ML20005G422
| ML20005G422 | |
| Person / Time | |
|---|---|
| Site: | Byron, 05000000 |
| Issue date: | 11/17/1989 |
| From: | COMMONWEALTH EDISON CO. |
| To: | |
| Shared Package | |
| ML20005G418 | List: |
| References | |
| NUDOCS 9001190135 | |
| Download: ML20005G422 (16) | |
Text
{{#Wiki_filter:... 4 f i (< n. t ATTACHMENT B l 1 l: PROPOSED CHANGES TO APPENDIX A TECHNICAL SPECIFICATION OF FACILITY i OPERATING LICENSE NPF-66 L, - l-Revised pages: 3/4 4-34 4- 3/4 4-36 3/4 4-39 3/4 4-40 03/4 4-8 B3/4 4-11 83/4 4-12 B3/4 4-15 B3/4 4-16 4 35 891337 YhjiljjjCKosooo433 P PDC
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L l 11 ll l 1'l' II I I l I I Af I I6: I , i l 4 i (Javono ta'st uEnt l ii l i g amissis6 ruersan vanis ',_ IACCEPTABLE FORI
\ Mf DRO. TE5Til tearseLLies eestra sat et v,Pi,tt' i IntLL le . less 1i I hf i '/ ij j tePPts teattats teeltavatiotL7 ,y } f ( g assunto as e.se vt t ,f g g y
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2000 I g 1/st. isl4 g- sta t. te% i i l I- - l 1 , 3 , 1 l
,vevt APPLicaskt Fee stafve aatts i J
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' vP to see t/me ran fut stevitt ,
Presa et to as tret. ano contains j j E ;
'"~ 'usatius er iser Ane se Pass Fee '
fl / Passiett instevutar taeans d 1/ lI / I iI ( I / 4 I UNACCEPTABLI ' f / ACCEPTABLE-- ' 7 OPtuiION N-i 1000 0 PEUT!0lt f ; i - I. jjg i f i f stator natas W / I' F i ve to see'r/me N / I" l 1t i cf pd 1 l 1, IiI I
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n.eeesstarit
. o en instavict-
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i tI i < n' itst+- - ! t'astitatiyt 4 Ii ttwetaatuet (aft *r) . L l r m int stevict I i' uniti l i l3 Ptsies vP,te nettre ' l t 1 ll ll l l ll l l i,g- i ll l ) .li il i l l 11'I O
- 100 200 300 E00 inicarro itwturns (*P)
FIGURE 3.4-2A REACTOR COOLANT SYSTEN HEATUP LIMITATIONS APPLICABLE UP TO 32 EFPYI(UNIT 1) ' 4- Areliukky Au beea redad BYRON - UNITS 1 & 2 3/4 4-33 y a,. uei.n mz h 21.5yevs.
TNGRT .1
* . N HRTI2IAL MElPEm 84113 Controlling Meterial : ifald Metal Copper Content
- Ccaservatively ass esd to be .06 trfl Phosphorus content (aetuna costant a 0.06 nff5)
- 0.009 WF5 Initial N g Conservettvely seemed 40*F R g Aft.or 16 EFPr : 1/ E* '
period . 1 e , r i.et . te. . ,to ,,,, ,e,. 9 to 16 EPPF and oesteams e poseatte instament errore and 60 pass for l l l-i i I l l l l l l l 'l '! ggg, LEM TUT LDUT I j i l l I , I E r r - J J
- I I
I f l~ l gg IEATUP UP - i i 10 100 F/W % " - au',gpfagy t . l? N/ l orsaar20s _
-.. .DimxisPtaaLa . 2 -
-.. .opgggitog r r ;
Itase
> z I I I . _ _ .
i 01 11' 1 I @1T i.1321, i . m -_ x - ase s, aN"f ,
... .. i.
~ .. .e . .n N ;
FIGURE 3.4-2b REACTOR COOLANT SYSTEM HEATUP LIMITATIONS APPLICABLE UP TO 16 EFPY (UNIT 2) BYRON - UNITS 1 & 2 3/4 4 34 l l a--,-,-.-e-, --.,.,,-n....,--..,,--a - - - . . _ . , - ,
r' NATERIAL' PROPERTY BASIS i, CONTROLLING MATERIAL: CIRCUMFERENTIAL WELD RT AFTER 16 EFPY: 1/4T,146.5'F NDT 3/4T,122.8'F I-CURVES APPLICABLE FOR HEATUP RATES UP .TO 100'F/HR FOR THE SERVICE PERIOD UP TO - 16 EFPY. CONTAINS MARGIN OF 10'F AND 60 PSIG FOR POSSIBLE INSTRUMENT ERRORS. i 2500 u - -
- i iiiiii, ,,,,,, ,
, ii ' i 1 iiiiiii i,,,,, , '
Leak.lest - i i
+
l
~ ' ' '
2250 - Limit r , i I I
- i i V J B 2000 ,- ,-
i i
.i i , , ,
1750 l l:,
.: 'l '
/ Acceptable ' '
, i G 15c0 ,
l Operation ,,;, ,;- E ll' ' ' l:
~ / !;;'
Unacceptable f , , ll,; ; l. i l3350 ;
- ll Operation , ; ,f' f li i-g i., ,, , , , , 1, ., .
10*0 .' ' ' ' ' ' ' ' ' l .,; 3- ; ,' Heatup Rates ' l 'l,, , Up To r i > . 4 - 750 inn'F/Hr E l .. / '
- f '
e
, 'i Criticality Limit 5 Based on Ir. service t !
500 .
/ Hydrostatic Test --
Temperature (292'F) ~: 250 for the Service :: Period Up to 16 EFPY- - 0 50 100 150 200 250 300 350 400 450 500 lN01 Cafe 0 TEMPERATURC (DEC.F) 1 FIGURE 3.4 2b REACTOR COOLANT SYSTEtt HEATUP LIMITATIONS APPLICABLE UP TO 16 EFPY (UNIT 2) BYRON - UNITS 1 & 2 3/4 4-34
~3tuttT 2 '
i 9 d
,llE!IIIAL Pet)Pt m Batta t i l .
Soutrell Noterial : Meld listol Copper : Conserntstely assend to to 0 ' Wrs ftionssieres Content W oostest e 0.06 Wrs) !
- 0.000Wrs ,
ID1%181 Efg Caservetteely esamed to 40 % en m ee w apyr ) g ***1 Bar * - ' i 3 period g te W EFyr one emassas artaas or, and 80 psig fer ! poss8 tile $astemaet arters sees
/ o i ,.
)
I I l !. see, l'
,. { .
K nk KA , ll l0WHX)N r i Itese r J 5 r l m ansl% ggg j (PERLT32f F I .. g g62 ses
- ec a-X'. '
= g,.
j ; __ E j j m as EE NE .pu' .N E <uel < N MI <H ui G Ni W j m M-FIGURE 3.4-3b REACTOR COOLANT SYSTEM C00LD0WN LIMITATIONS APPLICABLE UP TO 16 EFPY (UNIT 2) BYRON - UNITS 14 2 3/4 4-36 s
i i r MATERIAL PROPERTY 8 ASIS ! i CONTROLLING MATERIAL: CIRCUMFERENTIAL WELD ; RT NDT AFTER 16 EFPY: 1/4T 146.5'F ; 3/4T,122.8'F CURVES APPLICABLE FOR COOLDOWN RATES UP TO 100'F/HR FOR THE SERVICE PERIOD UP l TO 16 EFPY. CONTAINS MARGIN OF 10*F AND 60 PSIG FOR POSSIBLE INSTRUMENT : ERRORS. 2500 m - ,, , ,, , , , , , l 1 8 r . i t ! 2250 l l l , l, - E , i. 4 , r t i , I i j
, e i i i )
toot , , - l l l ,' lll ,
, , , 1, ,
. i t f ' I r
, , e 6 1 a ,l 3 6 I ] 6 i i! e ,
1760 l l, l
'f r
l ie ll,,' i i . , l
, , , , ii ii , i e e i a ! i t II i a f
7 160o,,,,l', , l / l, , , i . ,, ,
.; i , ,, , i I i i .
e.,, . ,, ,, ; , lll 1 , - w 1450l Unacceptable a ,i 4 Operation -
/- Acceptable l l
$ ,,,, H}II} ,; , ;
!.,,'. Operation l li, i p c tesol -
, , , , , . . ,,..,i .
i . .. s,i4 . . i ,,, o t i,I ii i
. i i 6 a i 3 , i i6 ,! ;
,M, e
taJ
- .. , i , . , , i 4 ., ,
. . . , -i, , . . . . . . .
Y 75,p Cooldown i , - - 4 + . +i i . . . . i ll
! c Rates ,2 7 l l l l l l s->, , , . .
u oppg7 . 500', r
, , O e'x wer e s r ,
20 ,-., , , , , . 250 40 - -
, , . o0 3
100 i
+
. i , ii4 ,
- i 4 .
l i,l ! e I f 0 50 100 150 200 250 300 350 400 450 500 > lNDICATED TCWPCRATURC (0CC.r) I FIGURE 3.4-3b REACTOR COOLANT SYSTEM C00LDOWN LIti!TATIONS APPLICABLE UP TO 16 EFPY (UNIT 2) BYRON - UNITS 1 & 2 3/4 4-36 L
l l REACTOR COOLANT SYSTEM OVERPRES$11RE PROTECTION SYSTEMS 1 LIMITING CONDITION FOR OPERATION 3.4.9.3 At least one of the following Overpressure Protection Systems shall be OPERABLE: l!
- a. Two residual heat removal (RHR) suction relief valves each with a Setpoint of 450 psig
- 1%, or i
- b. Two power-operated relief valves (PORVs) with lift Setpoints that vary with RCS temperature which do not exceed the limit established in Figure 3.4-4^or ce uea 1 (Fi< pre L4 -% & uwd Q, J
- c. The Reactor Coolant System (RCS) depressurized with an RCS vent of greater than or equal to 2 square inches.
APPLICABILITY: MODES 4 and 5, and MODE 6 with the reactor vessel head on. ACTION: 1
- a. With one PORV and one RHR suction relief valve inoperable, either I restore two PORVs or two RHR suction relief valves to OPERABLE status ;
within 7 days or depressurize and vent the RCS through'at least a ' 2 square inch vent within the next 8 hours,
- b. With both PORVs and both RHR suction relief valves inoperable, '
depressurize and vent the RCS through at least a 2 square inch vent ; within 8 hours. ~
- c. In the event the PORVs, or the RHR suction relief valves, or the RCS vent (s) are used to mitigate an RCS pressure transient, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 30 days. The report shall describe the circumstances initiating the transient, the effect of the PORVs, or the RHR suction relief valves, or RCS vent (s) on the transient, and any corrective action necessary to prevent recurrence. ;
- d. The provisions of Specification 3.0.4 are not applicable, BYRON - UNITS 1 & 2 3/4 4-39
s psec- 3 --
\, - . _
.\ --
.g __ .
\s _l . I- --
g 900 1
\
[ ; i- t- - g\ i j N _ Selected Points on the Curve: e PS '
!. RTD MAX . ._ .
L ' \\ 70 535 97 535 550 2 800_
\147- 227 .
580 : ; 2 277 s 800 ;; 327 800 : L g i
\ .
1 S :: tim .
-? :
. t
;; 4____
k 700.
.f
, 1
- a. ..
= , ,!
g : . i ! g , 1 i x .: y
, ,r ,
x 600 ',' ' i i , 1 , 4 E ,# i -- x ,- n _ - ..
/ _ . . _ , .
__i - # __ 1 T. -,_ _ _ . . _ _ _ _ 500 - 1
'~
i
/ 50 100 200 300 400 T
RTD - LOWEST COMS RTD TEMPERATURE (*P)
/
/ FIGURE 3.4-4 NOMINAL PORV PRESSURE RELIEF SETPOINT VERSUS
/ RCS TEMPERATURE FOR THE COLD OVERPRESSURE PROTECTION SYSTEM 1
/ APPLICABLE UP TO 10 EFPY -
,j BYRON - UNITS 1 & 2 / -40 MN %. O
W- ! i i i n d
- _.e
._. _q . .
, m e_- ..._..s.-.. ;
-~ - - , - . . _ , . . l
_ . . a _ _ . . . . _- ,.,....... .., i
' 4 ' - - -- '
900 --
-- -- -- -- --+--
1
...i ..
1 -. . _ _ _ _ , . _ . . .
..r.. ~ ~ ~ ' ~~ ~ '--~ ~
. . Selected Points on the Curve:' . "~ - . _ . . . ,
l
--+ PS
- - - - - + - - - + - - - -
1 L- TtTD MAX . .. ... , 1 5 J
~ ~ ~ ~ ~ ~ ~
70 535 . ..--_._. l
} 1 97 535 - .....
..+.- ~- .
147 550 - - - - _ soo
- !! , 227 580 : _, _ _; --
g - 277 800 1 _ p .-- 327 800 :
. . + - - _
b
, +-
l : -+-- -
.E '
- E i
__ r e
; ; - _s 700__ =
, ; i 5 i -; ,
1 . 1 1 I N _ i- -- ,- 1 F I f _4 -
- 4 i
;- +_~Ef .
1
. . _. .r _ T t
600 ; ! ,' '
, ,f ~
---r-1: - _ . . 1r t --
i . _ _ -
.we_. a q - _.
-n __
-..4e. . . .
l
, _ _ . . ..._: _... . _ _ . . . . . _ .. _ . . _ . . . . .s _
r [ ..
. _ .. . . _ . ._4
..; ..+ _.._ ...
4 . 500
..y_. _ _ . - , . ...'y..- - - .. ~~
_..-.,. _.y- , _ g. .... - . _ . l
..r .
. . . . ,. _ _ _. .. _ .T.. . ,.. .
50 100 200
- 300 400 1
T,,, - we m m mum ( m i FIGURE 3.4-4a NOMINAL PORY PRESSURE RELIEF SETPOINT VERSUS RCS TEMPERATURE FOR THE COLD OVERPRESSURE PROTECTION SYSTEM APPLICABLE UP TO 10 EFPY (UNIT 1) BYRON - UNITS 1 & 2 3/4 4 Da
- - - A a---.a s a eo,_.e4 - -
- -e
)
l l
- == =:.==== = = = =b - =3__ ~
i.
== b .-
E g;3 = =CC'3 bb bb b==- == -- =: = = bb - -
= = - __===
-. = = ===:'.======d=:;=g = . 'g; =ll=l:E. _ _
[ __ -. 4 = == :;l= == -
- u ~
tA ?5^'?' 5 5 5 5 5'5 M 5555 e . . l EEEEE: b= ==: - m =r = = = = = = ==:== === ===: :: Sii$5-EE x= ;~ i s= =MaaEE E M'-. OsaEsism
- == -
- =;:EE;EEEEiEsiE= g m = = -
_9:=u
.. =============__====e j r= ~--i @ S 5 =' :E M Z E:"EE
~ ' -
M55-5E . . l Ei=4 5 EE _ a = s- E=f p =E &h ME=,ss=E===~~-~ a=2 g= a a := w= w= a e an =m t _ i T_E'=;;~= [ - U L _ Z--. E.-l E E 22 s- E s -E:-1.Z9! ^
]
E = ii= =2 g --. . . . _ . _ . .
-- .' : . . . =.: 7:_. . .W E= 4. . %_ _2 k__5_
[ 3 -- . . [ . . ' 4m . I k b :! h N bk =J. 5 :~ 9 I.5 I - - :.I hi b b k[ I J . -[i - I
=== a. = E=_ s a s em a. s. = m Es. a :: i.== E- a. r e H: r .m = - = : E7 s a s a s. __ ; .. -
L.' "- -
'l
? .i'- fi-- E ET) - ~-_'- i ' [ . . .II. [ = :. " '-~-E " . ;7 : n3 f I
3
- n s:..= E[$ d c- 5' .-i : r:h_ is := k- : :-
= :- = = = =
' .: bN. b b b T b '_ [ ~ : [..i._.__
5 . - . . _ '. ;= _bd [ _ . _ . . . - i . =. . (( .s -i ; = -.; -- . N ---_ Er = = = !==. x : = - . . -- -:: . m L :i- r -I ; - v c y -- -. mwys=m=
-- i ==. :._i _ = ==d Z.
==
t s=s: ma i== c i-: = = : : = - s = r [: : nv se ra m asn#soso:ssrs:orz. . .w- . . . .E e ;
~=':=;..... 2 .==.
n
..c..,.
s a s a s s = =_- .=: 4
-- .. : .. __ = _
{
.. {g = = .J =r r= = ; . C ' . . . '- 2 . { . . a. ;._ . j g g: gg -- g _
- i: - :. . ' ' ' _ f 7 1- -- '-: =~ =' ='=== = r '
-g 9 .= : = = . = = _ _ _ _ .'
=-- - il_-i_^ E - r . F. J; -_ E J- . 11: r ': c
==== ::7 = = = := : : :p = : - . r= r= = ; ; : m . = :,= . . m . -- n . r ,
l -~
=
==shb
-2===..___
N52 - bN5 5-N F~ E S'i M- 5
===:==p=y = =. =.=. g =. =t = c k . --. b be E b b U b ba bE b -*-- b b b EE.U .
E' I N d N b N5bb 7 s :== b .E.; EE 3E d5 _ __ _ z===---=
. . = =. ===:-
a 1 = = === 3 555 g- 53 ---_ . ; 5:._ 5_ 3 3 :;_3 p: 3 ;=:; _ _. . _ ; 3 g g 55: q = 7 y 9 .__ _. _ .- . === == . =E = = = : . = : 7 .m.;;._c_._..=c-2.z.==.=-.=_=:-=:==--p------___-. , ha --' _== = =5- B =
=---w_==----=:--=-==--=======: &: E ' ---- = si = = c i := = E--= --br - = == - = _ - - . - - _-da;;=~---. - "_
. * . .l..-- .- --
_ ~ _E ( _
- : .. .- .. .. . . :. : . :: == . ::_ .
y =- =: = - - r __. =======.==========_===-====*w===ss
~
N bbb bh5M 55.M i55- I'3 hb =~ h5 b - b U ' 1) 5bhh55h 1__ _. I
~~ hR sssM=M-55 55E E= E5EEEP =EE 5= 2+ = EE NE NE= _SEEM- ~~R I
c
===.=_--=-i === Eses ms= a 2 wa ms_ r _.E z a a :_a =s . -w - =a - am - -==-== ==.. e>-
(. =~^x-----=-y=r_..:...---==-----_---m.~==;rn-=---.=. g- y gg g a g =; g _=. g w .= ig ; g = ug.==_ - =:=. _- = --- t
= = _ __
- == E== =- :3:~ s E E E E a: ~_ E E s~:= - E s E s EE E E _
_ _ -- - =~ . =: E== : E =E E E 'l~~ ==;--E===E'~'-.=====__ :12
- d= Ei 55 EE i= - _= Z ._ i== =s == ;; E _f. iE=EE ==-E == 5= ---
. E sEr'.C.Z ~
= = = = = = _ _ = = - . - - _
..x===m.::.;====~=========
.'__~~~J
===
a:Ei { 5 -
---W*-~~--'
b M Eb b-b N b .5bbbNb5 5-bNb5 E!E ~ENE' bh5 Ebb--__N E 500samaE=sa =5b-b Nb=NIN5bb@N s=a =e s s:E=awsi ss 5NDN55bIU es E== gg e s as=E s ._: 2 a EEEE EEi== EE'EE s- EE is =-i zE= EE E= 5 =-- E-
=-=--===-~====;==- 2. EE E i.= E rid E Ei EE a Esi=i E=iEE
==r----=-------=-~=: = -m--- o -
- ----b - . . . .
a=_ ==== _ s= a sees a ss s= _ e a Esa s EE Es a si a sE s= a s= a a sa Eas===:=N
=---_.-=------r--r=-==-=----------=--=-=======____-
Ml===
= ==s :
o esE E Es sEEE EE EE a sa sea Es a s a sEs Esm ----======r==---
--r-==---====--=-------------=-m-= E. =-a s a Es"s"= a sE Eas.==!=s =- r- ==. e==sh 50 100 200 300 400 l
TRTDLOWESTCONSRTDTEMPERATURE(DEGF) FIGURE 3.4-4b NOMINAL PORY PRESSURE RELIEF SETPOINT VERSUS RCS TEMPERATURE FOR THE COLD OVERPRESSURE PROTECTION SYSTEM l (UNIT 2) BYRON - UNITS 1 & 2 3/4 4-40b
gadidio,, EmbrdI"'"9 /kachI- N REACTOR COOLANT SYSTEM ; { gj Mrjah BASES
~
l PRESSURE / TEMPERATURE LIMITS (Continued) Heatup and cooldown limit curves are calculated using the most limiting value of the nil-ductility reference temperature, RTNDT' *t th' '"d 'I l 32 effective full power years for Unit 1 (16 effective full power years for Unit 2) of service life. The 32 EFPY for Unit 1 (16 EFPY for Unit 2) service L
> l life period is chosen such that the limiting RTNDT at the 1/4T location in the core region is greater than the RT f the limiting unirradiated material. f
{ NDT ; The selection of such a limiting RT assures that all components in the HDT Reactor Coolant System will be operated conservatively in a'ccordance with applicable Code requirements. The reactor vessel materials have been tested to determine their initial RT ; the results of these tests are shown in Table B 3/4.4-1. Reactor opbtionandresultantfastneutron(Egreaterthan1MeV)irradiationcan cause an increase in the RT Therefore, a adjusted reference temperature, NDT. based in question,uponcan thebefluence, predictedcopper content using Figur B 3/4.4-1 an(d Mmsphents and content of the mate ' ART NDT computedbyeitherRegulatoryGuide1.99, Revision)thelargestvalueo ,
"4f,fe$tysy/ ,
DesMdat'Didebts/ofPgrtd ete(AaWat4n4Mi4e9046acteAss&. Mat 4615" l or the Westinghouse Copper Trend Curves shown in Figure B 3/4.4-2. The heatup ; and cooldown limit curves of Figures 3.4-2 and 3.4-3 include predicted adjustments for this shift in RT NDT at the end of 32 EFPY for Unit 1 (16 EFPY for Unit 2) as well as adjustments for possib temperature sensing instrume ts.}.7N>rgrgl le errors in the pressure and Values of ART determined in this manner may be used until the results NDT , from the material surveillance program, evaluated &ccording to ASTM E185, are available. Capsules will be removed in accordance with the requirements of r ASTM E185-73 and 10 CFR Part 50, Appendix H. The surveillance specimen with-i drawal schedule is shown in Table 4.4-5. The lead factor represents the rela-tionship between the fast neutron flux density at the location of the capsule and the inner wall of the reactor vessel. Therefore, the results obtained from the surveillance specimens can be used to predict the future radiation damage to the reactor vessel material by using the lead facto.r and the withdrawal time of the capsule. The heatup and cooldown curves must be recalculated when the ART NDT determined from the surveillance capsule exceeds the calculated ART NDT for the equivalent capsule radiation exposure. Allowable pressure-temperature relationships for various heatup and cooldown rates are calculated using methods derived from Appendix G in Section III of the ASME Boiler and Pressure Vessel Code as required by l- Appendix G to 10 CFR Part 50, and these methods are discussed in detail in l WCAP-7924-A, " Basis for Heatup .and Cooldown Limit Curves," April 1975. BYRON - UNITS 1 & 2 B 3/4 4-8 . 1 I _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . .- - -- . . _ - - . , - - - - - - . - - - -
y
. INSERT 4 Revised heatup'and cooldown curves have been generated for Unit 2.In accordance.with Regulatory Guide 1.99 Revision 2. For Unit 1, the curves remain;the same. However, the applicability date has been reduced per RG 1.99 Revision 2 to 29.5 EFPY for heatup. The Byron Unit 1 applicability date of 32 EFPY for cooldown remains the same.
^ Y) s
/scl:0386T:10
TABLE 8 3/4.4-la' REACTOR VESSEL TOUGHNESS
. , (UNii 1).
E M Average' Upper Shelf Energy w Normal to e- 50 ft-1b ~ Principal- Princip m y 35 mil Cu P MOT RT "'III"9 ""I COMPONENT Heat No. Temp. NOT Direction Grade -{%} (*F) Ofrecti< Closure Head Dome C3486-1 A5338 CLI .10 1%}
.016 -10 Q*] 1"F] (ft-Ib) { f t-1(b
< 40 -10 151 Closure Head Ring IV4566 A508 CL2
.11 .007 20 < 80 20 125 Closure Head Flange 124K358VA1 " ---
j .011 60 <100 60 Vessel Flange " 145 --- 123J219VA1 ---
.012 10 < 70 10-Inlet Nozzle " 152 ---
, IV4684/3V1320 .12 .008
" -10 < 40 -10 117 y "
IV4684/3V1320 "
.12 .008
" " -20 < 40 -20 116 ---
[ IV4695 "
.13 .007 y " " -20 < 10 -20 116 ---
IV4695 "
.12 .006 -20 < 10 -20 119 Outlet Nozzle IV4656 "
.11
" .007 0 < 10 0 131 j, "
IV4656 "
.11 .007
" -20 < 10 -20 131 2V2557 "
.11 .007 -20 < 10 -20
" 112 ---
2V2557 .11 1
.008 -10 50 -10 94 Nozzle Shell 123J218 "
.05 .010 20 < 70 20-Upper Shell 4 * "
138 184 SP-5933 .05 .010 (75) 40 <100 40 139 Lower Shell W1 SP-5E51 " 156
.04 .014 / H) 10 < 70 10 Botton Head Ring 150 160 IV4fd2 " ---
.012 0 < 60 0 115 ---
80ttom Head Dome C2815-1 A5338 CLI .19 .009 -30 40 -20 Upper to tower 118 --- WF336 ---
.024-Shell Girth Weld H .010 (,70)-30 30 -30 77* ---
- Normal to Principal Welding Direction
,,o,
. men e _ .
f h h V Y
. . . ~. .. . - __ _ _ . _ _ _ _ _ _ _ _ . .
t ss TABLE 8 3/4.4-1b o z
. REACTOR VESSEL TOUGHNESS
,(UNIi 2)-
E O v$ Average Upper.. o- _ Shelf Energy to 50 ft-Ib Normal to Component 3b "iI Principal Heat No. Grade Cu P T NOT " Princit Closure Head Dome C43/5-2 A533 8, C1.1 Q .@ ("F) Temp. RT NOT Direction "9
- Closure Head Ring 48C1300-1-1
.12 .013 @
<20
("F) _(ft-Ib) Directi _( f t-it Closure Head Flange A508 C1.3 .05 114 2029-V-1 .007 -30 -C A508 C1.2 <30 -30 Vessel Flange --
.011 108 124L556YA1 " 0 <60~
Inlet Nozzle --- 0 157 51-2979 .008 30 -- tn " " <90 30
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REACTOR COOLANT SYSTEM BASES L PRESSURE / TEMPERATURE LIMITS (Continued) j,,p m fj ThefnotchinthecooldowncurveofFigure3.4-3)Wduetotheaddedcon-straint on the vessel closure flange given in Appendix G of 10 CFR 50. This hydrostatic test pressure, the flange regions that are highly s bolt preload must exceed the RT i NDT of the material by at least 120'F. The flange RT required.NhD + 120'F impinges on the cooldown curves and therefore the notch is n izfS 4pm /)gc 1 hb /mn de/em,M sklics/cr -/hd Mc wm/ c/cwn /l%je p AnKy. _ Three for finite separate heatup calculations are required to determine the limit curves rates. As is done in the cooldown analysis, allowable pressure-temperature relationships are developed for steady-state conditions as well as finite heatup rate conditions assuming the presence of a 1/4T defect at the inside of the vessel wall. produce compressive stresses at the inside of the wall that alleviate theThe the tensile stresses produced by internal pressure. crack tip legs the coolant temperature; therefore, the KThe metal temperature at the during heatup is lower than the K IR for the 1/4T crack IR for the 1/4T crack during steady-state conditions.at the same coolant temperature. During heatup, especially at the end of the transient, conditions may exist such that the effects of comp'ressive thermal stresses and different KIR's for steady-state and finite heatup rates conditions no longer represents a lower bound of all similar cu heatup rates when the 1/4T flaw is considered. Therefore, both cases have to be analyzed in order to assure that at any coolant temperature the lower value of the allowable pressure calculated for steady-state and finite heatup rates is obtained. Thesecondportionoftheheatupanalysisconcernsthecalculdionof pressure-temperature limitations for the case in which a 1/4T deep outside surface flaw is assumed. Unlike the situation at the vessel inside surface, the thermal gradients established at the outside surface during heatup produce stresses present. stresses which are tensile in nature and thus tend to reinforce any pressure These thermal stresses, of course, are dependent on both the ramp.rate of heatup and the time (or coolant temperature) along the heatup Furthermore, since the thermal stresses, at the outside are tensile and increase with increasing heatup rate, a lower bound curve cannot be defined. Rather, each heatup rate of interest must be analyzed on an individual basis. Following the generation of pressure-temperature curves for both the steady-state produced and finite heatup rate situations, the final limit curves are as follows. point comparison of the steady state and finite heatup At rate any data.A composite c given temperature, the allowable pressure is taken to be the lesser of the three values taken from the curves under consideration. BYRON - UNITS 1 & 2 B 3/4 4-15 . l J
REACTOR C00LANT'$YSTEM BASES PRESSURE / TEMPERATURE LIMITS (Continued) The.ure of the composite curve is necessary to set conservative heatup limitations because it is possible for conditions to exist such that over the course of the heatup ramp the controlling condition switches from the inside to the outside and the pressure limit must at all times be based on analysis of the most critical criterion. i I. Finally, the composite curves for the heatup rate data and the cooldown ' rate data are adjusted for possible errors in the pressure and temperature 1 sensing instruments by the values indicated on the respective curves. L Although the pressurizer operates in temperature range's above those for which there is reason for concern of nonductile failure, operating limits are provided to assure compatibility of operation with the fatigue anaiysis ! performed in accordance with the ASME Code requirements.
+
The OPERABILITY of two PORVs, or two RHR suction valves, or an RCS vent opening of at least 2 square inches ensures that the RCS will be protected from pressure transients which could exceed the limits of Appendix G.to 10 CFR Part 50 when one or more of the RCS cold legs are less than or equal to 350*F. Either PORV has adequate. relieving capability to protect the RCS from overpres- . surization when the transient is limited to either: (1) the start of an idle RCP with the secondary water temperature of'the steam generator less than or equal to 50'F above the RCS cold leg temperatures, or (2) the start'of a centrifugal charging pump and its injection into a water solid RCS. These two scenarios are analyzed to determine the resulting overshoots assuming a single PORV actuation with a stroke time of 2.0 seconds from full closed to full open. Figure 3.4-4 is based upon this analysis and represents the maximum allowable PORV variable setpoint such that, for the two overpres-surization transients noted, the resulting pressure will not exceed the cop 4*al-10-e#f to dkka :tive %e"pa-
\ ;;&yearsdEfPY) p m K,c aw M Appendi)x eq , G reactor vessel NDT limits (nW RHR RCS suction isolation valves 8701A and 8702A are interlocked with an "A" train wide range pressure transmitter and valves 8701B and 87028 are inter-locked with a "B" train wide range pressure transmitter. Removing power from i
valves 87018 and 8702A, prevents a single failure from inadvertently isolating both RHR suction relief valves while maintaining RHR isolation capability for both RHR flow paths. ' t 3/4.4.10 STRUCTURAL INTEGRITY i L.- The inservice inspection and testing programs for ASME Code Class 1, 2 and 3 components ensure that the structural integrity and operational readiness ! of these components will be maintained at an acceptable level throughout the life of the plant. These programs are in accordance with Section XI of the ! j ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50.55a(g) except where specific written relief has been granted by the Commission pursuant to 10 CFR 50.55a(g)(6)(1). BYRON - UNITS 1 & 2 B 3/4 4-16
~
- 8 ATTACHMENT C EVALUATION OF SIGNIFICANT HAZARDS CONSIDERATION Commonwealth Edison has evaluated this proposed amendment and determined that it involves no significant hazards consideration. According to 10CFRSO.92(c),
a proposed amendment to an operating license involves no significant safety hazards considerations if operation of the facility in accordance with the proposed amendment would not:
- 1. Involve a significant increase in the probability or consequences of an accident previously evaluated; or
- 2. Create a possibility of a new or different Lind of accident from any accident previously evaluated; or
- 3. Involve a significant reduction in a margin of safety.
The probability of an occurrence or the consequence of an accident, or malfunction of equipment important to safety, as previously evaluated in the FSAR, has not increased because the revised heatup, cooldown, and Low-Temperature Overpressurization (LTOP) setpoint curves are more conservative than the current curves. In addition, these curves have been revised per Regulatory Guide 1.99 Revision 2 and thus meet the NRC standards of methodology for calculating heatup and cooldown curves. By maintaining the revised heatup and cooldown limits during operation, due to the increased conservation of the limits, the probability or consequence of an accident or malfunction of equipment important to safety is not increased. In addition, none of the transients described in the FSAR will result in pressures or temperatures exceeding the limits of the revised heatup and cooldown curves. The Power Operated Relief Valve (PORV) LTOP setpoint enable temperatures remain the same while the Unit-2 LTOP lift setpoints have been revised to become more restrictive than the previous LTOP setpoints. Thus, the PORV cold
'overpressurization protection system provides protection per Technical Specification 3.4.9.3, MODES 4, 5, and 6. Additionally, the PORVs and safety valves provide overpressurization protection during power operation. Finally, the heatup and cooldown curves are not an initiating event for any accident.
Therefore, the probability of an occurrence or the consequence of an accident, or malfunction of equipment important to safety as previously evaluated in the FSAR has not increased because the heatup, cooldown, and L10P setpoint curves are more conservative than the current curves. 4
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ur i The possibility of a new or different kind of accident from any previously l evaluated is not created because operation within the limits of the revised j . heatup, cooldown, and LTOP setpoint curves provide additional conservatism in ' operation compared to the present curves. The curves have been reviewed in accordance with RG 1.99 Revision 2 to ensure that the heatup, cooldown, and LTOP setpoint limits are conservative. The heatup and cooldown curve change does not result in any modification. Operating with new limits does not introduce any new variables. Operating with the revised curves has no effect on the creation of the possibility for an accident or. malfunction of a different type of accident or. malfunction than any previously evaluated in the FSAR. Operation will_be maintained within the pressure and temperature limits
- based on minimum RTND temperatures and reactor vessel materials during heatup and cooldown, thus, ensuring reactor vessel structural integrity.
The proposed Technical Specification change does not involve a significant reduction in the margin of safety since the revised heatup, cooldown, and LTOP setpoint curves are more conservative than the present curves and are based on the RG 1.99 Revision 2 methodology which maintains or increases the margin of safety for operations. Since the revised heatup, cooldown, and LTOP setpoint limits are more conservative than the present curves, they ensure that reactor vessel structural integrity will be maintained based on minimum RTND temperatures and reactor vessel materials. Thus, Commonwealth Edison believes that the proposed amendment involves a no significant hazards consideration.
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r> m ATTACHMENT D ENVIRONMENTAL ASSESSMENT I The proposed Technical Specification amendment involves several changes. Byron Station has evaluated the proposed amendment against the criteria for and identification of licensing and regulatory actions requiring environmental assessment in accordance with 10CFR51.21. It has been determined that the
'b proposed changes meet the criteria for a categorical exclusion as provided for j under 10CFR51.22(c)(9). The following is a discussion of the changes and how L they meet the criteria for categorical exclusion.
! The proposed change to the Technical Specifications revises the heatup, L cooldown, and LTOP setpoint curves in accordance with RG 1.99 Revision 2 methodology. The revised curves are more conservative than the present curves
, and ensure that brittle fracture failure does not occur based on minimum RTND and reactor vessel materials. The proposed change involves no significant I hazards consideration. It does not result in a significant change in the types or a significant increase in the amounts of any effluents that may be released offsite. The proposed change does not result in any increase in individual or cumulative occupational radiation exposure, as the change has no
'effect on releases of radiation or exposure but only provides additional conservation in operation.
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