Text

Marked Up Pages for Proposed STS 3/4.6,3/4.7 & 3/4.9
	ML19326B363
Person / Time
Site:	Arkansas Nuclear
Issue date:	01/01/1977
From:	; ARKANSAS POWER & LIGHT CO.
To:	;
Shared Package
ML19326B362	List: ML19326B361; ML19326B363;
References
	NUDOCS 8004150740
	Download: ML19326B363 (156)
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INDEX , I l l LIMITING CCNDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PA_f B&W - ATMOSPHERIC TYPE CONTAINMENT 3/4.6 Rynewt Bowswm0NTAIN" NT-SYSTEMS. o _ n%sau. me ava. .,wwa \ 3/4.6.1 r nu. .v. ...v.v

                                                        .,u-u
                           . m.u.nu.
                                 .mm                              r                                                                                        l paw Rai -

vont yeh'innen tBuild.'n.te g r i ty . . . . . . . . . . . . . . . . . . . . . . . . . .3/4 . . .6-1J

                                                                                                                                               ...         l
                         ,.c a t a i nce n t Le a ka g e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 6-2J grocto- 5<    Te3eitka-?crat" = ...................................... -3/4 04-Cen te it.mun i- Sfruc tural In tegri ty. . . . . . . . . . . . . . . . . . . .                             3/4 6-9J O
   '      3/4.6.2
                         ,DEP.,           u SSURILATIONANDCOOLINGSYSTEMS   .

System. .Og % eN r J v.u n w g .. i.a ; L...mc. t
r. pr~ay .4.m - n .....+ L .,,,/,..................
3/4 6-llJ e M6*4 s.<, tve system................................ 3/4 6-13J go.e s+
.ca ta i ni.an t tool in g Sysi.em. . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 6-15J
_m...- ,, - ...... -..- . .. sf,.u.a abuina ut.wnn v r , , , , , , , asarco................................ ,, , u .vu REAc'rvR BU"@ !MS 3/4.6.4 {0NTAI""ENT--ISOLATION VALVES......................... 3/4 6-18J 3/4.6.5 COMSUSTIBLE GAS CONTROL Hydrcgen Analyzers................................... 3/4 6-21J
-Cl ectric l!yJi vg.. 'cc;-Binec5 '
n' ................... 3/' S 22J Hyd rogen Pu rge frkeme Sy s tem . . . . . . . . . . . . . . . . . . . . . . . . 3/4 6-23J
-- "ydrogen "ixing Systcr................................ -2/^ S 25]
VEN Ytt. h 7"lDN 3/4.6.6 PENETRATION ROOM CX"AUST AIR CLEANUF SYSTEM.......... 3/4 6-26J __,n.., e,
...-..m.. .n uv v. . u.
n-,,-- m m.tna.......... in _ , , , , - .
....................... s,, e vo n1 O_
e v - . . . .-- -
\_
B&W-STS 8 0 0 415 0 740 *I-d U^""*"Y l' I977
i l l
% l
?fO J V) INDEX l 1 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS l SECTION PAGE 3/4.7 PLANT SYSTEMS l 3/4.7.1 TURBINE CYCLE - Safety Valves........................................ 3/4 7-1 p,.c4,i
;C c r} Fe ed wa te r Sy s t em . . . . . . . . . . . . . . . . . . . . . . .3/4 . . . 7-4 Condensa te S torage Tan k. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 7-6 Ac t i v i ty . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/ 4 7 - 7 Bio c.k Main Steam u nc Mc!:tica Valves..................... 3/4 7-9 c___ m . .. ,<- u- , ,. , ,m swww.svusy ri u w s . u.nu m sa wi f . . .. . . .. ... .. . .. . . . .. . . . ... .. -v/ t i av 9 r4 7 6 e+r,u r. r s , r n a vn n nnre-.in- ,+--In.~..- . ...-- ---.. - .- s,__
si T . s .b a i c ru v wiv uiv i i u n i e u., J J V nL/ I Lnr Lnn i UNC L AflA 4 n i A UTT.. . . . . . J/'+ / 4J M 14 7 9 ranunnurue ennt vue i t A * ,- a. raue?ru w/ r .s u V4 44 y ai6Ig I _ 9 r .4 7 $ . bvm 6 A 51 Ad FT FT 4 6 4 4 Jtwy b I. . . . . ................... 4% Wg 7 1
/'] %/
3/4.7.4 SERVICE WATER SYSTEM.................................
/A F J') f *t . "I . C Egrasewcy co.at.sser meo J J l t vT. t.1DJ -i A T.
w s r i f, rw.i1 M"" f* t ti tr w4 sin. . .................................. J 3/4 7-15 p/ 'A # 1f 9 t "F su
\'
f* f ^^^ M/A.7.e y - a nn^-"Cb6Aun.......................j .............. wvvu i nV 4
- ^ * * *
912 a/ 7 's 'i l G3M >#9**eMJW49/A/K k! List 'rroW 3/4.7.7 CONTROL ROOM EMERGENCY AIR - CL;,"NUr ",'l';T;". . . pys.v.W. . 3/4 7-18
_,y , n -,.,,,,..m me-.. - . . . . . . . - ..- - . - . . . . . - -..- -.. ,, , ,,
. , , , . , .m m%s r uar avun unnugi nin uuunnar aia u.n............. s , ,. . noc a vor n s. s, , u. nn ,x, , ,, s, anenc r r Re,,ssoer (s u s w e.n < ci 3/4.7.9 . . su. vm.om ..
ru.vummms.................................... 3 4 7-24 3/4. 7.10 SEALED SOURCE CONTAMINATION. . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 7-28 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1 A.C. SOURCES O p e ra t i n g . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/ 4 8 - 1 c- u,. a _. _
~r.............................................. -ai ,. a-a 3/4.8.2 ONSITE POWER DISTRIBUTION SYSTEMS A.C. Distribution - 0perat Sg........................ 3/4 8-6 *r ,i...
n:..
-.s.
m..:__
.-.. m..u.......................... w4 or c.u. sum D.C. Distribution Cp m ti v........................ 3/4 8-8 g m..,. no. . .... _ . - -.m.. .c u, _. " - n............................
3/4 8-10 p jl l B&W-STS VII January 1, 1977
t i f % \~ _INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREME _SECTION PAGE 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 pc rmt StHt-DM6 3/4.9.4
. ON . A liuG T PEN ETRATIONS . . . . . . . . . . . . . . . . . . . 3/4 . . . 9-4 3/4. 9. 5 COMMUNICATIONS...................................... 3/4.9-5 a-sn, . ms.v- ,r.-,
uuu
. . . ~ , ....~u.,.,.~ nn, ge n n~ u n n.. e. .n.a.n v. i , , u. . . . . . . . . . . . . . . . . . . . n, . ~i , nru m 3/4.9.7 AH t /L ih R'Y CRANE TRAVEL - 5"""' "UCL "TORAGE TGGL BUILDING..... 3/4 9-7 3/4.9.8 COOLANT .=:~ . _ _ p:r CIRCULATION.................................
r- 3/4 9-8 r . a r m i n i-v &,,.m_n*m = o_ si n, . ^s . 2 a -un miv. ,s, v. .u u- mi t n o. .n- en,w a ,, u.unu-, .. ....tnu
. . . a ev.erui. .....
u'i n' va
^^ , ,- m s i ,. .s. .,-- . - , , - . n-.--,,,
v -o n n um.uu , , - - - - , nunw.vn , , , n ,
., ,, _. .ussa........................ si , r i O-s f ,. ..a. , , i.;r s, m n . ,. -
sivnnuu ,nvus n mnn.un
,,. ,-n . - , , - . -,i1 n-3, uuiw............................ 3, 3, -
3/4. 9.12 'T0".""" Ptsu HANDun MEA VENTot. ArMN "00L ^.!" CL"ANUF SYSTEM. . . . . . . . . . . . . . . . . . .3/4 . . 9-12 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 GROUP HEIGHT, INSERTION AND POWER DISTRIBUTION LIMITS......................... 3/4 10-1 3/4.10.2 PHYSICS TESTS....................................... 3/4 10-2 I 3/4.10.3 REA CTOR COOLANT LOOPS . . . . . . . . . . . . . . . .3/4 . . 10-3
. . . . . . .l . . . . .
l 3/4.10.4 SHUTDOWN MARGIN..................................... 3/4 10-4 t/4.10. 5 mtwMurt Tcm PCR A ruRE Poe, c htrs cau T v . . . . . . . . . . . . 3/4 20 .S
\ \
w i ' ( , B&W-STS VIII
\
January 1,1977
){v ~
INDEX BASES SECTION_ _ PAGE 3/4.0 APPLICABILITY.......................................... B 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTR0L..................................... B 3/4 1-1 3/4.1.2 B0 RATION SYSTEMS...................................... B 3/4 1-2 3/4.1. 3 MOVABLE CONTROL ASSEMBLIES........................... B 3/4 1-3 3/4.2 POWER DISTRIBUTION LIMITS.............................. B 3/4 2-1 A i 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION............ B 3/4 3-1 3/4.3.2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION.................................... B 3/4 3-1 3/4.3.3 MONITORING INSTRUMENTATION........................... B 3/4 3-2 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT L00PS................................ B 3/4 4-1 3/4.4.2 and 3/4.4.3 SAFETY VALVES............................ B 3/4 4-1 3/4.4.4 PRESSURIZER........ ................................. B 3/4 4-2 3/4.4.5 STEAM GENERATORS..................................... B 3/4 4-2 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE....................... B 3/4 4-4 3/4.4.7 CHEMISTRY............................................ B 3/4 4-5
') 3/4.4.8 SPECIFIC ACTIVITY.................................... B 3/4 4-5 Q((V 3/4.4.9 PRESSURE / TEMPERATURE LIMITS.......................... B 3/4 4-6 g 3/4.4.10 STRUCTURAL INTEGRITY................................. B 3/4 4-11 B&W-STS IX June 1, 1976
9- INDEX
, BASES SECTION PAGE 3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3/4.5.1 CORE FLOOD ING TAN KS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 5-1 3/4.5.2 and 3/4.5.3 ECCS SUBSYSTEMS.......................... B 3/4 5-2 l 3/4.5.4 BORATED WATER STORAGE TANK .......................... B 3/4 5-2 1
1 Ns-B&W-STS X June 1, 1976
Pm l f g V INDEX BASES , t B&W-ATMOSPHERIC TYPE CONTAINMENT l SECTION f PAGE 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 fif7MTu*g,.mau.t.o. ..w.s.
.n.....u................................... . . . . . ~
B 3 4 6-lJ 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS................ B 3/4 6-3J
. .c . . ,,,,...,......-,.m,,. ......m u .. .
arnewn nun w** ..._............................. ,,,,c. -,, I 3/4.6.4
- CONTAI;;;EiiT ISOLATION VALVE. . . . . . . . . . . . . . . . . . . . . . . . .
B 3/4 6-4J 3/4.6.5 COMBUSTIBLE GAS _ ,s/,4. .,c.,.G PENC CONTR0L............'................. B 3/4 6-4J
<< . . . . n eur,.TR.AT, .-.m.n. m . . m. - ION, -m ROOM. . VEN..Til A r.t.cN JY.Swin . . o v v. . um.nue.
nm -~m.., w ,v. es ,2/,4, 6-45, 4 s.a.un......... _n. -, .e m I
$f. , 7 ., .".. 9.ioMmf*tut Im. ,F t ,P**
a. A. ,, m m *8.**a un u o""^.._..............................
, _9 9, ,* 41 s . s , ,A . ..
a %O J) s.-
., B&W-STS XI-J September 1,1976 -
i.
/' T -
4%*) INDEX BASES SECTION PAGE 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE.................................... B 3/4 7-1 9 e4 5 a .r t f= A 1A PeteFMAv^n M M P e e t in f* iTf*1An r* n A v1 in F t T ha T T A T T AB.f n *9 IA *? 9
.# / "T . d .L J u n.a i s wbsik nFT 5 Vs 4 a aw.s w w a mg a ks se hanaUn6 6&45& Isi 6 VI Y . .W W/ 7 d W *) rA v 6 f.naannaa r ant eAne v ans. i n n e r= n e u eveta _ n 9 a =r e sg i.s.4 w vu su vn 6 31 5 vvva. A s vu vvn a h.n .JBJ54.a....... ........... __ U .J f 'T d .#
3/4.7.4 S ERV I CE WATER SYSTEM. . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 7-3 e m u euer cm.ms ru, g SfA 3/4.7.5 i tt v f sm a v == w 6sAsIn t 6 I s 6n s t e r" A T P f e n tf
.rtivn............................... w J/ *t =J a " *At 9 i4 *? e et ^^m n no v e- ev . a ma n a a a = = .s g T . s.W s 6VVw a n w 4 b Y 4 i v il . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . W af 't / " 't am counm*uan/da meressriew 3/4.7.7 CONTROL ROOM EMERGENCY -AI,", CLEANU^ SYSTEM. . . . . . . . B 3/4 7-4 - I 3
3 g - n .-rer nom.m n n nan evn e m o ne, a.- -- -
...a -ua- .. n - - - - - ., ,r , . , . m mems < vor nuvo un.nus, nin vec.nnur aia. c o........ o af , e-a l .
(' SHoct surecesson.s (snuesun') 3/4.7.9 HYDRAULIC "NuBSERS............................... B 3/4 7-5 3/4.7.10 SEALED SOURCE CONTAMINATION...................... B 3/4 7-6 3/4.8 ELECTRICAL POWER SYSTEMS , , , , , , , , , , , , , , , , , , , , , , B 3/+ if-l n 1 A P P h t 1 R P r* tt n R D t% 1 9,f s ,M .v.. n.m. svunees..................................... u a f ,A v.
,, n , n.,..v- nn . n n . e v n . M . ,. . ,, . . -m,-v-... M ,, n ,
s i ,A .v.c vnasiu runun viainivviivn aiascoa................ v si ,A vs 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 prAc-me sumite:r l 3/4.9.4 C0TAINi'ENT- PENETRATIONS . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 9-1 l 3/4.9.5 COMMUNICATIONS................................... B 3/4 9-1 l
/~T
[QJ %. x/ B&W-STS XII September 1, 1976
>~
[J INDEX BASES ' 1 SECTION PAGE m n c . . - , . . . . . . . . . , . mm.n, ---.m.. J ,, 7,
.-u -.--
s.w i u t i, innnuuanu unavuw vrtn.wawaas................... u J/ 4 7-4 fin so Lin RY 3/4.9.7 CRANE TRAVEL - S";';T ."'J:' ST0"ACC BUILDING. . . . . . . . . B 3/4 9-2 3/4.9.8 C00LANT CIRCULATION................................ B 3/4 9-2 asa m n mese=aH> anu we m a n ne - * * - a v. s . s te w , ent n _ vr=wii n kny e r im f. e nwni euv a w v un J/ 7 s.J I wnuw ns1 U hsu inwa 4 a Tv.vi nt,e e v t* ? ? 11_ w viwia.. ..
,fA w sf 7 M &
s a.
,,,o s. - . . . .- -- .-i,,, m_.---- ..,y, . men sg 7 J.6U EIllu J , m /.4.J.6i- . .MM6Ln . ....._ LLv6L nLM'wtun v6JJLL ne i u
_n
n. o_
s i unnur. ruv vensLn ,,4 L6v66........................... u v, - M M 94hN D Lm~l%G....A, RC A WWTIL ATib H 3 /4. 9.12 , a,.m . -- svnnwu ruww nan wwLnnur
. . - . . . . . cme,u n ,, ,
JTJitn.................... D J/ 7 , Jo_J 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 GROUP HEIGHT, INSERTION AND POWER DISTRIBUTION LIMITS........................ B 3/4 10-1 3/4.10.2 PHYSICS TESTS...................................... . B 3/4 10-1 3/4.10.3 REACTOR COOLANT L00PS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 10-1 l , 3/4.10.4 SHUTDOWN MARGIN.................................... B 3/4 10-1 ; 3/4. so. S minorsum rtnesAArutzt for cRmcaury 6 W4 lael } t t
~ \.- j r
B&W-STS XIII January 1, 1977
)
r D i
E i 1 g INDEX ' 1
' , ) 's DESIGN FEATURES SECTION PAGE 5.1 SITE Exclusion Area............................................ 5-1 Low Population Zone....................................... 5-1 l
wa r,m a.mmm: m-nu,c
5. 2 um . m. eu.rn. mr mn Configuration............................................. 5-1 Design Pressure and Temperature........................... 5-4 l
5.3 REACTOR CORE Fuel Assemblies........................................... 5-4 Control Rods.............................................. 5-4 P 1 5.4 REACTOR COOLANT SYSTEM De s i g n Pres sure a nd Tempera tu re. . . . . . . . . . . . . . . . . . . . . . . . . . . 5-5 Volume.................................................... 5-5 l z..r ..mm. ciu me , m umumm. m .n.
. m..,..,u.
o,. m -vom. vn.............................. M l I 1 l 5.6 FUEL STORAGE l Criticality............................................... 5-5 t' 0rainage.................................................. 5-5 i Capacity.................................................. 5-5
...m,,-... ! ,.e cunruntni u n.u u un ..- s.un, ,,o.,,........................
m.. . 6 ' l l l h./"'T,w.] -
'- B&W-STS XIV June 1, 1976 l
l i
0 o SECTION 3/4.6J CONTAINMENT SYSTEMS SPECIFICATIONS FOR BABC0CK AND WILCOX l ATMOSPHERIC TYPE CONTAINMENT 1 9
\.
I
RtncroR BmLotu& s 3/4.6 CCNTAU,MCNT SYSTEMS REACTOR BurLDING-3/4.6.1 TRIMARY C0iiTAIIe;ENT REncroR BsitL0tNG C0iiiAlie;Elii INTEGRITY LIMITING CONDITION FOR OPERATION RcacrpR BulLQ1NG 3.6.1.1 Primary CCNTAU;MCNT INTEGRITY shall be maintained. APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: RcAcroR BLitoING- REACT 0L BUILDTNfr Without prina;y CONTAIN".CNT INTEGRITY, restore CCNTAIieiEiii 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
6. REAcreR ButLbsHce
4.f. l .1 Primary CONTAIN"CNT INTEGRITY shall be demonstrated:
a. At least once per 31 days by verifying that:

1. All penetrations
not capable of being closed by 0PERABLE reuk Wilfq entai= cat automatic isolation valves and required to be l closed during accident conditions are closed by valves, blind flanges, or deactivated automatic valves secured in their positions, except as provided in Table 3.6-1 of Specification 13.6.4.ll. I
2. "il equip ::t h:tche: Or: cle::d :nd :: led.
ceaclerboEss
b. By verifying that each -- '"---' akr lock is OPERABLE per Specification 13.6.1.3l.

c. rat ep;pment hatek 's verIWed c.los<J and scalec0 a.#er eack oP**bg or a& least once per 24 monfks.
*Except valves, blinglggand deactivated automatic valves which are located inside thel......_... and are locked, sealed, or otherwise secured in the closed position. These penetrations shall be verified closed during each COLD SHUTDOWN except that verification of these penetrations being closed need not be performed more often than once per 92 days.
(' B&W-ATMOSPHERIC 3/4 6-lJ January 1, 1977 1
I 1 REACT 0kBUILbN & O GONfA NMEM SYSTEMS I gracre+A cuitow V CC:!A :;;Ct;T LEAMGE LIMITING CONDITION FOR OPERATION Reader WAJ'iy 3.6.1.2 C;atsi.;c...-leakage rates snall be limited to:
a. An overall integrated leakage rate of:
nachnhellLy
1. 1 L a, k0.20} percent by weight of the centch-en. air per 24 hours at P,, -(-50-) psig, or O re.Jw ba'Abg

2. l i t , {0.10k percent by weight of the centai ..-cat air per 24 hours at a reduced pressure of Pt ' PS 9' 40

b. A combined leakace rate of < 0.60 :.4 for all penetrations an'd valves subject to Type B anif C tests, when pressurized to P3 . I APPLICABILITY: MODES 1, 2, 3 and 4.
ACTION: With either (a) the measured overall integrated : nt in-;nt leakage rate ( exceeding 0.75 La or 0.75 Lt, as applicable, or (b) with the measured " combined leakage rate for all penetrations and valves subject to Type B and C tests exceeding 0.60 La, restore the leakage rate (s) to within the limit (s) prior to increasing the Reactor Coolant System temperature above 200 F. SURVEILLANCE REQUIREMENTS . ruchn %E\hg 4.6.1.2 The ;;atei.... cat leakage rates shall be demonstrated at the following test schedule and shall be determined in conformance with the l criteria specified in Appendix J of 10 CFR 50 using the methods and provisions of ANSI N45.4 - 1972:
a. Three Type A tests (Overall Integrated 0;at;iarcat Leakage Rate) shall be conducted at 40 + 10 month intervals during shutdown at either Paf60-) psig, or at Pt f#fi&) psig, duiing w shall be conducted during the shutdown for the 10-year plant
^ inservice inspection.
5 eack Jo year service perlo8. The bird test of each set b) pj - k- B&W-ATMOSPHERIC 3/4 6-2J September 1, 1976
REAc7OR BulLDlWG-C0';TAIM O T SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) NQ periodic Type A test fails to n:eet either .7
.75 L schedule for subsequent T ests8allbe revieh,~dandappro the Comn . If two consecutive Type A tests fail to mee L or .75 L , a Type A test shall be pe at least every hst until two consecu pe A tests meet either .75 L or . which e above test schedule may be resume 8.
b. 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, or 0.25 L t* I
2. Has a duration sufficient to establish accurately the change in leakage between the Type A test and the supplemental test.
O.
3. Required the quan esular%O L $
( orbledfromthe,t,1,tgagaginjectedintothe-eentcimat
=...mc... during the supplemental test to be equivalent to at least 25 ercent of the total measured leakage rate at Pa '
s psig, or P t ' N PS1 ' 9 c /. Type B and C tests shall be conducted with gas at P k psig at intervals no greater than 24 months except for t$sts involvingt air \oc.ks,
1. fi- leckt. l

2. o enetratiene using centinueur !eekage writering syst e 1

3. Y:ht; p=::eri:cd =f th fluid f= 0c:1 sy;t=.
dp. Air locks shall be tested and demo,nstrated OPERABLE per Surveillance Requirement f 4.6.1.30
f. y dic tests are not required for ons continuously mo th ent Isolation Valve i and Channel We zat vided the systems E per Surveillance Requirement y
B&W-ATMOSPHERIC 3/4 6-3J January 7, 1977
RncTOR *aktt ONC-i l ) i G0MA MMEM SV3TEMS ' v l SURVEILLANCE REQUIREMENTS (Continued)
<; . kage from isolation valves that are sealed with flui a sea tem may be excluded, subject to the pro ns of Appendix J, ion III.C.3, when determin e combined leakage rate prov the seal syst valves are pre:isurized to at least 1.10 P i the seal system capacity is adequatetomaintain,s( pr e for at least 30 days.
h. Type B te or penetrations employing a c uous leakage mo ng system shall be conducted at P '
at l ntervalsnogreaterthanonceper3yearl,(50 . e f. All test leakage rates shall be calculated using observed data converted to absolute values. Error analyses shall be performed to select a balanced integrated leakage measurement system.
f. [ f. The provisions of Specification 4.0.2 are not applicable. l O\
1 l l ( IBall-ATMOSPHERIC 3/4 6-4J January 1, 1977
~ ^ -' - ,--r -vM -'
l l p, I 1
) ilEAC7bt 8MILDl%
b/ EON""NENT SYSTEMS
)StncTax Bull 3tN6 ONTM N't:t3 AlR LOCKS LIMITING CONDITION FOR OPERATION 3 . 6 .1. ') Each E$$ En rt ir le;k shall be OPERABLE with:
a. Both doors closed except when the air lock is being used for normal transit entry and exit through the c ntaim:n*Y"G'en at least one air lock door shall be closed, and **b ?

b. An overall air lock leakage rate of 1 0.05 L, at P,, psig).
A,PPLICABILITY: MODES 1, 2, 3 and 4. ACTION:
,. .ter bu#dly
a. Wi th one se .ta ......s..t air lock door inoperable:

1. Maintain at least the OPERABLE air lock door closed and'
{ l either restore the ir perable air lock door to OPERABLE ( status within 24 hours or lock the OPERABLE air lock door closed.
2. Operation may then continue until performance of the next required overall air lock leakage test provided that the OPERABLE air lock door is verified to be locked closed at least once per 31 days. -

3. Otherwise, be in at least HOT STANDBY within the next six hours and in COLD SHUTDOWN within the following 30 hours.

4. The provisions of Specification 3.0.4 are not applicable.
, reuter k;iding
b. With the contair: cat 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 six hours'and in COLD SHUTDOWN within the following 30 hours.
SURVEILLANCE REQUIREMENTS 4,.6.1.3 Eachn.ew cc ;tainunha cat gir lock shall be demonstrated OPERABLE: ( I sf ARKANSAS-UNIT / 3/4 6-/
p { VEACroRBMnonss. b !CC"T'"i".:"T- SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
a. *Af ter each opening, except when the airlock is being used for multiple entries, then at least nce per 72 hours, by verify-ing seal leakage no greater tha cc/ minute whtn the volume between the door seals is pressurized to > P the rate is determined by either pressure de8ay( for at M sig) and least 15 minutes or by precision flow measurement when measured for at least 30 seconds with the volume between the door seals at a constant pressure of g psig,

b. At least once per 6 conths by conducting an overall air lock leakage test at P (6#$sig)andbyverifyingthattheoverall air 1.ockleakagefateiswithinitslimit,and

c. ^t le::t On:: p:r 5 nth by ::'fy .g th:t erly en: deer i-
P Of- le:E ::r be Opened :t : time. ;
i p 1
=
Exemption to Appendix "J" of 10 CFR 50.
O-v ~ (. l ARKAtl5AS-UtilT / 3/4 6-5E T
O CONTAINMENT SYSTEMS 00".!"5 "' !!^L*?!0" ""L"E ""O 0"a"". L MELO Pn:CC"n!!"T!0" 0YSTEMS (OPT!On".L) MITING CONDITION FOR OPERATIC
~
1
/
N 3.6.1.4 The contain a nt isolation valve and channel weld pressuri tion systems s 11 be OPERABLE. APPLICABILIT . MODES 1, 2, 3 and 4. ACTION: With the containment olation valve or channel weld ressurization system inoperable, res re the inoperable system OPERABLE status within 7 days or be in a least HOT STANDBY wit the next 6 hours and in COLD SHUTDOWN withi the following 30 rs. e SURVEILLANCE REQUIREMENTS
, s 4.6.1.4.1 The conta ment isolation valve pressur tion system shall be demonstrated OPERA E at least once per 31 days by v ifying that the system is pressu zed to > 1.10 P (55) psig, and has dequate capacity to maintain sy empressureforal,least30 days.
4.6.1.4.2 he containment channel weld pressurization syst shall be demonstr ed OPERABLE at least once per 31 days by verifying t t the syst s pressurized to > P (50) psig, and has adequate capac to main in system pressure Tor,a,t least 30 days. O (
~-
B&W-ATMOSPHERIC 3/4 6-6J June 1, 1976 l l l .. -._ -__ -_---. .-- - - - - - - - - -
l l i 0 REAcut suunme f' GONTMf#tfMtf SYSTEMS d INTERNAL PRESSURE LIMITING CONDITION FOR OPERATION i Tha rauhr %hMn1 D- N y ce tai rnt internal pressure shall be maintained 3.6.1.5 between and peig. 12.o ps; (s.s in. 90 and n,7 p3l, ( a.o ps;y, APPLICABILITY: MODES 1 2 3 :nd 0. K a,st ACTION: ceseh>r ha'.ld n With the centif ent $nternal pressure outside of the limits above, restore tha internal pressure to within the limits within 1 hour.or be in at least HOT STAND 2Y within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. 1 I F SURVEILLANCE REQUIREMENTS reacfor bs % 4.6.1.5 TPc pr'n y ::nt: h;gr.t internal pressure shall be determined to within the limits at least or.ce per 12 hours. t l l Is.
1 CONTAINMENT SYSTEMS - I l NRTEMPERATURE LI JNG CONDITION FOR OPERATION
\ /
3.6.1.6 imary containment average air temperature shall not xceed
*F.
APPLICABILITY. MODES 1, 2, 3 and 4. ACTION: '
~
With the containmen average air temperature > *
, reduce the average air temperature to w thin the limit within 8 hours or be in at least HOT STANDBY within th next 6 hours and in COLD S TDOWN within the following 30 hours.
l f SURVZILLANCE REQUIREMENTS
/ 1 4.6.1.6 The primary contai ment average r temperature shall be the arithmetical sverage of t temperatures a the following locations and shall be determined at ast once per 24 hou :
Location a. b.
c. !

d. !
e. ! , ,~ k B&W-ATMOSPHERIC 3/4 6-8J June 1, 1976 l
,.rW" )
q ~'st i . REAcuR iklLOM s / t w . ., m ,. ...
. ,. .m.... . .S Y Sl.Etu.-
v Rene.rog gwwmq C."'::::'" ST7UCTURAL INTEGRITY (Tpi al L._.'- LIMITING CONDITIONS FOR OPERATI0ff nae buith 3.6.1.7 The structural integrity of the m..to,.ai =nt Yhall be maintained at a level consistent with the acceptance criteria in Specification 4.6.1.7. APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: m.w b i/h n With the structural integrity of the cc7 twat nut conforming 'to the cbove requirements, restore the structural integrity to t:ithin the limits within 24 hours or be in at least H0T STANDBY tiithin the next G hours and in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE REQUIREMENTS _ Ree sto,. BuiWin nseto
-' 4.5.1.7.1 1*-# it Tandons The cc.:..ri.-bu;Ih i 7t t.ndens' structurci integrity shall be anmenstrated ct the end of one, three cnd five yacr3 follcuing the initial ccatcinment structural integrity tast cad at five year intervals theraafte. . The tendons' structural integrity shall b:
da.T.onstrat:d by: N The lift -eff forse eS P c. Cat:rrair.i.m M a r:.praccatative cer.pio d cc 10:30 :'l 1:::d:n.:
'l (5 d:=, 5 vortic::1, cnt' 10 hc0?) e:-r.; .' s t ~ ~ .~ P . ; -
w..-., i .- . r :. . . " .., . , . . .,. . ., ... .a. s. pl .
, , e . . . . .., . . ... . _ .
l incledo D ~Inlo: ding cycla in t:hich each cf case t:ndens ia
i aatar inn.:d M datcraina if any viire: cr ctrcnd; are braher er ji t.anca::. K :he lif t cff icrce ;f :ay en- trden in c':: 0 tal
;l Oga.fic90,1 tic.: i.c qut of ' a pred.ict0d c'auni:MT : ' ~ -~ .. , . . . . . . . . . .';; i.;l 'd,jC Cnd t'.nd:r Or 00ch .3 :" 't ci' 't.d inr . '. . . . V C. aM r. <.3T1c:. n'.- '. if
l ir.. -
c, . uh03 ..andon ce; found . ;:sthiu, t'.; . acrysi;ic:::a ...eyca cay prococ.d ccn:idarin; 20 singic dac's
; cicacy c.: er'aut cad .cccptcble. More tha.4 cna d2fcctive tar: don I tica is cvid3nce of cbar;: 1 h cru ci dr criginal dagr?.df.tica c the "*eg"ie Dngdenct'r Uc ! r[ :.Une: .i.9 d.cg, '.hi Sa cf i:.N'_* D ".Y. Y lcss xct.n'ctha"; du.'ingin miidence tha ? rst thc. 2 tact; of the .dcas, th: num* ace cf to.:fonc c!.2ck4 for lift off force during obsequent tests mcy be reduced P.o 1
( ' representative sample of at least 9 tendons (3 doma, 3 verticd s G]/ : and 3 hoop). B&W-ATMOSPHERIC 3/4 6-9J September 1, 1976 1 I l
o 'd Rmrod outtoiws l46tf*ftHE+ff SYSTEMS SURVEILLANCE REQUIREMENTS (Continced)
b. Removing one tiire or strand from each of the doma, vertical and hoop tandons checked for lift off force and determining -
that over the entire length of the removed wire or strand:
1. The tendon wires or strands are free of corrosion.

2. There are no changes in physical appearance of the sheathing filler grease.

3. A minimum tensile strength value of ;:. (guaranteed ultimate strength of the tendom materiai) for at least three utre or strand samples (one frcm each end and one at mid-length) cut frcm each removed wire or strand.
Failure of any one of the tendon samples to meat the minimum tensile strength test is evidence af abnormal degradation of the ccc.t i n nt structure. reseter M Idig
'4.G.1.7.2 End Aachcranes and Adjccent Concrete Surfcces The structural integrity of the end anchorages and adjacent concreta surfccas shall ba , demonstrated by determining through inspection that no apparent changes have occurred in the visual appearance of the end cnchoraga concreta exterior surfaces or the concrete crcck patterns adjacent to tha and Lcnchorages. Inspections of the concreta shall be performad during tha jTypa A -:e"' W m t lockaga rcto tests (reforanca Specific: tier. 4.G.1.2) l t:hile the rene1 4:.m.e.. U. tt2 i: ct its m=imca test pressure. " .E r r ! 4.3.1.7.3 Liner picta The ctra:turci integr.ty ci tha ;, e.fa.. L.c...-.
r 6 disI i:r.r plate shat: b: dotamined drir.; the chutric.;n fcr cach Ty?: A cc.'n M...w 10chcg2 rr.ta t03t (referca.:n Sp;0ificati:n 0.G.1.2) by c
; vtual inspa tion cf th pit.tc ca. cri.'/ir.g r.3 caparcut ch?.ngas in apparence or c:tcr r*:acr.c.1 <!acr.dati:n. ' ' nactor "' balld.**p
i.3.'. ~ . an t .' .. . .ny c' .::.ud . . .irLde;;o.
r .
4 1,- r tuc; t.at;;c.r. + r:g ;.; ; ct:ay: . y, :rc. Mres : d in::prti r.c
'. '.'. c: rt: ~
.;- ' m.n . c i c. :r "," a r.p: W cui.:n - . /,. iz. 3. .
~
Yic ci : .a : . 2 . i . . . i c: :... .:nd x ceni:.:c. ...;
! con..;.1;r..at.1 dit:::a O .
c:ncr.n: (ccpecif .'/ c c.atn anchoraur): the id: pac-tica src.ricc2: t!.a '.olor'.nc:s '.;! .acking ca the cycractive ::tiens taken.
;j ga. tltotive a~ f YfEc*I r*?*'T ** " & Yk* '*
If'

f ** c h th t a b o * *-
,a reyuir<d inspeetto.e .ae d be >ab~ctte.t to the ca,.ai.,rion pursu se to '\ spesiC.'es tion 4.11. 4 as d shall **Mnst , if eglien ble , breitss wives ,
7% frece- b s in.d het Fee *=y tem
tan , ug,, sagny,g tag tw en e,,d,.
,-.~.n.
u
,,,. a ,m e- ~ -,,._ _r,. ~ , . ~ > %
cwa *** *aye <.td e As y u ie tesdon se.ditiens F;//ee p y. v1;*r._ B&W-ATMOSPHERIC 3/4 6-10J ar, skJune rathi;1976 1,
7 RE Acrat sun.0% F
,,/ CC"!AI;,,I.NT SYSTEMS 1 i
I (, 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS l (-] pa:mtamoma
.0""!""C"i-SPRAY SYSTEM - opsg4 rius j
I LIMITING CONDITION FOR OPERATION
,... to. 6 *ildia f 3.6.2.1 Two independent 0:ntair.cri pray systems shall be OPERABLE .
with each spray system capable of tuking suction from the BWST on an 6S '
suction nt f r:entto the:prqgug,tgp cen. .~ c.._ signal and cut: br Tump.-on =:ticallyFtransferring ted .;;ts :tr:;: t:" 12.
T ear = 6/e a f ' 3 l
-level ;ign:1. E;;h : pray cycts fic- path fcea, th nt in st = rg=cy k
=p chali bc .i rr CPERACLE Jece,:, mat cool c.- '
.d
i APPLICABILITY: MODES 1,^ 2,-- oc,J 4.- p ACTION: .
,. n te h ital . .
With one-contei...s..i. yspray system inoperable, restore the inoperable spray system to OPERABLE status within 72 hours or be in at least H0T i STANDBY within the next 6 hours; restore the inoperable spray system L to OPERABLE status within the next 48 hours or be in COLD SHUTDOWN l within the next 30 hours. r 3 SURVEILLANCE REOUIREMENTS U reae % k ilaf.'sg 4.6.2.1 Each enoinment spray system shall be demonstrated OPERABLE:
a. At least once per 31 days by verifying that each valve ( us!, ,
p= r Opc ted = ut;;;ti;b in the flow path, hat is not locked, sealed or otherwise secured in position, is in its ' 1 i reIulce l e rrcet position. w*f ca Y. l aes.wira 4/' * *fPq'fn y & r*/a *"l*#'N' n pc
b. By verifying. ge are .b.h.ty a F...at er ei'!=, cuhtier each pump daval;;; ;
-d i s aoi se pic25ucc cf ; pri; when tested pursuant to I '
Specification 4.0.5. t
c. At least once per 18 months, during shutdown, by: 1

1. Verifying that each automatic valve in the flow path !
actuates to its correct position on an(0 ntain at spr:y ESU l tathsignal, < j the espablIsTf *h f6
2. Verifying -the4.each spray pumpYstarty automatically on I a4(cv. i.o ;. ..m..t. 5 pre, te5t'r signal . <
E T A s- '. OT l
..- B&W-ATMOSPHERIC 3/4 6-11J January 1, 1977 e .
l 1 E y w pg.q ;.3...n y g 7 g.g. g g y ,_ g
. . . . . ., . , a . . . ~ - = - - -
l
Q REALTDR BMlLblNG 001TAUJ OiT SYSTEMS
\(eh 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS R ocrcR Bon stkG CC*Tf "Pi:9T SPRAY _
SYSTEM - SHUT 00WN LIMITING CONDITION FOR OPERATION one cautoc GA L g 3.6.2.1 -Twe-independent cente! men. spray systemf shall be OPERABLE with each spray system capable of taking suction from the BWST on an ES COntzimentspr:yaqtuationsignalandci.TEtiN1'ytransferring suction to the ::EI En 9 sump.cr : 5 r:ted .::tcr :t r:;; tank 1 1 ';;l ign:1. E::h :pr:3 :y:t:- fic.; p;th f .T. th ::nt:i= nt cc:rgency
=p th:!' be ti: :n 0"E"f"LE d:::y h::t :n:lcr. , .
APPLICABILITY: MODES3 4i 3 and 4. ACTION: na' reuFce h4\N-q With en: : r'21= nt spray system hoperable, rc:::r: the in:p;r:ble p-sy cytt-- te OPEPM'.E st:tu: e'tFi- 72 houn er b; ir, :t 1 ::t "0T STMiOOY within th; neat 5 h gr ; restore the inoperable spray system to OPERABLE status within the next 40-hours or be in COLD SHUTDOWN Q within the next 30 hours. 72-Q' SURVEILLANCE REQUIREMENTS 4.6.2.1 The Eachreefer
t:! = L'am0 S W spray system shall be demonstrated OPERABLE:

a. At least.once per 31 days by verifying that each valve (rzru:1,
-powc eper:ted er : t- stic) in the flow path hat is not locked, sealed or otherwise secured in position, is in its cerrg$ position. noF capMe of aJc~bJ ~1" ac%'A % hs **i" " P' " c^
ne e,peckNh3 o4
b. pump de"elep: a By verifying:that di::h:r; p.. ure of ;
0 r:<f p:!; celatie-when tested#1er, pursuantQm- to specification 4.0.5.
c. At least once per 18 months, during shutdown, by: 1

1. Verifying that e'ach automatic valve in the flow path actuates to its cerrr#* position on an(centi =:nt spr:y E5DS 4ee Q signal. "*"'
ne capaVN.h3 of f*e to
2. Verifying 4het +aeh spray pumpYstartf automatically on an(::ch ! =:nt :pr y t :t)-signal.
ESA5 S
).-
[ B&W-ATMOSPHERIC 3/4 6-11J. January 1, 1977 l
REACfbt BulL0 SNG
;0NTA!M" INT SYSTEMS v 5URVEILLANCE REQUIREMENTS (Continued) . t least once per 18 months by verifying a total leak r I lons per hour for the system at:
1. Normal operati ssure of > psig for those parts of the system downstre the pump suction isolation I
valve.
2. > psi;; for the piping from the containmen e'ncy sump isolation valve to the pump suction isola lon va d/. At least once per 5 years by performing an air or smoke flow test through each spray header and verifying each spray nozzle I
is unobstructed. O q v (. B&W-ATMUSPHERIC 3/4 6-12J , January 1, 1977
I I REAcroA Bau.osure CONTA!N""N SYSTEMS I SCDtdn HYD RosIDE
-GERA#. ADDITIVE SYSTEM (OPTIO"AL)-
LIMITING CONDITION FOR OPERATION a sohm % len;0e 3.6.2.2 The spray additive system shall be OPERABLE with spray addgt9/c . tankf centtia%g at !cact: hav:ng ;a 'indicded a level of .34 0! fe. of /FM weighFa per cen A solu a caateiaed +;onefofbet::en ue!'- sod .m11,300) hg (roxMe. and ( ) gallen ^#' ion containing between (188,300) and ( of I sodium ide (Na0H).
b. A contained volume o twe 500) and ( ) gallons of solution containin .en 0) and ( ) ppm of sodium thios Na2 2 03 ),
S between A45) 2nd ( ) ppn of bo , and between (5,700) and ( ) ppm odium hydrox-
, u.,nu, -..s.
APPLICABILITY: MODES 1 2, 3 a nd . ACTION: f,cdam k3 Arc,;Je With the spray additive system inoperable, restore the system to OPERABLE status withg74;tyours restore or be in at least H0T STANDBY within the next 6 hours; the'_pr 3'ditivesystemtoOPERABLEstatuswithinthenext48 d hours or be in COLD SHUTDOWN within the next 30 hours. SURVEILLANCE REQUIREMENTS ud;. h 3r,, Je 3 4.6.2.2 The 4 pray additive system shall be demonstrated OPERABLE:
a. At least once per 31 days by verifying that each valve (=nt:1, pc.::r Operated er : tratic) in the flow path hat is not locked, o sealed or othenvise secured in position, is i ts ewe'dt position. no+ copble of anfonkhce acWob gWs e.,pired pesa

b. At least once per 6 months by:

1. Verifying the contained solution volume in the tank /. I

2. Verifying the concentration of the NaOH and ":zS0 z3 solution $ by chemical analysis.
O ' V January 1, 1977 w B&W-ATMOSPHERIC 3/4 6-13J
l l REMnk GKswN' 7 EMMidMEftt SYSTEMS (7' v l l SURVEILLANCE REQUIREMENTS (Continued) 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 ane nteirrt :pr:; t::t signal. e.s A s 2
^t 1:::t On:: per 5 ye:r:
by ::ri fying ::F : htier 'le! -"- determined during pre-operational tests) fr following connections in the spray e system:
1. (Drain line 1 ati i gpm.

2. inelocation) i I
1 i i I v i l l i i w p I 's s O; B&W-ATMOSPHERIC 3/4 6-14J June 1, 1976
4 gggypA stutoiWG C0;;TA;:eI:47 SYSTEMS Q>-E r RcRc70R B0\L O" Q CC"'*!""I"T COOLING SYSTEM in n.r v. en --ua' s, LIMITING CONDITION FOR OPERATION rexlcrbu'AOln4 3.6.2.3 At least two independent : nt:f =nt cooling units shall be OPERABLE. APPLICABILITY: MODES 1, 2 and 3. ACTION: reuhr- LOUn'ng With one of the above required :::t ixat cooling units inoperable, restore at least two units to OPERABLE status within 72 hours or be in HOT SHUTDOWN within the next 12 hours. SURVEILLANCE REQUIREMENTS O U 4.6.2.3 At least the above required cooling units shall be demonstrated OPERABLE:
a. At least once per 31 days er ST".CC5"E0 TEST "^ SIS by:

1. Starting (unless already operating) each unit from the control room, l

2. Verifying that each unit operates for at least 15 minutes,

3. "crifying : :: l'ng ::t:r '? = r:t: Of ;
;p- ' :::5 erft ^^!:r.
b. At least once per 18 months by v'erifying that each unit starts automatically (:n l= ;;r ,o upon receipt of an E6 tee 6 signal.
~
l . r I i i J n'
\ .
84W-ATMOSPHERIC 3/4 6-15J January 1, 1977
CONTAINMENT SYSTEMS 3/'.5.3 !00!"E CLE.^"Lm SYSTEu (OPTIO"^1)
\IMITING CONDITION FOR OPERATION /
3.6.3. Two independent containment iodine cleanup systems shall l OPERABL APPLICABIL Y: MODES 1, 2, 3 and 4. ACTION: With one iodine eanup system inoperable, restore the i operable system to I OPERABLE status wi in 7 days or be in at least HOT ST DBY within the next 6 hours and in COLD 'UTDOWN within the following 30 ours. SURVEILLANCE REQUIREMENTS x , 4.6.3.1 Each iodine cleanup syst sha be demonstrated OPERABLE: ! v
a. At least once per 31 days a STAGGERED TEST BASIS by initiating, from the control room, fl rough the HEPA filters and charcoal adsorbers and verifying hat t system operates for at least 10 hours with the hea rs on.

b. At least once per months or (1) fter any structural main-tenance on the HE filter or charco adsorber housings, or (2) following p nting, fire or chemic release in any venti-lation zone c unicating with the syste by:

1. Verif ng that the cleanup system sati ies the in-place test g acceptance criteria and uses the test procedures of egulatory Positions C.S.a. C.S.c and 5.d of Regula-t y Guide 1.52, Revision 1, July 1976, an he system low rate is cfm +10%.
2 Verifying within 31 days after removal that a la ratory analysis of a representative carbon sample obtain in accordance with Regulatory Position C.6.b of Regula ry l Guide 1.52, Revision 1, July 1976, meets the laborato ' testing criteria of Regulatory Position C.6.a of Regul tory Guide 1.52, Revision 1, July 1976. C B&W-ATMOSPHERIC 3/4 6-16J September 1, 1976
CONTAINMENT SYSTEMS RVEILLANCE REQUIREMENTS (Continued) ,/
3. Verifying a system flow rate of cfm 110% durin system operation when tested in accordance with AN N510-1975.

c. A r every 720 hours of charcoal adsorber operati by verify-ing ithin 31 days after removal that a laborator analysis of a rep esentative carbon sample obtained in acco ance with Regula ry Position C.6.b of Regulatory Guide .52, Revision 1, July 76, meets the laboratory testing c teria of Regula- !
tory Posi on C.6.a of Regulatory Guide 1. , Revision 1 July ! 1976,
d. At least once r 18 months by: l l

1. Verifying th the pressure d p across the combined HEPA l filters and c rcoal adsorb banks is < (6) inches Water Gauge while ope ting the stem at a flow rate of l cfm i l . !

2. Verifying that the s em starts on either a safety injection test sign r on a containment pressure
- high test signal
3. Verifying that e filter oling bypass valves can be opened by ope tor action.

4. Verifying at the heaters dis ate +
kw i when tes d in accordance with A. I N510-1975. ,
e. After each omplete or partial replaceme of a HEPA filter !
bank by v rifying that the HEPA filter ban s remove > 99% of ; the DOP en they are tested in-place in ac rdance iiith ANSI N510- 5 while operatirq tbc system at a fl rate of cfm i 10%.
f. ter each complete or partial replacement of a c rcoal dsorber bank by verifying that the charcoal adsor s remove
> 99% of a halogenated hydrocarbon refrigerant test s when they are tested in-place in accordance with ANSI N510- 75 while operating the system at a flow rate of cfm i 10%.
u. B&W-ATMOSPHERIC 3/4 6-17J September 1,1976 l l l
( i __ _ _ -
gracr## Wil.DmG t COTf = E*Y SYSTEMS m. REncron son.omc-3/4.6.4 f"!"5" ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3.6.4.1 cmbeM% The centaix::t 1 solation valves specified in Table 3.6-1 shall be OPERABLE.with itehthe th:: : :500:- 5 T:b h 3.5 1. APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: With one or more of the isolation valve (s) specified in Table 346-1 inoperable, aiaintain 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 48 hours, or

b. Isolate each affected penetration within 48 hours by use of at least one deactivated automatic valve secured in the isolation position, or

c. Isolate each affected penetration within 48 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 4.6.4.1.1 The isolation valves specified in Table 3.6-1 shall be demonstrated OPERABLE prior to returning the valve to service after maintenance, repair or replacement work is performed on the valve or its associated actuator, control or power circuit by perfonnance of a cycling test :nd : r f h:thn cf be h t hn t he. i l g B&W-ATMOSPHERIC 3/4 6-18J January 1, 1977 l
REACTOR BELLOl % O ***aaat*+ sysTeas '( SURVEILLANCE REQUIREMENTS (Continued) 4.6.4.1.2 ' Each isolation valve specified in Table 3.6-1 shall be demonstrated OPERABLE +ri ; th: COLD SM'JT00"M er REPJEUNO MCC at least once per 18 nenths by: E5
a. Verifying that on an:^-tai- :nt kohtien t :t signal, each automatic isolation valve actuates to its isolation position.
k
b. Ver fyi g th:t en : 00 tai m:nt rad h th: high t :t Sign:1, i
5 purg: :nd enh:::t cut-. ti: V:he ::tuct : t its heh t hn petithr.

b. Ve r'sf png +6.+ each manua.\ g\ve. 4kJ 'is no}- Jockek seale[,
or okkerwlse secured 'iv. posi+;on , 'es in i+s ce9uired posi+ lon.
.S.'.l.3 The h0 h th: ti= 0' ::Ch pe ^r Op;r ted er : te=ti' V;h: .' c. . .r.s t ,. .. ..e.
u. u... . .2... _4. ;
... .. u.. . . . . 4. . u. 4 . 4. . , i4_4... ._ u.....
pur::::t to Sp::i'k: tion '.0.5. 1 b O s' L, B&W-ATMOSPHERIC 3/4 6-19J January 1,1977
( TABLE 3.6-1 Riideros haiws- r E -GONTMNMEM ISOLATION VALVES i f VALVE NUMBER FUNCTION ISOLATION TIME "o E ( ) seconds B
" I A. CONTAINMENT ISOLATION 1. \
O
2. Q ff 3 42 ?
B. CONTAINMENT PURGE j O)i OW f , AND EXHAUST 60 fy 1. 2. 6"y *, g g o .1 VSAR l C. MANUAL gg( I'
i. .
2. p lQ D. OTHER I-g 2. I j *May be opened on an intemittent basis under administrative control.
#Not subject to Type C leakage tests.
i 1
~ )
i
W W W.- B SCsDg -25m m m i ,
'~7Ecs( -~ t - /
[':e tres Y{r. %(<a do. %. (Laceties to T. B. ) .
$. CToK fM/t.3 Wt$ .TSN #7"MM 78 CV-Ic54,Ig45 Qateui '& k Gu $yla (sos 4.n,Me ; og45- 0.t rik 4 c d- 1110,1111, KCPs To i 4kter { on so, it'st < !**rs o !* 73 -Zus/4 d I1*I1, c172,12.14 L o t W - da s'clde )
ll c y - 484 s, 4to4 Vee 7 // enter (4703 Tus-lJa. ; 4 s w - d a rtrl/e, l
/4 cv- 1:4,ontt, id7Vowa f* % ( x14 , /L I4 - E ssf*4 )
IL no i e n 16 - de fr**de ) 15 cv- uss,7454 l'r Ltis Lie blfor (14ssyr-s:de; Wrf- Afrfr Ve] w
& si ev- Joes Garwei % k. t % ;m. uLtee Syty (Gats *de) 46 c t - 14 tt , seis. Fin 4er- (st/-dah*/e ; sca npVe) y n %* 4i C y - 1447 Wifrey en .57p/r ( On 6rde) 47 c v - 1.1ss cWv coa /,*s, Ala fre (daf;iVe) -
I .sy c y - 42*2. ChoY/e/ JJatee tu. Gvkr.1 (def.tn/*) S2. cv - 2a.r4 xcP l Go//y AN!>r { 0<tside) l f4 cy-12s2 %. Cen/ impt o loffm Co/er.r (defyside) of ct -424, 4 a.42 car */h,/ Mfor {isor-Antsu'Je, esosth'/s.> 6e ct .use s.nzzi ZC tJ (as ao - datelde -2221.E.pVe) j 3 4.1 c v - uit , .nass .rc sJ ( 11. s + - db'de ; s.sss' - % 'd e ) i 68 er - 4es, m pp s'a~p De<;~ (1446 -r-side; 444*-oortsWe) l 7e cv- avs, ws2 QuenA Ta' l *Pra:s (** cs -dartide , too -2,,A > l i.
I I w 5 GA y
'} i2 y i s 2 ^
4 v 4 y' $ M
. .?
5 .!
e 1
N t
& M +~ ~%! v J e i s 'Y E k 4 d
d d ? b Mg w hf. 4 Y I , w e 1 < L g g $ On n e 3.v k (
% y e :
h 3 $ i
~%
w Q N b s _A s b 4-uti , I __ - - - . , ~ - _ _ _
PN ta'a~ d. Vot< e M *. f** b"*Y**' E. kJATwz BWILDws PuxW Ao ExdAusT t-n cv r4e1 Tueter E<.'id.*y fe7 < 0 fcM ) - c v- 7to4
" ~ .D. sWe Fesst*
6";Usy f*y t (0= Yvo'de v- 1 cv - 1+oI * (zus,'de)b cv >< n -
$$ $ h $
cv - 1944 Z% s.'d e ) , Ma cv-,44s 4d,Y a- 7 rp a ida tMe cv ><se
~ ( T, side ss A cr-w+s, NH t/yf,,n %e4 bM2 -0.-rk; wt4-Xsge) rs a ev- w, mr - /,A.p kg a. hwi - Atah : mr t.ade )
4 O
LJW. h 4. V I e 6. Ect w (L.ccf.*.a is 7. K)
c. Nh' ost 7A mss- 41 Psr. I KC Sg te (Odrida) ss- M 4, 10 1 *-:: z- X G. Sv.y le (Sofro'de]
l2 cr-2. CF~T Tay e l f Bled (0* thv'de)
~ /9 SF - 42 , 43 Fuel Tia n fer Csul Racirc. L:~a (sr-n Ox7 c,*de) l 44 (sr-41,-H .Tn/Je) 31 Mg-2SA C FT ' A ' Fit / ~
(detride) M L - 31, - I t l , C F r /~h Teyly (Tmside) *
-12 3 n na - ssa c n r 'a' Fdi (Ga hide) . \
Q :r Y :: TgG (:-n'.!d 4 3- 1d-l4 , yd-Lo PArwi A&wRuy Reforn. . ( nt-t 1 -Zvuds ; WM -defssWe) $ 43 14 ~ t , x 4 Sr vis a kr Tsyly (A 2- !a1 side ;i- rusoVe) - 44 KA - s 1 , If .2~mstranes1~ be %g ly (ss- 2%p'de ; 31-45r71r lt'b AW W-/7, M-tr ?/nf $afly Sepply {/?- AM ;/W-Arl4 h 31 HV-tse,/ff KG. Woo.Aren (sys- n s.'de , if/- J etesWe) M NY - /29 , /N 16 64u/m,t [Is9 -7, r,Ve , /44 -/dp We) if nas - 4 , is, ts, Acs pra;s leo-s+,-aarmWe ) 4,7A,96 (rem -11, /d ,4 , 7), 7 6 11,BC,td fd,96 , s e - L. ,'h ,\. 9Bs 31 1D ,2 8
r
k. t; s L. Valcetfo. V.au , (Gantes to R.s;. )
3>. O Ne x. (Cauca Vavas) J FW 'l 6 kais F u t. x s. E $4 4 ( OeTride) 4 eu- u 4:- ru t.
r a. e as e < o t,-ide )
17 r w- os A E~.ser. Fe) to .:;on. E24 A (datside ) 31 Mu- se A caea. Fm ha.L'4'nq (nriga ) Mz 4 Gea Fle.J 'rsa 'A ' Na S pysy (- v ,,i J . ) ( d a t Ide) 32 M1-4 Core P h. d ' L A . ~ S ~ FtII { Z ,.;J )(GaYsid N MM-368 Gec Flo.A kk. 'c Ms Sepply (ruide) .
.sf e s - 3. 6 d}.asa r. a s h . << Ate, 5 gtr (M )
41 M1-22 YItroyau S,y/y [r> side.E
% 47 xesJ s a cRD Cae//y Akro - (Ar:Je )
s/ 4 c- 4o CAilled Wetee tu Grleo a (Em'de) S2 xcw-1s RC P Cao/;y Wa te,- (.Tus;de) 34 scu-s 4 .Z~~ tur. Cooliy to Lets e Codea (na,'de) i 45 ru -la a E-ar. rs) to s.s. g 2.4 s (s tride) I
REA470A Suite:W& s fMf*fftMEftf SYSTEMS (3 g 3/4.6.5 COMBUSTIBLE GAS CONTROL HYDROGEN ANALYZERS LIMITING CONDITION FOR OPERATION reacto r bu.Nm 3.6.5.1 Two 4-dependent tenut an4 hydrogen analyzers shall be OPERABLE. APPLICABILITY: MODES 1 and 2. ACTION: With one hydrogen analyzer inoperable, restore the inoperable analyzer to OPERABLE status within 30 days or be in at least HOT STANDBY within the next 6 hours. SURVEILLANCE REQUIREMENTS v 4.6.5.1 Eachhydrogganalyzershallbedemonstrated.OPERABLEatleast once per 92 d:y:#:E . .T^.CCSC TES' "# SIS - by performing a CHANNEL CALIBRATION using sample gases ten ui i ;: a.aepde -k r c&b ec.+Io n cuer +he ro.wge of 4he in s f ra m e n t.
a. One v h r perc--t hydre; n. Salen : ettr:g: .

5. F: - : lem p;r; :t hydr ;;;, b:10n:: nitre;;n.
1 l B&W-ATMOSPHERIC 3/4 6-21J January 1, 1977 l
ps CONTAINMENT SYSTEMS ELECT"!C "Y E N "EC^""!ME"S
\lMITING CONDITION FOR OPERATION ,/
3.6. Two independent containment hydrogen recombiner systems all be OPE BLE. APPLICABI TY: MODES 1 and 2. ACTION: With one hydrog recombiner system inoperable, restore he inoperable system to OPERABL status within 30 days or be in at ast HOT STANDBY within the next 6 urs. SURVEILLANCE REQUIREMENTS X / 4.6.5.2 Each hydrogen recombin syst shall be demonstrated OPERABLE:
a. At least once per 6 mont y verifying during a recombiner system functional test t the minimum heater sheath temper-ature increases to > 7 F ithin 90 minutes and is maintained for at least 2 hours.

b. At 1. east once per 1 months by:

1. Performing CHANNE CALIBRAT N of all recombiner instrumen tion and control cir. its.

2. Verify g through a visual examina on that there is no evid ce of abnormal conditions with y the recombiners (1. ., loose wiring or structural con etions, deposits o foreignmaterials,etc.).

3. Verifying during a recombiner system funct nal test that the heater sheath temperature increases to 1200 F within 5 hours and is maintained for at least 4 holir .

4. Verifying the integrity of the heater electrical ircuits by performing a continuity and resistance to grou test following the above required functional test. The sis-tance to ground for any heater phase shall be > 10, O ohms.
s B&W-ATMOSPHERIC 3/4 6-22J January 1,1977
,y RGMroR betwtup '
E6NY*fftMENT SYSTEMS HYDROGEN PURGE CLE"L'P SYSTEM (If le:: th:r 2 hydrogen re erbiner: : vail:ble) LIMITING CONDITION FOR OPERATION reac b r h d h 3.6.5.3 Acentai~enthhdrogenpurgecleanupsystemshallbeOPERABLEand capable of being powered from a minimum of one OPERABLE emergency bus. APPLICABILITY: MODES 1 and 2. ACTION: . reuscc kOh With the co-tai ~" ntbydrogen purge c! up system inoperable, restore the
~
hydrogen purge cle2 up system to OPERABLE status within 30 days or be in at least HOT STANDBY within the next 6 hours. SURVEILLANCE REQUIREMENTS
' 6.5.3 The hydrogen purge cleanup system shall be demonstrated OPERABLE:
a. At least once per 31 days 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 main-tenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire or chemical release in any venti-lation zone comunicating with the system by:

1. Verifying that the cleanup system satisfies the in-place testing acceptance criteria and uses the test procedures of Regulatory Positions C.5.a. C.S.c and C.S.d of Regula-tory Guide 1.52,. Revision 1, July 1976, and the system flow rate is 50 cfm + 10f

2. Verifying within 31 days Sit te ,mial that a laboratory analysis of a represen9t 9y v s-Mn cample obtained in accordance with Regulatpy Mdjon C.6.b of Regulatory Guide 1.52, Revision 1. July 1976, m.Sts the laboratory testing criteria of Regulatory Positich C.6.a of Regula-tory Guide 1.52, Revision 1 -July 1976.
t B&W-ATMOSPHERIC 3/4 6-23J January 1, 1977 l
W Acrea. W Lot h C0t' TAR.M:t?T SYSTEbis i l SURVEILLANCE REQUIREMENTS (Continued)
3. Verifying a system flow rate of _50 cfm i 10% during system operation when tested in accordance with ANSI N510-1975.

c. After every 720 hours of charcoal adsorber operation by verify-ing within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Positior. C.6.b of Regulatory Guide 1.52, Revision 1, July 1976, meets the laboratory testing criteria of Regula-tory Position C.6.a of Regulatory Guide 1.52, Revision 1, July 1976.

d. At least once per 18 months by:

1. Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is </446-) inches Water Gauge while operating the system at a flow rate of 50 cfm i10%.
, u__,.,4._ .u.. .u. ,n .- ---- . --- u ..,
( u...,... m..... -.. u. l
=..n zur- ' --- - -- --~ --
3. Verifying that the beaters dissipate 1.0 kw 1 JQfg, Jar when tested in accordance with ANSI N510-1975,

e. After each complete or partial renlacement of a HEPA filter bank by verifying that the HEPA filter banks remove > 99% of the 00P when they are tested in-place in accordance with ANSI N510-1975 while operating the system at a flow rate of so cfm i 10%.

f. After each complete or partial replacement of a charcoal adsorber bank by verifying that the charcoal adsorbers remove
> 99% of a halogenated hydrocarbon refrigerant test gas when they are tested in-place in accordance with ANSI N510-1975 while operating the system at a flow rate of _EQ_ cfm i 10%.
& l v
's B&W-ATMOSPHERIC 3/4 6-24J January 1, 1977 l
h CONTAINMENT SYSTEMS "YDROCE" "!!"C SYSTE." (OPTIC"^1) MMITINGCONDITIONFOROPERATION 3.6.5. Two independent hydrogen mixing systems shall be OPE LE. APPLICABIL TY: MODES 1 and 2. ACTION: With one hydroge mixing system inoperable, restor he inoperable system to OPERABL tatus within 30 days or be in t least HOT STANDBY within the next 6 ho rs. l SURVEILLANCE REQUIREMENTS
, s 4.6.5.4 Each hydroge mixing system shall b demonstrated OPERABLE:
a. At leas once per 92 days on a STAGGE TEST BASIS by;

1. tarting each system from the contro room.
. Verifying that the system operates for a least 15 minutes. . At least once per 18 months by verifying a system ow rate of at least cfm.
N . O l t' j B&W-ATMOSPHERIC 3/4 6-25J January 1, 1977 l l l l l
REAc roR IktLbtM& Ci-cana+*MenSvSTEMS " VEN TIL ATION O 3/4.6.6 PENETRATION ROOM EX"AUST AI" CLEANUP SYSTEM (0PTIONAL' I LIMITING CONDITION FOR OPERATION : 1 ve n h lah'o n l 3.6.6.1 Two independent c^= tai- rt penetration room enh:::t ci cle:nup 1 systems shall be OPERABLE. l l APPLICABILITY: MODES 1, 2 and 3. l 1 ACTION: l v e n Fila +'io n l With one cent:f : nt penetration room enh:uct ci :10:nup system inoperable, ! restore the inoperable system to OPERABLE status within 7 days or be in i at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within l the following 30 hours. 1 l SURVEILLANCE REQUIREMENTS i I 4.6.6.1 Each ceta!- rt penetration room ^venHla.Wn ^'"-+ '4- '-*-" system shall v be demonstrated OPERABLE: :
a. At least once per 31 days on a STAGGERED TEST BASIS by 4-iti:. 7;,
skrHngexh V e n +i la-Ho > l unW from the control room, c;. th- u;F the MEP" '"ter: :nd charce:1 1
terber and verifying that the system operates for at least 1 4 hour.: tith the heters c-

b. At least once per 18 months or (1) after any structural main-tenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire or chemical release in any venti-lation zone comunicating with the system by:
rifying th:t with tb . y-t= epr: tin; :t : #1 e r:*^
fm + 10% and exhausting through th Iters and charcoa ers, the total ow of the system to the facil uding leakage through the system diver a ves, is < n the system is tested tting cold DOP at the system . (For -
, . _ r: with divertia; valves _)
{ l ,- Ce B&W-ATMOSPHERIC 3/4 6-26J January 1, 1977 l
i l RFACTOR &Ril. DIN (r l bq E0ftf*fftMEftf SYSTEMS 1 0 SURVEILLANCE REQUIREMENTS (Continued)
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.S.c and C.S.d of Regula-tory Guide 1.52, Revision 1, July 1976, and the system flow rate is aooo efm 110%.

3. Verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory '
Guide 1.52, Revision 1, July 1976, meets the laboratory testing criteria of Regulatory Position C.6.a of Regula-tory Guide 1.52, Revision 1. July 1976.
4. Verifying a system flow rate of acoo cfm i10% during system operation when tested in accordance with ANSI N510-1975.

c. After every 720 hours of charcoal adsorber operation by verify-ing within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with v Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 1, July 1976, meets the laboratory testing criteria of Regula-tory Position C.6.a of Regulatory Guide 1.52, Revision 1, July 1976.

d. At least once per 18 months by:

1. Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is < (6) inches Water Gauge while operating the system at a flow rate of 2o00 cfm i 10%.

2. Verifying that the system starts on an::fety 'r.*::ti:r ES test signal.

3. Verifying that the filter cooling bypass valves can be manually opened, a
m.4,.a. .u.. .u. u...... o ,.,4-... 1 w
-TiF . . . . i. . ". L. ;"'. '. .!."' ~G. ;',m,, .,. ",..3 ' . . .' .1 ". '. Z. , , .. ,.o., m 7..
e. After each complete or partial replacement of a HEPA filter bank by verifying that the HEPA filter banks remove > 99% of the D0P when they are tested in-place in accordance with ANSI
,y {
N510-1975 while operating the system at a flow rate of aooo cfm i 10%.
' B&W-ATMOSPHERIC 3/4 6-27J January 1, 1977
t O RE K TOR SWLDM & h) 40MMMMEM SYSTEMS v i SURVEILLANCE REQUIREMENTS (Continued)
f. After each complete or partial replacement of a charcoal i adsorber bank by verifying that the charcoal adsorbers remove l
> 99% of a halogenated hydrocarbon refrigerant test gas when they are tested in-place in accordance with ANSI N510-1975 while operating the system at a flow rate of nooo cfm + 10%.
1 l 1 l I l O U v l l l 0, B&W-ATMOSPHERIC 3/4 6-28J January 1, 1977 l l
O CONTAINMENT SYSTEMS 3/?.5.7 "ACUU" "ELIEF VALVES (0"T!0"AL) , RMITING CONDITION FOR OPERATION ,/ N 3.6.7.1 e primary containment to atmosphere vacuum relief ives shall I be OPERABL ith an actuation setpoint of < psid. APPLICABILITY: DES 1, 2, 3 and 4. ACTION: With one primary contain t to atmosphere v uum relief valve inoperable, restore the valve to OPERABL status withi 4 hours or be in at. least HOT STANDBY within the next 6 urs and n COLD SHUTDOWN within th.e following 30 hours. SURVEILLANCE VIREMENTS s x 4.6.7 No additional Surveillance Requirements other than those equiied by ecification 4.0.5.
O w 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE SAFETY VALVES LIMITING CONDITION FOR OPERATION Foarkeen (10 3.7.1.1 -A M-main steam line code safety valves shall be OPERABLE.w&4h.
" #t cettia0: : peci' icd '- T d h 3.'
APPLICABILITY: MODES 1, 2 and 3. ACTION: less Nn /ouckeen (414) With - main steam line code safety valves 4noperable, cper: tic. 4- "^0ES 1, 2 :nd 3 may pre ::d previd d, that "fth* ' heur:, either the
,4 c;;rd!c vehe h rettered tc OPEPf.P' E ct:tu: er the Much:r Ov:rp "cr hip Setpe4at it educed per T e k 3.' 1; cthe mi :, be b at k ;t ll0T STa90BY "ith4" the " ext S heur trd 4- COLD SML'T00uu m4 a4. the <g77em4 7 30 5:urs. The prevhic : Of Speci#ketica 3 0 ' are "^t =piic %
re sFo re the inope rA\e va\ve (s) fe OPER ABLE s f a.t a s w,' fL'm 24 howes, or pla.c e %e reu+or in at lea.5 + por srAnosy m n a&&'Mena\ q kours , o.nB h COLO Shur 0 NN m'E'm an ads.Kona.\ 72 hour s . SURVEILLANCE REQUIREMENTS 4.7.1.1 No additional Surveillance Requirements other than those required by Specification 4.0.5. l l V' '~ B&W-STS 3/4 7-1 January 1, 1977
c O 00
.n-mu s.,-r s m
E -NAHMM-AtiOWABEE-h96tfAR-0VERPOWER TRIP SETPOINT WITH IN0%otE s, -5 TEAM LINE 5AFETY VALVES
-s us
Maximum Allowable Nuclear Maximum Number of Inoperable Safety Overpower Trip Setpoint Valves on Any Steam Generator (Percent of RATED THERMAL POWER) 1 ( )
2 w 3 ( ) s b N 8 N C 3 CD a maa N O
( ( (
D .e O 00
.. . , , . l g . - . .
STEM LINc SAFETY VALVES "ER ST = GENERATOR v, [MBER LIFT SETTING (i 1%)* OR SIZE l
a. psig

b. psig

c. psig w
1 d. psig
?
i i
*The lift setting pressure shall correspond to ambient conditions of the valve at nomin j operating temperature and pressure.
N' a w
~
f
PLANT SYSTEMS O E/'lER GENCV
^1'v !L I A DV FEEDWATER SYSTEM LIMITING CONDITION FOR OPERATION e,n e rge n ccp 3.7.1.2 Two independent ste? geae. ater aux"hry feedwater pumps and associated flow paths shall be OPERABLE with:
a. On $$x ' Sy#'"'Nte fee pump ::p:ble Of bei7; p= r:d 'r0r an-0PERABLE.: r;:: y besv I

b. Oneeme:::i rgenc
. ., y; feedwater pump capable of being powered from an OPERABLE steam supply system.
APPLICABILITY: MODES 1,2and37 ACTION: With one aux emerYencf 4 43r feedwater system inoperable, restore the inoperable system to OPERABLE status within 72 hours or be in HOT SHUTDOWN within the next 12 hours. y s SURVEILLANCE REQUIREMENTS 4.7.1.2 Each aux"iry feedwater system shall be demonstrated OPERABLE: 91
a. At least once per 3+ days on a STAGGERED TEST B/, SIS by:
Ae
1. Verifying that-eash steam turbine driven pump develops a discharge pressure of > H S psig :t : ' lee Of ;,
;;"' wde o n recircAKon flow when the secondary steam reaa'y pressure is gr.eate- thea avo psigdio%. *M* IS no f ca pde of ^*ho maI cO;W acWiev'ing 'i+s rep ired pos s e 3 n 2.
Verifying)that
teriti,. in the eachflowvalve ' ::::1,ispnot path that r r locked, Oper:'-dsealed er or otherwise secured in position, is in its correct position.

3. 4. %', n g dat h vnokoc-at'iven eine rgency eeJwo.+er d pu mp develops a. cLsc.bege pressare of a es psig wherson re e's t e w\cMo n flo w, n '
f The pcovisio n of 5pec',0;ca.+;o n 3. o. 4 ma
O a\\ow operat'onc\ 4es F',ng of -Se steam- d r'y be suspe nded l
B&W-STS 3f4 7-4 January 1, 1977 l \
PLANT SYSTEMS U SURVEILLANCE REQUIREMENTS (Continued) I
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 on ay( :::"bry feed
-ater actuati^n test signal.
Os steam- dr'nve n
2. Verif ing that eaebpump starts automatically upon receipt of a , aux"Mry f : int:r ::t;; tic;. test)-signal.
l B&W-STS 3/4 7-5 January 1, 1977
! h;PLANTSYSTEMS v l CONDENSATE STORAGE TANK LIMITING CONDITION FOR OPERATION l 3.7.1.3 The condensate storage tank (CST) shall be OPERABLE with a , minimum contained volume of 9 11 ens c' wate . g.3 feet bo7,ooo pile.o) o / w, der. l APPLICABILITY: MODES 1, 2 and 3. ACTION: ) With the condensate storage tank inoperable, within24 hours either:
a. Restore the CST to OPERABLE status or be in HOT SHUTDOWN

within the next 12 hours, or 1
servoce waker syslem
b. Den:onstrate the OPERABILITY of the (altern:t :: r;;) as a backup supply to the E
fi$ feedwater system and restore the condensate storage tank to OPERABLE status within 7 days or be in HOT SHUTDOWN within the next 12 hours.
SURVEILLANCE REQUIREMENTS 4.7.1.3.1 The condensate storage tank shall be deinonstrated OPERABLE at least once per 12 hours by verifying the contained water volume to be l within its limits when the tank is the supply source for the :::"i:ry emergencg feedwater pumps, 4.7.1.3.2 The .......... (secua.e
-..:rwOer sqsbem
r;:) shall be demonstrated OPERABLE at least once per 12 hours by (mthed de-andent uper :!ternat: 00er:0) t:::; r th: (:1ter.:t .:ter :: r::) h th: ::;;1y :: r;; f;r th:
ser 'i:ry f::&:ter pur? . verdfmg hat ak \ east one service wafer i loop s's c>perabcg cma MV the serve'ce wo.+er qsfem- e mergenc g
/eec0 ader Sgrfe/n tSofcdlon ettIve.s are eleer oren or CPER A8LE whenever 4ke seru;ce tacher sydem ls Ne suppl; y scarce f r N e.
Q
\,J pma rge nc q 4eeAos.+e r p umys. s B&W-STS 3/4 7-6 June 1, 1976
O PLANT SYSTEMS ACTIVITY ! LIMITING CONDITIOh FOR OPERATION 3.7.1.4 The specific activity of the secondary coolant system amil be .l < ede uCi/ gram DOSE EQUIVALENT I-131. i
- o.17 APPLICABILITY: MODES 1, 2, 3 and 4.
j ACTION: o.l7 With the specific activity of the secondary coolant system > ede uC1/ gram l DOSE EQUIVALENT I-131, be in at least HOT STANDBY within 6 hours and in COLD SHUTDOWN within the following 30 hours. i SURVEILLANCE REQUIREMENTS 4.7.1.4 The spacific activity of the secondary coolant system shall be determined to be.within the limit by perfomance of the sampling and
analysis program of Table 4.7-2.
l l l 1 B&W-STS 3/4 7-7 June 1, 1976 l
TABLE 4.7-2 SECONDARY COOLANT SYSTEM SPECIFIC ACTIVITY l SAMPLE AND ANALYSIS PROGRAM TYPE OF MEASUREMENT SAMPLE AND AND ANALYSIS ANALYSIS FREQUENCY l
1. Gross Activity Determination At least once per 72 hours l

2. Isotopic Analysis for DOSE a) 1 per 31 days, whenever )
EQUIVALENT I-131 Concentration the gross activity determina-tion indicates iodine e a.cFivef'g l tr:ti :: ;r;;tcr tPr.10% > 5 F.m e 5
- f th: :ll =b': 'i-ft.
bc.%ro and . b) 1 per 6 months, whenever the gross activity determination indicates iodine ::rcentration: w,. u , ,.... , w m, +w. miimm 34.4+ e v N mes ack rounb. v l l
- B&W-STS 3/4 7-8 June 1, 1976
i PLANT SYSTEMS BLOCK MAIN STEAM L:".E ISOLAT!O" VALVES i ( LIMITING CONDITION FOR OPERATION 6\ock 3.7.1.5. Each main steam l'n: 1::12 tier valve shall be OPERABLE. APPLICABILITY: MODES 1, 2 and 3. ACTION: h0C
"OE 1 With one main steam "n: ice!: tier valve inoperable, POWER OPERATION may continue provided the inoperable valve is either restored to OPERABLE status or closed within 4 hours.
Otherwise, be in HOT SHUTDOWN within the next 12 hours.
"0 0 E S 2-2nd 3 i th ::: 1.:f
te= 1in: i: !:ti:n v:1v inoper:ble, sub: quent Oper: tier * "^0ES 1, 2 er 2 = y pre:: d prc'.id d:
n e 4 eper:ble 1:0! tier '!:! f
2. it m intai ed cleted.
Othcr i c, be in l'OT C'"JT00'..'" within th: 9:nt 12 heurt.
b. The prc'ficient of Spe:!'!:: tion 2.0." re n t :pplicable.
SURVEILLANCE REQUIREMENTS 6\ock 4.7.1.5 Each main steam lin: i:Ol: tier vaive shall be demonstrated OPERABLE by verifying full closure within gy_ seconds when tested purstant to Specification 4.0.5. O (J ( n' B&W-STS 3/4 7-9 January 1, 1977
g PLANT SYSTEMS SEC0"Of" "?TEP CHE"!! Tov W ITING CONDITION FOR OPERATION , 3.7.1.6 T secondary water chemistry shall be maintained 1 thin the limits of Tab 3.7-2. APPLICABLITY: MODE 2 and 3. ACTION: (To be detemined in a manner se or in the bases and to be imposed by a change to this Specification. SURVEILLANCE REQ EMENTS
/ ~
O 4.7.1. The secondary water chemistry shall be determined to be thin th '.imits by analysis of those parameters at the frequencies speci.* d v Table 4.7-3. t l i l 0 GI' B&W-STS 3/4 7-10 June 1, 1976 e w
t 1l
-w _ -tw _
(t =u s u J J
-u -y -e u
B
-T -T -U 15 13 -V EB LBJ 4 A b I -I -V T I -I fV -J -L 15 Ab = u v
r
-I fI y -n_a - n --V_ g s , a -y.n ,, -ytU ,i ir -w 3 ,. _3 -5 3 . g 2g ;u .n Ju_U ,i - - % U 1 .il ,c eo
_
.- . s L5 -J _i . u 25 _b ,s .r r _J . .c e 15 Lu_G ,s .n t sw;u b e Ew .t -n m 5
_i 5 u a m aw__g 6
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n -i a mu_y _ n
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_s nw_y
.n e- .-u u
b
- e . i_ J .m bw_c l, ib
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s. _i l
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n. U T
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. _U 6w _P e .w l *wsI p -ui5 mo ai 1s r5 u
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, D. m m,;g.,4w .- LNt. MIaJ= . C3@ a* "&NG l l 1l' i:l11 jq!4 4
l O v C, C:
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3: e-e- 4.C - t) J: 4J I' () li Ad) 2 *t (J 4h
#l 1:
w-2C L1 () F- 1. 1 ". 3 119-L IL 3 .} 8393 kJ IJ CC 1:J3
>= *s-a- "3 1-lr 93(J J 1J<i C J: 1J 4193 L n-( -- 1:,
J3i s aC 3 1-
-J .#,in .J <1 <3 ** qJ 1.
Of r
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LJ et
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1-IJ J (J 34,,3 13nJ IL v 4f Ll dJ
.u aa44 tJ :- <-
J cC ' 5 ,, i:
.c3 >- :- <J o a[ O == iL >- .L74 c'u + iJ :3
i:
tJ <J,3
: 3 <>
o < 1ris ce i J2e LJ *~ e- o
*C 93si
i: i :
43(J Cd , , ,o - aC 1... (3 in
c <J 1
() 4 Jq.A o -:: LJ < 3t) o :. n. s1 i 3.--
<, i..
1:
<3C) ,.J:
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< >c i . ... q; 83 i.<> ,-uin: .o.-a o :-i3. :r i *I o <
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J i B&W-STS 3/4 7-12 June , 1976 l l 1 l I
PLANT SYSTEMS /3 V 2/' '.2 STEf" CENE""TCP "RESSURE/TE.""EP1,TURE L!"!"T!3N-ITING CONDITION FOR OPERATION ,/ 3.7.2.1 he temperature of the secondary coolant in the steam gen ators shall be > 110) F when the pressure of the secondary coolant iythe steam generator is (237)psig. / APPLICABILITY: all times. ACTION: With the requirements of e above specificatio not satisfied:
a. Reduce the steam gen ator press e to < (237) psig within 30 minutes.

b. Perform an engineering ev tion to determine the effect of overpressurization on e str tural integrity of the steam
, generator. Determin that the eam generator remains acceptable
. for continued oper ion prior to creasing its pressure above (237) psi .
SURVEILLANCE RE EMENTS
/ ~
4.7.2 The temperature of the secondary coole-t in each steam nerator j sha be determined to be > (110) F at least once per hour when se dary ! p ssure in the steam generator is > (237) psig and T,yg is < 200*F. i l l (v - B&W-STS 3/4 7-13 January 1, 1977
PLANT SYSTEMS
3/'.'.3 C0"a0F.EF C^^' IP.C SSTE" SYSTE" hMITINGCONDITIONFOROPERATION
~
3.7.3. Two independent component cooling water loops shall be O RABLE. APPLICABIL Y: MODES 1, 2, 3 and 4. ACTION: With one component ooling water loop inoperable, re ore the inoperable loop to OPERABLE sta within 72 hours or be in a east HOT STANDBY within the next 6 hour nd in COLD SHUTDOWN wit n the following 30 hours. SURVEILLANCE REQUIREMENTS s - 4.7.3.1 Each component cooling wa loop shall be demonstrated OPERABLE:
a. At least once per 31 ays by ve fying that each valve (manual, power operated or tomatic) serv ing safety related equipment that is not lock , sealed or othe se secured in position, is v in its correc osition.

b. At least ce per 18 months, during shutd , by:

1. rifying that each automatic valve in t flow path actuates to its correct position on an ESF test :
signal. I
2. Verifying that each component cooling water emerg cy pump starts automatically on an ESFAS test signal.
I
~
s ( , B&W-STS 3/4 7-14 June 1, 1976 l
O e'^at svsteas 3/4.7.4 SERVICE WATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.4.1 Two independent service water loops shall be OPERABLE. l APPLICABILITY: MODES 1, 2, 3 and 4. ! I ACTION: , J With one service water loop inoperable, restore the inoperable loop to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE REQUIREMENTS l 1 4.7.4.1 Each service water loop shall be demonstrated OPERABLE: l i
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.
I
b. At least once per 18 months, during shutdown, by[4.0. S Surve%c e Re ;
"- 4#y ng .._t i tgutremenks 02:5 tut itiof Specillca.Kon
've '- the #! :: path .
tutt:: t: it: ::r- :t p::iti:n :: :: ESF"S t::t- ;
ti;n:1.
.+-. -- ---., .. ,+. +,
_o u 4 ,.,4 +w.+ - -w ....4,.
.' 3. .-_...' + i. !. R. . ,'. . :. ~ _; . 6 '.b.a_J . . .. . . . .... ' .:.:. ' . i _ . . i. . "- ~' " ' ' '~ --
l
~
June 1, 1976 C B&W-STS 3/4 7-15
1 l J -
\
PLANT SYSTEMS- - L E /0 E R GE N C.'( Coo L I N G. Po tJ O 3/4.7.5 ULT !"a TE "E."' S P"'
, (OPT!0"?,L)
LIMITING CONDITION FOR OPERATION 3.7.5.1 Theeinere! tikencq cool',nNone$ ate heat shall be OPERABLE with: ,
nd:cdeA cQepfh of S }ee f (7D etcre-deen A U"h 1 '*O' ''

a. A minimum water leve! at er abcvc clevation ( ) Mc n Sc Level , USCS da tur. or 3 f ,+ wi A a n;+ 2 'in AloA,5 5 or 6. I

b. Anaveragewatertemperatureof<[JoSf*F. /
APPLICABILITY: MODES 1 2 A.
k. 2a and .
ACTION: With the requirements of the above specification not satisfied, be in at least HOT STANDBY within 6 hours and in COLD SHUTDOWN within the following 30 hours. ' p - h3 SURVEILLANCE REQUIREMENTS 4.7.5.1 The emer ultiyency cool lnfi"poncI
-ste heat shall be determined OPERABLE :t leert Once per 2' 50ur; by crifying th: ;verage .;;tcr tcmperaturc :nd water levcl to be with'n their limit .
ct . At least once per ac/ hours bg verd ging 4he po,d's depb mee+c the rey.remen ks of S pec P,'eJ:on 3 7. 5. I a. .
b. M lea 5 F once per a9 Ju,ars dar',.3 4Lepericdefjune1nrcqk Cepteder 30 Q veM inq %ak he pond's nuerttge wa.+er +empera}m re ls wdL'un ;+s \' m'ks , g

d. Ab leas4 once per yectr b y mak u'ng sonnc0l,tg s c/ Nue penc0 ctnd
" r%ccaf.3 .5pe io n+%e m,n 3,7. 5.
rQa;remenh<od
)a , m a in Jep fl, mee + s % e. )
4 n k/ m' l BtJ,STE ( 3/4 7-16 Imary t, 1977
h PLANT SYSTEMS 2/d S cL000 "",0TECT!0" (0? TIC"?L*) LIMITING CONDITION FOR OPERATION
3. 6.1 Flood protection shall be provided for all safety related sys s, components and structures when the water level of the (usua ly the ultimate heat sink) exceeds Mean Sea Level USGS d um, at APPLICABIL TY: At all times.
ACTION: With the water vel at above elevation Mean a Level USGS datum:
a. (Be in at ast H0T STANDBY within 6 hour and in COLD SHuiDOWN within the 11owing 30 hours.)

b. Initiate and co lete within h the following flood protection measu- s.
bp 1. (Plantdepende i )
2. (Plantdependent)
SURVEILLANCE REQUIREMENTS , s 4.7.6.1 The water level t shall be d ermined to be within the limit by:
a. Measureme at least once per 24 hours hen the water level is below el ation Mean Sea Level USGS atum, and

b. Meas ement at least once per 2 hocrs when he water level is equ to or above elevation Mean Sea Lev USGS datum.
This pecification not required if the facility desi has adequate pas Ive flood control protection features sufficient accommodate t Design Basis Flood identified in Regulatory Guide 1. 9, ugust 1973.
C, ) t
/B&W-STS - 3/4 7-17 January 1,1977' l
~
h Q PLANT SYSTEMS 7 3/4.7.5 CONTROL ROOM EMERGENCY AIR CONDITIONING AND AIR FILTRATION SYSTEM LIMITING CONDITION FOR OPERATION 7 3.7.5.1 Two independent control room emergency air conditioning and air filtration systems shall be OPERABLE. APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: With one control room emergency air condittoning or air filtration system inoperable, restore the inoperable system tc OPERABLE status wtthtn 7 days or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE REQUIREMENTS 7 ( 4.7.5.1.1 Each control room emergency air conditioning system shall be i demonstrated OPERABLE: E* a. At least once per 31 days on a STAGGERED TEST BASIS by:
1. Starting each unit from the control room, and

2. Vertfying that each unit operates for at least 1 hour and maintains the control room air temperature < 84*F D.B.

b. At least once per 18 months by vertfying a system flow rate of 9900 cfm + 10%.
7 4.7 4.1.2 Each control room emergency air filtration system shall be demonstrated OPERABLE:
~
a. At least'once per 31 days on a STAGGERED TEST BASIS by inttf ating, from the control room, flow through the HEPA filters and charcoal adsorbers and verffying that the system operates for at least 15 minutes.
+
b. At least once per 18 months or (1) af ter any structural main-tenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire or chemical release in any venti-lation zone comunicating with the system by:
s2 ARXANSAS - UNIT $ 3/4 7- F gg 311978 i 27mm.
nnarT Untu r - PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
1. 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.S.d of Regula-
tory Guide 1.52, Revision 2. March 1978, and the system l flow rate is 2000 cfm 110%.
2. Verifying within 31 days after removal that a laboratory analysis of a representative carbon sample cbtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory l testing criteria of Regulatory Position C.6.a of Regula-tory Guide 1.52, Revision 2, March 1978. l

3. Verifying a system flow rate of 2000 cfm 110% during system operation when tested in accordance with ANSI N510-1975.

c. After every 720 hours of charcoal adsorber operation by verify-
ing 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 Regula-tory Position C.6.a of Regulatory Guide 1.52 Revision 2 March 1978. ,

d. At least once per 18 months by:

1. Verifying that th' e pressure drop across the combined HEPA filters and charcoal adsorber banks is < 6 inches Water
' Gauge while operating the sy tem at a flow rate of 2000 cfm + 10%.
2. Verifying that on a control room high radiation or high chlorine test signal, the system automatically isolates the control room within 10 seconds and switches into a recirculation mode of operation with flow through the HEPA filters and charcoal adsorber banks.

e. After each complete or partial replacement of a HEPA filter . ~
bank by verifying that the HEPA filter banks remove > 99% of the DOP when they are tested in-place in accordance with ANSI N510-1975 while operating the system at a flow rate of-2000
~
cfm i 10%. ARKANSAS - UNIT I 3/4 7 MAY 151976
kb5 EH MI E ' PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) f. Af ter each complete or partial replacement of a charcoal ' adsorber bank by verifying that the charcoal adsorbers remove
> 99.95% of a halogenated hydrocarbon refrigerant test gas
Een they are tested in-place in accordance with ANSI N510- .
l 1975 while operating the system at a flow rate of 2000 cfm i 10%. i i
+
to ARKNISAS - UNIT 3 3/4 7-tg MAY 15 B7B -
PLANT SYSTEMS (] ) 3/'.'.S ECCS "U"" "^^" Ey".^.UST ^ !" CLEMUo SYSTEu l LIMITING CONDITION FOR OPERATION 7.8.1 Two independent ECCS pump room exhaust cleanup systems trains sha be PERABLE. APPLI ILITY: MODES 1, 2, 3 and 4. ACTION: ! 1 With one ECCS ump room exhaust air cleanup system inoperable restore the inoperable syst to OPERABLE status within 7 days or be in least HOT l STANDBY within th next 6 hours and in COLD SHUTDOWN withi the following ! 30 hours. I l SURVEILLANCE REQUIREMENTS 4.7.8.1 Each ECCS pump roem exh st air el nup system shall be demonstrated l D OPERABLE: i
a. At least once per 31 days or STAGGERED TEST BASIS by initiating, from the control room, fl th ugh the HEPA filters and charcoal ,
adsorbers and verifying at th system operates for at least : 10 hours with the heat s on. I
b. At least once per months or (1) af r any' structural main- 1 tenance or. the H filter or charcoal sorber housings, or ;
(2) following p inting, fire or chemical lease in any venti-lation zone c unicating with the system b :
1. Verif ing that with the system operating a flow rate of cfm + 10% and exhausting through th HEPA filters a charcoal adsorbers, the total bypass flow f the ystem to the facility vent, including leakage rough l the system diverting valves, is < 1% when the sys em is tested by admitting cold DOP at the system intake. (For .
systemswithdivertingvalves.) C B&W-STS '3/4 7-21 January 1, 1977 . 1
p( - PLANT SYSTEMS hRVEILLANCEREQUIREMENTS(Continued)
2. Verifying that the cleanup system satisfies the in-plac testing acceptance criteria and uses the test procedur s of Regulatory Positions C.S.a, C.5.c and C.S.d of Re la-tory Guide 1.52, Revision 1, July 1976, and the sy em flow rate is cfm +10%.

3. Verifying within 31 days after removal that a boratory nalysis of a epresentative carbon sample o ained in cordance with Regulatory Position C.6.b o Regulatory Gu'de 1.52, Revision 1, July 1976, meets e laboratory tes ing criteria of Regulatory Position .6.a of Regula-tory utde 1.52, Revision 1, July 1976.

4. Verifyi a system flow rate of cfm 110% during system o ration when tested in ac rdance with ANSI N510-1975.

c. After every 720 h rs of charcoal sorber operation by verifying within 3 days after r val that a laboratory fm analysis of a repre entative c bon sample obtained in accordance
's with Regulatory Posit on C.6.b f Regulatory Guide 1.52, Revisior. "
C 1, July 1976, meets th labor tory testing criteria of Regulatory Position C.6.a of Regula or Guide 1.52, Revision 1, July 1976.
d. At least once per 18 mon by:
1 Verifying that t press re drop across the combined HEPA filters and ch coal adso er banks is < (6) inches Water Gauge while o rating the stem at a flow rate of cfm i10%.
2. Verifyin that the system star on a safety injection test si .a l .

3. Veri ing that the filter cooling pass valves can be man ally opened.

4. erifying that the heaters dissipate +_
kw when tested in accordance with ANSI N51 1975. . f
e. ter each complete or partial replacement of a EPA filter ank by verifying that the HEPA filter banks rem e > 99% of _
the 00P when they are tested in-place in accordanc with ANSI N510-1975 while operating the system at a flow rate of ,p cfm i 10%. , . VJ
~
L B&W-STS 3/4 7-22 January 1, 1977 1
PLANT SYSTEMS SURVEILLisNCE REQUIREMCNTS (Continued)
f. h complete or partial replacement of a ch adsorber ban ing that the sorbers remove
> 99% of a halogenated efrigerant test gas when they are -p ace in accordance N510-1975 --wtrtTF operating the system at a flow rate of __ /
O\ I,-l ' C 3/4 7-23 January 1,1977 B&W-STS . l l
O,PLANTSv51 EMS . I l 3/4.7.9 HYDRAULIC SML'SSERS 6 HOC K SUPPR ESSORS [S Nd6BE R5) LIMITING CONDITION F0'R OPERATION shak satyressors 3.7.9.1 All hydraulic Enebter listed in Table 3.7-3 shall be OPERABLE. APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: s)ml suppressers inoperable, replace or restore the With one or more hy(draulic OnubberinoperableYEbTe7 s) to OPERABLE status within 7 least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE REQUIREMENTS s[nl suppreswrs 4.7.9.1 Hydraulic nubters will be demonstrated OPERABLE by performance of the following augmented inservice inspection program and the require-9 ments of Specification 4.0 5. stuk suppressors
a. Each hydraulic esber with seal material fabricated from ethylene propylene or other materials demonstrated compatible with the operating environment and approved as such by the HRC, shall be detennined OPERABLE :t 10::t One: '.f ter net le :
than ' ent! but Nithi- 5 enth: Of in it':1 critic:lity :nd in accordance with tne inspection schedule of Table 4.7-4
-there:f.ter. by a visual insp,ection.cf the sa"bM Visual inspections of the T u W E shall include, but are not necessarily limited to, inspection of the hydraulic fluid reservoirs, fluid connections, and linkage connections to the piping and I anchors. !riti:ti:r Of th; T:b b '.' ' faa s Lien ; 3edule-
h:!' be - 8 a..uminy Jm un.6 woe e m ;vu>iy 46 cne o moncn 4 :;'ectie interv ! r s}xlwnre wr

b. Each hydraulic :nubber with seal material not fabricated from ethylene propylene or other materials demonstrated compatible l with the operating environment shall be determined OPERABLE at least once per 31 days by a visual inspection of the snubber.
Visual inspections of the snubbers shall include, but are not necessarily limited to, inspection of the hydraulic fluid reservoirs, fluid connections, and linkage connections to the piping and anchors. b ( " !( I B&W-STS 3/4 7-24 January 1,1977 l
l PLANT SYSTEMS snoes .suret.sJo45 HYDRAULICt "UCOGO (Continued) SURVEILLANCE REQUIREMENTS (Continued)
c. At least once per 18 months during shutdown a representative I sample of at least 10 hydraulic snubbers or at least 10% of all snubbers listed in Table 3.7-3, whichever is less, shall l be selected and functionally tested to verify correct piston movement, lock up and bleed. Snubbers greater than 50,000 lbs capacity may be excluded from functional testing requirements.
Snubbers selected for functional testing shall be selected on a rotating basis. Snubbers identified in Table 3.7-3 as I either "Especially Difficult to Remove" or in "High Radiation Aree h" may be exempted from functional testing provided .these snubbers were demonstrated OPERABLE during previous functional tests. Snubbers found inoperable during functional testing - shall be restored to OPERABLE status prior to resuming opera-tion. For each snubber found inoperable during these functional tests, an additional minimum of 10% of all snubbers or 10 snubbers, whichever is less, shall also be functionally tested until no moretested. functionally failures are found or all snubbers have been ( ( BW-STS 3/4 7-25 January 1, 1977 l
)
v v t Sdd OYoW!ng TABLE 3.7-3 985, [Ag65
.s g g.fuPPR g
SAFETY RELATED HYDRAULIC -St".'SSE".S' F ' 4 v. ACCESSIBLE OR HIGH RADIATION ESPECIALLY DIFFICULT SNUBBER SYSTEM SNUBBER INSTALLED INACCESSIBLE ZONE ** TO REMOVE NO. ON, LOCATION AND ELEVATION (A or I) (Yes or No) (Yes or No) M a
? -
lX c
Snubbers may be added to safety related systems without prior License Amendment to Table 3.7-3 provided that I
E a revision to Table 3.7-3 is included with the next License Amendment request. E Q ** Modifications to this table due to changes in high radiation areas shall be submitted to the NRC as part of
--. the next L,1 cense Amendment request.
G 0 ( ( k
[ .-
.1- 3 Table -3.15 1- -
SAFETY RELATED SHOCK SUPPRESSORS (SNUBBERS)" Snubber No. Location Elevation Snubber in High Snubbers Snubbers Snubbers Radiation Area Especially Inaccessible Accessible During Shutdown ** Difficult During Normal During Normal
, to Remove Operation Operation HS-1 Decay Heat Line B '329' 1" X X HS-2 Decay Heat Line A 322' 11-3/8" X X HS-49 Decay IIeat Line A 329' 1" X X HS-50 Decay Heat Line A 322' 11-3/8" X X HS-8 Pressurizer Spray Line 408' 7-11/16" X X HS-9 Pressurizer Spray Line 408' 7-11/16" X X HS-51 Pressurizer Spray Line *373' 0" X X X HS-52 Pressurizer Spray Line 373' 0" X X X HS-53 Pressurizer Spray Line 382' 0" X X X HS-54 Pressurizer Spray Line 381' 6" X X X HS-55 Pressurizer Spray Line 398' 6" X X X HS-56 Pressurizer Spray Line 398' 0" X X X HS-57 Pressurizer Spray Line 406' 10" X X HS-58 Pressurizer Spray Line 408' 7-11/16" X X HS-59 Pressurizer Spray Line 408' 7-11/16" X X HS-60 Pressurizer Spray Line 408' 7-11/16" X X HS-61 Pressurizer Spray Line 408' 7-11/16" X X HS-62 Pressurizer Spray Line 408' 7-11/16" X X HS-63 Pressurizer Spray Line 408' 7-11/16" X X --. . .i . . _ _ _ _ _ _ _ _
Snubbers may be added to safety related systems without prior License Amendment to Table 3.7-3 provided that a revision to Table 3.7-3 is included with the next License Amendment request. l
** Modifications to this table due to changes in high radiation areas shall be submitted to the flRC as part of the next License Amendment request.
s
/
Amendment flo. 23 - 1
. 7-3 Tcblo 3.lf,-i l SAFETY RELATED Sil0CK SUPPRESSORS (SNUBBERS)
Snubber in liigh Snubbers Snubbers Snubbers Sber No. Location Elevation Radiation Arca Especially Inaccessible

Accessible During Shutdowd'*Dif ficult to During Norr.a1 Durin: Normal Remove Operation Operation WS-10 Pressur'Izer Relief Line 409' 2-3/4" X X
WS-Il Pressurizer Relief Line 410' 2-3/4" X WS-12 Pressurizer Relief Line 410' 2-3/4" X NS-13 Pressurizer Relief Line 400' 0" X X WS-14 Pressurizer Relief Line 400' 0" X X NS-66 Pressurizer Relief Line 41G' 2-3/4" X HS-67 Pressurizer Relief Line 410' 2-3/4" X NS-68 Pressurizer Relief Line 410' 2-3/4" X X NS-69 Pressurizer Relief Line 410' 2-3/4" - X HS-70 Pressurizer Relief Line 391' 0" X X
~
X NS-71 Pressurizer Relief Line 367' 6" X X ' X
S-72 Pressurizer Relief Line 357' 0" X X X WS-88 Pressurizer Relief Line 370' 0" X X X N-A-1 Pressurizer Relief Line 400' 0" X X X

-A-2 Pressurizer Relief Line 399' 0" X X X

-8-1 Pressurizer Relief Line 400' 0" X X X

-B-2 Pressurizer Relief Line 391' 0" X X X N-C-1 Pressurizer Relief Line 410' 2-3/4" X X 1(-C-2 Pressurizer Relief Line 394' 0" X X HS-3 Main Steam Line A 4?S' 0" X X

S-4 Main Steam Line A 4C8' 6" X NS-5 Main Steam Line A 423' .0" X WS-7 Main Steam Line B 42,' 0" X 400' 6" Z HS-IS Main Steam Line A X NS-16 Main Steam Line 8 423' 2" X 2" X X HS-17 Main Steam Line B 4 T '. ' -
406- 6" I HS-18 Main Steam Line B NS-19 Main Steam Line B 396- 0" X X WS-20 Main Steam Line B (08' 6" x HS-22 Nin feedwater Header B 376' 4-11/16" X X ifs-23 Ma'in Feedwater Header B 376' 4-11/16" x I .
Snubbers may be added to safety related systems without prior License Amendaent to Table 3.7-3 provided that I
,: ew + , Ta bl a 1 7 1 is inc1:2ded with the next License Amendment request.
c N, - ( 1 Tcbis -5. . ) 37.7 ()p/ ( SAFETY RELATED Sl!OCK SUPPRESSORS (SNUBBERS)# , Location Elevation Snubber in High Snubbers Snubbers Snubbers -: bber No. Radiation Area Especially Inaccessible
Accessible :
. .
During Shutdow@ Difficult to During Normal DurinaNorsa:!
Remove Operation Operation
. 2 4 Main Feedwater Header C 376' 4-11/16" X X X 376' 4-11/16" X X X 2 5 Main Feedwater Header B 15-26 Main Feedwater Header B 376' 4-11/16" X X l 'S-27 Main feedwater Header B 376' 4-11/16" X X j
15-28 Main Feedwater Header B 376' 4-11/16" X X X iS-29 Main feedwater Header B 376' 4-11/16" X X X iS-30 Main feedwater Line A 361' 0" X d5-31 'iain Feedwater Header A 376' 4-11/16" X X . (5-32 Main feedwater Header A 376' 4-11/16" X X !
!S-33 Main feedwater Header A 376' 4-11/16" X X 'i 15-34 Main Feedwater Header A 376' 4-11/16" X X HS-35 Main Feedwater Heaier A 376' 4-11/16" X X :
85-36 Main Feedwater HeaJer A 376' 4-11/16" X X X dS-37 Main feedwater He:. der A 376' 4-11/16" X X d5-38 Main feedwater Header A 376' 4-11/16" X X dS-21 Emergency Feedwater Line B 394' 0" X X X l. Reactor Coolant Pump A 390' 10" ' 1A X X
?A Reactor Coolant Pump A 390' 10" X X IB Reactor Coolant Pump B 390' 10" X X 2B Reactor Coolant Pump B 390' 10" X X IC Reactor Coolant Pump C 390' 10" X X i Reactor C0olant Pump C 390' 10" '
2C X X 10 Reactor Coolant Pump D 390' 10" X X 2D Reactor Coolant Pump D 390' 10" X X t :
Snubbers may be added to safety related systems without prior License teendment to Tabic 3.7-3 provided that , j l .
a revision to Table 3.7-3 is included with the next License Amendment request.
** Modifications to this table due to changes in high radiation areas shall be submitted to the flRC as part of !
the next License Amendment request. ,
0 e O oo TABLE 4.7-4 5 HOC L 34 PPRES302 na 12 HYDRAULIC SN000ER INSPECTION 3CHEDULE _ b y snar wrussons NUMBER OF iNOBBERE FOUND IN0PERABLE NEXT REQUIRED INSPECTICN INTERVAL ** DURING INSPECTION OR DURING INSPECTION INTERVAL
1 0 18 months + 25%
12 months T 25% 1 2 6 months T 25% 3 or 4 124 days [25% 5, 6, or 7 d2 days + 25%
>8 31 days + 25%
R. N bl
Snubbers may be categorized into two groups, " accessible" and " inaccessible". This categorization shall be based upon the snubber's accessibility for inspection during reactor operation. These two groups may b inspected independently according to the above schedule.
gf ** The required inspection interval shall not be lengthened more than one step at a time and the provisions g of Specification 4.0.2 are not applicable.
=
M N
~
l
PLANT SYSTEMS 3/4.7.10 SEALED SOURCE CONTAMINATION O- - I LIMITIN3 CONDITION FOR OPERATION 3.7.10.1 Each sealed source containing radioactive material either in excess of 100 micrccuries of beta and/or gama emitting material or 5 microcuries of alpha emitting material shall be free of > 0.005 micro-curies of removable contamination. APPLICABILITY: At all times. ACTION:
a. Each sealed source with removable contamination in excess of the above limit shall be imediately withdriwn from use and:

1. Either decontaminated and repaired, or

2. Disposed of in accordance with Commission Regulations.

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable, e
y ) SURVEILLANCE REQUIREMENTS v ( 4.7.10.1.1 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. 4.7.10.1.2 Test Frequencies - Each category of sealed sources shall be tested at the frequency described below,
a. Sources in use (excluding startup sources and fission detectors previously subjected to core flux) - At least once per six months for all sealed sources containing radioactive material:
@ l B&W-STS 3/4728 January 1, 1977
~
(. l 1 l
{}- PLANT SYSTEMS O SURVEILLANCE REQUIREMENTS (Continued) l
1. With a half-life greater than 30 days (excluding Hydrogen 3) and .

2. In any form other than gas.

b. . Stored sources not in use - Each sealed source and fission
' detector shall be tested prior to use or transfe* 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.
4.7.10.1.3 Reports - A report shall be prepared and submitted to the Comission on an annual basis if sealed source or fission detector leakage tests reveal the presence of > 0.005 microcuries of removable contamination. s ( B&W-STS 3/4 7-29 Ja- y 1,1977 l l
D 3/4.8 ELECTRICAL POWER SYSTEMS '( ,, I \ V 3/4.8.1 A.C. SOURCES 0":",^7. m0 i 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 trans-mission network and the onsite Class lE distribution system. I

b. Two separate and independent diesel generators each with:
J
1. A Separate day end engine-a.eunted fuel tank / containing a minimum volume of 160 gallons of fuel. I eine rge nc y wlM level

2. A separateV fuel storage system containing a minimum volur 1
of gellens of fuel .13 7 inches (ao oco gatte.Q c/ 6el. 7 3. A separate fuel transfer pump.
4. A s Fac W c o mp re.s s o r.
A APPLICABILITY: MODES 1,9 cdc2, 3 and 4. ACTION:
a. With either an t.iffsite circuit er dit:01 generator of the above required A.C. electrical power sources inoperable, demonstrate the OPERABILITY of the remaining A.C. : Orte: by cM5de circd, performing Surveillance Requirements 4.8.1.1.1.a and.4.8.l.l.2.a.4 within one hour and at least once per"$ hours thereafter; Nal'5 5 N restore at least two offsite circuits :nd two di:0:1 g:ncr:ter: Ts'd1I.',
ove r*
to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 Lours. 1, . _ ;; m , _ , 'm E .t . L a Jm , a- ~ 4 1 1.
wn . t .- A t i+' en Off1t circuit and on: dic ci gc. crater f -d cD'M: r the OPERAB A a.C. electric:1 pO: r : ur:^ of the remaining A.C. sour i noper:b'- .;trate y performing 4.8.1.1.2.a.4 within Surveillance Requi nts 4.8.1.1.1 ours thereafter; restor'e at]f,"f",j)f,
" -~
one hour and at least on least one of the inoper sou to OPERABLE status within is*1h
~
in the next 6 #" "" ' N ' 12 hours or be i east HOT STANDB hours and D SHUTDOWN within the follow 1 hours. ": tere 2 t ' ^ ' . t :: 0ff;it cireuit: and two dic;cl gencr:-. ' ' ' ~ d L stata: within 72 hour; fre.. T the ti= cf 4-itibt-O. 3/4 8-1 January 1, 1977 B&W-STS
3/4. 8 ELECTAICAL Pol.JER SYSTE/45
,l _
f\C T 1 Ol\l? N'outlunu ech \ s NES a bG e tel q e n a rcz. 4 c r o} hl, e a hc v e reg u o r ech N. C , elec 4clcN power scucce s incye cct h\p. , b e me n s britke 4Lo o pc R A n. s u ry o f 41, e cc,v,o ln t og cksgl y,,ero hr hq Qe s horm t ec Surve,llcr ace Sega's re sten b s N ?. l. ). /. ct OMS t udes5 Oe S'esel ge nercdor Is cd read g operaf,, , 4.3,\.1.2.a.4 w K n one koar aaD a4 leasf onc e per on kcacs MereaNer : reskre a+ lea s+ Jwo die sc I cjenera? ors h OPER ACLG sfakas wikks'n 7% Lcars or he ln a+ lea s h HOT S TRA/08 V wi N, .'n 4 L e n e d <s' hours ass ts COL O a ttr Dowu wifl, .'n 4he 1bilowlng 301, cur s. O, U'UN one oNsl5e &cc& ad o,re 8,ese\ ge,recalcr of 4Le c'bove re9airec( M.o. elec+rica\ peteer sources
'noperdAe3 ee m ons+ cede 4ke OPER A BILITY cf Mte re n1J n 1*g A. C. scu rces bg puefor,n;ag fa rve ;lL nc e Rey;ce men 4 s M. I.U a sse an \ess 4he < Die se l generaser 's a h eab g cme rak' m il . ?. t. ). 2, a . L}
w/A'in one kour amJ c.+ onc e se r 24 L.xt s &ae-a% c ; res 4c re cA \ ea s + one of -he lacperet\de ceucces 4u OPERA 8tE clafas w Aln 12 k e ar s o r be in. oA- leas h JJ0T STAN btV u R'~s khe ned 4. Lev.cs a ak 'a COLO St 4 T 6 0 a w w e'll '.^ 4 k e fo//owIng 30 kours, t I l 3h F-lo. c
ELECTRICAL POWER SYSTEMS v l l ACTION (Continued)
-Or 50 4- ;t lc;;t M0T ST? NOSY within th; ncxt S hour; and in COLO-5"L"00MN ithin the f0110 wing 30 heur;.
d. With of the above required offsite A.C. circuits inoperable, demonstrate the OPERABILITY of two diesel generators by performing Surveillance Requ.irement 4.8.1.1.2.a.4 within one hour and at least once per3 7 hours thereafter, unless the diesel generators are already operating; restore at least one e off of hours the inopjrab}lt' or c Ti fe,;,s,igs.gyces
;. ...T ST* to OPERABLE NOSY within thestatus next 6within hours.24 With Only On; cff;it; ;;urc; rc;t;r:d, rc;ter; t lc;;t tw0 Off;it: circuit; te OPCRACLE :t;tu; within 72 h;ur; fra. tir.:
Of *-iti:1 10 : Or be in at ic;;t M0T STff:DSY within the next , 5 heuc; and in COLD SHUT 00WN within the felle.,ing 20 hours. ' e sl. With of the above required diesel generators inoperable, demonstrate the OPERABILITY of two offsite A.C. Circuits by performing Surveillance Requirement 4.8.1.1.1.a within one hour , and at least once pe87 hours thereafter; restore at least one 1 of the inoperable diesel generators to OPERABLE status withi'n
6/ hours or be in at least H0T STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. Re tecc l I at lcast twe diesel genereter; te OPERACLC ;tatu; within 72 "
hcur; frec,. ti.,e of initial 1;;; or bc in at ic;;t "0T ST,*J:0SY '
"ithin th; n;xt 5 hour; ;nd in COLD SM"TOOWN within the felle ing .S ti k. ... . m ...
1 SURVEILLANCE REQUIREMENTS 4.8.1.1.1 Each of the above required independent circuits between the I offsite transmission network and the onsite Class lE distribution system shall be:
/
a. Determined OPERABLE at least once per 7 days by verifying correct breaker alignments and indicated power availability. I

b. Demonstrated OPERALLE at least once per 18 months during shutdown by transferring (manually and automatically) unit power supply from the nonnal circuit to the alternate circuit.
1
\
4.8.1.1.2 Each diesel generator shall be demonstrated OPERABLE: '
a. At least once per 31 days on a (STAGGERED TEST BASIS} by:

f -
s (~ B&W-STS 3/4 8-2 January 1,1977
ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIRLMENTS (Continued)
1. Verifying the fuel volume in the day and engir mounted I
feel tank. emergency
2. Verifying the fuel volume in the fee storage tank. l

3. Verifying the fuel transfer pump can be started and transfers fuel from the storage system to the day and engin =ented tank. l

4. Verifying the diesel starts from ambient condition and l
accelerates to at least (9007 rpm in 1g seconds.
5. Verif
-(-50-$ying kw, andthe generator operates is synchronized, for 1 60 minutes. loaded to t l350l S. '!:rifying the dic::1 ;;n:rster i: cli;n;d to provid; Standby p= r to th: :::: icted = rg:ncy bu :: . l 7 '!er"ying that the Out : tic lead quence timer i OPCRACL: with the interval betw;;n ::ch lead block i; within i 10% Of it; dC;ign intCr=1.
b. At least once per 92 days by verifying that a sample of diesel T fuel obtained " ::: rd:ne with AS"i 0270-5'h- from the -f-ee4- l emerreg storage tank is within the acceptable-limits specified in Table 1 of ASTM D975-i+when checked for viscosity, water l and sediment. 'S

c. At least once per 10 :::Or.ths during shutdown by:
i
1. Subjecting the diesel to an ' inspection in accordance with !
procedures prepared in conjunction with its manufacturer's recomendations for this class of standby service. i
2. Verifying the generator capability to reject a load of B S2a 1 (large:t 19;10 =crg:::y 10:d, kw without tripping. I

3. Simulating a loss of offsite power in, conjunction with a I safety injection actuation test signal, and: ,
a) Verifying de-energi:ation of the eriergency busses and load shedding from the emergency busses. I b) Verifying the diesel starts from ambient condition on the auto-start signal, energizes the emergency busses with permanently connected loads, energizes O the auto-connected emergency loads through the load (;s.) sequencer and operates for > 5 minutes while its ,K generator is loaded with the emergency loads. l B&W-STS 3/4 8-3 January 1, 1977 L
1 I p ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) i i t c) Verifying that all diesel generator trips, except i' engine overspeed and generator differential, are automatically bypassed up:n 100: Of ";;1 tag: on the emergency bus and/or a safety injection actuation g test signal. l j 4. Verifying the diesel generator operates for > 60 minutes while loaded to ; (100)! ht.. f. ll rwtfS9
5. '!:rifyir; th t th: : t: : rn::t:d 10:d: t ::F dic::1
. -; ncr:ter de net :::::d th: ?000 50 r r: ting of hw.
m i
f I,
-~) ,
J (. B&W-STS 3/4 8-4 June 1,1976 ,
ELECTRICAL POWER SYSTEMS 5"'JT T ';*
\IMITINGCONDITIONFOROPERATION 3.8.1. As a minimum, the following A.C. electrical power sources sh 1 be OPE E:
a. O circuit between the offsite transmission network d the ons e Class lE distribution system. I

b. One di el generator with:

1. Day a engine-mounted fuel tanks cont ning a minimum volume f gallons of fuel. l

2. A fuel sto ge system containing minimum volume of gallons of 1. I

3. A fuel transfer ump.
.O APPLIC' ABILITY: MODES 5 and 6.
1 ACTION: With less than the above mini require A.C. electrical power sources OPERABLE, suspend all opera ons involving RE ALTERATIONS or positive reactivity changes until e minimum require .C. electrical power sources are restored to ERABLE status.- SURVEILLANCE UIREMENTS
, s 4.8.1.2 The above required A.C. electrical power sources sh be demon rated OPERABLE by the perfonnance of each of the Surve nce Re rements of 4.8.1.1.1 and 4.8.1.1.2, except for requirement 4 .1.1.2a.5.
l l B&W-STS 3/4 8-5 January 1, 1977 l 1
ELECTRICAL POWER SYSTEMS
' v /4.8.2 ONSITE POWER DISTRIBUTION SYSTEMS 4.C. DISTRIBUTION ^1 MTU40 L.IMITING CONDITION FOR OPERATION 3.8.2.1 The following A.C. electrical busses shall be OPERABLE and I energized with tie breakers open between redundant busses:
f41607 volt Emergency Bus # S3
'(4160) volt Emergency Bus # /N T480T volt Emergency Bus # .RE T480) voit Emergency Bus # B6 f2407 vuii A.C. iilai Bus i -t240) voit A.C. Vitsl Ous #
(1207 volt A.C. Vital Bus # R51 h 11201 volt A.C. Vital Bus # RS2 s ( ' fl201 volt A.C. Vital Bus # R$3 117.31 volt A.C. Vital Bus # R5Y . l APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: With less than the above complement of A.C. busses OPERABLE, restore the inoperable bus co OPERABLE status within 8 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN witt.in the following 30 hours. SURVEILLANCE REQUIREMENTS 4.8.2.i The specified A.C. busses shall be determined OPERABLE with l tie breakers open between redundant busses at least once per 7 1 days by verifying correct breaker alignment ard indicated power availability.
~
s B&W-STS 3/4 8-6 January 1, 1977 L
ELECTRICAL POWER SYSTEMS
^ C. O!ST"!""T!^"
S!!"T00'.?" hMITINGCONDITIONFOROPERATION 3.8.2.2 s a minimum, the following A.C. electrical busses shall b OPERABLE: , I 1 - (4160 volt Emergency Bus 1 - (480) volt mergency Bus 1 - (240) volt A. Vital Bus 2 - (120) volt A.C. Vit Busses. APPLICABILITY: MODES S and 6. ACTION: With less than the above compleme of A. busses O'ERABLE and energized, establish CONTAINMENT INTEGRITY ithin 8 ho . i I : SURVEILLANCE R IREM_ENTS
- x - - - -
4.8.2. The specified A.C. busses shall be determined OPERABLE at ast once er 7 days by verifying correct breaker alignment and indicated l p r availability. I i
l I
l B&W-STS 3/4 8-7 January 1, 1977
)
l I
f
, ELECTRICAL POWER SYSTEMS v D.C. DISTRIBUTION - CI Z T X LIMITING CONDITION FOR OPERATION 3.8.2.3 The following D.C. bus trains shall be energized and CPERABLE with tie breakers between bus trains open:
TRAIN "A" consisting of it50fl251-volt D.C. bus No.1, (E50/1251-volt D.C. battery bank No.1 and a full capacity charger. TRAIN "B" consisting of fE50/125f-volt D.C. bus No. 2, 4050/1257-volt D.C. battery bank No. 2 and a full capacity charger. APPLICABILITY: MODES 1, 2, 3 and 4. ACTION:
a. With one f250fl25)-volt D.C. bus inoperable, restore the
'noperablebustoOPERABLEstatuswithin6/hoursorbeinat least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN D within the following 30 hours.
b. With one 4050/125)-volt D.C. battery and/or its charger inoperable, restore the inoperable battery and/or charger to OPERABLEstatuswithin6/hoursorbeinatleastHOTSTANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
SURVEILLANCE REQUIREMENTS 4.8.2.3.1 Each D.C. bus train shall be detennined OPERABLE and energized with tie breakers open at least once per 7 days by verifying correct breaker alignment and indicated power availability. 4.8.2.3.2 EachfE50/125{-voltbatterybankandchargershallbe demonstrated OPERABLE:
a. At least once par 7 days by verifying that:

1. The electrolyte level of each pilot cell is between the I
minimum and maximum level indication marks. V January 1, 1977
, B&W-STS 3/4 8-8 t
l l
ELECTRICAL POWER SYSTEMS
\ / SURVEILLANCE REQUIREMENTS (Continued)
2. The pilot cell specific gravity, corrected to (77J'F and full electrolyte level, is > 1.a o , I ,

3. The pilot cell voltage is > 2.13 volts.
l
4. The overall battery voltage is > fB50/125) volts.

b. At least once per 92 days by verifying that:

1. volts under The float voltage of each charge. cad connected h:2 act dcccca cdcell is >t..y:n morc volte frc- the v; he obscrved during the original acccptanct
+ett. I
2. The specific gravity, corrected to T771*F and full electrolyte level, of each connected cell is > 1.20.and-h: net decrc :cd =re than frc- the v E: observed during the previcus tc:t. I

3. The electrolyte level of each connected cell is between the minimum and maximum level indication marks.

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 deterioration. I

2. The cell-to-cell and terminal connections are clean, tight and coated with anti-corrosion material. l

3. The battery charger will supply at least /l,9 amperes at a minimum of 105 volts for at least 0 h;urs. so wates,

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 emergency loads for 181 hours 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 l rating when subjected to a performance discharge test. This l performance discharge test shall be performed subsequent to l the satisfactory completion of the required battery service test.
O \ - B&W-STS 3/4 8-9 June 1, 1976 l
ELECTRICAL POWER SYSTEMS v 0.0. O!ST"I"UTIO" S"UT00M kMITINGCONDITIONFOROPERATION 3.8.2.4 As a minimum, the following 0.C. electrical equipment and s shall be ergized and OPERABLE: 1 - (250 25)-volt D.C. bus. I 1 - (250/12 volt battery bank and charger supplyi the above D.C. bus. APPLICABILITY: MODES nd 6. _ ACTION: With less than the above compi nt of D. . equipment and bus OPERABLE, establish CONTAINMENT INTEGRITY w hin hours. O v SURVEILLANCE REQUIR ENTS 4.8.2.4.1 e above required (250/125)-volt D.C. bus shal be determined OPERABLE d energized at least once per 7 days by verifying orrect breaker lignment and indicated power availability.
4. .4.2 The above required (250/125)-volt battery bank and char r
. all be demonstrated OPERABLE per Surveillance Requirement 4.8.2.3. /
l l (, B&W-STS 3/4 8-10 January 1, 1977
,_ ~ . . , _ _ - . - - . _ _ _
Q 3/4.9 REFUELING OPERATIONS TOV BORON CONCENTRATION LIMITING CONDITION FOR OPERATION 3.9.1 With the reactor vessel head unbolted or removed, the boron concen-tration of all filled perti n:- of theTeactor'goolant Syste :nd the re-i fueling c:::1 shall be maintained uniform and sufficient to ensure that the more restrictive of the following reactivity. conditions is met:
a. Either a K scrv:tiv !!!:"of::::
0.95f:r or rless, Wtic:,
rt:19 ch ' clud or : : 1% ak/k cen-1800

b. A boron+<m.concentration pp , which ppm. includ:: : 50 pp-
. <_ of .-..+.4.+<
1.(.1050)
,.m.,o_ .3,~....
I 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 1(25 7 gpm of Truol ppm
' boron or its equivalent until K is reduced to < 0.95 or the boron concentration is restored to > iib 50)"$$n, whichever is the more restrictive.
_ l The provisions of Specification 3.0.3 are not applicable. 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. [

b. Withdrawal of any safety or regulating rod in excess of 3 feet from its fully inserted position within the reactor pressure vessel.
4.9.1.2 The boron concentration of the reactor coolant :y:t: and the refueling canal shall be determined by chemical analysis at least once each 72 hours. t
*The. reactor shall be maintained in MODE 6 when the reactor vessel head is unbolted or removed.
l s- B&W-STS 3/4 9-1 January 1, 1977
O aeructING OeERATI0ns v INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.9.2 As a minimum, two source range neutron flux monitors shall be cperating, each with continuous visual indication. '- th: ::ntr:1 r;=
=d On: with =dib!: *ndi : tier *- th: :::t:t =nt : d th: cent 01 re- .
APPLICABILITY: MODE 6. ACTION: With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS or positive reactivity changes. The provisions of Specification 3.0.3 are not applicable. h SURVEILLANCE REQUIREMENTS v 4.9.2 Each source range neutron flux monitor shall be demonstrated OPERABLE by performance of:
a. ACHANNELFUNCTIONALTESTatleastonceperfdays. I

b. ACHANNELFUNCTIONALTESTwithinhhourspriortothe I
initial start of CORE ALTERATIGNS.
c. A CHANNEL CHECK at least once per 12 hours during CORE ALTERATIONS.
l 1OQU a B&W-STS 3/4 9-2 January 1,1977
REFUELING OPERATIONS DECAY TIME LIMITING CONDITION FOR OPERATION 3.9.3 The reactor shall be subcritical for at least:(100) :eur;. I
a. 73. howrs when d'iscbeg'mg 6 07 'irraS'Med o.ssemb\les WIN 4 413 'irro.d'sa.4ect c~ssembMes 'm the, Spent del poo\, or s

b. 175 hours when cEischargI ng 177 Irra8 o.kecl cosSem Wes evik Hl3 'it' ras'io.4e8 a.s.remhIIe5 'in bc
.s p e n k fu.e } pool .
APPLICABILITY: During movement of irradiated fuel in the reactor pressure vessel. ACTION: . Eke doe reguired fime, With the reactor subcritical for less than (100) Pe;.r;. suspend all operations involving movement of irradiated fuel in the reactor pressure vessel. The provisions of Specification 3.0.3 are not applicable. l C SURVEILLANCE REQUIREMENTS 4.9.3 The reactor shall be determined to have caen subcritical for at leastjl^0} h =: by verification of the date and time of subcriticality oprior 'to movement of irradiated fuel in the reactor pressure vessel. Q Yke. +lme, rep'wect bg S pe6-flc.a. Hon s . ') . 3 s g o.wa A wo.mber of ;rcoLAed assemh 'in 4ke. sped -/uel poo J B EW-sr5 3/q w geg,c l , (47y i l
- -- - ----- -,i.y y wg- +--c'- - --
i REFUELING OPERATIONS (3 REACTOR Bu lt.2t Net td . CC"TAl"",C"! PENETRATIONS LIMITING CONDITION FOR OPERATION rueter hi/ho 3.9.4 The contair: cat-penetrations shall be in the following status:
a. The equipmentN closed and heid in place by a minimum of four bolts. l

b. A minimum of one doortA in s
persmst

b nd **eq**'Y "M*
ci-lock closed. I <
+ -c. t E::b p:nctr: tion pr viding direct :::::: frca, th: E n B n_ m::#
cat
t 0:pher; t; the Out td; t= :phcr; : hall bc cith;r:
- 1. Cl ::d by :n i:;lc.ti;n v.lva, blind flangc, or =nu;l v:1v:, or
2. 30 C;poblC Cf bcing Cl0^Gd by Gn CICISOLE autesatic-
-cent:f ::nt purg: :nd exhau:t i:01stien ve.lve.
(~ APPLICABILITY: V] During CORE ALTERATIONS or movemen't of irradiated fuel within tha est, sin =nt. rs"6 b d'y - ACTION: With the requirements of the above specification not satisfied, immediately suspend all operations -' involving CORE ALTERATIONS or movement of irradiated fuel in the""to applicable. -^-+"- d .y.The provisions of Specification 3.0.3 are not SURVEILLANCEREQUlftENTS r w illia
; 4.9.4 Each of the above required c.uteron ta i n=nt- netrations shall be determined to be either in its closed / isolated condition or capable of l being closed by :n OPEPl.S'.E cut ^ Oti centat= nt purg: :nd exhau:t volv; 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 i
containment by/ l
f. Verifying the penetrations are in their isolated condition, or

b. T : ting th: ::nt:f m nt purg: :nd exh:::t v:1v:: per the A
,d - pplicabic partion of Sp;cificatica '4.0.4.1.2). \
B&W-STS 3/4 9-4 January 1, 1977 l 1
O v REFUELING OPERATIONS d COMMUNICATIONS LIMITING CONDITION FOR OPERATION 3.9.5 Direct comunications shall be maintained between the control room and personnel at the refueling station. 1 APPLICABILITY: During CORE ALTERATIONS. ACTION: When direct comunications between the control room and personnel at the refueling station cannot be maintained, suspend all CORE ALTERATIONS. The provisions of Specification 3.0.3 are not applicable. 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 durin~g CORE ALTERATIONS. I B&W-STS 3/4 9-5 June 1, 1976
REFUELING OPERATIONS l ru:L :?Not:nc cn:cc: cr:nAc:L:TY b1MITING CONDITION FOR OPERATION 3.9.6 e fuel handling bridges shall be used for movement of cont rods or 1 assemblies and shall be OPERABLE with:
a. Ah st minimum capacity of pounds. I

b. A hoist verload cutoff limit of pounds.
APPLICABILITY: Durin movement of control rods or el assemblies within the reactor pres re vessel. ACTION: With the requirements for bridg OPERABI Y not satisfied, suspend use of any inoperable bridge from ope tio involving the movement of control rods or fuel assemblies wit the reactor pressure vessel. The provisions of Specification 3.0.3 e ot applicable. I v SURVEILLANCE REQUIREME
- s 4.9.6 Each fu handling bridge used for movement of ntrol rods or fuel assembl s within the reactor pressure vessel shal e demonstrated OPERABLE w' hin 100 hours prior to the start of moving con ol rods or fuel ass blies by performing a hoist load test of at least pounds and d onstrating an automatic load cutoff when the hoist loa xceeds pounds. -
l I l l l l A (J ' .' s (' January 1,1977 B&W-STS 3/4 9-6 i l l A
/
REFUEL.ING OPE. RATIONS AutiLIRRV (m w)' CRANE TRAVEL - SPENT I'JCL STC^ ACE POOL BUILDING LIMITING CONDITION FOR OPERATION 3.9.7 Loads in excess of sooo pounds shall be prohibited from travel over fuel assemblies in the :tcrc- pool. I M** 5 . spen 4 Ael APPLICABILITY: With fuel assemblies and water in the steregc pool. I ACTION: With the requirements of the above specification not satisfied, place the crane load in a safe condition. The provisions of Specification 3.0.3 are not applicable. SURVEILLANCE REQUIREMENTS 4.9.7 The crane electrical power disconnect which prevents crant travel over the spent fuel pool' shall be verified open under administrative control at least once per 7 days, or the cranc travel interlock which prevents crane travel over the spent fuel pool shall be demonstrated OPERABLE within 4 hours prior to each use of the crane for lifting loads in excess of 2000 pounds. 8 m . 7 ~
!AAR '~ 373 ARKANSAS - UNIT 3 3/4'9-/
g w REFUELING OPERATIONS a s COOLANT CIRCULATION e ' LIMITING CONDITION FOR OPERATION 3.9.8 At least one decay heat removal loop shall be in operation. , l APPLICABILITY: MODE 6. ! I ACTION:
a. With less than one decay heat removal loop in operation, except as provided in b below, suspend all operationc involving an increase in the reactor decay heat load or a reduction in ufn6.UM,boronconcentrationoftheReactorCoolantSystem. Close all ntain=nt penetrations providing direct access from the udn b41JD;;ntain, Tent atmosphere to the outside atmosphere within 4 hours.

b. The decay heat removal loop may be removed from operation '
for up to 1 hour per 8 hour period during the performance of I CORE ALTERATIONS.in th; vicinity of the rc::ter prc::ur g T;;;;l (het) leg:. I I
c. The provisions of Specification 3.0.3 are r.]t applicable.
SURVEILLANCE REQUIREMENTS 4.9.8 A decay heat removal loop shall be demonstrated to be operating and circulating reactor coolant at a flow rate of > (2000) gpm at least once per 24 hours. 1500 l lJ B&W-STS 3/4 9-8 September 1, 1976 l l
)
G ' Q REFUELING OPERATIONS C0!!'"!':"Et:T """CE ," 0 C"""ST !50LAT!0^ SYST[". MMITINGCONDITIONFOROPERATION 3.9.9 containment purge and exhaust isolation system shall be OPERABLE. , APPLICABILITY: MODE 6. ACTION: With the containment pu e and exhaust isolation tem inoper'able, close each of the purge a exhaust penetration providing direct access from the containment tmosphere to th utside atmosphere. The provisions of Specificati 3.0.3 are t applicable. I l SURVEILLANCE REQUIREMENT 4.9.9 The con nment purge and exhaust isolation s tem shall be demonstra d OPERABLE within 100 hours prior to the art of and at least onc er 7 days during CORE ALTERATIONS by verifyin that contai nt purge and exhaust isolation occurs on manual in iation and a high radiation test signal from each of the containm t radia-t monitoring i'nstrumentation channels.
- I /
I O i ()e k B&W-STS 3/4 9-9 June 1, 1976 I
I REFUELING OPERATIONS l v P19 f*1 t t A T. P n--.,t--.-- 1 1 r c e f"t D..... r *\ P T. h. n. . -_- LIMITING CONDITION FOR OPERATION s - 3.9. As .. minimum, 23 feet of water shall be maintained over th l top of 3diated fuel assemblies seated within the reactor pre re vessel. APPLICABILITY: uring movement of fuel assemblies or co 01 rods within the reacto ressure vessel while in MODE 6. ACTION: With the requirements of t above specifica n not satisfied, suspend all operation involving movem t of fuel a emblies or control rods within the reactor pressure ves . T rovisions of Specification 3.0.3 are not applicable. v SURVEILLANCE REQ REMENTS 1
~
4.9.10 e water level shall be determined to be at least its m imum l requi d depth within 2 hours prior to the start of and at least on pe 4 hours during movement of fuel assemblies or control rods withi e reactor pressure vessel.
/ ~'
f a s B&W-STS 3/4 9-10 October 1,1975
REFUELING OPERATIONS STOPlCC P^0L L'ATEP, L:" L LIMITING CONDITION FOR OPERATION
/
3.9. As a minimum, 23 feet of water shall be maintained over t top of irr iated fuel assemblies seated in the storage racks. APPLICABIL : Whenever irradiated fuel assemblies are in e storage pool. ACTION: With the requirement o the specification not s 1sfied, suspend all movement of fuel and cran operations with lo s in the fuel storage area and restore the water vel to within s limit within 4 hours. The provisions of Specificatio 3.0.3 ar not applicable. J SURVEILLANCE REQUIREME S e x 4.9.11 The wa r level in the storage pool shall be det ined to be at least it inimum required depth at least once per 7 da when irradiate uel assemblies are in the fuel storage pool. iJ } B&W-STS 3/4 9-11 October 1. 1975 I
l 1 REFUELING OPERATIONS l O i e LH -- mmm TV,E.mm.nHDLTNG AR.EA
. VENflL.ATID.W. ~- . . . .
wivnnuw e vv6 nan v6Lnnur JIJl c ,, i v I
\
i I LIMITING CONDITION FOR OPERATION The Ael hanJIlng avec ven+l/af,'on ; 3.9.12 Twoindcpcadcat.ucisteragepeeleircleanupsystem/shallbe OPERABLE. handl,'n g APPLICABILITY: Whenever^ irradiated fuel 4s in the storage pool. ACTION:
a. 'o'ith enc fuel storagc pool air cicanup systa.r. inoperabic, fuci F
- .cicant within the storagc p;;l or crenc spcration with 1;;d:
cycr thc storegc pool .T,ay precocd provided the 0"E"AOLE n fuel-stor gc peci air cicanup ;y-t = i: in opcration and discharging i through at lasst er.; train of llE"A filters and charcoal ed crb r . N Withnefuelstora%ge 6 41 e e a u n Fil d en vi peal air cicanup system *0PERABLE, suspend I
a. #.
all operations involving movement ofy,fug]s w ithin the tcr:ge spenk ith 1;;d: ^ the stcr gc p 1 until del pool or r;nc opcratio.n..i.w.y,cas m the c,e . o 3 i c .sr.c
<a,.me; 4"c..rfuel
_a co..u see, is r system age pee restored I g to OPERABLE status. v d I b.g'. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. SURVEILLANCE REQUIREMENTS 4.9.12 The above required fuel otr:g: handl;ng;; r larea veJ,1*+;on air cicanup system;l shall be l demonstrated OPERABLE:
. n. ,. .. .. . . - - ,, s.... -- . e,nenenen ,ee, n.e,e u.. s u . u --
u. mw ewuJw vi sw w ywa wa wuj w vse u w s owwk na.w .hwa wrwJ & . wy a a a s b I u k i a 53 y fr= th; :::tr:1 c;=, f1= th=;h th "EP,^. filter: :nd ch;rce !
Od::rb r: :nd verifying th:t th: y-t= ;per .t:- for ;t it;,t 10 hour; with the h::t r: On.
o. g. At least once per 18 months or (1) after any structural main-tenance on the HEPA filter or charcoal adsorber housings, or (2). following painting, fire or chemical release in any venti-lation zone communicating with the system by:
1 s ' (. B&W-STS 3/4 9-12 September 1, 1976
\
l REFUELING OPERATIONS SURVEILLANCE REQUIREMENTS (Continued)
1. Verifying that with the system operating at a flow rate ofvo oco cfm + 10% and exhausting through the HEPA filters and charcoal adsorbers, the total bypass flow of the system to the facility vent, including leakage through the system diverting valves, is < 1% when the system is tested by admitting cold D0P at the system intake. (For systems with diverting valves.)
venKlaMcn
2. Verifying that the cicanup system satisfies the in-pl6ce testing acceptance criteria and uses the test proced. ares of Regulatory Positions C.5.a. C.S.c and C.S.d of Regula-tory Guide 1.52, Revision 1, July 1976, and the system flow rate is no ooo cfm +10%.

3. Verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 1, July 1976, meets the laboratory testing criteria of Regulatory Position C.6.a of Regula-tory Guide 1.52, Revision 1, July 1976.
V 4. Verifying a system flow rate of w eco cfm +10% during system operation when tested in accordance with ANSI N510-1975.
b. 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 1, July 1976, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 1, July 1976.
c g. Atleastoncepec18monthsby/g
/ k[lerifying that the pressure drop across the coabined HEPA filters and charcoal adsorber banks is < {6T inches Water Gauge while ope' rating the system at a flow rate of w,eco cfm + 10%.
2. V., ; ry i..u tim t v , o hid, edietion test signe., the system Oute. tically starts (unless already opereting) end d+reets its cxhaust flew through the lC;'A filters end ch: rec:1 derber b=ks.
lO . l L1 ' September 1,1976 l( B&W-STS 3/4 9-13 ) i
1 rm (1l( REFUELING OPERATIONS l SURVEILLANCE REQUIREMENTS (Continued)
. ver ify ing Uia c die sys i.cm ma inila isia dic av ent Twel 5ter g po ea at a negative pressure of > (1/4) inc er Gauge re e to the outside atmospEere g system operation. + Verifying that the r r coo ' bypass valves can be manually ope . % + Ifying that the heaters dissipate + - hen tested in ew v,Jance wius AN31 r;51^-i^ $.
d J./ After each complete or partial replacement of a HEPA filter bank by verifying that the HEPA filter banks remove > 99% of _ the D0P when they are tested in-place in accordance with ANSI N510-1975 while operating the system at a flow rate of vooco cfm i 10%. e[. After each complete or partial replacement of a charcoal adsorber bank by verifying that the char oal adsorbers remove l > 99% of a halogenated hydrocarbon refrigerant test gas when I they are tested in-place in accordance with AiiSI N510-1975 v while operating the system at a flow rate of so ooo cfm i 10%.
~
d v (. B&W-STS 3/49-14 September 1, 1976 i i l _ -
O. . 3/4.10 SeECiAt TEST ExCEeTIONS D l GROUP HEIGHT, INSERTION AND POWER DISTRIBUTION LIMITS (~J LIMITING CONDITION FOR OPERATION 3.10.1 The group height, insertion and power distribution limits of Specifications 3.1.3.1, 3.1.3.2, 3.1.3.6, 3.1.3.7, 3.1.3.8, 3.2.1 and I 3.2.4 may be suspended during' the performance of PHYSICS TESTS provided:
a. The THERMAL POWER is maintained < 85% of RATED THERMAL POWER. I

b. The Nuclear Overpower Trip Setpoint is < 10% cf Rated Thermal Power higher than the THERMAL POWER at Rich the test is performed, with a maximum setting of 90% of RATED THERMAL I
POWER.
u. The I L.. . L v i Speci fluo mune 0.2.2 e.id 0.2.3 e. e .no ...toined
=d dctcraincd ;t th: frcqu acics specificd in ?.10.1.2 bcle .
APPLICABILITY: MODE 1(durlag PHYS ICS TES T). n ACTION: (/) w ne bemd pwer > ef% of iMej Thermd hwer a With en, cf the liinits of Spccifications 3.2.2 cr 3.2.3 being cxcccded
.hilc t',c rcquirc .cnts of Spccifications 2.1.0.1, 0.1.2.2, 3.1.0.C,
3. .3.7, 3.1.3.S 3.2.1 cc J.2.4 arc su;p=d:d, cithcr-to
25 % o) Ra%S rurmA Power , o r
a. Reduce THERMAL POWER sufficicntly to ;;tisfy the ACTION requir=cnt; cf Spccificcti;n; 3.2.2 =d 3.2.0, ;r

5. Be in at least HOT STANDBY within 6 hours.
SURVEILLANCE REQUIREMENTS 4.10.1.1 The Nuclear Overpower Trip Setpoint shall be determined to be set within the limits specified within 8 hours prior to the initiation of ans-ae lent once pcr S hcur; during PHYSICS TESTS. 4.10.1.2 Thc Suryciliar.= Pequirc.T.cr.t; cf Spccific; tion; ?.2.2 =d 4.2.0 shell bc pcrformcd ;t inst ena pcr t= hour; during P"YS!CS TESTS. ,v B&W-STS 3/4 10-1 January 1, 1977
O,SPECIALTESTEXCEPTIONS i l ) PHYSICE TESTS LIMITING CONDITION FOR OPERATION 3.10.2 The limitations of Specifications 3.1.1.3, 3.1.3.1, 3.1.3.2, I 3.1.3.6, 3.1.3.7, and 3.1.3.8 may De suspended during the performance of PHYSICS TESTS provided: l
a. The THERMAL POWER does not exceed 5% of RATED THERMAL POWER.

b. The reactor trip setpoints on the OPERABLE Nuclear Overpower Channels are set at < 25% of RATED THERMAL POWER.

c. The nuclear instrumentation Source Range and Intermediate Range high startup rate control rod withdrawal inhibit are 0FERABLE.
APPLICABILITY _: MODE 2(durbgP RV5 tC 5 TE.f r ) , ACTION: I v g With the THERMAL POWER > 5% of RATED THERMAL POWER, immediately open the control rod drive trip breakers. . SURVEILLANCE REQUIREMENTS 4.10.2.1 The THERMAL POWER shall be determined to be < 5% of RATED THERMAL POWER at least once per hour during PHYSICS TESTS. 4.10.2.2 Each Source and Intermediate Range and Nuclear Overpower Channel shall be subjected to a CHANNEL FUNCTIONAL TEST within 12 hours prior to initiating PHYSICS TESTS. i r y ' (C) June 1, 1976 B&W-STS 3/4 10-2
\ ~
SPECIAL TEST EXCEPTION (0PTIONAL)
/mT D REACTOR COOLANT LOOPS !
LIMITING CONDITION FOR OPERATION 3.10.3 The limitations of Specification -ih+t3.1.).4 o.A 3.3.f may be suspended during the performance of stertoy and PHYSICS TESTS provided:
a. The THERMAL POWER does not exceed 5% of RATED THERMAL POWER, en

b. The reactor trip setpoints on the OPERABLE Nuclear O'verpower channels are set < 25% of RATED THERMAL POWER.
APPLICABILITY: During :tartup and PHYSICS TESTS. ACTION: With the THERMAL POWER greater than 5% of RATED THERMAL POWER, imediately Optr th: : r.tr:1 red drhe tHp brc:kers +69 h rea.<+or. 9 SURVEILLANCE REQUIREMENTS ' 4.10.3.1 The THERMAL POWER shall be determined to be < 5% of RATED l THERMAL POWER at least once per hour during :tertup crZ PHYSICS TESTS. 4.10.3.2 Each Nuclear Overpower Channel shall be subjected to a CHANNEL FUNCTIONAL TEST within 12 hours prior to initiating t:rtup er PHYSICS TESTS. (J . (' B&W-STS 3/4 10-3 January 1,1977 l
k[ SPECIAL TEST EXCEPTION v SHUTOOWN MARGIN LIMITING CONDITION FOR OPERATION 3.10.4 The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 may be suspended for measurement of control rod worth and shutdown margin provided:
a. Reactivity equivalent to at least the highest estimated control rod worth is available for trip insertion from OPERABLE control rod (.s). I

b. All axial power shaping rods arc withdrawn to at least 90%
(indicated position) and OPERABLE. APPLICABILITY: MODE 2. ACTION:
a. With cny safety or regulating control rod not fully inserted and with less than the above reactivity equivalent available for trip insertion or the axial powcr shaping rods not within their withdrawal limits, immediately initiate and continue D boration at 1D5 f gpm of 19m $ ppm boric acid solution or d$
its equivalent until the SHUTDOWN MARGIN required by Specifi-cation 3.1.1.1 is restored. v
b. With all safety or regulating control rods fully inserted and the reactor subcritical by less than the above reactivity equivalent, immediately initiate and continue boration at 1125) gpm of TFw) ppm boric acid solution or its equivalent until the SHUTDOWN MARGIN required by Specification 3.1.1.1 is restored.
SURVEILLANCE REQUIREMENTS 4.10.4.1 The position of each safety, regulating, and axial power shaping rod either partially or fully withdrawn shall be determined at least once per 2 hcurs. l l 4.10.4.2 Each safety or regulating control 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 SHUTDCWN MARGIN to less than the limits of Specification 3.1:1.1. 4.10.4.3 each moving The axial power axial powershaping rodsrod shaping shall be pmo
> 6.5% (Sdifat'Egstrated OPERABLE byO ;itio hours prior to reducing the SHUTDOWN'l4ARGIN to less than the limits of Specification 3.1.1.1.
() VJ s B&W-STS 3/4 10-4 January 1, 1977
2- \c e SPECIAL TEST EXCEPTIONS - MINIMUM TEMPERATURE FOR CRITICALITY LIMITING CONDITION FOR OPERATION 3.10.5 The minimum temperature for criticality limits of Specification 3.1.1.g may be suspended during low temperature PHYSICS TESTS provided:
a. The THERMAL POWER does not exceed 5% of RATED THERMAL POWER, tJ u c le a.c dver power i 'b. The reactor trip setpoints on the OPERABLE Linear Icwcr L;.cl .
High neutr:n flux :0 nit:rSg channels .re set at 1 20% of RATED THERMAL POWER, and b
. Thc b :ter C001:nt Sy;ter temper:tur: :nd pr:::ure rei: tion
' : hip i; maintained within th acc;ptabic r:gica of sparetion thou On Figur: 3.02. j APPLICABILITY: During startup and PHYSICS TESTS. l ACTION: a. }
.g a. With the THERMAL POWER > 5 percent of RATED THERMAL POWER, l 5 imediately open the reactor trip breakers.
4r r ~ L'ith th; I;::.ctor C001 nt Sy; tem t;mperature and prcs;ur; rcle.ti-- fj s1 . e region of unacceptable operation
l
,0 , 3.4-2, imediate reactor ers and restore 19 the temperature-pressura .
within its limit g within 30 , perform the engineerin y : :fi;; tion 2.t.0.1 ;ricr .. .. .. g ev, L.; -t, . , _ , ;ri;ic:...,. h l l ,g l L.C SURVEILLANCE REQUIREMENTS . ,y u Ys 4.10.5.1 T; c Rceute Cecient S.rs t e.a ten.peretuie end pre 55ure eietion5 hip [,id shall be verificd to bc within the cc cptabl r:sf0r f0r Oper:tiOr Of r Fi 3 u. c 3.4-2 ai. it:as unce per c. cur. p f, ., L< 4.10.5.8 The THERMAL P06.R shall be determined to be 1 5% of RATED THERMAL POWER at least once per hour. Ade e ov.rpowec t 4.10.5.1 Each Log:rithic ?;wer L;v:1 :nd Lincar P; ;r L2.el channel L. shall be subjected to a CHANNEL FUNCTIONAL TEST within 12 hours prior to initiating low temperature PHYSICS TESTS. i P' . D , ARKANSAS - UNIT 3 3/4 10-5 - l l J
) \
1 l l BASES j
)
i > FOR LIMITING CONDITIONS FOR OPERATION AND l SURVEILLANCE REQUIREMENTS
l O 'l O l NOTE The summary statenents contained in this section provide the bases for the specifications of Sections 3.0 and 4.0 and are not considered a part of these technical specifications as provided in 10 CFR 50.36. ! 3 i
3/4.0 APPLICABILITY n BASES The specifications of this saction provide the general requirements applicable to each of the Limiting Cont:ons for Operation and Surveil-lance Requirements within Section 3/4. l 3.0.1 This specification defines the applicability of each specifica- l tion in terms of defined OPERATIONAL MODES or other specified conditions l and is provided to delineate specifically when each specification is applicable. 3.0.2 This specification defines those conditions necessary.to constitute compliance with the terms of an individual Limiting Condition for Operation and associated ACTION requirement. 3.0.3 This specification delineates the ACTION to be taken for circumstances not directly provided for in the ACTION statements and whose occurrence would violate the intent of the specification. For example, Specification 13.5.11 calls for each Reactor Coolant System
\c 'j core flooding tank to be OPERABLE and provides explicit ACTION require-ments when one tank is inoperable. Under the tenns of Specification 3.0.3, if more than one tank is inoperable, the facility is required to be in at least HOT STANDBY within 1 hour and in COLD SHUTDOWN within the following 30 hours. \
3.0.4 This' specification provides that entry into an OPERATIONAL l MODE or other specified applicability condition must be made with (a) the full complement of required systems, equipment or components OPERABLE and (b) 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 number of specifications when startup with inoperable equipment would not affect plant safety. The:e exceptions are stated in the ACTION statements of l the appropriate specifications. l ( B&W-STS B 3/4 0-1 June 1,1976 i
APPLICABILITY v 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 Conditions 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 Surveillence Requirements. 4.0.2 The provisions of this specification provide allowable toler-ances 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 signifi- " cantly 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 Operations. 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 surveillance activities have'been satisfactorily demonstrated on a current basis as required to meet the OPERABILITY requirements of the Limiting Condition for Operation. i ( BW-STS B 3/4 0-2 January 1, 1977
O L' APPLICABILITY O BASES Under the te :,s of this specification, for example, during initial plant startup or rollowing extended plant outages, the applicable surveillance activities must be perfonned within the stated surveillance interval prior to placing or returning the system or equipment into OPERABLE status. 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 period-ically 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 specification. Under the terms of this specification, the more restrictive require-ments 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 prededence over.the ASME Boiler and Pressure Vessel '1 Code provision which allows pumps to 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 is declared inoperable and takes precedence over the ASME Boiler and Pressure Vessel provision which allows a valve to be incapable of perfortning its specified function for up to 24 hours before being declared inoperable. I i l I ! i ; I, j i
)
m' t~ B&W-STS B 3/4 p-3 January 1, 1977 l
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i
3/4.1
} REACTIVITY CONTROL SYSTEMS BASES _
3/4.1.1 B0 RATION CONTROL 3/4.1.1.1 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. During Modes 1 and 2 the SHUTDOWN MARGIN is known to be within limits if all control rods are OPERABLE and withdrawn to or beyond the insertion i limits. SHUTDOWN MARGIN requirements vary throughout core life as a function of fuel depletion, RCS boron concentration and RCS T a The most atnoT8a.d operating
~
restrictive temperature, condition occurs at E0L, and is associated with awith T}Ned steam line break accident postu and resulting uncontrolled RCS cooldown. In the analysis of this accident a minimum SHUTDOWN MARGIN of 'C."l% ak/k is initially required to control the reactivity transient. ccordingly, the SHUTDOWN MARGIN required is based upon this limiting condition and is consistent with O FSAR safety analysis assumptions. (urrra) 3/4.1.1.2 BORON DILUTION 150 0 A minimum flow rate of at least C000) GPM provides adequate mixing, l prevents stratification and ensures that reactivity changes will be gradual through the Reactor Coolant System in the core during boron concentratgreductionsintheReactorCoolantSystem. A flow rate of at least LyyPM will circulate an equivalent Reactor Coolant System l 1 volume of Of..vv) cubic feet in approximately -30'ininutes. The reactivity l change rate associated with baron concentration reduction will be within i the capability for operator recognition and control. ) 3 / 4.1.1. 3 MODERATOR TEMPERATURE COEFFICIENT The limitations on moderator temperature coefficient (MTC) are provided to ensure that the assumptions used in the accident and transient analyses remain valid through each fuel cycle. The surveillance require-ment for measurement of the MTC each fuel cycle are adequate to confirm l the MTC value-since this coefficient changes slowly due principally to the reduction in RCS boron concentration assaciated with fuel burnup. The confinnation that the measured MTC value is within its limit provides assurance that the coefficient will be maintained within acceptable values l throughout each fuel cycle. l B&W-STS B 3/4 1-1 June 1, 1976 l \
I l b8 f REACTIVITY CONTROL SYSTEMS U l l BASES 3/4.1.4.4 MINIMUM TEMPERATURE FOR CRITICALITY This specification ensures that the reactor will not. be made critical with the Reactor Coolant System average temperature less than 525 F. This limitation is required to ensure +) the moderator temperature coefficient is within its analyzed temperature range. 2) the pretective instru;;nt; tion 1 i: within it; n;r;;;l ;pcr; ting rangc, 3) thc prc;surizer i: ;;p;bi: Of bei9g " 2r CPE".^3LE :t;tu: ..ith c. :t::: bd bic, ;nd 4) th; r;;; tor prc: ure v :::1 i; d eve it: 'nimum "T *EC C' NDT
/4.1.2 80 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) makeup or DHR pumps, 3) separate flow paths, 4) boric acid pumps, 5) associated heat tracing systems, and 6) an emergency power suppl" from OPERABLE emergency busses.
With the RCS average temperature above 200 F, a minimum of two v separate and redundant borori injection systems are provided to ensure single functional capability in the event an assumed failure renders one of the systems inoperable. Allowable out-of-service periods ensure that minor component repair or corrective action may be completed without undue risk to overall facility safety from injection system failures during the repair period. The boration. capability of either system is sufficient to provide a SHUTDOWN MARGIN from expected operating conditions of 1.0 %Ak/k after xenon decay and cooldown to 200*F. The maximum expected boration capability requirement occurs at E0L g full power equilibrium xenon conditions and re acid Lgi..fes either550# dt tanks or upp,b.gallons
.J of 8700 ppmppm of a270 barated water borated from water thethe from boric borated water storage tank.
3.50,000 The requirements for a minimum contained volume of (102,500) gallons ( 35.9 Lf.e.*+el of borated water in the borated water storage tank ensures the capa- %O bility for borating the RCS to the desired level. The specified quantity of borated water is consistent with the ECCS requirements of Specification 3.5.4. Therefore, the larger volume of borated water is specifiesd.
"itP the ".CS temp:r:tur: 5:1~: 200 c. ::: 'njection y:t-- i:
ptdl; "ithout singl f:!! r :cn;iderati:n 0; th: L :i: cf th:
~
E B&W-STS B 3/4 1-2 January 1, 1977
p REACTIVITY CONTROL SYSTEMS yd i \ V' BASES 3/4.1.2 BORATION SYSTEMS (Continued)
,tt2bl^ . ti"4ty :cedi'4 ' c' th: _ tcr and th additi;na: a tr ic ti r.;
f,prc'ibiti th; ;i.gicLi.~9E m ;ti;n*.LTED"T!^C;;. tc;. bcc;;.x ac.j in:p;c;bl;. g . t , cc r;;;ti ,i n i n3 ; ' n th; . . : m i: :Sici:nt tc ;" n 2 r u,~"-rggbec,.. uap2 ~ .,",g '.t -. .j.: , ;. 4c.cd %1 ;. 200 e ;ng 3;;.,, 373 % reor m , . ~ , - ,. -
- .~ . .m.2 s v u u , m . .. . . ; m3u . ; c- c . v . .c .- , s _ _ , , _ - - _, , , , re _.
t , w..voa v. s . - , _ ,, v , r,. Lv. e ced os tc. i. um L.m ov... ou J 5 Lv. og sys ta.. e. (
) g;.llcns of (1300) ppa. Lo.ated oatm. 7 . 1., tL be, mead mate. ste.agc tank.
The contained water volume limits include allowance for water not available because of discharge line location and other physical 'charac-teristics. The limits on coptained water vo'ume, and boron concentration ensure cBnTMn(e.,gg within g pH valu between '$tM) and M/ of theThe solution sprayed l l n a7ter a design basis accident. pH band minimizes the evolution of iodirie and minimizes the effect of chloride and caustic stress corrosion cracking on mechanical systems and components. 7'c OPEP'3!LI7v '# c: b= in Mtic- cy:t - dring 1E* 'OC i n= rc; that this cytter i; r/ ilable for reacti.ity =ntrc! i!c #-
"CCE 5.
3/4.1.3 M0VABLE CONTROL ASSEMBLIES The specifications of this section (1) ensure.that acceptable power distribution limits are maintained, (2) ensure that the minimum SHUTDOWN MARGIN is maintained, and (3) limit the potential effects of rod misalign-ment on associated accident analyses. OPERABILITY of the control rod position indicators is required to determine control rod positions and thereby ensure compliance with the control rod alignment and insertion i limits. 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. For example, misalignment of a I safety or regulating rod requires a restriction in THERMAL POWER. In addition, those accident analyses affected by a misaligned rod are l reevaluated to confirm that the results remain valid during future opera tion. The position of a rod declared inoperable due to misalignment should not be included in computing the average group position for determining the OPERABILITY of rods with lesser misalignments.
~ \,
B&W-STS B 3/4 1-3 January 1, 1977
\
REACTIVITY CONTROL SYSTEMS
\
v BASES 3/4.1.3 MOVABLE CONTROL ASSEMBLIES (Continued) g, ,,4 gg Y assumed The maximum rod drop ti.ne permitted is consistent with the rod drop time used in the safety analyses. Measurement with T
> TS25PF and with reactor coolant pumps operating ensures thalVEhe hieasured drop times will be representative of insertion times experienced during a reactor trip at operating conditions.
l Control rod positions and OPERABILITY of the rod position indicators l are required to be verified on a nominal basis of once per 12 hours with frequent verifications required if an automatic monitoring channel is inoperable. that the applicable LCO's are satisfied.These verification frequencies are adeq
\
The limitation on THERMAL POWER based on xenon reactivity is with specified rod insertion limits satisfied.ne:essary to ensure that powe I b Gj
. ~J i
N l l l
\
i t
\
i V- B&W-STS .- B '/4 1-4 June 1, 1976 l
g u 3/4.2 POWER DISTRIBUTION LIMITS ( ) v' 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
> (1.32/1.301 during normal operation and during short tenn transients, (b) maintaining the peak linear power density s g kw/ft dg normal operation, and (c) maintaining the peak power density 1 L.sr kw/ft daring short tenn transients. In addition, the above criteria must be met in order to meet the assumptions used for the loss-of-coolant accidents.
The power-imbalance envelope defined in Figures (3.2-1,aM- 3.2-2),awa 3.1-3, and the insertion limit curves, Figures (3.1-1, 3.1-2, 3.1-3, 3.1-4,aM 3. -spa 3.i 3.1 5), are based on LOCA analyses which have defined the maximum linear heat rate such that the maximurr. clad temperature will not exceed the Final Acceptance Criteria of 2200 F fcllowing a LOCA. Operation outside of the power-imbalance envelope alone does not constitute a situation that would cause the Final Acceptance Criteria to be exceeded should a LOCA occur. The power-imbalance envelope represents the boundary of operation limited by the Final Acceptance Criteria only if the control rods are at the inser-
-- tion limits, as defined by Figures (3.1-1, 3.1-2, 3.1-3, 3.1-4 eM 3.1-5T, nwa 3.i-/
and if a f+/$'ercent QUADRANT POWER TILT exists Additional c'onservatism is introducted by application of: ks,imabne cush w,.p, ag eger
a. Nuclear uncertainty factors. eapneering anA uncerWg fach t s aho m+ &',r l' dh.

b. Thermal calibration uncertainty. '

c. Fuel densification effects.

d. Hot rod manufacturing tolerance factors.

c. ne\ vso %owh y.
The ACTION statements which permit limited variations from the b=. sic requirements are accompanied by additional restrictions which ensures that the original criteria are met. The definitions of certain design limit nuclear power peaking factors I as used in these specifications are as follows: F Nuclear Heat Flux Hot Channel Factor, is defined as the maximum 0 local fuel rod linear power density divided by the average fuel rod linear power density, assuming nominal fuel pellet and rod dimensions. Y B&W-STS B 3/4 2-1 January 1, 1977
l l POWER DISTRIBUTION LIMITS v BASES N F Nuclear Enthalpy Rise Hot Channel Factor, is defined as the AH ratio of the integral of linear power along the rod on which minimum DNBR occurs to the average rod power. It has been determined by extensive analysis of possible operating power shapes that the design limits on nuclear power peaking and on minimum DNBR at fullJower are met, orovided: l 1 1.hT g .18 g Fq 1:(W; F H Ib* b Power Peaking is not a directly abservable quantity and therefore l limits have been established on the bases of the AXIAL POWER IMBALANCE produced by the power peaking. It has been determined that the above hot channel factor liaits will be met provided the follcwing conditions are maintained.
1. Control rods in a single gro'up move together with no individual rod insertion differing by mare than _ ( )% (indicated pcsitica) i from the group average height. r <t, o ', ow, ,
2. (q) Regulating rod groups are sequenced with overlapping groups as required in Specification 3.1.3.6. v
3. The regulating rod insertion limits of Specification 3.1.3.6 I are maintained.
4 AXIAL POWER IMBALANCE limits are maintained. The AXIAL POWER IMBALANCE is a measure of the difference in power between the top and bottom ha'ives of the core. Calculations of cora average axial peaking factors for many plants and measurements from operating plants under a variety of operating conditions have been correlated with AXIAL POWER IMBALANCE. The correlation shows that the design power shape is not exceeded if the AXIAL POWER IMBALANCF is maintained betwear ' (10) pcrcent and - (15) percent at PUE0 SiER"X PO'in wh b f.d b ,# flyre s 3. 2- # , 3. 2 -2,aJ 3. 2 Se desig.. '.iait pc cr pca. king f;; tor; are the mcct rc trict!'fe calculac full power for the range from all control rods ful drawn to minimum allona control rod insertion and are the co .- design i basis. Therefore, for tion at a fraction of ~ HERMAL POWER, the design limits are met. When u 'ncore d + ors to make power distribu-tion maps to determine F and F g: g
a. The measurema total peaking factor, s hall be incre -
y (3) percent to account for makufacturi . rances and further increased by (5) percent to account N for :::;ucacnt e..ei. '
.T B&W-STS B 3/4 2-2 June 1, 1976 l%
POWER DISTRIBUTION LIMITS bw
/O /
i BASES v
i. .0 ..C0ZTC~Cnt ;f Cr.thalpj riSJ h0t ChC r.Cl faci 0r, 3. .k l '
E ..ccaned b, 'l) pcr:cnt 2 = = unt f:r m= =r r :t r*:'
#9.9 For Condition II events, the core is protected from exceeding (2;.~' ,
kw/f t locally, and from going below a minimum DNBR of (1. :/l.30), by automatic protection on power, AXIAL POWER IMBALANCE, pressure and tempera ture. Only conditions 1 through 3, above, are mandatory since the AXIAL POWER IMBALANCE is an explicit input to the Reactor Protection System. The QUADRANT POWER TILT limit assures that the radial power distribution satisfies the design values used in the power capability analysis. Radial power distribution measurements are made during startup testing and periodically during power operation. 3Al The QUADRANT POWER TILT limit of 4 4% at which corrective action is required provides DNB and linear heat generation rate protection with x-y plana power tilts. d 1...;'.ing tilt of (0.5).; can ic biccet:d Scfn 0; ma rg ., f a r z.cc e tc. i n ty i n P i =plct;d. The limit ofW% was selected F. to provide an allowance for9 the uncertainty associated with the indicated power tilt. In the event the tilt is not corrected, the margin for
' uncertainty on Fn is reinstated by reducing the power by 2 percent for each percent of Yilt in excess of (1) . 3.m C ,
3/0.2.5 0.'C P Tf'.:TERS-The 1i .'.t; ;r th; CLC r:1;t;d p;r:=cter; :=urc th:t ud of th^ paran rs are maintained within the normal steady state envelo f operation umed in the transient and accident analyses. ' consistent wi limits are e FSAR initial assumptions and have n analytically demonstrated adequa l o maintain a minimum DNB 1.30f throughout each analyzed transient. , The 12 hour periodic surveil 4 of these parameters through instru-ment readout is sufficient t sure thau e parameters are restored within their limits fol og load changes a ther expected transient operation. The 18 periodic measurement of is adequate to - ect flow degradation and ensure correCS total flow rate 6 ion of the flow indi on channels with measured flow such that the in ted perc flow will provide sufficient verifi, cation of flow rate on _ . . 5xi;.
,2 e ,,B&W-STS B 3/4 2-3 June 1, 1976 m
) 3/4.3 INSTRUMENTATION s O
BASES 3/4.3.1 and 3/4.3.2 REACTOR PROTECTION SYSTEM (RPS) AND ENGINEERED SAFETY FEATURES ACTUATION SYSTEM (ESFfS) INSTRUMENTATION l The OPERABILITY of the RPS and ESFAS instrumentation systems ensure that 1) the associated ESFAS action and/or RPS trip will be initiated when the parameter monitored by each channel or combination thereof exceeds its setpoint, 2) the specified coincidence logic is maintained,
3) sufficient redundancy is maintained to permit a channel to be out of service for testing or maintenance, and 4) sufficient system functional capability is available for RPS and ESFAS purposes from diverse parameters. l The OPERABILITY of these systems is required to provide the overall reliability, redundance and diversity assumed available in the facility design for the protection and mitigation of accident and transient con-ditions. The integrated operation of each of these systems is consistent with the assumptions used in the accident analyses.
The surveillance requirements specified for these systems ensure ei that the overall system functionii capability is maintained comparable to the original design standards. The periodic surveillance tests performed at the minimum frequencies are sufficient to demonstrate this capability.
"c mca;urcment of respen5e time at the ;pecified frequencies pro . assurance that the RPS and ESFAS action function asso with l each channe completed within'the time limit assumed * ~
e safety analyses. No creu taken in the analyses f se channels with response times indicated applicabl Response time may be rated series of sequential, overlapping or tota ~ nel test measurements ded that such test demonstrate tal channel response time as define , sor response time ication may be demonstrated by either 1) in place, or
c. .;itc tc;t mca;urca.nt; er 2) utilizini; replacement sensers oith -
certified scapon:t timcs. l l B&W-STS B 3/4 3-1 September 19, 1975
.v .
\
3/4.3 INSTRUMENTATION O. v
}5
(,) BASES 3/4.3.3 MONITORING INSTRUMENTAT'.0N 3/4.3.3.1 RADIATION MONITORING '.!NSTRUMENTATION 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 INCORE DETECTORS The OPERABILITY of the incore detectors ensures that the measurements obtained from use of this system accurately represent the spatial neutron flux distribution of the reactor core. See Bases Figures 3-1 and 3-2 for examples of acceptable minimum incore detector arrangements. 3/4.3.3.3 SEISMIC INSTRUMENTATION ( D I' The OPERABILITY of the seismic instrumentation ensures that suffi- cient capability is available to promptly determine the magnitude of a seismic event so that the response of those features important to safety may be evaluated. 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 l 10 _ , -CFR Part 100. -Tl.6 ia m u= n weivo is consiscenc wico J.c . cuv...oc aun m . _ _ , _ _ . . . i uns
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Th OP"",AC:LITY cf th ::t r:1;;i::1 in:trument: tion :n:ure: th:4 cuf#f rient metecrcic;ical d:t: i vail:Slc for c: tit. ting pct ntial
- diatier dere to the public : : result of rcutinc cr :::idental-re!c::: of r:dic :tive matert !: tc the ate::phere. i; ::p:Si'ity i requi cd te ev !uate the need for #-itiating protective me::ure; t l p-etect the 'e!!!" nd e fety of the public. 7'i #nctrumcr.t:ti:n in centi: tent "ith the rece m:ndation: cf Regul tery Cuide '.23 "Oncit: "et:Or:10;ical Prc;r ::," T bruary 10'2.
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( ., B&W-STS B 3/4 3-2 January 1,1977
l l l l I l 1 1w w 3/4.3 INSTRUMENTATION BASES n,,w eunw r .r .i riv.iy nm v uiu, v. .u ru
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r,, .u. ,,~, 4 ., _ n m. , ,m ., - m , a e m L _ . . , , j. , n, , e, , r f th - l C 1 B&W-STS B 3/4 3-3 January 1, 1977 1 1 1
J' '%
/ - . . ..a r i u v n _ . -.. . va a na s s urd UN L Y ;? ' A . . . .x r e r, m n ,, . . u r, b
LACK RADI AL
' SYMMETRY / = -- N- '
l ,. o l - k ,( ~~ _ - o ) TOP AXf AL CORF
, HALF j
i
.a W %~ ~~~~~~-
e. (M
$ 7 -.---
k -- E, C L [ o[- O g\ AXtAL PLANE $ \ e E y
\
x
~. / }
s 8 E C r --.~ o N
/- %<> ^;nd ! <> \ j BOTTOM AXIAL CORE k f / HALF /
Q- - - -
,/
EG -~
/ % / % / \,
Bases Figure 3-1 incore instrumentation Specification Acceptablo Minimum AXIAL POWER IMBALANCE Arrangement
~~
B&W-STS B'3/4 3-4 September 1,1976
O k'g -- _T;,. r auunc r un iLLus . H A e a um vs.6- ?* m avi ust FOft OPEHAllum ' e RADI AL _ _ _ _ _ _ _ _ _ _ _ _ _ 2_1/4 _________ CO R E.___ 1
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[
/ % \ RADIAL SYMMETRY IN TH!S PL ANE \ w /
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!=
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( / / IN THIS PL ANE
/ ./
x%% ~l i _ +s -
/ ,/ \,
Bases Figure 3-2 Incore Instrumentation Specification Acceptable Minimum QUADRANT POWER TILT Arrangement
. n' x_. B&W-ST:i B 3/4 3-5 September 1,1976
1 3/4.4 REACTOR COOLANT SYSTEM BASES t 3/4.4.1 REACTOR COOLANT LOOPS ' The plant is designed to operate with both reactor coolant loops in i operation, and maintain DNBR above (1.02/1.30} during all normal opera-tions and anticipated transients. With one reactor coolant pump not in operati:n in one or both loops, THERMAL POWER is restricted by the Nuclear Overpower Based on RCS Flow and AXIAL POWER IMBALANCE and the Nuclear Overpower Based on Pump Monitors trip, ensuring that the DNBR will be maintained above (1.22/1.307 at the maximum possible THERMAL POWER for the number of reactor coolant pumps in operation-ci th: Ical quality at thc pelat of Tiniasm DNOR equal te (22/10)%, whicheser is acre re trictisc. A single reactor coolant loop proviaes sufficient heat removal capability for removing core decay heat while in HOT STANDBY; however, single failure considerations require placing a DHR loop into operation in the shutdown cooling mode if component repairs and/or corrective actions cannot be made within the allowable out-of-service time. (m 3/4.4.2 and 3/4.4.3 SAFETY VALVES The pressurizer code safety valves operate to prevent the RCS from being pressurized above its Safety Limit of.{2750) psig. Each safety valve sel seis'sdesigned 5etpcint. to relieveJooooolbs noV 3reabr -Aw 3% per hour akove N of saturated valve's se +psteam*M. at se a prejsce The relief capacity of a single safety valve is adequate to relieve any overpressure condition which could occur during shutdown. In the event that no safety valves are OPERABLE, an operating DHR loop, con-nected to the RCS, provides overpressure relief capability and will prevent RCS overpressurization. , Durir.g operation, all pressurizer code safety valves must be OPERABLE to prevent the RCS from being pressurized above its safety limit of
$2750)psig. The combined relief capacity of all of these valves is greater than the maximum surge rate resulting from any transient.
Demonstration of the safety valves' lift settings will occur only during shutdown and will be performed in accordance with the provisions of Section XI of the ASME Boiler and Pressure Code. I B&W-STS B 3/4 4-1
- June 1, 1976
o (A fx REACTOR COOLANT SYSTEM v ! \ V BASES 3/4.4.4 PRESSURIZER A steam bubble in the pressurizer ensures that tne RCS is not a hydraulically solid system and is capoble of accommodating pressure surges during operation. The steam bubble also protects the pressurizer code safety valves and power operated relief valves against water relief. The low level limit is based on providing enough water volume to prevent a pressurizer low level or a reactor coolant system low pressure condition that would actuate the Reactor Protection System or the Engineered Safety Feature Actuation System as a result of a reactor scram. The high level limit is based on providing enough steam volume to prevent a pressurizer high level as a result of any transient. The power operated relief valves and steam bubble function to relieve RCS pressurc during all design transients. Operation of the power operated relief valves minimizes the undesirable opening of the spring-loaded pressurizer code safety valves. O 3/4.4.5 STEAM GENERATORS The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS will be maintained. The program for inservice inspection of steam generator tubes is based on a modification of Regulatory Guide 1.83, Revision 1. Inservice inspection of steam generator tubing is essential in order to maintain surveillance of the conditions of the tubes in the event that there is evidence of mechanical damage or progressive degradation due to design, manufacturing errors, or inservice conditions that lead to corrosion. Inservice inspection of steam generator tubing also provides a means of characterizing the nature and cause of any tube degradation ' so that corrective measures can be taken. The plant is expected to be 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 chemistry l limits, localized corrusion may likely result in stress corrosion cracking. ' The extent of cracking during plant operation would be limited by the limitation of steam generator tube leakage between the primary coolant systen and the secondary coolant system (primary-to-secondary leakage = 1 GPM). Cracks having a primary-to-secondary leakage less than this limit during operation will have an adequate margin of safety w withstand the loads Uj B&W-STS B 3/4 4-2 June 1,1976
pp REACTOR COOLANT SYSTEM A Q BASES imposed during normal operation and by postulated accidents. Operating plants have demonstrated that primary-to-secondary leakage of 1 GPM can be detected by monitoring the secondary coolant. Leakage in excess of this limit will require plant ',nutdown and an unscheduled inspection, during which the leaking tubes will be located and plugged. Wastage-type defects are unlikely with proper chemistry treatment of I the secondary coolant. However, even if a defect should develop in service, it will be found during scheduled inservice steam generator tube examinations. Plugging will be required for (40B of the tube nominal wall thickness. Steam generator tube inspections of operating plants have demonstrated the capability to reliably detect degradation that has pene-trated 20% of the original tube wall thickness. Whenever the results of any steam generator tub.ing inservice inspec-tion fall into Category C-3, these results g,ijl be prnmptly reported to the Commission pursuant to Specification 6.s.. prior to resumption of plant operation. Such cases will be considered by the Commission on a case-by-case basis and may result in a requirement for analysis, laboratory % examinations, tests, additional eddy-current inspection, and revision of the Technical Specifications, if necessary. The rter ;er. crater = ter ic';:1 '4-it: arc acnciatent "ith the iaiti la are pticar ia the CS? % B&W-STS B 3/4 4-3 _ January 1, 1977
7 REACTOR COOLANT SYSTEM (Q t v BASES I 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE l 3/4.4.6.1 LEAKAGE DETECTION SY;TEMS The RCS leakage detection systems required by this specification are provided to detect and monitor leakage from the Reactor Coolant Pres *;ure Boundary. !Leae detuction 5,5tv.u3 oco wen 5;atent nith the cecca...cadaci;n; cf Rcgulatory Cuidc 1.45, "Rc;ctor Ccolant Prcasucc Coundary Lcak;;; Octcction Cy';tems," May 1972. l 3/4.4.6.2 OPERATIONAL LEAKAGE PRESSURE B0UNDARY 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. Industry experience has shown that, while a limited amount of leakage { is expected from the RCS, the UNIDENTIFIED LEAKAGE portion of this can be , i reduced to a threshold value of less than 1 GPM. This threshold value is v sufficiently low to ensure early detection of additional leakage. ; i The total steam generator tube leakage limit of 1 GPM for all steam generators ensures that the dosage contribution from tube leakage will be limited to a small fraction of Part 100 limits in the event of either a steam generator tube ru&ture or steam line break. The 1 GPM limit is consistent with the assumptions used in the analysh of these accidents.
-!Lu (C.C) GPM icoko3c 1.uii t go. 2 6u. 3euecetcc en3mcca that stcom gcacr; tor tub: catcgrity i; m;intaincd in the cycat of ; main stc;; linc rupturc cr undcr LOCA conditions.
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 leakage detection systems.
-The CONTROLLCC LCA 300 liuiit of ( ) CPM rcstricts opc*; tion ;!th :
total 200 ic;kagc te eli RC puinp see 3 in excess of ' ) CPF. l L ll w B'&h-STS B 3/4 4-4 January 1,1977 s j l 1
1 O) REACTOR COOLANT SYSTEM u, y ; U BASES _ _ - - 1 1 3/4.4.7 CHEMISTRY l The limitations on Reactor Coolant System chemistry ensure that corrosion of the Reactor Coolant System is minimized and reduce the pat?ntial for Reactor Coolant System leakage or failure due to stress corresion. Maintaining the chemistry within the Steady State Limits shown on Table 3.4-1 provides adequate corrosion protection to ensure the structural integrity of the Reactor Coolant System over the life of the l plant. The associated effects of exceeding the oxygen, chloride and fluoride limits are time and temperature dependent. Corrosion studies 1 show that operation may be continued with contaminant concentration ! levels in excess of the Steady State Limits, up to the Transient Limits, l for the specified limited time intervals without having a significant i 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 concentrations to within the Steady State Limits. The surveillance requirements provide adequate assurance that con-4, centrations in excess of the limits will be detected in sufficient tima 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 the Part 100 limit following a steam generator tube rupture accident in ccajunction with an assumed steady state primary-to-secondary scnam generator leakage rate of 1.0 GPM.
'4 it:
Thc 504 valu;s uper : par for r:trf: th; limits cr. sp;;iff;
" le tfen by the "RCe;tivity Of '"repres;.
'... ...e g location . values are conservative in t ic site para-meters of the site, . ite bou
cation and meteorological conditions, were not consi valuation. The NRC is finaliz-ing site specifi ia which will be use basis for the reevalt the specific activity limits of this s .
. ma.untion wy-ec: alt in higher 14-it . ,: B&W-STS B 3/4 4-5 June 1, 1976
--REACTOR COOLANT SYSTEM
(_SJ J (n) v BASES The ACTION statement permitting POWER OPERATION to continue for limited time periods with the primary coolant's specific activity > 1.0 pCi/ gram DOSE EQUIVALENT I-131, but within the allowable limit shown on Figure 3.4-1, accomodates possible iodine spiking phenomenon which may occur following changes in THERMAL POWER. Operction with ;pecific ;c-
+wMy icvcis cxcccding 1.0 yC /grc.; 005: EQU:"Ab NT :-121 but within the limits shown cn rigucc 2.0 i sust bc rc;tricted to ac merc than 10 pcccent cf the unit; yearly oper; ting tiac sincc the ectivity levels elicacd bj-figurc 2.1 1 increc;c the 2 hour thyroid dc;c ct th; ;it boundcry by c facter cf up to 20 following a pcstuicted stcaa gcneretcr tubc rupturc.
Reducing T a to<k500hFpreventsthereleaseofactivityshou.lda steamgeneratorV8berupturesincethesaturationpressureoftheprimary coolant is below the lift pressure of the atmospheric steam relief valves. The surveillance requirements provide adequate assurance that I excessive specific activity levels in the primary coolant will be de-tected in sufficient time to take corrective action. Informatica obtained
.ee-4cdine ; piking will bc uscd to asscas thc parameters essccieted idr -5pib ng pnenc...sne.
A . eduction in f. cqucacy cf isctcpic analy;c , fc'%+- ing pcWCr chansc0 may bc PCimi33ibic if J U 3 ti f i C d by t h e d a t c c h t a i r.C u .= d 3/4.4.9 PRESSURE / TEMPERATURE LIMITS All components in the Reactor Coolant System are designed to with-stand the effects of cyclic loads due to system temperature and pressure changes. These cyclic loads are introduced by normal load transients, reactor trips, and startup and shutdown operations. The various categories of load cycles used for design purposes are provided in Sectice ( ) ei the FSAR. During heatup and cooldown, the rates of temperature and pressure changes are limited so that the maximum speci-fied heatup and cooldown rates are consistent with the design assumptions and satisfy the stress limits for cyclic operation. During heatup, the thermal gradients in the reactor vessel wall produce thermal stresses which vary from compressive at the inner wall to l tensile at the outer wall. These thermal induced compressive stresses tend to alleviate the tensile stresses induced by the internal pressure. Therefore, a pressure-temperature curve based on steady state conditi ns (i.e., no thermal stresses) represents a lower bound of all similar curves for finite heatup rates when the inner wall of the vessel is treated as the governing location.
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p b B&W-STS B 3/4 4-6 September 1, 1976
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(, B&W-STS B 3/4 4-8 September 1,1976
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l q V REACTOR COOLANT SYSTEM j v ; BASES l The heatup analysis also covers the detennination of pressure-temperature limitations for the case in which the outer wall of the vtssel becomes the controlling location. The thennal gradients estab-lished during heatup produce tensile stresses at the outer wall of the ! vessel. These stresses are additive to the pressure induced tensile stresses which are already present. The thermal induced stresses at the outer wall of the vessel are tensile and are dependent on both the rate cf heatup and t."e tire along the heatup ramp; therefore, a lower bound curve similar to that described for the heatup of the inner wall cannot i be defined. Consequently, for the cases in which the outer wal1 of the vessel becomes the stress controlling location, each heatup rate of interest must be analyzed on an individual basis. curveswhichwerepreparedbydeterminingthemostconser)vativecase,Th with either the inside or outside wall controlling, for any heatup rate u to 100'F per hour. The cooldown limit curves, Figures 13.4-2) anc (p3.4-35arecompositecurveswhichwerepreparedbaseduponthesame type analysis with the exception that the controlling location is always ,e the inside wall where the cooldown thermal gradients tend to produce tensile stresses while producing compressive stresses at the outside v ( wall. The heatup and cooldown curves were prepared based upon the most limiting value of the predicted adjusted reference temperature at the endof(2hEFPY. The reactor vessel materials have been tested to determine their initial RT /ther:wlt: of 'hese test 5 e e 5how. lo aldE5 Tsie 4-i. ,bd reu h c "cact;c opbtion and resultant fast neutron (E>l Mev) irradiation will cause an increase in the RT Therefore, an adjusted reference tem-perature,baseduponthefl$c.e and copper content of the material in __ A Baced question, can be predicted using OASES Figwes 4-1 cad 4-2. rThe heatup w; t c.c . refi include pre- I R,,por t , andcooldownlimitcurves,ofFigures43.4-2)and13.4-31}2{ dicted adjustments for this shift in RT at the end of EFPY, as well gp asadjustmentsforpossibleerrorsint$Tpressure and temperature sensing instruments. M #' The actual shift in RTkDT of the vessel material will be established l periodically during operation by removing and evaluating,J" accertn;c mith ";T". [105-70, reactor vessel material irradiation surveillance sg qimens i "ccriS!!!,nstalled Since thenc;r th; in;id neutron spectrawell of irradiation at the th; r:::t:rsamples versel and in the h*-8ene N* 8 M" I yessel inside the radius are essentially identical, the measured tran- ! sition shift for a sample can applied wi'th confidence to the adjacent section of the reactor vessel. The heatup and cooldown curves must be B&W-STS B 3/4 4-10 June 1, 1976 s-
REACTOR COOLANT SYSTEM BASES recalculated when the ART determined from the surveillance capsule is differentfromthecalcul$NdART NDT f r the equivalent capsule radiation exposure. The pressure and temperature limits shown on Figures 3.4-2 and 3.4-3 for reactor criticality and for inservice leak and hydrostatic testing have been provided to assure compliance with the minimum temperature requirements of Appendix G to 10 CFR 50. The number of reactor vessel irradiation surveillance specipens and the frequencies for removing and testing these specimens are provided in Table 4.4-3.to assure compliance with the requirements of Appendix H to 10 CFR Part 50. The limitations' imposed on pressurizer heatup and cocidown and spray water temperature differential are provided to assure that the pressurizer is operated within the design criteria assumed for the fatigue analysis performed in accordance with the ASME Code requirements.
) 3/4.4.10 STRUCTURAL INTEGRITY
( The inspection programs for ASME Code Class 1, 2 and 3 components, except steam generator tubes, ensure that the structural integrity of these components will be maintained at an acceptable level throughout the life of the plant. To the extent applicable, the inspection program for these components is in compliance with Section XI of the ASME Boiler and Pre:;sure Vessel Code. The internals vent valves are provided to relieve the pressure generated by steaming in the core following a LOCA so that the core remains sufficiently covered. Inspection and manual actuation of the internals vent valves 1) ensure OPERABILITY, 2) ensure that tha valves are not stuck open during normal operation, and 3) demonstrates that the valves t,q;in t Op:n :nd are fully open at the forcey equivalent to the differential pressures assumed in the safety analysis. B&W-STS B 3/4 4-11 June 1, 1976 v
.__ ~- y 3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)
QLs b BASES 3/4.5.1 CORE FLOODING TANKS 1 The OPERABILITY of each core flooding tank ensures that a sufficient volume of borated water will be imediately forced into the reactor vessel in the event the RCS pressure falls below the pressure of the tanks. This initial surge of water into the vessel provides the initial cooling mechanism during large RCS pipe ruptures. The limits on volume, boren concentration and pressure ensure that the assumptions used for core flooding tank injection in the safety I analysis are met. The tank power operated isolation vaives 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 tank isolation valves fail to meet single failure criteria, removal of power to the valves is required.
The limits for operation with a core flooding tank inoperable for l l any reason except an isolation valve closed minimizes the time exposure l < of the plant to a LOCA event occurring concurrent with failure of an additional tank which may result in unacceptable peak cladding tempera-tures. If a closed isolation valve cannot be imediately opened, the full capability of one tank is not available and prompt action is required to place the reactor in a mode where this capability is not required. 1 l 1 1
~.
B&W-STS B 3/4 5-1 October 1, 1975 l l l
EMERGENCY CORE COOLING SYSTEMS J BASES 3/4.5.2 and 3/4.5.3 ECCS SUBSYSTEMS The OPERAB L TY of two independent ECCS subsystems with RCS average temperature > -
'F 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 operatNg in conjunction with the core flooding tanks is capable of suppb g sufficient core cooling to maintain the peak cladding tempera-tures 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. 3so With the RCS temperature below +30&)'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. The Surveillance Requirements provided to ensure OPERABILITY of each component ensures, that, at a minimum, the assumptions used in the l'n) safety analyses are met and that subsystem OPERABILITY is maintained. Power is required to be removed from any valve which fails to meet single failure criteria. The decay heat removal system leak rate # ( surveillance requirements assure that the leakage rates assumed for the system during the recirculation phase of the low pressure injection will not be exceeded. 3/4.5.4 BORATED WATR STORAGE TANX The OPERABILITY ef the barated water storage tank (BWST) as part of the ECCS ensures that a safficient supply of borated water is available for injection by the ECCS in the event of a LOCA. The limits on BWST minimum volume and boron concentration ensure that 1) sufficient water is available within containment to permit recirculaticn cooling flow to the core, and 2) the reactor will remain subcritical in the cold condi-t tion following mixing of the BWST and the RCS water volumes with all control rods inserted except for the most reactive control assembly. These assumptions are consistent with the LOCA analyses. The cont'ained water volume limit includes an allowance for water not usable because of tank discharge line location or other physical characteristics. The limits on contained water volume, and boron eofbetweenj8.5 hand concentration tion sprayed within ensurgaggp,3 cer. m r:cr after a design basis (accident
. The 11.0{ofthesolu-I pH band minimizes the evolution of iodine and minimizes the effect of #. components. chloride and caustic stress corrosion cracking on s mech B&W-STS V.)
B 3/4 5-2 September 1, 1976 1
O~ O l l BASES ! FOR l 1 SECTION 3/4.6J CONTAINMENT SYSTEMS SPECIFICATIONS i l FOR BABC0CK AND WILC0X STS ATMOSPHERIC TYPE CONTAINMENT ! l n'
M'~} v 3/4.6 CONTAlfiMEtiT SYSTEMS BASES REAcroR Bou. Dis l 3/4.6.1 5:"ARY CONTAI J: NT Murog svaosus- ' 3/4.6.1.1 y. . ., . ... . .. . I NTEGR I TY RL ncroR ew.p m te F . . . ..o . ,. CGNTAIN:: NT materials from the 'O"'."Ur,1,yEGRITY ensures
. E t atmosphere that will bethe release to restricted of those radioactive leakage paths and associated leak rates assumed in the riafety 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 1 during accident conditions.
PlacieR BJILDMd-3/4.6.1.2 0..in :.:NT LEAKAGE re u k e b.alL' ca,,33d T limitations on : cat 2im.cnhleakage rates ensure that the total u . . . o .. . T leakage volume will not exceed the value assumed in the safety analyses at the peak accident pressure of 59 psig, P . As an added conser/atism, the measured overall integrated leakage r8te is further limited to < 0.75 Lu or < 0.75 L manceoftheperiodicteststoaccountfbr,asapplicable,duringperfor- possible degradation of the l r.c:"..:. =:,t ieakage barriers between leakage tests. r ute r b4V .$ The surveillance testing for measuring leakage rates are consistent with the requirements of Appendix "J" of 10 CFR 50. RacroR 8eitomo-3/4.6.1.3 CU;i!A....-::N! AIR LOCKS reu lock The limitations on closure and leak rate fg thp"bb2% a i .....uit air ar Mt"a(TriEn,e En1 leak required rate given. to meetSurveillance the restrictions on of testing wfifLdNb INTEGRITY the air lock seals and provide assurance that the overall air lock leakage will not become excessive due to seal damage during the intervals between air lock leakage tests. l 3/1.5.1.; CCNTAINP NT ISCLATION '!AL'!E AND CNAnnue acuu rncaavn u i wr I M C (0?Tf0NAL) ! ine 6FERA5RITY uf the i:,uia tiun vaive and cunia inmcni. doinici ncid I prca uricction syster.: i; required tc mcct the rc trictient er ever21' contci mcat le k rcte cc umed " the accident analytic. The Su me!'!ance-
?c:;uic .cata fa - Atecioning CPERACILITY ere con 5istent. itl AppendL "J" ul :C CTR :G. ~
B&W-ATMOSPHERIC B 3/4 6-lJ June 1, 1976
CONTAINMENT SYSTEMS O. 3 v I
\ s BASES 3/4.6.1.5 INTERNAL PRESSURE reahr \;
cmi,,g limitations on cMtai W=3. internal pressure ensure that 1) the c;nte... J t :tructur;this:espect pressure differential prevented fromatmosphere to the outside exceeding its design neg(ativ of 3.0) psigand2)theUP.1g:"d3peakpressuredoesnotexceedthedesign pressure of g psig during LOCA conditions, sli The maximum peak pressure obtained from a LOCA event is Mr) psig. The limit of (31 psig for igitial positive c nt:f : cat pressure will limit the total pressure to M psig which is less than the design pressure and is consistent with the safety analyses.
-_2 / ' . S . l . E ;T T:"?:?aT'JPI
_ e tv TSC '#.itat#'n; ,n ;;nt:.ir-'3t .;;r;g; ;ir t:,per;tur; =:gr; th;t the Over " ;:,ntc.ir;;nt av;.q: cir t;mper;tu.m d::: not ex:: d the hitial temperature conditi:n : umcd in the :: ident ;n ly i; f r a G , 406A,- 3/4.6.1.7 REACTOR BuaDm0-CC." TAI,:::ii STRUCTURAL INTEGRITY This limitation ensures that the structural integrity of the geb ',Lg t:f rm:nt will be maintained comparable to the original design standards for the life of the facility. Structural integrity is required to I ensure psig in thethat the of event containment a LOCA. will withstand the maxirgugessure of M54J - The measurement of Ido m.ou M t tendon lift off force, the visual and metallurgical examination of tendons, anchor-ages and liner, and the Type A leakage tests are sufficient to demon-strate this capability. i l i l l l
\ .'
y . J g B&W-ATMOSPHERIC B 3/4 6-2J June 1, 1976 l l
e d . CONTAINMENT SYSTEMS (J ) BASES 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS REncrest suanse, 3/4.6.2.1 CONTAINtENT SPRAY SYSTEM The OPERABILITY of the rc,..,raLM:y ontainm.at spray system ensures that coni.ce.Aaim r U.LC f me*t- depressurization and cooling capability will be available in the event of a LOCA. The pressure reduction and resultant lower ccatcir=ntem W k,,, leakage rate are consistent with the assumptions used in the safety analyses. The leak rate surveillance requirements assure that the leakage rates assumed for the system during the recirculation phase will not be exceeded.
.soown avesosior fpom#
3/4.6.2.2 o,. - ... SYSTEM (ort!ONAL) 2.% q u,')e edo % The OPERABILITY of thehpray_ additive systg ~ ensures that suf-ficient of Na0H a LOCA. -The end
c. i Ne n . . @m. C a,Sa?c added to theTon.a;nEnt spray in the event Ge,C,0, is aieilabic 2c cded0, volum and c"ccganic conccatratica iodinc cn;urc; ;uf'ici fcc.m the ccatairacnt .~ ct -'
atrac;, 2n8 Petur- it to the spray ::ter. The limits on contained sodium hydroxide p solution volume and concentration, cad conta ' ' gelutica vclume and concentration ensu "g-a% tu r '" 5) W
; h!b,;cdium_
thic;ulfate-
#pli mli.0) of the solution sprayed within _..EniM.iiMr a design basis I occident. TheTpHTaM minimizes the evolution of iodine and minimizes the effect of chloride and caustic stress corrosion cracking 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 assuwd in the safety analyses.
RCn4TCR NLDING 3/4.6.2.3 CONTAInti:NT- COOLING SYSTEM (OIT!0NAL) em er o.& s The OPERABILITY of the etsia eal cooling system ensures that 1) the rm4(*7' containant air temperature will be maintained within limits during normal "~ operation, and 2) adequate heat removal capacity is available when operated in conjunction with the cratain=nt spray system / during post-LOCA conditions. - ructs7,a-**g?
/4.0.3 100 N: CLEANur SYSTE". (OITIONAL) 1 - % GFERABILITY ur um vun io n.umn i. .vdinm cleonup sjstem. - - - u that I sufficient oval capability will be ava e event of l a LOCA. The reouction in nventory reduces the i resulting site bounda on dose ed with containment i o leakage. ation of this system and resultan moval l [# ] . %w ty arc ccoistcat ;;ith thc a;;umption; u;cd in thc LOCA ar,a.,.mt.
B&W-ATMOSPHERIC B 3/4 6-3J September 1, 1976 l l
CONTAINMENT SYSTEMS v BASES REr7CrcRf%tLbm & 3/4.6.4 CONTAILMENT ISOLATION VALVES rese r b&# The OPERABILITY of the containmcat isolation valves ensures that the re ferbA 9 ' containment atmosphere will be isolated from the outside environment in the event of or atmosphere a release of radioactive pressurization atLr.i.a.l of thec cc.&a t t msinecne.bth.e c.cn.t.ai=cnt e isolation re.Mer kJ bO; .o -
<r . . . . . 1 man .a i nre%n 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.
3/4.6.5 COMBUSTIBLE GAS CONTROL The OPERABILITY of the equipment and systems required for the detection and control of hydrogen gas ensures that this equipment will I be available to maintain the hydrogen concentration within containmcat A reube k.Lnp - 1 below its flammable limit during post-LOCA conditions. Eithcr rcccabiner unit (or The7 Fge cic;nup systemT is capable of controlling the expected I h drogen generation associated with 1) zirconium-water reactions, 2 radiolytic decomposition of water and 3) corrosion of metals within +ke rm be LM% contcin cnt. These hydrogen control systems are consistent with the 9 recomendations of Regulatory Guide 1.7, " Control of Combustible Gas Concentrations in Containment c ollowing a LOCA", March 1971. v The hydrogcn T. M ag ;ystem: arc provided tc ca:urc adcquat; mixing of the containmcat atmosphere felicwing & LOCA. Thi5 mLin3 od;cn : 11 p-event localiced accumulation; cf hydrog n from exccading the fle.....able MnH-VE N rsL A T w H 3/4.6.6 PENETRATION ROOM HHAUCT AIR CLEANUP SYSTEM (OPTIONAL) l The OPERABILITY of the penetration room edc a.o.x ~eust air " nup system i ensures that gioactive materials leaking frnm the ?$!";I$$;3 atmosphere through Ed!Mr-St penetrations following a LOCA are filtered and adsorbed ! prior to reaching the environment. The operation of this system and i the resultant effect on offsite dosage calculations was assumed in the ! LOCA analyses. I 3/4.0.7 VACUUM RELIEF VALVES (GPTIONAL) The OPERA 0!LITY cf the primary containrie M -Otm;;p.cre. v;cuu: relicf valvc: cn urc; that the contain& cat internal pressucc dacs act Lccomc accc negativc than psig. This ccad Mica is necessar W Srevent exceeding the ccat in=.nt de:f sn limit for internal vacum-of p:fg.
.j ' ~
s
's B&W-ATMOSPHERIC B 3/4 6-4J September 1, 1976 l
l l tQ ,n 3/4.7 PLANT SYSTE_MS f BASES I I 3/4.7.1 TURBINE CYCLE 3/4.7.1.1 SAFETY VALVES The OPERABILITY of the main steam line code safety valves enrares that the secondary system pressure will be limited to within its design pressure of $os4 psig during the most severe anticipated system operational transient. The maximum relieving capacity is associated with a turbine trip from 100% RATED THERMAL POWER coincident with an assumed loss of condenser heat sink (i.e., no steam bypass to the condenser). The specified valve lift settings and relieving capacities are in accordance with the re uirements of Section III of the ASME Boiler and Pressure Vessel Code, 1971) Editi I valves on all of the steam lines i io'Thelbs/hr total which relieving is (capac~ity uf) percentfor all of the total secondary steam flow o (/ ) lbs/hr at 100% RATED THERMAL POWER. A minimom of 2 0":"A"LC ;;f;ty velvc; pcr :t=m gencr; tor en;urc; that ;uffici;nt rclieving capacity i; availabic for the allc;;;ble Tll R"AL P0"CR ratricti:n in T;ble 2.' 1. 9 S9RTUP and/or "^":", 0" "ATION i; allc.;;ble ..ith nfety cle-inoperab within the limitations of the ACTION requirements on e basis of the redu ion in secondary system steem flow and THERMAL P ER required by the reduce eactor trip settings of the Nuclear Overpo channels. The reactor trip tpoint reductions are derived on the lowing bases: SP M - (Y)(V) x ,f,'.D..,.sih where: SP = reduced Nuclear Overpo - r rip Setpoint in percent of RATED THERMAL POWER V = maximum number of operable s ty valves per steam generator
,N. 9 _1
(-105.5} = Nuclear 0v ower Trip Setpoint specif in Table 2.2.1 X = Total lieving capacity of all safety valve er steam gen tor in lbs/ hour
s. 4.34 7N lbs/Ao r.
Y= aximum relieving capacity of any one safety valve i lbs/ hour
IL IO .59.1 %s/kr O
, V ~.
. B&W-STS B 3/4 7-1 September 1, 1976
h PIANT SYS1 EMS v 9 BASES ElnERGENCY 3/4.7.1.2 AUXILIARY FEEDWATER SYS' EMS The OPERABILITY of the #[$NNy feedwater system ensur Reactor Coolant System can be cooled down to less thai $ M*es that the F from normal operating conditions in the event of a total loss of offsite power. r/,, ,no/,c c Cad cicatric driven =yY,c acyary feedwater pump is capable of delivering a total feedwater entrance of the steamflow of f350-)?Tpm generators. at a pressure of Eachsteamdriven#Ny. . g"_ pig to the feedwater pump is capable of d g pressure of (il 37"$elivering a total of sig to the <ntran:.e feedwater the steamflow of t?se-)h,pm generators. This at a capacity is sufficient to ensure that adequate feedwater flow is available y to removg less than #(305PF where the Decay 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 for cooldown of v the Reactor Coolant System to less than*t305PF in the event of a total loss of offsite power or of the main fesdwater system. The minimumj.5HuTbou)N water volume is sufficient to maintain the RCS at HOT STAT:00Pconditions for b.5{ hours with steam discharge to atmosphere concurrent with 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. 3P.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 steani line rupture. This dose includes the effects of a coincident 1.0 GPM primary to secondary tube leak in the steam generatar of the affected steam line. These
. values are consistent with the assumptions used in the safety analyses. -
BLOCX 3/4.7.1.5 MAIN STEAM LItc ISOLATI0t?-VALVES
&ck The OPERABILITY of the main steam line isciathr, 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 lp I .
B&W-STS B 3/4 7-2 June 1, 1976 i i
PLA'lT SYSTEMS "N BASES positive reactivity effects of the Reactor Coolant System cooldown associated with the blowdown, and 2i limit th7 pressure rise within centainnent in the event the steam line rupture occurs within -cent im
-menh The OPERABILITY of the main steam i:ci:tica valves within the closure times of the surveillance r equirements are consistent with the assumptions used in the safety analyses.
hi.7.'.6 5EC0iiCARY WATER CHEMISTRY s s . ,\ r , n r. L _ '\\ U, m 3 73 % a rem gen e rare , rew v.,tre, mi"- I, d,% < 4 MU s 5%dl - ' ce,, te\ O ,,d rc[4c io n o f' o l' g c o r r o s ,'ve ;mp wri ' s ie Fle <.F a,n
._ 1 _ n - , %, , e, , s ,
t. y ,--,a mv- - ,, rn . n s m a e uu c acj , c.c;
,s. /
2/' ' ? STE? CENC'" TOR MEL' J n /T "PE;'ATM L:"! TAT:0N The li.T.taticr 3 ca ;/T, ger.e.eter pr ;;s\re nd tc..,cretur; cn;ur-- 1 that ressure indu d stresses in the steam erators de n ceed the naximum owabl fracture toughness stress li ts. T mitations of (70)*F and ( sig are based on a steam genera T NDT OI ( } and are sufficie 'to ent brittle fracture. 3/4.1.3 CWCli[iT COOL!NC WATG siSim.. TF OPERABILI the component cooling wa stem ensu 's that , suffi ies.t coo ', capacity is available for continue ration i sa ty r ed equipment during normal and accident conditi Th l rJ ' ant cooling capacity of this system, assuming a single fai is , g"N
.casistent with the ;;;umption; u;cd in th; ;;fcty analy:c;. ,
1
\
( 9' . B&W-STS B 3/4 7-3 June 1,1976 j
PLANT SYSTEMS qG v BASES 3f4.7.4 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 safety analyses. EMERGENC / cOCUNG POND 3/4.7.5 ULTIi4ATC HEAT CINk (OITIONAL) e ,a n < c e cce\' g pc. J J Pfk The limitations on the ulti(..;t) heat-sink ic.cl 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. te k Jeb de Thelimitationsonminimumw;tcrIrsyk l and maximum temperature are based on providing a 30 day cooling water supply to safety related equip-ment without exceeding their design basis temperature and is consistent with the recommendations of Regulatory Guide 1.27, " Ultimate Heat Sink 4i for Nuclear Plants", March 1974. v 3/4.7.0 ILOOD PROTECTION (0PTIONAL' ThC l!GiIGtiCn Gr. flGGd prGtictiGn in3sie5 het IdCil A ty psvicutivc 00tiOCS Will hC takCT. (and GECrGtion Will DC tCrainet6d) in thi iv66t Of fIced c00diti0n0. ThO li.it Of Ol0VOIi0n ( ) Ian 500 LCiCl i3 UG3Cd er the = i u: clevation at which facility Hocd centrol wee 3nres picvida protection to ;;fct3 c; lated equip;cnt. yy0l(TONING ANb All. FnTR ATIcel 3/4.7.7 CONTROL ROOM EMERGENCY AIR CLEANUP SYSTEM i ce,J,F:c43 uj Ar f.lk8ca The OPERABILITY of the controi room emergency air cl enup-system I ensures that 1) thc embicnt air tcapcrotur; d;;; not cn;cd th; allowabic tc perature for continuou; duty r; ting f;r th C;uipant ;nd instruantatien cccled by thi: ;y;tcm and 2) the control room will remain habitable for operations personnel during and following all credible accident conditions. 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 equiva-lent. This limitation is consistent with the requirements of General Design Criterion 19 of Appendix "A", 10 CFR 50. I ,( B&W-STS B 3/4 7-4 September 1, 1976
PLANT SYSTEMS BASES 3/4.7.3 ECCS FUMP ROOM EEAUST AIR CLEAiiUF SYSTEM (OFTIONALP l The OPERABILITY vi uiu [CCS pump .ces exheu5t dir cleenup systcar I cn,urcs that radisective meteriels leskir.g frem the CCCS equir, ant H +'in the pump rc= following ; LOCA arc filtered-peier-to reaching the enu rcamen t. Thc operation of this system and th; resultent effcct on off;ite do;;gc calculaticas werc assured in the safety analy ce. 3/4,7.9 HYDRAULIC CNUCCERS JHon Jurarssoas (suageggs) La smussm The hydraulic snubber; are required OPERABLE to ensure that the structural integrity of the reactor coolant system and all other safety lowin a seismic or other relatedinitiating event systems is maintained dynaaic loads. during and fol,d6%g,' excluded from this The only'?h inspection program are those installed on nonsafety related systems and then only if their failure or failure of the system on which they are installed, would have no adverse effect on any safety related system. N The inspection frequency applicable tc :nubbcr; ccr,tcining scal fabriceted f.cm retcrials which havc bcca dcmonstrated compatible with (. they of sn;,u gr,fing envirencent is based upon maintaining a constant level protection. Therefore, the required inspection interval ith the observed'3i$ M & Tailures. Thggumber97 varies inveyspgw'r?"found inoperable'Cnu _ during an inspection of these 'dubW,, deter-mines the time interval for the next required inspection,of thc;c :nubbcr;. 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 elapsed (nominal time less 25%) may not be used to lengthen the required inspection interval. Any inspection whose results require a shorter inspection interval will override the previous schedule. siaGgaw To provide further assurance of :n i cr reliability, a representa-tive sample of the installed'TSM'Tl'will be functionally tested during plant shutdowns at 18 month intervals. Thesa tests will include stroking of the"fdMUTto verify proper gi ton3 movement, lock-up and bleed. Observed failures of these sample w0bT5Ff'will requirg functional testing of additional units. To minimize personnel exposures,'EnMIE installed in high radiation zones or in especially difficult to remove locations may be exempted from these functional testing requirements provided the OPERABILITY eT these snobbecs was demonstrated during functional testing at either the completion of their fabrication or at a subsequent date.
~-
( B&W-STS B 3/4 7-5 September 1, 1976
PLANT SYSTEMS o V BASES 3/4.7.10 SEALED SOURCE CONTAMINATION The limitations on removable contamination for sources requiring leak testing, including alpha emitters, is based on 10 CFR 70.39(c) limits for plutcatum. This limitation will ensure that leakage from byproduct, source, and special nuclear material sources will not exceed allowable intake values. ( B&W-STS B 3/4 7-6 June 1, 1976
)
[g) 3/4.8 ELECTRICAL POWER SYSTEMS i [ SASES 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 cor.ditions within the facility. The minimum specified indepen-dent and redundant A.C. and D.C. power sources and distribution systems , satisfy the requirements of General Design Criterion 17 of Appendix "A" l l to 10 CFR 50. l l The ACTION requirements specified for the levels of degradation of I the power sources provide restriction upon continued facility operation 1 comensurate with the level of degradation. The OPERABILITY of the I power sources are consistent with the initial condition assumpt' ions of , the safety analyses and are based upon maintaining at least one of each ' of the onsite A.C. and D.C. power sources and associated distribution l systems OPERABLE during accident conditions coincident with an assumed l loss of offsite power and single failure of the other onsite A.C. source. ' Thc 0"CRACILITY ef the oinicomo specified A.C. end 0.0. pc ;r S,
= :: :nd =ecist;d distribstien ajateina during shutde n and cafucling ensurcs that 1) the fecility cen be meintain;d in the shutdr.c Or
)
r;fucling conditien fe extended time pected; end 2) suffici;nt in-tcu= tetien end cent.vl supebility is eveileble fer .aenitering and maint f ' i the resility 5tutua. l l .y, \ O (. B&W-STS B 3/4 8-1 June 1, 1976 l l l
3/4.9 REFUELING OPERATIONS (},, \ V 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 unifonn boron concentration is maintained for reactivity control in the water volumes having direct access to the reactor vessel. These limita-tions are consistent with the initial conditions assumed for the baron dilution incident in the accident analysis.
3/4.9.2 INSTRUMENTATION The OPERABILITY of source range neutron flux monitors ensures that redundant monitoring capability it 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 i 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 safety analyses. REMMR OdH.DMG-3/4.9.4 CONTAIN"ENT PENETRATIONS rmte bA8: The requirements on c atainmca$tpenetrationclogureanL9PERABILITY ensure that a release of radioactive material within 'idKilhacE! will be restricted from leakage to the environment. The OPERABILITY and closure requirements are sufficient to restrict radioactive material release from a fuel element rupture based upon the lack ofF"ccate" U'at pressuriza-tion potential while in the REFUELING MODE. ' 1 3/4.9.5 COMMUNICATIONS The requirement for communications capability ensures that refueling station personnel can be promptly informed of significant changes in the facility status or core reactivity condition during CORE ALTERATIONS. B&W-STS B 3/4 9-1 January 1,1977
REFUELING OPERATIONS Q+ g BASES T _$., ,/ 4 . o . f. Pflet IJ A ttfit *in ?fier" r% ri r'.n u a ,mATv u tnt e.. . . . . T.RIM . vu umawuu vs Ls T k. . a. A. D. -. E . M A.D. m T. t T T. ._mm.4._..~..s.,. 91 ,m,r,
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RH %C il.l A RY 3/4.9.7 CMNE TRAVEL - C""NT IU~L ST0"",C" 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 poul ensures that in the event this load is dropped (1) the activity release will be limited to that contained in a single fuel assembly, and (2) any possible distortion of fuel in the storage racks will not result in a critical array. This assumption is consistent with the activity release assumed in the accident analyses. 3/4.9.8 COOLANT CIRCULATION
?
The requirement that at least one decay heat removal loop be in 5 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 reactor core to minimize the effect of a baron dilution incident and prevent boron stratification. . , . _ , , . . , . , c.-,,- -o y w , }$
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p '\ \ s'." ).* 'M)t.[ } T,$. MN.g.,s b I i j I J.,,/Ti .jd.ss - d % k '\g g';f'{ Ny} t r- %. - LOW POPULATION ZONE FIGURE 5.1-2 l .( B&W-STS 5-3 October 1, 1974
q' \ DESIGN FEATURES v' DESIGN PRESSURE AND TEMPERATURE S.2.2 The reactor containment building is designed and shall be maintained for a maximum internal pressure of 5 4 psig and a temperature of g F. I 5.3 REACTOR CORE FUEL ASSEMBLIES 5.3.1 The reactor core shall contain 177 fuel assemblies with each fuel assembly containing aof fuel rods clad with {Zircaloy -4f. Each fuel rod shall have a nominal active fuel length of /44 inches and contain a maximum total weight of n?/ grams uranium 7he initici cccc-lo; ding hcil h n ; ; m a imum caric.tcat of wcight pcrccat L' 2&- Reload fuel shall be similar in physical design to the initial core loading and shall have a maximum enrichment of 2.5 weight percent U-235. The first cycle fuci lecding :hcil contcia buracble pci;cn rod n:cablic; with ccch ;;;=bly contcining up to burncble poi:cn rod; cf ;intcred Al C, z LC cle.d with Zirccicy ?. CONTROL RODS 5.3.2 The reactor core shall contain el safety and regulating and 7 axial power shaping (APSR) control rods. The safety and regulating control rods shall contain a nominal 134 inches of absorber material. The APSR's shall contain a nominal 36 inches of absorber material at their lower ends. 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. t l l 1 June 1,1976
~
B&W STS 5-4
(). v ry DESIGN FEATURES 5.4 REACTOR COOLANT SYSTEM DESIGN PRESSURE AND TEMPERATURE 5.4.1 The reactor coolant system is designed and shall be maintained:
~~a. In accordance with the code requirements specified in Section~~
~~'f.ifE-EF of the FSAR, with allowance for normal degradation pursuant to applicable Surveillance Requirements.~~
b. For a pressure of 2 500 psig, and

c. For a temperature of Lso*F, except for the pressurizer and .
~~pressurizer surge line which is 670 *F. VOLUME 5.4.2 The total =ter and :ter volume of the reactor coolant system is cubic feet, et ; neaiias T~~
~~& er ':::N"r.~~
I JI,9o o 5.5 METEOROLOCICAL T0'[R LOCATION (5.1,,Thenieteecelegicelteershellbeiscetedesshownonrigure w ....,. 5.6 FUEL STORAGE CRITICALITY 5.6.1 The new cnd spent fuel sterage racks are designed and shall be maintained with a nominalal.b inch center-to-center distance between fuel assemblies placed in the :tcr:g racks to ensure a k eq
< 0.90 with the Ctcr:g; pool filled with unborated $$(er.uivalent -Thc k toOf eff 0.95 Schde a cen:crvativ; ellewence of (3.0)% ak/k fer encerw.ntic Tc describcd in Sectica (0.1) of the T5AR. The spent eel ru.ks are des,j neJ ancE M\ % ma',s +d.-ed wih o. non,,md 13.5 :nc k ce der +o cen+e r LVa s ce he kee n bl aswmbl;n phued in %e rads +o ensure es L, c.I W u;lanF b 6 0.95 w;R & rel DRAINAGE #.It,J wak wWeersed waer.
5.6.2 The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation a9 7 #4 O j ' B&W STS 5-5 January 1, 1977
O oeston reATuaes O CAPACITY 5.6.3 The spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than g fuel assemblies.
~~- - ~ . . . . , . . . . ,. -- , . . . . , . . . , , , , , , , -~~
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~~v. u_mt i. m_ .,., ..~~
~~v# 7 1 lv B&W STS 5-6 January 1, 1977~~
"SLE 5.7 1 l to CCMPCNCNT CYCLIC OR TRA'iSICNT LIMITS b-Compo nt or System Cycle or Transient Limit Design Cycle or Tra ent
~~1. Reac or Coolant (240) Heatup and Cooldown (70 F to 557 F 70 F)~~
~~Sys m Cycles~~
~~2. Reactor olant (160) Step Load Reduction (100% to 8" RTP*)~~
~~System Cycles (Resulting from turbine trip)~~
~~3. Reactor Coolant (150) Step Load Reduction ('O to 8% RTP*)~~
~~System Cycles (Resulting from electrical load rejection)~~
4. Reactor Coolant ( ) Reactor Trip Cycles Reactor Trip System esulting from loss e ctric power to al RC ps)

5. Reactor Coolant (160) Rea or Trip ycles Reactor Trip System (Resul 'ng f m turbine trip wit automatic O control ion)
~~V' 6. Reactor Coolant (40) Reac r Trip ycles Reactor Trip System (Res ting from d wi drawal accide )~~
7. Once Through (8 Reactor Trip Cycle Reactor Trip Steam Generator (Resulting from comple loss of all main feed-water)

8. Cnce Through (40) Reactor Trip Cycles R ctor Trip l (Resulting from loss of Steam Gene tor station power) '

9. Once Th ugh (20) Reactor Trip Cycles Reactor rip j Stea Generator (Resulting from loss of \ '
~~feedwater to one steam l generator) l~~
~~RATED THERMAL POWER O '~~
f,) ' B&W-STS 5-7 September 1, 1976
kt T=L: S.7 : 'c;ntinued) I "
~~%pc-+ c- Sys ter Cyc!~~
~~cr 'rcncient Lim il Oc ; i c,7 C, ; a :w 4,~~
10. Once Through (10) Reactor Trip Cycles Reactor Trip am Generator (Resulting from stuck open turbine bypass valve)

11. Reactor Cool System (80) Rapid Depressurization (^ psig to 300 psig in one hour)

12. Reactor Coolant (2 hange of Flow es Loss of one or more System RC pumps

13. Reactor Coolant (20) Hydr at Test Pressurized to >
~~System (3125) psig~~
~~14. Once Through 5) Hydrostatic Tests Pressurized to >~~
~~Steam Generat 125) psig~~
15. Reactor olant (480) Test Transients High P sure Sy m Injection t

1. Reactor Coolant (240) Test Transients Core Flooding Chet.
~~System Valve Test s' s B&W-STS 5-8 September 1, 1976}}~~

ML19326B363

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