ST-HL-AE-1271, Forwards Rev 5 to Draft Tech Spec,Offsite Dose Calculation Manual & Process Control Program.Exemption Requested from 10CFR50,App J Re Leakage Test of Containment Air Lock
| ML20127E405 | |
| Person / Time | |
|---|---|
| Site: | South Texas  |
| Issue date: | 06/17/1985 |
| From: | Dewease J HOUSTON LIGHTING & POWER CO. |
| To: | Thompson H Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20127E410 | List: |
| References | |
| CON-#385-519 OL, ST-HL-AE-1271, NUDOCS 8506240540 | |
| Download: ML20127E405 (500) | |
Text
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Company nos,-umio,w_ m m,a= n.mi.,,,cimsmi mmmn June 17, 1985 ST-HL-AE-1271 File No: G9.6 Mr. Hugh L. Thompson, Jr. , Director Division of Licensing Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555 South Texas Project Units 1 & 2 Docket Nos. STN 50-498, STN 50-499 South Texas Project Electric Generati .g Station Technical Specifications Offsite Dose Calculation Manual Process Control Program
Dear Mr. Thompson:
Houston Lighting & Power Company (HL&P) hereby submits draft Technical Specifications for the South Texas Project Electric Generating Station (STPEGS). These draft Technical Specifications refer to the Offsite Dose Calculation Manual (0DCM) and Process Control Program (PCP); these documents are included in this submittal. As discussed during a meeting with the NRC staff on April 19, 1985, this submittal is for the purpose of initiating a review of the draft STPEGS Technical Specifications using available staff resources from cognizant NRC branches in advance of the formal Technical Specification review process which is scheduled to begin in November 1985. The NRC staff approved of this approach because of the additional time if would afford for review of the STPEGS three-train safety system design. HL&P ' will submit an update to these draft Technical Specifications in November, 1985, which will coincide with the scheduled start of NRC's formal review. As recommended by the NRC staff the attached Technical Specifications are submitted as a markup of Draft Revision 5 to the Standard Technical Specifications (STS) for Westinghouse plants. Opposite each Technical Specification page are explanations for the differences between the STPEGS Technical Specifications and the STS. Thest. comments are intended to aid NRC's review and understanding of HL&P's reasoning behind each Technical Specification. These explanations are not intended to provide justification for the design; justification for design is provided in the Final Safety Analysis Report and supporting documents. j 8506240540 850617 j PDR ADOCK 05000498 A PDR J00 S8/NRC3/k i
ST-HL-AE-1271 File No: G9.6 Page 2 In conjunction with this submittal of the draft STPEGS Technical Specifications HL&P, hereby requests an exemption to 10CFR50 Appendix J paragraph III.D.2.(b)(ii) as shown in the STS (see STPEGS Technical Specification 4.6.1.3.b.2) for the STPEGS Units 1 and 2 to require a leakage test of the containment air lock prior to establishing containment integrity only when maintenance has been performed on the air lock that could affect seating capability. Appendix J would require a leakage test of the air lock prior to establishing containment integrity anytime the air lock was opened. As approved by the NRC Safety Evaluation Report on the Westinghouse i Owners Group (WOG) Technical Specification Optimization Program (TOP) for Reactor Trip System Instrumentation, WCAP 10271, these draft Techncial Specifications include the following changes to the STS:
- 1) An increase in the surveillance interval for Reactor Protection System (RPS) analog channel operational tests to once per quarter;
- 2) An increase in the time which an inoperable RPS analog channel may be maintained in the untripped condition to six hours;
- 3) An increase in the time an inoperable RPS analog channel may be bypassed to allow for testing of another channel in the same function to four hours; and
- 4) RPS analog channel testing in a bypassed condition.
HL&P makes the following commitments relative to the conditions placed on utilities implementing the above changes.
- 1) HL&P will implement a staggered test plan for the RPS analog channels which will be tested quarterly as specified in the attached Technical Specifications.
- 2) HL&P will implement plant procedures to require a common mode evaluation for failures in those RPS analog channels which are tested quarterly and HL&P will perform additional testing or other appropriate remedial action if common mode failure is suspected.
S8/NRC3/k
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ST-HL-AE-1271 File No: G9.6 Page 3
- 3) HL&P is investigating installation of hardware to support testing of the RPS analog channels in bypass. The attached Technical Specifications authorize testing in bypass for those RPS analog channels with the required hardware. The attached Technical Specifications are written such that the installation of additional hardware will not require a Technical Specification change and that testing in bypass without the appropriate hardware is not authorized.
- 4) The attached Technical Specifications require that portions of RPS analog channels which provide input to Engineered Safety Features Actuation System (ESFAS) channels be tested at the frequency specified for the ESFAS channel.
- 5) HL&P,will collect RPS setpoint drift data over a twelve month period to demonstrate the adequacy of existing setpoints and allowable values once quarterly testing has begun. -Any necessary changes to the setpoints and allowable values will be made based on a review of the above data.
If you have any questions, please contact Mark McBurnett at 512-972-3651 extension 6054. Very truly yours, l J M4D
.. G. Dewease ice President Nuclear Plant Operations MAM/bjf Attachments: Technical Specifications Offsite Dose Calculation Manual Process Control Program S8/NRC3/k
g. Ilouston Lighting & Power Company ST-HL-AE-1271 File No.: G9.6 Page 4 cc: Robert D. Martin Brian E. Berwick, Esquire Regional Administrator, Region IV Assistant Attorney General for Nuclear Regulatory Commission the State of Texas 611 Ryan Plaza Drive, Suite 1000 P. O. Box 12548, Capitol Station Arlington, TX 76011 Austin, TX 78711 N. Prasad Kadambi, Project Manager Lanny A. Sinkin U.S. Nuclear Regulatory Commission 3022 Porter Street, N.W. #304 7920 Norfolk Avenue Washington, D.C. 20008 Bethesda, MD 20814 Oreste R. Pirfo, Esquire Claude E. Johnson Hearing Attorney Senior Resident Inspector /STP Office of the Executive Legal Director c/o U.S. Nuclear Regulatory Commission U.S. Nuclear Regulatory Commission P. O. Box 910 Washington, DC 20555 Bay City, TX 77414 Charles Bechhoefer, Esquire M. D. Schwarz, Jr. , Esquire Chairman, Atomic Safety & Licensing Board Baker & Botts U.S. Nuclear Regulatory Commission One Shell Plaza Washington, DC 20555 Houston, TX 77002 Dr. James C. Lamb, III J. R. Newman, Esquire 313 Woodhaven Road Newman & Holtzinger, P.C. Chapel Hill, NC 27514 1615 L Street N.W. Washington, DC 20036 Judge Frederick J. Shan Atomic Safety and Licensing Board Director, Office of Inspection U.S. Nuclear Regulatory Commission and Enforcement Washington, D. C. 20555 U.S. Nuclear Regulatory Commission Washington, DC 20555 Mr. Ray Goldstein, Esquire 1001 Vaughn Building E. R. Brooks /R. L. Range 807 Brazos Central Power & Light Company Austin, TX 78701 P. O. Box 2121 Corpus Christi, TX 78403 Citizens for Equitable Utilities, Inc. c/o Ms. Peggy Buchorn H. L. Peterson/G. Pokorny Route 1, Box 1684 City of Austin Brazoria, TX 77422 P. O. Box 1088 Austin, TX 78767 Docketing & Service Section Office of the Secretary J. B. Poston/A. vonRosenberg U.S. Nuclear Regulatory Commission City Public Service Board Washington, DC 20555 P. O. Box 1771 San Antonio, TX 78296 S8/NRC3/a Revised 5/22/85
pm-o ST-HL-AE-1271 File No.: G9.6 Page 5 cc: Mr. David K. Lacker,. Chief Bureau of Radiation Control Texas Department of Health
-1100 West 49th Street Austin, Texas 78756 S8/NRC3/a
Technical
- Specifications Sou
- 1 exes 3ro'ec: .
.li~:s'&2 i
i l COMPARISON WITH ; NUREG 0452, REVISION 5 (Draft) i WE NGH U E PR RI D 1 R R ACT RS I l l I
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( l e i l I I l 4 l . O NUREG 0452/STPEGS COMPARISON
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l __.....__..t_.. . . . , ~ _ _ O ( . INDEX DEFINITIONS SECTION PAGE 1.0 OEFINITIONS L 1 ACTI 0N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-1 L 2 ACTUATION LOGIC TE57. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-1 1.3 ANALOG CHANNEL OPERATIONAL TEST.............................. 1-1 L4 AXIAL FLUX 0 !FFERENCE. . . . . . . .,. . . . . . . . .'. . . . . . . . . . . . . . . . . . . . . . . . 1-1 , 1.5 CHANNEL CALIBRATION........................................... 1-1 L6 CHANNEL CHECK................................................. 1-1 1.7 CONTAINMENT INTEGRITY......................................... 1-2
. .-..- ..r. . . = : -_ -..=.~.............................................
- _ r_ r 1.9 CORE ALTERATION............................................... 1-2 L10 005E EQUIVALENT I-13L . . . . . . . . . .~ . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-2 Lil 1-AVERAGEDISINTEGRATIONENERGY.............................'. 1-2
- s. *
'. 1.12 ENGINEERED SAFETY FEATURES RESPONSE TIME..................... 1-3 .
L13 FREQUENCY N0TATION........................................... 1-3 1.14 IDENTIFIED LEAXAGE........................................... 1-3 L 15 MASTER RELAY TE5T. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-3 .* 1.16 MEM8ER(5)0FTHEPU8LIC...................................... . 1-3 1.17 0FFSITE DOSE CALCULATION MANUAL.............................. 1-3 1.18 QPERA8LE - OPERA 81LITY....................................... 1-4 , 1.19 CPERATIONAL MODE - M00E...................................... 1-4 L20 PHYSICS TEST 5................................................ 1-4 1.21 PRESSURE BOUNDARY LEAKAGE.................................... 1-4 1.22 PROCESS CONTROL P R0 GRAM. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-4 > 1.23 PURGE - PURGING.............................................. 1-4 ,! 1.24 QUADRANT POWER TILT RATI0.................................... 165 1.25 RATED THERMA L P0WER. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-5 i L 26 REACTOR TRIP SYSTEM RESPONSE TIME............................ 1-5 1.27 REPORTA8LE EVENT............................................. 1-5
- 1. .i- ^iittt9-99tt9tMG=tNTEGRITi. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-5 I
- 1. SHUTDCWN MARGIN.......................,....................... 1-5
( 1. SITE 80UNCARY................................................ 1-5
- 1. g S uvE RE uv TEST. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-s y-sTs I NUREG 0452/STPECS ,
COMPARISON
.. .. . . . . . . . . . ~ . . . . . . . . . . . . - . .. .. . .... . . . .
C
\.
\. INDEX .
DEFINITIONS SECTION . PAGE I 1 5 SOLIDIFICATION............................................... 1-6 1.3)a SOURCE CHECK................................................. 1-6'
- 1. STAGGERED TEST BASI 5......................................... 1-6 1 4 THERMAL P0WER................................................ 1-6 1.3 TRIP ACTUATING DEVICE OPERATIONAL TE5T. . . . . . . . . . . . . . . . . . . . . . . 1-6 1.37 UNIDENTIFIED LEAKAGE......................................... 1-&
- 1. UNRESTRICTED AREA............................................ 1-7 1.3 VENTILATION EXHAUST TREATMENT 5YSTEM......................... 1-7 1 VENTING...................................................... 1-7
- 1. [ WASTE GAS HOLDUP SY5 TEM...............'....................... 1-7 TA8LE 1.1 FREQUENCY N0TATION...................................... 1-8
' TABLE 1.2 OPERATIONAL M0bES.............:......................... 1-9 J V.
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PSTS II NUREG 0452/STPEGS COMPARISON
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SAFETY LIMITS AND LIMITING SAFETY SYSTEM SET /INGS . SECTION PAGE 2.1 SAFETY LIMITS , 2.1.1 REACTO R C0 RE. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-1 2.1.2 REACTO R COO LANT. SYSTEM PRE 55URE. . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-1 . FIGURE 2.1-1 REACTOR' CORE SAFETY LIMIT - FOUR LOOPS IN OPERATION.. 2-2 fid:di 2. _-2 EU.GT05 C055 id.'ih d.-ii - Tadci s ?i la 0Fissilva. 2--! 2.2 LIMITING SAFETY SYSTEM SETTINGS
- 2. 2.1 REACTOR TRIP SYSTEM INSTRUMENTATION S ETP0!NTS. . . . . . . . . . . . . . - 2-4 TA8LE 2.2-1 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP,SETPCINTS.... 2-5 BASES
( SECTION PAGE .. 2.1 SAFETY LIMITS . 2.1.1 R EACTO R C0 R E. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 2-1 2.1.2 REACTOR COOLANT SYSTEM PRE 55URE............................. 8 2-2
- 2. 2 LIMITING SAFETY SYSTEM SETTINGS 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION'5ETPOINTS............... 8 2-3 l
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]f-STS III NUREG 0452/STPEGS
! COMPARIS0N . l
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. . . ' " .- , . , .. D ".' .~ - - - - -- --. .-
\ INDEX .
LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENT 5' SECTION PAGE 3/4.0 APPLICA81LITY............................................... 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 80 RAT!0h CONTROL Shutdown Margin - T avg Greater Than 200*F................ 3/4 1-1 Shutdown Margin - T avg Less Than or Equal to 200*F....... 3/4 1-3 Moderator Temperature Coe ffi ci ent. . . . . . . . . . . . . . . . . . . . . . . . 3/4 1-4 Minfaus Temperature for Criticality...................... 3/4 1-6 3/4.1.2 80 RATION SYSTEMS Flow Path - Shutdown..................................... 3/4 1-7
,, Flow Paths - Operating................................... 3/4 1-8 Chargi ng Pump - 5hutdown. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 1-9 g ., Charging Pumps - Operating............................... 3/4 1-10
( Berated Water Source - 5hutdown.......................... 3/4 1-11 Borated Water Sources - 0perating........................ 3/4 1-12
.t 3/4.1.3 MOVA8LE CONTROL ASSEMLIES - '
Group Height............................................. 3/4 1-14 TA8LE 3.1-1 ACCIDENT ANALYSES REQUIRING REEVALUATION IN THE EVENT OF AN IN0PERA8 LE NR00. . . . . . . . . . . . . . . . . . . 3/4 1-16 Position Indication Systems - Operating. . . . . . . . . . . . . . . . . . 3/4 1-17 Position Indication System - Shutdown.................... 3/4 1-18 Rod Orop Time............................................ 3/4 1-19 Shutdown Rod Insertion Limit............................. 3/4 1-20 Control Rod Insertion Limits............................. 3/4 1-21 FIGURE 3.1-1 R00 BANK INSERTION LIMITS VER$US THERMAL POWER FOUR- LOOP 0P ERATION. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 1-22 _- ..- ::: : : _ : 22n 2 . . . .:_:: =..n._.. : = : . - - - - - - r_: r- . r_::
. - - . . . . . . - -- - .i -
STMRI.____.._..-..__--................................7
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NUREG 0452/STPEGS W-STS
- IV COMPARISON
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INDEX - LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENT 5- _; SECTION . PAGE * ' 3/4.2 POWER DISTRI!!#10X LIMITS 3/4.2.1 AXIAL FLUX O!FFERENCE. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 2 FIGURE 3.2-1 AXIAL FLUX DIFFERENCE LIMITS AS A FUNCTION OF RATED TEERMAL P0WER...................................... 3/4 2-3 3/4.2.2 NEAT FLUX MOT CHANNEL FACT 0R. . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 2-4 FIGURE 3.2-2 K(Z) - NORMALIZED F (Z) . q AS A FUNCTION OF CORE MEIGNL 3/4 2-5 3/4.2.3 RCS Flaw RATE AND NUCLEAR ENTHALPY RISE NOT CHANNEL FACT 0R................................................. 3/4 2-8 iiws65 ..s . 5.a I5 *
- DS':0PERA M C m ............................................ 2/ 2 , , 3/4.2.4 QUADRANT POW G TILT RATI0................................ 3/4 2-11 3/4.2.5 DN8 PARAMETERS....................................,....... . 3/4 2-14 95:_ L F1 =E :i:5 r :i._... ............................... . j/; 25--l
{ 3/4.3 INSTRUMENTATION
- k. 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION...................... 3/4 3-1 TA8LE 3.3-1 REACTOR TRIP ' SYSTEM INSTRUMENTAIIGN. . . . . . . . . . . . . . . . . . . 3/4 3-2
~'
TA8LE 3.3-2 REACTCR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES.... 3/4 3-9 TA8LE 4.3-1 REACTOR TRIP SYSTEM INSTRUMENTATICM SURVEILLANCE REQUIREMENTS...........................;.........'........ 3/4 3-11 3/4.3.2 ENGINEERED SAFETY FEATURES ACTUATION SYL'EM INSTRUMENTATION........................................ 3/4 3-16 TA8LE 3.3-3 ENGINEERED SAFETY FEATURES MTUATION S'tSTOt INSTRUMF.NTATION.......................................... 3/4 3-18 TA8LE 3.3-4 ENGINEERED SAFETY FEATURES ACTUATT 3 SYSTEM INSTRUMENTKi!ON TRIP SETPOINTS... v...................... 3/4 3-30 TA8LE 3.3-5 ENGINEERED SAFETY FEATURF' it M y d T*.M*.S............. 3/4 3-37 TA8LE 4.3-2 ENGINEERED SAFETY FEATUk% TCW. f ;N SYSTEM INSTRUMENTATION SURVEILLANCE ',tEQUIhinLNT5. . ; . . . . . . . . . . . . . 3/4 3-42 3/4.3.3 MONITORING INSTRUMENTATION R'adiation Moni toring For Plant Operations. . . . . . . . . . . . . . . . 2/43-47l ( V-STS V NUREG 0452/STPEGS O( l COMPARISON
[. . _ _ . . . _ _ . . . _ _ . _ . . _. ._ : .' e L RLO.E] . LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS
*' SECTION m
. s
- . PAGE TA8LE 3-3-6 RADIATIONMhNITORINGINSTRUMENTATION - l
\ FOR PLANT OPERATIONS..................................... 3/4 3-48 l
- TA8LE 4.3-3 RADIATION MONITORING INSTRUMENTATION FOR PLANT
- e s OPERATIONS SURVEILLANCE REQUIREMENTS..................... 3/4 3-50
\3 Movable Incore Detectors................................. 3/4 3-51 Seismic Instrumentation.................................. 3/4 3-52 TABLE 3.3-7 SEISMIC MONITORING INSTRUMENTATION.................... 3/4 3-53 J TA8LE 4.3-4 SEISMIC MONITORING INSTRUNENTATION SURVEILLANCE REQUIREMENTS............................................. 3/4 3-54 Meteorological Instrumentation........................... . 3/4 3-55 TABLE 3.3-8 METEOROLOGICAL MONITORING INSTRUMENTATION. . . . . . . . . . . . . 3/4 3-56 TA8LE 4.3-5 METEOROLOGICAL MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS............................................. 3/4 3-57 i
Remote Shutdown Instruentation.......................... 3/4 3-58 TA8LE 3.3-9 REMOTE SHUTDOWN MONITORING INSTRUMENTATION............ 3/4 3-59
.s TA8LE 4.3-6 REMOTE SHUTDOWN MONITORING INSTRUMENTATION
( SURVEILLANCE REQUIREMENTS................................ 3/4 3-60 Accident" Monitoring Instrumentation......'................ 3/4 3-61 TABLE 3.3-10 ACCIDENT M NITORING INSTRUMENTATION.................. 3/4 3-62
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TA8LE 4.3-7 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE i REQUI R E TENT 5. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 3-64
= = w __..... .. _............................... .. . . ..
Fire Detection Instrumentation........................... 3/4 3-67 TA8LE 3.3-11 FIRE DETECTION INSTRUMENTATION....................... 3/4 3-68 l Loose-Part Detection 5ystas.............................. 3/4 3-69
... Radioactive Liquid Effluent Monitoring Instrumentation... 3/4 3-70 TABLE 3.3-12 RAO!0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION 3/4 3-71 TABLE 4.3-8 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS................ 3/4 3-74 Radioactive Gaseous Effluent Monitoring Instrumentation.. 3/4 3-77 TA8LE 3.3-13 RADI0 ACTIVE GASEDUS EFFLUENT MONITORING INSTRUMENTATION.......................................... 3/4 3-78 TA8LE 4.3-9 RADI0 ACTIVE GASE0US EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS................ ,
3/4 3-85
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INDEX . LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION . PAGE
.. 3/4.4 REACTdRCOOLANTSYSTb4 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATIQN .
Startup and Power 0peration. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 4-1 Hot Standby.............................................. 3/4 4-2 Ho t 5hutdown. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 4-3 Cold shutdown - Loops F111ed............................. 3/4 4-5 Cold Shutdown - Loops Not Ft11ed......................... 3/4 4-6 I-.? L _ __ ,.. .........................................
.u , , r=
___ ____ ___, .~-.-_,............... .................... 3/4.4.2 SAFETY VALVES 5hutdown............................................... 3/4 4-9 . Operating............................................. 3/4 4-10 3/4.4.3 PRE 55URIZER.............................................. 3/4 4-11 g h 3/4.4.4 RELIEF VALVES............................................ 3/4 4-12 y ( 3/4.4.5 ST u M GENERATOR 5......................................... 3/4 4-14 TA8LE 4.4-1 MINIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTED DURING INSERVICE INSPECTION............................. 3/4 4-19 .. TABLE 4.4-2 ST DM GENERATOR TUBE INSPECTION....................... 3/4 4-20
. 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection 5ystass................................' 3/4 4-21 Op e rati o nal Le a kage. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 4-22 TA8LE 3.4-1 REACTOR COOLANT SYSTEM PRESSURE ISCLATION VALVES...... 3/4 4-24 3/4.4.7 CHEMISTRY................................................ 3/4 4-25 TA8LE 3.4-2 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS............ .. 3/4 4-26 TA8LE 4.4-3 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS SURVEILLANC REQUIREMENTS...,......................................... 3/4 4-27 3/4.4.8 SPECIFIC ACTIVITY........................................ 3/4 4-28 FIGURE 3.4-1 DOSE EQUIVALENT I-131 REACTOR COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL PCWER WITH THE REACTOR COOLANT SPECIFIC ACTIVITY > 1pCf/ gram DOSE EQUIVALENT I-131.................................... 3/4 4-30 TABLE 4.4-4 REACTOR COOLANT.5PECIFIC ACTIVITY SAMPLE AND ANALYSIS PR0 CRAM.................................................. 3/4 4-31
( - O W-STS
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VII NUREG 0452/STPEGS COMPARISON .
r INDEX . N. LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION . PAGE 3/4.4.9 PRESSURE / TEMPERATURE LIMITS Reactor Coolant Systas.............. .................... 3/4 4-33 FIGURE 3.4-2 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS - APPLICA8LE UP 70 EFPY.............................. 3/4 4-34
. FIGURE 3.4-3 REACTOR C00LANT SYSTEM C00LDOWN LIMITATIONS - ,
APPLICABLE UP TO EFPY.............................. 3/4 4-35 TABLE 4.4-5 REACTOR YESSEL MATERIAL SURVEILLANCE PROGRAM - WITHORAWAL SCHEDULE...................................... 3/4 4-36 Pressurizer.............................................. 3/4 4-37 Overpressure Protection Systems.......................... 3/4 4-38 3/4.4.10 STRUCTURAL INTEGRITY..................................... 3/4 4-40 3/4.4.11 REACTOR COOLANT SYSTEM VENTS............................. 3/44-41l E5 EMERGENCY CORE COOLING SYSTEMS
,s 3/4.5.1 ACCUMULATORS............................................. 3/4 5-1
(. 3/4.5.2 ECCS SU85YSTEMS - Tavg GREATER THAN OR EQUAL TO 350*F.... 3/4 5-5
.t 3/4.5.3 ECCS SUS $YSTEMS - Tavg LESS THAN 350*F................... 3/4 5-9 3/A.5. 555U" 5"U5'_ : b" b2:""
-s ' - **
. :1 _c-3/4.5.5 REFUELING WATER STORAGE TANK............................. 3/4 5-13 w-STS VIII
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NUREG 0452/STPEGS i COMPARISON I l l
)
- ~ .
. .,s INDEX ,
gJ. y. . LIMITING CON 0!TIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE W - ATMOSPHERIC TYPE CONTAIMENT . 3/4.6 CONTAINMENT SYSTEMS. 3/4.6.1 PRIMARY CONTAIMENT
. Containment Integrity.................................... 3/4 6-1A Containment Laakage...................................... 3/4 6-2A Contai nment Ai r Locks. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 6-5A
" p.-i._ _ ; _
- !:-i!:: ^i:1:: == ^^ :: : ! 2-1--
.... ;.;; :.:n :y: _ _................................. _,_ _ _,c : :-
Air Temperature.......................................... 3/4 6-9A
-tontai.. .- -i ;ssi 5;ruc.urai integri;y. . . . . . . . . . . . . . . . . . ./ e aca j Containment Ventilation 5ystas........................... 3/4 6-17A ,
3/4.6.2 DEPRESSURIZATION AND COOLING SYSTENS
\
Containment Spray 5ystaa................................. 3/4 6-19A ,, Spray Additive Systas.................................... 3/4 6-22A Containment Cooling 5ystas............................... 3/4 6-23A l 2/=.5.2 1001-5 0.E.a0F 5h m .................................... J/- 5 .;5a 3/4.6.4 CONTAIMENT ISC LATION *4 VALVE 5. . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 6-27A TA8LE 3.6-1 CONTAIMENT ISO LATION VALVE 5. . . . . . . . . . . . . . . . . . . . . . . . . . 3/46-29Al 3/4.6.5 COMBUSTIBLE GA5 CONTROL Hydrogen Nonitors........................................ 3/4 6-30A El ectric Hydrogen Roccabi ners. . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 6-31A y . . . . . . .; - - - . . -;. : - __ __ __ 3:': ^-32 2,- ;- 20 I"; 37 :;- ___ .................... . . 3[ 5-3-- ,
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LIMI CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION *
. PAGE
.. M-ICECONDENSERTYPECONTAINMENT 3/4.6 CONTAINM SYSTEMS 3/4.6.1 PRIMARY NTADetENT Containeen Integrity............................ ....... 3/4 6-18 Containment akage............................ ......... 3/4 6-28 TABLE 3.6-1 SECONDARY AINMENT SYPA55 LEAKAGE PATHS ........... 3/4 6-58 l Containment Air ks...................'... ............. 3/4 6-68 Containment Isolat n Valve and Channel Id Pressurization Sys ............... ................. 3/4 6-88 i
, Internal Pressure..... .............. ................... .
3/4 6-98
! Air Temperature......... ......'.... ..................... 3/4 6-108 s I Containment Vessel Structu al In grity.................. 3/4 6-118
^
Shield Building Structural ity..................... 3/4 6-128 5.\ Shield Building Air Cleanup ten....................... ( 3/46-138l Containment Ventilation 5y tes.. ........................ 3/4 6-158
^
3/4.6.2 DEPRESSURIZATION AND NG SY
. .'s Containment Spray 5 ............. ................... 3/4 6-178 Spray Additive Sys .................. ............'..... 3/4 6-208 Containment Cool g 5ystas............... ............... 3/4 6-218 3/4.6.3 I0 DINE CLEANU 5YSTEM....................... ............ 3/4 6-248 2
3/4.6.4 CONTAINM ISOLATION VALVE 5................... .......... 3/4 d-268
- TABLE 3.6,-2 CONTA ISCLATION VALVE 5................. ........
i 3/4 6-288 l l 3/4.6.5 C0feu BLE GAS CONTROL Hy gen Monitors.................................... ... 3/4 6-298 octric Hydrogen Recombiners. . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 6-308 i Hydrogen Control Distributed Ignition Systas............. 3/46-318l Hydrogen Purge Cleanup 5ystas............................ 3/4 6-328 Rydrogen Mixing 5ystas.............................. ... 4 6-348
- y-STS IX-8 NUREG 0452/STPEGS COMPARISON
--,,_---.,_-.,_____--___,_,_,,._____,..,_,__my,, . , - _ _ _
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- ~
l D\SORD :" INDEX . LIMITING ONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE y-SU8ATH05PHERICTYPECONTAINMENT 3/4.6 CONTAINMENT SYSTEMS ,'
. 3/4.6.1 PRIMARY CO INH 5NT,,
Containment I agrity.............................. ..... 3/4 6-1C Containment Lea go......,........................ ....... 3/4 6-2C Conta f neent At r Lo ks. . . . . . . . . . . . . . . . . . . . . . . . . ......... 3/4 6-5C
. Containment Isolatio Valve and Channel Weld Pressurization Syst s................... ............. 3/4 6-7C
~
Internal Pressure....... ............... ................ 3/4 6-8C
'~ FIGURE 3.6-1 MAXIMUM ALLOWA8LE PRIMARY CONTAINME AIR PRESSURE VERSUS RIVER WATER TEMPERAfuRE AND WATER TEMPERATURE.................. ..... ..................... 3/4 6-9C
~
.Afr Temperature................ . ..'..................... 3/4 6-10C h, FIGURE 3.6-2 MINIMUM ALLOWA8LE PRIMARY NTAINMENTAVERAGEAIR TEMPERATURE VERSUS RIVER WATE TEMP RATURE............... 3/4 6-11C -
(~ Containment Vessel Structur l'Integr ty.................. 3/4 6-12C Containment Ventilation stas......... ................. 3/4 6-1SC 3/4.6.2 DEPRES$URIZATION AND LING SYSTEMS ,, Containment Quench pray Systas............. ............ 3/4 6-21C i Recirculation Spr y.Systas.................... .......... 3/4 6-22C l l Spray Additive ystas........................... ........ 3/4 6-23C l 3/4.6.3 CONTAINMENT SOLATION VALVES....................... ..... 3/4 6-24C TA8LE 3.6-1 CONTAI ENT ISOLATION VALVES...................... ... 3/4 6-26C l 3/4.6.4 CCMBU BLE GAS CONTROL Hyd g e n Mo n i to rs . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 6-27C E ctri c Hydroge n Recomb iners. . . . . . . . . . . . . . . . . . . . . . . . . . . . /4 6-28C drogen Purge Cleanup 5ystaa............................ 3 6-29C
'Mydrogen Mixing Systas................................... 3/4 -31C 3/4. 5 SUBATHOSPHERIC PRESSURE CONTROL SYSTEM Staas Jet Air Ejector..............'...................... ,3/4 6-3 3/4 6-33C
, Mechanical Vacuus Pumps..................................
( 3/4 6-34C 3/,4.6.6 VACUUM RELIEF VALVES..................................... W-STS IX-C
- NUREG 0452/STPEGS COMPARISON
r 1
. i
' "~
, - . - -- .- . - - - , . . -- - - "~ . .~~ T.~:e < . a. . .L . . ~ ... --.l-.- . T ~ C.- --
INDEX ,
~
IMITINGChNDITIONSFOROPERATIONANDSURVEILLANCEREQUIREMENTS SECTI . AGE M - DUAL TYPE CONTAINMENT 3/4.6 CONTA NMENT SYSTEMS 3/4.6.1 PR CONTAIMENT Contai nt Integrity................................ .... 3/4 6-1D Containee Leakage.....l......................... ...... 3/4 6-20 TA8LE 3.5-1 SE CONTAINMENT SYPASS LEAXAGE PATHS... ........ 3/46-50l Containment Af Locks........................ ........... 3/4 6-60 Containment Isol tion Valve and Channel Wel Pressurization 5) stems................................ 3/4 6-80 Internal Pressure.... ................. ................. 3/4 6-90 Air Temperature........ ............. ................... 3/4 6-100 ,- Containment Vessel Strue ral Inte ity.................. 3/4 6-11D
',N Containment Ventilation System.. ........................ 3/4 6-12D -
I' 3/4.5.2 DEPRESSURIZATION AND COOLING 157 EMS
\
Containment Spray Sys tem. . . . . . . . . . . . . . . . . '. . . . . . . . . . . . . . 3/4 6-140 , , Spray Additive System... .......... ..................... 3/4 6-17D Containment Cooling Sy .......... ................... 3/4 6-180 3/4.6.3 IODINE CLEANUP SY ................... ................ 3/4 6-200 3/4.6.4 CONTAINMENT ISO ON VALVES............... ............. 3/4 6-220 TA8LE 3.5-2 CONTA! ISCLATION VALVES.............. ........... 3/4 6-240 l 3/4.6.5 CopeUSTIB GAS CONTROL Hydrogen nitors................................. ..... 3/4 6-250 j Electr c Hydrogen Recombiners. . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 6-260 Hyd en Purge Cleanup Systas.......................... . 3/4 6-27D drogen Mixing 5ystas................................... 3/4 6-290 3/4.6.6 PENETRATION ROOM EXHAUST AIR CLEANUP SYSTEM............... /4 6-300 . 1 3/4. VACUUM RELIEF VALVES...................................... 3/ 6-320 ; i 1 NUREG 0452/STPEGS W-575
- IX-0 COMPARISON J
, ) i
* "~~~~
... ,.: - - - - . . . - . . . . . . . ....-....7.'.'~. '.'- .
( INDEX , L NG CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION .. . g
. 3/4.6.8 SECONDARY CONTAINM Shield Building Air C up 5ystas................... .... 3/4 6-330 Shield Buildfng Integrity... ..................... ...... 3/4 6-350 Shield Building Structural In ty............ ......... 3/4 6-360 g .
6 _ G. l l (. - PSTS X-D NUREG 0452/STPEGS O COMPARISON
r ] l l
.. . - ~- .. ~~- - - - ;
O.,
\
i . INDEX . @8 Pl ' MITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.6.6 PEN ION ROOM EXHAUST AIR CLEANUP SYSTEM.. ........... 3/4 6-358 3/4.6.7 ICE CONDENS . Ice 8ed.......... ................. .................... 3/4 6-378 Ice led Temperature nitoring staa...........*......... 3/4 6-398 Ice Condenser Doors...... . ............................ 3/4 6-408 InletDoorPositionMpa r 5ystes.................... . 3/4 6-428 Divider 8arrier Perjonnel Access ors and ... EquipmentHatepas..................................... 3/4 6-438 ContainmentfffRecirculationSystems..................g 3/4 6-448 Fl oor Djr ms . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ........... 3/4 6-458 Refupling Canal Drains............. ............ ....... 3/4 6-468
~
(91 der 8arrier Sea 1................................ ... 3/4 6-478 O' \ s, TA8 .6-3 OIVIDER 8ARRIER SEAL ACCEPTA8LE PHYSICAL PROPERTIES... 3/46'488l -
/
- 4) 4.6.8 VACUUM RELIEF VALVE 5..................................... 3 -49B t
NUREG 0452/STPEGS PSTS x-8 COMPARISON
~--
i 1
.. - . . =
( IN0Ex DRTI i.INITING CONDITIONS FOR OPERATION ANO SURVEILLANCE REQUIREMENTS SECTION PAGE l 3/4.7 PLANT SYSTEMS (
- 3/4.7.1 TURSINE CYCLE Safety Va1ves..............*............................. 3/4 7-1 J
TABLE 3.7-1 MAXIMUM ALLOWA8LE POWER RANGE NEUTRON FLUX HIGH . SETPOINT WITH INOPERA8LE STEAM LINE SAFETY VALVES OURING FOUR LOOP 0PEAATION...................................... 3/4 7-2
--meta-er+=*-nan amn-attwwaste-ruw en-nere-n ev e nun-rtva-mwt--
= 88TPotNT:W3Dt-tMCPERA8 tc-39 09 ?ji 0- s. i-1 l
TA8LE 3.7-[ 5 TEAM LINE SAFETY VALVES PER L00P. . . . . . . . . . 3/4 . . . 7-3 i - Auxilia Feedwate r 5y s taa. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 3/4 7-4 FK E DWKTT8-y=uAn[Wg$torageTank.....<............................
... .. 3/4 7-6 Specific Activity........................................ 3/4 7-7 TA8LE 4.7-1 SECONDARY COOLANT SYSTEM SPECIFIC ACTIVITY SAMPLE
ANO ANALYSIS PR0 GRAM..................................... 3/4 7-8 s I s Main Staam Line Isolation Va1ves................'......... 3/4 7-9 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION.......... 3/4 7-10 3/4.7.3 COMPONENT COOLING M TER SY5 TEM........................... 3/4 7-11 '
essegrem c.couW 3/4.7.4 = ts y TER 5YSTEM.................................*.... 3/4 7-12 l 3/4.7.5 ULTIMATE HEAT 51NK....................................... 3/4 7-13
=ar .c.- . . . -s -----____.
r ws rirs t es ; w;;. . . . . . . . . . . . . . .
. y .,., 4...............
3/4 7-15 3/4.7.7 CONTROL ROOM EMERGENCY 4 Y5 TEN................ gusr a. 3/4.7.8 -post,g e p u m sowo -50M=mECC5 EXHAUST.-R =At J . . r ; ; ; ; ;.r; . . . . . . . . . . . . . . . . 3/4 7-18 3/4.7.9 SNUB 8ERS................................................. 3/4 7-20 FIGURE 4.7-1 SAMPLE PLAN 2) FOR SNUBBER FUNCTIONAL TEST. . . . . . . . . . . 3/4 7-25 l 3/4.7.10 SEAtE0 SOURCE CONTAMINATION.............................. 3/4 7-28 O NUREG 0452/STPEGS W-STS XI COMPARISON .
7
' = 0RH LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.7. N FIRE SUPPRESSION SYSTEMS Fire Suppression Water Systas................... ........ 3/4 7-28
- 5 prey ,and/or Sprinkl er 5ystems. . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 7-31 Ha1on 5ystems............................................ 3/4 7-35
- Fire Hose 5tations....................................... 3/4 7-36 3/4'7-37l TA8LE 3.7-4 FI RE HOS E STATION 5. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
.... _ ,r. - . . - . . . ,,, , ,, t 3/4.7 S FIRE RATED A55E)SLIES.................................... 3/4 7-40 a 11- - 9 i _-- .- i- --3 = E --i . . . . . . . . . . . . . . . . . . . . . . . . . . . . .,'; ;.
. - .., . an n . - - . . . mn..n. ........................... . , , , ,. .
.\ 3/4.8 ELECTRICAL POWER SYSTEMS
(. 3/4.8.1 A.C. SOURCES ,
. . ei Operating................................................ 3/4 8-1 TABLE 4.8-1 DIESEL GENERATOR TEST SCHEDULE. . . . . . . . . . . . . . . . . . . . . . . . 3/4 8-8 l Shutdown........ 3/4 8-9 o ps enig g - o re s :T d.To. . ou . . .sivt.
. . . . .otsvaisirrion
. . . . . . . . . . . .. .... . .. ... ,. . s. ... .. ... . . .. . 6. J/4 8-Sa.,
5hatd m - awastrs. To o*85m WTened . . . . . . . . . .. . . . J/4 8 -5h 3/4.8.2 D.C. SOURCES 0perating................................................ 3/4 4-10 TABLE 4.8-2 8ATTERY SURVEILLANCE REQUIREMENTS..................... 3/4 8-12 l Shutdown................................................. 3/4 8-13 3/4.8.3 ONSITE POWER DISTRIBUTION Operating................................................ , 3/4 8*14 Shutdown......................'........................... 3/4 8-16 W-STS XII - NUREG 0452/STPECS COMPARISON
1
,. . . a . . . ~. ... C : . T.:. ': " ~ T.~.? ~~
g . 1 LIMITING CON 0!TIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS
- l
. \
SECTION PAGE 3/4.4.4 ELECTRICAL EQUIPN S T PROTECTIVE DEVICES l M inary-C.-- i. -.n................. d- i-i.7^ ; Containment Penetration Conductor Overcurrent Protective 0evices..................................... 3/4 8-18 TA8LE 3.8-1 CONTAINMENT PDETRATION CONOUCTOR OVERCURRENT
..,, PROTECTIVE 0EVICES....................................... 3/4 4-20
- - ---- u ,: = = - :::: - :-- :..ur:- rer - - m e ...
reu
-tant-e-+-e-t-tems.nessaven-va+ve+-vwomat-ewsentan-eenteretav----
*- =AN0/04:4YSA$8:08VtC85.................................... 2/- 5-22 3/4.9 REFliELING OPERATIONS 6'
( 3/4.9.1 BORON C0NCBTRATION...................................... 3/4 9-1 s 9 3/4.9.2 INSTRUMENTATION.......................................... 3/4 9-2 3/4.9.3 DECAY TIME............................................... 3/4 9-3 3/4.9.4 CONTAINMENT 8UILDING P D ETRATI0NS........................ 3/4 9-4 3/4.9.5 CCMJN I CATIO NS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 9-5 3/4.9.8 MANIPULATOR CRANE.REFM f*4.W 9. .M4 W.W: . . . . . . . . . . . . . . . . 3/4 9-6 FunLymoduq euu.bhei . 3/4.9.7 CRANE TRAVEL g6 Hat-Fvii.--sivRi==-iws.-iviutimrv. . . . . . . . 3/4 9-7 3/4.9.8 RESIDL'AL HEAT RDOVAL AND COOLANT CIRCULATION Hf,gh Water Leve1......................................... 3/4 9-8 Low Water Leve1.......................................... 3/4 9-9 ( NUREG 0452/STPEGS O W-STS XI!! COMPARISON ,
,:- . . * .. s .. = .... .. . .. . . .
- n. -
INDEX f., LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS
.,t SECTION PAGE 4 t/i " ^ ^8NTAINMMT-Mien 8-AND-HHAUST-T ADEAT40M-SYSTX _
- / 0a 9
- 3/4.9.2dk WATER LEVEL - REACTOR VESSEL..............a.............. 3/4 9-n
/c .
- )t
. 3/4.9.22 WATER LEVEL - STORAGE POOL .............................. 3/4 9-12 l .s .
3/4.9.t* FUEL STORAGE P0OL AIR CLEANUP SYSTEM. . . . . . . . . . . . . . . . . . . . . 3/4 9-13 l .
.' 3/4.'10 SPECIAL TEST EXCEPTIONS
-3/4.10.1 SHUTDOWN MARGIN...............................'........... 3/4 10-1 3/4.10.2 GROUP HEIGHT, INSERTION, AND POWER DISTRIBUTION LIMITS... 3/4.10-2 O s an.10.2 PuYSICS TESTS............................................ 3/4 10-3 3/4.1'J. 4 REACTO R COO LANT L00 P S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 10-4 3/4.10.5 POSITION IN0! CATION SYSTEM - SHUTD0WN.................'... 3/4 10-5't j 3/4. u RADICACTIVF EFFLUENTS I -
] 3/4.11.1 LIQUID EFFLUENTS
- ! Concentration............................................ 3/4 n-1 TABLE 4. u-1 RADI0 ACTIVE LIQUID WASTE SAMPLING AND ANALYSIS j PR0 GRAM.................................................. 3/4 11-2 Dose..................................................... 3/4 n-5 Liquid Radwasta Treatment Systaa......................... 3/4 n-6
,- Liquid Holdup Tanks...................................... 3/4 11-7 w.STS XIV NUREG 0452/STPEGS
~
COMPARIS0N
.;~-'-..~ . . .
. . . ;1, .
.7....~... .; - - T * ~~- * '~ :~ . ..
0 ( *. INDEX , LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.u.2 GASE005 EFFLUENTS . Oose Rate................................................ 3/4 u-8, TA8LE 4. u-2 RADICACTIVE GASE0US WASTE. SAMPLING AND ANALYSIS PR0 GRAM.................................................. 3/4 u-9 Ocse - Noble Gases....................................... 3/4 u-u Dose - Iodine-131, Iodine-133. Tritium, and Radioactive Material in Particul ate Fo rs. . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 u-13 Gaseous Radwasta Treatment Systaa........................ 3/4 u-14 Explosive Gas Mixture...*................................. 3/4 u-15 Gas storage Tanks........................................ 3/4 u-16 3/4.u.3 SO LIO RADI0 ACTIVE WASTES. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3'/4 u-17
-. 3/4.u.4 TOTA L 00 5 E. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 u-18 ,
(--. 3/4.12 RA0!0LOGICAt. ENVIRONMENTAt. MONITORING 3/4. U.1 MONITORING PR0 GRAM....................................... 3/4 u-1 TA8LE 3.12-1 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM........ 3/4 u-3 TA8LE 3.12-2 REPORTING LEVELS FOR RADI0 ACTIVITY CONCENTRATIONS IN ENVIRONMENTAL 5 AMP LES. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 12-9 TABLE 4.12-1 DETECTION CAPA81LITIES FOR ENVIRONMENTAL SAMPLE ANALYS!$,............................................ 3/4 u-10 3/4.12.2 LAND USE CEN5US.......................................... 3/4 U-13 l 3/4.12.3 INTERLABORATORY COMPARISON PR0 GRAM....................... 3/4 12-15 l I l
' O l
i l NUREG 0452/STPEGS I W-STS XV
- COMPAR.ISON l
L
l
. : . . -. ~... ."
~7 ~~ ; ~ ~ ~ -- ~ .. :. . -
- -... .. . .,... ~ . . - . . .
l
~ '
.l!EE . .
8ASES SECTION . PAGE
. 3/4.0 APPLICA8!LITY............................................... 8 3/4 0-1
. . 3/4.1 REACTIVITY CONTROL SYSTEMS -
.. 3/4.1.1 80 RATION CONTR0L..........................................
8 3/4 1-1
. 3/4.1.2 80 RATION SYST9t$.......................................... 8 3/4 1-2 3/4.1.3 MOVASLE CONTROL A33FMBLIES. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 1-3
. . l 3/4.2 POWER O!$TRIBUTION LINIT5. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . , 8 3/4 2-1 3/4.2.1 AXIAL FLUX 0!FFERENCE...............................~...... 8 3/4 2-1 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR and RCS FLOW RATE ANO NUCLEAR ENTHAL7Y RISE NOT CHANNEL FACTOR....... 8 3/4 2-2
._ FIGURE 5 3/4.2-1 TYPICAL IN0!CATED AXIAL FLUX O!FFERENCE VER$US THERMAL P0WER............................................. 8 3/4 2 3/4.2.4 QUADRANT POWER TILT RATI0..............................,.. 8 3/4 2-5 3/4.2.5 DNS PARAMTER5............................................ B 3/4 2-5 3/4.3 INSTRUMENTATION 4
. 3/4.3.1 and 3/4.3.2 REACTOR TRIP SYSTEM and ENGINEERED SAFETY FEATURES ACTUATION SV5 TEN INSTRUMENTATION............... 8 3/4 3-1 i
l . 3/4.3.3 MONITORING INSTRUMENTATION................................ 8 3/4 3-3 1 4
~
N..7 . .--- 5MR-wT EM3f"EttP-fNU i 5 w i . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5/*a i 'l 9 i NUREG 0452/STPECS PSTS XVI COMPARISON _wm_
_ _ _w__ a - . _aa- a _2 0 i l .I l i i
- O i
i I .l J l 5 l i I l. l I i i I O 1 r
r s . INDEX BASES SECTION . pAgg 3/4.4 REACTOR COOLANT SYSTEM . 3/4.4.1 REACTOR COOLANT LOOP 5 AND COOLANT CIRCULATION............. 8 3/4 4-1, 3/4.4.2 SAFETY VA LVES. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . * . . . . . . . . . . - 8 3/4 4-2 3/4.4.3 PRE 55URIZER............................................... B 3/4 4-2 3/4.4.4 RE LI EF VA LVE5. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . I 3/4 4 3 . 3/4.4.5 STEAM GBERAT0R$.......................................... I 3/4 4 3 3/4.4.8 REACTOR COOLANT SYSTEM LEAXAGE...........,................. 8 3/4 4-4 3/4.4.7 CHEMISTRY................................................. 8 3/4 4-5 3/4.4.8 SPECIFIC ACTIVITY......................................... 8 3/4 4-5 . 3/4.4.9 PRES $URE/ TEMPERATURE LIMIT 5............................... 8 3/4 4-7 s .- TABLE 8 3/4.4-1 REACTOR VESSEL 700GHNESS.......................... 8 3/4 4-9 . i - FIGURE 8 3/4.4-1 FAST NEUTRON FLUDCE (D1MeV) AS A FUNCTION OF FULL POWER SERVICE LIFE.................................. 8 3/4 4-10 FIGURE 8 3/4.4-2 EFFECT OF FLUENCE AND COPPER CONTENT ON SHIFT OF RT FOR REACTOR VESSELS EXPOSED TO 550*F............ 8 3/4 4-11 MT 3/4.4.10 STRUCTURAL INTEGRITY..................................... I 3/4 4-16 3/4.4.11 REA'CTOR COOLANT SYSTEM VENT 3............................. 8 3/4 4-16 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATOR $....................... ...................... 8 3/4 5-1 3/4.5.2 and 3/4.5.3 ECCS SUS $Y5TEMS............................... 8 3/4 5-1 3/4.5.4 BORON INJECTION SYSTEM.................................... 8 3/4 5-2 3/4.5.fREFUELINGWATERSTORAGETANK.............................. I 3/4 5-2 w.373 xygg NUREC 0452/STPEGS COMPARISON ,
.. . . . _ . . . . 2:. 2.
~
G 1N.gj,X, 8ASES SECTION - g M-ATMOSPHERICTYPECONTAINMENT . 3/4'6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMAAY CONTAINMENT....................................... 3 3/4 6-1A 3/4.6.2 OEPRES$URIZATION AND COOLING SYSTEMS...................... 3 3/4 6-3A
/- +. : 1001.-E CLE- = 5Y::: ..................................... 5 3/4 5-5.---
3/4.6.4 CONTAINMENT !$0LATION VALVES.............................. B 3/4 6-5A , 3/4.6.5 COMUSTIB LE GAS CONTR0 L. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 3/4 6-5A 3/i.5.5 FE-di.--TIOii ~= 5iEU5T hie C Li:-S? 5Y:: 6. . . . . . . . . . . . . . . 5 3/- 5-E - l
- . . --1---u.--.u-2............... .. . . . ._ . ___ 5 3l- +-+-
f,.
- i
- 1 WSTS
" XVIII-A NUREG 0452/STPEGS ,
COMPARISON
I , E { i
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EllE . SAsts l N erion 140T APPUCA61' gg i \ j M - ICE CONOENSEA TYPE CONTAIMENT . i 1 J 4 3/4.6 CONTA!WENT , j 3/4.6.1 PRIMARY CONTAI . .................................... 8 3/ , 3/4.6.2 OEPRES$URIZATION AND COOL $YSTEMS................. .. 8 3/4 6-48 3/4.6.3 100!NE CLEANUP SYSTEM............. ...... ............. B 3/4 6-58 i 3/4.4.4. CONTAINNENT ISOLATION VALVES.... ...... ............... 8 3/4 6-58 3i4.6.8 c0N8usr18a as ........................ ........ 8 v4 6-68
. 1 3/4.4.4 PENETRATI EXHAUST AIA FI'LTRATION SYSTEM......... . l'3/4 6-68 P
3/4.6.7 EN512............................................. 4 6*-68 . i ( ,.' .6.8 VACUUM RELIEF VALVES.................................,.... 8 3/4 4-l
.', I I
- I
.. )
l I l l I l i l l l NUREG04Si/5TPEGS W-ST3
- XVIII-8 c0MPARISON
l
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BASE 5 .
- 5'ECT10N . N DT A~991,\C.h3LE. -
g y,-SUSATHOSPHERIC'TYPECONTAINMENT 4.6 CONTAINMENT SYSTEMS - l m .. b n C0nfu. ext. . ......................................
' 8m
. m ... o vRE55UM M T 0 C00 MNG SYSTEM 5.................. . M 6-3C l 3/4.6.3 CONTAINMENT !$0LATION VAL ~
. ...... ............... B 3/4 6-4C l
3/4.6.4 CONSU5TIBLE GAS CONTROL.
............. 8 3/4 6-4C l
l 3/4. 6. 5 5USATM05P I RE55URE CONTROL 5YSTEM. . . . . . . . . . . . . . . . . 8 3/4 6-4C l l 3/ ACUUM AELIEF VALVES.. ................................... 83 C (,, ., ( 1 W-STS XVI!! C NUREG 0452/STPECS COMPARIS0N . e e e ogs e a MW ,9 .9@ # ., e '88 ** .
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. - . - ~ -
. . . . . . . . . - : ~~.~'.-. . :.~. .*; .
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E .. BASES ucTroN = f,gg
. ' 3/4.7 PLANT SYSTEMS .
3/4.7.1 TUR8!NE CYCLI. ........................................... 8 3/4 7-1 6 3/4.7.2 STEAM GENERATOR PRES $URE/ TEMPERATURE LIMITATION........... 8 3/4 7-3 3/4.7.3 COMPONENT COO LING WATER 5YSTEM. . . . . . . . . . . . . . . .. . . . . . . . . . . . 8 3/4 7-3 I #
, 3/4.7.4 B ii E WATER SYSTEM...................................... 8 3/4 7-3 J3 3/4. 7. s ULTIMATE NEAT 5 !Nx. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 3/4 7-3
^!= J. 5 =V-:+ % i - i 6 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
5 ?i= ?-- ; vsNT wATio4 3/4.7 CONTROL 200M EMERGENC SYSTEM................. 8 3/4 7-4 War.pmouse son.t 4 (ya si creusy ;- - - i 3/4.7. ------------"---73-" 8 3/4 7-4 l 3/4.7.fSNU88EAS.................................................. 8 3/4 7-5 (' ' u.7.45E4tEOSOURCCONTAa!NATI0N............................... 83/478 lC . 3/4.7 33 FIRE SUPPRE55!ON SYSTIM5.................................. 8 3/4 7-7 3/4. 7.b FI AE RATED A55tM LIES. . . . . . . . . . . . . . . . . . . . . . . .8.3/4 . . .7-7'
; 3/ . i . = - r4-awaivmma-maa =im . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 fi '- -
4 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1, 3/4.8.2, and I/4.s.3 A.C. SOURCES, D.C. 500RCts, and ONSITE POWEA DISTRIBUTION............................... 8 3/4 8-1 3/4.8.4 ELICTRICAL EQUIPMENT PROTECTIVE OtVIC!5................... 8 3/4 8-3 (. W-5TS
- XIX NUREC 0452/STPEGS COMPARISON 6
r ; l 1 0 ly LN9.5X. l ~ BASES \ l i SECTioN }{OT AFPt.tc. AB LC Pact I M - DUAL TYPE CONTAINMENT . ! 4.6 CONTAINMENT SYSTEMS l m k NAa e0 m i ENr...............:...................... 8 3i / l 3/4.5.2 OEPRE55 I TION /NO COOLING SYSTEMS................. . ' . B 3/4 6-30 3/4.6.3 ICDINE REMOVAL ..................... ............ 8 3/4 6-40 , 3/4.6.4 CONTAIMENT IS0tATION VAL ... ....................... 8 3/4 6-40 l m .6.s ConsuSTInt aA5 Co m o ............ ................... . m 6 30 3/4.6.6 PENETRATION EXHAUST AIR CLEANUP SYSTEN... . . . . . . . . . ' 8 3/4 6-50 l
- v. m.6.7 v nu u vA tvE5. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . i m 6-50 .
(
.8 SECONCAAY CONTAINMENT..................................... 8 60 l
l .
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( NUREG 0452/STPECS e COMPARISON W STS XVI!!-0 ,
~- =
. . ~~~.. . . . . . . . . . . - . - .7.': h.'. . "~.;. ......:.. ...: ' :~~.*.~.-*~
sAsts M .P.Ed 3/4.9 REFUELING CPERATIONS 3/4.9.1 00RON CONCDmtAT10N. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 9-1 3/4.9.2 INSTRUMENTATION........................................... B 3/4 9-1 3/4.9.3 OECAY T1M. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 3/4 9-1 3/4.9.4 CONTAll9 TENT SUILDING PENETRATIONS......................... B 3/4 F 1 3/4.9.5 C0f01UN I CAT!0NS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 83/49-1
'R E Prd 4 L M M m A w'W E. . .
3/4.9.43..,-...----;u_......................................... I 3/4 9-2 prvEt AAs_tx'N4 Evan'iR4 (FNg) 3/4.9.7 CRANE TRAVEL - .- ; .-- r. = =: = = =................ B 3/4 F 2
^
,3/4.9.8 RESIOUAL NEAT REMOVAL AND COO LANT CIRCULATION. . . . . . . . . . . . . B 3/4 9-2 ,
3/4.9.9 ....... I 3/4 F 2 3/4.9. and'3/4.9 TER LEVEL - REACTOR VESSEL and 8 3/4 F3 , i STORAGE P00L............................................
- evest;ts-sWNtRet m ................'.......... 8 3/4 F3 3/4.f.t Futt. MaMou% wasom e, t, ras) sesmust 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN MARGIN........................................... 83/410-1 3/4.10.2 GROUP HEIGHT,, INSERTION, ANO POWER O!5TRIBUTION LIMITS.... 8 3/4 10-1 3/4.10.3 PHYSICS T1375............................................. t 3/4 10-1 3/4.10.4 REACTOR COOLANT L00PS..................................... B 3/4 10-1 3/4.10.5 POSITION IM0! CATION $YSTEM - SHUTD0WN..................... B 3/4 10-1 NUREG 0452/STPECS COMPARISON XX PSTS , ,
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S 15E . BASES . 3/4.15 RADICACTIVE EFFLUENTS 3/4.u.1 LIQUID EFFLUENT3........................................ B 3/4 u-1 3/4. n.1 GASE0d5 EFF LUENT3. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B . . 3/4
. . . u-3 3/4.u.3 SO LIO RADICACTIVE WASTE 5. . . . . . . . . . . . . .'. . . . . . . . . . . . . . . . . .
B 3/4 u-6 3/4.u.4 TOTA L 00 $ E . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 3/4 u-6 3/4.M RA0 tot.0GICAL ENVIRONMENTM. MONITORING , 3/4.12.1 MONITORING PR0 GRAM. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 3/4 12-1 3/4.12.2 LANO USE CEN5US......................................... 8 3/4 12-1 [. 3/4.12.3 INTERLA80AATORY COMPAA!5ON PR0GAAM..........'............ B 3/4 12-2 s e
' O XXI NUREG 0452/STPEGS PSTS COMPARISON
ut - . .y .: . . - . % -S .. . .=:: . . .~;' .'. -- T . . . ......:...... ' ' ' " t,. . DRAFT jf9.E .. DESIGN FIATURf5 ggg '. gg-L.LJ.III .. 3.1.1 IXCLus!ON ARIA.............................................. 5-1 5.L 2 LOW POPULATION 20NE......................................... 5-1* 5.1.3 MAPS DEFINING UNRESTRICTED ARIAS ANO $1TE SOUNDARY FOR RA010 ACTIVE GA8E0US AND LIqu!O EFFLUENTS. . . . . . . . . . . . . . . . . . . . 5-1 FIGURE 5.1-1 EXCLUSION ARIA....................................... 5-2 FIGURE 5.1-2 LOW POPULATION 20NE.................................. 5-3
- FIGURE 5.1-3 UNRISTRICTED ARIA ANO $!TI SOUNCARY FOR RA010 ACTIVE GAst005 EFFLUENT 3........................ 5-4 FIGURE 5.1-4 UNRESTRICTED AREA ANO $!TI SOUNDARY POR RA010 ACTIVE LIqu!O EFFLUENT 5......................... 5-5 5.2 CONTAlffWff ,
5.2.1 CONFIGURATION............................................... 5-1 5.2.2 OESIGN PRESSURE AND TEMPERATURE............................. 5-1 g\ . ( 5.3 REACTOR CORE 5.3.1 PUEL AS$0SLIES............................................. 5-6 5.3.2 CONTROL 200 AE80 SLIIS...................................... 5-6 5.4 REACTOR COOLANT SYSTIN 5.4.1 DESIGN PRESSURE AND TEMPERATURE............................. 5-6 5.4.2 V0usE...................................................... 5-6 5.5 METt0R0 LOGICAL TOWER LOCAT10N................................. 5-6 5.8 FUEL ITORAGE 5.6.1 CRITICALITY................................................. 5-7 5.8.2 0RAINAGE..................................................... 5-7 5.8.3 CAPACITY....................................................
/ 5 -7 5.7 COMPONENT CYCLIC OR TRAN5! TNT LIMIT........................... 5-7 TABLE 5. 7-1 COMPONENT CYCLIC OR TRANs!!NT LIMIT 3. . . . . . . . . . . . . . . . . . 5-4 1
NUREG 0452/STPECS I i COMPARISON I i
. . _ _ _ . . . .. _. ... ~. . . _ _ . , .,...
O DMU I!!EU. ..
~~
ADMINISTRATIVE CONTROLS 4 I EEQ!! , PAGE 6.1 RESPONS!8!LITY.............................................. 6-1
- 6. 2 0Rea Iu r!On................................................ 61
~
6.1 1 0FFSITE................................................ .. 5-1
- 6. 2. 2 UNIT STAFF..,.............................................. 4-1 FIGURE 6.2-1 0FFSITE ORGANIZATION...............................
- 6-3
, FIGURE 6.2-2 UNIT ORGANIZATION.................................. -
6-4 TA8LI 6. F 1 NINIMuM SHIFT CAEW COMPOSITION..........,............ 4-5
- 6. 2. 3 INDEPENDENT SAFETY ENGINifAING GROUP Functfon.........................................'......... 6-6 Cogo s f t f on . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-6 . .
Responsfbf1ftfes.......................................... E-6 Records................................................... 4-6 ( 6. 2. 4 5HIFT TECHNICAL ADV!50R. . . . . . . . . . . . . . . . . . . . . . , . . . . . . . . . . . . 4-6 6,3 UNIT STAFF QUALIFICATIONS..................................l 6-6 .. - 8 6.4 TRA!NING.................................................... 647 AU0!T............................................ 6.5 REVIEV AND 6-7 6.5.1 UNIT 4tEVIDFGROUlc NMT ON" "C V'" C##""*C" Functfon.................................................. 6-7 Comp o s i t i o n. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-7 Alternates................................................ 6-7 Meeting Frequency......................................... 6-8 quorum.................................................... 6-8 Responsibf1ities.......................................... 5-6 Records................................................... ,6-9 ( W.5TS
- XXI!!
0 NUREG 04.12/STPEGS . COMPARISON
. --- .. . . . . . . _ w.
....4 . . . . . .. . . . . . . . . . . . . . . . . . . . . . .. .... . . .._. . . . . .
n
\
INgg .. ADMIN!$7R TIVE CONTROLS Nuct.sgs secry Rsview bom (Msd.15) 6.5.2 GQNF - Function.................................................. 5-9 Ceaposition.....................................'......... . 5-10 Alternates................................................ 5-i0 Censu1tants............................................... 5-10 Meeting Frequency......................................... 5-10 Quorm.................................................... 5-11 Reviow.................................................... 5-11 , Aud1ts...............................................'..... 5-11 Reco rds'. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .5-12 ....... t 4 6.6 9 fPORTA8 LE EVENT ACTI0N. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-13 6.7 SAFETY LIMIT VIOLATION...................................... 5-13 ( , 6. 8 PROCEDURES AND PR0 GRAMS..................................... 6-13 6.9 REPORTING REQUIREMENTS .,,, 6.9.1 ROvTINE Rf70RT3........................................... 6-16 5tWrtupReport............................................ 5-16 Annual Reports............................................ 4-16 Annual Radiological Environmental Operating Report........ ' 6-17 Sealannual Radfoactive Effluent Release Report............ 5-18 Monthly Opereting Report.................................. 6-20 Radial Peaking Factor Limi t Report. . . . . . . . . . . . . . . . . . . . . . . . 6-20
- 6. 9. 2 S P EC I A L REPO RT3. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-21 6.10 R ECO RD R ETENTION. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-21 A
ySTS XXIV NUREG 0452/STPECS C0!iPARISON , __,.,__-_-__---.-,.-----------------+-m - - - = = ' - - = - - ' ' ' ' ' " ' ' - - - - ' - ' ' ' - - ' - -
g
..:..........--.....=. --- -
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I ADMINISTRATIVE CONTROLS SECTION . 6.11 RADIATION PROTECTION PR0 GRAM............................... 6-22 6.12 MIGH RA01ATION AREA........................................ '6-6.13 PROCESS CONTROL PROGRAM (PCP).......... ................... N2 6.14 0FFSITE 005 E CALCULATION Malt AL (00CM) . . . . . . . . . . . . . . . . . . . . . 6-24 i . 6.15 MAJOR CHANGES TO LIQUID. GASEOUS. AND SOLIO RA0 WASTE TREATMENT 5YSTEMS................................. 6-25
.O
.e e b NUREG 0452/STPEGS PSTs uv co.VPARISON
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SECTION 1.0 DEFINITIONS m
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\ s. 1 NUREG 0452/STPEGS COMPARISON l l _
Prge No. 1 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE # NOTES co 1.0 PAGE: 1.0 1- 1.0 OPERATING MODE: J.23456 DEFINITIONS DESIGN 1) 1.3 ANAICG CHANNEL OPERATIONAL TEST definition will be modified IATER to include generic definition of channel operational test and digital operational test. 6
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O . J 1.0 DEFINITIONS ~
. i 9 .
The defined terms of this section appear in capitalized type and are applicable throughout these Technical Specifications. i . ACTION ,
!3- .
1.1 ACTION shall be that part of a Technical Specification which prescribes l ,
.i remedial measures required under designated conditions.
l
?.
- ACTUATION LOGIC TEST I 1.2 An ACTUATION LOGIC TEST shall be the application of various simulated
! input combinations in conjunction with each possible interlock logic state and
' verification of the required logic output. The ACTUATION LOGIC TEST shall 1 include a continuity check, as a minimus, of output devices.
.a. ,
4 ANALOG CHANNEL OPERATIONAL TEST I 1.3 An ANALOG CHANNEL OPERATIONAL TEST shall be the infection of a simulated
. signal into the channel as close to the sensor as practicable to verify O CPERASILITY of alam, interlock and/or trip functions. The ANALOG CHANNEL
.0PERATIONAL TEST shall include adjustments, as necessary, of the alarm, inter-lock and/or Trip 5etpoints such that the Setpoints are within the required l l
. range and accuracy.
- 1 AXIAL FLUX OIFFERENCE ,s t
1.4 AXIAL FLUX DIFFERENCE shall be the difference in normalized flux signals between the top and bottom halves of a two section excore neutron detector. CHANNEL CALIBRATION :
.1. 5 A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel such that it responds within the required range and accuracy to known l l values of input. The CHANNEL CALIBRATION shall encompass the entire channel )
including the sensors and alars, interlock and/or trip functions and say be performed by any series of sequential, overlapping, or total channel steps i such that the entire channel is calibrated. i
', CHANNEL CHECK \
l 1.6 A CHANNEL CHECK shall be the qualitative assessment of channel behavior i during operation by observation. This determination shall include, where
.*- possible, comparison of the channel indication and/or status with other Indications and/or status derived from independent instrument channels measuring th,e same parameter.
O
- l NUREG 0452/STPEGS COMPARISON W-STS 1-1
P;ge No. 2 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE # NOTES oo 1.0 PAGE: 1.0 1- 2.0 DEFINITIONS OPERATING MODE: A23456 DATA 1) 1.7, 1.10 STSEGS specific data provided (brackets deleted). DESIGN 2) 1.8 CONTROLLED LEAKAGE Deleted. The CVCS is not associated with SI for the STPEGS design. ED 3) 1.11 E AVERAGE DISINTEGRATION ENERGY ' sample' changed to ' reactor coolant at the time of sampling' for clarity. ED 4) 1.11 E AVERAGE DISINTEGRATION ENERGY 'the radionuclides in the sample' changed to ' isotopes,
.- other than iodines, with half lives greater than 15 minutes, making up at least 95% of the total noniodine activity in the coolant' for clarity.
ED 5) 1.9 - 1.11 Renumbered 1.8 - 1.10 due to deletion of 1.8. - ED 6) 1.10 DOSE EQUIVALENT I-131. l STPEGS options selected.
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. . . .T . - . . ~ . . . . . :~.2 . . : '. ~ T ~~
O ggg(3 OEFINITIONS . CONTAINMENT INTEGRITY 1.7 CONTAINMENT INTEGRITY shall exist when: ,
- a. All penetrations required to be closed during accident conditions are either; .
- 1) . Capable of being closed by an OPERA 8LE containment automatic isolation valve system, or .
- 2) Closed by manual valves, blind flanges, or deactivated automatic valves secured in their closed positions, except as provided in ,
. Table /3.6-1(of Specification /3.6.3F" ,
- b. All equipment. hatches'are closed and sealed, 4
- c. Each air lock is in compliance with the requirements of Specification r(3. 6.1.3Y
- d. The containment leakage rates are within the limitsof Specification !
v{3.6.1.2Yand
- e. The sealing mechanisa associated with each penetration (e.g., welds, l bellows, or 0-rings) is OPERA 8LE. ;
-tehRE0ttED-tEMME-i! -
e t:5!t t- tt:=i -5! =t=: != cM-j i- S+ -+=^-
!!"""U "_ _ _ LE".""" . , ,
_ _ _ . _ ___ u___ CORE ALTERATION i 1.9. CORE ALTERATION shall be the movement or manipulation of any component within the reactor pressure vessel with tne vessel head removed and fuel in the vessel. Suspension of CORE ALTERATION shall not preclude completion of l sovement of a component to a safe conservative position. ! 00$E EQUIVALENT I-131 l 1.1.0 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcurie / gram) wh.-h alone would produce the same thyroid dose as the quantity and isotopic ) h mixture of I-131. I-132. I-133, I-134, and I-135 actually present. The thyroid i dose conversion factors used for this calculation shall be those listed in v(Table III of TID-14844, " Calculation of Distance Factors for Power and Test Reactor Sites" .. !#.,1w " 7 -d FE;4.g kwr74u;4.-ir1GM.-ia 1,
% *tubeM9PPh I - AVERAGE DISINTEGRATION ENERGY M I 1.11 I shall be the average (weighted in proportion to the concentration of '
d each radionuclide in theAsample) of the sum of the average beta and gassa energies per disintegration (MeV/d) for the radionuc44 des =4st: thew 3=- 15cTOPf5. ETNER TMN 100ntis, sysTN HALF LIVES eRIATEM ~1 HAM 15 MmlTGS, MKEN6 ut AT LEASr 90= NUREG O M /STPEGS or 7mr 7prar. usu-iopwr AcTuirY W TW cos M T.
- 1 M-575 1-2 -
l
P;ge No. 3 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE I NOTES ca 1.0 PAGE: 1.0 1- 3.0 DEFINITIONS OPERATING MODE: J.23456 NONE 1) No changes. t O t 0
.9 I
l. l l l l 9 l
4 {', , DEFINITIONS ENGINEERED SAFETY FEATURES RESPONSE TIME 1.12 The'ENGINEtRED SAFETY FEATURES (ESP) RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its E5F Actuation Setpoint l
- at the channel sensor until the E5F equipment is capable of perfoming its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include .
' diesel generator starting and seqvence loading delays where applicable. j
' FREQUENCY NOTATION ..
1.13 The FREQUENCY NOTATION sp'ecified for the performance of Surveillance l Requirements'shall correspond to the intervals defined in Table 1.1. l j IDENTIFIED LEAKAGE 1.14 IDENTIFIED LEAKAGE shall be:
- a. Leakage (except CONTROLLED LEAKAGE) into closed systems, such as pump ,
seal or valve packing leaks that are captured and conducted to a sump or collecting tank, or Leakaga into.the containment atmosphere from sources,that are both ON b. specifically located and known either not to interfere with the l operation of Leakage Detection Systems or not to be FRE55URE BOUNDARY l
. LEAKAGE, or
- c. Reactor Coolant System leakage through a' steam generator to the ~
Secondary Coolant Systas. .... l MASTER RELAY TEST
. ~
1.15 A MASTER RELAY TEST shall be the Snergization of each master relay and
. verification of OPERABILITY of each rein. The MASTER RELAY TEST shall include a continuity check of each associated slave relay.
i _ MEMBER (5) 0F THE PUBLIC 1.16 MEMBER (5) 0F THE PUBLIC shall include all persons who are not occupa-tionally associated with the plant. This category does not include employees of the licensee, its contractors, or vendors. Also excluded from this category are persons who enter the site to service equipment or to make deliveries. This category does include persons who use portions of the site for recre-ational, occupational, or other purposes not associated with the plant. OFF5ITE 005E CALCULATION MANUAL 1.17 The OFFSITE DOSE CALCULATION MANUAL (00CM) shall contain the methodology and parameters used in the calculation of offsite doses due to radioactive 4 gaseous and itquid affluents, in the calculation of gaseous and liquid effluent monitoring Alam/ Trip 5etpoints, and in the conduct of the Environ-mental Radiological Monitoring Program. NUREG 0452/STPEGS c0 SON ySTS 1-3
Paga No. 4 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE f NOTES oo 1.0 PAGE 1.0 1- 4.0 DEFINITIONS OPERATING MODE: 123456 OTHER 1) 1.22 PROCESS CONTROL PROGRAM (PCP) definition - revised. Consistant with STSRev4 and recently approved Byron Tech Spec. O
)
e O O
4 O DEFINITIONS OPERA 8LE - OPERA 8ILITY , 1.18 A systas, subsystem, train, component or device shall be OPERA 8LE or have OPERA 8ILITY when it is capable of perfoming its specified function (s),
'and when all necessary attendant instrumentation, controls, electrical power, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component, or _ device to perform its .
function (s) are also capable of performing their related support function (s). t OPERATIONAL MODE - MODE ,
,1.19. An'0PERATIONAL MODE (i.e., MODE) sha11* correspond to any one inclusive combinat. ion of core reactivity condition, power level, and average reactor 1 coolant temperature specified in Table 1.2. l PHYSICS TESTS 1.20 PHYSICS TESTS shall be those tests performed to measure the fundau ntal I
. nuclear characteristics of the reactor core and related instrumentation:
(1) . described 'in ChapterM14.07of the FSAR, (2) authorized under the provisions of 10 CFR 50.59, or (3) othenvise approved by the Commission. PRES 5URE 80UNOARY LEAKAGE 1.21 PRES $URE SOUNDARY LEAKAGE shall be leakage (except steam generator tube leakage) through a nonisolable fault in a Reactor Coolant Systas component %O body, pipe wall, or vessel wall. PROCESS CONTROL PROGRAM - PROV!510NS To ASWRf TW 1 The PROCESS ' CONTROL PROGRAM (PCP) shall contain theA;-. . x. "-E-T,.22w ,116g, e6alyseartestsrand-determinatiene te-be-made=to ensure-that= pecessing.and __n__ a __- packaging-of-solid-cadioactive-was* _s __ >_ e> - e j--d e= d-y-. . _ _ _ _ _-_P_-d- .. 'p
- n _4 -___:__ _.
7.---- ..., -. - - . -. -- ;-- - - - - -
---y - - - - -
--- 5 -f :-- :: =- y 3 ,..7,;--- -__-,--y_--___- m -- __ 7___
-__.--..==.:-7.,------ .: ;- - --.- - - . - - - - - - - - -
PURGE - PURGING 1.23 PURGE or PURGING shall be any controlled process of discharging air or gas ) l from a confinement to maintain temperature, pressure, humidity, concentration I or other operating condition, in such a manner that replacement air or gas is required to parify the confinement. j l SOLIDIFICATION of wet radioactive wastes results in a vaste forn with properties that meet the requirements of 10 CFR Part 61 and of low-level l
. radioactive waste disposal sites. The PCP shall identify process parameters
! influencing SOLIDIFICATION such as pH, oil content, 5:0 content, solids content, ratio of solidification agent to waste and/or necessary additive,s for each type of anticipated waste, and the acceptable boundary conditions for the process parameters shall be identified for each vaste type, based on l 1aboratory scale and full scale testing or experience. The PCP shall also l include an identification of conditions that must be satisfied, based on full
- hv,
' I scale testing, to assure that devatoring of
- bead resins, powdered resins, and filter sludges will result in volumes of free water, at the time of disposal, within the limits of 10 CFR Part 61 and of low-level radioactiv's weste disposal sites. NUREG 0452/STPEGS 1
l w-STS 1-4 COMPARISON )
. l
P ge No. 5 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE # NOTES 00 1.0 PAGE: 1.0 1- 5.0 OPERATING MODE: ,1 2 3 4 5 6 DEFINITIONS DATA 1) 1.25 RATED THERMAL POWER. STPEGS specific data provided. DESIGN- 2) 1.28 SHIELD BUILDING INTEGRITY deleted. Not applicable to STPEGS design. DESIGN- 3) 1.28 SHUTDOWN MARGIN reference to ' full-length' rod clusters deleted since STPEGS does not have any partial length rod clusters. ED 4) 1.29 and 1.30 renumbered 1.28 and 1.29 due to deletion of 1.28.
.e l
l
. . f ." , , . - .: . . . . . .-- ... .
. "- i . . . .'.-' . . . _ _ _ _ .
DEFINITIONS l
. QUADRANT POWER TILT RATIO .
. 1.24 QUADRANT POWER TILT RATIO shall be the ratio of the maximum upper excore detector calibrated output to the average of the upper excore detector cali- ,
brated outputs, or the ratio of the maximum lower excore detector calibrated !
. output to the average of the lower excore detector calibrated outputs, whichever l
. .is greater. With one excore detector inoperable, the remaining three detectors. '
shall be used for computing the average. RATED THERMAL POWER 1.25 RATED THERMAL POWER shall be a total reactor core heat transfer rate to i the reactor coolant of MMWt.
. REACTOR TRIP SYSTEM RESPONSE TIME
~
. 1.26 The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval from
- when the monitored parameter exceeds its Trip 5etpoint at the channel sensor l until loss of stationary gripper coil voltage.
REPORTABLE EVENT 1.27 A REPORTA'8LE EVENT shall be any of those conditions specified in O- Section 50.73 of 10 CFR Part 50.
- 2. :- 3 , " _ , ",
- ^ :;q : ;g : " xj "
- ;Q:d n 71 -%,- y-- - :- -
m- -g :: = :xg ;;;; :;- = == : u =:_;xu; ==::. z-- : , _ _ -j= - : ^ ;!' L- _1 - '; 1
^
, 4 . ': E^":1' 3:!! !x; !!L.;t";n Ey; i= !; rE"L" ELE, _.._ l
. . . .._. ..,_....._.... . wow ;.n .;_x i
7........... ....., ......, , __...-., .. .--......j u w . . . -- --__ ; . , l SHUTDOWN MARGIN 1.2fSHUTDOWNMARGINshallbetheinstantaneousamountofreactivitybywhich the reactor is subcritical or would be suberitical from its present condition assuming all N rod cluster assemblies (shutdown and control) are fully inserted except for the single rod cluster assembly of highest reactivity worth which is assumed to be fully withdrawn. I SITE BOUNDARY s.9 1.30 The SITE BOUNDARY shall be that line beyond which the land is neither I owned, nor leased, nor otherwise 1:entro11ed by the licensee. l NUREG 0452/STPEGS COMPARISON W _-STS 1-5
P2ge No. 6 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE i NOTES co 1.0 PAGE: 1.0 1- 6.0 OPERATING MODE: 223456 DEFINITIONS ED 1) 1.31 - 1.37 renumbered 1.30 - 1.37 due to deletion of 1.28. O
.t I
l l 6 9 O
o DEFINITIONS DMR SLAVE RELAY TEST 0
- . 1.3% A SLAVE RELAY TEST shall be the energization of each slave relay and verification of CPERA8ILITY of each. relay. The SLAVE RELAY TEST shill include a continuity check, as a minimum, of associated testable actuation devices.
SOLIDIFICATION I 1.3$ SOLIDIFICATION shall be the convdrsion of wet wastes into a form that meets shipping and burial ground requirements. ,, SOURCE CHECK .' 1.3 A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a source of increased radioactivity. STAGGERED TEST BASIS 1.3f A STAGGERED TEST 8 ASIS shall consist of:
- a. A test schedule .for n systems, subsystems, trains, or other
> designated components obtained by dividing the specified test interval into n equal subintervals, and l
- b. The testing of one system, subsystem, train, or other designated component at the beginning of each subinterval.
THERMAL POWER 4 "" 1.$ THERMAL POWER shall be the tctal reactor core heat transfer rate to the reactor coolant. TRIP ACTUATING CEVICE OPERATIONAL TEST 5 1.3% A TRIP ACTUATING DEVICE OPERATIONAL TEST shall consist of operating the Trip Actuating Device and . verifying OPERA 8ILITY of alarm, interlock and/or trip functions. The TRIP ACTUATING DEVICE OPERATIONAL TEST shall include l adjustment, as necessary, of the Trip Actuating Device such that it actuates 1 at the required Setpoint within the required accuracy. l j i UNIDENTIFIED LEAKAGE 6 1.37 UNIDENTIFIED LEAXAGE shall be all leakage which is not IDENTIFIED LEAKAGE l or CONTROLLED LEAXAGE. l
)
l l l O NUREG 0452/STPEGS PSTS 1-6 C0FARISON
\
- - - . .-----,-e-w. m-- . -. - -, - , . - . , - .~-.- - m- - - , - - - - - - - . - - . - -
Page No. 7 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE i NOTES 00 1.0 PAGE: 1.0 1- 7.0 OPERATING MODE: 123456 DEFINITIONS ED 1) 1.38 - 1.41 renumbered 1.37 - 1.40' due to deletion of 1.28.
~
O l l l t i l l l l l I O I
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.- . =...=..:.- =: . .- . . .- ==
i DEFINITIONS -
) .
$. UNRESTRICTED AREA .
.k 1.3g An UNRESTRICTED AREA shall be any area at or beyond the SITE BOUNDARY
. access. to which is not controlled by the licensee for purposes of protection of l
. individuals from exposure to radiation and radioactive materials, or any area ;
j ; within the SITE BOUNDARY used for residential quarters or for industrial, ' j ,, commercial, institutional, and/or recreational purposes. .
! VENTILATION EXHAUST TREATMENT SYSTEM i
1.3[ A VENTILATIU' N EXHAUST TREATMENT SYSTEM shall be any systes designed andI 1 3 installed to reduce gaseous radiciodine or radioactive material in particulate l . . form in effluents by passing ventilation.or vent exhaust gases through charcoal - j adsorbers and/or HEPA filters for the purpose of removing iodines or particu- ,
$<- 1ates from the gaseous exhaust stream prior to the release to the environment. ,
j f. Such a systes is not considered to have any effect on noble gas effluents. l .' Engineered Safety Features Atmospheric Cleanup Systems are not considered l l ? to be VENTILATION EXHAUST TREATMENT SYSTEM components. i j VENTING
! i 39 i-1.40 VENTING shall be the controlled process of discharging air or gas from a' '
. confinement to maintain temperature, pressure, humidity, concentration, or other ,
i operating condition, in such a manner that replacement ' air or gas is not pro- l
.'vided or required during VENTING. Vent, used in system names, does not imply a VENTING process. .
4 WASTE GAS HOLDUP SYSTEM
- , 0 . ..<<
; 1.4f A WASTE GAS HOLDUP SYSTEM _shall be any system designed and installed to c reduce radioactive gaseous effluents by collecting Reactor Coolant Systes
!. < offgases from the Reactor Coolant System and providing for delay or holdup i J for the purpose of reducing the tctal radioactivity prior to release to the , ,i environment. - i } s'I 1 1 . ?
- i. .
- .) -
1 i NUREG 0452/STPEGS 1-7 COMPARISON j PSTS ~ l I
Page No. 8 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE # NOTES oc 1.0 PAGE: 1.0 1- 8.0 OPERATING MODE: ).23456 DEFINITIONS NONE 1) Table 1.1 Frequency Notation. No Change s - e O. l l
.t O
l
T l , ;. , c . ; .T, . --- ; .---~~ :C . .
. . .. ....:...~....... ..
'~
O .- . DRW TABLE 1.1 FREQUENCY NOTATION NOTATION FREQUENCY 5 At least once per 12 hours. O At least once per 24 hours. W At least once per 7 days. M At least once per 31 days. Q At least once per 92 days. 5A At least once per 184 days. . R At least once per 18 months. 5/U - Prior to each reactor star 6ap. N.A. Not applicable.
\. P Completed prior to each release.
. \
O W-STS
~
1-8 NUREG 0452/STPEGS CopARISON
Page No. 9 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE i NOTES 00 1.0 PAGE: 1.0 1- 9.0 OPERATING MODE: 123456 DEFINITIONS DESIGN 1) Table 1.2 Operational Modes. Average Coolant Temperature for refueling changed to <=150 F to reflect STPEGS rapid refueling design. s
. O
.O e
l O
~
l i
. .=.----
. _ .. I
~
1 l TABLE 1.2 OPERATIONAL MODES l i
% RATED AVERAGE COOLANT REACTIVITY TEMPERATURE MODE CONDITION. K ff THERMAL POWER
- l
> 0.99 > 5% > 350*F l
- 1. POWER OPERATION
> 0.99 > 350*F
- 2. STARTUP 1 5%
- < 0.99 0 > 350*F
- 3. HOT STAND 8Y
~
< 0.99 0 350'F > T"VU
- 4. HOT SHUTDOWN > 200*F
~
5, COLD SHUTDOWN < 0.99 . O 1 200*F
- 6. REFUELING ** $ 0.95 0
i NF .
- o
- Excluding decay heat.
** Fuel in the reactor vessel with the vessel head clost.re bolts less than fully tensioned or with the head removed.
I 1
. l e ,
l l l i l l i 1 O 1-9 NUREG 0452/STPEGS PSTS COMPARISON l v
l I e' . DRAFT ! t ( l t l i SECTION 2.0 ; 5 t SAFETY LIMITS
- i LIMITING SAFETY SYSTEM SETTINGS
= 9s - i
*9
. I L
l
)
I l NUREG 0452/STPEGS COMPARISON
Page No. 10 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE # NOTES 00 2.0 PAGE: 2.0 2- 1.0 OPERATING MODE: 12---- SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS DESIGN- 1) 2.1.1 Reactor Core. Reference to Figure 2.1-2 and n-1 loop operation deletad,since STPEGS will not pursue n-1 operation. DESIGN- 2) 2.1.1 Reactor Core - Action. Reference to loop i ' operation' deleted since STPEGS will not pursue i n-1 operation. 4
. IID l
l l I l O I i
._ . _ _ = _ -_ - .- . . _ . - - _ _ _ . ._
DMR
~
O
- l 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 1
( l 2.1 SAFETY LIMITS REACTOR CORE ; 2.1.1 -The combination of THERMAL POWER, pressurizer pressure, and the highest ' - operatina loop coolant temperature (T ,,g) shall not " exceed the limits shown in *
^' -
^^^^ ^'
Figu g 2.1-1,
^ -
- APPLICABILITY
- M00E5 1 and 2.
ACTION: Whenever the point defined by the combination of the highest M 1oop
- average temperature and THERMAL POWER has exceeded the appropriate pressurizer pressure line, be in HOT STANOBY within 1 hour, and comply with the require-ments of Specification 6.7.1. -
- REACTOR COOLANT SYSTEM PRESSURE 2.1.2 The Reactor Coolant Systes pressure shall not exceed 2735 psig. 'O APPLICA8ILITY: MODES 1, 2, 3, 4, and 5.
(, ACTION: l MODES 1 and 2: W enever the Reactor Coofant Systes pressure has exceidad 2735 psig, be'.'. in HOT STAN05Y with the Reactor Coolant Systas pressure within its limit
- within I hour, and comply with the requirements of Specification 6.7.1.
1 MODES 3, 4 and 5: l W enever the Reactor Coolant Syster pressure has exceeded 2735 psig, reduce the Reactor Coolant Systes pressure to within its limit within 5 minutes, and cogly with the requirements of Specification 6.7.1. O s NUREG 0452/STPECS W-STS 2-1 COMPARISON l
Page No. 11 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE f NOTES co 2.0 PAGE: 2.0 2- 2.0 OPERATING MODE: J. 2 3 4 5 - SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS DATA 1) Figure 2.1-1 Reactor Core Safety Limit - Four Loops in Operation. Replaced with STPEGS figure. ( i . O l l
.t l
O
l sn FSM 12470-1 680 660 A
, ,, ,3,,
{0' 66 (0.67,643.7) (0, 652.7) 640 N (0.955, 628.8) ((i.20,598.6) g 620 '
$ (0, 6 25 . 6) N (0.985, 620.2) 1860 PSI A (1.05, 603.2 g5"
~N 1
(1.10,590.8)
\
0
/
(1.20,591.8)- (1.20, 579.8)- 560 540 0 20 40 60 80 100 12 0 POWER (PERCENT) O SOUTH TEXAS PROJECT O UNITS 1 & 2 i FSAR N LocWillbEGtiM52/STPEGS COMPARIS0N Figwe 2.11 2-2
Page No. 12 06/33/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE i NOTES co 2.0 PAGE: 2.0 2- 3.0 OPERATING MODE: 12345-SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS DESIGN- 1) Figure 2.1-2 Reactor Core Safety Limit - Three Loops in Operation. Deleted. Not applicable to STPEGS design. STPEGS will not pursue n-1 loop operation. O l .e O
,. .. m _- .. .. ..
n _._.1..._.
. - s, .. . lt ua b
=i
( UNACCEPTABLE-660
.-__ ,,2400 PS.IA1 g
,w ..
OPERATION ' _( ,~
. _ a._ ,
.6
_ . .. t
,, _ .... <a:
,6 _
-%_..Te.,-_ .
iA D _ h
'h
% - t._.
g - m x f_
- '2250 PSlA' N -
g .
. g
-m
.r.
1_ . _: 7 _'""- __ u ._ N.
. ,.m .,
-- ___ ... . n
_, _:w _. :
..g 7e . .
% s L, jC._- . ' __ . . _
x - _ 2000 PS tr_. ,_ ..._. . _. ._
._. h_ .;
4 _
- u. ...= ._
_* _1 . .:".
-+--
_ DHS MURE POR lu.utTRM10N ONW 4 -
% '~ *-
~
g 00 pHTT USE FOR OPER. ATION p. W... g 600 __ . ., r. ,, . . g_._ . . . . 1
. ... .g
_ _..1700 PSI . ... A.. . . . . w_ , p _ o. _.< ,._2
.< i.
_ %. :=. .
=
_., g . ..
._ A . _. ', '
y _ _ _. _ _ ;..._ _ _ -._. -= . % , y, . y 580 % ACCEPTABLE . .__ . _. OPERATION - -----
....i. _ ..
. ._ _ .w. . _.
_t
==
__ %_ _.1 m. _v_-
-t- . ._.i\_. _ .
5M - _.._,. _ _ .t_..
.t ..... __.. ._ .. ... . ... ...
.m,
. . _ ... . . . .- . t .. ._...- . . . .
0 0.2 0.4 0.6 0.8 FRACTION OF RATED THERMAL POWER FIGURE 2.1-2 REACTOR CORE SAFETY LIMIT - THREE LOOPS IN OPERATION NOT APPLICABLE s NUREG 0452/p,TfEG{ D W-5TS 2-3 COMPARISM
Page No. 13 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE f NOTES oc 2.0 PAGE: 2.0 2- 4.0 OPERATING MODE: 123456
~
SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS ED 1) 2.2.1 Reactor Trip System Instrumentation Satpoints. Action b.2 reference clarified. s e O 4 O [ L
. .. .. .. . .: . ..... .:-" v :' -
L SAFETYLIMITSANDLIM{pNGSAFETYSYSTEMSETTINGS .
)
2.2 LIMITING SAFEW SYSTEM SETTINGS ,
- l
' REACTOR TRIP SYSTEM INSTRUMT._NTATION SETPOINTS 2.2.1 The R'esctor Trip System Instrumentation and Interlock Setpoints shall be set, consistent with ths Trip Setpoint values shown in Ttbla 2.2-1.
APPLICABILITY: .As shown for each channel in Table 3.3-1. ACTION: ,
- a. With a Reactor Trip System Instrumentation or Interlock Setpoint less conservative than the value shown in the Trip ietpoint column but more c.onservative than th value shown in the Allowable Value i coluan of Table 2.2-1, adjust the Setpoint consistant with the Trip
) !
Setpoint value. .
- b. With the Reacter Trip System Instrumentation or Intarlock Setpohnt less conservative than the value shown in the Allowable Values -
column of . Table 2.2-1, either:
- 1. Adjust the Setpoint consistent with the Trip Setpoint value of
' Table 2.21 and determine within 12 hours that Equation 2.2-1 was satisfled for the affected channel, or
- 2. De:1tra tN channel inoperable and apply the k:able ACTION TA SLE 3.5-I statement rwquirement of W h h 3.2.1 ~f.he channel is restored to OPERACLE status with its Setpoint adfusted. 3 .',
censistent with the Trip Setpoint value. Equation 2.2-1 , Z + R + 5 ,< TA Whers: I = The value from Column I of Table 2.2-1 for the affected channel, R = The "as measured" valua (th percent span) of rack error for the affected channel, t S = Either the "as measured" value (in percent span) of the sensor e ror, or the value from Colum S (Senser Error) of Table 2.2-1 for the affected channel, and TA = The value from Column TA (Tetat A11ovance) of Table 2.2-1 for the affected channel. NUREG 0452/STPEGS COMPARISON
- r Page No. 14 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS.
NOTE TYPE NOTE i NOTES __________-- ------ -----------------------------------------~~~~~---- oc 2.0 PAGE: 2.0 2- 5.0 OPERATING MODE: 123456 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTING 3 DATA 1) Table 2.2-1 Reactor Trip System Instrumentation Trip Setpoints. Preliminary STPEGS values provided. O-
.t
\
9 e
.. -. - . _ ~ - _ _.-.- . .- .. - .-. - - -- . -- _. - -
O . o . o. i . TABLE 2.2-1 l i REACTOR TRIP SYSTEM INSTRUNENTATION TRIP SETPOINTS -1 4 'Q s'
SENSOR l ERROR TOTAL.
ALLOWANCE (TA) Z (S) TRIP SETPOINT ALLOWA8tE VALUE .l
! FUNCTIONAL UNIT N.A.
'I Manual Reactor Trip N.A. N.A. N.A. N.A.
1. Power Range, Neutron Flux ist. I
- 2. g109N of RTP** $ 111.2 % of RTP**
- a. High Setpoint J7.5Y 14.56T 0 27.3 14.561 0 SE25N of RTP** gtP&N of RTPa*
- b. Low Setpoint 18.31 4. 3 I .Ce Power Range, Neutron Flux. -ff-et 10.51 O g5N of RTP** with ${& 8 N of RTP** with
- 3. a time constant a time constant High Positive Rate g2} seconds lf2} seconds
- 10. 5}- a g5N of RTP** with $ N of RTPa* with
- 4. Power Range, Neutron Flux, . }-
a time constant High Negative Rate a time constant g27 seconds 'g2} seconds l et If3 of RTP**
- 5. Intermediate Range, 117.07 18.4h0 g25N of RTP**
Neutron Flux l Source Range, Neutron Flux J10.0hD ffl05}' cps gl.4 x 105} cps .
- 6. J17.0T (o.B 4.4 2.3 +04 l i
.E2.73)- {&St See Note 1 See Note 2 ,.
- 7. Overtemperature AT J6.4F
}- 40.2} See Note 3 See Note 4 i
- 8. Overpower AT l 3.1 l610 8850.$
.{&S}- .{0.71T E1.571(4906} psig gl^^.'Ppsig ;l
- 9. Pressurizer Pressure-tow 239f.7 g 2300 t
w Pressurlzer Pressure-lligh 43.1-1 .{0.71}' .fl.5} $[43867 psig $2396T6sig gy 10.
- 11. Pressurizer Water Level-High f5.0} .{2.18711.5} g92N of instrument g93.SN of instrument .
h . span span . g et. l 0.fo M .( .
@$ -{4:0}- {-Ir6}- M90N of loop 1[49.2K of loop
>; g 12. Reactor Coolant flow-Low J2. 5} - design flow
- design flow" Q 'j g * -
ii
- Loop des (On flow = 195,700]' gpm ,
""RTP = RATED THERMAL POWER t
Page No. 15 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE # NOTES 00 2.0 PAGE: 2.0 2- 6.0 OPERATING MODE: J23456 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS DATA 1) Table 2.2-1 Reactor Trip System Instrumentation Trip Satpoints. Preliminary STPEGS values provided. DESIGN- 2) Table 2.2-1 Reactor Trip System Instrumentation Trip Satpoints. Item 14 Steam /Feedwater Flow Mismatch and Note (**) deleted. Not applicable to STP design. ED 3) Table 2.2-1 Reactor Trip System Instrumentation Trip Satpoints. Item 15 - 18 renumbered 14 - 17 due to deletion of item 14. DESIGN 4) Table 2.2-1 Reactor Trip System Instrumentation Trip Satpoints. Item 16.a. Changed to ' Low Emergency Trip Fluid Pressure' to reflect STP design.
- 5) Table 2.2-1 Reactor Trip System Instrumentation Trip Satpoints. Item 16. Turbine Trip values to be provided IATER.
ED 6) Table 2.2-1 Reactor Trip System Instrumentation " Trip Satpoints. Item 17. 'ESF' changed to 'ESFAS' for clarity.
O 6, o, i I i 1 ty TABLE 2.2-1 (Continued) REACTOR TRIP SYSTEN INSTRUNENTATION TRIP SETPOINTS h SENSOR TOTAL; ERROR Z (S) TRIP SETPOINT ALLOWABLE VALUE FUNCTIONAL UNIT All0WANCE (TA)
- 13. Steam Generator Water .{3Gre} {27.10f.(1.5}' 1ld$3N of narrow 1 M ]E of narrow 15.0 f2.lo range instrument range instrument Level Low-Low span span
~
l gg [16.0) [13.24T [1.5] g40]X of full e am/feedwater f4ow steam flow at,RTP # g42.5]% of fullsteam flow at RTPa* Mismatch / ' p./ Colneident With Steam Gene or Water 2.0] [9.18] [1,5]' 1L32.3]% of narrow A g 0.4]% of narrow , range instru M range instrument Level ow-Low maan gr
~
/o4 0. 3 /o,Soo - 9P/6 34 1{4760}-volts
- 15. Undervoltage - Reactor -t2:41 Fir 29} 1{48367 volts y i y, Coolant Pumps SA o.ol O S7.2 .
es .JP -(Gr11- g5h63 Hz
- 36. Underfrequency - Reactor .(J:57 M57.11Hz Soolant Pumps i
go,en) LM tMER- UNMt- UWBE- M 4'L MrA: -N-A. 3(800]ptfg -t(750]'pftg l ! a. Lodiluid 0$t" Pressure #rA- ' LATER- Ll}1ER. LM;82. LA7tE. . L4kE.
- b. Turbine Stop Valve -MrA. -M A:- N.A. 2(J]% open 3(J]% opetr ,
Closure s'1
- N. A.
J8". Safety Injection Input M.A. M.A. N.A. N.A. g from ESFA9 **
- n 88 -
, i5 g .
e:, b_ N , O h f RTP = RATED-ittERMA . M 8 e
Page No. 16 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE # NOTES oc 2.0 PAGE: 2.0 2- 7.0 OPERATING MODE: J23456 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS ED 1) Table 2.2-1 Reactor Trip System Instrumentation Trip Satpoints. Item 19 - 21 renumbered 18 - 20 due to deletion of item 14. DATA 2) Table 2.2-1 Reactor Trip Systems Instrumentation Trip Satpoints. STPEGS specific preliminary values provided. s e O
.t O
e c 1 - TA8tE 2.2-1 (Continued) 19 REACTOR TRIP SYSTEN INSTillstENTATION TRIP $ETPOINTS h . 5tN50R . TOTAL ERROR Z _(5) TRIP SETPOINT ALLOWARLE VAluE ,\ FtRICTIONAL UNIT ALLOWANCE (TA) l
/KJM Reactor Trip System ,
Interlocks N.A. N.A. N.A. 1 x 10 88 amp 1[6 x 10 88) amp .
- a. Intermedlate Range Neutron Flux, P-6 l
- b. Low Power Reacter Trips !
Block, P-7 N.A. N.A. N.A. _0 of RTP** _ 2.2% of RTPa*
- 1) P-10 input M.A. N.A. < 0 RTPa* Turbine 2.2 RTP** Turbine
- 2) P-13 input N.A. Impulse Pressure :
Tapelse Pressure y - Equivalent Equivalent . Power Range Neutron N.A. N.A. N.A. < 8 of RTPa* 1 50.2'JE of RTPa* , c. Flux, P-8 M.A. N.A. of RTPa* 1 2.2 of RTP**
- d. Power Range Neutron M.A. _
Flux, P-9 Power Range Neutron M.A. N.A. N.A 1OkofRTP** 1 8NofRTPa* e. Flux, P-10 g f. Turbine impulse Chamber N.A. N.A. N.A. h RTPa* TurbineImpulse lopulse Pressure I 2.2 RTPa* Turbine Pressure ig o Pressure, P-13 Equivalent Equivalent -l gg . n 4,2tf. Reactor Trip Breakers M.A. N.A. N.A N.A. N.A. :l N.A. i N.A. N.A. N.A. M.A. E'lo 2f. Automatic Trip and Interlock
- Q .
" Logic ,_
2D
# .I aaRIP = Ratto TiitRMAL POWER .
] ! '
Page No. 17 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECM. SPECS. NOTE TYPE NOTE i NOTES 00 2.0 PAGE: 2.0 2- 8.0 OPERATING MODE: 123456 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTING'S DATA 1) Table 2.2-1 Reactor Trip System Instrumentation Trip Satpoints. Preliminary STPEGS values provided.
- 2) Table 2.2-1 Reactor Trip System Instrumentation Trip setpoints. Delts T definition deleted. RTD manifold instrumentation under review. Delta T definition will be provided LATER.
O
.t e
t C
/
'l i
TABLE 2.2-1 (Continued) . TABLE HDTATIONS l 9 'l' M
** NOTE 1: DVERTEtrERATINtE AT I
[{ [T (1 f v 5) - T'] + Ks(P - P') - f (AI)I AT { (3 , ,,5) i AT , N - Ks I
'l j g ( Lsle.d ^^
_ ;=:_ -
^ ^ ^
Idhere: si = . _ _ _ =7 f;-^ _=__ ___ l i i f{ = Lead-lag compensator on measured AT; I ants utilized in lead-lag cespensator for AT :: 8 p. re Is = Ti c gf,,5 = Lag compensator en seasured AT; O Ya
= Time constants utilized in the lag desponsator "for AT. Ya " E W '
I . l re
= Indicated AT at RATED THERMAL POWER; l s's AT, K.
= EEEEEk I.0fd . .
K, = EEiW5NIR!9:; 0.0 875 e
= The function; generated by the lead-lag compensator for T8 '8 I { **l dynamic compensation; 33
) r., is = Time constants utlitzed in the lead-lag compensator for T8, v4 = 3357.se,
- v sg4Qsas.
T = Average temperature. *F; l nM oo I l*8 **"P'"'***' ** ****"'*d I avg i 1 + s.5 , . O 3atA . ] $N v. = Time constant. utilfred in the measured T,,g tag compensator, t. = h j zg . C i 8 - j i ' m
Page No. 18 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE f NOTES l oo 2.0 PAGE: 2.0 2- 9.0 OPERATING MODE: 123456 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS I l DATA 1) Table 2.2-1 Reactor Trip System Instrumentation Trip Satpoints. Preliminary STPEGS specific values j provided. ! Table 2.2-1 Reactor Trip System Instrumentation 2) l Trip Satpoints. Delta T definition deleted. RTD 1 manifold instrumentation under review. Delta T ! definition to be provided LATER. l i-O
.t O
" r O . 0 .. - O . I I i TABLE 2.2-1 (Continued) , I 'T TA8tE NOTAil0NS (Continued)
- j M
NOTE 1: (Continued) . 593.0 - T' <
~ [5es.5)*F (Nominal T"'* at RATED THERMAL POWER);
- o. coots 7
. Ra
= EEEEEEEE3/psfg; .
= Pressurizer pressure, psig; .
P
'I.
- P' = 2235 psig (Nominal RCS operating pressure); .
, 5 = Laplace transfem operator, s 8; l and f (AI) is a functlen of the indicated difference between top and bottom detectors of the ' power-range'vlaa6rers los chambers; with gains to be selected based on measured instrument response during plant startup tests such that: 34 10.0 l . (1) For g g g between -EasgE and + mix, f,(at) = o, where g g and g are percent RATED THERMAL ! E POWER in the top and botten halves of the core respectively, and gg+gb is total THERMAL 4 POWER in percent of RATED THE m 4L POWER; l (2) Foreachpercentthatthemagnitudeof(qg g)nceeds 35 the AT Trip 5etpoint shall f be automatically reduced by of its value at RATED THE m4L POWER; and i ! 10 . (3) For each percent that the gly of(qg g) exceeds *EE35,theATTripsetpointsha11 be automatically reduced by EfSEii]E of its value at PATED THERMAL POWER. j g NOTE 2: The channel's maximum Trip setpoint sha11 not exceed its computed Trip setpoint by more.than . E5 o AT sma. i M
-~
i 82 - ; l z :;! . B . t m ~
Page No. 19 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE # NOTES 00 2.0 PAGE: 2.0 2-10.0 OPERATING MODE: J23456 SAFETY LIMITS AND LIMITING SAFETY' SYSTEM SETTINGS DATA 1) Table 2.2-1 Reactor Trip System Instrumentation Trip Satpoints. Preliminary STPEGS specific values provided. DESIGN 2) Table 2.2-1 Reactor Trip System Instrumentation Trip Satpoints. Delta T definition deleted. RTC maniford instrumentation under review. Delta T definition will be provided LATER. ED 3) Table 2.2-1 Reactor Trip System Instrumentation Trip Setpoints. ' rate-lag' changed to ' lead-lag' to make nomenclature consistent with Note 1. s - G
.I t
{ i l [ O l L
O r O ~ O,
- A i
k
- ,t!
TASTE 2.2-1 (Centinued) . e
'T TABLE IIDIAT10005 (Continued) .
n, - IRTE 3: OWERPOWER AT
- I T - 4 [T f1 f v.5 - D ~ # 8(AIII .I AT
- fl+153 i AT ,(K. - MS f r,5 v5
,i
- g. = C t.a.l=Q f _ _ _ _ _
,i .
* = As defined in liste 1 r e , v: = As defined'la Ilote 1, l
= As defined in liste 1, 1 153
\ l u Ys
= As defined in IIste 1, l .'.
o y AT,
= As defined I.n Ilote 1, l
K. = EEEEE5,t.070 .
.l
= 0.02 for increasing average temperature and 0 for decreasing average i Ks .
te. ,e,atore. . i uAo g ,'s,5 = The function generated by the esta-lag compensator for T,q dynamic , compensation. r, = Tlee constants utilized in the N -lag compensate'r for T,q , y, b le ,- g . I = As defined in hte 1 . f R g , ,,3 ; GR i
- om vo
= As detined in 10ote 1 Q I
! Z :;l
- E m s I
. jj t-I
1 l Page No. 20 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE # NOTES oc 2.0 PAGE: 2.0 2-11.0 OPERATING MODE: J23456 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS, DATA 1) Table 2.2-1 Reactor Trip System Instrumentation Trip Setpoints. Preliminary STPEGS values provided.
~
- e.
.9 r
l o
.F j
. i 1
TAttE 2.2-1 (Continued) 9 . TABLE IIDTATlofts (Continued) .
} I
! IIOTE 3: (Continued) \ o.00135 K. = EEBRIES/*F for T > T* and Ke = 0 for T 1 T", l T = As deffned in Hote 1, i' = Indicated T #9 at. RATED 593,0 THERMAL POWER (Calibratten temperature for AT I l instrumentation, 1 F888:M*F ) , I 5 = As defined in pote 1, and l l i f (AI) = 0 for all AI.
- i i
l The channel's maximas Trip Setpoint shall not exceed its cesputed Trip 5etpelat by more than i y NDTE 4: - l p sasnL. 3 9 mccur.sr4y. j .
-i
.. ~.
h - l @9 . C29
'OE -
e5 .- 3 8 L. . Ii
- i
,,;.. .l '
.,... ..... T : . . . . T.. . . . * ~:. . ;< . . . . ~ _
2 4
. . NOTE
.The 8ASES contained in succeeding pages summarize
. the reasons for the Specifications in Section 2.0, but in accordance with 10 CFR 50.36 are no,t part of these Technical Specifications.
. g 8 9
?
l l I I e 6 O NUREG 0452/STPEGS COMPABISON A.UG 7 1980 (
; .:- .. . .. ..h= :
O 2.1 SAFETY LIMITS - BASES e
~
2.1.1 REACTOR CORE The restrictions of this Safety Limit prevent overheating of the fuel . and possible cladding perforation which would result in the release of fission products to the reactor coolant. Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the ' heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature. , Operation above the upper ' boundary of the nucleate boiling regime coul.d result in excessive cladding. temperatures because of the onset of departure from nucleate boiling (ON8) and the resultant sharp reduction in heat transfer coefficient. ON8 is not a directly measurable parameter during operation and therefore THERMAL POWER and reactor ecolant temperature and pressure have been related to DN8 through the W-3 correlation. The W-3 DN8 correlation has been developed to predict the DN8 flux and the location of DN8 for axially uniform and nonuniform heat flux distributions. The local DN8 heat flux ratio (DN8R) is defined as the ratio of the heat flux that would cause DN8 at a particular core location to the local heat flux and is indicative of the sargin to DN8.
}
The minimum value of the DN8R during steady-state operation, nonsal operational transients, and anticipated transients is limited to 1.30. This value corresponds to a 95% probability at a 95% confidence level that DNB l will not occur and is chosen as an appropriate margin to DN8 for all operating g conditions. The curves of Figure 2.1-1 a d @ i-M show the loci of points of THERMAL POWER, Reactor Coolant Systes pressure and average temperature for which the minimum DN8R is no less than 1.30, or the average enthalpy at the vessel exit is equal to the enthalpy of saturated liquid. I' N These curves are based on an enthalpy hot channel factor, F N ofA1755 and a reference cosine with a peak of 1.55 for axial power shape. AYa,llowanceis included for an increase in F"g at reduced power based on the expression: ' I.52 0.3 ! F"g = 2555 [1+ G52 (1-P)]
' Where P is the fraction of RATED THERMAL POWER.
These limiting heat flux conditions are higher than those calculated for the range of all control rods fully withdrawn to the maximum allowable control rod insertion assuming the axial power imbalance is within the limits of the f t(AI) function of the Overtemperature trip. When the axial power imbalance l 1s not within the tolerance, the axial power imbalance effect on the Over-temperatureATtripswillreducetheSetpointstoprovideprotectionconsistentl with core Safety Limits. NUREG 0452/STPECS COMPARISON i W-STS B 2-1 . l .
. :.- - .; - - > . .. . _ , . . . . . . . . . . . . ,,, ,,_, , n._
SAFETY LIMITS . , 8ASES 2.1.2 REACTOR COOLANT SYSTEN 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.maximus transient pressure of 110% (2735 psig) of design pressure. The Safety Limit of 2735 psig is therefore consistant with the design criteria and associated Code requirements. SIO7 The entire RCS is hydrotested at 125% (3129 psig) of design pressure, to demonstrate integrity prior to initial operation. O. . 9 D f NUREG 0452/STPECS COMPARISON l if-STS 8 2-2 L
i
-.. ~ .; . ... .... . . . , . . ... . : * . . . . . . . . . .
-- ~ - - -
~
~ '
lO 2.2 LIMITING SAFETY SYSTEM SETTINGS gg
! BASES
- 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS The Reactor Trip 5etpoint Limits specified in Table 2.2-1 are the nominal .
values at which the Reactor trips are set for each functional unit. The Trip
- 5etpoints have been selected to ensure that the core and Reactor Coolant
- System are prevented from exceeding their safety limits during normal operation j, and design basis anticipated operational occurrences and to assist the Engi-neered Safety Features Actuation Systas in sitigating the consequences of accidents. The setpoint for a Reactor Trip System or interlock function is considered to be adjusted consistent with the nominal valud when the "as
; measured" setpoint is withinL the band allowed for calibration accuracy.
To accommodate the instrument drift assumed to occur between operational tests and the accuracy to which Setpoints can be esasured and calibrated, Allowable Values for the Reactor Trip 5etpoints have been specified in Table 2.2-1. Operation with Setpoints less conservative than the Trip Set-point but within the Allowable Value is acceptable since an allowance has been l ande in the safety analysis to accommodate this error. An optional provision has been included for determining the OPERABILITY of a channel when its Trip , 1 Setpoint is found to exceed.the Allowable Value. The methodology of this . ) option utilizes the "as measured" deviation from the specified calibration ' i point for rack and sensor components in conjunction with a statistical combin-ation of the other uncertainties of the instrumentation to measure the process variable and the uncertainties in calibrating the instrumentation. In Equa- ] tion 2.2-1, I + R + 5 i TA, the interactive effects of the errors in the rack .'.
; and the sensor, and the "as esasured" values of the errors are considered. 2, as specified in Table 2.2-1, in percent span, is the statistical suonation of errors assumed in the analysis excluding those associated with the sensor and 1 rack drift and the accuracy of their 'neasurement. TA or Total Allowance is
] the difference, in percent span, between the Trip 5etpoint'and the value used j in the analysis for Reactor trip. R or Rack Error is the "as measured" devia-tion, in percent span, for the affected channel from the specified Trip Set-point. 5 or Sensor Error is either the "as measured" deviation of the sensor from its calibration point or the value specified in Table 2.2-1, in percent span, from the analysis assumptions. Use of Equation 2.2-1 allows for a i, sensor drift factor, an increased rack drift factor, and provides a threshold value for REPORTA8LE EVENTS. - 1 . 3 The methodology to derive the Trip 5etpoints is based upon combining all of the uncertainties in the channels. Inherent to the determination of the Trip 5etpoints are the magnitudes of these channel uncertainties. Sensors and other instrumentation utilized in these channels are expected to be capable of operating 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. Seing 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 d
; more serious problems and should warrant further investigation. ! NUREG 0452/STPECS COMPARISON 4 PSTS B 2-3
~ ~ - - - *
...........~...,..w.-.~. ....;,-.. *
. ; .:- ~.~~~ . ~ ~ - -
l . .
~ - - . -
r
, LIMITING 5AFETY SYSTEM SETTINGS t
BASES W REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS (Continued) ' i .
! The various Reactor trip circuits automatically open the Reactor trip breakers preset orwhenever calculateda Ievel. condition monttored by the Reactor Trip System reaches a .
In addition to redundant channels and trains the i design approach provides a Reactor Trip System which monitors numerous sys, tem
, variables, therefore providing Trip System functional diversity. The functional
-*; capability at the speciffed trip setting is required for those anticipatory or diverse Reactor trips for which no direct credit was assumed in the safety analysts to enhance the overall reliability of the Reactor Trip System. The
,; Reactor initiated.TripThis System initiates a Turbine trip signal whenever Reactor trip is gravents the reactivity insertion that would otherwise result
. from excessive' R9 actor Coolant System cooldown and thus avoids unnecessary actuation of the Engineered Safety Features Actuation System.
Manual Reactor Trio The Reactor Trip System includes manual Reactor trip capability. l Power Rance. Neutron Flux In each of the Power Range Neutron Flux channels there are two independent histables, each with its own trip setting used for a High and Low Range trip * ', a setting. The Low Setpoint trip provides protection during subcritical and low.
) ; power operations to mitigate the consequences of a power excurston beginning 4
from low power, and the High Setpoint trip provides protection during power 1 power levels.to sitigate the consequences of a reactivity excursion from all operations
- The Low Setpoint trip may be manually blocked above P-10 (a power level of approximately below the P-10 Setpoint. 10X of RATED THERMAL POWER) and is automatically reinstated Power Rance. Neutron Flux. High Rates The Power Range Positive Rate trip provides protection against rapid flux 3
Specifically, this trip complements the Power Range Neutr trips to ensure that the criteria are met for rod ejection from mid power. accidents. The Power Range Negative Rate trip provides protection for control rod drop At high power a single or multiple rod drop accident could cause local flux peaking which could cause an unconservative local DN8R to exist. The i Power Range Negative Rate trip will prevent this from occurring by tripping the j reactor.~ No credit is taken for operation of the Power Range Negative Rate trip for those control rod drop accidents for which DN8Rsc will lbe grgtgr gjgg 3-STS B 2-4 COMPARISON O
------nn--,. ..n,_n-,---m- - - - ,,. . , ---,--,,,_n,___ -,c_-ne_..?,.r.,--.,-
- _ _- _~ =- - _ - - - ._ . __ . _---- _
~
l l I - 1 ... . : ..~ l _. .......,-. - - -% ;,----- n . u . ...:=- v.-- - . . :- . . - . : : - -- - "=.;* *- ~
-~"= * ' ~ -
l j LIMITING SAFETY SYSTEM SETTINGS I 8ASES
. I Intermediate and Source Rance. Neutron Flux , f The Int.ormediata and Source Range, Neutron Flux trips provide core l protection during reactor startup to mitigate the consequences of an uncon- -
trolled rod cluster control assembly ' bank withdrawal from a subcritical condition. These trips provide redundant protection to the Low Setpoint trip l of the Power Range, Neutron Flux channels. The~5ource Range channels will , initiate a Reactor trip at about los counts per second unless manually blocked when P-6 becomes active. The Intermediate Range channels will initiate a Reactor trip at a current level equivalent to approximately 25% of RATED THERMAL POWER unless manually blocked when P-10 becomes active. No catoir *45 TEEN van 09tRatioW CF rd TR*PS A550csATYD'sef t m tifMR ralIMttRMEDIATC QR Qvertemperature AT sovece RawGe crawNta N rue scenoeur Awatysts; pees vfR, r#M C#cr yAnturY AT rns .spccarico rase urrskos as nuevosto si runs The Overtempera IAY'Ff,y ro suaswer rw oortAu. ntuAssuTY ov rne firneron Taar S l combinations of pressure, power, coolant temperature, and axial power distribu-tion, provided that the transient is slow wifa respect to piping transit delays from the core to the temperature detectors (about 4 seconds), and pressure is within the range between the Pressurizer High and Low Pressure trips. The Set-point is automatically varied with: (1) coolant temperature to correct for [ temperature induced changes in density and heat capacity of water and includes ' dynamic compensation for piping delays from the core to the loop temperature detectors, (2) pressurizer pressure, and (3) axial power distribution. With normal axial power distributton, this Reactor trip Ifmit is always below the core Safety Limit as shown in Figure 2.1-1. If axial peaks are greater than . design, as indicated by the difference between top and botton' power range -
- nuclear detectors, the Reactor trip is automatically reduced according to the notations in Table 2.2-1.
Gotionai-ior--Giant. -;.m ; i,Wd -o-1-L , -Le- -t b l
=0perstbi-wit;.-. , ^-- - =1 .t 1:; e -1 ef sei 'ce be!c- !?= (=) ' v -
j *=4-Se;pointasoes -m % .ir -Re ,i=r Trip 5y.4es-5.'W-i =;! ^ !;;i!;-- i 3ecause the-*-*"etpoint and associatedWp etti prevent-ONS curing (--l^; toop operation exclusiv - i--th.-Overkw.r.i.or. ip- i. (-1) Z apard '- h ve-ths-(n)- Mep-8-! d-Epei-t != p n=see kie-eit:r r:::id x; er
- st2;=f= pet ir i^=-Overtemper.i.or. M t?===els an* ~ 412; B: - 5 ir= =
* -++ -*=-"-*oop salv._ In Bi: =-J: ef ever d en, th: ?-5 4 -trH r- ^ ==
;-:p -- _ t ! _ = : e 21 @ 2 uirer F h Er!? :L i^ie:re==t:-i p = :r 1:::1-Overpower AT The Overpower AT trip provides assurance of fuel integrity (e.g., no fuel pellet selting and less than IX cladding strain) under all possible l overpower conditions, Ifmits the required range.for Overtemperature AT trip, and provides a backup to the High Neutron Flux trip. The Setpoint
- 1s automatically varied with
- (1) coolant temperature to correct for tempera-J ture induced changes in density and heat capacity of water, and (2) rate of
- \
NUREG 0452/STPEGS B 2-5 COMPARISON PSTS
\ . -
vy-yv-,--- - r ,w- c -
.,,-.m %.re-,,,---- ,,.u., -,,-nm ,,,_-,_m
i .. ,
;. ,- . c..:. .. .: ..w ev $.r -*:' N - "- '-
8 LIMITINGSAFETYSiSTEMSETTINGS BASE 5
. t
, Overpower AT (Continued) *
- ( change of temperature for dynamic compensation for piping delays from the core ,
, p to the loop temperature detectors, to ensure that the allowable heat genera-
.] tion rate (kW/ft) is not exceeded. The Overpower AT trip provides protection to mitigate the consequences of various size steam breaks as reported in WCAP-9225, " Reactor Core Response to Excessive Secon,dary Steam Releases."
Pressurizer Pressure 3 In each of the pressurizer pressure channels, there are two independent f:* histables, each with its own trip setting to provide for a High and Low Pressure trfp thus limiting the pressure range in which reactor operation is permitted. ' The Low Setpoint trip protects agafnst low pressure which could lead to DN8 by 4 tripping the reactor in the event of a foss of reactor coolant pressure. On decreasing power the Low Setpoint trip is automatically blocked by P-7
. (a power level of approximately 105 of RATED THERMAL POWER with turbin chamber pressure at approximately 105 of full power equfvalent); and on increasing power, automatically reinstated by P-7.
The High Setpoint trip functions in conjunction with the pressurizer
- j relief and safety valves to protect the Reactor Coolant System against system overpressure.
Pressurizer Water Level j .i 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-mately 105 of RATED THERMAL POWER with4a turbine impulse chamber pressure at^o" l
approximately 10% of full power equivalent); and on increasing power, auto-matica11y reinstated by P-7.
\
Reactor Ccolant Flow
~ !
1 The Low Reactor Coolant Flow trips provide core protection to prevent DN8 l t by mitigating the consequences of a loss of flow resulting from the loss of one or more reactor coolant pumps. 1
, - On increasing power above P-7 (a power level of approximately 105 of
' RATED THERMAL POWER or a turbine tapulse chamber pressure at approximately 10%
of full power equivalent), an automatic Reactor trip will occur if the flow in g .more than one loop drops below 905 of nominal full loop flow. Above P-8 (a t power level of approximately]435 of RATED THERMAL POWER) an automatic Reactor l
' trip loop full will occur flow. if the flow in any single loop drops below 90% of nominal Conversely, on decreasing power between P-8 and the P-7 an automatic Reactor trip will occur on low reactor coolant flow in more than one loop and below P-7 the trip function fs automatically blocked.
PSTS NUREG 0452/STPEGS 8 2-5 COMPARISON e 9 b
. ,. .. 5- .n..
. .. . . -1 -:m - ~~ ~
" " ~ ~ . . . . .
LIMITING SAFETY SYSTEM SETTINGS 8ASES v^vifensi-hv41 ants =#ereft*hi- ._c.; C.,.r.ete. - 4he=p=*-$dy.6L tr!p ='11 ;m...; ??-e =Eintaus=weWR= free. going =6elow 140:during:neraal= operational = transients- c4- ek'., when={n-tF4eope=ere 4n-eperation end4he4vertem ted --t....e4,,e edjusted-t., th=velve=epecified=fot<=elMe ys ';g.sttn. perature4T 0 '.p.' .^. % ' Uith!:';1:.. t l gperethr4%=P=4= trip =et=U6X =AATE04HERMAL-POWER ^'
......l':'_
wehd th=0NSR=9 pen goingAntow=ba0=dur';; :; ,a1 :;;;;t' :21 -t. ;';;.-
=end ent h @ t d '.. ..iente=wi*l. !. 1: Ing '
,.;t' n. I
, Jt,eam Generator Water La' vel ~
1ess of heat sinkThe Steam Generator Water Level Low-Low trip protects the re' a < resulting from fos.in the event of a sustained steam /feedwater flow afsmatch I s of normal feedwater. allowances for starting delays of the Auxiliary Feedwater Systas.The specifi . 6 Stese/Feedwetr 't. Mesatc' ; .4-tow-6 team "e
. . rem.. _.... .....
' ,,,,,,gMggg __ 1 kow&L. L;;;l= trip =4s=not== sed 4n=the= transient 1 ^. ';
=and=ecci '
l 4mcieded=in4able4J-4=to= ensure:the-functional =capaht!'Ey ;f I?= :. r:!"M
- trip:vettir.y. c.4=thereby-*+%:overa)4=rel4abit it; =f i' ; "=n t ;- -!p --
' 1
%.'=-:-3Ms=
%. :,. . trip =is=sedundant=to=the=5 team /seneratoc=Wata= L;ui ' n '__-
The= Steam /Feedwaten=F-low = Mismatch portion of=tM5 try ! _ i!-i __ when=the:eteam: Slow-exceeds =the:feedwater=$1ow:by greater =then- a c = ==? i.
)
T+-'2-.4G9:tts/te.r tr!,; !5 Jhe=5 team-Senerat.c4 L vated *--*%~e+=r !weL4s.ys:6elow: : Water = level p. i': e? ^1
- nerrow: range-instr mt G5]E res 'ed k=ted 'i_ 1 "h se-trip =ealues= include:suf*4cfent=e!!: ._=L _ N
,emcess w=4eaet-r-tri =f --real= operating: values terpreclude=spardous= tripe L;t *" '
?_!-i tapaci+y e+p-before=the: steam generators-are-4.j.---Therefera, r I
n _ etarting= time =requir- : ate af=the=auxiliaryJeedwata- - T n^2; . ,_' .1 i-.44he resulting: thermal =transtea+-*a-tM:Asac^cr "eek s' _ ... seisse te M=f:'1 L e--Ty-+==-===& Undervoltace and Underfrecuency - Reactor Coolant Pumo Busses
. The Undervoltage and Underfrequency Reactor Coolant Pump aus trips pro-flow. core protection against ON8 as a result of complete loss of forcedl coolant vide The specified Setpoints assure a Reactor trip signal is generated before the Low Flow Trip 5etpoint is reached.
the from Underfrequency momentary electrical and powerUndervoltage transients. trips to prevent spurious Reactor trip For undervoltage, the delay is set so that the time required for a signal to reach the Reactor trip breakers I I'I following the simultaneous breakers shall not exceedfCt;: trip of two or more reactor coolant pump bus circuit 4] seconds. so that the time required for a signal to reach the Reactor trip breakersFor un aftertheUnderfrequencyTrip5etpointisreachedshallnotexceed[tlcSifsecond NUREG 0452/STPEGS
)t-STS 8 2-7 " S
_ _ = - . . .
-. . . . . . . . _ --- :.. : .. : . . .;. . : . . E.~ _
~
.: , .s.
t 1 O., ' LIMITING SAFET SYSTEM SETTINGS r
, 8ASES Undervoltace and Underfrecuency - Reactor Coolant Pump Busses (Continued) 4
' Bus trips _tre automaton decreasing power the Undervoltage and Underfrequenc
/
i '" or NATED THERMAL NhiEA with a turbine impulse chamber 10% try P-7.of full power equi alent); and on increasing power, reinstated automatic v ' i
' Turbine Trip ,
A Turbine trip initiates a Reactor trip.
.. , trfp from the Turbine trip is automatically blocked by P-9 (a power level ofOn
( approximatelyby50% automatically P-9. of RATED THERMAL POWER); and on increasing power, reinsta t Safety Iniection input from ESF . System instrumentationIf a Reactor trip has not already been generated by the Re a Reactor trip upon any, signal which initiates a Safety Injection.the E t i O
- instrumentation Table 3.3-3. channels which initiate a Safety Injection signal are shown in The ESF Reactor Trio System Interlocks The Reactor Trip System interlocks perform the following functfons: '
., P-6 On tripincreasing (i.e., prevents power P-6 allows the manual block of the Source Ran '
'ge
^
i.,, M i ".. .". premature block of Source Range trip) (;. :__ ._mswS ! ?. the high voltage to thevdetectors... -A a . "..;.. ... "'MMM andfdeenertslIm , On decreasing power, Source
- ange Level trips are automatically reactivated and.high voltage restored.
1 i ' P-7 t On increasing power P-7 automatically enables Reactor trips on low ) I j l } flow in more than one reactor coolant loop, reactor coolant pump j bus undervoltage and underfrequency, Tat-f _ t. L, pressurizer low
.' pressure and pressurizer high level.
l i
, listed trfps are automatically blocked.On decreasing power, the above E
I e lO MSTS 8 2-8 NUREG 0452/STPEGS COMPARISON 9
. . . . > esp . %- * -, . _ _ . ~ _ , _ _ _ _ - - . - . _ _ , - ..~
' . ,$ . ,, . . s u . . .... -- " '.----- , .if1 h- - . .:.
. . - . 2.- --- -- -- - . , . . . .
LIMITING SAFETY SYSTEM SETTINGS 8ASES
~
Reactor Trio System Interlocks (Continued) P-8 On increasing power, P-8 automatically enables Reactor trips on low . l flow in one or more reactor coolant loops, ha
- L.^ ;74reakerm On decreasing power, 'the P-8 auto-matica11y 61ocks the #: : : :: :';-E. UW AEACToft COOLANT FLOW TRIP 8hl CNE Loop. .
P-9 On increasing power, P-9 automatically enables Reactor trip on . Turbine trip. On decreasing power, P-9 automatically blocks Reactor trip on Turbine trip.
- P-10 on increasing power, P-10 allows the manual block of the Intermediate Range trip and the Low Satpoint Power Range trip; and automatically blocks the Source Range trip and deenergizes the Source Range high
- voltage power. On decreasing power, the Intermediate Range trip and the Low Setpoint Power Range tripsare automatically reactivated.
Provides input to P-7. , P-13 Provides input to P-7. .
. 't e
G 6 0 0 . NUREG 0452/STPEGS COMPARISON y-5TS 8 2-9 6
1 1 l O l i i l L I f I l 1 l r l t l l r l t l 9w - e i I i I i i l l l t l t f I O r j NUREG 0452/STPECS COMPARISON i I
Pzgo No. 1 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE # NOTES co 3/4.3.1 PAGE: 3/4 3- 1.0 INSTRUMENTATION - REACTOR TRIP SYSTEM INSTRUMENTATION NINE 1) 3/4.3.1 No change. s O-l i i i l O I
m.rra.u .. . . . . . . . . v. '.-- --. . =. . . . w . . ,e .. .. . - ~ . mn O s . DRAFT , 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.1 As a minimum, the Reactor Trip System instrumentation channels and interlocks of Table 3.3-1 shall be OPERA 8LE with RESPONSE TIMES as shown in Table 3.3-2.
*APPLICA8ILITY: As,shown in Table 3.3-1.
ACTION: As shown in Table 3.3-1. SURVEILLANCE REQUIREMENTS 4.3.1.1 Each Reactor Trip Systan instrumentation channel and interlock and l \ the automatic trip logic shall be demonstrated OPERABLE by the performance of the Reactor Trip system Instrumentation Surveillance Requirements specified in l Table 4.3-1. . 4.3.1.2 The REACTOR TRIP SYETEM RESPONSE TIE of each Reactor trip function l
- shall be demonstrated to be within its Itait at least once per 18 months. .. '
Each test shall include at least one train such that both trains are tested at least once per 36 months and one char.nel per function such that all channels are tested at least once every N times 18 months where N is the total number - of redundant channels in a specific Reactor trip function as shown in the l
" Total No. of Channels" column of Table 3.3-1.
O W-STS
~
3/4 3-1 i NUREG 0452/STPEGS l
- COMPARISON
f P g3 No. 2 06/13/85 COMPARISON CF NURF3 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE # NOTES co 3/4.3.1 PAGE: 3/4 3- 2.0 INSTRUMENTATION - REACTOR TRIP: SYSTEM INSTRUMENTATION ED 1) Table 3.3-1 Item 1. Action items renumbored, changed from 'll' to '10'. DESIGN 2) Table 3.3-1 Item 6.c. Added per WCAP-10271 and. Essociated SER. DESIGN 3) Table 3.3-1 Item 7A. Overtemperature Delta-T Three Loop Operation deleted. STPEGS plans no n-1 loop operation. DESIGN 4) Table 3.3-1 Item 7B. Overtemperature Delta-T Three Loop Operation deletet. STPEGS is a four-loop plant. y v O
, e s
4 F ' w _ _ _ _ _ _ _ _ _ _ _ m.
~
l
.f-x
'l TABLE 3.3-1 .
d
'T .i REACTOR TRIP SYSTEN Ill5TRistENTATION MINIIRM - TOTAL IIS. CHANNELS CHAlelELS APPLICABLE TO TRIP OPERABLE MODES ACTION ' 'I OF CHAiGIELS FullCTIONAL UNIT 2 1 2 1 2 1 -
- 1. Manual Reactor Trip 3 , 4* , 5*
I 41 10 2 1 2 ,-
~
- 2. Pouer Range Neutron Flux 28 4 2 3 1, 2 ,-
- a. High Setpoint 18##, 2 2#
4 2 3
- b. Lou Setpoint ,
- n. I .
4 2 3 1, 2 2#
- 3. Pouer Range, Neutron Flux ;I, High Positive Rate . s i.
3 1, 2 2#' *i Pouer Range,lieutron Flux, 4 2
. 4.
High Negative Rate .
.l'
)
1 2 1###, 2
- 3 -
- 5. Intermediate Range, lieutron Flux 2
- 6. Source Range, Neutron Flux 2 1 2 2## .' 4
- a. Startup ,
5 -
- b. ,Shutdoun 2 1 2 3,4,5 .
2, 3 f,2, (o#
- 7. Overtemperature AT 4 -
fl
.i
-Four=toop-Plant: 55
- -__L,,__=1?= 4 2 3 ~1, 2
;:-=i!= i 1*' 3 ir-2 9 l!
i;..._ L_, }' ) l L. 7 6 L _, n ant: , M i;.. _. L , ^, _ . .ii ---------3--------- 2------ -- 2 1r2 . 75
- 1** 2 1.-'?
9 '. nM T= 1_ , , _ _ =1 ? = 5 e
. on '
2 NOTE: c M lk u *
, ,ETC.- SEE END eTAuE. > !
e$ ACTION Srursrurs-2s swo or TAME. . .
".1 E,
y f: - I i ,- s , ,
r Page No. 3 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE # NOTES 00 3/4.3.1 PAGE: 3/4 3- 3.0 INSTRUMENTATION - REACTOR TRIP SYSTEM INSTRUMENTATION DESIGN 1) Table 3.3-1 Item 8A. Overtemperature Delta-T Three Loop Operation deleted. STPEGS plans no n-1 loop operation. DESIGN 2) Table 3.3-1 Item 8B. Overtemperature Delta-T Three Loop Operation deleted. STPEGS is a four-loop plant. DESIGN 3) Table 3.3-1 Item 9B. Pressurizer Pressure-Low Three Loop plant deleted. STPEGS is a four-loop plant. DESIGN 4) Table 3.3-1 Item 10B. Pressurizer Pressure-High Three Loop plant deleted. STPEGS is a four-loop plant. DESIGN 5) Table 3.3-1 Item 11. Pressurizer Water Level-High Total No. of Channels changed from '3' to '4' to reflect STPEGS design. , DESIGN 6) Table 3.3-1 Item 11. Pressurizer Water Level-High Minimum Channels operable changed from '2' to '3' to reflect STPEGS design. DESIGN 7) Table 3.3-1 Item 11. Pressu112er Water Level-High l Action changed from '7f' to '6f' since Action 6 l applies to this logic. OTHER 8) Table 3.3-1 Note (***) added per WCAP-10271 instructions. DESIGN 9) Table 3.3-1 Item 12. Action changed to '6#' per WCAP-10271 and associated SER. l i l l l O l l
..<..i.....:.,.-:* .,
.. .. . . s. . . .
, )
8L TABLE 3.3-1 (Continued) h REACTOR TRIP SYSTEM INSTRUNENTATION . MININIM ., TOTAL NO. ^ CHANNELS CHANNELS APPLICABLE - FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE M00ES ACTION a 4 E 3 1, E 6, g -l
- 8. Overpower AT '
5
- . nc ioop=ebt=
four=L _, .0; _. die i 2 3 1. 2 7 5 twas-i_ _, C, _. =i!= 4 1" 3 .2 2 d b* 5 -- *
-,r sq
- == 5,, 5 y _7 --
5
; = . _ _ , .c . .. ; = a 1- 2.- : 7
- 9. Pressurizer Pressure--Low 4 2 3 1 6
=. F-_.-Loop: Plani i 2 3 1 55 f ) b. Thres: Loop-Phi . - - -- - -- I 7F 6M "J, 10. Pressurizer Pressure--High 4 2 3 f 2,
- . F__.-Loop-Plasik ----
i n----- = == 2 ---=--- Id 5*- L. Thi-- L:_, ? L.i 3- ---?-- --
?-- 1--I ??- i .'
- p. -
- 11. Pressurizer Water Level--High -R 4 2 t3 1 77 6 i i
- 12. Reactor Coolant Flow--Low # '
- a. Single Loop (Above P-8) 3/ loop 2/ loop in 2/ loop in 1 9F 6 I
, any oper- each oper- ;
ating loop ating loop 6 '
- b. Two Loops (Above P-7 and 3/ loop 2/ loop in 2/ loo'p 1 $$ 6* l below P-8) two oper- each oper- **
h . gg ating loops ating loop E5 o ll; $; :.
- e en;g ,
g .- I
P;g3 No. 4 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE # _______________ ______ NO ________________________TES co 3/4.3.1 PAGE: 3/4 3- 4.0 INSTRUMENTATION - REACTOR TRIP SYSTEM INSTRUMENTATION DESIGN 1) Table 3.3-1 Item 14. deleted. Steam Generator Water Level Not applicable to STPEGS design. ED 2) Table 3.3-1 Item 15 - 17. Renumbered 14 - 16 due to deletion of item 14. DESIGN 3) Table 3.3-1 Item 14 (was 15) Undervoltage - Reactor Coolant Pumps. Three Loop Plant deleted. STPEGS is a four-loop plant. DESIGN 4) Table 3.3-1 Item 15 (was 16) Underfrequency - Reactor Coolant Pumps. Three Loop Plant' deleted. STPEGS is a four-loop plant. DESIGN 5) Table 3.3-1 Item 16b (was 17) Turbine Trip - Channels to Trip changed from '4' to '2' to reflect STPEGS design. DESIGN 6) Table 3.3-1 Item 16b (was 17) Turbine Trip - / Minimum Channels Operable changed from '1' to '3' ' to reflect STPEGS design. . DESIGU 7) Table 3.3-1 Item 16b (was 17) Turbine Trip -
- Action changed from '12f' to '6#' since action applies to this logic.
DESIGN 8) Table 3.3-1 Item 16a (was 16a) Turbine Trip - Change title from " Low Fluid Oil Pressure" to " Low Emergency Trip Fluid Pressure" to reflect correct title for STPEGS. Action changed per WCAP-10271 and associated SER. OTHER 9) Table 3.3-1 Note (***) added per WCAP-10271 and associated SER. O i
n e
- l' t
?
. I TABLE 3.3-1 (Continued) !
19 M
. REACTOR TRIP SYSTEM INSTRWlENTAT10" '.i MININUM TOTAL N0. CHANNELS CHANNELS Ai '98tf FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE ACTION
- 1
- 13. Steam Generator Water 3Nsta. gen. 2/sta. gen.
~
3/sta. gen. 1. 2
- Level--Low-Low in any oper- each oper-ating sta4 ating sta. !
gen., gen. . i e
.m a. _ _ . - _ _ ta_.__ a_ _. E __ e _a_ ___ e _ . _ _ ___ e _._ ___ S e AT. a..w - -__._.us.w. u.w .m . .ww. m wwou. . e oeu.. . a w .u.. p... mg a atm.
.u m a a = an .a.w w - . . . ,_n 4- . _ _ R_
.- . __A S kW...--__. W .a-aW e.____, .wyu. _ _ _
.uwe5 hw . .W .www5
_. , - --__m _
. . . __ .. _ . ,___._m _ . - .. , _,m__. _ . . l .
. ._ . .... . .- .._ . .,... . . _ . . . . . . . . .,...wu ,
w_
. .-.,-__- ..w ._
_,___._m s. i
-._-w.. .- g' y -- e ^ . C _ ^ ", ' .. ;- e ^ u.. . :
+ a. .. :--^ e' .
; ^ ,..
N .wwwg
. . . . >-_____ f _-
e . 1 ._
+ , . .,..w - .
..w .us u _ .bH
'a :'- e' .
w .. . .. l4 ) s. Undervoltage--Reactor Pumps Coolant *g m
.';_: :.::; ?'_;.".
,m___ .
4-1/ bus 2 3 1
- e. . .....w a.wwy - . .. ....
- u. J ar e.w. & & & #W j
15p. Underfrequency--Reactor Pumps Coolant . . -
';;; '.::; Th ."_. -
2 3 1 6# *
. ,n___ ,___ ._ . 4-1/.
. bus . -
g . . .. . w w __....... .,.
. w. . . .
.. . , . . , 3
. Turbine Tri EMERGENCY TRIP .
- a. L luid bet. Pressure 3 2 2 1 WG# - .
- v. b. Tu ine Stop yalve closure 4 5,?, K3 1 1&(J# t"*3 s
m .n om 2: H N M . i tvl o .w .
Page No. 5 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE # NOTES oc 3/4.3.1 PAGE: 3/4 3- 5.0 INSTRUMENTATION - REACTOR TRIP SYSTEM INSTRUMENTATION ED 1) Table 3.3-1 Item 18 - 21 Renumbered 17 - 20 due to deletion of item 14, page 3/4 3-3.
- 2) Table 3.3-1 Item 17 (was 18) Safety Injection Input from ESF. Action changed from '10' to '9' since actions are renumbered.
ED 3) Table 3.3-1 Item 19 (was 20) Reactor Trip Breakers. Action changed from '10' to '9'. Actions were renumbered. ED 4) Table 3.3-1 Item 19 (was 20) deactor Trip Breakers. Action changed frore 'll' to '10'.
** Actions were renumbered.
ED 5) Table 3.3-1 Item 20 (was 21) Automatic Trip and Interlock Logic. Action changed from 'll' to '10' and '10 to '9'. Actions were renumbered.
- 6) Table 3.3-1 Item 17 (was 18) Safety Injection Input - identify as coming from "ESFAS" rather than "ESF" for clarification.
l l l O \ _ _
' 4
~
. . . . . . . . . . . ....n. .s. .. .
i i TABLE 3.3-1 (Continued) .
'T REACTOR TRIP SYSTEM INSTRistENTATION - ,
y - . sa i MINIMUM ! CHANNELS CHANNELS. APPLICABLE TOTAL NO. j OPERABLE ND0ES ACTION - OF CHANNELS To TRIP _ FUNCTIONAL UNIT ! 17% Sat.ty injection input 2 1, 2 Js'9 ., 2 1 l from ESFAS " I% Reactor Trip System Interlocks
'! a. Intermediate Range
- 2 1 2 2fJ 8 l .
Neutron Flux. P-6 , ,
- b. Low Power Reactor l Trips Block, P-7 3 1 8 .
P-10 Input 4 2 j. or - 1 8 P-13 Input 2 1 2 g l ll
",j, c. Power Range Neutron 1 8 !
4 2 3 Flux. P-8 2 3 1* - 8 sg
- d. Power Range Neutron : 4 1 Flux P-9 ?
l .:
- e. Power Range Neutron 1,2 8 I !
4 2 3 Flux. P-10 il t Turbine lopulse Chamber l
- r. 1 2 1 8 {
Pressure, P-13 2 1 2 1 2 M % Reactor Trip Breakers 2 2 1 2 3I . 4", 5* K - . I i nM OO
^
.ERRO K Automatic Trip and Interlock 2 1 2 I 2 38'9 f Loggle 2 . 1 2 '3g , 4*, 5* JP 10 g !
kM N i hk;g > . l
]
l . . 8 I
P;ge No. 6 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE f . NOTES co 3/4.3.1 PAGE: 3/4 3- 6.0 . INSTRUMENTATION - REACTOR TRIP SYSTEM INSTRUMENTATION DESIGN 1) Table 3.3-1 Note (**) deleted. Note applies to n-1 loop operation. Not applicable to STPEGS
. design.
DESIGN 2) Table 3.3-1 Action Statement 2a - Change 1 hour to
+: read 6 hours. Per WCAP-10271 and associated SER.
DESIGN 3) Table 3.3-1 Action Statement 2b - Change period of bypass. Per WCAP-10271 and associated SER. OTHER 4) Table 3.3-1 Note (***) added per WCAP-10271 and associated SER. e
. G O
O
. . .. . . . . ..--~~ =~~ . .". ::."- . .. ~ . - - . .
TABLE 3.3-1 (Continued) . TABLE NOTATIONS 80nly if the Reactor Trip Systas breakers happen to be in the cle' sed position and the Control Rod Drive System is capable of rod withdrawal.
' R ; d ;.... i;'5) 555
=t-{ ' t;'- {;e i{jd '=[ gtj j]-. ' . - ^. - ii- 5 , ,
- .;;;;;===;;;.-;;;=g ; = ; ; g ; === . =- .. ,..
WlTH ~19tE Planblog$ of SMtg T1 V h*h fhefhy 3 FThe provisions of Specification 3.0.4 are not applicable. ' M8elow the P-6 dntemediate Range Neutron Flux Interlock) Setpoint. l N#8elow the P-10 (Low setpoint Power Range Neutron Flux Interlock) Setpoint. l i ACTION STATEMENTS l l ACTION 1 - With the number of CPERA8LE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to CPERABLE status within 48 hours or be in HOT STANDBY within' the next 6 hours. - ACTION 2 - With the number of OPERABLT channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION say proceed provided the following conditions are satisfied:
- a. The inoperable channel is placed in the tripped condition within hourg il
'z - -
b. h ?>E ' . .-; . .} %M C...i-;. .Qti- xt is E f . t; ,- t5-Q .-..
., M},;=.c-i
}.'. , 5
.; ==; _;; . ; ;;g ; ; ;...-. ;=a;.;;; g; ;g; . . ;; ;;=
- c. Either, THERMAL POWER is restricted to less than or equal to 75*: of RATED THERMAL POWER and the Power Range Neutron Flux Trip setpoint is reduced to less than or equal to M854.ofRATEDTHERMALPOWERwithin4 hours;or,the QUADRANT POWER TILT RATIO is monitored at least once per 12 hours per Specification 4.2.4.2.
b, hR CHAWAlff$ WrTH BY0%$$ FAPARLLtTY, Cyr ADDir/#lAl CNA4Nfl MAY Br le _ BYPASSED fot of 70 k konnts fot SERVflLLANCE Tf5TnN6 PG 4.%I.'l PtoviDen THE sNcPERABLE CHAyNCL LS IN TW TRifMD CoNDITied. 2.f0M tHtWWELS wrfu Na BYrAM CAPANuTY,74 MINIMyM egAW4 cts OPERABLE THf IN0PCRA6Li CMyNEL MAY BG bYPAS$fD UP t$ M1*] 70 2.Nou R$WEMR,SVRVEILLANCE FCR TE.S DMG 0F oTHER CHANNCL$ fEA PCcsF) C ,
- 4. 3. l. l .
NUREG 0452/STPEGS COMPARISON W-STS 3/4 3-6
- _ , , - _ . , , _ m , _ . . . - , , , . , . _ _ . - , _ . . . _ - _ . _ - . , - . _ - . _ - . . . - _ _ - . _ _ _ . , - _
f P ge No. 7 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE # NOTES oc 3/4.3.1 PAGE: 3/4 3- 7.0 INSTRUMENTATION - REACTOR TRIP SYSTEM INSTRUMENTATION DESIGN 1) Table 3.3-1 Action 5. Revised per WCAP-10271 and associated SER. DESIGN 2) Table 3.3-1 Action 6a. Placed in trip condition with 'l' hour changed to within '6' hours. *Per WCAP 10271 and~ associated SER. DESIGN 3) Table 3.3-1 Action 6b. Bypassed for up to 2 hours changed to 4 hours. *Per WCAP 10271 and associated SER. DESIGN 4) Table 3.3-1 Action 7 Deleted. *Per WCAP-10271 and
,. associated SER.
. G-e 9
O
i
*~~. * - - = . r . . T-l . . . u ... - .. - . . .. ... . . . . :.- . -. - -- - . i t
TABLE 3.3-1 (Continued) ,
~
ACTION STATEMENTS (Continued) ACTION 3 - With the number of channels OPERA 8LE one less thah the Minimum
- Channels OPERA 8 3 requirement and with the THERMAL POWER level:
' Below the P-6 (Intermediate Range Neutron Flux Interlock) a.
restore the inoperable channel to CPERA8LE - Setpoint,ior status pr to increasing THERMAL POWER above the P-i *
- . Setpoint, and l 8u" .
3*dU b. Above the P-6 (Intermediate Range Neutron F' lux Interlock) i d1Ce 'setpoint but below 105 of RATED THERMAL POWER, restors the l 1 k15 *g inoperable channel to CPERA8LE status prior to increasing v1 THERMAL POWER above 105 of RATED THERMAL POWER. I gn g! I tg .' ACTION 4 - With the number of CPERA8LE channels one less than the Minfaum ! eg w Channels 0PERA8LE requirement, suspend all operations involving - posi,tive reactivity changes. g j . u
% 0l & TION 5 - With the number of OPERA 8LE channels one less' than the Minimum l requir.eme.nt. =W = " .'=p - et?e Channels..O. P. ERA 8LE ~. .-' . - ?
U., u, g3 p, . . - -
.._.._....r-__ -, ,--- - ,, r
- g
^
---e-_ . ! : ! ^.- .: -l -; " - Y:?... 'i M EO ,... d ;;__ ....
g
*
- _ _ _ _ _ " " . ,. . " i. : . . . 2. " . . C. . . ^. . . . . y.
a u 4 lA u s I u ACTION 6 - With the number of CPERA8LE channels one less than the Total Number of Channels STARTUP and/or POWER OPERATION say proceed provided the following conditions are satisfied: = l
? * {
" a. The inoperable channel is placed in.the tripped condition 3
} 5.! withingur,and l l 3": b. 7:= " = "' -_._!= ^^~"""'I =+7_- -----t != --tM_._..
i g a g- -
. ,8 5, w .._,_. !- ..'._.___1 =;- te by, :: - -:= ;; = . =;
2 - -
?=:_ _ - _ e :- - ? 2_ _- . . _ - - - - _ _ - - _ _
~~ '-
gg
' ' ' ~ ~ ~ ~ - ~
'2 i~ i;, - . E. ..:. . . _._. T. .: .H_. _. - ' '
vi . $> _- ,m u . , ,, .._.p.-..,__.._.;.._._.5, . _g
; g _ _ .. . _. . . . . _ .
"""I
.$ N * "
.'. ? ' ' " "
f ' _ "l ' E ' E T '
~ "*
P :-.1_E T ' ~~ _ _ : T_~ ~. z -' 1 _~ : ?_ ..: 4 .u 2 31j 3m ACTION 8 - With less than the Minimun Number of Channels OPERABLE, within 1 hour determine by observation of the associated permissive annunciator window (s) that the interlock is in its required stat for the existing plant condition, or apply Specification 3.0,.3a 3 ---- g 4oemov44. cx4uuri. my so syPASSED r0A VPTD 4nev43 F08
- b. l. Fon
%ayeswe.amurts w,rn resras rzasyrus c4ranum,5,.I.i enevsoro rurs uoressais emun ss sg ratasenn Co*W.
wrearnesrsou 2,Few emuais mm we ernss carsstarr,rus mamm epaurais orgapatE asau,ae,.aur as Nsr ueneven, THE tsortanew eHAwust may at synsseo ran or su 4 saans von Sanvenanuca Ts.s} sus ce erses emuns rust Seregricarsaw 4.3.I.I. NUREG 0452/STPEG W STS 3/4 3-7 COMPARISON
i l Pcge No. 8 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE i NOTES 00 3/4.3.1 PAGE: 3/4 3- 8.0 INSTRUMENTATION - REACTOR TRIP SYSTEM INSTRUMENTATION DESIGN 1) Table 3.3-1 Action 9 deleted. Applies to n-1 loop operation, not applicable to STPEGS design. ED 2) Table 3.3-1 Action 10, 11, and 12 renumbered 9,
, 10, and 11 due to deletion of Action 9.
O n e O
4
.-.-..--:. .. --.---.-:- ...$...... i
~
O DWH TA8LE 3.3-1 (continued) , ACTION STATENENTS (Continued)
. .. .m
_. z -.- ,;. ..
.. z .-- -,
. . . . . , , . . . . . . . - . . . . _ . _ - - . - - __ -. 7,
- ,- 1;-. .. ; r - ;- - ;; -
i
. - - - - . - - . .. ,7.-
55:N? e~_f'~-_~.' _N.',~ _, _ :,. ,, _ ,-___ T _ -
". . ".'. X
..' __ -_- ' ~ .____
L E L : f.~1 -....; ~ ' I '_ , -_
;;:- :; :: : ::= : . . . _ . .
ACTION 3 -Channels With the number of OPERAELE channels one.less than the Minimum CPERA8LE requirement, be in at least HOT STAN08Y within 6 hours; however, one channet say be bypassed for up to
- 4.X hours for surveillance testing per Specification 4.3.1.1, i
provided the other channel is CPERA8LE. , i ACTION - With the number of CPERABLE channels one less than the Minimum l Channels CPERA8LE requirement, restore the inoperable channel to OPERA 8LE status within 48 hours or open the Reactor Trip System breakers within the next hour. ACTION ll - With one of.the diverse trip features (undervoltage or shunt trip attachment) inoperable, restore it to OPERABLE status within 48 hours or declare the breaker inoperable and apply 9 ACTIONA3. The breaker shall not be bypassed while one of the diverse trip features is inoperable except for the time required for performing maintenance to restore the breaker to OPERABLE status.
. 14CT10N d-With-the =of Channelsi=operationraay:continee number =ef=0PERA8LE=channele=less=then=the=Teta* provided-the=4aoperabts- *-
4 shannels are placed=in-the-tripped:<ondition wiiJiin-i .L T. 4 i 4 o4 - NUREG 0452/STPECS g-STS 3/4 3-8 COMPARISON l
1 Pzgo No. 9 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE # NOTES co 3/4.3.1 PAGE: 3/4 3- 9.0 INSTRUMENTATION - REACTOR TRIP SYSTEM INSTRUMENTATION
- 1) Table 3.3-2 Values verified for STPEGS. Delete parenthesis.
- 2) Table 3.3-2 Note (*) - Delete parenthetical l statement as it is not applicable to STPEGS.
l e - O l O 1 i I
t
- i TARLE 3.3-2 REACTOR TRIP SYSTEN INSTRUNENTATION RESPONSE TIMES ,
3 e:- RESPONSE TIME . . g. FUNCTIONAL UNIT N.A. ! - f
- 1. Manual Reactor Trip
- 2. Power Range, Neutron Flux iko.5 econd*
l [
- 3. Power Range. Neutron Flux, ,
N.A. l .. High Positive Rate -
?j I-
- 4. Power Range, Neutron Flux, -
x_ _f ;
*High Negative Rate $ 10.57 second*
il Intermediate Ranga Neutron Flux N.A. .,. (; 5. N.A. y 6. Source Range. Neu. on Flux , 4 econds* l h
- 7. Overtemperature aT
{ D4deconds* ( ,
- 8. Overpower AT 2bconds !
- 9. Pressurizer Pressure--Low
- 10. Pressurizer Pressure--Nigh , h2Neconds Pressurizer Water Level--Nhgh N.A.
l 11. i
~
E I ' nE i oo , E6 o . > i $$m . ,
~
'U E " Neutron detectors are exempt from response time testing. Response time of the neutron flux signal portion 1
* :;! of the channel shall be measured from detector output or input of first electronic component in channel.
.;'5- t= -??e --- _ = =ri=. 1_. -., 2, 1975. Ex: 1 . i t.,
h ii 1.115,
=g
" ^
' " ' - ^
M {
~~~~~' .
h
Page No. 10 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE f NOTES 00 3/4.3.1 PAGE: 3/4 3-10.0 INSTRUMENTATION - REACTOR TRIP SYSTEM INSTRUMENTATION DESIGN 1) Table 3.3-2 Item 14. Steam Generator Water Level-Low. Deleted. Not applicable to STPEGS design. ED 2) Table 3.3-2 Item 15 - 21 renumbered 14 - 20 due to deletion of Item 14.
- 3) Table 3.3-2 Item 16a (was 17a) Turbine Trip.
Change title from " Low Fluid Oil Pressure" to " Low Emergency Trip Fluid Pressure" to reflect correct title for STPEGS design.
*- 4) Table 3.3-2 Item 17 (was 18) Safety Injection -
identify as " Input from ESFAS to clarify. DATA 5) Table 3.3-2 Values verified for STPEGS delete . parathensis. i e e O J
- TABLE 3.3-2 (Continued) 8T 8 REACTOR TRIP SYSTEM INSTRIBENTATION RESPONSE TINES ,
h 1 a RESPONSE TIINE FUNCTIONAL UNIT '
- 12. Reactor Coolant Flow--Low
- a. Single Loop (Above P-4) [1 econd l ,,
f I j j b. Two Loops (Above P-7 and below P-8) 3J1TTecond D2decends
- 13. Steam Generator Water Level--Low-Law
.l
)
+
-*sumptartssumssa:UtsIF-j
- - ss= l i D1.5deconds i
,% undervoltage - Reactor Coolant Pumps '
I ] ! '% underfrequency - Reactor Coolant Pumps Sho.6decond bR Turbine Trip
- CMEAnotY TitiP ~-
Fluid 1b94 Pressure $1. A.
- a. LowA .!
- b. Turbine Stop Valve closure N.A.
q
% Safety Injection Input free ESFAS N.A.
.i
! 8IM. Reactor Trip System Interlocks N.A. . 1 l NM. Reactor Trip Breakers L. kA. j Automatic Trip and Interlock Loefc N.A. ! i M24
- 5 -
. sg -
g o I ls i . . E : ! R 21 F
.(
us l:.
r
=
l Page No. 11 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE # NOTES l _______________ ______ __________________________________________________ i 00 3/4.3.1 PAGE: 3/4 3-11.0 - INSTRUMENTATION - REACTOR TRIP SYSTEM INSTRUMENTATION
- 1) Table 4.3-1 Item 2a, 2b, 3, 4, 5, 6, 7, 8, 9, 10, 11, 12. Analog Channel Operational Test frequency changed from 'M' to 'Q'. Per WCAP-10721 and l associated SER.
DESIGN 2) Table 4.3-1 Item 7 Overtemperature Delta - T - Channel Calibration changed from "R (13)" to "R". Deletion of the RTD bypass manifold is under review for STPEGS. DESIGN 3) Table 4.3-1 Item 6, Source Range Neutron Flux -
Channel Calibration changed from "R(4,5,12)" to "R(4,5)" Notation 12 is not applicable.
DESIGN 4) Table 4.3-1 Item 2a,- Semiannual (R) Channel Calibration Actions expanded to include Action 6. Now reads "R(4,5,6) ". This change results from WCAP-10271 and associated SER. S e O
^
t TARLE 4.3-l' ,
'j
) 39 ! 3 REACTOR TRIP SYSTEN INSTRtBENTAT1tAl SURVEILLANCE REQUIREMENTS TRIP l ANALOG ACTUATING DEVICE MODES FOR lAIICH . CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUAT1001 SURVEILLANCE 'l CHAltlEL LOGIC TEST IS REQUIRE 0 i-i CALIRRATION TEST TEST CHECK t FUNCTIONAL UNIT N.A. R(10 N.A. 1, 2, 3*, 4*, 58 - Manual Reactor Trip N.A. N.A.
- 1. I i -
1 2. Power Ran08 Neutron Flux 5(2, 4), Jt'O(,3) N.A. . N.A. 1, 2 l i
- a. High Satpoint S M(3,4), ,
! 4(4.6),Q, R(4,5 3 1***, :I l l 5 R(4)
)l"g(is) N.A. N.A. 2 l
, w b. Low Setpoint
> [ <,s) N.A. N.A. 1, 2 ;!
\
- 3. Power Range, Neutron Flux. N.A. .R(4)
) u
' High Positive Rate N.A. 1, 2 i
- 4. Power Range, Neutron Flux, N.A.
- R(4) Jt'Q(IS) N.A. ,,
'l:
High Negative Rate Q(IS) 1***, 2 N.A. N.A. I
- 5. Intermediate Range. 5 R(4. 5) S/U(1)/ ,
1 i Neutron Flux Q R(4, 5) _10 S/U(1),p(9Xis) N.A. N. A. ~ 2**, 3, 4, 5 l tg
- 6. Source Range, Neutron Flux 5 l
k.A. N.A. 1, 2 l l
- 7. Overtemperature AT S R(331 Jt'kI85} if
'n N.A. 1, 2 N.A.
Overpower AT S R Jr'(ss) g8. 1 l 8 E 9. Pressurizer Pressure--Low 5 R [(IS.f6) N.A. N.A. 5o N;A. 1, 23 i
' $ $; 10. Pressurizer Pressuie--High S R [(IS'IG) N.A.
- N
@ $ 11.
Pressurizer Water sevel--High 5 R / Q(IS) N.A. N.A. 1 C ' N Reactor Coolant Fl..w--Low S }t )f'Q(IS) N.A. . N.A. I m l i8
,1 l
g 12. 3:=' . 1____________ -. ___ _ _ _
I Pr.ge No. 12 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. ' i NOTE TYPE NOTE # NOTES cc 3/4.3.1 PAGE: 3/4 3-12.0 - INSTRUMENTATION - REACTOR TRIP SYSTEM INSTRUMENTATION l DESIGN 1) Table 4.3-1 Item 14 Steam Generator Water Level - Low deleted. Not applicable to STPEGS design. ED 2) Table 4.3-1 Item 15 - 19 renumbered as Items 14 - 18 due to deletion of Item 14. DESIGN 3) Table 4.3-1 Item 13, 18a, 18b, 18c, 18d Analog Channel Operational test frequency changed from "M" to "Q". Per WCAP-10271 and associated SER. DESIGN 4) Table 4.3-1 Item 14 & 15 Trip Actuating Device
.. Operational Test frequency changed from "M" to "Q" . Per WCAP-10271 and associated SER.
- 5) Table 4.3-1 Item 16a - Change title from " Low Fluid Oil Pressure" to " Low Emergency Trip Fluid Pressure" to reflect correct title for STPEGS.
- 6) Table 4.3-1 Item 17 - Safety Injection Input -
Identify as coming from "ESFAS" rather than "ESF" for clarification. e
O :^O , o TABLE 4.3-1 (Continued)
- ty .
l $ REACTOR TRIP SYSTEM INSTRtMNTATION SURVEILLANCE REQUIREENTS , m TRIP i ANALOG ACTUATING N00ES FOR CNANNEL DEVICE %AtICN g l CNANNEL CNAlelEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE LOGIC _ TEST IS REQUIRE 0 gj - CALISRATlotl TEST TEST q FullCTIONAL UNIT CHECK _ l ! 13. Steam Generator Water Level-- S R /kI'#' RGt) N.A. 1, 2 Ig Low-Low , H. i- . _ _ _ _1 - , L _ _j
. . . 1, !
~
sa.
% Undervoltage - Reactor Coolant N.A. R N.A. . KQ (Is,IG) N.A. 1 <
- i
, 1:' Pumps i
- l5M, N.A. R N.A. K Q(IS,l(,) N.A. 1 i' l y Underfrequency - Reactor -
j g Coolant Pumps , 1 Ito l A 1 Turbine,J$Ec.cercy Tase N.A. R N.A. S/U(1,10) N.A. 1
- a. LowpFluid 1811 Pressure Turbine Stop Valve N.A. R N.A. S/U(1,10) N.A. 1 i i ~
- b. I Closure i .
i l'1 % Safety Injection Input from N.A. N.A. N.A. R N.A. 1, 2 ,[ ESFAS j
% . Reactor Trip System' Interlocks g a. Intermediate Range g .
' ' 2**
g I i j ng Neutron Flux, P-6 N.A. R(4) X (is) N.A. N.A. I
,l b. Low Power Reactor l g :- Trips Block, P-7 N.A. R(4) . )((Q 8 )(15)' N.A. N.A. 1
-{
Power Range Neutron Q C'3 MO c. N.A. R(4) )((8)DS) N.A. N.A. 1 2 l ;
@ l'l Flux, P-8 < p .
Power Range Neutron
@ d.
N.A.
)(15) N.A. N.A. 'l ,!
lj 8 Flux, P-9 _R(4) l ; i' I
Page No. 13 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE # NOTES co 3/4.3.1 PAGE: 3/4 3-13.0 INSTRUMENTATION - RCCTOR TRIP SYSTEM INSTRUMENTATION DESIGN 1) laule 4.3-1 Item 18e, 18f Analog Channel Operational Test frequency changed from "M" to "Q" . Per WCAP-10271 and associated SER. ED 2) Table 4.3-1 Item 20 - 22 renumbered 18 - 20 due to deletion of item 14. s e O-e O
l ,.' . . . . .. . l ( . i ( ~ TARLE 4.3-1 (Continued) . REACTOR TRIP SYSTEN INSTRWENTATION SURVEILLANCE RE@lREMENTS 3 TRIP . ANALOG . ACTUATING N00E5 FOR CNANNEL DEVICE WICN - CNANNEL CNAMEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE .d . FUNCTIONAL UNIT CHECK CALIRRATION TEST TEST LOGIC TEST IS REWIREO . l ) 24. Reactor Trip systes Interlocks (Continued) .i
- e. Power Range -
1, 2 f { Neutron Flux, P-10 N.A. R(4) M8)OS) N.A. N.A.
} '
- f. Turbine Impulse Chamber QUSI l l Pressure, P-13 N.A. R k8) N.A. N.A. I g)
$ h . Reactor Trip Breaker N.A. N.A. N.A. N(7 N.A. 1, 2, 3*, 48, 5*l Y
- C3 T 2a2% Automatic Trip and=#nteriesk N.A. N.A. N.A. . tL A. M(7) 1, 2, 3*, 4*, 5*
, U Logic -J
- 21. REACTOR TRIP 8YPASS MREAKEg M.A. d. A. #. A. M03),R04) NA. 1. 2. , 3 *,4*, 5
- Y e
5 : . 85 i - l Rn U siig . cs pts .- i
@ i -
M !-
~
3 .
P;ge No. 14 l 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. I NOTE TYPE NOTE # NOTES oc 3/4.3.1 PAGE: 3/4 3-14.0 INSTRUMENTATION - REACTOR TRIP SYSTEM INSTRUMENTATION DESIGN 1) Table 4.3-1 Notation 2. Revised per GL 83-28. DESIGN 2) Table 4.3-1 Notation 3. Revised per GL 83-28. DESI'GN 3) Table 4.3-1 Notation 6. Revised per GL 83-28. ED 4) Table 4.3-1 Notation 8. ' Analog Channel' deleted per GL 83-28. ED 5) Table 4.3-1 Notation (9). Last sentence (inclusion of Boron Dilution Alara Setpoint verification) deleted. Not applicable to STPEGS design. G O I l , l T l l O
i
.~. . ~ ~. r ~ ~ ' ~
. .. . C .' 7" . ..... .. ... T.~.
TABLE 4.3-1 (Continued) '. TA8LE NOTATIONS r Only if the Reactor Trip Systas breakers happen to be closed and the Control Rod Drive System is capable of rod withdrawal. l Selow P-6 (Intermediate Range Neutron Flux Interlock) Setpoint. l Selow P-10 (Low Setpoint Power Range Neutron Flux Interlock) Setpoint. * (1) If not p'erformed in previous 7 s. [ (2) --!: , _ . 2 c. ' x(- ';;;i"; t. we. . ,- ;c ". !..i!wAabove 15X of RATED THERMAL POWER. Adjust escorer channel vic.; cxistentwit ;;% ;.:c"; 9ewen if absolute difference is greater than 25. ". . , . . . ! - : .. . . i .
" _:1^!u ti=c-' & ^ h = & 14cakl t ==L.; "-i= =^^* 1 ;- 1.
(3) ""..; L ,.!;t : ,x!xc. ;C MPARf'fi
. A ncore to excore AXIAL FLUX OIFFERENCE above 15X of RATED THERMAL POWER. Reca11brate if the absolute difference is greater than or equal to 35. .h s u ic i_.. ;f 4pecificat" 'dA ;ce=aotapplicabledoc=ent;j ';'s"M " . "
(4) Neutron detectors may be excluded from CHANNEL CALIBRATION. s (5) Detector plateau curves shall be obtainer, and evaluated and compared to manufacturer's data. For the Interseufate Range and Power Range Neutron Flux channels the provisions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1. , e (6) Incore - Excore Calibration. .above=555:sf-AATEB= THERMAL-POWE4;:==ms: ' provisM -ef-Witication-2&2- re=not=epp14 cable,';. .c.; , "..^.
"" 2 gy..
(7) Each train shall be tasted at least every 62 days on a STAGGERED l TEST SASIS. (8) With power greater than or equal to the Interlock setpoint the required WIALOG=CHANNEtF 0PERATIONAL TEST shall consist of verifying that the interlock is in the required state by observing the permissive annun-ciator window. barmy (9) 4 tooth 4yAsurveillance in MODES 3". 4*, and 5" shall also include l verification that peruf ssives P-6 and P-10 are in their required i state for existing plant conditfons by observation of the pemissive l annunciator window. 7;_x?y . . [ i _ g _ ~-{ ..?___...}^.__..... ( __ __m_. ._-.__g ___y_.,_ , _ . . l l l ' NUREG 0452/STPEGS COMPARISON i W-575
.,,, 3/4 3-14 i
P;ge No. 15 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE # NOTES 1 00 3/4.3.1 PAGE: 3/4 3-15.0 - INSTRUMENTATION - REACTOR TRIP SYSTEM INSTRUMENTATION DESIGN 1)~ Table 4.3-1 Notations (15) and (16). Added per WCAP-10271 and associated SER. h O\ l . l l i i l l l l O
I
- v. ,.. T l-- - ..T.*
O L TA8LE 4.3-1 (Continued) . TA8LE NOTATIONS (Continued) (10) Setpoint verification is not applicable. - (11)- The TRIP ACTUATING DEVICE DaERATIONAL TEST shall independently verify the OP'ERABILITY of the undervoltage and shunt trip circuits for the Manual Reactor Trip Function. The test shall also verify the OPERABILITY of the Bypass Breaker trip circuit (s). (12)- The TRIP ACTUATING DEVICE OPERATI0flAL TEST shall independently verify the OPERABILITY of the undervn1tane and shunt trip attachments of the Reactor Trip Breakers. - (13) - Local manual shunt trip prior to placing breaker in service. (Or for plants that do not actuate the shunt trip attachment of the bypass breakers on a manual reactor trip): Remote manual undervoltage trip when breaker placed in service. (14) - Automatic undervoltage trip. (IS) Chen cHMMEL $ HALL SC MsTED EVERv 92 DAYS oH A STAGGERED. TT$r 5A515. O.\ QG)
- Comes.v worn rue Sv'avestinuce Requinemixrs or SercuricAnou 4.3.2.t rea AW PotmCM of rME CHMNEL MEQUInto ro .au OPERABLE av SPEcirscAneN 3.3.2.
.i t
o N M EG OO 2/SU EGS Jt ST5 3/4 3-15 COMPARISON
~
l ,
' i s -
7 Jr [_ P ge No. 15 e j - 06/13/85 _ COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS.
/
~ ' ' '
NOTE TYPE NOTE # NOTES _______________ ______ _____________________________ ,________.e_____,___ , 00 3/4.3.2 PAGE: 3/4 3-16.0 _ INSTRUMENTATION - ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INS j FONE. /
- 1) 3/4.3.2 No Changes.
+ ,
+
P \ 4 N m
/
vu
/
# S'
? , p 8
e f e
./
g h
/
o y e b% y l
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.9
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9 m. a
- - /
m A____ . m._M.
f I I ..
-~~..' . . . . . . . . - . . . . .
INSTRUMENTATION
. 3/4.3.2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMEN
- LIMITING CON 0! TION FOR OPERATION l 3.3.2 The Engineered safety Featuns Actuation system (ESFAS) .
Setpoints set consistent with the values shown in the Trip 5etpoint column of . Table 3.3-4 and with REspCMSE TIMES as shown in Table 3.3-5. ApPLICA81LITY: As shown in Table 3.3-3. E .
- a. With an ESFAS Instrumentation or Interlock Trip 5etpoint trip less c6nservative than the val ~ue shown in the Trip Setpoint column but more conservative than the value shown in the Allowable Value column of Table 3.3-4, adjust the Setpoint consistent with the Trip Setpoint value. .
- b. With an ESFAS Instrumentation or Interlock Trip setpoint less conserva-tive than f.he value shown in the Allowable Value column of Table 3.3-4, either:-
Adjust the Setpoint consistent with the Trip Setpoint value of
-( 1.
Table 3.3-4,'and determine within 12 hours that Equation 2.2-1 ( was satisfied for the affected channet, or
.5
- 2. Declare the channel inoperable and apply the applicable ACTION statement requirements of Table 3.3-3 until the channel is restored to 0'dAA8LE status with its Setpoint adjusted consistent with the Trip 5etpoint value. ,
Equation 2.2-1 I + R + 5 i TA Where: I = The value from Column I of Table 3.3-4 for the affected channel, R = The "as measured" value (in percent apan) of rack error for the affected channel, in percent syn) cf the sensor 5 =error, Either thevalue or the "asfrom seesured" Column value (5 (Sensor Error) - of Table 3.3-4 for the affected channel, and TA = The value from Column TA (Total Allowance) of Table 3.3-4 for the affacted channe1.
- c. With an ESFAS instrumentation channel or interlock inoperable, take O
the ACTION shown in Table 3.3-3. , NUREG 0452/STPEGS W-STS 3/4 3-16 COMPARISON w - . _
P2ge No. 17 i 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE # NOTES co 3/4.3.2 PAGE: 3/4 3-17.0 - INSTRUMENTATION,- ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INS t l DESIGN 1) 4.3.2.2 second sentence, change to read "... such , i that all trains are tested at least once per 54 i months ..." STPEGS has 3-Train logic. (3x18=54) l i f s e O e
- O
. . . . . . . . - . . ~ - -
. 'IN'STRUMENTATION
- t.
SURVEILLANCE REQUIREMENTS
,. 4.3.2.1 Each ESFAS instrumentation channel and interlock and tlin automatic f I
. actuation logic and relays shall be demonstrated OPERABLE b the ESFAS Instrumentation Surveillance Requirements specifi d n Tab e 4.3-2. .
l
!' 4.3.2.2 The ENGINEERED SAFETY FEATURES RESPONSE TIME of each ESFAS _ Attfunction
- i. ; shall be demonstrated to be within the limit at least once per 18 sonths.
!. Each test shall include at least one train such that b e kXtrains are tested at 84 least once perig months and one channel per function such that all channels are tested at least once per N times 18 months where N is the total number
.of redundant channels.in a specific ESFAS function as sr.own in the " Total
- No. of Channels" column of Table 3.3-3.
O
\.
( eg
~
- O .
NUREG 0452/STPEGS W _-STS 3/4 3-17 COMPARISON
P ge No. 18 06/13/85
, COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS.
NOTE TYPE NOTE # NOTES 00 3/4.3.2 PAGE: 3/4 3-18.0 INSTRUMENTATION - ENGINEERED SAFETY FEATURES ACTt1ATION SYSTEM INS DESIGN 1) Table 3.3-3 Item 1-Change title:" Safety Injection (Reactor Trip Feedwater Isolation, Control Room Emergency Ventilation, Start Standby Diesel Generators, Reactor Containment Cooling Fans and Essential Cooling Mater" to reflect STPEGS nomenclature
- 2) Table 3.3-3 Item lb - Break out into two separate line items, 1) Automatic Actuation Logic and 2)
Actuation Relays. Line Item 1 has 2 channels, 1 is required to trip and 2 are required to be operable. l ~ 2A) Item 2 has 3 channels, 2 channels to trip and 3 to be operable. STPEGS 3-Train design.
- 3) Item le - delete " Differential Pressure between Steam Lines High". STPEGS does not have this feature.
- 4) Item la - Add new item, " Compensated Steam Line Pressure Low". Change channels to trip to read ,
"2/ steam line in any steam line" and Minimum channels operable to read "2/ steam line in each steam line". Applicable modes are 1,2,&3.
4A) Action statement 15 applies STPEGS design.
- 5) Item if - Add new item " Compensated T_ cold - Low Low". Total channels "3/ loop", Channels to trip "2/ loop in each loop" and applicable modes are 1,2&3. Action statement 15 applies. STPEGS design.
- 6) Item la - Change Action statement from "19" to "18". Editorial.
. 7) Item ld - Change Action statement from "20 to "19". Editorial.
l O
- n. .#
i TABLE 3.3-3 iy '
- I ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUENTATION h * ..,
MINIMUM CHANNELS APPLICABLE TOTAL NO. CMANNELS 0F CHANNELS TO TRIP OPERA 8tE M00ES ACTION ',; ! FUNCTIONAL UNIT' i l 1 Safety injection (Reactor * ! E* 6 y Trip, Feessater Isolation. ; sy I N g Control RooellsW4stime! St .. ) bleselGeneratorsg,ontainment FEAcTOR
- i Cooling Fans, and Essential Coetswa h Water). '
/8 :
- a. . Manual Initiation 2 1 2 1,2,3,4,, Jd f 1,2,3,4 14 !
A b. Automatic Actualon lossc 2 . 1 2 1 taste-and Actu.t.1on RnAYS 3 2, 3 j , 2, 3,4 l4 ' j *;* * , liil Atl#35 .
- c. Containment 3 2 2 1,2,3 158 !
); Pressure--Nigh-t
- d. Pressurizer 4
- 2 3 1, 2, 3# # i j Pressure--Low
- l-1
= -aitferential s-*=mer .j:
Reassura-setussa I, , j Steamati _ - ::!:' :l ) :ty_.- L _ " . Maat 4 - F 3/ steam line 2/ steam ling 2/ steam linefM J,2,3 158 ., l sw any stsas line Eacn Srtw L NE .. ., .'
- , e*n$4,vlPEMSAWD pgsg.g Operatin!r- St#1 put
- s.oept Lete 2 LwP I4 LecP I4 9 4 -
; oa I,2,3 g(/S l*
2/epsta44mg l l g pgp ps 6.g TILN E DRqts MED 3/eperst$tg s h anyLeoP 4 SM EAO4 Loop y l. e-apSN tiMg 1, h D steasiltRe ,
. 8j I
i- - a Q: : L
l P;ge No. 19 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE # NOTES , co 3/4.3.2 PAGE: 3/4 3-19.0
- INSTRUMENTATION - ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INS DESIGN I
- 1) Table 3.3-3 Item le 2) " Differential Pressure Between Steam Lines - High. Three Loop Plant.
Deleted. Not applicable to STPEGS four loop plant.
- 2) Table 3.3-3 Item if 1) " Steam Line Pressure - Low, l Four Loop Plant". Deleted. STPEGS does not have this feature.
l e -
. IID-
.e 1
O I t
O "O O TA8LE 3.3-3 (Continued)
'T ENGINEEREDSAFETYFEATURESACTUAT10NSYSTENINSTRUNINTATION 3 -
- NININLM -
~ ' '
ELS* CHANN LICAB N00E ACTI FUNC ONAL IT F LS T TRIP OPER LE
, s.
Saf y Inj tion (Re ter . Tr p, Fee ater Iso ation - ntrol oom Iso' lon, art. - Diesel nerator con innen
- Cool g Fans, u tss ntial Ser ce Water (C..n nJed) ,
Q 2) Thre Loop P nt .
- 158 w T ee too 3/ steam ine 2/ eam line' 2/ steam ine A eratis ice and 1/ '
team line . Two oops 3/ erating 28**/st 2/ rating 1 0 rating , eam line 11ne t ice eam line . 3 in a her '
. ope jing
. st am,line 1, 3**
- f. team Lin Pressur--Lbw ,
- 1) F ur Loop lant 1 pressu 1 pr ssure 1 ressure ,15*
E Four to ps *
- Opera ng loop an 2 loops y ,3 loops g@
Thr e Loops 'l pre sure/ ** pressure 1 pressur in lg 0 rating oper ting n any oper- .in any 2 gg :1 ating loop operati loops r; ;3 gg p"3
=c
. < . p g ,
W
***** [
Page No. 20 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE # NOTES 00 3/4.3.2 PAGE: 3/4 3-20.0 . INSTRUMENTATION - ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INS
- 1) Table 3.3-3 Delete Item 1.2) "Three Loop Plant".
N/A to STPEGS. DESIGN 2) Table 3.3-3 Item 2a " Manual Initiation" - change to read - Total number of channels " 2 pair", channels to trip to "1 pair with 2 switches operated simultaneously," as minimum channels i operable to "2 pair". STPEGS design. ED 2A) Change Action statement to read "18" to reflect. Editorial.
*' Table 3.3-3 Item 2b - Break out into two separate 3) line items, 1) Automatic Actuation Logic and 2)
Actuation Relays. Line Item 1 has 2 channels, 1 is required to trip and 2 are required to be . operable. 3A) Line Item 2 has 3 channels, 2 channels to trip, 3 to be operable, applicable to Modes 1,2,3&4 and Action 14 applies. This reflects the STPEGS 3-Train design. . O O
- u --
/
O . I TABLE 3.3-3 (Continued) .
. *1 .
l 3 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUNENTATION 4 i n NINun TOTAL No. CHANNELS CHANNELS APPUCABLE OF CHANNELS TO TRIP _ OPERA 8LE M00ES ACTION . l gucTianAtuNIT . 3 $ - - a a a am a
.. . . .,? ..-.. . , t = ; = = .: = , . 2. 2 - .,
e.____, . *
. e j ___ -__ ____-_ ,
w -...- w. _ -- - --- ,...g ----3 l
;;;_ _::: =- -- - - -- - .- r k
.; ;--. ;; 1 -. - - - - ,
j
- n. . .. ,i . , ......." 7.--.-
I:99=-- e W - i _ l . < ~ ^ y.--.--- K L_ __ ^- . . - 7.---- - .
, . 7 -
- 8 r~'""""W
, - - . - - . . . , -- r __',__3,r-" "."*3 9,
p - t O Containment Spray i 2. f a. Manual Initiation 2 PAlft 1 ith 2 SLR 1,2,3,4 klf t 2h ' switches ceGRAWD SIMMANEm*5LY 1,2,3,4 14
~
- b. Automatic Actuation f.o61c 2 1 2 g A 3 /, 2, ~3, e/ /1 togtumund Actuation RraAys 3 SEREM l
containment Pressure-- 4 2 3 1,2,3 17
- c.
High-3 + , 1 ! 1 5 ' 'OM00 l rd o O N. O
, - ~ l C/3 %
om , z:3M , a .- 1 M i l
Pzga No. 21 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE # NOTES oc 3/4.3.2 PAGE: 3/4 3-21.0 INSTRUMENTATION - ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INS
- 1) Table 3.3-3 Item 3.a.2) Break out into two separate line items, 1) Automatic Actuation Logic and 2) Actuation Relays. Line Item 1 has 2 channels, 1 is required to trip and 2 are reqcired to be operable.
DESIGN 1A) Line Item 2 has 3 channels, 2 channels to trip, 3 to be operable, Modes 1,2,3&4 are applicable and Action 14 applies. This reflects STPEGS 3-Train design.
- 2) Table 3.3-3 Item 3b " Phase 'B' Isolation" -
,. delete. STPEGS does not have this feature.
- 3) Table 3.3-3 Item 3c " Purge and Exhaust Isolation"
- Change title to " Containment Ventilation Isolation" and renumber as Item 3b. STPEGS nomenclature.
DESIGN 4) Table 3.3-3 Item 3b 1) " Manual Initiation" - Delete number of channels etc and add note "See Manual Containment Spray and Manual Phase 'A' ' Isolation above". Containment ventilation is manually actuated in Phase A or CSS.
- 5) Table 3.3-3 Item 3b 2) Break out into two separate line items,1) Automatic Actuation Logic and 2)
Actuation Relays. Line Item 1 has 2 channels, 1 l is required to trip and 2 are required to be I operable. 5A) Line Item 2 has 3 channels, 2 channels to trip and 3 to be operable. Both line items - Applicable Modes "1,2,3,4&6##" and Action "17" to reflect STPEGS 3-Train design.
- 6) Table 3.3-3 Item 3.b.4) - Add new item "RCB Purge Radioactivity - High". Total number of channels "2", channels to Trip "1", Minimum channels operable "2", applicable Modes"6##" and Action statement "17". This reflects STPEGS design.
O
O t TA8tE 3.3-3 (Continued) #- 9 h ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION . . I. MINIMUM ;
; TOTAL NO. CHANNELS CHANNELS APPLICABLE i OF CHANNELS TO TRIP _ OPERABLE MDDES ACTION .
l j FUNCTIONAL UNIT " s 4 l 1 l 3. Containment Isolation t (
- a. Phase "A" Isolation -
t i Manual Initiation 2 1 2 1.2,3,4 4918
- 1) ,
l 2 1,2,3,4 14
- 2) Automatic ActuationLme 2 1 i Logic and Actuation striavs 3 2, 3 1. 2, 3,4 14 [
} . j Relays l SeeItem1.aboveforallSafetyInjectioninitiatingfunctionsandl ; j u 3) Safety Injection
) requirements.
j Y m _. ...___ _ _ 1-
._-._.. 6
. . . - - ..,$$$at$;G 1 5Y& -- -~ b--- - --
3-b,be k "$ . bh N .:
=***== :
l 1
~-2
~-
1, . 3, i n j 21 2 l ^ --!! _ .^_-jr-} j- = s_,._ _.. . . . _ . . ) 21 r _.n:: ..; i- -- 2 - - - - -
-1, :, :. 17 l
l ....._......-
^
- b. e. Co,gnWMDITJparitATNW g._. ,_ _ - _ __ -
35*3*ti*" See won C=rnwwnr Staav ano humat Pasr"i Isournov esove. ; l 1, 2, 3 i 15
- 1) Manual Initiation g 2 1 2 ,
- l -
E 2) Automatic Actuationlase 2 1 2 1, 2, 3, 4,6 * .11F l7 nM b;!- _ " Actuation itoAf5 3 2.
- 3 1,2,3.4,6 17 g
a Surtays c: . h za
- 3) Safety Injection #
See Item 1. above for all Safety Injection initiating functions and requirements. 6 'l7 h [:
- 4) Rc8 PWR6E -2, -
1 I i, m
~
RADsoAcTsVfTY- HsGH , i i ) -
T P ge No. 22 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE # NOTES oc 3/4.3.2 PAGE: 3/4 3-22.0 . INSTRUMENTATION - ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INS DESIGN 1) Table 3.3-3 Item 4.a.1) - Change total number of channels to "2/ steam line", Minimum channels operable to "2/ steam line" . This reflects STPEGS design. Action statement change to "23" is editorial.
- 2) Table 3.3-3 Item 4.a.2) - Change Action to "22".
Reflects deletion and renumbering of Action statements.
- 3) Table 3.3-3 Item 4b - Break out into two separate line items, 1) Automatic Actuation Logic and 2)
Actuation Relays. Line Item 1 has 2 channels, 1 is required to trip and 2 are required to be operable. DESIGN 3A) Line Item 2 has 3 channels, 2 channels to trip and 3 to be operable. Both line items applicable Modes "1,2,3" to reflect STPEGS 3-Train design. Action statement change to "21" is editorial. ! 4) Table 3.3-3 Item 4c " Containment Pressure High e j -2 " replaced with" Steam Pressure - Rate high - 2. Total number of channels "3/ steam line", channels operable "3/ steam line", applicable Mode "3###" and Action "19*". 4A) This reflects STPEGS 3-Train design and applicable mode applies only in Mode 3 below P-11. l 5) Table 3.3-3 Item 4.d " Steam Flow in Two Steam l Lines - High" Deleted. STPEGS does not have this I feature.
- 6) Table 3.3-3 Item 4.d - New Item added " Safety Injection" with a statement to see item 1 for all Safety Injection Functions. STPEGS has Steam Line Isolation upon Safety Injection.
l
. TABLE 3.3-3 (Continued) '
ly ; ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTIMENTATION h E MINIMUM r CHANNELS CHANNELS APPLICABLE - TOTAL NO. OPERABLE N00ES ACTION j. OF CHANNELS TO TRIP ' FUNCTIONAL UNIT l
- 4. Steam Line Isolation .
- a. Manual Initiation I
/ operating .1, 2, 3 :St 2 3 .
- 1) Individual A'1/ steam Ilne 1/ steam line steam line tyE2, 2 1 2 1,2,3
- 2) System 1 2 1,2,3 2221 .
- b. Automatic Actuation Losic 2 f , 2, 3 3 2. 3 21 i s _. Actuation En*Ys agg #
Selays - STrw 2/57Em var f3/STE4*l W. 3 15819 '.'
- c. ContainmentAP ressure- RATE , 3/sfEAa4 uve LIM -
} ANY SITAnt Und !
Svery INucTiod kse High-2 SEE ITEM l ASoVE roR ALL SAFETY lamcnad latITIATim JWCTIMS Ape REdi/ff(EMENT.S.
- d. 2 : ,
M
't i !Mi= != ,i--f ,
.._m... . . . .
-p . - , _ . - _
- 1 /s t: = i i .- n-1m i_ ,- _ 2/staam:ti= 1/ t =, i * .-
^ ^ ~;==
, _. -_!. 2 4tM5F
' s
-@ l
!-s t
_ _ _ _ . - - ' ~ -
-Efew -- ". _ _ , _ Of r _" I .- - 'u_ #b 3 W g g4_ -;, _
p g_
- staan=t4aw E, .. .. . .. . .
-, "'""" -- r " ' ' ' -
[T3 17-- 2 /ste===t i ne==1/st.an=H-1/ste===h b ee 5hr -t _ , -
-:, _ . &-- -.., 2 &-
O b mi- L ,- 2/ operating laaa/any- r/aperatii ,. up
$i -
= prati--- =t.am-ta= -. .c.;-
^
= _ us -
+L = T iw
Page No. 23 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE # NOTES cc 3/4.3.2 PAGE: 3/4 3-23.0 . INSTRUMENTATION - ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INS
- 1) Table 3.3-3 Item 4d " Steam Line Isolation - Steam Flow in Two Steam Lines - High. Coincident with T(avg) Low-Low" Deleted. Not applicable to STPEGS design.
s -
. G l
.I l
l O
/
b . .. . . .. . . g TABLE 3.3-3 (Contineed) , 39 i 3- ENGINEERED SAFETY FEATURES ACTUATION SYSTEM ill5TRINGENTATIoll b l - ( I; j [ FullCJ1000AL IT TAL OF CHAleffLS CHAINkLS T0ARIp OpfdABLE L5 APPL M60E5 ACT10d - 3 1 j Ste Line I atten ti ) , 7 'I j 5 eam F1 in Two L s--N . stact t With: .
.g
--L Low . 1, 3 : ,
,,g I
- 1) F Plant .
.I i u 1T ,'
j 1 Four s 1 ,,/1 asty 1 T,,, '
.i
! *;' @' I"8 - 2 s 3 leeps
. g U L s 1T r- *** T in 1 T,y / in 16 l i . T
'** ating loop any opefating tdoper ing ,
' loop oops ! Th Loop ant . 1 l hree L s 1 T ,, oop T,,,any 1 T ,, g M '**I . loops 2 oops 16
- Two s 1 av0
/oper- 1*
- T , T ,,in .
l "8 ing loop in oper- operati loop . I g . i ng atiploop ,,
; i j
' ga >2
. * - i
] ?Y C3 ! 83 . ! *4 . l E I _ _ . - _ _ _ _ _ _ _ _ _ _ _ - - _ -
Pzga No. 24 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE i NOTES oc 3/4.3.2 PAGE: 3/4 3-24.0 INSTRUMENTATION - ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INS
- 1) Table 3.3-3 Item 4e " Steam Line Isolation - Steam Line Pressure-Low" Deleted. Not applicable to STPEGS design.
- 2) Table 3.3-3 Item 4f1) " Steam Line Isolation -
Steam Line Pressure-Negative Rate-High" Deleted. Not applicable to STPEGS design. e O
.I e
O
TABLE 3.3-3 (Continued)
,9 ENGINEERES SAFETY FEATURES ACTUATION SYSTEM INSTRUENTATION f h .
(
' INIDAM CHANNEL 1 CABLE i
/ TOTAL . [CHANNEh N00ES ACT! .;
- FUNCTI0dALUNI[
0F C LS/ 10 TRfP _ OPERA 8 _ (team bIsola on (
- 4. laued) j
} ,
- e. team Li press e--Law 1. 2. 38"
) i - i 1) sur L Plant ressure 1p ssure 1 pres are 158 Four[ oops I ani 2 loops any 3'Ioops ) Oper6 ting foop / *** pressure 1pbssurein 16 l ! Three L 1 pressure / 3 Operati oper4 ting in any r- any2oper-/-
/ loop att hting 1 s l ; o
< s
- Three/ Loop p ant I ,
1 1 / l Three Loop / 1 pressure 15* y s 1 pres ure/ pressure ' / operating loop / acy 2 Igop any loops Tuo Loops
! 1 ressure/ 1*** pressure 1 pressure 16 l -
- in any/oper- any oper-op,efating rating j
loop ating loop ating loop -
- f. Steam Line P ssure - tive 3* ** /
I l Rate--High /
) Fou Loop P1 t
,l l
I F rLoop/ 3/ steam line Ine 2/ eam line 15 - l rating' 2/ste!as any st ii-
- s w
a - 3/ ating 2 team line 2/ steam line 16 . ! $E Thr Loops
*in each/oper-
'stpilne n any oper-go rating ting steam ating team g l gg ""* IIA
- Z i!
i $2 > d M- ' l E .
- .I
Pzga No. 25 06/13/05 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE # NOTES 00 3/4.3.2 PAGE: 3/4 3-25.0 - INSTRUMENTATION - ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INS
- 1) Table 3.3-3 Item 4f.2) " Steam Line Isolation - 3 Loop Plant" Delete. STPEGS is a 4 loop plant -
this item is not applicable.
- 2) Table 3.3-3 Item'5a. Break out into two separate line items, 1) Automatic Actuation Logic and 2)
Actuation Relays. Line Item I has 2 channels, 1 is required to trip and 2 are required to operable. DESIGN 2A) Line Item 2 has 3 channels, 2 channels to trip and
, 3 to be operable. Both line items: applicable Modes "1,2,3". The reflects the STPEGS 3-Train design. Action statement change to "24" is editorial.
DESIGN 3) Table 3.3-3 Item 6.a. " Auxiliary Feedwater - , Manual Initiation". Column values should read "1/ pump", "l/ pump", "1/ pump", Modes "1,2,3". This reflects STPEGS 3-Train design. The change to Action '22' is editorial.
- 4) Table 3.3-3 Item 6.b. Break out into two separate line item, 1) Automatic Actuation Logic and 2)
Actuation Relays. Line Item has 2 channels, 1 is required to trip and 2 are required to be , operable. 4A) Line Item 2 has 3 chtnnels, 2 channels to trip and 3 to be operable. Both line items: Applicable Modes "1,2,3". This reflects STPEGS 3-Train design. Change to Action '21' is editorial. DESIGN 5) Table 3.3-3 Item 6.c.1) " Start Motor Driven Pumps"
- Delete. STPEGS AFW actuation on Low-Low Steam i Generator Level does not differentiate between motor and turbina driven pump starts, therefore
( , this line item is not required. ED 5A) Table 3.3-3 Item 6.c.1 Change to Action '19' is editorial. O
- . , . : - . . . . ,. *:. e l. * * . ;.~'.' .
i t TABLE 3.3-3 (Continued) . _
,9 l h ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRWENTITION * '
'l
' MINIMUM CHANNELS CHANNELS APPLICABLE l TOTAL No. MODES _ ACTION ! OPERABLE OF CHANNELS TO TRIP _ FUNCTIONAL UNIT
- ,. . . _ _ _m__
n; = ;-- = ::;;-. _-..-- -; E r-- -- E , 21 =1 l 2/ steam ii . 18: L -l ,,.- 3/mi aur-fine-2/ des iin.
- Operatheg-
# 6 1"
' 'I 3/operati- 2/ steam:lin 1/staan=11r.-
4es-t_ - !;; w;;.5
" d;;= -li;;; ! ;; _. ., , . _r_.
_.4! g -tiny-. h
.e g t= =
. =*bw h .
2 Turbine Trip and
. 5.
- e. .
4 Feedwater Isolation , u 1 2 1, 2,3 25 24 .l
- a. Automatic Actuation Logic 2 24
-togic=asuk Actuation RaaYs 3 2, 3 1, 2;,, 3
*elays-4/sta. gen. 2/sta. gen. 3/sta. gen. 1, 2 10*19 * 'i
- b. Steae Generator in each in any oper-Water Level-- operating j
ating sta. gen. High-High (P-14) sta. gen. , y 1NSERT 3/4 3-25
) 6. Auxiliary Feedwater 23 2,2, E f/PvMP 1/ PUMP f1/pdMP 1, 2, 3 l J
g a. Manual Initiation 2 1,2,3 2F21 . nN b. Automatic Actuation Logic 2 1 21 .l 4 3 f, 2, 3 .
^;
g, and Actuation Relays 3 ' i c. Sta. Gen. Water Level-- L a e a a
,$g$; tow-tow @
! p ,
! 4 %=Sta G L 4/sta. gen. . 2/sta.' gen.
3
'3/sta. gen. 1, 2, 3 . goal 9 l 8 sciven-F r - in any oper- in each .
1
" i:
ating sim. gen. operating sim. gen. j l. I*
T P ge No. 26 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE i NOTES oc 3/4.3.2 PAGE: 3/4 3-25.5 . INSTRUMENTATION - ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INS
- 1) Table 3.3-3 Item Sc. Add new item " Low Compensated T_ Cold". Column values should read "3/ loop", "2/SG in any steam generator", "2/SG in each steam generator", Modes "1ffff,2,3" and Action "15*". This reflects STPEGS 3-Train design.
- 2) Table 3.3-3 Item 5d. Add new item "Feedwater Flow-High coincident with either following in 2 of 4 loops". Column values should read "3/ loop",
"2/SG in any steam generator", "2/SG in each steam generator", Modes "19###,2,3" and Action "15*".
2A) This reflects STPEGS 3-Train design.
- 3) Table 3.3-3 Item 5.d). Add new Item " Reactor Coolant Flow-Low". Column values should read "3/ loop", "2/ loop in any loop" and "2/ loop in each loop". There are no values for Applicable Modes and Action. This reflects the STPEGS 3-Train design.
- 4) Table 3.3-3 Item 5.d). Add new Item "T_ avg. -
Low". Column values should read "1/ loop", "1/ loop in any 2 loops" and "1/ loop". This reflects the STPEGs 3-Train design.
- 5) Table 3.3-3 Item 5.e. Add new Item " Safety Injection"". There are no column values, add Note "See Item 1 above for all Safety Injection initiating functions and requirements. This is consistent with previous treatment of Safety Injection line items DESIGN 6) Table 3.3-3 Item 5.f. Add new Item 'T-avg - Low Coincident with Reactor Trip ** (P-4)'. Column values are '4','2','3', Modes '1,2,3' and Action
'19*'. This reflects STPEGS 3-Train design.
O
1 O 0 _ 0 1NEZ 3.3-3 (Qintimm0 acne 5ED SAFEIT PER11ME ACfutrim ST511M DEmtM9ftATIN , MDtDtM 10D010. OWeEIS OWNFIS APPLICAME OF O WeEIS 101 RIP OFUWEE IGES ACTIN MICTIGeht,INIT
- 5. ItMBI!Et1 RIP NO 755R11!K 150UU10t (CCW7:)
WR b (em W3 SG 2/hnp. in any SG 2/3mer. in each INN 2. 3 IS* 3/ loop L --I c. Im ce T ""M assysrsaai 6rv. sma Gr#. g
- d. Feekster Flow - High 3/ loop 2/ . in 2/h. in each INN. 2. 3 15*
coincident with either asspSTawnM lampstr e 6Df-followirg in 2 of 4 loops: ammeter coolane Flow - tm 3/Icop 2/ in my 2/ loop,ineach 1 cop LOOP Or 1/ loop 1/ loop. in my 1/ loop 5 T "E
-tm 2 joops nM oo $ to SofttyInbtCtion SEE .DEM l fBOVE FM ALL SartrYZvxcited intArsw Racnots AND REQUIREntENTS. = + $2 f. T -tm coincident with 4 2 3 1.2.3 19* *O a:Ecoe Trir* e-4) - .
g - m
P;ga No. 27 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE i NOTES co 3/4.3.2 PAGE: 3/4 3-26.0 INSTRUMENTATION - ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INS
- 1) Table 3.3-3 Item 6.c.2 Delete. Explanation given in 11) on page 3/4 3-25.
- 2) Table 3.3-3 Items 6d, 6f, 6g and 6h. Delete as they are not applicable to STPEGS design.
- 3) Table 3.3-3 Items 6e. Delete " Start Motor Driven Pumps". STPEGS does not differentiate between motor and turbine driven pumps. Leave words
" Safety Injection and renumber as 6.d.
3A) Add note in place of column values to "See Item 1 above for all Safety Injection initiating functions and requirements.
- 4) Table 3.3-3 Item 7.a. Break out into two separate line items, 1) Automatic Actuation Logic and 2)
Actuation Relays. Line Item I has 2 channels, 1 is required to be trip and 2 are required to be operable. - 4A) Line Item 2 has 3 channels, 2 channels to trip and 3 to be operable. Both line items: Applicable.* Modes "1,2,3,4" and Action "14." The above reflects the STPEGS 3-Train design. I t l l \ l l O l
\ -
( I TABLE 3.3-3 (Continised) ;
,9 ElmiiMEEREO SAFETY FEATURES ACTUATICII SYSTEle IIISTALDENTATICII ) 3 I i IIIIIIIEM j TOTAL 110. CHAlelELS CHANIIELS APPLICABLE OF CHAlelELS TO TRIP - OPERABLE N00ES ACTIO11 l I
FUIICTICIIAL tallT
- 1
- 6. Assaillary Fee &sater (Continued) ,
l -;__--.
=.--=..-
..=.. . . . . .
{ 1
- I e=enderweteugspeeP= ..
-'BenME=9enstar
- l
.= .=.= 3 ,. .
-....- .= .-
! t:* le safety Injection , , .
[See item 1. above for all Safety Injection initiating fianctions and - y - - requirements. -
) M - ,
m ., .... ., ........ ..... l $turt=stetsrertsuus=- l _,_,e -l ,,. ..,si i.ii 5_:_ $l
! ;[ ~
-start =sestes===- .
i
;...-.._... s. . , , . . , , . . , , . ., .
.i _- _e . , . , _ .. : E 7. Automatic Swltchever ta
'i
,o i Contalament Sump
,8 ig "}T27 % .
1 2 1.2.3.4 14 [ t +. Automatic Actisation tosec 2
> t.
$$ heeW Actuation cruys 3 2 's i, 2, 3, 4 14 Q l t
82 zs
'#*48F5 . - I
*o '.
m . O -~ . j' i . i
P;ge No. 28 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE i NOTES co 3/4.3.2 PAGE: 3/4 3-27.0 INSTRUMENTATION - ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INS
- 1) Table 3.3-3 Item 7b - Delete " Consistent with Sump Level-High". Not applicable to STPEGS.
ED 2) Table 3.3-3 Item 8a change to "4.16kv ESF Bus Undervoltage - Loss of Voltage", Item 8b to "4.16kv ESF Bus Undervoltage - Tolerable Degraded Voltage plus S.I. or Sustained Degraded Voltage". STPEGS nomenclature. Change to Action '19' is editorial. DEDIGN 3) Table 3.3-3 Item 9d " Steam Generator Water Level P-14" Deleted. Adi new Item 9d " Power Range
.- Neutron Flux Input to Excessive Cooldown, P-15".
This change reflects STPEGS design.
. (Il l
O i l L
- . ....- : . . - . ... *'. ' ' :s %: ... 2 . 2 :U
. o f *:.Ti.' - . -
TABLE 3.3-3 (Continued) l 'T ' ENGillEERED SAFETY FEATURES ACTUATicN SYSTEM INSTRt2ENTATION ) 3 . i MINitt41 { CHAlelELS CHANNELS APPLICABLE TOTAL hD. ACTION ; l OPERA 8tE N00ES OF CHAlstFLS TO TRIP _ l FisiCTIO11AL talIT I 2 7. Automatic Switchover to !' j Containment Sump (Continued) i 4 2 3 1,2,3,4 17 h [ b. RWST Level-Lew-Lew ' !, . Coincident Wit M 2 & -1, ^ . ^. , - 17- [ 4ents! l: . L. _I - C. 4 . l . See Item 1. above for all Safety Injection initiating functions t* Safety Injection l and requirements.
,,, r.
N 8. Less of Power g ]
- E5F
- 2/ bus -
3/ bus 1, 2, 3, 4 20*19 , I a.4r64 LV3 5us Gn.-der- . 4/ bus l af i y voltage-LossofVoltage s l I-b.4.864 kh Undervoltage-Tacnastr 1, 2, 3, 4 -20*.19 I :it [ aded Voltage 4/ bus 2/ bus 3/ bus * ' Gs44 - pt.us es sostwco or64actD VsTM,E I
- 9. Engineered fety Features -
- i e
Actuation System Interlocks
- 2 2 1,2,3 2T ZO
- a. Pressurizer Pressure, 3 .
P-11 ,l 2 3 1, 2, 3 11 20
- b. 4 l Low-Low Tavg, P-12 232,Z +i 2 2 1,2,3
! c. Reacter Trip, P-4 2 ... i 5t:s ^ . _ . .WWeter3/ ster-gen.--- -2/sta.=,.... 2/sta=, ... 17 2r- 2' E :d. inw+ .
-tu- .7-
- n ai tuve m *- ,_..iing= [i l @ ','
w tine - st = - ... . i.. ,_ .. c-g 2, 3 I,2,3 20 1 y5 &. Powen Raw,e Nwrnet A- , z:;l Vs.us sNrer To EXCfsSNC . E CosDown PROTEcTal, P-IS -
. j .-
P g3 NO. 29 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE # NOTES co 3/4.3.2 PAGE: 3/4 3-27.5 INSTRUMENTATION - ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INS
- 1) Table 3.3-3 Item 10 " Control Room Ventilation" added including Item 10.a " Manual Initiation",
Item 10b " Safety Injection", Item 10c " Automatic Actuation Logic and Actuation Relays" and Item 10d
" Control Room Make-up Air Radioactivity-High".
lA) Values provided to reflect STPEGS design.
- 2) Table 3.3-3, Item 11 "FHB HVAC" added including Item lla " Manual Initiation", Item 11b " Automatic Actuation Logic and Actuation Relays (BOP)", Item 11c " Safety Injection" and lid " Spent Fuel Pool Exhaust Radioactivity-High".
2A) value provided to reflect STPEGS design. ED 3) Table 3.3-3 Item 9 Change to Action '20' and '22'
- is editorial.
9 O
. .,,/ J )
j . . 1as h 3.3-3 P " -i . aconsen saperr numse acustrum semn sumammarum - I
\ unna ,
anuanz 4
- 1 pus.se. ossees ausses n e c n a nt. u a t er oneses in inar creustz uses acrum
]
~
i ! 10.-S. anner.unsinurttArIGE , l
- s. Ma mal intesatsen 3(1/ train) 2(t/trata) 3(t/ train) All 24
$gr Tste I name fee nst Saarry Tauressoar surntssind featinous Akk REearsRGMENTS.
- h. Safety 1sdectlen A -
W c. A* -=ric Areneh Losac 72 KI It All 24 2+ z h p W Acen=tlan REtAtS (SOP) 5 A 3 All-
- W ~
- l A d. Omtrol naamM Air 2 I . 2 All 24 .
j y a ds ctivity - nuk
- 11. nm unc
- a. neemi Initiatsen 3(t/ train) 2(t/trase) 3(t/trase) 1, 2, 3, 4, p42 with irrwe==a 2
fieri se spent
!ji fuel pool , .
j 8E i; is o b. u -** Actus. san 3 2 3 I, 2, 3, 4, e_
# 24 -
l .u -- wit r ! g" Halays (IDF) fumi sn ayant
- l. 8M f=1 rool
\ 3 . j S c. Safety Injection Sgg irsn i saan Font Als. SAvtry InstTsoar 1411*sATIM FodcTsavs suo Weaggtserars,
- d. Spent Ruel Fool Edumet 2 1 2 with irr=as.ted p 24 l a ds ettwicy - ass > fuel s y t
P g3 N3. 30 ' - 06/13/85 COMPARISON OF NUREG 04$2, REV.5, AND STPEGS 'IECH. SPECS. NOTE TYPE NOTE f ' NOTES
~
co 3/4.3.2 . PAGE: 3/4 3-28.0 INSTRUMENTATION - ENGINEERED
, SAFETY FEATURES 7+CTUATION SYSTEM INS 1}' Table 3.3-3 Table Notations. **,'***, ****
Deleted. Raplaced with it, fif, #fff, and new **,
***, ****, based on STPEGS design and
, modifications to Table 3.3-3.
2). Table 3.3-3 Table Notations. Insert for page 3/4 3 :-28 .
- 2) Table 3.3-3 Table Notations. Action 16 Deleted since it is not applicable to STPEGS design.
ED 3) Table 3.3-3 Table Notations. Action "17 and 18"
~ renumbered 16.
"16 and 17" due to' deletion of Action 9
0
.I A'
E I t 1 O
. . .e. =. . . . . .
9..- . TABLE 3.3-3 (Continued) TABLE NOTATIONS ; p*TheprovisionsofSpecification3.0.4arenotapplicab1's. ..
> # Trip function say be blocked in this MODE below the P-11 (Pressurizer Pressure Interigck) Setpoint.
- I
= - ; --_ a = - - -= +r-e:-es;=s
;1= e; -u eu J
% l qngspquuig 535p5HE . .
- , . . . . . . . . . . . . _ _ _ . ___. ... .- r - -- -- r;---- - . ; .rr-- - - - - .
. . p. i......... ......_-..-..g ..w.-.,- ___.. . -- ....-g __ . . . . _ _ _ _ _
r . . . . . . . . . . , . . . . . . . . . . . . . . . . . _ _ _ ..7.._._.. I NSERT 3/4- 3-28 --+ .
. ACTION STATEMENTS ACTION 14 - With the number of OPERA 8LE channels one less than the Minimum a
, Channels 0PERA8LE requirement, be in at least HOT STAN08Y within 6 hours and in COLD SHUTDOWN within the following g- ,
- 30 hours; however, one channel may bh bypassed for up to 2 hours for surveillance testing per Specification 4.3.2.1, provided the other channel is 0PERA8LE.
ACTION 15 - With the number of OPERA 8LE channels one less than the Total Number of Channels, operation say proceed until performance of the next required ANALOG CHANNEL OPERATIONAL TEST provided the inoperable channel is placed in the tripped condition within , I hour. W.www ..w.. 5
. ... i. . p. w . .W...g .w.p ..7i.. . 5.g FV..W.i. .. F. .w.. ....w.=.ww .
- ;;: ._. . . . _^ h 0.= i .=,.- 5.5 . . . . . .. . . . .
!. _ _ . . . . M " - 2 ' _ _ _ ^, : 27 07 " : 2 ^' _: .. . " .... . . , . - . . . . . . .
....,.......... .. ..._.....................i,........ . . . .
"JP'***** **'
-_.._.--._.____.____.-....7_. .. '- r "'* J ==
;".........'T: . .,, ,. . -, .. ; . ; . . . . . . . J . ^ . 2. ^a L: :. . . . . " . . ... .
ACTION - With the number of OfftASLE caannels one less tnan tne Total
. Number of Channels, operation may proceed provided the inoperable channel is placed in the bypassed condition and the Minimum Channels OPERA 8LE requirement is set. One additional channel say be bypassed for up to 2 hours for surveillance testing per Specification 4.3.2.1. .
IT ACTION 21 - With less than the Minimum Channels OPERABLE requirement, operation say continue provided the containment purge supply . and exhaust valves are maintained closed.
. NUREG 0452/STPEGS COMPARISON M-STS 3/4 3-28
O .
**Feedwater Isolation only.
*** Function is actuated by either actuation train A or actuation train B. -
, Actuation train C is not used for this function.
**** Automatic switchover to Containment sump is accomplished for each train using the corresponding RWST level transmitter.
hDuringCOREALTERATIONSormovementofirradiatedfuelwithinContainment.
- fifWhen below the F-11 (Pressuriser Pressure Interlock) setpoint.
s ifffTrip function is blocked in MODE 1 above the F-15 (excessive cooldown protection) setpoint.
. e, O !
3' "uate o432/srercs
/#5EF 3/4 coxgiarsos .
P;ge No. 31 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE i NOTES cc 3/4.3.2 PAGE: 3/4 3-29.0 - INSTRUMENTATION - ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INS ED 1) Table 3.3-3 Table Notations. Action "19 and 24" renumbered "18 and 23" due to deletion of Action 16. 2 ).. Table 3.3-3 Table Notations. Insert new Action 24 which identifies the action to be taken for the STPEGS FHB HVAC. 1 O 4
)
9 e e O
. . . . . . - - . . . . ~ .:..~....~~.~~~ '
- . .. ~. "
o TABLE 3.3-3 (Continued) .
" "l ACTION STATEMENTS (Continued) .
If . ACTION 2k - With the number of OPERABLE channels one less than the Minimum Channels OPERA 8LE requirement, restore the inoperable channel to OPERA 8LE status within 48 hours or be in at least HOT STAN08Y
- within the next 6 hours and in COLD SHUTDOWN within the following
; . 30 %urs. - -
. 19 .
ACTION as - With the number of OPERA 8LE channels one less than the Total e i * ,.
. Number of Channels, STARTUP and/or POWER OPERATION may proceed-provided the following conditions are satisfied: .
3- a. The inoperab'le channel is placed in the tripped condition within 1 hour, and l
- b. The Minimus Channels OPERA 8LE requirement is met; however,
,one additional channel may be bypassed for up to 2 hours for surveillance testing of other channels per Specification 4.3.2.1. .
ACTION & - With less than the Minimum Number of Channels OPERA 8LE, within I hour determine by observation of the associated pemissive l
',* annunciator window (s) that the interlock is in its required state for the existing plant condition, or apply Specification 3.0.3.
21 ACTION at - With the number of CPERA8LE channels one lesc than the Minimum l Channels OPERABLE requirement be in at least HOT STANOBY ,, within 6 hours and in at least HOT SHUTDOWN within the following 6 hours; however, one channel may be bypassed for up to 2 hours for surveillance testing per specification 4.3.2.1 provided the l . other channel is OPERA 8LE. ' 22. j - ACTION 23:- With the number of OPERA 8LE channels one less than the Total Number of Channels, restore the inoperable channel to OPERABLE 3 status within 48 hours or be in at least HOT STAN08Y within
; 6 hours and in at least HOT SHUTDOWN within the following l 6 hours.
93 i ACTTnN tat - With the number of OPERA 8LE channels one less than the Total ' ~ Huster of Channels, restore the inoperable channel to OPERABLE status within 48 hours or declare the associated valve inoperable and take the ACTION required by Specificationf3.7.1.5pr-- ACTION 25 - With the number of OPERA 8LE channels one less than the Minimum Channels OPERA 8LE requirement, be in at least NOT STAN08Y within 6' hours; however, one channel may be bypassed for up to 2 hours for surveillance testing per specification 4.3.2.1 provided the other. channel is OPERABLE. O ACTION 24 - With the number of OPERABLE channels one less than the Minimum i Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours or initista and maintain operation or Twr smeno FHB HyAc. 5ysret w w Em6mcr Mr. g g 3/4 3-29 COMPARISON y-sTS w .. - , - _ - . - . . - ----- . _ _ - - - _ _ _ . . _ _ . _
l f J P ge No. 32 06/13/85 . COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE i NOTES co 3/4.3.2 PAGE: 3/4 3-30.0 INSTRUMENTATION - ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INS DESIGN 1) Table 3.3-4 Item 1 Changed to " Safety Injection" (Reactor Trip, Feedwater Isolation, Control Room Emergency ventilation, Start Standby Diesel Generators, Reactor Containment Cooling Fans and Essential Cooling Water)" to reflect STPEGS specific titles DATA 2) Table 3.3-4 Item 1 and 2 Preliminary values provided. DESIGN 3) Table 3.3-4 1.e. Differential Pressure Between Steam Lines - High deleted. Not applicable to
,. STPEGS design.
ED 4) Table 3.3-4 1.f. Renumbered 1.e due to deletion of Item 1.e. DESIGN 5) - Table 3.3-4 1.f. Low-Low Compensated T (c) added.
. Based on STPEGS design.
e O
TARLE 3.3-4 ' 4 i l 3 ENGINEERED SAFETY FEATURES ACTUATION SYSTEN INSTRtalENTATION TilIP SETPOINTS SENSOR TOTAL ERROR (5) TRIP SETPOINT ALLOWASLE VALUE FUNCTIONAL UNIT _ . ALLOWANCE (TA) Z !
- 1. Safety In.jection (Reactor Trip, .
Feeduater Isolation, Control _3 ,,f . [' E*** ,J.J M x; StartjDiesel _gacWWL ~ Varra g,,,,,g,,,,gg,,g,g,,,,, Cooling c ( Fans, and Essentialfsssetse Wate3 . ,ws . N. A. N.A. N.A. N.A. j
- a. Nanual Initiation .N.A.
N.A. ! Automatic Actuationi Logic N.A. N.A. N.A. N.A. R b. S.o S.T 1 AND AcnnTooM REL4YS 3,35 5. e '.;
* ~
Containment Pressure--High 1 D0.71] [1E5] $ @ psig $ Cat'] psig y c.
- d. Pressurizar Pressure--Low D13.1]# D 10.71 [,.<{1.5 [ Sk 1850 [psig i k psig' .
Is.s] Eorer}==tivs/ i E5rry.5 5 [tes}=pst= j i re==ottferenttuM... .. _
-=sts}- .
4etween=stees=t:._. :::,. -
. j kl.$ $ psig 1 [$ psig*
6.f. k Y ea [L N Pressure--Low i s.s o.s s,9 z.ssa p z 33o.1 [
. CAPEMArto e T -low
- 2. ontainment Spray .
N.A. N.A. N.A. N.A.
- a. Manual Initiation N.A.
N.A. N.A. N.A. N.A.
- b. Automatic Actuati.on Logic M.A.
and Actuation Relays . psig . h c. Cantainment Pressure--High-3 h0.71 1.5[ $ psig i . OE c: 3 :. eg -
. h G- l
. i
Page No. 33 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE # NOTES 00 3/4.3.2 PAGE: 3/4 3-31.0 INSTRUMENTATION - ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INS DESIGN 1) Table 3.3-4 3.b Phase "b" Isolation deleted. Not applicable to STPEGS design. ED 2) Table 3.3-4 3.c renumbered b. due to deletion of 3.b.
- 3) Table 3.3-4 3.b " Purge and Exhaust Isolation" changed to " Containment Ventilation Isolation" to reflect STPEGS specific title.
- 4) Table 3.3-4 3.b.4) Containment Radioactivity -
High added to reflect STPEGS design.
,. Instrumentation Trip Setpoint values to be provided LATER.
l 9 l l
. . .x....... . . . . . .< : '
.. ..' 'T
. . 3
.i TABLE 3.3-4 (Continued) .
h FNGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS , SENSOR j 3 TOTAL ERROR ' ALLOWANCE (TA) Z (S) TRIP SETPOINT ALLOWA8LE VALUE . FUNCTIONAL UNIT .
- 3. Containment Isolation
- a. Phase "A" Isolation .
N.A. ; N.A. N.A. N.A. N.A. .
- 1) Manual Initiation 1 N.A. ;
- 2) Automatic Actuation Logic N.A. . N.A. N.A. N.A. ,
and Actuation Relays il
- 3) Safety Injection See Item 1. above for all Safety Injection Trip Setpoints and Allowable Values.
y 7
. I a;> ". ,
. ; .. "&'" isolatiesp ,
- Ems
" Erh. M.4. ,Erh. 5.4. W.A.
Cr.=i-inklatir. T; 5;tematic: Actuation N,A. -W.A. - N. h. - WrA. W. A.- .'
-togic=and=Actestion- .
-Relays:
[3:0] [03 1] = {t S] i (3h05r]-p.is-5_{12 311-y. iip I,
$ W m inment:Pressurs -
#1M CONTAINMENT WNTfLATIoAl
- c. 7_. _.2 E :=^.Al solation M.A. N.A. N.A. N.A. N.A.
g 1) Manual Initiation , I OE N.A. M.A. N.A. N.A. N.A.
- 2) Automatic Actuation -
- l Eg Logic and Actuation
'hg Relays
- 3) Safety Injection See Item 1. above for all Safety. Injection Trip Setpoints and Allowable Values.
*i
*I tri O (LATGtt)
=
(LATER) (LATER) (t.ATER) (LATER) O e 5
- 4) CONTntNMENT M i P+oloursvart -HsH M
- 9 d
P;go No. 34 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE i NOTES
------~~~------ ------ --------------------------------------------------
oc 3/4.3.2 PAGE: 3/4 3-32.0 INSTRUMENTATION - ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INS DATA 1) Table 3.3-4 4.a - c. Preliminary values provided. DESIGN 2) Table 3.3-4 4.d Steam Flow in Two Steam Lines - High deleted. Not applicable to STPEGS design.
- 3) Table 3.3-4 4.e Steam Line Pressure - Low deleted.
Not applicable to STPEGS design. ED 4) Table 3.3-4 4.f renumbered c.
- 5) Table 3.3-4 4.c " Containment Pressure High - 2" deleted. Not applicable to STPEGS. .
i-
- 6) Table 3.3-4 Item 4.d " Safety Injection" added.
STPEGS design. O e O
O t 1 2 TABLE 3.3-4 (Continued). . . if . Ih ENGINEERED .AFETY FEATURES ACTUATION SYSTEN INSTRUMENTATION TRIP SETPOINTS SENSO.
' TOTAL ERROR ALLOWANCE (TA) Z (S) _ TRIP SETP0 INT ALLOWABLE VALUE FUNC110NAL UNIT ,
- 4. Steam Line Isolation N.A. N.A. N.A. N.A. N.A.
- a. Manual Initiation Automatic Actuatimi Logic N.A. N.A. . N.A. N.A. H.A.
- b.
- and Actuation Relays
,_____, . . . _ . 4
.._L. ; G - -.J- - n -
2. 6; 6 y.. . =_6.6 .__... W in
, Z;_.!:;; ;;r -- ; =
r---- ^ 5 - ----7-- ' 44 i
--- . O '- ? 4;-
_ . . _ _ _ 6 --
- ^r- ?!= !- _ _ !rg ? !---
. _ _ _ __..= _; ;- _ _ . . _ _ , _ . . _ . . . ,
i r _ _- -_ __: :_ ::_ = _; ___n
. . .._ . p j ...--..,_= *
'". = ; ;::; . .
t
;-j
^
__- ___ [-L ij i - _-j : -- - :_-- i-___:_ ,-~; f i ' - ---2 [_1E 5- x _ _- i j [i.;j 3 ;.; p_ g 3; , M O Sf. Stems Line Pressure - h0.5/ DOM 1 h pal /s i psi s** , i ! if) o Negative Rate--High !. '
> s- '
- xs in See ]Les 1. above for 31] Safety Injection Trip setpoints and Allowable WValues.
M t* d. S A FETY [WJ Ft. TION . '
@N - .
3:= ,- M 8 . Ig .
Pzga No. 35 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE i NOTES 00 3/4.3.2 PAGE: 3/4 3-33.0 INSTRUMENTATION - ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INS DATA 1) Table 3.3-4 Item 5 Turbine Trip and Feedwater Isolation. Preliminary values provided.
- 2) Table 3.3-4 Item 5c " Low Compensated T-cold", Item 5d "High Feedwater Flow", Item Se " Safety
^
Injection" and Item 5f "T-avg - Low" added. STPEGS design. DESIGN 3) Table 3.3-4 6.d Auxiliary Feedwater, Undervoltage
- RCP deleted. Not applicable to STPEGS design.
DATA 4) Table 3.3-4 6.c Auxiliary Feedwater preliminary
.- values provided.
0
. t O
ll i ' TABLE 3.3-4 (Continued) i If - 3 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUENTATION TRIP SETPOINTS i SENSOR ' ' ERROR *: TOTAL , FUNCTIONAL UNIT ALLOWANCE (TA) Z (S) TRIP SETPOINT ALLOWA8LE VALUE i Turbine Trip and Feedwater
- 5. '
't
' Isolation '
f
- a. Automatic Actuation Logic N.A. N.A. N.A. N. A.- N.A. ; !
t Actuation Relays . I q.o 27.0 ST.I St94;335 of narrow l
- b. Steam Generator Water Level--High-High (P-14)
[S;03 h 2.18 h l.5 [narrow $ Mrange of range instrument instrument span. . u e span. ,
$ 6. Auxiliary Feedwater .
Y a. Manual Initiation N.A. N.A. 'N.A. N.A. N.A. u p
- b. Automatic Actuation Logic N.A. N.A. N.A. M.A. N.A.
and Actuation Relays I 15.0 12,. lf 33.0 33 3
- c. Steam Generator Water [40;83 g (32;*35 of g ta0;*ls of narrow 1 Level--Low-Low
[$75485)1.5 narrow range range instrument I-instrument span. f
! .- 5 Pan. .
* . L
~ " ^rvoltage : ? th. N.A. W.i. 70 .. : 69 ?,
E _ ::ii_.__ T_[_ ::f. ECP ii_._. . ,',
~ I
- c. In, n=y====*d T 3.3 0.5 I.9 ' 2538*F 15%.4*F l E d. n@ Fementer d 4.2 I.6 3.0 130.at n = 6.30.92 naw 8E ensacident with l8 i.,as rw 2.5 2.1 0.6 290.cz r1= 2ae.6xtra, g i
ss
) =
4.0 1.1 1.s 2538 r z.5 x.2 r *s M r.,3 _ i , a e. ser.cy injectsen See Irm I ame = au suanImccriou Tor &rroars e Amr (4wes. i
- f. T - to ,cate- h with 40 1.1 1.9 2538'r 25%.2*F
( p-- + Trip'
Pzga No. 36 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE # NOTES cc 3/4.3.2 PAGE: 3/4 3-34.0 INSTRUMENTATION - ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INS
- 1) Table 3.3-4 Item 6.e renumbered to 6.d. Editorial
- 2) Table 3.3-4 Item 6.f " Loss of Offsite Power", item 6.g " Trip of All Main Feedwater Pumps", and Item 6.h " Suction Transfer on Low Pressure" deleted.
Not applicable to STPEGS design. '
- 3) Table 3.3-4 Item 7.a " Automatic Switchover to Containment Sump, RWST Level-Low-Low" total allowance values to be provided LATER. Other preliminary values provided.
- 4) Table 3.3-4 Item 8.a and 8.b - Title changed from "4kv" to "4.16kv". Consistant with STPEGS
. nomenclature.
- 5) Table 3.3-4 Item 8 STPEGS values provided.
l l l l l l O
i '
, , , f 6 e' ,
-(
i TA8LE 3.3-4 (Continued) 9 ', h ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUDENTATION TRIP SETPOINTS , SENSOR . TOTAL ERROR
- 4 (S) TRTP SETPOINT ALLOWA8LE VALUE ALLOWANCE (TA) 2 ;-
] FUNCTIONAL UNIT .
- 6. Aux 111ary Feedwater (Continued) .
l See Item 1. above for all Safety Injection Trip Setpoints and Allowable Values. l d,p. Safety Injection R^ R.- ,- [9=3W - {4692]t i. 1 ) i. i.rs -ei-6ficii: . _ r- r -- 9 4. I; M.A. M.5. 5.A. 5.5. 5.5. T-1p- *-4.%%:Na!- T _ ' _tr- I
*"'P'*
--- i-:-- 2i Mi "i- i E*'*3 ?! i E""13 ?! f .
l E S-- !!-- T- ?- e fressure:. ' - u s *
) 7. Automatic Switchover to Containment Sump l
N.A. N.A. I!. A. N.A. g
- a. Automatic Actuation Logic N.A.
and Actuation Relays ' 19 (LATEld l (LATKR} l. 2. f$ ~ 2 A- WA. Mv4. 1 [183% i $383%
- b. RWST Level--Low-Low !
l . Coincident With __ _ , i .:= i'-- {-i -__ , 3_imi :=- 2 ^ tonta! .i ? , L _i ..- ___._ i:--.; a ;,=;-;e- .
=end=- !
i Safety Injection See Item 1. above for all Safety Injection Trip Setpoints and Allowable Values. I
- 8. Loss of Power '
k.A. N.A. N.A. <b5760 5652} olts
- a .4.16* kV Bus Undervoltage Ultha<M0.g75[
(Loss of Voltage) voltswith[ second Ilse - h i . gg a D 0.25 delay. ! ig a second time delay. 8 - gg ' b.4.169 kV Bus Undervoltage N.A. N.A. N.A. <b6576[olt k6511 dolt
- Githa<%3.3 ith a (I[3.3 M
*U (Grid Oegraded Voltage) 'second time I l '
second itse l E delay. delay. { t 1
P;go No. 37 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE # NOTES ; _______________ ______ __________________________________________________ j i co 3/4.3.2 PAGE: 3/4 3-35.0 INSTRUMENTATION - ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INS '
- 1) Table 3.3-4 Item 9.d " Steam Generator Water Level, P-14" deleted. Not applicable to STPEGs.
- 2) Table 3.3-4 Item 9.d " Intermediate Range Neutron Flux P-6", Item 9.e " Low Reactor Trip Block P-7, P-10 input and P-13 input", Item 9.f " Power Range Neutron Flux P-8, Item 9.g " Power Range Neutron Flux P-10.
2A) Item 9.h " Turbine Inpulse Chamber Pressure P-13" and Item 9.1 " Excessive Cooldown Protection P-15" added. This reflects STPEGS design. i- 3) Table 3.3-4 Items 9.a and 9.b - Values provided and verified for STPEGS.
. 9 8
l l l
~
l i O I l i l
i ' l ... , . .. . i i t: i , TARLE 3.3-4 (Continued) - j 'T l l 3 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUENTATION TRIP SETPOINTS i ! 5.i.550R l TOTAL E*ROR .' ! FUNCTIONAL UNIT AL10WANCE (TA) 2 5) TRIP SETPo!NT ALLOWABLE VALUE l 9. Engineered safety Features Actuatten System Interlocks ! a. Pressurizer Pressure, P-11 N.A. N.A. N.A. 1985 sig ,, 1996 fg . N.A. N.A. D550.6]*F& i
- b. Low-Low T,,,, P-12 N.A.
b553Q
. c. Reacter Trip, P-4 N.A. N.A. N.A. N.A. N.A. ,
?
! $ l s: M s N NA NA NA 21 X IN ags ?6 Y I N smps b / /C Intermediate Rarge , j g Neutron ri - u 4: 12.1 percent of lumD ' NA NA NA i-10 percent of RKIED l j,f. Im Pbuer Reactor P-10 fryut DENmt.IOER VEHLIDER j Trip Black - P-7
' P-13 input NA NA NA 610 percent %nbine 612.1 percent 2nbine Impulse Pressure Impulse Pr===re Fgdalent Fgdalent ,
1 NA NA NA 6 40 percent of RAlfD 642.1 percert of IUWB i M Pbuer Rarge Neutron Fim - P 8 MEDW.PGER MENmt.IDER j NA NA NA 210 percent of RATED 27.9 percent of R@ h p.II. Pbuer Rarge Neuturm Pitst - P-10 DEHLIDER DE3HLIDER
> Qg .
.2 '.
U _f,f. hnbine Impulse dumber NA NA NA 610 percent %nbine i12.1 perrent Rnbine 4 ED Pressure - P-13 Impulse pressure Impulse Presstre Q Fedalent ,Fedulent
" 'NA U10 percent of RATED d12.1 percent of IUWB
- 5. If. Pressive Cooldoun NA NA Protection P-15 MDMAI. IU4R V EEMAI. I U f R 1
l
I pig 3 No. 38 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE # NOTES
)
cc 3/4.3.2 PAGE: 3/4 3-35.5 l INSTRUMENTATION - ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INS l 1
- 1) Table 3.3-4 Item 10 " Control Room Ventilation" added including Item 10.a " Manual Initiation",
Itam 10.b " Safety Injection", Item 10.c " Automatic Actuation Logic and Actuation Relays (BOP)", and Item 10.d " Control Room Make-up Air , Radioactivity-High". ' 1A) This addition makes Table 3.3-4 consistent with Table 3.3-3. , 1
- 2) Table 3.3-4 Item 11 "FHB HVAC" added including 1 Item 11.a " Manual Initiation", Item 11.b
** " Automatic Actuation-Logic and Actuation Relays (BOP)", Item 11.c " Safety Injection" and Item 11.d
" Spent Fuel Pool Exhaust Radioactivity-High.
O O O
.o i .
7 ,o J 1 TABLE 3.3-4 (Continued) ENGINEERED SAFETY FEATURES ACTiiAT1001 SYSTEM INSTRiSEllTATI001 TRIP SETPo!NTS I SENSOR ' l TOTAL ERROR (S) TRIP SETPOINT ALLOWABLE VALUE FUNCTIONAL UNIT AttolaulcE (TA) 2 i
/0.t. ommut samt viserIiAFBWI N.A. N.A N.4 - d. A .
- a. Meanni Initiation N. A.
- b. Safety Im p etim -StrIrfml Aemt,r Fwt Au. Sarcrr IuJacruw 14mATIM6 FuNcTMS MD REgouttMEMTS.
w ~ N.A. N. A. N.A. MA-
- c. A* -atic Actuation ler N: N.A.
AND lagteaul AchW RELAG (SOP) ' T . . w ' i in N.A. N.A. N.A. N.A.
- d. r$ntrol Room Mahe 44p Air N.A.
Radioactivity-mish .
- 11. ne lmc N.A. N.A. N.A. N.A.
- a. Menin1 1nitiatica 'S.A. ,
N, A
- N.A. N.A. N.A. N.A.
9a b. A
- m ic Actuell e E6 o Igic and Actustlen
$0 nelsys(nor)
-w 82 z
- c. SafetyIn1ection SEE inn I An*K Font ALL SMCn hMTow IMinATIMG Twdcnaws Ao Rgmtggmas,
;8 M.A. -
- d. Spent Mael Pool thhaust N.A N.A . N.A. N.A. .
' Radioactivity - High
?
P ge No. 39 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE # NOTES oc 3/4.3.2 PAGE: 3/4 3-36.0 . INSTRUMENTATION - ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INS NONE; 1) Table 3.3-4 Table Notations. No Changes. 4 *
~
e 9 I l l l l l l 1 l . 1 1
2
- TABLE 3.3-4 (Continued) .
TA8tE NOTATIONS
- Time constants utilized in the lead-lag controller for Steam Line Pressure-Low are z 21 [503 seconds and t2 g [5] seconds. CHANNEL CALIBRATION shall ensure that these time constants are adjusted to these values.
**The time constant utilized irl the rate-lag controller for Steam Line Pressure-Negative Rata-High is less than or equal to [503 seconds. CHANNEL CALIBRATION shall ensure that this time constant is adjusted to this value.
O \
~
e
~~ .
t O . NUREG 0452/STPEGS COMPARISON PSTS 3/4 3-36
~
Page No. 40 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE i NOTES oc 3/4.3.2 PAGE: 3/4 3-37.0 INSTRUMENTATION - ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INS
- 1) Table 3.3-5 Item 1 " Manual Initiation" - delete Item 1.d " Phase 'B' Isolation", Item 1.e " Purge and Exhaust Isolation", Item 1.1 " Reactor Trip" and Item 1.m " Start Diesel Generator".
lA) These items are either not required in this part of the table or are not applicable to STPEGS. DESIGN 2) Table 3.3-5 Revised to reflect STPEGS design and nomenclature. Items renumbered as required. STPEGS preliminary values provided.
- 3) Table 3.3-5 Items 2.a.9) delete " Start Diesel i- Generators" since the diesel start and loading times are included in the values provided as indicated by the notes.
- 4) Table 3.3-5 Item 2 add new title " Steam Line Isolation" and identify as Item 2.j. The response time will be provided LATER.
l l O l
1 1Q J11 5
' TABLE 3.3-5 .
" i ENGINEERED SAFETY FEATURES RESPONSE TIMES RESPONSE TIME IN SECONOS ;
. INITIATION SIGNAL AND FUNCTION
- 1. Manual Initiation .
- a. Safety Injection-gcc5) Pomes N.A. .
- b. Containment Spray M.A.
O Y5A I ofation N A.
- h. htIaIeGoumuMENT hfNTILATION $ A. 150LAtt0N
-n = e I;d;t:_..
m
-e@
F e -j E-nedet I5--!ed--- .ltWt; d, : _ Steam Line Isolation ** N.A. ef. f, g. , Feedwater Isolation N.A. Auxiliary Feedwater PtJMP5 N.A. - 9.W. '
- h. P. E'ssenthaT b Water N.A.
i, f. fdstainment GeoWag Fan [CootrRs M.A. j.X Control RokeN bs I EMER6EMcY STANTuP N.A.
\
X. dtIIM7et((skvwvs }{VAC fMERGENCY Sr44tur StWiesel=4eneratoe. Nwati (
- 2. Containment Pressure--High-1
- a. Safety Injection :$50C3)::PtJMP5 f(27hp12h , ,,
6, 1) Reactor Trip (Faom S.I.) D2V
- c. 2) Feedwater Isolation ANDTune:Ne TAlp PT7M
. d. Y kl7M)/27M
.e. 3)h A @a"sN
- 4) SYEWlhsolation Isolationh 25 k f(10 M M60rv' f 5) Auxiliary Feedwater Pomps g, 6)Essenti . . := Water Sysrsg h32M)/ 7]@
h55k1)/{40M h.
- j. 8) 7)TfoEinment ControlRhoe#[;7$ C::: :b=Fand CoatsRs fA:fRGENCY J7MKTvP N. A.
, ar. art 01evut=Gunem g-1@
,J. Stram Luc Isaariert Currd) .
NUREG 0452/STPEGS COMPARISON W-STS 3/4 3-37 .
-- - . . - - - - , - - - - -.-,-i--~------,.----.----w--.-
P;ge No. 41 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE # NOTES 00 3/4.3.2 PAGE: 3/4 3-38,0 INSTRUMENTATION - ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INS
- 1) Table 3.3-5 Revised to reflect STPEGS nomenclature and design. Items renumbered as required. STPEGS preliminary values provided. -
- 2) Table 3.3-5 Item 3.a.9) delete " Start Diesel Generators" since the diesel start and loading a times are included in the values provided as indicated by the notes.
- 3) Table 3.3-5 Add new item 3.j " Steam Line Isolation" and new Item 3.k " Component Cooling Water System". These items are required by STPEGS design.
6 - DESIGN 4) Table 3.3-5 Add new Item 4.1 " Component Cooling l Water". Required by STPEGS design. O k I O
I l
.;.: . ' .. . .-, .. 7 . .'.- . . .T .. _,
TA8LE 3.3-5 (Continued) . ENGINEERED SAFETY FEATURES RESPONSE TIMES . 4 RESPONSE TIME IN SECONOS g INITIATING SIGNAL AND FUNCTION r: 93. Pressurizer PressugLow*
- a. SafetyInjection3([CCS) ik27d1)p12[3
.S. 2) Reactor Trip.(Fbi S.I.) ih2T*- .
~
'$ n 27 Feedwater Isolation / Tunswr TAtP SI7
$g g J_.4) Iei" Isolation , <h1742)j[7g dd A #) Mfs tYo[ 'T25k)/dOO.)
J 5 /.R Auxilia Fedwaterfears th60M
.A) Essenti Sumisa Water SysTrm 1447[)/ 2h h }f$#y , ,h . E0inment 69 tag FanfCoetsas ihssh)/ *
. J. 8) Contro1 Room A3MGia VENTKATieu Emngen agn.$ 25.0 /s0.0
%4 " 8'"' EME!
^
O _.
^
^
\ - - - - - - - - .--
N 2, ____u- -, t , , _ ,
; (?;
w -#++- 2 L'U. ~ ~ E ~ l ', e, j W n'l -;-5 i J 5j - , J. --' E; ---- U i _ , - i -_t --- \ g .E f W- : - '.
; (?Ej:; fj
I : ^ ^. 7. l _; _N" ---- -' :- , 21 ^ i ; i--j -~[- j'--
--;- _ e-- - : - : :-- !+2 -: - :.. =-
-^
-, ~- M-*" ': - :::: : _r: 1 ; [ r; CoMnuse S
4f. A team Line Pressure--
- a. SafetyInjection[Ci.5) ik12k)/J22h J. &) Reactor Trip (FkoM S I.) ih2]--
l f,. 2) Feedwater Isolation <h7h
- 8. ;97 a N A" Isolation 417N)/ ]$1)
.L. 4) Sur.s- NatiN h25k)/hh l Auxiliary Feedwater PvMP5 1360}r-
- f. .M l g.43 Essentia1A -- Water 5 Tm . D32O)/[7h
.f.f) o$binnent eenMuu Fan eetens SM5503/J40f2)
J. Coarwur Coouw Wang 4.sren , 474%$.o@ NUREG 0452/STPEGS i COMPARISON pSTS 3/4 3-38
. - , , - , _ ~ _ - - - , _ , . - . -
_ _ . - - _ _ . , _. . _ _ . , , , ,_ _ . _ . , _ _ , - , , . _ , , . . _ , , , , . _ _ _ _ _ _ _ _ , _ . _ _ _ . , . , - - . _ , - . - . - _ - - , - _, 7,-_
P ge No. 42 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE # NOTES co 3/4.3.2 PAGE: 3/4 3-39.0 , INSTRUMENTATION - ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INS
- 1) Table 3.3-5 Revised to reflect STPEGS nomenclature and design. Items renumbered as required. STPEGS preliminary values are provided.
- 2) Table 3.3-5 Item 6 " Containment Pressure-High 3" -
. change to Item 5 and delete Item 5.b " Phase 'B' Isolation" since it is not applicable to STPEGS.
- 3) Table 3.3-5 Item 7 " Containment Pressure-High 2" l
and Item 8 " Steam Flow in two Steam Lines - High" delete as they are not applicable to STPEGS.
- 4) Table 3.3-5 Item 12 "Undervoltage RCP", Item 13
" loss of Offsite Power" and Item 14 " Trip of all
, Main Feedwater Pumps". Delete as they are not applicable to STPEGS design. . I l l l l l O
7.~~ . . . . - . . . - T - K- - - - - . - X. *.T. '-.- ~.'T T - ' p 6L t
\
- TA8LE 3.3-5 (continued) .
ENGINEF 9 SAFETY FEATURES RESPONSE TIMES
'. INITIATING SIGNAL AND FC RESPONSET[MEINSECONOS ntinued) 4/. SteamLinePressy' ttenfurg4mey.Srw#.gegr.32.go74o.od
[ d'. J$ e- - Control
-f se.es c
~
g Y
' #de,steamLint 17.0 A k . Containment Pr
'L"h 6M
, ~
- a. Containr : Hi:
_ _ rn - n N g i;;;' ;- 3' - M *=*== M A255r _m M - L..
~
m N d"tnam w 6 i L T -' i 6.% staan lifan Pressure =desguabes. Rate--High Steam Line Isolation .gfMS E f70 7* % staan Generator Water J,evel--HighvHigh 2.5[
- a. Turbine Trip
- b. Feedwater Isolation ,,,
7h f.2L Staan Generator Water Level-L,ow-Low
- a. % Auxiliary Feedwater Pumps
( i L60 F , t ;_-.ir= _-_- - - - - - - ---------j t _ :___i._ i i . I Neoffuto-seensc=- t \ . 7._.--_- -=r : : = = : ^ 2 .- - ! =- - 9 2: :: = 3 :p: l 7e T-4= -s at: - -: - w
- 2 l .
- - 2._-?: _g g-4ggwate*--w: = -
NUREG 0452/STPEGS w-sTs 3/4 3-39 COMPARISON
P;ge No. 43 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE # NOTES 00 3/4.3.2 PAGE: 3/4 3-40.0 INSTRUMENTATION - ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INS
- 1) Table 3.3-5 Revised to reflect STPEGS design and nomenclature. Items renumbard as required.
STPEGS preliminary values provided.
- 2) New Item 10.c added "4.16kV ESF Bus Undervoltage (Sustained Degraded Voltage)". This reflects STPEGS design. Response time to be provided LATER.
DESIGN 3) Table 3.3-5 New Item 11 added " Compensated T cold
~
Low-law" added to reflect STPEGS design. 41> 1 O l
. .b ' - ~ ~ " - -- - . ." . . . . . .. . . ... ' T. . . .~
TA8LE 3.3-5 (Continued) . ENGINEERED SAFETY FEATURES RESPONSE TINES , RESPONSE TINE IN SECON05 INITIATING SIGNAL AND FUNCTION __ _ _ _ 7 __ *
' ~- - .
Cowour wiw Grsty fuJicrew g% RWST Lavel-Low-Low
- a. Automatic Switchover to Cantainment Sump ,N.A.
- b. Coincident with Containment Sump Level--High and Safety Injection
'(Automatic Switchover to Containment Sump) h250()/265()
! A s n ervoltage g\l0[
(Loss of Voltage) ( b. 4.16 TA kV M 4 us f.SF 8 *
< 10 [
~ '
Undervoltage Oegraded VoltageM) (TcLtAABir OATER) .5 C. 4.16 KV ESF Bus UNDERVOLTAGt(SaswMD DEGRADED VoLTka). Compensated T Low-Low g /.
- a. Safety Injection Pumps liEEer 6 23. o //A.d
- b. Reactor Trip h & J.o
- c. Feedvater Isolation 3; ster s 7. o 0 1~uS B !M E 7~M I P .
b 2. 0 (n> Containment Isolation - Phase A. -lingsr 6 2.7, d###//7.c e f. v e
- 1- 43
+ f. Containment turge Isolation &aser- * .s v. o/a.o (*)
4 q f. Auxiliary Feedwater Pumps tunt. A 40.o Reactor Containment Fan Coolers tiggr 6 S'4~ 0 /+0.o" 47
.1, jf. Essential Cooling Water System ilmenrr 6 + '/, [ '# /.r z . o j p. Component Cooling Water System sacar 3 +"). 0 > r . s'" #
NUREG 04 STPEGS
- COMPARISON W-STS 3/4 3-40
Pcge No. 44 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. l NOTE TYPE NOTE # NOTES 1 co 3/4.3.2 PAGE: 3/4 3-40.5 INSTRUMENTATION - ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INS DESIGN 1) Table 3.3-5 New Items 11 - 17 added. Addition of these items required by STPEGS design. DESIGN 2) Table 3.3-5 New Items 11 - 17 Preliminary values provided for STPEGS. e-
~
e O
~ ;~. . .;. . 7.7.
. .~ ..,=. .~. . - - ~ . 7 ~'... X .
- C ;~ . ~ .. ~; T.--*...
~. .
9 TA8LE 3.3-5 (Continued) .
; DRdTI ENGINEERED $AFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS Il ConsesusATED , Lw. Lose &
Control Room Ventilation Emergency Sta tup d) / ca. l
. toest f +0.cf.;rf.o
% Steam Line Isolation 44 + p, o
- f. Turbine Trip (Later) j Containment Airborne Radioactivity - Hiah l i
- a. Containment Ventilation Isolation * (Later)
G l E Low Compensated T o3d ! f a. Turbine Trip tusse i 2.5 O)
- b. Feedwater Isolation tuser 17.0 IN h, High Feedvater Flow - High Coincident with 2 of 4 Loops
- Having Either Reactor Coolant Flow - Lov or T - Low ev; l
- a. Turbine Trip - Reactor Trip setor fE LS .
- b. Feedwater Isolation tueur 6 7.
4 - Low Coincident with Reactor Trip ! 0)
- a. Feedwater Isolation tsar 67 0 1 14 A Control Room Makeup Air Radioactivity - Hiah
- a. Control Room Ventilation Emergency Startup (I. ster) l l
- 17. Spent Fuel Fool Exhaust Radioactivity - Hiah
- s. FEB EVAC Emergency Startup (Later) l uvRzc 0452/STPEGS COMPARISON F575 3/4 3-40A
,-.-_._,.,,,___-_,__,-,.,,,,,.,,,.i__.#,.--...._.,_..e,
- __,.._.-,,.-,,--.n_ . - - __ - - , , . , , ,
l Prge No. 45 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE # NOTES oc 3/4.3.2 PAGE: 3/4 3-41.0 INSTRUMENTATION - ENGINEERED SAFETY FEATURES ACTtTATION SYSTEM INS DESIGN 1) Table 3.3-5 Table Notations. Items 4 & 5. Change "RHR pumps" to " Low-head Safety Injection pumps".
- STPEGS does not utilize RHR for low head safety injection.
O 9 l O i
l __ .. ...,....._... . .. .. T_.
. ... . .. . " : . . ~..
O .- ... y
. TABLE 3.3-5 (Continued) . M
* ~ *
. TA8LE NOTATIONS (1) Diesel generator starting and sequence loading delays kncluded.
.- (2) Diesel generator starting and sequence loading delay not included.
Offsite power available. {3) Air-operated valves. Low HEAoSann.TmeerwN
,.(4) Diesel generator starting and sequence loading delay included. Atut
. pumps g included. .
(5) Diesel generator starting and sequence loading delays not included. AIR'/ pumps M included. A ~ Low fluo Svert lwJtenoa
\.
W-STS
~
3/4 3-41 NUREG 0452/STPEGS COMPARISON
~
N . _ _ _ _ _
r - P2g3 No. 46 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. l NOTE TYPE NOTE # NOTES ' 00 3/4.3.2 PAGE: 3/4 3-42.0 INSTRUMENTATION - ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INS ED 1) Table 4.3-2 Item 1. Changed to ' Safety Injection (Reactor Trip, Feedwater Isolation, Control Room Emergency Ventilation, Start Standby Diesel Generator, Reactor Containment Fan Coolers, Essential Cooling Water) ' to reflect STPEGS titles. DESIGN- 2) Table 4.3-2 1.e. ' Differential Pressure Between Steam Lines - High' deleted. Not applicable to STPEGS design. ED 3) Table 4.3-2 1.f. Changed to 1.e ' Low Compensated Steam Line Pressure'. Editorial. DESIGN 4) Table 4.3-2 1.f. Low-Low Compensated T(cold) - Low-Low added to reflect STPEGS design. O f l 1 0 (
i t I; TARLE 4.3-2 .I 39 h ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTatN NTATION SURVEILLANCE REQUIntntals *jl ! TRIP , .j . ' ANALOG ACTUATING MODES CHANNEL DEVICE MASTER SLAVE FOR 14tICN h' CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE L ! CHANNEL CHECK CALIRRATION TEST TEST LOGIC TEST TEST TEST IS REQUIRED d FinICTIONAL UNIT i 1 l'; l 1. Safety Injection (Reactor Trip, : Feedwater Isolation Control SW , ! Roomilastatine; Start [0lesel l. Generators, Containment Coolina . ' l Fans, and Essentialf5eentes Water) g yg .
- a. Manual Initiation N.A. N.A. N.A. R . N.A. N. A.' N.A. 1,2,3,4 f
l , 1,2,3,4 I b. Automatic Actuatic t. A. N.A. . N.A. N.A. M(1) M(1) Q u h,
) Logic and Actuat' 1 3 w Relays 1
k c. Containment Pressure- S R H N.A. N.A. N.A. N.A. 1. 2, 3 l k High-1
- d. Pressurizer Pressure- S R H N.A. N.A. N.A. N.A. 1,2,3 Low Eo Eo E^ :
!{'
e Bisi m .Aiei_.^. m -- - E -
^^
2. I 3 t
-R$
- ca gle c s4rro }
5 H N.A. N'. A. N.A. N.A. 1, 2, 3
- .2 4 53 team Lfne R
'" S R M MA. M. A . KA. W A- lo h 3 :
h 2.g. CON *$o'*}br.t,v.t.a ay g Coatalnment i N.A. N.A. N.A. R N.A. N. A'. N.A. 1,'A, 3 4 .l I h.ManualInitiation 1 , 2 , 3. '4'
. O k. Automatic Actuation N.A. N.A. N.A. N.A. M(1) M(1) Q fI 5g Logic and Actuation .
l - hg Relays N.A. i-E k . Containment Pressure- S R H N.A. N.A. N.A. 1, 2, 3@C l *8 1
*y High-3 -
p i o 3 pI . l
- j
e-Peg: No. 47 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE # NOTES ' oc 3/4.3.2 PAGE: 3/4 3-43.0 INSTRUMENTATION - ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INS DESIGN 1) Table 4.3-2 3.b ' Containment Isolation Phase "B" Isolation' deleted. Not applicable to STPEGS design. DESIGN 2) Table 4.3-2 3.c.1), 2), 3) renumbered b.1), 2), 4 and 3), respectively. Item renamed Containment Ventilation Isolation. Note added: Manual actuation is accomplished by manual Phase A isolation and manual containment spray actuation. DESIGN 3) Table 4.3-2 3.b.4) RCB Purge Radioactivity - High added to reflect STPEGS design. e o O l l e
- O i
..L..... .
- . ' .. -l '
y l i TARLE 4.3-2 (Continued) , if : ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUENTATION sa SURVEILLANCE REQUIRENENTS
, ]
i TRIP . ANALOG ACTUATING NDOES CHANNEL DEVICE NASTER SLAVE FOR nellCN CHANNEL CNAIRIEL CNANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE (-
- FUNCTIONAL UNIT CHECK CALIRRATION TEST . TEST LOGIC TEST TEST TEST 15 REQUIRED j l 3. Containment Isolation l ,'
i a. Phase "A" Isolation l N.A. N.A. N.A. R N.A. N.A. N.A 1 2,3,4 i
- 1) Nanual Initiation N.A. N.A. N.A. N.A. M(1) M(1) Q 1,2,3,4
- 2) Automatic Actuation .
J , Logic and Actuation Rclays ,.- j ) , See Ites 1. at a for all Safety Injection Surveillance Requirements. ..
"a w
- 3) Safety injection l -
4 D=9tannatstatti=*'_ -- i ;.i
^ ^ ^ - ^ ^ ^
-_ ^^_ -- E^ ( ;; ; :
=
l' -et=automati-: A-r = =6 - s_= 7.= w_=_ == =.J3 --.=- . E. :. - a
% : .^__ _ _ _ : _..
-astays.
ll'
-= =
-- A - - - == == : :
; 2) raura-
^
- = == . - ,
j . . _ _ _ _ _ _ III ,I. I 8 ,e s lation I i6 o ,
,jl; $; 1) Manual Initiation SEE 119t:
Irrms'ttEME. 2.a.0m "$1EE. 3.a.0w T a mc. ~.90:3C. wnw huh W N. herme.
"Jt:2. 1:
.i Hw 4 .
hy 2) Automatic Actuation Logic and Actuation N.A. N.A. N.A. N.A. M(1) N(1) Q 1, 2, 3, 4 g Relays ,-, , g3 N -l
~
i
- 3) Safety Injection See item 1. above for all Safety Injection Survelliance Requlrements, p t;
- 4) RCS Paest Roomnenvaf-See SncincAnw +.s.3.I w sveveauwe &outemeur.s. i. '
uness
I \ l Page No. 48 i 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE - NOTE i NOTES oc 3/4.3.2 PAGE: 3/4 3-44.0 INSTRUMENTATION - ENGINEERED SAFETY FEATURES ACTt1ATION SYSTEM INS DESIGN- 1) Table 4.3-2 4.c Containment Pressure - High -2 deleted. Not applicable to STPEGS design. DESIGN- 2) Table 4.3-2 4.d Steam Flow in Two Steam Lines-High deleted. Not applicable to STPEGS design. DESIGN- 3) Table 4.3-2 4.e Steam Line Pressure - Low deleted. Not applicable to STPEGS design. ED 4) Table 4.3-2 4.f Changed to 4.c ' Steam Negative Pressure Rate - High' to reflect STPEGS nomenclature. DESIdN 5) Table 4.3-2 4.d ' Safety Injection' added to reflect STPEGS design. f 1 l l O
s TABLE 4.3-2 (Continued) .. 9 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRtRENTATION Q SURVEILLANCE REqd1REMtnI5 ,l.
. I*
. TRIP '
ANALOG ACTUATING MDOES
- CHANNEL DEVICE MASTER SLAVE FOR WICN :
RELAY RELAY SURVEILLANCE :.. CHANNEL CNANNEL CNANNEL OPERATIONAL OPERATIONAL ACTUATION CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST _IS REQUIRE 0 ' ? FUNCTIONAL UNIT
- 4. Stoan Line Isolation l g
- a. Manual Inttiation N.A. N.A. N.A. R N.A. N.A. N.A. 1, 2, 3 lf 1, 2, 3 4'
- b. Automatic Actuation it. A. N.A N.A N.A. M(1) M(1) Q .
Logic and Actuation [ Relays m -_=r j- --- a ;- _ _ - { e E Ee Ee Ee Ee i, =, i_ i-mee i . R ==cuam-rj _- ; -- te - e a- g g_e ae ee ae i. _=. =- : e
- tatacideatztsttIE
.;_ ;- - : --- -_. = E ==__ M- e_&_ M4 1, i, i = l
; 2 2 24- : i- := i- .- LL2 eteam-i i-- -
6ArwE b T. team time Pressure- S R H N.A. N. A. Bl. A. N.A. 3 g'NegativeRate-Hip
^ '### #
- 5.TierblnNerfpaSdYeedwafeY j s Isolation . : ,
- a. Automatic Actuation N.A. N.A. N.A. N.A. M(1) .M(1) Q 1, 2,3 l, Q Logic and Actuation ,
M Relays
.M N.A. N.A. N.A. N.A.
1, 2 - j,
- b. Steam Generator Water S R ,
tevel-High-High l
$ C Auxiliary Feedwater F4 N .
- a. Manual Initiation N.A. N.A. N.A. R N.A. N.A. N.A. 1, 2, 3 g
;?, b. Automatic Actuation N.A. N.A .N.A. N.A. M(1) M(1) - Q 't, 2, 3 % '
g and Actuation Relays : >
- e
- _ . - - )
P ge No. 49 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE # NOTES cc 3/4.3.2 PAGE: 3/4 3-44.5 , INSTRUMENTATION - ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INS DESIGN 1) Table 4.3-2 5.c Turbine Trip and Feedwater Isolation, Compensated T(cold) - Low added to reflect STPEGS design. DESIGN 2) Table 4.3-2 5.d Turbine Trip and Feedwater Isolation, Feedwater Flow - High added to reflect GTPEGS design. DESIGN 3) Table 4.3-2 5.e Turbine Trip and Feedwater Isolation, Safety Injection added to relfact STPEGS design. DESJGN 4) Table 4.3-2 5.f Turbine Trip and Feedwater Isolation, T(avg) - Low added to reflect STPEGS design. O l l l 0 l l
O 3,3 3 hl. - - -- lIE d i i llR- i : I ll = - p
,hl, e I-0 rl 8 '
~
l5-3 1 lll, t l .
\
l lg - . a. g r 8-1 ?$ lf
,g=Jp' ! :.9 4 1 l
11 i > .o ruser w 3 % = g-
-,,---,-.,-m..ma
Pcge No. 50 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE # NOTES 00 3/4.3.2 PAGE: 3/4 3-45.0 INSTRUMENTATION - ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INS ED 1) Table 4.3-2 Item 6d "Undervoltage RCP", Item 6.f
" Loss of Offsite Power", Item 6.g " Trip of all Main Feedwater Pumps" and Item 6.h " Suction Transfer on Low Pressure" deleted. Not applicble to STPEGS design.
ED 2) Table 4.3-2 Item 6.e " Safety Injection - renumber to Item 6.d Editorial. DESIGN 3) Table 4.3-2 Item 7.b - Delete " Containment Sump Level High". Not applicable to STPEGS. O I l O
! O .O o q
TABLE 4.3-2 (Cont'inued) h ENGINEERED SAFETY FEATURES ACTUATION SYSTEN INSTRUMENTATION SURVEILLANCE REQUIRENENTS 1 - i TRIP '
~
ANALOG ACTUATING A00ES i
- CHANNEL DEVICE MASTER SLAVE FOR WHICH l
CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE i CHANNEL FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST IS REQUIRED l
- 6. Auxiliary Feedwater (Continued) l
- c. Steam Generator Water S R H N.A. N.A. N.A N.A 1, 2, 3 1
j Level-Low-Low
= =-_=2==-- 2i -- ;g _
2_ i_2. M -2 -2 =2 : ,- l
- g. Safety Injection See Iten 1. above for all Safety Injection Surveillance Requirements. l 4 i---- - f-- ---- E =_ E F a. ,M 4. .a . M.5. M.i 1, 2, 3 -
*t
--f 0 isit
- M;= E=
l g w eir._r.;_ M ;_ m t.._f n - M_# #;+ M_ _ E E_= 1. = - l "
= .:- : : xx : xr r_.
- m -
l
- 7. Automatic Switchover to .
Containment Sump
- a. Automatic Actuation N.A. N.A. N.A. N.A.- M(1) M(1) Q 1, 2, 3, 4 l
i logic and Actuation ! hRelays i OE R M N.A. N.A. N.A. N.A 1,2,3,4 ig h RWST Level-Low-Low S , ll;$ Coincident With mu
-e E=
j h f .1-; _ -?- E M M-e E= :
.=
CJ e s =tevet- ?p-- , WiiI e Safety Injection See Item 1..above for all Safety Injection Surveillance Requirements.
P go No. 51 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE # NOTES 00 3/4.3.2 PAGE: 3/4 3-46.0 . INSTRUMENTATION - ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INS ED 1) Table 4.3-2 Item 8 - Change title to read " Loss of Power-Diesel Generator Start", Item 8.a change to read "4.16kV ESF Bus Undervoltage (Loss of voltage)", Item 8.b to read "4.16kV ESF Bus Undervoltage (Grid Degraded Voltage) + S.I. DESIGN 2) Add new Item 8.c "4.16kV ESF Bus Undervoltage (Sustained Degraded Voltage)".- This reflects correct STPEGS design and nomenclature. DESIGN 3) Table 4.3-2 Item 9.d delete " Steam Generator Water Level P-14" and add " Power Range Neutron
- Flux Input to Excessive Cooldown Protection P-15".
This reflects STPEGS design. O e N O
O :O On 3. l 9 TABLE 4.3-2 (Continued)- '.
?,
3 ENGINEERES SAFETY FEATURES ACTUATIO11 SYSTEM Ill5TRl4ENTATICII SURVEILLAllCE REQUIREMENTS .
- ?l- .
TRIP , : ANALOG ' ACTUATING N00E5 ' CHANNEL DEVICE MASTER SLAVE FOR 1AIICH l CH40mlEL CHAle1EL CHANNEL OPERATIONAL OPERATIONAL ACTunTICII RELAY RELAY SURVEILLANCE 1-FUNCTIONAL UNIT CHECX CALIBRATION TEST TEST LOGIC TEST TEST TEST 15 RE0ulaED ?
- s. e of Pouer-assa 6*aE***' l :
.th ESF t
- a. 43kVapus II. A. R N.A N N.A. N.A. N.A. 1,2,3,4 underveltage (Less of Voltage) ..
A.lb ESF ' i
- b. Ta kVaSus W.A. R. N.A. N N.A. N.A. II. A. 1, 2, 3, 4 .
Undervaltate (Grid Degraded Voltage)+S.I. - ', I
. . s .
- s. 4.16 nY ESF Bos y,p. R. N. A. M g. A. H. A. N.A I.2, 5. 4 UdoCM7esE -
(Susrasto DEstepfD 16L74
*t 9. Engineered Safety '
;I g Features Actuattaa '
System Interlocks
- a. Pressurizer N.A. R H N.A.
- N. A. N.A. N.A. 1, 2, 3 Pressure, P-11
- h. Low-Law T,,,, P-32 II. A. A N N.A. N.A. N.A. II. A. 1, 2, 3 g c. iteacter Trip, P-4 ll. A. N.A N.A. _R N.A. N.A. N.A.
- i, 2, 3 - !
Eh :--- - - - - -
- = = =_ _x g
- s. -
-. --. 5 - 5 .
~'
OE gn A.Twxf%weercsttoDKEssent M M $$A R M pl g.a* gg* gg* ;* 2' 3 ' y coowownIkrscrseu.P-ts ..
!6 I
)
c
]
P'ge No. 52 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE i NOTES l l 60 3/4.3.2 PAGE: 3/4 3-46.5 . INSTRUMENTATION - ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INS DESIGN 1) Table 4.3-2 Add new Item 10 " Control Room Ventilation" and new Item 11 "FHB HVAC". These reflect STPEGS design. i 8 e O o O
TAIE2 4.3-2 (Contivanad) sarmeumen SAFE 2Y FD3tBE ACH3G1GI ST5tBt ilEMEMMUG1GI SElvl51DNE urqui ! 1 RIP ANAIDG ACitNTDC NIIES OwtEL IlEVIM MA!mR SAVE IUt WHOf OWHL OWN!L OPHINTIGIRL OPFRATIGIAL ACIUG10N IIEI M IIEIX SMUEDUNE RSCt104AL 13RT OHK _CALIIRATIGI 1TST TEST II)GlC 115T 115r TEST IS IIH)JDIED A: comet gameuurnmum
- a. Mammal 1=tein ri= N.A N.A N.A R N.A N.A N.A AIL
- b. Awa==rie San N.A N.A N.A N.A N(1) N.A N.A Att legic amul Arhantim .
Relays (Bur) e 3 ,g g y nag See Iran I abase for all Safety injection Surwillance , "-- ;- J.Ceannas. Roe MAsu# AIR KApMnet stognef fSee Speciffention 4.3.3.I for pH1 mace segsinments
- 11. MS IRIAC LA3
~
- s. Manaal Inteiselm N.A. N.A. N.A. R N.A. N.A. N.A. 1, 2, 3, 4, with 8
A irradiated fuel in the spent fuel pool
- b. hea==rie ach== rim N.A. N.A. N.A. N.A. M(1) N.A. N.A. 1,2,3,4,with irradiated fuel in the spent s fuel pool gy c. Safety Injection See Itan 1 above for all Safety Injection Surveillance Requinments y d. Spent Ram 1 7bol See Specification 4.3.3.1 fee merwill-ww nequirnames g f 1 2 mmet - "- ;" 2 7-EA %
g .
~~
8 , s TABLE IISTATION (1) Each trais shall be tested at. least every 62 days on a STAGGERED TEST BASIS.
1 l l P ge No. 53 ) COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE i NOTES i oc 3/4.3.3.1 PAGE: 3/4 3-47.0 INSTRUMENTATION - MONITORING INSTR RADIATION MON-ITORING DESIGN 1) ' 3.3.3.1 Delete entire L.C.O. Replace with L.C.O. consistant with Spec 3.3.2 and Table 3.3-6. a i* O I l . l l l O
;b;.5. '- & . : . .. . . . . ~ . '. ' . .. . :. :.r.' ~ ' ' ~ ' '
~ ; !
INSTRUMENTATION MontroRING INSTRUMENTATION 3/4. 3.11MetMEERED= SAFETY = FEATURES-ACTUAff0ft:4YSTDktivstRUMDs***tC* ' , RADIATION MONtTORING FOR PLANT OPERATIONS JMITING CONDITION FOR OPERATION .
}.3.3.1_Tns,RAapnox, gwyg 14srgonsgrn ogpMygg(E5FA s sq t qAA r cro:4rygs
..... ; ra _, . . _ _ . _ _ ;; . . ., ...tum s=Actuat .m= , ;...__...;.;..-
- 3. 3 - 6 h sNews in Table be OPfAABLE with theIr Trip Setpoints set consistent with the values shown in the Trip 5etpoint column of .
Table 3.3- . APPt.ICA81LITY: As shown in Table 3.3 . ACTION: n% WITH A RADIATied MoMITORING CNANWEL FoR PLANT OPERATICWS w :1 __ -- m - e- -e--___ -+ - ' Q- Trip Setpoint Atp*p. less conservative than the value shown in the Trip Satpoint column but as,re conservative than the value shown in the Allowable Value column of Table 3.3-A,' adjust the Setpoint consistant with the Trip 5etpoint value. 4 . A RAotArio4 McN ToRIN6 CNANNEl.
- b. With4 . Trip 5etpoint less conserva-tive than f.he value shown in the Allowable Value column of Table 3.3 , either:
O i.
- 1. Adjust the Setpoint consistent with the Trip 5etpoint value of 3.16 ~ TableA3::3=*, and determine within 12 hours that Equation 2.2-1 I
was satisfied for the affected channel, or
- 2. Declare the channel inoperable and apply the applicable ACTION 3*S-4 statement requirements of Table 32x3/gntt i sne snannel is
- restored to OPERA 8LE status with its setpoint adjusted consistent with the Trip 5etpoint value. ,
Equation 2.2-1 Z + R + 5 f,TA Where: (o Z = The value from Colisen Z of Table 3.3-/ for the affected channel, R = The "as measured" value (in percent span) of rack error for the affected channel, 5 = Either the "as measured" value (in percent span) of the sensor error,orthevaluefromColumn5(SensorError)ofTable3.3-g for the affected channel, and TA = The value from Column TA (Total Allowance) of Table 3.3-A for the affected channel. '*
- c. With an ESFA5 instrumentation channel or interlock inoperable, take o theACTIONshowninTable3.3f.
47 PSTS 3/4 3-38 NUREG 0452/STPEGS COMPARISON
Page No. 54 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE # NOTES co 3/4.3.3.1 PAGE: 3/4 3-47.5 INSTRUMENTATION - MONITORING INSTR RADIATION MONITORING
- 2) 4.3.3.1 Change " Analog Channel Operational Test" to read " Digital Channel Operational Test". This reflects the STPEGS instrumentation. See Definition Section.
a
. O' e
i
,, ,- - .-..~..-, ._.
h IL)I U f "IN'STRUMENTATION - J SURVEILLANCE REQUIREMENTS 4.3.3.1 Each radiation monitoring instrumentation channel for plant operations p e rgg. mshall be demonstrated CPERA8LE by the performance of the CHANNEL CHECK, CHANNEL '
' UALIBRATIONAandjAWAGG CHANNEL OPERATIONAL TEST for the MODES and at the } frequencies shown in Table 4.3-3.
1 l O ' (. 4 l
. 1 l
.i
. I
. 1
- l l
1
.~
1
^~, I 1
O MIREG 04!2/STPECS PSTS 3/4 3-47 A COMPARISON b
P;ge No. 55 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE # NOTES 00 3/4.3.3.1 PAGE: 3/4 3-48.0 INSTRUMENTATION - MONITORING INSTR RADIATION MONITORING
- 1) Table 3.3-6 Delete old table 3.3-6 and replace with a revised Table reflecting the STPEGS monitoring instrumentation for 1) Containment, 2)
Fuel Handling Building and.3) Control Room / Electrical Auxiliary Building. lA) Column Headings for Total Allowance, 2, Sensor Error, Trip Setpoint, Allowable Value, Total number of channels, channels to Trip, Minimum channels operable, Applicable Modes and Action. See Specification 3.3.3.1 for explanation. s e O e l O
~, - - - , ,- , , - . . - . - - , ,
. O .
O . O> .g TA8tE 3.3-6
'1 * * -
RADIATION MONITORING INSTRUNENTATION FOR PLANT OPERATIONS h MINIMUM CHAWlELS CHANNELS APPLICA8LE ALARN/ TRIP TO TRIP /ALARN OPERA 8LE M00ES SETPOINT ACTION FUNCT10NAL UNIT
- 1. Contal at
*1
- a. Containee Atmosphere 1 2 All 5 [2] mR/h :
Radioactivit igh ,
- b. RCS Leakage Detect ' '
N.A. I 1, 2, 3 N.A. 29
- 1) Particulate Radioact ty 1, 2, 4 N.A. 29
- 2) Gaseous Radioactivity N.A. I
$ 2. Purge and Exhaust Ventilation ,
I a. Particulate Radioactivity 1 2 All 26 ,,
- b. Gaseous Radioactivity 1 All
- 26 C '
i rti :'
- 3. Fuel Storage Pool Areas n (T) }
,. a. Radioactivity-High , q Gaseous Radioactivity 1. 2 na , < [2] mR/h 27 , [i
- b. Criticality-Radiation L 1 2 ***
1 15 mR/h 28 {.
- 4. Control Room ;
88 a. Air Intake- lation Level 1/ intake 2/ intake All 1 [2] 27 . s i6 o All 2
- . ll; $; b. Contre Room Atmosphere 1 2 1 [2] mR/h Gt' Radiatlon-High eM / .-
i a .
!. t;.
! O -'_ , l . ; V
. o i
f P2g3 No. 56 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND 3TPEGS TECH. SPECS. NOTE TYPE NOTE # NOTES 00 3/4.3.3.1 PAGE: 3/4 3-48.5 INSTRUMENTATION - MONITORING INSTR RADIATION MONITORING DESIGN 1) Table 3.3-6 Replaces STSRev5 page 3/4 3-48 to j reflect STPEGS design. O
.t l
l 1
~
s
/
TABLE 3.3-4 g RADIAT100B lenutTORINC INSTRt90ENTAfl008 FOR PIANT OPERATIONS 4* - sfusan in TSTAL Estee 10 tat 30. osseets auseets Aryuounz puncTtomat aut? AttSunsett (TA) { ($) TRIP $firellft R L a nett utgf CP O*88813 101 RIP tytRnstz pass acruns
- 1. Centelement z i 2, 4 e6
- e. RC rurse Ceees.e (tArri) (sAeA) (unn) 7.5 x ed*-/c<. (tane)
- Radteactivity - Righ (AIRA-RE-8012 CIRA-RE-4013)
- 6. Ses Atese,here Ceoes.. g.A. n.m . g.g M A. EA. l #4. l /,2, 3,4- 29 l g Radioactivity - RCPS 8- Leehage Detectlen y (atRA-RE-80 IC)
E # A. I f.2, A 4 29
- c. Res At se,here rectacolate #4 N. A. N.A. N. A. I N. A .
Radleectivity - RCPS Leakene Detectice (WIRA-RE-80llA)
- 2. FueI NeedIing Dutidlug see./b e.es Eehe... j 2 i 2 28 Casease RadioactiV y - Mlah (tArra) (taTrn) (tarfe) 8.8 xid (r.4TFe)
(AIRA-RE-8035. CIRA-RE-8036)
- 3. Centrol Reen/Electrieel Ameillery g Rutidtas n $ e. Centret RosaltAR Air latehe (LATEM) (TATER) (14TER) I 3X10 m (L4TER) 2, f 2 A s.t 27 O C1 caseeme Radteocttwity - Neph ho (AIRA-RE-8033, CIRA-RE-En14)
%e&
8N m
- P1
?
(n
PCgo No. 57 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE # NOTES 00 3/4.3.3.1 PAGE: 3/4 3-49.0 INSTRUMENTATION - MONITORING INSTR RADIATION MONITORING ED 1) Table 3.3-6 Table Notations - Delete
- and ***
statements and change ** to
- which reads "... fuel in the spent fuel pool". This reflects requirements for STPEGS.
- 2) Table 3.3-6 Action 26 - Change to read "....
a containment purge valves are maintained closed". This reflects STPEGS design and nomenclature.
- 3) Table 3.3-6 Action 27 Changed to identify the STPEGS system as " Control Room Envelope HVAC System in the emergency mode". This reflects
, ,# STPEGS design.
- 4) Table 3.3-6 Action 28 change to read ".... is provided in the spent fuel pool exhaust or the FHB HVAC System is operated in the filtration mode".
This reflects STEGS design.
.t O
I O _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _______.-.________J
- - - - - - --~~ .
DRAFT TA8LE 3.3-6 (Continued) . TABLE NOTATIONS
.. _.. _ _ a a.. ____a.. __ __ .. . . __...__'____ _ -
With irradiated fuel in the{ fuel stsump pool.gEEEE
. ACTION STATEMENTS .
f ACTION 26 - With less than the Minfeum Channels OPERA 8LE requirement,
. operation may continue provided the containment purge met valves are maintained closed.
. ACTION 27 - With the number of CPERABLE channels one less than the Minimum Channels OPERA 8LE requirement, within 1 hour fus8ess: M initiate operation
* '* ** bY "** '" *h*
EnrmsucY , ACTION 28 - With less than tV Minimum Channels OPERA 8LE requirement, opera-tion may continue fnr up to 30 days provided an appropriate O ., smT _ portable continunss monitor with the same Alarm Setpoint' is provided in thek/ual wesamye poo1Jamma. Restore une inoperable monitors to CPEDSLE status within 30 days or suspend all f operations involvf.y fuel movement in theVfuel M pool, asseo saur ; ACTION 29 - Must satisfy the A'. TION requirement for Specification 3.4.6.1. , gyg4pg og in f00 WN Sqg y ss ofERJpTO IM W E
. VsLTRATsaN MODE.
i i l l
\
I
- O NUREG 0452/STPEGS COMPARISON y-STS 3/4 3 49
,e ,,nr, - - . - - - - - - , ,--- aw,,,,r,.,,,,,myv.---n_---.--__--. -.-,_w__-__~na-- .,,. -,a,,,-,,_ _,, ,.,_a_.,_,____,,---n,, , - - - - - - - , - . - - - , ,
Pzg3 No. 58 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. t NOTE TYPE NOTE f NOTES co 3/4.3.3.1 PAGE: 3/4 3-50.0 INSTRUMENTATION - MONITORING INSTR RADIATION MONITORING l
- 1) Table 4.3-3 Revised the Functional Unit column to be consistent with the revised Table 3.3-6.
- 2) Table 4.3-3, Notations - Delete "(*)" since it is not applicable to STPEGS and change "(**)" to
"(*)" . In addition change wording of new "(*)" to read "With irradiated fuel in the spent fuel pool." This reflects STPEGS design.
O
.t O
r i TABLE 4.3-3 RASIATION MONITORilIG lil5TRISENTATIO11 FOR PLANT - 3 OPERATIOll5 3 aILLANCE REQUIuEntnis
,,,,,, i~*
CHAlelEL plDOES FOR 4R0100 , CHAISIEL CHAlelEL OPERATIONAL SURVEILLAllCE . CHECK CALIRRATI0ll TEST 15 REQUIRES ,;; FtBICTIGIIAL 181IT " 1 Cantainment a.RCB lbnGCGuea.5 '
" '" "& ~
RacnontTwsrf(AIRA RE-yol2, CIRA-Rf-ct: 3wienctivityettle 5 R M E6 . b,RtB Armstvtef t=aaC$rtesRags-amem 1,2,3,4 !i fC 5 tT 1)Eterth=h*= 5 R M (g tRp-RE-follA) *CLI*ILY l
* *** "Y *
~ ~
^
c.RCS PJ" " *** . w futs(ULATC __- _,-_ __ _ g____.______g- . - . . - - . . . w EPS LEA 80%E 5 R M JETI,&3,4 J. Intr o v m e m ,t: w - e = n 3;3-
*(NLRA-RE-TollA)2. FUE_L HANDLIQG St]ILDIN5 a.Snwr FA% &&y f Euwasr GAwas hh*=#7 5 R M *(~
g , w (41RA- ' RE-fd35,CMA-RE-fax'-ggge.gf,-p 3-gag,,; , 3.t. Centre) Room / ELECT AUX lLIARf "- DOILQllV[RICAL s ^" i-- ; 5 R H All a.Cwraxb EAB Aic Lan*E s m - aussatseesss-
%,ggmay --u - - - -- - n ,
' (R$-RE-fo33, TABLE IIDTATIosis
,g,y,3g '
as m Trs= *==-semi = + - : - - - O ! 82 C a with irradiated fuei in theif41 semesse poei.neemo .
*4 swer m :
5 h m I
P ge No. 59 06/13/85
. COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS.
NOTE TYPE NOTE # NOTES 00 3/4.3.3.2 PAGE: 3/4 3-51.0 INSTRUMENTATION - MOVABLE INCORE DETECTORS - NONE 1) 3/4.3.3.2 No Changes. 4 4
~
o
.I 4
2 S e
7 l O
. INSTRUMENTATION
, UN 1
. MOVA8LiINCOREOETECTORS i LIMITING CONDITION FOR OPERATION I -
3.3.3.2 The Movable Incore Detection Systen shall be OPERABLE with: i
- s. At least 75% of the detector thimbles, ;
- b. A minimue of two detector thimbles per core quadrant, and l
- c. Suf ficient movable detectors, drive, and readout equipment to map l these thiebles. ;
APPLICA8!LITY: When the Movable Incere Detection System is used for: l
- a. Recalibration of the Excore Neutron Flux Detection System, or l f
- b. Monitoring the QUADRANT POWEA TILT RATIO, or
~
- c. MeasurementofFh,F(Z)andF q m, g ACTION.
With the Movable Incore Detection System inoperable, do not use the systte for l l the above applicable monitoring or calibration functions. The provisions of ) Specifications 3.0.3 and 3.0.4 are not appitcable. l
<! j k* \
\
' SURVEILLAACE REQUIREMENTS j 4.3.3.2 TheMovableIncoreDetectionSystemshallbedemonstratedOPERABLEatl least once per 24 hours by norinalizing each detector output when required for:
. a. Recalibration of the Excore Neutun Flux Detection System, or l
- b. Monitoring the QUADRANT POWER TILT RAT 10, or
- c. Measurement of Fg N , Fq(Z), and F,y.
O' W
,,-STS 3/4 3-51 NUREG 0452/STPEGS COMPAltISON 1
I
P ge No. 60 ' 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE # NOTES oc 3/4.3.3.3 PAGE: 3/4 3-52.0
- INSTRUMENTATION - SEISMIC INSTRUMENTATION
- 1) 3.3.3.3 Action a. Inoperable period changed from
'30' days to '31' days. For consistency HL&P
, proposes to use 31 days throughout the Tech Specs as a monthly requirement, see Table 1-1.
DESIGN- 2)- 3.3.3.3 Note (*) deleted. Not applicable to STPEGS design e-O-
.t l -
l t f , O
i j
- ~~... . 2 :~~L.: :- -
.-.3 -
BRAFT INSTRUMENTATION ) . SEISMIC I STRUMENTATION 1 . . i LIMITING CON 0! TION FOR OPERATION 1
! 3.3.3.3 TheseisaicmonitoringinstrumentationshowninTable3.3-7"shallbe.l
- OptAASLE.
APpLICA8!LTTY: At all times. . 1 LGQN: . j a. With one or more of the abdve required sofsaic monitoring instruments i ~ '51 - Report Inoperante for more than 30 days, prepare and submit a Special to the Commission); pursuant to Specification 6.9.2 within the next 10 days outlining the cause of the salfunction and the plans
- for restoring the instrument (s) to CPEAASLE status.
I b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. i SURVEILLANCE REQUIREMENTS l I l - 4.3.3.3.1 fach of the above required seismic monitoring instruments shall be t l ! demonstrated OPERA 8LE by the performance of the CHANNEL CHECX, CHANNEL CALI- , i SRATION, and ANALOG CHANNEL OPERATIONAL TEST at the frequencies shown in l ! ! Table 4.3-4. l \
- 4.3.3.3.2 Each of the above required seismic nitert instruments actuated '
i duri a seismic event greater than or equal t 0.01 shall be restored'to i , OP LE status within 24 hours and a CHANNEL CALISRA I performed within l 10 days following the seismic event. Oata shall be retrieved from actuated i instruments and analyzed to determine the magnitude of the vibratory ground , estfon. A Special Report shall be prepared and submitted to the Commission '. pursuant to Specification 6.9.2 within 14 days describing the segnitude, l frequency spectria, and resultant effect upon facility features important to
- safety. i i
I I t
- F-- ?-;ir-- _..;.u-- ;, b: shared:with=edditienebanite=et : : - -- - . I : M: l p ; ? ::-seiset-; e- ;;. _ :.,;ation ..? w . ..,,,,416,-Twim ical .^,, ; .";... l .... 1 I l
l
- seec=the reconnendatione of-Regulatory: Guide-iriir4*VTITO.. i, ,,,,,-. i ;ii . .i l l
I l i l 1 i NUREG 0452/STPEGS
- y575 3/4 3-52 COMPARISON (
i-
P;ga No. 61 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE # NOTES co 3/4.3.3.3 PAGE: 3/4 3-53.0 INSTRUMENTATION - SEISMIC INSTRUMENTATION .
- 1) Table 3.3-7 Instruments and ranges provided for Table 3.3-7. These reflect STPEGS design and nomenclature.
DESIGN
- 2) Table 3.3-7 Note (*) change to read "With reactor control room indication and alarm". This reflects STPEGS design.
ED 3) Table 3.3-7 Add note (**) which reads "At Seismic Monitoring Panel in Control Room, Unit 1". This reflects STPEGS design. O
.4 e
4
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G
(
. ~~
. ~~....... . ...~~~.--
. . . .. . n.
v . . TA8t.E 3.3-7 .
~
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SEISMIC MONITORING INSTRUMENTATION MINIMUM
+' O -
MEASUREMENT INSTRUMENTS
= ? OPERA 8t.E RANGE
=
M INSTRUMENTS AND SENSOR t.0 CATIONS a reas
- a
- 1. Triaxial Time-History Accele
, Qe o.,9 &
$ "b 6 1 i%.h9 *#
- a. FRCE ftEt.O CMThMMEN!"
(7twow Gaarn blOG fMNO r40N
, 3tt.-54} 9) 2%%
V 1 k 6. Ovrstor Fac.t of Garsinuter 4 39" 1
't $ c. ntit(ft,61-0")
56 UPPER LATGRAL ~
- 6. SuoMr(CL&4'-7K") +b% '
1
- 2. Triaxial Peak Accelerographs
- a. s.YiSEd'rDo") t3p 1 vispry 4 39
- 1
- 6. r(aacrostRCS EL 610 )
COLO LEG Cf t 31 1
. c. RC httWG '
htCEsE80(M9 A 3../< .T (stascs carataco
% ear,w ,y revoe TRIAMAL e oeu,cr 439
- 1 O_'
2 t.. - % '- 9 "). 4). riaxial seismic switches ##'* 13$ ,
# n
S.4 Tk AxlAL Sfesasc 7"n c4EA "# T33 . La 4 7 i
. (o,4. RESPoysE $PEC TitvM ' ANA L12[R $6 1*
-;--;d C = : a n = =:-_-- 1 : -
f 7.4. MMucrie T'aprRecogores*
- tse ,pe ro 5
7.,4. Puv8Aek SYs trss # # 259 1 A e 4 4 0' l $
*With reactor control room indication AW ALAKM.
O'
- 4 Ar sfisMic McNonnikG PAnst IN GNitas Room YW# A* o 3/4 3-53 NUREGOkbMPbd F575 COMPARISON
\
Page No. 62 06/13/85 COMPARI3ON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE f , NOTES 00 3/4.3.3.3 PAGE: 3/4 3-54.0 . INST'.UMENTATION t - SEISMIC INSTRUMENTATION DATA 1) Table 4.3-4 Instruments and ranges provided for s Table 4.3.4. This reflects STPEGE design. ,
- 2) Table 4.3-4 Delete notes (*) and (**) since they are not applicable to Table 4.3-4.
O l l l l e i O l
t
~~~ ~ ~ ~ ~ ~
l y -- ;..- . . . . -- : . . . . . . . . . . i TA8LE 4.3-4 , j - - SEISMIC MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS
- e. FHB FouxonTacW M R 3.A ANALCG i M R SA CNANNEL J. MEAS CHANNEL CHANNEL OPERATIONAL.
I INSTRUMENTS AND SENSOR LOCATIONS CHECK CALIBRATION TEST MGTER
- 1. Triaxial Tfee-History Accelerogmapa$s .
- a. FRrr*Firto R SA
- b. CearAwnnir BLtd. Vev4DATICM R $A
- s. 001stor Vect erletraosterS4t!t R SA
- . S/6 Upptr Lartaat brNr R SA
- 2. Triaxfal Peak Accelerographs
- a. SPrvf krt. / Int MX N.A. R r'4. A.
i
- b. REAcmR ktrL N.A. R N.A.
- s. CoLO llo er RC Pirtud N.A. R N.A. -
4 N.A. . R N.A.
- 3. p Sri!Courniero Thurson't Acentrestart m:snb1
- R ws. s A
- k. e,-.+cve4e4-4e4ewh -
4., s. 7kl4 x/A L 8(/f Ar tC [wiff # f M R $A
- s. # .w- w .
.S t. TRIARIAL Scisarte 7konart W N R sA ao
,p. Y W- T: "M:
- 4. 7.l- ld L .,..... % ;^.. Socorderes 4 w. REstetsr Srrerem Anyzra W N R SA l
4 wm e w
- 7. W. Mad ur7/c T4rt freccor#9 1 ream R $A
- 4. 1 tid. ** 9E.
RA. 8-4VBA(R Systrm ' 1 ETA. M R SA
- 9. / nt3. s' An.
o' l l
, =hWWW ,
-e!!Wittersectae: content --- irdeath, i
O F575 3/4 3 54 NU45#SEJWrPEcs COMPARISON .
Page No. 63 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE # NOTES 00 3/4.3.3.4 PAGE: 3/4 3-55.0 INSTRUMENTATION - METEOROLOGICAL INSTRUMENTATION
- NONE 1) 3/4.3.3.4 No changes.
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. w. v. . . . . . . . - . .
. .s.. . . . _ .
I INSTRUMENTATION .. J'l NETt0R0 LOGICAL INSTRUMENTATION , LINITING CON 0! TION FOR OPERATION 3.3.3.4 The meteorological monitoring instrumentation channels shown in Table . - 3.3-4 shall be OPEAA8LI. APPLICA8!LITY: At all tfaes. ..<- M:
- a. With one or more required meteorological monitoring channels inoperable for more than 7 days, prepare and submit a Special Report to the Commission pursuant to Specification 8.9.2 within the next 10 days outlining the cause of the malfunction and the plans for restoring
- the channel (s) to CPEAA8LE status. ,
I b. The previsions of SpecI1fications 3.0.3 and 3.0.4 are not app 1tcable. g- . SURVEILLANCE REQUfREMENTS . 4.3.3.4 fach of the above seteorological monitoring instrumentation channels ' # shall be demonstrated CPEAA8LI by the performance of the CHANNEL CHECK and CHANNEL CALIBRATION at the frequencies shown in Table 4.3-5. l a e b e O , 75T5 3/4 3 55 NUREG 0452/STPECS COMPARISON
Page No. 64 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE # NOTES co 3/4.3.3.4 PAGE: 3/4 3-56.0 , INSTRUMENTATION - METEOROIDGICAL INSTRUMENTATION DATA 1) Table 3.3-8 Meteorological Monitoring Instrumentation Instruments and locations provided.
- 2) Table 3.3-8 Note (*) 'The delta temperature channel may be taken out of service for up to eight hours provided a wind direction channel is operable to allow calculation of atmospheric stability.'
. 2A) The atmospheric stability is calculated by the method presented in the Bases Section.
s O
pv .. . _ TA8LE 3.3-8 ,, DRH
. METEOR 0LCGICAL HONITORING INSTRUMENTATION .
. MINIMUM INSTRUMENT LOCATION g
- 1. Wind Speed l PamARY en Guw *
- a. Mrrteeterucat bra Neefnel Elev. ~63 FrtT , 1
- g. MEntanke+w, b
- 2. Wind Ofrectfen -
rmmAny on DAIXup l
- a. M tTre c te re4r % vta Noetnal Elev. 13 MT 1 W. Mm. f
- 3. Afr Temperature - AT l faim44y
- a. wrfeeweis Twte Noetnal' Elev. 33- 195 FGT 1*
- # 4 flan 6 math, # .
i ( -
*The delta temperature channel may be taken out of service for up to eight hours provided a wind direction channel is operable to allow calculation of staospheric stability.
0 NUnto 0452/strEcs 3/4 3 sg COMPARISON r$Ts
Page No. 65 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE i NOTES co 3/4.3.3.4 PAGE: 3/4 3-57.0 , INSTRUMENTATION - METEOROLOGICAL INSTRUMENTATION
- 1) Table 4.3-5 Values provided consistent with those presented in Table 3.3-8.
. O i
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. . . ...... .. .. . . . .. ... . : .::-- :.;; . *: :~..
( - \
. TA8LE 4.3-5 .
METEOROLOGICAL MONITORING INSTRUMENTATTON
$URVEILtANCE REQUIREMENTS ,
CHANNEL CHANNEL IMsTRUMtwT CHECK CALIBRATION -
. 1. W1nd speed l
.I a. Noetnal Elev. D3MT O SA 3
, .-_ .. t-e.... . -
,1 2. Wind Ofrectfen
[ a. Meetnel tiew. 33 f*m* O SA . 8 y
- 3. Af r Temperature - AT O'
i, s. Meetnal Elev. ' 3 5- 195 4tr O SA s l t. 4 5 e e psis 3/4 3 87 NUREO 0452/STPECS COMPARISON
P ge No. 66 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE i NOTES co 3/4.3.3.5 PAGE: 3/4 3-58.0 , INSTRUMENTATION - REMOTE SHUTDOWN SYSTEM
- 1) 3.3.3.5 Action b. Deleted. Not applicable to STPEGS design.
, 2) 3.3.3.5 Action a renumbered b. due to deletion of Action b.
- 3) 4.3.3.5.2 Surveillance deleted. Not applicable to STPEGS design.
s e 4 IID-e 8 4 l l l l l 0 I
~
".................t;.... '*-~~...'..
~
g . . . .. . ..... . \ . INSTRUMENTATION , REM 0ft $NUTDQWN $Y$ TEM LIMITING CON 0! TION FOR OPERATION .
- 3.3.3.5 The Remete Sh6tdown System transfer switches, power, controls and ,
senttaring fnetrumentatten channels shown in Table 3.3-9 shall be OPERA 8LE.
. APPLICABILITY: M0083 1, 2, and 3.
Mt ,
- s. With the number of OPERA 8LE reesta shutdown sanf tering channels less
- than the Minfews Channels OPERA 8LE as required by Table 3.3 9, restere' the inoperable channel (s) to CPERA8LE status within 7 days, er be in HOT SHUTDOW within the neat 12 hours. ,
w==tHe ess:ene e tre. waere:4 emete-thetd, ree4te.4 , own:9yetee tteeneferrow+4eheeppemeeve
,t.re.=tes,e e . = w f(Pruit(e):te-optRA8LE status:within=Peeyel=eete=tfr*tIWfaf?IllMWWY*
C withtw:enerneet:38:6 b.% The previstens of Specf fication 3.0.4 are not applicable.
.g SURVIILLANCE REQUIREMENTS 4.3.3.5.1 fuh remote shutdown menttaring fnetroentatten channel shall be demonstrated CPERA8LE by performance of the CHANNEL CHt:K and CHANNEL CALIBRAf!0N operattens at the frequencies shown in fatte 4.3 6.
W 97879r8Stech'Aenetetshutdown-Systas*trenefee+ewitehr-powe mead-:: r '-- Ttfevit*tM1941pgtthe ectuated+eenpenentspaghef kte comenstretee4PEAmett' w >eee6*enee-p :-L'.eenenee f PsTs 3/4 3 la NUnto 0452/strtcs C0)irARISON ,
Page No. 67 , COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE f NOTES co 3/4.3.3.5 PAGE: 3/4 3-59.0 , INSTRUMENTATION - REMOTE SHUTDOWN SYSTEM
- 1) Table 3.3-9 Replaced with Table insert specifio to STPEGS.
e
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-0...:...G::-~.....'. - .
. -l. - . . .
. .. ~
O . DRAFT
. . i es
) ELE E .
3 .. kk I llw 1 unn; nun a
!L] l was nun m85 ,
3- { {en
. 45 unu suunn d$ tt unn O e 5
!. 55" en e!
l 1 il 1 8
-e - ! l g5 -
ll - s j3 -
> >l e.
- > s se ,
1 n iil si &
; t' 55 I I"5 1e 513{3 aj ~j5P4 .!
1 1 j] *5 2 1 )!'- t i !)d 531x.s , ~i 3 . tj
.i33 1
p,e
=s E 3 3tt 3 3 1 =E ._
s >\ .8j3
=
i 3 A,laati 2' c-3. E j g, F:1l JEaa 5' l, fulfill]a 883 Ja a js j 0 !$ - aussasa.uga s e sa_la u 1 u ass g away t<Unto 0452/strtes
, P5Ts 3/4 3 33 conrAnisoft
Page No. 68 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECM. SPECS. NOTE TYPE NOTE f NOTES co 3/4.3.3.5 PAGE: 3/4 3-59.5 . INSTRUMENTATION - REMOTE SHUTDOWN SYSTEM ED 1) Table 3.3-9 Title changed to ' Remote Shutdown Monitoring Instrumentation'. This reflects STPEGS nomenclature. DATA 2) Table 3.3-9 Instrumentation provided. (STS provides illustration rather than requirement.) DESIGN- 3) Table 3.3-9 Transfer Switches deleted. Pot applicable to STP design. DESIGN- 4) Table 3.3-9 Control Circuits deleted. Not applicable to STP design.
- 5) Table 3.3-9 Identify as South Texas Project specific.
. G O
4
' TAR 2 3.M RPMfft SoffDrtet N?ffMtfM: INmtWNIATT(N Sourn TOAs Psaatt1 .Sno voc ' MortMM IWAtrtff 1tXAI.180 OWNIIA 14C4fffW W OWM25 WIMR2 WHE9ff_ asp'- e rs
- 2 (start , note) ;L
- 3. h h- rha 2 oktended amp) 2 Asr - ers
- 2. anseeer trt, nr.dwr wicette amertar tet, 'A &
Mtshener el Asr - tpes
- 3. imense om Let t , .. = ?.A.
Asup t/ tan,
- a. m ta Asr - ers I/ta, 1/In, 1/lsm,
- b. He las Asr - @ rs +
Wisseen Asp - ties + 3 2,
- 4. amareer castet Pnman - MAe near Wiesten ei 2 ses,e
- 5. Presortser humane W - @rs 4 2 Asr - pre + 4 &
- 6. Poemmetser meer im=1 wtanwee for 2 shovela
\ Asr - grs 4/etam 11,= 2/mes 11,e 7., seem the P wane S.* seems esemeer meer tent. Aar - prs e 4()/sc) $* udemsgp wieswn p w er Flow Ante
- 9. Austil e y ued Asr -Ipre + 4Q/sc) 3" e wt e.,.
Aar - e rs 3 g
- 98. Asd11ery pum6meer soorge 1 6 useer let
\ HUnto 0452/8TPtcs C0ftPAR180N
Page No. 69 06/13/85 COMPARISON OF NURIG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE f NOTES co 3/4.3.3.5 PAGE: 3/4 3-60.0 . INSTRUMENTATION - REMOTE SHUTDOWN SYSTEM
- 1) Table 4.3-6 Revised to reflect Table 3.3-9.
Surveillance intervals are consistent with the RTS requirements established by WCAP-10271 and associated SER. O O.
.. . . . .. ~ 7 . . .. 7.7 . . ~ * ' '
- ~ " '
(o. . * . ,; . . . ........!7...*.**..... . . . . . . . . . .
*
- f g .il <il 4 g .,ia s......,,,, , ,,.e 8
i
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4 , W w g 5
- I :5 .e li g l :
b- 3 -
; . 5 i:
l i l i . I ll n b i t O i t u O I'.l jag}}'l ' ' lb hk[fff ll I!$ lll d i N Mtc 0452/87PEC5 ' plTS 3/4 3=H COMPAR150tl ____._______________._____t____________.___.___.._
Page No. 70 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. EPECS. NOTE TYPE NOTE # NOTES co 3/4.3.3.6 PAGE: 3/4 3-61.0 . INSTRUMENTATION - ACCIDENT MONITORING INSTRUMENTATION
- 1) 3.3.3.6 Action a. Changed to 'With the number of
... Table 3.3-10, perform the action statement as
- indicated in Table 3.3-10.' This is required to accurately reflect appropriate action statements to specific instruments.
- 2) 3.3.3.6 Action b. Changed to 'With the number of... Table 3.3-10, perform the action statement as indicated in Table 3.3-10.' Required to accurately reflect appropriate action statements to specific instruments.
DE'SkGN- 3) 3.3.3.6 Action c. Deleted. Not applicable to STP design. ED 4) 3.3.3.6 Action d. renumbered c. due to deletion of C. s I O
4 INS ENTATION , ACCIDE T MONITORING INSTRUMENTATION
=_
LIMITING CONDITION FOR OPERATION 3.3.3.6 The accident monitoring instrumentation channels shown in Table 3.3-2.3 -
? shall be OPERA 8LE.
I- APPLICA8ILITY: MODES 1, 2, and 3. ACTION: e a.' With the number of CPERA8LE accident monitoring instrumentation channels less than the Total Number of Channels shown in
, l Table 3.3-10. - .;t.. tM 5- ; rabt. A !(-) L; ?E?f.5LE
^ ?= =1thi%. e- t-59_t=least= HOT =43AN08V=w$th4w=ths:
nuxt:$ heurs and in at:4 east NOT:8 HUT 00WN stte.in .n: i;i i - .--
-4=heeEsp PERFORM THr ACTION starracur As #NOiCAfra IM .7A3LT 3.3=j . *
- b. With the number of OPERA 8LE accident monitoring instrumentation '
channels wti tt-e-.eit . .:.M.';t :x;; xt % ii . -.:'_. , .... ..._
- i. *sM m" ' ' ' -e m.-e.4ta+ ! ee _.. ! t . U;;-- _;; t:1.___.. . i .,:. . . !; 3
. g e.;Lw isti s nitorv=end=the n--ter -:;-:-!_..; cd'_ti;; L._!
less than the Minimus Channels OPERA 8LE requirements of
. Table 3.3-10, westore=4he-4noperabi,-c.ba. WW#EE .L:-_
N-4-.t- least-NOT=$7AN087 ;! U-!= L: .._ i 4-me -in-w east-HOT =$HUTDOWM-wRhi ; t- f-il;-i- 5 : ; .-.. PERFORM THE ACTION srAr Ment As wortArro w Taur 3.3-10. e r fhg t gr p nt a i ya n 1 , r acto ' r dia io 1 aan e , n uf y f
. i'ry ns (( y t,i11 ea a te d popri r (e ( 7 s d s Ih r J hn 1s P s. w th n s r la it p i.1 R t e islon p f c f. i
., /t 4d : t r 'de etwn l a 0p .! '
- 44. The provisions of Specification 3.0.4 are n'o t applicable.
SURVEILLANCE REQUIREMENTS
'4.3.3.6 Each accident monitoring instrumentation channel shall be demonstrated OPERA 8LE by performance of the CHANNEL CHECK and CHANNEL CALIBRATION at the l frequencies shown in Table 4,3-7.
W STS 3/4 3-61 NUREG 0452/STPEGS COMPARISON
Page No. 71 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE i NOTES co 3/4.3.3.6 PAGE: 3/4 3-62.0 , INSTRUMENTATION - ACCIDENT MONITORING INSTRUMENTATION DATA 1) Table 3.3-10 Accident Monitoring Instrumentation provided. (STS Instruments listed as illustration only not requirements.) DATA 2) Table 3.3-10 ' Action' Column added to table in order to provide proper Action statements for STPEGS. See 3.3.3.6, page 3/4 3-61.
- 3) 3.3-10 Delete line Items 10-Boric Acid Tank Level, 13-PORV Position Indicator, 14-PORV Block Valve Position Indicator, 15-Safety Valve Indicator.
Not applicable to STPEGS.
- 4) 3.3-10 Add new Item 11 - Auxiliary Feedwater Flow and renumber remaining items. This reflects STPEGS design. <
O
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TABLE 3.3-10 - - iy p
- CCIDENT NONITORING INSTRtBENTATION a
en ' 1 TOTAL . MINilt90 . l
. No. Or CNAmELS .
INSTRUNENT iiii-ir
- CHANNELS OPERABLE ACTION
- 1. Containment P: . f3 1 l 2
- 2. Reactor Coo' .w alaIm1 Teererature - TH0T (Wide Range) El/ Loop 1/ Loop 3
~
- 3. Reactor Co. ,nt b 1 sperature - TCOLD (Wide Range)' El/ Loof 1/Loor 3 ;
I 'I i
- 4. Reactor Coo' 7t Pressure - Wide Range /I . 1
- 5. Pressurizer *er Level
. f3 ,
1 . /
- 6. Steam Line f ,sure 3f/ steam generator 1/ steam generator / *
{
'i' 7. Steae Generstor Water Level - Narrow Range 1/steae Osnerator 1/ steam generater I C
! 8. Steam Generator Water Level - Wide Range 1/staan'generatar 1/ steam generator 3
- 9. Refueling V ' r Storage Tank Water Leve? '# 3 1 /
10;:=8srts:Asts: ! ' Inn L. _i 2 i j{ TANK jfANK / 10.11. Aux 114aru Feer **'* 2... . w T4uKLuft - -
. . 1= - u ? r ; :--t=
II. Alnut . FLost */FEEli LINC 1/ FEED LIAlf 3
- 12. React .i 5ystem Subcooling Margin Monitor 2 1 3 ,
u --ruhit . len idic:te- 2/veiv: 1/vei - l tt m=-W 55-3 V i : ?=:ii!-- 5-Ji_:irr 2/ :i : 1/ val:: , l 1 . OO 35. icizi; 9:i:: b iiiz !adt_ i=r I/:=I-- E':-I-- l . I6 o Zi
- gg
> *'l3.Mr. Containment Water Level (Narrow Range) ,
2 1 un - O hl4.W. Containment Water Level (Wide Range) 43 1, / , I trj
- I 8 ,
Page No. 72 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE i NOTES 00 3/4.3.3.6 PAGE: 3/4 3-63.0 . INSTRUMENTATION - ACCIDENT MONITORING INSTRUMENTATION DATA 1) Table 3.3-10 Accident Monitoring Instrumentation provided. (STS instruments provided as illustration only, not requirements.) DATA 2) Table 3.3-10 ' Action' required added to table as page 3/4 3-63A.
- 3) 3.3-11 Delete line items 19 - Unit Vent - High Range Noble Gas Monitor and item 23 - Reactor Coolant Radiation Level Monitor. Not applicable to STPEGS.
- 4) 3.3-11 Add line items 19 - Neutron Flux (Extended Range), 22-RCS Pressure (Extended Range) , 23-Steam Generator Blowdown Radiation Level and 24-Neutron Flux, Startup Range. This reflects STPEGS design.
. IID-
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~
1
. . . :,. . . . . . .c . ..
.s .
TABLE 3.3-10 (Continued) $ ACCIDENT DENT!TORING. IN57213ENTATION . ' vs , TOTAL MINIIRM e 5
- No. OF CHAltlELS
- insinunEur tmm-mm=vanyT= cnAnufts SPEaA8tE ACTION Exar j 1538. Sa core 4Thermocouples -
4/esumquadrant/Ta4v2/earaquadrant/ TRAL #f .,
-15. 5.it h..t - hi;;h = - ^ L- :: ::_..ii ;- 5.5. i .
LINE
- f(pW. Steama tettsf 400 gly:Ausgs: Radiation Monitor EE.1/Straaf Lius 1/ steam line 3 l 174 contalament$$EEIMrs - High Rage Radiation Monitor **. 2 1 2 [
I f.fE. Reactor Vessel 1dutur Level IusrainrvrAT10# SYSTEpt(RVI.15) 2 . 1 2 f u
-::. :=: .:- -:--h-- - i:- i= _i . ..;-- =.e. I 1 k 19 NEurnow Ftux (Exrswoco RAnas) 2. I 2. -
l w 20. CourhuMMENT HYDR 06EW CowcENTRATioW S 'l . E zl. COMTAIMMENr PesssvRE (EXTEWDED RANGE) E I 2
- 22. RCS PRessune (Exrtuneo RAust)' 2. I 2.
- 25. 5inM GenennToR BtowDown RADIAT10N lEygt. I l 3 i 24. NEUTRON lQUy - STARTUP SMGE E I 2 ,, ..
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Page No. 73 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE i NOTES 60 3/4.3.3.6 PAGE: 3/4 3-63.5 INSTRUMENTATION - ACCIDENT MONITORING INSTRUMENTATION DATA 1) Table 3.3-10 ACTION STATEMENTS added to Table. s - e O
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l l 1 TABLE 3.3-10 (Continued) l ACTION STATEMENTS ACTION 1 - s. WiththenumberofOPERABLEchannelsonebessthanthe I Required Number of Channels requirements, restore the ! inoperable channel to OPERABLE status within 14 days, or be in at least HOT SHUTDOWN within the next 12 hours.
- b. With the number of OPERABLE channels two less than the Required Number of Channels requirement, restore at least one inoperable channel to OPERABLE status within 7 days, or i be in at least HOT SHUTDOWN within the next 12 hours. l 1
- c. With the number of OPERABLE channels less than the Minimum Channels Operable requirement, restore at least one ,
inoperable channel to OPERABLE status with 48 hours or be in l HOT SHUTDOWN within the next 12 hours. ACTION 2 -
- a. With the number of OPERABLE channels one less than the i Required Number of Channels requirements restore one inoperable channel to OPERABLE status within 7 days, or be
*in at least HOT SHUTDOWN within the next 12 hours.
- b. With the number of OPERABLE channels less than the Minimum ;
Channels Operable requirements, restore at least one ' inoperable channel to OPERABLE status within 48 hours, or be l in HOT SHUTDOWN within the next 12 hours. ., g l ACTION 3 - With the number of OPERABLE channels less than the Minimum i Channels Operable requirement, restore at least one inoperable ! channel to OPERABLE status within 48 hours, or be in HOT SHUTDOWN within the next 12 hours. I l l NUREG 0452/STPEGS COMPARISON 3/4 3-(o3 A
I Page No. 74 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE # NOTES oc 3/4.3.3.6 PAGE: 3/4 3-64.0 . INSTRUMENTATION - ACCIDENT MONITORING INSTRUMENTATION DATA 1) Table 4.3-7 Revised to be consistent with Table 3.3-10.
- 2) Table 4.3-7 Add new Item 11 - Auxiliary Feedwater Flow and renumber remaining Items. This reflects STPEGS design.
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TABLE 4.3-7 ,' 39 3 ACCIDENT NONITORIIIG INSTMOENTATION SilRVEILLAIICE RE0utREsturs CHAlelEL CNA001EL - litSTRUNENT iii_i ;1: ? ".2,) CNECK - CALIBRATIell ,
- 1. Containment Pressure .
N R l SYSTEM M A
- 2. Reactor coolantaamsta Temperature - Tygy (Wide Range)
- 3. Reactor Coolant M R j Temperature - Tg (Wide Range)~
4 Reactor Coolant Pressure - Wide Range N - 3
- 5. Pressurizer Water Level N R
- 6. Steam Line Pressure N 3
{ , l Y 7. Steam Generator Water Level - Na.rrow Range M R 6s D 8. Steam Generator Water Level - Wide Ran08 N R - i N
- l
. 9. Refueling Water Storage Tank Water Level , 3 -
E if. _. 3 . .t ! f ? - - 5 : ! :i !--. L . _? N - 10.11. Auxiliary Feedwater h SrogAGr TAAIK LutL M a -l 8l. AussLsnRY retowartR Fran! .
- 12. Reactor Coolant Systes Subcooling Margin Monitor N R i i 'I
- . :- = ?= i ti == imi= = tr-- N -
k- j' E li :: - E? ^- 9:i:: ?::ii!-- ! ^! 21-- N-- -
-u- ,. {
nE
. 35. 5:fri- 9:i:; Lii!== i-le_ ir: ==cN- 3: , jl 35 o .
. ,i k hl3,4fr. Containment Water Level (Narrow Ranes) ,
N R .
$2 2: :;j 14.97. Containment Water Level (Wide Range) N R.
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l l Page No. 75 L 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE # NOTES I l l oc 3/4.3.3.6 PAGE: 3/4 3-65.0 , INSTRUMENTATION - ACCIDENT MONITORING INSTRUMENTATION DATA 1) Table 4.3-7 Revised to be consistant with Table 3.3-10. ! DESIGN- 2) Table 4.3-7 Add line Items 19 - Neutron Flux (Extended Range), 20-Containment Hydrogen Concentration, 21-Containment Pressure (Extended Range), 22-RCS Pressure (Extended Range), 23-Steam Generator Blowdown Radiation Level. I 2A) Add 24-Neutron Flux, Startup Range. This reflects STPEGS design. DESIGN 3) Table 4.3-7 Note (*) deleted. Not applicable to STPEGS design. o O
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. 1 1 - TABLE 4.3-7 (Centlemed) ,
ACCleENT teNITORIlgi IIIsistBENTATION SURWEILLANCE REQUIREENTS , 1 **
. CIWWeEL. CNMeEt.
i INSTSUMENT 6 T CHECK CALIOEATION ,. 1 her '
- I j I M . in Cere4Thermecouples
- M E ?
I n. - _ ;; . [.. . . - - - It == - - - - -- " R 2-LsNG 1 IOEF. Steasjauttef -WRadiatten llentter M R . 17.g. Cantainment$ATM-e - uteh anos endiatten Montter M l e , J Y.sE neacter vesui issear Level Iummrvreov Sysnrn(RvuS) M a e 1 l ::. :::n- -: --- =: ~ ^:i'= ^ . _ i ^^ . " i- M----=2=- = = n =. , w k 19 Neurnow Ft.ox (Exrtuoco Riuss) n e . . u l
- M R-i 20. COWTNMMENT HYOR06EW COwCENTRADOW
! b 21 CarAIMMEur Pktssune (erraoro Ruse) M R l - M # l 22. RCS Pnessune (Esrtuneo RAust)
^i R
- 23. S7EAM GENERMDR bwpoWM MADMnog l.EVEL l
! 24. Neurnon Vun- t Srnaror RmGe W R
'1.
l 5 O$ . . . 2 . , .-- .$.^-__7 L i' _ ..,. i _i.". ^ "
^
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P;ge No. 76 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPEOS. NOTE TYPE NOTE i NOTES co 3/4.3.3.7 PAGE: 3/4 3-66.0 . INSTRUMENTATION - CHLORINE DETECTION SYSTEMS
~
DESIGN- 1) 3/4.3.3.7 Deleted. Not applicable to STPEGS design. STPEGS Control Room does not have isolation upon chlorine detection.
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O.. . . : . . u. . ESTRUMENTATION [ (( / b. CHLORINE DETECTION SYSTEM 5 LIMITING CONDITION FOR OPERATION l 3.3.3.7 Two independent Chlorine Detection Systems, with their Alara/ Trip Setpoints adjusted to actuate at a chlorine concentration of less than or .l equal to 5 ppe,.shall be OPERA 8LE. APPLICA8ILITY: All MODES. ACTION: ,
- a. With one Chlorine Detection System inoperable, restore the inoperable l
, system to OPERA 8LE status within 7 days or within the next s hours initiata and maintain operation of the control Roos Emergency Ventilation System in the recirculation mode of operation.
- b. With both Chlorine Detection Systaes inoperable, within 1 hour .
inttfate and maintain operation of the Control Room Emergency - Ventilation System in the recirculation mode of operation.
\
- c. The p'rovisions of Specification 3.0.4 are not appitcable. .
t SURVEILLANCE REQUIREMENTS 4.3.3.7 Each Chlorine Detection Systas shall be deeonstrated OPERABLE by l
. performance of a CHANNEL CHECK at least once per 12 hours, an ANALOG CHANNEL OPERATIONAL TEST at least once per 31 days and a CHANNEL CALIBRATION at least once per 18 months.
l 9
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. I PSTS 3/4 3-66 NUREG 0452/STPEGS COMPARISON
f Page No. 77 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE # NOTES 00 3/4.3.3.8 PAGE: 3/4 3-67.0 , INSTRUMENTATION - FIRE DETECTION INSTRUMENTATION
- 1) 3/4.3.3.7 " Fire Detection Instrumentation". This r system is currently under development. The information will be provided LATER.
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INSTRUMENTATION - FIRE DETECTION INSTRUMENTATION LINITING CONDITION FOR OPERATION . 7
'. 3.3.3./ As a minimum, the fire detection instrumentation for each fire detection
, zone shown in Table 3.3- n shall be OPERA 8LE. -
APPLICA3ILITY: Whenever equipment protected by the fire detection instrument
'T is required to be OPERA 8LE. . ; .) ACTION:
- a. With any, but not more than one-half the total in any fire zone, Function A fire detection instruments shown in Table 3.3-11 inoperable, restore the inoperable instrument (s) to OPERABLE status within 14 days or within the next I hour establish a fire watch patrol to
- inspect the zone (s) with the inoperable instrument (s) at least once .
per hour, unless the instrument (s) is located inside the containment, then inspect that containment zone at least once per 8 hours (or
!. . ~ monitor the containment air temperature at least once per hour at -
thelocationslistedinSpecification4.6.1.g.
- b. With more than one-half of the Function A fire detection instruments in any fire zone shown in Table 3.3-11 inoperable, or with any .
Function 8 fire detection instruments shown in Table 3.3-11 inoperable, or with any two or more adjacent fire detection instruments shown in Table 3.3-11 inoperable, within I hour establish a fire watch patrol' to inspect the zone (s) with the inoperable instrument (s) at least *' once per hour, unless the instrument (s) is located inside the contain-ment, then inspect that containment zone at least once per 8 hours
] (or monitor the containment air temperature at least once per hour atthelocationslistedinSpecification4.6.1.f). E
,i c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS
- 4. 3. 3./.7 1 Each of the above required fire detection instruments which are accessible.during plant operation shall be demonstrated OPERABLE at least once
, per 6 months by performance of a TRIP ACTUATING DEVICE OPERATIONAL TEST. Fire
;j detectors which~are not accessible during plant operation shall be demonstrated OPERA 8LE by the performance of a TRIP ACTUATING DEVICE OPERATIONAL TEST during 4
each COLD SHUTDOWN exceeding 24 hours unless performed in the previous 6 months.
, 4.3.3./.7 2 The NFPA Standard 72D supervised circuits supervision associated with the detector alarus of each of the above required fire detection instruments shall be demonstrated OPERA 8LE at least once per 6 months. -
97974rer8=h, .-vyervised-cin.utts; associated with-detector. alarusr-betwesw e-- j e=i.rument-and:the- control room sham-be-demonstrated-CPERA8Lf:st =tesst _..a c. x,__ NUREG 0452/STPEGS
- ysTs 3/4 3-67 c0MPARISON
, , - _ ,,,._______r_ _ _
_._m, , . , _ _ , . . _ , _ _ _ .
Page No. 78 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE f NOTES co 3/4.3.3.8 PAGE: 3/4 3-68.0 . INSTRUMENTATION - FIRE DETECTION INSTRUMENTATION
- 1) Table 3.3.11 " Fire Detection Instruments". The Fire Dett10 tion System is currently under development. The information will be provided (LATER).
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FIRE DETECTION INSTRUMENTS
**** . . . TOTAL NUMBER .
INSTRUMENT LOCATION OF INSTRUMENTS * . [ Illustrative] NEAT FLAME SMOKE Tx79) TiUy} . Tx7j7
- 1. Containment **
- a. Zone 1 Elevation -
- b. Zone 2 Elevation - ~ '
- 2.
- Control Roos
- 3. Cable Spreading -
- s. Zon's 1 Elevation
- b. Zone 2 Elevation
- 4. Computer Room
- 5. Switchgear Room THis svsTEM 15 CURRENT 1Y UNDER DHELCMNT.
THE INFORMATscN WILL BC PROVIDGD (LATER),
- 6. '* Remote Shutdown Pane 1s
- 7. Station Battery Rooms -
- 8. Turbine ..
- a. Zone 1 Elevation
- b. Zone 2 Elevation
- 9. Diesel Generator
- a. Zone 1 Elevation .,
- b. Zone 2 Elevation ,
- 10. Safety-Related Pumps -
- a. Zone 1 Elevation
- b. Zone 2 Elevation '
I
- 11. Fuel Storage
- a. Zone 1 Elevation !
- b. Zone 2 Elevation j
[ List all detectors in areas required to ensure the OPERA 8ILITY of safety-related equipment.J
*(x/y): x is number of Function A (early warning fire detection and notification only) instruments.
y is number of Function 8 (actuation of Fire Suppression O Systems and early warning and notification) instruments.
**The fire detection instruments located within the containment are not 1
required to be OPERA 8LE during the performance of Type' A containment leakage
- rate tests. NUREG 0452/STPEGS p5TS 3/4 3-68 COMPARISON
- , .-y- - - , , . , - , . - - - _ _ - - - - - , ,
Page No. 79 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE i NOTES co 3/4.3.3.9 PAGE: 3/4 3-69.0 . INSTRUMENTATION - LOOSE-PART DETECTION SYSTEM
- 1) 3.3.3.9 renumbered 3.3.3.8 due to deletion of 3.3.3.7.
- 2) 4.3.3.9 renumbered 4.3.3.8 due to deletion of 4.3.3.7.
- 3) 4.3.3.8.b. Changed tio "An ANAIDG CHANNEL CALIBRATION TEST except for verification of setpoint at least once per 31 days. This is consistant with the approved Test Specs of other
. plants (SNUPPS, Catawba & McGuire).
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5 I 1 Page No. 81 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE # NOTES
** 3/4.3.3.10 PAGE: 3/4 3-70.0
- INSTRUMENTATION - RADIOACTIVE LIQUID EFFLUENT MONITORING INSTR.
ED 1) 3/4.3.3.10 renumbered 3/4.3.3.9 due to deletion of 3/4.3.3.7.
- 2) 4.3.3.10 renumber to " Analog Channel Operational Test" to read " Digital Channel Operational Test".
STPEGS design provides a digital system. See Definitions Section. l. { 1 6 l l l l l l l NUREG 0452/STPEGS ) COMPARISON l
. ~..- .............:...
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INSTRUMENTATION .
. LOOSE-PART OETECTION SYSTEM LIMITING CONDITION FOR OPERATION 1 :
3.3.3./ The Loose-Part Detection Systes shall be OPERA 8LE. . APPLICA8ILITY: MODES 1 and 2. ACTION:
- a. With one or more Loose-Part Detection Systes channels inoperable for }
acre than 30 days, prepare ,and subait a Special Report to the - Comeission pursuant to Specification 6.9.2 within the next 10 days outlining the cause of the sa1 function and the plans for restoring the channel (s) to OPERA 8LE status.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 2 . 4.3.3./ Each channel of the Loose-Part Detection Systems shall be demonstrated, OPERA 8LE by performance of:
- a. A CHANNEL CHECK at least once per 24 hours, .
-
- gegprrop ytrgATiw oF SMW
'l b. An ANALOG, CHANNEL CPERATIONALgTEST at,least once per 31 days, and g I
- c. A CHANNEL CALIBRATI.1N at least once per 18 months.
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~-
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- INSTRUMENTATION ,
. LOOSE-PART DETECTION SYSTEM LIMITING CON 0! TION FOR OPERATION 1 :
3.3.3./ The Loose-Pert Oetection Systes shall be OPERA 8LE. . APPLICASILITY: MOES 1 and 2. . t With one or more Loose-Part Detection system channels inoperable for l s. sore than 30 days, prepare ,and subett a Special Report to the - Commission pursuant to Specification 8.9.2 within the next 10 days outlining the cause of the salfunction and the plans for restoring the channel (s) to CPERASLE status. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. b.
- 0. SURVEILLANCE REQUIREMENTS
.o 5-R .
4.3.3/ Each channel of the Loose-Part Oetection Systems shall be demonstrated, OPERA 8LE by performance of:
- a. A CHANNEL CHECK at least once per 24 hours, .
- gem wsnificAhey of M*MW \
'l b. An ANALOG, CHANNEL QPERATIONALgTEST at,least once per 31 days, and g
I c. A CHANNEL CALIBRATION at least once per la months. f
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,,,- 3/4 3-69
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Page No. 81 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. - NOTE TYPE NOTE f NOTES - co 3/4.3.3.10 PAGE: 3/4 3-70.0
- INSTRUMENTATION - RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTR.
ED 1) 3/4.3.3.10 renumbered 3/4.3.3.9 due to deletion of 3/4.3.3.7.
- 2) 4.3.3.10 renumber to " Analog Channel operational Test" to read " Digital Channel Operational Test".
STPEGS design provides a digital system. See Definitions Section. n e, le l l t ( l i l I NUREG 0452/STPEGS COMPARISON i - - - - - - _ _ _ _ _ _
S
. . . . s . . .. :: .t.: s v
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- TABLE 3.3-12 .,
.f l$
g RADIOACTIVE L10073 EFFLUENT NDIIITotlNG IIISTAtBENTATICII sn MINilast CHANNELS IIISTAtsfMT OPERABLE ACT1000
- 1. Radleactivity Monitors Providias Alarm and Automatic Terminetten of Release l
- a. Liquid Raesasta Effluent Line 1 52,6 .
fft _ i. - 1 M .' L_14es-SutWEleer: trains) ' . : Effluent-t _ i E-
- q AtarEF95$10BG=7TW9dtNW: I y 2. - - :=iivity : 51 . .^. -
a -gggemogwm .
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s.,
;f
-. " . 6cw=tseter=Systus-iffE _1-iis i .;
- t. - ;-- T-Ceeltag tenter:$ystaEFifflanNE 11=. i 3?-
i i
- 2. -- _- -
, - =- . - - = _ , ___ ___ ____
+r--6t44GFSEREF3 tar:S%sudeus-$ff%eent=%5ms: '
faiteewats=turttaEFliWF) i 5 ,
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15Meysats 85 ites L cc) _-
= .
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Flaw Rate Measurement Devices . l Oo 2.f. ' Eos~ a. Liquid Radweste Effluent Line 1* N29 l e
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N O CA M 6 fflaunt tine i ses- .
'Z. *4 -
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Page No. 83 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE i NOTES c0 3/4.3.3.10 PAGE: 3/4 3-72.0 INSTRUMENTATION - RADIOACTIVE LIQUID EFFLUENT MohITOhING INSTR. DESIGN- 1) Table 3.3-12 Item 5 Radioactivity Recorders deleted. Not applicable to STPEGS design. 4
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l Page No. 84 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE # NOTES co 3/4.3.3.10 PAGE: 3/4 3-73.0 . INSTRUMENTATION - RADICACTIVE LIQUID EFFLUENT MONITORING INSTR.
- 1) Table 3.3-12 Action Statements. Action Statements
. not applicable to STPEGS have been deleted and the remaining Action Statements have been renumbered.
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TA8LE 3.3-1* (Continued) , DMFI.
, ACTION STATEMENTS .
28
- ACTION SE - With the number of channels CPERA8LE less than required by the Mfnfaus Channels OPERA 8LE requirement, effluent releases via this pathway any continue for up to 14 days provided that prior to inttiating a release:
's. At least two independent samples are analyzed in accordance with Specification 4.11.1.1.1, and
- b. At least two technically qualifted members of the facility
.. staff independently verify the release rate calculations
.'. and discharge Ifne valving.' .
Otherwise, suspend release of radioactive effluents via this pathway. . ACT30N:38:m: Wf}d /r e ch4nds PtfA8 1 ss a n f y'yhe nf an ' Is PEpASI qu nt of lu t i v4 f p as co ti ve ro td g ab .es re y d . o t to tiv ty or up ys at I r fai e
- to to of th 1 7 ic Cu fe/ 1:
- e. A 1 t c pe 12 ,ou en he spe ff a tv ty of e o a c la t i g at an 0 01 ic oc r1 /gr S I t I- 1, er
- b. A le t c pe 24 u s Qen aneof.e equal oc ft ac fv ty f%
ec c elapt i leu to
- 0. ac uri /gr SE IA I 131 N Wi umb r e chehn OPRA) I se a n uf red yte nf 1s CPE n ynent of lu t I se vi
; hf p as to in f to dpys re ide t t, t
. le e per ur 1 91 to a an ly d ti ity t e tfo of ne u'$e/ .
1/mit f th n ACTION = With the number of channels OPERA 4LE less than nquind by the Minfa n Channels OPERA 8LE requi n ment, effluent releases via this pathway any continue for up to 30 days provided the flow rate is estimated at least once per 4 hours during actual n1 eases. Pump perfonnance curves generated in place may be used to estf asta flew.
*Ct902 Sk- Wf e uebe of he/nel QP % 1 se h n ut d i n' s t LE mq r nt of 1 t ole ses f p y y on n to up d ys ro de th
- g. ;,,. y d, . n. at
.., . .. ., ,. ,s .,,n NUREC 0452/STFECS r$7s 3/4 3-73 COMPARISON
Page No. 85 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE 4 - NOTES co 3/4.3.3.10 PAGE: 3/4 3-74.0 . INSTRUMENT 1. TION - RADIOACTIVE LIQUID EFFLUENT MONITORING INSTR.
- 1) Table 4.3-8. This Table have been revised to correspond to Table 3.3-12. There are no changes in the requirements.
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ll e t> ... l i el.D 8- 5" i. ay Ib . .hl a .i,. u, , , , , ,. y ].E
. ip
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e=* M til 4 t NUREG 0452/STPEGS 3/4 3-74 c0MPARISON rsTs a
Page No. 86 G6/13/35 - COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE f NOTES co 3/4.3.3.10 PAGE: 3/4 3-75.0 , INSTRUMENTATION - RADIOACTIVE LIQUID EFFLUENT MONITORING INSTR.
- 1) Table 4.3-8. This Table has been revised to correspond with Table 3.3-12. There are no changes in the requirements.
e
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- em eo e
Page No. 87 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE i NOTES co 3/4.3.3.10 PAGE: 3/4 3-76.0 INSTRUMENTY.' ION - RADIOACTIVE LIQUID EFFLUENT MohITORING INSTR. t
- 1) Table 4.3-8 Notation (1) - Change "THE ANALOG CHANNEL OPERATIONAL TEST....." to read "THE DIGITAL CHANNEL OPERATIONAL TEST....". STPEGS has
, Digital Radiation Monitoring Instrumentation.
- 2) Table 4.3-8 Notation (2) - Delete. STPECS has Digital Radiation Monitoring Instrumentation.
- 3) 4.3-8 Notations (3) and (4) renumbered (2) and (3) due to deletion of notation (2).
. G
.t t.
O t O l l 1 ?
I TABLE 4.3-8 (Continued) . TA8LE NOTATIONS DIGITAL * (1) The m RCOG 3 CHANNEL OPERATIONAL TEST shall also demonstrate that automatic isolation of the following of this pathwayconditions andexists: control room alars annunciation occur if any
- a. Instrument indicatas measured levels above the Alars/ Trip 5etpoint, op
- b. Circuit failure, or'
- c. Instrument indicates a downscale failure, or
- d. Instrument controls not set in operata mode.
r r. _ :=.
~_1. __Z
- =:ec = : --- _._- _ _. _; ._ 7 , 7 1 __g
_ _~ T '_ ' ] _- __ _ _ _ . _ _ _ _ _. ._ .__ M. _ - ~. _ _ _ ~_1_2.Z ~_ - -_ _.,__'~{._.4._.]..~..,'__".._4..';._.._--'_-,: _. . . ._.____.._._;a_;
; m - g : - , ---
.. .e....n.. . .e... , _._._
i; ^ -- .
- l b M5 5 2; .: . ' '_ M _. _ , _ .
The initial CHANNEL CALIBRATION shall be perfomed using one or more of the reference standards certified by the National Bureau of Standards (N85) or using standards that have been obtained from suppliers that participate b in sensurement assurance activities with N85. These standards shall permit
' calibrating the system over'its intended range o'f energy ~and measurement range. .For subsequent CHANNEL CALIBRA ON, sources that have been related to the initial calibration shall be us -
3 . dQ CHANNEL CHECX shall consist of verifying indication of flow during periods of. release. CHANNEL CHECX shall be made at least once per 24 hours on days on which continuous, periodic, or batch releases'are made. l l 1 O : l W STS 3/4 3-76 NUREG 0452/STPEGS COMPARISON
Page No. 88 06/13/85 COMPARISO1T OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE f NOTES 00 3/4.3.3.11 PAGE: 3/4 3-77.0 . INSTRUMENTATION - RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTR.
, 1) 3.3.3.11 Renumbered 3.3.3.10 due to deletion of
,- 3.3.3.7 iy
- 2) 4.3.3.11 Renumber 4.3.3.10 due to deletion of 3.3.3.7.
l l l l s O
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9
~
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,, , .. . . . . . .. . . . . . . . . . . . . . - . - . . T.C ..~ . . .L._'.-
INSTRUMENTATION DRAFI* RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION ". to 3.3.3. A The radioactive gaseous effluent monitoring instrumentation channels
- shown in Table 3.3-13 shall be OPERA 8LE with their Alarm / Trip 5etpoints set to ,
ensure that the limits of Specifications 3.11.2.1 and 3.11.2.5 are not exceeded. The Alare/ Trip setpoints of these channels meeting Specification 3.11.2.1 shall be determined and adjusted in accordance with the methodology and
. parameters in the 00CM.
~
APPLICA8ILITY: As shown in Table 3.3-13 ACTION:
., s. With a radioactive gaseous effluent monitoring instrumentation channel Alare/ Trip setpoint less conservative than required by the above specification, immediately suspend the release of radioactive gaseous effluents monitored by the affected channel, or declare the channel inoperable.
- b. With the number of OPERA 8LE radioactive gaseous effluent monitoring
- instrumentation channels *1ess than the Minimum Channels OPERA 8LE, take the ACTION shown in Table 3.3-13. Restore the inoperable instrumentation to OPERA 8LE status within the time specified in the ACTION, or explain in the next Semiannual Radioactive Effluent Release Report pursuant to Specification 6.9.1.4 why this inoperability was not corrected within the time specified. s i
- c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS
. 10
. 4.3.3.$ Each radioactive gaseous affluent monitoring instrumentation channel shall be demonstrated OPERA 8LE by performance of the CHANNEL CHECX, SOURCE CHECK, CHANNEL CALIBRATION and ANACOGjCHANNEL OPERATIONAL TEST at the^ DIGITAL frequencies shown in Table 4.3-9.
O
- PSTS 3/4 3-77 NUREG 0452/STPEGS i COMPARISON l
- - . _ - - - - , _ _ _ . _ ~ , _ . , _ _ _ _ _ _ - _ _ _ - _ . - . . . _ _ _ _ _ . _ _ _ _ _ . _ - - - - . . . _ . _ . _ . _
Page No. 89 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE # NOTES co 3/4.3.3.11 PAGE: 3/4 3-78.0 . INSTRUMENTATION - RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTR.
- 1) Table 3.3-13 Table revised, those items not applicable to STPEGS have been deleted and the Action Statements renumbered as the unnecessary Action Statements have been deleted. There are no changes in the requirements.
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+ NUREG 0452/STPEGS PSTS 3/4 3-78 COMPARISON
Page No. 90 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE i NOTES co 3/4.3.3.11 PAGE: 3/4 3-79.0 . INSTRUMENTATION - RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTR.
- 1) Table 3.3-13 Table revised, those items not applicable to STPEGS have been deleted and the Action Statements renumbered as the unnecessary Action Statements have been deleted. There are no changes in the requirements.
t c O l 1 O
' /~
s i: 39 TABLE 3.3-13 (Continued)- , h RADI0 ACTIVE GASE005 EFFLUENT MONITORING INSTRIS H TATION P
~
MINIMUM CHAltlELS INSTRIMENT OPERABLE APPLICABILITY ACTION 'u o
;- . .11- . =. =
==e-
-.-..-e.. _ _ _ _ _ _ _ _
,- : -- f-= - r-- ---- - y,,,-,,
7-- ,
- - . - - - - - - - ;= = ;_ -- ;.;=.== =- -
= . . , - . _ . _ _ ,._ . _ ..
i
- r. $$, -. _ _ ."-.. $ $- . i (.".-" 1! __..".--I, 2 26, 32-
, M
- s N
Y f.l. Condenser Evacuation System p m l
- a. Noble Gas Activity Monitor 1 :*
- 47 31 -
- b. Iodine Sampler 1 * . 53 32.
- c. Particulate Sampler 1
- 31 32.
- d. Flow Rate Monitor
- 1 w 30
- e. Sampler Flow Rate Montter 1
- 44 30 ,
E nm .' eo eo -
>*~
- o s.n H eJ .
t/3 % O t/3
=
m .
- {
O * (J3 . .* * . - O 9
Page No. 91 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE i NOTES 00 3/4.3.3.11 PAGE: 3/4 3-80.0 . INSTRUMENTATION - RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTR.
- 1) Table 3.3-13 Table revised, those items not applicable to STPEGS have been deleted and the Action Statements renumbered as the unnecessary Action Statements have been deleted. There are no changes in the requirements.
. lll>-
. t l
l e j
C s O vt
.I I
. i TABLE 3.3-13 (Cantinued)
If i RADI0 ACTIVE GASEGUS EFFLUENT DENITORIllG IllSTRtSEIITAT1018
~. .
MlliittM CMANIGELS . , IItSTRUMENT OPERABLE APPLICABILITY- ACTICII link .
%,4- AVant h ullEEESPEE h , u
- a. Itable Gas Activity leoniter l'
- 31 #
i I b. Iodine SE4E9EPdoAliTOR 1, a 37gg l l -
- c. Particulate Sampler 1 *-
3281 l d. Flow Rate Stealter 1 .* 3o 46 .
- y e. Sampler Flow Rate Monitor 1
- 3o48 y _.
' ' _~_~___
O
- - ~ ~ ' ~ '
" " ~ ' * '
= -- w w wW Ww w mv .- W WwW y ww 6%EF IGNI$NEttugFWf:95L-- 1
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^
l == ,?: :?-- Ext: :^_ww!i- 1 = 4 C M ** .I ' oM
- aOO i
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4 1
Page No. 92 06/13/85 COMPARISON OF NUREG 0452, RE7.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE i NOTES oc 3/4.3.3.11 PAGE: 3/4 3-81.0 . INSTRUMENTATION - RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTR.
- 1) Table 3.3-13 Table revised, those items not applicable to STPEGS have been deleted and the Action Statements renumbered as the unnecessary Action Statements have been deleted. There are no changes in the requirements.
s e O
.t O
i
DRAFT s
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ii < ! .ii il i lIb PSTS 3/4 3-81 NUREG 0452/STPEGS COMPARISCN L
Page No. 93 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE f NOTES 00 3/4.3.3.11 PAGE: 3/4 3-82.0 INSTRUMENTATION - RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTR.
- 1) Table 3.3-13 Table revised, those items not applicable to STPEGS have been deleted and the Action Statements renumbered as the unnecessary l
Action Statements have been deleted. There are no changes in the requirements. i l i I 1 . O 9 l 1 i O \ l l
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< N <14 NUREG 0452/STPEGS PST5 3/4 3-82 COMPARISON P
Page No. 94 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE # NOTES 00 3/4.3.3.11 PAGE: 3/4 3-83.0 INSTRUMENTATION - RADIOACTIVE GASEOUS EFFLUENT MDNITORING INSTR.
- 1) Table 3.3-13 Note ** deleted because it is not a release point. STPEGS design.
- 2) Table 3.3-13 Action 45, 48, 49 and 50 are deleted.
Not applicable to STPEGS.
- 3) Table 3.3-13 Action 46 and 47 are renumbered as Action 30 and 31 respectively. Editorial.
s-O e 9 O
l
. . . .. .-. I O:
TA8LE 3.3-13 (Continued)
. TA8LE NOTATIONS
- At all times. ,
-sa-Sur4ay:W45"-*>l EL^^3^ Fi= -i m. _u-ACTION STATEMENTS
'assar=at Wi ferj6f c i d s 1 sf ify i
n s) b P$ 1be ev t Jm e on en s 1f ie s n 1 11 i le s . A le ee et e f nk s na b At a hi 1 q 1f rs of h ci it s f p n vr e se ra e 1 1 to s rwse;[u e o c) o t we efflu i is y y/
. ACTION M - With the number of channels OPERA 8LE less than required by the .
Minimum Channels CPERA8LE requirement, effluent releases via O this pathway may continue for up to 30 days provided the flow ! rata is estimated at least once per 4 hours. ;
. ACTIONh- With the number of channels OPERA 8LE less than required by the ;
Minimum Channels OPERA 8LE requirement, effluent releases via ! this pathway may continue for up to 30 days provided grab samples are taken at least once per 12 hours and these samples. ,, , are analyzed for radioactivity within 24 hours.
- N 'eAus6e o cilan 1 Pf (E e ttfanre ire b
- po h'n s OPJ di , ia el to e
, / G r di ac,tf n vi is at a.
*Cf3tlpt:46 - 1 en pf e n 1 OP E es th n equ o by th '
i C nneig 0 AASLE ' 1r , er i of th s
$i y/conine ov de gr s 1s 1 ed at JEM eas c her 4 ur a an ly ed it n e o1 ur/.
"YIETftNf:99:- W 1Ig of h n Is OP 8 on I ss an re irp b 3h :tyfousCh OP qui t, oper ti n o t is s ma id g ab es re ak aryd 1y ed
. st/ponpa cpn 4 ur . fth bot ch nne s i operab ,
6n pay in p ovi ed rab les re tak n a al zed c 4 usd in dog s g t j o s ri g o er per i s erjfo a a 1 st O W-STS _ 3/4 3-83 NUREG 0452/STPEGS COMPARISON s..... .. .. . . . .
Page No. 95 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE f. NOTES co 3/4.3.3.11 PAGE: 3/4 3-84.0 INSTRUMENTATION - RADIOACTIVE GASEOUS EFFLUENT MDNITORING INSTR. DESIGN- 1) Table 3.3-13 Action Statement 52 deleted. Not applicable to STPEGS design. ED 2) Table 3.3-13 Action Statement 51 renumbered 32.
,t Editorial.
s O O
e.,_.. .- .. .. . .... . . . . . . . . - - - - - - . . - - - -
- - X - 2~.: = ~
J -
, TA8tf 3.3-13 (Continued) .
TA8LE NOTATIONS (Continued) 3E . ACTION -SE- With the number of channels OPERA 8LE less than required by the Minimum Channels OPERA 8LE requirement, effluent releases via tha_affected pathway may continue for up to 30 days provided samples are continuously collected with auxiliary sampling equipment as required in Table 4.11-2.
= -:-f -d :n.wis ^"'"* ' : : ?::: 9:
- -i:2 5, the-Mfwieundhennele4PERA8LE 4 0. _ ..^., ;;;,. . -
a,.... - ,,1 7 h
- O e a e
9 e 8 e 9 0h PSTs 3/4 3-84 NUREc 0452/STPEGS COMPARISON
Page No. 96 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE i NOTES 00 3/4.3.3.11 PAGE: 3/4 3-85.0 , INSTRUMENTATION - RADI0 ACTIVE GASEQUS EFFLUENT MONITORING INSTR.
- 1) Table 4.3-9 Table revised and made consistent with Table 3.3-13. Those items not applicable to STPEGs have been deleted and the Action Statements renumbered as the unnecessary Action Statements have been deleted. No changes in requirements.
e O e t O
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Page No. 97 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE # NOTES co 3/4.3.3.11 PAGE: 3/4 3-86.0 . INSTRUMENTATION - RADI0 ACTIVE GASEOUS EFFLUENT HONITORING INSTR.
- 1) Table 4.3-9 Table revised and made consistent with Table 3.3-13. Those items not applicable to STPEGS have been deleted and the Action Statements renumbered as the unnecessary Action Statements have been deleted. No reduction in requirements s
O-O
*j TAeLE 4.3-9 (Continued)
'T ^
g BAtleACTIW GASE805 EFFLtlENT SONi1 BRING IIISTEIBENTATION $NRWEILLANCE K0tilAEleff5 . Anthl0G .
~
CNeteIEL IO SES Fa t 1stitst - CNAGEEL SOUSCE OteleIEL OPERATICIIAL Ill5TmeefT
- SURWEILLAIICE .
_ CHECK CNCE CAtlBSAT1018~, TEST 15 N0tflKO E_ ^^^ ,TE ME^~_^'_ E5ME- b !__: _ E- '= 3- e 27 - -- U-- , , , - - ' 7
- c. ::, _^.
.- flunit- .- ---- -- -E- isIA. ll(4) - --- ~ - vi
~
(Autametic Centrol, redundant) - -
^ : , ^. . 1 6 ters R
r- i w.-n.--iltgi-or- qtsi N ^^ - W r-dusti k f. Condenser Evacuatten System
- a. Noble Gas Activity Stonitor S N R([)
()(/) * -
- k. Iodine Sampler W N.A. N.A. * '
N.A.
- c. Particulate Sampler W N.A. M.A. *
. N. A. -
- d. Flow Rate Montter D N.A. R . q *
- e. Sampler Flow Rate Stentter 8 N.A. R
, q * -
^
t" ~, : ^ _ 52.f. . 2. nE oo a. Noble Gas Activity Dennitor 8 N R(/) * . hR b. Iedine-*P McNirog y D :It:2.M Jt::1.
== At:A.Q
- 8A C i zg .
m - b 1
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Page No. 98 06/13/85 COMPARISON OF NURIG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE f NOTES co 3/4.3.3.11 PAGE: 3/4 3-87.0 . INSTRUMENTATION - RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTR.
- 1) Table 4.3-9 Table revised and made consistent with Table 3.3-13. Those items not applicable to STPEGS have been deleted and the Action Statements renumbered as the unnecessary Action Statements have been deleted. No reduction in requirements e
O e t e
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Page No. 99 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE 9 NOTES co 3/4.3.3.11 PAGE: 3/4 3-88.0 . INSTRUMENTATION - RADI0 ACTIVE GASEQUS EFFLUENT MONITORING INSTR. DESIGN- 1) Table 4.3-9 Table revised and made consistent with Table 3.3-13. Those items not applicable to STPEGS have been deleted and the Action Statements renumbered as the unnecessary Action Statements have been deleted. No change in requirements. o O e I l 1 l i O l l l
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Page No. 100 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE i NOTES 00 3/4.3.3.11 PAGE: 3/4 3-89.0 , INSTRUMENTATION - RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTR.
- 1) Table 4.3-9 Table revised and made consistant with Table 3.3-13. Theae items not applicable to STPEGS have been deleted and the Action Statements renumbered as the unnecessary Action Statements have been deleted. No changes in requirements.
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Page No. 101 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE # NOTES 00 3/4.3.3.11 PAGE: 3/4 3-90.0 , INSTRUMENTATION - RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTR.
- 1) Table 4.3-9 Notations (**) delete. It is not a release point. STPEGS design.
- 2) Table 4.3-9 Notations 1, 4 and 5 Delete. STPEGS design does not include Analog Channels with automatic isolation or Hydrogen / Oxygen monitors.
ED 3) Table 4.3-9 Notations 2 and 3 - Change to 1 and 2. Editorial.
- 4) Table 4.3-9 Notation 2 - Change " Analog Channel i-Operational Test" to " Digital channel Operational Test". STPEGS provides digital radiation monitoring instrumentation.
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-- T~ $ l ..'.. . . -- X TA8LE 4.3-9 (Continued) .
DRAFT TA8LE NOTATIONS
- At all times. '
ser r.g W . sic O.5-mi^^J ^^i. . ; m - . . ^.h . . . (1) TP/ ANd E OPER4 ha)Ial ns rata tha autdnaic del to i pa y la ann cia ion cu i of oil co iti ns ,is .
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. ns f i s as evel abo e Al p tpo nt o
- b. C f u ,o
. In t1 icates seal fai ure or st n co 1s not et i ope ate de.
(/) The $NAtOG EL OPERATIONAL TEST shall atseLdemonstrate that control room alarm annunciation occurs if any of the following conditions exists:
- a. Instrument indicates measured levels above the Alaru Setpoint, or
- b. Circuit failure, or
,O .
- c. Instrueent indicates a downscale failure, or * '
- d. Instrument controls not set in operate mode.
(/) The initial CHANNEL CALIBRATION shall be performed using one or more of " the reference standards certified by the National Bureau of Standards (N85) or using standards that have been obtained from suppliers that participate in esasurement assurance activities with N35. Thesa-standards shall permit calibrating the system over its intended range of energy and seesurement range. For subsequent CHANNEL CALIBRATION,. sources that have been related to the initial calibration shall be used. M T)le RA 6N 6el in a se fs a ss le on in ina :
- a. ol rc tr .e- , la an to n,
/
. s F ur olum/e pereen gen anc ni ge NEL LI TI s 11 1 u of rd s sample
/ co ing inal
/ '
3
. One vol ent oxygen, la e ni ro ,
. F ur luey pery t o ger), bg ance/ nit gen.
iO W-STS 3/4 3-90 NUREG 0452/STPEGS C0JfPARISON , l
~
Page No. 102 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE i NOTES 00 3/4.3.4 PAGE: 3/4 3-91.0
- INSTRUMENTATION - TURBINE OVERSPEED PROTECTION DESIGN 1) 3/4.3.4 Delete Specification. The STPEGS design has the turbine oriented such that turbine missiles have very low probability of impacting Safety Related equipment or structures. This Tech Spec is not necessary to assure safe operation.
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. 3/4.3.4 TURBINE OVERSPEED PROTECTION i
- l LIMITING CONDITION FOR OPERATION 3.3.4 At least one Turbine Overspeed Protection Systas shall be OPERA 8LE.
,' APPICA8ILITY: 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 pra'ssure turbine steam ifne inoperable, restore the inoperable valve (s) to OPERA 8LE status within 72 hours, or close at least'one valve in the affected steam line(s) or isolata j the turbine from the steam supply within the next 6 hours.
I b. With the above required Turbine Overspeed Protection System otherwise .
. inoperable, within 5 honrs isolate the turbine from the staas 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 Systaa shall be demonstrated OPERA 8LE: i'
- a. At least once per [ days by cycling each of the following valves e througtr at least one completa cycle from the running position: g
.' 1) [Four] high pressure turbine stop valves, .
[Four] high pressure turbine governor valves, ! 2)
- 3) [Four] low pressure turbine reheat stop valves, and l
- 4) [Four] low pressure turbine reheat intercept valves. l
~
- b. At least once per 31 days by direct observation of the movement of each of the above valves through one completa cycle from the running positien, l
- c. At least once per 18 months by performance of a CHANNEL CALIBRATION on the Turbine Overspeed Protection Systems, and l
- 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 unacceptable flaws or excessive corrosion. If unacceptable flaws or excessive corrosion are found, all other valves of that type shall be inspected.
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,,-STS 3/4 3-91 NUREG 0452/STPEGS COMPARISON
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4 ? i .li j I e t t NUREG 0452/STFEGS COMPARISON
e Page No. 63 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE # NOTES
** 3/4.4.1.1 P?.GE: 3/4 4- 1.0 OPERATING MODE: 12----
e RCS - REACTOR COOLANT LOOPS & COOLANT CIRC. - STARTUP & POWER OP. NONE 1) No change. l l l 1 t I O
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3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION . STARTUP AND POWER OPERATION LIMITING CONDITION FOR dPERATION 3.4.1.1 All reactor coolant loops shall be, in operation. . APPLICABILITY: , MODES 1 and 2.* ACTION: With less than the above required reactor coolant loops in operation, be in at least HOT STAN08Y with'in 6 hours. t SURVEILLANCE REQUIREMENTS 4.4.1.1 The above required reactor coolant loops shall be verified in operation and circulating reactor coolant at least once per 12 hours. l
~
l "See Special Test Exceptions Specification 3.10.4. NUREG 0452/STPEGS l
)g-STS 3/4 4-1 COMPARISON
Page No. 64 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NO1E' # NOTES
** 3/4.4.1.2 PAGE: 3/4 4- 2.0 OPERATING MODE: --3---
RCS - REACTOR COOLANT LOOPS & COOLANT CIRC. - HOT STANDBY DATA 1) 3/4.4.1.2 STPEGS specific nomenclature provided. DATA 2) 4.4.1.2.2 Steam generator secondary side water level provided as 10% narrow range. This level ensures that the SG is available as a heat sink. O l l 1 I 1 O
I REACTOR COOLANT SYSTEM i r' HOT STANDBY l LIMITING CONDITION FOR OPERATION
- At*1 east two of the reactor coolant loops listed below shall be 3.4.1.2 OPERABLE with two reactor coolant loops in operation when the Reactor Trip System breakers are closed and one reactor coolant loop in operation c- when1 the Reactor Trip System breakers are open:"
- a. ReactorCoolantLoopkA nd its associated steam generator and l reactor coolant pump,
- b. ReactorCoolantLoophB[anditsassociatedsteamgeneratorand reactor coolant pump, -
l
- c. .R,eactor Coolant LoopMC[and its associated steam generator and.
reactor coolant pump, and
- d. ReactorCoolantLoophDdnditsassociatedsteamgeneratorand -
reactor coolant pump. APPLICABILITY: MODE 3. ACTION: .. With less than the above required reactor coolant loops OPERABLE, (- . a. yestore the required loops to OPERABLE status within 72 hours or be
, in HOT SHUTDOWN within the next 12 hours,
- b. With only one reactor coolant loop in operation and the Reactor Trip System breakers in the closed position, within I hour open the Reac
- Trip System breakers. , .s
- 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 REQUIREMENTS At least the above required reactor coolant pumps, if not in 4.4.1.2.1 operation, shall be determined OPERABLE once per 7 days by verifying correct breaker' alignments and indicated power availability. 4.4.1.2.2 The reipiire'd steam generators shall be determined OPERABL secondary side water level to be greater than or equal _ 10%toVEsag v4eAoW ranee at least onc per 12 hours. 4.4.1.2.3 The required reactor coolant loops shall be verified in' operation and circulating reactor coolant at least once per 12 hours. O 9 O *All reactor coolant pumps may be deenergized for up 'to I hour provided: (1) no operations are permitted that would cause dilution of the Reactor Coolant System boron concentration, and (2) core outlet temperature is maintained at least 10*F below saturation temperature. NUREG 0452/STPEGS
- COMPARISON 3/4 4-2 W-STS
---,-,m .,a- - -,.v.,-- . - - - - - , , , - . - - - , . - . , - . , - - - , , - -, , - , , - - - , , - - . , - - - , - - - , , , , . , - . - , .,-,,,--,------,-,n,- - , , -
Page No. 65 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE # NOTES
** 3/4.4.1.3 PAGE: 3/4 4- 3.0 OPERATING MODE: ---4--
RCS - REACTOR COOLANT LOOPS & COOLANT CIRC. - HOT SHUTDOWN DATA 1) 3/4.4.1.3 Maximum RCS cold leg temperature and secondary water temperature to be provided LATER (awaiting resluts of Cold Overpressure Mitigation Analysis). DESIGN 2) 3/4.4.1.3 Residual Heat Removal loops identified to reflect STPEGS design. s l l O
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i s REACTOR COOLANT SYSTEM HOT SHUTDOWN . LIMITING CONDITION FOR OPERATION - At.least two of the loops listed below shall be OPERA 8LE'an 3.4.1.3 one of these, loops shall be in opecetfom** a.. Reactor Coolant Loop A and its associated steam generator and
- reactor coolant pump,*
and its associated steam generator and
- b. Reactor Coolant Loop *8 . .
. reactor coolant pump, c.
Reactor Coolant Loop C and its associated steam generator and
. reactor coolant pump,*
and its associated steam generator and
- d. Reactor Cnolant Loop *D -
reactor coolant pump, 8. d.wolW Y remouaA WP L* v APPLICABILITY: MODE 4. r ACTION: ggeres coeuwt AND/e# RNN With less than the above requiredA l cops OPERABLE, israediate
- a. initiate corrective action to return the required loops to CPE status as soon as possible; if the remaining CPERABLE lo loop, be in CDLD SHUTDOWN within 24 hours. l' REAcTES With noA CMMT OR W oop in operation, suspend all operations involvin
- b. l tion in boron co0 centration of the Reactor Coolant System and
, immediately initiate corrective action to re to operation.
I hour l M All reactor coolant pumps anW"E Opumps"may be desnergized t re Reactor Coolant System boron concentrati provided: is maintained at least 10'F below saturation temperature. Reactor A reactor coolant pump shall not be started with'Fone unless Coolant System cold leg temperatures less than or, ;hanequal &*Fmore or to ( o the secondary water tamperature of each steam generator is less l above each of the Reactor Coolant ,5ystes cold leg temperatures. 3.- 3/4 4-3 NUREG 0452/STPEGS W-STS COMPARISON
o-Page No. 66 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE # NOTES
** 3/4.4.1.3 PAGE: 3/4 4- 4. 0 OPERATING MODE: ---4--
RCS - REACTOR COOLANT LOOPS & COOLANT CIRC. - HOT SHUTDOWN DATA 1) 4.4.1.3.2 Steam generator secondary side water level provided as 10% narrow range. This level ensures that the SG is available as a heat sink. DESIGN 2) 4.4.1.3.4 Moved from STSEGS Section 4.5.3.d with no changes in requirements since RHR pumps are not part of the ECCS. . DATA 3) 4.4.1.3.4 RHR interlock setpoints provided. b.) is set so that spurious pressure spikes will not close the valves & damage the RHR pumps. O
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_._ _ _ _ m -__ . m.. . . _ , . _ g.g J s .4 M REACTOR COOLANT SYSTEM
' SURVEILLANCE REQUIREMENTS .
4.4.1.3.1 The required reactor coolant pump (s) indicated power availability. V l 4.4.1.3.2 The required steam generator (s)mdf 10% Nggew sha J 1 east once per 12 hours.
. At least one reactor coolant or RHR loop shall be verified in 4.4.1.3.3 operation and circulating reactor coolant at least once per 12 hours.
4.4.1.3.4 At least once per 18 months by verifying automatic isolation a interlock action of the RHR system fr that: I a. With simulated or actual Rese, tor Coolant System pressure ERsignal greater than or equal to .LHr/psis the, interlocks will cause the R SuctionIsolationvalvesfromb5eopened, and V AzoS) actor Coolant System pressure signal 'O b. With a simulated or actual psig the interlocks will cause the RHR greater than or equal toSuction Isolation valves to automatically f c 1 1
~
' N0ulfrom 4.5.3.d.
y 4 [ I i l 1 . 6t O <- NUREG 0452/STPEGS COMPARISON 3/4 4-4
., W-STS .
Page No. 67 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE # NOTES
** 3/4.4.1.4.1 PAGE: 3/4 4- 5.0 OPERATING MODE: --- RCS - RC LOOPS & COOLANT CIRC.-COLD SHUTDOWN, LOOPS FILLED DATA 1) 3.4.1.4.1 Minimum RCS cold leg temperature and secondary side water temperature to be provided LATER (awaiting results of Cold Overpressure Mitigation Analysis).
DESIGN 2) 3.4.1.4.1 Action a. and Note (**) changed to reflect STPEGS design of 3 Puut trains. DATA 3) 3.4.1.4.1 Steam generator secondary side water level provided as 10% narrow range. This level ensures that the SG is available as a heat sink. O.
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REACTOR COOLANT SYSTEM COLD $HUTDOWN - LOOPS FILLED . _ LIMITING CONDITION FOR OPERATION ' 3.4.1.4.1 At least one residual heat removal (RHR) loop shall be OPERABLE and in operation *, and either: l
- a. One additional RHR loop shall be OPERA 8LE**, or
- b. The secondary side water level of at. least two steam generators g shall -be greater than [{3% oF the no.ecow ru.nge..
MODE 5 with reactor coolant loops filled ***. l-APPLICABILITY: ACTION:
- a. With*bLb sne of the RHR loops inoperable and with less than the required steam aenerator water level, immediately initiate corrective action to returnAthe inoperable RHR loop to OPERABLE status or restore the oni required steam generator water level as soon as possible.
- b. With no RHR loop in operation, suspend all operations involving a O s reduction in boron concentration of the Reactor Coolant System and isusediately initiate corrective action to return the requfred RHR loop to operation.
SURVEILLANCE REQUIREMENTS t 4.4.1.4.1.1 The secondary side water level of at least two steam generators when required shall be determined to be within limits at least once per 12' hours. 4.4.1.4.1.2 At least one RHR loop shall be determined to ta in operation and circulating reactor coolant at least once per 12 hours. 9 (1) no operations
*The RHR pump may be deenergized for up to I hour provicac:
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. I l
'Two RHR loopsmay be inoperable for up to 2 hours for surveillance testing
**0ne provided the other RHR loop is OPERA 8LE and in operation.
f the Reactor
***A reactor coolant pump shall not be started with one or mor F unless Coolant System cold leg tertperatures less than or equal tothe above each of the Reactor Coolant System cold leg temperatures.
O NUREG 0452/STPEGS 3/4 4-5 COMPARISON l W-STS
~
Page No. 68 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE # NOTES
-~~~-~~~---- ------ _________________________________ ** 3/4.4.1.4.2 PAGE: 3/4 4- 6.0 OPERATING MODE: --- RCS - RC LOOPS & COOLANT CIRC.-COLD SHUTDOWN, I4f0PS NOT FII1ED DESIGN 1) 3.4.1.4.2 Note (*) changed to reflect STPEGS design of 3 RHR trains.
9:
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O ' DRAR l I REACTOR COOLANT SYSTEM COLD SHtfTDOWN - LOOPS NOT FILLED ~ I LIMITING CONDITION FOR OPERATION Two residual heat removal (RNR) loops shall be OPERABLE" and at
.l: i 3.4.1.4.2 least one RHR loop shall be in operation.**
M00E 5 with reactor coolant loops not filled. APPLICABILITY: ACTION: a. With less than the above required RHR loops OPERABLE, 'insediately 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. , O~ SURVEILLANCE REQUIREMENTS 4.4.1.4.2 At least one RHR loop shall be determined to be in operation and circulating reactor coolant at least once per 12 hours. '
* *d$fRHRloopsaybeinoperableforupto2hoursforsurveillancetesting provided the other RHR loop is OPERA 8LE and in operation.
*?. (1) no opera-
**The RHR pump may be deenergized for up to I hour provided:
- 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.
[ O. NUREG 0452/STPEGS 3/4 4-6 COMPARISON PSTS
Page No. 69 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE # NOTES
** 3/4.4.1.5 PAGE: 3/4 4- 7.0 OPERATING MODE: 12345-RCS - RC LOOPS & COOLANT CIRC.- ISOLATED LOOP (OPTIONAL)
DESIGN- 1) 3/4.4.1.5 Section deleted. STPEGS does not plan to operate n-1 loops. O t I l ? O
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~
DMR REACTOR COOLANT SYSTEM _.E ISOLATED LOOP (OPT'IONAL).* _6 J ._ LINITING CONDITION FOR OPERATION ; i 3.4.1.5 The boron concentration of an isolated loop shall be maintained l greater than or equal to the boron concentration of the operating loops. 1 APPLICA81LITY: MODES 1, 2, 3, 4, and 5. . A,qlgN: ~' With the requirements of the above specification not satisfied, do not open the isolated loop's stop valves; either increase the b'oron concentration of the isolated loop,to within the limits within 4 hours or be to a SHUTDOWN MARGIN equivalent to at le,ast 35 Ak/k at 200*F. s
$URVEILLANCE REQUIREMENTS .,i i
4.4.1.5 The boron concentration of an isolated loop shall be determined to b'e greater than or equal to the boron concentration of the ope ' leg or cold leg stop valves of an'isolatad loop. O 3/4 4-7 NUREG 0452/STPEGS PSTS COMPARISON ( --- - _ - _ _ _ - _ _ _ - - - - - - ~ . . - . . _ _ _ , , _ _ _ _ _ _ _ _ _
Page No. 70 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE # NOTES
** 3/4.4.1.6 PAGE: 3/4 4- 8.0 OPERATING MODE: 123456 RCS - RC LOOPS & COOLANT CIRC.- ISOLATED LOOP STARTUP (OPTIONAL)
DESIGN- 1) 3/4.4.1.6 Section deleted. STPEGS does not plan to operate n-1 loops. b
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ISOLATED. LOOP STARTUP [0PTIONAL] I LIMITING CONDITION FOR OPERATION
. i 3.4.1.5 A reactor coolant loop shall remain isolated until:
- a. The isolated loop has been operating on a recirculation flow of j gpa for at least 90 minutes and the greater than or a:[ual totemperatureatthecold1Toftheisolatedloopisw the highest cold leg temperature of the operating loops, and l.
The reactor is subcritical by at least 2% Ak/k. l I. b. APPLICA81LITY: All MODES. g:
- With the requirements of the above specification not satisfied, suspend startup of the isolated loop. -
O, SURVEILLANCE REQUIREMENTS g 4.4.1.E.1 The isolated loop cold leg temperature shall be determined to be . within 20*F of the highest cold leg temperature of the operating loops within
- 30 minutes prior to opening the cold leg stop valve.
4.4.1.6.2 The reactor shall be detaruined to be suberitical by at least l 1% Ak/k within 30 minutes prior to opening the cold leg stop valve. 4 O 3/4 4-8 NUREG 0452/STPEGS PSTS COMPARIS'ON I
Page No. 71 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. . NOTE TYPE NOTE # NOTES
** 3/4.4.2.1 PAGE: 3/4 4- 9.0 OPERATING MODE: -- RCS - RC LOOPS & COOLANT CIRC. - SAFETY VALVES, BHUTDOWN NONE 1) 3/4.4.2.1 No Changes.
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O - gyg REACTOR' CO0LANT SYSTEM
. 3/4.4.2 SAFETY VALVES .
SHUTOOWN LINITING CONDITION FOR OPERATION 3.4.2.1 A minimum of one pressurizer Code safety valve shall be OPERA 8LE with a lift setting of 1485 psig
- IL*
APPLICABILITY: MODES 4 and 5. ,. ACTION: , , With no pressurizer Code safety valve OPERABLE, imediately suspend all operations. involving positive reactivity ch'anges and place an CPERA8LE RHR loop into operation in the shutdown cooling mode.
~
O SURVEILLANCE REQUIREMENTS
.I i
4.4.2.1 No additional requirements other than those required by Specification 4.0.5. .
*The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.
O W-STS 3/4 4-9 NUREG 0452/STPEGS COMPARISON
Page No. 72 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE i NOTES
** 3/4.4.2.2 PAGE: 3/4 4-10.0 OPERATING MODE: 123---
RCS - RC LOOPS & COOLANT CIRC. - SAFETY VALVES, OPERATING NONE 1) 3/4.4.2.2 No Changes. O l I .i l O
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REACTOR COOLANT SYSTEM -
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OPERATING - LIMITING CONDITION FOR OPERATION 3,4.2.2 All pressurizer Code safety valves shall be OPERA 8LE with a lift setting of 2485 psig 2 25.* AreLICA81LITY: MODES 1, 2, and 3. EIIB With one pressurizer Code safety valve inoperable, either restore the inoperable valve to CPERA8LE status within 15 minutes or be in at least NOT STAN08Y within 5 hours and in at least NOT SHUTDOWN within the following 6 hours. . O s.
$URVEILLANCE REQUIREMENTS 4.4.2.2 No additional requirements other than those required by Specification 4.0.5. ...
"The lift setting pressure shall correspond to ambient conditions of the vitive at nominal operating temperature and pressure.
O MS 3/4 4-10 NUREG 0452/STPEGS COMPARISON
Page No. 73 06/13/85 l COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. 1 NOTE TYPE NOTE i NOTES l c* 3/4.4.3 PAGE: 3/4 4-11.0 OPERATING MODE: 123--- RCS - PRESSURIZER. DATA 1) 3.4.3 STPEGS specific values provided. DESIGN 2) Bases 3.4.3 Clarify minimum water volume as 'or 92% of span' since control panel reads % of span not cubic feet. DESIGN- 3) 4.4.3.3 Deleted. This surveillance requirement was to demonstrate capability to power pressurizer heaters from lE buses. STPEGS 2 groups of backup pressurizer heaters are normally powered from lE buses. O
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s z . REACTOR COOLANT $YSTEM - 3/4k.3 PRESSURIZER - LIMITING CONDITION FOR OPERATION , 18t t eq.4.3 The pressurizer shall_ be OPERABLE with a water volume of less than orf 3 7 each uai having tolgtest4. a capacitycubic feetI of at and leastytt563 at ieasr.kW. swo groups orpressurizer neaters podQ* BACKUP \0E J l75 7 '.- e**
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APPLICABILITY: MODES 1, 2, and 3. - ACTION: - - BACgfp INOPERABLE - a With only one group of A pressurizer heatersA02f2ASti, restore at least l Or BACKUP two groupshto OPERABLE status within 72 hours or be in at least HOT HEATERS STAN08Y within the next 6 hours gnd in HOT SHUTDOWN within the
-' following 6 hours,
- b. With the pressurizer othemise inoperable, be in at least HOT STANOBY with the Reactor Jr.ip System breakers open within 6 hours and in HOT l , ,,
, SHUTDOWN within the following 6 hours, l' t r j .
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$URVEILLANCE REQUIREMENTS 4.4.3.1 The pressurizer water volume shall be detemined to be within its Itait at least once per 12 hours.
4.4.3.2 The capacity of each of the above required groups o s urizer heaters shall be ,beriffed by energizing the heaters and measuring circuit current at least once per 92 days. '
. .=.; = -.h,;-- _ ;
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hed=CPERA8L&- x .x. ,..c -18 months i, w- ? i, ^.. .... .'.. . :., i i _. ' --- ti;+ -essi te t'- _ .crge. _,- pe-- .- -
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3/4 4-11' NUREG 0452/STPEGS l - COMPARISON
Page No. 74 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. - NOTE TYPE NOTE # NOTES
** 3/4.4.4 PAGE: 3/4 4-12.0 OPERATING MODE:* 123---
RCS - RELIEF VALVEF DESIGN 1) 3.4.4 Action d. 'its associated solenoid valve' changed to 'the PORV' to reflect STPEGS design which has PORV's without solenoids. 9
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drift O . 3/4.4.4~ RELIEF VALVES LINITING' CONDITION FOR OPERATION . 3.4.4 All power-c stated relief valves (PORVs) and their associated
- valves shall be OPERA 8LE.
APPLICA8ILITY: ICDES 1, 2, and 3. ACTION: inoperable, because of excessive seat .
- a. With one or more PORV(s)ither restore the P0RV(s) to CPERABLE status,
. leakage, within 1 hour e
'er close the associated block valve (s); oth ,
following 30 hours. b. With one PORY inoperable due to causes other than excessivej , leakage, within I hour either restore the PORV to CPERA8LE close the associated block valve and remove power from th valve; restore the PORV to OPERABLE status within.the foll 72 hours or be in NOT STAND 8Y wtihin the n SHUTDOWN within the following 30 hours. O-
- c. With both.P0RV(s) inoperable due to cau status or close their associated block valve and COLD SHUTDOWN within the following 30 hours.
d. With one or more block valve (s) inoperable, within I hour: (1) restore the block valve (s) to OPERABLE^status, or c block valve (s) and remove power ' ' ' ' from the blo"ck valve (
--- and f'
thePORVandremovepowerfromdove,asappropriate,fortheisolated (2) apply the ACTION b. or c. PORV(s). 3.0.4 are not applicable.
- e. The provisions of Specificattor j
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t i i O . 3/4 4-12 NUREG 0452/STPEC5 W.5TS
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COMPARISON
l l l Page No. 75 1 06/13/85 ' COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE # NOTE'S
** 3/4.4.4 PAGE: 3/4 4-13.0 OPERATING MODE: 123---
RCS - RELIEF VALVES DESIGN- 1) 4.4.4.3 Deleted. This requirement was to demonstrate the capability to power the PORVs from lE buses. STP PORVs are normally powered from lE buses. l l
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l. l 0 . 3/4.4.4 RELIEF VALVES . SURVEILLANCE REQUIREMENTS 4.4.4.1 In addition to the requirements of Specification 4.0.5, each PORY shall be demonstrated CPERABLE at least once ,per 18 months by:
- a. Performance of a CHANNEL CALIBRATION, and
- b. Operating the valve through one complete cycle of full travel.
4.4.4.2 Each block valve shall be demonstrated OPEAABLE at least once per 92 days by operating the valve through one complete cycle of full travel unless the block valve is closed with power removed in order to meet the requirements of ACTION b. or c. in Specification 3.4.4. . wm _ . . m . __. . . . __ e W O l O . NUREG 0452/STPEGS 3/4 4-13 COMPARISON PSTS
l Page No. 76 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE # NOTES
** 3/4.4.5 PAGE: 3/4 4-14.0 OPERATING MODE: 1234--
RCS - STEAM GENERATORS NONE 1) 3/4.4.5 No Changes. 6 0
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O . DRTT REACTOR COOLANT SYSTEM 3/4.4.5 STEAM GENERATORS - LIMITING CONDITION FOR OPERATION 3.4.5 Each steam generator shall be OPEAABLE. APPLICABILITY: MODES 1, 2. 3, and 4. 3: . With one or more steam generatorsabove inoperable, 200*F. restore the inoperable to CPERABLE status prior to increasing T ,g SURVEILLANCE REQUIREMENTS 4.4.5.0 Each steam generator shall be demonstrated OPERA 8LE by perfor the following augmented inservice inspection program and the requirem Specification 4.0.5. O 4.4.5.1 Staan Generator Sample SeTection and Inspection - Each steam genera shall be detenmined OPERABLE during shutdown by selecting and inspe least the minimum number of steam generators specified in Table 4.4-1. 4.4.5.2 Steam Generator Tube Samole Selection and Inspection - The generator tune minimum sample size, inspection result The classificat corresponding action required shall be as specified in Table 4.4-2. inservice inspection of steam generator tubes shall be performed at the quencies specified in Specification 4.4.5.3 and the inspected The tube
. verified acceptable per the acceptance criteria of Specification 4.4.5.4.
i tubes selected for each inservice inspection shall include at least 3% o
. total number of tubes in all steam generators; the tubes selected for these inspections shall be selected on a random basis except:
- a. Where experience in similar plants with similar tubes inspected shall be from these critical areas; b.
The first sample of tubes selected for each inservice inspection (subsequent to the preservice inspection) of each steam genera shall include: j 3/4 4-14 W-STS
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NUREG 0452/STFEGS COMPARISON
.,-.._..-.r-,-_ ._y
Page No. 77 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE # NOTES
** 3/4.4.5 PAGE: 3/4 4-15.0 OPERATING MODE: 1234--
RCS - STEAM GENERATORS NONE 1) 3/4.4.5 No Changes, O
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l REACTOR COOLANT SYSTEM - 1 STEAM GENERATORS SURVEILLANCE REQUIREMENTS (Continued)
- 1) ..All nonplugged tubes that previously had detectable wall penetrations (greater than 205),
- 2) Tubes in those areas where experience has indicated potential problems, and
- 3) A tube inspection (pursuant to Specification 4.4.5 4 8) shall -
be performed on each selected tube. If any selected tube does not permit the passage of the eddy current probe for a tube inspection, this shall be recorded and an adjacent tube shall be selected and subjected to a tube inspection.
- c. The tubes selected as the second and third samples (if required by Table 4.4-2) during each inservice inspection say be subjected to a partial tube inspection provided:
- 1) The tubes selected for these samples include the tubes from l'
those areas of the tube sheet array where tubes with imperfections were previously found, and _. (
- 2) The inspections include those portions of the tubes where imperfections were previously found.
The results of each sample inspection shall be classified into one of the ., following three categories: Catecory
' Insoection Results C-1 Less than 5% of the total tubes inspected are degraded tubes and none of the inspected tubes are defective.
f C-2 One or more tubes, but not more than 1% of the total tubes inspected are defective, or between 55 and 105 of the total tubes inspected are degraded tubes. C-3 More than 10% of the total tubes inspected are degraded tubes or more than 1% of the inspected tubes are defective. I Note: In all inspections, previously degraded tubes must exhibit i significant (greater than 101) further wall penetrations to be included in the above percentage calculations. O W-STS 3/4 4-15 NUREG 0451/STPEGS COMPARISON
Page No. 78 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE # NOTES
** 3/4.4.5 PAGE: 3/4 4-16.0 OPERATING MODE: 1234--
RCS - STEAM GENERATORS NONE 1) 3/4.4.5 No Changes. O l O I
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7e . . . . .. ...l REACTOR COOLANT SYSTEM STEASGENERATORS SURVEILLANCE REQUIREMENTS (Continued) 4.4.5.3 Inspection Frecuencies - The above required inservice inspections of . steam generator tubes shall be performed at the following frequencies:
- a. The first inservice inspection shall be performed after 6 Effective f
- Full Power Months but within 24 calendar months of initial criticality.
Subsequent inservice inspections shall be performed at intervals of not less than 12 nor more than 24 calendar months after the previous
,- inspection. If two consecutive inspections, not including the preser-vice inspection, result in all inspection results falling into the C-1 4
- category or if two' consecutive inspections demonstrate that previously
' observed degradation has not continued and no additional degradation has occurred, the inspection interval may be extended to a maximum of l
once per 40 months; ! b. If the results of the inservice inspection of a s' team generator conducted in accordance with Table 4.4-2 at 40-month intervals fall in Category C-3, the inspection frequency shall be increased to at i least once per 20 months. The increase in inspection frequency shall apply until the subsequent inspections satisfy the criteria of Specification 4.4.5.3a.; the interval may then be extended to a i 4 I maximum of once per 40 months; and 4
- c. Additional, unscheduled inservice inspections shall be performed on each steam generator in accordance with the first sample inspection specified in Table 4.4-2 during the shutdown subsequent to any of *a*
the following conditions:
' Reactor-to-secondary tubes leak (not including leaks originating l 1) from tube-to-tube sheet welds) in excess of the limits of l Specification 3.4.6.29,or*'
2) A seismic occurrence greater than the Operating Basis Earthquake? # A loss-of-coolant accident requiring actuation of the Engineered l 3) Safety Features @or ! 4) A main steam line or feedwater line break. i l O- . 3/4 4-16 NUREG 0452/STPEGS W-STS
- COMPARISON
Page No. 79 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE i NOTES
** 3/4.4.5 PAGE: 3/4 4-17.0 OPERATING MODE: ,1234--
RCS - STEAM GENERATORS DESIGN 1) 4.4.5.4.a.6)and Note (*) Nominal tube wall thickness plugging under review. Limit to be provided LATER. O-
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STEAM CENERATOR. '
$URVEILLANCE REQUIREMENTS (Continued) .
4.4.5.4 Acceptance Criteria .
- a. Asusedinthis/pecification:
- 1) Imperfection means an exception to the dimensions, finish, or contour of a tube from that required by fabrication drawings or -
specifications. Eddy-current testing indications below 20% of the nominal tube wall thickness, if detectable, may. be considered as imperfections."
- 2) Oeoradation means a service-induced cracking, wastage, wear, or general corrosion occurring on either inside or outside of a tube."
- 3) Decraded Tube means a tube containing imperfections greater than or equal to 205 of the nominal well thickness caused by degradation. "
- 4) 5 Decradation means the percentage of the tube wall thickness
- s affected or removed by degradation "
i ! 5) Defect means an imperfection of such severity that it exceeds # the plugging limit. A tube containing a defect is defective. '
- 6) pluccina Limit means the imperfection depth at or beyond which the tube shall be removed from service and is equal to (Larra) g of the nominal tube wall thickness.'r-
- 7) Unserviceable describes the condition of a tube if it leaks or contains a defect large enough to affect its structural integ-y rity in the event of an Operating Basis Earthquake, a loss-of-coolant accident, or a staan line or feedwater ifne break as l
t specified in Specification 4.4.5.3c. , above." r
- 8) Tube Inspection means an inspection of the steam generator tube from the point of entry (hot leg side) c etely around the U-bend to the top support of the cold le l
l l m.....
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i 3/4 4-17 NUREG 0452/STPEGS W-STS COMPARISON l ~ i
Page No. 80 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. . NOTE TYPE NOTE # NOTES
** 3/4.4.5 PAGE: 3/4 4-18.0 OPERATING HODE: 1234--
RCS - STEAM GENERATORS NONE 1) 3/4.4.5 No Changes. O
.t 4* ; O
l l O REACTOR COOLANT SYSTEM , f STEAM GENERATOR SURVEILLANCE REQUIREMENTS (Continued) Preservice Inspection means an inspection of the
- full length of l
- 9) each tube in each steam generator performed by eddy current techniques' prior to service to establish a baseline condition .
of the tubing. This inspection shall be performed prior to initial POWER OPERATION using*the equipment and techniques expected to be used during subsequent inservice inspections.
- b. The steam generator shall be determined OPERA 8LE after completing .
the corresponding actions (plug all tubes exceeding the plugging limit and all tubes containing through-wall cracks) required by l Table 4.4-2. 1 4.4.5.5 Reports - .
! a. Within 15 days following the completion of each inservice inspectfun .
of steam generator tubes, the number of tubes plugged in each stone generator shall be reported to the Commission in a Special Report 4 pursuant to Specification 6.9.2;
- b. The complete results of the steam generator tube inservice inspectfon j shall be submitted to the Commission in a Special Report pursuant to Specification 6.9.2 within 12 months following the completion of the inspection. This Special Report shall include:
l
- 1) Number and extent of tubes inspected, ,,
- 2) Location and percent of wall-thickness penetration for each indication of an imperfection, and .
l
- 3) Identification of tubes plugged.
- c. Results of steam generator tube inspections which fall into Category
, C-3 shall be reported in a Special Report to the Commission pursuant l ' to Specification 6.9.2 within 30 days and prior to resump plant operation. l gations conducted to determine cause of the tube degradation and corrective measures taken to prevent recurrence. 4 WREG 0452/SIPEGS W-STS 3/4 4-18 COMPARISON
\
Page No. 81 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE # NOTES
- 3/4.4.5 PAGE: 3/4 4-19.0 OPERATING MODE: 3.234--
RCS - STEAM GENERATORS DESIGN 1) Table 4.4-1 Table revised to reflect STPEGS design of 4 loop plant. 6 O O
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I TABLE 4.4-1 f IC * . MINIMUM NUMBER OF STEAM GENERATORS TO BE . 1 INSPECTED DURING INSERVICE INSPECTION-i : 1 4 t 1 l ! 7 > t i Yes j, No Preservice inspection T 1, 5 _NFour T- I. r Four i No. of Steam Generators per Unit All -0. T2 Two First inservice 14Mian a' a-%S f onei , f-mm %-,a inser. ice is { l TABLE NOTATIONS h d l encompassing 3 N % of the tubes - i i dicate that i' j
- 1. The inservice inspection eney be limised i to one steem generator l o y all steem generators are performing in a like menner. i Note that under some c rcumther steem generato ,
g di i s <
. one or more steem generators may be found so be more severe ; ted. The third and e then i' tho
, g2. The other steam generator not inspected during the first insonnce inspection shall be w = . l ( ; =g+-tions should follow the instructions described in 1 abogve i h l i 2,gsecond Eachand ofthird theinspections. other two steam generators not inspected hi tructions described in 1 above. The lourth and subsequent inspections shall follow t e ns during the' fi l 'l 8
= ..
bM
- 00
= ::2:3 8% - 3::= '
Q k 1 G e -- --
Page No. 82 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE # NOTES
** 3/4.4.5 PAGE: 3/4 4-20.0 OPERATING MODE: 1234--
RCS - STEAM GENERATORS NONE 1) Table 4.4-2. No Changes. 0
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Page No. 83 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE i NOTES c* 3/4.4.6.1 PAGE: 3,'4 4-21.0 OPERATING MODE: 1234-- RCS - REACTOR COOLANT SYSTEM LEAKAGE, LEAKAGE DETECTION SYSTEMS DESIGN OTHER 1) 3/4.4.6.1 Revised to reflect STPEGS design and is consistent with Callaway Tech Spec and RG 1.45 (Reactor Coolant Pressure Boundary Leakage Detection System, May 1973). O
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DRTT REACTOR COOLANT $YSTEM . 3/4.4.6 REACTOR C0OLANT SYSTEM LEAKAGE LEAKAGE DETECTION $YSTEMS LIMITING CONDITION FOR OPERATION
- - l
! f 3.4.'6.1 The following Reactor Coolant SystesJoakage fetectionfystaas shall l 1 be OPERA 8LE: .
.1...-_. z ,..__.___ - y
- s. Thefontainment/taosphereI"-----damal* cWwes, _
fonitoring fystandraerscw47* **
- b. The [ontainmenY h fusp veland/ low)fonitoringgystem,and j
& ron ontaineen condensate flow rateJ.asius.
- c.
- fonitoringfystem. #" '# #"
l . APPLICA81LITY: M00E5 1* 2* 3* and 4*
"en"a**dams "n"#Auses esonric dwe comomWT ! ATenefresRS t$ PERWD UhlN6 TWf S T bCCIDdW ,
wt,g 4,,,, ACTION:
' Detection Systems OPERA 8LE, l With only two of the above required Leaka rovided grab samples of the contain-operation say continue for up to 30 days g ment atmosphere are obtained and analyzed,at least once per 24 hours when the required Gaseous or Particulate Radioactive Monitoring System is inoperable; l
.g otherwise, be in at least NOT STAN08Y within the next 6 hours and in COLD 5HUTDOWN within the following 30 hours. .-
l SURVEILLANCE REQUIREMENTS 1 4.4.6.1 The /eakage [etection posingM.smarsfystems(pnnrot* shall be wodemonstrated Gmews emwas- CPERA8LE by:
- a. taospherea Containment /ofCHANNELCHECK,CHANNELCALIBRATION,andANALOGCH performanceOPERATIONAL TEST at the frequencies specified in Table 4.3-3, netMAL Containment4 4mahat.jspgevelandflow nitoring[ystesperformance (
- b. 18 months, and of CHANNEL CALIBRATION at least once p
- r-- -.c-
_ _~= 'T - E f~_- ___ w m owneton M M%~ C. cmuarr Reu.ru PERCOMnMer Cmauxmr or Vw CmetCllbNNEL CavaRAreN l i NUREG 0452/STPEGS 3/4 4-21 COMPARISON W-STS
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S Page No. 84 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE # NOTES c* 3/4.4.6.2 PAGE: 3/4 4-22.0 OPERATING MODE: 1234-- RCS - OPERATIONAL LEAKAGE DESIGN 1) 3.4.6.2.c Item modified to reflect STPEGS design which has no loop stop valves. DESIGN- 2) 3.4.6.2.e Deleted. Not applicable, CVCS is not associated with SI on STPEGS design. DESIGN 3) 3.4.6.2.f and Note (*) It is not possible to test valves at full system pressure prior to entry into Mode 4 since full system pressure cannot be attained until Mode 4. O
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. REACTOR COOLANT SYSTEM
~
OPERATIONAL LEAKAGE -
- LINITING CONDITION FOR OPERATION
~
3.4.5.2 Reactor Coolant System leakage shall be Ifeited to:
- s. No PRES $URE B0UNDARY LEAKAGE, l
' b. 1 gpa UNIDENTIFIED LEAKAGE, 1 gpa total reactor-to-secondary leakage through all staas
[-
- c. ^ h and ner_sJors
~
500T ga11ons per day through any one steam generator, m
], .
l
- d. 10 gpa IDENTIFIED LEAKAGE from the Reactor Coolant Systas, ._-
ms==mpan==c 2235 2 20 psig l
#% 1 pa leakage at a Reactor Coolant Systaa pressure of fromanyReactogCoolantSystas/ressure)'solationfalvespecified in Table 3.4-1.
j ' APPLICA8ILITY: MODES 1, 2, 3, and 4. E
- a. With any PRESSURE SOUNDARY LEAKAGE, be in at least NOT STAN
. within 5 hours and in COLD SHUTDOWN within the following 30 hours, i
- b. With any Reactor Coolant Systes leakage greater than any one of the PRES 5URE ICUNDARY LEAKAGE and leakage from reduce the leakage abovelimits, Reactor Coolantexcluding System /ressure/ solation /alves,leastNOTSTAND
- rate tothe within within nextIfeits within 6 hours and4inhours COLD orSHUTDOWN be in at within the following 30 hours.
- c. With any Reactor Coolant System /ressure / solation /alve leakage greater than the above limit, isolate the high p use of at least two closed manual or deactivated automatic valves, or be in at least NOT STAN0BY within the next 4 ho SHUTDOWN within the following 30 hours.
- j k T 4 MEM TE M f> W W
^<m rar =*"' * "
J25$
# Tssi entssuans LIS$ DAW msaa na m os,,,co wm, s , .r z2ss nwam sur awar ro ar wemy m Marmat To PM W !
OsFff fiat ro THE ONf*WAlf fbwfR, 3/4 4-22 NUREG 0452/STfECS y5TS COMPARISON t
Page No. 85 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE # NOTES
** 3/4.4.6.2 PAGE: 3/4 4-23.0 OPERATING MODE: 1234--
RCS - OPERATIONAL LEAKAGE
- ED 1) 4.4.6.2.1.b ' pocket' sump changed to ' normal and secondary' sump, STPEGS nomenclature.
DESIGN 2) 4.4.6.2.1.c Deleted. CVCS is not assolcated with SI for STPEGS design. Consistent with 3.4.6.2.e. G.
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.. . .. - .~ 7.
~
( DRH REACTOR COOLANT $YSTEM OPERATIONAL LEAKAGE SURVEILLANCE REQUIREMENTS 4.4.6.2.1 Reactor Coolant System leakages shall be demonstrated to be within each of the above Itatts by: .
- a. Monitoring the containment atmosphere N particulate ( l radioactivity monitor at least once per 12 hours.~
* - vo euk w o secovoser
- b. Monitoring the containment ssemur sump inventory and discharge at .
feast once per 12 hours." 3 l
= . . . . . . . . . .
8W!8muuuk oolant System water inventory balance at C.% performanceofaReactor)e least once per 72 hours, l Monitoring the feactor /ead /lange feakoff [ystem at least once per l dw 24 hours.
\
4.4.6.2.2 Each Reactor Coolant Systes fressure Taale3.4-1shallbedemonstrated0PERA8LEbyve/ rifying leakage to be solation within [alvesp its Itatt: ..
- a. At least once per la months,
- b. Prior to entering MODE 2 whenever the plant has been in COLD SHUTDOWN for 72 hours er more and if leakage testing has not been performed in the previous 9 months, 9 Prior to returning the velve to service following maintenance, c.
repair er replacement work on the valve, and
- d. Within 24 hours following valve actuation due to automatic or manual action or flow through the valve. , ,
The provisions of Specification 4.0.4 are not applicable for entry into M00E 3 l i er 4 l 1 3/4 4-23 NUREG 0452/STPEcs ! {t 575 COMPARI'30N l t
Page No. 86 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE # NOTES c* 3/4.4.6.2 PAGE: 3/4 4-24.0 OPERATING MODE: 1234-- RCS - OPERATIONAL LEAKAGE DATA 1) Table 3.4-1. Reactor Coolant System Pressure Isolation Valves. STPEGS specific valves provided. G.
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* - .. :~ - - - - . . . .
DRH I TA8LE 3.4-1 .
- RDCTOR COOLANT SYSTEM PRESSURE 150LAT!0N VALVES FUNCTION VALVE NUM8ER .
NHSI Cold Les Injection Check Valves 15I0007 A. 5. C (RCS Loops 1, 2, 3) BHSI Cold Leg Recirculation Check Valves ISI0009 A. 5, C (RCS Loops 1, 2, 3) LMSI/NHSI Mot Leg Recirculation Check Valv 1810010 A, 5, C (RCS Loops 1, 2, 3) s' 020 A, 5. C LHSI Not les Recirculation Check Valves (RCS Loops 1, 2, 3) LRSI Cold Leg Injection Check Valves IRH0032 A, 5, C (RCS Loops 1, 2, 3) LMSI/HMSI/ Accumulator Cold Les Injection
* %S10038 A, 5. C Chect Valves (RCS Loops 1, 2, 3)
Accumuistor Cold Leg Injection Check Valve 5, C 1510046 A (RCS Loops 1, 2, 31
.. s STP %0Ves prouldsA.
e ) 3/4 A 24 3-5T5 NUREG 0452/STPECS COMPARISON l l *
- L -
Paga No. 87 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPECS TECH. SPECS. NOTE TYPE NOTE f NOTES
** 3/4.4.7 PAGE: 3/4 4-25.0 OPERATING MODE: ,123456 RCS - CHEMISTRY NONE 1) 3/4.4.7 No Changes.
O.
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~ .'**
---~..:.....-...:..--....
O DWT REACTOR COOLANT SYSTEM . 3/4.4.7 CHEMISTRY LIMITING CONDITION FOR OPERATION 1 I 3.4.7 The Reactor Coolant Systes chemistry shall be saintained within the . I limits specified in Table 3.4-2. APPLICA81LITY: At all times. .. ACTION: . MODES 1, 2, 3, and 4:
- a. With any one or more chemistry parameter in excess of its Steady-State Limit but within its Transient Lielt, restore the parameter to within its Staa@-Stata Limit within 24 hours or be in at least NOT
- STAN08Y within the next 6. hours " and in COLD SHUTDOWN within the following 30 hours. M l
- b. With any one or more chemistry parameter in excess of its Transient l Limit, be in at least NOT STAND 8Y within 6 hours and in COLD SHUTDOWN O- within the following 30 hours. t At All other Times:
With the concentration of either chloride or fluoride in the Reactor Coolant Systas in excess of its Steady-State Limit for more than 24 hours ,, or in excess of its Transient Limit, reduce the pressurizer pressure to less than or equal to 500 psig, if applicable, and perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the Reactor Coolant System; detamine that the Reactor Coolant System remains acceptable for continued operation prior to increasing the pressurizer pressure above 500 psig or prior to proceeding to M00E 4. SURVi!LLANCE REQUIREMENTS 4.4.7 The Reactor Coolant System chemistry shall be determined to be within 1 the limits by analysis of those parameters at the frequencies specified in Table 4.4-3. (
, O - .
W-575 3/4 4-25
** NUREG 0452/STPEGS COMPARISON
Page No. 88 l 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE # NOTES
** 3/4.4.7 PAGE: 3/4 4-26.0 OPERATING MODE: ,123456 RCS - CHEMISTRY NONE 1) Table 3.4-2 Reactor Coolant System Chemistry Limits. No Changes.
s O l l 1 e 0
DRAFT TABLE 3.4-2 .
, ~ . REACTOR COOLANT SYSTEM CHEMISTRY LIMITS .
STEADY-STATE TRANSIENT
- LIMIT LIMIT ,
PARAMETER , I Ofssolved Oxygen
- 1 0.10 pps i 1.00 pps 5.0.15 ppai 5 1.50 pps Chloride ,
. 1 0.15 ppe $ 1.50 ppe ,
Fluoride
. ~ \.
less Wn or egal u 04.
- Limit not applicable with T,yg .
-,l
\ l : i 3/4 4-26 W-STS
- NUREG 0452/STPEGS COMPARISON
- v. ,
Page No. 89 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE # NOTES e* 3/4.4.7 PAGE: 3/4 4-27.0 OPERATING MODE: 123456 RCS - CHEMISTRY NONE 1) Table 4.4-3 Reactor Coolant System Chemistry Limits Surveillance Requirements. No Changes. O
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r , : , .. DR:H i . E E_4.4-3 -
~
REACTOR COOLANT SYSTEM - CHEMISTRY LIMITS SURVEILLANCE REQUIREMENTS SAMPLE AND i ANALYSIS FREQUENCY _ PARAMETER At 14ast once per 72 hours Dissolved Oxygen
- a.. At least once per 72 hours Chloride .
At least once per 72 hours Fluoride O . N less than or equal to 250*F
*Not required with T ,yg O
3/4 4-27 NUREG 0452/STPEGS PSTS COMPARISON
~~_
q Page No. 90 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE i NOTES ______ __________________________________________________ {
** 3/4.4.8 PAGE: 3/4 4-28.0 OPERATING MODE: ,12345-RCS - SPECIFIC ACTIVITY ED 1) 3.4.8 Note (*) omitted from STSRev5 appears to be a typo Inserted for STPEGS based on STSRev4.
0 1 ( .e I O
~
- - - - ,--- - ; ,.~.--. . .,. .
REACTOR COOLANT SYSTEM , 3/4.4.6 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 microfurie per gras 005E EQUIVALENT I-133, and
- b. Less than or equal to 100/E aferocuries per gram of gross radioactivity. -
APPLICA81LITY: M00E5 1, 2, 3, 4, and 5. ACTION: MODES 1, 2 and 3*:
- a. With the specific activity of the reactor coolant greater than 1 mic rie per gras DOSE EQUIVALENT I-131 for more than 48 hours ne continuous time interval or for more than 800 hours in during any con'secutive 12-month period, or exceeding the Ifmit line shown O. on Figure 3.4-1, be in at least HOT STAND 8Y with Tavg 1ess than 500*F within 6 hours; and With the gross specific activity of the reactor coolant greater than 1
b. 100/tmicrocuriespergrasofgrossradioactivity,beinatleastNOT
, STAN08Y with T,,,less than 500*F within 6 hours.
MODES 1, 2, 3, 4, and 5: - s With the specific activity of the reactor coolant greater than 1micrq(uriepergras005EEQUIVALENTI-131orgreaterthan100/lmicro-
. furies per gram of gross radioactivity, perform the sampling and analysis
- requirements of Itas 4.a) of Table 4.4-4 until the specific activity of
. the reactor coolant is restored to within its Ifaits.
e 1 pgA-Ee 7ngt CR EG'AL re MCY. f y, O W
,-STS 3/4 4-28 NUREG 0452/STPEGS COMPARISON
Page No. 91 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. - NOTE TYPE NOTE # NOTES c* 3/4.4.8 PAGE: 3/4 4-29.0 OPERATING MODE: 12345-RCS - SPECIFIC ACTIVITY NONE 1) 3/4.4.8 No Changes. i
/
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O' DRWT REACTOR C00LANT SYSTEM . SURVEILLANCE REQUIREMENTS 4.4.8 The specific activity of the reactor coolant shall be determined to be , within the limits by performance of the sampling and analysis program of Table 4.4-4. . O N 9 3/4 4-29 l NUREG 0452/STPEGS PSTS COMPARISON
7 Page No. 92 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE # NOTES
** 3/4.4.8 PAGE: 3/4 4-30.0 OPERATING MODE: ,12345-RCS - SPECIFIC ACTIVITY NONE 1) Figure 3.4-1 No Changes.
O
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l i STP FSAR l 12470-5
. 250 < '
G
.3 2
5 _a b 200 5 UNACCEPTABLE . p OPERATION N D 8 g 150 5 8 C* q - l m E 100 a.
~
m
'_I w ACCEPTABLE b 50 CPERATION t
g . 5 w sn 1 8 - l 0 l 20 30 40 50 60 70 80 90 100
~
PERCENT OF RATED THERMAL POWER SOUTH TEXAS PROJECT UNITS 1 & 2 (\ Dose EquivaJent 1131 Primary Cooient Specific Activity Limit versus Percent of Rated Thermal Power with the Primary Coolant Specific Activity >1.0pci/ gram dose Equivalent i 131 Figure 3.HREG 0452/STPECS
~
COMPARISON 3/4.4 30 L
l i 93 Page No. 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE # NOTES
** 3/4.4.8 PAGE: 3/4 4-31.0 OPERATING MODE: 12345-RCS - SPECIFIC ACTIVITY
- NONE 1) Table 4.4-4 Reactor Coolant Specific Activity Sample and Analysis Program. No Changes.
O
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O- -
)
s
- TABLE 4.4-4
'T REACTOR COOLANT SPECIFIC ACTIVITY SAMLE 'i.
AIS ANALYSIS iGGRAN ' Q *l SAM LE AND ANALYSIS MDOES IN 1AllCH sal #LE TYPE OF NEASURENENT Als ANALYSIS FREQUENCY AND ANALYSIS REQUIRED e
., 'g ,
r Gross Radioactivity At least once per 72 hours. 1,2,3,4 !. 1. Determination * .
- 2. Isotopic Analysis for DOSE EqulVA- 1 per 14 days. I j LENI I-131 Concentration -
1 per 6 months *** 1 ?.
- 3. RadiocheetcalforEDeterminatten**
- a) Once per 4 hours, 18, 28, 38, 48, 58
- 4. Isotopic Analysis for ledine i
Including I-131, I-133, and I-135 whenever the specific e activity exceeds 1 l ,! pCl/gran 005E t8 j i EQUIVALENT I-131 ! T or 200 4 pC1/ gram of l :,
- gross radioactivity, and t*
l 6 1,2,3 .-
- b) One sample between 2
! h and 6 hours following }
.i l . a THERMAL POWER change
- i exceeding 15% .
of the RATED TIEW4AI. i-POWER.within a 1-hour f. period. -
, i
! i. S E ; 8o ,- i i 4o ! %C hb . Cl23 a m -
. .a ,
tcs a i
. j5
f Page No. 94 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE f NOTES
** 3/4.4.8 PAGE: 3/4 4-32.0 OPERATING MODE: 12345-RCS - SPECIFIC ACTIVITY
- NONE 1) Table 4.4-4, Notations. No Changes.
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DRST s TABLE 4.4-4 (Continued) , TABLE NOTATIONS "A gross radioactivity analysis shall consist ofspecific the quanl The total 2 with half-lives less than 10 minutes and all radiofodines. activity shall be the sus of the degassed beta-gamma Detemination of is taken and extrapolated back to when the sample was taken. the contributors to the gross specific activity shall be based upon thoseThe latest a energy peaks identifiable with a 955 confidence level. l data say be used for pure beta-emitting radionuclides. '
**AradiochemicalanalysisforIshallconsistofthequantitativemeasurement of the specific activity for each radionuclide, except for radionuclides with half-lives less than 10 minutes and all radiofodines, which is identified in The specific activities for these individual radio-the reactor coolant.
nuclides shall be~ used in the determination of E for the reactor. sample. Detamination of the contributors to I shall be based upon those energy peaks identifiable with a 95X confidence level. 20 days of POWER OPERATION
*** Sample to be taken after a sinimum of 2 EFPD andhave elap
#Until the specific activity of the Reactor Coolant Systes 'is restored g
within its limits. l l
.u .
e 4 O . NUREG 0452/STPECS
- COMPARIS6N m -
1 Page No. 95 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE # NOTES
** 3/4.4.9.1 PAGE: 3/4 4-33.0 OPERATING MODE: J. 2 3 4 5 6 RCS - PRESSURE / TEMPERATURE LIMITS DATA 1) 3/4.9.1 STPEGS specific values provided.
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; l i REACTOR C0OLANT SYSTEM .
3/4.4.9 PRESSURE / TEMPERATURE LIMITS 1 REACTOR COOLANT SYSTEM 4 LIMITING C0'NDITION FOR OPERATION . . i t i 3.4.9.1 The Reactor Coolant system (except the pressurizer) temperature and l pressure shall be Ifaited in accordance with the limit lines shown on Figures ? l 3.4-2 and 3.4-3 during heatup, cooldown, criticality, and inservice leak and i hydrostatic testing with: Amaximumheatupof\100 in any 1-hour period, a. in any 1-hour period, and .
- b. Asaximumcooldownof\100 10] in any
" c. AmaximumtemperaturechangeoflessthanorequaltDo 1-hour period during inservice hydrostatic and leak tasting operations above the heatup and cooldown Ifeit curves.
APPLICA81LITY: At all times. [ M: With any of the above liefts exceeded, restore the temperature and/or pressure to within the Ifeit within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition on ~# the acceptable for' continued operation or be in at least NOT STAN08Y within the next 6 hours and reduce the RCS T,,,and pressure to less than 200*F and . 500 psig, respectively, within the following 30 hours. t SURVEILLANCE REQUIREMENTS The Reactor Coolant System tamparature and pressure shall be 4.4.9.1.1 detaruined to be within the limits at least once per 30 minutes during system heatup, cooldown, and inservice leak and hydrostatic testing operations. 4.4.9.1.2 The reactor vessel saterial irradiation surveillance specimens shall be removed and examined, to determine changes in material properties, as required by 10 CFR Part 50, Appendix H in accordance with the schedule The results of these examinations shall be used to update in Table 4.4-5. Figures 3.4-2 and 3.4-3. O 3/4 4-33 NUREG 0452/STPEGS F5TS COMPARISON
I Page No. 96 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE # NOTES
** 3/4.4.9.1 PAGE: 3/4 4-34.0 OPERATING MODE: 123456 RCS - PRESSURE / TEMPERATURE LIMITS DATA 1) Figure 3.4-2. Figure for STPEGS provided.
O' o O
STP FSAR 12470-6 0 . , 3000 - MATERIAL PROPERTY BASIS CONTROLLING MATERI AL - RV INTERMEDI ATE SHELL R-1606-3
, LEAK TEST LlHIT COPPER CONTENT: CONSERVATIVELY ASSUMED AS 0.10 WT $
f RT MDT INITIAL: 100F ' RT NDT AFTER 32 EFPY I /4T, 910F 7 2000 3/4T,64'F E CURVE APPLICABLE FOR HEATUP 0",, 0 RATES UP T0100 F/HR FOR' THE SERVICE PERIOD UP TO 32
$$ EFgY AND CONTAINS MARGINS OF g 10 F AND 60 PSIG FOR POSSIBLE g INSTRUMENT ERRORS E
O '- S s e f 1000 f f <, CRITICALITY LIMIT BASED ON INSERVICE HYDROSTATIC TES TEMPERATURE (242 F) FOR HEATUP RA ES THE SERVICE PERIOD UP TO UP TO 100 F/HR 32 EFPY 0 0 100 200 300 400 illDICATED TEMPERATURE (OF) SOUTH TEXAS PROJECT UNITS 1 & 2 Reactor Coolant System Heatup Limitations Applicable to 32EFPY FigureNUREcp452/STPEGS w.u nnm., 3/4.4- 34
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Page No. 97 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. ' NOTE TYPE NOTE #
------------ NOTES i ** 3/4.4.9.1 PAGE: 3/4 4-35.0 l OPERATING MODE: 1 2 3 4 5 6 RCS - PRESSURE / TEMPERATURE LIMITS l DATA 1) i Figure 3.4-3 Figure for STPEGS provided. '
l 9-= I l l k O l l i l- J
t STP TS lO j
- i. <
N00 , W Trasat pseetnTY satis tesit0LLle6 mafttt&L = tv 18718:59:171 OstLL 0 4000 3 COPPtt testtsT 80s8ttVAfttELY AssuM10 8T A8 0.00 8 lIIAE8 88 IW'T5
~
801 'FTER 82 EFFT1 BTggy A 1187. Gl't sist,gn't F
- 2000 StutLt 400L90R Gutet RATES APPLICABLE 5910800{0tffER pet Tag Slavitt Prai40 se TO E
- 32 EfPY. As0 Cesfales matelst
- OF le'f 450 00 P814 fM 905818LE IssfanMts? Ettet8
( I . l - l000 800L90ft SAtt8 t
---.$F.U.?L*L /M .- -
- SS let l l(sao'r)
,a e
9
- 100 30
- 800 INDICAftp TEMPERATURE ('F) ngAt DRAFT February 1985 O SOUTH TEXAS, PROJECT :
UNIT 1 es=2dev= South Texas Project 3/4 4-35 s..eteres Limitations m go to 3MM Figure 3.4-3 _j L j
r- , Page No. 98 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. ' NOTE TYPE NOTE # NOTES
** 3/4.4.9.1 PAGE: 3/4 4-36.0 OPERATING MODE: 123456 RCS - PRESSURE / TEMPERATURE LIMITS DATA 1) Table 4.4-5 STPEGS specific values provided.
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l O BR1FT TABLE 4.4-5 REACTOR VESSEL MATERIAL SURVEILLANCE PROCRA VESSEL LEAD CAPSULE
- FACTOR WITHDRAVAL TIME (ETPY)
NUMBER LOCATION 4.00 1st Refueling U 58.5* 5 241' 3.69 Y 9 61* 3.69 V 15 238.5* 4.00 X Standby 121.5* 4.00 W Standby 301.5* 4.00 2 N
. ei Q .
3/4 4-36 PSTS NUREC 0452/STPECS COMPARISON
[ Page No. 99 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE # NOTES
** 3/4.4.9.2 PAGE: 3/4 4-37.0 OPERATING MODE: 123456 RCS - PRESSURERIZER -
DATA 1) 3.4.9.2 STPEGS specific values provided. i l l . l O J
REA TOR COOLANT SYSTEM PRE 55URIZER' LIMITING CONDITION FOR OPERATION 3.4.9.2 The pressurizer erature shall be Itaited to: .
- a. A maximum heatup of 100h in any 1-hour period,
- b. Amaximumcooldownofk200in any 1-hour period, and Amaximumspraywatertemperaturedifferentialofk320 F.
c. APPLICABILITY: At all t{ses. ACTION:
- With the pressurizer temperature limits in excess of any of the above limits, restore the temperature to within the Itsits within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the pressurizer; determine that the pressurizer .
remains acceptable for. continued operation or be in at least NOT STAN08Y within O( the next 6 hours and reduce the pressurizer pressure to less than 500 psig within the following 30 hours.
$URVEILLANCE REQUIREMENTS l
4.4.9.2 The pressurizer tasperatures shall be determined to be within the The limits at least once per 30 minutes during system heatup or cooldown. spray water temperature differential shall be determined to be within the limit at least once per 12 hours during auxiliary spray operation. f O NUREG 0452/STPEGS COMPARISON 3/4 4-37 y-STS t
Page No. 100 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE i NOTES
** 3/4.4.9.3 PAGE: 3/4 4-38.0 OPERATING MODE: ---456 RCS - OVERPRESSURE PROTECTION SYSTEMS
- 1) 3.4.9.3 PORV maximum lift setting, RCS vent size, and RCS cold leg temperature to be provided LATER (awaiting results of Cold overpressure Mitigation Analysis).
6 l l 6 I i I O I
O ptACTOR COOLANT SYSTEM B 1FTil OVERPRESSURE PROTECTION SYSTEMS
. LIMITING CONDITION FOR OPERATION
~
3.4.9.3 At least one of the following Overpressure Protection Systems shall
. be OPERA 8LE:
- a. Two power-operated reifef valves (PCRVs) with a lift setting of I
.. less than or equal to p psig, or ~
- b. The Reactor Coolant System (RCS)# d epressurized with an RCS vent of greater' than or equal to / g square inches. l APPLICA811.ITY: MODE 4 when the temperature of any RCS cold leg is less than
- or equal u L3Mj'F, MODE 5 and M00E 6 with the reactor vesset head on. -
l LATER. ACTION:
- a. With one.PORY inoperable, restore the inoperable PORV to OPERA 8LE status within 7 days or depressurize and vent the RC5 through at i least a4 M square inch' vent within the next 8 hours. l 1
- b. With both PORVs inoperable, depressurize and vent the RCS through at least a utrE58. square inch vent within 8 hours. l
- c. In the event either the PORVs or the RCS vent (s) are used to mitigate
+ an RCS pressure transient, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within
- O i
30 days. The report shall describe the circumstances initiating the l transient, the effect of the PORVs or RCS vent (s) on the transient,
' and any corrective action necessary to prevent recurrence.
l d. The provisions of Specification 3.0.4 are not applicable. O . 3/4 4-38 W-STS
- NUREG0452/STPEGS COMPARISON L
I Page No. 101 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE # NOTES
** 3/4.4.9.3 PAGE: 3/4 4-39.0 OPERATING MODE: :--456 RCS - OVERPRESSURE PROTECTION SYSTEMS DESIGN 1) 4.4.9.3.1.c ' isolation valve' changed to ' block valve'. STPEGS nomenclature.
O I f O l' ' -r -- -
~ - --- - . .
O ^ D1AFf REACTOR COOLANT SYSTEM CVERPRE55URE* PROTECTION SYSTEM .
$URVEILLANCE REQUIREMENTS 4.4.9.3.1 Each PORV shall be demonstrated CPERABLE by: ' a. Performance of an ANALOG CHANNEL OPERATIONAL TEST on the actuation channel, but excluding valve operation, within 31 days prior to entering a condition in which the PORY is required OPERABLE .
and at least once per 31 days thereafter when the PORY is required OPERA 8LE."
- b. Perfomance of a CHANNEL CALIBRATION on the PORY actuati at least once per 18 month
- c. Verifying the PORV hmth.,va ve is open at least once per 72 hours when the PORV 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 hours" when the vent (s) is being used for overpressure protection. O N. 1
*Except when the unt pathway is provided with a valve which is locked, sealed or otherwise secut ?d in the open position, then verify these valves open at least once per 31 days.
9 0 O NUREG 0452/STPECS 3/4 4-39 COMPARISON W-5TS
i l Page No. 102 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE # NOTES
** 3/4.4.10 PAGE: 3/4 4-40.0 OPERATING MODE: J. 2 3 4 5 6 RCS - STRUCTURAL INTEGRITY NONE 1) 3/4.4.10 No Changes.
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REACTOR COOLANT SYSTEM . 3/4.4.10 STRUCTURAL INTEGRITY _ t LIMITING CONDITION FOR OPERATION
.. The structural integrity of ASME Code Class 1, 2, and 3 components 3.4.10 shall be maintained in accordance with Specification 4.4.10.
APPLICABILITY: All MODES.
. M: '
a. With the structural integrity of any ASME Code Class 1 compone not conforming to the above requirements, restore the structural' integrity of the affected component (s) to within its limit or isola the affected component (s) prior to increasing the Reactor Coolant Systen temperature more than 50*F above the minimum temperatu
, required by NOT considerations.
b. With the structural integrity of any ASME Code Class 2 compone not conforming to the above requirements, restore the structural integrity of the affected component (s) to within its limit or isol the affected component (s) prior to increasing the Reactor Coolant Systes temperature above 200*F. s, c. With the structural integrity of any ASME Code Class 3 compon not conforming to the above requirements, the affected component (s) free service. I
- d. The provisions of Specification 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.4.10 In addition to the requirements of Specification 4.0.5, each rea coolant pump flywheel shall be inspected per the recommendat Position C.4.h of Regulatory Guide 1.14, Revision 1, August 1975. O - 3/4 4-40 W-STS
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NUREG 0452/SW EGS COMPARISON L _
x 9 Page No. 103 4 06/13/85 COMPARISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE # NOTES
** 3/4.4.11 PAGE: 3/4 4-41.0 OPERATING MODE: ,1234-- l RCS - REACTOR COOLANT SYSTEM VENTS l l
l DATA 1) 3/4.4.11 STPEGS specific values provided. I DESIGN OTHER 2) 3.4.11 LCO and Actions revised to reflect STPEGS design and STS changes identified in GL 83-37 (TMI) and consistent with Byron Tech Spec. OTHER 3) 4.4.11.1 Deleted, and 4.4.11.2 modified since LCO requires valves to be closed. 9-O
l 2 ( , REACTOR COOLANT SYSTEM - l . 3/4.4.11 REACTOR COOLANT SYSTEM VENTS * - LINITING CONDITION FOR OPERATION 7' *N ga orLaMTr 3.4.11 h jeactor Coolant System nt path 5 4 consisting ofdtwol1sut V4LES emergency busens shall/be OPERABLE M 2E3
....._ . - red f and closed.
- - -- _. __- __. ..____=..
APPLICA8ILITY: MDOES 1, 2, 3, and 4. - ACTION:' urss n HEAD
- a. With one of the above /eactorA N vent paths inoperable.
STARTUP and/or POWER OPERATION say' continue provided the inoperable vent path is maintained-'closed with power removed from the valve O( actuatorsof all the - vent path; restore the inoperable vent path to OPERA 8LE status within 30 days, or, be in HOT STAN08Y within 6 hours and in valves in the inoperable COLD SHUTDOWN within the following 30 hours, er int neovE wptt HEAD ,', vent paths inoperable;
-D. WithtwoeamessefeactorA_- -- ,_ _
maintain nthe inoperable vent pathsclosed with power removed from the valve actuators of all the sust asissementetsuk valves in the y - inoperable vent paths, and restore at least M of the vent paths to OPERA 8LE status within 72 hours or be in HOT STAN08Y l within 5 hours and in COLD SHUTDOWN within the following 30 hours. 1P SURVEILLANCE REQUIREMENTS __ . . .s___.. _ .
- 7 mm a m .__.a_ a. _
vent path shall be demonstrated 4.4.11.[Eachfeactor OPERA 8LE at least once per la months by:
- a. Verifying all manual isolation valves in each vent path are locked in the open position, em rw vswrperu
- b. Cycling each vent valve 3through at least one complete cycle of full travel from the control room,J:mut Dew. COLD .StWTDOWN ca O~ c.
REFUELING. Verifying flow through the
' venting ounma COLDN .=
y 7 ystem vent paths during SHUTDow[eactorjoolant TG rufL . NUREG 0452/STPEGS 3/4 4-41 COMPARISON ! PSTS 4 b e
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f l NUREG 0452/STPEGS I 4 COMPARISON t
Page No. 104 06/13/85 COMP 7RISON OF NUREG 0452, REV.5, AND STPEGS TECH. SPECS. NOTE TYPE NOTE # NOTES
** 3/4.5.1 PAGE: 3/4 5- 1.0 OPERATING MODE:
- 123---
EMERGENCY CORE COOLING SYSTEMS - ACCUMULATORS DATA ED 1) 3/4.5.1 ' Reactor Coolant System accumulator' name changed to ' Safety Injection accumulator'. STPEGS specific values provided. OTHER 2) 3.5.1 item b. Required borated water volume given as 4-level in addition to the required gallons since water volume is measured as 4-level on the control boards. STPEGS specific values to be provided LATER. DATA 3) 3.5.1 Minimum boron concentration to be provided LATER. i DESIGN 4) 3.5.1 Actions: Separate action statement provided for each LCO. This reflects the significance of deviation from each LCO. DESIGN 5) 4.5.1.1.a.1 Add 'or indications'. l O
l l DRAFT
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( 3/4.5 EMERGE' ICY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS i . LINITING CONDITION FOR OPERATION 4
- Ciseb w @b,.System fra Gef accumulator shall be OPERA 8LE with:
- 3.5.1 Eacpx... .x'....
- a. The isolation valve open, ,
SMO A containedtoborated water volume ofEbetween l
- b. (wre.pending tutwun (terre)% #
ed. O.Amo , level)465}}