ML20087N608
| ML20087N608 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 03/30/1984 |
| From: | PUBLIC SERVICE CO. OF NEW HAMPSHIRE, WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML20087N607 | List: |
| References | |
| NUDOCS 8404040078 | |
| Download: ML20087N608 (589) | |
Text
{{#Wiki_filter:- _ _ _ _ r I I INDEX
- EFINITIONS SECTION PAGE 1.0 DEFINITIONS 1.1 ACTI0N........................................................ 1-1
- 1. 2 ACTUATION LOGIC TEST......... ...................... .. ..... 1-1
- 1. 3 ANALOG CHANNEL OPERATIONAL TEST............... .............. 1-1
- 1. 4 AXIAL FLUX DIFFERENCE......................................... 1-1
- 1. 5 CHANN E L CA LI B RATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-1
- 1. 6 CHANNEL CHECK............................ .................... 1-1
- 1. 7 CONTAINMENT INTEGRITY....................... ............... 1-2 1.8 C O NT R O L L E D L E A KA G E . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-2 1.9 CORE ALTERATION............ .. ... . . ... . ........ .... 1-2
- 10 DOSE EQIVALENT I-131.... .. . ... .... .
.. . .. .. ...... .... 1-2 1.11 E-AVERAGE DISINTEGRATION ENERGY............ .............. .. 1-2 1.12 ENGINEERCO SAFETY FEATURES RESPONSE TIME.. ............... . 1-3 1.13 FREQUENCY N0TATION........................................... 1-3 1.14 IDENTIFIED LEAKAGE. . . . . . . . . . . . . . . . ...................... 1-3 1.15 MASTER RELAY TEST...... ..................................... 1-3 1.16 MEMBER (S) 0F THE PUBLIC.... . ............. ............... 1-3 j 1.17 0FFSITE DOSE CALCULATION MANUAL. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-%3 j 1.18 OPERABLE - OPERABILITY...................... ......... ..... 1-%9 1,19 OPERATIONAL MODE - MODE.... ... ............ ...............
1 'S.1 1.20 PHYSICS TESTS................ ........... ................... 1-4 1.21 PRESSURE BOUNDARY LEAKAGE.................................... 1-4 1.22 PROCESS CONTROL PR0 GRAM...................................... 1-4 1.23 PURGE - PURGING.............................................. 1-W 1.24 QUADRANT POWER TILT RATI0....................... ............ 1-5 1.25 RAT ED TH E RMA L P0WE R. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-5 1.26 REACTOR TRIP SYSTEM RESPONSE TIME..........*.................. 1-5 1.27 REPORTABLE eestFREtteE. EV.WT. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-5 COASTA t MH6A!Y EAM LobuM 1.28 -C::I:L SUILDI'0 INTEGRITY.................................... 1-5 1.29 SHUTDOWN MARGIN.............................................. 1-% 5-1.30 SITE ECUNDARY.............. ................................. 1 'E S'~ j 1.31 SLAVE RELAY TEST............................................. 1-6 W-STS I 8404040078 840330 l PDR ALOCK 05000443 - -- - -- ' A PDR
INDEX DEFINITIONS S SECTION PAGE 1.32 SO LI D I F I C AT I O N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-6 1.33 SOURCE CHECK................................... ............. 1-6 1.34 STAGGERED TEST BASIS........................ ................ 1-6 1.35 THERMAL P0WER.............................. ... ... ...... 1-16 1.36 TRIP ACTI.'ATING DEVICE OPERATIONAL TEST....................... 1-% 6 1.37 UNIDENTIFIED LEAKAGE.......... ...... ................... ... 1-% 6 i 1.38 UNRESTRICTED AREA............. ............. . ... ....... 1-7 - 1 1.39 VENTILATION EXHAUST TREATMENT SYSTEM. . . . . . . ..... . ... .. 1-7 1.40 VENTING....... ........... . ............ .. .. . ... . .. 1-7 WASTE TRMTM 4^'SYSTEM.(M#A#BD. . . . h C.;[" 2
.. ... .. ... . 1-7 i 3.g41x wA5rs Tasamsar- seren TAsLE 1."2, FREQUENCY NOTATION.cueuip) 1-7 .
1-8 l TAELE 1.31 GPERATIONAL MCDES...... ..... .. . . .. 1-53 j t V.-STS II _4
INDEX SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS SECTION PAGE 2.1 SAFETY LIMITS 2.1.1 PEACTOR C0RE................................................ 2-1 2.1.2 kEACTOR COOLANT SYSTEM PRESSURE. . . . . . . . . . . ............ ... 2-1 FIGURE 2.1-1 REACTOR CORE SAFETY LIMIT - FOUR LOOPS IN OPERATION.. 2-2
"* :'J r.: 2.12 RCACTCR ;;; CAT:7 L :'.:7 TORE: w::T; I:, CT: RAT :N :;
2.2 LIMITING SA~ETY SYSTEM SETTINGJ 2.2.1 REACTOR TRIP ' SYSTEM INSTRUMENTATION SETP0INTS............... 2-A3 TABLE 2.2-1 REACTOR TRIP SYSTEM INSTRUMENTAT!CN TR!D SETPOINTS.... 2-54!, BASES SECTION PAGE 2.1 SAFETY LIMITS 2.1.1 REACTOR C0RE.......................... ... ........ ....... B 2-1 2.1.2 REACTOR COOLANT SYSTEM PRESSURE............................. B 2-2 2.2 LIMITING SAFETY SYSTEM SETTINGS 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETP0INTS............... E 2-3 W -STS III
INDEX LIMITING CONDITIONS FOR OPERATION AND SURVE!LLANCE REOUIREMENTS SECTION PAGE 3/4.0 APPLICABIL"TY............................................... 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL Shutdown Margin - T,yg > 200*F........................... 3/4 1-1 Shutdown Margin - T,yg < 200*F........................... 3/4 1-3 Moderator Temperature Coefficient........................ 3/4 1-4 Minimum Temperature for Criticality.. . . . . ..... .... .. 3/4 1-6 3/4.1. 2 BORAT:0N SYSTEMS Flow Path - Shutdown..................................... 3/4 1-7 Flow Paths - Operating....... ...... ......... ...... . 3/a 1-S Charging Pump - Shutcown........ ........... ... .. . ... 3/4 1-9 Charging Pumps - Operating............................... 3/4 1-10 Berated Water Source - Shutdown. .. .............. ...... 3/4 1-11 Borated Water Sources'- Operating.. . . . . . . .. ... . 3/a 1-12 3/4.1.3 MOVABLE CONTROL ASSEMBLIES Group Height.............................. . . . . ... ... 3/4 1-la TABLE 3.1-1 ACCIDENT ANALYSES REOU: RING REEVALUAT:0N :N THE EVENT OF AN INOPERABLE FULL-LENGTH R00................... 3/4 1-16 Position Indication Systems - Operating................ . 3/4 1-17 Position Indication System - Shutcown..... . . . ........... 3/4 1-18 Rod Drop Time............................. . ........... 3/4 1-19 Shutdown Rod Insertion Limit............................. 3/4 1-20 Control Rod Insertion Limi ts. . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/a 1-21 FIGURE 3.1-1 ROD BANK INSERTION LIMITS VERSUS THERMAL POWER FOUR-LOOP 0PERATION...................................... 3/4 1-22 e,-, , o r 4 e wwism
,,a
.e a.
aan nauw vueeneva _. . . w ii h Arld 4 J V E.5d bid -- ,- -- 4 4 8 6 &b ar b r hd FT-- 4.IV Vi t M P P t MPM M M PM A *T P La _
- f. M_**
a6 r.be hw w s wv m31 15 6ws'. .................................... ed 7 & hw W-STS IV W
INDEX LIMITING CONDITIONS FOR OPERAT!CN AND SURVEILLANCE REOUIREMENTS SECTION PAGE 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE.................................... 3/4 2-1 FIGURE 3.2-1 AXIAL FLUX DIFFERENCE LIMITS AS A FUNCTION OF RATED TH E RMA L P0WER . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 2-3 3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR.... ........................ 3/4 2-4 FIGURE 3.2-2 K(Z) - NORMALIZED F q(Z) AS A FUNCTION OF CORE HEIGHT. 3/4 2-17l 3/4.2.3 RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACT 0R............................ ....... ............ 3/4 2-8 FIGURE 3.2-3 RCS TOTAL FLOW RATE VERSUS R, ?' o - FOUR LOOP 99& 0 P E RAT I O N . . . . . . . . . . . . . . . . . . . .'f. . 3/4 . . 2. f'SID ......
- 22": 2. 2 ' ?.:: 2:L *:N1L'" ": l T... T::: : EUPSP.... '
. . . . . -:/t 2 10 1
3/4.2.4 QUAD RANT POWER TI LT RATIO . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 2-1311' 3/4.2.5 CNB PARAMETERS........................................... 3/4 2-1514 TABLE 3.2-1 CNE PARAMETERS........................................ 3/4 2-1618; 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATICN...................... 3/4 3-1 TABLE 3.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION.................. 3/4 3-2 TABLE 3.3-2 . REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES.... 3/4 3-10 l TABLE 4.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE l REQUIREMENTS....................................... ..... 3/4 3-12 1 3/4.3.2 ENGINEERED SAFETY FEATURES ACTUATICN SYSTEM INSTRUMENTATION........................................ 3/4 3-16 i
. TABLE 3.3-3 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION.......................................... 3/4 3'its l TABLE 3.3-4 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM
~
, INSTRUMENTATION TRIP SETPCINTS........................... 3/4 3-112T TABLE 3.3-5 ENGINEEREDSAFETYFEATURESR5FPONSETIMES............. 3/4 3-37.Af TABLE 4.3-2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS. . . . . . . . . . . . . . . . 3/4 3-4240 ; 3/4.3.3 MONITORING INSTRUMENTATION.............................. ,
; -. 17 Radiatien Monitoring E I b NE ffEt W s................ 3/4 3-M O W-STS V
em e m * . .we e mme v -- - , w e vn - o
( INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REOU:REMENTS SECTION PAGE TABLE 3-3-6 RADIATION MONITORING INSTRUMENTATION FO R F L?f" C P : F,.">T I G n 3. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 3-4GY7
~
TABLE 4.3-3 RADIATION MONITORING INSTRUMENTATION TOR Funni
-GAEAAMete SURVEI LLANCE REQUIREMENTS. . . . . . . . . . . . . . . . . . . . . 3/4 3'52S1 Movable Incore Detectors................................. 3/4 3'545.3
. Seismic Instrumentation........................... .... . 3/4 3-5657 TABLE 3.3-7 SEISMIC MONITORING INSTRUMENTATION.................... 3/4 3-355 TABLE 4.3-4 SEISMIC MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS................................... . ... .. 3/43-5RS%d Meteorological Instrumentatien............ . . . . . . . . . . . 3/4 3'5ET7 TAELE 3.3-8 METEOROLOGICAL MONITORING INSTRUMENTATION.., . . . . . . 3/4 3-59J17 TABLE 4.3-5 METEOROLOGICAL MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS............................ . . . . . . . . . . .
3/4 3-4653, Remote Snutdown Instrumentation............... .......... 3/4 3-0s40
, TABLE 3.3-9 REMOTE SHUTDOWN MONITORING INSTRUMENTATION............ 3/4 3-B2?/ !
TABLE 4.3-6 REMOTE SHUTDOWN MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS................................ 3/4 3-Bi 42-Accident Monitoring Instrumentation...................... 3/4 3-B443 TABLE 3.3-10 . ACCIDENT MONITORING INSTRUMENTATION.... . . . . . . . . . . . . 3/4 3-5s4Y TABLE 4.3+7 ACC! DENT MONITORING INSTRUMENTATION SURVEILLANCE l REQUIREMENTS.......... ................. ................ 3/4 3-?sLI
-CL .'.;; 0;t;;t-:... j:t;;;. . . . . . . . . . . 3/
- a-;?
- Fire Detection Instrumentation.................. . . . . . . . . 3/4 3-?&44 TABLE 3.3-11 FIRE. . DETECTION INSTRUMENTATIQN
%t peks c.he.s zeb.eh.hr . . . . .................. 3/4 3'5440I
. 3/4 7-4r Radioactive Liquid Effluent Monitoring Instrumentation... 3/4 3-?G47 TABLE 3.3-12 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION 3/4 3-7170 TABLE 4.3-8 RADI0 ACTIVE LIQUID EFFLUENT MONITORING -
f - INSTRUMENTATION SURVEILLANCE REQUIREMENTS................ 3/4 3-7s72 i Radioactive Gaseous Effluent Monitoring Instrumentation.. 3,/4 3 ' TAB.LE 3.3-13 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION.......................................... 3/43 7'6 TABLE 4.3-9 RADICACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SU:VEILLANCE REQUIREMENTS................ 3/4 3- EU
, L;;;;- P;rt Detection Syrt:- ............................ 3/ * :-02 --
l 3/4.3.4 TURBINE OVERSPEED PROTECTION.............................. 3/4 3- $2T* i l l W-STS VI O
, r. p _ g 3._,.__, .--- ..- . .,._ . - - _ , _ , - , . _
INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RECUIREMENTS SECTICN PAGE 3/4.4 REACTOR CCOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION Startup and Powe r Ope rati o n. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 4-1 Hot Standby.............................................. 3/4 4-2 Hot Shutdown..................... ....................... 3/4 4-3 Col d Shutdown - Loops Fi 11 ed. . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 4-5 Col d Shutdown - Loops Not Fi11 ed. . . . . . . . . . . . . . . . . . . . . . . . . 3/4 4-6
..... . ...,. ........................... o,-
Irri;ted L;;; St;.;.p. .................... : 3-9 3/4.4.2 SAFETY VALVES Shutdown..,...................... ..................... 3/44'S7 0perating............................................. 3/4 4-157 3/4.4 3 PRESSURIZER.............................................. 3/4 4-149 3/4.4.4 RELIEF VALVES................ ........................... 3/4 4-1230 3/4.4.5 STEAM GENERATORS................... .. ... ............. 3/4 4-Isil TABLE 4.4-1 MINIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTED ! DURING INSERVICE INSPECTION............................. 3/4 4-1910 TABLE 4.4-2 STEAM GENERATOR TUBE INSPECTION.... .................. 3/4.4-1911 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE ~ Leakage Detection Systems................................ 3/4 4'2GlW Ope rati o n al Le a ka ge . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 4-2119 TABLE 3.4-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES. . . . . . 3/44-332\l 3/4.4.7 CHEMISTRY................................................ 3/4 4-2411L TABLE 3.4-2 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS. . . . . . . . . . . . . . . 3/4 4-2613 TABLE 4.4-3 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS SURVEILLANCE REQUIREMENTS............................................. 3/4 4-2624 3/4.4.8 SPECIFIC ACTIVITY.....................'................... 3/4 4-27 2dT' FIGURE 3.4-1 DOSE EQUIVALENT I-131 REACTOR COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER
. WITH THE REACTOR COOLANT SPECIFIC ACTIVITY > 1pCi/ gram DOSE EQUIVALENT I-131................ ................... 3/4 4'262)
TABLE 4.4-4 REACTOR COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PR0 GRAM.................................................. 3/4 4-1927 W _-STS VII
. m.. e
INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLA%.E R{pUIREMENTS SECTION PAGE
. 3/4.4.9 PRESSURE / TEMPERATURE LIMITS Reactor Coolant System................................... 3/4 4-M27 FIGURE 3.4-2 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS - ,
APPLICABLE UF TO //,) EFPY................ ......... ... 3/4 4-M3d FIGURE 3.4-3 REACTOR COOLANT SYSTEM C00LDOWN LIMITATIONS - AP P LI CAB LE U P TO e //o ) E F PY. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 4-M31 TABLE 4.4-5
"""" """~' " " " * " g APhy.
SURVEILLANCE rh "E - L'IT::01"L'" L 00:::0U LC.EEMVa.4.S.cM f.4L.4.E . . . . . . . . . . . . . . . . 3/4 4-N31. Pressurizer........... ............. .......... .. ... 3/4 4-5E33 Overpressure Protection Systems.......................... 3/4 4-% 3T FlWAE '3 4-4 Rcs cos.o even reassurr secrecnos serkmr> q 4 -4 3C 3/4.4.10 STRUCTURAL INTEGRITY................... .. ............. 4 4 4-N J7i 3/4.4.11 REACTOR COOLANT SYSTEM VENTS............... .......... 3/4 4-%i? i 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACdUMULAT0RS.................... ........ ....... . ... 3/" 5-; 3/4.5.2 ECCS SUESYSTEMS - T,yg > 350*F...... ................ ... 3/4544 3/4.5.3 ECCS SUBSY!TEMS - T,yg < 350*F.......... .. .. . 3/a5-N 3/4.5.4 BORON INJECTION SYSTEM Boron Injection Tank................................ .... 3/4 5-1!O Heat Tracing............................................. 3/4 5-1Q 11 l l . 3/4.5.5 REFUELING WATER STORAGE TANK......................'....... 3/4 5-b11'L W-STS VIII :
INDEX I. LIMITING CONDITIONS F0E OPERATION AND SURVEILL*NCE REQUIREMENTS SECTION PAGE W - DUAL TYPE CONTAINMENT 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT Containment Integrity.................. ................. 3/4 6-1D W _3',d Containment Leakage............... ........... . . . . . . . . . 3/4 6-20 SEcoupMy courn*unGur avnss s.ssoom ww 2/4 4-so Containment Air Locks.................................. . 3/4 6-6D o,_._2___m ,__,_.2__ , _ , __a em___ _. , . _ . _ . . . . . . . . . . ....... . . . . . . . . . . . . . m. . . U C C '_' - 4 h i-. $g .s..a.......................... . . . . . . U " E-92 Internal Pressure...................................... . 3/46-(D Air Temperature.......................................... 3/46-ND Containment Vessel Structural Integrity. . . . . . . . . . . . . . . . . . 3/46-3SD u Containment Ventilation System. . . ... ............. . . . 3/4 6- MD 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS Containment Spray System................................. 3/46-b Spray Additive System......................... . . . . . . . . . 3/46-ND Cr t 9---+ Na' ' ;; fy:t-- ....... . . . . . . . . . . . . . 2, , 5 ^.D
.nn,ue _ ,. ,_,
. ._ _.._ _r,r..
a m.o.s _ _ . . . . a m r.n.............................. . . . . . ... . ..-.- l 3/4. 6.'Q CONTAINMENT ISO LATION VALVES. . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/46-ND Tass.t 1.6-2 CONTnovnwr rsa 4itoo VAL.ves 3lq s -t:0 3/4.0.'54 COMBUSTIELE GAS CONTROL - Hydrogen Monitors........................................ 3/46-kD El ectri c Hydrogen Recombi ners. . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/46-ND
. . , - . ,- . __o...,_- -
..s...,,,,,c.,..__
Hydroge n Mi xi ng System. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/46-b o,..~ rwc.m.... . . _ _ . ... ._ . .... -m
. ,., n o nnnu ev u . , , e,.
m-r s... ~.................
- . - - . e r,
_. e_ee_- _9 / A g "P traetutu m e , T e t* tras.see m 7 . 9,
-. . . . . . . - _ - _ . . . . . . . .................................... ,/ s . . ' '
W _-STS IX-D AUS 6 1981'
IN:Et ( LIMITINGC5NDITIONSFCROPERATIONANDSURVEILLANCEREOUI;EMENTS SECTION PAGE 5" . . 3/4.6.h 5:: NO""' CONTAINMENT aruci.osus. s u n.p e uc,.
,o_,2 o ,,m __ ,,_ ,, , . _ .
.,,,w --
m --- t Le ta.a. . . . _ . . . . . . . . ,sa, w ha. u..
.';;. cuuiicing Integrity.........................
Aq
....... 3/4 6- N SH e'1 Building Structural Integrity.......... *
........... 3/4 6-36D 5%<~<c Gdasee. [ M O@w<d Gv.losu*c. Ems gency EA2V h4 E4desw.. Coolm9System y t- alp l
l l l I 1 t I i i l l W _-STS X-D
#R 3 0 m3 1
I
~
TNDEX
/
LIITING CONDITIONS FOR OPERATION AND SURVEILLAN*E RECUIREMENTS SECTION PAGE 3/4.7 PLANT SYSTEMS
'3/4.7.1 TURBINE CYCLE Safety Valves....................'....... ................ 3/4 7-1 TABLE 3.7-1 MAXIMUM ALLOWABLE POWER RANGE NEUTRON FLUX HIGH -
SETPOINT WITH INOPERABLE STEAM LINE SAFETY VALVES DURING FOUR LOOP 0PERATION...................................... 3/4 7-2
..u..m e 3___.i_., u,vrunu n , ., _
nm n , e' n nt .- e _ _ . . . . . . . . . . . e w-e _ _ i.ei rr e n e e, n_v m...-. "
- m. . - - - . . - , . . - . _ _ _
_,,.,,....,......----...--..fbo.
..1 6 F1 e 3ii 41.IieT N .6 es i .
- e. . a v . Y 7 4. V b .2 w . Th. il 6. I fe s3 E E '
=, y. 6
___. .,...an....
. ............................................. w,, ,. , . .
TABLE 3. 7-li'.i. STEAM LINE SAFETY VALVES PER L00P. . . . . . . . . . . . . . . . . . . . . 3/4 7-3 P=".rs wy', = ,7-Feedwater System............................... 3/4 7-4 Condensate Storage Tank.................................. 3/4 7-6 pri'i: Activity........................................ 3/4 7-7 TA5*.E 4.7-1 SECONDARY COOLANT SYSTEM SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PR0 GRAM.................................... 3/47-5li
< Main Steam Line Isolation Va1ves......................... 3/4 7-9 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION. . . . . . . . . . 3/4 7-10 uema #, - -
3/4.7.3 ' W.v.PONENT COOLING WATER SYSTEM.......................... ~3/4 7-11 3/4.7.4 SERVICE WATER SYSTEM..................................... 3/4 7-12 3/4.7.5 ULTIMATE HEAT SINK...................................... 3/4 7-13
,o _ _ - rn...
3/4. 7.b CONTROL ROOM .Mo.W... .f C v AIR CLEA:,'UP SYSTEM. . . . . . . . . . . . . . . . 3/4 7-N.IV
, , , , . gg ,.. ..- - . . , ,
.,,.3 . . . . . - ...e . . .
m. m.m, . . . m .e ,.
. m . .s .m.~. _
3/4.7.N l SNUBBERS................................................. 3/4 7-M 16 TABLE 3.7-ka SAFETY-RELATED HYDRAULIC SNUBBERS.................... 3/47-Maf TABLE 3.7'kb SAFETY-RELATED MECHANICAL SNUBBERS................... 3/4 7-M L FIGURE 4.7-1 SAMPLE PLAN 2) FOR SNUBBER FU."CTIONAL TEST...... .... 3/4 7-NE 3/4.7.}QS SEALED SOURCE C0NTAHlMATION............:................. 3/4 7'26eN
. 4 4.-
W -STS XI m
-+ . ~ , - , - - , - - ,. , g ,
1NDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REOUIREMENTS . SECTION PAGE 3/4.7.119 FIRE SUPPRESSION SYSTEMS Fire Suppression Wated System............................ 3/4 7-}946 Spray and/or Spri nkl er Systems. . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 7-?s27
>- 2 ... -
ww -g -
............................................. .- - ,i, , w,,e t'9-Co,+--.- _$ .f a , 7_ a ?
11u.. s- -- -............................................ - . , Fire Hose Stations.J:.................... ............... 3/4 7'SRJL TABLE 3.7-$ FIRE HOSESTATIONS.................................... 3/4 7-}G3 2h Yard Fire Hydrants and Hydrant Hose Houses... ...:....... 3/4 7-4C 33 TABLE 3.7-i YARD FIRE HYDRANTS AND ASSOCIATED HYDRANT 3/47-4s3k HOSE HOU 3/4.7.1210 FIRE RATED ASSEMELIES.......................... ......... 3/4 7-4C1Y-3/4.7.btllAREA TEMPERATURE MONITORING.............................. 3/4 7 34.2G TABLE 3.7-I'AREATEMPERATUREMONITORING.........,.................. 3/47-4s39'.i 3/4.8 ELECTRICAL POWER SYSTEMS . 3/4. '8.1 ' ~A.C.
~
5,0URCES '.' , Operating................................ .......... .... 3/4 8-1 TABLE 4.5-1 DIESEL GENERATOR TEST SCHEDULE.. ........
.. . .. . 3/4 S'il '1
.: _ : 1. m & svnw o w-_, . .nu y s c.m .... ._..._
-<~m Shutdown................................................. 3/4 8-8 3/4.8.2 0.C. SOURCES 0gerating................................................ 3/4 S-9 TABLE 4.8'S BATTERY SURVEILLANCE REQUIREMENTS..................... 3/4 8-11 l Shutdown................................................. 3/4 S-12 3/4.8.5 ONSITE POWER DISTRIBUTION SYSTEMS 0perating................................................ 3/4 8-13 Shutdown........................ .................... .. 3/4 8-15 Trip Ck
- t V.ewsAv 12 A 3lH 8-lb 46 W-STS XII
INDEX ( LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REOUIREMENTS SECTION PAGE 3/4.8.4 ELECTRICAL EQUIFMENT PROTECTION DEVICES A.C. Ci rcuits Inside Primary Containment. . . . . . . . . . . . . . . . . n 3/4 8-K Containment Penetration Conductor Overcurrent gp Protective Devices..................................... 3/4 8-17 TABLE 3.8-1 -es998. CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT 2o PROTECTIVE DEVICES.................................... 3/4 8-h Motor-0perated Valves Thermal Overload Protection. P.c.v.m,. 4. _ 3/4 8-2C '
. w er a ,
S2 TABLE 3.8-2 MOTOR-OPERATED VALVES' THERMAL OVERLOAD PROTECTION yy
"'D/C P 0'f rA;; DEVICES. cf.eM4M . Ar. 4% .TM*.4.V : . . . . . .... 3/4 8-h.
3/4.9 REFUELING OPERATIONS 3,'4.9.1 BORON CONCENTRATION................................ ..... 3/4 9-1 3/4.9.2 INSTRUMENTATION......................... .............. . 3/4 9-2 3/4.9.3 DECAY TIME............................................... 3/4 9-3 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS.. ... ........... .. 3/4 9-a 3/4.9.5 COMMUNICATIONS....................... .. ................ 3/4 9-5
~
REFREIf44 M4C.t1/v 3/4.9.6 """IPULCC.7 C?X. . . . .E............. .... ............... 3/4 9-6 l l 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE POOL BUILDING.......... 3/4 9-7 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION High Water Level......................................... 3/4 9-S - Low Water Level.......................................... 3/4 9-9 l l W-STS XIII .
. . . . . _ . _ ____ ..= . . . _ _ _ _ _ _ _ . . . . _ . _ _
INDEX i LIMITINGC$NDITIONSFOROPERATIONANDSURVEILLANCERE0VIREMENTS SECTION PAGE 3/4.9.9 CONTAINMENT PURGE AND EXHAUST ISOLATION SYSTEM........... 3/4 9-10 . 3/4.9.10 WATER LEVEL - REACTOR VESSEL............................. 3/4 9-11 3/4.9.11 WATER LEVEL - STORAGE POOL ............................. 3/4 9-12 fuCL swMGd BWLDl416 EMBA&MC 3/4.9.12 STO .?.CE "00L ^.!R C'_E"9 SYSTE". .f . . . . . . . . . . . . . . . . . . . . . . . . 3/4 9-13 nIA CLE4NINC SYSTB*1 3.*.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUT'DOWN MARGIN............................... .. . . . . 3/4 'C-1 O 3/4.10.2 GROUP HEIGHT, INSERTION AND POWER DISTRIBUTION LIMITS.... 3/4 10-2 3/4.10.3 PHYSICS TESTS...................... . . . . . . . . . . . 3/4 10-3 3/4.10.4 REACTOR COOLANT L00PS............................. ...... 3/4 10-4 3/4.10.5 POSITION INDICATION SYSTEM - SHUTDOWN.... . . . . . . . . . . . . . . 3/4 10-5 3/4.11 RADI0 ACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS Concentration............................................ 3/4 11-1 TABLE 4.11-1 RACI0 ACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PR0 GRAM.................................................. 3/4 11-2 Dose...................................................... 3/4 11-56 i Liquid Radwaste Treatment System......................... 3/41141 Liquid Holdup Tanks...................................... 3/4 11- M W _-STS XIV
** * * * * * . *** **" m *Em em es
- e3 96- . . . **.e= * * .eems e # .e.em p w* *- *N
-, ,, ,- -,--p- , ,,--g, n. , - ,-,-, - yy- g
INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEIU ANCE RECUIREMENTS SECTION PAGE 3/4.11.2 GASEOUS EFFLUENTS , Dose Rate................................................ 3/411'Bf} f I TABLE 4.11-2 RADIOACTIVE GASEOUS WASTE. SAMPLING AND ANALYSIS : PR0 GRAM.................................................. 3/4 1139/d! Dose - Noble Gases....................................... 3/4 11-12/y Dose - Iodine-131, Iodine-133, Tritium, and Radicactive l Materi al i n Parti cul ate Fo rm. . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 11-1GLf; Gaseous Radwaste Treatment System. . . . . . . . . . . . . . . . . . . . . . . . 3/4 11-541& Explosive Gas Mixture.................................... 3/4 11-15 Q:
... w.moy ...m............................................ .,, . ._to
. I i
, 3/4.11.3 SOLID RADI0 ACTIVE WASTE.................................. 3/4 11-2417!
3/4.11.4 TOTAL 00SE................................... ........... 3/4 11-19 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING i l 3/4.12.1 ' MONITORING PR0 GRAM....................................... 3/4 12-1 TABLE 3.12-1 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM........ 3/4 12-3 : TABLE 3.12-2 REPORTING LEVELS FOR RADIOACTIVITY CONCENTRATIONS IN ENVIRONMENTAL SAMPLES................................. 3/4 12-9 TABLE 4.12-1 DETECTION CAPABILITIES FOR ENVIRONMENTAL SAMPLE } ANALYSIS............................................. 3/4 12-10 } 3/4.12.2 LAND USE CENSUS.......................................... 3/4 12-13 3/4.12.3 INTERLABORATORY COMPARISON PR0 GRAM....*................... 3/4 12-14 l W-STS XV
INDEX
/
BASES SECTION PAGE 3/4.0 APPLICABILITY......................... ..................... B 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTR0L.....,.................................... B 3/4 1-1 3/4.1.2 BORATICN SYSTEMS.......................... ..... ........ B 3/4 1-2 3/4.1.3 MOVABLE CONTROL ASSEMBL!ES... ...... ...... . ...... ..... B 3/4 1-2 3 '4. 2 DCWER DISTRIEUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE............... ... .. .. .... .... B 3/4 2-1 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR anc RCS FLCW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR..... B 3/4 2-2 FIGURE B 3/A.2-1 TYPICAL INDICATED AXIAL FLUX DIFFERENCE VERSUS I THERMAL POWER.( Fvn . F.tM.T. .@&4) . . . . . . . . ..... . B 3/4 2-3 j 3/4.2.4 QUADRANT PCWER TILT RATIO.... .................. . ....... B 3/4 2-5 6 l 3/4.2.5 DNB PARAMETERS..................... ......... ... ... .... B 3/4 2'54 ( 3/4.3 INSTRUMENTATION l 3/4.3.1 and 3/4.3.2 REACTOR TRIP SYSTEM and ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION................ B 3/4 3-1 3/4.3.3 MONITORING INSTRUMENTATION................................ B 3/4 3-2 3/4.3.4 TURBINE OVERSPEED PROTECTION......*......... .. .......... B 3/4 3-5 l FLC. TARE S Jly l 'L T V P Kas. T A* 0 t C A TEJ 4x1 A d. Em DIF/d4EN(.d VEcy g J/y l y j THt4M*L towf A (sansr cons 7seco wp/Mr4 And Reto40 (aAE.5) l l i i W-STS XVI
INDEX [ BASES SECTION PAGE 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION. . . . . . . . . .3. .3/4 . 4-1 3/4.4.2 SAFETY VALVES........................... ................. B 3/4 4-2 3/4.4.3 PRESSURIZER........................................... ... B 3/4 4-2 3/4.4.4 RELIEF VALVES............................................. E 3/4 4-3 3/4.4.5 STEAM GENERATORS..................................... .... B 3/4 4-3 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE. . . . . . . . . . . . . . . . . . . . . . . . . . . S 3/4 4-4 3/4.4.7 CHEMISTRY.................................. ......... E 3/4 4-5 3/4.4.8 SPECIFIC ACTIVITY................. ..................... . B 3/4 4-5 3/4.4.9 PRESSURE / TEMPERATURE LIMITS.............. ............... 5 3/4 4-5 i t TABLE B 3/4.4-1 REACTOR VESSEL TOUGHNESS.......................... B 3/4 4-9 FIGURE 9 3/4.4-1 FAST NEUTRON FLUENCE (E>1MeV) AS A FUNCTICN OF FULL POWER SERVICE LIFE.................. ........ .... . 5 3/4 4-10 _ '""^; --u cecerr ne e,"-o-- ------o-o- -u co-e-
'~6go,_3 -
__ ___:- /;::.T*;"e;ppe;g ::".::a. - -
.NDT'*""'^"'*"'"----~ - - - - ' - ~ ~ ~ '
Trunema-".c
. _ - -.,o.
3/4.4.10 STRUCTURAL INTEGRITY..................................... B 3/A A-15.tei i 3/4.4.11 REACTOR COOLANT SYSTEM VENTS. . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 4-It,ldI ( 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 AC C UMU LATO R S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 5-1 3/4.5.2 and 3/4.5.3 ECCS SUBSYSTEMS............................... B 3/4 5-1 3/4.5.4 BORON INJECTION SYSTEM.................................... B 3/4 5-2 3/4.5.5 REFUELING WATER STO RAGE TANK. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 5-2 a W-STS
- XVII
l I INDEX BASES
- SECTION PAGE W - DUAL TYPE CONTAINMENT 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT................................ . ... B 3/4 6-1D 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS. ..... . . .... ..... B 3/4 6-3D
,, e._.;
r, . . -- -
..siat nt.muvAL sraicna................... ................ ..;,,,. ;. .- -
3/4. 6.4.3 CONTAINMENT ISOLATION VALVES. . . . . . . . . . . . . . . . . .... .. S 3/4 6-4D 3/4.6.*E'{COMSUST!BLE GAS CONTROL... ............. .... .... ..... E 3/4 6-ED
-- -- .... . ~ . . . . c............rw...
~...5 r . . n . :. . . . . . . . . .. .. . . . - . .__ .... .. .. . ..
u e, ,, ~ --,,-- . --
,, ._en .
, ~ow.. . . . . . . .nuvca..................... . ....... ... .. _ , .z 3/4.6. '- N:7 ~ CONTAINMENT. f^#.C.WMd . .SU.40[MG. . . ... E 3/4 6-ED
\ W-STS XVIII-D APR 3 0 EIS
.~
f INDEX BASES SECTION PAGE 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE................................ ............ B 3/4 7-1 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION........... B 3/4 7-3 . ARMi 3/4. 7Q.3 COMPONENT COOLING WATER SYSTEM............................ B 3/4 7-3 3/4.7.4 SERVICE WATER SYSTEM....... .............................. B 3/4 7-3 3/4.7.5 ULTIMATE HEAT SINK.......................... ............. 8 3/4 7-3
.............. ....... . ............... .... =i,s-HAoGu.P 3/4. 7.N4 CONTROL ROOM .".:R :."C'l AIR CLEANL'? SYSTEM. . . . . . . . . . . . ,
... B 3/4 7-4
,,. e.., ---. -. . -. ... -. . ._--
., ..e- ....-,....
m, uwwu anon... n . c. .. anni..n 2:sicn. .... ..... e o/* e - 3/4.7l$) SNUBBERS.................................................. B 3/4 7-5 3/4.7.1CISEALED SOURCE CONTAMINATION. ... ............ . ...... . E 3/4 7-6 3/4.7.119FIRESUPPRESSIONSYSTEMS.................................. B 3/4 7-5 3/4. 7.13 CFIRE RATED ASSEMBLIES. . . . . . . . . . . . . . ................... B 3/4 7-7 3/4.7.13llAREA TEMPERATURE MONITORING................ .............. B 3/4 7-7 l i 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1', 3/4.8.1 and 3/4.8.3 A.C. SOURCES, D.C. SOURCES AND l ONSITE POWER DISTRIBUTION SYSTEMS....................... B 3/4 8-1 ! 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTION 0EVICES................... B 3/4 8-3 PSTS XIX EV. 2 1981
INDEX t EASES SECTION PAGE 3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION...................................... B 3/4 9-1 3/4.9.2 INSTRUMENTATION........................................... B 3/4 9-1 3/4.9.3 DECAY TIME................................................ B 3/4 9-1 3/4.9.4 CONTAINMENT BUII. DING PENETRATIONS........ .............. B 3/4 9-1 3/4.9.5 COMMUNICATIONS............................................ B 3/4 9-1 REFRELJ Af f- rf 4 chi 4/E 3/4.9.6 " A"I P U LATC P, C F ^ " E . . . . . . . . . . . . . . . . . . . . . . ............. E 3/4 9-2 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE BUILDING.... ........... B 3/4 9-2 3/4.9.8 RESIDUAL HEAT REMOVAL AND COCLANT CIRCULATION....... .... E 3/4 9-2 3/4.9.9 CONTAINMENT PURGE AND EXHAUST ISOLATION SYSTEM............ B 3/4 9-2 2/4.9.10 and 3/4.9.11 WATER LEVEL - REACTOR VESSEL and STO RAG E P 0 0 L . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 9-3 e.rme F LLEi. ,STO M.GC,, B.iA.u..o. ia.S Ene&cevcy
,.4. w a w nr,w e mee
, m
,. / 4. es . .,
1 www a u a e , a 4 w.i ru r u .--o
.... ............... ........ .. o ,/ *v u
- e. 2 AIR CL.64AllA7G. 5457FF1 3/4.10 SPECIAL TEST EXCEPTIONS l 3/4.10.1 SHUTDOWN MARGIN........................................... B 3/4 10-1 l
3/4.10.2 GROUP HEIGHT, INSERTION AND POWER DISTRIBUTION LIMITS..... B 3/4 10-1 1 . ! 3/4.10.3 PHYSICS TESTS............................................. B 3/4 10-1 l 3/4.10.4 REACTOR COOLANT L00PS..................................... B 3/4 10-1 1 1 3/4.10.5 POSITION INDICATION SYSTEM - SHUTU0WN..................... B 3/4 10-1 l t l . W-STS xx SEP.15 1981 l
INDEX i EASES 3/4.11 RADI0 ACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS........................................ B 3/4 11-1 3/4.11.2 GASEOUS EFFLUENTS....................................... B 3/4 11-2 3/4.11.3 SOLID RADIOACTIVE WASTE................................. B 3/4 11-5 3/4.11.4 TOTA L 00 S E . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 11-6 3/a.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/a.12.1 MONITORING PROGRAM........ ... ........................ B 3/4 12-1 3/4.12.2 LAND USE CENSUS........................................ B 3/4 12-2 3/4.12.3 INTERLABGRATORY COMPARISON PR0 GRAM. . . . . . . . . . . . . . . . . . . . . . B 3/4 12-2 l i 1 _W-STS XXI
INDEX l DESIGN FEATURES SECTION . PAGE 5.1 SITE , 5.1.1 EXCLUSION AREA.............................................. 5-1 5.1. 2 LOW POPULATION 20NE......................................... 5-1 5.1. 3 #M. . n.M. . &. . .M. . .S.%.. .#. .e.b. ..@. . eWM... e fnfd.H#.n.^.f.f4. .u. e_ c_a._e r_
. - , v ..w.
n . . , n, e-e,,, m ,. S. I H -.m stre. nouss o . n.,. ,rm. e . s..s...... 5_1 ero. ot. ca u.t o k. f .h. ..........................
.ueNr.s .y:. s FIGURE 5.1-1 "XCLU"'0" ."".CA. .s ra .+p. AEW4.H.Mc+. seuwsdAv . 5-2 FIGURE 5.1-2 LOW PO P U LATI O N 20 N E . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-3 FIGURE. 5.1 H. @.. ...~mm.
.P,,,E f.W .m 8 W.,DM unam WM rrwo M W 7 %^.'......... ....... 5_ d.
.~..m . .,.,__ w e,.:_ e,,m.n.env en,
-_ . . . n ,,n,,,-.,
w ,,,.o..e.............
... - . . . . q. .
5.2 CONTAINMENT 5.2.1 CONFIGURATION............ ..................... . . . . . 5-1 i-5.2.2 DESIGN PRESSURE AND TEMPERATURE................. ......... . 5-11 . l 5.3 REACTOR CORE 5.3.1 FUEL ASSEMBLIES.... . ... .......... . . . . . . . . . 5 '!iS' 5.3.2 CONTROL ROD ASSEMBLIES................ ..... ... . ......... 5 '5S 5.4 REACTCR COOLANT SYSTEM
- 5. 4.1 DESIGN PRESSURE AND TEMPERATURE..... ................ . . . . 5-%S'
! 5.4.2 V0LUME.......................................... ... ....... 5-7 S
- 5.5 METEOROLOGICAL TOWER L0 CATION................................. 5-1T-
- 5. 6 FUEL STORAGE
- 5. 6.1 CRITICALITY.......................................... ...... 5-M
! 5.6.2 0RAINAGE.................................................... 5-%/, 5.6.3
~
CAPACITY.................................................... 5-M l , 5.7 COMPONENT CYCLIC OR TRANSI'ENT LIMIT........................... 5-1 d TABLE 5.7-1 COMPONENT CYCLIC CR TRANSIENT LIMITS.............. ... 5 '87 W-STS XXII ee. mee *6
. e
- INDEX
( AD"INISTRATIVE CONTRCLS SECTION PAGE 6.1 RESPONSIBILITY............................................... 6-1 6.2 ORGANIZATION ............................................... 6-1 6.2.1 0FFSITE................................................... 6-1 M7NW 6.2.2 # E ~c STAFF.............~................................... 6-1 FIGURE 6.2-1 0FFSITE ORGANIZATION............................... .6-3 FIGURE 6.2-2 STATION ORGANIZATION........................ ...... 5-4 TAELE 6.2-1 MINIMUM SHIFT CliEW CCMPOSITI0h...................... 6-5 wOP u - (m eA+D_< av4e,4 o . 9. 3 . . . . . . . . . . . V. CNG.TNCCP..Th'C.
. _. . . . . ',{gcyfod
.'.... ................. A. %." ".7 Function........... ...................................... 6-E 7 Codositien..................................... . ......
6- V R e s ; c r. s i t: i l i t i e s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5- 5s7 4 t c . +.
. N. .*_ n_ . __ ................................................ 6.b7 6.2.4 SHIFT TECHNICAL ADVISDR......................
......... .. 6-$ '7 STkT1on) 6.3 44: STAFF 0VALICICATIONS................. ...... .... . . . 6 '5.'7 6.4 TRAINING.................................................... 6'%.f-t 1
6.5 REVIEW AND AUDIT..................................... ..... 6- V sinTicM n e,,, er, , Pen.kTled R.EVtEW com ITME 6.5.1 , , , , , - vna. m.*u unovP......................................... c.xs V l Function.................................................. 6- M - . Composition............................................... 6-%f' Al t e r n a t e s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-V Meeting Frequency......................................... 6-1Lj Qucrum.................................................... 6-11
-- R e s p o r} s i b i 1 i ti e s . . . . . . . . . . . . . . . . . . . . . . ..................
6-R T As%eh Records................................................... 4m 6-3/0 1 W-STS XXIII 6
INDEX ( ~ ADMINISTRATIVE CONTROLS SECTION . pagg.29k. $Pf6T"'1 petDIT MO 6.5.2 ,. m m . . . . - R6VIC*', Cones ATEE
....nn. nuver.an ktvitw anu n...,, ue.. unsur.................... 6-%IO Function.................................................. 6-3 /0 Composition...................................... . ......'
6- 3011 Alternates...... ......................................... 6- M f( Consultants.................. .................... ...... 6-30it Meeting Frequency......................................... 6-10 11 Quorum....................................... ............ 6-Mil Review........................ ............. ....... ..... 6-M 12. Audits. 6-K l"4-Am Records. n eriH.............................. ...... ... ......... 1, ; .+
; ;j }
GN E AIT 6.6 REPORTABLE OC: r :NCE ACTION............ ..... ... ...... . 6-M /y 6.7 SA ETv LIMIT VICLATION.......... .. .. .... . .. .. . ... 5-5 /'f 6.8 PROCEDURES AND PROGRAMS................ ........ .. .. .... 6- 3 14 6.9 REPCRTING REQUIREMENTS................ ........... ....... 6-% I 7 6.9.1 ROUTINE REPORTS AND wuraeus a f aurs .. . .. .. 6-Nil Startup Report.......... ................. ..... . . .. 6- S il Annual Reports............................................ 6-% O Annual Radiological Environmental Coerating Report. ...... 6- N Semiannual Radioactive Effluent Release Report... ........ 6-?S 16 i Monthly Operating Report.................................. 6-M W l i;,; ; ;;m 0;; . . s . .m s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . G l ." . m..g nu u i s .L.w.. -M. "-
- an M 'n g..... .......,.... -e-tt-
,u...o,.
,c w
. . , ... .n nogw. w................................ 4 l Radial Peaking Factor Limit Report. . . . . . . . . . . . . . . . . . . . . . . . 6-M 19 i
6.9.2 SPECIAL REP 0RTS........................................... 6-N2.3 6.10 RECORD RETENTION........................................... 6 '24 ~2 3 W-STS XXIV l' . . . . . . . . ...
. $.g.,
h-- .
, , . . . . - ~ . . . . ..
7 INDEX (
.AOMINISTRATIVE CONTROLS -
SECTION . 6.11 RADIATION PROTECTION PR0 GRAM............................... 6-%2.9 ~ 6.12 HIGH RADIATION AREA-(0pt h.,.U............................. 6-25 6.13 PROCESS CONTROL PROGRAM.................. ..... . . . . . . . . . . 6-26 6.14 0FFSITE DOSE CALCULATION MANUAL......................... .. 6-MN 6.15 MAJOR CHANGES TO RADIOACTIVE LIOUID GASEOUS. AND 20 LID WASTE TREATMENT............................................ 5- [
/
wW k.-STS XXV
" ^ * ' " * * * ' * * * " **
et + e . . e. e .,e e e. ' ** *
- .- _ .,w _ _ ,
9, , ,.,. , , , _ _ . - . - , - - , . , , y,. -
--rm-- -~- -m- ---- - v-wu -- e-- y-- r-
e
/
SECTION 1.0 DEFINITIONS e S
.k
)
0
* * * ** " ' * -* . = --
_ _ . - ._ , , y - _ _
,,y ,
JUSTIFICATION Section 1.0 In the text of Section 1.0 where a capital letter with a circle around it appears, please refer to the letter below for the appropriate justification. A. Site Boundary - Additional clarification in order for the definition to apply to the gaseous effluent dose calculations. B.- Solidification - Seabrook Station specific. C. Unrestricted Area - A specific definition was needed to address liquid effluent dose calculations. D. Clarifies types of waste systems and streams as well as specific site receiving Seabrook waste. E. seabrook Station does not dewater resins for shipment as they are incorporated within the binder agent for SOLIDIFICATION as stated in the PCP for the permanently installed solidification system. F. Required per Generic letter 83-43 G. Seabrook Station plant specific information. l. l l
(
- 1. 0 DEFINITIONS The defined terms of this section appear in capitalized type and are applicable throughout these Technical Specifications.
ACTION 1.1 ACTION shall be that part of a Technical Specification which prescribes remedial measures required under designated conditions. ACTUATION LOGIC TEST -
- 1. 2 An ACTUATION LOGIC TECT shall be the application of various simulated input combinations in conjunction with each possible interlock logic state and verification of the recuired logic output. The ACTUATION LOGIC TEST sna11 include a continuity check, as a minimum, of output cevices.
ANALOG CHANNEL OPERATIONAL TEST 1.3 An ANALOG CHANNEL OPERATIONAL TEST shall be the injection of a simulatec signal into the channel as close to the sensor as oracticatie to verit- ' OPERAEILITY of alarm, interlock and/or trip functions. Tre AA ALOG Cda','iEL OFERATIONAL TEST shall incluce adjustments, as necessary, of tne alarm, inter-lock and/or Trip Setpoints such that the Setpoints are within the required range and accuracy.
.~ AXIAL FLUX DIFFERENCE l 1.4 AXIAL FLUX DIFFERENCE shall be the difference in normali:ed 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 with the required range and accuracy to known l values of input. The CHANNEL CALIBRATION shall encompass tne entire cnannel including the sensors and alarm, interlock and/or trip functions and may be performed by any series of sequential, overlapping, or total channel steps ,
such that the entire channel is calibrated, CHANNEL CHECK . t. l - 1.6 A CHANNEL CHECK shall be the qualitative assessment of channel behavior 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 indepencent instrument channels measuring the same parameter. W-STS 1-1 I i
CECINITIONS I 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 OPERABLE 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-\) of Specification (3.6.3).
2
- b. All equipment hatches are closed and sealed,
- c. Each air lock is in compliance with the requirements of Specification (3.6.1.3),
- d. The containment leakage rates are within the limits of Specification r (3.6.1.2), and
- e. Tne sealing mechanism associated with eacn cenetration (e.g., weics, bellows, or 0 rings) is OPERABLE.
' CONTROLLED LEAKAGE 1.8 CONTROLLED LEAKAGE shall be that seal water flow su; plied to tne reacter coolant pump seals.
CORE ALTERATION
- 1. 9 CORE ALTERATION shall be the movement or manipulation of any cc peneht within the reactor pressure vessel wi*h the vessel head removed and fuel in the vessel. Suspension cf CORE ALTERATION shall not preclude completion of movement of a component to a safe conservative position.
DOSE EOUIVALENT I-131 1.10 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (micrcCurie/ gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present. The thyrcid dose conversion factors used for this calculation shall be those listed in NAC {Tib't III e' TID 'i"?4, "0;;;;1: tier ar nietanc*. r-r+a-e ene paw r and-T;;; j R;;;ter Cit;;" Or Tit!: E-7 ef ""C "r;;i.;ery Ouid: 1.109, Revi:i;r. 1, ,
^:tQu, 15777. Iteleases, Rey \ den & Re& EHlue*&s!'Gw'de i.'% "Gicu% & Annul ases k% .r , k Y - AVERAGE DISINTEGRATION ENERGY
- 1.11 I shall be the average (weighted in proportion to the concentration of i each radionuclide in the sample) of the sum of the average beta and gamma
!. energies per disintegration (MeV/d) for the radionuclides in the sample. l l W-STS 1-2
DEFINITIONS p ENGINEERED. SAFETY FEATURES RESPONSE TIME 1.12 The ENGINEERED SAFETY FEATURES RESPONSE TIME shall be that time interva from when the monitored parameter exceeds its ESF Actuatior. Setpoint at the channel sensor until the ESF equipment is capable of performing 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 sequence loading delays where applicable. FREQUENCY NOTATION 1.13 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1/3s 2 IDENTIFIED LEAKAGE 1.14 IDENTIFIED LEAKAGE shall be:
- a. Leakage (except CONTROLLED LEAKAGE) into closed systems, such as pumo seal or valve packing leaks that are captured and conducted to a sumo or collecting tank, or
- b. Leakage into the containment atmeschere frem sourcer that a e bet-soecifically located and known either n:t to interfere vith t.a ::eraticr cf Leakage Detection Systems or not to be PRESSURE SOUNDARY LEAKAGE, or
- c. Reactor Coolant System leakage through a steam generator to the Secondary Coolant System.
MASTER RELAY TEST
- 1. 15 A MASTER RELAY TEST shall be the energi:ation cf each master re'ay and verification of OPERABILITY of each relay. The MASTER RELAY TEST shall include a continuity check of each associated slave relay.
MEMBER (S) 0F THE PUBLIC 1.16 MEMBER (S) 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-l ational, occupational, or other purposes not associated with the plant. . OFFSITE DOSE CALCULATION MANUAL 1.17 The OFFSITE DOSE CALCULATION MANUAL'(ODCM) shall contain the methodology and parameters used in the calculation of offsite doses due to radioactive gaseous and liquid effluents and in the calculation of gaseous and liquic effluent monitoring Alarm / Trip Setpoints, and in the conduct of the Environ-mental Radiological Monitoring Prngram. W-STS 1-3
- - - , ._ _% o_ -m--,,
CE:INITIONS
/ .
OPERABLE - OPERABILITY 1.18 A system, subsystem, train, component or cevice shall be OPERAcLE or have OPERABILITY when it is capable of performing 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 ccpable of performing their related support function (s). OPERATIONAL MODE - MODE
- 1. 19 An OPERATIONAL MODE (i.e., MODE) shall correspond to any one inclusive combination of core reactivity condition, power level, and average reactor coolant temperature specified in Table 1.'R.
l PHYSICS TESTS 1.20 PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation: (1) cescribed in Chapter (14.0) of the FSAR, (2) authorized under :ne provisions of 10 CFR 50.59, or (3) otherwise approved by the Commission. PRESSURE BOUNDARY LEAKAGE 1.21 PRESSURE BOUNDARY LEAKAGE shall be leakage (except steam generator tube leakage) through a nonisolabid fault in a Reactor Coolant System ccmponent body, pipe wall, or vessel wall. PROCESS COMTROL PROGRAM See L3ert _c he PROCESS CONTROL PROGRAM (PCP) shall contain the provisions to re I that the 'FICATION of wet radioactive wastes results in a wae* .orm with properties that .
- the requirements of 10 CFR Part 61 and , ow-l eve l radioactive waste dis I sites. The PCP shall icen+# j process parameters influencing SOLIDIFICATIO 5 as pH, oil cent , d p0 content, solics con-tent, ratio of SOLIDIFICATION a +
wa + nd/or necessary additives for each type of anticipated waste, and +
- table boundary conditions for the process parameters shall be i ied for ea . = te type, based on labora-tory scale and full-sca sting or experience. : P shall also include an identification onditions that must be satisfied, on full-scale testing, to re that dewatering of bead resins, powdered r , and filter slud 1 result in volumes of free water, at the time of disco '
within imits of 10 CFR Part 61 and of low-level radioactive waste disposa s. PURGE - PURGING 1.23 PURGE or PURGING shall be any controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other cperating condition, in such a manner that replacement air or gas is required to purify the confinement. W-STS 1-4
i Tnscyt I Y
) PROCESS CONTROL PROGRAM lksaain coveeunets s a nesia swemmy e
, n m7e .ydt, iu+.ueJ sodoimieriou sys6, W l.2.2. The PROCESS CONTROL PROGRAM'shall contain the orovisions to assure that the SOLIDIFICATION of wet radioactive wastesfresults in a waste form with properties that meet the requirements of 10 CFR Part 61 and of low-level radio-active waste disposal sites 3 The PCP shall identify process parameters influencing SOLI)IFICATION such as pH, oil content, H 2 O content, solids content, ratio of solidification agent to waste and/or necessary additives for each type of anticipated waste, and the acceptable boundary conditions for the process parameters shall be identified for each waste type, based on laboratory scale a4h* -ase. full scale testing or experience. The PCP shell elsv inciede an id nti' '
4 4;ti0r Of canditicas that :::t b; :etisfied, bered c,a full scele test.ing, te -
- ure tht de-ai.ering-cf- berd *= cine , pc.;d: red re:in:, and filter :hdges will -
emen1t in vole ==s M f=ee weter, :t the time cf di:pssal, ithin the li .its sf 10 CF" Part 51 ead of icrl: vel redirective we:te di:p;;:1 :ites. Vor a mobile Sox tosfscaitow av per,guse.nt.ssnrsons serva< h be irsed ln l'h-04 ut, sba.hbn p<rman ently 'NShtlle). COL.lDiplcn7tod sysl err's, the Selet!ed vtsbOV avnush have an. accepfabis "T'opscal PCP hv wkuch he un de*wb' k-tkt pevtu ne& condshons of IOCPR Part- 4l ad -k Idas+c. Dssr os al Sih:S ve net. f a kJh k Yhc 664.bPodb 6hfbn SQs Waffe, ) 1 i l i
T E::NITIONS ( --- CUADRANT COWER TILT Pt.TIO 1.24 QUADRANT POWER TILT RATIO sr. ail be the ratio of the maximum ec er exc:re cetector calibrated output tc the average of the upper excore cete .:r cali-b ated outputs, or tne ratio of the maximum lower excore detector cali: rated cutpet to the average of the lower ercore detecto* calibrate: outpats, whicns er is greater. With one execre detector inoperable, the remaining three detectors snail be usec for ccm;:uting tne average. RATED THEPMAL PCWER 1.25 RATED THERMAL POWER shall te a total reactor core heat transfer rate tt, the rea: o- coolant cf g 4t. g iEACTOR t TRIP SYSTEM PESPONSE TIME
'. 2 E The REACTOR TRIP 3YITE.M RESFONSE TIME sha'l be the time ir. tern'. from
-9er ne renit: red parameter exceecs its Tri: Set:cint at -he enan ei senser
.." . .il ioss of s tationary gripoer coil voltage.
EVEA/T-T E00RTAELE --:~T" O:: Ev6NT g '
. . ,. r;- g...:
m.
. _ . . ... - _.~... c.c --
.m.
, . n., < .. ....u.
. . . . . . . . . . . . = . .a e, m
ir r 4:ni; ~ C. O. ~.. .; n 5. ^.1.1'.-- Section Soa3 -h, lo cFR Part 50 Coumnunna/r EucLDSU A.E
" " T._ : :'.': .:::-: *!TEGRI- -
, , courm. a_n, :1_...
su.r. . .a. .c.w,a.
._..: . m r.s.u c,2 A. .s.. .. .,i . axi s has ,.1
- e - . . ,
n----' wren: b c. Esch ccor in each access opening is c!:ced ey:ect when Pe acces: coeniv; g is being used for aorcal tra sit enir. an: exit, the1 c. itas. : e :cor snail be closec, cashnimenemt Eulosure.
- b. Tne T- c"'u' ; Filtration Systec is OPERAELE, and
- c. The sealing Techani:m ass 0:iatec with ea:n cenetration (a.g. , we %s, bellows, er 0-rings) is CPERA3cE.
SHUTCGWN MARGIN o.
. 25 SHUTDOWN MARGIN shall be the instantaneeus amcunt of recctivity Oy whi;n the reacter is s.;b:ritical or wouic ce surcritical from its. p-esent concitior.
assuming all full-lengts r:d cluste* assemblies (shutdown and centrol', are ' fully *nserted except for tne single rod cluster assembly of highest reactivity worin wnien is assumed to be fully withdrawn. SITi 90LNDAW 1.30 The SITE BCUNOARY shall be that line beyond which the land is neitner ea. dst % *e owned, ncr leased, site. boundary usel nor etnerwise contrelied 9p recra.ational purposesbybylicersee. Nertscas g,ghw.f48Ltc sp E skalt be h consi4<.ed. +e 1,e beyend ti< site boundary Gr papeses A me<Hn9 9:3e m sose. specWcah6M. (ReakId. occ.upany .Cacier$ ska tt he suppin'ed at w loca+ tint W-STS f,L. peposed of dase calcula-hov1-5 l 1
CEFINITIONS
}
SLAVE RELAY TEST 1.31 A SLAVE RELAY TEST shall be the energization of each slave relay and verification of OPERABILITY of each relay. The SLAVE RELAY TEST shall include- _ a continuity check, as a minimum, of associated testable actuation devices. SOLIDIFICATION . 1.32 SOLIDIFICATION shall be the immobili ation of wet radioactive wastes such as evaporator bottoms, spent resins'+ ludges, and r:::r:: :::::i: con-
- @ wwwAwe4es-as a result of a process of thoroughly mixing the waste type with a SOLIDIFICATION agent (s) to foam a free-standing monolith with chemical and physical characteristics specified in the PROCESS CONTROL PROGRAM.
SCLtRCE CHECK
~
1.33 A SOURCE CHECK shall b'e the qu'alitative assessment of channel response when te channel sensor is exposed to a source of increased radicactivity. STAGGERED TEST BASIS 1.34 A STAGGERED TEST BASIS shall consist of: ,
- a. AtesEschedulefornsystems, subsystems, trains,orotherdesignated
) , components obtained by dividing the specified test interval into n equal subintervals, and
. b. The testing of one system, subsystem, train, or other designated component
- at the beginning of each subinterval.
T_9ERMAL POWER 1.35 THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant. TRIP ACTUATING DEVICE OPERATIONAL TEST 1.36 A TRIP ACTUATING DEVICE OPERATIONAL TEST shall consis't of operating the Trip Actuating Device and verifying OPERABILITY of alarm, interlock and/or trip functions. The TRIP ACTUATING DEVICE OPERATIONAL TEST shall include adjustment, as necessary, of the Trip Actuating Device such that it actuates
. at the required Setpoint within the required accuracy.
UNIDENTIFIED LEAKAGE 1.37 UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED LEAKAGE cr CONTROLLED LEAKAGE.
)
\ W-STS 1-6 l'
f DE:!NITIONS UNRESTRICTED AREA gee I g e,+ E g a NRESTRICTED AREA shall be any area at or beyond the SITE BOUNDAPv access to w trolled by the licensee for pur . ,, o ction of individuals from exposure to *
. materials, or any area within the S residentia r for industrial,
, institutional, and/or recreational purposes.
VENTILATION EXHAUST TREATMENT SYSTEM l 1.39 A VENTILATION EXHAUST TREATMENT SYSTEM shall be any designed and installed l to reduce gaseous radiciodine or radioactive material in particulate form in effluents by passing ventilation or vent exhaust gases through charcoal adsorbers and/or HEPA filters for the purpose of removing iodines or particu-late's from the gaseous exhaust stream prior to the release to the environment. Engineered Safety Features (ESF) Atmospheric Cleanup Systems are net censicerec to be VENTILATION EXHAUST TREATMENT SYSTEM components. VENTING ! 1.40 VENTING shall be the controll.ed process of discharging air or gas from a Confinement to maintain temperature, cressure, hu-idity, c ': sntrat' r := cther , l ::erating c:ndition, in such a cancer that re;iate. ment air er scs is : : :- l viced or required during VENTING. Vent, used in system names, coes not imply a VENTING process. TREAT 11 EA/1" l WASTE C _" "C'Z? SYSTEM (Gaseous) s cssEQus R&DV4576 TREAThEUG 1.41 A "*"~" ^^" '^' " SYSTEM shall be any system designed and installed to pli reduce radioactive gaseous effluents by collecting Reacto- Coolant Systems offgases from the Reactor Coolant System and providing for celay or nelcue fi l for the purpose of reducing the total radioactivity prior to release :: the 9 environment. l [ WA ST'E TRE ATHEAn SVSTEM (l.Ipid)
\H2 A L GLAiD RADu)ASTR TREATMENT SyS 75H ghall 9g
,g condtnatio,i 44 comportedy designed and insEallel to Fedacc
~
b raltoaene ma4-erAah in ft p t1 e Hit 4ritts by passtai . process e.Gf twe,d strea,m3 -throg any one. of several possidt evapora+ov 3 ddjo,. geot,,.e,.dje,3 prg .fo release. -fo -the eAv'trcm e,cb l W-STS 1-7
'.%.* .- e A, p_'. f. . - " _ >- . ,,,i
. L.' . . -
"f.'.>'e, r,. * , L'-.f *
',Q q ;7 g ,. ..; . .R, E . L. . , . g. . . *. ,
k..,- r\. ;c.~. e-r. ? e, ~ 5 ,\ .c. g, %. , .*;g , *.; _ ** . _; %. y. c ~ .., ., -
. -G y .. ,, ~ #s , z
. sq , . 3, ,
- g. ..; . ; ,, ,. * ,
~,r-
.su * *i .,e ,
n.. . . o
.+ ,,,..
.. . s.~S ~
. n...
.. .~ ,
. .~, .
.4.,..
.g *
,s
- w n. .- . a .. ... .,, ,. g..
* .* . ., . . .,. ..r<.
~g.
. .,.. + , .
]
~....3.- . . 1
, n .~. . - . . .
.- ~ ,, ..
- ,x._... :_ n.
,m__. .
.. _.. m-..... ,m...
_. rnse w_ x . _. . pF *" # [4 ,.s g/8h = ca b*8 .T,,].i.,4.w'9 .. 88 / ' * * *, It . ** h ** . ," # 9
- l m. _.. ~ . .%.
, . . . . c..h. ;..* . .v. >..-se , . w . '.~.' ... ..-;.. .. 4.s.- s .. s. ~. ,_s, z, , ..
, . . Q ,' . m %.. j * .-- . -. ~ ,
~ % . ~ ,
. - .. e
...42dk Resrasc.renl'". l.nsn (w.IQ. uWeFnnds sly . ..
r ~. % , . . n .4 e. #
,. 1.% i..t
. m_.m...i.
.2
- m. .
, .~.,_
, . , . - . , . . .y. . . = . , .- .. ..
a . m_
' , ; -m._.,.
y
. = . . .
c ,g (,..o3 .
.-, Bu'r
,a . # . '
=w. .:r.3 2. - se%nmucreo. , .
._ . - . 3 ,. . . .
A"* w" r.s>9un . ews> -rw 'th e
, m .
- . ;t 3ke ca%IaQskille'de%iL&se bepnd he wface
- 2. 2. e%e AA: mikalaadv [ eneL wk eHlue4> % & . . . . .
-. - L sdmeqM.hmulhjortJi%e &&ey. han unde y e.
~
. . . & +.. . .A.% (ass ud. k he .ma. -
w ,
* - ( ;.
.o..
, . . ,. _ . . ~ . . . .
> ...-. .~.
. . . . . . . . . -- .. .. -~.m,.~.-.' . . .
., . . . . ....a - . .
.n...._
- 7. . . . . ..I.... . . . _ . , . . .. . . ..
. +-
. .- . . . . . . . , .. ..-.r- .s . ,..,.. ,,,da . . . .
~ s i
s
.,e.
.:. .- .. . r T
^ ' , . . .
- - . . - , . , , , . . . . ,., ..,n.y. .-- , . . . . f.,
..q
- '1
..m... m, . . ,-
- e i,a , . . .i - <
- s
* '= f. * -, ,
s . ,. .
....n . .. .
m a. .-.ew ., .. . , ....
.. .;. . . ..,.,.2.. s, .,..vn. n,. b.y .x . , s a p_. _.n
. .s... ._-
r-. -
*-=*
_v. * *. ,
* .< . : , r b-",, <
** .e--. . , n. ,',Mr .g d s.
,~l' , > , f .' . Y ' f .
- .- ,;... . . . -n . . . . ,.-
4 .4 y
. .. .~ w . ,
- t o .
,.w, k
0 _ ,. ..._s_ . . . _ ... . ~ . .
.m ,
a f.-... U ,. '
* .~. ,
4 F i e = .
._,. . . ' .w
- -. .L.
. - . .i.,,. ....... ..
~.
?... . .a. ? Q . . , : Q. . .. 7 . . . . . . . .
g , g . y o b-..
$ . Q
. e .s.e. hp%e.. . e.w
- e .gde-., .,.m.e .- 'D-q e. ,9.' p %e . ,h..-ww .e e-.m ep -e .
.g.m.-
*, "
- a .
. n %
. . . . . ~ . . . . . ~ . . , . .. . . - . - . . - . ~ 2.,.L :,4
._ . - ..._ ...Sa, n ...'.:... ;; -a... -~ . .. , -
~ ,
f' . ..
- n. , ..
ei 9
...A- .e.y . .e . , *. -.e.-e .*= < . gm--b EM- 'e-e..
s . .e. se ,. - . , A y
& a F . .. . . - . . . . . . . J . d(a .. ( . 4 - .., ,,..i... .. .- . . ,.
r
-A l
. . . ~ . . . , . . - .. .. , . . :: ,. . . " .'n n ; %M . . . . , i n . :. . ; . .,..u . .. . .
, "* y. .
p & s-
. 4-
.. , p[#'r (h
, p 4- g , [ ** .c- + e , 4
,..'[.*% *8 ' * * .# .
...-7,,,7...q,....,,
.w.'
q .u . .s.ey.,. . y. : f y . f . _ s .,.e . . . c ac,. . h. . .: " , , z .. .. . . .,. ,. , . . . I %> . ' *
- p , s g
.." - .'g? "* g*) .
e , _
+
g . b pe , N~
\* , . '. (,
..c.,m' , .
W
'N e . .g 6 . -
-.ee
. . +
.=w.-.ew.,'==.=r.a.Mw, g
.4
,.y.p.,n'%,.J.s....w.*.
*-.w . . . ,i... . - . - .
a e ?
- y,.
.,A
.
- i . ,
# .., *9* T b* ,". F g
' * *g i 4 f' F s'T '*4
. .p- ',.o,,5, L.-.v.-,..f L, M' #4..'* a. N ...y .1;.- ...a. .. . .
- ',ve,.*4[zl,'...
g'y. -e t( w* qS*}8 "v 4 '-'7 jy,' ' '
.' #' ~
. , )' .O . %.-)-
s.
, '-[ ~,j} ' ".' j.N*t' .*. .* ' , .
. .g ,
.t,e.'-'
'., . - a . ? ;; ., . . .y. .['~,,. p
% s g
A % s, ,
. -.= 6-n..
.q.._ . .
4e .. . . .L aw....
- ,. . , . . 9p, wp - + e.eae e = % -
- p. .e
. , .. ..e., ,)e .
,p...eg .
g
. . . , . . ,..,.,,.as.,.4
_ ~g .J ' .
.""'%.i.... w...,...
',g .'
...g.. . .
]
..t.,4.
- .~ .'
...".....m
+ s . ,
e
.,sg ? . , - ,
-.3..,.n.,
,, e . p
. T* ,,,
y,
.,r .$ . . . - -
+..7,. ,
._. , , . ~
lq, ,l , * *
. f . ,,. ^b- . b. d .,~ '
. * . 2,o .,, - .
# y n .,, .
-y , s . :, . ,, , .
;s , >.e. p ;-p*. .*. s-s .,.. ,x^em (,. w ,.n., p. m. , ,,.,., ;e.s- ..r.a.~m.-.--....s., .
- ~ . - . ~ ~ ~ . . . . . . - . . . . . - - .. . . - -
7* '
%- Q*.{
.(
g
..+ '. .- a;a s'. r x
,w ,* i*
g s, *. '(e, g. a , , - p i :.- - *
- f j-
- 4.I , Wt 7 f,".s -w.- 3/, - .. e
*.a,,,-..., * . -.c '
- S O N. , . )=ae. .. , ,{.'.y., * ; .4e > , ?.v...*e .J ?" .m. g ,E p3*L esE.'w.. Wv +, - d ..*= h N, .y.ge- ***' *..,.g* .r.,. ,-w .>+ ..
,* c .f. .,;- .
'- y.
b.'*.4 a=A .- ' - ....cw. n-
/., %'.,E.t-e u e. <
*eg'-
# e qg ham,* _v-) * ' f g s. m<g p g . g g ._-Q* ,o t i , - % S g.m
o.
..3
".% "g .y' g
(;j . ,,,7, . =.1. w.y+ .".e ,-.O f. : ,.. c (a'-- y' ' ' p c y ;4'* .,
=-'
. . ,f.:
* ; y,. .y. . . . -o. . 6. 3, . ' -
. ".s u . k.
, O { 4 44,
-f .e)- (g. . -a. ' , .* ,g .* -we y ..>.s -%.*m..gg' .'. N
- g
.e t , *. .
*
- L ./.7a*v wm h,,, is'y. @ !- *A ** A. . ..~,... ,Qa3* 4 .n e* .%3 4*
MatA.n :a. ,agg.W!-'..,**
, 4 f..
ua.4%,.w n--g,M
.. . , .,.1 4-m 4 M'Mg
. .F ~.'.
[
TABLE 1.1 f OPERATIONAL MODES REACTIVITi % RATED AVERAGE COOLANT MODE CONDITION, K THERMAL POWER
- TEMPERATURE
;ff
- 1. POWER OPERATION 10.99 > 5% 2(350*F)
- 2. STARTUP 10.99 { 5% 1 (350*F)
- 3. HOT STAND 8Y < 0.99 0 1 (350*F)
- 4. HOT SHUTDOWN < 0.99 0 (350"F)
> (200 F)> T"V9
- 5. COLD SHUTDOWN ( 0.99 0 { (200*F)
- 6. REFUELING ** { 0.95 0
{ (140*F)
" Excluding decay heat.
** Fuel in the reactor vessel with the vessel head closure bolts less than fully tensioned or with the head removed.
l
,, TABLE 1.2 ~
f FREQUENCY NOTATION ~
, NOTATION , FREQUENCY S At least once per 12 hours.
D At least once per 24 hours. l , W At least once per 7 days. l M At least once per 31 days. Q At lea'st once per 92 days. SA At least once per 184 days. R At least once per 18 months. S/U Prior to each reactor startup. P* Completed prior to each release. (. N.A. Not applicable.
" Applies only to RETS, not STS PWR-STS 1-)% 8/7/80
(
JUSTIFICATIONS Section 2.0 and Bases - In the text of Section 2.0 and Bases, where a capital letter with a circle around it appears, please refer to this. sheet for the appropriate justification. A. N-1 loop operation is not allowed at Seabr,ook'. B. NRC change allowed for Calloway Plant. C. Seabrook Station specific plant data. D. Westinghouse terminology. E. Not applicable to Seabrook. F. Provides clarification to aid in understanding o'f Reactor Trip Setpoint limits of Section 2.2.1.
.~
r e b
, , , - - - - , - , - - , - , , . . - .- - - - - - . - - , . -, - - , - < - - , - , ,, , - - . > - - - a
~~.
( 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS ] REACTOR CORE 2.1.1 The combination of THERMAL POWER, pressurizer pressure, and the highest operating loop coolant temperature (T 8 g FigureX 2.1-1 trd -1 U :);shall not exceed the limits shown Y /eafin. ::; : tire %fo 7 : d 2.1 2 fr
;r etier operah,an APPLICA8ILITY: MODES 1 and 2. ,
ACTION: Whenever the point defined by the combination of the highest operating loop average temperature and THERMAL POWER has exceeded the appropriate pressurizer pressure line, be in HOT STAN08Y within 1 hour, and comply with the require-ments of Specification 6.7.1. REACTOR COOLANT SYSTEM PRESSURE
.'. 2.1.2 The Reactor Coolant System pressure shall not exceed 2735 psig.
I APPLICABILITY: MODES 1, 2, 3, 4, and 5. ACTION: MODES 1 and 2 Whenever the Reactor Coolant System pressure has exceeded'2735 psig, be in HOT STANDBY with.the Reactor Coolant System pressure within its limit within 1 hour, and comply with the requirements of Specification 6.7.1. MODES 3, 4 and 5 Whenever the Reactor Coolant System pressure has exceeded 2735 psig, reduce the Reactor Coolant System pressure to within its limit within 5 minutes, and comply with the requirements of Specification 6.7.1. , W-STS 2-1 SEP 2 81981
.,-e - ,,,: ,,. ,+.--r e e~ ~ -, -, n.-sn-.-. er , ,--.-_--,v,- . . - - , . -- , , - , ,----~,,.e .,,e,--- ,, , ,--,, m,--,--, -- . --, -e,,-
680 . .k - : OPERATION l UNACCEPTABLE 660 -- - ----- 400psk i
. i 640
' 2250 P SIA , ,
l r
~
i l l
. r 620 2000 PSIA <
S60PSg4 l i LL. d I g 600 '
+
th ! u - i cx j l 580
-- ACCEPTABLE OPERATION
~
i 560 !
~
l i
~
~
i i 540 ,
. I ; i l
520 O.0 0.20 OA0 0.60 080 1.00 1.20 ( FRACTION OF RATED THERMAL POWER) i' FIGURE 2.1-1 REACTOR COOLANT SAFETY LIMIT FOUR LOOPS IN OPERATION 2-2
s' S*FETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS
- 2. 2 LIMITING SAFETY SYSTEM SETTING'S ,
REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS
- 2.2.1 The Reactor Trip System Instrumentation and Interlocks Setpoints shall be set consistent with the Trip Setpoint values shown in Table 2.2-1.
APPLICABILITY: As shown for each channel in Table 3.3-1. ACTION: ! a. With a Reactor Trip System Instrumentation or Interlock Setpoint . less conservative than the value shown in the Trip Setpoint column of Table 2.2-1, adjust the Setpoint consistent with the Trip Setpoint vrlue,
- b. With the Reactor Trip System Instrumentation or Interlock Setpoint less conservative than the value shown in the Allowable Values
' column of Taole 2.2-1, n' a-a +ae c~me' " t": tri;; d ::mditier h ithin 1 '.:r , rc within the f: lining 12 hours eitner:
- 1. Determine that Equation 2.2-1 was satisfied for.tne affected channel and adjust the Setpcint consister,t with the Tri: i Setpoint value of Table 2.2-1, or
- 2. Declare the channel inocerable and a: ply the a;:licable ACTION statement requirement of Specification 3.3.1 until the channel is restored to OPERAELE status with its Setooint adjustec 3 consistent with the Trip Setpoint value.
Equation 2.2-1 , Z + R + 5 < TA Where: I Z= the value for Column Z of Table 2.2-1 for the affected channel, R= the "as measured" value (in percent span) of rack error for the affected channel, S= either the "as measured" value (in percent span) of the. sensor error, or the value for Column S (Sensor Error) of Table 2.2-1 for the affected channel, and
. m % ef 5 Pan).
TA = the value for Column TA (Total A11'owancey of Table 2.2-1 for ( the affected channel.
~
3 W-STS 2-\
~
M I I AllLE 2. 2- 1 J, REACTOR TRIP SYSTEH INSTRilMINIATION HilP SEIPOINf5 Y
~ SENSOR T0iAl ERROR FUNCTIONAL UNIT All0WANCE (IA) Z
{S) _ TRIP SETPOTNT ALLOWABLE VAllH;
- 1. Manual Reactor Trip .N.A. N.A. N.A. N.A. N.A.
- 2. Power Range, Neutron Flux
- a. liigh Setpoint (7.5) (4.56) 0 $(109)% of RTP $(111.2)% of RTP
- b. Low Setpoint (8.3) (4.5G) 0 1(25)% of RTP 1(27.2)% of RfP h
- 3. Power Range, Neutron Flux, (0.5) 0 liigh Positive Rate 1(5)% of RTP with 1(6.d)% of RTP with a time constant a time constant 1(2) seconds 1(2) seconds 16 3
- 4. Power Range, Neutron Flux, (F-0.) (0.5) 0 5(5)% of RTP with liigh Negative Rate $(6.li)% of RIP with a time constant a time constant y 1(2) seconds Vf .c 1(2) seconds
- 5. Intermediate Range, (11.0) (8.41) 0 $(25)% of RTP Neutron Flux 1(31)% of RIP
- 6. Source Range, Neutron Flux (17.0) 5 (10.01) 0 4 1(10 ) cps $(1.4 x 105) cps l S. g 2.*I3 l 3,t ab
- 7. Overtemperature Al M (+-Pt) (M) See Note 1 See Note 2
' 12.
- 8. Overpower AT (4.3) (1.3) (M) See Note 3 See Note 4 3.3 ff45 14 3V
- 9. Pressurizer Pressure-Low (6-83 (O./l) (1.5) >(4944) psig 1(4886) psig
- 10. Pressurizer Pressure-liigh (3.1) (0.71) (1.5) 1(2385) psig 1(2396) psig 4
- 11. Pre .surizer Water level-liigh (5.0) (2. Ill) (1.5) $(92)% of instrument 1(93.8)% of instrument span span
- 12. Low Reactor Coolant flow (2.5) (1.0) (l.S) g(90)% of loop 1(89.2)% of loop i
design flow
- design flow
- aloop design flow = (95,700) gpm 4r Sec 4/o/c 5-RIP = RATED TilERMAL POWER l
t
; if -
m TA010 2.2-1 (t:ontinae( N )% of narrow Level Low-Low if 0 / a. s if range instrument range instrument span span M. Si:c.a/imeda ter T1:u (4&re) (12. 2^-) (4r&) f('0)% af foi1- 1(12. 5)% ;i ' "
-M!:r:tch Ce!nc!& nt "!!h ste:: '!:a et 'li? ster: fica et "!P
-Ste:r Cen:r::ter h' ter (42-et (4-18) (4-6-) '(92. 3)% of narre.; <
level-Lua Lua 72=;; in:t-- ent ?(20.f,% a of ..;crea
- n;;: nstr nt
'Pd*- 'Pa#-
later lakr 70 7, e f N,=a=1 la+er 19%. Undervoltage - Reactor (-2-0) (4-e8) lake ->(1920) ;;!t:- -(4?E9)
'? Coolant Pumps Sus Velfee N .
lakr 6ter lafer 554 lake 15"N. Underfrequency - Reactor (J,&) -e- (e-1) Z W ) Hz Coolant Pumps 1M Hz l&N. Idrbine Trip
. lakr
- a. Low Fluid Oil Pressure N.A. N.A. N.A. 1(800) psig 1(-750-) psig
- b. Turbine Stop Valve later H.A. N.A. N.A. 1(1)% open 1(4-)4 open
! Closure
, I'l16. Safety Injection input' N.A. N.A. N.A. N.A. H.A. from ESF l ITM. General Warning Alarm N.A. N.A. N.A. N.A. N.A. l RIP = RAT [6 TilERMAL POWER t 1
N 1 IF c m gj Q TAllLE 2.2-1 (Continumi) REACTOR 1 RIP SYST[H INSTRUHl_ NI,ATION _lRIP SETPOINTS SENSOR total Erit 0R - FUNCTIONAL UNIT ALLOWANCE (TA) Z (S) TRIP SETPOINT ALLOWABLE VALUE 1920. Reactor Trip System f Interlocks
- a. Intermediate Range Neutron Flux, P-6 N.A. N.A. N.A. 1(1 x 10 80) amps 1(6 x 10 ) ainps
- b. Low Power Reactor Trips Block, P-7 later
- 1) P-10 input N.A. N.A. N.A. $(10)% of RTP $(4Gre.)% of RTP
~
later
- 2) P-13 input N.A. N.A. N.A. $(10)% RTP Turbine f(43re)% RTP Iurbine n3 Impulse Pressure Impulse Pressure .
j- Equivalent Equivalent 49
- c. Power Range Neutron N.A. H.A. N.A. $(48)% of RTP $(60,3-)% of RIP ,
g Flux, P-8 i i 20 later i
- d. Power Range Neutron N.A. N.A. N.A. -< 0be-)% o f RT P -<(-43 4-)% o f R I P Flux, P-9
- e. Power Range Neutron later N. A. N.A. H.A >(10)% of RTP >64.&)% of RIP
, Flux, P-10 h Ecr
- f. Turbine Impulse Chamber N.A. N.A. N.A. <(10)% RTP Tubrine 1610-e)% RIP Turbine Pressure, P-13 Impulse Pressure Impulse Pressure Equivalent Equivalent l 2,024. Reactor Trip Breakers N.A. N.A. H.A N.A. H.A. .
L I 2.1 12. Automatic Trip and Interlock N.A. N.A. N.A. N.A. N.A. l Logic rip = RAII0 TilERHAL POWER
C N s c s e c s e s
)
, ; 3 )
. s 3 0t1 c c ( (
] e e
) s s 2.s =
l a ) , s ( 3 8 ( o1)L ( t i f k = g r
- g y
y o 3 a t
) r T , T a 1 s
'P , , r n T T r o e
- A A o f p f m P r r r o
( o o r e c 3 f f e l g K l l r r l o a
+ e o o r l
; l t r t g
] n l a t n o o s n o ,
'T i t t r n e
o c c T
- a n p g t o m g a d
) n c o a l e r
5 e ; g c ; l d u i R -
}
4 m i r T A l a g a E W d a a e I s a d t d - l O e l g e e s e d ; P l v m u S l n r a T e e a n N I I u e A h L e h T e i 0 s l t A h t h t l I d a d H t d t n i I l e n e n R y n e r n o A o m i r i E i u C T f u i l b ; s i ( O i n d s d T . d e f a d N
- n o e a e d e; z e e 1 a z e z D m z
- E M r i m i E t n i e
2 l o l l T ao l r n i B ) I t i n i A ri i u o l 2 A 1 T a t o t R et na t i T R s u u es i t t a r t u E n r t l 2 y e s o s a gn s; r o _ B K b p t ; t t e t s e t t A m nc a n T np nc p a n 1 - T o ae s a A omo ae m e s n a _ A c t s n t i t s s e t s i s e s d i t c n) t p n
. K d g n) p n e ,) c
( e 'a o3 m o t r;2St3 nc o4 e m o r c( o c ui c( g o c
, u l
c a c 19 )9 s! f m c _ T A s a d a e m }~ g e m i d 09 o. 0 en a e1 m are g e m e e i2 a i nlL( o0- ( hy Td i3 A v a. i T
<_ M L TT L T I T 1 1
)
3 = = = = = = = = = = = = = 3 5 g - T 1 S. 3 s A T 2 t 1 i 3 I E ( , * + , , R { , 1 l T i 3 T i 2 4 s i r A .K K 1 1 T ) i T A y T 3 1 A
@ R E
P M : E l y e > T f r R e - E T h V D A W , _ ~ _ - 1 _ E s 1 0 _ N J-f U i TW
~ i I ' ' I lj e
\
-l ly IAllt f 2.2-1 (Continued} vs g IABLE NOIATIONS (Continued) NOTE 1: (Continued) T'
< (588.5)"f (Nominal Tavo
- K3 = (. 'Cb P = Pressurizer pressure, psig; P' = 2235 psig (Nominal RCS operating pressure); i S = laplace transform operator, sec 8; ;
and f (AI) is a function of the indicated difference between top and bottom detectors of the power-range nuclear ion chambers; with gains to be selected based on measured instrument response during plant STARitlP tests such that: h (1) For qt Ob be ween -(3'a)% and + h, f (AI) = 0, where q and q are Percent RATED THERMAL i t b POWER in the top and bottom halves of the core respectively, and qt*9b is total THERMAL POWER in percent of RAllD IlllRMAL POWER; (ii) For each percent that the magnitiale of q q, exceeds -(35)%, the AT Trip Setpoint shall i - be automatically reduceil by ( )% of its value at RATED TilERMAL POWER; and l (iii) For each percent that the magnitude of qL y h exceeds *k)%,theATTripSetpointshall Io49
! be aittomatically reduced by (4,&)7. of its v ilue at RAILD TilERHAL POWER.
i NOTE 2: Ze
' The channel's maximum Trip Selpoint shall not exceed its computed Trip Setpoint by more th.in (3-a)%.
ef ATinsirramed syan. i I. - I . ..
s s c e g e s l
) a r )
I e 0 A v 1 ( a ( f 2 g c n i m d
- i a 7 s n 1 a y e d ,
r c
- e y d , T
- T r r S. o r o t f f
o f 0 r r e 1 d e l f n l l a l o T o r [ e r t r t n
. u n o K t o c a c r g e g a T p a l
)
d m l - e s e - e u t e t
) o t a d n t e a r e i g r u t a e n n r c i i o 1 e l i
i t t C v L n ( a i i
- o y i
( S g l i ( N 5 n l O i d e i 1 I t , , , , , , s e z , ,
- T a 2 A T
f 1 e 1 e 1 e 1 e 1 e 1 e r t a r i l 1 1 2 O N f [ t o t o t o t o t o l o c n n e i t u t e o t e o I N N H N H H i e N N I E 5 g s D I K n n n n n n r , t n n A B i i i i i i o , nn i r i i T A - f e oo a T d d d d d d r i i t d d
. e e e e e e f a
t t t s e e M n p n n n n " t ca n n n i i i i i i , / a ns o i i f f f f f f ) ) r un c f f e e e e e e 9 2 e f e e e - T d d d d d d 0 0 p p e l e d A s s s s s s 1 0 e m em h o i m s s 1 A A A A A A ( ( t 1 c T A A 3 = = = = = = = = = r- = =
,a f 3 3 3 2
3 T 5 f 7
. 1 , , [ ,
T s a T 4 3 7 e,
@. T A
R E
- W 3
A g I y T A K K y t 3 t e O y P r R e - E T h - V A W O - . 3 - E T - . O N nh* ~ - . . i ,
! i l
e s I 8T TART [ 2.2-1 (Continue @ us
,- d 1ARLE N01ATIONS (Continued)
NOTE 3: (Continued) . K. = (0.00128)/"F for T > T" and Ks = 0 for T 1 I", T = As defined in Note 1, T" = Indicated I ,g at RATED TilERHAL POWER (Calibration temperature for AT instrtmentation,1 (588.5)"F), S . = As defined in Note 1, and f 2(AI) = 0 for all AI. 1 rf NOTE 4: The channel's maxista Trip setpoint shall not exceed its computed Trip Setpoint by more than g5 (he)%. of sp n. 2.T NOTE C: 13. Es % sens*< u ror Co ' Tasy s sa U. S y i ts % S en,s, e,,,, p,,,,u p,h, Paessace. . These values av k "as ne"S"'*d" se*'so- e#-ov *" a*ey co bssa h on
& ws i b< Asbed to deter de S Coe Eg ua % a. 2-I, fo > ts,e e , J n .' e i disc *fcAperahve 4 T ch a n nel.
2 4 6 i i I
s. i BASES FOR SECTION 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS ( ( j . G G
I i ( NOTE The BASES. contained in succeeding pages summarize the reasons for the Specifications in Section 2.0, but in accordance with 10 CFR 50.36 are not part of these Technical Specifications. t I r AUG 7 1980 l
..e*
O
- , , - , , , _-_,--r--r+--e---=
-r-e-- ,--ene - - - w t--- r ,, , ,,4- m, yw-e -m -ww- w---ga m--ee--w i-------w-- -*-- ------ r----
1 2.1 SAFETY LIMITS BASES 2.1.1 REACTOR CORE The restrictions of this Safety Limit prevent overheating of the fuel and possible cladding perforation which would result in the release of fission products to the reactor coolant. Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature. Operation above the upper boundary of the nucleate boiling regime could result in excessive cladding temperatures because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient. DNB is not a directly measurable parameter during operation and @ therefore THERMAL F0WER and Reactor Coolant Temperature and Pressure have been related to DNB through the M-fcorrelation. The-W-#DNB correlation nas been 8 developed to predict the DNB flux and the location of DNB for axially uniform and non uniform heat flux distributions. The local DNB heat flux ratio, DNBR, defir.ed as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux, is indicative of the margin to DNB. The minimum value of the DNBR during steady state operation, normal operational transients, and anticipated transients is limited to 1.30. This value corresponds to a 95 percent probability at a 95 percent confidence level that DNB will not occur and is chosen as an appropriate margin to DNB for all operating conditions. The curves of Figurey, (2.1-1) W (2. '-2)- show the loci of points of THERMAL POWER, Reactor Coolant System pressure and average temperature for which the minimum DNBR is no less than 1.30, or the average enthalpy at the vessel exit is equal to the enthalpy of saturated liquid. These curves are based on an enthalpy hot channel factor, F N of 1.55 and areferencecosinewithapeagof1.55foraxialpowershape. Abg,llowanceis a included for an increase in Fg at reduced power based on the expression: Fh=1.55[l+0.2(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 allcwable control rod insertion assuming the axial power imbalance is within the limits of the . ft (delta I) function of the Overtemperature trip. When the axial power , imbalance is not within the tolerance, the axial power imbalance effect on the Overtemperature delta T trips will reduce the setpoints to provide protection consistent with core safety limits. . W-STS B 2-1 MAR 151979 -
SAFETY LIMITS BASES 2.1.2 REACTOR COOLANT SYSTEM PRESSURE The restriction of this Safety Limit protects the integrity of the Reactor Coolant System from overpressurization and thereby prevents the release of radionuclides contained in the reactor coolant from reaching the containment atmosphere. , Rea c,,6+ Ss.f e.= c.-ee =u+5mi a cl4 a. e, I ta9 eval'cSead Min @ f TheVreactor pressure vessel-and-pressurizer, are designed to Section III m of the ASME Code for Nuclear Pow,er Plant which permits a maximum transient W pressure of 110% (2735 psig) of design pressure. Th: Reecte.- C;;?;nt Sy;te pfpfng, ;;;.e. and fittin;:, rr: de;ign:d te 'tSI ; 3i. i Editica, r5fch ;;r;ft: e -ex'rur tr a:!:nt pr:::;r; cf 120% (2^05 poig) er cumpuneni d::f;n pre;;ur:. The Safety Limit of 2735 psig is therefore consistent with the design criteria and associated code requirements. The entire Reactor Coolant System is hydrotested at 3107 psig, 125% of design pressure, to demonstrate integrity prior to initial operation. 4 l l l
?-STS B 2-2 MAR 151979 l 1
._1
2.2 LIMITING SAFETY SYSTEM SETTINGS {} Bases 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS The Reactor Trip setpoint limits specified in Table 2.2-1 are the nominal values at which the reactor trips are set for each parameter. The setpoints have been selected to ensure that the reactor core and reactor coolant system are prevented from exceeding their safety limits during normal operation and design basis anticipated operational occurences, and to assist the Engineered Safety Features Actuation System in mitigating the consequences of accidents. The methodology to derive the trip setpoints is based upon statistically combining all of the uncertainties in the channels. Inherent to the deter-aination of the trip setpoints is the determination of the magnitudes of the channel uncertainties. Sensors and other instrumentation utilized in these channels are expected to be capable of operating within the allowances of these uncertainty magnitudes. To accommo'date the instrument drift assumed to occur between operational tests and the accuracy to which setpoints can be measured and calibrated, i Allowable Values for the reactor trip setpoints have been specified in Table 2.2-1. Operation with setpoints less conservative than the Trip Setpoint but within the Allowable Value is acceptable since an allowance has been made in-the selection of that setpoint to accommodate this error without exceeding the value used in the safety analysis. ' ~ Rack drift which results in the parameter exceeding it's Allowable Value indicates that the rack has not met its allowance. However, the channel ,may B 2-3 I
. - . - . . - . . _=__ _ ..
J be determined to be OPERABLE, as long as equation 2.2-1 is satisfied. The methodology of this determination utilizes the "as measured" deviation from f the specified calibration point for rack and sensor components in conjunction with a statistical combination of the other uncertainties of the instrument-tation used to measure the process variable and the uncertainties in 4 calibrating the instrumentation. It allows increased flexibility in those cases where the difference between the selected trip setpoint and the value assumed in the safety analysis is significantly greater than the statistical summation of uncertainties. In Equation 2.2-1, Z + R + S < TA, the interactive effects of the errors in the rack and the sensor, and the "as measured" values of the errors are considered. Z, as specified in Table 2.2-1, in percent span, is the statistical summation of errors assumed in the analysis excluding those associated with the sensor and rack drift and the accuracy of their measurement. TA or Total Allowance is the difference, in percent span, between the Trip Setpoint and the value used in the analysis for reactor trip. R or Rack Error is the "as measured" deviation, in percent span, for the affected channel from the specified Trip Setpoint. S or Sensor Error is either the "as measured" deviation of the sensor from its calibration point or the value specified in Table 2.2-1, in percent span, from the analysis
- assumptions.
Satisfying equation 2.2-1 indicates that even though the rack may have drifted.nore than was originally allowed, either the sensor drifted in an opposite direction to compensate for the rack drift (if sensor drift was measured) or sufficient margin existed between the selected trip setpoint and the values used in the safety analysis to accommodate the excess rack drift. B 2-4
The value of the use of Equation 2.2-1 is that it allows for the use of a sensor drift factor, it allows an increased rack drift factor, and it provides a threshold value for REPORTABLE EVENT which may be larger than the difference between the Trip Setpoint and the Allowable Value. The setpoint for a reactor trip or interlock function is considered to be adjusted consistent with the nominal value when the "as measured" setpoint is within the band allowed for calibration accuracy. For example, if a bistable has a trip setpoint of f,100%, has a span of 125%, and has a calibration accuracy of + 0.50% of span, then the bistable is considered to be adjusted to the trip setpoint as long as the "as measured" value for the bistable is 1 1100.62%. The various Reactor trip circuits automatically open the Reactor trip breakers whenever a condition monitored by the Reactor Trip System reaches a preset or calculated level. In addition to redundant channels and trains, the design approach provides a Reactor Trip System which monitors numerous system variables, therefore providing Trip Systes functional diversity. The functional capability at the specified trip setting is required for those I anticipatory or diverse Reactor trips for which no direct credit was assumed in the accident analysis to enhance the overall reliability of the Reactor Trip System. The Reactor Trip System initiates a Turbine trip signal whenever Reactor trip is initiated. This prevents the reactivity insertion that would otherwise result from excessive Reactor Coolant System cooldown and thus l avoids unnecessary actuation of the Engineered Safety Features Actuation System. I i B 2 .
--,--,,-.------.-m.
LIM?TfNG SAFETY SYSTEM SETTINGS t SASES T M CIDR TRIP SYSTEM INSTRUMENTATION SETPOINTS (Continued) The var Reactor trip circuits automatically open the Reac+a $ rip breakers wheneve.- ndition monitored by the Reactor Tri em reaches a preset or calculated 1 . In addition to redunda
- nnels and trains, the oesign approach provides a tor Trip Syste - ch monitors numerous system variables, therefore providing s unctional diversity. The functional capability at the specified trip 's required for those anticipatory or diverse Reactor trips for wh no direct c 4 was assumed in the accident analysis to enhance the rall reliability of the or Trip System. The Reactor Trip Syste itiates a Turbine trip signal whene eactor trip is initiated. ' prevents the reactivity insertion that would o .. . #se result from ex ive Reactor Coolant System cooldown and thus avoids unnecess .
a- acion of the Engineered Safety Features Ac uation System. . Manual Reactor Trio The Reactor Trip System includes manual Reactor trio cacability. Power Rance, Neutron Flux In eacn of the Power Range Neutron Flux c annels there are twc incepencent bistables, eacn with its own trip setting used for a High and Low Range tric setting. Tne Low Setpoint trip.provices protection curing succritical and low power operations to mitigate the consequences of a power excursion beginning from low power, and the High Setpoint trip provices procection during pcwer operations to mitigate the consequences of a reactivity excursion frcm all pcwer levels. The Low Setpoint trip may be manually blocked above P-10 (a power level l of approximately 10% of RATED THERMAL POWER) and is automatically reinstated below the P-10 Setpoint. Power Range, Neutron Flux, High Rates The Power Range Positive Rate trip provides protection against rapic flux increases which are characteristic of a rupture of a control rod drive housing. Specifically, this trip complements the Power Radge Neutron Flux High and Low trips to ensure that the criteria are met for rod ejection from mid power. The Power Range Negative Rate trip provides protection for control rod drop accidents. At high power a single or multiple rod dros accident could cause local flux peaking which could cause an unconservative local DNSR to exist. The Power Range Negative Rate trip will prevent this from occurring by tripping the reactor. No credit is taken for operation of the Power Range Negative Rate trip for those control rod drop accidents for which DNBR's will be greater thang. i m Ldd valu e3 y-STS B 2- @
l l L?MITfNG SAFETv SYSTEM SETTINGS BASES Intermediate and Source Rance, Neutron Flux The Intermediate and Source Range, Neutron Flux trips provide core protection during reactor STARTUP to mitigate the consequences of an uncon-trolled rod cluster control assembly bank withdrawal from a suberitical condition. These trips provide redundant protection to the Low Setpoint trip of the Power Range, Neutron Flux channels. The Source Range channels wil'1 initiate a Reactor trip at about 105 counts per second unless manually blocked i when P-6 becomes active. The Intermediate Range channels will initiate a Reactur trip at a current level equivalent to ap' proximately 25% of RATED THERMAL POWER unless manually blocked when P-10 becomes active. Overte cerature aT The Overtemperature Delta T trip provides core protection to prevent DNS f:r all combinaticns of pressure, pc.er, coolant temperature, and axial power distribution, pr:vided that the transient is slow with respect to pioing transit delays from the core to the temperature detectors (about 4 seconds), and pressure is within the raTge between the Pressuri:er Hign and Low Pressure
- trips. The Setooint is automatically varied with: (1) coclar,t te.?:erature te 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) pressuri:er pressure, and (3) axial cower distribu-tion. With normal axial power distribution, this Reactor trip limit is always belew 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 bottom power range nuclear detectors, the Reactor trip is autcmatically reduced according to the notations in Table 2.2-1.
Over: ewer af The Overpower Delta T trip provides assurance of fuel integrity (e.g., no fuel pellet melting and less than 1% claading strain) under all possible overpower conditions, limits the required range for Overtemperature Delta T trip, and provides a backup to the High Neutron Flux trip. The setpoint is automatically varied with: (1) coolant temperature to correct for tempera- . ture induced changes in density and heat capacity of water, and (2) rate of change of temperature for dynamic compensation for piping delays frem the core to the loop temperature detectors. .to ensure that the allowable heat genera-tion rate (kW/f t) is not exceeded. The Overpower AT trip provides protection to mitigate the consequences of various size steam breaks as reported in WCAP 9226, " Reactor Core Response to Excessive Secondary Steam Break." 7 W-STS B 2-1
LIMITING SAFETY SYSTEM SETTINGS
/
BASES Pressurizer Pressure In each of the pressurizer pressure channels, there are two independent bistables, each with its own trip setting to provide for a High and Low Pressure trip thus limiting the pressure range in which reactor operation is permitted. The Low Setpoint trip protects against low pressure which could l lead to DNS by tripping the reactor in the event of a loss of reactor coolant l pressure. On decreasing power the Low Setpoint trip is automatically blocked by P-7 (a power level *of approximately 10% of RATED THERMAL POWER with turbine impulse chamber pressure at approximately 10% of full power equivalent); and on increasing power, automatically reinstated by P-7. The High Setpoint trip functions in conjunction with the pressurizer relief and safety valves to protect the Reactor Coolant System against system overpressure. C*essurizer Water Level The Pressurizer High Water Level trip is provided to prevent water relief through the pressurizer safety valves. On decreasing power tr.e Pressuri:er High Water Level trip is automatically blocked by P-7 (a power level of approximately 10% of RATED THERMAL POWER with a tureine imoulse chammer I pressure at approximately 10% of full equivalent); and on increasing power, i automatically reinstated by P-7. Low Reactor Coolant Flow l The Low Reactor Coolant Flow trips provide core protection to prevent DNB by mitigating the consequences of a loss of flow resulting from the loss of one or more reactor coolant pumps. On increasing power above P-7 (a power level of approximately 10% of RATED THERMAL POWER or a turbine impulse chamber pressure at approximately 10% Oc ' '"" a **r *a"iv*'*" ) ^" '" =^ tic "**ctor tria *'" cc"" more than one loop drops below 90% of nominal. full loop flow. Above P-8 (a t"* * '" g power level of approximatelyy of RATED THERMAL POWER) an automatic Reactor trip will occur if the flow in any single loop drops below 90% of nominal full loop flow. 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 is automatically blocked. 98-1 for Plants permitted N-1 Loco Ooeration l I h goin The P-8 setpoint tri ,, vent the m..:- - nlue of the DNBR from ng normal operational transients and anticipe tr--eiants T W-STS B 2 '!i
LIMITING SAFETY SYSTEM SETTINGS
/
BASES Antional for Plant Permitted N-1 Loop Operation (Continued) when (n-1) loop .p setpoint is adjusted to the va in operation and theall Overtemperature r'a, tion 0g112 With 'the T ified for lo Ojg . Overtemperature Delta T Trip ed to the value specified for (n-1) loop operation . rip at s THERMAL POWER will prevent the minimum ' the ONBR from going below 1. O am ..; aarmal operational tra o and anticipated transients with (n-1) loops in oper l Steam Generator Water Level The Steam Generator Water Level Low-Low trip protects the reactor from loss of heat sink in the event of a sustained steam /feedwater flow mismatch resulting from loss of normal feedwater. The specifiec Setpoint crovices allowances for starting delays of the Auxiliary Feecwater System.
=&Ltam/Feedwater Flow Mismatch and Low Steam Generator hate- Level
.The e reeawater Flow Mismatch in coincidence with a Steam Generat Low Water Level is not used in the transient and at:icent analyses .ut is included in Tatie 2.c . *c ensure the functional cacabili ty of tne .cified
.. trip settings and thereby nca the overall reliability of t eactor Trip System. This trip is redundant 6 he Steam Generator ha evel Lcw-Low trip. The Steam /Feedwater Flow Misma . ortion of
- s trip is activated when the steam flow exceeds the feedwater .' g eate" than or equal to (1.42 x 105) lbs/ hour. The Steam Gener - ow '-ter level cortion cf the
@ trip is activated when the water drops below .%,
trip values include suf. d ent allowance in as .incicateo cy the narrow range instrument. TF excess of normal oper **. values to preclude spurious trics 't will 'nitiate a Reactor trip b e the steam generators are dry. The efore, . edvired capacity an arting time requirements of tne auxiliary feec.a er ou.... are reduce the resulting thermal transient on the Reactor Coolant System generators is minimized. Undervoltace and Underfrecuency - Reactor Coolant Pumo Busses l \ i The Undervoltage and Underfrequency Reactor Coolant Pump Bus trips pro- ^
- vide core protection against DNB as a result of complete loss of forced coolant f
flow. The specified Setpoints assure a Reactor trip signal is generated before the Low Flow Trip setpoint.is reached. Time delays are incorporated in the Underfrequency and Undervoltage trips to prevent spurious Reactor trips from momentary electrical power transients. For undervoltage, the delay is set so that the time required for a signal to reach the Reactor trip breakers following the simultaneous trip of two or' more reactor coolant pump bus circuit } breakers shall not exceed (1.2) seconds. For underfrecuency, the dalay is set so that the time required for a signal to reach the Reactor trip breakers after the Underfrecuency Trip Setpoint is reached shall not exceed (0.3) seconds. On decreasing power the Undervoltage and Underfrequency Reactor Coolant Pump Bus trips are automatically blocked by P-7 (a power level of approximately 10% W-STS B 2 %'9 SEP l 5 Eb
f
' IMITING SAFETY SYSTEM SETTINGS SASES Undervoltage and Underfrequency - Reactor Coolant Pumo Busses (Continued) of RATED THERMAL POWER with a turbine impulse chamber pressure at approximately 10%
by P-7. of full power equivalent); and on increasing power, reinstated automatically Turbine Trio A Turbine trip initiates a Reactor trip. On decreasing power the Reactor ($) trip from the lurbine trip is automatically blocked by P-9 (a power level of approximately,)C% of RATED THERMAL POWER); and on increasing power, reinstated automatically by P-9. Safety Iniection Inout from ESF If a Reactor trip has not already been generated by the Reactor Trip System instrumentation, the ESF automatic actuation logic channels will initiate a Reactor trip upon any signal which initiates a Safety Injection. The ESF instrumentation channels which initiate a Safety Injection signal are shown in Tatle 3.3-3. General Warnine Alarm ' A General Warning Alarm in both Solid State Tri: System trains j initiates a Reactor trip. The General Warning Alarm is activated in each train of the Solid State Trip System when the train is being testec or is otherwise inoperable. The General Warning Alarm trip provides protection for conditions under which both trains of the Trip System may be rencered inocerable. Reactor Trio System Interlocks
. The Reactor Trip System interlocks perform the following functions:
P-6 On increasing power P-6 allows the manual block of the Source Range trip (i.e. , prevents premature block of Source Range trip), p ';id : 0 b:chup bleck fer Scurce Range Neutr;r " lux doubling, and d: en:rgire: thf high volt:g: to the detect;r;. On decreasing power, Source Range Level trips ,are automatically reactivated and . high voltage restored. P-7 On increasing power P-7 automatically enables Reactor trips on low flow in more than one reactor coolant loop, reactor coolant ' pump (E) bus undervoltage and underfrequency,-Twebine trip, pressurizer low pressure and pressurizer high level. On decreasing power, the above listed trips are automatically blocked. 10 W-STS B 2-B. I
I LlMlTING SAFETY SYSTEM SETTlNGS (' '
, BASES Reactor Trip System Interlocks (Continued)
P-8 On increasing power P-8 automatica .f enables Reactor trips on low' flow in one or more reactor coolant loops. :nd en: Or ::r: r:::tcr ([) :::1:nt p;;; br;;h:re crea On decreasing power, the P-8 auto-matica11y blocks the above li sted trips. I I P-9 On increasing power P-9 automatically enables Reactor trip on I Turbine trip. On decreasing power, P-9 automatically blocks Reactor j trip on Turbine trip. P-10 On increasing power P-10 allows the manual block of the Intermediate Range trip and the Flow Setpoint Power Range trip; and automatically blocks the Source Range trip and de energizes the Source Range hign voltage power. On decreasing power, the Intermediate Ra ge tric and the Low Setpoint Power Range trip are automatically reac-ivated. Provices input to P-7. P-13 Provides input to P-7. l l l l i
~
i 11 y-STS B 2-1 9 w + -
6
/
SECTIONS 3.0 AND 4.0 LIMITING CONDITIONS FOR OPERATION AND SURdEILLANCEREQUIREMENTS (
( 3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.0 APPLICABILITY LIMITING CONDITION FOR OPERATION 3.0.1 Compliance with the Limiting Conditions for Operation contained in the succeeding Specifications is required during the OPERATIONAL MODES or other conditions specified.therein; except that upon failure to meet the Limiting Conditions for Operation, the associated ACTION requirements shall be met. 3.0.2 Noncompliance with a Specification shall exist when the requirements of the Limiting Condition for Operation and associated ACTION requirements are not met within the specified time intervals. If the Limiting Condition for Operation is restored prior to expiration of the specified time intervals, completion of the Action requirements is not required. 3.0.3 When a Limiting Condition for Operation is not met, except as provided in the associated ACTION requirements, within one hour action shall be initiated to place the unit in a MODE in which the Specification does not apply by placing it, as applicable, in:
- 1. At least HOT STANDBY within the next 6 hours,
/' 2. At least HOT SHUTDOWN within the following 6 hours, and
- 3. At.least COLD SHUTDOWN within the subsequent 24 hours.
Where corrective measures are completed that permit operation under the ACTION requirements, the ACTION may be taken in accordance with the specified time limits as measured from the time of failure to meet the Limiting Condition for Operation. Exceptions to these requirements are stated in the individual Specifications. This Specification is not applicable in MODES 5 or 6. 3.0.4 Entry into an OPERATIONAL MODE or other specified condition shall not be made unless the conditions for the Limiting Condition for Operation are met without reliance on provisions contained in the ACTION requirements. This provision shall not prevent passage through or to OPERATIONAL MODES as required to comply with ACTION requirements. Exceptions to these requirements are
. . stated in the individual Specifications. .
- I W-STS 3/4 0-1 JUL 2 71981
- - .,,-.m - - , . - , -
l 1 APPLICA8ILITY , t
! $URVEILLANCE REQUIREMENTS l
4.0.1 Surveillance Requirements shall be met during the OPERATIONAL MODES
! or other conditions specified for individual Limiting Conditions for Operation unless otherwise stated in an individual Surveillance Requirement.
- 4.0.2 Each Surveillance Requirement shall be performed within the specified 1
time interval with:
- a. A maximum allowable extension not to exceed 25% of the surveillance
- interval,'but
- b. The combined time interval for any 3 consecutive surveillance intervals shall not exceed 3.25 times the specified surveillance interval.
4.0.3 Failure to perform a Surveillance Requirement within the specified time interval shall constitute a failure to meet the OPERABILITY requirements for a Limiting Condition for Operation. Exceptions to these requirements i are stated in the individual Specifications. Surveillance Requirements do not have to be performed on inoperable equipment. 4.0.4 Entry into an OPERATIONAL MODE or other specified condition shall not be made unless the Surveillance Requirement (s) associated with the Limiting j Condition for Operation have been performed within the stated surveillance l interval or as otherwise specified. 4.0.5 Surveillance Requirements for inservice inspection and testing of ASME Code Class 1, 2, and 3 components shall be applicable as follows:
- a. Inservice inspection of ASME Code Class 1, 2, and 3 components and inservice testing of ASME Code Class 1, 2, and 3 pumps and valves shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50, .Section 50.55a(g), except where specific written relief l has been granted by the Commission pursuant to 10 CFR 50, Section 50.55a(g)(6)(1).
I ~ W-STS 3/4 0-2 JUL 2 31980
r APPLICABILITY SURVEILLANCE REQUIREMENTS (Continued) 4.0.5 (Continued)
- b. Surveillance intervals specified in Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda for the inservice inspection and testing activities required by the ASME Boiler and Pressure Vessel Code and applicable Addenda shall be applicable as-follows in these Technical Specifications:
. ASME Boiler and Pressure Vessel Required frequencies for Code and applicable Addenda performing inservice terminology for inservice inspection and testing inspection and testing activities activities Weekly At least once per 7 days Monthly At least once per 31 days Quarterly or every 3 months At least once per 92 days Semiannually or every 6 months At least once per 184 days Every 9 months At least once per 276 days Yearly or annually At least once per 356 days
- c. The provisions of Specification 4.0.2 are applicable to the above required frequencies for performing inservice inspection and testing activities.
- d. Performance of the above inservice inspection and testing activities shall be in addition to other specified Surveillance Requirements.
- e. Nothing in the ASME Boiler and Pressure Vessel Code shall be construed to supersede the requirements of any Technical Specification.
W-STS 3/4 0-3 'NOV 2 61980
.-_ ~ --_ . _- . . .
JUSTIFICATIONS Section 3/4.1 J In the text of Section 3/4.1 where a capital letter with a circle around it appears, please refer to this sheet for the appropriate justification. A. Seabrook Station plant specific data. B. The requirement to consider the most reactive rod to be fully withdrawn is applicable when credit is being taken for withdrawn rods to meet shutdown pargin requirements, i.e., during critical operation. However, in those instances when all rods are known to be fully inserted the requirement to assume the most reactive rod is' withdrawn constitutes an unnecessary burden on plant operations and the needless processing of primary coolant. C. These changes allow the use of boron concentration limits as well as RCCA withawal limits to maintain the NTC less positive than o pcm/*F. D. Heat tracing is not required on any CVCS components which contain 4 we. percent boric acid providing these components are located in a building maintained at 65'F or higher and has redundant temperature indication and alarms. E. Clarifies ACTION statement that only one flow path is required in MODE 4. F. Clarifies ACTION statement that only one pump is required in MODE 4. G. There are no part length rods at Seabrook Station. H. As long as the shutdown margin, rod insertion limits, and peaking factors are within the limits, no further ACTION is necessary. I. To clarify the LCO for Digital Rod Position Indication J. A surveillance interval of 24 hours is sufficient since other surveillance requirements (OPTR, Spec. 3.2.4 and Rod Misalignment, Spec 3.1.3.1) contain sufficient monitoring . K. N-1 loop operation is not permitted at Seabrook Station. L. This note was added to identify, to the operator, a move limiting Action statement regarding the inoperability of one charging pump while in Mode 4. m
- --- , - . . , - , , , , s 4 - , - . , , 3 , , ~ - , - - . - - , . - - + , .
3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL SHUT 00WN MARGIN - T >200*F LIMITING CONDITION FOR OPERATION i.3 3.1.1.1 The SHUTDOWN MARGIN shall be greater than or equal to (b6%) delta k/kforg)loopoperation. APPLICA8ILITY: MODES 1, 2*, 3, and 4. ACTION: i.3 With the SHUTDOWN MARGIN less than ( h6%) delta k/k, immediately initiate and continue boration at greater than or equal to 30 gpm of a solution containing greater than or equal to 7000 ppm boron or equivalent until the required SHUTDOWN MARGIN is restored. SURVEILLANCE REQUIREMENTS 4.1.1.1.1 The SHUTOOWN MARGIN shall be determined to be greater than or equal to (h6%) delta k/k: 13
- a. Within one hour after detection of an inoperable control rod (s) and at least once per 12 hours thereafter while the rod (s) is inoperable.
If the inoperable control rod is immovable or untrippable, the above required SHUTDOWN MARGIN shall be verified acceptable'with an increased allowance for the withdrawn worth of the immovable or untrippable control rod (s).
- b. When in MODE 1 or MODE 2 with K,ff greater than or equal to 1.0 at least once per 12 hours by verifying that control bank withdrawal is -
within the limits of Specification 3.1.3.6. l l c. When in MODE 2 with K,7f less than 1.0, within 4 hours prior to achieving reactor criticality by verifying that the predicted critical control rod position is within the limits of Specification 3.1.3.6. .
- d. Prior to initial operation above 5% RATED THERMAL POWER after each
. fuel loading, by consideration of the factors of e below, with the control banks 'at the maximum insertion limit of Specification 3.1.3.6.
I
*See Special Test Exception 3.10.1.
3/4 1-1 W-STS
.NOV 2 01980
REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
- e. When in MODES 3 or 4, at least once per 24 hours by consideration of the following factors:
- 1. Reactor coolant system boron concentration,
- 2. Control rod position,* '
- 3. Reactor coclant system average temperature,
- 4. Fuel burnup based on gross thermal energy generation,
- 5. Xenon concentration, and
- 6. Samarium concentration.
4.1.1.1.2 The overall core reactivity balance shall be compared to predicted values to demonstrate, agreement within i 1% delta k/k at least once per 31 Effective Full Power Days (EFPD). This comparison shall consider at least those factors stated in Specification (4.1.1.1.1.e), above. The predicted reactivity values shall be adjusted (n: n itzed) to correspond to the actual core conditions prior to exceeding a fuel burnup of 60 Effective Full Power Days after each fuel loading. l W l
- Wdh all conbl rods known 4, be, Gily m3erfed , s s Aroown h44MN g defemma,4 ion may inke crejit for %e yeae.+ivi41 wor % af th con 4rol l rol assum.J io he 6 th u, N r wn.
OCT I B75 3/4 1-2 W-STS
REACTIVITY CONTROL SYSTEM 3 i SHUTDOWN MARGIN - T,yg < 200*F LIMITING CONDITION FOR OPERATION 3.1.1.2 The SHUTDOWN MARGIN shall be greater than or equal to 1.0% delta k/k. APPLICABILITY: MODE 5. . ACTION: With the SHUTDOWN MARGIN less than 1.0% delta k/k, immediately initiate and continue boration at greater than or equal to A gpm of a solution containing greater than or equal tomppm boron or equivalent until the required SHUTDOWN MARGIN is restored. g SURVEILLANCE REQUIREMENTS 4.1.1.2 The SHUTDOWN MARGIN shall be determined to be greater than or equal to 1.0% delta k/k: i
- a. Within one hour after detection of an inoperable control rod (s) and at least once per 12 hours thereafter while the rod (s) is inoperable.
If the inoperable control rod is immovable or untrippable, the SHUTDOWN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the immovable or untrippable control rod (s). l b. At leas.t once per 24 hours by consideration of the following factors:
- 1. Reactor coolant system boron concentration,
- 2. Control rod position, *
- 3. Reactor. coolant system average temperature, j 4. Fuel burnup based on gross thermal energy generation, S. Xenon concentration, and .
- 6. Samarium concentration.
4 Wdh all co.nbl vods k.nown lo Ise. fully inse*+cd., S HuTDoufo M A R r,.ars o h deferwnaMon may kke cre8t+ -for %e re.4c+iMy wor % of iiie con b l a ,,_a 4o u u, .wa_,. W-STS 3/4 1-3 NOV 2 01980 e
l i
~
REACTIVITY CONTROL SYSTEMS MODERATOR TEMPERATURE COEFFICIENT LIMITING CONDITION FOR OPERATION i 3.1.1.3 The moderator temperature coefficient (MTC) shall be: i Less positive than (0) delta k/k/*F for the all rods withdrawn, l a. . I cle life (BOL), hot zero THERMAL POWER condition. beginningofcy$ 54 4 l b. Less negative than -(&:4) x 10 delta k/k/'F for the all rods withdrawn, end of cycle life-(EOL), RATED THERMAL POWER condition. , APPLICABILITY: Specification 3.1.1.3.a - MODES 1 and 2* only#. ( Specification 3.1.1.3.b - MODES 1, 2, and 3 only#. ACTION:
- a. With the MTC more positive than the limit of 3.1.1.3.a above, opera-tion in MODES 1 and 2 may proceed provided:
, o C:nt.pe.caf ing1 r d -.thdr:r ! limits are established and maintained 1. sufficient to restore the MTC to less positive than 0 delta
@ k/k/'F within 24 hours or be in HOT STANDBY within the next 6 hours. 'here eithdr:r ! 'frit: che!' be d-dditier te the
'n :rtier. 'frit c' Sp::f'icatier ?.'.2.5.
! 2. The27 ! : : a e maintained "ithi- th: withdrau:1 limits c:t blish d at=: until a subsequent calculation verifies that i h the MTC has been restored to within its limit for the all rods withdrawn condition.
- 3. In lieu of any other report required by Specification 6.9.1, a Special Report is prepared and submitted to the Cr-amission t
pursuant to Specification 6.9.2 within 10 days, describing the value of the measured MTC, the interim centre! red .ithdruc1 oPo'a+8h l
@ limits, and the predicted average core burnup necessary for restoring the positive MTC to within its limit for the all rods withdrawn condition.
- b. With the MTC more negative than the limit of 3.1.1.3.b above, be in l
, HOT SHUTDOWN within 12 hours.
"With K,ff greater than or equal to 1.0. #See Special Test Exception 3.10.3.
W-STS 3/4 1-4 MAY 15 880
l - REACTIVITY CONTROL SYSTEMS f l
~
SURVEILLANCE REQUIREMENTS-l l 4.1.1.3 The MTC shall be determined to be within its limits during each fuel cycle as follows: -
- a. The MTC shall be measured and compared to the BOL limit of Specifi-cation 3.1.1.3.a, above, prior to initial. operation above 5% of RATED THERMAL POWER, after each fuel loading. ,
b_. The MTC shal Q4.} _} -(TvG) x 10 } be measured at any THERMAL POWER and compared delta k/k/*F (all rods withdrawn, RATED THERMAL POWER condition) within 7 EFPD after reaching an equilibrium boron concen-tration of 300 ppe. Intheevent.ghiscomparisonindicatestheMTC is more negative than -(i x 10 delta k/k/*F, the MTC shall be remeasured, and compared'\o)the EOL MTC limit of specification 3.1.1.3.b, .at least once per 14 EFPD during the remainder of the fuel cycle. q, t o W-STS 3/4 1-5 AUG 1 3 73
-- , , , - . - - , , , . - , , . . . ,-n--..,---
i REACTIVITY CONTROL SYSTEMS MINIMUM TEMPERATURE FOR CRITICALITY LIMITING CONDITION FOR OPERATION 3.1.1.4 The Reactor Coolant System lowest operating loop temperature (T**9) shall be greater than or equal to (SM)*F. APPLICA8ILITY: MODES 1 and 2 . ACTION:
. less than With a Reactor Coolant to System operating within loop its limit temperature within (T*89)be in HOT 15 minutes FSI (444-)*F, STANDBY within the restore T*XExt 15 minutes.
SURVEILLANCE REQUIREMENTS 4.1.1.4 The Reactor Coolant System temperature (T'#9) shall be determined to i be greater than or equal to (-SM)*F: t Sft
- a. Within 15 minutes prior to achieving reactor criticality, and
- b. At least once per 30 minutes when the reactor is critical and the Reactor Coolant System T DeviationAlarmnotreseU9 is less than (554)*F 541 with the T**9-T
1
#With K greater than or equal to 1.0.
*SeeSpINa1TestException3.10.3.
l ' W-STS 3/4 1-6 15 M
REACTIVITY CONTROL SYSTEMS 3/4.1.2 BORATION SYSTEMS FLOW PATH - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.2.1 As a minimum, one of the following boron injection flow paths shall be OPERABLE _ and capable of being powered from an OPERABLE emergency power [ source.
- a. A flow path from the boric acid tanks via either a boric acid transfer pump or a gravity feed connection and a charging pump to the Reactor Coolant System if the boric acid storage tank in Specification (3.1.2.5a) is OPERABLE, or
- b. The flow path from the refueling water storage tank via a charging pump to the Reactor Coolant System if the refueling water storage tank in Specification-(3.1.2.5b) is OPERABLE.
APPLICABILITY: MODES 5 and 6. ACTION: With none of the above flow paths OPERABLE or capable of being powered from an OPERABLE emergency power source, suspend all operations involving CORE ALTERATIONS or positive reactivity changes. ( SURVEILLANCE REQUIREMENTS l 4.1.2.1 At least one of the above required flow paths shall be demonstrated OPERABL
)s. At l=:t en:: per ' day: by verifying that the t::perature of the-hat traced pertion af the flew path is grater th= cr equal t0-(55)*F rher : ' lee path % = the beric acid tark: i: u:ed.
@ Y.
At least once per 31 days by verifying that each valve (manual, power . sealed, or otherwise secured in position, is in its correct (position. J [ W-STS 3/4 1-7 JUL
- 71981 l
REACTIVITY CONTROL SYSTEMS FLOW PATHS - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.2 At least two of the following three boron injection flow paths shall be OPERA 8LE:
- a. The flow path from f.he boric acid tanks via a boric acid transfer pump and a charging pump to the Reactor Coolant System.
- b. Two flow paths from the refueling water storage tank via charging l pumps to the Reactor Coolant System. ,
APPLICABILITY: MODES 1, 2, 3, and 4 . ACTION: With only one of the above required boron injection flow paths to the Reactor Coolant System OPERABLE, restore at least two baron injection flow paths to the Reactor Coolant System to OPERABLE status within 72 hours or be in at least HOT STANDBY and borated to a SHUTDOWN MARGIN equivalent to at least 1% delta k/k at 200*F within the next 6 hours; restore at least two flow paths to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours. SURVEILLANCE REQUIREMENTS l 4.1.2.2 At least two of the above required flow paths shall be demonstrated OPERABLE: 4t leret :::: p..- 7 days by verifyin;; that the t:rperature f the h::t tg;ed p;rti:n Of the '?;.; p th fr;; the Leiiu asid tanks is grerter ther e eg :1 t: (SS)*C ch;n it is e requiced -etec source. ! o. b. At least once per 31 days by verifying that each valve (manual, power operated, or automatic) in the flow path that is not locked, sealed. or otherwise secured in position, is in its correct position. b.t At least once per 18 months during shutdown by verifying that each automatic valve in the flow nath actuates to its correct position on a Qst signal. lsafety injec}ToD @ , j C1L At least once per 18 months by verifying that the flow path required by Specification 3.1.2.2.a delivers at least 30 gpm to the Reactor Coolant System. 0nly one boron injection flow path is required to be OPERABLE whenever the i temperature of one or more of the RCS cold legs is less than or equal to l (4E*F. If 1% Pa% 16 inoPecable., %e alieve AcTiom stafe,.ent shall aN P I
@ 30 7except Eat egly one flew pa% need be wesforcJ io OPER ABLE stalus wlMin
-the stated he 16 ifs W-STS 3/4 1-8 h1AR 111981
f REACTIVITY CONTROL SYSTEMS CHARGING PUMP - SHUTOOWN LIMITING CONDITION FOR OPERATION 3.1.2.3 One charging pump in the boron injection flow path required by Specification (3.1.2.1) shall be OPERABLE and capable of being powered from an OPERABLE emergency power source.
~
APPLICA8ILITY: MODES 5.ind'6. ACTION: With no charging pump OPERABLE or capable of being powered from an OPERABLE emergency power source, suspend all operations involving CORE ALTERATIONS or positive reactivity changes. ,9 . SURVEILLANCE REQUIREMENTS
- t l
-4.1.2.3.1 The above required charging pump shall be demonstrated OPERABLE by verifying, that on recirculation flow, the pump develops a discharge pressure
, of greater than or equal to aqso psig when tested pursuant to Specification ! 4.0.5. g 4.1.2.3.2 All charging pumps, excluding the above required OPERABLE pump, i shall be demonstrated inoperable at least once per 12 hours, except wher. the reactor vessel head is removed, by verifying that the motor circuit breakers have been removed from their electrical power supply circuits. l
~
l
- i I
l lg' W-STS 3/4 1-9 JUL 2 71981 y , - - - - , - ,,,-c-----~. - , ,,,--- ,-- e, n.-, .. ,,...., . .-- , ,. - - - - - - , - - - - - . , - - - - - - - -
REACTIVITY CONTROL SYSTEMS i CHARGING PUMPS - OPERATING L!MITING CONDITION FOR OPERATION 3.1.2.4 At least two charging pumps shall be OPERABLE. APPLICABILITY: MODES 1, 2, 3 and 4 . El Also see Se.ckion 3.7 3, ACriori s].
~
ACTION With only one charging pump OPERABLE, restore at least two charging pumps to OPERABLE status within 72 hours or be in at least HOT STANDBY and borated to a SHUTDOWN MARGIN equivalent to at least 1% delta k/k at 200*F within the next 6 ,. hours; restore at least two charging pumps to OPERABLE status within the next l 7 days or be in COLD SHUTDOWN within the next 30 hours, l l l SURVEILLANCE REQUIREMENTS
~
4.1.2.4.1 At least two charging pumps shall be demonstrated OPERABLE by verifying, that on recirculation flow, each pump develops a discharge pressure of greater than or equal to aygo psig when tested pursuant to Specification 4.0.5. g 4.1.2.4.2 All charging pumps, except the above required OPERABLE pump, shall ! be demonstrated inoperable at least once per 12 hours whenever.the temperature of one or more of the RCS cold legs is less than or equal to DM)*F by verifying that the motor circuit breakers have been removed from thei lectrical power supply circuits. 30 g
#A maximum of one centrifugal charging pump shall be OPERABLE whenever the
- temperature of one or more of the RCS cold legs is less than or equal to
)*F. If %s pump gs mope,dle , %e 4.booe. A c.TIOM s+aie=e.d shail ae AI t
(
@ exced %+ only one e.kar63 pump ned Ae resforel +o oeeRmE sta+ e wWo We 34a4e2 %e h'mii.
W-STS 3/4 1-10 MAY l 5 57d
, --,n ,x - , ,- -
1 REACTIVITY CONTROL SYSTEMS BORATED WATER SOURCE - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.2.5 As a minimum, one of the following borated water sources shall be OPERABLE: a. with Aboricacidstoragesystem:nd:t1::t:re:h::cf:ted5::ttacing l
- 1. A minimum contained borated water volume of VFOO gallons,
- 2. Between (21000 27
,000) and (?_,_fhR)
__ ppm of boron, and 4dF
- 3. A minimum solution temperature of (446)*F.
- b. The refueling water storage tank with:
- 1. A minimum contained borated water volume of 24,500 gallons, i 2. A minimum boron concentration of (2000) ppm, and
- 3. A minimum solution temperature of (35)*F.
APPLICABILITY: MODES 5 and 6. i ACTION: With no borated water source OPERABLE, suspend all operations involving CORE t ALTERATIONS or positive reactivity changes. SURVEILLANCE REQUIREMENTS l 4.1.2.5 The above required borated water source shall be demonstrated OPERABLE:
- a. At least once per 7 days by:
- 1. Verifying the boron concentration of the water, ,
- 2. Verifying the contained borated water volume, and
- 3. Verifying the boric acid storage tank solution temperature when it is the source of borated water.
- b. At least once per 24 hours by verifying the RWST temperature.eher f t i: th: ::urc; ;f be sted ::ter :nd th -{ stsido) ei.- t;;;;r:tur is-12;; than (35) r.
W-STS 3/4 1-11 NOV 2 01980
REACTIVITY CONTROL SYSTEMS 80 RATED WATER SOURCES - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.6 As a minimum, the following borated water source (s) shall be OPERABLE as required by Specification 3.1 2.2: .
- a. A boric acid storage system Y I+ 'ee't en ::::icted h::t tracing-
,,,.. _ m<.wm
\
- 1. A minimum contained borated water volume of 20,:L90 gallons ,
7000 7700
- 2. Between (20,000-) and (22,50^) ppe of boron, and 3.
GS~ A minimum solution temperature of (t+5)*F. &
- b. The refueling water storage tank with:
- 1. A contained borated water volume of between v79,000 and 'ig5iooo gallons,
- 2. Between (2000) and (2100) ppm of boron, and
- 3. A minimum solution temperature of (35)'F. .
i APPLICABILITY: MODES 1, 2, 3 and 4. l ACTION:
- a. With the boric acid storage system inoperable and being used as one of the above required borated water sources, restore the storage system to OPERABLE status within 72 hours or be in at least HOT STANOBY within the next 6 hours and borated to a SHUTDOWN MARGIN equivalent to at least 1% delta k/k at 200*F; restore the boric acid storage system to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours.
'b . With the refueling water storage tank inoperable, restore the tank to OPERABLE status within one hour or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. -
W-STS 3/4 1- 12 .N0y 2 0 $80 v - ..
REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS 4.1.2.6 Each borated water source shall be demonstrated OPERABLE:
- a. At least once per 7 days by:
- 1. Verifying the boron concentration in the water, a5A;: 2. Verifying the contained borated water volume of the water
! source, and
- 3. Verifying the boric acid storage system solution temperature when it is the source of borated water.
- b. At least once per 24 hours by verifying the RWST temperature. whee g the (c'* side) ri- t: p:r:tur; i; 1;;; tt.;r, (25) F U
^
l i W-STS 3/4 1-13
-NOV 2 01980
_ , _ , _ . _ . . . ~ _ . _ . - . . . _ - . . - . . , . . _ . _ c _ .y- . . . . -.
REACTIVITY CONTROL SYSTEMS
~
3/4.1.3 MOVABLE CONTROL ASSEMBLIES GROUP HEIGHT LIMITING CONDITION FOR OPERATION 3.1.3.1 All full length (shutdown and controlj rodsy :nd all pact ler.gth r:d; which are inserted in the core, shall be OPERABLE and positioned within 212 steps (indicated position) of their group step counter demand position. APPLICABILITY: MODES 1* and 2*. ACTION:
- a. With one or more full length rods inoperable due to being immovable as a result of excessive friction or mechanical interference or known to be untrippable, determine that the SHUTDOWN MARGIN require-ment of Specification 3.1.1.1 is satisfied within 1 hour and be in HOT STANDBY within 6 hours.
- b. With more than one full er p;rt length rod inoperable or misaligned from the group step counter demand position by more than i 12 steps (indicated position), ginb HOT STANDBY within-6 hours.
- c. With one full er p.. ; Wgth rod trippable but inoperable due to causes other than addressed by ACTION a, above, or misaligned from its grcup step counter demand height by more than i 12 steps (indicated position), POWER OPERATION may continue provided that within one hour either:
- 1. The rod is restored to OPERABLE status within the above alignment requirements, or
- 2. The rod is declared inoperable and the remainder of the rods in the group with the inoperable rod are aligned to within i 12 steps of the inoperable rod while maintaining the rod sequence and insertion limits of Figure $ (3.1-1) red (3.1-2). The THERMAL Q3)
POWER level shall be restricted pursuant to Specification (3.1.3.6) during subsequent operation, or ! 3. The rod is declared inoperable and the SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is satisfied. POWER OPERATION may then continue provided that: a) A reevaluation of each accident analysis of Table 3.1-1 is performed within 5 days; this reevaluation shall confirm that the previously analyzed results of these accidents remain valid for the duration of operation under these l conditions. b) The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 ! is determined 'at least once per 12 hours. i "Seg Special Test Exceptions 3.10.2 and 3.10.3. W-STS 3/4 1-14 'NOV 2 1987 e
t i REACTIVITY CONTROL SYSTEMS ACTION (Continued) c) Apowerdistributionmapisobtagnedfromthemovable incore detectors and F (Z) and F a withintheirlimitswikhin72hobYs.reverifiedtobe
)is -The T:lC'"'".L "'".'C", level is r;d:::d t: 1::: th:n er eRe-' te 75% :f "",TE" T;;CR"AL "0WCR ithi.. ih: n:xt 5:ur l p4 : d rithi- the f:11; air,g ? h ur; the high neutrer *'ex
-t**? ??+" cia + 4e r:duced to le;a ih .. er ;;;:1 t: 95%
- f PAT 50 S C".,".L .'0WCR.
SURVEILLANCE REQUIREMENTS 4.1.3.1.1 The position of each full and p:rt length rod shall be determined to be within the group demand limit by verifying the individual rod positions at least once per 12 hours except during time intervals when the Rod Position , Deviation Monitor is inoperable, then verify the group positions at least once l per 4 hours. l 4.1.3.1.2 Each full length rod not fully inserted and :::P p:rt 1Ength'r:d l uht:h i; '7:;rt:d '- th: :: e shall be determined to be OPERABLE by movement l of at least 10 steps in any one direction at least once per 31 days. i l i l l I, t I SO'l 2 8 81 W-STS 3/4 1-15 l l l
TABLE 3.1-1 ACCIDENT ANALYSES REQUIRING REEVALUATION IN THE EVENT OF AN INOPERABLE FULL 02 1'A;T LENGTH ROD (}} [ Rod Cluster Control Assembly Insertion Characteristics Rod Cluster Control Assembly Misalignment l Loss Of Reactor Coolant From Small Ruptured Pipes Or From Cracks In Large Pipes Which Actuates The Emergency Core Cooling System Single Rod Cluster Contr.o1 Assembly Withdrawal At Full Power Major Reactor Coolant System Pipe Ruptures (Loss Of Coolant Accident) Major Secondary System Pipe Rupture Rupture of a Control Rod Drive Mechanism Housing (Rod Cluster Control Assembly Ejection) l . l W-STS 3/4 1-16 OCT 1 1976 \
,- - - , ,.w-. , _ .,- -.,e. y,_._. . ..y__. ,.-.g. ym,. _- , ,. ~ _ . ._,7.. v. . &
REACTIVITY CONTROL SYSTEMS POSITION INDICATION SYSTEMS-OPERATING LIMITING CONDITION FOR OPERATION 3.1.3.2 The ""' ' - aa+-a' =ad a -+ '--a+' control rod position indication system and the demand position indication system shall be OPERABLE and capable
' OI of' Odetermining the control rod positions within i 12 steps dea cea+=l banks he+ ween ma& 2AT steps. Fe. ska+Jewn ba=ks,b posi+ ion mus4 ac dake ;iek te N range be.4 ween o ad 19 APPLICA8ILITY: MODES 1 and 2. 5 '* P ' * *
- 8 ' ' *"d * * ' 5 +* **
- ACTION:
- a. With a maximum of one rod position indicator per bank inoperable
- either:
- 1. Determine the position of the non-indicating rod (s indirectly by the movable incore detectors at least once per hours and immediately after any motion of the non-indicating rod which exceeds 24 steps in one direction since the last determination of the rod's position, or
- 2. Reduce THERMAL POWER TO less than 50% of RATED THERMAL POWER f within 8 hours.
- b. With a maximum of one demand position indicator per bank inoperabie either:
- 1. Verify that all rod position indicators for the affected bank are OPERABLE and that the most withdrawn rod and the least withdrawn rod of the bank are within a maximum of 12 steps of each other at least once per 8 hours, or
- 2. Reduce THERMAL POWER to less .than 50% of RATED THERMAL POWER within 8 hours.
SURVEILLANCE REQUIREMENTS 4.1.3.2 Each rod position indicator shall be determined to be OPERABLE by
~
( verifying that the demand position indication system and the rod position ! indication system agree within 12 steps at least once per 12 hours except l~ during time intervals when the Rod Position Deviation Monitor is inoperable, then compare the demand position indication system and the rod position indica-tion system at least once per 4 hours. l ( l W-STS 3/4 1-17 - JUL 151979
..---,..-n. . _ , = . , , , , . - y _ - . - , . , , . - , - . _ _ , --..-,,.,,.-n -
,g, , , - , , , , . _ , , . . ., .-n,.,, . . , , - +-,m - - - -
REACTIVITY CONTROL SYSTEMS POSITION INDICATION SYSTEM-SHUT 00WN LIMITING CONDITION FOR OPERATION 3.1.3.3 One rod position indicator (excluding demand position indication) shall be OPERABLE and capaole of determining the control rod position within fg each-chutena cogtrolera s 6,r OI iM12lLh step $ d Ya C U U N I Na7.'0'be'e$ert 'e--th rod not fulig,igserted."((.fM,Uk'A'/[ 'af APPLICABILITY: MODES 3*#, 4*# and 5*#. ACTION: With less than the above required position indicator (s) OPERABLE, immediately open the reactor trip system breakers. SURVEILLANCE REQUIREMENTS 4.1.3.3 Each of the above required rod position indicator (s) shall be determined to be OPERABLE by performance of a -C:lANNEL TU::CT!CN^,L TEST at least once per 18 months. Avetos cuauutu orsaAmonun. Test
*With the reactor trip system breakers in the closed position. #See Special Test Exception 3.10.5.
NOV 2 01980 W-STS 3/4 1-18 l I
e REACTIVITY CONTROL SYSTEMS ROD OROP TIME LIMITING CONDITION FOR OPERATION ($) 33 3.1.3.4 The individual full length (shutdown and control) rod drop time from the fully withdrawn position shall be less than or equal to (070-) seconds from beginning of decay of stationary gripper coil voltage to dashpot entry with: 500
- a. T,yg greater than'or. equal to (5++)*F, and
- b. All reactor coolant pumps operating.
APPLICABILITY: MODES 1 and 2. ACTION:
- a. With the drop time of any full length rod determined to exceed the above limit, restore the rod drop time to within the above limit prior to proceeding to MODE 1 or 2.
X. With the acd drop ti;;s withia '#-it; but determined with n ' e=c+^e 0001:nt p" p: eperating, Oper:ti:n ;;y ptec::d pr ;ided TuEDMM - POWE" i r :tricted t0.
*k L::: th:n er equ:1 t: (003% f "^ TEE T" R"AL POWER -oen th: @ r:: ter cre!=a+ step v:!ve: 4- the n neper: ting leep :re open;-
e 1L L :: th:n er equa! t: (70)% ef RAT 0 T" ""a.L PCMEo when the r::cter ree'ent e+ap " !ve: i- the nen pec Ling ::p re cle:;d. SURVEILLANCE REQUIREMENTS 4.1.3.4 The rod drop time of full length rods shall be demonstrated through measurement prior to reactor criticality: l I
- a. For all rods following each removal of the reactor vessel head,
! b. For specifically affected individual rods following any maintenance on. or modification to the control rod drive system which could affect the drop time of those specific rods, and
- c. At least once per 18 months.
W-STS 3/4 1-19 1 . 1976 r n -- - ---r-- - mw
j REACTIVITY CONTROL SYSTEMS SHUTDOWN ROD INSERTION LIMIT i LIMITING CONDITION FOR OPERATION 3.1.3.5 All shutdown rods shall be fully withdrawn. APPLICABILITY: MODES 1* and 2*#. l ACTION: With a maximum of one shutdown rod not fully withdrawn, except for. surveillance testing pursuant to Specification (4.1.3.1.2), within 1 hour either:
- a. Fully withdraw the rod, or l
- b. Declare the rod to be inoperable and apply Specification (3.1.3.1).
i SURVEILLANCE REQUIREMENTS 4.1.3.5 Each shutdown rod shall be determined to be fully withdrawn: l a. Wi. thin 15 minuter prior to withdrawal of any rods in control banks A, B, C or D during an approach to reactor critica.lity, and l
- b. At least once per 12 hours thereafter.
*See Special Test Exceptions 3.10.2 and 3.10.3. #With K,ff greater than or equal to 1.0.
( W W-STS 3/4 1-20 'NOV 2 01980 i
. REACTIVITY CONTROL SYSTEMS CONTROL R00 INSERTION LIMITS LIMITING CONDITION FOR OPERATION 3.1.3.6 The control banks shall be limited in physical insertion as shown in Figures (3.1-1) : d (2.'-2). (g)
APPLICA8ILITY: MODES 1* and 2*#. l l ACTION: With the control banks' inserted beyond the above insertion limits, except for surveillance testing pursuant to Specification (4.1.3.1.2), either:
- a. Restore the control banks to within the limits within 2 hours, or
- b. Reduce THERMAL POWER within 2 hours to less than or equal to that fraction of RATED THERMAL POWER which is allowed by the group position using the above figuret, or
- c. Be in at least HOT STANOBY within 6 hours.
,y SURVEILLANCE REQUIREMENTS h
4.1.3.6 The position of each control bank shall be datermined to be within the insertion limits at least once per 12 hours except during time intervals when the Rod Insertion Limit Monitor is inoperable, then verify the individual i rod positions at least once per 4 hours. l l
"See Special Test Exceptions 3.10.2 and 3.10.3.
#With K,ff greater than or equal to 1.0.
.f W-STS 3/4 1-21 -N0'Y 2 01980 P fw y- -r,----e .n-. ->n. ---- . , . , ----.-,. - - - , , - -,,,m -
---r-,,y ,_.- ,.--, m-,. -, - ;-,--,y . - , ,- ,
228-- 220 (0.30.'228) ,/ 0.844,258) 200 / f l 180 ' .0,16 0 160 s M 140_ / / (1.0, 146) 120-8 p 100 / >
/
E [ 80 / 60 . O [(0.0,49) o m 40-20 0 * 0) i i . . . . i . . , .
.0 0.2 0.l4 0.6 0.8 1.0 FRACTION OF RATED THERMAL POWER
! FIGURE 3.1-1 CONTROL ROD INSERTION LIMITS AS A FUNCTION OF POWER 3/4 1-22
~. _
R CTIVITY CONTROL SYSTEMS PAR ENGTH ROD INSERTION LIMITS (OPTIONAL) LIMITING NOITION FOR OPERATION
/
3.1.3.7 The p et length control rod bank shall be:
- a. Limi in physical insertion as shown on Figure 3.1-3), and
- b. Limited from c ering any axial segment of the fuel assemblies for a l period in exces of (18) out of any 30 Equivalent ull Power Days.
APPLICA8ILITY: MODES 1 and 2*. ACTION:
- a. With the part lengt control rod ba inserted beyond the insertion limit of Figure (3.1- ), either:
- 1. Withdraw the part ngth co trol rod bank to within the limit within 2 hours, or
- 2. Reduce THERMAL POWER wi in 2 hours to less than or equal to l that fraction of RATE e RMAL POWER which is allowed by the l bank position using e ab e figure, or in 6 hours.
- 3. Be in at least HOT TANDBY wi l
- b. With the neutron abs er section of e part length control rod I bank covering any al segment of the uel assemblies for a period exceeding 18 out o any 30 consecutive E D period, either:
- 1. Reposition e part length control rod roup to satisfy the l above lin' within 2 hours, or
- 2. Be in least HOT STAND 8Y within the next hours.
SURVEILLANCE REQUI EMENTS \
/
4.1.3.7 The sition of the part length control rod bank shall be etermined at least onc per 12 hours.
*See ial Test Exceptions 3.10.2 and 3.10.3.
W-STS 2/' 1- - NOV 2 01980
NEAC11VITYCONTROLSYSTEMS P LENGTH R00 INSERTION LIMITS (if required by DN8 considerations LIMIT CONDITION FOR OPERATION s , 3.1.3.7 All art length rods shall be fully withdrawn. APPLICA8ILITY: MODES 18 and 2*. ACTIOy: With a maximum of o part length rod not fully w thdrawn, within 1 hour either:
- a. Fully withdr the rod, or
- b. Be in at least T STAND 8Y withi the next 6 hours.
SURVEILLANCE REQUIREMENTS A 4.1.3.7 Each part length rod shall determined to be fully withdrawn by:
- a. Verifying the po tion of the art length rod prior to increasing THERMAL POWER ve 5% of RATE THERMAL POWER, and
- b. Verifying, a least once per 31 da s, that electric power has been dist.onnect from its drive mechani by physical removal of a breaker fr the circuit.
l
- See Special est Exceptions 3.10.2. and 3.10.3.
[ l PSTS, , 2/01 5 NOV 2 01980
JUSTIFICATIONS Section 3/4.2 In the text of Section 3/4.2 where a capital letter with a circle around it appears, please refer to this sheet for the appropriate justification. A. Seabrook Station plant specific data. B. There are no part length rods at Seabrook Station. C. Procedures exist which enable the reduction to be performed while at power (i.e., placing a channel in the trip mode) and thus avoid an unnecessary shutdown. l D. Seabrook Station does not have an APDMS. E. There is no single indication that provides total RCS flow. The four individual loop flow rates must be summed. The change is justified as a clarification as to what must actually be done. F. R2 is deleted from the Technical Specifications per Westinghouse current methodology. G. PSNH letter SBN-502, dated 4/25/83, to the NRC provided information on this change. H. Using just the 4 pairs of symmetric thimbles is limiting in that a plugged thimble or inoperable detector could p'reclude performance of this
. surveillance. The use of a full core flux map (option b) is acceptable in that the data from this map used in conjunction with pretious full core flux maps (taken to satisfy surveillance' requirements 4.2.2.2 and 4.2.3.2) is sufficient to confirm that the Quadrant Power Tilt Ratio is-acceptable. Also, use of a full core flux map is consistent with the bases for the specification on movable detectors (STS 3.3.3.2).
I. N-1 loop operation is not permitted at Seabrook Station. J. Changed value from psia to psig t 1 l 4 w w w , - ,
.e w -- -,- - - . _ m - , - - - - - - -
e
c i y
)
3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE LIMITING CONDITION FOR OPERATION 3.2.1 TheindicatedAXIALFLUXDIFFERENCE(AFD)shallbemaintainedwithinh
.etEr)fr target bands (flux difference units) about the target flux difference.
S ee 3ns e r+ .r APPLICABILITY: MODE 1 above 50% of RATED THERMAL POWER *. ACTION: repired
- a. With the indicated AXIAL FLUX DIFFERENCE outside of the 464% target band about the target flux difference and with THERMAL POWER:
- 1. Above 90% of RATED THERMAL POWER, within 15 minutes either:
a) Restore the indicated AFD to within the target band limits, or l b) Reduce THERMAL POWER to less than 90% of RATED THERMAL POWER. I 2. Between 50% and 90% of RATED THERMAL POWER: a) POWER OPERATION may continue provided: . l l
- 1) The indicated AFD has not been outside of the b g target band for more than 1 hour penalty deviation cumulative during the previous 24 hours, and
- 2) The indicated AFD is within the limits stiown on j Figure (3.2-1). Otherwise, reduce THERMAL' POWER to I
less than 50% of RATED THERMAL POWER within 30 minutes and reduce the Power Range Neutron Flux-High Trip Setpoints to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours. b)' Surveillance testing of the Power Range Neutron Flux l Channels may be performed pursuant to Specification i (4.3.1.1) provided the indicated AFD is maintained within the limits of Figure 3.2-1. A total of 16 hours operation may be accumulated with the AFD outside of the target band during this testing without penalty deviation.
- b. THERMAL POWER shall not be increased above 90% of RATED THERMAL POWER unless the indicated AFD is within the ef!P)%' target band and ACTION a~.2.a) 1), above hat, been satisfied. veguirel "See Special Test Exception 3.10.2.
1
)(-STS 3/4 2-1 NOV 2 0 E80
y y.- - . , .-e, i.- 9* s -*,. 3 ., m.iv. . ; .y 7- . , , . wm . 4
. , - g - O,
,y,,- ., ,
,..e , g ,, .
~,,
v hee ., , oa,.e *. e9 A 4 ..o a J a
. e p
g a t.e*4 . g
. t s
, . ... ,
- g *
, s a e ., . .= ,
-4 6
s s. 6,., 3 F
'ss
. s*. .,,
. . , ., ; a
- 4
*
- r a,,
*( ,
~
4 e e -.-. , . , . . . .
,ss -
s a
" g./,
. ' ~ . 4.. .' .k ,. p,
,e,
' f'* e h--.
4 l .-
; .* p-[
- s. - .
g '^ g
~
g ' ,e * ,.
**g* %
'.M.".
.:~.
s e -.- . , . . e e ,a y .x. , s w._ ...* ., ,o s s .* ,
, . ......_...,.-..vy-
. ,+
- . . . . . _ g..t.... --w.,.<- .,..v . - . . ,4r . . -
. . . . . . . . ~ -_ - - - - . . . . .
. . s g.* * .m.;
r .
,.4. .
.n ,- . .<, ,-s.v ,,u. .7
^ - _- A, ,
h _ r e
.%..ep,? .. .
n%g-*?: L.... s e.
* .* .#. s, , -s . m . : ,,'
s
--, C y.
,-. krmwp :o .i - 3 OOC _AWD . M71L.. . . . . . . .. . _ _ _ _ . _ . _ _ . . _ . _ _ . . _
,..g. ,-
.<. .,% :b. -
- 'm.- . . . '
r: . -.-. ,,_; . e... . .
..S d. u $b *4 $$ OV.A k % L .COV$ h ..M t t M t_Q uwm u KhC{. .. ._. - - . - .
.,a
-m = .s . , . s- q y ,
... ~ , . . . . .Y. , . _ _ Y,?' ...-,'. yp& .g f_' .'.y QR f'} y$ Q g [& .. -_ . . _ _ _ . _ _ -_ .. . . .__
s .. .
.. . , ,.s..- e s . . -
~~
.w.. . .< _ . , -
- ...-.. . . - ~ c..
1 . .3 7. . mea . . for., re.lo . co r.e s. . .. ... . . . . . .- . . _ - ... . ..
. ,,4 . .
s .-
.. s. , , . , ... s s, *
< * .. . . . . .. .aw ,.
- y. . , s - -.....-.
--.. -~. . - ~, s ,,.. - . -..,.-..- . . - - , . , ,
. .,, s,e r. r- .~ ..- - ~...*.. -. - . - - . . . - ,. .
c-n s ,p n s
, ., P a
;. ., .~ - . . . _ _ ..- .. .. - .- - - --
. . -. .c
.d .
4 i
-.-n.. - . ..--.-. ~ . . .
. . . . , , a. r
~
b
. 4 n*'P, p
) .. > ,. , ~.+
, .n... - .~...'m 5-- .~- ~ s.. k. L- ' . . - -L- . . . . . - - - - . . . - - - -. . -+ -- ~
. S
'. 9
- .......1--- -, - . ' . . . . . v .G .. . 4 :. --.s --. , . ,, . . . - . . ....... _ . . .. . . - - .
. +
l . . . , . . . ' . . . . _ . . . - . . . - - .. ~
.% ~.. . . > . . - . 2 .e..4 a
.s
- e e g ., ,. . .
- n, , ..
..-4 .m - . . . . _%..-.
. . . ..-- .%+,.x.
y a.. s. .
..s..
t ,
.. * . . , . . ,. ~.*7 ,. ~n , .;_ -
..A s .. * *. -
- l. ,p .m_
l
., - _; ..y ,.. .\* - , , _ ,_ ,
.-. v ,: . -. . - . -- ~
, , _~ r _
- ---.-_.-e .a -. . ; ,.. < . l. J n L 9.s.. , . .e e
a.- I
.. - .-~ - -
.' _ ~ . - - . . _ ._
. _ . . . - . . ~ ~_.4.-.. . - . , . - .
e-
'.,. . .. .-- _4 ~.. --._ . . .~
% . s s o. : ,
.g< . . s
._~.n..
, . . . , . . ~ . - . . - . . . - , - .. . . . . ;. -- :. -. . . - _ . - .. - . . . -
, . .. , ; * ~
, ~ ^ .,
- x . -.q_m- . ,
,n.o -
- r L . _. . ._. - . .
%..,.7...
.v ...- ..-.. .. ,.;... *.. +.....- C.%
... L ._.w - . . . __ . .
..Q,.
.,,~
3 e. l . . . 4,'**, . a ' ,e .c ( + ] '- l
. . - . _s.
y.. .: - _ .
.4 ... - . - . . . . + , _ . . - - _ . . . . . . . - . . . . - - - - - .
l 6 %. *', ,
. s 1
- - , - . . . . . . _.. . - +-... - .. - . . - .. -
L *
..~^+'
i p y- f
. . .- ,, . . . - .w-. .
. . _ . -> .e..,.--,.-...n.s--..-..-..-.--.........-~. '
. t' s ,,
-~... .- . - -
g
~.-- ~ . ..
a '
. y'** p ,, :
l'"_ r._ y . ,
,.s* . -
-Q s
~J w
*4.
4
. t~ '.c s .L . e n. ' -v- _
* '. .w_. . e
.a .e ",+ *- -s.....'...4.~'.. ,
...--+m -4
...%. w w' 4.~.c,r..-.> .. .1 %., .A .-.w..w+
e..
+ - . .-- ..-- .. - -- . ~ - -. - .--._ .
r v _4v .' '
,,,_';x...
- .~*- , ' , . , t
.' ;, t - ' A['3. ~ , *~ ' *- b . J .*: ,
'.,7, * ' ' ' . ' ,
.. ., ~~' s,.~'.,4 ' ,. ., y
- v. w Ne -
]
y, , , u .- . t
' ,. -......_..--.m- .
l'. sA, .
. . . n
.- s n
. 4 A ~ ,.,,.g. * '-. * ..4a,,,..
*' a 4.
s ,
.s. . ~
. * * . .h
, . ,r '. -
. ,.* * . ., .e r, . -
"* *. > " ... s.
. %' .. , ~ .. ,,
.. ~%.
.s y ,
- - < = . . - - . . . - . - - . _ , . _ - - -
s o. a, ' s% r A- ,
, 's, . i 1
-'.o .
*' . ' ,. . , is g
~
a'4 . .*
' ,,. K . 1' ,T, e p 4._ a ~ _, s.
t...
+ .
,v.. t .a.. 4.,
.,,, y, - , ,
y . , 4 ~'. e Q. . .e.- . ,, s.
- J v. _- ,
, , , . - .-^,
e r *
. . N.:g;,.s-...' 3..
. . - . ,a
.. . .. a... . -
, =.
? .c, *+,,. . n . .
h..1 [$u *,' w -:.
~ '
*f
.* #. ./ +pe*' 'T ~ m- b *1 * +?- ~ . ' **. , , . -
y* *[, ~f2 a, g.:A x , :1 - ~ ge.a.. ,- , -. .. s e y,,. _ .. ;,,.,-
-,.w. a
,c e- .
. .,y .
g (4
. .v a.* e > ,,-....- . . . , . . . , .g., ., , =.; ..~-, . . . . . . * * ;
. a a
e av . t .
*-f .. m ._^q
.p N* 'O s'
- . -*' ' * ' , p
*. b,
.J -g 9-9 w,,,,
h
- a
.*- p f.,=.,.,& '-*^%. %* ' . * - -4 '* * ?' $*,. -.
w Ni 4..f ' ? .e n ,a ,g t.n w - N . . g4g' m.. ,. cMag,,e ,,c.c. s . .,'.'.u w.m..u w m a2t g,m.nw_g.'v.nh' .:., m'Vtc h .o: m . w s'.m m'y"e..,4.2:
. - - = __ - _ _ _ _ .
POWER DISTRIBUTION LIMITS ( ACTION (Continued)
- c. THERMAL POWER shall not be increased above 50% of RATED THERMAL .
POWER unless the indicated AFD has not been outside of the greguired target band for more than 1 hour penalty deviation cumulative during the previous 24 hours. Power increases above 50% of RATED THERMAL ' POWER do not require being within the target band provided the accumulative penalty deviation is not violated. SURVEILLANCE REQUIREMENTS 4.2.1.1 The indicated AXIAL FLUX DIFFERENCE shall be determined to be within its limits during POWER OPERATION above 15% of RATED THERMAL POWER by:
- a. Monitoring the indicated.AFD for each OPERABLE excore channel:
- 1. At least once per 7 days when the AFD Monitor Alarm is OPERABLE, and
- 2. At least once per hour for the first 24 hours after restoring the AFD Monitor Alarm to OPERABLE status.
- b. Monitoring and logging the indicated AXIAL FLUX DIFFERENCE for each OPERABLE excore channel at least once per hour for the first 24 hours and at least once per 30 minutes thereafter, when the AXIAL
' ~ FLUX DIFFERENCE Monitor Alarm is inoperable. The logged values of the indicated AXIAL FLUX DIFFERENCE shall be assumed to exist during the interval preceding each logging.
% regutred 4.2.1.2 The indicated AFD shall be considered outside of it: :(.}*' target band when 2 or more OPERABLE excore channels are indicating the AFD to be outside the target band. Penalty deviation outside of the W target band @
shall be accumulated on a time basis of: velutred
- a. One minute penalty deviation for each 1 minute of POWER OPERATION
- outside of the target band at THERMAL POWER levels equal to or above l 50% of RATED THERMAL POWER, and
~
- b. One-half minute penalty deviation for each 1 minute of POWER OPERATION outside of the target band at THERMAL POWER levels between 15% and 50% of RATED THERMAL POWER.
4.2.1.3 The target flux difference of each OPERABLE excore channel shall be j determined by measurement at least once per 92 Effective Full Power Days,*i4h-11' ;:rt ';ngth ;;ntre! ret 'u!!y H th9 er The provisions of Specification g 4.0.4 are not applicable. 4.2.1.4 The target flux difference shall be updated at least once per 31 Effective Full Power Days by either determining the target flux difference pursuant to 4.2.1.3 above or by linear interpolation between the most recently , measured value and 0 percent at the end of the cycle life. The provisions of Specification 4.0.4 are not applicable. W-STS 3/4 2-2 NOV 2 01980 y .y- r - .-~ s.- - - , . -%=.-,-------.,-m , ,,----.-,,y.,,.-.,-.-,--.--.--m. . , - --
,--..-,,..or--- - - - - - = . - - . - -- - - - - - - - - - - -
( 120 8 Q # E 0 E
!5 100 * >
UNACCEPTABLE (-11,90) (11,90) UNACCEPTABLE OPERATION OPERATION 1 ll l - ( \ ACCEPTABLE OPERATION (-31,50) (31,50) 7 40 l 20 0 40 -30 -20 -10 0 10 20 30 40 50 FLUX DIFFERENCE (41) % ( ) FIGURE 3.2-1
\ AXIAL FLUX DIFFERENCE LIMITS AS A FUNCTION OF RATED THERMAL POWER 3/4 2-
i l POWER DISTRIBUTION LIMITS I 3/4.2.2 HEATFLUXHOTCHANNELFACTOR-Fg i LIMITING CONDITION FOR OPERATION 3.2.2 F q(Z) shall be limited by the following relationships: ( Fq (Z) 1 [2.32] [K(Z)] for P > 0.5. - l P
.Fq (Z) 5 [(4.64)] [K(Z)] for P 5 0.5 where P = THERMAL POWER RATED THERMAL POWER l
I and K(Z) is the function obtained from Figure (3.2-2) for a given core height location. ( APPLICABILITY: MODE 1 ACTION:, With Fq (Z) exceeding its limit:
% C: ;'; cith :ither ;f the f;?l:wi.;; ACTIONS.
i 4,% l Reduce THERMAL POWER at least 1% for each 1%0F (Z) exceeds the l limit within 15 minutes and similiarly reduce the Power Range i Neutron Flux-High Trip setpoints within the next 4 hours; POWER i OPERATION may proceed for up to a total of 72 hours; subsequent l POWER OPERATION may proceed provided the Overpower delta T Trip l Setpoints have been reduced at least 1% for each 1% Fg(Z) l exceeos the limit. The Oveierer d:?t: ' Trip Sctp;i..t r; i:- 1
-ti r. :h:!' be perfer;;d ith th; r::cter " et i=..i iiOT STANDSi.
1 Ewuvue T:: SIT.Ai "0WCE .. nece==..j t; ;;;t th; li.T.ite ef Spswi-
-fie,i.on_(0.2.6) u.ing th; AP0"5 iti. O,e latest inuvi a map and-
! @ +pdrted " (ATCei$ el...Le eniy)
- b. Ide,nt,ify and correct the cause of the out of limit condition prior to. increasing THERMAL POWER above the reduced limit required by a, above; THERMAL POWER may then be increased provided F (Z) is demon-strated through incore mapping to be within its limit 9
( W-sis 3/4 2-4 SEP 151979
( POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS 4.2.2.1 The provisions of Specification 4.0.4 are not applicable. 4.2.2.2 F xy shall be evaluated to determine if F 9(Z) is within its limit by:
- a. Using the movable incore detectors to obtain a power distribution map at any THERMAL POWER greater than 5% of RATED THERMAL POWER.
- b. Increasing the measured.F. component of the power distribution map by 3% to account for manuf,Xcturing tolerances and further increasing the value by 5% to account for measurement uncertainties.
- c. Comparing the F c mputed (Fx) btained in b, above to:
xy
- 1. The F limits for RATED THERMAL POWER (F P) for the appropriate xy measured core planes given in e and f below, and
- 2. The relationship:
l [ F xy =FRTP x [l+0.2(1-P)] + L where F*Y'is the limit for fractional THERMAL POWER operation expressed as a function of F RTP and P is the fraction of RATED xy THERMAL POWER at which F was measured. xy
- d. Remeasuring F xy according to the following schedule:
RTP
- 1. When F is greater than the F x limit for the appropriate measured core plane but less than the F*Y relationship, additional RTP power distribution maps shall te taken and F xyC compared to F xy and F L:
a) Either within 24 hours after exceeding by 20% of RATED THERMAL POWER or greater, the THERMAL POWER at which F x was last determined, or
, b) At least once per 31 EFPD, whichever occurs first.
~k W-STS 3/4 2-5 MAY- 15 M6
POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued) C RTP
- 2. When the F xy is less than or equal to the F xy limit for the appropriate measured core plane, additional power distribution maps shall be taken and F C compared to F RTP and F l at least xy xy xy once per 31 EFPD.
RTP
- e. The F xy limits for RATED THERMAL POWER (Fxy ) shall be provided for all core planes containing bank "D" control rods and all unrodded ,
l core planes in' a Radial Peaking F. actor Limit Report per Specifica- I tion 6.9.1.'10. i
~1
- f. The F limits of e, above, are not applicable in the following core planeFregions as measured in percent of core height from the bottom of the fuel:
, 1. Lower core region from 0 to 15%, inclusive. 1
- 2. Upper core region from 85 to 100%, inclusive. '
1
- 3. Grid plane regions at 17.8 1 2%, 32.1 1 2%, 46.4 i 2%, 60.6 1 2%
and 74.9 i 2%, inclusive (17 x 17 fuel elements).
- 4. Core plane regions within i 2% of core height (i 2.88 inches) about the bank demand position of the bank "D" or part length control rods.
C
- g. With F exceeding F :
C 5f: F (Z)q'irit :h ll be red.ced :t 1:::t 1% fe; ..d,1% r
- d- ::d (f:r F 1:: : it.i r G) less i.:... 2.;2 ::d q
'R. The effects of F on*Q F (Z) shall be evaluated to determine if 1
W t D - Fq (Z) is within its limits. 4.2.2.3 When F (Z) is measured for other than F determinations, an overall measuredF(Z)ShallbeobtainedfromapowerdiHributionmapandincreased by 35 to a0 count for manufacturing tolerances and further increased by 5% to account for measurement uncertainty. _W-STS 3/4 2-6 NOV 2 1981 J
-e n--v--- - ~,n -- --v- - - , . - - - - . . - - - - - -
'i 1.2 l
(6.0,1.0) 1.0 "' "* ' 7
/
G
~
u. 0 0.8 S u A 3 0.6 e o (12,0,0.65) 2 l R g 0.4 l' O.2 0.0 0 2 4 6 8 10 12 f CORE HEIGHT (FT) I l-FIGURE 3.2-2
=
ik K(Z)- NORMAllZED FOI2) AS A FUNCTION OF CORE HEIGHT 7 -
. 3/42'5 i
i
,e.,
POWER DISTRIBUTION LIMITS 3/4.2.3 RCS FLOW RATE AND NUCLEAR ENTHALY RISE HOT CHANNEL FACTOR LIMITING CONDITION FOR OPERATION 3.2.3 RCS fata.! flee =4e. h The combination of 'ndireted R:::t:r C::1:nt Sy:t:r (RCS' iviol ficw ea4e and R , K shall be maintained within the region of allowable operation shownonFlgure3.2-3for4loopoperation. Where: N F 3g a* R
- 1 = 1.49 [1.0 + 0.2 (1.0 - P)]
k*[1-R (SU)] - THERMAL POWER b% P = RATED THERMAL POWER N N c.1L Fg = Measured values of Fg obtained by using the movable incore detectors to obtain a power distribution map. The measured N values of F g shall be used to calculate R since Figure 3.2-3 includes measurement uncertainties of M % for flow and 4% for N 2. 0 incore measurement of FAH, and Q
% R0P (SU) = R:d 90" P;nelty :: ' : functi;n of r:gier :verag" burnu es
-:hrr- in Figur: 3.2 0, where a region is defined 0; those
@ eer " 110: with th; 3;..; 1;; ding date (reic:d:) er enri:bment-(fir:;t ;;re).
APPLICABILITY: MODE 1. ACTION: t With the combination of RCS total flow rate and R), % ou,tside the region of acceptable operation shown on Figure 3.2-3: 0 i a. Within 2 hours either:
- 1. Restore the combination of RCS total flow rate and Rj ,
g % to within the above limits, or
- 2. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER and reduce the Power Range Neutron Flux - High trip setpoint to less than or equal to' 55% of RATED THERMAL POWER within the next 4 hours.
! ~ W-STS 3/4 2-8 NOV 2 01980 t l
POWER DISTRIBUTION LIMITS ACTION: (Continued)
- b. Within 24 hours of initially being outside the above limits, verify through incore flux mapping and RCS total flow rate comparison that the combination of R , and RCS total flow rate are restored to
@ i withintheabovelimits%orreduceTHERMALPOWERtolessthan5%of RATED THERMAL POWER within the next 2 hours.
- c. Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER above the reduced THERMAL POWER limit required by ACTION items a.2. and/or b. above; subsequent POWER ,
@ OPERATION may proceed provided that the combination ofj R , \ and indicated RCS total flow rate are demonstrated, through IncoFe flux mapping and RCS total flow rate comparison, to be within the region of acceptable operation shown on Figure 3.2-3 prior to exceeding the following THERMAL POWER levels:
- 1. A nominal 50% of RATED THERMAL POWER,
- 2. k nominal 75% of RATED THERMAL POWER, and
- 3. Within 24 hours of attaining greater than or equal to 95% of
^ RATED THERMAL POWER. SURVEILLANCE REQUIREMENTS 4.2.3.1 The provisions of Specification 4.0.4 are not applicable. of RCS Mal flamk. @ 4.2.3.2 The combination :f ic.;ic.ted RCS tete; fi n nt: and R determinedtobewithintheregionofacceptableoperationofFlgur83.2-3: h@shallbe
- a. Prior to operation above 75% of RATED THERMAL POWER after each fuel loading, and
- b. At least once per 31 Effective Full Power Days.
4.2.3.3 The 4ad4 sated RCS total flow rate shall be verified to be within the region of acceptable operation of Figure 3.2-3 at least once per 12 hours when h the most recently obtained value% of Ry % , obtained per Specification 4.2.3.2, eee assumed to exist. 4.2.3.4 The RCS kE $ b W k Y $M(([a shall be subjected to a CHANNEL . CALIBRATION at least once per 18 months. 4.2.3.5 The RCS total flow rate shall be determined by measurement at least once per 18 months. W-STS -3/4-2-5 JUL .151979
MEASUREMENT UNCERTAINTIES OF 2.0% FOR FLOW AND 4.0% FOR INCORE MEASUREMENT OF F"g 3 AREINCLUDED IN THIS FIGURE. l 4 48-ACCEPTABLE UNACCEPTABLE OPERATION REGION OPERATION REGION 46 ,
^ .
l 2 L 44 0' O C W . l F 42 E
- 3 o
N 40 I
< ( 1.,3 9.0)
F O F co 38 O l 36 34 O.90 0.95 1.00 1.05 1.10 ] , R= R F[g /1.4 i 0+0.2(1.0-k FIGURE 3.2-3 ~ RCS TOTAL FLOW RATE VS. R SEABROOK UNIT 1-FOUR LOOP OPERATION 3/4 2-10
POWER DISTRIBUTION LIMITS 3/4.2.4 QUADRANT POWER TILT RATIO 4 LIMITING CONDITION FOR OPERATION 3.2.4 The QUADRANT POWER TILT RATIO shall not exceed 1.02. APPLICABILITY: MODE 1 above 50% of RATED THERMAL POWER *. ACTION:
- a. With the QUADRANT POWER TILT RATIO detarmined to exceed 1.02 but less than or equal to 1.09:
- 1. Calculate the QUARANT POWER TILT RATIO at least once per hour until eithe?:
a) The QUADRANT POWER TILT RATIO is reduced to within its limit, or b) THERMAL POWER is reduced to less than 50% of RATED THERMAL POWER.
- 2. Within 2 hours either: ,
a) Reduce the QUADRANT POWER TILT RATIO to within its ! limit, or b) Reduce THERMAL POWER at least 3% from RATED THERMAL POWER for each 1% of indicated QUADRANT POWER TILT RATIO in excess of 1.0 and similarly reduce the Power Range Neutron Flux-High Trip Setpoints within the next 4 hours.
- 3. Verify that the QUADRANT POWER TILT RATIO is within its limit within 24 hours after exceeding the limit or reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within the next 2 hours and reduce the Power Range Neutron Flux-High Trip setpoints to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours.
l 4. Identify and correct the cause of the out of limit condition - prior to increasing THERMAL POWER; subsequent POWER OPERATION above 50% of RATED THERMAL' power may proceed provided that the QUADRANT POWER TILT RATIO is verified within its limit at least once per hour for 12 hours or until verified acceptable at 95% or greater RATED THERMAL POWER.
"See Special Test Exception 3.10.2.
\\
W-STS 3/4 2-14 fl0V 2 01930
POWER DISTRIBUTION LIMITS ACTION: (Continued)
- b. With the QUADRANT POWER TILT RATIO determined to exceed 1.09 due to misalignmentofeitherashutdowngcontr p t M t'-rod:
- 1. Calculate the QUADRANT POWER TILT RATIO at least once per hour until either: .
a) The QUADRANT POWER TILT RATIO is reduced to within its limit, or b) THERMAL POWER is reduced to less than 50% of RATED THERMAL POWER.
- 2. Reduce THERMAL POWER at least 3% from RATED THERMAL POWER for each 1% of indicated QUADRANT POWER TILT RATIO in excess of 1.0, within 30 minutes.
- 3. Verify that the QUADRANT POWER TILT RATIO is within its limit within 2 hours after exceeding the limit or reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within the next
/ 2 hours and reduce the Power Range Neutron Flux-High trip Setpoints to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours.
- 4. Identify and correct the cause of the out of limit condition prior to increasing THERMAL POWER; subsequent POWER OPERATION above 50% of RATED THERMAL POWER may proceed provided that the QUADRANT POWER TILT RATIO is verified within its limit at least once per hour for 12 hours or until verified acceptable at 95%
or greater RATED THERMAL POWER,
- c. With the QUADRANT POWER TILT RATIO determined to ex eed 1.09 due to causes other than the misalignment of either a shutdown control 4r hk) ;;rt I;..;th rod: Ir
- 1. Calculate the QUADRANT POWER TILT RATIO at least once per hour until either:
a) The QUADRANT POWER TILT RATIO is reduced to within its limit, or b) THERMAL POWER is reduced to less than 50% of RATED THERMAL POWER. it. W-STS 3/4 2-ya NOV 2 01980
.A _ _ . _ . _ . _ _ - _ _
POWER DISTRIBUTION LIMITS I ACTION: (Continued)
- 2. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 2 hours and reduce the Power Range Neutron Flux-High Trip Setpoints to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours.
l
- 3. Identify and correct the cause of the out of limit condition l prior to increasing THERMAL POWER; subsequent POWER OPERATION above 50% of RATED THERMAL POWER.may proceed provided that the QUADRANT POWER TILT RATIO is veriffcd within its limit at least once per hour for 12 hours or unti\ verified at 95% or greater RATED THERMAL POWER.
- d. The provisions of Specification 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.2.4.1 The QUADRANT POWER TILT RATIO shall be determined to be within the limit above 50% of RATED THERMAL POWER by:
- a. Calculating the ratio at least once per 7 days when the alarm is OPERABLE.
l
- b. Calculating the ratio at least once per 12 hours during steady state l operation when the alarm is inoperable.
4.2.4.2 The QUADRANT POWER TILT RATIO shall be determined to be within the limit when above 75 percent of RATED THERMAL POWER with one Power Range Channel inoperable by using the movable incore detectors to confirm that the ec -e"zad , O x tri; pr r di:tributi;n, :Strined *r:: th; t p ir; ef ;yr:tric thiath l Mc; tic.:, i :n;i;t;nt with the indicated QUADRANT POWER TILT RATIO at least once per 12 hours. by ei k . CL.M5I M Oc 4* P iirs of sm etric. %% Ale l,eg,y ,,.
^ @ b. (A sij We w o n ble i ncore. d ele <.A ab,i 3 y , te m 4, ,,,, ,4,,, _g c du AD eAur po w ee. T u:r % ri o sa6 3 M g g, reg y,,,,,,,+,,
l 04 St ecaft'ca.fth 3.L 3.2 13 l PSTS 3/4 2-14 SEP 101980
l l l l POWER DISTRIBUTION LIMITS 3/4.2.5 DN8 PARAMETERS l LIMITING CONDITION FOR OPERATION 3.2.5 The following DN8 related parameters shall be maintained within the limits shown on Table 3.2-1: '
- a. Reactor Coolant System T,yg.
! b. Pressurizer Pressure. APPLICABILITY: MODE 1. ACTION: With any of the above parameters exceeding its limit, restore the parameter to within its limit within 2 hours or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 4 hours. SURVEILLANCE REQUIREMENTS l 4.2.5 Each of the parameters of Table 3.2-1 shall be verif' fed to be within their limits at least once per 12 hours. l l i l l(
- 14 W-STS 3/4 2-15 NOV 2 01980 9
.. , < - _ - - - _ _ _ -. y
TABLE 3.2-1 s c. 21 DNB PARAMETERS
<n LIMITS 9 H-i uses; Ir Op:r:- -M-1 L;;;; In Op:: -
16 Loops In t!:n i L::; Step -t!:n i Le;p St:p PARAMETER Operation Value: 0; a Y:P;:: 01 ::d-kk) dIFY ' Reactor Coolant System T avg < M*F _f- (50^)"F -< (570)'F Pressurizer Pressure > (MEG)W _1 (2220) p;i;*- 1 (2220)* pda-Isor psig ca N b 7
%~9 t.
2
- ar
! **
- Limit not applicable during either a THERMAL POWER ramp in excess of (5%) of RATED THERMAL POWER per minute or a THERMAL POWER step in excess of (10)% of RATED THERMAL POWER.
GB E!
Section 3/4.3 In the text of Section 3/4.3 where a capital letter with a circle around it appears, please refer to this sheet for the appropriate justification. A. No three loop operation.
- B. Four loop operation only
'C. Seabrook specific information.
D. Action statement 6 applies to functions that have 3 " minimum channels operable". Action statement 7 applies to functions that have 2 " minimum channels operable". E. Not applicable to Seabrook. F. Reduced to major functional unit for clarity. G. Action statement 13 applies to functions that have 3 " minimum channels operable". Action statement 12 applies to functions that have 2 " minimum channels operable". Hl. Exemption under R.G. 8.12. Two area monitors located in area. Also See FSAR 9.4.2, 6.5.1, 15.7.4. H2. Alarm / Trip Setpoint. To prevent inadvertent trip when bridge is moved. H3. See R.G. 1.97 and NUREG-0737. This monitor reads photons only. H4. Not n'eeded in Mode 6 because of area monitor in 1.a H5. See RAI 460.35. I. All building discharges are routed to unit plant vent. Condenser exhaust system in Modes 1, 2, 3, 4'goes to unit plant vent. i
- J. To keep this table consistent with Table 3.3-6.
l K. No Waste Gas Holdup System or Hydrogen Monitoring System. L. Change per requirements of GL 83-37. M. See RAI 460.35. A e
, c m,
N. Due to deletion of non-applicable ACTION statements, numbers have been revised to maintain them in the correct order. O. Per discussion with SNUPPS, this change recommended by NRC. P. The Surveillance requirements are deleted from this Technical Specification. Public Service of New Hampshire will develop a Turbine 3 Overspeed Protection Reliability Program to administratively control this activity. This Program is currently under development and upon its completion will be submitted to the NRC for review along with proper justification, including a Probabilistic Assessment, as suggested by Section 3.1 of NUREG-1024. 5 O
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 OPERABLE with RESPONSE TIMES as shown in Table 3.3-2. APPLICABILITY: As shown in Table 3.3-1. ACTION: As shown in Table 3.3-1. I SURVEILLANCE REQUIREMENTS 4.3.1.1 Each reactor trip system instrumentation channel and interlock and the automatic trip logic shall be demonstrated OPERABLE by the performance of the reactor trip system instrumentation survaillance requirements specified in Table 4.3-1. . 4.3.1.2 The' REACTOR TRIP SYSTEM RESPONSE TIME of each reactor trip function shall be demonstrated to be within its limit 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 channel 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
" Total No. of Channels" column of Table 3.3-1.
1 l l l . l W-STS 3/4 3-1 l SEP,1.5 19'81 l . l t
TABLE 3.3-1
\$
- y. REACTOR TRIP SYSTEM INSTRUMENTATION MINIMUM l TOTAL NO. CHANNELS CHANNELS APPLICABLE i OF CHANNELS TO TRIP OPERABLE MODES ACTION .
FUNCTIONAL UNIT
- 1. Manual Reactor Trip 2 1 2 1 2 1 2 1 2 3 A, 4*, 5* TSID @
- 2. Power Range, Neutron Flux - High 4 2 3 1, 2 2 Setpoint ggy y Low 4 2 3 1 ,2 2 Setpoint
- 3. Power Range, Neutron Flux 4 2 3 1, 2 2 High Positive Rate t' 4. Power Range, Neutron Flux, 4 2 3 1, 2 2
# High Negative Rate
- 5. Intermediate Range, Neutron Flux 2 1 2 1 ,2 3
- 6. Source. Range, Neutron Flux gy 2 2 4 A. Startup 2 1 B. Shutdown 2 1 2 3*, 4*, 5* M ID @
C. Shutdown 2 0 1 3, 4, and 5 5 3 1 : 2. c, a
- 7. Overtemperature AT @ q 2 Moo, 'aan Plant
' Four Loop D ai;;- 4 2 3 1, 2 9 Three Loop Operati g 1 Three Loop Plant M B. Three Loop p peration 3 3 2 1** 2 2 A[ 1, 2 . , , l 3 o
TABLE 3.3-1 (Continued)
'T '
l REACTOR TRIP SYSTEM INSTRUMENTATION MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE OF CHANNELS TO TRIP OPERABLE MODES ACTION FUNCTIONAL UNIT
- 8. Overpower AT @ 4 1 3 1,1 4
! . Loop Plant ,
Four ion 4 2 3 1, 2 1** 9 g Three loop Operati 4 , B. Three Loop Plant 7, Three Loop Operat 3 2 2 T a on 3 , 1** 2 1, 2 .l # I
- 9. Pressurizer Pressure-Low el 2. 3 i t
- Teus Leep Plant.--- + - .
) t'
+
b 4. B_ Three Lcep Plant. &B- -. 4-- Pressurizer Pressure--High 2. 3 8 . 2- 4 # Y 10. s1
" 4. Ter Lcep P-lant-- + -1, 2 :
1 -S. Three-toop-PJant t F -1, 2
- 11. Pressurizer Water Level--High 3 2 2 1 7 l 12. Loss of Flow 7, A. Single Loop (Above P-8) 3/ loop 2/ loop in 2/ loop in 1 any oper- each oper-ating loop ating loop B. Two Loops (Above P-7 and 3/ loop 2/ loop in 2/ loop 1 7 below P-8) two oper- each oper-ating loops ating loop m
i M , t & I e
1 TABLE 3.3-1 (Continued) y REACTOR TRIP SYSTEM INSTRUMENTATION ' MINIMUM ; TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CilANNELS TO TRIP OPERABLE MODES ACTION
- 13. Steam Generator Water @ Y/sta. gen. 2/sta. gen. /sta. gen. 1, 2 [4 # @
Level--Low-Low in any oper- each oper-ating sta. ating sta. gen. gen. T <*aa= Generator Water Level - Low 2 stm. gen. 1 sta. gen. I stm. gen. 1, 2
~
team / level and level coin- level and
@ Coinciden Feedwater Flow Misma 2 sts/ feed- cident with 2 st ya/
flow mismatch low aismatch in tu. 1 sta./ flowj feed yhin same sim, ismatc gen. ame sta. gen. or 2 sta. R gen. en. level and tar / - T flow mismatch in same steam gen. pj E Undervoltage-Reactor Coolant y -2/6us. Pumps
/h. . 2 ,. ,a i er L A 6uses 6e A -Four ivup Pler,t - 4 1/uub ~3- +
- 4. T;i.ee Luup Pler.t +/tms- t ~2~
- h I M Underfrequency-Reactor Coolant 4-Mh, Pumps IMs **
5,n,s ,,, 2,= ou 6., g j#
^
f^"r l^^p Diant ' - l /bG- @ + --}"'- B. n rea Laap D!an* +/ims- & -} - v> Q /6 R Turbine Trip y i A. Low Fluid Oil Pressure 3- 2 2 1 7
- u. B. Turbine Stop Valve Closure 4 4 4 1 7#
G3 b l
/
. TABLE 3.3-1 (Continued)
'T g REACTOR TRIP SYSTEM INSTRUMENTATION MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF' CHANNELS TO TRIP OPERABLE MODES ACTION 17tfL. Safety Injection Input from ESF 2 1 2 1, 2 hO
%. Rea d u. C;;! -+ pnm_n Breaker _
4/ breaker i ove 8 ci - - su B; Above P-7 an reaker 2 eaker 1 11, y ating loop 1926. Reactor Trip System Interlocks R A. Intermediate Range
- Neutron Flux, P-6 2 l 1 2 2,, 8 Y
B. Low Power Reactor Trips Block, P-7 P-10 Input 4 2 3 1 8 or P-13 Input 2 1 2 1 8 C. Power Range Neutron ., Flux, P-8 4 2 3 1 8
$o
$11 Yo e
TABLE 3.3-1 (Continued) y REACTOR TRIP SYSTEM INSTRUMENTATION MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION D. Low Setpoint Power Range Neutron Flux, P-10 4 2 3 1, 2 8 E. Turbine Impulse Chamber Pressure, P-13 2 1 2 1 8 19 7 %. Reactor Trip Breakers 2 1 2 1, 2 4r 9 2 1 2 3*, 4*, 5* & 8o 24M. Automatic Trip Logic 2 1 2 1, 2 M9 2 1 2 3*, 4*, 5* &I0 ! U. Y m . o e-0
TABLE 3.3-1 (Continued) TABLE NOTATION With the reactor trip system breakers in the closed position, the control rod drive system capable of rod withdrawal.
@ !$ . E$.',5.55..!.h.32 7.'f._
.n .s uv 50005 ". s$._". f b) f_"!.,P'ry..
"h??'d_!_".,f."."'ti",
The provisions of Specification 3.0.4 are not applicable.
##B elow the P-6 (Intermediate Range Neutron Flux Interlock) setpoint. ###B elow the P-10 (Low Setpoint Power Range Neutron Flux Interlock) Setpoint.
ACTION STATEMENTS , ACTION 1 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours or be in HOT STANDBY within the next 6 hours. ACTION 2 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:
- a. The inoperable channel is placed in the tripped condition within 1 hour.
- b. The Minimum Channels OPERABLE requirement is met; however, the inoperable channel may be bypassed for up to 2 hours for surveillance testing of other channels per Specification 4.3.1.1.
- c. Either, THERMAL POWER is restricted to less than or equal
- to 75% o.f RATED THERMAL POWER and the Power Range Neutron Flux trip setpoint is reduced to less than or equal to (85)% of RATED THERMAL POWER within 4 hours; or, the QUADRANT POWER TILT RATIO is monitored at least once per 12 hours per Specification 4.2.4.2.
l l i W-STS 3/4 3-7 SEP 151981 l
TABLE 3.3-1 (Continued) ACTICN STATEMENTS (Continued) ACTION 3 - With the number of channels OPERABLE one less than the Minimum Channels OPERABLE requirement and with the THERMAL POWER level:
- a. Below the P-6 (Intermediate Range Neutron Flux Interlock) setpoint, restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above the P-6 Setpoint.
- b. Above the P-6 (Intermediate Range Neutron Flux Interlock) setpoint but below 10 percent of RATED THERMAL POWER, restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above 10 percent of RATED THERMAL POWER.
ACTION 4 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement suspend all operations involving positive reactivity changes. ACTION 5 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, verify compliance with the SHUTDOWN MARGIN requirements of Specif.ication 3.1.1.1 or 3.1.1.2, as applicable, within 1 hour and at least once per 12 hours thereafter. ACTION 6 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP ad/or POWER OPERATION may proceed provided the following conditions are satisfied:
- a. The inoperable channel is placed in the tripped condition within 1 hour.
- b. The Minimum Channels OPERABLE requirement is met; however, the inoperable channel may be bypassed for up to 2 hours for surveillance testing of other channels per Specification 4.3.1.1.
( ACTION 7 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed l until performance of the next required OPERATIONAL TEST provided the inoperable channel is placed in the tripped condition within 1 hour. ACTION 8 - With less than the Minimum Number of Channels OPERABLE, within one hour determine by observation of the associated permissive annunciator window (s) that the interlock is in its required state for the existing plant condition, or apply Specification 3.0.3. i I I W-STS 3/4 3-8 SEP 151981
TABLE 3.3-1 (Continued) ACTION STATEMENTS (Continued)
- With a channel associated with an operating loop inopernhln ,
e the inoperable channel to OPERAR' c ;toLui~Within 2 hours or be in '
-et HOT S innin the next 6 hours. One channel ass i erating loop may be bypassed for u ours for surveillance teau;. 3 p:r S"erification 4.3.1.1.
N OE. m. iici, l W he number of OPERABLE Channels one less than the Minim e uirement, restore the inoper l Channe s s G70.".^ ' nel to OPERABLE status wi s or L POWER to below the P-8 (Power Ran lock) setpoint within the n rs. Operation below the P-u se6vu ^* ==y, c ursuant to ACTION 11. nCTIO:: ' - with the number of OPERABLE Channels one less than tha u 4 _.;7 - Channels OPERAutt icyu:F;rra+ ona"'ti;.. may continue provided the inoperable eb--..'. is placed in the trippsd-enndition w u r. -- ACTION R -)9With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, be in at least HOT STANDBY within 6 hours; however, one channel may be bypassed for up to 2 hours for surveillance testing per Specification 4.3.1.1, provided the other channel is OPERABLE. 10 ACTION 14 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPEARABLE status within 48 hours or open the reactor trip breakers within the next hour. l l W-STS 3/4 3-9 SEP 151981
I TABLE 3.3-2
'T -
REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES "l RESPONSE TIME i FUNCTIONAL UNIT
- 1. Manual Reactor Trip Not Applicable
- 2. Power Range, Neutron Flux 1 (0.5) seconds *
- 3. Power Range, Neutron Flux, High Positive Rate Not Applicable 1
- 4. Power Range, Neutron Flux, jv,t ahal'eakIe High Negative Rate 1(ES)-trendr*6!)
- 5. Intermediate Range, Neutron Flux Not Applicable ti 6. Source Range', Neutron F' lux Not Applicable
+
- Y 7. Overtemperature AT $ (4.0) seconds * ,
U$
- 8. Overpower AT Not Applicable
- 9. Pressurizer Pressure--Low 1 (2.0) seconds
- 10. Pressurizer Pressure--High 1 (2.0) seconds
- 11. Pressurizer Water Level--High Not Applicable l
i
=
ui rn Neutron detectors are exempt from response time testing. Response time of the neutron flux signal portion
of the channel shall be measured from detector output or input of first electronic component in channel.
[ fThis prevision is net applicsle-to-EP%ecketed-after-January-A 1978r-See Reguiotcry Guide i.ii0,
" ::;bcr 1477 )
!, - U_S.
/
, -s TABLE 3.3-2 (Continued)
{j REACTOR TRIP SYSTEM' INSTRUMENTATION RESPONSE TIMES FUNCTIONAL UNIT RESPONSE TIME
- 12. Loss of Flow A. Single Loop (Above P-8) 1 (1.0) seconds B. Two Loops (Above P-7 and below P-8) 1 (1.0) seconds
- 13. Steam Generator Water Level--Low-Low 1 (2.0) seconds
- 14. Sice; Os..m. ;t r 'date; L : !-Lt. C;lr. cider.t ith
@ E+e=-/roaameter r!:; ,",;,_;tch "o; App;j ;tle infer 14T5. Undervoltage-Reactor Coolant Pumps < 64-6-) seconds
**151E. Underfrequency-Reactor Coolant Pumps < (0.6) seconds T
LI I4TS. Turbine Trip A. Low Fluid Oil Pressure Not Applicable B. Turbine Stop Valve Not Applicable 17 TS. Safety Injection Input from ESF Not Applicable ($);3. Rcomie. C;;lar.t ra; p 0 moke, rea . i;v., T. .y Het f.pp!!::Sl;
- 13 29. Reactor Trip System Interlocks Not Applicable 19244 Reactor Trip Breakers Not Applicable w .10% Automatic Trip Logic Not Applicable m
"U e--
M VD
5 o
~
TABLE 4.3-1 !
'T REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS v
TRIP r ANALOG ACTUATING MODES FOR CHANNEL DEVICE WlICH CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST IS REQUIRE 0
- 1. Manua'l Reactor Trip N.A. N.A. N.A. R N.A. 1, 2, 3*, 4*, S*
- 2. Power Range, Neutron Flux High Setpoint S(9) D(2,4), M N.A. N.A. 1, 2 M(3, 4),
Q(4, 6), R(4,5) low Setpoint S(9) R(4) M N.A. N.A. 1,,,, 2 R 3.. Power Range, Neutron Flux, N.A. R(4) M N.A. N.A. 1, 2
- liigh Positive Rate w
h 4. Power Range, Neutron Flux, N.A. R(4) M N.A. N.A. 1, 2 liigh Negative Rate S. Intermediate Range, S(9) R(4,5) S/U(1),M N.A. N.A. 1 ,2 Neutron Flux
- 6. Source Range, Neutron Flux S(9) R(4,5) S/U(1),M(9) N.A. N.A. 2 , 3, 4, S
- 7. Overtemperature AT S R H N.A. N.A. 1, 2
- 8. Overpower AT S R H N.A. N.A. 1, 2
- 9. Pressurizer Pressure--Low S R M N.A. N.A. 1
- 10. Pressurizer Pressure--High S R H N.A. N.A. 1, 2 m
o m
" 11. Pressurizer Water Level--High 5 R H N.A. N.A. 1
- 12. Loss Of Flow 5 R H N.A. N.A. 1
]
e e
l TABLE 4.3-1 (Continued) It J. REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS - d TRIP ANALOG ACTUATING MODES FOR CHANNEL DEVICE WHICH CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST IS REQUIRE 0
- 13. Steam Generator Water Level-- S R H N.A. N.A. 1, 2 Low-Low I". Sica. Gwiciaio. U;ter L:::!---
Lcw Caineide..i with Sic;;/
-S-- -R- .
-M- ' -N-A- -- -kA-- W
@ Tecdaetcr ." low Mio-atch tyli. Undervoltage - Reactor Coolant N.A. R N.A. M N.A. 1 Pumps btyT6L Underfrequency - Reactor N.A. R N.A. H N.A. I u, Coolant Pumps '
0
/4R Turbine Trip A. Low Fluid Oil Pressure N.A. N.A. N.A. S/U(1,10) N.A. 1 B. Turbine Stop Valve N.A. N.A. N.A. S/U(1,10) N.A. 1 Closure il Safety Injection Input from N.A. N.A. N.A. R N.A. 1, 2 ESF U. .Lm;r Cec!:d a" ;' beder -N-A. -N-A- M^ --R--- #-A- -
Pc;itica '-ip hl 21L Reactor Trip System Interlocks lQ A. Intermediate Range 's Neutron Flux, P-6 N.A. R(4) M N.A. N.A. 2,, ['$ B. Low Power Reactor cg Trips Block, P-7 N.A. R(4) M (8) N.A. N.A. 1 = C. Power Range Neutron Flux, P-8 N.A. R(4) H (8) N.A. N.A. 1
TABLE 4.3-1 (Continued)
'T g REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS
- TRIP ANALOG ACTUATING MODES FOR CHANNEL DEVICE %4tICH CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST IS REQUIRED D. Low Setpoint Power Range Neutron Flux, P-10 N.A. R(4) M (8) N.A. N.A. 1, 2 E. Turbine Impulse Chamber Pressure, P-13 N.A. R M (8) N.A. N.A. 1 M R. Reactor Trip Breaker N.A. N.A. N.A. M (7) N.A. 1, 2, 3* , 4* , 5*
102fL Automatic Trip Logic H.A. N.A. N.A. N.A. M (7) 1, 2, 3*, 4*, 5* R
+
Y
~ >=
U i t
- l TABLE 4.3-1 (continued)
TABLE NOTATION With the reactor trip system breakers closed and the control rod drive system capable of rod withdrawal. N - Below P-6 (Intermediate Range Neutron Flux Interlock) setpoint.
#N -
Below P-10 (Low Setpoint Power Range Neutron Flux Interlock) setpoint. (1) - If not performed in previous 7 days. (2) - Heat balance only, above 15% of RATED THERMAL POWER. Adjustchannel i if absolute difference greater than 2 percent. (3) - Compare incore to excore axial flux difference above 15% of RATED THERMAL POWER. Recalibrate if the absolute difference is greater than or equal to (3) percent. , (4) - Neutron detectors r.ay be excluded from CHANNEL CALIBRATION. (5) - Detector plateau curves shall be obtained and evaluated. For the Intermediate Range and Power Range Neutron Flux Channels the provisions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1. (6) - Incore - Excore Calibration. (7) - Each train shall be tested at least every 62 days on a STAGGERED TEST BASIS. (8) - With power greater than or equal to the interlock setpoint the required i OPERATIONAL TEST shall consist of verifying that the interlock is in
. the re' quired state by observing the permissive annunciator window.
(9) - Monthly Surveillance in MODES 3*, 4* and 5* shall also include verification that permissives P-6 and P-10 are in their required state for existing plant conditions by observation of the permissive annunciator window. (10) - Setpoint verification is not applicable. l W-STS 3/4 3-15 NGV f 1931
INSTRUMENTATION l 3/4.3.2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION u!MITING CONDITION FOR CPERATION 3.3.2 The Engineered Safety Features Actuation System (ESFAS) in strumentation channels ar.d interlocks shown in Table 3.3-3 shall be OPERABLE with their i Tr.ip Setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3-4 and with RESPONSE TIMES as shown in Table 3.3-5. ADPLICABILITY: As shown in Table 3.3-3 ACTION:
- a. With an ESFAS Instrumentation or Interlock Trip Setpoint trip less conservative than the value shown in the Trip Setpoint eclumn of Table 3.3-4, adjust the Setpoint consistent with the Trip Setpoint
) value.
- b. With an ESFAS Instrumentation or Interlock Trip Setpoint less conservative than the value shown in the Al'cwable Values Column of (j{} +ne Table t
3.3-4, p :rr
**e c":- :1 '..m .. .rr;: ::ntiti;. ;e.... In;,
within t"e ':!!: ' ; 12 hours either:
- 1. Determine that Ecuation 2.2-1 was satisfied for the affected channel and adjust the Se point consister.: with the Trip Setpoint va'.ue of Table 3.3-4, or
- 2. Declare the channel inoperable and apply the applicable ACTION statement requirements of Table 3.3.3 until the cnannel is restored to OPERABLE status with its Setpq, int adjustec consistent with the Trip Setpoint value.
I ! Equation 2.2-1 2 + R + 5 < TA Where: 2 = the value from Column Z of Table 3.3-4 for the affected channel, R = the "as measured" value (in percent span) of rack error for the affected channel, S = either the "as measured" value (in percent span) of the sensor error, or the value for Column S (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 affected channel. W-STS 3/4 3-16 e
- NSTRUMENTATION
/
3 /4. 3 INSTRUMENTAT*CN S'JRVEILLANCE RECUIREMENTS 4.3.2.1 Each ESFAS instrumentation channel and interlock and the automatic actuation logic and relays shall be demonstrated OPERABLE by performance of the Engineered Safety Features Actuation System Instrumentation Surveillance Requirements specified in Table 4.3-2. 4.3.2.2 The ENGINEERED SAFETY FEATURES RESPONSE TIME of each ESFAS function shall be demonstrated to be within the limit at least once per 18 months. Each test shall incluce at least one train such that both trains are tested at least once per 36 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 cf reduncant channels in a specific ESFAS function as shown in the " Total No. of Channels" Column of Table 3.3-3. 9 t I l w-STS 3/4 3-17 e re y .e
... y
..o.
g eggem,, em es =e pS * *""NW
l TABLE 3.3-3 If i y ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION
- 1. SAFETY INJECTION,-lEAGMR
!"I", cernuaire ic<u ,;:cg,
@ = CONTROL =t cE ROOM - ISOLATION, START-eest-ING TANS AN; Es5ENTIAL-C E Du f f* f" tfATrn
- a. Manual Initiation 2 1 2 1, 2, 3, 4 -
Mir
- b. Automatic Actuation 2 1 2 1,2,3,4 M li R*
Logic and Actuation Relays i T
!$5 c. Containment 3 2 2 1,2,3 Ni*11 h
Pressure-High
- d. Pressurizer 4 2 3 1,2,3 # M*l6 Pressure - Low
- e. M.M.
o .%.L[m.
.. ?. i . . .P 55 u e -W -
j 54* " I' . 1, 2, 3 12 *
"ccssur::_Bettera 3ld.'"a taa*
2./sfeaa liac
- u. u.u - e : d<=a liae
@c.__. . ... - .
o Plant 4 Four Loops 3se 2/ steam line 2/ steam line Operating a line us @ ### Q Three Loops 3/ operating /aeam 2/ opera 16
- Operating st line any steam line o' operating go steam line
TABLE 3.3-3 (Centinuid)
, 'T -
~
g ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CilANNELS .T0 TRIP OPERABLE ' MODES CTION e SAFETY INJECTION, REACT 0 RIP, FEEDWATER ISOLATION CONTROL 00M ISOLATION, START DIESEL GENER CONTAINMENT COOLING FANS ANO RS ESSENTIAL SERVICE WATER (Continued)
- ii) Three Loop Plant Three Loops 3/ steam li 2/ steam line 2/st de line IS*
Operating twice and 1/3 s am lines
}
/st f- Two Loops 3/ operating 2 2/ operating 16 O Operating steam line line tyice steam line i in either
. ope'ratirig team line 1, 2, 3 0 , f. Steam Flow in Two l Steam. Lines-High l i) Four Loop Plant ! Four Loops 2/ steam line 1/ steam line 1/steat line 15* Operating any 2 steam lines - l h Three oops ting 2/ operating steam-line 1 /any operating 1/ operating steam line 16 Op l en steam line e
TABLE 3.3-3 (Continued) y ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION SAFETY INJECTION, REACTOR P, FEEDWATER ISOLATION, CONTROL ISOLATION, START DIESEL GENERAT CONTAINMENT COOLING FANS AND ESSE L SERVICE WATER (Continued) ii) Three Loop Plant Three Loops 2/ steam line 1/ steam line 1 eam line 15* y Operating ny 2 steam , li s
)
Two Loops 2/ operating I E a 1/ operating 16 f,
!) Operating steam line eratin steam line I steam line Coincident With Either T,yg-Low-Low 1,2,3 N i). Four Loop Plant Four Loops 1 T,yg any 15*
1 T,yg/ loop 1 T,yg any Operatin 2 loops 3 loops ' Thr oops 1 T,yg/ 1 E T,yg in 1 T,yg in any 16 y operating any operating two operating o loop loop loops M
TABLE 3.3-3 (Continued) , y ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE H00ES ACTION SAFETY INJECTION, REACTOR TR FEEDWATER ISOLATION, CONTROL R0 ISOLATION, START DIESEL GENERATOR . CONTAINHENT COOLING FANS AND ESSENTIAL SERVICE WATER (Continued) li) Three Loop Plant any 15* Three Loops 1 T,yg/ loop 1 T,yg any avg Operad ng 2 loops 21 ps
### in any
$> Two Loops 1 T"#9/ 1
"#9 in 1T
#U 16 Operating operating loop iy operati operating loop loop Or, Coincident With Steam Line Pressure-Low 1, 2, 3 i) Four Loop Plant Four Loop 1 pressure / 1 pressure 1 pressure 15*
@ Operatitj loop any 2 loops any 3 loops T ree Loops 1 pressure / l### pressure i pressure 16 perating operating in any oper- in any 2 w loop ating loop operating loops 5
O e
v TABLE 3.3-3 (Continued) y ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION SAFETY INJECTION -;CACTOR TRIP,
-TEE'". ATER ISCLATIO::, C0;4 TROL ROOM
- -ISOLAliSN, 5 TART DIC5EL GENERATORS
-C004 TAI:"E!T COGLihG FAN 5 AND ESSE!!TIAL SE'".'ICC WATER (Centinued)-
- 88) T ree Loop Plant Three Loops e/ 1 pressure 1 pressure ir Operating loop oops any 2 loons --
w 7
) Two Loops 1 pressure
# #~ #
pressure 1 press 16 T Operating in any oper- any operating loop
% ating loop
- 2. CONTAINHENT SPRAY
, a. Manual . tmh'aW 2 1 with 2 1,2,3,4 Mir 2 coincident switches
- b. Automatic Actuation 2 1 2 '1, 2, 3, 4 M 11 Logic and Actuation @
Relays
; c. Containment Pressure-- 4 2 3 1,2,3 W13 g "igh "igh Hi -3
, u>
Q 3. CONTAINMENT ISOLATION
; a. Phase "A" Isolation ro 1) Hanual Tadinhbri 2 1 2 1,2,3,4 M I 5'
- 2) Safety Injection See 1 above for all Safety Injection initiating functions and requirements i
TABLE 3.3-3 (Continued) ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION !
, {
- MINIMUM TOTAL NO. CllANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION CONTAINMENT ISOLATION (continued)
- 3) Automatic Actuation 2 1 2 1,2,3,4 M ti Logic and Actuation Relays
- b. Phase "B" Isolation
- 1) Manual 2 1.with 2 1,2,3,4 1915~
2 coincident g switches . T
- 2) Automatic Actuation Logic and Actuation 2 1 2 1,2,3,4 M 11 g
$ Relays
- 3) Containment 4 2 3 1,2,3 1% 13 Pressure- "!H+i 4"igh
- c. Purge and Exhaust Isolation
@ 1) Automatic Actuation 2 1 2 1,2,3,4 M l'f Logic and Actuation Relays
- 2) Containment OnlicPuey 4 1 1, 2, 3, 4 M I'f Radioactivity-High w 3) Safety Injection See 1 above for all Safety Injection initiating functions and Q requirements M
I
TABLE 3.3-3 (Continu2d) y ENGINEERED SAF8TY FEATURE ACTUATION SYSTEM INSTRUMENTATION MINIMUM i TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION
- 4. STEAM LINE ISOLATION
- 2. 2.
@ a. Manual .In' Q 4 m Vsteam line 1/ steam line Voperating 1,2,3 24 Z0
- steam line
. b. Automatic Actuation 2 1 2 1,2,3 29 12 Logic and Actuation Relays ; c. Containment Pressure-- 4 2 3 1,2,3 Tit 13 @ ~
li!;;h uigh Hi-2. R eE M, ijb._5t.s.am 8 'ia* P"55mm Rdc 3/sb li,e.2.jsb li.e
= ; ^ h?e '
-1 ;-2, 3, 9 314
+ .
g,f,A /' ** S te e.T. Lj.c;-"igh
; y gg g d- 5fea Lsae Presp*pw 3 l s6 tia*
==1 s6 li.4 2fsb ha' 12
- l 2./s6 li.e Ii li3
! any d e lu.is Four Loops 2/stea 1/ steam line 1/ steam line . Operating an lines Three Loops ng 1 #/any 1/ operating Operat' steam line operating steam line steam line
%) Three Leep Pier.t ,
" m Loops 2/ steam line 1/ steam line 1/ steam line 'c*-
Operating any 2 steam ,
,_. 'e_
"_, Two Loops /any 1/ operating a ing 1 _
operating steam line
~
$ Oper steam line
~
steam line
. )
TABLE 3.3-3 (Continued) y ENGINEERED SAFETY FEATURE' ACTUATION SYSTEM INSTRUMENTATION MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES TION STEAM LINE ISOLATION (Continu Coincident With Either T,yg--Low-Low 1 ,3 i) Four Loop Plant , Four Loops 1 T,yg/ loop - 1T any 1 T, ny 15* Operating 2h,ygps 3 ops Three Loops 1 T,yg/oper- 1 T,y 1T avg in any 16 6, Operating ating loop any op tin two operating g loo ops i ii) Three Loop Plant Three Loops 1 T,yg/l p 1 T,yg any 1 T,yg an 15*
@ Operating 2 loops 2 loops Two loops T,yg/oper- l ### T avg 1 T,yg in any 16 Operating ating loop in any oper- operating loop ating loop M
e-*
. u. ,
in 1 4
sac TABLE 3.3-3 (Continued) . S y ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION MINIMUM
. TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT ,
0F CHANNELS TO TRIP OPERABLE MODES ACTION AM LINE ISOLATION (Continued)
, oincid,ent With Steam ' e Pressure-Low 1, 2, 3 i), Four Loop nt Four Loops ressure/ 1 pressure 1 pre e 15*
l @ Operating loop any 2 loops loops w Three Loops 1 pressure / pre e 1 pressure in 16
'A Operating operating loop in er- any 2 oper-w ing loop ating loops ii) Three Loop Plant Three Loops pressure / 1 pressure 1 pressure 1S*
h Operating Two Lo loop 1 pressure / any 2 loops any 2 loops l# pressure 1 pressure Op ing operating in any oper- any operating loop ating loop loop
- 5. TURBINE TRIP & I FEE 0 WATER ISOLATION
- a. Steam Generator /stm. gen. 2/stm. gen. /stm. gen. 1, 2 N* h Water Level--
High-High in any oper-ating stm gem. in each oper-ating stm. gen. g IF
,_ b. Autoraatic Actuation 2 1 2 1, 2, 22 l ce Logic and Actuation is Relay
O & - _ g s p u
- a .
4
- D, i,
d 89 a a Y af "
- +
5 t
< d 2
e 5
= .
W hyw
- g gs* .o
? *01
- k
=
f*$2k
-g, .4. 3 e* o y e
z 52f *rJ n-< g g 8 C zuo 4 5 z m c 5 e 4* - a p Sm 4 74 7 I 5 $ $5 4 FT d t 4 g". A g w m. u- <n-pE r m W a 5 w o$ e a 5 k ' I .$ e d@ EW & _ I
.b #$ j o -e a
m av
~g 4, r-
+e S
5 W
- T, 8
w 3 . a , me g t b 3 'it
$s E
~
8k H b Ur.f l g w e& Ym# - r
= kI kd )
Q S. 3 es 9 )f d
- 2 25 m g
w o $ . 7 n a w e W-STS d 3/4 3-A3
TABLE 3.3_-3 (Centinued) y ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION i MINIMUM l' TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION Et16%EuC1 TE. AUM MARY FEEDWATER o rianet I4tMsa,g i it.hr Dei enPw k I h 1, 2, 3 MM Al Mea = Det ses Paw 2. I ~2., Ii 2.i 3 11
- b. Automatic Actuation Logic 2 1 2 1,2,3 22 IF and Actuation Relays
- c. Sim. Gen. Water Level-Low-Low w
1 Start Motor- y w 1. Driven Pumpgand hsta. gen. 2/sta gen. 1/sta. gen. 1, 2, 3 h* b 13 y' p h 6 e p rs~v e s P a p in any opera- in each . r'4 ting sta gen. operating stm. gen. bine- ~ Driven Pump tm. gen. 2/stm. gen. 2/stm. gen ' 1, 2, 3 ', 7 n anu ' 2 operaung :;^-= tina stm. gen. stm. gen
-d . Unde, vul1 ;; -PIP -
Stzrt Turbi"e-Dr i Jer Pr p J-1/be; --e- - & 1, 2 - f98-- cn d% Safety Injection ~ Q Start Motor-Driven Pump % . - and Turbine-Dr.iven Pump See 1 above for all Safety Injection initiating functions and en requirements ro E B Y. Station Blackout Start Motor-Driven Puma and Turbine-Driven Pump 2 1 2 1, 2, 3 19
TABLE 3.3-3 (Continued) - 4 i h ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION 1 MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION i
"'"'!'.!".7,Y
. T:C^/ ;.TER 6....t ' n;;d)
- m. Trip of Main i F eeuwa i... P" 7e Start Motor- %-
I Driven Pumps and Turbine-Drivaa a -- z/ pump 1/ pump 42 l . 1/ pump 19 i -
$% AUTOMATIC SWITCHOVER TO w CONTAINMENT SUMP 5 b w a. RWST Level - Low 4 2 3 1,2,3,4 W 13
, Coincident With Ccate'n.T.;nt L...p Level .9igh - -E- - + 1, 2, 3, i M Safety Injection See 1 above for Safety Injection initiating functions and requirements-1 i
- b. Automatic Actuation 2 1 2 1,2,3,4 M ll '
Logic and Actuation Relays ui
$ TIL LOSS OF POWER
- 2. I l&
m a. 4 kv Bus 4/ Bus 2/ Bus VBus 1, 2, 3, 4 288 gg
-U Loss of Voltage 4 kv 2. t f (,
- b. Gt4d-Degraded Voltage 4/ Bus 2/ Bus 'S/ Bus 1, 2, 3, 4 '29
- TABLE 3.3-3 (Continued) y ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION, MINIMUM' TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION 10k ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INTERLOCKS Pressurizer Pressure, 3 2 2 1,2,3 hM h . a. P-ll
- 4. Lu. La.; T avg' ~12 ~4'- 2 ~3-~ -I' 2' 3 ~EI-' @
- c. Reactor Trip, P-4 2 2 2 1,2,3 M11 R
+
Y NU e
-o e-U9 9
1 l l TABLE 3.3-3 (Continued) TABLE NOTATION
#Trip function may be blocked in this MODE below the P-ll (Pressurizer Pressure Interlock) setpoint.
##Trip function may be blocked '.i this MODE below the P-12 (Low-Low T Interlock) setpoint. avg The che.....i(s) ...wc Led w i t;. the protecti;c fu~ '.t cr.; de i ved f rc. . the est of ervico seactor Coolant cuup ;h;?' be p!:::d ir, the tripped ;;de.
*The provisions of Specification 3.0.4 are not applicable.
l ACTION STATEMENTS ACTION With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, be in at least HOT STANDBY within 6 hours and in COLD SHUTDOWN within the following 30 hours; however, one channel may be bypassed for up to 2 hour for surveillance testing per Specification 4.3.2.1, provided the other channel is OPERABLE. 11 ACTION hi With the number of OPERABLE channels one less than the Total Number of Channels, operation may proceed until performance of the next required OPERATIONAL TEST provided the ino'erable p channel is placed in the tripped condition within 1 hour. With a channel associated with an operating loop inope , he inoperable channel to OPERABLE st hin 2 hours or east HOT S'ANDBY T the next 6 hours and in at least HOT SH ~ llowing 6 hours. One channel associa n operating loop may for u or surveillance testing per Specification 4. . . . . ACTION With the number of OPERABLE channels one less than the Total Number of Channels, operation may proceed provided the inoperable ' l channel is placed in the bypassed condition and the Minimum Channels OPERABLE requirement is met. One additional channel may be bypassed for up to 2 hours for surveillance testing per Specification 4.3.2.1. p-ACTION h - With .less than the Minimum Channels OPERABLE requirement, operation may continue provided the centainment purge supply and exhaust valves are maintained closed. l W-STS 3/4 3- SEP i 5198k
I TABLE 3.3-3 (Continued) ACTION STATEMENTS (Continued) If ACTION TG - With the number of OPERA 8LE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours or be .in at least HOT STANOBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. . Ir ' ACTIONDI-WiththenumberofOPERABLEchannelsonelessthantheTotal Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:
- a. The inoperable channel is placed in the tripped condition within 1 hour.
- b. The Minimum Channels OPERABLE requirements is met; however, one additional channel may be bypassed for up to 2 hours
- for surveillance testing of other channels per Specification i 4.3.2.1.
ACTION 4 - With less than the Minimum Number of Channels OPERABLE, within one hour determine by observation of the associated permissive annunciator window (s) that the interlock is in its required state for the existing plant condition, or apply Specification 3.O.3. _ 19 l ACTION DS - With the number of OPERABLE Channels one less than the Minimum Channels OPERABLE requirement, be in at least HOT STANDBY l l 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 other channel is OPERABLE. ACTION With the number of OPERABLE channels one less than the Total Number of Channels, restore the inoperable channel to OPERABLE status within 48 hours or be in at least HOT STANDBY within 6 hours and in at least HOT SHUTDOWN within the following 6 hours. I 20 ACTION 24 With the number of OPERABLE channels one less than the Total Number of Channels, restore the. inoperable channel to OPERABLE ~ status within 48 hours or declare the associated valve inoperable and take the ACTION required by Specification (3.7.1.5). l Af SEP 151981 W-STS 3/4 3-)Q
~ - , - - - -
C IAlllE 3.3-4 lCe Q en ENGINEERED sal'ETY flAIURLS AClllAll0N SYSIIH INSiltilMENIATION TRIP SEIPOINTS StN50R 10lAL [ Rit0R - FUNCTIONAL UNIT At10WANCE (IA} 7 (S) TRIP SEIP0lNT ALLOWADLE VAlUE
- 1. Safety Injection (i c ts; Tr:p,-
F::i::tcr h:!: tim, Cutre4--
"::- h;!-ti r., Si r t O!i;;;
C:....oiu ;, C::i ! :::t C^e!8ng r m , ...J E..c..;..u! Cercic: " ' '_ e - } '
- a. Manual Initiation N.A. N.A. N.A. N.A. N.A.
m b. Automatic Actuation Logic N.A. fl. A. N.A. N.A. N.A. 1 ai .3 later
, c. Containment Pressure-liigh 1 (3.0) (0.71) (1.5) $ (-hfr) psig 1 (M) psig N$ d. Pressurizer Pressure--Low (13.1) (10.71) (1.5) 1 (1850) psig 1 (1839) psig
-e. C ! l Tc. c..i!.; T. c o au. < - (44) ( " " ? )- H I ("?) ;'t 5-i Be*.:: r St 2'8ne: '. . u; >
-1 CCC) r !
3-'4-) iM 2. SF S- 14fer { 64 Steamline Pressure--Low (iH-e) (10 71) (1.5) $ (67&) psig 1 (698)r psig*
- 2. Containment Spray
- a. Manual Initiation H.A. II. A. N.A. N.A. N.A. .
- b. Automatic Actuation Logic fl. A. H.A. H.A. N.A. N.A.
and Actuation Relays y.o 3 y.o 81 I
- c. Containment Pressure-liigh-3 (+-ft) (0.71) (1.5) $ (it-06) psig 5 (+0--44 ) p s i g I
- l l
f I
'T TAHIE 3.3-4 (Continue:1}
- ENGINEERED SAFETY EEAltlRES ACIUATION SYSll'M INSTRUMENTATION TRIP SETPOINTS SENSOR 10TAl iRROR FUNCTIONAL UNIT Att0WANCE (IAJ Z_ (S) TRIP SETPOINT All.0WA81.E VALUF
. 3. Containment Isolation l*e a.
Phase "A" Isolation
.i
- 1) Manual Initiation N.A. N.A. N.A. N.A.
I' N.A.
- 34) Automatic Actuation Logic N.A. N.A. N. A.
and Actuation Relays N.A. N.A. w 2.1) Safety injection k See Item 1. above for all Safety Injection Setpoints and Allowable Values.
, b. Phase "B" Isolation .
- 1) Manual. Initiation N.A. N.A. N.A. N.A.
N.A.
- 2) Autom,atic Actuation N.A. N.A. N.A.
Logic and Actuation H.A. N.A. Relays - 4.o
- 3) Containment Pressure- (4-0) 13 0 11.t High-3 (0.71) (1.5) $ (4iH &) psig 5 (14.-34) psig I
i
- c. Purge and Exhaust Isolation i i) Me;..;al P ! t bib; M-A-l #-A-- 4-A-- 44-A,-
-H-A--
t 1) Automatic Actuation ! N.A. N.A. N.A. N.A. Logic and Actuation N.A. Relayt j 3) Safety Injection See Item 1. above for all Safety Injection Setpoints and Allowable ..
- 2) baba *dd On b8C leder Pege. Mte citvd y- gy g,. 4 3 g&
%*" l H t%I t i
I , 39 { TADIE 3.3-4 (Coni.Inuedl h ENGINIIRLD SAff 1Y FEAlllRf 5 ACillAlION SYSilH INSiltill11NIAl10N IRIP SEIPolNIS StH50R 10lAL iRROR FUNCTIONAL IINIT Al l.0WANCI (IA) / _ _ ( '> ) _ IRIP SEIPolNI ALLOWAllli VAllll;
- 4. Steam Line Isolation
- a. Manual Initiation N. A. H.A. N.A. N.A. N.A.
- b. Automatic Actuation Logic N.A. fl. A. N.A. N.A. N.A. *
- and Actuation Relays
- c. Containment Pressure-liigh-2 (N) 0.71 1.5 $ N i
, t. W am Flow in Two Steamlines- (20.0) (13.16) (1.5/ 5 A function i A function defi R iligh, o with 1.5) defined as as follows: p
- fo11cws: A corre ing to 44%
y AP corres- ull steam flow g"
' te ponding to between 0% and loatl
- 40% i1 and then a Ap cam flow increasing linearly
. > 0% to a op correspondint and 20% lo to 114.0% of full and then a s flow at full I
, Ap increasing load.
i linearly to 110% of full load. i i T
- avg
- L, Lu - M t 1. ;G ( L.2) j t S 2 ) ." - $ 5 5^. C )"'
d .e: Steam line Pressure - 1.ow (N) (10.71) (1.5) 1( k psig $ psig a Hs ioo ter-c.r. Lu qlt. . c Steam line Pressure (8.0) (0.5) (0) 5(44&) psihee 1 ( . ) ps i hre'^ Rate lli p. - e l 1
! l .. ! l
4-T-h-te2M "[}h t
= ct l u 6 8' -k g e 4 e e n.T* a e ig sie vi fs at # f la 2 = -
2 x E - l 5 E
*
- vi
- r#
+'5 ik 't I*
C g ,s + < e g l (*< >t 4 +* f +5 m w w q p
= nsE R $ 1I h !
6)4 a e
- f y, ; ;
aa ai s s 28s a E = = m' e g _I 5E0
== i h
*t %
+
6 4 q 4 2 4 2
-6 c m1 -.s i ' 5; '
% B 4 t b g 4
- g E G d
- i a 2 2 -
A
, 5 2 ~
h E, S I di 5 8 i * < E G WE 9 h ' 4 4 9 m az w c -4 e a, 4 t E l a w l b i :>- i - e E l 1 e [ .o .$
= Y. c- 8 i,8 i I
r, I.3 >- e t 31 5 _3. .V -s& 3 E -S 3 c ax
=
2 eo a . o w l - , 32 c . . .v.p . m , h O T 4 da dE
=*= f
.k.
<.*-+-
4&J w %, u. o s I i . 6 4 s s v.E e .-t'- 4 .o4 v 1 M i .1 .! ! a y -s # 8 H >- Aeg E3 E . j2 j7 g oj 05 .{ M j] j3 v ' a s .y 4 4 2 < g i '.< w $y s2 5 gGs e c5 g jw 3i e if r{ es J < G .3 r e 2 m e e v r e o u ,. g u. m
- u e r V-STS y t - - -
3/4 3-32
C gg I Allll' 3.1-4_(Cont inner,l.) ik d ENGINEERID SAFETY FEATilRES ACTtlATION SYSilli INSTRilHlilTATION TRIP SETPOINTS SINSOR 10lAt IHHOR FUNCTIONAL. l! NIT b ev Al l 0WANCl: ( I A_} 7 _Q)_ 1 RIP SETPalHT All0WARI E V,Allif W. ^; , ; 'i;tecy my Feedwater (Continued) , , d4 Safety Injection See Item 1. above for all Safety Injection Setpoints and Allowable Values.
, e %. Station Blackout H.A. it. A. N.A. N.A. N.A.
i
-; T-!; cf " '- T ;,,ete.- " =p; -LA- ;4. A. ^! ". - *! ". . 1. ". . -
- 4.. Z t'cr. 'r
- .cf - . m. L s, -
W .i.A.
~
;4. A. ; (2) ft ; ("1) " -
!'; c ; xr R
- D. Automatic Switchover ty Containment Sump
?
W Automatic Actuation Logic H. A. N.A. N.A. N.A. N.A. and Actuation Relays III y u Iq}er q 1i. RWST Level Low Lw 1 N.A. N.A. N.A. $ -(4834- geb
- Coincident With: $ M Gd
^ :1 9 m t c g t_fr' .wA. M-A-.- -N-A- : (20") <m'"'
and ' I 2 ' ' Safety Injection See Item 1 above for all Safety Injection Setpoints and Allowable Values. R Loss of Power i a. 193$ ledt v 4 kV Bus l'..i -^1'aca it. A. N.A. N.A. (< nrm &) < (#.rffJf vol ts ll.oss of VoltageJ iiolts with U i th a < (4HW-I ' a 5 (e-e5)(der second time second time delay
. delay
- b. 332 6 Intr.-
4 kV Bus L:<..- L um H.A. fl. A. fl. A. < (4HWr) volts < (691+) volts
] Grid Degraded Voltage) wi th a 5 (-3-3) , wi th a < (9-t) second timel '" second time
4 C I Alll i :1. .l-4 (Cont ir:ucil) I t.
, ENGil4LFRID SAFLlY FEAlllRI S AClllAll0N SYSitM INSIRilHLNIAIION TRIP SEIP0lHIS l
r
~
SiNNOR i IDIAL iRHOR i FilNCTIONAl. IINIT 4 Al l 0WANCli_(TA) / _f S) 1 HIP SETPOINT ALLOWABLE VAlllE , foi. Entlineered Safety Feature i Actuation System Interlocks en uk NA MSO lAStY a Pressurizer Pressure, P-11 (:h+) ( 9-7+) ( 1-4) (198'r) psig - 1 (49M-) psig & I f4996-) psig
-b- 4eu- c T**U , i 12 -f4-0) -(4-44) -(4-e) -(552)"D "
-' (550.5)"'
, - (555.') l'
, w b la. Reactor Trip, P-4 N.A. fl. A. fl. A. N.A. N.A. A t : Y ! ! #I 1 0 .; 1-- l 't
~
I e a
1 l TABLE 3.3-5 ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS
- 1. Manual
- a. Safety Injection (ECCS) Not Applicable Feedwater Isolation Not Applicable Reactor Trip (SI) Not Applicable Containment Isolation-Phase "A" Not Applicable Containment Vent and Purge Isolation / Not Applicable Emergency Feedwater Pumps Not Applicable Service Water System Not Applicable Containment Air Recirculation Fan Not Applicable
- b. Containment Spray Not Applicable Containment Isolation-Phase "B" Not Applicable Containment Vent and Purge Isolation Not Applicable
- c. Containment Isolation-Phase "A" Not Applicable Containment Vent and Purge Isolation Not Applicable
- d. Steam Line Isolation < Not Applicable
- 2. Containment Pressure 6 Hrg g
- a. Safety Injection (ECCS) < (27.0)M/-(42-)N -
- b. Reactor Trip (from SI)
< (2.0)
I l
- c. Feedwater Isolation (7.0) N
- d. Containment Isolation-Phase "A" <(17.0)es/(27.0)(r
- e. Containment Ven't and Purge Isolation (25.0) /(10.0) M
- f. [IN$$FeedwaterPumps f
- g. EL nti:1 Service Water System '-< ((60.0)
M.0)( [ /(47.0) g # n.
- h. E$0:W(([O kr7 *"
(k0) /(M.0)
- 1. $ rye'o7Ec[I$NtYer $ct^;;'ic 1:s90 I -
35~ W-STS 3/4 3-X SEP 151981
- g, _
I TABLE 3.3-5 (Continued) , ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS
- 3. Pressurizer Pressure-Low g 25.o
- a. Safety Injection (ECCS) 1 (47,4) 12.0) m
- b. Reactor Trip (from SI) 1 (2.0)
~
- c. Feedwater Isolation 1 (7.0) N
- d. Containment Isolation-Phase "A" 1 (17.0)$ /(27.0,di)
- e. _ Containment Vent and Purge Isolation f (2';.0) "10. 0) ^' 4 kmugey
^= 4 - j Feedwater Pumps f.
-< (60.0) d-
- g. E::enti:1 Service Water System < (47.0) /(Mm.o N
- h. ...E.__.d_."......'....,....-
_(< /M 4r f: .t :1 ";= I::1:ti:n Not ;,ppliceble-4 'Ya SD% OU"---- u-u
- 1.o 22.o
- a. Safety Injection (ECCS) 1 (4 erg)$)/(-le-e-)y
- b. Reactor Trip (from SI) 1 (2.0)
- c. Feedwater Isolation 1 (7.0) N %
- d. Containment Isolation-Phase "A" 1 (17.0) /(27.0) N I
- e. Containment Vent and Purge Isolation ' (25. 0)N /(10. 0}N) ^'4
- f. "W9$YFeedwaterPumps (60.0) 5 %
- g. E:::at%1 Service Water System < (M) /(47.0)
- h. #- .D.. . . . . Y. . "_ . . . . , . . . . .
" a"m'" . ". ,'O)- 7 0 C: .t--! " x !;eleti:n N t f.pplic ble
% M aam Flow in Two Steam Lines - High Coincident with IavD "
- a. Safe ,jection (ECCS) 1 (24.0)(4) . )(5)
- b. Reactor Tr rom SI) 1 (4
- c. Feedwater Isolati (9.0)(3) f
- d. Containment Isolation-P "A" (19.0)(2)/(29.0)(1) ]
- e. Containment Vent and Purge Iso n 1 (27.0)(1)/(12.0)(2)
- f. Auxiliary Feedwater Pump 5 (60.0)
- g. Essential Service er System _ 34.0)(2)/(49.0)(1)
- h. Steam Line ation 3) 1 (9.
- 1. Cont nt Cooling Fans 1 (57.0)( .0)(2)
(, J. trol Room Isolation Not Applicable , W-STS 3/4 3- SEP J 5198;
TABLE 3.3-5 (Continued) ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL ANO FUNCTION RESPONSE TIME IN SECONDS Steam Flow in Two Steam Lines-High Coincident with eam Line Pressuae-Low
- a. ety Infection (ECCS) 1 (12.0)(5) .0)(4)
- b. Reac rip (from SI) <
- c. Feedwater ion 1 (7.0)(3) l d. Containment Isolat -Phase "A" 1 (17.0)(2)/(27.0)(1)
- e. Containment Vent and Pu ion ~< (25.0)(1)/(10.0)(2)
- f. Auxiliary Feedwater Pum 1 (60.0)
- g. Essential'Servic er System 1 (32.0)(2)/(47.0)(1)
- h. Steam Lin ation < (9.0)(3)
- 1. Con ent Cooling Fans 1 (1)/(40.0)(2) j ontrol Room Isolation Not Applica 14 i '2 .
fK Containment Pressure--Migh "igh
- a. Containment Spray < (45.0) N/(57.0) N b, Containment Isolation-Phase "B" GC) ~M
- c. h. -@_ _.T'O_ .*.'" v"& 6&. L"'% Fan
$(55)
-_ s . n, ra . .M)( 4 eo,oo 6 %, - Steam Generator Water Level--High-High
- a. Turbine Trip 5 (2.5)
- b. Feedwater Isolation 1 (7 ^)N <. 7.o ._
~/K Steam Generator Vater Level - Low-Low
- a. ";ter-driven Auxilicrf 5;;icter "'is E=*tarFuJWe- P -ps 1 (60.0) l u ,-w ,s._
-- n... m ', Y
. . 1' .'u.. . ' .L-v:
;.2 '
- _.an.-,
n, ton une Pa-ge, 9% Containment Radioactivity - High dr it.
- a. Purge and Exhaust Isolation 1 (25.0) E/(10.0) N l
l 6 l3 PSTS 3/4 3- SEP 151981
C TABLE 3.3-5 (Continued) l ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS.
- 22. n-a. Levei tuw .uui.,s.2__. . . . . . .. < ,__ _. . . . . . . . . . . . . . . . . . . . _ _...g .
i;.el iiivh ...c 0;f;ty Injectiv.. -
]
- e. Avt^;; tic 0.itwhuvui te C;.,tsh.asnt a - --
.-r ,<,9t_t_n
/_ ,
N/ r_oct
-- N, i
mm < is. ,,_2__.,u,...we w w * - n u..... n- 3 9 2. . . ,- - r e:fa;ter ," gap; i (An n) c3 71h Station Blackout E=a +e g tac y
- a. A-mil;;ci eedwater F Pumps 1 (60.0)
T-i- ef Meia c::da;te. Tumg.
m Agv414 3qy C33f;;t;,- pg,p3 gg; App]j;;b]; 30 W Loss of Power
- a. 4.16 kV Emergency Bus 1 (10)
Undervoltage (Loss of Voltage)
- b. 4.16 kV Emergency Bus 1 (10)
Undervoltage (Degraded Voltage) 3% W 3/4 3-49 SEP 151981 _-STS e 9
' - S* ($) TABLE 3.35 (Continued) TABLE NOTATION
- Diesel generator starting and sequence loading delays included.
Response time limit includes opening of valves to establish SI path and attainment of discharge pressure for centrifugal charging pumps, SI and RRR pumps.
# Diesel generator starting and sequence loading delay not included.
Off-site power available. Response time limit includes opening of valves to establish SI path and attainment of discharge pressure for centrifugal charging pumps.
## Diesel generator starting and sequence loading delays included.
Response time limit includes opening of valves to establish SI path and attainment of discharge pressure for centrifugal charging pumps. e 15 E [ W-STS 3/4 3-
TABLE 4.3-2
. r
, ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION ;
ip vi SURVEILLANCE REQUIREMENTS Y i
. TRIP ANALOG! ACTUATING MODES CHANNEL DEVICE MASTER SLAVE FOR WHICH CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST IS REQUIRED
- 1. SAFETY INJECTION, SEAGMR-TMF FEEa"^TER ISOLATI0", 00"T':0L g R00;4 ISOLAiioW 5iAni GiE5EL GE;4CRATOR5, CGhinsretni C00LI M .
TA"5 A"O ESSE"TIAL SER"!CE ATE"
- a. Manual, Initiation N.A. N.A. H.A. R N.A. N.A. N.A. 1, 2, 3, 4
- b. Automatic Actuation N.A. N.A. N.A. N.A. M(1) M(1) Q 1, 2, 3, 4 Logic and Actuation Relays
.A c. Contairment Pressure- S R M N.A. N.A. N.A. N.A. 1, 2, 3 y High I f!@ d. Pressurizer Pressure- S R M N.A. N.A. N.A. N.A. 1, 2, 3 Low .
- e. e. .-easuresi 0.s.fm. .6.
aW.P. spet tw S R H N.A. N.A. N.A. N.A. 1, 2, 3 _Betueen Ste:: Line -- Higtr-
- f. St::: r!:e fa Tua W= < -R- -M - -M.-A ----- .AA. - -N-A--- -fi . A . - -1, 2, 3 s@ Linn i;ip.Cair.cii.nt W4h-Either
.gg mv , ,r $- e 46 -M-A-- .H-A-- -N-A- R-A:- 1, 2, 3
-2 . Ste : Liae_ -S- -R- W -M-A- -N-A- N-*- -N-A - 1, 2, 3 -
ore :ure--Lau
- 2. CONTAINMENT SPRAY Mu a. Manual Initiation N.A. N.A. N.A. R N.A. N.A. N. A. 1, 2, 3, 4
_ b. Automatic Actuation N.A. N.A. N.A. N.A. M(1) M(1) Q 1, 2, 3, 4 l c.n Logic and Actuation m Relays E c. Containment Pressure-- S R H N.A. N.A. N.A. H.A. 1, 2, $ l l+194-Htt)F Hi S I i l
TABLE 4.3-2 (Centinued)
'T ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION g SURVEILLANCE REQUIREMENTS
@ TRIP ANALOG ACTUATING MODES CHANNEL DEVICE MASTER SLAVE FOR WHICH CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST IS REQUIRED
- 3. CONTAINMENT ISOLATION
- a. Phase "A" Isolation
~
- 1) ManualIndE4Ea N.A. N.A. N.A. R N.A. N.A. N.A. 1, 2, 3, 4
- 2) Safety Injection See 1 above for all Safety Injection Surveillance Requirements
- 3) Automatic Actuation N.A N.A. N.A. N.A. M(1) M(1) Q 1,2,3,4 w Logic and Actuation
) Relays s b. Phase "B" Isolation
- 1) Hanual IItNIE9 N.A. N.A. N.A. R N.A. N.A. N.A. 1, 2, 3, 4
- 2) Automatic Actuation N.A. N.A. N.A. N.A. M(1) M(1) Q 1, 2, 3, 4 Logic and Actuation Relays
- 3) Containment S R H N.A. N.A. N.A. N.A. 1, 2, 3 Pressure j!gg ll gh l c. Purge and Exhaust Isolation
- 1) Automatic Actuation N.A. N.A. N.A. N.A. M(1) M(1) Q 1,2,3,4 Mo Logic and Actuation Relays
" d=c Ib-u s
- 2) ContainmenV Hadio- S R H N.A. N.A. N.A. N.A. 1, 2, 3, 4 m lerical-High 2 adoity
- 3) Safety Injection See 1 above for all Injection Surveillance Requirements.
l l l
y i TABLE 4.3-2 (Continued) x ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS TRIP ANALOG ACTUATING MODES CHANNEL DEVICE MASTER SLAVE FOR WHICH CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST _ LOGIC TEST TEST TEST IS REQUIRED
- 4. STEAM LINE ISOLATION
- a. Manual Idh'abn N.A. N;A. N.A. R N.A. N.A. N.A. 1, 2, 3
- b. -Automatic Actuation N.A. N.A. N.A. N.A. M(1) M(1) Q 1, 2, 3 Logic and Actuation
- Relays
- c. Containment Pressure-- S* R H N.A. N.A. N.A. N.A. 1, 2, 3 i;.J. iii 'H -Z
- d. NEE bM4'MY5 S R H N.A. N.A. N.A. N.A. 1, 2, 3 R Lir.c - "ip Cc,i.. cia..t Wit;. E; User -
[
- 1. T,yg--Lcw-Lou er S R H N.A. N.A. N.A. N.A. 1,2,3
- 2. Sica. Line S R H N.A. N.A. N.A. N.A. 1, 2, 3 6 Hsa $5555$ ke"u..e R 4e. s R M A'A v4 &4 A'A 3 a 'l S. TURBINE 1 RIP AND FEE 0 WATER ISOLATION
- a. Steam Generator Water S R H N.A. N.A. N.A. N.A. 1, 2 Level--High-High
- b. Automatic Actuation N.A. N.A. N.A. N.A. _M(1) M(1) Q 1, 2 g and Actuation Relay
- 6. A"XILIARY FEEDWATER
- a. Manual TEbd** N.A. N.A. N.A. R N.A. N.A. N.A. 1, 2, 3
$ b. Automatic Actuation N.A. N.A. N.A. N.A. M(1) M(1) Q 1, 2, 3
' logic and Actuation Relays c c. Steam Generator Water S R H N.A. N.A. N.A. N.A. 1, 2, 3 m Level--Low-Low 1
a Z u- d m3C$ d Wa&* E S$
- c- .
*w<
w W> g j w mWW 4 .t s e m o W> g % MSM 4 T ' f*W - % % $ 4 ze H ::: T C MH q \ } sz 08
<a m
w 0 m e . i-
~
= a= 1 CWC ! q. 4 *F
- 5 0 a.EM5- a g H i M 5 E05t$ y g g 5 -<co-
. 3 z a j 8 28 1
- d S en a2
~ Ew 8We -
A WW #550 f E 7 4 w 5 558W s
$ E$ z d i 5E S S se f
? WE oc T .8 e it d 5 55 -
S a ,
$ Ew v W ES m W W 5 EE .
m
+
k $ b 9 +g .7 b 3 . c.? .e
$ c h, .i 8 3* .
r' s .- n H b
- c _, t<
E
- N o Fe s s' s3 a I
2 e m = .3 343 c 't
'A j g e o o
' ~
l W-SIS .s 3fcfJ-43
~ .
l
I TABLE 4.3-2 (Continued)
- T { ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION
$, SURVEILLANCE REQUIREMENTS u
TRIP
. ANALOG ACTUATING MODES CHANNEL DEVICE MASTER SLAVE FOR WHICH CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST 15 REQUIRED
-AUMM4ARW. FEEDWATER (Continued)
Crim sM.y 4- 'Jad .. ..I t;;; """
#-A. -A-- W-A- 4-A:- -N-A - -M-A- -+-
d s. Safety Injection See 1 above for all Safety Inject W Surveillance Requirements e 4. Station Blackout N.A. N.A. N.A. R N.A. N.A. N.A. 1, 2, 3
- g. Trip of va ia F::i;ter W-A. --M-*-. -N-A--- -R- M-A-- 4-A-- -M-A - -lv-2--
Pt$*PS t
%
- AUTOMATIC SWITCHOVER TO Y CONTAINHENT SUMP
#1 a. RSWT Level - Low 5 R H N.A. N.A. N.A. N.A. 1, 2, 3, 4 Coincident With "
Ce n ta i --^ n t E ,- Lc ; 1 ---G- --M--- -N-A-- R -N-A-- #-A- -1, 2, 3, T Figh 'nd Safety Injectiori See 1 above for all Safety Injection Surveillance Requirements
- b. Automatic Actuation N.A. N.A. N.A. N.A. M(1) H(1) Q 1, 2, 3, 4 Logic and Actuation
- Relays 1L.i LOSS OF POWER
- a. 4.16 kV En rg:ncy Bus N.A. R N.A. R N.A. N.A. N.A. 1, 2, 3, 4
'hderee!!:g: { Loss of p Voltage) o
- b. 4.16 kV E::rge"c"--Bus N.A. R N.A. R. N.A. N.A. N.A. 1, 2, 3, 4 m Hada~e!!:geJDegraded g~
Voltage? E e
TABLE 4.3-2 (Centinued)
'i jj (- ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS TRIP ANALOG ACTUATING MODES CHANNEL DEVICE MASTER SLAVE FOR WHICH CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST- TEST LOGIC TEST TEST TEST IS REQUIRED 10 1 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INTERLOCKS
- a. Pressurizer Pressure, N.A. R H N.A. N.A. N.A. N.A. 1,2,3 P-11 b.- Law,Lew T,yg," 12- -N-A- -R-- -M - & N. A. - N.A. --M-A- 1, 2, 3 - .
- 4 E. Reactor Trip, P-4 N.A. N.A. N.A. R N.A. N.A. N.A. 1,2,3
, kt Y
8k$ e 9 o S 19 5 4
TABLE 4.3-2 (Continued) TABLE NOTATION (1) Each train shall be tested at least every 62 days on a STAGGERED TEST BASIS. I l 6 PSTS 3/4 3-49 SEP 151981
INSTRUMENTATION r 3/4.3.3 MONITORING INSTRUMENTATION RADIATION MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.1 The radiation monitoring instrumentation channels shown in Table 3.3-6 shall be OPERABLE with their alarm / trip setpoints within the specified limits. APPLICABILITY: As shown in Table 3.3-6. < ACTION:
- a. With a radiation monitoring channel alarm / trip setpoint exceeding the value shown in Table 3.3-6, adjust the setpoint to within the l limit within 4 hours or declare the channel inoperable.
- b. With one or more radiation monitoring channels inopera' ole, take the ACTION shown in Table 3.3-6.
- c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4 4.'3.3.1 Each radiation monitoring instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and ANALOG CHANNEL OPERATIONAL TEST operations for the MODES and at the frequencies shown in Table 4.3-3. L l
~
41 W-STS 3/4 3-M SEP 151C81 rg,, - w= -o- -y - ,-- yy--- .w ---y- -w- - - - - w w -rv, w -- - - - - - - - e-
7 r TABLE 3.3-6 y g RADIATION MONITORING INSTRUMENTATION MINIMUM
, CHANNELS APPLICABLE ALARM / TRIP MEASUREMENT INSTRUMENT OPERABLE MODES SETPOINT RANGE ACTION
- 1. AREA MONITORS 1 r" ' u nraae Pool Area
- i. Criticality Monitu, (IT $ 1 G d - G0 10 ) 4/;i, Z~i
- 11. Ventilation Svsta=
5 f . auon (1) (1 2 x backgrouna) ti 10 ) ,, 27 _5 afr ~/ Ar or so-8-jog m 14
%. Containment - Purge & (1) 6 (< 2 x background) (4 - 10 ) cp 2821.
Exhaust Isolation A"*'S"**'" i ttune.lato. C ra s e-Control Room Isolation (1) All MODES
$ ,x c pn e,
< 2 x background) ~I jo#s
- 19 )b 2R 2.3
$ d $"lMd'-5NN[ O) Ag3 t10 DES f(Awhen.
(_ 'Mn a k * ((10
'Y iL Containment Area,P.t h 2 1, 2, 3 & 4 ( ) rad /hr M rad /hr 392'f lakev to*.-10'
$4 2. PROCESS MONITORS
. a. ""-1 (;tnraae Pool Area -
Ventilation System isui :;r. ^ 0#1 i. Gaseous Activity (i) ?v h =c kn eniind i
, (1 - 10") g cpm 27 y ";. mis.uiote Activity (1) **
([2xbackground) (1 - ;G ) cp 77 - a.,h. Containment
- i. Gaseous Activity
-a) Per;;c 1 Exhaert Iselet-ion- 5 flP G -(< 2 x baskgreend)
- -(1 - 10_) _pg_
4.h) RCS Leakage Detection (l) 1, 2, 3 & 4 N/A (1 - 10 ) cia 2a
<n M ii. Particulate Activity ### ~ /#N/CC-i $
a-}-Perna 1_Fxha nt
-Isolat-ion fl-}- .(4-2 + backgrcund)- fl--10 -}-cpm5 h
m <Cb) RCS Leakage Detection (1) 1, 2, 3 & 4 N/A (1 - 10 ) cpm f924 l ' f 10'*-16% c,.lcc
- With fuel in the storage pool or building
** With irradiated fuel in the storage pool l
h b . f6 3 Compue t cooling Wcder S- As bathsed Ib ~L l* 3*('Y'd
*T r L-oe ,s ry . (,) su nop2S -
a Lcv p 6 (g ) r}u tM DES g y ke.tzqed gg-'_;,3-5 3 g $- c
TABLE 3.3-6 (tentinued) vi RADIATION MONITORING INSTRUMENTAT70N-MINIMUM CHANNELS APPLICABLE ALARM / TRIP MEASUREMENT INSTRUMENT OPERABLE MODES SETPOINT RANGE ACTION PROCESS MONITORS (Continued)
- c. Noble Gas Effluent Monitors 1 Da &aste Building Exhaust Sys h 1 1, 2, 3 & 4 N.A. 1-102 uCi/cc 30 _
ii. Auxiliary Buildina ' h% p Sy:, tem, 1 1, 2, 3 & 4 N.A. 1-10 uc u u. % UN3EtrM bMS t'
~
i d [ ' '~ 1/ valve 1, 2, 3 & 4 N.A. 10 3 ug9 qc
,,392.j g
- tr'-10 % )k,-
" h "=ncnheric Steam Ni Dump Valve Discharge 1, 2, 3 & 4 N.A. 1-103 uCi/cc
- v. Shield Building Exhaust System 1 1, 2, 3 & 4 1-104 uCi/cc 30 vi. Containment Purge &
5 Exhaust System , 2, 3 & 4 N.A. uCi/cc 30 vii. C xhaust ystem 1 1, 2, 3 & 4 N.A. 1-105 uCi/cc
$o e-*
t,11 5
TABLE 3.3-6 (Continued) ACTION STATEMENTS AL uun ca _: +ha number of OPERABLE channels less thma "..% Channels Ortnao a . m.:c: :-' "' -.m area surveys of the nitoring ...1-"- ntation at ACTION - With the number of OPERABLE channels less than the Minimum Channels OPERABLE requirement, comply with the ACTION l i requirements of Specification (3.4.6.1). l l m.il;;. " "'" +h= n'imher of OPERABLE channels less than the Minimum a g Channels OPERABLE requirement - _ , . , . . . . _
==a+- .." Sp=u ncar. ion (3.9.12).
" TION require-ACTION - With the number of OPERABLE channels less than the Minimum Channels OPERABLE requirement, comply with the ACTION require-ments of Specification (3.9.9).
23 ACTION 29,- With the number of OPERABLE channels less than the Minimum Channels OPERABLE requirement, within I hour initiate and maintain operation of the control room emergency ventilation system in the recirculation mode of operation. 1*4
- ACTION )&'- With the number of OPERABLE Channels less than required by the Minimum Channels OPERABLE requirements, initiate the preplanned l alternate method of monitoring the approriate parameter (s), within 72 hours, and: . .
h 1) either restore the inoperable Channel (s) to OPERABLE - status within 7 days of the event, or
- 2) prepare and submit a Special Report to the Commission l pursuant to Specification 6.9.2 within 14 days following I the event outlining the action taken , the cause of the
( inoperability and the plans and schedule for restoring I the system to OPERABLE status. ACT[pM:LS~- W4%A %A4. O d OPERABLE M M b nh u oeease 17i.w, p
$ 4 m & & A Q W %. recto d SW bA b ~
W-STS 3/4 3- SEP 151c81
TABLE ,.3-3 19 U RADIATION HONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS ANALOG CHANNEL MODES FOR WHICH CHANNEL CHANNEL OPERATIONAL SURVEILLANCE IS INSTRUMENT CHECK CALIBRATION TEST REQUIRED
- 1. AREA MONITORS H. .L '. Si:r:^ paal Area
- i. Criticality Monitor 3 "
__ . M ii. Ventilation System
! wiauon S R H A %. Containment - P' urge & Exhaust Isolation i 11ampaattu. Ga e- S R H 6 w
@ %. E sr roe ~mIsolation Cnb*rol"" *=y S s
R H All MODES g n
'E % d$nTATn"m'enlTrea- S R. H ass 1,2,3 nooe[E-4 T 2. PROCESS HONITORS M & N 1 Storage Pool Area - Ven- -
tilation Sysi.em i=vio;.. .. - Gaseous Activity %_
- i. S " **
ii. Partisuhte Acuvity S R H ~' ~ cCh. Containment
- i. Gaseous Activity a) Purge-& 4xhauste bolat; ion- + 4-- -H- -6~
cCh) RCS Leakage Detection .S R H 1, 2, 3, & 4 ii. Particulate Activity a)Jurge & E,xhaust. 4s61atien. -S -R-- -H - --
<Cid RCS Leakage Detection S R H 1, 2, 3, & 4 v,
m *With fuel in the storage pool or building. 52 **With irradiated fuel in the stirage pool.
- b. Prl Mr1 Co* P8"'# O"I*j ht" g i Loop *8 . 5 R M au nooc a Loop .S R M ALLHCOES
TABLE 4.3-3 (Centinued)
$ RADIATION HONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS ANALOG CHANNEL MODES FOR milch CilANNEL CHANNEL OPERATIONAL SURVEILLANCE IS INSTRUMENT CHECK CALIBRATION TEST REQUIRED PROCESS MONITORS (Continued)
- c. Noble Gas Effluent Monitors t "-' ~ ta Building Exhaust System 9 1- * "
R M 7__' - y , S R H Hatn 5+ea m Lae t' iit. Sten Safety-Valve- - { A -Bheharge S R H 1, 2, 3 & 4
- h. ^*=mnheric Steam Dump Va v e S R H 1. 2. 1 1 4 Shield Building v.
Exhaust System S M 1, 2, 3 & 4 vi. Containment Purge & ' ' ' ' ' Exhaust System 3 R H ' ~~ 1, 2, 3 & 4 i v gserExhaustSystem S R H 1,2,3
=o M
(
INSTRUMENTATION MOVABLE INCORE DETECTORS LIMITING CONDITION FOR OPERATION 3.3.3.2 The movable incore detection system shall be OPERABLE with:
- a. At least 75% of the detector thimbles,
- b. A minimum of 2 ditector thimbles per core quadrant, and
- c. Sufficient movable detectors, drive, and readout equipment to map these thimbles.
APPLICABILITY: When the movable incore detection system is used for:
- a. Recalibration of the excore neutron flux detection system,
- b. Monitoring the QUADRANT POWER TILT RATIO, or
- c. MeasurementofFh,F(Z)andF q xy
. ACTION:
With the movable incore detection system inoperable, do not use the system for the above applicable monitoring or calibration functions. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. SURVEILLANCE REQUIREMENTS 4.3.3.2 The movable incore detection system shall be demonstrated OPERABLE at least once per 24 hours by normalizing each detector output when required for:
- a. Recalibration of the excore neutron flux detection system, or
- b. Monitoring the QUADRANT POWER TILT RATIO, or
- c. MeasurementofFh,F(Z),andF q xy*
l l W-STS 3/4 3 SEP 151981
INSTRUMENTATION SEISMIC INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.3 The seismic monitoring instrumentation shown in Table 3.3-7 shall be OPERABLE. . APPLICABILITY: At all times. ACTION:
- a. With one or more seismic monitoring instruments inoperable for more than 30 days, prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within the next 10 days outlining the cause of the malfunction and the plans for restoring the instru-ment (s) to OPERABLE status.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.3.3.3.1 Each of the above seismic monitoring instruments shall be demon-strated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and ANALOG CHANNEL OPERATIONAL TEST operations at the frequencies shown in Table 4.3-4. 4.3.3.3.'2 Each of the above seismic monitori Jninstruments actuated during a seismic event greater than or equal to (4:41 Fg shall be restored to OPERABLE status within 24 hours and a CHANNEL CALIBRATION performed within 5 days following the seismic event. Data shall be retrieved from actuated instru-ments and analyzed to determine the magnitude of the vibratory ground motion. A Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 10 days describing the magnitude, frequency spectrum , and resultant effect upon facility features important to safety. - k W-STS 3/4 3- 5 . SEP 151981
b TABLE 3.3-7 SEISMIC MONITORING INSTRUMENTATION MINIMUM MEASUREMENT INSTRUMENTS INSTRUMENTS AND SENSOR LOCATIONS RANGE OPERABLE
- 1. Triaxial Time-History Accelerographs -
6
- a. }-Sn.wi -A joo rece Re)) -Fod Af . r l9 '
1N R.m Aw rn+w k t
- b. I- S H - x r - t. ,o t to ,+. reu s (,;, n .t l9 1N
- c. I- SH - Y Y- 4 7sn ol' d. n o . XIv.
- t. la ik I
- 2. Triaxial Peak Accelerographs
- a. I-S H-x R -/ 7DL Rearlev ties Sao. n-20 Hg i
- b. 1-41 ~ X R-l7D E 1?n elne (%I. Ap. o-20 Hy 1
- c. I-sn - Y R - /,70y pccio Asi9 o - 2 6 Hu 1 e
e =*-- t e +
- 3. Triaxial Seismic Switches orTrigger3
- a. I-Sh- Xs -4780 Ree field NA 1*
- 6. I-SM ~ K S - 4201 ro d. F*ou a A . A/A 1*
- c. 1- C h - X S -4709 [pt. l~oud. NA 1*
- d. l- G h - YS -le'll n fon f. Os. f/r. NA 1*
- 4. Triaxial Response-Spectrum Recorders
- a. I~ Sn- t R -470S~ Con +. Found. /~3 0 Ns- "
1* i
- b. I ~6H- Y A - L70/, S.G. Ils Sy. 1-30 Hy 1
- c. I- S h - X R ~/,7tn Pri. Aux. Aly. l- 3 d Np 1
- d. l- $ H ~ YR -DO R %.W. Pa no k , l-30 h 1
. +
4r & ( "With reactor control room indication sf W-STS 3/4 3-b6 SEP 15 G21
= ,vv w
TABLE 4.3-4 SEISMIC MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS ANALOG CHANNEL CHANNEL CHANNEL OPERATIONAL INSTRUMENTS AND SENSOR LOCATIONS CHECK CALIBRATION TEST
- 1. Triaxial Time-History Accelerographs
- a. I- S M - x1 - L mo F',.ce_ Cs e.lk M* R SA 4 Casr cent. Rm Asrt. En h see.A R y y y b ae t -s n - x i - L > n I rad. Gu d. A k H* R SA C.48 .L-SM- V'~ h')'o Cent 09 f/r
- N M* R SA
- 2. Triaxial Peak Accelerographs
- a. l-SH-XR - 4 702. Remin. l/es . spp NA R NA
- b. l~Gh-xR-L*103 Q v fn,l. Ak . NA R NA
- c. l- 6 H-xR ~1,7MV
#cck) A,o,$ a NA R NA
+ 4WP e e P M= + 4Wr
./
g 3. Triaxial Seismic Switches
- a. I-S H~Ys- n'?oo F~ree fie'Id ** H R SA b.1-cM-rS-47B/ <%f. E d ** H R SA
- c. 1 -K M -Ys ~4 7A9 fa,J . A mt. *
- H R SA
- d. l~ <M- 2 s -A 710 font. go. p/r. ** H R SA
- 4. Triaxial Response-Spectrum Recorders
- a. l- SM-XA -470r 64.' & ,J . ** H R SA
- b. I-GH yQ-f184 .% HB n may NA R % Wh r
- c. I A h - YR - A20'1 Pri. Aux . N da NA R 5k2 t's
\ d. 1 c,n- Y R -s7f)g xgo. L,s }lre. NA R W /Jfr
-e w + W
.de . 9Wr e WP t
"Except seismic trigger
**With reactor control room indications.
s l % l W-STS 3/4 3-N 3 p l 3 gg ; 1
~.
INSTRUMENTATION METEOROLOGICAL INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.4 The meteorological monitoring instrumentation channels shown in Table 3.3-8 snall be OPERABLE. APPLICABILITY: At all times. ACTION:
- 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 6.9.2 within the next 10 days outlining the cause of the malfunction and the plans for restoring the channel (s) to OPERABLE status.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.3.3.4 Each of the above meteorological monitoring instrumentation channels shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3-5. 57 W-STS 3/4 3-ha' SEP i 51981
l l, h TABLE 3.3-8 METEOROLOGICAL MONITORING INSTRUMENTATION NINIMUM INSTRUMENT LOCATION OPERABLE l
- 1. WIND SPEED
- a. bwu level, Nominal Elev. '/1 f f 1
~
- b. %cr Lcue.(, Nominal Elev. 2o t f r- 1
- 2. WIND DIRECTION
- a. low er level Nominal Elev. 93 f f- 1
- b. f ieur 2, eve /,
Nominal Elev. 204 f/- 1
- 3. AIR TEMPERATURE - DELTA T
- a. Lower Lue/, Nominal Elev. f3#- /SD ff 1
- b. /l m r 1ew/, Nominal Elev. Y3 fr-Jof ft 1 u
l l l ( W-STS 3/4 3- SEP 151981
TABLE 4.3-5 ' METEOROLOGICAL MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL CHANNEL , INSTRUMENT CHECK CALIBRATION
- 1. WIND SPEED
- a. Nominal Elev. 4.? f f- D SA
- b. Nominal Elev. 969 ff D SA
- 2. WIND DIRECTIDN ,
- a. Nominal Elev. Y? [I D SA
- b. Nominal Elev. 2 69 ff- D SA
- 3. AIR TEMPERATURE - DELTA T
- a. Nominal Elev. 'f? [t-/S0 ff- D SA
- b. Nominal Elev. YJf /-J os f f D SA i
l 3/4 3-M SEP 151981 PSTS
^
l l INSTRUMENTATION REMOTE SHUTDOWN INSTRUMENTATION l LIMITING CONDITION FOR OPERATION 3.3.3.5 The remote shutdown monitoring instrumentation channels shown in
, Table 3.3-9 shall be OPERABLE with readouts displayed external to the coatrol l
room. APPLICABILITY: MODES 1, 2, and 3. l ACTION: ! a. With the number of OPERABLE remote shutdown monitoring channels less than required by Table 3.3-9, restore the inoperable channel (s) to OPERABLE status within 7 days, or be in HOT SHUTDOWN within the next 12 hours.
- b. The provisions of Specification 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.3.3.5 Each remote shutdown monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK and CHANNEL l CALIBRATION operations at the frequencies shown in Table 4.3-6. l W-STS 3/4 3- SEP 151981 l 1 l
TABLE 3.3-9 st < h un REMOTE SHUTDOWN MONITORING INSTRUMENTATION MINIMUM READOUT MEASUREMENT CHANNELS INSTRUMENT LOCATION RANGE OPERABLE
- 1. "; cr "- 7 " -'::r " u- -
l 1 Intermediate Range Nuclear Flux CP-toyA *6 /F'-2co% fWit ] Source Range Nuclear Flux c Atos A
- S 8 l - JDY C PS j 2.%
4.
-cter Tri; Bre:Ecr Ind!r: tie- -0"E"-CLOSE 1/t r ip h-a='---
1
- 35. Reactor Coolant Temperature - c P- tos*
- S o-7eo p 1 Average t'
+
'I 5% Gen. E F td Fleia c.P-lot **B 0-380 GFN l T 5. Pressurizer Pressure c t-tor A65 o-15oo psi 3 1 f (, lh Pressurizer Level c.f-ior4 0 o- f oo % 1 2
- 73. Steam Generator Pressure C /-lot A*f3 o -1500 psi 3 '/ te r ;c r-?*a -
g M. Steam Generator Level c t-losA *6 1/ tbr ;cnerater - g,
- 11. Cor. tral Red P :itie" I iacerti^^ lirit Liait Switche critch/.od tih--RHR-fisw Rat:- m m
m 13. ";;R T' :p e r=+ " ra _ +- [ 14. A =i'iary F :1;ater Flecr 9:te- -t-- o - f oo 7. E 1 Gonc. Ac.ia Tanh. Leul c P-t o rd *6 g
TABLE 4.3-6 REMOTE SHUTDOWN MONITORING INSTRUMENTATION
- SURVEILLANCE REQUIREMENTS CHANNEL CHANNEL INSTRUMENT CHECK CALIBRATION
- 1. Derer tr.;2 Lelear r h.x -M-- t I lt. Intermediate Range Nuclear Flux M N.A.
~
2.1 Source Range Nuclear Flux M H.A. .
- 4. Rmeet:r " !p ?r::Ecr !~'! rat!cn 4t-- 4HW-35 Reactor Coolant Temperature - Average M R qg i h...EFA S._.f % fh.W._.-
M R N
- TK Pressurizer Pressure M R Y
g44 Pressurizer Level M R w D ~I % Steam Generator Pressure H R
? TR. Steam Generator Level M R 11.- Cer.teni enri pacitinn L4-it Eitet.; , + --R.
- 12. ;;;; nsw ti -M--- --R-
- 12. RiiRT qcr2t= - -M- -ft-m 4' a r4142ru re mp3+or rinw pato
_g-_ _g tvl ' 6ene. Ac;J Tank Leve.l
- 9. t1 R v.
(D
INSTRUMENTATION ACCIDEN'T MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 303.3.6 The accident monitoring instrumentation channels shown in Table 3.3-10 shall be OPERABLE. APPLICABILITY: MODES 1, 2, and 3. ACTION:
- a. With the number of OPERABLE accident monitoring instrumentation channels less than the Required Number of Channels shown in Table 3.3-10, restore the inoperable channel (s) to OPERABLE status within 7 days, or be in at least HOT SHUTDOWN within the next 12 hours.
- b. With the number of OPERABLE accident monitoring instrumentation channels less than the MINIMUM CHANNELS OPERABLE requirements of Table 3.3-10, restore the inoperable channel (s) to OPERABLE status within 48 hours or be in at least HOT SHUTDOWN within the next 12 hours.
- c. The provisions of Specification 3.0.4 are not applicable.
SURVEILLA'NCE REQUIREMENTS 4.3.3.6 Each accident monitoring instrumentation channel shall be demonstrated OPERABLE by, performance of the CHANNEL CHECK and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3-7. - l W-STS 3/43h SEP 151981 l
TABLE 3.3-10 9 y ACCIDENT MONITORING INSTRUMENTATION REQUIRED MINIMUM NO. OF CHANNELS INSTRUMENT (Illustrational Only) CHANNELS OPERABLE
- 1. Containment Pressure l
{ '
- 2. Reactor Coolant Outlet Temperature - TH0T (Wide Ralnge) 2\ 1
- 3. Reactor Coolant Inlet Temperature - TCOLD (Wide Range) 2 1
- 4. Reactor Coolant Pressure - Wide Range 2 1 S. Pressurizer Water Level 2 1
- 6. Steam Line Pressure EO 1/st generator
, _-( ,* 2/ steam ge erator s p
[ 7. Steam Generator Water Level - Harrow Range a. f 1/ steam gene ator 1/ eam generator k 8. Steam Generator Water Level - Wide Range w E ;5 , 1/ steam genera or / steam generator
- 9. Refueling Water Storage Tank Water Level 4O2
-s 1
- 10. Boric Acid Tank Solution Level 2 1 Emeqrocy +
- 11. Auxil; ry Feedater Flow Rate Pq 2/ steam gener or 1/ steam generator Y.
- 12. Reactor Coolant System Subcooling Margin Monitor D 2 1
- 13. PORV Position Indicator b' 2/ Valve 1/V Ive
- 14. PORV Block Valve Position Indicator
$Y G- 2/ Valve 1/ Val $ 15. Safety Valve Position Indicator p 2/Valv 1/ Valve N (6 e-
[ 16. Containment Water Level (Narrow Range) E2 1
$ 17. Containment Water Level (Wide Range) 2 1
- 18. In Core Thermocouples / core quadrant 2/ core quadrar.
TABLE 4.3-7
'?
y . ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL - CHANNEL INSTRUMENT (Illustrational Only) CHECK CALIBRATION
- 1. Containment. Pressure
- 2. Reactor Coolant Outlet Temperature - THOT (Wide Range) R
- 3. Reactor Coo,lant Inlet Temperature - TCOLD (Wide Range) M R
- 4. Reactor Coolant Pressure - Wide Range M R
- 5. Pressurizer Water Level M R s- b
- 6. Steam Line Pressure 8 0 M R R
- 7. Steam Generator Water Level - Harrow Range 1%
hM R Y 8. Steam Generator Water Level - Wide Range E * ,_ . M R
~>
ls 9. Refueling Water Storage Tank Water Level M R
- 10. Boric Acid Tank Solution Level Em e.r D4 (f
-d.
M R
- 11. ^r"geuy.ary Feedwater Flow Rate h}
M R
- 12. Reactor Coolant System Subcooling Margin Monitor 0-pM R
- 13. PORV Position Indicator h. g, M R
- 14. PORV Block Valve Position Indicator M R o
- 15. Safety Valve Position Indicator g- F t s
- 16. Containment Water Level (Harrow Range) p :
m 17. Containment Water Level (Wide Range) M F
- 18. In Core Thermocouples M i
(
s INSTRUMENTATION FIRE DETECTION INSTRUMENTATION LIMITING CONDITION FOR OPERATION 7 3.3.3.1 As a minimum, the fire detection instrumentation for each fire detection zcne shown in Table 3.3-11 shall be OPERABLE. . APPLICABILITY: Whenever equipment protected by the fire detection instrument is required to.be OPERABLE.
~
ACTION: With the number of OPERABLE fire detection instrument (s) less than the minimum t number OPERABLE requirement of Table 3.3-11:
- a. Within 1 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 the
- containment at least once per 8 hours or (monitor the containment air temperature at least once per hour at the locations listed in Specification 4.6.1.6).
- b. Restore the inoperable instrument (s) to OPERABLE status within 14 days, or in lieu' of any other report required by Specification 6.9.1, prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within the next 30 days outlining the action taken, the cause of the inoperability and the. plans and schedule for restoring the instrument (s) to OPERABLE status.
- c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 7 , 4.3.3.i.1 Each of the above required fire detection instruments which are l accessible during plant operation shall be demonstrated OPERABLE at least once per 6 mon'th's by performance of a TRIP ACTUATING DEVICE OPERATIONAL TEST. Fire i detectors which are not accessible during plant operation shall be demonstrated l OPERABLE by the performance of a TRIP ACTUATING DEVICE OPERATIONAL TEST during each COLD SHUTOOWN exceeding 24 hours unless performed in the previous 6 months. 4.3.3. 2 The NFPA Standard 720 supervised circuits supervision associated with the detector alarms of each of the above required fire detection instruments shall be demonstrated OPERABLE at least once per 6 months. ! 7 l 4.3.3.i.3 The nonsupervised circuits,. associated with detector alarms, between the instrument and the control room shall be demonstrated OPERABLE at least once per 31 days. W-STS 3/4 3-1HI l SEP.1 '5 1981 v.. - -
TABLE 3.3-11 FIRE DETECTION INSTRUMENTS INSTRUMENT LOCATION (Illustrative **) MINIMUM INSTRUMENTS OPERABLE
- HEAT FLAME SMOKE
- 1. Containment Zone 1 Elevation Zone 2 Elevation f ' '
I 2. Control Room O ts irtfo r e a fts a b b C-
- 3. Cable Spreading supplies af a la+c dake-Zone 1 Elevation Zone 2 Elevation ,
- 4. Computer Room
- 5. Switchgear Room
- 6. Remote Shutdown Panels
- 7. Station Battery Rooms Zone 1 Elevation Zone 2 Elevation
- 8. Turbine Zone 1 Elevation Zone 2 Elevation .
- 9. Diesel Generator Zone 1 Elevation
- Zone 2 Elevation
- 10. Diesel Fuel Storage
- 11. Safety Related Pumps Zone 1 Elevation Zone 2 Elevation
- 12. Fuel Storage Zone 1 Elevation Zone 2 Elevation "The fire detection instruments located within the Containment are not required to be OPERABLE during the performance of Type A Containment Leakage Rate Tests.
** List all detectors in areas required to insure the OPERABILITY of Safety related equipment and indicate instruments which automatically actuate fire suppression systems.
47 W-STS 3/4 3-M SEP . 51981
-- -4.,
t INSTRUMENTATION LOOSE-PART DETECTION INSTRUMENTATION _ _ LIMITING CONDITION FOR OPERATION . 3.3.3. The loose part detection system shall be OPERABLE. APPLICABILITY: MODES 1 and 2 , l ACITON:
- a. With one or more loose part detection system channels inoperable for more than 30 days, prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within the next 10 days outlining the cause of the malfunction and the plans for restoring the channel (s) to OPERABLE status.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
J SURVEILLANCE REQUIREMENTS + ' 8 4.3.3.'1 Each channel of the loose part detection systems shall be demonstrat.ed OPERABLE by performance of:
- a. A CHANNEL. CHECK at least once per 24 hours,
- b. An ANALOG CHANNEL OPERATIONAL TEST at least once per 31 days, and
- c. A CHANNEL CALIBRATION at least once per 18 months.
PSTS 3/43-h .NOV 2 1981
INSTRUMENTATION RADIOACTIVE LIOUID EFFLUENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 9
- 3. 3. 3.M The radioactive liquid effluent monitoring instrumentation channels shown in Table 3.3-12 shall be OPERABLE with their alarm / trip setpoints set to ensure that the limits of Specification 3.11.1.1 are not exceeded. The alarm /
trip setpoints of these channels shall be determined and adjusted in accordance with the methodology and parameters in the OFFSITE DOSE CALCULATION MANUAL (00CM). APPLICABILITY: At all times. . ACTION:
- a. With a radioactive liquid effluent monitoring instrumentation channel alarm / trip setpoint less conservative than required by the above Specification, immediately suspend the release of radioactive liquid effluents monitored by the affected channel, or declare the channel inoperable.
- b. With less than the minimum number of radioactive liquid effluent monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 3.3-12. Restore the inoperable instrumentation to OPERABLE status within the time specified in the ACTION or, ' 'i= ef ;
-Lin .;;; C..c.; ".,a t, explain in the next Semiannual Radioactive Effluent Release Report why this inoperability was not corrected within the time specified. .
- c. The provisions of Specifications 3.0. 0.4, ed 5.9.1.9.l are not applicable.
SURVEILLANCE REOUIREMENTS 4.3.3. Each radioactive liquid effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION and CHA;4:4-(L TW4CTIC :AL TZ? operations at the frequencies shown in Table 4.3-TS. I meto6 A camust. opeurww . Test 9 . 1 l 1 I l - l l 69 l PWR-STS-REIS . 3/4 3-h 1/4/83
l l
$ kkN k c
U n 2 16 a 50 x ;4 A
*C WW JJ Z w cc i C Z *C m m l
i w W
= Z CE Z eC W e.4 c=4 M w.4 e
f 4 - ::lll A
> EUC Z
w ca: 'N N W
- 2
~
*3
- .3 K o# .$ $ , o
!E
~
e w" D i v e e e a > . U u N r D tr c o -4 R= ! M - Z w *+= OJ
=
CE c W c U
.o s [
m i o C:-5 - 3
. E % W J O M '
o M
+ N, Q %
w C
- .: a a w Q 3 = LL W o c u. as it o- O T si a
Z w 3 c Z d c: m W 3 W 2 m - c hq ' --c: g,
,.I CE3 4
W
-.a 6
et w
- l CD W
w
-A ==
C p a L (1) s 6 2 '.%gp *I W E C % w A a e eJ 1: w cz: b C: ct: - a E s o 4(. -
-+-
<J ci eoi < W c bqu m
c :I a :,2 4 Y
= <WW <1 -* Cl ef".
W W e.& C C QJ 3 Q> c; et-w 3 C C %= *. CP c aC ZW w=
- 4 LJ fil.
L. C ec
*2W
-a QJ 3
Z eC W w,
- l
*g
{ 9O Q% J -a c:i - %= 4 w Cw ea c cw % W C w .i C W wm ll> t"
<J 3
O 1. C> m ct; W 6 CJ 3 W 3 C eU e w U g: C I 3 C $ 3
=9 Q
- CW :3 7 C) O6 J a T C. . = 7 3 e
- - =o -- 3 -- cx o E .,o c 3 E- > -- 2 o ..
l U Z c. %- o u m
.cJ "E E o WG
= w c
e- o
- * =
- S -+- d asC W Z %- - M.., 2 .a eC -E %. f C
- E WC cm -
WJ CE2 WC ei cm - m Cl C W J CE2 .Q.a ei a QJ
>=
WJ ::52 \J wm C cK QH W L C QH W ** W Le c ** Z W L -. g GJ eC >= H et 48 C - >= *E" - >- C **= w +J nl g M W +d C CC E m .OJ :', CE wZ m *J V -Z Z~ O C W - ed w to n: 3 Z Z~ to en - 90 *J to 4
- OE 2 6 == oE U C 6 - cz: 3 6 c '
w E M. "C3 43 3 E CE C 4J W D t c 3T w to C cn w +J E C +J 3 .WJ W re = ci 5 o
>= H CK W >= H C C
- Qt r0 r5 aC CK QJ C1 d .et H C4ll' 4J H 47 QJ U CC C W G Li 3
mU C wU U C 5= b E 7 r8 U ll> w .T.= 3 - lll> w w= c W 5 4; . QJ w= E ::$ wH 3 r0 c3 m >== > C. 3 f5 +4 *J W = tO Q y.= e H C" 4A b <= eC 5= E O QJ - - P- C" CJ @ U w ed ? t.) GJ O D e.J r0 90 *C n.- e.J =* d aut" o-J W * *C ca W W Z 4.h wa Ct: J W C$1 [g m3 w - *3 C eC C Z C eC . . eC . . o . . J . . . $ O(,J ck: 4J
, to .lll2 U FO .O 6 rU . ll2 U $g 70 PWR-STS-RETS 3/4 3-X 1/4/83
TABLE 3.3-12 (Continued) I - T RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION 0 T MINIMUM A d CilANNELS OPERABLE ACTION INSTRUMENT S. RADluatiiviT "renonres*
- a. Liquid Radwaste Effluent Line 1 28
- b. Steam Gener n ffluent Line 1 larm/ trip set point is based on recorder-controller.
{
- Required on y ' ~
u l .
/
/
,/
t 8 l l
TABLE 3.3-12 (Continutd) TABLE NOTATION b ACTION'5'8- With the number of channels OPERABLE less than r:;uir:d by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 14 days provided that prior to initiating a release:
- a. At least two independent samples are analyzed in accordance with Specification 4.11.1.1.1, and
- b. At least two technically qualified members of the Facility Staff independently verify the release rate calculations and discharge line valving; Otherwise, suspend release of radioactive effluents via this
. pathway.
.D ACTION iML - With the number of channels OPERABLE less than -requirci by- the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided grab samples are analyzed for radioactivity for up to 30 days at a lower limit of detection of no more than 10 7 microcurie /ml:
- a. At least once per 12 hours when the specific activity of the-secondary coolant is greater than 0.01 microcurie / gram DOSE EQUIVALENT I-131.
- b. At least once per 24 hours when the specific activity t of the secondary coolant is less than or equal to 0.01 microcurie / gram DOSE EQUIVALENT I-131.
l With the n. umber of channels OPERABLE less than required by th Mini OPERABLE requirement, effluent a a this pathway may c u rovided that, at least once per 12 h sa llected and analyzed for ~ radi
. at a lower limit of detection t crocurie/ml.
2.8 ACTION lis - With the number of channels OPERABLE less than equi.;J Ly the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided the flow rate is estimated at least once per 4 hours during actual releases. Pump performance curves generated in place may be l used to estimate flow.
'- With the. number of channels OPERABLE less than required b PERABLE requirement, efflue '
via this pathway may con inu ovided the radioacti is determined at least once pe h l actual releases. PWR-STS-RETS 3/4 3 1/4/83 Ji
J o 7 ii TABLE 4.3- h - 2 7 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRtMENTATION SURVEILLANCE REQUIREMENTS
- h. CHANNEL FUNCTIONAL -
h , CHANNEL SOURCE CHECK CHANNEL CALIBRATION TEST i INSTRtMENT CHECK 1
. 1. RADI0 ACTIVITY MONITORS PROVIDING ALARM AND AUTOMATIC TERMINATION OF RELEASE I Test m k pischan c 2.
- a. Liquid Radwaste mii'oc..;. Li ..e D P R(M Q(1)
Fla Ef'gk't~aak. Dens'n 1-.
- b. Steam Generator Blowdown .rrnt LP- D M ROL) Q(1)
- c. Turbine Building (F!cer 0 =i:0 Sumps 2_ '
w Effluent Line D H R(N Q(1) D RADI0 ACTIVITY MONITORS PROVIDING ALARM BUT
?J r OVIDING AUTOMATIC TERMINATION OF REL 1
- a. Service Water Sy Effluent Line D M R(3)
- b. Component Cooling Water Sy Effluent Line D M 3) Q(2) 1
- 3. CONTINUOUS COMPOSITE SAMPLERS AND SAMPLER FLOW HONITOR
- a. ' Steam Generator Blowdown Eff e d ne D N. . R Q 4 ,(alternate to item 1.
U N.A. N'
- R b. Turbine ing Sumps Effluent Line D R '
I
$ nate to item 1.c)
Y
. TABLE 4.3-W (Continued) o h
n h RADI0 ACTIVE LIQUID EFFLUENT HONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS T g CilANNEL FUNCTIONAL CilANNEL SOURCE CHANNEL - d CilECK CllECK CALIBRATION TEST INSTRUMENT 2.4. FLOW RATE HEASUREMENT DEVICES T<st Taft DecAaSc 3
- a. Liquid Radwaste-Effl;;nt Lin: D(t) N.A. R (2-) Q O)
FlashTaalt. > min 3
- b. Steam Generator Blowdown Ef'! ent L n: D(O N.A. R Q-) Q 6)
- c. "i;charg: Car.al- Gduldk hkfcuDuk9 , Dt44 N.A. -R-N4 %#4
;. aaninacTIVITY RECORDERS
- x y
~
Liquid Radwaste Effluent Line n ,,___ J o _ [ a. _ o N.A. n 0
- b. Ste @ eratoPBT6Fdo' n D _
*he 'ectad a nn ps;c 2/d 2 - 7 2 .-
farp davves am uhlhjej }o eshn/wflow t 9
TABLE 4.3- (Continued) TABLE NOTATION (1) The CHANNEL FUNCTIONAL TEST shall also demonstrate that automatic isolation of this pathway and control room alarm annunciation occur if any of the following conditions exists:
- 1. Instrument indicates measured levels above the alarm / trip setpoint.
- 2. Circuit failure.
- 3. Instrument indicates a downscale failure.
- 4. Instrument controls not set in operate mode.
e CHANNEL FUNCTIONAL TEST shall also demonstrate that control ala iation occurs if any of the following conditio xists: ( 1. Instrument indic sured levels above alarm setpoint.
- 2. Circuit failure.
- 3. Instrument ind' a downscale failure.
4 rument controls not set in operate mode. 2.
^(%) The initial CHANNEL CALIBRATION shall be performed using one or more of the reference standards certified by the National Bureau of Standards or using standards that have been obtained from suppliers that participate in measurement assurance activities with NBS. These standards shall permit calibrating the system over its intended range of energy and measurement range. For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration shall be used.
( CHANNEL CHECK shall consist of verifying indication of flow during periods
- of release. CHANNEL CHECK shall be made at least once per 24 hours on I days on which continuous, periodic, or batch releases are made.
l l l l PWR-STS-RETS 3/4 3- 1/4/83 l . l l
- s l l
i INSTRUMENTATION I RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION
. _ _ _ _ _ LIMITING CONDITION FOR OPERATION 3.3.3. The radioactive gaseous effluent monitoring instrumentation channels shown in Table 3.3-13 shall be OPERABLE with their alarm / trip setpoints set to ensure that the limits of Specification 3.11.2.1 are not exceeded. The alarm / trip setpoints of these channels shall be determined and adjusted in accordance with the methodology and parameters in the ODCM.
l APPLICABILITY: As shown in Table 3.3-13 ACTION:
- a. With a radioactive gaseous effluent monitoring instrumentation channel alarm / 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 less than the minimum number of radioactive gaseous effluent monitoring instrumentation channels OPERABLE; take the ACTION shown in Tacle 3.3-13. Restore the inoperable instrumentation to OPERABLE status within the time specified in the ACTION or, 4- icu Of :-
Liceart: E;e..t Repert, explain in the next Semiannual Radioactive Effluent Release Report why,this inoperability was not corrected r within the time specified. l ! c. The provisions of Specifications 3.0.3, 3.0.4, and 6.9.1.9.b are not applicable. SURVEILLANCE REOUIREMENTS 10 4.3.3.14, Each radioactive. gaseous effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations at the frequencies shown in Table 4.3-19L . 3 l 7f 1/4/83 PWR-STS-RETS . 3/4 3-7g
TABLE 3.3-13 I T RADI0 ACTIVE GASE0US EFFLUENT MONITORING INSTRUMENTATION *
l T
MINIMUM CHANNELS 3 INSTRUMENT OPERABLE APPLICABILITY ACTION
. TE GAS Il0LDUP SYSTEM
- a. Not[Je Livity Monitor -
Providing Ala d Automatic Termination of Rele 1 *- [ b. Iodine Sampler . 1 41
- c. Particulate Sampler
- 41 m d. Effluent System Flow Rate
) Measuring Device 1 36 w
p er Flow Rate Measuring Device 1 3 h' 2. fMaso4crwe.o.4s wasre X. WA';IE CfS l'^LO'J" SYSTEM EXPLOSIVE GAS MONITORING SYSTEM (for systems designed h to withstand the effects of a hydrogen explosion)
- c. !!yd;cgcr WPiter (f,uta;;;atic ccatrel)- -I " +
llyd.ogcn er Oxygen Monitor (Process) ** 4% 1 39 31 ASTE GAS ll0LDUP SYSTEM EXPLOSIVE GAS I STEM (for systems not designed to wi effects of _ _ - y a hydrogen explosion) _ 4
'h a. Ilydrogen Monitors (Automatic control, redundant)
Q*% %
** 40, 42 OxytJEn Monitors (Process, **
- b. 2
TABLE 3.3-13 (Continued) N RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSiRUMENTATION m T g MINIMUM CilANNELS APPLICABILITY ACTI OPERABLE d INSTRUMENT
- 3. CONDENSER EVACUATION SYSTEM
- 37
- a. ble Gas Activity Monitor 1
- 41
- b. Iodin ampler 1
- c. Particulate ampler 1 41
- 36
- d. Flow Rate Monito 1
- 36
, e. Sampler Flow Rate Moni r 1
- 4. VENT llEADER SYSTEM
- 37
- a. Noble Gas Activity Monitor
- 41
- b. Iodine Sampler 1
- 41
- c. Particulate Sampler 1
- 36
- d. Flow Rate Monitor 1
- 36
- e. Sampler Flow Rate Monitor 1
- 5. CONTAINMENT PURGE SYS1 t' a. Noble Gas Act' ity Monitor - Providing
- Alarm and omatic Termination of Release 1 38 R
O -
- 41
- b. Iodi Sampler 1
- 4
- c. articulate Sampler 1 s
TABLE 3.3-13 (Continued) I y RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION
~
n m E MINIMUM CilANNELS . 3 INSTRUMENT OPERABLE APPLICABILITY ACTION ST CCTa mMFNT PURGE SYSTEM (Continued)
~
f g d. Flow Rate Monitor
'
- 36 1 2' _
t LE AUXILIARY CUILDING VENTILATIGN-
-S;'ST04- Pte ur veArr-m a. . Noble Gas Activity Monitor 1 Wo s
- h. Iodine Sampler 1 4LS2 ha c. Particul' ate Sampler 1
% 31
- d. Flow Rate Monitor 1 %%9
* %2,9
- e. Sampler Flow Rate Monitor 1 4
.L STORAGE AREA VENTILATION SYSTEM
- a. Noble Gas Ac i or 1
- b. Iodine Sampler 1 41
@ c. Particulate Sampler 1 41 C d. Flow Rate Monitor 1
^
36 R oo
- e. r low Rate Monitor 1
TABLE 3.3-13 (Centinued) I y RAD!0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION 4 T
= MINIMUM CilANNELS 3 INSTRUMENT OPERABLE APPLICABILITY ACTION -
. ASTE AREA VENTILATION SYSTEM
- a. Noble Ga vity Monitor 1
- b. Iodine Sampler 1 41
@ c. Particulate Sampler 1 41
- d. ' Flow Rate Monitor 1 36 m e. Sam w Rate Monitor 1 36 s
Y Neo % a3 0 tiler EXilAUST AND VENT SYSTEMS such as:
" TEA". GENE"f. TOR GLOWOOWN "[N! SYST[i;,
TURBINE GLAND SEAL CONDENSER EXilAUST
@ 2. McH e Cn Au;.i.ity "enits- - -i--
Q,t. Iodine Sampler 1 %32. , b% Particulate Sampler 1
% 32.
-d ."lew " ate ugggter -*--
- c, x. Sampler Flow Rate Monitor 1 3429 C ,
t S
TABLE 3.3-13 (Continued) I TABLE NOTATION
- At all times.
** During WASTE GAS HOLDUP SYSTEM operation.
ON 35 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, the contents of the tank (s) may be released to the environment for up to 1 ys vided that prior to initiating the release:
- a. At two independent sampi the tank's contents are analyz and
- b. At least two techn' ualified members of the Facility Staff indepen y verify release rate calculations and disc valve lineup; Othe e, suspend release of radioactive effluents + is Way.
ACTION 31ii - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided the flow rate is estimated at least once per 4 hours. ACTION N - With the number'of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided grab samples are taken at least once per 12 hours and these samples are analyzed for gross activity within 24 hours. ka m ' '~ the number of channels OPERABLE les .' . immediately suspend
@ Minimum ,
oac 1ve effluents'via this pa w . 31 ACTION 3(~ - With the number of channels OPERABLE less than required by the i Minimum Channels OPERABLE requirement, operation of this WASTE I GAS HOLDUP SYSTEM may continue for up to 30 days provided grab samples are collected at least once per 4 hours and analyzed l within the following 4 hours. ith the number of channnels OPERABLE one less than the OPERABL ion of this h system ma
. With two channels e, be in at least HOT STANDBY wi .
c PWR-STS-RETS 3/4 3-M 1/4/83 1
i l TABLE 3.3-13 (Continued) TABLE NOTATION 37-ACTION 1R(- With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via the effected pathway may continue for up to 30 days provided samples are continuously collected with auxiliary sampling i equipment as required in Table 4.11-2. ccT::" do - With the numhar of channels OPERA 8LF na^ ' . . . o ai . isQuirea oy
~ ~ ' ' ' -
the Minimum Channeic r-auirement, suspend oxygen supply
@ b .n= recomoiner.
l - 1 l 1 i PWR-STS-RETS 3/43-If . 1/4/83 l
9 TABLE 4.3-X 2 7 RADIDACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS ,
-i m CHANNEL MODES IN WHICH g CHANNEL SOURCE CHANNEL FUNCTIONAL SURVEILLANCE CHECK CALIBRATION TEST REQUIRED d '
INSTRUMENT CHECK
. E GAS HOLDUP SYSTEM
- a. Noble ivity Monitor -
Providing Alar' utomatic Termination of Release' ~. P P R(3) Q(1) h b. Iodine Sampler ~ 'N ' N.A. N.A. N N.A.
- c. Particulate Sampl'er W .
x N
- d. Effluent System Flow Rate P ..
~ - -
.. Q Measuring Device w e. Sampler F asuring D N.A. R Q k e Y 2A. WASTE GAS Il0LDUP SYSTEM EXPLOSIVE k GAS MONITORING SYSTEM (for systems designed to withstand the effects of a hydrogen explosion) h -2. "ydregen ."sa'ter (.^uto eLiu
-centre!)
-B-- -N-A- - Q(4) - -H-A.% lydr: gen er Oxygen Monitor D N.A. Q(k)orQ(k) M (Process)
- M M AS HOLDUP SYSTEM EXPLOSIVE .
GASMONITORTflG=6YSTEMforsystems g not designed to withsta7(Pthr%. ,_ g effects of a hydrogen explosion) % _ U " # # " %
5 $ 2i" % "S E't$^"' "* 1' i "
c ._
- b. Ilydr gesper"Diigenlo$it rs D N.A. Q(4) or Q(5) M ~
rocess, dual)
9 TABLE 4.3-h (Ccntinued) I y RADI0ACTICE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS n
- T g CllANNEL MODES IN milch CilANNEL FUNCTIONAL SURVEILLANCE d CllANNEL CllECK SOURCE CilECK CALIBRATION TEST REQUIRED INSTRUMENT ,
CONTAINHENT PURGE SYSTEM
- a. No Activhy Monitor -
Providing and Automatic Termination of Re , D P R(3) Q(1)
- b. Iodine Sampler N.A. N.A. N.A.
- c. Particulate Sampler W . . .A. N.A.
$ d. Flow Rate Monitor D N.A. R Y e er Flow Rate Monitor D N.A. R Q N%N
~2 1 t,UXIL!?"Y SU!LO!;;G VO47:LAT10G
-SMTEM- PLAAir VEM
- a. Noble Gas Actvity Monitor D H R(h Q( )
h b. Iodine Sampler W N.A. N.A. N.A. *
- c. Particulate Sampler W N.A. N.A. N.A.
g d. Flow Rate Monitor D N.A. R Q
- e. Sampler Flow Rate Monitor D N.A. R Q t
@ 4 TABLE 4.3-h (Continued) 2 .
7 RADI0 ACTIVE GASEQUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS "I T
$ CilANNEL Moff IN WHICH d CHANNEL SOURCE CilANNEL FUNCTIONAL SDRVEILLANCE INSTRUMENT CHECK CHECK CALIBRATION TEST REQUIRED
- 3. CONDENSER EVACUATION TEM
- a. Noble Gas Activity Moni r D M R(3) 2.)
- b. Iodine Sampler W N.A. H.A. N.A.
- c. Particulate Sampler N.A. N.A. N.A.
- d. Flow Rate Monitor D H.A. Q b Sampler Flow Rate Monitor N.
_ e. D R Q
< > /
8 ( o" 4. VENT llEADER SYSTEM i
- a. Noble Gas Activity Monitor D M R(3) Q(2)
- b. Iodine Sampler W N.A. N.A. .
- c. Particulate Sampler W N.A. H.A. N.A.
- d. Flow Rate Monitor D N.A. R Q
- e. SamplerFlowRpt[ Monitor D N.A. R Q t
5 f
f TABLE 4.3-M (Continuid) 2 7
@ RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS i'l T " CHANNEL MODE N WHICH
- h. CHANNEL SOURCE CHANNEL FUNCTIONAL juRVEILLANCE IESTRUMENT CHECK CHECK CALIBRATION TEST / REQUIRED
- 7. Fu . ORAGE AREA VENTILATION SYSTEM
- a. Noble Gas 'vity Monitor D M R(3) -- Q(2)
- b. Iodine Sampler W N.A. N.A. N.A.
- c. Particulate Sampler W N.A. N.A N.A.
/' *
- d. Flow Rate Monitor N.A. R Q
- e. Sampler Flow Rate Monitor D N.Af
/
R Q
- 8. RADWASTE AREA VENTILATION SYSTEM
- a. Noble Gas Activity Monitor 0 M R( Q(2)
/
- b. Iodine Sampler /W N.A. N.A. N.A *
- c. Particulate Sampler W N.A. N.A. N.
l
- d. Flow Rate Monitor D H.A. R Q
- e. SamplerF1owRat[ Monitor D N.A. R Q l t l
TABLE 4.3- -(Ccntinued) 2 y RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS "I T - CHANNEL MODES IN WHICH A CC) CilANNEL FUNCTIONAL SURVEILLANCE CilANNEL SOURCE d CilECK CllECK CALIBRATION TEST REQUIRED INSTRUMENT 3
- h. 0 tiler EXilAUST AND VENT SYSTEMS such as: .
iTE" CENERATOR OLOWCOWN VENT-SYSTEe;, TURBINE GLAND SEAL CONDENSER EXilAUST
-c. ";M : C:: "ctivity ;L, iter
. -&- -M-- M - Q ' 2 ) --
- 4. K Iodine Sampler W N.A. N.A. N.A w *
' .b 4. Particulate Sampler W N.A. N.A. N.A.
E$ + 4 r!ee Date "er!ter t-- 4-A- h -Q--- CK Sampler Flow Rate Monitor D N.A. R Q t 8 .
9 TABLE 4.3-M (Continued) TABLE NOTATION
- At all times.
RhDI0hCTWE GAswMTE SysTEH
@ ** During 'JITE C?i "0LO"I 0'f5T:" aperation.
The CHANNEL FUNCTIONAL TEST shall also demonstrate that automatic ' tion his pathway and control room alarm annunciation occurs if of the follo conditions exists:
- 1. Instrument in ' s measured levels ab e alarm / trip setpoint.
g 2. Circuit failure. Instrument i es a downscale failure. 3.
- 4. rument controls not set in operate mode.
1 OL) The CHANNEL FUNCTIONAL TEST shall also demonstrate that control room alarm annundiation occurs if any of the following conditions exists:
- 1. Instrument indicates measured levels above the alarm setpoint.
- 2. Circuit failure.
- 3. Instrument indicates a downscale failure.
- 4. Instrument controls not set in operate mode.
2 M) The initial CHANNEL CALIBRATION shall be performed using one or more of the reference standards certified by the National Bureau of Standards or using standards that have been obtained from suppliers that participate in measurement assurance activities with NBS. These standards shall permit calibrating the system over its. intended range of energy and measurement ran'ge. For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration shall be used. 3 04.) The CHANNEL CALIBRATION shall include the use of standard gas samples containing a nominal: ,
- 1. One volume percent hydrogen, balance nitrogen, and
- 2. Four volume percent hydrogen, balance nitrogen.
4(5) The CHANNEL CALIBRATION shall include the use of standard gas samples containing a nominal:
- 1. One volume percent oxygen, balance nitrogen, and
- 2. Four volume percent oxygen, balance nitrogen.
S4 PWR-STS-RETS 3/4 3-M 1/4/83
INSTRUMENTATION 3/4.3.4 TURBINE OVERSPEED PROTECTION LIMITING CONDITION FOR OPERATION 3.3.4 At least one turbine overspeed protection system shall be OPERABLE. APPICABILITY: MODES 1, 2, and 3. ACTION: krm.ed con b l
- a. Withonestopvalveorone;r::r.:rvalvegerhighpressureturbine steam lead inoperable and/or with one n... t stop valve or one reheat intercept valve per low pressure turbine steam lead inoperable, restore the inoperable valve (s) to OPERABLE status within 72 hours, or close at least one valve in the affected steam lead (s) or isolate the turbine from the steam supply within the next 6 hours.
- b. With the above required turbine overspeed protection system otherwise inoperable, within 6 hours isolate the turbine from the steam supply.
SURVEILLANCE REQUIREMENTS
$ce. 7tas4tftca.Y}cr1 h
.4.1 The provisions of Specification 4.0.4 are not applicable.
4.3.4. The above required turbine overspeed protection system 11 be demonstra OPERABLE:
- a. At le t once per 7 days by cycling each of the ollowing valves through least one complete cycle from the nning position.
- 1. (Four) h pressure turbine stop va s.
- 2. (Four) high ressure turbine gove or valves.
l 3. (Four) low pres e turbine r at stop valves.
- 4. (Four) low pressure rbi reheat intercept valves.
- b. At least once per 31 days di observation of the movement of each of the above valve brough on omplete cycle from the running position.
I l c. At least once p 18 months by performance of CHANNEL CALIBRATION on the turbi overspeed protection systems.
- d. At leas once per 40 months by disassembling at least on f each of the ve valves and performing a visual and surface inspect of v e seats, disks and stems and verifying no unacceptable flaw r l orrosion.
W-STS 3/4 3- NOV 2 1981 t _ _ _____-._ _ _
JUSTIFICATIONS Section 3/4.4 In the text of Section 3/4.4 where a capital letter with a circle around it appears, please refer to this sheet for the appropriate justification. A. Seabrook Station plant specific data. B. Manual transfer from " normal" to " emergency" power at Seabrook is not applicable. Certain of the pressurizer heater are permanently powered from an IE bus. C. The power supply for the block valves is always on an emergency bus which will automatically transfer to the diesel generator. PORV's are provided with a IE direct current power supply. D. Added to provide criteria for sleeving or plugging E. To clarify acceptance criteria for 4.4.5.4.a.10 F. This is not required for nor applicable to the Seabrook design. G. There are no loop isolation valves in the Seabrook design.
~
H. Allowable leakage is 13 gpa per reactor coolant pump seal for a total of 52 gpa. I. It appears this was meant for plants with automatic modulating valves. Seabrook does not utilize automative modulating valves but instead has-manual throttle valves. J. This statement added for clarification. Normally, the plant will be i passing through Modes 3 and 4, either heatin8 up or cooling down, and would not be in a steady state condition during this time. K. Seabrook design utilizes a variable PORV getpoint for LTOP which is dependent on RCS temperature. The 3.2 in vent was determined during transient analysis. L. Added per requirement of Generic Letter 83-37 5
3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION STARTUP AND POWER OPERATION LIMITING CONDITION FOR OPERATION 3.4.1.1 All reactor coolant icops 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 STANDBY within 1 hour. SURVEILLANCE REQUIREMENT 4.4.1.1 The above required reactor coolant loops shall be verified to be in operation and circulating reactor coolant at least once per 12 hours. l "See Special Test Exception 3.10.4. i W-STS 3/4 4-1 MAY 151980 L
REACTOR COOLANT SYSTEM HOT STANOBY LIMITING CONDITION FOR OPERATION 3.4.1.2 a. At least two of the Reactor Coolan't loops listed below shall ( l be OPERABLE: l 1. Reactor Coolant Loop (A) and its associated steam generator and Reactor Coolant pump,-
- 2. Reactor Coolant Loop (B) and its associated steam generator and Reactor Coolant pump,
- 3. Reactor Coolant Loop (C) and its associated steam generator and Reactor Coolant pump,
- 4. Reactor Coolant Loop (D) and its associated steam generator i and Reactor Coolant pump.
- b. At least one of the above Reactor Coolant loops shall be in operation.*
i APPLICABILITY: MODE 3 ACTION:
- a. With less than the above required Reactor Coolant loops OPERABLE, restore the required loops to OPERABLE status within 72 hours or be in HOT SHUTDOWN within the next 12 hours.
- b. With no Reactor Coolant loop in operation, suspend all operations
( involving a reduction in boron concentration of the Reactor Coolant L System and immediately initiate corrective action to return the required Reactor Coolant loop to operation. SURVEILLANCE REQUIREMENTS 4.4.1.2.1 At least the above required Reactor Coolant pumps, if not in operation, shall be determined to be OPERABLE once per 7 days by verifying correct breaket a'lignments and indicated power availability. 4.4.1.2.2 The required steam generators shall be determined OPERABLE by verifying secondary side water level to be greater than or equal to (475) at least once per 12 hours. 11.1 7, (@) 4.4.1.2.3 At least one Reactor Coolant loop shall be verified to be in operation
. and circulating reactor coolant at least once per 12 hours.
"All Reactor Coolant pumps may be de energized for up to I hour provided i (1) no operations are permitted that would cause dilution of the reactor I
coolant system boron concentration, and (2) core outlet temperature is maintained at least 10'F below saturation temperature. l-W-STS 3/4 4-2 NOV _2 61980 _ _ . _ _ . _ _ _ . _ _ - - - _ . . _ _ _ . _ - . - - - - _ . _ - ~ -.
, s REACTOR COOLANT SYSTEM HOT SHUTDOWN LIMITING CONDITION FOR OPERATION 3.4.1.3 a. At least two of the Reactor Coolant and/or residual heat removal ,
(RHR) loops listed below shall be OPERABLE:
- 1. Reactor Coolant Loop (A) and its associated steam generator and reactor coolant pump,*
- 2. Reactor Coolant Loop (B) and its associated steam generator and reactor coolant pump,* ,
- 3. Reactor Coolant Loop (C) and its associated steam generator and reactor coolant pump,*
- 4. Reactor Coolant Loop (D) and its associated steam generator and reactor coolant pump,*
- 5. Residual Heat Removal Loop (A),
- 6. Residual Heat Removal Loop (B).
l ( b. At least one of the above Reactor Coolant and/or RHR loops shall be in operation.** APPLICABILITY: MODE 4. AC'iION: j a. With less than the above required Reactor Coolant and/or RHR loops OPERABLE, immediately initiate corrective action to return the required loops to OPERABLE status as soon as possible; if the remaining OPERA 8LE loop is an RHR loop, be in COLD SHUTDOWN within 24 hours.
- b. With no Reactor Coolant.or 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. coolant loop to operation.
"A Reactor Coolant pump shall not be started with one or more<pf the Reactor h
Coolant System cold leg temperatures less than or equal to (W5-) F unlass
- 1) the pressurizer water volume i-s less than /400 cubic feet or 2) the secondary water temperature of each steam generator is less than _gg_ F j above each of the Reactor Coolant System cold leg temperatures.
**All' Reactor Coolant pumps and residual heat removal pumps may be de-energized for up to-1 hour provided 1) no operations are permitted that would cause 7 dilution of the Reactor Coolant System boron concentration, and 2) core l outlet temperature is maintained at least 10*F below saturation temperature.
W-STS 3/4 4-3 JUL 2 71981 i
REACTOR COOLANT SYSTEM . SURVEILLANCE REQUIREMENTS 4.4.1.3.1 The required Reactor Coolant pump (s), if not in operation, shall be determined to be OPERABLE once per 7 days by verifying correct breaker alignments and indicated power availability. 4.4.1.3.2 The required steam generator (s) shall be determined OPERABLE by verifying least oncesecondary side water level to be greater than or equal to (19-)% at g per 12 hours. 18 1 4.4.1.3.3 At least one Reactor Coolant or RHR loop shall be verified to be in operation and circulating reactor coolant at least once per 12 hours. l t l l 1 W-STS 3/4 4-4 N0i 2 01980
REACTOR COOLANT SYSTEM COLD SHUTDOWN 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:
- a. One additional RHR loop shall be OPERABLE #, or
- b. The secondary side water level of at least two steam generators l shall be greater than %. g APPLICABILITY: MODE 5 with Reactor Coolant loops filled" ACTION:
- a. With less than the above required loops OPERABLE or with less than the required steam generator level, immediately initiate corrective action to return the required loops to OPERABLE status or to restore the required level as soon as possible.
.. b. With no RHR loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required RHR , loop to operation. SURVEILLANCE REQUIREMEN'TS I 4.4.1.4.1.1 The required RHR loop shall be demonstrated OPERABLE pursuant to Specification 4.0.5. 4.4.1.4.1.2 The secondary side water level of at least two steam generators when required shall be determined to be within limits at least once per l 12 hours. 4.4.1.4.1.3 At least one RHR loop shall be determined to be in operation and circulating reactor coolant at least once per 12 hours.
# One RHR loop may be inoperable for up to 2 hours for surveillance testing provided the other RHR loop is OPERABLE and in operation.
l l y A Reactor Coolant pump shall not be started with one or more,pf the Reactor 0 Coolant System cold leg temperatures less than or equal to (W4-)*F unless
- 1) the pressurizer water volume is less than ff,pp_ cubic feet or 2) the
~
secondary water temperature of each steam generator is less than 50 *F above each of the Reactor, Coolant System cold leg temperatures.
**The RHR pump may be de energized for up to I hour provided 1) no operations are permitted that would cause dilution of the Reactor Coolant System baron concentration, and 2) core outlet temperature is maintained at least 10*F below saturation temperature. -
W-STS 3/4 4-5 JUL 2 71981
- , . - - - - - . , - - a -- . - -, ., , . - - , - , , - . . - + . . - , - - - - - - , , -r - - - - - - -
REACTOR COOLANT SYSTEM
' COLD SHUTDOWN - LOOPS NOT FILLED LIMITING CONDITION FOR OPERATION
. 3.4.1.4.2 Two residual heat removal (RHR) loops shall be OPERABLE # and at least one RHR loop shall be in operation."
APPLICABILITY: MODE 5 with Reactor Coolant loops not filled. , ACTION: *
- a. With less than the above required loops OPERABLE, immediately initiate corrective action to return the required 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.
s SURVEILLANCE REQUIREMENTS 4.4.1.4.2.1 The required RHR loops 5 hall be demonstrated OPERABLE pursuant to Specification 4.0.5. 4.4.1.4.2.2 At least one RHR loop shall be' determined to be in operation and circulating reactor coolant at least once-perr12 hours. One RHR loop may be inoperable for up to 2 hours for surveillance testing provided the other RHR loop is OPERABLE snd in operation. i
- The RHR pump may be de energized for up to 1 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.
6
/
I W-STS _ 36 66 NOV 2- 1981 r- - , ,-- - , , , , , - - - , - - - = - - - - - - - - - - - - - - - - - - - - - - - - -
I
/
/
/
/
RE TOR COOLANT SYSTEM ISOLA LOOP (OPTIONAL) .
\TIONFOROPERATION LIMITING C0 G
\ /
3.4.1.5 The boro concentration of an isolated loop shall e maintained greater than or eq 1 to the boron concentration of the o rating loops. APPLICA8ILITY: MODES 2, 3, 4, and 5. - ACTION: . With the requirements of t above specification satisfied, do not open the isolated loop's stop valv s; either increase he boron concentration of the isolated loop to within t limits within 4 ours or be in at least HOT STANDBY within the next 6 hours ith the uniso ted portion of the RCS borated to a SHUTDOWN MARGIN equivalent at least 1 delta k/k at 200*F. i
/
. s. SURVEILLANCE REQUIREMENTS
/ \
l 4.4.1.5 The baron concentr ion of an isolated loop all be determined to be greater than or equal to t boron concentration of the operating loops at least once per 24 hours a within 30 minutes prior to o ning either the hot leg or cold leg stop val es of an isolated loop. i l I$
- STS 3/4 '-7 ' f!GV 2 1981 l
t l
/
ACTOR COOLANT SYSTEM
/
IS TED LOOP STARTUP (OPTIONAL) / LIMITING NDITION FOR OPERATION b /
/
- 3. 4.1. 6 A reac r coolant loop shall remain isolated unti .
- a. The isola d loop has been operating on a rec culation flow of greater tha or equal to gpm for at lea 90 minutes and the temperature the cold leg of the isolate loop is within 20*F of the highest co leg temperature of the o erating loops. -
- b. The reactor is sub ritical by at least percent delta k/k.
APPLICABILITY: ALL MODES. ACTION: With the requirements of the above spe ication not satisfied, suspend startup of the isolated loop. . SURVEILLANCE REQUIREMENTS I \ 4.4.1.6.1 The isolated 1 op cold leg temperature shall b determined to be within 20'F of the high t cold leg temperature of the ope ting loops within 30 minutes prior to op ing the cold leg stop valve. 4.4.1.6.2 The reac r shall be determined to be subcritical b at least 1 percent delta k/ within 30 minutes prior to opening the cold g stop valve. W-STS iW4 4-5
!!:'f/ 2 1981 l
i
REACTOR COOLANT SYSTEM 3/4.4.2 SAFETY VALVES - SHUT 00WN LIMITING CONDITION FOR OPERATION 3.4.2.1 A minimum of one pressurizer code safety valve shall be OPERABLE with a lift setting of 2485 PSIG i 1%.*
~
APPLICA8ILITY: MODES 4 and 5. ACTION: With no pressurizer code safety valve OPERABLE, immediately suspend all operations involving positive reactivity changes and place an OPERABLE residual heat removal loop into operation in the shutdown cooling mode. SURVEILLANCE REQUIREMENTS 4.4.2.1 No additional Surveillance Requirements other than those required by Specification 4.0.5. - l "The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure. 3/4 4- II 2 03I PSTS e
REACTOR COOLANT SYSTEM p OPERATING - LIMITING CONDITION FOR OPERATION 3.4.2.2 All pressurizer code safety valves shall be OPERABLE with a lift setting of 2485 PSIG i 1%.* ' APPLICABILITY: MODES 1, 2, and 3. ACTION: With one pressurizer code safety valve inoperable, either restore the inoperable valve to OPERABLE status within 15 minutes or be in at least HOT l STANOBY within 6 hours and in at least HOT SHUTDOWN within the following l 6 hours. SURVEILLANCE REQUIREMENTS 4.4.2.2 No additional Surveillance Requirements other than those required by Specification 4.0.5. I 1
"The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.
NCV 2 G81 I W-STS 3/4 4
, REACTOR COOLANT SYSTEM 3/4.4.3 PRESSURIZER LIMITING CONDITION FOR OPERATION
@ Gnib 3.4.3 The pres equal to (b)surizercubicshall feet,be OPERABLE and at least two with a water groups volume of of pressurizer less than or heaters each having a capacity of at least (150) kw.
APPLICABILITY: MODES 1, 2, and 3. l ACTION:
- a. With one group of pressurizer heaters inoperable, restore at least two groups to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours,
- b. With the pressurizer otherwise inoperable, be in at least HOT STANDBY with the reactor trip breakers open within 6 hours and in HOT ~ SHUT 00WN within the following 6 hours.
SURVEILLANCE REQUIREMENTS 4.4.3.1 The pressurizer water volume shall be determined to be within its limit at least once per 12 hours. 4.4.3.2 The capacity of each of the above required groups of pressurizer heaters shall be verified by measuring circuit current at least once per 92 days.
' ' 2.2 The :::rgeacy p;wer ;upply fer the pre;;uMzer h::ters ch ll be de-^ net-eted OD{Dag(( ;t ];;;; ;379 797 13 7;3th; hj ;;33;))y tp3pg hrc hg_
p;;;r fr:: the ee-;e! te tM : regeacy power supply and caergizing th
@ 2:: tee 5.-
- l l
W-STS 3/4 4- #2
I REACTOR COOLANT SYSTEM 3/4.4.4 RELIEF VALVES, LIMITING CONDITION FOR OPERATION 3.4.4 All power-operated relief valves (PORVs) and their associated block valves shall be OPERABLE. APPLICABILITY: MODES 1, 2, and 3. ACTION:
- a. With one or more PORV(s) inoperable, within I hour either restore the PORV(s) to OPERABLE status or close the associated block valve (s) and remove power from the block valve (s); otherwise, be'in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours,
- b. With one or more block valve (s) inoperable, within 1 hour either restore the block valve (s) to OPERABLE status or close the block valve (s) and remove power from the block valve (s); otherwise, be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
- c. The provisions of Specification 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.4.4.1 In addition to the requirements of Specification 4.0.5, each PORV shall be demonstrated OPERt.3LE 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 OPERABLE 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 Specification 3.4.4.a. -4.4.0.3 ne :- g-ary nnwar copply <ci the p0RY: :nd bicck velves shell be-trn-tr:t^d OPER^3LE :t least Onta per 18 enth; by:
-e: " ne:11y transfer-ing =ativ: :nd centr:1 per:r fro- the nc--a! te-the : : ;=acy p~ee =@p!y,and b Cp;TatiGs th; v0lV00 thro"Oh 3 CC"pl^t" cycl ^ Of #"Il t"a"^l i f
80 W-STS 3/4 4 'N g .2 1981
REACTOR COOLANT SYSTEM 3/4.4.5 STEAM GENERATORS - LIMITING CONDITION FOR OPERATION 3.4.5 Each steam generator shall be OPERABLE. APPLICABILITY: MODES 1, 2, 3 and 4. ACTION:
' With one or more steam generators. inoperable, restore the inoperable generator (s) to OPERABLE status prior to increasing T,yg above 200*F.
SURVEILLANCE REQUIREMENTS 4.4.5.0 Each' steam generator shall be demonstrated OPERABLE by performance of the following augmented inservice inspection program and the requirements of Specification 4.0.5. 4.4.5.1 Steam Generator Sample Selection and Inspection - Each steam generater ! shall be determined OPERABLE during shutdown by selecting and inspecting at least the minimum number of steam generators specified in Table 4.4-1. 4.4.5.2 Steam Generator Tube Sample Selection and Inspection - The steam generator tube minimum sample size, inspection result classification, and the corresponding action required shall be as specified in Table 4.4-2. The. inservice inspection of steam generator tubes shall be performed at the fre-quencies specified in Specification 4.4.5.3 and the inspected tubes shall be verified acceptable per the acceptance criteria of Specification 4.4.5.4. The tubes selected for each inservice inspection shall include at least 3% of the 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 water chemistry indicates critical areas to be inspected, then at least 50% of the 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 generator shall include: .
\1 1981 NOV 2 PSTS 3/4 4- R 4
I REACTOR COOLANT SYSTEM
, f SURVEILLANCE REQUIREMENTS (Continued) verEwed b l
- 1. All nonp h;;:d tubes that previously had detectable wall penetrations (greater than 20%).
- 2. Tubes in those areas where experience has indicated potential problems.
- 3. A tube inspection (pursuant to Specification 4.4.5.4.a.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 may be subjected to a partial tube inspection provided:
- 1. The tubes selected for these samples include the tubes from, those areas of the tube sheet array where tubes with imperfections were previously found.
- 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: Category Inspection Results C-1 Less than 5% of the total tubes inspected are degraded tubes and none of the inspected tubes are defective. C-2 One or more tubes, but not more than 1% of the total tubes inspected are defective, or between
- 5% and 10% of the total tubes inspected are j degraded tubes. ,
C-3 More than 10% of the total' tubes inspected are I degraded tubes or more than 1% of the inspected tubes are defective. Note: In all inspections, previously degraded tubes must exhibit l significant (greater than 10%) further wall penetrations to be included in the above percentage calculations. l i I2 W-STS 3/4 4- K 'NOV 2 1981
I REACTOR COOLANT SYSTEM t SURVEILLANCE REQUIREMENTS (Continued) ! t 4.4.5.3 Inspection Frequencies - 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 Full Power Months but within 24 calendar months of initial criticality.
Subsequent inservice inspections shall be performed at intervals of l not less than 12 nor more than 24 calendar months after the previous l inspection. If two consecutive inspections following service under AVT conditions, not including the preservice inspection, result in
- all inspection results falling into the C-1 category or if two
, consecutive inspections demonstrate that previously observed degra-1 dation has not continued and no additional degradation has occurred, ;
i the inspection interval may be extended to a maximum of once per 40 months. l
- b. If the results of the inservice inspection of a steam 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 least once per 20 months. The increase in inspection frequency shall apply until the subsequent inspections satisfy the criteria of Specification 4.4.5.3.a; the interval may then be extended to a '
maximum of once per 40 months.
- c. Additional, unscheduled inservice inspections shall be performed on each steam ger. orator in accordance with the first sample inspection specified in Table 4.4-2 during the shutdown subsequent to any of
! the following conditions:
- 1. Primary-to-secondary tubes leaks (not including leaks originating from tube-to-tube sheet welds) in excess of tne limits of l l Specification 3.4.6.2.
l
- 2. A seismic occurrence greater than the Operating Basis Earthquake.
- 3. A loss-of-coolant accident requiring actuation of the engineered safeguards.
l
- 4. A main steam line or feedwater line break.
i
*I NOV 2 1981 W
_-STS 3/4 4-
s REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 4.4.5.4 Acceptance Criteria
- a. As used in this Specification:
- 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. l 2. Degradation means a service-induced cracking, wastage, wear or j general corrosion occurring on either inside or outside of a tube.
- 3. Degraded Tube means a tube containing imperfections greater than or equal to 20% of the nominal wall thickness caused by degradation.
- 4. ~% Degradation means the percentage of the tube wall thickness affected or removed by degradation.
- 5. Defect means an imperfection of such severity that it exceeds the s h;;d A. limit. A tube containing a defect is defective.
OE 6. gw Edf Limit means the imperfection depth at or beyond which l the tube shall be removed from service and is eaual to (40)%* of the nominal tube wall thickness.
- 7. Unserviceable describes the condition of a tube if it leaks or contains a defect large enough to affect its structural integ-rity in the event of an Operating Basis Earthquake, a loss-of-
! coolant accident, or a steam line or feedvater line creak as specified in 4.4.5.3.c, above. l
- 8. Tube Inspection means an inspection of the steam generator' tube from the point of entry (hot leg side) completely around the U-bend to the top support of the cold leg. -
"Value to be determined in accordance with the recommendations of Regulatory Guide 1.121, August 1976;
( 14 fiOV 2 1981 y-STS 3/4 4 'Ni \
i l REACTOR COOLANT SYSTEM ( I SURVEILLANCE REQUIREMENTS (Continued)
- 9. Preservice Inspection means an inspection of the full length of 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 after the ;
, field hydrostatic test and prior to initial POWER OPERATION t using the equipment and techniques expected to be used during subsequent inservice inspections. 5 ee. TaseA I l b. The steam generator shall be determined OPERABLE after completing , the corresponding actions (plug all tubes exceeding the ?! u;;Sgvepirable. 1 limit and all tubes containing through-wall cracks) required by Table 4.4-2. Os-4.4.5.5 Reports
- a. Within 15 days following the completion of each inservice inspection of steam generator tubes, the number of tubes plugged in each steam generator shall be reported to the Commission in a Special Report pursuant to Specification 6.9.2.
- b. The complete results of the steam generator tube inservice inspection shall be submitted to the Commission in a Special Report pursuant to
{ Specification 6.9.2 within 12 months following the completion of the inspectiori. This Special Report shall include:
- 1. Number and extent of tubes inspected.
- 2. Location and percent of wall-thickness penetration for each indication of an imperfection.
- 3. Identification of tubes p!:;;d. ee.paleed @
- c. Results of steam generator tube inspections which fall into Category C-3 and require prompt notification of the Commission shall be reported pursuant to Specificati,on 6.9.1 prior to resumption of plant operation. The written followup of this report shall provide a description of investigations conducted to determine cause of the tube degradation and corrective measures taken to prevent recurrence.
I 16 W-STS 3/4 4-M 'NOV 2 1981
7 _ . ,' y e , > : - Q.ts. .mn,. - - - ,. .,.<
~~- --
. * , , . ya * :;;W. 4.'4r**5 "p y y c.c 3,7 .. = - ,
. .. g -~,7. . . -
.v: . a ,r.,,.
n
*:* $ ,. * .# 3 s=..... ,, , . .- .ww. . . - . : .
_. a - , a, c . - . ~
.- r--------
,--;t . ,.,
n ,
- s e.
- s. - ..s .., . ,
, ,. , , p. ,,.; ,
u ,
<.. a,-.
." , ., .,..s. .a~s s ',.-.., . + e
,. .p+,.
*9. .. ..w .> . - - ' .+*> . . *,.6--. *. 1< ..: % .. 'Y t* * .. .
. ,s*e .gn . 4.,o-
. n.m4. . ai . 1. . , .-,.v,~., -
-*-* m -
.p,v. *. ...3.==,..,3..,',:.*.
, ~ .n s, W. .- # - * , .4. .
.n; *. . .,,,r.. - - m~ c., . , . -. ;. , -%. ,3.c ., . ,. -s e ,,
;4 . s .
. 9 <
.e ,.w,.. ,s p,. - ,- ,, , ~,
.%.> , , ., . , .' t . , , g . #, , ,
.e.- ,s*,.
. . ~. . i . a s .4 - . *< s .,- + j-
.,,,,$.~..*,.9,..
- 1 '.g.,_.,.. , . ..#,,. *-e . . .e . . *% . : ,. - .<. .
v . , e . . a
. .- #4. ., , ..
a
..s. ..3 ,
.,,,c..
e
.,.~.~&,. ..--..._m.
.. . .4
. ~ ..n.. - ., _ _ . . . . . . _
s- , 6 . , .
.p
"; -e 4- .. ; y . , r; % s s ;,
{ , W,,3o.y vy ..%s y , , ggy 4,.; . , .,4. 5,4 > m .-- r ., y .g f , ' . n. e ,, . .
.++
..; = ;.,.-. ,
- v. w ,,.; ~. ....nA .~ *).,
s*. .
. etr: . . v;
., !--.' .:' J=s.*T'***'****'**'**'**#"*"9"******[...
^
."#*"t*******.-^
. ' ' * * * ' ' ' - " * '- - * * * * * * ~ * - - " ~ " * * " - ' ' - ~ '* ' " - * * * * "' - " * ** " *"*-' .
.t - - .N . .' .. . . ,. -
.. . ' J
, ,Y. .. e . .,-,' . ., ya.. z._
, - g w ~ ., , p , s . a ur . .. y3.,t. . ,_,,.
~ xg ,.. Q ., o. ] . o - e *. " ? .- : ,s4 . q , . ,y l
. , .- a f*,. m =
f e. . '
'%. 3-e , .
~.- .f' n-3. ' 7 ... .
.+a . - ., _.e.. ".
, . . ~ , e..,~..,,.,.-...-,.,.,--$-.. . ..
,.' ^ s s,,
T-n
. 4
. P'- y g.,. .g . ,-
; a., y y-. , = .
3 -
.f.
. - - - . - . - ~ - - . - , . - . - . . . . . . - -
.;h . .% .3 *
; * ,,. 7,.
9 Oc .a s > ~ . . . Jl ; . d, _.w'~., , .- ,.g g _. 1 -^n. ,o g y.Q_ : '. ;mleg'c .._ j.( Q. .g gl l ' '_.' g. ..
. Gs- p}.'.~ggho..* pgg ..p, g _,,.f g f g_fp.. _ . .
i
. . , .. _ , ~* * , ,. . .y
. e-S-' , < .
r
...,.s.......s~ . - . . ..g 8 g.
,# n g
*I e g'h
.i.,. . -y *a .o P , -'
gi,
' s. * ~~. '
, :, , . ,e.
- , , . ev . e.s , . ~4,,.*,,A. , , . e.b 4,.. -
. . w. , . t 4 4 ..
n, -.33nyggg '.. gaygg gam. du... ho.-u,g- aggg f a ay. .. . . - -. . . . . ..
, - - ..ya.. .
g e 4.,.e. s
' = .. . " 1 * .) %..+
..*.y. '; ygg$.g ,e - s =
.,, * 'sb,'... -a ,,, , at
..ss 7
* - , *. . ', J%**.#,.. w ,.. m .- . $,. . .v 5,
,,.* o g.,. , , ' ."f', .
e.- =. ....
., . 7
. <f -..
*f 5
I
.~
Coj ' , u ;., l . 'y. e
'^
."i M i . Cn . ..
':.l.'.. a-~. 1: l[ .'.. . l. : ~ . .- - . . . . . . . . . . -- . -. . -
, , .-* . ,. . , ,*- . t .
ee. a s
*.6...
.\. .,=. .'.4.. -f .A- . .
a.%.+;.7'
, .s^
.*,C,'e.'7..s.*
.e
**-8
*e'S$
' es = .. ... ,,*'. , l-
,t o
a w. 4. ya8e.g.e'# * . s
- g;.' ' ,.f..
- .i.. . * * - - - - I-s'k . d
*=..- -
, .* ' p ** I -
, , ' 4.- , %Gi .. ** O y
. it'. $ .
4 s ".o' p} ^d8 i g
,'" n '
w , g d.
/'*,,'s "'.,'y
^ ' $s
,p
.g t,. I e '%
'.N g .
./*, . .a--* , e4 4 . ._ &.=. 4 e p ampe. WM,. .e.4.e g een e m a ---
, ='. *g d. ..g *4 e+ s.wi.e. , ,+ , ,. = =.m ute,=b er 6.%. .
,ies-- . * , - . . . . . . . - - . . . - ....
- 4 , ..M ,
=.#'**'%. g,-%3 4 ,, t * , '=' L/
*%,"u#
9 '=
# .A. *g
, f ., j$ " , e A g
- g
** i 4 g *' %. / e, f -
*IB
. 4
\*,
s b
. 4 4 O , " . . .
- g* ,
.-al e
,e . ea -de ^
- WeMee! w # ,.,v e,ue o.4. .. . e.e . e . e e .-4.-.. . . - - - .- - ...
8 .g-'_
, h 9 [ bg 9
. . g7 .
4
= A 0 *f &
- A i
* ' .h b.e.Q sw a g, m. =e-6 es.6M.. * - = - - * .-M---
-M ee+ wi. --he.
M MS- *.*g.e-eN-W'
*, p-
. SM h
,.'b(q.*% .* . . . *' MM OM M M.MO
. % 8 8 h'b a dm Ng I .I ,'* * %
g
).
.}Q O.'_ ",g * ",
-{.* g. s v
, , eg .- f. * *- . *g".'
. _...m.'s , e 6~~s-,*
. . . . . . . . . . . - - - ._. ~.
(. ,
- .. .
- p
# g "*a
. e .s,.
-==.ae..'-mww=.--
, . e .
- - , . e .. 4 sp.. .e.o. ab e.
u e-.
.
- m.m e.-..y.e e - e : _ .me w e .o e .' e.e 6 . . . . . . . .- .- . -
*
- I g-4 , . % -4 .
4,,. H 4-e gk. .& s 44, y.., h 4 .' , ..
. mr ,,
y ., ,* % f ,,
. ,m. e - go.is. .= e e e eg..me.g. es 'p.e4. .- .sp -m .-est .wa u i - e . .e. eo .. # . w....4 ..-. ..- .= . . - -. -
. ,'d_* y g _ #
* .g e.,f* % - g % %
' " t .*
y .4 *
- y
, 4.- . . . - , ._ , . -
4 *-~ *
,_ 4 e
~.--a - -- ..-. . -E....",.- ,
. - r ~ . 4- -
....6... -
...r._. .. ....
,y "'= h*.
^
,a, -e s , , ,.a; ** ~
n '.
- N*. .<- , _1, .- ,
~
-~~s.t-g_""=*I., !. . ~ . m. ~
a.1.F
-s
,.g 4
s , \f I
- e k, e.
-.g
-eew -
' .' %" af " - , h
- s 5 '* *
*N,
_ ew s_, m ,
.+--A_
s
"'a.
. .. - + . .4 - - - . .
. . . , . -9..L- a--*
a.n,c
,, *.,g; .s *^d.. s.. ,,;.*-y., . . * .
.s
,*.p..
,#,,.aw ,s .s,- .'. ,.' 1+ '.m. '
* '/ A s . .
s.- . ,
* * * = , e 9e .,.8 ., - ,
emge.m, p.=m . -ee.4 6. -
- osm = 4 me ,
_.e T. +
-g. ._. _ op..='* es %
-en e m m- . e.... .-- . . . - - - . - . . . . < . .
a , gm- O ,,* *
. p
,- ' = .<O'_JTp.t 4
.m. ,..#,, d %-- - . . . . ,
,- , . g.
#..4 a
* *. * - .") %.
, ' *- d*'
D'
,F* 6 . , . . . , . -
. . - . - . . ~ ,Mf- 4., .-.M
-..gs..-
.y
.-g,..*-e..-.--,.e-L-..'c
- u-...- .. . . . .. . . .. . . . .
s -.=a.** -: ..
, , i
'# # ,,,"" - 6
. ,- , .' . },.,a. .v . *. g
,,'g .. -. .. $ *
, g. . ' . = $' - . . . ' g.t
% g.
.-----.w
. -. s
~
- m. . , *; .
., ,n .-. - ~ .. -. ,
- .. . .~. . . ......
.,.v... e. & <- (
<,x...n..,,.<.
. m.. .
,_-; ; .. **,4-~ * : ?. y.':,'s t;.
.4 _ .
- , . - - - - - . = . - _ . . . . _ . _ . _ ..
3.-
-_
- 4 r' -
e
*
- 4
% .- , %j9 3 *
@ ,p e p ? * ? , _
; a, .
. a ,
. . a
& * = ( *, w 9
# *% . w
.. h- . . . _--------,_4
% *
- e e
,e . **
. - - .a, 8 4 *
-- - a +- -
.e: -- = ~ ~ . .
..- .- . - - - ~ ~..- - .-- - .. 4
- , s e v ~ > . . =*, ..
- g q . ., . , ,,y.j.e-
'a<,ge',4,'
~
"3#%.,x
.< s, e.
*; 4. < , . , , ,
* . ""*_ ,y, yg ',g* 4 m , .,g , , . , g
-e-..o- - - . . e.- ..-...-e. -e e - -4e e-d ww se W e -
-. ..-.-g --
*f e'.e a9 ,* ' . ,, '- # eb egP._. #
g , F *& F 4
, ,.r- .. .,%-,.
e.<.5.,'
- b
. .,"4 g. -
* , . . s
.. ~ , , . . . .- . _ . ,
j ..,, 4. . it ..wl 4 g as ': r * ! , a,, . , ', ^ A g ,,* y # g,. ,
*h
."r:",. . ,
. . , .f:-
., u yw s e.o
.. ~ . . - . . - . . - + - - . . -- ---. .~....- y
_.-y.s,,. a- *',.e..-.% _ _
*;, W s. s , a- c 4 -.. .t ;.
,.~.,',.;,yp~,,*..'-;
y
. ,,
- a
, .g
;e,.. f-
=g.'- . ,m'*.%. g,
- g. + % ,s
-*,. p f
, %m i.. J e },
.**N..# g q .; .',e ....? -
., +
E.,*,
,,j.,s v.,_. * #a,
.M
.M M.O
.,.M -
e*.m.w.-=W. * .m.-=*..e
*+.e., ,,,
.s,,-
- . ' * ,* J y .g. vf*V] ...,..'.*15/,3,. t.
4,a
.i, o .
e * , 4 s .- - ,
;.,e,, , , t .
,9'"N, . > ' ,s.. * . ' , * * ' .
*s fn a, e e., d e
,m.m ee.- . . . ev. _... .. -- . .,..
.go. peo
=e7 gp gewh --e. e,w..,y..e - . ,4.eew - - -. ee . .a
\-
..+ s.
g , , ,' * ,
- rA ,< ,+s , , u ,
y ,*g
% , w _N 4q 1.,
.. . ,*.S . s
,e..,,.s. , .
h.%*8"' '4,/...f. p..,,,.i. s. s. . . . a._ ,,...
'""*'.=**'***..**e.***,Mm.:_eq.g.
*******'. 4,,,- . , ,
f., .4.ma=*. .
~'f . '"
y s'A'
',. )
s g
- s. ,,
a ...6 3 s u .. *
* ,,
- g,
*.A . 94. .
*/
. , ..m g
.A f.. *./.*',... .* f
..,* =
g , , . 1 . 4
~,#4 p6.*..W."
M@,4 bMM 4 $ 6M h k '
. g peg M 4 W. . MM . ^ -..i4gE".h -,.6 F'
Wegew 4sg.-#h.. -w e e.w p.e.9 7-9.g.94.i-,. a -', .
' ~
'.- . '.( ,',,
,.s.n's
_4
%,* .. ' ' s. . . c s , . ~ .
r , e.'
- a, .s- . < ,
'. '~ $. % u, ...t. _t . , a f .-- ~s .. - : i . ,
r. t c .y . v -*
- 7c.t . ,,.-t-. ,
, Da.
. * *. . . - . -.. .- . . . . ,.- .a-ar % + ,,a.--+.-.6+--..-..-*e.-++......L.
w *-e.
..,y,-A,.
. Q v .* ' *' g4"o
.%,***9.*8'"l y.a, 4
.+ J. _ .a- ## _
*-%, __ , =g ,g,. U^ .%...
M;..,
=* d.6. i
,g.'..-. ,g*.,.
.-**4 e ( , ', ,
. '- *A
, . ,', 5
^
.**8 g ' - * .,4 > #, *) .?. . . ,+ 3e '* vp .
%2'kxm' b., DM,,,4a:,n.iLs, Q.4. M %'.h.va.M *.2<, 1 M M M M, ki^-D.:.G N + - u . h .>..a a c.. A .G O. p
[ TABLE 4.4-1 MINIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTED DURING INSERVICE INSPECTION Preservice Inspection
- Yes No. of Steam Generators per Unit Four First Inservice Inspection Two Second & Subsequent Inservice Inspections Onel,2 Table Notation:
- 1. The inservice inspection may be limited to one steam generator on a rotating schedule encompassing 12% of the tubes if the results of the first or previous inspections indicate that all steam generators are performing in a like manner. Note that under some circumstances, the operating conditions in one or more steam generators may be found to be more severe than those in other steam generators. Under such circumstances the sample sequence shall be modified to inspect the most
(.' severe conditions.
- 2. Each of the other two steam generators not inspected during the first j inservice inspections shall be inspected during the second and third l inspections. The fourth and subsequent inspections shall follow the instructions described in 1 above.
l t 9 4 l( , t 3/4 4 . e
1 1 TABLE 4.4-2 h STEAM CENERATOR TUBE INSPECTION t IST SAMPLE INSPECTION 2ND SAMPLE INSPECTION 3RD SAMPLE INSPECTION Sample Action Action Action i Size Result Required Result Required Result Required A minimus C-1 None N/A N/A N/A N/A l of S Tubes per S. G.- gepg,. C-2 4Abog defective C-1 None N/A N/A tubes and in-spect addi- ge$*;,, ak we defective C-1 ,Noge tional 2S tubes tubes and in- C-2 de fec-in this S.C. C-2 spect additional tive tubes 45 tubes in this S.C. Perform ac-C-3 tion for C-3 4 result of first sample i . Perform action ( C-3 for C-3 result N/A N/A
of first sample C-3 Inspect all All other tubes in this S. G.s are None N/A N/A'
,' S.G.Y$$khde- C-1 fective tubes and inspect 2S Some S.C.s Perform action N/A N/A tubes in each C-2 but no for C-2 result , other S.G. additional of second l S.G. are sample Prompt noti- C-3 fication to
- NRC pursuant to specifi-
[ cation 6.9.1 Additional Inspect all l S.G. is tubes in each C-3 S.G. and %" defective tubes N/A N/A Prompt notifi-cation to NRC . l pursuant to . specification i 6.9.1 S = AE% Where n is the number of steam generators inspected during an i (- inspection 3/4 4-17
REACTOR COOLANT SYSTEM 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION SYSTEMS LIMITING CONDITION FOR OPERATION 3.4.6.1 The following Reactor Coolant System leakage detection systems shall be OPERABLE:
- a. Thecontainmentatmospherefgstee: _
(particulate) radioactivity monitoring system, .
- b. Avoange The containment pc9 e_ sump level and #!ce monitoring system, and bc. St. &.... b"?$W,$.%%. .*._G*'.h. .*mSkce_,_,,,,,
. -- _ . . . . . .. so.....____ ,w , , s, .o
@ 4 == a+
=+=a P are-(;n:::: erparticulat:)rdicactivity-g
.;.it;-'r.g :y:t =.
APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: With only @ of the above required leakage detection systems OPERABLE, operation may continue for up to 30 days provided grab samples of the contain-ment atmosphere are obtained and analyzed at least once per 24 hours when the required gaseous or. particulate radioactive monitoring system is inoperable; otherwise, be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE REQUIREMENTS 4.4.6.1 The leakage detection systems shall be demonstrated OPERABLE by:
- a. Containme,nt atmosphere (gaseous and/or particulate) monitoring system performance of CHANNEL CHECK, CHANNEL CALIBRATION and ANALOG CHANNEL OPERATIONAL TEST at the frequencies specified in Table 4.3-3,
- b. Contaiament p Wsump level rd 'i= monitoring system performance of CHANNEL CALIBRATION at least once per 18 months, "O
i.!E[5[$_35.. 3.E...!.2N,._!.2E!N23
. ...... ., E2N '5 DEI O t , W-STS 3/44-k NOV 2 1981
, REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION 3.4.6.2 Reactor Coolant System leakage shall be limited to:
- a. No PRESSURE BOUNDARY LEAKAGE,
- b. 1 GPM UNIDENTIFIED LEAKAGE,
- c. 1 GPM total primary-to-secondary leakage through all steam generators :t h:!sted #-^= +he 9:::ter 0;eler.t Sp ter and g00)galpnsperdaybroughanyonesteamgenerator,act4elatedg d.@10 GPM IDENTIFIED LEAKAGE from the Reactor Coolant System, and
- e. 52 GPM CONTROLLED LEAKAGE at a Reactor Coolant System pressure of Tl35 1 20 psig.
- f. 1 GPM leakage at a Reactor Coolant System pressure of 2235 20 psig from any Reactor Coolant System Pressure Isolation Valve specified in Table 3.4-1.
APPLICABILITY: MODES 1, 2, 3, and 4. ACTION:
- a. With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT STANDBY within 6 hours and in COLD SHUTDOWN within the following 30 hours.
.b. With any Reactor Coolant System leakage greater than any one of the above limits, excluding PRESSURE BOUNDARY LEAKAGE and leakage from Reactor Coolant System Pressure Isolation Valves, reduce the leakage rate to within limits within 4 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours'.
- c. With any Reactor Coolant System Pressure Isolation Valve leakage greater than the above limit, isolate the high pressure portion of -
the affected system from the low pressure portion within 4 hours by use of at least two closed manual or deactivated automatic valves, or be in at least HOT STANDBY within the next 6 hours and in COLD ) . SHUTDOWN within the following 30 hours. l t l 19 jf-STS 3/4 4-M GS1 NOV 2
l l REACTOR COOLANT SYSTEM l SURVEILLANCE REQUIREMENTS I 4.4.6.2.1 Reactor Coolant System leakages shall be demonstrated to be within
; each of the above limits by:
- a. Monitoring the containment atmosphere (gaseous or particulate) radioactivity monitor at least,once drama p@er 12 hours.
- b. Monitoring the containment pa % gesump inventory and discharge at l least once per 12 hours.
- c. Measurement of the CONTROLLED LEAKAGE to the reactor coolant pump seals when the Reactor Coolant System pressure is 2235 1 20 psig at least once per 31 days.:ith t h n il:ti::g ; : Ell., gr- The l
provisions of Specification 4.0.4 are not applicable for entry into l MODE 3 or 4.
- d. Performance of a Reactor Coolant System water inventory balance at de d af y least once per 72 hours.Jaci.35pAhho o vyeM4o q.o.y eve net 11 odes 3 ** 'tapptr.aue (%, en+a+c
- e. Monitoring the reactor head flange leakoff system at least once per 24 hours.
4.4.6.2.2 Each Reactor Coolant System Pressure Isolation Valve specified in l Table 3.4-1 shall be demonstrated OPERABLE pursuant to Specification 4.0.5, except that in lieu of any leakage testing required by Specification 4.0.5, each valve shall be demonstrated OPERABLE by verifying leakage to be within its limit:
- a. At least once.per 18 months.
- b. Prior to entering MODE 2 whenever the plant has been in COLD SHUTDOWN for 72 hours or more and if leakage testing has not been l performed in the previous 9 months.
- c. Prior to returning the valve to service following maintenance, repair or replacement work on th6 valve.
- 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 MODE 3 or 4. . t I NOV 2 1981 W-STS 3/4 4-2g,
i' TABLE 3.4-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES VALVE NUMBER FUNCTION Thu inb. 4 ion will be supph'ed 4 a lato dae. e a D e % e PSTS 3/4 4- NOV 2 1991 1: -
)
l REACTOR COOLANT SYSTEM 3/4.4.7 CHEMISTRY LIMITING CONDITION FOR OPERATION 3.4.7 The Reactor Coolant System chemistry shall be maintained within the limits specified in Table 3.4-2. APPLICABILITY: 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 Limit, restore the parameter to within its Steady State Limit within 24 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
- b. With any one or more chemistry parameter in excess of its Transient Limit, be in at least HOT STANDBY within 6 hours and in COLD SHUTDOWN within the following 30 hours.
At All Other Times: With the concentration of either chloride or fluoride in the Reactor Coolant System 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; determine that the Reactor Coolant System remains acceptable for continued operation prior to increasing the pressurizer pressure above 500 psig or prior to proceeding to MODE 4. SURVEILLANCE REQUIREMENTS 4.4.7 The Reactor Coolant System chemistry shall be determined to be within the limits by analysis of those parameters at the frequencies specified in Table 4.4-3. l n, NOV 2 1081 W _-STS 3/4 4-N k I
TABLE 3.4-2 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS STEADY STATE TRANSIENT PARAMETER LIMIT LIMIT DISSOLVED OXYGEN * $ 0.10 ppa 1 1.00 ppm CHLORIDE 5 0.15 ppm 1 1.50 ppm FLUORIDE 1 0.15 ppm 5 1.50 ppm
" Limit not applicable with T,yg less than or equal to 250*F.
l 23 W-STS 3/4 4-26 NGV 2 1981
l IABLE 4.4-3 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS SURVEILLANCE REQUIREMENTS SAMPLE AND PARAMETER ANALYSIS FREQUENCY DISSOLVED OXYGEN
- At least once per 72 hours CHLORIDE At least once per 72 hours FLUORIDE At least once per 72 hours i
"Not required with T,yg less than or equal to 250*F l
l l ( l 2A l y-STS 3/4 4-26 .ncy 2 1981
~ ~ w n w e-,., . , - , , , ~ - , w--- - - o e. e ,..-m- w.+ - - >
REACTOR COOLANT SYSTEM t 3/4.4.8 SPECIFIC ACTIVITY LIMITING CONDITION FOR OPERATION 3.4.8 The specific activity of the primary coolant shall be limited to:
- a. Less than 'or equal to 1.0 microcurie per gram DOSE EQUIVALENT I-131, and
- b. Less than or equal .to 100/E microcuries per gram.
APPLICABILITY: MODES 1, 2, 3, 4, and 5. ACTION: MODES 1, 2 and 3*:
- a. With the specific activity of the primary coolant greater than 1.0 microcurie per gram DOSE EQUIVALENT I-131 but within the allowable limit (below and to the left of the line) shown on Figure 3.4-1, operation may continue for up to 48 hours provided that the cumula-tive operating time under these circumstances does not exceed 800 hours in any consecutive 12-month period. With the total cumulative operating time at a primary coolant specific activity greater than 1.0 microcurie per gram DOSE EQUIVALENT I-131 exceeding 500 hours in any consecutive 6-month period, prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within 30 days indicating the number of hours above this limit. The provisions of Specification 3.0.4 are not applicable.
- b. With the specific activity of the primary coolant greater than 1.0 microcurie per gram DOSE EQUIVALENT I-131 for more than 48 hours during one continuous time interval or exceeding the limit line shown on Figure 3.4-1, be in at least HOT STANDBY with T less than 500 F avg within 6 hours. . .
j c. With the specific activity of the primary coolant greater than 100/E microcuries per gram, be in at least HOT STANDBY with T less than avg 500*F within 6 hours.
" With T,yg greater than or equal to 500*F.
l 25' W-STS 3/4 4-M li?/ .a 1981 l l
I REACTOR COOLANT SYSTEM ACTION: (Continued) MODES 1, 2, 3, 4, and 5:
- a. With the specific activity of the primary coolant greater than 1.0 microcurie per gram DOSE EQUIVALENT I-131 or greater than 100 6 microcuries per gram, perform the sampling and analysis requirements of item 4a of Table 4.4-4 until the specific activity of the primary coolant is restored to within its limits. A REPORTABLE OCCURRENCF shall be prepared and submitted to the Commission pursuant to Speciti-cation 6.9.1. This report shall contain the results of the specific activity analyses together with the following information:
- 1. Reactor power history starting 48 hours prior to the first sample in which the limit was exceeded, l 2. Fuel burnup by core region,
- 3. Clean-up flow history starting 48 hours prior to the first sample in which the limit was exceeded, l
l
- 4. History of de gassing operations, if any, starting 48 hours prior to the first sample in which the limit was exceeded, and
! 5. The time duration when the specific activity of the primary coolant exceeded 1.0 microcurie per gram DOSE EQUIVALENT I-131. SURVEILLANCE REQUIREMENTS l 4.4.8 The specific activity of the primary coolant shall be determined to be l within the limits by performance of the sampling and analysis program of Table 4.4-4. - i t l M .NOV 2 1981 W-STS 3/4 4- M l
l l 300 3 N b 250 E a C . s P k 200 o. E UNACCEPTABLE 8 OPERATION E U b E E
\
E g 100 7
$ ACCEPTABLE
,4 OPERATION 5 50 8
8 0 20 30 40 50 60 70 80 90 100 PERCENT OF RATED THERMAL POWER FIGURE 3.4-1 DOSE EQUIVALENT l-131 PRIMARY COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER , WITH THE PRIMARY COOLANT SPECIFIC \ ACTIVITY >1.0 Ci/ gram DOSE EQUIVALENT l-131 3/4 4- M
TABLE 4.4-4 1
'T i
\. $
u' PRIMARY COOLANT SPECIFIC ACTIVITY SAMPLE
, AND ANALYSIS PROGRAM i
TYPE OF MEASUREMENT SAMPLE AND ANALYSIS MODES IN WHICH SAMPLE AND ANALYSIS FREQUENCY AND ANALYSIS REQUIRED
~
- 1. Gross Activity Determination At least once per 72 hours 1, 2, 3, 4
- 2. Isotopic Analysis for DOSE EQUIVA- 1 per 14 days 1 LENT I-131 Concentration '
' ~
- 3. Radiochemical for E Determination 1 per 6 months
- 1
- 4. Isotopic Analysis for Iodine a)' Once per 4 hours, # # # # #
l,2,3,4,5 Including I-131, I-133, and I-135 whenever the specific w activity exceeds 1.0
) pCi/ gram DOSE
. EQUIVALENT I-131 l g' p or 100/E pCi/ gram, and
- P oa b) One sample between 2 1,2,3 and 6 hours following a THERMAL POWER change exceeding 15 percent of the RATED THERMAL POWdR within a 1-hour
- period.
fUntilthespecificactivityoftheprimarycoolantsystemisrestoredwithinitslimits. i
- Sample to be taken after a minimum of 2 EfPD and 20 days of POWER OPERATION have elapsed since reactor was last subcritical for 48 hours or longer.
E to
! to 2
, REACTOR COOLANT SYSTEM 3/4.4.9 FRESSURE/ TEMPERATURE LIMITS REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION 3.4.9.1 The Reactor Coolant System (except the pressurizer) temperature and pressure shall be limited in accordance with the limit lines shown on Figures 3.4-2 and 3.4-3 during heatup, cooldown, criticality, and inservice leak and hydrostatic testing with:
- a. A maximum heatup of (100)*F in any 1-hour period,
- b. A maximum cooldown of (100)*F in any 1-hour period.
- c. A maximum temperature change of less than or equal to (10)*F in any 1-hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves.
APPLICABILITY: At all times. ACTION: With any of the above limits exceeded, restore the temperature and/or pressure to within the limit within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the Reactor Coolant System; determine that the Reactor Coolant System remains acceptable for continued operation or be in at least~ HOT STANDBY within the next 6 hours and reduce the RCS T and pressure to less than 200*F and 500psig,respectively,withinth8V90llowing30 hours. SURVEILLANCE REQUIREMENTS l i 4'.4.9.1.1 The Reactor Coolant System temperature and pressure shall be I determined to be within the limits at least once per 30 minutes during system l heatup,.cooldown, and inservice leak and hydrostatic testing operations. 4.4.9.1.2 The reactor vessel material irradiation surveillance specimens shall be removed and examined, to determine changes in material properties, as required by 10 CFR 50, Appendix H in accordance with the schedule in-
. Table 4.4-5. The results of these examinations shall be used to update '
l Figures 3.4-2 and 3.4-3. !( l l 29 l W-STS 3/4 4-M NOV 2 1981
=
COPPER CONTENT : CONSERVATIVELY ASSUMED TO BE . 0.10 WT% (ACTU AL CONTENT = 0.06 WT%) RT INITIAL : 40*F NDT RT AFTER 16 EFPY : 1/4T, 110 *F NDT 3/4T, 87'F 1 l CURVE APPLICABLE FOR HEATUP RATES UP TO 60*F/HR FOR THE SERVICE PERIOD UP TO 16 EFPY AND CONTAINS MARGINS OF 10*F AND 60 PSIG FOR POSSIBLE INSTRUMENT ERRORS, 3000 LEAK TEST LIMIT ~
.a %
E e b w 2000 2 D M M W C / 0.
/
l 2 / - CRITICALITY
% / LIMIT BASED e y ON INSERVICE > 'iOOO / HYDROSTATIC
, g 7 F HEATUP EMPERATURE
$: CURVE- (255'F) FOR
~
THE SERVICE O PERIOD UP O TO 16 EFPY
=
O F 0 0 0 100 200 300 400 500
@ INDICATED TEMPERATURE (DEG.F) l FIGURE 3.4-2 SEABROOK UNIT I REACTOR COOLANT SYSTEM HEATUP LIMITATIONS' APPLICABLE TO 16 EFPY 3/4 4-30
MATERIAL PROPERTY BASIS COPPER CONTENT, : CONSERVATIVELY ASSUMED TO BE - 0.10 WT% (ACTUAL CONTENT = 0.06 WT%) RT INITIAL : 40*F NOT . RT NOT AFTER 16 EFPY : 1/4T,110*F , 3/4T, 87'F CURVE APPLICAB'LE FOR COOLDOWN RATES OP TO 100*F/HR FOR THE SERVICE PERIOD UP TO 16 EFPY AND CONTAINS MARGINS OF 10*F AND 60 PSIG FOR POSSIBLE INSTRUMENT ERRORS -
,/
3000-to b i g . h 2000
!3 E
Q LU F-
<C ^
9 C E 1000 COOLDOWN RATES (*F/HR) 0- j
,'40 20_ y/
100- / 0 0 100 200 300 400 500' INDICATED TEMPERATURE (DEG.F) FIGURE 3.4-3 SEABROOK UNIT I REACTOR COOLANT SYSTEM COOLDOWN LIMITATIONS APPLICABLE TO 16 EFPY 3/4 4-31
TABLE 4.4-5
- !T . . _ _ . .
SURVEILLANCE CAPSULE REMOVAL SCHEDULE 1 \ Orientation i Capsule of Lead Removal Expected Capsule
'. Identification Capsulosial Factorlbl Time Fluence (nicm2)
U 58.5* 4.00 1st Refueling 3.3 x 1018 Y 241* 3.89 5 EFPY 1.2 x 101Mc) V 61* 3.69 9 EFPY 2.2 x 101Mdl X 238.5* 4.00 15 EFPY 3.9 x 1019
, W 121.5* 4.00 Stand-By w Z 301.5* 4.00 Stand-By 1
a
- a. Reference irradiation Capsula Assembly Drawing, Fgure 2-4.
- b. The factor by which the capsule fluence leads the vessels maximum inner war fluence.
- c. Approximate Fluence at % wall thicknass at End-of-Life.
- d. Approximate Fluence at vessel inner wall at End-of-Ufe.
to
s: . wma REACIOR COOLANT SYSTEM
. PRESSURIZER l I
LIMITING CONDIT.ON FOR OPERATION ] 3.4.9.2 The pressurizer temperature shall be limited to:
- a. A maximum heatup of (100)*F in any 1-hour period,
- b. A maximum cooldown of (200)*F in any 1-hour period, and
- c. A maximum spray water temperat'ure differential of (320)*F.
APPLICABILITY: At all times. ACTION: With the pressurizer temperature limits in excess of any of the above limits, restore the temperature to within the limits 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 HOT STANDBY within the next 6 hours and reduce the pressurizer pressure to less than 500 psig within the following 30 hours. ( SURVEILLANCE REQUIREMENTS 4.4.9.2 The pressurizer temperatures shall be determined to be within the . limits at least once per 30 minutes during system heatup or cooldown. The l spray water temperature differential shall be determined to be within the i limit at least once per 12 hours during auxiliary spray operation. l .
. N 33 PSTS 3/4,4 '34 .NOV 2 1981 l .
i ~ l l
' REACTOR COOLANT SYSTEM i 0VERPRESSURE 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 relief valves (PORVs) with a lift setting of .
less than or equal to (15^) pW, - A madmum se.fpimt defmed by Fi y e. 3 9 -9 or g
- b. The Reactor Coolant System (RCS) depressurized with an RCS vent of greater than or equal to (3 2) square inches.
APPLICA8ILITY: MODE 4 when the temperature of any RCS cold leg is less than or equal to (4 Mr)*F, MODE 5 and MODE 6 with the reactor vessel head on. Sof nn ACTION: W
- a. With one PORV inoperable, restore the inoperable PORV to OPERABLE status within 7 days or depressurize and vent the RCS thrcugh a
%mfain the. RCS ,m *.
(3.2) vendedsquare inch conf Jian M4(( vent (s)PostVes both within the havenext I,ece8reste.c hours,4 4. oftg4sur sfalus . 1
- b. With both PORVs inoperable, depressurize and vent the RCS through a (3 2) square inch vent (s) within 8 hoursa niai.6*m N Acs in a. ve.ded ie.4;+;u unH1 hath Poggi. Moe heca res4ed 4 opt 4A8:4 .A.As .
- 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 30 days. The report shall describe the. circumstances initiating the transient, the effect of the PORVs or vent (s) on the transient, and any corrective action necessary to prevent recurrence.
- d. The provisions of Specification 3.0.4 are not applicable.
e e 34 W-STS 3/4 4-M NOV 2 1981
,- REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS 4.4.9.3.1 Each PORV shall be demonstrated OPERABLE by:
- a. ~ Performance of a ANALOG CHANNEL OPERATIONAL TEST on the PORV actuation channel, but excluding valve operation, within 31 days
, prior to entering a condition in which the PORV is required OPERABLE and at least once per 31 days thereafter when the PORV is required OPERABLE.
- b. Performance of a CHANNEL CALIBRATION on the PORV actuation channel at least once per 18 months.
- c. Ve'rifying the PORV isolation valve is open at least once per 72 hours when the PORV is being used for overpressure protection.
- d. Testing pursuant to Specification 4.0.5.
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.
(
"Except when the vent pathway is provided with a valve'which is locked, sealed, or otherwise secured in the open position, then verify these valves open at least once per 31' days.
l l l t l( l W-STS 3/4 4- MV2 198;
~ '
1[ M - 2500 , , , , , , VALID FOR THE FIRST 3.9 EFPY. SETPOINT CONTAINS MARGINS OF 10*F AND 60 PSI FOR POSSIDLE INSTRUMENT ERRORS AND
$ 50'F FOR TRANSIENT EFFECTS.
2000 - - 5 E
~ '
O P = 500 PSIG; T s 190*F P = 364 + 5.752e.01666T; T > 190
- F h
m 2 O n.
, [ 1500
^
k Cc t m" m x>
* $P 'O
$Y m 2 Q a .
5 3 x 1000 - - z in 4 G N 8 E N 500 - - 20 I I I I I l 50 100 150 200 250 300 350 RCS TEMPERATURE (*F)
l REACTOR COOLANT SYSTEM u. 3/4.4.10 STRUCTURAL INTEGRITY LIMITING CONDITION FOR OPERATION 3.4.10 The structural ' integrity of ASME Code class 1, 2 and 3 components shall be maintained in accordance with Specification 4.4.10. APPLICABILITY: All MODES. ACTION:
- a. With the structural integrity of any ASME Code Class 1 component (s) not conforming to the above requirements, restore the structural integrity of the affected component (s) to within its limit or isolate the affected component (s) prior to increasing the Reactor Coolant System temperature more than 50'F above the minimum temperature required by NDT considerations. ,
- b. With the structural integrity of any ASME Code Class 2 component (s) not conforming to the above requirements, restore the structural integrity of the affected component (s) to within its limit or isolate the affected component (s) prior to increasing the Reactor Coolant System temperature above 200*F. -
- c. With the structural integrity of any ASME Code Class 3 component (s) not conforming to the above requirements, restore the strudtural integrity of the affected component (s) to within its limit or isolate the affected component (s) from service.
- 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 reactor coolant pump flywheel shall be inspected per the recommendations of Regulatory Position C.4.b of Regulatory Guide 1.14, Revision 1, August 1975. we l 31 Ld-STS 3/4 4-3El ,NOV 2 1981 i
~ ~
um -
e REACTOR COOLANT SYSTEM 3 /4, a/11 REACTCR CCOLANT SYSTEM VENTS LIMITING CONDITION FOR OPERATION o ne. 3.4.11 At least one Reactor Coolant System vent path consisting of (4we9 vent valvenCa0 and (one) block valve powered from emergency buses shall be OPERABLE and closed at each of the following locations:
- a. jReactor vessel head),
- b. '
(Pressurizer steam space), -and-(,'L. parmhel pathi) 56 (9: :te- cre!::t sy:::m 'igh ;;i,: . APDLICAE!LITY: MODES 1, 2, 3, and 4. ACTION:
- a. With one of the above Reactor Coolant System vent paths inopera:le, !
STARTUP and/or POWER OPERATION may continue provided the inoperable , vent path is maintained closed with, power removed from the valve , actuator of all the vent valves ano block valves in tne inoperaole .
. vent patn; restore the inoperaole vent path to 0?ERABLE status within 30 days, or, be in HOT STANDBY within 6 hours and in COLD SHUTDOWN within the following 30 hours.
- b. With two or more Reactor Coolant System vent patns ino:erable; ,
maintain the inoperable vent path closed with power removec from the valve actuators of all the vent valves and block valves in the
- inoperable vent paths, and restore at least (two) of the vent paths to OPERABLE status witnin 72 hours or be in HOT STANC5Y -
within 6 hours and in COLD SHUTDOWN within the following 3C hcurs. SURVEILLANCE REOUIREMENTS 4.4.1.1 Each Reactor Coolant System vent path block vavle not required to be closed by ACTION a. or b. , above, shall be demonstrated OPERABLE at least once per 92 days by operating the valve through one complete cycle Of full travel from the control room. 4.4.11.2 Each Reactor Coolant System vent path shall be demonstrated OPERABLE at least once per 18 months by:
- a. Verifying all manual isolation valves in each vent path are locked
, in the open position,
- b. Cycling each vent valve through at least one complete cycle of full travel from the control room, and 33 W-STS 3/4 4- M
- E : TOR : n :NT SYSTEM 5 R'.'EILL 'CE RECUIREMENTS (Cer.tinued) t I
- c. Verifying flow through the Reactor Coolant System vent paths duri-ing venting.
e G p W e o S 39 PSTS 3/4 4-M - e e 9
^
JUSTIFICATIONS Section 3/4.5 In the text of Section 3/4.5 where a capital letter with a circle around it appears, please refer to this sheet for the appropriate justification. A. Seabrook Station plant specific data. B. These valves are powered from MCC's and they will deenergize theca valves. C. This section added to cover Modes 4 and 5. The disabling of the accumulator isolation valves is necessary to assure that low temperature overpressurization does not occur. See RAI 440.105. D. At Seabrook, the automatic interlock is set at 365 psig, and the automatic isolation signal has been changed to 660 psig. E. See Seabrook SER Section 5.2.2.2 on LTOP and RAI 440.105 for justification of 305'F temperature. l l I r i .
3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS LIMITING CONDITION FOR OPERATION 3.5.1.t Each reactor coolant system accumulator shall be OPERABLE with:
- a. The isolation valve open,
- b. A contained borated water volume of between 4/25~ and 4400 gallons,
- c. A boron concentration of between (1900) and (2100) ppe, and
- d. A nitrogen cover pressure of between 440 and 4 79 psig.
APPLICA8ILITY: MODES 1, 2, and 3*. ACTION:
- a. With one accumulator inoperable, except as a result of a closed 4 e isolation valve, restore the inoperable accumulator to OPERABLE status within 1-hour or be in at least HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours.
- b. With one accumulator inoperable due to the isolation valve being closed, either immediately open the isolation valve or be in at least HOT STANDBY within 1-hour and in HOT SHUTDOWN within the following 12 hours.
SURVEILLANCE REQUIREMENTS 4.5.1.1 Each accumulator shall be demonstrated OPERABLE:
- a. At least once per 12 hours by:
- 1. Verifying, by the absence of alarms, the contained borated water volume and nitrogen cover pressure in the tanks, and
- 2. Verifying that each accumulator isolation valve is open.
" Pressurizer pressure above 1000 psig.
I W-STS 3/4 5-1 .NOV 2 01980
l EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
- b. At least once per 31 days and within 6 hours after each solution volume increase of greater than or equal to (1% of tank volume) by verifying the boron concentration of the accumulator solution. ,
- c. At least once per 31 days when the RCS pressure is above 2000 psig by verifying that power to the isolation valve operator is disconnected by ::: :' 0' the beeeker 'rce th: circuit. (@)
- d. At least once per 18 months by verifying that each accumulator isolation valve opens automatically under each of the following conditions:
- 1. When an actual or a simulated RCS pressure signal exceeds the P-11 (Pressurizer Pressure Block of Safety Injection) setpoint,
- 2. Upon receipt of a safety injection test signal.
- 4. 5.1. 2 Each accumulator water level and pressure channel shall be demonstrated OPERABLE: I
- a. At least once per 31 days by the performance of a ANALOG CHANNEL I
OPERATIONAL TEST.
- b. At least once per 18 months by the performance of a CHANNEL CALIBRATION.
i ( 4 4 6 6 e SEP 151981 W-STS 3/4 5-2 e
--. p . - - ,,..->,-,-..,,.--,_,e_-----_w ~.-D. g -. . ,,,..m,t r--'- -m *
'*w-- v---- - ~ '+ -v - - -
EMERGENCY CORE C001ING SYSTEMS 3/4.5.1 ACCUMULATORS ( LIMITING CONDITION FOR OPERATION l 3.5.1.2 Each reactor coolant system accumulator isolation valve shall be shut with power removed ' from the valve operator. APPLICABILITY: MODES 4* and 5 ACTION: With one or more accumulator isolation valve (s) open and/or power available tc *he valve operator (s), immediately close the accumulator ' isolation val, and/or remove power from the valve operator (s). SURVEILLANCE REQUIREMENTS
,. 4.5.1.2 Each accumulator isolation valve will be verified shut with power f removed from the valve operator at least once per 31 days.
- When one or more of the RCS cold. legs is less than or equal to 3050 F.
( . 3 3/4 5'3s 40
l i EMERGENCY CORE COOLING SYSTEMS 3/4.S.2 ECCS SUBSYSTEMS - T,yg > 350*F LIMITING CONDITION FOR OPERATION 3.5.2 Two independent Emergency Core Cooling System (ECCS) subsystems shall be OPERABLE with each subsystem comprised of:
- a. One OPERABLE centrifugal charging pump,
- b. One OPERABLE safety injection. pump-{#cer !ce; ;!:nt: cr!y),
- c. One OPERABLE residual heat removal heat exchanger,
- d. One OPERABLE residual heat removal pump, and
- e. An OPERABLE flow path capable of taking suction from the refueling water storage tank on a safety injection signal and automatically transferring suction to the containment sump during the recirculation phase of operation.
APPLICABILITY: MODES 1, 2, and 3. ACTION:
- a. With one ECCS subsystem inoperable, restore the inoperable subsystem to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours.
- b. In the event the ECCS is actuated and injects water into the Reactor Coolant System, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 90 days describ-ing the circumstances of the actuation and the total accumulated actuation cycles to date. The current value of the usage factor for each affected safety injection nozzle shall be provided in this Special Report whenever its value exceeds 0.70.
l n l l 4 W-STS 3/4 5-16 NOV 2 01980 l .
)
! 1 l
EMERGENCY CORE COOLING SYSTEMS - SURVEILLANCE REQUIREMENTS 4.5.2 Each ECCS subsystem shall be demonstrated OPERABLE:
- a. At least once per 12 hours by verifying that the following valves are in the indicated positions with power to the valve operators removed:
Valve Number Valve Function Valve Position a. See a.Tnsev4 I a. h
- b. b. b.
- c. c. c.
- b. At least once per 31 days by:
- 1. Verifying that the ECCS piping is full of water by venting the ECCS pump casings and accessible discharge piping high points, and
- 2. Verifying that each valve (manual, power operated or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.
- c. By a visual inspection which verifies that no loose debris (rags, trash, clothing, etc.) is present in the containment which could be transported to the containment sump and cause restriction of the pump suctions during LOCA conditions. This visual inspection shall be performed:
- 1. For all accessible areas of the containment prior to establish-inE* CONTAINMENT INTEGRITY, and
\PRIMAny '
- 2. Of the areas affected within containment at the compl'etion of
@ d.
g ach containment entry when , CONTAINMENT INTEGRITY is established. At least once per 18 months by:
- 1. Verifying automatic isolation and interlock action of the RHR system from the Reactor Coolant System by ensuring that:
- a) with a simulated or actual Reactor Coolant System pressure signal greater than or equal to 405,psig the interlocks i prevent the valves from being opene',
d and @ b) with a simulated or actual Reactor Coolant System pressure , signal less than or equal to 60ft.psig the
- erlocks will cause the valves to automatica1Ty close. 440 @
- 2. A visual inspection of the containment sump and verifying that .
the subsystem suction inlets are not restricted by debris and that the sump components (trash racks, screens, etc.) show no evidence of structural distress or abnormal corrosion. i l r W-STS
~
3/4 5-A
'NOV 2 1981 l
l l 1
v .. ., , . . n .~ m
... --e. .
u.
,..s. .
.. .s... qr^..*.J*v I
. . q .
.t ;g,p. ,q..w. ~*
. .. < W,. f ,.;* .i y.7. D r..f,- - .,.- - s1
.. q. .. ,.. ,,.7e v. e i
.. s l a. . 1 . .u ._
l .., , . .
. . y. .,. ~. .. ..
s.,. .
. .. . . . , . . . ~ . . . . .
..- :. w _-
r s.
. . ..' _i . .....,' . y ._. [ l . . uf.,. ,V.m.. _J. . fM Cf , f. -_ .- . ,~ , '. . _ .,.,.. . . . .
.u
...3- # .
.i , .
. - <: n . *.
. m, ,m .n. w . ,~ j.c- _os.
,A+ . , . . . . , ,
.ce. o -m .. ?,y...m ;w r. ;,w. ; p* . .
-n.,._. m~._,&..
~y. .. ..7c.. ,, :.e.. ; .e ~
s
. 3;._, . : ; e ., .
.+ n. ..t y--
s.
, .-..,.. %., -..,.a......,_
~. - 4. , - ...
. . ....- . . , ~
j.y -
.. y*In. pos c.V.
l c .g .y~9 . _.Wagwe *.gu} r3.. .. . . ri
;,, . e1
. .% . .-.,3 q f- ..;ys... y _
.> ,rn m._~ , ,Q ** e
_g pn 3 . m.. .. .. m _ . f.3.y- ,
- .~ .. -
p e g n' 4.. m .g,v. p.. ... ,.c.,
.,g q. . g-. ,- - a , . 3: . y . -., -
s' .
. Q
- IL ' . .. [ l ..m;..~
;,n . ,.. M 9" M.. .,. l ~.} m= l ?. L . .Qfen w . v. - . . .e. g. m. -
V,..,e W .:.%.n... a ,.- a, - .
,.,.~.,,,>.~.w.r,. -
.s..
n,. . . ....-. . . + >.m. s ,..m.. .s .,
- . n . .v -.u _ ., n. ..
w..
--.:.. 4 - ,S x y _ g y.. m,.- y .. .. m,,..:. m. -..y,
~
- p. o.
. .. nn; . . . r.:, ns, ..t-of,n
- , , , . . ,- , . . ~
i .
. . . s 3..e .. : a, w, , s. ..
...O SI.V
.:.m. w ... '.vii .
! .. V .x.. .91..,.. .S. W..a.s a. -.
+ ..c f .. .
10 -
.Oftrl ...
.. . ,~ r . 7.,.. . . .
- n. w ,a, ; . -
.. , -,L_ a~ .x
-.n ;.w....,c~...
-. .- - . , . . 3 .7
- 1s.. .: a.-. 7m:. . .- ...--,...s...,.
.; .y 3 . .. . .
.e .
n . ,a. ,, r,' ,c- .. .- m. ..y. c. .:n . n . .e
--- W.,.- ~..)Ws'
~
.... ...+ cgs..y .w. ; .
s sr. Pmp rsoldon . . Open
.. . , . c :. . _
_ . . J_..._,.c as_-V...fi. _. . , M.
, ..m .._ .; ,.
- it
.. _.. n . _ . .m, . . , n , . . _ . . . .
g pen
.x 5 4
. 4.j.*.
4 .
.>. 5... .
<+ _
,..,.,.n-...,
.- . . .- . . 3.. .
.,. . . . ~ . - . . . . .
n^ ..'. .
...S.r Pmp fa c Mt.e9 Isolabn d en
.S. I,..m V.- ... 114 - -. . , ~ . ,. ,.- . .
. s- .. s ww
. ._ . . _ . _ . . , . .......t . m..
.....m.,>
a . . ~ . . . . _ . . . .- ,
.. - ,. . R H ,.V..,14,~.,'
. .. n. '.M. W R, HR A.p h @l.e3.Is,lahon .
. - . - Open
..,c e.. .m ; > .,. . . , ~
. . . .~ . . . . .
.. , . ~ . - . . , .
- r. _ p . Q , .::. + 7 y ,y ,... ~;.,,y. y.,.;. k /. ,.. ,..
, _ . y*; . gg: .. he; .;.g.
. . . ,%.~e m i .
w s.. w ., u ;. - _ , .s .,-....-.t.,~,_....
. oq
. . ,, n, , . ,. u. .
- m. .~. .. . . .
...e,e ,r... ,j .y A..
..oa -
* ' a .e 3. .'.f .;y ,.M . ." ,. ,,y s s
.g.,
_* ^- -
..- n
_ m. .,._,~...,~.%..,c...~,.%.n~.._.
~
.,. v. .: ~ . ,. .
~
.,. ~ .
_ _. a '_._"_._'2R MV- S i _ _i-e<
-.r k__ .R E V ikiLe=J IsoI fno,i - - . . . Clued
~ . .
, .., . e ,.- ..
.......~- .. . ..
-J. . m .~. . -Q,*._M.x.t , *w
- V. . c .' oseJ_
47 .: . . .. .J. ." '. ' .O.. . ..' r
....-.- .a.- - .
m 1. .(h f R. ,5*
'n . .V ., *'
7.0 Ab. QA U f .
-E[,-N'}y 9 , O , g", $# ,Y' e k.4N"' ,
' a *
,a ., i
,.1' f
- c.m.,, m . ,f.m, w n w. n. ~m',, ;..,- e ,,. . . .. -. . . . ,
y . [*d . .i, , _ w . ,, r ,#, : .y .
.-.,L,..e s
...y , .
- s. .e.,g *>..L.;- ,m..r.c
.. s .
,.c 7 . .y . . . ;. .; .; ,. .
~
~v. . . . . - .
u,, . . ....,s.......,y ... . . w t . ~.u.s :. St . . V .Y.1. %..-.. +. Le rseldkn .. close
. m. w:pg.m..
.. . ., .Sr..ko. * :. : -
_m - .
. e ,e r. . 3
~g-
.u.
A t
,,,3
- , ,. sg 4. - s.
c=
.g- <
,;-,.p ; +w4. %,
cjpgej
. <t , .. ., /, ; -
.- w.. - . . . . , .9
. . g. y_ g o z ..-m.cp.2 .s4.;g,y.;
e . - - . . n. .: .n - --..,-A.a
- n. .,.,,.-
.a . . - - .
r s~..' s w 1..,e
~ , , . + . - r- .
.. e.
.* . ' ,. * . , .+.a .%.A',., ,y ;g.%.
r
-- w
. u.
- - -e .g . , .. . , . . .. , n.
L,, <
...s. s .o a .
'..r..
~
....,e a
.. . r.' . . ,...k._,%..
,,,, i v. . ., .a
,~s - . , , ,,
- 5. *
- p. . .
..w,,s . ...._, ~... w. 4 .% . . ~ . . . . . . . -
- . , . . "4w%,.,._. %..%.p ,..,.9. .a ,
a r . %c 4 . . , . p>.. s.
,- ...,%.,s. .
r . ~
- 5 - g .;3f: . . 8 , ., -. .e
, t< f. ; p 4. ,.,*g.;w . ., q. .4 . gp 4.. . s . a, ,9 ,, ,7 s
. . ... .. fg ... .
-. . . ., -..- (..
-..~g , p .,s. ..g. ,.
m
- g. % , w . , , , .. ,
-.I
,.-.. . .-e ; #- .-,
r..waw
., x .
v 4 .= q. .. a t c ,,4, a. . , . ,,- . .-. . ...w-y - .% - .-
., . . g. -
s e &.,,.u. e n. . *m : .y,. ..* -. ..,Q . . w.p f
.-e* .~ - . . ,; n .6 . s.
Q, y s.. e ,p,#' .w v (.4-e
+ . ,s,. - ".
s
*.'m. _
P .4, * **eJ' .j. r,s'*,,..e.,.. p.: c .. , .
'". 4 7 , ? 3. .- '. V , , '. C
. g r t.1 * , .
, , . -c , J s ~r - *.
-.:.y ~ 4. Q~ .,: Q.Q 1 y. Q. .+: +~M*Cf.s.u.:,.g:',.g ,
~
.; :.. y --:..- .
., ;. . . s*a(.-J m;L..,%. . i. , , . , . _ ..- .- .
, a.
.> v ~. .
.t ~a ww .s .w C* *...
; ,+
, y. , . ' .
..4.,.
&Nd *
, , f m.. . c,e y b. A.,..... -A , e r ..r , '.,.s * '.w*.< .*.
g * . . . <
, . 4
...v~..L
, .- .t ..-.s. , .
.e 4
- g. -4..y. py -w _, -
,.y ,,.%, (._ .y .u.,
._... ,y _. . . . . _ . . . . . . ,
.,'w..-.s..,-a .%.m. . g ,s .~v . a;
- g.:..,,..s
.f. . w a .y d
. g* , + ,.,,, g-c.+ 2 m y'q 7 ,..:A ,. c. ..
5 c .. ., r *- t .. . , .. 4. g .,w 2 , , - .
+.*s*
- g#+ 3 , . ; .n.; n.
.. c
.4 *..,.;,.s,.e. : :, ..,
.. ru o. .v...
.,,j;f ; ,.i,- A . (, . P. ,g. +
, -y y c .m.,c y . .
- x. .
.~s e, ..a. e ,a
. $..'y; ..
4 "4 - e C = ,,, y - _.,,r
,s+ _ .,
f. 1 4 y
,,j a f9 7'.. ,S
.e., u. v. , . .. %g , . .p.. .,
./ .w s . .u, 4 m r.,. f .. ..A% g ,., 4 ,-. > . . ..,..a
. . ~ - . - . . - ,.. .
,- : ...t
]* t l; . . , ' v.% _ ?.+
'."s- - , + ' , , .',
3 -. , . . k. Q s$..l.* 4g,
- * , ..e . .: } c '
- y* C;t sf s.,.I.
1,
*"'".y,.".e9.,
.A .t
-m,..-..x .,,m.. . . , . . .w .*' .. .,. ..nc. , - .m w
7t..',t. y #,...s....r-.. .<
.- ~ * ,
. -" o
..-.4 s
5.,,- , y. # .a.,., ,- 4 s . . .
.. 2 4$. .s , s. .,.%..v. .
y N. .e . .- v s. e - t
. / . t.v".,...g. .-,,g'
. . . . .; ,v.,.*.9..,e(1.,** .* .
j g
. . . gOu,- ',. %.="'.y ;
.f ., .s-= % ,[,.a,r'
. . 3.,, .hf . .
a
.,,5 ,4 . ;3..
g., a.g.Ife e,pe.14.eei.e.-.. 3* g fr g
.J Mp enge q.ee.g.> r -
@we , . c.
M e.,.
*. ~, r s. .
..... ' r. .. g , .e $.,e7.,.*
e e y ebf and ?a e. * =up.m. M 8E.4.~.-*=*- - - * - . -.
+ - . - W qnse., e, .e e,m, ..a**=.**
, . .- . - ~ <
s *t ***.e-fe* j L.e
. . j. ,
+
- p. s.;*.
; ~.A
..,,.5 4 ,. ...,,4. a
..s..# , : ...g ,. h. r-. .. . .-.3 .-a4.,.,
~:r.
.m .. ~ , w,. ,..- e ,;. , ., , .g*.s.. m. f ., fe h.e%<..,.,, . ,.
<. m,. ... , ,a ..
. s + . M. .%e. p ,.. .- s..m. s -
n,..,.. 6
*Iff M*
pw
. . M.w' * . V ., a **.=/Ie %- A. .* s .'C L.b w..- 2
-. . ~y e# :n n'u..+a w s w h .w x m[kw a. s.m:t,. vp. kx .g.. w eg ,s
~<.n. w ..w.c. c;9..e.n y ec y . Y:.tsaw.wd.m;h3.w.a n; ..;m e
.2
~~e . . aw ,.xs m .c.;;s.&,c .w%
, a w . w v.c . .a.. p;z . f O
l l EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
- e. At least once per 18 months, during shutdown, by:
- 1. Verifying that each automatic valve in the flow path actuates to its correct position on (safety injection actuation and automatic switchover to containment sump) test signals.
1 ! 2. Verifying that each of the following pumps start automatically upon receipt of a safety injection actuation test signal: a) Centrifugal charging pump b) Safety injection punip c) Residual heat removal pump . i 9^" ^cap ^We of deI*f Mi %f
- f. By verifying that each of the following pumps g th: #ndi :t:d di;;he.;e pressure ^a Specification 4.0.5:
--hWe4'
' - ! :Latfr bc 'l:e when tested pursuant to
/ 1. Centrifugal charging pump 1 2470 psik @ M gp m
- 2. Safety Injection pump 1 / 4 4.5- psik @ 4 S SP "
d
- 3. Residual heat removal pump 1_ 173 psig @ SM 3P*
- g. By verifying the correct position of each electrical and/or mechanical i position stop for the following ECCS throttle valves:
j 1. Within 4 hours following completion of each valve stroking operation or maintenance on the valve when the ECCS subsystems are required to be OPERABLE.
- 2. At least once per 18 months. ' .
W k Head $de.fyI4ed. t Lterwaiafe HeadSafdy Ta.)de Mp64 System -t#A System Valve Number Valve Number 1
- a. st-U-le a. SI y-go -'
- b. SI-g -141 b. si-V- 95*
- c. s r.v - I f I c. sr-V- 104
.d . si.u- 155 d. sr- v- 1oq
- c. SI-V-t q f- Sr v-t2a 3- sz-v-125
- h. .sr V - 111 i
b W-STS 3/4 5-1 .NOV 2 01980 l
EMERGENCY CORE COOLING SYSTEMS , SURVEILLANCE REQUIREMENTS (Continued)
- h. By performing a flow balance test, during shutdown, following com-pletion of modifications to the ECCS subsystems that alter the subsystem flow characteristics and verifying that:
- 1. For centrifugal charging pump lines, with a single pump running:
a) The sum of the injection line flow rates, excluding the highest flow rate, is greater than or equal to 33[L gpm, and b) The total pump flow rate is less than or equal to j;gg gpm.
- 2. For safety injection pump lines, with a single pump running:
a) The sum of the injection line flow rates, excluding the highest flow rate, is greater than or equal to $cfg: gpm, l and b) The total pump flow rate is less than or equal to fugg gpm.
- 3. For residual heat removal pump lines, with a single pump running, the sum of the injection 1.ine flow rates is greater than or equal to gg2&59Pm-1 l
l l . l l l 7 W-STS 3/4 5-K g gg; 4
EMERGENCY CORE COOLING SYSTEMS 3/4.5.3~ ECCS SUBSYSTEMS - T, < 350*F LIMITING CONDITION FOR OPERATION 3.5.3 As a minimum, one ECCS subsystem comprised of the following shall be OPERABLE:
- a. One OPERABLE centrifugal charging pump,#
One OPERABLE residual heat removal heat exchanger, b.
- c. One OPERABLE residual heat removal pump, and
- d. An OPERABLE flow path capable of taking suction from the refueling water storage tank upon being manually realigned and transferring suction to the containment sump during the recirculation phase of operation.
APPLICABILITY: MODE 4. ACTION:
- a. With no ECCS subsystem OPERABLE because of the inoperability of either the centrifugal charging pump or the flow path from the refueling water storage tank, restore at least one ECCS subsystem to OPERABLE status within 1 hour or be in COLD SHUTDOWN within the next l 20 hours,
- b. With no ECCS subsystem OPERABLE because of the.inoperability of either the residual heat removal heat exchanger or residual heat removal pump, restore at least one ECCS subsystem to OPERABLE status or maintain the Reactor Coolant System T**9 less than 350*F by use
.of alternate heat removal methods.
l l
- c. In the event the ECCS is actuated and injects water into the Reactor Coolant System, a Special Report shall be prepared and submitted to the Commission pursuant,to Specification 6.9.2 within 90 days describ-ing the circumstances of the actuation and the total accumulated i
actuation cycles to date. The current value of the usage factor for I each affected safety injection nozzle shall be provided in this Special Report whenever its value exceeds 0.70. i
# A maximum of one centrifugal charging pump and one safety injection pump shall be OPERABLE whenever the temperature of one or more of the RCS cold legs is less than or equal to (FAr)*F. m 305' W ,
l i 8 W-STS 3/4 5-% MAY 151980 l r
EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS 4.5.3.1 The ECCS subsystem shall be demonstrated OPERABLE per the applicable Surveillance Requirements of 4.5.2.1 4.5.3.2 All charging pumps and safety injection pumps, except the above required OPERABLE pumps, shall be demonstrated inoperable at least once per 12 hours whenever the temperature of one or more of the RCS cold legs is less than or equal to (996)*F by verifying that the motor circuit breakers have been removed from heir electrical power supply circuits. 305 l l i 9 W-STS 3/4 5'A MAY 15197? 9
EMERGENCY CORE COOLING SYSTEMS 3/4.5.4 BORON INJECTION SYSTEM BORON INJECTION TANK LIMITING CONDITION FOR OPERATION 3.5.4.1 The boron injection tank shall be OPERABLE with:
- a. AcontainedboratedwatervolumeofY[N$ "'8ff/#"980 _
b gallons, l
- b. A boron concentration of between 20,000 and 22,500 ppm, and
- c. A minimum solution temperature of 145 F.
APPLICABILITY: MODES 1, 2 and 3. ACTION: With the boron injection tank inoperable, restore the tank to OPERABLE status within 1 hour or be in HOT STANDBY and borated to a SHUTDOWN MARGIN equivalent to 1% delta k/k at 200*F within the next 6 hours; restore the tank to OPERABLE status within the next 7 days or be in HOT SHUT 00WN within the next 12 hours. . SURVEILLANCE REQUIREMENTS i 4.5.4.1 The boron injection tank shall be demonstrated OPERABLE by:
- a. Verifying the contained borated water volume at least once per 7 L. days,
.b. Verifying the boron concentration of the water in the tank at least once per 7~ days, and
- c. Verifying the water temperature at least once per 24 hours.
1 10 W-STS 3/4 5-1 APR 15197B I r - - + -
EMERGENCY CORE COOLING SYSTEMS i HEAT TRACING LIMITING CONDITION FOR OPERATION 3.5.4.2 At least two independent channels of heat tracing shall be OPERABLE for the boron injection tank and for the heat traced portions of the associ-ated flow paths. . APPLICABILITY: MODES 1, 2 and 3. dCTION: With only one channel of heat tracing on either the boron injection tank or on the heat traced po'rtion of an associated flow path OPERABLE, operation may continue for up to 30 days provided the tank and flow path temperatures are verified to be greater than or equal to (145) F at least once per 8 hours; otherwise, be in at least HOT STANDBY within 6 hours and in HOT SHUTDOWN within the following 6 hours. t SURVEILLANCE REQUIREMENTS 4.5.4.2 Each heat tracing channel for the boron injection tank and associated flow path shall be demonstrated OPERABLE:
- a. At least once per 31 days by energizing each heat tracing channel, and
- b. At least once per 24 hours by verifying the tank and flow path temperatures to be greater than or equal to (145)*F. The tank temperature shall be determined by measurement. The flow path
( temperature shall be determined by either measurement or recircula- ' tion flow until establ.ishment of equilibrium temperatures within the tank. I l (
\1 W-STS 3/4 5-1q SEP 15 579
'~
w , -. --.
EMERGENCY CORE COOLING SYSTEMS 3/4.5.5 REFUELING WATER STORAGE TANK LIMITING CONDITION FOR OPERATION 3.5.5 The refueling water storage tank (RWST) shall be OPERABLE with:
- a. A contained borated water volume of between 471,oooand gggggallons, @
- b. A boron concentration of between. (2000) and (2100) ppm of boron, and
- c. A minimum water temperature of (35)*F.
APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: With the refueling water storage tank inoperable, restore the tank to OPERABLE status within 1 hour or be in at least HOT STANDBY within 6 hours and in COLD SHUTDOWN within the following 30 hours. i - SURVEILLANCE REQUIREMENTS i 1 4.5.5 The RWST shall be demonstrated OPERABLE:
- a. At least once per 7 days by:
i 1. Verifying the contained borated water volume in the tank, and l
- 2. Verifying the boron concentration of the water. .
- b. ~_At least once per 24 hours by verifying the RWST temperature whee-
@ -the (;;t;id;) ':f - t::p:r?+.'ra fe 1ere th:r 25 c -
i f l 12, W-STS 3/4 5- R APR 15 G78 l l
JUSTIFICATIONS Section 3/4.6 In the text of Section 3/4.6 where a capital letter with a circle around it appears, please refer to this sheet for the appropriate justification. A. The value of Pa = 46.1 psig is Seabrook specific for peak containment accident pressure. See FSAR Section 6.2.6. B. The value of La is changed to 0.75 per FSAR Section 6.2.6.2. C. Section 6.2.6.1 of Seabrook FSAR does not include this method of testing as an alternative. D. Add per commitment in FSAR Section 6.2.6.2. E. The Action statement does not permit increasing temperature to over 200*F if the LCO is not met. F. ANSI - N45.4-1972 changed to ANSI /ANS 56.8-1981 for latest standard that meets all FSAR commitments. 4.6.1.2.a is replaced from a direct requirement of ANSI /ANS 56.8-1981. Partial pressure testing is not allowed by ANSI /ANS 56.8-1981. G. Seabrook design does not have this system. H. These changes will make this Section compatible with ANSI /ANS 56.8-1981 for containment leak rate testing. I. .The Seabrook design is a 36 inch containment purge, and the 1000 hours is per the Seabrook SER pages 6-10, 6-12, and 6-16. l J. The Seabrook design does not use educators, but rather has gravity feed. ! In addition, there is no means to measure flow. However, there is 3/4" drain connection to verify flow. K. Per requirements of Generic Letter 82-16. i L. Seabrook Station plant specific data. O
3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT l CONTAINMENT INTEGRITY LIMITING CONDITION FOR OPERATION l 3.6.1.1 Primary CONTAINMENT INTEGRITY shall be maintained. l. APPLICA8ILITY: MODES 1, 2, 3 and 4. L ACTION: . C s pri vs Without primary CONTAINMENT INTEGRITY, restord CONTAINMENT INTEGRITY within one hour or be in at least NOT STANOBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
. SURVEILLANCE REQUIREMENTS g
4.6.1.1 Primary CONTAINMENT INTEGRITY shall be demonstrated: (' a. At least once per 31 days by verifying that all penetrations
- not
- capable of being closed by OPERABLE containment automatic isolation -
valves and required to be closed during accident conditions are closed by valves, blind flanges, or deactivated automatic valves secured in their positions except as provided in Table 3.6-2 of Specification 3.6.{
- b. By verifying that each containment air lock is OPERABLE per Specification 3.6.1.3.
- c. After each closing of each penetration subject to Type 8 testing, except the containment air locks, if opened following a Type A or B test,byleakratetestingthesealwithgasatP,(kpsig)and ,
b verifying that when the measured leakage rate for these seals is added to the leakage rates determined pursuant to Specifica-tion 4.6.1.2.d for all other Type 8 and gCenetrations, the combined leakage rate is less than or equal to 0.M L,. @ t l = Except valves, blind flanges, and deactivated automatic valves which are i located inside the containment and are locked, sealed or otherwise secured i, in the closed position. These penetrations shall be verified closed during each COLD SHUTDOWN except that such verification need not be performed more often than once per 92 days. W _-DUAL 3/4 6-10 c.E : i :931
. - -m-.,we,-, , ,-w,--+v.e--y,,--g ye -.ya -p -
g---w, , .*--,--w,.-.,..,. ,es--e---y ry- s- w- - - ---S#-.- ---- - -r
,-s-- - ,,--ww
CONTAINMENT SYSTEMS CONTAINMENT LEAKAGE LIMITING CONDITION FOR OPERATION l 3.6.1.2 Containment leakage rates shall be limited to: 1 l a. An overall integrated leakage rate of:
- 1. Less than or equal to La , (0.20) percent by weight of the con-w.
tainment air per 24 hours at P,, (M psig)..bs.
% L::: thea av aq"=1 te L , (0.10) p:r:: t by wei;;ht :f the : r-t & :::t :fr per 24 heur; :t : redu;;d pr ;sur Of "t' (25 W&
75 b
- b. A combined leakage rate of less than or equal to 0.40 L, for all penetrations and valves subject to Type B and C tests, when pressurized
. Ne tdv b l pench dion w'It be alleusca +o exced .r*/, of f4e to tohP,l a llow e d (p os .
t .)
- c. A combined bypass leakage rate of less than or equal to (0.10) L, for l
all penetrations identified in Table 3.6-1 as secondary containment bypass leakage paths when pressurized to P,. APPLICABILITY: MODEX 1, 2, 2 :M ' 5- (or p,1., b ,,h1 *+ N '* # ) @ ACTION: With either (a) the measured overall integrated containment leakage rate exceeding 0.75 L, Or 0.75 gL , :: :p;M ::M , or (b) with the measured combined leakage rate for all penetrations and valves subject to Types B and C tests b exceeding 0. [ L,, or (c) with the combined bypass leakage rate exceeding
- (0.10) L,, restore the overall integrated leakage rate to less than or equal to 0.75 L, Or 1
- :: t.'na,. : : ; 1 to 0.70 L , tes ;pp i::ble, the combined leakage rate for all penetrations and valves subject to Type B and C tests to lessthanorequalto0.NL,,andthecombinedbypassleakageratetolessthan
! or equal to (0.10) L, prior to increasing the Reactor Coolant System temperature ( above 200*F. \ l t W-DUAL 3/4 6-20 OCT I 1979
^
, CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS
- 4. 6.1. 2 The enntainment leakage rates shall be demonstrated at the following test schedule and shall be determined in conformance with the criteria specified in Appendix J of 10 CFR 50 using the methods and provisions of ."f:5! M15.t (1072).
Ausr/us rg.e-t9e1 F See Tnsed Z
- a. T'r:: Type A +==+e (nu. 11 Int ge:t:d ;;,; tai,, ;-t Le p3ge a,te)-
ch:11 5: rea+"-t-d at 10 ; 10 ui,it, iiii.. .el: during :Futda r. et eith:r P, (50 ;:f;) r :t "t (25 ;;f;) du-i-; ==^ 10 y :r ::-"ic - peried- Th= +hird t::t -f :::5 --t k= 5: ;;. ducted dur'n; the _:hu+anun fn th. in ,uome ni r$t in - re4c ;;;;ti; ;, l l b. If any periodic Type A fails to meet enheed.75 L,-07 .75 Lt , the test schedule for subsequent Type A tests shall be reviewed and approved by the Commission. If two consecutive Type A tests fail to g
,- meet-cith;rJ.75 L, er M ' t, a Type A test shall be performed at least every 18 months until two consecutive Type A tests meet either O.75 L, er .75 Lt at which time the above test schedule may be resumed.
- c. The accuracy of each Type A test shall be verified by a supplemental test which:
- 1. Confirms the accuracy of the Type A test.by verifying that the difference between supplemental and Type A test data is within 0.25 L,,-cr 0.25 Lg; i 2. Has a duration sufficient to establish accurately the change in l leakage rate between the Type A test and the supplemental test.
- 3. Requires the quantity of gas injected into the containment or bled from the containment during the supplemental test to be equivalent to at least 25 percent of the total measured leakage at P, psig),c- Pg (25 p;ig).-
@ w.i
- d. Type B and C tests shall be conducted with gas at Pa (M psig) at intervals no greater than 24 months except for tests involving: .
- 1. Air locks,
- 2. Penetrations using continuous leakage monitoring systems, and l W-DUAL 3/4 6-30 NOV 15 B77 l
Y '"# ' [ 't k .g *1[ * .( k s'P 'e' D2. 5 , * , ~.11 ' ..
, 3..
- e ['
. . , .'.J ,
[ ',.T. ;[ d. s . <) ,o.. , W*s .1 . , . .e.+ w,e.(. ; u. ,.e: . , . , % . , , .,. -,q. ;w. .e% s .
. a; -~
-.g y . s - .
,..r.-%,.,'e.,.,.. M< ~,m- ._: ..y..v.cc. . , . " . .s,.- .
g ,-. .eq ., q y .ar ,.,.*,,,,. s , s.
- .,y 7.g.,4 ...e m, , g..
,u a, . ,f.. .<. .
. . .s...w, x, p . - < - o... s.,
.....s<
;f ;. ' . . '. ;
-o . *L,+ .
- .; ,. .e ;+av*~p.,,-=%, .,;,Q .
.sq'. .v. W. ;., y,6,.s T t .t., -. ., -
s; : . ., . . x . - , ~ , c . S. , s ~, .f,. 2..,.
. i. .t** ; ,h e cc..
. f. . v .M , m. . e, v . 3..q
. W -.x , . . Ast'p, e:s' n '.'. , . .
'. ..s. .
.e
.S..
. . n . . p f... . - . , , .,..
.,n. -
...-.e 4.- ,-
i N,.....g ;
,,, N' &, W+y -, #, .ru .%)j!4 , , '* !
- c ,,' . ,. - ~ . ; ,n -e .
r- *
. v. . ,; ..
- o. . w.,- w. , - s < g . .
u- . -
, ,- . 4- .
f '.Q'ln sp+Nj,;rQ & ~
.. ;a i d,)* 1h }{ ~ ~~f* ,l
'h i k- pR. 4.,.u ,y ,.s --,,? ,,*u
. .,. w . . ; , . . . . . . - - . . , , . . .. . . . ..
r d nw.,m...m.< O, E';% '$ y f'sQW4M* ?e5 W)* OS-i e .:.. . . < , .
. ~?;s .v.y s.
"&.^>.' n? Q
"& W Gy* . f- , --
.. s
* - n *
&.,. e _a
. ,._s.~.w,2. :', .C .
.r,, . ea. w :p., . 4.%. % ,
9.as 6 s.m ~
-."..r.p..a..,c.Q.-m.m w % u. .-.w # -_, . y. ,....;. _e - -
. , . , ~c . m ;.,.
2 e s
' .. . , . - ,1 . .. - , a s- ! . , A. -
a h,(*' ;,$'[.. " "
^
t Y . ,- w.[ " , -
/
[ 7
,.- 4( {- C. 'dOh M d .. O- d,,( N .//$/g
-&y.
s,. y~. e.m...o. ~ q,,. m..f.p, s.* .. . ' r a
.y, m%. % .p y t
~,.. c s.., '..t.nn . . . . .:;) . M. . . p~. . , ,.. .n..
.s
, c, .f.M h.~,.._..,.
,. . ,.w..p;
.a. . ..;g.s . a, 4.
.A .< , a,. a.: s -. , . , y, .v , . . s< t. .-a,,.mn
.b
. . ;,-. ... . . . d1. ~ . . ,,; -
~ ..
.L .. . , -
/
y~ < . - - . dP?O C #2. MXl"c . y671*.5.
~-
,,, s S * ' ,
. O $b O $ f YY .
U ' 1_ .s, . V L. ~l' x , * : ,w i& ..alW-e ' .:-s . ' .' . - a
. s
- , -: n< ., =
eus.$.cke,a-$' e ;.::(' . .l.y y'.
.:- m.:. W,l.,2.W."1;C.w?t*y'b.u;u.,s.,%m^ a.- A.0 '.h- .o ,.:x :.'. ..
~ .~. ., .x * :
a ..m*1e.x - ceeE. . . . .-
, s c+wn, . ,. % > ,4 . ,.
.m.,. 4,%t . % ms
.s m u a .m
.1; .q. ,4* . :. . -w5_ m .. t. . r ? c< ,- %a, ..~.
3.
- ,. w .
..:.A
.W%. . .e . . i.
. _ ,, =i.t w
-- n,
- 1. , . . .,
. .s-; n. ,
%.-3.c . , .,@ < y,.-
o ,,.;.p
,t
..,g ;.
,.i.p a. ~,, W. %.wbG.
. f ., - ,?.e \a. .v s, .,_e y.' .,7. . , .
m , .. m y ..,, ,, 5 . t ,. *. , * <.
..n.e. c*'..N'Y y . -~. .%,j e % >= -
*e-
- > g,..-" 4 , W -. . 5 f, _ .
-. ,- ~- - ~ ,3 , - ,
~-- A '", - A h. vW-Y-A $n.. .. ?='h.,Q*&,, .'La.A t..w.e,'.<+.-
a -
-k-~&, , - 's.k, .c + U . k *. u *. --
+ ./. , ,m. $.~.Iir. *'.,s.
=<;1 s ..P. %.
' ~
e,*,i et 4 . -et..6,.,*p......W w. . # . 4 ' ,3 . . =/ '... ,,A ..- .- - m..
,. - - .,s. g *.,v
[ , ., A
- r. .
[, [.' 4*.4 .?.,.,...,
. ,,3 nd. '.
. . . . . ., M, 4 r.,'r.' "..L 4.6 d,,~J, 4* $gM, W :+4,1 C. .I, $.~.
, . ~.g ..e,.e,q p m..-[2,.
'. 1.
- ., . .. f.a.> -.
v ..f ... .=4.y,.' v. a . .
',s
. , .N* N .,*ya .e
.P N
,.."'s(.., ,'g 7.e= ;' g - -
. , , p.
+ : ., , .; s.;<,,.., .
. . s -; - , , .-: ,
y.
....,.,,m......,3.%...,_,.s.w..,p-,,.,.3 ._,,,.a .
~ . , _ . , , . - , . . _ ... . . , . . .
a .. ,
-.t-
* ) ,-. . se .1..
.- _ , , . . gy f r.
.3 ,,
, .f;r a, .f m ~ -
n$ -q.9,g e,. ".M qr% .,. O , ; s .-W g t-yc < % ,, .,. .;. ,,v. ~ ~ ,,s. g j U ', - -
. -. . s. . 3.s q.7. y n?.f' t ,p' t .. .,R.,;,i.'.._w.,...,p_....,....
> 1
. .t,:,. , , ,_ ]. *
. ..~.
-y .a .. : . . . -
, o .- . .. ~/,L .* - : r;, .- ,v~
' ~ ' -
- . :' , 'z* -
.,~2i,- 4
, -.. ~ ,.. . - ~
. .. ._- - . . .. . .A -
4.. ..
?.,: ,. , .
g m. . .,.. . . ,.o,
- v.h. ,, p. ..q . . .?.. . , .#...,.- . . . . . ..
. _L.,y.
.. .. ,- . . . - o
... - _ . . N,.. .
-c.'*.e. x.
.. . - .*. . N,
-a.
5 q ~. .
.,*y 3 ; ,-
3 .~ , - m-I - -. ,.
+- m
. w 4
- a. ..,-.-
.-Q. .
-. e
.s .
.4
. g. ,
.P.* a . , <fP.S
- , . . .. ~ . ,'
.,r **
o.. ++ ,
.,.r...
.4.,.. .2 t
, - . , _.,.~e - * , ,.' ,p - ', .* .,
I .
.<.e.,,'*-
i
,s. N~. r
. , ' ! ". ; . ,g. rt ,.h '. N.;.',.... w N,. . >. .~, ,s; u ...;.#, m.s, .CN n> W,;y<, x ;.g.'f .,s., .
{
, ,y...... . . . . . .. ,. . ., ,
. , - .z ....:.e.,,,. .v a. ., .~....,- u..- -n, . . . . . . .
a~%.... .: - .,a,a,y.L
.m, 's .;, . . . y
. . , , . , . .a ,o. .m. - .s.y.g<
.'..,.,.,rN..,..
y, c..
>. , w. .w.; w. .
m..;J...r_.g.,.n
,. p . , .. .,
m s .-< s
- y. a
. , . s . ,- .,. . ,.
- q. c . %,gw,eN .. \,.,,. . a ...m . a :, s .e. .we e,q; ,n , . .4$ +%.t %.s. ,s .,., c ..
a. y .~ . - ; .
. ; 4
*. .. p. .&. .- - 3c. -
L veym. s
. . . -.. .,-.-4.e..%.., ,%.;w-
,...r
..a,
.. ...,p,.. *=. . a. . .. , e m era ,A.
t
-y. t, . : 4 ,.,. . ._ _
,.a,n,. .p. ?. .. .. ,u.
.n,._.,...~..,a.. ;..u vn.
u,
.. s .
.'ae,. ..gf .pp 4 ., a sr. ,..'[';' m..*
, g * #
h ,' s ,/ .'g *g *. . . , ,
- y. ,., .w.r. ,.~
.;- a v n +
._. . .. . s 1. 9a....,..~..'s,.7 p .
-a.-
4.,. ...,,.;,. . , . ,
. w'
, .. r . v .a .-y.-
, - y . a *.. a, ,.' a~e. ,.. 4
...< .. . " ; .w # :., . n . M. C ,~e *'., c .: , -- ', .*. -
- ..q..=. r p. -
. . -- *...y'...n..
. -.- . . . - ~ - , . , , ~. ' A.s .s.,ne.,.,e.
, . ss..,
x . ._..-9,.,- . w-- s . s. ,e y.s ,
,.....u
.a n . - . . . - .. .. .
.m
*',"p_:.
..- ~ . * " ' , ,D
* . p' .y ? ..r . 2 ; n.::.... . f i . m, \c* :A. ,. .
., ,. ...9. ,..~,2~.. ,,s ,,,s w: s* .-
- - . . i,
,_y <
, . . e <e.
6A..,<.;>_,.*,*.,. m v'.m s , o _ y, e
. - . . < . ~ . . ,
.$,%. .w.,s . .a 4-. ,n ....- .-. _ .
- : ,. . ; y s.%.9 + ' . .. .n--
". * ?- ..~. - -
s.
~
, ,. . J,, ~, , ;, e ,..n- ;. .
...,.. . n se. e. r , . . ~ .,>.. .x.. , y .*,W .. s
.- * > s
._,,,.,,.~,;*a.
w%. ... .
.~_p*
w.....'~..-.-s
.i .. -
,
- wJ. * *$ ",4',,1**.. *A% _': . s ' - . l, , ' . '. 2.- . = , , . * - a
. .+.*-.e6J.*,4.'.-..,pa+.
- e- >.w
- . s p' '6.
g
- . , .& .e-e4 $
,,' ' * *.a.
#2 W.(
,* a # ' "
P
" . < - ' . , , ' .' . . :a6 .. ..
, ,-s.."[,'.'..' . <*. . .. ,.
.I y>. e .
., . **t
.s v :. .' r *
.,.L,-
i
. .; 1 3
; .m 1
l ~ - . . p .a> w. .~ V.,r'e ,e , , s 4 .s pn....,.,e. ,
.4. .e.......
..-f.*
,;,:..,../ - . . . - ..v._-
. . ,h ,4.,. ,. . . ., w
. e. s.... - . .
..('
I i s
, /.
e.
. i >, . ,. .,,.g,' ' ' . ,a.
, ya
. g.
<+ e ..v...,e.,',s. a..
- w.,.,,a am . . . .p , w -
_,, . . -*4
-t. '"t,.* t 3 *s. ,
3
*,e.....e. ..'sp.
. x . .
I ... . . . - ,.g..=.--me,,,4 1 y e ,,. .- * ' - + i 4 - *. e h ..em*g. g e . b eh /,.* . s=b. . ' 9m, a. .,,..*.a
..,,-,i.p.#
.. .+ .
-**..4
- f
+o. p c., -/ .- 8 ,. e i e
z m , ..* . s ,.
.se . .1' .+ .,e n a
.e
,.'--*m%,._ jpp e.r.%-s .f* ['. , +'.4 .
g 4
, ,e.- , ..J' - .,.s .
.g .- .$_
...-e-
- ai -
o 'w. ., , e. 3e i+, ' , p r eegpa g g.h.=me . p. t
.v. ~ '+ .ge e e>=w ,.e e- ee 4 .
,'A-yo.d'es.y.
,m
.g qsg , > - -
.-...e.=pe .e s g . . ...-.a e .. . . O J . ', s s * * *%.
' ** { e
- u. ..s' < ,e 1,a ,
.,,.yf~'....f^*y'...'.',,".
,., ~ a. ;x , . , n...: - x a ..
. - .. . ~ . . v
.p m =
-<.. - - ... vn.a _..,,s,,c.Q.. m e r t
.+-~. +w......w
%,-. =*.e.--i-.s.,,,,,..u.%..e
" w 4 e r-r .,e.. ,.a .v
- y j.
.a es.J.u %, s. .s_ ..i -m ,;
A.n,+s.-a.*
%er,,,.,-
-;-. s. - -
-- Du,'l h';% .,* * ., ( *. J g een fr., .:yN. /. .*? yw < -}% l; *'.'Qu.' . N U.
e
-, .+' .. . .n.e.., , g _',..i ;*.
- N a. **Y-,S ( f ,. ,e
~.~, .. .. -.w . , , . .,.c.,e .
*. ,.v
. w ,
m
~
1 D.a,
.._ s'.....?.. '*y# _f.y,.^.p.a ,.+
*;a M 6 w,%.* t, .' . Y
- wm"-.'.p(.,
e
.O a
- r~.,4';,
' .'.,3.-? --
, g.
8' y
, <* ,'. 7
.- + va r.g_.n no ' - . . ..-. , - - . -
.7 e ' .h : a
./,.i . . ' ." , J
..s,p'
* ,. ,, % sa % a - sw -N ', /
. 3,-j ., L*
.,;, ,s . ,
a,; . =4 4. ar.. ac --a c.'g# _(v,' e . p
- . .. - . ..s
.~.W ).
I p. y y . . , ,e, p.
- b WW e gb y y .y c, ..,_y -
s f u _ e
,.g s
a
- y.1
*. , ..[.,Jq'e.h"a; .*t M...i., ...4d+ &-* ,wwe"v,- e e., . y 6
_-.er 1- ,s sy.as s,~a.,,g .s.3,
- a. , .,
..wy.%,
ks =f. ~or aig. m n~ e y,- .',~f.%.s .s,.,, e
- 9. ~. .~ m ;.',c_ ,v y g. a.o. . a p . ~ ..,.,.s n
m ;s . .e o., p ic- en.e ne - 4
's
.-e- sl .g. :.3 =m* , k.L , ** *y+s s g ;l:%*nr. .K,,g,. .g. , w
.j w'
. no. pg s% n j% *
/ - . '.h.
=w1' m', (*
-ne f
.4, ,.
+*+, ,. ,
r _-- v., w. , ., .
./.
,..s.-_.3.*...
g,.e,,
- . , ,wi , .,. . ',- - , p y, y; . . , '* p.m .,7 c 2..; y.., ;> , .,. * , , ,
w, n.* v. -
"* - , '** a -
- 4 ..- . - * * -
p es' et.'4". p* .e,* -.,*=*.ar y y =- oo.er se p saism.s w ;y--sN= *o ch.'%-@* *= <e- -em 4 e- db' 8. p .. ;. *== . 8 * % .d k w e4e--*-*.-e-.
. , - - - . . - = . --.;~. -
.,. * * " - * -ff ",
- g' d ;. " ,
- 4,* + *. " * -' *
*[-
k, ,. . ? ,y%
,9.
er. ,. i*
.. o. ,5w.- . , u.
- 0 ;a"*
. .s 3 .,g a . ,.e. p. A .> < . 44.
3 t
.q . -
, . W
. . .. %= .. .. 4 h
,k 6
,s .,,..
. .. 7. a . . . . .g
,.,3 s,4,.~-,%., g -
3
'e -
n . .. N.,m .: . r,p . #. ..-f,,.s,.,b-.e. ., , , r: . *j .. a n e .% ' ,.g, r- a. . ~ . ,6 . .
. . . . . - . ~ . . , . . . .
.~.4,. g ,,,V,.4,, 4. --**,; . % ,9.
.* -n. . pow., .e.4, m%, s]. . a3,.,..t,A.,,,,.,,p.,=;,'..,.,,
a..
. .. -- r .s .n y g.44. , e, : 3s ,, -- ,; , ., , , _ * = * .
.g ., ,j .
-^;e.
=
.r -
- 6. .,y a j
,- [m
' s4 [ g ' ',.~ . . J.'. ~ T. *,. . m"
- vt ' ", g L + ' [,b ' ,). 3..
e . r a %, .}' vos.n,
= ,,
s' [* f.a j, L., ' ' ,A., 5 *4 .' .A. * ,%.
- b. . . , b,. ,.J,..n.g~ ,
+
- h. ~- . -. . .
7 '.. . ~ , . . . - . . , 4 .. .~ .o .s . a
-- 4 .*
b-
'L:-
- y y'..,_,;=*- .:. p>
d.,. 4.' '
-y v*s .. -
h g,, :.'r . ' n ,, ,. :. ' ; , . <; ;
...e.. .
. *_ . . ~. % 4 .C.A y.
I 6 -. r , . ,Q...w. %.ns. ,e4
*** c ,r,%;s.s ,m. .a
% *g,.m-' . +*4 w ,/,,a,m W* ,-.g.-<p--- .;. . s-w -...,w r- _.
6 n ... - *
.3
,, . e ,** . ,
j . / h fe = ; c4 -. * = 3(. . . . ~ I "' s .% s y > e(." , g + ,7 h , f.[-*, ,;. ,, I. p ,
'.,,,,'-,-..,w*"',
..,~sj
.~m
<A; .-.,y m 3P .w.a ;
-e . ,, e,/ ..,.v. .g
,..g,- r.+-3 ~.;, ay arp+ ~*Y.es ---.s. +* --ewa.y,.4 .44 e .fe.x %+,.'r,p n%.~f ,fg. ; e** cp.. . 9.- p ., -
Q , g 44 *% y, ; wf,jg O -
*~s. %.o- r . . j~ ..Nfe , 3 a -
.=
p q p.~a,q A % . c,, ., .,=, , %y.
.cJr.,g* . ,~,.* ' ,2 ,,r *,* p ,.* :. .*. * - , ,
..y
&$$y g $.5_bl~.'YphDL QY..b,Q.%. w 4.
% h : $ ;$ D -, &{h&;_ N$hj%.,. - , ,fYl4
.e g sJ p . ~.+ * -
p efk.$h .14+&m'pY . YYNy QW$Yh -
s CONTAINMENT SYSTEMS 7-SURVEILLANCE REQUIREMENTS (Continued)
- 3. Valves pressurized with fluid from a seal system.
- e. The combined bypass leakage rate shall be determined to be less than or equal to (0.10) L, by applicable Type 8 and C tests at least once per 24 months except for penetrations which are not individually testable;penetrationsnotindividuallytestab,1ghagb4 egerginedd tohavenodetectableleakagewhentestedwitnre:pcurr::dwhileg thecontainmentispressurizedtoP,(kpsig)duringeachTypeA test.
- f. Air Locks shall be tested and demonstrated OPERABLE per Surveillance Requirement 4.6.1.3.
'E. Typ: peri:dic t;;t: are n t r;quir;d f:r p:nctr tion: continuou:ly meritered by the Centeinmuni IsviaLun Vol.e end Cheonci Mcid P.
i:: tier Sy:ter: provided th: syst c~ er; OPERASLE p;r Sur.eillence esser h '
-R;;uir;;;nt 1.5.1.0.
3 'ts Leakage from isolation valves that are sealed with fluid from a seal system may be excluded, subject to the provisions of Appendix J, Section III.C.3, when determining the comoined leakage rate provided the seal system and valves are pressurized to at least 1.10 Pa (55 psig) and the seal system capacity is adequate to maintain system pressure for at least 30 days. h% Type B tests for penetrations employing a continuous leakage monitoring system shall be conducted at Pa (% psig) at intervals no greater than once per 3 years. g4 l k. 1 The provisions of Specification 4.0.2 are nnt applicable. l t i 1, , W-DUAL 3/4 6-4D E " i
TABLE 3.6-1 SECONDARY CONTAINMENT BYPASS LEAKAGE PATHS PENETRATION RELEASE LOCATION e t g-P e Eo a s 8 i 0-P P+~ 4- S'
? 3 d-w G
tn
.C E
o H O vs a m
l
. l l
CONTAINMENT SYSTEMS CONTAINMENT AIR LOCKS I LIMITING CONDITION FOR OPERATION 3.6.1.3 Each containment air lock shall be OPERABLE with: a.'- Both doors closed except when the air lock is being used for normal transit entry and exit through the containment, then at least one air lock door shall'be closed, and
- b. An overall air lock leakage rate of less than or equal to 0.05 L, at P,, (M psig).
@ 44.I APPLICABILITY: MODES 1, 2, 3 and 4.
ACTION:
- a. With one containment air lock door inoperable:
- 1. Maintain at least the OPERABLE air lock door closed and either restore the inoperable air lock door to OPERABLE status within 24 hours or lock the OPERABLE air lock door. closed.
- 2. Operation may then continue until performance of the next required overall air lock leakage test provided that the OPERABLE air lock door is verified to be locked closed at least once per
- 31. days.
- 3. Otherwise, be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
- 4. The provisions of Specification 3.0.4 are not applicable.
- b. With the containment air lock inoperable, except as the result of an inoperable air lock door, maintair, at least one air lock door closed; restore the inoperable air lock to OPERABLE status within 24 hours
( or be in at least HOT STAN0BY within the next 6 hours and in COLD l SHUTOOWN within the following 30 hours. i W-DUAL 3/4 6 MAR 151978 _.
CONTAINMENT SYSTEMS SURVEILLANCE REQIJIREMENTS 4.6.1!3 Each containment air lock shall be demonstrated OPERABLE:
- a. Within 72 hours following each closing, except when the air lock is being used for multiple entries, then at least once per 72 hours, by verifying no detectable seal leakage by pressure decay when the volume between the door seals is pressurized to greater than or
@equaltoP,44.I(M psig) for at least 15 minutes, .
- b. By conducting overall air lock leakage tests at not less than P ,
W./ (30 psig), and verifying the overall air lock leakage rato is within its limit:
- 1. At least once per 6, months,# and spet= Aa,
- 2. Prior to establishing' CONTAINMENT INTEGRITY when maintenance has been performed on the air lock that could affect the air lock sealing capability."
- c. At least once per 6 mcnths by verifying that only one door in each
, air lock can be opened at a time.
I The provisions of Specification 4.0.2 are not applicable.
- Exemption to Appendix J of 10 CFR 50.
PDUAL 3/4 6-7D SEP 15181
ONTAINMENT SYSTEMS C INMENT ISOLATION VALVE AND CHANNEL WELD PRESSURIZATION SYSTEMS (OP DNAL) LIMITING ONDITION FOR OPERATION ,/ x 3.6.1.4 The c tainment isolation valve and channel weld pressurization f systems shall be PERABLE.
/
l APPLICABILITY: M00 1, 2, 3 and 4. ACTION: With the containment.isol ion valve or channel weld pressurization system inoperable, restore the ino rable system to OPERABLE' status within 7 days or be in at least HOT STAN08Y wi in the next 6 hours 'nd in COLD SHUTOOWN within the following 30 hours. SURVEILLANCE REQUIREMENTS X 4.6.1.4.1 The containment isolation va ve p essurization system shall be
/
demonstrated OPERABLE at least once per 31 day by verifying that the system j is pressurized to greater than or equal to 1.10 (55 psig) and has adequate i capacity to maintain system press re for at'least 3 days. 4.6.1.4.2 The containment ch nel weld pressurization s stem shall be demon- ! strated OPERABLE at least o e per 31 days by verifying th t the system is ! pressurized to greater th or equal to P, (50 psig) and has dequate capacity i to maintain system pres re for at least 30 days. ( l l l l l MAR 151978
-0UAL - 2/' f 00
? .
; CONTAINMENT SYSTEMS INTERNAL PRESSURE LINITING CONDITION FOR OPERATION 3.6.1.k Primary containment internal pressure shall be maintained between f.gigt, and Igigy psig.
A'PPLICABILITY: MODES 1, 2, 3 and 4. l ACTION: - With the containment internal pressure outside of the limits above, restore the internal pressure to within the limits within 1 hour or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTOOWN within the following 30 hours. SURVEILLANCE REQUIREMENTS 4.6.1. The primary containment internal pressure shall be determined to be within the limits *t least once per 12 hours. l l l l
?
PDUAL 3/4 6-10 , MAR 151978
CONTAINMENT SYSTEMS AIR TEMPERATURE LIMITING CONDITION FOR OPERATION 5 3.6.1.4 Primary containsa...t average air temperature shall not exceed /20 'F APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: With the containment average air tempera'ture greater than /20 *F, reduce the average air temperature to within the limit within 8 hours, or be in at least HOT STANOBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE REQUIREMENTS . r wpi hted 4.6.1.li The primary containment average air temperature shall be the -an,_g:ticci- .. average of the temperatures -:t th: #:awiaa ' ace +4aae and shall be determined at least once per 24 hours: 7 .
'As deteemmed by a. mani,nu,m of }
L:::ti:n v7, of -Ae tempen_bc de.techs dse A fo, Type A confammenf +ests h b. c.
- d. N e.
1 i
/ -
3/4 6- JUL 151979 PDUAL . n -- - .n..,- . - - . . ,.
} CONTAINMENT SYSTEMS CONTAINMENT VESSEL STRUCTURAL INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.1. The structural integrity of the containment vessel shall be maintained atalevelconsistentwiththeacceptancecriteriainSpecification4.6.1.)
APPLICABILITY: MODES 1, 2, 3 and 4. 1 ACTION: With the structural integrity of the containment vessel not conforming to the above requirements, restore the structural integrity to within the limits prior to increasing the Reactor Coolant System temperature above 200*F. SURVEILLANCE REQUIREMENTS 4.6.1. The structural integrity of the containment vessel shall be determined during the shutdown for each Type A containment leakage rate test (reference
'.' Specification 4.6.1.2) by a visual inspection of the exposed accessible interior and exterior surfaces of the vessel. This inspection shall be performed prior to the Type A containment leakage rate test to verify no apparent changes in appearance of the surfaces or other abnormal degradation. Any abnormal degrada-l tion of the containment vessel detected during the above required inspections j shall be reported to the Commission pursuant to Specification 6.9.1.
l e. ( 3/4 6- DEC 151978 W-0UAL
~
CONTAINMENT SYSTEMS $ CONTAINMENT VENTILATION SYSTEM LIMITING CONDITION FOR OPERATION 7 3.6.1.% The ( inch) containment purge supply and exhaust isolation valves shall be sealed closed. Operation with the (8 inch) purge supply and/or exhaust isolation valves open shall be limited to less than or equal to (39) hours per 365 days. sooo APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: With the ( inch) containment purge supply and/or exhaust isolation valve (s) open, or with the (8 inch) purge supply and/or exhaust isolation valve (s) open g, _for more or be in than g ) HOT STANOBY within the next 6 hours and in COLD SHUTDOWNhours per 365 d at least within the following 30 hours. SURVEILLANCE REQUIREMENTS 3 4.6.1. I b(, m<A, containment purge supply and exhaust isolation valves shall be verified to betleded c\osed at least o nce per 31 days. g
-0. C M :d :t least sa;; ;;r 24 P.;;r:.
-t. k oled - cle::d :t 1 ::t ence pn 31 d;y--
4.6.1.k.2 The cumulative time that the (8 inch) purge supply and exhaust isolation valves have been open during the past 365 days shall be determined at least once per 7 days. l 4.6.1. 3 At least once per 6 months on a STAGGERED TEST BASIS each sealed l closed ( W inch) containment purge supply and exhaust isolation valve shall be ( demonstrated OPERABLE by verifying that the measured leakage rate is less than or equal to (0.05) L,. 7 4.6.1.14 At least once per 3 months each (8 inch) containment purge supply and exhaust isolation valve shall be demonstrated OPERABLE by verifying that l the measured leakage rate is less than or equal to (0.05) L,. t
\\
W-DUAL 3/4 6-120 SEP 2 81981
CONTAINMENT SYSTEMS 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS CONTAINMENT SPRAY SYSTEM ( ndit t & r 'er iedin: rrev a' )- LIMITING CONDITION FOR OPERATION 3.6.2.1 Two independent con'tainment spray systems shall be OPERABLE with each spray system capable of taking suction from the RWST and transferring suction to the containment sump. APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: With one containrent spray system inoperable, restore the inoperable spray system to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hou.as; restore the inoperable spray system to OPERABLE status within the next 48 hours or be in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE REQUIREMENTS 4.6.2.1 Each containment spray system shall be demonstrated OPERABLE:
- a. At least once per 31 days by verifying that each valve (manual, I power operated or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.
- b. By verifying. + b+ aa -eci c & tica c','eachpump'$:$8ia a l disee p pressure of greater than or equal to later psig when l tested pursuant to Specification 4.0.5.
- c. At least once per 18 months during shutdown, by:
i 1. Verifying that each automatic valve in the flow path actuates to its correct position on a ,j ,_ __ test signal. l 2. Verifying that each spray pump starts automatically on a
.. _ _ _ _ _ test signal.
- d. At.least once per 5 years'by performing an air or smoke flow test
- through each spray header and verifying each spray nozzle is unobstructed.
12 JAN 151979
-W-DUAL 3/4 6-M D ,
e e i I
i
\
l TAINMENT SYSTEMS 3/4 2 DEPRESSURIZATION AND COOLING SYSTEMS 1 CONTAI ENT SPRAY SYSTEM (No credit taken for iodine removal) LIMITING C ITION FOR OPERATION 1 3.6.2.1 Two in 3endent containment spray systems shall be O RABLE with each spray system capa e of taking suction from the RWST and tra sferring suction j to the containment ump. APPLICABILITY: MODES , 2, 3 and 4. ACTION:
- a. With one containme spray system inoper, le and at least (four) containment cooling ans OPERABLE, res e the inoperable spray system to OPERABLE st us within 7 da or be in at least HOT STANDBY within the next 6 hours nd in COLD S TDOWN within the following 30 hours.
- b. With two containment spray s tem. inoperable and at least (four) containment cooling fans OPERA , , restore at least one spray system to OPERABLE status within 72 hq6 or be in at least HOT STANDBY within the next 6 hours and in' COL SHUTDOWN within the following 30 hours. Restore both spray s of initial loss or be in av,y' stems t OPERABLE least. HOT status within 7 days ANDBY within the neXt 6 hours and in COLD SHUTDOWW within the fo owing 30 hours.
- c. With one containment s y system inoperabl and one group of required containment cooling fyns inoperable, restore ither the inoperable spray system or the fnoperable group of coolin fans to OPERABLE status within 72 h @ rs or be in at least HOT STA BY within the next 6 hours and in CO @ SHUTDOWN within the following 0 hours. Restore both the inopera e spray system and the inoperable roup of cooling fans to OPERAB status within 7 days of initial los or be in at least HOT STA BY within the next 6 hours and in COLD UT00WN within the f lowing 30 hours.
l l SURVEILLANCE REQUI EMENTS s 4.6.2.1 Each containment spray system shall be demonstrated OPERABLE:
- a. least once per 31 days by verifying that each valve (manual, power operated or automatic) in the flow path that is not locked sealed or otherwise secured in position, is in its correct position.
DUAL - ;/; C 5 MAR 151978
INMENT SYSTEMS h ,, y SURVEILLANCE IREMENTS (Continued)
- b. By verifying, that recirculation flow, ea.ch' pump develops a discharge pressure of ater than or equ,al'to psig when tested pursuant to Specification 0.5. ,, ' ,
- c. At least once per 18 months, dur' shutdown, by:
- 1. Verifying that each automatic valv in the flow path actuates to its correct pos' on on a signal, and
- 2. Verifying th each spray pump starts.automati lly on a est signal.
- d. At least ce per,5 years by performing on air or smoke flow st throu each spray header and verifying each spray nozzle is uno tructed.
f 1 l n (, y-0UAL 2, '-inn JUL 151979 O
I CONTAINMENT SYSTEMS SPRAY ADDITIVE SYSTEM (OPTIONAL) LIMITING CONDITION FOR OPERATION 3.6.2.2 The spray additive system shall be OPERABLE with:
- a. A spray additive tank containing a volume of between 1500 and gallons of between 19 and Al percent by weight NaOH Q 33ga_
solut ion,and
% Tee :pr:y rdd4+4ue educter: ::ch ::pable Of ddia; he0" selutivo
-fre ti,e c;m ica! cdditive t: 9 te : nt:in: cat !?"ey syst e ?""? - g fl e .-
APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: With the spray additive system inoperable, restore the system to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours; restore the spray additive system to OPERABLE states within the next 48 hours or be in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE REQUIREMENTS ! 4.6.2.2 The spray additive system shall be demonstrated OPERABLE:
- a. At least once per 31 days by verifying that each valve (manual, power operated or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.
- b. At least once per 6 months by:
- 1. Verifying the contained solution volume in the tank, and
- 2. Verifying the concentration of the NaOH solution by chemical I
analysis.
- c. At least once per 18 months, during shutdown, by verifying that each automatic valve in the flow path actuates to its correct position on a g______ test signal.
f d. At least once per 5 years by verifying b solution flow r:t (t be d:t: mi.ed dur'ng pr: eparatiermi io.L;)-from the feller %g dreia
- ----+4 , 4-
'" spray additive cy t::: bk and -the RWsr,
- 1. 'Orsir 'nc ivuauun) ., ; p: 7 % 4 cas-vm 6vm,(em s46 l
- 2. -{0 e4a "ne !:::tien) ; cmn TAmg c33.ym4vic (%gg l
G MAR 151978 W-DUAL 3/4 6-MD .
1 1
, 40NTAINMENT SYSTEMS CO NMENT COOLING SYSTEM (OPTIONAL) (Credit taken for iodine removal b spray ystems)
LIMITING NDITION FOR OPERATION s , 3.6.2.3 (Two independent groups of containment cooling fans s all be OPERABLE with (two) fan ystems to each group. (Equivalent to 100% co ing capacity.) APPLICABILITY: DES 1, 2, 3~and 4. ACTION:
- a. With one gro of the above required contai ent cooling fans inoperable and both cont inment spray systems OPERABL , restore the inoperable group of cooli fans to OPERABLE status ithin 7 days or be in at least HOT STAND within the next 6 hou and in COLD SHUTDOWN within the follow g 30 hours.
- b. With two groups of he above require containment cooling fans inoperable, and both ontainment s ay systems OPERABLE, restore at least one group of co ing fans t OPERABLE status within 72 hours or be in at least HOT ANDBY wi hin the next 6 hours and in CCLD SHUTDOWN within the foll wing hours. Restore both above required 7
groups of cooling fans to PE BLE status within 7 days of initial loss or be in at least HOT ANDBY within the next 6 hours and in COLD SHUTDOWN within the f owing 30 hours.
- c. With one group of the ab er uired containment cooling fans inoperable and one containment spr syste inoperable, restore the inoperable l spray system to OPERAB status ' thin 72 hours or be in at least
! HOT STANDBY within t next 6 hou and in COLD SHUTDOWN within the following 30 hours. Restore the in erable group of containment cooling fans to OP ABLE status withi 7 days of initial loss or be in at least HOT ANDBY within the nex 6 hours and in COLD SHUTDOWN within the foll ing 30 hours. SURVEILLANCE REQUIREME S l 4.6.2.3 Each grou of containment cooling fans shall be monstrated OPERABLE: l l a. At lea t once per 31 days by:
- 1. Starting each fan group from the control room an erifying that each fan group operates for at least 15 minute
. Verifying a cooling water flow rate of greater than o equal to gpm to each cooler.
At least once per 18 months by verifying that each fan group I starts automatically on a test signal. t t
--DUAL -
2/' 9-PD MAR 15 7 l l l l
/
/
/
CO INMENT SYSTEMS CONTA ENT COOLING SYSTEM (OPTIONAL) (No credit taken for iodine remo al by spray sy ems) LIMITING C0 TION FOR OPERATION - 1 3.6.2.3 (Two) 1 ependent groups of containment cooling fans ,s 11 be OPERABLE with (two) fan sys as to each group. (Equivalent to 100% cooling capacity.) APPLICABILITY: MODES , 2, 3 and 4. ACTION:
- a. With one group of he above required contajhment cooling fans inoperable and both containmen spray systems OPERABLE, restore the inoperable group of cooling fans to OPERABLE statur/within 7 days or be in at least HOT STANOBY with the next 6 ho 'rs and in COLD SHUTDOWN within the following 30 ours.
- b. With two groups of the abo requir d containment cooling fans inoperable, and both contain ent ppray systems OPERABLE, restore at least one group of cooling fa o OPERABLE status within 72 hours or be in at least HOT STANDBY y hin the next 6 hours and in COLD SHUTDOWN within the following/30 urs. Restore both above required groups of cooling fans to OPERABLE tatus within 7 days of initial loss or be in at least HOT TANDBY w hin the next 6 hours and in
- COLD SHUTDOWN within the llowing 30 ours.
I c. With one group of the a ve required con inment cooling fans inoperable and one containment s y system inoperab , restore either the inoperable group of tainment cooling fan or the inoperable spray system to OPERA 8LE (atus within 72 hours or e in at least HOT STANDBY within the ext 6 hours and in COLD SH TDOWN within the following 30 hour . Restore both the'inoperabl group of containment cooling fans an the inoperable spray system to 0 ERABLE status within 7 days initial loss or be in at least H0 STANDBY within the next 6 ho s and in COLD SHUTOOWN within the fo owing 30 hours. i SURVEILLANCE REQUIRE,NTS
, s j
4.6.2.3 Each gro of containment cooling fans shall be demonstrate OPERABLE:
\
- a. At 1 st once per 31 days by:
~
I
- 1. Starting each fan group from the control room verifying th t
. each fan group operates for at least 15 minutes.
l
- 2. Verifying a cooling water flow rate of greater than or equal o gpm to each cooler.
i
. At least once per 18 months by verifying that each fan group
! starts automatically on a test signal.
- L , 6-166 MAR 151979
_-0UAL
/
/
e 4 C TAINMENT SYSTEMS 3/4 IODINE CLEANUP SYSTEM (OPTIONAL) LIMITING NOITION FOR OPERATION 3.6.3 Two in endent containment iodine cleanup systems
/
.all be OPERABLE.
APPLICA8ILITY: ES 1, 2, 3 and 4. ACTION: With one iodine cleanu system inoperable, restore,the inoperable system to OPERA 8LE status within days or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTD0 within the fo11owingf30 hours. SURVEILLANCE REQUIREMENTS 4.6.3 Each iodine cleanup system allj demonstrated OPERABLE: .
- a. At least once per 31 days o /a STAGGERED TEST BASIS by initiating; from the control room, flgs rough the HEPA filters and charcoal adsorbers and verifying f. hat t e system operates for at least 10 hours with the heaters on. .
/ '
l b. At least once per 18 Aonths or (1) fter any structural maintenance l on the HEPA filter or charcoal adsor r housings, or (2) following painting, fire or emical release in ny ventilation zone communicating with the system b :
- 1. Verifyin hat the cleanup system sat fies the in place testing accepta criteria and uses the test p cedures of Regulatory Positi C.5.a C.5.c and C.5.d of Regu tory Guide 1.52, Revisi n 2, March 1978, and the system flo rate is cfm i1 .
- 2. V ifying within 31 days after removal that a 1 oratory analysis a representative carbon sample obtained in ac rdance with egulatory Position C.6.b of Regulatory Guide 1.5 Revision 2, March 1978, meets the laboratory testing criteria o Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, arch 1978.
- 3. Verifying a system flow rate of cfm 1 10% during syst l
operation when tested in accordance with ANSI N510-1975. W-DUAL 4'4 6-isu JUN 1 1979
._. ~ . . _. . _ . . . . - ~ . _ . _ - - . . . ~ - _ . . _ . _ . . _ _ _ _ _ . _ _ _ . _ . _ _ . . _ _ . . _
~
/
TAINMENT SYSTEMS SURVEIL E REQUIREMENTS (Continued)
- c. After very 720 hours of charcoal adsorber operation by ver fying within 1 days after removal that a laboratory analysis o a repre-sentativ carbon sample obtained in accordance with Reg atory Position .b of Regulatory Guide 1.52, Revision 2, M ch 1978, meets the 1 oratory testing criteria of Regulatory P sition C.6.a of Regulatory Guide 1.52, Revision 2, March 1978.
- d. At least once p 18 wonths by:
- 1. Verifying tha the pressure drop across e combined HEPA filters and ch coal adsorber banks is ess than (6) inches Water Gauge whil operating the syste at a flow rate of cfm 1 10%.
- 2. Verifying that the s stem starts either a Safety Injection Test Signal or on a C tainment ressure -High Test Signal.
- 3. Verifying that the filte coo ing bypass valves can be opened by operator action.
- 4. Verifying that the heater 'ssipate 1 kw when tested in accordance wi AN N510-1975.
- e. After each complete or pa ial repla ment of a HEPA filter bank by verifying that the HEPA ilter banks r ve greater than or equal to (99.95)%" of the 00P w n they are test in place in accordance with ANSI N510-1975 ile operating the stem at a flow rate of cfm 1 1 ..
- f. After each compl e or partial replacement o a charcoa] adsorber bank by verify g that the charcoal adsorbers move greater than or l equal to 99.9 of a halogenated hydrocarbon re igerant test gas when they a tested in place in accordance with NSI N510-1975 while oper ting the system at a flow rate of fm 1 10%.
l l R 99.95% plicable when a filter efficiency of 99% assomed in the safe 6 l analy s; 99% when a filter efficiency of 90% is assumed. s i APR 15197 W-DUAL C/4 0 200 % a w -- , - - _ , , _ _ _ _ _ - _
CONTAINMENT' SYSTEMS 3/4.6l CONTAINMENT ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3
- 3. 6.'4 The containment isolation valves specified in Table 3.6-2 shall be OPERABLE with isolation times as shown in Table 3.6-2.
MODES 1, 2, 3 and 4. APPLICA8ILITY: ACTION: With one or more of the isolation valve (s) specified in Table 3.6-2 inoperable, maintain at least one isolation valve OPERABLE in each affected penetration that is open and either:
- a. Restore the inoperable valve (s) to OPERABLE status within 4 hours, or
- b. Isolate each affected penetration within 4 hours by use of at least one deactivated automatic valve secured in the isolation position, or O
- c. Isolate each affected penetration within 4 hours by use of at least one closed manual valve or blind flange; or
- d. Be in at least HOT STAND 8Y within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
SURVEILLANCE REQUIREMENTS 4.6. 1 The isolation valves specified in Table 3.6-2 shall be demonstrated OPERABLE prior to returning the valve to service after maintenance, repair or replacement work is performed on the valve or its associated actuator, control or power circuit by performance of a cycling test and verification of isolation time.
- (
14 W-DUAL 3/4 6-hD JUN 1 1979
T CONTAINMENT SYSTEMS , SURVEILLANCE REQUIREMENTS (Continued)
- 4. 6.' 2 Each isolation' valve specified in Table 3.6-2 shall be demonstrated OPERABLE during the COLD SHUTDOWN or REFUELING MODE at least once per 18 months by:
- a. Verifying that on a Phase A containment isolation test signal, each Phase A isolation valve actuates to its isolation position.
- b. Verifying that on a Phase 8 containment isolation test signal, each Phase 8 isolation valve actuates to its isolation position.
- c. Verifying that on a Containment Purge and Exhaust isolation test signal, each Purge and Exhaust valve actuates to its isolation position.
3
- 4. 6.'4. 3 The isolation time of each power operated or automatic valve of Table 3.6-2 shall be determined to be within its limit when tested pursuant to Specification 4.0.5.
[ l l l l - W _-DUAL 3/4 6-SEP 2 81981 a
- s. y ,
r b [> TABLE 3.6-2 CONTAINMENT ISOLATION VALVES
- MAXIMM VALVE NUMBER FUNCTION ISOLATION TIME (Seconds)
A. PHASE "A" ISOLATION 1. 2- . D TS" y B. PHASE "B" ISOLATION P
- 1. _, 3 w
s 2- +' n i i
- C. 9 CONTAINMENT PURGE AND p?
EXHAUST k Y. 1. f2 2.' g+ D. MANUAL 1. [ m
- 2. 3 e
~
E. OTHER o-
- 1. (*
r> - 2. E LO ,- $w "May be opened on an intermittent-basis under administrative control. y j 8 Not subject to Type C leakage tests.
**The provisions of Specification 3.0.4 are not applicable.
l l
CONTAINMENT SYSTEMS , 3/4.6.' COMBUSTIBLE GAS CONTROL HYDROGEN MONITORS LIMITING CONDITION FOR OPERATION
- 3. 6.'5.1 Two independent containment hydrogen monitors shall be OPERABLE.
l APPLICABILITY: MODES 1 and 2. ACTION: With one hydrogen monitor inoperable, restore the inoperable monitor to OPERABLE status within 30 days or be in at least HOT STANDBY within the next 6 hours. SURVEILLANCE REQUIREMENTS i 4.6.4.1 Each hydrogen monitor shall be demonstrated OPERABLE by the performance of a CHANNEL CHECK at least once per 12 hours, a ANALOG CHANNEL OPERATIONAL TEST at least once per 31 days, and at least once per 92 days on a l STAGGERED TEST BASIS by performing a CHANNEL CALIBRATION using sample gas containing:
- a. One volume percent hydrogen, balance nitrogen.
- b. Four volume percent hydrogen, balance nitrogen.
l 17 W-DUAL 3/4 6 '24D SEP 151981
CONTAINMENT SYSTEMS ELECTRIC HYDROGEN RECOMBINERS - W LIMITING CONDITION FOR OPERATION 61 3.6.1.2 Two independent containment hydrogen recombiner systems shall be OPERABLE. APPLICABILITY: MODES 1 and 2. ACTION: With one hydrogen recombiner system inoperable, restore the inoperable system to OPERABLE status within 30 days or be in at least HOT STANDBY within the next 6 hours. SURVEILLANCE REQUIREMENTS 9 4.6.4.2 Each hydrogen recombiner system shall be demonstrated OPERABLE: L f
, a. At least once per 6 months' by verifying during a recombiner system functional test that the minimum heater sheath temperature increases to greater than or equal 700'F within o0 minutes. Upon reaching 700 F, increase the power. setting to maximum power for 2 minutes and verify that the power meter reads greater than or equal to 60 kw.
l b. At least once per 18 months by:
- 1. Performing a CHANNEL CALIBRATION of all recombiner instrumentation and control circuits,
- 2. Verifying through a visual examination that there is no evidence of abnormal conditions within the recombiner enclosure l
(i.e., loose wiring or structural connections, deposits of i foreign materials, etc.), and
- 3. Verifying the integrity of all heater electrical circuits by performing a resistance to ground test following the above required functional test. The resistance to ground for any heater phase shall be greater than or equal to 10,000 ohms.
1 l l 1
@ 0 1981 W-DUAL 3/4 6-25D AUG
CO AINMENT SYSTEMS HYOR PURGE CLEANUP SYSTEM (If less than two hydrogen recombiners available LIMITING CO ITION FOR OPERATION 1 - 3.6.5.3 and capable ofAcontanmenthydrogenpurgecleanupsystemshallbeOPERAB)s. being wered from a minimum of one OPERABLE emergency .
. APPLICABILITY: MOD 1 and 2.
ACTION: With the containment hyd gen purge cleanup system inoper. le, restore the 4 hydrogen purge cleanup sys em to OPERABLE status within days or be in at
. least HOT STANDBY within 6 ours.
SURVEILLANCE REQUIREMENTS X 4.6.5.3 The hydrogen purge cleanu system sha be demonstrated OPERABLE:
- a. At least once per 31 days b init ing, from the control room, flow through the HEPA filters and ha coal adsorbers and verifying that the system operates for at lea 10 hours with the heaters on.
- b. At least once per 18 months r (1 after any structural maintenance on the HEPA filter or char al adso ber housings, or (2) following t painting, fire, or chemic release any ventilation zone l communicating with the stem by:
- 1. Verifying that e cleanup system sa isfies the in place testing accept ce criteria and uses e test procedures of l
Regulatory P itions C.5.a C.S.c and .d of Regulatory Guide 1.52, evision 2, March 1978, and e system flow rate is cfm 10%.
- 2. Verifyi g within 31 days after removal that a aboratory analy s of a representative carbon sample obt 'ned in acco dance with Regulatory Position .C.6.b of Reg latory Gui e 1.52, Revision 2, March 1978, meets the lab ratory t ting criteria of Regulatory Position C.6.a of R ulatory uide 1.5.2, Revision 2, March 1978.
- 3. _ Verifying a system flow rate of cfm i 10% during s stem l
' operation when tested in accordance with ANSI N510-1975.
g' AUG 6 81 DUAL - 3/4 C ' ~
,. , - - - - - - - . - ,- - . - - + , - - , - - , - - -
NTAINMENT SYSTEMS SURVE LANCE REQUIREMENTS (Continued) , s
- c. ter every 720 hours of charcoal adsorber operation by ve fying wi hin 31 days after removal..that a laboratory analysis o a repre-sen tive carbon sample obtained in accordance with Reg atory Posi on C.6.b of Regulatory Guide 1.52, Revision 2, M rch 1978, meets he laboratory testing criteria of Regulatory P sition C.6.a of Reg atory Guide 1.52, Revision 2, March 1978.
- d. At least ce per 18 months by:
- 1. Verifyi g that the pressure drop across e combined HEPA filters nd charcoal adsorber banks is ess than (6) inches Water Gau while operating the syste at a flow rate of cfm 1 10%.
- 2. Verifying tha the filter coolin ypass valves can be manually opened.
- 3. Verifying that th heaters di sipate + kw when tested in accordanc with A I N510-1975.
- e. After each complete or par a replacement of a HEPA filter bank by verifying that the HEPA filt banks remove greater than or equal to (99.95)%* of the DOP when e are tested in place in accordance with ANSI N510-1975 while pera ing the system at a flow rate of cfm 1 10%.
- f. After each complete o partial repi cement of a charcoal adsorber bank by verifying th t the charcoal sorbers remove greater than or equal to 99.95% of halogenated hydr arbon refrigerant test gas when they are tes d in place in accor nce with ANSI N510-1975 while operating he system at a flow rat of cfm 1 10%.
99.95% appl cable when a filter efficiency of 99% is assumed in he safety an yses; 99% when a filter efficiency of 90% is assumed.
-DUAL 3/ " 4-220- --
_ - _ - - _ _ _ _ _ _ ___---_______._-.___._.-___________-___.____.9_ ~_
CONTAINMENT SYSTEMS g HYOROGEN MIXING SYSTEM (Optional) LIMITING CONDITION FOR OPERATION 3.6. Two independent hydrogen mixing systems shall be OPERABLE. APPLICABILITY: MODES 1 and 2. ACTION: With one hydrogen mixing system inoperable, restore the inoperable system to OPERABLE status within 30 days or be in at least HOT STANDBY within the next 6 hours. SURVEILLANCE REQUIREMENTS
- 4. 6. .i Each hydrogen mixing system shall be demonstrated OPERABLE:
- a. ' At least once per 92 days on a STAGGERED TEST BASIS by starting each system from the control room and verifying that the system operates for at least 15 minutes. .
- b. At least once per 18 months by verifying a system flow rate of at least lq4e cfm.
l l l i r O W-DUAL 3/4 6- 0 AUG 6 1981 t
/
/
/
ONTAINMENT SYSTEMS - 3 6.6 PENETRATION ROOM EXHAUST AIR CLEANUP SYSTEM (OPT'IONAL)
/
LIMITI CONDITION FOR OPERATION 3.6.6 Two in ependent containment penetration room exhaust 'ir cleanup
/
systems shall OPERABLE. APPLICABILITY: ES 1, 2, 3 and 4. ACTION: With one containment pe tration room exhaust air g eanup system inoperable, restore least HOT theSTANDBY inoperable sy tem within t to OPERABLE next 6 hours and statu/s 1 COLDWithin 7 days within SHUTDOWN or be the in at following 30 hours. SURVEILLANCE REQUIREMENTS
~
s 4.6.6
/
Each containment penetration r m exhaust air cleanup system shall be demonstrated OPERABLE:
- a. on a GGERED TEST BASIS by initiating, At fromleast theonce perroom, control 31 day _ low throu the HEPA filters and charcoal
- adsorbers and verifyf that the sy tem operates for at least 10 hours with the heat s on.
- b. At least once per ji8 months or (1) afte any structural maintenance on the HEPA filtpr or charcoal adsorber usings, or (2) following painting, fire or chemical release in any entilation zone communicating with the syst by:
- 1. Verify g that with the system operating t a flow rate of l cfm 0% and exhausting through the HEPA ilters and charcoal adsorbers, the total bypass flow of the sys em to the facility vept, including leakage through the system 'verting valves, is 1,ess than or equal to 1% when the system is t sted by admitting (For systems w th diverting l
/ cold DOP at the system intake.
valves.) l 2 Verifying that the cleanup system satisfies the i place testing acceptance criteria and uses.the test procedures of Regulatory Positions C.5.a, C.5.c and C.5.d of Re latory Guide 1.52, Revision 2, March 1978, and the system fic rate is cfm + 10L l I _-DUAL - 2/46-au JUN 1 37
___ _ . _ _ - _ - - = - - - ~ - - - - . f # kONTAINMENTSYSTEMS SURV LLANCE REQUIREMENTS (Continued)
/
Verifying within 31 days after removal that a laborat y analysis of a representative carbon sample obtained in accord ce with Regulatory Position C.6.b of Regulatory Guide 1.52, evision 2,
.rch 1978, meets the laboratory testing criteria f Regulatory P ition C.6.a of Regulatory Guide 1.52, Revisio 2, March 1978.
- 4. Veri ing a system flow rate of cfm + 1 during system operat' n when tested in accordance with XN 'N510-1975.
- c. After every 72 hours of charcoal adsorber ope tion by verifying within 31 days a er removal that a laborato nalysis of a repre-l sentative carbon le obtained in accordance with Regulatory Position C.6.b of R ulatory Guide 1.52, R ision 2, March 1978,
- meets the laboratory sting criteria of R ulatory Position C.6.a of Regulatory Guide 1. , Revision 2, Ma h 1978.
- d. At least once per 18 month by:
- 1. Verifying that the press e dr across the ccmbined HEPA filters and charcoal adso er anks is less than (6) inches Water Gauge while operating he system at a flow rate of cfm + 10%.
4 i
- 2. Verifying that the syst starts n a Safety Injection Test
! Signal.
- 3. Verifying that the f iter cooling byp s valvss can be manually opened.
- 4. Verifying that e heaters dissipate + kw when tested in acco dance with ANSI N510-1 W ~
l
- e. After each comp 1 e or partial replacement of HEPA 11ter bank by verifying that e HEPA filter banks remove greater han or equal to (99.95)%* of e DOP when they are tested in place in ccordance with ANSI N5 -1975 while operating the system at a f1 rate of cf + 10%. '
- f. After e ' complete or partial replacement of a charcoal ad rber bank b verifying that the charcoal adsorbers remove greater han or .
. equal o 99.95% of a halogenated hydrocarbon refrigerant test s who they are tested in place in accordance with ANSI N510-1975 wh e operating the system at a flow rate of cfm + 10%. -
N 99.95% plicable when a filter efficiency of 99% 1.s assumed in the safety analyses; 99% when a filter efficiency of 90% is assumed. ( p at e s.,nn _ APR 151978 s y *- - ,. ,. ..-..,y..,,....y-- ,,y----.-m.,,,,. . . . . _ . , _ , , . . . .-m.w_.-_,.~,___.-_.m ,_- - _ . .. - - -- - . _ . . . . - - .
d
/
TAINMENT SYSTEMS N h /
,/
3/4 G M VACUUM RELIEF VALVES ,/
/
LIMITING CO ITION FOR OPERATION /
/
3.6.7 The prima containment to atmosphere vacuum relieVvalves shall be OPERABLE with an ac ation set point of less than or equal to psid. APPLICABILITY: MODES 2, 3 and 4.
/
ACTION: With one primary containment atmosphere vac'uum relief valve inoperable, restore the valve to OPERABLE s tus withinj4 hours or be in at least HOT STANDBY with the next 6 hours and n COLDfSHUTDOWN within the'following 30 hours. SURVEILLANCE REQUIREMENTS j f 4.6.7 No additional Su eillance Requirements other t. n those required by Specification 4.0.5. L { PDUAL ' /6 MI 1973 w - " = ~ ~ e , -- ,,--- ,---s ,
l l .- CONTAINMENT SYSTEMS _h g M 3/al.C.C CONTh/Mt1EAYT UNCLOSURE QUILDING-courmuneArt KNCLDsutt stus.p/No- rNTEGRITy i LIMITING CONDITION FOR OPERATION
=
teourniuneur sucasuari 3.6. M BUILDING INTEGRITY shall be maintained. APPLICABILITY: MODES 1, 2, 3 and 4. ' ACTION: scouve m eur ausLo m ; qecurs,annrur a ncLC$UARVJ Without-C;!!Lf BUILDING INTEGRITY, restore RiK LO' BUILDING INTEGRITY within 24 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN
- .<within the following 30 hours.
~
SURVEILLANCE REQUIREMENTS S'. I ' ' #"*"""#" # #" " " # E < l 4.6.1.1 SiHEt#8UILDING INTEGRITY shall be demonstrated at least e nce per 31 days by verifying that each door in each access opening is closed except when the access opening is being used for normal transit entry and exit, then at least one door shall be closed.~ , .
., / .
f i s [ l l J. < l . jj i ? . l
', fg :
^, ,L.r', $.
..< . ,/,
f , .wa
)
\. j' p ww f 10 .~
W-DUAL 3/4 6-MD a SEP 1 1979 l > (
' s* * * ' #
CONTAINMENT SYSTEMS ( 0/t. .0 CONC'" CONT *J NENT 5"!:LO ""!LDIE ^.!" CLE.'""" SYSTC",- cournwntur ENLL osuRE EntAMDCY EKHAMsT AAC ENcLosutL cooLINC- S45TCH LIMITING CONDITION FOR OPERATION S 2. t< sa..a.a <>eto, ee e am. e, ewt.a <.t?e 3.6. M Two independent :M::: tut:d Mg :: c ema"Psystems shall be OPERABLE. APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: i c=4c', ,# culosw . e,, <r e= A4 W ***l'd b With one sa ::: system inoperable, restore the inoperable
~
rm !c!ag ei ::= system to OPERABLE status within 7 @ days or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE REQUIREMENTS 52 t cab".. cat ene.losure u.os.cy ulaud . 4c=tk , 4.6.1L.\ Each :hte!d buiMkg air c :=u;"! system shall be demonstrated OPERABLE:
- a. At least or.ce per 31 days on a STAGGERED TEST BASIS by initiating, from the control room, flow through the HEPA filters and charcoal adsorbers and verifying that the system operates'for at least 10 hours with the heaters on.
- b. At least once per 18 months or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire or chemical release in any ventilation zone communicating l with the system by:
(
- 1. Verifying that with the system operating at a flow rate of Idcr cfm + 10% and exhausting through the HEPA filters and charcoal adsorbers, the total bypass flow of the system to the facility vent, including leakage through the system diverting valves, is l less than'or equal to 1% when the system is tested by admitting cold 00P at the' system intake. (For systems with diverting valves).
- 2. Verifying that the cleanup system satisfies the in place testing acceptance criteria and uses the test procedures of Regulatory Positions C.5.al C.S.c and C.S.d of Regulatory Guide 1.52, Revision 2, March 1978, and the system flow rate is h cfm + 10%.
1 1 l .11 W-0UAL 3/4 6-MD MAR 151978 i
4
, ONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
- 3. Verifying within 31, days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978.
I 4. Verifying a system flow rate of Jdgg cfm + 10% during system . operation when testad in accordance with KNSI N510-1975.
- c. After every 720 hours of charcoal adsorber operation by verifying within 31 days after removal that a laboratory analysis of a repre-sentative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, l meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978.
- d. At least once per 18 months by:
- 1. Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than (6) inches lC Water Gau'ge while operating the system at a flow rate of W cfm + 10%.
- 2. Verifying that the system starts on a Safety Injection Test
( Signal.
- 3. Verifying that the filter cooling. bypass valves can be manually opened.
- 4. - Verifying that each system produces a negative pressure of greater than or equal to (0.25) inches W.G. in the annulus
, within (1) minute after a start signal.
- 5. Verifying that the heaters dissipate gl v 1 kw when tested in accordance with ANSI N510-1975.
- e. After.each complete or partial replacement of a HEPA filter bank by verifying that the HEPA filter banks remove greater than or equal to '
(99.95)%* of the DOP when they are tested in place in accordance with ANSI N510-1975 while operating the system at a flow rate of h cfm 1 10%.
- f. After each complete or partial replacement of a charcoal adsorber bank by verifying that the charcoal adsorbers remove greater than or
- equal to 99.95% of a halogenated hydrocarbon refrigerant test gas when they are tested in place in accordance with ANSI N510-1975 while operating the system at a flow rate of h cfm + 10%.
N 99.95% applicable when a filter efficiency of 99% is assumed in the safety analyses; 99% when a filter efficiency of 90% is assumed. 2A W-DUAL 3/46-4 JUL 15 379
, - . , , . . , . . - , , . , . ..-.-- _-. - . , . , - . . . - _ . . - , - - , _ . , . - - - -- -- n. -.-
CONTAINMENT SYSTEMS (COMTA/WNseT EocLosta AFs
;;;KL"' BUILDING STRUCTURAL INTEGRITY h LIMITING CONDITION FOR OPERATION g ,ceskinme<+ eulowe, 3.6 4 3 The structural integrity of the rM e? ? building shall be maintained at a level consistent with the acceptance criteria in Specification 4.6.1L3.
S APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: t en&SnmeAf enclosure, With the structural integrity of the :h' !?' building not conforming to the above requirements, restore the structural integrity to within the limits within 24 hours or be in at least HOT STANDBY within the next 6 hours and in CO,LD SHUTDOWN within the following 30 hours, i SURVEILLANCE REQUIREMENTS
- 4. 6.g t cenWtamentenelowe i 1.3 The structural integrity of the :M:?fbuilding shall be determined i during the shutdown for each Type A containment leakage rate test (reference l Specification 4.6.1.2) byyi,suaQsp,egn of the exposed accessible 'nterior and exterior surfaces of we :=:::suui sums and verifying no apparent changes l
in appearance of the concre g su abnormal degradation of N : T g:1- ace or,other abnormal detected degradation. during the above 'Any required inspections r, hall be reported to the Commission pursuant to Specification 6.9.1. I W-DUAL 3/4 6- ApR 3 0 m
n JUSTIFICATIONS SECTION 3/4.7 In the text of Section 3/4.7 where a capital letter with a circle around it appears, please refer to this sheet for the appropriate justification. A. No loops at Seabrook can be isolated. B. Seabrook is not authorized to operate with less than four loops in operation. C. Seabrook Station terminology is " emergency" rather and " auxiliary" feedwater system or component. D. There are only two emergency feedwater pumps at Seabrook (1 motor driven and 1 steam driven). E. The emergency feedwater pumps only take suction from the condensate storage tank. There is no alternate source of water. F. Seabrook Station plant specific data. G. The STS only checks for water level and temperature of the ultimate heat sink. With two service water pumps running, the flow would be about 21,000 gym. If there is a problem with the water flow to the service
- water pump house, water level would decrease and the plant would enter an ;
; LCO because of level, not flow.
H. Action items a & b were essentially the same. This clarifies what action asst be taken if the pump house becomes inoperable. I. The change to ,72 hours brings this into line with the Section on Service Water Operability, while the change to 24 hours is consistent with the STS. 4 J. The STS is not applicable to the Seabrook design. Seabrook relies on two remote air intakes rather.than filtering. Therefore, the existing Seabrook Tech Specs , as revised, are used.
! K. The fire pumps capacity is 1500 gpa each.
L. Seabrook has two diesel driven fire pumps and one motor driven fire pump. s M. ANI has accepted a yearly frequency for system flush. N. As a minimum, enough fuel should be available for eight hours of pump operation. i P. Added per commitment letter SBN-399 dated December 3,1982 to the NRC. l l
Q. This Sections intent is to prove the snubbers are functional as. designed (operable). Step f.5 does not provide any acceptance criteria for snubber operability but rather, goes into structural integrity of the hanger and connacting pins which is not addressable in this Specification. R. ASTM-D270-65 is to be dropped from use in 1984 and replaced with ASTM-D4057-81. This change is made to take that action into account at this time. I \ l I i i l -
q' 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE l SAFETY VALVES LIMITING CONDITION FOR OPERATION 3.7.1.1 All main steam line code safety valves associated with each steam generator Of an uni::1:ted reerter ceelant 1:Op shall be OPERABLE with lift settings as specified in Table 3.723. (@) z _ APPLICABILITY: MODES 1, 2 and 3. ACTION: 4
- a. With Cn) reactor coolant loops and associated steam genarators in oper'a tion and with one or more main steam line code safety valves inoperable, operation in MODES 1, 2 and 3 may proceed provided, that within 4 hours, either the inoperable valve is restored to OPERABLE status or the Power Range Neutron Flux High Trip Setpoint is reduced per Table 3.7-1; otherwise, be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
- 16. With (n-1) rc::ter ceelent 1::p: :nd :::::iated :tes; generatcrs in-Op;rsti:n 2nd with :n; er ;;r: ;;in stee; line ced: : fety v !v :
-::se:f ated with . epereting 10:p 'n per:b!:, Oper:tien in MODES 1, 2 end ,3 ::y preceed' pre"ided, th:t withi ' h;;r:, either th:
(h) #reper:ble volv: i: r;;ter:d t: OPERABLE st:tu: Or th: . :w;r R:nge---
";otren flux "igh Trip Sotpeint is r:duced pe- Table ?.7-2; otherwi::,
4; in et loest HOT STAND 6V wiuiin Uiv nen 6 iivui s and in CCLC
- ""T00MM within the following 30 h: r .
b NL The provisions of Specification 3.0.4 are not applicable. SURVEILLANCE REQUIREMENTS 4.7.1.1 No additional Surveillance Requirements other than those required by
-Specification 4.0.5.
! W-STS 3/4 7-1 MAR 151979
)
l I l l ( l TABLE 3.7-1 ; l MAXIMUM ALLOWABLE POWER RANGE NEUTRON FLUX HIGH SETPOINT WITH INOPERABLE STEAM LINE SAFETY VALVES DURING % LOOP OPERATION 4 , Maximus Number of Inoperable Maximum Allowable Power Range l Safety Valves on Any Neutron Flux High Setpoint Operatina Steam Generator (Percent of RATED THERMAL POWER) l 1 (87) 2 (M)65 g l 3 (W)43 l - l l TC LE 2.7 2 l \ MAXIMUM ALLOWABLE POWER RANGE NEUTRON FLUX HIGH SETPOINT ]pY$ M PERABLE STEAM LINE SAFETY VALVES DURING N-1 LOOP OPJ, RATION Maximum Number of In rable Maximus All e Power ~ Range Safety Valves on Any Neutro ux High Setpoint Operatina Steam Generator (Perce of RATED THERMAL POWER) l 1 (52) l ! 2 (38) 3 (25) l I
"At least tw safety valves shall be OPERABLE on the-non-operating am i generato G
1 l l l W-STS 3/4 7-2 MAR 151979
= - +
- w y o.-2+ - , - , . - , , . . , , _ , , _
TABLE 3.7-y .
@ STEAM LINE SAFETY VALVES PER LOOP v.
VALVE NUMBER LIFT SETTING (1 1%)* ORIFICE SIZE
- a. V6.V21 V3(,,VSO / I FS- psig R (/4 set in 3 b ., V1,V23; V3% Vsl /203 psig "R h4.5g M
- c. VS, V24. V39; VS2. 12.1 0 psig "R //4 sa ler 3
- d. V9 V29 V39;VS3 3 /23 9 ps1g "R'# [/4 9 ds)
M
& c. V ' * > V*"' V40 : v su 12sr p,3 R (14 % in )
w "The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure. w H I CJT 5 U 1 P
I i l PLANT SYSTEMS i IEt1GRG gac.Y l AWH tf*RY FEEDWATER SYSTEM LIMITING CONDITION FOR OPERATION two @ emc cacy @ 3.7.1.2 At least -ecee- independent steam generator sur'q':rj l feedwater pumps and associated flow paths shall be OPERABLE with:
- a. One Iwo mo@ tor-driven zuremeyggacy .ej feedwater pumps,-each capable of being
, poweredfromreg:t: emergency bussw , and
- b. One steam turbine-drivenewegfacy zur ::rj feedwater pump capable of being powered from an OPERABLE steam supply system.
APPLICABILITY: MODES 1, 2, and 3. ; ACTION: With one emerur'ytncy bi:rj feedwater pump inoperable, restore em the req a. feedwater pumps to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours.
- b. With two $N' tin [ feedwater pumps inoperable, in at least HOT STANDBY witin 6 hours and in HOT SHUTDOWN within the following 6-hours. }
'litt,
. trru ;;,,i'fr ~ #-- " ' r -'--- '----- r
-- mmediately initiate Icorrective action to restore at'least one I W i": j U feedwater pump to OPERABLE status as soon as possible.or l
SURVEILLANCE REQUIREMENTS eme 4.7.1.2 Each cur *qency ^.:rj feedwater pump shall be denonstrated OPERABLE:
- a. At least once per 31 days by:
- 1. Verifying that each motor driven pump develops a discharge pressure of greater than or equal to ,[qfgypsig at a flow of grl eater than or equal to later gpm.
- 2. Verifying that the steam turbine driven pump develops a discharge pressure of greater than or equal to & psig at a flow of greater than or equal to h gpm when the secondary stea:n supply pressure is greater than hpsig. The provisions of g Specification 4.0.4 are not applicable for entry into MODE 3.
l W-STS 3/4 7-4 AUG 7 1980
PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
- 3. Verifying that each non-automatic valve in the flow path that is not locked, sealed, or otherwise secured in position is in its correct position.
(em ergehn
- 4. Verifying that each automatic valvejin the flow path is in the fully open position whenever the sex'irry feedwater system is placed in automatic control or when above 10% RATED THERMAL POWER.
- b. At least once per 18 months during shutdown by:
- 1. Verifying that each automatic valve in the flow path actuates to its correct position upon receipt of an =x"f rj feedwater actuation test signal. * *M **' Y
- 2. Verifying that each -$$$IN) feedwater pump starts as designed automatically upon receipt of an tur"is y feedwater actuation test signal. cat *3ency
- c. Gt lead once per W -on%s duwtng shutdown , or followiny compleMon of mod [Ssc'edt'on3 to %c eft 4 system s whic4 alfers the systems Slosu charackertsites s 6y v ertCys'ng 4lo w S vo m he Con d ensafe.
Skovo.9e Tanit 4o each skeam g eneva.+or, i l l l W-STS 3/4 7-5 Jul. 3 31980
PLANT SYSTEMS CONDENSATE STORAGE TANK LIMITING CONDITION FOR OPERATION 3.7.1.3 The condensate storage tank (CST) shall be OPERABLE with a contained water volume of at least mo,ooogallons of water. APPLICABILITY: MODES 1, 2, and 3. ACTION: With the condensate storage tank inoperable, within 4 hours eith:r:
'k. Restore the CST to OPERABLE status or be in at least HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours, or i
K 0-. n:tr:t: th: 0."E""."ILI" c' th: (:lternate water :: rce) :: :
- ;,.aug -;;'y +a +ke ?"Mi'f: j f:: trter ;"-as and rettere the
_::nd:n::te ster ;;: tent t: 0"E"^."LE statu; v' th* day: Or be ir :t
';;st "0T ST."."0"'f it.'.in the n;;t S heurs and 4- unT <wi rTnnw within
. ' _ < . ,. , u o,,. c 6.. . . . -
...w . _-
SURVEILLANCE REQUIREMENTS 4.7.1.3.1 The condensate storage tank shall be demonstrated OPERABLE at least once per 12 hours by verifying the contained water volume is within its limits when the tank is the supply source for the auxiliary feedwater pumps.
. , , . . ru. < _ , . - - _ _ _ - _ . _ . _ __.---_s m,, <_ > - . > - --- - -
'To F e 5 e er a ey yvm gW5 4p5 3 3 34 WE u.rw wg y gF W 4W o % s / .a u sua a a ass sa 153 b5 G bCLA V r 4.imDi.L Gb w upw tF W s 5 E h45 5 .3 Uf K 5 hhsss =asw hb[v . I - , r. s vhI the (-i +. m+. te --..me) 9 +w. ..moiy . .. ,.. fe the 2.=<,52 y f:: e:te.
-fMd8Pe.
1 1 W _-STS 3/4 7-6 El 2 0 E
r PLANT SYSTEMS ACTIVITY LIMITING CONDITION FOR OPERATION 3.7.1.4 The specific activity of the secondary coolant system shall be less than or' equal to 0.10 microcuries/ gram DOSE EQUIVALENT I-131. APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: With the specific activity of the secondary coolant system greater than 0.10 microcuries/ gram DOSE EQUIVALENT I-131, be in at least HOT STANDBY within 6 hours and in COLD SHUTDOWN within the following 30 hours. t SURVEILLANCE REQUIREMENTS 4.7.1.4 The specific activity of the secondary coolant system shall be determined to be within the limit by performance of the sampling and analysis program of Table 4.7-1. l l 1 . W-STS 3/4 7-7 JUL 15 379'
t TABLE 4.7-1 SECONDARY COOLANT SYSTEM SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM TYPE OF MEASUREMENT SAMPLE AND ANALYSIS AND ANALYSIS FREQUENCY <
- 1. Gross Activity Determination At least once per 72 hours.
i
- 2. Isotopic Analysis for DOSE a) I per 31 days, when-EQUIVALENT I-131 Concentration ever the gross activity determination indicates iodine concentrations greater than 10% of the allowable limit.
b) 1 per 6 months, when-ever the gross activity determination indicates iodine concentrations below 10% of the allow-able limit. ' l M-STS 3/4 7-8 MAY 151976
l 1 1 I PLANT SYSTEMS MAIN STEAM LINE ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3.7.1.5 Each main steam line isolation valve shall be OPERABLE. APPLICABILITY: MODES 1, 2, and 3. ACTION: MODE 1 - With one main steam line isolation valve inoperable but open, POWER OPERATION may continue provided the inoperable valve is restored to OPERABLE status within 4 hours; otherwise reduce power to less than or equal to 5 percent of. RATED THERMAL POWER within 2 hours. MODES 2 - With one main steam line isolation valve inoperable, subsequent and 3 operation in MODES 2 or 3 may proceed provided:
- a. The isolation valve is maintained closed.
- b. The provisions of Specification 3.0,4 are not applicable.
Otherwise, be in HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours. SURVEILLANCE REQUIREMENTS 4.7.1.5 Each main steam line isolation valve shall be demonstrated OPERABLE by verifying full closure within 5:0 seconds when tested pursuant to Specification 4.0.5. W-STS 3/4 7-9 AUG 6 1981 L
1 PLANT SYSTEMS I 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION LIMITING CONDITION FOR OPERATION 3.7.2 The temperatures of both the primary and secondary coolants in the steam generators shall be greater than (70)*F when the pressure of either ! coolant in the steam-generator is greater than (200) psig. APPLICABILITY: At all times. ACTION: kith the requirements of the above specification not satisfied:
- a. Reduce the steam generator pressure of the applicable side to less than or equal to (200) psig within 30 minutes, and
- b. Perform an engineering evaluation to determine the effect of the overpressurization on the structural integrity of the steam generator.
Determine that the steam generator remains acceptable for continued operation prior to increasing its temperatures above 200*F. t SURVEILLANCI REQUIREMENTS 4.7.2 The pressure in each side of the steam generator shall be determined to be less than 200 psig at least once per hour when the temperature of either the primary or secondary coolant is less than 70*F. I
~
\
hFSTS 3/4 7-10 JUN 1 3 73
~' '
PLANT SYSTEMS s PRIMARY 1 3/4.7.3 ' COMPONENT COOLING WATER SYSTEM
}
LIMITING CONDITION FOR OPERATION s PnNW # 3.7.3 ^t 1:::t two independent' component cooling water loops shall be OPERABLE. APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: 1 With only' k h$mponent cooling water loop OPERABLE, restore at least two loops to OPERABLE status within 72 hours or be in at least HOT STAN0BY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE REQUIREMENTS an-a yi 4.7.3 ^.t 1:::t two' component cooling water loops shall be demonstrated OPERABLE:
- a. At least once per 31 days by verifying that each valve (manual, power operated or automatic) servicing safety related equipment that is not locked, sealed, or otherwise secured in position, is in its correct position.
- b. At least once per 18 months during shutdown, by verifying that each automatic valve servicing safety related equipment actuates to its correct position on a ,,,,,,,,,,, test signal.
t W-STS 3/4 7-11
l PLANT SYSTEMS t 3/4.7.4
- SERVICE WATER SYSTEM
( LIMITING CONDITION FOR OPERATION I 1 3.7.4 At least two independent service water loops shall be OPERABLE. APPLICABILITY: MODES 1, 2, 3 and 4. . ACTION: With only one service water loop OPERA 8LE, restore at least two loops to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE REQUIREMENTS 4.7.4 At least two service water loops shall be demonstrated OPERABLE:
- a. At least once per 31 days by verifying that each valve (manual, power operated or automatic) servicing safety related equipment that t is not locked, sealed, or otherwise secured in position, is in its correct position.
- b. At least once per 18 months during shutdown, by verifying that each automatic valve servicing safety related equipment actuates to its correct position on a pf,jf,u test signal.
l ( ? W-STS 3/4 7-12 MAY 15 576 l.
l h f I 4 (' PLANT SYSTEMS i 3/4.7.5 ULTIMATE HEAT SINK LIMITING CONDITION POR OPERATION 3.7.5 The ultimate heat sink shall be OPERABLE with:
- a. A service water pumphouse water level at or above minus 37'-0" Mean Sea Level, USGS datue, and - fir p:*k en ek- A*1:::i: 0::::
ef :sfficient i:: tr pre ide e f1r r::: cf ;; ::::: th:2 21, ^^0 - -
;? , 22
- b. A mechanical draft cooling tower comprised of two cooling tower fans and a contained basin water volume of equal to or greater than 4 x 106 gallons at an average temperature of less than 900F.
APPLICABILITY: MODES 1, 2, 3 and 4. sc w.a+8,Ac P" U "" '"* # *'." * '"4' ' **'" * ' " r ru'* P bou3e ACTION ** Afk e.o N e ru'us peus +.w n 7 s"'8 hours ** h e m 44 'least k & s+==d.hy w@' Mt4 A e a cyt (, hea . ana is cas.o s wiooewu anWe -N nes.t u hon s .
/ a. "ith ::: creli ; t=:: f r. i..e g ::b!:, ::: tere et 1:::: tr: f:.
te 0?!?^=L"_ ::::e: ri*ki- 72 5:,,s :: 5: in :: 1:::: ;;0T STA"D"Y h eithi :he :::t 6 heer= :=' in CO' n *=,'TO^'.-' Itiil.. the fell :in;; _ 30 herre,
- b. With the mechanical draft cooling tower inoperable, restore the cooling tower to OPERABLE status within hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
I i SURVEILLANCE REQUIREMENTS l l 4.7.5.)The ultimate heat sink shall be demonstrated OPERABLE: l a. 2'i At least once per M hours by verifying the service water h pumphouse water level and the cooling tower basin water level and average temperature to be within their limits,
~
- b. At least once per 31 days by verifying the operation of each cooling tower fan.
. c. At least once per 18 months by verifying the automatic operation of each cooling tower fan in a Tower Actuation test signal.
f 3/4 7-13 . e v.e., . . . . , , . . ~ _ , . , , < - , _ - . - - - - - - _,, . _ , - , - , , _ _ _ _ , , , ,,,,y- ,,,--,-.-.--.-.,s-e-- ,- - , , -
1 I 1 I ( PLANT SYSTEM AIR 3/4.7.6 CONTR L ROOM AHL- MAKEUP SYSTEM LIMITING CONDITION FOR OPERATION 3.7.6 The control room a4e make@up system shall be -OPE 4WrBtt comprised of: l a. Two OPERABLE remote air intakes. l
- b. Two OPERABLE makeup fans and their associated discharge dampers,
- c. An OPERABLE flow path capable of transferring air from the remote air intakes to the control room.
APPLICABILI_TY: ALL MODES ACTION: MODES 1, 2, 3 AND 4 With any control room e makeup N system redundant component inoperable, I restore the inoperable component to OPERABLE status within 7 days or be in ct least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. MODES S AND 6 ta>) 2'itt ;;j ;;; rci iss eir ;;i;;; 27:::: r:9-d a r : p r e n *- i..ey;;;bi;, ;;;;;r: th; ir.;p;cch!; :: ;0r::t :: 0"S".'."LE ;;;t;; eithir ' dry; er i;Iti;;; ;;d .intain s y;;:*ia^ af the centr:E L rc r rir rehe.y .g L. ir the r::ircul; tic: ;;i . g ai, avahable ireslote codeol roo% *aatte.up ai,. wctb;g 3 0 kows OPf ftri- With bee control room 44r makeupV y-t-- cre-rMe, suspend all operations involving N ALTERATIONS or positive reactivity changes. C#8C i ft-) S e pr^"iri::: s f Sp;;ifis. Liv,.. .2.0.2 er; ;;; cppli;;b!; A M00: 5.- I 3/4 7-14 l
{ PLArt SYSTEM SURVEILLANCE REQUIREMENTS 4.7.6 / The control room 44e make ystda shall be demonstrated OPERABLE: 4.E. At least once per 12 hours by verifying that the control room air temperature is less than or equal to 1200F.
- b. At least once per 31 days by verifying operation of each makeup fan. ,
- c. At least once per 18 months by:
- 1. Verifying that each makeup fan flow rate exceeds 1200 cfm with both remote air intakes open.
- 2. Verifying the automatic i-IM k~ i,Q4-- of each makeup fan and its associated discharge damper on a high radiation test signal.
- 3. Verifying that each remote air intake manual isolation valve can be ---^" ' to the closed position.
f- e perated l l S I O 3/4 7-15 1 L_
Not ppb cable .fo Sede,A h NT SYSTEMS ( 3 hw 7. 6 FLOOD PROTECTION (OPTIONAL *) LIMI ING CONDITION FOR OPERATION 1 3.7.6 Floo protection shall be provided for all safety related sys ms, components a structures when the water level of the (usual the ultimate heat ink) exceeds Mean Sea Level USGS datum, at . APPLICABILITY: A all times. ACTION: With the water level at above elevation M n Sea Level USGS datum: a .' (Be in at least H0 STANOBY within 6 hours nd in at least COLD SHUTDOWN within the 11owing 30 hours) d
- b. Initiate and complete w hin h rs, the following flood protection measures:
- 1. (Plant dependent)
- 2. (Plant dependent)
( SURVEILLANCE REQUIREMENTS
, x 4.7.6 The water level at sh 1 be determine to be within the Ifmits by:
- a. Measurement at least nce per 24 hours who the water level is below elevation Me Sea Level USGS datum, d
- b. Measurement at ast once per 2 hours when th water level is equal to or above el vation Mean Sea Level USG datum.
- This speci ication not required if the facility design has adeq te passive flood co rol protection features sufficient to accommodate the sign Basis Flood entified in Regulatory Guide 1.59, August 1973. ,
i l l W-STS 2/' 14 NOV 151977 t w- -
Not Apphcatale 4a seavoo k NLANTSYSTEMS 3 7.7 CONTROL ROOM EMERGENCY AIR CLEANUP SYSTEM LIMITI CONDITION FOR OPERATION 1 / 3.7.7 Two in pendent control room' emergency air cleanup systems shall be OPERABLE. APPLICABILITY: A MODES ACTION: MODES 1, 2, 3 and 4: With one control room emerg cy air cleanup system ino erable, restore the , inoperable system to OPERA 8L status within 7 days or be in at least HOT STAN08Y within the next 6 hour and in COLD SHUTD0 within the following
- 30 hours.
MODES 5 and 6:
- a. With one control room emergen air leanup system inoperable, restore the inoperable system 0 RABLE status within 7 days or initiate and maintain operation ~
the remaining OPERABLE control room emergency air cleanup syste the recirculation mode.
- b. With both control room emerge y air eanup systems inoperable, or with the OPERABLE control ro emergenc air cleanup system, required to be in the recirculation de by ACTIO (a), not capable of being powered by an OPERABLE em gency power sou e, suspend all operations involving CORE ALTERATI0 or positive reac 'vity changes.
SURVEILLANCE REQUIREMENTS
/ \
4.7.7 Each control room rgency air cleanup system shall be emonstrated OPERABLE: -
- a. At least on e per 12 hours by verifying that the control com air temper ture is less than or equal to (120)*F.
- b. At 1.ea once per 31 days on a STAGGERED TEST BASIS by initia 'ng, from control room, flow through the HEPA filters and charco 1 adso ers and verifying that the system operates for at least 10 ours with the heaters on.
l W- S 3/' 7-15 JUL 2 71981
/
Not Ayldable A Seab rook / I PLANT SYSTEMS i
\ SURVEILLANCE REQUIREMENTS (Cnntinued) /
. At least once per 18 months or (1) after any structural mai enance on the HEPA filter or charcoal adsorber housings, or (2) 11owing painting, fire or chemical release in any ventilation zo e nicating with the system by:
i 1. Verifying that with the system operating at a f ow rate of ( m 1 10% and exhausting through the HEPA fil rs and charcoal a rbers, the total bypass flow of the syst to the facility von including leakage through the system iverting valves, is less an or equal to 1% when the system tested by admitting cold D at the system intake. (For sys s with diverting valves.)
- 2. Verifying hat the cleanup system sati ies the in place testing acceptance riteria an.f uses the test rocedures of Regulatory Positions C. a, C.S.c and C.3.d of egulatory Guide 1.52, Revision 2, M ch 1978, and the sy em flow rate is cfm 1 10%.
- 3. Verifying, withi 31 days after emoval, that a laboratory analysis l of a representativ carbon s e obtained in accordance with Regulatory Position .6.b of gulatory Guide 1.52, Revision 2, March 1978, meets the labora cry testing criteria of Regulatory i Position C.6.a of Regu to Guide 1.52, Revision 2, March 1978.
- 4. Verifying a system flow a e of cfm 1 10% during system operation when tested ac rdance with ANSI N510-1975.
- d. After every 720 hours off:harcoal sorber operation by verifying within 31 days after repval, that a laboratory analysis of a represen-tative carbon sample ostained in acco ance with Regulatory Position i
C.6.b of Regulatory ide 1.52, Revisi 2, March 1978, meets the laboratory testing iteria of Regulato Position C.6.a of Regulatory Guide 1.52, Revisi 2, March 1978.
- e. At least once p 18 months by;
- 1. Verifyi that the pressure drop across e combined HEPA filter and charcoal adsorber banks is les than (6) inches Water auge while operating the system at a flow rate of cfm _ 10%.
- 2. V ifyingthatonacontainmentphaseAisokatontestsignal, i e system automatically switches into a recirc ation mode of l operation with flow through the HEPA filters and harcoal adsorber banks.
. Verifying that the system maintains the control room t a positive pressure of greater than or equal to (1/4) inch W.G. r ative .
to the outside atmosphere during system operation. \
- 4. Verifying that the heaters dissipate 1 kw wh tested in accordance with ANSI N510-1975.
y., S
-~ - , -. _ c' ' ' " MAY 15 560 l
i
/
Alof Anh6ble. 6 S&wk 7 TLANT SYSTEMS ,- S ILLANCE REQUIREMENTS (Continued) f. fter each complete or partial replacement of a HEPA filJer bank by v ifying that the HEPA filter banks remove greater than or equal to (99.5)%*oftheDOPwhentheyaretestedinplacein/accordance with SI N510-1975 while operating the system at a flow rate of cfm + 10%.
- g. After each complete or partial replacement of charcoal adsorber bank by ver ying that the charcoal adsorbers emove greater than or equal to 99. of a halogenated hydrocarbon efrigerant test gas when they are sted in place in accordanc with ANSI N510-1975 while operating e system at a flow rate f cfm + 10%.
1 m 99.9 applicable when a filter efficiency of 99% is assumed in the s.a ty anal es; 99% when a filter efficiency of 90% is assumed. l l _-STS , m 7-17 - MAY 151980
l Ald Ay$caWe 4 Seskole
- SYSTEMS 3/4. ECCS Ptw ROOM EXHAUST AIR CLEANUP SYSTEM LIMITING ITION FOR OPERATION s ,
3.7.8 Two i nt ECCS pump room exhaust air cleanup sy ens shall be OPERABLE. '
- APPLICA8ILITY
- MODES , 2, 3 and 4.
ACTION: With one ECCS puep room st air cleanup system noperable, restore the inoperable system to OPERABLE tatus within 7 da or be in at least HOT STAN08Y within the next 6 hours nd in COLD SHU OWN within the following 30 hours. SURVEILLANCE REQUIREMENTS E 4.7.8 Each ECCS pump room exhaust a r clea system shall be demonstrated , OPERA 8LE:
- a. At least once per 31 s on a STAGGE D TEST BASIS by initiating, from the control room flow through the HEPA filters and charcoal adsorbers and verify ng that the system erates for at least 10 hours with the hea rs on.
- b. At least once pe 18 months or (1) after any tructural maintenance on the HEPA fil r or charcoal adsorcer housi s, or (2) following painting, fire or chemical release in any venti tion zone communicati with the system by:
- 1. Veri ng that with the system operating at a. low rate of cfm 10% and exhausting through the HEPA filte s and charcoal ad rs, the total bypass flow of the system t the facility ve , including leakage through the system diverti g valves, is ss than or equal to 1% when the system is test.ed admitting old DOP at the system intake. (For systems with di rting valves.) ,
- 2. Verifying that the cleanup system satisfies the in pla testing acceptance criteria and uses the test procedures of Reg atory Positions C.S.a C.5.c and C.S.d of Regulatory Guide 1.52 Revision 2, March 1978, and the system flow rate is fm
+ 10%.
( N _ STS 2/0 7-16 ~ MAY 151990
. . . _ . - - ~ __ _ _ _ . _ . _ . . _ _ _ . _ _ _ _ _ _ _ . _ . . _ . _ _ _ _ _ _ _ _ _ _ _ _ . _ . _ _ _
Aja+ Apphca 4le le Seakook, LANT SYSTEMS / S ILLANCE REQUIREMENTS (Continued) -
/
- 3. Verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with '
Regulatory Position C.6.b of Regulatory Guide 1.52,. Revision 2, . March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revisforv'2, March 978. .
- 4. V ifying a system flow rate of cfm 1 10% during system op ation when tested in accordance with ANSI'NS10-1975.
/
- c. After eve 720 hours of charcoal adsorber operation by verifying within 31 d s after removal that a laboratory' analysis of a representativ carbon sample obtained in accordance with Regulatory Position C.6.b f Regulatory Guide 1.52, Redision 2, March 1978, meets the labora ory testing criteria of Regulatory Position C.6.a of Regulatory Gui 1.52, Revision 2, March 1978.
- d. At least once per 18 nths by:
- 1. Verifying that th pressure d across the combined HEPA filters and charco adsorber ks of less than (6) inches system at a flow rate of WaterGaugewhileoprating[.t cfm 1 10%.
- 2. Verifying that the syst tarts on a Safety Injection Test Signal.
- 3. Verifying that the filter ooling bypass valves can be manually opened.
- 4. Verifying that the aters d isipate 1 kw when tested in accordanc with ANSL N510-1975.
- e. After each complete partial repla nt.of a HEPA filter bank by verifying that the ftPA filter banks we greater than or equal to (99.95)%* of the DGP when they are tes in place in accordance with ANSI N510-1975 le operating the sys at a flow rate of cfm 1 10%.
- f. After each cpaplete or partial replacement f a charcoal adsorber bank by verffying that the charcoal adsorbe remove greater than or equal 99.95% of a halogenated hydroca n refrigerant test gas when the are tested in place in accordance wi ANSI M510-1975 while
( operat g the system at a flow rate of cf 1 10%. A 99.95% appl [ cable when a filter efficiency of 99% is . assumed n the safety analyses; 9% when a filter efficiency of 90% is assumed. f W STS e, n MAY 1 0 I
,_,,._,v,, ---_y-,,..---,.-.,,-,-g * - -
PLANT SYSTEMS S /4. 7.1 SNUBBERS LIMITING CONDITION FOR OPERATION
- 3. 7.' AllsnubberslistedinTables3.7-faand3.7-bshallbeOPERABLE.
APPLICABILITY: MODES 1, 2, 3, and 4. MODES' S and 6 for snubbers located on systems required OPERABLE in those MODES. ACTION: With one or more . snubbers inoperable, within 72 hours replace or restore the inoperable snubber (s) M OPERABLE status and perform an engineering evaluation i per Specification 4.7. M. on the attached component or declare the attached system inocerabTe and follow the appropriate ACTION statement for that system. SURVEILLANCE REOUIREMENTS 7 0.7.1L Each snubber shall be deconstrated CDERABLE by cer'ornance of the
'ollowing augmented inservice insoection program and the recuirements of Specification 4.0.5.
- a. Insoection Tvoes As used in this specificatien, tyre o' sr,tbber shall tea snu::ers of the same design and manufacturer, irrespective of capacity. '
- b. Visual Insoections Snubbers are categori:ed as inaccessible or ac:essible during reactor i operation and may be treated independently. The accessibility of j each snubber shall be determined and approvec by the (Station Health :
- Physicist) or qualified designee prior to perfo ming each visual '
inspection. The determination shall be based u:en the then existing 4 radiation levels in each snuboer locaticn and tne excetted time to I perform the visual inspection and shall be in a:Cordar.ce witn tne ; i recommendations of Regulatory Guides 8.8 and 8.10. Snubbers access-ible during reactor operation shall be inspected in accordance with the schedule stated below. Snubbers inaccessible curing reactor operation shall be inspected during each reactor shutdown greater than 48 hours unless previously inspected in accordance with the l schedule stated below. The first inservice visual inspection of each type of snubcer shall be performed after 4 months but within 10 months af commencing 3 ^ POWER OPfRATION and shall include all snubbers listed in Tables 3.7-4a and 3.7-N . If less than two snubbers of each type are found i inoperable during the first inservice visual inspection, the second inservice visual inspection shall be performed 12 months 2 25% from the date of the first inspection. Otherwise, subsequent visual inspections shall be performed in accordance with the following schedule:
~
I6 W-STS 3/4 7-20
n- g f,
, .: ;, ~
- _ ANT SY5 E"$ _
l
* . p 3;;VEILL:sCE RECCIREMENT! (Continued)
Nc. Inoperacie Snubcers of each Subsecuent Visual type per Insoection Period Insoection Period *# t 0 ~ 18 months 1 25% r , 1 12 months 2b%. - 2 6 months 2'25%~ / 3,4 - 124 days 2 25% 5,6,7 /~ 62 days 1 25% 8 or more 1/~ , 31 days 1-25%
- c. Refuelino Outage Inspections' !
/
At least once per 18 months,an inpection shgil be performed o't all l the snubbers listed in Tables 3:7 9a and 3. M b attached to sections of safety systems piping that have experienced unexpected, potentially oamaging transients as determined from a review of operationdl data and a visual inspection of the systems. In addition to satisfying the visual inscection acceotance criteria, freedom-of motion of mechanical snucbers shall be verified using at least one ef.tde following: (i) manually inouced snutter movemon ;-(ii);ovalestion of in-place snubber piston setting; or (iii) strexing t6e m'achanical snutoer througn its full range of travel.
./
- d. Visual InsDection Acceptance Criteria
,4 /
*'isual inspections shall verify:
(1) that Ahere are na visible incications of camage or impaired OPEDASILITY and (2)'Sttacreer.ts to the foundation or supporting structure are secure. Snubbers whicn appear inoperable as a result of visual inspections may ce determined
~
CSERAELE for the purpose of establishing the next victal ins:ecticn
'nterval, provicec that: (1) the cause of the rejection is c'atrly established and remedied for that particular snubber and for other snubbers irrespective of type that may be generically suscectible; and (2) the affected snubber is functionally tested in (he as found concition and cetermined OPERABLE per fpecification 4.7.%f. bher. a fluid port of a hydraulic snubcer is found to .be uncovered ne snu:ber sna11 be ceclared inoperaele and snall not be cetermined OPERABLE via functional testing unless the test is started witn tne piston in the as-found setting, extending the piston rod in the tension mode direction. All snubbers connected to an inoperable common hydraulic fluid' reservoir shall be counted as inoperable snubbers.
i e. Functional Tests During the first refueling shusiewn,and at least once per 18 months thereafter during shutdown, a representative sample of snubbers shall "The inspection interval for each type of snubber shall not be lengthened more than one step at a time unless a generic proclem has been identified l and corrected; in that event the inspection interval may be lengthened one l , step the first time and two steps thereafter if no inoperable snutbers of that type are found. The provisions of Specification 4.0.2 are not applicable.
-W-STS 3/4 7-21 17 l
, . _ . . - . . , _ . . . ~ _ ~ . - - _ . . . . _ ._- .-_ -
i i oLANT SYSTEuS l SU '.E:. LANCE REQUIREMENTS (Continued) e. Functional Tests (Continued) l be tested using one of the following sample plans. The sample plan shall bethe during selectcd test period.prior to the test period and cannot be changed The NRC Regional Administrator shall be notified in writing of the sample plan selected prior to the test period implemented: or the sample plan used in the prior test period shall be
- 1) At least 10% of the total of each tyce of snubber in use in the plant shall be functionally tested eitner in place or in a bench test. For each snubber of a type that does not meet the functional test acceptance criteria of Specification a.7.lkf. ,
an additional 10% of that type of snuteer shall be functionally tested of that until ty:e have no more been failures are founc functionally or until tes:ed; or all snubbers
- 2) A representative sampie of eacn type of snubber shall be func-tionally tested in accorcance wit 9 Fi gure 4. 7-1. ";" is tne total numoer of snubbers of a type feua- ;
tance requirements of Specification 4.7)kf. net meetic; the accep-The cumulative number of snubbers of a type tested is cenoted by "N". At the end of each day's testing, tr.e naw va!.ea of "N" anc "C" (pre-vious day's total plus current cay s ' cremer. s) sr.aii ce plotted on Figure 4.7-1. If at any time the point plotted falls in the " tested. functionally Reject" region all snubbers of that type shall be If at any time the point plotte falls in the " Accept" terminated. regicn, testing of sne:bers of tnat type may be , When the point plotted lies in the " Continue Testing" region, additional snubbers of that type shall be l I tested until the peint fails in the "Acce::" region or tne i
" Reject" region, or all the snubcers cf that ty;e have teen tested. }
Testing ecuipment failure during functicnal testing , may invalidate tnat cay's testing and allow tnat cay's testing i to resume anew at a later time, provicing all snubbers tested with the failed equipment during the cay of equipment failure are retested. The representative sample selected for the functional test sample plans shall be randomly selected from the snubbers of each type and reveiwed before beginning the testing. The review shall ensure as far as practital that they are representative of the various configu-rations, operating environments, range of size, and capacity of snubbers of each type. l- Snubbers placed in the same locations as snubbers which failed the previous functional test shall be retested at thethe sampletimeplan. of the next functional test but shall not be included in If during the functional testing, additional sampling is required due to failure of only.one type of snubber, the l functional testing results shall be reviewed at that time to deter-mine if additional samples should be limited to the type of snubber which has failed the functional testing. W-STS 18 3/4 7-32 l l l
t
- LANT SY!'ES SURVEILLANCE REOUIREMENTS (Centinued)
- f. Functional Test Acceptance Criteria The snubber functional test shall verify that:
- 1) Activation (restraining action) is achieved within the specified range in both tension and compression, except that inertia dependent, acceleration limiting mechanical snubbers, may be tested to verify only that activation takes place in both directions of travel;
- 2) Snubber bleed, or release rate where required, is present in '
both tension and compression, within the specified range;
- 3) Where required, the force required to initiate or maintain motion of the snubber is within the specified range in bcth l direction cf travel; ;
5
~
- 4) For snubbers specifically required net to dis; lace under continuous load, the ability of the snuteer to withstand lead without displacement; and y ~
n te.+ To. e'.^.nr :-t of th: ;.;;t:r t: tP: c c m., c a .7,-- ;
@ t h . ^ .1.t ;m am..w aws ai s newie.-
. .;=
Testing methods may be used to measure parameters indirectly or
- arameters other than these specified if these results can e c:rrelated to the speci.fied parameters througn established metneds.
- g. Functional Test Failure Analvsis An engineering evaluation shall be made of each failure to mee . tne functional test acceptance criteria to determine the cause of tne failure. The results of this evaluation shall be used, if a:plicable, i.
in selecting snuboers to be tested in an effert to determine tne ) i OPERAEILITY cf other snuceers irrespective of tyce wnich may be su: Ject to the same failure mode. j For the snubbers found incperable, an engineering evaluation shall i be performed on the components to which the incperable snubbers are t attached. The purpose of this engineering evaluation shall be to determine if the components to which the inoperable snubbers are attached were adversely affected by the inoperability of the snub ers in order to ensure that the component remains capable of meeting the designed service, i If any snubber selected for functional testing either fails to lockup or fails to move, i.e., frozen-in place, the cause will be evaluated and if caused by manufacturer or design deficiency all snubbers of the same type subject to the same defect shall be evaluated in a manner to ensure their OPERABILITY. This testing requirement shall be , Scecification 4.7he. independent for snubbersofnotthemeeting requirements stated 'intest the functional acceptance criteria. w-sTs M 3/4 7-M
-- n - - - ,., s r--
3LsN S*S EwS SURVEILL'NCE DEOUIREMENTS (Continued)
- h. Functional Testinc of Reoaired and Reclacec Snubbers Ii Snubbers which fail the visual inspection or the functional test acceptance criteria shall be repaired or replaced. Replacement snubbers and snubbers which have repairs which might affect the functional test result shall be tested to meet the functional test criteria before installation in the unit. Mechanical snubbers shall have met the acceptance criteria subsequent to their most recent service, and the freedom-of-motion test must have been performed within 12 months before being installed in the unit.
- i. Snubber Seal Reolacement Procram The seal service life of hydraulic snubbers shall be monitored to -
ensure that the service life is not exceecec ce: ween surveillance inspections. Tr.e maximum expectec service life for tne va-icus ; seals, seal materials, and acolicatiens shall be dete-mirec and i estaclished basec on engineering information ano the seals snali ce replacec so that the maximum service life will not ce exceecec during a period when the snubber is required to be OPERABLE. The seal replacements shall be cccumented ar.d the cocumentatier. shall te etained in accc-dance with 5:ecificaticn 6.10.2. A 20 W-STS 3/4 7-24'
TAELE 3.7- a SAFETY-REL*TED HYORAULIC StUEBER5 (Manufacturer) .- SYSTEM" SIZE-(Kips) Small Mecium Large ( ), ( ) ( ) (' ) ( ) ( ) i This mb =4 ion +o loe suppliel . I I at a Inte d=4c. e
. l t.
~
5 : : tai-1 Su. total-2 TOTAL f
"A listing of individual snubbers and more dethiled information shall be available for NRC review at the ( ) facility.
e 9 O H 2.1 W -STS 3/4 7-M
- . _ , - r- - , , , .- - . - - . - - - _ , , y
i TAELE 3.7- b SAFETY-RELATE: MECHANICAL SNUEEERS (Manufacturer) ._ SYSTEM
- SIZE-(Kios)
Small Mecium Large ( )()
, ( ) (' ) ( )( )
S ldormd!Ori to be suppflej af a. later dake.
- I s..
Se: total-1 Subtotal-2 ' TOTAL "A listing of individual snubbers and more detailed information shall be available for NRC review at the ( ) facility. 2.2. V-STS 3/4 7-26. e
t l 10, 9 8 7 - REJECT 6 G f 3 2 / CONTINUE ES M G g /
,b I
2 f*/ I ' 3 ,/ ACCEPT : 0 10 20 30 40 60
. 50 70 80 90 100 1
[ N' . I [. FIGURE 4.7-1 SAMPLE PLAN 2) FOR SNUBBER FUNCTIONAL TEST PSTS 3/4 7.k
--. _ , p. _ , . _ . - - _ - , - - - _ _ . , . , , _ , . - - - - - - - - - -
PLANT SYSTEMS 3/4.7. k SEALED SOURCE CONTAMINATION LIMITING CONDITION FOR OPERATION 8 3.7. h Each sealed source containing radioactive material either in excess of 100 microcuries of beta and/or gamma emitting material or 5 microcuries of alpha emitting material shall be free of greater than or equal to 0.005 microcuries of removable contamination. APPLICABILITY: At all times. ACTION:
- a. With a sealed source having removable contaminstion in excess of the above limits, immediately withdraw the sealed source from use and either:
- 1. Decontaminate and repair the seale.? source, or
- 2. Dispcse of the sealed source in accordance with Commission Regulations.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILL#.NCE REQUIREMENTS F 4.7.M.1 Test Requirements - Each sealed source shall be tested for leakage and/or contamination by:
- a. The licensee, or
- b. Other persons specifically authorized by the Commission or an Agreement State. ,
t The test method shall have a detection sensitivity of at least 0.005 I -}}