ML20076B833
| ML20076B833 | |
| Person / Time | |
|---|---|
| Site: | 05000363 |
| Issue date: | 10/04/1982 |
| From: | Baskin K SOUTHERN CALIFORNIA EDISON CO. |
| To: | Kerrigan J Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8210070102 | |
| Download: ML20076B833 (36) | |
Text
{{#Wiki_filter:_ ._ _ _. _ _ _ _ _ . . . . _ _ . . - i Southern California Edison Company ,5CE P. O. B OX 8 00 . 2244 WALNUT GROVE AVENUE , ROSTMEAD. CALIFORNI A 9 9770 K. P. B ASKIN Tata.wong saa v,ano oc wo Director, Office of Nuclear Rea'ctor Regulation Attention: Ms. Janis D. Kerrigan, Acting Branch Chief Licensing Branch No. 3 U. S. Nuclear Regulatory Commission Washington, D.C. 20555 Gentlemen:
Subject:
Docket No. 50-363 San Onofre Nuclear Generating Station Unit 3 During the week of September 13, 1982, meetings were held at San Onofre Nuclear Generating Station (SONGS) with D. Brinkman and D. Hoffman of the NRC to discuss Unit 3 Technical Specifications. Fourteen items required for Unit 3 technical specifications remained open at the conclusion of the meetings with responsibilities for resolution of these items as indicated in Attachment 1. The purpose of this letter is to transmit the information required to resolve those items listed in Attachment 1 as Southern California Edison Company's (SCE) responsibility. This information is summarized as follows: Item 1 Clarification of Action Requirements for items 19, 20, 21 and 22 on Table 3.3-10. Table 3.3-10 Action 22 in Unit 2 Technical Specifications included a circular reference to specification 3.3.3.6 of which it is a part. Table 3.3-10 and associated Action statements have been revised as shown in Attachment 2 to eliminate this circular reference. Item 3 Verify Applicability of Table 4.4-5 and provide additional wording for the bases to explain " Lead Factor". Table 4.4-5 has been revised to be applicable to Unit 3 and is included as Attachment 3. The bases 3/4.4.8 currently contains an adequate explanation of the purpose of the " Lead Factor". No additional changas to the bases are required. - 0 0l f #ili "J 8 8
- Jd" 88 5 ~ l A PDR -
i 4 , Ms. Janis D. Kerrigan October 4,,1982 ) Item 4 - Provide heat up and cool down curves (Figures 3.4-2 and 3.4-3) applicable to Unit 3. - Figures 3.4-2 and 3.4-3 have been. revised for Unit 3 and are included in Attachment 4. Additionally, corresponding changes to pages 3/4 4-3, 4-27, 4-32, 4-33, to Bases pages B 3/4 4-1, 4-6, 4-7, 4-9, and to Bases Table B 3/4, 4-1 are included in Attachment 4 i reflecting differences from Unit 2. Item 5 Resolution of correct pressure to be used in 4.5.1.e.l. ; The correct pressure to be used in 4.5.1.e.1 is 715 psia which is the pressure above which the safety injection tanks (SIT) are required to be operable. Although the set point for automatic opening of the SIT isolation valves is 515 psia, no credit is.taken in the accident analysis for the automatic opening of these valves because they are locked opened when RCS pressure is greater than 715 psia (by locking open the breaker) during normal operation, the initial state in the accident sequences analysed. Item 6 Pump performance data for Sections 4.5.2.f and g. Section 4.5.2.f and 4.5.2.g have been revised with Emergency Core Cooling System pump performance data applicable to Unit 3 and are included as Attachment 5. Item 7 Provide justification in the Refueling Machine bases 83/4.9.6 for the exception in 3.9.6 for four finger CEAs. Revised bases B3/4.9.6 addressing the exception fy not using the refueling machine to move the four finger CEAs is included as Attachment 6. Item 8 Revise bases B3/4.10.1 clarifying the reason for entry into MODE 3 during performance of CEA worth measurement tests. i A paragraph is added to bases B3/4.10.1 as shown in Attachment 7 l which explains the reason for entry into MODE 3 during CEA worth measurement tests. Item 11 Revise bases to explain that tank volumes required to be maintained by technical specification are usable volumes. The bases for technical specifications 3/4.1.2 (Boration Systems), 3/4.5.4 (Refueling Water Storage Tank) and 3/4.7.1.3 (Condensate Storage Tanks) have been revised as shown in Attachm6nt 8 to explain the usable volume considerations in the technical specifications. Bases 3/4.6.2.2 (Iodine Removal System) was not revised because the contained volume and usable volume of the tank are the same because the discharge line is at the bottom of the tank and there are no internal structures to reduce the usable volume. ;
..- . .--.,..m ~ - .
4 Ms. Janis D. Kerrigan October 4, 1982 Item 12 Identify location of seismic monitoring instrumentation. Notes as shown in Attachment 9 have been added to Tables 3.3-7 and 4.3-4 to indicate the location of common seismic instrumentation. The information provided in this letter should complete SCE's open items enabling the issuance of final draft Unit 3 Technical Specifications. Should you have any questions regarding the informati.n provided in this letter, please call me. Very truly yours,
/)+1.V -fo s. kPB Enclosure cc:
Harry D. Brinkman, Rood,NRC NRC(to(to bebe opened opened by addresse by addressee only.only.) ) l l l l
,. . s _ . . , . . , . . . . _ , . . . - - - .__ ...-_y.,_. .
y 4 O e S e ATTACHMENT 1
- .x = - - .: -.: - . .
4 SAN ON0FRE UNIT 3 TECHNICAL SPECIFICATIONS ,' ction No. Action Responsibility Completion Date 1- Clarify Action Requirements SCE October 4, 1982 For Items 19, 20, 21, 22
- p. 3/4 3-53 Table 3.3-10 2 Determine if Additional NRC October 1, 1982 wording in 4.4.4.3 c. 4 is accountable p. 3/4 4-11
- 3. Verify Lead Factor and Previous SCE October 4, 1982 Additional Wording for Bases Table 4.4-5 p. 3/44-28
- 4. Previous Figures 3.4-2 and 3.4-3 SCE October 4, 1982 (on graph paper) pps 3/4 4-29 and 3/4 4-30
- 5. Resolve Safety Analysis Number SCE October 4,1982 for 4.5 1.e.1 (715 or 515)
- p. 3/4 5-2
- 6. Provide Pump Performance Draft SCE October 4, 1982-for 4.5.2.f and 4.5.2.g
- p. 3/4 5-5 and 3/4 5-6
- 7. Provide Wording for Bases SCE October 4,1982 Justifying the Exceptions for the Four Finger CEAs for Specification 3.9.6 p. 3/4 9-6
- 8. Provide Wording for Bases SCE October 4,1982 Clarifying Reason for Entering MODE 3 When Performing Special Test Exception 3/4 10.1
- p. 3/4 10-1
- 9. Provide Pages 3/4 11-11 Through NRC October 1, 1982
. 3/4 11-19 as they were :
Inadvertently Deleted
E 1 Action No. , Action Responsibility Completion Date
- 10. Determine Wnat the ** was NRC
- October 1, 1982 Initially Intended For in -
Technical Specification 3/412.2
. p. 3/4 12-11
- 11. Provide Bases for Usable Tank SCE October 4, 1982 '
Volumes and How Determined for a Number of Technical Specifications in the Boration Systems, Plant Systems, ECCS Systems and Containment Systems in the Bases
- 12. Identify Location of Seismic SCE October 4, 1982 Monitoring Information and Provide Proposed Technical Specification for Technical Specification 3.3.3.3 pps 3/4 3-42, 3.-43 and 3-44
- 13. Monitor Technical Specifications NRC October 1, 1982 for Seismic Monitoring, Meteorological Monitoring and the Control Room Emergency Air Cleanup System to Reflect that these are Shared Systems Between Units 283
- Technical Specifications 3.3.3.3, and 5.7.5
- 14. Add Missing Prompt Notification NRC October 1, 1982 Items J and K to Technical Specification 6.9.1.12 p. 6-20 and Revise Thirty Day Written Reports Technical Specification t 6.9.1.13 p. 6-21 for Possible Missing Items.
I
y~ ~ e, i n e ) d O e O ATTACHMENT 2 I
l - ~. 3 g TABLE 3.3-10 g - ' ACCIDENT DONITORING INSTRUMENTATION (CONTINUE 0) 7 . REQUIRED MINIIRSI O 5 l NUPRER OF CHAISIELS
-4 INSTRtK NT ^
CHANNELS OPERABLE ACTION 4 U %
- 17. Containment Water Level - Wfde Range 2 1 20, 21 .
- 18. Core Exit Thermocouples 7/ core 4/ core 20, 21 quadrant quadrant u 7.0, 2 l A
, 19. Containment Area Radiation - High Range 2 1 -
M
- 20. Main Steam Line Area Radiation 1/ steam line N.A. M 2.0 ,
o 21. Condenser Evacuation System Radiation 1 N.A. JP- 30 g , Monitor - Wide Range ,
- 22. Purge / Vent Stack Radiation Monitor - 2 5 MM' 22, W w Wide Range * -
- 23. Cold Leg HPSI Flow 1/ cold leg N.A. ,
20
^
- 24. Hot Leg HPSI Flow ,
1/ hot leg N.A. 20 l l l ISTES: l
*The two required channels are the Unit 2 monitor and the Unit 3 monitor.
- 9 0g6 On0
. r
( TABLE 3.5-10 (Continued) ,' ACTION STATEMENTS _ , ACTIONh0- With the neber of OPERABLE accident monitoring channels less than the Required Number of Channels, either testore the inoperable channel to OPERABLE status within 7 drys, or be in HDT SHUTDOWN within the next 12 hours. ACTION 21 - With the mmber of OPERABLE accident monitoring channels less , than the Minimum Channels OPERABLE requirement, either restore the inoperable channel (s) to OPERABLE status within 48 hours or be in at least HDT SHUTDOWN within the next 12 hours.
- Wi t of OP L s i ON-ro C n , I w h 1 i on .3 .
7b Ac.c..Jee f Mon. lor.n3 num ber ef ACTION 4fr- With the number of OPERABLEAChanr.els less than^ required 3y t h
-- 2 nnf y ---
' - - - -- ^
eitner restore the c hanne.h 3 inoperable Channel (s) to OPERABLE status within 72 hours, or:
- 1) Initiate the preplanned alternate method of monitoring the appropriate parameter (s), and
- 2) Prepare and submit a Special Report to the Commission l
pursuant to Specification 6.9.2 within 14 days following the event outlining the action taken, the cause of the _- inoperability and the plans and schedule for restoring the system to OPERABLE status. s I S e e \\ SAN ONOFRE-UNITg 3/4 3-53a
7 _. 3 . s b 9 e 9 9 ATTACHMENT 3 ( ) )' 1 - -
m o - e
.l i !
i.- TABLE 4.4-5 l REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM - WITHDRAWAL SCHEDULE '
.. !I
$ CAPSULE VESSEL LEAD l g NUMBER LOCATION FACTOR WITHDRAWAL TIME i ,
2 i , T 1 83* .4-17 1.T Standby i c , 3 2 97* M l.f 6.6 l'EFPY M 3 104* . hts" 1.f 1T.2.ts:5 EFPY 4 284* 4,4!P 3,S 24 EFPY i 5 263* 4-99" J. F Standby
- j 6 ,
277* M 1,5 Staney * - 1 I
- e. i M
5
' I fP .
=
,9
O e 5 1 4 3 I e e ) ATTACHMENT 4 e
..-.-.._.._~n.-
REACTOR COOLANT Sv5 TEM NOT SHUTDOWN ,
~
LIMITING CONDITION FOR OPERATION 3.4.1.3 a. At least two of the loop (s)/ train (s) listed below shall be OPERABLE:
- 1. Reactor Coolant Loop 1 and its associated steam generator dnd at least one assoCisted Reactor Coolant pump,**
- 2. Reactor Coolant Loop 2 and its associated steam generator and at least one associated Reacter Ccolant pump,**
- 3. Shutdown Cooling Train A,
- 4. Shutdown Cooling Train B.
- b. At ' Sast one of the above Reactor Coolant loops and/or shutdown cooling trains shall be in operation."
APPLICABILITY: MODE 4 ACTION: ( a. With less than the above required Reactor Coolant loops and/or shutdown cooling trains OPERABL E, immediately initiate correc-tive action to return the required loops / trains to OPERABLE status as soon as possible; if the remaining OPERABLE loop is a shutdown cooling train, be in COLD SHUTDOWN within 24 hours. ;
~
- b. With no Reactor Coolant loop or shutdown cooling train in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immedi-ately initiate corrective action to return the requ'. red coolant loop / train to operation.
=
All Reactor Coolant pumps and shutdown cooling pumps may be de-energized for up to I hour provided (1)' no operations are permitted that would cause dilution of the Reactor Coolant System boron concentration, and (2) cort outlet temperature is maintained at least 10*F below saturatgn }yerature. l A Reactor Coolant pump shall not be started with one or more the Reactor Coolant System cold leg temperatures less than or equal to F unless
- 1) the pressurizer water volume is less than 900 cubic feet or 2) the secondary water temperature of cach steam generator is less than 100*F above each of the Reactor Coolant Systen cold leg temperatures.
k 3/4 4-3 SANONOFRE-UNIT /*3 p
~- __ __ _ _ ._ _ _ __ _ _
_ _3,_ , . . _
* . . . . . . . . . . . . _ . . . j l
REACTOR COOLANT SYSTEM 3/4.4.8 PRESSURE / TEMPERATURE LIMITS ." REACTOR COOLANT SYSTEM - l.'IMITING CONDITION FOR OPERATION 3.4.8.1 The Reactor Coolant System (except the pressurizer) temperature and pressure shall be limited in accordance with the limit lines shown on Figure 3.4-2 and Figure 3.4-3 r'uring heatup, cooldown, criticality, and inservice leak and hydrostatic testing with:
^~
~4. A senirs5~nnprne op Mho aus sent HanA NRose edme h came s,5M g
tan Tseu near. A maremmai norrw er se*> > Amay aMr med Mrsee wrfs #C 48'*8 484n *fW8'tMRAfid48 6483rr5R, 74#A# #ff 8/r gegy ggg gggg }NlP 'P. A swenava Mf8Fru# em 6egr /W Aary sur assie4.psbee wryg g gese ws,waggggg 73mpt
& 4 #1 sai cmM GP 18'Pr An# Antr ANS #4m( AWfded a#t7N dCao adE 7)>ampbrat(d uss ymna logy'. A ngessneanit sneseead** #F 9eYlp m amos neex Na mMTN $C. H gg, .rewinguagtteed g4gytTwyt "rmona e die '8r & " **
- N ee seeaprpa/4.#v eue asmat retop warse# ceam#3s Taueu 2009'. salprrax vmov 2a098 ws,1m>=werrissa ;
- c. A maximum tamperature$ of 1msemusumissingsmemmi 10*F in any one hour period during inservice hydrostatic and leak testing operations above the heeta , and cooldown limit omrves. -
APPLICABILITY: At all times. ( ACTION: With any of the above limits exceeded, restore the temperature and/or pressure to within the limit within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the Reactor Coolant System; determine th41 the Reactor Coolant System remains acceptable for continued operations or be in at least HOT , STANDBY within the next 6 hours and reduce the RCS T and pressure to less than200*Fand500 psia,respectively,withinthefoiT8 wing 30 hours. SURVEILLANCE REQUIREMENTS 4.4.8.1.1 The Reactor Coolant System temperature and pressure shall be determined to be within the limits at least once per 30 minutes during system heatup, cooldown, and inservice leak and hydrostatic testing operations. 4.4.8.1.2 The reactor vessel material irradiation surveillance specimens shall be removed and examined, to determine changes in material properties, at the intervals required by 10 CFR 50 Appendix H in accordance with:the schedule in Table 4.4-5. The results of these examinations shall be used to update Figures 3.4-2 and 3.4-3. Recalculate the Adjusted Reference Temperature based on the greater of the following: g ,6Se&' I a. The actual shift in reference temperature for platef fr44944 as determined by impact testing, or
- b. 2-2 2.- 2M B k-The predicted shift in reference temperature for weld seams M.03A
.>r. 2-2.o3c Ns determined by Regulatory Guide 1.99, " Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials."
$AN ONOFRE-UNIT / 3 3/4 4-27
': c
-r - '
e
. .L , [ [. : *i~ .
[h:.) a if jij: [jf h fi h'6 7:t [' [; *;f h ' J f- .J .
.i.
T
'.II lij .h
'f, -{ t :'-"
~- ~-
'k
.,I j
i
- {
ll'
"*T l
p-. 3[3. 7
. Jj *** ti if
'j!:
l . - -1 t.ijp.- i .
.L... :. l 3
n: .u: - ... .
-!- @...j : .
.n. :.:.... :- -
4.
. . .. .:t.s.:. :. .t. n. ..._
- :. . t.h .: j. . ;: :.
-. 4 11 12 h ..a i.
M: :t : 4J . : a . ..:3.:. g.tr !.t{.2 .. -.. I-nn: ,I ":)::. . '
. . .g.. .
mym n-y -~ qm u
.2t!.
n- &. p i ~ -
- p. .a .
3 n u :: -it; ATm
+
- j. :hli;. ,
.a.41.!y
.n= .I a. .y- . .:. ..if I::ji. jih .- .~ aj
. M .!!ni50*M i:NOc7-4 y!! [nl.
- 5.h ni }l' 'in _
'i!i l' - -
Ifi m*nt 0% !
.i 0:i :iiidr u.
.1 _4 :i: . . .
Ij fifi .- . . E~
-. -.! _.1
- iii @i I: I iff Y[ ! t !It .
l ili- ~l Jia 7 jjj;
. . v... .: . T-
$4 n:1.
?j. 2: i J :. . ;
.~ ._, y li:
[il tii: l,.>:.]2F.
.t c . : . : :t ij.- 3:i 'i,[ Ct J -} . .\ <
.n ._' -! ,
. 4. ._ .u. m. - - ~ ~ --
- .7
~,-a.. i
.t{. :u;
~
- . .: ;;l.p
- n n -
n . .; f i
- p. n: ? .
m scum . t . I
.,n. F...
- n. e p:
a
.' o: j... i ,i ! p
+ ..
- i. -
.. r y
t : w..y . ,In.io. . .. : l :1 ,n; ,, ,. : ,t-T{ $"lh 3.. ,, t i
}i.:
a t..
- .- . . ,p .
- 3. . . . 1 . 2;p : -
- t. i.,p. -
- q. __' :
- 4. r,. n, i. n. . :.a.p. .j -. : :n
..n
;n .:: .d J'. j:j!
4 - ii- - [ . :! ll 1 l i'1 iF}
;;p
;; l:
.l.- . -
;; e E:! ..:1 " . . L. .O.NE'ST g. l. n. . ; j 4
.g .t gjj ;i. .t :
p. o
,,o "
g m 1
.a n
a- ' 7.. % .:
.I;2 r. .i.
2 4 ,p . .i n :' . : g. .. j
.%gl . iifu.T OP.-20MF; p, y l
g q .
.:li '.
di: .:: t.i. :. ilp. i..:.!
.j. . u.
- i .j .
. .. .. ,1.. '.
- p
!. : p.
n.. M ; il h R Qajl id7 [v "- ti 8. ilji
.= ._1. l: ....
~ ~
liH ij P~
- ~
v
*:. dsFl :j :
- .&; - l-E i
~
y ._ j 2:-
; . .: t
- r n1
- l' i r j n:
- .y- ' j::..
1:- M c - " - - 4r .: nt t: !!:: .it: :. i W . iF. . :: 1._ .i 4fl.i,i. :.. g j!.!. p'- .g[i y,i. !!" .!/,,jjji! .: . . . 4 ' . i.F!F. m j;,; . i.j.ii. i
- !LE. c g . .. .
. ,j
.' .ji h
"i' iji[jjj. 4 j!!: gl lq [:.ji@![
gjj g Hip ih. pg i- i gj i
~
r p: i. :M ::: - ' ;- .t i
.! t!!. ;11 r a; .
y ~~ - i c.qp:- ... c
- -. .n . .: . . . . . 4*:1 .:.
.::- . ' . "r : -
g .. .
. n. t .t i.:. :ttr.E. i -
. ' n .[_t . ,. t . .:.!".
!Ni-- ' l .. 1 'I }i .
W ! di i
- l}I UN lii: tl. :N! ii !! id- .!!
i 1 li N 20 0, 0- ; i .jp :: .:- -- pp; 1.. .: .!!; ; i p- n. 1 !ji .
..in a .;.p:.e. . :t..r
- :. ;. - . .a n .: . a .- .a s -:dAua !-
- :3, . . .t.: .:2:2 2 u..
g i ~: -
..: n -4 :. n: -
t: r - - c - t : - til! :
- tr. . 1-- n
- ". ~. ;::.: "1 .
e ti r 7) h
- h. ! ih. tjp ;l(( i.!! !: d . I j' h':. p p.
g ..
} g
,~
e.
. p u. ...-
.:asu: :.
I-
.1~. .. .:
l l*.: - . . i
- . .a,e M ' '
I I Ni .'l p, jj:i' b > j ;. . ;: ! ..l .Ih...' .G : p :i
!l ; . .k--
W 1600-l.:a1.a i . . 'C. -i- i;.': C =.- r.
-! in: :.:. p: .!.
g :
- 1. 2
- . n b. -
1-i, . -
; . :. . .u. a a.i ;:.. ._l-:
1 u.. ..
- t A ! : ,n.. i' .ii _.:.
d , U llg ~il 1:- b: l- ! i..ih i g i]a
. :. ._ L - "
~
'u t* . . . . . _ . . . . ,. .
.!- -- :p ..l *:.
a :; [.'e .n i
.- : ;t . . I e
O -- 1-
- - l -- :
f u:. :t 4j o 1 [- 4
,i :: h. -
4
. .e
- pl . .: . le n:
W I
- r. t
-y; .. .
'!) -
A l'
-h- ! i: ! i L io ! , ::
'. .. il! ! ,
a: p 1200- n.. . .. . .
,.i.- . . . . .
- r. ..4-; -i t : - -. 2. ?
% . ja r; C- .: .:..33 a u .j " : : .n : .t I1 :
. J _. -g2"- .. .
J.u. t, .cg .t.._ L' :. r - - - -
.ac -
i-- 1 ** j- i.. 3y i ..: :2: e- a :
- 2. .., . .t - i O , . . ..
,. l ; a .-.i n "r -1:h: ;; - "-
- r 1 aIf+f: 2 .: tit: ..."t t. i. !:t t:- : 1: '
g ! .! 5 e o m i
-- F' -
i ! - . Ii .
.!!Ri :d ! rii- u li :
. fi ii:
11 1 11 r
;!! :! !F j i! !i w .
2 w!, '
- l. :... :.
k.it.t. :t. :
- t. .
1 ll:
.. d. . .. {:. :
- j. . -[1] .: j.- .:..' j .' - ., 4 y, _
F# -
! t t t o : tEATpP. ....iliidi li
'. .. ifI - -
ti !!Ji! I li 111 "I -
.209E' a !'
3 h; [:'0 *:F/MR10*FW kl kn$:[j
; . il Wii ? i: !i! .
;p t 4 1j u 'N .}ij '. ! 'y I t.
-{ .. .:.! I-j g :: I t lU L . ,
ey I . sya- +- .;.,j.
"- in ,
- . I:,lg:_
i
- I ;, -
1 _ _I.
- " ~+ m s.: M8# r- :i:: tn i
7{. ut .n n - L;.,j] ..q :. jp.
.i F.
",.3 .
.3....
l 1 O E -t-fi *ili &'$ 'jl
' b. .L N!! }E. --- iib. !!I. i: !
.ii! 5.di! t .:
='-
. l . . i-
, .. r- y -: y- - -
. ... -- t .
.a . . .
..! i ! ! $.!} -
t'
,,_,. :i ! i !!' f*ii if if N5 ! fi I ' '
- . . :,: r :j i..
; in : j; ". . . j!
. ; . t' ": i..
- .: :t :!
t . II. . l ..: 3"j;' ::t: 1. - j:{n jj :.{N"::
- i
'g.. ,
l
.. . . . T p s;:-
I '~{.. ," "'
.P n. . **i':
*- TI' "Y.
tr '
'!".~.
".j"~
] . U. :... :
.3 .,r: ~~I'. "
t.t.4 i. !) : :; *t.k. ... . tj t. p. 1 l
,,. jo~ -- -
5)- J U.
- d. . . -+
ut '
- l!!i- 16 0.-:r dii --
200.ct;c . i,iii ;20.d.g! t... a ~i 13 6
,n.
0 : . ..
...r iliii fi- ih- dit r : itn ! : 11' 1r l-l.
- 8 ! "t
!!F 4
,. . ..: ,. to. . .i .i i. .t.... .. t.
t . gyg {.. a+ _ __ _ _.2.. = :u.l gs .gr$MPERATQR EHa@)N H)M 4+, - l ii : i n- e :nt m te :n " n .an Y:iI: t h- . . ., HEAT L.m m,5 ~PHE4,ep./ g mRATO. RE LIMIT (J1..0p-u:r pR$ict.m p-TYR5
.s .. . .
- n
, . . . . . - . . .l- .
~~
m g .c
- ~
au ..u n:. - n+ r aa. .dun .- iat.
.p: .4 f..
} ~~
l : 1 :- . l.: ~
- ! n,. d :* nr " .r .-
atf ti. r i r l:; :nt 2 1 :dt: ..h 2 : n. .
.a
.;:j 3p: i en : :~ .n .
l .. . l-at h.. t a .:t} :!. g.p{:;'" ;*l;:
- . :n. : al.
- t : -h- 4 :3 3 L
- .
- . i.cr. i
*y g :g . . nr t
.. . . . t- ... - . . . . ... -
- 1. ..} .
. . . . . .. ..t". r, p. : :i
.Jt "I .
H *i.t 0. 8 l.J.,l. a. ."1. : ... n,p+f.-
. 2 ...: : ;
..i :- tt !. .*-J' a. j .-
;'.-[ :- l .
SAN ONOFRE-UNIT 3 3/4 4-29
..j 4 ....; g.
.h:
.:. . ja
. .. ; ; .. .a. p- 3,:.. : .p7t y I.-
- = ~, i !
.:. w : '- -
;j.
l.
;_ ]; a. =.}"
.: . . p,t ... y. : .3n.. . , - @' j , ;'..h. r.: ---..:..
'. j
- a. . a. .
. . n.,;; . ,
- , .j ", ;..n.p:-
.e i.
- u. . . ..
E nii E :i F : Il[fi!! [If; til:st: , . I2 !!!- _j r - S- i3 ?- E! Ei ali' 'i- .iM".- t-7
- e -~
~:: " n. ,2 :4 -".
. gag!
go i a g. g .. ... &" t.*t r.
. .n
. ...: n..
- ,.3...... :. . c. .
tt :
- n. -
-t - l :..
+2g ;J 5:TW r s: r
,=-- - i. -
y:'7- . a n;
.n"i -; ti;e.;!.m::.t
;;8; + -
m ,. a: .
,t
-- r -
.. .-.---- :.- .- m, =
30: .. ' l. . t _. .: :. ._;;L1 . .-
. 0. t3.t. :.
.d.. :. .. .. . '.l.3 11
- ... a. +.:l :t.:
u.: : ..
. l= :E i: I :ii :iii li:l !! 4 Ili: :si
. M.: ' . i lii; iii!
m !i.. .iiii: 5
" .-? 0..;.y i i!! : :i sii
..; Mi 2 :: ; 7g n j;p pv
[. ... .. ._
. 4_7 . . .yp.ch ;g.y n;; 4. .: g; ;;: . ' , j3. .g.
q, . .
-,, ijn x ill"~ u,:p e ii :
+p;..;ip i- m ti . -i fin
- .+ :12 .
z . l .. a =.: a=;:. .:.: au -
- z. : . -: : ;. i .: ~
.:- y w(.
. !!!! .i ' in it!i id5:!-.. ! 1 i t y!!!
- ,h,_CDDLD0wp:-
Q .Ir :;t ': h PM > Y w.: if:
.i-~~ . j; !. ! .!h- i i :8 '-
I'l
.! . _ __1 2_ - ] =. g.;
L,.
.2 at a __
y .t.hi n:g. ;.EPYHR @'E . . - -
' l' !! ifH 11 0 :t: !!F ,
!!i !, ,.
- ih: *Ii' -
il . _jPf HR ( 1 tt'1 *f "I
- p t, ; -
i~ it * - a j':-l .j I . ' } a: .;l..
- l ! :it: ji
-l
': }- :n l :-
..t. .!.g,; . .' :. .l:.
... .... .. ... .. t
.M.
I.. t_::
; ,l.. !. :, .!.1 .4....:h '
- 1. : - - ~1 :.,,,_ dt.. l. 'j . ;r.{:. : . ;l. t. ,l :
. :.-.t o ;. 4. p"- ;p:. en i 3 ;:. t 1 : . ... . .... ~
u) ;. p....:: i: .3-4,1! l ! ;h, t . a. !. .! i :,' 1. ii" . m a g, . i
- i. .
i > p::! :- - m -
.;y
. I.
i .a e at r.',
- : r 4t .r.
~
- , ; i. ,:
,i .
v a tii
,, j- 8 -
g ji; WES'I' - ib..W
- e +; :
gi :ii . I i i ja : U : , f g3 g
; f
#F % :"
P. .u,-N...O g I j;T .. U ii:i:ih,.
+
..o... ..I 7 t[
- :::- : : rs :r ::. .
' j i- eg;
. ; .u- :
-+- - -
W ;F ' E l_ ...' l :l:". - d:
,,4,_ ', n e-Ei1.HL2O2*.Ein
$ l
;I :- ': ": ;- .4 ' ..,
n;f' il. i
.:.] j'6: I j. ::j4 ".i l'- l . :1 1
- e i -
!:h n ;... L) . ; :
- r i ~
n - .- :
.- ; i Il ..v. ..
- p. gg ..,. p. . a. ! ..! 3.
.9... "4 e
...l. g
- !! ' ::7.;. ! g -- l
- 1. . .. . .: 1 1:
m' t---
- r e' .-
..h .. ' . .tj-w :r :- rr; tr!
E ih. - W2000-iii - t:
~! ii ufi ihi-W ii:
F hi: s. f! :.: - 0 !Yi! : 11
!.1.
,4 h...r.
.i ' ' II.
N y. I .i,; -'
, . . " ., . " . 3_; " ..l', -
- 4 r". l,. _!:
E, "."
- l:-
u m ..: n: . . . . j : :::: a . g: '_:.2
. t
~
i-
+}; . _ ,[' .-7_ : .
C -l F:-
": rtt. -
t tt:: ta: n+!t~-
? _
- " t1
- . i. r ._ -
D iii 31h' j iii
?! 'L * ;I jii; il ii %b j i.i'
-8 ' i 3 M4" -jit !!D ! i
- '$. n .: r-
;L. ._ : _ ._ : . :; : t
,_ : 12: . r " " - - - -
v g)
- u.1 _ :':;-- . ":.
.t
- ; :. :- .- r-in
+--
ar
- n:
- n
-n r
- . i .
". 1' : :.t -..
- r :::
.t --t - : r:-
- [.h:
u -
- c!:n W i ; i: I@ l b S I
.IE }.l IU !Ni !- [ M*i . 'EI I fil l !!
E + . . . . 3 1 Q I - id ..i ii: }
.Ni i :-
- g. .
1..1 ' ;1i 'i ir tr .I 3 : il i @ : I jb.
-: ,) l- , i' ! il: '!3
- n. d:i. 4
- Fi i-
.!ff+h 9 3 lil i '
} }li .i- hb q;. 9h
. . p. .:.
- a. i . 1.
Q - :
' 7
; I: .a p. i .t* .- :- tt-c.
ni m -) u!
- . :- : : E':
i h: . : ,. . :t - - e- W F204 ! 4
< F - :
..;;p .
. . t.
- i. p. .
- n. !!. ": .r:.li . :. b.:. .-
. I q. h..
M yI: -i= ;H : : thq-Ih; Ce ua. 2-+ J m; 2 4 --.i - - - . .
" . " , i .S.'1 .: p.:
sv g () : , I. i.
.].: n.t-.
.a
+
;s.::.
.; l.'.r r
14. 7 i. fo :. !!!: ..1 h C: 01 IE. , '.n n LE .i i f O .._... . . . . _ . . _ . - . . _.. .. . : _. .__
$. .t. ..
1 13 t a 4 2: ! ....! i:1 ~ i! !.! . '. s. - ' :.!j l. l ~ lh. .{ i.
; 'i I.II. L. !.ij Lj'.i. fi Ji.:
y e- _ g . ap
. i. i . : :
- rI
.i .g i l 'u " l :;l o ! .
.COOLDOWN:-neb " II' -
2 L ..a lel . L. - i . lOPHR< ll0*F ili) :i li. . Jil i ii !H ii! :'- ih j !"!i! " .s :i !E ! .! -.E l
--1 rl.
. g ..
1" .T- ! .,p . !
. . 3. y.1 I .4 m :.ji x3 i
- [. :.n ....: m : 1 ::: -{
q- ..!.. t ._.4... p- n s ,..m. g g:a ma m.'a .: ;aa ~a
, :- p. ,: n: :m :pl.;a - ti - n . - jn :, - :.l W 400-1 1 3
- n u h. . .- -
b:l '
. r: ;. - -
- : n :. .
- y. . .
n:: :qf.i..: c. d. :p- p-:- : :r.n
% cJ.- '
.22
.q .;
..n i : ~tr-.: : n. + - - -
.. h -
i m; u c - m.:n.
- ,- :i.
,- r, m :-. :r un : ;,
- rp :. . ,or,. t" + : r : n -
I: i : i. : b . s - i -.r u e tp: . r at ni.a . e- t. l i. jii dj.
- g
-j - -
iii : 1: ii .! i iii i. 4 '!: l.i[ l!y
- h].:
i- : i
.. .j . . ,
y
.). .
~. . . . _ .
g (- g...
. j ...
7.4;;. ., 4
- n. ;t; g.i.4:,, .
-:4[
em - ...
- .7 .at i-
.u. .: : i .a
- . 2 -
"q , .. :r i . ..
. .e$ .
.1
. :. _.o -
an. : :":p: 4;.p 2
~-
i : -
"n":15,0 200 2E 2" p- ng : :
r r:: j:n :r.00L=u/~1 :
. _[':2. :- ;;;:- t t . .
1 . : t .:h -
. ... : a. .- t.r; 4. . - : r nr :
+" 'I :p: .::: p; i tt -
. g a.i 4"
"r M[
i 1 . . , : . o ' 3- --b - --- RCOiOEM PER AlI'T@?!E M*F.t
-ENO@tidkT&
t' l : t .M :*! 1:: !!r ! l ? -
- a.v a
..L ...,; . :.. . Opt 10 0 nn T.nPERg(RE LIM 1 Iwpty m =. :.W.N ntgu. -5YRy jl
_ u... -
%. . ...e .M .
ql-. "! '
. : U 2
1l: rlj: : :j J: [' [ ':1 u.,- i .;p ~ !j i. ' j n' g: ]: . :t
- ~
I -
'l: 4 P . . .:. f. ! .' +1 : J
? " 1' t.t- : ,: ':. .. .
l . . .. ii 5 ! li i!!!
#3 hb i l" 4.tU 1 P.
- i i fji : ~+[ ii '* f . 1 i'1 ! !'
- . ;. j ! r .. .h"T*:&h...: 3'f:3-g
. : p. '. !. :
- n. it. :::. .
.-3, p. mih;gl.
1
. l.4.3 $ :
.] . t. j {. {. l.i.ft. !. _.
- I
. h!...
SAN ONOFRE -UNIT 3 3/4 4-30
N REACTOR COOLANT SYSTEM ( OVERPRESSURE PROTECTION SYSTEMS 7.-65~*1: RCS TEMPERATURE I M ' \*
~
LIMITING CONDITION FOR OPERATION 3.4.'d.3.1 At least one of the following overpressure protection systems shall be OPERA 8LE:
- a. The Shutdown Cooling System Relief Valve (PSV9349) with:
- 1) A lift setting of 406 1 10 psig" .and
- 2) Relief Valve isolation valves 337,M339,h9377and 3 $ 9378 open, or, 3 3 '
- b. The Reactor Coolant System depressurized with an RCS vent of greater
, than or equal to 5.6 square inches.
APPLICABILITY: MODE 4 when the temperature of any one RCS cold leg is less than or equal to 486'F; MODE 5; MODE 6 with the reactor vessel head on. l 2BS' ' ACTION:
- a. With the SDCS Relief Valve inoperable, reduce T,,, to less' than 200*F, depressurize and vent the RCS through a greater than or equal to 5.6 square inch vent within the next 8 hours.
(
- b. With one or 'oth SDCS Relief Valve isolation valveg in a single SDCS Relief Valve isolation valve pair (valve pair 3 3 pfV9339 or valve pair J fiV9377 andYHV9378)open closed, %9337 and the closed valve (s) within 7 days or reduce T,y, to less than 200*F. depres-surize and vent the RCS through a greater than or equal to 5.6 :
inch vent within the next 8 hours.
- c. In the event either the SDCS Relief Valve or an RCS vent is used to sitigate an RCS pressure transient, a Special Report shall be prepared and submitted to the Commission pursuant to Specificution 6.9.2 within 30 days. The report shall describe the circumstances initi-ating the transient, the effect of the SDCS Relief Valve or RCS vent on the transient and atty corrective action necessary to prevent recurrence.
- d. The provisions of Spccification 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS _ 4.4.8.3.1.1 The SDCS Relief Valve shall be demonstrated OPERA 8LE byr
- a. Verifying at least once per 72 hours when the SOCS Relief Valve is being used for overpressure protection that SOCS Relief Valve isolation valves 9337, 9339, NV9377 and 9378 are open. \
( "For valve temperatures less than or equal to 130*F. SAN ON0FRE-UNIT 3/4 4-32
... . ... . .... . . . _ . . . . . . . . - - - . - . . . . ... .. .. - _ ;;;; U ~i' M
_ - . . - . . . . - _ . - - - - ----= i REACTOR COOLANT SYSTEM ( . i i OVERPRESSURE PROTECTION SYSTEMS
- l 285'F ~ ; 8 RCS TEMPERATURE > W . ' '
LIMITING CONDITION FOR OPERATION 3.4.8.3.2 At least one of the following overpressure protection systems shall be OPERA 8LE:
- a. The Shutdown Cooling System Relief Valve (PSV9349)'with:
- 1) A lift setting of 406 i 10 psig*, and
- 2) Relief Yalve isolation valves h 9337, M 339, V9377 and p9378open,or, 3 3
- b. A sinTous of one pressurizer code safety valv.e with a lift setting of 2500 psia + 1%**.
~
7 225~ APPLICABILITY: MODE 4 with RCS temperature above 259*F. ACTION:
- a. With no safety or relief valve OPERA 8LE, be in COLD SHUTDOWN and vent the RCS through a greater than or equal to 5.6 square inch vent within the
- next 8 hours.
- b. In the event the SDCS Relief Valve or an RCS vent is used to mitigate an
( RCS pressure transient, a Special Report shall be prepared and submitted j to the Commission pursuant to Specification 6.9.2 within 30 days. The l report shall describe the circumstances initiating the transient, the
- effect of the SDCS Relief Valve code safety valve or RCS vent on the
! transient and any corrective action necessary to prevent recurrence. j SURVEILLANCE REQUIREMENTS 4.4.8.3.2.1 The SDCS Relief Valve shall be demonstrated OPERABLE by: a. Verifying at least nce pe 72 hour thatthejoCSReliefValve isolation valves 9337,3 HV9339,3 HV9377and/HV9378areopenwhen l the SDCS Relief Ive is b ing use for overpressure protection. L
- b. Verifying relief valve setpoint at least once per 30 months when tested pursuant to Specification 4.0.5.
4.4.8.3.2.2 The pressurizer code safety valve has no additional surveillance requirements other than those required by Specification 4.0.5. I 4.4.8.3.2.3 The RCS vent shall be verified to be open at least once per 12 hours when the vent is being used for overpressure protection, except when the vent pathway is provided with a valve which is locked, sealed, or otherwise secured in the open position, then verify these valves open at least once per 31 days.
*For valve temperatures less than or equal to 130*F.
**The lift setting pressure shall correspond to ambient conditions of the valve
( at nominal operating temperature and pressure. SAN ONOFRE-UNIT ft :5 3/4 4-33 "C.~ -- 2-l l - - - M"
~
- 3/4.4 REACTOR COOLANT SYSTEM C- 8ASES ~
3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION - 4 The plant is designed to operate with both reactor coolant loops and g Al*' b' a'ssociated reactor coolant pumps in operation, and maintain DNBR ehementuH l'24 during all normal operations and anticipated transients. As a result, in MODES 1 and 2 with one reactor coolant loop not in operation, this specification requires that the plant be in at least HOT STANDBY within I hour since no safety analysis has been conducted for operation with less than 4 reactor coolant pumps or less than two reactor coolant loops in operation. In MODE 3, a single reactor coolant loop provides sufficient heat removal capability for removing decay heat; however, single failure considerations require that two loops be OPERA 8LE. In MODE 4, and in MODE 5 with reactor coolant loops filled, a single reactor coolant loop or shutdown cooling train provides sufficient heat removal cap hility for removing decay heat; but single failure considerations require that at least two loops / trains (either RCS or shutdown cooling) be OPERABLE. In MODE 5 with reactor coolant loops not filled, a single shutdown cool-ing train provides sufficient heat removal capability for removing decay heat; but single failure considerations, and the unavailability of the steam genera' tors as a heat removing component, require that at least two shutdown cooling trains be OPERABLE. The operation of one Reactor Coolant Pump or one shutdown cooling pump C provides adequate flow to ensure alxing, prevent stratification and produce gradual reactivity changes during boron concentration reductions in the Reactor Coolant System. The reactivity change rate associated with boron reductions will,therefore,hewithinthecapabilityofoperagogecognitionandcontrol. The restrictions on starting a Reactor CoolantJPump in Modes 4 and 5 with one or more RCS cold legs less than or equal to 348'F are provided to prevent RCS pressure transients, caused by energy additions from the secondary system, which could exceed the limits of Appendix G to 10 CFR Part 50. The RCS will be protected against overpress'ure transients and will not exceed the limits of Appendix G by either (1) restricting the water volume in the pressurizer and thereby providing a volume for the primary coolant to expand into or (2) by l restricting starting of the RCPs to when the secondary water temperature of each steam generator is less than 100*F above each of the RCS cold leg temperatures. 3/4.4.2 SAFETY VALVES The pressurizer code safety valves operate to prevent the RCS from being pressurized above its Safety Limit of 2750 psia. Each safety valve is designed t l to relieve 4.6 x 105 lbs per hour of saturated steam at the valve.setpoint plus 3E accumulation. The relief capacity of a single safety valve is adequate
- to 2% relieve old leg anytemperature overpressure condition greater which than W F.could occur during shutdown with RCS In the event that no safety valves are OPERABLE and for RCS cold leg temperature less than or equal to S*F, the .
pressure relief capability and will prevent RCS overpressurization. operatin k 8 3/4 4-1 t SAN ONOFRE-LW e
. : = . w . .~ = - - - - .- - -. = ..:. - -- - - --
i REACTOR COOLANT SYSTEM BASES SPECIFIC ACTIVITY (Continued)
. L Reducing T to less than 500*F prevents the release of activity should l asteamgeneratN8thbe rupture since the saturatier. pressure of the primary coolant is below the lift pressure of the atmospheric steam relief velves.
i The surveillance requirements provide adequate assurance that excessive specific activity levels in the time to take corrective action. primaryInformationcoolant will beondetected obtained in sutficient iodine spiking will be used to assess the parameters associated with spiking phenomena. A reduction in frequency of isotopic analyses following power changes may be permissible if justified by the data obtained. 3/4.4.8 PRESSURE / TEMPERATURE LIMITS
- All components in the Reactor Coolant System are designed to withstand 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 Section 3.9.1.1 of the FSAR. During startup and shutdown, the rates of temperature and pressure changes are ( limited so that the maximum specified 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 tensile at the outer wall. These*wemal sinduced compressive stresses tend to alleviate ( the tensile stresses induced the internal pressure.' Therefore, a pressure-temperature curve based on steady state conditions (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. The heatup analysis also cones the determination of pressure-temperature limitations for the case in which the outer wall of the vessel becomes the controlling location. The thermal gradients established 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 Seemisf induced stresses at the outer wall of the vessel are tensile and are l dependent on both the rate of heatup and the' time along the heatup ramp; therefore, a lower bound curve similar to that described for the heatup of the inner wall cannot be defined. Consequently, for the cases in which the outer well of the vessel becomes the stress controlling location, each heatup rate of interest must be analyzed on en individual basis. SAN ONOFRE-UNIT g 3 8 3/4 4-6 MEEOsalFr-l
REACTOR COOLANT SYSTEM ( BASES - _ PRESSURE / TEMPERATURE LIMITS (Continued) - The heatup and cooldown limit curves (Figures 3.4-2 and 3.4-3) are composite cu'rves which were prepared by determining the most conservative case,_with , = either the inside or outside wall controlling, for any heatup rwk of we +o 60*//h ' or coolkua nu=. er ve += soo'r%e . The heatup and cooldown curves were prepared basea upon l the most limiting value of the predicted adjusted reference temperature C, the end of the service period indicated on Figure 3.4-2 and 3.4-3. The reactor vessel materials have been tested to determine their initial RT the results of these test are shown in Table B 3/4.4-1. ti$T;ndresultantfastneutron(Egreaterthan1Mev)irradiationwillcauseReactor a opera-an increase in the RT Therefore, an adjusted reference temperature, based upon the fluence and h p.er and phosphorous content of the material in question, can be predicted using FSAR Table 5.2-5 and the recommendations of Regulatory Guide 1.99, Revision 1, " Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials." The heatup and cooldown limit curves, Figures 3.4-2 and 3.4-3, include predicted adjustments for.this shift in RT L at the end of the applicable service period, as well as adjustments for NOT f possible errors in the pressure and temperature sensing instruments. , S' The actual shift in RT of the vessel material will be established
- periodicallyduringoperatigTby removing and evaluating, in accordance with
. ASTM E185-73 and 10 CFR Appendix H, reactor vessel material irradiation sur-
% veillance specimens installed near the inside wall of the reactor vessel in l the ccre area. The surveillance specimen withdrawal schedule is shown in Table 4.4-5. Since the neutron spectra at the irradiation samples and vessel inside radius are essentially identical, the measured transition shift for a
, sample can be applied with confidence to the adjacent section of the reactor vessel taking into account the location of the sample closer to the core than '
_ the vessel wall by means at the Lead Factor. The heatup and cooldown curves 7 must be recalculated when the delta RT determined from the surveillance i capsuleisdifferentfromthecalculatEdeltaRT for the equhalent capsule y radiation exposure. HDT The pressure-temperature limit lines shown on Figure 3.4-2 and 3.4-3 for r 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 maximus RT for all reactor coolant system pressure-retaining materials, with the Ngception of the reactor pressure vessel, has been deter-mined to be 90*F. The Lowest Service Temperature limit line shown on
= Figure 3.4-2 and 3.4-3 is based upon this RT since Article NB-2332 (Summer e Addenda of 1972) of Section III of the ASME NIler and Pressure Vessel Code 7 requires the Lowest Service Temperature to be RT 100*F for piping, pumps y andvalves.Belowthistemperature,thesystemp$s+uremustbelimitedtoa maximum of 20% of the system's hydrostatic test pressure of 3125 psia.
The limitations imposed en the pressurizer heatup and cooldown rates and spray water temperature differential are provided to assure that the pres-surizer is operated within the design criteria assumed for the fatigue analysis performed in accorciance with the ASME Code requirements. SAN ONOFRE-UNIT g 3 8 3/4 4-7 gyG~ L _ _ _ _ _ ___ _ __ _- - ----- --------- - - - - - - - - - - - - - - - - -- ------
. .s .
TAstr8SN.4-l .
, REACTOR VESSEL TOUGl01E55 -
SAll ONOFRL ISIIT 3 i Temperature of IHalsene W ! Charpy V-Ilotch Shelf Cv energy i E 9 30 0 50 for Longitudinal Piece lie. Code lie. IIsterial Vessel Location [p)_ F ft - Ib-ft - Ib Direction - ft Ib j 215-01 C-6801-1 A533GROCLI W Shell Plate -20 28 64 lif l 215-01 C-6801-2 A533GRBCLI tipper h 11 Plate -20 -6 34 8# i l 215-01 C-6801-3 A533GRBCLI tapper h 11 Plate -20 IS 36 lif i 215-02 C-6002-4 A533GleCLI Lower Shell Plate -30 32 62 115 : 215-02 C-6002 5 A533GRBCLI Lauer Shell Plate 6 36 64 Ile i l 215-02 C-6802-6 A533GRBCLI Lower Shell Plate -40 32 100 ,, 30 [ , 215-03 C-6002-1 A533GRCCLI Intermediate h il -20 56 ISO g5 I 215-03 C-6802-2 A533GRBCLI Intermediate ~ 5 hell -20 40 66 113 215-03 C-6802-3 A533GRSCLI Intermediate hil -10 44 80 101 l 203-02 C-6823 A500CL2 Vessel Flange Forging 0 -30 -15 WA 20g-02 C-E824-1 A50lKt2 Closure plead Flange -40 -No -lea #4 Forging 205-02 C-6829-1 A500CL2 Inlet florale Forging 10 -35 -5 . leg a 205-02 C-6829-2 A508CL2 Inlet flozzle Forging 0 - 55 -M 156 205-02 C-4829-3 A500CL2 . Inlet IIstate Forging 10 -W - 35 112 205-02 C-6829-4 A500CL2 Inlet flozzle Forging 10 -30 15 108 205-06 C-6830-1 A500CL2 Outlet flezzle Forging -10 -10 -85 125 205-06 C-6030-2 A500CL2 Outlet listale Forging -10 -18 -5 131 232-01 C4040-1 A533ERBCLI Oettom lined Tones -50 - 30 o id7 ' ' 232-82 C-0041-1 A533metti Oettom Ilsed Omme -40 le 20 11
)
. ' . : 1. ,
TAaLE 6 3/*l.Y-l .
. IIEACTOR VESSEL TOUGHESS ,
5AN OWFilE UNIT 3 Temperature of Minlaus W 4 Charpy V-Itotch Shelf Cv energy , for Longitudinal Place its. Code lie.' lhterial Vessel Location PT 1 e 30 950 ft - Ib-ft - Ib Birecteen - ft Ib i 205-03 C-4831-1 A500CLI Inlet IIsrzle Ferging S/E -20 11 40 124 205-03 C-6831-2 A508CLI Inlet llorale Forging S/E -20 11 40 124 205-03 C-6831-3 A508CLI Inlet IInzzle Forging 5/E -20 -15 50 114 i 205-03 C-6838-4 A500CLI Inlet flozzle Forging 5/E -20 -15 50 114 255-0F C-4032-1 A5003.1 Outlet lisaale Fergleg $/E -20'- -20 9 15't 205-07 C-6832-2 A508CLI Outlet florale Forging 5/E -20 -20 0 ,152 231-01 C-4833-1 A$33GABCLI Closure Ilsed Peel - WP. 9 M M 231-01 C-4834-1 A533GilBCLI Closure Head Peel - 30 se m M l i 231-02 . Cc4835-1 A533GABCLI Closure llead Dame - 4e le M M l l ' l . M = list Available O e
e REACTOR COOLANT SYSTEM (o . BASES PRESSURE / TEMPERATURE LIMITS (Continued) i The OPERABILITY of the Shutdown Cooling System relief valve or a RCS vent opening of greater than 5.6 square inches ensures that the RCS will be protected from pressure transients which could exceed the limits of Appendix G to 10 to GPF.CFR Part 50 when one or more of the RCS cold legs are less than or equal I The Shutdown Cooling System relief valve has adequate relieving g.C cap _) ability to protect the RCS from overpressurization when t limited to either (1) the start of an idle RCP with the secondary water temperature of the steam generator less than or equal to 100*F above the RCS cold leg;temperatures HPSI F: or (2) inadvertant safety injection actuation with two 1-M " . .. ,___5 and latdown isolated. T into a eater famp solid RCS with full charging capacity i n3ec 7m. 3/4.4.9 STRUCTURAL INTEGRITY The inservice inspection and testing programs for ASME Code Class 1, ' 2 and 3 ccaponents 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 Cv. Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda l as required by 10 CFR Part 50.55a(g) except where specific written relief hes l been granted by the Commission pursuant to 10 CFR Part 50.55a (g) (6) (1). Components of the reactor coolant system were designed to provide access to pennit inservice inspections in accordance with Section XI of the ASME Boiler and Pressure Vessel Code, 1974 Edition and Addenda through Summer 1975. l l SAN ONOFRE-UNIT f 8 3/4 4-9 ' 3 ! f
~
ehe,-p%W MJmm g,_ .M Ne
=ee,- 4ews %, 44 4 ,w - -e 4
6 e 0 e e e ATTACHMENT 5 0
j
= .,
[ EMERGENCY CORE COOLING SYSTEMS . SURVEILLANCE REQUIREMENTS (Continued) -
- 2. A visual inspection of the containment sump and verifying that the subsystem suction inlets are not restricted by debris and i that the sump components (trash racks,1::reens, etc.) show no evidence of structural distress or abnormal corrosion.
- e. At least once per 18 months, during shutdown, by:
- 1. Verifying that each automatic valve in the flow path actuates to its correct position on SIAS and RAS test signals.
- 2. Verifying that each of the following pumps start automatically upon receipt of a Safety Injection Actuation Test Signa:
- a. High-Pressure Safety Injection pump.
- b. Low-Pressure Safety Injection pump.
- c. Charging pump.
- 3. Verifying that on a Recirculation Actuation Test Signal, the containment sump isolation valves open and tha recirculation vaises at the refueling water tank close.
- f. By verifying that each of the following pumps develops the indicated developed head and/or flow rate when tested pursuant to Specification 4.0.5:
- 1. .High-Pressure Safety Injection pumps developed head, at an indi-cated flow rate of 650 gpm, greater than or equal to M feet for P017, M feet for P018 and 2g55 for P019. *#T3 2.I'52- 2.of f
- 2. Low-Pressure Safety Injection pump developed head greater than or equal to 4ES S 9 c-E feetk.a.t m bii fis.s
- 3. Charging pump flow rate greater than or equal to 40 gps.
- g. By performing a flow balance test, during shutdown, following completion of modifications to the ECCS subsystems that alter the subsystem flow characteristics and verifying the following flow rates:
- 1. For High-Pressure Safety Injection pump cold, leg injection with a single pump running:
- a. The sum of the injection lines flow rates, excluding the highest flow rate, is ater than or equal to 44'1152 gpa for P017 running, gpa for P018 running and 5 5 gpa for P019 running and 4 t. ,' gg6
- b. The total pump flow rate is greater than or equal to C82.6 for P017 running, gpa for P018 running and gpa for P019 running. 99 E
SANONOFRE-UNIT [3 3/4 5-5
~
( , . I l lb EMERGENCY CORE COOLING SYSTEMS l l
' SURVEILLANCE REQUIREMENTS (Continued) l
- l
- 2. For a single High-Pressure Safety Injection pump hot / cold leg injection.
- a. The sum of the c*old leg infection flow rates is greater '
than or equal to 385 gpe, and
- b. The hot leg injection flow rate is greater than oi equal to 385 gpe.
- c. The combined total het/ cold legs injection flow rate is greater than or equal to 896 gpa.
- 3. For the Low-Pressure Safety Injection pump with a single pump running:
- a. The flow through 2ach injection leg shall be greater than or equal to 3000 gpm when tested individually and corrected to the same pump suction source and leg back r.ressure conditions. The difference between high and law flow legs
( shall be less than or equal to 100 gpe. , b. The total ECCS flow through 2 cold leg injection lines i shall be greater than or equal to 4450 gpm when corrected for elevation head. e ( SAN ONOFRE-UNIT 3/4 5-6 .
4.> aps: y a rp o a- sp- e e 4w de e e m w e ar ee>ep omumW - much W M -- Whmmens
- 4 , m e.-swp1m se ga w-B
-.-e. a . ,.
l O e O e e ATTACHMENT 6 S
- e 9
.. .. ...._ i..._ _
a --_..1.. - . . .1 -. BASES 3/4 9.6 REFUELING MACHINE The UPERABILITY requirements for the refueling machine ensure that: 1)the refueling machine will be used for movement of all fuel assemblies including those with a CEA inserted 2) each machine has sufficient load capacity to fift a fuel assembly including those with a CEA, and 3) the core internals and pressure vessel are protected from excessive lifting force in the event they are inadvertenly engaged during lifting operation. With the exception of the four finger CEA's, CEA's are removed from the . reactor vessel along with the fuel bundle in which they are inserted utilizing the refueling machine. The four finger CEA's are inserted through the upper guide structure (UGS) with two fingers in each of two adjacent fuel bundles in the periphery of tne core. The four finger CEA's are either removed with the UGS and lift rig or can be removed with separate tooling prior to UGS removal utilizing the auxiliary hoist of the polar crane. t {
( h i 0
- e 0
f ATTACHMENT 7 0 T
1 BASES FOR 3/4 10.1 SHUTDOWN MARGIN This special test exception provides that a minimum amount of CEA worth is immediately available for reactivity control when CEA worth measurement tests are performed. This special test exception is required to perdit the periodic yerification of the actual versus predicted core reactivity condition !
-occurring as a result of fuel burnup or fuel cycling operations.
Although CEA worth testing is conducted in MODE 2, during the performance of these tests sufficient negative reactivity is inserted to result in temporary entry into MODE 3. Because the intent is to immediately return to MODE 2 to continue CEA worth measurements, the special test exception allows limited operation in MODE 3 without having to borate to meet the shutdown margin requirements of Technical Specification 3.1.1.1. l
. e
,py 4d**OW O
e 0
=
S e B ATTACHMENT 8 I f l
.a .-.w.- .~. - - . u. x - . -. ,. --
(, BASES FOR 3/4.1.2 (B0 RATION SYSTEMS) The water volume limits are specified relative to the top of tie highest suction connection to the tank. (Water volume below this datum is not
. considered recoverable for purposes of this specification.) Vortexing, internal structures and instrument error are considered in determining the
. tank level corresponding to the.specified water volume limits.
BASES FOR 3/4.5.4 (REFUELING WATER STORAGE TANK) : Change First Paragraph on p. 3/4.5-3 for clarity: 1 The water volume limits are specified relative to the top of the highest l suction connection to the tank. (Water volume below this datum is not considered recoverable for purposes of this specification.) The specified volume limits consist of the minimum volume requirea for ECCS injection above the Recirculation Actuation Signal (R.'.S) setpoint, plus the minimum volume ' required for the transition to ECCS recirculation below the RAS setpoint, plus the volume corresponding to the range of the RAS setpoint, including RAS l instrument error high and low. Vortexing, internal structure, and instrument error are considered in determining the tank level corresponding to the specified water volume limits. BASES FOR 3/4.7.1.3 (CONDENSATE STORAGE TANKS) ! The OPERABILITY of condensate storage tank T-121 with the minimum water volume ensures that sufficient water is available to maintain the RCS at NOT STANDBY conditions for two hours followed by cooldown to shutdown cooling initiation, with steam discharge to atmosphere with concurrent loss cf off-site power and most limiting single failure. The OPERABILITY of condensate storage tank T-120 in conjunction with tank T-121 ensures that sufficient water is available to maintain the RCS at HOT STANDBY conditions for 24 hours including cooldown to shutdown cooling initiation, with steam discharge to atmosphere with concurrent loss of off-site power and most limiting single failure. The' contained water volume limits are specified relative to the highest auxiliary faedwater pump suction inlet in the tank for T-121, and to the T-121 cross connect siphon inlet for T-120. (Water volume below these datum levels is not considered recoverable for purposes of this specification). Vortexing, , internal structure and instrument error are considered in determining the tank levels corresponding to the specified water volume limits. Prior to achieving 100% RATED THERMAL POWER, Figure 3.7-1 is used to determine the minimum required water volume for T-121 for the maximum power level (hence maximum decay heat) achieved.
. e l
l
= =4 .a. 4 - - +% . . . mea...w...swwe,,,,g- . . , , ,
4 N e 0 e S 0 ATTACHMENT 9 O 6 9 6
l 1 l( TABLE 3.3-7 SEISMIC W ITORING INSTRUMENTATION - i .
^
< Minimum Measurement Instrument f Instrumerits & Sensor locations
- Rance Operable _
- 1. Triaxial Time-History Strong Motion Accelerometers .
- a. Steam Generator Base Support -2 to +2g 1
- b. Pressurizer Base Support
-2 to +2g I
- c. Reactor Coolant Pump -2 to +2g 1
- d. Containment Base in Tendon Gallery -2 to +2g i
- e. Containment Operating Level -2 to +2g i
- f. Unit #1 Free Field -1 to +1g 1 2
- g. Control Building Basement -2 to +2g 1
- h. Control Building Roof -2 to +2g 1
- 1. Safety Equipment Building Base Slab -2 to +2g 1
- j. Safety Equipment Building Piping Support -2 to +2g 1
- k. Radwaste Building Equipment Support -2 to +2g 1
- 2. Triaxial Peak Reading Accelerographs .
- a. Control Building-Control Room -2 to +2g i
- b. Control Building Base -2 to +2g I
- c. Top of Containment Structure -5 to +5g f d. Reactor Coolant Piping i
-2 to +2g 1
- 3. Seismic Triggers
- a. Containment Base in Tendon Gallery +0.00,5 to +0.05g 1
- b. Containment Operating Level +0.005 to +0.05g 1 i
- 4. Seismic Switches
- a. Steam Generator Base Support Set pt. 0.45 Horz/0.30 Vert. 1**
- b. Containment Base in Tendon Gallery Set pt. 0.40 Horz/0.50 Vert. 1**
- 5. Seismic Alarm Annunciator (4a & 4b are sensors)
- a. Control Room Panel L-167
- 6. Peak Shock Recorder
- a. Containment Base in Ten &n Gallery 2 to 25.4 Hz 1**
, 1.6 to 90g
- 7. Peak Shock Annunciator 2to25.4$z 1 1.6 to 90g
- a. Control Room Panel L-167 j All sum;c. ens km u+a h.n es loca+ad on us.1 2. w . tk L s < ep k m o f ilem I .+ .
With control room indication l 3/4 3-43 SANONOFRE-UNIT /3
. . = - [LL 6
TABLE 4.3-4 SEISMIC MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS
~
. CHANNEL CHANNEL FUNCTIONAL FUNCTIONAL
_ INSTRUMENTS AND SENSOR LOCATIONS CHECK CALIBRATION TEST
- 1. Triaxial Time-History Strong Motion Accelerometers
- a. Steam Generator Base Support M* R SA
- b. Pressurizer Base Support M* R SA
- c. Reactor Coolant Pump N' R SA
- d. Containment Base in Tendon Gallery M* R SA
- e. Containment Operating Level M* R SA
- f. Control Building Basement M* R SA
- g. Control Building Roof M* R SA
- h. Safety Equipment Building Base - M* R SA
- 1. Safety Equipment Building Piping Support M* R SA
- j. Radwaste Building Equipment Support M* R SA
- 2. Triaxial Peak Recording Accelerographs
- a. Control Building-Control Room N/A R N/A
- b. Control Building Base N/A R N/A
- c. Top of Containment Structure N/A R N/A
- d. Reactor Coolant Piping N/A R N/A
- 3. Seismic Triggers
- a. Containment Base in Tendon Gallery M R SA
- b. Containment Operating Level M R S/U***
- 4. Seismic Switches
- a. Steam Generator pase Support M R** SA**
- b. Containment Base in Tendon Gallery M R** SA**
- 5. Seismic Alare Annunciators (4a & 4b are sensors) .
- a. Control Room Panel L-167 M R SA
- 6. Peak Shock Recorder
- a. Containment Base in Tendcn Gallery N/A R** N/A
- 7. Peak Shock Annunciator *
- a. Control Room Parsel L-167 N/A R**
N/A o Except seismic trigger an With Control Room indication 808 Need not be performed more frequently g than once per 6 months. W AII se sm.e ;;s ya,,,,,,4 +,,,, , . ja h J i s, ; y,, . r & I qm_@MER-%%mfLE) cwLswn -
* '}}