ML20129B035
| ML20129B035 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 07/26/1985 |
| From: | George Thomas PUBLIC SERVICE CO. OF NEW HAMPSHIRE |
| To: | Knighton G Office of Nuclear Reactor Regulation |
| References | |
| SBN-846, NUDOCS 8507290039 | |
| Download: ML20129B035 (555) | |
Text
- ._
, George S. Thomas
,)('d 7
F- -- - Vice President-Nuclear Production
\j rutee s.ne. or N wompee.
July 26, 1985 New Hampshire Yankee Division SBN-846 T.F. B7.1.2 United States Nuclear Regulatory Commission Washington, DC 20555 Attention: Mr. George W. Knighton, Chief Licensing Branch No. 3 Division of Licensing
Reference:
(a) Construction Permits CPPR-135 and CPPR-136, Docket Nos. 50-443 and 50-444 (b) Letter from J. DeVincentis (New Hampshire Yankee) to G. W. Knighton (NRC) dated May 29, 1985, "Seabrook Station Technical Specification Improvement Program Subj ect : Technical Specifications for Seabrook Station
Dear Sir:
Enclosed for your review is a copy of the proposed Seabrook Station draft Technical Specifications. This marked-up version supersedes any other version of the proposed Seabrook Technical Specifications previously submitted. As discussed in our April 9, 1985, meeting, the submittal is a copy of the Westinghouse Standard Technical Specifications which has been marked-up to reflect Seabrook-specific design and to implement changes to improve the usefulness of the document. For each change from the Standard, we have provided a justification as indicated by the circled capital letters.
In addition to the justifications provided, we are preparing and will submit by August 16, 1985, a risk based analysis that will further support the technical changes. This will consist of two sections comprising the following: (1) analyses that examine the risk importance of Seabrook systems and, for five systems of high risk importance, evaluates the related Technical Specifications to assure that they are optimized with regard to risk; and (2) an analysis of the Turbine Overspeed Protection System that justifies removing the Specification and reducing the frequency of the surveillance based on risk and cost-benefit evaluation. Each section will provide a quantitative basis for the technical changes that have been made.
8507290039 850726 PDR 0l A ADOCK 05000443 hD PDR l l P.O. Box 300 Seabrook,NH03874 Telephone (603)474-9521 M'
United States Nuclear Regulatory Commission July 26, 1985 Attention: Mr. George W. Knighton Page 2 We understand that this submittal begins your formal review process,in accordance with the schedule provided at our April 9 meeting, and will culminate in a final Seabrook Station Technical Specification document to be issued with the Operating License.
Please address any questions or comments to Mr. Warren Hall at (603) 474-9574, extension 4046.
Very truly yours, j
'Ge'orge S. Thomas GST/cjb cc: ASLB Service List
INDEx
- E~INITIONS l 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 CHANNEL CALIBRATION........................................... 1-1
- 1. 6 CHANNEL CHECK................................................. 1-1
- 1. 7 CONTAINMENT
., ,. m ,- m , - - . ~ ,.--
INTEGRITY.......................................'.. 1-2
.. . ..~.s u m m ............................................ -
1*C
l.1O h CORE ALTCRATION'6idi14C d Win!NtL'dn 1-2 1.M COS E EQIVA LENT I-131. . . . . . . . . . . . . . . . . . . . . . X4tWN41.**Yh;si * *
- I-2.
................... 1-2
- 1. k l-AVERAGE DISINTEGRATION ENERGY.............................. 1-2
'.. k ENGINEERED SAFETY FEATURESTIME..................... RESPONSE 1-3 1.s! FREQUENCY N0TATION........................................... 1-3 1.1T IDENTIFIED LEAKAGE...... .................................... 1-3 1._k MASTER RELAY TEST............................................ 1-3
- 1. b MEMBER (S) 0F THE PUBLIC...................................... 1-3
- 1. N OFFSITE 00SE CALCULATION MANUAL.............................. 1-%3 i
- 1. D. OPERABLE OPERABILITY.......................................
- 1-%9 1.Nt OPERATIONAL MODE -
M0DE...................................... 1 'S,1 1.ld PHYSICS n TESTS................................................ 1-4
- 1. M PRESSURE BOUNDARY LEAKAGE.................................... 1-4
- 1. M PROCESS CONTROL PR0 GRAM......................................1-4
- 1. # PURGE - PURGING.............................................. 1-W
- 1. k QUADRANT POWERRATI0....................................
TILT 1-5
- 1. # RATED THERMAL P0WER.......................................... 1-5 l
- 1. k REACTOR TRIP SYSTEM RESPONSE TIME...'.......'.................. 1-5
- 1. N REPO RTAB LE OCCURRENC E. l",V.WT. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .1-5 e
, courm. .u,min m.ncmsune c n. i 6 . ..wu.r,
+
INTEGRIiY....................................
.. . . ~ . . ~
15
- 1. N SHUTDOWN MARGIN..............................'................ 1-% 5' 1 S SITE ECUNDARY................................................
BL 1.M SLAVE RELAY TEST.............................................
1-6l 1-6 i
r W-STS I h
t .- .
L
INDEX .
DEFINITIONS
- SECTION pAgg
- 1. k SOLIDIFICATION............................................... 1-6
- 1. R S0uRCE CsECx................................................. 1-6
- 1. k STAGGERED TEST BASIS......................................... 1-6 I N THERMAL P0WER................................................ 1-16
- 1. k TRIP ACTUATING DEVICE OPERATIONAL TEST....................... 1-% 6
- 1. N UNIDENTIFIEDLEAKAGE......................................... 1-16 1.N UNRESTRICTED AREA............................................ 1-7 !.
- 1. k VENTILATION EXHAUST TREATMENT SYSTEM......................... 1-7 1.#d VENTING..................................... ................ 1-7
- 1. k WASTE d. N C 3 .E S YS TEM. (f dAWM). . . . . . . . . . . . . . . . . . . . 1-7 J .41 wA5rs rnr4rnear- 59sr-en s.:eut p) 1-1 T A B L E 1."2, F R EQ U E N C Y N 0 TAT I O N . (. . . . . . . . . . . . . . . . . . . . . . . . . . . . .
1-8
, TAELE 1.3l OPERATIONAL MCDES....................................... 1-13 W-STS II
INCEX SAFETY LIMITS AND LIMITING SArETY SYSTEM SETTINGS SECTION FAGE 2.1 SAFETY LIMITS 2.1.1 REACTOR C0RE................................................ 2-1 2.1.2 REACTOR COOLANT SYSTEM PRESSURE............................. 2-1 FIGURE 2.1-1 REACTOR CORE SAFETY LIMIT - FOUR LOOPS IN OPERATION. . 2-2 "ICLT.: 2.12 REACTOR ;;R: CAT:TY LI::IT Tha;; m;C7; I.; 07:TiATIO:: -e 2.2 LIMITING SAFETY SYSTEM SETTINGS 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETP0INTS............... 2-\3 TABLE 2.2-1 REACTOR TRIP SYSTEM INSTRUMENTATION TRID SETPOINTS.... 2-14 f BASES SECTION PAGE 2.1 SAFETY LIMITS 2.1.1 REACTOR C0RE................................................ B 2-1 2.1. 2 REACTOR COO LANT SYSTEM PRESSURE. . . . . . . . . . . . . . . . . . . . . . . . . .B. .2-2
- 2. 2 LIMITING SAFETY SYSTEM SETTINGS 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS............... B 2-3 W-STS III -
INDEX LIMITING CONDITIONS FOR ODERATION AND SURVE!t'.ANCE REOUIREMENTS SECTION PAGE 3/4.0 APPLICABILITY...........................
.................... 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 5 200 F........................... 3/4 1-3 Moderator Temperature Coefficient........................ 3/4 1-4 Minimum Temperature for Criticality........... .......... 3/4 1-6 3/4.1.2 BORATION SYSTEMS Flow Path - Shutdown.....................................
3/4 1-7 Flow Paths - Operating...................................
' 3/4 1-8 .
Charging Pump - Shutdown...................... .......... 3/4 1-9 Charging Pumps - Operating...............................
3/4 1-10 Borated Water Source - Shutdown.......................... 3/4 1-11 Borated Water Sources - Operatine..... ... .. ........... 1/A 1-12 W Beaahatea},,.Soueco - 5 % W uin 3/4.1.3 MOVABLE CONTROL ASSEMBLIES 3N - 19 Group Height.............................................
3/4 1-}415 TABLE 3.1-1 ACCIDENT ANALYSES REQUIRING REEVALUAT!CN IN THE EVENT OF AN INOPERABLE FULL-LENGTH R00................... 3/4 1-1617 Position Indication Systems - Operating.................. 3/4 1-17 f %
1 Position Indication System - Shutdown....................
3/41-1RIT Rod Orop Time............................................ 3/4 1-1410 Shutdown Rod Insertion Limit............................. 3/4 1-24 1(
Control Rod Inserti.on Limits............................. 3/4 1-3( 11 FIGURE 3.1-1 ROD BANK INSERTION LIMITS VERSUS THERMAL POWER
-, , o c , ,_,
FOUR-LOOP 0PERATION...................................... 3/4 1'22.~Li 4 4 w w ri w. a a. onn oiuv m_ eent,mm 6. 4. F6
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.................................. _ . ,# 4,9_a*
W-STS
_ IV D
m memeen--h
INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLUlCE 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 THERMAL PLANAR P0WER...................................... 3/4 2-3 3/4.2.2 -HE?T FLUXRA .'OTDIA t. l'EAkWC-C"ANNCL ."ACT0R. rN.7wt.
. . . . .pxy
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....................... 3/4 2-4 wna ... - en ,.n
.ro..-,,,,..
qsm, or oso, ovnunm .. .r n wo. ..a - uvg g-p.
vc ,r.
m .
A. ,TL 3/4.2.3
-RCS FLOW "AT' AND- NUCLEAR ENTHALPY RISE HOT CHANNEL F A C T O R . - E 4M . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3/4 . . 2 'B[7
.4uuns a.c a nwe .winw uun nn.a vanava n . -
N ^ TION ...................
...... b.................
- r0UR LOOPI- 3/" 2 'id 9
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.- a.. ... u - .. ~ , v
- . . . . - . . , , . ~..
-- e,. y >- ..............
_2. ,f ,
-s 3/4.2.4 QUADRANT POWER TILT RATIO............... ................ 3/4 2-ts7 8 3/4.2.5 DNS PARAMETERS........................................... 3/42-15795l
-E.AE'r '__'-'_
"""."m'a"n'u"~~.m"a"s"........................................
' . < ' , " ." .. . d - i.
"q 3/J.3 INSTRUMENTATIOf1 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATICN...................... 3/4 3-1 TASLE 3.3-1
..e,. , .
REACTOR TRIP SYSTEM INSTRUMENTATION...................
..- --.-.- - . ,,- .m . .... -. --
3/4 3-2 i
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sia cn anain c. ...... r.ao . m ,. . m.. naarut6c ....- .... -,
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TABLE 4.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVE REQUIREMENTS................................ILLANCE ............. 3/4 3-12 3/4.3.2 ENGINEERED SAFETY FEATURES ACTUATICN SYSTEM INSTRUMENTATION........................................ 3/4 3-16 TABLE 3.3-3 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION.......................................... 3/4 3'17Er TABLE 3.3-4 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETP0INTS....'................ ...... 3/4 3-342T
- --TABLE-3r3-5--ENGINmE0 SAFEF FEATURE 4-R5SPONSE TIC . .
3/" 3-3733 --
TABLE 4.3-2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS. . . . . . . . . . . . .3/4 . . . 3-4238; 3/4.3.3 MONITORING INSTRUMENTATION.............................. "
Radiatien Monitoring ME E*. b s................ 3/43-445' y-STS v
I INDEX LIMITING CON 0!TIONS COR OCERATI'N AN? SURVEILLANCE REOU:REMEN~5 SECTION PAGE TABLE 3-3-6 RADIATION MONITORING INSTRUMENTATION TCR PLa"* 0":?,ATICN5..................................... 3/4 3-4G9EI TABLE 4.3-3 RADIATION MONITORING INSTRUMENTATION-TOR FcAnT
.G46"erriefd SURVEILLANCE REQUIREMENTS. . . . . . . . . . . . . . . . . . . . . 3/4 3'5(47 Movable Incore Detectors................................. 3/4 3'5449 Seismic Instrumentation..................................
3/4 3-5Fds?
TABLE 3.3-7 SEISMIC MONITORING INSTRUMENTATION.................... 3/4 3-568
. TABLE 4.3-4 SEISMIC MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS................ ............... ... .... ... 3/4 3-54S4, Meteorological Instrumentation........................... 3/4 3'SL:T3 TAELE 3.3-8 METEOROLOGICAL MONITOR!NG INSTR" MENTATION...... .. . . 3/4 3-hS5Y
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..wii n.,-n-, . .
- - s-,......_- .ca. r..................................
- c w... .......... .
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Remote Snutdown Instrumentation.......................... 3/4 3-ei!3F
, TABLE 3.3-9 REMOTE SHUTDOWN MONITORING INSTRUMENTATION............ 3/4 3-B249$
.,,-. , _e - ~~-- e . x. _ - _ _ n , u ..---...
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... ._ . n...... -
-..wn.no . . . . _. .. .ma -
en.m.....na.w u ---..,_v ig.. - _ ,.e . ..... ................ ........
./ 7 . ==
s-Accident Monitoring Instrumentation...................... 3/4 3-B4d'7 TABLE 3.3-10 ACCIDENT MONITORING INSTRUMENTATION.................. 3/4 3-Bs SP
.,3 . nee,.- u-utyne,,f-ve--o.o- ~-.. . . --
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m..- - - _ _ m .a....w.--m, m....m.-
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,ws . ..
. . .. ....................................... _ .r. _.
wuswe ..m. n. %ww.wo ..__
.j,._... . . ......
J/* J w r-Fire Detection Instrumentation........................... 3/4 3-?&57 TABLE 3.3-11 FIRE DETECTION INSTRUMENTATIQN ......................
s, ssa w t=ps.+ss4.ses z u s. m s k t_. 3/4 3 64 do v u ? -- !L Radioactive Liquid Effluent Monitoring Instrumentation... 3/43-?s4,2-TABLE 3.3-12 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION 3/4 3-14'43 TABLE 4.3-8 RADIOACTIVE LIQUID EFFLUENT MONITORING -
l 4
INSTRUMENTATION SURVEILLANCE REQUIREMENTS................ 3/4 3-7S 5~
Radioactive Gaseous Effluent Monitoring Instrumentation.. 3/4 3- 4F TABLE 3.3-13 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION.......................................... 3/4 3' 49 TABLE 4.3-9 '
RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIF,EMENTS. . . . . . . ..... .. 3/4 3 7Y
-boete4e- t D e t ec t i o n Sy s t em . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . -3/ " 2 ds--
--3/4-3-4-TU R B-I N E-OVE RS P E ED-P R OTE Cf!ON . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .i/4-3-782---
W-STS VI
INCEX LIMITING CONDITIONS FOR OPERATICN AND SURVEILLANCE REOUIPEMENTS SECTION PAGE 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION Startup and Power Operation..............................
3/4 4-1 Hot Standby..................... ........................ 3/4 4-2 Hot Shutdown............................................. 3/4 4-3 Cold Shutdown - Loops Filled............................. 3/4 4-5 Cold Shutdown - Loops Not Filled......................... 3/4 4-6
.. .aw - ...,. ................. ....... - . . _ -
ai ,
I01;ted L;;R Stai;_p. ..................... 2. '
- 3/4.4.2 SAFETY VALVES
~
Shutdown............................................... 3/44'S7 0perating............................................. 3/4 4-157 3/4.4.3 PRESSURIZER.............................................. 3/4 4-119 3/4.4.4 RELIEF VALVES............................................ 3/4 4-1410 e.. ,o;s .....,.
. m auan.va a, . .. . .. .
af , -
-iA5tE 4. 4-1 HINiMuo-RUMBER-0~ STEAM-CEMERATOR5 TO EE in5sECTED '
..m.
- DURING-INSERVIC" INS-PECTION. . . .........................
....m ....-, ,,, . . . - -
-9/4 F1tO-inouw ,.. , ,6 ...,,-.m. --
aiann uuaunniva ivum *aaru. .va.... ..................
s,> -249r' 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE 4
. Leakage Detection Systems................................ 3/4 4 '2C 18' Operational Leakage......................................
3/4 4-dil9
-TAOLC 3.4-1 "EAE-TOR-600 TANT SYSTEM-PRES'StTRE n0cAMON-VAL-V"S. . . .
/4-4-NUh 3/4.4.7 CHEMISTRY................................................ 3/4 4-24R.7L TABLE 3.4-2 REACTOR COO LANT SYSTEM CHEMISTRY LIMITS. . . . . . . . . . . . . . . 3/4 4-2613
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. 3/4.4.8 SPECIFIC ACTIVITY.....................'................... 3/4 4-24 74 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.................................... 3/44'?f2dp TABLE 4.4-4 REACTOR COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS l PR0 GRAM.................................................. 3/4 4-}c2; W-STS VII
. INDEX
.:M:T!NG CdNOITIONS FOR OPERATICN AND SURVEILLMCE REOUIR SECTION PAGE 3/4.4.9 PRESSURE / TEMPERATURE LIMITS Reactor Coolant System................................... 3/4 4-N.23
' FIGURE 3.4-2 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS -
APPLICABLE UP TO /4 ) EFPY..... ......................... 3/4 4 'N.h FIGURE 3.4-3 REACTOR COOLANT SYSTEM COOLDOWN LIMITATIONS -
. , e ,_
APP LICAB LE UP T0 t //o ) EFPY. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 4- M o
,..,e e-.-y mm stHP.,e.r, , ncE.. _#ff.HRE--
. m.m u. .e o.-
- . ~ ~ . . . . e
- e-r , s.m._--... .-
v:T:==L :Cu::UL:.cmWatrspMM9.tc. . . . . . . . . . . . . +/4-4-x3(-
Pressurizer.............................................. 3/4 4-M32.
FlGute 3 4.Ove rpres s ure P rotecti on Sys tems . . . . . . . . . . . . . . . . . . . . . . . . . . 3/44-%3 4 Rcs cos.o ouenPMessuna reort=criosi 3 */ -'I .35*3 3/4.4.10 STRUCTURAL INTEGR *TY. . . . . . . . . . . . . . . .serpristr>
..................... 3 4 4 '5a yg VENTS.............................
3/4.4.11 REACTOR COOLANT SYSTEM 3/4 4-M37 3/c.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS............................................. 3/4 5-1 3/4.5.2 ECCS SUBSYSTEMS - T >
350*F........................... 3/4 5-1 4 3/4.5.3 ECCS SUBSYSTEMS - T < 350 F........................... 3/4 5-\ff
, , , , . . _ o ,, n ,, o . . . . . . . . - . . . . _ - - . .
w / 7. w . T ww..-. anws.bsswa J i e 6 673
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-m - -
wsvia a ng sw w a ws. T. _t.3 ...................................
..n M /.
wi E V .a n w 4 v u_ a -
i rw a w asOL$D F............................................
_ S ,/ *E*^- 4% .I '
I
, 3/4.5.1 REFUELING WATER STORAGE TANK......................:...... 3/4 5 'N. lO W-STS VIII
' INOEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REOUIREENTS SECTION PAGE y - DUAL TYPE CONTAINMENT 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT Containment Integrity....................................
3/4 6-10
- . . . . . , , _ . Containment
'~- - - weepMy -eecra. Leakace......................................
ora. o yrm em-r4 ras- 3/4 6-20
.-a/4-*50--
Containment Air Locks.................................... 3/4 6-60
,__._,____ ,__._ .__ m , _ _ _m em___ . , , .
---wM B # 7i%I6 w . e7 V I ww .w 6 W LA 8 F% % 664. hp83M4 Iw s -'% 3 b.
..__-<___.. ,,ws..u.................................
_ ,! . e.e-__
Internal Pressure........................................ 3/46-(0 Air Temperature..........................................
3/46-$kD Containment Vessel Structural Integrity.................. 3/46-580 Containment Ventilation System........................... 3/46-$bD 3/4.6.2 DEPRESSUR11ATION AND CCOLING SYSTEMS Containment Spray System................................. 3/46dhh Spray Additive System.................................... 3/46-fD
-C:,-t:f----* "A^.; Sy:t-- ..... . . . . . . . . . . . . . . . . . . . . :,4 0 .7;-
.,*.o.a ,.___
r n e--u __e c_. .wr . c,. . _ _ ,,. ,_ , m-5iastn.................................... . , . .. -
3/4.6.'%) CONTAINMENT ISO LATION VALVES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 6-
~~Thes.6- h ' rwTnwnsuT ,nm.+Ttow-vat;ves
--9l%=4 %:.
^
3/4.6 '54 COMBUSTIBLE GAS CONTROL Hydrogen Monitors........................................ 3/46-hfu Electric Hydrogen Recombiners............................ 3/46-2hD i-.;;;r "urg: "le-a"" <ytte-..:...... .................. - /" C 0^
Hydrogen Mixing System................................... 2o 3/4 6-280
.,4. , v, s ,... ,,
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onno ev o . ,,e. . ., - -. -
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a_ u e , o .
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m_ m_o me,,ee o,.m-, , , . , , , -
y-STS Ix-D AUG 6 1981'
L:5'! TING C5.'C: :0N5 FOR OPERATION AC SUR'/EIL>NCE REOU:REME
_SECTION PAGE 3-' . .
3/4.6.h C:00 Of '"' CONTAINMENT racLO54 8.E Sult o suo.
._,,e_,,o__ , , _ , ,
m.. ..._ __,...
Qns , e + s a..u. e. .
., ..m...,. ..... ... .......................
,,..m
_n;;..;euiicing Integrity................................. 21 3/4 6- NO C H 9 d. Building Structural Integrity. . . . . . . . . . . . . . . . . . . . . . 3/46-NO IGetsamed EAdesus e. ]
@ ^ *'=
kksuec. Eme gency E.kam&
Coolm 9 5'.gsfe m 4-&l.2.
l W-sTs
- X-0 pp. 3 0 uts
___________-----_m_-.___._
~
INDEX i
LIITING CCNDITIONS 'O' ODER' TION AND SURVEILL**CE RECUIREF'ENTS SECTION
. PAGE 3/4.7 PLANT SYSTEMS
'3/4.7.1 TURBINE CYCLE Safety Va1ves....................'........................ 3/4 7-1 TABLE 3.7-1 MAXIMUM ALLOWAELE POWER RANGE NEUTRON FLUX HIGH -
SETPOINT WITH INOPERABLE STEAM LINE SAFETY VALVES DURING FOUR LOOP 0PERATION...................................... 3/4 7-2
'T1** Ay__C" _'). "I _ )
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- o. m...
.n. er. - wur..................... _ - ; . ., -
ei, m f=%rsucy . % F e e dwa t e r Sy s t em . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 7-4 Condensate Storage Tank.................................. 3/4 7-6 5 M 'io Activity........................................ 3/4 7-7 TABLE 4.7-1 SECONDARY COOLANT SYSTEM SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PR0 GRAM..................................... 3/4 7-6 ;
Main Steam Line Isolation Va1ves......................... 3/4 7-9
- 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION.......... 3/4 7-10 LPRmast fi -- -
3/4. 7. 3 W.~.PONENT COO LING 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
~ -i .... c a v r.w . .~o......................................... ., , ., , . , . .
3/4.7.% CONTROL ROOM .je%..KpPf .. C JM" AI R C LCAN UP- SYSTEM. . . . . . . . . . . . . . . .
3/4 7-Elt/
.,,.. 3 .. . -... . . . . . . .. . .__. .., .. .., ., , ,o.. . . . , - . . . . ,... , ... . . .-.
3/4.7.0 SNUBBERS................................................. 3/4 7-N /5
- v. s..e. _i e .,. . ,,1. _e x e. ce. v. _ _e e n , r. en o .
. .~. u r.4,.n. om_n n e ,e .e . . . . . . . . . . . . . . . . . . . .
w
.m
.m . . ,,1. _e . e.er. v. _ m,.. .n e , , ,.e n or,.......-..
m.som m ne on.
-.---- m . . . _
~...................
+IGU" 4 7 i S AMFt-E--Pt-AR 2 ) - FOR-SNU3BER-FtlNCYIONAt-TEti. . . . . . . . . . .
3/4+N2f--
. 3/4.7.N S SEALED SOURCE CONTAMINATION......'......:................. 3/4 7 'N#~4-4 WIW .>
h'- STS XI
INDEX LIMITING CONDITIONS FOR OPERATION AND SL'RVEILL*NCE REOUIREMENTS .
SECTION PAGE 3/4.7.149 FIRE SUPPRESSION SYSTEMS Fire Suppression Water 5ystem............................ 3/4 7-79;U/
Sp ray and/or Spri nkl e r Sys tems. . . . . . . . . . . . . . . . . . . . . . . . . . .
mm e. .-_,
3/4 7-?s27
- 2 '- -~********************************************* . ,,. - --
~~ ' -~
"c i ; r Sy r t : :: . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 ^* '
Fire Hose Stations. ' ....................................
.m . ..,, ,M.i -.--. . n c .r.v~o m- s-,,..-.l. 3/4 7'SR;U7
- /,, ,, s, 3b_.
n.4vna.................................... - -
Yard Fire Hydrants and Hydrant Hose Houses.......:....... 3/4 7-4031 w s.r ,J~ v.n.,,
- r. .
-.-- . . - 1
..nc o wRANf5--ANv saavu- 6n.Et-ttYGm. .._ ._ _ _ . . . _ _ -
,, I nua:. nsua:.a. - m4 , s, ., 32;:-
3/4.7.1210 FIRE RATED ASSEMELIES.................................... 3/4 7-44.33 3/4.7.lil) AREA TEMPERATURE MONITORING..............................
3/4 7-L(36F TAC L" 3. 7-I AREA 4EM2ERATURE4&'4-TORINC. . . . . . 3/' .
... ... . . 7-%36, 3/4.8 ELECTRICAL POWER SYSTEMS .
s 3/4.'8.1 ^ 'A.C. SOURCES
- ~ '
Operating................................................ 3/4 8-1 TAELE 4.E-1 DIESEL GENERATOR TEST SCHEDULE........................
^'CL
- .w e ...- -- .._ - . - -
3/4 S'il ' t Lvnw e6sviaw.nu n.6.... . . . ......
e Shutdown..........................
w/,
/'l
...................... 3/4 8-8 3/4.8.2 D.C. SOURCES 0gerating................................................ 3/4 8-9 TABLE 4.8-4 BATTERY SURVEILLANCE RE
. . . . . . . . . Q. U I R EM E N3/4 S h u t d own . . . . . . . . ..............................
3/4 8-11 TS6-12 . . . l. . .
3/4.8.3 ONSITE POWER DISTRIBUTION SYSTE 0perating......................MS ..... .................... 3/4 8-13 Shutdown............................ .................... 3/4 8-15 Trip Chad r.c m,k. .r.1 A 3/ y &-/6 Me
- w. e e
W-STS XII
INDEX i
LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REOUIREMENTS SECTION PAGE 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTION DEVICES A.C. Circuits Inside Primary Containment................. 3/4 8- %.g Containment Penetration Conductor Overcurrent jp Protective Devices..................................... 3/4 6-N
--TABLE 3.8-1 h.cVNI AlNMENT PENETRATION-CON 0ttCTOR-0VERCURREttT -
-PROTEGMVE-DEVICES a ............................,..
Motor-Operated Valves Thermal Overload Protection. Pn.W
-3/4-8-b
.-w wa ,---
3/4 8-M.
r2
-tAGLE 3.0 MOTOR-OPERkTED-VAbVEf-THERMAt-0VERtCAO-PROTEC-MON -
"'O/03 SYPA : DEVitES- a.F.eM48i4~.4.T71.Y ! . W4.7 : . . . . . . E. . . -1/4-8-bp-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 9-4 3/4.9.5 CCMMUNICATI0HS........................................... 3/4 9-5 R 6 F w tl.tAlb. H4CH
." / N.* ?L'LATO,1 CF.":L . ivg.,
3/A.9.6 ...................................... 3/4 9-6 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE POOL BUILDING.......... 3/4 9-7 3/4.9.8 RESIOUAL HEAT REMOVAL AND COOLANT CIRCULATION High Water Level......................................... 3/4 9-8 Low Water Level.......................................... 3/4 9-9 W-STS XIII .
d.9
INDEX i
t!MITINGCbHDITIONSFOROPERATIONANDSURVEILLANCEREOUIREMENTS 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 fact STv84M BWLDM16 MBA4RNCf 3/4.9.12 STCR.'.CE POOL ?.IR CLE " 9 nIA CLEANINC. SYsrB+1 SYSTE".......................... 3/4 9-13 3 ' " .10 SPECIAL TEST EXCEPTIONS
, 3/4.10.1 SHUT 00WN MARGIN.......................................... 3/4 10-1
- 3/4.10.2 GROUP HEIGHT, INSERTION AND POWER DISTRIBUTION LIMITS.... 3/4 10-2 3/'.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 - SHUT 00WN.................... 3/4 10-5 3/4.11 RADIOACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS Concentration............................................ 3/4 11-1 TABtE-4 M1-1 -RADICACTIVE-L-IQUIO-WASTE SAMPtfNO-AND-ANALYS15 PROGRAM.................................................. -3/4--l 2-- -
0ose......................................................
Liquid Radwaste Treatment System......................... 3/4 11-56 Liquid Holdup 3/4 11-5)
Tanks...................................... 3/4 11- M W -STS XIV g na ea *W
INDEX
(
LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.11.2 GASEOUS EFFLUENTS .
t Dose Rate................................................
3/411'80j i
t
-TABLE 4.11-2 -RA&ICACTIVE-GASE005-WAS-TE-SAMPtING-~AND ANAtYSI5 -
i
---PR0 GRAM.................................................. -$f4-11MT/dh t
Dose - Noble Gases....................................... i 3/4 11-12/y !
Cose - Iodine-131, Iodine-133, Tritium, and Radicactive Material in Particulate Form............................. 3/4 11-M L5';
Gaseous Radwaste Treatment System........................ 3/4 11-M I6-Explosive Gas Mixture.................................... 3/4 11- N FIi
... mow.y ....m.v.......................................... .~,, -
.2 Ac i
1 i
t 3/4.11.3 SOLID RADI0 ACTIVE WASTE.................................. 3/4 11-N 17!
3/4.11.4 TOTAL 00SE............................................... 3/4 11-19 i
3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING .I l
3/4. 12. 1 MONITORING PR0 GRAM....................................... 3/4 12-1 j
-T^.SLE 3.12=1==AADIMOGICAC ENVIRONMENTAL--MONFTORING PROGRAM. ...... --3/4-12+3
-4ABL' J.12-2 REFORENG LE'!ELS- r0P RAD-10ACTIVIW CONOE N TIONS
-4N-ENVERONMENTAt-S AMPLESr. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . #4-12 --TAB L E-4r12-1-D ETECTION-G A P ABRI-T-1ES-FO R-E NVIRONMENT A L-S AMPL E AN A LY SrIS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . : . --3/4 10 3/4.12.2 LAND USE CENSUS..................c....................... 3/4 12-13 3/4.12.3 INTERLABORATORY COMPARISON PR0 GRAM....*................... 3/4 12-14 W-STS XV
INDEX l
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 BORATION SYSTEMS.......................................... B 3/4 1-2 3/4.1.3 MOVABLE CONTROL ASSEMBL!ES.. ............................. B 3/4 1-3 s .
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 and RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEt. FACTOR....... B 3/4 2-2 i
FIGURE B 3/4.2-1 TYPICAL INDICATED AXIAL FLUX DIFFERENCE VERSUS l TH E RMA L POWE R . ( Fv n . F.t as.T. .W S.9 ) . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 2-3 3/4.2.4 QU AD RANT P OWE R TI LT RATI 0. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 2 '5.6 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 'R3
-G/h-3.4 itfRSfNE6SMEG-PROT: CT IO N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . u 4/4 4 '!iT.
Ftc.u.ste 6 3l41~1 T 't P h a t. Z ^r o t c A TE.h . A v.i n t. Flux OnffettGArc.C veg.sq g ]l12. -y tnt 4M *t. Powsg (nnsr cou >Jooo nwo/Mrn and Reto40 (aAe_s)
W-STS XVI
. . . mg y e ^ - - -
INDEX i
BASES SECTION PAGE 3/4. 4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION........... B 3/4 4-1 3/4.4.2 SAFETY VALVE 5............................................. B 3/4 4-2 3/4.4.3 PRESSURIZER............................................... B 3/4 4-2 3/4.4.4 RELIEF VALVES............................................. B 3/4 4-3 3/4.4.5 STEMTENERMORS. . .
................................... E 3/4 4 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE............................ B 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 LIMIT 5............................... B 3/4 4-5 g
TABLE B 3/4.4-1 REACTOR VESSEL TOUGHNE55.......................... B.3/4 4-9 FIGURE B 3/4.4-1 FAST NEUTRON FLUENCE FULL POWER SERVICE............................LIFE......(E>1MeV) AS A FUNCTICN B 3/4 4-10 O
_. Auv5w;
- c etu. . _ .3 e e:re r ne c'.". .". .- -- --
-.____-. no. . . . -. -. .- . ..o
- . . " . . -." .. e r -~
ne e7
.---.e eyone_, -- .
---e-
. g iwa nLn64-A ew .... _. .... 4 . ... +
reu, e n , . ,- -
.nm6.aw...
i..___ .
3/4.4.10 STRUCTURAL INTEGRITY..................................... B 3/4 4-?ileI 3/4.4.11 REACTOR COOLANT SYSTEM VENT 5. . . . . . . . . . . . . . . B. .3/4 . . .4-15.UI 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS.............................................. B 3/4 5-1 3/4.5.2 and 3/4.5.3 ECCS SUBSYSTEMS............................... B 3/4 5-1 as ,.a .,. --,nu uun-a n ,--.. m,--u a nu uu s 1&.ana.uu......................
............. - u -i,,--
1 3/4. 5.1L REFUELING WATER STORAGE TANK. . . . . . . . . . . . .B. 3/4 . . . 5-2 W-STS XVII e _ wm em *
'NCEX BASES SECTION PAGE y - DUAL TYPE CONTAINMENT 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT................................ ...... B 3/4 6-ID 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS. .. ................. B 3/4 6-30 C.,, c
. . s A is t MLMOVAL DTheCDJ.................................... ; e, ;' .
3/4. 6.%d5 CONTAINMENT ISO LATION VALVES. . . . . . . . . . . . . . . . . . . . . . . . . B . .3/4
. 6-4D 3/4. 6.'584 COMBUSTIBLE GAS CONTROL. . . . . . . . . . . .. ... ... ............ B 3/4 6-E0
.._...-...Ts."....+
.,,...g F C. t 6 i. - . . . . h . .' e C . . .n . . . ..
. J, - _.
.mu.ma......................................
__. . . . , m.... . . . . . . . _
,. c_.-
3/4.6. -
-::::N: - CONTAINMENT. M .<. @.M 4 ..M 40[ W ............. E 3/4 6-EC
_ y-STS XVIII-0 APR 3 01975
t- 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 i PAm M.Yt 3/4.7.3
- 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........................................ B 3/4 7-3 u.. - ~-...-. .a .................. . ............. ........ a 41 , , ,-
MARELOP 3/4.7.'R4 CONTROL ROOM E.".:.r. :N ;'. A I R C L:N:'.'? S Y ST EM . . . . . . . . . . . . . . . . B 3/4 7-a f
ei ~.n -
-....u, L .e ..
nn-.
n..
-v .. -
6....nw.. .--.
non --.-.a c i d s tre. . . . .
. i rsn a 4 .. .... O J/ 9 /*
3 /4 . 7 .'t 7 S N U B B E R S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 7-5
- -/4.7.3CISEALED SOURCE CONTAMINATION., ~
............................ B 3/4 7'E )
3/4.7.149FIRESUPPRESSIONSYSTEMS.................................. B 3/4 7'E'l 3/4.7.13 EIRE RATED ASSEMBLIES..... ............................... B 3/4 7'RF 3/4.7.1311 AREA TEMPERATURE MONITORING................ .............. B 3/4 721LT 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1, 3/4.8.2 and 3/4.8.3 A.C. SOURCES, D.C. SOURCES AND ONSITE POWER DISTRIBUTION SYSTEMS....................... B 3/4 6-1 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTION DEVICES................... B 3/4 8-3 W-STS XIX BDY. 2 1981
.e . . w M c- n m. , -~mm - ~ = '
l i
I INDEX
/
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 BUI'LDING PENETRATIONS.......... .............. B 3/4 9-1 3/4.9.5 COMMUNICATIONS............................................ B 3/4 9-1 REFRELJ 8J C- M4cHf 4]E 3/4.9.6 "ANIPULATCI CR*ME......................................... 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............. I 3/4 9-2 3/4.9.9 CONTAINMENT PURGE AND EXHAUST ISOLATION SYSTEM............ B 3/4 9-2 3/4.9.10 and 3/4.9.11 WATER LEVEL - REACTOR VESSEL and STORAGE P00L.............................. ............. E 3/4 9-3 FuE -eST0,,A, A, GE. , B..ut -- ,LD. in ,Ett EA.6E.._lvC,y
,/ 4. 9. ,_2 s ,. ,L
, c uru . ,vvm .
u..._,.....
+ . . . .............................. a
. .,/,, . .
a.,
hlR CL.G Ah!ING. sqbTSP1 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN MARGIN........................................... B 3/4 10-1 3/4.10.2 GROUP HEIGHT, INSERTION AND POWER DISTRIBUTION LIMITS..... B 3/4 10-1 3/4.10.3 PHYSICS TESTS............................................. B 3/4 10-1 3/4.10.4 REACTOR COOLANT L00PS..................................... B 3/4 10-1 3/4.10.5 POSITION INDICATION SYSTEM - SHUTD0WN..................... B 3/4 10-1 W-STS XX SEP.15 19 81 b
@Gh.@*ee eggg9 4 me 4mW4 e e g m e mee .we gem , m
INDEX i .
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 RADI0 ACTIVE WASTE................................. B 3/4 11-5 3/4.11.4 TO TA L 0 0 S E . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 11-6 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING V
- 3/4.12.1 MONITORING PR0 GRAM...................................... B 3/4 12-1 3/4.12.2 LAND USE CENSUS. . . . . . . . . . . . ................. .......... B 3/4 12-2 3/4.12.3 INTERLABGRATORY COMPARISON P R0 GRAM. . . . . . . . . . . . . . . . . . . . . . B 3/4 12-2
_W-STS XXI
= m__ _
INCEX DESIGN FEATURES SECTION PAGE 5.1 SITE 5.1.1 EXCLUSION AREA.............................................. 5-1 5.1.2 LOW POPULATION ZONE.......... ..............................
5-1 5.1.3 .hQ. .e, ..nQQ. u^T'o.?.%.i . ,
0 .e%.
.u.n.a.NS r r-T. m m W.,Me m._ S AflUk.
en_ o ean n.W,a.ch. Tr_*
u.u_e.e_ m_ .t r
i, uu --
a . rreineu-r
- .T ww
-o e r n. m .
g* q SLTE. grount O R.'tmvens.r........................................
rott. L.lCRU.L O EFFLu2N1~5 ,e, _
FIGURE 5.1-1 y:. g EXC'.L':IO" AR:A. .s s ret .eep. Amins.oN.oM4. spu.vsrs.tt . .
5-2 FIGURE 5.1-2 LOW POPULATION 20NE..................................
LIQwtD ,E,m,,f.l.trern~ f Di$thtA%E Lot.A 5-3 F1 UP.e. S.1_
. .e .., y r. w -.~mu. ~ . u- unu.w
..r~r u ~Tip p
n o.................. -
.-v--
.. .... m ev.r.
--. e. n,.~.n.s e v e n. _c.s
. . n , v m . -. --
- -- io oe
_o .<............. ..... .
5.2 CONTAINMENT 5.2.1 CONFIGURATION................................... .......... 5-1 5.2.2 DESIGN PRESSURE AND TEMPERATURE................. .......... 5-11 .
5.3 REACTOR CORE 5.3.1 FUEL ASSEMSLIES. .. .. .. . . ..... . .... ....... . ....... 5-15' 5.3.2 CONTROL ROD ASSEMBLIES...... ................. ........... 5-1 7.
- 5. 4.1 DESIGN PRESSURE AND TEMPERATURE....................... ..... 5-%.5' 5.4.2 V0LUME................................................ ..... 5-%S' 5.5 METEOROLOGICAL TOWER L0 CATION................................. 5-LS-
- 5. 6 FUEL STORAGE 5.6.1 CRITICALITY.......................................... ...... 5-16 5.6.2 DRAINAGE.................................................... 5-%l, 5.6.3 CAPACITY.................................................... 5-M 5.7 COMPONENT CYCLIC OR TRANSIENT LIMIT........................... 5-45 TABLE 5.7-1 COMPONENT CYCLIC CR TRANSIENT LIMITS............ .....
5-17 W-STS XXII
It40EX
(
t'_"It:IST UTIVE CCNTROLS SECTION PAGE
__ 6.1 RESPONSIBILITY................... ....... ................. 6-1
- 6. 2 ORGANIZATION;...............................................
6-1 6.2.1 0FFSITE................................................... 6-1 m 780A#
6.2.2 Gi 7- S T A F F . . . . . . . . . . . . .~. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-1 FIGURE 6.2-1 '0FFSITE ORGANIZATION...... ........................ .6-3 FIGURE 6.2-2 STATION ORGANIZATION....... ..... ... ............ E-4.
TABLE 6.2-1 MINIMUM SHIFT CREO COMP'SITIOS................
0 ..... 6-5 a
.o. n.--- o
- w. .. .n..,_.. m.i. ~.
o m .--.v r --
.- e_ m_ .t e--u..N r-a-on h' vl aw
^- -
b _ b (~~~'
"F-UHLL50 P~. ..............................................
. 64 --
.e. :-~- 3
-E.K ,,
p o :r s vt u u . . . . ,
..........................a........ ......
M C..-iDil'Me W ............ ....... ................... i'7 M_ Th. ,e tf',- ") s
[ . r, - - .- --../................................................ -
- 6. 2. ; SHIFT TECHNICAL ADVISOR....... ..............:........
" . STAFF \0VALIkICATINf..................................
V V V h 6 4 TRO NIhG . .
v v k' V.p9EVFW ND)I6DI:..
v o.................................... W v v n
s pgaAmon R.sureuRg rTTSE N IC,,.rpriorpi
_,,,,,-,,---..P............. ............................
...,--u.- -
V
'F).Rc}n ................................. ..............
tc%s to . ............................................. '6- r
[A Q Cpa y ............................................. f-pigFesnc ........................... ............ 6-7
, j u & Q . ,..............................................
6-5. .
1 Res i As %pon&biliies...........
./
6-3i 4/w f.eCofds. .......... v........... ....... .. ... ............
}-%Io
/
W-STS
_ XXIII u-
INDEX
( .
t .
- 0MINISTRATIVE CONTRCLS
'SECTION
- n n
/ png,.c9. - 9 FEMA $1T,MO
~ , . . .
R2V G
. j,vo. .g ovv HtVity mu n...-. vg# j_ N.., contV...............
<tYM 8 ]unw w./.... -gg n 6 4 d'nhtfo nb . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .- . /o ....
Com b itTbne, ............. ... ...... ....... .
........ 5-X)ll ytepte%,.................. ..
......................... 6,M1(
Conettantso ............... ............................ -x a eeti @ Ereque
..................................... . 6-M Il gu6 tup.e..... ............................................ e- v il Reyfhw
- .................................................. 6-M 12.
udits..
As1kee . .... ............................ .............. 6-M lt.
toYd .............. ....... (optf g
evervr
- 6. 6 REPORTAB LE OC:'.;ME'::E ACTION. . . . . . . . . . . . . . . . . . . .
/
........... E-likly
- 6. 7 S AFETY lit'IT VIOLATION. . . . . . ....
.. .... ...... .. ...... 5 'S /4 6.8 PROCEDURES AND PR0 GRAMS....................... .......... .. 5 '1S. 14
- 6. 9 REPCRTING REQUIREMENTS............. . ..................... 6-K l ]
6.9.1 ROUTINE REPORTS AND 8,EPoKTA01.E E U Earr$ .............. 6-% ll Startup Report.............................. .. . ..... 5- 5 11 Annual Reports............................................ 6 'M G Annual Radiological Environmental Coerating Report........ 6-1$ 0 Semiannual Radioactive Effluent Release Report............ 6-2S 2.1 Monthly Operating Report.................................. 1 n____m, 6-M W
..on.cs....................................
. , . . . . - n. __ _ ,r , . , .
s 1' . ...g L im i i m o m . .
- 0 "- % n F n 1 ' n ;; . . . . . . . . . . . . . ,. . . . M
.<. m_. , , _ _
> . ,n... o., . .7 . . uun nuev..6a................................ f-EC-- -
Radial Peaking Factor Limit Report........................ 6-N 19 6.9.2 SPECIAL REP 0RTS.................................... 2 6'S43 6.10 RECORD RETENTION................................ .. ....... 6-7423 W-STS XXIV
. - ' . . e e e .g> e.m.e ,e e e.e ew _
e=
= * * * " * * -
- 9L - -
INDEX i
AD!NISTRATIVE CONTROLS -
SECTION ,
6.11 RADIATION PROTECTION PR0 GRAM............................... 6-h523] ~
6.12 HIGH RADIATION AREA-(0ptier,al)............................. 6-25 6.13 PROCESS CONTROL PR0 GRAM.................................... 6-26 6.14 0FFSITE DOSE CALCULATION MANUAL.......... ................. 6-2%2d 6.15 MAJOR CHANGES TO RADIOACTIVE LIOUID. GASEGUS. AND SOLID WASTE TREATMENT............................................ 6-ig)7
/
i I
l l
-W-STS XXV .
l L_.
f SECTION 1.0 DEFINITIONS t
1 f
t l
l s
JUSTIFICATION
(
Section 1.0 1
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. This definition added to describe the testing of actual digital signals.
B. Clarifies types of waste systems and streams as well as specific site receiving Seabrook waste.
C. 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.
D. Seabrook Station plant specific information.
E. Required per Generic letter 83-43 F. Site Boundary - Additional clarification in order for the definition to apply to the gaseous effluent dose calculations.
G. Unrestricted Area - A specific definition was needed to address liquid I
effluent dose calculations.
k
i 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. l ACTUATION LOGIC TEST
- 1. 2 An ACTUATION LOGIC TEST shall be the application of various simulated input combinations in conjunction with each possible interlock logic state and verification of the required logic output. The ACTUATION LOGIC TEST shall
. include a continuity check, as a minimum, of output devices.
ANALOG CHANNEL OPERATIONAL TEST 1.3 An ANALOG CHANNEL OPERATIONAL TEST shall be the injection of a simulated signal into the channel as close to the senser as practicable to verify '
OPERASILITY of alarm, interlock and/or trip functions. Tne ANALOG CHANNEL OPERATIONAL TEST shall incluce adjustments, as necessary, of the alarm, inter-
, lock and/or Trip Setpoints such that the Setpoints are within the required range and accuracy.
AXIAL FLUX DIFFERENCE 1.4 AXIAL FLUX DIFFERENCE shall be the difference in' normalized flux signals between the top and bottom halves of a two section excore neutron detector.
CHANNEL CALIBRATION -
1.5 A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel such that it responds with the required range and accuracy to known values of input. The CHANNEL CALIBRATION shall encompass the entire channel 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 .
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 independent instrument channels measuring the same parameter.
E-STS 1-1
t CErINITIONS
/
CONTAINMENT INTEGRITY
- 1. 7 CONTAINMENT INJEGRITY 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-\[ofSpecification(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 cf Specification (3.6.1.2), and
- e. The sealing mechanism associated with each penetration (e.g., welds, beilows, or 0-rings) is OPERABLE.
( --CONTROLL"O LEAKAGE-
- 1. 0 CnNTRn' LEG--LEAKAGE--shal' be thet--seml watar #1cs :ue M tc Me en - re
--ceMar.t pump = seel a .
CORE ALTERATION
- 1. 9 CORE ALTERATION shall be the movement or manipulation of any component within the reactor pressure vessel with the vessel head remov(d and fuel in the vessel. Suspension of CORE ALTERATION shall not preclude completion of movement of a component to a safe conservative position.
See Insert 3' DOSE EOUIVALENT I-131 h
- 1. k DOSE EQUIVALENT I-131 shall be that concentration of I-131 (micrcC 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 thyroid
- dose conversion factors used for this calculation shall be those listed in NRC
-(T ele III cf TID 'iS, "Calcul: tier er nistanca nr+nre fnr onwar ana h
.Reeeter-S+tes" or T:bic E-7 ef P C Reguisterj Cuide 1.109, Revisic. 1, ,
-Oct26i 13777. Rcylden Gashi t% "calculahde & Annual kse.s k% .r.o ,2cuhne Releases a Re& E#lueds/'
E - AVERAGE DISINTEGRATION ENERGY
- 12 l.K, l shall be the average (weighted in proportion to the concentration of each radionuclide in the sample) of the sum of the ayerage beta and gamma energies per disintegration (MeV/d) for the radionuclides in the sample.
W-STS 1-2
t INSERT I DIGITAL CHANNEL OPERATIONAL TEST 1.10 A DIGITAL CHANNEL OPERATIONAL TEST shall consist of exercising the digital computer hardware using database manipulation and injecting simulated process data to verify OPERABILITY of alarm and/or trip functions.
a J
l f
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t f
r.
1 t
l I
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L- .. . _ _ . -
DEFINITIONS t-i ENGINEERED. SAFETY FEATURES RESPONSE TIME l3
- 1. 3C The ENGINEERED SAFETY FEATURES RESPONSE TIME shall be that ti from when the monitored parameter exceeds its ESF Actuation 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' 14 1.14 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1/3s 2.
IDENTIFIED LEAKAGE ISI 1.14 IDENTIFIED LEAKAGE shall be:
a.
Leakage (except CONTROLLED LEAKAGE) into closed systems, such as pumo seal or valve collecting packing tank, or leaks that are captured and conducted to a sump or b.
- Leakage into the containment atmosc'here from sourcer that a e bcth
- sDecifically located and known either not to interfere vi th ina ;eratien cf Leakage Detection Systems or not to be PRESSURE BOUNDARY LEAKAGE, or i c.
ReactorSystem.
Coolant Coolant System leakage through a steam generator to the Secondary MASTER RELAY TEST 1.hAMASTERRELAYTESTshallbetheenergi:ationofeachmasterrela 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.
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-
. ational, occupational, or other purposes not associated with the plant.
OFFSITE DOSE CALCULATION MANUAL 18 1.17 The DFFSITE 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 liquid ef fluent monitoring Alarm / Trip Setpoints, and in the conduct of the Environ-mental Radiological Monitoring Program.
W-STS 1-3
CEFINITIONS
(.
OPERAELE - OPERABILITY 1.
A system, subsystem, train, compenert or cevice shall be OPERABLE 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 capable of performing their related support function (s).
OPERATIONAL MODE - MODE 2.0 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.
I PHYSICS TESTS Al
- 1. bq PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation:
(1) described in Chapter (14.0) of the FSAR, (2) authorized under the provisions of 10 CFR 50.59, or (3) otherwise approved by the Commission.
PRESSURE BOUNDARY LEAKAGE 2.2.
1.24 PRESSURE BOUNDARY LEAKAGE shall be leakage (except steam generator tube
, f leakage) body, throuch pipe a nonisolable wall, or vessel wall. fault in a Reactor Coolant System component PROCESS CONTROL PROGRAM se e J g e,.y 31-he PROCESS CONTROL PROGRAM (PCP) shall contain the provisions to
' re that the properties that FICATION of wet radioactive the. requirements of 10 wastes results in a wae- orm with CFR Part 61 an radioactive waste dis - ow-level 1 sites. The PCP shall iden
- j process parameters influencing SOLIDIFICATI0h as pH, oil cent tent, ratio of SCL.'0IFICATION a , H7 0 content, solids con-
. +o wa=* nd/or necessary additives for each type of anticipated waste, and atable boundary conditions for the process parameters shall be i- . Tied for ea . ste type, based on labora-tory scale and full-sca sting or experience.
an identification P shall also include onditions that must be satisfied, on full-scale testing, to re that dewatering of bead resins, pcwdered r slud a s, and filter 1 result in volumes of free water, at the time of dispo within imits of 10 CFR Part 61 and of low-level radioactive waste disposai **s.
PURGE - PURGING
- 24 1.24 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 hscyt E Y
\) l(uauiu couteurses s.ia nesia swanig]
PROCESS CONTROL PROGRAM as i n m 7--..edt, iss+.tieJ soupiFac4 riou sySIeef l St2. 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 di__sposal sites 3 The PCP shall identify process parameters influencing SOLIDIFICATION such as pH, oil content, H 2 O content, solids content, ratio of solidification agent to waste and/or necessary additive 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 a-Q/o* -eM. full scale testing or experience. The PCP she I e1.0 ...clude en identi
. -catiar Of conditions that .u t bc stisfied, based- on fuli scele testing, tc as;ure that dcwai.ering cf bead racinc, perdered retir. , end filter ;1udges will-recnit in volomac M f"ee water, at the ti?.e of di;posal, within--the limits cf 10 CP Part $1 end of ic.clevel radicactive e:Ste dispos;! ;ites.
~
For a mobile sos.no tfscarton or pengs,es.ni.s sgrion geyvGe to be gasej n'n I,ci,
- 4 % chx. hon p<rman edly n\skliel CoLIDificArtod sysfem tke selecled i vessov must have an acceptabl< ToptEl PCP hv wluck he an demonshale.
thf Pane & condthbns a4 10 CPR Psvt- 4l nd -the klas}c Dnsposal S ts ce net l__ .
ha klbto k O c 6ed>Odl Sta.fth sh,k coasfe, . ]
O
?
- E
- :'.': IONS i -
UAD" ANT COWER TILT Ft.TIO AS-
- 1. M QUADP. ANT POWEe TILT RATIO shall be the ratio of the maximum uceer excere cetector calibrated output tc the average of the upper excere detecter cali-brated outputs, or tne ratio of the maximum lower excore detector chiitrated outpet to the average of the lower ercore detector calibrated outputs, whichever is greater. With one excore detector inoperable, the remaining three detectors shall be used for computing the average.
RATED THEPMA' POWER 2.6 1.M RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 34h Kdt. g REACTOR TRIP SYSTEM RESPONSE TIME
- 1. The REACTOR TRIP. SYSTEM RESFCNSE TIME shall be the time interval from hen tne cenitored parameter exceeds its Tric Setroint at the enannel sensor
..ntil loss of station'ary gripoer coil voltage.
7, e .3 r g . : p. . : _. E.V
... - .. E
- alt-
.28 EVENT l'. N A RE704 AE E ~'-"~~~.:: S.a bi 7.ny :f tnese c:.nciti:ns speci'ied in i;'c#: n n. C.C.1.1: : u 5. E.1. :1T Sec+ ton 50.~l3 -fo lo CFR Part 50 CONTslannu1~ EucLOSURE
!"' '_: :'J: _::.' : !NTEGRI '
coursonnem- SWcLOSURE 1.M -S":E'M FU:'_::5:: :NTE:-RITY shall axis . when:
'e .' 2SEEch door in each access opening is cicsed except when the access opening is being used for normal traasit er.try !.nc exit, then at leas. cne coor snail be closed, codain.nent Eulowee.
y b. The !"4- "i v : Filtration Systec is OPERAELE, and
- c. The sealing mec5nnism associated with each penetration (e.c. , ulcs, bellows, er 0-rings) is GPERA3t.E.
StiUTDOWN MARGIN 30
- 1. M SHUTDOWN MARCIN shall be tne instantaneous amount cf reactivity by wht:n
. the reactor is scbcritical o" would ce sutcritical from its present condition assuming all fuli-lenct- r:,d cluste assemblies (shutdown and control'; are fully inserted except for the single rod cluster assembly of highest reactivity
_ worth whicn is assumed to bc- fully withdrawn.
SITI B00NDA?
- l. The SITE BCUNDARY she'l De that iine beyond which the land is neitner osta. uhNh 'the owned, site. boundary ner leased, usel foi. norrecte.nhonal etnerwi,se controlied by NeMacas purposes big licersee. a.whE.f48MC.
sp slialt bc
- ,, cia..a. % a w,o,a o em sowe. co n, Poses ,uca se m (R) dose. specWc4hlutr. (Reahdd occ.upWy 4 rs shall be supplied ai- 17<,loca+2dnr W-STS 6.b p-poses of Jose.calculato%-5 g __ . . . _ .
- DEFINITIONS SLAVE RELAY TEST
- l. A SLAVE RELAY TEST shall be the energization of each slave relay and verification of OPERA 8ILITY of each relay. The SLAVE RELAY TEST shall include a continuity check, as a minimum, of associated testable actuation devices.
SOLIDIFICATION -
- 1. SOLIDIFICATION shall be the conversion of wet wastes into a form that meets shipping and burial ground requirements. ,,
SOURCE CHECK .
- 1. A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a source of increased radioactivity.
STAGGERED TEST BASIS
- 1. NSTAGGEREDTESTBASISshallconsistof:
- a. A test schedule .for n systems, subsystems, trains, or other designated components obtained by dividing the specified test interval into n equal subintervals, and [
( b. The testing of one system, subsystem, train, or other designated component at the beginning of each subinterval.
THERMAL POWER
- 1. THERMAL. POWER shall be the total reactor core heat transfer rate to the reactor coolant.
TRIP ACTUATING DEVICE OPERATIONAL TEST
- 1. A TRIP ACTUATING DEVICE OPERATIONAL TEST shall consist 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.
l UNIDENTIFIED LEAKAGE l 1. UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED LEAKAGE j or CONTROLLED LEAKAGE.
l l
W-STS
- 1-6
,-n-y_ _-_
,.,m ,,. , ._ , . -,, ,.,, _ ,.m,_ ,,,,_,,,,,r -
f DE :NITIONS UNRESTRICTED AREA ge e I g e,y+ g a
NRESTRICTED AREA shall be any area at or beyond the SITE BOUNDARV access to w ,
+
trolled by the licensee for pur e^- .. r %ectio'n of individuals from exposure to ~
d ~
materials, or any area within the SI
residentian r for industrial, a ,
institutional, and/or recreational purposes.
VENTILATION EXHAUST TREATMENT SYSTEM
- 1. A VENTILATION EXHAUST TREATMENT SYSTEM shall be any designed and installed to reduce gaseous radioiodine or radioactive material in particulate form in effluents by passing ventilation or vent exhaust gases through charccal 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 not considerec to be VENTILATION EXHAUST TREATMENT SYSTEM components.
VENTING
- 1. k VENTING shall be the controlled process of discharging air or gas from confinement to maintain temperature, pressure, humidity, c:ncentrati:r or other cperatir.; ccndition, in such a manner that replacement air er gas is -- . ; ro-vided or required during VENTING. Vent, used in system names, does not imply a VENTING process.
TREAT M EA/1-WASTE C TL"L? SYSTEM (Gaseous) q1.t caseous f%OWAsre. TRenThEUG
- 1. M A "* " ^*" "^'~ " SYSTEM shall be any system designed and installed to reduce radioactive gaseous effluents by collecting Reactor Coolant Systems offgases from the Reactor Coolant System and providing for delay or holdup i for the purpose of reducing the total radioactivity prior to release to the {
environment, i WA ST'E TRE ATHEA/T' SYSTEFt (L.ipid.)
1k '
A lisiAi D RftDu1ASTE TREATAfEA/T SYS7FM dAll be aq comtYuwkion 64 cormportod> hsi 9ned an& ins 4allel to vesace rdtoac.five mater $al5 in 1%utd eMureth by passInq
~ .
process, e.Mluecf strea,m3 throg any one of several possible e*PoredWS dnd/or Je=tneahjer_s prior 40 release Yo %e endtecm-e,&
(
W -STS 1-7
_. . . . - . . _ . _ . . . . _ . . . . _ . . , _ . . _ . . _ _ _ _ _ . ~ _ -
INSERT III UNRESTRICTED AREA (for liquid effluents only)
~
1.39 .The UNRESTRICTED AREA for liquid effluents, for the dose calculations, shall be defined at or beyond the surface edge of the- initial mixing zone'where effluents from the submerged multiport diffuser discharge have undergone prompt dilution (assumed to be 10:1).
(
L
B TABLE 1.1
(
, f OPERATIONAL MODES i
REACTIVITY % RATED MODE AVERAGE COOLANT CONDITION, K THERMAL POWER
- df TEMPERATURE
- 1. POWER OPERATION 10.99 > 5% 1(350*F)
- 2. STARTUP 10.99 { 5% 1(350*F)
- 3. HOT STANDBY ( 0.99 0 1(350F)
- 4. HOT SHUTDOWN ( 0.99 0 (350 F) > T avg
> (200*F)
- 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.
TABLE 1.2 -
(*
(
FREQUENCY NOTATION
, NOTATION , FREQUENCY S
At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
D At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
W At least once per 7 days.
M At least once per 31 days.
Q At least 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
r 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 Seabrook.
B. Seabrook Station specific plant data.
C. Westinghouse terminology.
D. Not applicable to Seabrook.
E. Provides clarification to aid in understanding of Reactor Trip Setpoint limits of Section 2.2.1.
F. The I hour time limit to be in HOT STANDBY increases the likelihood of human error causing an unnecessary plant trip or some other state of less safe plant conditions. Investigation and discussion has found that when operating at' full power, the minimum time to reach HOT STANDBY is about 50 to 55 minutes. Forcing a MODE change that takes almost the entire allowable time places undue pressure on the operator and increases the probability of a potential error creating more unnecessary problems.
/
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 shall not exceed the limits shown in g FigureX 2.1-17 =d 2.1-2 fcr rd vi"Y9;)p Operation, rc pect?'/ely_fo,. y /,,7 operah, APPLICABILITY: 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 STANDBY within1 hour $, and comply with the require-ments of Specification 6.7.1. P REACTOR COOLANT SYSTEM PRESSURE 2.1.2 The Reactor Coolant System pressure shall not exceed 2735 psig.
(,
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 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, 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, j 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 1
680 I -
I
(-
l UNACCEPTABLE l OPERATION l'
- 660 --
._.n_____
! i I400 pS,4 640 #0 Psts l
620 SI4 19 e
,. 6o ps;
- i j
- u. i d '
l g 600 .
sn d ~
f-1 580 _
~
ACCEPTABLE OPERATION 560 1 !
~
l i
j 540 ;
t
~
l l
520 ' '
O.0 0.20 0.40 0.60 0.80 1.00 1.20 l (FRACTION OF RATED THERMAL POWER) l l
' \
' i i
l li~ FIGURE 2.1-1 REACTOR COOLANT SAFETY LIMIT l FOUR LOOPS IN OPERATION l 2-2 i i I $
t
SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS
- 2. 2 LIMITING SAFETY SYSTEM SETTINGS
~ REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS 2.2.1 The Reactor Trip System Instrumentation and Interlock 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 but more conservative than the value shown in the Allowable Value column of Table 2.2-1, adjust the Setpoint consistent with the Trip Setpoint value. .
- b. With the Reactor Trip System Instrumentation or Interlock Setpoint less conservative than the value shown in the Allowable Values column 'of Table 2.2-1, either:
- 1. Adjust the Setpoint consistent with the Trip Setpoint value of Table 2.2-1 and determine within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> that Equation 2.2-I was satisfied for the affected chanr.el, or
- 2. Declare the channel inoperable and apply the applicable ACTION statement requirement of Specification 3.3.1 until the channel is restored to OPERABLE status with its Setpoint adjusted consistent with the Trip Setpoint value.
Equation 2.2-1 , Z + R + S < TA Where:
Z = The value frem Column Z of Table 2.2-1 for the affected channel, R = The "as measured" value (ih 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 from Column S (Sensor Error) of Table 2.2-1 for the affected channel, and t . m */, +f s en s TA=ThevaluefromColumnTA(TotalAllowanchhofTable2.2-1for g
the affected channel.
I 3
W-STS 2-\ ,
,-~
TA8LE 2.2-1 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS ()
Total Sensor Functional Unit Allowance (TA) Z Drift (S) Trip Setpoint Allowable Value
- 1. Manual Reactor Trip NA NA NA NA NA
- 2. Power Range, Neutron Flux, 7.5 4.56 0 $ 109% of RTP High Setpoint $ 111.1% of RTP Low Setpoint 8.3 4.56 0 5 25%.of RTP $ 27.1% of RTP
- 3. Power Range, Neutron Flux, 1.6 0.5 0 5 5% of RTP with a time High Positive Rate 5 6.3% of RTP with a time constant 2 2 seconds constant 2 2 seconds y 4. Power Range, Neutron Flux, 1.6 0.5 0 5 5% of RTP with a time High Negative Rate 1 6.3% of RTP with a time t
i constant > 2 seconds constant > 2 seconds
- 6. Source Range, Neutron Flux 17.0 10.01 0 5 105 cps 51.6 x 105 cps 3 73 I od r o To
- 7. Overtemperature AT 6.5 4h"NE 1.3S:0.47# See note 1 See note 2 1 92 o.12
- 8. Overpower AT 4.8 JLs0k 4}pHE See Note 3 See Note 4
- 9. Pressurizer Pressure - Low 3.1 0.71 1.69 2 1945 psig 1 1935 psig
- 10. Pressurizer Pressure - High 3.1 0.71 1.69 5 2385 psig 5 2395 psig
- 11. Pressurizer Water Level-High 8.0 2.18 1.82 5 92% of' instrument span 5 93.8% of instrument span a s~ t ez 10.0 .
- 12. Low Reactor Coolant Flow 4h4- Ehf1F 0.6 2-90-4% of loop design 2 % of loop design flow
- flow *
- Loop design flow = 95,700 gpm # See Note 5 RTD = RATED THERMAL POWER
TABLE 2.2-1 (Continued)
REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS Total Sensor Functional Unit Allowance (TA) 2 Drift (S) Trio Setpoint Allowable Value
- 13. Steam Generator Water 17.0 15.28 1.76 117.0% of narrow range 115.9% of narrow range Level - Low-Low -
instrument span instrument span
- 14. Undervoltage - Reactor 15.0 1.39 0 1 10200 Volts t 9822 volts Coolant Pumps
- 15. Underfrequency - Reactor 2.9 0 0 1 55.5 H'z 1 55.3 Hz Coolant Pumps
- 16. Turbine Trip A. Low Trip System Pressure 2 (800) psig B. Turbine Stop Valve Closure NA NA NA 2 later NA NA NA 1 (1) % open 2 later
- 17. Safety Injection Input NA NA NA NA NA from ESF
- 18. General Warning Alarm NA NA NA NA NA
- 19. Reactor Trip System Interlocks
- a. Intermediate Range NA NA NA 2 lx10-10 amps 16x10-Il amps Neutron Flux, P-6
- b. Low Power Reactor Trips Block, P-7
- 1) P-10 Input NA NA NA 5 10% of $ 12.1% of Rated Thennal Power Rated Thermal Power i
TABLE 2.2-1 (Continued)
REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS NOTATION Total Sensor Functional Unit Allowance (TA) Z Drift (S) Trip Setpoint Allowable Value
- 2) P-13 Input NA NA NA 5 10% RTP 5 12.3% RTP Turbine Turbine Impulse Impulse Pressure Pressure Equivalent Equivalent
- c. Power Range Neutron NA NA NA 5 50% of Rated 5 52.1% of Rated Flux, P-8 Thermal, Power Thermal Power
- d. Power Range Neutron NA NA NA 5 20% of Rated Flux, P-9 5 22.1% of Rated Thermal Power Thermal Power
- e. Power Range NA NA NA 2 10% of Rated 2 7.9% of Rated Neutron Flux, P-10 Thermal Power Thermal Power
- f. Turbine Impulse Chamber NA NA NA 510% Rated Thennal 5 12.3% Rated Thermal Pressure, P-13 Power Turbine Impulse Power Turbine Impulse Pressure Equivalent Pressure Equivalent
- 20. Reactor Trip Breakers NA NA NA NA NA
- 21. Automatic Trip and Inter- NA NA NA NA NA lock 1.ogic
r~
TABLE 2.2-1 ' (Continued)
TABLE NOTATIONS h
NOTE 1: OVERTEMPERATURE AT 1+T IS 1+T S I I AT () , ,p3) I + *35 o 2 1+ 3 Y- )-b I 5 I + '6 Where: AT = Measured AT by RTD Manifold Instrumentation; 1 + T)S 1+T 2S
= Lead-lag compensator on measured AT; T1, T2 =.
Time constants utilized in lead-lag controller for AT, Tj, 2 8 sec.,
T2 $ 3 sec;
- 1 U 1+T 3S
Lag compensator on measured AT; T3
Time constants utilized in the lag compensator for AT, T3 = 0 secs; ATo = Indicated AT at RATED TIERMAL POWER; K1 = 1.0995; K2 = 0.0112; 1+T 4S
=
1+T 53 The function generated by the lead-lag controller fora Ig dynamic compensation;
=
T4, T5 Time constants utilized in lead-lag controller for Tavg, T4 2 33 secs.,
TS $ 4 secs; T = Average temperature. *F; 1
= Lag compensator on measured T 1+T 36 _ avg I
T TABLE 2.2-1 (Continued)
TABLE NOTATIONS (Continued)
, NOTE 1: (Continued)
=
T6 Time constant utilized in the measured Tavg lag compensator,16 = 0 secs; T' =
5 588.5'F (Nominal Tavg at RATED THERMAL POWER);
K3 = 0.000519;
) P = Pressurizer Pressure, psig; P' =
2235 psig (Nominal RCS operating pressure) and t-S = Laplace transform operator, sec-l; y and f1 (al) 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 startup tests such that:
, (i) For qt - 9b between -35% and +8%, fj(AI) = 0, where qt and qb are percent RATED THERMAL POWER in the top and bottom halves of the core respectively, and qt + 9b is total THERMAL POWER in percent of RATED THERMAL POWER; (ii) For each percent that the magnitude of qt - Ab exceeds -35%, the AT Trip Setpoint shall be automatically reduced by 1.09% of its value at RATED THERMAL POWER; and (iii) For each percent that the magnitude of qt - Ab exceeds +8%, the AT Trip Setpoint shall be automatically reduced by 1.00% of its value at RATED THERMAL POWER.
NOTE 2: The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by laore than 2.4% of AT instrument span.
i
TABLE 2.2-1 (Continued)
TABLE NOTATIONS (Continued)
NOTE 3: OVERPOWER AT 0 ( + 'l 3) ( 1 )
I IK 4 ~ (*7 1+T 75
) ~I l l+T 5 6 1 )
U(1+T 6 S 2 I
(1 + T 2) I + '3S 5 6
Where: AT = As defined in Note 1, 1+T jS
= As defined in Note 1, 1+T S 2 T1, T2 = As defined in Note 1, 1
= As defined in Note 1, 1+T 3 3 N
T3
= As defined-in Note 1,
.h AT o = As defined in Note 1, K4
= 1.09, K5
= 0.02/*F for increasing average temperature and 0 for decreasing average temperature, 1+T 7S
=
1+T S The function generated by the rate-lag controller for T,yg dynamic compensation; 7
T7
= Time constants utilized in rate-lag controller for Tavg. T7 > 10 secs.,
= As defined in Note 1, 1+T 63 T6
= As defined in Note 1,
n TABLE 2.2-1 (Continued)
TABLE NOTATIONS (Continued) ()
NOTE 3: (Continued)
K6 =
0.00128/*F for T > T" and K6 = 0 for T 5 T",
T = As defined in Note 1, ,
T" =
Indicated Tavg at RATED THERMAL POWER (Calibration temperature for AT instrumentation, < 588.5'F),
S = As defined in N0tE 1, and .
f 2(AI) = 0 for all AI NOTE 4: The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than 3.4% of span.
NOTE 5: 1.38 is the sensor error for Ta v and 0.47 is the sensor error for Pressurizer Pressure.
These values or the "as measuredg sensor error or any combination must be added to determine S for Equation 2.2-1, for the entire Overtemperature AT channel.
s,
(
BASES FOR SECTION 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 4
h.
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.
( AUG 7 1980 ee
/
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 h
therefore related to DNB THERMAL POWER through '.he M-if and The correlation. Reactor U- TDNBCoolant correlation Temperature has been an 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 ratic, ONBR, defined 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.
g The curves of FigureX (2.1-1 eM (2. '-2)- show the loci of points of THERMALPOWER,ReactorCoolantSy)stempressureandaveragetemperatur 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, FN of 1.55 and a reference cosine with a peag of 1.55 for axial power shape. A$g,llowanceisa included for an increase in F g at reduced power based on the expression:
F
= 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 allowable control rod insertion assuming the axial power imbalance is within the limits of the fj (delta I) function of the Overtemperature trip. When the axial power id. balance 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
-- '~~~
.. 7'. .. . . .__
SAFETY LIMITS BASES I
2.1.2 REACTOR COOLANT SYSTEM PRESSURE TherestrictionofthisSafetyLimitprotectstheintegrityoftheReactorl.
Coolant System (RCS) from overpressurization and thereby prevents .the release 4
of radionuclides contained in the reactor coolant from reaching the containment atmosphere.
The reactor vessel, pressurizer, and the RCS p?,ing, valves and fittings are designed to Section III of the ASME Code for Nuclear. Power Plants which permits a. maximum transient pressure of 110% (2735 psig) of design pressure.
, The Safety Limit of 2735 psig is therefore consistent with the design criteria and associated Code requirements.
The entire RCS is hydrotested at 125% (3110 psig) of design pressure, to demonstrate integrity prior to initial operation.
4
, O9 9
f I
i W-STS 8 2-2
2.2 LIMITING SAFETY SYSTEM SETTINGS (f)
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-mination of the trip setpoints is the determination of the magnitudes of the I
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 accommodate the instrument drift assumed to occur between operational tests and the accuracy to which setpoints can be measured and calibrated, 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
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
.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 calibrating the instrumentation. It allows increased flexibility in those
-cases where tLe 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 i
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 4
i l
assumptions.
1 Satisfying equation 2.2-1 indicates that even though the rack may have
! drifted more than was originally allowed, either the sensor drifted in an i
i opposite direction to compensate for the rack drift (if sensor drift was
. measured) or sufficient margin existed between the selected trip setpoint and l l the values used in the safety analysis to accommodate the excess rack drift.
l 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 drif t 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 j( 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 histable is f_100.62%.
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 System functional diversity. The functional capability at the specified trip setting is required for those 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 avoids unnecessary actuation of the Engineered Safety Features Actuation System.
B 2-5
LIC TING SAFETY SYSTEM SETTINGS t
EASES 1tf* ROR TRIP SYSTEM INSTRUMENTATION SETPOINTS (Continued) 4 The var Reactor trip circuits automatically open the Reac" rip breakers whenever ndition monitored by the Reactor Tri em reaches a preset or calculated 1 .
In addition to redundan ' nnels and trains, the design approach provides a tor Trip Syster' ch monitors numerous system variables, therefore providing i Ss unctional diversity. The functional l capability at the specified trip s .
s required for those anticipatory or diverse Reactor trips for whi no direct c '
was assumed in the accident analysis to enhance the rail 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 .. Ase result from e .ive Reactor Coolant System cooldown and thus avoids ur.necess .
a- ution of the Engineered Safety Features Actuation System.
J Manual Reactor Trio The Reactor Trip System includes manual Reactor trip capability.
Power Rance, Neutron Flux In eacn of the Power Range Neutron Flux cnannels there are twc incependent bistables, each with its own trip setting used for a High and Low Range trip setting. Tne Low Setpoint trip provides protection curing subcritical and low power operations to mitigate the consequences of a power excursion beginning from low power, and the High Setpoint trip provides protection during power cperaticns to mitigate the consequences of a reactivity excursion from all power levels.
The Low Setpoint trip may be manually blocked above P-10 (a power level of approximately 10% of RATED THERMAL POWER) and is automatically reinstated below the P-10 Setpoint.
Power Ranoe, 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 Rarige 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 drcp accident could cause local flux peaking which could cause an unconservative 1ccal DNBR to exist. The Pcwer 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 t .
- for those control rod drop accidents for which DNBR's will be greater thang.
b %nu e W-STS B 2-4 Q
LIM: TING SAFETv SYSTEM SETTINGS EASES Intermediate and Source Range, 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 subcritical condition. These trips provide redundant protection to the Low Setpoint trip of the Power Range, Neutron Flux channels. The Source Range channels will initiate a Reactor trip at about 105 counts per second unless manually blocked
.when P-6 becomes active. The Intermediate Range channels will initiate a Reactor trip at a current level equivalent to approximately 25% of RATED THERMAL POWER unless manually blocked when P-10 becomes active.
Overtemcerature aT The Overtemperature Delta T trip provides core protection to prevent DNE for all combinaticns of pressure, pcwer, coolant temperature, and axial power distribution, provided that the transient is slow with respect to piping transit delays from the core to the temperature detectors (about 4 seconds),
and pressure is within the ran*;e between the Pressurizer High and Low Pressure
- The Setooint is automatically varied with: (1) coolant temcerature to trips.
correct for temperature induced changes in density and heat capacity of water and includes dynamic compensation for piping delays from the core to the loop temperature detectors, (2) pressurizer pressure, and (3) axial power distribu-tion. With normal axial power distribution, this Reactor trip limit is always below the core Safety Limit as shown in Figure 2.1-1. If axial peaks are greater than design, as indicated by the difference between top and bottom power range nuclear detectors, the Reactor trip is automatically reduced according to the notations in Table 2.2-1.
Overcower AT The Overpower Delta T trip provides assurance of fuel integrity (e.g. , no fuel pellet melting and less than 1% cladding 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 from 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."
i 7
W-STS B 2 'i
LIMITING SAFETY SYSTEM SETTINGS
.f EASES Pressurizer Pressure In each of the pressurizer pressure channels, there are two independent bistables, each with its own trip setting to provide for a High and Low Pressure trip thus limiting the pressure range in which reactor operation is permitted. The Low Setpoint trip protects against low pressure which could lead to DNB by tripping the reactor in the event of a loss of reactor coolant pressure.
On decreasing power the Low Setpoint trip is automatically blocked by P-7 (a power level 'of approximately 10% of RATED THERMAL POWER with turbine impulse chamber pressure at approximately 10% of full power equivalent); and on increasing power, automatically reinstated by P-7.
The High Setpoint trip functions in conjunction with the pressurizer relief and safety valves to protect the Reactor Coolant System against system overpressure.
Dressurizer Water Level The Pressurizer High Water Level trip is provided to prevent water relief through the pressurizer safety valves. On decreasing power the Pressurizer High Water Level trip is automatically biccked by P-7 (a power level of approximately 10% of RATED THERMAL PCWER with a turbine impulse chamber pressure at approximately 10% of full equivalent); and on increasing power, automatically reinstated by P-7.
Low Reactor Coolant Flow The Low Reactor Coolant Flow trips provide core protection to prevent DNS by mitigating the ccnsequences 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%
@ of full power equivalent), an automatic Reactor trip will occur if the flow in more than one loop drops below 90% of nominal . full loop flow. Above P-8 (a gd power level of approximatelyp 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 Oceration
,h The P-8 setpoint trin u"' Ant the im.:-- - n lue of the DNBR from goino bal-- 1.00 uuring normal operational transients and anticipo - t- ~ 4ents e
W-STS B 2-5
LIMITING SAFETY SYSTEM SETTINGS
/
EASES 4Lqtional for Plant Permitted N-1 Loop Ooeration (Continued) when (n-1) loop in operation and the Overtemperature De t' ' h-tp Tetpoint -
is adjusted to the va cified
- for all leo " ' e ration. With the Og Overtemperature Delta T Trip e- s ed to the value specified for (n-1) loop operation -
rip at /
FD THERMAL POWER will prevent t'e minimum v o the DNBR from going below 1.3 .
m al operational
..s and anticipated transients with (n-1) loops in operation.
tra 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 provices allowances for starting delays of the Milicy Feecwater System.
b ergency
- Q am/Feedwater Flow Mi'smatch and Low Steam Generator Wate- Level
.The . 'Feedwater Flow Mismatch in coincidence with a Steam Generat-Low Water Level '- is not used in the transient and accident analyses .,ut is -
included in Table 2.2 . *e ensure the functional capability of the =cified
.. trip settings and thereby . nca the overall reliability of t eactor Trip System. This trip is redundant he Steam Generator Wa evel Lcw-Low trip. The Steam /Feedwater Flow Misma d oortion of * .s trip is activated when the steam flow exceeds the feedwater .' ,f greate* than or equal to (1.42 x 105) lbs/ hour. The Steam Gener * . Low ter level cortion of the
@ trip is activated when the water i drops below ? %, as indicated Dy the narrow range instrument. TF . trip values include suf..J ent allowance in excess of normal oper + ' g values to preclude spurious trips 't will initiate a Reactor trip b e the steam generators are dry. Therefore, . required capacity an earting time requirements of tne auxiiiary feecwater pc..., are reduce d the resulting thermal transient on the Reactor Coolant System
+
generators is minimized.
Undervoltace and Underfrecuency - Reactor Coolant Pumo Busses The Undervoltage and Underfrequency Reactor Coolant Pump Bus trips pro-vide core protection against DNB as a result of complete loss of forced coolant
- flow. The specified Setpoints assure a Reactor trip signal is generated before the Low Flow Trip Setpoint -is reached. Time delays are incorporated in
.the Underfrequency and Undervoltage trips to prevent spurious Reactor trips from momentary electrical power transients. For undervoltage, the delay is set so that the time required for a signal to reach the Reactor trip breakers following the simultaneous trip of twa or' more reactor coclant pump bus circuit breakers shall not exceed (1.2) seconds. For underfrequency, the delay is set so that the time required for a signal to reach the Reactor trip breakers af ter 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%
9 PSTS B 24 SEP1 5IE
f LIMITING SAFETY SYSTEM SETTINGS i
BASES Undervoltace and Underfrecuency - Reactor Coolant Pumo Busses (Continued) of RATED THERMAL POWER with a turbine impulse chamber pressure at approximate 10%
by of full power equivalent); and on increasing power, reinstated automatically P-7.
Turbine Trio A Turbine trip initiates a Reactor trip. On decreasing power the Reactor O6 trip from the Turbine trip is automatically blocked by P-9 (a power level of h approximate 1yyK% 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 autcmatic 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 Table 2.3-3.
General Warnino Alarm '
A General Warning Alarm in both Solid State Trio System trains initiates a Reactor trip. The General Warning Alarm is activated in each train of the Solid is otherwise State Trip System when the train is being tested or inoperable. The General Warning Alarm trip provides protection for conditions inoperable. under which both trains of the Trip System may be rendered 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), previd :
@ b::ku;. block f r Scurce Range Neutr;r "= d;ubling, and de cr.crgi;a: tha high voltage to t" detectcra 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 h bus undervoltage and underfrequency, Turbine trip, pressurizer icw pressure and pressurizer high level. On decreasing pcwer, the above listed trips are automatically blocked.
i 10 W-STS B 2-E
~ ~~
LIMITING SAFETY SYSTEM SETTINGS
- (,
i '
BASES Reactor Trip System Interlocks (Continued)
P-8 On fl increasing power P-8 automatically enables Reactor trips on low ccowler.t inpump one orbr:more hcr reactor epen. coolant loops, end :n Or ::r: re::tc7
((h On decreasing power, the P-8 auto-matica11y blocks the above 1Arted trip %.
P-9 On increasing power P-9 automatically enables Reactor trip on Turbine trip. On decreasing power, P-9 automatically blocks Reactor trip on Turbine trip.
P-10 On increasing power P-10 allows the manual block of the Intermediate Range trip and the Flow Setpoint Power Range trip; and automatically voltagethe blocks power.Source Range trip and de energizes the Source Range high On decreasing power, the Intermediate Range trip and the Low setpnint Power Range trip are automatically reactivated.
Provides input to P-7.
P-13 Provides input to P-7. *
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e H
W-STS B 2 '%
- - - ' ~ ^ ~ ~~ \
/
SECTIONS 3.0 AND 4.0 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS
/
d
'i.
I 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 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, T 2. At least HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and
- 3. At.least COLD SHUTDOWN within the subsequent 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
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.
(,
W-STS 3/4 0-1 JUL 2 71981
r APPLICABILITY .
SURVEILLANCE REQUIREMENTS 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 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 S'urveillance 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 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 Condition for Operation have been performed within the stated surveillance 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 has been granted by the Commission pursuant to 10 CFR 50, Section '
50.55a(g)(6)(i).
8 E-STS 3/4 0-2 JUL 2 31980
i 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 366 days f
- 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.
f W-STS 3/4 0-3 'NOV 2 61990
JUSTIFICATIONS Section 3/4.1 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 margin 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. The standard specification calls for a plant shutdown if the EOL HTC limit is exceeded. This change would allow the plant staff to develop operating limits that would restore its limit (e.g., reduce RCS temperature). This is similar to what is currently allowed if violations of the BOL case occur.
D. Heat tracing is not required on any CVCS components which contain 4 wt.
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. This note added as a result of the boron dilution event analysis.
F. This specification added as a result of the boron dilution event analysis.
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 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 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 change incorporated as a result of the cold overpressure mitigation analysis.
l 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 Ef0 RATION CONTROL SHUTDOWN MARGIN - T >200*F LIMITING CONDITION FOR OPERATION 3.1.1.1 l3 The SHUTDOWN MARGIN shall be greater than or equal to (4-6%) delta k/k for ) loop operation.
APPLICABILITY: MODES 1, 2*, 3, and 4.
ACTION:
13 With the SHUTOOWN MARGIN less than (4-6%) delta k/k, immediately initiate and continue boration at greater than or equal to So gpm of a solution containing greater than or equal to_7oon ppm boron or equivalent until the required SHUTOOWN MARGIN is restored.
SURVEILLANCE REQUIREMENTS 4.1.1.1.1 The SHUTOOWN MARGIN shall be determined to be greater than or equal to (4 r6%) delta k/k:
l3
- a. Within one hour after detection of an inoperable control rod (s) and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 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 MOON 2 with K,ff greater than or equal to 1.0 at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by verifying that control bank withdrawal is within the limits of Specification 3.1.3.6.
- c. When in MODE 2 with K,ff less than 1.0, within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> 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 Ifmit of Specification 3.1.3.6.
- See Special Test Exception 3.10.1.
W-STS 3/4 1-1
'NOV 2 01980
REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
- e. When in MODES 3 or 4, at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by consideration of the following factors:
- 1. Reactor coolant system boron concentration,
- 2. Control rod position, *-
- 3. Reactor coolant 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 (nc m:1! zed' to correspond to the actual ;
core conditions prior to exceeding a fuel burnup of 60 Effective Full Power Days after each fuel loading.
- Wdh all confrel rods known 4o be. Glly inserfed , s Huroown #Mecrin g '
deler*inalien way irke crejit for %e yeae.+iudy worth of the conitol rol assumed io be Gily w &),,mn.
W-STS 3/4 1-2 4
REACTIVITY CONTROL SYSTEMS
/
SHUTOOWN MARGIN - T $ 200*F LIMITING CONDITION FOR OPERATION 3.1.1.2 The SHUTDOWN MARGIN shall be greater than or equal to(.1 4-0% delta k/k.
APPLICABILITY: MODE 5. ,
ACTION:
1M With the SHUTDOWN MARGIN less than 4-0% delta k/k, immediately initiate and continue boration at greater than or equal to Jpq_ gpm of a solution containing greater than or equal to_2gp ppm boron or equivalent until the required SHUTOOWN MARGIN is restored.
SURVEILLANCE REQUIREMENTS - 4.1.1.2 The SHUTDOWN MARGIN shall be determined to be greater than or equal to delta k/k:
I a. Within one hour after detection of an inoperable control rod (s) and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 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),
- b. At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by consideration of the following factors:
- 1. Reactor coolant system baron concentration,
- 2. Control rod position, **
- 3. Reactor. coolant system average temperature,
- 4. Fuel burnup based on gross thermal energy generation,
- 5. Xenon concentration, and
- 6. Samarium concentration.
T_h RCS.n.boven conc.e,d raba dall be 2.2.000 em when ce*lant (**F d*e ** A d'48E85 P
44 With all conbel vods knota n 40 be blly 56Seded, S HLATDoulo HARGin h deferminafson mayt alze crelst -for %e rede+I Aiy worth of tiie confrol ved assumed 4o be fully w'rth d va w n-W-STS 3/4 1-3 NOV 2 01980
REACTIVITY CONTROL SYSTEMS MODERATOR TEMPERATURE COEFFICIENT LIMITING CONDITION FOR OPERATION 3.1.1.3 The moderator temperature coef ficient (MTC) shall be:
- a. Less positive than 0 delta k/k/'F for the all rods withdrawn, beginning of cycle life (BOL), hot zero THERMAL POWER condition.
- b. Less negative than -5.6 x 10 ' detta k/k/'F for the all rods
~
withdrawn, and 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:
With the MC more positive than the limit of 3.1.1.3.a above, or less negative than the limit of 3.1.1.3.b above, operation in MODES 1 and 2
- may proceed provided:
- 1. Operating limits are established and maintained sufficient to restore the MC to within its limit within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
- 2. The operating limits shall be maintained untti a subsequent calcu-lation verifies that the MC 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 Commission pursuant to Specification 6.9.2 within 10 days, describing the value of the measured M C, the interim operating limits, and, if applicable, the predicted average core burnup necessary for restoring the positive MTC to within its limit for the all rods withdrawn condition.
- With ef f greater than or equal to 1.0.
- See Special Tes t Exception 3.10.3.
3hlJI
REACTIVITY CONTROL SYSTEMS t
i SURVEILLANCE REQUIREMENTS 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~
@ co(#rG) delta k/k/*F (all rods withdrawn, RATED THERMAL POWER x 107 EFPD ndition) within } be after measured reaching anatequilibrium any THERMAL POWER boron concen-tration of 300 ppm. Intheevent.ghiscomparisonindicatestheMTC is more negative than -(-3-0-) 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, o
W-STS 3/4 1-5 AUG I 1973
REACTIVITY CONTROL SYSTEMS
( MINIMUM TEMPERATURE FOR CRITICALITY LIMITING CONDITION FOR OPERATION 3.1.1.4 The Reactor Coolant System lowest operating loop temperature (Tavg) shall be greater than or equal to (44t)*F.
APPLICABILITY: MODES 1 and 2 .
ACTION:
less than With a Reactor CoolanttoSystem within operating its limit loop temperature within 15 minutes(T*89)be in HOT SSI STANDBY (4H-)*F, restore T*XExt 15 minutes.
within the SURVEILLANCE REQUIREMENTS 4.1.1.4 The Reactor Coolant System temperature (T**9) shall be determined to be greater than or equal to (-5++)*F:
- a. Within 15 minutes prior to achieving reactor criticality, and i
- b. At least once per 30 minutes when the reactor is critical and the Reactor Coolant System T is less than (654)*F with the Tavg-Tref DeviationAlarmnotreseU9 54/
- With K greater than or equal to 1.0.
- SeeSpIbalTestException3.10.3.
W-STS 3/4 1-6 15 M
REACTIVITY CONTROL SYSTEMS 3/4.1.2 BORATION SYSTEMS FLOW PATH - SHUT 00WN
. 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 5. 4,S',o.nd 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 4.1. 2.1 At least one of the above required flow paths shall be demonstrated OPERABL
, ). At ica:t cnce per ' day: by verifyir.g that the temperatura of the-hcat trocad partier. Of the flow path is grcater thar er equal to-(55) F when : 'lce path %= the beric acid tanks is usedr X 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. J t
W-STS 3/4 1-7 JUL 2 71981
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 OPERABLE:
- a. The flow path from the 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 pumps to the Reactor Coolant System.
APPLICABILITY: MODES 1, 2, 3f-and 4.P @
ACTION:
With only one of the above required boron injection flow paths to the Reactor Coolant System OPERABLE, restore at least two boron injection flow paths to the Reactor Coolant System to OPERA 3LE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY and borated to a SHUTDOWN MARGIN equivalent to at least 1% delta k/k at 200 F within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; restore at least two flow paths to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the -
( next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
SURVEILLANCE REQUIREMENTS 4.1.2.2 At least two of the above required flow paths shall be demonstrated OPERABLE:
X ^t le::t once per 7 day: by verifying that the te=perature of-tfre--
heat trac @r4 ton-of-the-f4ew-path frc: the bor-ic-ac-hl-tenks-is-
@ grantar than er equc] te (65)CC when it is 3 aquired wctcp ccurce
- o. )(. 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 18 months during shutdown by verifying that each automatic valve in the flow .ath ac"uates to its correct position on a
W test signal. s= c h ti e(dio3 @
At least once per 18 months by verifying that the flow path required C. K. by Specification 3.1.2.2.a delivers at least 30 gpm to the Reactor Coolant System.
N;1y ci.c beren injacHea " e" path ic required to bc 0"E"ASLE whcnever the @
v
- tvagrature of en^ er =nra nLthe RCS celd les; i; Ic;c than-cr equal te (275)"IJ W-STS 3/4 1-8 gna , t jgg;
e REACTIVITY CONTROL SYSTEMS t
CHARGING PUMP - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.2.3 One charging pump in the boron injection flow path required by Specification (3.1.2.1 OPERABLE emergency powe)r source.shall be OPERABLE and capable of being power APPLICABILITY: MODES i =d C. ,5and 6 ACTION: -
ca 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. ,
- b. Tke grovtsioots e-f SPectCtca[thn 3 0 9 are ^*Y "PPh caMe for e4*y into HoDE 'l f*om HdDE 3-
,.. SURVEILLANCE REQUIREMENTS 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 29go psig when tested pursuant to Specification 4.0.5. g 31 clam 4.1.2.3.2 All charging pumps, excluding the abov equired OPERABLE pump, shall be demonstrated inoperable *It least once per ,2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />;, except when the reactor vessel head is removed, by verifying that the motor circuit breakers
% fe been rere'?:d fic., thei c!:ctrical power supply + cuits_
o.v e Tecv ed in ibe open posifada, g (Jith(n 11 hours1.273148e-4 days <br />0.00306 hours <br />1.818783e-5 weeks <br />4.1855e-6 months <br /> Prter Yo .eMv1 into h0DE 3 from Hot >E 4 , ba% ch-h pmps, reg ude) by Spectficah6n 31 M shall be o Pege gy:,
O w -* Cln inopera.kle pu%p ~ay be enegi3e1 -few +c563 P<outdel The dtsc ha*<3 e of the Pump has been isolaled fvo, & Rcs by a. clesed
.(
- iso \dn'on va\ve tuntk % pousev vemoved Evo,n se valve op eraiov, l' ov by a manual isolation valve sectgect in g closej posiS ,
W-STS 3/4 1-9 JUL 2 71981
REACTIVITY CONTROL SYSTEMS CHARGING PUMPS - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.4 At least two charging pumps shall be OPERABLE.
APPLICABILITY: MODES 1, M2, 3 and +4- .
h .
ACTION:
With only one charging pump OPERABLE, restore at least two charging pumps to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY and borated to a SHUTDOWN MARGIN equivalent to at least 1% delta k/k at 200 F within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; restore at least two charging pumps to OPERABLE status within the next 7 days or be in COLD SHUTOOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
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 2_qn psig when tested pursuant to Specification 4.0.5. @
-4.1.2.4.2 ^11 charging p"=ne, axecpt the ebeve requires npraan' E- pump, chc!1
-tc demonstretcd #neper:ble et lea;t crea Par 1? Scure whencycr thc temper 2t"re of crc cr mere of the RCS ce!d legs is le:: then er equal te (275) r by verifying--
-thet thc motor ci.cuit breakcr; have been remcycd frcm their electrical pcuer
- uppl; ci cuits.
~#I mcximee of ora cantrif" gal chaeging pump shall be OPERASLE whenever the
--temp e r-e t "ra af ene e" =cre ef the RCS cold legs is Icsa then or equal te -
' m" , .,
s c h o .c . -
\
s W-STS 3/4 1-10 MAY 1 5 1978
REACTIVITY CONTROL SYSTEMS BORATED WATER SOURCE - SHUTOOWN LIMITING CONDITION FOR OPERATION 3.1.2.5 As a minimum, one of the following borated water sources shall be OPERABLE:
- a. A boric acid storage system and td &_t 1:::t One 2:ccciated h::t tracing
-cyste- "ith-1.
A minimum contained borated water volume of WOOgallons,
- 2. A mm _ boron conce drak.t af 700a ppm, md
-Ects::cr (20,000) and (22,500) pp. c' beren, and
- 3. 46 A minimum s,olution temperature of (M53 F. A
- b. The refueling water storage tank with:
1.
A minimum contained borated water volume of 2't.5oo gallons,
- 2. A minimum boron concentration of (2000) ppm, and i 3. A minimum solution temperature of (35)*F.
APPLICABILITY: MODES 5 and 6.
ACTION:
With no borated water source OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes.
SURVEILLANCE REQUIREMENTS 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 i:-th: Ocuronce c eper 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> r bamted " terby verifying and-thc the RWST (cutside) temperature.-ter ci- temperature ii.c it g Jc:: than (25) I.
k s W-STS 3/4 1-11 $0V 2 01980
I REACTIVITY CONTROL SYSTEMS B0 RATED WATER SOURCES - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.6 As a minimum, the following borated water source (s) shall be OPERABLE as required by Specification 3.1.2.2:
A boric acid storage system /p- c_ ct least@ene arreciated heat-traci a.
aycter :fitF-
- 1. A minimum contained borated water volume of 20,200 gallons, b
- 2. 0
". LY._.*w?,T
-- o,vov n?n?,
,~ - "$"N&Y,EY,
, vu re.. *?2 9?.
- .O?_
P?.~T* **A 47
- 3. A minimum solution temperature of (-ME) F. g
- b. The refueling water storage tank with:
- 1. A contained borated water volume of between #79400 and WE000 gallons, f1 ,m s au m .m ho-on concentva.f5n c4 3000ffm> a nd
(, 2. Set:::er (2000) and (2100) pp= af h a a n "1 (
- 3. A minimum solution temperature of (35) F.
APPLICABILITY: MODES 1, 2, 3 and 4.
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 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and 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 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
- 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 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
t W-STS 3/4 1-12 N0y 2 0 880
- 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,
- 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 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by verifying the RWST temperature.h g th- (e'itsMe) cir, temperaturc i; le;; ther. (25) I yy ,
-(
s
\
W-STS 3/4 1-13 NOV 2 01980
~
REACTIVITY CONTROL SYSTEMS UNBORATED WATER SOURCES - SHUTDOWN LIMITING COKDITION FOR OPERATION 3.1.2.7 The Boron Thermal Regeneration System shall be rendered incapable of performing its dilution function by:
- a. Removing power from the Chiller Compressor (CS-E-18), and
- b. Positioning the three-way valve (HCV-387) to bypass all flow around the Thermal Regenerative Demineralizers.
APPLICABILITY: MODES 4, 5, and 6 ACTION:
- a. With either an OPERABLE Chiller Compressor or flow path alignment to the Thermal Regenerative Demineralizers, immediately remove power from o
the Chiller Compress'r and/or align the flow path to bypass the demineralizers.
- b. In the event that testing is required during or after maintenance, operation of the Chiller Compressor or the three-way bypass valve is permitted as long as the two components are not out of Specification 3.1.2.7 simultaneously. During such testing the unaffected component shall be verified to be in compliance with this specification throughout the test. When the test is completed all conditions for Specification 3.1.2.7 shall be re-established.
SURVEILLANCE TIEQUTREMENTS 4.1.2.7 The above required ccaditions shall be verified at 1 cast once per 31 ,
days. '
i
(
, (
l
( 3/4 1-14
REACTIVITY CONTROL SYSTEMS i
3/4.1.3 MOVABLE CONTROL ASSEMBLIES GROUP HEIGHT LIMITING CONDITION FOR OPERATION 3.1.3.1 All full length { shutdown and controlj rodsy, and all pert length rod; which are inserted in the core, shall be OPERABLE and positioned within 12 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 I hour and be in HOT STANDBY within 6. hours.
- b. With more than one full-er part-length rod inoperable or misaligned from the group step counter demand position by more than i 12 steps b
(indicated pocition), g in HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
- c. With one full-ci pert Mgth rod trippable but inoperable due to causes other than addressed by ACTION a, above, or misaligned from
( its group step counter demand height by more than i 12 steps (indicated position), POWER OPERATION may continue provided that within 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 2 12 steps of the inoperable rod while maintaining the rod sequence and insertion limits of Figuret (3.1-1) and (? 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 conditions.
b) The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is determined at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
( *See Special Test Exceptions 3.10.2 and 3.10.3.
II W-STS 3/4 1-14 4;CV 2 1981
REACTIVITY CONTROL SYSTEMS ACTION (Continued) c) Apowerdistributionmapisobtagnedfromthemovable incore detectors and F (Z) and F are verified to be withintheirlimitswikhin72hohgrs.
)l1 The T" P"^.L POL' : level L .:due:d +: 1 :: ther er equal te 75M Of "ATCO T"CRMAL PCWER ; thin ih: n:xt h ar end eithi- the fellowing 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> the high neutrer uy
@ -trip setpa4at fe r:duced to less than er equ:1 to SS%
of "^TED T" "Jt"; POWER.
- d. See I nsert I SURVEILLANCE REQUIREMENTS
/
i 4.1.3.1.1 The position of each full --d 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 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> except during time intervals when the Rod Position Deviation Monitor is inoperable, then verify the group positions at least once e per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. W 4.1.3.1.2 Each full length rod not fully inserted 2nd :: 5 p:rt !:rgth red
=hich i; in;;rted '- th: cene shall be determined to be OPERABLE by movement of at least 10 steps in any one direction at least once per 31 days.
k W-STS 3/4 1- Id3Y 2
1 INSERT I
-(.
- d. With more than'one rod trippable but inoperable due to causes other than addressed by ACTION a above, POWER OPERATION may continue provided that:
- 1. . Within one hour, the remainder of the rods in the bank (s) with the inoperable rods are aligned to within + 12 steps of the inoperable rods while maintaining the rod sequence and insertion limits of Figure 3.1-1. The THERMAL POWER level shall be restricted pursuant to Specification 3.1.3.6 during subsequent operation, and
- 2. The inoperable rods are restroed to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
(
TABLE 3.1-1 ACCIDENT ANALYSES REQUIRING REEVALUATION IN THE EVENT OF AN INOPERABLE FULL OR r"A^T LENGTH ROD ( Rod Cluster Control Assembly Insertion Characteristics Rod Cluster Control Assembly Misalignment Loss Of Reactor Coolant From Small Ruptured Pipes Or From Cracks In Large Pipes Which Actuates The~ Emergency Core Cooling System Single Rod Cluster Control Assembly Withdrawal. At Full Power Major Reactor Coolant System Pipe Ruptures (Loss Of Coolant Accident) , Major Secondary System Pipe Rupture Rupture of a Control Rod Drive Mechanism Housing (Rod Cluster Control Assembly Ejection) ( ( 0 W-STS 3/41-$ OCT 1 1976
REACTIVITY CONTROL SYSTEMS POSITION INDICATION SYSTEMS-OPERATING LIMITING CONDITION FOR OPERATION 3.1.3.2 The rhett e , e-+*e' end pa*+ 'ength control rod position indication system and the demand position indication system shall be OPERABLE and capable OI of' 0determining a4L "JA7 steps.the controlbanks For sh+Jewa rod positions
,% pasdionwithin i 12 steps dor conwel h=& he+ ween mu.4 6e determiaet for m range be.4 ween oandW APPLICABILITY: MODES 1 and 2. Steps and ato and 217 s+ces.
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 motio.n of the non-indicating red which exceeds 24 steps in one direction since the last determination of the rod's position, or
- 2. Reduce THERMAL POWER T0 less than 50% of RATED THERMAL POWER
/ within 8 hours.
- b. With a maximum of one demand position indicator per bank inoperable 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 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. h W-STS 3/4 1- JUL 151979
REACTIVITY CONTROL SYSTEMS ( POSITION INDICATION SYSTEM-SHUTDOWN LIMITING CONDITION FOR' OPERATION 3.1.3.3 One rod position indicator (excluding demand position indication) shall be OPERABLE and capable of determining the control rod position within r ea control er aart 'e i i12ste8..sfEs.wch-chutd:un6 F. c,o u..,.
- w. m
*i4 oa ai==+ a,ed**=>
4.+r. - .' .a to. weLe 6
= ageamel bbera aw,eea o. 4 iom,nath
, + ,y rod not full.lasy.i ns.tcps.
a e ,-pi g Fy m in ba
.a aio ,,4 asy , ,,,.
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 REOUIREMENTS ( 4.1.3.3 Each of the above required rod position indicator (s) shall be determined to be OPERABLE by performance of a -CilAsii4EL TU: TIO"^1 TEST at least once per 18 months. OtG.lTAL c>m uuri. orcannovat.Tesr "With the reactor trip system breakers in the closed position.
#See Special Test Exception 3.10.5.
i W-STS 3/41-k NOV 2 01980
i REACTIVITY CONTROL SYSTEMS R00 DROP 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 (2.2) secondc from beginning of decay of stationary gripper coil voltage to dashpot entry with: 500
- a. T,yg greater than or, equal to (5+4)*F, and
- b. All reactor coolant pump; 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. "itt the -ed drop times witP'- 'i-it; but deter. Tined with n-1 ea-+^e 00!:nt pumps operating, Operation m;y prc:: d pr .ided TurDMM - _n ~ ,e n e. ....:..a . ( 1s L :: than Or equal tc (00)% of ^TED Ti: R"AL POWER -Len th: re :ter cecient step v !ve '" the ncneperating !cep cre cpen,
-se.
1L. be:: ther c eque' te (75)% of RAT 0 T":"."'.L PC';!ER when the rc::ter reelen+ e+0p v21v:: '^ the nen;peroin,3 1:0; cre !;;;d. SURVEILLANCE REQUIREMENTS 4.1.3.4 The rod drop time of full length rods shall be demonstrated through measurement prior to reactor criticality:
- a. For all rods following each removal of the reactor vessel head,
- b. For specifically affected individual rods following any maintenance on or modification to the control rod drive system which could affect the drop time of those specific rods, and
- c. At least once per 18 mor.ths.
\
20 y-STS 3/4 1-lQ GCT 1 9 75
REACTIVITY CONTROL SYSTEMS i , SHUTOOWN R00 INSERTION LIMIT LIMITING CONDITION FOR OPERATION 3.1.3.5 All shutdown rods shall be fully withdrawn. APPLICABILITY: MODES 1* and 2*#. 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
- b. Declare the rod to be inoperable and apply Specification (3.1.3.1).
( SURVEILLANCE REQUIREMENTS 4.1.3.5 Each shutdown rod shall be determined to be fully withdrawn:
- a. Within 15 minutes prior to withdrawal of any rods in control banks A, B, C or 0 during an approach to reactor criticality, and
- 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.
( 11 W-STS 3/4 1'2Q 'NOV 2 01980
REACTIVITY CONTROL SYSTEMS CONTROL ROD 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) :nd (3.1-2). (g) APPLICABILITY: MODES 1* and 2*#. 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 figure %, or
- c. Be in at least HOT STANDBY within 6 hours.
1 SURVEILLANCE REOUIREMENTS 4.1.3.6 The position of each control bank shall be determined 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 rod positions at least once per 4 hours.
*See Special Test Exceptions 3.10.2 and 3.10.3. #With K,ff greater than or equal to 1.0.
( 11 . W-STS 3/4 1-M. NOV 2 01980
228 220 / (0.30.' 228) /(0.844,228) 200 /
/ /
180
.0,164) g 160 5
w 140- / [ / Il O' 146) 120-8 E 100- / > /
/
( w 80- l 60- - - O f(o.o, 49) O oc 40-20-
.31, 0)
.g i . .
o' O.2 0.4 0.0 0.8 1.o FRACTION OF RATED THERMAL POWER FIGURE 3.1-1 CONTROL ROD INSERTION LIMITS ( AS A FUNCTION OF POWER 3/4 1-
R CTIVITY CONTROL SYSTEMS PAR \(ENGTH ROD INSERTION LIMITS (OPTIONAL) (((} LIMITING NDITION FOR OPERATION
\ /
3.1.3.7 The p rt length control rod bank shall be:
- a. Limite in physical inscrtion as shown on Figure 3.1-3), and
- b. Limited from c ering any axial segment of the fuel assemblies for a period in exces of (18) out of any 30 Equivalent ull Power Days.
i APPLICABILITY: 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 that fraction of RATE RMAL POWER which is allowed by the bank position using e ab e figure, or I
- 3. Be in at least HOT TANDBY wi in 6 hours. -
- b. With nhe neutron abs ber section of e part length control rod bank covering any a al segment of the uel asseiblies 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 above lin' withia 2 hours, or
- 2. Be in least HOT STANDBY within the next hours.
SURVEILLANCE REQU EMENTS \
/
4.1.3.7 The sition of the part length control rod bank shall be etermined at least onc per 12 hours. .
*See S ial Test Exceptions 3.10.2 and 3.10.3.
i s w-STS 2/" 1-'d -- .NOV 2 01980
i EAC11VITY CONTROL SYSTEMS g PAR LENGTH R0D INSERTION LIMITS (if required by DNB considerations LIMITI CONDITION FOR OPERATION s e 3.1.3.7 All art length rods shall be fully withdrawn. APPLICABILITY: MODES 1* and 2*. ACTION: With a maximum of o e part length rod not fully w thdrawn, within 1 hour either:
- a. Fully withdr w the rod, or
, b. Be in at least 'T STAND 8Y withi the next 6 hours. f 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 a ve 5% of RATE THERMAL POWER, and '
- b. Verifying, a least once per 31 da s, that electric power has been disconnecte from its drive mechani by physical removal of a breaker fr the circuit.
- See Special est Exceptions 3.10.2. and 3.10.3.
i l i l A
)
I l W-STS :/: 1 s- N0y 2 o 1980 l
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. The standard specification calls for the LCO on FO to be met by performing a surveillance on FXY. This change is made to reflect actual practice and does not involve any technical requirements. D. The standard specification has lost its significance with the elimination of the Rod Bow factor (R2 ). Thus, the specification has been rewritten to apply only to the Nuclear Enthalpy Rise Hot Channel Factor (FA H). The RCS total flow rate specification is covered under Specification 3/4.5.2. E. The Action Statement has been rewritten and simplified so that the action is to verify that peaking factors are within limits. The QPTR does not have a safety limit, it merely indicates that something abnormal is happening in the core that warrants investigation. F. Using just the 4 pairs of symmetric thimbles is limiting in that a plugged thimble or inoperable detector could preclude 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 previous 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). e G. A better determination of change in QPTR can be made on a 24 hour basis since QPTR is not expected to change significantly over a 12 hour period. H. Table 3.2-1 can be deleted and the parameters placed in the Specifica-tion. The additional limits are not applicable since N-1 loop operation is not permitted at Seabrook. J. The AFD is continuously monitored by the Station computer during power operation. The software for this calculation will be tested prior to station operation and will be controlled by administrative procedures. The only failure that will cause a loss of AFD monitoring is a total failure of the Station computer, and this is not considered likely. Should this occur, AFD monitoring will be performed manually until the computer system is restored to service. When the Station computer system is operable, there is no technical justification f or manually monitoring the AFD. K. The RCS total flow limits and sarveillance requirements have been moved from 3/4.2.3 to this specification.
3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX OIFFERENCE LIMITING CONDITION FOR OPERATION 3.2.1 The indicated AXIAL FLUX DIFFERENCE (AFD) shall be maintained within 45)fr target bands (flux difference units) about the target flux difference. S ee 3nse r4 I APPLICABILITY: MODE 1 above 50% of RATED THERMAL POWER *. ACTION: reg,uire b
- a. With the indicated AXIAL FLUX DIFFERENCE outside of the 45-)% 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 b) Reduce THERMAL POWER to less than 90% of RATED THERMAL POWER.
- 2. Between 50% and 90% of RATED THERMAL POWER:
a) POWER OPERATION may continue provided: . ves wel
- 1) The indicated AFD has not been outside of the @_ -
target band for more than 1 hour penalty deviation cumulative during the previous 24 hours, and
- 2) The indicated AFD is within the limits shown on Figure (3.2-1). Otherwise, reduce THERMAL POWER to 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 P0WER within the next 4 hours.
b) Surveillance testing of the Power Range Neutron Flux Channels may be performed pursuant to Specification (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 4(+)% target band and ACTION a.2.a) 1), above has been satisfied. veguirel
*See Special Test Exception 3.10.2.
W-STS 3/4 2-1 NOV 2 01980
1 INSERT I
- a. + 5% for a fresh core with average accumulated burnup of i 3000 S D/MTU.
- b. + 3%, -12% for a fresh core with average accumulated burnup of >3000 MWD /MTU.
- c. +3%, -12% for reload cores.
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 *(+}Wr r eguid target band for more than I 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 P0 DER by:
- a. Monitoring the indicated AFD for each OPERABLE excore channel \
% f t least once.per 7 days when the AFD Monitor Alarm is OPERABLE,
-and.
!% At 1 r+ nnce opr- h- f;r th: W t ?4 heur aft:r rc ter5g:
1- th ^FC ."srit:r f.la.a. Le Or:RACLE .LuLua.
- 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.
4.2.1.2 % reguired The indicated AFD shall be considered outside of it (.4% target band when 2 or more OPERABLE excore channels are indicating the AFD to be outside the target band. Penalty deviation outside of the e(-5-)J6 target band @ shall be accumulated on a time basis of: 'C$ wired
- a. One minute penalty deviation for each 1 minute of POWER OPERATION outside of the target band at THERMAL POWER levels equal to or above 50% of RATED THERMAL POWER, and
- b. One-half minute penalty deviation for each 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 determined by measurement at least once per 92 Effective Full Power Days,w-trh. 21' part length--cont, rob-cod: fully 'ithdrawn. 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
I 120 8 E # E !5 l 1M 0 8 HA
$h UNACCEPTABLE (-11,90) (11,90) UN AOCEPTABLE OPERATION OPERATIGN f l ACCEPTABLE OPERATION
/
(-31,50)
\(31,50) 7 40 .
20 0 40 -30 -20 -10 0 10 20 30 40 50 FLUX DIF FERENCE (al) %
) FIGURE 3.2-1
( AXIAL FLUX DIFFERENCE LIMITS AS A FUNCTION OF RATED THERMAL POWER 3 3/42-\
POWER DISTRIBUTION LIMITS 3/4.2.2 PLANAR RADIAL PEAKING FACTOR - XY LIMITING CONDITION FOR OPERATION RTP L F F F
.3.2.2. XY shall be less than XY and XY where RTP-F XY is the XY limit for RATED THERMAL POWER for applicable core planes RTP containing . bank "D" control rods and unrodded core planes. XY is provided in the Radial Peaking Factor Limit Report per specification 6.9.1.7.
s L RTP XY = XY [1 + .2(1 - P)] F where P is the fraction of RATED THERMAL POWER at which XY was measured. APPLICABILITY: MODE 1 ACTION: RTP
- a. With XY is greater than the XY limit for appropriate measured . core F
plane but less than the XY relationship, additional power distribution p p RTP y L maps shall be taken and XY compared to XY and XY
- 1. Either within 24 hours af ter exceeding by 20% of RATED THERMAL FC POWER or greater, the THERMAL POWER at which XY was las t deter-mined, or
- 2. At least once per 31 EFPD, whichever occurs first.
3/4 2-4
r POWER' DISTRIBUTION LIMITS 3/4.2.3 NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR - FAH () LIMITING CONDITION FOR OPERATION
- 3. 2.3 ' FAH shall be less than 1.49[1.0 + .2(1 - P) }
where P is the fraction of RATED THERMAL POWER at which FAH was measured. APPLICABILITY: MODE 1-ACTION: With FaH exceeding-its limit a.. Within 2 hours reduce THERMAL POWER to the level where the LIMITING CONDITION FOR OPERATION is satisfied.
- b. Identify and correct the cause of the out of limit condition prior to increasing THERMAL POWER above - the reduced limit required by a, above; THERMAL- POWER may' then be increased provided FaH is demonstrated through incore mapping to be within its limits.
SURVEILLANCE REQUIREMENTS 4.2.3.1 The provisions of specification 4.0.4 are not applicable. 4.2.3.2 FaH shall be demonstrated to be within its limit prior to operation l above 75% RATED 'DiERMAL POWER af ter each fuel loading and at leas t once
- per 31 EFPD Lther,eaf ter by l .
j a. Using the moveable incore detectors to obtain a power distribution I map 'at 'any THERMAL POWER greater than 5% of RATED THERMAL POWER.
- b. ITsing ' the measured value of FaH since it contains 4% FaH measure-
' ment uncertainty.
l. i. l I 3/'I20
.. = _
POWER DISTRIBUTION LIMITS
- 1. Lower core region from 0 to 15%, inclusive.
- 2. Upper core region from 85 to 100%, inclusive.
- 3. Grid plane regions at 17.8 f 2%, 32.112%, 46.4 + 2%,
60.612% and 74.9 f 2% inclusive.
- 4. Core plane regions.within i 2% of core height (1 2.88 inches) about the bank demand position of the bank "D" or part length control rods.
/
b f
.~
POWER DISTRIBUTION LIMITS () F
- b. With XY exceeding XY, reduce THERMAL POWER a t leas t 1% for each 1%
F XY exceeds the limit within- 15 minutes and similarly reduce the Power Range Neutron Flux-High Trip Setpoints within the next 4 hours; POWER OPERATION may proceed for up ' to a total of 72 hours; subsequent POWER OPERATION may proceed provided the Overpower delta T Trip ' Setpoints have been reduced at least 1% for each 1% F xy exceeds the limit.
- c. Identify 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 FXY is demons trated through incore mapping to be within its limit.
SURVEILLANCE REQUIREMENTS 4.2.2.1 The provisions of specification 4.0.4 are not applicable. 4.2.2.2 XY shall be demonstrated to be within its limits at least once per 31 EFPD by
- a. Using the movable incore detectors to obtain a power distribution map at any THERMAL POWER greater than 5% of RATED THERMAL POWER.
F
- b. Increasing the measured XY component of the power distribution map o
by 3% to account for manufacturing tolerances and further increasing the value by 5% to account for measurement uncertainties.
- c. The XY limits are not applicable in the following core planes regions as measured in percent of core height from the bottom of the fuel:
/ ,
3/428
r - POWER DISTRIBUTION LIMITS 3/4.2.4 QbADRANT POWER TILT RATIO LIMITING CONDITION FOR OPERATION 3.2.4 The QUADRANT POWER TILT RATIO shall not exceed 1.02. APPLICABILITY: MODE 1 above 50% of RATED THERMAL POWER *. ACTION:
- a. With the QUADRANT POWER TILT RATIO determined to exceed 1.02 within 24 hours and every 7 days thereaf ter, verify that FXY and FAH are within their limits by performing Surveillance Requirements 4.2.2.2 and 4.2.3.2. THERMAL' POWER and setpoint reductions shall be in accordance with the ACTION statements of Specifications 3.2.2 and 3.2.3.
SURVEILLANCE REOUIREMENTS 4.2.4.1 The QUADRANT POWER TILT RATIO shall be determined to be within the limit above 50% of RATED THERMAL POWER by:
- a. Calcula ting the ra tio at least once per 7 days when the alarm is OPERABLE.
- b. Calculating the ratio at least once per 12 hours during steady state operation when the alarm is inoperable.
4.2.4.2 The QUADRANT POWER TILT RATIO shall be determined to be within the limit when above 75 percent of RATED THERMAL POWER with one Power Range Channel inoperable by using the movable incore detectors to confirm indicated QUADRANT POWER TILT RATIO at least once per _24 hours by either g)
- a. Using the four pairs of symmetric thimble locations or
- b. Using the moveable incore detection system to monitor the QUADRANT POWER TILT RATIO subject to the requirements of Specification 3.3.3.2.
*See Special Test Exception 3.10.2 3/4 L?
POWER DISTRIBUTION LIMITS 3/4.2.5 DNB PARAMETERS b( LIMITING CONDITION FOR OPERATION 3.2.5 The following DNB related parameters shall be maintained within the following limits:
- a. Reactor Coolant System T, < 594 'F
- b. Pressurizer Pressure J,2205 psig*
- c. Reactor Coolant System Flow > 391,000 GPM**
APPLICABILITY: MODE 1 ACTION: With any of the above parameters exceeding its limit, re s tore 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 shown above, 4.2.5.1 Each of the parameters f T:b1; 3.2-1 shall be verified to be within their limits at least once per 12 hours. 4.2.5.2 The RCS total flow rate shall be determined by measurement at least once per 18 months. 4.2.5.3 The RCS flow rate indicators shall be subjected to a CHANNEL CALIBRATION at leas t once per 18 months.
- 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% RATED THERMAL POWER.
** Includes a 2.1% flow measurement uncertainty.
3/4 2^ 9
,, .~
TABLE 3.2-1 m @
~
sn DNB PARAMETERS LIMITS N-1 Loops In Opera- N-1 Loops I pera-N Loops In tion & Loop Stop tion & p Stop PARAMETER Operation Valves Open Valv s Closed Reactor Coolant System a 1 (581) F 1 (569) F i (570) F Pressurizer Pressure > (2220) psia
- 1 (2220) psia
- 1 (2220)* psia w
~
\
h
~_
~
- Li not applicable during either a THERMAL POWER ramp in excess of (5%) of RATED THERMAL
" WER per minute or a THERMAL POWER step in excess of (10)% of RATED THERMAL POWER.
G3 3
m JUSTIFICATIONS 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. See RAI 460.35. , J. To keep this table consistent with Table 3.3-6. K. No Waste Gas Holdup System or Hydrogen Monitoring System. L. Change per requirements of GL 83-37. M. See RAI 460.35. N. Due to deletion of non-applicable ACTION statements, numbers have been revised to maintain them in the correct order. P. This Technical Specification is deleted. New Hampshire Yankee will i develop a Turbine 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 Risk Assessment, as suggested by Section 3.1 of NUREG-1024.
R. This addition made to comply with NRC requirements to implement WCAP-10271 and Supplement 1 of the same. T. This added to indicate that instrument (s) tested by the DIGITAL CHANNEL
!' OPERATIONAL TEST.
U. Table 4.3-5 deleted and surveillance requirements stated in the , Specification. V. Table 4.3-6 (same as in U). I 4 W. Table 4.3-7 (same as in U). - X. Provides clarification of the exact type of test to be performed. Y. Table 3.3-2 "RTS Instrumentation Response Times" is removed from the Technical Specifications and will be placed in a licensee maintained and controlled document. The Surveillance Requirement 4.3.1.2 remains in the Specification and, per TS 4.0.3, failure to meet the surveillance causes 4 the RTS Instrumentaiton to be declared inoperable. Thus, the intent of the Specification is not changed. Z. Table 3.3-5 " Engineered Safety Features Response Times" is removed from the Technical Specifications and will be placed in a licensee maintained and controlled document. The Surveillance Requirement 4.3.2.2 remains in the Specification and, per TS 4.0.3, failure to meet the surveillance requirements causes the Engineered Safety Features Actuation System Instrumentation to be declared inoperable. Thus, the intent of the "
- Specification is not changed.
AA. Currently this Specification requries an eventual plant shutdown if the
- number of operable thimbles drops from 43 to 42. The revised Action j statement provides an intermediate area where operation is still permitted provided that an evaluation of the operable thimble distribution be performed. This evaluation determines how the reduced number of thimbles i affects the ability to calculate peaking factors and determine, what if, 4 any increased uncertainties need to be applied to peaking factor i measurements. The lower limit of 50% represents a reasonable lower bound i for'this evaluation. The Surveillance Requirements were removed because, l' as written, it has little to do with tF.e intent of the Specification.
j Furthermore the Tech. Spec. definition of OPERABLE covers detector normalization and plateau determination. 7 BB. This Technical Specification is deleted. The reporting requirements in i the Action Statement and the Surveillance Requirements will be included in a licensee maintained and controlled document. l l t i l l l l
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 RE r0NSE TI"E; as shown-in-
-Tebic 3. 3-2. g APPLICABILITY: As shown in Table 3.3-1.
ACTION: As shown in Table 3.3-1. SURVEILLANCE REQUIREMENTS 4.3.1.1 Each reactor trip system instrumentation ch5nnel and interlock and the automatic trip logic shall be demonstrated OPERABLE by the performance of the reactor trip system instrumentation surveillance 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.
W-STS
- 3/4 3-1 SEP i 51981
TABLE 3.3-1 12" REACTOR TRIP SYSTEM INSTRUMENTATION MINIMUM TOTAL N0. CilANNELS CHANNELS APPLICABLE-FUNCTIONAL Uf1IT OF CHANNELS TO TRIP OPERABLE MODES ACTION
- 1. Manual Reactor Trip 2 1 2 1 2 1 2 1 '2 3 A, 4*, 5* TSID 6
- 2. Power Range, Neutron Flux - liigh 4 2 3 1, 2 2 Setpoint ggg g Low 4 2 3 1 ,2 2 Setpoint
- 3. Power Range, Neutron Flux 4 2 3 1, 2 2 liigh Positive Rate t' 4. Power Range, Neutron Flux, 4 2 3 1, 2 2
# liigh Negative Rate
- 5. Intermediate Range, Neutron Flux 2 1 2 1 ,2 3
- 6. Source, Range, Neutron Flux gg Startup 2 2 2 4 A. 1 B. Shutdown 2 1 2 3h 4\, 5\ M 10 6 C. Shutdown-. 2 + -l-- - 3, 4, end 5 +
3 I i 2. c, n
- 7. Overtemperature AT @ ti 2
~
Four Loop pe ', 4 2,, 3 1, 2 9 g Three Loop Operation 1 1 - , . _ B. Three Loop Plant peration 2 e_.
TABLE 3.3-1 (Continued) 1:C h REACTOR TRIP SYSTEM INSTRUMENTATION MINIMUM TOTAL NO. CilANNELS CilANNELS APPLICABLE OF CilANNELS TO TRIP OPERABLE MODES ACTION FUNCTIONAL UNIT
- 8. Overpower AT @ of 1 3 1,1 4"
- 1. rnne Loop Plant
** 9 g hree 4 i, B. Three Loop Plant ?
g Three Loop Operatin 3 2 2 7 N
, operaticn 3 1** 2 1, 2
- 9. Pressurizer Pressure-Low el 2. 3 i t ##
Feu. Loop Plant ~ - - + M Mp t'
+
b 4.B. Three4oop4ht. - 4 --. - 1 Y 10. Pressurizer Pressure--liigh 4 2. 3 1, 2. 4m"
" Tcur Locp Plant 1, 2 -6 g - A. 4- --I , 2 4
-O. Three4oop-511 ant ~3- #
Pressurizer Water Level--liigh 3 2 2 1 7 11.
- 12. Loss of Flow #
A. Single Loop (Above P-8) 3/ loop 2/ loop in 2/ loop in 1 7 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,
w C.II (D
TABLE 3.3-1 (Continued) y REACTOR TRIP SYSTEM INSTRUMENTATION ' MINIMUM TOTAL NO. CilANNELS CilANNELS APPLICABLE FUNCTIONAL UNIT OF CilANNELS TO TRIP OPERABLE MODES ACTION
@ Y/stm. gen. 2/stm. gen. /stm. gen.
- 13. Steam Generator Water Level--Low-Low in any oper-ating stm.
each oper-ating stm. 1, 2 [4 h@ gen. gen. T N am Generator Water Level - Low 2 stm. gen. I stm. gen. I stm. gen. 1, 2 Coincident team / level and level coin- level and b Feedwater flow Misma 2 stm/ feed- cident with 2 s ytm/ flow mismatch 1 stm./ feed f-4tnrmismatch in tm. flow ism 3Ich in same stm. gen. ame stm. gen. or 2 stm. gen, en. level and { T flow mismatch in same steam gen. 19 % Undervoltage-Reactor Coolant y -2/6us Pumps a/bs on z .. ..c i f# V Four Luup Plent-
' 'fuus keg 5"5'5 $ 3 /
- 4. Th.eu Luup Plant 4-+/b s- t + , , ,
f h/M Underfrequency-Reactor Coolant Pumps 4-2/h3 IMSe g,,m 6. ,s aan ou su g j# g
^
Fcur teca D! ant 4 1/ bus -e- . l-
-2. Thnce Le'ep n!ent 34/Lua +- - --l - [-6 m
Q //, R Turbine Trip g - A. Low Fluid Oil Pressure 3 2 2 1 7 u, B. Turbine Stop Valve Closure 4 4 4 1 7# rd i
. TABLE 3.3-1 (Continued) 'T y REACTOR TRIP SYSTEM INSTRUMENTATION MINIMUM-TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION 1 JT8 Safety Injection Input cf from ESF 2 1 2 1, 2 W b nu. Reauu. R & d lymp_ Breaker Position Trip ~ _
A. Above P-8 1/breanc. 1_c --- -1/ breaker ~-- i su g B. Above P-7 and bel peake F -~2 eaker 1 11 per oper-ating loop 1928. Reactor Trip System Interlocks R A. Intermediate Range gg
- Neutron Flux, P-6 2 ! 1 2 2 8 ,
E 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 D. Power Range. Neabn 4 2. 3 1 E Flux, 9 c} Ma e-. C,11
' TABLE 3.3-1 (Continued) y REACTOR TRIP SYSTEM INSTRUMENTATION MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP .0PERABLE MODES ACTION
-EA low Setpoint Power Range Neutron Flux, P-10 4 2 3 1, 2 8
. F E. Turbine Impulse Chamber Pressure, P-13 2 1 2 1 8 19 ?k Reactor Trip Breakers 2 1 2 1, 2 Jr 9 2 1 2 3*, 4*, 5* & to 2.0'2il Automatic Trip Logic 2 1 2 1, 2 .PE 9 2 1 2 3*, 4*, 5* &lo h
Y m u e-M .
1 l l l TABLE 3.3-1 (Continued) TABLE NOTATION
@ *With therod control reactor drive trip system system breakers capable of rod withdrawal. in the closed position, the .
-TNtfaENhas'socNtkNU i. hbp 7vt$uY. TNti Dic -
e ettt-of Cerv-ics haut.u. Ces4ent Leop-shall be pMced ii. j.he , tripped
-condition.- dannel vcpired 4, be orErM 8tE by SPeoLmhon 7.3 2..
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% of 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.
j W-STS 3/4 3-7 SEP 151981
- ,-_.w .
I TABLE 3.3-1 (Continued) ACTION 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 SHUTOOWN MARGIN requirements of Specification 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 .a 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. W-STS
- 3/4 3-8 SEP 151981
TABLE 3.3-1 (Continued) ACTION STATEMENTS (Continued)
- With a channel associated with an operating loop inopernhl< ,
e the inoperable channel to OPE within 2 hours or be in OT ST in the next 6 hours. One channel ass i erating loop may be bypassed for u ours for surveillance es u..g ;;r Saec4fication
" 4.3.1.1.
N ACliUn ' - th the number of OPERABLE Channels one less than the Minim Channe equirement, restore the inoper nel to OPERABLE status wi rs or L POWER to below the P-8 (Power Ran
- lock) setpoint within the n rs. Operation below the - seuru e+ ==y, co ursuant to ACTION 11.
nCi;CM - with the number of OPERABLE Channels one less than _th, ut : ;7 - Channels OPERAdLt requifemen+ ana"' tion may continue provided the inoperable - . ' . is placed in the tripped cnndition wi u r. - 9 ACTION 14 - With 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 }S - 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. W-STS
- 3/4 3-9 SEP i 51981
f TABLE 3.3-2 y REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES FUNCTIONAL UNIT _ RESPONSE TIME
- 1. Manual Rea Trip Not Applicable 2.
o'N Power Range, Heutron Flux i (0.5) seconds *
\
- 3. Power Range, Neutron Flux, liigh Positive Rate Not A licable
- 4. Power Range, Neutron Flux, Not affh'ca ble liigh Negative Rate 1 (0.5) recen&* @
- 5. Intermediate Range, Neutron Flux Not Applicable R 6. Source Range, Neutron Flux Not Applicable
~
Y 7. Overtemperature AT
/ i (4.0) seconds
- o
- 8. Overpower AT Not Applicable
- 9. Pressurizer Pressure--Low i (2.0) seconds
- 10. Pressurizer Pressure--lii0h 5 (2.0) seconds
- 11. Pressurizer Water Level--liigh Not Applicable
$ NeutrondetIctorsareexemptfromresponsetimetesting. Response time of the neutron flux signal portion of the,ctiannel shall be measured from detector output or input of first electronic component in channel. [ -(-Thj r 'provi s i an is-net applic h. to CP's deskete&ef ter January !,1978. Se Reuulatery-Guide-ttiO, N
-Nacticr4477 )
$ \
TABLE 3.3-2(Continued] y REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES FUNCTIONAL UNIT _ RESPONSE TIME
- 12. Loss of Flow A. Single Loop (Above P-8) < (1.0) seconds B. Two Loops (Above P- nd below P-8) 3(1.0)s'econds
/
- 13. Steam Generator Water Level- ow-Low <, (2.0) seconds
- 14. Stca... Ce&u. utar L'atc; Leuc!-!c.. C;!&cidu..t with
@ 4 t e = Maad e t e r!!!" ^/. 3 .;tch Not Applicable fder-14Ti. Undervoltage-Reactor Coolant Pumps < (4-6-) seconds R
- 15%.
Y Underfrequency-Reactor Coolant Pumps [ < (0.6) seconds Z t/, Tit. Turbine Trip A. Low Fluid Oil Pressure Not Applicable B. Turbine Stop Valve Not Applicable 17M. Safety Injection Input from ES Not Applicable
/ 'N
@l3. Rea d ci Ccalar.t Puci.p Ci cau. Tes ii. ;vn T ;p
/
Net f.pp!icablc IS 20. Reactor Trip Syste vInterlocks Not Applicable 19% Reactor Trip,Br/ eakers Not Applicable w 20R Automatic Trip Logic Not Applicable Q / x l to e
,/
/ N.
e
TABLE 4.3-1
'T REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS ]
TRIP ANhl0G ACTUATING MODES FOR CllANNEL DEVICE WilICH CllANNEL Cl!ANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE FUNCTIONAL UNIT CilECK CALIBRATION TEST TEST LOGIC TEST IS REQUIRED
- 1. flanual Reactor Trip N.A. N.A. N.A. R N.A. 1, 2, 3*, 4*, 5*
- 2. Power Range, Neutron Flux liigh Setpoint STS) 0(2,4), %G,00 N.A. H.A. 1, 2 M(3,4),
Q(4, 6), R(4,5) #g W Low Setpoint STM R(4) 'S/tf (1h) N.A; N.A. l , 2 t' 3.. Power Range, Neutron Flux, N.A. R(4) KQQQ N.A. N.A. 1, 2 n., liigh Positive Rate N.A. N.A. 1, 2 b 4. Power Range, Neutron Flux, N.A. R(4) MQ (,iG liigh Negative Rate - m
- 5. Intermediate Range, 51 % R(4, 5) Mit 03) N.A. N.A. 1 , 2 Neutron Flux
- 6. Source Rance, Heutron Flux SM R(4, 5) b N.A. N.A. 2 , 3, 4, 5
- 7. Overtemperature AT S R % QQl) N.A. N.A. 1, 2
- 8. Overpower AT S R KQ00 N.A. N.A. 1, 2
- 9. Pressurizer Pressure--Low 5 R K G,QQ N.A. N.A. 1
- 10. Pressurizer Pressure--liigh S R % Q(ll) H.A. N.A. 1, 2 m
] 11. Pressurizer Water Level--liigh 5 R K G. Cl) N.A. N.A. 1 $ 12. Loss .~, Flow S R % GOh N.A. N.A. 1 'S_.
l l TABLE 4.3-1 (Continued) 1:c J. REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS o u - Q$) TRIP ANALOG ACTUATING MODES FOR CHANNEL DEVICE WHIC!! CHANNEL CilANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST IS REQUIRED
- 13. Steam Generator Water Level--
Low-Low S R M(II.I'd N.A. N.A. 1, 2
- 14. Sicaa Genuiaiu. "ater Leuc! - -- -R--- W -N--A--- h 4 ;-2--
ic.; Cciacident ith Sicn/
@ fccd etcr fle "ise.ci-ch t y h. Undervoltage - Reactor Coolant N.A. R N.A. MG{lbl9 N.A. 1 Pumps 15%. Underfrequency - Reactor N.A. R N.A. KCR(ttit 4 N.A. I w -
Coolant Pumps w -
" 14R 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. J S/*(1,10) N.A. 1 Closure l'1 %. Safety Injection Input from N.A. N.A. N.A. R N.A. 1, 2
@ ESF <
- e. n.2 c#.,tP=pe=s= -*-4. -w ,a. + w Pu ition ' ip W 2fL Reactor Trip System Interlocks Mu Intermediate Range A.
Neutron Flux, P-6 N.A. R(4) KQOQ N.A. N.A. 2 M
$ B. Low Power Reactor G(f 1t)
Trips Block, P-7 N.A. R(4) X (fL) N.A. N.A. 1 m. c C. Power Range Neutron Q ,%In) Flux, P-8 N.A. R(4) K((8) N.A. .N.A. 1 D. Swe %=3e Meub9 A/. A. g(4 a(y, p) 9. A . N. A. I Aut P-9 . A s -- -- -
TABLE 4.3-1 (Continued) < 85 y REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS TRIP ANALOG ACTUATING MODES FOR CHANNEL DEVICE WHICH ~ CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST 15 REQUIRED E1 Low Setpoint Power Range GL (.010 Neutron Flux, P-10 N.A. R(4) h,(1L) N.A. N.A. 1, 2 F l. Turbine Impulse Chamber G.(fi Dl Pressure P-13 N.A. R K (8) N.A. N.A. 1 t9 N,. Reactor Trip Breaker . N.A. H.A. N.A. M (7) N.A. 1, 2, 3*, 4*, 5* 262is Automatic Trip Logic H.A. N.A. N.A. N.A. M (7) 1, 2, 3*, 4*, 5* R o m M
TABLE 4.3-1 (Continued) TABLE NOTATION With the reactor trip system breakers closed and the control rod drive system capable of rod withdrawal. M T# - Below P-6 (Intermediate Range Neutron Flux Interlock) setpoint. M *## - Below P-10 (Low Setpoint Power. Range Neutron Flux Interlock) setpoint. 2.
. (1) -
If not performed in previous days. (2) - Heat balance only, above 15% of RATED THERMAL' POWER. Adjustchannel 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 may 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-OPERATIONAL TEST shall consist of verifying that the interlock is in the re' quired state by observing the permissive annunciator window. , Quaried (9) - -Mont4FyySurveillance 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. (g) - EcuA channel shall be tedcJ at lensi eve *y 92. Jays on a STAGCERE D TEST B A SIS (.\ 2.) ~ Comply udtk surveidauc regunremen43 sf spadChalion +311 Co-any perhon of -the ckannel rep *ed 4o ke *punble by Spec'Wddehn 332. (13) - U nob p erformed in 1he previous 31 days. W-STS 3/4 3-15 Ngy e 1931
INSTRUMENTATION 3/4.3.2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.2 The Engineered Safety Features Actuation System (ESFAS) instrumentation channels and interlocks shown in Table 3.3-3 shall be OPERA 8LE with their Trip Setpoints set consistent with the values shown in the Trip Setpoint column of . Table 3.3-4.-and-wt4h-RESPONSE-TIMES as-shown f r T.51
' ~~
APPLICA8ILITY: 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 column but more conservative than the value shown in the Allowable Value column of Table 3.3-4, adjust the Setpoint consistent with the Trip Setpoint value. .
- b. With an ESFAS Instrumentation or Interlock Trip Setpoint less conserva-tive than the value shown in the A11cwable Value column of Table 3.3-4, either:
- 1. Adjust the Setpoint consistent with the Trip Setpoint value of Table 3.3-4, and determine within 12 hours that Equation 2.2-1 was satisfied for the affected channel, or
- 2. Declare the channel inoperable and apply the applicable ACTION statement requirements of Table 3.3-3 until the channel is restored to OPERABLE status with its Satpoint' adjusted consistent with the Trip Setpoint value.
. Equation 2.2-1 I + R + 5 < TA Where:
Z = 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 from 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,
- c. With an ESFAS instrumentation channel or interlock inoperable, take the ACTION shown in Table 3.3-3. .
(
'f- STS 3/4 3-16
- NSTRUME'o A CN 5 JR'.'E I LL2'.C E RECUI REvE'. 5 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 Reauirements 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. t W-STS 3/4 3-17 e .r,e I *e w
...i Ud
. O G4 e e m ee se e
- TABLE 3.3-3 19 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION u,
MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION t
- 1. SAFETY INJECTION,-RCACTOR TRIP, CErnuATFD T (n LATION ,-
@ CONTROL ROO" ISOLATION, START-
--DIESEL CENEP^ TOP,5, CONIAinc;ENT-COOLING TANS AND E55ENTIAL--
4ERVICC ATER -
- a. Manual Initiation 2 1 2 1,2,3,4 N if
- b. Automatic Actuation 2 1 2 1,2,3,4 M li M*
Logic and Actuation Relays v @ G c. Containment 3 2 2 1,2,3 M*11 Pressure-liigh-l
- d. Pressurizer 4 2 3 1,2,3 # '20*l6 Pressure - Low Steam Line. %su.c-te.a
- e. Diffecmc.t;el- -
- 5 4 *
- I . 1, 2,1 12
- 4'c e55 urc 4ettf e en_ 3[5I.'""
l'a'
. ;L/5fea liac.
_ m 3 de=* liae
@c.___
m -. 1. ,. . u ' = - no Plant 3/s e
~
Four Loops 2/ steam line 2/ steam line Operating a line u> @ ### Q Three Loops 3/ operating
~
am 2/ opera 16
- Operating ste line any steam line o' operating g
steam line
TABLE 3.3-3 (Continued) , lG < f' w ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE OF CHANNELS TO TRIP -OPERABLE MODES CTION FUNCTIONAL UNIT SAFETY INJECTION, REACTO RIP, FEEDWATER ISOLATION CONTROL 00M ISOLATION, START DIESEL GENER CONTAINMENT COOLING FANS AND RS ESSENTIAL SERVICE WATER (Continued) ii) Three Loop Plant Three Loops 3/ steam li 2/ steam line 2/st a line 15* Operating twice and 1/3 M s am lines
.w .
2/ operating Two Loops 3/ operating 2###/ st 16 7 M Operating steam line line t 'ce steam line in ~ her o rating team line
~
- f. Steam Flow in Two 1, 2, 3 Steam Lines-High
_i) Four Loop Plant
~
Four Loops 2/ steam line 1/ steam l'ine 1/ steam line 15*
@ Operating any 2 steam lines h Three cops 2/ operating l ###/any 1/ operating 16 Op ting steam line operating steam line e.,. steam line _
5
TABLE 3.3-3 (Continued) 4 v' ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION
/
MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION SAFETY INJECTION, k' ACTOR IP, FEEDWATER ISOLATION, CONTROL OM ISOLATION, START DIESEL GENERAT CONTAINMENT COOLING FANS AND ESSEN L SERVICE WATER (Continued) ii) Three Loop Plant Three Loops 2/ steam line 1/ steam line 4/ steam line 15* y Operating ny 2 steam , y li s F Two Loops 2/ operating l### a 1/ operating 16 U Operating steam line erating steam line i steam line Coincident With Either 1,2,3 ## T,yg--Low-Low i) Four Loop Plant 15* Four Loops 1 T,yg/ loop 1 T,yg any 1 T,yg any Operatin 2 loops 3 loops Thr Loops 1 T,yg/ l ### T,yg in 1 T,yg in any 16 m m
"#^t'"9 operating any operating two operating m loop loop loops
$19
\
\ /
Z O E K M d 40 W LD
& e - e e U
w
.a %
- LC O %
<W M UC =
mQ N
.JK Z Q.
Q Q. M < m m H
< h o C. C.
o
>=
Z C C C LaJ rU m m & E h G C. G3 D C C C) L o 6 m 2: M LLJ r0 e C 3 o aNC H EJJ m e m- m e M D LLJ CI3 C1 C. C +J m m h *J Z EZ< > o > ro as M eC rc L rg 6 M M Zm r0 C r0 L L Z < uJ m >= Q C. h Q. GJ Z M Z Q. C. C C C.
^ w EUC N - Q - ac -e o T P=
W W 3 >= GI C W L
- h C - m 3 L
+2 Z C *e +J G Q. m as C.
C O to to L o m a. o O M M CL 3 0 4) O C U >== J Q. Om > c m- 6 - V 4 LLJ M > C. f5 C. m C. A 3 ZM t0 0 47 N C G
- M >== Z >= H % C. L % rU C I O < - % hC Q. h % e m 4 = 0 % o C% C +J
. UH - N - - - rc me tc M uJ m
LaJ D C. J >= 0 CI3 4 M Q. O N N
< LaJ =1 C O
- 43 L
4; L C1
>= 6 O LLJ ZZ w r"*) 3 3 C
>= Z N N C o me
>= J< C3 C) - m m +J LLJ <Z > > +J GS G) to e L HU #0 r0 to L C. L L C.
< O >= >= L C. O C. GD C W >= 1 C) O Q. O C w e C. me e oe C ^ Q w V E m 47 3 w o 3 o t&J OD= C +J J Z %< e C I +J M a-a Q:: *J #O C C 1 C '
E J LLJ C C - L rc Z >-- C Z Z C Q. 3 - LaJ CE: LLJ < U m m C m M >= C V C. C. m C. CZ LA o o C1 m C3 0 c. C. 9 o C1
>- O J Z M C O C C. C L C o O C U U uJ < uJ J J e oe Q O o- Je
< to 6 >- 4 0 +J l J +J 4
= LaJ < C) C r0 J r0 CJ rU C #U m #'Z M C <*C LLJ GJ CJ L L .C
+2 e C L 3
L L 3 O cL C) OCZ L L Q O U
=m e-a LLJ C .C c. 3 C. == J o o c. c.
Z >= H - J U >- .H Q >- O 3 LL. 6C H 6- o< J<C> mom E M +J ro 2 >- O >- U M ^ C CJ D UWM w *- Q +J ^ LaJ w >= L/> - "O M e
=Z e J e3 U < Z d Z LLJ J U Z mWGZ< C o P.- - Z m e M >-<>-->= U 0
>= >.- 2 < < =
U La.J c I >= LaJ
= u.waZm -
D <weoe O L. A M L M U LaJ
-W-STS 3/^ 3-20 SEP 151981 i
e i, - , - - - - - .. - - _ . ._. _.,,,,, , ,.
3 TABLE 3.3-3 (Continued)
'T y ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION 111HIMUM
' TOTAL N0.
CHANNELS CHANNELS APPt! CABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION SAFETY INJECTION, ,"EACTOR TRIP,
-FEED'JATEP ISOLATION, CONTROL ROOM
--MMATI0ti, 5 TART DIE 5EL GENERATORS-
-CONTAIN!!ENT C00LiiiG FAii5 AiiD
-ESSENTI^L SERICC WATER (Conti ved)=
) _T_hree Loop Plant N 1 pressure Three Loops iecc~ e/ 1 pressure Operatin0 loop oops any 2 loo
'# Two. Loops 1 pressure MH essure 1 pressu 16 T Operating in any oper- any operating
% ating loop loop
- 2. CONTAINMENT SPRAY
- a. Manual . tmh'aFon 2 1 with 2 1, 2, 3,'4 W l5-2 coincident switches
- b. Automatic Actuation 2 1 2 1,2,3,4 N 11 Logic and Actuation ,
Relays
- c. Containment Pressure-- 4 2 3 1,2,3 W 13 g 44gh "! h0 H i -3 0 3. CONTAINMENT ISOLATION
; a. Phase "A" Isolation m 1) Manual TE4Eh' ort 2 1 2 1,2,3,4 M I 6' ~
- 2) Safety Injection See 1 above for all Safety Injection initiating functions and requirements 1
TABLE 3.3-3 (Continued) ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION
]
MINIMUM TOTAL NO. CHANNELS CllANNELS APPLICABLE FUNCTIONAL UNIT OF CilANNELS TO TRIP OPERABLE MODES ACTION
, CONTAltlMENT ISOLATION (continued)
- 3) Automatic Actuation 2 -
! 2 1,2,3,4 M it Logic and Actuation Relays
- b. Phase "B" Isolation
- 1) . Manual Iniblion 2 1 with 2 1,2,3,4 M if 2 coincident ,
a switches . Y
- 2) Automatic Actuation Logic and Actuation 2 1 '2 1,2,3,4 M 11 g Jtf$ Relays
~
- 3) Containment 4 2 3 1, 2, 3 1% 13 Pressure- "!4 Hi-6"!gh
- c. Purge and Exhaust Isolation
- 1) Automatic Actuation 2 1 2 1,2,3,4 M 14 Logic and Actuation Relays
- 2. I
- 2) Containment Ond=<.Purg 4 % 1, 2, 3, 4 M, t 'i Radioactivity-liigh .
m 3) Safety Injection See 1 above for all Safety Injection initiating functions and q requirements CJ1 VD
TABLE 3.3-3 (Continued) ENGINEERED SAFLTY FEATURE ACTUATION SYSTEM INSTRUMENTATION MINIMUM TOTAL NO. CilANNELS CllANNELS APPLICABLE FUNCTIONAL UNIT OF CilANNELS TO TRIP OPERABLE MODES ACTION
- 4. STEAM LINE ISOLATION
- 2. 2.
@ a. Manual In{4&4im
'Usteam line 1/ steam line 4/ operating 1,2,3 24 Z0 steam line
- b. Automatic Actuation 2 1 2 1,2,3 29 IS Logic and Actuation Relays
- c. Containment Pressure--
High "igh Hi-2. 4 2 3 1,2,3 M 13 h E 1d e e M. M'1bM..
.: :*o*w'-I"r*'P"55""-
Stcom L;r.c;--!!igh "e Rede1ls6'3 /56h.oel Ac.gfsg
.y su g;,,
j;, i , 2, 3 y %g N'"In '[* nnn5 fan ~t 35 kae 2jsb liny ay sh imc gf,h,, l[.e I2i3 i 12
- e at ng i lines -
Three Loops ng 1 #/any 1/ operating to Operat' steam line operating steam line steam line
%) -Three L00p Plant ,
' N an 1. oops 2/ steam line 1/ steam line 1/ steam line JL Mo Operating any 2 steam __
'_" Two Loops a ing l ###/any 1/ operating % ig Oper
~
steam line operating steam line
~ -
steam line
TABLE 3.3-3 (Continued) y ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION MINIMUM ~ TOTAL NO. CHANNELS CilANNELS APPLICABLE FUNCTIONAL UNIT OF CllANNELS TO TRIP OPERABLE MODES TION STEAM LINE ISOLATION (Continu Coincident With Either T --L w-L w 1, ,3 avg i) Four Loop Plant Four Loops p- 1T avg any 1 T, ny 15* 1Tavg/l Operating 2 ops 3 ops o s l ### T av, 1, Three Loops 1Tavg/ per- 1T avg in any 16 q Operating ating loop any op atin two operating i loo ops ii) Three Loop Plant Three Loops p 1T avg an 15* 1Tavg/l 1 T,yg any
@ Operating 2 loops 2 loops Two loops Tavg/oper- l ### T avg 1T avg in any 16 Operating .
ating loop in any oper-operating loop ating loop
$o w
t,,11
g N O *
- h I 5 6 5 1R T
C 1 1 1 22 A E L BS AE 3 CD IO , , LM 2 2 2 N P O P , , , I A 1 1 1 T . A n T n g - e rg N i n E s - s s i .e M ep erp ep et np . U R SE
- oo upo reo ro ra uo ur eom g t T MLL l sol sl se hs S UEB s s sp .c 2 N MNA e e2g e2 eo mag I INR r r n r r p t en NAE p pyi py pyo s i M IHP n nt n no /nt ) E MCO 1 j1 aa 1 a ! al ia d T e S u Y e e .
n S - r- m i s r s ur - e t N ep ep ep sep .rg n O ro o ro spo ne o I S uo e o uo eoo epm C T LP sl r l sl r l got
- ( A EI s p s py s 1 U NR e2 g e2 ng .y 3 T NT r # n r # an mng
- C A py# i py# i t an 3 A HO n# n n# nt s i
. CT 1 a i 1 a li a / nt 3 E 2ia R
E U p L T o B A S / /o / / . A E .L e el e e rg n T F OE r r r e NN u ug u un g Y N s sn s si T LA s si s st 2 E AH e rp ra et e rp rrp ea m. F TC t - A O o pr po peo s S TF o e o po / O l 1 p l 1 ol D o E R w E ) o E d L t N e - t n I u e n a nn G n r l oo N i u P ii E t s s s tt n s s p p p g r aa _ o h e p pg og o og ; i o uu C t r o on on o on n 0 t - tt ( i P o o i. L i L Li oi ! a- cc N W e L Lt a ea t e t ea L o T A rl ee AA O t r rr er e er & L nvh cd I n u ue re r re o 0 eeg in _ T e d, o op h p h h p wp P $ GLi t a - A F FO TO T TO TO I ! H a T L i m R mr- mcy I O c a T" aeh oia N S n e E et g t gl U I i t ) ) E T. t ai uoe o S i i N ? SWH AlR L E i I h A N O I N I L B 0 R E U E TT a b T M - C A N U . F 5 y wD w sQ u w ;nB
TABLE 3.3-3 (Continued) . O
" ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION MINIMUM TOTAL NO. CilANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES AC1f0N
(, . FEEDW ATER ZSotA TION
- a. su Ge e,ab w 4c,a teve 1 9/3 6 7e, 2/54 Sc4 al 3N ye,, i, t /3
- hig-u;g
- b. Low ges T., C.;,aa,.a i 4.y ope-rab, 36 re4 in ead ore-a% sN ye, h
<1 1 3 1, 2_ /4 '
i=> W Reack- T if C- Safd Ttideda'on S e e. [ ag '
,c, f,, ,ll 3 pe4 7"Jc'E##
T
'"'18"I 'N -func.4 soss y 4.n L
- e stwe ~, e,&,
CO S Y oa
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 ErtsEEuCY 7% AUXILIARY FEEDWATER o riavat2 4 m 9 s i riab Deben Pa y k 1 h 1,2,3 MN 11 Steam Drtsco Paw 1 1 2. I, ~2. i 3 - li
- b. Automatic Actuation Logic 2 1 2 1,2,3 22 15' and Actuation Relays
- c. Stm. Gen. Water Level-Low-Low w
2 1. Start Motor- q l$ w Driven Pumpga.nd 3/stm. gen. 2/stm. gen. /stm. gen. 1, 2, 3 %* b Tu.<bbse Dn'oes ibp in any opera- in each ting stm gen. operating stm. gen. g d E*'-+ Turbine-Driven Pump n. 2/stm. gen. 2/stm. gen 1, 2, 3 --1F __in any_ ... mow 2 operaunu --eposatinn__ stm. gen, stm. gen
- d. Unuervult-age-RCP-4t-art Turbine-
=Dr ver Pump- i h - - & 1, 2 -- f0*-
m d %. Safety Injection
$ Start Motor-Driven Pumpi - and Turbine-Driven Pump See 1 above for all Safety Injection initiating functions and m requirements in E B Y. Station Blackout Start Motor-Driven Pumps and Turbine-Driven Pump 2 1 2 1,2,3 hlF
.v" TABLE 3.3-3 (Continued)
ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION h MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION
-A'JX!L!?.rJ.' IEE0 WATER ' o..ti .u:d)
Q. Trip of Main I- e eawo ic. Ptapt Start Motor- ~ _ Driven Pumps and Turbine-Drivpn Dog g/pymp ]/pymp ]/pymp 1
]9 8% AUTOMATIC SWITCHOVER TO y CONTAINMENT SUMP y a. RWST Level - Low 4 2 3 1,2,3,4 M 13 i Coincident With
-sentebment-Gump-Level - High & P 1, 2, 3, '! -9
-And Safety Injection See 1 above for Safety Injection initiating functions and requirements
- b. Automatic Actuation 2 1 2 1,2,3,4 K 11 Logic and Actuation Relays en Qh LOSS OF POWER
- 2. 1 16 en a. 4 kv Bus 4/ Bus 2/ Bus 1/ Bus 1, 2, 3, 4 W rd Loss of Voltage 3 4 Av 2. t 16
- b. GHd-Degraded Voltage N/ Bus 2/ Bus 3/ Bus 1,2,3,4 N*
y >- - 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 10 1. ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INTERLOCKS
- a. Pressurizer. Pressure, 3 2 2 1,2,3 M il
@ P-ll
-b Low Lo T avg, -l? - - - -1, 2, 3 -M-b %. Reactor Trip, P-4 2 2 2 1,2,3 MM w
D c. Steeg Genedor 3/sfm.ye4. g/ N yen 2 /sb ycy 1,2,3 J7 Water Level, P-ly i4. y apud47 se e 4 spem^y Y gy Sl~ gen sk- g e n W C.A O_. 0 0
TABLE 3.3-3 (Continued) TABLE NOTATION Trip function may be blocked in this MODE below the P-ll (Pressurizer Pressure Interlock) setpoint.
-Mp4enet4en-may-be41ocked-in-thi c MODE- below--the P-12 (Low-Low T
-4nteHeck}-setpointr avg The che.wicl(3) osacciated wi th the pretccti /c function; de, > <cd frcm the
--out vf aervica Keactor CoolantTuup shc?' be placed in the trippcd mcde.
*The provisions of Specification 3.0.4 are not applicable.
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. 12. ACTION M. 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 inoperable channel is placed in the tripped condition within 1 hour. With a channel associated with an operating loop inope" , he inoperable channel to OPERABLE sta thin 2 hours or
~
e_ast HOT STANDBY the next 6 hours and in at least HOT SHuluuw : N +ha following 6 hours. One channel associat - n operating loop may oe M u or surveillance testing per Specification 4.3.4. m 12 ACTION 'hI With the number of OPERABLE channels one less than the Total Number of Channels, operation may proceed provided the inoperable 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. I' ACTION h - With .less than the Minimum Channels OPERABLE requirement, operation may continue provided the containment purge supply and exhaust valves are maintained closed. I W-STS 3/4 3- SEP 15198k
-- . . = . . - . . - . . _ ~ .
TABLE 3.3-3 (Continued) ACTION STATEMENTS (Continued) 15~ ACTION }9 - 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 at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. ACTION - 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 requirements is met; however, one additional channel may be bypassed for up to 2 hours for surveillance testing of other channels per Specification 4 4.3.2.1.
ACTION - 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 f state for the existing ~ plant condition, or apply Specification 3.O.3. .___ 19 ACTION IS - 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 at least HOT SHUTDOWN within the following r 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. 2.0 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). Ac. TION 21 Withie umbe- of CPERA6LE. cho.nne.h one /ess f/;q,,74, lisnsonan C.kanneb, OPERABLE re.guiremedo be on a} /ea.sf NOT STrwbby suihin 4 hours) however, one channel may he bypassed fe. up h a hea,s Co- sme>L,ce bsL,, pec spec,(,da M , y 3.2 1 f'0Vt btb %e ,tff}}et C ne$ is O PNA /3 tE. N SEP 151981 W-STS 3/4'3-M
TABLE 3.3-4 ENGINEERED SAFETY FEATURE ACTUATION _ SYSTEM TRIP SETPOINTS Total Sensor Functional Unit Allowance (TA) 2 Drift (S) Trip Setooint Allowable Value
- 1. SAFETY INJECTION A. Manual Initiation NA NA NA NA NA B. Automatic Actuation Logic NA NA NA NA NA C. Containment Pressure - High-1 4.2 0.71 1.67 D. Pressurizer Pressure - Low 5 4.3 psig 5 5.3 psig 13.1 10.71 1.69 2 1850 psig 2 1840 psig E. Steamline Pressure - Low 13.1 10.71 1.63 2 585 psig 2 568 psig (Note a)
- 2. CONTAINMENT SPRAY ta A. Manual Initiation NA NA NA 2 B. Automatic Actuation Logic NA NA NA NA NA NA w NA C. Containment Pressure - 4.0 0.71 1.67 5 18.0 psig A High-3 5 19.1 psig
_o
- 3. CONTAINMENT ISOLATION A. Phase "A" Isolation
- 1. Manual Initiation NA NA NA NA NA
- 2. Automatic Actuation NA NA NA NA NA Logic
- 3. Safety Injection See Item 1 above for all Safety Injection Trip Setpoints/ Allowable Values
TABLE 3.3-4 (Continued) (h) ENGINEERED SAFETY FEATURE ACTUATION SYSTEM TRIP SETPOINTS _ Total Sensor Functional Unit Allowance (TA) 2 Drift (S) Trio Setpoint Allowable Value B. Phase "B" Isolation
- 1. Manual Initiation NA NA NA NA NA
- 2. Automatic Actuation NA NA NA NA NA
- 3. Containment Pressure - 4.0 0.71 1.67 5 18.0 psig 5 19.1 psig i
High-3 C. Purge and Exhaust Isolation
- 1. Automatic Actuation NA NA NA NA NA Logic
- 2. Containment On Line Later Later Later 5 2 x background NA Purge Radioactivity -
to High i CE 3. Safety Injection See Item 1 above for all Safety Injection Trip Setpoints/ Allowable Values i va
- 4. STEAM LINE ISOLATION ,
A. Manual Initiation NA NA NA NA NA B. Automatic Actuation Logic NA NA NA NA NA l C. Containment Pressure - 5.2 0.71 1.67 5 4.3 psig 5 5.3 psig High-2 D. Steamline Pressure - Low 13.1 10.71 1.63 1 585 psig 2 568 psig (Note a)
- E. High Steam Pressure Rate 3.0 0.5 0 5 100 psi 5 123 psi (Note b)
TABLE 3.3-4 (Continued) ENGINEERED SAFETY FEATURE ACTUATION SYSTEM TRIP SETPOINTS Total Sensor Functional Unit Allowance (TA) 2 Drift (S) Trip Setpoint Allowable Value
- 5. TURBINE TRIP
- a. Steam Generator Water 4.0 2.18 'l.76 5 86.0% of narrow range $ 87.2% of narrow range Level - High-High instrument span instrument span
- b. Automatic Actuation Logic NA NA NA NA NA
- 6. FEE 0 WATER ISOLATION i
- a. Steam Generator Water 4.0 2.18 1.76 $ 86.0% of narrow range $ 87.2% of narrow range Level - High-High f . ag- instrument span instrument span
- u b. Low RCS Tavg Coincident 4.6 1.12 4,8& 564*F 444,6*F
! 'c With Reactor Trip M . 2_ u; c. Safety Injection NA NA NA NA NA ,II 7. EMERGENCY FEE 0 WATER A. Manual Initiation NA NA NA NA NA B. Automatic Actuation Logic NA NA NA NA NA C. Steam Generator Water 17.0 15.28 1.76 117.0% of narrow range 115.9% of narrow range Level - Low-Low instrument span instrument span D. Safety. Injection See Item 1 above for all Safety Injection Trip Setpoints/Allowble Values E. Station Blackout NA NA NA NA NA
- 8. Automatic Switchover to Containment Sump i A. RWST level - low NA NA NA 2 ll(,, <f y g Sels 2 119, 7~71 SaIS Coincident with Safety See Item 1 above for all Safety Injection Trip Setpoints/ Allowable values i
Injection B. Automatic Actuation Logic NA NA NA NA NA And Actuation Relays (a) Time constants utilized in the lead-lad controller for steam pressure-low are T1 > 50 seconds and T <5 seconds. (b) The time constant utilized in the rate-lao enntrnilor fnr hinh ctaam nracesira rata - cn corona,
.k-.
TABLE 3.3-4 (Continued) _ ENGINEERED SAFETY FEATURE ACTUATION SYSTEM TRIP SETPOINTS Total Sensor Functional Unit Allowance (TA) Z Drift (S) Trip Setpoint Allowable Value
- 9. Loss of Power
- a. 4 Kv Bus, NA NA NA 5 2975 volts with a 5 (later) volts Loss of Voltage 5 (later) second time with a 5 (later) delay second time delay
- b. 4 Kv Bus NA NA NA 5 3325 volts with a 5 (later) volts Grid Degraded Voltage 5 (later) second time with a 5 (later) delay second time delay
- 10. Engineered Safety Features Actuation System Interlocks
- a. Pressurizer Pressure, NA NA NA 5 1950 psig 5 1960 psig Y
P-il
- b. Reactor Trip, P-4 NA NA NA NA NA F
c- 5 fem Genecake" 6ee T&c m 5 above Son S+.< n Gen e4. %+e > L e.. \ T .y SefroOP W+e Level , P-14 ,,1 p tj, wale %lue i l l 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 Feedwate,r Isolation Not Applicable Reactor Tri Not / Applicable Containmen\p(SI)
Isolation-Phase "A" N,ot' Applicable Containment Vent and Purge Isolation / Not Applicable Emergency Fee'd, water Pumps Not Applicable Service Water System Not Applicable Containment Air' Recirculation Fan Not Applicable
- b. ContainmentSpray\ Not Applicable Containment Isolation-Phase "B" Not Applicable Containment Vent andg 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-44eh.grg g
- a. Safety Injection (ECCS) _ 27.0) MA-N-)$ U
- b. Reactor Trip (from SI) < (2.
- c. Feedwater Isolation < (7.0)
- d. Containment Isolation-Phase "A" (17.0) (2 .0)
Containment Ven' {ndPurgeIsolation
- e. ~<
(25.0) /( ^.0) ^'d
^Emm
- f. .=i' peyr ur Feed ter Pumps -< (60.0)
- n. 1
- g. [3acatin' Ser/vice Water System < ( R. /(47.0) g #
- h. EEIOm! CY[1Ya kE2 < /(M.0) 1.
22WnNhh??*%?'??,, L(~.0) u ,, ,. ; , - ~. x90
/ .
i ! , N l l 33 W-STS 3/4 3-M SEP 151981 L
I 1 h /
/
l b TABLE 3.3-5 (Continued)
/
/
l ENGINEERED SAFETY FEATURES RESPONSE TIMES
\ /
INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS
- 3. Pres'surizer Pressure-Low I g
250
- a. Safety Injection (ECCS) - (W ,4) (12.0) M-
- b. Rea\ ctor Trip (from SI) <(.)
- c. Feedwater Isolation (7[0)~
- d. Containment Isolation-Phase "A" /(17.0) "s, . 0)O)
- e. _ Containment Vent and Purge Isolation , - (25.0) ^ 0) N ^'4 ey h mx p' i a rj \Feedwater Pumps f.
/
/ - (60.
n.o e)
- g. -Eccenti:1-Se\rv, ice Water System / <( .0) /(-32-G) h 24d^U" '..3- "A
.o...,-
/
M) /(M
-i- Cent 01 Rec::: I:chtien- [ det Applic 21e
- 4. _ $Ien '
45!EE theN Sta= L- .u m ,
- '28 22 0
- a. Safety Injection (ECC'S) / _ 42-0)(4.)/(-124)ffp
- b. ReactorTrip(fromSI)\ $(.)
- c. Feedwater Isolation k, 1 (7.0)
- d. Containment Isolation-Pha's "A" < (17.0) 7.0) N
- e. Containment Vent and Pdrge Isolation (25.0) /'.". 0) N) * -
- f. h N : @ Feedwater, Pumps (60.0 a
/ %
n.
- g. E::catie! Service. Water System <( )(g.)/(47. 0) h E^NM- "
_' .t . ", M "w' s a.".,-M ~7. 0
.;_ r. - . _ , n.-- , ,-...-
t appucau
% 4taam Flow inSteam Two/Lines - High Coincident with lavgM*-' */
- a. Safe jection (ECCS) 1 (24.0)(4)/ ' . )(5)
/
- b. Reactor Tr ' rom SI) < (4
/
- c. Feedwater Isolati -s(9.0)(3)
- d. Containment Isolation-P "A" / (19.0)(2)/(29.0)(1)
Containment Vent and Purge Iso
~
- e. n (27 0.)(1)/(12.0)(2)
- f. Auxiliary Feedwater Pump $(60.0k
- g. Essential Service .er System
- h. Steam Line ation
_ 34.0)(2)/(\49.0)(1) (3) 1 (9.
- i. Cont ' nt Cooling Fans 5 (57.0)( \' 0)(2)
- j. ntrol Room Isolation Not Applicable ,
'\
W-STS 3/4 3-M SEP J 5 iS3; 34
2) C /
/
TABLE 3.3-5 (Continued) - ENGINEERED SAFETY FEATURES RESPONSE TIMES l INITIATING SIGNAL AND FUNCTION / RESPONSE TIME'IN SECONDS
\ /
Steam Flow in Two Steam Lines-High Coincident with eamsLine Pressure-Low / j/
- a. 'fety Injection (ECCS)
< (12.0)(5) . 0)(4)
- b. Reac rip (from SI) </
- c. Feedwater ion < (7.0)(3)
- d. Contain ent Isolat -Phase "A" (17.0)(2)/(27.0)(1)
- e. Containment Vent and Pu s 1on
- f. Auxiliary Feedwater Pump (25.0)(1)/(10.0)(2)
/ < (60.0)
- g. Essential'Sehvic 6er System !
s (32.0)(2)/(47.0)(1)
- h. Steam Lin ation ! < (9.0)(3)
\- /
- i. Con ment Cooling Fans / <
(1)/(40.0)(2) s / - j ontrol Room Isolat'i,on / Not Applica W i 2. 7% Containment Pressure-fich2! !ich.
- a. Containment Spray \/
(45.0) N/(57.0) N
- b. Containment Isolation-Pila e "B" "E) A (75)
- c. E 'Y ' b l* 00
.C) b%
Steam Generator Water l'evel--High-Hich N - a. Turbine Trip [ < (2.5)
- b. Feedwater I lation ; (7.0) ') _ 7. 0
~7% Steam Generator' Water Level - Low-Lew
- a. I
-Motor-dgi cer. Auxilicry--
. Fee 6 tater P"=s bWar Fe eJMe- P.-f s < (60.0
- 5. Turbjibdri.c 5"isry Fccictcr n;.ps- 1 (6_. 0) -
watne Pese, \ 9% Containment <Racioactivity - High 4
- a. Purge and Exhaust Isolation (25/0) N /(10.0) g
\
'\ \
/
/ \
t N N
-W-STS 3/4 3-h SEP 151981 s 35- \.
~
~2.
~
b TABLE 3.3-5 (Continued)
/
/
/
/ .
G..INEERED SAFETY FEATURES RESPONSE TIMES _EN /
,i
/
INITIATING SIGNAL AND FUNCTION
/
RESPONSE TIME IN SECONDS
\ '
AA. nwai Ltevel Luw vvInvident wius m i m iin.4ci 6 .i u. . v EeVil - higin aAd. C&f Gly Inj dC1 A un-MtitefhdtiC $ditsjivves iG CeLtd 5 6. Tis 5; e
- - . ..y- < /,9 t M_ , / /,9_A_
_, t;
-I 2. 'Gderielt5sc d SCE
-d. Iuiuiht us$ vin AUXilidij
- r cedactar Tu..p- .
< (60 n) c{T3- Station Blackout e < gue,
- a. Awxiliery Feedwater Pu,mps g 1 () .0)
" T-ip of u si
/
credsater.T ug3
= Auv414= y FeedIter T .p5 - Net Applicet:10 ta W Loss of Power
- a. 4.16,k/EmergencyBus s (10)
Undervoltage (Loss of g Voltage)
- b. 4.16 kV Emergency Bus 5 (10)
Undervoltage (Degraded Voltage) f 4
\
\
-W-STS 3/4 3.s9 SEP 151931 3G i
s . h j'
,/ ~
h TABLE 3. 5 (Continueo) TABLE NOTATION -
/
* ~I
/
Diesel generator starting and sequence loading delays included. Response time limicNncludes opening of valves.'to establish SI path and attainment of dischargbpressure for centrif6 gal charging pumps, SI and RHR pumps.
/
/
/
Diesel generator starting and sequenceQoading delay not included. Off-site power available./ Response timeN1imit includes opening of ' valves to establish SI-p'ath and attainment'of discharge pressure for centrifugal charging' pumps. 1
## Diesel generator starting and sequence loading delaysNincluded.
time limit includes opening of valves to estab1'ish SI path and
Response
attainmen ,t of discharge pressure for centrifugal charging pumps. W-STS 3/43-g SEP 151981 l 31
TABLE 4.3-2 er ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION J. SURVEILLANCE REQUIREMENTS w us
. TRIP ANALOG! ACTUATING MODES CHANNEL DEVICE MASTER SLAVE FOR WillCH CilANNEL CilANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE FUNCTIONAL UNIT CllECK CALIBRATION TEST TEST LOGIC TEST TEST TEST IS REQUIRED ' 1. SAFETY INJECTION, REACTOR Ta!P FEED'cMTER ISCLATIC", CONT"0L RSOM ISCLAliGii 5TARI DIESEL-g -GE;iERATOR5, C0iiikumtni CBBt4#G- .
EA"5 AND ESSENTIAL SERVICC-WATE"
- a. Manual Initiation N.A. N.A. N.A. R N.A. N.A. N.A. 1, 2, 3, 4 N.A. N.A. M(1) M(1) 1, 2, 3, 4
- b. Automatic Actuaticn N.A. N.A. Q Lo0ic and Actuation Relays 2 c. Containment Pressure- S R H N.A. N.A. N.A. N.A. 1, 2, 3 y liigh !
/N d. Pressurizer Pressure-Low S R H N.A. N.A. N.A. N.A. 1, 2, 3
- e. I i 2ieYYI5h $ S R H N.A. N.A. N.A. N.A. 1, 2, 3
.Estween Steam Lines--
Hi1TP
- f. Stc= r7 c.,.
a 7mn %, s _g. 49 _ _y,A, .gg _ .g_A__ 4t-A-_ 4-2, 3 @ Liiiua "igh Cciacident-- Wt4>-Ei ther i.
. avg uuo cu o ,ar T-- -R- + -N-A- 41-A-- -N-A- N* 1, 2, 3 e----Stc = Li na. P -R-- tt - -N d 4t-*- N-A- 4t-A - - 1, 2, 3 P-rc ure--Lee-
- 2. CONTAINMENT SPRAY Manual Initiation N.A. N.A. N.A. N.A. N.A. N.A. 1,2,3,4 Mu a. R
- b. Automatic Actuation N.A. N.A. N.A. N.A. M(1) M(1) Q 1, 2, 3, 4
- c. Logic and Actuation a Relays co
~
- c. N.A. N.A. N.A. N.A. 1, 2, 3 Containment Pressure-- S R H "ich-!!!gh us-3,
TABLE 4.3-2 (Continued) 8T ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION i'i m SURVEILLANCE REQUIREMENTS - TRIP ANALOG ACTUATING- MODES CHANNEL DEVICE MASTER SLAVE FOR WHICH CilANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE FUNCTIONAL UNIT CllECK CALIBRATION TEST TEST LOGIC TEST TEST TEST IS REQUIRED - 3. CONTAINMENT ISOLATION
- a. Phase "A" Isolation
- 1) Manual ridk+[oa 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 m logic and Actuation 1 Relays
- b. Phase "B" Isolation V -=. ..
- 1) Manual Insh dt59 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 P re s s u re-4' Hi H3;-- l'i gh
- 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 m Logic and Actuation m Relays
- e. ~ .. . , @
- 2) ContainmenF Radio- S R M(2) N.A. N.A. N.A. N. A. 1, 2, 3, 4 G k
$ ey eud:-ical-Hi0h
- 3) Safety Injection See 1 above for all Injection Surveillance Requirements.
1 TABLE 4.3-2 (Continued) m ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS TRIP - ANALOG ACTUATING MODES CllANNEL DEVICE MASTER SLAVE FOR WillCll CilANNEL CilANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE FUNCTIONAL UNIT CllECK CALIBRATION TEST TEST LOGIC TEST ' TEST TEST IS REQUIRED
- 4. STEAM LINE ISOLATION
- a. Manual Idh'al-ion 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 M N.A. N.A. N.A. N.A. 1, 2, 3 issub llishHZ
- d. N SY N E' h T E itY$$ S R H N.A. N.A. N.A. N.A. 1, 2, 3 M Line;--!!igh Ceincident-With E;Uier-
[ 3 .e 1. T avg
-- Lc.;- Lc.. c r - -S- -R-- -M- -M-A. M -thA. -N-A- -1, 2, 3-go
- 2. Stcom U ne 5 -R- -H- -N-A-- -N-A- -N-A- -4h-A . -1, 2, 3 6 Hmh s[EliY.o N"m., Raf e. s R t1 *4 vd *4 "4 3
- 5. TURBINE TRIP-M!D FEE 0 WATER
-I SOL AT ION--
- a. Steam Generator Water S R H N.A. N.A. N.A. N.A. 1, 2 Level--liigh-iligh
- b. Automatic Actuation N.A. N.A. N.A. N.A. M(1) M(1) Q 1, 2 g gand Actuation Relay lh AUXILIAl1Y FEEDWATER
- a. Manual n iNali*a N.A. N.A. N.A. R N.A. N.A. N.A. 1, 2, 3
$u b. Automatic Actuation N.A. N.A. N.A. N.A. M(1) M(1) Q 1, 2, 3 Logic and Actuation Relays 5, c. Steam Generator Water S R M N.A. N.A. N.A. N.A. 1,2,3 m level--Low-Low of
TABLE 4.3-2 (Continued)
F ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION y SURVEILLANCE REQUIREMENTS TRIP ANALOG ACTUATING MODES CHANNEL DEVICE MASTER SLAM FOR WHICH CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST IS REQUIRED L . FEEpwr1TER ISOLATION
- o. 5 6 Gcou do, wate,Icel g M N4 ^' A NA N4 ' s 2.
S Il$-WQ
- b. Low R es r.y coa,.;a,.d S ({ f1 A//}- /A
/ A/# /t/A- / , L.
W M Reado- n ip g,,, qg g 34 gm,,e t) ,nc, Reg u i,e - e 4+s . g c . Saf at En[c cUos gec l gom 7 g,,g ,,, .c b3 I
TABLE 4.3-2 (Continued) 8f { 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 IS REQUIRED
-AUXILIARE FEE 0 WATER (Continued) scienc. m 's 6 Undi vGltogC "C" M-A. -R-- -N-A:- --R- h -N-A-- -tf-A - +
d s. Safety Injection See 1 above for all Safety Injection Surveillance Requirements P 4. Station Blackout N.A. N.A. N.A. R N.A. N.A. N.A. 1, 2, 3 g Tr4Nf t!aia Ferie ter W-A. -fHb -N-A--- -R-- N-A-- A-- -N-A-- M Pt*PS R
% 9 AUTOMATIC SWITCil0VER TO Y CONTAIN!!ENT SUMP
- a. RSWI Level - Low S R M N.A. N.A. N.A. N.A. 1, 2, 3, 4 Coincident With E-entainment S=p Level 0 -R- -M--- -N-A-- - R -H-A-- 14-A--- -1, 2, 3,T J!igh And-Safety Injection See 1 above for all Safety Injection Surveillance Requirements
- b. Automatic Actuation N.A. N.A. N.A. N.A. ' M(1) M(1) Q 1,2,3,4 Logic and Actuation Relays
- 1. 'l LOSS OF POWER
- a. 4.16 kV E crgency Bus N.A. R N.A. R N.A. N.A. N.A. 1,2,3,4 underveltag { Loss of u> VoltageY
-u - - b. 4.16 kV Emm}ancy. Bus H.A. R N.A. R. N.A. N.A. N.A. 1, 2, 3, 4 m "ndame!tage30egraded g Voltage?
TABLE 4.3-2 (Continued) 'T ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION h b SURVEILLANCE REQUIREMENTS TRIP ANALOG ACTUATING MODES CilANNEL DEVICE MASTER SLAVE FOR WilICH CHANNEL CllANNEL 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. -M-A- -R----- -M - h -N.A.- N.A. -H-k- 1, 2, 3 -
Lv., Lv. Tgg, P b t,. Reactor Trip, P-4 N.A. N.A. N.A. -R N ,9 -Ndr.R N.A. N.A. 1,2,3 NA h0) & 1L3
$ c. Seam Gene %%r S R 11 11 O )
Y 00de.- Level, P-14 MD e-- W
TABLE 4.3-2 (Continued). TABLE NOTATION
-(1) Each. train shall be tested at least every 62 days on a STAGGERED TEST BASIS.
(2) A DIG sTA L C H ANNEL OP2KA T/OsvAL - TEST tuill b e f e'50' ~ @ es%U. instrunientdt'en i 6 W-STS 3/4 3-M SEP 151981
INSTRUMENTATION 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 limit within 4 hours or declare the channel inoperable.
- b. With one or more radiation monitoring channels inoperable, 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 4.3.3.1 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. t I. l - gg W-STS . 3/4 3-44 SEP 151981
m TABLE 3.3-6 S d RADIATION HONITORING INSTRUMENTATION f0R PLANT-GPERAH0ftS-
@ HINIMUH CilANNELS CilANNELS APPLICABLE ALARH/ TRIP FUNCTIONAL UNIT TO TRIP / ALARM OPERABLE H0 DES SETPOINT ACTION
- 1. Containment 1,2.,M4 lafe7
- a. Containment Atmosphere ,
1 2 --AM- d i23 E."/h -8& M Radioactivity-N10h ( ost Local
- b. RCS Leaka0e Detection
- 1) Particulate Radioactivity N.A. I 1, 2, 3, 4 N.A. '23-22.
- 2) Gaseous Radioactivity N.A. I 1, 2, 3, 4 N.A. 2.?-
$ 1 C.Pur0e and Exhaust fE iYah on Criampside- Cra-ne_)
g g 5 Q AiS aR/hr d 2.3 o-s 2xb.ctrqvou
-a. Partici, late .7:dioacti's t ty- 1 2 h 7,'$c*,." &
Ha- S Fea L;ne. ta,pq
- 2. . d__Jaseous mHadioact,1vity 1/(/alve ll/l/afsc -Alh n 31 2 - 2. 5
- 3. SN h EV' [cYYt2r4 e-eedioaetivtty-!! intr Le,c,7 A .2x bacho wad 5
- 4. -6steotte-Radleac t-i > ! ty '
1 'R] '**fIl 1.{4bmR/h -$72.26
- b. - Looe Cri t L.alBi ty Radiat4or:- Lc ;ci 1 il "All ' usan a-8&- 26 1JS=mR/tr ;
- 4. Control Room Isolation t Eas-i- Air Tekkr., cgoo c uto. '
~
p x hac h.q v8"^
- a. -Air-4nteke-Radiat4en4 eve 4- 1/ intake lX/ intake All 1 9 3 ="/h LST;,gpfM
' whacket1Ce West 4trInbelte .
- b. Eontro4--Room-Atansphere- 1[sn}alte, lifiabde. All $f2]:mR/g- #2.1
-Radiat4on4fgh- 5 loo certo *-
4 2 x t,adr*""
~M esevis.S" d
TABLE 3.3-6 (Continued) ACTION STATEMENTS AL,iluN a .J:9 +ba number of OPERABLE channels less than +q ue , ,,,g Channels OPtxao d . + i n :-+ na A m area surveys of the
@ monitored area with_Mnitoring . 2 "nentation at Jav' :... per 24 hours.
X2. I'- ACTION 2(i - With the number of OPERABLE channels less than the Minimum Channels OPERABLE requirement, comply with the ACTION requirements of Specification (3.4.6.1). Acn;;, C t"+k +ho ntimber of OPERA 5LLE channels less than thp Minimum a g Channels OPERABLE requir,e_3RDL.M- 7 ggghs-vr opeui & Cit ~1on (3.9.12).
-.. M req W e-33
- L ACTION M '- With the number of OPERABLE channels less than the Minimum Channels OPERABLE requirement, comply with the ACTION require-ments of Specification (3.9.9).
- 14 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.
25' jsi ACTION 38'- With the number of OPERABLE Channels less than required by the Minimum Channels OPERABLE requirements, initiate the preplanned 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 d.ays of the event, or
- 2) prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within 14 days following the event outlining the action taken , the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status.
l^ , ACT[OtJ 2b'- DJdA Zit. O el 6PERABE M M bN M L M otcRAsw & , f
$ u p w co n % D M I.; W L . rce w J s u)
& I a~%r> bM -
l 47 W-STS 3/4 3-!KL SEP 151981 i L
TABLE 4.3-3 st J, RADIATION HONITORING INSTRUMENTATION FOR Ptt"' d OPER",TIONS SURVEILLANCE REQUIREMENTS E!fGD h CilANNEL H0 DES FOR WillCil CilANNEL CilANNEL OPERATIONAL SURVEILLANCE FUNCTIONAL UNIT CllECK CALIBRATION TEST IS REQUIRED
- 1. Containment
- a. Containment Atmosphere R Pas-t dioactivttv-lif0h Loc 47 S R H' -4tt.1. 2,3 'N
- b. R S Leakage Detectiori .
- 1) Particulate Radio- S R H 1,2,3,4 activity ,
- 2) Gaseous Radioactivity S R H ' 1, 2, 3, 4 Isolah*en R 1. C.. Purge and Exhaust . Vent 41st! er s R P1 5 u (nuspA6 ce4 m 2- D="+iru!:te. Rad!cartiv!ty- -S- -R- --H- -AH-kcu 2. M 2d Es~RakiES tivity S R H' -AH- ! ,2i3 '9 -
- 3. .ftMi,2pakeTaYdeek'Y"
.a--Radicac t ! vit-y-High -
ct. -Ca$$ ads edicactivitr S R H
'**~ A l l
- b. Crl$fra!ty-Radiaticr.Lcvcl- S R H + All
- 4. Control Room Is ola4- to'1
- a. nk[Mih 1-- S R H All
- b. M t oh Ye M N Pphere- '
--Radiet-ion-HitIh-- S R H All
-J/10;LNai A Ufms._
*--Wrtfr-fuel-in-the-fttel-storage-pool-area ---
"*-Witb irradiated fual 4a the fuel storage peal areas
#* e
INSTRUMENTATION MOVABLE INCORE DETECTORS LIMITING CONDITION FOR OPERATION 3.3.3.2 The movable incore detection system shall be OPERABLE with:
- a. At leash 75% of the detector thimbles,
- b. A minimum of 2 detector thimbles per core quadrant 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. Measurement of A or XY ACTION:
- a. With less than the above required detector thimbles but more than 50%
of the detector thimbles available either:
- 1. Perform an evaluation of the operable thimble dis tribution. This evaluation shall consider the effects of the reduced number of detector thimbles on the applicable monitoring or calibration fune-tions and shall determine what, if any, increased uncertainty shall be applied to incore measurements. The results of this evaluation shall be submitted to the Commission in a Special Report 15 days prior to using the moveable incore de tection sys tem for the appli-cable monitoring or calibration functions, or
- 2. Do not use the system for the above applicable monitoring or calibration functions.
- b. Wi th le ss than 50% of the detector thimbles available do not use the system for the above applicable monitoring or calibration functions.
- c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS Tke. prou hio ns of SpectG ca.+5 n it.o.1 as e i;ot o p h ka li/e 3l 4 3 - 4'1
i 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. ACTIONi
- 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 REOUIREMENTS 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,noJinstruments actuated during a seismic event greater than or equal to (4 M Fg shall be restored to OPERABLE status within 24 hours and a CHANNEL CALIBRATION performed within 5 days fo.llowing 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. So W-STS 3/4 3-M SEP ; 51981
C - TABLE 3.3-7 SEISMIC MONITORING' INSTRUMENTATION
- MINIMUM MEASUREMENT INSTRUMENTS INSTRUMENTS AND SENSOR LOCATIONS RANGE OPERABLE
- 1. Triaxial Time-History Accelerographs -
=a=
- a. f-SM-Y T- 760 /~r e r Rcil -End anI~ Z* l9 1k b As- Ltakr
- b. f - SH - x r - 4 70 t co nt. Fi.u ,Jo (,a n r le 4 14
- c. f- SH - X T- 47s n fod. n o. fly _ r. Ia 18 I
- 2. Triaxial Peak Accelerographs
- a. I-S h-Y R */a 7D 2. R eado v Ws Sa,e . n-20 Hg i
- b. I-91 ~ x n-l7o s wek< G1. Ao. 6-ao Hy 3
- c. l~SH - Y R - /,709 PCc1D AsrW o - 2 6 Hu i e
e &
- t. e +
- 3. Triaxial Seismic Switches orTrig e r3
- a. I-Gh- XG -6700 free Reid N4 1*
- b. I-S M .K S -l.2 0 I [s n + . l~eu n A . NA 1*
- c. 1- s h - x S -DO9 [sarf. Cou,rl. NA 1*
- d. 1- G h -YS-fo71 n (nnh Co. f/r. NA 1*
- 4. Triaxial Response-Spectrum Recorders
- a. I~ SM - Y R -470'S" Con +. Fnu nd. /-30 fly 1*
- b. l~691- Y R - /78/s S.G. lib Sy. l-30 Ny 1
- c. I-5h-XR-Dn2 l>ri. A u x . bid . l- 3d Hv ,
1
- d. l- $ fi- YR -l ~)0 R R.U) Pump Hsc. l 3G fu i
.,, +
-f= e
\
( *With reactor control room indication SI W-STS 3/4 3-% SEP 15193i
l l TABLE 4.3-4 l SEISMIC HONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS ANALOG CHANNEL
~ CHANNEL CHANNEL OPERATIONAL-INSTRUMENTS AND SENSOR LOCATIONS CHECK CALIBRATION TEST
- 1. Triaxial Time-History Accelerographs
- a. I- S M - >c r - noo Gee f u e.lf M* R SA 4 Easr <anf. Pm th12. rn & se c. k R y baar I -5 H - X T- f. '7 n I fe d . l~ou nd. b k M* R SA M*
C.% l -s h- k'r' h'7 / 0 Cen.f. 09. f/r. N N R SA
- 2. Triaxial Peak Accelerographs
- a. l-SH-XR - /: 702. Readne l/es . supp NA R NA
- b. l'GH-XR-4~103 Q v ri,ol. Ay . NA R NA
- c. l- 6 h -X R -47M e/ e cc(4/ A ,o r h,,
NA R NA
& W & &
T W + WP i g 3. Triaxial Seismic Switches
- a. I-S H -Ys - /r '700 free fieM ** H R SA
- b. J gm.ys 62Dr (~nn-f. & nel ** H R SA
- c. 1 -m -ys ~4 7/19 fu l . b d. *
- H R SA
- d. t - <H- y s -A J /O fon t. go. p/r. *
- H R SA
- 4. Triaxial Response-Spectrum Recorders
- a. I- SH XR -1705~ (x,0. & n d . ** H R SA
- b. I-GH -yQ - A ]DL .gr,, 1;g s u mf* NA R G Alh .
- c. 1 c h - YR -D0'l Pri . A u y . kdq NA R W NA
- d. I ~ % n - Y R -47D9? q gj. Amp Nse. NA R % Allt e 1R + W 4 .
Wr & W
^Except seismic trigger
**With reactor control room indications.
1 52-W-STS 3/4 3-S cep l ' g..
' :1. '
INSTRUMENTATION METEOROLOGICAL INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.4 The meteorological moni.toring instrumentation channels shown in Table 3.3-8 shall 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. -hmeteorological monitoring instrumentation channels inTale3 3-2 shall be demonstrated OPERABLE:by th; performance nf * " c u ^ "" c ' """'" 'nd
-GHANNEt:= tat 49 RAT!n" ;craticr.% t the f equenciac chnun in Tabic 1. 1 5.
O g it*S** N <. 8 9 0- U" 01 C 6t ArMN E.t, c.htECK, M
@ a.
- b. 03 Vev6' * (* **
- Se*" o W Ct*d^>at cai,tsterroo.
1 53 W-STS 3/4,3-h8- SEP 151981
C TABLE 3.3-8 METEOROLOGICAL MONITORING INSTRUMENTATION MINIMUM INSTRUMENT LOCATION OPERABLE
- 1. WIND SPEED
- a. I-owu Lese.I, Nominal elev. L/.s 5t- 1
- b. Ut>per Lcv e,( , Nominal Elev. 2 0'1 f r 1
- 2. WIND DIRECTION
- a. lnw e< Leve.l Nominal Elev. 93 W 3
- b. (Agg<r Levc/, Nominal Elev. 204 4- 1
- 3. AIR TEMPERATURE - DELTA T
- a. Lowu Lue/, Nominal Elev. f3 ff- /S0 Ef 1
- b. U ur deve/ , Nominal Elev. Y3 ce-Jof ft 1
/
V
\
f4 W-STS 3/4 3-M SEP 151931
TABLE 4.3-5 METEOROLOGICAL MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL CHANNEL INS UMENT CHECK CALIBRATION .
- 1. WI SPEED
- a. N inal Elev. 43 -f P D SA
- b. Nomi al Elev. 9 69 [f D A
- 2. WIND DIRECTI .
- a. Nominal Ele Y3 N D SA
- b. Nominal Elev. 7 69 ~ f /~ D SA
- 3. AIR TEMPERATURE - DELTA
- a. Nominal Elev. W [t-/ ff- 0 SA
(
- b. Nominal Elev. YJ f/-> n 9 D SA I
1 l l W-STS g SEP 151991
, INSTRUMENTATION REMOTE SHUTDOWN INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.5 The remote shutdown monitoring instrumentation channels shown in Table 3.3-9 shall be OPERABLE with re a douts displayed external to the control room.
APPLICABILITY: MODES 1, 2, and 3. 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 REOUIREMENTS (i~nTable 3.3-9 ) 4.3.3.5 Each remote shutdown monitoring instrumentation channe 9shall be demonstrated OPERABLE:by perfe--ance af +he CH.^.PNEL C"ECK.and CP^NNEt GAMBRAHO" cparatier.: at the ' req"crcies threa 4- hb! " 3-57 d.. Eveq 31 da.3s by eufovdaue af a c HiM" vet cureg, ut ,
- b. Evcey 18 +ae.dtu by per"or%e,tc<
r of a cHAN"EL Gall 8R4Tl04) N
- M C"^" C d 8 C dd $ounc Range neudmn h mdyumednYtbn b ndt'
; YC p uir e E -ko kade C H R AINEL CD U 6 R ATl0N op e tKho u .
W-STS 3/4 3-et SEP 151981
- m
. .. e
~
TABLE 3.3-9
@ REMOTE SiluTDOWN HONITORING INSTRUMENTATION TOTAL NO. HINIMUM READOUT OF CHANNELS INSTRUMENT LOCATION CHANNELS OPERABLE
- 1. Intermediate Range Neutron Flux Ab CP-109 4 +B
- 2. Source Range Neutron Flux c.P-to8A*B 5h 4 0
- 3. Reactor Coolant Temperature - CP-10F A 'B [, j pp Aver % e. +- -
4 Stears Gess, - CP-tog A*B p9 E.%cqeac.y t:ceboates. Flow , 9 9 0 FA Pressurizer Pressure cP-ior 468 m N
#P-3884'8 P p Pressurizer Level 7,g
$. 6 1 M 7% Steam Generator Pressure CP-198A66 g oo Il Steam Generator Level CP-1094 68 '+-
o
"+
T Bonc Acid Ltt Level C P-108 4 + B 4-O tT O
& 9
TABLE 4.3-6 REMOTE SHUT 00WN HONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CilkNNEL CilANNEL Ill5TRUMEllT CHECK CALIBRATION
- 1. Intermediate Range Neu on Flux H N A-
- 2. Source Range Neutron Flux H 4
- 3. Reactor Coolant Temperature - live e H R
- 54. Pressurizer Pressure H R 4L Pressurizer Level H R Steam Generator Pressure 'R
'l '6.
" H R Steam Generator Level a, 8 'A. N H R 84 Ste % Gene % 6 . Eme-$ emy Feel,de. Flow G
. . . ... g , s. , ... .
. . . . ....g..
3
'e. -
O e 0
INSTRUMENTATION ACCIDENT MONITORING INSTRUMENTATION LIMIiiNG CONDITION FOR OPERATION 3.3.3.6 The accident rr.onitoring 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.
SURVEILLANCE REQUIREMENTS hTal,le 3.'5 -19 4.3.3.6 Each accident monitoring instrumentation channelfshall be demonstrated OPERABLE by-performance ef the -CHANNEL C"CCK ar+GHANNEL -CAL-IBRAHON-eperaticns et4he-f requenc4es-shown--in Tabic 4. 2-7. a.. E uevy 3 1 A q s by perfo%uc < g w CHAMNEL CHECR, a nd
- b. Eveg l S' %cntiu b y Pedrovmuet of A Cli49a EL Ca u B A.r} Tid ^l-O 57 W-STS 3/4 3-e4 SEP 151981
TABLE 3.3-10
'T y ACCIDENT MONITORING INSTRUMENTATION REQUIRED MINIMUM NO. OF CHANNELS INSTRUMENT CllANNELS OPERABLE
?
- 1. Containment Pressure ,
1 1
- 2. Reactor Coolant Outlet Temperature - Til0T (Wide Range) 2 1
- 3. Reactor Coolant Inlet Temperature - TCOLD (Wide Range) 2 1
- 4. Reactor Coolant Pressure - Wide Range 2 1
- 5. Pressurizer Water Level 2 1 EO 1/st m generator
- 6. Steam Line Pressure -- 0 , 2/ steam ge erator s p Steam Generator Water Level - Narrow Range 1/ eam generator
[ 7. o c [ 1/ steam gene ator s p_
%% 8. Steam Generator Water Level - Wide Range n ,,
1/ steam genera or / steam generator R 7
- 9. Refueling Water Storage Tank Water Level
(( 2 1
- 10. Boric Acid Tank Solution Level [ 2 1 Em ugesc1 +
- 11. Auxil ary Feedater Flow Rate Pg 2/ steam gener or 1/ steam generator 3
Reactor Coolant Systen Subcooling Margin Monitor D
- 12. g 2 1
- 13. PORV Position Indicator b' 2/ Valve 1/V lve
- 14. PORV Block Valve Position Indicator
$IG- 2/ Valve 1/ Val E
m 15. Safety Valve Position Indicator o [,- 2/Valv 1/ Valve c-m 16. Containment Water Level (Harrow Range) a 2 1 $ 17. Containment Water Level (Wide Range) 2 1
- 18. In Core Thermocouples / core quadrant 2/ core quadran
TABLE 4.3-7 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS
]
CHANNEL CHANNEL INSTRUMENT (Il trational Only) CHECK CALIBRATIO
- 1. Containment Pre ure 1
- 2. R Reactor Coolant Outle Temperature - THOT (Wide Range)
- 3. Reactor Coolant Inlet Temp ture - TCOLD (Wide Range) M R
- 4. Reactor Coolant Pressure - Wide ge M R
- 5. Pressurizer Water Level M R
- 6. Steam Line Pressure R E 7. Steam Generator Water Level - Narrow Range M R
, 8. Steam Generator Water Level - Wide Range M R
- 9. Refueling Water Storage Tank Water Level M R
- 10. Boric Acid Tank Solution Level M R Em er
- 11. Aux"gencyury Feedwater Flow Rate M R
- 12. Reactor Coolant System Subcooling M gin Monitor M R M R
- 13. PORV Position Indicator
- 14. PORV Block Valve Position dicator M R cn 15. Safety Valve Positio ndicator F 1
- 16. Containment Wa Level (Harrow Range)
\
m 17. Containme Water Level (Wide Range) M F 19
- 18. In r e Thermocouples M i t Nx s
NSTRUMENTATION C RINE DETECTION SYSTEMS LIMITING ONDITION FOR OPERATION
/
3.3.3.7 Two i ependent chlorine detection systems, with their afarm/ trip
/
setpoints adjust to actuate at a chlorine concentration of lest than or equal to 5 ppm, s 11 be OPERABLE. . APPLICABILITY: ALL DES ACTION:
- a. With one chlorin detection system inoperable restore the inoperable detection system OPERABLE status within 7 days or within the next 6 hours initiate a maintain operation of he control room emergency ventilation system'1 the recirculation m de of operation.
- b. With both chlorine det tion systems i operable, within 1 hour initiate and maintain o ration of t control room emergency ventilation system in the ecircula on mode of operation.
- c. The provisions of Specificat n 3 0.4 are not applicable.
SURVELLIANCE REQUIREMENTS 4.3.3.7 Each chlorine detect system shall be de nstrated OPERABLE by performance of a CHANNEL CHE at least once per 12 urs, an ANALOG CHANNEL OPERATIONAL TEST at least ce per 31 days and a CHANN L CALIBRATION at least once per 18 months. I W- TS A/4 1 67- SEP 151C8
INSTRUMENTATION FIRE DETECTION INSTRUMENTATION LIMITING CONDITION FOR OPERATION
-7 3.3.3.4 As a minimum, the fire detection instrumentation for each fire detection zone 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 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 contain' ment, 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 4.3.3. 1 Each of the above required fire detection instruments which are accessible during plant operation shall be demonstrated OPERABLE at least once per 6 months'by performance of a TRIP ACTUATING DEVICE OPERATIONAL TEST. Fire detectors which are not accessible during plant operation shall be demonstrated OPERABLE by the performance of a TRIP ACTUATING DEVICE OPERATIONAL TEST during each COLD SHUTDOWN 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.
; 4.3.3. 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.
, 6'l W-STS 3/4 3-b8 SEP 151981 v w. -
-y- - - - - - - . + - -
y , . , _ . , , , _ __ , _ , ._.,__p, _ _ _ y ,_ _ _ , . - _ _ - , , . _ . _ _ _ _ ~ _ _ _ _ , , _ . _ , - - y
TABLE 3.3-11 FIRE DETECTION INSTRUMENTS INSTRUMENT LOCATION __ MINIMUM INSTRUMENTS OPERABLE
- HEAT FLAME SM0KE
- 1. Containment Zone 1 Elevation ~
Zone 2 Elevation Ots idmaf[v6 b b <-
- 2. Control Room
- 3. Cable Spreading
. suppIies a.+ a late aa4e-Zone 1 Elevation Zone 2 Elevation ,
- 4. Computer Room
- 5. Switchgear Room
- 6. Remote Shutdown Panels
- 7. Station Battery Rooms 4 Zone 1 Elevation Zone 2 Elevation
- 8. Turbine Zone 1 Elevation 4
Zone Z 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.
i t ** List all detectors in areas required to insure the OPERABILITY of Safety related equipment and indicate instruments which automatically actuate fire suppression systems, p W-STS 3/4 3-BG SEP . 51981
/
/
INSTRUMENTATION / LOOSE-PART DETECTION INSTRUMENTATION ,
\ ~/
LIMITING CONDITION FOR OPERATION g ,I 3.3.3 4 Th se part detection system shall be OPERABL .' APPLICABILITY: MODES 1 and 2 / ACITON: f
,i .
- a. With one or mores loose part detection sy' stem channels inoperable for more than 30 days',sprepare and submit a Special Report to the Commission pursuantqo Specification'6.9.2 within the next 10 days outlining the cause ofs the malfunction and the plans for restoring the channel (s) to OPERABLE status /
b.
\
The provisions of Specification /s 3.0.3 and 3.0.4 are not applicable. Y. SURVEILLANCE REOUIREMENTS / \ 4.3.3.'R y Each channel / of the loose part systemi detection \ shall be demonstrated f OPERABLE by performanc <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.
\
\
s\
\
\
s bl W-STS 3/4 3-)0 1981 NOV 2
INSTRUMENTATION RADI0 ACTIVE 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 00SE CALCULATION MANUAL (ODCM). APPLICABILITY: At all times. ACTION:
- a. With a radioactive liquid effluent monitoring instrumentation channel alarm / trip setpoint less conservative than required by the above Specification, immediately suspend the release of radioactive liquid effluents monitored by the affected channel, or declare the channel inoperable. .
- b. With less than the minimum number of radioactive liquid effluent monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 3.3-12. Restore the inoperable instrumentation to OPERABLE status within the time specified in the ACTION or, " 'i= c' a
-Licera;; Dent Repert, 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, r.d 6.9.1.9.5 are not applicable.
SURVEILLANCE REOUIREMENTS 4.3.3. Each radioactive liquid effluent monitoring :nstrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION and CHANN"L IUNCTICNAL TC:T operations at the frequencies shown in Table 4.3-TS* f WA/ Lob dHWNSL. dPCMrMAML TES T S 62. PWR-STS-RETS 3/4 3'M. 1/4/83 t
TABLE 3.3-12 A - T RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION ? @ g MINIMUM CHAllt1ELS d IllSTRUMENT OPERABLE ACTION
- 1. RADI0 ACTIVITY MONITORS PROVIDIllG ALARM AND AUTOMATIC TERMINATION OF RELEASE
- a. Liquid Radwaste Ef flucat Line Tesf Tad / Oisc/rarpe 1 -s8 27
- b. Steam Generator B1owdown-EH4uent-t-me Flasl i fa,ik Drash 1 29.29 O
Turbine Building ff4 oar Daains-} Sumps Ef fluent Line 29,19
- c. 1 m .. RADI0 ACTIVITY MONITORS PROVIDING ALARM BUT 110T PROVIDING g , IATIC TERMINATION OF RELEASE
- a. Service S stem Effluent Line 1
^k>G Effluent Line 30
- b. Component Cooling Water
% CONTINU0US COMPOSITE SAMPLERS AND SAMPLER FL
- a. Steam Generator Blowdown E Line 29 (alternate to item .
- b. T uilding Sumps Effluent Line (alternate to item 1.c) 1 M FLOW RATE MEASUREMENT DEVICES
- a. Liquid Radwaste Ef f!as tine 7esi Iad DEc/tarpe 1 3 29 .
- b. Steam Generator Blowdown EfIlueat-t4ne Flos/ Thd DminM I >L 19 h w
- c. 04scIurge e ," H C L m la h' $ Wae. Ots%c'1 --l-M4 EN4 10 "
a,1 em cavves n.e
. . ...s a%Ided
% .edwde r r s . . . . . + 1. .. .
l TABLE 3.3-12 (Continued) RADI0 ACTIVE LIQUID EFFLUEtlT M0tlITORIllG It1STRUMENTATI0tl = T HittIMUM A CilAtitlELS d OPERABLE ACTI0tl It4STRUMEllT S. RAD 10 At t i v ii'. " cnonrns*
- a. Liquid Radwaste Effluent Line 1 28
;^
- b. Steam Gener uuwn Effluent Line 1 d N
- Required only ' larm/ trip set point is based on recorder-controller.
0
~
e
TABLE 3.3-12 (Continued) TABLE NOTATION A2 ACTION M - With the number of channels OPERABLE less than-rcquired 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 .
5 At least two technically qualified members of the ec%c 'hon
- b. ' i ty -
Staff independently verify the release rate calculations and discharge line valving; Otherwise, suspend release of radioactive effluents via this
, pathway.
ACTION'2% - With the number of channels OPERABLE less than requir:d by- the Minimum Cnannels 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 00SE EQUIVALENT I-131.
- b. At least once per 24 hours when the specific activity of the secondary coolant is less than or equal to 0.01 microcurie / gram DOSE EQUIVALENT I-131.
With the number of channels OPERABLE less than required by th Minim .. OPERABLE requirement, effluent a a this pathway may co '
.. uo o rovided that, at least once per 12 ho sam 11ected and analyzed for radio *' .y at a lower limit of detection t
- ..icrocurie/ml.
~
1 29 ( ACTION K - With the number of channels OPERABLE less than requiced Ly-the Minimum Channels OPERABLE requirement, effluent releares via this pathway may continue for up to 30 days provided the flow rate is estimated at least once per 4 hours curing actual releases. Pump performance curves generated in place may be used to estimate flow. nT ::" " - With the number of channels OPERABLE less than required by + m un.~. "~1< OPERABLE requirement, effluent c '. x via this pathway may continue vor - _ :: , 4 orovided the radioacti 4 *; ' ei is cetermined at least once par , o - _ -
.4 9 actual releases.
PWR-STS-RETS' 3/4 3 ~% 1/4/83 Vf
L A N OT I S TE CT N U
)
1 ( Q
)
1 ( Q
)
1 ( Q [)(2 Q Q F S T N E M E R N I O U LI Q ET E NA R NR ) 3 )3 Q ) ) LAB 3 3 E I ( ( (j ( i - C CL R R R ;
. R R N A A C L
L I E V R L . . U EE S NCK . A. NRC P M M M M N N N AUE O l iOl l I CSC T A T N E M 8 ?.h
- R U
L 3 T E
. S NK 4 N NC D D D D D D 0 I AE E l ii l
L G CC B N n A I i T R O a r t T e T n e I N . Dn s p U B e u R E n e O c nL.i m l L i n . M qin S u M RN e n f f P M i L a at T N E h c Ten k" Or. AO LI AT i L E A S t n U N s A c D e GO i . e
. af s! GN n N u L
F NI D .lf i a NI e A f f l F E IT DA I t L FE C r I DR M l u S y S E f f I N = t n IE f R n E ~ D VI a w r VT f r w I OM T .a n o c c O RC E e "L oI s) p m c. U RR d t P d . Q PE ti si w ? PI a M w1 I S T ei o F SA T W A S o m u1 L E RC OI TE l B ( g RM OO y g n E l Be t S ge m V TT e r ne TT S i T ri nt I I A t o in IU l I o ii T NM s t di NA r o S t o C OO a a l L O e o O at o A MT w r i MG t C P r t 0 U d e ut N a MR ee I D YA T E R a n e Bn e YI TD W t n OO CT nt ea e t A IDS G eu II e e I Gn ea R VNA d nl VV , c ne SN r nn IAE i u m if I O i v on UO me i T L a bf TRL pi OM at b CME q e rE C E r mL U el r ARR i t u A R e o NU t a u T 0A L S T 0 S C I O S( T N I LF I F TL E DAO D O NF M A . . . A . . O . . U R a b c R a b C a b R T S N . . I 1 3 I7hBMd i wD CR$ T
Y TABLE 4.3-T3 (Continued) m h h RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS T r"n FUNCTIONAL CilANNEL SOURCE CilANNEL - d CilECK CilECK CALIBRATION TEST INSTRUMENT 2.4. FLOW RATE HEASUREMENT DEVICES T<st Teft DscAaSe_ N.A. Q(2.)
- a. Liquid Radwaste -Ef f'acnt Lin: D (4) R (3)
FlashTid. Dmin
- b. Steam Generator Blowdown Ef" cent Lin: D Gd N.A. R O) 'Q NA
- c. O!xh;rge Canal G@ulafh Qfe Dirda9e Dt44 N.A. -R-N A % N4
, L. "'" INACTIVITY RECORDERS *
, s y
~
[ a. Liquid Radwaste Effluent Line - n_ _N h -- _ x
@ b.
51cALGe.neratw=BTNdown Ef fluent Line D N.A. n O_
+
Ah; fontnnto nn n7ge 3/M 3 -73 ,- Pa~p davves am uN$sje} fo eshmhElow t t 8
TABLE 4.3- (Continued) p TABLE NOTATION (t) Th. DIG 1TA L C H AN AIE L O PER ATIOwA L TEGr sk // be pe,~ formed . The CHANNEL FUNCTIONAL TEST shall also demonstrate that automati ion
+his pathway and control room alarm annunciation occur ny of the folio conditions exists:
- 1. Instrument icates measured 1 ' above the alarm / trip setpoint.
- 2. Circuit failure.
- 3. Instrur indicates a downscale #
ure. 4 nstrument controls not set in operate mode. 2 Th e RN N.OG- C H A NNEL. O PERA TIONA L TEST s % )l he em el. e CHANNEL FUNCTIONAL TEST shall also demonstrate that control m ala. ciation occurs if any of the following conditio xists:
- 1. Instrument indica sured levels above + alarm setpoint.
- 2. Circuit failure.
- 3. Instruaent ind '
- a dooscale failure.
4 rument controls not set in operate mode.
@ The initial CHANNEL CALIBRATION shall be performed using one or more of the reference standards certified by the National Bureau of Standards 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.
b CHANNEL CHECK shall consist uf verifying indication of flow during periods of release. CHANNEL CHECK shall be made at least once per 24 hours on days on which continuous, periodic, or batch releases are made. b] PWR-STS-RETS 3/4 3-M 1/4/83 u
INSTRUMENTATION RADI0 ACTIVE 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. 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 Table 3.3-13. Restore the inoperable instrumentation to OPERABLE status within the time specified in the ACTION or, " lieu cf c-Licearce Ever.t Repert, explain in the next Semiannual Radioactive Effluent Release Report why,this inoperability was not corrected within the time specified.
- c. The provisions of Specifications 3.0.3, 3.0.4 , rd 5.9.1.9.5 are not applicable.
SURVEILLANCE REOUIREMENTS 4.3.3. 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- M. 9 B. PWR-STS-RETS 3/43-74 1/4/83
TABLE 3.3-13 2 - i' RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION T . A MINIMUM CilANNELS d INSTRUMENT OPERABLE APPLICABILITY ACTION
. TE GAS 110LDUP SYSTEM
- a. Noble tivity Monitor -
Providing Ala d Automatic Termination of Rele 1 { b. Iodine Sampler 1 41
- c. Particulate Sampler
- 41 w a. Effluent System Flow Rate
) Measuring Device 1 36 w
- e up er Flow Rate Measurin0 Device 1- 3 I* RAplo4crwe_o.As wAsrs M. -WASTE CAS !!0 LOUP SYSTEM EXPLOSIVE GAS MONITORING SYSTEM (for systems designed h to withstand the effects of a hydrogen explosion) e----HydretgwHeaitar-fAutomaLic chl-)- --l -
" -3 9--
4.% !!yd.sgcr cr Oxygen Monitor (Process) 1 3$ 37_. lASTE GAS Il0LDUP SYSTEM EXPLOSIVE GAS M STEM (for systems not designed to wi i effects of C a hydrogen explosion) # R **
$ a. ilydrogen Monitors (Automatic control, 40, 42 redundant)
- b. Ilydror e en Monitors (Process, 2 **
~WQ
TABLE 3.3-13 (Continued) RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION U 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 1 41
- c. Particulate ampler
- 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 SYST .
t' a. Hoble Gas Act' ity Monitor - Providing ^ omatic Termination of Release 38 R Alarm and 1 cn
- 41
- b. Iodi Sampler 1
- 4
- c. articulate Sampler 1
~
TABLE 3.3-13 (Continued) , E
= RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION
~
nl d HINIMUM CllANNELS k OPERABLE APPLICABILITY ACTION d INSTRUMENT
- 5. :::'"umrHT PURGE SYSTEM (Continued)
~
g d. Flow Rate Monitor f 36 Y_
@m rnow Iiate Monitor 1 LL AtHHt4ARY-BtHtfHHG-VENTI LATIGN -
-SYSTEM- PI A ur v e Arr-
- a. Noble Gas Activity Monitor 1 M 31 m
x
- b. Iodine Sampler 1
- R,33 NN c. Particulate Sampler 1
% 33
- d. Flow Rate Monitor 1
%:30
^ 36 30
- e. Sampler Flow Rate Monitor 1 El STORAGE AREA VENTILATION SYSTEM
- a. Noble Gas Ac i or 1
- 41
- b. Iodine Sampler 1
@ c. Particulate Sampler 1 41 w
^
5 s d. Flow Rate Monitor - 1 36 Monitor 1
TABLE 3.3-13 (Continued) 2 RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION y
"I i
A MINIMUM CilANNELS OPERABLE APPLICABILITY ACTION - d INSTRUMENT
. WASTE AREA VENTILATION SYSTEM
- a. Noble Gas "vity Monitor 1 7
- b. Iodine Sampler 1 41
@ c. Particulate Sampler 1
- 41
- 36
- d. Flow Rate Monitor 1
- 36 m e. Sam w Rate Monitor 1 g 0 tiler EXilAUST AND VENT SYSTEMS
[9 g 3% such as:
--ME-A;; GE;;["ATOR GLGWCOWi; VC;;T SYSTEi4, -
TURBINE GLAND SEAL CONDENSER EXilAUST O 2. "^h!c Cn Auu..ty ::enitcr -1
-*- - at. Iodine Sampler 1
% 33
* % 33 b %. Particulate Sampler 1
- M--
, -d flow "2tc Meafter 1 c, x. Sampler Flow Rate Monitor 1 M 30 t 8
TABLE 3.3-13 (Continued) 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 la mays nvided that prior to initiating the release:
- a. At + two independent sample . the tank's contents are analyz and
- b. At least two techn' qualified members of the Facility Staff indepen y verify release rate calculations and disc valve lineup; Othe e, suspend release of radioactive effluents ' is
, way. N 3D ACTION '31i - With the number of channels OPERABLE less than eeq"4"ad hy- 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 M - With the number of channels OPERABLE less than-rcquir W 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. Liiun Z W th the number of channels OPERABLE less th=a ~g. - .,, - . . c : .
@ Minimum Lnanness ur c n- immediately suspend
"." ":S3 vi raaloactivu effluents via this patnway.
32.- '
,/ ACTION N - With the number of channels OPERABLE less than r;auired by-the Minimum Channels OPERABLE requirement, operation of this WASTE GAS HOLDUP SYSTEM may continue for up to 30 days provided grab samples are collected at least once per 4 hours and analyzed.
within the following 4 hours. n ;T:^" *^ - With the number of channnels OPERABLE one less than r "."4- d i, the Minn.m Wanals OPERABLE ran"4 ~m, vperation of this OE system may <nn+# 1 Tu r up w I' ^ys - With two channels parenble, be in at least HOT STANDBY witnio a ovm. : , t 71 PWR-STS-RETS 3/4 3-M 1/4/83
TABLE 3.3-13 (Continued) TABLE NOTATION 33 ACTIONTA[- With the number of channels OPERABLE less than r; qui :d 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. ACT::" N - With the m mber_of channels OPERABLF nee C euan iequFria oy the Minimum Channetj' 1- " 'St r pauirement, suspend oxygen supply OE- A one recomoiner. % ______ O O I l I
. l l
33 PWR-STS-RETS 3/4 3-b4 1/4/83
9
'. nBLE 4. 3-R
- o RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS
= MODES IN WillCil T SURVEILLANCE g CilANNEL SOURCE CilANNEL FUNCTIONAL TEST CllECK CilECK CALIBRATION REQUIRED d "'
INSTRUMENT ,
. E GAS 110LDUP SYSTEM
- a. Noble ivity Monitor -
Providing Alarn Automatic Termination of Release . P P R(3) Q(1) h b. Iodine Sampler , H.A. N.A. N
- c. Particulate Sampler W -
N N.A.
- d. Effluent System Flow Rate P .. Q Measuring Device w e. Sampler Fln casuring D N.A. R Q
) 9errce b
lb
/. WASTE GAS Il0LDUP SYSTEM EXPLOSIVE GAS MONITORING SYSTEM (for systems designed to withstand the effects of a hydrogen explosion) h a. !!y<lrogcn !!sai tor (htoi..8Lic
-centrc! ) -
-D-- -H-A-- -Q(4)-- :
a_% "ydroger, er Oxygen Monitor D N.A. Q(4) or Q(6) MO) (Process)
. iAS Il0 LOUP SYSTEM EXPLOSIVE .
GAS MON SYil for systems not designed to withsta - ~ ~ __
) effects of a hydrogen explosion) '
__p W ** h a. Ilydrogen Monitors ( Automatic' D control, redundant) '
- b. Ilydr geygen Monitors D N.A. Q(4) or Q(5) M rocess, dual)
~
9 TABLE 4.3-M (Continued) A 7 RADI0ACTICE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS "I - T MODES IN WillCil E CllANNEL FUNCTIONAL SURVEILLANCE CilANNEL SOURCE d CllECK CllECK CALIBRATION, TEST REQUIRED INSTRUMEllT CONTAINMElli PURGE SYSTEM
- a. Nob Activity 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 . . 4. A. i N.A.
$ d. Flow Rate Monitor D ti. A. R Y gu e. . er Flow Rate Monitor D H.A. R Q F% 2. E -AUXhlMY-BEE 0liiG VENT:LAT10i?-
-GY&TEtt- PLAwr VE A7r-
- a. Noble Gas Actvity Monitor D M R(3J Q(2.)
h b. Iodine Sampler W N.A. N.A. N.A. *
- c. Particulate Sampler W N.A. N.A. H.A.
g d. Flow Rate Monitor D N.A. R Q (il k
- g e. Sampler Flow Rate Monitor D N.A. R Q (1) w e
n ,
@ q TABLE 4.3- h (Continued)
I T RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS T l8 CilANNEL MOD - IN WillCil RVEILLANCE d CllANNEL SOURCE CllANNEL FUNCTIONAL TEST INSTRUMENT CllECK CllECK CALIBRATION REQUIRED
- 3. CONDEllSER EVACUATION (TEM ,
- a. Noble Gas Activity Moni r D M R(3) 2)
- b. Iodine Sampler W N.A. N.A. N.A.
- c. Particulate Sampler N.A. N.A. N.A.
- d. Flow Rate Monitor D N.A. Q L ^
3 e. Sampler Flow Rate Monitor D N. R Q lJ c 8 f' o" 4. VENT llEADER SYSTEM I . L .i a. Noble Gas Activity Monitor D M R(3) Q(2)
- b. Iodine Sampler W N.A. N.A. .
i
- c. Particulate Sampler W N.A. N.A.
jN.A.
^
- d. Flow Rate Monitor D N.A. R Q
, e. Sampler Flow R e Monitor D N.A. R Q C t 8
1 TABLE 4.3- (Continued) a 7
@ RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS "I
T A CllANNEL MOD i WilICil d CilANNEL SOURCE CilANNEL FUNCTIONAL ,50 EILLANCE _ STRUMENT CilECK CilECK CALIBRATION TEST / REQUIRED
- 7. FU ORAGE AREA VENTILATION SYSTEM
- a. Noble Gas A '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.
Flow Rate Monitor N.A. ( 2 d. t Q
< , e. Sampler Flow Rate Monitor D N.Ap / R Q 8
- 8. RADWASTE AREA VENTILATION SYSTEM
- a. Noble Gas Activity Monitor D M R( Q(2)
- b. Iodine Sampler W N.A. N.A. N.A f
- c. Particulate Sampler W N.A. N.A. N.
- d. Flow Rate Monitor D N.A. R Q
^
g e. Sampler Flow R e Monitor D N.A. R Q l sm e
TABLE 4.3- (Continued) 2 ? RADI0 ACTIVE GASE0US EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS T g MODES IN WilICll CilANNEL SOURCE CllANNEL FUNCTIONAL SURVEILLANCE d CllECK CALIBRATION TEST REQUIRED INSTRUMENT CilECK 3
- h. 0 tiler EXilAUST AND VENT SYSTEMS such as:
-STE^" CEi;ERATOR CLOW00Wl; VE;;T-SYSTEM, TURBINE GLAND SEAL CONDENSER '
EXilAUST
-c. l chic Cn ".ctivity ;;o.,itor -fP -M-- R(3)- -Q(2)-
RK Iodine Sampler N.A. N.A. N.A $ W w *
' b 's. Particulate Sampler W N.A. N.A. N.A.
N,'a c + a r! ct. oe te uen!ter- -e- 4Hr.- -a - -q-- C. %. Sampler Flow Rate Monitor D N.A. R Q(1) t 5
9 TABLE 4.3-M (Continued) TABLE NOTATION
- At all times.
RADICocTWE GAswASTE SYST2F1 @ ** During WACTE Ct.S HOLOLT SYST:", operation.
@D) The The AivALoc. CH A NNeu oPsRATioivAL TEST s'ha // be P'A **** N -
CHANNEL FUNCTIONAL TEST shall also demonstrate that automatic ' tion his pathway and control room alarm annunciation occurs if of the follow' conditions exists:
- 1. Instrument in tes measured levels ab ' ..e alarm / trip setpoint.
- 2. Circuit failure. '
es a downscale failure.
~
- 3. Instrument i c
- 4. rument controls not set in operate mode.
&(2.) Tke DIGt T;*)L 'C HANNEL cme &T/DNAL r&s7 shall be Peh'n ed-CHANNEL FUNCTIONAL TEST shall also demonstrate that control arm annu 'ation occurs if any of the following conditions exi * .
- 1. Instrumen ~ dicates measured levels ab ne alarm setpoint.
- 2. Circuit failure.
- 3. Instrumen cates a downscale failure.
A nstrument controls not set in operate mode. (3) 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. (.9) 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.
(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.
'll PWR-STS-RETS 3/4 3-M 1/4/83
bec. SG Ca404
. INSTRUMENTATION
\ /
3/4'N3. 4 TURBINE OVERSPEED PROTECTION / LIMITING CONDITION FOR OPERATION / 3.3.4 At least one turb ne overspeed protection system shall be OPERABLE. APPICABILITY: MODES 1, 2, an'dx3. ACTION: , er ed b "- I
- a. With one stop valve or one,conftvlge'icmcrglveper high pressure tur steam inoperable and70r with one re..,_at stop valve or one hne reheat intercept valve'per low pressure lurbine steam s able, restore the inoperab'le valve (s) to OPERABLE' status within 72 hours, or close at le,ast one valve in the affected steams r isolate theturbine/romthesteamsupplywithinthenext6'o s. ju
- b. With the/above required turbine overspeed protection systemsotherwise inope'rable, within 6 hours isolate the turbine from the steam'N upply.
, SURVEILLANCE REOUIREMENTS
/
.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 d 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) 'gh pressure turbine stop val s.
- 2. (Four) high ressure turbine gove or valves.
'3 . (Four) low pres e turbine r eat stop valves. ,
- 4. (Four) low pressure rbi reheat intercept valves.
- b. At least once per 31 days dir t observation of the movement of each of the above valve brough on emplete cycle from the running position.
- c. At least once p 18 months by performance of CHANNEL CALIBRATION on the turbin overspeed protection systems.
- d. At leas once per 40 months by disassembling at least on f each of the ove valves and performing a visual and surface inspect. of l v e seats, disks and stems and verifying no unacceptable flaw or l orrosion.
*18 W-STS 3/4 3-7i NOV 2 1981
l i I 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. This change incorporated as a result of the cold overpressure mitigation analysis. F. This is not required for nor applicable to the Seabrook design. G. There are no loop isolation valves in the Seabrook design. J. This statement added for clarification. Normally, the plant will be passing through Modes 3 and 4, either heating up or cooling down, and would not be in a steady state condition during this time. K. Seabrook design utilizes a variable PORV setpoint for LTOP which is dependent on RCS temperature. Tne 3.2 in2 vent was determined during transient analysis, i L. Added per requirement of Generic Letter 83-37. N. Deleted Table 4.4-3 and stated surveillance requirements in the Specifications. P. This change made to clarify the intent of the LCO. R. A 92 day surveillance interval is unusually short. Heaters are not designed to change drastically in power output over an 18 month time frame. Additionally, this surveillance test is time consuming and involves personnel hazard. S. This Technical Specification is deleted. The Action Statement is covered by the Action Statements in 3.4.1.1, 3.4.1.2, and 3.4.1.3. The Surveillance Requirements will be included in the ISI/IST program under the scope of a licensee maintained and controlled document. T. This surveillance is unnecessary. With the station in MODES I and 2 the operator is constantly observing his control boards which contain alarms, indications, and graphs. Loss of operation of any reactor coolant loop will be immediately obvious to the operator by alarms and indications. Performance of a surveillance of this type is not necessary and only provides additional paperwork and unnecessary diversion of the operators.
U. This Surveillance Requirement is deleted from the Technical Specifications and included in the ISI/IST program under the scope of a licensee maintained and controlled document. V. Limits for Controlled Leakage are removed because controlled leakage is not real leakage but is the flow into the RCP seals. Almost all of this flow is recovered back into the reactor coolant system. W. This table is deleted and will be contained in a licensee maintained and controlled document.
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 loops shall be in operation. APPLICABILITY: MODES 1 and 2.* ACTION: With less than the above required reactor coolant loops in operation, be in at least HOT STAND 8Y within'S$ hours. 6 SURVEILLANCE REQUIREMENT
- ~
- -4
'I' -ep.4.1.1 eratien :ndThecirculeting Obeve -eqef eed -reacter-cecl:nt reacter ccclant at leastIcep; once-per th ll bc 12verified hours. tu be in T
l t l. l l l ? f l l i f, "See Special Test Exception 3.10.4. l i . W-STS 3/4 4-1 MAY 151980 i f
REACTOR COOLANT SYSTEM HOT STANOBY - LIMITING CONDITION FOR OPERATION 3.4.1.2 At least two of the reactor coolant loops listed below shall be OPERABLE with two reactor coolant loops in operation when the Reactor Trip System breakers are closed and one reactor coolant loop in operation when the Reactor Trip System breakers are open:* -
- a. Reactor Coolant Loop [A] and its associated steam generator and reactor coolant pump,
- b. Reactor Coolant Loop [B] and its associated steam generator and reactor coolant pump,
- c. Reactor Coolant Loop [C] and its associated steam generator and r'eactor coolant pump, and
- d. Reactor Coolant Loop [D] and itis associated steam generator and reactor coolant pump. -
APPLICA8ILITY: 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 only one reactor coolant loop in operation and the Reactor Trip System. breakers in the closed position, within 1 hour open the Reactor
~
Trip System breakers. ~
- c. With no reactor coolant loop in operation,. suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required reactor coolant loop to operation.
SURVEILLANCE REOUIREMENTS . l 4.4.1.2.1 At least the above required reactor coolant pumps, if not in l operation, shall be determined OPERABLE once per 7 days by verifying , correct breaker' alignments and indicated power availability. ' 4.4.1.2.2 The reqdited steam generators shall be determined OPERABLE by verifying secondary side water level to be greater than or equal to [17%] at least once per 12 hours. 4.4.1.2.3 The required reactor coolant loops shall be verified in operation and circulating reactor coolant at least once per 12 hours.
*All reactor coolant pumps may be deenergized for up'to 1 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.
%E Sec. TreciAlTest Exceptwo 5)ecicahthan .r.10. '1 W-STS 3/4 4-2
. 7 :: . . . .. . ". - - .
.- ~.
REACTOR COOLANT SYSTEM HOT SHUTDOWN LIMITING CONDITION FOR OPERATION 3.4.1.3 At.least two of the loops listed below shall be OPERABLE and at least one of these loops shall be in operation:" a. Reactor Coolant Loop [A] and its associated steam generator and reactor coolant pump,** b. Reactor Coolant Loop [8] and its associated steam generator and reactor coolant pump,** c. Reactor Coolant Loop [C] and its associated steam generator and reactor coolant pump,** d. Reactor Coolant Loop [0] and its associated steam generator and reactor coolant pump,"*
- e. RHR Loop [A], and
- f. RHR Loop [8].
APPLICA8ILITY: MODE 4.* ACTION: a. With less than the above required loops OPERABLE, immediately initiate corrective action to return the required loops to OPERA 8LE status as soon as possible; if the remaining OPERABLE loop is an RHR loop, be in COLD SHUTOOWN within 24 hours. b. l With no loop in operation, suspend all operations involving a reduc- l l tion in baron concentration of the Reactor Coolant System and immediately to operation. initiate corrective action to return the required loop "All reactor coolant pumps and RHR pumps may be deenergized for up to 1 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.
**A reactor coolant pump shall not be started uith cr.a ac ;cre af tin:
-Eco+ ant-Syste= ccid leg temper:turce -less than er cqual tu [275] F unless Mce 'h the secondary water temperature of each steam generator is less than 50 'F above each of the Reactor Coolant System cold leg temperatures.
g gee. Speclal kt 6cepfios SpecIft/Alion 3 lM ' l W-STS 3/4 4-3 l.
REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS 4.4.1.3.1 The required reactor coolant pump (s), if not in operation, shall be determined 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 secondary side water level te be greater than or equal to [17]% at l least once per 12 hours. 4.4.1.3.3 At least one reactor ecolant or RHR loop shall be verified in operation and circulating reactor coolant at least once per 12' hours. o
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Si- e.T C- Cla_.4L a
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 l
- b. The secondary side water level of at least two steam generators shall be greater than [17]%.
[ APPLICABILITY: MODE 5 with reactor coolant loops filled ***. l l ACTION:
- a. With one of the RHR loops inoperable and with less than the required steam generator water level, immediately initiate corrective action to return the inoperable RHR loop to OPERABLE status or restore the required steam generator water 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 requi' red'RHR (
loop to operation. SURVEILLANCE REOUIREMENTS 4.4.1.4.1.1 The secondary side water level of at least two steam generators when required shall be determined to be within limits at least once per 12 hours. 4.4.1.4.1.2 At least one RHR loop shall*be determined to be in operation and circulating reactor coolant at least once per 12 hours.
"The RHR pump may be deenergized for up to 1 hour provided: (1) no operations are permitted that would cause dilution of the Reactor Coolant System boren concentration, and (2) core outlet temperature is maintained at least 10*F below saturation temperature.
**0ne RHR loop may be inoperable for up to 2 hours for surveillance testing provided the other RHR loop is OPERABLE and in operation.
***A reactor coolant pump shall not be started 4 th m e er me m of the Reacre
-Ecelsa+ sye+e cold 1:; *a par.2turer !cce *han ne age,! ty [275]oc unless the secondary water temperature of each steam generator is less than JTE) 'F above each of the Reactor Coolant System cold leg temperatures.
4 See Specid Ted Exccf on h' SP ec0CicafS$ 310 4 ( W-STS 3/4 4-5
-- _- ,.~.,m ,
- "-- --~' -
. . . . - -.- . . - n:. / . . rL_ . :.- .-~ . _. ~. !
i l REACTOR C00LAf.T SYSTEM COLD SHUTDOWN - LOOPS NOT FILLED j 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.** l APPLICA8ILITY: MODE 5 with reactor coolant loops not filled. -
, ACTION:
- a. With less than the above required RHR loops OPERABLE,'immediately initiate corrective action to return the required RHR loops to OPERABLE status as soon as possible.
- b. With no RHR loop in operation, suspend all operations involving a reduction in baron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required RHR loop to operation.
SURVEILLANCE REOUIREMENTS 4.4.1.4.2 At least one RHR loop shall be determined to be in operation anc circulating reactor coolant at least once per 12 hours.
*0ne RHR loop may be inoperable for up to 2 hours for surveillance testing !
provided the other RHR loop is OPERABLE and in operation.
- . P
**The RHR pump may be deenergized for up to 1 hour provided: (1) no opera- l tions are permitted that would cause dilution of the Reactor Coolant Systtm boron concentration, and (2) core outlet temperature is maintained at least 10*F below saturation temperature. .i W-STS 3/4 4-6
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RE TOR COOLANT SYSTEM ISOLA LOOP (OPTIONAL) LIMITING CON TION FOR OPERATION O 3.4.1.5 The boro concentration of an isolated loop shall aintained greater than or equ 1 to the boron concentration of the o , rating loops. APPLICABILITY: MODES 2, 3, 4, and 5. ACTION: With the requirements of th above specification n satisfied, do not open the isolated loop's stop valv s; either increase he boron concentration of the isolated loop to within th 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 t at least 1 delta k/k at 200 F. SURVEILLANCE REOUIREMENTS
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- 4. 4.1. 5 The boron 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 i \ _-STS 3/4 4 -t'C'! 2 1981
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ACTOR COOLANT SYSTEM ,/ IS TED LOOP STARTUP (OPTIONAL) ' j LIMITING NDITION FOR OPERATION
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3.4.1.6 A reac r coolant loop shall remain isolated unti .
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- a. The isola d loop has been operating on a rec' culation flow of greater tha or equal to gpm for at leas 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
/ \
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 V4' W 1:2! 2 1981
r REACTOR COOLANT SYSTEM 3/4.4.2 SAFETY VALVES SHUTOOWN LIMITING CONDITION FOR OPERATION l 3.4.2.1 A minimum of one pressurizer code safety valve shall be OPERABLE with i a lift setting of 2485 PSIG 1%.* APPLICABILITY: 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. i i SURVEILLANCE REQUIREMENTS l 4.4.2.1 No additional Surveillance Requirements other than those required by Specification 4.0.5. -
"The lift setting pressure shall correspond to ambient conditions of the valve
- - at nominal operating temperature and pressure.
(
'~#2 I93I W-STS 3/4 4- '-
REACTOR COOLANT SYSTEM . 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 STANDBY within 6 hours and in at least HOT SHUTDOWN within the following 6 hours. SURVEILLANCE REQUIREMENTS 4.4.2.2 No additional Surveillance Requirements other than those required by Specification 4.0.5.
"The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.
! i F .hCV 2 1981 W-STS 3/4 4-1Q
-~.
REACTOR COOLANT SYSTEM 3/4.4.3 PRESSURIZER LIMITING CONDITION FOR OPERATION 92 % *f pres su.en 3 er leveI} @ 3.4.3 The pressurizer shall be OPERABLE with a water volume of less than or equal to -(3 55c) cubic feet, and at least two groups of pressurizer heaters each having a capacity of at least (150) kw. APPLICABILITY: MODES 1, 2, and 3. ACTION: p W
- a. QM O PERABLE With one group of pressurizer heaters ireperable, 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 STANOBY t with the reactor trip breakers open within 6 hours and in HOT SHUTDOWN < within the following 6 hours. ,
t
.y SURVEILLANCE REQUIREMENTS 4.4.3.1 The pressurizer water volume shallsbe determined to be within its limit at least once per'72 hours.
4.4.3.2 The capacity of each of the above required groups of pressurizer heaters h ll be verified by measuring circuit current at least once per Og uif~,ostk3sman.sa t m ~ ~;
' l1 1. 3 7 The emergency e-supply for-the-pressuciter-heaters shall be-
' demcactrated-ORERABLE at lead once per-48-months-by-manually transferrmg-
-pcuer frem the-neemal to the emergency--pcwcr supply and energizing th heaters, '
i
, i' m
b NOV 2 1981 W-STS 3/4 4-11
r l l l 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 valvas'shall be OPERABLE. K. APPLICABILITY: MODES 1, 2, and 3. ' ACTION:
?.
c , a. 'With "one or more PORV(s) inoperable because of sicessive seat leakage,
~s t within 1 hour either restore the PORV(s) to OPERABLE status or close the associated block valve (s); otherwise be in at least HOT STAND 8Y within 6 hours and in COLD SHUT 00WN within the following 30 hours.
- b. With one PORV inoperable due to causes other than excessive seat leakage, within 1 hour either restore the PORV to OPERA 8LE status
, or close the associated block valve and remove power from the block J-' valve; restore the PORV to OPERABLE status within the following 72 hours or be in HOT STAND 8Y within the next 6 hours and in COLD SHUTOOWN within the following 30 hours.
- c. With both PORV(s) inoperable due to causes other than excessive seat leakage, within 1 hour either restore each of the PORV(s) to OPERABLE .
status or close their associated block valve (s) and remove power from the clock valve (s) and be in HOT STAND 8Y within the next 6 hours and in COLD SHUTOOWN within the following 30 hours.
- d. With one or more block valve (s) inoperable, within 1 hour:
- 1) restore the block valve (s) to OPERABLE status or close the block
- ._ ___,. valve (s) and remove power from the block valve (s),_cr..close the PORV -
and remove power from its associated solenoidTaTveTali~d 27 agigily the ACTION of b. or c. above, as appropriate for the isolated PORV(s).
- e. The provisions of Specification 3.0.4 are not applicable..
SURVEILLANCE REQUIREMENTS l 4.4.4.1 In addition to the requirements of Specification 4.0.5, each PORV shall be demonstrated OPERABLE 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. i 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 i- the block valve is closed with power removed in order to meet the requirements of ACTION b. or c. of Specification 3.4.4. ! ( 3/4 4-L.
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NREACTOR COOLANT SYSTEM / S j/ 3/4.4.5 STEAM GENERATORS /
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LIMITING CONDITION FOR OPERATION /
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3.4.5 Each steam generator shall be OPERABLE.
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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 Tavg above 200 El SURVEILLANCE REQUIREMENTS
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4.4.5.0 Each steam generator shall be demonstrated OPERABLE by performance of the following augmented inservic'e inspection / program and the requirements of Specification 4.0.5. 4.4.5.1 Steam Generator Samole Selection and Inspection - Each steam generator l shall be determined OPERABLE during shutdown by selecting and inspecting at enerators specified in Table 4.4-1. leasttheminimumnumberofsteamg/ \ 4.4.5.2 Steam Generator Tube Samole Selection and Inspection - The steam generatortubeminimumsamplesize, inspect \onresultclassification,andthe corresponding action required shall be as specified in Table 4.4-2. The inserviceinspectionofsteam/generatortubes'shallbeperformedatthefre-quenciesspecifiedinSpecification4.4.5.3and'theinspectedtubesshallbe verified acceptable per the acceptance criteria o'f Specification 4.4.5.4. The tubes selected for each in' service inspection shall 'nclude at least 3% of the total number of tubes ip'all steam generators; tb t'ubes selected for these inspections shall be se'lected on a random basis except-
- a. Where exp'e/ \
rience in similar plants with similar water chemistry s indicates critical areas to be inspected, then at least 50% of the tubes illspected shall be from these critical areas.
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- b. Theffirst sample of tubes selected for each inservice inspection (subsequent to the p'.eservice inspection) of each st'eam generator sh'all include:
k W-STS 3/4 4- N
i b REACTOR COOLANT SYSTEM
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N / SURVEILLANCE RE0UIREMENTS (Continued) /
\1. meaa e tubes that previously had detectable /
All nonphgged wall penetrations (greater than 20%).
- 2. ,ubes in those, areas where experience has indicated potential problems. /
3.
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A tube inspection (pursuant to Specification 4.4.5.4.a.8) shall be pe'r. formed on each selected tube. If not permit the. passage of the eddy curre,dny nt probe selected for a tubetube does inspection, this shall be recorded ar)d an adjacent tube shall be select'ed and subjected to a tube inspection.
- c. The tubes selec d Table 4.4-2) durin\as the second g each inserviceand third samples inspection may be(ifsubjected required to bya partial tube inspectioq provided: /
- 1. The tubes selected'for these/ samples include the tubes from those areas of the tube she'et array where tubes with imperfections were prev' iou' sly found.
The inspections include
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- 2. se imperfections were e'viousl) portior.s nd. of the tubes where The results of each sample inspection shall be c assified into one of the following three categories: '
Category Inspection Results C-1 Less than 5% of the tota 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 inspe'cted ..e degraded tubes. C-3' More than 10% of the total tubes inspected are degraded tubes or more than 1% of the inspected tubes are defective. Note: In all inspections, previously degraded tubes must exhibit significant (greater than 10%) further wall penetrati'ons to be included in the above percentage calculations. 12- ' W-STS 3/4 4-14 'NOV 2 1981 t
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\REACTOR COOLANT SYSTEM
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SURVEILLANCE REQUIREMENTS (Continued) /
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4.4.5.3 Inspection Frequencies - The above required inservice inspections of steam generator tubes shall be performed at the following frequ6ncies:
- a. The firs nservice inspection shall be performed a ter 6 Effective Full Power Months but within 24 calendar months of initial criticality.
Subsequent in' service inspections shall be perfor'med at intervals of not less than 12 nor more than 24 calendar months after the previous inspection. If'two consecutive inspections ,following service under AVT conditions, not, including the preservice inspection, result in all inspection result,s falling into the C-1 category or if two consecutive inspectionss demonstrate that / previously observed degra-dation has not continued and no additiolial degradation has occurred, the inspection interval'may be extende'd to a maximum of once per 40 months.
- b. If the results of the inservi.ce inspection of a steam generator conducted in accordance with TabIe 4.4-2 at 40 month interval, fall inCategoryC-3,theinspectiofifrequencyshallbeincreasedtoat g
least once per 20 months. Tije increase in inspection frequency shall apply until the subse
< Specification 4.4.5.3.a; th'quent e interval inspections satisfy the may then be extended to acriteria of maximum of once per 40 m n'ths.
c.* Additional, unscheduled inservice inspections shall be performed on each steam generatorfi' n accordance with s'the first sample inspection specified in Table ,4.4-2 during the shutdown subsequent to any of the following conditions:
- 1. Primary-to.- condary tubes leaks (not including leaks originating from tube-to-tube sheet welds) in excess'of the limits of l Speci c'ation 3.4.6.2.
- 2. A seismic occurrence greater than the Operating Basis Earthquake.
CC na%% @ \ 3 A4 1oss-of-coolant accident requiring actuation df the engineered fafeguards (9. x..a,A 4 A' main sr.eam line or feedwater line break. s 13 f;0V 2 1981 W-STS 3/4 4 'Hi
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\REACTOR COOLANT SYSTEM 5
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N SURVEILLANCE REQUIREMENTS (Continued) /
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4.4.5.4
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Acceptance Criteria /
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- a. As used 1 this Specification: /
1.
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Imperf tion means an exception to the dimens, ions, finish or contour ofg a tube from that required by fabrication drawings or specifications. Eddy-current testing indicdtions below 20% of the nominal t,ube wall thickness, if detect'able, may be considered as'icperfections. /
- 2. Degradation mea ba service-induced cr cking, wastage, wear or 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 nom'inal wall thickness caused by degradation.
- 4. % Degradation means the percentage of the tube wall thickness affected or removed by de 'i
- 5. Defect means an imper ction of suche/ gradat
\ severity that on.
it exceeds Oc the -pl ugg&Es'i- limit. A tube contain'i g a defect is defective. ngw
- 6. d3M'; Limit means the imperfection depth at or beyond which the tube shall be removed from service a d is equal 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 st,ructural integ-rity in the event of an Operating Basis Earthquake a loss-of-coolant/accident, or a steam line or feedwater line, break as speci Ied in 4.4.5.3.c, above.
- 8. Tube Inspection means an inspection of the steam gene dromthepointofentry(hotlegside)completelyarou(atortube f
bend to the top support of the cold leg. d the "Valuejto be determined in accordance with the recommendations of Regulatory Guide 1.121, August 1976. 14 NOV 2 1981 W-STS 3/44-M
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$ ,I REACTOR COOLANT SYSTEM /
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SURVEILLANCE REQUIREMENTS (Continued) /
- 9. Preservice Inspection means an inspection of the/ul f each tube in each steam generator performed by t,echniques prior to service to establish a bas /eline eddycondition current of tthe tubing. This inspection shall be performed after the field hydrostatic test and prior to initial' POWER OPERATION usind\theequipmentandtechniquesexpectedtobeusedduring subsequent inservice inspections. /
G ee. Tnse*+ T /
- b. The steam gen \ shall be determined OPERABLE after completing erator the correspondind\ actions (plug all tubes' exceeding the pluggSgyepkable.
limit and all tubes containing through-wall cracks) required by g Table 4.4-2. 4.4.5.5 Reports
- a. Within 15 days following the ccepletion of each inservice inspection of steam generator tubes, t'h I1 umber of tubes plugged in each steam generator shall be reported o the Commission in a Special Report pursuant to Specification 6 9.i.
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Thecompleteresultsoftheste\mgeneratortubeinserviceinspection
- b. f a shall be submitted to ,the Commissioq in a Special Report pursuant to Specification 6.9.2 within 12 monthsTfollowing the completion of the inspectiori. This Sp'ecial Report shall include:
- 1. Number and e tent of tubes inspected.
- 2. Location and percent of wall-thickness penetration for each
! indication of an imperfection. c.
- 3. Ident fication of tubes pluggei re.polred
( Results of steam generator tube irispections which fak into Category l i C-3 an,d require prompt notification of the Commission shall ho s rep'orted pursuant to Specificati,on 6.9.'@Wior toMption of 'gg,3 - p/ lant' operation. The written followup of this report shallgravide . adescription of investigations conducted to determine cause of the ! 'fube degradation and corrective measures taken to prevent recurrence. L
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i II W-STS 3/4 4-h7 NOV 2 1981
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INSERT I .,
- 10. Repair means the action of restoring the tube to useful service condition when shown acceptable, or plugging the tuba. ,
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{ , TABLE 4.4-1 MINIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTED DURING INSERVICE INSPECTION '
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l Preservice Inspection. *
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Yes N No. of Steam Generators per Unit . /
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Four First Inservicespection In\ / Two Second & Subsequent I ervice Inspections Onel,2
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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 ,tfiat all steam generators are performing in a like manner. Note that under some circumstances, the operating conditions in one or mo're ,s' team 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 (h severe conditions.
- 2. Each of the other two steam generators no(inspected during the first inservice inspections chall,be inspected during the second and third
' inspections. The fourth and subsequent inspections shall follow the instructions described in/I above.
1
. (.
N \ V 3' 3/4 4-16
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TABLE 4.4-2 STEAM GENERATOR TUBE INSPECTION / IST SAMPLE INSPECTION 2ND SAMPLE INSPECTION 3RD, SAMPLE INSPECTION
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Sample Size
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Re su,l t Action Required Result Action Required
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Result Action Required A minimum C-1
\ None N/A N/A / N/A N/A of S Tubes /
per S. G. ge p /
.. C-2 -P h g e fec tive C-1 None / N/A N/A tube and in- ge -
spect addi-s de fec tive C-1 ,Noge tional 2S gtubes tubey'and in- C-2 de fec-in this S.G. C-2 spect additional tive tubes 4S /tubes in N s / this S.G. C-3 Perform ac-tion for C-3 result of first sample Perform action (~ C-3.! for C-3 result N/A N/A A
/ of first sample C-3 Inspect all tubes in this A'11 other S. G.s are
\None N/A N/A S.G.' M de C-- I fective tubes and inspect'2S Some S.G.s Perform action N/A N/A tubes in j each C-2 but no for C- 2 result other S.G. additional of second
[ Prompt noti-S.G. are C-3 sample fic,ation to NRC pursuant to Specifi-
/ cation 6.9.1 Additional Inspect all S.G. is tubes in each C-3 S.C. and M '
de fec t ive tubes N/A x N/A Prompt notifi-cation to NRC pursuant to g Specification s 6.9.1 \ S = 12'% Where n is the number of steam generators inspected during an ( ins pec t iot. 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. The containment atmosphere (gesen"e e(particulate) radioactivity monitoring system, h b .
duo The containment pec @.e sump level-cr.d '!ce monitoring system, and Ce nht ty& gaAnbeu,+jve Gas Hoddor Either
- c. .._s cente! ament al" cooler condense +a 00':1 retc) cr c
< cr.tci ment ctmc pherc (gccccu3 or pcrticulctc) rcdiccctivity-- g monitoring system.-
APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: With only two of the above required leakage detection systems OPERABLE, ( operation may centinue 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 demorstrated OPERABLE by:
- a. Containment atmosphere (gaseous and/or particulate) monitoring system performance of CHANNEL CHECK, CHANNEL CALIBRATION and ANALOG CHo""" OPERATIONAL TEST at the frequencies specified in Table 4.3-3, clvai
- b. Containr-nttePYesump level-end <!ce monitoring system performance of CHAkti-L CALIBRATION at least once per 18 months, h -(-Spetrty epprepriate cueveilloiiue test.5 depending upon the type of-
-@ 4eakage-detectinn system-selccted.)-
( , s
\S W-STS 3/44-h NOV 2 1981 i
REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION 3.4.6.2 Reactor Coolant System leakage hall be limited to:
- a. No PRESSURE BOUNDARY LEAKAGE, -
- b. 1 gpm UNIDENTIFIED LEAK 5GE,
- c. 1 gpm tota'l reactor-to-secondary leakage through all steam generators net i: !sted 're- the 0-ecter Cee? cat Systa= and-500 gallons per day thrcugh any one steam generator r ----
- d. 10 gpm IDENTIFIED LEAKAGE from the Reactor Coolant System, y m - enconi i en e e^XAGE-- at a ":::*s aahat r --Syst== a'*ess s e &
-2235-*-20 psig, and-631. 1 gpm leakage at a Reactor Coolant System pressure of 2235 2 20 psig
' from any Reactor Coolant System Pressure Isolation Valve specified in Teble ?.'-1.*
la+cr 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.SHUTOOWN 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 STAND 8Y within the next 6 hours and in COLD SHUTOOWN within the following 30 hours.
- c. With any Reactor Coolant System Pressure Isolation Valve leakage
.. ... greater than the above limit, reduce the leakage _ rata. to within. limits -
** -- =*
within 4 hours, or be in at least HOT STAN08Y Dithio~the dekt 6"hdurs ~ ""' and in COLD SHUTOOWN within the following 30' hours.
^ Test-pressuresJass than 2235-ps4g-but-greater 4han-350-psig-are-attwed-
- -Ob s e rved -4 e a ka ge-s h a l-1-ce-a d j us ted--f o r-the-actua-t-t e s t-p re s su re -up-to-2235-p s i g-
. -assuming-the-leakage-to-be-directly-properational-to-pressure-differential' to-
-the-one-4alf-po sr.
i 13 3/4 4-24
REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE SURVEILLANCE. REQUIREMENTS 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 per 12 hours; l
@ b. Adne Monitoring the containment-pec-. sump inventory and discharge at least once per 12 hours; l
')l, -Measurement cf die-60NTROLLED--LEAKAGE-to-the reactor coo + ant-pump h - least-once-per-31-days; tith th: seals-when
- ct!:tf the_pen.
rg 9ve- ful-ly-4 Reactor-Coo The 4ent-S
-provisions-of-SpecR-ication-4ror4-are not appttcab'le-for entry into---
-H00E-3 or-4; C .ti, Performance of a Reactor Coolant System water inventory balance at g least once per 72 hourst-and-d e b Slead Y 5Me*Pc'at'54 neProu ..
e+ S'Pecificertsb=s 4 0 4 are not apphcable for en+vy snko H0063 3 ** 4. '5'#f c),g. Monitoring the Reactor Head Flange Leakoff System at least once per 24 hours. l i 4.4.6.2.2 Each Reactor Coolant System Pressure Isolation Valve-specified in--
--Tab!: 3.M shall be demonstrated OPERA 8LE by verifying leakage to be within its limit:
- a. At least once per 18 months, I
- b. Prior to entering MCDE 2 whenever the plant has been in COLD SHUTDOWN for 72 hours or more and if leakage testing has not been performed in the previous 9 months, l
- c. Prior to returning the valve to service following maintenance, repair or replacement work on the valve, ana I
- 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 10 W-STS 3/44-h
i. n i L L TABLE 3.4-1 .' s
- j. N REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES -
\s VALVE NUMBER FUNCTION
,p o \ . .
L TRg inbma4 ion will be suppsed at d \t & e.< A a e . : N
\ /
l- % / u N. ,e
\ /
s /
\/ s/
l.> X l. \
,./
f f
/
/ -
/
c
/
/
/ 'x \
\
\\
\
\.,
i-
'\ \
\-s i . S.
k kk * ', W-STS '3/4 4-M. fiO'! 2 1991 i
. , , , , . , - - .m- ,,, , - . - , , _ _ . _ _ _ . . _ , . . < , , _ _ _ . _ , . . . , _ , , , , . , _ _ . , _ , - ~ , _ - . _ , , , - _ _ . ~ . . . , , , . - , , - - , . _ , , , , , . _ , _ _ . .
i 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. h. With any one or more chemistry parameter in excess of its Transient Limit, he 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 chluride 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 pressuri7er 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 frcawende; speci# icd iS Mab h 1.4-3. specified inMle 3.ti-2. 4+ leasi enee. per ~n 1. kow s** ( N g NOV 2 1991 PSTS 3/4 4-74
TABLE 3.4-2 I REACTOR COOLANT SYSTEM CHEMISTRY LIMITS STEADY STATE TRANSIENT 4 PARAMETER LIMIT LIMIT DISSOLVED OXYGEN
- 1 0.10 ppm 5 1.00 ppm CHLORIDE 1 0.15 ppm 5 1.50 ppm FLUORIDE $ 0.15 ppm 5 1.50 ppm 4
" Limit not applicable with T,yg less than or equal to 250*F.
t 23 W-STS 3/4 4-26 NGV 2 1981
lABLE 4.4-3 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS SURVEILLANCE REQUIREMENTS SAMPLE AND PARAM R ANALYSIS FREQUENCY DISSOLVED XYGEN* At least once p 72 hours CHiORIDE At least onc per 72 hours FLUORIDE At least ce per 72 hours (
"Not required with T avg less tha or equal 250'F
- t. l
-W-W-STS -3A 4-In- .ga / 2 1981
. REACTOR COOLANT SYSTEM 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 mi.s continue for up to 48 hours provided that the cumula-
!- tive operti.v.g 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 within 6 hours. avg
- c. With the specific activity of the primary coolant greater than 100/E microcuries per gram, be in at least HOT STANDRY with T less than 500*F within 6 hours. avg
- With T,yg greater than or equal to 500 F.
I 2.9 W-STS
- 3/4 4-N C/ .0 19S1
I REACTOR COOLANT SYSTEM ACTION: (Continued) MODES 1, 2, 3, 4, and 5:
- a. Witn the specific activity of the primary coolaret graater 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 OCC"RRENCE EVENT shall be prepared and submitted to the Commission pursuant to Specifi-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,
- 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, t 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 4.4.8 The specific activity of the primary coolant shall be determined to be within the limits by performance of the sampling and analysis program of Table 4.4-4.
- t. (
- 25 NOV 2 1981 W-STS 3/4 4- M
300 E . : .
$u e . .
250~
;E ** * * * * *
't-
- a ; .
t . : . o . . . o 200- - t u : : UNACCEPTABLE OPERATION z . . . .
.c . . . .
o i30 . . . . .... .. . :. . . o : : : : o . . . . . . i : i i : i m .00 ................
- ........,. ..3 .... . .. ..:. . ....,... .. .
m : : : : 1 ACCEPTABLE i . . E OPERATION i i i w . . J . . . . .
< ,0 .... .. .... . .... ..... ... .. ... .. ... . .
a . . : : : : : o . . . . . . u : : : : : : m . . . . . m : : : : : : o . . . . . o : : 0 . . . . . . . 20 30 40 50 60 70 80 90 100 PERCENT OF R ATED THERM AL 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 >1.0yCi/ gram DOSE EQUIVALENT l-131 Seabrook - Units 1 & 2 3/4 4-26
o , TABLE 4.4-4 sc 1 $) o' PRIMARY COOLANT SPECIFIC ACTIVI~ SAMPLE AND ANALYSIS PROGRA' t 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 8
- 3. Radiochemic31 for E Determination 1 per 6 months
- 1 Isotopic Analysis for Iodine # # # # #
- 4. a) Once per 4 hours, l,2,3,4,5 Including I-131, I-133, and I-135 whenever the specific u, activity exceeds 1.0
); pCi/ gram DOSE
,, EQUIVALENT I-131 '
'g g or 100/E pCi/ gram, and b) One sample between 2 1,2,3 and 6 hours following a THERMAL POWER change exceeding 15 percent of the RATED THERMAL POW 2R within a 1-hour period.
l fUntilthespecificactivityoftheprimarycoolantsystemisrestoredwithinitslimits.
- 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.
5 80 enemmimme
, REACTOR COOLANT SYSTEM 3/4.4.9 PRESSURE / 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,withinthPfollowing30 hours. SURVEILLANCE REQUIREMENTS 4.4.9.1.1 The Reactor Coolant System temparature and pressure shall be determined to be within the limits at least once per 30 minutes during system heatup, cooldown, and inservice leak and hydrostatic testing operations.
---4r4.-9.L2-The-reacto r-v es s el-ma te r-i a l-i r rad i a t-i o n-s u rv e ille nce-s p eci me ns--
--shalLbe removed-and-examined;-to-determine-changes in material-properties, as-required-by-10-CFR-50;-Appendirx-H-in-accordance-with-the-schedule-in-
@ Table-4-4-5r--The-results-of-these-examinations- snall-be used-to-update----
-Figures-3,4-2-and-3r4 -3r-19 W-STS 3/4 4-M ,fl0V 2 19 8 !
COPPER CONTENT :C O N S E R V ATIVE LY AS SU M E D TO B E 0.'10 WT% (ACTU AL C ONTE NT=0. 0 6 WT%) RT NDT INITI AL :40 F RT NDTAFTE R 16 EFPY :1/4T.110 F 3/4T.87 F CURVE APPLIC AB LE FOR HEATUP R ATES UP TO 60 F/HR FOR THE SERVICE PERIOD UP TO 16 EFPY AND CONTAINS M ARGINS OF 10 F AND 80 PSIO FOR POSSIBLE INSTRUMENT ERRORS n LE AK'TE ST m LI M IT C. . tu , C
- D .
M : W 2000- - - e . : O. l 2 ' tu : . H : u) . .
>- l l M : :
H : : : 2; : :
<( .
O 1000- - - C R ITIC A LITY g/ O LIMIT B ASE D ON INSERVICE C HYD R O STATIC E HEATUP~ TEST CURVE . l TEM PER ATURE O : : (2 5 $ F) FOR H . THE DERVICE O
. PERIOD UP 4 ,
TO 18 EFPY uJ . . . m . : 0- i i i i 0.0 100.0 200.0 300.0 400.0 500.0 INDICATED TEMPER ATURE (DEG. F) FIGURE 3.4-2 SEABROOK UNIT 1 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS APPLICABLE TO 16 EFPY Seabrook - Units 1&2 3/4 4-29
r l M ATERIAL PROPERTY BASIS COPPER CONTENT
- CONSERVATIVELY ASSUMED TO BE 0.10 WT% ( ACTU AL C O NTENT=0.0 6 WT%)
[ RTNDTINITI AL :40 F RTNDTA FTER 16 EFPY .1/4T.110 F 3/4T.07 F CURVE APPLICABLE FOR COOLDOWN RATES UP TO 100 F/HR FOP. THE SERVICE PERIOD UP TO 16 EFPY AND CONTAINS M ARGINS OF 10 F 3 C, w w AND 60 PSIG FOR POSSIBLE INSTRUMENT ERRORS 9 2000-- - - .-: - - - - - -* - - U) . . a v u : : : m . . D, v m : : : : u g . .
- a. : :
a - m . . n - 2 : : a . . Z 1000~ ***
'?' **
COOLDOWN R ATES ( F/HR) : . . : 40 NNN .. ! i 6o- . : 100-- , I
, {
0- , , , , 0 100 200 300 400 S00 IN DICATED TEMPER ATURE (DEG. F) FIGURE 3.4-3 . SEABROOK UNIT 1 REACTOR COOLANT SYSTEM COOLDOWN LIMITATIONS APPLICABLE TO 16 EFPY L b Seabrook - Units 1&2 3/4 4-30
TABLE 4.4-5
!? - - - . .
'^
SURVEILLANCE CAPSULE REMOVAL SCHEDULE i
\
Capsule Orientation of Lead Homoval Expected Capsule identificatio'nN Capsuloslal FactorIbl Time Fluence (nicIn2) U \58.5* 4.00 1st Refueling 3.3 x 1D1a Y 241' 3.69 5 EFPY 1.. 251019[c] V B1' 3.69 9 EFPY 2.2 x 1019[d] X 238.5* 4.00 15 EFPY 3.9 x 1019
. W 121.5* MOO Stand-By w Z 301.5* 4.00\ Stand-By b
. N Os-g- ; a. Reference Irradiation Capsu's Assem Drawing > Figure 2-4.
- b. The factor by which the capsule fluence leads the vessels maximum inner wal1 fluence.
- c. Approximate Fluence at % wa!! thickness at End-of-Ufeh
- d. Approximate Fluence atNe'ssel inner wall at End-of-Life.
\N x *r
'\
s, N r
~ , .
IO p g
/
gi - N m
REACIOR COOLANT SYSTEM PRESSURIZER LIMITING CONDITION 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 temperature differential of (320) F.
APPLICABILITY: At all times. ACTION: With the pressurizer temperat~reu 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 spray water temperature Jifferential shall be determined to be within the limit at least once per 12 hours during auxiliary spray operation. Y - l k 32, ! W-STS 3/4 4-% .fl0V 2 1981 l t
1 REACTOR COOLANT SYSTEM OVERPRESSURE PROTECTION SYSTEMS LIMITING CONDITION FOR OPERATION 3.4.9.3 At least one of the following overpressure protection systems shall be OPERABLE: 4.b Two power operated relief valves (PORVs) with a lift setting of . less than or equal to (A50) psf 1 a" A max /mu m selfpo'ini definef ksf Fipec 3 9-9 or
%C The Reactor Coolant System (RCS) depressurized with an RCS vent of g
greater than or equal to (3 2) square inches. APPLICABILITY: MODE 4 when the temperature of any RCS cold leg is less than or equal to (47&)*F, MODE 5 and MODE 6 with the reactor vessel head on. 319 m ACTION: W
- a. With one PORV inoperable, restore the inoperable PORV to OPERABLE status within 7 days or depressurize and vent the RCS through a ,
(3.2.) square inch vent (s) within the next 8 hours,d ma'inf 4oain the RCSsfahn CPEg4ser to A ,. vedea confiUon tm4sl both PORV's have been res4o,c i
- b. With both PORVs inoperable, depressurize and vent the RCS through a (3 2) square inch vent (s) within 8 hours maida*m a th< Acs in c. ve 4e4 canal + loo, unUI both PORQse kave been reshores +o opeppes.g ,k+us .
- 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.
a.. ho rc$sdud ker.f n ~*vsl (RWO sach'on relief valves e.a.<A A h A. 'e+ P* int 4 , ps,3 ,e f 7, ( o r l 1 33 W-STS 3/4 4-X NOV 2 1981
REACTOR CCOLANT SYSTEM OVERPRESSURE PROTECTION SYSTEM SURVEILLANCE REOUIREMENTS 4.4.9.3.1 Each PORV shall be demonstrated OPERABLE by: ' a. Performance of an ANALOG CHANNEL OPERATIONAL TEST on the PORV actuation channel, but excluding valve operation, within 31 days prior to entering a condition in which the PORV is required OPERABLE and at least once per 31 days thereafter when the PORV is required OPERABLE; l b. Performance of a CHANNEL CALIBRATION on the PORV actuation channel at least once per 18 months; and l
- c. Verifying the PORV isolation valve is open at least once per 7? hours when the PORV is being used for overpressure protection.
1 3 4.4.9.3.Y The RCS vent (s) shall be verified to be open at least once per 12 hours" when the vent (s) is being used fcr cverpressure protection.
"Except when the vent pathway is provided with a valve which is locked, sealed, or otherwise secured in the open po;ition, then verify these valves open at least once per 31 days.
I4.4.9.3.2 Each RHR suction relief valve shall be demonstrated OPERA 8G when the RHR suction relief valves are being used for cold overpressure protection t as follows: C ,
- a. For RHR suction relief valve :
By verifying at least once per 31 days that RHR RCS 'S cIion" ~
~
1) Isolation Valve is open with power to the valve---. - operator removed, and { 2) By verifying at least once per 12 hours that -is open. - -.
- b. For RHR suction relief valve
~
! 1) By verifying at least once per 31 days that is open ( with power to the valve operator removed, and ~ - ~ ~ - ~ ~ - - 4 2). By verifying at least once per 12 hours that (isopen. _
- c. Testing pursuant to Specification 4.0.5. -
3l'i 4 3'l
,' .1 0
6 3 0
.0 t
- 3
)
0 F i6( 2 E R U T A _ R E 0P r0 2 M _ E T TID N S ISN F C OPA 0 P0 S 9 R T E 6R I SDO > T _
.NR 0 YARES. T; 5 P
TF E TC 0NE T F 6 6 6 1 1 91Ef - 3FMf E 00 - 9 e OU T SSTN RT 1s 52 T RNSE 7 lI G; fGNI S I 5 ERIN E S 0 HALA P + 0 TMDR iT 0 4 1 RSS R 0 6 5 3 iI NS = = OO - lTP A F P P lNR F AOO0 VCF5 0 6 0 0 0 0 0 0 0 0 0 0 0 5 0 0 5 0 5 2 2 2 1 1 G asz5Eg> a1a2N2 nCC2m 0.>:* OM OOIC O<mr,5 mMmC=m 0=O;mO$OZ WmiTOEin - -
- C MQC) ,o C :
CD_G ._ "@N o D* A $ ,c l
l t REACTOR COOLANT SYSTEM 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 structural 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 REQUIREMElRS 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.
, t 36 W-STS
- 3/4 4- M
'NOV 2 1981 J
REACTOR ^?0LANT SYSTEM i
- .... r. LuulANT SYSTEM VENTS LIMITING CONDITION FOR OPERATION O
0%C 3.4.11 At least one Reactor Coolant System vent path consisting of (4we3 vent valvebCaC 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-(,Two parattel paths) 4L (R:2cto- Ccel:nt Sy: tem high point).
APPLICAE*LITY: MODES 1, 2, 3, and 4. ACTION:
- a. With cne of the above Reactor Ccolant System vent paths incperable, .
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 anc block valves in tne inoperaole i
. vent patn; restore tne inoperable vent path to OPERABLE status .
. within 30 days, or, be in HOT STANDBY within 6 hours and ir, j COLD SHUTDOWN within the following 30 hours. ; i
- b. With two or more Reactor Coolant System vent patns inoperable; 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 hcurs or be in HOT STANCEY -
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 rou:,, and ,
1 i 37 l W _-STS 3/4 4-44 1
- : :. - s6 :. --n...,
- .. .-- c. v.e. .r e v.
.-~VE : L:. 'C E :E X':REMENTS (^or.-i ued) l l
- c. Verifying flow through the Reactor Coolant System vent paths duri-ing venting.
6 e an*
- a 39 W-STS 3/4 4-M
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 these valves.
C. This Specification 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. l D. At Seabrook, the automatic interlock is set at 365 psig, and the automatic
- isolation signal has been changed to 660 psig.
I E. This change incorporated as a result of the cold overpressure mitigation I analysis. F.- The boron injection tank has been deleted from the Seabrook design. G. This changes the allowed outage time to 8 hours, which will allow most repairs to accumulators to be completed before requiring a shutdown. In addition, accumulators are needed in accident analysis only for large LOCA which is a relatively unimportant risk contributor. H. The standard 12 hour surveillance interval is too frequent, occupying operator time and attention unnecessarily. Also, the operator will respond to alarms and indications. J. This changes the allowed outage time to 7 days, which will allow most repairs to be accomplished before requiring a shutdown. Also, the extended allowed outage time does not significantly effect ECCS system reliability. , K. The most likely inoperable conditions for the RWST are level and boron concentration slightly out of range, which would have a minimal effect on l the ability of the RWST to function. However, to restore these conditions could require mora than the standard I hour allowed outage time. Thus, the extended allowed outage time required by Action a permits most inoperable conditions to be restored before requiring a plant shutdown. 1 i L-
l i I 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS LIMITING CONDITION FOR OPERATION
- 3.5.1.1 Each reactor coolant system accumulator shall be OPERABLE with:
- a. The isolation valve open,
. b. A contained borated water volume of between 6/25' and 4460 gallons,
- c. A boron concentration of between (1900) and (2100) ppm, and
, d. A nitrogen cover pressure of between 480 and 4 7 9 psig.
APPLICABILITY: MODES 1, 2,~ and 3*. ACTION:
- a. With one accumulator inoperable, except as a result of a closed
/ isolation valve, restore the inoperable accumulator to OPERABLE status within l-hour or be in at least HOT STANDBY within the next 6 hours and i 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 hours by: (bh
- 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.
s W-STS 3/4 5-1 NOV 2 01980
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 rc=cvel cf the breaker 'rce the circuit- @
- d. At least once per 18 months by verifying that each accumulator isolatiou 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 infection test signal.
1."2. 4.5.1.% Each accumulator water level and pressure channel shall be demonstrated OPERABLE:
- a. At least once per 31 days by the performance of a ANALOG CHANNEL OPERATIONAL TEST.
- b. At least once per 18 months by the performance of a CHANNEL CALIBRATION.
SEP 151981 W-STS 3/4 5-2
f EMERGENCY CORE COOL.ING SYSTEMS 3 /4 . 5. I ACCUMULATORS LIMITING CONDITION FOR OPERATION 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: Q.. With one or more accumulator isolation valve (s) open and/or power available to the valve operator (s), immediately close the accumulator isolation valve 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 t, removed from the valve operator at least once per 31 days.
a 6, The y revisn' n,s oSP S "Ninkbn 3 O'H Ars nd' *9elic"bic 4** **b 1 m)., MODE Lt from HODE 3.
> w u po-, e ,: s e s . , . + n o o , a r.. , , , , , a ce peuun3er pre ssu.r e 7 tooo psg ed aux,,mt 4,,, lut /.4 4 g,/ge shal1 6e opm p,s rey t v e.1 b 3 s p ec f d ,p g , 3 , g,f,g 4,,
*n m, = ^r rc -r 15 ocs wie ics: i- != :F- eq=! to 307 F.
p 3 3/4 5-h 46
e EMERGENCY CORE COOLING SYSTEMS 3/4.5.2 ECCS SUBSYSTEMS - T > 350 F ava 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-(feur !ce; p!:nt: cr!y),
- c. One OPERABLE residual heat removal heat exchanger, o 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: (11 Jans
- a. With one ECCS subsystem inoperable restore the inoperable subsystem to OPERABLE status within 72 he_r. 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.
i N W-STS 3/4 5-3* NOV 2 01980
EMERGENCY CORE COOLING SYSTEMS - SURVEILLANCE REQUIREMENTS 4.5.2 Each ECCS subsystem shall be demonstrated OPERABLE: M
- a. At least once per )E hours @by verifying that the following valves are in the indicated positions with power to the valve operators r.emoved:
Valve Number Valve Function Valve Position a. Seea.Tnsed 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 Lad'sh ano "n+cl eerd el san,h4]
V
- 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-i in9+ CONTAINMENT INTEGRITY, and
\PRmuy l 2. Of the areas affected within containment at the completion of g ach containment entry when , CONTAINMENT INTEGRITY is established.
- d. 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 4ES,psig the interlocks prevent the valves from being opened, and @
l b) with a simulated or actual Reactor Coolant System pressure l signal less than or equal to 60& psig the erlocks will 2.
'cause the valves to automatically close. 440 A visual inspection of the containment sump and verifying that h
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. 5* W-STS 3/4 5-A
'NOV 2 1981 L
Insed- I
. Wive.. Numbe.s Valve Fanc. hon llatue. Pos14wn SI- y.- 3 . _ _ __
Acca,,mlab. Iso Ia& ton . ._.. .. ... _ Op en .
-- SL'hfl . -- - ..-. . . .._H '* . . . . - _ . ..Open . ..
_ _ _ . _ _ _s r .-v 3 2. - 8i
' l .. _ _ . . . . ._ _ .. _ __ . O p e n . _ . .
Sr-V_ _H'l In in . 0 p en . . _ _ _... C SS V-43 RWsT/sr Pamp rsolaion...__ . ._ . O p en . .. ._ _ . _.. . _ . c G S-V- El ti n 8' dren_____. SI-hlitL.___.___._..S r Pump fo cold I.e9 IsoLJw,: Ope n .__. ___
._ _ R H-V- I'l _.RHR ru ,p.bo Gd LeyIsolahon. .__ Open .. _ , .
RH-V_2(o " M 8t l' dpen. _ . . _ _
- R W _ 3.2 -
Rag Ao let t.e3.Isoldion clos.el _ R n-V- 70 " 4 c I osel._._ _... .. .
.5%-3-17 .-- . 5E.Fo Hot _ Leg.rsolakon _ closes ... .
GI- V l O.'l tu n 4 n closed. _ - Wh - . & Ohf..er. .h w.4.+e<i4 w.m m ...-.. a 4.m. e e- . * - = ..* . &
+ v- '$ M. e *4.e6 b.
I t l .
EMedGENCY 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.
- 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 , is ca 1hr
- f. By verifying that each of the following pumps d="agable of bevelof ep: the m1
#ndicated-df(edj "pe:cification S 4.0.5:i charge pressure er , rgf reelgtice '!:u when tested pursuant to
( l. Centrifugal charging pump 1 2470 psik
- 2. Safety Injection pump 1 /4 45- psid cl
- 3. Residual heat removal pump >_ 173 psigj
- g. By verifying the correct position of each electrical and/or mechanical position stop for the following ECCS throttle valves:
- 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.
Hetk Hed SJdg r,3,j,, Ederwbs'de HubSaMy EnJechbn
-HAGI System RM- Sys tem Valve Number Valve Number
- a. SIV-N3 a. SIVFo
- b. SI V -147 b. S t-V- 95~
- c. s r.v - ts l c. s r- V- lott
- d. SI u- 14y d. sr.v.109
. C. sz-V- \ q f- Sr V- 13d i SI- V - 11S'
- h. .sr V - 129 6
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, folicwing 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 m gpm, and b) The total pump flow rate is less than or equal to Egg gpm.
- 2. For safety injection pump lines, with a single pump running:
The sum of the injection line flow rates, excluding the a) highest flow rate, is greater than or equal to yy g gpm, and b) The total pump flow rate is less than or equal to ffo , ( gpm.
- 3. For residual heat removal pump lines, with a single pump running, the sum of the injection line flow rates is greater than or equal to gggggpm.
\
7 W-STS 3/4 5-K g.,j gg L.
EMERGENCY CORE COOLING SYSTEMS 3/4.5.3' ECCS SUBSYSTEMS - T,yg < 350*F LIMITING CONDITION FOR OPERATION 3.5.3 As a minimum, one ECCS subsystem comprised of the following shall be OPERABLE: ,
- a. OneOPERABLEcentrifugalchargingpumpX
- b. One OPERABLE residual heat removal heat exchanger,
- 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 4f ACTION:
- a. With no ECCS subsystem OPERABLE because of the inoperability of either the centrifugal charging pump or the flow path from the i
refueling water storage tank, restore at least one ECCS subsystem to OPERABLE status within I hour or be in COLD SHUTDOWN within the next i- 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 avg less than 350 F by use
.of alternate heat removal methods,
- 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 1
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. ! d. The provs*s s%s of SP'ctFNahbs 3 0 H 0 ^DY afPI A ble Eu" C*bV ido M O D E '{ .from f10DE 3 f ^ :: hr c' ^"e-centri fug21-Gharging-pump--and-cne s afety injaot4cr pu-n
-shal' be ORERABtE-whenever-the te perettere of er,e er incre-cf the 9CS _ce!d
-1;gs -i; les.-th.: c equal--te (27i)oN onJ
# D>tthih 12 hours prior to envy n,$whopar'3 fue hoDE 9 bath chq *3 pu,n both s&sts nn'l"W*A Pu =PS Aut**A h't Sp eisika be6,s 3.S. 2, shell be 096RAe,L2.
8 W-STS 3/4 5-% MAY 151980 L
EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS 4.5.3.1 The ECCS subsystem shall be demonstrated OPERA 8LE per the applicable requirements of Specification 4.5.2. ' 4.5.3.2 All charging pumps and Safety Injection pumps, except the above allowed OPERABLE pumps, shall be demonstrated inoperable by verifying that the motor circuit breakers are secured in.the open position at least once per _ 12 hours,wh;n:v;r th t::per:ture of 0.9e c= = ara a# th: 9C5 celd 1 g:-i: les
- thea or equal t: [27"]"I.-
9 W-STS 3/4 5-1
/
EMERGENCY CORE COOLING SYSTEMS 4.5.4 BORON INJECTION SYSTEM BOR INJECTION TANK LIMITI NDITION FOR OPERATION s - 3.5.4.1 The b ron injection tank shall be OPERABLE with:
- a. Acont*r.edboratedwatervolumeofY[tYI[r, ,' 8YS '
gallons,
- b. A boron con entration of between 20,000 and/22,500 ppm, and
- c. A minimum sol ion temperature of 145 F.
APPLICABILITY: MODES 1, 2 and 3. ACTION: With the boron injection tank i operable, estore the tank to OPERABLE status within 1 hour or be in HOT STAND and b rated to a SHUTOOWN MARGIN equivalent to 1% delta k/k at 200*F within t ne 6 hours; restore the tank to OPERABLE status within the next 7 days or be i HOT SHUT 00WN within the next 12 hours. SURVEILLANCE REQUIREMENTS
, s 4.5.4.1 The boron inje ion tank shall be demon rated OPERABLE by:
- a. Verifying t e contained barated water vol me at least once per 7 days,
- b. Verifyf g the baron concentration of the wate in the tank at least once r 7' days, and c, Ver fying the water temperature at least once per '4 hours.
l l 1 i ta W-STS =st: a APR 1519 1 k
ERGENCY CORE COOLING SYSTEMS HE TRACING LIMI CONDITION FOR OPERATION x 3.5.4.2 At ast two independent channels of heat tracing sh I be OPERABLE for the boron jection tank and for th.e heat traced portion of the associ-ated flow paths. APPLICABILITY: M0 5 1, 2 and 3. ACTION: With only one channel of eat tracing on either th boron injection tank or on the heat traced po'rtion of n associated flow pat OPERABLE, operation may continue for up to 30 days p vided the tank an flow path temperatures are verified to be greater than o equal to (145)* at least once per 8 hours; othe mise, be in at least HOT S NDBY within hours and in HOT SHUTDOWN within the following 6 hours. I SURVEILLANCE RE0VIREMENTS
, s 4.5.4.2 Each heat tracing annel for the boron i ection tank and associated flow path shall be demonst ated OPERABLE:
- a. At least once er 31 days by energizing each eat tracing channel, and
- b. At least ce per 24 hours by verifying the tank nd flow path temperat es to be greater than or equal to (145) . The tank temper ure shall be determined by measurement. Th flow path tempe ture shall be determined by either measuremen or recircula-ti< n low until establ.ishment of equilibrium temperat es within the ta .
f k 44-W-STS 3/' 5- % SEP 15 sg
EMERGENCY CORE COOLING SYSTEMS
-3/4.5.h REFUELING WATER STORAGE TANK LIMITING CONDITION FOR OPERATION
- 3. 5. The refueling water storage tank (RWST) shall be OPERABLE with:
- a. A contained borated water volume of between 47toooand aff53ggallons, @
- b. A boron concentration of between (2000) and (2100) ppm of boron, and
- c. A minimum water temperature of (35) F.
APPLICABILITY: H0 DES 1, 2, 3 and 4. ACTION:
-W4th-t he-ref ueM ng-wa t e r-s tora ge-ta n k--inoperaM erre s tore-the-t a n k-to-O PE RA 8 tE~-
--status-within-bhour_cr-be-4n-at-least-HOT-STANDBY-within-6-hours and-in-COL +
-6HUTDOWN-wi t hi n-th e-fo M owi ng,30-hou rs; I '
SURVEILLANCE REQUIREMENTS 4.5. The RWST shall be demonstrated OPERABLE:
- a. At least once per 7 days by:
- 1. Verifying the contained borated water volume in the tank, and
- 2. Verifying the boron concentration of the water.
- b. ,At least once per 24 hours by verifying the RWST temperature,whe+
@ -the (at:i&) 9 t=p--+"ra b 4e e t" 35"- -
E th % R m s r m o p e r a y e ,; llons L'd /c" S tun t% (n%coc 4 codaked.1,omfel
<pttee on a 6eronwn6 velu~e, 3,egfu conae.dration tkwugooeen
.yade,.%, ,p,,oco 9,Mess l 74.,
2000 fem o resfwe. O 64 h, O AcRAds e SM,o 4,14g e Acuo o rb 'M un srmbhy u;t% i kou > m.1 is GM ct+urpoton uima the Gilms4 30 "'
- b. W A a eedah,1 lu.mtcA mfe wlu-e less b Al31; coo ggIs ao ed eren g
8 cen<edws Isti Iecs th.sn 16' w tY
- I e R wale-4e""H'*I*".la.& s, M o e,* d*
- ihe bk h, CPf!R Acte d,bss 5thi.e l hes
- Le in af /cd//vT "*"#7 *' '
6 hans and in cot.D Sn47touw Se7p'/h 4//04"7 3 # #*' e' s _ 3/4 5;M
~
W-STS N ___ .. Apg 15 G78
JUSTIFICATIONS Section 3/4.6 i i 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 requiremect 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.
!!. 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. J. The Seabrook design does not use educators, but rather has gravity feed. In addition, there is no means to measure flow. Ilowever, there is 3/4" drain connection to verify flow. K. Per requirements of Generic Letter 82-16. L. Seabrook Station plant specific data. H. This Table is removed and will be contained in a licensee maintained and controlled document. N. This changes the allowed outage time to 7 days which will allow most repairs to be completed before requiring a shutdown. Also, the allowed outage time does not significantly affect the Containment Spray System reliability.
P. The Spray Additive System is of very low importance compared to the Containment Building Spray. The action statement has been modified to allow outages of 31 days, reflecting the low importance. The risk from a plant shutdown caused transient is much greater than the risk from having SAS inoperable. R. The surveillance period was changed to 6 months because the valve alignment is very unlikely to change over 6 months and because of the low system importance of the SAS. S. The Channel Check frequency was changed to every 7 days to be consistent with the relatively low importance of Ilydrogen Monitors as reflected by the 30 day allowed outage time. T. The system function test frequency was changed to every 18 months to avoid unnecessary cycling of equipment. U. The allowed outage time was changed to 7 days to reflect the low risk significance of the enclosure building. A detailed evaluation of this change will be provided in the risk based analysis to be submitted in mid August. i 9 I
t 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT
^
CONTAINMENT INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.1.1 Primary CONTAINMENT INTEGRITY shall be maintained. APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: . @ Without primary CONTAINMENT INTEGRITY, re NTAINMENT INTEGRITY within . one hour or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE REQUIREMENTS . 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.p
- 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 8 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 B and gc* enetrations, the combined leakage rate is less than or equal to 0.M L,. @
"Except valves, blind flanges, and deactivated automatic valves which are '
located inside the containment and are locked, sealed or otherwise secured in the closed position. These penetrations shall be verified closed during each COLD SHUT 00WN except that such verification need not be performed more often than once per 92 days. W-DUAL 3/4 6-10
- m. : i :ai
l l l lf CONTAINMENT SYSTEMS CONTAINMENT LEAKAGE ! LIMITING CONDITION FOR OPERATION 3.6.1.2 Containment leakage rates shall be limited to:
- a. An overall integrated leakage rate of:
- 1. Less than or equal to L,, (0.20) pegcent by weight of the con-tainment air per 24 hours at P,, (M psig)..'tm.
% te:: th a ar an"O te L , (0.!0) p:rcent by a:fght of the co w t @ :ent cir per 24 heur et a reduced pressur; cf "g, Z WC-75~
- b. A combined leakage rate of less than or equal to 0.9a L f r all penetrations and valves subject to Type B and C tests, when pressurized
. Ne NMvMual pucht'on will be allowea +o cu: ecd .5 7, of t4e i to P,I allowed ..(o or LS) tofa. .
- c. A combined bypass leakage rate of less than or equal to (0.10) L, for all penetrations identified fa Td!: 3.0 1 as secondary containment bypass leakage paths when pressurized to P,. @
APPLICABILITY: MODEX 1, 2, 2 :nd ' f (or p ,1. , lo e d 'T *I #1*d' @ ACTION: With either (a) the measured overall integrated containment leakage rate exceeding 0.75 L, er 0.75 Lg , : :pp!!::ble, or (b) with the measured combined leakage rate for all penetrations and valves subject to Types B and C tests bexceeding0.[L,,or(c)withthecombinedbypassleakagerateexceeding (0.10) L,, restore the overall integrated leakage rate to less than or equal , to 0.75 L,er 1::: than ;r :qe:1 to 0.75 L ,gas ap;lic 2!e.-the combined leakage rate for all penetrations and valves subject to Type B and C tests to , less than or equal to 0.N L,, and the combined bypass leakage rate to less than or equal to (0.10) L, prior to increasing the Reactor Coolant System temperature ( above 200*F. , n W-DUAL 3/4 6-20 007 1 mg
CONTAINMENT SYSTEM ( SURVEILLANCE REQUIREMENTS
- 4. 6.1. 2 The containment 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 fE ! ".15.0 '1072h MC/AMs 5-6.8 818[ &
- a. -T'r:: he* 7+*W(n arm ate;r:t:d Cer.t:i = = t tr 9 =;- Ste)-
;h:11 ': cc- *cted et 10 ; 10 w,,u, ;,,t...;1 c rir.; :h:t e r, at Ofth:r ", (50 ; fg) Or :t at I EII) ^2 ^ 7### 00 2^
perfed. Th= 6'i rd *-' Of :--h et k= 1: :;r.dztzd dur*q the _ tu+ a~n fn, th. in y.. ni e+ <- y!ce : :; nti = ,
- b. If any periodic Type A fails to meet e4%e,6.75 L, :r .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 f meet-:!th:ro.75 L,er ?5 Lg, a Type A test shall be performed at least every 18 months until two consecutive Type A tests meet eft'er C.75 L, r .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,,_:r 0.25 Lt.
- 2. Has a duration sufficient to establish accurately the change in 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, (M psig),er *g (25 ;;ig}r d.
h W,/ Type B and C tests shall be conducted with gas at P, (M psig) at intervals no greater than 24 months except for tests involving:
- 1. Air locks.
- 2. Penetrations using continuous leakage monitoring systems, t and t0UAL 3/4 6-30 NOV 15 577
c. INSERT I
- a. Periodic Type A tests shall be conducted to Pa(46.1 psig) at the first refueling shutdown but not more than three years subsequent to the preoperational test and at intervals not to exceed five (5) years thereafter.
I
i CONTAINMENT SYSTEMS < 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 B and C tests at least once per 24 months except for penetrations, which are not individually testable; penetrations not individually testab,l g ha g g egergined to have no detectable leakage when tested witn-se:; ::::25 while g thecontainmentispressurizedtoP,kpsig)duringeachTypeA test.
- f. Air Locks shall be tested and demonstrated OPERABLE per Surveillance Requirement 4.6.1.3.
'tA. Typ: 9 periedic test: are net required for p:nctratien: continu u;1y
- ritered by the Centeinment 1svioL;vn Live end-Chenuci Lid Ice 33mr- h ,.
!: tier Ey tem: previded th: ;y3tein; ere OPE'?ACLE per arveillence $.
j acquiren nt 4.5.1.4. 3% 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 combined leakage rate provided the seal system and valves are pressurized to at least 1.10 Pa so.'s (% 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 than once shall per be conducted at P,(M 3 years. 4) psig) at intervals no greater g i1 The provisions of Specification 4.0.2 are not applicable. W-DUAL 3/4 6-40 M '
@ j
/
/
/
5 p 5 s m M E 5 2 d w Ci sc 3
- Tins 'i So ba4 ion 4o be saPph'ed -,. 4 m
E a af a l er da+e. j Q E 5 5 5 a m E E 8 m 5 C C 5 c. N
\
\
\
W-0UAL 3/4 6-SD MAR 151978
CONTAINMENT SYSTEMS CONTAINMENT AIR LOCKS 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 cf less than or equal to 0.05 L, at (SG psig). P,, 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 Icck 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 STAND 8Y within the next 6 hours and in COLD SHUTOOWN 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, maintain at least one air lock door closed; restore the inoperable air lock to OPERABLE status within 24 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
J > W-DUAL 3/4 6-60 MAR 151978 ,
r l l l CONTAINMENT SYSTEMS SURVEILLANCE RE0UIREMENTS
- 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,V41(54 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 rate is within its limit:
- 1. At least once per 6, months,# and s_ pcsmAar
- 2. Prior to establishing' CONTAINMENT INTEGRITY when maintenance has been performed on the air lock that could affect the air lock sealing capability.*
/ c. At least once per 6 months by verifying that only one door in each
, air lock can be opened at a time.
The provisions of Specification 4.0.2 are not applicable.
- Exemption to Appendix J of 10 CFR 50.
\
W DUAL 3/4 6-70 ; SEP 151981 l
/
\CONTAINMENTSYSTEMS /
C .AINMENT ISOLATION VALVE AND CHANNEL WELD PRESSURIZATION SYSTEMS (OP 1AL) LIMITING ONDITION FOR OPERATION s 3.6.1.4 The c tainment isolation valve and channel weld pressu zation systems shall be PERABLE. APPLICABILITY: MOD 1, 2, 3 and 4. ACTION: With the containment,isol ion valve or channel weld ressurization system inoperable, restore the ino rable system to OPERABL status within 7 days or be in at least HOT STANDBY wi in the next 6 hours nd in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE REQUIREMENTS - s , 4.6.1.4.1 The containment isolation v ve p essurization system shall be demonstrated OPERABLE at least once p r 31 day by verifying that the system is pressurized to greater than or e ual to 1.10 (55 psig) and has adequate capacity to maintain system press re for at least 3 days. 4.6.1.4.2 The containment ch nnel weld pressurization 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 Pa (50 psig) and has dequate capacity to maintain system pres re for at least 30 days, s MAR 151978
-DUAL - 2/4 E C D --
- - . - . . . . - . _ _ . = _ _ - - _
. CONTAINMENT SYSTEMS
{ INTERNAL PRESSURE LIMITING CONDITION FOR OPERATION
.3. 6.1. ( Primary containment internal pressure shall be maintained between lakr and hths psig.
A'PPLICABILITY: MODES 1, 2, 3 and 4. ACTION: ! i 1 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 SHUTDOWN within the following 30 hours. 1 f SURVEILLANCE REQUIREMENTS I 4.6.1. The primary containment internal pressure shall be determined to be within the limits at least once per 12 hours. ( 7 W-DUAL 3/4 6-10 MAR 151978 '; ~
CONTAINMENT SYSTEMS AIR TEMPERATURE 8 LIMITING CONDITION FOR OPERATION 5 3.6.1.4 Primary containment average air temperature shall not exceed /20 F APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: O 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 STAND 8Y within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE REOUIREMENTS g 4.6.1.1 The primary containment average air temperature shall beweihcd r the....caritg +ticci average of the temperatures et the fcD cwia; leceHaae and shall be determined at least once per 24 hours: 7 1s cle+ermmed by o. minima,m of 3
-Lecation ir% of -the temperadwe. de.tec.hn s dsect kr Type. A conbn.nen4 +esh, h
b.
- c. \
d. e.
/
( 1 ! l W-DUAL 3/4 6- JUL 151979
CONTAINMENT SYSTEMS CONTAINMENT VESSEL STRUCTURAL INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.1. The structural integrity of the containment vessel shall be maintained at a level consistent with the acceptance criteria in Specification 4.6.1.'%. 6 APPLICABILITY: MODES 1, 2, 3 and 4. 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 'r 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-tion of the containment vessel detected during the above required inspections shall be reported to the Commission pursuant to Specification 6.9.1 2.
I 10 W-0UAL 3/4 6-MO DEC 151978 I
I CONTAINMENT SYSTEMS 1 CONTAINMENT VENTILATION SYSTEM ' LIMITING CONDITION FOR OPERATION 7 3.6.1.\ Each containment purge supply and exhaust isolation valve shall be OPERABLE and: 36
- a. Each [4R,-inch] containment shutdowr. purge supply and exhaust isolation valve shall be closed and sealed closed, and '
- b. The [8-inch] containment purge supply and exhaust isolation valve (s) may be open for up to [1000] hours during a calendar year provided no more than one pair (one supply and one exhaust) are open at one time.
APPLICABILITY: MODES 1, 2, 3; and 4. ACTION:
- a. With a ( .-inch] containment purge supply and/or exhaust isolation valve open or not sealed closed, close and/or seal close that valve or isolate the penetration (s) within 4 hours, otherwise be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
- b. With the [8-inch] containment purge supply and/or exhaust isolation valve (s) open for more than [1000] hours during a calendar year, close the open [8-inch] valve (s) or isolate the penetration (s) within 4 hours, otherwise be in at least HOT STANDSY within the next 6 hours, and in COLD SHUTDOWN within the following 30 hours.
- c. With a containment purge supply and/or exhaust isolation valve (s) having a measured leakage rate in excess of the limits of Specifica-tions 4.6.1.8.3 and/or 4.6.1.8.4, restore the inoperable valve (s) to OPERABLE status within 24 hours, otherwise be in at least HOT STANOBY within- the next 6 hours, and in COLD SHUTDOWN within the following 30 hours.
W-00AL t\ 3/4 6-120
( CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS 3(, w h 7 4.6.1.1 1 ntainment purge supply and exhaust isolation valves shall be verified to beilocked closed at least o nce per 31 anys. g
-0. O k cd at lee 5t once per 24 heues.
- 4. Scaled cle ed :t least crice pe. 21 dcyn 7
4.6.1.1.2 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.
- 4. 6.1. .3 t least once per 6 months on a STAGGERED TEST BASIS each sealed closed (42 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,.
4.6.1. 4 At least once per 3 months each (8 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,. 3\L{ V0
CONTAINMENT SYSTEMS 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS CONTAINMENT SPRAY SYSTEM LIMITING CONDITION FOR OPERATION 3.6.2.1 Two independent containment 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- ' GM) With one containment spray syste (inoperable, restore the inoperable spray system to OPERABLE status within uns or be in at least HOT STANDBY within the next 6 hours; 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, power operated or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.
ns ec.
- b. By verifying. that nn recirrlatica 'hu , each pump & vp! cps aaMe *C develop %g discharge pressure of greater than or equal to ider psig when tested pursuant to Specification 4.0.5.
- c. 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 a , m.. _. .,....est
__ _t signal.
- 2. Verifying that each spray pump starts automatically on a
,._y, - ~ m 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. ,
( l [3 JAn 151979 W-DUAL 3/4 6-MD
l C TAINMENT SYSTEMS 3/4 .2 DEPRESSURIZATION AND COOLING SYSTEMS CONTAI ENT SPRAY SYSTEM (No credit taken for iodine removal) LIMITING C0 DITION FOR OPERATION
/
3.6.2.1 Two in Sendent containment spray systems shall be 0 RABLE with each spray system capa e of taking suction from the RWST and tra sferring suction 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, rest e the inoperable spray system to OPERABLE st us within 7 day 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 h or be in at least HOT STANDBY within the next 6 hours and i COL SHUTDOWN within the following 30 hours. Restore both spray s stems t OPERABLE status within 7 days of initial loss or be in a least HOT ANDBY within the neXt 6 hours and in COLD SHUTDOWV within the fo owing 30 hours.
- c. With one containment sp ay system inoperabl and one group of required containment cooling f s inoperable, restore ither the inoperable spray system or the 'noperable 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 C0 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 UTOOWN within the f 11owing 30 hours.
SURVEILLANCE REQUI EMENTS
/
\
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 - :/4 O le h!AR 151978
A
'e CbMIAINMENT SYSTEMS
,s',,-
/
SURVEILLANCE UIREMENTS (Continued)
/
- b. By verifying, that /l develops a recirculation flow, each pump discharge pressure of ater than or equ t'to psig when tested pursuant to Specification 0.5.
- c. At least once per 18 months, dur' shutdown, by:
- 1. Verifying that each a tematic valv
- n the flow path actuates
, to its correct pos' on on a t signal, and 1!. Verifying th each spray pump starts automati lly on a est signal.
- d. At least ceper,5yearsbyperf5rmingonairorsmokeflow st throu each spray header and verifying each spray nozzle is uno tructed.
( t l l l l l i
's W-DUAL 3/' s-1sn JUL 15 B79 1
i CONTAINMENT SYSTEMS f 1 SPRAY ADDITIVE SYSTEM LIMITING CONDITION FOR OPERATION 3.6.2.2 The spray additive system shall be OPERABLE with: ~
- a. A spray additive tank containing a volun.e of between 1500 and 9150 gallons of between 19 and al percent by weight NaOH g
solution, and
% -Twc cpray additive eductor-r--eaeh-eepable of cdding NatWscluuva
-ficin tiic chendcal cdditive tank te centcinment cprey syste p"=p -
fisw. APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: g h dditive system inoperable, restore the system to OPERABLE Withthesprayh.2heurcorbeinatleastHOTSTANDBYwithinthenext6ho status within urs; restore the spray additive system to OPERABLE status 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. (om At least once per M-days by v@erifying 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 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.
- d. At least once per 5 years by verifying b solution flow -rcte (tc bc deterined#-dur#ng pre-c eretic..ai i.c a d from the -fe!!c'? ng drain I
-cennectient the spray additive cyctcm- bk and -the RWST.
- 1. -{0rair ' 41: Ivuauun) ; gpm %4 ces-Vm *v%(+m S47)
- 2. -(Orain 'i ne locatien) +
C " TAmd cas-vy +VW. (% gws7) W _-DUAL 3/4 6- 5M
40NTAINMENT SYSTEMS C0 INMENT COOLING SYSTEM (OPTIONAL) (Credit taken for iodine removal b spray ystems) LIMITING ONDITION FOR OPERATION 3.6.2.3 (T 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'ng 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 COLD
, SHUTDOWN within the foll wing hours. Restore both above required if groups of cooling fans to PE ABLE 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 e re uired containment cooling fans inoperable and one containment spr systei inoperable, restore the inoperable spray system to OPERAB E status ' thin 72 hours or be in at least HOT STANDBY within t next 6 hour 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 REOUIREME S
/ \
4.6.2.3 Each grou of containment cooling fans shall be monstrated OPERABLE:
- 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.
b At least once per 18 months by verifying that each fan group
', starts automatically on a test signal.
_-DUAL - 2 " 91'D IdAR I 5 L
. i
/
a CO INMENT SYSTEMS CONTAIhMENT COOLING SYSTEM (OPTIONAL) (No credit taken for iodine remov'al by spray sy tems) LIMITING CON TION FOR OPERATION
\
3.6.2.3 (Two) i. ependent groups of containment cooling fans hall be OPERABLE with (two) fan sys ms to each group. (Equivalent to 100% co ling capacity.) APPLICABILITY: MODES , 2, 3 and 4. ACTION:
- a. With one group of he above required conta' ment cooling fans inoperable and both containmen spray systems OPERA E, restore the inoperable group of cooling fans to OPERABLE statu within 7 days or be in at least HOT STANDBY with the next 6 ho rs and in COLD SHUTDOWN within the followin'g 30 ours,
- b. With two groups of the abo requir d containment cooling fans inoperable, and both contain ent ray systems OPERABLE, restore at least one group of cooling fa o OPERABLE status within 72 hours or be in at least HOT STANDBY hin the next 6 hours and in COLD SHUTDOWN within the following 0 curs. Restore both above required groups of cooling fans to OP RABLE 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.
- c. With one group of the a ve required con inment cooling fans inoperable and one containment sp ay system inoperab , restore either the inoperable group of c ntainment cooling fan or the inoperable spray system to OPERABLE atus within 72 hours or e in at least HOT STANDBY within the ext 6 hours and in COLD SH TDOWN within the l following 30 hour . Restore both the inoperabl group of containment
! cooling fans an the inoperable spray system to C ERABLE status within 7 days initial loss or be in at least H0 STANDBY within the next 6 ho s and in COLD SHUTDOWN within the fo owing 30 hours. SURVEILLANCE REQUIRE NTS
/ \
4.6.2.3 Each gro p of containment cooling fans shall be demonstrate OPERABLE:
- a. At 1 st once per 31 days by:
- 1. Starting each fan group from the control room verifying th t each fan group operates for at least 15 minutes.
- 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.
~ _-DUAL - ;, 4 o- 180 MAR 151979
/
C TAINMENT SYSTEMS / 3/4 3 IODINE CLEANUP SYSTEM (OPTIONAL) /
)
LIMITING NOITION FOR OPERATION 3.6.3 Two in endent containment iodine cleanup systems h 1 be OPERABLE.
~
APPLICABILITY: ES 1, 2, 3 and 4. ACTION: With one iodine cleanu system inoperable, restore the inoperable system to OPERABLE status within days or be in at least BdT STANDBY within the next 6 hours and in COLD SHUTDOW within the following/30 hours. SURVEILLANCE REOUIREMENTS x .
/
/
/
4.6.3 Each iodine cleanup system hal e demonstrated OPERABLE:
- a. At least once per 31 days 9 a STAGGERED TEST BASIS by initiating,
, from the control room, fl 96 rough the HEPA filters and charcoal adsorbers and verifying J, hat t e system operates for at least 10 hours with the heaters vn. .
b.
/
At least once per 18jhonths or (1) fter any structural maintenance on the HEPA filter of charcoal adsor er housings, or (2) following i painting, fire'or' hemical release in ny ventilation zone communicating with the system b :
- 1. Veri fying hat the cleanup system sat fies the in place testing acceptan e criteria and uses the test p cedures of Regulatory Positio s C.5.a. C.S.c and C.S.d of Regu tory Guide 1.52, Revisi n 2, March 1978, and the system flo rate is cfm 1 10*.
- 2. Ve ifying within 31 days after removal that a 1 boratory 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 m operation when tested in accordance with ANSI N510-1975.
W-DUAL 4 /, 6-isu - JUN 1 1979
/
TAINMENT SYSTEMS
)
SURVEILL CE 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 .6.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 months 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 move 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 o stem at a flow rate of cfm 1 10*.
- f. After each compi e or partial replacement o a charcoal adsorber bank by verify' g that the charcoal adsorbers' emove greater than or 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%.
A 99.95% plicable when a filter efficiency of 99% assumed in the safe l analy s; 99% when a filter efficiency of 90% is assumed.
\ .
APR 15197 ! FOUAL N /4 0 200 %
i CONTAINMENT SYSTEMS 3 3/4.6.% CONTAINMENT ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3 later s
- 3. 6.% The containment isolation valves specified in " O " ' shall be VF OPERABLE with isolation times as shown in Table-376 2.
_fo,- em vaiv e.. APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: With one or more of the isolation valve (s) specified-irr-fatrteWr-t 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
- 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 STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
SURVEILLANCE REQUIREMENTS 4.6. 1 The isolation valves pecif-ied ir. Table 3.5 1 shall be demonstrated OPERABLE prinr 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. k if W-DUAL 3/4 6- 3 JUN 1 1973
CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 3
- 4. 6.4. 2 Each isolation' valve specified i Table 3.0 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 containm.ent 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 B 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 td"- Tebiu 3.0-2 shall be determined to be within its limit when tested pursuant to Specification 4.0.5. (. i l l i l l l IG W-DUAL 3/4 6-2ED SEP 2 81981
_. ~s s, , o E 'N TABLE 3.6-2 e K CONTAINMENT ISOLATION VALVES j,
,/
MAXIMUM f VALVE NUMBER FUNCTION ISOLATION TIME (Seconds) A. PHASE "A" ISOLATION 1 \ 2- \ %., O f 3-
- 8. P PHASE "B" ISOLATION N.
-1. N _/3
'N A- -6
- 2. ' '/[%
C. CONTAINMENT PURGE ANO \ T EXHAUST p %. . g's 1. ft 2 ,/ - g '%. D. MANUAL l' /
/ [
/ m '
- 2. / b ',
E. OTHER j N 1.
/
/ L %
, 2/ .
E / ' .s , no / ' [ /*May be opened on an intermittent basis under administrative control. w .- y s\ 8 j Not subject to Type C leakage tests. \
/ **The provisions of Specification 3.0.4 are not applicable.
l
/
l
r CONTAINMENT SYSTEMS . 3/4.6. COMBUSTIBLE GAS CONTROL HYDROGEN MONITORS LIMITING CONDITION FOR OPERATION 4 3.6.4.1 Two independent containment hydrogen monitors shall be OPERABLE. 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 REOUIREMENTS tj 4.6.4.1
@ d$^4D Each hydrogen monitor shall be demonstrate l PERABLE by the h'
performance of a CHANNEL CHECK at least once per uns, a ANALOG CHANNEL OPERATIONAL TEST at least once per 31 days, and at least once per 92 days on a 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 19 W-0UAL 3/4 6- N SEP 151981
CONTAINMENT SYSTEMS ELECTRIC HYDROGEN REC'OMBINERS - W LIMITING CONDITION FOR OPERATION il 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 4.6. 2 Each hydrogen recombiner system shall be demonstrated OPERABLE:
- a. At least once per X months by verifying during a recombiner system functional test that the minimum heater sheath temperature increases to greater than or equal 700*F within 90 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.
- b. Atleastonceper18moni.hsby:
- 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 (i.e., loose wiring or structural connections, deposits of foreign materials, etc.), and i
- 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.
! 3 W-DUAL 3/4 6- AUG 6
t CO AINMENT SYSTEMS HYDR PURGE CLEANUP SYSTEM (If less than two hydrogen recombiners available j LIMITING C0 ITION FOR OPERATION 4
~
i 3.6.5.3 A conta'nment hydrogen purge cleanup system shall be OPERAB and , capable of being owered from a minimum of one OPERABLE emergency s. j , APPLICABILITY: MOD 1 and 2.
- ACTION
A With the containment hyd gen purge cleanup system inoper le, restore the l hydrogen purge cleanup sys em to OPERABLE status within days or be in at
. least HOT STANDBY within 6 ours.
! SURVEILLANCE REQUIREMENTS 1 <
/
/
4.6.5.3 The hydrogen purge cleanu system sha be demonstrated OPERABLE: 1
- a. At least once per 31 days b init' ting, from the control room, flow j , through the HEPA filters and ha coal adsorbers and verifying that i
the system operates for at lea 10 hours with the heaters on. i i 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 painting, fire, or chemic release any ventilation zone communicating with the stem by: . 1. Verifying that e cleanup system sa isfies the in place 4 testing ac.ept ice criteria and uses e test procedures of Regulatory P itions C.5.a, C.5.c and .5.d of Regulatory
- Guide 1.52, svision 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 i
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 j uide 1.52, Revision 2, March 1978.
- 3. Verifying a system flow rate of cfm i 10% during s stem 4
operation when tested in accordance with ANSI N510-1975. i 1 t, AUG 6 81 DUAL - 3/; ; 'yn - 1 l
NTAINMENT SYSTEMS SURVE LANCE s REQUIREMENTS (Continued)
- 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 saraple 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. Verify 1 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 + 10%.
- 2. Verifying tha the filter coolin ypass valves can be manually opened.
- 3. Verifying that th heaters di sipate 1 kw when tested in accordanc with A I N510-1975.
i
- e. After each complete or par a replacement of a HEPA filter bank by I
verifying that the HEPA filt r 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 repl 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%.
R 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. i \ _-0 VAL 3/? W
CONTAINMENT SYSTEMS HYOR0 GEN MIXING SYSTEM LIMITING CONDITION FOR OPERATION 3.6.'4.3 i.4 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 43 4.6.'S.4 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 infer efm.
l I ( 10 W-0UAL 3/4 6-2Ru 6 1931 AUG
/
[ ONTAINMENT SYSTEMS [ 3 6.6 PENETRATION ROOM EXHAUST AIR CLEANUP SYSTEM (OPTIONAL) LIMITIN CONDITION FOR OPERATION 3.6.6 Two in ependent containment penetration room exhaust cleanup systems shall OPERABLE. APPLICABILITY: . ES 1, 2, 3 and 4. ACTION: With one containment pe tration room exhaust air eanup system inoperable, restoretheinoperablesytemtoOPERABLEstatu/s ithin 7 days or be in at least HOT STANDBY within tN next 6 hours and i COLD SHUTDOWN within the following 30 hours. T SURVEILLANCE REQUIREMENTS x ,
/
4.6.6 Each containment penetration r m exhaust air cleanup system shall be demonstrated OPERABLE: f
- a. At least once per 31 day on a GGERED TEST BASIS by initiating, from the control room, ow throu the HEPA filters and charcoal adsorbers and verifyi that the sy tem operates for at least 10 hours with the heate on.
- b. Atleastonceper/8monthsor(1)afte any structural maintenance on the HEPA filtpf or charcoal adsorber usings, or (2) following painting, fire Y chemical release in any entilation zone com.aanicating with the syst by:
- 1. Verify g that with the system operating t a flow rate of cfm 0% and exhausting through the HEPA ilters and charcoal adso ers, the total bypass flow of the sys em to the facility vepf., including leakage through the system 'verting valves, is less than or equal to 1% when the system is t sted by admitting dold 00P at the system intake. (For systems w th diverting valves.)
2 Verifying that the cleanup system satisfies the in place testing acceptance criteria and uses the test procedures of Regulatory Positions C.S.a, C.5.c and C.5.d of Re latory Guide 1.52, Revision 2, March 1978, and the system flo rate is cfm + 10%. l. _ 0 VAL 4-au JUN I in i
* /
/
kONTAINMENTSYSTEMS ( SURV LLANCE REQUIREMENTS (Continued)
/
Verifying within 31 days after removal that a laboratg y analysis of a representative carbon sample obtained in accord hce with Regulatory Position C.6.b of Rcgulatory Guide 1.52, evision 2, arch 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 10 during system operat'on when tested in accordance with ANS N510-1975.
- c. After every 72 hours of charcoal adsorber oper tion by verifying within 31 days a ter removal that a laboratory analysis of a repre-sentative carbon s mple obtained in accordan with Regulatory Position C.6.b of R ulatory Guide 1.52, Re ision 2, March 1978, meets the laboratory sting criteria of gulatory 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 combined HEPA filters and charcoal adsor er anks is less than (6) inches Water Gauge while operating he system at a flow rate of (
cfm i 10%.
- 2. Verifying that the syst starts n a Safety Injection Test Signal.
- 3. Verifying that the f iter cooling byp ss valves can be manually opened.
- 4. Verifying that e heaters dissipate 1 kw when tested in acco dance with ANSI N510-1975.
- e. After each compl e or partial replacement of HEPA ilter bank by verifying that he 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 fl rate of cf 10%. -
- f. After ea 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 whe they are tested in place in accordance with ANSI N510-1975 wh e operating the system at a flow rate of cfm 1 10%. 99.95% plicable when a filter efficiency of 99% is assumed in the i safety analyses; 99% when a filter efficiency of 90% is assumed. I W- AL e-M16- C -
/
.f TAINMENT SYSTEMS h j
) , 3/4 VACUUM RELIEF VALVES [ l LIMITING C0 ITION FOR OPERATION 3.6.7 The primar containment to atmosphere vacuum relie valves shall be OPERABLE with an ac ation set point of less than or egyll to psid. , APPLICA8ILITY: MODES 2, 3 and 4. / - ) ACTION: I With one primary containment atmosphere v uum relief valve inoperable, restore the valve to OPERABLE s tus within hours or be in at least HOT STAND 8Y with the next 6 hours and 'n COLD HUTDOWN within the'following l- 30 hours. i SURVEILLANCE REQUIREMENTS
- s
?
( 4.6.7 No additional Su eillance Requirements other t, n those required by Specification 4.0.5. i r .a i l J r i 1 2 L( , 4 t i W-DUAL __ m 641D-- JUN 1 3 79
.m___.-____ _ _ _ _ _ = - - -
t CONTAINMENT SYSTEMS h ,
-Suygtc93;t3733Ig7gg;773/,.f,g,5- CONTA/Mt16ATT ENCLOSURE 6 u t t D/N(>-
COMTA D Mt1ENT GNCLOS MR2 Su lLD/Nto- ZNTEGRITY LIMITING CONDITION FOR OPERATION y, g teourmuneur eactosuasi
- 3. 6.1. '& ';;;IELG BUILDING INTEGRITY shall be maintained.
APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: s couranneur cNeLw21 \ ecurs,angur cNCLOSUA() ~J bg 3 Without-S"!ELDVBUILDING INTEGRITY, restore Sti!ELO'8UILDING INTEGRITY within,M vhours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN
.( .within the following 30 hours.
~
SURVEILLANCE REOUIREMENTS 5., , tcourninuser Enscuosu As)
- 4. 6.'&.'8 V 5"IELO BUILDING INTEGRITY shall be demonstrated at least once 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.
4 7,1 W-DUAL 3/4 6- M SEP 1 1979
CONTAINMENT SYSTEMS
--3/-L G _ E 3E00%AF/ C0!MIM4EMT--
-SHIELO B"ILDING AIR CLEANU? 25 TEM -
COMTFtl N rlEM T E N C1. cst d E. E/1 cst 4-Eucy EX H At4ST 4 aid C,dokr46- S YSTF>f LIMITING CONDITION FOR OPERATION - S' 1_ tcodain-eteacM5"*e e=evev cAusAJ cool 8"9 / 3.6. $ Two independent Shield Ouilding f#- Cle:nufSystemsshallbeOPERABLE. APPLICABILITY: MODES 1,2,i,and4. ACTION: ves%nned enefasuri ememency thaudand wel**91 With one SPicid Suilding ^4- Cleanu?Systeminoperable,restoretheinoperable l system to OPERABLE status within 7 days or be in at least HOT STAN08Y within the next 6 hours and in COLD SHUTOOWN within the following 30 hours. SURVEILLANCE REOUIREMENTS uoak-Amc.d wiosun e ewm ency <*uvf ad co tuiy 1 4.6.8.1 Each ......_ __......, " c!e:ne;Mystem shall be demonstrated OPERABLE: .
- a. At least once 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 continuous hours,with the 50:ters a p r: ting;-
- 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 communi-cating with the system by:
- 1) Verifying that the cleanup system satisfies the in place pene-tration and bypass leakage testing acceptance criteria of less
*]% and uses the test procedure guidance in Regulatory Positions C.S.a. C.S.c, and C.5.d of Regulatory Guide 1.52, Revision 2, March 1978, and the system flow rate is later cfm 10%;
- 2) Verifying, within 31 days after removal, that a laboratory analysis of a representative carbon sample obtained in accord-ance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revi-sion 2, March 1978, for a methyl iodide penetration of less than
*]%; and ,
- 3) Verifying a system flow rate of Igh.r cfm 10% during system j cperation when tested in accordance with ANSI N510 1 Z .
IWO W-0UAL 3/4 6 K@L
4 CONTAINMENT SYSTEMS D _ SURVEILLANCE REQUIREMENiS (Continued) 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, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978, for a methyl , iodide penetration of less than [**]%:
- d. At least once per 18 months by: *
- 1) Verifying that the pressure drop.across the combined HEPA filters and charcoal adsarber banks is less than (-6-3 inches Water cfm + 10%, Gauge while operating f.he system at a flow rate of Mgy-I
- 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 I greater than or equal to [0.25] inch Water Gauge in the annulus within % g nute after a start signal, and l
-!P)- -Vcrifyi6g t.iet the hauer= n W pe*=
terted 4. ;;cerder.cc with "S! Z10-1575.- 1 I i,u %7, g e. > After each complete or partial replacement of a HEPA filter bank, by verifying that the cleanup system satisfies the in place pene r ion and bypass leakage testing acceptance criteria of less than ( % in d accordance
, the system atwith ANSI a flow ra N510-M74 for a 00P test aerosol while operating o jgig,cfm210%;and l
- f. After each complete or r ial replacement of a charcoal adsorber bank, by verifying that the cleanup system satisfies the in place 0"
tha netration and bypass leakage testing acceptance criteria of less
*]% in accordance with ANSI N51049Mr for a halogenated hydro-carbon refrigerant test gas while operatin the system at a flow rate of la te.v cfm 2 10%, tigo I
*0.05% value applicable when a HEPA filter or charcoal adsorber efficiency of 99% is assumed, or 1% when a HEPA filter or charcoal adsorber efficiency of 95% or less is assumed in the NRC staff's safety evaluation. (Use the value assumed for the charcoal adsorber efficiency if the value for the HEPA filter is different from the charcoal adsorber efficiency in the NRC staff's safety evaluation).
**Value applicable will be determined by the following equation:
P=1 -E , when P equals the value to be used in the test requirement (%), E is efficiency assumed in the SER for methyl iodide removal (%), ( and SF is the safety factor to account for charcoal degradation between tests (5 for systems with heaters and 7 for systems without heaters). ! W-00AL 'A3 3/4 6-240 I
CONTAINMENT SYSTEMS s COAlTAINt16MT E MC Lost 4 A F_f
-5HIELC BUILDING STRUCTURAL INTEGRITY h LIMITING CONDITION FOR OPERATION 5- s conk'nment eclowe e ,
3.6 4 3 The structural integrity of the chic!d building V shall be maintained at a level consistent with the acceptance criteria in Specification 4.6.1 3. s' APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: uontainment enclosctre, With the structural integrity of the d ie4d' building not conforming to the above requirements, restore the structural integrity to within the limits within EL4= hour:s or be in at least HGT STANDBY within the next 6 hours and in CO,LD SHUTDOWN ^within the following 30 hours. 3 Eas{s SURVEILLANCE REOUIREMENTS f teen 6'inment enelowr e, t 4.6.4.3 The structural integrity of the :Pic!fbuilding shall be detemined during the shutdown for each Type A containment leakage rate test (reference Specification 4.6.1.2) byyisu,ahj,nsgegon of the exposed accessible interior and exterior surfaces of we u :::muui euing and verifying no apparent changes in appearance of the con Any abnormal degradation of i.ne ,cggs,ugf,acg,ogo,ther abnormal er m:moui numg detected degradation. during the above required inspections shall be reported to the Commission pursuant to Specification 6.9.% v's % LS' A a<js L I. M W-DUAL 3/4 6- b ApR 3 0 m
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 gpm. 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 must-be taken if the pump house becomes inoperable.
- 1. The extended allowed outage time for the cooling tower is based on its low importance relative to the service water pumphouse. A detailed evaluation of this change will be provided in the risk-based analysis to be submitted in mid August.
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 gpm each. L. Seabrook has two diesel driven fire pumps and one motor driven fire pump. 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. P. Added per commitment letter SBN-399 dated December 3, 1982 to the NRC. N .
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. S. The Surveillance Requirements are deleted and will be included in a licensee maintained and controlled document. T. This Table'is deleted and will be included in a licensee maintained and controlled document. t. U. .This surveillance is deleted in order to avoid unnecessary diversion of the operators. The condition of 120*F in the Contr'ol Room will be obvious
- to the operators without any surveillance.
V. This change. serves to clarify the action statement to reflect the design of the Service Water System at Seabrook. 4
3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE SAFETY VALVES LIMITING CONDITION FOR OPERATION 3.7.1.1 All main steam line code safety valves associated with each steam generator-cf 2n unicciated reacter ccclant !cep shall be OPERABLE with lift settings as specified in -Table =3=7e'S? @
~'~
leher c APPLICABILITY: MODES 1, 2 and 3. W ACTION:
- a. WithOk)reactorcoolantloopsandassociatedsteamgeneratorsin operation 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) reacter coolant loop: and :::cciated stccm gancrators in-
-operaticr 2nd "ith one or marc main steam line ccdc safety valve -
-associated with ai, cper& ting loop ncper:ble, cperatica in MODES 1, i
2 and 3 may proceed provided, that eith4- ' hcur;, cither the (h) i ncperable valv is restored to OPEP^.BLE ct:tu cr the Pcwcr Pang;
-Mcutrcn flux "igh Trip Setpcint is reduced pe" Table 3.7-2;-ctherwise, . . ._
u .. , ,_ un, , , . . . - . . . .. . - - nn,n w a si uw acc34 IIV O JIMOUQ3 W I billll bild Ilt A L Q lluul 3 Gi f u its b u t.U 4""TDOWN within the fellcwing 30 heur:. b NL The provisions of Specification 3.0.4 are not applicable. SURVEILLANCE REQUIREMENTS W&i-Ho-addit 4ona4-Survei+1ance REqvf rements-otherthart-t4tose-required- by
--Specif-ica t-ion-4:0-5.
1 l 1
-W-STS 3/4 7-1 l MAR 151979 l l
l
1 i TABLE 3.7-1 ' MAXIMUM ALLOWABLE POWER RANGE NEUTRON FLUX HIGH SETPOINT WITH INOPERABLE STEAM LINE SAFETY VALVES DURING % LOOP OPERATION 4 Maximum Number of Inoperable Maximum Allowable Power Range Safety Valves on Any Neutron Flux High Setpoint Operating Steam Generator (Percent of RATED THERMAL POWER) 1 . (87) 2 (%) fos 3 (42-) 4 3 TACLE 3.7 2
\ MAXIMUM ALLOWABLE POWER RANGE NEUTRON FLUX HIGH SETPOINT M '
i :-
- MPERABLE STEAM LINE SAFETY VALVES OURING N-1 LOOP OPJRATION s Maximum Number of In erable Maximum Allo e Power Range Safety Valves on Any Neutro ux High Setpoint Operating Steam Generator (Perce of RATED THERMAL POWER) 1 (52) 2 -(38) 3 (25)
, *At least tw safety valves shall be OPERABLE on the non-operating am generato
-\ k W-STS 3/4 7-2 MAR 151979
. ,, m:c .-
.m D TABLE 3.7-STEAM LINE SAFETY VALVES-PER LOOP'
\
VALVE NUMBER LIFT SETTING.(1 1%)* ORIFICE SIZ
\
- a. V(, V2L V3bV50 / IFS- psig "R . /4 s2 in )
's. ,
b ., V 1, V23; V3% VSI 's - /2O 3 psig "R [//, se /o ) c. [ i VF, V24. V3fP) VS2. 'N s /2.10 psig y' R //4 sa v in ) I d.
/<
y V9, V2S, V39,. V53 123ff /psig "R[/Le.4)
/
Q c. vio,vas,vqo: v54 /ai$ ' g,3 " R (/4 %iol w y - e
/*
./
"The lift setting pressure shall correspond to ambient conditions of the valve'at nominal operating temperature and pressure. ,/
l
/ '
,r
.N.,
,r %
f . * ., M i i '.. 1 E 1 2n
- o j * '
& a' i
PLANT SYSTEMS l'NGRG Cac.1 J"!LI ARY FEEDWATER SYSTEM LIMITING CONDITION FOR OPERATION , two @ erncv @ 3.7.1.2 At least -44 wee- independent steam generator cux9encyicry feedwater pumps and associated flow paths shall be OPERABLE with: One eme
@ tor-driven aux,vgenc.y
- a. Iwo mo inry feedwater pumps,-each capable of being poweredfromsegrat: emergency buss 1!$, and entv
- b. One steam turbine-driven aux qqncy::ry feedwater pump capable of being powered from an OPERABLE steam supply system.
APPLICABILITY: MODES 1, 2, and 3. ACTION:
- a. With one emeaux;ygency $ pump inoperable, restore the required cux;.ge.ncy
- ry feedwater , e>ner
,;;ry 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.
emevq ency
- b. Withtwozuxiliarjfeedwaterpumpsinoperable,[beinatleastHOT STANDBY witin 6 hours and in HOT SHUTDOWN within the following 6 hours. j
,% With thrcc cuxi'i ~ 4+2ter = n ir. :cr
fcorrective action to restore at' least one Y?j E?@r feedwaterkmmediately pump ini I to OPERABLE status as soon as possible.or - t SURVEILLANCE REQUIREMENTS emer ency 4.7.1.2 Each aux'gisry feedwater pump shall be demonstrated 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 greater than or equal to lofer gpm.
- 2. Verifying that the steam turbine driven pump develops a discharge pressure of greater than or equal to h psig at a flow of greater than or equal to lqfer gpm when the secondary steam supply pressure is greater than leder psig. The provisions of (
Specification 4.0.4 are not applicable for entry into MODE 3. l W-STS 3/4 7-4 AUG 7 1980
f 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.
- 4. Verifying that each automatic valve [in 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 :=iliary feedwater actuation test signal. * *'*"i C^C Y
- 2. Verifying that each eme;gtncy::ry nux feedwater pump starts as designed automatically upon receipt of an eu "is y feedwater actuation test signal. e=cv3ency C.
Gt lead once per 18 mon %s durtng shutdown , or following comple.flon of vnodiCical 'ons to he EFU) s y>fe m s to bich a l f er s n e sysfems flow cAaro.ch ertsN c3 , b y verifyin9 flow 4 rom the condensafe.
*"96 tan So eat.$ Sh ea m 9 tntvakov.
1 o l k , y-STs 3/4 7-s Jut. 2 31980 l l
}
l PLANT SYSTEMS i 1 CONDENSATE STORAGE TANK I LIMITING CONDITION FOR OPERATION 3.7.1.3 The condensate storage tank (CST) shall be OPERABLE with a contained water volume of at least 2to,000 gallons of water. APPLICABILITY: MODES 1, 2, and 3. ACTION: With the condensate storage tank inoperable, within 4 hours either-1L. 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 16 - 00:en:trate the OPERABILITY cf the (:! ternate water : urec) 2: ;-
-beukup-tepply +n the eux 4 '!:ry feedwater pu ps end rectere the
-cond n::te sterage-tank-te OPERASLE :tatu: with*
- day er 50 #a at leest "0T STf. NOSY within th; next C hsurs and i, unT qunTnnWN within
- the f0!!e"ing A heure. -
SURVEILLANCE REOUIREMENTS . 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.
-4.71.2.2 'h: (alternate water :surce) shell be J ~n>L.oted CPE"ABLE ot le :t once per 12 heu = Ly (=uthed dependent upcn alternate scurec) -tentacr-({}the(e1+arnatewate-seurce)is+ha cunnly enneca fer the aux #'i:ry feedente-9mP8.
s. W-STS 3/4 7-6 JUL 2 3 E60
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. 5 SURVEILLANCE REQUIREMEN15 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. W-STS 3/4 7-7 JUL 151979 i
i TABLE 4.7-1 SECONDARY COOLANT SYSTEM SPECIFIC ACTIVITY i SAMPLE AND ANALYSIS PROGRAM > TYPE OF MEASUREMENT SAMPLE AND ANALYSIS AND ANALYSIS FREQUENCY -
- 1. Gross Activity Determination At least once per 72 hours.
- 2. Isotopic Analysis for DOSE a) 1 per 31 days, when-EQUIVALENT I-131 Concentration ever the gross activity determination indicates iodine concentrations greater than 10% of the allowable limit, b) I per 6 months, when-ever the gross activity determination indicates iodine concentrations below 10% of the allow- .
able limit, t l l t r l l k W-STS 3/4 7-8 MAY 151976 l
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 nours 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;O seconds when tested pursuant to . Specification 4.0.5.Th 7 odtslom o g speeic{c 1,*e, ii.e.'l au e notgltcalle
- 4. e4y wb M00 E 3 9
W-STS 3/4 7-9 AUG 6 I!al
PLANT SYSTEMS 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: With 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 deterdine 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 SURVEILLANCE 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. l i i ! l W-STS
~ 3/4 7-10 JUN 1 1979
~
PLANT SYSTEMS 3/4.7.ghe R'fl 3 COMPONENT COOLING WATER SYSTEM LIMITING CONDITION FOR OPERATION
- s pn-wa-4 ,
3.7.3 ^.t le :t Iwo independentVcomponent cooling water loops shall be OPERABLE. APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: With only one Qwr component cooling water loop OPERABLE, restore at least two
- loops to OPERABLE status within 72' hours or be in at least HOT STANDBY within l the next 6 hours and in COLD SHUTOOWN within the following 30 hours.
SURVEILLANCE REQUIREMENTS
- 4. 7.3 9.t le::t twoe component
'r-a,vf 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 ,s. . .,........ test signal. [ l l l W-STS 3/4 7-11 MAY 151976
PLANT SYSTEMS i 3/4.7.4 SERVICE WATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.4 At least two independent service water loops shall be OPERABLE. 7 APPLICABILITY: MODES 1, 2, 3 and 4.
- ACTION:
With only one service water loop OPERABLE, restore at least two loops to OPERA 0LE status within 72 hours or be in at least HOT STANOBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE REQUIREMENTS 4.7.4 At lea:t 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 gg. g u test signal.
Ead sev vice wa.te> loor shall tnc.lu.de. eih. . . 7
- a. one semitce waf e* eamp bin ud tne serviteuxilev pumpkouge , og
- b. One coolO3 % )cv Wme kin Ad'The ccolty bu&
l l W-STS 3/4 7-12 MAY 151976
l' t PLANT SYSTEMS 3/4.7.5 ULTIMATE HEAT SINK LIMITING CONDITION FOR 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 datum, and-e !! e p th en eho Atlenti 0: an l
cf :uffici:at si:: te provida e fle" ret cf g ::ter th n 21,000-
-g, , :=d
- b. A mechanical draf t cooling tower comprised of two cooling tower fans and a contained basin water volume of equal to or greater than 4 x 10 6 gallons at an average temperature of less than 900F.
l APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: Wi% % servke wde" P"W kd"** 'W *'.= We, neb e % s e,vl< c to 0PEM 6LE m,+< ru,p ke=s e s+=+u s wh(n 71 hours e* h e en af l cast k f Han d y wsMt; M4 meyt ' t, hen,6 maa in cos.o s wroawu u Wn -m ne,,t so koa,3, [ a. WitF :n: ecoling teuer fan--inop:::ble, :::te e et !c::t-tec fan ,
-to-ORERABLE-et-ete: eith!= 72 has.-; ;; he in a: 1e::: ;0T4TANDay. h i
withir - the- ne::t 6 heure ead in C0in mf!00"" id in the fc!!cring 30 her:: .
- 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.
"% k-SURVEILLANCE REQUIREMENTS 4.7.5 The ultimate heat sink shall be demonstrated OPERABLE:
a. 14 h At least once per h2 hours by verifying the service water pumphouse water Icvel 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 on h a Tower Actuation test signal.
s I t 3/4 7-13 -
I PLANT SYSTEM 3/4.7.6 CONTROL ROOM MR- MAKEUP SYSTEM LIMITING CONDITION FCR OPERATION W system shall be-OP "s ir., comprised of: 3.7.6 The control room e4+ makeup
- a. Two OPERABLE remote air intakes.
- 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.
APPLICABILITY: ALL MODES ACTION: MODES 1, 2, 3 AND 4 With any control room eir makeup system redundant component inoperable, [' restore the inoperable component 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. MODES 5 AND 6 tan) 2'! :F anj ::atrol . -- air ;;kcup r;:::: reb " der cc penen* :
-inspcr:b!c, sc;tcre th; inspc.ot!; :: per::t : 0"EPJ.31.E ;;; t s: -
eithia ' 2:y; er iritiat: and meintain e,cr-*'a" ^' t'e c en t r a l-
.rc= tir rehc p y L. Er the r:eircul;;ic: ;;Jm.
n, gai, av h kle,ves4oee.een4 olroo % **keupal wdM 3okou'*8"f fir)- With bet 4,- control room 44 makeupVeye *=r ep- able , suspend all operations involving N ALTERATIONS or positive reactivity changes. Co M 6:-) %% rouisicen of Opccifiw iiv.. 3.0.3 eic not applicabi; A
-tiODC S-3/4 7-14
4 { PLANT SYSTEM SURVEILLANCE REQUIREMENTS 4.7.6 / The control room e make ystdm shall be detronstrated OPERABLE: g KE. 'At-hnrst ence yn 12-hours-by-ver-i-f-yingahat-theaantr.ol rocarair--- tcape ra t ure-i s-l e s t--th an--o r-equ a l-t o-1200 F 4 h. At least once per 31 days by verifying operation of each makeup fan. b x. 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-so%" gn^a 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 m - ed to the closed position.
- f. e p erded 4
5 i 3/4 7-15
\LANTSYSTEMS 3 7.6 FLOOD PROTECTION (OPTIONAL *)
LIMI G CONDITION FOR OPERATION x e 3.7.6 Floo protection shall be provided for all safety related syst 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 Me n Sea Level USGS datum:
- a. (Be in at least H0 STANDBY within 6 hours nd in at least COLD SHUTDOWN within the llowing 30 hours) d
- b. Initiate and complete w hin h rs, the following flood protection measures:
- 1. (Plant dependent)
- 2. (Plant dependent)
( SURVEILLANCE REQUIREMENTS I \ 4.7.6 The water level at sb 1 be determine to be within the limits by;
- a. Measurement at least nce per 24 hours whe t.he 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 spec ication not required if the facility design has adeq te passive flood c rol protection features sufficient to accommodate the sign Basis Flood i entified in Regulatory Guide 1.59, August 1973.
l l y-STS 2/' ' u - NOV 151977
Nst &phca61e. % se&och NLANTSYSTEMS 3 7.7 CONTROL ROOM EMERGENCY AIR CLEANUP SYSTEM LIMI CONDITION FOR OPERATION
\
/
3.7.7 Two in ependent 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 OPERABL status within 7 days or be in at least HOT STANOBY within the next 6 hour and in COLD SHUTDOW within the following 30 hours. MODES 5 and 6: /. a. With one control room emergen air leanup system inoperable, restore the inoperable system OP 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 ode by ACTIO (a), not capable of being powered by an OPERABLE emq gency power sou e, suspend all operations involving CORE ALTERATI0 S or positive react'vity changes.
SURVEILLANCE REQUIREMENTS 4.7.7 Each control room ergency 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 lea t once per 31 days on a STAGGERED TEST BASIS by initia ng, from e 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.
s
&S 3/d 7-15 JUL 2 71981
r A/of App lNa/>lc h Seab Voole ! PLANT SYSTEMS r
\ SURVEILLANCE REQUIREMENTS (Continued)
. 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 mmunicating with the system by:
- 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 ad rbers, the total bypass flow of the syst m to the facility vent 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 ms with diverting valves.)
- 2. Verifying that the cleanup system satt les the in place testing acceptance riteria and uses the test rocedures of Regulatory Positions C. a, C.S.c and C.5.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 of a representativ carbon sam 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 (
Position C.6.a of Regu tor Guide 1.52, Revision 2, March 1978. .
- 4. Verifying a system flow a e of cfm 1 10% during system operation when tested ac ordance with ANSI N510-1975.
- d. After every 720 hours of harcoal a sorber operation by verifying within 31 days after re val, that a laboratory analysis of a represen-tative carbon sample o tained in acco ance with Regulatory Position C.6.b of Regulatory ide 1.52, Revist 2, March 1978, meets the laboratory testing iteria of Regulato Position C.6.a of Regulatory Guide 1.52, Revist 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 ifying that on a containment phase A isolat on test signal, e system automatically switches into a recirc ation mode of 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.
W- 5 A'4 7-16 MAY 15 GEO
Alof /)fpll u Yc. k Sat $* con TLANT SYSTEMS - S ILLANCE REQUIREMENTS (Continued) / l
- f. fter each complete or partial replacement of a HEPA fil) r bank by v ifying that the HEPA filter banks remove greater thpo or equal to (99. 5)%* of the 00P when they are tested in place i /accordance with SI N510-1975 while operating the system at a low rate of cfm 1 10%. ,
- g. After each complete or partial replacement of charcoal adsorber bank by ver ying that the charcoal adJorbers 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 1 10%.
{ h l 99.95Y applicable when a filter efficiency of 99% is assumed in the sa ty anal es; 99% when a filter efficiency of 90% is assumed. _-sis , m '-17 _. MAY 151980
/
Nd ApplicaWe 4 SesL<oo4 i P SYSTEMS 3/4. ECCS PUMP ROOM EXHAUST AIR CLEANUP SYSTEM LIMITING C ITION FOR OPERATION 3.7.8 Two indep dent ECCS pump room exhaust air cleanup sy ems shall be OPERABLE. APPLICABILITY: MODES , 2, 3 and 4. ACTION: With one ECCS pump room exh st air cleanup system inoperable, restore the inoperable system to OPERABLE tatus within 7 da or be in at least HOT STAND 8Y within the next 6 hours and in COLD SHU OWN within the following 30 hours. SURVEILLANCE REQUIREMENTS A 4.7.8 Each ECCS pump room exhaust a r clea p system shall be demonstrated (- OPERA 8LE:
- a. At least once per 31 d s on a STAGGE D TEST BASIS by initiating, from the control room flow through the HEPA filters and charcoal adsortpers and verify ng that the system erstes 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 adsorber housin s, or (2) following painting, fire or chemical release in any venti tion zone communicati with the system by:
1.. Verify ng that with the system operating at a. Iow rate of cfm 10% and exhausting through the HEPA filte s and charcoal ads rs, the total bypass flow of the system t the facildty . ve t, including leakage through the system divert g valves, is 1 ss than or equal to 1% when the system is tested admitMng old 00P 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.S.c and C.S.d of Regulatory Guide 1.5?.
Revision 2, March 1978, and the system flow rate is _ fm
+ 10%.
\
,, sis .3/4-M 8 -~ ~ MAY 151990
/
/ i
&qh A fflCQ YC $CQbV00 h
[ LANT SYSTEMS EILLANCE REQUIREMENTS (Continued) / 3.
/
Verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordan'ce with Regulatory Position C.6.b of Regulatory Guide 1.52,Ifevision 2, March 1978, meets the laboratory testing criteria pf Regulatory Position C.G.a of Regulatory Guide 1.52, Revisic /2, March 978.
- 4. V ifying a system flow rate of cfm 1 10 during syst'em opeationwhentestedinaccordancewithANSpN510-1975.
- c. After ever 720 hours of charcoal adsorber op tion by verifying within 31 da s after removal that a laborato analysis of a representativ carbon sample obtained in acc dance with Regulatory Position C.6.b f Regulatory Guide 1.52. R9/isipn 2, March 1978, meets the labora ory testing criteria of egulatory Position C.6.a of Regulatory Gui 1.52, Revision 2, Ma h 1978.
- d. At least once per 18 nths by:
- 1. Verifying that th pressure dro across the combined HEPA filters and charco- adsorber anks of less than (6) inches Water Gauge while op rating t e system at a flow rate of __,
< cfm +- 10%.
- 2. Verifying that the syst m tarts on a Safety Injection Test Signal.
- 3. Verifying that the filt r ooling bypass valves can be manually opened.
- 4. Verifying that the ty aters d sipah 1 kw when tested in accordan e with ANS N510-1975.
- e. After each complete 9 partial repla ment of a HEPA filter bank by verifying that the iEPA filter banks move greater than or equal to (99.95)%* of the D P when they are tes d in place in accordance with ANSI H510-1975 w le operating the systs. at a flow rate of cfm 1 10%.
- f. After each e plete or partial replacement f a charcoal adsorber bank by ver,fying that the charcoal adsorber remove greater than or equal 6 99.95% of a halogenated hydrocarb n refrigerant test gas when the are tested in place in accordance wi h ANSI N510-1975 while opor g the system at a flow rate of cf i 10%.
99.95% appl [cablewhenafilterefficiencyof99%isassumed n the safety analyses; J9% when a filter efficiency of 90% is assumed. ' i W.STS 3/; Tm - MAY 15 20
7.._ , , PLANT SYSTEMS 3/4.7. SNUBBERS LIMITING CONDITION FOR OPERATION 3.7.MAllsnubbersshallbeOPERA8LE. The only snubbers excluded from the requirements are those installed on nonsafety-related systems and then only if their failure of failure of the system on which they are installed would have no adverse effect on any safety-related system. .
. APPLICA8ILITY: MODES 1, 2, 3, and 4. MODES 5 and 6 for snubbers located on j systems required OPERABLE in those MODES.
ACTION: With one or more snubbers inoperable on any system, within 72 hours replace or re-storetheinoperablesnubber(sh.toOPERABLEstatusandperformanengineeringeval-ua ti on ,.. . . .. . . . . . .on--i . 7."; on the attached component or declare the attached system inoperable and follow the appropriate ACTION statement for that system. SURVEILLANCE REQUIREMENTS
- 4. '/W7 Each snubber shall be demonstrated OPERA 8LE by performance of the follo'w of Spec {ng augmented inservice inspection program in lieu of the requirements
/
igcation4.0.5.
- a. Insoection Tyoes As'used in this specification, type of snubber shall,mean snubbers of thessame design and manufacturer, irrespective of capacity.
- b. VisualI)spections Snubbers are' categorized as inaccessible or accessible during reactor
@ operation. Ea'cA of these groups (inaccessible and accessible) may be inspected independently according to'the schedule below. The first inservice visual inspection of performed after 4 months but within/e'ach type of snubber shall be 10 months of commencing POWER OPERATION and shall indlude all sn'ubbers. If all snubbers of each type r.. ..., .,......, are 'found.0PERABLE during the first inservice visual inspection, the seco'nd' inservice visual inspection -C f that-
- ey: M ] shall be performed'it'the first refueling outage. Othemise.
subsequent visual inspections (of sa given system] shall be performed in accordance with the'following schedule:
\
No. of Inoperable Sn/ubbers of Each Type Subsequent Visual [on Any System] pdr Insoection Period Inspection Period * ** 0 18 months 2 25% 1 12 months ! 25% 2 6 months 2 25% 3,4 124 days e 25% 5,6,7 62< days 2 25% 8 or more 31 'da 2 25%
'The inspection interval for each type of snubberien e ginn syst)'-]-shall not be lengthened more than one step at a time unless a generic problem has been identified and corrected; in that event the inspection interval may be 1engthened one step the first time and two steps thereafter if no inop\ erable
, snubbers of that type are found fon-that.syytam]. - \
**The provisions of Specification 4.0.2 are not applicable.
PSTS 3/4 7-
PLANT SYSTEMS SURVEILLANCE REOUIREMENTS (Continued)
- c. Visual Insoection Acc;eotance Criteria Visual inspections shall verify that: (1) there are no visible /
indications of damage or impaired OPERABILITY, (2) attachments to the foundation or supporting structure are functional, and (3) fasten-er,s for attachment of the snubber to the component and to the' snubber ~ anchorage are functional. . Snubbers which appear inoperable 'as a ' resU(tofvisualinspectionsmaybedeterminedOPERABLEforthe purpose of establishing the next visual inspection interval, provided that: '(1) the cause of the rejection is clearly established and remediedyor that particular snubber and for other snubbers irrespec-tive of type @a-that cy: tem] that may be generically susceptible; and (2) the\affected snubber is functionally tested in the as-found condition and determined s OPERABLE per Specification 4.7.if. All snubbers connected to an inoperable common hydraulic fluid reservoir shall be counted as inoperable snubbers. --Efor-these-snubbees-commefr-to-morethan-onehysteerthe-OPERABILITY vf-such-snubbers-sha-H-be--
-considered ia a "e s s ing-the-sdr-vs411 "ra <'hadule-for each-of-the--
-f'el ? t ad Sy ?
- a" ] ,
- d. Transient Event Insoection '
j
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An inspection shall be performed of all. snubbers attached to sections .
+
of systems that have experienced unexpected, potentially damaging transients as determined frc'm a review of operational data and a
~
visual inspection of the systems within 6 months following such an event. In addition to satisfying the visual inspection acceptance criteria, freedom-of motion,of mechanical snubbers shall be verified using at least one of the ,following: (1) manually induced snubber movement; or (2) evaluatjon of in p1, ace snubber piston setting; or (3) stroking the mechan,1 cal snubber through its full range of travel,
- e. Functional Tests /
/
Duringthefirstrefuelingshutdownandat\leastonceper18 months thereafter during shutdown, a representative sample s of snubbers of each type shall'be tested using one of the following sample plans. The sample plan for each type shall be selected prior to the test period and csnnot be changed during the test period. The NRC Regional Administrator shall be notified in writing of the ' sample plan selected for each jnubber type prior to the test period or the sample plan used in'the prior test period shall be implemented:
- 1) At least 10% of the total of each type rofshall snubbe\
be functionally tested either in place or in a bench test. For each snubber of a type that does not meetythe functional test acceptance criteria of Specification 4. 7.1f. , an additional 10% of that type of snubber shall be functionally tested until no more failures are found or until all snubbers of that type have ceen functionally tested; or ' . .k N ( y-STS 3/4 7-
( PLANT SYSTEMS
,\
SURVEILLANCE REOUIREMENTS (Continued)
\ -
- e. Functional Tests (Continued)
- 2) A representative sample of each type of snubber shall be func tionally tested in accordance with Figure 4.7-1. "C" is th'e total number of snubbers of a type fouqd not meeting the' accept-ance requirements of Specification 4.7.if. The cumulative .
number of snubbers of a type tested is denoted by "N". At the end of each day's testing, the new values of "N" and "C" (pre-vi'ous day's total plus currr.nt day's increments),shall be - plotted on Figure 4.7-1. If at any time the point plotted falls \i the " Reject" region, all snubbers of.that type shall be func,ntionally tested. If at any time the point plotted falls in the " Accept" region, testing of snubbers'of that type may be terminated.g When the point plotted lies,in the " Continue Testing" region, additional snubbers ofjthat type shall be tested-until the point falls in the " Accept" region or the
" Reject" region, or all the snubbers.of that type have been tested; or /
- 3) An initial representative sample of 55 snubbers shall be func-tionally tested. Fors each snubber type which does not meet the functional test acceptance criteria, another sample of at least one-half the size of th'e initial sample shall be tested until the total number tested 'is' equal to the initial sample size multiplied by the factor,^1 + C/2, where "C" is the number of snubbers found which do'notxmeet the functional test acceptance criteria. The results' from this sample plan shall be plotted using an " Accept" line which fo11cws the equation N = 55(1
+ C/2). Each snubber point should be plotted as soon as the snubber is tested / If the point \ plotted falls on or below the
" Accept" line, testing of that type of snubber may be terminated.
If the point plotted falls above the " Accept" line, testing must continue 'until the point falls 'in the " Accept" region or allthesnubbersofthat'typehavebee\ b tested. Testing equipment failure during functional testing may invalidate that day's testing and allow- that day's testing \to resume anew at a later time N the day of'siiovided all snobbers tested with the equipment failure are retested. The representative(ailed equipment sample during selected'for the functional test sample plans tha11'be randomly selected from th'e snubbers of each typo and reviewed before be' ginning the testing. The review shall ensure, as far as practicable, that they are represen-tative of the various configurations, operating environments, range of size, and capacity of snubbers of each type. Snubbers placed in the same location as snubbers which failed the previous functional test
/<shall be ratested at the time of the next functional test butsshall
// not be included in the sample plan. If during the functional testing, additional sampling is required due to failure of only one type 'of snubber, the functional test results shall be reviewed at that time to determine if additional samples should be limited to the type of x snubber which has failed the functional testing. N y
. . . . . r- ~. --
PLANT SYSTEMS
. SURVEILLANCE REQUIREMENTS (Continued)
\. Functional Test Acceotance Criteria The snubber functional test shall verify that:
/
Activation (restraining action) is achieved within the / 1)\specifiedrangeinbothtensionandcompression;j// '
- 2) Sn,ubber bleed, or release rate where required, is pres 4nt in
.. both tension and compression, within the specified range;
- 3) For \ mechanical snubbers, the force required to initiate or maintain motion of the snubber is within the specified range in both\ directions of travel; and /
- 4) For snubbers specifically required not to disp' lace under continuous load, the ability of the snubberjto withstand load without displacement.
- n /
/ .
Testing methods may be used to measure parameters indirectly or parameters other than'those specified if those results can be correlated to the speci'fied parameters thr'ough established methods.
- g. Functional Test Failuro A alysis '
\
An engineering evaluation shal1 be ,made of each failure to meet the s functional test acceptance criteria to determine the cause of the failure. The results of this eva,1uation shall be used, if applicable, in selecting snubbers to be testeds in an effort to determine the OPERABILITY of other snubbers / irrespective of type which may be subject to the same failur mode. For the snubbers found inoperable, an engineering evaluation shall be performed on the comp,onents to which the inoperable snubbers are attached. The purpose' of this engineering' evaluation shall be to determine if the ccmponents to which the inoperable snubbers are attached were advepsely affected by the inoperability of the snubbers in order to ensur's that the component remains c'apable of meeting the designed service. \
/ \ fails to If any snubber selected for functional testing either
~
s lock up or' fails to move, f.e. , froien-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 shd11 be func-tiona,11y tested. This testing requirement sup11 be independent of s the, requirements stated in Specification 4.7.Me. for snubbers not meeting the functional test acceptance criteria. \
\
N\ l
\ l W-STS R \. !
3/47-N
PLANT SYSTEMS SURVEILLANCE REOUIREMENTS (Continued) h Functional Testing of Reoaired and Reolaced Snubbers j S ubbers which fail the visual inspection or the functional test ! acce'ptance criteria shall be repaired or replaced. Replacement snubbe'rs and snubbers which have repairs which might affect the function'ai test results shall be tested to meet the functional test . criteria b'efore 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 monthssbefore being installed in the unit' .
\
\
- i. Snubber Service Life Procram l
* 's The service life of hydraulic and mechanical. snubbers shall be monitored to ensure that\ the service life is'not exceeded between surveillance in'spections. NT,he maximum expe'cted service life for various seals, springs, and other critical parts shall be deter-mined and established based on' engineer,ing information and shall be extended or shortened based on monitored test results and failure history. Critical parts shall be replaced so that the maximum service life will not be exceeded during a period when the snubber is required to be OPERABLE. The parts'r,eplacements shall be docu-mented and the documentation shall be retained in accordance with
/
Specification 6.10.} / r
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4 W-STS 3/47-K
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10, s -- 9
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N 8
\ , -
'N /
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REJECT j
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?s ..
4 ef 'N f
, / CONTINUENN TESTING ,
[ / 'N
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S '
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/ ACCEPT
( N x s 0 10 20 30 40 50 60 70 80 90 '100 N s k FIGURE 4.7-1 SAMPLE PLAN 2) FOR SNUB 8ER FUNCTIONAL TEST W-STS 3/4 7-
PLANT SYSTEMS 3/4.7.}8Q SEALED SOURCE CONTAMINATION LIMITING CONDITION FOR OPERATION 8 3.7.71L 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 contamination in excess of the above limits, immediately withdraw the sealed source from use and either:
- 1. Decontaminate and repair the sealed source, or
- 2. Dispose of the sealed source in accordance with Commission Regulations.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REOUIREMENTS 8 4.7.19 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.
The test method shall have a detection sensitivity of at least 0.005 microcuries per test sample. 4.7.k.2 Test Frequencies - Each category of sealed sources (excluding startup sources and fission detectors previously subjected to core flux) shall be tested at the frequency described below.
- a. Sources in use - At least once per 6 months for all sealed sources containing radioactive materials:
- 1. With a half-life greater than 30 days (excluding Hydrogen 3),
k and
- 2. In any form other than gas.
2?- W-STS 3/4 7-24 fiOV 2 01950
PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
- b. Stored sources not in use - Each sealed source and fission detector shall be tested prior to use or transfer to another licensee unless tested within the previous six months. Sealed sources and fission detectors transferred without a certificate indicating the last test date shall be tested prior to being placed into use. .
- c. Startup sources and fission detectors - Each sealed startup source and fission detector shall be tested within 31 days prior to being subjected to core flux or installed in the core and following repair or maintenance to the source.
8 4.7.TQ.3 Reports - A report shall be prepared and submitted to the Commission
. on an annual basis if sealed source or fission detector leakage tests reveal the presence of greater than or equal to 0.005 microcuries of removable contamination.
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k 13 w-STS 3/4 7-21 'NOV 2 01980
PLANT SYSTEMS c 3/4.7.'N , FIRE SUPPRESSION SYSTEMS FIRE SUPPRESSION WATER SYSTEM LIMITING CONDITION FOR OPERATION C 3.7.141 The Fire Suppression Water System shall be OPERABLE with: !
- a. At least [two] fire suppression pumps, each with a capacity of
'g #5oo[-0500-3 gpm, with their discharge aligned to the fire suppression header, , b. Separate water supplies, each with a minimum contained volume of . 2k90,ooo gallons, and a
Qre. wab 3
- c. An OPERABLE flow path capable of taking suction from the V_ tank
-and the +=nb and transferring the water through distribution piping witn OPERABLE sectionalizing control or isolation valves to the yard hydrant curb valves, the last valve ahead of the water flow (h) alarm device on each sprinkler or hose standpipe, and the last valve ahead of the deluge valve on each Deluge or Spray System required to be OPERABLE per Specifications 3.7.R,2, 3.7.M.% and 3.7.'11.%
9 R3 99 APPLICABILITY: At all times. ACTION: goals Qewrable availa.ble, b
- a. With one pumpvand/or one wat.er supply incpe-eble, restore the inoper-able equipment to OPERABLE status within 7 days or provide an alter-nate backup pump or supply. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
- b. With the Fire Suppression Water System otherwise inoperable, establish a backup Fire Suppression Water System within 24 hours.
t 29
'd-STS 3/4 7-14
PLANT SYSTEMS SURVEILLANCE REQUIREMENTS 4 .1.1 The Fire Suppression Water System shall be demonstrated OPERABLE: I a. At least once per 7 days by verifying the contained water supply volume,
,e I b.
AtQeast once per 31 days.-on-a.-S"CCERED TEST-SA%5-by start ngM0h b electric on motor-driven recirculation flow,pump and operating it for at leas,t 15 minutes
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c. N At least once per 31 days by verifying that each va/ (manual, power I N lve operated, or automatic) in the flow path is in its correct position,
- d. JAtleastonceper \ 12. / ' l months by performanc of a system flush,J @ l
- e. At least once per 12 months by cycling 'each testable valve in the flow path through at le'ast one complete cycle of full travel,
- f. N At least once per 18 months by per /
which includes simulated automatic, forming a system functional test l actuation of the system throughout its operating sequence, and: [
- 1) Verifying that each automatic va. ve in the flow path actuates to its correct positi'on, 2)
/ L 500 0 Verifying that each pump develops at least [4500-] gpm at a system head o [250] feet, l
- 3) Cycling each valve in the flow path that is not testable during plant ope, ration through at least one complete \ cycle of full travel and ,
- 4) Veri,fying that each fire suppression pump starts [ sequentially]
to,, maintain the Fire Suppression Water System pressuresgreater than or equal to kkp psig. g.
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Atfleast once per 3 years by performing a flow test of the systemyn accordance with Chapter 5, Section 11 of the Fire Protection Handbook, 14th Edition, published by the National Fire Protection Association. t l W-STS N 3/4 7-29
- l PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
.1.2 The fire pump diesel engine shall be demonstrated OPERABLE:
ton a STAGGERED TEST hasn'S;
- a. At least once per 31 daysNby verifying:
1. The fuel storage tank contains at least A00 gallons of fuel, and j/'
- 2. The diesel starts from ambient conditions an/d operates for at lea'st 30 minutes on recirculation flow.
b.
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At least oncesper 92 days by verifying that a sample of diesel fuel from the fuel stora k w th*- the Sciep'ge i
..u__ - u _ _ o _ _, ,__ -
tank, t:ble-obtained 1imit; in accordance specifi$d ir, Tchl; i cfwitl#
^.S ASTM DE70 G5, "075 c.
N , _, ,,, gg ;;;, y At least once per 18 months,-durir.g/
,,7 y shutdcr., by subjecting the diesel to an inspection in accord'ance with procedures prepared in conjunction with its manufacturer's reco'mmeridations for the class of service.
9 /\ 4.7. k l.3 The fire pump diesel starting 24-volt battery bank and charger
- l shall be demonstrated OPERABLE
- /
- a. At least once per 7 days by / verifying that:
- 1. The electrolyte / level of each batte is above the plates, and
- 2. x The overall b ttery voltage is greater than or equal to 24 volts.
- b. At least once pe/r92daysbyverifyingthatthes\ pecific gravity is appropriate foi continued service of the battery.
- c. Atleastonc/ e per 18 months by verifying that: -
- 1. The atteries,cellplatesandbatteryracksshownov\ isual indication of physical damage or abnormal deterioration, an,d
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- 2. The battery-to-battery and terminal connections are clean, tight, free of corrosion and coated with anti-corrosion material.
H-STS 3/4 7- MAY 151980
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INSERT I / f
.I ASTM - D4057-81, has '
j N1. - A kinematic viscosity, at 40 C, of greater than or, equal to 1.9 centistrokes but less than or equal to 4.1 centis gravity was not determined by comparison with the,trokas, suppliers if certification. / 2. N bright appearance with proper color A clear when /tested in accordance with ASTM-D4176-82. /
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' PLANT SYSTEMS SPRAY AND/0R SPRINKLER SYSTEMS LIMITING CONDITION FOR OPERATION 9 3.7.h.2 The following Spray and/or Sprinkler Sys,tems shall be OPERABLE:
- a. [ Plant dependent - to be listed.by name and location.]
- b. $ee L ser+ Y c.
APPLICABILITY: Whenever equipment protected by the Spray / Sprinkler System is l required to be OPERABLE. ACTION:
- a. With one or more of the above required Spray and/or Sprinkler Systems inoperable, within 1 hour establish a continuous fire watch with backup fire suppression 4quipment for those areas in which redundant .
systems or components could be damaged; for other areas, establish an hourly fire watch patrol.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS A C7>K. 2 Each of the above required Spray and/or Sprinkler Systems shal -be demonstrated OPERABLE: . fl
- a. east once' per 31 days by verifying that each valve (manual, power-b.
operat'ed,Nor automatic) in the flow path / is in its' correct position,l At least once'per 12 months by cycling each' testable valve in the flow path throuijh at least one complete-c'ycle of full travel, l
- c. At least once per anths:
- 1) By performing a system functional test which includes simulated l automatic actuation o'f the' system, and:
/ the automa\
Verifying'that a) tic valves in the flow path actua,tetotheircorrectpositionson4,mm _ _ _ _ test signal, and b)/ Cycling each valve in the flow path that is not testable
' g during plant operation through at least one complete cycle -
of full travel. N W-STS T 3/4 7 '!L}
INSERT II I
- a. Cable Spreading Room b.- Computer Room e
- c. ' Electrical Tunnels
- 1. Control Building to Containment
- 2. Control Building to PAB
- d. -Diesel Generator Building
- 1. Fuel Oil Storage Tanks
- 2. Fuel Oil Pipe Trenches
- 3. Fuel Oil Day Tanks
- e. Primary Auxiliary Building
- 1. Electrical Chases
-+
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- =. -
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SURVEILLANCE REQlJIREMENTS (Continued) ' N)
. 2 By a visual inspection of the dry pipe spray andIprinkler headers to verify their integrity; and / ~
- 3) By a visual inspection of each nozzle's spray' area to verify the . [
spray pattern is not obstructed. j .
- d. At least, once per 3 years by performing an air' flow test through each open, head spray / sprinkler header and verifying each open head spray / sprinkler nozzle is unobstructed.
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's 0.9 W-STS 3/47-h
PLA T SYSTEMS NgY A //lCa.b[e
/ h fea$ rook CO2 S TEMS LIMITING NDITION FOR OPERATION i
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3.7.h.3 The llowing high pressure and low pressure CO 2 sys .ns shall be OPERABLE.
- a. (Plant d endent - to be listed by name and loca on.)
b. c. APPLICABILITY: Whenever e ipment protected by the 02 systems is required to be OPERABLE. ACTION: a '. With one or more of the ove requir d C0psystems inoperable, within one hour. establish contin us fire watch with backup fire suppression equipment for th e a eas in which redundant systems or components could be damaged; r other areas, establish an hourly fire watch patrol. Restore th ystem to OPERABLE status within 14 g days or, in lieu of any other ep t required by Specification r 6.9.1, prepare and submit a etia Report to the Commission pursuant to Specification 6.9.2 wit n the ne t 30 days outlining the action
, taken, the cause of the i perability nd the plans and schedule for restoring the system to ERABLE status,
- b. The provisions of Spe fications 3.0.3 and .0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.7.11.3.1 Each of the ove required C07 systems shall be o monstrated OPERABLE at least once er 31 days by verTfying that each valv (manual, power operated or automatic in the flow path is in its correct positi n. 4.7.11.3.2 Each of the above required low pressure CO systems sh I be demonstrated OPE LE: 2
- a. At 1 st once per 7 days by verifying the CO,, storage tank 1 el to be reater than and pressure to be greater than psi and
- b. [ least once per 18 months by verifying:
- 1. The system valves and associated ventilation dampers and fire door release mechanisms actuate manually and automatically,
, upon receipt of a simulated actuation signal, and
- 2. Flow from each nozzle during a " Puff Test."
_-STS M 7-34 MAY I 5 lo80 1
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NT SYSTEMS Mof A// // e b# SURVE LANCE REQUIREMENTS (Continued)
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4.7.11.3. Each of the above required high pressure CO systems sha be demonstrate OPERABLE: 2
- a. At le t once per.6 months by verifying the CO stora tank weight to be a least 90% of full charge weight 2
- b. At least o e per 18 months by:
- 1. Verifyin the system, including associated entilation dampers and fire door elease. mechanisms, actuates ma ally and automatically, upon receip of a simulated actuation si nal, and
- 2. Performance of flow test through he ers and nozzles to assure no blocka .
l W STS 1/' ' '5 MAY 1519S0
PL T SYSTEMS lYd ~ Affltca Ye -lo Sea b Voo h HALON STEMS LIMITING DITION FOR OPERATION
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3.7.11.4 The f lowing Halon systems shall be OPERABLE.
- a. (Plant endent - to be listed by name and location b.
c. APPLICABILITY: Whenever uipment protected by the Hal n system is required to be OPERA 8LE. , ACTION:
- a. With one or more of.th above required alon systems inoperable, within 1 hour establish a continuous f re watch with backup fire suppression equipment fo those area in which redundant systems or components could be damage ; for ot er areas, establish an hourly fire watch patrol. Restore the s tem to OPERABLE status within 14 days or, in lieu of any o her eport required by Specifica-tion 6.9.1, prepare and submi Special Report to the Commission i pursuant to Specification 6.9. within the next 30 days outlining the action taken, the cause t inoperability and the plans and schedule for restoring the ystem o OPERABLE status.
- b. The provisions of Specif ations 3.0. and 3.0.4 are not applicable.-
SURVEILLANCE REQUIREMENTS
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4.7.11.4 Each of the above quired Halon systems sh 11 be demonstrated OPERABLE:
- a. At least once p 31 days by verifying that ea valve (manual, power operated or automatic) in the flow path 1 in its correct position.
- b. At least o e per 6 months by verifying Halon storag tank weight to be at ecct 95% of full charge weight (or level) a pressure to be at leas 90% of full charge pressure.
- c. At le st once per 18 months by:
- 1. Verifying the system, including associated ventilation mpers and fire door release mechanisms, actuates manually and a to-matically, upon receipt of a simulated actuation signal, a d
( 2. Performance of a flow test through headers aild nozzles to assure no blockage. ; _-STS -3M i - J o NOV 2 01950 s 1 l
PLANT SYSTEMS FIRE HOSE STATIONS LIMITING CONDITION FOR OPERATION q.3 Ider 3.7.K.'5, The fire hose stations given in ?&ic-3.7-? shall be OPERABLE.
. APPLICABILITY: Whenever equipment in the areas protected by the fire hose stations is required to be OPERABLE.
- ACTION:
- a. With one or more of the fire hose stations-gwea 3-T2!c 3.7=4 inoperable, provide gated wye (s) on the nearest OPERABLE hose station (s). One outlet of the wye shall be connected to the standard length of hose provided for the hose station. The second outlet of the wye shall be connected to a length of hose sufficient to provide coverage for the area left unprotected by the inoperable hose station. Where it can be demonstrated that the physical routing of.the fire hose would result in a recognizable hazard to operating technicians, plant equipment, or the hose itself, the fire hose shall be stored in a roll at the outlet of the OPERABLE hose station. Signs shall be mounted above the gated wye (s) to identify the proper hose to use. The above ACTION requirement shall be accomplished within 1 hour if the inoperable fire hose is the primary means of fire suppression; otherwise route the additional hose within 24 hours.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REOUIREMENTS D7 Each of the fire hose station ven in Table 3.7-4 shall be demonst ated OPERABLE:
- a. tqeast once per 31 days, by a visual inspection of the fire' ose stations accessible during plant operations to assure alld equired equipmenb is at the station.
N. b. Atleastonce,peQ8 months,by:
- 1) Visual inspection of the stations not accessible during plant !
operations to asidre all required equi'pment is at the station,
- 2) Removing the hose for N inspection /and re-racking, and i l
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- 3) Inspecting all gaskets and' replacing any degraded gaskets in l the couplings. /
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- c. At least once per 3 years, by:
- 1) Partially opefiing each hose station valve tosverify v'alve l OPERABILITY and no flow blockage, and \
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- 2) Conducting a hose hydrostatic test at a pressure of'150 psig or l at least 50 psig above maximum fire main operating pressure, whichever is greater. \
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'f-STS 3/4 7-hi
i l ( ' l 4 / TABLE 3.7-4 7'
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FIRE HOSE STATIONS
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LOCATION * - ELEVATION HOSE RACK #
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This ( orecd}On E bC SN e aidet d ele. [ 88 '
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._-STS 3/4 7- % NOV 0 0193']
PLANT SYSTEMS YARD FIRE HYDRANTS AND HYORANT HOSE HOUSES LIMITING CONDITION FOR OPERATION 94 3.7.h.1 The yard fire hydrants and associated hydrant hose houses given in
...,m .. e shall be OPERABLE.
Id cr , APPLICABILITY: Whenever equipment in the areas protected by the yard fire hydrants is required to be OPERABLE. ACTION:
- a. With one or more of the vard fire hydrants or associated hydrant hose houses giv = in Tam e-3.J-5 inoperable, within 1 hour have sufficient additional lengths of 2 1/2 inch diameter hose located in an adjacent OPERABLE hydrant hose house to provide service to the unprotected area (s) if the inoperable fire hydrant or associated hydrant hose house is the primary means of fire suppression; otherwise, provide the additional hose within 24 hours.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4Q. . Each of the yard f, ire hydran nd associated hydrant hose houses given able 3.7-5 shall be demonstrated OPERABLE:
- a. At least once per 31 days, by visual. inspection of the pydrant hose house to assure all required equipment is at the hose house, i N /
- b. At least once s per 6 months (once during March, April, or May and -
once during3 September, October, or November), each yard fire hydrant and verifying that the,bf visually hydrant inspecting barrel is dry 4 and that the hydrant is not damaged, and /
- c. At least once per lhanths by:
- 1) Conductingahosehdrostatictestatapressureof150psigor I at least 50 psig above'maiimum fire main operating pressure, whichever is greater,/
- 2) Inspectingallthe[asketsandreplacinganydegradedgaskets l inthecouplings7.and
- 3) Performing,adlowchakofeachhydrantto'verifyits l OPERABILITY.
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._ .- . .-. .- . . . . . . = . .. . .. .-.. ..
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S~ / TABLE 3. 7-Ya / YARD FIRE HYDRANTS AND ASSOCIATED HYDRANT HOSE HOUSES ,/
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LOCATION
- HYDRANT NUMBER ,
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Tkts irsor 4s'on be foe s uPPI A af a. later dcde. i
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- List all. Yard Fire Hydrants and Hydrant Hose Houses required to ensure the
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W-STS 3/4731, M
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i PLANT SYSTEMS 3/4.7. FIRE RATED ASSEMBLIES
. LIMITING CONDITION FOR OPERATION 10
- 3. 7. 'h2 All fire rated assemblies (walls, floor / ceilings, cable tray enclosures,'
and other fire barriers) separating safety-related fire areas or separating portions of redundant systems important to safe shutdown within a fire area and all sealing devices in fire rated assembly penetrations (fire doors, fire windows, fire dampers, cable, piping, and ventilation duct penetration seals shall be OPERABLE. APPLICABILITY: At all times. ACTION:
. a. With one or more of the above required fire rated assemblies and/or sealing devices i~noperable, within 1 hour either establish a
. continuous fire watch on at least one side of the affected assembly, or verify the OPERABILITY of fire detectors on at least one side of the inoperable assembly and establish an hourly fire watch patrol,
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not acplicable.
SURVEILLANCE REOUIREMENTS
'4 .1 At least once per 18 month e above required fire rated assem ' e s' and penetration sealing devices shall be verified OPERABLE by performi g visual i'nspection of:
a.
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The exposed surfaces of each fire rated assembly,
- b. Each fire window / fire damper and associated-hardware, and c.
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At least 10% of each type of'se(1ed, penetration. If apparent changes in appearance or abnorma.1-cegtadations are found, a visual inspection of an additional 10%'of eachNyp.e of sealed penetration shall be made. This inspection process shalNontinue until a 10% sample with no apparent-changes in appearance orN bqormal degradation is found. Samples sha'll be selected such that each penetration will be inspected every'15 years. 1 33 h-STS 3/4 7- %
( PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 4 72 'Each of the above required fire doors shall be verified OPERABLE by / inspecting the automatic hold-open, release and closing mechanism and la_tches at least once per 6 months., and by verifying: / a. N ,/ The position of each closed' fire door at least once per 24 hours.
- b. A /
That doors with automatic hold open, and'riii_ ease mechanisms are free of obstructions at least once per-24' hours. c.
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The position of each locked'c sed fire door at'least once per 7 days. /
- d. The OPERABILITY of the fire door supervision system by performing a TRIP ACTUATING DEVICE OPERATIONAL TEST at least once per 31 days.
i 4 34 g-STS 3/4 7-M qicy 2 19si
PLANT SYSTEMS I\ 3/4.7.M AREA TEMPERATURE MONITORING LIMITING CONDITION FOR OPERATION i Ider @ 11
+
- 3. 7.h The temperature of each area shown in-TaMeshPt shall be maintained t
I within the 4"it: ' r.dicated '- T;b k 3. 7 *F. id t'ca.+ed lib'tti. I 1 APPLICABILITY: Whenever the equipment in an affected area is required to be OPERABLE. l ACTION: l With one or more areas exceeding the temperature limit (s)-shown-i-n--
-Table-3r7-Tr
- a. For more than ++g9ht hours, prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within the next 30 days providing a record of the amount by which and the cumulative time the temperature in the affected area exceeded its limit and an analysis to demonstrate the conti.nued OPERABILITY of the affected equipment. The Provisidts of SF"t6c" Nov 3 0 3 ed 3 o-N "# ^# 4/j','gf,'
- b. By more than 30*F, in addition to the Special Report required above, within 4 hours either restore the area to within its temperature limit or declare the equipment in the affected area inoperable.
SURVEILLANCE REQUIREMENTS
- 4. 7?M Il .
he temperature in each of the areas shown in Table 3.7-
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determi -to be within its limit at least once per 24 hours. / .shall be
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b / 6 TABLE 3.7-1 AREA TEMPERATURE MONITORING AREA TEMPERATURE LIMIT (*F) 1. 2.
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NOV 2 test
JUSTIFICATIONS Section 3/4.8 In the text of Section 3/4.8 where a capital letter with a circle around it appears, please refer to this sheet for the appropriate justification. A. Seabrook Station diesels do not have engine mounted fuel day tanks. B. The Seabrook Station plant specific data. C. The Seabrook diesels operate at 514 RPM not 900 D. The Seabrook diesel and emergency power sequencer are designed to load in 120 seconds. E. The wording clarifies when this step is to be used and uses wording to match Seabrook Station diesel procedures. F. Commitment made to NRC in RAl-430.81. G. Emergency power sequencer is the Seabrook specific name for load sequencer. H. This step deleted per NRC Generic Letter 83-30. I. This is an SER commitment to perform this testing. J. Plant specific design parameters for Seabrook Station. K. The Seabrook "short time" rating is the same as the }[ 2000 hour rating. L. The bases section of the Tech Specs explains the technical reasons for not starting the diesel at operating temperature with an auto-start signal. Therefore, we call our Surveillance Requirement 4.8.1.1.2.f.14 rather than f.6.b. M. Seabrook barring device is y_ turning gear, and the Seabrook diesel has differential lockout relay rather than an emergency stop. N. These steps were added as a result of SER Commitments. O. This Section changed to be plant specific for Seabrook Station 4 battery DC System. P. Seabrook batteries have 59 cells rather than 60 thus our lower (128 volt) float charge. Q. Clarification statement for interconnection cables acceptance criteria. Also, the Seabrook battery chargers are designed to deliver 150 amps at , 132 volts for a minimum of 8 hours. l l
R. Note b change reflects that this applies to the float charge current while Note e applies to temperature readings on each cell. The temperature correction should be for the specific cell rather than the average temperature of all cells. S. Changes in this Specification reflect plant specific design of Seabrook Station. T. Added clarification of testing method when current transformers are involved. Alsc, we only have 13.8 kV medium breakers at Seabrook. U. The penetration withstand times are much longer than motor protection
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times. This clarifies which time in Table 3.8-1 we have to meet. Also, called out specific action to be taken if penetration withstand time is acceptable but motor protection time fails. Also, added overload devices and clarified test surveillance. V. Seabrook does not have bypass devices. The surveillance requirements have been changed to reflect NRC acceptance of the Seabrook philosophy of replacing thermal overload devices with pre-calibrated devices. W. This Section added as the result of an SER commitment. X. This table is deleted and will be contained in a licensee maintained and controlled document. Y. This change to the allowed outage time will allos,most repairs to an offsite circuit or a diesel generator to be made before shutting down the plant. In addition, this change has a minimal effect on the reliability of the AC Electrical System. A more detailed justification based on risk analyses will be submitted August 16, 1985. l l l , i l l
t 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1 A.C. SOURCES OPERATING LIMITING CONDITION FOR OPERATION 3.8.1.1 As a minimum, the following A.C. electrical power sources shall be OPERABLE:
- a. Two physically independent circuits between the offsite transmission network and the onsite Class lE distribution system, and
- b. Two separate and independent diesel generators, each with:
g 1. Separate day and engine mount s fuel tanks containing a minimum volume of & gallons of fuel,
- 2. A separate fuel storage system containing a minimum volume of la4 er gallons of fuel, and
- 3. A separate fuel transfer pump.
APPLICABILITY: MODES 1, 2, 3, and 4. ACTION:
- a. Wi~th either an offsite circuit or diesel generator of the above required A.C. electrical power sources inoperable, demonstrate the OPERABILITY of the remaining A.C. sources by performing Specification 4.8.1.1.la within 1 hour and at least once per 8 hours thereafter and Specification 4.8.1.1.2.a.4 within 24 hours; restore at least two offsite circuits and two diesel generators to OPERABLE status within
-Qvjs ?2 hers or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
- b. With one offsite circuit and one diesel generator of the above required A.C. electrical power sources inoperable, demonstrate the OPERABILITY of the remaining A.C. sources by performing Specifications 4.8.1.1.la within 1 hour and at least once per 8 hours thereafter and Specification 4.8.1.1.2.a.4 within 8 hours; restore at least one of the inoperable sources to OPERABLE status within 12 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. Restore at least two offsite circuits _and two diesel generators to OPERABLE status within C ha r dfrom th time of initial loss or be in at least HOT STANDBY within the next 6 hours and in COLO SHUTDOWN within the following 30 hours. h
- c. With one diesel generator inoperable in addition to a or b above, verify that: -
(1) all required systems, subsystems, trains, components and devices that depend on the remaining OPERABLE diesel generator as a source of emergency power are also OPERABLE, and
-W-STS 3/4 8-1 JUL 2 71991
f ELECTRICAL POWER SYSTEMS ACTION: (Continued) emerc (2) When in MODE 1, 2, or 3, the steam-driven aux"u}encyry feed pump is OPERABLE.
' If these conditions are not satisfied within 2 hours be in at least HOT STAND 8Y within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
- d. With two of the above required offsite A.C. circuits inoperable, demonstrate the OPERABILITY of two diesel generators by performing Specification 4.8.1.1.2a.4) within 8 hours, unless the diesel gener-ators are already cperating; restore at least one of the inoperable offsite sources to OPERABLE status within 24 hours or be in at least HOT STANDBY within the next 6 hours. With only one offsite source restored, restore at.least two offsite circuits to OPERABLE status with W P "=c from time of initial loss or be in at least NOT STANDBY
<}o.4p} within the next 6 hours and in COLD SHUTDOWN with
@ e.
30 hours. With two of the above required diesel generators inoperable, demonstrate the OPERABILITY of two offsite A.C. circuits by perform-ing Specification 4.8.1.1.la. within 1 hour and at least once per
- 8 hours thereafter; restore at least one of the inoperable diesel generators to OPERABLE status within 2 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the
, followina 30 hours.
Restore at least two diesel generators to (htp) OPERABLE status withiM2 mm from time of initial loss.or be in least HOT STAND 8Y within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE REQUIREMENTS 4.8.1.1.1 Each of the above required independent circuits between the offsite transmission network and the onsite Class 1E distribution system shall be:
- a. Determined OPERABLE at least once per 7 days by verifying correct breaker alignments, indicated power availability, and
- b. Demonstrated OPERABLE at least once per 18 months during shutdown by transferring (manually and automatically) unit power supply from the normal circuit to the alternate circuit.
4.8.1.1.2 Each diesel generator shall be demonstrated OPERABLE: i
- a. In accordance with the frequency specified in Table 4.8-1 on a STAGGERED TEST BASIS by:
t g ,
- 1. Verifying the fuel level in the day and engine : ented fuel tank, W-STS 3/4 8-2 JUL 2 71931
m ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
- 2. Verifying the fuel level in the fuel storage tank,
- 3. Verifying the fuel transfer pump starts and transfers fuel from the storage s stem to the day-and- cnginc ; cur 4ed tank, a C
- 4. Verifying th iesel arts from ambient condition and accelerates to at least (4G9-) rpm in less than or equal to (10) seconds.*
The generator voltage and frequency shall be (4160) + (420) volts and (60) + (1.2 The dieseT gener)ator shall be started for this test by usingHz with one of the following signals:
~
a) Manual, b) Simulated loss of offsite power:by itself.
~
sr c) Simulated loss of offsite power in conjunction with an-E4F actuation test signal. sr d) An-EM actuation test signal by itself.
- 5. Verifying the generator is synchronized, loaded to greater than r equal t nti" ce: 2 ting) in less than or equal to (4e-)/2O 60F3kW{ seconds? and operates with a load greater than or equal to Dentinucur "aHng) for at least 60 minutes, ,
\ AAer cowletion oC s+eps" veviCv%t, i reluvaed 4o swipe't s7ams ad W
- 6. v VVcr;;y;ng the diesel generator enc is aligned to provide standby @
power to the 2::cciated assoctare oemerLc eneE 4" busses.
- b. At least once per 31 days and be cTc peretica cf the dic:e1 whece the period of operation was greater than or equal to 1 hour by checking for and removing accumulated water from the day and enginc-
- cented fuel tanks.
Spe Inse.4 I F
-c, -At 1000, once per 92 c ys and"rce new fuc. cil prioP Lu oddition te-
-thc stcrage tanks by verifying that a ; ample cbtMned 4" accordance-with AS -0270-1975 ha; a water :nd cdiment centent of Iccr t ha n--se-
-cqual tc .05 vuiume peicent end a kinematic viccc ity 0 40"C vi-
-greater than er equal to 1.0 but ic;; than cr equal to 4.1 when-
-tected in accordance with ASTM C075 77, and an impurity level of-
- le:: than 2 mg. of in3clubles per 100 al. when tested in accordance-etMSTM-0227A 70
-F
% At least once per 18 months, during shutdown, by:
- 1. Subjecting the diesel to an inspection in accordance with I procedures prepared in conjunction with its manufacturer's
, recommendations for this class of standby service,
- 6 ee Inse rt.II-1 W-STS 3/4 8-3 .t!OV 2 1981 l
1
g INSERT I I L
- c. At least once per 92 days by checking for and removing accumulated water from the fuel oil storage tanks.
- d. By sampling new fuel oil in accordance with ASTM D4057-81 prior to I addition to storage tanks and:
(1) By verifying in accordance with the tests specified in ASTM D975-81 prior to addition to the storage tanks that the sample has: (a) An API gravity of within 0.3 degrees at 60* or a specific gravity of within 0.0016 at 60/60*F, when compared to the supplier's certificate or an absolute specific gravity at 60/60*F of greater than or equal to 0.81 but less than or equal to 0.89 or an API gravity of greater than or equal to 28 degrees but less than or equal to 42 degrees, (b) A kinematic viscosity at 40*C of greater than or equal to 1.9 centistokes, but less than or equal to 4.1 centistokes, if gravity was not determined by. comparison with the supplier's certification, (c) A flash point equal to or greater than 125'F, and (d) A clear and bright appearance with proper color when tested in accordance with ASTM D4176-82. (2) By verifying within 30 days of obtaining the sample that the other properties specified in Table 1 of ASTM D975-81 are met when tested in accordance with ASTM D975-81 except that the analysis for sulfur may be performed in accordance with ASTM D1552-79'or ASTM D2622-82, and the analysis for carbon residue may be performed in accordance with ASTM D524-76 or ASTM D189-76 and converted by FIG X 3 of ASTM D524-76 to Ramsbottom Carbon Residue. (a) Fuel oil shipments which fail to meet the criteria in 4.8.1.1.2.d.2 shall require the following:
- 1. Storage tanks which contain this oil shall be verified or restored to meet the crite'ria specified in 4.8.1.1.2.d.2 within 30 days of fuel oil receipt.
- e. At least once every 31 days by obtaining a sample to fuel oil in accordance with ASTM D2276-78, and verifying within 7 days of sampling, that total particulate contamination is less than 10 mg/ liter when checked in accordance with ASTM D2276-78, Method A.
k
h
.(,
INSERT II
*These diesel ~ generator starts from ambient conditions shall be performed only once per 184 days in these surveillance tests and all other engine starts for the purpose of this-surveillance testing shall.be preceded by an engine prelube period and/or other warmup procedures recommended by the manufacturer so that the mechanical stress and wear on the diesel is miminized..
4r I L i
=
O 1 (. g t . . . , . . . .
ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 47a
- 2. Verifying the generator capability to reject a load of greater than or equal to (-!: gert cingle crergency inad) kw while maintaining voltage at (4160) 1 (420) volts and frequency at (60) 1 (1.2) Hz (!ert then-cr equel te 75 cf the difference-
-bettecer ncmin:1 speed and the everspeed trip setpoint, er 15%
deec na.Linel whichever is ic33).while nel e.xceedti '*+*' R I' 't - 6083
- 3. Verifying the generator capability to reject a load of (-continuous l Tating) kw without tripping. The generator voltage shall not )
exceed (4784) volts during and following the load rejection. l l
- 4. Simulating a loss of offsite power by itself, and:
l l a) Verifying de-energization of the emergency busses and load 1 shedding from the emergency busses. b) Verifying the diesel starts on the auto-start signal, energizes the emergency busses with permanently connected
@ loads within (10) seconds, energizes the auto-connected
@crgen]eowe 4 shutdown loads through the Qoad. sequencer and operates for greater tnan or equal to 5 minutes while its generator is loaded with the shutdown loads. After energization, the steady state voltage and frequency of the emergency busses shall be maintained at (4160) + (420) volts and (60) 1 (1.2) Hz during this test.
s1
- 5. Verifying that on an E4F actuation test signal, without loss of offsite power, the diesel generator starts on the auto-start signal and operates on standby for greater than or equal to 5 minutes. The generator voltage and frequency shall be t
(4160) 1 (420) volts and (60) 1 (1.2) Hz within (10) seconds after the auto-start signal; the steady state generator voltage and frequency shall be maintained within these limits during this test. l E -Veri fyi ng u.st cn :imulated ! cts of the diece! genereter, with cffsit; pv.c. nut a;;i! bla, the 'n=de ara <had 6 n= the-- h -cmcrgency busscs end Uiet ;ubsequan+ inadinn af th; dicwi
-generatcr i; in accordancc with design rcquiresnt;.
Sr 6 %. Simulating a loss of offsite power in conjunction with an 'ESE actuation test signal, and 4 owe ac+aaAion +csF shaal(r#),ual a) Verifying de-energization of the emergency busses and load shedding from the emergency busses.
\
W-STS 3/4 8-4 OCT 2 31980
ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) (e e gency po w er CT b) Verifying the diesel starts on the auto-start signal, energizes the emergency busses with permanently connected loads within (10) seconds, energizes the auto-connected emergency (accident) loads through the -leadhequencer and operates for greater than or equal to 5 minutes while its generator is loaded with the emergency loads. After energization, the steady state voltage and frequency of the emergency busses shall be maintained at (4160) +~ (420) volts and (60) 1 (1.2) Hz during this test. h c) Verifying that all automatic diesel generator trips, 33 3 except engine overspeed 4 and generator differential, are Chat,o-volt bus %lt automatically bypassed upon loss of voltage on the 3 emergency bus concurrent with a safety injection actuation , signal.
- 78. Verifying the diesel generator operates for at least 24 hours.
During shall bethe firstto2 greater loaded hours of thisortest, than the equal to diesel generat (2 t ur rati m @kw and during the remaining 22 hours of this test, the diesel f' generator shall be loaded to greater than or equal to (-sen-4083 ti'U000 P2tirC) kw. The generator voltage and frequency shall be (4160) 1 (420) volts and (60) 1 (1.2) Hz within (10) seconds after the start signal; the steady state generator voltage and frequency shall be maintained within these limits during this test. Within 5 minutes after completing this 24-hour test, perform Surveillance Requirement 4.8.1.1.2.
.g 8 'R. Verifying that the auto-connected loads to each diesel generator do not exceed the 2000-h. cur ratin ofg g kw.
l s hod + me. ) Cl M. Verifying the diesel generator's capability to: a) Synchronize with the offsite power source while the generator is loaded with its emergency loads upon a simulated restoration of offsite power, b) Transfer its loads to the offsite power source, and c) Be restored to its standby status. 10M. Verifying that with the diesel generator operating in a test mode, connected to its bus, a simulated safety injection signal overrides the test mode by (1) returning the diesel generator to standby operation and (2) automatically energizing the emergency loads with offsite power. W-STS 3/4 8-5
i l l 1 i 1 ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 1 lb. Verifying that the fuel transfer pump transfers fuel from each fuel storage tank to the day and engin; mcunted tank of each diesel via the installed cross-connection lines. IAM. Verifying that the Etc *E Tc l(([ I sequ ce timer is OPERABLE with the interval between each load block within i 10% of its design interval. 13M. Verifying that the following diesel generator lockout features prevent diesel generator starting only when required: a) (turr.i.;g gccr Orgaged) ba.rwing debtce engaged h b) -(=q:=y step) differe,dial lockoui ve.lg 9 Gee. Inser+ 'IIC
- h. At least once per 10 years or arter any modifications which could affect diesel generator interdependence by starting both diesel generators simultaneously, during shutdown, and verifying that both diesel generators accelerate to at least (400-) rpm in less than or equal to (10) seconds. 634 ,
h -
%. At least once per 10 years by:
- 1. Draining each fuel oil storage tank, removing the accumulated sediment and cleaning the tank using a sodium hypoclorite solution, and
- 2. Performing a pressure test of those portions of the diesel fuel l
oil systes designed to Section III, subsection ND of the ASME l Code at a test pressure equal to 110 percent of the system l design pressure. . . . (parstw& bo 5"peet(tet:. Jan M 2- wlft in 30 0" 3 4.8.1.1.3 Reports - All diese l bereportedtotheCommission.glgeneratorfailures,validornon-valid,shall
, rre nt is Spccificatier. 5.0.1. Reports of diesel generator failures shall include the information recommended in Regulatory Position C.3.b of Regulatory Guide 1.108, Revision 1, August 1977. If the i number of failures in the last 100 valid tests (on a per nuclear unit basis) is greater than or equal to 7, the report shall be supplemented to include the additional information recomended in Regulatory Position C.3.b of Regulatory Guide 1.108, Revision 1, August 1977.
W _-STS 3/4 8-6 CCT 2 31980 l [.
, INSERT III 14.
- a. Verify that the diesel starts with the engine initially at normal generating temperature equilibrium (jacket water and lube oil temperature within i 10*F (5 1/2 "C) of normal operating temperature) on manual start signal, and reaches rated voltage 4160 i 420 volts
() and frequency 60 i 1.2 Hz within 10 seconds.
- b. Verify the diesel generator is manually synchronized, loaded to greater than or equal to 6083 kW in less than or equal to 120 seconds, and operates for greater than or equal to 5 minutes.
- 15. Verifying that the undervoltage load shed relays are bypassed during diesel generator load sequencing, and that the load shedding bypass is reinstated after the EPS RM0 pushbutton is depressed.
- 16. Simulating a Tower Actuation signal (TA) while the diesel generator is loaded with the permanently connected loads and auto-connected emergency (accident) loads, and verifying that the service water pump automatically trips, the cooling tower pump and fan automatically starts. After energization the steady state voltage and frequency of the emergency buses shall be maintained at 4160 i 420 volts and 60
- 1.2 Hz.
- 17. While diesel generator 1A is loaded with the permanently connected loads and auto connected emergency (accident) loads, manually connect the 1500 hp startup feedwater pump to 4160 volt bus E5. After energization the steady state voltage and frequency of the emegency bus shall be maintained at 4160
- 420 volts and 60 i 1.2 Hz.
(
Table 4.8-1 ( DIESEL GENERATOR TEST SCHEDULE _ NUMBER OF FAILURES IN LAST 20 VALID TESTS
- TEST FREOUENCY _
11 At least once per 31 days
>2 At least once per 7 days **
s.
" Criteria for determining number of failures and number of valid tests shall be in accordance with Regulatory Position C.2.e of Regulatory Guide 1.108, Revision 1, August 1977, where the number of tests and failures is determined on a per diesel generator basis. For the purposes of this test schedule, only valid tests conducted after the completion of the preoperational test requirements of Regulatory Guide 1.108, Rev 1, Aug 1977, shall be included in the computation of the "last 20 valid tests."
**This test frequency shall be maintained until seven consecutive failure free demands have been performed and the number of failures in the last 20 valid demands has been reduced to one or less.
( s 3/48-)
ELECTRICAL POWER SYSTEMS A.C. SOURCES SHUTDOWN LIMITING CONDITION FOR OPERATION 3.8.1.2 As a minimum, the following A.C. electric'al power sources shall be OPERABLE:
- a. One circuit between the offsite transmission network and the onsite Class lE distribution system, and
- b. One diesel generator with:
- 1. Day and enginc-mcented fuel tanks containing a minimum volume of & . gallons of fuel,
- 2. AfuelstoragesystemcontainingaminimumvolumeofJ_qf{gy gallons of fuel, and
- 3. A fuel transfer pump.
.\ APPLICABILITY: MODES 5 and 6. l ACTION: With less than the above minimum required A.C. electrical power sources OPERABLE, immediately suspend all operations involving CORE ALTERATIONS, positive reactivity changes, movement of irradiated fuel, or crane operation with loads over the fuel storage pool, and within 8 hours, depressurize and vent the Reactor Coolant System through a greater than or equal to (3.2) square inch vent. In addition, when in MODE 5 with the Reactor Coolant loops not filled, or in MODE 6 with the water level less than 23 feet above the reactor vessel flange, immediately initiate corrective action to restore the required sources to OPERABLE status as soon as possible. SURVEILLANCE REQUIREMENTS 4.8.1.2 The above required A.C. electrical power sources shall be demonstrated OPERABLE by the performance of each of the Surveillance Requirements of 4.8.1.1.1, 4.8.1.1.2 (except for requirement 4.8.1.1.2.a.5), and 4.8.1.1.3. ( W-STS 3/4 8-8 JUL g 7 geg;
ELECTRICAL POWER SYSTEMS 3/4.8.2 0.C. SOURCES OPERATING LIMITING CONDITION FOR OPERATION 3.8.2_.1,lAs a minimum the following D.C. electrical sources shall be OPERABLE:
- a. Train A: 1. 125-volt Ba t t e ry bank No . lA or 1C
- 2. One full capacity battery charger on Bus llA
- 3. One full capacity battery charger on Bus 11C
- b. Train B: 1. 125-volt Battery bank No. 1B or'lD
- 2. One full capacity battery charger on Bus llB
- 3. One full capacity battery charger on Bus llD APPLICABILITY: MODES 1, 2, 3, and 4. .
ACTION: the mmi.nu- regui,el bdoy1%ks seder bin I*f"*b
- a. With en: Of t.t repr:d battery banb in ; rs% restore the inoperable battery bank to OPERABLE status within 2 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within.the-following 30 hours.
- b. With one of the required full capacity chargers inoperable, demonstrate the OPERABILITY of its associated battery bank by performing Surveil-lance Requirement 4.8.2.1.a.1 within one hour, and at least once per
't hourt thereafter. If any Category A limit in Table 4.8-2 is not met, declare the battery inoperable.
> j
- x This s c1 ication in ended f r e on pla s 'th two vi 'ons of p.C. power nly. odific ions may be ecess ry, o a p1 t-uni ue q is, to a o adate diffe en s designs.
V y N N v SURVEILLANCE REQUIREMENTS 4.8.2.1 Each (950/125)-volt battery bank and charger shall be demonstrated OPERABLE:
- a. At least once per 7 days by verifying that: -
- 1. The parameters in Table 4.8-2 meet the Category A limits, and
- 2. The total battery terminal voltage is greater than or equal to (25C/120)-volts on float charge.
l'AB l @ W-STS 3/4 8-9 JUL 2 71981
)
ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
- b. At least once per 92 days and within 7 days after a battery discharge with battery terminal voltage below f2f&/110)-volts, or battery overcharge with battery terminal voltage above (-300/150)-volts, by verifying that:
- 1. The parameters in Table 4.8-2 meet the Category B limits,
- 2. There is no visible corrosion at either terminals or connectors,
' or the connection resistance of these items is less than (150 x 10 8) ohms, -an4(on conacchas be.been eac A5 sub+=d Ae resistance ,
c4 % M-cwonneehy causdrow-M cell-lo-cell reaMsg o sei I connech.on 'es'slante ad
- 3. The average electrolyte temperature of (e represent:the a'"+er)/t, af' connected cells g above (60 F)., gq cens Pe" oh
- c. ! At least once per 18 months by verifying that:
- 1. The cells, cell plates, and battery racks show no visual indication of physical damage or abnormal deterioration,
- 2. The cell-to-cell and terminal connections are clean, tight, and coated with anti-corrosion material, t
- 3. The resistance of each cell-to-cell and terminal connection is lessthanoregalto(150x108)"h
*
- end-(en connecFoe behoceg*rocA5s
* " "" * '
- U # N iN The YADio2 battery charger le'Esu5eYI ill supply at least (400-)oamperes w,hf**",,. ms at cu 4.
wmm,m at (125/25&)-volts for at least (8) hours. /50 132 h
- d. At least once per 18 months, during shutdown, by verifying that the battery capacity is adequate to supply and maintain in OPERABLE status all of the actual or simulated emergency loads for the design duty cycle when the battery is subjected to a battery service test.
- e. At least once per 60 months, during shutdown, by verifying that the battery capacity is at least 80% of the manufacturer's rating when subjected to a performance discharge test. Once per 60 month interval, this performance discharge test may b.e performed in lieu of the battery service test.
- f. At least once per 18 months, during shutdown, by giving performance discharge tests of battery capacity to any battery that shows signs of degradation or has reached 85% of the service life expected for the application. Degradation is indicated when the battery capacity drops. more than 10% of rated capacity from its average on previous performance tests, or is below 90% of the manufacturer's rating.
W-STS
~ 3/4 8-10 JUL 2 71981
( TABLE 4.8-2 . BATTERY SURVEILLANCE REOUIREMENTS CATEGORY A(1) CATEGORY B(2) l PARAMETER LIMITS FOR EACH LIMITS FOR EACH ALLOWABLE (3) DESIGNATED PILOT CONNECTED CELL VALUE FOR EACH CELL CONNECTED CELL Electrolyte > Minimum level > Minimum level Above tcp of Level indication mark, indication mark, plates, and < %" above and < " above and not maximum level maximum level overflowing indication mark- indication mark I lFloatVoltage > 2.13 volts > 2.13 volts (6)
, > 2.07 volts !. Not more than 1
i 0.020 below the average of all l Specific > 1.200(5) > 1.195 connected cells
! Gravity (4) i l Average of all Average.of all connected cells connected cells
> 1.205
> 1.195(5)
TABLE NOTATIONS (1) For any Category A parameter (s) outside the limit (s) shown, the battery may be considered OPERA 8LE provided that within 24 hours all the Category 8 measurements are taken and found to be within their allowable values, and provided a11' Category A and 8 parameter (s) are restored to within limits within the next 6 days. (2) For any Category 8 parameter (s) outside the Ifmit(s) shown, the battery may be considered OPERABLE provided that the Category 8 parameters are within their allowable values and provided the Category 8 parameter (s) are restored to within limits within 7 days. (3) Any Category 8 parameter not within its allowable value indicates an inoperable battery. (4) Corrected for electrolyte temperature and level. (5) Or battery charging current is less than [2] amps when on charge. g g (6) Corrected for average electrolyte temperature.
'f-STS il 3/4 8-X
1 ELECTRICAL POWER SYSTEMS 0.C. SOURCES SHUT 00WN LIMITING CONDITION FOR OPERATION , two 3.8.2.2 As a minimum, one EGGG/125]-volt battery bank and 4ts arreciated full-capacity chargensghall be OPERABLE. j {f % same4rav,1} @ APPLICABILITY: MODES 5 and 6. ACTION: With the required battery bank and/or full-capacity charger inoperable, immediately suspend all operations involving CORE ALTERATIONS, positi.ve reactivity changes, or movement of irradiated fuel; initiate corrective action to restore the required battery bank and full-capacity charger to OPERA 8LE status as soc'n as possible, and within 8 hours, depressurize and
, vent the Reactor Coolant System through a 3 21 square inch vent.
( SURVEILLANCE REOUIREMENTS 4.8.2.2 The above required EG50/125]-volt battery bank and full-capacity charger shall be demonstrated OPERABLE in accordance with Specification 4.8.2.11 1
'f-STS 3/4 8-)2 4
. _. - , . , - . --. - . . - _ .-. .- . . _ . - . . - - . . . _ = . . . - . .
/ ELECTRICAL POWER SYSTEMS 3/4.8.3 ONSITE POWER DISTRIBUTION OPERATING LIMITING CONDITION FOR OPERATION 3.8.3.1 The following electrical busses shall be energized in the specified mannen eith tie breeker eper (both) betwcca .edmident-busses within the unit fend between units at the sam: ;tatica);
vivlsion n A.C. Emergency Busses consisting of: '
. (4160) volt Emergency Bus #
- 2. (480) volt Emergency Bus #
- i. Divisi #2 A.C. Emergency Busses consisting of:
- 1. (416 volt Emergency Bus #
he 2. (480) vo Emergency Bus # Insed.DZ . (120) volt A.C. 1 Bus # ener ' zed from its associated inverter connected t .C. Bus # .
- f. (120) volt A.C. Vital Bus [ energized from its associated inverter connected to D.C # .
). (120) volt A.C. Vi - us # energizedfromitsassociated inverter conne to D.C. Bus # .
(120) vol .C. Vital Bus # energi d from its associated / inver connected to D.C. Bus # .
! .. /125) volt D.C. Bus #1 energized from Batte ank #1.
L (250/125) volt D.C. Bus #2 energized from Battery Ban #2. APPLIC' ABILITY: MODES 1, 2, 3, and 4. ACTION: trcu.ns (%Q F
- a. With one of the required di"in c = of A.C Emergency busses not fully energized, re-energize the di'>i A ithin 8 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within e
th.u..followip.g#,e30 u.gJu h u mim sus. su Bu3asweetw
+6 bc.+wea ne etod an "$".$,
r 'f;Na w -* coa %.** Sam + aia Mt 9 ud
- b. With one A.C. Vital _$es either not energized from its associated (fanc.) 71nverter.
Bus: or with the inverter not connected to its associated D.C.(1) re energiz least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours; and (2) re-energize the A.C. Vital from its associated inverter connected to its associated D.C. Bus within 24 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
- c. ized Ms associated Battery Bank 4 With one D.C.
re-energize Bus Lus the D.C. not fenerfffdYs associated Battery Bankswitin 2 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. m Two inverters may be disconnected from their D.C. Bus for up to 24 hours as necessary, for the purpose of performing an equalizing charge on their associated
, battery bank provided (1) their vital busses are energized, and (2) the vital busses associated with the other battery bank are energized from their associated inverters and connected to their associated D.C. Bus.
W-STS
- 3/4 8-13 SOY 2 1931
4 INSERT IV (-
- a. Train A A.C. Emergency Busses consisting of:
- 1. 4160-volt Emergnecy Bus #ES.
'2. 480-volt Emergency Bus #E51.
- 3. 480-volt Emergency Bus #E52.
- b. Train'B A.C. Emergency Busses consisting of: ,
- 1. 4160-volt Emergency Bus #E6.
- 2. 480-volt Emergency Bus #E61.
- 3. 480-volt Emergency Bus #E62.
- 4. 480-volt Emergency Bus #E64.
i
- c. 120-volt A.C. Vital Panel #1A energized from its associated inverter connected to D.C. Bus #11A.'
2
- d. 120-volt A.C. Vital Panel #1B energized from its associated inverter connected to D.C. Bus #11B.*
- e. 120-volt A.C. Vital Panel #1C energized from its associated inverter connected to D.C. Bus #11C;"
- f. 120-volt A.C. Vital Panel #1D energized from its associated inverter connected to D.C. Bus #11D.*
- g. 120-volt A.C. Vital Panel #1E energized from its associated inverter connected to D.C. Bus #11A.*
- h. 120-volt A.C. Vital Panel #1F energized from its associated inverter connected to D.C. Bus #11B.#
- i. Train A 125-volt D.C. Busses consisting of:
[ 1. 125-volt D.C. Bus #11A energized from Battery Bank 1A or IC. . 2. 125-volt D.C. Bus #11C energized from Battery Bank 1A or IC.
- j. Train B 125-volt D.C. Busses consisting of:
- 1. 125-volt D.C. Bus #11B energized from Battery Bank 1B or ID.
- 1. 125-volt D.C. Bus #11D energized from Battery Bank 1B or ID.
} i \ r
ELECTRICAL POWER SYSTEMS I . SURVEILLANCE REQUIREMENTS n a.% 4.8.3.1 The specified busses shall be determined energized in the required manner at least once per 7 days by verifying correct breaker alignment and - indicated voltage on the busses.
+
e d 4
/
o S a (, ' M-STS 3/4 8-14 fl0 V 2 1991
ELECTRICAL POWER SYSTEMS f ONSITE POWER DISTRIBUTION SHUTDOWN LIMITING CONDITION FOR OPERATION 3.8.3.2 As a minimum, the following electrical busses shall be energized in the specified manner: M %
- a. One div:.\n: ice of A.C. Emergency Busses consisting of one (4160) volt and 44 480) volt A.C. Emergency Busses lis+ct in 3.r 5 5 M aad G)(c. acNW, "UIN,% cy ac a sW. ,A ,c,,,,
- b. Two t*w A.C. Vital w ,., m e,,e,r,gized,from n their associated inverters connected to their respective D.C. Busses.
C- one, e41ke. ho t2o vett A c val Panels IE o*1F dt. One (45e/125) volt D.C. Bus energized fraas i" 5lied:in w, peci 3.tr.s.:.:ccieted bette f APPLICABILITY: MODES S and 6. ACTION: 14d eanels With any of the above required electrical busses not energized in the required f manner, immediately suspend all operations involving CORE ALTERATIONS, positive reactivity changes, or movement of irradiated fuel, initiate corrective action 1 to energize the required electrical bussegin the specified manner as soon as possible, and within 8 hours depressurize[and vent the RCS through a (s.2) square inch vent. SURVEILLANCE REQUIREMENTS
\ GMpaneb 4.8.3.2 The specified bussec shal[l be determined energized in the required manner at least once per 7 days by verifying correct breaker alignment and indicated voltage on the busses.
l l l 4, fl0V 2 1981 W-STS 3/4 8-15 l
_. _ _ _ _ _ _ . - . . _ _ _ . _ _ _ _ _ _ ~ . _ . ~ . . . . . . _ - - . ._ l ELECTRICAL POWER SYr.TEMS ONSITE POWER DISTRIBUTION i I TRIP CIRCUIT FOR INVERTER I-2A {^ %f LIMITING CONDITION FOR OPERATION _ , , 3.8.3.3 The safety related trip circuit which trips the D.C. feed from D.C. Bus 11C to inverter I-2A af ter 15 minutes of discharge from the battery shall be OPERABLE. Note - this LIMITING CONDITION FOR OPERATION is applicable only when D'.C. Bus 11C is required to be OPERABLE. APPLICABILITY: MODES 1, 2, 3, 4, 5 and 6 ACTION: With this safety related trip circuit inoperable, restore the trip circuit to OPERABLE status within 7 days or de-energize the D.C. feed to inverter I-2A by tripping the D.C. circuit breaker in D.C. Bus llc. Verify that this breaker is open once per 7 days thereaf ter. SURVEILLANCE REQUIREMENTS 4.8.3.3 The safety related trip circuit shall be demonstrated operable at least once per 18 months. I . I l' . l l l 3/4 8- 16
.%(
ELECTRICAL POWEP SYSTEMS 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES A C cl Rc tr iT,".5,.n IN cnurgp.ue t.-- pg y i -v 5I
- m. DE cm .m.Al P MMy con n1,co cn o c e.7.,7. H NhEA/ T.c' cen3 non,.ecers
, yr-, v e g i
LIMITING CONDITION FOR OPERATION 3.8.4.I The circuit breakers feeding the following loads inside containment shall be padlocked in the open position: REFuGAlam CANAL- S K let*t E R Pu ** P l -SF- P.17 2. POLAR GANTRY CRANE (MM-CR-3) DISTRIBUTION PANEL PP-7A DISTRIBUTION PANEL PP-7B 800 CourRok Ct.turGR, c ttMcE Ptx Twtg l-F H - R E s1 l APPLICABILITY: MODES 1, 2, 3, and 4. i
? l hE_XCEPTION: If any of the above mentioned loads are required for brief dura- l tions (not to exceed 72 hours) during plant operation, the per-tinent circuit breaker can be unlocked and be closed for the required duration provided this change in breaker position becomes part of the applicable operating procedure used for the work inside containment. i SURVEILLANCE REQUIREMENTS 4.8.4.I Verify at least once per'3I days that the circuit breakers listed in Specs h fio 3.8.4.i are padlocked in the open position. !. l \
n,,1,3,4,png g low g,g an1 ,g g Ag,ye ,c$u,,,j ,,,,,u,4, Asso ctnied etnu'il b r eaker ts) in ne speciGA panelce w//44
~ '
1 kour. _ 3/4 8- O
\
. l ELECTRICAL POWER SYSTEMS I f
-3 /A . S . ? ELECTRIGAL-EQUIPMENT PROTECTIVE DEVICES '
CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES LIMITING CONDITION FOR OPERATION 2 3.8.4.1 All containment penetration conductor overcurrent protective devices shown in Table 3.8-1 shall be OPERABLE. lder X APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: With one or more of the containment penetration conductor overcurrent protective device (s) chce " T;ble 3.0-1-inoperable:
- a. Restore the protective device (s) to OPERABLE status or de-energize the circuit (s) by tripping the associated backup circuit breaker within 72 hours, declare the affected System or component inoperable, and verify the backup circuit breaker to be tripped at least once per 7 days thereafter; the provisions of Specification 3.0.4 are not applicable to overcurrent devices in circuits which have their backup ..
circuit breakers tripped, or '
- b. Be in at least HOT STANDBY within the next 6 hours and in COLD SHUTOOWN within the following 30 hours.
SURVEILLANCE REOUIREMENTS 2. 4.8.4.\ All containment penetration conductor overcurrent protective devices A r- H T;ble 2.0 i shall be demonstrated OPERABLE:
- a. At least once per 18 months: .
13 9
- 1. By verifying that the medium voltage (4-M kV) circuit breakers are OPERABLE by selecting, on a rotating basis, at least 10% of the circuit breakers ^' ^" " "--^ '-"a and performing the following: ,
(a) A CHANNEL CALIBRATION of the associated protective relays,
-and See 'Enscet 3E l (b) An integrated system functional test which includes simulated automatic actuation of the system and verifying that each relay and associated circuit breakers and control circuits function as designed and :: :pecifiedgk-Tabic 3.0-1.
l rwe h t exe,eA q %e pene.tmbon w/4,fa4 f4 d @ IS e.y o" 199; W-STS 3/4 8-14
INSERT V. l (due'to the large currents involved, it is impractical to inject primary side I signals to current transformers. Therefore, the channel calibration will be performed by injecting a signal on the secondary side of these transformers at
- their test plug), and b
I-
ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) (c) For each circuit breaker found inoperable during these functional tests, an additional representative sample of at least 10% of all the circuit breakers of the inoperable type shall also be functionally tested until no more failures are found or all circuit breakers of that type have been functionally tested. See Insert E g
- 2. By selecting and functionally testing a representative sample of at least 10% of each type of lower voltaae circuit breakers, Circuit breakers selected for functional testing shall be
. selected on a rotating basis. For the lower voltage circuit N ove' load devices) breakersVthe nominal trip setpoint and short circuit response times are listed in J ai s3 @ 4 3 Tc: ting-of the:c circuit
( brecke"s shal' centist Of 'njccang & current in excess cf-the h breakers nemiaal setpoint and me:Suring the rc: pen:c time. The S c L sert Q meccured respen:0 time wi be ccmpcred te the mcnufccturcr's d2t: tc i nsure that it i: Ic:: then er caus! te : vcluc speci-
#!ed by tha man"fecturer. Circuit breakers #found inoperable i
during functional testing shall be restored to OPERABLE status
; enJ/or eve,loaJ Jevue3 pri r to resuming operation. For each circuit breaker +found inoperable during these functional tests, an additional repre-sentative sample of at least 10% of all the circuit breakers +of the inoperable type shall also be functionally tested until no
- more failures are found or all circuit breakers +of that type
' have been functionally tested. seeInsen3zI ~ g
- 3. By selecting and functionally testing a representative sample of each type of fuse on a rotating basis. Each representative sample of fuses shall include at least 10% of all fuses of that type. The functional test shall consist of a non-destructive resistance measurement test which demonstrates that the fuse meets its manufacturer's design criteria. Fuses found inoper-able during these functional tests shall be replaced with OPERABLE fuses prior to resuming operation. For each fuse found inoperable during these functional tests, an additional representative sample of at least 10% of all fuses of that type shall be functionally tested until no more failures are found ,
or all fuses of that type have been functionally tested.
- 4. S ee. Ins e *+ ~v i i i .
- b. Atleastonceper60monthsb@ysubjectingeachcircuitbreakertoan inspection and preventive maintenance in accordance with procedures prepared in conjunction with its manufacturer's recommendations.
l l 19 W-STS 3/4 8-N W2 GSI 1 1
INSERT VI i O If the trip time does not exceed the penetration withstand time but is outside the verification time listed in-Table-3*t, the device shall be replaced or recalibrated. No additional testing is required. INSERT VII The functional test shall consist of injecting a current input at the specified trip setpoint to each selected circuit breaker and/or overload device and verifying that each circuit breaker and/or overload device is OPERABLE by tripping at a time value not to exceed the penetration withstand time shown in-Tebic-- 3.8 L.- INSERT VIII Corrective actions for any generic degradation of overcurrent protective devices, such as setpoint drift, manufacturing deficiencies, material defects, etc., shall be applicable to all (Class IE and non-Class 1E) protective devices of identical design. \
i
, . , -- ; . j , e ---- ,
t, G TABLE 3.8-1 CONhAINMENTPENETRATIONCONOUCTOR OVERCURRENT PROTECTIVE DEVICES s DEVICE NUMBER + SYSTEM-AND LOCATION POWERED
- 1. 6900 VAC Reactorboolantpump (Primary breaker) 1
/
(Backup breaker) j' 2 3
/ 4
- 2. 480 VAC from M0A0 Centers /
Listall;primarybreakers\ / . Backup breakers / Backup breakers
/
- 3. 480 VAC from MCC
/
List all; primary breakers / \ ( ' Backup breakers Backup breakers /
/ \
- 4. 125V DC Lichtinq List all; primary breakers /
/
Backup breakers Backup breakers
- 5. 440 VAC CROM Power y
Primary breakers Backup breakers Backup breake'rs s N. V l N I
\
W-STS 3/4 8-20
ELECTRICAL POWER SYSTEMS MOTOR-OPERATED VALVES THERMAL OVERLOAO PROTECTION , LIMITING CONDITION FOR OPERATION I 3.8.4.3 The thermal overload protection of each valve'given in T2 .c p+er 3.0-2 shall be OPERABLE. APPLICABILITY: Whenever the motor-operated valve is required to be OPERABLE. ACTION: With the thermal overload protection for one or more of the above required valves inoperable? [ continuously] bypass the inoperable thermal overload within 8 hours [; restore the inoperable thermal overload to OPERABLE status within 30 days] or declare the affected valve (s) inoperable and apply the appropriate ACTION Statement (s) for the affected system (s). SURVEILLANCE REOUIREMENTS 4.8.4.3 The thermal overload protection for the above required valves shall a h bedemonstratedOPERABLEg p onthemotorstarterby...Q;r.st,,onceper18monthsandfollowingmaintenance
. re.ar.cc o f ; 0:!A=:L C*.LICC"!C!1 of a represen-valves. Thue. thermal overlead aevtcestative will sample of at least he v=placed wit /s25% of all pecca/ 'Amthermal overloa aevlees.
- A i h e r ,m n / , ,e rl o 4 J p r ,[ e c H ,,) d evi; c 15 coris,'de. d i ,,pem 6Ie iF Me tc\a<g e p c en fia,y -tsne is not- in acatedar,a win th crolces'a l csfnLInslieJ e A Ifi , afplica& rel<uf curvc.
l W-STS 3/4 3-22 i
( I 4 TABLE 3.8-2 -
/
MOTOR-OPERATED VALVES THERMAL OVERLOAO PROTECTION BYPASS DEVICE / SYSTEM (S) VALVE NUMBER (Continuous)(Accident Conditions)(No) AFFECTED ,
/
,/ .
* /
. /
/
/
N \ /
/
s /
/ -
/
/
,/
s N ( / t
/ \
/
j \
. '\
/ N
. f .
x N NN l - ! \, s S. ! / i l i
\
\
W-STS 3/4 8-23
<.s - - _ - . - - - - _ _ - - - _ . . - . _ _ . _ - - _ - - - _ - - _- - _ _ _ - _ _ - . _ _ _ _ - _ _ . - _ _ - - - . _ _ _ - - _ - _ . . _
JUSTIFICATIONS I Section 3/4.9 In the text of Section 3/4.9 where a capital letter with a circle around it appears, please refer to this sheet for the appropriate justification. A. Core alterations should be allowed to continue if one source range channel is operable and a temporary detector is operable with containment or control room indication. B. The wordiag changes and number value changes on this sheet, (except as shown by other juo r.ifications), are plant specific for Seabrook Station and come from the manufacturers technical manuals. C. The Seabrook Station crane has only limit switches to prevent this function. - D. Plant specific system designation for Seabrook Station as used throughout the FSAR. E. The Fuel Storage Building Emergency Air Cleaning System is operating prior to any fuel movement, therefore, there is no automatic feature to stnrt this system on a high radiation alarm. Thus, this paragraph can be deleted. F. Seabrook Station plant specific data. k t
3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION LIMITING CONDITION FOR OPERATION 3.9.1 -Mith th: r;;;ter vessel-h: d-closure belts less tha. fully L ..5iened e.-
~
-with-the heed "a T/ d, [he boron concentration of all filled portions of the' Reactor Coolant System and the refueling canal shall be maintained uniform and sufficient to ensure that the more restrictive of the following reactivity conditions is met:
- a. Either a K,ff of 0.95 or less
- b. A boron concentration of greater than or equal to (2000) ppm.
APPLICABILITY: MODE 6*. ACTION: With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS or positive reactivity changes and initiate and continue boration at greater than or equal to (30 gpm , of a solution containing greater than or equal to (7ooo) ppm boron or its (f() equivalent until K,77 is reduced to less than or equal to 0.95 or the boron concentration is restored to greater than or equal to (2000) ppm, whichever is the more restrictive. SURVEILLANCE REQUIREMENTS 4.9.1.1 The more restrictive of the above two reactivity conditions shall be determined prior to: *
- a. Removing or unbolting the reactor. vessel head, and
- b. Withdrawal of any full length control rod in excess of 3 feet from its fully inserted position within the reactor pressure vessel.
4.9.1.2 The baron concentration of the reactor coolant system and the refueling canal shall be determined by chemical analysis at least once per 72 hours.
"The reactor shall be maintained in MODE 6 whenever fuel is in the reactor vessel with the vessel head closure bolts less than fully tensioned or with
(. the head removed. W-STS 3/4 9-1 SEP 151981
REFUELING-OPERATIONS 3/4.9.2 INSTRUMENTATION l LIMITING CONDITION FOR OPERATION 3.9.2 As a minimum, two source range neutron flux monitors shall be OPERABLE each with continuous visual indication in the control room and one with audible indication in the containment and control room. , APPLICABILITY: MODE 6. ACTION: l a. With one of the above required monitors inoperable-cr not oper: ting, l 4 ::dictcly urpend s epe*etion: 4 velving CORE ALTERATIONS,er l
@ _ 4+4u.
.;r.u
.. +4o4+o ,.w,-,.
.s . u.;.w.;ww.mc,_.w,g,,,,.,,,,,.a - - - ~ - - ~ ~ --~ m if , w ,, .y a,6u % ,i,o n g o i.e -ts <-+.ime+
, cha-S 3 l
- b. With both of the above required monitors inoperable or not operating,
. determine the boron ' concentration of the reactor coolant system at I least once per 12 hours. ,
I ( SURVEILLANCE REQUIREMENTS l 4.9.2 Each source range neutron flux monitor shall be demonstrated OPERABLE by performance of:
- a. A CHANNEL CHECK at least once per 12 hours,
- b. A ANALOG CHANNEL OPERATIONAL TEST within 8 hours prior to the initial I
start of CORE ALTERATIONS, and
- c. A ANALOG CHANNEL OPERATIONAL TEST at least once per 7 days.
t I ! W-STS 3/4 9-2 I 0 00I
REFUELING OPERATIONS [ 3/4.9.3 DECAY TIME LIMITING CONDITION FOR OPERATION l 3.9.3 The reactor shall be subcritical for at least 100 hours. APPLICABILITY: During movement of irradiated fuel in the reactor -
, pressure vessel.
ACTION: With the reactor subcritical for less than 100 hours, suspend all operations involving movement of irradiated fuel in the reactor pressure vessel. SURVEILLANCE REQUIREMENTS 4.9.3 The reactor shall be determined to have been subcritical for at least 100 hours by verification of the date and time of subcriticality prior to movement of irradiated fuel in the reactor pressure vessel. 4 w.STS 3/4 9-3 UL 2 71991
1 { I REFUELING OPERATIONS t 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS LIMITING CONDITION FOR OPERATION 3.9.4 The containment building penetrations shall be in the following status:
- a. The equipment door closed and held in place by a minimum of four bolts,
- b. A minimum of one door in each airlock is closed, and
- c. Each penetration providing direct access from the containment atmosphere to the outside atmosphere shall be either:
- 1. Closed by an isolation valve, blind flange, or manual valve, or
- 2. Be capable of being closed by an OPERABLE automatic Containment Purge and Exhaust isolation valve.
APPLICABILITY: During CORE ALTERATIONS or movement of irradiated fuel within the containment. ( ACTION: ( With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS or movement of irradiated fuel in the containment building. ' SURVEILLANCE REQUIREMENTS 4.9.4 Each of the above required containment building penetrations shall be determined to be either in its closed / isolated condition or capable of being closed by an OPERABLE automatic Containment Purge and Exhaust isolation valve J within 100 hours prior to the start of and at least once per 7 days during [ CORE ALTERATIONS or movement of irradiated fuel. in the containment building by:
- a. Verifying the penetrations are in their closed / isolated condition, or
- b. Testing the Containment Purge and Exhaust isolation valves per the applicable portions of Specification 4.0.4.2.
3617 i W-STS 3/4 9-4
REFUELING OPERATIONS l 3/4.9.5 COMMUNICATIONS LIMITING CONDITION FOR OPERATION 3.9.5 Direct communications shall be maintained between the control room and personnel at the refueling station. APPLICABILITY: During CORE ALTERATIONS. ACTION: When direct communications between the control room and personnel at the refueling station cannot be maintained, suspend all CORE ALTERATIONS. l SURVEILLANCE' REQUIREMENTS
- 4.9.5 Direct ~ communications between the control room and personnel at the refueling station shall be demonstrated within one hour prior to the start of
}. and at least once per 12 hours during CORE ALTERATIONS. I r i l . L JUL 2 71961 W-STS 3/4 9-5 l h
t REFUELING OPERATIONS 3/4.9.6 "Al I!'ULATOR CRA;;E RE FuE LING. t1ACHIME LIMITING CONDITION FOR OPERATION 3.9.6 The$*:sf EcNEineandauxiliaryheintshallbeusedformovementof drive rods or fuel assemblies and shall be OPERABLE with: a. men hdist The acr.ipu,.ois. crane used for movement of fuel assemblies having:
- 1. A minimum capacity of (2750) pounds, ar.d 2.
An overload cutoff limit less than or equal to (2700) pounds. b. The auxiliary hoist used for latching and unlatching drive rods having:
- 1. A minimum capacity of (610) pounds, and 2.
A load indicator which shall be used to prevent lifting loads in excess of (600) pounds. APPLICABILITY: During movement of drive rods or fuel assemblies within t the reactor pressure vessel. ACTION: tre L % ~oekke, With the requirements for cranf and/.or hoist OPERABILITY not satisfied, suspend use of any inoperable maripulat:F"PrThe and/or auxiliary hoist from operations involving the movement of drive rods and fuel assemblies within the reactor pressure vessel. I SURVEILLANCE REQUIREMENTS 4.9.6.1
-ah has+ )
Each tanipulatar cranc used for movement of fuel assemblies within the reactor pressure vessel shall be demonstrated OPERABLE within 100 hours prior to the start of such operations by performing a load test of at least l 4000 60769) pounds and demonstrating an automatic load cutoff when the crane load exceeds (2700) pounds. 4.9.6.2 Each auxiliary hoist and associated load indicator used S for movement of drive rods within the reactor pressure vessel shall be demonstrated OPERABLE within 100 hours prior to the start of such operations by performing a load test of at least (610) pounds. l l W-STS 3/4 9-6
REFUELING OPERATIONS I 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE POOL BUILDING LIMITING CONDITION FOR OPERATION 3.9.7 Loads in excess of .2m, pounds shall be prohibited from travel over fuel assemblies in the storage pool. APPLICABILITY: With fuel assemblies in the storage pool. ACTION: With the requirements of the above specification not satisfied, place the crane load in a safe condition. ', SURVEILLANCE REQUIREMENTS t D 4.9.7 Crane interlocks -end physical :tep; which prevent crane travel with loads in excess of 2/oo pounds over fuel assemblies shall be demonstrated OPERABLE within 7 days prior .to crane use and at least once per 7 days thereafter during crane operation. k JUL 2 71981 W-STS 3/4 9-7
( REFUELING OPERATIONS 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION HIGH WATER LEVEL LIMITING CONDITION FOR OPERATION
. 3.9.8.1 At least one residual heat removal (RHR) loop shall be OPERABLE and in operation.* ,
APPLICABILITY: MODE 6, when the water level above the t'op of the reactor vessel flange is greater than or equal to 23 feet. ACTION: With no RHR loop OPERABLE and in operation, suspend all operations involving I an increase in the reactor decay heat 1. cad or a reduction in boren concentration of the Reactor Coolant System and immediately initiate corrective action to return the required RHR loop to OPERABLE and operating status as soon as possible. Close all containment penetrations providing direct access from the containment atmosphere to the outside atmosphere within 4 hours. ( SURVEILLANCE REOUIREMENTS 4.9.8.1 At least one RHR loop shall be verified in operation and circulating , reactor coolant at a flow rate of greater than or equal to [2800] gpm at g least once per 12 hours.
"The RHR loop may be removed from operation for up to 1 hour per 8-hour period
( during the performance of CORE ALTERATIONS in the vicinity of the reactor vessel hot legs. W-STS 3/4 9-8
l REFUELING OPERATIONS I l LOW WATER LEVEL ; LIMITING CONDITION FOR OPERATION 3.9.8.2 Two independent residual heat removal (RHR) loops shall be OPERABLE, and at least one RHR loop shall be in operation.* APPLICABILITY: MODE 6, when the water level above the top of the reactor vessel flange is less than 23 feet. ACTION:
- a. With less than the required RHR loops OPERABLE, immediately initiate corrective action to return the required RHR loops to OPERABLE status, or to establish greater than or equal to 23 feet of water above the reactor vessel flange, 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. Close all containment penetrations providing l direct access from the containment atmosphere to the outside I
- atmosphere within 4 hours.
~
SURVEILLANCE REOUIREMENTS 4.9.8.2 At least one RHR loop shall be verified in operation and circulating reactor coolant at a flow rate of greater than or equal to [2800] gpm at least once per 12 hours.
- Prior to initial criticality, the RHR loop may be removed from operation for up to 1 hour per 8-hour period during the performance of CORE ALTERATIONS in the vicinity of the reactor vessel hot legs.
$ (
W-STS 3/4 9-9
REFUELING OPERATIONS 3/4.9.9 CONTAINMENT PURGE AND EXHAUST ISOLATION SYSTEM LIMITING CONDITION FOR OPERATION 3.9.9 The Containment Purge and Exhaust Isolation System shall be OPERABLE. APPLICABILITY: During CORE ALTERATIONS or movem.ent of irradiated fuel within the containment. ACTION: I With the Containment Purge and Exhaust isolation System inoperable, close each of the Purge and Exhaust penetrations providing direct access from the contain-ment atmosphere to the outside atmosphere. The provisions of Specification 1s.3 and 3.0.4 are not applicable. SURVEILLANCE REQUIREMENTS 4.9.9 The Containment Purge and Exhaust Isolation System shall be demonstrated OPERABLE within 100 hours prior to the start of and at least once per 7 days during CORE ALTERATIONS by verifying that containment Purge and Exhaust isolation occurs on manual initiation and on a high radiation test signal from each of the containment radiation monitoring instrumentation channels. ( I W-STS 3/4 9-10
REFUELING OPERATIONS 3/4.9.10 WATER LEVEL - REACTOR VESSEL f LIMITING CONDITION FOR OPERATION 3.9.10 At least 23 feet of water shall be maintained over the top of the
- reactor pressure vessel flange.
APPLICABILITY: During movement of fuel assemblies or control rods within the reactor pressure vessel when either the fuel assemblies being moved or the fuel assemblies seated within the reactor pressure vessel are irradiated. ACTION: With the requirements of the above specification not satisfied, suspend all operations involving movement of fuel assemblies or control rods within the pressure vessel. I i SURVEILLANCE RE0VIREMENTS 4.9.10 The water level shall be determined to be at least its minimum required depth within 2 hours prior to the start of and at least once per 24 hours thereafter during movement of fuel assemblies or control rods. l \ l l -k JUL 2 71931 W-STS 3/4 9-11
REFUELING OPERATIONS ( 3/4.9.11 WATER LEVEL-STORAGE POOL LIMITING CONDITION FOR OPERATION 3.9.11 At least 23 feet of water shall be maintained over the top of irradiated fuel assemblies seated in the storage racks. APPLICABILITY: .Whenever irradiated feel assemblies are in the storage pool. ACTION: With the requirements of the above specification not satisfied, suspend all movement of fuel assemblies and crane operations with loads in the fuel storage areas and restore the water level to within its limit within 4 hours. The. provistbns aff S ecWea hhn 3.o.3 a,id F.o.y are ci /9 dj4 ( SURVEILLANCE REQUIREMENTS 4.9.11 The water level in the storage pool shall be determined to be at least its minimum required depth at least once per 7 days when irradiated fuel assemblies are in the fuel storage pool. T ! JUL 2 71981 V-STS 3/4 9-12
. REFUELING OPERATIONS F(A E L. STOR 4G L S u lL.DI M G- EhERG enc 4 RIR cl.EANIN&- S vS TEH 3/4.9.12 ST0!! ACE POOL ^JPs CLE.^NUP SYST "
LIMITING CONDITION FOR OPERATION 3.9.12 Two independent fuel storagebilai peclc.r q exAnud Olferblns c!canup sy; ten shall be OPERABLE. APPLICABILITY: Whenever irradiated fuel is in the storage pool. ACTION: (buitas, exhaus+ C1+cr +vdno9
- a. With one fuel storage iccl ci c!nnup ;y; tem inoperable, fuel
- j. movement within the storage pool or crane operation with loads over l the storage pool may proceed provided the OPERABLE fuel storage. pee 1 Kir ciccnup systc; is capable of being powered from an OPERABLE emergency power source and is in operation and discharging through at least one train of HEPA filters and charcoal adsorbers.
, b. With no fuel storage,pcc! cir ciccnup ;y;te: OPERABLE, suspend all l operations involving movement of fuel within the storage pool or l crane operation with loads over the storage pool until at least one i spent fuel storage poc! ci cle'nup cystem is restored to OPERABLE l / status. O
\. .D.3 a nA y l i c. The provisions of Specification 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.9.12 The above required fuel storage pool air cleanup systems shall be demonstrated OPERABLE:
- a. At least once 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 I communicating with the system by: l l l l JUL 2 71981 W-STS 3/4 9-13
_ . _._.y . . _ _ _ -. - REFUELING OPERATIONS SURVEILLANCE REQUIREMENTS (Continued)
- 1) Verifying that the cleanup system satisfies the in place penetration and bypass leakage testing acceptance criteria
([fhf][) of less tha6'SF]% and uses the test procedure guidance in . Regulatory Positicns C.S.a. C.S.c, and C.S.d of Regulatory Guide 1.52, Revision 2, March 1978, and the system flow rate is kler cfm i 10%;
- 2) Verifying, within 31 days after removal, that a laboratory analysis of a representative carbon sample obtained in accor- I dance with Regulatory Position C.6.b of Regulatory Guide 1.52, ,
Revision 2, March 1978, meets the laboratory testing criteria 3 of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978, for a methyl iodide penetration of less than [ ]%; and
- 3) Verifying a system flow rate of jdgy cfm 10% during system l operation when tested in accordance with ANSI N510-1999.
1960
- c. After every 720 hours of charcoal adsorber operation by verifying, within 31 days after removal, that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978, for a methyl iodide penetration of less than [**]%.
- d. At least once per 18 months by:
- 1) Verifying that the pressure drop across the ccmbined HEPA /b tb " l filters and charcoal adsorber banks is less than EG}' inches l Water Gauge while operating the system at a flow rate of
& cfm 10%, I
-B} -Yeri fying that er, -a Migh Dadiatier test :ign21, tha system- l automatic:!!y : tart: (urle:: :!r:cdy operating) and dirccts=WL6 (b) exheU5t flew th-cugh the ugpa <<!ter; :nd c3c7c;;7 esenrso, b;nks, e l
1 4 1 E'STS 3/4 9-14
l 1 f REFUELING OPERATIONS SURVEILLANCE REQUIREMENTS (Continued) 2.10 Verifying that the system maintains the spent fuel storage pool area at a negative pressure of greater than or equal to [l/4] inch Water Gauge relative to the outside atmosphere during systen operation,
, l 3'4) Verifying that the filter cooling bypass valves can be manually opened, and 9 '5) Verifying that the heaters dissipate 95I // kW when l tested in accordance with ANSI N510-197&.
146%)
- e. After each camplete or partial replacement of a HEPA filter bank, by verifying that the' cleanup system satisfies the in place penetr ~ n and bypass leakage testing acceptance criteria of less than [ % in
- accordance with ANSI N510- for a 00P test aerosol while operating the system at a flow rate 10%.
_Ldzr cfm
- f. After each complete or partial replacement of a charcoal adsorber bank, by verifying that the cleanup system satisfies the in place (
t, penetration and bypass-leakage testing acceptance criteria of less \ o og than5f]%inaccordancewithANSIN510- for a halogenated j nydrocarbon refrigerant test gas while ating the system at a flow rate of h cfm 10%. ( (.
'f-STS 3/4 9-15
JUSTIFICATION Section 3/4.10 In the text of Section 3/4.10 where a capital letter with a circle around it appears, please refer to this sheet for the appropriate justification. A. All rods at Seabrook Station are full length rods. There are no part length rods. . B. Specification 3.1.3.7 is for part length rods. Since there are no part length rods at Seabrook Station this Specification is not necessary. C. USNRC Regulatory Guide 1.68 Rev. 2 requires RCCA drop testing to be performed at hot and cold, full flow and no flow conditions in order to bound the situations where a scram might occur. The current Technical Specifications only allow the required pumps to be de-energized for 1 hour. Since one hour is not enough time to acquire a complete set of RCCA drop times, compliance with the Reg. Guide requires multiple unnecessary start-stop cycles on the reactor coolant pumps and/or the residual heat removal pumps. In order to facilitate the required start-up testing, Special Test Exceptions 3.10.4.1 should be extended to apply to Sections 3.4.1.2, 3.4.1.3, and 3.4.1.4. Testing performed within the restrictions of T.S. 3.10.4.1 will not result in increased risk to the core or the public. D. Seabrook Station plant specific data. E. The design of the Seabrook DRPI will not allow this Surveillance Requirement to be performed on the shutdown banks. F. Deletion of the requirement to flux map is based on the fact that the specific tests that might invoke this Special Test Exception are done such that full core flux maps are of little or no value.
- 1. Rod worth and rod drop time testing is done at HZP.
- 2. Dropped, stuck and ejected rod worth tests require the performance of flux mapping to verify test predictions and are performed at 50%
power or less.
- 3. Incore/excore calibrations are usually performed within the confines of Specification 3.2.1 and its ACTION statement a.2.a.2. Only on an infrequent basis will incore/excore calibrations be performed under this test exception. In either case quarter-core flux mapping is performed every 2-4% change in AFD as part of the test procedure.
Furthermore full core mapping with the core in an AFD transient would Produce questionable results due to the length of time involved. l l
i ( 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN MARGIN LIMITING CONDITION FOR OPERATION 3.10.1 The SHUT 00WN MARGIN requirement of Specification 3.1.1.1 may be suspended for measurement of control rod worth and shutdown margin providedithe.7
- 4. eactivity equivalent to at least the highest estimated control rod7 worth is available for trip insertion from OPERABLE control rod (s),
hd J
% All part-!cngth red: 2r wi thd rat:1 t at les:t the 180-step pc:ition and OPERABLE.
APPLICABILITY: MODE 2. ACTION:
- a. With any full-length control rod not fully inserted and with less than the above reactivity equivalent available for trip insertiongr the part-
, !cngth r:d: 10t with" their withdraw:1 li-itt, immediately initiate and ', continue boration at greater than or equal to (30) gpm of a solution containing greater than or equal to (7ood ppm baron or its equivalent until 4 the SHUTDOWN MARGIN required by Specification 3.1.1.1 is restored. UJ)
- b. With all full-length control rods fully inserted and the reactor subcritical by less than the above reactivity equivalent, immediately initiate and continue boration at greater than or equal to (3o) gpm of a solution containing greater than or equal to (m 4 ppm boren or its equivalent until the SHUTDOWN MARGIN required by Specification 3.1.1.1 is restored.
SURVEILLANCE REQUIREMENTS 4.10.1.1 The position of each full-length 2nd part-length rod either partially or fully withdrawn shall be determined at least once per 2 hours. 4.10.1.2 Each full-length rod not fully inserted shall be demonstrated capable of full insertion when tripped from at least the 50% withdrawn position within 24 hours prior to reducing the SHUTDOWN MARGIN to less than the limits of Specification 3.1.1.1. The part-!ength "ede shall be demen:trated OPERABLE by mcving each
@ p'a.10.1.2 rt-length red greater than e- equal to 10 steps within 4 hem 5 prier t: -
7:ducing--the-SHUT 00WM _uanc;g ;g ;g33 ;gaa gg; 35_j;g of 3p;;;;;c3;;cn 3,7 ;,7,_
\
W-STS 3/4 10-1 NOV 2 01980 1
SPECIAL TEST EXCEPTIONS 3/4.10.2 GROUP HEIGHT, INSERTION, AND POWER DISTRIBUTION LIMITS F < LIMITING CONDITION FOR OPERATION 3.10.2 The group. height, insertion, and power distribution limits of Specifications 3.1.3.1, 3.1.3.5, 3.1.3.6,-3.'.3.7, 3.2.1, and 3.2.4 may be suspended during the performance of PHYSICS TESTS providedi h J
% IheTHERMALPOWERismaintainedlessthanorequalto85%ofRATED ERMAL POWER,.and-
,'b'. -The-1-imits-of Specificat-ions-3-2r2-and-3-2-3-are-maintained and: determine _d=at the f"equencies-speci-f-ied-in-Speci ficot.iusi
-4A0dr24elow-APPLICABILITY: MODE 1.
ACTION: . Thc A L POW EP, 3 reMe." M Tf 7, 86T6D M64ML. PdOJER e,ibr: With -anyM the4imi-ts-e. Specif-icat4cos 3.2.2 er 3.2.3-being-exceeded while -
.-the-requTrements-obSpeci ficat-ions 3. i . 3. i , 3. i . 3. 3, 3.1. 3. S , 3.1. 3. 7, 3. 2.1, -and 3,214-ere-suspended;-either:- j
- a. Reduce THERMAL POWER to less suf#icient hn owthe ts :eYi:fy em.l to E7, R&TED
^.CTION 7ttfetAt.
requi-amant-s- (
-ef Sywuii,icat.iuns 3.2.2 end 3.2.0, en .
PouoER toque l kou.r y av
- b. Be in HOT STANDBY within 6 hours.
SURVEILLANCE REQUIREMENTS 4.10.2.1 The THERMAL POWER shall be determined to be less than or equal to 85% of RATED THERMAL POWER at least once per hour during PHYSICS TESTS. 10.2.2 The-Surveillencs-Requir-ements-of-the-below li:ted-SpeeMc2Hane
-be performed-atMeast--once-per-12-hours-dur-ing 2HYSICS-TESTS.
- s. Speci-f-icat-ica 4. 2. 2. 2-and ' . 2. 2. 3-
-tr:---Speci-f-ica tion-4. 2. 3. 2
( ( W _-STS 3/4 10-2 NOV 2 01980
SPECIAL TEST EXCEPTIONS 3/4.10.3 PHYSICS TESTS LIMITING CONDITION FOR OPERATION 3.10.3 The limitations of Specifications- 3.1.1.3,3.1.1.4,3.1.3.1,3.1.3.5,d.nd 3.1.3.6, ed 3.1.2.7 may be suspended during the performance of PHYSICS TESTS provided:
- a. The THERMAL POWER does not exceed 5% of RATED THERMAL POWER,
- b. The reactor trip setpoints on the OPERABLE Intermediate and Power Range Nuclear Channels are set at less than or equal to 25% of RATED THERMAL POWER, and c.
The Reactor is greater Coolant than System or equal lowest operating loop temperature (T,y9) to (&31-)"F. S't .1 APPLICABILITY: MODE 2. ACTION:
, a. With the THERMAL POWER greater than 5% of RATED THERMAL POWER, immediately l,
open the reactor trip breakers. b. With a Reactor Coolant System operating loop temperature (T,yg) less than
@( )*F, restore T,yg to within its limit within 15 minutes or be in at least HOT STANDBY within the next 15 minutes.
SURVEILLANCE REQUIREMENTS 4.10.3.1 The THERMAL POWER shall be determined to be less than or equal to 5% of RATED THERMAL POWER at least once per hour during PHYSICS TESTS. 4.10.3.2 Each Intermediate and Power Range Channel shall be subjected to a ANALOG CHANNEL OPERATIONAL TEST within 12 hours prior to initiating PHYSICS TESTS. 4.10.3.3 The Reactor Coolant System temperature (T,yg) shall be determined to be greater than or equal to (6%) F at least once per 30 minutes during PHYSICS TESTS. 5yl - SEP 151981 W-STS 3/4 10-3
i l SPECIAL TEST EXCEPTIONS , ,, _, 3/4.10.4 REACTOR COOLANT LOOPS _ _ _ . _ . LIMITING CONDITION FOR OPERATION 3.10.4 The limitations of the following requirements may be suspended: a.
~ s 3.u. t.2. ,'S 3. l.3, ua 3 9. l.y1 @
Specifications 3.4.1.1V- During the performance of startup and PHYSICS TESTS in MODE 1 or 2 provided:
' 1) The THERMAL POWER does not exceed the P-7 Interlock Setpoint, .e and -
- 2) The Reactor Trip Setpoints on the OPERABLE Intermediate and Power Range channels are set less than or equal to 25% of RATED THERMAL POWER.
- b. Spec'i fication 3.4.1.2 - During the performance of hot rod drop time measurements in MODE 3 provided at least two reactor coolant loops as listed in Specification 3.4.1.2 are OPERABLE..
APPLICABILITY: Ouring operation below the P-7 Interlock Setpoint. or performance of hot rod drop time measurements. ACTION:
, a. With the THERMAL POWER greater than the P-7 Interlock Setpoint during the performance of startup and PHYSICS TESTS, immediately open the Reactor trip breakers.
- b. With less than the above required reactor coolant loops OPERABLE during performance of hot red drop time measurements, immediate-ly open the reactor trip breakers and comply with the provisions of the ACTION statements of Specification 3.4.1.2. .
SURVEILLANCE REOUIREMENTS _ _ , 4.10.4.1 The THERMAL POWER shall be determined to be less than P-7 Interlock Setpoint at least once per hour during startup'and PHYSICS TESTS. 4.10.4.2 Each Intermediate and Power Range channel, and P-7 Interlock shall be subjected to an ANALOG CHANNEL OPERATIONAL TEST within 12 hours prior to initiating startup and PHYSICS TESTS. 4.10.4.3 At least the above required reactor coolant loops shall be determined OPERABLE within 4 hours prior to initiation of the hot rod drop time measure-ments and at least once per 4 hours during the hot rod drop time measurements by verifying correct breaker alignments and indicated power availability and by verifying secondary side narrow range water level to be greater than or (3 equal to 41%.
- 3/4 10-4 i
l I i
'. SPECIAL TEST EXCEPTIONS 3/4.10.5 POSITION INDICATION SYSTEM - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.10.5 The limitations of Specification 3.1.3.3 may be suspended during the performance of individual full length (shutdown and control) rod drop time measurements provided;
- a. Only one shutdown or control bank is withdrawn from the fully inserted position at a time, and
- b. The rod position indicator is OPERABLE during the withdrawal of the rods.*
APPLICABILITY: MODES 3, 4, and 5 during performance of rod drop time measurements. ACTION: With the position indication system inoperable or with more than one bank of rods withdrawn, immediately open the reactor trip breakers. I SURVEILLANCE RE0VIREMENTS 4.10.5 The above required rod position indication systems shall be determined to be OPERABLE within 24 hours prior to the start of and at least once per 24 hours thereafter during rod drop time measurements by verifying the demand position indication system and the rod position indication systems agree:
- a. Within 12 steps when the rods are stationary,-e,wk
-tn uitg4- 3a teps dur 47; red :: tion E
*This requirement is not applicable during the initial calibration of the rod position indication system provided (1) K,ff is maintained less than or equal to 0.95, and (2) only one shutdown or control rod bank is withdrawn from the fully inserted position at one time.
i ( W-STS 3/4 10-5 NOV 2 0 B00
JUSTIFICATION Section 3/4.11 In the text of Section 3/4.11 where a capital letter with a circle around it appears, please refer to this sheet for the appropriate justification. A. Concentration limits on discharge are based on 10CFR20 values which relate to approximately two orders of magnitude greater in potential dose commitments than the dose objective of other Tech Spec sections which implement 10CFR50 Appendix I. Therefore, UNRESTRICTED AREA as used in these Tech Specs should be applied only to dose calculations (not , l concentration limits) associated with those Tech Specs which implement Appendix 1. and as such can take credit for dilution in the ocean which can be associated with the initial mixing zone. For concentration limits of 10CFR20, this additional dilution in the ocean is not appropriate and, therefore, the term UNRESTRICTED AREA should not be used in 3/4.11.1. The release point from the discharge structure should be stated to indicate where concentration calculations will be made. The NRC's standard model RETS use of UNRESTRICTED AREA could be interpreted to mean that both concentration limits of 10CFR20 and dose objectives of 10CFR50, Appendix I be determined at the same point of discharge, which would be unnecessarily restrictive to our ODCM dose calculations. B. This information deleted and will be contained in a licensee maintained and controlled document. C. Seabrook Station plant specific information. l E. Dose assessments will be determined utilizing actual release data and parameters not anticipated or projected conditions. F. Plant design does not include any tanks unprotected that will normally contain radioactive material. Temporary tanks, may however, be used on occasion for special evolutions which may have radioactive material in the process. t H. The Waste Gas System is designed to withstand a hydrogen explosion, therefore, it is only required the oxygen be monitored and be controlled to less than 4%. I. Testing each container should not be a requirement when alternate process l meets burial ground and shipping requirements. l i J. The Process Control Program shall define surveillance requirements for verification of plant specific solidification process which will be initially approved by the NRC. l i
K. Specification 3.11.4 implies that each individual licensee is directly reeponsible and accountable for releases from portions of the uranium fuel cycle beyond his own facility. It is not necessary for each individual power plant licensee to be burdened, either directly or by implication, with the possibility of assessment of dose contributions from uranium fuel cycle facilities far removed from his own, and which he has neither any direct knowledge of or any operational control over. This conclusion is based on EPA's stated intent concerning the implementation of the 40 CFR Part 190 standards. The EPA has determined that in the vast majority of situations, the sum of all reasonably postulable contributions from sources beyond 10 miles of a particular site will be small compared to these standards and should be ignored in assessing compliance. The EPA has stated that, "it would not be reasonable to attempt to incorporate into compliance assessment doses which are small fractions of the uncertainties associated with the determination of doses from the primary source of exposure" (Referenced d). The EPA has also determined as part of the promulgation of 40 CFR Part 190 standards that, except under highly improbable circumstances, conformance to the criteria of Appendix 1 to 10 CFR Part 50 should provide reasonable assurance of compliance with the standards for
.up to five reactor units at one site.
Since Seabrook has only two reactors at the site, and there are no other uranium fuel cycle facilities in the vicinity of the plant, it is concluded, based on EPA determinations, that the only dose commitments which need to be considered in Seabrook's case are those resulting from on site (station) sources.
3/4.11 RADI0 ACTIVE EFFLUENTS 3/4.11.1 LIOUID EFFLUENTS CONCENTRATION LIMITING CONDITION FOR OPERATION 3.11.1.1 The concentration of radioactive material released in liquid effluents & tc UNRESTRICT:0 ARCAS (see Figure 5.1-1) shall be limited to the concentrations specified in 10 CFR Part 20, Appendix B, Table II, Column 2 for radionuclides other than dissolved or entrained noble gases. For dissolved or entrained noble gases, the concentration shall be limited to 2 x 10 4 microcuries/ml total activity. g APPLICABILITY: At all times. { t h point of disc 6Se inna M mMport Je#*(se ACTION:
% With the concentration of radioactive material released in liquid effluents W
-tc,UNRESTRICTE: AREA exceeding the above limits, immediatelyarestore the concentration to within the above limits. T g uy,n.43 1
%. The prc.icient c' Scci'icct* A C M .b crc c+ =pp'4:2'c.
SURVEILLANCE REOUIREMENTS Radioact.i,veliquidwastesshallbesampledandanalyzedacco'5Iin
~
to the sampling and ana19 sis.. program of I 2'; A .-
~~...,_ _11_ ., li -
4.11.1.1.2 TheresultsoftheradioactivityEnalyses.shallbeusedin accordance with the methodology and parameters in the ODCM~to assure that the concentrations at the point of release are maintained within the limits ,of Specification 3.11.1.1.
N
~
B N y-PWR-STS-RETS 3/4 11-1 1/4/83 1
TABLE 4.11-1 RADI0 ACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM
\
'\ Lower Limit Liquid Release k Sampling Minimum Analysis Type of Activity ofDetectjon (LLD)
Type
\ Frequency Frequency Analysis (pCi/ml)
A. Batch Waste \ P P Releage Each Batch Each Batch Principa} Gamma 5x10'7 Ten's (t 'euiJ Raawaste Q Emitters Tut Tulu ) I-131 1x10
-6 P '\ M Dissolved and lx10 -5 One Batch /M Entrained Gases
\ (Gamma Emitters)
P \ M H-3 1x10
-5 d e Each Batch Cor5posite f
\/ Gross Alpha 1x10 P
h r-89, SrSO Q 'ds 5x10 Each Batch Composite
/ \ Fe-55 lx10
-6 s
~7 B. Continuogs W /
W P'rincipa{ Gamma 5x10 Releases " - - + 4 -"- " g/ C :p;;it:9 Emitters Ge*b Saapk (LGe hp EffhedGlag ) -6
/ I-131\ 1x10
-5 M /
Grab Sample M Dissolvedand 1x10 Entrained Gases @ / (Gamma Emitters)
-5
/M Continuou f M
C mpetitj _ H-3 1x10 Grc4 Sampk Gross Alpha 1x10
-7 O Q p Sr-89, Sr-90 Sh10-8 Genti nuous-f C:: petite .-
7 Grab sample, Fe-55 lx10.6
/
/ 's
>\
/ \
PWR-STS-RETS 3/4 11-2 1/4/83
\
.n TABLE 4.11-1 (Continued)
RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM
's
\ Lower Limit Minimum of Detection Liquid Release Sampling Analysis Type of Activity Type. (LLD)
Frequency Frequency Ana lys i s , ( Ci/ml)a C. Steam Generator Blowdown
\bN Grab sample Nhf
-Grch carple Principal Cacma Emittersf 5x10-7g Flash Tank I-131 1x10-6 MM 3(hf ,
Dissolved and 1x10-5 Grab Sample M '
/ Entrained Cases N (Camma Emitters)
WM \Mhi H-3 1x10-5 Crab Sample M a#
\ Gross Alpha 1x10-7 s
\
WY Qhf Sr-89, Sr-90 5x10-8 Grab Sample, Crch 0;plc g
/ Fe-55 1x10-6
,7 .
./'
/ ,
/
/
/ 's.
/ N N
\s
/
/ '\ \
/ \
/
/
j
/
/
~
3/4 11 - 5
TABLE 4.11-1 (Continued) ( TABLE NOTATION a The LLD is defined, for purposes of these specifications, as the smallest concentration of radioactive material in a sample that will yield a net count, above system background, that will be detected with 95% probability with only 5% probability of falsely concluding that a blank observation represents a "real" signal. N For a,particular measurement system, which may include radiochemical separation: LLD = E V 2.22 x 106 Y exp (-Aat) Where: . . LLO is the 9a priori" lower limit of detection as defined above, as microcuries,per unit mass or volume,
\
h is the standard deviation of the background counting rate or of s tne counting rat',of e a blank sample as appropriate, as counts per minute, y E is the counting eff.iciency, as counts per disintegration, V is the sample size \ insunits of mass or volume, 2.22 x 108 isthenumber\ of disintegrations per minute per microcurie,
/ \
Y is the fractional radiochemical yield, when applicable,
/
A is the radioactive decay co\nstant for the particular radionuclide, and / at for/ plant effluents is the elapsed time between the midpoint of sample collection and the time of counting.
/ \
Typ'ical values of E, V, Y, and at shou d, e used in the calculation.
/ S ee Ins e r+ T. D
-It :hteld be* recognized that th; LLD is de.4.ed :: :n : prinei (before the ' fact) "-it representhg- the capability of a ses:9erent cySte :nd--
net' :: an a pesterin"4 (af ter the fact) limi4 - for c'partienlar mane""emant__ bA l batch release is the discharge of liquid wastes of a discrete volume. jPrior to sampling for analyses, each batch shall be isolated, and [. / then thoroughly representative sampling. mixed by a method described in the ODCM ' s
/ q l / 1 / \.
PWR-STS-RETS 3/4 11-1 1/4/83
d i INSERT I
'\
The 'value of Sb used in the calculation of the LLD for a detection system
,' shall be, based on the actual observed variance of the background counting rate of.the blank samples (as appropriate) rather than on an unverified theoretically predicted variance. Analysis shall be performed in such a i
manner that the stated LLD's will be acheived under routine conditions. Occassionally background fluctuations, unavoidably small sample sizes, prolonged decay times, or other uncontrollable circumstan'ces may render the LLD's unavailable. \It should be recognized that the LLD is defined as an a , priori (before the fa'ct) limit representing the capability of a measurement j system and not as an a 'posteriori (af ter the fact) limit for a particular I measurement. This does not preclude the calculation of an after the fact LLD for the sample in question'and appropriate decay correction parameters. i 4 1
/
/
\ .
Ns / d
/\
/ N,
./
4 [ !
< )s\g 4
/ N
/ \
- /
/ \
\
\ .
\
/ \
/ x
\
\
N N,
\
\.
^
\
h, I i
?
/
/
/
4
,,vy --w--n- em- -e,,:.* e -- ----r = ,, , - - , -,-,--.+--.cw<r---=-re---*-m ,-r-9 e- - - - - - - - - - -
TABLE 4.11-1 (Continued) N\ TABLE NOTATION C
'gThe principal gamma emitters for which the LLD specification applies ex1clusively are the following radionuclides: Mn-54, Fe-59, Co-58, Co-60, Zn 65, Mo-99, Cs-134, Cs-137, Ce-141, and Ce-144 This list does not mean N
that only these nuclides are to be considered. Other gamma peaks that are idbntifiable,
\
together with those of the above nuclides, shall also be analyzed and reported in the Semiannual Radioactive Effluent Release Report pursuant to Specification 6.9.1. . See _Tosert IE d A composite sa'mple is one in which the quantity of liquid sampled is proportional to 'the quantity of liquid waste discharged and in which the h method of sampling employed results in a specimen that is representative of the liquids relea' sed.
/
'A continuous release is\the discharge of liquid f wastes of a nondiscrete volume, e.g. , from a volume of a system that has an input flow during the continuous release. /
/
# Tc be representative c' the ;"entitics and concentration: c' radicactiva m.cterials in liquid eff+uents, iamp'lc: chcIl--bc collected contintme!y 4"
.pcopertier to the rate of ficw of(the affluent :tr:2r 'icr tc analyses, c11 sempicWken--for-the-composite, sha!' be thercughly mixed-in order-
'cr the ccmpositt cample te be' rep ~ ^a+'+4"a ^ ' + * ^ ^ " ' " ^ " - ^ ' ^ " ^
/ \
cum genceabo o- htowdown Ys onl ys heet % samling and anal y sis as
. Micahd duct 3 periods 'wken all or pa,t\ of A blowdown efflued is b ein3 dischgr9ed 4o y eny,*,,nment in3)ead af tAe normal P"8ce55 of re cycllg er use./in ne sfaffon. 'N
\\
\
\
k
\
\
\
\
\ s
\,
\
/
/ .
/
/
s* / PWR-STS-RETS 3/4 11-4 1/4/83
i-INSERT II
\ %.
Nuclides which are below the LLD for the analysis shall be reported as "Not Detecte'd" g in the Nuclides LLD, and shall not be reported as being present at the LLD level for that nuclide. The "less than" values.shall not be used in the' require'd. dose calculation. When unusual circumstances result in LLD's ~ , higher than required, the reasons shall be documented in th'e Semiannual Radioactive Effluent Release Report.. /
/
. /
/
/
/
/
\ '
j'
\ /
/
,/
/
j\ 4 .
/ \
/ \s
~/ \
/ \
.- a x
6
\
y. e
i RADI0 ACTIVE EFFLUENTS DOSE LIMITING CONDITION FOR OPERATION 3.11.1.2 The dose or dose commitment to a MEMBER OF THE PUBLIC from radioactive materials in liquid effluents released, from each reactor unit, to UNRESTRICTED AREAS (see Figure 5.1-3) shall be limited:
- a. During any calendar quarter to less than or equal to 1.5 mrems to the total body and to less than or equal to 5 mrems to any organ, and
- b. During any calendar year to less than or equal to 3 mrems to the total body and to less than or equal to 10 mrams to any organ.
AFPLICABILITY: At all times. ACTION:
- a. With the calculated dose from the release of radioactive materials in liquid effluents exceeding any of the above limits, in lieu cf
'icenscc Event Repc"t, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report that identifies the cause(s) for exceeding the limit (s) and defines the corrective actions that have been taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits. (This Special
-Repont_sha!' :!:o inc!ede (1) tha result: cf redinlegi m1 analycac--
-ef-the-dninidagaster-- s c ue-a and (2) the--cadiologica! 4mpec+ cr NeW/ t'"IC I
--f4nkhed-dr-ieking-water-supplies with eegarri to the rannirements b $cdrealt --cf 10 CFR Pcrt 141. )*
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE RE0VIREMENTS s 4.11.1.2%m lative dose contributions from liquid eJJ.luents for the current calendar quarter nQhecurrentcalendaryearshallbedeterminedin accordance with the methodology and parame.ters'in the 00CM at least once per 31 days.
*Thi+-sentenca is-app 4-ieable-oMy i f dr4nk4ng-water-stipp ,y i:,-token Trum Uie
+eceiving4ater-body-ethit 3 -mi l e s - o f--4he-- pl ant discharge . I? thegase &
-rw 3 miie3-downstream-onbj,--
f}er4it-ed-phnts-this i5 6 PWR-STS-RETS 3/4 11 'E 1/4/83
RADI0 ACTIVE EFFLUENTS LIQUID RA0 WASTE TREATMENT SYSTEM LIMITING CONDITION FOR OPERATION (J eter sned) 3.11.1.3fTheliqudradwastetreatmentsystemshallbeOPERABLEandappropriate portions 40f the system shall be used to reduce releases of radioactivity when the-pr0jc;ted doses due to the liquid effluent, from each reactor unit, to UNRESTRICTED AREAS (see Figure 5.1-3) would' exceed 0.06 mrem to the total body or 0.2 mrem to any organ in a 31 day period. APPLICABILITY: At all times. ACTION: phich could reJue the ra$oacMise hguid wask bc6 9ed
- a. With radioactive liquid waste being discharged without treatment and in excess of the above limits and any portion of the liquid radwaste treatment system *not in operation, h lieu of : Lican;;; E;cnt 9epc + ,
prepare and submit to the Commission within 30 days pursuant to Specification 6.9.2 a Special Report that includes the following information:
- 1. Explanation of why liquid radwaste was being discharged without treatment, identification of any inoperable equipment or subsystems, and the reason for the inoperability,
- 2. Action (s) taken to restore the inoperable equipment to OPERABLE status, and
- 3. Summary description of action (s) taken to prevent a recurrence.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS [dc4c= med) T
/g , . _ -
4.11.1.3.1 Doses due-t liquid rek sn from each reactor unit-to-UNRESTRICTED AREASshallbe-project:h.at~1g earl. ;p3Ldays-in'aEo7 dance with the methodology and parameters in th Q Dts. , 4.11.1.3.2 Tie l ins't id Radwaste t'em , hall be demon-strated OPERABLE by meeting Specifications 3.11.1.1 and 3.11.1.2.
'7 PWR-STS-RETS 3/4 11-5 1/4/83
RADIOACTIVE EFFLUENTS LIOUID HOLDUP TANKS
- LIMITING CONDITION FOR OPERATION 3.11.1.4 The quantity of radioactive material contained in cac.. of the felicwing ay +emPot"'1 unprotected outdoor tankK shall be limited to less than or equal to /0 curies, excluding tritium and dissolved or entrained noble gases.
- c. A V0utside teliiptmaq tank APPLICABILITY: At all times.
ACTION: ternpor wyu 4"8 M ed " O *"
- a. With the quantity of radioactive material in any of th; ab;v: 'icted tankt exceeding the above limit, immediately suspend all addit. ions of radioactive material to the tank, within 48 hours reduce the tank contents to within the limit, and describe the events leading to this condition in the next Semiannual Radioactive Effluent Release Report.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REOUIREMENTS
/
m h poca, ' "y unemfe # b 4.11.1.4 Th'e7uantity of radioactive material contained in -- ' - " " - listed tank % shall be' determined to be within the above limit by-andlyzing a representative sample of the tankts fontents at least once,per'7 days when radioactive materials are being added'tosthe tank. ' N,/ ../'.. - p.- N
- Tanks included in this-spscification are those outdoor tanks that'are not
~
N surrounded by 1iners', dikes, or walls capable of holding the tank contents and that do-not have tank overflows and surrounding area drains connected to the ,liduid radwaste treatment system. \
/-
s PWR-STS-RETS 3/4 11 '% 1/4/83
RADI0 ACTIVE EFFLUENTS 3/4.11.2 GASEOUS EFFLUENTS DOSE Rt.TE LIMITING CONDITION FOR OPERATION
~
3.11.2.1 The dose rate due to radioactive materials re1 eased in gaseous ef fluents from the site to areas at and beyond the SITE B0UNDARY (see Figure 5.1-5) shall be limited to the following: I
- a. For noble gases: Less than or equal to 500 mrems/yr to the total body and less than or equal to 3000 mrems/yr to the skin, and
- b. For iodine-131, for iodine-133, for tritium, and for all radionuclides in particulate form with half lives greater than 8 days: Less than or equal to 1500 mrems/yr to any organ.
APPLICABILITY: At all times. ACTION:
- a. With the dose rate (s) exceeding the above limits, imme rs the release rate to within the above limit (s).
- b. he pervisier.: cf Speci#icat4n R9'ob s-a act applic210.-
SURVEILLANCE REOUIREMENTS s @ - 4.1h2 1.1 The dose rate due to noble gases in gaseous effluents shal1 e determined'to be within the above limits in accordance with the.meth'odology and parameters insthe 00CM. N /, N 4.11.2.1.2 The dose rate-due to iodine-131, iodine-133, tritium, and all radionuclides in particulate' form. with half li,ves' greater than 8 days in gaseous effluents shall be determined s ~ to beMithin the above limits in accordance with the methodology andjaraliieters in the ODCM by obtaining representative samples and performing analysis'in,'accordance with the sampling and analysis program specified'in-Tel: M1 2. -- N N s N s
.- s .
./ l
./
/ l 1
.C
~
9 PWR-STS-RETS 3/4 11-1 1/4/83
n TABLE 4.11-2
@ RADI0 ACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM /
J. m s
- s. /
/
T Minimum Lower Limit of d
'\
Gaseous Release Type Sampling Frequency Analysis Frequency Type of Activity Analysis /
/ etection D
(pCi/ml) (LLD)" Condusee N P P b' -4 C %. h te Cas Stprege Ead Tank Eeeh-Tank- Principal Gamma Emitters 1x10
.Tenk Evaceho A Grab Sample H-3 / Ixfo-'
' P P c c / b -4 h 1. Containment PURGE 'Each PURGE Each PURGE Principal Gamma, Emitters 1x10 or VE"T Grabyample -6
\ M 11-3 (oxide) 1x10 8 M Plant Vent MC 'd \ 1x10
~4 s Princiti$1GammaEmittersb Grab Sample N H ,'
s% -6 ll-3 1x10 R C.2 ruel Storagc Arca @ N Princ! pal C rT.; Emitterb he10-Vcatilatics 4eab S=ple -M- N, 4 y / N H 4*10-4 NincipalCommaEsitter? M" C.3 nuxiliary Oldg,- eati ast Arca,
-tt--'
-Grab Sa p l e p/ +t- N ,' g-) )
-SGB-Vent i-Ot4 tees / '
Ali Relca;c Types ContinuoIis 8 WM
" -12
-&. I-131 1x10 as '!:ted in A, 0, Charcoal s*
-G-above . Sample
' Continuous O WM
-11 Principal Gamma Eniittersf 1x10 Particulate Sample
# 5e -11
/ Continuous M Gross Alpha -1x10
/ Composite Par-ticulate Sample
- ContinuousPe -11~ Q Sr-89, Sr-90 1x10
/ Composite Par-R g ticulate Sample
-6 Continuous
- Noble Gas Noble Gases 1x10 Monitor Gross Beta or Gamma
s
~
,l o
TABLE 4.11-2 (courmgan) RADI0 ACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM
, ,'/
N. T s Minimum ,, Lower Limit of A Sampling Analysis Type of ,f Detection (LLD)* d Gaseous Release Type Frequency Frequency Activity Analysis / (pCi/ml)
/ .
/
,/
~. '
/
/
\,. .,
. 'N., ,/
't- -
b
' ~ /,/ s. . O Es : / . x N.
/
/
E WI I-131 3 D. Turbine G. qq Continuous 1x10'I2 S team Packsy . Charcoal . Eshus t-er Sample N [ Continuous WA Particulate Principal Gamma Emitters b ' 1x10
~II
/ Sample s4 Continuous *U' KQ Gross Alpha 1x10'II Composite Par-ticulate Sample q Continuous Q Sr-89, Sr-90 1x10'II R Composite Par-
$ ticulate Sample
TABLE 4.11-2 (Continued) TABLE NOTATION a The LLD is defined, for purposes of these specifications, as the smallest oncentration of radioactive material in a sample that will yield a net count, above system background, that will be detected with 95% probability with only 5% probability of falsely concluding that a blank observation represents a "real" signal. N For a p, articular measurement system, which may include radi6 chemical separation: 4.66 s b E V 2.22 x 106 Y exp (-Aat) Where: LLD is the "a priori" lower limit of detection as defined above, as microcuries 'per unit mass or volume,
\
s h is the standard\deviation"of the background counting rate or of tne counting rate of a blank sample as appropriate, as counts per minute, E is the counting efficiency as counts per disintegration, V is the sample size (in u\of mass or volume, nits
\
2.22 x 108 is thefnumber of disintegrations per minute per microcurie,
\
Yisthefractidnalradiochemicalyield,whenapplicable,
/ \
A is the radioactive decay constant for the particular radionuclide, and / \\ at for pl nt effluents is the elapsed time between the midpoint of sample, collection and the time of counting. l '\ Typical values of E, V, Y, and at should b'e used in the calculation.
/
ItshouldberecognizedthattheLLDi_sdefineda\ s an a priori (before the fact) limit representing the capability of a' meas.urement system and not /as an a posteriori (after the fact) limit for a pa'rticular measurement.
\
/
/ \'
/
/
/
12 PWR-STS-RETS 3/4 11-M 1/4/83
~
~
TABLE 4.11-2 (Continued) TABLE NOTATION The principal gamma emitters fo' which the LLD specification applies include the following radionuclides: Kr-87, Kr-88,/Xe-133, Xe-133m, Xe-135, and Xe-138 in noble gas releases and Mn-54,' Fe-59, Co-58, Co-60, 2nh 65, Mo-99, I-131, Cs-134, Cs-137, Ce-141 and Ce-144 in iodine and part'iculate releases. This list does not mean that only these nuclides aret(beconsidered. Other gamma peaks that,a're identifiable, together with those of the above nuclides, shall also.be analyzed and reported in the Semiannual Radioactive Effluent Release _ Report pursuant to Speci fication 6. 9.1.'M. See Insert E /
\ 9 e c
Sampling and analysis shall also be performed following shutdown, startup, or a THERMAL POWER change exceeding 15 ' percent of RATED THERMAL POWER within a one hour N period. Sec ensert/III-e d \
/
Tritium grab samples shall be taken at least once per 24 hours when the refueling canal is flooded. / N. MA Dvent itium grab ;cmple: th!' bet $en at least once per 7 day #re: the-44ation--e*Asust frch the spent fuel paci erea, whenever spent fuel is in- the spent fucl pool '\ e .e n The ratio of the sample flow ra.te to the sampled stream flow rate shall be known for the time per'iod covered by each dose or dose rate calculation made in accordance with' Specifications 3.11.2.1, 3.11.2.2 and 3.11.2.3.
/
#4Samplesshallbechandedatleastsonceper7daysandanalysesshallbe
\
completed within 48, hours after changing, or after removal from sampler. Sampling shall also' be performed at \ least once per 24 hours for at least 7 days following e'ach shutdown, startup or. THERMAL POWER change exceeding 15 percent of RATED THERMAL POWER in one hour and analyses shall be completed within'48 hours of changing. Yhen samples collected for 24 hours are analyzed,,t'he corresponding LLDs may be increased by a factor of 10. This requirement does not apply if (1) analysis shows that the DOSE
. EQUIVALENT I/131 concentration in the primaryscoolant has not increased more than a' factor of 3; and (2) the noble gas smonitor shows that effluent activity h5s not increased more than a factor o'f .
j x ' 3 He g scouS wa.sfe samplin7 and anch /5 /5 pggfay aes nd explicihy vege Sa>np% and analyss ai a sp ect f, ed G. L D -6 C se%& % z-I3 3 release. Eskaks ef I-133 releaS es 8
- s a.h be defcvmtned kl <* S ^ f Ac ** M O "*'8
.)nss vs -u , r-u,) aa a~e a "+ utenefc s
'Ihc VC.ked5c. P& log
- s PWR-STS-RETS 3/411-k 1/4/83
t INSERT III
-\unless; .1) . analysis shows that the DOSE EQUIVALEtfr 1-131 concentrations in the primary coolant has not increased more than a factor of 3,2) the noble gas
~
acti'vity monitor for the plant vent has not increased by more than a factor of
- 3. Fo'r containment purge, requirements apply only when purge is in operation.
/
l f 4 Y
- /
I s t r
/
/
A l
/9
/ \
/
L
/
/
/ ,
j \
/ \ '
/
l ..
./
k N [': t-L i
- y. 3 .- - , . . . . _ , , , - . . - - . . . . . - , _ . . - , . _ . , . , , - . _ . . . . . , _ - . _ . . ~ . _ , ,_.m,,, .-, . - _ , - . . . -.,_m-.. .
RADI0 ACTIVE EFFLUENTS
^
DOSE - NOBLE GASES LIMITING CONDITION FOR OPERATION 3.11.2.2 The air dose due to noble gases released in gaseous effluents, from each reactor unit, to areas at and beyond the SITE BOUNDARY (see Figure 5.1-1) shall be limited to the following: l'
- a. During any calendar quarter: Less than or equal to 5 mrads for gamma radiation and less than or equal to 10 mrads for beta radiation and,
- b. During any calendar year: Less than or equal to 10 mrads for gamma radiation and less than or equal to 20 mrads for bsta radiation.
APPLICABILITY: At all times. ACTION
- a. With the calculated air dose from radioactive noble gases in gaseous
. effluents exceeding any of the above limits, i 'i = cf : Licensee Event Sprt, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report that identifies the cause(s) for exceeding the limit (s) and defines the corrective actions that have been taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases l will be in compliance with the above limits.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS
%4.11.2 P Cumulative dose contributions for Nhe current calendar quarter and /
current calendai' year-forsnoble gases shall be determined in accordance with , the methodology and parametTers-in the ODCM at least once per-31' days.
,/'
'/N.C'N ,
N i 14 r PWR-STS-RETS 3/4 11-}A 1/4/83
m
)
I i l RADIOACTIVE EFFLUENTS l DOSE - 10 DINE-131, 10 DINE-133. TRITIUM, AND RADIONUCLIDES IN PARTICULATE FORM 1 LIMITING CONDITION FOR OPERATION 3.11.2.3 The dose to a MEMBER OF THE PUBLIC from iodine-131, iodine-133, tritium, and all radionuclides in particulate form with half-lives greater than . 8 days in gaseous effluents released, from each reactor unit, to areas at and beyond the SITE BOUNDARY (see Figure 5.1-1) shall be limited to the following: I
- a. During any calendar quarter: Less than or equal to 7.5 mrems to any organ and,
- b. During any calendar year: Less than or equal to 15 mrems to any organ. ,
APPLICABILITY: At all times. l ACTION:
- a. With the calculated dose from the release of iodine-131, iodine-133, tritium, and radionuclides in particulate form with half lives greater than 4"
8 days, in gaseous effluents exceeding any of the above limits,
'feu Of 2 Lic2ncee E'/ea.t Rep;rt, prepare and submit to the Com-mission within 30 days, pursuant to Specification 6.9.2, a Special Report that identifies the cause(s) for exceeding the limit and. defines the' corrective actions that have been taken to reduce the releases and the proposed corrective actions to be taken to assure that. subsequent releases will be in compliance with the above limits. ,
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REOUIREMENTS x4.11.2.3 Cumulative dose contributions@for the current-calendar 7 quarter and current calendar-year for iodine-131, iodine-133; tritium, and radionuclides in particulate form with half-1.ives greaterethan 8 days shall be determined in accordance with the methodology anTpaFimeters in the ODCM at least once per 31 days. /, %~
/ N N
/
15-PWR-STS-RETS 3/4 11-11 1/4/83
c 1 RADI0 ACTIVE EFFLUENTS GASEOUS RADWASTE TREATMENT LIMITING CONDITION FOR OPERATION
@mulatED (N previous)
I 3.11.2.4 The VENTILATION EXHAUST TREATMENT SYSTEM and the WASTE GAS HOLDUP j SYSTEM shall be OPERABLE and appropriate portions of thes systems shall be j used to reduce releases of radioactivity when the projectec-doses in6S1 days - due to gaseous effluent releases, from each reactor unit, to areas at and beyond the SITE BOUNDARY (see Figure 5.1-1) would exceed either: I
- a. 0.2 mrad to air from gamma radiation, or
- b. 0.4 mrad to air from beta radiation, or
- c. 0.3 mrem to any organ of a MEMBER OF THE PUBLIC.
APPLICABILITY: At all times. ACTION:
- a. With radioactive gaseous waste being discharged without treatment and in excess of the above limits, " 'ieu of e Licer.ce: E scr;t ocpe-t, prepare and submit to the Commission within 30 days; pursuant to Specification 6.9.2, a Special Report that includes the following information:
- 1. Identification of any inoperable equipment or subsystems, anc the reason for the inoperability,
- 2. Action (s) taken to restore the inoperable equipment to OPERABLE status, and
- 3. Summary description of action (s) taken to prevent a recurrence.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REOUIREMENTS ca a@
~x4.11.2.4.1' Doses,due to gaseous releasespf @ /
rom each reactor unit to areas'at and beyond the SITE B00NDARYsshal.1 be tre,ccted-at least once per. 31' days in accordance with the methodology and parameters in the ODCM.
~ N~ .
4.11.2.4.2 The installed Gaseous Radwaste Treatment Systim'shall be demon-i strated OPERABLE by meeting Spec,ification 3.11.2.1 and 3.11.2.2 or 3.11.~2.3. j ,. - ~ aM s Ib PWR-STS-RETS 3/4 11-M 1/4/83
RADIDACTIVE EFFLUENTS f EXPLOSIVE GAS MIXTURE LIMITING CONDITION FOR OPERATION 3.11.2.5 The concentration of oxygen in-the WASTE GAS HOLDUP SYSTEM shall be limited to less than or equal to 2% by volume.uh;n;ver the hydroger cer.:entr: tion ex;;ad; t% by .;l; c. APPLICABILITY: At all times. ACTION: Mr. 'Jith the cunwentratien cf oxyger ia the us.37g gag ggtggp gy;7 3_
-greater *har 2% by v;lume but 'er: th r er equ:1 to 4% by veluce,-
-reduce the ;xyger cer:ent-etier te the :b v; lirit; with4-40 hen. 5.
c, tt. With the concentration of oxygen in the WASTE GAS HOLDUP SYSTEM greater than 4% by volume-cr.d the hydr: gen : r.:entr: tier grect:r _th r 4% by lume, immediately suspend all additions of waste gases to the system and reduce the concentration of oxygen to less than or equal to 4% by volume without delay,ther take the ^.CTION .n
- 2. :b 'f e.
b 'g. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. SURVEILLANCE REOUIREMENTS . r h ..ge "
. 4.11.2.5 Theconcentrationsofhydrogenandioxygen;intheWASTEGASHOLDUP SYSTEM shall be determined to be within-the'~above limiss by monitoring the wastegasesintheWASTE.GASHOLDUPSYSTEMwiththehydrogen:n?oxygensmonitors N required OPERABLE.by' Table'3.3-13 of Specification 3.3.% H. %
.3.to pC Wr.r-r' '~
l' l i ! i 1 I l 11 i PWR-STS-RETS 3/4 11-75. 1/4/83 l
10 ACTIVE EFFLUENTS got /)pf3hcable do GAS AGE TANKS $cMWdk
- LIMITING COND ' ION FOR OPERATION x ,
3.11.2.6 The quantit of radioactivity contained in each s storage tank shall be limiteu to less than or equal to curiesno))egases(considered as Xe-133). APPLICABILITY: At all times. ACTION:
- a. With the q'antity u of radioactiv material in any gas storage tank exceeding the above limit,j'dmedia ly suspend all additions of radioactive material to 6e tank and thin 48 hours reduce the tank contents to within the imit,
- b. The' provisions of pecifications 3.0.3 and 3. are not applicable.
SURVEILLANCE REOUIREME TS
/ \
4.11.2.6 The uantity of radioactive material contained in each gas stor e tank shall determined to be within the above limit at least once per 24 hours en radioactive materials are being added to the tank. O w l l l PWR-STS-RETS :,, ;; E 1/4/83 L
, RADIOACTIVE EFFLUENTS 3/4.11.3 SOLID RADI0 ACTIVE WASTE LIMITING CONDITION FOR OPERATION ,
3.11.3 Radioactive wastes shall be SOLIDIFIED or dewatered in accordance with the PROCESS CONTROL PROGRAM to meet shipping and transportation requirements during transit, and disposal site requirements when received at the disposal site. APPLICABILITY: At all times. ACTION:
- a. With SOLIDIFICATION or dewatering not meeting disposal site and shipping and transportation requirements, suspend shipment of the inadequately processed wastes and correct the PROCESS CONTROL PROGRAM, the procedures and/or the solid waste system as necessary to prevent recurrence. (Jae .ad. NNde r ocewed easia*
- bo=a ef M FhmCabl%g
- b. WithSOLIDIFICATIONordewatering not performed in accordance with the PROCESS CONTROL PROGRAM, (1) Jest the mpr:per!y prece::ed weste i
in each container to ensere thet it meet) burial ground and shipping
. requirements and (2) take appropriate administr'ative action to prevent recurrence.
- c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REOUIREMENTS
% ve.hw .+ souotac4Tio4/ .< rep eso.kHoc. test specuas + wet wa.ste ye o~el 4.1C3s {0LIDIFICFInw nr 3t 1 ::t en, r nr.c.ntativ. t.et Specimen %m ,3 t633gt every ter.th batch-of erh tuna a# wet radicact4": '::ter (- a ,f % %
spe^t -acinc ?avancester bct h , beric cid selution: nd ::d4enHaMet4 I vecified in a m d PROCESS CCNTRaC aacCn,^y. soht4ent)-: hall'b: by's%4 win tu Peocsssncecoarraot. Paoobm. failsta to acwaIaace with( the the
- as .e 4 ui,e a.'
! If any test pecimen verify SOLIDIFICATION SOLIDIFICAT ON f the batch under test shall be suspended untiVsuch time as
~
t1 . I test specimens can be obtained, alternative SOLID a ION parame s can be determined in accordance'with the ESS CONTROL PROGRAM, a a subsequ'ent test verifies' SOLI 0IFI ON. SOLIDIFI-CATION of the tch may then be resum'ed usi ne alternative SOLIDIFICATION pa eters determined by PROCESS CONTROL PROGRAM.
- b. If the initial test sp
/%
men fromT atch of waste fails to verify SOLIDIFICATION, the PROCE CON L PROGRAM shall provide for the l collection and testing'of re entative test specimens from each consecutive batch of the e ty of wet was'te until at least 3 con-secutive initial test ecimens dem trate SOLIDIFICATION. The PROCESS CONTROL PRO M shall be modif as. required, as provided in Specification .13, to assure SOLIDIFI ION of subsequent batches of waste. ,
/
With the i talled equipment incapable of meeting ~ ecification 3.11.3 1 l c. l M r deci ed inoperable, restore the equipment to OPEn LE status
/ or p vide for contract capability to process wastes as essary to s sfy all applicable transporation and disposal requiremen ,
l l 18 PWR-STS-RETS 3/4 1147 1/4/83 i
RADI0 ACTIVE EFFLUENTS 3/4.11.4 TOTAL DOSE LIMITING CONDITION FOR OPERATION 3.11.4 The annual (calendar year) dose or dose commitment to any MEMBER OF , THE PUBLIC due to releases of radioactivity and to radiation from traniur fe:LSixfwo
-cycle sources shall be limited to less than or equal to 25 mrems to the total body or any organ, except the thyroid, which shall be limited to less than or equal to 75 mrems.
APPLICABILITY: At all times. ACTION:
- a. With the calculated doses from the release of radioactive materials in liquid or gaseous effluents exceeding twice the limits of Specifica-tion 3.11.1.2.a, 3.11.1.2.b, 3.11.2.2.a, 3.11.2.2.b, 3.11.2.3.a, or 3.11.2.3.b, calculations shall be made including direct radiation contributions from the reactor units and from outside storage tanks to determine whether the above limits of Specification 3.11.4.thave been exceeded. If such is the case, " 'f eu of ; '.f cen 00 Event
% pert, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report that defines the corrective
, action to be taken to reduce subsequent releases to prevent recur-rence of exceeding the above limits-and incitdes the schedule for-achieving conformance with the above limits. This Special Report, as defined in 10 CFR Part 20.405c, shall include an analysis that m estimates the radiation exposure (dose) to a MEMBER OF THE PUBLIC S Fd-wg f rom drium fuel cyc'e- sources, including all effluent pathways and direct radiation, for the calendar year that includes the release (s) covered by this report. It shall also describe levels of radiation and concentrations of radioactive material involved, and the cause of the exposure levels or concentrations. If the estimated dose (s) exceeds the above limits, and if the release condition resulting in violation of 40 CFR Part 190 has not already been corrected, the Special Report shall include a request for a variance in accordance with the provisions of 40 CFR Part 190. Submittal of the report is considered a timely request, and a variance is granted until staff action on the request is complete.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
. SURVEILLANCE REOUIREMENTS
( x l 4.1174?14umulative dose contributio(rr@from liquid and, gaseous effluents shallf be determined'i cordancewithSpecificationsJ.11:1!2,4.11.2.2,and4.11.2.3, and in accordance ith ado a_nd-par ~ameters in the ODCM.
. 4.11.4.2 Cumulative dose, contributions f.r_om direct radiation from the reactor units and from radwaste storage tanks shall be'determg ed in accordance with the methodology Tnd parameters in the OCCM. This requ dhment ds able only-under conditions set forth in Specification 3.11.4.a.
PWR-STS-RETS 3/4 11- 1/4/83
JUSTIFICATION Section 3/4.12 In the text of Section 3/4.12 where a capital letter with a circle around it appears, please refer to the letter below for the appropriate justification. A. Action statements and surveillance requirements deleted and will be contained in licensee maintained and controlled documents. B. T.S. 3.12.2. Seabrook does not have an elevated release as defined in R.G.I.111. C. T.S. 3.12.3. Only the radioactive material referenced by the Interlaboratory Program should have to be analyzed.
3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.1 MONITORING PROGRAM LIMITING CONDITION FOR OPERATION 3.12.1 The radiological environmental monitoring program shall be conducted as specified in-T:b': 2.12 1. laW APPLICABILITY: At all times. /,
/
ACT, ION:
/} /
\ With the radiological environmental monitoring program.nb being a.\ conducted as specified in Table 3.12-1, in ' ice of L'iceme: E=n t -
"cp;rt, prepare and submit to the Commission, in the Annual Radio-logical Environmental Operating Report required by Specification 6.9.N1. M V a description of the reasons for not< conducting the program a recurrence.
as recuired Em nNwi of and the olans for h IderaAery preventing analy sQ $ b.[Witntnesevelofradioactivityastheresultofplanteffluentsin
, an environm' ental sampling medium at a,specified location exceeding the reportingslevels of Table 3.12-2,when averaged over any calendar quarter,Ja 'icu Of ;--Licensee Sent ".eport, prepare and submit to i the Commission Vithin 30 dayk pur'suant to Specification 6.9.2, a 5pecial Report that eidentifies,the cause(s) for exceeding.the. limit (s) and defines the corrective N
actions to be taken to reduce radioactive effluents so that the potential annual dose
- to A MEMBER OF THE PUBLIC is less than the calendar y' ear limits of Specifications 3.11.1.2, 3.11.2.2, and 3.11.2.3. ,1When more than one of the radionuclides in Table 3.12-2 are detected 'in the sampling medium, this report shall be submitted if: '
/
concentration (1) concentration (2)
. +
reporting level (1) repor, ting level (2) ***2 1.0
/
When radionuclides other than e inthos\ Table 3.12-2 are detected and are the result of plant effluents, this report shall be submitted if the potenti,a'l annual dose
- to A MEMBER OF,THE PUBLIC is equal to or greater than the calendar year limits of Sp'ecifications 3.11.1.2, 3.11. 2. 2/and 3.11. 2. 3. This report is not required if the measured level of radioactivity was not the result of'p.lant effluents; however, in such an event, the condition shall be reported and described in the Annual Radiological Environmental Operating Report.
/ \ -
- c. With milk or fresh leafy vegetable samples unavailable from one or
( N ( If ava,da.hk) m' ore ofobtaining tionsVfor the sample locations replacement required samples and by addTable 3.12-1, them to loca-the radidentify'iologic
~
environmental monitoring program within 30 days. The specific
*The' methodology and parameters used to estimate the potential annual dose to a, MEMBER OF THE PUBLIC shall be indicated in this report. \
PWR-STS-RETS 3/4 12-1 1/4/83 l I
JADIOLOGICAL ENVIRONMENTAL MONITORING
,/
ACTION: (Continued)
\
locationsfromwhichsampleswereunavaIlablemaythenbedeleted from the monitor.ing program. Ir !ieu of a Lic:nsee Oscn Rcpc,rt :nd-ation 6.9.1.M 9 identify the cause of the unavail-Pursuant to Specific ability of samples a _iidsidentify the new location (s) for ob replacement samples in Release Report and also,the next in include Semiannual the reportRadioactive Effluent a revised figure (s) and table for the 00CM r'eflecting the new location (s).
,/ N
- d. The provisions'of Specifications 3.0.3 'and 3.0. are not applicable.
/
SURVEILLANCE REQUIREMENTS
.e radiological
[ s 4.12.1 environmental monitoring samples shall be cdliected pursuant to Table 3.12-1 from the specific locations given in the table a'nd.s figure ('s) in the ODCM, and shall be analyzed pursuant to the requirements of \ Table' 3.12-1 and the detection capabilities required by Table 4.12-1. N
/
PWR'-STS-RETS 3/4 12-2 1/4/83
N. TABLE 3.12-1
/
/
o . h RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM \ / m x' -4
- m. N g y Number of ,
H N sReiresentative m Exposure Pathway ' Samples and Sampling and Type and Frequency and/or Sample Sample Locations a Collection Frequency of Analysis
- 1. DIRECT RADIATION 4^ rcutine ;;n'icr'n; tztien; Quarterly Gamma dose quarterly.
inni_nDAn) ~_:i-isus. n- - - ..: Tun. . . . , - i . t.s.. a...... .m._ a_,;__a__ v..._
.v i b s su .s ..wbL. s uu
__ . .: aL vu 3 bp U U IU M S M tons wth Y 580 inctvimant fne -^
.\ - ". . 8. .-- .u3, -,o*
'.'.u^.. ^.u^ ' ' . . ^.'.."y" .uuua'. ,' ,
^ ^ ^ ^ ~ ~ ^
OP PaLOve dosten t h CF S " ^ ' _ ' ^ " .'4 I. ". ',' _n l me n ti --ne N, pj ateg tg
- f a 1 1. ~.- . ~ .
Ce n t.ca.tN c. P M35 . 2r 'rner - g of stat,.ca;, ca: . - dou& % plast ,,,,.n. ....su.u
. . _ .. . _ ,. _ _ i. _ u..---
___a-_
- - -i - --- " -"a i.e-y M .m,
, .. .m o. m.u ._1, u,ou
_r u. at-onu r nna nun a nu_
- v. T. r uvununn. -
s 4 Inni_nnic_). g u .4 4 u. e g g o* BJ g e _n n_ nii tm c c i n r.i n. f.
. e_ t_ _s e. l. a. a.u..m
. . ,, , - - i. n
. m suu'n
..am t.,a. r_u.-n_ I. n_ n i.- r_ s i.
c a.. r e.n In - T Shahon$ tast'i6 YLOO ( g en_ .o.L. e. _m n. .. n - C...... *A..,_, ce m e v t d OS[sige k c. 5
,46m
- i .-
/s.DD17 11D'19 b
f laced . d cond rol . 1._ boC c.a,,ance nf tho <tdinne i enn n unn,a ujn scu. un {0Cedt045PoPtaltLT18" , tunas p 4 C C u, ' " . _ c .s ,. : .i :_.___,a -_____ _.._m. (( A CFh g $ M4h 9
.,,,,,,,1,,.,
--s -u s._ , 'm.. ,
su - v. %u... . sa - _ a, e y-
,. u p .
-J 1 N$lOtW%C f $ -^ews .um. In n ~e~.n e , r_ c_ h. s_s n_ i. ,, ,unu
. I.._.
w
^- s u n o .e f_g .
.f.
-.f% F - .N. . *~A_ _*1 ( f_ gg
_ r. n . m,a^
% ua
, * ,a
.v.w= : n. n. . .e ._
e g ( [
- .n F,,-aca -i-? suay v u s y' O O ,'h s.O.
- ^ ^
.ARE u...-
%O ) N.b O, $_$ ,I O. ~ y.
. _. kN .N.".o'
^
. u s u s. . u . .
^
O. _ aium N.$. ~ u , .O~ ,b. -.O.O... . -- M,____ b Eg,. ,s. b s.
. . .u. sHWE gA5 L, .g . _s . _ . -
ro_ u-a8bG us Tu s i l L 8 5 cdLil Usabs y ua a i :_,n u l, g i s . w - ., t. n.. s___t wuuua s.a
,s.,
w .u.-~ ,,.e.o.;,. e. . c e
~.
i
,r_a___:-_ rr n ,,L..-.
- LU W c 5 ct. s,y. . e s.. .ua: . asuu n.i.%m s .. . : .._: ,n_
N h. .., m m ,a _ e..o,.. o s u s m ' ' a v sy e s ar : s_ uuse _ _ s i. j. , e_ n_ n. t_ r.i n. . . e n_ e n_ g w n u u r s . .w u r - - - ' " '" J a s tu b r wy . - g -tmh yhh:c ' ' dese and shoislet ho i nt-l ine f o rt i- ti,u aanspl i;;g "r0gro.T.. Ilic C6dc Icl' C r', in-par 0a.th000C , O.g.- w _Do! A1, nrnuir4 nna way nf rMfi-ing uc'au IC 00;' P k 'GC6 tie'B i 'i li> > a ' I ' t' i I*C2 tine-' hat c2n SC e,c d L-o- -. - 41n .n.. t_ i_ T. ,s, e L.- e,,n a_ r i. f i. r_ i. n. e .,. *. i. a,
. ,, * . . a. t . w- ... . ,,gsj I 4 - n. ,.1 . 6_2_ h i n_
i- * *. _ n
.... uus.
ni . -
1 TABLE 3.12-1 (Continued) m 2 RADIOLOGICAL ENVIRONMENTAL HONITORING PROGRAM a N Number of - d N Representative 5 Exposure Pathway Samples and Sampling and Type and Frequency 0 and/or Sample x Sample Locations a Collection Frequency of Analysis
- 2. AIRBORNE \ j/
Radiciodine and ' Samples from 5 locations (Al-AS). Continuous sampler Radioiodine Cannister: Particulates N operation with sample I-131 analysis weekly. 3 samgiles -(Al ^.3) f rom close collection weekly, or / to the 3' SITE B0UNDARY locations, more frequently if f in different~ sectors, of the required by dust Particulate Sampler: highest calculated annual average loading. Gross beta radioactivity ground-level D/Q. N, ,/ analysis folloping
/ filter change;
, I sample 4A4-) from the yk'cinity ' Gamma isotopic analysis of a community having the highest _/ of composite (by R
calculated annual average ground x location) quarterly. o level D/Q. N x - T 1 sample 4AS-} from a control \- location, as for example 15-30 km distant.and i" the least preva- /
/ 'N N Qv
@ ? c ".t .;..d directicr. ./'
3. WATERBORNE
- a. Surface g mthe arm ot tac descAa,pe. Grab samp}e-mo<thly - d g I sample epstre r (Wal) f Campo;ite campje cecr NGamma isotopic analysis I sample f,= de-stream W22L} .
'f.leca.fto* Imenth period monthly. Composite for
- a. co.dro tritium analysis quarterly.
Samples from 1 or'$ sources d
- b. Ground Quarterly Gamma' isotopic and tritium 4Wbl , '.."a), on ly i f ' i he!" - to be analysis qua terly.
affected . ' U Peterbah @ w Orinkig -1 ;;;plc cf nach nf 1 te 3 (tfc1 - f~ p;; j tc 3 a,i,p j g j 131 ;a;}y;j; on c;ch Mc3) of i le ..earc:t mter ever ?-ucek pe,;ed- composite he.. il,a Jose t supplics that cau!d le. eher I-131 analysis- .ca!c"'2ted-foi um i. ei, s umps t G. affectcJ by s i.:, d ischacgc. is perfer:cd, ;;nthly Sion-of t he ef 2 t e r is-arete r
/ -eempc i te ethe r.e i se- ,,,,,,,,m,,, ,_ ,, _ h c.,_...
bl$[J Tc I (b e q. [O w-- g .,s O h r s b .a a[ s loc,R-icre ('A4 ); -yamma--iwtopic mi: 1 y',u
. mon %1y. Compositt sur 1: i t i::m '" :!y< i - quatw4y.
TABLE 3.12-1 (Continued) 8%. m N RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM a
-4 N
x m x
- N s Number of /
f g N.s Representative ,-
- Exposure Pathway' s, Samples and Sampling and Type and frequendy and/or Sample ' x Sample Locations a Collection Frequency of Analysis' N <
- d. Sediment -l' sample from downstream area Semiannually Gamma isotopic analysis d from with existing or potential semianntially.
shoreline recreational value. fwd 4-). /
~. ,'
- 4. INGESTION 'N '
N,.
- a. Hilk Samples from milking. animals Semimonthly when , Gamma isotopicd and 1-131 in 3 locations (!al '!a3} within animals are on ,/ analysis semimonthly when 5 km distance having the' highest pasture, monthly at animals are on pasture; dose potential. If there are s other times- monthly at other times.
m none, then, 1 sample from milking ' 2 animals in each of 3 areas -(441. - , y -hs)- between 5 to 8 km distant X, y where doses are calculated to be s u' ' greater than 1 mrem per yr.%e' -' - s 1 sample from milking animals ' at a control location 4444-), 15-30 km distant 2-d the-g
!c2ct prc'12!ent wind direct!om g
- b. Fish and I sample of -eEdd c ercially Sc=pic in seasc w -Gamma isotopic analysis d Inverte- and recreationally important Semiannually +4--they on. edible portions.
brates species in vicinity of plant =---*--"aa* discharge area. -(4bl - R_).
/ 1 sample,of same species in areas not influenced by plant dis-charge -(4MO - !b_);
g
%. ftmtt--
Ermhmg. h=M,e e r a?ch p i-ripa4-class nr rnm nrneyc e s r m. 3ny grea At--t-inc e f- harves k Gamma-i+otep i c--an-a4ysM
.3_mgi t ;,y,- t ;
w that is ;rrigated by eater i v
.M ic!' 1 g+4d i,loni -oate have b haan discharge:' (!cl - !c ). N
1 1 N
!, I 3
1 3 1 I
/ y c
d n a d n a
- n e d d
/ u qs i c
p i c
/F'd ei p rs o o y t t l o . o .
a a ss ss nn ii ii a aA s s
- ay ay ef ml ml po y
ma ma an an T Ga Ga
/
y c
/ /
n M e
/
A u % R dq ' .~ G ne gn gn O R ar f no , no s P G i g nn o i s ds ra ue / 'ds i s ra ue
,N
)
N I R l i pt mc yg l n
'yl ng N.
d e u O T I Sl ae l hi t w no hiN' t w no i t n N O M C o Mg or Mgorf n o L - / ' C A t ' ( T - o k 1 N E n dn n
'ra i l
- M wt us .
l t m - 2 N sont oi' i s 1 O R n g r e g g' d resr i mnaf wei . 3 I i eh n sol d . V k nf gei . r ne E N ofi gl ) egeom / L E tiihap3 h hi r B ntd ?ml t ntt o / A L ea f ea c o cf T A rt oovsI fi ner C I a eew' a f gt s k - oti re a i p f G s f e nl l httd O n i v,. oiuml f' oai0 cen t L e o d agado . O v i f t n c eet nn) t 3 ah anf cccaiP. vsi I i 3 D tda s oFa iws2 f A R ft ao anc f! o ao el dQd dt i : on L d e/e eemng es s : t etDm l ! k en-rsee ee st c r p li eel l l reiil o md 0 al 0 brpp pt m rsd ef a: 3 v p2 mpmm af evr se - em c ueaa af ef ree r5 raI e NRSS S o o pl p Ib 1 ps( O s s) L s
,' y ae td c'
u wl d 'a hp oe t m drc N' aa PS F oP4 o er ro u/ . sd c on pa N' E x s
\ ,hUTAd R* Ca >8
h, x E h tj, H
\ % T_ABLE 3.12-1 (Continued) l T ', TAllLE NOTATION
./,.
g \ ,/
~* a j
Specific param'eters of distance and direction sector from the centerline of one reactor, and' additional description where' pertinent, shall be provided for each and every sample location in Table'3.12-1 in a table and figure (s) in'the ODCH. Re fer to NU"EG 0133, "Pi e,,r: tier c' Radic!;gic;! E f f!=nt Tect-ical
@ ."Spc9 ficat inne far Muc!:;r Perc- p!2ntc," Octcher 1978, and te ";dic;cgic:? Sce n :cht Branch Technicci c;itica, " sisicr. 1, "cx :':cr 1"7".
Deviations are permitted from the required-sampling schedule if specimens are unobtainable due to circumstances such as hazardous conditioris, s'easonal unavailability, and malfunction of automatic sampling' equipment. If specimens are unobtainable due to sampling equipment malfunction, ef fort shall be made to coinplete corrective action prior to the end of the next sampling period. All deviations from the sampling schedule shall be documented'in the Annual Radiological Environ-mental Operating Report pursuant to Specification 6.9.1.h8. It is ,rscognized that, at times, it may not be possible or practicable to continue to obtain samgiles of the media'of choice at the most. desired location or time. In these instances suitable alternative media and locations may be chosen for the particular R
- pathway in question and appropriate substitutions made within'30 days in the radiological environmental ~
monitoring pro 0 ram. -I.. ii^" d_a_tiran<aa ruant nepe
- 7 d jursuant to Specification 6.9.1.h2T identify .b*. ' OC M ' the cause of the unavailability of samples for that pathway and identify the new location (s)40r obtaining
., replacement samples in the next Semiannual Radioactive' Effluent Release Report and also include in the report a revised figure (s) and table for the ODCM reflecting the new location (s).
b ' One or more instruments, such as a pressurized,lon chamber, for measuring and recording dose rate continuously may be used in~ place of, or in, addition to, integrating dosimeters. For the purposes of this h table, a thermoluminescent dosimeter (TLD)'is considered to be one phosphor; two or more phosphors in a
~
packet are considered as two or more dosimeters. Film badges shall not be used as dosimeters for measuring direct radiation. {Thc 10 statier.3 i s nui. an obselui: - ' ' - - h^ '-- ' ' " - ^ - * - ' '" = ' i a a
-^^'^"J---
5tet;una may bc .mJoccd acccrding 16 geegraphic;I !!aitetion;, e.g., e' 2a eca=a ci'a, see cer'ers '
-Le uvci waier su i.het the ""-hcr 6' desieetcr; 2;y be redeced ;cccrdingly. ha franancy af'mae!ytic cr -
rcadout 'r- TLD system "i11 depea.d upc- the chcoacteri', tic; cf the :peci'ic syste 3:cd 2nd-shetAl be-- ce!ccted to obtairm'yt h"7 dese : for ;tica with air,isel fuding.) C Airborne particulate, sample filters shall be analyzed for gross beta radioactivity 24 hours or more af ter y sampling to allow for radon and thoron daughter decay. If gross beta activity in air particulate samples
) is greater than' ten times the yearly mean of control samples, gamma isotopic analysis shall be perf ormed ' .
sw on the individual samples.
/ .
s.
./
t
/
/
,2
's - - - _TABi r. 3. t?-ljContinued)
,/
/
It-DLL _1_101 AT1011 d
% Gama isotopic analysis means the identification aral quantification of gamma emitting raelionuclide's d that may be attributable s to the effluents from the f.tcility.
%_ . , .. '.-- ._ u , . .,... _m
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ont i n I i uu m.s. u.
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iii..,. _ . . ,. . . m ,. n , ., , ; , , o
,....>.s.
ss, u 8 x
- s. m....... _ . .m. .. . . - - . . . _. , . . , , ...,i.i. .,_ w. kon sihn, thic - . . - :- m. , ,r . ' ' i - -.d.. ~ ."m '. nu u . ' .- .w--.-
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- - - ' ' - - r.r---- . " - --'
g _. ._.. ... . . . . . . : ,.;,.ou.cnt . - - . m.
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u - v i vei i i. . u . are : u i o m, . .s.. .u. . m . . u.a . n o i. . um ,__ _ . - -e to s we - xs The tinse shall be calculated for the maximum organ and age grotip, insing ~tlie methodology and parameters in the ODCH.
,- s g
D f h ,r* os t nrestre more t h;in nnr a $ p; , ,; pi l e y e. h., f t 8 -- p c ;- l c,, ,;nt n in e- i .p . 'O d'-.*a , h -- ;; ,; 'r t - - , c , *, , - , o r e. r n n, i rn erstic. l u < ;.er ,, l i n n, chill I_s o evi , n t_ f . f .,_, ^ ^ tc ,*, h;,p h;il l Im n;. i rl i n i n t- I n ri i n ri e,;tmn f o t n f-- O$ . , ..
. . ,.- . . . _ . . . . . . . . > < ,-~u. . . ..n,e..,.,
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\ r=4 N 3 W (J
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< ^ .:
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C. - ' ,CO- O. Q ag
< - 6N C C C C C r=* N N CC U *= M *-e M e-e N e >- > U C>
N Z Q C. 3 c-e w J w + e U M Z tt3 Cl"A 2n Q C W U - c4
}
J +J ny C3 > b r=4
<. W O
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s C N si ciag 43 Wi
=
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W M did' ( c1 .Cr o C 01 C Q. j e er w OC) e o e e 33 i , on e o po o e e L. nf 3 M / b% QC) C C O O O C M LO C Cp' E:
/ 4J **= QC>. C. v Q M M V N +>
+J U . cr ea l
90 CL 3v CJ 0J M r=4 :;%
.'}3 Cno C: ( L ,o. \
,,, .e.
a: C3 e\ O C: C3 #
/ m m ac"
+--C3
.- C". r-e t. = *[
e N f h er C's C3 O m e
.O r=4 er m M 8
#U 93C) r1 c6.
A s m m m m W to Z M r=4 r=* J L. si f0 m e e e e e i M i 4 e C)% = c(. C e C C C C C L- I 141 m 'O WG
< ::'*. E 6 V U N N *-* U U C3 ft M PWR-STS-RETS 3/4 12-9 1/4/83
TABLE 4.12-1 2 x DETECTION CAPABILITIES FOR ENVIRONMENTAL SAMPLE ANALYSI5a ,gi 7 N -
'O N \ f' T LOWER LIMIT OF DETECTION (LLD)'
A m
\ .
/
- Airborne Particulate Fish Milk Food Pro' ducts Sediment Water N Analysis (pCi/t) N or Gas (pCi/m3 ) (pCi/kg, wet) (pCi/1) (pC,i/kg, wet) (pCi/kg dry)
I gross beta 4
'N 0.01 ,/
N 3 coo ' H-3 2000* s.
~
'/
N / Hn-54 15 ', 130 /
/
Fe-59 30 260 , /
/
Co-58,60 15 ' -1'30 x '-
- 30 /
Zn-65 260 '- g u y
,/
x. 4 o Z r-Hb-95 15 's n I-131 [h 0.07 1-60 Cs-134 15 /y0.05 130 15 - 6'O 150 Cs-137 18 0.06 150 18 80 180 l Ba-La-140 15" b # 15 '* l * ! f r.: dri nkirl;; .-_ __. ,,. .. , e-is ts , ; valuc of 3GGG pCi/i .ey tm u,;1 _ Detedi%'lkit fo, e-4 khA9 w<de* pahay
/ O s '
* . +
TABLE 4.12-1 (Continued)
/
/
TABLE NOTATION .
/
a / Thi s iist does not mean that only these nuclides are to be considered. Other peaks that are identifiable, together with those'of the above nuclides, shall also be analyzed and reported in the' Annual Radiological Environm' ental Operating Report pursuant to Specific'ation 6.9.1.KS. b Required detection capabilities for thermoluminescent dosimeters used i for environm' ental measurements shall be in acc'ordance with the 1 recommendations of Regulatory Guide 4.13. / c
'\ [
The LLO is define'd, for purposes of these' specifications, as the smallest concentration of radioactive material in a sample that will yield a net count, above system background, that will be detected with 95% probability with only 5% probability of falsely concluding that a blank observation represents a "real" signal. For a particular measuremeilt system, which may include radiochemical separation: \ /
/
4.66/'Nsb' LLO = s E - V - 2.22 - Y 's exp(-Aat) Where:
/ \
/ 's
/ \
LLO is the "a prfori" lower limit of detection s as defined above, as picoeuries / pe,r unit mass or volumet'\ h is the standard deviation of the background counti.ng rate or of s tne counting' rate of a blank sample as appropriate, as counts per minute, E is the ' counting efficiency, as counts per disintegration, V is the sample size in units of mass or volume,\\ 2.22/s i the number of disintegrations perxminute per picoeurie, Y!sthefractionalradiochemicalyield,whenapplicable,
\
is the radioactive decay constant for the particular rddionuclide, and
\ '
at for environmental samples is the elapsed time between sample collection, or end of the sample collection period, and time of, counting N Typical values of E, V, Y, and at should be used in the calculation. F PWR-STS-RETS 3/4 12-11 1/4/83
/
TABLE 4.12-1 (Continued) - TABLE NOTATION It should be recognized that the LLD is defined as an a priori (before j the fact) limit representing the capability of a measurement system and not as an a posteriori (after the fact) limit for a partic'ular measurement. Analyses ,shall be performed in such a manner that the s,tated LLDs will be achieved under routine conditions. Occasionally background fluctuations, unavoidable 'small sample sizes, the presence of interfering nuclides, or other uncontr61,lable circumstances may render these'LLDs unachievable. In such cases, the contributing factors shall be, identified and described in the Annual Radiological Environmental Operating Report pursuant to Specification 6.9.1. M g j t
/
Ln m u _u .,m,,.;._,, n - 2_ .d . . . . - " ' ' " " " ' 'u" ,., ..,. Iuw'X l T, e_?'._.'~" T. ,. " J. .,.4.,.
,i. , &...,..J."",".',7,"'""""',,,,s..-..
. d_ 1 ., 21
- h WS Joes net predude. the. calculedios of aq/a po.sf e,. o -
'" **Skr u t w4 Lase.4 en g g, , 44,
' ac.tuh p. . e4.g3 g,, N sample in %ucs4/en and "Perotrtaic de% cer*edeon paramelers'sueg as gecay undele !agsa andpl .
aurin <g analp6 \f
@0 Po.re,A only l\
e.The Bu-No LLD and concesi aSon can 1.hdeter-inel by1he analysis of f3 5 kod -lioed daugkier pro'hui L. - 190 s \ub se gued +o an & day potoA followth colledron, TAe ca/w/a//ov da// I,e predicled on ne gor,,/ ng ,gtj eguahons kr a Ave.d daupfer sifudion and he assa=/ {og f 1 hat any unsu/pr/ed k-lW tn%e sample. would have decayed to an ins /quf, ceil a oud(,/ / ear / u% g' value.), The Introw1A ogua-/Sn.s wf// assu,me.-f4,t the suppoM e4 ils ort' inal La.-tHo aclivshh <1 -% tw of colledbr is 3ero\. s 9 N PWR-STS-RETS 3/4 12-12 1/4/83 l -
RADIOLOGICAL ENVIRONMENTAL MONITORING t l 3/4.12.2 LAND USE CENSUS t LIMITING CONDITION FOR OPERATION 3.12.2 A land use census shall be conducted and shall identify within a distance of 8 km (5 miles) the location in each of the 16 meteorological sectors of the nearest milk animal, the nearest residence and the nearest garden" of greater than 50 m2 (500 ft 2) producing broad leaf vegetation. -(Ser-- devet.cd reicsses c; dewed % Regulatcry Guide 1 111. %v sie- 1, Aly 27', i j -the laad---use census W 11 eise identify within a distanc; cf 5- k- (3 -ilc:) i i -the lecction; in cach cf the meeocehgic21 sector: cf c ilk s W a!; and-c1' gardes--cf grec.tcr ther 50 ,2 p. educing e voa ic:f eccetsticr.) j APPLICABILIJ(: At all times. ACTION: hilead 24%) Q
- a. With a land use census identifying 'a location (s) that yields a-calculated dose or dose commitment V greater than the values cu'rrently being calculated in Specification 4.11.2.3, ia ' ice of : Licences s Event Repe n., identify the new location (s) in the next, Semiannual Radioactive
\ Effluent Release Report, pursuant to Specification 6.9.1.1 6.
- b. /
With a land use census identifying a location (s) that yields a calc'sulated dose or dose commitment (via the same/ exposure pathway) . l 20 percent greater than at a location from which samples are cur-Q rently beiigt obtained in accordance with Specification 3.12.1, add r EPeiS5'oa the new location (s) to the radiological environmental monitoring program within 30 daysp The sampling location (s), excluding the from W ousner 1 control stationglocation, having the losest calculated dose or dose
+2 oued sa 91 ey commitment (s), v1a the same exposuredathway, may be deleted from ma u oL % ea' this monitoring pr,ogram after (Octotier 31) of the year in which this 4 gy,gg land use census was ccnducted. Ir' 'ieu of a Licentec Event Report i , y ,, aM Sursuant to Specification ,6f9.1.129, identify the new location (s) "' #* " b' in the next Semiannual Ra'dioactive Effluent Release Report and also include in the report a rev'ised figure (s) and table for the 00CM reflecting the new locatioi1(s)(
- c. The provisions of Specifications 3:0.3 and 3.0.4 are not applicable.
SURVEILLANCE REOUIREMENTS -
\
4.12.2 The land use censu/s shall be conducted during w sc leastonceper12monthsusingthatinformationthatgt;negrowingseasonat w4 provided the be t by cikr c ult:, :uch at by- or by consulting
@ local agriculturethorities.
au,a' door-to-door The results of the survey, aerial land use , census included in the Arinual Radiological Environmental Operating Re' port pursuant to survey,s shall be Specification 6.'9.1.1L6
/
- Broad leaf' vegetation sampling ef 2t least thr-ee-d4"aert 'ind:m' '/cgetctiw may be,.p'erformed at the site boundary in each of two dif ferent directiori sectors with,the highest predicted D/Qs in lieu of the garden census. Specifications foy broad leaf vegetation sampling in Table 3.12-1.4c shall be followed, \
including analysis of control samples.
/
/ PWR-STS-RETS 3/4 12-13 1/4/83
RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.3 INTERLABORATORY COMPARISON PROGRAM LIMITING CONDITION FOR OPERATION referestel $ 3.12.3 Analyses shall be performed on -en radioactive materials, supplied as part of an Interlaboratory Comparison Program that has been approved by the Comission, that correspond to samples required by Tab!: 3.12-1. - Icdcr APPLICABILITY: At all times. ACTION: ith a'alyses n not being performed as required above, report-th'e corr.ective actions taken to prevent a recurrence to the-Commission in th'e' Annual Radiological Environmental Operating' Report pursuant to Specification 6.9.1.MS.
- b. The provisions of Spec.ifications 3.0'3 and 3.0.4 are not applicable.
, SURVEILLANCE REQUIREMENTS
@ Tk d. mag ,e+/ e 3 -ne nac appcovel rsseAahomke x ,, ,L,gs se AkJ I.% 00ch 4.12.3 -The Ir,tey bcr:t ry Co miser "regr= sh:!'y ce,,,praa,, erbe 2 :rjb__ '- the 00:M.
A summary of ,the results obtained as part of the above require'd'Interlaboratory Comparison Program shall be included in the Annual Radiological Envi'ronmental Operating Report pursuant to Specification 6.9.1.R N l l PWR-STS-RETS 3/4 12-14 1/4/83
JUSTIFICATIONS Section 3/4 Bases In the text of Section 3/4 Bases, 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. No part length rods in Seabrook core. ' C. This information provided to show the range of target bands allowed at Seabrook. E. This page(s) added as a result of Radiological Environmental Tech Specs. F. No such system exists at Seabrook. G. Seabrook cannot operate with less than 4 loops in operation. H. Provides clarification for electrical design of Seabrook. J. This change incorporated as a result of the boron dilution event analysis. L. Physics tests are done in MODE 2 when there is no nuclear heat (heat is generated by reactor coolant pumps). Therefore, rod motion has no impact on Tavg. Thus, rod motion will not cause Tavg to change or fall below the limits of Specification 3.1.1.4. M. No drinking water pathway exists. P. This provides consistency with the cold overpressure mitigation analysis that the pressurizer is assumed not to be solid, thus there is no need to reference pressurizer level. R. This change provides consistency with the FSAR discussion concerning the safety analysis. S. This change incorporated as a result of the cold overpressure mitigation analysis.
9 f BASES FOR - SECTIONS 3.0 AND 4.0 LIMITING CONDITIONS FOR OPERATION AND _ SURVEILLANCE REQUIREMENTS Of
+
1 NOTE The BASES contained in succeeding pages. summarize the reasons for the Specifications in Section 3.0 and 4.0, but in accordance with 10 CFR 50.36 are not part of these Technical Specifications. P-AUG 7 $80
3/4.0 APPLICABILITY BASES The specifications of this section provide the general requirements applicable to each of the Limiting Conditions for Operation and Surveillance Requirements within Section 3/4. In the event of a disagreement between the requirements stated in these Technical Specifications and those stated in an applicable Federal Regula-tion or Act, the requirements stated in the applicable Federal Regulation or Act shall take precedence and shall be met. 3.0.1 This specification defines the applicability of each specification in terms of defined OPERATIONAL MODES or other specified conditions and is provided to delineate specifically when each specification is applicable. 3.0.2 This specification defines those conditions necessary to constitute compli-ance with the terms of an individual Limiting Condition for Operation and associated ACTION requirement. 3.0.3 The specification delineates the measures to be taken for those circum-stances not directly provided for in the ACTION statements and whose occurrence would violate the intent of a specification. For example, Specification 3.5.2 requires two independent ECCS subsystems to be OPERABLE and provides explicit ACTION requirements if one ECCS subsystem is inoperable. Under the requirements of Specification 3.0.3, if both the required ECCS subsystems are inoperable, within 1 hour measures must be initi-ated to place the unit in at least HOT STANDBY within the next 6 hours, and in at least HOT SHUTDOWN within the following 6 hours. As a further example, Specification 3.6.2.1 requires two Containment Spray Systems to be OPERABLE and provides explicit ACTION ! requirements if one Spray System is inoperable. Under the requirements of Specifica-tion 3.0.3, if both the required Containment Spray Systems are inoperable, within 1 hour measures must be initiated to place the unit in at least HOT STANDBY within the next 6 hours, in at least HOT SHUTDOWN within the following 6 hours, and in COLD SHUTDOWN within the subsequent 24 hours. It is acceptable to initiate and complete a reduction in OPERATIONAL MODES in a shorter time interval than required in the ACTION statement and to add the unused portion of this allowable out-of service time to that provided for operation in subsequent lower OPERATION MODE (S). Stated allowable out of-service times are applicable regardless of the OPERATIONAL MODE (S) in which the inoperability is discovered but the times provided for achieving a made reduction are not applicable if the inoperability is discovered in a made lower than the applicable mode. For exam-pie if the Containment Spray System was discovered to be inoperable while in STARTUP, the ACTION Statement would allow up to 156 hours to achieve COLD SHUTDOWN. If HOT STANDBY is attained in 16 hours rather than the allowed 78 hours, 140 hours would still be available before the plant would be required to be in COLD SHUTDOWN However, if this system was discbvered to be inoperable while in HOT STANDBY, the 6 hours provided to achieve HOT STANDBY would not be additive to the time available to achieve COLD SHUTDOWN so that the total allowable time is reduced from 156 hours to 150 hours. 3.0.4 This specification provides that entry into an OPERATIONAL MODE or otner specified applicability condition must be made with: (1) the full complement of required systems, equipment, or ccmponents OPERABLE and (2) all other parameters as specified in the Limiting Conditions for Operation being met without regard for allowable deviations and out of-service provisions contained in the ACTION statements. The intent of this provision is to ensure that facility operation is not initiated with either required equipment or systems inoperable or other specified limits being exceeded. Exceptions to this provision have been provided for a limited numcer of specifi- ! cations when startuo with inoperable equicment would not affect plant safety. These exceptions are stated in the ACTION statements of the apprcoriate specifications.
'f-STS B 3/4 0-1 l
APPLICABILITY BASES 4.0.1 This specification provides that surveillance activities necessary to insure the Limiting Conditions for Operation are met and will be performed during the OPERATIONAL MODES or other conditions for which the Limiting Condi-tions for Operation are applicable. Provisions for additional surveillance activities to be performed without regard to the applicable OPERATIONAL MODES or other conditions are provided in the individual Surveillance Requirements. Surveillance Requirements for Special Test Exceptions need only be performed when the Special Test Exception is being utilized as an exception to an individual specification. 4.0.2 The provisions of this specification provide allowable tolerances for performing surveillance activities beyond those specified in the nominal surveillance interval. These tolerances are necessary to provide operational flexibility because of scheduling and performance considerations. The phrase "at least" associated with a surveillance frequency does not negate this allowable tolerance value and permits the performance of more frequent surveillance activities. The tolerance values, taken either individually or consecutively over 3 test intervals, are sufficiently restrictive to ensure that the reliability , associated with the surveillance activity is not significantly degraded beyond ' that obtained from the nominal specified interval. 4.0.3 The provisions of this specification set forth the criteria for determination of compliance with the OPERABILITY requirements of the Limiting Conditions for Operation. Under this criteria, equipment, systems or components are assumed to be OPERABLE if the associated surveillance activities have been satisfactorily performed within the specified time interval. Nothing in this provision is to be construed as defining equipment, systems or components OPERABLE, when such items are found or known to be inoperable although still meeting the Surveillance Requirements. 4.0.4 This specification ensures that the surveillance activities associated with a Limiting Condition for Operation have been performed within the specified time interval prior to entry into an OPERATIONAL MODE or other applicable condition. The intent of this provision is to ensure that surveil-lance activities have been satisfactorily demonstrated on a current basis as required to meet the OPERABILITY requirements of the Limiting Condition for Operation. Under the terms of this specification, for example, during initial plant startup or following extended plant outages, the applicable surveillance activities must be performed within the stated surveillance interval prior to placing or returning the system or equipment into OPERABLE status. , i !
~
W-STS B 3/4 0-2 e ---=y- .mep - - + ~ - _, s-. .-w-- ,_ ~ - - -
i APPLICABILITY BASES 4.0.5 This specification ensures that 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 will be performed in accordance with a periodically updated version of Section XI of the ASME Boiler and Pressure Vessel Code and Addenda as required by 10 CFR 50.55a. Relief from any of the above requirements has been provided in writing by the Commission and is not a part of these Technical Specifications. This specification includes a clarification of the frequencies for performing the inservice inspection and testing activities required by Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda. This clarification is provided to ensure consistency in surveillance intervals thoughout these Technical Specifications and to remove any ambiguities relative to the frequencies for performing the required inservice inspection and testing activities. Under the terms of this specification, the more restrictive requirements of the Technical Specifications take precedence over the ASME Boiler and Pressure Vessel Code and applicable Addenda. For example, the requirements of Specification 4.0.4 to perform surveillance activities prior to entry into an OPERATIONAL MODE or other specified applicability condition takes precedence over the ASME Boiler and Pressure Vessel Code provision which allows pumps to i be tested up to one week after return to normal operation. And for example, the Technical Specification definition of OPERABLE does not grant a grace period before a device that is not capable of performing its specified function it declared inoperable and takes precedence over the ASME Boiler and Pressure Vessel Code provision which allows a valve to be incapable of performing its specified function for up to 24 hours before being declared inoperable. JUL 2 7133; W-STS B 3/4 0-3
3/4.1 REACTIVITY CONTROL SYSTEMS BASES 3/4.1.1 BORATION CONTROL 3/4.1.1.1 and 3/4.1.1.2 SHUTDOWN MARGIN A sufficient SHUTDOWN MARGIN ensures that 1) the reactor can be made subcritical from all operating conditions, 2) the reactivity transients associated with postulated accident conditions are controllable within acceptable limits, and 3) the reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition. SHUTDOWN MARGIN requirements vary throughout core life as a function of fuel depletion, RCS boron concentration, and RCS T avg. The most restrictive condition occurs at EOL, with T,yg at no load operating temperature, and is associated with a postulated steam line break accident and resulting uncon-
) trolled RCS gjoldown. In the analysis of this accident, a minimum SHUTDOWN MARGIN of (-+-6%) delta k/k is required to control the reactivity transient.
l Accordingly, the SHUTDOWN MARGIN requirement is based upon this limiting condition and is consistent with FSAR safety analysis assumptions. 44?th 7 avg
-I c s s -th a n 200 C, the "eacti Jity transfer.t: resultir.g froa 6 pc-bulated stccm riine bi cok wclde-i. arc -inimal and a 1T Alta b/b SHUTDCUM "ARGIN pmuide.s
.-adcquite prctectica 6 ce. L S ed T.
3/4.1.1.3 MODERATOR TEMPERATURE COEFFICIENT The limitations on moderator temperature coefficient (HTC) are provided to ensure that tne value of this coefficient remains within the limiting condition assumed in the FSAR accident and transient analyses. The MTC values of this specification are applicable to a specific set of plant conditions; accordingly, verification of HTC values at conditions other than those explicitly stated will require extrapolation to those conditions in order to permit an accurate comparison. The most negative MTC value equivalent to the most positive moderator density coefficient (MDC), was obtained by incrementally correcting the MDC ! used in the FSAR analyses to nominal operating conditions. These corrections W-STS B 3/4 1-1 NOV 2 01980
~
INSERT I With T avg less than 200 F, the most restrictive condition is associated with an inadvertent boron dilution event and resulting loss of shutdown margin. In the analysis of this event, a minimum SHUTDOWN MARGIN of 1.2% delta k/k is required to permit sufficient time for the operator to terminate the dilution prior to loss of all shutdown margin. Accordingly, the SHUTDOWN MARGIN requirement with Tay less than 200 F is based upon this limiting condition andisconsistentwi$hFSARsafetyanalysisassumptions.
REACTIVITY CONTROL SYSTEMS BASES MODERATOR TEMPERATURE COEFFICIENT (Continued) involved subtracting the incremental change in the MDC associated with a core condition of all rods inserted (most positive MDC) to an all rods withdrawn condition and, a conversion for the rate of change of moderator density with temperature at RATED THERMAL POWER conditions. This value of the MDC was then transformedintothelimitingMTCvalue(-N)x10~4 delta k/k/ F. The MTC Ha 4 value of (-3<6) x 10 delta k/k/ F represents a conservative value (with corrections for burnup and soluble boron) at a core condition of 300 ppm '~ equilibrium boron concentration and is obtained by making these corrections to thelimitingMTCvalueof(M)x1~ k/ F. The surveillance requirements for measurement of the MTC at the beginning and near the end of the fuel cycle are adequate to confirm that the MTC remains within its limits since this coefficient changes slowly due principally to the , reduction in RCS boron concentration associated with fuel burnup. 3/4.1.1.4 MINIMUM TEMPERATURE FOR CRITICALITY 5'51 This specification ensures that the reactor will not be mad critical with the Reactor Coolant System average temperature less than (!'r44) F. This limitation is required to ensure 1) the moderator temperature coefficient is within it analyzed temperature range, 2) the protective instrumentation is
" within its normal operating range, 3) the P-12 interlock is above its setpoint,
- 4) the pressurizer is capable of being in an OPERABLE status with a steam bubble, and 5) the reactor pressure vessel is above its minimum RTNDT temperature.
3/4.1.2 B0 RATION SYSTEMS The boron injection system ensures that negative reactivity control is available during each mode of facility operation. The components required to perform this function include 1) borated water sources, 2) charging pumps,
- 3) separate flow paths, 4) boric acid transfer pumps, 5) cc:ac;oicd Luot- tiscire systau, and p an emergency power supply from OPERABLE diesel generators.
m MODES t , L a.nd 3 With the RCS-avcrcg tcy ciotse e L e 200@ % a minimum of two boron injection flow paths are required to ensure single functional capability in the event an assumed failure renders one of the flow paths incperable. The boration capability of either flow path is sufficient to provide a SHUTDOWN 4 W-STS B 3/4 1-2 .NOV 2 0 E50
REACTIVITY CONTROL SYSTEMS BASES BORATION SYSTEMS (Continued) a mini-* ~ co 4a m d U8I*
. + 2 0,20 0 l3 MARGIN from expected operating conditions of h4% delta k/k after xenon decay and cooldown to 200 F. The maximum expected boration capability requirement, O occurs at E0L from full power equilibrium xenon conditions and requires 1
gallons ofM ppm borated water from the boric acid storage tanks or V2,52" gallons of 2000 ppm borated water from the refueling water storage tank. N(g -wi--- ce.dai e4 volu ~ic of (17 9.0o 0 ) Uith th "CS tcmperature be!cu 200 I, cae injecticn systc- ir accept:ble
-w+1_hnyt gingle fafk rg ccg3ideratjCG-on the Lua it-O f thO StablO PO2Cti"4 ty--
-. condition of the ranctnr and the additional rc;trictions prohibiting CORE $ee
-hTERAMONS-and-pos i ti v e . eseth+ty-changes in the event the 5irrg-le in jecti m Sycter becomee i nep e rabl e-l Truer t 3
-The !i-itation for 2 maximum of cae-centrifugal charging p ap tc be -
OPERACLE and the Sus veillerree-Requirement to vecify all charging pu pe aveop+ the required OPERABLE pump Lu be inspcrablebelow(2'5{'Fprovides;;;urence-
-that a mass eddition-pressure transien+ --
nf a 4 ingle PORV.- c2nie .el. M y thc operat4on a sic %. co.Jamea vota e, of 4800]
~ The boron capability required below 200 F is sufficient to provide a SHUTDOWN MARGIN of 1.2% delta k/k after xenon decay and cooldown from 200 F to Q 140"F. Inis condition requires eithe W( ) gallons of Shee0%pm borated water from the boric acid storage tanks or ) gallons of 2000-ppm borated _
water from the refueling water storage tank. A w ., h J Ei b ed volu ~e of 2%IS) The contained water volume limits include allowance for water not available because of discharge line location and other physical characteristics. The limits on contained water volume and boron concentration of the RWST also ensure a pH value of between (8.5) and (11.0) for the solution recirculated within containment after a LOCA. This pH band minimizes the evolution of iodine and minimizes the effect of chloride and caustic stress corrosion on mechanical systems and components. The OPERABILITY of one boron injection system during REFUELING ensures that this system is available for reactivity control while in MODE 6. 3/4.1.3 MOVABLE CONTROL ASSEMBLIES The specifications of this section ensure that (1) acceptable power distribution limits are maintained, (2) the minimum SHUTDOWN MARGIN is main-tained, and (3) limit the potential effects of rod misalignment on associated accident analyses. OPERABILITY of the control rod position indicators is required to determine control rod positions and thereby ensure compliance with I the control rod alignment and insertion limits. 1 The liMfaf8hu *a ofex4aiuly ,& th gas in 110005 9,r ,,4 t, ensu,, At Il e baron dtI i rh ccm ao f exceed N vet lu e a s sw.i e d si w tra ns,ckt' ana /ysh . W-STS B 3/4 1 C 'NOV 2 0 G20 l l
INSERT II The limitation for a maximum of one-centrifugal charging pump to be OPERABLE and the Surveillance Requirement-to verify all charging pumps except the required OPERABLE pump to be inoperable in MODES 4,5, and 6 provides assurance that a mass addition pressure transient can be relieved by operation of a single PORV. As a result of this, only one boron injection system is available. This is acceptable on the basis of the stable reactivity condition of the reactor, the emergency power supply requirement for the OPERABLE charging pump and the additional restrictions prohibiting CORE ALTERATIONS and positive reactivity-changes in the event the single injection system becomes inoperable. i s i. r e
't- e i )
t. i.
- ,- y 9 . 9 -. -o. - - . .-e- e r s,,- e,,,ww,.m. ,,%- ,
w.. . ., m. ,_m,..-- r-.. , _ . - - ._ym
REACTIVITY CONTROL SYSTEMS BASES MOVABLE CONTROL ASSEMBLIES (Continued) The ACTION statements which permit limited variations from the basic requirements are accompanied by additional restrictions which ensure that the original design criteria are met. Misalignment of a rod requires measurement of peaking factors and a restriction in THERMAL POWER. These restrictions provide assurance of fuel rod integrity during continued operation. In addition, those safety analyses affected by a misaligned rod are reevaluated to confirm that the results remain valid during future operation. The maximum rod drop time restriction is consistent with the assumed rod drop time used in the safety analyses. Measurement with T greater than or equalto($k$)Fandwithallreactorcoolantpumpsoperatngensuresthat the measured drop times will be representative of insertion times experienced during a reactor trip at operating conditions. Control rod positions and OPERABILITY of the rod position indicators are required to be verified on a nominal basis of once per 12 hours with more frequent verifications required if an automatic monitoring channel is inoperable. These verification frequencies are adequate for assuring that the applicable LCO's are satisfied. l (ALTEP'!AT: ;r .equ4 red by 0"o censiecration:)- The restriet, inn nenhibiting pa rt lergt))-FOd--iMertie" onern ee that adupeco
-power shape: and . opid lccal power-changes which asy -##a-+ n'" ----#d---+4-a-
-de-not-occur as e resulteet%th red imer+4an d" ring operation.-
Not app [l cable +o Seo.lovoo ft l l l W-STS B 3/4 1-4 'fl0V 2 01980 e
3/4.2 POWER DISTRIBUTION LIMITS BASES ~ The specifications of this section provide assurance of fuel integrity during Condition I (Normal Operation) and II (Incidents of Moderate Frequency) events by: (a) maintaining the minimum DNBR in the core greater than or equal to 1.30 during normal operation and in short-term transients, and (b) limiting the fission gas release, fuel pellet temperature, and cladding mechanical properties to within assumed design criteria. In addition, limiting the peak linear power density during Condition I events provides assurance that the initial conditions assumed for the LOCA analyses are met and the ECCS acceptance criteria limit of 2200 F is not exceeded. _ ,. .- A~~~ k, The definitions of~certain hot Scr Ga: and channel EcMpeaking factors as used in 3 thes'e-specifications are as follows: Fq (Z) Heat Flux-liot Channel Factor, is defined as the maximum,loca'l heat flux on theQurface of a fuel rod at core eleyation Z divided by the average fuel' rod heat flux, allowing ,for' manufacturing tolerances N on fuel pellets and rod
/N f
F
# Nuclear Enthalpy Rise Hot Cha~nnel Facto ms is defined as the ratio of I the integral of linear 10wer aiong the rod with'the shighest integrated power to the,aversge rod power.
F (Z) Radil a king Factor, is defined as the ratio of peak power density to average power density in the horizontal plane at core elevation Z. ] i
) .
3/4.2.1 AXIAL FLUX DIFFERENCE The limits on AXIAL FLUX DIFFERENCE (AFD) assure that the F (Z) upper bound 9 envelope of 2.32 times the normalized axial peaking factor is not exceeded during either normal operation or in the event of xenon redistribution following power changes. Target flux difference is determined at equilibrium xenon conditions.witA
@ & prt-!ength cont-col--rodsr-w-thdrm:~ frca the-cone-- The full-length rods may be positioned within the core in accordance with their respective insertion limits and should be inserted near their normal position for steady-state operation at high power levels. The value of the target flux difference obtained under these conditions divided by the fraction of RATED THERMAL POWER is the target flux difference at RATED THERMAL POWER for the associated core burnup conditions. Target flux differences for other THERMAL POWER levels are obtained by multiplying the RATED THERMAL POWER value by the appropriate fractional THERMAL POWER level. The periodic updating of the target flux difference value is necessary to reflect core burnup considerations.
W-STS B 3/4 2-1 WOV 2 019E0
i i i POWER DISTRIBUTION LIMITS BASES AXIAL FLUX DIFFERENCE (Continued) rep d 6ecu6cdtod 3.2@ Although it is intended that the plant will be operated with the AFD within the T(-5-)% target bandPabout the target flux difference, during rapid plant THERMAL POWER reductions, control rod motion will cause the AFD to deviate outside of the target band at reduced THERMAL POWER levels. This deviation will not affect the xenon redistribution sufficiently to change the envelope of peaking factors which may be reached on a subsequent return to RATED THERMAL POWER (with the AFD within the target band) provided the time duration of the deviation is limited. Accordingly, a 1-hour penalty deviation limit cumulative during the previous 24 hours is provided for operation outside of the target band but within the limits of Figure (3.2-1) while at THERMAL POWER levels between 50% and 90% of RATED THERMAL POWER. For THERMAL POWER levels between 15% and 50% of RATED THERMAL POWER, deviations of the AFD outside of the target band are less significant. The penalty of 2 hours actual time reflects this reduced significance. Provisions for monitoring the AFD on an automatic basis are derived from the plant process computer through the AFD Monitor Alarm. The computer determines the 1-minute average of each of the OPERABLE excore detector outputs and provides an alarm message immediately if the AFD for 2 or more OPERABLE excore channels are outside the target band and the THERMAL POWER is greater c than 90% of RATED THERMAL POWER. During operation at THERMAL POWER levels between 50% and 90% and between 15% and 50% RATED THERMAL POWER, the computer outputs an alarm message when the penalty deviation accumulates beyond the limits of I hour and 2 hours, respectively. r- r _. . . . o e ra a * - - - m. p .a a u. * - at hand ""* M
@ 4e th dif fo.J target ba=Js Provided T
"***'Sfor' f " E Gab ka t 34h oS i* U Or e h U a O n'el E 2 2 3/4.2 2'alids3/4.2.3 HEAT FLUX.HOTxCHANNEL. FACTOR, and'R 9 ELOWRATE-AN NUCLEAR ENTHALPLRISE HOTsCHANNEL FACTOR / Ne v
/ ,,.-
N The limits on heat flux hot channel factor, RCS flowrate, and nuclear enthalpy rise hot channel factor ensure that 1) the design limits on' peak local powersdensity and minimum DNBR are not exceeded and 2) in the event of a LOCA the peakNfuel clad temperature will not exceed the 2200 F ECCS acceptance criteria limit. f e' g ygg/c., .py x / s e n :l.i.L Each of these is measurable but will normally only be determined periodically as specified in Sp'ecifications'4.2.2 and 4.2.3. This periodic surveillance is sufficient to insure 3that the limits are maintained provided:
/ %
- a. ControlrodsinasinglegroupmovetNstherwithnoindividualrod s insertion differing by more than + M steps, ihdicated, from the group demand position. 12. N
- b. Control rod gr oups are sequenced with overlapping groups as Nscribed .
in Specification 3.1.3.6. NN W-STS B 3/4 2-2 HOV 2 S 1980
l I I 5% 5% g 1.0 ,--- W i 3 0.9 $ t O l Fon einer cone l 000 J
- 0. 8 - i
<( i 2 ' O.7 - ' 1 i . __ TAR G ET FLU X l>J i / DIFFERENCE 3- 0.6 ' i / }- 'A' C 0. 5 - I W i <i- 0.4 ' i T i i l.l I O 03 ', Z 4 O 0.2 = i 5 ! O 0.1 - i << i T ! 1 0.0 , . i , , ,
-30 -20 -10 0 10 20 30 INDICATED AXIAL FLUX DIFFERENCE FIGURE B 3/4 2-1 TYPICAL INDICATED AXIAL FLUX DIFFEllENCE VEflSUS THERMAL POWER Seabrook - Units 1&2 B 3/4 2-3 .
i l g 1.0 , , w i + 3 % ,- 12 % 3 g'g _ ', i N Fi m CORE O ' a i > 3000 g 0.8 - 1 MWD /MTU
< i RELOAD 2 0.7 - i CORES E '!
i W ' _- TARGET FLUX 0.6-
% i' DIFFERENCE Q 0.5 - i w i F<-- 0*4-5 1 i
-u_ i O 0.3 - ; Z i O 0.2 - ', I ' O 0.1 - '
< i C I u- 0.0 , , , , , , ,
DIC TED A lAL FLUX Di FEREN E (%) FIGURE B 3/4 2-2 TYPICAL INDICATED AXIAL FLUX DIFFERENCE VERSUS THERMAL POWER Seabrook - Units 1&2 B 3/4 2-4
3/4.2 POWER DISTRIBUTION LIMITS BASES 3/4.2.2 & 3/4.2.3 PLANAR RADIAL PEAKING FACTOR and NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR The planar radial peaking factor is defined as the ratio of peak power density to average power density in a horizontal plane at core elevation Z. The nuclear enthalpy rise hot channel factor is defined as the ratio of the integral of linear power along the rod with the highest integrated power to the average rod power. The limits on planar radial peaking factor and nuclear enthalpy rise hot channel factor ensure tha t 1) the design limits on peak local power density and minimum DNBR are not exceeded and 2) in the event of a LOCA the peak fuel clad tem-perature will not exceed the 2200*F ECCS acceptance criteria limit. Each of these is measurable but will normally only be determined periodically as specified in Specifications 4.2.2 and 4.2.3. This periodic surveillance is suf-ficient to insure tha t the limits are maintained provided:
- a. Control rods in a single group move together with no individual rod insertion differing by more than + 12 s teps , indicated, from the group demand position.
- b. Control rod groups are sequenced with overlapping groups as described in Specifica tion 3.1.3.6.
- c. The control rod insertion limits of Specifications 3.1.3.5 and 3.1.3.6 are maintained.
- d. The axial power dis tribution, expressed in terms of AXIAL FLUX DIFFERENCE, is maintained within the limits.
The limit on peak local power density is of ten referred to in terms of the heat flux hot channel factor . Fq, F Q is defined as the maximum local heat flux on the surface of a fuel rod divided by the average fuel rod heat flux. For Seabrook the LOCA analysis justifies a Q F limit of 2.32 at 1007. RATED THERMAL POWER. 3/4.2.4 QUADRANT POWER TILT RATIO The purpose of this specification is to de tec t gross changes in core power dis tribution be tween monthly incore flux maps. During normal opera tion the QUADRANT POWER TILT RATIO is set equal to zero once acceptability of core peaking factors has been es tablished by review of incore maps. The limit of 102 is established as an indication tha t the power distribution has changed enough to warrant further inves tiga tion.
3/4.2 POWER DISTRIBUTION LIMITS BASES 3/4.2.5 DNB PARAMETERS The limits on the DNB-related parameters assure dhat each of the parameters are maintained within the ' normal steady-s tate envelope of operation assumed in the transient and accident analyses. The limits are consis tent with the -initial FSAR assumptions and have been analytically demonstrated adequate to maintain a minimum DNBR of 1.30 throughout each analyzed transient. The 12-hour periodic surveillance of these parameters through instrument readout is sufficient to ensure tha t the parameters are restored within their limits following load changes and other expected transient operation. a
3/4.3 INSTRUMENTATION BASES 3/4.3.1 and 3/4.3.2 REACTOR TRIP SYSTEM and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION The OPERABILITY of the Reactor Trip System and the Engineered Safety Features Actuation System instrumentation and interlocks ensures that: (1) the associated ACTION and/or Reactor trip will be, initiated when the parameter monitored by each channel or combination thereof reaches its Setpoint, (2) the specified' coincidence logic is maintained, (3) sufficient redundancy is main-tained to.. permit a channel to be out-of-service for testing or maintenance, and (4) sufficient system functional capability is available from diverse parameters. The OPERABILITY of these systems is required to provide the overall reliability, redundancy, and diversity assumed available in the facility design for the protection and mitigation of accident and transient conditions. The integrated operation of each of these systems is consistent with the assumptions used in the safety analyses. The Surveillance Requirements speci-fied for these systems ensure that the overall system functional capability is maintained comparable to the original design standards. The periodic surveil-lance tests performed at the minimum frequencies are sufficient to demonstrate this capability. - The Engineered Safety Features Actuation System Instrumentation Trip Setpoints specified in Table 3.3-4 are the nominal values at which the bistables are-set for each functional unit. A Setpoint is considered to be adjusted consistent with the nominal value when the "as measured" Setpoint is within the band allowed for calibration accuracy. To accommow e the instrument drift assumed to occur between operational tests and the accaracy to which Setpoints can be measured and calibrated, Allowable Values for the Setpoints have been specified in Table 3.3-4. Opera-tion with Setpoints less conservative than the Trip Setpoint but within the Allowable Value is acceptable since an allowance has been made in the safety analysis to accommodate this error. An optional provision has been included for determining the OPERABILITY of a channel when its Trip Setpoint is found to exceed the Allowable Value. The methodology of this option utilizes the "as measured" deviation from the specified calibration point for rack and sensor components in conjunction with a statistical combination of the other uncertainties of the instrumentation to measure the process variable and the uncertainties in calibrating the instrumentation. In Equation 3.3-1, ! Z + R S < TA, the interactive effects of the errors in the rack and the i sensor, and the "as measured" values of the errors are considered. Z, as scecified in Table 3.3-4, 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, R or Rack Error is the "as measured" deviation, in the percent span, for the affected channel from the scecified Trip Setpoint. 5 or Sensor Error is either the "as measured" deviation of l S 3/4 3-1 y-STS
INSTRUMENTATION ' BASES REACTOR TRIP SYSTEM and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION (Continued) the sensor from its calibration point or the value specified in Table 3.3-4, in percent span, from the analysis assumptions. Use of Equation 3.3-1 allows - for a sensor drift factor, an increased rack drift factor, and provides a threshold value for REPORTABLE EVENTS. The methodology to derive the Trip Setpoints is based upon combining all of the uncertainties in the channels. . Inherent to the determination of the Trip Setpoints are the magnitudes of these channel uncertainties. Sensor and rack instrumentation utilized in these channels are expected to be capable of operating within the allowances of these uncertainty magnitudes. Rack drift in excess of the Allowable Value exhibits the behavior that the rack has not met its allowance. Being that there is a small statistical chance that this will happen, an infrequent excessive drift is expected. Rack or sensor drift, in excess of the allowance that is more than occasional, may be indicative of more serious problems and should warrant further investigation. - The measurement of response time at the specified frequencies provides
- assurance that the Reactor trip and the Engineered Safety' Features actuation associated with each channel is completed within the time limit assumed in the safety analyses. No credit was taken in the analyses for those channels with [
response times indicated as not applicable. Response time may be demonstrated by any series of sequential, overlapping, or total channel test measurements provided that such tests demonstrate the total channel response time as defined. Sensor response time verification may be demonstrated by either: (1) in place, onsite, or offsite test measurements, or (2) utilizing replacement sensors with certified response time. The Engineered Safety Features Acutation System senses selected plant parameters and determines whether or not predetermined limits are being exceeded. If they are, the signals are combined into logic matrices sensitive to combina-tions indicative of various accidents, events, and transients. Once the required logic combination is completed, the system sends actuation signals to those Engineered Safety Features components whose aggregate function best serves the requirements of the condition. As an example, the following actions may be initiated by the Engineered Safety Features Actuation System to mitigate the consequences of a steam line break or loss-of coolant accident: (1) Safety Injection pumps start and automatic valves position, (2) Reactor trip, (3) feed-water isolation, (4) startup of the emergency diesel generators, (5) containment spray pumps, start and automatic valves position, (6) cont nment isolation (7) steam line isolation, (8) Turbine trip, (9) e '1'a0 eedwater pumps g ' ,,,, q ,nc7) start and automatic valves position, (10) containment cooling fans start and automatic valves position, (11) essential service water pumps start and auto-matic valves position, and (12) Control Room Isolation and Ventilation Systems start. - ( W-STS B 3/4 3-2
INSTRUMENTATION BASES REACTOR PROTECTION SYSTEM AND ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION (Continued) The Engineered Safety Feature Actuation System interlocks perform the following functions: P-4 Reactor tripped - Actuates turbine trip, closes main feedwater valves on T,yg below setpoint, prevents the opening of the main feedwater valves which were closed by a safety injection or high steam generator water level signal, allows safety injection block so that components can be reset or tripped. Reactor not tripped prevents manual block.of safety injection. P-11 On increasing pressurizer pressure, P-11 automatically reinstates safety injection actuation on low pressurizer pressure. On decreasing pressure, P-11 allows the manual block of safety injection actuation on low pressurizer' pressure. i I 12 g On .nu coeinc pr e "Y C00hnt 1;op tempcroturc,- P-12 outematically--
-rcin t?.tss safety injectica ectuoGun un high stc = ficw coincident with =4+ her l'/w low I ec lew stees line g e m e, e.M previt cm P-N on me,' easing 3 fen, Senerd.- unfe level F-lhdo=4hcal/7 i As All fe ed wa.bev isola:f-ion valVes and sn hiin% -Feelwdco codvol valve s modata.bn.
3/4.3.3 MONITORING INSTRUMENTATION 3/4.3.3.1 RADIATION MONITORING INSTRUMENTATION The OPERABILITY of the radiation monitoring channels ensures that 1) the radiation levels are continually measured in the areas served by the individual channels and 2) the alarm or automatic action is initiated when the radiation level trip setpoint is exceeded. 3/4.3.3.2 MOVABLE INCORE DETECTORS The OPERABILITY of the movable incore detectors with the specified minimum complement of equipment ensures that the measurements obtained from use of this system accurately represent the spatial neutron flux distribution of the reactor core. The OPERABILITY of this system is demonstrated by irradiating each detector used and determining the acceptability of its voltage curve. N For the purpose of measuring Fq (Z) or F AH a full incore flux map is used. Quarter-core flux maps, as defined in WCAP-8648, June 1976, may be used in recalibration of the excore neutron flux detection system, and full incore flux maps or symmetric incore thimbles may be used for monitoring the QUADRANT
\ POWER TILT RATIO when one Power Range Channel is inoperable.
W-STS B 3/4 3 SEP 151981
INSTRUMENTATf0H < l ( BASES 3/4.3.3.3 SEISMIC INSTRUMENTATION The OPERABILITY of the seismic instrumentation ensures that sufficient capability is available to promptly determine the magnitude of a seismic event and evaluate the response of those features important to safety. This capability is required to permit comparison of the measured response to that used in the design basis for the facility to determine if plant shutdown is required pursuant to Appendix "A" of 10 CFR Part 100. The instrumentation is consistent with the recommendations of Regulatory Guide 1.12, " Instrumentation for Earthquakes," April 1974. 3/4.3.3.4 METEOROLOGICAL INSTRUMENTATION . The OPERABILITY of the meteorological instrumentation ensures that sufficient meteorological data is available for estimating potential radiation doses to the public as a result of routine or accidental release of radioactive materials to the atmosphere. This capability is required to evaluate the need for initiating protective measures to protect the health and safety of the public and is consistent with the recommendations of Regulatory Guide 1.23, "Onsite Meteorological Programs," February 1972. 3/4.3.3.5 REMOTE SHUTDOWN INSTRUMENTATION The OPERABILITY of the remote shutdown instrumentation ensures that sufficient capability is available to permit shutdown and maintenance of HOT STANDBY of the facility from locations outside of the control room. This capability is required in the event control room habitability is lost and is consistent with General Design Criteria 19 of 10 CFR 50. 3/4.3.3.6 ACCIDENT MONITORING INSTRUMENTATION The OPERABILITY of the accident monitoring instrumentation ensures that sufficient information is available on selected plant parameters to monitor and assess these variables following an accident. This capability is consistent with the recommendations of Regulatory Guide 1.97, " Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant Conditions During and Following "TMI-2 Lessons Learned Task Force an Accident," Status December Report and 1975 Short-Term and NUREG 0578,#.I Recommendations Il l c W-STS B 3/4 3 .l NOV 2 01980
INSTRUMENTATION BASES 4h'4 .3.3.7 CHLORTHE DETECTION SYSTEMS - The OPERA 51Liii ui Lhe chivi."e de tec L i u" sy=Lem e"su. ca ihai sufficient e=a'5414ty i: aveilauie to pruuagily detect cnd iaitista protccti e acLica in - 4he-event-of-en acc ide"Loi chicrine rcicase. This wepability ;s required Lu - 0F -pectact-contrci .vem ger v""ci and is ce">i>te"L -ith the reccameneeticas cf Regulatory-Guide 1.35, "ProceuLicamf Sclear Pcwer Pie"t Contrci Ree Opcestcra Againet-en-Aee4 dental Chiurine Release," Icbruecy 1075. 7 3 /4. 3. 3.1L FIRE DETECTION INSTRUMENTATION OPERABILITY of the fire detection instrumentation ensures that adequate warning capability is available for the prompt detection of fires. This capability is required in order to detect and locate fires in their early stages. Prompt detection of fires will reduce the potential for damage to safety-related equipment and is an integral element in the overall facility fire protection program. In the event that a portion of the fire detection instrumentation is g inoperable, the establishment of frequent fire patrols in the affected areas is required to provide detection capability until the inoperable instrumentation is restored to OPERABILITY. . 3/4.3.3. LOOSE-PART DETECTION INSTRUMENTATION see JusMtca.4,t @ W Seebron 3/4J - The OPERABILITY of the'lodse part. detection instrumentation-ensUris that sufficient capability is available to defeEt-loose metallic parts in the
~~
~
primary system and avoid or mitigate damage'to jrimary-system components.
~
The allowable out-of-service times.and' surveillance requirements'are-consistent
~
with the recommendations"3f Regulatory Guide 1.133, " Loose-Part Det'EEtion Progm for-thTPrimary System of Light-Water-Cooled Reactors," May 1981. i l t N l
$I
.W-STS B 3/4 3-4 uCV 2 1981
l INSTRUMENTATION BASES l 9 3/4.3.3.K RADI0 ACTIVE LIOUID EFFLUENT MONITORING INSTRUMENTATION The radioactive liquid effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in liquid effluents duririg actual or potential releases of liquid effluents. The alarm / trip setpoints for these instruments shall be calculated and adjusted in accordance with the methodology and parameters in the ODCM to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20. The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63 and 64 of Appendix A to 10 CFR Part 50. The purpose of tank level indicating devices is to assure the detection and control of leaks that if not controlled could potentially result in the transport of radioactive materia-1s to UNRESTRICTED AREAS. . 10 3/4.3.3. K RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION The radioactive gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases of gaseous effluents. The alarm / trip setpoints for these instruments shall be calculated and adjusted in accordance with the methodology and parameters in the ODCM to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20. This instrumentation also includes provisions for monitoring (and controlling) the concentrations of potentially explosive gas mixtures in the WASTE GAS HOLDUP SYSTEM. The OPERABILITY and use of this instrumentation is consistent with the requirements of General, Design Criteria 60, 63 and 64 of Appendix A to 10 CFR Part 50. . 3/4.3.4 TURBINE OVERSPEED PROTECTION gg. p gg* n 3h.3 This specification is provided to ensure that the turbine erspeedov. / protection instrumentatDand-theyrbine speed contr61 valves are OPERABLE and will protect the turbine from exc.essiveovenpeed. Protection from turbine excessive overspeed is requiged-sitiEe excessive overspeed-of thqurbine could generate potentially damaging missiles which could impact and damag'e' safety related componen~ts, equipment, or structures. 6 PWR-STS-RETS B3/43-) 1/4/83
3/4.4 REACTOR CCOLANT SYSTEM BASES 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION The plant is designed to operate with all reactor coolant loops in operation and maintain DNBR above 1.30 during all normal operations and anticipated transients. In MODES 1 and 2 with one reactor coolant loop not in operation this specification requires that the plant be in at least HOT STAND 8Y . within 6 hours. I In MODE 3, two reactor coolant loops provide sufficient heat removal capability for removing core decay heat even in the event of a bank withdrawal accident; however, a single reactor coolant loop provides sufficient heat removal capacity if a bank withdrawal accident can be prevented, i.e., by opening the Reactor Trip System breakers. Single failure considerations require that two loops be OPERABLE at all times. In MODE 4, and in MODE 5 ~with reactor coolant icops filled, a single reactor coolant loop or RHR loop provides sufficient heat removal capability for removing decay heat; but single failure considerations require that at least two loops (either RHR or RCS) be OPERA 8LE. ~ In MODE 5 with reactor coulant loops not filled, a single RHR Icop provides sufficient heat removal capability for removing decay heat; but single failure considerations, and the unavailability of the steam generators as a heat removing component, require that at least two RHR Icops be OPERA 8LE. The operation of one reactor coolant pump (RCP) or one RHR pump provides adequate flow to ensure mixing, prevent stratification and produce gradual reactivity changes during baron concentration reductions in the Reactor Coolant System. The reactivity change rate associated with baron reduction will, therefore, be within the capability of operator recognition and control. na M50GS Hc The restrictions on starting an RCP with-or.c co-more.KrS~ "CS caid icg: !::;- @
-than-cr equal iv [275] I ??? provided to prevent RCS pressure transients, caused by energy additions from the Secondary Coolant System, which could exceed the
- limits of Appendix G to 10 CFR Part 50. The RCS will be protected against overpressure transients and will not exceed the limits of Appendix G-by e4 t"er-
-(l) wirining-t-hc ater veh=c in the pr-essurire" "d-weretj gr=idi .g4 -vclu.T.e for-the-reaetar-coolent to egend-htc. cr (2)- by restricting starting of the RCPs to when the secondary water temperature of each steam generator is less than fjl*F above each of the RCS cold leg temperatures.
_ [7nrtnun 3
-4 greater than uirement or '*to maintain the baron concentration of an isolated ler the baron concentration of the operatin , ensures that no reactivity additio . + e core could occur durinc .up of an isolated loop. Verification of the .. concentr'*'.. in an idle loop prior to opening the stop valves provides a reassi- f the adecuacy of the boron concentration in the isolated loop. " ating the iso m loco on recirculating flow for at least 00 minutes to opening its stop valves e.. < adequate s
mixing of the coolant '
..is loop and prevents any reactivity effects uw '
, baron concent stratifications, y-5 " B 3/4 4-1 i i
3/4.4 REACTOR COOLANT SYSTEM BASES
-REACTOR CCOLANT LOOPS eD CGUuiii CIRCiiLATION-(Cui.Linued)
(-OFTIONAL) - of an idle loop will inject cool water from the loo
- core. The reactivi
- resulting from this c . injection is minimized by delaying isolated loop '
its temperature is within 20 F of the operating loo g the reactor s 1 prior to loop startu any power spike which could result from wa - uced reactivity transient. I 3/4.4.2 SAFETY VALVES i The pressurizer code safety valves operate to prevent the RCS from being pressurized above its Safety' Limit of 2735 psig. Each safety valve is designed W8#3 to relieve N lbs per hour of saturated steam at the valve setpoint. The h condition which could occur during shutdown. In therelief eventcapacity of a single safety valve is that no safety valves are OPERABLE, an operating RHR loop, connected to the RCS, provides overpressure relief capability and will prevent RCS overpressurization. In addition, the Overpressure Protection System provides a diverse means of protection against RCS overpressurization at low temperatures. j During operation, all pressurizer code safety valves must be OPERABLE to prevent the RCS from being pressurized above its safety limit of 2735 psig. The combined relief capacity of all of these valves is greater than the maximum surge rate resulting from a complete loss of load assuming no reactor trip until the first Reactor Protective System trip setpoint is reached (i.e. , no credit is taken for a direct reactor trip on the loss of load) and also assuming no operation of the power-operated relief valves or steam dump valves. Demonstration of the safety valves' lift settings will occur only during { shutdown and will be performed in accordance with the provisions of Section XI i of the ASME Boiler and Pressure Code. 3/4.4.3 PRESSURIZER 6ce Tosert H
-The-1-imi-t-on-the-maximua-water--vokme-in-the-pressurizer assures-that-the
.-parameter __is-maintained-w+thm-the-normal-steady-stete-envelope of operat4cn assumed-in-the-SAR. The ligiit is cerisis-tentwi-th the-inith4-SAR assu=p+innt
-Thq 12 houi periodic surveillance ic ("fficiant tn oneuvo that +ho nar, motor _
c45-restered tn within its limit fnlinwing eypected-transient oper-atier The
-BaMtifft-Wats" t!Olu"9--al SO-ensures-that 2 stonm huhhle is formod and t h i_, e the-
-RGS-4s-not-a-hydrauliculiy sui idiystem Tho rpnnirement that 3 -inimum '
-number-of- pressur42er-heate a be OPERASLE cabanees the ennahility nf-the plant-
-to-control-Reactor Caslant Spteat-pres,surn and establish natural circulation _
l W-STS B 3/4 4-2 NOV 2 6 GEO
INSERT III 3/4.4.3 PRESSURIZER During normal operation, the pressurizer level is maintained about a programmed setpoint either by the operation of the pressurizer level control system or manual operation. The pressurizer level setpoint is a function of average coolant temperature and varies between approximately 25 and 60 percent of span. The intent is to have an essentially constant RCS mass regardless of -
. power level. The initial FSAR assumption of pressurizer level 5% above normal programmed level accounts for appropriate measurement uncertainties. The limit on the maximum water volume in the pressurizer of 92% is not bounded by the FSAR but ensures that a steam bubble is formed and thus the RCS is not a hydraulically solid system. The 12-hour periodic surveillance is sufficient to ensure that the parameter is restored-to within it's limit following expected transient operation. The requirement that a minimum number of pressurizer heaters be OPERABLE enhances the capability of the plant to control Reactor Coolant System pressure and establish natural circulation.
I
REACTOR COOLANT SYSTEM BASES 3/4.4.4 RELIEF VALVES The power-operated relief valves (PORVs) and steam bubble function to relieve RCS pressure during all design transients up to and including the design step load decrease with steam dump. Operation of the PORVs minimizes the undesirable opening of the spring-loaded pressurizer code safety valves. Each PORV has a remotely operated block valve to provide a positive shutoff capability should a relief valve become inoperable. 3/4.4.5 STEAM GENERATORS Sec_ 3 ash thtCILbOd bedior\ b N ) ' The Surveillance Requirements for inspection of the steam generator tub s ensure that the structural integrity of this portion of the RCS will be ma'in-tained. The program for inservice inspection of steam generator tubes /is based on modification of Regulatory Guide 1.83, Revision 1. Inservice inspection of steam generator tubing is essential in order to maint'ain surveil-lance of the' conditions of the tubes in the event that there is e'vidence j of mechanical dama,ge o progressive degradation due to design, manufacturing errors,orinservYceconditionsthatleadtocorrosion. Inservice inspection i I of steam generator tilbing also provides a means of characterizing the nature I and cause of any tube degradation so that corrective mea'sures can be taken.
\ /
The plant is expected to be x operated in a manner such that the secondary coolant will be maintained within those chemistry limits found to result in
~
negligible corrosion of the steam \ generator tubes. If the secondary coolant chemistry is not maintained within these limits, localized corrosion may likely result in stress corrosion cracking.' The extent of cracking during plant operation would be limited by the.1 imitation of steam generator tube leakage between the primary coolant system \and the secondary coolant system (primary-to-secondary leakage = 500 gallons p'eq day per steam generator). Cracks having a primary-to-secondary leakage less, than this limit during operation will have an adequate-inargin of safety to withstand the loads imposed duringnormaloperationandby~postulatedaccidentsk0peratingplantshave demonstrated that primary-to-secondary leakage of 500 ga,llons per day per steam generator can readily be detected by radiation monitors of steam generator blowdown. Leakage in ex'ess c of this limit will require pihn,t shutdown and an unscheduled inspection', during which the leaking tubes will be located and plugged. /
/
Wastage-typ'e defects are unlikely with proper chemistry treatment of the secondary coolant. Howev e , even if a defect should develop in service it will be foun'd during scheduled inservice steam generator tube examinatio,ns. plugging, limit of (40)% of the tube nominal wall thickness. Plugging w'ill be required f Steam generator, tube inspections of operating plants have demonstrated the capability to rel,iably detect degradation that has penetrated 20% of the original tube wal thickness.
/
W-STS B 3/4 4-3 NOV 2 01980
REACTOR COOLANT SYSTEM BASES SNE'AM* GENERATORS (. Continued) Whenever the r hh 30 dat a _ df-anyQteam/generatortubingiryrviceinspection fall into Category C-3, these resultsfti.Q be promptly reported to the Commission pursuanttoSpecification6.9.Thiortojesump'tknofplantoperation. Such cases result inwill be considered by%_ lysis, laboratory examinatiTrisdteststhe a requirement-for additional Comm eddy-current insp 6 cn, and revision of the Technical Specificati Es,> if cessary7 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.6.1 LEAKAGE DETECTION SYSTEMS The RCS leakage detection systems required by this specification are provided to monitor and detect leakage from the Reactor Coolant Pressure Boundary. These detection systems are consistent with the recommendations of Regulatory Guide 1.45, " Reactor Coolant Pressure Boundary Leakage Detection Systems," May 1973. 3/4.4.6.2 OPERATIONAL LEAKAGE Industry experience has shown that while a limited amount of leakage is expected from the RCS, the unidentified portion of this leakage can be reduced to a threshold value of less than 1 GPM. This threshold value is sufficiently low to ensure early detection of additional leakage. The Surveillance Requirements for RCS Pressure Isolation Valves provide added assurance of valve integrity thereby reducing the probability of gross valve failure and consequent intersystem LOCA. Leakage from the RCS Pressure Isolation Valves is IDENTIFIED LEAKAGE and will be considered as a portion of the allowed limit. The 10 GPM IDENTIFIED LEAKAGE limitation provides allowance for a limited amount of leakage from known sources whose presence will not interfere with the detection of UNIDENTIFIED LEAKAGE by the lea,kage detection systems. The CONTROLLED __ LEAKAGE S 3us4tCncal-ten (Y opera) limiti(1on restricts Secftotion n W+4 when the totaLf. low
- supplied to the reactor coolant pump seals.exce_eds (52.) GPM.with'this modulating valve in the supply line fully open at a, nominal"RCS pressure of 2235 psig.
, This limitation ensu g that-in-the-~liEnt of a LOCA, the safety'injectic will notJe-less than assumed in the accident analyses. The total steam generator tube leakage limit of 1 GPM for all steam generators not i clated-f.cm the RCS-ensures that the dosage contribution from the tube _ leakage will be limited to a small fraction of Part 100 limits in the event of either a steam generator tube rupture or steam line break. The 1 GPM limit is consistent with the assumptions used in the analysis of these accidents. The 500 gpd leakage limit per steam generator ensures that steam generator tube integrity is maintained in the event of a main steam line rupture or under LOCA conditions. W-STS 8 3/4 4-4 'NOV 2 01920
REACTOR COOLANT SYSTEM BASES OPERATIONAL LEAKAGE (Continued) PRESSURE BOUNDARY LEAKAGE of any magnitude is unacceptable since it may be indicative of an impending gross failure of the pressure boundary. Therefore, the presence of any PRESSURE BOUNDARY LEAKAGE requires the unit to be promptly placed in COLD SHUTDOWN. 3/4.4.7 CHEMISTRY The limitations on Reactor Coolant System chemistry ensure that corrosion of the Reactor Coolant System is minimized and reduces the potential for Reactor Coolant System leakage or failure due to stress corrosion. Maintaining the chemistry within the Steady State Limits provides adequate corrosion protection to ensure the structural integrity of the Reactor Coolant System over the life of the plant. The associated effects of exceeding the oxygen, chloride, and fluoride limits are time and temperature dependent. Corrosion studies show that operation may be continued with contaminant concentration levels in excess of the Steady State Limits, up to the Transient Limits, for the specified limited time intervals without having a significant effect on the structural integrity of the Reactor Coolant System. The time interval
.' permitting continued operation within the restrictions of the Transient Limits provides time for taking corrective actions to restore the contaminant concen-trations to within the Steady State Limits.
The.furveillance @equirements provide adequate assurance that concentrations in excess of the limits will be detected in sufficient time to take corrective action. 3/4.4.8 SPECIFIC ACTIVITY The limitations on the specific activity of the primary coolant ensure that the resulting 2-hour doses at the site boundary will not exceed an appropriately small fraction of Part 100 limits following a steam generator tube rupture accident in conjunction with an assumed steady state primary-to-secondary steam generator leakage rate of 1.0 GPM. The values for the limits on specific activity represent limits based upon a parametric evaluation by the NRC of typical site locations. These values are conservative in that specific site parameters of the (Seabrook ) site, such as site boundary location and meteorological conditions, were not considered in this evaluation. W-STS B 3/4 4-5 NOV 2 01980
REACTOR COOLANT SYSTEM BASES SPECIFIC ACTIVITY (Continued) The ACTION stat'ement permitting POWER OPERATION to continue for limited time periods with the primary coolant's specific activity greater than 1.0 microcuries/ gram DOSE EQUIVALENT I-131, but within the allowable limit shown on Figure 3.4-1, accommodates possible iodine spiking phenomenon which may occur following changes in THERMAL POWER. Operation with specific activity levels exceeding 1.0 microcuries/ gram DOSE EQUIVALENT I-131 but within the limits shown on Figure 3.4-1 must be restricted to no more than 800 hours per year (approximately 10 percent of the unit's yearly operating time) since the activity levels allowed by Figure 3.4-1 increase the 2-hour thyroid dose at the site boundary by a factor of up to 20 following a postulated steam generator tube rupture. The reporting of cumulative operating time over 500 hours in any 6 month consecutive period with greater than 1.0 microcuries/ gram DOSE EQUIVALENT I-131 will allow sufficient time for Commission evaluation of the circumstances prior to reaching the 800-hour limit. Reducing T,y to less than 500 F prevents the release of activity should a steam generator tube rupture since the saturation pressure of the primary coolant is below the lift pressure of the atmospheric steam relief valves. The surveillance requirements provide adequate assurance that excessive specific activity levels in the primary coolant will be detected in sufficient time to take corrective action. Information obtained on iodine spiking will be used to assess the parameters associated with spiking phenomena. A reduction in frequency of isotopic analyses following power changes may be permissible if justified by the data obtained. 3/4.4.9 PRESSURE / TEMPERATURE LIMITS The temperature and pressure changes during heatup and cooldown are limited to be consistent with the requirements given in the ASME Boiler and Pressure Vessel Code, Section III, Appendix G.
- 1) The reactor coolant temperature and pressure and system heatup and cooldown rates (with the exception of the pressurizer) shall be limited in accordance with Figures 3.4-2 and 3.4-3 for the first full power service period.
a) Allowable combinations of pressure and temperature for specific temperature change rates are below and to the right of the limit lines shown. Limit lines for cooldown rates between those presented may be obtained by interpolation. M-STS B 3/4 4-6 NOV 2 01980
REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (Continued) b) Figures 3.4-2 and 3.4-3 define limits to assure prevention of non-ductile failure only. For normal operation, other inherent plant characteristics, e.g. , pump heat addition and pressurizer heater capacity, may limit the heatup and cooldown rates that can be achieved over certain pressure-temperature ranges.
- 2) These limit lines shall be calculated periodically using methods provided below.
- 3) The secondary side of the steam generator must not be pressurized above 200 psig if the temperature of the steam generator is below 70 F.
- 4) The pressurizer heatup and cooldown rates shall not exceed 100 F/ hr and g " _ 200 5 respectively. The spray shall not be used if the temperature difference between the pressurizer and the spray fluid is greater than 320*F.
- 5) System preservice hydrotests and in-service leak and hydrotests shall be performed at pressures in accordance with the requirements of ASME Boiler and Pressure Vessel Code, Section XI.
The fracture toughness properties of the ferritic materials in the reactor vessel are determined in accordance with the NRC Standard Review Plan, ASTM E185-73, and in accordance with additional reactor vessel require-ments. These properties are then evaluated in accordance with Appendix G of the 1976 Summer Addenda to Section III of the ASME Boiler and Pressure Vessel Code and the calculation methods described in WCAP-7924-A, " Basis for Heatup and Cooldown Limit Curves, April 1975." Heatup and cooldown limit curves are calculated using the most limiting value of the nil-ductility reference temperature, RTNDT'.at the end of (/d effective full power years of service life. The (/M EFPY service life
~
period is chosen such that the limiting RTNDT at the 1/4T location in the core region is greater than the RT f the limiting unirradiated material. NDT The selection of such a limiting RTNDT assures that all components in the Reactor Coolant System will be operated conservatively in accordance with applicable Code requirements. W-STS d 3/4 4-7 .NOV 2' 01950
l l REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (Continued) The reactor vessel materials have been tested to determine their initial RTNDT; the results of these tests are shown in Table B 3/4.4-1. Reactor operation and resultant fast neutron (E greater than 1 MEV) irradiation can cause an increase in the RTNDT. Therefore, an adjusted reference temperature, based upon the fluence and copper content of the material in question, can be predicted using Figure B 3/4.4-1 and the recommendations of Regulatory Guide 1.99, Revision 1, " Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials." The heatup and cool-down limit curves of Figures 3.4-2 and 3. 4-3 include predicted adjustments for this shift in RT NDT at the end of ('/2) EFPY (as well as adjustments for possible errors in the pressure and temperature sensing instruments). See &s4&c&o 1 @ Sedwn 3h-4 V ues of ART NDT determined in this manner may be used until the resul.ts f from the ma'eria surveillance program, evaluated according,to' ASTM E185, are available. Capsule removed in danc /e with the require-ments of ASTM E185-73 and 10 CFR p ppe. dix H. The surveillance specimen withdrawal schedule is shewn-in' Table 4.4-5. The-heatup and cooldown curves must be reca.culat the ART NDT determined from sur,eillance capsule egceeds the calculated ARTNDT for the equivalent capsule radiation exposure. Allowable pressure-temperature relationships for various heatup and cooldown rates are calculated using methods derived from Appendix G in Section III of the ASME Boiler and Pressure Vessel Code as required by l Appendix G to 10 CFR Part 50, and these methods are discussed in detail in i WCAP-7924-A. The general method for calculating heatup and cooldown limit curves is based upon the principles of the linear elastic fracture mechanics (LEFM) technology. In the calculation procedures a semi-elliptical surface defect with a depth of one quarter of the wall thickness, T, and a length of 3/2T is assumed to exist at the inside of the vessel wall as well as at the outside of the vessel wall. The dimensions of this postulated crack, referred to in Appendix G of ASME Section III as the reference flaw, amply W-STS B 3/4 4-9 NOV .'0 850
~
I? 4 w TABLE B 3/4.4-1' REACTOR VESSEL T0tlGilNESS ASME 50 FT-LB/35 RT
~
COMP MATERIAL CU P NDTT MIL TEMP *F NOT FT-LB COMPONENT CODE TYPE % % *F LONG TRANS *F LONG TRANS P+ s. M - w b Q D
?
e I $3 (b k O-_ 4, p+ e J - G > h r n
> CA E R
- m ~
to g G-- o
20 1 10 _
^
eq - 1/4T E _. O N E 10' _ . w : . O - 3/4T Z W . D - J u_ Z o 10'
- _
E - F-- : . D - W - Z _ 17 - 10 i i i i , . 0 5 10 15 20 25 30 35 EFFECTIVE FULL POWER (YEARS) FIGURES B 3/4 4.1 FAST NEUTRON FLUENCE (E<1mev) As a Function of Full Power Service Life. Seabrook - Units 1&2 B 3/4 4-10
REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (Continued) exceed the current capabilities of inservice inspection techniques. Therefore, the reactor operation limit curves developed for this reference crack are conservative and provide sufficient safety margins for protection against non-ductile failure. To assure that the radiation embrittlement effects are accounted for in the calculation of the limit curves, the most limiting value of the nil ductility reference temperature, RTNDT, is used and this includes the radiation-induced shift, ARTNDT, c rresponding to the end of the period for which heatup and cooldown curves are generated. The ASME approach for calculating the allowable limit curves for various heatup and cooldown rates specifies that the total stress intensity factor, Ky , for the combined thermal and pressure stresses at any time during heatup or cooldown cannot be greater than the reference stress intensity factor, K 7g, 7g is obtained from the reference for the metal temperature at that time. K fracture toughness curve, defined in Appendix G to the ASME Ccde. The K yg curve is given by the equation: K IR = 26.78 + 1.223 exp [0.0145(T-RTNDT + 160)] (1) where K IR is the reference stress intensity factor as a function of the metal temperature T and the metal nil ductility reference temperature RT NDT. Thus, the governing equation for the heatup-cooldown analysis is defined in Appendix G of the ASME Code as follows: C K;g + kit g KIR (2) Where, K yg is the stress intensity factor caused by membrane (pressure) stress. K g is the stress intensity factor caused by the thermal gradients. l W-STS B 3/4 4-11 NOV 2I)IE30
REACTOR COOLANT SYSTEM _B_AS E S PRESSURE / TEMPERATURE LIMITS (Continued) K IR is provided by the code as a function of temperature relative to the RT NOT f the material. C = 2.0 for level A and 8 service limits, and C = 1.5 for inservice hydrostatic and-leak test operations.
.At any time during the heautp or cooldown transient, K IR is determined by the metal temperature at the tip of the postulated flaw, the appropriate value for RTNDT, and the reference fracture toughness curve. The thermal stresses resulting from temperature gradients through the vessel wall are calculated and then the corresponding thermal stress intensity factor, KIT, f r the reference flaw is computed. From Equation (2) the pressure stress intensity factors are obtained and, from these, the allowable pressures are calculated.
COOLDOWN f i For the calculation of the allowable pressure versus coolant temperature during cooldown, the Code reference flaw is assumed to exist at the inside of f the vessel wall. During cooldown, the controlling location of the flaw is l always at the inside of the wall because the thermal gradients produce tensile l stresses at the inside, which increase with increasing cooldown rates. Allowable pressure-temperature relations are generated for both steady-state and finite cooldown rate situations. From these relations, composite. limit curves are constructed for each cooldown rate of interest. The use of the composite curve in the cooldown analysis is necessary because control of the cooldown procedure is based on measurement of reactor coolant temperature, whereas the limiting pressure is actually dependent on the material temperature at the tip of the assumed flaw. During cooldown, the 1/4T vessel location is at a higher temperature than the fluid adjacent to the vessel ID. This condition, of course, is not true for the steady-state situation. It l follows that at any given reactor coolant temperature, the delta T developed i r NOV 2 0 ISSO W-STS B 3/4 4-12 f l
REACTOR COOLANT SYSTEM BASES
=-
PRESSURE / TEMPERATURE LIMITS (Continued) during cooldown results in a higher value of K 7g at the 1/4T location for finite cooldown rates than for steady-state operation. Furthermore, if conditions exist such that the increase in K IR exceeds kit, the calculated allowable pressure during cooldown will be greater than the steady-state value. The above procedures are needed because there is no direct control on temperature at the 1/4T location; therefore, allowable pressures may unknowingly be violated if the rate of cooling is decreased at various intervals along a cooldown ramp. The use of the composite curve eliminates this problem and assures conservative operation of the system for the entire cooldown period. HEATUP Three separate calculations are required to determine the limit curves g for finite heatup rates. As is done in the cooldown analysis, allowable pressure-temperature relationships are developed for steady-state conditions as well as finite heatup rate conditions assuming the presence of a 1/4T defect at the inside of the vessel wall. The thermal gradients during heatup produce compressive stresses at the inside of the wall that alleviate the tensile stresses produced by internal pressure. The metal temperature at the crack tip lags the coolant temperature; therefore, the K f r the 1/4T crack IR during heatup is lower than the K f r the 1/4T crack during steady state IR conditions at the same coolant temperature. During heatup, especially at the end of the transient, conditions may exist such that the effects of compressive thermal stresses and different K IR 's f r steady-state and finite heatup rates do not offset each other and the pressure-temperature curve based on steady-state conditions no longer represents a lower bound of all similar curves for finite heatup rates when the 1/4T flaw is considered. Therefore, both cases have to be analyzed in order to assure that at any coolant temperature the lower value of the allowable pressure calculated for steady-state and finite heatup rates is obtained.- l W-STS B 3/4 4-13 .NOV 2 01950 l
REACTOR COOLANT SYSTEM '1 BASES PRESSURE / TEMPERATURE LIMITS (Continued) The second portion of the heatup analysis concerns the calculation of pressure-temperature limitations for the case in which a 1/4T deep outside surface flaw is assumed. Unlike the situation at the vessel inside surface, the thermal gradients established at the outside surface during heatup produce 4 stresses which are tensile in nature and thus tend to reinforce any pressure
, stresses present. These thermal stresses, of course, are dependent on both the rate of heatup and the time (or coolant temperature) along the heatup ramp. Furthermore, since the thertral stresses, at the outside are tensile and increase with increasing heatup rate, a lower bound curve cannot be defined.
l Rather, 'each heatup rate of interest must be analyzed on an individual basis. Following the generation of pressure-temperature curvt; for both the L steady state and finite heatup rate situations, the final limit curves are produced as follows. A composite curve is constructed based on a point-by-point comparison of the steady-state and finite heatup rate data. At any given temperature, the allowable pressure is taken to be the lesser of the three values taken from the curves under consideration. The use of the composite curve is necessary to set conservative heatup limitations because it is possible for conditions to exist such that over the course of the heatup ramp the controlling condition switches from the inside
' to the outside and the pressure limit must at all times be based on analysis of the most critical criterion. ,
4 Finally, the composite curves for the heatup rate data and the cooldown rate data are adjusted for possible errors in the pressure and temperature sensing instruments by the values indicated on the respective curves. Although the pressurizer operates in temperature ranges above those for which there is reason for concern of non-ductile failure, operating limits are provided to assure compatibility of operation with the fatigue analysis performed in accordance with the ASME Code requirements. The OPERABILITY of two PORVs or an RCS vent opening of greater than (h,7-. square inches ensures that the RCS will be protected from pressure transients which could exceed the limits of Appendix G to 10 FR Part 50 when one or more of the RCS cold legs are less than or equal to (Fr:r F. Either PORV has N adequate relieving capability to protect the RCS from overpressurization when s the transient is limited to either (1) the start of an idle RCP with the secondary water temperature of the steam generator less than or equal to ( )*F above the RCS cold leg temperatures or (2) the start of a-HilM pump and its injection into a water solid RCS, - A
'So&t3 Imechos orchayp l
l 1 ! W-STS B 3/4 4-14 NOV 2 01980
REACTOR COOLANT SYSTEM BASES 3/4.4.10 STRUCTURAL INTEGRITY The inservice inspection and testing programs for ASME Code Class 1, 2, - and 3 components ensure that the structural integrity and operational readiness
.. of these components will be maintained at an acceptable level throughout the life of the plant. These programs are in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50.55a(g) except where specific written relief has been granted by the Commission pursuant to 10 CFR 50.55a(g)(6)(i).
Components of the Reactor Coolant System were designed to provide access to permit inservice inspections in accordance with Section XI of the ASME Boiler and Pressure Vessel Code, tak< Edition and Addenda through later . 3/4.4.11 REACTOR COOLANT SYSTEM VENTS Reactor Coolant System vents are provided to exhaust noncondensible gases and/or steam from the Reactor Coolant System that could inhibit natural circulation core cooling. The OPERABILITY pf least one Reactor Coolant System vent path from the (reactor vessel head]," lac [Rc:ctor Cociant Systea high feiet-3, the (pressurizer steam space]ar.d th; [iscletica caadenaci M gh @
=
pciati-ensures that the capability exists to perform this function. The valve redundancy of the Reactor Coolant System vent paths serves to minimize the probability of inadvertent or irreversible actuation while ensuring that a single failure of a vent valve, power supply, or control system does not prevent isolation of the vent path. The function, capabilities, and testing requirements of the Reactor Coolant System vents are consistent with the requirements of Item II.B.1 of NUREG-0737,
" Clarification of TMI Action Plant Requirements," November 1980.
.1 E-STS B 3/4 4 A'
i 4 3/4.5 EMERGENCY CORE COOLING SYSTEMS BASES 3/4.5.1 ACCUMULATORS The OPERABILITY of each Reactor Coolant System (RCS) accumulator ensures that a sufficient volume of borated water will be immediately forced into the
; reactor core through each of the cold legs in the event the RCS pressure falls i
i below the pressure of the accumulators. This initial surge of water into the core provides the initial cooling mechanism during large RCS pipe ruptures. The limits on accumulator volume, boron concentration and pressure ensure that the assumptions used for accumulator injection in the safety analysis are met. The accumulator power' operated isolation valves are considered to be ,
" operating bypasses" in the context of IEEE Std. 279-1971, which requires that bypasses of a protective function be removed automatically whenever permissive , conditions are not met. In addition, as these accumulator isolation valves ; , fail to meet single failure criteria, removal of power to the valves is required.
i ,
- The limits for operation with an accumulator inoperable for any reason except an isolation valve closed minimizes the time exposure of the plant to a LOCA event occurring concurrent with failure of an additional accumulator 4
whicn may result in unacceptable peak cladding temperatures. If a closed isolation valve cannot be immediately opened, the full capability of one accumulator is not available and prompt action is required to place the , reactor in a mode where this capability is not required. 3/4.5.2 and 3/4.5.3 ECCS SUBSYSTEMS
- The OPERABILITY o two independent ECCS subsystems ensures that sufficient emergency core cooling capability will be available in the event of a LOCA assuming the loss of one subsystem through any single failure consideration.
Either subsystem operating in conjunction with the accumulators is capable of supplying sufficient core cooling to limit the peak cladding temperatures within acceptable limits for all postulated break sizes ranging from the , double ended break of the largest RCS cold leg pipe downward. In addition, each ECCS subsystem provides long-term core cooling capability in the recirculation mode during the accident recovery period. With the RCS temperature below 350 F, one OPERABLE ECCS subsystem is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the limited core cooling requirements. y-STS 8 3/4 5-1 4l0V 2 01980
EMERGENCY CORE COOLING SYSTEMS BASES ECCS SUBSYSTEMS (Continued) The limitation for a maximum of one centrifugal charging pump-ahe-c fcty #njccticr p n to be OPERABLE and the Surveillance Requirement to verify all charging pumps and safety injection pumps except the required OPERABLE charging pump to be inoperable -bdewCP"T provides assurance that amassadditionpressuretransientcanberelievekbytheoperationofa single PORV. {ngoog3q,,d59 The Surveillance Requirements provided to ensure OPERABILITY of each component ensures that at a minimum, the assumptions used in the safety analyses are met and that subsystem OPERABILITY is maintained. Surveillance requirements for throttle valve position stops and flow balance testing provide assurance that proper ECCS flows will be maintained in the event of a LOCA. Maintenance of proper flow resistance anu pressure drop in the piping system to each injection point is necessary to: (1) prevent total pump flow from exceeding runout conditions when the system is in its minimum resistance configuration, (2) provide the proper flow split between injection points in accordance with the assumptions used in the ECCS-LOCA analyses, and (3) provide an acceptable level of total ECCS flow to all injection points equal to or above that assumed in the ECCS-LOCA analyses. 3d Q .4 BORON INJECTION SYSTEM The OP ITY of the boron infection system as part of the ECC sures that sufficient ne *ive reactivity is injected into the core to .teract any positive increase i "eactiv,ity caused by RCS system coo on. RCS cooldown can be caused by inadverten epressurization, a loss-of solsilt accident, or a steam line rupture. The limits on injection tank mini.. concentration ensure that the assumpti c ained volume and boron in the steam line break analysis h are met. The contained water vol ' imit inc as an allowance for water not usable because of tank disch " line location or ce r physical characteristics. The OPERABILIT the redundant heat tracing channels ciated with the boron in en system ensure that the solubility of the boron lution will be ntained above the solub'lity limit of 135 F at 22,500 ppm bo 3/4.5. REFUELING WATER STORAGE TANK The OPERABILITY of the Refueling Water Storage Tank (RWST) as part of the ECCS ensures that a sufficient supply of borated water is available for injection by the ECCS in the event of a LOCA. The limits on RWST minimum volume and boron concentration ensure that 1) sufficient water is available within containment to permit recirculation cooling flow to the core, and 2) the reactor will remain subcritical in the cold condition folicwing mixing of the RWST and the RCS water w-STS B 3/4 5-2 NOV 2 01950
1 l l EMERGENCY CORE COOLING SYSTEMS BASES REFUELING WATER STORAGE TANK (Continued) volumes with all control rods inserted except for the most reactive control assembly. These assumptions are consistent with the LOCA analyses. The contained water volume liinit includes an allowance for water not usable because of tank discharge line location or other physical characteristics. The limits on contained water volume and boron concentration of the RWST also ensure a pH value of between (8.5) and (11.0) for the solution recirculated within containment after a LOCA. This pH band minimizes the evolution of iodine and minimizes the effect of chloride and caustic stress corrosion on mechanical systems and components. y r W-STS 8 3/4 5-3 NOV 2 01980 l
3/4.6 CONTAINMENT SYSTEMS BASES 3/4.6.1 PRIMARY CONTAINMENT 3/4.6.1.1 CONTAINMENT INTEGRITY Primary CONTAINMENT INTEGRITY ensures that the release of radioactive materials from the containment atmosphere will be restricted to those leakage paths and associated leak rates assumed in the accident analyses. This restriction, in conjunction with the leakage rate limitation, will limit the site boundary radiation doses to within the limits of 10 CFR 100 during accident conditions.
^
3/4.6.1.2 CONTAINMENT LEAKAGE The limitations on containment leakage rates ensure that the total containment leakage volume will not exceed the value assumed in the accident analyses at the peak accident pressure, P,. As an added conservatism, the measured overall integrated leakage rate is further limited to less than or equal to 0.75 L,-e- 0.75 Lg, 2: 2pplicebia during performance of the periodic tests to account for possible degradation of the containment leakage barriers between leakage tests.
~
The surveillance testing for measuring leakage rates are consistent with the requirements of Appendix "J" of 10 CFR 50. 3/4.6.1.3 CONTAINMENT AIR LOCKS The limitations on closure and leak rate for the containment air locks are required to meet the restrictions on CONTAINMENT INTEGRITY and containment leak rato. Surveillance testing of the air lock seals provide assurance that the overall air lock leakage will not become excessive due to seal damage l during the intervals between air lock leakage tests. i l l W-DUAL B 3/4 6-10 MAY 151976 s
CONTAINMENT SYSTEMS BASES
!,, C'A rnNTAINMENT ISOLATION VALVE AND CHANNEL WELD PRESSifo'7 E r SYSTEMS (0PTIONAL)
The OPERABILITY of the isola b e and containment channel weld pressurization systems i ed to meet .c ** ctions on overall contain-ment leak r . in the accident analyses. The ance Requirements fp.de mining OPERABILITY are consistent with Appendix "J" of 5 3/4.6.1.kINTERNALPRESSURE
\/o. lues b %o s ed.en will be s uppl/ed af a /deM-The limitations on containment internal pressure ensure that 1) -the.
. containment structurn is nrovnnted from exceeding its design negative pressure
-d444eeent4abwi4h-respesb4c the annulus atrnosphere-of (1,0) pd , and 2) the containment peak pressure does not exceed the design pressure of (%) psig
, during (LOCA or steam line break) conditions. The maximum peak pressure expected to be obtained from a (LOCA or steam line break) event is (%) psig. The limit' of (%) psig for initial positive containment pressure will limit the total pressure to (N ) psig which is less than the design pressure and is consistent with the accident analyses. 3/4.6.1. AIR TEMPERATURE (aven g on w The limitation k containment average air temperature ensures that the oveml] containment# peak air temperature does not exceed the design temperature of (Ide#F during (LOCA or steam line break) conditions and is consistent with the accident analyses. 3/4.6.1.fCONTAINMENTVESSELSTRUCTURALINTEGRITY s. This limitation ensures that the structura integrity of the containment steel vessel will be maintained comparable to the original design standards for the life of the facility. Structural integrit is required to ensure that the vessel will withstand the maximum pressure of ( M ) psig in the event of a (LOCA or steam line break accident). A visual inspection in conjunction with Type A leakage tests is sufficient to demonstrate this capability. 3/4.6.1.kCONTAINMENTVENTILATIONSYSTEM The(k-inch)containmentpurgesupplyandexhaustisolationvalvesare required to be closed during plant operation since these valves have not been demonstrated capable of closing during a LOCA or steam line break accident. Maintaining these vaives closed during plant operations ensures that excessive quantities of radioactive materials will not be released via the containment purge system. The use of the containment purge lines is restricted to the (8-inch) purge supply and exhaust isolation valves to ensure that the site boundary dose guidelines of 10 CFR Part 100 would not be exceeded in the event of a loss-ofmeht accident during purging operations. LOC.4 W-DUAL B 3/4 6-20 SEP 151991
1 CONTAINMENT SYSTEMS BASES 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS 3/4.6.2.1 CONTAINMENT SPRAY SYSTEM The OPERABILITY of the containment spray system ensures that containment depressurization and cooling capability will be available in the event of a LOCA. The pressure reduction and resultant lower containment leakage rate are consistent with the assumptions used in the accident analyses. ({redit taken--for-4od4ne-eemovaM-t YWo independard f ThW containment spray systemsand the c etM - + coo! %g systen are redundant to each other in providing post accident cooling of the containment atmosphere. However, the containment spray system also provides a mechanism for removing iodine from the containment atmosphere and therefore the time requirements for restoring an inoperable spray system to OPERABLE status have been maintained consistent with those assigned other inoperable ESF equipment. ( it taken for iodine removal) The containmer s stem and the containment cool' _. redundant to each other in pro ost a
- ing of the containment atmosphere. Since no credit has . r iodine removal by the contain-ment spray system, t a e o.2 e out of service . uirements for the containme y system and containment cooling system have nterrelated usted to reflect this additional redundancy in cooling capabi .
3/4.6.2.2 SPRAY ADDITIVE SYSTEM (CPI""_F The OPERABILITY of the spray additive system ensures that sufficient Na0H is added to the containment spray in the event of a LOCA. The limits on NaOH volume and concentration ensure a pH value of between (8.5) and (11.0) for the solution recirculated within containment after a LOCA. This pH band minimizes the evolution of iodine and minimizes the effect of chloride and caustic stress corrosion on mechanical systems and components. The contained water volume limit includes an allowance for water not usable because of tank discharge line location or other physical characteristics. These assumptions are consistent with the iodine removal efficiency assumed in the accident analyses. l l W-00AL 8 3/4 6-30 MAR 151978 t
~. . -- -. . _ = _ - - ._
l l CONTAINMENT SYSTEMS BASES X4.6.2.3 CONTAINMENT COOLING SYSTEM (OPTIONAL) he OPERABILITY of the containment cooling sys;.em ensures that (1) e contai ent air temperature will be maintained within limits during n mal ; operatio and (2) adequate heat removal capacity is available whe operated in conjuncti with the containment spray systems during post-LOCA onditions. (Credit taken or iodine removal by spray systems) The containme cooling system and containmen pray' system are redundant to each other in pro ding post accident coolin f the containment atmosphere. As a result of this re dancy in cooling ca ility, the allowable out-of-service time requirements or the containm t cooling system have been appropriately adjusted. Ho er, the owable out-of-service time require-ments for the containment spra ys), . have been maintained consistent with that assigned other inoperable E equipment 'since the containment spray system also provides a mechanism for movi iodine from the containment atmosphere. (No credit taken for to 'ne removal by spr systems) The containe t cooling system and the conto' ment spray system are redundant to wh other in providing post accidert c ing of the containment atmosphere Since no credit has been taken for iodine r val by the contain-ment spr system, the allowable out-of-service time require. ts for the cont ment spray system and containment spray system have been errelated an adjusted to reflect this additional redundancy in cooling capaci 4#1.JL3 IODINE REMOVAL SYSTEM (OPTIONAL) - The OPERABIL .he containment iodine filter ensures that sufficient iodine removal c 'lity will be a e in the event of a LOCA. The reduction in containment iodin y reduces the resulting site boundary radiation doses associ . with inment leakage. Cumulative operation of the system w he heaters on for rs over a 31-day period is sufficie reduce the buildup of moisture on . dsorbers and HEPA fil . . The operation of this system and resultant io i moval capac are consistent with the assumptions used in the LOCA analyses. 3/4.6. CONTAINMENT ISOLATION VALVES The OPERABILITY of the containment isolation valves ensures that the containment atmosphere will be isolated from the outside environment in the event of a release of radioactive material to the cont'ainment atmosphere or pressurization of the containment. Containment isolation within the time limits specified ensures that the release of radioactive material to the environment will be consistent with the assumptions used in the analyses for a LOCA. W-DUAL B 3/4 6-40 MAY 151980
CONTAINMENT SYSTEMS BASES 4 3/4.6.'E COMBUSTIBLE GAS CONTROL The OPERABILITY of the equipment and systems required for the detection and control of hydrogen gas ensures that this equipment will be available to maintain the hydrogen concentration within containment below its flammable limit during post-LOCA conditions. Either recombiner unit (or the purge system) is capable of controlling the expected hydrogen generation associated with (1) zirconium-water reactions, (2) radiolytic decomposition of water, and (3) corrosion of metals within containment. (Cumulative operation of the purge system with the heaters on for 10 hours over a 31 day period is sufficient to reduce the buildup of moisture on the adsorbers and HEPA filters.) These hydrogen control systems are consistent with the recommendations of Regulatory Guide 1.7,." Control of Combustible Gas Concentrations in Containment Following a LOCA," March 1971. The hydrogen mixing systems are provided to ensure adequate mixing of the containment atmosphere following a LOCA. This mixing action will prevent localized accumulations of hydrogen from excee' ding the flammable limit. t%6. 6 PENETRATION ROOM EXHAUST AIR FILTRATION SYSTEM (OPTIONAL) The OP TY of the penetration room exhaust s sures that radioactive materia 'ing from the contain a osphere through containment penetrations following a L v '+ rior tc reaching the environment. Cumulative operation of the -
- m. with +ers on for 10 hours over a 31-day period is suffic o reduce the buildup ot e on the adsorbers and HEPA fi s. The operation of this system and the res effect on offe osage calculations was assumed in the LOCA analyses.
O, c ] _ VACUUM RELIEF VALVES (OPTIONAL) The OPERABILITY of ' ontainment sphere vacuum relief valves ensures that the containment assure differential does not This conditic become more negative than i. essary to prevent e reeding the
- nt design limit for internal pressure ai 3/4.6. -5EC0"D^RV CONTAINMENT ENctosuRE Bu/LD/b 3/4.6.h #ENTILATION SY5 TEM CoVTA/A>ME/VTggugr 4g6gl44Y Mf
( m6n-a t cAdesure emu 9edy_ oGJ 4J coo ener, g) The OPERABILITY of the +Me4d-buHdmg-vanM4aMonNsystema ensures that l' 9 containment vessel leakage occurring during LOCA conditions into the annulus will be filtered through the HEPA filters and charcoal adsorber trains prior to discharge to the atmosphere. Cumulative operation of the system mith-the
.hcaten aHf,.10 how3 over a 31-day period is sufficient to reduce the buildup of moisture on the adsorbers and HEPA filters. This requirement is necessary to meet the assumptions used in the accident analyses and limit the site boundary radiation doses to within the limits of 10 CFR 100 during LOCA conditions.
MS.Y 15 1900 W-DUAL B 3/4 6-50
E l CONTAINMENT SYSTEMS BASES 5.t
- 6 " C '85" "8' 8 " "^#'~/
3/4. 6.1B.1 CONTAINMENTNINTEGRITY t e,ocesune 8(tiLDr^16- I j {ccendcry-CONTAINMEN1VINTEGRITY ensures that the release of radioactive materials from the primary containment atmosphere will be restricted to those leakage paths and associated leak rates assumed in the accident analyses, i This restriction, in conjunction with operation of the secondary containment ventilation system, will limit the site boundary radiation doses to within the limits of 10 CFR 100 during accident conditions. t couxswncm- GNewsuae_; 3/4.6. V 3 -SPIELD BUILDING STRUCTURAL INTEGRITY t This limitation ensures that the structural integrity of the containment endosum
;Mc?d building will be maintained comparable.to the original design standards for the life of the facility. Structural integrity is required to provide 1) protection for the steel vessel from external
- missiles, 2) radiation shielding in the event of a LOCA, and 3) an annulus surrounding the steel vessel that can be maintained at a negative pressure during accident conditions. A visual inspection is sufficient to demonstrate this capability.
l
. t l
l l I ! i l l l l W-0UAL B 3/4 6-6D MAR 151978 l
\
r 3/4.7 PLANT SYSTEMS BASES , 3/4.7.1 TURBINE CYCLE 3/4.7.1.1 SAFETY VALVES The OPERABILITY of the ain steam line code safety valves nsures that the secondary system pressur will be limited to within 110% ( psig) of its design pressure of (40 ps10 during the most severe anticipated system operational transient. The maximum relievin0 capacity is associated with a turbine trip from 100% RATED THERMAL POWER coincident with an assumed loss of condenser heat sink (i.e., no steam bypass to the condenser). The specified valve lift settings and relieving capacities are in accordance with the requirements of Sectiori III of the ASME Boiler and Pressure Code, 1971 Edition._ The total relieving capacity for all valves o all of the steam lines is ( *) lbs/hr which is (!!0) percent of the total M m r/0 9 secondary steam flow of ( g) ids /hr at 100% RATED THERMAL POWtM. A minimum ^or 2 OPERABLE safety valves per steam generator ensures that sufficient relieving capacity is available for the allowable THERMAL POWER restriction in Table 3.7-2. STARTUP and/or POWER OPERATION is allowable with safety valves inoperable within the limitations of the ACTION requirements on the basis of the reduction in secondary system steam flow and THER!iAL POWER required by the reduced reactor trip settings of the Pcwer Range Neutron Flux channels. The reactor trip setpoint reductions are derived on the following bases: For loop operation SP = (X) X- (Y)(V) x (109)
!cnN Lacop-operet4en-
.sp = 0 M~w-(46-)
-X---
Where: l SP = Reduced reactor trip setpoint in percent of RATED THERMAL POWER l V = Maximum number of inoperable safety valves per steam line MNa44 mum-number-of-i n op e rab l e-sa f e ty-v alv e s - pe r--ope re t tng-
-steam-line-W-STS B 3/4 7-1 NOV 151977 L
~
c f PLANT SYSTEMS BASES SAFETY VALVES (Continued)
= H (109) Power Range Neutron Flux-High Trip Setpoint for (h) loop operation
(%-) = 42xi m percent of o^TED THERMAL POWER pcraissibic by-
-P-C Satpcint-fc.- (N--1) locp cpcration, -
X = Total relieving capacity of all safety valves per steam line in lbs/ hour Y = Maximum relieving capacity of any one safety valve in ' lbs/ hour
- EHEMEMC.Y .
3/4. 7.1. 2 ".L'XILI APY FEEDWATER SYSTEM cmeqcu.y The OPERABILITY of the a n ! 12ry feedwater system ensures that the Reactor Coolant System can be cooled down to less than (350)"F from normal operating conditions in the event of a total loss of offsite power. b electric driven *2N@[8j feedwater pump is capable of delivering ufues a total feedwater flow of (450-) gpm at a presge of (4444) psig to the entrance will be. of the steam generators. The steam driven 2r"$$ feedwater pump is capable upphed of delivering a total feedwater flow of (MG) gpm at a pressure of (444ii.) psig O' M to the entrance of the steam generators. This capacity is sufficient to O' ensure that adequate feedwater flow is available to remove decay heat and reduce the Reactor Coolant System temperature to less than (350)*F when the Residual Heat Removal System may be placed into operation. 3/4.7.1.3 CONDENSATE STORAGE TANK The OPERABILITY of the condensate storage tank with the minimum water volume ensures that sufficient water is available to maintain the RCS at HOT STANDBY conditions for (l# hours with steam discharge to the atmosphere concurrent with total loss of offsite power. The contained water volume limit includes an allowance for water not usable because of tank discharge line location or other physical characteristics. 3/4.7.1.4 ACTIVITY The limitations on secondary system specific activity ensure that the resultant offsite radiation dose will be limited to a small fraction of 10 CFR Part 100 limits in the event of a steam line rupture. This dose also includes the ef fects of a coincident 1.0 GPM primary to secondary tube leak in the steam generator of the affected steam line. These values are consistent with the assumptions used in the accident analyses. W-STS B 3/4 7-2 NOV 2 01980
I PLANT SYSTEMS . i BASES 3/4.7.1.5 MAIN STEAM LINE ISOLATION VALVES The OPERABILITY of the main steam line isolation valves ensures that no ! more than one steam generator will blowdown in the event of a steam line ' rupture. This restriction is required to 1) minimize the positive reactivity effects of the Reactor Coolant System cooldown associated with the blowdown, and 2) limit the pressure rise within containment in the event the steam line rupture occurs within containment. The OPERABILITY of the main steam isolation valves within the closure times of the surveillance requirements are consistent with the assumptions used in the accident analyses. 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION The limitation on steam generator pressure and temperature ensures that the pressure induced stresses in the steam generators do not exceed the maximum allowable fracture toughness stress limits. The limitations of (70) F and (200) psig are based on a steam generator RT I IM') F and are sufficient to' prevent brittle fracture. NDT AAMg 3/4 COMPONENT COOLING WATER SYSTEM The OPERABILITY of the component cooling water system ensures that sufficient cooling capacity is available for continued operation of safety related equipment during normal and accident conditions. The redundant cooling capacity of this system, assuming a single failure, is consistent with the assumptions used in the accident analyses. 3/4.7.4 SERVICE WATER SYSTEM The OPERABILITY of the service water system ensures that sufficient cooling capacity is available for continued operation of safety related equipment during normal and accident conditions. The redundant cooling capacity of this system, assuming a single failure, is consistent with the assumptions used in the accident conditions within acceptable limits. 3/4.7.5 ULTIMATE HEAT SINK-(OPTIO.P.L)- The limitations on the ultimate heat sink level and temperature ensure that sufficient cooling capacity is available to either 1) provide normal cooldown of the facility, or 2) to mitigate the effects of accident conditions within acceptable limits. W-STS B 3/4 7-3 MN 15 two
PLANT SYSTEMS 6 BASES ULTIMATE HEAT SINK (Continued) t % -e aa deal d.<H coothq owev l j h leder _The limitations onVminimum water level and maximum temperature are based on providing MX-day cooling water supply to safety related equipment without exceeding their design basis temperature and is consistent with the recommend-ations of Regulatory Guide 1.27, " Ultimate Heat Sink for Nuclear Plants," March 1974. . H/4. 7 " Ft000 PROTECTIOJ (OPTIONAL) _ The limi etion ensures sty protective actions will be taken (and operation w erminated) in the event of flood h conditions. The limit of elev ) Mean 1 is based on the maximum elevation at which f rlood control measures prov1 on to
+
safety-ra - uipment. 6 Mhketse 3/4. 7.% CONTROL ROOM OiERCENCY AIR -C+EAN@ SYSTEM rma h.cu air h I The OPERABILITY of the control room w ti h pfor system ensures that 1)
. O the ambient air temperature does not exceed the allowable temperature for continuous duty rating for the equipment and instrumentation cooled by this system and 2) the control room will remain habitable for operations personnel during and following all credible accident conditions. Cumulativ: ^p0 rat!0r
.^ f-t" g"te- ith the hcater cr-fer-10-hours gee, o Cl-day period is sufficier.t-
~
- te reduce the-bui4<iup cf meisture--on-the adserters cr.d-HEPA fil tern The OPERABILITY of this system in conjunction with control room design provisions is based on limiting the radiation exposure to personnel occupying the control room to 5 rem or less whole body, or its equivalent. This limitation is consistent with the requirements of General Design Criteria 19 of Appendix "A",
10 CFR 50. _ 8 ECCS PUMP R00'i EXHAUST AIR FILTRATION SYSTEM The OPERABIL ECCS pump room exhaust air filtra .,ystem ensures that radioactive mater leaking from the quipment within the pump room following a LOCA are filtere reaching the environment, Cumulative operation of the s s e hea . for 10 hours over a 31 h cay period is sufficio reduce the buildup of moistur e adsorbers and HEPA filt e operation of this system and the resultant n of age calculations was assumed in the accident analyses. W-STS B 3/4 7-4 MAY I 51HO
PLAtlT SYSTEMS BASES 7 3/4.7.h SNUB 8ERS All snubbers are required OPERA 8LE to ensure that the structural integrity of the Reactor Coolant System and all other safety-related systems is main-tained during and following a seismic or other event initiating dynamic loads. Snubbers are classified and grouped by design and manufacturer but not by size. For example, mechanical snubbers utilizing the same design features of the 2-kip, 10-kip and 100-kip capacity manufactured by Company "A" are of the same type. The same design mechanical snubbers manufactured by Company "B" for the purposes of this Technical Specification would be of a different type, as would hydraulic snubbers from either manufacturer.
~
A list of individual snubbers with detailed information of snubber location and size and of system affected shall be available at the plant in accordance with Section 50.71(c) of 10 CFR Part 50. g accessibility of each snubber shall be determined and approved by the w.. . k, iew a cug]. The determination shall be based upon the existing radiation levels and the expected time to perform a visual inspection in each snubber location as well as other factors associated with accessibility during plant operations (e.g., temperature, i l atmosphere, location, etc.), and the recommendations of Requiatory Cuides 8.8 and 8.10. The addition or deletion of any hydraulic or mechanical snubber shall be made in accordance with Section 50.59 of 10 CFR Part 50. The visual inspection frequency is based upon maintaining a constant level of snubber protection to each safety related system during an earthquake or severe transient. Therefcre, the required inspection interval varies inversely with the observed snubber failures on a given system and is determined l by the number of inoperable snubbers found during an inspection of each system. In order to establish the inspection frequency for each type of snubber on a safety-related system, it was assumed that the frequency of snubber failures and initiating events is constant with time and that the failure of any snubber on that system could cause the system to be unprotected and to result in failure during an assumed initiating event. Inspections performed before that interval has elapsed may be used as a new reference point to determine the next inspection. However, the results of such early inspections performed before the original required time interval has elaosed (nominal time less 25%) may not be used to lengthen the required inspec'icn interval. Any inspection whose results require l ' a shorter inspection interval will override the previous senedule. l l The acceptance criteria are to be used in the visual inspection to determine l OPERABILITY of the snubbers. For example, if a fluid port of a hydraulic snubber is found to be uncovered, the snubber shall be declared inoperable and shall not be determined OPERABLE via functional testing. 1
- To provide assurance of snubber functional reliability, one of three '
functional testing methods is used with the stated acceptance criteria: I y-STS 8 3/4 7-5
PLANT SYSTEMS BASES SNUB 8ERS (Continued) l
- 1. Functionally test 10% of a type of snubber with an additional 10% tested for each functional testing failure, or
- 2. Functionally test a sample size and determine sample acceptance or rejectionrunng-F-igure-4J -l;-or 3.
Functionally test a representative sample size and determine sample acceptance or rejection using the stated equation. Figure 4.7-1 was developed using "Wald's Sequential Probability Ratio 7 Plan" as described in " Quality Control.and Industrial Statistics" by l Acheson J. Duncan. i Permanent or other exemptions from the surveillance program for individual l snubbers may be granted by the Commission if a justifiable basis for exemption i is presented and, if applicable, snubber life destructive testing was performed to qualify the snubbers for the applicable design conditions at either the com-pletion of their fabrication or at a subsequent date. Snubbers so exempted shall be listed in the list of individual snubbers indicating the extent of the exemptions. The service life of a snubber is established via manufacturer input and information through consideration of the snubber service conditions and associated installation and maintenance records (newly installed snubbers, seal replaced, spring replaced, in high radiation area, in high temperature area, etc.). The requirement to monitor the snubber service life is included to ensure that the snubbers periodically undergo a performance evaluation in view of their age and operating conditions. These records will provide statistical f. t bases for future consideration of snubber service life. i
'f- STS B 3/4 7-6
P_LANT SYSTEMS BASES 3/4.7. SEALED SOURCE CONTAMINATION The limitations on removable contamination for sources requiring leak testing, including alpha emitters, is based on 10 CFR 70.39(c) limits for plutonium. This limitation will ensure that leakage from byproduct, source, and special nuclear material sources will not exceed allowable intake values. Sealed sources are classified into three groups according to their use, with surveillance requirements commensurate with the probability of damage to a
- source in that group. Those sources which are frequently handled are required to be tested more often than those which are not. Sealed sources which are continuously enclosed within a shielded mechanism (i.e., sealed sources within radiation monitoring or boron measuring devices) are considered to be stored and need not be tested unless they are removed from the shielded mechanism.
3/4.7. FIRE SUPPRESSION SYSTEMS The OPERABILITY of the fire suppression systems ensures that adequate fire suppression capability is available to confine and extinguish fires occurring in any portion of the facility where safety-related equipment is located. The fire suppression system consists of the water system, spray, and/or sprinklers, CO , lMun, fire hose stations, and yard fire hydrants.
.Thecollectivecapabi$ityofthefiresuppressionsystemsisadequateto minimize potential damage to safety-related equipment and is a major element in the facility fire protection program.
In the event that portions of the fire suppression systems are inoperable, alternate backup fire-fighting equipment is required to be made available in the affected areas until the inoperable equipment is restored to service. When the inoperable fire-fighting equipment is intended for use as a backup means of fire suppression, a longer period of time is allowed to provide an alternate means of fire fighting than if the inoperable equipment is the primary means of fire suppression.
~l W-STS B 3/4 7-Ye fiOV 2 01980
PLANT SYSTEMS BASES FIRE SUPPRESSION SYSTEMS (Continued) The surveillance requirements provide assurance that the minimum OPERABILITY requirements of the fire suppression systems are met. -An cllcwanee n-made--for-ensttr4n<r-c cuf f 4 c 4 ant vohme cf "eivn i n i.he Hahn-storage-tanks-.
-by--ver+fying-e+ther-the-weight-or-the-hvel-of-the-tanks,-4+vcl mecsur4 ment s -
are-made-by-either 2 II f ru- E M_ --approved :cthedr In the event the fire suppression water system becomes inoperable, immediate corrective measures must be taken since this system provides the major fire suppression capability of the plant. The requirement fcr c
- tucr.ty-fcur hcur repcrt te the Cer-iccier, pre, ides for--prcept coeluctier of th; cccep+=h414ty cf the ccrrective mccsures tc prc'.ide edcquate firc - -cupprcccicr cepcbi'ity for the ccntinucd pr0+act_4rn-of-the suc1;cr p1;nt, 10 ^
3/4.7.12 FIRE RATED ASSEMBLIES The OPERABILITY of the fire barriers and barrier penetrations ensure that fire damage will be limited. These design features minimize the possibility of a single fire involving more than one fire area prior to detection and i extinguishment. The fire barriers, fire barrier penetrations for conduits, cable trays and piping, fire windows, fire dampers, and fire doors are periodically inspected to verify their OPERABILITY.
\\
3/4.7.14 AREA TEMPERATURE MONITORING The area temperature limitations ensure that safety-related equipment will not be subjected to temperatures in excess of their environmental qualification temperatures. Exposure to excessive temperatures may degrade equipment and can cause a loss of its OPERABILITY. The temperature limits include an allowance for instrument error of Gtk)'F. 8 W-STS B 3/4 7'% NOV 2 1931
3/4.8 ELECTRICAL POWER SYSTEMS BASES 3/4.8.1, 3/4.8.2 AND 3/4.8.3 A.C. SOURCES, D.C. SOURCES AND ONSITE POWER DISTRIBUTION SYSTEMS The OPERABILITY of the A.C. and D.C power sources and associated distribution systems during operation ensures that sufficient power will be available to supply the safety related equipment required for 1) the safe shutdown of the facility and 2) the mitigation and control of accident condi-tions within the facility. The minimum specified independent and redundant A.C. and D.C. power sources and distribution systems satisfy the requirements of General Design Criterion 17 of Appendix "A" to 10 CFR 50. The ACTION requirements specified for the levels of degradation of the power sources provide restriction upon continued facility operation commensurate with the level of degradation. The OPERABILITY of the power sources are consistent with the initial condition assumptions of the safety analyses and are based upon maintaining at least one redundant set of onsite A.C. and D.C. power sources and associated distribution systems OPERABLE during accident conditions coincident with an assumed loss of offsite power and single failure < of the other onsite A.C. source. The A.C. and D.C. source allowable out-of-service times are based on Regulatory Guide 1.93, " Availability of Electrical Power Sources", December 1974. When one diesel generator is inoperable, there is an additional ACTION requirement to verify that all required systems, subsystems, trains, components and devices, that depend on the remaining OPERABLE diesel geneFator as a g ce of emergency power, are also OPERABLE, and that the steam-driven E!"ur., feedwater pump is OPERABLE. This require-ment is intended to provide assurance that a loss of offsite power event will not result in a complete loss of safety function of critical systems during the period one of the diesel generators is inoperable. The term verify as used in this context means to administratively check by examining logs or other information to determine if certain components are out-of-service for maintenance or other reasons. It does not mean to perform the surveillance requirements needed to demonstrate the OPERABILITY of the component. The OPERABILITY of the minimum specified A.C. and D.C. power sources and associated distribution systems during shutdown and refueling ensures that
- 1) the facility can be maintained in the shutdown or refueling condition for extended time periods and 2) sufficient instrumentation and control capability is available for monitoring and maintaining the unit status.
The Surveillance Requirements for demonstrating the OPERABILITY of the diesel generators are in accordance with the recommendations of Regulatory Guides 1.9, " Selection of Diesel Generator Set Capacity for Standby Power Supplies", March 10,1971,1.108, " Periodic Testing of Diesel Generator Units ) Used as Onsite Electric Power Systems at Nuclear Power Plants", Revision 1, l August 1977, and 1.137, " Fuel-Oil Systems for Standby Diesel Generators", l Revision 1, October 1979. W-STS S 3/4 8-1 JUL 2 7 1981
ELECTRIC POWER SYSTEMS BASES A.C. SOURCES, D.C. SOURCES AND ONSITE POWER DISTRIBUTION SYSTEMS (Continued) See Ensed I!r QH The Surveillance Requirement for demonstrating the OPERABILITY of the Station batteries are based on the recommendations of Regulatory Guide 1.129,
" Maintenance Testing and Replacement of Large Lead Storage Batteries for '
Nuclear Power Plants," February 1978, and IEEE Std 450-1980, "IEEE Recommended Practice for Maintenance, Testing, and Replacement of Large Lead Storage Batteries for Generating Stations and Substations." Se c Tns ent I (D Verifying average electrolyte temperature above the minimum for which the batterywassized,totalbatteryterminalvoltageonfloatcharge, connection resistance values and the performance of battery service and discharge tests ensures the effectiveness of the charging system, the ability to handle high discharge rates and compares the battery capacity at that time with the rated capacity. Table 4.8-2 specifies the normal limits for each designated pilot cell and each connected cell for electrolyte level, float voltage and specific gravity. The limits for the designated pilot cells float voltage and specific gravity, greater than 2.13 volts and .015 below the manufacturer's full charge specific gravity or a battery charger current that had stabilized at a low value, is characteristic of a charged cell with adequate capacity. The normal limits for each connected cell for float voltage and specific gravity, greater than 2.13 volts and not more than .020 below the manufacturer's full charge specific gravity with an average specific gravity of all the connected cells not more than .010 below the manufacturer's full charge specific gravity, ensures the OPERABILITY and capability of the battery. Operation with a battery cell's parameter outside the normal limit but within the allowable value specified in Table'4.8-2 is permitted for up to 7 days. During this 7 day period: (1) the allowable values for electrolyte level ensures no physical damage to the plates with an adequate electron transfer capability; (2) the allowable value for the average specific gravity of all the cells, not more than .020 below the manufacturer's recommended full charge specific gravity, ensures that the decrease in rating will be less than the safety margin provided in sizing; (3) the allowable value for an individual cell's specific gravity, ensures.that an individual cell's specific gravity will not be more than .040 below the manufacturer's full charge specific i gravity and that the overall capability of the battery will be maintained j within an acceptable limit; and (4) the allowable value for an individual j cell's float voltage, greater than 2.07 volts, ensures the battery's capability l to perform its design function.
]
W-STS B 3/4 8-2 M 27 E 1
i i INSERT IV The following is a clarification to the Regulatory Guide 1.108 requirement pertaining to the functional capability of the diesel generator at full load temperature conditions: The functional capability of the diesel generator at full load temperature is demonstrated by manually starting the diesel within 5 minutes after completion of testing which brought the diesel to full load temperature conditions and verifying that the required voltage and frequency are automatically attained within acceptable limits and time. Furthermore, the diesel generator is manually loaded to the continuous rating and is operated for at least 5 minutes. There is no need for automatic starting of the diesel by simulating loss of AC or for automatic load sequencing. The I fact that the diesel generator is at full-load temperature conditions has no effect either on the circuitry which starts the diesel generator upon loss of AC or on the circuitry which controls the diesel generator upon loss of AC or on the circuitry which controls the loading sequence. Functional capability of those circuits is verified elsewhere in the surveillance program. 1 . 1 4 l k f'
INSERT V The minimum requirement for one battery per train is acceptable because bus ties are intentionally provided between the two buses of each train so that either battery can serve as the source of power for the combined load. S
ELECTRICAL POWER SYSTEMS BASES 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES TN i 0" Containment electrical penetrations and penetration conductors are protected by either deenergizing circuits not required during reactor operation or by demonstrating the OPERABILITY of primary and backup overcurrent protection circuit breakers during periodic surveillance. N feThe :: surveillance requirements applicable to lower voltage circuit breakers,f-- provide assurance of breaker %ed-fuse reliability by testing at least one and/r representative sample of each manufacturer's brand of circuit breaker, p fuse. Each manufacturer's molded case and metal case circuit breakers
-end hr fuses are grouped into representative samples which are then tested on,(
a rotating basis to ensure that all breaker,s If a wide variety exists within any manufacturer' sfeer'+r brandtuses are breakers of circuit tested. *:nd/cm fuses, it is necessary to divide that manufacturer's breakercecnc'cr tuses into groups and treat each group as a separate type of breakergee fuses for surveillance purposes. The OPERABILITY of the motor operated valves thermal overload protection
-end bg a;; de'Mcm ensures that these devices will not prevent safety-related valves from performing their function. The Surveillance Requirements for demonstrating the OPERABILITY of these devices are in accordance with Regulatory Guide 1.106, " Thermal Overload Protection for Electric Motors on Motor Operated Valves," Revision 1, March 1977.
SEP 151981 W-STS B 3/4 8-3
3/4.9 REFUELING OPERATIONS BASES . 3/4.9.1 BORON CONCENTRATION The limitations on reactivity conditions during REFUELING ensure that:
- 1) the reactor will remain subcritical during CORE ALTERATIONS, and 2) a uniform boron concentration is maintained for reactivity control in the water volume having direct access to the reactor vessel. These limitations are consistent the accidentwith the initial conditions assumed for the boron dilution incident in analyses. The value of 0.95 or less for Keff includes a 1 percent delta k/k conservative allowance for uncertainties. Similarly, the boron concentration value of (2000) ppm or greater includes a conservative uncertainty allowance of 50 ppm boron.
3/4.9.2 INSTRUMENTATION ' The OPERABILITY of the source range neutron flux monitors ensures that redundant monitoring capability is available to detect changes in the reactivity condition of the core. 3/4.9.3 DECAY TIME The minimum requirement for reactor subcriticality prior to movement of irradiated fuel assemblies in the reactor pressure vessel ensures that sufficient time has elapsed to allow the radioactive decay of the short-lived fission products. This decay time is consistent with the assumptions used in the accident analyses. 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS The requirements on containment building penetration closure and OPERABILITY ensure that a release of radioactive material within containment will be restricted from leakage to the environment. The OPERABILITY and closure restrictions are sufficient to restrict radioactive material release from a fuel element rupture based upon the lack of containment pressurization potential while in the REFUELING MODE. 3/4.9.5 COMMUNICATIONS The requirement for communications capability ensures that refueling station personnel can be promptly informed of significant changes in.the facility status or core reactivity conditions during CORE ALTERATIONS. W-STS B 3/4 9-1 SEP 151981
REFUELING OPERATIONS BASES ReFL4ELIMG NAC HIME 3/4.9.6 -MANIPULATOR CRANE
- k rchehm ~cMAC veggh p adth The'0PERACILITY requirements for the reampumtcr cranes = ensure that:
- 1) _ rip"hter ceanes will be used for movement of drive rods and fuel assemblies, g ) each'%rane has sufficient load capacity to lift a drive rod or fuel assembly, and 3) the core internals and pressure vessel are protected from excessive lifting force in the event they are inadvertently engaged during lifting operations.
3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE BUILDING The restriction on movement of loads in excess of the nominal weight of a fuel and control rod assembly and associated handling tool over other fuel assemblies in the storage pool ensures that.in the event this load is dropped
- 1) the activity release will be limited to that contained in a single fuel rac.
assembly, and 2) 6..y possible distortion of fuel in the storage-rock %. rill not result in a critical array. This assumption is consistent with the activity release assumed in the accident analyses. 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION The requirement that at least one residual heat removal (RHR) loop be in operation ensures that (1) sufficient cooling capacity is available to remove decay heat and maintain the water in the reactor pressure vessel below 140 F as required during the REFUELING MODE, and (2) sufficient coolant circulation is maintained through the re' actor core to minimize the effect of a boron dilution incident and prevent boron stratification. The requirement to have two RHR loops OPERABLE when there is less than 23 feet of water above the reactor pressure vessel flange ensures that a single failure of the operating RHR loop will not result in a complete loss of residual heat removal capability. With the- reactor vessel head removed and 23 feet of water above the reactor pressure vessel flange, a large heat sink is available for core cooling. Thus, in the event of a failure of the operating RHR loop, adequate time is provided to initiate emergency procedures to cool the core. 3/4.9.9 CONTAINMENT PURGE AND EXHAUST ISOLATION SYSTEM The OPERABILITY of this system ensures that the containment vent and purge penetrations will be automatically isolated upon detection of high radiation levels within the containment. The OPERABILITY of this system is required to restrict the release of radioactive material from the containment atmosphere to the environment. W-STS B 3/4 9-2 SEP i 51981
REFUELING OPERATIONS BASES 3/4.9.10 and 3/4.9.11 WATER LEVEL - REACTOR VESSEL and STORAGE POOL The restrictions on minimum water level ensure that sufficient water depth is available to remove 99% of the assumed 10% iodine gap activity released from the ruptu'e r of an irradiated fuel assembly. The minimum water depth is consistent with the assumptions of the accident analysis. S T O R A G C 6tA L LDING. enexc.e,v. ,, fa g c.,.,_, m F t4 E t,. t s 3, r,y 3/4.9.12 -STOP, ACE POOL VENTILATION SYSTEM W t cm,I d.. m e be t6 *ceY aA" h M * * ' I The limitations on the ;torage poc: vent 1!atler syctesVensure that all radioactive material released from an irradiated fuel assembly will be filtered through the HEPA filters and charcoal adsorber prior to discharge to the atmos-phare. Cumulative operation of.the system with the heaters on for 10 hours over a 31-day period is sufficient to reduce the buildup of moisture on the adsorbers and HEPA filters. The OPERABILITY of this system and the resulting iodine removal capacity are consistent with the assumptions of the accident analyses. W-STS B 3/4 9-3 SEP 15 I981
, 3/4.10 SPECIAL TEST EXCEPTIONS BASES 3/4.10.1 SHUTDOWN MARGIN This special test exception provides that a minimum amount of control rod
. worth is immediately available for reactivity control when tests are performed for control rod worth measurement. This special test exception is required to permit the periodic verification of the actual versus predicted core reactivity condition occuring as a result of fuel burnup or fuel cycling operations.
3/4.10.2 GROUP HEIGHT, INSERTION, AND POWER DISTRIBUTION LIMITS This special test exception permits individual control rods to be positioned outside of their normal group heights and i.nsertion limits during the performance of such PHYSICS TESTS as those required to 1) measure control rod worth and 2) determine the reactor stability index and damping factor under xenon oscillation conditions. 3/4.10.3 PHYSICS TESTS This special test exception permits PHYSICS TESTS to be performed at less than or equal to 5% of RATED THERMAL POWER with the RCS T slightly lower than normally allowed so that the fundamental nuclear chaMteristics of the reactor core and related instrumentation can be verified. In order for various characteristics to be accurately measured, it is at times necessary to operate outside the normal restrictions of these Technical Specifications. For instance, to measure the moderator temperature coefficient at 80L, it is necessary to Q position the various control rods at heights whigh may not normally be allowed by Specification 3.1.3.6 "ich " tur ry c2Nhe RCS T tc fall slightly nnay b e, below the minimum temperature of Specification 3.1.1.4 3.ukthe measa,e-ed. 3/4.10.4 REACTOR CCOLANT LOOPS This special test exception permits reactor criticality under no flow conditions and is required to perform certain startup and PHYSICS TESTS while at low THERMAL POWER levels. 3/4.10.5 POSITION INDICATION SYSTEM-SHUTOOWN l This special test exception permits the position indication systems to be inoperable during rod drop time measurements. The exception is required since l the data necessary to determine the rod drop time is derived from the induced ! ; voltage in the position indicator coils as the rod is dropped. This induced voltage is small compared to the normal voltage and, therefore, can .:ot be observed if the position indication systems remain OPERABLE. W-STS B 3/4 10-1 DEC 51979
I i 3/4.11 RADI0 ACTIVE EFFLUENTS @ BASES 3/4.11.1 LIQUID EFFLUENTS
~
3/4.11.1.1 CONCENTRATION This specification is provided to ensure that the concentration of radioactive materials released in liquid waste effluents to UNRESTRICTED AREAS will be less than the concentration levels specified in 10 CFR Part 20, Appen-dix B, Table II, Column 2. This limitation provides additional assurance that the levels of radioactive materials in bodies of water in UNRESTRICTED AREAS will result in exposures within (1) the Section II.A design objectives of Appen-dix I, 10 CFR Part 50, to a MEMBER OF THE PUBLIC and (2) the limits of 10 CFR Part 20.106(e) to the population. The concentration limit for dissolved or entrained noble gases is based upon the assumption that Xe-135 is the control-ling radioisotope and its MPC in air (submersion) was converted to an equivalent concentration in water using the methods described in International Commission on Radiological Protection (ICRP) Publication 2. This specification applies to the release of radioactive materials in liquid effluents from all reactor units at the site. The required detection capabilities for radioactive materials in liquid waste samples are tabulated in terms of the lower limits of detection (LLDs). Detailed discussion of the LLD, and other detection limits can be found in HASL Procedures Manual, HASL-300 (revised annually), Currie, L. A. , ." Limits for Qualitative Detection and Quantitative Determination - Application to Radiochemistry," Anal. Chem. 40, 586-93 (1968), and Hartwell, J. K., " Detection Limits for Radioanalytical Counting Techniques," Atlantic Richfield Hanford Company Report ARH-SA-215 (June 1975). 3/4.11.1.2 DOSE This specification is provided to implement the requirements of Sections II.A, III.A and IV.A of Appendix ~I, 10 CFR Part 50. The Limiting Condition for Operation implements the guides set forth in Section II. A of Appendix I. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appen-dix I to assure that the releases of radioactive material in liquid effluents to UNRESTRICTED AREAS will be kept "as low as is reasonably achievable." Also, + fresh "atea sites with dri ding-water supplic:, th & ccr be potenti:lly-
-Offected-by plant oparatient, thereisr-easonableaccurancethattheoperation-h nr m rw-Clity will nnt racnit 4a radionue4+de cancentraticas in the finished
--dMn44ng-water that are 4a exc+se nf the-requisemenia cf t0 CFo Part 141. The dose calculation methodology and parameters in the ODCM implement the require-ments in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data, such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The equations specified in the ODCM for calculating the doses due to the actual release rates of radioactive materials in liquid effluents are consistent with the methodology provided in PWR-STS-RETS B 3/4 11-1 1/4/83
RADI0 ACTIVE EFFLUENTS h BASES Regulatory Guide 1.109, " Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," Revision 1, October 1977 and Regulatory Guide 1.113, " Estimating Aquatic Dispersion of Effluents from Accidental and Routine Reactor Releases for the Purpose of Implementing Appendix I," April 1977. This specification applies to the release of radioactive matericls in liquid effluents from each reactor at the site. For units with shared radwaste treatment systems, the liquid effluents from the shared system are proportioned among the units sharing that system. 3/4.11.1.3 LIOUID RADWASTE TREATMENT SYSTEM The OPERABILITY of the liquid radwaste treatment system ensures that this system will be available for use whenever liquid effluents require treatment prior to release to the environment. The requirement that the appropriate portions of this system be used when specified provides assurance that the releases of radioactive materials in liquid effluents will be kept "as low as is reasonably achievable". This specifica' tion implements the requirements of 10 CFR Part 50.36a, General' Design Criterion 60 of Appendix A to 10 CFR Part 50 and the design objective given in Section II.D of Appendix I to 10 CFR Part 50. The specified limits governing the use of appropriate portions of the liquid radwaste treatment system were specified as a suitable fraction of the dose design objectives set forth in Section II.A of Appendix I, 10 CFR Part 50, for liquid effluents. This specification applies to the release of radioactive materials in liquid effluents from each reactor unit at the site. For units with shared radwaste treatment systems, the liquid effluents from the shared system are proportioned among the units sharing that system. 3/4.11.1.4 LIOUID HOLDUP TANKS b $nks listed " tH : Speci'icati r . include all those outdoor radwaste tanks that are not surrounded by liners, dikes, or walls capable of holding the tank contents and that do not have tank overflows and surrounding area drains connected to the liquid radwaste treatment system. Restricting the quantity of radioactive material centained in the specified tanks provides assurance that in the event of an uncontrolled release of the tank's contents, the resulting concentrations would be less than the limits of 10 CFR Part 20, Appendix B, Table II, Column 2, at the nearest potable water supply and the nearest surface water supply in an UNRESTRICT 0 AREA. T 3/4.11.2 GASEOUS EFFLUENTS
%3Md Q 3/4.11.2.1 DOSE RATE This specification is provided to ensure that the dose at any time at and beyond the SITE BOUNDARY from gaseous effluents from all unitsign the site will be within the annual dose limits of 10 CFR Part 20 to UNRESTRICTED AREAS. The annual dose limits are the doses associated with the concentrations of 10 CFR Part 20, Appendix B, Table II, Column 1. These limits provide reasonable PWR-STS-RETS B 3/4 11-2 1/4/83
1 RADIOACTIVE EFFLUENTS @ i
' BASES assurance that radioactive material discharged in gaseous effluents will not result in the exposure of a MEMBER OF THE PUBLIC.-in- r U4 RESTRICTED AREA, either within or outside the SITE BOUNDARY, to annual average concentrations exceeding the limits specified in Appendix B, Table II of 10 CFR Part 20 (10 CFR Part 20.106(b)). For MEMBERS OF THE PUBLIC who may at times be within the SITE B0UNDARY, the occupancy of that MEMBER OF THE PUBLIC will usually be sufficiently low to compensate for any increase in the atmospheric diffusion factor above that for the SITE BOUNDARY. Examples of calculations for such MEMBERS OF THE PUBLIC, with the appropriate occupancy factors, shall be given in the ODCM. The specified release rate limits restrict, at all times, the corresponding gamma and beta dose rates above background to a MEMBER OF THE PUBLIC at or beyond the SITE BOUNDARY to less than or equal to 500 mrems/ year to the total body or to less than or equal to 3000 mrems/ year to the skin.
These release rate limits also restrict, at all times, the corresponding thyroid dose rate above background to a child via the inhalation pathway to less than or equal to 1500 mrems/ year. ' This specification applies to the release of radioactive materials in gaseous effluents from all reactor units at the site. The required detection capabilities for radioactive materials in gaseous waste samples are tabulated in terms of the lower limits of detection (LLDs). Detailed discussion of.the LLD, and other detection limits.can be found in HASL Procedures Manual, HASL-300 (revised annually), Currie, L. A. , " Limits for Qualitative Detection and Quantitative Determination - Application to Radio-chemistry," Anal. Chem. 40, 586-93 (1968), and Hartwell, J. K. , " Detection Limits for Radioanalytical Counting Techniques," Atlantic Richfield Hanford Company Report ARH-SA-215 (June 1975). 3/4.11.2.2 DOSE - NOBLE GASES This specification is provided to implement the requirements of Sections II.8, III.A and IV.A of Appendix I, 10 CFR Part 50. The Limiting Condition for Operation implements the guides set forth in Section II.B of Appendix I. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I sog g assure that the releases of radioactive material in gaseous effluents-tc.at h RCSTRICTED AREAS will be kept "as low as is reasonably achievable." The Sur-w q eillance v Requirements implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational proce-dures based on models and data such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially under-estimated. The dose calculation methodology and parameters established in the ODCM for calculating the doses due to the actual release rates of radioactive noble gases in gaseous effluents are consistent with the methodology provided in Regulatory Guide 1.109, " Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," Revision 1, October 1977 and Regulatory Guide 1.111,
" Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water Cooled Reactors," Revision 1, July 1977.
PWR-STS-RETS B 3/4 11-3 1/4/83
1 I l RADICACTIVE EFFLUENTS @ BASES The ODCM equations provided for determining the air doses at and beyond the SITE BOUNDARY are based upon the historical average atmospheric conditions. This specification applies to the release of radioactive materials in gaseous effluents from each reactor unit at the site. For units with shared radwaste treatment systems, the gaseous effluents from the shared system are proportioned among the units sharing that system. 3/4.11.2.3 DOSE - 10 DINE-131, 10 DINE-133, TRITIUM, AND RADIONUCLIDES IN PARTICULATE FORM t -the S ITE BOUNDARY) This specification is provided to implement the requirements of Sections II.C, III.A and IV.A of Appendix I, 10 CFR Part 50. The Limiting } Conditions for Operation are the guides set forth in Section II.C of Appendix I. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive materials in gaseous effluents -tc U=E'iTRICTED AREAS-will be kept "as low as is reasonably achievable." The ODCM calculational methods specified in the Surveillance Requirements implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data, such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The ODCM calculational method-ology and parameters for calculating the doses due to the actual release rates of the subject materials are consistent with the methodology provided in Regula-tory Guide 1.109, " Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," Revision 1, October 1977 and Regulatory Guide 1.111, " Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors," Revision 1, July 1977. These equa-l tions also provide for determining the actual doses based upon the historical average atmospheric conditions. The release rate specifications for iodine-131, tritium, and radionuclides in particulate form with half lives greater than 8 days are dependent upon the existing radionuclide pathways to man in the areas at and beyond the SITE BOUNDARY. The pathways that were examined in the development of these calculations were: 1) individual inhalation of airborne radionuclides, 2) deposition of radionuclides onto green leafy vegetation with ! subsequent consumption by man, 3) deposition onto grassy areas where milk l animals and meat producing animals graze with consumption of the milk and meat by man, and 4) deposition on the ground with subsequent exposure of man. l This specification applies to the release of radioactive materials in gaseous effluents from each reactor unit at the site. For units with shared radwaste treatment systems, the gaseous effluents from the shared system are proportioned among the units sharing that system. 3/4.11.2.4 GASEOUS RADWASTE TREATMENT The OPERABILITY of the WASTE GAS HOLDUP SYSTEM and the VENTILATION EXHAUST TREATMENT SYSTEM ensures that the systems will be available for use whenever gaseous effluents require treatment prior to release to the environment. The requirement that the appropriate portions of these systems be used, when specified, provides reasonable assurance that the releases of radioactive PWR-STS-RETS B 3/4 11-4 1/4/83
RADI0 ACTIVE EFFLUENTS h BASES materials in gaseous effluents will be kept "as low as is reasonably achievable". This specification implements the requirements of 10 CFR Part 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50 and the design objectives given in Section II.D of Apoendix I to 10 CFR Part 50. The specified limits governing the use of appropriate portions uf the systems were specified as a suitable fraction of the dose design objectives set forth in Section II.B and II.C of Appendix I, 10 CFR Part 50, for gaseous effluents. This specification applies to the release of radioactive materials in gaseous effluents from each reactor unit at the site. For units with shared radwaste treatment systems, the gaseous effluents from the shared system are proportioned among the units sharing that system. 3/4.11.2.5 EXPLOSIVE GAS MIXTURE This specification.is provided to en.sure that the concentration of potentially explosive gas mixtures contained in the WASTE GAS HOLDUP SYSTEM is maintained below the flammability limits of hydrogen and oxygen. (Automatic control features are included in the system to prevent the hydrogen and oxygen concentrations from reaching these flammability limits. These automatic control features include isolation of the source of hydrogen and/or oxygen, automatic diversion to recombiners, or injection of dilutants to reduce the concentration below the flammability limits.) Maintaining the concentration of hydrogen and oxygen below their flammability limits provides assurance that the releases of radioactive materials will be controlled in conformance with the requirements of General Design Criterion 60 of Appendix A to 10 CFR Part 50. ~ 4/A.11.2.6 GAS STORAGE TANKS This sp '4 ation considers postulated radioactive releases due to a was gas system leak or re, and limits the quantity of radioactivity ned in each pressurized gas st e tank in the WASTE GAS HOLDUP S
~
o assure that a release would be substan lly below the guidelin 10 CFR Part 100 for a postulated event. Restricting the quantity of radioa *4 1 contained in each gas storage tank provides assurance that in t vent of an . trolled release of the tank's contents, the result ~ otal body exposure to "MBER OF THE PUBLIC at the nearest exclus' rea boundary will not exceed 0.5 This is con-sistent with St rd Review Plan 11.3, Branch Technical Positio B 11-5, "Postula adioactive Releases Due to a Waste Gas System Leak or Fai " i G-0800, July 1981. 3/4.11.3 SOLID RADI0 ACTIVE WASTE This specification implements the requirements of 10 CFR Part 50.36a and General Design Criterion 60 of Appendix A to 10 CFR Part 50. The process param-eters included in establishing the PROCESS CONTROL PROGRAM may include, but are not limited to waste type, waste pH, waste / liquid / solidification agent / catalyst ratios, waste oil content, waste principal chemical constituents, and mixing and curing times. PWR-STS-RETS B 3/4 11-5 1/4/83
RADI0 ACTIVE EFFLUENTS (() BASES 3/4.11.4 TOTAL DOSE This specification is provided to meet the dose limitations of 40 CFR Part 190 that have been incorporated into 10 CFR Part 20 by 46 FR 18525. The specification requires the preparation and submittal of a Special Report whenever the calculated doses from plant generated radioactive effluents and direct radiation exceed 25 mrems to the total body or any organ, except the thyroid, which shall be limited to less than or equal to 75 mrems. For sites containing up to 4 reactors, it is highly unlikely that the resultant dose to a MEMBER OF THE PUBLIC will exceed the dose limits of 40 CFR Part 190 if the individual reactors remain within twice the dose design objectives of Appendix I, and if direct radiation doses from the reactor units and outside storage tanks are kept small. The Special Report will describe a course of action that should result in the limitation of the annual dose to a MEMBER OF
~
THE PUBLIC to within the 40 CFR Part 190 limits. For the purposes of the Special Report, it may be assumed that the dose commitment to the MEMBER OF THE PUBLIC from other uranium fuel cycle sources is negligible, with the exception that dose contributions from other nuclear fuel cycle facilities at the same site or within a radius of 8 km must be considered. If the dose to any MEMBER OF THE PUBLIC is estimated to exceed the requirements of 40 CFR Part 190, the Special Report with a request for a variance (provided the release conditions resulting in violation of 40 CFR Part 190 have not already been corrected), in accordance with the provisions of 40 CFR Part 190.11 and 10 CFR Part 20.405c, is considered to be a timely request and fulfills the requirements of 40 CFR Part 190 until NRC staff action is completed. The variance only relates to the limits of 40 CFR Part 190, and does not apply in any way to the other requirements for dose limitation of 10 CFR Part 20, as addressed in Specifications 3.11.1.1 and 3.11.2.1. An individual is not considered a MEMBER OF THE PUBLIC during any period in which he/she is engaged in carrying out any operation that is part of the nuclear fuel cycle. O PWR-STS-RETS B 3/4 11-6 1/4/83
3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING g BASES 3/4.12.1 MONITORING PROGRAM The radiological environmental monitoring program required by this specification provides representative measurements of radiation and of radio-active materials in those exposure pathways and for those radionuclides that lead to the highest potential radiation exposures of MEMBERS OF THE PUBLIC resulting frcm the plant operation. This monitoring program implements Section IV.B.2 of Appendix I to 10 CFR Part 50 and thereby supplements the radiological effluent monitoring program by verifying that the measurable concentration ~s of radioactive materials and levels of radiation are not higher than expected on the basis of the effluent measurements and the modeling of the environmental exposure pathways. Guidance for this monitoring program is provided by the Radiological Assessment Branch Technical Position on Environ-mental Monitoring. The initially specified monitoring program will be effective for at least the first three years of commercial operation. Following this period, program changes may be initiated based on operational experience. The required detection capabilities for environmental sample analyses are tabulated in terms of the lower limits of detection (LLDs). _The LLDs required-by-Tab!c d.12 0 are considered optimum for routine environmental measurements in industrial laboratories. It should be recognized that the , LLD is defined as an a priori (before the f act) limit representing the capa- I bility of a measurement system and not as an a posteriori (after the fact) limit for a particular measurement. Detailed discussion of the LLD, and other detection limits, can be found in HASL Procedures Manual, HASL-300 (revised annually), Currie, L. A. , " Limits for Qualitative Detection ana Quantitative Determination - Application to Radiochemistry," Anal. Chem. 40, 586-93 (1968), and Hartwell, J. K., " Detection Limits for Radioanalytical Counting Techniques," Atlantic Richfield Hanford Company Report ARH-SA-215 (June 1975). PWR-STS-RETS B 3/4 12-1 1/4/83
3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING @ t BASES 3/4.12.2 LAND USE CENSUS This specification is provided to ensure that changes in the use of areas at and beyond the SITE BOUNDARY are identified and that modifications to the radiological environmental monitoring program are made if required by the results of this census. The best information from the door-to-door survey, from aerial survey or from consulting with local agricultural authorities shall be used. This census satisfies the requirements of Section IV.B.3 of Appen-dix I to 10 CFR Part 50. Restricting the census to gardens of greater than 50 m2 provides assurance that significant exposure pathways via leafy vege-tables will be identified and monitored since a garden of this size is the minimum required to produce the quantity (26 kg/ year) of leafy vegetables assumed in Regulatory Guide 1.109 for consumption by a child. To determine this minimum garden size, the following assumptions were made: 1) 20% of the garden was used for growing broad leaf vegetation (i.e. , similar to lettuce and cabbage), and 2) a vegetation yield of 2 kg/m2, 3/4.12.3 INTERLABORATORY COMPARISON PROGRAM The requirement for participation in an approved Interlaboratory Ccmparison Program is provided to ensure that independent checks on the precision and accu-racy of the measurements of radioactive material in environmental sample matrices are performed as part of the quality assurance program for environ-mental monitoring in order to demonstrate that the results are valid for the purposes of Section IV.B.2 of Appendix I to 10 CFR Part 50. PWR-STS-RETS B 3/4 12-2 1/4/83
I O SECTION 5.0 DESIGN FEATURES t r b I l k i
JUSTIFICATIONS Section 5.0 Items 5.1.3 and 5.1.4 were added to satisfy the requirements of the RETS. All other marked-up data in this Section is Seabrook Specific data. b i
5.0 DESIGN FEATURES 5.1 SITE EXCLUSION AREA 5.1.1 The exclusion area shall be as shown in Figure (5.1-1).
~
LOW POPULATION ZONE 5.1.2 The low population zone shall be as shown in Figure (5.1-2). S ee Insert 1 5.2 CONTAINMENT CONFIGURATION 5.2.1 The reactor containment building is a steel lined, reinforced concrete building of cylindrical shape, with a dome-roof and having the following design features:
- a. Nominal inside diameter = 140 feet.
- b. Nominal inside height = [ feet.
- c. Minimum thickness of concrete walls = L/ feet.(oiMe:S
- d. Minimum thickness of concrete M'= 3 feet. /o e'ncAe3
- e. Minimum thickness of concrete floor pad = & .be.w md anA 4 foot- dbvd
- f. Nominal thickness of steel 1.iner = Nhes- ifq, J/g, a,j t/2 ancA fo r-&e
. floor, watti ad dome vespec+tve ly
- g. Net free volume = cubic feet.
;Li 70%ooo DESIGN PRESSURE AND TEMPERATURE -
5.2.2 The reactor containment building is designed and shall be maintained for a maximum ir.ternal pressure of 92. psig and a temperature of 2_%. F. A heasuve)-from h top sitrface. &f-L dild 4p -fke inigde, oMhc done. W-STS 5-1 MAY 151976
- INSERT I SITE BOUNDARY FOR CASE 00S EFFLUENTS 5.1.3 The s'ite boundary for gaseous effluents shall be as shown in Figure 5.1-1. ,
DISCHARGE BOUNDARY FOR LIQUID EFFLUENTS 5.1.4 The discharge boundary for liquid effluents shall be as shown in Figure 5.1-3. b i l I i i l l l l l r. I i h f
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l DESIGN FEATURES 5.3 REACTOR CORE l l FUEL ASSEMBLIES f 5.3.1 The reactor core shall contain l9 3 fuel assemblies with each fuel assembly containing ;u,y fuel rods clad with (Zircaloy -4). Each fuel rod shall have a nominal active fuel length of / W inches and contain a maximum total weight of g grams uranium. The initial core loading shall have a weight percent U-235. Reload fuel shall be similar maximum enrichment in physical design to the of 3_j_ initial core loading and shall have a maximum enrichment of & weight percent U-235. CONTROL R0D ASSEMBLIES 5.3.2 The reactor core shall contain 57 full length end } control rod assemblies. The full length control rod assembTies' shall part length contain a nominal 142 inches of absorber material. , The part !cngth centrol red-
-:::c.blic; cha!' centain a nort.inal 36 inche: of abscrber material--et-theit -icwer cads. The nominal values of absorber material shall be 80 percent silver, 15 percent indium and 5 percent cadmium. All control rods shall be clad with stainless steel tubing. -The--balance Of the vcid -lengtM&the paet- -!cngth red: chall contair cluriner oxide.
4 5.4 REACTOR COOLANT SYSTEM DESIGN PRESSURE AND TEMPERATURJ 5.4.1 The reactor coolant system is designed and shall be maintained:
- a. In accordance with the code requirements specified in Section (5.2) of the FSAR, with allowance for normal degradation pursuant to the applicable Surveillance Requirements,
- b. For a pressure of 215 psig, and
- c. For a temperature of g50 F, except for the pressurizer which is
#O F.
VOLUME 5.4.2 The total water and steam volume of the reactor coolant system is /2,350
+ cubic feet at a nominal T 625r ava of (gg,) F.
5.5 METEOROLOGICAL TOWER LOCATION 5.5.1 The meteorological tower shall be located as shown on Figure (5.1-1). (
-W-STS 5- '
OCT 151978
l OESIGN FEATURES i 5.6 FUEL STORAGE CRITICALITY
- 5. 6.1.1 The spent fuel storage racks are designed and shall be maintained with:
- a. A k,ff equivalent to less than or equal to 0.95 when flooded with '
l ( unborated water, which includes a conservative allowance of l
, t 2h ) delta k/k for uncertainties as described in Section (4.3) of the FSAR.
/D.35 g b.
A nominal ()() inchscenter-to-center distance between fuel assemblies placed in the storage racks. 5.6.1.2 The k,ff for new fuel for the first core loading stored dry in the 0 98-spent fuel storage racks shall not exceed (u>sa) when aqueous foam moderation is assumed. DRAINAGE 5.6.2 The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation /4 feeh f ihches CAPACITY 5.6.3 The spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than 1230 fuel assemblies.
- 5. 7 COMPONENT CYCLIC OR TRANSIENT LIMIT 5.7.1 The components identified in Table 5.7-1 are designed and shall be maintained within the cyclic or transient limits of Table 5.7-1.
W-STS 5s .ggy 2 0 E60
TABLE 5.7-1 COMPONENT CYCLIC OR TRANSIENT LIMITS , T 4 " CYCLIC OR DESIGN CYCLE COMPONENT TRANSIENT LIMIT OR TRANSIENT 2co Reactor Coolant System ke50)Seatupcyclesat5100F/hr Heatup cycle - T,y9 from-5 200 F and ( cooldown cycles at to > 550*F.
< 100 F/hr. CooTdown cycle - T avg from
> 550 F to 5 200 F.
200 (459) pressurizer cooldown cycles Pressurizer cooldown cycle at 5 200 F/hr. temperatures from > 650 F to 5 200 F. 80 (400-) 1% s of load cycles, without > 15% of RATED THERMAL POWER to immediate turbine or reactor trip. D% of RATED THERMAL POWER.
'/O l T (5&) cycles of loss of offsite Loss of offsite A.C. electrical l C A.C. electrical power. ESF Electrical System. l 90 , .
(400-) cycles of loss of flow in one Loss of only one reactor reactor coolant loop. coolant pump. 400 (609) reactor trip cycles. 100% to 0% of RATED THERMAL POWER. (10) inadvertent auxiliary spray Spray water temperature differential actuation cycles. > 320 F. (50) leak tests. Pressurized to > (2485) psig. (5) hydrostatic pressure tests. Pressurized to > (3100) psig. Secondary System (1) steam line break. Break-in a > 6 inch steam line. s (5) hydrostatic pressure tests. Pressurized to > (1350) psig. G Di
l STANDARD TECHNICALSPECIFf, CATIONS SECTION 6.0
- ADMINISTRATIVE CONTROLS l
s
JUSTIFICATIONS Section 6.0 In the text of Section 6.0 where a capital letter with a circle around it appears, please refer to this sheet for the appropriate justification. A. Seabrook Station plant specific information. B. PSNH letter to NRC Serial SBN-492 dated 3/24/83. C. NRC Generic letter 82-12 and 83-14 require this information. D. This Section deleted and is included in Chapter 13 of the FSAR. E. As provided for by NUREG-0472 Rev. 3 Draft. G. Requirements of 10 CFR 50.54 (x) and (y). H. Seabrook Station does not have a " Health Physicist" perform this task. Station procedures for preparing the RWP assigns this responsibility to a
" Health Physics" qualified technician.
I. When health physics personnel are stationed to exercise direct control of an area,-the flashing light is unnecessary and a distraction, in addition to possibly causing longer stay times while it is placed and removed. J. Per requirements of NRC Generic Letter 83-43. K. Per requirements of NRC Generic Letter 82-16. L. See Attachment I M. Liquid effluents are not released inside the site boundary. 1 l i i
nitschanenTb. ( In the Table, footnote "b" instructs that all mean concentrations calculated will use only detectable measurements. Such a mean can be significantly biased in the positive direction. Such a treatment of data, for the purposes'of the Annual Report, has been recognized by the health physics and environmental professions as being incorrect. The proper way to calculate a mean is to'use all data, whether positive or negative or whether greater or less than the LLD. (See references 1 and 2 below.) The remainder of the table's format is acceptable and is currently being used by the other operating plants in New England. Ref. 1 - NUREG 0475. Radiological Environmental Monitoring by NRC Licensees for Routine Operations of Nuclear Facilities. October 1978. (See Page 9.) Ref. 2 - Upgrading Environmental Radiation Data. Health Physics Society Committee Report HPSR-1. 1980. .(See Chapter 6.) i 1 's l 1 i
ADMINISTRATIVE CONTROLS 6.1 RESPONSIBILITY stuhon Hae.ne er- sfahon 6.1.1 The (P! ant Superintendent) shall be responsible for overall-un-i-t- opera-tion and shall delegate in writing the succession to this responsibility during his absence. Sperbtenh,d-6.1.2 The Shift 4cpcrvison (or during his absence from the Control Room, a designated individual) shall be responsible for the Control Room command function. A management directive to this effect, signed by the (Mghc:t Ic';el-ef ccrperate management} shall be Missued to all station personnel,er er ernua! b ;5;' (ynce-Preudeni,Nutea.v %Ldsn) 6.2 ORGANIZATION OFFSITE ddwn . 6.2.1 The offsite organization for-tmM management and technical support shall be as shown in Figure 6.2-1. - STATioAl
-UN-T--STAFF STNMDA/
6.2.2 The -Ga44, organization shall be as shown in Figure 6.2-2 and:
- a. Each on duty shift shall be composed of at least the minimum shift crew composition shown in Table 6.2-1.
- b. At least one licensed Reactor Operator shall be in the Control Room when fuel is in the reactor. In addition, while the unit is in MODE 1, 2, 3 or 4, at least one licensed Senior Reactor Operator shall be in the Control Room.
- c. A health physics technician # shall be on site when fuel is in the reactor.
- d. All CORE ALTERATIONS shall be observed and directly supervised by either a licensed Senior Reactor Operator or Senior Reactor Operator Limited to Fuel Handling who has no other concurrent responsibilities during this operation.
- e. A site Firg Brigade of at least 5 members shall be maintained onsite7 at f*d d all times The Fire Brigade shall not include the Shift Supervi cr,T "
and the (3X) other members of the minimum shift crew necessary for safe shutdown of the unit and any personnel required for other essential functions during a fire emergency. The health physics technician and Fire Brigade composition may be less than the minimum requirements for a period of time not to exceed 2 hours in order to accommodate unexpected absence provided immediate action is taken to fill the required positions. ALL STS 6-1 , ggj
ADMINISTRATIVE CONTROLS UNIT STAFF (Continued) .
- f. Administrative procedures shall be developed and implemented to limit the working hours of staff who perform safety-related functions (e.g., licensed Senior Operators, licensed Operators, health physicists, auxiliary operators, and key maintenance personnel).
skMon - [The amount of overtime worked by-unit staff members performing safety-related functions shall be limited in accordance with the NRC Policy Statement on working hours (Generic Letter No. 82-12).] or [ Adequate shift coverage shall be maintained without routine heavy use of overtime. The objective shall be to have operating personnel work a normal 8-hour day, 40-hour week while the W is operating. Cggag However, in the event that unforeseen problems require substantial amounts of overtime'to be used, or during extended periods of shut-down for refueling, major maintenance, or major plant modification, on a temporary basis the following guidelines shall be followed:
- 1. An individual should not be permitted to work more than 16 hours straight, excluding shift turnover time.
- 2. An individual should not be permitted to work more than 16 hours in any 24-hour period, nor more than 24 hours in any 48-hour period, nor more than 72 hours in any 7-day period, all excluding shift turnover time.
- 3. A break of at least 8 nours should be allowed between work periods, including shift turnover time.
- 4. Except during extended shutdown periods, the use of overtime should be considered on an individual basis and not for the entire staff an a shift.
Any deviation from the above guidelines shall be authorized by the kHNSNPlant Superint6ent} or his deputy, or higher levels of manage-ment, in accordance with established procedures and with documenta-tion of the basis for granting the deviation. Controls shall be included in the procedures such that individual overtime shall be reviewed monthly by th W Plcr.t Superintcndant-] or his designee to assure that excessive hours have not been assigned. Routine deviation from the above guidelines is not authorized.] k
'f-STS 6-2
4-uitti 6e supplie.1 d a lcder 4d e. . FIGURE 6.2 - 1 g3 OFFSITE ORGANIZATION
l Qn aptoLje organ >3ab charF will be swepli&A a+ a. lafer day. b f-H
i Table 6.2-lh MINIMUM SHIFT CREW COMPOSITION TWO UNITS WITH TWO SEPARATE CONTROL ROOMS
~
WITH UNIT 2 IN MODES 5 OR 6 OR DE-FUELED POSITION NUMBER OF INDIVIDUALS REQUIRED TO FILL POSITION MODES 1, 2, 3 & 4 MODES 5 & 6 a SS l f rene SR0 1 Hene i RO 2 . 1 9 A0 2 2 STA 1C None WITH UNIT 2 IN MODES 1, 2, 3 OR 4 POSITION NUMBER OF INDIVIDUALS REQUIRED TO FILL POSITION MODES 1, 2, 3 & 4 MODES 5 & 6 a SS 1" l SRO 1 None R0 2 1 A0 2 1 STA l 'C None a/ Individual may fill the same position on Unit 2 b/ One of the two required individuals may fill the same position on Unit 2. d This Position rep [ red if no SenI* Lbensed opodor on shift W an STA cer6(kahhn. ALL STS 6- AUG 12 B80 4
- :bic C.2 1 41:NIMON SUIri CL CC".CCITION
("!CLC U"IT I?C:L!TY
'n run i : NUMBER OF INDIVIDUALS REQUIRED TO FILL POSITI F
~
&6 SS 1 1 SRO None R0 2 A0 2 1 1 None s Su erid a SS -
Shif t 58per"e e.dise with a Senior Reactor Operators License on Unit 1 - SR0 - Individual with a Senior Reactor Operators License on Unit 1 RO - Individual with a Reactor Operators License on Unit 1 AO - Auxiliary Operator STA - Shift Technical Advisor
%pcNden&e,d Except for the Shift Super"icer, the Shift Crew Composition may be one less i than the minimum requirements of Table 6.2-1 for a period of time not to exceed 2 hours in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore the Shift Crew Composition to within the minimum requirements of Table 6.2-1. This provision does not permit any shift crew position to be unmanned upon shift change due to an oncoming shift crewman being late or absent.
S pe,suge4 S***N * *S l During any absence of the Shift aperci:cr from the Controll Room while the unit is in MODE 1, 2, 3 or 4, an individual (other than -t4WShift Technical Advisor) with a valid SR0 license shall be designated to assume the Control _ 4%g Room command function. During any absence of the Shift Super"iccM from the Control Room while the unit is in MODE 5 or 6, an individual with a valid SR0 l or R0 license shall be designated to assume the Control Room command function. l l b ALL STS 6 '4 AUG 6 m81
t ADMINISTRATIVE CONTROLS
'q o PeR nrionen L secr/oA> /
%2. 3 INDE"ENDEN' SAFETY ENGINEERING-GRGUp- (-MEG-) @ES) /,e x
FUNCTION
,/g O ES /
6.2.3.1 The " shall function to examine plant operating ch,aracteristics, NRC issuances,h qdustry advisories, Licensee Event Reports and other sources of plant design and operating experience information, including plants of similar design, which may indicate areas for improvingilant safety. COMPOSITION ,, OES 6.2.3.2 The M EG shall be composed of at ast le / five, dedicated, full-time engineers located on site. /
- 7 RESPONSIBILITIES -
/
o ss cog niya.nce 6.2.3.3 The MEG shall be responsible or maintaining curve.llence-of plant activities to provide indepefident verific'ation* that these activities are performed correctly [and that human errors are \ reduced as much as practical. AUTHORITY
/
6.2.3.4 The shall make detailed reccmmendations fa revised procedures, equipment f modifications, maintenance activities, operatio activities or other means of improving plant safety to 'e "igh !c'/c! corpcrete ^fficisi
- il -in j ticMM211v oHantad psi'icr. who iInct in the unegement clair, fer_.
, , ~ -. .,. m.,-a..,.+v....,.
, y),,e 7 ,; t_ g ,cu- (, ; A ,,t, 6.2.4 SHIFT TECHNICAL ADVISOR s gra b dent The Shift Technical Advisor shall provide technical support to the Shift Super v4eer in the areas of thermal hydraulics, reactor engineering and plant analysis with Gu when regard to the no SRO ** safe shM-toperation of the unit.The STA shil be asared to the opuakm3 kas sta cerh'#mfab,i,
.6_3d M STAFF QUALIFICATIONS ./
N ,wn 7 minimum qualifi-6.3.1 Each member of the sait stafLshallmeet_or-exceed't~he .
' cations of (ANSI "18.1-19711.R.c..-l.PRFl% 7 chary lO7pxcerthf- MsI/M5 31 NF ws/l be. used eu__thc--staxda#74Tieu o4 ause arc.: - #us. ~@ _
ifications for members of the unit staff sha e y use of an overall q
- atement referen sNSI Standard agreed to by the NRC staff) or alternately by spe *
- ual position qualifications.
Generally, the first method ~ erable; however, tne - thod is adaptable to th staffs requiring special qualification sta . . . b a unique organizational structure.
"Not responsible for sign-off function.
ALL STS 6-7 t i
ADMINISTRATIVE CONTROLS t [Trum, Cen4ce Manager ]
'674m TRAINING .
, Iic ens ci,
.N 6.4.1 A retraining.and replacementvtrainingjrogram-for'th's.un+g_taf g --
bs f shall be maintained under tWdirection of_the_(-pe:it'icr title) and shall meet or
~
exceed the requirements and recommendations of Section (5,5) of (an ANSI 181 19H Standard-agreed to by,the1GC staf f)- and Appendix "A"'of-10 sCFR Part 55 and the supplemental-requirements specified in Section A and C of' Enclosure 1 of theMarch-2871980 NRC letter to all licensees shall include' familiarization w.ith' relevant industry operational experience identified by the M EG- N OES
'6. 5 REVIEW AND AUDIT Lb* * "'" 3
\ s . s The . , b which independent review and audit of facility operati
- f accomplisheo . ke one of several forms. The licensee m er assign /
this function to an ~ tional unit separate and
- pendent from the group having responsibility for uni 'on or m i ize a standing committee composed of i'ndividuals from within a e the licensee's organization.
Irrespective of the
\ .
used, the licensee shall spe . he. details of each functi ' ement provided for the independent review an 't process as ' rated in the'following s example specifications. ,e
\ '
6.5.i STATICA1 cPErtATrous REVIER/ co/yn f 77eg (56RC.) UNIT REVIEW CROUP (URG) , FUNCTION ' So RC. l 6.5.1.1 The (Urit Revice Group) sha1Q function to advise the (Plant Super-Staho^ H"'"fe'
&ntentfent-) on all matters related to nuclear' safety.
COMPOSITION 6.5.1.2 The @ nit-Revir.; Croup.)- shall be composed of the: s+abn Hana a \ Chairman:"**6*" (Plant Supe intendent) ,e-Member: ,/(0perations Super"icerf$,a,,g,a.sa nager wm Member: ,- (Technical 4eperviscr)sev iu5 H""'t'" . Member: / (Maintenance Supervi:cr)DeP**** 6"d""8"" .
. Member: / (Maat Instrument ^%d Contr'ol Engineer)Dep facai svesvi.5er Member: (P!cnt Nuclear Enginec=3feuldca Member: (-Health-Phys 4c4r# lica.f m PWsio\va<edat DepaAnc,* Spevvisor ceM+<atNo Mcaber; '
5 E #5 D'F""+'" *# T'""" #" rie=be r: em ine e ri't *'" cu,,,os(,7 ocparte Supervis*" \ ' ALTERNATES '
- ./ Sonc.
- 6. 5.1. 3 All alternate members shall be appointed in writing by the -@RG-)-
Chairman,to serve on a temporary basis; however, no more than two alternates shall par'ticipate as voting members in @ RG) activities at any one time.
/ so R,C. '
'\
ALL STS 6-8 SEP 161980
. INSERT I '.O .except that Training Center Staff instructors who have not obtained a facility SRO license may be certified for specific subjects in which they have ' demonst, rated SRO license level comprehension and skill. In addition, guest lecturers considered to be experts by nature of their work responsibilities may.be.us'ed on a limited basis to supplement the training center staff ins tructors'. The guest lecturers are exempt from senior operator criteria.
The program.... N
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ADMINISTRATIVE CONTROLS 't
\ MEETING FREQUENCY SORC-6.\
5.1.4 The (4RG-) shall meet at least once per calendar month and as convened by thes(-URfr) s Chairman or his designated alternate. sORC. QUORCH So R C-. 6.5.1.5 The minimum quorum of the -(4RG) necessary for the performance of the GORC(URG) responsibi,lity and authority provisions of these Technical Specifications shall consist of\the Chairman or his designated alternate and four members
. including alternat'es.
J RESPONSIBILITIES 6.5.1.6 Th e '" " "^ " # -" " aup-) shall be responsible for:
- a. Review of 1) all\procedures required by cification Spe/ 6.8 and changes
'thereto, 2) all progr,ams required oy Specification 6.8 and changes thereto, 3) any other proposed procedure's or changes thereto as deter-d 4 mined by the (s"+larit Su er ntendent)-to' affect nuclear safety, dwn danager ~,
- b. Review of all proposed te'sts and experiments that affect nuclear safety. N /
c.
\/
Review of all proposed changes to Appendix "A" Technical Specifications.
\
- d. Review of all proposed changes o'r s modifications to unit systems or equipment that affect nuclear safety.
@c P,es dext,hlea fW=l)w
,e 4
- e. Investigation of all violations of the Technical Specifications including the preparatjon and forwarding of reports covering evalua-tion and recommendations to prevent recurrence to the (4uper4ntenden.
eHcuar "180t5} and 'to the (Cc; pan"eG ^~ Rcricu Nu^ht e,newand-Audit e-m.mdicW" " c" AiMQ udea,hf all RE.PORiA8L . x
- f. Peview of event +-requ+ra.E
-a ev cur-wrRtenn+ 4 74 "+ 4cn-.to-the.
4cr-iesinn. /
- g. Review ofshum&e' on operations to detect potential nuclear safety hazards.
\
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- h. Performance' of special reviews, investigations or analyses a $9e ,.
reports , thereon as requested by the (Plant Superinte-dem+ r the (Compan Nucicar Review and ^.udit Croup)NSe9AC, \
,%b o'n \
i. Revie/y w of the ecurity Plan and implementing procedures and shall submit recommended changes to the (Gempany Ncclear "cview and Aud-i-t ' Grcup).- NSIM. L.. i j/. / Review submit of the Emergency recommended Plantoand changes theimplementing procedures mpany-Nuc-leae-Rev i ew-and shall and-Audit
/ .creup+. S n r<.c. .
LL TS 6-9 686 DS SEP 101980
[ \ INSERT II
\
- k. ' Review of any accidental, unplanned or uncontrolled radioactive release including the preparation of reports covering evaluation, rec'mmendations o and disposition of the corrective action to prevent recurrence and the forwarding of these reports to the (Vice President, Nuclear % Production) and to the NSARC. ,
- 1. Review o changes to the PROCESS CONTROL PROGRAM and.the OFFSITE DOSE CALCULATION MANUAL. *
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ADMINISTRATIVE CONTROLS s AUTHORITY @ SORC
\ The (Unit Ret'iew Crcup)-shall:
6.5.1.7
\ Sto. hon Hana
- a. \ Recommend in writing to the (Shatr-Super'gerntendent) approval or disapproval of items considered under 6.5.1.6(a) through (d) above.
- b. R nder determinations in writing with regard to whether or not each item. considered under 6.5.1.6(a) through (e) above constitutes an unreviewed safety question. .
\ , sca. pres, bed *, % Ica Pra chon
- c. Provide ' written notification within 24 hours to the (Superintendent-
.-<..n..-. o i. s .,,. +~,. s, and the y..
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ScRC. ~ disagreement Detween tne M and the (Plcat Scperintend...., patio 4 **f however, the\(Picnt Super ntendentkshall have responsibility for/ i resolution of 'such disagreements pursuant'to 6.1.1 above.
\
RECORDS
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Come. 6.5.1.8 The (4htitSODev!ee R.C Grcup).shall maintain written minutes pf each {41RQ meeting that, at a minimum, docuriient the results of all (4%)"aTrivities performed under the responsibility'and authority provisions of these technical specifications. f-Copies shall be provided to the (Super 4"tendent of ocwcr4&c#res& /c8" h*J" <E*' Plants) and the ggny Nuclear Rct-i,c cnd ^.udit Grcup). 6.5.2 CC"?AN'l NUC'.:M "EVIP "O ^UDF 'CRCUP (C""ACL Alucuen SAFsiy nuosr Aivo Review connstreE (W0 h FUNCTION / \
/ \
A/ 5 A R.g f s 6.5.2.1 The (Ec p:n3 ..uclear "c.ie" and %di+. Crcup)- shall function to provide independent review and audit of designated activities in the areas of: a,
/ \
nuclear power plant operations \
- b. nuclear engin/ eering x
/
- c. chemistry and radiochemistry j' y
- d. metallurgy '
/
- e. instrumentation and control s
/ s
/
f. radiological safety
/
- g. ,/ mechanical and electrical engineering
/
l
}
ALL STS 6-10 SEP 101980
ADMINISTRATIVE CONTROLS e
's h. quality assurance practices N
N 4r (4ther-appropriate fields asscciated with the unique charactcristics-N cf the nuclear power plant)-
'N
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COMPOSITION 6.5.2.2 The\NSo g c,'CNR?,C)- shall be composed of~ the+ af /.wf G<- eusons. ne da s d er.6 he appo W sit w 6Fa'ny Chateman 44 a$ e4ces inddingby the senio Vn.cfresJent. Co ttec.+q shall be ompehuif 40 condet reweu:4 ident tfred di (.5'. .t i . Each 4n e.m(rer *^all c
%eef *e fyellficd.Fions of ANSr- 3.1 - MW, Sechton 4 7. / .
N ALTERNATES \ -
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6.5.2.3 All alternate members shall be appointed in writing by the-(CNR^.G)- Senice~ Vtcc 6ts5QT ' Direct 0 to serve on a temporary basis; however, no more than tt:0 :! te rnate:4 miino'dy shall participate as voting members \in (CNRAC) activitie: at any one time.
., . . AlsMt.c. achushes
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CONSULTANTS .
. .NSAAC 6.5.2.4 Consultants shall be utilized as determined by the (C.N.G) Cirector--
to provide expert advice to the (-C-NRA&). MSNLL-
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MEETING FREQUENCY /
#54 AL, /
6.5.2.5 The (CNRAC) shall mee't at least once per calendar quarter during the initial year of unit operation following fuel loading and at least once per '
' thew eb l six months-thereafter3 r 2.5' *7o . @
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QUORUM $,. [y cu,a NSA "' Nsan.c
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- 6. 5. 2. 6 h The -niaiaum./ quorum f the (CPRAC) necessary for the performance of the (CNRAC) review'and audi functions of these Technical Specifications shall consistoftheBysector or u dod enn+ ed altermte and gat least 4 GNRA6-)M544(- h members including alternates. No more than a minority of the quorum shall have line responsibility for operation of the unit.The. \ltee. caw %,.n ana/or a4cende. cd pacticipale as an IJ S A A.c. *emb er iden ev er % Chairmn (or ademale) is m Al%Aence.
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f dALL STS 6-11 SEP 101c80
ADMINISTRATIVE CONTROLS I H R$ VIEW s NMAL 6.5.2.7 The (C""AC) shall review: x
- a. \The safety evaluations for 1) changes to -prccedurci, equipment or@
systems and 2) tests or experiments completed under the provision of Section 50.59, 10 CFR, to verify that such actions did not constitute an lunreviewed safety question. . ,
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- b. Propose'd changes to procedures, equipment or systems which involve an unreviewed safety question as defined in Section 50.59, 10 CFR.
'N
- c. Proposed tests or experiments which involve an unreviewed safety question as defined in Section 50.59, 10 CFR. f
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- d. Proposed changes to Technical Specifications or,t is Operating License. \..\ ,
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- e. Violations of codes, regulations, orders, Technical Specifications, license requirements, or'of internal procedures or instructions having nuclear safety significance.
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- f. Significant operating abnorm'alities or deviations from normal and ,
expected performance of sh.+too c%uip ment- ht a.Hect nuc9 ear safety
- g. 4 n Nrbh cur -itter M5t4 R etier te the r - icc on. i
, s
- h. All recognized indications of an unanticipated deficiency in some aspect of design or operation of structures, systems, or components.
ht c.ou.la o.ffect- nuclear redefy/ and \ s
- i. Reports and meetings minutes of the #'W~"^"- - Crcup).
SdRCA AUDITS . i slo. host - 6.5.2.8 Audits of-uni.t activities shall be performed under the cognizance of the (C" pac). These audits'shall encompass:
/ daSt '
- a. The conformance of ur,4t, operation to provisions contained within the Technical Specifications and applicable license conditions at least once per ,12 months. i
- b. The performance, training and qualifications of the entire.dati$n .
' .un+tr staff at least once per 12 months.
- c. g he results of actions taken to correct deficiencies occurring.in M equipment, structures, systems or method of operation that
' affect nuclear safety at least once per 6 months. .
, Ehe aaM3 dali be perfor-ed witu%c spec 6ed he shfen'al w'4 4 a'a n "'" "'
allm aa Lie d esisa lio.e & r K'7o protisded thf 1se corn hs ' ed ide* val fse * *Y @ 3 cons.eesiT(og a di- idervals does nd e x ce e d 3 27 t o m es th < .yiecs CNd \J./ 6-12 fLISTS SEP 101990
ADMINISTRATIVE CONTROLS X
- d. The performance of activities required by the Operational Quality Assurance Program to meet the criteria of Appendix "B", 10 CFR 50, at least once per 24 months.
- e. The Emergency Plan and implementing procedures at least once' per gmonths. ,/
- f. TheSecurityPlanandimplementingproceduresatleastInceper s
'24. months.
- g. yo er area of operation considered appropriate b the CMAG (usarc.) or the (Vice Prc;idcat Operetiens}.Sc.uer Ve Pres ed
- h. The Fire Pr t,ection Program and implementing procedures at least once per 24 months. ,
N /
- i. An independent fire protection and loss prevention inspection and audit shall be performec' s annually utilizing either qualified offsite licensee personnel or an outside fire srotection firm.
- j. An inspection and aud t,0f the fire protection and loss prevention program shall be performe,d by an.outside qualified fire consultant at intervals no greater than 3 years.
s See Insert Br AUTHORITY ,[ , Ns4 Ac. / . senior %cc P eulent-6.5.2.9 The (CNMC) shall report'to and adyise the (Vicc Prcsident Opcraticas) on those areas of responsibility'specified in Sections 6.5.2.7 and 6.5.2.8. RECORDS ,/ NSAAC. 6.5.2.10 Records of (C N AC) activities shall be p epared, approved and dis-tributed as indicated below: nsnee. Minutes of,e/ach (CNPAC}
- a. meeting shall be prepared, apprc' icd and forwarded to the (j'gc,,Pgigeg^ gal.ivn;) 'within g days following each meeting.
- b. Repo,rt's of reviews encompassed by Section 6.5.2.7[e bove, ; hell be prcpared, approvcd and fenscrded to the (Vice within tt days following completion of the review. 5"*" U'"O*'* President Operetiensb 3o h k Denae/Audit l
c. g toreth7 (ports encompassed by Vice Prc;idcat-0peratiens) and to the management, positions hall Section be forwarded 6.5. responsible for the areas audited within 30 days after completion of
/. the audit by the auditing organization. _ \
l chall b e melwJecl' i.i -the. rmindea dere applicable *
- d wa"ded
! ; kAdte sepereded cove r t @ ne ce r,setr y l [
/
/ ALL STS 6-13 SEP 101980
~ _ . - - - - , -
,-y - _ _ - -
%. INSERT III I
s .
. k '. The radiological environmental. monitoring program and the results k thereof at least once per 12 months. f
- 1. The OFFSITE DOSE CALCULATION MANUAL and implementing procedures at least once'per 24 months.- ,/
m.
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The PROCESS CONTROL PROGRAM and implementing procedures for processing and packaging of radioactive wastes at least once per 24 months. n. N
-The performance of activities required by the Qu lity Assurance Program for effluent and environmental monitoring at least once per 12 months..
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y
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ADMINISTRATIVE CONTROLS EVEUT .#. 6.6 REPORTABLE @GatRREttCf-ACTION (.T) E VEA/T.S 6.6.1 The following actions shall be taken for REPORTABLE 0CC' RRENCES. J
- a. The Commission shall be notified and/or a report submitted pursuant to the requirements of Speci'icetion-C.S.s.eJ<.a 50.7J of loca. ArtSO,ad
- b. EachREPORTABLEE$L6ENCE-rcqui" 4 hcur acti'icitier to--t-he Cc--istier shall be reviewed by the fBRG) and submitted to the (GNRAG) and the (Superintendent of Dewe* Pipnt e usac thee. Pres usea+, Waeteae Machon 6.7 SAFETY LIMIT VIOLATION 6.7.1 The following actions shall be taken in the event a Safety Limit is violated:
Qnes*ae.6, waetea 9eoLaQ a. TheNRCOperationsCentershallbenotifiedbytelephoneassoonas] possible and in all cases within one hour. The(Super %tendentef( Pc er P hnts) and the (CNPaC)-shall be notified within 24 hours. Ns n ec.,
- b. A Safety Limit Violation Re ort shall be prepared. The report shall be reviewed by the 6fR8')F his report shall describe (1) applicable circumstances preceding the violation, (2) effects of the violation upon facility components, systems or structures, and (3) corrective b
action taken to prevent recurrence.
- c. The Safety Limit Violation Report shall be submitted to the Commission, the g g,.mc, d lCNRAC) and the (Superintendent Of Pcwer Picat4 within 14 days ie violation. %cc WesiJed,Nudcar deJac6on
- d. Critical operation of the unit shall not be resumed until authorized
' by the Commission.
6.8 PROCEDURES AND PROGRAMS 6.8.1 Written procedures shall be established, implemented and maintained covering the activities referenced below:
- a. The applicable procedures recommended in Appendix "A" of Regulatory Guide 1.33, Revision 2, February 1978.
-b. Refue!!ng-operations.
- c. Surve444ance-end-tost activities--of afaty2eFated-equ4pment--
ALL STS 6-14 SEP 2 81981
ADMfNISTRATIVE CONTROLS q @ E. b Security Plan implementation. g.C Emergency Plan implementation. p Y. Fir: . "retectier. ."regrc: imp k; ntatiem Sonc & Gee Tnsca + S h, , , , 6.8.2 ach procedure of 6.8.1 above, and changes thereto shall g,,,,be reviewed by the and approved by the (944nt-Super %tendent,-prior to implementation and reviewed periodically as set forth in administrative procedures. 6.8.3 Temporary changes to procedures of 6.8.1 above may be made provided:
- a. The intent of the original procedure is not altered.
- b. The change is approved by two members of the plant management staff, at least one of whom holds a Senior Reactor Operator's License on the unit affected. ,
- c. The change is documented, reviewed by the h and approved by the (Phn,t Super'a+endent) within 14 days of implementation.
5 kho.s $Aqu -
@ xexnsu+ %
6.8.4 The following programs shall ba established, implemented, and maintained:
- a. Primary Coolant Sources Outside Containment A program to reduce leakage from those portions of systems outside containment that could contain highly radioactive fluids during a serious transient or accident to as low as practical levels systems include (the rc:fr:rhti:n spray, safety injection',Y.chemical The and volume control, ips stripperPnd hyd*eger "ece@iners). The program shall include the followin h g;g g 4 c,,g ,_ , g (i) Preventive maintenance and periodic visual inspection requirements, and (ii) Integrated leak test requirements for each system at refueling cycle intervals or less,
- b. In-Plant Radiction Monitoring A program which will ensure the capability to accurately determine the airborne iodine concentration in vital areas under accident conditions. This program shall include the following:
(i) Training of personnel, (ii) Procedures for monitoring, and (iii) Provisions for maintenance of sampling and analysis equipment. l ALL STS 6-15 SEP 161980 i
INSERT IV I ! d. PROCESS CONTROL PROGRAM implementation.
- e. OFFSITE DOSE CALCULATION MANUAL implementation.
- f. Quality Assurance Program for effluent and environmental monitoring.
INSERT V
- d. In an emergency, reasonable action that departs from a procedure, license condition or a technical specification may be taken when this action is immediately needed to protect public health and safety and no action consistent with procedures, license conditions and technical specifications that can provide adequate or equivalent protection is immediately apparent.' ,
Such action shall be approved, as a minimum, by a licensed senior reactor operator prior to taking the action.
ADMINISTRATIVE CONTROLS i c. Secondary Water Chemistry -(-P':!9: caly)- A program for monitoring of secondary water chemistry to inhibit steam generator tube degradation. This program shall include: (i) Identification of a sampling schedule for the critical variables and control points for these variables, (ii) Identification of the procedures used to measure the values of the critical variables, (iii) Identification of process sampling points, which shall include monitoring the discharge of the condensate pumps for evidence of condenser in-leakage, (iv) Procedures for the recording and management of data, (v) Procedures defining corrective actions for all off-control point chemistry conditions, and (vi) A procedure identifying (a) the authority responsible for the interpretation of the data, and (b) the sequence and timing of administrative events' required to initicte corrective action,
- d. Backuo Method for Determining Subcooling Margin
-(, J.9: aitt, e eing!: charn:1 ;f merit:r ng in;trum:nt:tfen)-
i I A program which will ensure the capability to accurately monitor the Reactor Coolant System subcooling margin. This program shall include the following: (i) Training of personnel, and (ii) Procedures for monitoring.
- e. Postaccident Samoling A program which will ensure the capability to obtain and analyze reactor coolant, radioactive iodines and particulates in plant gaseous effluents, and containment atmosphere samples under accident conditions. The program shall include the following:
(i) Training of personnel, (ii) Procedures for sampling and analysis, (iii) Provisions for maintenance of sampling and analysis equipment: ALL STS 6-15 M f.F. 6 1031
ADMINISTRATIVE CONTROLS
/ 'l 6.9 REPORTING REOUIREMENTS e-v c err s ROUTINE REPORTS AND REPORTABLE 0CC= RENCES 6.9.1 In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following reports shall be submitted to the Circ;tcr cfWRC the Regicncl RemoulOfMmms fige p+fn+ce Inspection-and Oforcement unless otherwise noted.
STARTUP REPORT
- 6. 9.1.1 A summary report of plant startup and power escalation testing shall be submitted following (1) receipt of an operating license, (2) amendment to the license involving a planned increase in power level, (3) installation of fuel that has a different design or has been manufactured by a different fuel supplier, and (4) modifications that may have significantly altered the nuclear, thermal, or hydraulic performance of the plant.
6.9.1.2 The startup report shall address each of the tests identified in the FSAR and shall include a description of the measured values of the operating conditions or characteristics obtained during the test program and a comparison of these values with design predictions and specifications. Any corrective actions that were required to obtain satisfactory operation shall also be described. Any additional specific details required in license conditions based on other commitments shall be included in this report. 6.9.1.3 Startup reports shall be submitted within (1) 90 days following completion of the startup test program, (2) 90 days following resumption or commencement of commercial power operation, or (3) 9 months following initial criticality, whichever is , earliest. If the Startup Report does not cover all three events (i.e., initial criticality, completion of startup test program, and resumption or commencement of commercial operation) supplementary reports shall be submitted at least every three months until all three events have been completed. ANNUAL REPORTSM dak.a 6.9.1.4 Annual reports covering the activities of the e 44 as described below for the previous calendar year shall be submitted prior to March 1 of each year. The initial report shall be submitted prior to March 1 of the year following initial criticality. ' 6.9.1.5 Reports required on an annual basis shall include:
- a. A tabulation on an annual basis of the number of station, utility, and other personnel (including contractors) receiving exposures i
M Asingle submittal may be made for a multiple unit station. The submittal should combine those sections that are common to all units at the station. ALL STS 6-17 APR 2 1931
ADMINISTRATIVE CONTROLS I greater than 100 mrem /yr and their assnciated manrem exposure according to work and job functions,2/ e.g., reactor operations and surveillance, inservice inspection, routine maintenance, special maintenance (describe maintenance), waste processing, and refueling. The dose assignments to various duty functions may be estimated based on l pocket dosimeter, TLD, or film badge measurements. Small exposures ! totalling less than 20 percent of the individual total dose need not be accounted for. In the aggregate, at least 80 percent of the total whole body dose received from external sources should be assigned to specific major work functions. Docu~iedab 4 au challengp 4 ne pmsuri3er pouar oe = M
- b. The resulta uMhc core barre, acvement menHar'ng acti"ities-per-4ermed-dur-in[-the-reportperiod.t i @
ve.hef valves. f0 M s) or saf et volves-(CE un ts caly) _
<. (Any--ether unit unique reports required en en ennual basis.-)
MONTHLY OPERATING REPORT 6.9.1.6 Routine reports of operating statistics and shutdown experience, including documentation of all challenges to the PORVs or safetv valves, shaJJ be submitted on a monthly basis to the Director, Office of* Management-en+ 458" 9 Fragra;n Aa:1ye4e, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555, with a copy to th ";gior;' Of fic; cf In;pectf en and Enfer c =t, no later than the 15th each month followino the calendar month covered by the report. N R C-kenio%IAdm mstrafo O 4E?OP"3LE OCCURRENCES 4.0.1.7 T5: "DARTABLE OCCURRENCES of Specifications 6.9.1.8 and 6 0 'C . below, including corrective e d : - measures to r currence, shall Supplemental re e uired to fully describe
@ be reported final to theof resolution NRC.
occurr . ase of corrected or s tal reports, a licensee apart shall be completed and reference shall be ma report date.
^^^""' "07?"!C?TIO:, -ITH nRITTEN ICLLC'.fJP 4 1.8 The types of events listed below shall be reported within 24 hours telep o onfirmed by telegraph, mailgram, or facsimtie transmiss o the Director of ional Office, or his designate no late + e first
@ working day following the with a written foll eport within 14 days.
The written followup report shall i inimum, a completed copy of a , licensee event report form. on pro n the licensee event report form shall be a rr' . m e , as needed, by additionai ve material to
@ omplete explanation of the circumstances surrounding . nt.
2/
- This tabulation supplements the requirements of 520.407 of 10 CFR Part 20.
l i ALL STS 6-18 APR 2 1981 w _-_____- ___
r-ADMINISTRATIVE CONTROLS f
, Failure of the reactor protection system'or other systems subjec to l limiting safety system settings to initiate the required prote ive function by the time a monitored parameter reaches the setpoi ecified as the limiting safety system setting in the tech cal s scifications or failure to complete the required protect ve fu tion.
- b. Opera ion of the unit or affected systems when any par ..eter or operati n subject to a limiting condition for operat' n is less conserva ive than the least conservative aspect of e Limiting Condition or Operation established in the Technic Specifications.
- c. Abnormal deg dation discovered in fuel claddin , reactor coolant pressure boun ry, or primary containment.
- d. Reactivity anomali s involving disagreement ith the predicted value of reactivity balan under steady state c nditions during power operation greater tha or equalIto 1% Ak/ ; a calculated reactivity balance indicating a S TDOWN MARGIN les conservative than specified in the Technical Specifi tions; short erm reactivity increases that correspond to a react period of less than 5 seconds or, if subcritical, an unplanned re tivity nsertion of more than 0.5%
Ak/k; or occurrence of any unp anne criticality.
- e. Failure or malfunction of one or re components which prevents or could prevent, by itself, the fu f1 1 ment of the functional require-ments of system (s) used to cope with ecidents analyzed in the SAR.
- f. Personnel error or procedura inadequac which prevents or could prevent, by itself, the ful illment of t, functional requirements of systems required to cop with accidents analyzed in the SAR.
- g. Conditions arising from atural or man-made e ents that, as a direct result of the event re ire unit shutdown, ope ation of safety systems, or other pro ctive measures required Technical Specifications.
- h. Errors discovered 'n the transient or accident anal es or in the methods used for uch analyses as described in the sa ety analysis report or in th bases for the Technical Specification that have or could have per tted reactor operation in a manner less onservative than assumed n the analyses.
- 1. Performanc of structures, systems, or components that requ res remedial ction or corrective measures to prevent operation a manner ss conservative than assumed in the accident analyse in the sa ety analysis report or Technical Specifications bases; disco ery during unit life of conditions not specifically consi red j in e safety analysis report or Technical Specifications that i
~
re ire remedial action or corrective measures to prevent the e istence or development of an unsafe condition. A R STS N APR 2 1991
l ADMINISTRATIVE CONTROLS
- -THIRTY CA" '!:ITT:N REF^aTS -
.9.1.9 The types of events listed belov shall be the subject of wri n re rts to the Director of the Regional Office within thirty days o occu ence of the event. The written report shall include, as a nimum, a complet copy of a licensee event. report form. Information pr ided on the licensee ent report form shall be supplemented, as needed, additional narrative ma rial to provide complete explanation of the rcumstances
, surrcunding th vent.
- a. Reactor prote ion system or engineere afety feature instrument settings which a found to be less nservative than those esteb-lished by the Techn' al Specifica 'ons but which do not prevent the fulfillment of the fun ional r uirements of affected systems.
1
@ b. Conditions leading to oper in a degraded mode permitted by a l Limiting Condition for op tion- plant shutdown required by a Limiting Condition fo peration.;
- c. Observed inadequ es in the implementation o dmi'nistrative or i procedural con als which threaten to cause redu 'on of degree of redundancy p vided in reactor protection systems o ngineered safety fe re systems,
- d. Abno ..al degradation of systems other than those specified 1
- 6. .l.8.c above designed to contain radioactive material resul
rom the fission process. RADIAL PEAKING FACTOR LIMIT REPORT ('d caly) N 6.9.1. h The F xy limit for Rated Thermal Power (F P) shall be provided to the ei+ectenf the Regional nffke af Inspect on-and4nforc=nt, with a copy to the Director, Nuclear Reactor Regulations, Attention Chief of the Core Performance Branch, U. S. Nuclear Regulatory Commission, Washington, D.C. 20555 for all core planes containing bank "D" control rods and all unrodded core planes at least 60 days prior to cycle initial criticality. In the event that the limit would be submitted at some other time during core life, it will be submitted 60 days prior to the date the limit would become effective unless otherwise exempted by the Commission. Any information needed to support F xRTP will be by request from the NRC and need not be included in this report. l l3 ALL STS 6-24 'NOV 2 1981
ADMINISTRATIVE CONTROLS ANNUAL RADIOLOGICAL EN'/IRONMENTAL OPERATING REPORT
- l 6 6.9.1.14 Routine Radiological Environmental Operating Reports covering the operation of the unit during the previous calendar year shall be submitted prior I to May 1 of each year. The initial report shall be submitted prior to May 1 of the year following initial criticality.
The Annual Radiological Environmental Operating Reports shall include summaries, interpretations, and an analysis of trends of the results of the radiological environmental surveillance activities for the report period, including a compar-ison with preoperational studies, with operational controls as appropriate, and with previous environmental surveillance reports, and an assessment of the observed impacts of the plant operation on the environment. The reports shall also include the results of land use censuses required by Specification 3.12.2. The Annual Radiological Environmental Operating Reports shall include the results of analysis of all radiological environmental samples and of all environmental radiation measurements taken during the period pursuant to the locations speci-fied in the Table and Figures in the ODCM; as well as summarized and tabulated results of these analyses and measurements.in the ferm:t af the t;tle in th; JWdinicgical ^.::::: ment Erench Tec5aic:1 Pcsition, 9 visivu 1, ;ic.mnter 1:73; In the event that some individual results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted as soon as possible in a supplementary report. The reports shall also include the following: a summary description of the radiological environmental monitoring program; at least two legible maps ** covering all sampling locations keyed to a table giving distances and directions from the centerline of one reactor; the results of licensee participation in the Interlaboratory Comparison Program, required by Specification 3.12.3; discussion of all deviations from the sampling schedule of Table 3.12-1; and discussion of all analyses in which the LLD required by Table 4.12-1 was not achievable.
*A single submittal may be made for a multiple unit station.
**0nemapshallcoverkY$IEc'Ys near the site boundary; a second shall include the more distant stations.
Loca Ho45 l PWR-STS-RETS 6-) 1/4/83
ADMINISTRATIVE CONTROLS SEMIANNUAL RADICACTIVE EFFLUENT RELEASE REPORT
- Cl 6.9.1.X Routine Radioactive Effluent Release Reports covering the operation of the unit during the previous 6 months of operation shall be submitted within 60 days after January 1 and July 1 of each year. The period of the first report shall begin with the date of initial criticality.
The Radioactive Effluent Release Reports shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit as outlined in Regulatory Guide 1.21, " Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes and Releases of Radioactive Materials in Liquid and Gaseous Ef fluents from Light-Water-Cooled Nuclear Power Plants," Revision 1, June 1974, with data summarized on a quarterly basis following the format of Appendix B thereof. The Radioactive Effluent Release Report to be submitted within 60 days after January 1 of each year shall include an annual summary of hourly meteorological data collected over the previous year. This annual summary may be either in the form of an hour-by-hour listing on magnetic tape of wind speed, wind direc-tion, atmospheric stability, and precipitation (if measured), or in the form of joint frequency distributions of wind speed, wind direction, and atmospheric stability.** This same report shall include an assessment of the radiation doses due to the radioactive liquid and gaseous effluents released from the crit er station during the previous calendar year. This same report shall also include an assessment of the radiation doses from radioactive liquid 2nd gaseous @ effluents to MEMBERS OF THE PUBLIC due to their activities inside the SITE - B0UNDARY (Figure 5.1-h) during the report period. All assumptions used in making these assessments, i.e. , specific activity, exposure time and locaticn, shall be included in these reports. The meteorological conditions concurrent with the time of release of radioactive materials in gaseous effluents, as determined by sampling frequency and measurement, shall be used for determining the gaseous pathway doses. The assessment of radiation doses shall be performed in accordance with the methodology and parameters in the OFFSITE DOSE CALCULATION MANUAL (0DCM). The Radioactive Effluent Release Report shall also include once a year an assessment of radiation doses to the likely most exposed MEMBER OF THE PUBLIC from reactor releases and other needy urar.ium fut' cych source:, including doses from primary effluent pathways and direct radiation, for the previous calendar year to show conformance with 40 CFR Part 190, Environmental Radiation Protection Standards for Nuclear Power Operation. Acceptable methods for
*A single submittal may be made for a multiple unit station. The submittal should combine those sections that are common to all units at the station; however, for units with separate radwaste systems, the submittal shall specify the releases of radioactive material from each unit. **In lieu of submission with the first half year Radioactive Effluent Release Report, the licensee has the option of retaining this summary of required meteorological data on site in a file that shall be provided to the NRC upon request.
21 PWR-STS-RETS 6-4 1/4/83
l' ADMINISTRATIVE CONTROLS (5) ( ' calculating the dose contribution from liquid and gaseous effluents are given in Regulatory Guide 1.109, Rev. 1, October 1977. The Radioactive Effluent Release Reports shall include the following information for each class of solid waste (as defined by 10 CFR Part 61) shipped offsite during the report period:
- a. Container volume,
- b. Total curie quantity (specify whether determined by measurement or estimate),
- c. Principal radionuclides (specify whether determined by measurement or estimate),
- d. Source of waste and processing employed 44 9 dewatacad spent erin, sceperted dry weste, ev peratae-bettom:),
- e. ,
Type of container (e.g. , LSA, Type A, Typa R Large Quantity), and
- f. Solidification agent or absorbent (= 0 camant, ura form:!dehyde).
The Radioactive Effluent Release Reports snal:@<s4m4Fa
)nciuae a 4<<aI) list and descri of unplanned releases from the site to L'N"EST.4.CTZ" A",CAO of radioactive materials in gaseous and liquid effluents made during the reporting period.
The Radioactive Effluent Release Reports shall include any changes made during the reporting period to the PROCESS CONTROL PROGRAM (PCP) and to the OFFSITE - DOSE CALCULATION MANUAL (ODCM), as well as a listing of new locations for dose calculations and/or environmental monitoring identified by the land use census pursuant to Specification 3.12.2. e 1 l h ' 17.
ADMINISTRATIVE CONTROLS SPECIAL REPORTS Special repus o w ha re uired covering inspections t .
. ance activities. These specia r '
n an individual basis for each unit and their n and submittal are in the Technical Specific NRC. Regional M ~ 's tn+oe 6.9.2 Special reports shall be submitted to the Direc-tor-cf the Office of-
-Inspection and E-nforcement-Regional-0M-ice-within the time period specified for each report.
6.10 RECORD RETENTION In addition to the applicable record retention requirements of Title 10, Code of Federal Regulations, the following records shall be retained for at least the minimum period indicated. - 6.10.1 The following records shall be retained for at least five years: I s k hon
- a. Records and logs of-une operation covering time interval at each I power level. '
- b. Records and logs of principal maintenance activities, inspections, repair and replacement of principal items of equipment related to nuclear safety.
Eusurs
- c. All REPORTABLE 4CC"RRENCES--:ubmitted to the Cc-irrien
- d. Records of surveillance activities; inspections and calibrations required by these Technical Specifications.
- e. Records of changes made to the procedures required by Specification 6.8.1.
- f. Records of radioactive shipments.
- g. Records of sealed source and fission detector leak tests and results.
- h. Records of annual physical inventory of all sealed source material of record.
ALL STS6 6-M $0V 2 1981
ADMINISTRATIVE CONTROLS 6.10.2 The following records shall be retained for the duration of the Unit Operating 1.icense: . SYaftoo g
- a. Records and drawing changes reflecting -wst- design modifications made to systems and equipment described in the Final Safety Analysis Report.
- b. Records of new and irradiated fuel inventory, fuel transfers and assembly burnup histories.
- c. udoloveall Records of radiation exposure for all individuals entering radiation y control areas.
coninfid
- d. Records of gaseous and liquid radioactive material released to the environs.
staNort
- e. Records of transient or operational cycles for those-unit components identified in Table 5.7-1. .
- f. Records of reactor tests and exp^eriments.
- g. Records of training and qualification for current members of the
-a st staff.
s w en
- h. Records of in-service inspections performed pursuant to these Technical Specifications.
- i. Records of Quality Assurance activities required by the QA Manual.
- j. Records of reviews performed for changes made to procedures or equipment or reviews of tests and experiments pursuant to 10 CFR 50.59.
SoRC NSA RC
- k. Records of meetings of the WRG) and the (CNRAC).
d
- 1. Records of the service lives of all $hwle&
bb r alisted mewaal snahe>s ir. Table: repsna % 2.7-4c speu4uam 7.7P " ' '* including the date at which the service life commences and associated installation and maintenance records. l
- m. Records of secondary water sampling and water quality.
s ! y / - ( 6.11 RADIATION PROTECTION PROGRAM Procedures for personnel radiation protection shall be prepared consistent with the requirements of 10 CFR Part 20 and shall be approved maintained and Jhered to for all operations involving personnel radiation ex,posure, f m
- n. ecords of analyses required by the radiological environmental monitoring program that would permit evaluation of the accuracy c the analysis at a later date. This should include procedures effective at specified times and QA records showing that these procedures were followed. j
.NOV 2 1981
ADMINISTRATIVE CONTROLS
/
I 6.12 HIGH RADIATION AREA 6.12.1 In lieu of the " control device" or " alarm signal" required by paragraph 20.203(c)(2) of 10 CFR 20, each high radiation area in which the intensity of radiation is greater than 100 mrem /h?"but less than 1000 mrem /hr
- shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a Radiation Work Permit (RWP)*. Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the following:
- a. A radiation monitoring device which continuously indicates the radiation dose rate in the area.
- b. A radiation monitoring device which continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received. Entry into such areas with this monitoring device may be made after the dose rate level in the area has been established and personnel have been made knowledgeable of them.
- c. A health physics qualified individual (i.e. , qualified in radiation protection procedures) with a radiation dose rate monitoring device who is responsible for providing positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified byathe facility Health Physi,cist in the RWP. @
'( s heain guiu pol.7.a me.~ cia.,
6.12.2 In addition to the requirements of 6.12.1, areas accessible to personnel with radiation levels such that a major portion of the body could receive in one hour a dose greater than_1000_ mrem shall_be provided _with locked doors to prevent unauthorized entry,5end-the4ey; shall te maintained "nde the e minic- see. '
-trative centrol Of the-Shift Foreman on duty 2nd/cr health phy:ic cuper"isier, ec Q Doors shall remain locked except during periods of access by personnel under an approved RWP which shall specify the dose _ rate levels in the immediate work area and6 he map mum allowele :tay time fer individuals in that arca. For See individual areas accessible to personnel with radiation levels such that a L5ed' E~-
major portion of the body could receive in one hour a dose in excess of ~ 1000 mrem ** that are located within large areas, such as PWR containment, where no enclosure exists for purposes of locking, and no enclosure can be reason-ably constructed around the individual areas, then that area shall be roped off, conspicuously posted and a flashing light shall be activated as a warning device 2 In lieu of the stay time specification of the RWP, direct or remote (such as use of closed circuit TV cameras) continuous surveillance may be made by personnel qualified in radiation protection procedures to provide positive ex
&posurecontrolovertheactivitieswithinthearea.
ov % area sats he ~Ae* se avect co& rot at healHe physics pe<so.@ @
" Health Physics personnel or personnel escorted by Health Physics personnel shall be exempt from the RWP issuance requirement during the performance of their assigned radiation protection duties, provided they are otherwise
, following plant radiation protection procedures for entry into high radiation [ (, areas.
** Measurement made at 18" from source of radioactivity.
26 ALL STS 6-N
'NOV 2 1981
INSERT VI t Access is administratively controlled through the issued RWP and key authorization by the Shift Superintendent and/or health physics qualified technician. INSERT VII Any additional radiological controls (e.g. staytimes, postings, barricades, etc.) to limit individual exposure. b
ADMINISTR.TTIVE CONTROLS h ( 6.13 PROCESS CONTROL PROGRAM (PCP) @ 6.13.1 The PCP shall be approved by the Commission prior to implementation. 6.13.2 Licensee initiated changes to the PCP:
- 1. Shall be submitted to the Commission in the Semiannual Radioactive Effluent Release Report for the period in which the change (s) was made. This submittal shall contain:
- a. Sufficiently detailed information to totally suport the rationale for the change without benefit of additional or supplemental information;
- b. A determination that the change did not reduce the overall conformance of the solidified waste product to existing criteria for solid wastes; and
- c. Documentation of the fact that the change has been reviewed and found acceptable by the
- 2. Shall become effective upon review and acceptance by the g 6.14 0FFSITE DOSE CALCULATION MANUAL (00CM) _@,
6.14.1 The ODCM shall be approved by the Commission prior to implementation. 6.14.2 Licensee initiated changes to the ODCM:
- 1. Shall be submitted to the Commission in the Semiannual Radioactive Effluent Release Report for the period in which the change (s) was made effective. This submittal shall contain:
- a. Sufficiently detailed information to totally support the rationale for the change without benefit of additional or supple-mental information. Information submitted should consist of a package of those pages of-the ODCM to be changed with each page
- 7 number;d and-prov % d "ith an approval and date bow, together with appropriate analyses or evaluations justifying the change (s);
@ ba+cA au codasm%g % vevsdon mdeD
- b. A determination that the change will not reduce the accuracy or-reliability of dose calculations or setpoint determinations; and
- c. Documentation of the fact that the change has been reviewed and found acceptable by tne .
M soAC-
- 2. Shall become effective upon review and acceptance by the (URQ s
N PWR-STS-RETS 6-) 1/4/83
i ADMINISTRATIVE CONTROLS j c Jt' l 6.15 MAJOR CHANGES TO RADI0 ACTIVE LIOUID, GASEOUS AND SOLID WASTE TREATMENT
. SYSTEMS ^
g 6.15.1 Licensee initiated major changes to the radioactive waste systems (liquid, gaseous and solid):
- 1. Shall be reported to the Commission in the Semiannual Radioactive Effluent Release Report for the period in which the evaluation was
. reviewed by the (Ur.it Rem e. C. cup). The discussion of each change shall contain:
s SORC- .
- a. A summary of the evaluation that led to the determination that the change could be made in accordance with 10 CFR Part 50.59.
- b. Sufficient detailed information to totally support the reason for the change without benefit of additional or supplemental information;
- c. A detailed description of I.he equipment, components and processes
- involved and the interfaces with other plant systems;
- d. An evaluation of the change, which shows the predicted releases of radioactive materials in liquid and gaseous effluents and/or
,c quantity of solid waste that differ from those previously predicted in the license application and amendments thereto;
- e. An evaluation of the change, which shows the expected maximum exposures to a MEMBER OF THE PUBLIC in the UNRESTRICTED AREA and to the general population that differ from those previously estimated in the license application and amendments thereto;
- f. A comparison of the predicted releases of radioactive materials, in liquid and gaseous effluents and in solid waste, to the actual releases for the period prior to when the changes are to be made;
- g. An estimate of the exposure to plant operating personnel as a result of the change; and --
- h. Occumentation of the fact that the change was reviewed and found acceptable by the fttRfff' CORC- go g
- 2. Shall become effective upon review and acceptance by the flR6t.
- Licensees may chose to submit the information called for in this Specification as part of the annual FSAR update.
(. \ 11 PWR-STS-RETS 6-X 1/4/83 A}}