIT 1 3/4 7-12
.